International Nuclear Information System (INIS)
Kaul, D.C.
1982-01-01
Throughout the last two decades many efforts have been made to estimate the effect of body self-shielding on organ doses from externally incident neutrons and gamma rays. These began with the use of simple geometry phantoms and have culminated in the use of detailed anthropomorphic phantoms. In a recent effort, adjoint Monte Carlo analysis techniques have been used to determine dose and dose equivalent to the active marrow as a function of energy and angle of neutron fluence externally incident on an anthropomorphic phantom. When combined with fluences from actual nuclear devices, these dose-to-fluence factors result in marrow dose values that demonstrate great sensitivity to variations in device type, range, and body orientation. Under a state-of-the-art radiation transport analysis demonstration program for the Japanese cities, sponsored by the Defense Nuclear Agency at the request of the National Council on Radiation Protection and Measurements, the marrow dose study referred to above is being repeated to obtain spectral distributions within the marrow for externally incident neutrons and gamma rays of arbitrary energy and angle. This is intended to allow radiobiologists and epidemiologists to select and to modify numbers of merit for correlation with health effects and to permit a greater understanding of the relationship between human and laboratory subject dosimetry
MPACT Subgroup Self-Shielding Efficiency Improvements
International Nuclear Information System (INIS)
Stimpson, Shane; Liu, Yuxuan; Collins, Benjamin S.; Clarno, Kevin T.
2016-01-01
Recent developments to improve the efficiency of the MOC solvers in MPACT have yielded effective kernels that loop over several energy groups at once, rather that looping over one group at a time. These kernels have produced roughly a 2x speedup on the MOC sweeping time during eigenvalue calculation. However, the self-shielding subgroup calculation had not been reevaluated to take advantage of these new kernels, which typically requires substantial solve time. The improvements covered in this report start by integrating the multigroup kernel concepts into the subgroup calculation, which are then used as the basis for further extensions. The next improvement that is covered is what is currently being termed as ''Lumped Parameter MOC''. Because the subgroup calculation is a purely fixed source problem and multiple sweeps are performed only to update the boundary angular fluxes, the sweep procedure can be condensed to allow for the instantaneous propagation of the flux across a spatial domain, without the need to sweep along all segments in a ray. Once the boundary angular fluxes are considered to be converged, an additional sweep that will tally the scalar flux is completed. The last improvement that is investigated is the possible reduction of the number of azimuthal angles per octant in the shielding sweep. Typically 16 azimuthal angles per octant are used for self-shielding and eigenvalue calculations, but it is possible that the self-shielding sweeps are less sensitive to the number of angles than the full eigenvalue calculation.
Monte Carlo validation of self shielding and void effect calculations
International Nuclear Information System (INIS)
Tellier, H.; Coste, M.; Raepsaet, C.; Soldevila, M.; Van der Gucht, C.
1995-01-01
The self shielding validation and the void effect are studied with Monte Carlo method. The satisfactory comparison obtained between the APOLLO 2 results of the self shielding effect and the TRIPOLI and MCNP results allows us to be confident in the multigroup transport code. (K.A.)
Self-Shielding Of Transmission Lines
Energy Technology Data Exchange (ETDEWEB)
Christodoulou, Christos [Univ. of New Mexico, Albuquerque, NM (United States)
2017-03-01
The use of shielding to contend with noise or harmful EMI/EMR energy is not a new concept. An inevitable trade that must be made for shielding is physical space and weight. Space was often not as much of a painful design trade in older larger systems as they are in today’s smaller systems. Today we are packing in an exponentially growing number of functionality within the same or smaller volumes. As systems become smaller and space within systems become more restricted, the implementation of shielding becomes more problematic. Often, space that was used to design a more mechanically robust component must be used for shielding. As the system gets smaller and space is at more of a premium, the trades starts to result in defects, designs with inadequate margin in other performance areas, and designs that are sensitive to manufacturing variability. With these challenges in mind, it would be ideal to maximize attenuation of harmful fields as they inevitably couple onto transmission lines without the use of traditional shielding. Dr. Tom Van Doren proposed a design concept for transmission lines to a class of engineers while visiting New Mexico. This design concept works by maximizing Electric field (E) and Magnetic Field (H) field containment between operating transmission lines to achieve what he called “Self-Shielding”. By making the geometric centroid of the outgoing current coincident with the return current, maximum field containment is achieved. The reciprocal should be true as well, resulting in greater attenuation of incident fields. Figure’s 1(a)-1(b) are examples of designs where the current centroids are coincident. Coax cables are good examples of transmission lines with co-located centroids but they demonstrate excellent field attenuation for other reasons and can’t be used to test this design concept. Figure 1(b) is a flex circuit design that demonstrate the implementation of self-shielding vs a standard conductor layout.
Uncertainty Analysis with Considering Resonance Self-shielding Effect
International Nuclear Information System (INIS)
Han, Tae Young
2016-01-01
If infinitely diluted multi-group cross sections were used for the sensitivity, the covariance data from the evaluated nuclear data library (ENDL) was directly applied. However, in case of using a self-shielded multi-group cross section, the covariance data should be corrected considering self-shielding effect. Usually, implicit uncertainty can be defined as the uncertainty change by the resonance self-shielding effect as described above. MUSAD ( Modules of Uncertainty and Sensitivity Analysis for DeCART ) has been developed for a multiplication factor and cross section uncertainty based on the generalized perturbation theory and it, however, can only quantify the explicit uncertainty by the self-shielded multi-group cross sections without considering the implicit effect. Thus, this paper addresses the implementation of the implicit uncertainty analysis module into the code and the numerical results for the verification are provided. The implicit uncertainty analysis module has been implemented into MUSAD based on infinitely-diluted cross section-based consistent method. The verification calculation was performed on MHTGR 350 Ex.I-1a and the differences with McCARD result decrease from 40% to 1% in CZP case and 3% in HFP case. From this study, it is expected that MUSAD code can reasonably produce the complete uncertainty on VHTR or LWR where the resonance self-shielding effect should be significantly considered
Uncertainty Analysis with Considering Resonance Self-shielding Effect
Energy Technology Data Exchange (ETDEWEB)
Han, Tae Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2016-10-15
If infinitely diluted multi-group cross sections were used for the sensitivity, the covariance data from the evaluated nuclear data library (ENDL) was directly applied. However, in case of using a self-shielded multi-group cross section, the covariance data should be corrected considering self-shielding effect. Usually, implicit uncertainty can be defined as the uncertainty change by the resonance self-shielding effect as described above. MUSAD ( Modules of Uncertainty and Sensitivity Analysis for DeCART ) has been developed for a multiplication factor and cross section uncertainty based on the generalized perturbation theory and it, however, can only quantify the explicit uncertainty by the self-shielded multi-group cross sections without considering the implicit effect. Thus, this paper addresses the implementation of the implicit uncertainty analysis module into the code and the numerical results for the verification are provided. The implicit uncertainty analysis module has been implemented into MUSAD based on infinitely-diluted cross section-based consistent method. The verification calculation was performed on MHTGR 350 Ex.I-1a and the differences with McCARD result decrease from 40% to 1% in CZP case and 3% in HFP case. From this study, it is expected that MUSAD code can reasonably produce the complete uncertainty on VHTR or LWR where the resonance self-shielding effect should be significantly considered.
Self shielding in cylindrical fissile sources in the APNea system
International Nuclear Information System (INIS)
Hensley, D.
1997-01-01
In order for a source of fissile material to be useful as a calibration instrument, it is necessary to know not only how much fissile material is in the source but also what the effective fissile content is. Because uranium and plutonium absorb thermal neutrons so Efficiently, material in the center of a sample is shielded from the external thermal flux by the surface layers of the material. Differential dieaway measurements in the APNea System of five different sets of cylindrical fissile sources show the various self shielding effects that are routinely encountered. A method for calculating the self shielding effect is presented and its predictions are compared with the experimental results
A new formulation for resonance self-shielding factors
Energy Technology Data Exchange (ETDEWEB)
Palma, Daniel A.P.; Martinez, Aquilino S.; Silva, Fernando C. da [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear]. E-mail: aquilino@lmp.ufrj.br
2007-07-01
The activation technique allows either absolute or relative very precise neutron intensity measurements. This technique requires the knowledge of the Doppler broadening function to determine resonance self-shielding factors. In the present work a new formulation is proposed for the self-shielding factors where the Doppler broadening function is calculated using the Frobenius's method and compared to the values obtained from the four-pole Pade method. This calculation method is shown to be effective from the point of view of accuracy. (author)
A new formulation for resonance self-shielding factors
International Nuclear Information System (INIS)
Palma, Daniel A.P.; Martinez, Aquilino S.; Silva, Fernando C. da
2007-01-01
The activation technique allows either absolute or relative very precise neutron intensity measurements. This technique requires the knowledge of the Doppler broadening function to determine resonance self-shielding factors. In the present work a new formulation is proposed for the self-shielding factors where the Doppler broadening function is calculated using the Frobenius's method and compared to the values obtained from the four-pole Pade method. This calculation method is shown to be effective from the point of view of accuracy. (author)
Neutron self-shielding with k0-NAA irradiations
International Nuclear Information System (INIS)
Chilian, C.; Chambon, R.; Kennedy, G.
2010-01-01
A sample of SMELS Type II reference material was mixed with powdered Cd-nitrate neutron absorber and analysed by k 0 NAA for 10 elements. The thermal neutron self-shielding effect was found to be 34.8%. When flux monitors were irradiated sufficiently far from the absorbing sample, it was found that the self-shielding could be corrected accurately using an analytical formula and an iterative calculation. When the flux monitors were irradiated 2 mm from the absorbing sample, the calculations over-corrected the concentrations by as much as 30%. It is recommended to irradiate flux monitors at least 14 mm from a 10 mm diameter absorbing sample.
RZ calculations for self shielded multigroup cross sections
Energy Technology Data Exchange (ETDEWEB)
Li, M.; Sanchez, R.; Zmijarevic, I.; Stankovski, Z. [Commissariat a l' Energie Atomique CEA, Direction de l' Energie Nucleaire, DEN/DM2S/SERMA/LENR, 91191 Gif-sur-Yvette Cedex (France)
2006-07-01
A collision probability method has been implemented for RZ geometries. The method accounts for white albedo, specular and translation boundary condition on the top and bottom surfaces of the geometry and for a white albedo condition on the outer radial surface. We have applied the RZ CP method to the calculation of multigroup self shielded cross sections for Gadolinia absorbers in BWRs. (authors)
RZ calculations for self shielded multigroup cross sections
International Nuclear Information System (INIS)
Li, M.; Sanchez, R.; Zmijarevic, I.; Stankovski, Z.
2006-01-01
A collision probability method has been implemented for RZ geometries. The method accounts for white albedo, specular and translation boundary condition on the top and bottom surfaces of the geometry and for a white albedo condition on the outer radial surface. We have applied the RZ CP method to the calculation of multigroup self shielded cross sections for Gadolinia absorbers in BWRs. (authors)
Radiation monitoring in a self-shielded cyclotron installation
International Nuclear Information System (INIS)
Capaccioli, L.; Gori, C.; Mazzocchi, S.; Spano, G.
2002-01-01
As nuclear medicine is approaching a new era with the spectacular growth of PET diagnosis, the number of medical cyclotrons installed within the major hospitals is increasing accordingly. Therefore modern medical cyclotron are highly engineered and highly reliable apparatus, characterised with reduced accelerating energies (as the major goal is the production of fluorine 18) and often self-shielded. However specific dedicated monitors are still necessary in order to assure the proper radioprotection. At the Careggi University Hospital in Florence a Mini trace 10 MeV self-shielded cyclotron produced by General Electric has been installed in 2000. In a contiguous radiochemistry laboratory, the preparation and quality control of 1 8F DG and other radiopharmaceuticals takes place. Aim of this work is the characterisation and the proper calibration of the above mentioned monitors and control devices
Unresolved resonance self shielding calculation: causes and importance of discrepancies
International Nuclear Information System (INIS)
Ribon, P.; Tellier, H.
1986-09-01
To compute the self shielding coefficient, it is necessary to know the point-wise cross-sections. In the unresolved resonance region, we do not know the parameters of each level but only the average parameters. Therefore we simulate the point-wise cross-section by random sampling of the energy levels and resonance parameters with respect to the Wigner law and the X 2 distributions, and by computing the cross-section in the same way as in the resolved regions. The result of this statistical calculation obviously depends on the initial parameters but also on the method of sampling, on the formalism which is used to compute the cross-section or on the weighting neutron flux. In this paper, we will survey the main phenomena which can induce discrepancies in self shielding computations. Results are given for typical dilutions which occur in nuclear reactors. 8 refs
Unresolved resonance self shielding calculation: causes and importance of discrepancies
International Nuclear Information System (INIS)
Ribon, P.; Tellier, H.
1986-01-01
To compute the self shielding coefficient, it is necessary to know the point-wise cross-sections. In the unresolved resonance region, the parameters of each level are not known; only the average parameters. Therefore the authors simulate the point-wise cross-section by random sampling of the energy levels and resonance parameters with respect to the Wigner law and the x 2 distributions, and by computing the cross-section in the same way as in the resolved regions. The result of this statistical calculation obviously depends on the initial parameters but also on the method of sampling, on the formalism which is used to compute the cross-section or on the weighting neutron flux. In this paper, the authors survey the main phenomena which can induce discrepancies in self shielding computations. Results are given for typical dilutions which occur in nuclear reactors
Revisiting the stamm'ler self-shielding method
International Nuclear Information System (INIS)
Hebert, A.
2004-01-01
The generalized Stamm'ler method is been used in lattice codes such as PHOENIX, WIMS-AECL and DRAGON-IST for computing self-shielded cross sections, prior to the main flux calculation. This method is handicapped by deficiencies, such as its low accuracy and its inability to represent distributed self-shielding effects in a fuel rod or across a fuel bundle. The paper describes improvements that could be made to the generalized Stamm'ler method in order to mitigate these two defects. A validation is presented for the case of 238 U nuclides located in different geometries. The isotopic absorption rates obtained with the proposed numerical scheme are compared with exact values obtained with a fine-group elastic slowing-down calculation in the resolved energy domain. (author)
Uranium self-shielding in fast reactor blankets
Energy Technology Data Exchange (ETDEWEB)
Kadiroglu, O.K.; Driscoll, M.J.
1976-03-01
The effects of heterogeneity on resonance self-shielding are examined with particular emphasis on the blanket region of the fast breeder reactor and on its dominant reaction--capture in /sup 238/U. The results, however, apply equally well to scattering resonances, to other isotopes (fertile, fissile and structural species) and to other environments, so long as the underlying assumptions of narrow resonance theory apply. The heterogeneous resonance integral is first cast into a modified homogeneous form involving the ratio of coolant-to-fuel fluxes. A generalized correlation (useful in its own right in many other applications) is developed for this ratio, using both integral transport and collision probability theory to infer the form of correlation, and then relying upon Monte Carlo calculations to establish absolute values of the correlation coefficients. It is shown that a simple linear prescription can be developed for the flux ratio as a function of only fuel optical thickness and the fraction of the slowing-down source generated by the coolant. This in turn permitted derivation of a new equivalence theorem relating the heterogeneous self-shielding factor to the homogeneous self-shielding factor at a modified value of the background scattering cross section per absorber nucleus. A simple version of this relation is developed and used to show that heterogeneity has a negligible effect on the calculated blanket breeding ratio in fast reactors.
Self Shielding in Nuclear Fissile Assay Using LSDS
International Nuclear Information System (INIS)
Lee, Yong Deok; Park, Chang Je; Park, Geun Il; Song, Kee Chan
2012-01-01
The new technology for isotopic fissile material contents assay is under development at KAERI using lead slowing down spectrometer(LSDS). LSDS is very sensitive to distinguish fission signals from each fissile isotope in spent and recycled fuel. The accumulation of spent fuel is current big issue. The amount of spent fuels will reach the maximum storage capacity of the pools soon. Therefore, an interim storage must be searched and it should be optimized in design by applying accurate fissile content. When the storage has taken effect, all the nuclear materials must be also specified and verified for safety, economics and management. Generally, the spent fuel from PWR has unburned ∼1 % U235, produced ∼0.5 % plutonium from decay chain, ∼3 % fission products, ∼ 0.1 % minor actinides (MA) and uranium remainder. About 1.5 % fissile materials still exist in the spent fuel. Therefore, for reutilization of fissile materials in spent fuel at SFR, resource material is produced through pyro process. Fissile material contents in resource material must be analyzed before fabricating SFR fuel for reactor safety and economics. In assay of fissile content of spent fuel and recycled fuel, intense radiation background gives limitation on the direct analysis of fissile materials. However, LSDS is not influenced by such a radiation background in fissile assay. Based on the decided geometry setup, self shielding parameter was calculated at the fuel assay zone by introducing spent fuel or pyro produced nuclear material. When nuclear material is inserted into the assay area, the spent fuel assembly or pyro recycled fuel material perturbs the spatial distribution of the slowing down neutrons in lead and the prompt fast fission neutrons produced by fissile materials are also perturbed. The self shielding factor is interpreted as that how much of absorption is created inside the fuel area when it is in the lead. Self shielding effect provides a non-linear property in the isotopic
Situations of potential exposure in self-shielding electron accelerators
International Nuclear Information System (INIS)
Rios, D.A.S.; Rios, P.B.; Sordi, G.M.A.A.; Carneiro, J.C.G.G.
2017-01-01
The study discusses situations in the industrial environment that may lead to potential exposure of Occupationally Exposed Individuals and Public Individuals in self-shielding electron accelerators. Although these exposure situations are unlikely, simulation exercises can lead to improvements in the operating procedure as well as suggest changes in production line design in order to increase radiation protection at work. These studies can also be used in training and demonstrate a solid application of the ALARA principle in the daily activities of radiative installations
Validation of calculated self-shielding factors for Rh foils
Jaćimović, R.; Trkov, A.; Žerovnik, G.; Snoj, L.; Schillebeeckx, P.
2010-10-01
Rhodium foils of about 5 mm diameter were obtained from IRMM. One foil had thickness of 0.006 mm and three were 0.112 mm thick. They were irradiated in the pneumatic transfer system and in the carousel facility of the TRIGA reactor at the Jožef Stefan Institute. The foils were irradiated bare and enclosed in small cadmium boxes (about 2 g weight) of 1 mm thickness to minimise the perturbation of the local neutron flux. They were co-irradiated with 5 mm diameter and 0.2 mm thick Al-Au (0.1%) alloy monitor foils. The resonance self-shielding corrections for the 0.006 and 0.112 mm thick samples were calculated by the Monte Carlo simulation and amount to about 10% and 60%, respectively. The consistency of measurements confirmed the validity of self-shielding factors. Trial estimates of Q0 and k0 factors for the 555.8 keV gamma line of 104Rh were made and amount to 6.65±0.18 and (6.61±0.12)×10 -2, respectively.
Insufficient self-shielding correction in VITAMIN-B6
International Nuclear Information System (INIS)
Konno, Chikara; Ochiai, Kentaro; Ohnishi, Seiki
2011-01-01
We carried out a simple benchmark calculation test with a multigroup cross-section library VITAMIN-B6 generated from ENDF/B-VI. The model of this test consisted of an iron sphere of 1 m in radius with an isotropic 20 MeV neutron source in the center. Neutron spectra in the sphere were calculated with an Sn code ANISN and VITAMIN-B6 or FENDL/MG-1.1. A calculation with MCNP and ENDF/B-VI was carried out as a reference. The neutron spectra with ANISN and FENDL/MG-1.1 agreed with those with MCNP, while those with ANISN and VITAMIN-B6 were at most 50% different from those with MCNP. We uncovered that the discrepancy came from insufficient self-shielding correction due to the followings; 1) The smallest background cross section of 56 Fe in VITAMIN-B6 is 1. 2) The weighting flux used in generating VITAMIN-B6 is not adequate. VITAMIN-B6 should be revised for adequate self-shielding correction. (author)
Resonance Self-Shielding Methodologies in SCALE 6
International Nuclear Information System (INIS)
Williams, Mark L.
2011-01-01
SCALE 6 includes several problem-independent multigroup (MG) libraries that were processed from the evaluated nuclear data file ENDF/B using a generic flux spectrum. The library data must be self-shielded and corrected for problem-specific spectral effects for use in MG neutron transport calculations. SCALE 6 computes problem-dependent MG cross sections through a combination of the conventional Bondarenko shielding-factor method and a deterministic continuous-energy (CE) calculation of the fine-structure spectra in the resolved resonance and thermal energy ranges. The CE calculation can be performed using an infinite medium approximation, a simplified two-region method for lattices, or a one-dimensional discrete ordinates transport calculation with pointwise (PW) cross-section data. This paper describes the SCALE-resonance self-shielding methodologies, including the deterministic calculation of the CE flux spectra using PW nuclear data and the method for using CE spectra to produce problem-specific MG cross sections for various configurations (including doubly heterogeneous lattices). It also presents results of verification and validation studies.
Resonance self-shielding calculation with regularized random ladders
Energy Technology Data Exchange (ETDEWEB)
Ribon, P.
1986-01-01
The straightforward method for calculation of resonance self-shielding is to generate one or several resonance ladders, and to process them as resolved resonances. The main drawback of Monte Carlo methods used to generate the ladders, is the difficulty of reducing the dispersion of data and results. Several methods are examined, and it is shown how one (a regularized sampling method) improves the accuracy. Analytical methods to compute the effective cross-section have recently appeared: they are basically exempt from dispersion, but are inevitably approximate. The accuracy of the most sophisticated one is checked. There is a neutron energy range which is improperly considered as statistical. An examination is presented of what happens when it is treated as statistical, and how it is possible to improve the accuracy of calculations in this range. To illustrate the results calculations have been performed in a simple case: nucleus /sup 238/U, at 300 K, between 4250 and 4750 eV.
The resonance self-shielding calculation with regularized random ladders
International Nuclear Information System (INIS)
Ribon, P.
1986-01-01
The straightforward method for calculation of resonance self-shielding is to generate one or several resonance ladders, and to process them as resolved resonances. The main drawback of Monte Carlo methods used to generate the ladders, is the difficulty of reducing the dispersion of data and results. Several methods are examined, and it is shown how one (a regularized sampling method) improves the accuracy. Analytical methods to compute the effective cross-section have recently appeared: they are basically exempt from dispersion, but are inevitably approximate. The accuracy of the most sophisticated one is checked. There is a neutron energy range which is improperly considered as statistical. An examination is presented of what happens when it is treated as statistical, and how it is possible to improve the accuracy of calculations in this range. To illustrate the results calculations have been performed in a simple case: nucleus 238 U, at 300 K, between 4250 and 4750 eV. (author)
URR-PACK: Calculating Self-Shielding in the Unresolved Resonance Energy Range
International Nuclear Information System (INIS)
Cullen, Dermott E.; Trkov, Andrej
2016-07-01
This report describes HOW to calculate self-shielding in the unresolved resonance region (URR), in terms of the computer codes we provide to allow a user to do these calculations himself. Here we only describe HOW to calculate; a longer companion report describes in detail WHY it is necessary to include URR self-shielding.
The problem of resonance self-shielding effect in neutron multigroup calculations
International Nuclear Information System (INIS)
Wang Qingming; Huang Jinghua
1991-01-01
It is not allowed to neglect the resonance self-shielding effect in hybrid blanket and fast reactor neutron designs. The authors discussed the importance as well as the method of considering the resonance self-shielding effect in hybrid blanket and fast reactor neutron multigroup calculations
International Nuclear Information System (INIS)
Nasrabadi, M.N.; Mohammadi, A.; Jalali, M.
2009-01-01
In this paper bulk sample prompt gamma neutron activation analysis (BSPGNAA) was applied to aqueous sample analysis using a relative method. For elemental analysis of an unknown bulk sample, gamma self-shielding coefficient was required. Gamma self-shielding coefficient of unknown samples was estimated by an experimental method and also by MCNP code calculation. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the gamma self-shielding within the sample volume is required.
Self-shielding for thick slabs in a converging neutron beam
Mildner, D F R
1999-01-01
We have previously given a correction to the neutron self-shielding for a thin slab to account for the increased average path length through the slab when irradiated in a converging neutron beam. This expression overstates the case for the self-shielding for a thick (or highly absorbing) slab. We give a better approximation to the increase in effective shielding correction for a slab placed in a converging neutron beam. It is negligible at large absorption mean free paths. (author)
Self-shielding models of MICROX-2 code: Review and updates
International Nuclear Information System (INIS)
Hou, J.; Choi, H.; Ivanov, K.N.
2014-01-01
Highlights: • The MICROX-2 code has been improved to expand its application to advanced reactors. • New fine-group cross section libraries based on ENDF/B-VII have been generated. • Resonance self-shielding and spatial self-shielding models have been improved. • The improvements were assessed by a series of benchmark calculations against MCNPX. - Abstract: The MICROX-2 is a transport theory code that solves for the neutron slowing-down and thermalization equations of a two-region lattice cell. The MICROX-2 code has been updated to expand its application to advanced reactor concepts and fuel cycle simulations, including generation of new fine-group cross section libraries based on ENDF/B-VII. In continuation of previous work, the MICROX-2 methods are reviewed and updated in this study, focusing on its resonance self-shielding and spatial self-shielding models for neutron spectrum calculations. The improvement of self-shielding method was assessed by a series of benchmark calculations against the Monte Carlo code, using homogeneous and heterogeneous pin cell models. The results have shown that the implementation of the updated self-shielding models is correct and the accuracy of physics calculation is improved. Compared to the existing models, the updates reduced the prediction error of the infinite multiplication factor by ∼0.1% and ∼0.2% for the homogeneous and heterogeneous pin cell models, respectively, considered in this study
Characteristic Determination Of Self Shielding Factor And Cadmium Ratio Of Cylindrical Probe
International Nuclear Information System (INIS)
Hamzah, Amir; Budi R, Ita; Pinem, Suriam
1996-01-01
Determination of thermal, epithermal and total self shielding factor and cadmium ratio of cylindrical probe has been done by measurement and calculation. Self shielding factor can be determined by dividing probe activity to Al-alloy probe activity. Due to the lack of cylindrical probe made of Al-alloy, self shielding factor can be determined by parabolic extrapolation of measured activities to 0 cm radius to divide those activities. Theoretically, self shielding factor can be determined by making numerical solution of two dimensional integral equations using Romberg method. To simplify, the calculation is based on single collision theory with the assumption of monoenergetic neutron and isotropic distribution. For gold cylindrical probe, the calculation results are quite close to the measurement one with the relative discrepancy for activities, cadmium ratio and self shielding factor of bare probe are less then 11.5%, 3,5% and 1.5% respectively. The program can be used for the calculation of other kinds of cylindrical probes. Due to dependency to radius, cylindrical probe made of copper has the best characteristic of self shielding factor and cadmium ratio
Self-shielding of hydrogen in the IGM during the epoch of reionization
Chardin, Jonathan; Kulkarni, Girish; Haehnelt, Martin G.
2018-04-01
We investigate self-shielding of intergalactic hydrogen against ionizing radiation in radiative transfer simulations of cosmic reionization carefully calibrated with Lyα forest data. While self-shielded regions manifest as Lyman-limit systems in the post-reionization Universe, here we focus on their evolution during reionization (redshifts z = 6-10). At these redshifts, the spatial distribution of hydrogen-ionizing radiation is highly inhomogeneous, and some regions of the Universe are still neutral. After masking the neutral regions and ionizing sources in the simulation, we find that the hydrogen photoionization rate depends on the local hydrogen density in a manner very similar to that in the post-reionization Universe. The characteristic physical hydrogen density above which self-shielding becomes important at these redshifts is about nH ˜ 3 × 10-3 cm-3, or ˜20 times the mean hydrogen density, reflecting the fact that during reionization photoionization rates are typically low enough that the filaments in the cosmic web are often self-shielded. The value of the typical self-shielding density decreases by a factor of 3 between redshifts z = 3 and 10, and follows the evolution of the average photoionization rate in ionized regions in a simple fashion. We provide a simple parameterization of the photoionization rate as a function of density in self-shielded regions during the epoch of reionization.
New Improvements in Mixture Self-Shielding Treatment with APOLLO2 Code
International Nuclear Information System (INIS)
Coste-Delclaux, M.
2006-01-01
Full text of the presentation follows: APOLLO2 is a modular multigroup transport code developed at the CEA in Saclay (France). Previously, the self-shielding module could only treat one resonant isotope mixed with moderator isotopes. Consequently, the resonant mixture self-shielding treatment was an iterative one. Each resonant isotope of the mixture was treated separately, the other resonant isotopes of the mixture being then considered as moderator isotopes, that is to say non-resonant isotopes. This treatment could be iterated. Recently, we have developed a new method that consists in treating the resonant mixture as a unique entity. A main feature of APOLLO2 self-shielding module is that some implemented models are very general and therefore very powerful and versatile. We can give, as examples, the use of probability tables in order to describe the microscopic cross-section fluctuations or the TR slowing-down model that can deal with any resonance shape. The self-shielding treatment of a resonant mixture was developed essentially thanks to these two models. The goal of this paper is to describe the improvements on the self-shielding treatment of a resonant mixture and to present, as an application, the calculation of the ATRIUM-10 BWR benchmark. We will conclude by some prospects on remaining work in the self-shielding domain. (author)
Computing Moment-Based Probability Tables for Self-Shielding Calculations in Lattice Codes
International Nuclear Information System (INIS)
Hebert, Alain; Coste, Mireille
2002-01-01
As part of the self-shielding model used in the APOLLO2 lattice code, probability tables are required to compute self-shielded cross sections for coarse energy groups (typically with 99 or 172 groups). This paper describes the replacement of the multiband tables (typically with 51 subgroups) with moment-based tables in release 2.5 of APOLLO2. An improved Ribon method is proposed to compute moment-based probability tables, allowing important savings in CPU resources while maintaining the accuracy of the self-shielding algorithm. Finally, a validation is presented where the absorption rates obtained with each of these techniques are compared with exact values obtained using a fine-group elastic slowing-down calculation in the resolved energy domain. Other results, relative to the Rowland's benchmark and to three assembly production cases, are also presented
New improvements in the self-shielding formalism of the Apollo-2 code
International Nuclear Information System (INIS)
Coste, M.; Tellier, H.; Ribon, P.; Raepsaet, V.; Van der Gucht, C.
1993-01-01
One important modelization of a transport code working on a coarse energy mesh is the self-shielding. The French transport code APPOLO 2, developed at the Commissariat a l'Energie Atomique, uses a self-shielding formalism based on a double equivalence. First a homogenization gives the reaction rates in a heterogeneous geometry, and then a multigroup equivalence gives, once the reaction rates are known, the self-shielded cross-sections. The homogenization is a very sensitive part because it is the one which requires physical modelizations. We have added a new model which allows us to treat numerous narrow resonances statistically distributed in the same group of the multigroup mesh. It is important to notice that for a narrow resonance isolated in a group, that new model is equivalent to the previous narrow resonance model (NR)
Calculation of the electron trajectory for 200 kV self-shielded electron accelerator
International Nuclear Information System (INIS)
Wang Shuiqing
2000-01-01
In order to calculate the electron trajectory of 200 kV self-shielded electron accelerator, the electric field is calculated with a TRAJ program. In this program, following electron track mash points one by one, the electron beam trajectories are calculated. Knowing the effect of grid voltage on electron optics and gaining grid voltage focusing effect in the various energy grades, the authors have gained scientific basis for adjusting grid voltage, and also accumulated a wealth of experience for designing self-shielded electron accelerator or electron curtain in future
SUBGR: A Program to Generate Subgroup Data for the Subgroup Resonance Self-Shielding Calculation
Energy Technology Data Exchange (ETDEWEB)
Kim, Kang Seog [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
2016-06-06
The Subgroup Data Generation (SUBGR) program generates subgroup data, including levels and weights from the resonance self-shielded cross section table as a function of background cross section. Depending on the nuclide and the energy range, these subgroup data can be generated by (a) narrow resonance approximation, (b) pointwise flux calculations for homogeneous media; and (c) pointwise flux calculations for heterogeneous lattice cells. The latter two options are performed by the AMPX module IRFFACTOR. These subgroup data are to be used in the Consortium for Advanced Simulation of Light Water Reactors (CASL) neutronic simulator MPACT, for which the primary resonance self-shielding method is the subgroup method.
International Nuclear Information System (INIS)
Patino, N.E.; Abbate, M.J.; Sbaffoni, M.M.
1990-01-01
The procedures employed in the treatment of the resonance shielding effect have been identified as one of the causes of the large discrepancies found in the neutronic calculation of high conversion light water reactors (HCLWRs), indicating the need for a revision of the self-shielding procedures employed. In this work some well known techniques applied in HCLWR self-shielding calculations are evaluated; the study involves the comparison of methods for the generation of group constants, the analysis of the impact of considering some isotopes as infinitely diluted and the evaluation of the usual approximations utilized for the treatment of heterogeneities
Self-shielding effect in unresolved resonance data in JENDL-4.0
International Nuclear Information System (INIS)
Konno, Chikara; Takakura, Kosuke; Ochiai, Kentaro; Sato, Satoshi; Kato, Yoshinari
2012-01-01
At International Conference on Nuclear Data for Science and Technology in 2007 we pointed out that most of unresolved resonance data in JENDL-3.3 have a problem related to self-shielding correction. Here with a simple calculation model we have investigated whether the latest JENDL, JENDL-4.0, was improved for the problem or not. The results suggest that unresolved resonance data in JENDL-4.0 have no problem, but it seems that self-shielding effects for the unresolved resonance data in JENDL-4.0 are too large. New benchmark experiments for unresolved resonance data are strongly recommended in order to verify unresolved resonance data. (author)
International Nuclear Information System (INIS)
Cullen, D.E.
1978-01-01
Bonderenko self-shielded cross sections and multiband parameters from the Lawrence Livermore Laboratory Evaluated-Nuclear-Data Library (ENDL) as of July 4, 1978 are presented. These data include total, elastic, capture, and fission cross sections in the TART 175 group structure. Multiband parameters are listed. Bonderenko self-shielded cross section and the multiband parameters are presented on microfiche
Theoretical evaluation of self-shielding factors due to scattering resonances in foils
International Nuclear Information System (INIS)
Selander, W.N.
1960-06-01
A semi-analytical method is given for evaluating self-shielding factors for activation measurements which use thin foils having neutron scattering resonances. The energy loss by scattering in the foil is taken into account. The energy-dependent neutron angular distribution is expanded as a double series, the coefficients of which are (energy dependent) solutions of an infinite set of coupled integral equations. These are truncated in some suitable manner and solved numerically. The leading term of the series is proportional to the average, or effective flux in the activation sample. The product of this terra and the neutron capture cross-section is integrated numerically over the resonance to give the resonance self-shielding correction. Figure 4 shows resonance self-shielding factors derived in this mariner for the 132ev resonance in Co-59 and figure 5 shows similar results for the two Mn-55 resonances at 337ev and 1080ev. Self-shielding factors for 1/v capture are not significantly different from unity. (author)
License Application Design Selection Feature Report: Waste Package Self Shielding Design Feature 13
International Nuclear Information System (INIS)
Tang, J.S.
2000-01-01
In the Viability Assessment (VA) reference design, handling of waste packages (WPs) in the emplacement drifts is performed remotely, and human access to the drifts is precluded when WPs are present. This report will investigate the feasibility of using a self-shielded WP design to reduce the radiation levels in the emplacement drifts to a point that, when coupled with ventilation, will create an acceptable environment for human access. This provides the benefit of allowing human entry to emplacement drifts to perform maintenance on ground support and instrumentation, and carry out performance confirmation activities. More direct human control of WP handling and emplacement operations would also be possible. However, these potential benefits must be weighed against the cost of implementation, and potential impacts on pre- and post-closure performance of the repository and WPs. The first section of this report will provide background information on previous investigations of the self-shielded WP design feature, summarize the objective and scope of this document, and provide quality assurance and software information. A shielding performance and cost study that includes several candidate shield materials will then be performed in the subsequent section to allow selection of two self-shielded WP design options for further evaluation. Finally, the remaining sections will evaluate the impacts of the two WP self-shielding options on the repository design, operations, safety, cost, and long-term performance of the WPs with respect to the VA reference design
Enhancement of thermal neutron self-shielding in materials surrounded by reflectors
International Nuclear Information System (INIS)
Cornelia Chilian; Gregory Kennedy
2012-01-01
Materials containing from 41 to 1124 mg chlorine and surrounded by polyethylene containers of various thicknesses, from 0.01 to 5.6 mm, were irradiated in a research reactor neutron spectrum and the 38 Cl activity produced was measured as a function of polyethylene reflector thickness. For the material containing the higher amount of chlorine, the 38 Cl specific activity decreased with increasing reflector thickness, indicating increased neutron self-shielding. It was found that the amount of neutron self-shielding increased by as much as 52% with increasing reflector thickness. This is explained by neutrons which have exited the material subsequently reflecting back into it and thus increasing the total mean path length in the material. All physical and empirical models currently used to predict neutron self-shielding have ignored this effect and need to be modified. A method is given for measuring the adjustable parameter of a self-shielding model for a particular sample size and combination of neutron reflectors. (author)
Success and prospects for low energy, self-shielded electron beam accelerators
International Nuclear Information System (INIS)
Laeuppi, U.V.
1988-01-01
The advantages of self-shielded, low energy, electron beam accelerators for electron beam processing are described. Applications of these accelerators for cross-linking plastic films, drying of coated materials and printing inks and for curing processes are discussed. (U.K.)
Advanced resonance self-shielding method for gray resonance treatment in lattice physics code GALAXY
International Nuclear Information System (INIS)
Koike, Hiroki; Yamaji, Kazuya; Kirimura, Kazuki; Sato, Daisuke; Matsumoto, Hideki; Yamamoto, Akio
2012-01-01
A new resonance self-shielding method based on the equivalence theory is developed for general application to the lattice physics calculations. The present scope includes commercial light water reactor (LWR) design applications which require both calculation accuracy and calculation speed. In order to develop the new method, all the calculation processes from cross-section library preparation to effective cross-section generation are reviewed and reframed by adopting the current enhanced methodologies for lattice calculations. The new method is composed of the following four key methods: (1) cross-section library generation method with a polynomial hyperbolic tangent formulation, (2) resonance self-shielding method based on the multi-term rational approximation for general lattice geometry and gray resonance absorbers, (3) spatially dependent gray resonance self-shielding method for generation of intra-pellet power profile and (4) integrated reaction rate preservation method between the multi-group and the ultra-fine-group calculations. From the various verifications and validations, applicability of the present resonance treatment is totally confirmed. As a result, the new resonance self-shielding method is established, not only by extension of a past concentrated effort in the reactor physics research field, but also by unification of newly developed unique and challenging techniques for practical application to the lattice physics calculations. (author)
International Nuclear Information System (INIS)
Tzika, F.; Stamatelatos, I.E.
2004-01-01
Thermal neutron self-shielding within large samples was studied using the Monte Carlo neutron transport code MCNP. The code enabled a three-dimensional modeling of the actual source and geometry configuration including reactor core, graphite pile and sample. Neutron flux self-shielding correction factors derived for a set of materials of interest for large sample neutron activation analysis are presented and evaluated. Simulations were experimentally verified by measurements performed using activation foils. The results of this study can be applied in order to determine neutron self-shielding factors of unknown samples from the thermal neutron fluxes measured at the surface of the sample
International Nuclear Information System (INIS)
Nakagawa, Masayuki; Ishiguro, Yukio; Tokuno, Yukio.
1978-01-01
The self-shielding factors for elastic removal cross sections of light and medium weight nuclides were calculated for the parameter, σ 0 within the conventional concept of the group constant sets. The numerical study were performed for obtaining a simple and accurate method. The present results were compared with the exact values and the conventional ones, and shown to be remarkably improved. It became apparent that the anisotropy of the elastic scattering did not affect to the self-shielding factors though it did to the infinite dilution cross sections. With use of the present revised set, the neutron flux were calculated in an iron medium and in a prototype FBR and compared with those by the fine spectrum calculations and the conventional set. The present set showed the considerable improvement in the vicinity of the large resonance regions of sodium, iron and oxygen. (auth.)
REPOSITORY LAYOUT SUPPORTING DESIGN FEATURE NO.13 - WASTE PACKAGE SELF SHIELDING
International Nuclear Information System (INIS)
Owen, J.
1999-01-01
The objective of this analysis is to develop a repository layout, for Feature No. 13, that will accommodate self-shielding waste packages (WP) with an areal mass loading of 25 metric tons of uranium per acre (MTU/acre). The scope of this analysis includes determination of the number of emplacement drifts, amount of emplacement drift excavation required, and a preliminary layout for illustrative purposes
International Nuclear Information System (INIS)
Noorddin Ibrahim; Rosnie Akang
2009-01-01
Full text: One of the major problems encountered during the irradiation of large inhomogeneous samples in performing activation analysis using neutron is the perturbation of the neutron field due to absorption and scattering of neutron within the sample as well as along the neutron guide in the case of prompt gamma activation analysis. The magnitude of this perturbation shown by self-shielding coefficient and flux depression depend on several factors including the average neutron energy, the size and shape of the sample, as well as the macroscopic absorption cross section of the sample. In this study, we use Monte Carlo N-Particle codes to simulate the variation of neutron self-shielding coefficient and thermal flux depression factor as a function of the macroscopic thermal absorption cross section. The simulation works was carried out using the high performance computing facility available at UTM while the experimental work was performed at the tangential beam port of Reactor TRIGA PUSPATI, Malaysia Nuclear Agency. The neutron flux measured along the beam port is found to be in good agreement with the simulated data. Our simulation results also reveal that total flux perturbation factor decreases as the value of absorption increases. This factor is close to unity for low absorbing sample and tends towards zero for strong absorber. In addition, sample with long mean chord length produces smaller flux perturbation than the shorter mean chord length. When comparing both the graphs of self-shielding factor and total disturbance, we can conclude that the total disturbance of the thermal neutron flux on the large samples is dominated by the self-shielding effect. (Author)
Self-shielding characteristics of aqueous self-cooled blankets for next generation fusion devices
International Nuclear Information System (INIS)
Pelloni, S.; Cheng, E.T.; Embrechts, M.J.
1987-11-01
The present study examines self-shielding characteristics for two aqueous self-cooled tritium producing driver blankets for next generation fusion devices. The aqueous Self-Cooled Blanket concept (ASCB) is a very simple blanket concept that relies on just structural material and coolant. Lithium compounds are dissolved in water to provide for tritium production. An ASCB driver blanket would provide a low technology and low temperature environment for blanket test modules in a next generation fusion reactor. The primary functions of such a blanket would be shielding, energy removal and tritium production. One driver blanket considered in this study concept relates to the one proposed for the Next European Torus (NET), while the second concept is indicative for the inboard shield design for the Engineering Test Reactor proposed by the USA (TIBER II/ETR). The driver blanket for NET is based on stainless steel for the structural material and aqueous solution, while the inboard shielding blanket for TIBER II/ETR is based on a tungsten/aqueous solution combination. The purpose of this study is to investigate self-shielding and heterogeneity effects in aqueous self-cooled blankets. It is found that no significant gains in tritium breeding can be achieved in the stainless steel blanket if spatial and energy self-shielding effects are considered, and the heterogeneity effects are also insignificant. The tungsten blanket shows a 5 percent increase in tritium production in the shielding blanket when energy and spatial self-shielding effects are accounted for. However, the tungsten blanket shows a drastic increase in the tritium breeding ratio due to heterogeneity effects. (author) 17 refs., 9 figs., 9 tabs
International Nuclear Information System (INIS)
Downar, T.
2009-01-01
The overall objective of the work here has been to eliminate the approximations used in current resonance treatments by developing continuous energy multi-dimensional transport calculations for problem dependent self-shielding calculations. The work here builds on the existing resonance treatment capabilities in the ORNL SCALE code system. The overall objective of the work here has been to eliminate the approximations used in current resonance treatments by developing continuous energy multidimensional transport calculations for problem dependent self-shielding calculations. The work here builds on the existing resonance treatment capabilities in the ORNL SCALE code system. Specifically, the methods here utilize the existing continuous energy SCALE5 module, CENTRM, and the multi-dimensional discrete ordinates solver, NEWT to develop a new code, CENTRM( ) NEWT. The work here addresses specific theoretical limitations in existing CENTRM resonance treatment, as well as investigates advanced numerical and parallel computing algorithms for CENTRM and NEWT in order to reduce the computational burden. The result of the work here will be a new computer code capable of performing problem dependent self-shielding analysis for both existing and proposed GENIV fuel designs. The objective of the work was to have an immediate impact on the safety analysis of existing reactors through improvements in the calculation of fuel temperature effects, as well as on the analysis of more sophisticated GENIV/NGNP systems through improvements in the depletion/transmutation of actinides for Advanced Fuel Cycle Initiatives.
AUTOSECOL: an automatic calculation of the self-shielding of heavy isotope resonances
International Nuclear Information System (INIS)
Grandotto-Biettoli, Marc.
The formalism is based on separating both types of resonance effects: local energy effects creating a fine structure in the flux, and bulk effects resulting in a slow variation in the flux. Effective reaction rates are defined that, used as tables in a multigroup calculation of cells with a large pitch in regard to resonance widths, allow an exact account of the dependence of the effective integral upon fast variations in the flux. These tables are used to introduce this phenomenon of resonance self-shielding in the multigroup Apollo program for solving the neutron transport equation, they are derived from nuclear data with using some parameters relating to the physical state of the resonant isotope inside the fuel medium. The AUTOSECOL system provides a library of effective reaction rates for taking account of the resonance self-shielding effect on the neutron flux in nuclear reactor cells. Its versatility in regard to the methods previously used for solving the same problem allows a rapid testing of the consequences of considering the self-shielding effect of new isotope resonances, a following up of the evolution in nuclear data evaluation, and rapidly studying the interest lying in new data. Results obtained with AUTOSECOL are compared with those obtained when using the SECOL code for computing the effective reaction rates of 235 U, 239 Pu, 107 Ag, 109 Ag, and 241 Pu [fr
Importance of self-shielding for improving sensitivity coefficients in light water nuclear reactors
International Nuclear Information System (INIS)
Foad, Basma; Takeda, Toshikazu
2014-01-01
Highlights: • A new method has been developed for calculating sensitivity coefficients. • This method is based on the use of infinite dilution cross-sections instead of effective cross-sections. • The change of self-shielding factor due to cross-section perturbation has been considered. • SRAC and SAINT codes are used for calculating improved sensitivities, while MCNP code has been used for verification. - Abstract: In order to perform sensitivity analyzes in light water reactors where self-shielding effect becomes important, a new method has been developed for calculating sensitivity coefficient of core characteristics relative to the infinite dilution cross-sections instead of the effective cross-sections. This method considers the change of the self-shielding factor due to cross-section perturbation for different nuclides and reactions. SRAC and SAINT codes are used to calculate the improved sensitivity; while the accuracy of the present method has been verified by MCNP code and good agreement has been found
International Nuclear Information System (INIS)
Zou Jun; He Zhaozhong; Zeng Qin; Qiu Yuefeng; Wang Minghuang
2010-01-01
A multigroup library HENDL2.1/SS (Hybrid Evaluated Nuclear Data Library/Self-Shielding) based on ENDF/B-VII.0 evaluate data has been generated using Bondarenko and flux calculator method for the correction of self-shielding effect of neutronics analyses. To validate the reliability of the multigroup library HENDL2.1/SS, transport calculations for fusion-fission hybrid system FDS-I were performed in this paper. It was verified that the calculations with the HENDL2.1/SS gave almost the same results with MCNP calculations and were better than calculations with the HENDL2.0/MG which is another multigroup library without self-shielding correction. The test results also showed that neglecting resonance self-shielding caused underestimation of the K eff , neutron fluxes and waste transmutation ratios in the multigroup calculations of FDS-I.
International Nuclear Information System (INIS)
Abe, Alfredo Y.; Santos, Adimir dos
1995-01-01
The present work summarizes the verification of the treatment of self-shielding based on Bondarenko method in HAMMER-TECHNION cell code for the Pu O 2 -U O 2 critical system using JENDL-3 nuclear data library. The results obtained are in excellent agreement with the original treatment of self-shielding employed by HAMMER-TECHNION cell code. (author). 9 refs, 1 fig, 9 tabs
An ''exact'' treatment of self-shielding and covers in neutron spectra determinations
International Nuclear Information System (INIS)
Griffin, P.J.; Kelly, J.G.
1995-01-01
Most neutron spectrum determination methodologies ignore self-shielding effects in dosimetry foils and treat covers with an exponential attenuation model. This work provides a quantitative analysis of the approximations in this approach. It also provides a methodology for improving the fidelity of the treatment of the dosimetry sensor response to a level consistent with the user's spectrum characterization approach. A library of correction functions for the energy-dependent sensor response has been compiled that addresses dosimetry foils/configurations in use at the Sandia National Laboratories Radiation Metrology Laboratory
Adaptive algorithms for a self-shielding wavelet-based Galerkin method
International Nuclear Information System (INIS)
Fournier, D.; Le Tellier, R.
2009-01-01
The treatment of the energy variable in deterministic neutron transport methods is based on a multigroup discretization, considering the flux and cross-sections to be constant within a group. In this case, a self-shielding calculation is mandatory to correct sections of resonant isotopes. In this paper, a different approach based on a finite element discretization on a wavelet basis is used. We propose adaptive algorithms constructed from error estimates. Such an approach is applied to within-group scattering source iterations. A first implementation is presented in the special case of the fine structure equation for an infinite homogeneous medium. Extension to spatially-dependent cases is discussed. (authors)
Self-Shielding Treatment to Perform Cell Calculation for Seed Furl In Th/U Pwr Using Dragon Code
Directory of Open Access Journals (Sweden)
Ahmed Amin El Said Abd El Hameed
2015-08-01
Full Text Available Time and precision of the results are the most important factors in any code used for nuclear calculations. Despite of the high accuracy of Monte Carlo codes, MCNP and Serpent, in many cases their relatively long computational time leads to difficulties in using any of them as the main calculation code. Usually, Monte Carlo codes are used only to benchmark the results. The deterministic codes, which are usually used in nuclear reactor’s calculations, have limited precision, due to the approximations in the methods used to solve the multi-group transport equation. Self- Shielding treatment, an algorithm that produces an average cross-section defined over the complete energy domain of the neutrons in a nuclear reactor, is responsible for the biggest error in any deterministic codes. There are mainly two resonance self-shielding models commonly applied: models based on equivalence and dilution and models based on subgroup approach. The fundamental problem with any self-shielding method is that it treats any isotope as there are no other isotopes with resonance present in the reactor. The most practical way to solve this problem is to use multi-energy groups (50-200 that are chosen in a way that allows us to use all major resonances without self-shielding. In this paper, we perform cell calculations, for a fresh seed fuel pin which is used in thorium/uranium reactors, by solving 172 energy group transport equation using the deterministic DRAGON code, for the two types of self-shielding models (equivalence and dilution models and subgroup models Using WIMS-D5 and DRAGON data libraries. The results are then tested by comparing it with the stochastic MCNP5 code. We also tested the sensitivity of the results to a specific change in self-shielding method implemented, for example the effect of applying Livolant-Jeanpierre Normalization scheme and Rimman Integration improvement on the equivalence and dilution method, and the effect of using Ribbon
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.
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.
Determination of self shielding factors and gamma attenuation effects for tree ring samples
International Nuclear Information System (INIS)
Dagistan Sahin; Kenan Uenlue
2012-01-01
Determination of tree ring chemistry using Neutron Activation Analysis (NAA) is part of an ongoing research between Penn State University (PSU) and Cornell University, The Malcolm and Carolyn Wiener Laboratory for Aegean and Near Eastern Dendrochronology. Tree-ring chemistry yields valuable data for environmental event signatures. These signatures are a complex function of elemental concentration. To be certain about concentration of signature elements, it is necessary to perform the measurements and corrections with the lowest error and maximum accuracy possible. Accurate and precise values of energy dependent neutron flux at dry irradiation tubes and detector efficiency for tree ring sample are calculated for Penn State Breazeale Reactor (PSBR). For the calculation of energy dependent and self shielding corrected neutron flux, detailed model of the TRIGA Mark III reactor at PSU with updated fuel compositions was prepared using the MCNP utility for reactor evolution (MURE) libraries. Dry irradiation tube, sample holder and sample were also included in the model. The thermal flux self-shielding correction factors due to the sample holder and sample for were calculated and verified with previously published values. The Geant-4 model of the gamma spectroscopy system, developed at Radiation Science and Engineering Center (RSEC), was improved and absolute detector efficiency for tree-ring samples was calculated. (author)
CREST : a computer program for the calculation of composition dependent self-shielded cross-sections
International Nuclear Information System (INIS)
Kapil, S.K.
1977-01-01
A computer program CREST for the calculation of the composition and temperature dependent self-shielded cross-sections using the shielding factor approach has been described. The code includes the editing and formation of the data library, calculation of the effective shielding factors and cross-sections, a fundamental mode calculation to generate the neutron spectrum for the system which is further used to calculate the effective elastic removal cross-sections. Studies to explore the sensitivity of reactor parameters to changes in group cross-sections can also be carried out by using the facility available in the code to temporarily change the desired constants. The final self-shielded and transport corrected group cross-sections can be dumped on cards or magnetic tape in a suitable form for their direct use in a transport or diffusion theory code for detailed reactor calculations. The program is written in FORTRAN and can be accommodated in a computer with 32 K work memory. The input preparation details, sample problem and the listing of the program are given. (author)
International Nuclear Information System (INIS)
Hebert, A.
1997-01-01
The subgroup method is used to compute self-shielded cross sections defined over coarse energy groups in the resolved energy domain. The validity of the subgroup approach was extended beyond the unresolved energy domain by partially taking into account correlation effects between the slowing-down source with the collision probability terms of the transport equation. This approach enables one to obtain a pure subgroup solution of the self-shielding problem without relying on any form of equivalence in dilution. Specific improvements are presented on existing subgroup methods: an N-term rational approximation for the fuel-to-fuel collision probability, a new Pade deflation technique for computing probability tables, and the introduction of a superhomogenization correction. The absorption rates obtained after self-shielding are compared with exact values obtained using an elastic slowing-down calculation where each resonance is modeled individually in the resolved energy domain
International Nuclear Information System (INIS)
Le Tellier, R.; Hebert, A.
2005-01-01
In this paper, we present the usage of the method of characteristics (MOC) with advanced self-shielding models for a fundamental lattice calculation on an ACR-type cell i.e. a cluster geometry with light water coolant and heavy water moderator. Comparison with the collision probability method (CP) show the consistency of the method of characteristics as implemented both in flux and self-shielding calculations. Acceleration techniques are tested in the different calculations and prove to be efficient. Comparisons with the Monte-Carlo code Tripoli4 show the advantage of a subgroup approach for self-shielding calculations : the difference in k eff is less than one standard deviation of the Tripoli4 calculation and in terms of total absorption rates, in the resolved resonances group, the maximum relative error is of the order of 3% localised in the most outer region of the central pin. (author)
GROUPIE2007, Bondarenko Self-Shielded Cross sections from ENDF/B
International Nuclear Information System (INIS)
2007-01-01
1 - Description of problem or function - GROUPIE reads evaluated data in ENDF/B Format and uses these to calculate unshielded group averaged Cross sections, Bondarenko self-shielded Cross sections, and multiband parameters. The program allows the user to specify arbitrary energy groups and an arbitrary energy-dependent neutron spectrum (weighting function). IAEA0849/15: This version include the updates up to January 30, 2007. Changes in ENDF/B-VII Format and procedures, as well as the evaluations themselves, make it impossible for versions of the ENDF/B pre-processing codes earlier than PREPRO 2007 (2007 Version) to accurately process current ENDF/B-VII evaluations. The present code can handle all existing ENDF/B-VI evaluations through release 8, which will be the last release of ENDF/B-VI. 2 - Modifications from previous versions: Groupie VERS. 2007-1 (Jan. 2007): checked against all ENDF/B-VII; increased page size from 120,000 to 600,000 points. 3 - Method of solution: All integrals are performed analytically; in no case is iteration or any approximate form of integration used. GROUPIE reads either the 0 deg. Kelvin Cross sections or the Doppler broadened Cross sections to calculate the self-shielded Cross sections and multiband parameters for 25 values of the 'background' Cross sections (representing the combined effects of all other isotopes and of leakage). 4 - Restrictions on the complexity of the problem: GROUPIE requires that the energy-dependent neutron spectrum and all Cross sections be given in tabular form, with linear interpolation between tabulated values. There is no limit to the size of the table used to describe the spectrum, so the spectrum may be described in as much detail as required. - If only unshielded averages are calculated, the program can handle up to 3000 groups. If self-shielded averages and/or multiband parameters are calculated, the program can handle up to 175 groups. These limits can easily be extended. - The program only uses the
International Nuclear Information System (INIS)
Chiba, Go; Tsuji, Masashi; Narabayashi, Tadashi
2014-01-01
In order to properly quantify fission reactor neutronics parameter uncertainties, we have to use covariance data and sensitivity profiles consistently. In the present paper, we establish two consistent methodologies for uncertainty quantification: a self-shielded cross section-based consistent methodology and an infinitely-diluted cross section-based consistent methodology. With these methodologies and the covariance data of uranium-238 nuclear data given in JENDL-3.3, we quantify uncertainties of infinite neutron multiplication factors of light water reactor and fast reactor fuel cells. While an inconsistent methodology gives results which depend on the energy group structure of neutron flux and neutron-nuclide reaction cross section representation, both the consistent methodologies give fair results with no such dependences.
Nuclear reactions and self-shielding effects of gamma-ray database for nuclear materials
Energy Technology Data Exchange (ETDEWEB)
Fujita, Mitsutane; Noda, Tetsuji [National Research Institute for Metals, Tsukuba, Ibaraki (Japan)
2001-03-01
A database for transmutation and radioactivity of nuclear materials is required for selection and design of materials used in various nuclear reactors. The database based on the FENDL/A-2.0 on the Internet and the additional data collected from several references has been developed in NRIM site of 'Data-Free-Way' on the Internet. Recently, the function predicted self-shielding effect of materials for {gamma}-ray was added to this database. The user interface for this database has been constructed for retrieval of necessary data and for graphical presentation of the relation between the energy spectrum of neutron and neutron capture cross section. It is demonstrated that the possibility of chemical compositional change and radioactivity in a material caused by nuclear reactions can be easily retrieved using a browser such as Netscape or Explorer. (author)
Design of a control system for self-shielded irradiators with remote access capability
International Nuclear Information System (INIS)
Iyengar, R.D.; Verma, P.B.; Prasad, V.V.S.S.; George, Jain R.; Das, Tripti; Deshmukh, D.K.
2001-01-01
With self-shielded irradiators like Gamma chambers, and Blood irradiators are being sold by BRIT to customers both within and outside the country, it has become necessary to improve the quality of service without increasing the overheads. The recent advances in the field of communications and information technology can be exploited for improving the quality of service to the customers. A state of the art control system with remote accessibility has been designed for these irradiators enhancing their performance. This will provide an easy access to these units wherever they might be located, through the Internet. With this technology it will now be possible to attend to the needs of the customers, as regards fault rectification, error debugging, system software update, performance testing, data acquisition etc. This will not only reduce the downtime of these irradiators but also reduce the overheads. (author)
Self-shielding flex-circuit drift tube, drift tube assembly and method of making
Jones, David Alexander
2016-04-26
The present disclosure is directed to an ion mobility drift tube fabricated using flex-circuit technology in which every other drift electrode is on a different layer of the flex-circuit and each drift electrode partially overlaps the adjacent electrodes on the other layer. This results in a self-shielding effect where the drift electrodes themselves shield the interior of the drift tube from unwanted electro-magnetic noise. In addition, this drift tube can be manufactured with an integral flex-heater for temperature control. This design will significantly improve the noise immunity, size, weight, and power requirements of hand-held ion mobility systems such as those used for explosive detection.
Nuclear reactions and self-shielding effects of gamma-ray database for nuclear materials
International Nuclear Information System (INIS)
Fujita, Mitsutane; Noda, Tetsuji
2001-01-01
A database for transmutation and radioactivity of nuclear materials is required for selection and design of materials used in various nuclear reactors. The database based on the FENDL/A-2.0 on the Internet and the additional data collected from several references has been developed in NRIM site of 'Data-Free-Way' on the Internet. Recently, the function predicted self-shielding effect of materials for γ-ray was added to this database. The user interface for this database has been constructed for retrieval of necessary data and for graphical presentation of the relation between the energy spectrum of neutron and neutron capture cross section. It is demonstrated that the possibility of chemical compositional change and radioactivity in a material caused by nuclear reactions can be easily retrieved using a browser such as Netscape or Explorer. (author)
Directory of Open Access Journals (Sweden)
GO CHIBA
2014-06-01
Full Text Available In order to properly quantify fission reactor neutronics parameter uncertainties, we have to use covariance data and sensitivity profiles consistently. In the present paper, we establish two consistent methodologies for uncertainty quantification: a self-shielded cross section-based consistent methodology and an infinitely-diluted cross section-based consistent methodology. With these methodologies and the covariance data of uranium-238 nuclear data given in JENDL-3.3, we quantify uncertainties of infinite neutron multiplication factors of light water reactor and fast reactor fuel cells. While an inconsistent methodology gives results which depend on the energy group structure of neutron flux and neutron-nuclide reaction cross section representation, both the consistent methodologies give fair results with no such dependences.
International Nuclear Information System (INIS)
Khattab, K.; Khamis, I.
2007-01-01
Measurement of the thermal self shielding coefficient ( Gth ) in the Syrian Miniature Neutron Source Reactor (MNSR) inner irradiation site using Dy foils is presented in this paper. The thermal self shielding coefficient is measured as a function of the foil thickness or numbers. The mathematical equation which calculates the average relative radioactivity (Bq/g) versus the foil number is found as well.
International Nuclear Information System (INIS)
Coste-Delclaux, M.; Aggery, A.; Huot, N.
2005-01-01
APOLLO2 is a modular multigroup transport code developed by Cea in Saclay. Until last year, the self-shielding module could only treat one resonant isotope mixed with moderator isotopes. Consequently, the resonant mixture self-shielding treatment was an iterative one. Each resonant isotope of the mixture was treated separately, the other resonant isotopes of the mixture being then considered as moderator isotopes, that is to say non-resonant isotopes. This treatment could be iterated. Last year, we have developed a new method that consists in treating the resonant mixture as a unique entity. A main feature of APOLLO2 self-shielding module is that some implemented models are very general and therefore very powerful and versatile. We can give, as examples, the use of probability tables in order to describe the microscopic cross-section fluctuations or the TR slowing-down model that can deal with any resonance shape. The self-shielding treatment of a resonant mixture was developed essentially thanks to these two models. The calculations of a simplified Jules Horowitz reactor using a Monte-Carlo code (TRIPOLI4) as a reference and APOLLO2 in its standard and improved versions, show that, as far as the effective multiplication factor is concerned, the mixture treatment does not bring an improvement, because the new treatment suppresses compensation between the reaction rate discrepancies. The discrepancy of 300 pcm that appears with the reference calculation is in accordance with the technical specifications of the Jules Horowitz reactor
Weld metal microstructures of hardfacing deposits produced by self-shielded flux-cored arc welding
International Nuclear Information System (INIS)
Dumovic, M.; Monaghan, B.J.; Li, H.; Norrish, J.; Dunne, D.P.
2015-01-01
The molten pool weld produced during self-shielded flux-cored arc welding (SSFCAW) is protected from gas porosity arising from oxygen and nitrogen by reaction ('killing') of these gases by aluminium. However, residual Al can result in mixed micro-structures of δ-ferrite, martensite and bainite in hardfacing weld metals produced by SSFCAW and therefore, microstructural control can be an issue for hardfacing weld repair. The effect of the residual Al content on weld metal micro-structure has been examined using thermodynamic modeling and dilatometric analysis. It is concluded that the typical Al content of about 1 wt% promotes δ-ferrite formation at the expense of austenite and its martensitic/bainitic product phase(s), thereby compromising the wear resistance of the hardfacing deposit. This paper also demonstrates how the development of a Schaeffler-type diagram for predicting the weld metal micro-structure can provide guidance on weld filler metal design to produce the optimum microstructure for industrial hardfacing applications.
Resonance self-shielding methodology of new neutron transport code STREAM
International Nuclear Information System (INIS)
Choi, Sooyoung; Lee, Hyunsuk; Lee, Deokjung; Hong, Ser Gi
2015-01-01
This paper reports on the development and verification of three new resonance self-shielding methods. The verifications were performed using the new neutron transport code, STREAM. The new methodologies encompass the extension of energy range for resonance treatment, the development of optimum rational approximation, and the application of resonance treatment to isotopes in the cladding region. (1) The extended resonance energy range treatment has been developed to treat the resonances below 4 eV of three resonance isotopes and shows significant improvements in the accuracy of effective cross sections (XSs) in that energy range. (2) The optimum rational approximation can eliminate the geometric limitations of the conventional approach of equivalence theory and can also improve the accuracy of fuel escape probability. (3) The cladding resonance treatment method makes it possible to treat resonances in cladding material which have not been treated explicitly in the conventional methods. These three new methods have been implemented in the new lattice physics code STREAM and the improvement in the accuracy of effective XSs is demonstrated through detailed verification calculations. (author)
A Wavelet-Based Finite Element Method for the Self-Shielding Issue in Neutron Transport
International Nuclear Information System (INIS)
Le Tellier, R.; Fournier, D.; Ruggieri, J. M.
2009-01-01
This paper describes a new approach for treating the energy variable of the neutron transport equation in the resolved resonance energy range. The aim is to avoid recourse to a case-specific spatially dependent self-shielding calculation when considering a broad group structure. This method consists of a discontinuous Galerkin discretization of the energy using wavelet-based elements. A Σ t -orthogonalization of the element basis is presented in order to make the approach tractable for spatially dependent problems. First numerical tests of this method are carried out in a limited framework under the Livolant-Jeanpierre hypotheses in an infinite homogeneous medium. They are mainly focused on the way to construct the wavelet-based element basis. Indeed, the prior selection of these wavelet functions by a thresholding strategy applied to the discrete wavelet transform of a given quantity is a key issue for the convergence rate of the method. The Canuto thresholding approach applied to an approximate flux is found to yield a nearly optimal convergence in many cases. In these tests, the capability of such a finite element discretization to represent the flux depression in a resonant region is demonstrated; a relative accuracy of 10 -3 on the flux (in L 2 -norm) is reached with less than 100 wavelet coefficients per group. (authors)
International Nuclear Information System (INIS)
Fitzpatrick, J.; Verrall, S.M.
1985-01-01
A comparative study has been carried out on the two philosophies for providing the radiological protection necessary for the transport and handling of packaged intermediate level wastes from their sites of origin to disposal. The two philosophies are self shielding and returnable shielding. The approach taken was to assess the cost and radiological impact differentials of two respective representative waste management procedures. The comparison indicated the merits of each procedure. As a consequence, a hybrid procedure was identified which combines the advantages of each philosophy. This hybrid procedure was used for further comparison. The results of the study indicate that the use of self shielded packages throughout will incur considerable extra expense and give only a small saving in radiological impact. (author)
International Nuclear Information System (INIS)
Ganesan, S.; Gopalakrishnan, V.; Ramanadhan, M.M.; Cullan, D.E.
1986-01-01
We investigate the effect of interpolation error in the pre-processing codes LINEAR, RECENT and SIGMA1 on calculations of self-shielding factors and their temperature derivatives. We consider the 2.0347 to 3.3546 keV energy region for 238 U capture, which is the NEACRP benchmark exercise on unresolved parameters. The calculated values of temperature derivatives of self-shielding factors are significantly affected by interpolation error. The sources of problems in both evaluated data and codes are identified and eliminated in the 1985 version of these codes. This paper helps to (1) inform code users to use only 1985 versions of LINEAR, RECENT, and SIGMA1 and (2) inform designers of other code systems where they may have problems and what to do to eliminate their problems. (author)
International Nuclear Information System (INIS)
Ganesan, S.; Gopalakrishnan, V.; Ramanadhan, M.M.; Cullen, D.E.
1985-01-01
The authors investigate the effect of interpolation error in the pre-processing codes LINEAR, RECENT and SIGMA1 on calculations of self-shielding factors and their temperature derivatives. They consider the 2.0347 to 3.3546 keV energy region for /sup 238/U capture, which is the NEACRP benchmark exercise on unresolved parameters. The calculated values of temperature derivatives of self-shielding factors are significantly affected by interpolation error. The sources of problems in both evaluated data and codes are identified and eliminated in the 1985 version of these codes. This paper helps to (1) inform code users to use only 1985 versions of LINEAR, RECENT, and SIGMA1 and (2) inform designers of other code systems where they may have problems and what to do to eliminate their problems
The new solid target system at UNAM in a self-shielded 11 MeV cyclotron
International Nuclear Information System (INIS)
Zarate-Morales, A.; Gaspar-Carcamo, R. E.; Lopez-Rodriguez, V.; Flores-Moreno, A.; Trejo-Ballado, F.; Avila-Rodriguez, Miguel A.
2012-01-01
A dual beam line (BL) self-shielded RDS 111 cyclotron for radionuclide production was installed at the School of Medicine of the National Autonomous University of Mexico in 2001. One of the BL’s was upgraded to Eclipse HP (Siemens) in 2008 and the second BL was recently upgraded (June 2011) to the same version with the option for the irradiation of solid targets for the production of metallic radioisotopes.
CO Self-Shielding as a Mechanism to Make 16O-Enriched Solids in the Solar Nebula
Directory of Open Access Journals (Sweden)
Joseph A. Nuth, III
2014-05-01
Full Text Available Photochemical self-shielding of CO has been proposed as a mechanism to produce solids observed in the modern, 16O-depleted solar system. This is distinct from the relatively 16O-enriched composition of the solar nebula, as demonstrated by the oxygen isotopic composition of the contemporary sun. While supporting the idea that self-shielding can produce local enhancements in 16O-depleted solids, we argue that complementary enhancements of 16O-enriched solids can also be produced via C16O-based, Fischer-Tropsch type (FTT catalytic processes that could produce much of the carbonaceous feedstock incorporated into accreting planetesimals. Local enhancements could explain observed 16O enrichment in calcium-aluminum-rich inclusions (CAIs, such as those from the meteorite, Isheyevo (CH/CHb, as well as in chondrules from the meteorite, Acfer 214 (CH3. CO self-shielding results in an overall increase in the 17O and 18O content of nebular solids only to the extent that there is a net loss of C16O from the solar nebula. In contrast, if C16O reacts in the nebula to produce organics and water then the net effect of the self-shielding process will be negligible for the average oxygen isotopic content of nebular solids and other mechanisms must be sought to produce the observed dichotomy between oxygen in the Sun and that in meteorites and the terrestrial planets. This illustrates that the formation and metamorphism of rocks and organics need to be considered in tandem rather than as isolated reaction networks.
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.
International Nuclear Information System (INIS)
Le Tellier, R.; Hebert, A.; Le Tellier, R.; Santamarina, A.; Litaize, O.
2008-01-01
Calculations based on the characteristics method and different self-shielding models are presented for 9 x 9 boiling water reactor (BWR) assemblies fully loaded with mixed-oxide (MOX) fuel. The geometry of these assemblies was recovered from the BASALA experimental program. We have focused our study on three configurations simulating the different voiding conditions that an assembly can undergo in a BWR pressure vessel. A parametric study was carried out with respect to the spatial discretization, the tracking parameters, and the anisotropy order. Comparisons with Monte Carlo calculations in terms of k eff , radiative capture, and fission rates were performed to validate the computational tools. The results are in good agreement between the stochastic and deterministic approaches. The mutual self-shielding model recently introduced within the framework of the Ribon extending self-shielding method appears to be useful for this type of assemblies. Indeed, in the calculation of these MOX benchmarks, the overlapping of resonances, especially between 238 U and 240 Pu, plays an important role due to the spectral strengthening of the flux as the voiding percentage is increased. The method of characteristics is shown to be adequate to perform accurate calculations handling a fine spatial discretization. (authors)
International Nuclear Information System (INIS)
Liu Haibo; Wu Yican; Zheng Shanliang; Zhang Chunzao
2004-01-01
Based on the Fusion-Driven Subcritical System (FDS-I), the 25 groups, 175 groups and 620 groups neutron nuclear data libraries with/without resonance self-shielding correction are made with the Njoy and Transx codes, and the K eff and reaction rates are calculated with the Anisn code. The conclusion indicates that the resonance self-shielding effect affects the reaction rates strongly. (authors)
International Nuclear Information System (INIS)
Karthikeyan, Ramamoorthy; Hebert, Alain
2008-01-01
A high conversion light water reactor lattice has been analysed using the code DRAGON Version4. This analysis was performed to test the performance of the advanced self-shielding models incorporated in DRAGON Version4. The self-shielding models are broadly classified into two groups - 'equivalence in dilution' and 'subgroup approach'. Under the 'equivalence in dilution' approach we have analysed the generalized Stamm'ler model with and without Nordheim model and Riemann integration. These models have been analysed also using the Livolant-Jeanpierre normalization. Under the 'subgroup approach', we have analysed Statistical self-shielding model based on physical probability tables and Ribon extended self-shielding model based on mathematical probability tables. This analysis will help in understanding the performance of advanced self-shielding models for a lattice that is tight and has a large fraction of fissions happening in the resonance region. The nuclear data for the analysis was generated in-house. NJOY99.90 was used for generating libraries in DRAGLIB format for analysis using DRAGON and A Compact ENDF libraries for analysis using MCNP5. The evaluated datafiles were chosen based on the recommendations of the IAEA Co-ordinated Research Project on the WIMS Library Update Project. The reference solution for the problem was obtained using Monte Carlo code MCNP5. It was found that the Ribon extended self-shielding model based on mathematical probability tables using correlation model performed better than all other models
CO Self-Shielding as a Mechanism to Make O-16 Enriched Solids in the Solar Nebula
Nuth, Joseph A. III; Johnson, Natasha M.; Hill, Hugh G. M.
2014-01-01
Photochemical self-shielding of CO has been proposed as a mechanism to produce solids observed in the modern, O-16 depleted solar system. This is distinct from the relatively O-16 enriched composition of the solar nebula, as demonstrated by the oxygen isotopic composition of the contemporary sun. While supporting the idea that self-shielding can produce local enhancements in O-16 depleted solids, we argue that complementary enhancements of O-16 enriched solids can also be produced via CO-16 based, Fischer-Tropsch type (FTT) catalytic processes that could produce much of the carbonaceous feedstock incorporated into accreting planetesimals. Local enhancements could explain observed O-16 enrichment in calcium-aluminum-rich inclusions (CAIs), such as those from the meteorite, Isheyevo (CH/CHb), as well as in chondrules from the meteorite, Acfer 214 (CH3). CO selfshielding results in an overall increase in the O-17 and O-18 content of nebular solids only to the extent that there is a net loss of CO-16 from the solar nebula. In contrast, if CO-16 reacts in the nebula to produce organics and water then the net effect of the self-shielding process will be negligible for the average oxygen isotopic content of nebular solids and other mechanisms must be sought to produce the observed dichotomy between oxygen in the Sun and that in meteorites and the terrestrial planets. This illustrates that the formation and metamorphism of rocks and organics need to be considered in tandem rather than as isolated reaction networks.
International Nuclear Information System (INIS)
Palma, Daniel A.; Goncalves, Alessandro C.; Martinez, Aquilino S.; Silva, Fernando C.
2008-01-01
The activation technique allows much more precise measurements of neutron intensity, relative or absolute. The technique requires the knowledge of the Doppler broadening function ψ(x,ξ) to determine the resonance self-shielding factors in the epithermal range G epi (τ,ξ). Two new analytical approximations for the Doppler broadening function ψ(x,ξ) are proposed. The approximations proposed are compared with other methods found in literature for the calculation of the ψ(x,ξ) function, that is, the 4-pole Pade method and the Frobenius method, when applied to the calculation of G epi (τ,ξ). The results obtained provided satisfactory accuracy. (authors)
Directory of Open Access Journals (Sweden)
Zeng Huilin
2014-10-01
Full Text Available In order to realize the automatic welding of pipes in a complex operation environment, an automatic welding system has been developed by use of all-position self-shielded flux cored wires due to their advantages, such as all-position weldability, good detachability, arc's stability, low incomplete fusion, no need for welding protective gas or protection against wind when the wind speed is < 8 m/s. This system consists of a welding carrier, a guide rail, an auto-control system, a welding source, a wire feeder, and so on. Welding experiments with this system were performed on the X-80 pipeline steel to determine proper welding parameters. The welding technique comprises root welding, filling welding and cover welding and their welding parameters were obtained from experimental analysis. On this basis, the mechanical properties tests were carried out on welded joints in this case. Results show that this system can help improve the continuity and stability of the whole welding process and the welded joints' inherent quality, appearance shape, and mechanical performance can all meet the welding criteria for X-80 pipeline steel; with no need for windbreak fences, the overall welding cost will be sharply reduced. Meanwhile, more positive proposals were presented herein for the further research and development of this self-shielded flux core wires.
Influence of preheating on API 5L-X80 pipeline joint welding with self shielded flux-cored wire
International Nuclear Information System (INIS)
Cooper, R.; Silva, J. H. F.; Trevisan, R. E.
2004-01-01
The present work refers to the characterization of API 5L-X80 pipeline joints welded with self-shielded flux cored wire. This process was evaluated under preheating conditions, with an uniform and steady heat input. All joints were welded in flat position (1G), with the pipe turning and the torch still. Tube dimensions were 762 mm in external diameter and 16 mm in thickness. Welds were applied on single V-groove, with six weld beads, along with three levels of preheating temperatures (room temperature, 100 degree centigree, 160 degree centigree). These temperatures were maintained as inter pass temperature. The filler metal E71T8-K6 with mechanical properties different from parent metal was used in under matched conditions. The weld characterization is presented according to the mechanical test results of tensile strength, hardness and impact test. The mechanical tests were conducted according to API 1104, AWS and ASTM standards. API 1104 and API 51 were used as screening criteria. According to the results obtained, it was possible to remark that it is appropriate to weld API 5L-X80 steel ducts with Self-shielded Flux Cored wires, in conformance to the API standards and no preheat temperature is necessary. (Author) 22 refs
International Nuclear Information System (INIS)
Li, J.; Nuenighoff; Pohl, C.; Allelein, H.J.
2010-01-01
The gas-cooled, high temperature reactor (HTR) represents a valuable option for the future development of nuclear technology, because of its excellent safety features. One main safety feature is the negative temperature coefficient which is due to the Doppler broadening of the (n,y) resonance absorption cross section. A second important effect is the spatial self-shielding due to the double heterogeneous geometry of a pebble bed reactor. At FZ-Juelich two reactor analysis codes have been developed: VSOP for core design and MGT for transient analysis. Currently an update of the nuclear cross section libraries to ENDF/B-VII.0 of both codes takes place. In order to take the temperature dependency as well as the spatial self-shielding into account the absorption cross sections σ (n,y) for the resonance absorbers like 232 Th and 238 U have to be provided as function of incident neutron energy, temperature and nuclide concentration. There are two reasons for choosing the Monte-Carlo approach to calculate group wise cross sections. First, the former applied ZUT-DGL code to generate the resonance cross section tables for MGT is so far not able to handle the new resonance description based on Reich-Moore instead of Single-level Breit-Wigner. Second, the rising interest in PuO 2 fuel motivated an investigation on the generation of group wise cross sections describing thermal resonances of 240 Pu and 242 Pu. (orig.)
International Nuclear Information System (INIS)
Nasrabadi, M.N.; Jalali, M.; Mohammadi, A.
2007-01-01
In this work thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing materials is studied using bulk sample prompt gamma neutron activation analysis (BSPGNAA) with the MCNP code. The code was used to perform three dimensional simulations of a neutron source, neutron detector and sample of various material compositions. The MCNP model was validated against experimental measurements of the neutron flux performed using a BF 3 detector. Simulations were performed to predict thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing solutes. In practice, the MCNP calculations are combined with experimental measurements of the relative thermal neutron flux over the sample's surface, with respect to a reference water sample, to derive the thermal neutron self-shielding within the sample. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the average thermal neutron flux within the sample volume is required
International Nuclear Information System (INIS)
Sudarshan, K.; Tripathi, R.; Nair, A.G.C.; Acharya, R.; Reddy, A.V.R.; Goswami, A.
2005-01-01
A simple method using an internal standard is proposed to correct for the self-shielding effect of B, Cd and Gd in a matrix. This would increase the linear dynamic range of PGNAA in analyzing samples containing these elements. The method is validated by analyzing synthetic samples containing large amounts of B, Cd, Hg and Gd, the elements having high neutron absorption cross-section, in aqueous solutions and solid forms. A simple Monte-Carlo simulation to find the extent of self-shielding in the matrix is presented. The method is applied to the analysis of titanium boride alloy containing large amount of boron. The satisfactory results obtained showed the efficacy of the method of correcting for the self-shielding effects in the sample
Kodama, Nao; Setoi, Ayana; Kose, Katsumi
2018-01-01
Spiral MRI sequences were developed for a 9.4T vertical standard bore (54 mm) superconducting magnet using unshielded and self-shielded gradient coils. Clear spiral images with 64-shot scan were obtained with the self-shielded gradient coil, but severe shading artifacts were observed for the spiral-scan images acquired with the unshielded gradient coil. This shading artifact was successfully corrected with a phase-correction technique using reference scans that we developed based on eddy current field measurements. We therefore concluded that spiral imaging sequences can be installed even for unshielded gradient coils if phase corrections are performed using the reference scans. PMID:28367906
Kodama, Nao; Setoi, Ayana; Kose, Katsumi
2018-04-10
Spiral MRI sequences were developed for a 9.4T vertical standard bore (54 mm) superconducting magnet using unshielded and self-shielded gradient coils. Clear spiral images with 64-shot scan were obtained with the self-shielded gradient coil, but severe shading artifacts were observed for the spiral-scan images acquired with the unshielded gradient coil. This shading artifact was successfully corrected with a phase-correction technique using reference scans that we developed based on eddy current field measurements. We therefore concluded that spiral imaging sequences can be installed even for unshielded gradient coils if phase corrections are performed using the reference scans.
International Nuclear Information System (INIS)
Mason, Grant W.; Spencer, Ross L.
2002-01-01
The 'self-shielding' m=1 diocotron mode in Malmberg-Penning traps has been known for over a decade to be unstable for finite length nonneutral plasmas with hollow density profiles. Early theoretical efforts were unsuccessful in accounting for the exponential growth and/or the magnitude of the growth rate. Recent theoretical work has sought to resolve the discrepancy either as a consequence of the shape of the plasma ends or as a kinetic effect resulting from a modified distribution function as a consequence of the protocol used to form the hollow profiles in experiments. We have investigated both of these finite length mechanisms in selected test cases using a three-dimensional particle-in-cell code that allows realistic treatment of shape and kinetic effects. We find that a persistent discrepancy of a factor of 2-3 remains between simulation and experimental values of the growth rate
Energy Technology Data Exchange (ETDEWEB)
Palma, Daniel A. [CEFET QUIMICA de Nilopolis/RJ, Rio de Janeiro (Brazil); Goncalves, Alessandro C.; Martinez, Aquilino S.; Silva, Fernando C. [COPPE/UFRJ - Programa de Engenharia Nuclear, Rio de Janeiro (Brazil)
2008-07-01
The activation technique allows much more precise measurements of neutron intensity, relative or absolute. The technique requires the knowledge of the Doppler broadening function psi(x,xi) to determine the resonance self-shielding factors in the epithermal range G{sub epi} (tau,xi). Two new analytical approximations for the Doppler broadening function psi(x,xi) are proposed. The approximations proposed are compared with other methods found in literature for the calculation of the psi(x,xi) function, that is, the 4-pole Pade method and the Frobenius method, when applied to the calculation of G{sub epi} (tau,xi). The results obtained provided satisfactory accuracy. (authors)
Self-shielding factors for TLD-600 and TLD-100 in an isotropic flux of thermal neutrons
International Nuclear Information System (INIS)
Horowitz, Y.S.; Dubi, A.; Ben Shahar, B.
1976-01-01
The applications of lithium fluoride thermoluminescent dosemeters in mixed n-γ environments, and the dependence of LiF-TL on linear energy transfer are both topics of current interest. Monte Carlo calculations have therefore been carried out to determine the thermal neutron absorption probability (and consequently the self-shielding factor) for an isotropic flux of neutrons impinging on different sized cylindrical samples of LiF TLD-100 and TLD-600. The calculations were performed for cylinders of radius up to 10 cm and heights of 0.1 to 1.5 cm. The Monte Carlo results were found to be significantly different from the analytic calculations for infinitely long cylinders, but, as expected, converged to the same value for (r/h) << 1. (U.K.)
Church, Cody; Mawko, George; Archambault, John Paul; Lewandowski, Robert; Liu, David; Kehoe, Sharon; Boyd, Daniel; Abraham, Robert; Syme, Alasdair
2018-02-01
Radiopaque microspheres may provide intraprocedural and postprocedural feedback during transarterial radioembolization (TARE). Furthermore, the potential to use higher resolution x-ray imaging techniques as opposed to nuclear medicine imaging suggests that significant improvements in the accuracy and precision of radiation dosimetry calculations could be realized for this type of therapy. This study investigates the absorbed dose kernel for novel radiopaque microspheres including contributions of both short and long-lived contaminant radionuclides while concurrently quantifying the self-shielding of the glass network. Monte Carlo simulations using EGSnrc were performed to determine the dose kernels for all monoenergetic electron emissions and all beta spectra for radionuclides reported in a neutron activation study of the microspheres. Simulations were benchmarked against an accepted 90 Y dose point kernel. Self-shielding was quantified for the microspheres by simulating an isotropically emitting, uniformly distributed source, in glass and in water. The ratio of the absorbed doses was scored as a function of distance from a microsphere. The absorbed dose kernel for the microspheres was calculated for (a) two bead formulations following (b) two different durations of neutron activation, at (c) various time points following activation. Self-shielding varies with time postremoval from the reactor. At early time points, it is less pronounced due to the higher energies of the emissions. It is on the order of 0.4-2.8% at a radial distance of 5.43 mm with increased size from 10 to 50 μm in diameter during the time that the microspheres would be administered to a patient. At long time points, self-shielding is more pronounced and can reach values in excess of 20% near the end of the range of the emissions. Absorbed dose kernels for 90 Y, 90m Y, 85m Sr, 85 Sr, 87m Sr, 89 Sr, 70 Ga, 72 Ga, and 31 Si are presented and used to determine an overall kernel for the
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)
International Nuclear Information System (INIS)
Marques, Andre Luis Ferreira; Ting, Daniel Kao Sun; Mendonca, Arlindo Gilson
1996-01-01
A calculation methodology of Flux Depression, Self-Shielding and Cadmium Factors is presented, using the ANISN code, for experiments conducted at the IPEN/MB-01 Research Reactor. The correction factors were determined considering thermal neutron flux and 0.125 e 0.250 mm diameter of 197 Au wires. (author)
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
Energy Technology Data Exchange (ETDEWEB)
Singh, Daya; Soltis, Patrick; Narayanan, Badri; Quintana, Marie; Fox, Jeff [The Lincoln Electric Company (United States)
2005-07-01
Self-shielded flux cored arc welding electrodes (FCAW-S) are ideal for outdoor applications, particularly open fabrication yards where high winds are a possibility. Development work was carried out on a FCAW-S electrode for welding 70 and 80 ksi yield strength base materials with a required minimum average Charpy V-Notch (CVN) absorbed energy value of 35 ft-lb at -40 deg F in the weld metal. The effect of Al, Mg, Ti, and Zr on CVN toughness was evaluated by running a Design of Experiments approach to systematically vary the levels of these components in the electrode fill and, in turn, the weld metal. These electrodes were used to weld simulated pipe joints. Over the range of compositions tested, 0.05% Ti in the weld metal was found to be optimum for CVN toughness. Ti also had a beneficial effect on the usable voltage range. Simulated offshore joints were welded to evaluate the effect of base metal dilution, heat input, and welding procedure on the toughness of weld metal. CVN toughness was again measured at -40 deg F on samples taken from the root and the cap pass regions. The root pass impact toughness showed strong dependence on the base metal dilution and the heat input used to weld the root and fill passes. (author)
Self-shielding and burn-out effects in the irradiation of strongly-neutron-absorbing material
International Nuclear Information System (INIS)
Sekine, T.; Baba, H.
1978-01-01
Self-shielding and burn-out effects are discussed in the evaluation of radioisotopes formed by neutron irradiation of a strongly-neutron-absorbing material. A method of the evaluation of such effects is developed both for thermal and epithermal neutrons. Gadolinium oxide uniformly mixed with graphite powder was irradiated by reactor-neutrons together with pieces of a Co-Al alloy wire (the content of Co being 0.475%) as the neutron flux monitor. The configuration of the samples and flux monitors in each of two irradiations is illustrated. The yields of activities produced in the irradiated samples were determined by the γ-spectrometry with a Ge(Li) detector of a relative detection efficiency of 8%. Activities at the end of irradiation were estimated by corrections due to pile-up, self-absorption, detection efficiency, branching ratio, and decay of the activity. Results of the calculation are discussed in comparison with the observed yields of 153 Gd, 160 Tb, and 161 Tb for the case of neutron irradiation of disc-shaped targets of gadolinium oxide. (T.G.)
International Nuclear Information System (INIS)
Koike, Hiroki; Kirimura, Kazuki; Yamaji, Kazuya; Kosaka, Shinya; Yamamoto, Akio
2018-01-01
A unified resonance self-shielding method, which can treat general sub-divided fuel regions, is developed for lattice physics calculations in reactor physics field. In a past study, a hybrid resonance treatment has been developed by theoretically integrating equivalence theory and ultra-fine-group slowing-down calculation. It can be applied to a wide range of neutron spectrum conditions including low moderator density ranges in severe accident states, as long as each fuel region is not sub-divided. In order to extend the method for radially and azimuthally sub-divided multi-region geometry, a new resonance treatment is established by incorporating the essence of sub-group method. The present method is composed of two-step flux calculation, i.e. 'coarse geometry + fine energy' (first step) and 'fine geometry + coarse energy' (second step) calculations. The first step corresponds to a hybrid model of the equivalence theory and the ultra-fine-group calculation, and the second step corresponds to the sub-group method. From the verification results, effective cross-sections by the new method show good agreement with the continuous energy Monte-Carlo results for various multi-region geometries including non-uniform fuel compositions and temperature distributions. The present method can accurately generate effective cross-sections with short computation time in general lattice physics calculations. (author)
International Nuclear Information System (INIS)
Zhang, Tianli; Li, Zhuoxin; Kou, Sindo; Jing, Hongyang; Li, Guodong; Li, Hong; Jin Kim, Hee
2015-01-01
The effect of inclusions on the microstructure and toughness of the deposited metals of self-shielded flux cored wires was investigated by optical microscopy, electron microscopy and mechanical testing. The deposited metals of three different wires showed different levels of low temperature impact toughness at −40 °C mainly because of differences in the properties of inclusions. The inclusions formed in the deposited metals as a result of deoxidation caused by the addition of extra Al–Mg alloy and ferromanganese to the flux. The inclusions, spherical in shape, were mixtures of Al 2 O 3 and MgO. Inclusions predominantly Al 2 O 3 and 0.3–0.8 μm in diameter were effective for nucleation of acicular ferrite. However, inclusions predominantly MgO were promoted by increasing Mg in the flux and were more effective than Al 2 O 3 inclusions of the same size. These findings suggest that the control of inclusions can be an effective way to improve the impact toughness of the deposited metal
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)
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)
DiJulio, D. D.; Cooper-Jensen, C. P.; Llamas-Jansa, I.; Kazi, S.; Bentley, P. M.
2018-06-01
A combined measurement and Monte-Carlo simulation study was carried out in order to characterize the particle self-shielding effect of B4C grains in neutron shielding concrete. Several batches of a specialized neutron shielding concrete, with varying B4C grain sizes, were exposed to a 2 Å neutron beam at the R2D2 test beamline at the Institute for Energy Technology located in Kjeller, Norway. The direct and scattered neutrons were detected with a neutron detector placed behind the concrete blocks and the results were compared to Geant4 simulations. The particle self-shielding effect was included in the Geant4 simulations by calculating effective neutron cross-sections during the Monte-Carlo simulation process. It is shown that this method well reproduces the measured results. Our results show that shielding calculations for low-energy neutrons using such materials would lead to an underestimate of the shielding required for a certain design scenario if the particle self-shielding effect is not included in the calculations.
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%
Nuth, Joseph A., III; Johnson, Natasha M.
2012-01-01
There are at least 3 separate photochemical self-shielding models with different degrees of commonality. All of these models rely on the selective absorption of (12))C(16)O dissociative photons as the radiation source penetrates through the gas allowing the production of reactive O-17 and O-18 atoms within a specific volume. Each model also assumes that the undissociated C(16)O is stable and does not participate in the chemistry of nebular dust grains. In what follows we will argue that this last, very important assumption is simply not true despite the very high energy of the CO molecular bond.
International Nuclear Information System (INIS)
Froehner, F.H.; Larson, Duane C.; Tagesen, Siegfried; Petrizzi, Luigi; Hasegawa, Akira; Nakagawa, Tsuneo; Hogenbirk, Alfred; Weigmann, H.
1995-01-01
A Working Party on International Evaluation Co-operation was established under the sponsorship of the OECD/NEA Nuclear Science Committee (NSC) to promote the exchange of information on nuclear data evaluations, validation, and related topics. Its aim is also to provide a framework for co-operative activities between members of the major nuclear data evaluation projects. This includes the possible exchange of scientists in order to encourage co-operation. Requirements for experimental data resulting from this activity are compiled. The Working Party determines common criteria for evaluated nuclear data files with a view to assessing and improving the quality and completeness of evaluated data. The Parties to the project are: ENDF (United States), JEFF/EFF (NEA Data Bank Member countries), and JENDL (Japan). Co-operation with evaluation projects of non-OECD countries are organised through the Nuclear Data Section of the International Atomic Energy Agency (IAEA). NEA/NSC Subgroup 15 has had the task to assess self-shielding effects in the unresolved resonance range of structural materials, in particular their importance at various energies, and possible ways to deal with them in shielding and activation work. The principal results achieved are summarised briefly, in particular: - New data base consisting of high-resolution transmission data measured at Oak Ridge and Geel; - Improved theoretical understanding of cross-section fluctuations, including their prediction, that has been derived from the Hauser-Feshbach theory; - Benchmark results on the importance of self-shielding in iron at various energies; - Consequences for information storage in evaluated nuclear data files; - Practical utilisation of self-shielding information from evaluated files. Benchmark results as well as the Hauser-Feshbach theory show that self-shielding effects are important up to a 4-or 5-MeV neutron energy. Fluctuation factors extracted from high-resolution total cross-section data can be
International Nuclear Information System (INIS)
Ganesan, S.
1978-01-01
A set of energy dependent fission widths of 1 + spin state corresponding to the recommended fission cross sections of Sowerby et al is evaluated by adjustment in the energy region 600 ev to 25 Kev. Corresponding to these mean fission widths of 1 + spin state, the intermediate resonance parameters based on Weigmann's formulation of Struitinsky's double humped fission barrier model are then obtained. Pseudorandom resonances are generated with and without the intermediate structure in the mean fission but leading to the same value of infinite dilution fission cross section. The effect of the intermediate structure on the self shielding factors was then investigated. (author)
Energy Technology Data Exchange (ETDEWEB)
Videira, Heber S.; Burkhardt, Guilherme M.; Santos, Ronielly S., E-mail: heber@cyclopet.com.br [Cyclopet Radiofarmacos Ltda., Curitiba, PR (Brazil); Passaro, Bruno M.; Gonzalez, Julia A.; Santos, Josefina; Guimaraes, Maria I.C.C. [Universidade de Sao Paulo (HCFMRP/USP), Sao Paulo, SP (Brazil). Faculdade de Medicina. Hospital das Clinicas; Lenzi, Marcelo K. [Universidade Federal do Parana (UFPR), Curitina (Brazil). Programa de Pos-Graduacao em Engenharia Quimica
2013-04-15
The technological advances in positron emission tomography (PET) in conventional clinic imaging have led to a steady increase in the number of cyclotrons worldwide. Most of these cyclotrons are being used to produce {sup 18}F-FDG, either for themselves as for the distribution to other centers that have PET. For there to be safety in radiological facilities, the cyclotron intended for medical purposes can be classified in category I and category II, ie, self-shielded or non-shielded (bunker). Therefore, the aim of this work is to verify the effectiveness of borated water shield built for a cyclotron accelerator-type Self-shielded PETtrace 860. Mixtures of water borated occurred in accordance with the manufacturer’s specifications, as well as the results of the radiometric survey in the vicinity of the self-shielding of the cyclotron in the conditions established by the manufacturer showed that radiation levels were below the limits. (author)
International Nuclear Information System (INIS)
Leal, L.C.; de Saussure, G.; Perez, R.B.
1990-01-01
The URR computer code has been developed to calculate cross-section probability tables, Bondarenko self-shielding factors, and self-indication ratios for fertile and fissile isotopes in the unresolved resonance region. Monte Carlo methods are utilized to select appropriate resonance parameters and to compute the cross sections at the desired reference energy. The neutron cross sections are calculated by the single-level Breit-Wigner formalism with s-, p-, and d-wave contributions. The cross-section probability tables are constructed by sampling by Doppler broadened cross-sections. The various self-shielding factors are computer numerically as Lebesgue integrals over the cross-section probability tables
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
International Nuclear Information System (INIS)
Pelloni, S.; Cheng, E.T.
1985-02-01
The Swiss LOTUS fusion-fission hybrid test facility was used to investigate the influence of the self-shielding of resonance cross sections on the tritium breeding and on the thorium ratios. Nucleonic analyses were performed using the discrete-ordinates transport codes ANISN and ONEDANT, the surface-flux code SURCU, and the version 3 of the MCNP code for the Li 2 CO 3 and the Li 2 O blanket designs with lead, thorium and beryllium multipliers. Except for the MCNP calculation which bases on the ENDF/B-V files, all nuclear data are generated from the ENDF/B-IV basic library. For the deterministic methods three NJOY group libraries were considered. The first, a 39 neutron group self-shielded library, was generated at EIR. The second bases on the same group structure as the first does and consists of infinitely diluted cross sections. Finally the third library was processed at LANL and consists of coupled 30+12 neutron and gamma groups; these cross sections are not self-shielded. The Monte Carlo analysis bases on a continuous and on a discrete 262 group library from the ENDF/B-V evaluation. It is shown that the results agree well within 3% between the unshielded libraries and between the different transport codes and theories. The self-shielding of resonance cross sections results in a decrease of the thorium capture rate and in an increase of the tritium breeding of about 6%. The remaining computed ratios are not affected by the self-shielding of cross sections. (Auth.)
International Nuclear Information System (INIS)
Mason, Grant W.; Spencer, Ross L.
2002-01-01
The 'self-shielding' m=1 diocotron mode in Malmberg-Penning traps has been known for over a decade to be unstable for finite length non-neutral plasmas with hollow density profiles. Early theoretical efforts were unsuccessful in accounting for the exponential growth and/or the magnitude of the growth rate. Recent theoretical work has sought to resolve the discrepancy either as a consequence of the shape of the plasma ends or as a kinetic effect resulting from a modified distribution function as a consequence of the protocol used to form the hollow profiles in experiments. Both of these finite length mechanisms have been investigated in selected test cases using a three-dimensional particle-in-cell code that allows realistic treatment of shape and kinetic effects. A persistent discrepancy of a factor of 2-3 remains between simulation and experimental values of the growth rate. Simulations reported here are more in agreement with theoretical predictions and fail to explain the discrepancy
International Nuclear Information System (INIS)
Komiya, Isao; Umezu, Yoshiyuki; Fujibuchi, Toshiou; Nakamura, Kazumasa; Baba, Shingo; Honda, Hiroshi
2016-01-01
The non-self-shield compact medical cyclotron and the cyclotron vault room were in operation for 27 years. They have now been decommissioned. We efficiently implemented a technique to identify an activation product in the cyclotron vault room. Firstly, the distribution of radioactive concentrations in the concrete of the cyclotron vault room was estimated by calculation from the record of the cyclotron operation. Secondly, the comparison of calculated results with an actual measurement was performed using a NaI scintillation survey meter and a high-purity germanium detector. The calculated values were overestimated as compared to the values measured using the Nal scintillation survey meter and the high-purity germanium detector. However, it could limit the decontamination area. By simulating the activation range, we were able to minimize the concrete core sampling. Finally, the appropriate range of radioactivated area in the cyclotron vault room was decontaminated based on the results of the calculation. After decontamination, the radioactive concentration was below the detection limit value in all areas inside the cyclotron vault room. By these procedures, the decommissioning process of the cyclotron vault room was more efficiently performed. (author)
1 1/2 years of experience with a 10 MeV self-shielded on-line e-beam sterilization system
International Nuclear Information System (INIS)
Lambert, Byron; Tang, Fuh-Wei; Riggs, Brian; Allen, Thomas; Williams, C.B.
2000-01-01
The Vascular Intervening Group of the Guidant Corporation (Guidant IV) has been operating a self-shielded, 10 MeV 4 kW, electron beam sterilization system since July of 1988. The system was designed, built and installed in a 70 square meter area in an existing Guidant manufacturing facility by Titan Scan Corporation and performance of the system was validated in conformance with 1S0-11137 standards. The goal of this on-site e-beam system was 'just in time' JIT, sterilization, i.e. the ability to manufacture, sterilize and ship, high intrinsic value medical devices in less than 24 hours. The benefits of moving from a long gas sterilization cycle of greater than one week to a JIT process were envisioned to be a) speed to market with innovated new products b) rapid response to customer requirements c) reduced inventory carrying costs and finally manufacturing and quality system efficiency. The ability of Guidant to realize these benefits depended upon the ability of the Guidant VI business units to adapt to the new sterilization modality and functionality and on the overall system reliability. This paper reviews the operating experience to date and the overall system reliability. (author)
International Nuclear Information System (INIS)
Gonzalez, Dania Soguero; Ardanza, Armando Chavez
2013-01-01
This paper describes the process of installation of a self-shielded irradiator category I, model ISOGAMMA LL.Co of 60 Co, with a nominal 25 kCi activity, rate of absorbed dose 8 kG/h and 5 L workload. The stages are describe step by step: import, the customs procedure which included the interview with the master of the vessel transporter, the monitoring of the entire process by the head of radiological protection of the importing Center, control of the levels of surface contamination of the shipping container of the sources before the removal of the ship, the supervision of the national regulatory authority and the transportation to the final destination. Details of assembling of the installation and the opening of the container for transportation of supplies is outlined. The action plan previously developed for the case of occurrence of radiological successful events is presented, detailing the phase of the load of radioactive sources by the specialists of the company selling the facility (IZOTOP). Finally describes the setting and implementation of the installation and the procedure of licensing for exploitation
Energy Technology Data Exchange (ETDEWEB)
Coste, M.
1994-01-01
This note gives in detailed way the self-shielding formalism which is used in the multigroup transport code APOLLO2. The self-shielded cross-sections are performed with the same scheme as in APOLLO1. We use two equivalencies, first an heterogeneous/homogeneous equivalence which gives the reaction rates and then a multigroup equivalence in order to obtain the cross-sections which preserve these reaction rates. However, numerous improvements were implemented, specially in the homogenization step. Homogenization can be performed group per group with different modelizations of the heavy slowing-down operator (statistical, intermediary and ``wide resonance`` models), which allows us to fit correctly the resonance shapes. Moreover, we can take exactly into account the spatial interferences between resonant isotopes with the background matrix model. Consequently, we are now able to perform, for instance, the radial distribution of the resonant absorption inside a fuel pin. (author). 7 refs., 3 figs.
Energy Technology Data Exchange (ETDEWEB)
Rios, D.A.S.; Rios, P.B., E-mail: denise@inovafi.com.br [Inovafi Física aplicada à Inovação Ltda, Sorocaba, SP (Brazil); Sordi, G.M.A.A.; Carneiro, J.C.G.G. [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)
2017-07-01
The study discusses situations in the industrial environment that may lead to potential exposure of Occupationally Exposed Individuals and Public Individuals in self-shielding electron accelerators. Although these exposure situations are unlikely, simulation exercises can lead to improvements in the operating procedure as well as suggest changes in production line design in order to increase radiation protection at work. These studies can also be used in training and demonstrate a solid application of the ALARA principle in the daily activities of radiative installations.
International Nuclear Information System (INIS)
Leal, L.C.; de Saussure, G.; Perez, R.B.
1989-01-01
The URR computer code has been developed to calculate cross-section probability tables, Bondarenko self-shielding factors, and self- indication ratios for fertile and fissile isotopes in the unresolved resonance region. Monte Carlo methods are utilized to select appropriate resonance parameters and to compute the cross sections at the desired reference energy. The neutron cross sections are calculated by the single-level Breit-Wigner formalism with s-, p-, and d-wave contributions. The cross-section probability tables are constructed by sampling the Doppler broadened cross-section. The various shelf-shielded factors are computed numerically as Lebesgue integrals over the cross-section probability tables. 6 refs
International Nuclear Information System (INIS)
Smith, Rachel L.; Young, Edward D.; Pontoppidan, Klaus M.; Morris, Mark R.; Van Dishoeck, Ewine F.
2009-01-01
Using very high resolution (λ/Δλ ∼ 95 000) 4.7 μm fundamental and 2.3 μm overtone rovibrational CO absorption spectra obtained with the Cryogenic Infrared Echelle Spectrograph infrared spectrometer on the Very Large Telescope (VLT), we report detections of four CO isotopologues-C 16 O, 13 CO, C 18 O, and the rare species, C 17 O-in the circumstellar environment of two young protostars: VV CrA, a binary T Tauri star in the Corona Australis molecular cloud, and Reipurth 50, an intermediate-mass FU Ori star in the Orion Molecular Cloud. We argue that the observed CO absorption lines probe a protoplanetary disk in VV CrA, and a protostellar envelope in Reipurth 50. All CO line profiles are spectrally resolved, with intrinsic line widths of ∼3-4 km s -1 (FWHM), permitting direct calculation of CO oxygen isotopologue ratios with 5%-10% accuracy. The rovibrational level populations for all species can be reproduced by assuming that CO absorption arises in two temperature regimes. In the higher temperature regime, in which the column densities are best determined, the derived oxygen isotope ratios in VV CrA are: [C 16 O]/[C 18 O] =690 ± 30; [C 16 O]/[C 17 O] =2800 ± 300, and [C 18 O]/[C 17 O]=4.1 ± 0.4. For Reipurth 50, we find [C 16 O]/[C 18 O] =490 ± 30; [C 16 O]/[C 17 O] =2200 ± 150, [C 18 O]/[C 17 O] = 4.4 ± 0.2. For both objects, 12 C/ 13 C are on the order of 100, nearly twice the expected interstellar medium (ISM) ratio. The derived oxygen abundance ratios for the VV CrA disk show a significant mass-independent deficit of C 17 O and C 18 O relative to C 16 O compared to ISM baseline abundances. The Reipurth 50 envelope shows no clear differences in oxygen CO isotopologue ratios compared with the local ISM. A mass-independent fractionation can be interpreted as being due to selective photodissociation of CO in the disk surface due to self-shielding. The deficits in C 17 O and C 18 O in the VV CrA protoplanetary disk are consistent with an analogous
Energy Technology Data Exchange (ETDEWEB)
Cooper, R.; Silva, J. H. F.; Trevisan, R. E.
2004-07-01
The present work refers to the characterization of API 5L-X80 pipeline joints welded with self-shielded flux cored wire. This process was evaluated under preheating conditions, with an uniform and steady heat input. All joints were welded in flat position (1G), with the pipe turning and the torch still. Tube dimensions were 762 mm in external diameter and 16 mm in thickness. Welds were applied on single V-groove, with six weld beads, along with three levels of preheating temperatures (room temperature, 100 degree centigree, 160 degree centigree). These temperatures were maintained as inter pass temperature. The filler metal E71T8-K6 with mechanical properties different from parent metal was used in under matched conditions. The weld characterization is presented according to the mechanical test results of tensile strength, hardness and impact test. The mechanical tests were conducted according to API 1104, AWS and ASTM standards. API 1104 and API 51 were used as screening criteria. According to the results obtained, it was possible to remark that it is appropriate to weld API 5L-X80 steel ducts with Self-shielded Flux Cored wires, in conformance to the API standards and no preheat temperature is necessary. (Author) 22 refs.
Energy Technology Data Exchange (ETDEWEB)
Gonzalez, Dania Soguero; Ardanza, Armando Chavez, E-mail: sdania@ceaden.edu.cu [Centro de Aplicaciones Tecnologicas y Desarrollo Nuclear (CEADEN), La Habana (Cuba)
2013-07-01
This paper describes the process of installation of a self-shielded irradiator category I, model ISOGAMMA LL.Co of {sup 60}Co, with a nominal 25 kCi activity, rate of absorbed dose 8 kG/h and 5 L workload. The stages are describe step by step: import, the customs procedure which included the interview with the master of the vessel transporter, the monitoring of the entire process by the head of radiological protection of the importing Center, control of the levels of surface contamination of the shipping container of the sources before the removal of the ship, the supervision of the national regulatory authority and the transportation to the final destination. Details of assembling of the installation and the opening of the container for transportation of supplies is outlined. The action plan previously developed for the case of occurrence of radiological successful events is presented, detailing the phase of the load of radioactive sources by the specialists of the company selling the facility (IZOTOP). Finally describes the setting and implementation of the installation and the procedure of licensing for exploitation.
CO self-shielding as the origin of oxygen isotope anomalies in the early solar nebula.
Lyons, J R; Young, E D
2005-05-19
The abundances of oxygen isotopes in the most refractory mineral phases (calcium-aluminium-rich inclusions, CAIs) in meteorites have hitherto defied explanation. Most processes fractionate isotopes by nuclear mass; that is, 18O is twice as fractionated as 17O, relative to 16O. In CAIs 17O and 18O are nearly equally fractionated, implying a fundamentally different mechanism. The CAI data were originally interpreted as evidence for supernova input of pure 16O into the solar nebula, but the lack of a similar isotope trend in other elements argues against this explanation. A symmetry-dependent fractionation mechanism may have occurred in the inner solar nebula, but experimental evidence is lacking. Isotope-selective photodissociation of CO in the innermost solar nebula might explain the CAI data, but the high temperatures in this region would have rapidly erased the signature. Here we report time-dependent calculations of CO photodissociation in the cooler surface region of a turbulent nebula. If the surface were irradiated by a far-ultraviolet flux approximately 10(3) times that of the local interstellar medium (for example, owing to an O or B star within approximately 1 pc of the protosun), then substantial fractionation of the oxygen isotopes was possible on a timescale of approximately 10(5) years. We predict that similarly irradiated protoplanetary disks will have H2O enriched in 17O and 18O by several tens of per cent relative to CO.
Multiband method for resonance self-shielding calculation of fuel assembly in arbitrary geometries
International Nuclear Information System (INIS)
Huang Weibin; Wu Hongchun; Cao Liangzhi; Yang Weiyan
2009-01-01
A formula to calculate the multiband parameters is derived based on the multiband method. Adopting the method combining two-band and three-band, and based on the WIMSD4-69 library, a code named RESCAL is developed. The validation shows that the results of RESCAL code are well in accordance with MCNP's, and the numerical errors meet the practical requirement. Due to the limitation of WIMSD4 69-group library and the method adopted to calculate multiband parameters, the precision of RESCAL code is highly affected by the ratio of water to uranium. (authors)
Unresolved resonance range cross section, probability tables and self shielding factor
International Nuclear Information System (INIS)
Sublet, J.Ch.; Blomquist, R.N.; Goluoglu, S.; Mac Farlane, R.E.
2009-07-01
The performance and methodology of 4 processing codes have been compared in the unresolved resonance range of a selected set of isotopes. Those isotopes have been chosen to encompass most cases encountered in the unresolved energy range contained in major libraries like Endf/B-7 or Jeff-3.1.1. The code results comparison is accompanied by data format and formalism examinations and processing code fine-interpretation study. After some improvements, the results showed generally good agreement, although not perfect with infinite dilute cross-sections. However, much larger differences occur when shelf-shielded effective cross-sections are compared. The infinitely dilute cross-section are often plot checked but it is the probability table derived and shelf-shielded cross sections that are used and interpreted in criticality and transport calculations. This suggests that the current evaluation data format and formalism, in the unresolved resonance range should be tightened up, ambiguities removed. In addition production of the shelf shielded cross-sections should be converged to a much greater accuracy. (author)
Defects detection on the welded reinforcing steel with self-shielded wires by vibration tests
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Crâştiu Ion
2017-01-01
Full Text Available The aim of this paper is the development and validation of a vibroacustic technique to welding defects detection, especially for welded reinforcing structures. In welded structures subjected to dynamic cyclic loads may appear and propagate fatigue cracks due to local structural damage. These cracks may initiate due to the technological parameters used in welding process, or due to environmental operating conditions. By the means of Finite Element Method (FEM, the natural frequencies and shape modes of more welded steel specimens are determined. The analysis is carried out in undamaged condition as well as damaged one, after artificially induced damages. The experimental measurement of the vibroacustic response is carried out by using a condenser microphone, which is suitable for high-fidelity acoustic measurements in the frequency range of 20 – 20.000 Hz. The vibration responses of the welded specimens, in free-free conditions, are carried out using algorithms based on Fast Fourier Transform (FFT, and Prony's series. The results are compared to modal parameters estimated using FE Analysis.
The new IBA self-shielded dynamitron accelerator for industrial applications
International Nuclear Information System (INIS)
Galloway, R.A.; DeNeuter, S.; Lisanti, T.F.; Cleland, M.R.
2004-01-01
Radiation Dynamics Inc. (RDI), currently a member of the IBA Group (Ion Beam Applications based Louvain-la-Neuve, Belgium), has been supplying accelerators since its founding in 1958. These systems supplied for both industrial processing and research application for electrons and ions have proven to be reliable and robust. Today's demands in the industrial sector have driven the design and development of a new version of our Dynamitron [reg] . This new system, envisioned to operate at electron energies up to 1.5 MeV, in many cases can be supplied with integral shielding providing a small footprint requirement for placement in a facility. In the majority of these lower energy applications this allows the appropriate material handling system to be installed inside the steel radiation enclosure. Designed to deliver beam power outputs as high as 100 kW, this new system is capable of servicing the high throughput demands of today's manufacturing lines. Still retaining the positive aspects of the industrially proven Dynamitron system, this compact system can be tailored to meet a variety of in-line or off-line processing applications
International Nuclear Information System (INIS)
Hong, Ser Gi; Kim, Kang-Seog
2011-01-01
This paper describes the iteration methods using resonance integral tables to estimate the effective resonance cross sections in heterogeneous transport lattice calculations. Basically, these methods have been devised to reduce an effort to convert resonance integral table into subgroup data to be used in the physical subgroup method. Since these methods do not use subgroup data but only use resonance integral tables directly, these methods do not include an error in converting resonance integral into subgroup data. The effective resonance cross sections are estimated iteratively for each resonance nuclide through the heterogeneous fixed source calculations for the whole problem domain to obtain the background cross sections. These methods have been implemented in the transport lattice code KARMA which uses the method of characteristics (MOC) to solve the transport equation. The computational results show that these iteration methods are quite promising in the practical transport lattice calculations.
International Nuclear Information System (INIS)
Choi, Sooyoung; Khassenov, Azamat; Lee, Deokjung
2016-01-01
This paper presents a new method of resonance interference effect treatment using resonance interference factor for high fidelity analysis of light water reactors (LWRs). Although there have been significant improvements in the lattice physics calculations over the several decades, there exist still relatively large errors in the resonance interference treatment, in the order of ∼300 pcm in the reactivity prediction of LWRs. In the newly developed method, the impact of resonance interference to the multi-group cross-sections has been quantified and tabulated in a library which can be used in lattice physics calculation as adjustment factors of multi-group cross-sections. The verification of the new method has been performed with Mosteller benchmark, UO_2 and MOX pin-cell depletion problems, and a 17×17 fuel assembly loaded with gadolinia burnable poison, and significant improvements were demonstrated in the accuracy of reactivity and pin power predictions, with reactivity errors down to the order of ∼100 pcm. (author)
International Nuclear Information System (INIS)
Tochon-Danguy, H.; Sachinidis, J.I.; U, P.; Egan, G.; Mukherjee, B.
1999-01-01
A series of health physics measurements was carried out at the IBA CYCLONE 10/5 Medical Cyclotron of the Austin and Repatriation Medical Centre, Melbourne. The neutron attenuation factor of the cyclotron shielding was estimated using the Superheated Bubble dosimeters. The neutron and gamma dose rates at various public access and radiation worker's area in the vicinity of the cyclotron facility were evaluated during the 11 C, 18 F, 13 N and 15 O production conditions. (authors)
Improved analysis on multiple recycling of fuel in prototype fast ...
Indian Academy of Sciences (India)
2011-08-03
Aug 3, 2011 ... ENDF/B-VII.0, and with the most recent specification of the fuel composition ... Fast breeder reactors; closed fuel cycle; fuel production and depletion; reprocessing .... set including self-shielding factors (SSF, i.e. self-shielded to ...
The background influence of cadmium detection in saline water using PGNAA technique
International Nuclear Information System (INIS)
Daqian Hei; Zhou Jiang; Hongtao Wang; Jiatong Li
2016-01-01
In order to solve the background influence of cadmium detection in saline water using prompt gamma neutron activation analysis (PGNAA) technique, a series experiments have been designed and carried out. Furthermore, a method based on internal standard was used to correct the neutron self-shielding effect, and the background influence has been decreased sequentially. The results showed a good linear relationship between the characteristic peak counts and the concentrations of cadmium after the neutron self-shielding correction. And in the detection of saline water by PGNAA technique, the proposed methodology can be used to reduce the influence of background with the self-shielding effect correction. (author)
Granular effect on the effective cross sections in the HTGR type reactors
International Nuclear Information System (INIS)
Almeida Ferreira, A.C. de.
1975-01-01
Effective cross section of bars for HTGR is studied from the point of view of heterogeneity. Microscopical heterogeneity due to grains is represented by a self-shielding factor, which is well determined [pt
Assessment of the MPACT Resonance Data Generation Procedure
Energy Technology Data Exchange (ETDEWEB)
Kim, Kang Seog [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Williams, Mark L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
2016-12-26
Currently, heterogeneous models are being used to generate resonance self-shielded cross-section tables as a function of background cross sections for important nuclides such as ^{235}U and ^{238}U by performing the CENTRM (Continuous Energy Transport Model) slowing down calculation with the MOC (Method of Characteristics) spatial discretization and ESSM (Embedded Self-Shielding Method) calculations to obtain background cross sections. And then the resonance self-shielded cross section tables are converted into subgroup data which are to be used in estimating problem-dependent self-shielded cross sections in MPACT (Michigan Parallel Characteristics Transport Code). Although this procedure has been developed and thus resonance data have been generated and validated by benchmark calculations, assessment has never been performed to review if the resonance data are properly generated by the procedure and utilized in MPACT. This study focuses on assessing the procedure and a proper use in MPACT.
Technological aspects of transmutation of technetium and iodine
International Nuclear Information System (INIS)
Konings, R.J.M.; Abrahams, K.; Franken, W.M.P.; Gruppelaar, H.; Kloosterman, J.L.; Thijssen, P.J.M.; Conrad, R.
1993-06-01
Pre-test calculations on the effect of self-shielding, experimental studies for sample chacterization and sample selection and the design of an irradiation facility for use in the HFR are discussed. (orig./HP)
Connection factor calculation for isotopic neutron flux measurements with foil detectors
International Nuclear Information System (INIS)
Avila L, J.
1987-01-01
Thermal and resonance neutron self-shielding factors, neutron flux distortion and edge effects as well as a connection factor for neutron flux profile around a foil detector have been calculated. A general expression for resonance self shielding factor is presented in order to take into account the most important resonances for a given isotope. A computer program SPRESYTER.BAS was written and results for In-115 and Au-197 foils are given
UNR. A code for processing unresolved resonance data for MCNP
International Nuclear Information System (INIS)
Hogenbirk, A.
1994-09-01
In neutron transport problems the correct treatment of self-shielding is important for those nuclei present in large concentrations. Monte Carlo calculations using continuous-energy cross section data, such as calculations with the code MCNP, offer the advantage that neutron transport is calculated in a very accurate way. Self-shielding in the resolved resonance region is taken into account exactly in MCNP. However, self-shielding in the unresolved resonance region can not be taken into account by MCNP, although the effect of it may be important in many applications. In this report a description is given of the computer code UNR. With this code problem-dependent cross section libraries can be produced for MCNP. In these libraries self-shielded cross section data in the unresolved resonance range are given, which are produced by NJOY-module UNRESR. It is noted, that the treatment for resonance self-shielding presented in this report is approximate. However, the current version of MCNP does not allow the use of probability tables, which would be a general solution. (orig.)
Heavy nucleus resonance absorption in heterogeneous lattices
International Nuclear Information System (INIS)
Coste, M.; Tellier, H.; Brienne-Raepsaet, C.; Van Der Gucht, C.
1992-01-01
To compute easily the neutron reaction rates in the resonance energy range, the reactor physicists use the self-shielding formalism and the effective cross-section concept. Usually, for these calculations, and equivalence process is used, in such a way that the absorption rate is correctly computed for the whole fuel pin. This procedure does not allow to preserve the spatial absorption rate distribution inside the pin. It is an important handicap if we want to reproduce the plutonium distribution in a spent fuel. To avoid this inconvenience, new improvements of the self-shielding formalism have been recently introduced in the new assembly calculation code of the French Atomic Energy Commission, APOLLO 2. With this improved formalism, it is now possible to represent the spatial and energetic dependence of the heavy nucleus absorption inside the fuel pin and to use a fine energy dependent equivalence process. As it does not exist clean experimental results for the spatial and energetic dependence of the absorption, the authors used reference calculations to qualify the self-shielding formalism. For the strongly self-shielded nuclei of interest in reactor physics, U238, Pu240 and Th232, the agreement between the self-shielding calculation and the reference ones is fairly good for the spatial and energetic dependence of the absorption rate
Validity of Hansen-Roach cross sections in low-enriched uranium systems
International Nuclear Information System (INIS)
Busch, R.D.; O'Dell, R.D.
1991-01-01
Within the nuclear criticality safety community, the Hansen-Roach 16 group cross section set has been the ''standard'' for use in k eff calculations over the past 30 years. Yet even with its widespread acceptance, there are still questions about its validity and adequacy, about the proper procedure for calculating the potential scattering cross section, σ p , for uranium and plutonium, and about the concept of resonance self shielding and its impact on cross sections. This paper attempts to address these questions. It provides a brief background on the Hansen-Roach cross sections. Next is presented a review of resonances in cross sections, self shielding of these resonances, and the use of σ p to characterize resonance self shielding. Three prescriptions for calculating σ p are given. Finally, results of several calculations of k eff on low-enriched uranium systems are provided to confirm the validity of the Hansen-Roach cross sections when applied to such systems
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)
Epithermal interrogation of fissile waste
International Nuclear Information System (INIS)
Coop, K.L.; Hollas, C.L.
1996-01-01
Self-shielding of interrogating thermal neutrons in lumps of fissile material can be a major source of error in transuranic waste assay using the widely employed differential dieaway technique. We are developing a new instrument, the combined thermal/epithermal neutron (CTEN) interrogation instrument to detect the occurrence of self- shielding and mitigate its effects. Neutrons are moderated in the graphite walls of the CTEN instrument to provide an interrogating flux of epithermal and thermal neutrons. The induced prompt fission neutrons are detected in proportional counters. We report the results of measurements made with the CTEN instrument, using minimal and highly self-shielding plutonium and uranium sources in 55 gallon drums containing a variety of mock waste matrices. Fissile isotopes and waste forms for which the method is most applicable, and limitations associated with the hydrogen content of the waste package/matrix are described
Validity of Hansen-Roach cross sections in low-enriched uranium systems
International Nuclear Information System (INIS)
Busch, R.D.; O'Dell, R.D.
1991-01-01
Within the nuclear criticality safety community, the Hansen-Roach 16 group cross section set has been the standard for use in k eff calculations over the past 30 years. Yet even with its widespread acceptance, there are still questions about its validity and adequacy, about the proper procedure for calculating the potential scattering cross section, σ p , for uranium and plutonium, and about the concept of resonance self shielding and its impact on cross sections. This paper attempts to address these questions. It provides a brief background on the Hansen-Roach cross sections. Next is presented a review of resonances in cross sections, self shielding of these resonances, and the use of σ p to characterize resonance self shielding. Three prescriptions for calculating σ p are given. Finally, results of several calculations of k eff on low-enriched uranium systems are provided to confirm the validity of the Hansen-Roach cross sections when applied to such systems. (Author)
Energy Technology Data Exchange (ETDEWEB)
Hathout, A M [National Center for Nuclear Safety and Radiation Control, NC-NSRC, Atomic Energy Authority, Cairo (Egypt)
1996-03-01
The narrow resonance approximation is applicable for all low-energy resonances and the heaviest nuclides. It is of great importance in neutron calculations, hence, fertile isotopes do not undergo fission at resonance energies. The effect of overestimating the self shielded group averaged cross-section data for a given resonance nuclide can be fairly serious. In the present work, a detailed study, and derivation of the problem of self-shielding are carried-out through the information of Hansen-roach library which is used for criticality safety analysis. The intermediate neutron flux spectrum is analyzed, using the narrow resonance approximation. The resonance self-shielded values of various cross-sections are determined. 4 figs., 3 tabs.
Study on 2D arbitrary geometry coupling resonance method
International Nuclear Information System (INIS)
He Lei; Wu Hongchun; Cao Liangzhi
2014-01-01
The paper firstly proposes a coupling resonance method in which subgroup method is employed in the serried peak energy region, and wavelet expansion method is employed in single peak energy region. The original subgroup model and wavelet expansion model are improved and coupled through the calculation of scattering source from subgroup to wavelet expansion, so that the self-shielding cross section in the whole energy region can be calculated accurately. To verify these theories and to prove the improvements, a PWR cell benchmark problem is calculated. It is demonstrated that, compared with other traditional multi-group resonance methods and continuous energy resonance method, this coupling resonance method has the ability to accurately calculate the whole energy region's self-shielding cross section while Keeping enough efficiency and finally has an ability to offer the accurate self-shielding parameters for latter transport, calculation. (authors)
Borated concrete for ZPPR fuel storage
International Nuclear Information System (INIS)
Gasidlo, J.M.
1985-01-01
Fuel handling at the Zero Power Plutonium Reactor (ZPPR) led to two requirements for storage of ZPPR fuel: a low neutron multiplication and shielded storage to minimize personnel doses. Boron-poisoned concrete was chosen as the storge medium with boron frit as the poisoning agent. The calculated effects of water content and boron concentration led to specifying a concrete with a water content that was higher than ordinary concrete. The finite size of the boron frit particles caused concern about reduced effectiveness due to self-shielding. The self-shielding was evaluated using optical path lengths for spheres and tabulated self-shielding for slabs. The results showed that the finite-sized particles were at least 80% as effective as infinitely dilute absorption. Neutron and gamma dose rates measured in the vault verified that personnel could work in the vault on a regular basis without exceeding personnel dose limits. 4 refs., 3 figs., 7 tabs
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
International Nuclear Information System (INIS)
Filbert, W.; Wehrmann, J.; Bollingerfehr, W.; Graf, R.; Fopp, S.
2008-01-01
The reference concept in Germany on direct final storage of spent fuel rods is the burial of POLLUX containers in the final repository salt dome. The POLLUX container is self-shielded. The final storage concept also includes un-shielded borehole storage of high-level waste and packages of compacted waste. GNS has developed a spent fuel container (BSK-3) for unshielded borehole storage with a mass of 5.2 tons that can carry the fuel rods of three PWR reactors of 9 BWR reactors. The advantages of BSK storage include space saving, faster storage processes, less requirements concerning technical barriers, cost savings for self-shielded casks.
Graphs of neutron cross sections in JSD1000 for radiation shielding safety analysis
International Nuclear Information System (INIS)
Yamano, Naoki
1984-03-01
Graphs of neutron cross sections and self-shielding factors in the JSD1000 library are presented for radiation shielding safety analysis. The compilation contains various reaction cross sections for 42 nuclides from 1 H to 241 Am in the energy range from 3.51 x 10 -4 eV to 16.5 MeV. The Bondarenko-type self-shielding factors of each reaction are given by the background cross sections from σ 0 = 0 to σ 0 = 10000. (author)
Testing of cross-section functionals in the unresolved resonance region
International Nuclear Information System (INIS)
Koshcheev, V.N.; Krivtsov, A.S.; Sinitsa, V.V.; Ukraintsev, V.F.
1994-01-01
The results of a comparison of the GRUCON, MMK and NJOY data processing codes in the treatment of evaluated neutron data in the unresolved resonance region are presented. The sets of average resonance parameters of 238 U and 239 Pu isotopes, which have been published by Munos-Cobos et al., and Ribon et al., were used in this exercise. Average cross-sections, self-shielding factors and Doppler broadening self-shielding factors are compared with the original results presented by the above-mentioned authors. Conclusions regarding the reliability of the neutron data processing codes are made. (author). 6 refs, 7 tabs
Design of incoming neutron-beam for detecting oil dirt
International Nuclear Information System (INIS)
Zhao Jingwu; Chen Xiaocheng; Alimujiang Naimaiti; Aierken Abuliemu
2012-01-01
For the technique of neutron back-scattering, the neutron counts are non-linear and have a tendency toward saturation because of the neutron self-shielding. As a result, the measurement accuracy is reduced and the measurement range is limited. Using a simply model and comparing with experimental data, it is shown that, in the measurement of the thickness of oil dirt, by adjusting the ratio of thermal to epithermal neutrons, the neutron self: shielding is weakened. As a result, the non-linearity can be reduced and the measurement accuracy and range can be improved. (authors)
Multiband discrete ordinates method: formalism and results
International Nuclear Information System (INIS)
Luneville, L.
1998-06-01
The multigroup discrete ordinates method is a classical way to solve transport equation (Boltzmann) for neutral particles. Self-shielding effects are not correctly treated due to large variations of cross sections in a group (in the resonance range). To treat the resonance domain, the multiband method is introduced. The main idea is to divide the cross section domain into bands. We obtain the multiband parameters using the moment method; the code CALENDF provides probability tables for these parameters. We present our implementation in an existing discrete ordinates code: SN1D. We study deep penetration benchmarks and show the improvement of the method in the treatment of self-shielding effects. (author)
Depletion optimization of lumped burnable poisons in pressurized water reactors
International Nuclear Information System (INIS)
Kodah, Z.H.
1982-01-01
Techniques were developed to construct a set of basic poison depletion curves which deplete in a monotonical manner. These curves were combined to match a required optimized depletion profile by utilizing either linear or non-linear programming methods. Three computer codes, LEOPARD, XSDRN, and EXTERMINATOR-2 were used in the analyses. A depletion routine was developed and incorporated into the XSDRN code to allow the depletion of fuel, fission products, and burnable poisons. The Three Mile Island Unit-1 reactor core was used in this work as a typical PWR core. Two fundamental burnable poison rod designs were studied. They are a solid cylindrical poison rod and an annular cylindrical poison rod with water filling the central region.These two designs have either a uniform mixture of burnable poisons or lumped spheroids of burnable poisons in the poison region. Boron and gadolinium are the two burnable poisons which were investigated in this project. Thermal self-shielding factor calculations for solid and annular poison rods were conducted. Also expressions for overall thermal self-shielding factors for one or more than one size group of poison spheroids inside solid and annular poison rods were derived and studied. Poison spheroids deplete at a slower rate than the poison mixture because each spheroid exhibits some self-shielding effects of its own. The larger the spheroid, the higher the self-shielding effects due to the increase in poison concentration
International Nuclear Information System (INIS)
Koshcheev, V.N.; Manturov, G.N.; Sinitsa, V.V.
1991-01-01
The effects of the neutron strength function uncertainties on the calculated values of the self-shielding factors and energy dependence of the total and capture 238 U cross-sections in the unresolved resonance region are investigated. (author). 26 refs, 5 figs
Impact of MCNP unresolved resonance probability-table treatment on uranium and plutonium benchmarks
International Nuclear Information System (INIS)
Mosteller, R.D.; Little, R.C.
1998-01-01
Versions of MCNP up through and including 4B have not accurately modeled neutron self-shielding effects in the unresolved resonance energy region. Recently, a probability-table treatment has been incorporated into a developmental version of MCNP. This paper presents MCNP results for a variety of uranium and plutonium critical benchmarks, calculated with and without the probability-table treatment
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
Use of the APOLLO2 transport code for PWR assembly studies
International Nuclear Information System (INIS)
Belhaffaf, D.; Coste, M.; Lenain, R.; Mathonniere, G.; Sanchez, R.; Stankovski, Z.; Zmijarevic, I.
1992-01-01
This paper presents some validation and application aspects of the APOLLO2, user oriented, modular code for multigroup transport assembly calculation which is developed at the French Commissariat a l'Energies Atomique. The main points approached in this paper are: the two dimensional collision probability convergence, critical leakage calculation schemes, self-shielding spatial discretization, and the equivalence procedure
Nuclear characteristics of Pu fueled LWR and cross section sensitivities
Energy Technology Data Exchange (ETDEWEB)
Takeda, Toshikazu [Osaka Univ., Suita (Japan). Faculty of Engineering
1998-03-01
The present status of Pu utilization to thermal reactors in Japan, nuclear characteristics and topics and cross section sensitivities for analysis of Pu fueled thermal reactors are described. As topics we will discuss the spatial self-shielding effect on the Doppler reactivity effect and the cross section sensitivities with the JENDL-3.1 and 3.2 libraries. (author)
SCALE Continuous-Energy Monte Carlo Depletion with Parallel KENO in TRITON
International Nuclear Information System (INIS)
Goluoglu, Sedat; Bekar, Kursat B.; Wiarda, Dorothea
2012-01-01
The TRITON sequence of the SCALE code system is a powerful and robust tool for performing multigroup (MG) reactor physics analysis using either the 2-D deterministic solver NEWT or the 3-D Monte Carlo transport code KENO. However, as with all MG codes, the accuracy of the results depends on the accuracy of the MG cross sections that are generated and/or used. While SCALE resonance self-shielding modules provide rigorous resonance self-shielding, they are based on 1-D models and therefore 2-D or 3-D effects such as heterogeneity of the lattice structures may render final MG cross sections inaccurate. Another potential drawback to MG Monte Carlo depletion is the need to perform resonance self-shielding calculations at each depletion step for each fuel segment that is being depleted. The CPU time and memory required for self-shielding calculations can often eclipse the resources needed for the Monte Carlo transport. This summary presents the results of the new continuous-energy (CE) calculation mode in TRITON. With the new capability, accurate reactor physics analyses can be performed for all types of systems using the SCALE Monte Carlo code KENO as the CE transport solver. In addition, transport calculations can be performed in parallel mode on multiple processors.
Importance of resonance parameters of fertile nuclei and of 239Pu isotope for fast power reactors
International Nuclear Information System (INIS)
Barre, J.Y.; Khairallah, A.
1975-01-01
The importance of resonance parameters of fertile nuclei and of 239 Pu isotope for fast power reactors will be restricted, in this presentation, to mixed oxide-uranium-plutonium fuelled sodium-cooled and uranium-oxide-sodium reflected fast reactors. The power range lies between 200 and 2000 MWe. Among the topics of this specialist meeting, the isotopes to be considered are, primarly 239 Pu then 238 U and 240 Pu. Resonance parameters are mainly used in fast power reactor calculations through the well-known concept of self shielding factors. After a short description of the determination and the use of these self-shielding factors, their sensitivities to resonance parameters are characterized from some specific examples: those sensitivities are small. Then, the main design parameters sensitive to the amplitude of self-shielding factors are considered: critical enrichment, global breeding gain. The relative importance of isotope, reaction rate and energy range are mentionned. In a third part, the Doppler effect, sensitive to the temperature variation of self-shielding factors, is considered in the same way. Finally, it is concluded that the present knowledge of resonance parameters for 238 U, 239 Pu and 240 Pu is sufficient for fast power reactors from a designer point of view [fr
Validation of the EFF-3.0 evaluation for 56Fe
International Nuclear Information System (INIS)
Hogenbirk, A.; Koning, A.J.; Gruppelaar, H.
1995-11-01
In this report the validation is described of the EFF-3.0 evaluation for 56 Fe. In this evaluation the unresolved resonance region is described in detail, using fluctuation factors derived from high-resolution measurements in Geel. In neutron transport problems the correct treatment of self-shielding is important for those nuclei present in large concentrations. Monte Carlo calculations using continuous-energy cross section data, such as calculations with the code MCNP, offer the advantage that neutron transport is calculated in a very accurate way. Self-shielding in the resolved resonance region is taken into account exactly in MCNP. Using data from the EFF-3.0 evaluation for 56 Fe the self-shielding in the unresolved resonance region can also be taken into account by MCNP. The innovation in methodology is twofold: Firstly the cross sections are represented by high-resolution data and secondly self-shielding in elastic and inelastic channels is taken into account. Benchmark calculations were performed using the high-quality data from the Fe slab experiment performed at the Technical University of Dresden. In order to show the effect of several approximations, a calculational benchmark was performed as well. It is concluded, that a very good description of experimental data is possible with the EFF-3.0 evaluation for 56 Fe. Therefore, this evaluation is recommended for accurate and reliable calculations. (orig.)
Heavy nucleus resonant absorption in heterogeneous lattices
International Nuclear Information System (INIS)
Tellier, H.; Coste, M.; Raepsaet, C.; Van der Gucht, C.
1992-11-01
The new self shielding formalism which is implemented in the transport code APPOLO 2 allows to compute the space and the energy dependence of the resonant absorption rate inside a fuel rod. As it does not exist any experimental result to check such a computation, we used very sophisticated reference calculations to check the self shielding formalism. Two kinds of reference calculation were used: a multi-group slowing down calculation with a very detailed energy mesh and a MONTE CARLO computation with point wise cross sections. The purpose of this study is a comparison of the self-shielding result with the ones of the two reference calculations. The geometry of the fuel element is the one of a light water-reactor and the resonant nuclei are the most important self-shielded nuclei: 238U, 232Th or 240Pu. The reactor agreement between the three kinds of calculations is very satisfactory for the nominal water density and for a reduced water density which simulates incidental operating conditions
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
New capabilities of the lattice code WIMS-AECL
International Nuclear Information System (INIS)
Altiparmakov, Dimitar
2008-01-01
The lattice code WIMS-AECL has been restructured and rewritten in Fortran 95 in order to increase the accuracy of its responses and extend its capabilities. Significant changes of computing algorithms have been made in the following two areas: geometric calculations and resonance self-shielding. Among various geometry enhancements, the code is no longer restricted to deal with single lattice cell problems. The multi-cell capability allows modelling of various lattice structures such as checkerboard lattices, a de-fuelled channel, and core-reflector interface problems. The new resonance method performs distributed resonance self-shielding including the skin effect. This paper describes the main code changes and presents selected code verification results. (authors)
Comparison of calculational methods for liquid metal reactor shields
International Nuclear Information System (INIS)
Carter, L.L.; Moore, F.S.; Morford, R.J.; Mann, F.M.
1985-09-01
A one-dimensional comparison is made between Monte Carlo (MCNP), discrete ordinances (ANISN), and diffusion theory (MlDX) calculations of neutron flux and radiation damage from the core of the Fast Flux Test Facility (FFTF) out to the reactor vessel. Diffusion theory was found to be reasonably accurate for the calculation of both total flux and radiation damage. However, for large distances from the core, the calculated flux at very high energies is low by an order of magnitude or more when the diffusion theory is used. Particular emphasis was placed in this study on the generation of multitable cross sections for use in discrete ordinates codes that are self-shielded, consistent with the self-shielding employed in the generation of cross sections for use with diffusion theory. The Monte Carlo calculation, with a pointwise representation of the cross sections, was used as the benchmark for determining the limitations of the other two calculational methods. 12 refs., 33 figs
A fast resonance interference treatment scheme with subgroup method
International Nuclear Information System (INIS)
Cao, L.; He, Q.; Wu, H.; Zu, T.; Shen, W.
2015-01-01
A fast Resonance Interference Factor (RIF) scheme is proposed to treat the resonance interference effects between different resonance nuclides. This scheme utilizes the conventional subgroup method to evaluate the self-shielded cross sections of the dominant resonance nuclide in the heterogeneous system and the hyper-fine energy group method to represent the resonance interference effects in a simplified homogeneous model. In this paper, the newly implemented scheme is compared to the background iteration scheme, the Resonance Nuclide Group (RNG) scheme and the conventional RIF scheme. The numerical results show that the errors of the effective self-shielded cross sections are significantly reduced by the fast RIF scheme compared with the background iteration scheme and the RNG scheme. Besides, the fast RIF scheme consumes less computation time than the conventional RIF schemes. The speed-up ratio is ~4.5 for MOX pin cell problems. (author)
Fast Breeder Blanket Facility FBBF. Annual report, January 1, 1981-December 31, 1981
International Nuclear Information System (INIS)
Clikeman, F.M.
1982-07-01
This annual report contains a summmary of fission rate, spectra, and gamma-ray heating rate measurements made in the first blanket of the Purdue Fast Breeder Blanket Facility. The first blanket consisted of aluminum clad, natural UO 2 fuel rods with a secondary cladding of stainless steel or aluminum. The blanket was arranged in two concentric regions around the neutron source and converter regions. A neutron diffusion code, 2DB, and a Monte Carlo code, VIM, both using homogeneous cross section groups have been used to calculate the reaction rates. Calculated to experimental values for a number of important reactions are presented. A modified method of applying Bondarenko self-shielding factors to correct for the self shielding of resonance energy neutrons in aluminum, stainless steel and UO 2 has improved the agreement between the calculations and experiment, but does not account for all of the differences
SWANS: A Prototypic SCALE Criticality Sequence for Automated Optimization Using the SWAN Methodology
International Nuclear Information System (INIS)
Greenspan, E.
2001-01-01
SWANS is a new prototypic analysis sequence that provides an intelligent, semi-automatic search for the maximum k eff of a given amount of specified fissile material, or of the minimum critical mass. It combines the optimization strategy of the SWAN code with the composition-dependent resonance self-shielded cross sections of the SCALE package. For a given system composition arrived at during the iterative optimization process, the value of k eff is as accurate and reliable as obtained using the CSAS1X Sequence of SCALE-4.4. This report describes how SWAN is integrated within the SCALE system to form the new prototypic optimization sequence, describes the optimization procedure, provides a user guide for SWANS, and illustrates its application to five different types of problems. In addition, the report illustrates that resonance self-shielding might have a significant effect on the maximum k eff value a given fissile material mass can have
MIT LMFBR blanket physics project progress report No. 7, July 1, 1975--September 30, 1976
International Nuclear Information System (INIS)
Driscoll, M.J.
1976-01-01
Work during the period was devoted primarily to a range of analytical/numerical investigations, including evaluation of means to improve external blanket designs, beneficial attributes of the use of internal blankets, improved methods for the calculation of heterogeneous self-shielding and parametric studies of calculated spectral indices. Experimental work included measurements of the ratio of U-238 captures to U-235 fissions in a standard blanket mockup, and completion of development work on the radiophotoluminescent readout of LiF thermoluminescent detectors. The most significant findings were that there is very little prospect for substantial improvement in the breeding performance of external blankets, but internal blankets continue to show promise, particularly if they are used in such a way as to increase the volume fraction of fuel inside the core envelope. An improved equivalence theorem was developed which may allow use of fast reactor methods to calculate heterogeneously self-shielded cross sections in both fast and thermal reactors
New computational methods used in the lattice code DRAGON
International Nuclear Information System (INIS)
Marleau, G.; Hebert, A.; Roy, R.
1992-01-01
The lattice code DRAGON is used to perform transport calculations inside cells and assemblies for multidimensional geometry using the collision probability method, including the interface current and J ± techniques. Typical geometries that can be treated using this code include CANDU 2-dimensional clusters, CANDU 3-dimensional assemblies, pressurized water reactor (PWR) rectangular and hexagonal assemblies. It contains a self-shielding module for the treatment of microscopic cross section libraries and a depletion module for burnup calculations. DRAGON was written in a modular form in such a way as to accept easily new collision probability options and make them readily available to all the modules that require collision probability matrices like the self-shielding module, the flux solution module and the homogenization module. In this paper the authors present an overview of DRAGON and discuss some of the methods that were implemented in DRAGON in order to improve on its performance
Heavy nucleus resonant absorption calculation benchmarks
International Nuclear Information System (INIS)
Tellier, H.; Coste, H.; Raepsaet, C.; Van der Gucht, C.
1993-01-01
The calculation of the space and energy dependence of the heavy nucleus resonant absorption in a heterogeneous lattice is one of the hardest tasks in reactor physics. Because of the computer time and memory needed, it is impossible to represent finely the cross-section behavior in the resonance energy range for everyday computations. Consequently, reactor physicists use a simplified formalism, the self-shielding formalism. As no clean and detailed experimental results are available to validate the self-shielding calculations, Monte Carlo computations are used as a reference. These results, which were obtained with the TRIPOLI continuous-energy Monte Carlo code, constitute a set of numerical benchmarks than can be used to evaluate the accuracy of the techniques or formalisms that are included in any reactor physics codes. Examples of such evaluations, for the new assembly code APOLLO2 and the slowing-down code SECOL, are given for cases of 238 U and 232 Th fuel elements
Effect of difference between group constants processed by codes TIMS and ETOX on integral quantities
International Nuclear Information System (INIS)
Takano, Hideki; Ishiguro, Yukio; Matsui, Yasushi.
1978-06-01
Group constants of 235 U, 238 U, 239 Pu, 240 Pu and 241 Pu have been produced with the processing code TIMS using the evaluated nuclear data of JENDL-1. The temperature and composition dependent self-shielding factors have been calculated for the two cases with and without considering mutual interference resonant nuclei. By using the group constants set produced by the TIMS code, the integral quantities, i.e. multiplication factor, Na-void reactivity effect and Doppler reactivity effect, are calculated and compared with those calculated with the use of the cross sections set produced by the ETOX code to evaluate accuracy of the approximate calculation method in ETOX. There is much difference in self-shielding factor in each energy group between the two codes. For the fast reactor assemblies under study, however, the integral quantities calculated with these two sets are in good agreement with each other, because of eventual cancelation of errors. (auth.)
Design of a software for gamma detector efficiency
International Nuclear Information System (INIS)
Lopez, G.
2011-01-01
Gamma spectroscopy with highly-pure-germanium detector is one of the most used method for qualitative and quantitative analysis of samples. Nevertheless Gamma spectroscopy results require to be corrected, first for taking into account the self-shielding effect that represents the absorption of the photons by the sample itself and secondly for correcting the fact that 2 photons emitted simultaneously with energy E 1 and E 2 are likely to be simultaneously detected and then counted as a single photon with an energy E 1 +E 2 . This effect is called gamma-gamma coincidence. A software has been designed to simulate both effect and produce correcting factors in the case of cylindrical geometries. This software has been validated on Americium 241 for the self-shielding effect and on Cesium 134 for gamma-gamma coincidence. (A.C.)
Update of WIMS-D libraries using JENDL-3.2, ENDF/B-VI.5 and JEF-2.2
International Nuclear Information System (INIS)
Gil, Choong Sup; Kim, Jung Do; Chang, Jong Wha
2001-01-01
The WIMS-D5 Libraries based on JENDL-3.2, ENDF/B-VI.5, and JEF-2.2 have been prepared and are being tested against the benchmark problems. Several sensitivity calculations for stabililty confirmation of the libraries were carried out such as the fission spectrum dependency, the self shielding effects of the elastic scattering cross sections, the self shielding effects of Pu -240 and Pu -242 capture cross sections below 4.0 eV, etc. The results of benchmark calculations with the libraries based on JENDL-3.2, ENDF/B-VI.5, JEF-2.2, and the '1986 library were intercompared. The predictions of criticalities and isotopic compositions with the updated libraries show good agreements with the measurements or the reference results. The multiplication factors with the library based on JENDL-3.2 are slightly higher than those of ENDF/B-VI.5 and JEF-2.2
Energy Technology Data Exchange (ETDEWEB)
Hong, U.; Choi, S. H.; Kang, H. D. [Kyungsan University, Kyungsan (Korea)
2000-03-01
The four kinds of the fluence monitor considered by self-shielding are design and fabricated for evaluation of neutron irradiation fluence. They are equipped with dosimeters consisting of Ni, Fe and Ti wires and so forth. The nuclear reaction rate is obtained by measurement on dosimeter using the spectroscopic analysis of induced {gamma}-ray. We established the nuetron fluence evaluating technology that is based on the measurement of the reaction rate considering reactor's irradiation history, burn-out, self-shielding in fluence monitor, and the influence of impurity in dosimeter. The distribution of high energy neutron flux on the vertical axis of the capsule shows fifth order polynomial equation and is good agree with theoretical value in the error range of 30% by MCNP/4A code. 22 refs., 50 figs., 27 tabs. (Author)
Energy Technology Data Exchange (ETDEWEB)
Luneville, L
1998-06-01
The multigroup discrete ordinates method is a classical way to solve transport equation (Boltzmann) for neutral particles. Self-shielding effects are not correctly treated due to large variations of cross sections in a group (in the resonance range). To treat the resonance domain, the multiband method is introduced. The main idea is to divide the cross section domain into bands. We obtain the multiband parameters using the moment method; the code CALENDF provides probability tables for these parameters. We present our implementation in an existing discrete ordinates code: SN1D. We study deep penetration benchmarks and show the improvement of the method in the treatment of self-shielding effects. (author) 15 refs.
A new e-beam application in the pharmaceutical industry
International Nuclear Information System (INIS)
Sadat, Theo; Malcolm, Fiona
2005-01-01
The paper presents a new electron beam application in the pharmaceutical industry: an in-line self-shielded atropic transfer system using electron beam for surface decontamination of products entering a pharmaceutical filling line. The unit was developed by Linac Technologies in response to the specifications of a multi-national pharmaceutical company, to solve the risk of microbial contamination entering a filling line housed inside an isolator. In order to fit the sterilization unit inside the pharmaceutical plant, a 'miniature' low-energy (200 keV) electron beam accelerator and e-beam tunnel were designed, all conforming to the pharmaceutical good manufacturing practice (GMP) regulations. Process validation using biological indicators is described, with reference to the regulations governing the pharmaceutical industry. Other industrial applications of a small-sized self-shielded electron beam sterilization unit are mentioned
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
Energy Technology Data Exchange (ETDEWEB)
Blaise, P.
1996-12-18
Resonance parameters for actinides play a significant role in the neutronic characteristics of all reactor types. All the major integral parameters strongly depend on the nuclear data of the isotopes in the resonance-energy regions.The author sets up a method for the adjustment of resonance parameters taking into account the self-shielding effects and restricting the cross section deconvolution problem to a limited energy region. (N.T.).
Fissile Content Assay of Spent Fuel Using LSDS System
International Nuclear Information System (INIS)
Jeon, Ju Young; Lee, Yong Deok; Park, Chang Je
2016-01-01
About 1.5 % fissile materials still exist in the spent fuel. Therefore, for reutilization of fissile materials in spent fuel at SFR, resource material is produced through the pyro process. Fissile material contents in the resource material must be analyzed before fabricating SFR fuel for reactor safety and economics. The new technology for an isotopic fissile material content assay is under development at KAERI using a lead slowing down spectrometer (LSDS). LSDS is very sensitive to distinguish fission signals from each fissile isotope in spent and recycled fuel. In an assay of fissile content of spent fuel and recycled fuel, an intense radiation background gives limits the direct analysis of fissile materials. However, LSDS is not influenced by such a radiation background in a fissile assay. Based on the decided LSDS geometry set up, a self shielding parameter was calculated at the fuel assay zone by introducing spent fuel or pyro produced nuclear material. When nuclear material is inserted into the assay area, the spent fuel assembly or pyro recycled fuel material perturbs the spatial distribution of slowing down neutrons in lead and the prompt fast fission neutrons produced by fissile materials are also perturbed. The self shielding factor is interpreted as how much of the absorption is created inside the fuel area when it is in the lead. The self shielding effect provides a non-linear property in the isotopic fissile assay. When the self shielding is severe, the assay system becomes more complex and needs a special parameter to treat this non linear effect. Additionally, an assay of isotopic fissile content will contribute to an accuracy improvement of the burn-up code and increase the transparency and credibility for spent fuel storage and usage, as internationally increasing demand. The fissile contents result came out almost exactly with relative error ∼ 2% in case of Pu239, Pu241 for two different plutonium contents. In this study, meaningful results were
Auditore, Lucrezia; Amato, Ernesto; Italiano, Antonio; Pagano, Benedetta; Baldari, Sergio
2017-01-01
The production of 18F, 11C, 13N, and 15O positron emitting radionuclides for PET imaging is usually accomplished in Nuclear Medicine Departments through direct nuclear reactions induced by protons accelerated by compact medical cyclotrons on liquid or gaseous targets. Messina University has funded the construction of a PET-cyclotron-radio-chemistry plant at the Messina University Hospital, equipped with a 11 MeV self-shielded cyclotron. We estimated the expected production yields of these nuc...
Current status of low energy EB machine
International Nuclear Information System (INIS)
Toshiro Nishikimi; Shuichi Taniguchi; Kenichi Mizusawa
1999-01-01
Electron beam processing systems have been in use in a variety of applications such as curing of paints and printing inks, crosslinking of PE products, treating of rubber tire and so on. Low energy electron processing systems have become popular as self-shielded machines, which are compact and easy to use and do not require special facility as an irradiation room. This manuscript introduces the status of low energy EB (electron beam) machine through Nissin's products current
Solution of the Doppler broadening function based on the fourier cosine transform
Energy Technology Data Exchange (ETDEWEB)
Goncalves, Alessandro da C [COPPE/UFRJ - Programa de Engenharia Nuclear, Universidade Federal do Rio de Janeiro, P.O. Box 68509, 21941-914 Rio de Janeiro, RJ (Brazil)], E-mail: agoncalves@con.ufrj.br; Martinez, Aquilino S.; Silva, Fernando C. da [COPPE/UFRJ - Programa de Engenharia Nuclear, Universidade Federal do Rio de Janeiro, P.O. Box 68509, 21941-914 Rio de Janeiro, RJ (Brazil)
2008-10-15
This paper provides a new integral representation for the Doppler broadening function {psi}({xi}, x), which is interpreted as being a Fourier cosine transform. This integral form allows the obtaining of an analytical solution in a simple and accurate functional manner as regards the elementary functions. The solution obtained through the new integral representation can be widely used in several applications such as the calculation of self-shielding factors and measurement corrections for the microscopic cross section through the activation technique.
Solution of the Doppler broadening function based on the fourier cosine transform
International Nuclear Information System (INIS)
Goncalves, Alessandro da C; Martinez, Aquilino S.; Silva, Fernando C. da
2008-01-01
This paper provides a new integral representation for the Doppler broadening function ψ(ξ, x), which is interpreted as being a Fourier cosine transform. This integral form allows the obtaining of an analytical solution in a simple and accurate functional manner as regards the elementary functions. The solution obtained through the new integral representation can be widely used in several applications such as the calculation of self-shielding factors and measurement corrections for the microscopic cross section through the activation technique
Nuclear performance optimization of the Be/Li/Th blanket for the fusion breeder
International Nuclear Information System (INIS)
Lee, J.D.; Bandini, B.R.
1985-01-01
More rigorous nuclear analysis, including treatment of resonance self-shielding effects coupled with an optimization procedure, has resulted in improved performance of the Be/Li/Th blanket. Net U-233 breeding ratio has increased 36% (to 0.84) while at an average U-233/Th ratio of 0.5 a/o average energy multiplication has increased only 12% (to 2.1) compared with earlier results
Progress report on the IAEA programme on the standardization of reactor dosimetry measurements
International Nuclear Information System (INIS)
Ertek, C.; Cross, B.; Chernyshev, V.
1979-01-01
This report briefly summarizes present activities, current status and procedures associated with neutron spectrum unfolding by activation technique within the IAEA programme on standardization of reactor radiation measurements. Experimental efforts and calculations related to unfolding are critically analyzed including the most recent techniques, interlaboratory cooperation, direct influence of recently measured cross-sections on the unfolded neutron flux density spectrum, re-evaluation of some cross-sections, neutron self-shielding factors and scattering effects. (author)
Energy Technology Data Exchange (ETDEWEB)
Lee, J.D.; Bandini, B.R.
1985-03-29
More rigorous nuclear analysis, including the treatment of resonance self-shielding effects coupled with an optimization procedure, has resulted in improved performance of the Be/Li/Th blanket. Net U-233 breeding ratio has increased 36% (to 0.84) while at a U/Th ratio of 0.25 a/o energy multiplication has increased 12% (to 2.1) compared with earlier results.
International Nuclear Information System (INIS)
Vertes, P.
1991-04-01
The FDMXPC program package was developed on the basis of the program system FEDMIX written for mainframe computers. The new program package for personal computers was developed for the interpretation of neutron transmission experiments and for producing group averaged infinite diluted and self-shielded cross sections, starting from evaluated data in ENDF format. The package was written for different FORTRAN compilers residing in personal computers under MS-DOS. (R.P.) 12 refs
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
The influence of rhodium burn-up on the sensitivity of rhodium self-powered neutron detectors
International Nuclear Information System (INIS)
Erben, O.
1980-01-01
Depression and self-shielding coefficients are presented for thermal and epithermal neutron flux densities. Functions are shown describing the distribution of beta particle sources on the emitter cross section for 0 to 50% rhodium burnup. The values are calculated of detector sensitivity to thermal and epithermal neutron flux densities for the said burnup for main types of rhodium SPN detectors made by SODERN. (J.B.)
International Nuclear Information System (INIS)
Arsenault, Benoit; Le Tellier, Romain; Hebert, Alain
2008-01-01
The paper presents the results of a first implementation of a Monte Carlo module in DRAGON Version 4 based on the delta-tracking technique. The Monte Carlo module uses the geometry and the self-shielded multigroup cross-sections calculated with a deterministic model. The module has been tested with three different configurations of an ACR TM -type lattice. The paper also discusses the impact of this approach on the efficiency of the Monte Carlo module. (authors)
International Nuclear Information System (INIS)
Lee, J.D.; Bandini, B.R.
1985-01-01
More rigorous nuclear analysis, including the treatment of resonance self-shielding effects coupled with an optimization procedure, has resulted in improved performance of the Be/Li/Th blanket. Net U-233 breeding ratio has increased 36% (to 0.84) while at a U/Th ratio of 0.25 a/o energy multiplication has increased 12% (to 2.1) compared with earlier results
SDSS-IV MaNGA: constraints on the conditions for star formation in galaxy discs
Stark, David V.; Bundy, Kevin A.; Orr, Matthew E.; Hopkins, Philip F.; Westfall, Kyle; Bershady, Matthew; Li, Cheng; Bizyaev, Dmitry; Masters, Karen L.; Weijmans, Anne-Marie; Lacerna, Ivan; Thomas, Daniel; Drory, Niv; Yan, Renbin; Zhang, Kai
2018-02-01
Regions of disc galaxies with widespread star formation tend to be both gravitationally unstable and self-shielded against ionizing radiation, whereas extended outer discs with little or no star formation tend to be stable and unshielded on average. We explore what drives the transition between these two regimes, specifically whether discs first meet the conditions for self-shielding (parametrized by dust optical depth, τ) or gravitational instability (parametrized by a modified version of Toomre's instability parameters, Qthermal, which quantifies the stability of a gas disc that is thermally supported at T = 104 K). We first introduce a new metric formed by the product of these quantities, Qthermalτ, which indicates whether the conditions for disc instability or self-shielding are easier to meet in a given region of a galaxy, and we discuss how Qthermalτ can be constrained even in the absence of direct gas information. We then analyse a sample of 13 galaxies with resolved gas measurements and find that on average galaxies will reach the threshold for disc instabilities (Qthermal 1). Using integral field spectroscopic observations of a sample of 236 galaxies from the Mapping Nearby Galaxies at APO (MaNGA) survey, we find that the value of Qthermalτ in star-forming discs is consistent with similar behaviour. These results support a scenario where disc fragmentation and collapse occurs before self-shielding, suggesting that gravitational instabilities are the primary condition for widespread star formation in galaxy discs. Our results support similar conclusions based on recent galaxy simulations.
Set up of a method for the adjustment of resonance parameters on integral experiments
International Nuclear Information System (INIS)
Blaise, P.
1996-01-01
Resonance parameters for actinides play a significant role in the neutronic characteristics of all reactor types. All the major integral parameters strongly depend on the nuclear data of the isotopes in the resonance-energy regions.The author sets up a method for the adjustment of resonance parameters taking into account the self-shielding effects and restricting the cross section deconvolution problem to a limited energy region. (N.T.)
International Nuclear Information System (INIS)
Hasegawa, A.
1992-01-01
JSSTDL 295n-104γ: A common group cross-section library system has been developed in JAERI to be used in fairly wide range of applications in nuclear industry. This system is composed of a common 295n-104γ group cross-section library based on JENDL-3 nuclear data file and its utility codes. Target of this system is focused to the criticality or shielding calculations in fast and fusion reactors using ANISN, DOT, or MORSE code. Specifications of the common group constants were decided responding to the request from various nuclear data users, particularly from nuclear design group in Japan. Group structure is decided so as to cover almost all group structures currently used in our country. This library includes self-shielding factor tables for primary reactions. A routine for generating macro-scopic cross-section using the self-shielding factor table is also provided. Neutron cross-sections and photon production cross-sections are processed by Prof. GROUCH-G/B code system and γ ray transport cross-sections are generated by GAMLEG-JR. In this paper, outline and present status of the JSSTDL library system is described along with two examples adopted in JENDL-3 benchmark test. One is for shielding calculation, where effects of self-shielding factor (f-table) is shown in conjunction with the analysis of the ASPIS natural iron deep penetration experiment. Without considering resonance self-shielding effect in resonance energy region for resonant nuclides like iron, the results is completely missled in the attenuation profile calculation in the shields. The other example is fast rector criticality calculations of very small critical assemblies with very high enrichment fuel materials where some basic characteristics of this library is presented. (orig.)
A study on physics parameters and flux behaviour for a fast critical facility using ''Baker'' model
International Nuclear Information System (INIS)
Abu-Leilah, M.M.; Hussein, A.Z.; Gaafar, M.A.; Hamouda, I.F.
1983-01-01
Comparative study was performed to emphasize the effects of using different nuclear data systems and methods on the various parameters of the fast reactor. Multigroup libraries as 11 (ANL-5800) and 26 (BNAB-64) energy group systems of nuclear data constants were used in the present work. The calculations were carried out for both infinite dilution (self-shielding factor F= 1) and self-shielded cross sections. Various computer codes were elaborated and derived to meet the conditional requirements for such calculations. The important output of these calculations are the neutron spectra, neutron balance, fission and capture rate distributions, critical mass, breeding ratio in each region and total breeding ratio of the reactor. Five different cases of study were considered employing two systems of constants, infinite dilution and self-shielded cross-sections and treating stainless steel of the reactor as to be substituted by iron. Moreover, calculations have been concerned for averaged one group nuclear data constants which were condensed from the 11 and 26 group systems. Comparisons of the multigroup results with those of the group were made. The condensation process for averaging to one group was done to estimate the effect of such physical simplification on the calculated parameters. The present work results have been compared with many published works. Fair agreements are obtained, which varified the consistance and completeness of the methods implemented and used
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
A Comparison of Monte Carlo and Deterministic Solvers for keff and Sensitivity Calculations
Energy Technology Data Exchange (ETDEWEB)
Haeck, Wim [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Parsons, Donald Kent [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); White, Morgan Curtis [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Saller, Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Favorite, Jeffrey A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2017-12-12
Verification and validation of our solutions for calculating the neutron reactivity for nuclear materials is a key issue to address for many applications, including criticality safety, research reactors, power reactors, and nuclear security. Neutronics codes solve variations of the Boltzmann transport equation. The two main variants are Monte Carlo versus deterministic solutions, e.g. the MCNP [1] versus PARTISN [2] codes, respectively. There have been many studies over the decades that examined the accuracy of such solvers and the general conclusion is that when the problems are well-posed, either solver can produce accurate results. However, the devil is always in the details. The current study examines the issue of self-shielding and the stress it puts on deterministic solvers. Most Monte Carlo neutronics codes use continuous-energy descriptions of the neutron interaction data that are not subject to this effect. The issue of self-shielding occurs because of the discretisation of data used by the deterministic solutions. Multigroup data used in these solvers are the average cross section and scattering parameters over an energy range. Resonances in cross sections can occur that change the likelihood of interaction by one to three orders of magnitude over a small energy range. Self-shielding is the numerical effect that the average cross section in groups with strong resonances can be strongly affected as neutrons within that material are preferentially absorbed or scattered out of the resonance energies. This affects both the average cross section and the scattering matrix.
TRITON, 3-D Multi-Region Neutron Diffusion Burnup with Criticality Search
International Nuclear Information System (INIS)
1974-01-01
1 - Nature of physical problem solved: TRITON is a multigroup diffusion depletion program in three dimensions (x,y,z). In addition to the straight K eff calculation, three types of criticality searches are possible - diluted control isotope search, region-wise smeared control isotope search, region-wise smeared control isotope search, region-wise smeared control isotope boundary search (the control isotope can be smeared over one region or over a group of regions called a control bank). The depletion equations are solved region-wise. More than one microscopic cross section library can be used in the various regions of the reactor. The same is true for self-shielding factors. Such sets of data can be changed at pre-determined time steps. 2 - Method of solution: The mathematical model employed for the solution of the finite difference equations, which is derived from a seven-point approximation of diffusion equations, is an on-line Chebyshev semi- iterative method. 3 - Restrictions on the complexity of the problem: Maximum number of: library sets: 1; self-shielding sets: 10; compositions: 100; self-shielding coefficients: 6000; groups: 10; fuel isotopes: 30; fission products: 29; isotopes: 50; burnable isotopes: 40; control banks: 100; mesh points: 15000; regions: 400; time steps: 100; control areas: 100; small time steps: 200; elements in the control list: 400; x planes: 100; y planes: 100; z planes: 100
APOLLO-2: An advanced transport code for LWRs
International Nuclear Information System (INIS)
Mathonniere, G.
1995-01-01
APOLLO-2 is a fully modular code in which each module corresponds to a specific task: access to the cross-sections libraries, creation of isotopes medium or mixtures, geometry definition, self-shielding calculations, computation of multigroup collision probabilities, flux solver, depletion calculations, transport-transport or transport-diffusion equivalence process, SN calculations, etc... Modules communicate exclusively by ''objects'' containing structured data, these objects are identified and handled by user's given names. Among the major improvements offered by APOLLO-2 the modelization of the self-shielding: it is possible now to deal with a great precision, checked versus Montecarlo calculations, a fuel rod divided into several concentric rings. So the total production of Plutonium is quite better estimated than before and its radial distribution may be predicted also with a good accuracy. Thanks to the versatility of the code some reference calculations and routine ones may be compared easily because only one parameter is changed; for example the self-shielding approximations are modified, the libraries or the flux solver being exactly the same. Other interesting features have been introduced in APOLLO-2: the new isotopes JEF.2 are available in 99 and 172 energy groups libraries, the surface leakage model improves the calculation of the control rod efficiency, the flux-current method allows faster calculations, the possibility of an automatic convergence checking during the depletion calculations coupled with fully automatic corrections, heterogeneous diffusion coefficients used for voiding analysis. 17 refs, 1 tab
Determination of boron over a large dynamic range by prompt-gamma activation analysis
International Nuclear Information System (INIS)
Harrison, R.K.; Landsberger, S.
2009-01-01
An evaluation of the PGAA method for the determination of boron across a wide dynamic range of concentrations was performed for trace levels up to 5 wt.% boron. This range encompasses a transition from neutron transparency to significant self- shielding conditions. To account for self-shielding, several PGAA techniques were employed. First, a calibration curve was developed in which a set of boron standards was tested and the count rate to boron mass curve was determined. This set of boron measurements was compared with an internal standard self-shielding correction method and with a method for determining composition using PGAA peak ratios. The advantages and disadvantages of each method are analyzed. The boron concentrations of several laboratory-grade chemicals and standard reference materials were measured with each method and compared. The evaluation of the boron content of nanocrystalline transition metals prepared with a boron-containing reducing agent was also performed with each of the methods tested. Finally, the k 0 method was used for non-destructive measurement of boron in catalyst materials for the characterization of new non-platinum fuel cell catalysts.
The SWAN-SCALE code for the optimization of critical systems
International Nuclear Information System (INIS)
Greenspan, E.; Karni, Y.; Regev, D.; Petrie, L.M.
1999-01-01
The SWAN optimization code was recently developed to identify the maximum value of k eff for a given mass of fissile material when in combination with other specified materials. The optimization process is iterative; in each iteration SWAN varies the zone-dependent concentration of the system constituents. This change is guided by the equal volume replacement effectiveness functions (EVREF) that SWAN generates using first-order perturbation theory. Previously, SWAN did not have provisions to account for the effect of the composition changes on neutron cross-section resonance self-shielding; it used the cross sections corresponding to the initial system composition. In support of the US Department of Energy Nuclear Criticality Safety Program, the authors recently removed the limitation on resonance self-shielding by coupling SWAN with the SCALE code package. The purpose of this paper is to briefly describe the resulting SWAN-SCALE code and to illustrate the effect that neutron cross-section self-shielding could have on the maximum k eff and on the corresponding system composition
Production of medical radioactive isotopes using KIPT electron driven subcritical facility
International Nuclear Information System (INIS)
Talamo, Alberto; Gohar, Yousry
2008-01-01
Kharkov Institute of Physics and Technology (KIPT) of Ukraine in collaboration with Argonne National Laboratory (ANL) has a plan to construct an electron accelerator driven subcritical assembly. One of the facility objectives is the production of medical radioactive isotopes. This paper presents the ANL collaborative work performed for characterizing the facility performance for producing medical radioactive isotopes. First, a preliminary assessment was performed without including the self-shielding effect of the irradiated samples. Then, more detailed investigation was carried out including the self-shielding effect, which defined the sample size and location for producing each medical isotope. In the first part, the reaction rates were calculated as the multiplication of the cross section with the unperturbed neutron flux of the facility. Over fifty isotopes have been considered and all transmutation channels are used including (n, γ), (n, 2n), (n, p), and (γ, n). In the second part, the parent isotopes with high reaction rate were explicitly modeled in the calculations. Four irradiation locations were considered in the analyses to study the medical isotope production rate. The results show the self-shielding effect not only reduces the specific activity but it also changes the irradiation location that maximizes the specific activity. The axial and radial distributions of the parent capture rates have been examined to define the irradiation sample size of each parent isotope
Production of medical radioactive isotopes using KIPT electron driven subcritical facility
Energy Technology Data Exchange (ETDEWEB)
Talamo, Alberto [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States)], E-mail: alby@anl.gov; Gohar, Yousry [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States)
2008-05-15
Kharkov Institute of Physics and Technology (KIPT) of Ukraine in collaboration with Argonne National Laboratory (ANL) has a plan to construct an electron accelerator driven subcritical assembly. One of the facility objectives is the production of medical radioactive isotopes. This paper presents the ANL collaborative work performed for characterizing the facility performance for producing medical radioactive isotopes. First, a preliminary assessment was performed without including the self-shielding effect of the irradiated samples. Then, more detailed investigation was carried out including the self-shielding effect, which defined the sample size and location for producing each medical isotope. In the first part, the reaction rates were calculated as the multiplication of the cross section with the unperturbed neutron flux of the facility. Over fifty isotopes have been considered and all transmutation channels are used including (n, {gamma}), (n, 2n), (n, p), and ({gamma}, n). In the second part, the parent isotopes with high reaction rate were explicitly modeled in the calculations. Four irradiation locations were considered in the analyses to study the medical isotope production rate. The results show the self-shielding effect not only reduces the specific activity but it also changes the irradiation location that maximizes the specific activity. The axial and radial distributions of the parent capture rates have been examined to define the irradiation sample size of each parent isotope.
Production of medical radioactive isotopes using KIPT electron driven subcritical facility.
Talamo, Alberto; Gohar, Yousry
2008-05-01
Kharkov Institute of Physics and Technology (KIPT) of Ukraine in collaboration with Argonne National Laboratory (ANL) has a plan to construct an electron accelerator driven subcritical assembly. One of the facility objectives is the production of medical radioactive isotopes. This paper presents the ANL collaborative work performed for characterizing the facility performance for producing medical radioactive isotopes. First, a preliminary assessment was performed without including the self-shielding effect of the irradiated samples. Then, more detailed investigation was carried out including the self-shielding effect, which defined the sample size and location for producing each medical isotope. In the first part, the reaction rates were calculated as the multiplication of the cross section with the unperturbed neutron flux of the facility. Over fifty isotopes have been considered and all transmutation channels are used including (n, gamma), (n, 2n), (n, p), and (gamma, n). In the second part, the parent isotopes with high reaction rate were explicitly modeled in the calculations. Four irradiation locations were considered in the analyses to study the medical isotope production rate. The results show the self-shielding effect not only reduces the specific activity but it also changes the irradiation location that maximizes the specific activity. The axial and radial distributions of the parent capture rates have been examined to define the irradiation sample size of each parent isotope.
MC^{2}-3: Multigroup Cross Section Generation Code for Fast Reactor Analysis
Energy Technology Data Exchange (ETDEWEB)
Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Yang, W. S. [Argonne National Lab. (ANL), Argonne, IL (United States)
2013-11-08
The MC^{2}-3 code is a Multigroup Cross section generation Code for fast reactor analysis, developed by improving the resonance self-shielding and spectrum calculation methods of MC^{2}-2 and integrating the one-dimensional cell calculation capabilities of SDX. The code solves the consistent P1 multigroup transport equation using basic neutron data from ENDF/B data files to determine the fundamental mode spectra for use in generating multigroup neutron cross sections. A homogeneous medium or a heterogeneous slab or cylindrical unit cell problem is solved in ultrafine (~2000) or hyperfine (~400,000) group levels. In the resolved resonance range, pointwise cross sections are reconstructed with Doppler broadening at specified isotopic temperatures. The pointwise cross sections are directly used in the hyperfine group calculation whereas for the ultrafine group calculation, self-shielded cross sections are prepared by numerical integration of the pointwise cross sections based upon the narrow resonance approximation. For both the hyperfine and ultrafine group calculations, unresolved resonances are self-shielded using the analytic resonance integral method. The ultrafine group calculation can also be performed for two-dimensional whole-core problems to generate region-dependent broad-group cross sections. Multigroup cross sections are written in the ISOTXS format for a user-specified group structure. The code is executable on UNIX, Linux, and PC Windows systems, and its library includes all isotopes of the ENDF/BVII. 0 data.
International Nuclear Information System (INIS)
Santamarina, A.
1991-01-01
A criticality-safety calculational scheme using the automated deterministic code system, APOLLO-BISTRO, has been developed. The cell/assembly code APOLLO is used mainly in LWR and HCR design calculations, and its validation spans a wide range of moderation ratios, including voided configurations. Its recent 99-group library and self-shielded cross-sections has been extensively qualified through critical experiments and PWR spent fuel analysis. The PIC self-shielding formalism enables a rigorous treatment of the fuel double heterogeneity in dissolver medium calculations. BISTRO is an optimized multidimensional SN code, part of the modular CCRR package used mainly in FBR calculations. The APOLLO-BISTRO scheme was applied to the 18 experimental benchmarks selected by the OECD/NEACRP Criticality Calculation Working Group. The Calculation-Experiment discrepancy was within ± 1% in ΔK/K and always looked consistent with the experimental uncertainty margin. In the critical experiments corresponding to a dissolver type benchmark, our tools computed a satisfactory Keff. In the VALDUC fuel storage experiments, with hafnium plates, the computed Keff ranged between 0.994 and 1.003 for the various watergaps spacing the fuel clusters from the absorber plates. The APOLLO-KENOEUR statistic calculational scheme, based on the same self-shielded multigroup library, supplied consistent results within 0.3% in ΔK/K. (Author)
Nuclear data adjustment methodology utilizing resonance parameter sensitivities and uncertainties
International Nuclear Information System (INIS)
Broadhead, B.L.
1983-01-01
This work presents the development and demonstration of a Nuclear Data Adjustment Method that allows inclusion of both energy and spatial self-shielding into the adjustment procedure. The resulting adjustments are for the basic parameters (i.e. resonance parameters) in the resonance regions and for the group cross sections elsewhere. The majority of this development effort concerns the production of resonance parameter sensitivity information which allows the linkage between the responses of interest and the basic parameters. The resonance parameter sensitivity methodology developed herein usually provides accurate results when compared to direct recalculations using existng and well-known cross section processing codes. However, it has been shown in several cases that self-shielded cross sections can be very non-linear functions of the basic parameters. For this reason caution must be used in any study which assumes that a linear relatonship exists between a given self-shielded group cross section and its corresponding basic data parameters. The study also has pointed out the need for more approximate techniques which will allow the required sensitivity information to be obtained in a more cost effective manner
Correction of rhodium detector signals for comparison to design calculations
International Nuclear Information System (INIS)
Judd, J.L.; Chang, R.Y.; Gabel, C.W.
1989-01-01
Rhodium detectors are used in many commercial pressurized water reactors PWRs [pressurized water reactor] as in-core neutron detectors. The signals from the detectors are the result of neutron absorption in 103 Rh and the subsequent beta decay of 104 Rh to 104 Pd. The rhodium depletes ∼1% per full-power month, so corrections are necessary to the detector signal to account for the effects of the rhodium depletion. These corrections result from the change in detector self-shielding with rhodium burnup and the change in rhodium concentration itself. Correction for the change in rhodium concentration is done by multiplication of the factor N(t)/N 0 , where N(t) is the rhodium concentration at time t and N 0 is the initial rhodium concentration. The calculation of the self-shielding factor is more complicated and is presented. A self-shielding factor based on the fraction of rhodium remaining was calculated with the CASMO-3 code. The results obtained from our comparisons of predicted and measured in-core detector signals show that the CASMO-3/SIMULATE-3 code package is an effective tool for estimating pin peaking and power distributions
Application of Bondarenko formalism to fusion reactors
International Nuclear Information System (INIS)
Soran, P.D.; Dudziak, D.J.
1975-01-01
The Bondarenko formalism used to account for resonance self-shielding effects (temperature and composition) in a Reference Theta-Pinch Reactor is reviewed. A material of interest in the RTPR blanket is 93 Nb, which exhibits a large number of capture resonance in the energy region below 800 keV. Although Nb constitutes a small volume fraction of the blanket, its presence significantly affects the nucleonic properties of the RTPR blanket. The effects of self-shielding in 93 Nb on blanket parameters such as breeding ratio, total afterheat, radioactivity, magnet-coil heating and total energy depositions have been studied. Resonance self-shielding of 93 Nb, as compared to unshielded cross sections, will increase tritium breeding by approximately 7 percent in the RTPR blanket and will decrease blanket radioactivity, total recoverable energy, and magnet-coil heating. Temperature effects change these parameters by less than 2 percent. The method is not restricted to the RTPR, as a single set of Bondarenko f-factors is suitable for application to a variety of fusion reactor designs
Computation of a Canadian SCWR unit cell with deterministic and Monte Carlo codes
International Nuclear Information System (INIS)
Harrisson, G.; Marleau, G.
2012-01-01
The Canadian SCWR has the potential to achieve the goals that the generation IV nuclear reactors must meet. As part of the optimization process for this design concept, lattice cell calculations are routinely performed using deterministic codes. In this study, the first step (self-shielding treatment) of the computation scheme developed with the deterministic code DRAGON for the Canadian SCWR has been validated. Some options available in the module responsible for the resonance self-shielding calculation in DRAGON 3.06 and different microscopic cross section libraries based on the ENDF/B-VII.0 evaluated nuclear data file have been tested and compared to a reference calculation performed with the Monte Carlo code SERPENT under the same conditions. Compared to SERPENT, DRAGON underestimates the infinite multiplication factor in all cases. In general, the original Stammler model with the Livolant-Jeanpierre approximations are the most appropriate self-shielding options to use in this case of study. In addition, the 89 groups WIMS-AECL library for slight enriched uranium and the 172 groups WLUP library for a mixture of plutonium and thorium give the most consistent results with those of SERPENT. (authors)
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)
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)
International Nuclear Information System (INIS)
Lee, Sang Chul; Kim, Won Seok; Han, Byong Su; Moon, Joo Hyun
2013-01-01
The increase of the operating NPPs and the superannuation of the equipment in NPPs cause a large amount of the metal radioactive waste. Presently the metal radioactive wastes are stored in the temporary storage facility in NPPs because of the delay of the construction of the final disposal facility. The radioactive level of general metal radioactive wastes is low, and the radioactive level can be lowered by the simple decontamination process. If the radioactive wastes are disposed as the industry waste, the disposal cost is diminished largely. For the disposal of the radioactive wastes as the industrial wastes, the radioactive level of the target wastes are evaluated. It is difficult to know the position of the source term for most of the metal radioactive and the source term is distributed non-homogeneously. And the self-shielding effect of the metal material makes the evaluation more difficult. In this study, the radioactive inventory evaluation system for the metal radioactive waste is developed. For the correction of the uncertainty of the position and the non-homogeneity of the source term, the 3D shape and multiple radiation measurement are used. The existing gamma-ray measurement system for the metal radioactive waste cannot reflect the position and the distribution of the source term and the effect of self-shielding. This evaluation system suggested in this system can calculate the reasonable value regarding to the position and the distribution of the source term and the effect of self-shielding. By the calculation of the partial inventory of the target metal waste, the advantage in the application of the clearance criteria can be obtained
Evaluation and validation of criticality codes for fuel dissolver calculations
International Nuclear Information System (INIS)
Santamarina, A.; Smith, H.J.; Whitesides, G.E.
1991-01-01
During the past ten years an OECD/NEA Criticality Working Group has examined the validity of criticality safety computational methods. International calculation tools which were shown to be valid in systems for which experimental data existed were demonstrated to be inadequate when extrapolated to fuel dissolver media. The spread of the results in the international calculation amounted to ± 12,000 pcm in the realistic fuel dissolver exercise n degrees 19 proposed by BNFL, and to ± 25,000 pcm in the benchmark n degrees 20 in which fissile material in solid form is surrounded by fissile material in solution. A theoretical study of the main physical parameters involved in fuel dissolution calculations was performed, i.e. range of moderation, variation of pellet size and the fuel double heterogeneity effect. The APOLLO/P IC method developed to treat latter effect, permits us to supply the actual reactivity variation with pellet dissolution and to propose international reference values. The disagreement among contributors' calculations was analyzed through a neutron balance breakdown, based on three-group microscopic reaction rates solicited from the participants. The results pointed out that fast and resonance nuclear data in criticality codes are not sufficiently reliable. Moreover the neutron balance analysis emphasized the inadequacy of the standard self-shielding formalism (NITAWL in the international SCALE package) to account for 238 U resonance mutual self-shielding in the pellet-fissile liquor interaction. Improvements in the up-dated 1990 contributions, as do recent complementary reference calculations (MCNP, VIM, ultrafine slowing-down CGM calculation), confirm the need to use rigorous self-shielding methods in criticality design-oriented codes. 6 refs., 11 figs., 3 tabs
A New Equivalence Theory Method for Doubly Heterogeneous Fuel
International Nuclear Information System (INIS)
Choi, Sooyoung; Lee, Deokjung
2014-01-01
The unique characteristics cannot be handled easily by conventional computer code. A new methodology is being developed to treat resonance self-shielding in a doubly heterogeneous system. The method first homogenizes the material in the fuel compact region using an analytical approximation for the disadvantage factor based on equivalence theory. The disadvantage factor accounts for spatial self-shielding of the resonance flux within the fuel grains. The doubly-heterogeneous effects are accounted by using a modified definition of background cross section, which includes geometry parameters and the cross sections of both the fuel grain and fuel compact regions. For the verification, the new DH methodology was implemented in deterministic transport code TICTOC developed at UNIST which uses equivalence theory for resonance treatment and Method of Characteristics (MOC) for the ray tracing. In previous research, this new methodology was verified for several pin cell problems but further verification is required to confirm the validity of the methodology for various situations. Therefore, in this study, 9 cases for unit pin cell problems are designed and the accuracy of the new DH method is compared to the Monte Carlo code, McCARD. The new method for doubly-heterogeneous self-shielding using equivalence theory was summarized and calculation procedure was presented. The new methods use analytical expression for the disadvantage factor therefore additional complicated module is not required. The new method was verified for 9 pin cell models. As a result, TICTOC with the new DH method predicts the eigenvalues within about 200 pcm error compared with Monte Carlo results for the most of problems
D'Aloisio, Anson
2017-08-01
In recent years, HST surveys such as CANDELS, HUDF, BoRG/HIPPIES, ERS, and the Frontier Fields, have made possible the first robust measurements of the rest-frame UV luminosity function of z =6-10 galaxies, spanning much of the redshift range over which reionization likely occurred. These measurements provide an estimate of the galactic ionizing photon output, addressing the critical question of whether these galaxies could have reionized the Universe. In addition, follow-up spectroscopy has measured the fraction of these galaxies that show Lyman-alpha emission. Interestingly, a dramatic decrease in this fraction above z 6 has been observed, and this evolution has (controversially) been interpreted as evidence that much of reionization happened over z=6-8 (as intergalactic neutral gas leads to large damping wings that scatter the Lyman-alpha line). The clumpiness of the IGM and how it self shields to ionizing photons impacts whether the observed population of galaxies can reionize the Universe, as well as the interpretation of the evolving Lyman-alpha emitter fraction. We propose to run fully coupled radiative-hydrodynamics simulations that are the first to resolve the evaporation of small structures by passing ionization fronts and, hence, to accurately assess the level of clumpiness and self-shielding from the IGM. Our study will nail down the clumping factor used to assess whether the observed population of galaxies can drive reionization, and it will address whether neutral self-shielding clumps in recently reionized regions can scatter galaxies' Lyman-alpha lines.
A Cross-Section Adjustment Method for Double Heterogeneity Problem in VHTGR Analysis
International Nuclear Information System (INIS)
Yun, Sung Hwan; Cho, Nam Zin
2011-01-01
Very High Temperature Gas-Cooled Reactors (VHTGRs) draw strong interest as candidates for a Gen-IV reactor concept, in which TRISO (tristructuralisotropic) fuel is employed to enhance the fuel performance. However, randomly dispersed TRISO fuel particles in a graphite matrix induce the so-called double heterogeneity problem. For design and analysis of such reactors with the double heterogeneity problem, the Monte Carlo method is widely used due to its complex geometry and continuous-energy capabilities. However, its huge computational burden, even in the modern high computing power, is still problematic to perform wholecore analysis in reactor design procedure. To address the double heterogeneity problem using conventional lattice codes, the RPT (Reactivityequivalent Physical Transformation) method considers a homogenized fuel region that is geometrically transformed to provide equivalent self-shielding effect. Another method is the coupled Monte Carlo/Collision Probability method, in which the absorption and nu-fission resonance cross-section libraries in the deterministic CPM3 lattice code are modified group-wise by the double heterogeneity factors determined by Monte Carlo results. In this paper, a new two-step Monte Carlo homogenization method is described as an alternative to those methods above. In the new method, a single cross-section adjustment factor is introduced to provide self-shielding effect equivalent to the self-shielding in heterogeneous geometry for a unit cell of compact fuel. Then, the homogenized fuel compact material with the equivalent cross-section adjustment factor is used in continuous-energy Monte Carlo calculation for various types of fuel blocks (or assemblies). The procedure of cross-section adjustment is implemented in the MCNP5 code
Measurement of neutron flux distribution by semiconductor detector
International Nuclear Information System (INIS)
Obradovic, D.; Bosevski, T.
1964-01-01
Application of semiconductor detectors for measuring neutron flux distribution is about 10 times faster than measurements by activation foils and demands significantly lower reactor power. Following corrections are avoided: mass of activation foils which influences the self shielding, nuclear decay during activity measurements; counter dead-time. It is possible to control the measured data during experiment and repeat measurements if needed. Precision of the measurement is higher since it is possible to choose the wanted statistics. The method described in this paper is applied for measurements at the RB reactor. It is concluded that the method is suitable for fast measurements but the activation analysis is still indispensable
Energy Technology Data Exchange (ETDEWEB)
Talamo, A.; Gohar, Y.; Nuclear Engineering Division
2007-05-15
Kharkov Institute of Physics and Technology (KIPT) of Ukraine has a plan to construct an accelerator driven subcritical assembly. The main functions of the subcritical assembly are the medical isotope production, neutron thereby, and the support of the Ukraine nuclear industry. Reactor physics experiments and material research will be carried out using the capabilities of this facility. The United States of America and Ukraine have started collaboration activity for developing a conceptual design for this facility with low enrichment uranium (LEU) fuel. Different conceptual designs are being developed based on the facility mission and the engineering requirements including nuclear physics, neutronics, heat transfer, thermal hydraulics, structure, and material issues. Different fuel designs with LEU and reflector materials are considered in the design process. Safety, reliability, and environmental considerations are included in the facility conceptual design. The facility is configured to accommodate future design improvements and upgrades. This report is a part of the Argonne National Laboratory Activity within this collaboration for developing and characterizing the subcritical assembly conceptual design. In this study, the medical isotope production function of the Kharkov facility is defined. First, a review was carried out to identify the medical isotopes and its medical use. Then a preliminary assessment was performed without including the self-shielding effect of the irradiated samples. Finally, more detailed investigation was carried out including the self-shielding effect, which defined the sample size and irradiation location for producing each medical isotope. In the first part, the reaction rates were calculated as the multiplication of the cross section with the unperturbed neutron flux of the facility. Over fifty isotopes were considered and all transmutation channels are used including (n,{gamma}), (n,2n), (n,p), and ({gamma},n). In the second part
Resonance parameters of the 6.67-, 20.9-, and 36.8-eV levels in 238U
International Nuclear Information System (INIS)
Olsen, D.K.; de Saussure, G.; Perez, R.B.; Difilippo, F.C.
1976-01-01
The ENDF/B-IV 238 U cross sections (MAT-1262) yield an effective capture resonance integral in strongly self-shielded situations which is too high. This situation suggests that the ENDF/B capture widths for the first few s-wave levels may be too large. Recent ORELA measurements of transmission through 238 U have been analyzed with a multilevel formula to determine the parameters of the 6.67-, 20.9-, and 36.6-eV levels. These three levels provide 86 percent of the infinitely dilute capture resonance integral
An investigation of tungsten by neutron activation techniques
International Nuclear Information System (INIS)
Svetsreni, R.
1978-01-01
This investigation used neutron from Plutonium-Beryllium source (5 curie) to analyse the amount of tungsten in tungsten oxide which was extracted from tungsten ores, slag and tungsten alloy of tungsten iron and carbon. The technique of neutron activation analysis with NaI(Tl) gamma detector 3'' x 3'' and 1024 multichannel analyzer. The dilution technique was used by mixing Fe 2 O 3 or pure sand into the sample before irradiation. In this study self shielding effect in the analysis of tungsten was solved and the detection limit of the tungsten in the sample was about 0.5%
Capability and limitation study of the DDT passive-active neutron waste assay instrument
International Nuclear Information System (INIS)
Nicholas, N.J.; Coop, K.L.; Estep, R.J.
1992-05-01
The differential-dieaway-technique passive-active neutron assay system is widely used by transuranic waste generators to certify their drummed waste for eventual shipment to the Waste Isolation Pilot Plant (WIPP). Stricter criteria being established for waste emplacement at the WIPP site has led to a renewed interest in improvements to and a better understanding of current nondestructive assay (NDA) techniques. Our study includes the effects of source position, extreme matrices, high neutron backgrounds, and source self-shielding to explore the system's capabilities and limitations and to establish a basis for comparison with other NDA systems. 11 refs
Theory and application of deterministic multidimensional pointwise energy lattice physics methods
International Nuclear Information System (INIS)
Zerkle, M.L.
1999-01-01
The theory and application of deterministic, multidimensional, pointwise energy lattice physics methods are discussed. These methods may be used to solve the neutron transport equation in multidimensional geometries using near-continuous energy detail to calculate equivalent few-group diffusion theory constants that rigorously account for spatial and spectral self-shielding effects. A dual energy resolution slowing down algorithm is described which reduces the computer memory and disk storage requirements for the slowing down calculation. Results are presented for a 2D BWR pin cell depletion benchmark problem
Theory and application of the RAZOR two-dimensional continuous energy lattice physics code
International Nuclear Information System (INIS)
Zerkle, M.L.; Abu-Shumays, I.K.; Ott, M.W.; Winwood, J.P.
1997-01-01
The theory and application of the RAZOR two-dimensional, continuous energy lattice physics code are discussed. RAZOR solves the continuous energy neutron transport equation in one- and two-dimensional geometries, and calculates equivalent few-group diffusion theory constants that rigorously account for spatial and spectral self-shielding effects. A dual energy resolution slowing down algorithm is used to reduce computer memory and disk storage requirements for the slowing down calculation. Results are presented for a 2D BWR pin cell depletion benchmark problem
Elements of transport and emplacement system
International Nuclear Information System (INIS)
1981-02-01
This report, undertaken to review proposals for transport, handling and emplacement of high-level radioactive wastes in an underground repository, appropriate to the U.K. context, falls under the headings: basic design concepts; waste block size and configuration; self-shielded or partially shielded blocks; concept of emplacement in long boreholes; concept of emplacement in short boreholes; concept of emplacement in tunnels; methods of emplacement; stages of disposal; repository access by adit, incline or shaft; handling techniques within repository; conventional and radiological safety; costs; areas for further research and development. (U.K.)
Large Sample Neutron Activation Analysis: A Challenge in Cultural Heritage Studies
International Nuclear Information System (INIS)
Stamatelatos, I.E.; Tzika, F.
2007-01-01
Large sample neutron activation analysis compliments and significantly extends the analytical tools available for cultural heritage and authentication studies providing unique applications of non-destructive, multi-element analysis of materials that are too precious to damage for sampling purposes, representative sampling of heterogeneous materials or even analysis of whole objects. In this work, correction factors for neutron self-shielding, gamma-ray attenuation and volume distribution of the activity in large volume samples composed of iron and ceramic material were derived. Moreover, the effect of inhomogeneity on the accuracy of the technique was examined
Radiobiological studies using gamma and x rays.
Energy Technology Data Exchange (ETDEWEB)
Potter, Charles Augustus; Longley, Susan W.; Scott, Bobby R.; Lin, Yong; Wilder, Julie; Hutt, Julie A.; Padilla, Mabel T.; Gott, Katherine M.
2013-02-01
There are approximately 500 self-shielded research irradiators used in various facilities throughout the U.S. These facilities use radioactive sources containing either 137Cs or 60Co for a variety of biological investigations. A report from the National Academy of Sciences[1] described the issues with security of particular radiation sources and the desire for their replacement. The participants in this effort prepared two peer-reviewed publications to document the results of radiobiological studies performed using photons from 320-kV x rays and 137Cs on cell cultures and mice. The effectiveness of X rays was shown to vary with cell type.
Comparisons of the MCNP criticality benchmark suite with ENDF/B-VI.8, JENDL-3.3, and JEFF-3.0
International Nuclear Information System (INIS)
Kim, Do Heon; Gil, Choong-Sup; Kim, Jung-Do; Chang, Jonghwa
2003-01-01
A comparative study has been performed with the latest evaluated nuclear data libraries ENDF/B-VI.8, JENDL-3.3, and JEFF-3.0. The study has been conducted through the benchmark calculations for 91 criticality problems with the libraries processed for MCNP4C. The calculation results have been compared with those of the ENDF60 library. The self-shielding effects of the unresolved-resonance (UR) probability tables have also been estimated for each library. The χ 2 differences between the MCNP results and experimental data were calculated for the libraries. (author)
International Nuclear Information System (INIS)
Konno, Chikara; Ikeda, Yujiro
2003-01-01
A new general purpose neutron incident nuclear data library, NMLIB, for neutron energy up to 3 GeV is generated by using NMTC/JAM for neutron cross sections above 20 MeV and JENDL-3.2 or -3.3 for neutron cross sections below 20 MeV. MATXS files (NMLIB/MG) and ACE files (NMLIB/MC) of NMLIB are also produced from HMLIB with the NJOY code. A simple test calculation suggests that NMLIB, NMLIB/MG and NMLIB/MC have no problem except for self-shielding. (author)
Development of a 1200 fine group nuclear data library for advanced nuclear systems
Institute of Scientific and Technical Information of China (English)
Jun Zou; Lei-Ming Shang; Fang Wang; Li-Juan Hao
2017-01-01
Accurate and reliable nuclear data libraries are essential for calculation and design of advanced nuclear systems.A 1200 fine group nuclear data library Hybrid Evaluated Nuclear Data Library/Fine Group (HENDL/FG) with neutrons of up to 150 MeV has been developed to improve the accuracy of neutronics calculations and analysis.Corrections of Doppler,resonance self-shielding,and thermal upscatter effects were done for HENDL/FG.Shielding and critical safety benchmarks were performed to test the accuracy and reliability of the library.The discrepancy between calculated and measured nuclear parameters fell into a reasonable range.
International Nuclear Information System (INIS)
1981-02-01
This paper is an 'executive summary' of work undertaken to review proposals for transport, handling and emplacement of high level radioactive wastes in an underground repository, appropriate to the U.K. context, with particular reference to: waste block size and configuration; self-shielded or partially-shielded block; stages of disposal; transportation within the repository; emplacement in vertical holes or horizontal tunnels; repository access by adit, incline or shaft; and costs. The paper contains a section on general conclusions and recommendations. (U.K.)
International Nuclear Information System (INIS)
Talamo, A.; Gohar, Y.
2007-01-01
Kharkov Institute of Physics and Technology (KIPT) of Ukraine has a plan to construct an accelerator driven subcritical assembly. The main functions of the subcritical assembly are the medical isotope production, neutron thereby, and the support of the Ukraine nuclear industry. Reactor physics experiments and material research will be carried out using the capabilities of this facility. The United States of America and Ukraine have started collaboration activity for developing a conceptual design for this facility with low enrichment uranium (LEU) fuel. Different conceptual designs are being developed based on the facility mission and the engineering requirements including nuclear physics, neutronics, heat transfer, thermal hydraulics, structure, and material issues. Different fuel designs with LEU and reflector materials are considered in the design process. Safety, reliability, and environmental considerations are included in the facility conceptual design. The facility is configured to accommodate future design improvements and upgrades. This report is a part of the Argonne National Laboratory Activity within this collaboration for developing and characterizing the subcritical assembly conceptual design. In this study, the medical isotope production function of the Kharkov facility is defined. First, a review was carried out to identify the medical isotopes and its medical use. Then a preliminary assessment was performed without including the self-shielding effect of the irradiated samples. Finally, more detailed investigation was carried out including the self-shielding effect, which defined the sample size and irradiation location for producing each medical isotope. In the first part, the reaction rates were calculated as the multiplication of the cross section with the unperturbed neutron flux of the facility. Over fifty isotopes were considered and all transmutation channels are used including (n,γ), (n,2n), (n,p), and (γ,n). In the second part, the parent
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
Thermal reactor benchmark testing of CENDL-2 and ENDF/B-6
International Nuclear Information System (INIS)
Liu Guisheng
1996-01-01
In order to test CENDL-2, ten homogeneous and eight heterogeneous thermal assemblies were used. Both of 123 group cross section libraries based on CENDL-2 and ENDF/B-6 were generated by a nuclear data processing system NSLINK, respectively. The calculations of resonance self-shielding, cell spectra, cell reaction rate ratios and effective multiplication factors (K eff ) of these assemblies have been performed by the modified PASC-1 code system. The calculated results using CENDL-2 show an excellent agreement with corresponding experimental values. However, for some assemblies the K eff values calculated by ENDF/B-6 data are underestimated. (7 tabs.)
Energy Technology Data Exchange (ETDEWEB)
Obradovic, D; Bosevski, T [Institut za nuklearne nauke Boris Kidric, Vinca, Beograd (Yugoslavia)
1964-07-01
Application of semiconductor detectors for measuring neutron flux distribution is about 10 times faster than measurements by activation foils and demands significantly lower reactor power. Following corrections are avoided: mass of activation foils which influences the self shielding, nuclear decay during activity measurements; counter dead-time. It is possible to control the measured data during experiment and repeat measurements if needed. Precision of the measurement is higher since it is possible to choose the wanted statistics. The method described in this paper is applied for measurements at the RB reactor. It is concluded that the method is suitable for fast measurements but the activation analysis is still indispensable.
ECNJEFI. A JEFI based 219-group neutron cross-section library: User's manual
International Nuclear Information System (INIS)
Stad, R.C.L. van der; Gruppelaar, H.
1992-07-01
This manual describes the contents of the ECNJEF1 library. The ECNJEF1 library is a JEF1.1 based 219-group AMPX-Master library for reactor calculations with the AMPX/SCALE-system, e.g. the PASC-3 system as implemented at the Netherlands Energy Research Foundation in Petten, Netherlands. The group cross-section data were generated with NJOY and NPTXS/XLACS-2 from the AMPX system. The data on the ECNJEF1 library allows resolved-resonance treatment by NITAWL and/or unresolved resonance self-shielding by BONAMI. These codes are based upon the Nordheim and Bondarenko methods, respectively. (author). 10 refs., 7 tabs
Calculation results and experimental testing of doppler feedback coefficients
International Nuclear Information System (INIS)
Yang Shunhai
1989-01-01
The Doppler feedback coefficients are calculated by the interpolation and group collapsing method from multigroup self-shielding factors and infinite dilution cross sections rather than effective resonance integrals by using resonance data base. Since many updated sets of multigroup data are in existence to be selected, the calculation process can be simplified. The heterogeneous effects are taken into account by equivalence relation. The computer code of Doppler feedback coefficients is created on computer CYBER-825 and PDP-11. The results calculated are in good agreement with the experiments
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
Thermal reactor benchmark testing of CENDL-2 and ENDF/B-6
Energy Technology Data Exchange (ETDEWEB)
Guisheng, Liu [Chinese Nuclear Data Center, Beijing, BJ (China)
1996-06-01
In order to test CENDL-2, ten homogeneous and eight heterogeneous thermal assemblies were used. Both of 123 group cross section libraries based on CENDL-2 and ENDF/B-6 were generated by a nuclear data processing system NSLINK, respectively. The calculations of resonance self-shielding, cell spectra, cell reaction rate ratios and effective multiplication factors (K{sub eff}) of these assemblies have been performed by the modified PASC-1 code system. The calculated results using CENDL-2 show an excellent agreement with corresponding experimental values. However, for some assemblies the K{sub eff} values calculated by ENDF/B-6 data are underestimated. (7 tabs.).
International Nuclear Information System (INIS)
Erben, O.
1980-01-01
The coefficients of thermal and epithermal neutron flux density depression and self-shielding for the SPN detectors with vanadium, rhodium, silver and cobalt emitters are presented, (for cobalt SPN detectors the functions describing the absorbtion of neutrons along the emitter cross-section are also shown). Using these coefficients and previously published beta particle escape efficiencies, sensitivities are determined for the principal types of detectors produced by Les Cables de Lyon and SODERN companies. The experiments and their results verifying the validity of the theoretical work are described. (author)
Results on the neutron energy distribution measurements at the RECH-1 Chilean nuclear reactor
Energy Technology Data Exchange (ETDEWEB)
Aguilera, P., E-mail: paguilera87@gmail.com; Romero-Barrientos, J. [Comisión Chilena de Energía Nuclear, Nueva Bilbao 12501, La Reina, Santiago (Chile); Universidad de Chile, Dpto. de Física, Facultad de Ciencias, Las Palmeras 3425, Nuñoa, Santiago (Chile); Molina, F. [Comisión Chilena de Energía Nuclear, Nueva Bilbao 12501, La Reina, Santiago (Chile)
2016-07-07
Neutron activations experiments has been perform at the RECH-1 Chilean Nuclear Reactor to measure its neutron flux energy distribution. Samples of pure elements was activated to obtain the saturation activities for each reaction. Using - ray spectroscopy we identify and measure the activity of the reaction product nuclei, obtaining the saturation activities of 20 reactions. GEANT4 and MCNP was used to compute the self shielding factor to correct the cross section for each element. With the Expectation-Maximization algorithm (EM) we were able to unfold the neutron flux energy distribution at dry tube position, near the RECH-1 core. In this work, we present the unfolding results using the EM algorithm.
JNC results of BN-600 benchmark calculation (phase 3)
International Nuclear Information System (INIS)
Ishikawa, M.
2002-01-01
The present work is the result of phase 3 BN-600 core benchmark problem, meaning burnup and heterogeneity. Analytical method applied consisted of: JENDL-3.2 nuclear data library, group constants (70 group, ABBN type self shielding transport factors), heterogeneous cell model for fuel and control rod, basic diffusion calculation (CITATION code), transport theory and mesh size correction (NSHEX code based on SN transport nodal method developed by JNC). Burnup and heterogeneity calculation results are presented obtained by applying both diffusion and transport approach for beginning and end of cycle
International Nuclear Information System (INIS)
Cullen, D.E.
1979-01-01
Program RECENT reconstructs energy-dependent neutron total, elastic, capture, and fission cross sections from a combination of resonance parameters and tabulated background cross sections in the ENDF/B format. Entire evaluations, not just cross sections, are written to the result file, which is in ENDF/B format. The output includes the original resonance parameters in a form that can be used in Doppler broadening and self-shielding calculations. A listing of the source deck is available on request. 5 figures, 5 tables
Statistical probability tables CALENDF program
International Nuclear Information System (INIS)
Ribon, P.
1989-01-01
The purpose of the probability tables is: - to obtain dense data representation - to calculate integrals by quadratures. They are mainly used in the USA for calculations by Monte Carlo and in the USSR and Europe for self-shielding calculations by the sub-group method. The moment probability tables, in addition to providing a more substantial mathematical basis and calculation methods, are adapted for condensation and mixture calculations, which are the crucial operations for reactor physics specialists. However, their extension is limited by the statistical hypothesis they imply. Efforts are being made to remove this obstacle, at the cost, it must be said, of greater complexity
Energy Technology Data Exchange (ETDEWEB)
Riley, Jr, J E; Lindstrom, R M
1987-01-01
Major levels of boron in borosilicate glasses were determined nondestructively by neutron activation analysis. The effects of neutron self-shielding by boron (1 to 8% by weight) are examined. Results of the analysis of a series of glasses with increasing boron composition are 1.150 +- .005% and 7.766 +- .035% for the low and high members of the series. Once analyzed, the glasses are useful as secondary standards for alpha track counting, and also ion and electron microprobe analyses of glasses. 12 refs.; 3 tables.
Estimation of "9"9Mo production rates from natural molybdenum in research reactors
International Nuclear Information System (INIS)
Blaauw, M.; Chakrova, Y.; Jacimovic, R.; Kling, A.
2017-01-01
Molybdenum-99 is one of the most important radionuclides for medical diagnostics. In 2015, the International Atomic Energy Agency organized a round-robin exercise where the participants measured and calculated specific saturation activities achievable for the "9"8Mo(n,γ)"9"9Mo reaction. This reaction is of interest as a means to locally, and on a small scale, produce "9"9Mo from natural molybdenum. The current paper summarises a set of experimental results and reviews the methodology for calculating the corresponding saturation activities. Activation by epithermal neutrons and also epithermal neutron self-shielding are found to be of high importance in this case. (author)
Uncertainty of Doppler reactivity worth due to uncertainties of JENDL-3.2 resonance parameters
Energy Technology Data Exchange (ETDEWEB)
Zukeran, Atsushi [Hitachi Ltd., Hitachi, Ibaraki (Japan). Power and Industrial System R and D Div.; Hanaki, Hiroshi; Nakagawa, Tuneo; Shibata, Keiichi; Ishikawa, Makoto
1998-03-01
Analytical formula of Resonance Self-shielding Factor (f-factor) is derived from the resonance integral (J-function) based on NR approximation and the analytical expression for Doppler reactivity worth ({rho}) is also obtained by using the result. Uncertainties of the f-factor and Doppler reactivity worth are evaluated on the basis of sensitivity coefficients to the resonance parameters. The uncertainty of the Doppler reactivity worth at 487{sup 0}K is about 4 % for the PNC Large Fast Breeder Reactor. (author)
New SCALE-4 features related to cross-section processing
International Nuclear Information System (INIS)
Petrie, L.M.; Landers, N.F.; Greene, N.M.; Parks, C.V.
1991-01-01
The SCALE code system has a standardized scheme for processing problem-dependent cross section from problem-independent waste libraries. Some improvements and new capabilities in the processing scheme have been incorporated into the new Version 4 release of the SCALE system. The new features include the capability to consider annular cylindrical and spherical unit cells, and improved Dancoff factor formulation, and changes to the NITAWL-II module to perform resonance self-shielding with reference to infinite dilute values. A review of these major changes in the cross-section processing scheme for SCALE-4 is presented in this paper
Theory of neutron resonance cross sections for safety applications
International Nuclear Information System (INIS)
Froehner, F.H.
1992-09-01
Neutron resonances exert a strong influence on the behaviour of nuclear reactors, especially on their response to the temperature changes accompanying power excursions, and also on the efficiency of shielding materials. The relevant theory of neutron resonance cross sections including the practically important approximations is reviewed, both for the resolved and the unresolved resonance region. Numerical techniques for Doppler broadening of resonances are presented, and the construction of group constants and especially of self-shielding factors for neutronics calculations is outlined. (orig.) [de
Mathematical aspects of the discrete space-time hypothesis
International Nuclear Information System (INIS)
Sardanashvili, G.A.
1979-01-01
A hypothesis of a microcosm space discreteness is considered from the theoretical-mathematical point of view. The type of topological spaces, which formalizes representations on the discrete space-time, is determined. It is explained, how these spaces arise in physical models. The physical task, in which the discrete space could arise as a version of its solution, is considered. It is shown that the discrete structure of space can arise with a certain interaction type in the system, for example, with its considerable self-shielding, which can take place, in particular, in the particles or in the cosmological and astrophysical singularities
Development of real-time low energy electron calorimeter
International Nuclear Information System (INIS)
Noriah Mod Ali; Smith, F.A.
1999-01-01
A low energy electron beam calorimeter with a thin film window has been fabricated to facilitate a reliable method of dose assessment for electron beam energies down to 200 keV. The system was designed to incorporate a data-logger in order that it could be used on the self-shielded 200 keV facility at MINT. In use, the calorimeter started logging temperature a short time before it passed under the beam and it continued taking data until well after the end of the irradiation. Data could be retrieved at any time after the calorimeter had emerged from the irradiator
Fast reactor core concepts to improve transmutation efficiency
International Nuclear Information System (INIS)
Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi
2015-01-01
Fast Reactor (FR) core concepts to improve transmutation efficiency were conducted. A heterogeneous MA loaded core was designed based on the 1000MWe-ABR breakeven core. The heterogeneous MA loaded core with Zr-H loaded moderated targets had a better transmutation performance than the MA homogeneous loaded core. The annular pellet rod design was proposed as one of the possible design options for the MA target. It was shown that using annular pellet MA rods mitigates the self-shielding effect in the moderated target so as to enhance the transmutation rate
Physical-chemical characterization of different hydrogels Formulations
International Nuclear Information System (INIS)
Rodriguez, Y.; Romero, M.; Soler, I.; Saldivar, D.
2001-01-01
They were carried out swelling studies at 37 0C of different hydrogels formulations whose composition was the following one: 3% PEG-1000, 1% Agar and concentrations of PVP and NVP 7, 10 and 14% and a maximum thickness of 2 mm, this formulations were irradiated to 25 kGy in a self-shield irradiator of 60Co and the same time studies of the mechanical properties. It was obtained that the hydrogels of PVP absorbs more water than those of NVP. It was described pseudo-Fickian s kinetic whose exponent diffusional is in the following range: 0.5 0.6
Energy Technology Data Exchange (ETDEWEB)
1981-01-01
A review is presented of proposals for transport, handling and emplacement of high-level radioactive waste in an underground repository appropriate to the U.K. context, with particular reference to waste block size and configuration; self-shielded or partially-shielded block; stages of disposal; transport by road/rail to repository site; handling techniques within repository; emplacement in vertical holes or horizontal tunnels; repository access by adit, incline or shaft; conventional and radiological safety; costs; and major areas of uncertainty requiring research or development.
International Nuclear Information System (INIS)
Newton, Tim; Hosking, Glynn; Hutton, Les; Powney, David; Turland, Brian; Shuttleworth, Ted
2008-01-01
The WIMS code provides a versatile software package for neutronic calculations, which can be applied to all thermal reactor types including mixed moderator systems. It can provide lattice cell and supercell calculations to produce the neutronic libraries for use in PANTHER or other whole core analysis codes; it can also be used for a complete analysis of small reactors. WIMS provides a range of flux solution methods, including diffusion, S-n, collision probability and the method of characteristics. It can also use the MONK Monte Carlo solver and includes a hybrid perturbation solution method, MAX. WIMS allows the user to assemble a calculational route using these solvers, and a set of cross-section self-shielding methods and utility routines for data transfers and manipulation (e.g. condensation). Key data are stored on WIMS interfaces that are updated by the various processes. Development of the WIMS code has continued to extend its capabilities and its range of application. The WIMS9 code was released in 2004, which included an improved treatment of resonance self-shielding, both using equivalence and sub-group methods. The latest version of the code, WIMS10, will be released in 2008. (authors)
Neutron cross sections in the unresolved resonance region
International Nuclear Information System (INIS)
Janeva, N.; Lukyanov, A.; Koyumdjieva, N.; Volev, K.
2005-01-01
In this work a development of the characteristic function model, created to reveal the resonance cross section structure in the unresolved resonance region is presented. The main advantage of this model is the calculation of resonance averaged self-shielding factors analytically. To determine average values of the cross sections and their functionals the function of joint statistical distribution of the R-matrix real and imaginary parts should be used. The characteristic function of such distribution is determined and the resonance ladder for the unresolved region is optimized to calculate the group averaged functionals in the same way as it is in the resolved resonance region. The main advantage of this model is the calculation of resonance averaged self-shielding factors analytically. The neutron width energy dependence leads to some deformation in the shape of resonances. This deformation is most apparent near the inelastic scattering threshold. For the case when the inelastic channel momentum is zero we present the formula for level shape bellow and over the inelastic threshold and the calculated resonance deformation in dependence of the position of the resonance in respect to the threshold. (authors)
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.
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.
Development of methods for burn-up calculations for LWR's
International Nuclear Information System (INIS)
Jaschik, W.
1978-01-01
This method is based on all burn-up depending data, namely particle densities and neutron spectra, being available in a burn-up library. This one is created by means of a small number of cell burn-up calculations which can easily be carried out and in which the heterogeneous cell structure and self-shielding effects can explicitly be accounted for. Then the cluster burn-up is simulated by adequate correlation of the burn-up data. The advantage of this method is given by - an exact determination of the real spectrum distribution in the individual fuel element clusters; - an exact determination of the burn-up related spectrum variations for each fuel rod and for each burn-up value obtained; - accounting for heterogeneity of the fuel rod cells and the self-shielding in the fuel; high accuracy of the results of a comparably low effort and - simple handling by largely automating the process of computation. Programed realization was achieved by establishing the RSYST modules ABRAJA, MITHOM, and SIMABB and their implementation within the code system. (orig./HP) [de
Investigating the physics and environment of Lyman limit systems in cosmological simulations
Erkal, Denis
2015-07-01
In this work, I investigate the properties of Lyman limit systems (LLSs) using state-of-the-art zoom-in cosmological galaxy formation simulations with on the fly radiative transfer, which includes both the cosmic UV background (UVB) and local stellar sources. I compare the simulation results to observations of the incidence frequency of LLSs and the H I column density distribution function over the redshift range z = 2-5 and find good agreement. I explore the connection between LLSs and their host haloes and find that LLSs reside in haloes with a wide range of halo masses with a nearly constant covering fraction within a virial radius. Over the range z = 2-5, I find that more than half of the LLSs reside in haloes with M test a simple model which encapsulates many of their properties. I confirm that LLSs have a characteristic absorption length given by the Jeans length and that they are in photoionization equilibrium at low column densities. Finally, I investigate the self-shielding of LLSs to the UVB and explore how the non-sphericity of LLSs affects the photoionization rate at a given N_{H I}. I find that at z ≈ 3, LLSs have an optical depth of unity at a column density of ˜1018 cm-2 and that this is the column density which characterizes the onset of self-shielding.
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
An overview of the current status of resonance theory in reactor physics applications
International Nuclear Information System (INIS)
Hwang, R.N.
1993-01-01
The neutron resonance phenomena constitute one of the most fundamental subjects in nuclear physics as well as in reactor physics. It is the area where the concepts of nuclear interaction and the treatment of the neutronic balance in reactor lattices become intertwined. The later requires the detailed knowledge of resonance structures of many nuclide of practical interest to the development of nuclear energy. The key issue of the resonance treatment in reactor applications is directly associated with the use of the microscopic cross sections in the macroscopic reactor cells with a wide range of composition, temperature,and geometric configurations. It gives rise to the so called self-shielding effect. The accurate estimations of such a effect is essential not only in the calculation of the criticality of a reactor but also from the point of view of safety considerations. The latter manifests through the Doppler effect particularly crucial to the fast reactor development. The task of accurate treatment of the self-shielding effect, however, is by no means simple. In fact, it is perhaps the most complicated problem in neutron physics which, strictly speaking, requires the dependence of many physical variables. Two important elements of particular interest are : (1) a concise description of the resonance cross sections as a function of energy and temperature; (2) accurate estimation of the corresponding neutron flux where appropriate. These topics will be discussed from both the historical as well as the state-of-art perspectives
Code development for eigenvalue total sensitivity analysis and total uncertainty analysis
International Nuclear Information System (INIS)
Wan, Chenghui; Cao, Liangzhi; Wu, Hongchun; Zu, Tiejun; Shen, Wei
2015-01-01
Highlights: • We develop a new code for total sensitivity and uncertainty analysis. • The implicit effects of cross sections can be considered. • The results of our code agree well with TSUNAMI-1D. • Detailed analysis for origins of implicit effects is performed. - Abstract: The uncertainties of multigroup cross sections notably impact eigenvalue of neutron-transport equation. We report on a total sensitivity analysis and total uncertainty analysis code named UNICORN that has been developed by applying the direct numerical perturbation method and statistical sampling method. In order to consider the contributions of various basic cross sections and the implicit effects which are indirect results of multigroup cross sections through resonance self-shielding calculation, an improved multigroup cross-section perturbation model is developed. The DRAGON 4.0 code, with application of WIMSD-4 format library, is used by UNICORN to carry out the resonance self-shielding and neutron-transport calculations. In addition, the bootstrap technique has been applied to the statistical sampling method in UNICORN to obtain much steadier and more reliable uncertainty results. The UNICORN code has been verified against TSUNAMI-1D by analyzing the case of TMI-1 pin-cell. The numerical results show that the total uncertainty of eigenvalue caused by cross sections can reach up to be about 0.72%. Therefore the contributions of the basic cross sections and their implicit effects are not negligible
International Nuclear Information System (INIS)
Hogenbirk, A.; Gruppelaar, H.; Nierop, D.
1994-07-01
The results of a careful check of the solutions as given in NJOY91.38 are presented. It appears that the conversion of DDX data to Kalbach parameters r and a as presented in NJOY91.38 is, in general, not entirely adequate. An improved subroutine was written, which yields a better description of the Kalbach-fit of the DDX data. Furthermore, the conversion of cross section data from centre-of-mass to laboratory system (which was needed for the processing of kerma data) appears to be not necessary anymore, as the HEATR-module of NJOY91.38 seems to operate correctly, also for centre-of-mass data. Problems still remain if DDX data for light isotopes are to be used in MCNP-calculations. This is illustrated using Be-9 from the EFF-2.2 evaluation as a sample case. Recommendations are given in order to solve the remaining problems. A computer code is presented, with which it is possible to create a problem-specific cross section library for use in MCNP-4, in which self-shielding in the unresolved resonance range is taken into account in an approximate form. It is shown, that the effects of neglecting self-shielding in the unresolved resonance range may be substantial in shielding calculations. This is especially relevant for the Fe-56 evaluation in EFF-2.2, in which a very extended unresolved resonance range is present. (orig./GL)
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
Improving shuffler assay accuracy
International Nuclear Information System (INIS)
Rinard, P.M.
1995-01-01
Drums of uranium waste should be disposed of in an economical and environmentally sound manner. The most accurate possible assays of the uranium masses in the drums are required for proper disposal. The accuracies of assays from a shuffler are affected by the type of matrix material in the drums. Non-hydrogenous matrices have little effect on neutron transport and accuracies are very good. If self-shielding is known to be a minor problem, good accuracies are also obtained with hydrogenous matrices when a polyethylene sleeve is placed around the drums. But for those cases where self-shielding may be a problem, matrices are hydrogenous, and uranium distributions are non-uniform throughout the drums, the accuracies are degraded. They can be greatly improved by determining the distributions of the uranium and then applying correction factors based on the distributions. This paper describes a technique for determining uranium distributions by using the neutron count rates in detector banks around the waste drum and solving a set of overdetermined linear equations. Other approaches were studied to determine the distributions and are described briefly. Implementation of this correction is anticipated on an existing shuffler next year
Energy Technology Data Exchange (ETDEWEB)
Costa, Ubirajara Pereira da [ITW Soldagem Brasil Miller-Hobart, Sao Paulo, SP (Brazil)
2003-07-01
Different welding process like SMAW, Semi-Automatic FCAW Gas-shielded and Self-shielded and Mechanized GMAW-MAG with Solid Wire are suggested to weld Transmission Pipelines. Presently, the largest extensions of Transmission Pipelines under construction, are in China like Lines West-East, Zong-Wu, Shan-Jing Fuxian and some others, totalizing about 8.000 km, and all using Semi-Automatic Self Shielded Flux Cored Arc Welding Process. Also, several papers and magazines that covers Transmission Pipelines Welding, not frequently mention Operational aspects of the process and some other variables like environment and site geography. This presentation intends to cover some of the Operational aspects of the Flux Cored Arc Welding and GMAW-Metal Cored in order to give sufficient information for Construction, Engineering, Projects e Contractors so they can evaluate these Process against the SMAW or even Mechanized Systems, considering the Operation Factor, Efficiency and Deposition Rate. We will not cover operational details of the GMAW Mechanized Systems but only suggest that be evaluated the possibility to replace the GMAW-Solid Wire by the GMAW-Metal Cored Wire. (author)
Resonance region neutronics of unit cells in fast and thermal reactors
International Nuclear Information System (INIS)
Salehi, A.A.; Driscoll, M.J.; Deutsch, O.L.
1977-05-01
A method has been developed for generating resonance-self-shielded cross sections based upon an improved equivalence theorem, which appears to allow extension of the self-shielding-factor (Bondarenko f-factor) method, now mainly applied to fast reactors, to thermal reactors as well. The method is based on the use of simple prescriptions for the ratio of coolant-to-fuel region-averaged fluxes, in the equations defining cell averaged cross sections. Linearization of the dependence of these functions on absorber optical thickness is found to be a necessary and sufficient condition for the existence of an equivalence theorem. Results are given for cylindrical, spherical and slab geometries. The functional form of the flux ratio relations is developed from theoretical considerations, but some of the parameters are adjusted to force-fit numerical results. Good agreement over the entire range of fuel and coolant optical thicknesses is demonstrated with numerical results calculated using the ANISN program in the S 8 P 1 option
Generation and validation of the WIMS-D5 library based on JENDL-3.2
International Nuclear Information System (INIS)
Gil, Choong-Sup; Kim, Jung-Do
2002-01-01
A WIMS-D5 library based on JENDL-3.2 has been prepared and is being tested against benchmark problems. Several sensitivity calculations for stability confirmation of the library were carried out, such as fission spectrum dependency and the self-shielding effects of the elastic scattering cross sections and the self-shielding effects 240 Pu and 242 Pu capture cross sections below 4.0 eV. The results of benchmark calculations with the libraries based on JENDL-3.2, ENDF/B-VI.5, JEF-2.2, and the 1986 WIMS-D library were intercompared. The multiplication factors for the thermal lattices are slightly underpredicted by all libraries (up to 1% with ENDF/B-VI.5). The k eff values with the library based on JENDL-3.2 are slightly higher than those of ENDF/B-VI.5 and JEF-2.2. The spectral indices for the lattices with JENDL-3.2 agree with the measured quantities within the uncertainties of the experiments. The calculated amounts of some isotopes such as 149 Sm, 237 Np, 238 Pu, 242 Cm and 243 Cm show large differences from the measured or reference values. (author)
International Nuclear Information System (INIS)
Kulisek, Jonathan A.; Anderson, Kevin K.; Casella, Andrew M.; Gesh, Christopher J.; Warren, Glen A.
2013-01-01
This study investigates the use of a Lead Slowing-Down Spectrometer (LSDS) for the direct and independent measurement of fissile isotopes in light-water nuclear reactor fuel assemblies. The current study applies MCNPX, a Monte Carlo radiation transport code, to simulate the measurement of the assay of the used nuclear fuel assemblies in the LSDS. An empirical model has been developed based on the calibration of the LSDS to responses generated from the simulated assay of six well-characterized fuel assemblies. The effects of self-shielding are taken into account by using empirical basis vectors calculated from the singular value decomposition (SVD) of a matrix containing the self-shielding functions from the assay of assemblies in the calibration set. The performance of the empirical algorithm was tested on version 1 of the Next-Generation Safeguards Initiative (NGSI) used fuel library consisting of 64 assemblies, as well as a set of 27 diversion assemblies, both of which were developed by Los Alamos National Laboratory. The potential for direct and independent assay of the sum of the masses of Pu-239 and Pu-241 to within 2%, on average, has been demonstrated
SWANS: A Prototypic SCALE Criticality Sequence for Automated Optimization Using the SWAN Methodology
Energy Technology Data Exchange (ETDEWEB)
Greenspan, E.
2001-01-11
SWANS is a new prototypic analysis sequence that provides an intelligent, semi-automatic search for the maximum k{sub eff} of a given amount of specified fissile material, or of the minimum critical mass. It combines the optimization strategy of the SWAN code with the composition-dependent resonance self-shielded cross sections of the SCALE package. For a given system composition arrived at during the iterative optimization process, the value of k{sub eff} is as accurate and reliable as obtained using the CSAS1X Sequence of SCALE-4.4. This report describes how SWAN is integrated within the SCALE system to form the new prototypic optimization sequence, describes the optimization procedure, provides a user guide for SWANS, and illustrates its application to five different types of problems. In addition, the report illustrates that resonance self-shielding might have a significant effect on the maximum k{sub eff} value a given fissile material mass can have.
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)
Sensitivity and uncertainty analysis for UO2 and MOX fueled PWR cells
International Nuclear Information System (INIS)
Foad, Basma; Takeda, Toshikazu
2015-01-01
Highlights: • A method for calculating sensitivity coefficients has been improved. • The IR approximation was used in order to get accurate results. • Sensitivities and uncertainties are calculated using the improved method. • The method is applied for UO 2 and MOX fueled PWR cells. • The verification was performed by comparing our results with MCNP6 and TSUNAMI-1D. - Abstract: This paper discusses the improvement of a method for calculating sensitivity coefficients of neutronics parameters relative to infinite dilution cross-sections because the conventional method neglects resonance self-shielding effect. In this study, the self-shielding effect is taken into account by using the intermediate resonance approximation in order to get accurate results in both high and low energy groups. The improved method is applied to calculate sensitivity coefficients and uncertainties of eigenvalue responses for UO 2 and MOX (ThO 2 –UO 2 and PuO 2 –UO 2 ) fueled pressurized water reactor cells. The verification of the improved method was performed by comparing the sensitivities with MCNP6 and TSUNAMI-1D. For uncertainty, calculation comparisons were done with TSUNAMI-1D, and we demonstrate that the differences are caused by the use of different covariance matrices
International Nuclear Information System (INIS)
Karthikeyan, R.; Tellier, R. L.; Hebert, A.
2006-01-01
The Coolant Void Reactivity (CVR) is an important safety parameter that needs to be estimated at the design stage of a nuclear reactor. It helps to have an a priori knowledge of the behavior of the system during a transient initiated by the loss of coolant. In the present paper, we have attempted to estimate the CVR for a CANDU New Generation (CANDU-NG) lattice, as proposed at an early stage of the Advanced CANDU Reactor (ACR) development. We have attempted to estimate the CVR with development version of the code DRAGON, using the method of characteristics. DRAGON has several advanced self-shielding models incorporated in it, each of them compatible with the method of characteristics. This study will bring to focus the performance of these self-shielding models, especially when there is voiding of such a tight lattice. We have also performed assembly calculations in 2 x 2 pattern for the CANDU-NG fuel, with special emphasis on checkerboard voiding. The results obtained have been validated against Monte Carlo codes MCNP5 and TRIPOLI-4.3. (authors)
Energy Technology Data Exchange (ETDEWEB)
Newton, Tim; Hosking, Glynn; Hutton, Les; Powney, David; Turland, Brian; Shuttleworth, Ted [Serco, Dorchester (United Kingdom)
2008-07-01
The WIMS code provides a versatile software package for neutronic calculations, which can be applied to all thermal reactor types including mixed moderator systems. It can provide lattice cell and supercell calculations to produce the neutronic libraries for use in PANTHER or other whole core analysis codes; it can also be used for a complete analysis of small reactors. WIMS provides a range of flux solution methods, including diffusion, S-n, collision probability and the method of characteristics. It can also use the MONK Monte Carlo solver and includes a hybrid perturbation solution method, MAX. WIMS allows the user to assemble a calculational route using these solvers, and a set of cross-section self-shielding methods and utility routines for data transfers and manipulation (e.g. condensation). Key data are stored on WIMS interfaces that are updated by the various processes. Development of the WIMS code has continued to extend its capabilities and its range of application. The WIMS9 code was released in 2004, which included an improved treatment of resonance self-shielding, both using equivalence and sub-group methods. The latest version of the code, WIMS10, will be released in 2008. (authors)
International Nuclear Information System (INIS)
Gil, Choong Sup; Kim, Jung Do
1998-01-01
The NJOY Nuclear Data Processing System is used to convert evaluated nuclear data in ENDF format into forms useful for applications. The NJOY code is a modular computer code designed to read evaluated data in ENDF format, transform the data in various ways, and output the results as libraries designed to be used in various applications. The modules are essentially independent programs, and they communicate with each other using input and output files, plus a very few common variables. The NJOY code can work with neutrons, photons, and charged particles. The evaluated nuclear data are reconstructed to pointwise cross sections from ENDF resonance parameters and interpolation schemes. The pointwise data can be Doppler broadened to the temperatures requested by users. The data in unresolved energy range can be computed to effective self-shielded pointwise cross sections. The cross sections and scattering matrices for free or bound scatterers in the thermal energy range are able to be produced. The self-shielded multigroup cross sections, group-to-group scattering matrices, photon-production matrices, and charged-particle cross sections from pointwise data can be generated for various application codes
Evaluation and validation of criticality codes for fuel dissolver calculations
International Nuclear Information System (INIS)
Santamarina, A.; Smith, H.J.; Whitesides, G.E.
1991-01-01
During the past ten years an OECD/NEA Criticality Working Group has examined the validity of criticality safety computational methods. International calculation tools which were shown to be valid in systems for which experimental data existed were demonstrated to be inadequate when extrapolated to fuel dissolver media. A theoretical study of the main physical parameters involved in fuel dissolution calculations was performed, i.e. range of moderation, variation of pellet size and the fuel double heterogeneity effect. The APOLLO/P IC method developed to treat this latter effect permits us to supply the actual reactivity variation with pellet dissolution and to propose international reference values. The disagreement among contributors' calculations was analyzed through a neutron balance breakdown, based on three-group microscopic reaction rates. The results pointed out that fast and resonance nuclear data in criticality codes are not sufficiently reliable. Moreover the neutron balance analysis emphasized the inadequacy of the standard self-shielding formalism to account for 238 U resonance mutual self-shielding in the pellet-fissile liquor interaction. The benchmark exercise has resolved a potentially dangerous inadequacy in dissolver calculations. (author)
Influence of resonance parameters' correlations on the resonance integral uncertainty; 55Mn case
International Nuclear Information System (INIS)
Zerovnik, Gasper; Trkov, Andrej; Capote, Roberto; Rochman, Dimitri
2011-01-01
For nuclides with a large number of resonances the covariance matrix of resonance parameters can become very large and expensive to process in terms of the computation time. By converting covariance matrix of resonance parameters into covariance matrices of background cross-section in a more or less coarse group structure a considerable amount of computer time and memory can be saved. The question is how important is the information that is discarded in the process. First, the uncertainty of the 55 Mn resonance integral was estimated in narrow resonance approximation for different levels of self-shielding using Bondarenko method by random sampling of resonance parameters according to their covariance matrices from two different 55 Mn evaluations: one from Nuclear Research and Consultancy Group NRG (with large uncertainties but no correlations between resonances), the other from Oak Ridge National Laboratory (with smaller uncertainties but full covariance matrix). We have found out that if all (or at least significant part of the) resonance parameters are correlated, the resonance integral uncertainty greatly depends on the level of self-shielding. Second, it was shown that the commonly used 640-group SAND-II representation cannot describe the increase of the resonance integral uncertainty. A much finer energy mesh for the background covariance matrix would have to be used to take the resonance structure into account explicitly, but then the objective of a more compact data representation is lost.
Cross-section covariance propagation for LWR fuel cells in one and two dimensions - 308
International Nuclear Information System (INIS)
Ball, M.; Novog, D.R.; Parisi, C.; D'Auria, F.
2010-01-01
Within the framework of the Uncertainty Analysis in Modeling (UAM) for Design, Operation and Safety Analysis of LWRs Benchmark sponsored by the OECD/NEA, a tool has been developed for the propagation of covariance uncertainty through resonance self-shielding and other neutron kinetics calculations using a direct, cross-section generation and substitution approach. The motivation behind the work described in this paper was to develop a portable uncertainty propagation tool that could be easily implemented with several neutron kinetics codes, without relying on detailed knowledge of the internal workings of those codes or access to adjoint solutions. Implemented initially with the SCALE code package, 'self-shielded' covariance matrices for common LWR fuel cells have been calculated, as well as contributions to K eff uncertainty by selected neutron cross-sections and processes in both one and two dimensions. The one dimensional results generated by the tool are compared against those obtained using the TSUNAMI-1D module of SCALE in order to verify the efficacy of the methodology. One-dimensional results show good agreement with TSUNAMI-1D, but there is also an indication that the loss of dimensionality corresponding to one-dimensional equivalent geometries of two-dimensional fuel cells may lead to significant changes in the calculated uncertainty on K eff arising from particular neutron-nuclide reactions. (authors)
International Nuclear Information System (INIS)
Hwang, R.N.; Toppel, B.J.; Henryson, H. II.
1980-10-01
Motivated by a need for an economical yet rigorous tool which can address the computation of the structural material Doppler effect, an extremely efficient improved RABANL capability has been developed utilizing the fact that the Doppler broadened line shape functions become essentially identical to the natural line shape functions or Lorentzian limits beyond about 100 Doppler widths from the resonance energy, or when the natural width exceeds about 200 Doppler widths. The computational efficiency has been further enhanced by preprocessing or screening a significant number of selected resonances during library preparation into composition and temperature independent smooth background cross sections. The resonances which are suitable for such pre-processing are those which are either very broad or those which are very weak. The former contribute very little to the Doppler effect and their self-shielding effect can readily be averaged into slowly varying background cross section data, while the latter contribute very little to either the Doppler or to self-shielding effects. To illustrate the accuracy and efficiency of the improved RABANL algorithms and resonance screening techniques, calculations have been performed for two systems, the first with a composition typical of the STF converter region and the second typical of an LMFBR core composition. Excellent agreement has been found for RABANL compared to the reference Monte Carlo solution obtained using the code VIM, and improved results have also been obtained for the narrow resonance approximation in the ultra-fine-group option of MC 2 -2
Neutronics characteristics of micro-heterogeneous ThO2-UO2 PWR cores
International Nuclear Information System (INIS)
Zhao, X.; Driscoll, M.J.; Kazimi, S.
2001-01-01
A new fuel concept, axially-micro-heterogeneous ThO 2 -UO 2 fuel, where ThO 2 fuel pellets and UO 2 fuel pellets are stacked in separate layers in the fuel rods, is being studied at MIT as an option to reduce plutonium production in LWR fuel. Very interesting neutronic behavior is observed: (1) A reactivity increase of 3% to 4% at EOL for a given 235 U inventory which results in a 20-30% increase in average core discharge burnup; (2) For certain configurations, a ''burnable poison'' effect is observed. Analysis shows that these effects are achieved due to a combination of changes in self-shielding, local fissile worth, and conversion ratio, among which self-shielding is the dominant effect at the end of a reactivity-limited burnup. Other variations of micro-heterogeneous UO 2 -ThO 2 fuel including duplex pellets, checkerboard pin distribution, and checkerboard-axial combinations have also been investigated, and their neutronic performance compared. It is concluded that the axial fuel micro-heterogeneity provides the largest gain in reactivity-limited burnup. (author)
International Nuclear Information System (INIS)
Garg, S.B.; Sinha, A.
1985-01-01
A 35 group cross-section library with P/sub 3/-anisotropic scattering matrices and resonance self-shielding factors has been generated from the basic ENDF/B-IV cross-section files for 57 elements. This library covers the neutron energy range from 0.005 ev to 15 MeV and is well suited for the neutronics and safety analysis of fission, fusion and hybrid systems. The library is contained in two well known files, namely, ISOTXS and BRKOXS. In order to test the efficacy of this library and to bring out the importance of resonance self-shielding, a few selected fast critical assemblies representing large dilute oxide and carbide fueled uranium and plutonium based systems have been analysed. These assemblies include ZPPR/sub 2/, ZPR-3-48, ZPR-3-53, ZPR-6-6A, ZPR-6-7, ZPR-9-31 and ZEBRA-2 and are amongst those recommended by the US Nuclear Data Evaluation Working Group for testing the accuracy of cross-sections. The evaluated multiplication constants of these assemblies compare favourably with those calculated by others
What FIREs Up Star Formation: the Emergence of the Kennicutt-Schmidt Law from Feedback
Orr, Matthew E.; Hayward, Christopher C.; Hopkins, Philip F.; Chan, T. K.; Faucher-Giguère, Claude-André; Feldmann, Robert; Kereš, Dušan; Murray, Norman; Quataert, Eliot
2018-05-01
We present an analysis of the global and spatially-resolved Kennicutt-Schmidt (KS) star formation relation in the FIRE (Feedback In Realistic Environments) suite of cosmological simulations, including halos with z = 0 masses ranging from 1010 - 1013 M⊙. We show that the KS relation emerges and is robustly maintained due to the effects of feedback on local scales regulating star-forming gas, independent of the particular small-scale star formation prescriptions employed. We demonstrate that the time-averaged KS relation is relatively independent of redshift and spatial averaging scale, and that the star formation rate surface density is weakly dependent on metallicity and inversely dependent on orbital dynamical time. At constant star formation rate surface density, the `Cold & Dense' gas surface density (gas with T 10 cm-3, used as a proxy for the molecular gas surface density) of the simulated galaxies is ˜0.5 dex less than observed at ˜kpc scales. This discrepancy may arise from underestimates of the local column density at the particle-scale for the purposes of shielding in the simulations. Finally, we show that on scales larger than individual giant molecular clouds, the primary condition that determines whether star formation occurs is whether a patch of the galactic disk is thermally Toomre-unstable (not whether it is self-shielding): once a patch can no longer be thermally stabilized against fragmentation, it collapses, becomes self-shielding, cools, and forms stars, regardless of epoch or environment.
An optimized ultra-fine energy group structure for neutron transport calculations
International Nuclear Information System (INIS)
Huria, Harish; Ouisloumen, Mohamed
2008-01-01
This paper describes an optimized energy group structure that was developed for neutron transport calculations in lattices using the Westinghouse lattice physics code PARAGON. The currently used 70-energy group structure results in significant discrepancies when the predictions are compared with those from the continuous energy Monte Carlo methods. The main source of the differences is the approximations employed in the resonance self-shielding methodology. This, in turn, leads to ambiguous adjustments in the resonance range cross-sections. The main goal of developing this group structure was to bypass the self-shielding methodology altogether thereby reducing the neutronic calculation errors. The proposed optimized energy mesh has 6064 points with 5877 points spanning the resonance range. The group boundaries in the resonance range were selected so that the micro group cross-sections matched reasonably well with those derived from reaction tallies of MCNP for a number of resonance absorbers of interest in reactor lattices. At the same time, however, the fast and thermal energy range boundaries were also adjusted to match the MCNP reaction rates in the relevant ranges. The resulting multi-group library was used to obtain eigenvalues for a wide variety of reactor lattice numerical benchmarks and also the Doppler reactivity defect benchmarks to establish its adequacy. (authors)
Energy Technology Data Exchange (ETDEWEB)
Hogenbirk, A; Gruppelaar, H; Nierop, D
1994-07-01
The results of a careful check of the solutions as given in NJOY91.38 are presented. It appears that the conversion of DDX data to Kalbach parameters r and a as presented in NJOY91.38 is, in general, not entirely adequate. An improved subroutine was written, which yields a better description of the Kalbach-fit of the DDX data. Furthermore, the conversion of cross section data from centre-of-mass to laboratory system (which was needed for the processing of kerma data) appears to be not necessary anymore, as the HEATR-module of NJOY91.38 seems to operate correctly, also for centre-of-mass data. Problems still remain if DDX data for light isotopes are to be used in MCNP-calculations. This is illustrated using Be-9 from the EFF-2.2 evaluation as a sample case. Recommendations are given in order to solve the remaining problems. A computer code is presented, with which it is possible to create a problem-specific cross section library for use in MCNP-4, in which self-shielding in the unresolved resonance range is taken into account in an approximate form. It is shown, that the effects of neglecting self-shielding in the unresolved resonance range may be substantial in shielding calculations. This is especially relevant for the Fe-56 evaluation in EFF-2.2, in which a very extended unresolved resonance range is present. (orig./GL).
International Nuclear Information System (INIS)
2001-01-01
Description of problem or function: MARS-ORNL is a selection of computer codes for the generation of problem-dependent multigroup cross section libraries. They are selected modules from the AMPX-2 system for AMPX interface format libraries, LASL codes for CCCC interfaces, and processing codes for libraries to be used by ANISN, DOT, or MORSE codes. The codes in the collection are used in connection with the following DLC data libraries: ZZ-LIB-IV (DLC-0040), ZZ-VITAMIN-C (DLC-0041), VITAMIN-4C (DLC-0053), ZZ-CLEAR/42B (DLC-0042), ZZ-CSRL/43B (DLC-0043), and EPRMASTER (DLC-0052). The functions of these processing codes are briefly described: A. AMPX Modules: AIM: Converts AMPX Master Interface Files from EBCDIC to binary form and back. AJAX: Merges, collects, assembles, re-orders, joins, and copies selected nuclides from AMPX Master Interfaces. BONAMI: Accesses Bondarenko factors from an AMPX Master Library and performs resonance self-shielding calculations. CHOX: Produces a coupled interface library in AMPX format by combining neutron libraries (generated by module XLACS), gamma libraries (generated by module SMUG), and photon production libraries (generated by module LAPHNGAS). CHOXM: Combines self-shielding factors as generated by the code SPHINX (PSR-0129) and an infinite dilution neutron master interface (generated by XLACS) to generate a self-shielded neutron AMPX Interface File. The interface produced by CHOXM is an input to the NITAWL module of AMPX. CHOXM is a modified version of CHOX. COMAND: Collapses ANISN cross section libraries. DIAL: Produces edits from AMPX Master Interfaces. ICE-II: Accepts cross sections from an AMPX working library and produces mixed cross sections in four formats: (1) AMPX working library format; (2) ANISN format; (3) group-independent ANISN format; (4) Monte Carlo processed cross section library format. NITAWL: Produces self-shielded and working cross section libraries in the formats required by the ANISN, DOT, or MORSE codes
MPI version of NJOY and its application to multigroup cross-section generation
Energy Technology Data Exchange (ETDEWEB)
Alpan, A.; Haghighat, A.
1999-07-01
Multigroup cross-section libraries are needed in performing neutronics calculations. These libraries are referred to as broad-group libraries. The number of energy groups and group structure are highly dependent on the application and/or user's objectives. For example, for shielding calculations, broad-group libraries such as SAILOR and BUGLE with 47-neutron and 20-gamma energy groups are used. The common procedure to obtain a broad-group library is a three-step process: (1) processing pointwise ENDF (PENDF) format cross sections; (2) generating fine-group cross sections; and (3) collapsing fine-group cross sections to broad-group. The NJOY code is used to prepare fine-group cross sections by processing pointwise ENDF data. The code has several modules, each one performing a specific task. For instance, the module RECONR performs linearization and reconstruction of the cross sections, and the module GROUPR generates multigroup self-shielded cross sections. After fine-group, i.e., groupwise ENDF (GENDF), cross sections are produced, cross sections are self-shielded, and a one-dimensional transport calculation is performed to obtain flux spectra at specific regions in the model. These fluxes are then used as weighting functions to collapse the fine-group cross sections to obtain a broad-group cross-section library. The third step described is commonly performed by the AMPX code system. SMILER converts NJOY GENDF filed to AMPX master libraries, AJAX collects the master libraries. BONAMI performs self-shielding calculations, NITAWL converts the AMPX master library to a working library, XSDRNPM performs one-dimensional transport calculations, and MALOCS collapses fine-group cross sections to broad-group. Finally, ALPO is used to generate ANISN format libraries. In this three-step procedure, generally NJOY requires the largest amount of CPU time. This time varies depending on the user's specified parameters for each module, such as reconstruction tolerances
International Nuclear Information System (INIS)
Goddard, Braden; Charlton, William S.; McDeavitt, Sean M.
2010-01-01
Research highlights: → HPGe detectors are suitable for UREX+3a real-time spectroscopy. → HPGe N-type detectors may be suitable for a reprocessing facility. → Gamma ray self-shielding does not occur for pipe diameters less than 2 in. - Abstract: Reprocessing nuclear fuel is becoming more viable in the United States due to the anticipated increase in construction of nuclear power plants, the growing stockpile of existing used nuclear fuel, and a public desire to reduce the amount of this fuel. A new reprocessing facility will likely have state of the art controls and monitoring methods to safeguard special nuclear materials, as well as to provide real-time monitoring for process control. The focus of this research was to create a proof of concept to enable the development of a detection strategy that uses well established gamma and neutron measurement methods to characterize samples from the Uranium Extraction Plus 3a (UREX+3a) reprocessing method using a variety of detector types and measurement times. A facility that implemented real-time gamma detection equipment could improve product quality control and provide additional benefits, such as waste volume reduction. In addition to the spectral analyses, it was determined by Monte Carlo N Particle (MCNP) simulations that there is no noticeable self-shielding for internal pipe diameters less than 5.08 cm, indicating that no self-shielding correction factors are needed. Further, it was determined that High Purity Germanium (HPGe) N-type detectors have the high gamma ray energy resolution and neutron damage resistance that would be required in a reprocessing facility. Finally, the gamma ray spectra for the measured samples were simulated using MCNP and then the model was extended to predict the responses from an actual reprocessing scenario from UREX+3a applied to fuel that had a decay time of 3 years. The 3-year decayed fuel was more representative of commercially reprocessed fuel than the acquired UREX+3a
International Nuclear Information System (INIS)
Nithyadevi, Rajan; Thilagam, L.; Karthikeyan, R.; Pal, Usha
2016-01-01
Highlights: • Use of advanced computational code – DRAGON-5 using advanced self shielding model USS. • Testing the capability of DRAGON-5 code for the analysis of light water reactor system. • Wide variety of fuels LEU, MOX and spent fuel have been analyzed. • Parameters such as k ∞ , one, few and multi-group macroscopic cross-sections and fluxes were calculated. • Suitability of deterministic methodology employed in DRAGON-5 code is demonstrated for LWR. - Abstract: Advances in reactor physics have led to the development of new computational technologies and upgraded cross-section libraries so as to produce an accurate approximation to the true solution for the problem. Thus it is necessary to revisit the benchmark problems with the advanced computational code system and upgraded cross-section libraries to see how far they are in agreement with the earlier reported values. Present study is one such analysis with the DRAGON code employing advanced self shielding models like USS and 172 energy group ‘JEFF3.1’ cross-section library in DRAGLIB format. Although DRAGON code has already demonstrated its capability for heavy water moderator systems, it is now tested for light water reactor (LWR) and fast reactor systems. As a part of validation of DRAGON for LWR, a VVER computational benchmark titled “Neutronics Benchmarks for the Utilization of Mixed-Oxide Fuel-Volume 3” submitted by the Russian Federation has been taken up. Presently, pincell and assembly calculations are carried out considering variation in fuel temperature (both fresh and spent), moderator temperatures and boron content in the moderator. Various parameters such as infinite neutron multiplication (k ∞ ) factor, one group integrated flux, few group homogenized cross-sections (absorption, nu-fission) and reaction rates (absorption, nu-fission) of individual isotopic nuclides are calculated for different reactor states. Comparisons of results are made with the reported Monte Carlo
Validation d'un nouveau calcul de reference en evolution pour les reacteurs thermiques
Canbakan, Axel
Resonance self-shielding calculations are an essential component of a deterministic lattice code calculation. Even if their aim is to correct the cross sections deviation, they introduce a non negligible error in evaluated parameters such as the flux. Until now, French studies for light water reactors are based on effective reaction rates obtained using an equivalence in dilution technique. With the increase of computing capacities, this method starts to show its limits in precision and can be replaced by a subgroup method. Originally used for fast neutron reactor calculations, the subgroup method has many advantages such as using an exact slowing down equation. The aim of this thesis is to suggest a validation as precise as possible without burnup, and then with an isotopic depletion study for the subgroup method. In the end, users interested in implementing a subgroup method in their scheme for Pressurized Water Reactors can rely on this thesis to justify their modelization choices. Moreover, other parameters are validated to suggest a new reference scheme for fast execution and precise results. These new techniques are implemented in the French lattice scheme SHEM-MOC, composed of a Method Of Characteristics flux calculation and a SHEM-like 281-energy group mesh. First, the libraries processed by the CEA are compared. Then, this thesis suggests the most suitable energetic discretization for a subgroup method. Finally, other techniques such as the representation of the anisotropy of the scattering sources and the spatial representation of the source in the MOC calculation are studied. A DRAGON5 scheme is also validated as it shows interesting elements: the DRAGON5 subgroup method is run with a 295-eenergy group mesh (compared to 361 groups for APOLLO2). There are two reasons to use this code. The first involves offering a new reference lattice scheme for Pressurized Water Reactors to DRAGON5 users. The second is to study parameters that are not available in
MPI version of NJOY and its application to multigroup cross-section generation
International Nuclear Information System (INIS)
Alpan, A.; Haghighat, A.
1999-01-01
Multigroup cross-section libraries are needed in performing neutronics calculations. These libraries are referred to as broad-group libraries. The number of energy groups and group structure are highly dependent on the application and/or user's objectives. For example, for shielding calculations, broad-group libraries such as SAILOR and BUGLE with 47-neutron and 20-gamma energy groups are used. The common procedure to obtain a broad-group library is a three-step process: (1) processing pointwise ENDF (PENDF) format cross sections; (2) generating fine-group cross sections; and (3) collapsing fine-group cross sections to broad-group. The NJOY code is used to prepare fine-group cross sections by processing pointwise ENDF data. The code has several modules, each one performing a specific task. For instance, the module RECONR performs linearization and reconstruction of the cross sections, and the module GROUPR generates multigroup self-shielded cross sections. After fine-group, i.e., groupwise ENDF (GENDF), cross sections are produced, cross sections are self-shielded, and a one-dimensional transport calculation is performed to obtain flux spectra at specific regions in the model. These fluxes are then used as weighting functions to collapse the fine-group cross sections to obtain a broad-group cross-section library. The third step described is commonly performed by the AMPX code system. SMILER converts NJOY GENDF filed to AMPX master libraries, AJAX collects the master libraries. BONAMI performs self-shielding calculations, NITAWL converts the AMPX master library to a working library, XSDRNPM performs one-dimensional transport calculations, and MALOCS collapses fine-group cross sections to broad-group. Finally, ALPO is used to generate ANISN format libraries. In this three-step procedure, generally NJOY requires the largest amount of CPU time. This time varies depending on the user's specified parameters for each module, such as reconstruction tolerances, temperatures
Effect of microalloying on precipitate evolution in ferritic welds and implications for toughness
International Nuclear Information System (INIS)
Narayanan, Badri K.; Kovarik, L.; Sarosi, Peter M.; Quintana, Marie A.; Mills, M.J.
2010-01-01
Ferritic weld metal deposited with a self-shielded arc-welding process has intentional additions of aluminum, magnesium, titanium and zirconium. This results in a complex precipitation process that has been characterized with a combination of electron microscopy techniques. This work indicates that the formation of a spinel oxide is critical for the nucleation of nitrides of zirconium and titanium and prevents the agglomeration of aluminum rich oxides and the formation of large aluminum nitrides. High-resolution transmission electron microscopy has been used to characterize the core/shell structure of the precipitates with microalloying additions. Thermodynamic modeling of the precipitate formation during solidification is consistent with the microstructural observations. The evolution of precipitate formation is critical to limit large inclusions and improve weld metal toughness.
Progress on China nuclear data processing code system
Liu, Ping; Wu, Xiaofei; Ge, Zhigang; Li, Songyang; Wu, Haicheng; Wen, Lili; Wang, Wenming; Zhang, Huanyu
2017-09-01
China is developing the nuclear data processing code Ruler, which can be used for producing multi-group cross sections and related quantities from evaluated nuclear data in the ENDF format [1]. The Ruler includes modules for reconstructing cross sections in all energy range, generating Doppler-broadened cross sections for given temperature, producing effective self-shielded cross sections in unresolved energy range, calculating scattering cross sections in thermal energy range, generating group cross sections and matrices, preparing WIMS-D format data files for the reactor physics code WIMS-D [2]. Programming language of the Ruler is Fortran-90. The Ruler is tested for 32-bit computers with Windows-XP and Linux operating systems. The verification of Ruler has been performed by comparison with calculation results obtained by the NJOY99 [3] processing code. The validation of Ruler has been performed by using WIMSD5B code.
Development of 350 keV electron accelerator
International Nuclear Information System (INIS)
Qin Jiuchang; Cui Shan; Zhou Wenzhen; Cui Zhipeng; Shi Zhenghu; Lu Zhongcheng; Chen Shangwen; Zhang Lifeng; Cui Zongwei; Huang Jun; Yin Meng
2007-01-01
The 350 keV electron accelerator is used for irradiation and production of plas- tic film of the medical infusion bags. The body structure of Van de Graft accelerator and the high voltage power supply of Cockcrof-Walton accelerator are adopted in the electron accelerator. The 350 keV DC power supply is supplied by the high frequency power supply with 14 kHz and 35 kW. The body and DC power supply of the electron accelerator are installed in the tank filled with 0.3 MPa SF 6 . The electron accelerator is compact, self-shielding and suitable for on-line electron beam processing. The main characteristics of the facility are terminal voltage 370 kV, electron beam power 7 kW (350 keV/20 mA), scaning width 70 cm, irradiation dose inuniformity ≤7%. (authors)
Propagation of dust electro-acoustic modes in dusty plasma
International Nuclear Information System (INIS)
Avinash, K.
2001-01-01
The propagation of the dust electro-acoustic (DEA) mode in dusty plasma with different electron and ion temperatures T e and T i and different ion species is studied. The critical ratio of the dust space charge to the ion space charge ε for the excitation of DEA mode is found to decrease with increasing T e /T i and increase with m i /m e (m i and m e are the ion and electron masses). Thus experiments with hydrogen plasma where electrons are sufficiently hotter than ions and where the reduction in the dust charge with ε is more than 50% are essential for the observation of self-shielding and the DEA mode
Multigroup neutron data base for nuclear geophysics
International Nuclear Information System (INIS)
Dworak, D.; Loskiewicz, J.
1989-01-01
The average group constants for the total, elastic, inelastic and capture cross sections as well as the average cosine of the scattering angle for elastic scattering and the average logarithmic energy decrement for elastic scattering have been obtained at two temperatures (300 and 400 deg K), using the ENDF/B-4 data and the IAEA-NDS pre-processing codes. The extended Abagyan group structure and the weighting spectrum of type 1/E were applied in course of the calculations. Self-shielding effect was not taken into account. All cross sections were Doppler broadened for both, 300 and 400 deg K temperatures. Under above assumptions, the average group constants were obtained for exactly 22 ENDF materials, which are of special importance for nuclear geophysics applications. 10 refs., 15 figs., 44 tabs. (author)
Energy Technology Data Exchange (ETDEWEB)
Weitman, J; Daaverhoeg, N; Farvolden, S
1970-07-01
In connection with fast neutron (n, {alpha}) cross section measurements a novel boron analysis method has been developed. The boron concentration is inferred from the mass spectrometrically determined number of helium atoms produced in the thermal and epithermal B-10 (n, {alpha}) reaction. The relation between helium amount and boron concentration is given, including corrections for self shielding effects and background levels. Direct and diffusion losses of helium are calculated and losses due to gettering, adsorption and HF-ionization in the release stage are discussed. A series of boron determinations is described and the results are compared with those obtained by other methods, showing excellent agreement. The lower limit of boron concentration which can be measured varies with type of sample. In e.g. steel, concentrations below 10-5 % boron in samples of 0.1-1 gram may be determined.
A program for calculating group constants on the basis of libraries of evaluated neutron data
International Nuclear Information System (INIS)
Sinitsa, V.V.
1987-01-01
The GRUKON program is designed for processing libraries of evaluated neutron data into group and fine-group (having some 300 groups) microscopic constants. In structure it is a package of applications programs with three basic components: a monitor, a command language and a library of functional modules. The first operative version of the package was restricted to obtaining mid-group non-block cross-sections from evaluated neutron data libraries in the ENDF/B format. This was then used to process other libraries. In the next two versions, cross-section table conversion modules and self-shielding factor calculation modules, respectively, were added to the functions already in the package. Currently, a fourth version of the GRUKON applications program package, for calculation of sub-group parameters, is under preparation. (author)
International Nuclear Information System (INIS)
Galli, M.
1989-01-01
This report is the user manual of the Teseo code which is a pre-processor code for MC 2 - II calculations MC 2 -II code, developed at ANL in the seventies, produces multigroup cross sections for fast reactor analysis, accounting in a fast and accurate way for resonance self shielding, Doppler broadening and overlap effects. It reads as input a set of files containing a problem-independent 2082 group cross section library, resonance parameters, energy and angular neutron distribution data etc. Tseo code produces this set of files by using ENDF-B or JEF basic nuclear data; it is written completely in Fortran-IV and has been implemented on IBM/370 compatible computers
An attempt to explain the uranium 238 resonance integral discrepancy
International Nuclear Information System (INIS)
Tellier, H.; Grandotto, M.
1978-01-01
Studies on uranium 238 resonance integral discrepancy were carried out for light water reactor physics. It was shown that using recently published resonance parameters and substituting a multilevel formalism to the usual Breit and Wigner formula reduced the well known discrepancy between two values of the uranium 238 effective resonance integral: the value calculated with the nuclear data and the one deduced from critical experiments. Since the cross section computed with these assumptions agrees quite well with the Oak-Ridge transmission data, it was used to obtain the self-shielding effect and the capture rate in light water lattices. The multiplication factor calculated with this method is found very close to the experimental value. Preliminary results for a set of benchmarks relative to several types of thermal neutron reactors lead to very low discrepancies. The reactivity loss is only 130 x 10 -5 instead of 650 x 10 -5 in the case of the usual libraries and the single level formula
International Nuclear Information System (INIS)
de Saussure, G.; Olsen, D.K.; Perez, R.B.
1976-01-01
The ENDF/B-IV prescription fails to represent correctly the 238 U total (and scattering) cross section between the levels of the resolved range. It is shown how this representation can be improved by properly accounting for the contribution of levels outside the resolved region to the cross section at energies inside the resolved region, and by substituting the more precise multilevel Breit-Wigner formula for the presently used single-level formula. The importance of computing accurately the minima in the total cross section is illustrated by comparing values of the self-shielded capture resonance integral computed with ENDF/B-IV and with a more accurate cross section model
Superconducting MRI system, MRT-50A
International Nuclear Information System (INIS)
Sugimoto, Hiroshi; Asahina, Kiyotaka
1987-01-01
The writers' developmental work on MRI (magnetic resonance imaging) was started in 1983. The model used first was MRT-15A (0.15 T). The next model was MRT-22A (0.22 T) which had a magnetic self-shield. As for the development of superconducting MRI system, they started clinical evaluation at Toshiba Central Hospital MRI Center in 1984 and got the official approval in 1985. For the model, use was made of MRT-50A (0.5 T) employing a superconducting magnet made by Toshiba. Herein represented are the basis of the images obtained through MRT-50 and the fact that the application fields of MRI are going on increasing (not only to brain and spine areas but also to the areas of the chest, abdomen and joints), and also the results of the work-in-progress of application software. (author)
International Nuclear Information System (INIS)
Takana, Hidemasa; Nishiyama, Hideya
2014-01-01
Detailed two-dimensional numerical simulations of a high energy loading nanosecond dc pulse DBD in a lean methane–air mixture were conducted for plasma-assisted combustion by integrating individual models of plasma chemistry, photoionization and energy loading. The DBD streamer propagation process with radical productions was clarified at 10 atm and 600 K as under the condition of actual internal engines at ignition. Energy is loaded to the streamer first by the formation of plasma channel and then ceased due to the self-shielding effect. Because of the inversed electric field in a discharge space during decrease in applied voltage, energy is loaded to the discharge again. It was found that higher energy is loaded to the DBD streamer for larger dielectric constant even at lower applied voltage, and higher number density of oxygen radical is produced at almost the same radical production efficiency. (paper)
Neutron shielding material based on colemanite and epoxy resin
International Nuclear Information System (INIS)
Okuno, K.
2005-01-01
In recent years, there has been a need for compact shielding design such as self-shielding of a PET cyclotron or up-gradation of radiation machinery in existing facilities. In these cases, high performance shielding materials are needed. Concrete or polyethylene have been used for a neutron shield. However, for compact shielding, they fall short in terms of performance or durability. Therefore, a new type of neutron shielding material based on epoxy resin and colemanite has been developed. Slab attenuation experiments up to 40 cm for the new shielding material were carried out using a 252 Cf neutron source. Measurement was carried out using a REM-counter, and compared with calculation. The results show that the shielding performance is better than concrete and polyethylene mixed with 10 wt% boron oxide. From the result, we confirmed that the performance of the new material is suitable for practical use. (authors)
Electron curing for high speed paper, film and foil converting
International Nuclear Information System (INIS)
Nablo, S.V.; Tripp, E.P.
1979-01-01
The status of self-shielded, compact electron processors for flexible web converting applications is reviewed. The uses of these units for a variety of laminating applications are described, with emphasis on the application techniques appropriate for low weight, (1 to 2 gm/m 2 ) 100% convertible adhesives. Performance data for electron cured adhesives with polyester/polyethylene systems is presented and compared with conventional urethane systems. The unique surface features of electron cured gravure coatings applied and cured at high speed are discussed, with reference to both paper and film substrates. An important advantage of electron curing of buried adhesive layers is the process quality control permitted by this 'all-electric' system. The performance characteristics of curing atmosphere control (inerting) for coatings are reviewed. Industrial experience with these processors has shown that effective inerting of coated flexible webs at speeds to 250 m/minute is both practical and economical. (author)
Directory of Open Access Journals (Sweden)
Lucrezia Auditore
2017-02-01
Full Text Available The production of 18F, 11C, 13N, and 15O positron emitting radionuclides for PET imaging is usually accomplished in Nuclear Medicine Departments through direct nuclear reactions induced by protons accelerated by compact medical cyclotrons on liquid or gaseous targets. Messina University has funded the construction of a PET-cyclotron-radio-chemistry plant at the Messina University Hospital, equipped with a 11 MeV self-shielded cyclotron. We estimated the expected production yields of these nuclides, accounting for target thickness, production of other radioactive nuclides, and time effects on the irradiated target purity. To this aim, both TALYS code (v. 1.8 and an analytical approach based on EXFOR experimental data were used. The general agreement between the two approaches, and with the available literature data, allows to assess the expected yields at the End of Bombardment, and relative target purities, to be used for further radiopharmaceutical preparation steps.
International Nuclear Information System (INIS)
Roussin, R.W.; Weisbin, C.R.; White, J.E.; Wright, R.Q.; Greene, N.M.; Ford, W.E. III; Wright, J.B.; Diggs, B.R.
1978-01-01
The Department of Energy (DOE) Division of Magnetic Fusion Energy (DMFE) and Reactor Research and Technology (DRRT) jointly sponsored the development of a coupled, fine-group cross-section library. The 171-neutron, 36-gamma-ray group library is intended to be applicable to fusion reactor neutronics and LMFBR core and shield analysis. Versions of the library are available from the Radiation Shielding Information Center (RSIC) at Oak Ridge National Laboratory in both AMPX and CCCC formats. Computer codes for energy group collapsing, interpolation on Bondarenko factors for resonance self-shielding and temperature corrections, and various other useful data manipulations are available. The experience gained in the utilization of this library is discussed. Indications are that this venture, which is designed to allow users to derive problem-dependent cross sections from a fine-group master library, has been a success
Li, Jianqi; Wang, Yi; Jiang, Yu; Xie, Haibin; Li, Gengying
2009-09-01
An open permanent magnet system with vertical B(0) field and without self-shielding can be quite susceptible to perturbations from external magnetic sources. B(0) variation in such a system located close to a subway station was measured to be greater than 0.7 microT by both MRI and a fluxgate magnetometer. This B(0) variation caused image artifacts. A navigator echo approach that monitored and compensated the view-to-view variation in magnetic resonance signal phase was developed to correct for image artifacts. Human brain imaging experiments using a multislice gradient-echo sequence demonstrated that the ghosting and blurring artifacts associated with B(0) variations were effectively removed using the navigator method.
International Nuclear Information System (INIS)
Leal, L.; Shibata, K.; Iwamoto, O.; Chiba, G.; Noguere, G.; Saint Jean, C. de; Sirakov, I.; Capote, R.; Sublet, J.C.; Coste-Delclaux, M.; Jouanne, C.; MacFarlane, R.E.; Kahler, A.C. Jr.; Lubitz, C.R.; Cullen, D.; Herman, M.; Ribon, P.
2011-01-01
This report summarises the work performed under WPEC Subgroup 32 (SG32) on issues pertinent to the methodology used in the unresolved resonance region (URR). The main purpose of SG32 was to verify the validity of the Single-level Breit-Wigner (SLBW) cross-section representation in the URR for self-shielding calculations. While SG32 work was under way, several other developments related to the URR on this subject came into play that had a direct impact on the results of calculations. The work described in this report focuses on: - testing of the SLBW formalism in the URR for fissile and fertile isotopes; - URR covariance representation; - interpolation issue with a URR resonance parameter for the infinitely dilute cross-section calculations; - ENDF URR parameter representation based on the LSSF = 0 or LSSF = 1 option
International Nuclear Information System (INIS)
Bode, P.; Koster-Ammerlaan, M.J.J.
2018-01-01
Pragmatic rather than physical correction factors for neutron and gamma-ray shielding were studied for samples of intermediate size, i.e. up to the 10-100 gram range. It was found that for most biological and geological materials, the neutron self-shielding is less than 5 % and the gamma-ray self-attenuation can easily be estimated. A trueness control material of 1 kg size was made based on use of left-overs of materials, used in laboratory intercomparisons. A design study for a large sample pool-side facility, handling plate-type volumes, had to be stopped because of a reduction in human resources, available for this CRP. The large sample NAA facilities were made available to guest scientists from Greece and Brazil. The laboratory for neutron activation analysis participated in the world’s first laboratory intercomparison utilizing large samples. (author)
High-field superferric MR magnet
International Nuclear Information System (INIS)
Huson, F.R.; Carcagno, R.; Colvin, J.
1987-01-01
Current large-bore (>20 cm), high-field (2-T) MR magnets have major implementation disadvantages, mostly related to the extensive stray field of traditional air-core superconducting magnets. To circumvent this problem, the authors designed, constructed, and tested a 30-cm prototype superconducting, self-shielded, high field magnet. This unshimmed superferric magnet can operate between 0.5 and 4 T with a field quality of about one part per million over one quarter of its aperture. The magnet can be ramped from one field strength to another in approximately 10 minutes. The 5-Gauss line extends less than 1 meter outside the magnet structure. Further details, including MR measurements and images, are demonstrated, as well as 1-meter bore scale-up projections
New ENDF/B-V nuclear data library for WIMS-D4M
International Nuclear Information System (INIS)
Deen, J.R.; Woodruff, W.L.; Costescu, C.I.
1994-01-01
A new 69-group 96-material library has been created for use in WIMS-D4M. The latest SUN version of NJOY (91.27) was used to generate the ENDF/B-V-based cross-section library. The library also includes ENDF/B-V based fission yields, energy fission and energy per capture data. The upper energy boundary has been extended from 10 to 20 MeV in order to model high energy neutron reactions. Additional fuel and moderator temperatures have been included to better predict temperature coefficients. More excess potential scattering points have been added to increase the accuracy of self-shielded resonance cross-sections. Several benchmark comparisons have been made to validate the new library. (author)
International Nuclear Information System (INIS)
Miller, L.F.; Williams, M.L.
1986-03-01
A neutronics library suitable for low-enrichment uranium (LEU) and high-enrichment uranium (HEU) fueled cores for both the Oak Ridge Research Reactor (ORR) and the Bulk Shielding Reactor (BSR) is documented herein. The library is obtained from version V of the Evaluated Nuclear Data File (ENDF/B-V) and contains 223 nuclides weighted over a variety of region-dependent neutron spectra. Self-shielding and zone-weighting effects are incorporated with 227-group calculations for several reactor-core configurations. Libraries are archived for both transport and diffusion theory seven-group calculations. Complete listings of processing details are included so that libraries with different specifications can be easily obtained. Results from validation calculations indicate that the neutronics libraries obtained from this effort are suitable for neutronics computations for the ORR and BSR. 12 refs., 5 figs., 15 tabs
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
Neutronics benchmarks of mixed-oxide fuels using the SCALE/CENTRM sequence
International Nuclear Information System (INIS)
Hollenbach, D.F.; Fox, P.B.
2000-01-01
The purpose of this study is to determine and document the reactor physics parameters (multiplication factors, spatially dependent flux ratios, and spacially dependent reaction rates ) for several distinct sets of problems using two distinct resonance cross-section processing techniques. In SCALE, by default, resonances are processed using NITAWL, which utilizes the Nordheim Integral Treatment. The results produced using this sequence are considered to be the base results. A second set of results are produced by replacing NITAWL with CENTRM/PMC. CENTRM produces point-wise fluxes for a given geometry configuration and set of isotopes. Using these fluxes, PMC produces problem-dependent self-shielding cross sections. Both sequences use ENDF/B-V cross-section data
A new expression for doppler broadening function based on Fourier Cosine Transform
Energy Technology Data Exchange (ETDEWEB)
Goncalves, Alessandro da C.; Martinez, Aquilino S.; Silva, Fernando C. da [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear]. E-mails: agoncalves@con.ufrj.br; aquilino@lmp.ufrj.br; fernando@con.ufrj.br
2007-07-01
The main objective of this paper consists of the derivation of an analytical solution for the Doppler broadening function {psi} ({xi}, x). The analytical solution is derived from a new integral expression for the {psi} ({xi}, x) function, which can be interpreted as a Fourier cosine transform. The expression obtained for {psi} ({xi}, x) in terms of elementary functions, proved quite simple and accurate, leading to a similar solution obtained through the differential equation for the {psi} ({xi}, x) function, using the methods of Frobenius and of parameter variation. The Doppler broadening function is widely used in applications related to the treatment of nuclear resonances, calculations of multigroup parameters and resonance self-shielding factors, and to correct microscopic cross section measurements through the activation technique. (author)
A new expression for doppler broadening function based on Fourier Cosine Transform
International Nuclear Information System (INIS)
Goncalves, Alessandro da C.; Martinez, Aquilino S.; Silva, Fernando C. da
2007-01-01
The main objective of this paper consists of the derivation of an analytical solution for the Doppler broadening function Ψ (ξ, x). The analytical solution is derived from a new integral expression for the Ψ (ξ, x) function, which can be interpreted as a Fourier cosine transform. The expression obtained for Ψ (ξ, x) in terms of elementary functions, proved quite simple and accurate, leading to a similar solution obtained through the differential equation for the Ψ (ξ, x) function, using the methods of Frobenius and of parameter variation. The Doppler broadening function is widely used in applications related to the treatment of nuclear resonances, calculations of multigroup parameters and resonance self-shielding factors, and to correct microscopic cross section measurements through the activation technique. (author)
The APOLLO assembly spectrum code
International Nuclear Information System (INIS)
Kavenoky, A.; Sanchez, R.
1987-04-01
The APOLLO code was originally developed as a design tool for HTR's, later it was aimed at the calculation of PWR lattices. APOLLO is a general purpose assembly spectrum code based on the multigroup integral transport equation; refined collision probability modules allow the computation of 1D geometries with linearly anisotropic scattering and two term flux expansion. In 2D geometries modules based on the substructure method provide fast and accurate design calculations and a module based on a direct discretization is devoted to reference calculations. The SPH homogenization technique provides corrected cross sections performing an equivalence between coarse and refined calculations. The post processing module of APOLLO generate either APOLLIB to be used by APOLLO or NEPLIB for reactor diffusion calculation. The cross section library of APOLLO contains data and self-shielding data for more than 400 isotopes. APOLLO is able to compute the depletion of any medium accounting for any heavy isotope or fission product chain. 21 refs
International Nuclear Information System (INIS)
de Saussure, G.; Perez, R.B.
1981-01-01
Several aspects of the measurement, analysis and evaluation of the cross sections of the fertile and fissile nuclides in the resonance regions are discussed. In the resolved range, for the fertile nuclides it is thought that the principal requirement for improved evaluations is for a practical methodology to deal with systematic errors and their correlations. For the fissile nuclides 235 U and 239 Pu, the ENDF/B-V evaluations are not consistent with ENDF/B procedures recommendations and fall short of the goals of resonance analysis. New evaluations of these two isotopes should be performed. In the unresolved resonance region it is shown that the ENDF/B representation is ambiguous and is not theoretically justified. A better representation may be desirable, and a validation of the representation with experimental self-shielding and transmission measurements is certainly required. 105 references
International Nuclear Information System (INIS)
Motoyama, Kazutaka; Morata, Oscar; Hasegawa, Tatsuhiko; Shang, Hsien; Krasnopolsky, Ruben
2015-01-01
A two-dimensional hydrochemical hybrid code, KM2, is constructed to deal with astrophysical problems that would require coupled hydrodynamical and chemical evolution. The code assumes axisymmetry in a cylindrical coordinate system and consists of two modules: a hydrodynamics module and a chemistry module. The hydrodynamics module solves hydrodynamics using a Godunov-type finite volume scheme and treats included chemical species as passively advected scalars. The chemistry module implicitly solves nonequilibrium chemistry and change of energy due to thermal processes with transfer of external ultraviolet radiation. Self-shielding effects on photodissociation of CO and H 2 are included. In this introductory paper, the adopted numerical method is presented, along with code verifications using the hydrodynamics module and a benchmark on the chemistry module with reactions specific to a photon-dominated region (PDR). Finally, as an example of the expected capability, the hydrochemical evolution of a PDR is presented based on the PDR benchmark
Effect of fuel particles' size variations on multiplication factor in pebble-bed nuclear reactor
International Nuclear Information System (INIS)
Snoj, L.; Ravnik, M.
2005-01-01
The pebble-bed reactor (Pbr) spherical fuel element consists of two radial zones: the inner zone, in which the fissile material in form of the so-called TRISO particles is uniformly dispersed in graphite matrix and the outer zone, a shell of pure graphite. A TRISO particle is composed of a fissile kernel (UO 2 ) and several layers of carbon composites. The effect of TRISO particles' size variations and distance between them on PBR multiplication factor is studied using MCNP code. Fuel element is modelled in approximation of a cubical unit cell with periodic boundary condition. The multiplication factor of the fuel element depends on the size of the TRISO particles due to resonance self-shielding effect and on the inter-particle distance due to inter-kernel shadowing. (author)
First results on study of gadolinium as burnable absorber
International Nuclear Information System (INIS)
Abbate, Maximo J.; Sbaffoni, Maria M.
2000-01-01
Following on with the work included in the 'Burnable absorbers research plan' several experiments were carried out oriented to determine Ga 2 O 3 burn up. Cold tests were performed and samples were irradiated in the RA-3 reactor. In this paper, some calculated values are presented together with their comparisons with experimental ones. The parameters foreseen for performing the experiments were verified and also the predictions on burn up of uranium and gadolinium isotopes concentrations. These results imply that the nuclear data of these isotopes included in the library are satisfactory. Next steps will be to measure other isotopes concentrations, gamma spectrum, and the irradiation of one pellet to determine self shielding effects in order to obtain effective cross sections i.e. for CAREM geometry. (author)
Criticality safety enhancements for SCALE 6.2 and beyond
International Nuclear Information System (INIS)
Rearden, Bradley T.; Bekar, Kursat B.; Celik, Cihangir; Clarno, Kevin T.; Dunn, Michael E.; Hart, Shane W.; Ibrahim, Ahmad M.; Johnson, Seth R.; Langley, Brandon R.; Lefebvre, Jordan P.; Lefebvre, Robert A.; Marshall, William J.; Mertyurek, Ugur; Mueller, Don; Peplow, Douglas E.; Perfetti, Christopher M.; Petrie Jr, Lester M.; Thompson, Adam B.; Wiarda, Dorothea; Wieselquist, William A.; Williams, Mark L.
2015-01-01
SCALE is a widely used suite of tools for nuclear systems modeling and simulation that provides comprehensive, verified and validated, user-friendly capabilities for criticality safety, reactor physics, radiation shielding, and sensitivity and uncertainty analysis. Since 1980, regulators, industry, and research institutions around the world have relied on SCALE for nuclear safety analysis and design. SCALE 6.2 provides several new capabilities and significant improvements in many existing features for criticality safety analysis. Enhancements are realized for nuclear data; multigroup resonance self-shielding; continuous-energy Monte Carlo analysis for sensitivity/uncertainty analysis, radiation shielding, and depletion; and graphical user interfaces. An overview of these capabilities is provided in this paper, and additional details are provided in several companion papers.
Energy-pointwise discrete ordinates transport methods
International Nuclear Information System (INIS)
Williams, M.L.; Asgari, M.; Tashakorri, R.
1997-01-01
A very brief description is given of a one-dimensional code, CENTRM, which computes a detailed, space-dependent flux spectrum in a pointwise-energy representation within the resolved resonance range. The code will become a component in the SCALE system to improve computation of self-shielded cross sections, thereby enhancing the accuracy of codes such as KENO. CENTRM uses discrete-ordinates transport theory with an arbitrary angular quadrature order and a Legendre expansion of scattering anisotropy for moderator materials and heavy nuclides. The CENTRM program provides capability to deterministically compute full energy range, space-dependent angular flux spectra, rigorously accounting for resonance fine-structure and scattering anisotropy effects
International Nuclear Information System (INIS)
Rios, Denise A. da S.; Sordi, Gian Maria A.A.
2014-01-01
This study aims to evaluate the radiological protection system in an installation where there is a self-shielded accelerator Category I, used for paint curing in polymer films, in order to optimize it in situations in which it is appropriate. For this a radiometric survey was conducted in the designated areas to check rate doses at different distances from the radiation focus, in which there is a possibility of presence or persons which can reliably detect any abnormality in the accelerator operation. The comparison of the obtained values related to the accelerator operating current with the natural background radiation showed that they are compatible, signaling that the principle of optimization is already being followed
Astronaut Exposures to Ionizing Radiation in a Lightly-Shielded Spacesuit
Wilson, J. W.; Simonsen, L. C.; Shinn, J. L.; Kim, M.-H. Y.; Cucinotta, F. A.; Badavi, F. F.; Atwell, W.
1999-01-01
The normal working and living areas of the astronauts are designed to provide an acceptable level of protection against the hazards of ionizing radiation of the space environment. Still there are occasions when they must don a spacesuit designed mainly for environmental control and mobility and leave the confines of their better-protected domain. This is especially true for deep space exploration. The impact of spacesuit construction on the exposure of critical astronaut organs will be examined in the ionizing radiation environments of free space, the lunar surface and the Martian surface. The computerized anatomical male model is used to evaluate astronaut self-shielding factors and to determine space radiation exposures to critical radiosensitive human organs.
ACTIV, Sandwich Detector Activity from In-Pile Slowing-Down Spectra Experiment
International Nuclear Information System (INIS)
Bozzi, L. and others
1978-01-01
1 - Nature of physical problem solved: Calculates the activities of a sandwich detector, to be used for in-pile measurements in slowing-down spectra below a few keV. The effect of scattering with energy degradation in the filter and in the detectors has been included to a first approximation. 2 - Method of solution: An iterative procedure is used: the calculation starts with a flux guess in which one assumes that each measured reactivity difference depends on the principal resonance only. The secondary resonance contribution is computed through the iterative process. For self-shielded cross-section calculations the model of Pearlstein and Weinstock (ref. 3) is used. The neutron spectrum can optionally be constant or 1/E inside each finite energy group relative to the resonance considered. 3 - Restrictions on the complexity of the problem: Maximum number of energy groups : 60
ETOA, ABBN Multigroup Constants from ENDF/B for Fast Reactors
International Nuclear Information System (INIS)
Nishimura, Hideo
1977-01-01
1 - Nature of physical problem solved: Production of ABBN type group constants up to 70 groups for fast reactor calculations, reading ENDF/B library as input. 2 - Method of solution: The multigroup method of Bondarenko et al. is used for processing basic nuclear data. Calculational algorithms for an unresolved resonance region are the same as those in the MC 2 code. For a resolved resonance region, an ultrafine energy structure dependent on a level scheme is adopted. 3 - Restrictions on the complexity of the problem: Maximum number of: energy groups: 70; sigma 0 values: 6; temperatures: 5. Self-shielding factors for an unrealistically low value of sigma 0 are not guaranteed because of the approximations used in the unresolved resonance region
International Nuclear Information System (INIS)
Pelloni, S.; Stepanek, J.; Vontobel, P.
1989-01-01
The capability of the advanced analysis of reactor engineering (AARE) modular code system and JEF-1-based nuclear data libraries to analyze light water high converter reactor (LWHCR) lattices is investigated by calculating the wet and dry cells of the PROTEUS-LWHCR phase II experiment. The results are compared to those obtained using several cell codes. Main features of the AARE code system, such as the self-shielding of resonance cross sections in the whole energy range, the generation of adequate fission source spectra, and the efficiency of the elastic removal correction,are investigated. In particular, it is shown that AARE results for the k ∞ void coefficient agree very well with the experiment, whereas other codes give larger deviations
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
CERN Library
2013-01-01
MicroShield® is a comprehensive photon/gamma ray shielding and dose assessment programme. It is widely used for designing shields, estimating source strength from radiation measurements, minimising exposure to people, and teaching shielding principles. Integrated tools allow the graphing of results, material and source file creation, source inference with decay (dose-to-Bq calculations accounting for decay and daughter buildup), the projection of exposure rate versus time as a result of decay, access to material and nuclide data, and decay heat calculations. The latest version is able to export results using Microsoft Office (formatted and colour-coded for readability). Sixteen geometries accommodate offset dose points and as many as ten standard shields plus source self-shielding and cylinder cladding are available. The library data (radionuclides, attenuation, build-up and dose conversion) reflect standard data from ICRP 38 and 107* as well as ANSI/ANS standards and RSICC publicat...
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
Utilization of plutonium in HTGR and its actinide production
International Nuclear Information System (INIS)
Karin, S.; Brogli, R.; Lefler, W.; Nordheim, L.
1976-01-01
The HTGR is a potential plutonium consumer. In this function it would burn plutonium, produce electricity and the valuable fissile isotope U-233. The advantages of this concept are discussed but particular attention is given to the production and the destruction of the higher actinides due to the high burnup achievable in such a system. The presence of the strong resonances in the plutonium isotopes demanded an extension of the methods for evaluation of self-shielding factors, a different structure for broad groups, and the adaptation of the reactor codes to these changes. Specifications for coated plutonium particles were developed. Also procedures were determined to evaluate the alpha ray and neutron emission rates of the actinide nuclides. First cycle calculations were carried out to establish in detail the characteristics of the plutonium reactors and their results are given
Effect of neutron anisotropic scattering in fast reactor analysis
International Nuclear Information System (INIS)
Chiba, Gou
2004-01-01
Numerical tests were performed about an effect of a neutron anisotropic scattering on criticality in the Sn transport calculation. The simplest approximation, the consistent P approximation and the extended transport approximation were compared with each other in one-dimensional slab fast reactor models. JAERI fast set which has been used for fast reactor analyses is inadequate to evaluate the effect because it doesn't include the scattering matrices and the self-shielding factors to calculate the group-averaged cross sections weighted by the higher-order moment of angular flux. In the present study, the sub-group method was used to evaluate the group-averaged cross sections. Results showed that the simplest approximation is inadequate and the transport approximation is effective for evaluating the anisotropic scattering. (author)
International Nuclear Information System (INIS)
Passard, Christian; Mariani, Alain; Jallu, Fanny; Romeyer-Dherbey, Jacques; Recroix, Herve; Rodriguez, Michel; Loridon, Joel; Denis, Caroline; Toubon, Herve
2002-01-01
The development of a passive-active neutron assay system for alpha low level waste characterization at the French Atomic Energy Commission is discussed. Less than 50 Bq[α] (about 50 μg Pu) per gram of crude waste must be measured in 118-l 'European' drums in order to reach the requirements for incinerating wastes. Detection limits of about 0.12 mg of effective 239 Pu in total active neutron counting, and 0.08 mg of effective 239 Pu coincident active neutron counting, may currently be detected (empty cavity, measurement time of 15 min, neutron generator emission of 1.6 x 10 8 s -1 [4π]). The most limiting parameters in terms of performances are the matrix of the drum - its composition (H, Cl...), its density, and its heterogeneity degree - and the localization and self-shielding properties of the contaminant
Resonance shielding in thermal reactor lattices
International Nuclear Information System (INIS)
Rothenstein, W.; Taviv, E.; Aminpour, M.
1982-01-01
The theoretical foundations of a new methodology for the accurate treatment of resonance absorption in thermal reactor lattice analysis are presented. This methodology is based on the solution of the point-energy transport equation in its integral or integro-differential form for a heterogeneous lattice using detailed resonance cross-section profiles. The methodology is applied to LWR benchmark analysis, with emphasis on temperature dependence of resonance absorption during fuel depletion, spatial and mutual self-shielding, integral parameter analysis and treatment of cluster geometry. The capabilities of the OZMA code, which implements the new methodology are discussed. These capabilities provide a means against which simpler and more rapid resonance absorption algorithms can be checked. (author)
Energy Technology Data Exchange (ETDEWEB)
Goncalves, Alessandro da Cruz
2010-07-01
An efficient and precise method for calculation of Doppler broadening function is very important to obtain average group microscopic cross sections, self shielding factors, resonance integrals and others reactor physics parameter. In this thesis two different methods for calculation of Doppler broadening function and interference term will be presented. The main method is based on a new integral form for Doppler broadening function {psi}(x,{zeta}) which gives a mathematical interpretation of the approximation proposed by Bethe and Placzek, as the convolution of the Lorentzian function with a Gaussian function. This interpretation besides leading to a new integral form for {psi}(x,{zeta}), enables to obtain a simple analytic solution for the Doppler broadening function. (author)
Verification of spectral burn-up codes on 2D fuel assemblies of the GFR demonstrator ALLEGRO reactor
International Nuclear Information System (INIS)
Čerba, Štefan; Vrban, Branislav; Lüley, Jakub; Dařílek, Petr; Zajac, Radoslav; Nečas, Vladimír; Haščik, Ján
2014-01-01
Highlights: • Verification of the MCNPX, HELIOS and SCALE codes. • MOX and ceramic fuel assembly. • Gas-cooled fast reactor. • Burnup calculation. - Abstract: The gas-cooled fast reactor, which is one of the six GEN IV reactor concepts, is characterized by high operational temperatures and a hard neutron spectrum. The utilization of commonly used spectral codes, developed mainly for LWR reactors operated in the thermal/epithermal neutron spectrum, may be connected with systematic deviations since the main development effort of these codes has been focused on the thermal part of the neutron spectrum. To be able to carry out proper calculations for fast systems the used codes have to account for neutron resonances including the self-shielding effect. The presented study aims at verifying the spectral HELIOS, MCNPX and SCALE codes on the basis of depletion calculations of 2D MOX and ceramic fuel assemblies of the ALLEGRO gas-cooled fast reactor demonstrator in infinite lattice
Aluminium-gold reference material for the k0-standardisation of neutron activation analysis
International Nuclear Information System (INIS)
Ingelbrecht, C.; Peetermans, F.; Corte, F. de; Wispelaere, A. de; Vandecasteele, C.; Courtijn, E.; Hondt, P. d'
1991-01-01
Gold is an excellent comparator material for the k 0 -standardisation of neutron activation analysis because of its convenient and well defined nuclear properties. The most suitable form for a reference material is a dilute aluminium-gold alloy, for which the self-shielding effect for neutrons is small. Castings of composition Al-0.1 wt.% Au were prepared by crucible-less levitation melting, which gives close control of ingot composition with minimal contamination of the melt. The alloy composition was checked using induction-coupled plasma source emission spectrometry. The homogeneity of the alloy was measured by neutron activation analysis and a relative standard deviation of the gold content of 0.30% was found (10 mg samples). Metallography revealed a homogeneous distribution of AuAl 2 particles. The alloy was certified as Reference Materials CBNM-530, with certified gold mass fraction 0.100±0.002 wt.%. (orig.)
Application of the self-powered detector concept in the design of a threshold gamma-ray detector
International Nuclear Information System (INIS)
LeVert, F.E.
1979-01-01
The self-powered detector concept has been utilized to develop an energy threshold gamma-ray detector. Gamma-ray energy discrimination is achieved by using a thick annular lead shield around the outer wall (emitter) of the detector in conjunction with a self-shielding central electrode (collector). Measurements conducted in the graphite pit of the Argonne Thermal Source Reactor have confirmed its ability to detect high-energy prompt fission gamma rays while discriminating against a significant flux of low-energy gamma rays from the decay of fission products. Also, auto-power spectral densities obtained with the detector were used to estimate the kinetic parameter, β/l, of the reactor
Neutron sensitivity of prompt-response self-powered neutron detectors and the interval rule
International Nuclear Information System (INIS)
Molina Avila, J.; Carmolopes, M.
1989-01-01
This paper is devoted to the calculation of thermal s th and epithermal s epi sensitivities of cobalt prompt-response Self-Powered Neutron Detectors (SPNDs). The thermal sensitivity was obtained for a Maxwellian neutron field, and the effect of scattering on the self-shielding correction was taken into consideration in the second-collision approximation. The dependence of s th on the emitter radius R was studied in a wide region of R (0.025 to 0.2 cm). The differential and global epithermal sensitivities were calculated using a simple expression for the first-collision neutron absorption probability. Finally, a criterion to evaluate the accuracy of the parameters of the model was established in the form of some Interval Rule which is very sensitive to the radial dependence of the flux perturbation correction and other parameters of the model in both the thermal and epithermal regions
The University of Pisa calculations for the Phase I of the OECD/NEA UAM Benchmark
International Nuclear Information System (INIS)
Ball, M.; Parisi, C.; D'Auria, F.
2009-01-01
In this paper we present the Univ. of Pisa preliminary results for the first exercise of the Phase I of the OECD/NEA Benchmark on the Uncertainty in Analysis and Modeling. The scope of exercise one is to address the uncertainties due to the basic nuclear data as well as the impact of processing the nuclear and covariance data, selection of multi-group structure and self-shielding treatment. DRAGON code and TSUNAMI code were employed, using the available covariance data matrix. The execution of DRAGON calculations required the use of ANGELO and LAMBDA codes for the extension of the covariance matrix from the original SCALE 44 group structure to DRAGON 69 group structure. The uncertainties for the main cross sections were evaluated and are presented here. (authors)
A validation of DRAGON based on lattice experiments
International Nuclear Information System (INIS)
Marleau, G.
1996-01-01
Here we address the validation of DRAGON using the Chalk River Laboratory experimental database which has already been used for the validation of other codes. Because of the large variety of information for different fuel and moderator types compiled on this database, the most basic modules of DRAGON are thoroughly tested. The general behaviour observed with DRAGON is very good. Its main weakness is seen in the self-shielding ,calculation where the correction applied to the inner fuel pin seems to be overevaluated with respect to the outer fuel pins. One question which is left open this paper concerns the need for inserting end-regions in the DRAGON cells when the heterogeneous B, leakage model is used. (author)
Radiation tests of the EMU spacesuit for the International SpaceStation using energetic protons
Energy Technology Data Exchange (ETDEWEB)
Zeitlin, C.; Heilbronn, L.; Miller, J.; Shavers, M.
2001-06-04
Measurements using silicon detectors to characterize theradiation transmitted through the EMU spacesuit and a human phantom havebeen performed using 155 and 250 MeV proton beams at the Loma LindaUniversity Medical Center (LLUMC). The beams simulate radiationencountered in space, where trapped protons having kinetic energies onthe order of 100 MeV are copious. Protons with 100 MeV kinetic energy andabove can penetrate many centimeters of water of other light materials,so that astronauts exposed to such energetic particles will receive dosesto their internal organs. This dose can be enhanced or reduced byshielding - either from the spacesuit or the self-shielding of the body -but minimization of the risk depends on details of the incident particleflux (in particular the energy spectrum) and on the dose responses of thevarious critical organs.
Preparation of the in-house neutron detectors and the software needed to process experimental data
International Nuclear Information System (INIS)
Haddad, Kh.; Haj-Hassan, H.; Helal, W.
2007-04-01
In - house neutron activation detectors were prepared in this work using pure commercial gold. The neutron self-shielding factors in the foils for both thermal and epithermal neutrons have been determined experimentally. The work shows good results repeatability and good agreement with certified activation monitors. the software KHW for neutron flux measurements using local and standards gold foils was designed and performed locally. it deals as well with irradiated uranium spectrums to calculate some important fission product ratios for neutron flux measurement. Some experiments were performed to investigate the possibility of using uranium, produced in the pilot plant, as fission neutron detector. The results shows the possibility of using fission product ratios to determine the cooling time of the samples. It shows also the possibility of using fission and activation product ratios as an indicators of neutron fluences ratios.(author)
Calculation of the Doppler broadening function using Fourier analysis
International Nuclear Information System (INIS)
Goncalves, Alessandro da Cruz
2010-01-01
An efficient and precise method for calculation of Doppler broadening function is very important to obtain average group microscopic cross sections, self shielding factors, resonance integrals and others reactor physics parameter. In this thesis two different methods for calculation of Doppler broadening function and interference term will be presented. The main method is based on a new integral form for Doppler broadening function ψ(x,ζ) which gives a mathematical interpretation of the approximation proposed by Bethe and Placzek, as the convolution of the Lorentzian function with a Gaussian function. This interpretation besides leading to a new integral form for ψ(x,ζ), enables to obtain a simple analytic solution for the Doppler broadening function. (author)
Study of a betavoltaic battery using electroplated nickel-63 on nickel foil as a power source
Energy Technology Data Exchange (ETDEWEB)
Uhm, Young Rang; Kim, Jong Bum; Son, Kwang Jae [Radioisotope Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Choi, Byoung Gun [Human Interface SoC Research Section, Electronics and Telecommunications Research Institute (ETRI), Daejeon (Korea, Republic of); Jeong, Dong Hyuk [Dongnam Institute of Radiological and Medical Science, Busan(Korea, Republic of)
2016-06-15
A betavoltaic battery was prepared using radioactive {sup 63}Ni attached to a three-dimensional single trenched P-N absorber. The optimum thickness of a 63Ni layer was determined to be approximately 2 μm, considering the minimum self-shielding effect of beta particles. Electroplating of radioactive {sup 63}Ni on a nickel (Ni) foil was carried out at a current density of 20 mA/cm{sup 2}. The difference of the short-circuit currents (I{sub sc}) between the pre- and postdeposition of {sup 63}Ni (16.65 MBq) on the P#-N junction was 5.03 nA, as obtained from the I-V characteristics. An improved design with a sandwich structure was provided for enhancing performance.
Energy Technology Data Exchange (ETDEWEB)
Ordonez, Robert E. Cooper; Silva, Jose Hilton F.; Trevisan, Roseana E. [Universidade Estadual de Campinas, SP (Brazil). Faculdade de Engenharia Mecanica. Dept. de Engenharia de Fabricacao
2003-07-01
Flux Core Arc Welding processes (FCAW) are beginning to be applied in pipeline welds, however, very limited experimental data regarding mechanical properties of pipeline weld joints with these processes are available in the literature. In this paper, the effects of preheat temperature and type of FCAW on mechanical properties (microhardness and tensile strength) of API X80 weld joint steel are presented. FCAW processes with gas protection and self-shielded were used. Multipasses welding were applied in 30'' diameter and 0,625'' thickness tubes. Influence factors were: FCAW type and preheat temperature. Acceptance criteria of welded joints were evaluated by API 1104 standard for tensile strength test and ASTM E384-99 for microhardness test. The results obtained showed that FCAW type and preheat temperature have no influence on mechanical properties of API X80 joint steel. (author)
Recent activities on clearance in IAEA and clearance automatic laser inspection system (CLALIS)
International Nuclear Information System (INIS)
Hattori, Takatoshi; Sasaki, Michiya
2005-01-01
Exemption levels for bulk amounts of materials have been described in RS-G-1.7 published as a safety guide in IAEA on August 2004. In Japan, the Nuclear Safety Commission adopted the RS-G-1.7 values as Japanese clearance levels after the careful review of dose assessment results. After completing revises of regulatory laws in relation to clearance level, solid wastes from decommissioning and operating nuclear power plants will be targets of clearance inspection. In CRIEPI, Clearance Automatic Laser Inspection System (CLALIS) has been developed, which can give high reliability and objectivity to the measurement data. The CLALIS is a new monitoring system coupling with 3D laser shape measurement, Monte-Carlo calculation and gamma measurement techniques, which can keep a high accuracy in the measurement data using an automatic correction technique for self-shielding effects of metal waste itself and is expected to apply as a practical use in actual clearance inspections. (author)
International Nuclear Information System (INIS)
Takeda, Toshikazu
2001-01-01
Recently, fuel assemblies of light water reactors have become complicated because of the extension of fuel burnup and the use of high-enriched Gd and mixed-oxide (MOX) fuel, etc. In conventional assembly calculations, the detailed flux distribution, spectrum distribution, and space dependence of self-shielding within a fuel pellet are not directly taken into account. The experimental and theoretical study of investigating these microscopic properties is named micro-reactor physics. The purpose of this work is to show the importance of micro-reactor physics in the analysis of MOX fuel assemblies. Several authors have done related studies; however, their studies are limited to fuel pin cells, and they are never mentioned with regard to burnup effect, which is important for actual core design. We used the subgroup method to treat the space dependence of the self-shielding effect of heavy nuclides, and we used the characteristics method to treat the angular dependence of neutron flux in a fuel pellet. Figure 1 compares the power distributions in MOX and UO 2 fuel cells at the beginning of burnup. The power is calculated with and without considering the space dependence of the self-shielding effect of the cross sections. For the MOX cell, the power distribution has a peak at the cell edge because of large Pu absorption especially when considering the spatial self-shielding effect. When a MOX rod is adjacent to UO 2 fuel rods, the flux distribution has an azimuthal dependence in addition to the radial dependence within a rod. For example, consider a 2x2 fuel assembly composed of three UO 2 rods and one MOX rod, with the mirror reflection boundary condition. A burnup calculation was done with the condition; the radius of the MOX pellet is divided into two regions, and the azimuthal angle is divided into eight. The number density of 239 Pu at 44 000 MWd/t for the MOX rod shows azimuthal dependence by 20%. The maximum burnup occurs in the direction of the UO 2 rods. This is
Micro-Reactor Physics of MOX-Fueled Core
International Nuclear Information System (INIS)
Takeda, T.
2001-01-01
Recently, fuel assemblies of light water reactors have become complicated because of the extension of fuel burnup and the use of high-enriched Gd and mixed-oxide (MOX) fuel, etc. In conventional assembly calculations, the detailed flux distribution, spectrum distribution, and space dependence of self-shielding within a fuel pellet are not directly taken into account. The experimental and theoretical study of investigating these microscopic properties is named micro-reactor physics. The purpose of this work is to show the importance of micro-reactor physics in the analysis of MOX fuel assemblies. Several authors have done related studies; however, their studies are limited to fuel pin cells, and they are never mentioned with regard to burnup effect, which is important for actual core design
MCB. A continuous energy Monte Carlo burnup simulation code
International Nuclear Information System (INIS)
Cetnar, J.; Wallenius, J.; Gudowski, W.
1999-01-01
A code for integrated simulation of neutrinos and burnup based upon continuous energy Monte Carlo techniques and transmutation trajectory analysis has been developed. Being especially well suited for studies of nuclear waste transmutation systems, the code is an extension of the well validated MCNP transport program of Los Alamos National Laboratory. Among the advantages of the code (named MCB) is a fully integrated data treatment combined with a time-stepping routine that automatically corrects for burnup dependent changes in reaction rates, neutron multiplication, material composition and self-shielding. Fission product yields are treated as continuous functions of incident neutron energy, using a non-equilibrium thermodynamical model of the fission process. In the present paper a brief description of the code and applied methods are given. (author)
CHAINS-PC, Decay Chain Atomic Densities
International Nuclear Information System (INIS)
1994-01-01
1 - Description of program or function: CHAINS computes the atom density of members of a single radioactive decay chain. The linearity of the Bateman equations allows tracing of interconnecting chains by manually accumulating results from separate calculations of single chains. Re-entrant loops can be treated as extensions of a single chain. Losses from the chain are also tallied. 2 - Method of solution: The Bateman equations are solved analytically using double-precision arithmetic. Poles are avoided by small alterations of the loss terms. Multigroup fluxes, cross sections, and self-shielding factors entered as input are used to compute the effective specific reaction rates. The atom densities are computed at any specified times. 3 - Restrictions on the complexity of the problem: Maxima of 100 energy groups, 100 time values, 50 members in a chain
Production of [11C]CO2 with gas target at low proton energies
International Nuclear Information System (INIS)
Sansaloni, Francesc; Lagares, Juan Ignacio; Llop, Jordi; Arce, Pedro; Díaz, Carlos; Pérez-Morales, José Manuel
2013-01-01
Nowadays the demand and the installation of self-shielded low-energy cyclotrons is growing, allowing the use of 11 C in many more centers. The aim of this study was the design of a new target and the evaluation of the production of 11 C as [ 11 C]CO 2 at low proton energies. The target was coupled to an IBA Cyclone-18/9 and the energy was decreased to 4–16 MeV. The newly designed target allowed the production of [ 11 C]CO 2 at different proton energies, and the results suggest that the cyclotron energy of Cyclone-18/9 is slightly higher than the nominal 18 MeV
Improved model for solar cosmic ray exposure in manned Earth orbital flights
International Nuclear Information System (INIS)
Wilson, J.W.; Nealy, J.E.; Atwell, W.; Cucinotta, F.A.; Shinn, J.L.; Townsend, L.W.
1990-06-01
A calculational model is derived for use in estimating Solar cosmic ray exposure to critical body organs in low-Earth orbit at the center of a large spherical shield of fixed thickness. The effects of the Earth's geomagnetic field and the astronauts' self-shielding are evaluated explicitly. The geomagnetic field model is an approximate tilted eccentric dipole with geomagnetic storms represented as a uniform-impressed field. The storm field is related to the planetary geomagnetic index K(sub p). The code is applied to the Shuttle geometry using the Shuttle mass distribution surrounding two locations on the flight deck. The Shuttle is treated as pure aluminum and the astronaut as soft tissue. Short-term, average fluence over a single orbit is calculated as a function of the location of the lines of nodes or long-term averages over all lines of nodes for a fixed inclination
Evaluation of the room shielding thickness of Hi-Art tomotherapy system
International Nuclear Information System (INIS)
Liu Haikuan; Wu Jinhai; Gu Naigu; Gao Yiming; Wang Li; Huang Weiqin; Wang Fengxian
2010-01-01
In this paper, we calculate and evaluate the room shielding thickness of a Hi-Art tomotherapy system, which is a new type of radiotherapy facility. Due to the self-shielding of the accelerator,only scattered beam and beam leakage were considered in calculating the room shielding thickness. The radiation field of the tomotherapy system was used as the basic data to calculate the shielding thickness of every 15 degree solid angle. The maximum shielding thickness required of each shielding wall was at the position with the angle of 15 degree, and the calculated shielding thickness were 1023, 975, 917, 1460, 1147 and 1189 mm for the east wall,south wall,west wall, north wall, the roof and the floor,respectively. According to the calculation results, all shielding walls, ceiling and floor could meet the requirement of the radiation protection, but the north wall thickness of 1200 mm was a little thinner. (authors)
Numerical Simulations of Finite-Length Effects in Diocotron Modes
Mason, Grant W.; Spencer, Ross L.
2000-10-01
Over a decade ago Driscoll and Fine(C. F. Driscoll and K. S. Fine, Phys. Fluids B 2) (6), 1359, June 1990. reported experimental observations of an exponential instability in the self-shielded m=1 diocotron mode for an electron plasma confined in a Malmberg-Penning trap. More recently, Finn et al(John M. Finn, Diego del-Castillo-Negrete and Daniel C. Barnes, Phys. Plasmas 6) (10), 3744, October 1999. have given a theoretical explanation of the instability as a finite-length end effect patterned after an analogy to theory for shallow water fluid vortices. However, in a test case selected for comparison, the growth rate in the experiment exceeds the theoretical value by a factor of two. We present results from a two-dimensional, finite length drift-kinetic code and a fully three-dimensional particle-in-cell code written to explore details of finite-length effects in diocotron modes.
LMFBR Blanket Physics Project progress report No. 2
International Nuclear Information System (INIS)
Forbes, I.A.; Driscoll, M.J.; Rasmussen, N.C.; Lanning, D.D.; Kaplan, I.
1971-01-01
This is the second annual report of an experimental program for the investigation of the neutronics of benchmark mock-ups of LMFBR blankets. Work was devoted primarily to measurements on Blanket Mock-Up No. 2, a simulation of a typical large LMFBR radial blanket and its steel reflector. Activation traverses and neutron spectra were measured in the blanket; calculations of activities and spectra were made for comparison with the measured data. The heterogeneous self-shielding effect for 238 U capture was found to be the most important factor affecting the comparison. Optimization and economic studies were made which indicate that the use of a high-albedo reflector material such as BeO or graphite may improve blanket neutronics and economics
Energy Technology Data Exchange (ETDEWEB)
Ivanova, T.; Laville, C. [Institut de Radioprotection et de Surete Nucleaire IRSN, BP 17, 92262 Fontenay aux Roses (France); Dyrda, J. [Atomic Weapons Establishment AWE, Aldermaston, Reading, RG7 4PR (United Kingdom); Mennerdahl, D. [E Mennerdahl Systems EMS, Starvaegen 12, 18357 Taeby (Sweden); Golovko, Y.; Raskach, K.; Tsiboulia, A. [Inst. for Physics and Power Engineering IPPE, 1, Bondarenko sq., 249033 Obninsk (Russian Federation); Lee, G. S.; Woo, S. W. [Korea Inst. of Nuclear Safety KINS, 62 Gwahak-ro, Yuseong-gu, Daejeon 305-338 (Korea, Republic of); Bidaud, A.; Sabouri, P. [Laboratoire de Physique Subatomique et de Cosmologie LPSC, CNRS-IN2P3/UJF/INPG, Grenoble (France); Patel, A. [U.S. Nuclear Regulatory Commission (NRC), Washington, DC 20555-0001 (United States); Bledsoe, K.; Rearden, B. [Oak Ridge National Laboratory ORNL, M.S. 6170, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Gulliford, J.; Michel-Sendis, F. [OECD/NEA, 12, Bd des Iles, 92130 Issy-les-Moulineaux (France)
2012-07-01
The sensitivities of the k{sub eff} eigenvalue to neutron cross sections have become commonly used in similarity studies and as part of the validation algorithm for criticality safety assessments. To test calculations of the sensitivity coefficients, a benchmark study (Phase III) has been established by the OECD-NEA/WPNCS/EG UACSA (Expert Group on Uncertainty Analysis for Criticality Safety Assessment). This paper presents some sensitivity results generated by the benchmark participants using various computational tools based upon different computational methods: SCALE/TSUNAMI-3D and -1D, MONK, APOLLO2-MORET 5, DRAGON-SUSD3D and MMKKENO. The study demonstrates the performance of the tools. It also illustrates how model simplifications impact the sensitivity results and demonstrates the importance of 'implicit' (self-shielding) sensitivities. This work has been a useful step towards verification of the existing and developed sensitivity analysis methods. (authors)
Energy Technology Data Exchange (ETDEWEB)
Rios, Denise A. da S.; Sordi, Gian Maria A.A., E-mail: denise@inovafi.com.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)
2014-07-01
This study aims to evaluate the radiological protection system in an installation where there is a self-shielded accelerator Category I, used for paint curing in polymer films, in order to optimize it in situations in which it is appropriate. For this a radiometric survey was conducted in the designated areas to check rate doses at different distances from the radiation focus, in which there is a possibility of presence or persons which can reliably detect any abnormality in the accelerator operation. The comparison of the obtained values related to the accelerator operating current with the natural background radiation showed that they are compatible, signaling that the principle of optimization is already being followed.
Electric field measurements in a nanosecond pulse discharge by picosecond CARS/4-wave mixing
Goldberg, Ben; Shkurenkov, Ivan; Adamovich, Igor; Lempert, Walter
2014-10-01
Time-resolved electric field measurements in hydrogen by picosecond CARS/4-wave mixing are presented. Measurements are carried out in a high voltage nanosecond pulse discharge in hydrogen in plane-to-plane geometry, at pressures of up to several hundred Torr, and with a time resolution of 0.2 ns. Absolute calibration of the diagnostics is done using a sub-breakdown high voltage pulse of 12 kV/cm. A diffuse discharge is obtained by applying a peak high voltage pulse of 40 kV/cm between the electrodes. It is found that breakdown occurs at a lower field, 15--20 kV/cm, after which the field in the plasma is reduced rapidly due to plasma self shielding The experimental results are compared with kinetic modeling calculations, showing good agreement between the measured and the predicted electric field.
International Nuclear Information System (INIS)
Bedregal, P.S.; Mendoza, A.; Montoya, E.H.; Cohen, I.M.; Universidad Tecnologica Nacional, Buenos Aires; Oscar Baltuano
2012-01-01
A new approach for analysis of entire potsherds of archaeological interest by INAA, using the conventional relative method, is described. The analytical method proposed involves, primarily, the preparation of replicates of the original archaeological pottery, with well known chemical composition (standard), destined to be irradiated simultaneously, in a well thermalized external neutron beam of the RP-10 reactor, with the original object (sample). The basic advantage of this proposal is to avoid the need of performing complicated effect corrections when dealing with large samples, due to neutron self shielding, neutron self-thermalization and gamma ray attenuation. In addition, and in contrast with the other methods, the main advantages are the possibility of evaluating the uncertainty of the results and, fundamentally, validating the overall methodology. (author)
Recent validation experience with multigroup cross-section libraries and scale
International Nuclear Information System (INIS)
Bowman, S.M.; Wright, R.Q.; DeHart, M.D.; Parks, C.V.; Petrie, L.M.
1995-01-01
This paper will discuss the results obtained and lessons learned from an extensive validation of new ENDF/B-V and ENDF/B-VI multigroup cross-section libraries using analyses of critical experiments. The KENO V. a Monte Carlo code in version 4.3 of the SCALE computer code system was used to perform the critical benchmark calculations via the automated SCALE sequence CSAS25. The cross-section data were processed by the SCALE automated problem-dependent resonance-processing procedure included in this sequence. Prior to calling KENO V.a, CSAS25 accesses BONAMI to perform resonance self-shielding for nuclides with Bondarenko factors and NITAWL-II to process nuclides with resonance parameter data via the Nordheim Integral Treatment
Study of a betavoltaic battery using electroplated nickel-63 on nickel foil as a power source
International Nuclear Information System (INIS)
Uhm, Young Rang; Kim, Jong Bum; Son, Kwang Jae; Choi, Byoung Gun; Jeong, Dong Hyuk
2016-01-01
A betavoltaic battery was prepared using radioactive 63 Ni attached to a three-dimensional single trenched P-N absorber. The optimum thickness of a 63Ni layer was determined to be approximately 2 μm, considering the minimum self-shielding effect of beta particles. Electroplating of radioactive 63 Ni on a nickel (Ni) foil was carried out at a current density of 20 mA/cm 2 . The difference of the short-circuit currents (I sc ) between the pre- and postdeposition of 63 Ni (16.65 MBq) on the P#-N junction was 5.03 nA, as obtained from the I-V characteristics. An improved design with a sandwich structure was provided for enhancing performance
Study of a Betavoltaic Battery Using Electroplated Nickel-63 on Nickel Foil as a Power Source
Directory of Open Access Journals (Sweden)
Young Rang Uhm
2016-06-01
Full Text Available A betavoltaic battery was prepared using radioactive 63Ni attached to a three-dimensional single trenched P–N absorber. The optimum thickness of a 63Ni layer was determined to be approximately 2 μm, considering the minimum self-shielding effect of beta particles. Electroplating of radioactive 63Ni on a nickel (Ni foil was carried out at a current density of 20 mA/cm2. The difference of the short-circuit currents (Isc between the pre- and postdeposition of 63Ni (16.65 MBq on the P–N junction was 5.03 nA, as obtained from the I–V characteristics. An improved design with a sandwich structure was provided for enhancing performance.
Decommissioning of an Irradiator MPX-γ - 25M and a neutron Irradiator
International Nuclear Information System (INIS)
Soguero, Dania; Guerra, Mercedes; Prieto, Enrique; Desdin, Luis
2013-01-01
In this paper a technology is developed with its procedures in radiation protection to ensure the safety of the process of decommissioning of two irradiators. Both processes are described, the process of decommissioning of a neutron Irradiator 4. 44·10 11 Bq, employed in the vegetal radio mutagenesis, and disassembling of an installation of gamma irradiation of 3.33 * 10 12 Bq, self-shielded of category I, model MPX - γ - 25 M. The specific objectives are: a) identify aspects of the contractual assurance, of human and technical resources, b) to evaluate the radiological situation of the process and c) analyze the potential radiological extraordinary events in each of the steps of the process, ensuring the right answers. Evaluation of radiological successful events described can be considered as reference to address the process of disassembling of other similar irradiators
Gamma ray auto absorption correction evaluation methodology
International Nuclear Information System (INIS)
Gugiu, Daniela; Roth, Csaba; Ghinescu, Alecse
2010-01-01
Neutron activation analysis (NAA) is a well established nuclear technique, suited to investigate the microstructural or elemental composition and can be applied to studies of a large variety of samples. The work with large samples involves, beside the development of large irradiation devices with well know neutron field characteristics, the knowledge of perturbing phenomena and adequate evaluation of correction factors like: neutron self shielding, extended source correction, gamma ray auto absorption. The objective of the works presented in this paper is to validate an appropriate methodology for gamma ray auto absorption correction evaluation for large inhomogeneous samples. For this purpose a benchmark experiment has been defined - a simple gamma ray transmission experiment, easy to be reproduced. The gamma ray attenuation in pottery samples has been measured and computed using MCNP5 code. The results show a good agreement between the computed and measured values, proving that the proposed methodology is able to evaluate the correction factors. (authors)
Nuclear Radiation Fields on the Mars Surface: Risk Analysis for Long-term Living Environment
Anderson, Brooke M.; Clowdsley, Martha S.; Qualls, Garry D.; Nealy, John E.
2005-01-01
Mars, our nearest planet outward from the sun, has been targeted for several decades as a prospective site for expanded human habitation. Background space radiation exposures on Mars are expected to be orders of magnitude higher than on Earth. Recent risk analysis procedures based on detailed dosimetric techniques applicable to sensitive human organs have been developed along with experimental data regarding cell mutation rates resulting from exposures to a broad range of particle types and energy spectra. In this context, simulated exposure and subsequent risk for humans in residence on Mars are examined. A conceptual habitat structure, CAD-modeled with duly considered inherent shielding properties, has been implemented. Body self-shielding is evaluated using NASA standard computerized male and female models. The background environment is taken to consist not only of exposure from incident cosmic ray ions and their secondaries, but also include the contribution from secondary neutron fields produced in the tenuous atmosphere and the underlying regolith.
Designing magnets with prescribed magnetic fields
International Nuclear Information System (INIS)
Liu Liping
2011-01-01
We present a novel design method capable of finding the magnetization densities that generate prescribed magnetic fields. The method is based on the solution to a simple variational inequality and the resulting designs have simple piecewise-constant magnetization densities. By this method, we obtain new designs of magnets that generate commonly used magnetic fields: uniform magnetic fields, self-shielding fields, quadrupole fields and sextupole fields. Further, it is worth noting that this method is not limited to the presented examples, and in particular, three-dimensional designs can be constructed in a similar manner. In conclusion, this novel design method is anticipated to have broad applications where specific magnetic fields are important for the performance of the devices.
Thermal neutron radiative capture cross-section of 186W(n, γ)187W reaction
International Nuclear Information System (INIS)
Tan, V H; Son, P N
2016-01-01
The thermal neutron radiative capture cross section for 186 W(n, γ) 187 W reaction was measured by the activation method using the filtered neutron beam at the Dalat research reactor. An optimal composition of Si and Bi, in single crystal form, has been used as neutron filters to create the high-purity filtered neutron beam with Cadmium ratio of R cd = 420 and peak energy E n = 0.025 eV. The induced activities in the irradiated samples were measured by a high resolution HPGe digital gamma-ray spectrometer. The present result of cross section has been determined relatively to the reference value of the standard reaction 197 Au(n, γ) 198 Au. The necessary correction factors for gamma-ray true coincidence summing, and thermal neutron self-shielding effects were taken into account in this experiment by Monte Carlo simulations. (paper)
Large packages for reactor decommissioning waste
International Nuclear Information System (INIS)
Price, M.S.T.
1991-01-01
This study was carried out jointly by the Atomic Energy Establishment at Winfrith (now called the Winfrith Technology Centre), Windscale Laboratory and Ove Arup and Partners. The work involved the investigation of the design of large transport containers for intermediate level reactor decommissioning waste, ie waste which requires shielding, and is aimed at European requirements (ie for both LWR and gas cooled reactors). It proposes a design methodology for such containers covering the whole lifetime of a waste disposal package. The design methodology presented takes account of various relevant constraints. Both large self shielded and returnable shielded concepts were developed. The work was generic, rather than specific; the results obtained, and the lessons learned, remain to be applied in practice
Fast Breeder Blanket Facility. Quarterly progress report, July 1, 1978--September 30, 1978
International Nuclear Information System (INIS)
Clikeman, F.M.
1978-09-01
This quarterly progress report summarizes work done at Purdue University's Fast Breeder Blanket Facility for the Department of Energy during the months July to September 1978. The summary includes reports on the models and methods used to characterize the FBBF facility. Using the reported models and calculational methods and computer codes a new cross section set has been generated, self-shielded for 300 0 K for use in all FBBF calculations using the 2DB computer code. The summary includes reports of the reproducability of foil activation data and measurements of the azmuthal symmetry of the facility. The status of the development of technique for the experimental measurements and preliminary foil activation measurements are also reviewed
Radiation Technician Scientist service
International Nuclear Information System (INIS)
Prieto Miranda, Enrique; Barrera Gonzalez, Gisela; Guerra Torres, Mercedes; Mora Lopez, Leonor; Altanes Valentin, Sonia; Rapado Paneque, Manuel; Plasencia Gutierrez, Manuel
2003-01-01
The irradiation service is part of the specialized technician scientist services of the Center of Technological Applications and Nuclear Development it belonging to the Radiobiological Department it provides a self shielded laboratory irradiator, PX y 30 type with Cobalt 60 sources, it destined for searches studies, so much basic as applying, in several branches of the science, like the radiobiology, the radiation chemistry, the solid state physics, the medicine, the agriculture and the Pharmaceutical- Medical Industry and besides offering the irradiation service properly with the which have been gotten significant economical outputs. The radiation processing is controlled by means of the dosimetric systems of Freckle, ceric cerous sulfate, Perspex (red, clear and Amber) and dose indicators
Randomly dispersed particle fuel model in the PSG Monte Carlo neutron transport code
International Nuclear Information System (INIS)
Leppaenen, J.
2007-01-01
High-temperature gas-cooled reactor fuels are composed of thousands of microscopic fuel particles, randomly dispersed in a graphite matrix. The modelling of such geometry is complicated, especially using continuous-energy Monte Carlo codes, which are unable to apply any deterministic corrections in the calculation. This paper presents the geometry routine developed for modelling randomly dispersed particle fuels using the PSG Monte Carlo reactor physics code. The model is based on the delta-tracking method, and it takes into account the spatial self-shielding effects and the random dispersion of the fuel particles. The calculation routine is validated by comparing the results to reference MCNP4C calculations using uranium and plutonium based fuels. (authors)
International Nuclear Information System (INIS)
Bechthold, W.; Closs, K.D.; Papp, R.
1988-01-01
Reference concepts for repositories in deep geological formations have been developed in several countries. For these concepts, emplacement configurations vary within a wide range that comprises drift emplacement of unshielded or self-shielded packages and horizontal or vertical borehole emplacement. This is caused by different parameters, criteria, and criteria weighting factors. Examples for parameters are the country's nuclear power program and waste management policy, its geological situation, and safety requirements, examples for criteria and repository area requirements, expenditures of mining and drilling, and efforts for emplacement and, if required, retrieval. Due to the variety of these factors and their ranking in different countries, requirements for a safe, dependable and cost-effective disposal of radioactive waste can be met in various ways
Benchmarking of EPRI-cell epithermal methods with the point-energy discrete-ordinates code (OZMA)
International Nuclear Information System (INIS)
Williams, M.L.; Wright, R.Q.; Barhen, J.; Rothenstein, W.
1982-01-01
The purpose of the present study is to benchmark E-C resonance-shielding and cell-averaging methods against a rigorous deterministic solution on a fine-group level (approx. 30 groups between 1 eV and 5.5 keV). The benchmark code used is OZMA, which solves the space-dependent slowing-down equations using continuous-energy discrete ordinates or integral transport theory to produce fine-group cross sections. Results are given for three water-moderated lattices - a mixed oxide, a uranium method, and a tight-pitch high-conversion uranium oxide configuration. The latter two lattices were chosen because of the strong self shielding of the 238 U resonances
ERRORJ. Covariance processing code. Version 2.2
International Nuclear Information System (INIS)
Chiba, Go
2004-07-01
ERRORJ is the covariance processing code that can produce covariance data of multi-group cross sections, which are essential for uncertainty analyses of nuclear parameters, such as neutron multiplication factor. The ERRORJ code can process the covariance data of cross sections including resonance parameters, angular and energy distributions of secondary neutrons. Those covariance data cannot be processed by the other covariance processing codes. ERRORJ has been modified and the version 2.2 has been developed. This document describes the modifications and how to use. The main topics of the modifications are as follows. Non-diagonal elements of covariance matrices are calculated in the resonance energy region. Option for high-speed calculation is implemented. Perturbation amount is optimized in a sensitivity calculation. Effect of the resonance self-shielding on covariance of multi-group cross section can be considered. It is possible to read a compact covariance format proposed by N.M. Larson. (author)
Analysis of the Rossendorf SEG experiments using the JNC route for reactor calculation
International Nuclear Information System (INIS)
Dietze, Klaus
1999-11-01
The integral experiments performed at the Rossendorf fast-thermal coupled reactor RRR/SEG have been reanalyzed using the JNC route for reactor calculation JENDL3.2/SLAROM/CITATION/JOINT/PERKY. The Rossendorf experiments comprise sample reactivity measurements with pure fission products and structural material in five configurations with different neutron and adjoint spectra. The shapes of the adjoint spectra have been designed to get high sensitivities to neutron capture or the scattering effect. The calculated neutron and adjoint spectra are in good agreement with former results obtained with the European route JEF2.2/ECCO/ERANOS. The C/E-values of the central reactivity worths of samples under investigation are given. Deviations in the results of both routes are due to the different libraries, codes, and self-shielding treatments used in the calculations. Results exceeding the experimental error are discussed. (author)
International Nuclear Information System (INIS)
Petrov, Nikolay; Todorova, Galina; Kolev, Nikola; Damian, Frederic
2011-01-01
The accurate and efficient MOC calculation scheme in APOLLO2, developed by CEA for generating multi-parameterized cross-section libraries for PWR assemblies, has been adapted to hexagonal assemblies. The neutronic part of this scheme is based on a two-level calculation methodology. At the first level, a multi-cell method is used in 281 energy groups for cross-section definition and self-shielding. At the second level, precise MOC calculations are performed in a collapsed energy mesh (30-40 groups). In this paper, the application and validation of the two-level scheme for hexagonal assemblies is described. Solutions for a VVER assembly are compared with TRIPOLI4® calculations and direct 281g MOC solutions. The results show that the accuracy is close to that of the 281g MOC calculation while the CPU time is substantially reduced. Compared to the multi-cell method, the accuracy is markedly improved. (author)
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
Kalpakkam multigroup cross section set for fast reactor applications - status and performance
International Nuclear Information System (INIS)
Ramanadhan, M.M.; Gopalakrishnan, M.M.
1986-01-01
This report documents the status of the presently created set of multigroup constants at Kalpakkam. The list of nuclides processed and the details of multigroup structure are given. Also included are the particulars of dilutions and temperatures for each nuclide in the multigroup cross section set for which self shielding factors have been calculated. Using this new multigroup cross section set, measured integral quantities such as K-eff, central reaction rate ratios, central reactivity worths etc. were calculated for a few fast critical benchmark assemblies and the calculated values of neutronic parameters obtained were compared with those obtained using the available Cadarache cross section library and those published in literature for ENDF/B-IV based set and Japanese evaluated nuclear data library (JENDL). The details of analyses are documented along with the conclusions. (author). 17 refs., 12 tabs
Li, Di; Heiles, Carl E.
2017-01-01
It is now well known that a substantial fraction of Galactic molecular gas cannot be traced by CO emission. The thus dubbed CO dark molecular gas (DMG) occupy a large volume of ISM with intermediate extinction, where CO is either not self-shielded and/or subthermally excited. We explore the utilities of simple hydrides, such OH, CH, etc., in tracing DMG. We mapped and modeled the transition zone cross a cloud boundary and derived emperical OH abundance and DMG distribution formulae. We also obtained absorption measurements of various species using Arecibo, VLA, ATCA, and ALMA. The absorption technique has the potential to provide systematic quantification of DMG in the next few years.
Sensitivity and uncertainty analysis of reactivities for UO2 and MOX fueled PWR cells
Energy Technology Data Exchange (ETDEWEB)
Foad, Basma [Research Institute of Nuclear Engineering, University of Fukui, Kanawa-cho 1-2-4, Tsuruga-shi, Fukui-ken, 914-0055 (Japan); Egypt Nuclear and Radiological Regulatory Authority, 3 Ahmad El Zomar St., Nasr City, Cairo, 11787 (Egypt); Takeda, Toshikazu [Research Institute of Nuclear Engineering, University of Fukui, Kanawa-cho 1-2-4, Tsuruga-shi, Fukui-ken, 914-0055 (Japan)
2015-12-31
The purpose of this paper is to apply our improved method for calculating sensitivities and uncertainties of reactivity responses for UO{sub 2} and MOX fueled pressurized water reactor cells. The improved method has been used to calculate sensitivity coefficients relative to infinite dilution cross-sections, where the self-shielding effect is taken into account. Two types of reactivities are considered: Doppler reactivity and coolant void reactivity, for each type of reactivity, the sensitivities are calculated for small and large perturbations. The results have demonstrated that the reactivity responses have larger relative uncertainty than eigenvalue responses. In addition, the uncertainty of coolant void reactivity is much greater than Doppler reactivity especially for large perturbations. The sensitivity coefficients and uncertainties of both reactivities were verified by comparing with SCALE code results using ENDF/B-VII library and good agreements have been found.
Gamma-ray imaging of the Quinby sources
International Nuclear Information System (INIS)
Gregor, J.; Hensley, D.C.
1996-01-01
The Quinby sources are alumina cylinders 7 inches in diameter and 8 inches high doped with weapons grade plutonium. We describe a computer tomography system for reconstructing three-dimensional images of these sources. Each reconstruction maps the spatial distribution of the internal [sup 241]Am gamma ray activity and is computed using an iterative, expectation-maximization algorithm with detection efficiencies based both on geometric model of the experimental setup and attenuation corrections. Constructed about a decade ago, the Quinby sources were to contain uniformly distributed material. However, for some of the sources we have found regions where the plutonium solution, tends to be concentrated. The ultimate goal of this work is to provide the basis for self-shielding corrections when analyzing differential dieaway neutron measurements
International Nuclear Information System (INIS)
Ljubenov, V.; Milosevic, M.
2003-01-01
A procedure for the neutron flux determination in a neutron field with an arbitrary energy spectrum, based on the using of standard methods for the measurement of irradiated foils activity and on the application of the SCALE-4.4a code system for averaged cross section calculation is described in this paper. Proposed procedure allows to include the energy spectrum of neutron flux reestablished in the location of irradiated foils and the resonance self-shielding effects in the foils also. Example application of this procedure is given for the neutron flux determination inside the neutron filter with boron placed in the centre of heavy water critical assembly RB at the Vinca Institute (author)
Energy Technology Data Exchange (ETDEWEB)
Soguero, Dania; Guerra, Mercedes; Prieto, Enrique; Desdin, Luis, E-mail: sdania@ceaden.edu.cu [Centro de Aplicaciones Tecnologica y Desarrollo Nuclear (CEADEN), La Habana (Cuba)
2013-07-01
In this paper a technology is developed with its procedures in radiation protection to ensure the safety of the process of decommissioning of two irradiators. Both processes are described, the process of decommissioning of a neutron Irradiator 4. 44{center_dot}10{sup 11}Bq, employed in the vegetal radio mutagenesis, and disassembling of an installation of gamma irradiation of 3.33 * 10{sup 12} Bq, self-shielded of category I, model MPX - {gamma} - 25 M. The specific objectives are: a) identify aspects of the contractual assurance, of human and technical resources, b) to evaluate the radiological situation of the process and c) analyze the potential radiological extraordinary events in each of the steps of the process, ensuring the right answers. Evaluation of radiological successful events described can be considered as reference to address the process of disassembling of other similar irradiators.
International Nuclear Information System (INIS)
Tzika, A.A.; White, D.L.; Aicher, K.P.; Moseley, M.E.; Price, D.
1990-01-01
This paper evaluates the biodistribution and clearance patterns of terrosomes as an intravascular magnetic susceptibility agent in a rat brain ischemia model. Sprague-Dawley rats were injected with Fe-59-labeled ferrosomes (0.05 mmol/kg). Blood and plasma Fe-59 clearance and T1 relaxivities were determined. Fe-59 organ distribution and excretion studies were also carried out. In a rat brain model of reversible middle cerebral artery occlusion, T2-weighted MR images were obtained with a 2.0-T General Electric CSI unit (with Acustar self-shielded gradient coils), before and after intravenous administration of the ferrosomes. The blood and plasma T1 relaxivity clearance was monoexponential (half life, 11 hours); plasma Fe-59 clearance was similar, but blood Fe-59 levels exhibited a delayed rise, indicating red blood cell radioactive iron incorporation
Temperature Calculation of Annular Fuel Pellet by Finite Difference Method
Energy Technology Data Exchange (ETDEWEB)
Yang, Yong Sik; Bang, Je Geon; Kim, Dae Ho; Kim, Sun Ki; Lim, Ik Sung; Song, Kun Woo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2009-10-15
KAERI has started an innovative fuel development project for applying dual-cooled annular fuel to existing PWR reactor. In fuel design, fuel temperature is the most important factor which can affect nuclear fuel integrity and safety. Many models and methodologies, which can calculate temperature distribution in a fuel pellet have been proposed. However, due to the geometrical characteristics and cooling condition differences between existing solid type fuel and dual-cooled annular fuel, current fuel temperature calculation models can not be applied directly. Therefore, the new heat conduction model of fuel pellet was established. In general, fuel pellet temperature is calculated by FDM(Finite Difference Method) or FEM(Finite Element Method), because, temperature dependency of fuel thermal conductivity and spatial dependency heat generation in the pellet due to the self-shielding should be considered. In our study, FDM is adopted due to high exactness and short calculation time.
Interfacial effects in fast reactors
International Nuclear Information System (INIS)
Saidi, M.S.; Driscoll, M.J.
1979-05-01
The problem of increased resonance capture rates near zone interfaces in fast reactor media has been examined both theoretically and experimentally. An interface traversing assembly was designed, constructed and employed to measure U-238 capture rates near th blanket--reflector interface in the MIT Blanket Test Facility. Prior MIT experiments on a thorium--uranium interface in a blanket assembly were also reanalyzed. Extremely localized fertile capture rate increases of on the order of 50% were measured immediately at the interfaces relative to extrapolation of asymptotic interior traverses, and relative to state-of-the-art (LIB-IV, SPHINX, ANISN/2DB) calculations which employ infinite-medium self-shielding throughout a given zone. A method was developed to compute a spatially varying background scattering cross section per absorber nucleus which takes into account both homogeneous and heterogeneous effects on the interface flux transient
Energy Technology Data Exchange (ETDEWEB)
Palau, J M [CEA Cadarache, Service de Physique des Reacteurs et du Cycle, Lab. de Projets Nucleaires, 13 - Saint-Paul-lez-Durance (France)
2005-07-01
This paper presents how Monte-Carlo calculations (French TRIPOLI4 poly-kinetic code with an appropriate pre-processing and post-processing software called OVNI) are used in the case of 3-dimensional heterogeneous benchmarks (slab reactor cores) to reduce model biases and enable a thorough and detailed analysis of the performances of deterministic methods and their associated data libraries with respect to key neutron parameters (reactivity, local power). Outstanding examples of application of these tools are presented regarding the new numerical methods implemented in the French lattice code APOLLO2 (advanced self-shielding models, new IDT characteristics method implemented within the discrete-ordinates flux solver model) and the JEFF3.1 nuclear data library (checked against JEF2.2 previous file). In particular we have pointed out, by performing multigroup/point-wise TRIPOLI4 (assembly and core) calculations, the efficiency (in terms of accuracy and computation time) of the new IDT method developed in APOLLO2. In addition, by performing 3-dimensional TRIPOLI4 calculations of the whole slab core (few millions of elementary volumes), the high quality of the new JEFF3.1 nuclear data files and revised evaluations (U{sup 235}, U{sup 238}, Hf) for reactivity prediction of slab cores critical experiments has been stressed. As a feedback of the whole validation process, improvements in terms of nuclear data (mainly Hf capture cross-sections) and numerical methods (advanced quadrature formulas accounting validation results, validation of new self-shielding models, parallelization) are suggested to improve even more the APOLLO2-CRONOS2 standard calculation route. (author)
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)
Nguyen, Van Do; Pham, Duc Khue; Kim, Tien Thanh; Kim, Guinyun; Lee, Manwoo; Kim, Kyung Sook; Kang, Heung-Sik; Cho, Moo-Hyun; Ko, In Soo; Namkung, Won
2011-01-01
The thermal neutron cross-section and the resonance integral of the 165Ho(n,γ) 166gHo reaction have been measured by the activation method using a 197Au(n,γ) 198Au monitor reaction as a single comparator. The high-purity natural Ho and Au foils with and without a cadmium shield case of 0.5 mm thickness were irradiated in a neutron field of the Pohang neutron facility. The induced activities in the activated foils were measured with a calibrated p-type high-purity Ge detector. The correction factors for the γ-ray attenuation ( Fg), the thermal neutron self-shielding ( Gth), the resonance neutron self-shielding ( Gepi) effects, and the epithermal neutron spectrum shape factor ( α) were taken into account. The thermal neutron cross-section for the 165Ho(n,γ) 166gHo reaction has been determined to be 59.7 ± 2.5 barn, relative to the reference value of 98.65 ± 0.09 barn for the 197Au(n,γ) 198Au reaction. By assuming the cadmium cut-off energy of 0.55 eV, the resonance integral for the 165Ho(n,γ) 166gHo reaction is 671 ± 47 barn, which is determined relative to the reference value of 1550 ± 28 barn for the 197Au(n,γ) 198Au reaction. The present results are, in general, good agreement with most of the previously reported data within uncertainty limits.
International Nuclear Information System (INIS)
Palau, J.M.
2005-01-01
This paper presents how Monte-Carlo calculations (French TRIPOLI4 poly-kinetic code with an appropriate pre-processing and post-processing software called OVNI) are used in the case of 3-dimensional heterogeneous benchmarks (slab reactor cores) to reduce model biases and enable a thorough and detailed analysis of the performances of deterministic methods and their associated data libraries with respect to key neutron parameters (reactivity, local power). Outstanding examples of application of these tools are presented regarding the new numerical methods implemented in the French lattice code APOLLO2 (advanced self-shielding models, new IDT characteristics method implemented within the discrete-ordinates flux solver model) and the JEFF3.1 nuclear data library (checked against JEF2.2 previous file). In particular we have pointed out, by performing multigroup/point-wise TRIPOLI4 (assembly and core) calculations, the efficiency (in terms of accuracy and computation time) of the new IDT method developed in APOLLO2. In addition, by performing 3-dimensional TRIPOLI4 calculations of the whole slab core (few millions of elementary volumes), the high quality of the new JEFF3.1 nuclear data files and revised evaluations (U 235 , U 238 , Hf) for reactivity prediction of slab cores critical experiments has been stressed. As a feedback of the whole validation process, improvements in terms of nuclear data (mainly Hf capture cross-sections) and numerical methods (advanced quadrature formulas accounting validation results, validation of new self-shielding models, parallelization) are suggested to improve even more the APOLLO2-CRONOS2 standard calculation route. (author)
Nuclear-data uncertainty propagations in burnup calculation for the PWR assembly
International Nuclear Information System (INIS)
Wan, Chenghui; Cao, Liangzhi; Wu, Hongchun; Shen, Wei
2017-01-01
Highlights: • The DRAGON 5.0 and NECP-CACTI have been implemented in UNICORN. • The effects of different neutronics methods on S&U results were quantified. • Uncertainty analysis has been applied to burnup calculation of PWR assembly. • The uncertainties of eigenvalue and few-group constants have been quantified. - Abstract: In this paper, our home-developed lattice code NECP-CACTI has been implemented into our UNICORN code to perform sensitivity and uncertainty analysis for the lattice calculations. The verified multigroup cross-section perturbation model and methods of the sensitivity and uncertainty analysis are established and applied to different lattice codes in UNICORN. As DRAGON5.0 and NECP-CACTI are available for the lattice calculations in UNICORN now, the effects of different neutronics methods (including methods for the neutron-transport and resonance self-shielding calculations) on the results of sensitivity and uncertainty analysis were studied in this paper. Based on NECP-CACTI, uncertainty analysis using the statistical sampling method has been performed to the burnup calculation for the fresh-fueled TMI-1 assembly, propagating the nuclear-data uncertainties to k_∞ and two-group constants of the lattice calculation with depletions. As results shown, for different neutronics methods, it can be observed that different methods of the neutron-transport calculation introduce no differences to the results of sensitivity and uncertainty analysis, while different methods of the resonance self-shielding calculation would impact the results. With depletions of the TMI-1 assembly, for k_∞, the relative uncertainty varies between 0.45% and 0.60%; for two-group constants, the largest variation is between 0.35% and 2.56% for vΣ_f_,_2. Moreover, the most significant contributors to the uncertainty of k_∞ and two-group constants varied with depletions are determined.
Zheng, Wangzhi; Cleveland, Zackary I; Möller, Harald E; Driehuys, Bastiaan
2011-02-01
When hyperpolarized noble gases are brought into the bore of a superconducting magnet for magnetic resonance imaging (MRI) or spectroscopy studies, the gases must pass through substantial field gradients, which can cause rapid longitudinal relaxation. In this communication, we present a means of calculating this spatially dependent relaxation rate in the fringe field of typical magnets. We then compare these predictions to experimental measurements of (3)He relaxation at various positions near a medium-bore 2-T small animal MRI system. The calculated and measured relaxation rates on the central axis of the magnet agree well and show a maximum (3)He relaxation rate of 3.83×10(-3) s(-1) (T(1)=4.4 min) at a distance of 47 cm from the magnet isocenter. We also show that if this magnet were self-shielded, its minimum T(1) would drop to 1.2 min. In contrast, a typical self-shielded 1.5-T clinical MRI scanner will induce a minimum on-axis T(1) of 12 min. Additionally, we show that the cylindrically symmetric fields of these magnets enable gradient-induced relaxation to be calculated using only knowledge of the on-axis longitudinal field, which can either be measured directly or calculated from a simple field model. Thus, while most MRI magnets employ complex and proprietary current configurations, we show that their fringe fields and the resulting gradient-induced relaxation are well approximated by simple solenoid models. Finally, our modeling also demonstrates that relaxation rates can increase by nearly an order of magnitude at radial distances equivalent to the solenoid radius. Copyright © 2010 Elsevier Inc. All rights reserved.
The EAGLE simulations: atomic hydrogen associated with galaxies
Crain, Robert A.; Bahé, Yannick M.; Lagos, Claudia del P.; Rahmati, Alireza; Schaye, Joop; McCarthy, Ian G.; Marasco, Antonino; Bower, Richard G.; Schaller, Matthieu; Theuns, Tom; van der Hulst, Thijs
2017-02-01
We examine the properties of atomic hydrogen (H I) associated with galaxies in the Evolution and Assembly of GaLaxies and their Environments (EAGLE) simulations of galaxy formation. EAGLE's feedback parameters were calibrated to reproduce the stellar mass function and galaxy sizes at z = 0.1, and we assess whether this calibration also yields realistic H I properties. We estimate the self-shielding density with a fitting function calibrated using radiation transport simulations, and correct for molecular hydrogen with empirical or theoretical relations. The `standard-resolution' simulations systematically underestimate H I column densities, leading to an H I deficiency in low-mass (M⋆ < 1010 M⊙) galaxies and poor reproduction of the observed H I mass function. These shortcomings are largely absent from EAGLE simulations featuring a factor of 8 (2) better mass (spatial) resolution, within which the H I mass of galaxies evolves more mildly from z = 1 to 0 than in the standard-resolution simulations. The largest volume simulation reproduces the observed clustering of H I systems, and its dependence on H I richness. At fixed M⋆, galaxies acquire more H I in simulations with stronger feedback, as they become associated with more massive haloes and higher infall rates. They acquire less H I in simulations with a greater star formation efficiency, since the star formation and feedback necessary to balance the infall rate is produced by smaller gas reservoirs. The simulations indicate that the H I of present-day galaxies was acquired primarily by the smooth accretion of ionized, intergalactic gas at z ≃ 1, which later self-shields, and that only a small fraction is contributed by the reincorporation of gas previously heated strongly by feedback. H I reservoirs are highly dynamic: over 40 per cent of H I associated with z = 0.1 galaxies is converted to stars or ejected by z = 0.
PLUTON, Isotope Generation and Depletion in Highly Irradiated LWR Fuel Rods
International Nuclear Information System (INIS)
Lemehov, Sergei; Motoe, Suzuki
2003-01-01
1 - Description of program or function: The PLUTON-PC is a three-group neutronic code analyzing, as functions of time and burnup, the change of radial profiles, together with average values, of power density, burnup, concentration of trans-uranium elements, plutonium buildup, depletion of fissile elements, and fission product generation in water reactor fuel rod with standard UO 2 , UO 2 -Gd 2 O 3 , inhomogeneous MOX, and UO 2 -ThO 2 . The PLUTON-PC code, which has been designed to be run on Windows PC, has adopted a theoretical shape function of neutron attenuation in pellet, which enables users to perform a very fast and accurate calculation easily. The code includes the irradiation conditions of the Halden Reactor which gives verification data for the code. Verification has been performed up to 83 GWd/tU, and a satisfactory agreement has been obtained. 2 - Methods: Based upon cumulative yields, the PLUTON-PC code calculates as a function of radial position and local burnup concentrations of fission products, macroscopic scattering cross-sections and self-shielding effect which is important for standard fuel (for Pu-242 mainly) and more importantly for homogeneous and inhomogeneous MOX fuel because of higher concentrations of fissile and fertile isotopes of plutonium. The code results in burnup dependent fission rate density profiles throughout the in-reactor irradiation of LWR fuel rods. The isotopes included in calculations have been extended to cover all trans-uranium groups (plutonium plus higher actinides) of fissile and fertile isotopes. Self-shielding problem and scattering effects have been revised and solved for all isotopes in the calculations for adequacy at high burnup, different irradiation conditions and cladding materials
Development and application of an on-line tritium production rate measuring method
International Nuclear Information System (INIS)
Yamaguchi, Seiya
1989-06-01
A highly sensitive on-line method for measuring the tritium production rate (TPR) of 6 Li was developed using the response difference of 6 Li and 7 Li-glass scintillators in a mixed neutron-gamma radiation field. A fitting method for subtracting the pulse height spectrum of 7 Li-glass from that of 6 Li-glass was introduced. The contribution of competing reactions such as 6 Li (n, n 'd) 4 He was estimated by kinematical analyses. An absolute value of the 6 Li content was determined by a chemical analysis. The thermal flux perturbation due to 6 Li-glass of various thickness and 6 Li contents was evaluated by measurement in a thermal neutron field and calculation by the modified Skyrme theory. A Monte Carlo calculation of the self-shielding effect was also made. The dependence of the self-shielding on neutron energy was examined by this Monte Carlo code. The edge effect, i.e., distortion of the pulse height spectrum due to partial energy deposition of the alpha and/or the triton, was investigated by measurement in a thermal neutron field and by a Monte Carlo simulation that was based on the scintillation mechanism and considered Bragg absorption and the ratio of contributions to luminescence by the alpha and the triton. The dependence of the edge effect on neutron energy was examined by this Monte Carlo code. This method was applied to the measurement of TPR distributions in simulated fusion blanket assemblies bombarded by D-T neutrons. Absolute values of the TPR were obtained with an experimental error of 3∼6 %. The measured results were compared with those of conventional β-counting methods and good agreement was obtained. An optical fiber system, using miniature lithium-glass scintillators, was fabricated for purpose of microminiaturization of detector size and adaption to strong electromagnetic field. Applicability of this system to a D-T neutron field was demonstrated. (author)
International Nuclear Information System (INIS)
Akino, Fujiyoshi; Takeuchi, Motoyoshi; Kitadate, Kenji; Yoshifuji, Hisashi; Kaneko, Yoshihiko
1980-11-01
As the core design for the Experimental Very High Temperature Gas Cooled Reactor progresses, evaluation of design precision has become increasingly important. For a high precision design, it is required to have adequate group constants based on accurate nuclear data, as well as calculation methods properly describing the physical behavior of neutrons. We, therefore, assembled a simulation core for VHTR, SHE-14, using a graphite-moderated 20%-enriched uranium Semi-Homogeneous Experimental Critical Facility (SHE), and obtained useful experimental data in evaluating the design precision. The VHTR is designed to accommodate burnable poison and control rods for reactivity compensation. Accordingly, the experimental burnable poison rods which are similar to those to be used in the experimental reactor were prepared, and their reactivity values were measured in the SHE-14 core. One to three rods of the above experimental burnable poison rods were inserted into the central column of the SHE-14 core, and the reactivity values were measured by the period and fuel rod substitution method. The results of the measurements have clearly shown that due to the self-shielding effect of B 4 C particles the reactivity value decreases with increasing particle diameter. For the particle diameter, the reactivity value is found to increase linearly with the logarithm of boron content. The measured values and those calculated are found to agree with each other within 5%. These results indicate that the reactivity of the burnable poison rod can be estimated fairly accurately by taking into account the self-shielding effect of B 4 C particles and the heterogeneity of the lattice cell. (author)
Analysis of the international criticality benchmark no 19 of a realistic fuel dissolver
International Nuclear Information System (INIS)
Smith, H.J.; Santamarina, A.
1991-01-01
The dispersion of the order of 12000 pcm in the results of the international criticality fuel dissolver benchmark calculation, exercise OECD/19, showed the necessity of analysing the calculational methods used in this case. The APOLLO/PIC method developed to treat this type of problem permits us to propose international reference values. The problem studied here, led us to investigate two supplementary parameters in addition to the double heterogeneity of the fuel: the reactivity variation as a function of moderation and the effects of the size of the fuel pellets during dissolution. The following conclusions were obtained: The fast cross-section sets used by the international SCALE package introduces a bias of - 3000 pcm in undermoderated lattices. More generally, the fast and resonance nuclear data in criticality codes are not sufficiently reliable. Geometries with micro-pellets led to an underestimation of reactivity at the end of dissolution of 3000 pcm in certain 1988 Sn calculations; this bias was avoided in the up-dated 1990 computation because of a correct use of calculation tools. The reactivity introduced by the dissolved fuel is underestimated by 3000 pcm in contributions based on the standard NITAWL module in the SCALE code. More generally, the neutron balance analysis pointed out that standard ND self shielding formalism cannot account for 238 U resonance mutual self-shielding in the pellet-fissile liquor interaction. The combination of these three types of bias explain the underestimation of all of the international contributions of the reactivity of dissolver lattices by -2000 to -6000 pcm. The improved 1990 calculations confirm the need to use rigorous methods in the calculation of systems which involve the fuel double heterogeneity. This study points out the importance of periodic benchmarking exercises for probing the efficacity of criticality codes, data libraries and the users
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)
Benchmark calculation of APOLLO-2 and SLAROM-UF in a fast reactor lattice
International Nuclear Information System (INIS)
Hazama, T.
2009-07-01
A lattice cell benchmark calculation is carried out for APOLLO2 and SLAROM-UF on the infinite lattice of a simple pin cell featuring a fast reactor. The accuracy in k-infinity and reaction rates is investigated in their reference and standard level calculations. In the 1. reference level calculation, APOLLO2 and SLAROM-UF agree with the reference value of k-infinity obtained by a continuous energy Monte Carlo calculation within 50 pcm. However, larger errors are observed in a particular reaction rate and energy range. The major problem common to both codes is in the cross section library of 239 Pu in the unresolved energy range. In the 2. reference level calculation, which is based on the ECCO 1968 group structure, both results of k-infinity agree with the reference value within 100 pcm. The resonance overlap effect is observed by several percents in cross sections of heavy nuclides. In the standard level calculation based on the APOLLO2 library creation methodology, a discrepancy appears by more than 300 pcm. A restriction is revealed in APOLLO2. Its standard cross section library does not have a sufficiently small background cross section to evaluate the self shielding effect on 56 Fe cross sections. The restriction can be removed by introducing the mixture self-shielding treatment recently introduced to APOLLO2. SLAROM-UF original standard level calculation based on the JFS-3 library creation methodology is the best among the standard level calculations. Improvement from the SLAROM-UF standard level calculation is achieved mainly by use of a proper weight function for light or intermediate nuclides. (author)
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)
Development of waste package designs for disposal in a salt repository
International Nuclear Information System (INIS)
Balmert, M.E.
1983-01-01
Three package design concepts were developed for CHLW and DHLW forms and spent fuel rods: (1) carbon steel overpack, borehole emplacement, (2) titanium clad, carbon steel reinforced overpack, borehole emplacement, and (3) carbon steel (self-shield) overpack, tunnel emplacement. For a DHLW canister with titanium clad overpack, the concept features a 9.5-cm-thick carbon steel overpack reinforcement supporting a 0.25-cm-thick titanium shell. The overall package dimensions are 84 cm diameter x 340 cm long weighing about 8.8 mtons. By contrast, a monolithic DHLW borehole package has a carbon steel overpack that is 10.4 cm thick, weighing about 9.3 mtons. The titanium clad/carbon steel reinforced borehole package is intended for remote emplacement in a vertical borehole in salt. The carbon steel overpack reinforcement provides structural integrity, primarily to resist external pressure, while the titanium overpack provides the necessary corrosion resistance to meet containment requirements. The carbon steel borehole package concept provides containment integrity for both external pressure and corrosion environments with a thicker carbon steel overpack in place of the titanium/carbon steel concept. A third concept utilizes an even greater thickness of cast steel or iron to resist external pressure and corrosion as well as reduce external shielding requirements. For example, a cast steel DHLW package would have overall dimensions of 125 cm diameter x 390 cm long, weighing 31 mtons. The purpose of this self-shield concept is to minimize handling and emplacement operations by reducing the package surface radiation dose to about 100 mrem/hr. In addition, it may serve as a shipping cask, thereby eliminating the need for a shielded hot cell at the repository for waste package assembly operations. 7 figures
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
Snead, C. J.; McKeegan, K. D.; Keller, L. P.; Messenger, S.
2017-01-01
The oxygen isotopic compositions of anhydrous minerals in carbonaceous chondrites reflect mixing between a O-16-rich and O-17, O18-rich reservoir. The UV photodissociation of CO (i.e. selfshielding) has been proposed as a mass-independent mechanism for producing these isotopically distinct reservoirs. Self-shielding models predict the composition for the CO gas reservoir to be O-16-rich, and that the accreting primordial dust was in isotopic equilibrium with the gaseous reservoir [1, 2]. Self-shielding also predicts that cometary water, presumed to represent the O-17, O-18-rich reservoir, should be enriched in O-17 and O-18, with compositions of 200 -1000per mille, and that the interaction with this O-17, O-18-rich H2O reservoir altered the compositions of the primordial dust toward planetary values. The bulk composition of the solar nebula, which may be an approximation to the 16O-rich gaseous reservoir, has been constrained by the Genesis results [3]. However, material representing the O-17, O-18-rich end-member is rare [4], and dust representing the original accreting primordial dust has been challenging to conclusively identify in current collections. Anhydrous dust from comets, which accreted in the distal cold regions of the nebula at temperatures below approximately 30K, may provide the best opportunity to measure the oxygen isotope composition of primordial dust. Chondritic porous interplanetary dust particles (CP-IDPs) have been suggested as having cometary origins [5]; however, until direct comparisons with dust from a known comet parent body were made, link between CP-IDPs and comets remained circumstantial. Oxygen isotope analyses of particles from comet 81P/Wild 2 collected by NASA's Stardust mission have revealed surprising similarities to minerals in carbonaceous chondrites which have been interpreted as evidence for large scale radial migration of dust components from the inner solar nebula to the accretion regions of Jupiter- family comets [6
Pen, Aranh
Since the discovery of anomalous sulfur isotope abundance in the geological record in sulfate and sulfide minerals (Farquhar et al., 2000), much effort has been put into understanding their origin to provide new insights into the environmental conditions on the early Earth (Farquhar et al., 2001; Pavlov and Kasting, 2002; Ono et al., 2003; Zahnle et al., 2006; Farquhar et al., 2007; Lyons, 2007; Lyons, 2008). This discovery gained immense interest because of its implications for both the lack of oxygen in the atmosphere during the Archean era 2.5-3.8 Gya (billion years ago), and for rise of oxygen, or the "Great Oxidation Event", that occurred 2.2-2.4 Gya (Holland, 2002). These signatures are believed to be produced in an anticorrelation to oxygen abundance in the early atmosphere, which will aid in quantifying the rate of oxygenation during the "Great Oxidation Event". According to Farquhar et al. (2000), the non-mass-dependent (NMD), or anomalous, fractionation signatures were produced by photochemical reactions of volcanic sulfur species in Earth's early atmosphere (> 2.3 Gya) due to the lack of an oxygen and ozone shield, resulting in an atmosphere transparent to solar ultraviolet (UV) radiation (Farquhar et al., 2001). Interpretation of the anomalous rock records, though, depends on the identification of (1) chemical reactions that can produce the NMD signature (Farquhar and Wing, 2003); and (2) conditions necessary for conversion of the gas-phase products into solid minerals (Pavlov and Kasting, 2002). The focus of my research addresses the first step, which is to determine whether the chemical reactions that occurred in Earth's early atmosphere, resulting in NMD fractionation of sulfur isotopes, were due to broadband UV photochemistry, and to test isotopic self-shielding as the possible underlying mechanism. In this project, our goals were to test isotopic self-shielding during UV photolysis as a possible underlying mechanism for anomalous sulfur isotopic
International Nuclear Information System (INIS)
Postma, H.; Blaauw, M.; Corvi, F.
2002-01-01
Neutron capture resonance analysis (NRCA) using a pulsed neutron beam and the time-of-flight (ToF) technique is a new method to determine the elemental compositions of artifacts. Neutron capture by an object can be observed by detecting the prompt capture gamma-radiation. Energies of resonance peaks in the ToF spectrum are the 'fingerprints' for elements. Since it is not necessary to determine the energy of the gamma-rays with any precision, it is possible to use a detector system with high detection efficiency. It is not necessary to take parts from an object for the analysis or to clean the surface or to do other things which might damage the object. Therefore NRCA is especially of interest for studying fragile, small or valuable objects from which one does not want to, or cannot take samples, or for which cleaning of even a small part of a surface is not desirable. Knowledge of the elemental composition of artifacts might be useful for archaeological or historical studies or to check the authenticity of an artifact. Recent experiments at the GELINA facility in Geel, Belgium show that indeed NRCA is a useful way to recognize elements on the basis of the energies of resonance in the ToF spectrum. We applied NRCA to several copper-alloy artifacts. In the studied objects very little activity was induced, which also disappeared quickly. Thus resonance energies allow us to recognize elements of an object. In addition, a quantitative analysis is possible on the basis of resonance areas. In the case of our artefacts the amounts of several elements (notably Sn, As, Zn, Fe, Sb, Ag, Au) were determined as ratios to copper. For a strong resonance it is necessary to take self-shielding into account. The effect of self-shielding made it possible to determine the absolute amount of copper by comparing the areas of a weak and a strong copper resonance, and thus also absolute amounts of the other components could be determined. The method of NRCA is discussed in relation to the
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)
Benchmark experiments on neutron streaming through JET Torus Hall penetrations
Batistoni, P.; Conroy, S.; Lilley, S.; Naish, J.; Obryk, B.; Popovichev, S.; Stamatelatos, I.; Syme, B.; Vasilopoulou, T.; contributors, JET
2015-05-01
Neutronics experiments are performed at JET for validating in a real fusion environment the neutronics codes and nuclear data applied in ITER nuclear analyses. In particular, the neutron fluence through the penetrations of the JET torus hall is measured and compared with calculations to assess the capability of state-of-art numerical tools to correctly predict the radiation streaming in the ITER biological shield penetrations up to large distances from the neutron source, in large and complex geometries. Neutron streaming experiments started in 2012 when several hundreds of very sensitive thermo-luminescence detectors (TLDs), enriched to different levels in 6LiF/7LiF, were used to measure the neutron and gamma dose separately. Lessons learnt from this first experiment led to significant improvements in the experimental arrangements to reduce the effects due to directional neutron source and self-shielding of TLDs. Here we report the results of measurements performed during the 2013-2014 JET campaign. Data from new positions, at further locations in the South West labyrinth and down to the Torus Hall basement through the air duct chimney, were obtained up to about a 40 m distance from the plasma neutron source. In order to avoid interference between TLDs due to self-shielding effects, only TLDs containing natural Lithium and 99.97% 7Li were used. All TLDs were located in the centre of large polyethylene (PE) moderators, with natLi and 7Li crystals evenly arranged within two PE containers, one in horizontal and the other in vertical orientation, to investigate the shadowing effect in the directional neutron field. All TLDs were calibrated in the quantities of air kerma and neutron fluence. This improved experimental arrangement led to reduced statistical spread in the experimental data. The Monte Carlo N-Particle (MCNP) code was used to calculate the air kerma due to neutrons and the neutron fluence at detector positions, using a JET model validated up to the
Neutron flux measurements in C-9 capsule pressure tube
International Nuclear Information System (INIS)
Barbos, D.; Roth, C. S.; Gugiu, D.; Preda, M.
2001-01-01
C-9 capsule is a fuel testing facility in which the testing consists of a daily cycle ranging between the limits 100% power to 50% power. C-9 in-pile section with sample holder an instrumentation are introduced in G-9 and G-10 experimental channels. The experimental fuel channel has a maximum value when the in-pile section (pressure tube) is in G-9 channel and minimum value in G-10 channel. In this paper the main goals are determination or measurements of: - axial thermal neutron flux distribution in C-9 pressure tube both in G-9 and G-10 channel; - ratio of maximum neutron flux value in G-9 and the same value in G-9 channel and the same value in G-10 channel; - neutron flux-spectrum. On the basis of axial neutron flux distribution measurements, the experimental fuel element in sample holder position in set. Both axial neutron flux distribution of thermal neutrons and neutron flux-spectrum were performed using multi- foil activation technique. Activation rates were obtained by absolute measurements of the induced activity using gamma spectroscopy methods. To determine the axial thermal neutron flux distribution in G-9 and G-10, Cu 100% wire was irradiated at the reactor power of 2 MW. Ratio between the two maximum values, in G-9 and G-10 channels, is 2.55. Multi-foil activation method was used for neutron flux spectrum measurements. The neutron spectra and flux were obtained from reaction rate measurements by means of SAND 2 code. To obtain gamma-ray spectra, a HPGe detector connected to a multichannel analyzer was used. The spectrometer is absolute efficiency calibrated. The foils were irradiated at 2 MW reactor power in previously determined maximum flux position resulted from wire measurements. This reaction rates were normalized for 10 MW reactor power. Neutron self shielding corrections for the activation foils were applied. The self-shielding corrections are computed using Monte Carlo simulation methods. The measured integral flux is 1.1·10 14 n/cm 2 s
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
NJOY 99/2001: new capabilities in data processing
International Nuclear Information System (INIS)
MacFarlane, Robert E.
2002-01-01
The NJOY Nuclear Data Processing System is used all over the world to process evaluated nuclear data in the ENDF format into libraries for applications. Over the last few years, a number of new capabilities have been added to the system to provide advanced features for MCNP, MCNPX, and other applications codes. These include probability tables for unresolved range self shielding, capabilities optimized for high-energy libraries (typically to 150 MeV for accelerator applications), options for detailed treatments of incident and outgoing charged particles, and a capability to handle photonuclear reactions. These new features and recent experience using NJOY99 for library production will be discussed, along with possible future work, such as delayed-neutron processing and capabilities to handle the new generation of photo-atomic, electro-atomic, and atomic-relaxation evaluations now becoming available in ENDF format. The latest version of the code, NJOY 2001, uses modern Fortran90 style, modularization, and memory allocation methods. The Evaluated Nuclear Data Files (ENDF) format has become the standard for representing nuclear data throughout the world, being used in the US ENDF/B libraries, the European JEF libraries, the Japanese JENDL libraries, and many others. At the same time, the NJOY Nuclear Data Processing System, which is used to convert evaluated nuclear data in the ENDF format into data libraries for nuclear applications, has become the method of choice throughout the world. The combination of these modern libraries of evaluated nuclear data and NJOY processing has proved very capable for classical applications in reactor analysis, fusion work, shielding, and criticality safety. However, over the last few years, new applications have appeared that require extended evaluated data and new processing techniques. A good example of this is the interest in accelerator-boosted applications, which has led to the need for data to higher energies, such as 150 Me
FISPACT-II: An Advanced Simulation System for Activation, Transmutation and Material Modelling
Energy Technology Data Exchange (ETDEWEB)
Sublet, J.-Ch., E-mail: jean-christophe.sublet@ukaea.uk [United Kingdom Atomic Energy Authority, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Eastwood, J.W.; Morgan, J.G. [Culham Electromagnetics Ltd, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Gilbert, M.R.; Fleming, M.; Arter, W. [United Kingdom Atomic Energy Authority, Culham Science Centre, Abingdon OX14 3DB (United Kingdom)
2017-01-15
Fispact-II is a code system and library database for modelling activation-transmutation processes, depletion-burn-up, time dependent inventory and radiation damage source terms caused by nuclear reactions and decays. The Fispact-II code, written in object-style Fortran, follows the evolution of material irradiated by neutrons, alphas, gammas, protons, or deuterons, and provides a wide range of derived radiological output quantities to satisfy most needs for nuclear applications. It can be used with any ENDF-compliant group library data for nuclear reactions, particle-induced and spontaneous fission yields, and radioactive decay (including but not limited to TENDL-2015, ENDF/B-VII.1, JEFF-3.2, JENDL-4.0u, CENDL-3.1 processed into fine-group-structure files, GEFY-5.2 and UKDD-16), as well as resolved and unresolved resonance range probability tables for self-shielding corrections and updated radiological hazard indices. The code has many novel features including: extension of the energy range up to 1 GeV; additional neutron physics including self-shielding effects, temperature dependence, thin and thick target yields; pathway analysis; and sensitivity and uncertainty quantification and propagation using full covariance data. The latest ENDF libraries such as TENDL encompass thousands of target isotopes. Nuclear data libraries for Fispact-II are prepared from these using processing codes PREPRO, NJOY and CALENDF. These data include resonance parameters, cross sections with covariances, probability tables in the resonance ranges, PKA spectra, kerma, dpa, gas and radionuclide production and energy-dependent fission yields, supplemented with all 27 decay types. All such data for the five most important incident particles are provided in evaluated data tables. The Fispact-II simulation software is described in detail in this paper, together with the nuclear data libraries. The Fispact-II system also includes several utility programs for code-use optimisation
Verification of High Temperature Free Atom Thermal Scattering in MERCURY Compared to TART
International Nuclear Information System (INIS)
Cullen, D E; McKinley, S; Hagmann, C
2006-01-01
This is part of a series of reports verifying the accuracy of the relatively new MERCURY [1] Monte Carlo particle transport code by comparing its results to those of the older TART [2] Monte Carlo particle transport code. In the future we hope to extend these comparisons to include deterministic (Sn) codes [3]. Here we verify the accuracy of the free atom thermal scattering model [4] by using it over a very large temperature range. We would like to be able to use these Monte Carlo codes for astrophysical applications, where the temperature of the medium can be extremely high compared to the temperatures we normally encounter in our terrestrial applications [5]. The temperature is so high that is it often defined in eV rather than Kelvin. For a correspondence between the two scale 293.6 Kelvin (room temperature) corresponds to 0.0253 eV ∼ 1/40 eV. So that 1 eV temperature is about 12,000 Kelvin, and 1 keV temperature is about 12 million Kelvin. Here we use a relatively small system measured in cm, but by using ρR scaling [6] our results are equally applicable to systems measured in Km or thousands of Km or any size that we need for astrophysical applications. The emphasis here is not on modeling any given real system, but rather in verifying the accuracy of the free atom model to represent theoretical results over a large temperature range. There are two primary objectives of this report: (1) Verify agreement between MERCURY and TART results, both using continuous energy cross sections. In particular we want to verify the free atom scattering treatment in MERCURY as used over an extended temperature range; by comparison to many other codes for TART this has already been verified over many years [4, 7]. (2) Demonstrate that this agreement depends on using continuous energy cross sections. To demonstrate this we also present TART using the Multi-Band method [8, 9], which accounts for resonance self-shielding, and Multi-Group method, without self-shielding [9
International Nuclear Information System (INIS)
Townsend, Lawrence W.; Zapp, E. Neal
1999-01-01
The true uncertainties in estimates of body organ absorbed dose and dose equivalent, from exposures of interplanetary astronauts to large solar particle events (SPEs), are essentially unknown. Variations in models used to parameterize SPE proton spectra for input into space radiation transport and shielding computer codes can result in uncertainty about the reliability of dose predictions for these events. Also, different radiation transport codes and their input databases can yield significant differences in dose predictions, even for the same input spectra. Different results may also be obtained for the same input spectra and transport codes if different spacecraft and body self-shielding distributions are assumed. Heretofore there have been no systematic investigations of the variations in dose and dose equivalent resulting from these assumptions and models. In this work we present a study of the variability in predictions of organ dose and dose equivalent arising from the use of different parameters to represent the same incident SPE proton data and from the use of equivalent sphere approximations to represent human body geometry. The study uses the BRYNTRN space radiation transport code to calculate dose and dose equivalent for the skin, ocular lens and bone marrow using the October 1989 SPE as a model event. Comparisons of organ dose and dose equivalent, obtained with a realistic human geometry model and with the oft-used equivalent sphere approximation, are also made. It is demonstrated that variations of 30-40% in organ dose and dose equivalent are obtained for slight variations in spectral fitting parameters obtained when various data points are included or excluded from the fitting procedure. It is further demonstrated that extrapolating spectra from low energy (≤30 MeV) proton fluence measurements, rather than using fluence data extending out to 100 MeV results in dose and dose equivalent predictions that are underestimated by factors as large as 2
The evolution of and challenges for industrial radiation processing—2012
International Nuclear Information System (INIS)
Berejka, A.J.; Cleland, M.R.; Walo, M.
2014-01-01
The evolution of industrial radiation processing is traced from Roentgen's discovery of X-radiation in 1895 by following the development of high current, electron beam accelerators (EB) throughout the twentieth century. Although Becquerel soon followed Roentgen with his discovery of what became to be known as radioactivity, electrical sources for ionizing radiation dominate industrial processing with there being more than ten times as many industrial installations using high current EB equipment than the facilities relying upon large concentrations of radioactive isotopes. In the 1950s, the discovery that ionizing radiation would enhance the value of what has become the world's largest volume commodity plastic, polyethylene (PE), opened the way for full scale commercial use of high current EB equipment. While the crosslinking of the PE insulation on wire became one of the first major industrial applications, other uses of EB processing soon followed. In the 1970s, low-energy, self-shielded EB equipment made the surface curing of inks, coatings and adhesives more industrially viable. In the early part of the twenty-first century, new market applications involving the low-energy EB surface decontamination of packaging materials emerged. This new area poses challenges for the metrology needed to control industrial processes, in that there is limited EB penetration into what have been used as dosimeters by industry. Major industrial use of radiation process is now over 50 years old. Because of the diversity of end-uses and the fact that the use of ionizing radiation in industry is a process technique, it is hard to quantify the value-added to numerous commercial products that benefit from this energy efficient process. It may be in excess of a trillion Euros in value-added to articles of commerce. In this milieu, there are some broad-based opportunities for research which are noted. - Highlights: ► The evolution of industrial radiation processing is traced from
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
International Nuclear Information System (INIS)
Budak, M.G.; Karadag, M.; Yuecel, H.
2010-01-01
The effective resonance energies E - bar r for the (n,γ) reactions of 152 Sm and 165 Ho isotopes were determined by using dual monitors ( 55 Mn- 98 Mo) due to their favourable resonance properties. The samples were irradiated in an isotropic neutron field obtained from 241 Am-Be neutron sources. The induced activities were measured with a high efficient, p-type Ge detector. The necessary correction factors for thermal neutron self-shielding (G th ), resonance neutron self-shielding (G epi ), self absorption (F s ) and true coincidence summing (F coi ) effects for the measured γ-rays were taken into account. Thus, the experimental E - bar r -values for above (n,γ) reactions are found to be 8.65 ± 1.80 eV for 152 Sm and 12.90 ± 2.69 eV for 165 Ho isotopes, respectively. The E - bar r -values for both 152 Sm and 165 Ho isotopes were also theoretically calculated from the newest resonance data in the literature. Theoretically calculated E - bar r -values are estimated to be 8.34 eV and 8.53 eV for 152 Sm by two different approaches, which are generally, much smaller than that the present experimental value by 1.4-3.6% for 152 Sm. In case of 165 Ho isotope, the theoretically calculated E - bar r -value of 8.63 eV from the first approach deviates substantially from the measured value by about 33%, whereas the theoretical E - bar r -value of 12.95 eV from the second approach agrees very well with our experimentally determined E - bar r -value. The results show that the present experimental E - bar r -values for 152 Sm and 165 Ho isotopes agree with the calculated ones from the second approach within limits of the estimated uncertainty if the recently evaluated resonance data are used. However, it is worth noting that the results for E - bar r -value calculated from the first approach are not satisfactorily accurate because of neglecting the neutron widths in that approach. Therefore, this study implies that it be regarded to the experimentally determined E - bar r
A new Tone's method in APOLLO3® and its application to fast and thermal reactor calculations
Directory of Open Access Journals (Sweden)
Li Mao
2017-09-01
Full Text Available This paper presents a newly developed resonance self-shielding method based on Tone's method in APOLLO3® for fast and thermal reactor calculations. The new method is based on simplified models, the narrow resonance approximation for the slowing down source and Tone's approximation for group collision probability matrix. It utilizes mathematical probability tables as quadrature formulas in calculating effective cross-sections. Numerical results for the ZPPR drawer calculations in 1,968 groups show that, in the case of the double-column fuel drawer, Tone's method gives equivalent precision to the subgroup method while markedly reducing the total number of collision probability matrix calculations and hence the central processing unit time. In the case of a single-column fuel drawer with the presence of a uranium metal material, Tone's method obtains less precise results than those of the subgroup method due to less precise heterogeneous–homogeneous equivalence. The same options are also applied to PWR UOX, MOX, and Gd cells using the SHEM 361-group library, with the objective of analyzing whether this energy mesh might be suitable for the application of this methodology to thermal systems. The numerical results show that comparable precision is reached with both Tone's and the subgroup methods, with the satisfactory representation of intrapellet spatial effects.
Directory of Open Access Journals (Sweden)
Volmert Ben
2016-01-01
Full Text Available In this paper, an overview of the Swiss Nuclear Power Plant (NPP activation methodology is presented and the work towards its validation by in-situ NPP foil irradiation campaigns is outlined. Nuclear Research and consultancy Group (NRG in The Netherlands has been given the task of performing the corresponding neutron metrology. For this purpose, small Aluminium boxes containing a set of circular-shaped neutron activation foils have been prepared. After being irradiated for one complete reactor cycle, the sets have been successfully retrieved, followed by gamma-spectrometric measurements of the individual foils at NRG. Along with the individual activities of the foils, the reaction rates and thermal, intermediate and fast neutron fluence rates at the foil locations have been determined. These determinations include appropriate corrections for gamma self-absorption and neutron self-shielding as well as corresponding measurement uncertainties. The comparison of the NPP Monte Carlo calculations with the results of the foil measurements is done by using an individual generic MCNP model functioning as an interface and allowing the simulation of individual foil activation by predetermined neutron spectra. To summarize, the comparison between calculation and measurement serve as a sound validation of the Swiss NPP activation methodology by demonstrating a satisfying agreement between measurement and calculation. Finally, the validation offers a chance for further improvements of the existing NPP models by ensuing calibration and/or modelling optimizations for key components and structures.
RHEIN, Modular System for Reactor Design Calculation
International Nuclear Information System (INIS)
Reiche, Christian; Barz, Hansulrich; Kunzmann, Bernd; Seifert, Eberhard; Wand, Hartmut
1990-01-01
1 - Description of program or function: RHEIN is a modular reactor code system for neutron physics calculations. It consists of a small number of system codes for execution control, data management, and handling support, as well as of the physical calculation routines. The execution is controlled by input data containing mathematical and physical parameters and simple commands for routine calls and data manipulations. The calculation routines are in tune with one another and the system takes care of the data transfer between them. Cross-section libraries with self shielding parameters are added to the system. 2 - Method of solution: The calculation routines can be used for solving the following physics problems: - Calculation of cross-section sets for infinite mediums, taking into account chord length. - Zero-dimensional spectrum calculation in diffusion, P1, or B1 approximation. - One-dimensional calculation in diffusion, P1, or collision probability approximation. - Two-dimensional diffusion calculation. - Cell calculation by THERMOS. - Zone-wise homogenized group collapsing within zero, one, or two-dimensional models. - Normalization, summarizing, etc. - Output of cross-section sets to off systems Sn and Monte-Carlo calculations
Large Sample Neutron Activation Analysis of Heterogeneous Samples
International Nuclear Information System (INIS)
Stamatelatos, I.E.; Vasilopoulou, T.; Tzika, F.
2018-01-01
A Large Sample Neutron Activation Analysis (LSNAA) technique was developed for non-destructive analysis of heterogeneous bulk samples. The technique incorporated collimated scanning and combining experimental measurements and Monte Carlo simulations for the identification of inhomogeneities in large volume samples and the correction of their effect on the interpretation of gamma-spectrometry data. Corrections were applied for the effect of neutron self-shielding, gamma-ray attenuation, geometrical factor and heterogeneous activity distribution within the sample. A benchmark experiment was performed to investigate the effect of heterogeneity on the accuracy of LSNAA. Moreover, a ceramic vase was analyzed as a whole demonstrating the feasibility of the technique. The LSNAA results were compared against results obtained by INAA and a satisfactory agreement between the two methods was observed. This study showed that LSNAA is a technique capable to perform accurate non-destructive, multi-elemental compositional analysis of heterogeneous objects. It also revealed the great potential of the technique for the analysis of precious objects and artefacts that need to be preserved intact and cannot be damaged for sampling purposes. (author)
Karadag, M; Tan, M; Oezmen, A
2003-01-01
Thermal neutron cross-sections and resonance integrals for the sup 7 sup 1 Ga(n,gamma) sup 7 sup 2 Ga and sup 7 sup 5 As(n,gamma) sup 7 sup 6 As reactions were measured by the activation method. The experimental samples with and without a cylindrical Cd shield case in 1 mm wall thickness were irradiated in an isotropic neutron field of the sup 2 sup 4 sup 1 Am-Be neutron source. The induced activities in the samples were measured by high-resolution gamma-ray spectrometry with a calibrated reverse-electrode germanium detector. Thermal neutron cross-sections for 2200 m/s neutrons and resonance integrals for the sup 7 sup 1 Ga(n,gamma) sup 7 sup 2 Ga and sup 7 sup 5 As(n,gamma) sup 7 sup 6 As reactions have been obtained relative to the reference values, sigma sub 0 =13.3+-0.1 b and I sub 0 =14.0+-0.3 b for the sup 5 sup 5 Mn(n,gamma) sup 5 sup 6 Mn reaction as a single comparator. The necessary correction factors for gamma attenuation, thermal neutron and resonance neutron self-shielding effects were taken into...
MULTIPLE ORIGINS OF NITROGEN ISOTOPIC ANOMALIES IN METEORITES AND COMETS
International Nuclear Information System (INIS)
Aleon, Jerome
2010-01-01
Isotopic fractionation and mixing calculations compared with coupled hydrogen and nitrogen isotopic composition of organic molecules from primitive chondrites, interplanetary dust particles (IDPs), and comets C/1995 O1 (Hale-Bopp) and 81P/Wild2 reveal that meteoritic and cometary organic matter contains three different isotopic components of different origins. (1) A major component of carbonaceous chondrites, IDPs, and comets Hale-Bopp and Wild2 shows correlated H and N isotopic compositions attributable to isotope exchange between an organic matter of solar composition and a reservoir formed by ion-molecule reactions at T 15 N-rich component having identical 15 N and D enrichments relative to the protosolar gas. Temperatures > 100 K deduced from the low D/H ratio and an anti-correlation between the abundance of this component and meteoritic age indicate a late origin in the solar protoplanetary disk. N 2 self-shielding and the non-thermal nucleosynthesis of 15 N upon irradiation are possible but unlikely sources of this component, and a chemical origin is preferred. (3) An interstellar component with highly fractionated hydrogen isotopes and unfractionated nitrogen isotopes is present in ordinary chondrites. A dominantly solar origin of D and 15 N excesses in primitive solar system bodies shows that isotopic anomalies do not necessarily fingerprint an interstellar origin and implies that only a very small fraction of volatile interstellar matter survived the events of solar system formation.
Error in interpreting field chlorophyll fluorescence measurements: heat gain from solar radiation
International Nuclear Information System (INIS)
Marler, T.E.; Lawton, P.D.
1994-01-01
Temperature and chlorophyll fluorescence characteristics were determined on leaves of various horticultural species following a dark adaptation period where dark adaptation cuvettes were shielded from or exposed to solar radiation. In one study, temperature of Swietenia mahagoni (L.) Jacq. leaflets within cuvettes increased from approximately 36C to approximately 50C during a 30-minute exposure to solar radiation. Alternatively, when the leaflets and cuvettes were shielded from solar radiation, leaflet temperature declined to 33C in 10 to 15 minutes. In a second study, 16 horticultural species exhibited a lower variable: maximum fluorescence (F v :F m ) when cuvettes were exposed to solar radiation during the 30-minute dark adaptation than when cuvettes were shielded. In a third study with S. mahagoni, the influence of self-shielding the cuvettes by wrapping them with white tape, white paper, or aluminum foil on temperature and fluorescence was compared to exposing or shielding the entire leaflet and cuvette. All of the shielding methods reduced leaflet temperature and increased the F v :F m ratio compared to leaving cuvettes exposed. These results indicate that heat stress from direct exposure to solar radiation is a potential source of error when interpreting chlorophyll fluorescence measurements on intact leaves. Methods for moderating or minimizing radiation interception during dark adaptation are recommended. (author)
Study on exposures in incidents situations involving X-ray generators of industrial use
International Nuclear Information System (INIS)
Rios, Paulo B.; Rios, Denise A. da S.
2014-01-01
Case studies of accidents involving radiation sources provides valuable information for the improvement of equipment and procedures to be adopted to prevent their recurrence. Therefore, they are recommended as part of the licensing process of radiative facilities. However, when the equipment has no radiological protection requirements or have no record of accidents, an effective method to appropriate this knowledge necessary to improvements of process and project is the investigation of reported incidents and the proposition limit situations where exposures unlikely and even low intensity can occur. This work reviews incidents in X-ray equipment and thickness meters. Emergency situations are also studied in this type of equipment and in an accelerator of electrons self-shielded for curing of paints and varnishes. The exposure estimates are calculated from data collected in workplaces on the distance and time for each proposed case. Although there is no severe consequences to occupational health, the results can be used to improve the quality of training for operators and maintenance technicians, for proposing of new alarm systems and emergency teams procedures and may serve as a warning in inadvertent and avoidable situations of unnecessary exposures
International Nuclear Information System (INIS)
Abánades, A.; Álvarez-Velarde, F.; González-Romero, E.M.; Ismailov, K.; Lafuente, A.; Nishihara, K.; Saito, M.; Stanculescu, A.; Sugawara, T.
2013-01-01
Highlights: ► TARC experiment benchmark capture rates results. ► Utilization of updated databases, included ADSLib. ► Self-shielding effect in reactor design for transmutation. ► Effect of Lead nuclear data. - Abstract: The design of Accelerator Driven Systems (ADS) requires the development of simulation tools that are able to describe in a realistic way their nuclear performance and transmutation rate capability. In this publication, we present an evaluation of state of the art Monte Carlo design tools to assess their performance concerning transmutation of long-lived fission products. This work, performed under the umbrella of the International Atomic Energy Agency, analyses two important aspects for transmutation systems: moderation on Lead and neutron captures of 99 Tc, 127 I and 129 I. The analysis of the results shows how shielding effects due to the resonances at epithermal energies of these nuclides affects strongly their transmutation rate. The results suggest that some research effort should be undertaken to improve the quality of Iodine nuclear data at epithermal and fast neutron energy to obtain a reliable transmutation estimation.
A non-terrestrial 16O-rich isotopic composition for the protosolar nebula.
Hashizume, Ko; Chaussidon, Marc
2005-03-31
The discovery in primitive components of meteorites of large oxygen isotopic variations that could not be attributed to mass-dependent fractionation effects has raised a fundamental question: what is the composition of the protosolar gas from which the host grains formed? This composition is probably preserved in the outer layers of the Sun, but the resolution of astronomical spectroscopic measurements is still too poor to be useful for comparison with planetary material. Here we report a precise determination of the oxygen isotopic composition of the solar wind from particles implanted in the outer hundreds of nanometres of metallic grains in the lunar regolith. These layers of the grains are enriched in 16O by >20 +/- 4 per thousand relative to the Earth, Mars and bulk meteorites, which implies the existence in the solar accretion disk of reactions--as yet unknown--that were able to change the 17O/16O and 18O/16O ratios in a way that was not dependent strictly on the mass of the isotope. Photochemical self-shielding of the CO gas irradiated by ultraviolet light may be one of these key processes, because it depends on the abundance of the isotopes, rather than their masses.
Andrianova, Olga; Lomakov, Gleb; Manturov, Gennady
2017-09-01
The neutron transmission experiments are one of the main sources of information about the neutron cross section resonance structure and effect in the self-shielding. Such kind of data for niobium and silicon nuclides in energy range 7 keV to 3 MeV can be obtained from low-resolution transmission measurements performed earlier in Russia (with samples of 0.027 to 0.871 atom/barn for niobium and 0.076 to 1.803 atom/barn for silicon). A significant calculation-to-experiment discrepancy in energy range 100 to 600 keV and 300 to 800 keV for niobium and silicon, respectively, obtained using the evaluated nuclear data library ROSFOND, were found. The EVPAR code was used for estimation the average resonance parameters in energy range 7 to 600 keV for niobium. For silicon a stochastic optimization method was used to modify the resolved resonance parameters in energy range 300 to 800 keV. The improved ROSFOND evaluated nuclear data files were tested in calculation of ICSBEP integral benchmark experiments.
NJOY99, Data Processing System of Evaluated Nuclear Data Files ENDF Format
International Nuclear Information System (INIS)
2000-01-01
1 - Description of program or function: The NJOY nuclear data processing system is a modular computer code used for converting evaluated nuclear data in the ENDF format into libraries useful for applications calculations. Because the Evaluated Nuclear Data File (ENDF) format is used all around the world (e.g., ENDF/B-VI in the US, JEF-2.2 in Europe, JENDL-3.2 in Japan, BROND-2.2 in Russia), NJOY gives its users access to a wide variety of the most up-to-date nuclear data. NJOY provides comprehensive capabilities for processing evaluated data, and it can serve applications ranging from continuous-energy Monte Carlo (MCNP), through deterministic transport codes (DANT, ANISN, DORT), to reactor lattice codes (WIMS, EPRI). NJOY handles a wide variety of nuclear effects, including resonances, Doppler broadening, heating (KERMA), radiation damage, thermal scattering (even cold moderators), gas production, neutrons and charged particles, photo-atomic interactions, self shielding, probability tables, photon production, and high-energy interactions (to 150 MeV). Output can include printed listings, special library files for applications, and Postscript graphics (plus color). More information on NJOY is available from the developer's home page at http://t2.lanl.gov/tour/tourbus.html. Follow the Tourbus section of the Tour area to find notes from the ICTP lectures held at Trieste in March 2000 on the ENDF format and on the NJOY code. NJOY contains the following modules: NJOY directs the flow of data through the other modules and contains a library of common functions and subroutines used by the other modules. RECONR reconstructs pointwise (energy-dependent) cross sections from ENDF resonance parameters and interpolation schemes. BROADR Doppler broadens and thins pointwise cross sections. UNRESR computes effective self-shielded pointwise cross sections in the unresolved energy range. HEATR generates pointwise heat production cross sections (KERMA coefficients) and radiation
Two-dimensional shielding benchmarks for iron at YAYOI, (1)
International Nuclear Information System (INIS)
Oka, Yoshiaki; An, Shigehiro; Kasai, Shigeru; Miyasaka, Shun-ichi; Koyama, Kinji.
The aim of this work is to assess the collapsed neutron and gamma multigroup cross sections for two dimensional discrete ordinate transport code. Two dimensional distributions of neutron flux and gamma ray dose through a 70cm thick and 94cm square iron shield were measured at the fast neutron source reactor ''YAYOI''. The iron shield was placed over the lead reflector in the vertical experimental column surrounded by heavy concrete wall. The detectors used in this experiment were threshold detectors In, Ni, Al, Mg, Fe and Zn, sandwitch resonance detectors Au, W and Co, activation foils Au for neutrons and thermoluminescence detectors for gamma ray dose. The experimental results were compared with the calculated ones by the discrete ordinate transport code ANISN and TWOTRAN. The region-wise, coupled neutron-gamma multigroup cross-sections (100n+20gamma, EURLIB structure) were generated from ENDF/B-IV library for neutrons and POPOP4 library for gamma-ray production cross-sections by using the code system RADHEAT. The effective microscopic neutron cross sections were obtained from the infinite dilution values applying ABBN type self-shielding factors. The gamma ray production multigroup cross-sections were calculated from these effective microscopic neutron cross-sections. For two-dimensional calculations the group constants were collapsed into 10 neutron groups and 3 gamma groups by using ANISN. (auth.)
Energy Technology Data Exchange (ETDEWEB)
Remsak, S [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)
1969-09-15
This paper deals with the affect od spatial interference between fuel elements and its influence on the resonance integral. General relation for correction of shielding dependent on the first collision portability was derived. Usual assumption about totally black resonance absorber was not taken into account and spatial distribution of elements was random. Shielding effect for two rods and self-shielding effect of tube was studied in detail. Analysis of standard Dancoff problem shows that the shielding correction function is dependent on all resonance parameters, temperature, size of rods and their mutual distance and properties of the moderator. Proposed method is based on the assumption of flat fluxes which enable analytical determination of first collision probability. Influence of interference scattering was taken into account afterwards and was not treated directly. System of balance equations was solved by applying IR approximation. In order to enable easier numerical calculation of shielding correction significant simplification of obtained results was done by applying P-L approximation of Bonaloumi. Obtained numerical results show good agreement with values obtained by applying Hellstrand empirical relation for resonance integral. A simple procedure is proposed for redefining some parameters and Asiu equations which enable more precise determination of transparency coefficient of tube element.
Specification of phase 3 benchmark (Hex-Z heterogeneous and burnup calculation)
International Nuclear Information System (INIS)
Kim, Y.I.
2002-01-01
During the second RCM of the IAEA Co-ordinated Research Project Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects the following items were identified as important. Heterogeneity will affect absolute core reactivity. Rod worths could be considerably reduced by heterogeneity effects depending on their detailed design. Heterogeneity effects will affect the resonance self-shielding in the treatment of fuel Doppler, steel Doppler and sodium density effects. However, it was considered more important to concentrate on the sodium density effect in order to reduce the calculational effort required. It was also recognized that burnup effects will have an influence on fuel Doppler and sodium worths. A benchmark for the assessment of heterogeneity effect for Phase 3 was defined. It is to be performed for the Hex-Z model of the reactor only. No calculations will be performed for the R-Z model. For comparison with heterogeneous evaluations, the control rod worth will be calculated at the beginning of the equilibrium cycle, based on the homogeneous model. The definitions of rod raised and rod inserted for SHR are given, using the composition numbers
Resonance analysis and evaluation of the 235U neutron induced cross sections
International Nuclear Information System (INIS)
Leal, L.C.
1990-06-01
Neutron cross sections of fissile nuclei are of considerable interest for the understanding of parameters such as resonance absorption, resonance escape probability, resonance self-shielding,and the dependence of the reactivity on temperature. In the present study, new techniques for the evaluation of the 235 U neutron cross sections are described. The Reich-Moore formalism of the Bayesian computer code SAMMY was used to perform consistent R-matrix multilevel analyses of the selected neutron cross-section data. The Δ 3 -statistics of Dyson and Mehta, along with high-resolution data and the spin-separated fission cross-section data, have provided the possibility of developing a new methodology for the analysis and evaluation of neutron-nucleus cross sections. The results of the analysis consists of a set of resonance parameters which describe the 235 U neutron cross sections up to 500 eV. The set of resonance parameters obtained through a R-matrix analysis are expected to satisfy statistical properties which lead to information on the nuclear structure. The resonance parameters were tested and showed good agreement with the theory. It is expected that the parametrization of the 235 U neutron cross sections obtained in this dissertation represents the current state of art in data as well as in theory and, therefore, can be of direct use in reactor calculations. 44 refs., 21 figs., 8 tabs
Emissions of chromium (VI) from arc welding.
Heung, William; Yun, Myoung-Jin; Chang, Daniel P Y; Green, Peter G; Halm, Chris
2007-02-01
The presence of Cr in the +6 oxidation state (Cr[VI]) is still observed in ambient air samples in California despite steps taken to reduce emissions from plating operations. One known source of emission of Cr(VI) is welding, especially with high Cr-content materials, such as stainless steels. An experimental effort was undertaken to expand and update Cr(VI) emission factors by conducting tests on four types of arc-welding operations: gas-metal arc welding (GMAW), shielded metal arc welding (SMAW), fluxcore arc welding, and pulsed GMAW. Standard American Welding Society hood results were compared with a total enclosure method that permitted isokinetic sampling for particle size-cut measurement, as well as total collection of the aerosol. The fraction of Cr(VI) emitted per unit mass of Cr electrode consumed was determined. Consistent with AP-42 data, initial results indicate that a significant fraction of the total Cr in the aerosol is in the +6 oxidation state. The fraction of Cr(VI) and total aerosol mass produced by the different arc welding methods varies with the type of welding process used. Self-shielded electrodes that do not use a shield gas, for example, SMAW, produce greater amounts of Cr(VI) per unit mass of electrode consumed. The formation of Cr(VI) from standard electrode wires used for welding mild steel was below the method detection limit after eliminating an artifact in the analytical method used.
Hybrid reduced order modeling for assembly calculations
International Nuclear Information System (INIS)
Bang, Youngsuk; Abdel-Khalik, Hany S.; Jessee, Matthew A.; Mertyurek, Ugur
2015-01-01
Highlights: • Reducing computational cost in engineering calculations. • Reduced order modeling algorithm for multi-physics problem like assembly calculation. • Non-intrusive algorithm with random sampling. • Pattern recognition in the components with high sensitive and large variation. - Abstract: While the accuracy of assembly calculations has considerably improved due to the increase in computer power enabling more refined description of the phase space and use of more sophisticated numerical algorithms, the computational cost continues to increase which limits the full utilization of their effectiveness for routine engineering analysis. Reduced order modeling is a mathematical vehicle that scales down the dimensionality of large-scale numerical problems to enable their repeated executions on small computing environment, often available to end users. This is done by capturing the most dominant underlying relationships between the model's inputs and outputs. Previous works demonstrated the use of the reduced order modeling for a single physics code, such as a radiation transport calculation. This manuscript extends those works to coupled code systems as currently employed in assembly calculations. Numerical tests are conducted using realistic SCALE assembly models with resonance self-shielding, neutron transport, and nuclides transmutation/depletion models representing the components of the coupled code system.
LEAD SLOWING DOWN SPECTROSCOPY FOR DIRECT Pu MASS MEASUREMENTS
International Nuclear Information System (INIS)
Ressler, Jennifer J.; Smith, Leon E.; Anderson, Kevin K.
2008-01-01
The direct measurement of Pu in previously irradiated fuel assemblies is a recognized need in the international safeguards community. A suitable technology could support more timely and independent material control and accounting (MC and A) measurements at nuclear fuel storage areas, the head-end of reprocessing facilities, and at the product-end of recycled fuel fabrication. Lead slowing down spectroscopy (LSDS) may be a viable solution for directly measuring not only the mass of 239Pu in fuel assemblies, but also the masses of other fissile isotopes such as 235U and 241Pu. To assess the potential viability of LSDS, an LSDS spectrometer was modeled in MCNP5 and 'virtual assays' of nominal PWR assemblies ranging from 0 to 60 GWd/MTU burnup were completed. Signal extraction methods, including the incorporation of nonlinear fitting to account for self-shielding effects in strong resonance regions, are described. Quantitative estimates of Pu uncertainty are given for simplistic and more realistic fuel isotopic inventories calculated using ORIGEN. A discussion of additional signal-perturbing effects that will be addressed in future work, and potential signal extraction approaches that could improve Pu mass uncertainties, are also discussed
Design of LSDS for Isotopic Fissile Assay in Spent Fuel
International Nuclear Information System (INIS)
Lee, Yongdeok; Park, Changje; Kim, Hodong; Song, Kee Chan
2013-01-01
A future nuclear energy system is being developed at Korea Atomic Energy Research Institute (KAERI), the system involves a Sodium Fast Reactor (SFR) linked with the pyro-process. The pyro-process produces a source material to fabricate a SFR fuel rod. Therefore, an isotopic fissile content assay is very important for fuel rod safety and SFR economics. A new technology for an analysis of isotopic fissile content has been proposed using a lead slowing down spectrometer (LSDS). The new technology has several features for a fissile analysis from spent fuel: direct isotopic fissile assay, no background interference, and no requirement from burnup history information. Several calculations were done on the designed spectrometer geometry: detection sensitivity, neutron energy spectrum analysis, neutron fission characteristics, self shielding analysis, and neutron production mechanism. The spectrum was well organized even at low neutron energy and the threshold fission chamber was a proper choice to get prompt fast fission neutrons. The characteristic fission signature was obtained in slowing down neutron energy from each fissile isotope. Another application of LSDS is for an optimum design of the spent fuel storage, maximization of the burnup credit and provision of the burnup code correction factor. Additionally, an isotopic fissile content assay will contribute to an increase in transparency and credibility for the utilization of spent fuel nuclear material, as internationally demanded
Feasibility studies on large sample neutron activation analysis using a low power research reactor
International Nuclear Information System (INIS)
Gyampo, O.
2008-06-01
Instrumental neutron activation analysis (INAA) using Ghana Research Reactor-1 (GHARR-1) can be directly applied to samples with masses in grams. Samples weights were in the range of 0.5g to 5g. Therefore, the representativity of the sample is improved as well as sensitivity. Irradiation of samples was done using a low power research reactor. The correction for the neutron self-shielding within the sample is determined from measurement of the neutron flux depression just outside the sample. Correction for gamma ray self-attenuation in the sample was performed via linear attenuation coefficients derived from transmission measurements. Quantitative and qualitative analysis of data were done using gamma ray spectrometry (HPGe detector). The results of this study on the possibilities of large sample NAA using a miniature neutron source reactor (MNSR) show clearly that the Ghana Research Reactor-1 (GHARR-1) at the National Nuclear Research Institute (NNRI) can be used for sample analyses up to 5 grams (5g) using the pneumatic transfer systems.
International Nuclear Information System (INIS)
Macku, K.; Jatuff, F.; Murphy, M.F.; Joneja, O.P.; Bischofberger, R.; Chawla, R.
2006-01-01
Different foil activation techniques have been used for measuring spatial distributions of the 63 Cu(n,γ) 64 Cu reaction within two pins of a SVEA-96 Optima2 boiling water reactor fuel assembly, at the critical facility PROTEUS. This reaction is of interest because its 1/v cross-section gives it a good representation of the 235 U fission rate. Initially, radial capture rate profiles were measured with mechanically punched copper foils. More detailed profiles were then determined by using a 0.2 mm copper wire spiral (∼200 μm resolution), as well as 5-, 10-, and 20-ring UV-lithography, electroplating, and molding (UV-LIGA) foils (up to a 100 μm resolution). For azimuthal measurements, apart from manually cut activation foils (into 8 sectors), 8- and 12-sector LIGA foils were used. The highly versatile LIGA foils have the additional advantage of being very easily separated into individual pieces after irradiation without the use of punches or other cutting tools. In order to account for the invasive character of the foil activation techniques, corrections to account for sample perturbations and for self-shielding effects were determined via simplified Monte Carlo (MCNP4C) modeling of the experimental setup. The final results from the various measurements of 63 Cu(n,γ) 64 Cu within-pin distributions have been compared with MCNP computations employing a detailed model of the full SVEA Optima2 fuel assembly
Large transport packages for decommissioning waste
International Nuclear Information System (INIS)
Price, M.S.T.
1988-08-01
This document reports progress on a study of large transport packages for decommissioning waste and is the semi-annual report for the period 1 January - 30 June 1988. The main tasks performed during the period related to the assembly of package design criteria ie those aspects of manufacture, handling, storage, transport and disposal which impose constraints on design. This work was synthesised into a design specification for packages which formed the conclusion of that task and was the entry into the final task - the development of package design concepts. The design specifications, which concentrated on the Industrial Package category of the IAEA Transport Regulations, has been interpreted for the two main concepts (a) a self-shielded package disposed of in its entirety and (b) a package with returnable shielding. Preliminary information has been prepared on the cost of providing the package as well as transport to a repository and disposal. There is considerable uncertainty about the cost of disposal and variations of over a factor of 10 are possible. Under these circumstances there is merit in choosing a design concept which is relatively insensitive to disposal cost variations. The initial results indicate that on these grounds the package with returnable shielding is preferred. (author)
Development of a program for calculating the cells of heavy water
International Nuclear Information System (INIS)
Calabrese, R.; Lerner, A.M.; Notari, C.
1991-01-01
We describe here a methodology to solve the transport equation i cluster-type fuel cells found in PHWR. The general idea is inspired in the English lattice code WIMS-D4 and associated library even if we have introduced innovations both in structure and contents. The different steps involved are the resonant calculation and the subsequent construction of the microscopic self-shielded cross sections for each isotope; the calculation of macroscopic cross sections per material and the condensation to a broader energy structure; the solution of the two dimensional discretized transport equation for the whole cell. The next step is the inclusion of a burn up routine. A program, ALFIN, was written in FORTRAN 77, and prepared in a modular structure. A sample problem is tested and ALFIN results compared to those produce by WIMS-D4. The discrepancies observed are negligible, except for the resonant region where the methods are different and in some aspect WIMS is clearly in error. (author)
The role of turbulence in star formation laws and thresholds
International Nuclear Information System (INIS)
Kraljic, Katarina; Renaud, Florent; Bournaud, Frédéric; Combes, Françoise; Elmegreen, Bruce; Emsellem, Eric; Teyssier, Romain
2014-01-01
The Schmidt-Kennicutt relation links the surface densities of gas to the star formation rate in galaxies. The physical origin of this relation, and in particular its break, i.e., the transition between an inefficient regime at low gas surface densities and a main regime at higher densities, remains debated. Here, we study the physical origin of the star formation relations and breaks in several low-redshift galaxies, from dwarf irregulars to massive spirals. We use numerical simulations representative of the Milky Way and the Large and Small Magellanic Clouds with parsec up to subparsec resolution, and which reproduce the observed star formation relations and the relative variations of the star formation thresholds. We analyze the role of interstellar turbulence, gas cooling, and geometry in drawing these relations at 100 pc scale. We suggest in particular that the existence of a break in the Schmidt-Kennicutt relation could be linked to the transition from subsonic to supersonic turbulence and is independent of self-shielding effects. With this transition being connected to the gas thermal properties and thus to the metallicity, the break is shifted toward high surface densities in metal-poor galaxies, as observed in dwarf galaxies. Our results suggest that together with the collapse of clouds under self-gravity, turbulence (injected at galactic scale) can induce the compression of gas and regulate star formation.
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
Hybrid reduced order modeling for assembly calculations
Energy Technology Data Exchange (ETDEWEB)
Bang, Youngsuk, E-mail: ysbang00@fnctech.com [FNC Technology, Co. Ltd., Yongin-si (Korea, Republic of); Abdel-Khalik, Hany S., E-mail: abdelkhalik@purdue.edu [Purdue University, West Lafayette, IN (United States); Jessee, Matthew A., E-mail: jesseema@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mertyurek, Ugur, E-mail: mertyurek@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN (United States)
2015-12-15
Highlights: • Reducing computational cost in engineering calculations. • Reduced order modeling algorithm for multi-physics problem like assembly calculation. • Non-intrusive algorithm with random sampling. • Pattern recognition in the components with high sensitive and large variation. - Abstract: While the accuracy of assembly calculations has considerably improved due to the increase in computer power enabling more refined description of the phase space and use of more sophisticated numerical algorithms, the computational cost continues to increase which limits the full utilization of their effectiveness for routine engineering analysis. Reduced order modeling is a mathematical vehicle that scales down the dimensionality of large-scale numerical problems to enable their repeated executions on small computing environment, often available to end users. This is done by capturing the most dominant underlying relationships between the model's inputs and outputs. Previous works demonstrated the use of the reduced order modeling for a single physics code, such as a radiation transport calculation. This manuscript extends those works to coupled code systems as currently employed in assembly calculations. Numerical tests are conducted using realistic SCALE assembly models with resonance self-shielding, neutron transport, and nuclides transmutation/depletion models representing the components of the coupled code system.
Hybrid reduced order modeling for assembly calculations
Energy Technology Data Exchange (ETDEWEB)
Bang, Y.; Abdel-Khalik, H. S. [North Carolina State University, Raleigh, NC (United States); Jessee, M. A.; Mertyurek, U. [Oak Ridge National Laboratory, Oak Ridge, TN (United States)
2013-07-01
While the accuracy of assembly calculations has considerably improved due to the increase in computer power enabling more refined description of the phase space and use of more sophisticated numerical algorithms, the computational cost continues to increase which limits the full utilization of their effectiveness for routine engineering analysis. Reduced order modeling is a mathematical vehicle that scales down the dimensionality of large-scale numerical problems to enable their repeated executions on small computing environment, often available to end users. This is done by capturing the most dominant underlying relationships between the model's inputs and outputs. Previous works demonstrated the use of the reduced order modeling for a single physics code, such as a radiation transport calculation. This manuscript extends those works to coupled code systems as currently employed in assembly calculations. Numerical tests are conducted using realistic SCALE assembly models with resonance self-shielding, neutron transport, and nuclides transmutation/depletion models representing the components of the coupled code system. (authors)
Calandrino, R; del Vecchio, A; Parisi, R; Todde, S; De Felice, P; Savi, A; Pepe, A; Mrskova, A
2010-06-01
The aim of this paper is to assess the activation phenomena and to evaluate the risk of external exposure and intake doses for the maintenance staff of two medical cyclotrons. Two self-shielded cyclotrons are currently operating in the facility for the routine production of (11)C and (18)F. Four radiochemistry laboratories are linked to the cyclotrons by means of shielded radioisotope delivery lines. Radiopharmaceuticals are prepared both for the PET Diagnostic Department, where four CT-PET scanners are operating with a mean patient workload of 40 d(-1) and for [(18)F]FDG external distribution, to provide radiopharmaceuticals for other institutions. In spite of the fact that air contamination inside the radiochemistry laboratories during the synthesis represents the largest 'slice of the pie' in the evaluation of annual intake dose, potential contamination due to the activated particulate, generated during cyclotron irradiation by micro-corrosion of targets and other components potentially struck by the proton beam and generated neutrons, should be considered. In this regard, the most plausible long-lived (T(1/2) > 30 d) radioisotopes formed are: (97)Tc, (56)Co, (57)Co, (58)Co, (60)Co, (49)V, (55)Fe, (109)Cd, (65)Zn and (22)Na. The results for the operating personnel survey has revealed only low-level contamination for (65)Zn in one test, together with minor (18)F intake, probably due to the environmental dispersion of the radioisotope during the [(18)F]FDG synthesis.
Energy Technology Data Exchange (ETDEWEB)
Abanades, A., E-mail: abanades@etsii.upm.es [Grupo de Modelizacion de Sistemas Termoenergeticos, ETSII, Universidad Politecnica de Madrid, c/Ramiro de Maeztu, 7, 28040 Madrid (Spain); Alvarez-Velarde, F.; Gonzalez-Romero, E.M. [Centro de Investigaciones Medioambientales y Tecnologicas (CIEMAT), Avda. Complutense, 40, Ed. 17, 28040 Madrid (Spain); Ismailov, K. [Tokyo Institute of Technology, 2-12-1, O-okayama, Meguro-ku, Tokyo 152-8550 (Japan); Lafuente, A. [Grupo de Modelizacion de Sistemas Termoenergeticos, ETSII, Universidad Politecnica de Madrid, c/Ramiro de Maeztu, 7, 28040 Madrid (Spain); Nishihara, K. [Transmutation Section, J-PARC Center, JAEA, Tokai-mura, Ibaraki-ken 319-1195 (Japan); Saito, M. [Tokyo Institute of Technology, 2-12-1, O-okayama, Meguro-ku, Tokyo 152-8550 (Japan); Stanculescu, A. [International Atomic Energy Agency (IAEA), Vienna (Austria); Sugawara, T. [Transmutation Section, J-PARC Center, JAEA, Tokai-mura, Ibaraki-ken 319-1195 (Japan)
2013-01-15
Highlights: Black-Right-Pointing-Pointer TARC experiment benchmark capture rates results. Black-Right-Pointing-Pointer Utilization of updated databases, included ADSLib. Black-Right-Pointing-Pointer Self-shielding effect in reactor design for transmutation. Black-Right-Pointing-Pointer Effect of Lead nuclear data. - Abstract: The design of Accelerator Driven Systems (ADS) requires the development of simulation tools that are able to describe in a realistic way their nuclear performance and transmutation rate capability. In this publication, we present an evaluation of state of the art Monte Carlo design tools to assess their performance concerning transmutation of long-lived fission products. This work, performed under the umbrella of the International Atomic Energy Agency, analyses two important aspects for transmutation systems: moderation on Lead and neutron captures of {sup 99}Tc, {sup 127}I and {sup 129}I. The analysis of the results shows how shielding effects due to the resonances at epithermal energies of these nuclides affects strongly their transmutation rate. The results suggest that some research effort should be undertaken to improve the quality of Iodine nuclear data at epithermal and fast neutron energy to obtain a reliable transmutation estimation.
Prospects for accelerator neutron sources for large volume minerals analysis
International Nuclear Information System (INIS)
Clayton, C.G.; Spackman, R.
1988-01-01
The electron Linac can be regarded as a practical source of thermal neutrons for activation analysis of large volume mineral samples. With a suitable target and moderator, a neutron flux of about 10 10 n/cm/s over 2-3 kg of rock can be generated. The proton Linac gives the possibility of a high neutron yield (> 10 12 n/s) of fast neutrons at selected energies. For the electron Linac, targets of W-U and W-Be are discussed. The advantages and limitations of the system are demonstrated for the analysis of gold in rocks and ores and for platinum in chromitite. These elements were selected as they are most likely to justify an accelerator installation at the present time. Errors due to self shielding in gold particles for thermal neutrons are discussed. The proton Linac is considered for neutrons generated from a lithium target through the 7 Li(p, n) 7 Be reaction. The analysis of gold by fast neutron activation is considered. This approach avoids particle self-absorption and, by appropriate proton energy selection, avoids potentially dominating interfering reactions. The analysis of 235 U in the presence of 238 U and 232 Th is also considered. (author)
Zanchi, Marta G.; Venook, Ross; Pauly, John M.; Scott, Greig C.
2010-01-01
The currents induced in long conductors such as guidewires by the radio frequency (RF) field in magnetic resonance imaging (MRI) are responsible for potentially dangerous heating of surrounding media, such as tissue. This paper presents an optically-coupled system with the potential to quantitatively measure the RF currents induced on these conductors. The system uses a self shielded toroid transducer and active circuitry to modulate a high speed LED transmitter. Plastic fiber guides the light to a photodiode receiver and transimpedance amplifier. System validation included a series of experiments with bare wires that compared wire tip heating by fluoroptic thermometers with the RF current sensor response. Validations were performed on a custom whole body 64 MHz birdcage test platform and on a 1.5T MRI scanner. With this system, a variety of phenomena were demonstrated including cable trap current attenuation, lossy dielectric Q-spoiling and even transverse electromagnetic wave node patterns. This system should find applications in studies of MRI RF safety for interventional devices such as pacemaker leads, and guidewires. In particular, variations of this device could potentially act as a realtime safety monitor during MRI guided interventions. PMID:19758855
An unshielded whole body radioactivity counter for monitoring persons after a radiation accident
International Nuclear Information System (INIS)
Katoch, D.S.; Somasundaram, S.
1979-01-01
An unshielded chair in which the subject sits, holding a 7.6 cm x 7.6 cm NaI (Tl) detector in his lap, was evaluated for monitoring of persons suspected of internal radioactive contamination following a radiation accident. The reduction in different energy bands of the background gamma-ray spectrum due to self-shielding of the subject was studied for two postures, designated ''upright'' and ''folding'' and the data were analysed in a CDC 3600 computer to obtain the best-fit regression equation relating the reduction factor with body weight and height. The response of the counter was evaluated using an in vitro method and the ranges of under/over-estimation of body burden resulting from assumption of partial/uniform distribution of activity were determined. Counting sensitivities were derived for 13 radioisotopes having gamma-ray energies in the range 145 keV-1.46 MeV. The results are presented and discussed. The study shows that this simple system may be used not only in radiation emergencies but also for operational monitoring of radiation workers for a number of radioisotopes of low and medium radiotoxicity. (auth.)
Cost evaluation of radiation vulcanization of natural rubber latex
International Nuclear Information System (INIS)
Makuuchi, K.
2003-01-01
Cost of radiation vulcanized NR latex was evaluated. The plant would be built in an existing dipping factory in an industrial area in a Southeast Asian country. One thousands dry tons of NR latex are vulcanized with a low energy electron accelerator. The electron accelerator is a self-shielding low energy type. The maximum accelerating voltage is 300 kV and the output power is 10 kW. The total construction cost of the plant is $400,000 including electron accelerator and other equipments. Costs of raw materials and utilities are $1.165 and $0.023 per one kg of product, respectively. The fixed costs of the plant consist of labor costs, labor overhead, maintenance, plant overhead, depreciation, and bank interest. It is $0.190/kg of product. The company overhead for operation including company management, R and D and insurance is $0.044/kg of product. Thus, the total production cost is estimated to be $1.422/kg of product. (author)
Measurements and applications of neutron multiple scattering in resonance region
International Nuclear Information System (INIS)
Ohkubo, Makio
1977-02-01
Capture yield of neutrons impinging on a thick material is complicated due to self-shielding and multiple scattering, especially in the resonance region. When the incident neutron energy is equal to a resonance energy of the material, capture probability of the neutron increases with sample thickness and reaches a saturation value P sub(CO). There is a simple relation between P sub(CO) and GAMMA sub(n)/GAMMA and the recoil energy by the Monte-Carlo calculation. To examine validity of the relation, P sub(CO) was measured for 19 resonances in 12 nuclides with thick samples, using a JAERI linac time-of-flight spectrometer with Moxon-Rae type gamma ray detector and transmission type neutron flux monitor. Results of the measurements confirmed the validity. With this relation, the GAMMA sub(n)/GAMMA or GAMMA sub(γ)/GAMMA value can be obtained from the measured P sub(CO), and also the level spins be determined by combining the transmission data. Because of the definition of P sub(CO), determination of the resonance parameters is not sensitive to the sample thickness as far as it is sufficiently thick. (auth.)
Energy Technology Data Exchange (ETDEWEB)
Kim, Min Jung; Go, A Ra; Kim, Kwang Pyo [Kyung Hee University, Yongin (Korea, Republic of)
2016-09-15
The object of this study was to evaluate the separation distance from a temporary storage facility satisfying the dose criteria. The calculation of ambient dose rates took into account cover soil thickness, facility size, and facility type by using MCNPX code. Shielding effects of cover soil were 68.9%, 96.9% and 99.7% at 10 cm, 30 cm and 50 cm respectively. The on-ground type of storage facility had the highest ambient dose rate, followed by the semi-ground type and the underground type. The ambient dose rate did not vary with facility size (except 5 × 5 × 2 m size) due to the self-shielding of decontamination waste in temporary storage. The separation distances without cover soil for a 50 × 50 × 2 m size facility were evaluated as 14 m (minimum radioactivity concentration), 33 m (most probably radioactivity concentration), and 57 m (maximum radioactivity concentration) for on-ground storage type, 9 m, 24 m, and 45 m for semi-underground storage type, and 6 m, 16 m, and 31 m for underground storage type.
Sadeghi, Mohammad Hosein; Sina, Sedigheh; Mehdizadeh, Amir; Faghihi, Reza; Moharramzadeh, Vahed; Meigooni, Ali Soleimani
2018-02-01
The dosimetry procedure by simple superposition accounts only for the self-shielding of the source and does not take into account the attenuation of photons by the applicators. The purpose of this investigation is an estimation of the effects of the tandem and ovoid applicator on dose distribution inside the phantom by MCNP5 Monte Carlo simulations. In this study, the superposition method is used for obtaining the dose distribution in the phantom without using the applicator for a typical gynecological brachytherapy (superposition-1). Then, the sources are simulated inside the tandem and ovoid applicator to identify the effect of applicator attenuation (superposition-2), and the dose at points A, B, bladder, and rectum were compared with the results of superposition. The exact dwell positions, times of the source, and positions of the dosimetry points were determined in images of a patient and treatment data of an adult woman patient from a cancer center. The MCNP5 Monte Carlo (MC) code was used for simulation of the phantoms, applicators, and the sources. The results of this study showed no significant differences between the results of superposition method and the MC simulations for different dosimetry points. The difference in all important dosimetry points was found to be less than 5%. According to the results, applicator attenuation has no significant effect on the calculated points dose, the superposition method, adding the dose of each source obtained by the MC simulation, can estimate the dose to points A, B, bladder, and rectum with good accuracy.
Development of irradiation capabilities to address the challenges of the nuclear industry
Energy Technology Data Exchange (ETDEWEB)
Leay, L. [The University of Manchester, Dalton Cumbrian Facility, Westlakes Science and Technology Park, Moor Row, Cumbria CA24 3HA (United Kingdom); The University of Manchester, School of Electrical and Electronic Engineering, Oxford Road, Manchester M13 9PL (United Kingdom); Bower, W. [The University of Manchester, Dalton Cumbrian Facility, Westlakes Science and Technology Park, Moor Row, Cumbria CA24 3HA (United Kingdom); The University of Manchester, School of Earth, Atmospheric and Environmental Sciences, Oxford Road, Manchester M13 9PL (United Kingdom); Horne, G.; Wady, P.; Baidak, A. [The University of Manchester, Dalton Cumbrian Facility, Westlakes Science and Technology Park, Moor Row, Cumbria CA24 3HA (United Kingdom); The University of Manchester, School of Chemistry, Oxford Road, Manchester M13 9PL (United Kingdom); Pottinger, M. [The University of Manchester, Dalton Cumbrian Facility, Westlakes Science and Technology Park, Moor Row, Cumbria CA24 3HA (United Kingdom); The University of Manchester, School of Electrical and Electronic Engineering, Oxford Road, Manchester M13 9PL (United Kingdom); Nancekievill, M. [The University of Manchester, School of Electrical and Electronic Engineering, Oxford Road, Manchester M13 9PL (United Kingdom); Smith, A.D. [The University of Manchester, Dalton Cumbrian Facility, Westlakes Science and Technology Park, Moor Row, Cumbria CA24 3HA (United Kingdom); Watson, S. [The University of Manchester, Dalton Cumbrian Facility, Westlakes Science and Technology Park, Moor Row, Cumbria CA24 3HA (United Kingdom); The University of Manchester, School of Electrical and Electronic Engineering, Oxford Road, Manchester M13 9PL (United Kingdom); Green, P.R. [The University of Manchester, School of Electrical and Electronic Engineering, Oxford Road, Manchester M13 9PL (United Kingdom); and others
2015-01-15
With the announcement of the U.K. new nuclear build and the requirement to decommission old facilities, researchers require bespoke facilities to undertake experiments to inform decision making. This paper describes development of The University of Manchester’s Dalton Cumbrian Facility, a custom built research environment which incorporates a 5 MV tandem ion accelerator as well as a self-shielded {sup 60}Co irradiator. The ion accelerator allows the investigation into the radiolytic consequences of various charged particles, including protons, alpha particles and a variety of heavier (metal and nonmetal) ions, while the {sup 60}Co irradiator allows the effects of gamma radiation to be studied. Some examples of work carried out at the facility are presented to demonstrate how this equipment can improve our mechanistic understanding of various aspects of the deleterious effects of radiation in the nuclear industry. These examples include applications in waste storage and reprocessing as well as geological storage and novel surveying techniques. The outlook for future research is also discussed.
Calibration of spent fuel measurement assembly
International Nuclear Information System (INIS)
Koleska, Michal; Viererbl, Ladislav; Marek, Milan
2014-01-01
The LVR-15 research reactor (Czech Republic) had been converted from the highly enriched IRT-2M to the low enriched IRT-4M fuel. For the possibility of the independent pre-transport evaluation of IRT-2M burnup, a spectrometric system was developed. This spectrometric system consists of the fuel holder, the collimator and the portable Canberra Big MAC HPGe (High Purity Germanium) detector. In order to have well reproducible and reliable experimental data for modeling of the measurement system, calibration with the 110m Ag isotope with known activity was performed. This isotope was chosen for having energies similar to isotopes measured in fuel assemblies. The 110m Ag isotope was prepared by irradiating of the silver foil in LVR-15 research reactor; its activity was evaluated in the LVR-15's spectrometric laboratory. From the measured data, an efficiency curve of the spectrometric system has been determined. The experimental data were compared to the calculation results with the MCNPX model of the spectrometric system. - Highlights: • Calibration of research reactor spent fuel measurement assembly. • On-site prepared 110m Ag isotope used for the measurement. • Calculated self-shielding factor for the IRT-2M fuel. • Applicable to other research reactor fuel geometries
Novel approach in k0-NAA for highly concentrated REE Samples.
Abdollahi Neisiani, M; Latifi, M; Chaouki, J; Chilian, C
2018-04-01
The present paper presents a new approach for k 0 -NAA for accurate quantification with short turnaround analysis times for rare earth elements (REEs) in high content mineral matrices. REE k 0 and Q 0 values, spectral interferences and nuclear interferences were experimentally evaluated and improved with Alfa Aesar Specpure Plasma Standard 1000mgkg -1 mono-rare earth solutions. The new iterative gamma-ray self-attenuation and neutron self-shielding methods were investigated with powder standards prepared from 100mg of 99.9% Alfa Aesar mono rare earth oxide diluted with silica oxide. The overall performance of the new k 0 -NAA method for REEs was validated using a certified reference material (CRM) from Canadian Certified Reference Materials Project (REE-2) with REE content ranging from 7.2mgkg -1 for Yb to 9610mgkg -1 for Ce. The REE concentration was determined with uncertainty below 7% (at 95% confidence level) and proved good consistency with the CRM certified concentrations. Copyright © 2017 Elsevier B.V. All rights reserved.
Pescarini, Massimo; Sinitsa, Valentin; Orsi, Roberto; Frisoni, Manuela
2016-02-01
Two broad-group coupled neutron/photon working cross section libraries in FIDO-ANISN format, dedicated to LWR shielding and pressure vessel dosimetry applications, were generated following the methodology recommended by the US ANSI/ANS-6.1.2-1999 (R2009) standard. These libraries, named BUGJEFF311.BOLIB and BUGENDF70.BOLIB, are respectively based on JEFF-3.1.1 and ENDF/B-VII.0 nuclear data and adopt the same broad-group energy structure (47 n + 20 γ) of the ORNL BUGLE-96 similar library. They were respectively obtained from the ENEA-Bologna VITJEFF311.BOLIB and VITENDF70.BOLIB libraries in AMPX format for nuclear fission applications through problem-dependent cross section collapsing with the ENEA-Bologna 2007 revision of the ORNL SCAMPI nuclear data processing system. Both previous libraries are based on the Bondarenko self-shielding factor method and have the same AMPX format and fine-group energy structure (199 n + 42 γ) as the ORNL VITAMIN-B6 similar library from which BUGLE-96 was obtained at ORNL. A synthesis of a preliminary validation of the cited BUGLE-type libraries, performed through 3D fixed source transport calculations with the ORNL TORT-3.2 SN code, is included. The calculations were dedicated to the PCA-Replica 12/13 and VENUS-3 engineering neutron shielding benchmark experiments, specifically conceived to test the accuracy of nuclear data and transport codes in LWR shielding and radiation damage analyses.
International Nuclear Information System (INIS)
2015-01-01
The objective of this Safety Guide is to provide recommendations on how to meet the requirements of the BSS with regard to irradiation facilities. This Safety Guide provides specific, practical recommendations on the safe design and operation of gamma, electron and X ray irradiators for use by operating organizations and the designers of these facilities, and by regulatory bodies. SCOPE. The facilities considered in this publication include five types of irradiator, whether operated on a commercial basis or for research and development purposes. This publication is concerned with radiation safety issues and not with the uses of irradiators, nor does it cover the irradiation of product or its quality management. The five types of irradiator are: - Panoramic dry source storage irradiators; - Underwater irradiators, in which both the source and the product being irradiated are under water; - Panoramic wet source storage irradiators; - Electron beam irradiation facilities, in which irradiation is performed in an area that is potentially accessible to personnel, but that is kept inaccessible during the irradiation process; - X ray irradiation facilities, in which irradiation is performed in an area that is potentially accessible to personnel, but that is kept inaccessible during the irradiation process. Consideration of non-radiation-related risks and of the benefits resulting from the operation of irradiators is outside the scope of this Safety Guide. The practices of radiotherapy and radiography are also outside the scope of this Safety Guide. Category I gamma irradiators (i.e. 'self-shielded' irradiators) are outside the scope of this Safety Guide
Industrial potential for application of radiation curing in Pakistan
International Nuclear Information System (INIS)
Ahmed, S.
1991-01-01
Potential applications of radiation curing of coating are in the field of wood and wood products, drying of printing inks, ceramics (roof and floor tiles) and textiles. Pakistan a 'timber deficit' country needs to improve her wood, plywood, hardboard and particle board to make for shortage of quality wood. Imports of wood and wood products are in excess of 3000 million rupees. Radiation curing can be applied and itexcels over heat treatment. Whereas costs of high energy units (500 KeV) with scanning type are rather high, low energy (100-175 KeV) flat beam self-shielded units costing 200,000 US$ are available. For developing countries ultraviolet (UV) curing is ideally suited because of its low price, flexibility and simplicity in handling. Alternately, multipurpose bunker type facility such as 500 KeV current mA can be utilized in carrying out heat-shrinkables production, irradiation of cable and wire and curing of coatings on wood and wood products. (author)
New laser technique revives old ideas for thermoluminescence neutron dosimetry
International Nuclear Information System (INIS)
Braeunlich, P.; Brown, M.; Gasiot, J.; Fillard, J.P.
1982-01-01
Laser heating is discussed as a means to evaluate thermoluminescence dosimeters in neutron dosimetry. Direct energy coupling from the photon beam to the phonons of the TL material permits heating of thin layers with rates of temperature increase exceeding 10 4 Ks - 1 . Rapid TLD evaluation will allow the design of dosimetry badges containing a number of different small thin film TLD elements in various orientations and behind appropriate filters, hydrogenous radiators, etc. Desired redundance is readily possible by using back-up TLDs for every specific task. Reading occurs with a scanning laser beam rather than by mechanically manipulating the TLD toward a fixed heat source. Improvements in the signal-to-noise ratio of up to a factor of 1000 are readily obtained. Thus, sensitive thin-film TLDs can be designed with negligible self-shielding for thermal neutrons in albedo applications and with known, nearly energy dependent cavity correction factors for dosimetry in mixed n-#betta# fields. Due to the greatly increased sensitivity possible with fast laser heating, significant advances are expected in the fast neutron dosimetry techniques which are based on hydrogeneous proton radiators or LET-dependent slow peak formation
The physics of nuclear reactors
Marguet, Serge
2017-01-01
This comprehensive volume offers readers a progressive and highly detailed introduction to the complex behavior of neutrons in general, and in the context of nuclear power generation. A compendium and handbook for nuclear engineers, a source of teaching material for academic lecturers as well as a graduate text for advanced students and other non-experts wishing to enter this field, it is based on the author’s teaching and research experience and his recognized expertise in nuclear safety. After recapping a number of points in nuclear physics, placing the theoretical notions in their historical context, the book successively reveals the latest quantitative theories concerning: • The slowing-down of neutrons in matter • The charged particles and electromagnetic rays • The calculation scheme, especially the simplification hypothesis • The concept of criticality based on chain reactions • The theory of homogeneous and heterogeneous reactors • The problem of self-shielding �...
Integrated Box Interrogation System (IBIS) Preliminary Design Study
Energy Technology Data Exchange (ETDEWEB)
DR. Stephen Croft; Mr. David Martancik; Dr. Brian Young; Dr. Patrick MJ Chard; Dr. Robert J Estop; Sheila Melton; Gaetano J. Arnone
2003-01-13
Canberra Industries has won the tendered solicitation, INEEL/EST-99-00121 for boxed waste Nondestructive Assay Development and Demonstration. Canberra will provide the Integrated Box Interrogation System (IBIS) which is a suite of assay instrumentation and a data reduction system that addresses the measurement needs for Boxed Wastes identified in the solicitation and facilitates the associated experimental program and demonstration of system capability. The IBIS system will consist of the next generation CWAM system, i.e. CWAM II, which is a Scanning Passive/Active Neutron interrogation system which we will call a Box Segmented Neutron Scanner (BSNS), combined with a physically separate Box Segmented Gamma-ray Scanning (BSGS) system. These systems are based on existing hardware designs but will be tailored to the large sample size and enhanced to allow the program to evaluate the following measurement criteria:Characterization and correction for matrix heterogeneity Characterization of non-uniform radio-nuclide and isotopic compositions Assay of high density matrices (both high-Z and high moderator contents)Correction for radioactive material physical form - such as self shielding or multiplication effects due to large accumulations of radioactive materials.Calibration with a minimal set of reference standards and representative matrices.THis document summarizes the conceptual design parameters of the IBIS and indicates areas key to the success of the project where development is to be centered. The work presented here is a collaborative effort between scientific staff within Canberra and within the NIS-6 group at LANL.
Measurement of mercury isotopic ratio in stone meteorites by neutron activation analysis
International Nuclear Information System (INIS)
Thakur, A.N.
1997-01-01
196 Hg and 202 Hg isotopes have been measured by neutron activation analysis in samples of twelve stone meteorites. Hg is extracted from an irradiated sample by stepwise heating. The mercury concentrations vary from 0.07 to 33 ppm. While most of the samples give 196 Hg/ 202 Hg ratios similar to terrestrial value within error limits, in some cases large anomalies are observed. A number of control experiments have been devised, that show the absence of experimental artifacts, during sample preparation, neutron irradiation, chemical separation and counting stages. Several anomalous and normal Hg distillate have been re-irradiated as Hg-diethyl-dithio-carbamate complex to eliminate the influence of neutron self shielding and interfering reactions from matrix elements. The isotopic ratio patterns persist in the distillates too proving that any artifacts during meteorite irradiation and measurement are essentially absent. Both positive and negative anomalies are observed: however, the negative anomalies are much more frequent and abundant. In an extreme case of fine grained magnetic particles of Ambapur Nagla the 196 Hg is apparently absent in the Hg released at 100 deg C. A 2σ 196 Hg/ 202 Hg value is only 6% relative to the monitor. This experiment shows the robustness of neutron activation analysis and suggest some constrains on the formation history of stone meteorites. (author)
Integrated Box Interrogation System (IBIS) Preliminary Design Study
International Nuclear Information System (INIS)
Croft, Stephen; Martancik, David; Young, Brian; Chard MJ, Patrick; Estop J, Robert; Sheila Melton; Arnone, Gaetano J.
2003-01-01
Canberra Industries has won the tendered solicitation, INEEL/EST-99-00121 for boxed waste Nondestructive Assay Development and Demonstration. Canberra will provide the Integrated Box Interrogation System (IBIS) which is a suite of assay instrumentation and a data reduction system that addresses the measurement needs for Boxed Wastes identified in the solicitation and facilitates the associated experimental program and demonstration of system capability. The IBIS system will consist of the next generation CWAM system, i.e. CWAM II, which is a Scanning Passive/Active Neutron interrogation system which we will call a Box Segmented Neutron Scanner (BSNS), combined with a physically separate Box Segmented Gamma-ray Scanning (BSGS) system. These systems are based on existing hardware designs but will be tailored to the large sample size and enhanced to allow the program to evaluate the following measurement criteria:Characterization and correction for matrix heterogeneity Characterization of non-uniform radio-nuclide and isotopic compositions Assay of high density matrices (both high-Z and high moderator contents)Correction for radioactive material physical form - such as self shielding or multiplication effects due to large accumulations of radioactive materials.Calibration with a minimal set of reference standards and representative matrices.THis document summarizes the conceptual design parameters of the IBIS and indicates areas key to the success of the project where development is to be centered. The work presented here is a collaborative effort between scientific staff within Canberra and within the NIS-6 group at LANL
Simulation of gamma-ray irradiation of lettuce leaves in a 137Cs irradiator using MCNP
International Nuclear Information System (INIS)
Kim, Jongsoon; Moreira, Rosana G.; Braby, Leslie A.
2010-01-01
Ionizing radiation effectively reduces the number of common microbial pathogens in fresh produce. However, the efficacy of the process for pathogens internalized into produce tissue is unknown. The objective of this study was to understand gamma irradiation of lettuce leaf structure exposed in a 137 Cs irradiator using MCNP. The simulated 137 Cs irradiator is a self-shielded device, and its geometry and sources are described in the MCNP input file. When the irradiation chamber is filled with water, lower doses are found at the center of the irradiation volume and the dose uniformity ratio (maximum dose/minimum dose) is 1.76. For randomly oriented rectangular lettuce leaf segments in the irradiation chamber, the dose uniformity ratio is 1.25. It shows that dose uniformity in the Cs irradiator is strongly dependent of the density of the sample. To understand dose distribution inside the leaf, we divided a lettuce leaf into a low density (flat) region (0.72 g/cm 3 ) and high density (rib) region (0.86 g/cm 3 ). Calculated doses to the rib are 61% higher than doses to the flat region of the leaf. This indicates that internalized microorganisms can be inactivated more easily than organisms on the surface. This study shows that irradiation can effectively reduce viable microorganism internalized in lettuce. (author)
Advances in X-Band and S-Band Linear Accelerators for Security, NDT, and Other Applications
Mishin, Andrey V
2005-01-01
At AS&E High Energy Systems Division, we designed several new advanced high energy electron beam and X-ray sources. Our primary focus has always been in building the world's most portable commercial X-band accelerators. Today, our X-band systems frequently exceed performance of the similar S-band machines, while they are more portable compared to the latter. The new designs of the X-band accelerators in the most practical energy range from 1 MeV to 6 MeV have been tested delivering outstanding results. Seventy 6 MeV X-band linacs systems have been produced. The most compact linac for security is used by AS&E in a self-shielded, Shaped Energy cargo screening system. We pioneered using the X-band linear accelerators for CT, producing high quality images of oil pipes and wood logs. An X-band linear accelerator head on a robotic arm has been used for electron beam radiation curing of an odd-shaped graphite composite part. We developed the broad-range 4 MeV to over 10 MeV energy-regulated X-band ...
Large sample neutron activation analysis of a reference inhomogeneous sample
International Nuclear Information System (INIS)
Vasilopoulou, T.; Athens National Technical University, Athens; Tzika, F.; Stamatelatos, I.E.; Koster-Ammerlaan, M.J.J.
2011-01-01
A benchmark experiment was performed for Neutron Activation Analysis (NAA) of a large inhomogeneous sample. The reference sample was developed in-house and consisted of SiO 2 matrix and an Al-Zn alloy 'inhomogeneity' body. Monte Carlo simulations were employed to derive appropriate correction factors for neutron self-shielding during irradiation as well as self-attenuation of gamma rays and sample geometry during counting. The large sample neutron activation analysis (LSNAA) results were compared against reference values and the trueness of the technique was evaluated. An agreement within ±10% was observed between LSNAA and reference elemental mass values, for all matrix and inhomogeneity elements except Samarium, provided that the inhomogeneity body was fully simulated. However, in cases that the inhomogeneity was treated as not known, the results showed a reasonable agreement for most matrix elements, while large discrepancies were observed for the inhomogeneity elements. This study provided a quantification of the uncertainties associated with inhomogeneity in large sample analysis and contributed to the identification of the needs for future development of LSNAA facilities for analysis of inhomogeneous samples. (author)
Contribution to the study of the unresolved resonance range of the neutrons cross sections
International Nuclear Information System (INIS)
Noguere, Gilles
2014-01-01
This document presents the statistical description of neutron cross sections in the unresolved resonance range. The modeling of the total cross section and of the 'shape - elastic' cross section is based on the 'average R-Matrix' formalism. The partial cross sections describing the radiative capture, elastic scattering, inelastic scattering and fission process are calculated using the Hauser-Feshbach formalism with width fluctuation corrections. In the unresolved resonance range, these models depend on the average resonance parameters (neutron strength function Sc, mean level spacing D c , average partial reaction widths Γ c , channel radius a c , effective radius R' and distant level parameter R-bar c ∞ ). The codes (NJOY, CALENDF...) dedicated to the processing of nuclear data libraries (JEFF, ENDF/B, JENDL, CENDL, BROND... ) use the average parameters to take into account the self-shielding phenomenon for the simulation of the neutron transport in Monte-Carlo (MCNP, TRIPOLI... ) and deterministic (APOLLO, ERANOS...) codes. The evaluation work consists in establishing a consistent set of average parameters as a function of the total angular momentum J of the system and of the orbital moment of the incident neutron l. The work presented in this paper aims to describe the links between the S-Matrix and the 'average R-Matrix' formalism for the calculation of Sc, R-bar c ∞ , ac and R'. (author) [fr
International Nuclear Information System (INIS)
Hattori, Takatoshi
2002-01-01
A New technique of precise and automatic measurements of gamma emitters in metal waste has been developed using 3D non-contact shape measurement and monte-carlo calculation techniques in order to confirm that specific radioactivity level of metal waste satisfies the clearance level and furthermore the surface contamination level of the metal waste is below the legal standard level. The technique can give a calibration factor every measurement target automatically and realize an automatic correction for reduction of background count rate in gamma measurements due to self-shield effect of the measurement target. The accuracy of the present method has been made clear using mock-metal wastes with various types of shape, number and size. Assuming the goal of the detection limit for practical use is 25OBq in radioactivity, a concept of the practical gamma monitor has been designed so as to be able to confirm both the clearance level and surface contamination level simultaneously and to cope with the metal waste at a speed of 2-10 ton a day. (author)
THE PRESSURE OF THE STAR-FORMING INTERSTELLAR MEDIUM IN COSMOLOGICAL SIMULATIONS
International Nuclear Information System (INIS)
Munshi, Ferah; Quinn, Thomas R.; Governato, Fabio; Christensen, Charlotte; Wadsley, James; Loebman, Sarah; Shen, Sijing
2014-01-01
We examine the pressure of the star-forming interstellar medium (ISM) of Milky-Way-sized disk galaxies using fully cosmological SPH+N-body, high-resolution simulations. These simulations include explicit treatment of metal-line cooling in addition to dust and self-shielding, H 2 -based star formation. The four simulated halos have masses ranging from a few times 10 10 to nearly 10 12 solar masses. Using a kinematic decomposition of these galaxies into present-day bulge and disk components, we find that the typical pressure of the star-forming ISM in the present-day bulge is higher than that in the present-day disk by an order of magnitude. We also find that the pressure of the star-forming ISM at high redshift is, on average, higher than ISM pressures at low redshift. This explains why the bulge forms at higher pressures: the disk assembles at lower redshift when the ISM exhibits lower pressure and the bulge forms at high redshift when the ISM has higher pressure. If ISM pressure and IMF variation are tied together, these results could indicate a time-dependent IMF in Milky-Way-like systems as well as a different IMF in the bulge and the disk
Evaluation of Uncertainties in the Determination of Phosphorus by RNAA
International Nuclear Information System (INIS)
Rick L. Paul
2000-01-01
A radiochemical neutron activation analysis (RNAA) procedure for the determination of phosphorus in metals and other materials has been developed and critically evaluated. Uncertainties evaluated as type A include those arising from measurement replication, yield determination, neutron self-shielding, irradiation geometry, measurement of the quantity for concentration normalization (sample mass, area, etc.), and analysis of standards. Uncertainties evaluated as type B include those arising from beta contamination corrections, beta decay curve fitting, and beta self-absorption corrections. The evaluation of uncertainties in the determination of phosphorus is illustrated for three different materials in Table I. The metal standard reference materials (SRMs) 2175 and 861 were analyzed for value assignment of phosphorus; implanted silicon was analyzed to evaluate the technique for certification of phosphorus. The most significant difference in the error evaluation of the three materials lies in the type B uncertainties. The relatively uncomplicated matrix of the high-purity silicon allows virtually complete purification of phosphorus from other beta emitters; hence, minimal contamination correction is needed. Furthermore, because the chemistry is less rigorous, the carrier yield is more reproducible, and self-absorption corrections are less significant. Improvements in the chemical purification procedures for phosphorus in complex matrices will decrease the type B uncertainties for all samples. Uncertainties in the determination of carrier yield, the most significant type A error in the analysis of the silicon, also need to be evaluated more rigorously and minimized in the future
Soenksen, L R; Kassis, T; Noh, M; Griffith, L G; Trumper, D L
2018-03-13
Precise fluid height sensing in open-channel microfluidics has long been a desirable feature for a wide range of applications. However, performing accurate measurements of the fluid level in small-scale reservoirs (sensor contact needs to be avoided. In particular, gravity-driven systems used in several microfluidic applications to establish pressure gradients and impose flow remain open-loop and largely unmonitored due to these sensing limitations. Here we present an optimized self-shielded coplanar capacitive sensor design and automated control system to provide submillimeter fluid-height resolution (∼250 μm) and control of small-scale open reservoirs without the need for direct fluid contact. Results from testing and validation of our optimized sensor and system also suggest that accurate fluid height information can be used to robustly characterize, calibrate and dynamically control a range of microfluidic systems with complex pumping mechanisms, even in cell culture conditions. Capacitive sensing technology provides a scalable and cost-effective way to enable continuous monitoring and closed-loop feedback control of fluid volumes in small-scale gravity-dominated wells in a variety of microfluidic applications.
The role of turbulence in star formation laws and thresholds
Energy Technology Data Exchange (ETDEWEB)
Kraljic, Katarina; Renaud, Florent; Bournaud, Frédéric [CEA, IRFU, SAp, F-91191 Gif-sur-Yvette Cedex (France); Combes, Françoise [Observatoire de Paris, LERMA et CNRS, 61 Av de l' Observatoire, F-75014 Paris (France); Elmegreen, Bruce [IBM T. J. Watson Research Center, 1101 Kitchawan Road, Yorktown Heights, NY 10598 (United States); Emsellem, Eric [European Southern Observatory, D-85748 Garching bei Muenchen (Germany); Teyssier, Romain [Institute for Theoretical Physics, University of Zürich, CH-8057 Zürich (Switzerland)
2014-04-01
The Schmidt-Kennicutt relation links the surface densities of gas to the star formation rate in galaxies. The physical origin of this relation, and in particular its break, i.e., the transition between an inefficient regime at low gas surface densities and a main regime at higher densities, remains debated. Here, we study the physical origin of the star formation relations and breaks in several low-redshift galaxies, from dwarf irregulars to massive spirals. We use numerical simulations representative of the Milky Way and the Large and Small Magellanic Clouds with parsec up to subparsec resolution, and which reproduce the observed star formation relations and the relative variations of the star formation thresholds. We analyze the role of interstellar turbulence, gas cooling, and geometry in drawing these relations at 100 pc scale. We suggest in particular that the existence of a break in the Schmidt-Kennicutt relation could be linked to the transition from subsonic to supersonic turbulence and is independent of self-shielding effects. With this transition being connected to the gas thermal properties and thus to the metallicity, the break is shifted toward high surface densities in metal-poor galaxies, as observed in dwarf galaxies. Our results suggest that together with the collapse of clouds under self-gravity, turbulence (injected at galactic scale) can induce the compression of gas and regulate star formation.
Criticality studies of fast assemblies with the new 27-group cross-section set
International Nuclear Information System (INIS)
Garg, S.B.; Shukla, V.K.
1976-01-01
A test of 27-group cross-section set (Garg-set) recently derived from ENDF/B library has been carried out in the criticality studies of the Pu 239 , U 235 and U 233 based metal, oxide and carbide fuelled fast critical assemblies. A total of twenty fast critical assemblies of different sizes and varying neutron spectra have been selected for analysis. Based on these analyses it has been observed that the Garg-set predicts well the criticality of uranium and plutonium based hard-spectra assemblies. In the soft-spectra systems it underpredicts criticality because of the following reasons: (a) It makes use of the higher capture cross-sections of structural and coolant elements given in ENDF/B - Version IV library. (b) It does not account for the resonance self-shielding effects of cross-sections. It has also been observed that the Garg-set gives better results than the MABBN-set for dense and dilute plutonium-based and the hard uranium-based assemblies. This superior trend of the Garg-set is slightly lost in the uranium-based dilute systems because of large differences in the capture cross-sections of structural elements of these two sets. (author)
Medical Isotope Production Analyses In KIPT Neutron Source Facility
International Nuclear Information System (INIS)
Talamo, Alberto; Gohar, Yousry
2016-01-01
Medical isotope production analyses in Kharkov Institute of Physics and Technology (KIPT) neutron source facility were performed to include the details of the irradiation cassette and the self-shielding effect. An updated detailed model of the facility was used for the analyses. The facility consists of an accelerator-driven system (ADS), which has a subcritical assembly using low-enriched uranium fuel elements with a beryllium-graphite reflector. The beryllium assemblies of the reflector have the same outer geometry as the fuel elements, which permits loading the subcritical assembly with different number of fuel elements without impacting the reflector performance. The subcritical assembly is driven by an external neutron source generated from the interaction of 100-kW electron beam with a tungsten target. The facility construction was completed at the end of 2015, and it is planned to start the operation during the year of 2016. It is the first ADS in the world, which has a coolant system for removing the generated fission power. Argonne National Laboratory has developed the design concept and performed extensive design analyses for the facility including its utilization for the production of different radioactive medical isotopes. 99 Mo is the parent isotope of 99m Tc, which is the most commonly used medical radioactive isotope. Detailed analyses were performed to define the optimal sample irradiation location and the generated activity, for several radioactive medical isotopes, as a function of the irradiation time.
Analysis of Homogeneous BFS-73-1 MA Benchmark Core
Energy Technology Data Exchange (ETDEWEB)
Kim, Yeong Il; Yoo, Jae Woon; Song, Hoon; Jang, Jin Wook; Kim, Yeong Il
2007-06-15
Analysis of BFS-73-1 critical assembly for MA transmutation has been carried out by using K-CORE system mainly, DIF3D code. All of measured data are compared with the results of analysis and sensitiveness of calculation conditions, for example, number of neutron energy groups, mesh size used, and analysis method, are assessed. Effective multiplication factor was in good agreement within experimental uncertainty in both transport and diffusion calculations. Fission rate distribution of U-235 and U-238 is also fairly good agreed with experimental results within maximum 5% in core region. But large discrepancy was seen in blanket region and it tends to increase as the location closes to core boundary. Largest error of relative reaction rate ratio was seen in Am-243 fission and U-238 capture. For the case of Am-243, the error lay on appropriate range considering the measurement uncertainty of that as 4.6%. Sample reactivity worths for scattering dominant isotope was greatly differ from the experimental results, which can be explained in terms of sample heterogeneity effect, sample self shielding and finally resonance bilinear correction effect. These effects will be evaluated as future study. C/E of effective delayed neutron fraction is within 4%, which is within the measurement uncertainty.
Energy Technology Data Exchange (ETDEWEB)
Ivkovic, M; Zdravkovic, Z; Sotic, O [Department of Reactor Physics and Dynamics, Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)
1966-04-15
A graphite block was calibrated for the thermal neutron flux of the Ra-Be source using indium foils as detectors. Experimental values of the thermal neutron flux along the central vertical axis of the system were corrected for the self-shielding effect and depression of flux in the detector. The experimental values obtained were compared with the values calculated on the basis of solving the conservation neutron equation by the continuous slowing-down theory. In this theoretical calculation of the flux the Ra-Be source was divided into three resonance energy regions. The measurement of the thermal neutron diffusion length in the standard graphite block is described. The measurements were performed in the thermal neutron region of the system. The experimental results were interpreted by the diffusion theory for point thermal neutron source in the finite system. The thermal neutron diffusion length was calculated to be L= 50.9 {+-}3.1 cm for the following graphite characteristics: density = 1.7 g/cm{sup 3}; boron content = 0.1 ppm; absorption cross section = 3.7 mb.
A closed cycle-cryostat for high-field Moessbauer spectroscopy
International Nuclear Information System (INIS)
Janoschka, A; Schuenemann, V; Svenconis, G
2010-01-01
A closed cycle-cryostat coupled to a Moessbauer spectrometer has been installed at the University of Kaiserslautern and is in full operation since march 2007. The setup is equipped with a low vibrating two-stage pulse tube cooler and has a cool down time of 48 h. The sample can be top loaded without the need to shut off the refrigerator. With the static helium exchange gas in the variable temperature insert the sample may be cooled down from room temperature to 50 K within several hours. Dynamic exchange gas with external supply of gaseous helium is used to cool the sample down to 2 K. The superconducting self-shielding split-coil generates a magnetic field of up to 5 Tesla and a stray field of ca. 60 mT at the outer cryostat walls. Moessbauer measurements can be performed in perpendicular or parallel field orientations. The sample holder and the Moessbauer drive are rigidly connected to the cryostat. In this way a line width of the two inner α-Fe lines of 0.32 mm/s has been currently achieved.
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
Medical Isotope Production Analyses In KIPT Neutron Source Facility
Energy Technology Data Exchange (ETDEWEB)
Talamo, Alberto [Argonne National Lab. (ANL), Argonne, IL (United States); Gohar, Yousry [Argonne National Lab. (ANL), Argonne, IL (United States)
2016-01-01
Medical isotope production analyses in Kharkov Institute of Physics and Technology (KIPT) neutron source facility were performed to include the details of the irradiation cassette and the self-shielding effect. An updated detailed model of the facility was used for the analyses. The facility consists of an accelerator-driven system (ADS), which has a subcritical assembly using low-enriched uranium fuel elements with a beryllium-graphite reflector. The beryllium assemblies of the reflector have the same outer geometry as the fuel elements, which permits loading the subcritical assembly with different number of fuel elements without impacting the reflector performance. The subcritical assembly is driven by an external neutron source generated from the interaction of 100-kW electron beam with a tungsten target. The facility construction was completed at the end of 2015, and it is planned to start the operation during the year of 2016. It is the first ADS in the world, which has a coolant system for removing the generated fission power. Argonne National Laboratory has developed the design concept and performed extensive design analyses for the facility including its utilization for the production of different radioactive medical isotopes. ^{99}Mo is the parent isotope of ^{99m}Tc, which is the most commonly used medical radioactive isotope. Detailed analyses were performed to define the optimal sample irradiation location and the generated activity, for several radioactive medical isotopes, as a function of the irradiation time.
Evaluation and analysis of nuclear resonance data
International Nuclear Information System (INIS)
Frohner, F.H.
2000-01-01
A probabilistic foundations of data evaluation are reviewed, with special emphasis on parameter estimation based on Bayes' theorem and a quadratic loss function, and on modern methods for the assignment of prior probabilities. The data reduction process leading from raw experimental data to evaluated computer files of nuclear reaction cross sections is outlined, with a discussion of systematic and statistical errors and their propagation and of the generalized least squares formalism including prior information and nonlinear theoretical models. It is explained how common errors induce correlations between data, what consequences they have for uncertainty propagation and sensitivity studies, and how evaluators can construct covariance matrices from the usual error information provided by experimentalists. New techniques for evaluation of inconsistent data are also presented. The general principles are then applied specifically to the analysis and evaluation of neutron resonance data in terms of theoretical models - R-matrix theory (and especially its practically used multi-level Breit-Wigner and Reich-Moore variants) in the resolved region, and resonance-averaged R-matrix theory (Hauser-Feshbach theory with width-fluctuation corrections) in the unresolved region. Complications arise because the measured transmission data, capture and fission yields, self-indication ratios and other observables are not yet the wanted cross sections. These are obtained only by means of parametrisation. The intervening effects - Doppler and resolution broadening, self-shielding, multiple scattering, backgrounds, sample impurities, energy-dependent detector efficiencies, inaccurate reference data etc - are therefore also discussed. (author)
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)
Implementation of a Gadolinium Burnable Absorber in the Carbide LEU-NTR
International Nuclear Information System (INIS)
Venneria, Paolo; Kim, Yonghee
2015-01-01
Among the most crucial are the rapid reactivity depletion during full-power operation and the positive reactivity insertion during the full-submersion criticality accident. In previous work, it has been suggested that both challenges can be mitigated through the successful implementation of a burnable absorber in the active core. Of the poisons previously surveyed, one of the most promising is Gadolinium in the form of Gadolina (Gd2O4). This paper explores the possibility of different methods by which the Gadolinia can be implemented in the core and makes a preliminary study of its effect on the full submersion criticality accident and the reactivity depletion during operation. The application of a Gadolinium neutron absorber in the active core region of the LEU-NTR has been shown to be neutronically feasible. It can be introduced into the core in various locations without resulting in core performance loss. The utility of the poison in terms of mitigating the full-submersion reactivity accident and the rapid change in reactivity during full-power operation have been preliminarily shown and the first steps towards eventual implementation made. Future work will consist of determining the maximum poison content in the core and tailoring the self-shielding effect in order to determine a specific Gd depletion rate
International Nuclear Information System (INIS)
Jallu, F.; Passard, C.; Mariani, A.; Ma, J.L.; Baudry, G.; Romeyer-Dherbey, J.; Recroix, H.; Rodriguez, M.; Loridon, J.; Denis, C.; Toubon, H.
2003-01-01
This paper deals with the PROMETHEE (PROMpt, epithermal and THErmal interrogation experiment) waste assay system for alpha low level waste (LLW) characterization. This device, including both passive and active neutron measurement methods, is developed at the French Atomic Energy Commission (C.E.A.), Cadarache Centre, in cooperation with COGEMA. Its purpose is to reach the requirements for incinerating alpha waste (less than 50 Bq[α], i.e. about 50 μg of Pu per gram of raw waste) in 118 litre- > drums. The PROMETHEE development and progress are performed with the help of simulation based on the Monte Carlo code MCNP4 [1]. These calculations are coupled with specific experiments in order to confirm calculated results and to obtain characteristics that can not be approached by the simulation (background for example). This paper presents the PROMETHEE measurement cell, its current performances, and studies performed at the laboratory about the most limiting parameters such as the matrix of the drum - its composition (H, Cl..), its density and its heterogeneity degree -the localization and the self-shielding properties of the contaminant. (orig.)
Energy Technology Data Exchange (ETDEWEB)
Jallu, F.; Passard, C.; Mariani, A.; Ma, J.L.; Baudry, G.; Romeyer-Dherbey, J.; Recroix, H.; Rodriguez, M.; Loridon, J.; Denis, C. [French Atomic Energy Commission (C.E.A./Cadarache), DED/SCCD/LDMN, Durance (France); Toubon, H. [COGEMA, VELIZY-VILLACOUBLAY (France)
2003-07-01
This paper deals with the PROMETHEE (PROMpt, epithermal and THErmal interrogation experiment) waste assay system for alpha low level waste (LLW) characterization. This device, including both passive and active neutron measurement methods, is developed at the French Atomic Energy Commission (C.E.A.), Cadarache Centre, in cooperation with COGEMA. Its purpose is to reach the requirements for incinerating alpha waste (less than 50 Bq[{alpha}], i.e. about 50 {mu}g of Pu per gram of raw waste) in 118 litre-<
Study of the Effect of Burnable Poison Particles Applying in a Pebble Bed HTR
International Nuclear Information System (INIS)
Wei Chunlin; Zhao Jing; Zhang Jian; Xia Bing
2014-01-01
In pebble bed high temperature gas cooled reactors (HTR), spherical fuel elements pass through the core several times to balance the burnup process in the fuel region, resulting in an acceptable shape and peak factor of power density in the simulation analysis. In contrast, when fuel elements pass through the core only once, the peak of power density occurs at the top of the core and its value is too high to be safe. These indicators/parameters can be improved by incorporating burnable poison in the fuel elements under certain conditions. In the current study, burnable poison particles (BPPs) in fuel elements are evaluated. In spite of the strong absorption capability of "1"0B, BPPs can decrease the depletion speed and increase the duration of "1"0B because of the self-shielding effect, resulting in improved shape and peak factor of power distribution. Several BPPs with different radius are discussed in power distribution, following the calculation for a full-scale reactor core with modified VSOP code. According the result, applying BPPs on fuel pebbles is an effective means to improve the distribution of the power density under one-through fuel load in HTR. (author)
S/sub n/ analysis of the TRX metal lattices with ENDF/B version III data
International Nuclear Information System (INIS)
Wheeler, F.J.
1975-01-01
Two critical assemblies, designated as thermal-reactor benchmarks TRX-1 and TRX-2 for ENDF/B data testing, were analyzed using the one-dimensional S/sub n/-theory code SCAMP. The two assemblies were simple lattices of aluminum-clad, uranium-metal fuel rods in triangular arrays with D 2 O as moderator and reflector. The fuel was low-enriched (1.3 percent 235 U), 0.387-inch in diameter and had an active height of 48 inches. The volume ratio of water to uranium was 2.35 for the TRX-1 lattice and 4.02 for TRX-2. Full-core S/sub n/ calculations based on Version III data were performed for these assemblies and the results obtained were compared with the measured values of the multiplication factors, the ratio of epithermal-to-thermal neutron capture in 238 U, the ratio of epithermal-to-thermal fission in 235 U, the ratio of 238 U fission to 235 U fission, and the ratio of capture in 238 U to fission in 235 U. Reaction rates were obtained from a central region of the full-core problems. Multigroup cross sections for the reactor calculation were obtained from S/sub n/ cell calculations with resonance self-shielding calculated using the RABBLE treatment. The results of the analyses are generally consistent with results obtained by other investigators
Energy Technology Data Exchange (ETDEWEB)
A. T. Till; M. Hanuš; J. Lou; J. E. Morel; M. L. Adams
2016-05-01
The standard multigroup (MG) method for energy discretization of the transport equation can be sensitive to approximations in the weighting spectrum chosen for cross-section averaging. As a result, MG often inaccurately treats important phenomena such as self-shielding variations across a material. From a finite-element viewpoint, MG uses a single fixed basis function (the pre-selected spectrum) within each group, with no mechanism to adapt to local solution behavior. In this work, we introduce the Finite-Element-with-Discontiguous-Support (FEDS) method, whose only approximation with respect to energy is that the angular flux is a linear combination of unknowns multiplied by basis functions. A basis function is non-zero only in the discontiguous set of energy intervals associated with its energy element. Discontiguous energy elements are generalizations of bands and are determined by minimizing a norm of the difference between snapshot spectra and their averages over the energy elements. We begin by presenting the theory of the FEDS method. We then compare to continuous-energy Monte Carlo for one-dimensional slab and two-dimensional pin-cell problem. We find FEDS to be accurate and efficient at producing quantities of interest such as reaction rates and eigenvalues. Results show that FEDS converges at a rate that is approximately first-order in the number of energy elements and that FEDS is less sensitive to weighting spectrum than standard MG.
Development of random geometry capability in RMC code for stochastic media analysis
International Nuclear Information System (INIS)
Liu, Shichang; She, Ding; Liang, Jin-gang; Wang, Kan
2015-01-01
Highlights: • Monte Carlo method plays an important role in modeling of particle transport in random media. • Three stochastic geometry modeling methods have been developed in RMC. • The stochastic effects of the randomly dispersed fuel particles are analyzed. • Investigation of accuracy and efficiency of three methods has been carried out. • All the methods are effective, and explicit modeling is regarded as the best choice. - Abstract: Simulation of particle transport in random media poses a challenge for traditional deterministic transport methods, due to the significant effects of spatial and energy self-shielding. Monte Carlo method plays an important role in accurate simulation of random media, owing to its flexible geometry modeling and the use of continuous-energy nuclear cross sections. Three stochastic geometry modeling methods including Random Lattice Method, Chord Length Sampling and explicit modeling approach with mesh acceleration technique, have been developed in RMC to simulate the particle transport in the dispersed fuels. The verifications of the accuracy and the investigations of the calculation efficiency have been carried out. The stochastic effects of the randomly dispersed fuel particles are also analyzed. The results show that all three stochastic geometry modeling methods can account for the effects of the random dispersion of fuel particles, and the explicit modeling method can be regarded as the best choice
International Nuclear Information System (INIS)
Santoro, R.T.; Maskewitz, B.F.; Roussin, R.W.; Trubey, D.K.
1976-01-01
The activities of the Radiation Shielding Information Center (RSIC) of the Oak Ridge National Laboratory are being utilized in support of fusion reactor technology. The major activities of RSIC include the operation of a computer-based information storage and retrieval system, the collection, packaging, and distribution of large computer codes, and the compilation and dissemination of processed and evaluated data libraries, with particular emphasis on neutron and gamma-ray cross-section data. The Center has acquired thirteen years of experience in serving fission reactor, weapons, and accelerator shielding research communities, and the extension of its technical base to fusion reactor research represents a logical progression. RSIC is currently working with fusion reactor researchers and contractors in computer code development to provide tested radiation transport and shielding codes and data library packages. Of significant interest to the CTR community are the 100 energy group neutron and 21 energy group gamma-ray coupled cross-section data package (DLC-37) for neutronics studies, a comprehensive 171 energy group neutron and 36 energy group gamma-ray coupled cross-section data base with retrieval programs, including resonance self-shielding, that are tailored to CTR application, and a data base for the generation of energy-dependent atomic displacement and gas production cross sections and heavy-particle-recoil spectra for estimating radiation damage to CTR structural components
Energy Technology Data Exchange (ETDEWEB)
Kotegawa, Hiroshi; Tanaka, Shun-ichi; Sakamoto, Yukio; Nakane, Yoshihiro; Nakashima, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1994-08-01
A comprehensive attenuation data of dose equivalent for point isotropic monoenergetic neutron sources up to 400MeV in infinite shields of water, ordinary concrete and iron has been calculated using the ANISN-JR code and a neutron-photon multigroup macroscopic cross section HIL086R. The attenuation factors were fitted to a 4th order polynomial exponent formula, making possible to use easily for point kernel codes. Additional data in finite shielding geometry was also calculated to correct the effect due to infinite medium, giving the maximum correction of 0.23 in the region for more 400 cm distance from neutron source of 400 MeV in iron shield. Effective attenuation length for monoenergetic neutrons have been studied in detail. Subsequently, it was shown that the attenuation length was strongly dependent upon the penetration length and the Moyer`s formula using a single attenuation length brought large error into the dose estimation behind thick shields for the intermediate energy neutrons up to 400 MeV. Furthermore, it was demonstrated that there was difference more than 50 % in the attenuation length of iron between the calculations with HIL086R and HIL086 because of the self-shielding effect. (author).
International Nuclear Information System (INIS)
Alam, S.B.; Lindley, B.A.; Parks, G.T.
2015-01-01
In this reactor physics study, we attempt to design a civil marine reactor core that can operate over a 10 effective-full-power-years life at 333 MWth using ThUO 2 and all-UO 2 fuel. We use WIMS to develop subassembly designs and PANTHER to examine whole-core arrangements, optimizing: subassembly and core geometry; fuel enrichment; burnable and moveable poison design; and whole-core loading patterns. We compare designs with a 14% fissile loading for ThUO 2 and all-UO 2 fuel in 13*13 assemblies with ZrB 2 integral fuel burnable absorber pins for reactivity control. Taking advantage of self-shielding effects, the ThUO 2 option shows greater promise in the final burnable poison design while maintaining low, stable reactivity with minimal burnup penalty. For the final poisoning design with ZrB 2 , ThUO 2 contributes 2.5% more initial reactivity suppression, although the all-UO 2 design exhibits lower reactivity swing. All the candidate materials show greater rod worth for the ThUO 2 design. For both fuels, B 4 C has the highest reactivity worth, providing 10% higher control rod worth for ThUO 2 fuel than all-UO 2 . Finally, optimized assemblies were loaded into a 3D reactor model in PANTHER. The PANTHER results show that after 10 years, the core is on the border of criticality, confirming the fissile loading is well-designed. (authors)
DESIGN OF LSDS FOR ISOTOPIC FISSILE ASSAY IN SPENT FUEL
Directory of Open Access Journals (Sweden)
YONGDEOK LEE
2013-12-01
Full Text Available A future nuclear energy system is being developed at Korea Atomic Energy Research Institute (KAERI, the system involves a Sodium Fast Reactor (SFR linked with the pyro-process. The pyro-process produces a source material to fabricate a SFR fuel rod. Therefore, an isotopic fissile content assay is very important for fuel rod safety and SFR economics. A new technology for an analysis of isotopic fissile content has been proposed using a lead slowing down spectrometer (LSDS. The new technology has several features for a fissile analysis from spent fuel: direct isotopic fissile assay, no background interference, and no requirement from burnup history information. Several calculations were done on the designed spectrometer geometry: detection sensitivity, neutron energy spectrum analysis, neutron fission characteristics, self shielding analysis, and neutron production mechanism. The spectrum was well organized even at low neutron energy and the threshold fission chamber was a proper choice to get prompt fast fission neutrons. The characteristic fission signature was obtained in slowing down neutron energy from each fissile isotope. Another application of LSDS is for an optimum design of the spent fuel storage, maximization of the burnup credit and provision of the burnup code correction factor. Additionally, an isotopic fissile content assay will contribute to an increase in transparency and credibility for the utilization of spent fuel nuclear material, as internationally demanded.