Ice Clouds Lidar Ratio and Multiple Scattering Factor in Version 4 CALIOP Level 2 Algorithm
Explore the source record for details and available documents.
A Garnier↗
SEARCH · Engineering Papers
Results for “MULTIPLICATION FACTORS”
Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.
Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.
A study on the criticality safety for the direct disposal of used nuclear fuel in Japan: Application of burnup credit to the criticality safety evaluation for the disposal canister - 14049
Japan has recently started the technical development of the direct disposal of used nuclear fuel (UNF) to prepare various disposal options. Unlike the disposal of radioactive waste from UNF reprocessing, criticality safety is important here because of the presence of a certain amount of fissile nuclides in UNF. This paper provides an outline of the research to be addressed in this field and the relevant studies in Japan. It especially includes the first result of a criticality safety evaluation of a disposal canister model adopting burnup credit. In this evaluation, it was assumed that a steel disposal canister loaded with fuel assemblies was surrounded by a clay buffer material, and the fuel assemblies and the disposal canister were intact. Conditions representative of discharged pressurized-water reactor (PWR) fuel were applied in this study - initial U-235 enrichment of 4.5 wt% and burnup of 45 GWd/t. The isotopic composition of the burnt fuel was evaluated using SWAT3.1 code. In the depletion calculation, conservative parameters were applied within the typical PWR operation range. The effective neutron multiplication factor (k{sub eff}) was then calculated using a continuous-energy Monte Carlo code MVP-2.0, adopting the JENDL-4.0 nuclear data library. Uncertainty in k{sub eff} due to depletion calculation errors was evaluated using the correction factors of isotopic composition, which were obtained via multiple comparisons between the calculated and measured isotopic concentrations of irradiated fuels. In addition, the effects of axial burnup profile and horizontal burnup gradient on k{sub eff} were evaluated using the open burnup distribution database. The k{sub eff} including the uncertainties and conservatism was found to be below 0.95 for both actinide-only and actinide and fission product considerations when the fuel assemblies and the disposal canister were assumed to be intact. (authors)
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
WFIRST Coronagraph Flight Performance Modeling
As it has for the past few years, numerical modeling is being used to predict the on-orbit, high-contrast imaging performance of the WFIRST coronagraph, which was recently defined to be solely a technology demonstrator. A consequence has been a realignment of modeling priorities and revised applications of modeling uncertainty factors and margins, which apply to multiple factors such as pointing and wavefront jitter, thermally-induced deformations, polarization, and aberration sensitivities. At the same time, the models have increased in fidelity as additional parameters have been added, such as time-dependent pupil shear and mid-spatial-frequency deformations of the primary and secondary mirrors, detector effects, and reaction-wheel-speed-dependent pointing and wavefront jitter.
Trauger, John↗
Generating Mixed Patterns of Residential Segregation: An Evolutionary Approach
The Schelling model of residential segregation has demonstrated that even the slightest preference for neighbors of the same race can be amplified into community-wide segregation. However, these models are unable to simulate mixed, coexisting patterns of segregation and integration, which have been seen to exist in cities. Using evolutionary model discovery we demonstrate how including social factors beyond racial bias when modeling relocation behavior enables the emergence of strongly mixed patterns. Our results indicate that the emergence of mixed patterns is better explained by multiple factors influencing the decision to relocate; the most important being the interaction of nonlinear, rapidly diminishing racial bias with a recent, historical tendency to move. Additionally, preference for less isolated neighborhoods or preference for neighborhoods with longer residing neighbors may produce weaker mixed patterns. Finally, this work highlights the importance of exploring the influence of multiple hypothesized factors of decision making, and their interactions, within agent rules, when studying emergent outcomes generated by agent-based models of complex social systems.
97 MATHEMATICS AND COMPUTING↗
Self-shielding models of MICROX-2 code
The MICROX-2 is a transport theory code that solves for the neutron slowing-down and thermalization equations of a two-region lattice cell. In the previous study, a new fine-group cross section library of the MICROX-2 was generated and tested against reference calculations and measurement data. In this study, existing physics models of the MICROX-2 are reviewed and updated to improve the physics calculation performance of the MICROX-2 code, including the resonance self-shielding model and spatial self-shielding factor. The updated self-shielding models have been verified through a series of benchmark calculations against the Monte Carlo code, using homogeneous and pin cell models selected for this study. The results have shown that the updates of the self-shielding factor calculation model are correct and improve the physics calculation accuracy even though the magnitude of error reduction is relatively small. Compared to the existing models, the updates reduced the prediction error of the infinite multiplication factor by approximately 0.1 % and 0.2% for the homogeneous and pin cell models, respectively, considered in this study. (authors)
97 MATHEMATICS AND COMPUTING↗
Dead-Time Correction of Pulsed Neutron Source Experiments with the Backward Extrapolation Method
This paper introduces a novel procedure to correct pulsed neutron source experimental data for dead-time effects. The procedure is based on the application of the backward extrapolation method to the reaction rate value in each time interval (bin) of the neutron counts distribution as a function of time. The theoretical framework has been successfully applied to the experimental data (timestamps) from the YALINA Thermal facility driven by a pulsed D-D neutron source. Dead-time effects may have more than 500 pcm (1 pcm=1 E-5) impact on the effective multiplication factor calculated by the area method. (authors)
07 ISOTOPE AND RADIATION SOURCES↗
Implementation and validation of the variational nodal expansion method core simulator
The variational nodal expansion method (VNEM) has been developed by the OECD Halden Reactor Project for the purpose of implementation into a full-core simulator. This version of the VNEM does not make use of the diffusion approximation, but instead provides a fast converged solution to the even-parity neutron transport equation in several minutes using nodal methods and variational techniques. This 'method to simulator' implementation has been completed and the VNEM-simulator is currently undergoing continued validation efforts to ensure that it can track more nuclear reactor cycles effectively. This paper presents the results of the most recent development on the 3D VNEM core simulator implementation and validation efforts. The method is briefly described and data is presented and discussed for three cases for the initial criticality tests at the beginning of life, hot stand-by state for the Ringhals-3 PWR. Also presented here are the results for one year of operation of the maiden cycle of this unit. The effective neutron multiplication factor, k{sub eff}, and relative neutron detector readings are compared showing excellent agreement. (authors)
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
The features of neutronic calculations for fast reactors with hybrid cores on the basis of BFS-62-3A critical assembly experiments
The different (U-Pu) fuel compositions are considered for next generation of sodium fast breeder reactors. The considerable discrepancies in axial and radial neutron spectra for hybrid reactor systems compared to the cores with UO{sub 2} fuel cause increasing uncertainty of generating the group nuclear constants in those reactor systems. The calculation results of BFS-62-3A critical assembly which is considered as full-scale model of BN-600 hybrid core with steel reflector specify quite different spectra in local areas. For those systems the MCNP 5 calculations demonstrate significant sensitivity of effective multiplication factor K{sub eff} and spectral indices to nuclear data libraries. For {sup 235}U, {sup 238}U, {sup 239}Pu the results of calculated radial fission rate distributions against the reconstructed ones are analyzed. Comparative analysis of spectral indices, neutron spectra and radial fission rate distributions are performed using the different versions of ENDF/B, JENDL-3.3, JENDL-4, JEFF-3.1.1 libraries and BROND-3 for Fe, Cr isotopes. For analyzing the fission rate sensitivity to the plutonium presence in the fuel {sup 239}Pu is substituted for {sup 235}U (enrichment 90%) in the FA areas containing the plutonium. For {sup 235}U, {sup 238}U, {sup 239}Pu radial fission rate distributions the explanation of pick values discrepancies is based on the group fission constants analyses and possible underestimation of some features at the experimental data recovery method (Westcott factors, temperature dependence). (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
A Chemistry Based Criticality Safety Evaluation of the HM-Process Mixer Settlers
The criticality safety of Savannah River Site's H Canyon two cycle uranium extraction and purification process, called the HM process, is examined. Explicit Monte Carlo models of the multi-stage mixer-settlers are generated for the first time using the KENO-VI module of the SCALE 6.1 code package and spanning normal operating conditions and a wide range of potential upset conditions with stagewise compositions based directly on SEPHIS Mod-4 chemistry models of the same conditions. To achieve industrial scale processing, mixer-settlers are often not geometrically favorable to prevent criticality so other means must be used for criticality safety, namely control of fissile concentration and mass. Historically, the worst case upset is a reflux condition resulting from a stream flow rate, temperature, or composition that is off from the normal operating conditions. A reflux is considered to be the state in which the fissile material is not proceeding in the desired direction, i.e. is stripping rather than extracting or vice versa. The criticality safety analyses which set the current operational limits were conducted many years ago with less sophisticated codes and only loosely tied criticality to chemistry. Those analyses conservatively assumed that a reflux changed the behavior of the mixer-settler such that the fissile content collected continuously in only three stages of the mixer-settler, at a rate based on the SEPHIS modeling, until criticality occurred. This implies that under some condition the distribution coefficient of the fissile material either goes to infinity or zero, which is not physically achievable. In such a scenario, the operator would have some amount of time to react and correct the problem before sufficient fissile material built up in the mixer-settler and caused a criticality. In reality, an upset in a stream would cause the process to transition to a new, though undesired, equilibrium state of operation. The distribution coefficients would change, but would not diverge. In this work, aqueous and organic streams are modeled for both the mixer and the settler region of each stage of each bank using chemical distribution data from SEPHIS and solving the chemical equations for a wide range of operational parameters spanning normal and upset conditions. Conservative assumptions are made regarding enrichment and reflection and the whole bank multiplication factor is evaluated using discrete stage-wise compositions. (authors)
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
HTTR criticality calculations with SCALE6: Studies of various geometric and unit-cell options in modeling
The fuel element of the High Temperature Engineering Test Reactor (HTTR) presents a doubly heterogeneous geometry, where tiny TRISO fuel particles dispersed in a graphite matrix form the fuel region of a cylindrical fuel rod, and a number of fuel rods together with moderator or reflector then constitute the lattice design of the core. In this study, a series of full-core HTTR criticality calculations were performed with the SCALE6 code system using various geometric and unit-cell options in order to systematically investigate their effects on neutronic analysis. Two geometric descriptions (ARRAY or HOLE) in SCALE6 can be used to construct a complicated and repeated model. The result shows that eliminating the use of HOLE in the HTTR geometric model can save the computation time by a factor of 4. Four unit-cell treatments for resonance self-shielding corrections in SCALE6 were tested to create problem-specific multigroup cross sections for the HTTR core model. Based on the same ENDF/B-VII cross-section library, their results were evaluated by comparing with continuous-energy calculations. The comparison indicates that the INFHOMMEDIUM result overestimates the system multiplication factor (k{sub eff}) by 55 mk, whereas the LATTICECELL and MULTIREGION treatments predict the k{sub eff} values with similar biases of approximately 10 mk overestimation. The DOUBLEHET result shows a more satisfactory agreement, about 4.2 mk underestimation in the k{sub eff} value. In addition, using cell-weighted cross sections instead of an explicit modeling of TRISO particles in fuel region can further reduce the computation time by a factor of 5 without sacrificing accuracy. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Bayesian estimation of cross-section and errors of experiment and calculation for accurate prediction of neutronic characteristics
In some experiments, there might be a correlation of experimental errors such as an error caused by fabrication. The correlation of calculation errors also might exist among analyses of experiments. In the present study, considering these correlations, the cross-section and errors of experiment and calculation are adjusted for the accurate prediction of neutronic characteristics. The adjustment method is derived based on Bayesian theory. The 'a posteriori' distributions of cross-section and errors of experiment and calculation are obtained by the derivation. The case study is performed by using the benchmarks of light-water moderated experiments reported in ICSBEP. The error from fuel rod fabrication and measurement are included in the experiment error, and the error from the simplification of the benchmark model is included in the calculation error. Since these errors have a positive correlation among the benchmarks, the prediction accuracy is expected to be improved by considering these correlations. For the sensitivity analysis of the correlation among benchmarks, the correlation coefficient of experiment error and that of calculation error is assumed from 0.0 to 0.9. The results indicate that the prediction accuracy of the multiplication factor can be improved by increasing the correlation coefficient of experiment error and that of calculation error. (authors)
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
MOCUM: A two-dimensional method of characteristics code based on unstructured meshing for general geometries
A transport theory code MOCUM based on the Method of Characteristics (MOC) as the flux solver with an advanced general geometry processor is developed for two-dimensional lattice and full core neutronics modeling. The core structure is represented by Constructive Solid Geometry (CSG) that uses Boolean operations to build complex geometries from simple polygons. Arbitrary-precision arithmetic is also used in the process of building CSG objects to eliminate the round-off error from the commonly used double precision numbers. Then, the constructed core frame will be decomposed and refined into a conforming Delaunay triangulation to ensure the quality of the meshes. The MOC solver kernel is fully paralleled using OpenMP and the developed numerical code is validated by several benchmarks representing various core geometries. For the cases modeled, the maximum percentage error for multiplication factor and the pin power compared to reference values are 0.1% and 0.7% respectively. (authors)
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Burnup calculation by the method of first-flight collision probabilities using average chords prior to the first collision
A technique to calculate the burnup of materials of cells and fuel assemblies using the matrices of first-flight neutron collision probabilities rebuilt at a given burnup step is presented. A method to rebuild and correct first collision probability matrices using average chords prior to the first neutron collision, which are calculated with the help of geometric modules of constructed stochastic neutron trajectories, is described. Results of calculation of the infinite multiplication factor for elementary cells with a modified material composition compared to the reference one as well as calculation of material burnup in the cells and fuel assemblies of a VVER-1000 are presented.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Study on neutronics calculation method of SCWR with annular fuel assembly - 26804
Many Supercritical Water Reactor designs have been proposed all over the world. Annular fuel was employed in SCWR by Xi'an JiaoTong University to achieve lower fuel center temperature and lower cladding surface temperature. Different analysis codes have been developed for these designs. In this article, a neutronics calculation code for SCWR with annular fuel is developed. This code is compared with Monte-Carlo code doing both assembly and core calculation. The results showed that the spectra difference of PWRs and SCWRs is the main reason for assembly calculation errors and conventional assembly homogenization method is not suitable for SCWRs. Thus the group-collapsing and spatial-averaging method were studied to improve the accuracy. The few-group structure which contains 14 energy groups is used. The 'super-assembly' model is employed to consider effect of reflector. The results showed that the accuracy of effective multiplication factor and radial power was improved. (authors)
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Towards application of a neutron cross-section uncertainty propagation capability in the criticality safety methodology - 14434
At the Paul Scherrer Institute (PSI) the methodology for criticality safety evaluations (CSE) used for Swiss applications is focused, following common practice, on establishing an upper subcriticality limit (USL) for the effective neutron multiplication factor (k{sub eff}), based on a statistical evaluation of the obtained validation results in terms of the calculated-to-benchmark k{sub eff} values. Recently, a new computational tool referred to as NUSS (Nuclear data Uncertainty Stochastic Sampling) was developed in order to propagate neutron-nuclear data (ND) uncertainties in calculations with Monte Carlo transport codes in combination with general-purpose ND libraries in ACE format. Thereby, this tool should now allow for a quantification of ND data uncertainties in CSE and to introduce these into the CSE criteria definition. This paper concerns an application of the NUSS methodology for criticality validation studies. Calculated k{sub eff} uncertainty estimations were obtained for the current PSI validation suite including in total 149 critical benchmark configurations for which in-house MCNP(X) models were developed, covering both uranium and mixed uranium-plutonium oxide fuels. The MCNPX-2.7.0 code was used together with the ENDF/B-VII.1 ND library while the ND variance-covariance matrixes (VCM) were taken from both SCALE-6.0 44-group libraries and also from ENDF/B-VII.1 files. Based on the newly obtained information on the uncertainties, a re-assessment of the lower tolerance bound (LTB) for k{sub eff} using the total uncertainties as weighting factors is proposed as means for the present PSI CSE methodology upgrade and some possible directions of further methodology enhancements are discussed next. (authors)
42 ENGINEERING↗
X-ray emission from relativistically moving electron density cusps
We report on novel methods to generate ultra-short, coherent, X-rays using a laserplasma interaction. Nonlinear interaction of intense laser pulses with plasma creates stable, specific structures such as electron cusps. For example, wake waves excited in an underdense plasma by an intense, short-pulse laser become dense and propagate along with the laser pulse. This is called a relativistic flying mirror. The flying mirror can reflect a counter-propagating laser pulse and directly convert it into high-frequency radiation, with a frequency multiplication factor of {approx} 4{gamma}{sup 2} and pulse shortening with the same factor. After the proof-of-principle experiments, we observed that the photon number generated in the flying mirror is close to the theoretical estimate. We present the details of the experiment in which a 9 TW laser pulse focused into a He gas jet generated the Flying Mirror, which partly reflected a 1 TW pulse, giving up to {approx} 10{sup 10} photons, 60 nJ (1.4 Multiplication-Sign 10{sup 12} photons/sr) in the XUV spectral region (12.8-22 nm).
07 ISOTOPE AND RADIATION SOURCES↗
Monte Carlo modeling and analyses of YALINA booster subcritical assembly, Part III : low enriched uranium conversion analyses.
This study investigates the performance of the YALINA Booster subcritical assembly, located in Belarus, during operation with high (90%), medium (36%), and low (21%) enriched uranium fuels in the assembly's fast zone. The YALINA Booster is a zero-power, subcritical assembly driven by a conventional neutron generator. It was constructed for the purpose of investigating the static and dynamic neutronics properties of accelerator driven subcritical systems, and to serve as a fast neutron source for investigating the properties of nuclear reactions, in particular transmutation reactions involving minor-actinides. The first part of this study analyzes the assembly's performance with several fuel types. The MCNPX and MONK Monte Carlo codes were used to determine effective and source neutron multiplication factors, effective delayed neutron fraction, prompt neutron lifetime, neutron flux profiles and spectra, and neutron reaction rates produced from the use of three neutron sources: californium, deuterium-deuterium, and deuterium-tritium. In the latter two cases, the external neutron source operates in pulsed mode. The results discussed in the first part of this report show that the use of low enriched fuel in the fast zone of the assembly diminishes neutron multiplication. Therefore, the discussion in the second part of the report focuses on finding alternative fuel loading configurations that enhance neutron multiplication while using low enriched uranium fuel. It was found that arranging the interface absorber between the fast and the thermal zones in a circular rather than a square array is an effective method of operating the YALINA Booster subcritical assembly without downgrading neutron multiplication relative to the original value obtained with the use of the high enriched uranium fuels in the fast zone.
43 PARTICLE ACCELERATORS↗