Macroscopic Absorption Cross Section Research Articles

Influence analysis of B4C content on the neutron shielding performance of B4C/Al

B4C/Al composites have been widely studied and used as neutron absorbing materials in various application fields because of its relatively simple preparation process, good mechanical properties, and low raw material prices. B4C content is the major factor affecting the neutron shielding performance of B4C/Al composites. A photo-neutron source driven by a 15 MeV electron linear accelerator has been used to measure the neutron shielding performance of the B4C/Al composite plates with B4C weight percentages of 16.85% and 31%, respectively. Based on the measurement result, the thermal neutron macroscopic cross sections of the B4C(16.85 wt%)/Al composite and B4C(31 wt%)/Al composite were 13.97 cm−1 and 25.58 cm−1, respectively. The measurement results showed that the B4C/Al composites have good shielding performance for neutrons with energies lower than 10 eV, especially for thermal neutrons, and the 1.02 mm thickness of B4C(31 wt%)/Al composite was able to absorb 90.91% of thermal neutrons. Experimental measurements were compared with calculation results. The experimental measurement result of the thermal neutron macroscopic absorption cross section was smaller than the calculated data, and the relative deviations of the experimental measurement results and calculation data of the thermal neutron macroscopic cross sections of B4C(16.85 wt%)/Al and B4C(31 wt%)/Al composites were −13.65% and −12.52%, respectively. The Monte Carlo N-Particle Transport code was used to calculate the neutron shielding performance of B4C/Al with different B4C content, and the shielding performance of the B4C/Al composites for low-energy neutrons was approximately linearly proportional to the B4C content, especially for the thermal neutron.

Nonlinear fractional diffusion model for space-time neutron dynamics

In this work, a new fractional multidimensional neutron diffusion model is introduced using Cattaneo equation instead of the commonly used Fick's law for two neutron energy groups. The model represents a coupled system of nonlinear hyperbolic fractional partial differential equations considering finite fast and finite thermal neutron velocities. The model provides new quantities that relate some of the most important nuclear parameters together producing some newly physical concepts. These quantities include fast and thermal neutron relaxation times, Riemann-Liouville fractional integrals of the neutron densities and derivatives of macroscopic absorption cross sections. This model can be useful for mathematicians and physicists since the anomalous subdiffusion process describes the neutron movement closer to reality. The effect of the anomalous subdiffusion exponent (0<γ≤1) on the neutron flux is discussed. The model is successfully applied to simple homogenous and heterogonous reactor cores with step, ramp and sinusoidal perturbations for different values ofγ to estimate the average reactor powers in the fuel and moderator regions. Also, the new fractional model is applied to LRA BWR benchmark problem including adiabatic heating and Doppler feedback. On the other hand and based on the concept of prompt jump, the proposed fractional model is simplified and then analytically solved using Mittag-Liffler function. Solving fractional models numerically consumes considerable computational time costs due to the persistent memory of fractional operators, especially for smaller values of γ. For this regard, adaptively multi-step differential transform MDTM method is implemented to solve the proposed fractional model. The MDTM is developed using step size control technique for which the length of the step integration varies according to the solution behaviour.

Radiation shielding properties of modified concrete mixes and their suitability in dry storage cask

Radiation shielding properties of 8 concrete mixes were studied in detail. The concrete mixes included 4 coarse aggregate types; dolomite, barite, ilmenite, and celestite, and 2 cementing material content. Important parameters accounted for in this study include: Mass attenuation coefficient that was calculated using WinXCom program then linear attenuation coefficient, half value layer, tenth value layer, and mean free path were deduced; Effective atomic number was calculated using Auto-Zeff software; Exposure build-up factor was calculated using the Geometric Progression approximation method up to 40 mean free path using Phy-X/PSD Software; Fast neutrons macroscopic effective removal cross-section and thermal neutron macroscopic absorption cross-section were calculated using MRCsC and JANIS-4 programs, respectively. Subsequently, the mixes were assessed for use in a dry storage cask using OpenMC code. The results showed that concrete mixes containing barite and celestite were the best in attenuating gamma rays emitted in narrow and broad beam situations.

Mechanical and radiation shielding properties of concrete containing commercial boron carbide powder

Boron-containing compounds have an excellent radiation shielding capability against slow neutrons. Moreover, sassolite in boron-containing compounds adversely affects cement hydration by retarding cement setting times. Accordingly, experimental and analytical investigations were carried out to examine the filler effect of commercial boron carbide (B4C) fine powder with a mean particle diameter of 2.64 μm on Portland cement hydration reaction, and concrete mechanical and radiation shielding properties. Isothermal calorimetry method was used to study the heat evolution and the cement degree of hydration of mixes containing 0, 25, 50 and 75% of B4C to cement by weight. Effects of adding B4C on concrete compressive strength were then investigated. Analytical investigations, specifically mass attenuation coefficient using WinXcom program and effective atomic number using Auto-Zeff, were carried out to evaluate the radiation shielding of corresponding concrete mixes for photon energies ranging from 10 keV to 10 MeV. Moreover, macroscopic effective removal cross-sections for fast neutrons and thermal neutron macroscopic absorption cross-sections were calculated using MRCsC program and JANIS-4 software, respectively. The results confirm that the presence of sassolite in the studied B4C powder retards the hydration reaction for the first 24 h and reveal significant improvements in concrete strength thereafter with the increase of boron carbide due to the filler effect. Furthermore, the addition of boron carbide yielded measurable improvement in neutron shielding capabilities of concrete mixes.

Preparation of boron carbide nanosheets with high neutron-shielding properties based on reduced graphene oxide aerogel

For this paper, boron carbide nanosheets were synthesized using a simple solid-phase thermal reaction with reduced graphene oxide (rGO) aerogel as a carbon source and amorphous boron powder as a boron source. The phase, structure, morphology, and particle size of the prepared B4C were characterized by X-ray diffraction (XRD), field emission scanning electron microscopy (FE-SEM), Fourier transform infrared spectroscopy (FT-IR), Raman spectroscopy, transmission electron microscopy (TEM), X-ray photoelectron spectra (XPS) and nanoparticle size distribution analysis. The effect of the reaction temperature, reaction time and B/C molar ratio on the crystal structure of B4C were also investigated. The results showed that the optimum conditions for the synthesis of B4C nanosheets were a reaction temperature of 1450 °C, a reaction time of 4 h and a B/C molar ratio of 4. The resulting nanosheets are hexagonal in shape, with edge lengths of 200–500 nm and an average thickness of about 100 nm. The crystal structure of the nanosheet is rhombohedral. The formation mechanism of the nanosheets is discussed on the basis of the experimental results. In addition, testing the neutron-shielding properties of the B4C composites showed that neutron transmittance with a B4C nanosheets content of 25 wt% was 94.83% lower (from 56.47% to 2.92%) than with epoxy resin, and the total macroscopic absorption cross-section was 518.41% greater (from 0.2857 ± 0.0023 cm−1 to 1.7668 ± 0.0004 cm−1). Hence, the results of this study identify a lightweight, effective material with excellent neutron-shielding properties.

The Adjoint Discrete Ordinates Solution Applied to Direct and Inverse Neutral Particle X, Y-Geometry Source-Detector Problems

Solving the adjoint transport equation allows obtaining maps of the neutral-particle importance to an objective function, e.g., a detector response. In source-detector problems, the adjoint interior source is defined as the macroscopic absorption cross section of the material the detector is made of. Determining the intensity of a source of particles in each energy group (causes) is possible by solving an inverse source-detector problem, given the source location and the detector response (effects). In this work, we present the application of an adjoint technique to solve energy multigroup X,Y-geometry direct and inverse transport problems. In order to obtain the importance maps, we have extended the adjoint spectral Green's function constant-nodal (SGF*-CN) method to numerically solve energy multigroup adjoint transport problems in the discrete ordinates formulation. Described here are the methodology to obtain the discretized SGF*-CNequations and the adjoint partial one-node block inversion scheme used to iteratively solve these equations. We present numerical results to two test problems to illustrate the accuracy of the present methodology.

Further development of methodology to model TRISO fuel and BISO absorber particles and related uncertainty quantification using SCALE 6

ABSTRACT In the 350 MW Modular High Temperature Gas-Cooled Reactor (MHTGR-350), not only is the fuel double heterogeneous but so are the lumped burnable poisons (LBPs). The LBPs are composed of Bi-Structural Isotropic (BISO) particles and the fuel is composed of Tri-Structural Isotropic (TRISO) particles. This work investigates different methods to model coated particles using KENO-VI and NEWT of SCALE 6. The most efficient way of modelling TRISO particles in terms of packing and randomization is established in continuous energy (CE) mode and its impact on is investigated. In the multi-group (MG) treatment, coated particles are modelled with the DOUBLEHET function which is only designed for particles that contain fuel. The LBP BISO particles could therefore not be modelled. Hence a method called the LBP Trace method is developed to model the LBP BISO particles using the DOUBLEHET function. It was found that changed by 1500 pcm compared to the conventional (homogenized) case, when using the LBP Trace method. However, no significant change was observed in the macroscopic absorption cross section that would be passed to a nodal core calculation. Furt hermore, the LBP Trace method showed small changes in the nuclear data uncertainty when compared to conventional case.

Analysis of neutronics benchmarks for the utilization of mixed oxide fuel in light water reactor using DRAGON code

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 (MCU) values of the benchmark. Maximum deviation of 1.8% in k∞ is observed for variants with spent fuel and for the states with control rod whereas all the other results are in par with the results reported in the benchmark document. The few and multi-group macroscopic cross-sections and flux of all the nuclides also compare well with the benchmark results except for the 11B macroscopic absorption cross section, which is further compared with the XNWLUP software. Inter-comparison of results with the generalized self-shielding model SHI of DRAGON code employing the traditional WIMSD formatted 172 group cross-section library has also been made to highlight the improvements made in computational schemes and cross-section library format.

Quantifying the effects of depletion parameters on the PWR spent fuel reactivity based on nuclide sensitivity coefficients

This study employs nuclide sensitivity coefficients to investigate the spent fuel reactivity of pressurized water reactors (PWRs) in terms of differences in predicted inventories under various depletion conditions. Considering a representative PWR spent fuel, the study began by selecting the top 20 nuclides of large macroscopic neutron absorption cross sections, which are expected to have a substantial effect on reactivity. The sensitivity coefficients (Δk∞/ΔNi) of each nuclide were first evaluated at various burnups and cooling times using MCNPX. These coefficients were used to facilitate a detailed analysis of the spent fuel reactivity and to identify which nuclides cause the effect in a quantitative manner. The approach was demonstrated by analyzing the effects on spent fuel reactivity of several depletion parameters, including the initial enrichment, specific power, operating history, boron concentration, fuel temperature, moderator temperature, control rod effect, and axial burnup profile. The results were generally in good agreement with those obtained from direct MCNPX criticality calculations. Two major advantages of this approach are: (1) the changes in spent fuel reactivity with various depletion parameters can be closely related to composition changes; and (2) it provides a quantitative estimate of the contribution of each nuclide to the reactivity effect based on the predicted inventory differences.

Using HED meteorites to interpret neutron and gamma‐ray data from asteroid 4 Vesta

AbstractHere, we construct a comprehensive howardite, eucrite, and diogenite (HED) bulk chemistry data set to compare with Dawn data. Using the bulk chemistry data set, we determine four gamma‐ray/neutron parameters in the HEDs (1) relative fast neutron counts (fast counts), (2) macroscopic thermal neutron absorption cross section (absorption), (3) a high‐energy gamma‐ray compositional parameter (Cp), and (4) Fe abundance. These correspond to the four measurements of Vesta made by Dawn's Gamma Ray and Neutron Detector (GRaND) that can be used to discern HED lithologic variability on the Vestan surface. We investigate covariance between fast counts and average atomic mass (&lt;A&gt;) in the meteorite data set, where a strong correlation (r2 = 0.99) is observed, and we demonstrate that systematic offsets from the fast count/&lt;A&gt; trend are linked to changes in Fe and Ni concentrations. To compare the meteorite and GRaND data, we investigate and report covariance among fast counts, absorption, Cp, and Fe abundance in the HED meteorite data set. We identify several GRaND measurement spaces where the Yamato type B diogenites are distinct from all other HED lithologies, including polymict mixtures. The type B's are diogenites that are enriched in Fe + pigeonite + diopside ± plagioclase, relative to typical, orthopyroxenitic diogenites. We then compare these results to GRaND data and demonstrate that regions north of ~70°N latitude on Vesta (including the north pole) are consistent with type B diogenites. We propose two models to explain type B diogenite compositions in the north (1) deposition as Rheasilvia ejecta, or (2) type B plutons that were emplaced at shallow depths in the north polar region and sampled by local impacts. Lastly, using principal component (PC) analysis, we identify unique PC spaces for all HED lithologies, indicating that the corresponding GRaND measurables may be used to produce comprehensive lithologic maps for Vesta.

Deformed hidden conformal symmetry for rotating black holes

The generic non-extremal Kerr-Newman black holes are holographically dual to hidden conformal field theories in two different pictures. The two pictures can be merged together to the CFT duals in general picture that are generated by $SL(2,\mathbb{Z})$ modular group. We find some extensions of the conformal symmetry generators that yield an extended local family of $SL(2,\mathbb{R})_L \times SL(2,\mathbb{R})_R$ hidden conformal symmetries for the Kerr-Newman black holes parameterized by one deformation parameter. The family of deformed hidden conformal symmetry for Kerr-Newman black holes also provides a set of deformed hidden conformal generators for the charged Reissner-Nordstrom black holes. The set of deformed hidden conformal generators reduce to the hidden $SL(2,\mathbb{R})$ conformal generators for the Reissner-Nordstrom black hole for specific value of deformation parameter. We also find agreement between the macroscopic and microscopic entropy and absorption cross section of scalars for the Kerr-Newman black hole by considering the appropriate temperatures and central charges for the deformed CFTs.

Propagation noise calculations in VVER-type reactor core

Neutron noise induced by propagating disturbances in VVER-type reactor core is addressed in this paper. The spatial discretization of the governing equations is based on the box-scheme finite difference method for triangular-z geometry. Using the derived equations, a 3-D 2-group neutron noise simulator (called TRIDYN-3) is developed for hexagonal-structured reactor core, by which the discrete form of both the forward and adjoint reactor dynamic transfer functions (in the frequency domain) can be calculated. In addition, both types of noise sources, namely point-like and traveling perturbations, can be modeled by TRIDYN-3. The results are then benchmarked in different cases. Considering the noise source as propagating perturbations of the macroscopic absorption cross sections, the induced neutron noise is calculated throughout the reactor core. For the first time, adjoint approach is applied and examined for modeling moving noise sources. Moreover, the space- and frequency-dependence of the propagation noise are investigated in this paper.

Neutron absorption constraints on the composition of 4 Vesta

AbstractGlobal maps of the macroscopic thermal neutron absorption cross section of Vesta's regolith by the Gamma Ray and Neutron Detector (GRaND) on board the NASA Dawn spacecraft provide constraints on the abundance and distribution of Fe, Ca, Al, Mg, and other rock‐forming elements. From a circular, polar low‐altitude mapping orbit, GRaND sampled the regolith to decimeter depths with a spatial resolution of about 300 km. At this spatial scale, the variation in neutron absorption is about seven times lower than that of the Moon. The observed variation is consistent with the range of absorption for howardite whole‐rock compositions, which further supports the connection between Vesta and the howardite, eucrite, and diogenite meteorites. We find a strong correlation between neutron absorption and the percentage of eucritic materials in howardites and polymict breccias, which enables petrologic mapping of Vesta's surface. The distribution of basaltic eucrite and diogenite determined from neutron absorption measurements is qualitatively similar to that indicated by visible and near infrared spectroscopy. The Rheasilvia basin and ejecta blanket has relatively low absorption, consistent with Mg‐rich orthopyroxene. Based on a combination of Fe and neutron absorption measurements, olivine‐rich lithologies are not detected on the spatial scales sampled by GRaND. The sensitivity of GRaND to the presence of mantle material is described and implications for the absence of an olivine signature are discussed. High absorption values found in Vesta's “dark” hemisphere, where exogenic hydrogen has accumulated, indicate that this region is richer in basaltic eucrite, representative of Vesta's ancient upper crust.

Compositional variability on the surface of 4 Vesta revealed through GRaND measurements of high‐energy gamma rays

AbstractMeasurements of the high‐energy gamma‐ray flux emanating from asteroid 4 Vesta by the Dawn Gamma‐Ray and Neutron Detector (GRaND) have revealed variability in the near‐surface elemental composition of the Vestan surface. These observations are consistent with the presence of large (≥8 × 104 km2) regions with distinct, HED‐like elemental compositions. The results agree broadly with other global measurements, such as the macroscopic neutron absorption cross section and spectral reflectance‐derived mineralogic maps. Two distinct regions with eucrite‐like elemental compositions have been identified, the first located primarily within the Lucaria and Marcia quadrangles and the second within Oppia quadrangle. The former region is collocated with some of the oldest, most heavily cratered terrain on Vesta. The interior of the 500 km diameter Rheasilvia impact basin is found to have a composition that is consistent with diogenite‐like material. Taken together, these observations support the hypothesis that Vesta's original crust was composed of basaltic outflows in the form of eucritic‐like material and that the Rheasilvia‐basin‐forming impact exposed lower‐crustal, diogenite‐like material. These measurements also constrain the maximum amount of mesosiderite‐like material to &lt;10% for each 15 × 15° surface element.

Effects of geochemical composition on neutron die-away measurements: Implications for Mars Science Laboratory's Dynamic Albedo of Neutrons experiment

Abstract The Dynamic Albedo of Neutrons (DAN) experiment, part of the scientific payload of the Mars Science Laboratory (MSL) rover mission, will have the ability to assess both the abundance and the burial depth of subsurface hydrogen as the rover traverses the Martian surface. DAN will employ a method of measuring neutron fluxes called “neutron die-away” that has not been used in previous planetary exploration missions. This method requires the use of a pulsed neutron generator that supplements neutrons produced via spallation in the subsurface by the cosmic ray background. It is well established in neutron remote sensing that low-energy (thermal) neutrons are sensitive not only to hydrogen content, but also to the macroscopic absorption cross-section of near-surface materials. To better understand the results that will be forthcoming from DAN, we model the effects of varying abundances of high absorption cross-section elements that are likely to be found on the Martian surface (Cl, Fe) on neutron die-away measurements made from a rover platform. Previously, the Mars Exploration Rovers (MER) Spirit and Opportunity found that elevated abundances of these two elements are commonly associated with locales that have experienced some form of aqueous activity in the past, even though hydrogen-rich materials are not necessarily still present. By modeling a suite of H and Cl compositions, we demonstrate that (for abundance ranges reasonable for Mars) both the elements will significantly affect DAN thermal neutron count rates. Additionally, we show that the timing of thermal neutron arrivals at the detector can be used together with the thermal neutron count rates to independently determine the abundances of hydrogen and high neutron absorption cross-section elements (the most important being Cl). Epithermal neutron die-away curves may also be used to separate these two components. We model neutron scattering in actual Martian compositions that were determined by the MER Alpha Proton X-Ray Spectrometer (APXS), as examples of local geochemical anomalies that DAN would be sensitive to if they were present at the MSL landing site. These MER targets, named “Eileen Dean,” “Jack Russell,” and “Kenosha Comets,” all have unusually high or low Cl or Fe abundances as a result of geochemical interactions involving water. Using these examples we demonstrate that DAN can be used not only to assess the amount of present-day hydrogen in the near-surface but also to identify locations that may preserve a geochemical record of past aqueous processes.

Chemical composition of the lunar surface from neutron leakage fluxes

The neutron leakage fluxes from the lunar surface are calculated by Monte Carlo transport code based on Geant4. The integral fluxes of fast neutrons, epi-thermal neutrons and thermal neutrons are analyzed. Numerical results for 20 kinds of lunar soils and 7 minerals show that the fast neutron fluxes are linearly related to the average atomic mass numbers of the lunar materials used in simulations. Meanwhile, the average atomic mass numbers are strongly modulated by the abundances of iron (Fe) and titanium (Ti), and a linear relationship between the average atomic mass numbers and the abundances of Fe and Ti is found. Furthermore, the results show that the ratios of epi-thermal to thermal fluxes for lunar soils are linearly related to the macroscopic absorption cross sections of lunar materials, and that the macroscopic absorption cross sections monotonically increase with the abundances of Fe and Ti by a simple function. Then we reach the conclusion that the neutron fluxes can provide the information about the Fe and Ti contents.

Constraints on Mercury's surface composition from MESSENGER neutron spectrometer data

The composition of Mercury's surface is poorly known, but the MESSENGER (MErcury Surface, Space ENvironment, GEochemistry, and Ranging) mission has provided a wealth of new data from three flybys. In particular, MESSENGER Neutron Spectrometer (NS) observations reveal a surface enriched in neutron absorbing elements, consistent with interpretations of color and albedo observations suggesting a surface composition enriched in Fe–Mg–Ti oxides. In this study, we have computed the neutron absorption cross sections for all of the available proposed surface compositions of Mercury and evaluated the plausibility of each surface composition based on the neutron absorption cross section observed by MESSENGER. For identified plausible compositions, the implications for the thermal and magmatic evolution of Mercury are discussed. The measured macroscopic neutron absorption cross section of Mercury is inconsistent with a crust formed from partial melting of plausible bulk mantle compositions, flotation in a magma ocean or adiabatic melting of upwelling cumulates during magma ocean overturn. However, the observed neutron absorption is consistent with model compositions of late-stage magma-ocean cumulates and some proposed compositions from spectral modeling and equilibrium modeling. This suggests that the enrichment of neutron absorbing elements may be indicative of the processes that acted to form Mercury's crust. The enrichment in neutron absorbing elements, in combination with spectral observations that constrain FeO in silicates (< 2 wt.%), offers strong evidence of a magma ocean on Mercury since global scale melting appears to be required to concentrate the major neutron absorbing elements while minimizing Fe in silicate minerals. We also find that iron plays a secondary role in the neutron absorption of plausible surface compositions and its variations within different Fe–Mg–Ti oxide solid solution series does not cause any overlap among the various oxide series in neutron absorption cross section. High-Fe oxides are not required and more Mg-rich oxides may even be favored as the Ti-contents can sufficiently account for the observed neutron absorption.

Sensitivity of Physics Parameters for Establishment of a Burned CANDU Full-Core Model for Decommissioning Waste Characterization

The sensitivity of parameters related with reactor physics on the source terms of decommissioning wastes from a CANDU reactor was investigated in order to find a viable, simplified burned core model of a Monte Carlo simulation for decommissioning waste characterization. First, a sensitivity study was performed for the level of nuclide consideration in an irradiated fuel and implicit geometry modeling, the effects of side structural components of the core, and structural supporters for reactive devices. The overall effects for computation memory, calculation time, and accuracy were then investigated with a full-core model. From the results, it was revealed that the level of nuclide consideration and geometry homogenization are not important factors when the ratio of macroscopic neutron absorption cross section (MNAC) relative to a total value exceeded 0.95. The most important factor affecting the neutron flux of the pressure tube was shown to be the structural supporters for reactivity devices, showing an 10% difference. Finally, it was concluded that a bundle-average homogeneous model considering a MNAC of 0.95, which is the simplest model in this study, could be a viable approximate model, with about 25% lower computation memory, 40% faster simulation time, and reasonable engineering accuracy compared with a model with an explicit geometry employing an MNAC of 0.99.

Sensitivity of Physics Parameters for Establishment of a Burned CANDU Full-Core Model for Decommissioning Waste Characterization

The sensitivity of parameters related with reactor physics on the source terms of decommissioning wastes from a CANDU reactor was investigated in order to find a viable, simplified burned core model of a Monte Carlo simulation for decommissioning waste characterization. First, a sensitivity study was performed for the level of nuclide consideration in an irradiated fuel and implicit geometry modeling, the effects of side structural components of the core, and structural supporters for reactive devices. The overall effects for computation memory, calculation time, and accuracy were then investigated with a full-core model. From the results, it was revealed that the level of nuclide consideration and geometry homogenization are not important factors when the ratio of macroscopic neutron absorption cross section (MNAC) relative to a total value exceeded 0.95. The most important factor affecting the neutron flux of the pressure tube was shown to be the structural supporters for reactivity devices, showing an 10% difference. Finally, it was concluded that a bundle-average homogeneous model considering a MNAC of 0.95, which is the simplest model in this study, could be a viable approximate model, with about 25% lower computation memory, 40% faster simulation time, and reasonable engineering accuracy compared with a model with an explicit geometry employing an MNAC of 0.99.

Study on the Dual Gamma Spectrum Saturation Logging Method Based on Pulsed Neutron Source and Numerical Simulation

AbstractThe dual gamma spectrum saturation logging based on a pulsed neutron source is a new technology that can record inelastic scattering and capture gamma ray spectra at the same time, along with gamma ray time spectrum at the far and near gamma ray detector by means of a special pulse and measurement timing design. It is found that counting statistics of inelastic scattering and capture gamma ray is superior to foreign instruments by using the large‐diameter instrument and BGO crystal detector. The gamma ray time spectrum after repeated pulse neutron emission still meets the exponential decay law, and the channel count of gamma time spectrum is cumulated, so its counting statistics is better than conventional neutron lifetime logging. In addition, the macroscopic absorption cross section obtained by this pulse and measure timing is nearly the same as the single pulse. It is found that the oil saturation can be determined with the cross‐plot technology of C/O and formation macroscopic absorption cross section under the condition of unknown formation porosity.