Fast Neutron Flux Distribution Research Articles

Core Optimization for Extending the Graphite Irradiation Lifespan in a Small Modular Thorium-Based Molten Salt Reactor

The lifespan of core graphite under neutron irradiation in a commercial molten salt reactor (MSR) has an important influence on its economy. Flattening the fast neutron flux (≥0.05 MeV) distribution in the core is the main method to extend the graphite irradiation lifespan. In this paper, the effects of the key parameters of MSRs on fast neutron flux distribution, including volume fraction (VF) of fuel salt, pitch of hexagonal fuel assembly, core zoning, and layout of control rod assemblies, were studied. The fast neutron flux distribution in a regular hexagon fuel assembly was first analyzed by varying VF and pitch. It was demonstrated that changing VF is more effective in reducing the fast neutron flux in both global and local graphite blocks. Flattening the fast neutron flux distribution of a commercial MSR core was then carried out by zoning the core into two regions under different VFs. Considering both the fast neutron flux distribution and burnup depth, an optimized core was obtained. The fast neutron flux distribution of the optimized core was further flattened by the rational arrangement of control rod channels. The calculation results show that the final optimized core could reduce the maximum fast neutron flux of the graphite blocks by about 30% and result in a more negative temperature reactivity coefficient, while slightly decreasing the burnup and maintaining a fully acceptable core temperature distribution.

Neutronic calculations for the VVER-1000 MOX core computational benchmark using the OpenMC code

The goal of this study is to perform neutronic calculations of the VVER-1000 MOX core computational benchmarks with an OpenMC code along with ENDF/B-VII.1 nuclear data library. The results of neutronic analysis using the OpenMC Monte Carlo code for the VVER-1000 MOX core, containing 30% mixed oxide fuel with low enriched uranium fuel, are presented in this study. As per the benchmark report, all six states are considered in the present study. The keff values, assembly average fission reaction rates, and pin-by-pin fission rates were calculated as per benchmark criteria. In addition, 2D thermal and fast neutron-flux distribution were also generated. The reactivity results and neutron flux distribution were compared with other results in which benchmark analysis was performed using the same core geometry and it showed great similarity with slight deviation. This shows that the modeling of the VVER-1000 MOX core was done successfully using OpenMC. Because OpenMC was successfully used for neutronics calculation of the VVER-1000 whole core, it may be mentioned here that OpenMC code can also be utilized for neutronics and other reactor core physics analyses of the VVER-1200 reactor which is to be commissioned in Bangladesh in the upcoming year.

A new structure design to extend graphite assembly lifespan in small modular molten salt reactors

Dimensional change of the graphite assemblies due to irradiation is a key factor in limiting the service lifespan of the reactor core in small modular molten salt reactors. The graphite assembly structure, determining the fast neutron flux distribution and temperature distribution, has a notable impact on its dimensional change and hence its lifespan. In this article, we put forward an innovative solid hexagonal prism assembly (HPA) surrounded by fuel salt to prolong the lifespan of the graphite assemblies and compare it with the traditional hexagonal round channel assembly (RCA) with fuel salt flowing through its central round channel. Under the single-cell model, the calculation results prove that the neutron flux distribution and temperature distribution in the HPA are more uniform than those in the RCA. The HPA deforms more slowly than the RCA under the same operating conditions. The lifespan of the RCA, defined as when it expands back to its original size, is 7.5 years under the 100 MW/m3 power density of fuel salt, while the lifespan of the proposed HPA is 8.4 years. The HPAs also allow a further expansion in the radial direction due to its noncontact arrangement in the core. It will take 14.9 years for the HPA assemblies to contact each other, which displays a significant improvement in lifespan.

Assessment of CASL VERA for BWR analysis and application to accident tolerant SiC/SiC channel box

Application of the Virtual Environment for Reactor Applications (VERA) to BWR analysis is assessed in this paper by comparing results to those calculated using other widely-used modeling tools, namely the U.S. Nuclear Regulatory Commission’s PARCS/PATHS and the Serpent Monte Carlo particle transport code. Additionally, VERA is used to calculate 3-D temperature and fast neutron flux distributions in silicon carbide (SiC) fiber-reinforced, SiC matrix composite (SiC/SiC) BWR channel boxes, which are being studied as an Accident Tolerant Fuel core structural material concept. The code-to-code comparisons were favorable, and the SiC/SiC channel box evaluation demonstrates the many advanced modeling features VERA provides while also highlighting the non-uniformity in fast neutron flux distributions that can play a role in potential SiC/SiC channel box deformation. Traditional BWR analysis tools do not have the calculation fidelity necessary for coupled assessment of flux and temperature gradients in a SiC/SiC channel box. VERA is a state-of-the-art modeling environment that was developed to increase the safety and economic competitiveness of nuclear power through improved modeling accuracy. While VERA has already been deployed in the nuclear industry for PWR applications, the current study is a vital initial step in the extensive development, validation, and verification that VERA must go through to be useful for BWR applications.

Verification and Validation of SuperMC3.2 Using VENUS-3 Benchmark Experiments

Methodological processes for nuclear power plant (NPP) pressure vessels' (PV) neutron fluence rate determination take the form of experimental measurement or theoretical calculations. However, the process of experimental measurement takes longer periods, as it requires the incorporation of surveillance capsules into a PV system undergoing normal NPP operation. Therefore, strong reliance on computation and modeling of radiation-induced degradation is given much attention. In this work, the VENUS-3 benchmark has been analyzed using SuperMC code, with the intention of validating SuperMC for accurate reactor neutronics; dosimetry response calculations for in-core/ex-core structural components, particularly with respect to the VENUS-3 configuration type pressurized water reactors (PWRs). In this work, complete three-dimensional (3D) geometry including the source modeling for VENUS-3 facility has been developed with SuperMC. Neutron transport and calculations of equivalent fission flux for the experimental target quantities, 115In (n, n′), 58Ni (n, p), and 27Al (n, α), are also achieved. The calculation results show good agreement with the experimental measurement. The greater majority of the calculated values (C/E) were within the required accuracy of ±10% for reactor components' dosimetry calculations. Most of the calculated values were contained within 5% deviation from the experimental data. Additional calculations and detailed analysis for fast neutron flux distribution and iron displacement per atom rate (dpa/s), including the characteristic effect of partial length shielded assembly (PLSA) on VENUS-3 core barrel, are also discussed. It is therefore evidenced that the effectiveness of SuperMC code for in-core/ex-core reactor neutronics computations has been convincingly demonstrated through the VENUS-3 benchmark testing.

Application of global variance reduction method to calculate a high-resolution fast neutron flux distribution for TMSR-SF1

The solid fuel thorium molten salt reactor (TMSR-SF1) is a 10-MWth fluoride-cooled pebble bed reactor. As a new reactor concept, one of the major limiting factors to reactor lifetime is radiation-induced material damage. The fast neutron flux (E > 0.1 MeV) can be used to assess possible radiation damage. Hence, a method for calculating high-resolution fast neutron flux distribution of the full-scale TMSR-SF1 reactor is required. In this study, a two-step subsection approach based on MCNP5 involving a global variance reduction method, referred to as forward-weighted consistent adjoint-driven importance sampling, was implemented to provide fast neutron flux distribution throughout the TMSR-SF1 facility. In addition, instead of using the general source specification cards, the user-provided SOURCE subroutine in MCNP5 source code was employed to implement a source biasing technique specialized for TMSR-SF1. In contrast to the one-step analog approach, the two-step subsection approach eliminates zero-scored mesh tally cells and obtains tally results with extremely uniform and low relative uncertainties. Furthermore, the maximum fast neutron fluxes of the main components in TMSR-SF1 are provided, which can be used for radiation damage assessment of the structural materials.

Neutronic simulation of CAREM-25 small modular reactor

CAREM is an Argentine project aimed to achieve the development, design and construction of an innovative small and integrated nuclear power plant. Some of the unique characteristics of CAREM-25 are integrated primary cooling system, self-pressurized, primary cooling by natural circulation, and safety systems based on passive specifications. In the present study neutronic evaluation of the CAREM-25 is discussed. Neutronic calculations are performed using WIMS, CITATION and MCNPX codes. Neutronic parameters such as effective multiplication factor, thermal and fast neutron flux distributions are discussed. Comparison of the results of CITATION and MCNPX codes show that the effective multiplication factor and the behavior of neutron flux distributions have an appropriate consistency.

Design of fast neutron channels for topaz irradiation

The aim of this work is to design fast neutron irradiation channels for topaz irradiation at Egypt Second Research Reactor (ETRR-2) by shielding the aluminum boxes, used for topaz irradiation, with B4C (boron carbide) or manufacturing the irradiation boxes from aluminum contains boron. This in turn suppresses the thermal neutrons inside the box resulting in reduction of the residual radioactivity in topaz which decreases the worker exposure and release time. Simulation of the fast irradiation channels using the code: MCNPX showed that the thermal neutron component can be suppressed to less than one-tenth. ETRR-2 has many irradiation positions. The fast neutron flux distributions along the irradiation positions and the required irradiation times were calculated. The limiting conditions for safe operation of movable experiments are considered in the selection of the irradiation positions. That is the change in reactor reactivity due to the B-10 shield is below the reactivity limit for safe irradiation experiments. Heat generated during irradiation in topaz and in the material that contains boron was considered, as well. Experimentally, the worker exposure has been reduced to less than one-tenth and a decrease of 12months in the release time of the irradiated topaz has been achieved. Also adequate cooling of the shielded box during irradiation has been demonstrated.

Neutronic analysis of a superheat BWR using high performance fuel rod technique

In this study, the neutronic analysis of conceptual design of a superheat BWR with annular fuel rods in steady state conditions is discussed. The purpose of this design, is to increase the superheated steam temperature to about 540 °C and plant efficiency to 40%. Neutronic calculations are carried out for a fuel assembly using MCNPX.2.7 Code. COBRA-EN code, based on Drift-flux two phase flow model, has been used to obtain thermal-hydraulic parameters. The proposed design for the fuel assembly includes the evaporator and superheater regions. The fuel assembly consists of sixty annular fuel elements in an 8 × 8 square array. Each annular fuel element is cooled externally by water and internally by steam. Finally, the neutronic results such as radial and axial power peaking factors, thermal and fast neutron flux distribution, fuel cycle estimation and void reactivity coefficient are investigated in the present article.

A Comparative Study of the HEU, LEU Fuels: Assessing the Effect of the MNSR Fuel Burnup on the Effective Multiplication Factor and Neutron Fluxes

The GETERA and MCNP4C codes have been used to perform the core conversion analysis of the MNSR from the existing HEU to four potential LEU fuels. The effect of fuel burnup on the effective multiplication factor (keff) and the neutron fluxes in the irradiation sites was also investigated. Results showed that the “keff“ dropped sharply in the first days of the reactor operation time (Xe effect), then decreased slowly due to the fuel consumption. The axial thermal, epithermal and fast neutron flux distributions in the irradiation sites exhibited similar flux patterns for all types of fuel. The values of the axial thermal neutron flux were slightly lower in the LEU fuel. For the HEU fuel, the fuel residence time in the core was shorter than that for the LEU fuel. For both HEU and LEU cores, the changes in the neutron fluxes in the irradiation sites with burnup were negligible, so no burnup effect on the neutron fluxes was observed.

Determination of fast neutron flux distribution in irradiation sites of the Malaysian Nuclear Agency research reactor

Determination of thermal to fast neutron flux ratio (f(fast)) and fast neutron flux (ϕ(fast)) is required for fast neutron reactions, fast neutron activation analysis, and for correcting interference reactions. The f(fast) and subsequently ϕ(fast) were determined using the absolute method. The f(fast) ranged from 48 to 155, and the ϕ(fast) was found in the range 1.03×10(10)-4.89×10(10) n cm(-2) s(-1). These values indicate an acceptable conformity and applicable for installation of the fast neutron facility at the MNA research reactor.

Investigation of fast neutron shielding characteristics depending on boron percentages of MgB2, NaBH4 and KBH4

The macroscopic cross-section Σ and average neutron fluence in matter Φ are usable factors to comment neutron shielding property of samples. In this paper, we have used MgB2, NaBH4 and KBH4 samples including different percentages of boron. Neutron macroscopic cross-section measurements of them have been done by using a source of mono-energetic neutrons (E eff = 4.5 MeV 241Am–Be). Average neutron fluence values and double differential fast neutron flux distributions of each samples calculated by using FLUKA Monte Carlo code. Also half value layers (HVLs) of samples are compared to paraffin which is one of the most neutron moderators. As a result, growing boron concentration can raise neutron shielding property of materials.

Analysis of a Neutronic Experiment on a Simulated Mercury Spallation Neutron Target Assembly Bombarded by Giga-Electron-Volt Protons

A neutronic benchmark experiment on a simulated spallation neutron target assembly was conducted by using the Alternating Gradient Synchrotron at Brookhaven National Laboratory and was analyzed to investigate the prediction capability of Monte Carlo simulation codes used in neutronic designs of spallation neutron sources. The target assembly consisting of a mercury target, a light water moderator, and a lead reflector was bombarded by 1.94-, 12-, and 24-GeV protons, and the fast neutron flux distributions around the target and the spectra of thermal neutrons leaking from the moderator were measured in the experiment. In this study, the Monte Carlo particle transport simulation codes NMTC/JAM, MCNPX, and MCNP-4A with associated cross-section data in JENDL and LA-150 were verified based on benchmark analysis of the experiment. As a result, all the calculations predicted the measured quantities adequately; calculated integral fluxes of fast and thermal neutrons agreed approximately within ±40% with the experiments although the overall energy range encompassed more than 12 orders of magnitude. Accordingly, it was concluded that these simulation codes and cross-section data were adequate for neutronics designs of spallation neutron sources.

Calculation and Benchmarking of an Azimuthal Pressure Vessel Neutron Fluence Distribution Using the Boxer Code and Scraping Experiments

The light water reactor BOXER code was used to determine the fast azimuthal neutron fluence distribution at the inner surface of the reactor pressure vessel after the tenth cycle of a pressurized water reactor (PWR). Using a cross-section library in 45 groups, fixed-source calculations in transport theory and x-y geometry were carried out to determine the fast azimuthal neutron flux distribution at the inner surface of the pressure vessel for four different cycles. From these results, the fast azimuthal neutron fluence after the tenth cycle was estimated and compared with the results obtained from scraping test experiments. In these experiments, small samples of material were taken from the inner surface of the pressure vessel. The fast neutron fluence was then determined form the measured activity of the samples. Comparing the BOXER and scraping test results have maximal differences of 15%, which is very good, considering the factor of 10{sup 3} neutron attenuation between the reactor core and the pressure vessel. To compare the BOXER results with an independent code, the 21st cycle of the PWR was also calculated with the TWODANT two-dimensional transport code, using the same group structure and cross-section library. Deviations in the fast azimuthal flux distribution weremore » found to be <3%, which verifies the accuracy of the BOXER results.« less

Fast-neutron flux distribution in ducted light and heavy concrete shields

This paper reports on a study of the effect of cylindrical air-filled ducts of different lengths and diameters on the distribution of fast neutrons in ordinary concrete ( ϱ = 2.3 g/cm 3) and ilmenite concrete ( ϱ = 4.6 g/cm 3). The source of neutrons was a collimated beam of reactor neutrons emitted from one of the horizontal channels of the ET-RR-1 reactor. The measurements were performed using phosphorus activation detectors. The data show that the presence of air-filled ducts tends to increase the neutron flux in the medium surrounding the duct, as compared with the corresponding values for solid concrete. A marked peak was observed at about 10 cm after the end of the duct. The value of the flux at these peaks is dependent on the duct length and diameter. Also, it was shown that the neutron flux at any point along the duct axis passing through ilmenite concrete is much lower than the corresponding value in ordinary concrete. In addition, solid ilmenite concrete is much better for fast-neutron attenuation than solid ordinary concrete, especially for large thicknesses. However, for small thicknesses (< 30 cm) ordinary concrete is the best. Semiempirical formulae have been derived to calculate the fast-neutron fluxes at different thicknesses along the axis of air-filled ducts of different diameters inserted in ordinary and ilmenite concretes and to calculate the fast-neutron flux distribution in solid concrete media. The derived formulae give results which are in good agreement with the measured values.

Application of the CR-39 detector for neutron measurements in a spent fuel storage pond

Application of the CR-39 solid state nuclear track detector for verification of spent CANDU (Canada Deuterium Uranium) fuel was investigated. A new method based on simultaneous measurement of thermal and fast neutron flux distributions in the lateral plane near the front end of the fuel bundle matrix is proposed. A pilot experiment was performed to determine the detector's response to thermal and fast neutrons in near real conditions. Thermal neutron flux spatial distribution was determined in the spent fuel pit of the KANUPP nuclear power plant (Pakistan). A strong near-neighbour effect was found. It was proved that one day of irradiation time is sufficient for in situ measurements. Further improvements of the method are proposed.

Analysis of neutron flux distributions generated by sources of fast neutrons in a range of bituminous coals

Monte Carlo calculations of fast neutron flux distributions have been made for 252Cf, 241Am/Be and 14 MeV monoenergetic neutron sources in a range of bituminous coals. The configurations examined were sources at the centre of a sphere, at the midpoint of the plane surface of a hemisphere and at or near the surface of a parallel-sided slab. The neutron fluxes were analysed in terms of the resultant yield of γ-rays from inelastic scattering on silicon, carbon and oxygen, which are the key reactions for determining the oxygen content of coal by processes yielding prompt γ-rays. The fast neutron spectra were found to undergo significant changes in shape as a function of distance from the source, in contrast with the assumptions made in the model developed by Wormald and Clayton. (1) However, the ratios of the response functions for inelastic scattering by the critical elements were found to be very insensitive to the type of coal, indicating that oxygen analyses could be performed for a given configuration and neutron source by using calibration factors derived from a single calculation or benchmark experiment on a coal of intermediate composition. The calculations described also provided information on the yield and spatial distribution of thermal neutrons from the selected sources.

Experiments and Analyses of Neutron and Gamma-Ray Streaming in the Cavity-Duct System of a Fast Neutron Source Reactor

Reaction rate and dose rate distributions due to streaming of neutrons and gamma rays were measured in the nearly cubical cavity and in the ducts leading to the cavity at the fast reactor YAYOI. The experimental configuration was arranged in a way that streaming through the ducts occurred only after scattering in the cavity. This was done by making the duct axes perpendicular to the source radiation from the reactor core. The spectrum of the source was modified from fast to thermal by putting moderators in the beam hole of the core assembly. The spatial distribution of the neutron reaction rates and gamma-ray dose rates in the cavity is almost flat except for a small decrease at the wall. The cadmium ratios at the duct inlets and the attenuation profiles in the ducts are almost identical irrespective of the spectrum of neutrons entering the cavity. The two-dimensional discrete ordinates transport calculation by TWOTRAN-II has shown that although the ray effect is observed in the fast neutron flux distribution in the cavity, the calculated results agree rather well with the experimental distributions in the cavity and with the streaming in the ducts with regard to the results measured with the l/v type neutron detectors. Also, for the gamma rays the agreement is rather good between the calculation and the experiment.

Neutron beam characterization by means of a radioactivation image transfer (RIT) technique

Fast and thermal neutron distributions in phantom have been recorded by means of neutron induced radioactivity in aluminum and copper sheets, respectively. The induced β-decays in the metal sheets are used to form radiographic images of the fast and thermal neutron distributions on diagnostic X-ray film via autoradiography. Film calibration then translates the optical density map into the fast or slow neutron flux distribution in phantom; and, since the neutron spectrum does not change significantly with depth, the fast neutron flux distribution also maps the fast neutron dose distribution. The present radioactivation-image transfer (RIT) technique has several distinct advantages over ionization chamber measurements for neutron flux distributions: (1) the RIT method requires two thirds less beam time than ion-chamber measurements, (2) the induced activity is insensitive to -/-rays, (3) the equipment requisite for data acquisition is far less costly than an automated ionization chamber apparatus and (4) a permanent analog record is produced which can be examined at any later date.

531. 高速中性子束測定用&lt;SUP&gt;232&lt;/SUP&gt;Th小型核分裂計数管

A 232Th miniature fission counter probe was perfected with use of the Th depositing technique. A sensitivity of approximately 1×10-6 counts/n/cmcm2 was obtained for reactor fast neutron flux with a typical cylindrical counter (6mm O.D., 10mm sensitive length). The counter was applied to the measurement of the fast neutron flux distribution within a research reactor core. The fast neutron flux distribution thus measured with the counter provided results identical with that obtained from the activation of threshold foils. Depleted U fission counters with Cd-cover have in the past been used for fast neutron flux measurements. The 232Th fission counter can, however, be used without Cd-filter, with excellent resolution of thermal and epithermal neutrons.