Volume 7 • Number 1 • June 2018
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Vol. 7No. 1pp. 1–9
High-energy muons generated from cosmic-ray particle showers have been shown to exhibit properties ideal for imaging the interior of large and dense structures. This paper explores the possibility of using a single portable muon detector in conjunction with image reconstruction methods used in nuclear medicine to reconstruct a 3-D image of the interior of critical infrastructure such as the Zero-Energy Deuterium (ZED-2) research reactor at Canadian Nuclear Laboratories’ Chalk River site. The ZED-2 reactor core and muon detector arrangement are modeled in GEANT4 and Monte Carlo measurements of the resultant muon throughput and angular distribution at several angles of rotation around the reactor are generated. Statistical analysis is then performed on these measurements based on the well-defined flux and angular distribution of muons expected near the surface of the earth. The results of this analysis are shown to produce reconstructed images of the spatial distribution of nuclear fuel within the core for multiple fuel configurations. This one-sided tomography concept is a possible candidate for examining the internal structure of larger critical facilities, for example the Fukushima Daiichi power plant, where the integrity of the containment infrastructure and the location of the reactor fuel is unknown.
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Vol. 7No. 1pp. 11–17
Multiple scattering has been well recognized as an important correction in neutron scattering cross-section measurements. The GEANT4 simulation toolkit includes a special thermal neutron scattering model and a corresponding data library at low neutron energies (<4 eV). A new method using GEANT4 to estimate the multiple-scattering effect in thermal neutron scattering experiments is presented. The method was applied to the double differential cross-section measurements of light water with various sample thicknesses under ambient conditions of temperature and pressure. The resulting scattering law for neutron energy transfer from 42.0 to 14.6 meV over scattering angles from 10° to 110° is presented and compared with the tabulated Evaluated Nuclear Data File (ENDF/B-VII).
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Vol. 7No. 1pp. 19–25
MCNP full-reactor models of the SLOWPOKE-2 and ZED-2 research reactors have been used for calculating gamma spectra and dose rates at locations where measurements were made. In addition to neutrons, MCNP criticality calculations are capable of tracking reactor prompt photons in space and energy, which can be tallied at locations of interest. Delayed photons from conceivable sources can be approximately included in the calculations, contributing to as much as one-third of the local photon intensity or committed dose rate in this study. Discrepancies between the MCNP calculated and measured results are analyzed for causes.
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Vol. 7No. 1pp. 27–35
A tissue equivalent proportional counter (TEPC) based on a gas electron multiplier (GEM) with a 2-dimensional (2-D) readout board for neutron dosimetry was developed and tested. This 2-D design of the anode structure presents multiple advantages over the traditional spherical TEPC design. This type of detector does not need a central wire to provide amplification as in traditional proportional counter design, making it possible to achieve higher efficiency (or sensitivity). Each charge-collecting pad in the 2-D structure represents a single miniature TEPC that can simulate the tissue size of a fraction of a micrometre. Such measurements could provide more precise dose-equivalent levels. In addition, the 2-D anode pattern provides the ability to reconstruct the 2-D projection of the ionizing particle’s track and it paves the way for the development of a time projection chamber for the full 3-D track reconstruction. This would prove an invaluable tool in future microdosimetry research. The simple design of GEM-based TEPCs permits them to be built relatively easily. Initially, a Monte Carlo study of the detector was conducted and the results of this study are presented. They show that the response of the GEM-based TEPC to a low-energy neutron would be greatly improved compared with that of spherical TEPCs. The prototype of the 2-D GEM-based TEPC has been constructed and tested with 252Cf neutron sources as well as with X-ray beams of various energies. The test with 252Cf was performed using moderated and nonmoderated neutrons. The lineal energy distribution spectra and the GEM-based TEPC’s dose-equivalent responses for these sources are presented in this paper.
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Vol. 7No. 1pp. 37–46
Dynamic parameters such as prompt neutron generation time are essential to safety analysis of a nuclear reactor. Currently, Monte Carlo based simulation methods are relied on for obtaining such dynamic parameters. There are known disagreements between different simulation codes. Further, the nuclear data that all simulations are based on cannot be considered accurate for all parameter ranges. An independent method for getting the parameters is desired for safety analysis and code verification purposes. A neutron flux perturbation experiment method to address this gap is introduced in this paper. The neutron flux variation is sampled together with a pseudo-random binary sequence perturbation input. A system transfer function and coherence from the input signal to the output signal are obtained through digital processing steps. The dynamic parameters that uniquely decide the system transfer function could be obtained by least-square fitting. In this work, 2 essential parameters, the prompt neutron generation time Λ and the effective delayed neutron fraction βeff, are obtained based on the experimental data. The results are in good agreement with Monte Carlo N-Particle code-based calculation. With room for improvement, the method introduced here provides an independent way of experimentally obtaining these dynamic parameters of a reactor core.
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Vol. 7No. 1pp. 47–66
The objective of this paper is to assess different correlations independently against a diversified databank—the Canadian Nuclear Laboratories multi-fluid and multi-geometry supercritical heat transfer databank. This databank was recently expanded by adding compiled and original experimental data obtained through collaboration with the Nuclear Power Institute of China. The databank was subjected to screening for outliers, duplicates, and unreliable data. In addition, inappropriate data, not satisfying certain conditions, were removed. Nevertheless, the used databank comprised more than 41 000 measurements of heat transfer to different fluids flowing vertically upward in different geometries. Following a literature review and a compilation of correlations, an assessment of the tabulated correlations was performed against the databank. In total, 24 correlations were considered and applied to the entire database for different fluids including water and different flow geometries including tube, annulus, and rod bundle. Graphical comparison of best-estimate correlations and representative experimental data is presented in this paper. In addition, statistical error analysis was performed and leading correlations were identified. Although the leading correlation showed a standard deviation of less than 6%, variation of predicted wall temperature and heat transfer coefficient with fluid temperature followed the scatter of the experimental data.
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Vol. 7No. 1pp. 67–83
The objective of this study is to evaluate round-tube data-based correlations depending on their applicability conditions including heat transfer mode and fluid. The present assessment of correlations was performed against the round-tube databases for vertical upward flow of water and CO2 from the Canadian Nuclear Laboratories multifluid, multigeometry databank of supercritical heat transfer. To categorize the data according to a representative heat mode, various criteria for onset of heat transfer deterioration were proposed. However, there is no consensus in the literature on a single approach. Two popular, semi-empirical criteria were chosen for screening the data for buoyancy- and acceleration-induced heat transfer deterioration. The experimental conditions of the data were screened for specified ranges of the 2 criteria indicating heat transfer deterioration. However, in many cases, the corresponding heat transfer mode of the experimental data was not appropriately predicted. In light of this inadequacy, improving accuracy of this method was deemed necessary. Therefore, each of the 2 criteria was empirically modified twice, once for water and once again for CO2. This paper presents these 2 original modifications for each of the fluids. Ultimately, each experimental data point was categorized by the modified criteria into 1 of 2 heat transfer modes, either normal or deteriorated. In total, 21 round-tube correlations were selected and applied to the categorized databases of normal and deteriorated heat transfer for water and CO2. Details of the assessment results are presented in this paper in tables of uncertainty numbers for each of the correlations and databases and in graphs, comparing best-estimate correlations with representative experimental data.
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Vol. 7No. 1pp. 85–94
External or soil-side corrosion poses a significant threat to buried piping in many industrial facilities, including nuclear power plants. Unlike aboveground piping, the assessment of external corrosion of buried piping is challenged not only by access restrictions, but also by the complex and highly variable external environment. The analysis is further confounded by the use of protective measures, such as coatings and cathodic protection, which, while mitigating against the onset of degradation, themselves also degrade and break down over time.This study uses extensive field data collected by the National Bureau of Standards at numerous test sites across the United States to estimate the distribution of external corrosion rates for buried carbon steel piping under different soil conditions. The analysis is based on the concept of a “steady-state” corrosion rate, which is estimated for each test site in a 2-step process using logarithmic regression. The resulting uniform (i.e., general corrosion) and localized (i.e., pitting) corrosion rates are grouped according to their soil texture classification and fitted by the log-normal probability distribution. A specific value of the distribution, e.g., 80th percentile upper bound, can then be used to predict the corrosion allowance, both in terms of leakage and structural integrity, over a given evaluation period as part of the fitness-for-service assessment process.The study results account for the general impact of environmental conditions on the external corrosion process, and are readily applicable to any existing site with buried carbon steel piping exhibiting (or suspected of experiencing) degradation (i.e., where coating and (or) cathodic protection are no longer deemed effective).
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Vol. 7No. 1pp. 95–102
One of the critical factors in the analysis of the in-vessel retention of corium during a postulated severe accident in a nuclear power plant is the interaction of corium constituents with the reactor vessel wall material. In a CANDU© reactor, after fuel channel disassembly, corium constituents would come into contact with the calandria vessel wall. Ablation of the wall due to physicochemical interaction with corium at high temperature could cause vessel failure, increasing the likelihood of radioactive release to the environment. Therefore, the interaction of the calandria vessel wall material (stainless steel 304L) with each of the main corium constituents (Zircaloy-4, Zr-2.5%Nb, Zircaloy-2, and UO2) was studied experimentally. Arrhenius equations that can be used for analysis of in-vessel corium retention were derived from experimental results. For alloys of Zr, measured rates of interaction were comparable with those reported for similar light-water reactor materials. For UO2, the rate of interaction at 1200 °C was negligible. Experiments performed with a zirconium oxide layer of 10 μm on the Zircaloy samples showed that the oxide acts as a protective barrier against high-temperature interactions with the vessel material. In all cases, if the temperature of the corium–vessel interface remains below the lowest Fe–Zr eutectic temperature of 928 °C, no significant ablation is observed after 24 h.
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Vol. 7No. 1pp. 103–117
Electrochemical oxidation on boron-doped diamond (BDD) electrodes is reportedly more effective than conventional chemical or electrochemical oxidation methods for the removal of recalcitrant organic contaminants from aqueous effluents. This work investigated the electrochemical oxidation of acetic acid and nuclear process effluents containing representative recalcitrant chlorinated and nitrogenated organics on a BDD anode using a DiaCell® apparatus. The removal of organic carbon mostly depended on applied current density, mass transfer effects, and the properties of organic compounds. The total organic carbon concentrations in the solutions were reduced to as low as 0.5 g·m−3, achieving mineralization of up to 99% of organic carbon. Dissolved oxygen significantly contributed to the removal of organic carbon at low concentrations. In particular, the BDD anodic oxidation system was capable of mineralizing organic carbon representing a mixture of mineral oil and carboxylic, aromatic, chlorinated, and nitrogenated organics. Complementing the experimental work, an electrochemical oxidation model was developed for simulating the reaction process on the BDD anode and for improving simpler literature-reported models. This model considered effects of fluid dynamic and operating parameters and introduced a retardation factor that addresses the effect of charges on the organic moieties on mass transfer and subsequently on the anodic oxidation rate. An expression for the retardation factor was formulated in terms of the ionization fraction of an organic compound, the charges on organic ions, the number of electrons transferred per organic carbon atom oxidized, and the effect of heteroatoms such as oxygen, nitrogen, and chlorine. Model predictions were verified with the experimental data obtained by varying pH, temperature, current density, volumetric flow, and dissolved oxygen. This model can be further exploited for optimization and design of improved BDD anodic oxidation processes.
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Vol. 7No. 1pp. 119–125
The microstructure and texture of Zr-2.5Nb pressure tubes is greatly influenced by the manufacturing route. Although the general nature of the microstructure remains consistent between different manufacturing routes, subtle differences in the relative size and aspect ratios of the elongated α-Zr grains differ with pressure tubes of different pedigree. These differences have been shown to correlate well with in-reactor deformation; however, the ability to consistently and efficiently characterize the microstructures hinders an understanding of the fundamental degradation mechanisms. This paper outlines a new approach to characterize Zr-2.5Nb pressure tubes using thin foils characterized with both diffraction contrast in a conventional transmission electron microscope (TEM) and transmission Kikuchi diffraction (TKD) in a scanning electron microscope (SEM). The combined approaches enable a characterization of the same region of material with both techniques and capitalize on the advantages of each approach. In addition to obtaining general microscopy from SEM-TKD, the localized texture is obtained and compared to texture from X-ray diffraction, which provides higher confidence that the grains examined in the TEM foils are representative of the bulk material.
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