Hasil untuk "Nuclear engineering. Atomic power"

Pengfei Zhang, Yang Zhang, Ting Wang et al.

The reactor vessel head is a core component of the nuclear reactor. Internally, multiple penetrations are connected through interference fit and welding, which are prone to generate Primary Water Stress Corrosion Cracking (PWSCC). ASME codes mandate non-destructive testing (NDT) of penetrations and their J-welds. However, the eccentricity between thermal sleeves and penetrations, coupled with the larger diameter of guide cones connected to thermal sleeves compared to penetrations, creates significant challenges in positioning inspection equipment for penetrations, adversely affecting NDT efficiency and quality.To address this issue, this paper proposes a precise positioning strategy for penetrations: 1) establishing a coordinate mapping relationship between the inspection equipment's coordinate system (X'O'Y′) and the reactor vessel head's coordinate system (XOY), (2) achieving millimeter-level of the guide cone positioning via machine vision-based object detection, and 3) executing final penetration alignment using the ultrasonic ranging sensor.Experimental validation on a mock-up demonstrates a positioning accuracy of ±0.5 mm, fully complying with NDT requirements. This methodology enhances the reliability of in-service inspection for reactor vessel head penetrations under PWSCC-sensitive conditions.

G. Federici, C. Bachmann, L. Barucca et al.

This paper describes the status of the pre-conceptual design activities in Europe to advance the technical basis of the design of a DEMOnstration Fusion Power Plant (DEMO) to come in operation around the middle of this century with the main aims of demonstrating the production of few hundred MWs of net electricity, the feasibility of operation with a closed-tritium fuel cycle, and maintenance systems capable of achieving adequate plant availability. This is expected to benefit as much as possible from the ITER experience, in terms of design, licensing, and construction. Emphasis is on an integrated design approach, based on system engineering, which provides a clear path for urgent R&D and addresses the main design integration issues by taking account critical systems interdependencies and inherent uncertainties of important design assumptions (physics and technology). A design readiness evaluation, together with a technology maturation and down selection strategy are planned through structured and transparent Gate Reviews. By embedding industry experience in the design from the beginning it will ensure that early attention is given to technology readiness and industrial feasibility, costs, maintenance, power conversion, nuclear safety and licensing aspects.

Taylor E. Grubbs, Igor A. Bolotnov

The two-phase flow within a Boiling Water Reactor steam separator is investigated using an interface capturing method. The simulations are focused on resolving the flow around the first pickoff ring which is the highest contributor to steam carryunder phenomenon. Multiple simulations are conducted of varying levels of resolution to evaluate the capabilities of interface capturing technique for this challenging problem. First, high-resolution simulations of the flow using a simplified <inline-formula><math xmlns="http://www.w3.org/1998/Math/MathML" display="inline"><semantics><mrow><mn>30</mn><mo>°</mo></mrow></semantics></math></inline-formula> wedge are conducted without a swirling velocity field present in the actual system. In order to understand the flow field generated by the separator swirler, secondary simulations of single-phase flow passing through a swirler model are conducted. Using this information, a coarse simulation of the full <inline-formula><math xmlns="http://www.w3.org/1998/Math/MathML" display="inline"><semantics><mrow><mn>360</mn><mo>°</mo></mrow></semantics></math></inline-formula> model was performed, which incorporated the effect of the swirler using a custom inflow boundary condition. Instantaneous carryunder/carryover along with void fraction and film thickness are evaluated at the pickoff ring entrance. Overall, these simulations demonstrate that interface capturing simulations can be an accurate tool for studying full-scale components within nuclear power plants.

Hangbok Choi, J. Bolin, Oscar Gutierrez et al.

Abstract The Fast Modular Reactor (FMR) is a 100-MW(thermal) gas-cooled fast reactor being developed by General Atomics Electromagnetic System with the goal of developing a FMR for flexible and dispatchable power to the U.S. electricity market in the mid-2030s. The conceptual design aims to develop and verify simplified design features. These include an inert helium gas coolant, pellet-loaded fuel rods, installations with air cooling as ultimate heat sink, and small and passive heat removal systems. The goal is to ensure the development of a safe, maintainable, cost-effective, and distributed nuclear energy-generating station. The baseline technologies selected to achieve this goal are a helium coolant that is an inert gas with no chemical reaction with structural components, not activated, single phase, enabling high-temperature operation and a high thermal efficiency Brayton cycle; conventional uranium dioxide (UO2) fuel, which is the most widely used and well-known fuel material, capable of high burnup (100 MWd/kg) and a long fuel life; and silicon carbide composite (SiGA®) cladding and internal structures that are chemically inert in the helium environment, exceptionally radiation tolerant, and being derisked by accident tolerant fuel technology development. The reactor was specifically designed with passive safety features, including high-temperature in-core materials and a reactor vessel cooling system consisting of cooling panels of naturally circulating water. The passive safety of the core was confirmed for the depressurized loss-of–forced cooling accident, which showed the peak cladding temperature at ~1600°C during the transient, which is below the current design limit of 1800°C. The conceptual design of the FMR has been conducted for the reactor system, vessel system, generator and turbomachine, instrumentation and control, residual heat removal system, plant service system, and containment, as well as pre-application licensing documents.

R. Adamson, C. Coleman, M. Griffiths

Abstract The fuel channels and fuel assemblies of all conventional nuclear reactors that generate power from the fission of uranium by thermal neutrons are made from zirconium alloys because of their low thermal neutron absorption cross-section. The dimensional stability, and the ability to predict dimensional changes, of components made from zirconium alloys is important to designers and operators of such reactors because deformation has a consequence for the operability or life of the reactor core. The dimensional changes in zirconium alloys due to neutron irradiation has been the subject of intense study since the inception of the thermal nuclear power reactor. During irradiation zirconium alloys behave differently from most other engineering alloys in that they resist swelling. They do exhibit anisotropic dimensional changes in the absence of an applied stress that depend on the microstructure; this process is called irradiation growth. Like any other material they also exhibit a dimensional response to an applied stress; this process is called irradiation creep. In this review the evolution in measurement methodologies (either from controlled experiments in materials test reactors or gauging of power reactor components) is described together with the results gleaned from such measurements. As measurements have improved and the amount of experimental and operational data has increased, the theoretical basis for modelling creep and growth has also evolved. The history of the evolution in understanding and the ability to predict dimensional changes in zirconium alloys over the past 60–70 years is described and discussed.

LIU Zhu, LONG Jiachen, GAO Yang, GUO Xianglong, ZHANG Lefu

Supercritical carbon dioxide is one of the most promising candidate working fluids for nuclear power plants. Structural materials using in this kind of system face severe challenges due to its high temperature and high pressure. Corrosion, especially oxidation and carburization, will occur, which leads to the failure of material. Traditional metallic materials like austenitic stainless steel, which is regarded as the candidate cladding materials, will break away because of their limited properties. To further improve the corrosion resistance of austenitic stainless steel as the candidate cladding material for supercritical carbon dioxide nuclear reactor, the general corrosion behavior of three kinds of alumina-forming austenitic stainless steel with different Al contents and their Al-free steel was investigated in supercritical carbon dioxide at 650 ℃/20 MPa. The purity of carbon dioxide was 99.99%. All samples were firstly ground, and then polished to eliminate the surface damage regions. Scanning electron microscopy equipped with backscattered electron, electron backscatter diffraction detector and energy dispersive spectroscopy was used for microanalysis. Focused ion beam system was used to prepare the cross-sectional transmission electron microscopy foils. Selected area electron diffraction was also carried out on the transmission electron microscopy foils for revealing the detailed microstructural characterization. Glow discharge optical emission spectrum was used for carburization analysis. The results show that the weight gain of materials decreases with the increase of Al content, and the weight gain of 3.5wt% Al steel is only about 0.022 mg/cm2. The weight gain against exposure time of different materials follows the parabolic growth law approximately, which indicates that the corrosion behavior is mainly controlled by diffusion. When the content of Al is less than 1.5wt%, a dual-layer Fe-rich oxide film is formed on the surface with poor protection. Some pores were observed within it, which can cause oxidation and carburization. Carburization occurs on Al-free steel and the depth of the carburization region can be up to about 12 μm. When the content of Al is higher than 2.5wt%, protective oxide films are formed on the surface. Its outer layer is mainly rich in Cr and the inner layer is rich in Al. The Al layer on 2.5wt% and 3.5wt% Al containing steel is more continuous. These oxides own better corrosion resistance than that formed on 1.5wt% steel. Although more protective oxides are formed, carburization still exists in the oxide film and matrix of this kind of materials, while the thickness of the carburization region decreases to about 6 μm. Higher Al content, above 2.5wt%, can effectively facilitate the formation of protective Al-rich oxide film, which hinders the outward diffusion of Fe and the inward diffusion of C. It further improves the oxidation and carburizing resistance of materials, which will be useful in the supercritical carbon dioxide nuclear power plant.

D. Poston, M. Gibson, T. Godfroy et al.

Abstract The Kilowatt Reactor Using Stirling TechnologY (KRUSTY) was a reactor design, development, and test program to demonstrate the nuclear operation of a Kilopower reactor. Kilopower systems are intended to provide between 1 and 10 kW(electric) in space, or on the surface of planets or moons, with a clear evolution to substantially higher power systems. KRUSTY was a prototype of a 1-kW(electric) highly enriched uranium–fueled Kilopower system. In March of 2018, KRUSTY successfully operated as a fission power system and was the first nuclear-powered operation of any truly new reactor concept in the United States in over 40 years. This paper discusses the design of the KRUSTY reactor along with the philosophy, goals, and engineering work that ultimately led to KRUSTY’s success.

T. Coissieux, J. Politello, C. Vaglio-Gaudard et al.

Abstract CABRI is an experimental reactor at the French Alternative Energies and Atomic Energy Commission (CEA) used to study fuel behavior during reactivity insertion transients. As these transients have a high level of multiphysics, it is important to develop suitable modeling and simulation tools to simulate them in order to be able to optimize testing and control of experimental conditions. This paper focuses on the development and validation of the neutron deterministic APOLLO3® calculation scheme that is included in the CABRI neutronic/thermal-hydraulic multiphysics coupled simulation tool; it represents the first stage of a stepwise validation process for the CABRI multiphysics simulation tool. The neutron calculation scheme is based on a classical two-step approach. The first step consists of a 281-energy-group flux calculation with the TDT-MOC (Method of Characteristics) solver for nuclear data space and energy collapsing for the different CABRI assembly clusters. The biases on a two-dimensional (2D) core neutron calculation due to self-shielding correction and collapsing on a restricted pattern are investigated by means of comparison with a direct full 2D calculation on a quarter core. The second step relies on a three-dimensional (3D) core calculation. Two approaches are presented. The first one consists of a best-effort approach corresponding to a 3D pin-scale description of the core, performing a transport calculation with the SN solver MINARET. And, the second one, a best-estimate approach, which will be implemented for kinetics calculations, relies on solving a simplified transport SPN equation in the solver MINOS with an exact 3D cell description of the core. The best-estimate calculation scheme is then used to analyze three experimental CABRI transients. A stepwise validation process is followed to quantify the calculation biases on physical parameters such as reactivity, reaction rates, and total core power at each step using static reference calculations with the stochastic code TRIPOLI4® or transient experimental data. The next development stage toward a multiphysics calculation scheme will be implementation and validation of coupling with a core thermal-hydraulic model.

Wooseong Park, Yong Hwan Yoo, Kyung Jun Kang et al.

Nuclear marine propulsion has been emerging as a next generation carbon-free power source, for which proper passive residual heat removal systems (PRHRSs) are needed for long-term safety. In particular, the characteristics of unlimited operation time and compact design are crucial in maritime applications due to the difficulties of safety aids and limited space. Accordingly, a compact and long-term-operable PRHRS has been proposed with the key design concept of using both air cooling and seawater cooling in tandem. To confirm its feasibility, this study conducted system design and a transient analysis in an accident scenario. Design results indicate that seawater cooling can considerably reduce the overall system size, and thus the compact and long-term-operable PRHRS can be realized. Regarding the transient analysis, the Multi-dimensional Analysis of Reactor Safety (MARS-KS) code was used to analyze the system behavior under a station blackout condition. Results show that the proposed design can satisfy the design requirements with a sufficient margin: the coolant temperature reached the safe shutdown condition within 36 h, and the maximum cooling rate did not exceed 40 °C/h. Lastly, it was assessed that both air cooling and seawater cooling are necessary for achieving long-term operation and compact design.

SHENG Jiangkun1,2;XU Peng1,*;QIU Mengtong2;DING Lili2;LUO Yinhong2;YAO Zhibin2;ZHANG Fengqi2;GOU Shilong2;WANG Zujun2

In order to investigate the influence of heavy ion fluence on single event upsets (SEU) and the SEU cross-section in NAND Flash memory, as well as the multiple-cell upsets (MCU) due to heavy ion irradiation, experimental studies were performed on two types of 25 nm NAND Flash memory devices. These experiments were conducted at the HI-13 Tandem Accelerator at the China Institute of Atomic Energy and the Heavy Ion Research Facility in Lanzhou (HIRFL). The experiment results revealed that bit upsets were randomly distributed across addresses, hereditary in two tests, reduced insignificantly after short-term annealing and reprogrammable, suggesting that the primary cause of bit upsets is charge loss of the floating gate in the memory cells. The memory cells with 25 nm feature size demonstrated a pronounced sensitivity to heavy ion irradiation, with a single ion strike on the sensitive volume capable of inducing a bit upset. Consequently, when the cumulative fluence remains significantly below the memory array’s density, the bit upset ratio exhibits a nearly linear correlation with the accumulation of heavy ion fluence. The study also identified that the severity of MCU events was proportional to the linear energy transfer (LET) of the incident ion, which led to a non-saturation effect in the SEU cross-section of NAND Flash memory, indicating that the Weibull fitting-based SEU cross-section evaluation method may not be appropriate for NAND Flash memory, or introduce significant errors. When programming the memory array to 55 h, the MCU topological patterns manifested as single vertical columns, double vertical columns, interval vertical columns, and so on, attributable to the interleaved bit line architecture of the memory array. Utilizing the ion track model, the effective radii of ion tracks for Ta, Ti, Al, and F were qualitatively estimated using experimental data from Ta. These estimated values align with the experimental observations of SEU cross-section and MCU statistics. It is postulated that in space applications, when the LET of the incident ion is sufficiently high or when adjacent bits within a byte are in a programming state, heavy ion incidence may induce multiple-bit upsets (MBU) within a single byte, posing additional challenges for error correction. The research further observed a decrease in the SEU cross-section with increasing fluence, particularly with Ti ions. This trend is attributed to the rapid decrease in the number of memory cells within the low voltage region of the programmed state threshold voltage distribution, which are centrally located on adjacent or nearby addresses, and the relatively unchanged growth rate of bit upsets in the remaining majority of memory cells. The relationship between SEU cross-section and heavy ion fluence suggests that irradiation at higher fluence may lead to an underestimation of the on-orbit bit upset rate for NAND Flash memory. Conversely, testing at lower fluence may yield a more conservative or accurate prediction of the on-orbit bit upset rate.

Zhenzhong Liu, Kaixuan Tan, Chunguang Li et al.

Laboratory experiments and point monitoring of reservoir sediments have proven that stable sulfate reduction (SSR) can lower the concentrations of toxic metals and sulfate in acidic groundwater for a long time. Here, we hypothesize that SSR occurred during in situ leaching after uranium mining, which can impact the fate of acid groundwater in an entire region. To test this, we applied a sulfur isotope fractionation method to analyze the mechanism for natural attenuation of contaminated groundwater produced by acid in situ leaching of uranium (Xinjiang, China). The results showed that δ34S increased over time after the cessation of uranium mining, and natural attenuation caused considerable, area-scale immobilization of sulfur corresponding to retention levels of 5.3%–48.3% while simultaneously decreasing the concentration of uranium. Isotopic evidence for SSR in the area, together with evidence for changes of pollutant concentrations, suggest that area-scale SSR is most likely also important at other acid mining sites for uranium, where retention of acid groundwater may be strengthened through natural attenuation. To recapitulate, the sulfur isotope fractionation method constitutes a relatively accurate tool for quantification of spatiotemporal trends for groundwater during migration and transformation resulting from acid in situ leaching of uranium in northern China.

M. Yaseen, Xu Wu

Abstract Recent performance breakthroughs in artificial intelligence (AI) and machine learning (ML), especially advances in deep learning, the availability of powerful and easy-to-use ML libraries (e.g., scikit-learn, TensorFlow, PyTorch), and increasing computational power, have led to unprecedented interest in AI/ML among nuclear engineers. For physics-based computational models, verification, validation, and uncertainty quantification (VVUQ) processes have been very widely investigated, and many methodologies have been developed. However, VVUQ of ML models has been relatively less studied, especially in nuclear engineering. This work focuses on uncertainty quantification (UQ) of ML models as a preliminary step of ML VVUQ, more specifically Deep Neural Networks (DNNs) because they are the most widely used supervised ML algorithm for both regression and classification tasks. This work aims at quantifying the prediction or approximation uncertainties of DNNs when they are used as surrogate models for expensive physical models. Three techniques for UQ of DNNs are compared, namely, Monte Carlo Dropout (MCD), Deep Ensembles (DE), and Bayesian Neural Networks (BNNs). Two nuclear engineering examples are used to benchmark these methods: (1) time-dependent fission gas release data using the Bison code and (2) void fraction simulation based on the Boiling Water Reactor Full-size Fine-Mesh Bundle Tests (BFBT) benchmark using the TRACE code. It is found that the three methods typically require different DNN architectures and hyperparameters to optimize their performance. The UQ results also depend on the amount of training data available and the nature of the data. Overall, all three methods can provide reasonable estimations of the approximation uncertainties. The uncertainties are generally smaller when the mean predictions are close to the test data while the BNN methods usually produce larger uncertainties than MCD and DE.

Mohammed A. Khalaf, C. C. Ban, M. Ramli

Abstract Promoting concrete quality is a logical inevitable choice for humanity in global revolutionary development. This study highlights one of the most important types of concrete, namely, the high density concrete, which is used widely in nuclear power plants and many other modern scientific and industrial applications. It can efficiently absorb and attenuate the dangerous types of radiation. It is low cost. These features make this type of concrete important in many relevant fields and applications. It is used as partition walls for rooms of examination in hospitals, laboratories and others. The attenuation degree of the biologically dangerous types of radiation (such as gamma rays, alpha rays, beta rays and X-ray) during the propagation of these rays through the mass of the shielding materials is in proportion with the atomic mass of the shielding material. Some of these rays can be stopped easily by a piece of paper (like α) or a thin layer of aluminium (such as β-ray). However, some of them (such as gamma rays) has ultrapenetration ability and can penetrate through the living bodies and cause destructive ionisation to biological cells. Thus, the concrete material that can properly shield from such rays is important and is the primary focus of discussion of this review. An overview of the works conducted to produce high density concrete for various engineering and radiation attenuation applications is provided. The important aspects are described in detail, such as the high-density concrete properties, materials used, physical and chemical material’s properties, mechanical properties and durability and the effect of the interfacial transition zone’s density on the properties of heavy weight concrete. The major advantages and contributions of this study are not restricted to the comprehensive review of the past works that led to the development this important type of concrete. The knowledge gap in this field of work is highlighted. An extensive review of literature reveals the difficulties related to the scarcity of work on the drastic reduction of the water/cement ratio to achieve high density concrete with reasonable workability using modern chemical admixtures, the approach of promoting high-density concrete through hybridisations of fibres and the reduction of their permeability by using the nanomaterials as mineral admixtures.

D. V. Shvets, I. A. Mikshin, A. A. Lapkis et al.

The research work considers and analyzes the problem of control valve failure in case of shaft-gear break during normal operation. In order to identify typical deviations and causes of failure of the electrically driven valves, the amplitude-frequency spectra of the current signal taken at one or more phases of the electric motor were analyzed. The method of spectral diagnostics adopted for analysis allows detecting hidden defects of reinforcement not detected in other types of analysis. In order to justify the study, spectral analysis is carried out on control valves installed in the main condensate and feedwater systems of power units WWER-1000 and WWER-1200. As a result of the study, the following relationships were established increase of amplitude speed of electric motor (EM) to -30 dB is a fact of EM operation at increased load, which will lead to wear of thrust-radial bearings; appearance of baseband frequencies in the region of 50 Hz frequency indicates that there are deviations in setting of electric drive torque limiters. The detected defects, in the future, will lead to the failure of the gear shaft and the shutdown of the power unit. The results of the performed work are used to supplement the existing catalog of defects of pipeline valves developed by RI NPE VETI NRNU MEPhI.

T. Pütterich, E. Fable, R. Dux et al.

In the present work, the tolerable impurity level and composition for a reactor plasma using several sets of model assumptions are evaluated. Special care was taken to evaluate a comprehensive and consistent set of atomic data for 35 different elements, such that the impurity level for various elements may be studied as a function of their nuclear charge. The data set may not only be useful for the presented work or for system codes which design fusion reactors, but also for interpretation of bolometric measurements. Additionally, the predictions of the spectral distribution of the radiated power is of high quality such that soft x-ray broadband measurements may be interpreted. In the present work the data is used for predicting the radiated power in a reactor plasma, using a 0D, several variants of a 0.5D model and a realistic 1D ASTRA modelling of a DEMO plasma, i.e. the EU DEMO1 2015 design. The maximal or appropriate impurity content of a reactor plasma for all models can be determined, such that the predictions from a simplistic 0D model can be compared to less simplistic models and a proper reactor simulation. These comparisons suggest that with the simplistic models the impurity content may be estimated within a factor of about 1.5, independent of the realization of the reactor plasma. At the same time this study underlines the sensitivity of the reactor performance on the impurity mixture and especially of the He content of the plasma. Additionally, an extended 0.5D model is presented which is able to predict variations of the fusion yield Q and the He concentration, when both is known for a reference scenario. These predictions prove to be of high accuracy when compared to the 1D ASTRA modelling and thus, allow the net impact of an increased dilution and a simlutaneous temperature rise at constant plasma pressure to be evaluated. Furthermore, the parameter space is scanned with more than 105 model reactor plasmas demonstrating that the use of a low-Z impurity diminishes the possibility of an economical feasible reactor plasma. The main results of the parameter scan are made available via scaling formulae.

Aditi Verma

It is my pleasure to introduce and frame this Nuclear Technology special issue, “The Nuclear, Humanities, and Social Science Nexus: Challenges and Opportunities for Speaking Across the Disciplinary Divides.” This special issue features 13 papers authored by humanists and social scientists who, though each rooted in distinct and rich scholarly traditions, share with each other an interest in the nuclear energy sector. These scholars adopt an intellectually diverse set of theoretical and methodological lenses to examine the work of nuclear energy practitioners and policy makers. The central aim of this special issue is to explore how research findings and insights from the humanities and social sciences can be used to shape and meaningfully inform the work of practitioners and policy makers in the nuclear energy sector and its corresponding areas of research and practice—all of which presently, in many ways, simultaneously face several challenges and opportunities and find themselves at a crossroads. Nuclear energy’s challenges are frequently (and have long been) described as having a significant “social” dimension. These challenges, as interpreted by nuclear engineers, include failures to site nuclear power plants and used nuclear fuel repositories, or more broadly, to secure support and approval for sustaining or expanding the use of nuclear energy. A negative perception of nuclear energy is frequently cited by nuclear engineers as the source of these challenges. Still other problems are believed to be the result of institutional failures and managerial difficulties. These include delays in construction projects and escalation of plant costs, the slow pace of development and commercialization of new nuclear energy technologies, and failures of regulatory institutions. In spite of, or perhaps because of, these challenges, organizations in the nuclear energy sector have, since their inception, proved to be rich research sites for scholars in the humanities and social sciences. In a significant and growing base of scholarship, researchers—historians, political scientists, sociologists, anthropologists, and science and technology studies (STS) scholars—have used a diverse, rich, and increasingly sophisticated set of theoretical and methodological approaches to examine the work of practitioners in nuclear organizations. Some concepts developed by social scientists have proved to be pivotal for the work of practitioners in the nuclear sector. For example, the idea of an organization that is capable of rapid and continuous learning—operationalized by the Institute of Nuclear Power Operations and World Association of Nuclear Operators for the nuclear industry—comes from a long line of sociological and management research on “high-reliability organizations.” Further, the idea that culture can play an important role in ensuring safety also finds its basis in a long tradition of sociological and anthropological research on culture. However, these concepts are often not used as the humanists and social scientists intended or used instrumentally. They undergo modification in their translation from research to practice, and their uptake and use by practitioners and policy makers in the nuclear sector have largely been serendipitous. Further, while social science scholars have produced a growing and increasingly relevant literature, it has not received significant attention from academic and practitioner nuclear engineers, nor has it really made its way into the intellectual canon of nuclear engineering and its pedagogy. This special issue as a whole, and each of the papers in it, seeks to bridge the intellectual divides between the nuclear sector and the researchers who have long studied its work and practices. The purpose of this introductory paper is to synthesize the findings across the papers that appear in this special issue, draw connections across them, and point to some potential areas of future collaboration across nuclear engineering, humanities, and the social sciences. *E-mail: aditi_verma@hks.harvard.edu NUCLEAR TECHNOLOGY · VOLUME 207 · iii–xv · SEPTEMBER 2021 © 2021 American Nuclear Society DOI: https://doi.org/10.1080/00295450.2021.1941663

Seoryeong Park, Jiwhan Boo, Mark Hammig et al.

The mask parameters of a coded aperture are critical design features when optimizing the performance of a gamma-ray camera. In this paper, experiments and Monte Carlo simulations were performed to derive the minimum detectable activity (MDA) when one seeks a real-time imaging capability. First, the impact of the thickness of the modified uniformly redundant array (MURA) mask on the image quality is quantified, and the imaging of point, line, and surface radiation sources is demonstrated using both cross-correlation (CC) and maximum likelihood expectation maximization (MLEM) methods. Second, the minimum detectable activity is also derived for real-time imaging by altering the factors used in the image quality assessment, consisting of the peak-to-noise ratio (PSNR), the normalized mean square error (NMSE), the spatial resolution (full width at half maximum; FWHM), and the structural similarity (SSIM), all evaluated as a function of energy and mask thickness. Sufficiently sharp images were reconstructed when the mask thickness was approximately 2 cm for a source energy between 30 keV and 1.5 MeV and the minimum detectable activity for real-time imaging was 23.7 MBq at 1 m distance for a 1 s collection time.

F. Kurniati, F. P. Krisna, J. Junios et al.

One treatment for cervical cancer is to use radioactive sources that directly target the cancer cell called brachytherapy. This study is aimed to determine dose distribution at phantom pelvis using the DOSXYZnrc Monte Carlo code. The phantom was derived from a CT scan image of the DICOM-type pelvis with a size of 50 × 50 × 28.8 cm obtained from Santosa Kopo Hospital. The source used was Ir-192, which makes an asymmetrical beam with a size of 0.45 × 0.09 × 0.09 cm. Monte Carlo simulation was performed to determine the dose distribution of the Ir-192 source on cervical cancer CT images based on the Manchester system. The Monte Carlo simulation was divided into two models with distance variations on the applicator. Model A used TPS data with a distance between sources of 0.9 cm, while model B had a distance between sources of 0.5 cm. The distribution of dose resulting from the Monte Carlo simulation was analyzed and compared with TPS data. The results showed that at the range of 50 %, dose distribution in model A reaches the end of 3.9 cm. When compared to the range of 50 % dose distribution at the TPS results that reaches the point of 4 cm, it produces a deviation value of 2.5 %, which is still within the tolerance range. Model A and Model B provide different dose distribution. In model B, it reaches 3.86 cm, resulting in a deviation of 1.02 %, which is still within the tolerance range. The resulting γ-index value for the 50 % dose distribution was 2.26, while the whole area's GPR value was 94.13 %. This indicates a difference in dose distribution between the two models. Therefore, the smaller the distance between the sources, the shorter the dose distribution range with relatively more uniform dose distribution.

O.S. Ifayefunmi, O.A. Mirseabasov, B.I. Synzynys

The response changes of the specific growth rate of Lemna minor duckweed was modeled using the logarithms of frond numbers on tritium activity concentration and gamma radiation dose from cobalt 60. The concept of average specific growth rate depends on the general exponential growth pattern, where toxicity is estimated based on the effect on the growth rate. One of the main questions of the effect of the radiation dose on duckweed is how to correlate the effect of beta radiation with the effect of any other radiation for modeling radiation on Lemna minor. Experimental data were extrapolated by utilizing the OECD guidelines.A linear relationship of absorbed dose and activity concentration was obtained for the average dependency growth rate of Lemna minor as D = (0.1257) · A0.585. The dose rate of gamma irradiation from 60Co increases with tritium activity dependence, on the specific growth rate of the Lemna minor duckweed. An increase in the tritium activity causes a decrease in the specific growth rate of the Lemna minor duckweed. It indicates that as the quantity of the beta radiation dose increase in Lemna minor duckweed, a higher quantity of gamma radiation will be required to cause the same effect in the specific growth rate of Lemna minor duckweed. The relation between the inhibition of the Lemna minor seedling growth and gamma and beta radiation dosage agrees roughly with that between the decrease of survival rate or fertility and dosage.

A. Jalil, A. Chetaine, H. Amsil et al.

The aim of this work is to help inform the decision for choosing a convenient material for the PGAA (Prompt Gamma Activation Analysis) collimator plug to be installed at the tangential channel of the Moroccan Triga Mark II Research Reactor. Two families of materials are usually used for collimator construction: a mixture of high-density polyethylene (HDPE) with boron, which is commonly used to moderate and absorb neutrons, and heavy materials, either for gamma absorption or for fast neutron absorption. An investigation of two different collimator designs was performed using N-Particle Monte Carlo MCNP6.2 code with the ENDF/B-VII.1 and MCLIP84 libraries. For each design, carbon steel and lead materials were used separately as collimator heavy materials. The performed study focused on both the impact on neutron beam quality and the neutron–gamma background at the exit of the collimator beam tube. An analysis and assessment of the principal findings is presented in this paper, as well as recommendations.