In this study, firstly, time-dependent neutronic analyses of a hybrid reactor with a 5 MW/m2 wall load and FLiNaK coolant using TRISO-coated CANDU spent fuel were performed. Neutronic analyses were performed using the XSDRNPM/SCALE nuclear code program. The effects of TRISO coated CANDU spent nuclear fuel on tritium breeding ratio (TBR), total fission rate (Σf), energy multiplication factor (M) and Burn up (BU) were investigated. Over the 48-month time period, the TBR value increased from approximately 1.082 at the beginning to approximately 1.14 at the end of 48th month. M, burn up and fissile fuel rates have also increased continuously during these 48-months. While the initial values of M was approximately 1.97, this value was obtained as 2.351 at the end of period. Also, the BU value was performed as 53.85 GWd/tM at the end of 48-month study period. Secondly, the potential of hydrogen production in the hybrid reactor integrated hydrogen production facility which has the magnesium chloride (Mg-Cl), vanadium chloride (V-Cl) and copper chloride (Cu-Cl) thermochemical cycles were investigated. As a result, the highest amount of the hydrogen production with Mg-Cl was obtained as 11.94 kg/s followed by the Cu-Cl as 5.62 kg/s depends on energy multiplication factor at the end of operation.
This chapter presents the activity conducted by the ITPA topical group (TG) on Diagnostics over about the last 15 years. Following a general introduction of the ITER Diagnostics led by their measurement roles, the document is organized in several subchapters detailing the design support, research and development activity conducted by each of the specialist working groups (WGs) of the TG. Please note that the magnetic diagnostics were supported at the TG without a specific WG. Their status is included in the general introduction. In the following some highlights of the subchapter’s contents are provided. Recent advances in ITER first wall (FW) diagnostics for the measurements of plasma-metallic wall interaction in support of the ITER research plan are reported. An InfraRed imaging Video Bolometer for ITER has been developed and tested on several tokamaks to measure the radiated power loss. A laser-induced breakdown spectroscopy (LIBS) technique which utilizes a pulsed laser beam to ablate locally by forming a crater, will measure local tritium inventory in the FW material. Real-time Residual Gas Analyzers will measure the neutral gas composition in a divertor port and an equatorial port during plasma operation. Due to the full metallic FW environment, the plasma-wall interaction in ITER will face several challenges such as the compromised radiated power and divertor heat flux measurements by reflection. Ray tracing and analysis codes have been developed to eliminate and correct the effects of reflection in the measurements. The characteristics of the reflecting surfaces depending on the roughness and angle of the incidence have been measured by dedicated experiments, and the results were applied to the reflection elimination. For the measurement of the metallic impurity radiation induced by eroded metallic atoms, a vacuum ultraviolet spectrometer has been developed and tested. An extensive thermonuclear diagnostic suite will be required to support the operation of ITER and the planned experimental program for future burning plasma experiments. Due to the harsh environmental conditions, the implementation of diagnostic systems in ITER is a major challenge. These conditions include high levels of neutron and gamma fluxes, neutron heating, particle bombardment. Therefore, the selection and design of diagnostic systems must take into account a number of phenomena previously unseen in diagnostic design. For this reason, the measurement of neutrons and confined or lost fast ions, with particular emphasis on alpha particles, is critical to ITER. The diagnostics associated with these measurements will be important for future plasma-burning experiments at ITER. The high neutron emission and very large plasma size in ITER make neutron diagnostics the main diagnostic method used to measure plasma parameters such as fusion power, fusion power density, ion temperature, energy of fast ions and their spatial distributions in the plasma core. Active spectroscopy techniques are methods where a neutral particle beam is injected into the plasma and information on plasma parameters is extracted from the measurement of line emission resulting from the beam-plasma interaction, either by plasma ions or by beam atoms. Spatial localization is achieved by crossing the beamline and multiple observation lines. The ITER plasma will be a high temperature, moderately dense, fully ionized collisional plasma. The plasma facing surfaces are principally metallic being fashioned from beryllium or tungsten but many other elements, arising from either structural or from operational needs, may enter this plasma. The energy range of the emitted photons range from meV (infra-red) to multi keV (x-rays) and originate from all areas of the plasma volume. The primary role of passive emission diagnostics is to identify what is in the plasma from spectral signatures. Extracting quantitative information from these measurements such as impurity content, ion temperature, rotation, degree of detachment and radiated power depends on calibrated instruments, a physics model of the atomic and molecular processes and plasma transport and an analysis workflow that takes into account environmental effects such as reflections. The particular needs for ITER have prompted a multi-machine, many-year effort to address all these aspects and this chapter reviews the work on diagnostic design, experiments and new analysis techniques. An overview of the laser diagnostics to be implemented on ITER is also provided in this paper. This includes descriptions of the Thomson scattering in the core, edge and divertor regions, polarimetry and interferometry diagnostics used for measuring plasma density and also measurements of helium density in the divertor using Laser Induced Flourescence. Techniques which can allow improvements on current measurements are also addressed in particular expanding poloidal polarimetry measurements to measure field fluctuations and proposed use of dispersion interferometery which has a number of advantages over existing methods. This paper identifies particular areas where further research and testing on existing tokamaks is useful even at this advanced stage to inform the design of diagnostics for ITER. Outstanding areas of concern for the implementation of laser diagnostics, in particular with a view to reliable operation are identified. An overview of the latest developments of microwave diagnostic systems and techniques is given. The primary focus is the contributions for ITER—the next step burning plasma experiment—which is supplemented by describing recent progress of techniques applicable for fusion experiments beyond ITER. The contributions are intentionally kept concise, and are being supplemented by a rich list of references for further studies. Radiation induced effects are receiving continuous and well-deserved attention of the ITER diagnostic community and they are in many cases one of the primary design drivers of the ITER diagnostic systems. The paper summarizes recent progress in this area focusing primarily on the ITER diagnostics but in some cases provides also outlook for the possible solutions for even more demanding radiation environment of fusion reactors beyond ITER. Despite advancements in the area of modeling and simulation of various radiation induced effects, experimental testing in a nuclear environment as close as possible to the target one is still seen as unavoidable for proper qualification of particular diagnostic functional elements. Recent advancement within three diagnostic areas: optical diagnostics, magnetics and bolometers is covered. Encouraging results on qualification of silica glass vacuum window assemblies are presented. In the area of magnetic sensors, progress of irradiation tests performed on ITER in-vessel LTCC inductive sensors is presented with outlook for novel technological approaches to inductive sensors utilizing thick printing and photolithography technologies being highlighted. Summary of advancements in the area of steady state magnetic field sensors based on Hall effect is given. New results of neutron irradiation test of the ITER borosilicate glass inserts for vacuum electrical feedthroughs are summarized finding negligible swelling at target level of neutron fluence. Off-line irradiation tests of fiber optic current sensors for plasma current measurement demonstrated that both for gamma doses up to 5 MGy and a total neutron fluence up to 1015 cm−2, radiation induced changes are still compatible with required measurement accuracy on ITER. The ITER bolometers are given as an example how considering radiation effects may influence the diagnostic design. Finally, outlook for future main R&D directions is outlined. All optical and laser-based diagnostics in ITER will be using mirrors to guide plasma radiation toward detectors, cameras and sensors. In the hostile plasma, radiation and particle environment the optical characteristics of diagnostic mirrors will degrade directly affecting the entire performance of involved diagnostic systems. An assessment of factors affecting mirror performance is provided. Among the prime adverse factors are deposition of plasma impurities, sputtering of mirror surface and steam ingress in the vicinity of mirrors. Within the International Tokamak Physics Activity with active support by ITER central team and domestic agencies, the structured research and development (R&D) program on mitigation of risks for diagnostic mirrors is underway. Within this program the mirror material development, the passive mitigation of mirror degradation by using diagnostic ducts and shutters along with an active mirror recovery program comprising the in-situ mirror cleaning and calibration is underway. Recent developments in diagnostic mirror R&D are described in this Chapter along with an example of their implementation of R&D solutions in ITER Infrared Thermography diagnostic. An assessment of still open engineering and physics questions, considerations on mirror risks during an early phase of ITER operation are given along with an overview of diagnostic mirror evolution in the late ITER operation stage toward the demonstration fusion power plant. Several crucial areas of diagnostic R&D outlined in ITER Research Plan are addressed. The basic control groups in a fusion reactor can be broken-down in five categories: (1) plasma position, magnetic configuration, and plasma current control, (2) profile control and confinement optimization, (3) MHD control and suppression, (4) edge dissipation control, radiation and plasma exhaust control and (5) break-down optimization. These categories are coupled via the physics (a control action in one domain will affect the other domains) and via shared actuators (e.g. ECRH for impurity accumulation avoidance, current density distribution control and MHD suppression). Consequently, a supervisory control system should determine the priority of the various
BackgroundBetavoltaic nuclear batteries, leveraging beta-emitting radioisotopes, offer inherent advantages such as long-term reliability, high energy density, compact form factors, and robust resistance to interference, positioning them as promising power sources for self-powered portable or embedded microdevices.PurposeThis study aims to enhance the conversion efficiency and output power of betavoltaic batteries with comprehensive consideration of the effects of backscattering, depletion region width, diffusion length, and electrode structure on charge collection efficiency, conversion efficiency, and output power.MethodsBy optimizing the device and electrode structure, i.e., introducing a PIN structure with "concentration gradient I- layer", optimizing the depletion region width, doping concentration and electrode materials, and increasing the spacing between electrode grid lines, 63Ni-SiC-based PIN junction betavoltaic batteries were successfully fabricated with higher overall conversion efficiency and output power. Both the Monte Carlo simulations and numerical computations were employed to obtain characteristic parameters of these developed batteries, and their performances were measured by experiments.ResultsThe fabricated batteries exhibit short-circuit currents, open-circuit voltages, output powers, and total conversion efficiencies ranging from 10.29 nA·cm-2 to 13.43 nA·cm-2, 1.32 V to 1.44 V, 11.66 nW·cm-2 to 14.69 nW·cm-2, and 2.24% to 2.82%, respectively. Compared with previous reported work, the open-circuit voltage, fill factor, and overall conversion efficiency increase by an average of 127.50%, 114.47%, and 512.10%, respectively. Moreover, the overall conversion efficiency is higher than those reported in the literature (0.5% to 1.99%).ConclusionsThese results indicate that the conversion efficiency and output power of betavoltaic batteries can be significantly improved by taking above-mentioned optimization measures, providing important theoretical guidance and experimental evidence for the design and fabrication of betavoltaic batteries.
Kęstutis Račkaitis, Povilas Poškas, Robertas Poškas
et al.
Interim storage of spent nuclear fuel is a very important part of the overall nuclear power generation cycle. At Ignalina NPP, spent nuclear fuel is stored in interim storage facilities in specially designed casks before being transferred to a geological repository. The internal structure of spent nuclear fuel casks and the processes involved are quite complex. Therefore, simplifying and optimizing simulations for evaluating decay heat removal from the cask is worthwhile. In this paper, the effect of computer model simplifications on the thermal characteristics of the CONSTOR RBMK-1500/M2 cask stored in building-type and open-type storage facilities is presented. The modeling was carried out using the ANSYS Fluent code. The analysis showed a substantial impact of solar insolation. Also, in the case of the homogenization of the SNF load in the basket, higher temperatures are obtained compared with the case when detailed modeling of the internal basket structure is performed. Hence, it was demonstrated that the homogenization model can be used in safety assessment as a conservative approach for the modeling of decay heat removal from the cask.
Based on academic research and industrial applications over more than 20 years, the Reactor Monte Carlo code (RMC) developed by the REAL (Reactor Engineering Analysis Laboratory) team at Tsinghua University since 2000 has become a powerful, innovative, and versatile simulation platform for nuclear reactor analysis, shielding simulations, criticality safety calculations, fusion neutronics analysis and beyond. Utilizing collaborative and agile development technology, advanced methods and the most cutting-edge algorithms can be tested and implemented in RMC quickly and efficiently. RMC has been deployed on many world-class supercomputers in China and played an irreplaceable role in the design and analysis of commercial nuclear power plants and newly designed types of advanced nuclear reactors. This paper reviews the state-of-the-art technologies developed in RMC in recent years, such as stochastic and continuous-varying media modeling, advanced transient simulation capability, more accurate energy deposition model, etc. Parallel acceleration on heterogeneous architecture supercomputers and machine learning algorithms would be incorporated in ongoing research and future development plans.
Thermal-fluid dynamics is fundamental to a wide range of subjects and technologies, such as energy and power generation, aerospace and astronautics, combined heating, cooling and power technology, nuclear engineering, automotive engineering, mechanical engineering, biological and medical engineering, energy saving and storage, renewable energy, environment, hydrogen and interdisciplinary subjects, such as net zero emission technologies, micro-and nano-fluidics, cooling of micro-electronics and propulsion systems, advanced thermal processes, waste heat energy recovery and others. It plays a crucial role in the development and breakthrough of scientific theories, innovative technologies and revolution of industry, clean energy and environmental issues
Since 2021, EAST tokamak has been operated with full tungsten divertors. Tungsten accumulation has been frequently observed in NBI-heated H-mode discharges, resulting in the degradation of plasma confinement performance. Control of tungsten impurity is thus critical for the maintenance of high-confinement plasmas. In this work the impact of the n = 1 resonant magnetic perturbation (RMP) on the behavior of intrinsic low- and high-Z impurities in EAST H-mode discharges are experimentally studied, utilizing high-performance extreme ultraviolet spectroscopic diagnostics. In the dedicated discharge, ELM mitigation, ELM suppression, H-L back transition, RMP penetration occurs in succession with increasing RMP current (IRMP). When IRMP is below the threshold for H-L back transition, IRMP_H-L = 2.29 kA, increasing influx of C2+ and C3+ ions and decreasing influx of C4+ and C5+ ions are observed simutaneously with enhancement of the RMP field. This opposite time behavior in the influx of C4+ and C3+ ion is then observed to be magnified during the RMP penetration phase. It indicates a impurity screening layer formed between the locations where C4+ and C3+ ions distribute during RMP application based on our previous analysis (W.M. Zhang et al 2024 Nucl. Fusion 64 086004). A large step of increase in C4+ influx after H-L back transition indicates C4+ ion mainly located at bottom of pedestal. A higher RMP coil currents threshold capable of impurity screening is found for high-Z impurity ions of Cu25+, Mo30+, W42+, i. e. 0.53–0.75 kA, than that for C4+ and C5+, i. e. 0.33 kA. Meanwhile, it is found that comparing to C4+ and C5+ ions the decontamination effect by this impurity screening layer is more efficient for these high-Z impurity ions in plasma core region, e.g. up to 70 % reduction in the impurity density, leading to a significant reduction of radiation power. Furthermore, the continuous reduction of core high-Z impurities level both in ELM mitigation and suppression phase proved that this impurity decontamination effect by RMP field is dominant over the impact of ELM activity to core high-Z impurities transport since tungsten is frequently observed to accumulate during original ELM-free phase. Experimental results from this work would contribute to further understanding of the underlying mechanism how the RMP field impacts the impurity transport.
This study is for exploring the diagnostic value of spiral CT combined with autoantibody tests for early lung cancer. In this paper, the total 340 samples is selected, which is divided into 3 groups: 140 early lung cancer patients, 100 patients with benign lung diseases and 100 healthy individuals for physical examination. Three groups of subjects were subjected to multi-slice spiral CT examination and 7 items of autoantibody examination, and the multi-slice spiral CT signs and 7 items of autoantibody levels were observed and compared. Based on the results of the two examinations, a binary logistic regression model was constructed, and this study conducted ROC curve analysis on its value in the diagnosis of early lung cancer. Out of 340 study subjects included in this study, 18 cases had missing clinical data >5%, and 322 valid data were ultimately collected, with a valid data recovery rate of 94.71%. The statistically significant difference did not exist in the halo sign, cavity sign, and Broncho vascular bundle sign among the three groups of subjects on multi-slice spiral CT (P > 0.05), while the statistically significant difference existed in the vacuolar sign, calcification sign, bronchial inflation signs, hairpin sign, lobulation sign, pleural indentation sign, and ground glass nodule sign (P < 0.05). The diagnostic results of multi-slice spiral CT showed 97 cases of benign lung lesions and 130 cases of lung cancer. The statistically significant differences existed in the levels of p53, PGP9.5, SOX2, GAGE7, GBU4-5, MAGE A1, and CAGE among the three groups of subjects (P < 0.05). Based on the results of multi-slice spiral CT and 7 autoantibody tests, binary logistic regression models F1 and F2 were constructed. The Omnibus test results showed that model χ2 = 254.004, P < 0.001. −2 log likelihood = 52.490, Cox & Snell R2 = 0.673, Nagelkerke R2 = 0.909, Hosmer Limeshow test χ2 = 3.311, P = 0.913, indicating good model fit. The ROC curve analysis results showed that multi-slice spiral CT with 7 individual autoantibodies, 7 combined autoantibodies, and 7 combined autoantibodies had good predictive value for early lung cancer (AUC = 0.645–0.989, P < 0.05), with model F2 having the highest predictive value (AUC = 0.989, P < 0.05). Multi-slice spiral CT and 7 autoantibody tests have high diagnostic value for early lung cancer, and combined examination can further improve diagnostic efficiency.
Medical physics. Medical radiology. Nuclear medicine, Nuclear engineering. Atomic power
The interaction between ions should be greatly modified under electronic excitation states, subsequently altering the interactions between materials. We perform a series of first-principles calculations to predict the solution and diffusion behaviors of interstitial hydrogen (H) in tungsten (W) under various electronic excitations. Qualitatively, the solution, diffusion, and trapping behaviors of H in W under various electronic excitation states are basically consistent with those in the ground state. However, it can be found that the solution energy and the migration energy barrier of H decreases as increasing the electronic temperature of system. The Pearson correlation coefficient study shows that there exists a perfect negative correlation between the lattice constant of W and H solution energy induced by lattice distortion. Besides, electronic excitations also make the binding energy of multiple H atoms decrease. That is, when the same number of H atoms are added to the vacancy, the binding energy decreases with increasing the electronic temperature of system. Based on these calculation results, we can infer that electronic excitations make dissolved H atoms more active in W system. This may, to some extent, allow dissolved H to migrate around and not aggregate so easily, thus reducing the production of H bubbles. Therefore, in quantitative terms, the electronic excited states have a certain effect on the H behavior in W.
Reactor design requires safety studies to ensure that the reactors will behave appropriately under incidental or accidental situations. Safety studies often involve multiphysics simulations where several branches of reactor physics are necessary to model a given phenomenon. In those situations, it has been observed that the neutron transport part is still a bottleneck in terms of computational times, with more than 80% of the total time. In the case of hexagonal lattice reactors, transport solvers usually invert the discretised Boltzmann equation by discretising the regular hexagon into lozenges or triangles. In this work, we seek to reduce the computational burden of the neutron transport solver by designing a numerical spatial discretisation scheme that would be more appropriate for honeycomb meshes. In our past research efforts, we have set up interesting discretisation schemes in the finite element setting in 2D, and we wish to extend them to 3D geometries that are prisms with a hexagonal base. In 3D, a rigorous method was derived to shrink the tensor product between 2D and 1D bases to minimum terms. We have applied these functions successfully on a reactor benchmark—Takeda Model 4—to compare and contrast the numerical results in a physical setting.
Estimation of both the surface tritium and tritium in bulk in key materials of fusion reactors is of great importance for tritium safety and the management of tritium-contaminated material. For the further quantitative analysis of tritium in solids, the elaborate BIXS spectra were calculated based on Monte Carlo simulation. Four types of tritium depth profile were considered to evaluate the quantitative estimation method. It is found that the attenuation of X-rays in tungsten depends on both the energy of X-rays and the depth of X-rays. The evaluation of tritium amount in surface layer indicated that the intensity of Ar(Kα) peak could be used to evaluate the surface tritium within 400 nm from the surface in most cases and the deviations were less than 9% in the calculation. The intensity of both W(Lα) X-rays and the high energy X-rays can be employed to roughly estimate the total tritium amount. For linearly decreasing and exponentially decreasing distribution, the maximum calculation deviations were 24.9% and 28.8%, respectively. While for Gaussian distribution, the maximum deviations were 146% and 53%, respectively. And it can also be used for tritium estimation in other materials.
A novel Fission Diffusion Synthetic Acceleration (FDSA) method is developed and implemented as a part of a hybrid neutronics method for source convergence acceleration and variance reduction in Monte Carlo (MC) criticality calculations. The acceleration of the MC calculation stems from constructing a synthetic operator and solving a low-order problem using information obtained from previous MC calculations. By applying the P1 approximation, two correction terms, one for the scalar flux and the other for the current, can be solved in the low-order problem and applied to the transport solution. A variety of one-dimensional (1-D) and two-dimensional (2-D) numerical tests are constructed to demonstrate the performance of FDSA in comparison with the standalone MC method and the coupled MC and Coarse Mesh Finite Difference (MC-CMFD) method on both intended purposes. The comparison results show that the acceleration by a factor of 3–10 can be expected for source convergence and the reduction in MC variance is comparable to CMFD in both slab and full core geometries, although the effectiveness of such hybrid methods is limited to systems with small dominance ratios.
A tungsten Langmuir probe exposed in the JET divertor during the ITER-like wall campaigns (ILW) has been studied to evaluate changes in mechanical properties and microstructure. The tip of the probe that was exposed to plasma was cross-sectioned and polished for post mortem analysis. Analysis involved a comparison with a non-exposed probe to determine the effect of plasma exposure on material microstructure and mechanical properties. Visually the probe appeared to have melted and re-solidified during its time in the vessel. Secondary electron (SE) images of cross sections showed the formation of bubbles near the exposed surface that ranged from 50 µm to sub-micron sized. Electron backscatter diffraction (EBSD) revealed that the average grain size had increased from 33 µm to 570 µm. The investigation also showed that hardness had increased from 5.2 to 6.1 GPa and pop-in behaviour was supressed after exposure. This was initially attributed to the uptake of deuterium (D) but nuclear reaction analysis (NRA) indicated that no deuterium remained in the sample and hinted that some other type of defect was modifying the mechanical properties.
The Uranium/Molybdenum metallic fuel has been proposed as promising advanced fuel concept especially in the dispersed fuel geometry. The fuel is manufactured in the form of small fuel droplets (particles) placed in a fuel pin covered by a matrix. In addition to fuel particles, the pin contains voids necessary to compensate material swelling and release of fission gases from the fuel particles. When investigating this advanced fuel design, two important questions were raised. Can the dispersed fuel performance be analyzed using homogenization without significant inaccuracy and what size of fuel drops should be used for the fuel design to achieve optimal utilization? To answer, 2D burnup calculations of fuel assemblies with different fuel particle sizes were performed. The analysis was supported by an additional 3D fuel pin calculation with the dispersed fuel particle size variations. The results show a significant difference in the multiplication factor between the homogenized calculation and the detailed calculation with precise fuel particle geometry. The recommended fuel particle size depends on the final burnup to be achieved. As shown in the results, for lower burnup levels, larger fuel drops offer better multiplication factor. However, when higher burnup levels are required, then smaller fuel drops perform better.
Nicholas Crowder, Joomyung Lee, Abhinav Gupta
et al.
Abstract Designing piping systems for nuclear power plants involves engineers from multiple disciplines (i.e., thermal hydraulics, mechanical engineering, and structural/seismic) and close coordination with the contractors who build the plant. Any design changes during construction need to be carefully communicated and managed with all stakeholders in order to assess risks associated with the design changes. To allow the quick assessment of building and piping design changes through a streamlined building-piping coupled analysis, this paper presents a novel interoperability solution that converts bidirectionally between building information models (BIMs) and pipe stress models. Any design changes during construction that are shown in an as-built BIM are automatically converted into a pipe stress model. Any further design changes due to building-piping interaction analyses are converted back to the BIM for the contractor and other designers to access the latest model. Two case studies are presented to illustrate the bidirectional conversion that allows an integrated coupled analysis of the building-piping system to account for their interactions.
Since the Fukushima nuclear accident, the safety of nuclear fuel under accident conditions has been paid more and more attention. As an accident-tolerant fuel, coated particle fuel has become a research hotspot. In this fuel element, a large number of fuel particles are dispersed in non-fissile matrix materials. The National Oak Ridge Laboratory of the United States proposed a design of fully ceramic micro-encapsulated fuel (FCM) that can be used in pressurized water reactors (PWR). As a promising accident-tolerant fuel, FCM is composed of tri-structural isotope (TRISO) particles and matrix. TRISO fuel particles are divided into five layers from the inside to the outside, followed by fuel core, carbon buffer layer, inner pyrolysis carbon, SiC ceramic layer and outer pyrolysis carbon. TRISO particles and non-fissile matrix are made into cylindrical fuel pellets and filled into the fuel rods. The fuel can withstand higher temperature and maintain the integrity of the fuel element under some extreme accident conditions. The double heterogeneity is introduced in the neutronic model of this fuel. The first heterogeneous effect consists of the fuel rod and its surrounding moderator. The second heterogeneous effect is caused by the random distribution of fuel particles in non-fission matrix. This DH effect brings great challenges to the existing resonance calculation methods. In fact, the Monte Carlo method has the ability to accurately simulate the DH problem. However, Monte Carlo calculation takes a lot of time and is difficult to be competent for time-related burnup calculation. In order to deal with resonance calculation under DH effect from the perspective of deterministic theory, researchers around the world have proposed many feasible methods. However, to implement these methods, existing lattice physics code need to be modified and validated. To solve the above problems, the traditional subgroup method was combined with the DH resonance integral table from the perspective of DH effect. The DH resonance integral tables for 238U and 235U were made respectively for FCM fuel. In the process of developing the DH resonance integral table, the Sanchez-MOC method was used for transport calculation. In order to obtain accurate resonance cross sections, the method of ultrafine group coupled with Sanchez-Pomraning was chosen to calculate the DH problem. In the process of transforming the DH resonance integral table into the subgroup parameters, the fitting method with Pade approximation was used to calculate the subgroup parameters. Finally, the sub-group parameters combined with the traditional MOC solver could recover the real XS under DH condition on the conventional fuel mixing problem. Numerical results show that considering the integral table generated in double heterogeneous system, under the same transport conditions and within the applicable range of the integral table, the absolute deviation caused by the subgroup resonance part to the keff calculation can be kept within 200 pcm. The significance of this work is that for conventional lattice physics code that are not suitable for modification, such as HELIOS, it is possible to directly perform the calculation of the dispersed particulate fuel problem by modifying the resonance integral table and subgroup parameters online.
The disposal area of a deep geological repository (DGR) for the disposal of spent nuclear fuels (SNFs) is estimated considering the spacing between deposition holes and between disposal tunnels, as determined by a thermal analysis using the decay heat of a reference SNF. Given the relatively large amount of decay heat of the reference SNF, the disposal area of the DGR is found to be overestimated. Therefore, we develop a computer program using MATLAB, termed ACom (Assembly Combination), to combine SNFs when stored in canisters such that the decay heat per canister is evenly distributed. The stability of ACom was checked and the overall distribution of the decay heat per canister was analyzed. Finally, ACom was applied to disposal scenarios suggested in the conceptual design of a DGR for SNFs, and it was confirmed that the decay heat per canister could be evenly distributed and that the maximum decay heat of the canister could be much lower than that of a canister estimated using a reference SNF. ACom can be used to improve the disposal efficiency by reducing the disposal area of a DGR for SNFs by ensuringg a relatively even distribution of decay heat per canister.
Gennady G. Kulikov, Anatoly N. Shmelev, Vladimir A. Apse
et al.
The objectives of the article are (1) to show the nuclear and physical causes of hard γ-quanta in the U-232 decay chain, (2) to propose tactics for handling uranium containing U-232, and (3) to assess the efficiency of its protective γ-barrier against uncontrolled proliferation. The authors show the general picture of the decay chains of U-232 nuclide transformations, on which the protection of uranium from its uncontrolled proliferation is based. During the decay of nuclei, their emission of α- or β-particles is only the first stage of the most complex process of rearrangement of both the internal structure of the nucleus itself, which consists in the rearrangement of the neutron and proton shells and the levels of its excitation, and in the rearrangement of the electron shells of the atom. As a rule, the daughter nucleus is in a highly excited state, which is removed by the emission of hard γ-quanta and internal conversion electrons. After the second case, the remaining excitation of the atom is removed by the emission of characteristic γ-quanta and Auger-electrons with characteristic γ-quanta. In addition, explanations are given for the quantum-mechanical reasons for the hard γ-radiation of Tl-208 and Bi-212, which complete the U-232 decay chain. The authors also proposed a tactic for handling uranium containing uranium-232. Since the hard γ-quanta of Tl-208 and Bi-212 appear only at the end of the U-232 decay chain, after its chemical purification from its decay products, U-232 itself does not pose a radiation hazard; therefore, at this time it is advisable to conduct all necessary operations for transporting the material to the plant, fabricating uranium-based fuel containing U-232, and transporting this fuel to the nuclear facility where it will be used.