Nuclear Systems Research Articles

Turbulent Lorentz heat flow visualization in radiative boundary layer regime

Modern nuclear energy systems often employ the magneto-hydrodynamic and feature radiative heat transfer. Motivated by studying the near-wall transport phenomena in such applications, the present article examines the simultaneous influence of thermal radiative flux and magneto-hydrodynamics on two-dimensional electrically conducting natural convective turbulent heat transfer about a vertical surface under a transverse static magnetic field using the low Reynolds number kinetic energy and dissipation (k – ε) model. An optimized Crank–Nicolson finite difference method is applied to solve the nonlinear and coupled system of Reynolds-averaged boundary layer equations. Numerical computations are conducted to visualize the streamlines and heatlines in laminar and turbulent regimes via mathematical stream and heat functions based on the Bejan approach. A detailed parametric study of the effects of the magnetic field, radiative flux, and turbulent Reynolds number on flow average velocity, temperature, kinetic energy, and dissipation rate is conducted. The simulations reveal that an increase in the magnetic field intensity reduces the average velocity and dissipation rate, whereas an increase in thermal radiation decreases the time mean temperature. The study also includes contour plots of turbulence energy and dissipation rate, along with skin friction coefficient and Nusselt number. The obtained numerical outcomes are compared to previous literature, and a good agreement is found.

Atomistic simulation of brittle-to-ductile transition in silicon carbide embedded with nano-sized helium bubbles

The tensile response of cubic silicon carbide (SiC) bulk containing cavities (voids and He bubbles) has been investigated using molecular dynamic simulations. The formation of cavities in SiC leads to a significant degradation in the mechanical properties of SiC with more influence on material fracture than initial elastic deformation. The brittle-to-ductile transition occurs in cavity-embedded SiC as the pressure in He bubbles increases. This is associated with the deformation mechanism that bond breaking at a low He bubble pressure transfers to extensive dislocation activities at a higher He bubble pressure. The cavities can effectively concentrate stress around them in the direction perpendicular to the tension, which leads to preferred cracking in the region with a higher tensile stress. The failure mechanism as revealed by this study improves understanding of property degradation in SiC that may be useful for applications of SiC in advanced nuclear energy systems.

Heat Transfer Performance of Water-cooled Heat Sink for I&C System Controller Chip in Nuclear Power Plant

To improve the heat dissipation reliability of the nuclear power control system chip, the structure of its water-cooled heat sink is compared and analyzed. Three schemes of water-cooled heat sinks are proposed, and CFD analyzes the heat dissipation performance of the heat sink. The simulation data of the heat transfer coefficient, pressure loss, and heat sink comprehensive coefficient are compared. The results show that the average temperature of the bottom surface of the three heat sinks tends to be flat with the flow velocity increase, and the heat dissipation coefficient Nu is also the same. The S-type heat sink has a higher heat transfer coefficient and lower average bottom temperature. The pressure loss of the S-type heat sink is much higher than that of the other two schemes with the increased flow rate, and the system resistance is large. By calculating the comprehensive coefficient of the heat dissipation process of the three heat sinks and considering the heat dissipation capacity and other factors, it is considered that the circular fin heat sink can be the optimal solution.

Production of nuclei and hypernuclei in relativistic ion reactions

We study the production of small nuclei, nuclei of intermediate mass, and hypernuclei in relativistic nuclear reactions by using dynamical and statistical models. We propose a novel mechanism responsible for the formation of complex nuclei from nucleons and hyperons produced in collisions: In this reaction the excited clusters consisting of baryons can stochastically be formed at a low sub-nuclear density after the dynamical expansion of the nuclear matter. One can describe the nucleation process within such clusters with the statistical models. We suggest the hybrid approach allowing for the consistent description of the experimental data measured in central collisions, which was not possible with other theoretical methods. The important consequence of this novel mechanism is the correlations of the produced nuclear species. The energy limits for describing the fragmentation and nucleation processes in excited finite nuclear systems with statistical models are also obtained.

In-reactor behaviour and economic assessment of enriched gadolinia burnable absorbers

The increasing demand for energy and rising energy prices have increased the importance of the efficiency and economic viability of operating nuclear power plants. One approach to improving the reactivity control in such cases is the use of burnable absorbers such as gadolinium oxide (Gd2O3). While Gd2O3 has proven effective in regulating reactivity in light water reactors, it has certain drawbacks, including the displacement of fissile isotopes from the fuel composition and residual reactivity penalties caused by isotopes other than the main neutron absorbers. In this study, the potential impacts of 157Gd-enriched Gd2O3 on the operation, safety, and economy of nuclear power plants were assessed. Fuel cycle analyses were conducted using the advanced nuclear design code system Studsvik CASMO4/SIMULATE3. Several aspects were analysed, including peaking factors, reactivity feedback parameters, power distribution, and shutdown margin. The results of the analysis revealed that by removing the isotopes responsible for residual reactivity from Gd2O3, it was possible to achieve the same fuel cycle length with lower uranium enrichment levels, enabling a higher concentration of fissile isotopes and yielding economic benefits without compromising safety.

Design and Operation Optimization of a Nuclear Heat-Driven District Cooling System

Carbon-free thermally driven district cooling systems (DCS) can effectively mitigate the excessive electricity consumption and carbon emissions associated with the cooling sector. This study proposes a DCS that employs nuclear heat as the primary energy source. The system comprises three main subsystems: heat station, heat transmission, and cooling station. A heat-only small modular reactor called Teplator, gas boilers, and heat storage are considered to supply the heat required to drive absorption chillers; cold storage and compression chillers are the supplementary units. The technoeconomic aspects of the system are formulated, and an algorithm is developed to determine the optimal design and operation. The method is examined for supplying a typical cooling demand profile with a peak of 2050 MWc. The resulting optimized design includes 11 nuclear plants (150 MWt each), 20 000 MWth heat storage, and 1.9 m diameter heat supply/return pipes. Absorption chillers with a total capacity of 1424 MWc are determined, covering 92% of the total cooling demand, and 244 MWc of compression chillers and 20 000 MWch of cold storage are found to cover the peak and enhance the load following. This system saved 69% of the electricity consumption and carbon emissions and 34% of the costs compared with an electric-based scenario.

Experimental research on liquid entrainment in the inclined up tee branch

The tee branch structure is broadly used in the nuclear power systems, and liquid entrainment in the tee branch has been studied in depth. However, most of the existing research focuses on the vertical tee branch, while the study about the entrainment phenomenon in the inclined tee branch hasn’t formed mature contents. This research aims to analyze the liquid entrainment in the inclined up tee branch. The liquid level measurement method of the Automatic Depressurization and Entrainment TEst Loop (ADETEL) was improved and supplementary experiments were conducted with single-end gas inlet of vertical tee branch. In addition, the test loop was modified to conduct visualization experiments on liquid entrainment experiments of the tee branch with a 45 degree angle. The experimental result reveals that the liquid entrainment phenomenon in the inclined tee branch is significantly different from that in the vertical tee branch. Based on the experimental phenomena and data, a new OLE (Onset of Liquid Entrainment) model is proposed which is well fitted with the existing data. Then, an entrainment rate model is proposed by nondimensionalizing the key parameters affecting the steady-state entrainment, with the error within ± 15% of the experimental data.

Conceptual design of megawatt-level mobile nuclear power system based on heat pipe cooled reactor

Mobile nuclear power system has the advantages such as the well adaptability, the strong sustainability, and the convenient transportation, which can satisfy the energy demand of special environments such as island, and polar region. In this paper, the conceptual design of Mobile Nuclear Power System named MNPS-1000 is proposed, which is consist of the modular heat pipe cooled reactor and combined cycle energy conversion system to achieve the 1MWe output. The design characteristics include the operation lifetime of 3000 days in full power mode, negative temperature reactivity feedback, redundancy in reactivity control, 33% energy conversion efficiency. Due to the efficient heat transfer capability of lithium heat pipes, total of 216 heat pipes can ensure the heat transfer of 3MWt. The modular design of reactor and heat pipe heat exchanger decreases the volume of the system to achieve the compact arrangement in the standard containers with 12.0m×2.5m×2.5m size. The reactor physics and thermal hydraulics analysis, heat pipe design, and energy conversion system analysis are taken out to show the feasibility and performances of this conceptual design.

Different approaches for stress index calculation in elbows of a fusion plant piping system

In the framework of the final verification for the mechanical assessment of a nuclear piping system, critical cross sections, which turned out to belong to an elbow, have been first selected: a study has been subsequently developed, focused on the evaluation of stress index and safety margins for primary stress, starting from what is set out by ASME B31.3 for Process Piping and by RCC-MRx design Code for mechanical components of nuclear installations. The aim was to move to the critical evaluation of the application of the approach “by rules” when finite elements analyses provide both results in terms of internal forces to obtain the nominal stress, and the detailed numerical evaluation of the local stress state and stress concentration. The results show the differences in stress computation that can be obtained by adopting the finite element modeling or by straight application of codes rules and allows to identify the different levels of conservatism and to choose the suitable criteria to be used in an elbow stress assessment.

Direct Bayesian inference for fault severity assessment in Digital-Twin-Based fault diagnosis

For applications in condition-based maintenance of nuclear systems, the assessment of fault severity is crucial. In this work, we developed a framework that allows for direct inference of the probability distributions of possible faults in a system. We employed a model-based approach with model residuals generated from analytical redundancy relations provided by physics-based models of the system components. From real-time sensor readings, the values of the model residuals can be calculated, and the posterior probability distributions of the faults can be computed directly using the methods of Bayesian networks. From the posterior distribution of each fault, one can estimate the fault probability based on a chosen threshold and assess the severity of the fault. By eliminating the discretization and simplifications in middle steps, this approach allows us to leverage the available computational resources to provide more accurate fault probability estimates and severity assessments.

Cladding Failure Modelling for Lead-Based Fast Reactors: A Review and Prospects

Lead-cooled fast reactors (LFRs) are considered one of the most promising technologies to meet the requirements introduced for advanced nuclear systems. LFRs have higher neutron doses, higher temperatures, higher burnup and an extremely corrosive environment. The failure studies of claddings play a vital role in improving the safety criteria of nuclear reactors and promoting research on advanced nuclear materials. This paper presented a comprehensive review of the extreme environment in LFRs based on the fuel performance analyses and transient analyses of reference LFRs. It provided a clear image of cladding failure, focusing on the underlying mechanisms, such as creep, rupture, fatigue, swelling, corrosion, etc., which are resulted from the motions of defects, the development of microcracks and accumulation of fission products to some extent. Some fundamental parameters and behavior models of Ferritic/Martensitic (F/M) steels and Austenitic stainless (AuS) steels were summarized in this paper. A guideline for cladding failure modelling was also provided to bridge the gap between fundamental material research and realistic demands for the application of LFRs.

Coupled neutronics, thermal-hydraulics, and fuel performance analysis of dispersion plate-type fuel assembly in a cohesive way

The nuclear reactor is a multi-physics and multiscale coupling system. The conditions of high temperature and high neutron flux in the reactor pose a significant challenge to the thermodynamic performance of the fuel. To evaluate the plate-type fuel's performance, a coupling calculation of neutron physics code OpenMC, fuel performance analysis code BEEs-Plates, and nuclear reactor system analysis code NUSAC has been implemented based on the open-source MOOSE platform. A novel and efficient method for mesh grids generation and data mapping between external codes and MOOSE was proposed to ensure accurate data exchange. This method can simplify the mesh grids generation process of external codes and accurately transfer data across different mesh grids. Finally, the coupling calculation of the JRR-3 M research reactor fuel assembly was carried out based on the above coupling code, and the neutronics parameters, thermal–hydraulic parameters, and fuel performance-related parameters of the fuel assembly were analyzed. The calculation results show that the maximum fuel burnup will reach 16.91% FIMA after 240 days. With the increase of fuel burnup, the peak power of the fuel assembly is significantly shifted towards both ends of the height direction. The VonMises stress, volume stress, creep strain, and displacement of the fuel will be considerable. Finally, the impact of the coupling calculation on the accuracy of the computational efficiency is preliminarily investigated.

Atlas of historical and proposed nuclear devices and power systems in the Earth-Moon System and wider Solar System

The 2010s witnessed a renewal in international interest of space operations extending outside geosynchronous orbit to encompass the cislunar environment and beyond with invigorated U.S., Chinese, and Russian lunar mission initiatives; planned commercial lunar projects; and coalescing international efforts to reach Mars and push deeper into the Solar System. With the space beyond the near-Earth orbital regimes likely to become competitive and congested in the coming decades, reliable and high-yield means of generating power and propulsion will be needed to enhance mission assurance and operational responsiveness. Nuclear systems represent a promising technology with a rich heritage and provenance of space-based applications to satisfy emerging requirements for electrical power and propulsive impulse for long-term human development of the Earth-Moon system and its planetary neighbors. In this work, an atlas for defining the varying historical or proposed uses of nuclear power in space operations and exploration is presented according to spatial zone within the Earth-Moon and wider Solar System. Depending on the spatial zone, nuclear power will be discussed with respect to applications such as electrical power generation, propulsion, resource extraction, and weapon testing and employment for the space control and planetary defense missions.

Highly efficient group separation of hexavalent actinides from lanthanides through a biphasic cooperative extraction system

The implementation of actinide (An) and lanthanide (Ln) group separation is one of the most critical tasks for the GANEX process to adapt nuclear fuel cycle in the Generation IV nuclear energy system. In this work, we proposed a biphasic cooperative extraction system to fulfill this aim, achieving selective extraction of hexavalent An to the organic phase by a hydrophobic extractant, N,N-di-(2-ethylhexyl)butylamide (D2EHBA), while masking trivalent or tetravalent Ln in the aqueous phase by a hydrophilic ligand, N,N,N’,N’-tetraethyldiglycolamide (TEDGA). Efficient group separation of actinides from lanthanides was achieved with remarkable single-stage An/Ln separation factors over 102. Moreover, the mechanism of An(VI)/Ln(III/IV) selectivity was analyzed by UV–Vis, FT-IR and MALDI-TOF-MS spectroscopic analyses, indicating more hydrophobic nature of An(VI)-D2EHBA complex than Ln(III/IV)-D2EHBA complex as well as much stronger masking effect of Ln(III/IV) than An(VI) by TEDGA contributing striking An/Ln separation.

Effects of cold rolling and heat treatment on microstructure and mechanical properties of 15Cr-15Ni ODS austenitic steel

Oxide dispersion strengthened (ODS) austenitic steels are considered as the promising candidate cladding materials application for advanced nuclear energy systems. However, the microstructure of ODS austenitic steels fabricated by mechanical alloying is usually not very homogeneous. Thermal mechanical treatment is an effective method for controlling microstructure of alloys. In this work, the effect of cold deformation and subsequent heat treatments on the improvement of microstructure homogeneous and mechanical property of a 15Cr-15Ni ODS austenitic steel was systemically investigated by means of electron backscatter diffraction, transmission electron microscopy and tensile test. Results show that the average grain size decreased and the distribution of grains was more homogeneous after optimized thermal mechanical treatment. The deformation of cold rolling and heat treatment temperatures shows obvious influence on the average size of dispersion particle, the proportion of high angle grain boundary, Schmidt factor and tensile property of 15Cr-15Ni ODS austenitic steel.

Effects of solutes on thermodynamic properties of (TMZrU)C (TM = Ta, Y) medium-entropy carbides: a first-principles study

High entropy carbide ceramics have garnered significant interest as a novel class of ultra-high temperature and superhard metallic materials. In the present work, a comparative investigation was conducted for the first time on the stability, mechanical, and thermodynamic properties of two medium entropy carbides (MECs), (TaZrU)C and (YZrU)C, using high-throughput first-principles calculations. Additionally, data from groups IV and V transition metal monocarbides were employed for comparison. The temperature-dependent thermodynamic properties, including bulk modulus (B), constant volume/constant pressure heat capacity (Cv/Cp), Gibbs free energy, volume, entropy, and thermal conductivity, were evaluated using the Debye-Gruneisen model. The results demonstrate that (TaZrU)C and (YZrU)C exhibit similar trends in their thermodynamic properties, with (YZrU)C displaying slightly superior performance as the temperature rises. This work provides valuable insights into the design of innovative high entropy fuels, holding significant implications for the advancement of MEC ceramic fuels in advanced nuclear power systems and nuclear thermal propulsion systems.

Roadblocks to Disarmament in the Nuclear Non-Proliferation Treaty System

Abstract For the last five decades, the states parties to the Nuclear Non-Proliferation Treaty (NPT) have come together every 5 years at Review Conferences to explore how they can advance the Treaty’s goals, including the pursuit of nuclear disarmament. Despite lengthy negotiations, the Review Conferences have failed to reach a consensus final document as many times as they have succeeded and even when a final document has been agreed it has contained few tangible gains for those working towards a nuclear weapon free world. This state of affairs played out again in 2022 when the NPT’s 10th Review Conference ended without agreement and with many states decrying the weakness of the disarmament provisions that had been under discussion. It has been common in the aftermath of Review Conferences for academics to consider the contemporary political issues that stymied progress. While such studies are important, it is also necessary to consider what deeper structural issues may be at play in preventing the goal of nuclear disarmament from being advanced in the NPT context. To this end, this article identifies and analyses three key structural issues with the NPT and NPT Review Conferences that create difficulties for those pursuing a nuclear disarmament agenda. First, it explores issues with the scope and content of Article VI and how it has been interpreted by the nuclear weapon states. Secondly, it looks at the difficulties that arise from the commitment to achieving consensus outcomes at the Review Conferences and, thirdly, it considers the limitations of who actually attends and contributes to Review Conferences.

Pyrocarbon Coating on Granular Al2O3 for HTGR-Type Power Reactor

Fourth-generation nuclear power systems are based on high-temperature gas-cooled reactors, in which the pebble fuel is the primary energy carrier. In this regard, applying protective pyrocarbon coatings on granulated fuel is an essential problem in ensuring the reliability of nuclear power plants. The article’s main idea is to research rational technological parameters of forming a pyrocarbon protective coating on the granules of a nuclear fuel model. For this purpose, granulated Al2O3 with the protective pyrocarbone coating was applied as a fuel model. The article’s aim is to study the effect of thermophysical parameters on applying a protective pyrocarbon coating on granulated Al2O3. During the experimental studies, thermal imaging of the pyrolysis process was used. The scientific novelty of the work is the equilibrium curves for the systems Al2O3:CH4, Al2O3:CH4:N2, and Al2O3:CH4:Ar. Their analysis allowed for evaluating rational thermochemical parameters of the pyrolysis process. As a result, rational thermophysical parameters of coating granulated Al2O3 with a pyrocarbon layer were evaluated, and the practical possibility of applying the pyrocarbon coating to granulated Al2O3 in the electrothermal fluidized bed was experimentally proven. It was shown that nitrogen does not significantly affect the target reaction product under a temperature of less than 1500 K. Also, the rational conditions for the pyrocarbon coating at a pressure of 0.1 MPa were realized at a temperature of 900–1500 K and using argon. Moreover, pyrocarbon was precipitated from hydrocarbon at 1073–1273 K. Overall, the need to add an inert gas for reducing the carbon black formation was proven to prevent a reduction of natural gas efficiency.

Benchmarking exercise of indigenous Monte Carlo neutron transport code PATMOC for hexagonal full core reactor geometry

An indigenous Monte Carlo neutron transport code, named PATMOC, has been designed and developed at Bhabha Atomic Research Centre (BARC) for reactor physics design applications. The code has been extensively tested for its algorithmic accuracy against numerous international numerical benchmarks, has also been validated on practical nuclear reactor systems like Advanced Heavy Water Reactor Critical Facility (AHWR-CF) and Apsara-U pool type research reactor located at BARC, and is now in regular use for day-to-day reactor physics studies. These reactor systems and most of the benchmark problems used for the verifications and validation of the code have square lattice arrangement in the core. In this paper, we present the results of verification of PATMOC code on whole-core reactor system with hexagonal lattice arrangement of fuel elements. In this exercise, we have analyzed the High Temperature engineering Test Reactor (HTTR) whole-core benchmark for integral and differential parameters at both lattice and core level. The results demonstrate the algorithmic accuracy of the code on hexagonal reactor systems also. This exercise will help make use of this code for reactor physics design studies also for the reactor systems with hexagonal arrangement.

Assessment of flow and heat transfer of triply periodic minimal surface based heat exchangers

An efficient intermediate heat exchanger is crucial for ensuring the safe operation, lifespan and energy efficiency of the advanced nuclear systems. Benefiting from the development of additive manufacturing technology, it is possible to design and optimize heat exchangers by introducing triply periodic minimal surface (TPMS) structures. With the excellent mechanical and thermal performance of TPMS, further improvements in the energy efficiency of advanced nuclear systems are achievable. In this study, I-WP surface, Primitive surface based heat exchangers and a printed circuit heat exchanger for accelerator driven subcritical systems are constructed in three dimensions. In order to obtain the potential enhancement of thermal performance of TPMS-based heat exchangers, the fluid flow and conjugate heat transfer characteristics in TPMS heat exchangers, especially for the specific working medium of lead-bismuth eutectic (LBE) are investigated. A parametric analysis is conducted on the key design variables including the solid volume fraction and hydraulic diameter. The results indicate that TPMS-based heat exchangers could achieve about 2–3 times the total heat transfer rate with half the volume of a printed circuit heat exchanger. The TPMS topologies contribute greater heat transfer enhancement on the Helium side than the LBE side, which helps to optimize the thermal resistance allocation in Helium-LBE heat exchangers. An increase of solid volume fraction enhances the convective heat transfer of Helium in I-WP heat exchanger, while homogenizing the local thermal performance. This study provides a database for the design and optimization of TPMS-based heat exchangers, which contributes to the sustainable future target.