Subjects: Energy Science >> Energy Science (General) Subjects: Nuclear Science and Technology >> Engineering Technology of Fission Reactor Subjects: Nuclear Science and Technology >> Particle Accelerator Subjects: Nuclear Science and Technology >> Reprocessing Technology of Spent Nuclear Fuel submitted time 2024-06-30
Abstract: Accelerator Driven sub-critical System (ADS) is considered to be the most important candidate for nuclear waste transmutation. We propose a Multi-Target Accelerator Driven System (MTADS) to resolve two longstanding challenges of ADS, namely heat removal and the associated target lifetime, and inhomogeneous power distribution that affects burn-up of the reactor. An 18 mA, 1 GeV proton beam is split into 12 beams by radio frequency cavities and injected to 12 compact targets inside the reactor. With beam power of 18 MW, the sub-critical reactor is driven to 1500 MW thermal power. The peaking factor of the reactor is reduced to 1.7 by optimization of targets number and position for Multi Target Accelerator Driven System. The maximum beam current density is also reduced to 18.5 μA/cm2, which prolongs the beam window lifetime to 12 months with T91 steel. Towards the next generation ADS, the concept of MTADS simplifies the sub-critical system and increases the transmutation efficiency.
Peer Review Status:Awaiting Review
Subjects: Nuclear Science and Technology >> Engineering Technology of Fission Reactor submitted time 2024-06-26
Abstract: Small modular reactors have received widespread attention owing to their inherent safety, low investment, and flexibility. Small pressurized water reactors (SPWRs) have become important candidates for SMRs owing to their high technological maturity. Since the Fukushima accident, research on accident-tolerant fuels (ATFs), which are more resistant to serious accidents than conventional fuels, has gradually increased. This study analyzes the neutronics and thermal hydraulics of an SPWR (ACPR50S) for different ATFs, BeO+UO2-SiC, BeO+UO2-FeCrAl, U3Si2-SiC, and U3Si2-FeCrAl, based on a PWR fuel-management code, the Bamboo-C deterministic code. In the steady state, the burnup calculations, reactivity coefficients, power and temperature distributions, and control-rod reactivity worth were studied. The transients of the control-rod ejection accident for the two control rods with the maximum and minimum reactivity worth were analyzed. The results showed that 5% B-10 enrichment in the wet annular burnable absorbers assembly can effectively reduce the initial reactivity and end-of-life reactivity penalty. The BeO+UO2 -SiC core exhibited superior neutronic characteristics in terms of burnup and negative temperature reactivity compared with the other three cases owing to the strong moderation ability of BeO+UO2 and low neutron absorption of SiC. However, the U3Si2 core had a marginally better power-flattening effect than BeO+UO2, and the differential worth of each control-rod group was similar between different ATFs. During the transient of a control-rod ejection, the changes in the fuel temperature, coolant temperature, and coolant density were similar. The maximum difference was less than 10◦C for the fuel temperature and 2 ◦ C for the coolant temperature.
Subjects: Nuclear Science and Technology >> Engineering Technology of Fission Reactor submitted time 2024-04-27
Abstract: The heavy-water-moderated molten-salt reactor (HWMSR) is a newly proposed reactor concept, in which heavy water is adopted as the moderator and molten salt dissolved with fissile and fertile elements is used as the fuel. Issues arising from graphite in traditional molten-salt reactors, including the positive temperature coefficient and management of highly radioactive spent graphite waste, can be addressed using the HWMSR. Until now, research on the HWMSR has been centered on the core design and nuclear-fuel cycle to explore the viability of the HWMSR and its advantages in fuel utilization. However, the core safety of the HWMSR has not been extensively studied. Therefore, we evaluate typical accidents in a small modular HWMSR, including fuel-salt inlet temperature-overcooling and -overheating accidents, fuel-salt inlet flow-rate decrease, heavy-water inlet temperature-overcooling accidents, and heavy-water inlet mass flow-rate decrease accidents, based on a neutronics and thermal-hydraulics coupled code. The results demonstrated that the core maintained safety during the investigated accidents.
Subjects: Nuclear Science and Technology >> Engineering Technology of Fission Reactor submitted time 2024-04-07
Abstract: Background : For high-fidelity simulations of fluid dynamics in molten salt reactor, even though a supercomputer is able to suppress the period of each simulation, the consequent expense is still prohibitively costly. A possible way to overcome this limitation is the use of Reduced Order Modelling (ROM) techniques. Purpose : Evaluating the accuracy of the ROM methods for reconstructing the velocity and pressure fields. Methods : Two ROM methods based on the Proper Orthogonal Decomposition (POD) with both Galerkin projection, namely FV-ROM (ROM based on Finite Volume approximation ) and SUP-ROM (ROM with supremizer stabilization ), are established for fluid dynamics of molten salt reactor. Then, both methods are tested on the unsteady cases of liquid-fueled molten salt reactor (LFMSR). Results : The FV-ROM demonstrates notable advantages in both velocity prediction and computational efficiency. For laminar and turbulent transient simulations, the average velocity L2 relative errors are less than 0.5% and 0.6%, respectively, with acceleration ratios of approximately 1500 and 1000 times for single time steps. Conversely, the SUP-ROM scheme demonstrates significant prowess in pressure prediction, achieving remarkably low pressure average L2 relative errors of 0.20% and 0.38% for laminar and turbulent transient scenario, respectively. Conclusions : The integration of the SUP-ROM and FV-ROM for fluid dynamics computations of molten salt reactor could significantly enhances computational efficiency and ensure reliability and accuracy of transient simulation.
Subjects: Nuclear Science and Technology >> Engineering Technology of Fission Reactor submitted time 2023-06-18 Cooperative journals: 《Nuclear Science and Techniques》
Abstract: Molecular dynamics method is used to investigate the displacement cascades in Ni-Mo binary alloy. Effects of the irradiation temperature, energy of the primary knock-on atoms and concentration of solute Mo atoms are taken into consideration on radiation damage to the Ni-Mo alloy. It is found that Mo atoms reduce production of the Frenkel pairs at 100 K, while they enhance defect production at 300 K and 600 K. Size of the largest defect clusters decreases with increasing concentrations of Mo atoms (CMo) at 100 K, but it increases with CMo at 300 K and 600 K. Most of the point defects get clustered in cascades leaving only a few vacancies and interstitials isolated.
Subjects: Nuclear Science and Technology >> Engineering Technology of Fission Reactor submitted time 2023-06-18 Cooperative journals: 《Nuclear Science and Techniques》
Abstract: To investigate the corrosion products of Cr in molten FLiNaK salt (46.5 mol% LiF–11.5 mol% NaF–42 mol% KF), the corrosion test of the pure metal Cr was performed in molten FLiNaK salt at 700 ℃ for 200 h. The FLiNaK salt after the corrosion test was thoroughly investigated by X-ray absorption near-edge structure spectroscopy, a transmission electron microscope, and X-ray diffraction. The results demonstrate that the predominant oxidation state of Cr in FLiNaK salt is Cr3+, and the main corrosion product in cooled FLiNaK salt is K2NaCrF6.
Subjects: Nuclear Science and Technology >> Engineering Technology of Fission Reactor submitted time 2023-06-18 Cooperative journals: 《Nuclear Science and Techniques》
Abstract: Based on the design of CLEAR (China LEAd-based Reactor), it is important to study the molten LBE (Lead-Bismuth Eutectic)/water interaction following an incidental steam generator tube rupture (SGTR) accident. Experiments were carried out to investigate the fragmentation behavior of the molten LBE/water contacting interface, with a high-speed video camera to record the fragmentation behavior of 300–600 ℃ LBE at 20 ℃ and 80 ℃ of water temperature. Violent explosion phenomenon occurred at water temperature of 20 ℃, while no explosion occurred at 80 ℃. Shapes of the LBE debris became round at 80 ℃ of water temperature, whereas the debris was of the needle-like shape at 20 ℃. For all the molten LBE and water temperatures in the present study, the debris sized at 2.8–5.0 mm had the largest mass fraction. The results indicate that the dominant physical mechanism of the molten LBE fragmentation was the Kelvin-Helmholtz instability between LBE/water direct contact interface.
Subjects: Nuclear Science and Technology >> Engineering Technology of Fission Reactor submitted time 2023-06-18 Cooperative journals: 《Nuclear Science and Techniques》
Abstract: In large-scale, loosely coupled systems, Monte Carlo criticality calculation always suffers from slow fission source convergence resulting from the high dominance ratio. The fission matrix acceleration method has shown its potential to accelerate the convergence of the fission source in many numerical simulations. In practice, however, instability of this method may be caused by imbalanced precisions of elements of the fission matrix. Hence, an improved method, in which the space mesh used to compute the fission matrix is defined adaptively based on the fission bank in each cycle, is introduced. The proposed method ensures balanced precisions of elements of the fission matrix, so is more stable than the existing fission matrix method.
Subjects: Nuclear Science and Technology >> Engineering Technology of Fission Reactor submitted time 2023-06-18 Cooperative journals: 《Nuclear Science and Techniques》
Abstract: Probabilistic safety assessment (PSA) is important in nuclear safety review and analysis. Because the design and physics of the fluoride salt-cooled high temperature reactor (FHR) differ greatly from the pressurized water reactor (PWR), the methods and steps of PSA in FHR should be studied. The high-temperature gas-cooled reactor (HTR-PM) and sodium-cooled fast reactors have built the PSA framework, and the framework to finish the PSA analysis. The FHR is compared with the PWR, HTR-PM and sodium-cooled fast reactors from the physics, design and safety. The PSA framework of FHR is discussed. In the FHR, the fuel and coolant combination provides large thermal margins to fuel damage (hundreds of degrees centigrade). The tristructural-isotropic (TRISO) as the fuel is independent in FHR core and its failure is limited for the core. The core damage in Level 1 PSA is of lower frequency. Levels 1 and 2 PSA are combined in the FHR PSA analysis. The initiating events analysis is the beginning, and the source term analysis and the release types are the target. Finally, Level 3 PSA is done.
Subjects: Nuclear Science and Technology >> Engineering Technology of Fission Reactor submitted time 2023-06-18 Cooperative journals: 《Nuclear Science and Techniques》
Abstract: As a potential candidate for generation IV reactors, lead-alloy cooled reactor has attracted much attentions in recent years. The China LEAd-based research Reactor (CLEAR) is proposed as the primary choice for the accelerator driven subcritical system project launched by Chinese Academy of Sciences. Lead-bismuth eutectic (LBE) is selected as the coolant of CLEAR owing to its efficient heat conductivity properties and high production rate of neutrons. In order to compensate the buoyancy due to the high density of lead-alloy, fixation methods of fuel assembly (FA) have become a research hotspot worldwide. In this paper, we report an integrated system of ballast and fuel element for CLEAR FA. It guarantees the correct positioning of each FA in normal and refueling operations. Force calculation and temperature analysis prove that the FA will be stable and safe under CLEAR operation conditions.
Subjects: Nuclear Science and Technology >> Engineering Technology of Fission Reactor submitted time 2023-06-18 Cooperative journals: 《Nuclear Science and Techniques》
Abstract: The reduction stripping behavior of Pu(IV) from 30%TBP/OK with hydroxysemicarbazide (HSC) was investigated, and the separation efficiency of HSC and DMHAN-MMH for U/Pu partitioning in Purex process was compared. The results show that HSC can effectively realize the separation of Pu from U; using mixer-settlers to simulate U/Pu separation in 1B bank of PUREX, from 16-stage counter current extraction experiment (in which 6 stages for supplemental extraction, 10 stages for stripping) with flow rate ratio (1BF:1BX:1BS) =4:1:1 in 1B contactor, good result was achieved that the yields are both more than 99.99% for uranium and Pu, the separation factor of plutonium from uranium (SFPu/U) is 2.8�104, and separation factor of uranium from plutonium (SFU/Pu) is 5.9�104. As a stripping reductant, HSC can effectively achieve the separation of Pu from U and the separation effect is nearly the same with DMHAN-MMH, which contributed to replace enough the latter with HSC in the U/Pu separation in Advanced Purex Process Based on Organic Reagent (APOR) process.
Subjects: Nuclear Science and Technology >> Engineering Technology of Fission Reactor submitted time 2023-06-18 Cooperative journals: 《Nuclear Science and Techniques》
Abstract: After the Fukushima Dai-ichi nuclear power plant accident on March 11, 2011, the radioactivity released from the accident was transported around the globe by atmospheric processes. The radioactivity monitoring program on atmospheric particulate in Lanzhou, China was activated by GSCDC to detect the input radionuclides through atmospheric transport. Several artificial radionuclides were detected and measured in aerosol samples from March 26 to May 2, 2011. The peaked activity concentrations (in mBq/m3) were: 1.194 (131I), 0.231 (137Cs), 0.173 (134Cs) and 0.008 (136Cs), detected on April 6, 2011. The average activity ratio of 131I/137Cs and 134Cs/137Cs in air were 13.5 and 0.78. The significant increase of 137Cs activity concentration, one order of magnitude higher than pre-Fukushima accident levels, in ground level aerosol was observed in 2013, as its resuspension from soil. The back-trajectory analysis simulated by NOAA-ARL HYSPLIT shows a direct transfer of the air masses released from Fukushima to Lanzhou across the Pacific Ocean, North America and Europe at the height close to 9000 m AGL. The value of effective dose for inhalation is close to one millionth of the annual limit for the general public.
Subjects: Nuclear Science and Technology >> Engineering Technology of Fission Reactor submitted time 2023-06-18 Cooperative journals: 《Nuclear Science and Techniques》
Abstract: The Chinese Academy of Science has launched a thorium-based molten-salt reactor (TMSR) research project with a mission to research and develop a fission energy system of the fourth generation. The TMSR project intends to construct a liquid fuel molten-salt reactor (TMSR-LF), which uses fluoride salt as both the fuel and coolant, and a solid fuel molten-salt reactor (TMSR-SF), which uses fluoride salt as coolant and TRISO fuel. An optimized 2 MWth TMSR-LF has been designed to solve major technological challenges in the Th-U fuel cycle. Preliminary conceptual shielding design has also been performed to develop bulk shielding. In this study, the radiation dose and temperature distribution of the shielding bulk due to the core were simulated and analyzed by performing Monte Carlo simulations and computational fluid dynamics (CFD) analysis. The MCNP calculated dose rate and neutron and gamma spectra indicate that the total dose rate due to the core at the external surface of the concrete wall was 1.91 μSv/h in the radial direction, 1.16 μSv/h above and 1.33 μSv/h below the bulk shielding. All the radiation dose rates due to the core were below the design criteria. Thermal analysis results show that the temperature at the outermost surface of the bulk shielding was 333.86 K, which was below the required limit value. The results indicate that the designed bulk shielding satisfies the radiation shielding requirements for the 2 MWth TMSR-LF.
Subjects: Nuclear Science and Technology >> Engineering Technology of Fission Reactor submitted time 2023-06-18 Cooperative journals: 《Nuclear Science and Techniques》
Abstract: Neutron-TPC (nTPC) is a fast neutron spectrometer based on GEM-TPC (Gas Electron Multiplier-Time Projection Chamber) and expected to be used in nuclear physics, nuclear reactor operation monitoring, and thermo-nuclear fusion plasma diagnostics. By measuring the recoiled proton energy and slopes of the proton tracks, the incident neutron energy can be deduced. It has higher n/γ separation ability and higher detection efficiency than conventional neutron spectrometers. In this paper, neutron energy resolution of nTPC is studied using the analytical method. It is found that the neutron energy resolution is determined by 1) the proton energy resolution (σEp/Ep), and 2) standard deviation of slopes of the proton tracks caused by multiple Coulomb scattering (σk(scattering)) and by the track fitting accuracy (σk(fit)). Suggestions are made for optimizing energy resolution of nTPC. Proper choices of the cut parameters of reconstructed proton scattering angles (θcut), the number of fitting track points (N), and the working gas help to improve the neutron energy resolution.
Subjects: Nuclear Science and Technology >> Engineering Technology of Fission Reactor submitted time 2023-06-18 Cooperative journals: 《Nuclear Science and Techniques》
Abstract: The present study is to develop a new user-defined function using artificial neural networks intent Computational Fluid Dynamics (CFD) simulation for the prediction of water-vapor multiphase flows through fuel assemblies of nuclear reactor. Indeed, the provision of accurate material data especially for water and steam over a wider range of temperatures and pressures is an essential requirement for conducting CFD simulations in nuclear engineering thermal hydraulics. Contrary to the commercial CFD solver ANSYS-CFX, where the industrial standard IAPWS-IF97 (International Association for the Properties of Water and Steam-Industrial Formulation 1997) is implemented in the ANSYS-CFX internal material database, the solver ANSYS-FLUENT provides only the possibility to use equation of state (EOS), like ideal gas law, Redlich-Kwong EOS and piecewise polynomial interpolations. For that purpose, new approach is used to implement the thermophysical properties of water and steam for subcooled water in CFD solver ANSYS-FLUENT. The technique is based on artificial neural networks of multi-layer type to accurately predict 10 thermodynamic and transport properties of the density, specific heat, dynamic viscosity, thermal conductivity and speed of sound on saturated liquid and saturated vapor. Temperature is used as single input parameter, the maximum absolute error predicted by the artificial neural networks ANNs, was around 3%. Thus, the numerical investigation under CFD solver ANSYS-FLUENT becomes competitive with other CFD codes of which ANSYS-CFX in this area. In fact, the coupling of the Rensselaer Polytechnical Institute (RPI) wall boiling model and the developed Neural-UDF (User Defined Function) was found to be useful in predicting the vapor volume fraction in subcooled boiling flow.
Subjects: Nuclear Science and Technology >> Engineering Technology of Fission Reactor submitted time 2023-06-18 Cooperative journals: 《Nuclear Science and Techniques》
Abstract: In atmospheric dispersion models of nuclear accident, the dispersion coefficients were usually obtained by tracer experiment, which are constant in different atmospheric stability classifications. In fact, the atmospheric wind field is complex and unstable. The dispersion coefficients change even in the same atmospheric stability, hence the great errors brought in. According to the regulation, the air concentration of nuclides around nuclear power plant should be monitored during an accident. The monitoring data can be used to correct dispersion coefficients dynamically. The error can be minimized by correcting the coefficients. This reverse problem is nonlinear and sensitive to initial value. The property of searching the optimal solution of Genetic Algorithm (GA) is suitable for complex high-dimensional situation. In this paper, coupling with Lagrange dispersion model, GA is used to estimate the coefficients. The simulation results show that GA scheme performs well when the error is big. When the correcting process is used in the experiment data, the GA-estimated results are numerical instable. The success rate of estimation is 5% lower than the one without correction. Taking into account the continuity of the dispersion coefficient, Savitzky-Golay filter is used to smooth the estimated parameters. The success rate of estimation increases to 75.86%. This method can improve the accuracy of atmospheric dispersion simulation.
Subjects: Nuclear Science and Technology >> Engineering Technology of Fission Reactor submitted time 2023-06-18 Cooperative journals: 《Nuclear Science and Techniques》
Abstract: The rapid development of nuclear technology has led to more liquid organic radioactive wastes. Different from the regular aqueous radioactive wastes, these liquids possess a higher hazard potential and cannot be disposed through the conventional methods due to their radioactivity and chemical nature. Spent extraction solvent is a kind of common liquid organic radioactive wastes. In this work, tri-butyl phosphate (TBP), which is more difficult to degrade in the spent extraction solvent, was used as the model compound. Influences of reaction conditions on total organic carbon (TOC) removal and the volume percentage of each gas component under supercritical water oxidation (SCWO) were studied. The SCWO behaviors of spent extraction solvent simulants were studied under the optimal conditions derived from the TBP experiment. The SCWO experiments were studied at 400–550 ℃, oxidant stoichiometric ratio of 0–200%, feed concentration of 1.5%–4% and pressure of 25 MPa for 15–75 s. The results show that the TOC removal of the simulants was greater than 99.7% and CH4, H2 and CO were not detected at 550 ℃, 25 MPa, oxidant stoichiometric ratio of 150%, feed concentration of 3%, and residence time of 30 s.
Subjects: Nuclear Science and Technology >> Engineering Technology of Fission Reactor submitted time 2023-06-18 Cooperative journals: 《Nuclear Science and Techniques》
Abstract: Tritium real-time measurement in glovebox or workplace is important to ensure safe operation of tritium. A novel tritium monitor system including an open-walled ionization chamber, an electrometer and an IPC (Industrial Personal Computer) has been developed to measure tritium in gaseous form. Using mesh walls, instead of sealed wall, the open-walled ionization chamber has less tritium absorption and lower memory effect. In addition, tritium gas can diffuse into the chamber’s sensitive region without the assistant of sampling system and ion trap, which are installed at the front-end of commonly used flow-through ionization chambers. Background signal of this monitor system is about 3.7×105 Bq/m3, and after exposed to tritium concentration at about 1011 Bq/m3 for 4h, background of the monitor can recover after purging it several times with dry air. It is suitable for longtime tritium measurements in both glovebox and workplace.
Subjects: Nuclear Science and Technology >> Engineering Technology of Fission Reactor submitted time 2023-06-18 Cooperative journals: 《Nuclear Science and Techniques》
Abstract: The safety analysis code SCTRAN for SCWR (Super Critical Water Reactor) is modified to own the capability to assess the radiation heat transfer with developing a two-dimensional heat conduction solution scheme and incorporating a radiation heat transfer model. The verification of the developed radiation heat transfer model is conducted through code-to-code comparison with CATHENA. The results show that the modified SCTRAN code is successful for that the maximum absolute error and relative error of the surface temperature between results of SCTRAN and CATHENA are 6.1 ℃ and 0.9%, which are acceptable in temperature prediction. Then, with the modified SCTRAN code, the loss of coolant accident with a total loss of emergency core cooling system (LOCA/LOECC) of Canadian-SCWR is carried out to evaluate its "no-core-melt" concept. The following conclusions are achieved: 1) in the process of LOCA, the decay heat can be totally removed by the radiation heat transfer and the natural convection of the high-temperature coolant, even without an intervention of ECCS (Emergency Core Cooling System); 2) The peak cladding temperature of the fuel pins in the inner and outer rings of the high power group are 1236 ℃ and 1177 ℃ respectively, which are much lower than the melting point of the fuel sheath. It indicates that the Canadian-SCWR can achieve "no-core-melt" concept under LOCA/LOECC.
Subjects: Nuclear Science and Technology >> Engineering Technology of Fission Reactor submitted time 2023-06-18 Cooperative journals: 《Nuclear Science and Techniques》
Abstract: A permanent magnet BLDC (brushless direct current) motor is used to move the control rod of a miniature neutron source reactor (MNSR). The BLDC motor drive is modeled using MATLAB/SIMULINK. Two main parts of the modeling are the inverter switching and the current control. Current control with chopping used to minimize the torque ripple of the MNSR control rod drive. Fuzzy logic current control together with soft chopping control shows the best response of all the three strategies. The prototype drive mechanism has an ATmega32 controller and power MOSFET switches. The simulation results are compared with experimental drive mechanism.