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Silicon carbide (SiC) is an important material for shielding nuclear fission products in tri-structural isotropic (TRISO) particles for high-temperature gas-cooled reactors. It has an important influence on the safety performance of TRISO particles. Therefore, it is necessary to study the irradiation behaviors and mechanical properties of the SiC layer of TRISO particles. The irradiation process and nanoindentation process can be described accurately by molecular dynamics simulations. It is helpful to analyze the irradiation behavior and mechanical properties of materials. In this paper, molecular dynamics simulations were used to study the irradiation behavior of equiaxed polycrystalline and long-axis polycrystalline SiC layers obtained from experiments. The calculated theoretical values of the irradiation swelling degree and mechanical properties of the SiC layer are in good agreement with the experimental values. It shows the strong applicability of the potential functions and other simulation system parameters used in this paper. The irradiation behavior of the SiC layer was studied in detail by swelling degree, density, atomic structure type, and point defect evolution. The results show that the grain structure of the SiC layer has no significant effect on its irradiation swelling and amorphization. The amorphization is not directly converted from the crystal structure to the amorphous structure during the irradiation process. The crystal structure is converted into an intermediate state structure and then converted into an amorphous structure. The point defects are mainly C vacancies, Si interstitial atoms, and C antisite atoms in the early stage of irradiation. The difference between the ratio of C point defects and Si point defects gradually disappears after the irradiation dose tends to be saturated. The evolution of amorphization and point defects tends to start nearby the grain boundaries. Irradiation reduces the mechanical properties of SiC layers. It no longer has a significant effect on the mechanical properties of the SiC layer after the irradiation dose tends to be saturated. The changes in the mechanical properties of the SiC layer before and after irradiation were analyzed by load-depth curve and stress & strain. The results show that the decrease in mechanical properties of the SiC layer is closely related to the mechanical behaviors under external force. For example, the bearing capacity and plastic deformation degree decrease, and the stress & strain distribution are disordered. The results provide a quantitative explanation for the irradiation behavior of the SiC layer of TRISO particles and the change of mechanical properties under irradiation. It is helpful to understand the irradiation behavior and relationship between the irradiation behavior and mechanical properties of the SiC layer.

The mathematical model of a radio frequency (RF) system incorporates crucial characteristic parameters of an RF cavity, including the cavity bandwidth, the resonant frequency, and Lorentz force detuning factors, which is very important for the cavity performance evaluation, the control system simulation, and the optimization of low-level control algorithms. A network analyzer is usually used to measure scattering parameters and subsequently construct mathematical models of a RF system. However, the measurement steps are tedious and fail to identify the system model online. Therefore, an advanced online identification algorithm was described for RF cavity systems in this paper, which simulated the operational principles of a network analyzer. This algorithm was integrated within a digital low-level RF (LLRF) system and consisted of a high-precision numerically controlled oscillator, a frequency control word generator module, an amplitude-phase calculation module, and a sweep frequency amplitude-phase processing module. With the assistance of the Matlab/Simulink platform, which provides an ideal environment for simulation and analysis, the deployment of this algorithm was ultimately realized on the field-programmable gate array (FPGA) chip of the low-level system, accomplishing the online identification of RF systems. Then, the sweeping algorithm was verified on a superconducting cavity of the China Accelerator Driven System Front-end Demo Linac at the Institute of Modern Physics. During the verification process, the half-bandwidth of all cavities in CM1 was accurately measured using an online identification algorithm, and the half-bandwidth of all cavities corresponding to CM1 was also measured utilizing a network analyzer. The results of half-bandwidth measured by the two methods are in good agreement, strongly supporting the validity of the online identification algorithm for cavity system models proposed in this paper. Furthermore, we successfully measured the distortion curve of cavity frequency response caused by Lorentz force detuning. Finally, the influence of Lorentz force detuning variation on the measurement of the critical cavity parameters was discussed in this paper. As the RF peak electric field increases, the radiation pressure within the cavity rises proportionately, leading to a corresponding increase in Lorenz force detuning. This article demonstrates that when the RF peak electric field of the cavity operating is less than 16 MV/m, Lorenz force detuning has a relatively limited impact on the bandwidth measurement of the cavity. Conversely, once the peak acceleration electric field surpasses this threshold, Lorenz force detuning exerts an unnegligible effect on the measurement of the half-bandwidth of the cavity. The specific mechanisms of the Lorenz force detuning’s influence on the half-bandwidth of the cavity will be further studied in future research.

The liquid metal cooled fast reactor has significant strategic significance for the innovative development of nuclear energy, and the development of fuel performance analysis programs suitable for liquid metal cooled fast reactors is of great significance for the design and safety analysis of fast reactors. Fuel elements are one of the most fundamental components of nuclear reactors, and analyzing their performance is one of the most challenging tasks in reactor development. The fuel element performance analysis program LoongCALF based on the multi-physics coupling platform MOOSE was developed in this paper. This program aimed to simulate the long-term service performance evolution of fuel elements under steady-state conditions, providing scientific basis for the design, service life prediction, and safety evaluation of LMFR fuel elements. The program was aimed at metal cooled fast reactor fuel elements, and based on finite element method and JFNK method, solved the thermal mechanical coupling equation of nuclear reactor fuel, and obtained the spatial distribution and temporal variation of physical quantities such as temperature, stress, strain, and fission gas release. The program used a 1.5-dimensional modeling method to perform multi-physics field coupling calculations on fuel rods. In addition, the program supported one-dimensional, two-dimensional, and three-dimensional mesh modeling calculations to meet the construction requirements of fuel element models of different sizes and shapes. The program adopted a modular design with separate material modules. Currently, the fuel types supported by the program include UO2 and MOX, the cladding materials include HT9 and 1515Ti, and the coolant materials supported sodium, lead, and lead bismuth. To verify the accuracy of the program, two benchmark examples were designed for numerical simulation and calculation in this paper. Example 1 used UO2 as the core material and 1515Ti as the cladding material; Example 2 used MOX as the core material and HT9 as the cladding material. By comparing and calculating with the Fiber-Oxide program of the Chinese Institute of Atomic Energy, it is found that the two programs show good consistency in key parameters such as core cladding temperature, displacement, stress, and fission gas release fraction. Although there are certain differences in the handling of repositioning and fracture models, resulting in some deviation in the initial calculation results, considering the limitations of the empirical models used in fuel performance analysis programs, this difference is still within an acceptable range. In summary, the LoongCALF program can accurately simulate the fuel behavior and key parameter evolution inside fuel elements under steady-state operating conditions of liquid metal cooled fast reactors. In the future, we will rely on more practical measurement cases to conduct deeper verification and optimization of the LoongCALF program, to ensure its accuracy and reliability in fast reactor design and safety analysis.

The deterministic two-step method for the fast reactor neutronics calculation, composed of cross-section homogenization and diffusion or transport core calculations, was widely applied in the fast reactor engineering design and analysis field. The homogenized cross-section calculation method based on Monte Carlo with continuous-energy and fine geometry can provide high-precision cross-sections for advanced fast reactors. The current development status and trends of coupling Monte Carlo-generated homogenized cross-sections with diffusion and transport core calculations were briefly reviewed in this paper. The methods discussed include the Monte Carlo flux-volume homogenization method, the superhomogenization equivalence technique (SPH), and the Monte Carlo flux-moment homogenization method (MHT). The MET-1000 metal fuel fast reactor is used as a benchmark. The SPH equivalent techniques are widely used to preserve the reaction rates of a reference heterogeneous model and a homogenous model. In this paper, the SPH was applied to the control rods' cross-section address to improve the diffusion core calculations. This equivalence technique reduces the overestimation of the control rod worth using the diffusion core solver from 13.5% to 0.35% and improves power distribution prediction accuracy. With SPH correction, the MC/diffusion in this work exhibits about

The method of characteristics has very strong geometric adaptability and can be used to solve the 3-D whole core problem of high temperature gas-cooled pebble-bed reactor. However, it suffers from the disadvantages of thousands of iterations and slow calculation speed. At present, there have been some attempts to reduce calculation time for method of characteristics, including coarse mesh finite difference (CMFD), general coarse mesh rebalance (GCMR) and general coarse mesh finite difference (GCMFD). The CMFD method is applied to realize the acceleration of regular region like squares. Furthermore, to break through the limitation of macroscopic regular geometry for the numerical acceleration method, researchers successively propose the GCMR method and the GCMFD method, but these methods are very complex to implement in irregular geometries and some approximation factors must be introduced. Based on the analysis of more dense tracks are arranged for the 3-D model, the sparse macro-track transport acceleration method was proposed, which greatly reduces the calculation amount of the acceleration equation and obtains a very good acceleration efficiency without reducing the accuracy. The 2-D method of characteristics divides the angular space into M azimuthal angles and P pole angles. The 3-D method of characteristics generally uses a level symmetric quadrature set. In 2-D geometry, the angle is relatively more dense at the poles. As a result, the 2-D macro-track transport acceleration can select a small number of azimuth angles with reflection symmetry and one pole angle as the solution directions of the acceleration equation. The sparse macro-track transport acceleration method implemented in this paper used the same angles as the method of characteristics. The number of tracks for 3-D problems is an order of magnitude higher than for 2-D problems. That is why reducing the number of 3-D tracks is a promising method for reducing the amount of accelerated equation calculation. This paper applied the macro-track transport acceleration method to 3-D whole core problem of high temperature gas-cooled pebble-bed reactor to solve the problem of irregular geometric numerical acceleration. The sparsity, macro-track length limit, and the maximum iteration step limit of the macro-track method are the main factors that affect the acceleration performance. The selection method of acceleration parameters was studied through two benchmark problems. Excellent speedup ratio can be obtained by setting the sparsity to 3-5, the macro-track length to around 2.0 cm, the acceleration iteration steps limit to 20-60. The results of 3-D small light water reactor benchmark and simplified pebble-bed reactor core model show that about 7 times the time speedup ratio and about corresponding 20 times the iterative step speedup ratio can be achieved.

The nuclear reactor system is complex and the operating environment is harsh, resulting in the complex phenomena of multi-physics coupling. The multi-physics coupling codes developed in the early stage shows limitations on codes' scalability and generality. Therefore, it is of great significance to build a multi-physics coupling framework and conduct research on key technologies in coupling problems, which may accelerate the development process of autonomous multi-physics coupling platform in China. In this paper, the multi-dimensional and multi-physics coupling finite element analysis platform for nuclear reactor developed by XJTU-NuTHeL was introduced. The main work consisted of the development of thermal-hydraulic model, the research of fuel performance analysis technology and the establishment of multi-physics coupling framework. In terms of thermal-hydraulic calculation, XJTU-NuTHeL conducted a series of studies on pressurized water reactors and advanced reactors grounded in the advanced multi-physics coupling framework, and developed the nuclear reactor system safety analysis code, NUSAC. In addition, a subchannel analysis model tailored for liquid metal fast reactors was established, and the fully coupled subchannel transient analysis code, FLARE, was developed. NUSAC and FLARE were then verified against relevant codes and experimental data. In the realm of fuel performance analysis, considering the wide application of finite element method in solid mechanics and its versatile modeling capabilities, XJTU-NuTHeL developed a fuel performance analysis code, BEEs, based on finite element method. The code could not only conduct multi-physics coupling analysis for traditional rod fuels under steady and transient conditions, but also extends its applicability to accident tolerant fuels and other fuels with diverse geometric shapes. This paper focused on the study and analysis of coated particle dispersed fuel and plate type fuel. The multi-scale simulation results of coated particle dispersed fuels, as well as the thermomechanical and corrosion behavior of plate type fuels were shown. In the context of multi-physics coupling analysis, the efficiency and accuracy of different mesh grid mapping schemes were studied and a multi-physics coupling framework was established. An example of the framework was then presented, showcasing the integration of the fuel performance code BEEs, the Monte Carlo neutron physics code OpenMC, and reactor system safety analysis code NUSAC. The keys parameters of mechanics, thermal-hydraulic and neutronics were obtained and analyzed through the coupling different codes. The multi-dimensional and multi-physics coupling finite element analysis platform built in this paper can provide a strong support for the high-fidelity numerical simulation of nuclear reactor multi-scale and multi-physics coupling.

To prevent severe safety accidents caused by primary coolant leakage in pressurized water reactors (PWRs), it is necessary to monitor the leakage rate of key parts of the reactor coolant pressure boundary (RCPB) in order to take timely countermeasures. The primary coolant of PWRs is usually high-pressure supercritical water containing boron, which can produce 18F in the high neutron flux environment of the reactor. The leak rate of the RCPB can be inferred by measuring the activity concentration of the radioactive 18F aerosol produced by the coolant leaking into the containment. A measurable sample source is usually obtained by extracting sample gas containing 18F aerosol from the containment and filtering it in the monitor. The measurement method used by the monitor is to screen the γ signal by using the β signal as the gate signal, which is actually a γ relative measurement method using the coincidence method to reduce the background. The electronic design of the monitor prevents the β-channel signal from being recorded, which makes it impossible to achieve the absolute measurement as with conventional coincidence methods. To ensure the accuracy of the monitoring results, the 18F aerosol monitor needs to be calibrated. In this paper the monitor was calibrated using a standard 18F surface source produced by the intermittent spot method on a filter paper base. By means of the decay of the 18F source and the automatic recording function of the monitor, the indication number of the monitor was measured at six source activities of different orders of magnitude, i.e. the count of the energy spectrum at the position of the 511 keV γ-window. The data were used to obtain the activity response and uncertainty of the monitor's readings and to verify the feasibility of such a calibration scheme on the monitor. Two standard point sources of 22Na and 68Ge/68Ga were also used to calibrate the monitor to verify the feasibility of using long-lived nuclides to make alternative sources for calibration. Sets of 511 keV γ-window count data from γ and coincidence channels were recorded to investigate the effect of the γ coincidence additive effect and different β emission spectra on the activity response, and to determine which long-lived positron nuclides should be used in the field calibration. Calibration with a standard 18F surface source shows that the monitor can give stable and reliable readings of this type of standard source over different orders of magnitude within the range. The extended uncertainty (k=2) of the displayed activity response is 6.0%, which meets manufacturing requirements. And the calibration of the long-lived positron nuclide standard point sources shows that the positron spectrum has a great influence on the activity response of the monitor, and the γ coincidence additive effect is not very significant. Therefore, it is recommended to use 22Na with β energy close to 18F to make point sources or surface sources for field calibration.

The focal spot size of X-ray tube is one of the key factors affecting the imaging resolution, and it is also an important index to characterize the working performance of the X-ray source. Typically, after the X-ray tube is manufactured, the focal spot size must be measured and marked. Nowadays, the most widely used measurement standards for X-ray tube focal spot size in the world are IEC 60336, EN 12543, and ASTM E1165. Between various standards, there are differences in the focal spot size measurement methods, scope of application, loading factors and imaging equipment. Among these, the slit method of EN 12543 standard is mostly utilized in industrial X-ray tubes. However, due to its complex and precise design requirements, the test equipment, i.e., the slit camera, is difficult to produce and process, and the cost is high, which is not conducive to promotion and application. This paper studied the influence and variation of the measurement of the focal spot size at different slit sizes by using double tungsten alloy cylinders instead of slit camera. Based on the calibration of distance and position parameters in the cylinder method, the measurement equipment was conveniently positioned, and the distance between the focal spot and the translation mechanism was measured simultaneously, which improves the position accuracy of the focal spot size measurement. Additionally, the formula for determining the focal spot size in the slit method was improved, so that the slit method can be applied to a wide range of slit widths. In particular, the geometric magnification of the system can be measured directly when the slit width is large, skipping the step of measuring the distance between the focal spot and the translation mechanism. The experiment scheme has low requirements on the centering conditions of the equipment and the distribution of the X-ray source. In the data processing process, dark field correction and bright field correction were used to eliminate the problem of inconsistency in the response characteristics of each detector pixel, and the Wiener Filtering was performed on the signal output by the detector to deal with Gaussian noise and uniformly distributed noise. The experiment results show that relatively accurate measurement values can be obtained by measuring MXR-225HP/11 with double tungsten alloy cylinders with the same spacing as the focal spot size. The measurement relative error is less than 10%. According to international standards, when the measurement results are within plus or minus 10% of the theoretical value, it can be used as a method to determine the focal spot size. Therefore, the improved method in this paper is reliable and does not require high testing equipment, allowing it to be widely used. In addition, the geometric magnification of the focus measurement device can be directly calculated when the slit width is larger than the focal spot size, and the calculation results are consistent with the results measured in the cylinder method.

This paper aims to comprehensively analyze the Alfven instability observed in the high βN discharge of EAST, considering both experimental observations and simulation studies. There are two purposes of this paper, the first one is to observe the AE activity in EAST, the second one is to apply and validate the FAR3d code in EAST. In the experimental analysis, the presence of a strong instability is detected using magnetic probes and electron cyclotron emission diagnostics. Two kinds of modes are observed:one's frequency ranging from approximately 70 kHz to 100 kHz, exhibiting an upward frequency sweep, and another frequency at around 60 kHz. These instabilities are found to destabilized near ρ=0.46 with a toroidal mode number of 2 calculated by toroidal Mirnov probe array. Complementing the experimental findings, simulations were performed using the eigensolver method of the FAR3d code. This is the first time that FAR3d code was applied on EAST, so the model needed to be validated by reproducing the AE activity. The simulation results successfully identify the dominant mode, toroidal Alfven eigenmode (TAE), occurring at ρ=0.45 with a frequency of 87 kHz, toroidal and poloidal number n/m=2/3, 2/4. Additionally, a sub-dominant mode, energetic particle mode (EPM), is identified at ρ=0.55 with a frequency of 62 kHz, n/m=2/4. The agreement between the simulation and experimental results confirms the accuracy of the simulation model in characterizing the instability types, as well as providing consistent location and frequency information. Furthermore, the FAR3d code is employed to evaluate the influence of the finite Larmor radius (FLR) effect. The simulation results demonstrate that the FLR effect has minimal impact on identifying low toroidal mode number (n) instability modes in the high βN discharge of EAST. Consequently, the FLR effect can be neglected during the mode identification process, allowing for accelerated calculations. While, in the simulation of high toroidal mode number, the FLR effects should be considered. In conclusion, this study presents a comprehensive analysis of the Alfven instability in the high βN discharge of EAST. By combining experimental observations with simulation results, the types of instabilities, their locations, and frequencies are successfully identified and found to be in good agreement. Additionally, the study confirms that the FLR effect has little influence on identifying low n instability modes, enabling faster calculations by excluding FLR effects during mode identification. These findings contribute to a better understanding of the Alfven instability in high βN discharges and provide valuable insights for future research in fusion plasma physics.

The diffusion coefficient of oxygen in zirconium alloy oxide film and matrix is an important parameter in corrosion kinetics of zirconium alloys. At present, there are some reports on the diffusion coefficient of oxygen in zirconium and oxide film, but the numerical values in the literature are quite different. In this paper, zirconium alloy was corroded in pure water to obtain a sample with an oxide film thickness of 1 μm. Through vacuum annealing test, the oxygen concentration distribution of the oxide film at four different temperatures was obtained, the diffusion coefficient of oxygen in zirconium alloy matrix was calculated, and the Arrhenius relation of the diffusion coefficient in this temperature range was obtained by fitting. Compared with the literature data, the trend is similar, but the values are different. The main reasons may be the following two aspects:One is the measurement deviation of oxygen concentration distribution curve, including spatial position deviation and concentration error, and the other is the different grain size of samples. The diffusion coefficient of oxygen in zirconium metal without irradiation obtained in this work can be mutually verified with the literature data, which has high reliability. Based on the corrosion weight gain curves of two zirconium alloys at four different temperatures, the diffusion coefficient of oxygen in the zirconium alloy oxide film was estimated by the equivalent diffusion model of the oxide film and the corrosion weight gain curve before the corrosion transition. It is found that the diffusion coefficient of oxygen in the zirconium alloy oxide film increases obviously with the increase of temperature, and the diffusion speed of atoms is significantly affected by the microstructure of the diffusion medium except temperature. Based on the research of two parts, it is found that the diffusion coefficient of oxygen in the oxide film is greater than that in zirconium metal when the temperature is below 440℃, and the diffusion coefficient of oxygen in the oxide film is more than twice that in zirconium metal when the temperature is below 400℃, which shows that the diffusion of oxygen in the oxide film of zirconium alloy is main at the interface at lower temperature (440℃ in this paper), and further shows that grain boundaries, pores and microcracks in the oxide film play a very important role in the corrosion behavior of the alloy. The above research provides important calculation parameters and basic research theory for the corrosion kinetic model of zirconium alloy. The results show that the diffusion coefficient of oxygen in Zr-Sn-Nb matrix follows a relationship as DZr(cm2/s)=0.18·exp(-180 000/RT). The diffusion coefficient of oxygen in oxide film, which follows a relationship as Dox(cm2/s)=3×10-7exp(-101 550/RT), is also evaluated by the weight gain profile in pre-transition stage.

The monolithic scintillation crystal detector has the ability to position the three-dimension interaction location of the gamma ray in the detector depending on accurate acquisition of scintillation light distribution. Depth estimation of light field can be solved by a lens array in light field technology. A new detector model was designed in the paper which introduced a lens array for positioning the interaction in the monolithic detector. The scintillation light is focused by each lens unit in the lens array. Photons in some discrete pixels of the photon counter array where the scintillation light is focus on are detected to form an image of multiple focus points instead of the scintillation light distribution detection based on continuous crystal. The position of the scintillation light source in the crystal can be reconstructed from the focus image through optical path inversion. The detector model consisting of a monolithic crystal, a lens array and a multi-pixel photon counter was established in the paper. The monolithic scintillator was a lutetium-yttrium silicate (LYSO) scintillator with dimensions 48 mm×48 mm×45 mm. The lens was hemispherical with a radius of 15 mm supposed to be uniformly distributed in 3×3 array and the material was as same as the crystal. The crystal and the lens array were considered as a whole in the simulation. The multi-pixel photon counter was set at 44.4 mm from the crystal surface. The pixel size of the photon counter array was 3 mm and the quantum efficiency was 30%. The axis of symmetry of the three parts was same. The reconstruction algorithm was based on optical path inversion from multiple focus points for locating the 3D position of gamma-ray interaction. The intersection points of the inverted optical paths and the set z plane with different depths in the crystal was calculated in the reconstruction process. When the intersection points converged towards the same point in the z0 plane, the depth of interaction was z0 and the intersection point in the z0 plane was the interaction point. The reconstruction results were analyzed to evaluate the performance of the monolithic scintillation crystal detector based on the lens array. The x/y resolution is better than 1.54 mm and the z resolution is better than 3.13 mm. The two interaction points when coincidence occurs are reconstruced in the condition that the energy is the same. The detector is feasible for the positioning of two interaction points by the reconstruction results of three groups of selected positions. The monolithic scintillation detector based on a lens array has both good positioning ability and position resolution. The detector is suitable for the position reconstruction of multiple interaction points.

The probabilistic safety assessment of nuclear containment under internal pressure is the key evaluation content for guaranteeing nuclear power plant safety, and the fragility assessment of the containment under internal pressure is the most crucial step in the probabilistic safety assessment of nuclear containment. At present, the internal pressure fragility assessment of the containment mostly adopts the simplified lognormal distribution model, which lacks a rigorous theoretical analysis. There are multiple uncertainty issues affecting structural responses of nuclear containment in the fragility analysis and the results of the fragility analysis should be on the basis of a rigorous capacity-demand model. Meanwhile, load parameters (i.e. internal pressure) are usually used as key information to assess the internal pressure failure probability of containment. However, damage and failure of containment are more intensively correlated to structural response parameters (e.g. structural acceleration, structural displacement, and damage indices) compared with load parameters. For this reason, this paper proposed an internal pressure fragility assessment method based on Bayesian theory for nuclear containment and a form of damage ratio was used as a parameter for fragility assessment of containment. The damage ratio was realized by defining the ratio of the damage area to the full area of the containment concrete. Specifically, the fragility assessment method first established a deterministic demand model for the containment under internal pressure, and then corrected the demand model by adding correction terms. After that, Bayesian estimation was used to gradually filter the correction terms, resulting in an accurate probabilistic demand model. Finally, a fragility assessment result was obtained by establishing a limit state equation for the containment. On this basis, the probabilistic model under internal pressure was introduced to realize the probabilistic safety assessment of the containment, and the results of probabilistic safety assessment were compared with those of conventional fragility assessment. The results show that the cumulative containment failure probability (CCFP) of the containment decreases gradually with the increase of the damage ratio and the coefficient of variation of the CCFP is obviously smaller than that of the conventional fragility assessment method. In this paper, Bayesian estimation is introduced for probabilistic safety assessment of internal pressure in containment, and the damage ratio is used as a key parameter for fragility assessment of containment, which can provide more accurate and conservative results for probabilistic safety assessment of internal pressure in containment.

Helicobacter pylori is a type of bacteria that is commonly found in the stomach lining and is known to cause gastritis, gastric ulcers, and gastric cancer. Detection and diagnosis of Helicobacter pylori infection are crucial for effectively eradicating the pathogen. Currently, the 14C-urea breath test is widely considered the gold standard for detecting the infection. During the test, patients ingest capsules containing 14C-labeled urea. It is important to identify the active ingredient, 14C, in the 14C-urea capsules to ensure the quality and safety of the medication. Unfortunately, there is no clear method provided by the Chinese Pharmacopoeia or the United States Pharmacopoeia for identifying the 14C nuclide. To address this gap, a method for identifying the 14C nuclide in 14C-urea capsules using liquid scintillation counter measurement was developed in this paper. In the experiment, a solution of 14C-urea was mixed with ethanol and low-quenching scintillation cocktails, and then argon gas was used to remove oxygen and prepare a low-quenching sample. The liquid scintillation counter was used to measured the β spectrum, and five characteristic points were selected for comparison with the β spectrum of a standard non-quenched 14C source. The identified characteristic points included the maximum apparent energy value, the apparent energy value at the right end of the half-peak height, the apparent energy value at the right end of the 90% peak height, the 90% peak energy width, and half-height width. Successful identification of the 14C nuclides is achieved by comparing the β spectrum of the low-quenching sample with the standard non-quenched 14C source. Oxygen and ethanol are the main sources of quenching during the preparation of low-quenching samples. However, the experimental study shows that the dissolved oxygen in the scintillation cocktails could be effectively removed by flowing argon gas for 10 minutes, which significantly reduces the quenching degree and extends the validity period of the sample to 30 minutes. Furthermore, adding less than 1.67% ethanol to the sample results in no significant quenching. To avoid introducing new quenching sources, the distribution of 14C-urea in the capsule was analyzed, and it is found that more than 90% of the 14C-urea is located at the bottom of the capsule. The established nuclide identification method was used to identify 14C in three batches of 14C-urea capsules from a company in Shanghai, and the results are all qualified, validating the effectiveness and feasibility of the established method. In summary, the nuclide identification method developed in this paper can provide valuable experience for identifying nuclides in other radiopharmaceuticals, enhancing the safety and effectiveness of radiopharmaceuticals. Accurate identification of active ingredients in medications is essential for ensuring the safety of patients and improving the quality of healthcare.

The benchmarking critical experimental data can be used as a reference, providing basic data support for the macroscopic test of nuclear data, the design method verification of new nuclear energy system and the nuclear critical safety analysis, etc. As a crucial link in the process of benchmarking critical experimental data, the uncertainty analysis and evaluation of critical experimental data have been widely concerned and studied at home and abroad. Traditional GUM (Guide to the Uncertainty in Measurement) department based on the uncertainty analysis of the process “one change strategy at a time” uncertainty analysis method in calculation efficiency is low, cannot consider interval probability distribution characteristics of the input parameter error, defects of inaccurate results. In this paper, an uncertainty analysis method for critical experimental data was proposed based on perturbation theory and combination sampling method, and the probability distribution characteristics of input parameters in its error interval were restored by efficient Latin hypercube sampling method, so as to improve the accuracy and rationality of the uncertainty analysis of critical experimental data. By means of conjugate flux perturbation theory, the efficiency of critical experimental data uncertainty analysis was improved without significantly increasing the calculation time. Moreover, the efficient Latin hypercube sampling method and the conjugate flux perturbation method were verified and the feasibility of their application to the uncertainty analysis of critical experimental data was proved. Then, the Latin hypercube sampling program and complex modeling program were compiled, and the compiled program was verified and analyzed on the international critical safety benchmark question, and the calculation results obtained by the traditional critical experimental data uncertainty analysis method and the results given on the international critical safety benchmark question were compared. The conjugate flux perturbation uncertainty analysis method based on MCM (Monte Carlo method) analysis process improves the computational efficiency, makes the analysis results more reasonable and accurate, and proves that the uncertainty program is accurate and feasible for the uncertainty analysis of critical experimental data. Finally, the program was applied to the uncertainty analysis of critical experimental data of neutron toxicological effects in uranium solution obtained by China Institute of Atomic Energy, and a good result was obtained. Compared with the traditional critical experimental data uncertainty analysis method, the standard uncertainty of core characteristic value keff is improved, and the efficiency of uncertainty analysis is greatly improved. The uncertainty analysis method of critical experimental data studied in this paper and the program of uncertainty analysis can provide a more efficient and reasonable analysis method for benchmarking uncertainty analysis and multi-parameter sensitivity and uncertainty analysis of subsequent batches of critical experimental data.

Taking the release of radioactive aerosols in containment to the environment through cracks during nuclear power plant accidents as the research background, an experimental facility was established to study the deposition characteristics of aerosols in capillary tubes and the experiments were then conducted using the facility. The capillary tubes used in the experiments are PEEKsil tube. The gas flow inside the capillary was driven by pressure, with a maximum pressure difference of up to 500 kPa on both sides. The aerosol particles used in the experiments were titanium dioxide, with a mass median diameter of 1 mm. Organic filter membrane was used to collect the leaked aerosol, and then inductively coupled plasma-atomic emission spectrometry (ICP-AES) was adopted to obtain the titanium mass, which will be used to obtain the aerosol mass concentration at the capillary inlet and outlet. A methodology to calculate the aerosol deposition velocity based on the experimental data was developed in this paper, and then the relationship between dimensionless deposition velocity and relaxation time was built. It shows that the variation of dimensionless deposition velocity and relaxation time of aerosols in capillaries can be divided into rising zone and plateau zone, and the dimensionless relaxation time boundary between two regions is approximately between 70 and 100. In the rising zone, the dimensionless deposition velocity increases with the dimensionless relaxation time. In the platform zone, the dimensionless deposition velocity remains basically unchanged and is less than 0.03. Compared with the experimental results of aerosol deposition characteristics in conventional pipelines, the deposition velocity in this paper is relatively small. To analyze the deposition characteristics of aerosols in pipelines, numerical analysis has been conducted using the Lagrange method and the Eulerian method, among which the widely used Eulerian method is the three-layer deposition model. The model assumes that: 1) The particle flux across the boundary layer is steady and unidimensional, perpendicular to the surface; 2) The fluid is well mixed so that the particle concentration gradient exists only very close to the deposition surface; 3) There are no sources or sinks of particles within the boundary layer; and 4) The surface is a perfect sink for particles. To investigate the reasons for the smaller particle deposition velocity in the capillary tubes, this paper first restores the three-layer deposition model, which agrees well with the previous experimental results, and then adds Saffman force to the three-layer deposition model. In the modified three-layer deposition model, this paper introduces an average lag coefficient that reflects the difference in axial velocity between particles and fluid, and then evaluates the impact of lag coefficient on particle deposition velocity. The results show that when the lag coefficient increases to a certain value, the effect of Saffman force is significantly enhanced, and the dimensionless deposition velocity sharply decreases. In the experimental study on the deposition characteristics of aerosols in capillary tubes conducted in this paper, a compressible flow driven by high pressure difference is present in the capillaries. The airflow continuously accelerates in the flow direction, resulting in the acceleration of aerosol particles. The Saffman force always exists during the aerosol transportation in the capillary tubes. When the difference between particle velocity and gas flow rate reaches the certain value, the Saffman force greatly weakens the wall deposition of particles.

Nitrogen pressurization is widely used in nuclear reactor (such as passive nitrogen-type accumulators and nitrogen regulator) because of passive mechanism. However, nitrogen pressurization brings new challenges. Loss of coolant accident (LOCA) can lead to rapid reduction of pressure in primary circuit of pressurized water reactor (PWR), putting dissolved nitrogen in the water into a supersaturated state, which contributes to nitrogen release. In order to prevent the accumulation of released nitrogen, which may act as a threat to reactor safety, the migration and release characteristics of nitrogen were studied. In this paper, a nitrogen wall release model was established, the physical significance of the release source term was explained and the specific process of nitrogen release was revealed. In the case of LOCA in primary circuit of a PWR, the pressure in the primary circuit decreases and the susaturated dissolved nitrogen degas from the liquid and form nitrogen bubbles on nucleate sites. The susaturated dissolved nitrogen continuously degas at the gas-liquid interface of the bubble, leading to the sustained growth of nitrogen bubble. After reaching departure diameter, the buoyancy force on the nitrogen bubble is greater than surface tension and it escapes from the wall finally. In this paper, the mathematical modeling of nitrogen dissolution equilibrium, dissolved nitrogen convection and diffusion, the susaturated dissolved nitrogen wall release and nitrogen bubbles upward floating was carried out. Combined with the one-dimensional lumped parameter method, the mass conservation of the control volume was analyzed and the one-dimensional nitrogen mass conservation equation was deduced. To solve the one-dimensional nitrogen mass conservation equation, a one-dimensional lumped parameter nitrogen migration solution program was written based on MATLAB. The program was verified by the experimental data of small break loss of coolant accident (SBLOCA). The trends of dissolved nitrogen concentration and void fraction under pressure relief condition during SBLOCA were simulated. The results suggest that susaturated nitrogen degas on the wall in the form of bubbles. The bubble grows at a rate proportional to t1/2 and also proportional to (C-Ci)12. The influence of the initial nucleation radius on the growth period of the bubble is negligible and the bubble rises uniformly in the main stream after departure. Under pressure relief condition during LOCA, the nitrogen wall release effect is significant and the diffusion process of dissolved nitrogen is weak. Controlled by nitrogen wall release, the average concentration of dissolved nitrogen decreased by 19.5% and the void fraction became 4.33 times of the original, which indicates that wall release has significant effect on circuit, potentially affecting reactor safety. The simulation results are in good agreement with experimental data of SBLOCA, proving the reliability of wall release model and program.

Gamma-ray spectrometry uses the X-ray and characteristic gamma-rays associated with 235U, 238U and their decay daughters in uranium samples for analytical calculations to achieve rapid and timely nondestructive detection of sample enrichment. As one of the methods for determining uranium enrichment by gamma-ray spectrometry, the relative detection efficiency calibration method has been commonly used in nuclear safeguards on-site inspections to measure the enrichment of samples. Its core is the relative efficiency self-calibration using multiple characteristic gamma-rays of 235U in the low energy region of 120-186 keV and the characteristic gamma-rays of 238U decay daughter 234Pam in the high energy region of 258-1 001 keV. In this paper, the relative detection efficiency of the detector was obtained using the Monte Carlo method based on the actual characterization results of a coaxial-type high-purity germanium detector, the fitting results and errors of three different types of fitting formulas in different energy region ranges were investigated. A comprehensive analysis of the effect of the fitting formulas from the low-energy region to the medium-high energy region on the relative efficiency of the characteristic peaks shows the logarithmic transfer series function performs better results among the three fitting formulas. During the on-site inspection, the aluminum and stainless steel shieldings of the sample and the self-absorption of the uranium dioxide and uranium hexafluoride body source have a certain impact on the relative detection efficiency curve. The specific influence and changing trend of these factors on the curve were studied, and the fitted curves for thicker shields and strong self-absorption were then corrected by adding correction factors. The results show that after adding the correction factor, the fitting error of the relative detection efficiency curve for different shielding and self-absorption cases is reduced from 2%-4% to 1%-2% at the characteristic peaks used for enrichment calculation. The results of this paper have reference value for the nondestructive assay of uranium enrichment in the on-site inspection of nuclear safeguards.

This paper presents a cold-end optimization design study of a nuclear energy power conversion system in the context of low-temperature heat sink applications. Based on the principles of simple recuperation cycles and recompression cycles, the study focuses on the supercritical carbon dioxide power cycle and its cold-end optimization for application in low-temperature heat sink environments. These strategies are designed to enhance the efficiency and reduce the operating pressure of the supercritical carbon dioxide power cycle under low-temperature heat sink conditions. Three different cold-end optimization approaches were proposed, which include: a non-condensing liquid-phase compressed supercritical carbon dioxide power cycle, a condensing liquid-phase compressed supercritical carbon dioxide power cycle, and a gas-phase compressed supercritical carbon dioxide power cycle. This research established a thermodynamic analysis program, which modeled turbine machinery such as compressors and expanders using isentropic models. Heat exchangers such as recuperations and coolers were simplified and modeled, with pressure drops and end temperature differences assumed. The study’s thermodynamic analysis program was compared with computational results from a Sandia National Laboratories research report, confirming the accuracy of the thermodynamic analysis methodology presented in this paper. For a 100 MW scale power generation unit, the thermodynamic analysis program was used to conduct computational analyses on both simple recuperation cycles and recompression cycles. These cycles were driven by a medium-to-high-temperature reactor heat source with a turbine inlet temperature of 550 ℃, and a medium-to-low-temperature reactor heat source with a turbine inlet temperature of 315 ℃. The study reveals that the optimal cold-end optimization approach varies based on the heat source temperature and cycle configuration. For the medium-to-high-temperature reactor heat source, the gasphase compressed supercritical carbon dioxide power cycle in the simple recuperation cycle configuration achieves the highest efficiency at 39%. In the recompression cycle configuration, the condensing supercritical carbon dioxide power cycle achieves the highest efficiency at 48%. For the medium-to-low-temperature reactor heat source, regardless of the cycle type, the condensing supercritical carbon dioxide power cycle yields the highest efficiency, the highest efficiency is 28% for the simple recuperation cycle and 33% for the recompression cycle. When comparing the cold-end optimized supercritical carbon dioxide power cycle to conventional supercritical carbon dioxide power cycles with the same maximum temperature and pressure, the optimized system exhibits a reduction in operating pressure drop and an efficiency improvement of 2% to 4%. The cold-end optimized configurations of the supercritical carbon dioxide power cycle hold promising applications for nuclear energy power conversion systems with low-temperature heat sink environments.

In order to realize the power supply of small nuclear reactors in remote areas, the sub-channel analysis program BRESA of supercritical carbon dioxide (S-CO2) fuel assembly based on Modelica language was developed in this paper. According to the fuel assembly model of 37 rods, the axial and radial physical and thermal parameters of S-CO2 fuel assembly were calculated, and the flow mixing effect between channels was introduced to obtain the distribution law of thermal hydraulics parameters in the axial and radial directions, The enthalpy rise and temperature rise changes of various typical channels were studied. The results show that the enthalpy rise and temperature rise changes of the channel in the middle of the fuel assembly are high, while the enthalpy rise and temperature rise changes of the channel at the edge of the fuel assembly are low. Through comparison with relevant literature, it is proved that the thermal hydraulic parameters of this study meet the design requirements and the correctness of the program is verified. On the basis of steady-state calculations, the transient conditions of flow rate changes were analyzed, and the response characteristics of various parameters of fuel assemblies under transient conditions of linear increase, linear decrease, and sudden change in flow rate were analyzed. The response characteristics of reactor power response and control system were investigated and analyzed, and the spatial variation trends of enthalpy and pressure with time were obtained. The results indicate that under transient conditions, the reactor power will vary with flow fluctuations, which is related to the negative feedback effect of the control system and temperature. The pressure and enthalpy values also fluctuate and change when the flow rate changes, which is influenced by the dual effects of flow rate and power changes. Under transient conditions of sudden flow changes, enthalpy and pressure undergo sudden and significant fluctuations, with the maximum fluctuation of outlet enthalpy reaching 12%, posing a challenge to the safe operation of the reactor. The research results of this paper provide theoretical support and technical guidance for the design and research of supercritical carbon dioxide reactor systems.

Under high-temperature conditions, supercritical carbon dioxide (S-CO2) power conversion systems have significant advantages, which show great potential in improving energy utilization efficiency, reducing environmental pollution, and realizing sustainable energy development. S-CO2, as a high-efficiency working fluid, can achieve high efficiency in higher temperature thermodynamic cycles, which helps to reduce fuel consumption and greenhouse gas emissions. Since the S-CO2 power cycle is significant to energy conversion and utilization, many researchers have conducted experimental studies on the convective heat transfer characteristics of S-CO2. However, existing studies mainly focus on low-temperature condition (10-200 ℃), and experimental data under hightemperature condition (200-500 ℃) are scarce. Most of the research conditions on the heat transfer characteristics of S-CO2 have not reached the optimal temperature range of the S-CO2 power conversion system, in order to master the critical data of the relevant design, it is necessary to carry out experimental research on the heat transfer characteristics of S-CO2 under high heat flux conditions, which holds a crucial significance in the development of associated power conversion systems. This paper first introduced the construction process of the S-CO2 forced circulation test bench, and evaluated the uncertainty of the relevant experimental data. The main temperature range of the experiment is 450-800 K, the pressure range is 7.40-10.22 MPa, and the heat flux range is 20-1 000 kW/m2. The heat transfer characteristics of S-CO2 in a vertical tube under high-temperature conditions were studied experimentally, and the effects of thermal acceleration and buoyancy on heat transfer were analyzed. The results show that the heat transfer trend of S-CO2 is similar to that of gaseous CO2 under high-temperature conditions, but the values are quite different, and the influence of thermal acceleration and buoyancy effect on the heat transfer of high-temperature S-CO2 is negligible. Finally, in order to develop a reliable prediction method of S-CO2 convection heat transfer under high-temperature conditions, this paper collected the empirical correlations of S-CO2 heat transfer in vertical heating circular tubes commonly found in the literature, and analyzed them under high heat flux, evaluated and analyzed their experimental prediction high-temperature conditions. It is found that there are some differences with the existing experimental data. Therefore, this paper proposes an empirical correlation suitable for high-temperature S-CO2 heat transfer. The newly proposed heat transfer correlation has the highest calculation accuracy, which calculates 96.65% of experimental data within ±10% error. It can achieve higher accuracy, which meets the actual engineering needs.

Boron neutron capture therapy (BNCT) is an advanced type of radiotherapy at cellular level, and it has attracted great attention as the next-generation cancer therapy and is booming worldwide in recent years. BNCT uses the short-range (4-10 μm) and high linear energy transfer particles produced from the 10B(n,α)7Li thermal neutron capture reaction on 10B to selectively kill tumor cells while suppressing exposure dose to normal tissues. Since the physical property and biological effect of neutron significantly change with neutron energy, the energy spectrum determination of BNCT neutron beam is of vital importance, and it is directly related to the safety and effectiveness of BNCT. In this paper, the multi-foil activation method was used to determine the energy spectrum of BNCT neutron beam. This method has the advantages of insensitive to γ-rays and little disturbance to the measurement field owing to the small size of activation foils. Considering the activation reaction cross section, resonance absorption energy, half-life, irradiation and radioactivity measurement time, disturbance to the measurement field, cost and other factors, seven materials, that is, InAl, AuAl, LaAl, MnNi, CuAl, W and Sc, were selected as the activation foils in this paper. The 640-group activation reaction cross section of each activation foil was prepared. By comparing the ratios of the simulated activation rate per atom, Rorder, to the ideal activation rate per atom, Rideal, for different arrangements of each activation foil, the optimal arrangement of the activation foils was determined, that is, W, InAl, AuAl, LaAl, MnNi, CuAl and Sc along the incident direction of the BNCT neutron beam. Finally, based on the simulation study of the neutron energy spectrum determination with multi-foil activation method, an effective unfolding method was established in this paper. The energy spectra of four BNCT neutron beams based on accelerators and reactors were determined by using the proposed unfolding method. The selected neutron beams of two accelerator-based BNCT (AB-BNCT) facilities are using p-7Li and D-T reactions, respectively, and the selected reactor-based BNCT neutron beams are produced from two still operating reactors, that is, Tsing Hua Open-pool Reactor (THOR) in Taiwan, China and Kyoto University Reactor (KUR) in Japan, which have been used clinically so far. The results show that the unfolded spectra are in good agreement with the initial spectra of the BNCT neutron beams, which causes the average relative deviations are not larger than 7%, and it meets the requirements of practical applications. The proposed neutron energy spectrum determination method is effective, practical and universal, and it provides a feasible method for determining the energy spectrum of BNCT neutron beam.

Metallic magnetic calorimetry holds significant promise in critical defense applications such as nuclear proliferation prevention, radionuclide classification, nuclear safety and anti-terrorism. Metallic magnetic calorimetry realizes energy resolution in the order of a few electron Volts by utilizing the significant contribution of heat capacity of a magnetic origin at milli-Kelvin temperatures in rare-earth doped Au films. For detecting gamma-ray photons with an energy of around 100 keV, the energy resolution is better than 0.1%, 10 times better than the high-purity germanium (HPGe) gamma spectrometer with the highest energy resolution. The metallic magnetic calorimeter can detect other radiation types than gamma rays such as alpha and beta emissions. Metallic magnetic calorimetry can also be used for precise measurement of astrophysical phenomena, cosmic rays and other high-energy physics phenomena. As an critical part of the metallic magnetic calorimeter, the pick-up coil plays a dual role in the sensor unit. It is not only responsible for converting magnetic flux changes into electrical signals, which is subsequently picked up by the SQUID chip, but also responsible for generating persistent excitation magnetic field and magnetizing the rare-earth doped sensor thin-film. The design configuration and electromagnetic parameters of pick-up coil can directly influence the energy resolution of the sensor. In the paper the influences of various parameters (e.g., number of turns of soil, duty ratio) on the electromagnetic characteristics using the finite element method was calculated. The actual pick-up coil size is in the order of micrometers. As direct measurement of such tiny devices is extremely difficult with conventional means and such measurements suffer high errors, in this paper scaled-up samples were measured, and the validity of the simulation model and parameters through comparison with measurement results was verified to provide a reliable foundation for design work of pick-up coils. The calculation results show that the influence of turn number of pick-up meander coil on inductance is more significant than other parameters such as duty ratio and loop pitch. With the increase of the duty ratio, the inductance shows a monotonic decrease trend. When the duty ratio increases from 30% to 70%, the inductance decreases by about 76.4%. On the other hand, with the increase of the turn number, the inductance shows an obvious increasing trend. When the turn number increases from 15 to 75, the inductance increases by about 253%. The simulation result also reveals the magnetic field distribution inside the metal magnetic calorimeter sensor pick-up coil, which provides critical information for the design of the sensor thin film.

In the 250 Wt RTPV (radioisotope thermophoto-voltaic) system, the heat source operates at a temperature above 1 000 ℃, which requires a long term stable thermal insulation devices. In this paper, three types of non-sealed high temperature MLI (multilayer insulator) were prepared by using spike, stainless steel wire mesh and aerogel as separator respectively. Each type was installed and tested in the RTPV prototype separately. The three MLIs are in the same size of 187 mm long, 183 mm wide and 50 mm thick. In MLI with spike separator, 50 metal foil reflectors were pierced several small holes, the spikes from the holes were used as the separator to make the adjacent reflectors in point contact. In MLI with stainless steel wire mesh separator, 50 stainless steel wire meshes were spot weld on each of the metal foil reflectors as the separator to make the adjacent reflectors in line contact. In MLI with aerogel separator, 18 polished metal plates were used as the reflectors, and slender strip aerogels in size of 50 mm×10 mm×2 mm were installed on four corners between adjacent reflectors as the separator. The main configuration of RTPV prototype tested in this paper was as follow. 1) An electric heating source with the size of 99 mm×95 mm×55 mm was used to simulate the radioisotope heat source. The heat source was supported in RTPV center by seven ceramic nails and coated with tungsten alloy radiation emitter. 2) Four optical filters were placed 34 mm away from the heat source and supported by four inner brackets. 3) Photovoltaic cells were installed on the side walls inside the housing with a distance of 14 mm from the optical filter. 4) Two high temperature MLIs were installed on the upper and lower end covers inside the housing. 5) Cold plates were used on the side walls outside the housing. Vacuum thermal performances of RTPV with different MLIs were carried out on a test platform. Based on the test results, some conclusions come out as follows. 1) The aerogel separator has the best performance in controllability of the assembly separating effect, which leads to the best thermal insulation performance consistency between different batches. 2) When the assembly separating effect is well controlled, the high temperature MLI with spike separator has a better insulation performance than stainless steel wire mesh separator. 3) The optical cavity inside the RTPV prototype is consisted of high-temperature MLI and the optical filters. A better closure of the optical cavity significantly contributes to the improvement of the heat utilization, and accesses to a higher heat source temperature. The optical cavity has the best closure performance when using the high temperature MLI with aerogel separator. While the RTPV prototype with aerogel separator input power is 250 Wt, the heat source temperature reaches 1 038 ℃.

Heat pipe reactor has the advantages of miniaturization, high inherent safety, and simple structure, which has extensive application prospects and research significance in the fields of universe exploration and distributed power generation. A megawatt heat pipe reactor is typically required to arrange thousands of heat pipes in a limited space to export the heat from the core, which increases system complexity and impedes engineering technology realization. To solve this problem, a novel design was proposed in this paper in which the flat heat pipe arrays and plate type fuel elements were tightly fitted to form an integrated core, which has the characteristics of compact structure, small and invariable thermal contact resistance, and stable operation under various operating conditions. Non-uniform heating is common in the actual heat transfer process of the core. The flat heat pipe should be able to achieve stable operation under the non-uniform heating condition to cool the core. Thus relevant research is necessary to clarify the influence of non-uniform heating on the flow and heat transfer characteristics inside the flat heat pipe and verify the feasibility of the flat heat pipe for core cooling. Based on this background, the three-dimensional thermal-flow-solid coupling analysis model of the minimum feature unit of the flat heat pipe array was established using COMSOL multiphysics software, and the research on the thermal-hydraulic and mechanical characteristics of the flat heat pipe under non-uniform heating conditions was conducted in this paper. The reliability of the numerical model was validated by comparing it with the literature. The relative deviation between the numerical simulation result and the experimental reference value is less than 5%. The rationality of a composite wick structure with special type arteries was analyzed using the above numerical models. The results show that the wick structure effectively reduces the flow resistance by changing the liquid return path, making the flat heat pipe work stably. The heat transfer and flow characteristics inside the flat heat pipe under various degrees of heating unevenness were studied, where the symbol Uf was used to define the degree of unevenness in this work. It was found through qualitative research and quantitative analysis that the temperature distribution on the wall surface of the evaporator of the flat heat pipe is affected by the non-uniform heating conditions, and the heat transfer and flow characteristics of the flat heat pipe are not, which suggests that the flat heat pipe could maintain good heat transfer performance and temperature uniformity. The thermal stress distribution of the flat heat pipe under different degrees of heating unevenness was studied furtherly. The results show that non-uniform heating increases the thermal stress of the shell of the flat heat pipe, the peak stress is on the vertex of the front end of the flat heat pipe; this does not cause material failure, and the working stress of the flat heat pipe is within the creep limit of the structural material at high operating temperature. Results of this study show the ability to realize the performance assessment of the flat heat pipe under non-uniform heating conditions, verify that the flat heat pipe designed in this paper could meet the demand of core cooling, and provide a reference method for the performance analysis of related untraditional heat pipes under special conditions.

After neutron irradiation, RAFM steel will produce two types of interstitial dislocation loops (1/2〈111〉 loop and 〈100〉 loop), resulting in radiation hardening and embrittlement after long service. And different types of dislocation loops may lead to different hardening phenomena. However, the formation and transformation processes of different types of dislocation loops are still obscure. Therefore, it is of great significance to clarify the formation and transformation processes of different types of dislocation loops for the evolution of materials’ microstructures and prediction of their properties. In this paper, we presented an improved parallel replica method (PRD), which improved the application of the original PRD method in dislocation loop evolution simulation. Through analyzing the mechanism characteristics of dislocation loop transition and extracting Burgers vector, the improved PRD method accurately identified the state of dislocation loops, and then calculated the simulation of accelerated transition process from 〈100〉 loop to 1/2〈111〉 loop. By combing the dislocation extraction algorithm in OVITO, the synchronous analysis of Burgers vector in the simulation process was realized. During the dislocation loop transforms, its habit plane will change and eventually affect the change of Burgers vector. So Burgers vector was used to redefine the adjudication for transition events in the original PRD method. We adjudicated whether the system undergoes a transition event, based on the existence of 1/2 〈111〉 Burgers vector in the simulated system and whether the proportion of 1/2〈111〉 Burgers vector increases. With such improvements, it would help us avoid lengthy cyclic iterations at some metastable state in our simulation and realize the simulation of accelerated transition phenomenon of 〈100〉 loop to 1/2〈111〉 loop. Finally, numerical experiments were performed to evaluate the effect of our work. The results show that the improved PRD algorithm successfully simulate the changing process of small size 〈100〉 loop to 1/2〈111〉 loop at 1 000 K, and the transition time is faster than the traditional molecular dynamics method. We also observe the coexistence phenomenon of 〈100〉 dislocation segments and 1/2〈111〉 dislocation segments in large size 〈100〉 loop simulation. So far as to the case of 〈100〉 loop with Ni solute elements, it is found that the complete transition to 1/2〈111〉 dislocation loop is happened. In addition, the accuracy of program simulation was verified, and the parallel performance were also analyzed and evaluated in this paper. We achieve about 90% parallel efficiency in the parallel calculating test, which is found much higher than the original PRD method. Moreover, our program has well-fined scalability, so it can be well expanded for the large-scale parallel calculation.

The LBE-cooled reactor is one of the important choices for sustainable and safe development of nuclear energy systems in the future. The safety issues involved with the steam generator tube rupture (SGTR) accident must be considered in the design and operation of such reactors. The essence of the evolutions during this SGTR accident for LBE-cooled reactors is focused on the interaction behaviors between high temperature liquid metal LBE and pressurized water. For such considerations, relevant researches on the mechanistic experiments and numerical simulations have been carried out at home and abroad. However, with respect to the numerical simulations of SGTR accident in LBE-cooled reactor, the traditional system analysis codes based on the typical two-fluid six-equations model, such as RELAP5, ATHLET, etc., are not appropriate here. And the targeted simulation codes used for SGTR in LBE-cooled reactors are still rare, researches in foreign countries are focused on the multi-phase, multi-components, multi-velocities based computational fluid dynamics (CFD) and neutronics coupled code SIMMER-Ⅲ which is developed to analyze the serious accidents of liquid metal cooled fast reactors, such as LBE-cooled reactors or sodium cooled reactors. While domestically, researches are focused on the multi-phase flow codes including NTC-2D and MC3D. It is absolutely of great significance to conduct further and in-depth studies on the fundamental models and numerical algorithms for thus SGTR accident. So, aiming at the interaction behaviors between LBE and water, which is one of the significant scientific issues involved with the accident evolution characteristics and behaviors of SGTR in LBE-cooled reactors, investigations on the basic theories, constitutive models and numerical algorithm were performed in this paper. A new three-fluid code including LBE, liquid water and vapor mixtures was developed, and its verification work was performed by using the published experimental data abroad including the water injecting into LBE liquid metal pool experiment in a small experimental apparatus by Japan JAEA and water injecting into LBE pool experiment in large experimental LIFUS5 series facilities by Italy ENEA. The results show that the new code developed in this paper agrees reasonably on the whole with the experimental results, but certain deviations still exist at some local details compared with the experimental results. On the whole, the relevant research achievements, including the critical related models, algorithms and code modules are able to simulate the interaction behaviors between LBE and water reasonably. The code developed in this paper is expected to provide certain theoretical and technical support for the accident analyses and safety evaluations of SGTR accident in LBE-cooled reactors. Meanwhile, more in-depth researches on the optimizations of the critical models and algorithms should be carried out in the future work.

Monte Carlo simulation has become an effective method for solving different physical problems, which is especially advantageous in the particle transport problems with complex geometric models. Charged particle transport simulation programs based on Monte Carlo method are widely used in medical therapy, radiation protection, accelerator design and other fields. Although the domestic neutron and photon transport simulation programs are well developed, there are few programs being able to do charged particle transport simulation. In addition, the energy range for programs based on the intranuclear cascade model and evaporative fission model is usually high, above several tens of MeV. For medium and low energy proton transport, the accuracy of nuclear reaction simulations based on nuclear databases is better than that of models. The biggest problem with Monte Carlo methods for simulating charged particle transport calculations is that the number of collisions of charged particles is too frequent. Therefore, this paper was based on the condensed history method to simulate the proton transport. In this study, with the application of key physical model of charged particle transport, such as the continuous slowing down approximation theory, multiple scattering theory and energy straggling model, a framework of heavy charged particles transport simulation program was established. Based on nuclear database of ENDF/B-Ⅶ.0 proton-nucleus interactions, the proton transport simulation function was implemented on the NPTS (neutron-photon transport simulation program). The extended NPTS program supported the simulation of proton transport problems in the energy range from 1 keV to 150 MeV and multi-coupling transport calculations of protons, neutrons, photons and electrons. A comparison of the stopping power of protons, deuterons and alpha particles in common materials with SRIM was carried out by the NPTS program, which proved the correctness of NPTS in calculating the stopping power of charged particles. The numerical results of neutron yield, average energy and angle for the thick target model agree well with GEANT4, MCNP6 and experimental results. In order to fully verify the program, a proton treatment chamber model with larger spatial dimensions and more complex geometry was used for the study. The neutron flux distribution inside the treatment chamber obtained by NPTS was basically consistent with GEANT4. The accuracy of the NPTS simulations of proton transport and multiparticle coupled transport was verified by several test models. Moreover, the proton energy and continuous slowing down approximation range curves obtained from the NPTS calculations can be used to quickly estimate the target thickness at maximum neutron yield, providing a favourable reference for BNCT (boron neutron capture therapy) target design. This paper effectively extends the application of the NPTS program in accelerator target design, shielding analysis, and radiation protection. The research effectively improves the application value of NPTS program, as well as enriching the research method of medium and low energy proton transport programs in domestic.

Recently, cement matrixes become a mainstream solution to address the stabilization of radioactive spent resins. Generally, the cumulative dose of spent resin cemented waste form will reach approximately 106 Gy over a disposal lifetime of 300 years. Due to the organic properties of spent resin, namely producing gas if being irradiated, it inevitably affects the mineral phase of cement waste forms. Therefore, the content and composition of the mineral phase could be key factors determining the performance of cement waste forms. With that in mind, in this paper, the influence of the above factors on ordinary Portland cement (OPC) was explored especially. Specifically, based on the Gibbs free energy minimization principle, the influence of three gases generated by irradiation of waste resin, namely H2, CH4 and CO2, on the mineral phase composition and total volume of ordinary Portland cement was discussed. Thereafter, the evolution of the mineral phase composition of ordinary Portland cement over time was investigated based on the rate of irradiated gas production. In this work, all the experiments were produced by GEMS software based on the thermodynamic Gibbs free energy minimization method. Meanwhile, the formulation for cement solidification of spent resin meets the standard of GB 14569.1-2011. From the experimental results, several important findings were obtained. Firstly, only a small amount of H2 and CH4 reacts with the cement hydration products, which causes the dissolution of the mineral phase with high Ca/Si, while the total volume of the mineral phase is almost invariant. Secondly, in the prior phase, CO2 first reacts with the portlandite in the cement hydration products. After the portlandite dissolute completely, the silicate minerals start to dissolve, forming a system composed of abundant carbonate and a few clay minerals, which decreases the total volume of the mineral phase. Thirdly, during the predicted 160 years, the effects of spent resin irradiation gases on the cemented waste form are consistently dominated by CH4 and H2, while the performance of the cemented waste form is stable. However, the risk of rupture and explosion caused by the accumulation of H2 and CH4 in the waste packaging containers and the risk of radioactive release due to carbonation of the cemented waste form by CO2 should not be overlooked. In summary, the above results reveal the evolution law of degradation of cement solidified body caused by waste resin irradiation gas. The results of this paper provide effective data support for the safety evaluation of waste resin treatment by cement solidified form.

The yields of 85Krm, 87Kr and 88Kr of neutron-induced fission of 239Pu are important nuclear parameters. They have a moderate half-life, are β- decay and emission γ-rays with a certain probability, so it is appropriate to use γ-ray spectrum method to determine their yield. At present, there are few experimental data with large deviation abroad and no experimental results have been reported in China. The yield measurements of 85Krm, 87Kr and 88Kr of thermal-neutron-induced fission of 239Pu were carried out based on the irradiation of plutonium target by Xi’an Pulsed Reactor rabbit system. The purified plutonium solution was then titrated and dried to prepare the silica-encapsulated plutonium target. Three kinds of measurement experiments were designed: γ-ray nondestructive method for the relative yield of 88Kr and 99Mo, gas-solid separation method for 85Krm, 87Kr and 88Kr and γ-ray nondestructive method for the relative yield of 85Krm, 87Kr and 88Kr. After the target irradiation and cooling, the fission products were measured by the above-mentioned methods. The efficiency of the three kinds of experimental samples was calibrated by the equivalent efficiency curve method. The plan source of organic glass package was used to equate the efficiency calibration of quartz tube source, and the plan source of stainless steel package with the same thickness and material as the measuring source was used to equate the efficiency calibration of 50 mL gas source. By Monte Carlo simulation, the suitable equivalent position was selected, and the efficiency value was fine corrected (less than 0.8%). The uranium target was prepared in the same way of plutonium target. The relative yields of 85Krm, 87Kr and 88Kr of 235U fission induced by thermal neutron measured by gas-solid separation after irradiation are consistent with the recommended values of the evaluation database, which indicates that the release rates of 85Krm, 87Kr and 88Kr in the plutonium target are the same. After modified by the influence of non-primary neutron field and non-primary isotope, the fission yields of 85Krm, 87Kr and 88Kr were obtained based on the measured relative yield with reference product 99Mo and the yield of 99Mo given by ENDF/BⅧ.0 evaluation database. The measured yields are (0.559±0.020)%, (0.992±0.034)% and (1.264±0.037)%, respectively, which are consistent with the evaluation value of the database and the uncertainty is significantly reduced compared with other literature value. In this paper, the retrieval and deduction methods of γ-ray energy peak interference ratio in measurement are universal, but the uncertainty of theoretical yield may be added. The experimental method in this paper can also provide reference for the measurement of fission product yield at different neutron energy points.

To meet the requirements of mitigating the climate change, a large scale deployment of nuclear energy is expected, which impulses higher efficiency in the utilization of uranium resources. As one of the promising solutions, the integral fast reactor featuring in a high internal breeding can reach a high burn-up, low reprocessing mass, and a long refueling period. In this paper, the relationship between the breeding characteristics and two main design parameters of the core, the rod diameter and the pitch-to-diameter ratio (P/D), was studied. The negative impact on internal breeding caused by the measures for reduction of sodium void reactivity was also analyzed. All calculations were carried out with the Monte Carlo code RMC and ENDF/B-Ⅶ.1 continous-energy ACE data libraries. In each case the breeding characteristics were shown by both breeding ratio of the fissile nuclides and the regeneration ratio of the TRU nuclides. The variation of keff, fissile nuclides, TRU nuclides along with burnup were also illustrated. The calculation results indicate that the internal breeding can be significantly enhanced by increasing the rod diameter and by decreasing the P/D. With the current U-Pu-10Zr fuel parameters, a satisfactory burnup reactivity curve can be realized with the rod diameter at 7.0 mm and the P/D at 1.14, taking into account a balance between neutronics and thermal-hydraulics. The positive sodium void reactivity effect makes one of the largest risks of sodium cooled fast reactor (SFR) under unprotected severe accidents. The sodium void reactivity effect can be reduced by setting an upper sodium plenum and reducing the core height-to-diameter ratio (H/D). However, both methods bring significant cost to the internal breeding, due to a higher leakage of neutrons under normal operation. In this paper it is found that to bring the whole-core sodium void reactivity down to zero, the H/D must be as low as 0.15 approximately under the existence of an upper sodium plenum at the height of 40 cm. Such a design results in a fast drop of the burnup reactivity, as well as a low breeding ratio, a low regeneration ratio, and the decreasing of the fissile mass. It can be concluded that in the integral fast reactor design a dilemma exists between the high internal breeding and the low sodium void reactivity effect. This paper suggests that the integral fast reactor should focus on the breeding characteristics and thus a high efficiency of the utilization of uranium resources, leaving the sodium void reactivity to be incorporated into the overall design of the reactor safety features.

Physical calculations of reactor cores are fundamental for reactor core design and nuclear safety analysis. The second generation of core calculation theory and methods based on advanced component homogenization theory and modern coarse grid block methods are commonly used in reactor core physics calculations. However, due to the fact that the two-step method of component homogenization does not accurately provide the grid rod power, the modified two-step method of cell homogenization based on the pin-by-pin homogenization theory of core calculations has gradually become a research hot spot in core calculations. Compared with the two-step method of component homogenization, the pin-by-pin calculation scheme only carries out the homogenization calculation for each cell in the component, which preserves the non-uniformity of different cells in the component, reduces the approximation and assumption in the core calculation, and can calculate the core number more accurately, which is conducive to the safety analysis of the core. However, the SP3/GSP3 methods currently in common use for pin-by-pin calculations still suffer from limitations in balancing computational accuracy and efficiency. In this regard, this paper aims to establish a core calculation method named quasi-diffusion theory which considers both computational efficiency and computational accuracy. First, the 3D quasi-diffusion equation was derived and established from neutron transport theory. Unlike the traditional diffusion equation, the formulation of the quasi-diffusion equation did not introduce a first-order approximation for the neutron angular flux, and the concept of the Eddington factor was introduced to deal with the leakage term of the neutron transport equation. The 3D quasi-diffusion equation was then formulated. In addition, a numerical solution model for the quasi-diffusion equation was developed by referring to traditional numerical solutions of the diffusion equation such as the transverse integral equation and using the nonlinear iterative cross section method. Second, since the Eddington factor was the key parameter in solving the quasi-diffusion equation, the solution of the quasi-diffusion equation depends on whether the Eddington factor was exact or not. In this paper, the Eddington factor was regarded as the homogenization parameter similar to other small group sections. The solution was obtained using Serpent, a modified version of the Monte Carlo program. Finally, several cases including the whole core model and the color-set assembly model were employed to validate the method, and the results were compared with the traditional diffusion and Monte Carlo reference solutions. By comparing the numerical output results under each model, the results show that for reactor cores with complex structures and strong heterogeneity, the accuracy of quasi-diffusion calculations is significantly better than that of traditional diffusion calculations, and the computationally efficient of the two methods is similar. In summary, the quasi-diffusion equation approximation, as a nonlinear method to approximate the neutron transport equation, is a computationally efficient and accurate method. It has great promise for new core calculations with strong neutron anisotropy and complex neutron energy spectrum.

Laser loading compression is one of the main research topics in the field of inertial confinement fusion. In order to simulate and explain the experimental phenomena of materials under high strain rates and pressures generated by laser shock, the submicron requirements for the adhesive layer thickness between the target multilayer materials in the state equation experiment were proposed in the physical experiments. Compared with the existing recombination methods, a novel adhesive recombination was reported in this paper. The poly glycidyl methacrylate (pGMA) film was prepared by initiating chemical vapor deposition method, and the liquid phase activation method was used to activate and cure the solid adhesive film. The submicron lossless bonding recombination of the state equation test target could be realized. In the activation process of pGMA film, the liquid phase activation method proposed in this paper will greatly improve the connection strength of multilayer materials without changing the thickness of the adhesive layer, which has not been reported. The thickness and morphology of the pGMA film were characterized, and the influence of process parameters on the film thickness was discussed. The deposition rate of pGMA film is about 10 nm/min in this paper. The white light confocal technology was used to indirectly measure the adhesive layer thickness of some nanoglue bonding samples. For the materials such as hydrocarbon and high-density carbon, the adhesive layer section of the nanoglue bonding samples was prepared by means of metallographic sample preparation combined with ion beam etching, and the adhesive layer thickness was directly characterized by scanning electron microscopy. The above results show that the adhesive layer thickness of the nanoglue bonding samples is submicron. And the thinnest adhesive layer of the nanoglue bonding samples is only 300 nm. The morphology of adhesive layer about the nanoglue bonding samples was characterized by optical microscope. The nanoadhesive layer connected the multilayer materials together evenly and seamlessly, and there were no gaps and air bubbles between the samples. The nanoglue bonding method reported in this paper has many advantages, such as precise control of the adhesive layer thickness, greatly improving the connection strength between the multilayer samples of the target in the equation of state experiment, realizing the precise assembly of the target with small size and complex configuration. This method will provide technical support for the improvement of the precision of laser loading compression physical experiment and the accurate interpretation of experimental results.

The fourth generation synchrotron radiation light source puts forward higher requirement for the alignment accuracy of the equipment, including the positioning accuracy of the equipment in a large space and the alignment monitoring accuracy of the equipment position after the machine is running. According to the alignment requirement of Hefei Advanced Light Source, the horizontal positioning accuracy within 30 m straight-line section is required to be better than 50 μm. In order to ensure the normal operation of the accelerator, new technical means must be used to monitor the lateral displacement of key components in the accelerator with high precision. The on-line alignment monitoring measurement is an important means to monitor the stability of particle accelerator system. Laser alignment with the characteristics of fast response and high accuracy, is an important data reference source for on-line alignment or adjustment. In this paper, a high-precision long-distance micro displacement monitoring system based on the similarity measure of laser spot was proposed. The lateral position offset of the measured system was monitored by measuring the displacement of laser imaging spot. Firstly, the structure and working principle of the laser light path control and detector assembly of the system were introduced, and the key errors of the system were analyzed, and then the similarity measurement algorithm of the light spot was used to monitor the lateral displacement. The similarity measurement algorithm used in this paper is an image processing algorithm that analyzes the similarity between the template and the search image. The similarity measurement algorithm takes advantage of the stability characteristics of the laser spot and uses interpolation function to calculate the minimum difference between the template and the search image, so as to obtain the position change value of the spot in the search image. Under the influence of random noise, the accuracy of the algorithm is better than ±0.17 pixels (pix). Finally, a 40 m long prototype system was built. Using the similarity measurement algorithm, the relative position deviation of the measured targets at different monitoring positions was obtained at different times. After testing, the horizontal monitoring accuracy of the system is better than 8 μm and the monitoring accuracy of the prototype is not related to the distance of the measured target. The successful development of the prototype system can provide basic data for the development of further micro displacement detection system, and can be applied to the on-line displacement monitoring of accelerator equipment such as fourth generation synchrotron radiation light sources, free electron lasers, linear colliders, etc.

An accurate two-phase flow model is an important basis for nuclear reactor safety analysis. The classical two-fluid six-equation model used in many current thermal-hydraulic programs has an ill-posed problem, if a small disturbance is introduced, the system parameters will oscillate greatly, resulting in unstable calculation results. In the existing researches at home and abroad, the most commonly used solutions to this problem are adding the virtual mass force or additional interphase interface pressure. However, the model can only be improved to a certain extent, and the ill-posedness of the model cannot be completely eliminated. The effective solution is to restore the two-phase pressure nonequilibrium and increase the volume transport differential equation to form a twopressure seven-equation model. In addition, a relaxation term is introduced to describe the multi-dimensional motion phenomenon of the interphase interface, and the concept of a relaxationtype twopressure model is proposed. This paper expounded the establishment method of the relaxation two-pressure model in LWR systems, discussed the processing methods and significance of the interphase interface pressurevelocity relaxation process, summarized the calculation method of the twophase mobile interphase interface, and introduced the numerical solution method of the model. Through eigenvalue analysis, the relaxed twopressure twofluid model studied in the paper is absolutely well-posed, which can solve the ill-posed problem of the original sixequation model. The relaxation term in the model can accurately describe the multi-dimensional motion on the interphase interface. There are two ways to obtain the relaxation coefficient, one is based on the relaxation time scale and the other is based on the acoustic impedance. The expression based on the acoustic impedance is only valid under high-speed flow conditions. For lowspeed flow, the first expression related to the traditional viscous resistance is more appropriate. The Riemann problem solution can be used to approximate the pressure and velocity on the interphase interface to solve the two-phase discontinuity at the interface. In the numerical solution of the model, a large number of existing researches use the step-by-step method to solve the problem, and the solution generates more intermediate variables and occupies a large space. There are also studies which solve the model as a whole, which preprocess the mass conservation equation and momentum conservation equation into the form of sumdifference differential equations to overcome the numerical instability problem caused by the disappearance of vapor/liquid phase. This paper summarized the existing research results and analyzed the future development direction of the relaxation twopressure twophase flow model, in order to provide theoretical reference for subsequent research.

Venturi-type bubble generator as well as its performance has been received wide attentions by engineers and researchers in recent decades. In this paper, experimental research was carried out to study flow regime and pressure drop of the gas-liquid (air-water) twophase through a Venturi bubble generator. The bubble generator was designed with a rectangular cross section, and the gas was injected into the high speed water flow from the throat section of the Venturi. Interface distribution and dynamic processes of bubbles were observed with a high speed camera, and the static pressures of inlet, throat and outlet of the bubble generator were measured by pressure sensors. Such three basic regimes as bubbly, slug and column are preliminarily identified in the bubble generator by visual experimental observations. A significant increase of the pressure drop of the bubble generator is found in the experiment when the flow regime transformed from slug into column. In order to identify the flow regime of the bubble generator, time frequency of pressure signal was analyzed, and it is found that the outlet section of the bubble generator is confirmed as the most favorable position for pressure signal on-line monitoring. A relative kurtosis of probability density function and a coefficient of variation of power spectrum density strongly associated with flow regime transition were extracted from the pressure signals, and they were applied to identify slug-column and bubbly-slug transition, respectively. Since the traditional pressure drop models can’t precisely predict the whole pressure drop of the Venturi bubble generator, a new pressure drop prediction method is proposed. The effects of partial liquid single-phase flow and flow regime transition on the pressure drop prediction are considered in the new correlation by introduced two modified parameters, ′ and θ′, respectively. Both of the two modified parameters were fitted from the current experimental results. Prediction results of the new correlation agree well with those of the experiments with a relative root mean square deviation of 10.74%. The new correlation is applied to the two-phase flow under conditions of normal temperature and 1-1 times of atmospheric pressure, while the liquid flow Reynolds number ranges from 1.7×104 to 7.0×104 and the low mass quality ranges from 0 to 5.1×10-5. The new method proposed in this paper for predicting pressure drop of the device regarding gas-liquid two-phase flow can provide theoretical bases and references for the online monitoring of the similar device.

The structural materials in the reactor and the corrosion products in the loop will be activated into radionuclides after being irradiated by strong neutrons. These nuclides and their decay products are the main sources of radiation hazards for workers. Therefore, the efficient and accurate calculation of the inventory of these radionuclides is of great significance for the design of reactor shielding protection, radioactive source items and waste management. In this paper, the neutron activation calculation method based on the mini-max polynomial approximation (MMPA) method was studied. Compared with the traditional neutron activation calculation methods such as the transmutation trajectory analysis method and exponential Euler method, MMPA method has the advantages of good numerical stability, high efficiency and no need to deal with short-lived nuclides alone. Furthermore, two solution strategies based on the MMPA method were studied for activation calculation, including the direct inversion solution strategy and the iterative solution strategy. The first solution strategy was proposed by Yosuke Kawamoto. This solution strategy completed the solution of the nuclide density by directly inverting the coefficient matrix and then calculating the multiplication and addition of multiple matrices. This solution strategy was computationally inefficient because it involved matrix inversion operations and a large number of matrix and vector multiplication operations. Therefore, a second solution strategy was proposed to improve the computational efficiency of the MMPA method. This solution strategy firstly transformed the solution expression of the MMPA method into an iterative expression after appropriate transformation, then obtained the substitution matrix for each iteration based on LU decomposition, and finally completed the solution of the nucleon density by adding multiple matrices. Compared with the first solution strategy, the solution strategy avoids the direct inversion of the matrix, and effectively reduces the number of multiplications of the matrix, so it has higher solution efficiency. Based on the self-developed nuclide inventory calculation code AMAC, the two solution strategies of MMPA method were completed. The correctness of MMPA method applied to neutron activation calculation under multiple conditions was preliminarily verified by typical material activation examples and self-constructed large-scale coefficient matrix examples under irradiation conditions, decay conditions and pulse conditions. The test results show that the calculation results of MMPA method are in good agreement with the calculation results of each reference solution, and the calculation accuracy is equivalent to that of 16 order Chebyshev rational approximation method (CRAM). In terms of computational efficiency, the solution efficiency of the iterative calculation strategy based on MMPA proposed in this paper is significantly higher than that based on direct inversion. The computational efficiency of this calculation strategy is equivalent to that of 16 order CRAM. The MMPA method is feasible in neutron activation calculation, and has good calculation accuracy and high calculation efficiency.

Rod-pinch diode is widely used in flash X-ray machines of various voltage levels. Its structure is very simple and the radiation dose is high with a small focal spot. The most important thing in the physics of rodpinch diode is to study the emission state of electrons and ions under different voltage-loading waveforms. At present, the theory of electron explosion emission has been mature, but there is a lack of specific research on the property of plasma emission except that anode plasma is indispensable. Therefore, the study of anode plasma emission of the rod-pinch diode is helpful to deeply understand the influencing factor of plasma on radiation dose. Improving the radiation imaging performance of rod-pinch diode by adjusting plasma emission, like pre-filled rod-pinch diode, PIC and Monte Carlo simulations were used in this paper. PIC simulation was used to obtain the peak current of the rod-pinch diode and the phase space file storing the electron bombarding the anode. The phase space file was used as the input source file of Monte Carlo to obtain the simulated radiation dose of the rod-pinch diode. Three X-ray devices of 450 kV, 1 MV and 4 MV were used to obtain the diode current and radiation dose at 1 m of the rod-pinch diode at the voltage level of 0.45-4 MV. By comparing the simulation result with the experimental result, the correct plasma emission function of the anode rod was obtained. According to the above methods, it is confirmed that the plasma emission function is closely related to the energy and power of electron bombarding the anode rod, and the ion emission function of anode rod with different diameters at the voltage level of 0.45-4 MV is obtained. At the same time, it is found that the position of the electron bombardment anode rod corresponds to the position of plasma emission. It shows that the heat transfer of electron deposition energy on the anode rod cannot be considered in the time range of X-ray emission. After the anode ion emission function obtained in this paper was added to the numerical calculation model, the relative error between the simulated peak current and radiation dose and the real experimental value is within 20%, which can accurately predict the diode current and radiation dose in the flash X-ray experiment. Using this model, the relationship between plasma density and electron bombarding anode rod was studied. Electrons bombarding the side of the anode rod with near grazing incidence and the end face of the anode rod can obtain a higher X-ray radiation dose. Combined with 3D modeling technology, the simulation method can fully evaluate the penetration ability of X-ray in a specific object, to scientifically select the input voltage level in the formal experiment. At the same time, combined with material science and related technology, the model can also be used for the design of a new X-ray diode.

Due to the small thermal neutron absorption cross section of 11B isotope, high abundance 11BF3 is regarded as an important ion implanted gas and P-type doping source, and is mainly used in semiconductor material manufacturing process to significantly improve the anti-radiation interference ability of semiconductor devices. In order to achieve the optimal design of the high-abundance 11BF3 isotope separation process, the mechanism of chemical exchange separation of 11BF3 isotopes was analyzed firstly, and the results show that the difference of intermolecular forces is responsible for the isotope separation. Then, the relative molecular weight and saturated vapor pressure data of isotope molecules were supplemented by self-defined components, and the parameters of the extended Antoine equation were regression extended in the Aspen Plus commercial chemical process simulation software. The process model of 11BF3 isotope separation based on Aspen Plus was established. RStoic reactor module was used for complexation and cracking reaction calculation, and Radfrac strict rectification module was used for exchange process calculation. Since the separation coefficient calculated by using NRTL-RK thermodynamic method was the closest to the experimental value, NRTL-RK thermodynamic method was selected for calculation in the exchange tower. In the complex tower and the cracking tower, the gas-liquid exchange process was not considered, so NRTL thermodynamic method was used. Next, based on the two sets of experimental data, the Murfree plate efficiency was used to correct the model in Aspen Plus, and the relative errors between the two sets of simulated and experimental values are -2.69% and 0.91%, respectively. It shows that the modified model can describe the process of 11BF3 isotope separation well. Finally, according to the established mathematical model, the effects of theoretical plate number, reflux ratio and column top pressure on the 11BF3 isotopic abundance at the top of the exchange column were analyzed. The results show that the decrease of the column top pressure, the increase of the reflux ratio and the increase of the theoretical plate number are all beneficial to the 11BF3 isotopic abundance. The maximum abundance of the exchange tower reaches 99.95% when the column top pressure is atmospheric pressure, the theoretical plate number is 720 and the reflux ratio is 300. The results of this paper can provide a theoretical basis for the subsequent optimization design. Subsequently, the 11BF3 isotope separation process model established in this paper can be further used to conduct multi-objective comprehensive analysis of investment cost, operating cost and abundance.

The accurate calculation of pressurized water reactor burned fuel neutron source intensity plays a significant role in the subcritical reactivity measurement. The calculation methods of both spontaneous fission source and (α, n) neutron source were researched in this paper. There were multiple sources of alpha particle in reactor core, including 238Pu, 239Pu, 240Pu, 241Am, 242Cm, 244Cm, etc. Among these sources, the dominant source was coming from 242Cm decay. 242Cm approximation method as well as ratio fitting method were proposed for obtaining (α, n) neutron source intensity. In 242Cm approximation method, all alpha particle sources were neglected with the exception of 242Cm. The ORIGEN code was used to analyze the ratio of (α, n) to the spontaneous fission neutron of 242Cm for different enrichment assemblies at different shutdown time. It is found that the ratio of (α, n) to the spontaneous fission neutron of 242Cm is fixed around 0.191. Therefore, the neutron number density of spontaneous fission neutron is multiplied by the factor of 0-191 to obtain (α, n) neutron source intensity due to 242Cm decay. On account of neglecting all (α, n) neutron source intensity with the exception of 242Cm, it is foreseeable that 242Cm approximation method would underestimate (α, n) neutron source intensity. In ratio fitting method, the ratio of (α, n) to the total neutron source intensity was analyzed. It is found that the ratio of (α, n) to the total neutron source intensity is linearly related to the burunp when the burnup is greater than 20 000 MW·d/tU. Therefore, two fitting methods were proposed to obtain the ratio of (α, n) to the total neutron source intensity. The polynomial fitting method was used with burnup less than 20 000 MW·d/tU, on the other hand, the liner fitting method was adopted with burnup greater than 20 000 MW·d/tU. Based on the in-house core design code package, the burned fuel neutron source calculation module was developed. In order to account for reactor core space effect and burnup history effect, the micro-depletion correction method in core design code package was used to obtain accurate number densities of actinide nuclides. The developed module was verified by calculating fuel assembly examples. Comparing with the reference code, the maximum errors of total source intensity is about 5% provided by both 242Cm approximation method and ratio fitting method. With the developed burned fuel neutron source intensity calculation module, both radial and axial distributions of burned fuel neutron source intensity in the reactor core are obtained. These distributions are of significant importance to ex-core detector signal measurements. The work in this paper is an important technology support for subcritical rod worth measurement development.

The nuclear power plant is a complex technical facility with potential radioactive hazards. Therefore, timely identification and treatment of accidents are necessary to avoid serious consequences. In order to quickly obtain information on nuclear power plant accidents and provide a new auxiliary means for the analysis of operator safety, it is necessary to study the accident diagnosis method and establish an accurate and reliable accident diagnosis model suitable for nuclear power plants. However, it is tough to establish a complete and accurate model of the nuclear power plant based on the model-based diagnosis method. The method based on expert knowledge needs a lot of expert experience knowledge which is difficult to obtain. Based on continuous data accumulated, the data-driven method can adaptively map the nonlinear relationship between symptoms and accidents, which is more suitable for fault diagnosis of the nuclear power plants. The data obtained by online monitoring for nuclear power plants are time series data with high data dimensions. However, there are few accident data, which makes it challenging to meet the training requirements of standard deep neural networks such as convolution neural network and time convolution network. Therefore, according to the data characteristics and high accuracy requirements of the nuclear power plant, a nuclear power plant accident diagnosis method based on time convolution capsule network was proposed in this paper. Firstly, the high-dimensional data obtained from the real-time monitoring of the nuclear power plant were input into the time convolution kernel to extract the depth of time-series features information of the data and reduce the difficulty of calculating. Then, the depth vector features of the data were mined through the capsule network to maximize the use of the data features of the operational information, which could also get a good convergence effect in the case of fewer accident data. Finally, the accident simulation data of the nuclear power plant were obtained by Fuqing simulator to train and test the model. The simulation data of regular operations and nine other accidents were collected to test the accident diagnosis model built in this paper. When the training set only accounts for 20% of the data set, it can still quickly and stably converge and accurately diagnose the accident information, and the diagnosis accuracy is more than 99%. Finally, the method in this paper was compared with other conventional depth learning methods through experiments. The results show that the time convolution capsule network model has better convergence and diagnostic accuracy than traditional machine learning and in-depth learning methods when the accident data of the nuclear power simulation system are not enough. The research content of this paper can provide some theoretical basis and technical reference for the application of the fault diagnosis method in nuclear power plants.

ASME 2021 code provides constitutive equation for high temperature creep of 316H stainless steel. The paper aims to help engineers and researchers correctly evaluate the strain and creep damage of high-temperature equipment by using the equation. Firstly, the physical mechanism of the items was explained. Secondly, the relationship between crucial parameters with temperature and stress was analyzed. Thirdly, the applicable stress and time range of the 316H creep constitutive equation was briefly introduced. And fourthly, the prediction values of the constitutive equation was compared with the data value from the isochronous stress-strain curve in ASME code. Results show that the constitutive equation is described by three items, in which the first creep stage is described by the rapid initial transient and transient items, and the second creep stage is described by the steady-state creep item. The equation’s applicability of evaluating the strain and creep damage of high-temperature equipment is limited by the starting time of the third creep stage. Moreover the key parameters of the equations, such as εs, εr, decay too rapidly to be accurately predicted at high stress. This research finds that due to item n increases with increasing temperature, the minimum creep rate m and transient creep rate constant r decrease in 1 000-1 075 ℉. Meanwhile, based on contrasting the rapid transient creep rate constants s of the original paper and ASME code, item D of parameter s is incorrect in the latter. Based on comparing predicted strain obtained by equation and data from ASME code isochronous stress-strain curves, the paper finds that predicted values at 950, 1 050 and 1 150 ℉ will agree with isochronous stress-strain curves, but larger than the datum from curve at 1 000 ℉, resulting in a relatively conservative strain prediction. Based on above four research and results, the paper concludes that the high temperature creep constitutive equation of 316H stainless steel in the ASME code can be used to evaluate high-temperature equipment structural integrity at 950, 1 050 and 1 150 ℉, but the result of 1 000 ℉ is conservatism.

Heat-pipe-cooled reactors offer many advantages, such as modularity and no flow loops. In recent years, it has attracted the attention of many scholars. Solid cores involve complex thermal-mechanical coupling behavior over their operating lifetime. The deformation of solid core structure has significant influence on its heat transfer performance. This paper introduced the gap heat transfer model by Ross and Stoute, which is an important part of the heat transfer process in the core. The creep and swelling behavior of materials used in solid core under high temperature and irradiation environment was briefly described. In this paper, a simplified finite element model of solid core was developed based on the secondary development of ABAQUS by programming subroutines, which considered the effect of gap heat transfer, creep and swelling behavior in the core. Based on the Ross-Stoute gap heat transfer model, the subroutine GAPCON was developed to simulate the gap heat transfer behavior in the core. The subroutine CREEP was developed to simulate the complex mechanical behavior of core components under different temperatures and radiation environments. Typical characteristics of temperature and stress distribution in each component during the normal operation period of 5 years were obtained. The thermal-mechanic coupling behaviors of solid core and the interaction between components were analyzed. The high temperature area appears in the center of the fuel pellets, and the temperature drops along the radial direction of the fuel pellets. This distribution of temperature field results in the inner extrusion and the outer tension of fuel pellets. The annular tensile stress region appears obviously outside the fuel pellets. These temperature and stress distribution characteristics can partly explain the radial cracking behavior of cylindrical fuel pellets. Furthermore, the designed size of the solid core was optimized based on finite element and machine learning to obtain lower temperature and stress peaks. The parametric modeling technique of ABAQUS was used to generate 5 800 sets of data, and the sample database was established for the training and testing of deep neural network machine learning. A surrogate model of core design parameters, maximum temperature and Mises stress in the reactor was established. The surrogate model had good predictive accuracy. Based on the surrogate model, pareto front was obtained by NSGA-Ⅱ algorithm and the optimized design parameters were obtained. The results of core temperature and stress optimized by surrogate model were verified by finite element method. Compared with the initial design, the maximum temperature decreases by 8.44% and the maximum stress decreases by 34.43%. The performance of the core is improved.

The construction of small modular reactor design is mostly for commercial power generation, and there are certain limitations in maneuverability and flexibility. In order to meet the demand of energy supply for special tasks, the vehicle-mounted very small modular reactor with more maneuverability arises at the right moment. In practical operation, vehicle-mounted reactor system will have multi-degree-of-freedom space movement, and the influence of vehicle movement on the reactor system is not clear at present. In this paper, the complex motion forms of the vehicle-mounted nuclear energy system running on the road were simplified into three types of operation forms. Based on the fluid force analysis in the non-inertial system, the additional force of vehicle-mounted motion conditions was derived from the momentum conservation equation, and a mathematical and physical model of vehicle-mounted motion conditions was established. The influence of additional force under motion condition on liquid lead-bismuth eutectic was analyzed reasonably, and corresponding additional force model was added into LETHAC program of transient thermal-hydraulic analysis program of lead-bismuth fast reactor to analyze thermal-hydraulic safety characteristics of LESMOR system under motion condition. The results show that the natural circulation flow of LESMOR system decreases significantly and the core cladding temperature and fuel temperature increase significantly under ramp driving, vertical vibration, pitching vibration and vertical coupled pitching vibration. Slope has little influence on the natural circulation of the system, and the change amplitude of system flow rate and temperature is small. The vertical vibration and pitching vibration will decrease the time domain average flow rate of the system and increase the time domain average temperature of the core outlet coolant. Due to the arrangement of the system, the pitching vibration has more influence on the heat exchanger than on the core. When the pitching vibration and vertical vibration are coupled, the average flow rate of the system in time domain decreases, and the vertical vibration plays a dominant role in the core flow rate. Under the limit condition, the transient drop of natural circulation flow of the system is not more than 30%, and the transient temperature of cladding and fuel is up to 411.3 ℃ and 442.4 ℃, and the rise is not more than 4 ℃, which is far below the safety limit of transient operation. The LESMOR system maintains stable and safe operation in both normal and possible driving limits. This paper provides a method to analyze thermal safety characteristics of reactor system under vehicle-mounted motion, which has certain reference significance.

Fast-spectrum micro-reactor has many technical advantages, such as high energy density, good mobility, fast deployment speed, long life and so on. Because of the characteristic of compact structure and high role of leakages in the neutron balance in fast-spectrum micro-reactor core, geometric deformation caused by temperature change and mechanical effect under external force is the main source of reactivity feedback. So it is very important to carry out transient analysis based on neutronics-thermo-mechanics coupling to accurately simulate the dynamic characteristics of fast-spectrum micro-reactor core. Based on the open source multi-physics coupling framework MOOSE, a three-dimensional coupled neutronics thermo-mechanics model for the dynamics analysis of fast-spectrum micro-reactors was developed in this paper. The thermo-mechanics coupling was solved by using the JFNK (Jacobian-Free Newton Krylov) method. Then the neutronics model was coupled with thermal-mechanics model by Picard iteration method through the MultiApp and Transfer system based on MOOSE framework. The neutronics solver can directly simulate the neutronics behavior of deformed nuclear reactor configurations including automatic adapting of the cross sections for account for density change. In this paper, this method was applied to the simulation of the supercritical transient of Goidva-Ⅰ with three-dimensional coupled neutronics thermo-mechanics model. Time evolution of fission rate, average temperature rise, surface displacement and surface velocity in the transient with an initial period of 16.2 μs were evaluated. The numerical result of fission rate during the transient was compared with the experimental data, and an overall good agreement was observed. From the numerical results, it can be seen that the fission rate increases, then reaches a maximum value with the positive reactivity introduced. Goidva-Ⅰ is a typical compact, fast-spectrum reactors, reactivity feedback is dominated by core deformation. The quickly accumulated thermal power causes the core temperature increasing, the reactor materials expand, resulting in a larger geometry and reduced material densities. Fission rate then drops due to loss of criticality caused by the thermal expansion. Due to the rapid increase and decrease of fission power, the core is in a state of compression and expansion one after another, and the thermal inertia effect leads to the oscillation of core surface displacement and velocity. At the same time, due to the change of core reactivity caused by deformation, the core power also produces periodic oscillations after 500 μs. The results show that this method can accurately consider the reactivity feedback effect caused by the thermal expansion, which lays a foundation for further multi-physics coupling safety analysis of fastspectrum micro reactors.

Flow boiling is widely encountered in many industrial fields such as the steam generator in nuclear reactors. Nowadays, a lot of studies are focused on the bubble dynamics parameters due to the close relationship between the heat transfer and bubble behaviors. Bubble departure frequency is one of the important bubble dynamic parameters which reflects the bubble cycle in time and thus directly influences the bubble evaporation heat. In order to improve the predicting accuracy of bubble departure frequency and thus improve the accuracy of heat transfer calculation under flow boiling, the affecting factors of bubble departure frequency firstly through 13 groups of flow boiling independent experiments of water (506 experimental data points) were analyzed in this paper. It is found that bubble departure frequency decreases with the increasing of liquid subcooling, while it increases with the increasing of channel size, mass flux, wall superheating and pressure in which the effect of channel size is often ignored by many researchers. Then the accuracy of five existing prediction models was evaluated. The results show that none of them have high prediction accuracy in all experimental data points. Cole model and Zuber model both perform badly and have the absolute errors of 163.2% and 218.2% respectively and the relative errors of 134.2% and 205.1% respectively, because they only consider the effect of physical properties and thus they are more suitable for pool boiling. Brooks model and Basu model both overestimate the results with the absolute errors of 280.2% and 102.1% respectively and the relative errors of 2352% and 37.1% respectively. The reason lies in that these two models are developed through their own experiment data and thus have limited application ranges. Chen model performs relatively better with the absolute error of 64.8%, but it largely underestimates the results with the relative error of -52.3% for the reason that this model is developed based on the pool boiling experiments of methane and ignored the effects of mass flux and liquid subcooling. Besides, the results show that most predicting models perform badly when the channel size is small which confirm the significant influence of channel size. Therefore, in this paper, a new prediction model was developed with higher predicting accuracy through dimensionless analysis which consideres the effects of channel size, physical properties, wall superheating and mass flux. This new model has the mean absolute error of 39.2% and mean relative error of -14.3% with the application ranges of bubble Reynold number 28-2 303, Bond number 1-9.272 and Jakob number 0.004-0.049.

In the two-phase flow system, the flow pattern affects the frictional resistance and heat transfer characteristics of the system. Accurate identification of different flow patterns is of great significance to the calculation of twophase flow. For the gasliquid flow in narrow channels, especially the flow pattern and flow pattern transition criterion in rectangular narrow channels, some researchers have done some experimental studies. For the study of flow pattern in narrow rectangular channels, the classification of flow patterns is basically consistent with that in conventional rectangular channels, including bubbly flow, slug flow, churn flow and annular flow. However, the performance of existing flow pattern transition criteria for twophase flow in narrow rectangular channels still needs further and comprehensive assessments. In this paper, a parameter representing the success rate of boundary separation of bubblyslug flow transition was introduced to quantitatively evaluate the performances of five typical bubbly-slug flow transition criteria and experimental data reported by 10 previous researchers, covering channel sizes ranging from 0.6 mm to 30 mm. Results show that the success rate of boundary separation of the Jones & Zuber criterion, Taitel criterion, MishimaIshii criterion and Xu criterion in predicting the experimental data varies from 61.86% to 98.10%. The variation mainly attributes to the assumption of constant critical void fraction in these criterions, since analyzing shows that the critical void fraction of flow pattern transition is related to the channel geometry. Although the HibikiMishima criterion considers the effect of channel size on the critical void fraction, the bubbly diameter (Db) is also needed in this model, which is difficult to obtain. The influence of channel size on the flow transition was quantitatively analyzed based on a dimensionless parameter of confinement number (Co). Finally, the flow pattern transition function (f(k)) was correlated with Co, and a new flow pattern transition criterion of bubbly-slug flows in narrow channel considering the physical properties of working fluid and the size of flow channel with higher accuracy and wider application range was established, which predicts the experimental data with success rates of boundary separation of 83.78% to 98.10% within the range of 0.486 57≤Co≤1.471 04, 0.016 m/s≤jg≤6.758 m/s, and 0.058 m/s≤jf≤5.059 m/s. The conclusions of this paper can provide references for the design and optimization of reactor heat exchanger components and compact heat exchangers.

Advanced nuclear reactors are currently moving toward longer fuel cycles and higher burnup levels. Besides, high burnup levels can improve fuel economy and reduce the total amount of spent fuel in reactors. The formation of high burnup fuel particle is closely related to the failure behavior of ceramic dispersion fuel, and it is of great significance to focus on the strength analysis and cracking behavior of high burnup fuel particle. The representative volume element of dispersion fuel was innovatively established in this paper, and the heterogeneity of pore size and position in high burnup fuel particle was considered, corresponding parametric modeling and mechanical simulation procedure of high burnup fuel particle was realized automatically. The impacts of mechanical property of matrix, distance between fuel particles, ambient hydrostatic pressure and damaged layer on the strength of the fuel particles were analyzed systematically. The results show that the distance between fuel particles increases, the less possibility of crack initiates from the fuel particle. The existence of damaged layer will increase the risk of cracking of the fuel particles slightly. Due to the stress interference of the heterogeneous pores, crack initiates from multiple sites within the fuel particle, and the danger zones are more possibility located at the outer space, which is consistent with the experimental observations. Besides, danger zones within the high burnup fuel particle were characterized by larger pores and pore strings. It can be inferred that the formation of larger pores is probably due to the continuous coalescence of small pores and their surrounding pores. The constraint pressure of matrix on the fuel particle is not a constant, specially, it fluctuates significantly, and its average value increases linearly with the ambient hydrostatic pressure. Increasing the elasticity modulus of the matrix will inhibit the cracking behavior of the fuel particle, which reveals that the matrix should be designed more stronger and tougher. The higher the ambient hydrostatic pressure, the less likely the fuel particle will crack. While shortening the distance between fuel particles, and the impact degree of ambient hydrostatic pressure reduces. This work realizes the accurate calculation of the internal stress field of high burnup fuel particle and establishes a high burnup fuel cracking strength evaluation model based on fine microstructure modeling. Besides, this paper provides an analytical method and numerical reference for the failure study and optimal design of dispersion fuel under high burnup conditions.

Nuclear thermalpropulsion reactor has attracted more and more attention because of its long service life, strong mobility and high concealment. However, in order to achieve higher mobility, nuclear thermalpropulsion reactor reduces its weight and does not have perfect radiation shielding measures, which will cause a lot of radiation to the external environment after operation. The radiation has an important impact on the application and development of nuclear thermal propulsion reactor. Therefore, it is necessary to study its radiation characteristics. This paper studies the radiation caused by the radioactive fission products produced after the operation of “Pluto” TORY ⅡC reactor published by the United States in the last century, ignoring the direct radiation of chain fission reaction, the radioactivity of component activation and the leakage of radioactive gas. Firstly in this paper, the burnup products of the reaction after 10 hours of full power operation were calculated by Monte Carlo method and the energy spectra of 106 radionuclides with high radioactivity were calculated, including Xray, gamma ray, bremsstrahlung and spontaneous fission gamma ray. Then the radiation field outside the reactor was calculated by pointkernel integral method. Because the pointkernel integral method has higher accuracy in calculating the simple structure, the reactor geometry was simplified in calculating the radiation field in this paper. The calculation results show that the reactor consumes 056% 235U after operation. The radioactivity increases gradually with time and finally reaches 1298×1021 Bq. Gamma radiation accounts for the main part. The maximum dose rate on the outer surface of TORY ⅡC shielding layer can reach 67185 Sv/h. Because the rear reflection layer is the thinnest and the side reflection layer is the thickest, the dose rate at the tail is the highest and the side dose rate is the lowest. At the same time, the calculated radiation field maintains a high dose rate in a large range. Therefore, this paper concludes that TORY ⅡC will produce a large amount of radiation after operation. Radiation shielding facilities must be installed to reduce the area of high radiation area, protect the environment and ensure personnel safety. At the same time, this paper calculates that the burnup products are regarded as uniform distribution and compares them with the previous accurate calculation results. It is found that there is little difference between the two calculation results in a far position. It can be inferred that when the distance is far enough, the two calculation results will be infinitely close. When the accuracy is not pursued, the burnup products can be regarded as uniform distribution to simplify the calculation and speed up the calculation time.

The complex cosmic thermal environment is a great challenge to the orbit injection of space reactors. Improper design will lead to liquid metal condensation and hinder the safe start-up of reactors. Therefore, for the safe start-up of liquid metal cooled space reactor, the liquid metal in the system must be insulated to prevent condensation on orbit. In this paper, the onorbit insulation strategy of space thermionic reactor was proposed and a model was established for calculation and verification. In order to simulate the influence of space thermal environment and verify the insulation strategy, this paper established a improved TOPAZ Ⅱ space thermionic reactor model based on the transient analysis code of space thermionic reactor system, TASTIN. The calculation model was divided into several control volumes. For each control volume, the governing equations were discretized based on the finite volume method (FVM), and the transient equations were solved by Gear algorithm. Moreover, the radiator heat transfer model (TASTIN-RAD) and the radiant heat transfer model of the heat shield (TASTIN-SHIELD) were developed for calculation of thermal environment (mainly solar heat flux) and different heat shields respectively. Then the spin and non-spin states of the space reactor on orbit were calculated with considering the solar heat flux. The results show that the spin state is better than the non-spin state in the first flight cycle, while in the subsequent flight cycle, the nonspin state is better than the spin state. After 24 hours’ flight of nonspin in orbit, the temperature of fluid is close to the freezing point and additional thermal insulation measures shall be taken. Thus, different insulation strategies including open and closed heat shields are calculated and analyzed in nonspin state for 24 hours flight. By adding an open heat shield, the solar heat flux has a significant heating effect on the fluid. The minimum temperature of the fluid after 24 hours of flight is 295 K, which can ensure that the fluid does not solidify. Under the closed heat shield scheme, the top of the heat shield has a special absorption emission ratio, which can heat the system circuit. After 24 hours orbit flight, the minimum temperature of the circuit fluid reaches 303 K. According to the calculation results, only depending on the flight attitude and the heating effect of solar heat flux is not enough to prevent the condensation of liquid metal working medium, so additional insulation measures must be added. Both insulation schemes proposed by this paper effectively ensure the long-term heat dissipation of the system and have high safety margin. The difference is that the closed shield strategy can make the fluid temperature rise slowly, while it decreasing in the open shield strategy, which means the close shield strategy has absolute security. The calculation results of this paper provide a basis for the insulation scheme design of space thermionic reactor.

In the process of long-term operation of steam generators, although the secondary side of the water quality is strictly controlled, in the complex thermal hydraulic conditions, impurities will also accumulate in the steam generator. It mainly includes the deposition of the heat transfer tube and the blocking of the tube support plate. Among them, the deposition of impurities on the supporting plate will cause the blocking of the tube support plate. The blockage of the support plate may lead to local highspeed flow and cross flow in steam generators, which will cause flowinduced vibration and tube cracks in some situations. Therefore, it is necessary to study the deposition of the tube support plate. In order to predict the deposit distribution on the tube support plate of the steam generator, the impurity migration model and the tube support plate flow blockage model were added to the threedimensional thermohydraulic code of the steam generator STAF developed by NuTHeL of Xi’an Jiaotong University. The accuracy of the code, which was based on a porous medium method and the drift flux model, has been fairly well verified. The deposition on the tube support plate with and without ageing for 150 days were predicted in the paper. Then the influence of ageing was studied. First of all, the results show that there are more deposits in the upper part of the hot side of the tube support plate, which suggests that boiling is the main factor leading to deposition. Compared with the situation without ageing, the mass of deposits after ageing is reduced, and the difference of the deposition mass between the hot and cold sides is also significantly reduced. In addition, ageing also has a certain effect on the distribution of deposits. After ageing, the distribution of deposits is closer to the two sides of the tube support plate, which is similar to the distribution of flow velocity. The reason for this phenomenon is that flash evaporation is more likely to occur where the flow rate is high, making the soluble precipitate and thus exacerbating the ageing. The research of this paper can provide reference for the design and cleaning of steam generator in nuclear power plant. In practical applications, we can take some measures through the program prediction to reduce the harm caused by the tube support plate clogging. In the future work, we will continue to study the influence of tube support plate blockage.