Shanghai High repetition rate XFEL (X-ray free electron laser) and Extreme light facility (SHINE) is a high-repetition-rate X-ray-free electron laser. The timing system of the beamlines and endstations must provide high-precision bunch IDs and a timing trigger for the equipment that works in single-pulse mode.

This study aims to design a data acquisition (DAQ) testing system to simultaneously acquire X-ray bunch IDs with their corresponding detector data package for subsequent data processing.

This DAQ testing system was developed on the Zynq UltraScale+ system-on-chip (SOC), and the White Rabbit protocol was employed for the timing system environment. A Bunch ID obtained from the FPGA mezzanine card (FMC) of the embedded White Rabbit node (WRN) was transferred to the server using a TCP protocol stack built on LwIP (light weight internet protocol). Finally, a Basler camera was employed to test this DAQ system, in which the pypylon library was applied to raw data acquisition software for camera snapshot whilst two channels of data were collected by an upper computer and saved to a database for comparison.

The number of bunch IDs obtained by this acquisition test system is the same as that of image frames taken by Basler camera, which demonstrates that the testing system can satisfy the requirements of bunch ID acquisition in SHINE beamlines and endstations.

The trend towards increasingly narrow apertures in multipole magnets poses a challenge to many conventional measurement methods. Consequently, these methods' applicability in small aperture multipole magnets is limited. However, the single stretched wire measurement technique has emerged as a promising alternative due to its minimal space requirements within the measurement domain. Therefore, this technique is well-suited for accurately measuring magnetic fields in small aperture magnets.

This study aims to introduce a novel approach for analyzing the gradient integral and multipole errors of the quadrupole magnet, to address the limitations associated with the current single stretched wire method (SSWM).

Firstly, a magnetic measurement system based on the single stretched wire method was constructed with two boasted key advantages: minimal space occupation within the measurement domain, and flexible motion modes. Then, leveraging these features, measurements of the four poles of a quadrupole magnet by employing a hyperbolic trajectory was acquired, and a new technique for analyzing both the gradient integral and multipole errors associated with the quadrupole magnet was developed. Finally, the feasibility of this SSWM was verified by comparing the results obtained from our system to those derived from the rotating coil method.

Measurement results of a quadrupole magnet with the inscribed radius of 11 mm and gradient of 100 T?m^-1 by SSWM show that the repeatability of three measurements is better than ±1.5×10^-4 which is less than one-third of the maximum value of multipole error of 5×10^-4, so it can meet the measurement requirements.

The methodology outlined in this study for constructing the measurement system and analyzing the resultant data offers a practical and effective solution for the future magnetic field measurements of small aperture magnets.

FeCrAl alloy cladding, as an accident tolerant fuel (ATF) mid-term commercial technology approach, has received extensive attention.

This study aims to investigate the effect of trace Y on the internal pressure burst and oxidation properties of FeCrAl alloy cladding.

Firstly, the crystalline grain size and micro-morphologies of FeCrAl and FeCrAlY alloy cladding samples were observed by optical microscope. Internal pressure burst and high temperature oxidation tests were carried out by burst test equipment and thermo-gravimetric analyzer with a moisture generator. Then, X-ray diffractometry (XRD), scanning electron microscope (SEM) and energy dispersive spectrometer (EDS) were employed to analyze the composition of oxidation products, surface and cross-sectional micro-morphologies of FeCrAl and FeCrAlY alloy cladding samples before and after high-temperature stream oxidation and the distribution of elements on the surface oxidation products.

The results show that trace Y is mainly dissolved in the FeCrAl alloy matrix, and no Fe-Y phase is formed. The inclusion of Y do not change the burst strength and the rupture opening morphology at room temperature (RT) to 1 000 ℃, and the high-temperature steam oxidation resistance of FeCrAl alloy cladding is significantly improved by the trace Y. Under the condition of steam oxidation at 800 ℃, 1 000 ℃ and 1 200 °C for 8 h, the oxidation weight gain of FeCrAlY alloy cladding decrease by 65.1%, 60.0% and 31.5%, respectively. Compared with the single Al₂O₃ oxide film on the surface of FeCrAl alloy cladding, the Y-containing composite oxide film with lower internal stress, higher compactness and better adhesion with the substrate is formed on the surface of FeCrAlY alloy cladding.

Therefore, the addition of trace Y do not change the burst properties of FeCrAl alloy cladding, however, the high-temperature steam oxidation resistance of FeCrAl alloy cladding is significantly improved.

The accurate acquisition and correlation coincidence of positron annihilation signal form the basis of the lifetime spectrum sensitive characterization of microscopic defects in materials. The complex radiation background interferes with the acquisition of positron annihilation signals, particularly in the study of neutron radiation damage of nuclear structural materials. The γ ray background generated by radionuclides induced by neutron activation affects the measurement results of positron lifetime spectrometer.

This study aims to investigate the effect of γ background on positron annihilation lifetime measurement.

First of all, the positron lifetime measurement system is built in a "fast-fast coincidence" manner, and radiation background simulation experiments are designed by selecting two typical nuclides, ⁶⁰Co and ¹³⁷Cs sources, with nearby feature γ photon energy for measuring positron annihilation lifetime. Then, the spectra under two typical activity ratios are compared with the activated neutron-irradiated samples.

The simulation results indicate that the double high energy γ rays generated by ⁶⁰Co sources are the primary factors affecting the spectrum shape and lifetime components. When the ⁶⁰Co/²²Na activity ratio is relatively low, 1.9, the peak-to-valley ratio of the spectrum significantly degrades, with the increase of random coincidence probability caused by radiation background. Further, at high activity ratio of 3.3, besides random coincidence, the false coincidence probability increases sharply, and the spectral shape is evidently distorted. For neutron-irradiated RPV steel, the lifetime value is reduced by 17% and 46% at low and high activity, respectively, compared with the non-irradiated samples.

Using the simulation method of radiation background sources and the influence rules of interference γ in this study, new techniques for eliminating γ background could be further explored in positron annihilation lifetime measurement.

Double perovskites have become a research hotspot in recent years due to their flexible structure, easy doping, and good thermal stability. Photoluminescence (PL) of rare-earth-doped double perovskite materials has been frequently reported, but few studies on thermoluminescence (TL) have been conducted.

This study aims to investigate the TL characteristics of Y_2-x-yBi_xEu_yMgTiO₆ (0≤x<1, 0≤y<1) phosphors.

Bi³⁺ and Eu³⁺ co-doped Y₂MgTiO₆ samples were synthesized by a high-temperature solid phase method, and the X-ray diffraction (XRD), PL, and TL of the samples were measured.

XRD analysis results show that the crystal structures of all samples are monoclinic P2₁/n, and Bi³⁺ and Eu³⁺ are doped into Y₂MgTiO₆ by substituting Y³⁺. The PL results show that Y_1.79Bi_0.01Eu_0.20MgTiO₆ has a strong red emission near 620 nm (corresponding to the ⁵D₀→⁷F₂ transition of Eu³⁺), which is accompanied by a long afterglow. The TL curves of the samples doped with different concentrations of Bi and Eu ions show that Y_1.79Bi_0.01Eu_0.20MgTiO₆ has the highest TL sensitivity, and the samples exhibits two significant TL peaks near 510 K and 610 K. The TL spectrum is more abundant than the fluorescence spectrum, and the ⁵D₀→⁷F_J (J =1,2,3,4) transition of Eu³⁺ can be observed. The TL intensity of the sample has a good linear relationship with the irradiation dose in the range of 2~1 000 Gy. The TL kinetic parameters of the samples are analyzed using two methods under different preheating temperatures (T_m and T_stop) and glow curve deconvolution. The analysis results show that the depth of the TL trap in the sample extends from 0.80 eV to 1.40 eV.

The results of this study indicate that the TL spectrum is richer than the PL spectrum and that Y_1.79Bi_0.01Eu_0.20MgTiO₆ may be used as TL dosimeter material for large dose detection.

The performance of solid oxide fuel cells (SOFCs) can be promoted by optimizing cathode materials.

This study aims to boost the electrochemical performances of cathodes for SOFCs by doping transition metal at the B-site of double perovskite.

Firstly, a series of B-site doped PrBa_0.8Ca_0.2Co₂O_5+δ(PBCC) oxides as cathodes for SOFCs were prepared by sol-gel. The effects of B-site doped content and doped elements on the crystalline structure of the cathodes were analyzed by X-ray diffraction (XRD) and scanning electron microscope (SEM). Then, the trends of conductivity and thermal expansion coefficient with B-site doped PBCC oxides were investigated. Finally, the electrochemical performances of cathodes with different B-site doped PBCC oxides were tested to find optimal doping element type and content.

Test results show that polarization is reduced and the electrochemical catalytic activity is improved when 5 mol% of Fe is doped on the B-site of the PBCC cathode. Compared to the PBCC cathode, the max power density of the full cell with a 5-mol% Fe-doped cathode increases from 988 mW?cm^-2 to 1 259 mW?cm^-2 at 700 ℃.

The electrochemical performances of SOFCs can be boosted by modifying the B-site of double perovskite using transition metal.

The China Spallation Neutron Source (CSNS) is a multidisciplinary research platform. Its high-energy 1.6 GeV proton beam serves various applications in aerospace devices and particle detector testing. However, certain irradiation applications and high-performance detectors require different beam energies. A degrader was designed to adjust the proton energy to a desired range.

This study presents a reasonable degrader scheme for the 1.6 GeV proton test beam at the CSNS.

The physical process of the 1.6 GeV high-energy proton beam passing through a degrader made of either of three different materials (iron, copper, and tungsten) was simulated using FLUKA, a Monte Carlo particle transport code. Parameters such as the degrader thickness, the energy deposition, the outgoing proton beam intensity, and the irradiation dose were determined through simulations. The optimal degrader material was identified. In addition, a continuously adjustable structure of the degrader was given.

Iron displays slight advantages in terms of energy deposition and radiation dose distribution, compared to copper and tungsten. Furthermore, the phase-space distribution of the outgoing proton beam and the secondary pion beam were also given, providing important references for future beam-line design.

An optimal degrader structure made of iron for the CSNS high-energy proton beam is proposed. The secondary pion test beam is also feasible at the proton test end station. This is a significant development for future engineering design.

The triple-to-double coincidence ratio-?erenkov (TDCR-?erenkov) method can be applied to the activity measurement of radionuclides by detecting the ?erenkov photons produced in a non-scintillation solution. The computation of the detection efficiency of this method is based on the premise that the energy of emitted β is completely deposited in the solution. However, this precondition is ideal and does not apply to the actual measurement because of the counting loss caused by the restrictions of a finite solution and the wall of the counting vial (i.e., wall effect).

This study aims to analyze the influence of the wall effect on the computation of detection efficiency.

The transport process of emitted β from the solution to the vial wall was analyzed in sections. Thereby the relationship between the number of ?erenkov photons and the deposition energy spectra of emitted β with different energies in different matrices was obtained. This relationship was used to further improve the calculation model of the TDCR-?erenkov method. Subsequently, the calculation model was simplified to reduce the required time. Geant4 calculated the deposition spectra of emitted β in different matrices, subsequently, the efficiency of different nuclides was calculated using curves of the number of ?erenkov photons vs. the energy of the emitted β. To verify the accuracy of the improved calculation model, measurements were carried out on a variety of pure β-emitters.

The results derived from the improved TDCR-?erenkov method are in good agreement with those of the TDCR-LS method. Especially for high energy β-emitters, the relative deviation of the results between the TDCR-?erenkov and TDCR-LS methods is reduced from 0.47% for the original method to 0.02% (⁹⁰Y), and 0.64% to -0.16% (³²P).

The TDCR-?erenkov method is more accurate when considering the wall effect in the activity measurement of high-energy β-emitters.

The passive residual heat removal (PRHR) system is an important innovative design of the advanced pressurized water reactor technology. Under accident conditions, PRHR system can transport decay heat in the form of natural circulation to ensure core cooling. However, the heat exchange function of PRHR system will be lost when the PRHR pipeline breaks. With development of the accident process, the coupling effect between different safety equipments of the passive core cooling system (PXS) will be affected. Besides, thermal hydraulic state of the reactor coolant system (RCS) will also be affected via complex interaction mechanism. As a result, the new thermal hydraulic phenomena occur, and thus ultimately affecting the accident mitigation capacity of the PXS.

This study aims to confirm the safety characteristics of PXS and identify the new thermal hydraulic phenomena of advanced passive nuclear power plant during accident with multiple failures.

A series of integral effect tests of loss of coolant accident (LOCA) were conducted on the advanced core-cooling mechanism experiment (ACME) facility. The influence of failure of PRHR HX flow and heat exchanging function on LOCA accident process were investigated on the basis of the test cases including PRHR pipeline break and cold leg (CL) break. The unique thermal hydraulic phenomena occurred during PRHR LOCA were explored, and their influence laws on the coupling effect among PXS safety equipments, and the influence laws on thermal hydraulic state of RCS were obtained.

The results show that there is a momentary reverse flow and heat transfer process in PRHR HX at the beginning of PRHR LOCA compared with typical CL LOCA. Besides, the natural circulation process between the core and steam generators (SGs) plays a critical role in cooling and depressurization of RCS, and its corresponding time-averaged heat transfer power is increased by about 30%. Besides, the asymmetric arrangement of PXS leads to a significant difference of transient thermal hydraulic state between the RCS branches, namely the PRHR cools the coolant via one RCS loop while two core makeup tanks (CMTs) inject the cold water to the core via the other RCS loop, and the pipeline resistance distribution shows a significant impact on the injection performance of safety equipment with low driven head such as CMTs.

The unique and important thermal hydraulic phenomena in the early stage of the accident, namely reverse flow and heat transfer process in PRHR HX and natural circulation process between core and SGs, are identified. The asymmetric arrangement effect will be more noticeable when the break occurs in PRHR pipeline.

As an innovative technology of nuclear power, magnetically suspended high-temperature molten salt canned motor pump (referred to as molten salt canned motor pump) can be used in the fourth generation molten salt reactor (MSR). Improving pump performance via hydraulic optimization design is significant to fourth-generation nuclear power technology.

This study aims to investigate the influence of different working fluids on the hydraulic optimization design of magnetically suspended high-temperature molten salt-canned motor pumps and provide suggestions for the optimal design of magnetically suspended high-temperature molten salt-canned motor pumps.

Firstly, ANSYS CFX software was employed to perform a numerical simulation of a magnetically suspended high-temperature molten salt canned motor pump. Based on response surface methodology (RSM), approximate models between significant parameters and optimization objectives were established. Then, taking the efficiency and head as optimization objectives, a non-dominated sorting genetic algorithm II (NSGA-II) was used to design the magnetically suspended high-temperature molten salt canned motor pump under molten salt and water.

Compared with water working fluid, the optimization space of the pump under molten salt working fluid is larger. When the efficiency of the optimization model under the two working fluids is the same, the impeller inlet diameter and the blade outlet placement angle of the molten salt optimization model are reduced, whereas the impeller outlet width and the diffuser throat plane width are increased. The efficiency of the finally determined molten salt optimization model is increased by 0.75% and the head is raised by 0.078 2 m whilst the efficiency of the water optimization model is increased by 0.55%, and the head is reduced by 0.035 9 m.

The research results of this study can be used to guide the hydraulic structure design of a magnetically suspended high-temperature molten salt-canned motor pump.

With the increase of complexity of reactor core design, the core modeling and calculation have brought challenges.

This study aims to implement the accurate modeling and calculation of unstructured geometry core.

Based on discrete ordinate nodal method for arbitrary triangular-z geometry, the precise modeling and mesh generation of unstructured core were established by constructive solid geometry (CSG), and Block-Jacobi parallel algorithm was employed to reduce calculation time of reactor core. Finally, based on the developed SARAX program, core physics calculations for new complex geometries of a space reactor and a heat pipe reactor were performed for accuracy verification by using Block-Jacobi parallel algorithm combining with established precise model and mesh.

The verification results show that the effective multiplication factor and radial power distribution agree with that of multi-group Monte-Carlo calculation. The calculation deviation of eigenvalues is less than 3.00×10^-3, and the relative deviation of radial power distribution is less than 1.5%.

Results of this study show that SARAX code has the ability of modeling and higher accuracy in the calculation of unstructured geometry core.

In order to accurately predict the friction pressure drop characteristics of liquid lead bismuth in the cross-section of the fuel assembly rod bundle, a suitable friction pressure drop model should be selected.

This study aims to investigate Friction pressure drop model for wire-wrapped rod bundles in full flow.

Eight different frictional pressure drop models within wire-wrapped rod bundles were evaluated their applicability by using statistical analysis. The prediction accuracy of experimental data from different models in different flow regimes was explored corresponding to laminar flow, transitional flow, and turbulence.

The analysis results show that the friction coefficient is not only related to the number of rod bundles (Nr) and the pitch-to-diameter ratio (P/D), but also related to the wire lead length-to-diameter ratio (H/D). The modified BDD model in the laminar flow range and this work model are more consistent with the experimental data. The modified BDD model, CTD model and this work model are relatively consistent with the experimental data in the transition flow range. The Rehme model, the UCTD model and this work model in the turbulent range are more consistent with the experimental data.

Therefore, the model presented in this study is suitable for predicting friction pressure drop in the cross-section of the fuel assembly bundle in the full flow state.

The propagation of pressure waves in nuclear energy systems will cause hydraulic load effects, and it is particularly important for the analysis of structural loads to accurately simulate the propagation process of pressure waves. System analysis codes such as RELAP5, TRACE, etc. are widely used in the simulation and analysis of reactor pressure wave propagation. But system analysis codes can only simulate one-dimensional pressure wave propagation behavior.

To cope with the multi-directional and multi-dimensional pressure wave propagation issue, corresponding model and algorithm study is carried out in this paper to investigate the two-dimensional pressure wave propagation behavior in two-phase steam-water flow condition.

By employing a four-step algorithm of time-step separation, and a non-equilibrium phase transition heat transfer model, a two- dimensional two-phase flow pressure wave propagation code (TPFPWPC-2D) is developed based on 2D axisymmetric cylindrical coordinate system. The code verification is carried out by using a typical benchmark of steam-water two-phase shock tube. Finally, in order to verify the ability of TPFPWPC-2D code to simulate the two-dimensional propagation of pressure waves, numerical simulations of the pressure wave propagation behavior in a cylindrical space region were conducted.

The results of code verification show that the new code proposed here agrees well with the two system analysis codes RELAP5 and WAHA. The 2D simulation application shows that the new code can capture the 2D propagation processes of pressure wave reasonably, especially the reflection and superposition characteristics.

From the results mentioned above, conclusions can be drawn that the new code developed in this paper can simulate the two-dimensional axisymmetric propagation characteristics reasonably in both quantitative and qualitative levels.

Many existing studies have shown that the use of suitable surface modification methods can enhance the boiling heat transfer effect of metal components, making it have a broad potential application prospect in the pressurized water reactor. However, for the weak alkaline environment of high temperature and high pressure in the reactor, little literature is reported on whether this enhanced effect can be maintained for a long time.

This study aims to explore the effect of corrosion on boiling heat transfer characteristics of metal specimens with micro-structure surface.

First of all, three micro-structures of micro-groove, micro-porous and micro-columns were processed on the surface of stainless steel plate specimens by laser processing. Then the specimens were placed in the high-temperature and high-pressure environment simulating the actual reactor conditions to carry out corrosion experiments for up to 200 d. Finally, the pool boiling experiment and visualization study of the specimens before and after corrosion were carried out for comparison.

The results show that the surface critical heat flux (CHF) of the three micro-structured metal specimens increases and then decreases with the increase of corrosion time, among which the micro-pores specimens have the largest bubble generation rate at the beginning of nuclear boiling, and the micro-groove specimens have the highest CHF.

The influence law and mechanism of long-term corrosion in pressurized water reactor on the enhanced heat transfer effect of different micro-structure surfaces are partially revealed by this study.