Resolution analysis of thermal neutron radiography based on accelerator-driven compact neutron source

Owing to the immobility of traditional reactors and spallation neutron sources,the demand for compact thermal neutron radiography(CTNR)based on accelerator neutron sources has rapidly increased in industrial applications.Recently,thermal neutron radiography experiments based on a D-T neutron generator performed by Hefei Institutes of Physical Science indicated a significant resolution deviation between the experimental results and the values calculated using the traditional resolution model.The experimental result was up to 23%lower than the calculated result,which hinders the achievement of the design goal of a compact neutron radiography system.A GEANT4 Monte Carlo code was developed to simulate the CTNR process,aiming to identify the key factors leading to resolution deviation.The effects of a low collimation ratio and high-energy neutrons were analyzed based on the neutron beam environment of the CTNR system.The results showed that the deviation was primarily caused by geometric distortion at low collimation ratios and radiation noise induced by highenergy neutrons.Additionally,the theoretical model was modified by considering the imaging position and radiation noise factors.The modified theoretical model was in good agreement with the experimental results,and the maximum deviation was reduced to 4.22%.This can be useful for the high-precision design of CTNR systems.

Measurement of the cavity-loaded quality factor in superconducting radio-frequency systems with mismatched source impedance

The accurate measurement of parameters such as the cavity-loaded quality factor(Q_(L))and half bandwidth(f_(0.5))is essential for monitoring the performance of superconducting radio-frequency cavities.However,the conventional"field decay method"employed to calibrate these values requires the cavity to satisfy a"zero-input"condition.This can be challenging when the source impedance is mismatched and produce nonzero forward signals(V_(f))that significantly affect the measurement accuracy.To address this limitation,we developed a modified version of the"field decay method"based on the cavity differential equation.The proposed approach enables the precise calibration of f_(0.5) even under mismatch conditions.We tested the proposed approach on the SRF cavities of the Chinese Accelerator-Driven System Front-End Demo Superconducting Linac and compared the results with those obtained from a network analyzer.The two sets of results were consistent,indicating the usefulness of the proposed approach.

Recent progress in nuclear data measurement for ADS at IMP

Recent progress in nuclear data measurement for ADS at Institute of Modern Physics is reviewed briefly.Based on the cooler storage ring of the Heavy Ion Research Facility in Lanzhou, nuclear data terminal was established.The nuclear data measurement facility for the ADS spallation target has been constructed, which provides a very important platform for the experimental measurements of spallation reactions. A number of experiments have been conducted in the nuclear data terminal. A Neutron Time-of-Flight(NTOF)spectrometer was developed for the study of neutron production from spallation reactions related to the ADS project.The experiments of 400 MeV/u ^(16)O bombarded on a tungsten target were presented using a NTOF spectrometer.Neutron yields for 250 MeV protons incident on a thick grain-made tungsten target and a thick solid lead target have been measured using the water-bath neutron activation method. Spallation residual productions were studied by bombarding W and Pb targets with a 250 MeV proton beam using the neutron activation method. Benchmarking of evaluated nuclear data libraries was performed for D-T neutrons on ADS relevant materials by using the benchmark experimental facility at the China Institute of Atomic Energy.

Preparation and verification of mixed high-energy neutron crosssection library for ADS

An accelerator-driven subcritical system(ADS)is driven by an external spallation neutron source, which is generated from a heavy metal spallation target to maintain stable operation of the subcritical core, where the energy of the spallation neutrons can reach several hundred megaelectron volts. However, the upper neutron energy limit of nuclear cross-section databases, which are widely used in critical reactor physics calculations, is generally 20 MeV.This is not suitable for simulating the transport of highenergy spallation neutrons in the ADS. We combine the Japanese JENDL-4.0/HE high-energy evaluation database and the ADS-HE and ADS 2.0 libraries from the International Atomic Energy Agency and process all the data files for nuclides with energies greater than 20 MeV. We use the continuous pointwise cross-section program NJOY2016 to generate the ACE-formatted cross-section data library IMPC-ADS at multiple temperature points. Using the IMPC-ADS library, we calculate 10 critical benchmarks of the International Criticality Safety Benchmark Evaluation Project manual, the 14-MeV fixed-source problem of the Godiva sphere, and the neutron flux of the ADS subcritical core by MCNPX. To verify the correctness of the IMPCADS, the results were compared with those calculated using the ENDF/B-VII.0 library. The results showed thatthe IMPC-ADS is reliable in effective multiplication factor and neutron flux calculations, and it can be applied to physical analysis of the ADS subcritical reactor core.

Development and validation of the code COUPLE3.0 for the coupled analysis of neutron transport and burnup in ADS

The analysis of the fuel depletion behavior is critical for maintaining the safety of accelerator-driven subcritical systems(ADSs). The code COUPLE2.0 coupling 3-D neutron transport and point burnup calculation was developed by Tsinghua University. A Monte Carlo method is used for the neutron transport analysis, and the burnup calculation is based on a deterministic method. The code can be used for the analysis of targets coupled with a reactor in ADSs. In response to additional ADS analysis requirements at the Institute of Modern Physics at the Chinese Academy of Sciences, the COUPLE3.0 version was developed to include the new functions of(1) a module for the calculation of proton irradiation for the analysis of cumulative behavior using the residual radionuclide operating history,(2) a fixed-flux radiation module for hazard assessment and analysis of the burnable poison, and(3) a module for multi-kernel parallel calculation, which improves the radionuclide replacement for the burnup analysis to balance the precision level and computational efficiency of the program. This paper introduces thevalidation of the COUPLE3.0 code using a fast reactor benchmark and ADS benchmark calculations. Moreover,the proton irradiation module was verified by a comparison with the analytic method of calculating the210 Po accumulation results. The results demonstrate that COUPLE3.0 is suitable for the analysis of neutron transport and the burnup of nuclides for ADSs.

Application of FLUKA and OpenMC in coupled physics calculation of target and subcritical reactor for ADS

The study of accelerator-driven subcritical reactor systems(ADSs) has been an important research topic in the field of nuclear energy for years. The main code applied in ADS research is MCNPX, which was developed by Los Alamos National Laboratory. We studied the application of the open-source Monte Carlo codes FLUKA and OpenMC to a coupled ADS calculation. The FLUKA code was used to simulate the reaction of highenergy protons with the nucleus of the target material in the ADS, which produces spallation neutrons. Information on the spallation neutrons, such as their energy, position,direction, and weight, can be recorded by a user-defined routine called FLUSCW provided by FLUKA. Then, the information was stored in an external neutron source file in HDF5 format by using a conversion code, as required by the OpenMC calculation. Finally, the fixed-source calculation function of OpenMC was applied to simulate the transport of spallation neutrons and obtain the distribution of the neutron flux in the core region. In the coupled calculation, the high-energy cross-section library JENDL4.0/HE in ACE format produced by NJOY2016 was applied in the OpenMC transport simulation. The OECD–ADS benchmark problem was calculated, and the results were compared with those obtained using MCNPX. It was found that the flux calculations performed by FLUKA–OpenMC and MCNPX were in agreement, so the coupling calculation method for ADS is reasonable and feasible.

Flow characteristics of natural circulation in a lead–bismuth eutectic loop

Lead and lead-alloys are proposed in future advanced nuclear system as coolant and spallation target.To test the natural circulation and gas-lift and obtain thermal-hydraulics data for computational fluid dynamics(CFD) and system code validation, a lead–bismuth eutectic rectangular loop, the KYLIN-Ⅱ Thermal Hydraulic natural circulation test loop, has been designed and constructed by the FDS team. In this paper, theoretical analysis on natural circulation thermal-hydraulic performance is described and the steady-state natural circulation experiment is performed. The results indicated that the natural circulation capability depends on the loop resistance and the temperature and center height differences between the hot and cold legs. The theoretical analysis results agree well with,while the CFD deviate from, the experimental results.

Neutronic analysis of deuteron-driven spallation target

Deuteron-driven spallation targets have garnered attention recently because they can provide high-energy neutrons to transmute long-lifetime fission products.In this study,the Geant4 toolkit was used to simulate the inter-action between a deuteron beam at 500 MeV and a com-posite target composed of alternating lead-bismuth eutectic(LBE)and water.The water was used because it may be employed as a target coolant.The energy spectrum,neu-tron yield,average energy,and total energy of the emitted neutrons were calculated for different thicknesses and thickness ratios between the LBE and water.For a constant target thickness,the neutron yield increases with an increasing thickness ratio of LBE to H 2 O,while the aver-age energy of the emitted neutrons decreases with an increasing in the aforementioned thickness ratio.These two aspects support the use of a pure target,either LBE or water.However,with an increasing LBE-to-H 2 O thickness ratio,the total energy of the emitted neutrons increases and then decreases.This result supports the addition of water into the LBE target.The angular distributions of the emitted neutrons show that the rear of the target is suit-able for loading nuclear waste containing minor actinides and long-lifetime fission products.

Analytical solution of modified point kinetics equations for linear reactivity variation in subcritical nuclear reactors adopting an incomplete gamma function approximation

The present work aims to achieve a fast and accurate analytical solution of the point kinetics equations applied to subcritical reactors such as ADS (Accelerator-Driven System), assuming a linear reactivity and external source variation. It was used a new set of point kinetics equations for subcritical systems based on the model proposed by Gandini & Salvatores. In this work it was employed the integrating factor method. The analytical solution for the case of interest was obtained by using only an approximation which consists of disregarding the term of the second derivative for neutron density in relation to time when compared with the other terms of the equation. And also, it is proposed an approximation for the upper incomplete gamma function found in the solution in order to make the computational processing faster. In addition, for purposes of validation and comparison a numerical solution was obtained by the finite differences method. Finally, it can be concluded that the obtained solution is accurate and has fast numerical processing time, especially when compared with the results of numerical solution by finite difference. One can also observe that the gamma approximation used achieve a high accuracy for the usual parameters. Thus we got satisfactory results when the solution is applied to practical situations, such as a reactor startup.

Microstructure and Mechanical Properties of Simulated Heat-affected Zones of EP-823 Steel for ADS/LFR

EP-823 steel is one of the candidate materials for accelerator-driven systems/lead-cooled fast reactors （ADS/LFR）. Its weldability was investigated by mechanical property tests and microstructure analysis on the enlarged heat-affected zones （HAZs） made by numerical and physical simulation. The finite element numerical simulation could simulate the welding thermal cycle of the characteristic regions in HAZs with extremely high accuracy, The physical simulation performed on a Gleeble simulator could enlarge the characteristic regions to easily investigate the relationship between the microstructure evolution and the mechanical properties of the HAZs. The results showed that the simulated partially normalized zone comprising tempered martensite, newly formed martensite and more tiny carbides has the highest impact energy. The fully normalized zone exhibits the highest hardness because of the quenched martensite and large carbides. The ductile property of the overheated zone is poor for the residual delta- ferrite phases and the quenched martensite.

Design of machine protection system for CAFe facility

Background In an accelerator-driven sub-critical system(ADS),a strong proton beam from a high-power proton accelerator irradiates a spallation target to produce very intense neutrons for driving a reactor.Currently,the China ADS front-end demo linac(CAFe)facility is being built to provide strong beam with the beam intensity of several mAs.Purpose This paper reports the machine protection system(MPS)of the CAFe facility,which prevents or minimizes the beam-induced damages or the damages from other sources to the machines in the accelerator.Method The MPS of CAFe facility works under the framework of experimental physics and industrial control system(EPICS).The developed machine protection system includes fast protection function and slow interlock function,where the fast function is performed by FPGA controller and the slow function is performed by PLC controller.Results Based on the reports on the status of all relevant subsystems,MPS will either perform an emergency shutdown of the beam or machine or will permit normal operation.When an emergency shutdown of the beam is needed,MPS switches off the ion source or switches on the chopper.In addition to its protective role,MPS will also trace the origin of an emergency shutdown.Conclusions The developed MPS has played an important role during the commissioning and successful operation of the CAFe facility at the end of 2019.