Neutron Damage Research Articles

Comprehensive Study on Microstructure and Mechanical Properties of Si Ion Irradiated Al 6063

Aluminum alloys are widely used as fuel cladding materials for research reactors. Neutron irradiation damage is a major concern in the nuclear industry, and it is of great interest to study irradiation damage by ions, which can cause similar damage to materials. In this study, Al 6063 alloy was irradiated with Si ions at 333 K up to 90 dpa to mimic the effects of both elemental Si, a transmutation product of thermal neutrons, and Frenkel pair formation, an effect of fast neutrons. The irradiation damage was investigated by transmission electron microscopy and the nanoindentation technique. The irradiated sample's net dislocation density was about 1.2·1014 m−2 in the vicinity of the voids and 2·1014 m−2 in the vicinity of the Si peak. The implanted Si peak concentration was ~ 6 at% at a depth of ~ 1100 nm. The formation of voids with an average size of 8 nm, and a peak number density of ~1022 m−3 was identified, and the void annihilation time dependence was found to be proportional to (1t)0.25.The amorphous to crystalline structure transformation temperature of the intermetallic Al8Fe2Si occurs at 723 K and follows the Johnson-Mell-Avrami diffusion control model. The hardness and estimated ultimate tensile strength of the irradiated surface layer, obtained by the nanoindentation measurements, increased by ~65% in comparison to the unirradiated sample. A similar increase in yield stress and ultimate tensile strength published values of neutron irradiation damage, occurs at much higher dpa (~260 dpa). This difference is suggested to be related to the high dpa rate and higher at%Si/dpa ratio in the present ion irradiation experiments.

Neutronic analyses of Upper port ECRH antenna system for CFETR

• 3D full neutronic model of CFETR ECRH was built to deliver the nuclear analysis. • The maximum neutron damage on ECRH antenna appear at the steering mirrors. • Irradiation of VV and TFC can meet the requirements after shielding optimization. • TBR of the whole reactor reduced to 1.137 after 3 ECRH antenna introduced. Electron Cyclotron Resonance Heating (ECRH) is one of the auxiliary heating systems designed for CFETR. The antenna of ECRH system is planned to be plugged into CFETR through 3 upper ports and 2 blanket modules of each port will be penetrated for the transmission of wave. In this study, three dimensional neutronic analysis has been performed to evaluate the shielding effect and conduct the optimization of ECRH antenna design. The simulation has been done on the integration of ECRH antenna system with 22.5-degree CFETR model. The particle transport has been done by MCNP with the support of FENDL-2.1 nuclear data library. An accurate neutron flux distribution has been obtained by employing the ‘on-the-fly’ global variance reduction technique. The neutron damage of TFC, VV and port was analysed to assess the shielding performance of ECRH antenna system. Initial results demonstrated that the fast neutron fluence and nuclear heating density of TFC and helium production rate of VV port were beyond the design limit. By adding shielding blocks, these parameters didn’t exceed the limits. The irradiation damage on antenna components were calculated for estimating its life after neutron irradiation and providing the support of cooling system design. The TBR decreased due to the blanket opening from original 1.149 to 1.137 in the whole reactor.

Characterising Ion-Irradiated FeCr: Hardness, Thermal Diffusivity and Lattice Strain

Ion-irradiated FeCr alloys are useful for understanding and predicting neutron damage in the structural steels of future nuclear reactors. Previous studies have largely focused on the structure of irradiation-induced defects, probed by transmission electron microscopy (TEM), as well as changes in mechanical properties. Across these studies, a wide range of irradiation conditions has been employed on samples with different processing histories, which complicates the analysis of the relationship between defect structures and material properties. Furthermore, key properties, such as irradiation-induced changes in thermal transport and lattice strain, are little explored.Here we present a systematic study of Fe3Cr, Fe5Cr and Fe10Cr binary alloys implanted with 20 MeV Fe3+ ions to nominal doses of 0.01 dpa and 0.1 dpa at room temperature. Nanoindentation, transient grating spectroscopy (TGS) and X-ray micro-beam Laue diffraction were used to study the changes in hardness, thermal diffusivity and strain in the material as a function of damage and Cr content. Our results suggest that Cr leads to an increased retention of irradiation-induced defects, causing substantial changes in hardness and lattice strain. However, thermal diffusivity varies little with increasing damage and instead degrades significantly with increasing Cr content in the material. We find significant lattice strains even in samples exposed to a nominal displacement damage of 0.01 dpa. The defect density predicted from the lattice strain measurements is significantly higher than that observed in previous TEM studies, suggesting that TEM may not fully capture the irradiation-induced defect population.

Direct observation of dislocation loops shrinkage upon annealing neutron-irradiated Fe–9Cr alloy

Post-irradiation annealing is an effective way to recover mechanical properties by removing neutron damage. Using in-situ transmission electron microscopy, a neutron irradiated Fe–9Cr alloy was annealed under specific time-temperature scheme. In general, dislocation loops density decreased and their mean size increased upon increasing annealing temperature and time. Concurrent shrinkage phenomenon was observed for both ½<111> and <100> type loops and is found to be vacancy diffusion driven. Essentially, loops’ neighboring microstructure dictates their shrinkage behavior. Shrinkage rate is both loop size and type dependent. Loop size defined vacancy concentration gradient and loop type associated elemental segregation are main deciding factors.

Likely U.S. Regulatory Considerations for D-T Fusion Power Reactors

The Electric Power Research Institute outlined three criteria important for a commercially viable fusion power plant: competitive electric power cost, regulatory simplicity, and public acceptance. In this paper we consider likely U.S. regulatory considerations for deuterium-tritium (D-T) fusion power reactors, relying on existing criteria and past actions by the U.S. Nuclear Regulatory Commission, which has asserted regulatory jurisdiction over U.S. commercial fusion reactors. We begin with consideration of a basic D-T fusion reactor, independent of plasma confinement approach. Because tritium and radioactivity are present, likely regulation will require containment structures and various safety measures for each component. Regulators are certain to require that all nuclear components of the system be housed in an overall containment vessel that must be held at less than atmospheric pressure to contain any leakage of tritium, radioactive corrosion products, radioactive coolant, and activated elements in the air. In addition, regulators are sure to require plant structure and operations that minimize the potential for clandestine plutonium breeding. Next, we add superconducting magnets and a plasma dump (divertor) to the basic system and recognize the small but nonzero probability of those magnets explosively quenching, potentially causing reactor damage and dramatically increasing containment vessel pressure. Finally, we consider ITER as prototypical of a D-T–fueled fusion power reactor. Because ITER-like systems are subject to damaging plasma disruptions, regulators are almost certain to require safeguards against such events significantly damaging first walls and subsystems. Finally, we believe that regulators are not likely to back off significantly in requirements related to the deuterium-deuterium and D3He fuel cycles even though the tritium production and the neutron damage in the latter fuel cycle are significantly below those in a D-T system. However, regulations for p11B and 3He3He fuel cycles are certain to be dramatically less demanding because of the lack of tritium and essentially no neutron production.

ARC reactor – Neutron irradiation analysis

Abstract Neutron irradiation is one of the most concerning issues to design plasma facing components and reactor inner structures in fusion devices, especially for high power density ones, like ARC reactor. This study addresses the main aspects of neutron irradiation on solid materials of ARC reactor. In particular it deeply analizes the effect of neutron induced activation proposing low activation structures, like vanadium alloys and different optimization methods like isotopic tailoring, detritiation and impurity control. Furthermore, irradiation damage issues and their dependence on the energy spectrum are highlighted. It resulted that V-Cr-Ti alloys dramatically reduce the radioactive inventory of ARC with respect to its baseline configuration, which proposes the application of Inconel 718. Such alloy is also optimizable through the tailoring of titanium isotopes and is virtually capable of hitting recycle limits in a couple of decades. Lastly, it shows a relatively growth of gas during irradiation. However, it is highlighted how experiments on neutron damage featuring fission neutrons risk to be able to tell very little about the behavior of the same materials under fusion neutrons, as damaging mechanisms seem to be different.

Attenuation of neutron and photon-induced irradiation damage in pressurized water reactor pressure vessels

The present work investigates the attenuation of neutron and photon-induced irradiation damage in a Pressurized Water Reactor (PWR)-like mock-up Reactor Pressure Vessel (RPV) using Tripoli-4® Monte Carlo simulations with a particular emphasis on the presence of a thick stainless steel heavy reflector between the core and RPV. Results show that the photon-induced damage is well described by the exponential law. The neutron-induced damage attenuates quicker than the exponential form near the outer surface of RPV. Nevertheless, an exponential form can represent the attenuation of neutron-induced damage within 5% discrepancy for penetration < 17 cm in a typical 22 cm thick RPV and within 20% in the outermost 3 cm. The exponential form with an additional negative term fits the attenuation of neutron-induced damage in RPV as the negative term considers the “leakage” of neutron near the outer surface. It is observed that the presence of a Gen III-like representative heavy reflector reduces the estimated neutron-induced damage by a factor of 2. On the other hand, this paper verifies that the neutron flux with energies above 0.5 MeV (ϕ>0.5) is more representative of the displacement damage attenuation than the flux above 0, 0.1, and 1 MeV in RPVs. The damage rate is approximatively equal to Kϕ>0.5 with K=9.5×10-22DPA/n∙cm-2 in PWR RPVs.

Neutronics related integration studies of EU-DEMO pellet injection system

• The problem of the integration of the pellet injection line system from neutronic and thermal point of view is presented. • The evolution of the design of the pellet injection system from inboard side of the machine is presented. • A neutronic preliminary assessment of an alternative solution for pellet injection system from the upper port of the machine has been studied. • All analyses have been done using MCNP-5 Monte Carlo code and ADVANTG hybrid has been used to implement specific variance reduction techniques. In the frame of the EU DEMO development program within the EUROfusion Consortium, the integration of the in-vessel components is crucial even at an early stage of the design process. The auxiliary, heating and fueling systems have to be integrated into the Breeding Blanket and, thus, they will undergo to a harsh nuclear environment during operation, and have a significant impact on the Tritium Breeding capability, loads and shielding performances. This work presents the neutronics and thermal analyses performed to support the integration studies of the pellet injection systems. This study is mainly devoted to optimize the design of a fueling line option consisting in a protrusion of the Vacuum Vessel (VV) in the inboard side supporting a guiding tube to drive the pellets along the ideal trajectory as close as possible to the First Wall. An alternative concept with a free-flight injector through the upper port, has also been studied evaluating the feasibility and the shielding needs. The three-dimensional neutronics simulations were carried-out with the MCNP5 Monte-Carlo code using a DEMO model with the integrated systems to calculate neutron fluxes, damage and nuclear heating distribution. Thermal analyses were carried-out using ABAQUS. Results of neutronics and thermal analyses on the fueling systems are presented and discussed.

Nuclear analyses for the design of the ITER-like plasma facing components vertical targets of the DEMO divertor

Abstract The Divertor Plasma Facing Components (PFCs) of a fusion reactor are the most loaded components in terms of high heat fluxes, which, combined with high neutron irradiation, can severely compromise their thermo-mechanical and physical properties as well as their heat removal capacity. Therefore, neutronic assessment plays a key role in the design of these critical components. The aim of this work is to perform a dedicated nuclear analysis for the European DEMO divertor PFCs placed on the vertical targets, aimed to provide significant outcomes in the PFCs selection concept. In particular, the present assessment is devoted to the reference ITER-like configuration under study within the EUROfusion WPDIV-PPPT programme. Three-dimensional neutronics analyses have been performed with the MCNP5 Monte Carlo code. This work presents detailed neutronics results with heterogeneous materials constitution and actual geometry of the PFC concept. High resolution data on the nuclear heating density and neutron damage of the ITER-like PFCs placed on the divertor vertical targets, including helium production, assessed for first time for the latest DEMO design, are presented and discussed.

Evaluation of the effects of neutron irradiation on first-generation corrosion mitigation coatings on SiC for accident-tolerant fuel cladding

High purity SiC and SiC/SiC composites coated with commercial TiN, Cr, CrN, or CrN/Cr multilayer coatings were irradiated in Ar or flowing PWR water in the Massachusetts Institute of Technology Nuclear Reactor Laboratory (MITR). Irradiation in Ar was performed in the core. In the water environment, identical samples were placed in one of three different locations: in-core, providing exposure to neutron damage and radiolysis-affected water; above-core, where samples were exposed to radiolysis-affected water but not neutron damage, or outside of the core, where samples were exposed to the coolant water without the effects of radiation. Radiation in Ar revealed significant cracking of all but the TiN coatings, attributed to differential swelling between the coating and substrate. Lattice swelling was not observed in any of the coatings, but 0.2% void swelling was observed in the Cr coating. All of the coatings failed during water exposures in the core. CrN/Cr spalled in each condition. Cr was protective, except under radiation damage as a result of cracking, and TiN severely degraded in the core with no coating was found following exposure. A SiC/coating ATF cladding system is expected to perform adequately following improvements in coating ductility and purity.

Fusion Materials Development at Forschungszentrum Jülich

Material issues pose a significant challenge for the design of future fusion reactors. From a historic point of view, the material mix used for the first wall of a fusion reactor has continually evolved, from original steel vessels to carbon and other low‐Z materials such as beryllium to tungsten as the primary candidate for a reactor's first wall armor and divertor material. For materials considered for fusion applications, a highly integrated approach is necessary. Resilience against neutron damage, good power exhaust, and oxidation resistance during accidental air ingress are design relevant issues while deciding on new materials or improving upon baseline materials. Neutron‐induced effects, e.g., transmutation adding to embrittlement, retention, and changes to thermomechanical properties, are crucial to material performance. In this contribution, the recent progress (2013–2019) in fusion materials development for current and future fusion devices, at Forschungszentrum Jülich GmbH, with activities focussing on advanced materials and their characterization is given. It is a continuation and extension of the work given by Coenen et al.

Microstructural examination of neutron, proton and self-ion irradiation damage in a model Fe9Cr alloy

Transmission electron microscopy (TEM) was used to compare the microstructural defects produced in an Fe9Cr model alloy during exposure to neutrons, protons, or self-ions. Samples from the same model alloy were irradiated using fission-neutrons, 2 MeV Fe+ ions or 1.2 MeV protons at similar temperatures (∼300 °C) and similar doses (∼2.0 dpa). The neutron-irradiated alloy contained visible interstitial dislocation loops with b = 111, and on average ∼5 nm in size. The density varied from 2±1 × 1020 m-3 (in the matrix far from dislocations and boundaries) to 1.2±0.3 ×1023 m-3 (close to helical dislocation lines). Chromium α′-phase precipitates were also identified at a density of 7.4±0.4 ×1023 m-3. Self-ion irradiation produced mostly homogeneously distributed dislocation loops (6–7 nm on average), and with a greater fraction of 100 loops (∼40%) than was seen in the neutron-irradiated alloy, and at a density of 6.8±0.8 ×1022 m-3. In contrast to the loops produced by neutron irradiation, the self-ion irradiated Fe9Cr contained only vacancy-type loops. Chromium also remained in solution. Proton-irradiated Fe9Cr contained interstitial dislocation loops close to helical-dislocation segments, similar to the neutron-irradiated sample. Chromium α′-phases were also identified in the proton-irradiated sample at a density of 2.5±0.3 ×1023 m-3, and large voids (up to 7 nm) were found at a density over 1022m−3. Like the neutron-irradiated sample, the density of dislocation loops was also heterogeneously distributed; far from grain boundaries and dislocation lines the density was 2.5±0.4 ×1022 m-3, while close to helical dislocation lines the density was 8.1±1.3 ×1022 m-3.

Modeling of neutron radiation-induced defects in silicon particle detectors

Radiation damage of the silicon detectors in future hadron colliders poses a major challenge for its reliable operation. It is crucial to investigate the neutron-induced radiation damage owing to its large flux in the calorimeters and the trackers. Measurement results on irradiated detectors are complemented by modeling for getting a deep insight into the device behavior. This work presents the development of neutron-induced radiation damage model in silicon detectors using TCAD simulation. A wide spectrum of measurement results available on neutron-irradiated silicon detectors, viz. full depletion voltage, leakage current, charge collection, and effective trapping times, helps in constraining the model parameters, resulting in a robust model for neutron damage. Modeling has been performed within the phase-space of the measurements, i.e. for different initial bulk resistivities, active thicknesses and for both polarities up to the fluence values of 9 × 1014 1 MeV neq cm−2 (where, 1 MeV neq refers to the equivalent fluence for monoenergetic neutrons of energy 1 MeV) and for two measurement temperatures, 253 and 263 K. The results obtained from the devised neutron damage model show a good agreement with the measurement results.

Newly developed BGO glasses: Synthesis, optical and nuclear radiation shielding properties

In this research paper, we studied the optical and nuclear shielding efficiency of newly developed BGO glasses with the following compositions (in wt%): 32Bi2O3–68GeO2, 42Bi2O3–58GeO2, 47Bi2O3–53GeO2, 52Bi2O–48GeO2, 62Bi2O3–38GeO2. BGO glasses were prepared by traditional melt quenching method. To obtain the band gap values of fabricated BGO glasses, optical absorption spectra were used for evaluation of optical properties. The mass attenuation coefficients (μ/ρ) were achieved for prepared glasses at 0.015–15 MeV photon energies employing MCNPX Monte Carlo code and WinXcom program. Moreover, broad-range of nuclear shielding parameters for gamma ray, neutrons and charged particles such as mass attenuation coefficient, half value layer, effective atomic number, buildup factors, mass stopping powers, projected ranges, fast neutron removal cross sections and damage factors were calculated. The refractive index is calculated from Eopt, As Bi2O3 concentration is enhanced, Eopt is also increased as well as the optical electronegativity and consequently the refractive index. In addition, the results showed that BIGE5 glass sample with highest Bi2O3 contribution has excellent nuclear radiation shielding ability among the other fabricated glass samples.

Conceptual Design and Analysis of In-Vessel Components for the Materials Plasma Exposure eXperiment (MPEX)

The materials plasma exposure experiment (MPEX) is a linear plasma divertor simulator currently undergoing conceptual design. The facility will expose material samples to steady-state plasma fluxes to examine plasma–material interactions (PMIs) that are expected in the next generation of fusion devices. The plasmas will be generated by a helicon source, with electron and ion heating sources of up to 800 kW possible. The peak heat fluxes of the target are expected to be up to 10 MW/m2. The facility will be capable of handling low-activation neutron-irradiated samples in order to examine the multivariate effects of neutron damage and plasma fluence. Neutron-irradiated samples are planned to be roughly of 10-mm diameter; however, plasma-facing components up to $60 \times 600$ mm can be accommodated. The steady-state nature of the device will require the magnetic confinement of the plasma to be achieved with superconducting magnets, with a maximum on-axis field of 2.5 T. In addition, since MPEX will be a steady-state device, in-vessel components need to be water cooled. The primary in-vessel components will be the target, the dump plate, the limiter, the skimmers, and the microwave absorber. The conceptual design of these components is presented here, including analyses that confirm that the designs are adequate to meet the requirements of MPEX operation.

Electron beam welding of oxide dispersion strengthened 9 Cr martensitic steel-an experimental and theoretical perspective

Abstract Advanced nuclear energy systems present significant improvements in economics, safety, reliability and sustainability over currently operating reactor technologies. Proper choices of cladding and structural material with a. high neutron damage resistance b. good mechanical properties at high operating temperatures and c. low activation properties are essential for the safe and reliable operation of the system. The 9-12% Cr transformable steels with lower carbon (∼0.1%) contents and addition of Mo, W, V, Nb present better high temperature strength than austenitic steels. The use of 9%Cr steel is considered as an optimum solution taking into consideration the present day requirements like environment, efficiency, elasticity, experience and economy that dominate considerations of modern power station design, layout, fabrication and operation. The pursuit of improved efficiency/power plant performance resulted in the design of system operating at higher temperature and pressure, which requires materials with appropriate high-temperature mechanical properties and metallurgical stability at operating temperatures. Oxide Dispersion Strengthened (ODS) 9% Cr steels have the potential to advance the high temperature performance of associated components in advanced nuclear energy systems with marginal oxidation resistance. The research focuses on Electron beam welding (EBW) of Hot Isostatically Pressed (HIP) ODS 9 Cr martensitic steel in as-received condition to arrive at the optimized weld process parameters. The weldments were characterized using Optical Microscopy for depth of penetration and width of weldments. The experimental results were theoretically verified using Minitab software

Neutron irradiation-induced microstructure damage in ultra-high temperature ceramic TiC

Titanium carbide (TiC) is an ultra-high temperature ceramic with potential as a structural material candidate for advanced reactor concepts. However, the irradiation tolerance of TiC is not well understood. Here, we reveal the key irradiation damage microstructure degradation processes in TiC using mixed spectrum neutron irradiations at dose of ∼2 displacements per atom (dpa) at temperatures of ∼220, 620, and 1115 °C, combined with state-of-art microstructure characterization using transmission electron microscopy (TEM) and high-resolution TEM (HRTEM). The helium injection (∼65 atomic parts per million) produced by neutron transmutation also occurred in the sample. TiC was observed to form irradiation-induced interstitial-type dislocation loops and He-stabilized cavities. At 220 and 620 °C, the analysis of the electron diffraction patterns, rel-rod imaging and HRTEM revealed that the dislocation loops were faulted Frank loops with Burgers vector bFrank = 1/3<111> lying on {111} planes. A detailed Burgers vector identification performed by the g.b technique revealed that the dislocation loops forming at 1115 °C were unfaulted, edge-type, with Burgers vectors b = a<100> and a/2<110> with corresponding {100} and {110} habit planes. Using continuum mechanics, we estimated the critical radius at which a faulted dislocation loop transitions to a perfect loop to be 9 nm. Further, no amorphization occurred in TiC under-investigated irradiation conditions while macroscopic swelling under point defect swelling regime was observed. He-stabilized cavities were detected at 1115 °C, but not at lower temperature irradiation. These observations indicate the onset of long-range vacancy migration in TiC occurs between 620 and 1115 °C.

Structural evolution and wear resistance of MoS2–Based lubricant films irradiated by heavy ions

In this work, a 2 MeV Au2+ ion irradiation was implemented to simulate the neutron damage on magnetron sputtered MoS2-based composite films. The irradiation damage mechanism and the wear resistance of films were accessed as a function of increasing Au2+ fluence. The XRD as well as the cross-section TEM images results reveal that the intrinsic crystallites of MoS2 in Mo-S-Ti composite films are more easily to be destructed. At the low fluence of 1 × 1014 Au2+/cm2, the irradiation damage peak was mainly distributed in depth of 0.45 μm – 0.7 μm in films. While, with the increase of the incident fluence, the disordering of MoS2 crystallites near the film surface were also become serious, and they are fully amorphized at the end of the studied fluence of 3.3 × 1015 ion/cm2 due to the accumulation effect induced by the ion collisions. Nonetheless, the tribotests in vacuum show that the lubricity of MoS2 phases would not be immediately failured due to their damage of intrinsic structure of MoS2 crystallites, and only enhancing the irradiation to the high dose (2, 5 and 1 dpa), a rapid degradation of lubricity will be observed.

Characterization of the neutron flux during production of 18F at a medical cyclotron and evaluation of the incidental neutron spectrum for neutron damage studies

The fast neutron spectrum was measured in an irradiation position adjacent to two 18F (fluorine-18) production targets at a medical cyclotron. The neutron spectrum was modeled using MCNP6 and experimentally determined using activation foils. A three-group neutron spectrum was determined using an over-determined least squares fitting method. The MCNP6 model results and measurements of the activation foils were used as input to unfold the neutron spectrum using the STAYSL PNNL code. The neutron spectrum adjacent to the 18F production target was evaluated for use in incidental experiments to evaluate neutron-induced radiolysis of polyurethane and radiation hardness testing of electronics. The fast neutron flux near the 18F target between the energies of 0.4 – 16.5 MeV was 1.8 – 3.0x109 n cm-2s-1. The neutron dose rate in polyurethane at the irradiation position was 2.8 cGy s-1– 5.8 cGys-1. The 1 MeV equivalent neutron flux was 2.4x109 n cm-2s-1 – 4.0x109 n cm-2s-1.

Ionizing and non-ionizing kerma factors in silicon for China Spallation Neutron Source neutron spectrum

The quantification of ionizing energy deposition and non-ionizing energy deposition plays a critical role in precision neutron dosimetry and in the separation of the displacement damage effects and ionizing effects induced by neutron radiation on semiconductor devices. In this report, neutrons generated by the newly built China Spallation Neutron Source (CSNS) are simulated by Geant4 in semiconductor material silicon to calculate the ionizing and non-ionizing kerma factors. Furthermore, the integral method is applied to calculate neutron-induced ionizing at the CSNS and non-ionizing kerma factors according to the standard neutron nuclear database and the incident neutron spectrum. In addition, thermoluminescence dosimeters are utilized to measure the ionizing energy deposition and six series of bipolar junction transistors are used to measure the non-ionizing energy deposition based on their neutron damage constants. The calibrated kerma factors that were experimentally measured agreed well with the simulation and integral calculation results. This report describes a complete set of methods and fundamental data for the analysis of neutron-induced radiation effects at the CSNS on silicon-based semiconductor devices.