Microstructural characterization and mechanical testing were performed on 400℃ thermally aged Z3CN20.09M cast austenitic stainless steels. A quantitative mathematical model is introduced for the relationship between aging defects and mechanical properties. Charpy impact energy was used as a characteristic mechanical parameter, and the short-range order of the Cr element segregation that induced by spinodal decomposition was chosen as the characteristic defect parameters. The macroscopic mechanical properties and the microscopic thermal aging defects was quantitatively correlated in this model, which has a great reference significance in safety assessment and lifetime prediction of the reactor structural materials.

Fretting wear behavior of the 690TT tube under three fretting regimes in high temperature pressurized (HTP, 288 °C and 8.4 MPa) water was studied based on fretting maps. The running condition fretting map was established in the ranges of 10 ∼ 100 N and 10 ∼ 100 μm. Fretting regime in this map can be divided into gross slip regime (GSR), mixed stick-slip regime (MSSR) and stick regime (SR). The GSR and MSSR ranges increase in HTP water compared to room temperature water, while the SR range decreases. According to the material response fretting map, the severe material degradation of 690TT tube under GSR is mainly caused by substrate delamination, whereas under MSSR can be attributed to the rapid corrosion. For SR, adhesive wear cannot cause material loss to 690TT tube.

With the development of high-temperature gas-cooled reactors, the coefficient of thermal expansion (CTE) of nuclear-grade graphite plays an increasingly important role in reactor design. A lower CTE enhances both the integrity of the graphite core structure and reactor efficiency. In this paper, we present a new isotropic graphite grade with a low CTE, utilizing spheroidized natural flake graphite (SFG) as a filler material. The ultralow isotropic CTE of 2.6-2.9×10−6 K−1 in the SFG-based graphite, owing to the ability of the slit-shaped pores within the SFG particles to accommodate cross-plane thermal expansion. To enhance the baking performance of the SFG-based graphite, hybrid fillers of SFG/coke or SFG/microcrystalline graphite (MG) were used to prevent cracking of the green bodies. In particular, the addition of MG prevents cracking without changing the low CTE value of the SFG-based graphite. This research contributes to the development of new graphite materials with low CTE that can be used in nuclear engineering, the semiconductor industry, and other high-temperature environments.

To address the challenges of increased beam power and target survivability associated with next-generation particle production beam lines, high dose, high-energy proton beam conditions are simulated using irradiation from low-energy ion beams. A low-energy ion irradiation study of POCO ZXF-5Q graphite under conditions similar to those of the NuMI NT-02 neutrino production target at the Fermi National Accelerator Laboratory is reported. Helium ion irradiation was performed at 100∘ C to a maximum damage level of 0.9 displacements per atom (DPA). Irradiation induced hardening, swelling of the irradiated region, inter-plane lattice expansion, and intraplane lattice contraction with increasing ion fluence was observed using micromechanical (nanoindentation, atomic force microscopy) and electron microscopy (high-resolution imaging, selected area diffraction) characterization. Similar changes were also observed in post-irradiation examination of the NT-02 target indicating that ion irradiation can be a valuable tool for estimating radiation damage in proton beam targets. Caution must be exercised though, because the hardening, lattice alteration, and swelling occur to different magnitudes for a given damage level. The observed hardening and embrittlement were greater for ion irradiated graphite. For He ion irradiated samples the lattice spacing changes were smaller at low damage levels (78% less expansion and 71% less contraction at 0.1 DPA) and larger at high damage levels (38% more expansion and 5% more contraction at 0.9 DPA) relative to that observed in the NT-02 target. The magnitude of swelling was 8.5× greater under ion irradiation which is influenced by the differing damage gradients and inclusion of implanted He ions in the region of interest.

The availability of irradiated material for the fracture-mechanical characterization of reactor pressure vessel steels in the context of surveillance programs is severely limited. Additional efforts are necessary to reduce the amount of material required for the determination of the reference temperature based on the Master Curve methodology. The objective of this study is the further development of a method for identifying the parameters of a cohesive zone model to simulate the fracture-mechanical behavior within the ductile-to-brittle transition region. A novel approach is proposed where statistically distributed numerical fracture toughness values are obtained by means of random spatial distributions of cohesive elements with either brittle or ductile fracture properties throughout the cohesive zone. Thereby, a numerical reference temperature is determined based on the ASTM E1921 standard and compared to the reference temperature obtained from tests on miniaturized CT specimens. It is shown that with the presented approach the reference temperature of a reactor pressure vessel steel can be predicted with high accuracy. Less material is required for the calibration of the model parameters than for an experimental determination of the reference temperature. Further development of the model is required to accurately predict the experimentally observed fracture toughness scatter within the transition region.

Irradiation can cause prominent damage to reactor concrete aggregates leading to amorphization, strength and modulus decrease, radiation induced volume expansion (RIVE) and micro-cracking, which limits their long-term performance. To develop an improved understanding of irradiation effects in concrete, three mineral analogues of concrete aggregates (limestone, marble and quartzite) were irradiated by 5.5 MeV He ions and 13 MeV Ni ions to surface doses of 0.011 displacements per atom (dpa) and 0.23 dpa, respectively, at room temperature. The two different ion species allow irradiation spectrum effects (ionizing and displacive) to be examined. Irradiation induced cracks were observed in He irradiated limestone and marble, and Ni irradiated quartzite. Full amorphization was observed in Ni irradiated quartzite with 14.3 % RIVE, and ∼25 % hardness and modulus decrease, while almost no change was observed in He irradiated quartzite except 4.35 % RIVE, revealing a possible ionization enhanced diffusion effect for high energy light ions. Furthermore, partial amorphization was observed in Ni irradiated marble and limestone matrix with a 12 % hardness decrease in marble while no amorphization was observed for He irradiation with a 20 % hardness increase in limestone matrix. The role of knock-on damage and irradiation spectrum on amorphization, volumetric expansion and mechanical property changes are discussed. Moreover, the onset and critical doses for amorphization and RIVE in quartz are obtained for ion irradiations at room temperature. The dose dependence of RIVE exhibits a delay compared to the amorphization behavior. The superior irradiation resistance of calcite phase compared to quartz phase implies there could be advantages to using calcareous aggregates and lowering the usage of siliceous aggregates for concrete in nuclear power plants for extended operation beyond 60 years. However, other effects such as corrosion, aging and reactions during severe accidents should also be considered, and further investigations are needed.

The ternary machine-learning interatomic potential for a Zr–C–Ag system was developed using first-principles calculations, moment tensor potential, and molecular dynamics simulations to calculate the diffusion coefficient of Ag in ZrC. The developed potential was utilized to investigate the vacancy formation energy, Ag substitutional energy, binding energy between Ag and vacancy, Ag interstitial energy, migration energy of Ag to an adjacent C vacancy in ZrC, and thermal expansion of ZrC. The results conformed to previously reported experimental and computational results, thereby validating the accuracy of the developed potential. The diffusion coefficients of Ag in ZrC0.94 and ZrC0.97 in the temperature range of 2800–3200 K were calculated using molecular dynamics simulations with the developed machine-learning interatomic potential. These calculation results can aid the safety analysis of radioactive 110mAg release in nuclear thermal propulsion reactors.