High-dose Neutron Irradiation Research Articles

Comparison of PM-HIP to forged SA508 pressure vessel steel under high-dose neutron irradiation

Powder metallurgy with hot isostatic pressing (PM-HIP) is an advanced manufacturing process that is envisioned to replace forging for heavy nuclear components, including the reactor pressure vessel (RPV). But PM-HIP products must at least demonstrate comparable irradiation tolerance than forgings in order to be qualified for nuclear applications. The objective of this study is to directly compare PM-HIP to forged SA508 Grade 3 Class 1 low-alloy RPV steel at two neutron irradiation conditions: ∼0.5–1.0 displacements per atom (dpa) at ∼270 °C and ∼370 °C. PM-HIP SA508 experiences greater irradiation hardening and embrittlement (total elongation) than forged SA508. However, uniform elongation and approximate toughness are comparable across all irradiated materials, suggesting irradiated PM-HIP SA508 exhibits superior ductility at maximum load-bearing capacity. The irradiation hardening mechanism is linked to composition rather than fabrication method. Since PM-HIP SA508 has higher Mn and Ni concentration, it is more susceptible to irradiation-induced nucleation of Mn-Ni-Si-P (MNSP) nanoprecipitates and dislocation loops, which both contribute to hardening. Conversely, the forged material nucleates fewer MNSPs, causing dislocation loops to control irradiation hardening. These results show promise for the irradiation performance of PM-HIP SA508 and can motivate future nuclear code qualification of PM-HIP fabrication for RPVs.

The irradiation effects on the nanoindentation hardness and helium bubbles evolution mechanism of Ni-based alloy

High-dose neutron irradiation could produce various kinds of microstructural defects, and the radiation resistance of nickel alloy is put forward for higher requirements under the high temperature. To simulate the (n, α) reaction, helium ions with an energy of 540 keV were implanted into nickel-based alloy 617 at 700 °C. The microstructure and mechanical properties were established by comparing the nanoindentation hardness before and after irradiation. For the low dose irradiation, the microstrain induced by vacancy-type defects like He-vacancy clusters and helium bubbles is slight, which would not result in a large increase of hardness. The hardening effect of high-temperature irradiation was much lower than that of room-temperature irradiation due to the recovery of matrix defects. The actual distribution of helium bubbles observed by Transmission Electron Microscope (TEM) was comparable to the SRIM simulation result. The helium bubbles were spherical near the surface of the alloy and polyhedral deeper in the matrix. Moreover, chromium carbide had a strong trapping effect on helium bubbles. The shapes and distribution of the helium bubbles were analyzed to evaluate irradiation effects on nickel-based alloys.

Optical transmission and dimensional stability of single-crystal sapphire after high-dose neutron irradiation at various temperatures up to 688 °C

The use of single-crystal sapphire optical fibers has been considered to extend fiber-optic sensing to the extreme temperature (>1000 °C) environments encountered in nuclear applications. However, before these sapphire fiber–based sensors can be deployed, their optical transmission and dimensional stability (which impacts drift of some sensors) must be characterized under representative testing conditions. Data regarding the optical transmission of sapphire following high-dose neutron irradiation at temperatures >100 °C is extremely limited. This work provides measurements of optical density (i.e., attenuation) and directional dimensional changes in bulk single-crystal sapphire materials irradiated to a fast neutron fluence of 2.4 × 10 21 n/cm 2 (3.5 displacements per atom) at temperatures ranging from 95 to 688 °C. Optical density measured after irradiation at 95 and 298 °C showed ultraviolet and visible absorption bands corresponding to known defect centers and temperature trends that were generally consistent with previous ex situ and in situ measurements made at much lower neutron fluence. However, optical density measured after irradiation at 688 °C was as much as two orders of magnitude higher, indicating that the fundamental mechanism for radiation-induced attenuation changes at this irradiation temperature. Additional analysis and comparison with previous works suggest that the attenuation may result from void formation, leading to increased Rayleigh scattering losses in the material and increased swelling that would also result in drift of Bragg grating-based sensors in sapphire fibers. These results pose serious questions regarding the feasibility of sapphire fiber–based sensors for high-temperature nuclear applications.

Mössbauer spectroscopy investigation of the effect of a high-dose neutron irradiation on the atomic redistribution in the industrial steel EP823

The authors studied radiation-induced structural-phase transformations that had occurred in EP823 steel of industrial application under conditions of its high-dose neutron irradiation in a fast fission reactor. The method of transmission Mössbauer spectroscopy with resonance detection and transmission electron microscopy was utilized. It has been observed that in ferritic-martensitic steel, under the neutron irradiation with doses above 50 dpa at temperatures of 570–660 °C, the decomposition of a BCC solid solution together with the release of the alloying elements chromium, molybdenum, and carbon from the metal matrix occurs. It is accompanied by the formation of an intermetallic χ phase and carbides of the Me23C6 type. The authors revealed the formation of the vacancy-clustered and supposedly gas-filled pores in the structure after irradiation.

Design and strategy for next-generation silicon carbide composites for nuclear energy

Silicon carbide (SiC) ceramic-based composites continue to be attractive material options for fusion in-vessel components and fission reactor core structures because of their exceptional high-temperature capability and favorable neutronic properties. As performance data accumulates, the limitations of the current generation of nuclear-grade SiC composites are becoming more apparent. These limitations mainly involve strength degradation during high-dose neutron irradiation. This paper discusses several options for improving the performance of the next generation of SiC composites to enhance the radiation resistance, along with new experimental results on neutron irradiation resistance. The main emphasis is on the selection of the fibers and the design and development of alternative interphase layers for advanced composites.

Investigation of a high-dose irradiated beryllium microstructure

The present paper studies the evolving Be microstructure after high-dose neutron irradiation at 660 K, 753 K and 873 K up to 34 dpa and generation of 5524 appm He and 596 appm 3H. At all three temperatures gas bubbles were observed whereby their average size is increasing with the temperature. In addition denuded zones (DZ) formed near grain boundaries. Their width increased with irradiation temperature and was found to be narrower for grains with a random orientation compared to those with a close to basal orientation with respect to the grain boundary. Furthermore, perceptions containing elements such as Al, Cr, Fe, Mg, Si as well as U were found inside single grains, often attached to dislocation lines, and in the vicinity of grain boundaries, mainly at 660 K and 753 K.

Research progress of hydrogen/helium effects in metal materials by positron annihilation spectroscopy

An important feature of the irradiation process in nuclear system is the formation of large displacement cascades, in which primary knock-on atoms and secondary particles formed by nuclear reactions generate a considerable number of defects such as dislocations, vacancies and transmutation gases. Predicting and mitigating the adverse effects of damage defect and transmutation hydrogen/helium produced by high-dose neutron irradiation on the mechanical properties of structural materials is the most significant challenge facing the current development of nuclear energy. To solve this problem, understanding the interaction mechanism between hydrogen/helium atoms and micro-defects is a very important breakthrough. Precursors of helium/ hydrogen bubble, small helium/hydrogen-filled vacancy complexes, may play an important role in realizing bubble nucleation, and the formation of these complexes is affected by many factors. However, only a little information about helium/hydrogen-vacancy clusters’ behavior has been obtained in metal/alloy materials. This is mainly limited by the characterization methods, such as the limited resolution of transmission electron microscope (TEM). Helium/hydrogen-vacancy clusters cannot be observed by TEM before the formation of helium bubbles. Applications of positron annihilation to the study of crystal lattice defects started around 1970s, when it was realized that positron annihilation is particularly sensitive to vacancy-type defects and that annihilation properties manifest the nature of each specific type of defect. In recent years, with the continuous development of slow positron beam and the improvement of various experimental testing methods based on slow positron beam, the application of positron annihilation technology has been extended to the research field of hydrogen/helium behavior in metal materials, which plays an important role in studying the hydrogen/helium radiation damage to metal materials. In this review, the basic principles of positron annihilation spectroscopy are briefly discussed and the three most important measurement methods used for hydrogen/helium effect studies are described (i.e. positron annihilation lifetime spectroscopy (PALS), Doppler broadening spectroscopy (DBS), coincidence Doppler broadening spectroscopy (CDBS)). In this paper, the application of positron annihilation spectroscopy to the study of hydrogen/helium behavior in metal materials is reviewed in combination with the reported relevant developments (including our research group’s achieve-ments). The advantages of three commonly used measurement methods in the following specific studies are highlighted: 1) The estimation of bubble size and concentration; 2) irradiation damage induced by hydrogen/helium; 3) the evolution behavior of irradiation-induced defects in the heat treatment process; 4) sy-nergistic effect of hydrogen and helium.

Progress in development of SiC-based joints resistant to neutron irradiation

This study fills a knowledge gap regarding neutron-irradiation resistance of SiC joints for nuclear applications, by investigating high-dose neutron irradiation effects on the strength of selected joints and low-dose neutron irradiation effects on recently developed joints fabricated by state of the art processing methods. The joining methods used for the high-dose radiation study included pressure-assisted liquid-phase sintering (LPS) of SiC nanopowder, pressureless calcia-alumina glass ceramics joining, and reaction sintering of Ti-Si-C powders with hot-pressing. The joints were neutron-irradiated at 530 °C to 20 displacements per atom (dpa). Other joining methods included low-pressure LPS of cold-pressed SiC green body, pressureless reaction sintered Ti-Si-C powder joint, spark plasma–sintered Ti diffusion bond, and hot-pressed Ti diffusion bond, which were irradiated at ∼500 °C to ∼2 dpa. There was no notable degradation of torsional strengths of the joints following the high-dose irradiation. The irradiation-induced degradation at low neutron dose was highly dependent on joint type.

A transmission electron microscopy study of EBR-II neutron-irradiated austenitic stainless steel 304 and nickel-base alloy X-750

The microstructure of EBR-II neutron-irradiated austenitic stainless steel 304 and nickel-base alloy X-750 was investigated. Both alloys were irradiated at low dose rates (∼2 × 10−8 dpa/s) to a neutron fluence of 6.9 × 1022 n/cm2 (E > 0.1 MeV) at 371–389 °C. Different types of defects, including Frank loops, cavities, and precipitates were characterized. The Frank loops in Type 304 stainless steel (SS) are larger in size (∼50 nm in diameter) and lower in number density (2.58 × 1021 m−3), compared to most previous higher dose rate neutron irradiation studies. The Frank loops in X-750 have an average size 26.0 nm of and a number density of 9.44 × 1021 m−3. In 304 SS and X-750, cavities are of ∼20 nm and ∼14 nm in diameter, respectively. The swelling of both alloys was found to be insignificant. In 304 SS, Ni and Si were found enriched at the cavity surfaces and Ni,Si-rich precipitates were also found. Multivariate statistical analysis using non-negative matrix factorization reveals that these Ni,Si-rich precipitates contain only ∼5.7 at.% Si, differing from the Ni3Si γ’ precipitates found in several previous studies. In X-750, L12-structured γ’ precipitates were found, and multivariate statistical analysis confirmed the 3:1 stoichiometry (Ni3(Ti,Al)) of the γ’ precipitates and the superlattice reflections confirmed the stability of the crystal structure of these γ’ precipitates, indicating higher-than-expected precipitate stability under high-dose neutron irradiation.

Neutron irradiation-induced modification of electrical and structural properties of GaN epifilms grown on Al2O3 (0001) substrate

The effects of fast neutron (up to 1.75 × 1020 f.n. cm–2) and fast plus thermal neutron (up to 3.5 × 1020 f.t.n. cm–2) irradiation on the electrical properties and crystal lattice of GaN films have been studied. It is shown that high-dose neutron irradiation induces a transition of n-type samples into a semi-insulating state with resistivity of ∼1010 Ω cm at room temperature, while p-type films undergo transition to a high-resistance state. The subsequent irradiation of the samples leads to a decrease of resistivity to ∼105 Ω cm at the final neutron fluence. Raman spectra show appearance of the disorder-activated Raman scattering modes at 210, 300 and 670 cm−1 linked to defects in Ga and N sublattices. X-ray diffraction measurements reveal an expansion of the GaN c-lattice parameter up to a saturation level of 0.42% at the final fluence, while the a-lattice parameter remains nearly unchanged. The initial value of c restores at 1000 °C annealing temperature, with the main annealing stage at 150 °C–400 °C. The significant role of thermal neutrons (E ≤ 0.1 MeV) in the damage build-up in GaN is shown.

Study of Radiation Properties of Structural Materials for Advanced Nuclear Energy Systems

This paper introduces briefly the development of nuclear energy systems from the GEN I to the advanced GEN IV, the progress of manufacturing radiation resistant materials associated with the development of nuclear energy systems and the new methods of investigating radiation properties of the structural materials for the GEN IV nuclear energy systems at first. Irradiation induced deterioration of materials properties is a bottle neck problem, which must be investigated and solved for the development of the GEN IV nuclear energy systems. Unfortunately, all the currently available neutron sources cannot meet the requirements of investigating radiation properties of structural materials irradiated by high dose neutron irradiation in the GEN IV nuclear energy systems. Therefore, two new methods of the accelerator heavy ion irradiation that simulates the high-dose neutron irradiation and the triple beam irradiation that mimics the real neutron irradiation environment in the GEN IV nuclear energy systems have been developed. These two methods are introduced in this paper. The present status of the study on radiation properties of structural materials for nuclear energy systems of the new generation and the near future development plan at China Institute of Atomic Energy (CIAE) are described also. The accelerator heavy ion irradiation facilities for different applications and the simultaneous triple beam irradiation platform with three separate accelerators or implanters have been established at the HI-13 tandem accelerator of CIAE. A series of structural materials for nuclear energy systems, such as the home-made modified austenic steel, CLAM steel, 1515 steel, Tantalum, Tungsten, etc. have been tested and investigated systematically. A superconducting linear accelerator and a variable energy heavy ion cyclotron have been imported from abroad for a better performance of the study. Combined with the currently existing facilities of 2×13 MeV and 2×1.7 MV tandem accelerators, 30 and 100 MeV proton cyclotrons, China experimental fast reactor, China advance research reactor, Miniature neutron source reactor, etc. a comprehensive and advanced system of experimental irradiation platform for structural materials of nuclear energy systems will be established in the near future for both domestic and foreign users.

Tritium and helium release from beryllium pebbles neutron-irradiated up to 230 appm tritium and 3000 appm helium

Study of tritium and helium release from beryllium pebbles with diameters of 0.5 and 1mm after high-dose neutron irradiation at temperatures of 686–968K was performed. The release rate always has a single peak, and the peak temperatures at heating rates of 0.017K/s and 0.117K/s lie in the range of 1100–1350K for both tritium and helium release. The total tritium release from 1mm pebbles decreases considerably by increasing the irradiation temperature. The total tritium release from 0.5mm pebbles is less than that from 1mm pebbles and remains constant regardless of the irradiation temperature. At high irradiation temperatures, open channels are formed which contribute to the enhanced tritium release.

High-dose neutron irradiation of Hi-Nicalon Type S silicon carbide composites. Part 2: Mechanical and physical properties

Nuclear-grade silicon carbide (SiC) composite material was examined for mechanical and thermophysical properties following high-dose neutron irradiation in the High Flux Isotope Reactor at a temperature range of 573–1073K. The material was chemical vapor-infiltrated SiC-matrix composite with a two-dimensional satin weave Hi-Nicalon Type S SiC fiber reinforcement and a multilayered pyrocarbon/SiC interphase. Moderate (1073K) to very severe (573K) degradation in mechanical properties was found after irradiation to >70dpa, whereas no evidence was found for progressive evolution in swelling and thermal conductivity. The swelling was found to recover upon annealing beyond the irradiation temperature, indicating the irradiation temperature, but only to a limited extent. The observed strength degradation is attributed primarily to fiber damage for all irradiation temperatures, particularly a combination of severe fiber degradation and likely interphase damage at relatively low irradiation temperatures.

Effects of neutron irradiation on optical and chemical properties of CR-39: Potential application in neutron dosimetry

Effects of high-dose neutron irradiation on chemical and optical properties of CR-39 were studied using FTIR (Fourier Transform Infrared) and UV–vis (Ultraviolet–Visible) spectroscopy. The primary goal was to find a correlation between the neutron dose and the corresponding changes in the optical and chemical properties of CR-39 resulted from the neutron irradiation. The neutrons were produced by bombarding a thick Be target with 22-MeV protons. In the FTIR spectra, prominent absorbance peaks were observed at 1735cm−1 (CO stretching), 1230cm−1(C–O–C stretching), and 783cm−1(C–H bending), the intensities of which decreased with increasing neutron dose. The optical absorbance in the visible range increased linearly with the neutron dose. Empirical relations were established to estimate neutron doses from these optical properties. This technique is particularly useful in measuring high doses, where track analysis with an optical microscope is difficult because of track overlapping.

Effects of manufacturing process on impact properties and microstructures of ODS steels

Oxide dispersion strengthened (ODS) steels are notable advanced alloys with durability to a high-temperature and high-dose neutron irradiation environment because of their good swelling resistance and mechanical properties under neutron irradiation. 9–12Cr-ODS martensite steels have been developed in the Japan Atomic Energy Agency as the primary candidate material for the fast reactor fuel cladding tubes. They would also be good candidates for the fusion reactor blanket material which is exposed to high-dose neutron irradiation. In this work, modification of the manufacturing process of 11Cr-ODS steel was carried out to improve its impact property. Two types of 11Cr-ODS steels were manufactured: pre-mix and full pre-alloy ODS steels. Miniature Charpy impact tests and metallurgical observations were carried out on these steels. The impact properties of full pre-alloy ODS steels were shown to be superior to those of pre-mix ODS steels. It was demonstrated that the full pre-alloy process noticeably improved the microstructure homogeneity (i.e. reduction of inclusions and pores).

Characteristics of microstructure, swelling and mechanical behaviour of titanium beryllide samples after high-dose neutron irradiation at 740 and 873 K

Titanium beryllide Be12Ti is one of the candidate materials for advanced neutron multiplier in the helium-cooled breeding blanket of Demonstration Power Plant (DEMO).Cylinder-shaped specimens (diameters of 3mm) consisting, mainly, of Be12Ti phase were investigated after high-dose neutron irradiation at 740 and 873K in High Flux Reactor in Petten, Netherlands.The sizes and shapes of radiation-induced gas bubbles as well as swelling of samples were evaluated by means of transmission electron microscopy (TEM). Mechanical behaviour of titanium beryllide specimens in initial and irradiated states was investigated using compression tests under constant loading.Obtained results are discussed and compared with available data on pure beryllium.

SiC/SiC and W/SiC/SiC Composite Heater by NITE-method for IFMIF and Fission Reactor Irradiation Rigs

For nuclear and fusion materials R&D, high dose neutron irradiation at very high temperature is one of the most important and challenging techniques. As the part of Japan/EU Broader Approach (BA) program for fusion, International Fusion Materials Irradiation Facility/Engineering Validation and Engineering Design Activities (IFMIF/EVEDA) project is on-going. Where, High Flux Test Module (HFTM) design aiming up to 1000°C has been intensively investigated. Innovative SiC/SiC and tungsten embedded SiC/SiC (W/SiC/SiC) heaters have been recognized potential and promising candidates. These heaters were fabricated successfully by NITE method and were evaluated. The SiC/SiC heater showed weak temperature dependence up to 1000°C on in-plane electrical conductivity. Irradiation capsule for heater performance evaluation under neutron irradiation in Belgian Reactor 2 (BR2) is also under preparation. The results of heater performance and basic properties evaluation and of underlying microstructure behavior analysis are provided.

Low cycle fatigue properties of reduced activation ferritic/martensitic steels after high-dose neutron irradiation

This paper focuses on the low cycle fatigue (LCF) behaviour of reduced activation ferritic/martensitic steels irradiated to a displacement damage dose of up to 70 dpa at 330–337 °C in the BOR 60 reactor within the ARBOR 2 irradiation programme. The influence of neutron irradiation on the fatigue behaviour was determined for the as-received EUROFER97, pre-irradiation heat-treated EUROFER97 HT and F82H-mod steels. Strain-controlled push–pull loading was performed using miniaturized cylindrical specimens at a constant temperature of 330 °C with total strain ranges between 0.8% and 1.1%. Comparison of the LCF behaviour of irradiated and reference unirradiated specimens was performed for both the adequate total and inelastic strains. Neutron irradiation-induced hardening may have various effects on the fatigue behaviour of the steels. The reduction of inelastic strain in the irradiated state compared with the reference unirradiated state at common total strain amplitudes may increase fatigue lifetime. The increase in the stress at the adequate inelastic strain, by contrast, may accelerate fatigue damage accumulation. Depending on which of the two effects mentioned dominates, neutron irradiation may either extend or reduce the fatigue lifetime compared with the reference unirradiated state. The results obtained for EUROFER97 and EUROFER97 HT confirm these considerations. Most of the irradiated specimens show fatigue lifetimes comparable to those of the reference unirradiated state at adequate inelastic strains. Some irradiated specimens, however, show lifetime reduction or increase in comparison with the reference state at adequate inelastic strains.

Effect of electron- and neutron-irradiation on Fe-Cu model alloys studied by positron annihilation spectroscopy

Electron- and neutron-irradiation effects on dilute Fe-Cu model alloys of nuclear reactor pressure vessel steels are studied by positron annihilation spectroscopy. We have found that, not only by high-dose neutron-irradiation but also by low-dose electron-irradiation, the aggregation of Cu atoms and vacancies takes place and the ultrafine Cu precipitates are formed after post-irradiation annealing at 400°C. In spite of large difference in the irradiation doses between the electron- and the neutron-irradiated samples, no significant difference is observed in the isochronal annealing behaviour above 400°C of positron annihilation and micro-hardness, indicating that small amount of extra vacancies enhance the aggregation of Cu atoms in Fe during the annealing-out process of the vacancies.

Formation of voids and secondary-phase precipitates in the Fe-16Cr-15Ni-2Mo-1Mn-Ti-Si steel under high-doze neutron irradiation and during post-irradiation annealing

The effect of high-dose neutron irradiation on the structural changes in Fe-16Cr-15Ni-2Mo-1Mn-Ti-Si austenitic steel have been investigated. Samples irradiated at temperatures of 390, 500, and 600° to damage doses of 46, 86, and 46 dpa, respectively, were analyzed by electron microscopy and dilatometry. The quantitative characteristics of radiation voids and secondary-phase precipitates formed under neutron irradiation are obtained. Their behavior upon heating to 700°C and annealing at this temperature for 2 h is studied. It is shown that annealing leads to the dissociation of small voids, which is accompanied by the growth of large ones. The secondary-phase precipitates are partially dissolved upon annealing, and their volume fraction decreases.