Recent activities in the field of nuclear materials and nuclear fuels

ABSTRACT

On the research and development of nuclear materials and fuels, many of outstanding papers have been presented in scientific journals, including the Journal of Nuclear Science and Technology. Some topics have been covered in this summary from the latest activities of nuclear materials and fuels.

KEYWORDS:

• Plasma facing materials
• reactor pressure vessel
• fuel cladding
• molten core materials
• thermal conductivity
• fission product
• accident tolerant fuels for LWRs

As a plasma-facing material (PFM) in fusion reactors, W has been proposed due to its outstanding properties. However, the degradation of the material properties such as permeation of hydrogen isotopes through and their trapping in W is one of important issues [1–4]. In order to clarify the key issues, polycrystalline W plate was implanted with 80 keV H²⁺ ions at room temperature [5]. Isochronal annealing revealed two hydrogen release stages which might be associated with the reported activation energies. In addition, H₂ blister formation was observed near the surface of the as-implanted W. The blister distribution remained unchanged after thermal annealing up to 600°C. It would be the critical issue for use of W as PFM. As a blanket materials, the investigation of MX precipitation behavior in the RAFM, F82H [6] revealed that the TaX (X: C and/or N) phase surely varied depending on the chemical composition of alloy and heat treatment; some TaX precipitates were unstable during tempering, MX precipitates were inter-granular, and sub-nano-metric MX were not found in the matrix.

On the fission reactor materials, the irradiation and annealing behavior of reactor pressure vessel steel was studied by positron annihilation doppler broadening (PADB) spectroscopy [7]. PADB measurement revealed the interaction and clustering of vacancies with solute clusters. The positron annihilation parameter and positron diffusion length showed the recovery of irradiation-induced defects.

In the issue of fuel cladding materials, the dependence of iodine partial pressure on its stress corrosion cracking behaviors of a Zr–Sn–Nb alloy was investigated by ring tensile tests [8]. Results showed that the maximum load, fracture displacement, tensile strength, and fracture energy of the Zr–Sn–Nb decreased monotonically with the iodine partial pressure. The critical iodine partial pressure of the Zr–Sn–Nb alloys was lower than 102 Pa under the present conditions.

In order to understand and predict the progression of core meltdown accidents at nuclear power plants, it is important to understand the behavior of molten core materials. The melting behavior of Ag–In–Cd alloys used in the control rods of pressurized water reactors, which are known to melt first at severe accidents, was investigated [9]. Although Ag is known as a material that has one of the highest thermal conductivities, the thermal conductivity of liquid Ag–In alloys is much lower than that of pure liquid Ag, but almost the same or less than that of liquid in all Ag1−xInx (x = 0.2–0.8) alloys.

On the other hand, the isothermal and cyclic corrosion behavior of the yttria (Y₂O₃)-stabilized zirconia (ZrO₂) in a LiCl-Li₂O molten salt were investigated [10]. The corrosion rate was very low in the molten salt of LiCl-Li₂O for 168 h and seven thermal cycles, while 10 times higher in the molten salt of LiCl-Li₂O-Li for 168 h. Additionally, in the molten salt of LiCl-Li₂O-Li for 168 h, a large amount of Li₂ZrO₃ was formed. The introduction of Y₂O₃-stabilized ZrO₂ was beneficial in increasing the hot corrosion resistance of the structural materials used to handle molten salts containing Li₂O at elevated temperature without forming lithium at the cathode during the electrolytic reduction process.

To predict the damage condition of reactor vessel, the dissolution behavior of core structure materials by molten metallic corium (stainless steel + B₄C) originated from control rod and its cladding was investigated [11]. Experiment results on the immersion suggested two types of dissolution mode in this system: (1) chemical dissolution by eutectic reaction between Fe and B, and (2) physical dissolution caused by the grains falling off from solid steel due to infiltration of molten metal. Moreover, on the basis of kinetic analysis, it was considered that the chemical dissolution in this system was slow. Therefore, the dissolution is ascribed to a main mechanism that physical dissolution precedes chemical dissolution. In addition, comparison between Na-bonded and He-bonded control rods with and without the shroud tube in a wide burn-up range [12] revealed that the diametric changes were smaller in the Na-bonded B₄C absorber pins than in the He-bonded B₄C pins. It is concluded that the Na-bonded B₄C pins are very effective for achieving long-life control rods.

In order to clarify the melting phenomena and relocation of the structural material in the core of the reactor, interaction and melting behavior among B₄C, 304SS (stainless steel), and Zircaloy-4 in atmosphere containing H₂/H₂O at 1473 K are investigated [13]. The results showed that the reaction at the interface between B₄C and 304SS under H₂O atmosphere was slow and an oxidation was observed after 3600s. Under H₂O/H₂ atmosphere, the concentration of B and C in the 304SS increased and the 304SS melted. Despite the atmosphere, an oxide layer formed on the surface of Zircaloy-4, and thus the reactions proceeded slowly when the Zircaloy-4 was contacted with B₄C and 304SS. Under H₂O atmosphere, continuous oxidation happened to 304SS, and 304SS was partially melted. Under H₂ atmosphere, the 304SS was also melted due to the diffusion of B and C from the B₄C. In addition, the oxidation of B₄C affected the oxidation behavior of 304SS and Zircaloy-4.

On the volatile fission products (FPs) such as Cs and I, the release mechanism from fuels under the accident is still not completely understood. In recent years, the wettability of liquid FPs against solid fuels has been focused because the interface between the fuel surface and the FPs becomes the migration pathway, which might have large influences on the release behavior of the FPs [14]. The sessile drop test for the wettability of liquid CsI and B₂O₃ on yttria-stabilized zirconia (YSZ) solid surface revealed that liquid CsI exhibited extremely high wettability against the YSZ surface with the contact angle of nearly zero [15]. This high wettability may act to suppress the FPs release. Furthermore, it was confirmed that the crystal orientation and surface roughness of the YSZ solids have large influences on the wettability of liquid B₂O₃. The present results contribute for deep understanding of the release behavior of the volatile FPs from fuels.

After the Fukushima Daiichi nuclear power plant accident, research on accident tolerant fuels has become a hot topic [16]. There are many candidates for cladding materials of accident tolerant fuels such as Zr alloys with improved oxidation resistance [17], silicon carbide based ceramic composites [18,19], and iron-based alloys [20,21]. Developments and characterizations of such materials are now ongoing to improve the safety of light water reactors under accident conditions.

Disclosure statement

No potential conflict of interest was reported by the authors.