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Radiation fields occur in nuclear power plant primary circuits, and decontamination is needed before the permanent shutdown of a reactor. Radiation field reduction in the primary coolant system of water-cooled reactors can be achieved by using a decontamination solution. As part of this, the corrosion behavior and characterization of oxide layers formed on different kinds of metal should be examined. In this study, NiFe2O4 thin film was deposited on substrates of stainless steel, inconel and carbon steel via electron beam evaporation. The reaction between nickel ferrite and different metal ions results in varying compositions of oxide layers. The phase, structure and morphology of materials were confirmed by SEM, XRD and XPS analysis, and the magnetic properties of each deposited sample were observed by using vibrating sample magnetometer (VSM). The characteristics of the surface-coated metals are the main focus of this work.

The behavior of dynamic strain aging (DSA) in a Zr-1.5Nb-0.4Sn-0.2Fe-0.1Cr alloy strip was investigated at temperature ranges of 25–600 °C via a tensile test. The tensile test was performed at two different strain rates 8.33 × 10-5 and 1.67 × 10-2 s-1. The shear stress of the alloy strip revealed a linear dependency on the test temperature when the specimens were tested under a higher strain rate (1.67 × 10-2 s-1). However, the linear relationship was broken due to DSA when the samples were deformed under a lower strain rate (8.33 × 10-5 s-1). The discrepancy was most significant at 400 °C. The trend in DSA behavior was similar irrespective of the orientation of the samples, i.e., rolling direction (RD) or transverse direction (TD). However, the effect of DSA was larger in the TD samples than the RD samples. The phenomena were interpreted to the variation in work hardening exponents and strain rate sensitivity. The value of the exponent decreased from 0.14 to 0.08 along a RD and from 0.1 to 0.07 along a TD, respectively. However, the smallest value was observed at 400–500 °C irrespective of the specimen orientation, which is consistent with the DSA behavior. It is suggested that the DSA was caused by an interaction of moving dislocations with solute atoms typically oxygen.

Zr alloy specimens were coated with Cr–Al alloy to enhance their resistance to oxidation. The coated samples were oxidized at 1200 °C in a steam environment for 300 s and showed extremely low oxidation when compared to uncoated Zr alloy specimens. The microstructure and elemental distribution of the oxides formed on the surface of Cr–Al alloys have been investigated by transmission electron microscopy (TEM) and X-ray photoelectron spectroscopy (XPS). A very thin protective layer of Cr2O3 formed on the outer surface of the Cr–Al alloy, and a thin Al2O3 layer was also observed in the Cr–Al alloy matrix, near the surface. Our results suggest that these two oxide layers near the surface confers excellent oxidation resistance to the Cr–Al alloy. Even after exposure to a high temperature of 1200 °C, inter-diffusion between the Cr–Al alloy and the Zr alloy occurred in very few regions near the interface. Analysis of the inter-diffusion layer by high-resolution transmission electron microscopy (HRTEM) and energy dispersive X-ray spectroscopy (EDS) measurement confirmed its identity as Cr2Zr.

Zirconium alloys are widely used to fabricate nuclear fuel claddings, and thus is desirable to improve the resistance of such alloys to corrosion and structural instability. In this study, Ta was used as an alloying element to improve the corrosion and oxidation resistance of zirconium alloys. The model alloy (TaZL) contained 0.03 wt% Ta and other elements with a proportion of less than 1 wt% in total (0.1 wt% Nb, 0.4 wt% Fe, 0.2 wt% Cr) in a zirconium base. The corrosion test involving pressurized water at 360 °C and oxidation test involving steam at 1200 °C indicated that TaZL exhibited the lowest weight gains among those of compared conventional and advanced Zr alloys. The corrosion and oxidation resistances of TaZL were respectively improved by 4 and 1.5 times compared to the corresponding values of Zircaloy-4. The microstructures of the oxide formed on TaZL were columnar along the oxide growth direction and did not change from columnar to equiaxed, which resulted in the high resistance of the alloy to corrosion and oxidation.

Anode materials providing a high specific capacity with a high cycling performance are one of the key parameters for lithium ion batteries’ (LIBs) applications. Herein, a high-capacity NiFe2O4(NFO) film anode is prepared by E-beam evaporation, and the effect of the heat treatment is studied on the microstructure and electrochemical properties of LIBs. The NiFe2O4 film annealed at 800 °C (NFO-800) showed a highly crystallized structure and different surface morphologies when compared to the electrode annealed at a lower temperature (NFO-600, NFO-700). In the electrochemical measurements, the high specific capacity (1804 mA g−1) and capacity retention ratio (95%) after 100 cycles were also achieved by the NFO-800 electrode. The main reason for the good electrochemical performance of the NFO-800 electrode is a high structure integrity, which could improve the cycle stability with a high discharge capacity. The NiFe2O4 electrode with an annealing process could be further proposed as an alternative ferrite material.

For enhancing accident tolerance in current operating LWRs, Zr-based claddings with a protective layer, which have outstanding resistance to high-temperature oxidation, are selected as strong candidates for accident-tolerant fuel claddings. One of the key factors determining the successful application of these coated claddings is their ability to maintain their coating layer during normal operation without significant damage. This study aimed to verify the scratch and fretting wear behaviors of coated Zircaloy-4 claddings deposited with a Cr–10Al binary alloy target (CrAl) by the arc ion plating method. In the scratch tests, the commercial Zircaloy-4 samples showed accumulated scratch grooves with ductile chipping caused by adhesive failure. However, there was no gross spallation, except for well-developed scratch traces caused by plastic deformation, on the entire stroke length of the CrAl-coated Zircaloy-4. The wear resistance of the CrAl-coated Zircaloy-4 against the as-received Zr grid was improved significantly because of its high load support capability owing to the coating layer and the formation of a load-bearing layer by the material transfer when compared with the results of the uncoated Zircaloy-4. Furthermore, when the Zr grid was pre-oxidized, the wear amount of the CrAl-coated Zircaloy-4 became negligible owing to the formation of a glaze layer by the interaction with the surface oxide of the Zr grid. According to the results of the high-temperature steam oxidation tests, the defected region of the CrAl coating layer caused by the scratch and fretting wear showed no further significant damage and still acted as a protective layer for enhancing the accident tolerance of the fuel cladding.

We report the development of magnetically steerable self-propelled micromotors that selectively remove radioactive Cs from contaminated water. Mesoporous silica microspheres were functionalized with the highly Cs-selective copper ferrocyanide, and half of the adsorptive particle surface was then coated with ferromagnetic Ni and catalytic Pt layers to fabricate Janus micromotors. The micromotor adsorbent displayed random propulsion in an H2O2 solution via catalytic bubble evolution from the Pt surface, and the micromotor adsorbent self-propulsion resulted in an 8-fold higher Cs removal compared to the stationary adsorbent within one hour. The ferromagnetism of the Ni layer allowed the micromotor adsorbent to be magnetically and remotely steerable, and the propulsion speed under a magnetic field was ˜11-fold greater than it was in the absence of the magnetophoretic force. The adsorption of Cs by the self-propelling micromotor adsorbent and the subsequent magnetic recovery of the adsorbent enabled the successful removal of radioactive 137Cs from aqueous solutions. More than 98% of the radioactive 137Cs ions were removed from solution, even in the presence of competing ions, such as Na+ (1000 ppm).

In the present work, oxide dispersion is used to improve the high-temperature strength of FeCrAl alloys, which have excellent high-temperature oxidation resistance. Laser-beam scanning is introduced as a new method for manufacturing oxide-dispersion-strengthened alloys. Various oxide powders were initially coated on the surface of the FeCrAl alloy, which was then surface-treated by a laser beam to cause the oxides to penetrate and disperse into the metal matrix. The microstructure of the dispersed oxide layer was investigated by scanning electron microscopy and energy dispersive X-ray spectroscopy to reveal the mechanisms involved in oxide dispersion during the laser-beam scanning process. The effect of the different powder materials on the strengthening of laser-beam-scanned FeCrAl was examined by tensile tests, which were carried out at room temperature, 380 °C, and 500 °C. Regardless of the powder material, the thin oxide-dispersed-strengthened specimens showed similar tensile strengths to that of untreated samples at room temperature, but showed improvements of up to 24% at elevated temperatures. The total elongation of the surface-treated samples was considerably lower than that of the untreated sample at 380 °C; however, this loss of ductility was not significant at 500 °C.

ATF claddings are defined as those that provide considerably increased response time in the event of a reactor accident while providing performance similar to or better than that of the commercial zirconium (Zr) cladding during normal operation. Surface modification technology consists of a coating technique for increasing oxidation resistance and an oxide dispersion strengthened (ODS) for improving high temperature strength. The CrAl alloy and Cr coating were applied by arc ion plating method and the FeCrAl alloy coating was applied by 3D laser coating. The ODS structure is prepared by 3D laser scanning technology using oxide powder. The prepared specimens were evaluated for corrosion/oxidation and mechanical behavior in normal and accident conditions. In the surface-modified Zr cladding concept, the oxidation resistance and strength at high temperature can be improved considerably over commercial Zr-based alloy cladding by a combination of two types of surface modification technology of the surface coating and partial oxide dispersion strengthened (ODS) treatment. From the various performance evaluations, the surface-modified Zr cladding concept shows promise as an ATF cladding.

In this study, the Fe3O4 thin film was prepared as anode materials to simultaneously provide higher reversible Li+ capacity compared to 2D materials with high cycle properties for lithium ion batteries. Nano-crystalline Fe3O4 thin film has been prepared by using an electron beam evaporation system followed by heat treatment. Electrochemical measurement showed that the as-fabricated Fe3O4 thin film showed the conventional charge–discharge voltage profiles of magnetite and good cycle performance. The initial reversible capacity of Fe3O4 thin film was maintained during the 100 cycles with no capacity fading in a potential range of 0.005–3.0 V (vs. Li+/Li).

To increase the mechanical strength of Zircaloy-4 cladding at high temperatures, partial oxide dispersion-strengthened (ODS) treatment of the cladding tube surface was achieved by using laser processing technology. The microstructural characteristics and stability of the ODS layer formed on the Zircaloy-4 cladding surface were analyzed at temperatures up to 1000 °C. Ring tensile and loss-of-coolant accident (LOCA) simulation tests were performed to evaluate the mechanical properties of the surface ODS treated Zircaloy-4 cladding tube. The formation and uniform distribution of Y2O3 particles formed in the Zr matrix were identified, and the stability of the particles was confirmed up to 1000 °C. When compared to the reference Zircaloy-4 cladding tube, the surface ODS treated Zircaloy-4 cladding tube showed improved mechanical properties at both room temperature and 500 °C, as well as under LOCA simulation conditions.

A helium cooled liquid lithium (or lead lithium) concept has been developed to design a liquid breeder blanket in South Korea. Ferritic–martensitic steel (FMS) was selected as a structural material for fusion reactors, and a commercial-scale advanced reduced-activation alloy (ARAA) has been developed. An experimental loop for a liquid breeder (ELLI), which we designed and fabricated ourselves, was constructed to validate the electromagnetic (EM) pump as well as to test the effects of the magnetohydrodynamics of a liquid metal flow, and to investigate the compatibility of PbLi with the structural materials of FMS. Thus far, performance tests on each component, such as the heaters and control systems used for heating the loop, were conducted, and characteristic tests using a magnet and an EM pump were carried out. Corrosion tests using ELLI were conducted using grade 91 FMS steel and ARAA for a 250-h period (150+100 h). In this paper, a commercial FMS with grade 91 and developed ARAA were used for corrosion test specimens to compare the corrosion characteristics in the flowing PbLi loop. A corrosion test was conducted for investigating the compatibility of PbLi using structural materials. The effects of the oxide coatings on the prevention of the corrosion progress were also investigated.

FeCrAl and Mo layers were cold-sprayed onto a Zr surface, with the Mo layer introduced between the FeCrAl coating and the Zr matrix preventing high-temperature interdiffusion. Microstructural characterization of the first-deposited Mo layer and the Zr matrix immediately below the Mo/Zr interface was performed using transmission electron microscopy, and near-interface elemental distributions were obtained using energy-dispersive X-ray spectroscopy. The deformation of the coated Mo powder induced the formation of microbands and mechanically interlocked nanoscale structures. The mechanical behavior of Zr with a coating layer was compared with those characteristic of conventional Zr samples. The coated sample showed smaller strength reduction in the test conducted at elevated temperature. The hardness and fracture morphology of the Zr matrix near the interface region were investigated to determine the effect of impacting Mo particles on the matrix microstructure. The enhanced hardness and cleavage fracture morphology of the Zr matrix immediately below the Mo/Zr interface indicated the occurrence of localized deformation owing to Mo particle impact.

An oxide-dispersion-strengthened (ODS) layer was formed on Zircaloy-4 tubes by a laser beam scanning process to increase mechanical strength. Laser beam was used to scan the yttrium oxide (Y2O3)–coated Zircaloy-4 tube to induce the penetration of Y2O3 particles into Zircaloy-4. Laser surface treatment resulted in the formation of an ODS layer as well as microstructural phase transformation at the surface of the tube. The mechanical strength of Zircaloy-4 increased with the formation of the ODS layer. The ring-tensile strength of Zircaloy-4 increased from 790 to 870 MPa at room temperature, from 500 to 575 MPa at 380°C, and from 385 to 470 MPa at 500°C. Strengthening became more effective as the test temperature increased. It was noted that brittle fracture occurred at room temperature, which was not observed at elevated temperatures. Resistance to dynamic high-temperature bursting improved. The burst temperature increased from 760 to 830°C at a heating rate of 5°C/s and internal pressure of 8.3 MPa. The burst opening was also smaller than those in fresh Zircaloy-4 tubes. This method is expected to enhance the safety of Zr fuel cladding tubes owing to the improvement of their mechanical properties. Keywords: Laser Surface Treatment, Microstructure, Oxide Dispersion Strengthened Alloy, Tensile Strength, Zirconium Alloy

Zircaloy-4 is a traditional zirconium-based alloy developed for application in nuclear fuel cladding tubes. The surfaces of Zircaloy-4 tubes were treated using a laser beam to increase their mechanical strength. Laser beam scanning of a tube coated with yttrium oxide (Y2O3) resulted in the formation of a dispersed oxide layer in the tube’s surface region. Y2O3 particles penetrated the Zircaloy-4 during the laser treatment and were distributed uniformly in the surface region. The thickness of the dispersed oxide layer varied from 50 to 140 μm depending on the laser beam trajectory. The laser treatment also modified the texture of the tube. The preferred basal orientation along the normal to the tube surface disappeared, and a random structure appeared after laser processing. The most obvious result was an increase in the mechanical strength. The tensile strength of Zircaloy-4 increased by 10–20% with the formation of the dispersed oxide layer. The compressive yield stress also increased, by more than 15%. Brittle fracture was observed in the surface-treated samples during tensile and compressive deformation at room temperature; however, the fracture behavior was changed in ductile at elevated temperatures.

NiFe2O4 thin film has been prepared by electron beam evaporation method and has been investigated as anode materials providing large reversible Li+ capacity with high cycling performance for lithium ion batteries. The NiFe2O4 thin film crystallized with annealing process after NiFe2O4 thin film deposition. In electrochemical measurements, the first discharge and charge capacity of NiFe2O4 thin film were found to be 1693 mAh g−1 and 1108 mAh g−1, which is above its theoretical capacity (915 mAh g−1). In addition, the cycle performance test of NiFe2O4 thin film also showed steady charge-discharge capacity at 0.1 C during 100 cycles compared to other NiFe2O4 materials, indicating that the empty space or porous surface of electrode was beneficial for accessibility of electrolyte and served as buffer to alleviate stress.

Tungsten was joined to ferritic-martensitic steel (FMS) for application in a plasma facing component. Zirconium foil was investigated as an interlayer material for the joining of W to FMS. Repeated...

The mechanical strength of Zircaloy-4 was enhanced by surface treatment consisting of laser beam scanning of an oxide-coated Zircaloy-4 sheet. An yttrium oxide (Y2O3) coating 7–55 μm thick was applied to Zircaloy-4 sheets. Then, a laser beam was used to form a dispersed oxide layer on the Zircaloy-4 metal surface. The thickness of the dispersed oxide layer varied from 80 to 200 μm depending on the process parameters. Tensile tests of samples having both surfaces treated were conducted at room temperature and 380 °C. The tensile strength of Zircaloy-4 at room temperature was increased by up to 20% with the formation of a thin dispersed oxide layer with a thickness

This study investigates protective coatings for improving the high temperature oxidation resistance of Zr fuel claddings for light water nuclear reactors. FeCrAl alloy and Cr layers were deposited onto Zr plates and tubes using cold spraying. For the FeCrAl/Zr system, a Mo layer was introduced between the FeCrAl coating and the Zr matrix to prevent inter-diffusion at high temperatures. Both the FeCrAl and Cr coatings improved the oxidation resistance compared to that of the uncoated Zr alloy when exposed to a steam environment at 1200 °C. The ballooning behavior and mechanical properties of the coated cladding samples were studied under simulated loss-of-coolant accident conditions. The coated samples showed higher burst temperatures, lower circumferential strain, and smaller rupture openings compared to the uncoated Zr. Although 4-point bend tests of the coated samples showed a small increase in the maximum load, ring compression tests of a sectioned sample showed increased ductility.

Zr-alloys are used as nuclear fuel cladding tubes in nuclear reactors. Recently, surface coating of conventional Zr-alloy tubes has been performed to enhance their resistance to corrosion and high-temperature oxidation. FeCrAl is a candidate coating material because of its extremely low oxidation rates and desirable mechanical properties. In this study, FeCrAl was deposited directly on a Zr-alloy using 3D printing technology. The processing conditions, such as laser beam power, scanning speed, and powder/gas feed rates, were optimized to attain a dense and adhesive FeCrAl layer. The minimum beam power required to melt the FeCrAl source powder was found to be 160 W. The FeCrAl layer coagulated when the scanning speed was low and the powder feed rate was high, and it became sparse and porous when the scanning speed was high and feed rate was low. The FeCrAl-coated Zr samples were tested for oxidation at 1200 °C. A thin Al2O3 (2–3 μm) formed on the FeCrAl surface, and a diffusion layer developed between the FeCrAl/Zr interfaces. By comparison, a thick ZrO2 (approximately 120 μm) layer formed in the Zr-alloy without coating. The oxidation resistance of the Zr-alloys increased by approximately 50 times by the formation of FeCrAl layer on their surfaces. This 3D printing method is expected to improve the safety of Zr fuel cladding tubes.

Joining technology for dissimilar materials was developed for the fabrication of an ITER blanket first-wall, which consisted of Be, CuCrZr, and stainless steel (SS). The Be/CuCrZr/SS joint was fabricated using a hot isostatic pressing (HIP) method. Beryllium armor was joined to the CuCrZr/SS block at 580 °C under 100 MPa. The optimal interlayer coatings of Cr/Cu and Ti/Cr/Cu were developed using an ion-beam assisted physical vapor deposition. Beryllium is also a candidate armor material for the TBM first-wall. Successful joining of Be to ferritic-martensitic steel (FMS) was accomplished using an HIP method by introducing the thick diffusion barrier. A thick diffusion barrier of a Cu foil(10 μm) limited the excessive diffusion and prevented the formation of brittle phases at the Be/FMS interface. Be and FMS were bonded at 650–850 °C; however, a temperature of lower than 750 °C was recommended to avoid material degradation of FMS. In addition, the joining of W to FMS has been developed. Tungsten is another armor material applicable to more severe plasma conditions. The large difference in the thermal expansion between W and FMS was resolved by introducing the Ti interlayer and Mo separator. Moreover, the double-staged HIP (the first stage at 900 °C and 100 MPa and the second stage at 750 °C and 70 MPa) was applied to suppress the edge delamination of W/FMS joints during thermal history.

FeCrAl and Zr alloys were joined by a hot isostatic pressing (HIP) method for fusion energy applications. The optimum conditions for the joining process were studied. The HIP temperatures were varied from 700 to 1050 °C. The mechanical properties of the HIPed samples were evaluated by four-point bending and tensile tests. The FeCrAl and Zr alloys HIPed at 700 °C showed higher joint strength than the other samples. The joint strength decreased with an increase in the HIP temperature from 700 to 950 °C and significantly dropped at 1050 °C. Transmission electron microscopy, scanning electron microscopy, and optical microscopy were used to characterize the joints and interface region of the HIPed samples. The joints appeared to be tightly bonded and no intermetallic compounds or gaps were observed at the interface for HIP temperatures up to 950 °C. A diffusion layer formed at the interface and its thickness increased with the HIP temperature. HIP at 1050 °C, on the other hand, resulted in significant inter-diffusion and formation of brittle inter-metallic compounds at the interface.

SiC-based ceramic composites are currently being considered for use in fuel cladding tubes in light-water reactors. The joining of SiC ceramics in a hermetic seal is required for the development of ceramic-based fuel cladding tubes. In this study, SiC monoliths were diffusion bonded using a Ti foil interlayer and additional Si powder. In the joining process, a very low uniaxial pressure of ∼0.1 MPa was applied, so the process is applicable for joining thin-walled long tubes. The joining strength depended strongly on the type of SiC material. Reaction-bonded SiC (RB-SiC) showed a higher joining strength than sintered SiC because the diffusion reaction of Si was promoted in the former. The joining strength of sintered SiC was increased by the addition of Si at the Ti interlayer to play the role of the free Si in RB-SiC. The maximum joint strength obtained under torsional stress was ∼100 MPa. The joint interface consisted of TiSi2, Ti3SiC2, and SiC phases formed by a diffusion reaction of Ti and Si.

The Na 0.6 Li 0.6 [Mn 0.72 Ni 0.18 Co 0.10 ]O 2 with well crystallized and high specific capacity is prepared by doping of Na and Li for Li and Na ion battery. The Na 0.6 Li 0.6 [Mn 0.72 Ni 0.18 Co 0.10 ]O 2 composite is easily synthesized by molten-salt method. The effects of annealing temperature, time, Na contents, and electrochemical performance are investigated. In XRD analysis, the substitution of Na-ion resulted in the P2–Na 2/3 MO 2 structure (Na 0.70 MO 2.05 ), which co-exists in the Na 0.6 Li 0.6 [Mn 0.72 Ni 0.18 Co 0.10 ]O 2 composites. The discharge capacities of cathode materials exhibited 284 mAh g −1 for Li and 237 mAh g −1 for Na ion battery with higher initial coulombic efficiency.

The corrosion behavior of Al 3 Ti-based intermetallic compounds was investigated under nuclear reactor normal operation conditions. The corrosion test was performed for Al–25Ti–10Cr and Al–21Ti–23Cr (at.%) in 633 K water and 673 K steam. The corroded surface was analyzed to identify the corrosion products. Both alloys showed a weight loss in 633 K water with no appreciable difference between the alloys. The corroded layer formed in water was shown to be the mixture of AlO(OH), TiO 2 , and Cr 2 O 3 . In 673 K steam, the corrosion behaviors of both alloys were similar with a small amount of weight gain. A thin, stable Al 2 O 3 layer was formed on the surface as result of oxidation in 673 K steam.

Following the Fukushima accident in March 2011, Korea Atomic Energy Research Institute (KAERI) has developed micro-cell fuel pellets for the purpose of reducing gas release, both stable and radioactive, not only under accident conditions, but during normal operation as well. The micro-cell pellet is composed of many cells covered with a ceramic or metallic wall material, whose function is either to lower or block the movement of gas atoms out of the cells, thereby reducing gas release out of the pellet. Based on both a Halden irradiation test of the micro-cell pellets up to 16 MWd/kgU and experience of fission gas release (FGR) in the conventional UO2 pellet, a simplified approach is proposed to estimate FGR in the micro-cell pellets when it is irradiated with a constant linear power of 30 kW/m up to 60 MWd/kgU under normal operating conditions. Considering the lack of measured data on diffusivity in the micro-cell pellets, sensitivity studies for FGR were performed for two cases: diffusivity in the micro-cell pellet is the same as and 10 times higher than in the UO2 pellet, respectively. Even for the 10 times higher diffusivity, FGR in the ceramic micro-cell pellet is at least comparable to or lower than in the UO2 pellet, and FGR in the metallic micro-cell pellet is clearly lower than in the UO2 pellet, due to its lower fuel temperature and the impact of its walls on gas movement. In real situations, the degree of reduction of FGR in the micro-cell pellets would be determined by the fraction of perfect walls and the magnitude of gas diffusivity. Until PIE results become available for the irradiated micro-cell pellets, the current approach provides only rough estimates and therefore its results need to be taken as preliminary.

A 3D laser coating technology using Cr powder was developed for Zr-based alloys considering parameters such as: the laser beam power, inert gas flow, cooling of Zr-based alloys, and Cr powder control. This technology was then applied to Zr cladding tube samples to study the effect of Cr coating on the high-temperature oxidation of Zr-based alloys in a steam environment of 1200 °C for 2000s. It was revealed that the oxide layer thickness formed on the Cr-coated tube surface was about 25-times lower than that formed on a Zircaloy-4 tube surface. In addition, both the ring compression and the tensile tests were performed to evaluate the adhesion properties of the Cr-coated sample. Although some cracks were formed on the Cr-coated layer, the Cr-coated layer had not peeled off after the two tests.

Ever since the Fukushima accident, accident tolerant fuel (ATF) has been widely studied, and a coating technology for a fuel cladding surface has been considered to decrease the high-temperature oxidation rate of zirconium-based alloy. In this study, arc ion plating technology was applied to coat a Cr layer on Zircaloy-4 as the oxidation-resistant coating layer for ATF cladding. A pure Cr phase was confirmed through an XRD analysis, and Cr film was also found to be homogeneous, dense, and uniform in the cross-sectional observation by SEM. Cr-coated Zircaloy-4 was tested in a high-temperature steam environment at 1473 K for 2000 s. The high-temperature oxidation tests showed that the Cr layer protected Zircaloy-4 from high-temperature oxidation in the steam. The stability of the Cr-coated layer deposited was maintained without peeling after a high-temperature oxidation test, and a Cr2O3 thin oxide layer was observed on the Cr coating layer. Furthermore, the TEM results showed that a Cr-Zr mixed layer was formed during the high-temperature oxidation tests. Based on the results obtained, the Cr film on Zircaloy-4 successfully acted as a corrosion protective layer, and arc ion plating was shown to be a corrosion protective coating system for ATF claddings.

Various oxides of Al 2 O 3 , ZrO 2 –Y 2 O 3 , and Ta 2 O 5 were coated on ferritic–martensitic steel for application as an environmental barrier layer. Sol–gel based coating was investigated to form the oxides by varying the coating parameters, such as the concentration of the precursors, the temperature of the curing, cycles of repeated runs, and additional heat-treatment. The obtained coatings revealed nano-sized granular structures. The surface morphologies were rough in alumina and zirconia, but appeared smooth in tantalum oxide. In the case of alumina and tantalum oxide, coated layers were mostly amorphous after pyrolysis at 750 °C. The crystalline phases were obtained after an additional heat-treatment at 950 °C. In the case of zirconia, a desirable oxide phase was formed when the samples were cured at 750 °C during the coating process. In addition to the heat-treatment after the coating, the repeated coatings were effective in crystallizing the coated layers and forming proper oxides.

The oxidation behaviour of the FeCrAl alloy under normal operating conditions of a light water reactor is reported. In a comparative study, Zr-based and FeCrAl alloy samples were corroded in pressurized 360 °C water for up to 500 days or oxidized at 1200 °C for 3000 s in a steam environment. Compared to Zr-based alloys, the FeCrAl alloy showed extremely low weight change after both tests. The microstructure of the oxides formed on FeCrAl alloys was characterized by various analysing techniques. It was concluded that depending on the test conditions, mostly different oxide phases of α-Fe 2 O 3 and α-Al 2 O 3 were formed.

SiC fiber-reinforced ceramic composites were fabricated using a polymer impregnation and pyrolysis process. To develop the low temperature process, the pyrolysis was conducted at 600 °C in air. Both a microstructural observation and a mechanical test were utilized for the evaluation of the impregnation. For the impregnation, two kinds of polycarbosilane having a different degree of cross-linking were used. The level of cross-linking affected the ceramic yield of the composites. The cross-linking under oxygen containing atmosphere resulted in a dense matrix and high density of filling. However, tight bonding between the matrix and fibers in the fully dense composite samples, which was obtained using a cross-linking agent of divinylbenzene, turned out to be deteriorative on the mechanical properties. The physical isolation of fibers from matrix phase in the composites was very important to attain a mechanical ductility. The brittle fracture was alleviated by introducing an interphase coating with metallic Zr. The combination of forming the dense matrix and interphase coating should be a necessary condition for the SiCf/SiC fiber-reinforce composite, and it is practicable by controlling the process parameters.

SiC plates were bonded using a laser beam scanning method. As an interlayer material, a Zr sheet 0.4 mm in thickness was used. Joining was performed at room temperature in air under atmospheric pressure. The interfacial microstructures along with their atomic compositions of the Zr/SiC reaction zone were analyzed. In the Zr side, diffused Si formed ZrSi 2 , Zr 5 Si 3 C x , and Zr 4 Si phases. Carbide transformation was observed (Zr + C → ZrC x ) near the Zr/SiC interface. No crystallographic orientation relationship was found in most of the grains; however, a boundary of ZrSi 2 [1 0 0] parallel to Zr 5 Si 3 C x [ 0 1 1 ¯ 1 ] was found in the observation. In the SiC side, penetrated Zr was found at up to 120 μm from the Zr/SiC interface. The diffused Zr formed a silicon-rich metastable phase of ZrSi y ( y > 3.3).

Oxidation of SiC in both air and water–vapor–rich environments was carried out at 1200 °C to examine the effects of different oxidation conditions on the early-stage oxidation behavior of SiC. Two different types of SiC oxidation behavior were found, passive or active, depending on the oxidation environment. All the samples possessed amorphous oxide layers, regardless of the oxidation environment. Three Si oxidation states (SiO, Si 2 O 3 , and SiO 2 ) were observed in this layer. The amorphous Si 2 O 3 state was dominant, and the ratio of the three different states changed with the test conditions.

Korea plans to test a Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) in ITER. The HCCR TBM adopts a four sub-module concept considering the fabricability and the transfer of irradiated TBM for post irradiation examination. Each sub-module has seven-layer breeding zone, including three neutron multiplier layers packed with beryllium pebbles, three lithium ceramic pebble bed packed tritium breeder layers, and a reflector layer packed with graphite pebbles. Based on this configuration, neutronic and electromagnetic calculations were performed and their results were applied for the conceptual design of HCCR TBM that considers manufacturing feasibility. Also, a design and safety analysis of HCCR Test Blanket System (TBS) was performed using integrated design tools modifying nuclear system codes for helium coolant and tritium behavior evaluation. The Advanced Reduced Activation Alloy (ARAA) is being developed as a structural material. A total of 73 candidate ARAA alloys were designed and their out-of-pile performance was evaluated. The graphite pebbles as the neutron reflector were fabricated by using mechanical machining and grounding method with the surface coated with SiC. The hydrogen permeation characteristics of structural materials were evaluated using the Hydrogen PERmeation (HYPER) facility. The recent design and R&D progress on these areas are addressed in this paper.

Blocks of tungsten and ferritic–martensitic steel (FMS) were joined without any interfacial defects or cracks. For the joining, two times of a hot isostatic pressing (HIP) were performed. The first HIP (900 °C, 100 MPa, 1.5 h) facilitates the diffusion bonding between W and FMS. The second HIP (750 °C, 70 MPa, 2 h) corresponds to a tempering process to retain the mechanical properties of the FMS. As an interlayer material, titanium foil that can mitigate the thermal expansion difference between W and FMS was used. In addition, a molybdenum foil was inserted to prevent an unwanted bonding of W to a canning material. The lateral cracks in W plates, which were usually observed in the case of a conventional HIP process, were not observed when the molybdenum separator was used. W/FMS joint mock-ups with a dimension of 50 mm × 50 mm × 32 mm (T) were successfully fabricated. The shear strength of the joints was 89 MPa on average.

The surface modification of engineering materials by laser beam scanning (LBS) allows the improvement of properties in terms of reduced wear, increased corrosion resistance, and better strength. In this study, the laser beam scan method was applied to produce an oxide dispersion strengthened (ODS) structure on a zirconium metal surface. A recrystallized Zircaloy-4 alloy sheet with a thickness of 2 mm, and Y2O3 particles of 10 μm were selected for ODS treatment using LBS. Through the LBS method, the Y2O3 particles were dispersed in the Zircaloy-4 sheet surface at a thickness of 0.4 mm, which was about 20% when compared to the initial sheet thickness. The mean size of the dispersive particles was 20 nm, and the yield strength of the ODS treated plate at 500°C was increased more than 65 % when compared to the initial state. This strength increase was caused by dispersive Y2O3 particles in the matrix and the martensite transformation of Zircaloy-4 matrix by the LBS.

The Fukushima accident has had a tremendous impact on Japan and the rest of the world in the areas of public health, economy, and nuclear energy policy. Thus, international consensus has been reach...

The ITER first wall (FW) includes beryllium armor tiles joined to a CuCrZr heat sink. The FWs are one of the critical components in an ITER machine with a surface heat flux of 4.7 MW/ ${\rm m}^{2}$ or above. The small-scale mockup shall be a part of the qualification tests and used to validate the performance of the dominant manufacturing technologies before the production of larger scale components, and this mockup shall be equipped with a hypervapotron heat sink and manufacturing processes developed for a semiprototype design. The small-scale mockup includes 48 beryllium armor tiles (12 mm $\times\,$ 12 mm) capable of withstanding the specified heat flux values. The tile thickness shall be 6 mm to minimize the beryllium surface temperature and evaporation under high thermal loads. The detailed fabrication process of semiprototype small-scale mockups was developed for a qualification test in Korea. For the CuCrZr and stainless steel, the canned materials are processed into an hot isostatic pressing (HIP) device. In the case of beryllium-to-CuCrZr joining, the HIP was conducted at 580 $^{\circ}{\rm C}$ and 100 MPa. For nondestructive tests of the fabricated semiprototypes, visual and dimension inspections were performed whenever needed during the fabrication process, and ultrasonic tests were performed using ultrasonic probes. Destructive tests for the qualification semiprototype were performed on a small-scale mockup, which was fabricated together with semiprototypes. The Korea heat load test facility using an electron beam system was constructed with an electron gun (maximum electric power of 800 kW) for a high heat flux application with a 300-kW high-voltage power supply and maximum accelerating voltage of 60 kV. This facility was operated to evaluate the performance test of plasma facing components. A cyclic heat flux test will be performed to evaluate the ITER qualification program.

To test the performance of nuclear fuels, coolant needs to be circulated through the test rig installed in the test loop. Because the pressure and temperature of the coolant is 15.5 MPa and respectively, coolant sealing is one of the most important processes in fabricating a nuclear fuel test rig. In particular, 15 instrumentation cables installed in a test rig pass through the pressure boundary, and brazing is generally applied as a sealing method. In this study, an induction brazing system has been developed using a high frequency induction heater including a vacuum chamber. For application in the nuclear field, BNi2 should be used as a paste, and optimal process variables for Ni brazing have been found by several case studies. The performance and soundness of the brazed components has been verified by a tensile test, cross section test, and sealing performance test.