Nozzles of fuel assembly play an important role in pressure water reactor (PWR) fuel assembly element. For a long time, ordinary processing technologies of nozzles of fuel assembly have the problems of difficult and complicated process, the low availability of material and long the development cycles of manufacturing. However, according to the study these issues can be well settled by using the additive manufacturing technology. This paper studies a nozzle of fuel assembly prepared by this additive manufacturing technology through slow-strain-rate tension (SSRT) test and microstructure observation experiment. The results of SSRT test show that yield strength of the nozzle of fuel assembly is about 401.5MPa, the extensional rigidity is about 673.5MPa and the ductility is about 45.7%. And the SEM fracture results of the SSRT sample indicate that the fracture microstructure contains a large number of dimples, and the way of fracture belongs to plastic. And the metallographic observation consequences manifest that the microstructure of nozzle of fuel assembly prepared by the additive manufacturing technology is composite tissue of both austenite and ferrite, and the grains are settled along the way of laser scanning and there are isometric with some kind of direction. This metallographic microstructure is different from the traditional morphology of the free carbide distributed in the matrix. The dual phase microstructure of austenite and ferrite can improve the mechanical properties of the matrix effectively, and avoid the free carbides which may lead to matrix fragmentation in the tensile deformation process. Moreover, the laser power could affect the microstructure and properties of nozzles of fuel assembly observably, and the high laser power could bring about the ablation of metal. Through the analysis of mechanical properties and microstructure, we have made it possible to make the laser additive manufacturing technology to be used for the fuel assembly nozzle preparation in the nuclear power area. This work not only presents the advantages of the laser additive manufacturing technology in the fuel element processing area of the nuclear power station, but also broadens the application range of the laser additive manufacturing technology. What’s more we provide the new thoughts for the fast and effective preparation of the fuel element especially for the fuel assembly nozzle in the nuclear power station.

Up to now, two kinds of filler metal with or without nickel element for submerged arc welding have been largely used in the reactor pressure vessel (RPV) manufacturing. In order to study the effect of nickel element on weld metal properties of SA-508 Gr.3 Cl.1, submerged arc welding material with nickel (AWS classification F8P4-EGN-F2N, F2 for short) and welding material without nickel (F8P4-EA3N-A3N, A3 for short) were used; and conventional mechanical properties, low-cycle fatigue test, and proton irradiation analysis of the two weld metals were studied. Results show that the mechanical properties of the two different weld metals are similar, except that the Charpy V-notch impact property of the weld metal with nickel is better than that without nickel; the micro-structures of F2 and A3 weld metals are both composed of ferrite base and granular bainite, but the columnar grain size of F2 weld metal is smaller relatively, which results in better impact property. In addition, the irradiated A3 weld metal has fewer dislocation loops than the irradiated F2 weld metal after the same proton irradiation dose; the irradiated weld metals both have higher micro-Vickers hardness than before.

In order to gain high strength, fine grain size, stronger anti-corrosion property, and especially low permeability, the material 022Cr19Ni10N was chosen to manufacture the Rod Travel Housing Forging (RTHF) for Control Rod Drive Mechanism (CRDM). But, cracks were found in some forgings failing to meet the requirements of ultrasonic testing (UT). The causes of the forging cracks of this austenitic stainless steel forging were investigated by means of metallography, scanning electron microscopy (SEM) and other experimental methodology. The results indicated that the second δ-ferrite phase leads to the forging cracks between γ-δ interface during the low temperature forging process, and finally leads to the forging failure. It’s found that the cracks are distributing along the stripe δ-ferrite, and almost distributing in the same area as the large size δ-ferrite by metallography & SEM microstructure observation. The δ-ferrite is firstly found in the electroslag ingot, and in which, the distribution and size is different from the case to the core. The largest size δ-ferrite is around the core area, and this characteristic passes on to the final forging microstructure, although the shape, quantity & distribution of the δ-ferrite changed during the manufacturing process. Most forging cracks were found around the core area of the forging by UT examination. In the final forging process, when the forging temperature drops to 750∼850°C, the δ-ferrite have been forged to stripe shape and hundreds-micron size while the plasticity of the austenite reduce. What’s more, there are large hot plasticity differences between the δ-ferrite and the austenite, so the forging cracks initiate between γ-δ interface and extend to the area around to be a long crack in the low temperature forging process. In order to avoid the forging cracks in the Rod Travel Housing Forging, it’s necessary to reduce the content of δ-ferrite or improve the final forging temperature. Improving the final forging temperature, to guarantee the plasticity of the δ-ferrite and austenite, is another process to reduce the cracks. But while the temperature improves, the grain size grows rapidly, and may form mixed structure. So the most effective mean to reduce the content of δ-ferrite is to redesign the chemical components, mainly by increasing the nitrogen content from 0.06 (wt, %) to 0.12(wt, %), which makes the low temperature forging process for fine grain size possible. In the high-nitrogen-content forging, the δ-ferrite distributed sporadically and no δ-ferrite strip is found. By increasing the austenite forming elements (especially nitrogen), the cracks during low temperature forging process are avoided. What’s more, owning to the optimization of chemical compositions and manufacturing processes, the Rod Travel Housing Forging got fine grain size, low relative permeability, and good comprehensive mechanical properties with the ultimate tensile strength up to 570MPa.

The complex microstructures of stainless steel (SS) cladding on low alloy steel (LAS) joint and the corrosion behavior in high temperature water environments were examined. Scanning electron microscopy, focused ion beam, transmission electron microscope and Raman spectroscopy were used. The heat-affected zone (HAZ) in the low alloy steel was mainly comprised of pearlite and ferrite, while the HAZ in the stainless steel was mainly comprised of austenite and ferrite. The HAZ in the low alloy steel was divided into overheated crystal region, complete recrystallization region and incompletely recrystallization region. A decarburization zone in the low alloy steel side and a carbon-enriched zone in the stainless steel side were identified. M 23 C 6 and M 7 C 3 precipitation were observed mainly in carbon-enriched zone. The surfaces of the weld after corrosion in simulated PWR primary water environment at 290 °C were significantly affected by dissolved oxygen. In aerated solution, the oxide film on A508III steel was mainly γ-Fe 2 O 3 , which becomes spinel oxide on the 309L/308L cladding. In deaerated solution, the oxide film was mainly Fe 3 O 4 on A508III steel, which was spinel oxide on the 309L/308L cladding.

The valve stem used in the main steam system of nuclear power plant is usually 17-4PH martensitic stainless steel. When it served in 300 C° for a long time, the thermal aging embrittlement of valve stem will be significant, with the performance of the ductile brittle transition temperature (DBTT) and the hardness increased, the upper stage energy (USE) decreased. It will seriously affect the safety and economic operation of nuclear power plant (NPP). It is important to study the thermal aging effect of the 17-4PH steel for safe operation of nuclear power plant. In this work, Three-Dimensional Atom Probe (3DAP), Energy Dispersive X-Ray Spectroscopy (EDX), Scanning Electron Microscope (SEM) and Optical Microscope (OM) are used to analyze the element distribution in 17-4PH steel. The results show that lath martensite will grow significantly under high temperature for a long time. More δ-ferrite will be found between lath martensite, and some carbide aggregates at its interface. In addition, the number density of Cu clusters in the17-4PH steel is increased. It is found that Ni and Mn have obvious segregation with the Cu cluster.

Cast austenitic stainless steels (CASS) possess excellent corrosion resistance and mechanical properties and are used alongside with wrought stainless steels (SS) in light water reactors for primary pressure boundaries and reactor core internal components. In contrast to the fully austenitic microstructure of wrought SS, CASS alloys consist of a dual-phase microstructure of delta ferrite and austenite. The delta ferrite is critical for the service performance since it improves the strength, weldability, corrosion resistance, and soundness of CASS alloys. On the other hand, the delta ferrite is also vulnerable to embrittlement when exposed to reactor service temperatures and fast neutron irradiations. In this study, the combined effect of thermal aging and neutron irradiation on the degradation of CASS alloys was investigated. Neutron-irradiated CASS specimens with and without prior thermal aging were tested in simulated light water reactor environments for crack growth rate and fracture toughness. Miniature compact-tension specimens of CF-3 and CF-8 alloys were tested to evaluate the extent of embrittlement resulting from thermal aging and neutron irradiation. The materials used are static casts containing more than 23% delta ferrite. Some specimens were thermally aged at 400 °C for 10,000 hours prior to the neutron irradiation to simulate thermal aging embrittlement. Both the unaged and aged specimens were irradiated at ∼320°C to a low displacement damage dose of 0.08 dpa. Crack growth rate and fracture toughness J-integral resistance curve tests were carried out on the irradiated and unirradiated control samples in simulated light water reactor environments with low corrosion potentials. While no elevated crack propagation rates were detected in the test environments, significant reductions in fracture toughness were observed after either thermal aging or neutron irradiation. The loss of fracture toughness due to neutron irradiation seemed more evident in the samples without prior thermal aging. Transmission electron microscope (TEM) examination was carried out on the thermally aged and neutron irradiated specimens. The result showed that both neutron irradiation and thermal aging can induce significant changes in the delta ferrite. A high density of G-phase precipitates was observed with TEM in the thermally aged specimens, consistent with previous results. Similar precipitate microstructures were also observed in the neutron-irradiated specimens with or without prior thermal aging. A more extensive precipitate microstructure can be seen in the samples subjected to both thermal aging and neutron irradiation. The similar precipitate microstructures resulting from thermal aging and neutron irradiation are consistent with the fracture toughness results, suggesting a common microstructural origin of the observed embrittlement after thermal aging and neutron irradiation.

Lead cooled fast reactor (LFR) is one of the next generation of nuclear plants. Tubes in LFR coolant system requires high temperature strength and Lead-Bismuth eutectic (LBE) corrosion resistance and. 9Cr1Mo steel - 12Cr2Si steel bimetallic billet that satisfies above properties has been produced by dual centrifugal casting. Bimetallic billet was normalized for 1 hour at 1050°C and air cooled. After normalizing, tempering was performed for 1 hour at 750°C and air cooled. Mechanical properties of 9Cr1Mo material satisfied ASME Code Section II Part A SA-426 CP91 standard. 9Cr1Mo steel has tempered martensite structure and 12Cr2Si steel has tempered martensite-ferrite mixed structure. The width of mixed zone was about 1000μm. The LBE corrosion resistance was indirectly verified by observation of oxide layer. After heat treatment, hardness of 9Cr1Mo steel was about 240Hv and 12Cr2Si steel was about 200Hv. As a result of high temperature tensile test, the suitable hot work temperature range was 1000°C and 1100°C. Thermo-Calc calculation result shows that the austenite phase fraction was major factor to the workability of the billet.

This paper is to compare the sections on materials between ASME-BVC and RCC-M code. This paper states the layout structure and the content about NPP materials specified in the two codes and shows the comparison results. This paper emphasizes particularly on the basic differences between two codes; ASME code focuses on fracture toughness requirements for pressure-retaining ferrite materials and RCC-M code focuses on Susceptibility to intergranular corrosion for austenitic and austenitic-ferrite stainless steel.

Nuclear scientists and engineers should consider adopting a more operational approach for the purpose of selecting their future materials. For each type of nuclear power generating reactor, for each coolant (water, helium or liquid metal), the next generation of specialists and decision-makers will need to choose and optimise the iron or nickel alloys, steels, ODS (oxide dispersed strengthened steels) and ceramics that are going to be used. It may well be considered that either each reactor type has its own, specific materials, or, in a complementary manner, that the efforts for improvements should be shared. At high temperatures, as found on fuel-cladding liners, heat exchangers or even tubes or tube liners, different types of steels and alloys may be envisaged. It is considered that austenitic steels provide a better creep resistance at high temperature but they must be stabilized by nickel, thereby becoming more expensive. Ferrite steels could be better as far as swelling, mechanical strength and thermal behaviour are concerned. To withstand corrosion, chromium or aluminium, ODS steels could turn out to be good solutions, if they can comply with stringent criteria. Concerning heat exchangers, choices must be made between iron and nickel alloys, according to proposed operating conditions. In the case of sodium-cooled rapid neutron reactors (RNRs), ferritic-martensitic alloys with 9%–12% chromium or chromium ODS steels could prove suitable, especially if we judge by their specific mechanical behaviour, up to at least 700°C. Nevertheless, behaviour of these steels — with respect to ageing, anisotropy, radiation induced segregation, radiation induced precipitation, reduction of activation products and welding — needs be better understood and qualified. Sodium heat exchanger materials should be carefully chosen since they have to withstand corrosion arising from the primary flow and also from the secondary or tertiary flow (either sodium or molten salts, gas or water); therefore, experimental loops are necessary to gain improved understanding and assessment of the designs envisioned. One way to improve alloys is through thermal, mechanical treatments or by surface treatments. A better way could, however, be to improve the nanostructure and mesostructure of the materials chosen at the drawing-board stage, for instance by nano-size cluster dispersion and grain size controls; experimental tests, microscope and spectroscope observations, multi-scale modelling and thermodynamics computing could also help calibrate and implement these improvements. Large, experimental databases and codes will be the keystone to defining more operational knowledge bases that will then allow us to determine terms of reference for the new materials. Failing this, time will be running out — within the next twenty years — to design and develop nuclear prototypes consistent with the criteria laid down for “Generation IV” reactors.

The author has investigated the characteristics of boron co-deposition with crud experienced in AOA and iron ferrite deposition in CDA. Corrosion product deposits found in cores with appreciable AOA have been reported in mostly nickel-based (as NiO or elemental nickel) as opposed to nickel ferrite deposits common to non-boiling cores. Significant quantities of meta-ZrO 2 and nickel iron oxyborates (bonaccordite), notably Ni 2 FeBO 5 have also been found in deposits on cores with AOA. On the basis of this general characterization information, the author has constructed a potential-pH diagram of Ni 2 FeB(OH) 10 , which is a hydrated state of FeNi 2 (BO 3 )O 2 as summarized in this paper. Although preliminary, the estimated E-pH diagram suggests some interesting observation, including: growth of bonaccordite “needles” on the clad is associated with a local anodic electrochemical reaction necessary to remove excess electrons from the system to a cathode. During the AOA cycle, the concentration of nickel and iron ions must have been unusually high as they should be for a significant amount of crud deposits. The author thinks such an acceleration of the anodic dissolution of metal cations is due to the effect of the long cell action corrosion mechanism. As early as 1949, an Italian scientist Petracchi demonstrated that electrochemical effects significantly influence the erosion rate. He constructed a flow nozzle with specimens kept under external electrical potential. Upon inducing as low as 0.1 mA/cm 2 of the positive current, the erosion rates were reported drastically increased. No erosion was observed by reversing the polarity of the potential. As discussed in a companion paper also presented at this conference [1], the author discusses various mechanisms (electrochemical cell configurations) that induce potential differences, including those differences in ionic concentration, aeration, temperature, flow velocity, radiation and corrosion potentials. In this paper, the author discusses how these potential differences are related to the AOA/CDA issues in PWR/VVER plants. The author is calling for further verification experiments regarding this corrosion mechanism as a joint international project.