[en] In 1987, the decision was taken to use in BN-350 and BN-600 reactors steels ChS68 cw and EP450 as standard materials for fuel cladding and FA wrappers, respectively. The application of EP-450 steel as 96 x 2 mm wrappers combined with CW ChS68 as 6.9 x 0.4 mm fuel claddings has reliably ensured the failure free operation of BN-600 fuel assemblies to a burnup of 11.2 at.% at a damage dose of ∼82 dpa. After the HT (HT is the normalization at 1050^oC and tempering at 720 deg. C, 1 h) the microstructure of EP450 wrappers consists of ferrite and tempered martensite grains with approximately 1:1 ratio. The microstructure is characterized by presence of rounded MC and M₂₃C₆ precipitates in the grain interior and along grain boundaries, with particle diameters ranging from 0.05-0.2 μm. The lath boundaries in grains of the tempered martensite contain some smaller (up to 0.1 μm) blocky precipitates of M₂₃C₆. Dislocations in the ferrite are uniformly distributed and their density is 2 x 10¹⁴ m^-2. The dislocation density in the tempered martensite is 1 x 10¹⁵ m^-2 and they are non-uniformly distributed intragranularly, forming low angle boundaries. Since the swelling behaviour of ChS68 steels limits the fuel burnup, it is desired that the cladding also be EP-450 but modified as an ODS option so that it retains sufficient strength at high temperatures. The method chosen is mechanically alloying. - Production of steel powder having the matrix composition (EP450) by the centrifugal atomization of the molten mass from a revolving crucible in high purity helium; - Mechanical alloying of the resultant powder with nano-particles (40-80 nm) of Y₂O₃ in vibrating high-energy attritor; - Vibro-filling the steel cans with the received powder blend, followed by degassing and sealing of cans; - Hot extrusion at ∼1150 deg. C of cans containing a powder blend to produce a hot-extruded bar at the drawing of no less than 10 and its subsequent working; - The quality of the final product is influenced by the production method of the yittria particles. In the initial production yittria of industrial production was used. The properties of the second production were improved by the use of yittria particles produced by special technology. Examples of particles are shown; - The different microstructures produced in bars are shown; - The microstructure of a finished tube are shown. The structure of the produced tubes consists of equiaxed subgrain sizes ranging from 0.1-3 μm. Uniformly distributed oxides are observable within grains and subgrains. The size distributions of the oxides are illustrated. The mean size of the oxide particles is about 7 mm, their concentration is ∼10¹⁶ cm^-3. The ODS samples are under irradiation in BN-600 reactor within two materials FAs up to the damage dose of ∼140 dpa beginning on 2010. (author)

[en] ODS ferritic steels, which is one of the innovative technologies in FaCT project, is the most promising material for fuel pin cladding tubes in commercialized SFR cores. JAEA has been developing ODS steels for the tubes since 1987. In design studies for large and medium scale SFR cores, discharge average burnup is targeted to be 150 GWd.t^-1 and peak burnup to be 250 GWd.t^-1. The tubes need to endure heavy displacement damages to 250 dpa. In order to improve thermal efficiency of SFR plants, the maximum outlet coolant temperature at the reactor vessel is set to 823 K, and the maximum (hot-spot) temperature of the tubes corresponds to be 973 K. Internal pressure increases with burnup due to noble gas accumulation, and approaches to be about 12 MPa in the highest burnup fuel pin after nine years of service. Mechanical properties of ODS steels are targeted to be over 300 MPa at 973 K for UTS, over 1% for uniform elongation (UE) and 120 MPa at 973 K for 10000 h for creep rupture strength against internal pressures. Historically, modified type SUS316 austenitic stainless steel named as PNC316 has been applied for fuel pin cladding tube, spacer wire and hexagonal duct in driver fuel pins or subassemblies in the experimental fast reactor JOYO MK-II core and in the PFBR MONJU. In conventional austenitic stainless steels such as PNC316, detrimental void swelling will emerge over 120 dpa. Since body centred cubic structure is more resistant against displacement damages than face centred cubic structure, ferritic steels are indispensable for the tubes. PH high strength ferritic-martensitic steel named as PNC-FMS has been developed since 1983 in JAEA. However, PH ferritic steels such as PNC-FMS will rapidly lose their strength over 923 K. In contrast, dispersion hardening by thermally stable oxide particles can be effective even over 923 K. (author)

[en] In the Republic of Korea, R and D activities on a sodium cooled fast reactor (SFR) were initiated in 2007 and focused on metallic fuel and FM steel as fuel assembly structural materials. As cladding tubes, high performance 9Cr-2W steels (Korean alloy) capable of reaching 250 dpa at 650^oC are being developed. The cladding tube will have an inner liner of V or Cr to minimize the chemical interaction with a metallic fuel. Suitable coating techniques for these candidate barrier materials for the inner surface of a cladding are also being studied. Modified 9Cr-1Mo steels are being considered as wrapper materials in a SFR. The draft road map for a SFR fuel cladding development is shown. On the basis of a grade 92 (9Cr-0.5Mo-1.8W-VNb), new alloy designs and an evaluation of the out of pile performance of these new alloys are scheduled until 2011. Large scale manufacturing of the FM steel cladding tubes will be initiated in 2011, and in-pile tests of these cladding tubes will be finished by 2022. R and D activities on ODS-FM steel are expected to start in 2010, followed by the same progress of the FM steel development programme. Ten kinds of batch 0 alloys were designed, and their nominal compositions are given. These alloy designs were mainly focused on the effects of B, C, Nb and Ta on the mechanical properties of cladding tubes. The alloy ingots, with 30 kg scale each, were prepared by a vacuum induction melting process at a Korean steel company, POSCO. The steel ingots were hot-rolled to a 15 mm thickness after a preheating at 1150 deg. C for 2 h, followed by a normalizing at 1050^oC for 1 h and a tempering at 750 deg. C for 2 h. The tensile and creep test results at 650 deg. for the batch 0 alloys and the reference steels such as grade 92 and HT-9 are shown. The results could be summarized in that some of the new alloys exhibited superior mechanical properties compared with the reference steels, and the additions of 0.17% B, 0.07% C, 0.13% Nb and 0.05% Nb-0.14% Ta led to enhanced tensile and creep properties. Based on these results of the batch 0 alloys, ten kinds of batch 1 alloys were designed. These alloy designs were focused on, not only optimized concentrations of V, Nb and Ta, but also the effects of additional alloying elements such as Ti, Zr, Pd, Pt and Nd. The tensile and creep test results at 650 deg. C of the batch 1 alloys and the reference alloys are shown. The B107 and B109 alloys were found to be superior in their tensile and creep properties. On the basis of these experimental results, batch 2 alloys were also designed, and their out of pile performances are being evaluated. The schematic flow of the fabrication process for the FM steels in a form of a plate with a 1 mm thickness is shown. The hot-rolled and normalized plates were tempered at 550 deg. C and 750 deg. C, respectively. The plate tempered at 750 deg. C was cold-rolled to a 1 mm plate with a reduction ratio of 75%, followed by an annealing at 700^oC for 30 min. The plate tempered at 550 deg. C was cold-rolled to a 1 mm thickness, and then subjected to a HT at 750 deg. C for 30 min. Twice cold rolling was conducted with an intermediate HT at 750 deg. C for 10 min after the first stage of a cold rolling. Three cold rolling passes with intermediate HTs were also performed. Images of extraction replicas for precipitates in these plates. M₂₃C₆ and MX precipitates were found in all specimens, and fine, uniform precipitates were observed in the CA3 specimen. These results closely correlated with the strain energy stored by a cold rolling. The tensile test results at 650 deg. C of these plates are shown. The CA3 specimen was found to be superior in its tensile strengths, being mainly attributed to the finely distributed precipitates.

[en] In addition to a grade 92 FM alloy currently under development, China is proceeding to develop a 13Cr-ODS alloy. The flow sheet is shown. Results of ion beam irradiation of ODS steels are also given. Mechanical properties and ion beam irradiation results of unirradiated 13Cr-ODS alloy: Several samples were prepared for different tests. The compositions of the sample are shown. Tensile tests were conducted and creep tests were performed. Ion beam irradiation tests were done and compared with austenitic steels. The tensile properties are listed and the creep test results are shown. The low creep strength of S2 was thought to be attributed to the formation of χphase in the matrix caused by the high content of Ti. The Ion irradiation simulations indicated that ODS alloys are superior to austenitic alloys in the resistance to irradiation swelling and segregation. (author)

[en] One of the IAEA's statutory objectives is to 'seek to accelerate and enlarge the contribution of atomic energy to peace, health and prosperity throughout the world'. One way this objective is achieved is through the publication of a range of technical series. Two of these are the IAEA Nuclear Energy Series and the IAEA Safety Standards Series. According to Article III.A.6 of the IAEA Statute, the safety standards establish 'standards of safety for protection of health and minimization of danger to life and property.' The safety standards include the Safety Fundamentals, Safety Requirements and Safety Guides. These standards are written primarily in a regulatory style, and are binding on the IAEA for its own programmes. The principal users are the regulatory bodies in Member States and other national authorities. The IAEA Nuclear Energy Series comprises reports designed to encourage and assist R and D on, and application of, nuclear energy for peaceful uses. This includes practical examples to be used by owners and operators of utilities in Member States, implementing organizations, academia, and government officials, among others. This information is presented in guides, reports on technology status and advances, and best practices for peaceful uses of nuclear energy based on inputs from international experts. The IAEA Nuclear Energy Series complements the IAEA Safety Standards Series. This report summarizes the results of two IAEA sponsored technical meetings, conducted in 2008 and 2011, and associated consultancies directed toward a common set of goals. These technical meetings and their venues were as follows: - IAEA Technical Meeting on Status and Trends of Stainless Steel Cladding and Fuel Assembly Materials and Components for LMR, held in Hyderabad, India, 2-4 July 2008; - IAEA Technical Meeting on Design, Manufacturing and Irradiation Behaviour of Fast Reactor Fuels, held in Obninsk, Russian Federation, 30 May-3 June 2011. One of the main objectives of these meetings was to 'share and exchange information on stainless steel structural materials for liquid metal cooled fast reactor fuel assemblies', producing a final report that would: - Identify the different varieties of austenitic, nickel based, ferritic-martensitic (FM) and oxide dispersion strengthened (ODS) steels having demonstrated success or potential improved performance as structural components of fast reactor fuel assemblies, with particular emphasis on fuel cladding; - Summarize the manufacturing processes of liquid metal fast reactor (LMFR) fuel cladding tubes, rods for end plugs, sheets, wrappers, wires, etc., starting from ingot preparation; - Summarize the irradiation behaviour of these steels in fast reactor service; - Focus in particular on the ODS variants of ferritic and FM steels as the path forward to achieving higher burnup of fuel in fast reactors. One major conclusion from this activity is that there is a need to develop a strong international capability to explore the radiation resistance, especially to void swelling, of ODS variants of ferritic and FM alloys using ion simulation techniques as surrogates for currently unavailable high flux fast reactors. This report represents the distillation and summary of the results of numerous past and ongoing activities conducted by IAEA Member States.