Takeji Kaito

Oxide Dispersion Strengthened (ODS) Fuel Pins Fabrication for BOR-60 Irradiation Test

In order to confirm and demonstrate Oxide Dispersion Strengthened (ODS) cladding fuel pin survival to high burn-up condition, an irradiation test in BOR-60 was conducted within the framework of JNC-Russian RIAR collaborative work. Two types of 9Cr martensitic and 12Cr ferritic ODS steel claddings were manufactured in JNC, and the upper end-plugs were welded by the pressurized resistance method. In the RIAR fuel production facility, the MOX and UO2 granulated fuels as well as uranium metal getter particles were vibro-packed into the ODS steel cladding. The lower end-plug was welded by the TIG end-face method. The inspection and quality control of the fabricated ODS fuel pins were done through X-ray analysis, gamma scanning, and leak testing etc. and it was confirmed the fuel pins satisfied the BOR-60 requirements. Those fuel pins were loaded into two dismountable experimental assemblies, and irradiation was started in the BOR-60 from June 26, 2003. The cladding middle wall temperature was kept within 700±20°C and 650±20°C. Currently, the attained burnup is 5 at.%. Toward the target burnup of 15 at.%, interim examinations will be made at intervals of 5 at.%.

Fuel Pin Irradiation Test at up to 5 at% Burnup in BOR-60 for Oxide-Dispersion-Strengthened Ferritic Steel Claddings

Oxide-dispersion-strengthened (ODS) ferritic steel has excellent radiation resistance and high temperature strength. In the fast reactor cycle technology development (FaCT) project carried out in the Japan Atomic Energy Agency (JAEA), ODS ferritic steel was identified as the only cladding tube material that promises achievements of a higher burnup of fuels and a higher coolant outlet temperature. The following two types of ODS claddings have been developed in JAEA: a 9Cr-ODS cladding with a basic chemical composition of Fe-0.13C-9Cr-2W-0.2Ti-0.35Y2O3 in mass% for obtaining a higher radiation resistance and a 12Cr-ODS cladding with a basic chemical composition of Fe-0.03C12Cr-2W-0.3Ti-0.23Y2O3 in mass% for obtaining a higher corrosion resistance. In order to confirm the irradiation behavior of ODS steels and thus judge their applicability to fuel claddings, ODS cladding fuel pins have been irradiated in BOR-60 of Research Institute of Atomic Reactors (RIAR) in Russia. These fuel pins contain vibro-packed MOX fuels with uranium metal getter particles that induce a lower oxygen potential inside the pins. This irradiation test was conducted as the JAEA-RIAR collaborative work. Results of the first irradiation test are reported in this paper with the focus on fuel and cladding chemical interaction (FCCI), as shown by postirradiation examination data.

In-reactor creep rupture properties of 20% CW modified 316 stainless steel

The in-reactor creep rupture tests of 20% cold worked modified 316 stainless steel were conducted in the temperature range from 878 to 1023 K using MOTA of FFTF, and were compared with the out-of-reactor tests. In-reactor creep rupture, lives become shorter than those of the out-of-reactor tests. In-reactor creep strain rate was significantly accelerated, and sufficient ductility appears to be maintained even under the irradiation. Considering 0.2% proof strength after neutron irradiation, sodium exposure or aging, the degraded rupture lives of in-reactor creep are ascribed to the enhanced dislocation recovery due to the neutron irradiation as well as to the solute elements dissolution into sodium under the sodium exposure environment.

Development of Two-Step Softening Heat Treatment for Manufacturing 12Cr-ODS Ferritic Steel Tubes

The 12Cr-oxide dispersion strengthened (ODS) ferritic steel is being developed as prospective cladding materials for attaining higher burnup and higher coolant outlet temperature in advanced fast reactor. Ductility and high temperature strength in the hoop direction can be attained by recrystallization in the manufactured tube of 12Cr-ODS ferritic steels. In this study, it was found that once recrystallization has taken place during manufacturing process with four passes cold rolling and intermediate heat treatment, the recrystallized condition can not be repeated at the final heat treatment. Thus, the unique two-step softening heat treatment was developed as the intermediate heat treatment; the first step heat treatment relieves the strain energy accumulated by cold rolling to raise the recrystallization temperature, and the subsequent second step heat treatment at the temperature, which is higher than the recrystallization temperature before relieving the work strain energy, accelerates the hardness reduction without recrystallization. This process leads to easy cold rolling and complete recrystallization at the final heat treatment.

ODS cladding fuel pins irradiation tests using the BOR-60 reactor

In order to confirm the irradiation behavior of ODS steels and thus judge their applicability to fuel claddings, fuel pin irradiation tests using 9Cr and 12Cr-ODS claddings developed by JAEA were conducted to burn-up of 11.9 at% and neutron dose of 51 dpa in the BOR-60. Superior properties of the ODS claddings concerning FCCI, dimensional stability under irradiation and so on were confirmed and indicated good application prospects for high burn-up fuel. On the other hand, anomalous irradiation behaviors, fuel pin failure and the microstructure change containing coarse and irregular precipitates, occurred in a part of the fuel pin with 9Cr-ODS cladding. This paper describes evaluation of the obtained irradiation data and the investigation results into the cause of the anomalous irradiation behaviors.

Charpy impact property related to {100} cleavage fracture in 15CrODS steel

Abstract Based on the former results that delamination along {100} cleavage fracture leads to the improved absorbed energy, {100} cleavage plane was formed in parallel to the rolled plate by recrystallisation and cold roll processing in 15CrODS steel plate. Charpy impact properties were investigated in terms of notch direction against {100} cleavage plane. Notch with normal direction against {100} cleavage plane induced delamination fracture, but provided similar absorbed energy to that of conventional extruded bar. The transverse notch direction against {100} cleavage plane served higher absorbed energy at low temperature. Higher upper shelf energy and lower DBTT were attained in conventional hot extruded bar, which contains any crystallographic planes rotated around <110> axis at the transverse direction. The delamination fracture leading to high absorbed energy is significantly affected by a crystallographic morphology of the {100} cleavage plane perpendicular to transverse plane.

Water Corrosion Resistance of ODS Ferritic-Martensitic Steel Tubes

Oxide dispersion strengthened (ODS) ferritic-martensitic steels have superior radiation resistance; it is possible to achieve a service temperature of up to around 973K because of their superior creep strength. These advantages of ODS steels facilitate their application to long-life cladding tubes in advanced fast reactor fuel elements. In addition to neutron radiation resistance, sufficient general corrosion resistance to maintain the strength of the cladding, and the stress corrosion cracking (SCC) resistance for spent-fuel-pool cooling systems and high-temperature oxidation for the fuel-clad chemical interaction (FCCI) of ODS ferritic steel are required. Although the addition of Cr to ODS is effective in preventing water corrosion and high-temperature oxidation, an excessively high amount of Cr leads to embrittlement due to the formation of a Cr-rich 0 precipitate. The Cr content in 9Cr-ODS martensite and 12Cr-ODS ferrite, the ODS steels developed by the Japan Atomic Energy Agency (JAEA), is controlled. In a previous paper, it has been demonstrated that the resistances of 9Crand 12CrODS ferritic-martensitic steels for high-temperature oxidation are superior to those of conventional 12Cr ferritic steel. However, the water corrosion data of ODS ferritic-martensitic steels are very limited. In this study, a water corrosion test was conducted on ODS steels in consideration of the spentfuel-pool cooling condition, and the results were compared with those of conventional austenitic stainless steel and ferritic-martensitic stainless steel.

Development of Manufacturing Process of PNC-FMS Wrapper Tube with SUS316 Short Joint

When a ferritic-martensitic stainless steel (PNC-FMS) wrapper tube having far greater swelling resistance against neutron irradiation is applied in the JOYO or MONJU reactor, it becomes necessary to weld it with SUS316 austenitic stainless steel (entrance nozzle and handling head). Such welding between PNC-FMS and SUS316 causes the delta (δ) ferrite formation at heat-affected zone, which leads to significant toughness degradation. In addition, bending of wrapper tube caused by their differential thermal expansion should be straightened. For preventing those problems, manufacturing process of the complex wrapper tube was developed. This process involves TIG-welding with SUS316 short pipe joints in 50mm length to both ends of a PNC-FMS round tube, and then performing the drawing and normalizing and tempering. Normalizing induces complete disappearance of the δ ferrite in the course of wrapper tube manufacturing. The mechanical properties of PNC-FMS/SUS316 welded zone were confirmed to be equivalent to those of the base metal even after thermal aging.

Characterization of recrystallization of 12Cr and 15Cr ODS ferritic steels

The recrystallization behavior of 12Cr and 15Cr oxide dispersion-strengthened (ODS) ferritic steels, which are the promising candidate materials for long-life core materials of the advanced fast breeder reactors, was investigated in terms of an intermediate softening heat treatment. It was clarified that keeping recovery structure at the intermediate heat treatment is indispensable for producing recrystallized structure at the final heat treatment. Prevention of repeating recrystallization is owing to the stable {100} 〈110〉 texture formation with less stored strain energy by the cold-rolling of the recrystallized structure. The two-step softening process was proposed to suppress the recrystallization and obtain adequate hardness reduction at the intermediate heat treatment. This process is effective for producing a stable recrystallized structure at the final heat treatment of the manufacturing process of ODS ferritic steel cladding.

Investigation of the cause of peculiar irradiation behavior of 9Cr-ODS steel in BOR-60 irradiation tests

Four experimental fuel assemblies (EFAs) containing 9Cr-ODS steel cladding fuel pins were previously irradiated in the BOR-60 to demonstrate the in-reactor performance of 9Cr-ODS steel for use as fuel cladding tubes. One of the EFAs achieved the best data, a peak burn-up of 11.9at% and a neutron dose of 51 dpa, without any microstructure instability or any fuel pin rupture. On the other hand, in another EFA (peak burn-up, 10.5at%; peak neutron dose, 44 dpa), peculiar irradiation behaviors, such as microstructure instability and fuel pin rupture, occurred. Investigations of the cause of these peculiar irradiation behaviors were carried out. The detection sensitivity in an ultrasonic inspection test was shown to be low for the metallic Cr and metallic Fe inclusions. The peculiar microstructure change reappeared with high-temperature thermal-aging of the 9Cr-ODS steel containing metallic Cr inclusions. The strength and ductility of the defective part containing metallic Cr inclusions were appreciably lower than those of a standard part without the inclusions. The combined effects of matrix Cr heterogeneity (presence of metallic Cr inclusions) and high-temperature irradiation were concluded to be the main cause of the peculiar microstructure change in 9Cr-ODS steel cladding tubes in the BOR-60 irradiation tests. They contributed to the fuel pin rupture.