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[en] A quantitative correlation of the swelling rate versus swelling is presented. Swelling data of 20% cold-worked 316 stainless steels were analyzed using the power law swelling equation. The prolonged transient region with keeping the suppressed swelling rate was clearly demonstrated for the improved 316 stainless steels like PNC316 and 316Ti. Rate theory analysed lead to the role of precipitates as point defect sinks for retardation of the void growth

[en] During Run-Beyond-Cladding-Breach (RBCB) operation in an oxide LMR, the performance of a breached fuel element is intimately associated with the formation of fuel-sodium reaction product (FSRP), Na₃(U/sub 1-y/Pu/sub y/)O₄. In-pile experiments coupled with destructive examinations of breached fuel have consistently revealed noticeable changes in fuel structure accompanying FSRP formation at the fuel surface. Previous analyses have also indicated a significant impact of FSRP on fuel centerline temperature. Successful modeling of breached fuel thermal behavior therefore requires a reasonably accurate knowledge of the thermal properties of the FSRP, especially its thermal conductivity. But laboratory investigations have been scarce and limited to the Na/UO₂ system because of the toxicity of plutonium and hygroscopicity of the FSRP. Hence, post-irradiation observations of fuel samples remain the most amenable way of deriving the thermal conductivity of the FSRP. Such work is a spin-off of the RBCB program in the Experimental Breeder Reactor-II (EBR-II), a program jointly sponsored by the US Department of Energy and the Power Reactor and Nuclear Fuel Development Corporation of Japan

[en] The run-beyond-cladding-breach (RBCB) operation of mixed-oxide liquid-metal reactor fuel pins has been studied for 6 yr in the Experimental Breeder Reactor II (EBR-II) as part of a joint program between the US Department of Energy and the Power Reactor and Nuclear Fuel Development Corporation of Japan. The formation of fuel-sodium reaction product (FSRP), Na₃MO₄, where M = U_1-yPu_y, in the outer fuel regions is the major phenomenon governing RBCB behavior. It increases fuel volume, decreases fuel stoichiometry, modifies fission product distributions, and alters thermal performance of a pin. This paper describes the morphology of Na₃MO₄ observed in 5.84-mm-diam pins covering a variety of conditions and RBCB times up to 150 effective full-power days (EFPDs)

[en] Focusing on the effect of temperature gradients across the cladding wall on void formation, observation of void distribution was carried out by means of electron microscopic techniques in the direction of the wall thickness in the irradiated fuel claddings made of P, Ti-modified 316 steel. A deformation analysis was also conducted including irradiation creep and swelling deformation, using a finite element method. At the beginning of swelling a secondary stress occurred at the cladding surface due to non-uniform swelling arising from the temperature gradient. Swelling appears to accelerate by this secondary stress, and the swelling developed a non-monotonic distribution, namely the swelling in mid-wall region was lower than that in the surface regions. This non-monotonic swelling distribution disappears with increasing fluence, however. In the deformation analysis, it became clear that swelling is not continuously accelerated by the secondary stress, because irradiation creep deformation relaxes the stress

[en] Paper deals with the results of the efforts to investigate into thermal conductivity of the MOX fuel fabricated following the UO₂ vibration compaction process undertaken in the Japan's Nuclear Fuel Cycle Development Institute (the JNC). Paper presents as well the results of thermal conductivity calculations on the basis of a thermal model. One investigated into the effect of uranium particles exposed and not exposed to melting to be added to the fuel

[en] Studies indicate there may be a large economic incentive to permit some continued reactor operation with breached fuel pin cladding. A major concern for this type of operation is the potential spread of contamination in the primary coolant system and its impact on plant maintenance. A study of the release and transport of contamination from naturally breached mixed oxide Liquid Metal Reactor (LMR) fuel pins was performed as part of the US Department of Energy/Power Reactor and Nuclear Fuel Development Corporation (DOE/PNC) Run Beyond Cladding Breach (RBCB) Program at EBR-II. The measurements were made using the Breached Fuel Test Facility (BFTF) at EBR-II with replaceable deposition samplers located approximately 1.5 meters from the breached fuel test assemblies. The effluent from the test assemblies containing the breached fuel pins was routed up through the samplers and past dedicated instrumentation in the BFTF before mixing with the main coolant flow stream. This paper discusses the first three contamination tests in this program. 2 refs., 5 figs., 2 tabs

[en] The run-beyond-cladding-breach (RBCB) operation of mixed-oxide LMR fuel pins has been studied for six years in the Experimental Breeder Reactor-II (EBR-II) as part of a joint program between the US Department of Energy and the Power Reactor and Nuclear Fuel Development Corporation of Japan. The formation of fuel-sodium reaction product (FSRP), Na₃MO₄, where M = U/sub 1-y/Pu/sub y/, in the outer fuel regions is the major phenomenon governing RBCB behavior. It increases fuel volume, decreases fuel stoichiometry, modifies fission-product distributions, and alters thermal performance of a pin. This paper describes the morphology of Na₃MO₄ observed in 5.84-mm diameter pins covering a variety of conditions and RBCB times up to 150 EFPD's. 8 refs., 1 fig

[en] Oxide dispersion strengthened (ODS) steels are being developing for application as advanced fast reactor cladding and fusion blanket materials, in order to allow increased operation temperature. Oxidation testing of ODS steel was conducted under a controlled dry air atmosphere to evaluate the high temperature oxidation behavior. This showed that 9Cr-ODS martensitic steels and 12Cr-ODS ferritic steels have superior high temperature oxidation resistance compared to 11 mass% Cr PNC-FMS and 17 mass% Cr ferritic stainless steel. This high temperature resistance is attributed to earlier formation of the protective α-Cr₂O₃ on the outer surface of ODS steels

[en] 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