Pre-irradiation Testing and Analysis to Support the LWRS Hybrid SiC-CMC-Zircaloy-04 Unfueled Rodlet Irradiation PDF Download
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Author: Publisher: ISBN: Category : Languages : en Pages :
Book Description
Nuclear fuel performance is a significant driver of nuclear power plant operational performance, safety, economics and waste disposal requirements. The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Pathway focuses on improving the scientific knowledge basis to enable the development of high-performance, high burn-up fuels with improved safety and cladding integrity and improved nuclear fuel cycle economics. To achieve significant improvements, fundamental changes are required in the areas of nuclear fuel composition, cladding integrity, and fuel/cladding interaction.
Author: Publisher: ISBN: Category : Languages : en Pages :
Book Description
Nuclear fuel performance is a significant driver of nuclear power plant operational performance, safety, economics and waste disposal requirements. The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Pathway focuses on improving the scientific knowledge basis to enable the development of high-performance, high burn-up fuels with improved safety and cladding integrity and improved nuclear fuel cycle economics. To achieve significant improvements, fundamental changes are required in the areas of nuclear fuel composition, cladding integrity, and fuel/cladding interaction.
Author: Publisher: ISBN: Category : Languages : en Pages :
Book Description
The purpose of the gamma irradiation tests conducted at the Idaho National Laboratory (INL) was to obtain a better understanding of chemical interactions and potential changes in microstructural properties of a mock-up hybrid nuclear fuel cladding rodlet design (unfueled) in a simulated PWR water environment under irradiation conditions. The hybrid fuel rodlet design is being investigated under the Light Water Reactor Sustainability (LWRS) program for further development and testing of one of the possible advanced LWR nuclear fuel cladding designs. The gamma irradiation tests were performed in preparation for neutron irradiation tests planned for a silicon carbide (SiC) ceramic matrix composite (CMC) zircaloy-4 (Zr-4) hybrid fuel rodlet that may be tested in the INL Advanced Test Reactor (ATR) if the design is selected for further development and testing.
Author: Publisher: ISBN: Category : Languages : en Pages :
Book Description
The Light Water Reactor Sustainability (LWRS) Program was established by the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) to develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors. The LWRS Program is divided into four R & D Pathways: (1) Materials Aging and Degradation; (2) Advanced Light Water Reactor Nuclear Fuels; (3) Advanced Instrumentation, Information and Control Systems; and (4) Risk-Informed Safety Margin Characterization. This report describes an irradiation testing readiness analysis in preparation of LWRS experiments for irradiation testing at the Idaho National Laboratory (INL) Advanced Test Reactor (ATR) under Pathway (2). The focus of the Advanced LWR Nuclear Fuels Pathway is to improve the scientific knowledge basis for understanding and predicting fundamental performance of advanced nuclear fuel and cladding in nuclear power plants during both nominal and off-nominal conditions. This information will be applied in the design and development of high-performance, high burn-up fuels with improved safety, cladding integrity, and improved nuclear fuel cycle economics.
Author: A. R. Causey Publisher: ISBN: Category : Languages : en Pages : 0
Book Description
This report presents the final safety analysis for the preparation, conduct, and post-test discharge operation for the full-length high temperature experiment-5 (flht-5) to be conducted in the l-24 position of the national research universal (nru) reactor at chalk river nuclear laboratories (crnl), ontario, canada. the test is sponsored by an international group organized by the u.s. nuclear regulatory commission. the test is designed and conducted by staff from pacific northwest laboratory with crnl staff support. the test will study the consequences of loss-of-coolant and the progression of severe fuel damage. an array of full-length lwr fuel rods will be subjected to conditions that simulate loss-of-coolant flow at decay-heat power level. the 12-position array includes one pre-irradiated pwr rod, 10 fresh fuel rods, and one instrument thimble containing a dummy steel rod. the boilaway of the coolant will permit heatup of the rods to the point of rapid cladding oxidation with concomitant cladding melting, partial fuel liquefaction, fuel oxidation, hydrogen generation, and fission product release. the hydrogen generation (from the oxidation reaction), the bundle temperatures, the liquid level, and the fission product release will be monitored during the test. the melt progression will be assessed using post test visual and metallographic examination results. experience from similar flht tests is combined with analysis results to show that: 1) high-temperature material will be contained within the shroud that surrounds the rod array, precluding damage to the nru loop pressure tube; 2) the released fission products and hydrogen will be contained and disposed of by the effluent control system in a manner that poses no unresolved safety hazard to operating personnel or to the public; and 3) the radiation exposure to operating personnel and to the public will remain below approved control limits.
Author: Margaret A. McGrath Publisher: ISBN: Category : Guide tube Languages : en Pages : 24
Book Description
Re-crystallized Zircaloy-4 guide tubes were irradiated in commercial pressurized water reactors (PWRs) at three different temperatures to fluences near 1 x 1022 n/cm2 E>1 MeV (15 displacements per atom), resulting in moderate corrosion and three different hydrogen contents (approximately 135, 240, and 700 parts per million). Sections of the guide tubes were re-irradiated in the Halden reactor to assess the irradiation creep and growth behaviors. Three conditions were applied: Bellows-loaded axial compression creep; zero stress growth; and zero stress, zero flux (control specimens). The guide tube sections were re-irradiated under simulated PWR conditions by utilizing a pressurized light water loop operating with normal PWR water chemistry at approximately 320C. Axial length changes were measured in-reactor by linear variable differential transformers (LVDTs), and post-irradiation hot cell measurements were done to confirm the LVDT elongation measurements. After minor corrections were made to account for reactor testing variables, it was shown that the LVDT measurements were accurate, thus creep and growth or free growth rates were established for each guide tube section. Hot cell examinations were also performed to establish the state of corrosion of each specimen, including hydrogen content, both before and after the re-irradiation in the Halden reactor. The results showed that stress free irradiation growth was different for each specimen and correlated qualitatively with the hydrogen content and commercial irradiation temperatures of the guide tubes. The higher hydrogen content, or higher commercial irradiation temperature, gave rise to higher subsequent growth rates. After subtracting the growth strain from the measured creep and growth strain values, the magnitude of creep and creep rates were essentially the same for all specimens: No effect of commercial reactor irradiation temperature or hydrogen content was observed. The results give important new data on irradiation creep and growth and on the correlation between hydrogen content and irradiation temperature on growth rates.
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Author: A. E. Mccorry
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Author: A. R. Causey
Author: A. W. Stewart
Author: R. M. Scott
Author: K. D. Cotnam
Author: Margaret A. McGrath
Author: J. E. McCauley