Prototypic irradiation testing of high-density U-Mo alloy dispersion fuels (original) (raw)
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Shielding Considerations for Advanced Fuel Irradiation Experiments
Journal of Nuclear Science and Technology, 2008
An in-pile test program for the development of a high burn-up fuel is planned for the HANARO reactor. The source term originates from a leakage of fission products from the anticipated failed fuels into the gas flow tubes and around the instrumentation and control system. In order to quantify the fuel composition in the event of a fuel failure, the isotope generation and depletion code ORIGEN 2.0 was used. The computer program Microshield 6.2 was used to calculate the doses from specific locations, where a high radioactivity is expected during an irradiation. The results indicate that the equivalent dose in the investigated working areas is less than the permitted dose rate of 6.25 Sv/hr. However, access to the area of a decay vessel may need to be limited, and the installation of a Pb wall with a 20.5 cm thickness is recommended. From the analysis of a radioactive decay with time, most of the concerned gaseous nuclides with short half-lives after 3 months, were decayed, with one exception which was Kr-85, thus it should be released in accordance with applicable government laws after measuring its activity in individual holding vessels.
Irradiation Testing of Structural Materials in Fast Breeder Test Reactor
Fast Breeder Test Reactor (FBTR) at Kalpakkam, India is a sodium cooled fast reactor with neutron flux level of the order of 10 15 n/cm 2 /s and temperature of coolant in the range of 600-720K (330-450°C), which is being used for the development of fuel and structural materials required for Indian Fast Reactor Programme. Irradiation performance testing on structural materials is being carried out by subjecting prefabricated specimens to desired experimental conditions as part of planned irradiation experiments, and by testing of material samples sourced from actual fuel clad tubes / fuel assembly wrapper tubes irradiated to various fuel burn up levels in FBTR. Pressurised capsules of Zirconium alloys and D9 alloy (modified stainless steel type 316 with controlled additions of titanium and silicon) have been developed to determine the in-reactor creep performance of indigenously developed zirconium alloys and D9 alloy. Pressurised capsules made of zirconium alloys were subjected to fluence levels up to 1.1 x 10 21 n/cm 2 (E> 1 MeV) in FBTR at temperatures of 579 to 592K and diameter measurements were carried out in the hot cell facility to determine the irradiation creep rate. Pressurised capsules of D9 alloy are currently undergoing irradiation at a temperature of 623K in FBTR along with small size tensile test specimens and shear punch test specimens of D9. Non-instrumented gas-gap type irradiation capsule has been developed to achieve higher irradiation temperatures (673 to 873K) of structural material specimens. The irradiation induced mechanical property changes in cold worked AISI Type SS316 fuel cladding of FBTR have been determined from tensile testing of portions of irradiated fuel clad tubes in the hot cells. Tests were carried out on clad tubes with dpa ranging from 13 to 83 at various test temperatures from ambient (300K) to irradiation temperature (790K). Shear punch tests have been used for characterizing the tensile property changes in cold worked AISI Type SS 316 wrapper material of FBTR fuel assemblies. From the results of shear punch tests on irradiated specimens, using correlation equations, the tensile properties of the wrapper material irradiated to various dpa ranging from 30 to 83 have been estimated. A considerable increase in the strength and decrease in the ductility of the wrapper material with increasing dpa was observed from the results. This paper discusses the salient features of irradiation facilities available at FBTR, irradiation experiments carried out on structural materials, and some of the important results obtained from tests on irradiated structural materials.
Frontiers in Nuclear Engineering
Metallic U alloys have high U density and thermal conductivity and thus have been explored since the beginning of nuclear power research. Alloys of U with modest amounts of Mo, such as U-10 wt % Mo (U-10Mo), are of particular interest because the γ-U crystal structure in this alloying addition shows prolonged stability in reactor service. Historically, radiation data on U-10Mo fuels were collected in Na fast reactors or lower temperature research reactor conditions, but little is known about irradiation behavior, particularly swelling and creep, at irradiation temperatures between 250 and 500°C. This work discusses the methodology and pre-irradiation characterization results from a U-Mo irradiation campaign performed in the High Flux Isotope Reactor at Oak Ridge National Laboratory. U-10Mo and U-17Mo samples irradiations are being completed at temperatures ranging from 250 to 500°C to three targeted fission densities between 2 × 1020 and 1.5 × 1021 fissions per cubic centimeter. Swe...
Irradiation tests and post-irradiation examinations of DUPIC fuel
Annals of Nuclear Energy, 2008
The technology of a DUPIC (Direct Use of spent PWR fuel In CANDU Reactors) fuel has been developed at KAERI for 10 years. To identify a robustness of the DUPIC fuel pellet, it has been irradiated for six times since August 1999 in HANARO. Among them, the first irradiation test was carried out with a simulated fuel. Therefore, a maximum burnup up to 6700 MWD/tHM was achieved through six irradiation tests of the DUPIC fuel. A remote instrumentation technology was also developed to obtain various on-line data including a centerline temperature and some remote devices had also been implemented. After irradiation tests of the DUPIC fuel, post-irradiation tests had been performed consecutively in the irradiation material examination facility (IMEF) at KAERI. A fuel performance code was also developed to compare the measured centerline temperatures for the fifth and the sixth irradiation tests.
Heavy ion irradiation of U–Mo/Al dispersion fuel
2006
The usage of high-density U-Mo/Al dispersion fuel for high burn up in research and test reactors seems to be limited by the unfavourable interdiffusion layer between the fuel and the Al-matrix, which develops during irradiation. This interdiffusion layer was observed up to now only after costly and time consuming in-pile irradiation and could not be created in out-of-pile experiments. This paper presents a new approach of creating such an interdiffusion layer out-of-pile by irradiation with heavy ions. An appropriate choice of heavy-ion irradiation simulates irradiation damage and deposition of fission fragments as it happens during in-pile irradiation and induces a diffusion process between the fuel and the Al matrix. An irradiation experiment and post-irradiation examinations are presented.
JOM, 2017
A low-enriched uranium U-10Mo monolithic nuclear fuel is being developed by the Material Management and Minimization Program, earlier known as the Reduced Enrichment for Research and Test Reactors Program, for utilization in research and test reactors around the world that currently use high-enriched uranium fuels. As part of this program, reactor experiments are being performed in the Advanced Test Reactor. It must be demonstrated that this fuel type exhibits mechanical integrity, geometric stability, and predictable behavior to high powers and high fission densities in order for it to be a viable fuel for qualification. This paper provides an overview of the microstructures observed at different regions of interest in fuel plates before and after irradiation for fuel samples that have been tested. These fuel plates were fabricated using laboratory-scale fabrication methods. Observations regarding how microstructural changes during irradiation may impact fuel performance are discussed.
Validation of the Physics Analysis used to Characterize the AGR-1 TRISO Fuel Irradiation Test
2015
The results of a detailed physics depletion calculation used to characterize the AGR-1 TRISO-coated particle fuel test irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory are compared to measured data for the purpose of validation. The particle fuel was irradiated for 13 ATR power cycles over three calendar years. The physics analysis predicts compact burnups ranging from 11.30-19.56% FIMA and cumulative neutron fast fluence from 2.21-4.39E+25 n/m 2 under simulated high-temperature gas-cooled reactor conditions in the ATR. The physics depletion calculation can provide a full characterization of all 72 irradiated TRISO-coated particle compacts during-and post-irradiation, so validation of this physics calculation was a top priority. The validation of the physics analysis was done through comparisons with available measured experimental data which included: 1) high-resolution gamma scans for compact activity and burnup, 2) mass spectrometry for compact burnup, 3) flux wires for cumulative fast fluence, and 4) mass spectrometry for individual actinide and fission product concentrations. The measured data are generally in very good agreement with the calculated results, and therefore provide an adequate validation of the physics analysis and the results used to characterize the irradiated AGR-1 TRISO fuel. Graphite holder Fuel compact stack Capsule 1 Nodes 41-48 Capsule 1 Capsule 6 Nodes 1-8 Capsule 5 Nodes 9-16 Capsule 4 Nodes 17-24 Capsule 3 Nodes 25-32 Capsule 2 Nodes 33-40
Annals of Nuclear Energy, 2021
UCO and UO 2 tristructural isotropic fuel compacts were irradiated in the AGR-2 experiment, conducted in the Advanced Test Reactor for 559.2 effective full power days. UCO and UO 2 compacts reached calculated peak burnups of 13.15 and 10.69% fissions per initial heavy-metal atom, and fast fluences of 3.47 Â 10 25 and 3.53 Â 10 25 n/m 2 (E > 0.18 MeV), respectively. The time-average volume-average temperatures ranged from 987 to 1296°C in UCO compacts and from 996 to 1062°C in UO 2 compacts. Fission product release-to-birth (R/B) ratios remained below 2 Â 10 À6 in UCO compacts and 10 À7 in UO 2 compacts during the first three irradiation cycles. R/B data then became unreliable due to mixing of the capsule gas flows which hindered the evaluation of fuel performance for the later portion of the irradiation. Post-irradiation examination of the irradiation capsules and fuel compacts is underway and will provide additional information on fuel performance.