• Proceedings of the Korean Nuclear Society Conference
• /
• 2005.10a
• /
• pp.280-281
• /
• 2005

• Proceedings of the Korean Nuclear Society Conference
• /
• 2005.10a
• /
• pp.173-174
• /
• 2005

• Proceedings of the Korean Nuclear Society Conference
• /
• 2005.10a
• /
• pp.258-259
• /
• 2005

• Nuclear Engineering and Technology
• /
• v.48 no.3
• /
• pp.613-623
• /
• 2016

Molybdenum-99 ($^{99}Mo$) is the most important isotope because its daughter isotope, technetium-99m ($^{99m}Tc$), has been the most widely used medical radioisotope for more than 50 years, accounting for > 80% of total nuclear diagnostics worldwide. In this review, radiochemical routes for the production of $^{99}Mo$, and the aspects for selecting a suitable process strategy are discussed from the historical viewpoint of $^{99}Mo$ technology developments. Most of the industrial-scale $^{99}Mo$ processes have been based on the fission of $^{235}U$. Recently, important issues have been raised for the conversion of fission $^{99}Mo$ targets from highly enriched uranium to low enriched uranium (LEU). The development of new LEU targets with higher density was requested to compensate for the loss of $^{99}Mo$ yield, caused by a significant reduction of $^{235}U$ enrichment, from the conversion. As the dramatic increment of intermediate level liquid waste is also expected from the conversion, an effective strategy to reduce the waste generation from the fission $^{99}Mo$ production is required. The mitigation of radioxenon emission from medical radioisotope production facilities is discussed in relation with the monitoring of nuclear explosions and comprehensive nuclear test ban. Lastly, the $^{99}Mo$ production process paired with the Korea Atomic Energy Research Institute's own LEU target is proposed as one of the most suitable processes for the LEU target.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1996.05b
• /
• pp.637-642
• /
• 1996

This paper presents the analysis results for the core degradation processes and the fission product release of the PHEBUS FPT0 experiment using MELCOR1.8.3. The objective of this study is to assess models associated with the core damage and fission product behavior in MELCOR. The calculation results were much improved through sensitivity studies. Thermal/hydraulic behavior in the core and the circuit was well predicted under the intact core geometry. In non-eutectic model case. the UO$_2$ dissolution model in the MELCOR always showed such a tendency that the resulting dissolved UO$_2$ mass was small at the highly oxidized condition due to the model logic. Total H$_2$ generation mass was underpredicted because the stiffner was not modeled and the liner in the shroud was not allowed to be oxidized in MELCOR. Some difficulties were found in modeling the activation product were solved by manipulating the RN input associated with the initial fission product inventory. These problem were occurred because there are no control rod model in MELCOR. Generally the fission product release ratio showed a similar trend compared with the measured data except the activation product. which have no model to simulate in MELCOR.

• Nuclear Engineering and Technology
• /
• v.2 no.4
• /
• pp.255-262
• /
• 1970

More detailed calculations of extension to general anisotropic transport equation with fission are studied. These calculations involve that the operator can be splitted into scattering and fission operators when we prove the completeness of general anisotropy. Applying these operators to the equation makes it easy to extract the slowing-down transient of zero-measure, and completely solves the transport equation. In addition, the number of the eigenvalues of the second anisotropy is classified with Cs unknown, B$_1$and B$_2$known constants.

• Proceedings of the Korean Nuclear Society Conference
• /
• 2004.10a
• /
• pp.1063-1064
• /
• 2004

• Proceedings of the Korean Nuclear Society Conference
• /
• 2005.10a
• /
• pp.300-301
• /
• 2005

• Nuclear Engineering and Technology
• /
• v.49 no.2
• /
• pp.434-441
• /
• 2017

Containment venting is one of several essential measures to protect the integrity of the final barrier of a nuclear reactor during severe accidents, by which the uncontrollable release of fission products can be avoided. The authors seek to develop an optimization approach to venting operations, from a simulation-based perspective, using an integrated severe accident code, THALES2/KICHE. The effectiveness of the containment-venting strategies needs to be verified via numerical simulations based on various settings of the venting conditions. The number of iterations, however, needs to be controlled to avoid cumbersome computational burden of integrated codes. Bayesian optimization is an efficient global optimization approach. By using a Gaussian process regression, a surrogate model of the "black-box" code is constructed. It can be updated simultaneously whenever new simulation results are acquired. With predictions via the surrogate model, upcoming locations of the most probable optimum can be revealed. The sampling procedure is adaptive. Compared with the case of pure random searches, the number of code queries is largely reduced for the optimum finding. One typical severe accident scenario of a boiling water reactor is chosen as an example. The research demonstrates the applicability of the Bayesian optimization approach to the design and establishment of containment-venting strategies during severe accidents.

• Nuclear Engineering and Technology
• /
• v.13 no.4
• /
• pp.254-263
• /
• 1981

Paper electrophoretic separation of fission products has been carried out by using the specially designed migration apparatus. In general, the isolation of rubidium, strontium, zirconium, ruthenium, cesium, cerium, molybdenum, and some short-lived fission products is more efficient under 0.1M HCl electrolyte as compared with 0.1M NaOH electrolyte. In addition to Np-239, 1-131∼135 were, in particular, observed with different iodine chemical species obtained by the paper-electrophoretic separation of short, neutron-irradiated uranyl nitrate solution.