• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2395-2407
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• 2022

The sodium-cooled fast reactor is among the innovative nuclear technologies selected in the framework of the development of Generation IV concepts, allowing the irradiation of uranium-plutonium mixed oxide fuels (MOX). A fundamental step for the safety assessment of MOX-fuelled pins for fast reactor applications is the evaluation, by means of fuel performance codes, of the integral thermal-mechanical behaviour under irradiation, involving the fission gas behaviour and release in the fuel-cladding gap. This work is dedicated to the performance analysis of an inner-core fuel pin representative of the ASTRID sodium-cooled concept design, selected as case study for the benchmark between the GERMINAL and TRANSURANUS fuel performance codes. The focus is on fission gas-related mechanisms and integral outcomes as predicted by means of the SCIANTIX module (allowing the physics-based treatment of inert gas behaviour and release) coupled to both fuel performance codes. The benchmark activity involves the application of both GERMINAL and TRANSURANUS in their "pre-INSPYRE" versions, i.e., adopting the state-of-the-art recommended correlations available in the codes, compared with the "post-INSPYRE" code results, obtained by implementing novel models for MOX fuel properties and phenomena (SCIANTIX included) developed in the framework of the INSPYRE H2020 Project. The SCIANTIX modelling includes the consideration of burst releases of the fission gas stored at the grain boundaries occurring during power transients of shutdown and start-up, whose effect on a fast reactor fuel concept is analysed. A clear need to further extend and validate the SCIANTIX module for application to fast reactor MOX emerges from this work; nevertheless, the GERMINAL-TRANSURANUS benchmark on the ASTRID case study highlights the achieved code capabilities for fast reactor conditions and paves the way towards the proper application of fuel performance codes to safety evaluations on Generation IV reactor concepts.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.715-722
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• 2009

A model to predict failure of designed-to-fail (dtf) fuel particles is discussed. The dtf fuel under study consisted of a uranium oxycarbide kernel coated with a single pyrocarbon seal coat. Coating failure was assumed to be due to fission gas recoil and knockout mechanisms and direct diffusive release of fission gas from the kernel, which acted to increase pressure and stress in the pyrocarbon layer until it ruptured. Predictions of dtf fuel failure using General Atomics' particle fuel performance code for HRB-17/18 and HFR-B1 irradiation tests were reasonably accurate; however, the model could not predict the failure for COMEDIE BD-1. This was most likely due to insufficient information on reported particle fuel failure at the beginning.

• Nuclear Engineering and Technology
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• v.41 no.2
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• pp.155-162
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• 2009

Since the 90's, EDF and AREVA-NP have irradiated, up to very high burnups, lead assemblies housing $M5^{(R)}$ cladded fuels. Post-irradiation examination of high burnup $UO_2$ pellets show an increase in the fission-gas release rate, an increase in fuel swelling, and formation of fission-gas bubbles throughout the pellets. Xenon abundances were quantified, and phenomena leading to this bubble formation were identified. All examinations provided valuable data on the complex state of the fuel during irradiation. They show the good behavior of these fuels, exhibiting various microstructures at very high burnups, none of which is likely to lead to problems during irradiation.

• Nuclear Engineering and Technology
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• v.2 no.4
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• pp.255-262
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• 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 Vacuum Society Conference
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• 2013.02a
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• pp.670-670
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• 2013

The concept of a fusion-driven transmutation reactor based on LAR (Low Aspect Ratio) tokamak as a neutron source is studied based on ITER physics and technology. The radial build of transmutation reactor components are self-consistently determined by coupling the systems analysis with radiation transport analysis and an optimal configuration of a transmutation reactor for aspect ratio, A in the range of 1.5 to 2.0 is found. The performance of a transmutation reactor is investigated and shows that a transmutation reactor with a neutron source producing fusion power less than 150 MW can destroy the transuranic actinides contained in the spent fuels produced from more than two 1 GWe PWRs with production of the fission power being greater than 2 GW.

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

• Nuclear Engineering and Technology
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• v.31 no.4
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• pp.365-374
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• 1999

A nuclear design of fission moly production targets for a research reactor, HANARO was peformed. It was found that the use of MCNP-4A, ORIGEN-2 code was reliable for the analysis of production characteristics of $^{99}$ Mo in a target fuel at an irradiation holes. A parametric study was done for the optimization of target location, target dimension, target shape and fuel materials. It was shown that a fuel thickness was the most sensitive parameters and electro-deposited target gave the highest 99Mo yield ratio. A pellet target with vibro-compaction powder, however, showed the largest production capacity and better engineering feasibility even with less yield ratio. Ten kinds of optimized target design for both LEU and HEU satisfied all the given design constraints. The most favorable design was the HEU ring-shaped electro-deposited target, considered the safety limit, production yield, chemical process easiness, yield ratio, and amount of radioactive waste.

• Nuclear Engineering and Technology
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• v.3 no.2
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• pp.77-97
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• 1971

The experimental fission cross-section data of U-238, Np-237 and Th-232, published up to the end of 1970, are reviewed and analyzed between their respective thresholds and 20.0 MeV. The results of a statistical analysis of the available data, performed with a weighted Least-squares Orthogonal Polynomial Pitting computer programme are presented in the form of point-wise cross-section values together with their uncertainties, and in the form of graphs of the fitted curves with an indication of a region of 95% statistical confidence level. An estimate of the fission spectrum weighted average cross-sections and their respective uncertainties is also given.

• Journal of the Korean Chemical Society
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• v.16 no.2
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• pp.59-63
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• 1972

In connection with two electron binding energy of each bond of saturated hydrocarbons, C-C bond fission and hydrogen abstraction from C-H bond are discussed by means of two center Huckel method. A beautiful correlation could be noticed between the observed bond dissociation energy and the calculated bond energy except for n-butane. Bond dissociation energies between C-C bond were also related to C-C bond fission. We could also find a very close relation between the relative easiness of hydrogen abstraction and the calculated binding energy of C-H bond. In other words, C-H bonds of tertiary hydrogen have been noticed as most weakely bonded and hence the tertiary hydrogen would most easily from the paraffins. In addition, the C-H binding energy is discussed applying ionic character of C-H bond which is derived from its dipole moment (0.4D)

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.193-198
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• 1998

Fission Moly 생산을 위한 조사시설로 HANARO가 사용점에 있어서 충분한 타당성을 갖는지에 대해 원자로 운전중 표적장전에 따른 노심반응도의 영향, 표적의 최대 표면 열출력밀도, Mo-99 생산능력 측면에서 평가하였다. 그 결과 운전중 표적 장전으로 인한 반응도가(reactivity worth)가 약 0.2 %Δ$\rho$정도로 제한치인 1.25 %Δ$\rho$보다 현저히 작고, 최대 표면 열출력밀도 또한 2.76 Mw/$m^2$보다 현저히 작음을 볼 수 있었다. 또한 OR공 한 개만을 사용한다. 할지라도 Mo-99의 상업생산 목표를 충분히 만족시킬 수 있음을 알 수 있었다 따라서 HANARO를 Fission Moly 생산을 위한 조사시설로 사용하는 것은 타당하다 할 수 있다.