• Proceedings of the Korean Institute of Electrical and Electronic Material Engineers Conference
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• 2002.07a
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• pp.62-66
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• 2002

The conventional floating zone(FZ) crystal and Czochralski(CZ) silicon crystal have resistivity variations longitudinally as well as radially The resistivity variations of the conventional FZ and CZ crystal are not conformed to requirement of dopant distribution for power devices and thyristors. These resistivity variations in conventional cystals limits the reverse breakdown voltage that could be achieved and forced designers of high power diodes and thyristors to compromise the desired current-voltage characteristics. So to produce high Power diodes and thyristors, Neutron Transmutation Doping(NTD) technique is the one method just because NTD silicon provides very homogeneous distribution of doping concentration. This procedure involves the nuclear transmutation of silicon to phosphorus by bombardment of neutron to the crystal according to the reaction $^{30}$ Si(n,${\gamma}$)longrightarrow$^{31}$ Silongrightarrow(2.6 hr)$^{31}$ P+$\beta$$^{[-10]}$ . The radioactive isotope $^{31}$ Si is formed by $^{31}$ Si capturing a neutron, which then decays into the stable $^{31}$ P isotope (i.e., the donor atom), whose distribution is not dependent on the crystal growth parameters. In this research, neutron was irradiated on FZ silicon wafers which had high resistivity(1000~2000 Ω cm), for 26 and 8.3hours for samples of HTS-1 and HTS-2, and 13, 3.2, 2.0 hours for samples of IP-1, IP-2 and IP-3, respectively, to compare resistivity changes due to time differences. The designed resistivities were approached, which were 2.l Ωcm for HTS-1, 7.21 Ω cm for HTS-2, 1.792cm for IP-1, 6.83 Ωcm for IP-2, 9.23 Ωcm for IP-3, respectively. Point defects were investigated with Deep Level Transient Spectroscopy(DLTS). Four different defects were observed at 80K, 125K, 230K, and above 300K.

• Nuclear Engineering and Technology
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• v.49 no.6
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• pp.1346-1353
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• 2017

The FISPACT-II inventory simulation platform is a modern computational tool with advanced and unique capabilities. It is sufficiently flexible and efficient to make it an ideal basis around which to perform extensive simulation studies to scope a variety of responses of many materials (elements) to several different neutron irradiation scenarios. This paper briefly presents the typical outputs from these scoping studies, which have been used to compile a suite of nuclear physics materials handbooks, providing a useful and vital resource for material selection and design studies. Several different global responses are extracted from these reports, allowing for comparisons between materials and between different irradiation conditions. A new graphical output format has been developed for the FISPACT-II platform to display these "global summaries"; results for different elements are shown in a periodic table layout, allowing side-by-side comparisons. Several examples of such plots are presented and discussed.

• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.1907-1916
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• 2020

In this work, the depletion capability implemented in Monte Carlo code MCS is investigated to predict the isotopic compositions of spent nuclear fuel (SNF). By comparison of MCS calculation results to post irradiation examination (PIE) data obtained from one pressurized water reactor (PWR), the validation of this capability is conducted. The depletion analysis is performed with the ENDF/B-VII.1 library and a fuel assembly model. The transmutation equation is solved by the Chebyshev Rational Approximation Method (CRAM) with a depletion chain of 3820 isotopes. 18 actinides and 19 fission products are analyzed in 14 SNF samples. The effect of statistical uncertainties on the calculated number densities is discussed. On average, most of the actinides and fission products analyzed are predicted within ±6% of the experiment. MCS depletion results are also compared to other depletion codes based on publicly reported information in literature. The code-to-code analysis shows comparable accuracy. Overall, it is demonstrated that the depletion capability in MCS can be reliably applied in the prediction of SNF isotopic inventory.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.949-953
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• 2019

The burnup equation of nuclides is one of the most equations in nuclear reactor physics, which is generally coupled with transport calculations. The burnup equation describes the variation of the nuclides with time. Because of its very stiffness and the need for large time step, this equation is solved by special methods, for example transmutation trajectory analysis (TTA) or the matrix exponential methods where the matrix exponential is approximated by CRAM. However, TTA or CRAM functions well when the flux is constant. In this work, a new method is proposed when the flux changes. It's an improved method compared to TTA or CRAM. Furtherly, this new method is based on TTA or CRAM, and it is more accurate than them. The accuracy and efficiency of this method are investigated. Several cases are used and the results show the accuracy and efficiency of this method are great.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.603-610
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• 2021

The major drawback of zirconia-based materials, in view of their applications as targets for minor actinide transmutation, is their poor thermal conductivity. The addition of MgO, which has high thermal conductivity, to zirconia-based materials is expected to improve their thermal conductivity. On these grounds, the present study aims at phase characterization and thermophysical property evaluation of neodymium-substituted zirconia (Zr_0.8Nd_0.2O_1.9; using Nd₂O₃ as a surrogate for Am₂O₃) and its composites with MgO. The composite was prepared by a solid-state reaction of Zr_0.8Nd_0.2O_1.9 (synthesized by gel combustion) and commercial MgO powders at 1773 K. Phase characterization was carried out by X-ray diffraction and the microstructural investigation was performed using a scanning electron microscope equipped with energy dispersive spectroscopy. The linear thermal expansion coefficient of Zr_0.8Nd_0.2O_1.9 increases upon composite formation with MgO, which is attributed to a higher thermal expansivity of MgO. Similarly, specific heat also increases with the addition of MgO to Zr_0.8Nd_0.2O_1.9. Thermal conductivity was calculated from measured thermal diffusivity, temperature-dependent density and specific heat values. Thermal conductivity of Zr_0.8Nd_0.2O_1.9-MgO (50 wt%) composite is more than that of typical UO₂ fuel, supporting the potential of Zr_0.8Nd_0.2O_1.9-MgO composites as target materials for minor actinides transmutation.

• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.407-412
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• 2014

In this research, we investigated the burnup characteristics and the conversion of fertile $^{232}Th$ into fissile $^{233}U$ in the core of a Sodium-Cooled Fast Reactor (SFR). The SFR fuel assemblies were designed for burning $^{232}Th$ fuel (fuel pin 1) and $^{233}U$ fuel (fuel pin 2) and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method) and TTA (Transmutation Trajectory Analysis) method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff) was between 0.964 and 0.954 and peaking factor is 1.88867.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.982-983
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• 2005

• Nuclear Engineering and Technology
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• v.25 no.3
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• pp.480-496
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• 1993

원자력발전 핵연료주기에서 고려하여야 할 중요한 요소의 하나는 사용후핵연료에서 비롯되는 고준위 방사성핵종이다. 고준위 방사성핵종의 처분 방법으로서 심지층처분방식은 가장 손쉬운 방법이기는 하나 매우 장시간의 감시가 필요하며, 특히 자연환경으로의 누출가능성이 커서 이의 대안으로서 외국 몇 나라에서는 소위 소멸처리(Transmutation)방법에 대한 연구를 활발히 하고 있다. 현시점에서 소멸처리 방법으로 가장 타당성이 있는 것으로 여겨지는 것은 원자로를 이용하는 것과 가속기 구동 미임계 시스템 (Accelerator-Driven Subcritical System)을 이용하는 방법이다. 본 기고문에서는 이들 방법을 중심으로 다양한 소멸처리 방법의 소개와 기술적인 문제점(특히 핵 특성관점에서)에 대한 고찰 그리고 향후 연구과제 등에 대하여 기술하고자 한다. 비록 소멸처리 시스템의 현실화를 위해서는 해결되어야 할 과제가 많이 남아 있지만, 기술적인 가능성과 방사능의 소멸이란 면을 고려할 때 소멸처리시스템은 궁극적인 방사성핵종 처리기술로서 연구·개발할 충분한 가치가 있는 것으로 판단된다.

• Nuclear Engineering and Technology
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• v.39 no.4
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• pp.237-248
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• 2007

Sodium-cooled fast reactor is considered in France as a potential candidate for a prototype of 4th generation system to be built by 2020. A detailed working program has been launched recently to identify by 2012 the potential improvement tracks for later industrial development of these reactors. The goals for innovation are first identified: Progress of the safety with a special attention to severe accidents risk minimization and mitigation (defense in depth approach); Economic competitiveness of the system mainly by reducing the capital cost, the investment risks by enhancing in service inspection and repair capacities, and raising the availability; Sustainability with fissile material management while reducing the proliferation risk; capacity for long-lived waste transmutation.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.351-359
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• 2016

The Korean Prototype Gen-IV sodium-cooled fast reactor (PGSFR) is supposed to be loaded with a relatively-costly low-enriched U fuel, while its envisaged transuranic fuels are not available for transmutation. In this work, the U-enrichment reduction by improving the neutron economy is pursued to save the fuel cost. To improve the neutron economy of the core, a new reflector material, PbO, has been introduced to replace the conventional HT9 reflector in the current PGSFR core. Two types of PbO reflectors are considered: one is the conventional pin-type and the other one is an inverted configuration. The inverted PbO reflector design is intended to maximize the PbO volume fraction in the reflector assembly. In addition, the core radial configuration is also modified to maximize the performance of the PbO reflector. For the baseline PGSFR core with several reflector options, the U enrichment requirement has been analyzed and the fuel depletion analysis is performed to derive the equilibrium cycle parameters. The linear reactivity model is used to determine the equilibrium cycle performances of the core. Impacts of the new PbO reflectors are characterized in terms of the cycle length, neutron leakage, radial power distribution, and operational fuel cost.