• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.4
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• pp.363-373
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• 2019

Waste characterisation associated with nuclear site decommissioning relies on radiochemical analysis of a diverse range of sample types, requiring extensive validation of analytical techniques using matrix-matched materials. The absence of relevant reference materials has hindered robust method development and validation. The paper discusses how method validation in support of nuclear waste characterisation can be achieved without using reference materials. The key stages in an analytical procedure are evaluated and a multi-stage approach is proposed with the ultimate aim of determining an operational envelope for an analytical procedure.

• Nuclear Engineering and Technology
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• v.51 no.8
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• pp.1964-1969
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• 2019

The safety and efficiency of nuclear reactors largely depend on the monitoring and control of nuclear radiation. Due to the unique nuclear-physical characteristics, Hf is one of the most promising materials for the manufacturing of the control rods and the emitters of neutron detectors. It is proposed to use the Compton neutron detector with the emitter made of Hf in the In-core Instrumentation System (ICIS) for monitoring the neutron field. The main advantages of such a detector in comparison the conventional β-emission sensors are the possibility of reaching of a higher cumulative radiation dose and the absence of signal delays. The response time of the detection is extremely important when a nuclear reactor is operating near its critical operational parameters. Taking Hf as an example, the general principles for calculating the chains of materials transformation under neutron irradiation are reported. The influence of ^179m1Hf on the Hf composition changing dynamics and the process of transmutants' (Ta, W) generation were determined. The effect of these processes on the absorbing properties of Hf, which inevitably predetermine the lifetime of the detector and its ability to generate a signal, is estimated.

• Korean Journal of Materials Research
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• v.21 no.11
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• pp.596-603
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• 2011

The very high temperature gas reactor (VHTR) is one of the next generation nuclear reactors for its safety, long-term stability, and proliferation-resistance. The high operating temperature of over 800$^{\circ}C$ enables various applications with high energy efficiency. Heat is transferred from the primary helium loop to the secondary helium loop through the intermediate heat exchanger (IHX). The IHX material requires creep resistance, oxidation resistance, and corrosion resistance in a helium environment at high operating temperatures. A Ni-based superalloy such as Alloy 617 is considered as a primary candidate material for the intermediate heat exchanger. In this study, the microstructures of Alloy 617 crept in pure helium and air environments at 950$^{\circ}C$ were observed. The rupture time in helium was shorter than that in air under small applied stresses. As the exposure time increased, the thickness of outer oxide layer of the specimens clearly increased but delaminated after a long creep time. The depth of the carbide-depleted zone was rather high in the specimens under high applied stress. The reason was elucidated by the comparison between the ruptured region and grip region of the samples. It is considered that decarburization caused by minor gas impurities in a helium environment caused the reduction in creep rupture time.

• Nuclear Engineering and Technology
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• v.55 no.11
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• pp.4002-4018
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• 2023

An integrated neutron interrogation system has been developed for non-destructive assay of highly-radioactive special nuclear materials, to accumulate knowledge of the method through developing and using it. The system combines a differential die-away (DDA) measurement system for the quantification of nuclear materials and a prompt gamma-ray analysis (PGA) system for the detection of neutron poisons which disturb the DDA measurements; a common D-T neutron generator is used. A special care has been taken for the selection of materials to reduce the background gamma rays produced by the interrogation neutrons. A series of measurements were performed to test the basic performance of the system. The results show that the DDA system can quantify plutonium of as small as 20 mg and it is not affected by intense neutron background up to 1.57 × 10⁷ s^-1 and gamma ray of 4.43 × 10¹⁰ s^-1. The gamma-ray background counting rate at the PGA detector was reduced down to 3.9 × 10³ s^-1 even with the use of the D-T neutron generator. The test measurements show that the PGA system is capable of detecting 0.783 g of boron and about 86.8 g of gadolinium in 30 min.

• 한국초전도학회:학술대회논문집
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• v.14
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• pp.95-95
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• 2004

• Proceedings of the Korean Powder Metallurgy Institute Conference
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• 2006.09b
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• pp.967-968
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• 2006

This research reports for the successful consolidation of $Al_2O_3$ powder with retained ultra-fine structure using MPC and sintering. Measurements in the consolidated $Al_2O_3$ bulk indicated that hardness, fracture toughenss, and breakdown voltage have been much improved relative to the conventional polycrystalline materials. Finally, optimization of the compaction parameters and sintering conditions will lead to the consolidation of $Al_2O_3$ nanopowder with higher density and even further enhanced mechanical properties.

• Journal of the Korean Ceramic Society
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• v.54 no.1
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• pp.9-22
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• 2017

Many new functional materials have been studied for efficient production and storage of energy. Many new materials such as sodium-based and sulfide-based materials have been proposed for energy storage, but research on Li batteries is still dominant. Due to the influence of environmental concerns regarding nuclear energy, interest in and research on fusion power are steadily increasing. For the commercialization of nuclear fusion, a design standard based on a considerable level of physical analysis and modeling is proposed. Nevertheless, limitations of existing materials in nuclear fusion environments limit practical applications. Tritium propagation material for continuous fusion reaction is one of the core materials, and therefore research on this material is being carried out intermittently. The key material for Li-based energy storage and tritium generation is the functional material Li-M-O. In this review, a structural description of functional Li-M-O system materials and technical trends for its applications are introduced.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.667-668
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• 2005

• Carbon letters
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• v.10 no.4
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• pp.300-304
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• 2009

Thermal emissivity of nuclear graphite was measured with its oxidation degree. Commercial nuclear graphites (IG-110, PECA, IG-430, and NBG-18) have been used as samples. Concave on graphites surface increased as its oxidation degree increased, and R value (Id/Ig) of the graphites decreased as the oxidation degree increased. The thermal emissivity increased depending on the decrease of the R (Id/Ig) value through Raman spectroscopy analysis. It was determined that the thermal emissivity was influenced by the crystallinity of the nuclear graphite.

• Journal of the Korean Ceramic Society
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• v.43 no.10 s.293
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• pp.613-618
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• 2006

Graphite is suitable materials as a moderator, reflector, and supporter of a nuclear reactor because of high tolerance to the high temperature and neutron irradiations. Because graphite is so weak to the oxidation, its oxidation study is essentially demanded for the operation and design of the nuclear reactor. This work focuses on the effect of the surface oxidation of graphite according to the surface treatment. With thermogravimeter (TG), oxidation characteristics of the isotropic graphite are measured at the three temperature areas, and oxidation ratio and amounts are estimated as changing the surface roughness. Furthermore, the polished graphite surface produced fom the surface treatment is investigated with the Raman spectroscopic study. Oxidation behaviors of the surface are also evaluated as elimination the polished layer by washing with strong sonication.