• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.111-112
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• 2004

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 요약집 춘계학술발표대회
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• pp.64-64
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• 2004

• 방사성폐기물학회지
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• 제16권2호
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• pp.195-202
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• 2018

원칙적으로, 지층 처분은 고준위 방사성 폐기물의 최종 처분을 위한 안전한 방법으로 간주된다. 그러나 사용후핵연료에 함유된 $^{99}Tc$ 및 $^{129}I$와 같은 일부 장수명 핵분열 생성물은 지하 환경에서 흡수성이 적은 음이온 핵종으로 이동성이 매우 크며 수백 keV 범위의 베타선 방출로 생태계에 피폭선량을 야기시킬 수 있다. 따라서 이 두 핵종을 효율적으로 분리하여 방사능으로 유해하지 않은 핵종으로 전환할 수 있다면 처분 안정성에 긍정적인 영향을 줄 수 있다. 이를 위한 하나의 방법은 이 두 가지 핵종을 원자로에서 수명이 짧은 핵종 또는 안정적인 핵종으로 변환하는 것이다. 이를 위해 두 핵종을 태우는 데 어느 원자로 유형이 더 효율적인지 평가하는 것이 필요하다. 본 연구에서는 경수로(PWR), 중수로(CANDU) 및 고속로(SFR, MET-1000)의 $^{99}Tc$ 및 $^{129}I$의 핵 변환 시뮬레이션 결과를 비교하고 고찰하였다.

• Nuclear Engineering and Technology
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• 제34권1호
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• pp.42-59
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• 2002

In order to transmute long-lived radioactive nuclides such as transuranics(TRU), Tc-99, and I- l29 in LWR spent fuel, a preliminary conceptual design study has been performed for the accelerator driven subcritical reactor system, called HYPER(Hybrid Power Extraction Reactor) The core has a hybrid neutron energy spectrum: fast and thermal neutrons for the transmutation of TRU and fission products, respectively. TRU is loaded into the HYPER core as a TRU-Zr metal form because a metal type fuel has very good compatibility with the pyre- chemical process which retains the self-protection of transuranics at all times. On the other hand, Tc-99 and I-129 are loaded as pure technetium metal and sodium iodide, respectively. Pb-Bi is chosen as a primary coolant because Pb-Bi can be a good spallation target and produce a very hard neutron energy spectrum. As a result, the HYPER system does not have any independent spallation target system. 9Cr-2WVTa is used as a window material because an advanced ferritic/martensitic steel is known to have a good performance under a highly corrosive and radiation environment. The support ratios of the HYPER system are about 4∼5 for TRU, Tc-99, and I-129. Therefore, a radiologically clean nuclear power, i.e. zero net production of TRU, Tc-99 and I-129 can be achieved by combining 4 ∼5 LWRs with one HYPER system. In addition, the HYPER system, having good proliferation resistance and high nuclear waste transmutation capability, is believed to provide a breakthrough to the spent fuel problems the nuclear industry is faced with.

• Nuclear Engineering and Technology
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• 제42권4호
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• pp.442-449
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• 2010

Recently, the neutron irradiation for large diameter silicon (Si)-ingots of more than 8" diameter is requested to satisfy the demand for the neutron transmutation doping silicon (NTD-Si). By increasing the Si-ingot diameter, the radial non-uniformity becomes larger due to the neutron attenuation effect, which results in a limit of the feasible diameter of the Si-ingot. The current evaluation method has a certain limit to precisely evaluate the radial uniformity of Si-ingot because the current evaluation method does not consider the effect of the Si-ingot diameter on the radial uniformity. The objective of this study is to propose a new evaluation method of radial uniformity by improving the conventional evaluation approach. To precisely predict the radial uniformity of a Si-ingot with large diameter, numerical verification is conducted through comparison with the measured data and introducing the new evaluation method. A new concept of a gradient is introduced as an alternative approach of radial uniformity evaluation instead of the radial resistivity gradient (RRG) interpretation. Using the new concept of gradient, the normalized reaction rate gradient (NRG) and the surface normalized reaction rate gradient (SNRG) are described. By introducing NRG, the radial uniformity can be evaluated with one certain standard regardless of the ingot diameter and irradiation condition. Furthermore, by introducing SNRG, the uniformity on the Si-ingot surface, which is ignored by RRG and NRG, can be evaluated successfully. Finally, the radial uniformity flattening methods are installed by the stainless steel thermal neutron filter and additional Si-pipe to reduce SNRG.

• Nuclear Engineering and Technology
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• 제46권3호
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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.

• Nuclear Engineering and Technology
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• 제51권4호
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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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• 제37권5호
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• pp.433-446
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• 2005

The development of accelerator science and technology has been accommodating ever increasing demand from scientific community of the beam energy and intensity of proton beams. The use of high-powered proton beams has extended from the traditional application of nuclear and high-energy physics to other applications, including spallation neutron source replacing nuclear reactor, nuclear actinide transmutation, energy amplification reactors. This article attempts to review development of proton accelerator, both linear and circular, and issues related to the proton beam energy, intensity as well as its output power. For related accelerator physics and technical review, one should refer to the recent article in the Reviews of Modem Physics [1]

• Nuclear Engineering and Technology
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• 제48권2호
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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.

• Nuclear Engineering and Technology
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• 제55권5호
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• pp.1871-1877
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• 2023

The main goal of the coordinated project development of therapeutic radiopharmaceuticals of Y-90 and Re-188 is to exploit advancements in radionuclide production technology. Here, direct and indirect production methods with medium reactor and cyclotron are compared to evaluate derived neutron flux and production yield. First, nano-sized ¹⁸⁶W and ⁸⁹Y specimens are suspended in water in a quartz vial by FLUKA simulation. Then, the solution is irradiated for 4 days under 9E+14 n/cm2/s neutron flux of reactor. Also, a neutron activator including three layers-lead moderator, graphite reflector, and polyethylene absorbent- is simulated and tungsten target is irradiated by 60 MeV protons of cyclotron to generate induced neutrons for ¹⁸⁸W and ⁹⁰Sr production via neutron capture. As the neutron energy reduced, the flux gradually increased towards epithermal range to satisfy (n/2n,γ) reactions. The obtained specific activities at saturation were higher than the reported experimental values because the accumulated epithermal flux and nano-sized specimens influence the outcomes. The beta emitters, which are widely utilized in brachytherapy, appeal an alternative route to locally achieve a rational yield. Therefore, the proposed method via neutron activator may ascertain these broad requirements.