• Applied Chemistry for Engineering
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• v.9 no.2
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• pp.238-242
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• 1998

Wood powder filled phenolic resin composites of different composition were prepared and their mechanical properties were investigated for optimum conditions. The composites showed maximum mechanical strength when the phenolic resin content was 45 wt%. Polyurethane prepolymer(PU) was evaluated as a modifier of the phenolic resin composites. Blocking of the isocyanates in the PU with phenol was necessary for homogeneous mixing of raw materials for the components. Maximum mechanical strength of the PU modified composites was observed when the PU content was 5 wt%. It was found that the mechanical strength of the composites cured at $210^{\circ}C$ were higher than those of the composites cured at $150^{\circ}C$.

• Nuclear Engineering and Technology
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• v.26 no.2
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• pp.247-256
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• 1994

The nuclide chain model used in SAV90 has been modified for the CASMO3/MEDIUM3 procedure. Since the default nuclide chain in SAV90, using 21 nuclides, is not sufficient to reproduce the CASMO3 results in the MEDIUM3 calculation, the extended nuclide chain models have been investigated and verified with various types of fuel assemblies. Among the extended nuclide chain models proposed, the 22 nuclide chain model, which contains only Pu238 additionally to the 21 nuclide chain, is recommended in terms of both accuracy and computing efficiency. Using this model core follow calculations for YGN-1 have been performed. The results showed good performance when compared to plant measurements.

• Nuclear Engineering and Technology
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• v.41 no.1
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• pp.99-106
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• 2009

The contents of transuranic elements ($^{237}Np$, $^{238}Pu$, $^{239}Pu$, $^{240}Pu$, $^{241}Am$, $^{244}Cm$, and $^{242}Cm$) in high-burnup spent fuel samples ($35.6{\sim}53.9\;GWd/MtU$) were determined by alpha spectrometry. Anion exchange chromatography and diethylhexyl phosphoric acid extraction chromatography were applied for the separation of these elements from the uranium matrix. The measured values of the nuclides were compared with ORIGEN-2 calculations. For plutonium, the measurements were higher than the calculations by about $2.6{\sim}32.7%$ on average according to each isotope, and those for americium and curium were also higher by about $35.9{\sim}63.1%$. However, for $^{237}Np$, the measurements were lower by about 52% on average for the samples.

• Nuclear Engineering and Technology
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• v.23 no.4
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• pp.387-400
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• 1991

An ocean compartment model simulating transport of nuclides by advection due to ocean circulation and intertaction with suspended sediments is developed, by which concentration breakthrough curves of nuclides can be calculated as a function of time. Dividing ocean into arbitrary number of characteristic compartments and performing a balance of mass of nuclides in each ocean compartment, the governing equation for the concentration in the ocean is obtained and a solution by the numerical integration is obtained. The integration method is specially useful for general stiff systems. For transfer coefficients describing advective transport between adjacent compartments by ocean circulation, the ocean turnover time is calculated by a two-dimensional numerical ocean model. To exemplify the compartment model, a reference case calculation for breakthrough curves of three nuclides in low-level radioactive wastes, Tc-99, Cs-137, and Pu-238 released from hypothetical repository under the seabed is carried out with five ocean compartments. Sensitivity analysis studies for some parameters to the concentration breakthrough curves are also made, which indicates that parameters such as ocean turnover time and ocean water volume of compartments have an important effect on the breakthrough curves.

• Elastomers and Composites
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• v.47 no.3
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• pp.238-245
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• 2012

Graphene oxide was synthesized from natural graphite, and its surface was modified using diisocyanatodicyclohexylmethane( $H_{12}MDI$). Isocyanate-graphene sheet(i-RGO) was obtained by reduction of surface modified GO. To select nanofiller having good dispersion with polyurethane, GO, i-RGO, natural graphite and thermal reduced graphite were analyzed, and then i-RGO was selected as a suitable nanofiller. PU/i-RGO nanocomposite was synthesized with various i-RGO contents to estimate effect of reinforcement on nanocomposite. Thermal stability, hardness, contact angle were increased with i-RGO contents due to i-RGO characteristic and crosslink bridge effect. But, tensile strength and elongation were decreased at i-RGO contents more than the 4 wt%. This phenomenon was interpreted by the excess formation of crosslink bridge.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 2000.11a
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• pp.173-181
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• 2000

IAEA자료에 의하면 원자력 발전용 원자로는 1998년말 현재 세계 32개국에서 434기가 운전중이며, 총 출력은 3억 4889만kW인 것으로 나타났고, 이는 세계 총 발전량의 17%를 담당하는 것으로 확인되었다. 그러나 농축 우라늄 고체 핵연료를 사용하는 발전로 개념은 근본적으로 핵물질 SEU(Slightly Enriched Uranium)를 생산하기 위한 235U 농축과 노내에서 238U의 중성자 포획으로 전환.생성되는 Pu의 누적에 따른 핵확산 우려, 고준위 방사성 폐기물로 취급되는 사용후 핵연료 처리.처분에 관한 정책적.기술적 장기 전망의 불확실성, 그리고 설계기준사고인 LOCA로부터 중대사고로 이어지는 안전성 문제 등이 대두되고 있다. Th$^{233}$ /U용융염 핵연료주기를 이용하는 발전로 개념은 원자력 발전이 안고있는 고유문제들을 배제 또는 완화할 수 있는 방안으로 고려되고 있다.(중략)

• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.299-312
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• 2014

An uncertainty propagation methodology based on the Monte Carlo method is applied to PWR nuclear design analysis to assess the impact of nuclear data uncertainties. The importance of the nuclear data uncertainties for $^{235,238}U$, $^{239}Pu$, and the thermal scattering library for hydrogen in water is analyzed. This uncertainty analysis is compared with the design and acceptance criteria to assure the adequacy of bounding estimates in safety margins.

• Nuclear Engineering and Technology
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• v.42 no.1
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• pp.79-88
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• 2010

Spent $UO_2$ nuclear fuel discharged from a nuclear power plant (NPP) contains fission products, U, Pu, and other actinides. Due to neutron capture by $^{238}U$ in the rim region and a temperature gradient between the center and the rim of a fuel pellet, a considerable increase in the concentration of fission products, Pu, and other actinides are expected in the pellet periphery of high burnup fuel. The characterization of the radial profiles of the various isotopic concentrations is our main concern. For an analysis, spent nuclear fuels originating from the Yeonggwang-2 pressurized water reactor (PWR) were chosen as the test specimens. In this work, the distributions of some actinide isotopes were measured from center to rim of the spent fuel specimens by a radiation shielded laser ablation inductively coupled plasma mass spectrometer (LA-ICP-MS) system. Sampling was performed along the diameter of the specimen by reducing the sampling intervals from 500 ${\mu}m$ in the center to 100 ${\mu}m$ in the pellet periphery region. It was observed that the isotopic concentration ratios for minor actinides in the center of the specimen remain almost constant and increase near the pellet periphery due to the rim effect apart from the $^{236}U$ to $^{235}U$ ratio, which remains approximately constant. In addition, the distributions of local burnup were derived from the measured isotope ratios by applying the relationship between burnup and isotopic ratio for plutonium and minor actinides calculated by the ORIGEN2 code.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2733-2742
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• 2020

This paper presents an assessment of applicability of the multigroup cross sections generated with Monte Carlo tools to the fast reactor analysis based on transport calculations. 33-group cross section sets were generated for simple one- (1-D) and two-dimensional (2-D) sodium-cooled fast reactor problems using the SERPENT code and applied to deterministic steady-state and depletion calculations. Relative to the reference continuous-energy SERPENT results, with the transport corrected P₀ scattering cross section, the k-eff value was overestimated by 506 and 588 pcm for 1-D and 2-D problems, respectively, since anisotropic scattering is important in fast reactors. When the scattering order was increased to P₅, the 1-D and 2-D problem errors were increased to 577 and 643 pcm, respectively. A sensitivity and uncertainty analysis with the PERSENT code indicated that these large k-eff errors cannot be attributed to the statistical uncertainties of cross sections and they are likely due to the approximate anisotropic scattering matrices determined by scalar flux weighting. The anisotropic scattering cross sections were alternatively generated using the MC²-3 code and merged with the SERPENT cross sections. The mixed cross section set consistently reduced the errors in k-eff, assembly powers, and nuclide densities. For example, in the 2-D calculation with P3 scattering order, the k-eff error was reduced from 634 pcm to -223 pcm. The maximum error in assembly power was reduced from 2.8% to 0.8% and the RMS error was reduced from 1.4% to 0.4%. The maximum error in the nuclide densities at the end of 12-month depletion that occurred in ²³⁷Np was reduced from 3.4% to 1.5%. The errors of the other nuclides are also reduced consistently, for example, from 1.1% to 0.1% for ²³⁵U, from 2.2% to 0.7% for ²³⁸Pu, and from 1.6% to 0.2% for ²⁴¹Pu. These results indicate that the scalar flux weighted anisotropic scattering cross sections of SERPENT may not be adequate for application to fast reactors where anisotropic scattering is important.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2803-2815
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• 2021

This paper addresses the uncertainty quantification and sensitivity analysis of a depleted light-water fuel assembly of the Turkey Point-3 benchmark. The uncertainty of the fuel assembly decay heat and isotopic densities is quantified with respect to three different groups of diverse parameters: nuclear data, assembly design, and reactor core operation. The uncertainty propagation is conducted using a two-step analysis code system comprising the lattice code STREAM, nodal code RAST-K, and spent nuclear fuel module SNF through the random sampling of microscopic cross-sections, fuel rod sizes, number densities, reactor core total power, and temperature distributions. Overall, the statistical analysis of the calculated samples demonstrates that the decay heat uncertainty decreases with the cooling time. The nuclear data and assembly design parameters are proven to be the largest contributors to the decay heat uncertainty, whereas the reactor core power and inlet coolant temperature have a minor effect. The majority of the decay heat uncertainties are delivered by a small number of isotopes such as ²⁴¹Am, ¹³⁷Ba, ²⁴⁴Cm, ²³⁸Pu, and ⁹⁰Y.