• Nuclear Engineering and Technology
• /
• 제30권2호
• /
• pp.91-98
• /
• 1998

Time history analysis is usually performed to characterize the nonlinear seismic behavior of a spent fuel storage rack(SFSR). In the past, the seismic analyses of the SFSR were performed with two-dimensional planar models, which could not account for torsional response and simultaneous multi-directional seismic input In this study, three-dimensional seismic analysis methodology is developed for the single SFSR using the ANSYS code. The 3D- Model can be used to determine the nonlinear behavior of the rack, i.e., sliding, uplifting, and impact evaluation between the fuel assembly and rack, and rack and the pool wall, This paper also reviews the 3-D modeling of the SFSR and the adequacy of the ANSYS for the seismic analysis. AS a result of the adquacy study, the method of ANSYS transient analysis with acceleration time history is suitable for the seismic analysis of highly nonlinear structure such as an SFSR but it isn't appropriate to use displacement time history of seismic input.

• 방사성폐기물학회지
• /
• 제12권4호
• /
• pp.287-297
• /
• 2014

사용후핵연료 또는 고준위폐기물의 안전한 처분을 위하여 지난 수십 년 동안 많은 나라들이 다양한 처분대안을 연구하여 왔다. 본 논문에서는 심지층처분기술에 있어서 사용후핵연료를 직접 처분하는 방안으로서 처분효율 향상을 위한 다양한 방안 중의 하나로 고려할 수 있는 PWR 사용후핵연료 집합체를 해체하여 연료봉을 밀집한 경우에 대한 처분 효율을 분석하였다. 이를 위하여, 우선 사용후핵연료 연료봉 밀집개념과 관련 처분용기 및 심지층처분 개념을 설정하였다. 이 개념에 근거하여 심지층 처분시스템의 공학적방벽 설계에 있어서 가장 중요한 요건인 완충재의 온도 제한요건을 만족시키는지 여부를 확인하기 위하여 각 처분개념 별로 열해석을 수행하였다. 그리고, 처분공 간격, 처분터널 간격 및 처분용기 열발산 면적에 따른 열해석 결과를 바탕으로, 단위처분면적 관점에서의 처분효율을 비교/분석하고 평가하였다. 또한, 사용후핵연료봉을 밀집시킨 경우에 있어서 냉각기간에 따른 처분개념을 분석하였다. 분석결과에 따르면 사용후핵연료봉을 밀집하여 심지층처분하는 경우 처분효율 측면에서 불리한 것으로 판단되었다. 다만, 사용후핵연료의 냉각기간을 70년 이상으로 장기화 할 경우 처분효율은 향상될 것으로 예상되지만, 사용후핵연료의 내구성 및 장기저장에 따른 조건 등 추가적인 분석이 필요하다.

• Nuclear Engineering and Technology
• /
• 제54권1호
• /
• pp.11-18
• /
• 2022

In this research, curium oxide (CmO₂) is studied as fuel for VVER-1200 reactor to get an attention to its energy value and possibilities. For this purpose, CmO₂ is used in fuel rods or integrated burnable absorber (IBA) rods with and without UO₂ and then compared with the conventional fuel assembly of VVER-1200 reactor. It is burned to 60 GWd/t by using SRAC-2006 code and JENDL-4.0 data library. From these studies, it is found that CmO₂ is competent like UO₂ as a fuel due to higher fission cross-section of ²⁴³Cm and ²⁴⁵Cm isotopes and neutron capture cross-section of ²⁴⁴Cm and ²⁴⁶Cm isotopes. As a result, when some or all of the UO₂ of fuel rods or IBA rods are replaced by CmO₂, we get a similar k-inf like the reference even with lower enrichment UO₂ fuels. These studies show that the use of CmO₂ as IBA rods is more effective than the fuel rods considering the initially loaded amount, power peaking factor (PPF), fuel temperature and void coefficient, and the quality of spent fuel. From a detailed study, 3% CmO₂ with inert material ZrO₂ in IBA rods are recommended for the VVER-1200 reactor assembly from the once through concept.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 2003년도 춘계학술발표대회 요약집
• /
• pp.301-301
• /
• 2003

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 2005년도 춘계학술발표회
• /
• pp.1269-1270
• /
• 2005

• 한국방사성폐기물학회:학술대회논문집
• /
• 한국방사성폐기물학회 2016년도 추계학술논문요약집
• /
• pp.77-78
• /
• 2016

• Nuclear Engineering and Technology
• /
• 제53권9호
• /
• pp.2816-2829
• /
• 2021

Decay heat residuals of spent nuclear fuel (SNF), i.e., the differences between calculations and measurements, were obtained previously for various spent fuel assemblies (SFA) using the Polaris module of the SCALE code system. In this paper, we compare decay heat residuals to their uncertainties, focusing on four PWRs and four BWRs. Uncertainties in nuclear data and model inputs are propagated stochastically through calculations using the SCALE/Sampler super-sequence. Total uncertainties could not explain the residuals of two SFAs measured at GE-Morris. The combined z-scores for all SFAs measured at the Clab facility could explain the resulting deviations. Nuclear-data-related uncertainties contribute more in the high burnup SFAs. Design and operational uncertainties tend to contribute more to the total uncertainties. Assembly burnup is a relevant variable as it correlates significantly with the SNF decay heat. Additionally, burnup uncertainty is a major contributor to decay heat uncertainty, and assumptions relating to these uncertainties are crucial. Propagation of nuclear data and design and operational uncertainties shows that the analyzed assemblies respond similarly with high correlation. The calculated decay heats are highly correlated in the PWRs and BWRs, whereas lower correlations were observed between decay heats of SFAs that differ in their burnups.

• 한국방사성폐기물학회:학술대회논문집
• /
• 한국방사성폐기물학회 2003년도 가을 학술논문집
• /
• pp.590-595
• /
• 2003

An Exponential experiment system which is composed of a neutron detector, a signal analysis system and a neutron source, Cf-252 has been installed in order to experimentally determine the neutron effective multiplication factor for a PWR spent fuel assembly. The axial background neutron flux is measured in a preliminary performance test. From the results, the spacer grid position is determined to be consistent with the design specifications within a 2.3% relative error. The induced fission neutron for four of the assemblies is also measured by scanning the neutron source, Cf-252 or the neutron detector. The exponential decay constants have been evaluated by the application of the Poisson regression to the net induced fission neutron counts. The measured keffs determined on the basis of the exponential decay constants of Cl5 appeared to be 0.541, 0.540, 0.597 and 0.556, respectively, which are comparable with 0.55195$\pm$0.00232 of the MCNP calculation.

• Nuclear Engineering and Technology
• /
• 제53권8호
• /
• pp.2454-2463
• /
• 2021

The objective of this research is to the use of americium (AmO₂) as a burnable absorber effectively instead of conventional gadolinium (Gd₂O₃) for VVER-1200 reactor by analyzing its impacts on reactivity, power peaking factor (PPF), safety factor, and quality of the spent fuel. The assembly is burned to 60 GWd/t by using SRAC-2006 code and JENDL-4.0 data library for finding the optimum amount and effective way of using AmO₂ as a burnable absorber. From these studies, it is found that AmO₂ can decrease the excess reactivity like Gd₂O₃ without changing the criticality life span and enrichment of ²³⁵U. A homogeneous mixture of the 0.20% AmO₂+ 4.95% enriched UO₂ fuel rod (model MF-4) decreases the PPF than the reference assembly. The use of AmO₂+UO₂ in the integral burnable absorber (IBA) rod or the outer layer could also decrease the PPF up to 10 GWd/t but increases rapidly after 30 GWd/t, which could be a safety threat. The fuel temperature coefficient and void coefficient of the model MF-4 are the same as the reference assembly. In addition, 22% of initially loaded Am are burning effectively and contributing to the power production.

• 대한기계학회논문집B
• /
• 제40권12호
• /
• pp.815-824
• /
• 2016

가압경수로에 장전되는 핵연료집합체는 연료 봉 다발과 지지격자 및 상하단 고정체로 구성되어 있다. 고온 고압의 냉각수는 원자로 하부로 유입되어 연료 봉 사이로 형성된 부수로를 따라 노심 상부로 흐른다. 경수로핵연료의 주요 열수력 성능인자는 정상운전시 압력강하 및 임계열속이며 사고시에는 급랭 시간이다. 한국원자력연구원에서는 경수로핵연료의 성능을 향상시키고 국산화를 위해 고성능 경수로핵연료, 이중냉각 핵연료 및 사고저항성 핵연료를 개발하였다. 경수로핵연료의 열수력 핵심기술을 개발하기 위해 압력강하 실험, 난류 유동혼합/열전달 실험, 임계열속 및 급랭 시험을 수행하였으며 전산유체역학 방법도 활용하였다. 더불어 사용후핵연료의 임시저장을 위한 건식저장 용기의 열유동에 대한 전산유체해석을 수행하였다. 한편, 경수로핵연료의 열수력 기반기술을 개발하고 실용화를 위해 대학 및 산업체와 협력연구도 진행하였다.