• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3449-3459
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• 2021

With the increasing availability of high-performance computing (HPC) platforms, uncertainty quantification (UQ) and sensitivity analyses (SA) can be efficiently leveraged to optimize design parameters of complex engineering problems using modeling and simulation tools. The workflow involved in such studies heavily relies on HPC resources and hence requires pre-processing and post-processing capabilities of large amounts of data along with remote submission capabilities. The NEAMS Workbench addresses all aspects of the workflows involved in these studies by relying on a user-friendly graphical user interface and a python application program interface. This paper highlights the NEAMS Workbench capabilities by presenting a semiautomated coupling scheme between Dakota and any given package integrated with the NEAMS Workbench, yielding a simplified workflow for users. This new capability is demonstrated by running a SA of a turbulent flow in a pipe using the open-source Nek5000 CFD code. A total of 54 jobs were run on a HPC platform using the remote capabilities of the NEAMS Workbench. The results demonstrate that the semiautomated coupling scheme involving Dakota can be efficiently used for UQ and SA while keeping scripting tasks to a minimum for users. All input and output files used in this work are available in https://code.ornl.gov/neams-workbench/dakota-nek5000-study.

• Nuclear Engineering and Technology
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• 제51권7호
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• pp.1835-1841
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• 2019

The mass attenuation coefficient ${\mu}_m$ for eight rock samples having different chemical composition was simulated using the MCNP 5 code in energy range($0.002MeV{\leq}E{\leq}10MeV$). Moreover, the ${\mu}_m$ for the studied rock samples was computed theoretically using XCOM database. The comparison between simulated and computed data for all selected rock samples showed a good agreement with differences varied between 0.01 and 8%. The highest ${\mu}_m$ was found for basalt rocks M2 and M1 and the lowest one is reported for limestone rocks Dike. The simulated values of the ${\mu}_m$ then were used to calculate other important shielding parameters such as the mean free path, effective electron density and effective atomic number. The exposure buildup factor EBF was also computed for the selected rocks with the contribution of G-P fitting parameters and the highest EBF attended by the basalt sample Sill and varied between 1.022 and 744 in the energy range between ($0.015MeV{\leq}E{\leq}15MeV$) but the lowest EBF achieved by basalt sample M2 and varied between 1.017 and 491 in the same energy range.

• Geomechanics and Engineering
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• 제25권2호
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• pp.111-121
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• 2021

Large deformation and rapid pressure propagation take place inside the rock mass under the dynamic loads caused by the explosives, on quarry faces in order to extract aggregate material. The complexity of the science of rock blasting is due to a number of factors that affect the phenomenon. However, blasting engineering computations could be facilitated by innovative software algorithms in order to determine the results of the violent explosion, since field experiments are particularly difficult to be conducted. The present research focuses on the design of a Finite Element Analysis (FEA) code, for investigating in detail the behavior of limestone under the blasting effect of Ammonium Nitrate & Fuel Oil (ANFO). Specifically, the manuscript presents the FEA models and the relevant transient analysis results, simulating the blasting process for three types of limestone, ranging from poor to very good quality. The Finite Element code was developed by applying the Jones-Wilkins-Lee (JWL) equation of state to describe the thermodynamic state of ANFO and the pressure dependent Drucker-Prager failure criterion to define the limestone plasticity behavior, under blasting induced, high rate stress. A progressive damage model was also used in order to define the stiffness degradation and destruction of the material. This paper performs a comparative analysis and quantifies the phenomena regarding pressure, stress distribution and energy balance, for three types of limestone. The ultimate goal of this research is to provide an answer for a number of scientific questions, considering various phenomena taking place during the explosion event, using advanced computational tools.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.2771-2782
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• 2022

When assessing the radiological consequences of postulated accident scenarios, it is of primary interest to determine the amount of radioactive fission gas accumulated in the fuel rod free volume. The state-of-the-art semi-empirical approach (ANS 5.4-2010) is reviewed and compared with a mechanistic approach to evaluate the release of radioactive fission gases. At the intra-granular level, the diffusion-decay equation is handled by a spectral diffusion algorithm. At the inter-granular level, a mechanistic description of the grain boundary is considered: bubble growth and coalescence are treated as interrelated phenomena, resulting in the grain-boundary venting as the onset for the release from the fuel pellets. The outcome is a kinetic description of the release of radioactive fission gases, of interest when assessing normal and off-normal conditions. We implement the model in SCIANTIX and reproduce the release of short-lived fission gases, during the CONTACT 1 experiments. The results show a satisfactory agreement with the measurement and with the state-of-the-art methodology, demonstrating the model soundness. A second work will follow, providing integral fuel rod analysis by coupling the code SCIANTIX with the thermo-mechanical code TRANSURANUS.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.19-29
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• 2022

The 2D/1D method has become the mainstream of the direct transport calculation considering the balance of accuracy and efficiency. However, the 2D/1D method still suffers from stability issues. Recently, a quasi-3D method has been proposed with axial Legendre expansion. Analysis and comparison of the 2D/1D and quasi-3D method is conducted in theory from the equation derivation. Besides, the C5G7 benchmark, the KUCA benchmark and the macro BEAVRS benchmark are calculated to verify the theory comparisons of these two methods with the direct transport code SHARK. All results show that the quasi-3D method has better stability and accuracy than the 2D/1D method with worse efficiency and memory cost. It provides a new option for direct transport calculation with the quasi-3D method.

• Nuclear Engineering and Technology
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• 제55권5호
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• pp.1775-1782
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• 2023

Several reactor physics commissioning experiments were conducted to obtain the neutronic parameters at the beginning of the G.A. Siwabessy Multi-purpose Reactor (RSG-GAS) operation. These parameters are essential for the reactor to safety operate. Leveraging the experimental data, this study evaluated the calculated core reactivity, control rod reactivity worth, integral control rod reactivity curve, and fuel reactivity. Calculations were carried out with Serpent 2 code using the latest neutron cross-section data ENDF/B-VIII.0. The criticality calculations were carried out for the RSG-GAS first core up to the third core configuration, which has been done experimentally during these commissioning periods. The excess reactivity for the second and third cores showed a difference of 510.97 pcm and 253.23 pcm to the experiment data. The calculated integral reactivity of the control rod has an error of less than 1.0% compared to the experimental data. The calculated fuel reactivity value is consistent with the measured data, with a maximum error of 2.12%. Therefore, it can be concluded that the RSG-GAS reactor core model is in good agreement to reproduce excess reactivity, control rod worth, and fuel element reactivity.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집E
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• pp.386-391
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• 2001

A finite element code based on P2P1 tetra element has been developed for the large eddy simulation (LES) of turbulent flows around a complex geometry. Fractional 4-step algorithm is employed to obtain time accurate solution since it is less expensive than the integrated formulation, in which the velocity and pressure fields are solved at the same time. Crank-Nicolson method is used for second order temporal discretization and Galerkin method is adopted for spatial discretization. For very high Reynolds number flows, which would require a formidable number of nodes to resolve the flow field, SUPG (Streamline Upwind Petrov-Galerkin) method is applied to the quadratic interpolation function for velocity variables, Noting that the calculation of intrinsic time scale is very complicated when using SUPG for quadratic tetra element of velocity variables, the present study uses a unique intrinsic time scale proposed by Codina et al. since it makes the present three-dimensional unstructured code much simpler in terms of implementing SUPG. In order to see the effect of numerical diffusion caused by using an upwind scheme (SUPG), those obtained from P2P1 Galerkin method and P2P1 Petrov-Galerkin approach are compared for the flow around a sphere at some Reynolds number. Smagorinsky model is adopted as subgrid scale models in the context of P2P1 finite element method. As a benchmark problem for code validation, turbulent flows around a sphere and a MIRA model have been studied at various Reynolds numbers.

• 경영과학
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• 제11권2호
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• pp.1-27
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• 1994

A reliability data processing MPRDP (Multi-Purpose Reliability Data Processor) has been developed in FORTRAN language since Jan. 1992 at KAERI (Korean Atomic Energy Research Institute). The purpose of the research is to construct a reliability database(plant-specific as well as generic) by processing various kinds of reliability data in most objective and systematic fashion. To account for generic estimates in various compendia as well as generic plants' operating experience, we developed a 'three-stage' Bayesian procedure[1] by logically combining the 'two-stage' procedure[2] and the idea for processing generic estimates[3]. The first stage manipulates generic plant data to determine a set of estimates for generic parameters,e.g. the mean and the error factor, which accordingly defines a generic failure rate distribution. Then the second stage combines these estimates with the other ones proposed by various generic compendia (we call these generic book type data). This stage adopts another Bayesian procedure to determine the final generic failure rate distribution which is to be used as a priori distribution in the third stage. Then the third stage updates the generic distribution by plant-specific data resulting in a posterior failure rate distribution. Both running failure and demand failure data can be handled in this code. In accordance with the growing needs for a consistent and well-structured reliability database, we constructed a generic reliability database by the MPRDP code[4]. About 30 generic data sources were reviewed and available data were collected and screened from them. We processed reliability data for about 100 safety related components frequently modeled in PSA. The underlying distribution for the failure rate was assumed to be lognormal or gamma, according to the PSA convention. The dependencies among the generic sources were not considered at this time. This problem will be approached in further study.

• Journal of Radiation Protection and Research
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• 제41권2호
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• pp.155-159
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• 2016

Background: Nuclear reactors produce a great number of antielectron neutrinos mainly from beta-decay chains of fission products. Such neutrinos have energies mostly in MeV range. We are interested in neutrinos in a region of keV, since they may take part in special weak interactions. We calculate reactor antineutrino spectra especially in the low energy region. In this work we present neutrino spectrum from a typical pressurized water reactor (PWR) reactor core. Materials and Methods: To calculate neutrino spectra, we need information about all generated nuclides that emit neutrinos. They are mainly fission fragments, reaction products and trans-uranium nuclides that undergo negative beta decay. Information in relation to trans-uranium nuclide compositions and its evolution in time (burn-up process) were provided by a reactor code MVP-BURN. We used typical PWR parameter input for MVP-BURN code and assumed the reactor to be operated continuously for 1 year (12 months) in a steady thermal power (3.4 GWth). The PWR has three fuel compositions of 2.0, 3.5 and 4.1 wt% $^{235}U$ contents. For preliminary calculation we adopted a standard burn-up chain model provided by MVP-BURN. The chain model treated 21 heavy nuclides and 50 fission products. The MVB-BURN code utilized JENDL 3.3 as nuclear data library. Results and Discussion: We confirm that the antielectron neutrino flux in the low energy region increases with burn-up of nuclear fuel. The antielectron-neutrino spectrum in low energy region is influenced by beta emitter nuclides with low Q value in beta decay (e.g. $^{241}Pu$) which is influenced by burp-up level: Low energy antielectron-neutrino spectra or emission rates increase when beta emitters with low Q value in beta decay accumulate Conclusion: Our result shows the flux of low energy reactor neutrinos increases with burn-up of nuclear fuel.

• 한국화재소방학회논문지
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• 제15권1호
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• pp.47-54
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• 2001

90년대 후반기부터 연이어 발생한 씨랜드 화재사고와 인천 인현동 라이브클럽 화재등의 동일한 유형을 가진 화재사건을 계기로 국내 다중이용업소의 실내장식재 및 장식물품에 사용하는 재료가 화재시 인명참사에 미치는 영향이 크다는 점이 발견되었다. 따라서 국내다중이용업소를 중심으로 실내장식재 및 장식물품의 사용과 관련한 화재안전규정 및 기준을 조사한 후에 유럽·미국 등의 선진국의 관련 규정을 조사 비교하여 그 문제점을 세부적으로 발췌하여 규제방향 및 규제기준의 상이점을 분석하여 대형인명참사를 초래하는 근본적인 원인을 찾아내었다. 고찰된 원인을 근간으로 실내장식재 및 장식물품에 사용하는 재료가 선진국 수준의 방재성능을 가질수 있도록 규제방향을 제시하고 제시된 규제방향을 뒷받침할 수 있는 효과적이고 검증된 시험방법을 연구제시하였다. 최종적으로 향후 발생가능한 대형인명참사를 효과적으로 방지하기 위해 다중이용업소를 중심으로 방화 및 방염제도의 효과적인 수립디 되도록 실내장식재 및 장식물품에 대한 세부규제안을 기술하여 방화 및 방염제도의 세부규안을 제시하였다.