• Nuclear Engineering and Technology
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• 제55권3호
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• pp.919-926
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• 2023

To analyze the cause of the destruction of thin, carbon-backed lithium fluoride targets during a measurement of the fusion of ⁷Li and ¹⁷O, we estimate theoretically the lifetimes of carbon and LiF films due to sputtering, thermal evaporation, and lattice damage and compare them with the lifetime observed in the experiment. Sputtering yields and thermal evaporation rates in carbon and LiF films are too low to play significant roles in the destruction of the targets. We estimate the lifetime of the target due to lattice damage of the carbon backing and the LiF film using a previously reported model. In the experiment, elastically scattered target and beam ions were detected by surface silicon barrier (SSB) detectors so that the product of the beam flux and the target density could be monitored during the experiment. The areas of the targets exposed to different beam intensities and fluences were degraded and then perforated, forming holes with a diameter around the beam spot size. Overall, the target thickness tends to decrease linearly as a function of the beam fluence. However, the thickness also exhibits an increasing interval after SSB counts per beam ion decreases linearly, extending the target lifetime. The lifetime of thin LiF film as determined by lattice damage is calculated for the first time using a lattice damage model, and the calculated lifetime agrees well with the observed target lifetime during the experiment. In experiments using a thin LiF target to induce nuclear reactions, this study suggests methods to predict the lifetime of the LiF film and arrange the experimental plan for maximum efficiency.

• Nuclear Engineering and Technology
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• 제55권8호
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• pp.2747-2756
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• 2023

Reinforced Concrete (RC) shear walls are one of the civil structures in nuclear power plants to resist lateral loads such as earthquakes and wind loads effectively. Risk-informed and performance-based regulation in the nuclear industry requires considering possible accidents and determining desirable performance on structures. As a result, rather than predicting only the ultimate capacity of structures, the prediction of performances on structures depending on different damage states or various accident scenarios have increasingly needed. This study aims to develop machine-learning models predicting drifts of the RC shear walls according to the damage limit states. The damage limit states are divided into four categories: the onset of cracking, yielding of rebars, crushing of concrete, and structural failure. The data on the drift of shear walls at each damage state are collected from the existing studies, and four regression machine-learning models are used to train the datasets. In addition, the bagging ensemble method is applied to improve the accuracy of the individual machine-learning models. The developed models are to predict the drifts of shear walls consisting of various cross-sections based on designated damage limit states in advance and help to determine the repairing methods according to damage levels to shear walls.

• Nuclear Engineering and Technology
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• 제48권3호
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• pp.758-764
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• 2016

Specimens of stainless steel reactor internals were irradiated with 240 keV protons and 6 MeV Xe ions at room temperature. Nanoindentation constant stiffness measurement tests were carried out to study the hardness variations. An irradiation hardening effect was observed in proton- and Xe-irradiated specimens and more irradiation damage causes a larger hardness increment. The Nix-Gao model was used to extract the bulk-equivalent hardness of irradiation-damaged region and critical indentation depth. A different hardening level under H and Xe irradiation was obtained and the discrepancies of displacement damage rate and ion species may be the probable reasons. It was observed that the hardness of Xe-irradiated specimens saturate at about 2 displacement/atom (dpa), whereas in the case of proton irradiation, the saturation hardness may be more than 7 dpa. This discrepancy may be due to the different damage distributions.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2975-2981
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• 2020

Radiation damage is one of the aging important causes in nuclear reactors. Radiation damage causes changes in material properties. In this study, this effect has been evaluated and analyzed on the clad of the Tehran research reactor (TRR). A grade 6061 aluminum is used as a clad in the TRR. The MCNPX code is used to designate the most sensitive location of the reactor and calculate neutron flux distribution. Then, a software using FORTRAN language programming is developed to process the particle track (PTRAC) output file of the MCNPX code. The SRIM code is used here to calculate the rate of displacement per atom. Moreover, the SPECOMP and SPECTER codes are also applied to estimate the displacement rate and compared with the results attained using the SRIM code. The rate of displacement per atom by the SPECTER and SRIM codes have been obtained 2.54 × 10^-7 dpa/s and 2.44 × 10^-7 dpa/s (QD method), respectively. Also, the mechanical properties have been evaluated using the RCC-MRx code and have been compared with experimental results. Finally, the change in the matter specification has been analyzed as a function of time.

• Nuclear Engineering and Technology
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• 제55권9호
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• pp.3229-3240
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• 2023

Due to neutron radiation, atomic displacement has a significant effect on material in nuclear reactors. A range of secondary displacement models, including the Kinchin-Pease (K-P), Lindhard, Norgett-Robinson-Torrens (NRT), and athermal recombination-corrected displacement per atom (arc-dpa) have been suggested to calculate the number of displacement per atom (dpa). As neutron elastic interaction is the main cause of displacement damage, the focus of the current study is to calculate the atomic displacement caused by the neutron elastic interaction in order to estimate the exact amount of yield strength in a WWER-1000 reactor pressure vessel. To achieve this purpose, the reactor core is simulated by MCNPX code. In addition, a program is developed to calculate the elastic radiation damage induced by the incident neutron flux (RADIX) based on different models using Fortran programming language. Also, due to non-elastic interaction, the displacement damage is calculated by the HEATR module of the NJOY code. ASME E-693-01 standard, SPECTER, NJOY codes, and other pervious findings have been used to validate RADIX results. The results showed that the RADIX(arc-dpa)/HEATR outputs have appropriate accuracy. The relative error of the calculated dpa resulting from RADIX(arc-dpa)/HEATR is about 8% and 46% less than NJOY code, respectively in the ¼ and ¾ vessel wall.

• 방사성폐기물학회지
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• 제19권1호
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• pp.75-85
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• 2021

Evaluating the quantitative damage to rocks through acoustic emission (AE) has become a research focus. Most studies mainly used one or two AE parameters to evaluate the degree of damage, but several AE parameters have been rarely used. In this study, several data-driven models were employed to reflect the combined features of AE parameters. Through uniaxial compression tests, we obtained mechanical and AE-signal data for five granite specimens. The maximum amplitude, hits, counts, rise time, absolute energy, and initiation frequency expressed as the cumulative value were selected as input parameters. The result showed that gradient boosting (GB) was the best model among the support vector regression methods. When GB was applied to the testing data, the root-mean-square error and R between the predicted and actual values were 0.96 and 0.077, respectively. A parameter analysis was performed to capture the parameter significance. The result showed that cumulative absolute energy was the main parameter for damage prediction. Thus, AE has practical applicability in predicting rock damage without conducting mechanical tests. Based on the results, this study will be useful for monitoring the near-field rock mass of nuclear waste repository.

• Nuclear Engineering and Technology
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• 제53권8호
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• pp.2682-2695
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• 2021

This paper aims to evaluate the structural dynamic responses and damage/failure of the nuclear fuel reprocessing plant under the free drop impact of spent fuel cask (SFC) and fuel assembly (FA) during the on-site transportation. At the present Part I of this paper, the large-scale SFC model free drop test and the corresponding numerical simulations are performed. Firstly, a composite target which is composed of the protective structure, i.e., a thin RC plate (representing the inverted U-shaped slab in the loading shaft) and/or an autoclaved aerated concrete (AAC) blocks sacrificial layer, as well as a thick RC plate (representing the bottom slab in the loading shaft) is designed and fabricated. Then, based on the large dropping tower, the free drop test of large-scale SFC model with the mass of 3 t is carried out from the height of 7 m-11 m. It indicates that the bottom slab in the loading shaft could not resist the free drop impact of SFC. The composite protective structure can effectively reduce the damage and vibrations of the bottom slab, and the inverted U-shaped slab could relieve the damage of the AAC blocks layer dramatically. Furthermore, based on the finite element (FE) program LS-DYNA, the corresponding refined numerical simulations are performed. By comparing the experimental and numerical damage and vibration accelerations of the composite structures, the present adopted numerical algorithms, constitutive models and parameters are validated, which will be applied in the further assessment of drop impact effects of full-scale SFC and FA on prototype nuclear fuel reprocessing plant in the next Part II of this paper.

• Nuclear Engineering and Technology
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• 제51권2호
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• pp.501-509
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• 2019

In this paper a body-centered cubic(BCC) crystal plasticity model based on microscopic dislocation mechanism is introduced and numerically implemented. The model is coupled with irradiation effect via tracking dislocation loop evolution on each slip system. On the basis of the model, uniaxial tensile tests of unirradiated and irradiated RPV steel(take Chinese A508-3 as an example) at different temperatures are simulated, and the simulation results agree well with the experimental results. Furthermore, crystal plasticity damage is introduced into the model. Then the damage behavior before and after irradiation is studied using the model. The results indicate that the model is an effective tool to study the effect of irradiation and temperature on the mechanical properties and damage behavior.

• 한국방사선학회논문지
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• 제10권8호
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• pp.583-590
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• 2016

본 연구는 부산광역시 거주민을 대상으로 원전주변 갑상선암 피해 소송 이후 원자력발전소 인근지역과 시내권에 거주하는 주민간의 인식 차이를 분석하여 국민들의 원자력에 대한 수용성 수준을 판별하고자 하였다. 설문조사는 총 551명(원전인근지역 269명, 시내권 282명)을 대상으로 대인면접을 통해 이루어졌으며, 분석 결과, 원전 주변 갑상선암 발병 피해 소송 사건을 계기로 국민들의 원자력에 대한 인식에 변화가 있는 것으로 나타났다. 원전인근지역 주민의 경우 사건 이후 원자력에 대한 불신과 불안감이 높아져, 원자력에 대한 부정적 인식이 강해졌음을 확인할 수 있었다. 반면 시내권 주민들은 원전 주변 갑상선암 발병 피해 소송 사건에도 불구하고 사건 이전보다 더욱 원자력 수용성에 긍정적인 인식을 갖는 것으로 분석되었다. 그러나 원자력 안전성과 신뢰성에 대한 인식은 부정적인 것으로 분석되어 원자력 수용성과 상반되는 이중적인 인식을 보였다. 이는 부산광역시 시내권 주민들의 경우 원자력의 편익, 필요성 등을 원칙적으로 인정하나 안전성에 대해서는 의심하는 것으로 판단되며 향후 원자력의 이용 확대를 지속적으로 추진하기 위해서는 국민의 올바른 이해와 신뢰 그리고 무엇보다 국민 또는 원전인근지역 주민들과 충분한 의사소통이 필요하다고 사료된다.

• Nuclear Engineering and Technology
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• 제52권7호
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• pp.1537-1544
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• 2020

Molecular dynamics (MD) simulations were conducted to investigate the temperature effects on the primary damage in gallium nitride (GaN) material. Five temperatures ranging from 300 K to 900 K were studied for 10 keV Ga primary knock-on atom (PKA) with inject direction of [0001]. The results of MD simulations showed that threshold displacement energy (E_d) was affected by temperatures and at higher temperature, it was larger. The evolutions of defects under various temperatures were similar. However, the higher temperature was found to increase the peak number, peak time, final time and recombination efficiency while decreasing the final number. With regard to clusters, isolated point defects and little clusters were common clusters and the fraction of point defects increased with temperature for vacancy clusters, whereas it did not appear in the interstitial clusters. Finally, at each temperature, the number of Ga interstitial atoms was larger than that of N and besides that, there were other different results of specific types of split interstitial atoms.