• Asia Pacific Journal of Corpus Research
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• v.1 no.1
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• pp.61-94
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• 2020

The present study examined vocabulary usage in elementary English classrooms in Japan using elementary school corpus. The authors used three wordlists to benchmark the lexical items for four classes in the corpus: the CEFR-J, the General Service List (GSL), and Academic Word List (AWL). The percentage of vocabulary usage belonging to the Level A1 in the CEFR-J was below 15% (Class A: 12.1%, Class B: 12.6%, Class C: 8.9%, and Class D: 13.6%) with no statistical difference between levels. The mean ratio of Level A2 vocabulary items was below 10%, and all classes showed less than 1% of vocabulary usage for the Levels B1 and B2. Over 70% of all vocabulary items in the corpus belonged to the most frequent 1,000-word band (level 1) of the GSL, while the next most frequent word band (level 2 of the GSL and AWL) accounted for less than 10%. The results suggest that elementary school English teachers should use more vocabulary items in the CEFR-J Level A1. The findings demonstrate that elementary school teachers are less likely to expose their pupils to grammatically well-structured sentences with an abundance of lexical items since the teachers repeatedly use the same lexemes in each class.

• Asia Pacific Journal of Corpus Research
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• v.4 no.2
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• pp.75-87
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• 2023

This paper presents a revised version of the vocabulary analyzer for self-reflection (VACSR), called VACSR v.2.0. The initial version of the VACSR automatically analyzes the occurrences and the level of vocabulary items in the transcribed texts, indicating the frequency, the unused vocabulary items, and those not belonging to either scale. However, it overlooked words with multiple parts of speech due to their identical headword representations. It also needed to provide more explanatory result tables from different corpora. VACSR v.2.0 overcomes the limitations of its predecessor. First, unlike VACSR v.1, VACSR v.2.0 distinguishes words that are different parts of speech by syntactic parsing using Stanza, an open-source Python library. It enables the categorization of the same lexical items with multiple parts of speech. Second, VACSR v.2.0 overcomes the limited clarity of VACSR v.1 by providing precise result output tables. The updated software compares the occurrence of vocabulary items included in classroom corpora for each level of the Common European Framework of Reference-Japan (CEFR-J) wordlist. A pilot study utilizing VACSR v.2.0 showed that, after converting two English classes taught by a preservice English teacher into corpora, the headwords used mostly corresponded to CEFR-J level A1. In practice, VACSR v.2.0 will promote users' reflection on their vocabulary usage and can be applied to teacher training.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2635-2649
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• 2022

CEFR is a small core-size sodium-cooled fast reactor (SFR) using high enrichment fuel with stainless-steel reflectors, which brings a significant challenge to the deterministic methodologies due to the strong spectral effect. The neutronic simulation of the start-up experiments conducted at the CEFR have been performed with a deterministic code system RAST-F, which is based on the two-step approach that couples a multi-group cross-section generation Monte-Carlo (MC) code and a multi-group nodal diffusion solver. The RAST-F results were compared against the measurement data. Moreover, the characteristic of neutron spectrum in the fuel rings, and adjacent reflectors was evaluated using different models for generation of accurate nuclear libraries. The numerical solution of RAST-F system was verified against the full core MC solution MCS at all control rods fully inserted and withdrawn states. A good agreement between RAST-F and MCS solutions was observed with less than 120 pcm discrepancies and 1.2% root-mean-square error in terms of k_eff and power distribution, respectively. Meanwhile, the RAST-F result agreed well with the experimental values within two-sigma of experimental uncertainty. The good agreement of these results indicating that RAST-F can be used to neutronic steady-state simulations for small core-size SFR, which was challenged to deterministic code system.

• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3888-3896
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• 2022

Nuclear data is the basis of reactor physics analysis. This paper aim at studying the influence of major evaluated nuclear data libraries, CENDL-3.2, ENDF/B-VIII.0, JEFF-3.3, and JENDL-4.0u, on the neutronics modelling of CEFR start-up tests. Results show these four libraries have a good performance and consistency in the modelling CEFR start-up tests. The JEFF-3.3 results exhibit only an 8 pcm k_eff difference with the measurement. The difference in criticality is decomposed by nuclide, which shows the large overestimation of CENDL-3.2 is mainly from the cross-section of ⁵²Cr. Except for few cases, the calculation results are within 1σ of measurement uncertainty in control rod worth, sodium void reactivity, temperature reactivity, and subassembly swap reactivity. In the evaluation of axial and radial reaction distribution, there are about 65% of relative errors that are less than 5% and 82% of relative errors that are less than 10%.

• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3897-3908
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• 2022

High-fidelity nuclear data libraries and neutronics simulation tools are essential for the development of fast reactors. The IAEA coordinated research project on "Neutronics Benchmark of CEFR Start-Up Tests" offers valuable data for the qualification of nuclear data libraries and neutronics codes. This paper focuses on the verification and validation of the CEFR start-up modelling using OpenMC Monte-Carlo code against the experimental measurements. The OpenMC simulation results agree well with the measurements in criticality, control rod worth, sodium void reactivity, temperature reactivity, subassembly swap reactivity, and reaction distribution. In feedback coefficient evaluations, an additional state method shows high consistency with lower uncertainty. Among 122 relative errors in the benchmark of the distribution of nuclear reaction, 104 errors are less than 10% and 84 errors are less than 5%. The results demonstrate the high reliability of OpenMC for its application in fast reactor simulations. In the companion paper, the influence of cross-section libraries is investigated using neutronics modelling in this paper.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4491-4503
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• 2023

This study aimed to verify and validate the transient simulation capability of the hybrid code system RAST-F for fast reactor analysis. For this purpose, control rod (CR) drop experiments involving eight separate CRs and six CR groups in the China Experimental Fast Reactor (CEFR) start-up tests were utilized to simulate the CR drop transient. The RAST-F numerical solution, including the neutron population, time-dependent reactivity, and CR worth, was compared against the measurement values obtained from two out-of-core detectors. Moreover, the time-dependent reactivity and CR worth from RAST-F were verified against the results obtained by the Monte Carlo code Serpent using continuous energy nuclear data. A code-to-code comparison between Serpent and RAST-F showed good agreement in terms of time-dependent reactivity and CR worth. The discrepancy was less than 160 pcm for reactivity and less than 110 pcm for CR worth. RAST-F solution was almost identical to the measurement data in terms of neutron population and reactivity. All the calculated CR worth results agreed with experimental results within two standard deviations of experimental uncertainty for all CRs and CR groups. This work demonstrates that the RAST-F code system can be a potential tool for analyzing time-dependent phenomena in fast reactors.

• Asia Pacific Journal of Corpus Research
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• v.1 no.1
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• pp.127-132
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• 2020

The purpose of this review is to provide local readers, more specifically, Korean student readers who are not all that familiar with the English language a general overview of research articles that have been published in Asia Pacific Journal of Corpus Research vol. 1, no. 1. A brief summary of each research article focusing on research methods and then followed by an overall review and some insights on research issues will be presented.

• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.2343-2351
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• 2024

Accurately predicting the thermal hydraulic parameters of a transient reactor core under different working conditions is the first step toward reactor safety. Mass flow rate and temperature are important parameters of core thermal hydraulics, which have often been modeled as time series prediction problems. This study aims to achieve accurate and continuous prediction of core thermal hydraulic parameters under instantaneous conditions, as well as test the feasibility of a newly constructed gated recurrent unit (GRU) model based on the soft attention mechanism for core parameter predictions. Herein, the China Experimental Fast Reactor (CEFR) is used as the research object, and CEFR 1/2 core was taken as subject to carry out continuous predictive analysis of thermal parameters under transient conditions., while the subchannel analysis code named SUBCHANFLOW is used to generate the time series of core thermal-hydraulic parameters. The GRU model is used to predict the mass flow and temperature time series of the core. The results show that compared to the adaptive radial basis function neural network, the GRU network model produces better prediction results. The average relative error for temperature is less than 0.5 % when the step size is 3, and the prediction effect is better within 15 s. The average relative error of mass flow rate is less than 5 % when the step size is 10, and the prediction effect is better in the subsequent 12 s. The GRU model not only shows a higher prediction accuracy, but also captures the trends of the dynamic time series, which is useful for maintaining reactor safety and preventing nuclear power plant accidents. Furthermore, it can provide long-term continuous predictions under transient reactor conditions, which is useful for engineering applications and improving reactor safety.