• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1718-1733
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• 2023

The virtual Frisch-grid method for room-temperature radiation detectors has been widely used because of its simplicity and high performance. Recently, side electrodes were separately attached to each surface of the detectors instead of covering the entire detector surface with a single electrode. The side-electrode structure enables the measurement of the three-dimensional (3D) gamma-ray interaction in the detector. The positional information of the interaction can then be utilized to precisely calibrate the response of the detector for gamma-ray spectroscopy and imaging. In this study, we developed a 3D position-sensitive 5 × 5 × 12 mm³ cadmium-zinc-telluride (CZT) detector and applied a flattening method to correct detector responses. Collimated gamma-rays incident on the surface of the detector were scanned to evaluate the positional accuracy of the detection system. Positional distributions of the radiation interactions with the detector were imaged for quantitative and qualitative evaluation. The energy spectra of various radioisotopes were measured and improved by the detector response calibration according to the calculated positional information. The energy spectra ranged from 59.5 keV (emitted by ²⁴¹Am) to 1332 keV (emitted by ⁶⁰Co). The best energy resolution was 1.06% at 662 keV when the CZT detector was voxelized to 20 × 20 × 10.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2246-2255
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• 2023

Evacuation time estimation (ETE) is crucial for the effective implementation of resident protection measures as well as planning, owing to its applicability to nuclear emergencies. However, as confirmed in the Fukushima case, the ETE performed by nuclear operators does not reflect behavioral features, exposing thus, gaps that are likely to appear in real-world situations. Existing research methods including surveys and interviews have limitations in extracting highly feasible behavioral features. To overcome these limitations, we propose a VR-based immersive experiment system. The VR system realistically simulates nuclear emergencies by structuring existing disasters and human decision processes in response to the disasters. Evacuation behavioral features were quantitatively extracted through the proposed experiment system, and this system was systematically verified by statistical analysis and a comparative study of experimental results based on previous research. In addition, as part of future work, an application method that can simulate multi-level evacuation dynamics was proposed. The proposed experiment system is significant in presenting an innovative methodology for quantitatively extracting human behavioral features that have not been comprehensively studied in evacuation. It is expected that more realistic evacuation behavioral features can be collected through additional experiments and studies of various evacuation factors in the future.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2088-2095
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• 2023

The safety of nuclear installations is largely determined by the tightness of fuel elements cladding. As the Fukushima nuclear accident showed, the main task in case of loss of power supply is to ensure reliable removal of residual heat release from spent fuel pool (SFP) with irradiated fuel assemblies (IFAs). The paper presents the results of calculated-experimental studies and thermal-hydraulic modeling of temperature storage modes of IFAs in SFP. Experimental studies of SFP's temperature regime and calculated evaluation of residual heat removal due to the thermal conductivity of building structures surrounding the SFP were performed. To ensure the safe operation of research reactors, it's necessary to know the IFA's residual heat power (RHP) in the reactor and SFP, which is determined depending on the operating time of fuel assemblies (FAs) and the IFAs calculated holding time. The FAs operating time depends on the reactor energy output. The IFAs calculated holding time is determined by the fuel burnup, U-235 mass in the fuel, and reactor utilization factor. The IFAs fuel burnup was calculated using the MCU-PTR program. Also presented are the RHP's calculation results using some of the empirical dependencies. The concept of a passive heat removal system (PHRS) based on thermosyphon's operating principle was proposed.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.1999-2010
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• 2023

Undesirable flapping motion of discs can cause the failure of swing check valves in nuclear passive safety systems. Time-resolved particle image velocimetry (PIV) was employed to investigate the flow characteristics around a free-to-rotate plate and the motion response, with the Reynolds numbers, based on the hydraulic diameter of the channel, from 1.32 × 10⁴ to 3.95 × 10⁴. Appreciable flapping motion (±3.52°) appeared at the Reynolds number of 2.6 × 10⁴ with the frequency of 5.08 Hz. In the low-Reynolds-number case, the plate showed negligible flapping. In the high-Reynolds-number case, the deflection angle increased with reduced flapping amplitude. The torque from the fluid determined the flapping amplitude. In the low-Reynolds-number case, Karman vortices were absent. With increasing Reynolds numbers, Karman vortices developed behind the plate with larger deflection angles. Strong interaction between the wake flow from the leading and trailing edge of the plate was observed. Based on power spectrum density (PSD) analysis, the vortex shedding frequency coincided with the flapping frequency, and the amplitude was positively correlated to the strength of the vortices. Proper orthogonal decomposition (POD) modes evince that, in the case of appreciable motion, coherent structures exhibited a larger spatial scale, enhancing the magnitude of the external torque on the plate.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2011-2017
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• 2023

Gas submerged jet is an outstanding thermohydraulic phenomenon in pool scrubbing of fission products during a severe nuclear accident. Experiments were performed on the hydraulic characteristics in the ranges of air mass flux 0.1-1400 kg/m²s and nozzle diameter 10-80 mm. The results showed that the dependence of inlet pressure on the mass flux follows a power law in subsonic jets and a linear law in sonic jets. The effect of nozzle submerged depth was negligible. The isolated bubbling regime, continuous bubbling regime, transition regime, and jetting regime were observed in turn, as the mass flux increased. In the bubbling regime and jetting regime, the air volume fraction distribution was approximately symmetric in space. Themelis model could capture the jet trajectory well. In the transition regime, the air volume fraction distribution loses symmetry due to the bifurcated secondary plume. The Li correlation and Themelis model showed sufficient accuracy for the prediction of jet penetration length.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1791-1801
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• 2023

The multi-physics coupled methodologies that have been widely used to analyze the complex process occurring in nuclear reactors have also been used to the R&D of numerical reactors. The advancement in the field of computer technology has helped in the development of these methodologies. Herein, we report the integration of ADPRES code and RELAP5 code into the SALOME-ICoCo framework to form a multi-physics coupling platform. The platform exploits the supervisor architecture, serial mode, mesh one-to-one correspondence and explicit coupling methods during analysis, and the uncertainty analysis tool URANIE was used. The correctness of the platform was verified through the NEACRP-L-335 benchmark. The results obtained were in accordance with the reference values. The platform could be used to accurately determine the power peak. In addition, design margins could be gained post uncertainty analysis. The initial power, inlet coolant temperature and the mass flow of assembly property significantly influence reactor safety during the rod ejections accident (REA).

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1630-1643
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• 2023

The correct situation awareness (SA) of operators is important for managing nuclear power plants (NPPs), particularly in accident-related situations. Among the three levels of SA suggested by Ensley, Level 3 SA (i.e., projection of the future status of the situation) is challenging because of the complexity of NPPs as well as the uncertainty of accidents. Hence, several prediction methods using artificial intelligence techniques have been proposed to assist operators in accident prediction. However, these methods only predict short-term plant status (e.g., the status after a few minutes) and do not provide information regarding the uncertainty associated with the prediction. This paper proposes an algorithm that can predict the multivariate and long-term behavior of plant parameters for 2 h with 120 steps and provide the uncertainty of the prediction. The algorithm applies bidirectional long short-term memory and an attention mechanism, which enable the algorithm to predict the precise long-term trends of the parameters with high prediction accuracy. A conditional variational autoencoder was used to provide uncertainty information about the network prediction. The algorithm was trained, optimized, and validated using a compact nuclear simulator for a Westinghouse 900 MWe NPP.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1911-1923
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• 2023

The nuclear safety assessment involving large transient simulations is forcing the community to develop methods for coupling thermal-hydraulics and neutronic codes and three-dimensional (3D) Computational Fluid Dynamics (CFD) codes. In this paper a set of dynamic boundary conditions are implemented in OpenFOAM^Ⓡ in order to apply zero-dimensional (0D) approaches coupling with 3D thermal-hydraulic simulation in a single framework. This boundary conditions are applied to model pipelines, tanks, pumps, and heat exchangers. On a first stage, four tests are perform in order to assess the implementations. The results are compared with experimental data, full 3D CFD, and system code simulations, finding a general good agreement. The semi-implicit implementation nature of these boundary conditions has shown robustness and accuracy for large time steps. Finally, an application case, consisting of a simplified open pool with a cooling external circuit is solved to remark the capability of the tool to simulate thermal hydraulic systems commonly found in nuclear installations.

• Nuclear Engineering and Technology
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• v.55 no.5
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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.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2373-2379
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• 2023

In sodium systems, leakage is one of the safety concerns; it can cause chemical reactions, which may result in fires. There are contact and non-contact types of leak detectors, and the conventional method of non-contact type detection is by gas sampling. Because of the complexity of this method, there has always been a need for a simple gas sensor, and the resistive-type nanostructure ZnO sensor is a promising option with various advantages. In this study, a ZnO sensor was fabricated, and the concept was tested as a leak detector using a dedicated experiment facility. The experiment results showed distinctive changes in resistance with the presence of sodium aerosol under various conditions. Replacing the conventional gas sampling with the ZnO sensors is expected to enable identification of the leakage location if used as a point-wise instrumentation and to greatly reduce the total cost, making the system simple, light, and effective. For further study, more tests will be performed to evaluate the sensitivity of key parameters under various conditions.