• Nuclear Engineering and Technology
• /
• v.54 no.6
• /
• pp.2213-2220
• /
• 2022

The maximum dose delivery at the end of the beam range provides the main advantage of using proton therapy. The range of the proton beam, however, is subject to uncertainties, which limit the clinical benefits of proton therapy and, therefore, accurate in vivo verification of the beam range is desirable. For the beam range verification in spot scanning proton therapy, a prompt gamma detection system, called as gamma electron vertex imaging (GEVI) system, is under development and, in the present study, the performance of the GEVI system in spot scanning proton therapy was predicted with Geant4 Monte Carlo simulations in terms of shift detection sensitivity, accuracy and precision. The simulation results indicated that the GEVI system can detect the interfractional range shifts down to 1 mm shift for the cases considered in the present study. The results also showed that both the evaluated accuracy and precision were less than 1-2 mm, except for the scenarios where we consider all spots in the energy layer for a local shifting. It was very encouraging results that the accuracy and precision satisfied the smallest distal safety margin of the investigated beam energy (i.e., 4.88 mm for 134.9 MeV).

• Nuclear Engineering and Technology
• /
• v.54 no.3
• /
• pp.888-895
• /
• 2022

Lately, ocean nuclear power plants have attracted attention as one of diverse uses of nuclear power plants. Because ocean nuclear power plants are movable or transportable, it is necessary to analyze the thermal hydraulics in a moving frame of reference, and computer codes have been developed to predict thermal hydraulics in large moving systems. The purpose of this study is to incorporate a three dimensional dynamic motion model into the SPACE code (Safety and Performance Analysis CodE) so that the code is able to analyze thermal hydraulics in an ocean nuclear power plant. A rotation system that describes three-dimensional rotations about an arbitrary axis was implemented, and modifications were made to the one-dimensional momentum equations to reflect the rectilinear and rotational acceleration effects. To demonstrate the code's ability to solve a problem utilizing a rotational frame of reference, code calculations were conducted on various conceptual problems in the two-dimensional and three-dimensional pipeline loops. In particular, the code results for the three-dimensional pipeline loop with a tilted rotation axis agreed well with the multi-dimensional CFD results.

• Nuclear Engineering and Technology
• /
• v.54 no.1
• /
• pp.283-292
• /
• 2022

Advanced radiation applications have been widely used and extended to many fields. As a result of this fact, choosing an appropriate shielding material based on the radiation application has become vital. In this regard, the integration of elements into polymer composites has been investigated and contributed to the quantity and quality of radiation shielding materials. This study reports photon attenuation parameters and electromagnetic shielding effectiveness of a novel polymer composite prepared with a matrix reinforced with three different proportions (5, 10, and 15 wt%) of niobium content. Addition of Nb dopant improves both photon attenuation and electromagnetic shielding effectiveness for the investigated composites. Therefore, Nb(15%) polymer composite with highest concentration has been found to be the best absorber for ionizing and non-ionizing radiations. Consequently, the performed analyzes provide evidences that the prepared Nb-reinforced polymer composite could be effectively used as photon radiation attenuator and electromagnetic shielding material.

• Nuclear Engineering and Technology
• /
• v.54 no.1
• /
• pp.30-35
• /
• 2022

In a nuclear reactor, gamma radiation is the dominant energy deposition in non-fuel regions. Heat is generated upon gamma deposition and consequently affects the mechanical and thermal structure of the material. Therefore, the safety of samples should be carefully considered so that their integrity and quality can be retained. To evaluate relevant parameters, an in-core gamma thermometer (GT) was used to measure gamma heating (GH) throughout the operation of the McMaster nuclear reactor (MNR) at four irradiation sites. Additionally, a Monte Carlo reactor physics code (Serpent-2) was utilized to model the MNR with the GT located in the same irradiation sites used in the measurement to verify its predictions against measured GH. This research aids in the development of modeling, calculation, and prediction of the GH utilizing Serpent-2 as well as implementing a new GH measurement at the MNR core. After all uncertainties were quantified for both approaches, comparable GH profiles were observed between the measurements and calculations. In addition, the GH values found in the four sites represent a strong level of radiation based on the distance of the sample from the core. In this study, the maximum and minimum GH values were found at 0.32 ± 0.05 W/g and 0.15 ± 0.02 W/g, respectively, corresponding to 320 Sv/s and 150 Sv/s. These values are crucial to be considered whenever sample is planned to be irradiated inside the MNR core.

• Nuclear Engineering and Technology
• /
• v.54 no.6
• /
• pp.2188-2197
• /
• 2022

To ensure the safety margin of a reactor pressure vessel (RPV) under normal operating conditions, it is regulated through the pressure-temperature (P-T) limit curve. The stress intensity factor (SIF) obtained by the internal pressure and thermal load should be obtained through crack analysis of the nozzle corner crack in advance to generate the P-T limit curve for the nozzle. In the ASME code Section XI, Appendix G, the SIF via the internal pressure for the nozzle corner crack is expressed as a function of the cooling or heating rate, and the wall thickness, however, the SIF via the thermal load is presented as a polynomial format based on the stress linearization analysis results. Inevitably, the SIF can only be obtained through finite element (FE) analysis. In this paper, simple prediction equations of the SIF via the thermal load under, cool-down and heat-up conditions are presented. For the Korean standard nuclear power plant, three geometric variables were set and 72 cases of RPV models were made, and then the heat transfer analysis and thermal stress analysis were performed sequentially. Based on the FE results, simple engineering solutions predicting the value of thermal SIF under cool-down and heat-up conditions are suggested.

• Nuclear Engineering and Technology
• /
• v.54 no.7
• /
• pp.2550-2563
• /
• 2022

This study performed a precise dynamic finite element time history elastic-plastic seismic analysis considering the welds, which have been not considered in design stage, on the nuclear components subjected to severe seismic loadings such as beyond-design basis earthquakes for sustainable nuclear power plants. First, the dynamic finite element elastic-plastic seismic analysis was performed for a general design practice that does not take into account the welds of the pressurizer surge line system, one of safety class I components in nuclear power plants, and then the reference values for the accumulated equivalent plastic strain, equivalent plastic strain, and von Mises effective stress were set. Second, the dynamic finite element elastic-plastic seismic analyses were performed for the case of considering only the mechanical strength over-mismatch of the welds as well as for the case of considering both the strength over-mismatch and welding residual strain. Third, the effects of the strength over-mismatch and welding residual strain were analyzed by comparing the finite element analysis results with the reference values. As a result of the comparison, it was found that not considering the strength over-mismatch may lead to conservative assessment results, whereas not considering the welding residual strain may be non-conservative.

• Nuclear Engineering and Technology
• /
• v.54 no.8
• /
• pp.2989-2998
• /
• 2022

Line contact structures, such as the contact between graphite brick and graphite tenon, widely exist in high-temperature gas-cooled reactors. Due to the stress concentration effect, the line contact area is one of the dangerous positions prone to failure in the nuclear reactor core. In this paper, the failure mechanism of line contact structures composed of IG11 nuclear graphite column and brick were investigated by means of experiment and finite element simulation. It was found that the failure process mainly includes three stages: firstly, the damage accumulation in nuclear graphite material led to the characteristic yielding of the line contact structure, but no macroscopic failure can be observed at this stage; secondly, the stresses near the contact area met Mohr failure criterion, and a crack initiated and propagated laterally in the contact zone, that is, local macroscopic failure occurred at this stage; finally, a second crack initiated in the contact area and developed in to a Y-shape, resulting in the final failure of the structure. This study lays a foundation for the structural design and safety assessment of high-temperature gas-cooled reactors.

• Nuclear Engineering and Technology
• /
• v.54 no.7
• /
• pp.2540-2549
• /
• 2022

For a pressurized water reactor power plant, the reactor coolant pump (RCP) is a kernel component. And for a canned motor RCP, the rotor system's properties determines its safety. The liquid coolant inside the canned motor RCP fills clearance between the metal shields of rotor and stator, forming a lengthy clearance flow. The influence of inlet preswirl on rotordynamic coefficients of clearance flow in canned motor RCP and their effects on the rotordynamic characteristics of the pump are numerically and experimentally investigated in this work. A quasi-steady state computational fluid dynamics (CFD) method has been used to investigate the influence of inlet preswirl. A vertical experiment rig has also been established for this purpose. Rotordynamic coefficients on different inlet preswirl ratios (IR) are obtained through CFD and experiment. Results show that the cross-coupled stiffness of the clearance flow would change significantly with inlet preswirl, but other rotordynamic coefficients would not change significantly with inlet preswirl. For the case of clearance flow between the stator and rotor cans, influence of inlet preswirl is not so significant as the IR is not large enough.

• Nuclear Engineering and Technology
• /
• v.54 no.6
• /
• pp.2329-2333
• /
• 2022

Base metal corrosion forms a significant issue during the chemical decontamination of the primary coolant loop in nuclear power plants as it is directly related to the economic and safety viability of decommissioning. In this technical note, potentiodynamic evaluations of several base metals (304 stainless steel, SA106 Grade B carbon steel, and alloy 600) were performed to determine their corrosion behavior during the hydrazine (N₂H₄)-based reductive ion decontamination (HyBRID) process. The results suggested that N₂H₄ protected the surface of the base metals in the HyBRID solution, which is primarily composed of H₂SO₄. The corrosion resistance of the carbon steel was further improved through the addition of CuSO₄ to the solution. The corrosion rate of carbon steel in the H₂SO₄-N₂H₄-CuSO₄ solution was lower than that exhibited in an oxalic acid solution, a commonly used reaction medium during commercial decontamination processes. These results indicate the superiority of the HyBRID process with respect to the base metal stability.

• Nuclear Engineering and Technology
• /
• v.54 no.8
• /
• pp.2828-2839
• /
• 2022

This study developed a safety assessment tool for geological disposal systems called APro, a systemically integrated modeling system based on modularizing and coupling the processes which need to be considered in a geological disposal system. Thermal, hydraulic, chemical, canister failure, radionuclide release and transport processes were considered in the current version of APro. Each of the unit processes in APro consists of a single Default Module, and several Alternative Modules which can increase the flexibility of the model. As an initial stage of developing the modularization concept and modeling interface, the Default Modules of each unit process were described, with one Alternative Module of chemical process. The computation part of APro is mainly a MATLAB workspace controlling COMSOL and PHREEQC, which are coupled by an operator splitting scheme. The APro model domain is a stylized geological disposal system employing the Swedish disposal concept (KBS-3 type), but the repository layout can be freely adjusted. In order to show the applicability of APro to the total system performance assessment of geological disposal system, some sample simulations were conducted. From the results, it was confirmed that coupling of the thermal and hydraulic processes and coupling of the canister failure and the radionuclide release processes were well reflected in APro. In addition, the technical connectivity between COMSOL and PHREEQC was also confirmed.