• Nuclear Engineering and Technology
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• v.43 no.6
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• pp.557-566
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• 2011

A cold spray coating (CSC) of copper was studied for its application to a high-level radioactive waste (HLW) disposal canister. Several copper coatings of 10 mm thick were fabricated using two kinds of copper powders with different oxygen contents, and SS 304 and nodular cast iron were used as their base metal substrates. The fabricated CSC coppers showed a high tensile strength but were brittle in comparison with conventional non-coating copper, hereinafter defined to as "commercial copper". The corrosion behavior of CSC coppers was evaluated by comparison with commercial coppers, such as extruded and forged coppers. The polarization test results showed that the corrosion potential of the CSC coppers was closely related to its purity; low-purity (i.e., high oxygen content) copper exhibited a lower corrosion potential, and high-purity copper exhibited a relatively high corrosion potential. The corrosion rate converted from the measured corrosion current was not, however, dependent on its purity: CSC copper showed a little higher rate than that of commercial copper. Immersion tests in aqueous HCl solution showed that CSC coppers were more susceptible to corrosion, i.e., they had a higher corrosion rate. However, the difference was not significant between commercial copper and high-purity CSC copper. The decrease of corrosion was observed in a humid air test presumably due to the formation of a protective passive film. In conclusion, the results of this study indicate that CSC application of copper could be a useful option for fabricating a copper HLW disposal canister.

• Nuclear Engineering and Technology
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• v.45 no.3
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• pp.401-414
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• 2013

HTR50S is a small modular reactor system based on HTGR. It is designed for a triad of applications to be implemented in successive stages. In the first stage, a base plant for heat and power is constructed of the fuel proven in JAEA's $950^{\circ}C$, 30MWt test reactor HTTR and a conventional steam turbine to minimize development risk. While the outlet temperature is lowered to $750^{\circ}C$ for the steam turbine, thermal power is raised to 50MWt by enabling 40% greater power density in 20% taller core than the HTTR. However the fuel temperature limit and reactor pressure vessel diameter are kept. In second stage, a new fuel that is currently under development at JAEA will allow the core outlet temperature to be raised to $900^{\circ}C$ for the purpose of demonstrating more efficient gas turbine power generation and high temperature heat supply. The third stage adds a demonstration of nuclear-heated hydrogen production by a thermochemical process. A licensing approach to coupling high temperature industrial process to nuclear reactor will be developed. The low initial risk and the high longer-term potential for performance expansion attract development of the HTR50S as a multipurpose industrial or distributed energy source.

• Nuclear Engineering and Technology
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• v.50 no.6
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• pp.937-943
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• 2018

Background: This work presents the performance of a novel electronic brachytherapy (EBT) device and radiotherapy (RT) experiments on both skin cancer cells and animals using the device. Methods and materials: The performance of the EBT device was evaluated by measuring and analyzing the dosimetric characteristics of X-rays generated from the device. The apoptosis of skin cancer cells was analyzed using B16F10 melanoma cancer cells. Animal experiments were performed using C57BL/6 mice. Results: The X-ray characteristics of the EBT device satisfied the accepted tolerance level for RT. The results of the RT experiments on the skin cancer cells show that a significant apoptosis induction occurred after irradiation with 50 kVp X-rays generated from the EBT device. Furthermore, the results of the animal RT experiments demonstrate that the superficial X-rays significantly delay the tumor growth and that the tumor growth delay induced by irradiation with low-energy X-rays was almost the same as that induced by irradiation with a high-energy electron beam. Conclusions: The developed new EBT device has almost the same therapeutic effect on the skin cancer with a conventional linear accelerator. Consequently, the EBT device can be practically used for human skin cancer treatment in the near future.

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.19-26
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• 2020

Small nuclear reactor features higher power capacity, longer operation life than conventional power sources. It could be an ideal alternative of existing power source applied for special equipment for terrestrial or underwater missions. In this paper, a 25kWe heat pipe cooled reactor power source applied for multiple use is preliminary designed. Based on the design, a thermal-hydraulic analysis code for heat pipe cooled reactor is developed to analyze steady and transient performance of the designed nuclear reactor. For reactor design, UN fuel with 65% enrichment and potassium heat pipes are adopted in the reactor core. Tungsten and LiH are adopted as radiation shield on both sides of the reactor core. The reactor is controlled by 6 control drums with B₄C neutron absorbers. Thermoelectric generator (TEG) converts fission heat into electricity. Cooling water removes waste heat out of the reactor. The thermal-hydraulic characteristics of heat pipes are simulated using thermal resistance network method. Thermal parameters of steady and transient conditions, such as the temperature distribution of every key components are obtained. Then the postulated reactor accidents for heat pipe cooled reactor, including power variation, single heat pipe failure and cooling channel blockage, are analyzed and evaluated. Results show that all the designed parameters satisfy the safety requirements. This work could provide reference to the design and application of the heat pipe cooled nuclear power source.

• Journal of the Korean Society for Precision Engineering
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• v.33 no.5
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• pp.333-339
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• 2016

In this paper, a fundamental experiment regarding the formation of porous 3D structures for personal safety products using 3D PPP (Porous Polymer Printing) was introduced for the first time. The filament was manufactured by mixing PP (Polypropylene) and CBA (Chemical Blowing Agent) with polymer extruder, and the diameter of the filament was approximately 1.75mm. The proposed 3D PPP method, combined with the conventional FDM (Fused Deposition Modeling) procedure, was influenced by process parameters, such as the nozzle temperature, printing speed and CBA density. In order to verify the best processing conditions, the depositing parameters were experimentally investigated for the porous polymer structure. These results provide parameters under which to form a multiple of 3D porous polymer structures, as well as various other 3D structures, and help to improve the mechanical shock absorption for personal safety products.

• Nuclear Engineering and Technology
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• v.47 no.7
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• pp.791-803
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• 2015

The effective cross sections (XSs) in the direct whole core calculation code nTRACER are evaluated by the equivalence theory-based resonance-integral-table method using the WIMS-based library as an alternative to the subgroup method. The background XSs, as well as the Dancoff correction factors, were evaluated by the enhanced neutron-current method. A method, with pointwise microscopic XSs on a union-lethargy grid, was used for the generation of resonance-interference factors (RIFs) for mixed resonant absorbers. This method was modified by the intermediate-resonance approximation by replacing the potential XSs for the non-absorbing moderator nuclides with the background XSs and neglecting the resonance-elastic scattering. The resonance-escape probability was implemented to incorporate the energy self-shielding effect in the spectrum. The XSs were improved using the proposed method as compared to the narrow resonance infinite massbased method. The RIFs were improved by 1% in $^{235}U$, 7% in $^{239}Pu$, and >2% in $^{240}Pu$. To account for thermal feedback, a new feature was incorporated with the interpolation of pre-generated RIFs at the multigroup level and the results compared with the conventional resonance-interference model. This method provided adequate results in terms of XSs and k-eff. The results were verified first by the comparison of RIFs with the exact RIFs, and then comparing the XSs with the McCARD calculations for the homogeneous configurations, with burned fuel containing a mixture of resonant nuclides at different burnups and temperatures. The RIFs and XSs for the mixture showed good agreement, which verified the accuracy of the RIF evaluation using the proposed method. The method was then verified by comparing the XSs for the virtual environment for reactor applicationbenchmark pin-cell problem, as well as the heterogeneous pin cell containing burned fuel with McCARD. The method works well for homogeneous, as well as heterogeneous configurations.

• Nuclear Engineering and Technology
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• v.43 no.2
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• pp.133-140
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• 2011

In this paper, a deformation measurement technology for nuclear fuel rods is proposed. The deformation measurement system includes a high-definition CMOS image sensor, a lens, a semiconductor laser line beam marker, and optical and mechanical accessories. The basic idea of the proposed deformation measurement system is to illuminate the outer surface of a fuel rod with a collimated laser line beam at an angle of 45 degrees or higher. For this method, it is assumed that a nuclear fuel rod and the optical axis of the image sensor for observing the rod are vertically composed. The relative motion of the fuel rod in the horizontal direction causes the illuminated laser line beam to move vertically along the surface of the fuel rod. The resulting change of the laser line beam position on the surface of the fuel rod is imaged as a parabolic beam in the high-definition CMOS image sensor. An ellipse model is then extracted from the parabolic beam pattern. The center coordinates of the ellipse model are taken as the feature of the deformed fuel rod. The vertical offset of the feature point of the nuclear fuel rod is derived based on the displacement of the offset in the horizontal direction. Based on the experimental results for a nuclear fuel rod sample with a formation of surface crud, an inspection resolution of 50 ${\mu}m$ is achieved using the proposed method. In terms of the degree of precision, this inspection resolution is an improvement of more than 300% from a 150 ${\mu}m$ resolution, which is the conventional measurement criteria required for the deformation of neutron irradiated fuel rods.

• Nuclear Engineering and Technology
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• v.42 no.6
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• pp.670-679
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• 2010

In this paper, a new approach using a pixel-based correction method was developed to fix the non-uniform responses of flat-bed type scanners used for radiochromic film dosimetry. In order to validate the method's performance, two cases were tested: the first consisted of simple dose distributions delivered by a single port; the second was a complicated dose distribution composed of multiple beams. In the case of the simple individual dose condition, ten different doses, from 8.3 cGy to 307.1 cGy, were measured, horizontal profiles were analyzed using the pixel-based correcton method and compared with results measured by an ionization chamber and results corrected using the existing correction method. A complicated inverse pyramid dose distribution was made by piling up four different field shapes, which were measured with GAFCHROMIC$^{(R)}$EBT film and compared with the Monte Carlo calculation; as well as the dose distribution corrected using a conventional method. The results showed that a pixel-based correction method reduced dose difference from the reference measurement down to 1% in the flat dose distribution region or 2 mm in a steep dose gradient region compared to the reference data, which were ionization chamber measurement data for simple cases and the MC computed data for the complicated case, with an exception for very low doses of less than about 10 cGy in the simple case. Therefore, the pixel-based scanner correction method is expected to enhance the accuracy of GAFCHROMIC$^{(R)}$EBT film dosimetry, which is a widely used tool for two-dimensional dosimetry.

• Nuclear Engineering and Technology
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• v.42 no.6
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• pp.620-635
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• 2010

In this paper we describe current activities on the project Multi-Scale Modeling and Analysis of convective boiling (MSMA), conducted jointly by the Paul Scherrer Institute (PSI) and the Swiss Nuclear Utilities (Swissnuclear). The long-term aim of the MSMA project is to formulate improved closure laws for Computational Fluid Dynamics (CFD) simulations for prediction of convective boiling and eventually of the Critical Heat Flux (CHF). As boiling is controlled by the competition of numerous phenomena at various length and time scales, a multi-scale approach is employed to tackle the problem at different scales. In the MSMA project, the scales on which we focus range from the CFD scale (macro-scale), bubble size scale (meso-scale), liquid micro-layer and triple interline scale (micro-scale), and molecular scale (nano-scale). The current focus of the project is on micro- and meso-scales modeling. The numerical framework comprises a highly efficient, parallel DNS solver, the PSI-BOIL code. The code has incorporated an Immersed Boundary Method (IBM) to tackle complex geometries. For simulation of meso-scales (bubbles), we use the Constrained Interpolation Profile method: Conservative Semi-Lagrangian $2^{nd}$ order (CIP-CSL2). The phase change is described either by applying conventional jump conditions at the interface, or by using the Phase Field (PF) approach. In this work, we present selected results for flows in complex geometry using the IBM, selected bubbly flow simulations using the CIP-CSL2 method and results for phase change using the PF approach. In the subsequent stage of the project, the importance of effects of nano-scale processes on the global boiling heat transfer will be evaluated. To validate the models, more experimental information will be needed in the future, so it is expected that the MSMA project will become the seed for a long-term, combined theoretical and experimental program.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.446-454
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• 2021

Critical heat flux (CHF) has traditionally been evaluated using look-up tables or empirical correlations for nuclear power plants. However, under complex moving condition, it is necessary to reconsider the CHF characteristics since the conventional CHF prediction methods would no longer be applicable. In this paper, the additional forces caused by motions have been added to the annular film dryout (AFD) mechanistic model to investigate the effect of moving condition on CHF. Moreover, a theoretical model of the natural circulation loop with additional forces is established to reflect the natural circulation characteristics of the loop system. By coupling the system loop with the AFD mechanistic model, a CHF prediction program called NACOM for natural circulation loop under moving condition is developed. The effects of three operating conditions, namely stationary, inclination and rolling, on the CHF of the loop are then analyzed. It can be clearly seen that the moving condition has an adverse effect on the CHF in the natural circulation system. For the calculation parameters in this paper, the CHF can be reduced by 25% compared with the static value, which indicates that it is important to consider the effects of moving condition to retain adequate safety margin in subsequent thermal-hydraulic designs.