• 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.40 no.1
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• pp.49-68
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• 2008

The purpose of this paper is to extensively review the condition monitoring (CM) techniques using empirical models in an effort to reduce or eliminate unexpected downtimes in general industry, and to illustrate the feasibility of applying them to the nuclear industry. CM provides on-time warnings of system states to enable the optimal scheduling of maintenance and, ultimately, plant uptime is maximized. Currently, most maintenance processes tend to be either reactive, or part of scheduled, or preventive maintenance. Such maintenance is being increasingly reported as a poor practice for two reasons: first, the component does not necessarily require maintenance, thus the maintenance cost is wasted, and secondly, failure catalysts are introduced into properly working components, which is worse. This paper first summarizes the technical aspects of CM including state estimation and state monitoring. The mathematical background of CM is mature enough even for commercial use in the nuclear industry. Considering the current computational capabilities of CM, its application is not limited by technical difficulties, but by a lack of desire on the part of industry to implement it. For practical applications in the nuclear industry, it may be more important to clarify and quantify the negative impact of unexpected outcomes or failures in CM than it is to investigate its advantages. In other words, while issues regarding accuracy have been targeted to date, the concerns regarding robustness should now be concentrated on. Standardizing the anticipated failures and the possibly harsh operating conditions, and then evaluating the impact of the proposed CM under those conditions may be necessary. In order to make the CM techniques practical for the nuclear industry in the future, it is recommended that a prototype CM system be applied to a secondary system in which most of the components are non-safety grade. Recently, many activities to enhance the safety and efficiency of the secondary system have been encouraged. With the application of CM to nuclear power plants, it is expected to increase profit while addressing safety and economic issues.

• Nuclear Engineering and Technology
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• v.38 no.1
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• pp.11-32
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• 2006

System thermal-hydraulic codes have been used in the past decades in the areas of design, operation, licensing and safety of Nuclear Power Plants (NPPs). The development and validation of these codes have reached a high degree of maturity, through the consideration of huge experiments and advanced numerical models. Nowadays, the analyses are based upon realistic approaches rather than the conservative evaluation models. However the applications of these computational tools require preliminary qualification issues. Although huge amounts of financial and human resources have been invested for the development and improvement of codes, the calculation results are still affected by errors. In the sophisticated nuclear technology, design and safety of NPP, these errors must be quantified. An overview of the state of the art of the current thermal-hydraulic system code is developed and the need of uncertainty analysis in code calculations is emphasized. Several sources of uncertainty have been classified and commented, and typical applications of such methods are shown.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.968-976
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• 2019

This paper presents a number of verification case studies for a recently developed sensitivity/uncertainty code package. The code package, ROMUSE (Reduced Order Modeling based Uncertainty/Sensitivity Estimator) is an effort to provide an analysis tool to be used in conjunction with reactor core simulators, in particular the Virtual Environment for Reactor Applications (VERA) core simulator. ROMUSE has been written in C++ and is currently capable of performing various types of parameter perturbations and associated sensitivity analysis, uncertainty quantification, surrogate model construction and subspace analysis. The current version 2.0 has the capability to interface with the Design Analysis Kit for Optimization and Terascale Applications (DAKOTA) code, which gives ROMUSE access to the various algorithms implemented within DAKOTA, most importantly model calibration. The verification study is performed via two basic problems and two reactor physics models. The first problem is used to verify the ROMUSE single physics gradient-based range finding algorithm capability using an abstract quadratic model. The second problem is the Brusselator problem, which is a coupled problem representative of multi-physics problems. This problem is used to test the capability of constructing surrogates via ROMUSE-DAKOTA. Finally, light water reactor pin cell and sodium-cooled fast reactor fuel assembly problems are simulated via SCALE 6.1 to test ROMUSE capability for uncertainty quantification and sensitivity analysis purposes.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.618-625
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• 2021

In this paper, the single-photon avalanche diodes (SPADs) featuring three different p-well implantation doses (∅_p-well) of 5.0 × 10¹², 4.0 × 10¹², and 3.0 × 10¹² atoms/cm² under the identical device layouts were fabricated and characterized to evaluate the effects of field enhanced mechanisms on primary dark pulses due to the maximum electric field. From the I-V curves, the breakdown voltages were found as 23.2 V, 40.5 V, and 63.1 V with decreasing ∅_p-well, respectively. By measuring DCRs as a function of temperature, we found a reduction of approximately 8% in the maximum electric field lead to a nearly 72% decrease in the DCR at V_ex = 5 V and T = 25 ℃. Also, the activation energy increased from 0.43 eV to 0.50 eV, as decreasing the maximum electric field. Finally, we discuss the importance of electric field engineering in reducing the field-enhanced mechanisms contributing to the DCR in SPADs and the benefits on the SPADs related to different types of radiation detection applications.

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.2734-2741
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• 2023

Brachytherapy is a special case of radiotherapy. It should be arranged according to some principles in medical radiation applications and radiation physics. The primary principle is to use as low as reasonably achievable dose in all ionizing radiation applications for diagnostic and therapeutic treatments. Dosimetric distributions are dependent on radioactive source properties and radiation-matter interactions in an absorber medium such as phantom or tissue. In this consideration, the geometrical structure and material of the seed capsule, which surrounds a radioactive material, are directly responsible for isodose profiles and dosimetric functions. In this study, the radiometric properties of capsule material were investigated on dose distribution in a water phantom by changing its nuclear properties using the EGSnrc Monte Carlo (MC) simulation code. Effective atomic numbers of hypothetic mixtures were calculated by using different elements with several fractions for capsule material. Model 6711 brachytherapy seed was modeled by EGSnrc/Dosrcnrc Code and dosimetric functions were calculated. As a result, dosimetric parameters of hypothetic sources have been acquired in large-scale atomic number. Dosimetric deviations between the data of hypothetic seeds and the original one were analyzed. Unit dose (Gy/Particle) distributions belonging to different types of material in seed capsule have remarkably differed from the original capsule's data. Capsule type is major variable to manage the expected dose profile and isodose distribution around a seed. This study shows us systematically varied scale of material type (cross section or effective atomic number dependent) offers selective material usage in production of seed capsules for the expected isodose profile of a specific source.

• Proceedings of the CALSEC Conference
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• 2003.09a
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• pp.45-53
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• 2003

This paper addresses the issues related to the development of product model and applications fer nuclear power plants based on STEP and PLIB standards. The ISO standards which can be applied are; STEP(Standard for the Exchange of Product Model Data) AP(application protocol) 221/231, AP 230/225, AP 227, ISO 13584 PLIB, ISO 15926 RDL. The data models of the AP's and ISO 15926 RDL are reviewed and an application system is proposed to exchange and share the design data of the nuclear power plant.

• International Journal of Reliability and Applications
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• v.2 no.2
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• pp.137-146
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• 2001

In general, analysis of the decision problems in nuclear system management involves a simultaneous consideration of various criteria and decision alternatives. Sometimes, it is a complex, unstructured, ill-defined process incorporating the multi-criteria and the data of impreciseness. To cope with this analysis, a fuzzy hierarchical analysis methodology is proposed and demonstrated with a simple example.

• Nuclear Medicine and Molecular Imaging
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• v.41 no.2
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• pp.85-90
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• 2007

Acute event in cerebrovascular disease is the second most common cause of death in Korea following cancer, and it can also cause serious neurologic deficits. Understanding of perfusion status is important for clinical applications in management of patients with cerebrovascular diseases, and then the attacks of ischemic neurologic symptoms and the risk of acute events can be reduced. Therefore, the normal vascular anatomy of brain, various clinical applications of acetazolamide-enhanced brain perfusion SPECT, including meaning and role of assessment of vascular reserve in carotid stenosis before procedure, in pediatric Moyamoya disease before and after operation, in prediction of development of hyperperfusion syndrome before procedure, and in prediction of vasospasm and of prognosis in subarachnoid hemorrahge were reviewed in this paper.

• Nuclear Engineering and Technology
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• v.31 no.6
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• pp.529-540
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• 1999

As part of IAEA CRP on "Compilation and evaluation of photonuclear data for applications", we evaluated photoproduction data of Mo, Zn, S and Cl isotopes for medical use and biological applications. Available experimental data were collected and their discrepancies were analyzed to select or reconstruct the representative data set. The photoabsorption cross sections were then evaluated tv applying the Giant Dipole Resonance (GDR) model for the energies below about 30 MeV and the quasi-deuteron model for energies below 140 MeV. The resulting representative photoabsorption data were given as input for the theoretical calculations for the emission process of light nuclei including neutron, proton, deuteron, triton, $^3He$, alpha particles and gamma rays by use of the Hauser-Feshbach and the preequilibrium model.