• Journal of Energy Engineering
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• v.25 no.1
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• pp.43-47
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• 2016

KHNP CRI has developed APR+ nuclear power plant since 2007, which is GEN III+ model with 1500 MWe capacity. To develop safer nuclear power plant than APR1400, we investigated advanced design features of ALWR being constructed in Korea and being developed/constructed in foreign countries. We applied the advanced design features and lessons learned from Fukushima accident to develop APR+ standard design suitable for both domestic construction and overseas construction business. One economic assessments have performed during safety design improvement phase(2013.1 ~ 2015.12) of APR+. The result of the economic analysis for APR+ safety inhancement design showed that APR+ N-th plant is about 39.2% more economical than coal-fired 1,000MW power plant. Also APR+ plant is more cost advantage over foreign advanced nation ALWRs.

• Journal of Radiation Protection and Research
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• v.28 no.2
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• pp.127-135
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• 2003

The primary contributors to the total occupational radiation exposure in operating nuclear power plants are operation and maintenance activities doting refueling outages. The Advanced Power Reactor 1400 (APR1400) includes a number of design improvements and plans to utilize advanced maintenance methods and robotics to minimize the annual collective dose. The major radiation exposure reduction features implemented in APR1400 are a permanent refueling pool seal, quick opening transfer tube blind flange, improved hydrogen peroxide injection at shutdown, improved permanent steam generator work platforms, and more effective temporary shielding. The estimated average annual occupational radiation exposure for APR1400 based on the reference plant experience and an engineering judgment is determined to be in the order of 0.4 man-Sv, which is well within the design goal of 1 man-Sv. The basis of this average annual occupational radiation exposure estimation is an eighteen (18) month fuel cycle with maintenance performed to steam generators and reactor coolant pumps during refueling outage. The outage duration is assumed to be 28 days. The outage work is to be performed on a 24 hour per day basis, seven (7) days a week with overlapping twelve (12) hour work shifts. The occupational radiation exposure for APR1400 is also determined by an alternate method which consists of estimating radiation exposures expected for the major activities during the refueling outage. The major outage activities that cause the majority of the total radiation exposure during refueling outage such as fuel handling, reactor coolant pump maintenance, steam generator inspection and maintenance, reactor vessel head area maintenance, decontamination, and ICI & instrumentation maintenance activities are evaluated at a task level. The calculated value using this method is in close agreement with the value of 0.4 man-Sv, that has been determined based on the experience aid engineering judgement. Therefore, with the As Low As Reasonably Achievable (ALARA) advanced design features incorporated in the design, APR1400 design is to meet its design goal with sufficient margin, that is, more than a factor of two (2), if operated on art eighteen (18) month fuel cycle.

• Nuclear Engineering and Technology
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• v.41 no.9
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• pp.1157-1170
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• 2009

A variety of Generation III/III+ reactor designs featuring enhanced safety and improved economics are being proposed by nuclear power industries around the world to solve the future energy supply shortfall. Nanofluid coolants showing an improved thermal performance are being considered as a new key technology to secure nuclear safety and economics. However, it should be noted that there is a lack of comprehensible design works to apply nanofluids to Generation III+ reactor designs. In this work, the review of accident scenarios that consider expected nanofluid mechanisms is carried out to seek detailed application spots. The Axiomatic Design (AD) theory is then applied to systemize the design of nanofluid-engineered nuclear safety systems such as Emergency Core Cooling System (ECCS) and External Reactor Vessel Cooling System (ERVCS). The various couplings between Gen-III/III+ nuclear safety features and nanofluids are investigated and they try to be reduced from the perspective of the AD in terms of prevention/mitigation of severe accidents. This study contributes to the establishment of a standard communication protocol in the design of nanofluid-engineered nuclear safety systems.

• Current Optics and Photonics
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• v.7 no.4
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• pp.419-427
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• 2023

The structural design of an optical-telescope assembly (OTA) for satellite laser ranging (SLR) is conducted in two steps. First, the results of a parametric study of the major design variables (e.g. dimension and shape) of the OTA part are explained, and the detailed structural design of the OTA is derived, considering the design requirements. Among the structural-shape concepts of various OTAs, the Serrurier truss concept is selected in this study, and the collimation of the telescope according to the design variables is extensively discussed. After generating finite-element models for different structural shapes, self-gravity analyses are performed. To minimize the deflection and tilt of the mirror and frame for the OTA under the limited design requirements, a parametric study is conducted according to design variables such as the shapes of the upper and lower struts and the spider vane. The structural features found in the parametric study are described. Finally, the OTA structure is designed in detail to maintain the optical alignment by balancing the gravity deflections of the upper and lower trusses using the optimal combination of the parameters. Additionally, thermal analysis of the optical telescope design is evaluated.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 1993.04b
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• pp.359-364
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• 1993

As advanced method for the automatic generation of parametric models in computer-aided design systems is required for most of two-dimensional model which is represented as a set of geometric elements, and constr- aining scheme formulas. The development system uses geometirc constrainis and topology parameters which are derived from feature recognition and grouping the design entities into optimal ones from pre-designed drawings. The aim of this paper is to present guidelines for the application and development of parametric design modules for the standard parts in mechaniscal system, the basic constitutional part of mold base, and other 2D features.

• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.293-305
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• 2015

In this paper, we suggest the conceptual design of a water-cooled reactor system for a low-pressure inherent heat sink nuclear desalination plant (LIND) that applies the safety-related design concepts of high temperature gas-cooled reactors to a water-cooled reactor for inherent and passive safety features. Through a scoping analysis, we found that the current LIND design satisfied several essential thermal-hydraulic and neutronic design requirements. In a thermal-hydraulic analysis using an analytical method based on the Wooton-Epstein correlation, we checked the possibility of safely removing decay heat through the steel containment even if all the active safety systems failed. In a neutronic analysis using the Monte Carlo N-particle transport code, we estimated a cycle length of approximately 6 years under 200 $MW_{th}$ and 4.5% enrichment. The very long cycle length and simple safety features minimize the burdens from the operation, maintenance, and spent-fuel management, with a positive impact on the economic feasibility. Finally, because a nuclear reactor should not be directly coupled to a desalination system to prevent the leakage of radioactive material into the desalinated water, three types of intermediate systems were studied: a steam producing system, a hot water system, and an organic Rankine cycle system.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.467-472
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• 1996

This study discusses special aspects of severe accident related recriticality modeling and analysis in the Advanced Neutron Source (ANS) reactor.$^{1, 2)}$ The analytical comparison of three elements core to former two elements case is conducted including evaluation of suitable nuclear cross-section sets to account for the effects of system configulation, fuel and moderator mixture temperature, material dispersion and the other thermal-hydraulics. Three elements core ANS reactor is the alternative core design which was proposed as a modified core design, with three fuel elements instead of two, that would allow operation with only 50% enriched uranium (former uranium fuel is the baseline design value of 93%) A comprehensive test matrix of calculations to evaluate the threat of a criticality event in the ANS is described. Strong dependencies still on geometry, material constituents, and thermal-hydraulic conditions are verified. Therefore, the concepts of mitigative design features are qualified.d.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.26 no.3
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• pp.207-212
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• 2013

APR1000(Advanced Power Reactor 1000) was developed to export 1000MW nuclear power plants by adding ADFs(Advanced Design Features) including 60 years design life, local frequency control operation, 0.3g SSE, etc. to OPR1000(Optimized Power Reactor 1000). In this paper, environmental fatigue analyses for the reactor vessel in APR1000 have been performed as per Reg. Guide 1.207. Outlet nozzle, which has a relatively high cumulative usage factor in the reactor vessel was evaluated and a structural integrity is maintained under the reactor coolant environment.

• Nuclear Engineering and Technology
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• v.27 no.5
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• pp.753-759
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• 1995

One of the important features of the advanced nuclear power plants is the system simplification. In this work, a model has been introduced to quantitatively evaluate the system simplification. A few models have been developed for quantitative evaluation of design simplification and the design enhancements of CVCS of the advanced reactors have been evaluated with models based on the entropy concept and the system availability. In addition, operational interface of CVCS with peripheral systems has been considered to develop a new evaluation model in this work. The quantification results for the design of the System 80+ and KSNPP indicate that the simplicity of the CVCS is primarily dependent on the type and number of charging pumps.

• 열병합발전
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• s.36
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• pp.16-21
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• 2003

For many years, T/G Supplier has constructed a number of thermal power plants and researched to improve the performance and the reliability of steam turbine, which are achieved by advances in design and materials technology. In recent, interest is renewed in advance steam condition as means of improving economy of thermal power plant and reducing environmental pollution. Improvements in the maximum power have been driven by the development of advanced rotor and bucket material and longer last stage bucket. Improvements in efficiency have been brought through advance in mechanical efficiency and thermodynamic efficiency. This paper describes a number of new steam path design features introduced to the steam turbine product. And also this paper describes new design technologies' development, new technologies' trend and technologies' development for ultra-super critical steam turbine.