• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2005년도 Proceedings of The 6th korea-china joint workshop on nuclear waste management
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• pp.76-85
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• 2005

The conceptual study of Accelerator Driven System (ADS) had lasted for about five years and ended in 1999 in China. As one project of 'the major state basic research program (973)' in energy domain, which is sponsored by the China Ministry of Science and Technology (MOST), a five years program of basic research for ADS physics and related technology has been launched since 2000 and passed national review last month. CIAE (China Institute of Atomic Energy), IHEP (Institute of High Energy Physics), PKU-IHIP (Institute of Heavy Ion Physics in Peking University) and other institutions are jointly carrying on the research. The research activities are focused on HPPA physics and technology, reactor physics of external source driven sub-critical assembly, nuclear data base and material study. For HPPA, a high current injector consisting of an ECR ion source, LEBT and a RFQ accelerating structure of 3.5MeV has been built. In reactor physics study, a series of neutron multiplication experimental study has been carried out and is being carrying on. The VENUS facility has been constructed as the basic experimental platform for the neutronics study in ADS blanket. It's a zero power sub-critical neutron multiplying assembly driven by external neutron produced by a pulsed neutron generator. The theoretical, experimental and simulation study on nuclear data, material properties and nuclear fuel circulation related to ADS is carrying on to provide the database for ADS system analysis. The main results on ADS related researches will be reported.

• Nuclear Engineering and Technology
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• 제45권4호
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• pp.477-488
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• 2013

The PLC (Programmable Logic Controller) has been widely used to implement real-time controllers in nuclear RPSs (Reactor Protection Systems). Increasing complexity and maintenance cost, however, are now demanding more powerful and cost-effective implementation such as FPGA (Field-Programmable Gate Array). Abandoning all experience and knowledge accumulated over the decades and starting an all-new development approach is too risky for such safety-critical systems. This paper proposes an RPS software development process with a platform change from PLC to FPGA, while retaining all outputs from the established development. This paper transforms FBD designs of the PLC-based software development into a behaviorally-equivalent Verilog program, which is a starting point of a typical FPGA-based hardware development. We expect that the proposed software development process can bridge the gap between two software developing approaches with different platforms, such as PLC and FPGA. This paper also demonstrates its effectiveness using an example of a prototype version of a real-world RPS in Korea.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1213-1224
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• 2024

The liquid lead-bismuth cooled fast reactor has been in a single-phase, low-pressure, and high-temperature state for a long time during operation. Considering the requirement of calculation efficiency for long-term transient accident calculation, based on a homogeneous hydrodynamic model, one-dimensional heat conduction model, coolant flow and heat transfer model, neutron kinetics model, coolant and material properties model, this study used the fully implicit difference scheme algorithm of the convection-diffusion term to solve the basic conservation equation, to develop the transient analysis program NUSOL-LMR 2.0 for the lead-bismuth fast reactor system. The steady-state and typical design basis accidents (including reactivity introduction, loss of flow caused by main pump idling, excessive cooling, and plant power outage accidents) for the ABR have been analyzed. The results are compared with the international system analysis software ATHENA. The results indicate that the developed program can stably, accurately, and efficiently predict the transient accident response and safety characteristics of the lead-bismuth fast reactor system.

• 한국콘크리트학회:학술대회논문집
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• 한국콘크리트학회 1998년도 가을 학술발표대회 논문집(III)
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• pp.649-655
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• 1998

In this study, the personal-computer program was developed to predict prestressing losses containment structures of Nuclear Power Plants by concrete creep and shrinkage. This program is constituted of three parts, which are pre-processor, calculation module and post-processor. Input data for his program are : material properties of concrete, rebar, liner and duct, test results of concrete creep and shrinkage, relative humidity, dimension of containment structures, and the number of prestressing tendon related on containment structures. To obtain better results, this program was made to reflect the prestressing losses due to influence that occurred after prestressing each tendon, thus it can predict prestressing losses and allowable prestressing forces of each tendon. As a case study, this program was applied to containment structures of Youngwang 3 & 4 NPP's and analytical result was compared with test results in In-service Inspection of containment structures. From this comparison, it was proved that this program could well predict prestressing losses by concrete creep and shrinkage.

• 한국압력기기공학회 논문집
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• 제6권1호
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• pp.65-71
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• 2010

The plugging rate of reheater tubes of Wolsung unit 1 nuclear power plant has been increased by corrosion and erosion since 1990. As the dimensions of the new first stage reheater bundle tubes which were supplied by Hanjung company to replace were different from old one, numerical calculations are carried out for flow and heat transfer in the reheater bundle tubes of the N.P.P. Numerical calculations consists of thermal performance, drain line pressure drop, flow change by pressure drop of line, stress analysis of finned tubes and analysis of flow induced vibration. Computational analysis using heat transfer research institute program is adopted to verify the results of the numerical calculations. It contains the evalution of performance in the system with view to location of the new reheater bundle and it shows the differences between the numerical calculation results and heat transfer research institute program output.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(2)
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• pp.285-290
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• 1998

KALIMER 고속로의 기본 핵연료인 U-l0Zr 핵연료봉의 노내 성능시험을 위해, 러시아의 BR-10 연구용 고속원자로에서 핵연료노내조사 Program이 1997년부터 수행되고 있다. 1 차년도에는 핵연료 시편의 설계 및 제조와 금속합금 핵연료의 균질도, 밀도, 열전도도 등의 노외 특성 시험이 수행되었다. U-l0Zr 핵연료심은 Arc 용해로 제조되었는데, 합금의 구성 원소들은 비교적 균일하게 분포되었다. 핵연료 시편은 2 개가 제작되었는데, BR-10 원자로에서 각각 연소도 1.08 % 및 2.15 %까지 연소된 후, 조사 후 검사가 수행될 것이다. 금속핵연료는 대개 낮은 연소도에서 급격한 변화틀 보이기 때문에, 본 핵연료 노내조사시험 Program의 결과는 금속핵연료봉의 성능해석 모델 개발에 활용될 수 있을 것이다.

• Journal of Radiation Protection and Research
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• 제26권3호
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• pp.215-224
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• 2001

In order to comply with revised national regulationson radiological protection and to implement recent nuclear data and dose conversion factors, KOPEC developed a new point kernel gamma and beta ray shielding analysis computer program. This new code, named VisualShield, adopted mass attenuation coefficient and buildup factors from recent ANSI/ANS standards and flux-to-dose conversion factors from the International Commission on Radiological Protection (ICRP) Publication 74 for estimation of effective/equivalent dose recommended in ICRP 60. VisualShieid utilizes graphical user interfaces and 3-D visualization of the geometric configuration for preparing input data sets and analyzing results, which leads users to error free processing with visual effects. Code validation and data analysis were performed by comparing the results of various calculations to the data outputs of previous programs such as MCNP 4B, ISOSHLD-II, QAD-CGGP, etc.

• 비파괴검사학회지
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• 제9권2호
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• pp.67-72
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• 1989

In nuclear power plants, nondestructive examinations(NDE) plays an important role in ensuring the integrity and reliability operation. As the number of plants and operational time increased, manual handling of voluminous data associated with PSI/ISI(preservice/inservice inspection) could result in many errors or mistakes in writing the examination plan or other reports. Several new approaches to process the data have been attempted and DBMS(Data Base Management System) has been well known concept with a faster and more accurate data processing. This paper proposes an application program, called NDTSYS designed with DBMS in micro computer. The program could be used for a tool to add new records to a data base, change existing records, delete records and request reports with the data base. It would be helpful to the user who manage the PSI/ISI data with minimal time and effort.

• Journal of Advanced Marine Engineering and Technology
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• 제22권5호
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• pp.609-616
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• 1998

Structural analyses are preformed for the high pressure steam turbine casings of the nuclear and the fossil power plants. An axisymmetric boundary element program for analysis of the casings is developed and applied in the process of practical structural design. To show the useful-ness and accuracy of the developed program results of the analysis are compared with those of the finite element analysis under hydrostatic test pressure, To check the validity of the axisymmetric numerical analysis of the casings the stresses resulting from the hydrostatic test pressure are measured using the strain gate. The results of the numerical analyses are compared and discussed with those of the experiments.

• Nuclear Engineering and Technology
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• 제33권1호
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• pp.62-72
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• 2001

The one-dimensional finite element program was developed to analyze the coupled behavior of heat, moisture, and air transfer in unsaturated porous media. By using this program, the simulation results were compared with those from the laboratory infiltration tests under isothermal condition and temperature gradient condition, respectively. The discrepancy of water uptake was found in the upper region of a bentonite sample under isothermal condition between numerical simulation and laboratory experiment. This indicated that air pressure was built up in the bentonite sample which could retard the infiltration velocity of liquid. In order to consider the swelling phenomena of compacted bentonite which cause the discrepancy of the distribution of water content and temperature, swelling and shrinkage factors were incorporated into the finite element formulation. It was found that these factors could be effective to represent the moisture diffusivity and unsaturated hydraulic conductivity due to volume change of bentonite sample.