• Nuclear Engineering and Technology
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• 제35권5호
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• pp.399-411
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• 2003

A model predictive control method is applied to design an automatic controller for thermal power control in a reactor core. The basic concept of the model predictive control is to solve an optimization problem for a finite future at current time and to implement as the current control input only the first optimal control input among the solutions of the finite time steps. At the next time step, the second optimal control input is not implemented and the procedure to solve the optimization problem is then repeated. The objectives of the proposed model predictive controller are to minimize the difference between the output and the desired output and the variation of the control rod position. The nonlinear PWR plant model (a nonlinear point kinetics equation with six delayed neutron groups and the lumped thermal-hydraulic balance equations) is used to verify the proposed controller of reactor power. And a controller design model used for designing the model predictive controller is obtained by applying a parameter estimation algorithm at an initial stage. From results of numerical simulation to check the controllability of the proposed controller at the $5\%/min$ ramp increase or decrease of a desired load and its $10\%$ step increase or decrease which are design requirements, the performances of this controller are proved to be excellent.

• Journal of Radiation Protection and Research
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• 제32권4호
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• pp.168-177
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• 2007

본 논문에서는 노심설계 단계에서 선정된 다양한 노심 장전모형 중에서 압력용기 중성자 조사취화 관점에서 가장 최적의 노심 장전모형을 선정할 수 있도록 신속하게 압력용기 취약위치에 대한 속중성자속을 예측할 수 있는 방법을 제시하였다. 인공신경회로망 기법을 통해 노심 반경방향 및 축방향 출력분포만을 이용하여 압력용기내벽 취약위치에서의 중성자 스펙트럼을 신속하게 평가할 수 있도록 중성자속 가중치를 생산하였고 데이터베이스를 구축하였다. 이 방법은 중성자 수송코드를 이용한 수송계산을 직접 수행하지 않고도 신속하게 압력용기 위치에서의 중성자 조사환경을 평가할 수 있으며 소송코드 결과와 비교하여 상대오차 3.4%이내의 정확도를 보였다.

• 한국환경과학회지
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• 제27권6호
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• pp.467-475
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• 2018

The lifetime of the electrode is one of the most important factors on the stability of the electrode. Since the lifetime of the DSA (Dimensionally stable anode) electrode is long, an accelerated lifetime test is required to reduce the test time. Beacuse there is no basis or standard method for accelerated lifetime testing, many researchers use different methods. Therefore, there is a need for basis and methods for accelerated lifetime testing that other researchers can follow. We designed a reactor system for accelerated lifetime testing and planned specific methods. Reactor system was circulating batch reactor. Reactor volume and cooling water tank were 12.5 L and 100 L, respectively. Electrode size was $2cm{\times}3cm$ (real electrolysis area, $5cm^2$). In order to maintain the harsh conditions, accelerated lifetime test was carried out in a high current density ($0.6A/cm^2$) and low electrolyte concentration (NaCl, 0.068 mol/L). Maintaining a constant temperature was an important operation parameter for exact accelerated lifetime test. As the accelerated lifetime test progressed, the active component of electrode surface was consumed and desorption occurred. At the point of 5 V rise, corrosion of the surface of the base material(titanium) also started.

• Nuclear Engineering and Technology
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• 제48권2호
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• pp.360-367
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• 2016

A windowless spallation target can provide a neutron source and maintain neutron chain reaction for a subcritical reactor, and is a key component of China's nuclear waste transmutation of coupling accelerator and subcritical reactor. The main issue of the windowless target design is to form a stable and controllable free surface that can ensure that energy spectrum distribution is acquired for the neutron physical design when the high energy proton beam beats the lead-bismuth eutectic in the spallation target area. In this study, morphology and flow characteristics of the free surface of the windowless target were analyzed through the volume of fluid model using computational fluid dynamics simulation, and the results show that the outlet cross section size of the target is the key to form a stable and controllable free surface, as well as the outlet with an arc transition. The optimization parameter of the target design, in which the radius of outlet cross section is $60{\pm}1mm$, is verified to form a stable and controllable free surface and to reduce the formation of air bubbles. This work can function as a reference for carrying out engineering design of windowless target and for verification experiments.

• 한국압력기기공학회 논문집
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• 제16권1호
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• pp.49-55
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• 2020

Creep rupture data for Alloy 690 steam generator tubes in a pressurized water reactor are essentially needed to demonstrate a severe accident scenario on thermally-induced tube failures caused by hot gases in a damaged reactor core. The rupture data were obtained using the tube specimens under different applied-stress levels at 650℃, 700℃, 750℃, 800℃, and 850℃. Important creep constants were proposed using various creep laws in terms of Norton power law, Monkman-Grant (M-G) relation, damage tolerance factor (λ), and Zener-Hollomon parameter (Z). In addition, a creep activation energy (Q) value for Alloy 690 tube was reasonably determined using experimental data. Creep behaviors such as creep strength, creep rates, rupture elongation showed the results of temperature dependence well. Modified M-G plot improved a correlation of the creep rate and rupture life. Damage tolerance factor for Alloy 690 tubes was found to be λ =2.20 in an average value. Creep activation energy for Alloy 690 tube was optimized for Q=350 (kJ/mol). A plot of Z parameter obeyed a good linearity, and the same creep mechanism was inferred to be operative in the present test conditions.

• 상하수도학회지
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• 제27권4호
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• pp.457-468
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• 2013

Based on the activated sludge model(ASM), a mathematical model which represents the aerobic sludge digestion by sequencing batch reactor(SBR) combined with ultrasonic treatment was composed and performed in this study. Aerobic digestion using sequencing batch reactor(SBR) equipped with ultrasound treatment was also experimented for the purpose of parameter calibration. Most of the presented kinetic parameters in ASM or ASM2 could be used for the aerobic digestion of sludge but the parameters related in hydrolysis and decay rate needed modification. Hydrolysis rate constant of organic matter in aerobic condition was estimated at $0.3day^{-1}$ and the maximum growth rate for autotrophs in aerobic condition was $0.618day^{-1}$. Solubilization reactions of particulate organics and nitrogen by ultrasonication was added in this kinetic model. The solubilization rate is considered to be proportional to the specific energy which is defined by specific ultrasound power and sonication time. The solubilization rate constant by ultrasonication was estimated at $0.202(W/L)^{-1}day^{-1}$ in this study. Autotrophs as well as heterotrophs also decomposed by ultrasonic treatment and the nitrification reaction was limited by the lack of autotrophs accumulation in the digester.

• 한국구조물진단유지관리공학회 논문집
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• 제5권2호
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• pp.147-154
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• 2001

The prestressed concrete containment is one of the most important structures in nuclear power plants, which is required to prevent release of radioactive or hazardous effluents to the environment even in the case of a severe accident. Numerical analyses are carried out by using the ABAQUS finite element program to assess the ultimate pressure capacity of the Y prestressed concrete containment with light water reactor at design criteria condition and aging condition considering varied properties of time-dependant materials respectively. From the results, it is verified that the structural capacity of the Y prestressed concrete containment building under the present, aging condition is still robust. In addition, the parameter studies for the reduction of the ultimate pressure capacity of containment building according to the degradation levels of the main structural materials are carried out. The results show that when the degradations of each materials are considered as individual and combined forms, the influence is large in the order of tendon, rebar and concrete degradation, and tendon-rebar, tendon-concrete and rebar-concrete degradation respectively.

• Nuclear Engineering and Technology
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• 제54권2호
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• pp.493-500
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• 2022

The objective of this study was to derive derived concentration guideline levels (DCGLs) reflecting the site-specific characteristics of KRR-1&2. A total of 7 nuclides (H-3, C-14, Co-60, Sr-90, Cs-137, Eu-152, and Eu-154) were selected for DCGLs derivation. Radiation dose at the sites was evaluated with RESRAD-ONSITE program. The dose contribution due to direct external exposure was the highest during the entire evaluation period. Ingestion had the second effect. The DCGLs of Co-60 was derived to be 0.051 Bq/g, and DCGLs of Cs-137 was 0.193 Bq/g. The DCGLs of H-3 showed the highest value of 129 Bq/g. The ratio of DCGLs derived by applying site-specific values and default values ranged from 0.27 to 19.6. For six nuclides excluding H-3, KRR-1&2 sites and the overseas NPP sites showed similar DCGLs. H-3 showed large differences in DCGLs from this study and overseas NPPs. The large difference resulted from input parameter values applied to the sites. In conclusion, it is critical to apply site-specific parameter values reflecting the site characteristics to derive DCGLs for decommissioned site clearance. The result of this study can be used as a reference for nuclide selection and DCGLs derivation reflecting the site characteristics when decommissioning nuclear facilities, including nuclear power plants in Korea.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2821-2835
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• 2020

The phenomenon identification and ranking table (PIRT) is an important basis in the nuclear power plant (NPP) thermal-hydraulic analysis. This study focuses on the importance ranking of the input parameters when lacking the PIRT, and the target scenario is the small break loss of coolant accident (SBLOCA) in a pressurized water reactor (PWR) CPR1000. A total of 54 input parameters which might have influence on the figure of merit (FOM) were identified, and the sensitivity measure of each input on the FOM was calculated through an optimized moment-independent global sensitivity analysis method. The importance ranking orders of the parameters were transformed into the Savage scores, and the parameters were categorized based on the Savage scores. A parameter importance ranking table for the SBLOCA scenario of the CPR1000 reactor was obtained, and the influences of some important parameters at different break sizes and different accident stages were analyzed.

• 유기물자원화
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• 제8권4호
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• pp.160-167
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• 2000

음식물 퇴비화를 위한 토양미생물제재의 퇴비화 효율을 평가하며, 토양미생물제재의 미생물수와 퇴비화과정에서 식종의 효과를 조사하였다. 분석대상 음식물은 본 연구원 구내식당에서 수거한 후 물리화학적특성을 분석하였다. 대상시료는 Bulking Agent로서 톱밥을 사용하여 함수율을 65%로 조정 후 반응기 B에 10%의 토양미생물제를 식종하였다. 토양미생물제제의 미생물은 호기성세균수가 $2.98{\times}10^9/g$이상, 방선균이 $3.93{\times}10^7/g$이상, 효모가 $1.21{\times} 10^5/g$, 균류가 $5.79{\times}10^7/g$ 이상이었다. 퇴비화 기간동안 최고온도는 Reactor A가 반응 10일후에 개시후 바로 급격한 변화를 보였으며, 반응이 종료된 후 두 Reactor 공히 pH8.9를 나타내었다. Reactor B의 경우 최대온도인 4일후에 $CO_2$농도도 역시 최대인 10.8%를 나타냈으며, Reactor A는 최대온도인 10일후에 6.1%를 나타내었다. 한편 Reactor A는 $r_{d}$(유기물질의 분해율)치가 0.35에서 0.41로 17.1%상승하였으며, Reactor B의 경우 0.31에서 0.51로 64.5%상승하였다.