• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1540-1553
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• 2019

A nuclear power plant (NPP) is a highly complex system-of-systems as manifested through its internal systems interdependence. The negative impact of such interdependence was demonstrated through the 2011 Fukushima Daiichi nuclear disaster. As such, there is a critical need for new strategies to overcome the limitations of current risk assessment techniques (e.g. the use of static event and fault tree schemes), particularly through simulation of the nonlinear dynamic feedback mechanisms between the different NPP systems/components. As the first and key step towards developing an integrated NPP dynamic probabilistic risk assessment platform that can account for such feedback mechanisms, the current study adopts a system dynamics simulation approach to model the thermal dynamic processes in: the reactor core; the secondary coolant system; and the pressurized water reactor. The reactor core and secondary coolant system parameters used to develop system dynamics models are based on those of the Palo Verde Nuclear Generating Station. These three system dynamics models are subsequently validated, using results from published work, under different system perturbations including the change in reactivity, the steam valve coefficient, the primary coolant flow, and others. Moving forward, the developed system dynamics models can be integrated with other interacting processes within a NPP to form the basis of a dynamic system-level (systemic) risk assessment tool.

• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1436-1445
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• 2021

To obtain the dynamic characteristics of reactor secondary circuit under transient conditions, the system analysis program was developed in this study, where dynamic models of secondary circuit were established. The heat transfer process and the mechanical energy transfer process are modularized. Models of main equipment were built, including main turbine, condenser, steam pipe and feedwater system. The established models were verified by design value. The simulation of the secondary circuit system was conducted based on the verified models. The system response and characteristics were investigated based on the parameter transients under emergency shutdown and overload. Various operating conditions like turbine emergency shutdown and overspeed, condenser high water level, ejector failures were studied. The secondary circuit system ensures sufficient design margin to withstand the pressure and flow fluctuations. The adjustment of exhaust valve group could maintain the system pressure within a safe range, at the expense of steam quality. The condenser could rapidly take out most heat to avoid overpressure.

• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.1998-2008
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• 2020

Solenoid operated valves (SOV) play important roles in industrial process to control the flow of fluids. Solenoid valves can be found in so many industries as well as the nuclear plant. The ability to be able to detect the presence of faults and predicting the remaining useful life (RUL) of the SOV is important in maintenance planning and also prevent unexpected interruptions in the flow of process fluids. This paper proposes a fault diagnosis method for the alternating current (AC) powered SOV. Previous research work have been focused on direct current (DC) powered SOV where the current waveform or vibrations are monitored. There are many features hidden in the AC waveform that require further signal analysis. The analysis of the AC powered SOV waveform was done in the time and frequency domain. A total of sixteen features were obtained and these were used to classify the different operating modes of the SOV by applying a machine learning technique for classification. Also, a deep neural network (DNN) was developed for the prediction of RUL based on the failure modes of the SOV. The results of this paper can be used to improve on the condition based monitoring of the SOV.

• Journal of the Korean Society of Systems Engineering
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• v.11 no.1
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• pp.1-7
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• 2015

The purpose of this paper is to present the results of reengineering of the YGN 3&4 (Yonggwang Nuclear Power Plant, Units 3&4) SDS (Safety Depressurization System) retrofit design and to make recommendations for the improvement in design and design procedure implementing the Systems Engineering (SE) process. YGN 3&4 is a basic model for OPR1000 (the Korean standard 1000 MWe plant). The basic model, herein, represents the reference plant for the OPR1000 development. In the middle of the YGN 3&4 construction, the Korean Nuclear Regulatory Body requested a retrofit of this plant with a means to rapidly depressurize the plant in conformance with a severe accident mitigation requirement. For the reengineering of the SDS in YGN 3&4, V-model and functional and physical architectures have been developed. A SE decision making method has been used for the selection of SDS valves. Finally, recommendations have been made to improve OPR1000 design for the improved operation and enhanced safety.

• Nuclear Engineering and Technology
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• v.34 no.4
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• pp.358-369
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• 2002

Abundant information about analyses of single steam generator tube rupture (SGTR) events is available because of its importance in terms of safety. However, there are few literatures available on analyses of multiple steam generator tube rupture (MSGTR) events. In addition, knowledge of transients and consequences following a MSGTR event are very limited as there has been no occurrence of MSGTR event in the commercial operation of nuclear reactors. In this study, a postulated MSGTR event in an APR1400 is analyzed using thermal-hydraulic system code MARSI.4. The present study aims to examine the effects of affected steam generator selection. The main steam safety valve (MSSV) lift time for four cases are compared in order to examine how long operator response time is allowed depending on which steam generate. (S/G) is affected. The comparison shows that the cases where two steam generators are simultaneously affected allow longer time for operator action compared with the cases where a single steam generator is affected. Furthermore, the tube ruptures in the steam generator where a pressurizer is connected leads to the shortest operator response time.

• Journal of Radiopharmaceuticals and Molecular Probes
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• v.2 no.2
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• pp.123-131
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• 2016

Increasing clinical demand for carbon-11 labeled radiopharmaceuticals has triggered technological advances in fields of radiochemistry and automated modules. Even though carbon-11 has a short half-life ($t_{1/2}=20.4min$), the consecutive second production of carbon-11 labeled radiopharmaceutical in one $^{11}C$-synthetic module should be delayed at least over 4 h to avoid the high radiation exposure. We herein aimed to produce two different carbon-11 labeled radiopharmaceuticals ([$^{11}C$]PIB and [$^{11}C$]methionine) by sharing of [$^{11}C$]methylation source in one $^{11}C$-synthetic module. The synthesis of $^{11}C$-labeling reagents ($[^{11}C]CH_3I$ or $[^{11}C]CH_3OTf$) is fully automated using the commercial TRACERlab $FX_{C-pro}$ module and is readily adaptable to $^{11}C$-labeling reactor for [$^{11}C$]PIB as well as another $^{11}C$-labeling apparatus for [$^{11}C$]methionine via the three-way valve. After completing the [$^{11}C$]PIB production, the re-synthesized $[^{11}C]CH_3I$ was passed through the three-way valve connected the polyetheretherketone (PEEK) line and loaded into the C18 Sep-Pak cartridge including the methionine precursor. The labeled product [^${11}C$]methionine was purified by a simple cartridge separation and reformulated into saline. The radiochemical yield of [$^{11}C$]PIB and [$^{11}C$]methionine were $5.3{\pm}0.6%$ and $18.7{\pm}0.8%$ (n.d.c.), respectively, with over 97% of radiochemical purity. The specific activity of [$^{11}C$]PIB was over $110GBq/{\mu}mol$. Total production time of two radiopharmaceuticals needs about 2 h from $1^{st}$ beam irradiation including quality control tests. Final [$^{11}C$]PIB and [$^{11}C$]methionine were satisfied all quality control test standards.

• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3745-3765
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• 2022

The concrete structures related to nuclear safety are threatened by accidental impact loadings, mainly including the low-velocity drop-weight impact (e.g., spent fuel cask and assembly, etc. with the velocity less than 20 m/s) and high-speed projectile impact (e.g., steel pipe, valve, turbine bucket, etc. with the velocity higher than 20 m/s), while the existing studies are still limited in the impact resistant design of nuclear power plant (NPP), especially the primary RC slab. This paper aims to propose the numerical simulation and theoretical approaches to assist the impact-resistant design of RC slab in NPP. Firstly, the continuous surface cap (CSC) model parameters for concrete with the compressive strength of 20-70 MPa are fully calibrated and verified, and the refined numerical simulation approach is proposed. Secondly, the two-degree freedom (TDOF) model with considering the mutual effect of flexural and shear resistance of RC slab are developed. Furthermore, based on the low-velocity drop hammer tests and high-speed soft/hard projectile impact tests on RC slabs, the adopted numerical simulation and TDOF model approaches are fully validated by the flexural and punching shear damage, deflection, and impact force time-histories of RC slabs. Finally, as for the two low-velocity impact scenarios, the design procedure of RC slab based on TDOF model is validated and recommended. Meanwhile, as for the four actual high-speed impact scenarios, the impact-resistant design specification in Chinese code NB/T 20012-2019 is evaluated, the over conservation of which is found, and the proposed numerical approach is recommended. The present work could beneficially guide the impact-resistant design and safety assessment of NPPs against the accidental impact loadings.

• Nuclear Engineering and Technology
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• v.47 no.4
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• pp.443-453
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• 2015

A hybrid safety injection tank (H-SIT) can enhance the capability of an advanced power reactor plus (APR+) during a station black out (SBO) that is accompanied by a severe accident. It may a useful alternative to an electric motor. The operations strategy of the H-SIT has to be investigated to achieve maximum utilization of its function. In this study, the master logic diagram (i.e., an analysis for identifying the differences between an H-SIT and a safety injection pump) and an accident case classification were used to determine the parameters of the H-SIT operation. The conditions that require the use of an H-SIT were determined using a decision-making process. The proper timing for using an H-SIT was also analyzed by using the Multi-dimensional Analysis of Reactor Safety (MARS) 1.3 code (Korea Atomic Energy Research Institute, Daejeon, South Korea). The operation strategy analysis indicates that a H-SIT can mitigate five types of failure: (1) failure of the safety injection pump, (2) failure of the passive auxiliary feedwater system, (3) failure of the depressurization system, (4) failure of the shutdown cooling pump (SCP), and (5) failure of the recirculation system. The results of the MARS code demonstrate that the time allowed for recovery can be extended when using an H-SIT, compared with the same situation in which an H-SIT is not used. Based on the results, the use of an H-SIT is recommended, especially after the pilot-operated safety relief valve (POSRV) is opened.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.403-410
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• 1995

A loss of the residual heat removal (RHR) accident during mid-loop operation of Yong-gwang Nuclear Units 3/4 was analyzed using the RELAP5/MOD3.1.2 code. In this work the following assumptions are used; (i) initially the reactor coolant system (RCS) above the hot leg center line is filled with nitrogen gas, (ii) two 3/4-inch diameter vent valves on the reactor vessel head and the top of pressurizer in the reactor coolant system are always open, and a level indicator is connected to the RMR suction line, (iii) the two steam generators are in wet layup status and the steam generator atmospheric dump valve assemblies are removed so that the secondary side pressure remains at nearly atmospheric condition throughout the accident, and (iv) the loss of RHR is presumed to occur at 48 hours after reactor shutdown. Findings from the RELAP5 calculations are (i) the core boiling begins at ∼5 min, (ii) the peak RCS pressure is ∼3.0 bar, which implies a possibility of temporary seal break, (iii) ∼94 % of the decay heat is removed by reflux condensation in the steam generator U-tubes in spite of the presence of noncondensable gas, (iv) the core uncovery time is evaluated to be 7.2 hours. Significant mass errors were observed in the calculations.

• Nuclear industry
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• no.6
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• pp.20-22
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• 1979

본문은, 일본의 통산성${\cdot}$에너지청이 4월 24일 일본의 원발20기에 대해서 행한 안전총점검의 중간보고를 요약한 것이다. 여기서, 보수관리상황을 점검한 주요설비는 다음과 같다. (1) 보조급수계 (2) 가압기방출 Valve (3) 비상용노심냉각장치 1) 고압주입계 2) 축압 Tank 3) 저압주입계 4) 격납용기내부 Spray계 5) 격납용기격리변 6) 비상용전원 등이며 주로 기동전 확인과 일상점검에 중점을 두었던 것이다.