• International Journal of Fluid Machinery and Systems
• /
• v.10 no.2
• /
• pp.138-145
• /
• 2017

The problem of non-uniform inflow exists in many practical engineering applications, such as the elbow suction pipe of waterjet pump and, the channel head of steam generator which is directly connect with reactor coolant pump. Generally, pumps are identical designs and are selected based on performance under uniform inflow with the straight pipe, but actually non-uniform suction flow is induced by upstream equipment. In this paper, CFD approach was employed to analyze unsteady hydrodynamic characteristics of reactor coolant pumps with different inflows. The Reynolds-averaged Naiver-Stokes equations with the $k-{\varepsilon}$ turbulence model were solved by the computational fluid dynamics software CFX to conduct the steady and unsteady numerical simulation. The numerical results of the straight pipe and channel head were validated with experimental data for the heads at different flow coefficients. In the nominal flow rate, the head of the pump with the channel head decreases by 1.19% when compared to the straight pipe. The complicated structure of channel head induces the inlet flow non-uniform. The non-uniformity of the inflow induces the difference of vorticity distribution at the outlet of the pump. The variation law of blade to blade velocity at different flow rate and the difference of blade to blade velocity with different inflow are researched. The effects of non-uniform inflow on radial forces are absolutely different from the uniform inflow. For the radial forces at the frequency $f_R$, the corresponding amplitude of channel head are higher than the straight pipe at $1.0{\Phi}_d$ and $1.2{\Phi}_d$ flow rates, and the corresponding amplitude of channel head are lower than the straight pipe at $0.8{\Phi}_d$ flow rates.

• Journal of Astronomy and Space Sciences
• /
• v.32 no.1
• /
• pp.81-89
• /
• 2015

In this paper, we describe the development of a bioreactor for a cell-culture experiment on the International Space Station (ISS). The bioreactor is an experimental device for culturing mouse muscle cells in a microgravity environment. The purpose of the experiment was to assess the impact of microgravity on the muscles to address the possibility of long-term human residence in space. After investigation of previously developed bioreactors, and analysis of the requirements for microgravity cell culture experiments, a bioreactor design is herein proposed that is able to automatically culture 32 samples simultaneously. This reactor design is capable of automatic control of temperature, humidity, and culture-medium injection rate; and satisfies the interface requirements of the ISS. Since bioreactors are vulnerable to cell contamination, the medium-circulation modules were designed to be a completely replaceable, in order to reuse the bioreactor after each experiment. The bioreactor control system is designed to circulate culture media to 32 culture chambers at a maximum speed of 1 ml/min, to maintain the temperature of the reactor at $36{\pm}1^{\circ}C$, and to keep the relative humidity of the reactor above 70%. Because bubbles in the culture media negatively affect cell culture, a de-bubbler unit was provided to eliminate such bubbles. A working model of the reactor was built according to the new design, to verify its performance, and was used to perform a cell culture experiment that confirmed the feasibility of this device.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1999.05a
• /
• pp.209-209
• /
• 1999

• Nuclear Engineering and Technology
• /
• v.41 no.8
• /
• pp.995-1004
• /
• 2009

The Advanced Power Reactor 1400 (APR1400) is an evolutionary advanced light water reactor (ALWR) based on the Optimized Power Reactor 1000 (OPR1000), which is in operation in Korea. The APR1400 incorporates a variety of engineering improvements and operational experience to enhance safety, economics, and reliability. The advanced design features and improvements of the APR1400 design include a pilot operated safety relief valve (POSRV), a four-train safety injection system with direct vessel injection (DVI), a fluidic device (FD) in the safety injection tank, an in-containment refueling water storage tank (IRWST), an external reactor vessel cooling system, and an integrated head assembly (IHA). Development of the APR1400 started in 1992 and continued for ten years. The APR1400 design received design certification from the Korean nuclear regulatory body in May of2002. Currently, two construction projects for the APR1400 are in progress in Korea.

• IE interfaces
• /
• v.23 no.1
• /
• pp.68-77
• /
• 2010

Digital reactor protection system which consists of many identical modules, is fault- tolerant to provide high safety. The modules themselves including DSP(digital signal processing) card are also fault-tolerant in nature. This paper assesses the safety for being-designed digital reactor protection system of 2-out-of-4 G structure with lockout. Some interesting design alternatives are compared. Fault tree analysis for assessing system safety is performed by Relex software. The selected reactor protection system fully satisfies EPRIURD stipulation of mean failure time of 50 years.

• Nuclear Engineering and Technology
• /
• v.54 no.11
• /
• pp.4373-4391
• /
• 2022

Marine reactor systems experience platform movement, and therefore, the system thermal-hydraulic analysis code needs to reflect the motion effect on the fluid to evaluate reactor safety. A moving reactor model for MARS-KS was developed to simulate the hydrodynamic phenomena in the reactor under motion conditions; however, its applicability does not cover the MULTID component used in multidimensional flow analyses. In this study, a moving reactor model is implemented for the MULTID component to address the importance of multidimensional flow effects under dynamic motion. The concept of the volume connection is generalized to facilitate the handling of the junction of MULTID. Further, the accuracy in calculating the pressure head between volumes is enhanced to precisely evaluate the additional body force. Finally, the Coriolis force is modeled in the momentum equations in an acceleration form. The improvements are verified with conceptual problems; the modified model shows good agreement with the analytical solutions and the computational fluid dynamic (CFD) simulation results. Moreover, a simplified gravity-driven injection is simulated, and the model is validated against a ship flooding experiment. Throughout the verifications and validations, the model showed that the modification was well implemented to determine the capability of multidimensional flow analysis under ocean conditions.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1997.10a
• /
• pp.351-356
• /
• 1997

In the des19n of KNGR (Korea Next Generation Reactor), we believe that economy is one of the most important factors to be considered Thus, we reviewed and evaluated the consequences of designing the head area into an integrated package from an economical point of view. The refueling outage durations of the nuclear power plants currently in operation In Korea, some having and others not having integrated head package, are compared. This paper discusses the characteristics of head area design and the critical design issues of KNGR head area to evaluate the effect of the head area characteristics on the outage duration.

• Nuclear Engineering and Technology
• /
• v.54 no.12
• /
• pp.4522-4533
• /
• 2022

For severe accident assessment of reactor pressure vessel (RPV), it is important to develop an accurate model that can predict transient thermo-mechanical behavior of the RPV lower head under the given condition. The present study revisits the lower head failure with two- and three-dimensional finite element models. In particular, we aim to give clear insight regarding the effect of the three-dimensionality present in the distribution of the thickness and thermal load of the lower head. For a rigorous validation of the result, both the OLHF-1 and the OLHF-2 tests are considered in this study. The result suggests that the three-dimensional effect is not negligible as far as the failure location is concerned. The non-uniformity of the thickness distribution is found to affect the failure location and time. The thermal load, which may not be axisymmetric in general, has the most significant effect on the failure assessment. We also observe that the creep property can affect the global deformation of the lower head, depending on the applied mechanical load.

• Nuclear Engineering and Technology
• /
• v.39 no.3
• /
• pp.193-206
• /
• 2007

The Korea Atomic Energy Research Institute has developed an advanced fast reactor concept, KALIMER-600, which satisfies the Generation IV reactor design goals of sustainability, economics, safety, and proliferation resistance. The concept enables an efficient utilization of uranium resources and a reduction of the radioactive waste. The core design has been developed with a strong emphasis on proliferation resistance by adopting a single enrichment fuel without blanket assemblies. In addition, a passive residual heat removal system, shortened intermediate heat-transport system piping and seismic isolation have been realized in the reactor system design as enhancements to its safety and economics. The inherent safety characteristics of the KALIMER-600 design have been confirmed by a safety analysis of its bounding events. Research on important thermal-hydraulic phenomena and sensing technologies were performed to support the design study. The integrity of the reactor head against creep fatigue was confirmed using a CFD method, and a model for density-wave instability in a helical-coiled steam generator was developed. Gas entrainment on an agitating pool surface was investigated and an experimental correlation on a critical entrainment condition was obtained. An experimental study on sodium-water reactions was also performed to validate the developed SELPSTA code, which predicts the data accurately. An acoustic leak detection method utilizing a neural network and signal processing units were developed and applied successfully for the detection of a signal up to a noise level of -20 dB. Waveguide sensor visualization technology is being developed to inspect the reactor internals and fuel subassemblies. These research and developmental efforts contribute significantly to enhance the safety, economics, and efficiency of the KALIMER-600 design concept.

• Nuclear Engineering and Technology
• /
• v.32 no.4
• /
• pp.379-394
• /
• 2000

As an in-vessel retention (IVR) design concept in coping with a severe accident in the nuclear power plant during which time a considerable amount of core material may melt, external cooling of the reactor vessel has been suggested to protect the lower head from overheating due to relocated material from the core. The efficiency of the ex-vessel management may be estimated by the thermal margin defined as the ratio of the critical heat flux (CHF）to the actual heat flux from the reactor vessel. Principal factors affecting the thermal margin calculation are the amount of heat to be transferred downward from the molten pool, variation of heat flux with the angular position, and the amount of removable heat by external cooling In this paper a thorough literature survey is made and relevant models and correlations are critically reviewed and applied in terms of their capabilities and uncertainties in estimating the thermal margin to potential failure of the vessel on account of the CHF Results of the thermal margin calculation are statistically treated and the associated uncertainties are quantitatively evaluated to shed light on the issues requiring further attention and study in the near term. Our results indicated a higher thermal margin at the bottom than at the top of the vessel accounting for the natural convection within the hemispherical molten debris pool in the lower plenum. The information obtained from this study will serve as the backbone in identifying the maximum heat removal capability and limitations of the IVR technology called the Cerium Attack Syndrome Immunization Structures (COASISO) being developed for next generation reactors.