• Nuclear Engineering and Technology
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• 제54권11호
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• pp.4373-4391
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• 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.

• Nuclear Engineering and Technology
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• 제53권3호
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• pp.763-775
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• 2021

The thermal hydraulic performances of Steam Generator (SG) under both steady and transient operation conditions are of great importance for the safety and economy in nuclear power plants. In this paper, based on our self-developed SG thermal hydraulic analysis code STAF (Steam-generator Thermalhydraulic Analysis code based on Fluent), an improved new version STAF-CT (fully Coupling and Transient) is developed and introduced. Compared with original STAF, the new version code STAF-CT has two main functional improvements including "Transient" and "Fully Three Dimensional Coupling" features. In STAF-CT, a three dimensional energy transferring module is established which can achieve energy exchange computing function at the corresponding position between two sides of SG. The STAF-CT is validated against the international benchmark experiment data and the results show great agreement. Then the U-shaped SG in AP1000 nuclear power plant is modeled and simulated using STAF-CT. The results show that three dimensional flow fields in the primary side make significant effect on the energy source distribution between two sides. The development of code STAF-CT in this paper can provide an effective method for further SG high fidelity research in the nuclear reactor system.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.72-83
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• 2022

This paper presents a numerical investigation of two-phase natural circulation flows established when external reactor vessel cooling is applied to a severe accident of the APR1400 reactor for the in-vessel retention of the core melt. The coolability limit due to external reactor vessel cooling is associated with the natural circulation flow rate around the lower head of the reactor vessel. For an elaborate prediction of the natural circulation flow rate using a thermal-hydraulic system code, MARS-KS1.5, a three-dimensional computational fluid dynamics (CFD) simulation is conducted to estimate the flow rate and pressure distribution of a liquid-state coolant at the brink of significant void generation. The CFD calculation results are used to determine the loss coefficient at major flow junctions, where substantial pressure losses are expected, in the nodalization scheme of the MARS-KS code such that the single-phase flow rate is the same as that predicted via CFD simulations. Subsequently, the MARS-KS analysis is performed for the two-phase natural circulation regime, and the transient behavior of the main thermal-hydraulic variables is investigated.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.666-671
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• 1998

Computational fluid dynamics(CFD) code FLUENT[11 was used to simulate the thermal hydraulic processes occuring in conceptual design of the accelerator-driven subcritical reactor(ADSR) liquid lead target. The purpose of the analysis is to investigate the thermal hydraulic characteristics of liquid lead as ADSR target material with various target geometries and injection locations of proton beam. In the calculation analysis, the local temperature of the liquid lead target rises to the boiling temperature very rapidly When the proton beam is injected from the bottom of the target system, the duration time to reach the boiling temperature is longer and the temperature distribution is flatter than other cases.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.639-644
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• 1997

The work reported in this paper identifies the thermal-hydraulic phenomena that are expected to occur during a number of key transients in SMART(System-integrated Modular Advanced ReacTor) which is under development at KAERI. The result of this effort is based on the current design concept of SMART integral reactor. Although the design is still evolving, the preliminary phenomena Identification and Ranking Table(PIRT) has been developed based on the experts' knowledge and experience. The preliminary PIRT has been developed by consensus of KAERI expert panelists and AHP(Analytical Hierarchy Process). Preliminary PIRT developed in this paper is intended to be used to identify and integrate development areas of further experimental tests needed, thermal hydraulic models and correlations and code improvements for the safety analysis of the SMART.

• 유체기계공업학회:학술대회논문집
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• 유체기계공업학회 2004년도 유체기계 연구개발 발표회 논문집
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• pp.546-551
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• 2004

An applying Thermodynamic method for the purpose of measuring hydraulic efficiency of pump-motor system, based on IEC60041 code, is not easy to adopt at field test. Even though there were splendid development in measuring technic in discharge measuring through the hydraulic machina lots of unsolved problems concerned in flow-rate are still remain in measuring hydraulic efficiency in hydraulic machine. The key point in measuring hydraulic efficiency is to measure exact flow-rate. So, Thermodynamic methode provides a good solution. This methode measures hydraulic efficiency by detecting the difference of temperature and pressure between the hydraulic process of machine, without measuring flow-rate of pump or turbine. By measuring temperature in mk level and absolute pressure in pascal, we can get a difference of thermodynamic specific energy in Moliere chart before and after of hydraulic process, md that difference is equal to hydraulic loses. Following the standard in proceeding Thermodynamic methode, I hope these trial and records make others be familiar to the thermal methode and make it easer to beginner for trial.

• 에너지공학
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• 제13권1호
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• pp.51-59
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• 2004

국내에 설치 운영중인 원전 훈련용 시뮬레이터의 핵 증기공급 계통 열수력 프로그램은 1980련 전후에 외국 벤더들이 개발하여 공급한 것으로 이들 열수력 프로그램은 핵 증기공급 계통 열수력 현상을 실시간으로 모의하기 위해 과도하게 단순화된 모델을 채택하고 있다. 그 결과 원자로 냉각계통에 복잡한 이상유동이 발생하는 사고를 모의하는 경우 정확도가 떨어질 수 있어 부정적인 훈련(Negative training)을 초래할 가능성이 있다. 이와같은 문제를 해결하기 위해 전력연구원에서는 RETRAN-3D코드를 기본으로 시뮬레이터용 핵 증기공급 계통 열수력 프로그램 ARTS코드를 개발하였다. RETRAN-3D코드를 기본으로 하는 ARTS코드는 거의 대부분의 사고를 실시간으로 모의할 수 있으며 계산의 건전성도 보장된다. 그러나, 대형냉각재 상실사고나 저압 저유속 상태의 장기 과도현상 등을 모의하는 경우에 발생하는 계산실패나 실시간 계산 지체등의 가능성이 있다. 이 경우 이를 자동으로 대체 보완할 수 있는 보조계산체계를 개발했다. 특히, ARTS코드의 실시간 계산 및 건전성 문제가 예상되는 대형냉각재 상실사고를 주모의 대상으로 간주했다. 계산 결과는 코드의 정확도, 실시간 계산능력, 건전성 및 운전원 교육등에서 최종안정성평가보고서 및 ANSI/ANS-3.5-1998$^{[1]}$ 시뮬레이터 소프트웨어 기준을 만족하는 것으로 평가되었다

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집D
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• pp.915-920
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• 2001

A spent fuel problem has prevented the nuclear power from claiming to be a completely clean energy source. The nuclear transmutation technology to incinerate the long lived radioactive nuclides and produce energy during the incineration process is believed to be one or the best solutions. HYPER(Hybrid Power Extraction Reactor) is the accelerator driven transmutation system which is being developed by KAERI(Korea Atomic Energy Research Institute). Some major feature of HYPER have been developed and employed. On-power fueling concepts are employed to keep system power constant with minimum variation of accelerator power. A hollow cylinder-type metal fuel is designed for the on-line refueling concept. Lead-bismuth(Pb-Bi) is adopted as a coolant and Spallation target material. HYPER is a subcritical reactor which needs an external neutron source. 1GeV proton beam is irradiated to Lead-bismuth(Pb-Bi) target inside HYPER, and spallation neutrons are produced. When proton beams are irradiated, much heat is also deposited in the Pb-Bi target and beam window which separates Pb-Bi and accelerator vacuum. Therfore, an effective cooling is needed for HYPER target. In this paper, we performed the thermal-hydraulic analysis of HYPER target using FLUENT code, and also calculated thermal and mechanical stress of the beam window using ANSYS code.

• Nuclear Engineering and Technology
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• 제53권2호
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• pp.455-465
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• 2021

The Molten Salt Reactor (MSR), one of the six advanced reactor types of the 4th generation nuclear energy systems, has many impressive features including economic advantages, inherent safety and nuclear non-proliferation. This paper introduces a system analysis code named TREND, which is developed and used for the steady and transient simulation of MSRs. The TREND code calculates the distributions of pressure, velocity and temperature of single-phase flows by solving the conservation equations of mass, momentum and energy, along with a fluid state equation. Heat structures coupled with the fluid dynamics model is sufficient to meet the demands of modeling MSR system-level thermal-hydraulics. The core power is based on the point reactor neutron kinetics model calculated by the typical Runge-Kutta method. An incremental PID controller is inserted to adjust the operation behaviors. The verification and validation of the TREND code have been carried out in two aspects: detailed code-to-code comparison with established thermal-hydraulic system codes such as RELAP5, and validation with the experimental data from MSRE and the CIET facility (the University of California, Berkeley's Compact Integral Effects Test facility).The results indicate that TREND can be used in analyzing the transient behaviors of MSRs and will be improved by validating with more experimental results with the support of SINAP.

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1860-1873
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• 2022

Nuclear thermal-hydraulic system analysis codes have been developed to comprehensively model nuclear reactor systems to evaluate the safety of a nuclear reactor system. For analyzing complex systems with finite computational resources, system codes usually solve simplified fluid equations for coarsely discretized control volumes with one-dimensional assumptions and replace source terms in the governing equations with constitutive relations. Wall boiling heat transfer models are regarded as essential models in nuclear safety evaluation among many constitutive relations. The wall boiling heat transfer models of two widely used nuclear system codes, RELAP5 and TRACE, are analyzed in this study. It is first described how wall heat transfer models are composed in the two codes. By utilizing the same method described in Part 1 paper, heat fluxes from the two codes are compared under the same thermal-hydraulic conditions. The significant factors for the differences are identified as well as at which conditions the non-negligible difference occurs. Steady-state simulations with both codes are also conducted to confirm how the difference in wall heat transfer models impacts the simulation results.