• Title/Summary/Keyword: SMART reactor

Search Result 119, Processing Time 0.023 seconds

Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

  • Bae, Hwang;Kim, Dong Eok;Ryu, Sung-Uk;Yi, Sung-Jae;Park, Hyun-Sik
    • Nuclear Engineering and Technology
    • /
    • v.49 no.5
    • /
    • pp.968-978
    • /
    • 2017
  • Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal-hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

Thermal-hydraulic Design of A Printed-Circuit Steam Generator for Integral Reactor (일체형원자로 인쇄기판형 증기발생기 열수력학적 설계)

  • Kang, Han-Ok;Han, Hun Sik;Kim, Young-In
    • The KSFM Journal of Fluid Machinery
    • /
    • v.17 no.6
    • /
    • pp.77-83
    • /
    • 2014
  • The vessel of integral reactor contains its major primary components such as the fuel and core, pumps, steam generators, and a pressurizer, so its size is proportional to the required space for the installation of each component. The steam generators take up the largest volume of internal space of reactor vessel and their volumes is substantial for the overall size of reactor vessel. Reduction of installation space for steam generators can lead to much smaller reactor vessel with resultant decrease of overall cost for the components and related facilities. A printed circuit heat exchanger is one of the compact types of heat exchangers available as an alternative to conventional shell and tube heat exchangers. Its name is derived from the procedure used to manufacture the flat metal plates that form the core of the heat exchanger, which is done by chemical milling. These plates are then stacked and diffusion bonded, converting the plates into a solid metal block containing precisely engineered fluid flow passages. The overall heat transfer area and pressure drops are evaluated for the steam generator based on the concept of the printed circuit heat exchanger in this study. As the printed circuit heat exchanger is known to have much larger heat transfer area density per unit volume, we can expect significantly reduced steam generator compared to former shell and tube type of steam generator. For the introduction of new steam generator, two design requirements are considered: flow area ratio between primary and secondary flow paths, and secondary side parallel channel flow oscillation. The results show that the overall volume of the steam generator can be significantly reduced with printed circuit type of steam generator.

Boundary condition coupling methods and its application to BOP-integrated transient simulation of SMART

  • Jongin Yang;Hong Hyun Son;Yong Jae Lee;Doyoung Shin;Taejin Kim;Seong Soo Choi
    • Nuclear Engineering and Technology
    • /
    • v.55 no.6
    • /
    • pp.1974-1987
    • /
    • 2023
  • The load-following operation of small modular reactors (SMRs) requires accurate prediction of transient behaviors that can occur in the balance of plants (BOP) and the nuclear steam supply system (NSSS). However, 1-D thermal-hydraulics analysis codes developed for safety and performance analysis have conventionally excluded the BOP from the simulation by assuming ideal boundary conditions for the main steam and feed water (MS/FW) systems, i.e., an open loop. In this study, we introduced a lumped model of BOP fluid system and coupled it with NSSS without any ideal boundary conditions, i.e., in a closed loop. Various methods for coupling boundary conditions at MS/FW were tested to validate their combination in terms of minimizing numerical instability, which mainly arises from the coupled boundaries. The method exhibiting the best performance was selected and applied to a transient simulation of an integrated NSSS and BOP system of a SMART. For a transient event with core power change of 100-20-100%, the simulation exhibited numerical stability throughout the system without any significant perturbation of thermal-hydraulic parameters. Thus, the introduced boundary-condition coupling method and BOP fluid system model can expectedly be employed for the transient simulation and performance analysis of SMRs requiring daily load-following operations.

Dynamic Characteristics on the CRDM of SMART Reactor (SMART 원자로 제어봉 구동 장치의 동특성해석)

  • Lee, Jang-Won;Cho, Sang-Soon;Kim, Dong-Ok;Park, Jin-Seok;Lee, Won-Jae
    • Transactions of the Korean Society of Mechanical Engineers A
    • /
    • v.34 no.8
    • /
    • pp.1105-1111
    • /
    • 2010
  • The Korea Atomic Energy Research Institutes has been developing the SMART (System integrated Modular Advanced ReacTor), an environment-friendly nuclear reactor for the generation of electricity and to perform desalination. SMART reactors can be exposed to various external and internal loads caused by seismic and coolant flows. The CRDM(control rod drive mechanism), one of structures of the SMART, is a component which is adjusting inserting amount of a control rod, controlling output of reactor power and in an emergency situation, inserting a control rod to stop the reactor. The purpose of this research is performing the analysis of dynamic characteristic to ensure safety and integrity of structure of CRDM. This paper presents two FE-models, 3-D solid model and simplified Beam model of the CRDM in the coolant, and then compared the results of the dynamic characteristic about the two FE-models using a commercial Finite Element tool, ABAQUS CAE V6.8 and ANSYS V12. Beam 4 and beam 188 of simplified-model were also compared each other. And simplified model is updated for accuracy compare to 3-D solid.