• Nuclear Engineering and Technology
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• 제55권12호
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• pp.4431-4446
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• 2023

A Multiphysics coupling framework, MPCORE, has been developed to analyze safety parameters using the best estimate codes. The framework contains neutron kinetics (NK), thermal hydraulics (TH), and fuel performance (FP) codes to analyze fuel burnup, radial power distribution, and coolant temperature (T_bc). Shuffling and rotation capabilities have been verified on the Watts Bar reactor for three cycles. This study focuses on two coupling approaches for TH and FP modules. The one-way coupling approach involves coupling the FP code with the NK code, providing no data to the TH modules but getting T_bc as boundary condition from TH module. The two-way coupling approach exchanges information from FP to TH modules, so that the simplified heat conduction solver of the TH module is not used. The power profile in both approaches does not differ significantly, but there is an impact on coolant and cladding parameters. The one-way coupling approach tends to over-predict the cladding hydrogen concentration (CHC). This research highlights the difference between one-way and two-way coupling on critical boron concentration, T_bc, CHC, oxide surface temperature, and pellet centerline temperature. Overall, MPCORE framework with two-way coupling provides a more accurate and reliable analysis of safety parameters for nuclear reactors.

• Nuclear Engineering and Technology
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• 제53권6호
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• pp.1769-1785
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• 2021

The Integral Pressurized Water Reactors (IPWRs) as the innovative advanced and generation-III + reactors are under study and developments in a lot of countries. This paper is aimed at the thermal hydraulic study of the hot and average fuel sub-channel in a Generation III + IPWR by loose external coupling to the neutronic simulation. The power produced in fuel pins is calculated by the neutronic simulation via MCNPX2.6 then fuel and coolant temperature changes along fuel sub-channels evaluated by computational fluid dynamic thermal hydraulic calculation through an iterative coupling. The relative power densities along the fuel pin in hot and average fuel sub-channel are calculated in sixteen equal divisions. The highest centerline temperature of the hottest and the average fuel pin are calculated as 633 K (359.85 ℃) and 596 K (322.85 ℃), respectively. The coolant enters the sub-channel with a temperature of 557.15 K (284 ℃) and leaves the hot sub-channel and the average sub-channel with a temperature of 596 K (322.85 ℃) and 579 K (305.85 ℃), respectively. It is shown that the spacer grids result in the enhancement of turbulence kinetic energy, convection heat transfer coefficient along the fuel sub-channels so that there is an increase in heat transfer coefficient about 40%. The local fuel pin temperature reduction in the place and downstream the space grids due to heat transfer coefficient enhancement is depicted via a graph through six iterations of neutronic and thermal hydraulic coupling calculations. Working in a low fuel temperature and keeping a significant gap below the melting point of fuel, make the IPWR as a safe type of generation -III + nuclear reactor.

• 한국강구조학회 논문집
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• 제23권5호
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• pp.629-636
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• 2011

본 연구는 압전 센서를 이용한 구조물의 실시간 모니터링과 강구조물 볼트접합부의 동적 컨트롤이 가능한 자가치유 시스템에 관한 것이다. 볼트접합부의 볼트 풀림 손상 탐색을 위해서 압전 물질의 전기역학적 커플링 성질을 이용하는 임피던스 기반의 구조물건전성평가 기법을 이용하였다. 계측된 임피던스 값을 기준치 값과 비교함으로써 볼트 풀림 손상 진단이 가능하다. 볼트 풀림 손상은 손상지수를 이용하여 정량적으로 평가되어지고, 손상이 진단되면 형상기억합금 와셔에 감겨있는 외부히터가 와셔에 열을 가하게 된다. 열이 가해진 형상기억합금 와셔는 축방향으로 팽창하고, 볼트접합부는 볼트풀림으로 인해 잃어버린 토크력을 회복하게 된다. 압전 센서를 이용한 임피던스 기반의 구조물건전성 평가기법과 형상기억합금 기반의 볼트접합부 동적 컨트롤 기능과 이를 이용한 자가치유 볼트접합부 시스템의 적용가능성과 성능을 실험을 통해 평가하였다.