• 방사성폐기물학회지
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• 제8권1호
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• pp.77-84
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• 2010

상용원자로에서 발생하는 산화물 사용후핵연료의 부피감용과 재활용을 위하여 산화물을 금속으로 환원시키는 공정에 대한 연구가 수행되어 왔다. 다양한 환원법 중에서, 한국원자력연구원은 LiCl-$Li_2O$ 용융염을 반응매질로 사용하는 전해환원공정을 현재 개발 중이다. 파이로 공정의 전단부에 해당하는 전해환원 공정은 PWR 산화물 연료 주기를 소듐냉각 고속로의 금속연료 주기에 연결시켜 준다. 이 논문은 금속전환 공정을 개발/개선하고, 용량 증대를 수행한 한국원자력연구원의 노력을 요약한다.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2635-2649
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• 2022

CEFR is a small core-size sodium-cooled fast reactor (SFR) using high enrichment fuel with stainless-steel reflectors, which brings a significant challenge to the deterministic methodologies due to the strong spectral effect. The neutronic simulation of the start-up experiments conducted at the CEFR have been performed with a deterministic code system RAST-F, which is based on the two-step approach that couples a multi-group cross-section generation Monte-Carlo (MC) code and a multi-group nodal diffusion solver. The RAST-F results were compared against the measurement data. Moreover, the characteristic of neutron spectrum in the fuel rings, and adjacent reflectors was evaluated using different models for generation of accurate nuclear libraries. The numerical solution of RAST-F system was verified against the full core MC solution MCS at all control rods fully inserted and withdrawn states. A good agreement between RAST-F and MCS solutions was observed with less than 120 pcm discrepancies and 1.2% root-mean-square error in terms of k_eff and power distribution, respectively. Meanwhile, the RAST-F result agreed well with the experimental values within two-sigma of experimental uncertainty. The good agreement of these results indicating that RAST-F can be used to neutronic steady-state simulations for small core-size SFR, which was challenged to deterministic code system.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2009년도 학술논문요약집
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• pp.51-52
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• 2009

Recently, several industrial countries using the fission energy have given attention to the Gen-IV SFR (sodium-cooled fast reactor) for achieving sustainable nuclear energy systems. In this context, an SFR is currently developed at the design concepts study stage in the Republic of Korea [Kim & Hahn 200909]. The sustainability of systems means economic, environment-friendly, proliferation-resistant, and safer systems. More specifically, this sustainability can be accomplished in terms of resource recycling and radioactive waste reduction. In the present work, the objective of fuel cycle cost modeling is to identify the impact of various conceptual options as a cost reduction measure for the Gen-IV SFR at the design concepts study stage. It facilitates the selection of several reasonable fuel cycle pathways for the future Gen-IV SFR from an economic viewpoint.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3207-3216
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• 2021

Three-dimensional structures of a vortical flow field and heat transfer characteristics in a partially blocked 7-pin fuel assembly mock-up of sodium-cooled fast reactor have been investigated through a numerical analysis using a commercial computational fluid dynamics code, ANSYS CFX. The simulation with the SST turbulence model agrees well with the experimental data of outlet and cladding wall temperatures. From the analysis on the limiting streamline at the wall, multi-scale vortexes developed in axial direction were found around the blockage. The vortex core has a high cladding wall temperature, and the attachment line has a low cladding wall temperature. The small-scale vortex structures significantly enhance the convective heat transfer because it increases the turbulent mixing and the turbulence kinetic energy. The large-scale vortex structures supply thermal energy near the heated cladding wall surface. It is expected that control of the vortex structures in the fuel assembly plays a significant role in the convective heat transfer enhancement. Furthermore, the blockage plate and grid spacer increase the pressure drop to about 36% compared to the bare case.

• Nuclear Engineering and Technology
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• 제53권3호
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• pp.932-941
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• 2021

This paper suggests a fracture simulation method for SFR metallic fuel pin under accident condition. Two major failure mechanisms - creep damage and eutectic penetration - are implemented in the suggested method. To simulate damaged element, stress-reduction concept to reduce stiffness of the damaged element is applied. Using the proposed method, the failure size of cladding can be predicted in addition to the failure time and failure site. To verify the suggested method, Whole-pin furnace (WPF) test and TREAT-M test conducted at Argonne National Laboratory (ANL) are simulated. In all cases, predicted results and experimental results are overall in good agreement. Based on the simulation result, the effect of eutectic-penetration depth representing failure behavior on failure size is studied.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.772-784
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• 2024

The hexagonal nodal code RENUS has been enhanced to handle irregularly deformed hexagonal assemblies. The underlying RENUS methods involving triangle-based polynomial expansion nodal (T-PEN) and corner point balance (CPB) were extended in a way to use line and surface integrals of polynomials in a deformed hexagonal geometry. The nodal calculation is accelerated by the coarse mesh finite difference (CMFD) formulation extended to unstructured geometry. The accuracy of the unstructured nodal solution was evaluated for a group of 2D SFR core problems in which the assembly corner points are arbitrarily displaced. The RENUS results for the change in nuclear characteristics resulting from fuel deformation were compared with those of the reference McCARD Monte Carlo code. It turned out that the two solutions agree within 18 pcm in reactivity change and 0.46% in assembly power distribution change. These results demonstrate that the proposed unstructured nodal method can accurately model heterogeneous thermal expansion in hexagonal fueled cores.

• 대한기계학회논문집A
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• 제35권9호
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• pp.1083-1090
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• 2011

소듐냉각 고속로를 개발함에 있어 최대 현안 중 하나가 증기발생기에서의 소듐-물 반응사고 가능성이다. 이를 개선하기 위해 지금까지 수십 종 이상 연구개발 되었지만 국가마다 그 사양이 다르고, 동일한 기종이 후속기에 다시 활용되지 못할 정도로 기술이 안정화 상태에 도달하지 못하였다. 최근 개발되고 있는 증기발생기의 공통적 목표는 소듐-물 반응사고의 조기감지 및 제어, 증기발생기의 검사 및 보수가 쉽게 용접개수를 줄이고 경제성을 높인 Benson 증기사이클을 적용하는 것이다. 이 논문에서는 지금까지 설계 또는 활용한 증기발생기들의 사양과 문제점을 비교, 분석하였고, 이를 토대로 현안 극복방안을 제시하였다.

• 대한기계학회논문집A
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• 제38권12호
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• pp.1421-1426
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• 2014

소듐냉각고속로(Sodium-cooled Fast Reactor, SFR)의 안전성 향상을 위해 고온 증기 Rankine 싸이클 대신 초임계 이산화탄소(Supercritical $CO_2$, $S-CO_2$) Brayton 싸이클을 전력변환 시스템에 사용하는 방안이 제시되고 있다. 이 경우, 중간 열교환기로는 확산 접합(Diffusion Bonding)에 의해 제작되는 미소채널형 열교환기인 PCHE(Printed Circuit Heat Exchanger)가 고려되고 있다. 따라서 본 연구에서는 PCHE 형 열교환기 후보재료인 다양한 오스테나이트계 합금의 확산접합 특성을 평가하였다. 후보재료별로 다양한 조건에서 확산접합부를 제작하고 상온에서 $650^{\circ}C$까지의 인장 특성을 평가하였다. 평가 결과 SS 316H와 SS 347H는 $550^{\circ}C$까지 모재와 유사한 특성을 보였지만 Fe-Ni-Cr 합금인 Incoloy 800HT는 모든 온도에서 인장특성이 감소하였다. 연신율 저하의 원인을 이해하기 위해 접합부 부근의 미세조직을 분석하였다.

• 대한기계학회논문집A
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• 제38권12호
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• pp.1415-1420
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• 2014

소듐냉각고속로(Sodium-cooled Fast Reactor, SFR)의 증기 Rankine 싸이클 발전시스템을 높은 열효율과 안전성을 가지는 초임계 이산화탄소(Supercritical carbon dioxide, $S-CO_2$) Brayton 싸이클로 대체하는 방안이 고려되고 있다. 다양한 오스테나이트계 합금이 고온 $S-CO_2$ 환경 열교환시스템 구조재료로 제시되고 있다. 구조재료는 장시간 고온 $S-CO_2$ 환경에 노출됨에 따라 미세구조에 변화가 일어나고, 나아가 기계적 특성의 저하가 발생할 수 있다. 본 연구에서는 오스테니틱 스테인리스강들과 Alloy 800HT를 고온 $S-CO_2$ 환경에 노출시키고, 그에 따른 부식특성 및 인장특성을 평가하였다. 그 결과 $650^{\circ}C$, 250시간 노출 후 316H SS와 800HT에서 큰 연신율 감소를 보였다. $S-CO_2$ 환경이 인장특성 변화에 미치는 영향을 표면 산화막 및 탄화거동을 통해 분석한 결과, 316H 와 800H 의 거동이 큰 차이를 보였다.

• Nuclear Engineering and Technology
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• 제52권8호
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• pp.1611-1625
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• 2020

We propose a scaled-down experimental model of vertical air-natural convection channels by applying the modified Ishii-Kataoka scaling method with the assistance of numerical analyses to the Reactor Vault Cooling System (RVCS) of the Proto-type Gen-IV Sodium-cooled fast reactor (PGSFR) being developed in Korea. Two major non-dimensional numbers (modified Richardson and Friction number) from the momentum equation and Stanton number from the energy balance equation were identified to design the scaled-down experimental model to assimilate thermal-hydraulic behaviors of the natural convective air-cooling channel of RVCS. The ratios of the design parameters in the PGSFR RVCS between the prototype and the scaled-down model were determined by setting Richardson and Stanton number to be unity. The friction number which cannot be determined by the Ishii-Kataoka method was estimated by numerical analyses using the MARS-KS system code. The numerical analyses showed that the friction number with the form loss coefficient of 2.0 in the scale-down model would result in an acceptable prediction of the thermal-hydraulic behavior in RVCS. We also performed experimental benchmarking using the scaled-down model with the MARS-KS simulations to verify the appropriateness of the scale-down model, which demonstrated that the temperature rises and the average air flow velocity measured in the scale-down model.