• Nuclear Engineering and Technology
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• 제47권3호
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• pp.227-239
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• 2015

The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, doublefault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperature profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain coolable. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel-coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.

• 한국지진공학회논문집
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• 제11권3호
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• pp.73-86
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• 2007

본 논문에서는 제4세대 소듐냉각고속로(Sodium-Cooled Fast Reactor)의 후보 노형으로 선정된 KALIMER-600에 대한 단순 지진해석모델을 개발하고 시간이력 지진응답해석을 수행하여 수평 면진설계(Seismic Isolation) 기술이 적용된 원자로건물의 주요기기 및 구조물에서의 지진응답 성능을 분석하였다. 개발된 단순 지진해석모델은 원자로건물, 원자로시스템, 주요 기기, 중간 열전달계통 배관, 그리고 면진장치를 포함하며 각각은 상세 유한요소해석을 통한 동특성 비교검증을 통하여 정확성을 검증하였다. 안전정지기준 0.3g의 설계인공지진 하중에 대한 시간이력 지진응답해석을 수행하여 면진설계와 비면진 설계조건에 따른 원자로 주요 부위에서의 층응답스펙트럼을 비교분석한 결과 KALIMER-600의 면진성능이 우수한 것으로 나타났다.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2395-2407
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• 2022

The sodium-cooled fast reactor is among the innovative nuclear technologies selected in the framework of the development of Generation IV concepts, allowing the irradiation of uranium-plutonium mixed oxide fuels (MOX). A fundamental step for the safety assessment of MOX-fuelled pins for fast reactor applications is the evaluation, by means of fuel performance codes, of the integral thermal-mechanical behaviour under irradiation, involving the fission gas behaviour and release in the fuel-cladding gap. This work is dedicated to the performance analysis of an inner-core fuel pin representative of the ASTRID sodium-cooled concept design, selected as case study for the benchmark between the GERMINAL and TRANSURANUS fuel performance codes. The focus is on fission gas-related mechanisms and integral outcomes as predicted by means of the SCIANTIX module (allowing the physics-based treatment of inert gas behaviour and release) coupled to both fuel performance codes. The benchmark activity involves the application of both GERMINAL and TRANSURANUS in their "pre-INSPYRE" versions, i.e., adopting the state-of-the-art recommended correlations available in the codes, compared with the "post-INSPYRE" code results, obtained by implementing novel models for MOX fuel properties and phenomena (SCIANTIX included) developed in the framework of the INSPYRE H2020 Project. The SCIANTIX modelling includes the consideration of burst releases of the fission gas stored at the grain boundaries occurring during power transients of shutdown and start-up, whose effect on a fast reactor fuel concept is analysed. A clear need to further extend and validate the SCIANTIX module for application to fast reactor MOX emerges from this work; nevertheless, the GERMINAL-TRANSURANUS benchmark on the ASTRID case study highlights the achieved code capabilities for fast reactor conditions and paves the way towards the proper application of fuel performance codes to safety evaluations on Generation IV reactor concepts.

• 한국방사선학회논문지
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• 제10권2호
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• pp.117-124
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• 2016

본 연구에서는 디지털 의료용 X선을 이용하여 소듐냉각고속로 금속연료봉 내부의 소듐 충전상태를 평가하고자 하였다. 의료용 X선은 디지털방사선(digital radiography, DR)시스템을 사용하였다. 최상의 공간분해능을 도출하기 위해 X선관의 초점크기와 디지털영상후처리 방법을 변화시켰다. 이때 텅스텐을 이용한 에지방법의 변조전달함수(modulation transfer function, MTF)를 사용하여 선예도(0.5 MTF)와 해상도(0.1 MTF) 를 평가하였다. 그 결과 소초점에서 영상후처리의 대조도를 강화시킨 조건에서 3.871 lp/mm (0.1 MTF)와 3.290 lp/mm (0.5 MTF)의 공간분해능을 얻었다. 결론적으로 이러한 결과는 디지털 의료용 X선을 이용하여 소듐냉각고속로 금속연료봉 내부의 소듐 충전상태를 관찰할 수 있었다.

• 기계저널
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• 제51권12호
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• pp.51-54
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• 2011

이 글에서는 제4세대 원자로로 다시 부각되고 있는 소듐냉각고속로(SFR: Sodium-cooled Fast Reactor)의 활용성, 계통설계 구성 및 공학적 안전설비에 대하여 가압경수로(PWR: Pressurized Water Reactor)와의 차이점을 위주로 소개한다.

• Korean Chemical Engineering Research
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• 제55권2호
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• pp.170-179
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• 2017

본 논문은 소듐냉각고속로 원형로 소듐-물 반응 압력완화계통의 급수배출부와 수소방출부의 설계요건 도출을 목적으로 한다. 증기발생기 전열관 누설에 의한 소듐-물 반응 발생 시, 증기발생기 내의 급수 증기를 신속하게 배출하는 조건을 도출하기 위해 급수덤프탱크 가스방출배관의 단면적과 증기발생기 급수배출배관의 수직길이를 변화시켜 연구를 수행하였다. 정상운전과 재장전운전에 대해 각각 계산을 수행하여 급수덤프탱크 가스방출배관의 단면적과 증기발생기 급수배출배관의 수직길이를 결정하였다. 정상운전 조건에서 소듐-물 반응 발생 시, 생성물인 수소에 의해 형성되는 과압이 소듐덤프탱크의 설계압력을 만족시킬 수 있도록 하는 가스방출배관의 직경을 도출하였고, 이 때 대기로 방출되는 수소의 유량과 농도를 계산하였다. 본 논문의 계산결과는 향후 소듐냉각고속로 원형로의 소듐-물 반응 압력완화계통의 설계요건으로 활용될 예정이다.

• Nuclear Engineering and Technology
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• 제52권7호
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• pp.1367-1379
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• 2020

The small modular sodium-cooled fast reactor (SMSFR) is an important component of Generation-IV reactors. The objective of this work is to improve the reactivity control in SMSFR by using innovative systems, including burnable poisons and optimized control rods. SMSFR with MOX fuel usually exhibits high burnup reactivity loss that leads to high excess reactivity and potential fuel melting in control rod withdrawal (CRW) accidents, which becomes an important constraint on the safety and economic efficiency of SMSFR. This work applies two types of burnable poisons in a SMSFR to reduce the excess reactivity. The first one homogenously loads minor actinides in the fuel. The second one combines absorber and moderators in specific assemblies. The influence of burnable poisons on the core characteristics is discussed and integrated into the analysis of CRW accidents. The results show that burnable poisons improve the safety performance of the core in a significant way. Burnable poisons also lessen the demand for the number, absorption ability, and insertion depth of control rods. Two optimized control rod designs with rare earth oxides (Eu₂O₃ and Gd₂O₃) and moderators are compared to the conventional design with natural boron carbide (B₄C). The optimized designs show improved neutronic and safety performance.

• Nuclear Engineering and Technology
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• 제47권1호
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• pp.47-58
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• 2015

Four recycling scenarios involving pyroprocessing of spent fuel (SF) have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR), KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU) SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro-SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. Thefsensitivity of cooling time before prior to pyro-processing was studied. As the cooling time sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC) decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs). If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE) isotopes. The RE isotope recovery factor should be lowered to ${\leq}20%$ in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

• Nuclear Engineering and Technology
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• 제39권3호
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• pp.193-206
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• 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
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• 제39권1호
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• pp.85-94
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• 2007

The Korean nuclear fuel cycle was modeled by the dynamic analysis method, which was applied to the once-through and alternative fuel cycles. First, the once-through fuel cycle was analyzed based on the Korean nuclear power plant construction plan up to 2015 and a postulated nuclear demand growth rate of zero after 2015. Second, alternative fuel cycles including the direct use of spent pressurized water reactor fuel in Canada deuterium uranium reactors (DUPIC), a sodium-cooled fast reactor and an accelerator driven system were assessed and the results were compared with those of the once-through fuel cycle. The once-through fuel cycle calculation showed that the nuclear power demand would be 25 GWe and the amount of the spent fuel will be ${\sim}65000$ tons by 2100. The alternative fuel cycle analyses showed that the spent fuel inventory could be reduced by more than 30% and 90% through the DUPIC and fast reactor fuel cycles, respectively, when compared with the once-through fuel cycle. The results of this study indicate that both spent fuel and uranium resources can be effectively managed if alternative reactor systems are timely implemented along with the existing reactors.