• Nuclear Engineering and Technology
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• 제39권1호
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• pp.9-20
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• 2007

Design study on the Gas Turbine High Temperature Reactor 300-Cogeneration (GTHTR300C) aiming at producing both electricity by a gas turbine and hydrogen by a thermochemical water splitting method (IS process method) has been conducted. It is expected to be one of the most attractive systems to provide hydrogen for fuel cell vehicles after 2030. The GTHTR300C employs a block type Very High Temperature Reactor (VHTR) with thermal power of 600MW and outlet coolant temperature of $950^{\circ}C$. The intermediate heat exchanger (IHX) and the gas turbine are arranged in series in the primary circuit. The IHX transfers the heat of 170MW to the secondary system used for hydrogen production. The balance of the reactor thermal power is used for electricity generation. The GTHTR300C is designed based on the existing technologies of the High Temperature Engineering Test Reactor (HTTR) and helium turbine power conversion and on the technologies whose development have been well under way for IS hydrogen production process so as to minimize cost and risk of deployment. This paper describes the original design features focusing on the plant layout and plant cycle of the GTHTR300C together with present development status of the GTHTR300, IHX, etc. Also, the advantage of the GTHTR300C is presented.

• 방사성폐기물학회지
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• 제21권4호
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• pp.571-576
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• 2023

The initial development plans for the six reactor designs, soon after the release of Generation IV International Forum (GIF) TRM in 2002, were characterized by high ambition [1]. Specifically, the sodium-cooled fast reactor (SFR) and very-high temperature reactor (VHTR) gained significant attention and were expected to reach the validation stage by the 2020s, with commercial viability projected for the 2030s. However, these projections have been unrealized because of various factors. The development of reactor designs by the GIF was supposed to be influenced by events such as the 2008 global financial crisis, 2011 Fukushima accident [2, 3], discovery of extensive shale oil reserves in the United States, and overly ambitious technological targets. Consequently, the momentum for VHTR development reduced significantly. In this context, the aims of this study were to compare and analyze the development progress of the six Gen IV reactor designs over the past 20 years, based on the GIF roadmaps published in 2002 and 2014. The primary focus was to examine the prospects for the reactor designs in relation to spent nuclear fuel burning in conjunction with small modular reactor (SMR), including molten salt reactor (MSR), which is expected to have spent nuclear fuel management potential.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회B
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• pp.2850-2855
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• 2008

Very High Temperature Reactor (VHTR) has been selected as a high energy heat source for a nuclear hydrogen generation. The VHTR heat is transferred to a thermo-chemical hydrogen production process through an intermediate loop. Both Process Heat Exchanger and sulfuric acid evaporator provide the coupled components between the VHTR intermediate loop and hydrogen production module. A small scaled Compact Nuclear Hydrogen Coupled Components test loop is developed to simulate the VHTR intermediate loop and hydrogen production module. Main objective of the loop is to screening the candidates of NHDD (Nuclear Hydrogen Development and Demonstration) coupled components. The operating condition of the gas loop is a temperature up to $950^{\circ}C$ and a pressure up to 6.0MPa. The thermal and fluid dynamic design of the loop is dependent on the structures that enclose the gas flow, especially primary side that has fast gas velocity. We designed and constructed a small scale sulfuric acid experimental system which can simulate a part of the hydrogen production module also.

• Journal of Radiation Protection and Research
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• 제33권3호
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• pp.113-120
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• 2008

이 연구에서는 국내 개발중인 초고온가스로 (VHTR: Very High Temperature Reactor)를 대상으로, 발생되는 삼중수소 양, 계통간 이송, 제거, 분포 그리고 최종적으로 생산된 수소에 대한 삼중수소에 의한 오염 준위를 예측할 수 있는 해석 모델인 TRITGO 코드를 소개하였고, 수소를 생산하는 IS (Iodine Sulfide) 계통으로의 삼중수소 투과양을 모의할 수 있도록 코드를 개선하였다. 또한 GT-MHR 600MW 열출력을 가정, 최종 수소 생산물의 삼중수소에 의한 오염치를 예측하였다. 예상 오염치는 약 0.055 Bq/$H_2-g$으로 일본 규제치 56 Bq/$H_2-g$에 약 1/1000 수준으로 낮게 예측되었다. 모의 결과 삼중수소 방출을 억제하기 위해서는 피복관의 건전성 유지 및 헬륨 냉각재와 흑연으로 구성된 반사체내 불순물인 $^3He$ 및 Li을 가능한한 낮은 준위로 유지하는 것이 필요함을 보여 주었다. 또한 냉각재내 불순물을 직접 제거할 수 있는 정화계통의 성능이 중요한 설계인자로 판단되었다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 춘계학술발표회
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• pp.443-444
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• 2005

• Corrosion Science and Technology
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• 제13권5호
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• pp.178-185
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• 2014

Alloy 617 is considered as a candidate Ni-based superalloy for the intermediate heat exchanger (IHX) of a very high-temperature gas reactor (VHTR) because of its good creep strength and corrosion resistance at high temperatures. Helium is used as a coolant in a VHTR owing to its high thermal conductivity, inertness, and low neutron absorption. However, helium inevitably includes impurities that create an imbalance in the surface reactivity at the interface of the coolant and the exposed materials. As the Alloy 617 has been exposed to high temperatures at $950^{\circ}C$ in the impure helium environment of a VHTR, the degradation of material is accelerated and mechanical properties decreased. The high-temperature strength, creep, and corrosion properties of the structural material for an IHX are highly important to maintain the integrity in a harsh environment for a 60 year period. Therefore, an alloy superior to alloy 617 should be developed. In this study, the mechanical and high-temperature corrosion properties for Ni-Cr alloys fabricated in the laboratory were evaluated as a function of the grain boundary strengthening and alloying elements. The ductility increased and decreased by increasing the amount of Mo and Cr, respectively. Surface oxide was detached during the corrosion test, when Al was not added to alloy. However the alloy with Al showed improved oxide adhesive property without significant degradation and mechanical property. Aluminum seems to act as an anti-corrosive role in the Ni-based alloy.

• 대한기계학회논문집A
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• 제35권10호
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• pp.1249-1255
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• 2011

공정열교환기는 초고온가스로로부터 생성된 $950^{\circ}C$ 정도의 초고온 열을 대량의 수소를 생산하기 위한 화학반응공정으로 전달하는데 필요한 핵심기기이다. Hastelloy-X 로 제작된 소형 공정열교환기 시제품이 한국원자력연구원에 있는 소형가스루프에서 성능시험이 계획되어 있다. 본 연구에서는 소형가스루프 시험조건하에서 소형 공정열교환기 시제품의 고온 구조건전성을 사전에 평가하기 위한 작업의 일환으로 소형 공정열교환기 시제품에 대한 고온 구조해석 모델링, 거시적 열 해석 및 탄 소성구조 해석을 수행하고 그 결과들을 정리한 것이다. 해석 결과는 공정열교환기 수정 시제품 성능시험장치 설계에 반영할 것이다.

• 대한기계학회논문집B
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• 제34권12호
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• pp.1093-1099
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• 2010

초고온원자로(VHTR) 설계에 있어 중간열수송루프(IHTL) 및 중간열교환기(IHX) 설계는 고온의 운전조건($950^{\circ}C$ 이상)으로 인하여 공학적으로 어려운 과제 중 하나로 알려져있다. 본 연구에서는 LiF, NaF 및 KF(46.5:11.5:42.0 mole %)의 공융혼합물인 Flinak molten salt 를 IHTL 의 열수송매체로 고려하였다. Flinak molten salt 의 세관에서의 열수력 특성을 평가하기 위하여 직경이 1.4 mm 인 원형관을 이용하여 고온의 가스와 Flinak 을 열교환할 수 있는 이중관식 열교환기를 구성하여 실험하였다. 실험 결과 층류유동에서 측정된 Flinak 의 마찰계수는 이론식인 64/Re 에 근접하였고 Nusselt 수는 일반적으로 3.66 에서 4.36 범위에 들었다.

• 대한기계학회논문집A
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• 제36권10호
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• pp.1283-1288
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• 2012

수소를 대량으로 생산하기 위한 원자력수소생산시스템에서 공정열교환기는 초고온가스로로부터 생성된 초고온 열을 화학반응공정으로 전달하는 핵심기기이다. 한국원자력연구원에 구축되어 있는 소형 가스루프에서 $Hastelloy^{(R)}$-X 로 제작된 중형 공정열교환기 시제품에 대한 성능시험이 계획되어 있다. 본 연구에서는 중형 공정열교환기의 고온구조건전성을 파악하기 위한 선행 연구로서 소형가스루프 시험조건하에서 중형 공정열교환기 시제품의 고온구조해석을 이전 연구에서 확립된 경계조건을 적용하여 수행하였다. 해석결과는 소형가스루프에서의 중형 공정열교환기 시제품에 대한 성능시험 결과와 비교할 예정이다.

• 한국압력기기공학회 논문집
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• 제10권1호
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• pp.64-69
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• 2014

Gr. 91 steel is used for the major structural components of Generation-IV reactor systems, such as a very high temperature reactor(VHTR) and sodium-cooled fast reactor(SFR). Since these structures are designed for up to 60 years at elevated temperatures, the prediction of long-term creep life is important for a design application of Gr. 91 steel. In this study, a number of creep rupture data were collected through world-wide literature surveys, and using these data, the long-term creep life was predicted in terms of three methods: the single-C method in Larson-Miller(L-M) parameter, multi-C constant method in the L-M parameter, and a modified method("sinh" equation) in the L-M parameter. The results of the creep-life prediction were compared using the standard deviation of error value, respectively. Modified method proposed by the "sinh" equation revealed better agreement in creep life prediction than the single-C L-M method.