• Nuclear Engineering and Technology
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• 제52권6호
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• pp.1340-1346
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• 2020

The aim of this paper is to explore the 8th-order Adams-Bashforth-Moulton (ABM8) method in the solution of the point reactor kinetics equations. The numerical experiment considers feedback reactivity by Doppler effects, and insertions of reactivity. The Doppler effects is approximated with an adiabatic nuclear reactor that is a typical approximation. The numerical results were compared and discussed with several solution methods. The CATS method was used as a benchmark method. According with the numerical experiments results, the ABM8 method can be considered as one of the main solution method for changes reactivity relatively large.

• Nuclear Engineering and Technology
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• 제49권4호
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• pp.700-709
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• 2017

This work looks at the effect of changes in kinetic parameters on simultaneous reactivity insertions and beam tube flooding in a typical material testing reactor-type research reactor with low enriched high density ($U_3Si_2-Al$) fuel. Using a modified PARET code, various ramp reactivity insertions (from $0.1/0.5 s to $1.3/0.5 s) plus beam tube flooding ($0.5/0.25 s) accidents under uncontrolled conditions were analyzed to find their effects on peak power, net reactivity, and temperature. Then, the effects of changes in kinetic parameters including the Doppler coefficient, prompt neutron lifetime, and delayed neutron fractions on simultaneous reactivity insertion and beam tube flooding accidents were analyzed. Results show that the power peak values are significantly sensitive to the Doppler coefficient of the system in coupled accidents. The material testing reactor-type system under such a coupled accident is not very sensitive to changes in the prompt neutron life time; the core under such a coupled transient is not very sensitive to changes in the effective delayed neutron fraction.

• Nuclear Engineering and Technology
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• 제41권3호
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• pp.307-318
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• 2009

KAERI (Korea Atomic Energy Research Institute) has developed the GAMMA+ code for a thermo-fluid and safety analysis of a VHTR (Very High Temperature Gas-Cooled Reactor). A key safety issue of the VHTR design is to demonstrate its inherent safety features for an automatic reactor power trip and power stabilization during an anticipated transient without scram (ATWS) accident such as a loss of forced cooling by a trip of the helium circulator (LOFC) or a reactivity insertion by a control rod withdrawal (CRW). This paper intends to show the ATWS assessment capability of the GAMMA+ code which can simulate the reactor power response by solving the point-kinetic equations with six-group delayed neutrons, by considering the reactivity changes due to the effects of a core temperature variation, xenon transients, and reactivity insertions. The present benchmark calculations are performed by using the safety demonstration experiments of the 10 MW high temperature gas cooled-test module (HTR-10) in China. The calculation results of the power response transients and the solid core temperature behavior are compared with the experimental data of a LOFC ATWS test and two CRW ATWS tests by using a 1mk-control rod and a 5mk-control rod, respectively. The GAMMA+ code predicts the power response transients very well for the LOFC and CRW ATWS tests in HTR-10.

• 한국에너지공학회:학술대회논문집
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• 한국에너지공학회 2000년도 추계 학술발표회 논문집
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• pp.115-120
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• 2000

The analysis of flux distribution under stead-state in large power reactors with assymetry reactivity insertions requires the use of three-dimensional diffusion calculations. For the purpose, consistently formulated modern nodal methods based on higher order interface techniques have become popular tools for flux distributions in large commercial nuclear reactors. Among the earlier developments, the nodal Green's function method obtains its nodal interface equation from the transverse-integrated integral diffusion equation using a finite-medium Green's function. In this method, the outgoing current from a node surface is formulated as a response of the incoming currents and the spatially integrated neutron source within the same node. The well-known nodal expansion method is also based on an interface partial current formulation. Nodal methods high-level interface variables, i.e., interface net current and flux, may be more computationally efficient than the nodal Green's function method because they have one fewer unknown per interface. The Analytic Nodal Method(ANM), which can be classified as an interface net current technique and, was faster in solving some standard benchmark problems than the other two methods.(omitted)

• 에너지공학
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• 제17권1호
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• pp.23-30
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• 2008

용융염 원자로는 고체핵연료를 사용하는 고전 원자로와는 달리 유동성을 갖는 액체핵연료를 장전하여 운전한다. 기존 동특성 코드는 핵연료의 유동으로 인한 동적 노물리 특성 영향을 고려하지 않기 때문에 용융염 원자로의 동특성 및 안전해석에 사용할 경우 신뢰성을 보장할 수 없다. 지금까지는 핵연료의 유동을 고려한 1점 동특성방정식을 이용하여 제한적으로 시스템안정성분석을 수행해 왔으나 이 경우 상세한 노심구조에서의 핵연료 및 중성자 거동에 대한 공간 종속성을 평가할 수 없다. 그러므로 핵연료의 유동 특성이 고려된 다차원 동특성 모델을 해석할 수 있는 컴퓨터 코드 개발이 필요하다. 본 논문은 용융염 원자로의 공간종속 중성자 동특성 해석을 위한 2군, 2차원 코드인 AMBIKIN2D의 개발 및 이에 수반하는 검증연구의 일환으로서 MSRE의 안정성실증실험을 모사하였다. 또한 비교 대상으로는 ORNL에서 개발한 Lumped parameter 방법을 사용한 일점 동특성 방정식에 의한 계산 결과를 포함하여 AMBIKIN2D의 정확성을 확인하였다.

• Nuclear Engineering and Technology
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• 제19권3호
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• pp.192-197
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• 1987

다량의 중수반사체를 가진 조밀한 노심에서는 핵분열시 발생하는ｒ선과 중수소와의 (ｒ,n) 반응에 의해 지발 광중성자가 다량 생성되므로 이러한 계통을 기술하기 위하여 광중성자와 그 모핵종의 공간적 분리에 역점을 두어 2-점노 운동방정식을 정립하였다. 여러 반응도를 주입하여 출력 천이를 모사계산하므로써 노심과 반사체사이의 관련 효과를 조사하였다. 이 모델에 의한 모사계산 결과와 공간 종속 운동방정식에 의한 계산결과를 비교하였다. 반사체 영역에서의 광중성자 효과가 포함되므로써, 이를 포함하지 않은 모델에 비해 출력 천이현상을 감소시켰다. 실제로 출력을 측정하는 계측기는 이러한 공간적 분리영 향을 제거하기 위하여 노심 내부에 위치하여야 한다.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.973-979
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• 2024

In the framework of the Generation IV research and development project, in which the French Commission of Alternative and Atomic Energies (CEA) is involved, a main objective for the design of Sodium-cooled Fast Reactor (SFR) is to meet the safety goals for severe accidents. Among the severe ones, the Unprotected Transient OverPower (UTOP) accidents can lead very quickly to a global melting of the core. UTOP accidents can be considered either as slow during a Control Rod Withdrawal (CRW) or as fast. The paper focuses on fast UTOP accidents, which occur in a few milliseconds, and three different scenarios are considered: rupture of the core support plate, uncontrolled passage of a gas bubble inside the core and core mechanical distortion such as a core flowering/compaction during an earthquake. Several levels and rates of reactivity insertions are also considered and the thermal-mechanical behavior of an ASTRID fuel pin from the ASTRID CFV core is simulated with the GERMINAL code. Two types of fuel pins are simulated, inner and outer core pins, and three different burn-up are considered. Moreover, the feedback from the CABRI programs on these type of transients is used in order to evaluate the failure mechanism in terms of kinetics of energy injection and fuel melting. The CABRI experiments complete the analysis made with GERMINAL calculations and have shown that three dominant mechanisms can be considered as responsible for pin failure or onset of pin degradation during ULOF/UTOP accident: molten cavity pressure loading, fuel-cladding mechanical interaction (FCMI) and fuel break-up. The study is one of the first step in fast UTOP accidents modelling with GERMINAL and it has shown that the code can already succeed in modelling these type of scenarios up to the sodium boiling point. The modeling of the radial propagation of the melting front, validated by comparison with CABRI tests, is already very efficient.