• Nuclear Engineering and Technology
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• 제43권6호
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• pp.523-530
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• 2011

In a previous study, the stochastic space-dependent kinetics model (SSKM) based on the forward stochastic model in stochastic kinetics theory and the Ito stochastic differential equations was proposed for treating monoenergetic space-time nuclear reactor kinetics in one dimension. The SSKM was tested against analog Monte Carlo calculations, however, for exemplary cases of homogeneous slab reactors with only one delayed-neutron precursor group. In this paper, the SSKM is improved and evaluated with more realistic and complicated cases regarding several delayed-neutron precursor groups and heterogeneous slab reactors in which the extraneous source or reactivity can be introduced locally. Furthermore, the source level and the initial conditions will also be adjusted to investigate the trends in the variances of the neutron population and fission product levels across the reactor. The results indicate that the improved SSKM is in good agreement with the Monte Carlo method and show how the variances in population dynamics can be controlled.

• 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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• 제53권6호
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• pp.1731-1735
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• 2021

The operation of the nuclear reactor requires accurate and fast methods and techniques for analysing its kinetics. These techniques become even more important when the MOX-fuel is used due to the lower value of delayed neutron fraction 𝛽 for ²³⁹Pu. Based on a Birth-and-Death process review, the mathematical model of thermal reactor core has been proposed different from existing ones. The analytical method for thermal point-reactor parameters evaluation is described within this work. The proposed method is applied for analysis of the unsteady transient processes taking place in a thermal reactor at its start-up or shutdown power change, as well as during small accidental power variation from the rated value. Theoretical determination of MASURCA reactor core reactivity through the analysis of experimental data on neutron time spectra was made.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.920-927
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• 2017

With ever-advancing computer technology, the Monte Carlo (MC) neutron transport calculation is expanding its application area to nuclear reactor transient analysis. Dynamic MC (DMC) neutron tracking for transient analysis requires efficient algorithms for delayed neutron generation, neutron population control, and initial condition modeling. In this paper, a new MC steady-state simulation method based on time-dependent MC neutron tracking is proposed for steady-state initial condition modeling; during this process, prompt neutron sources and delayed neutron precursors for the DMC transient simulation can easily be sampled. The DMC method, including the proposed time-dependent DMC steady-state simulation method, has been implemented in McCARD and applied for two-dimensional core kinetics problems in the time-dependent neutron transport benchmark C5G7-TD. The McCARD DMC calculation results show good agreement with results of a deterministic transport analysis code, nTRACER.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.102-107
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• 1997

A stiffness in a dynamical system is numerically studied to investigate a sensitivity of a reactor to the delayed neutron spectra with the Doppler feedback. To test numerical procedure, we adopted a case of a reactivity accident in a point reactor model. We found that the stiffness is sensitive to a reactivity insertion rate and the delayed neutron spectra in the Doppler feedback phase. Our numerical results show that global reactor characteristics are not very sensitive to the delayed neutron spectra even though their instantaneous ones are sensitive. We present the time evolution of each precursor group explicitly.

• Nuclear Engineering and Technology
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• 제47권1호
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• pp.94-101
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• 2015

This paper presents findings on the second-order sliding-mode controller for a nuclear research reactor. Sliding-mode controllers for nuclear reactors have been used for some time, but higher-order sliding-mode controllers have the added advantage of reduced chattering. The nonlinear model of Pakistan Research Reactor-1 has been used for higherorder sliding-mode controller design and performance evaluation. The reactor core is simulated based on point kinetics equations and one delayed neutron groups. The model assumes feedback from lumped fuel and coolant temperatures. The effect of xenon concentration is also considered. The employed method is easy to implement in practical applications, and the second-order sliding-mode control exhibits the desired dynamic properties during the entire output-tracking process. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness, and stability.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2000년도 추계학술대회 논문집 학회본부 D
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• pp.638-640
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• 2000

Reactor Noise is defined as the fluctuations of measured instrumentation signals during full-power operation of reactor which have informations on reactor system dynamics such as neutron kinetics, thermal-hydraulics, and structural dynamics. Reactor noise analyses of ex-core neutron detector internals such as fuel assembly and Core Support Barrel in Nuclear Power Plant. A real time mode separation technique have been developed and applied for the analyses. The analyses data base have been constructed for the continuous monitoring and diagnose of the reactor internals. Detailed design of diagnostic system reactor internal structures using neutron noise(RIDS).

• Nuclear Engineering and Technology
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• 제50권7호
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• pp.1051-1059
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• 2018

Fast periodic pulsed reactor is a type of reactor in which the fission bursts are formed entirely with external reactivity modulation with a specified time periodicity. This type of reactors could generate much larger intensity of neutron beams for experimental use, compared with the steady state reactors. In the design of fast periodic pulsed reactors, the time dependent simulation of the power pulse is majorly based on a point kinetic model, which is known to have limitations. A more accurate calculation method is desired for the design analyses of fast periodic pulsed reactors. Monte Carlo computer code MCNP6 is used for this task due to its three dimensional transport capability with a continuous energy library. Some new routines were added to simulate the rotation of the movable reflector parts in the time dependent calculation. Fast periodic pulsed reactor IBR-2M was utilized to validate the new routines. This reactor is periodically in prompt supercritical state, which lasts for ${\sim}400{\mu}s$, during the equilibrium state. This generates long neutron fission chains, which requires tremendously large amount of computation time during Monte Carlo simulations. Russian Roulette was applied for these very long neutron chains in MCNP6 calculation, combined with other approaches to improve the efficiency of the simulations. In the power pulse of the IBR-2M at equilibrium state, there is some discrepancy between the experimental measurements and the calculated results using the point kinetics model. MCNP6 results matches better the experimental measurements, which shows the merit of using MCNP6 calculation relative to the point kinetics model.

• 한국에너지공학회:학술대회논문집
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• 한국에너지공학회 1994년도 추계학술발표회 초록집
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• pp.141-145
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• 1994

A computer code for solving the 3-D time-dependent multigroup neutron diffusion equation by a coupled reactor kinetics method recently developed has been developed and for evaluating its applicability in CANDU transient analysis applied to a 3-D kinetics benchmark problem which reveals non-uniform loss of coolant accident followed by an asymmetric insertion of shutdown devices. The performance of the method and code has been compared with the CANDU design code, CERBERUS, employing a finite difference improved quasistatic method.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2001년도 하계학술대회 논문집 D
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• pp.2166-2168
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• 2001

Reactor Noise is defined as the fluctuations of measured instrumentation signals during full-power operation of reactor which have informations on reactor system dynamics such as neutron kinetics. The Reactor internal structures which consist of many complex components are subjected to flow-induced vibration due to high temperature and pressure in reactor coolant system. The above flow-induced vibration causes degradation of structural integrity of the reactor and may result in loosing mechanical binding component which might impact other equipment and component or cause flow blockage. It is important to analyze reactor noise signal for the early detection of potential problem or failure in order to diagnosis reactor integrity in the point of view of safety and plant economics. Detailed design of hardware diagnostic system reactor internal structures using neutron noise(RIDS).