• Nuclear Engineering and Technology
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• 제49권6호
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• pp.1326-1338
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• 2017

The continuous energy Monte Carlo neutron transport code, MC21, was coupled to the CTF subchannel thermal-hydraulics code using a combination of Consortium for Advanced Simulation of Light Water Reactors (CASL) tools and in-house Python scripts. An MC21/CTF solution for VERA Core Physics Benchmark Progression Problem 6 demonstrated good agreement with MC21/COBRA-IE and VERA solutions. The MC21/CTF solution for VERA Core Physics Benchmark Progression Problem 7, Watts Bar Unit 1 at beginning of cycle hot full power equilibrium xenon conditions, is the first published coupled Monte Carlo neutronics/subchannel T-H solution for this problem. MC21/CTF predicted a critical boron concentration of 854.5 ppm, yielding a critical eigenvalue of $0.99994{\pm}6.8E-6$ (95% confidence interval). Excellent agreement with a VERA solution of Problem 7 was also demonstrated for integral and local power and temperature parameters.

• 방사선산업학회지
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• 제4권1호
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• pp.33-38
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• 2010

The MCNP5 neutron, electron, photon Monte Carlo transport program was installed on the KISTI's SUN Tachyon computer using the parallel programming. Electron beam accelerators were modeled and shielding calculations were performed in order to investigate the reduction of computation time in the supercomputer environment. It was observed that a speedup of 40 to 80 of computation time can be obtained using 64 CPUs compared to an IBM PC.

• Nuclear Engineering and Technology
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• 제8권4호
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• pp.209-217
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• 1976

슬레브형로내의 중성자수송에 관한 적분 방정식을 유한요소법으로 기술했다. Hermite 내삽에 의한 1차 및 3차 다항식을 사용하여 얻은 수식으로 글레브형로의 임계치와 Milne 문제의 근사해를 계산하고 해석해와의 비교를 통하여 유한 요소법의 유용성을 논의했다.

• Nuclear Engineering and Technology
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• 제53권2호
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• pp.688-694
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• 2021

The aim of this work is the analysis, design and hardware implementation of the fractional-order point kinetics (FNPK) model along with its closed-loop controller. The stability and closed-loop control of FNPK models are critical issues. The closed-loop stability of the controller-plant structure is established. Further, the designed PI/PD controllers are implemented in real-time on a DSP processor. The simulation and real-time hardware studies confirm that the designed PI/PD controllers result in a damped stable closed-loop response.

• Nuclear Engineering and Technology
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• 제50권8호
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• pp.1339-1348
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• 2018

The aim of this paper is to determine neutronic performances of the light water reactor (LWR) spent fuel mixed with fertile thorium fuel in a FFHR. Time dependent three dimensional calculations for major technical data, such as blanket energy multiplication, tritium breeding ratio, cumulative fissile fuel enrichment and burnup have been performed by using Monte Carlo Neutron-Particle Transport code MCNP5 1.4, coupled with a novel interface code MCNPAS, which is developed by our research group. A self-sustaining tritium breeding ratio (TBR>1.05) has been kept throughout the calculations. The study has shown that the fissile fuel quality will be improved in the course of the transmutation of the LWR spent in the FFHR. The latter has gained the reusable fuel enrichment level conventional LWRs between one and two years. Furthermore, LWR spent fuel - thorium mixture provides higher burn-up values than in light water reactors.

• Nuclear Engineering and Technology
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• 제53권2호
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• pp.414-422
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• 2021

The aim of this work considers a second order point reactor kinetics model based on the P1 approximation of transport theory, called in this work as P1 point reactor model. The P1 point reactor model implicitly considers the time derivative of the neutron source which has not been thus considered previously. The inverse method to calculate the reactivity in nuclear reactors -chosen because its high accuracy- Matrix Formulation. The numerical results shown that the Matrix Formulation for the reactivity estimation constitutes a method with insignificant calculation errors.

• Nuclear Engineering and Technology
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• 제55권1호
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• pp.324-329
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• 2023

Different types of deterministic solution methods were used to solve neutron transport equations corresponding to half-space and slab albedo problems. In these types of solution methods, in addition to the error of the numerical solutions, the obtained results contain truncation and discretization errors. In the present work, a non-analog Monte Carlo method is provided to simulate the half-space and slab albedo problems with Inönü and Anlı-Güngör strongly anisotropic scattering functions. For each scattering function, the sampling method of the direction of the scattered neutrons is presented. The effects of different beams with different angular dependencies and the effects of different scattering parameters on the reflection probability are investigated using the developed Monte Carlo method. The validity of the Monte Carlo method is also confirmed through the comparison with the published data.

• Nuclear Engineering and Technology
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• 제55권4호
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• pp.1210-1217
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• 2023

The assessment of radiation fields in the lithium loop pipes and dump tank during the operation were performed for International Fusion Materials Irradiation Facility - DEMO-Oriented NEutron Source (IFMIF-DONES) in order to obtain the radiation dose-rate maps in the component surroundings. Variance reduction techniques such as weight window mesh (produced with the ADVANTG code) were applied to bring the statistical uncertainty down to a reasonable level. The biological dose was given in the study, and potential shielding optimization is suggested and more thoroughly evaluated. The MCNP Monte Carlo was used to simulate a gamma particle transport for radiation shielding purposes for the current Li Systems' design. In addition, the shielding efficiency was identified for the Impurity Control System components and the dump tank. The analysis reported in this paper takes into account the radiation decay source from and activated corrosion products (ACPs), which is created by d-Li interaction. As a consequence, the radiation (resulting from ACPs and Be-7) shielding calculations have been carried out for safety considerations.

• Journal of Radiation Protection and Research
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• 제18권2호
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• pp.47-59
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• 1993

본 연구에서는 사용후핵 연료를 안전하게 수송할 수 있는 수송용기의 여러 가지 설계 항목중에 수송용기 내부에 장전한 핵연료에 의한 핵임계반응을 방지하기 위한 핵임계해석을 수행하였다. 핵임계 해석에 사용한 HANSEN-ROACH-KENO-Va 전산시스템에 대한 검증계산을 수행하였고 수송용기의 핵임계측면에서의 안전성을 확보하기 위해 가능한 보수적인 가정을 하여 어떠한 경우에도 수송용기에 장전된 핵연료가 임계상태에 도달하지 않도록 수송용기 내부의 구조 및 적절한 핵임계 방지제를 선택하였고 정상수송 및 가상사고 조건 등에 대한 해석을 수행하였다. 그 결과 KSC-7 수송용기 의 설계조건을 만족하고 핵임계측면에서의 안전성을 보장할 수 있는 재료 및 구조에 대한 결론을 해석적으로 도출하였다.

• Nuclear Engineering and Technology
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• 제54권4호
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• pp.1253-1260
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• 2022

This paper reports the control method for the core power of the China initiative Accelerator Driven System (CiADS) facility. In the CiADS facility, an intense external neutron source provided by a proton accelerator coupled to a spallation target is used to drive a sub-critical reactor. Without any control rod inside the sub-critical reactor, the core power is controlled by adjusting the proton beam intensity. In order to continuously change the beam intensity, an adjustable aperture is considered to be used at the Low Energy Beam Transport (LEBT) line of the accelerator. The aperture size is adjusted based on the Proportional Integral Derivative (PID) controllers, by comparing either the setting beam intensity or the setting core power with the measured value. To evaluate the proposed control method, a CiADS core model is built based on the point reactor kinetics model with six delayed neutron groups. The simulations based on the CiADS core model have indicated that the core power can be controlled stably by adjusting the aperture size. The response time in the adjustment of the core power depends mainly on the adjustment time of the beam intensity.