• Nuclear Engineering and Technology
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• 제55권9호
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• pp.3450-3463
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• 2023

In the two-step analysis of Pebble-Bed type High-Temperature Gas-Cooled Reactor (PB-HTGR), the lattice physics calculation for the generation of homogenized cross-sections is based on the fuel pebble. However, the randomly-dispersed fuel particles in the fuel pebble introduce double heterogeneity and randomness. Compared to the deterministic method, the Monte Carlo method which is flexible in geometry modeling provides a high-fidelity treatment. Therefore, the Monte Carlo code NECP-MCX is extended in this study to perform the lattice physics calculation of the PB-HTGR. Firstly, the capability for the simulation of randomly-dispersed media, using the explicit modeling approach, is developed in NECP-MCX. Secondly, the capability for the generation of the homogenized cross-section is also developed in NECP-MCX. Finally, simplified PB-HTGR problems are calculated by a two-step neutronics analysis tool based on Monte Carlo homogenization. For the pebble beds mixed by fuel pebble and graphite pebble, the bias is less than 100 pcm when compared to the high-fidelity model, and the bias is increased to 269 pcm for pebble bed mixed by depleted fuel pebble. Numerical results show that the Monte Carlo lattice physics calculation for the two-step analysis of PB-HTGR is feasible.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2011년도 제42회 하계학술대회
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• pp.1027-1028
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• 2011

In this paper, three reliability calculation algorithms: Monte Carlo Simulation (MCS), Reliability Index Approach (RIA), and Sensitivity-based Monte Carlo Simulation (SMCS) are studied. Their efficiency and accuracy are validated by analytic test functions.

• 대한방사선방어학회:학술대회논문집
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• 대한방사선방어학회 2011년도 추계 학술발표회 및 심포지엄
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• pp.170-171
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• 2011

• Journal of Radiation Protection and Research
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• 제39권1호
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• pp.30-37
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• 2014

최근 표준한국인 성인남녀 복셀 팬텀 한 쌍인 High-Definition Reference Korean-Man (HDRK-Man)과 High-Definition Reference Korean-Woman (HDRK-Woman)을 개발한 바 있다. HDRK-Man과 HDRK-Woman은 각각 한국인 성인 남성과 여성을 대표하므로 이들을 이용하면 한국인 고유의 유효선량을 계산할 수 있다. 이러한 한국인 고유의 유효선량을 계산하기 위해서는 표준한국인 팬텀을 이용한 몬테칼로 선량계산을 수행하여야 한다. 하지만 이러한 몬테칼로 선량계산은 전산인체모델에 대한 이해와 고도의 몬테칼로 전산모사 기술이 요구되기 때문에 몬테칼로 전산모사 경험이 적은 연구자들에게는 접근성이 용이하지 않다. 최근 UFPE (Federal University of Pernambuco) 연구팀에서는 몬테칼로 선량계산 온라인 시스템인 CALDOSE_X를 운영하고 있다. CALDOSE_X는 사용자가 온라인상에서 직접 입력한 선원조건에 따라 곧바로 몬테칼로 선량계산을 수행하고 계산된 결과를 제공해주는 온라인 서비스이다. 따라서 CALDOSE_X를 이용하면 몬테칼로 전산모사 경험 없이도 누구나 쉽게 원하는 선원조건에 대한 선량계산을 수행할 수 있다. 하지만 CALDOSE_X에서 제공하는 선량값은 서양인 전산팬텀을 이용하기 때문에, 이를 한국인 종사자에 대한 방사선 방호목적으로 사용하기에는 여전히 한계가 있다. 이에 본 연구에서는 온라인상에서 사용자가 입력한 선원조건에 따라 실시간으로 표준 한국인에 대한 유효선량을 계산할 수 있는 웹 기반 표준한국인 선량평가 시스템을 구축하였다.

• 한국의학물리학회:학술대회논문집
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• 한국의학물리학회 2004년도 제29회 추계학술대회 발표논문집
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• pp.43-46
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• 2004

The commissioning of a treatment planning system of model-based dose calculation algorithm requires a lot of parameters to be selected to fit measured data, in which process physical insights for the parameters are often forgotten. We present the photon beam commissioning of Pinnacle$^3$ with the help of Monte Carlo (MC) simulation and evaluate the parameters Pinnacle$^3$ demands. Even though the MC calculation produces reasonable values for the commissioning, the thorough physical basis of the Pinnacles$^3$'s commissioning process is needed to use the MC derived parameters directly.

• Nuclear Engineering and Technology
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• 제46권3호
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• pp.407-412
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• 2014

In this research, we investigated the burnup characteristics and the conversion of fertile $^{232}Th$ into fissile $^{233}U$ in the core of a Sodium-Cooled Fast Reactor (SFR). The SFR fuel assemblies were designed for burning $^{232}Th$ fuel (fuel pin 1) and $^{233}U$ fuel (fuel pin 2) and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method) and TTA (Transmutation Trajectory Analysis) method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff) was between 0.964 and 0.954 and peaking factor is 1.88867.

• 조명전기설비학회논문지
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• 제28권6호
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• pp.7-14
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• 2014

We proposed a new algorithm to improve the calculation speed of forward ray tracing method which was quite a problem in conventional Monte Carlo algorithm. To verify the accuracy and the effect of improving calculation speed, we directly compared integrating sphere characteristics with conventional ray tracing algorithm under the same condition which was referred in a reference paper. By applying new algorithm to an absolute reflectometer, we calculated the degree of errors which were caused by baffle and port characteristics to find optimal system design condition.

• Nuclear Engineering and Technology
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• 제49권6호
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• pp.1172-1180
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• 2017

The application of Monte Carlo (MC) to large-scale fixed-source problems has recently become possible with new hybrid methods that automate generation of parameters for variance reduction techniques. Two common variance reduction techniques, weight windows and source biasing, have been automated and popularized by the consistent adjoint-driven importance sampling (CADIS) method. This method uses the adjoint solution from an inexpensive deterministic calculation to define a consistent set of weight windows and source particles for a subsequent MC calculation. One of the motivations for source consistency is to avoid the splitting or rouletting of particles at birth, which requires computational resources. However, it is not always possible or desirable to implement such consistency, which results in inconsistent source biasing. This paper develops an original framework that mathematically expresses the coupling of the weight window and source biasing techniques, allowing the authors to explore the impact of inconsistent source sampling on the variance of MC results. A numerical experiment supports this new framework and suggests that certain classes of problems may be relatively insensitive to inconsistent source sampling schemes with moderate levels of splitting and rouletting.

• 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.

• Nuclear Engineering and Technology
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• 제54권3호
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• pp.811-816
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• 2022

This work presents the methodology used to expand the capabilities of the Monte Carlo code OpenMC for the calculation of reactor kinetic parameters: effective delayed neutron fraction β_eff and neutron generation time L. The modified code, OpenMC(Time-Dependent) or OpenMC(TD), was then used to calculate the effective delayed neutron fraction by using the prompt method, while the neutron generation time was estimated using the pulsed method, fitting Λ to the decay of the neutron population. OpenMC(TD) is intended to serve as an alternative for the estimation of kinetic parameters when licensed codes are not available. The results obtained are compared to experimental data and MCNP calculated values for 18 benchmark configurations.