• Nuclear Engineering and Technology
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• 제53권7호
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• pp.2151-2161
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• 2021

Nuclear data (ND) sensitivity and uncertainty (S/U) quantification in shielding applications is performed using deterministic and probabilistic approaches. In this paper the validation of the newly developed deterministic program package ASUSD (ADVANTG + SUSD3D) is presented. ASUSD was developed with the aim of automating the process of ND S/U while retaining the computational efficiency of the deterministic approach to ND S/U analysis. The paper includes a detailed description of each of the programs contained within ASUSD, the computational workflow and validation results. ASUSD was validated on two shielding benchmark experiments from the Shielding Integral Benchmark Archive and Database (SINBAD) - the fission relevant ASPIS Iron 88 experiment and the fusion relevant Frascati Neutron Generator (FNG) Helium Cooled Pebble Bed (HCPB) Test Blanket Module (TBM) mock-up experiment. The validation process was performed in two stages. Firstly, the Denovo discrete ordinates transport solver was validated as a standalone solver. Secondly, the ASUSD program package as a whole was validated as a ND S/U analysis tool. Both stages of the validation process yielded excellent results, with a maximum difference of 17% in final uncertainties due to ND between ASUSD and the stochastic ND S/U approach. Based on these results, ASUSD has proven to be a user friendly and computationally efficient tool for deterministic ND S/U analysis of shielding geometries.

• International Journal of Naval Architecture and Ocean Engineering
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• 제3권1호
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• pp.9-19
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• 2011

The paper describes the validation of two time domain methods to simulate the behaviour of a destroyer operating in steep, stern-quartering seas. The significance of deck-edge immersion and water on deck on the capsize risk is shown as well as the necessity to account for the wave disturbances caused by the ship. A method is described to reconstruct experimental wave trains and finally two deterministic validation cases are shown.

• 한국안전학회지
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• 제39권3호
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• pp.75-83
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• 2024

Accurately assessing and managing risks in any endeavor is crucial. Risk assessment in engineering translates the abstract concept of risk into actionable strategies for systematic risk management. However, risk validation is met with significant skepticism, particularly concerning the uncertainty of probability. This study aims to address the aforementioned uncertainty in a multitude of ways. Firstly, instead of relying on deterministic probability, it acknowledges uncertainty and presents a probabilistic interval. Secondly, considering the uncertainty interval highlighted in OREDA, it delineates the bounds of the probabilistic interval. Lastly, it investigates how much explanatory power deterministic probability has within the defined probabilistic interval. By utilizing fault tree analysis (FTA) and integrating confidence intervals, a probabilistic risk assessment was conducted to scrutinize the explanatory power of deterministic probability. In this context, explanatory power signifies the proportion of probability within the probabilistic risk assessment interval that lies below the deterministic probability. Research results reveal that at a 90% confidence interval, the explanatory power of deterministic probability decreases to 73%. Additionally, it was confirmed that explanatory power reached 100% only with a probability application 36.9 times higher.

• 대한전기학회논문지:시스템및제어부문D
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• 제55권5호
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• pp.199-204
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• 2006

To develop a safety-critical network in nuclear power plants that puts more stringent requirements than the competitive commercial ones do, we establish four design criteria - deterministic communication, explicit separation/isolation structure, reliability, verification & validation. According to those design criteria, the fundamental design elements are chosen as follows - a star topology, point-to-point physical link, connection-oriented link control and fixed allocation access control. After analyzing the design elements, we also build a communication architecture with TDM (Time Division Multiplexing) bus switching scheme. Finally, We develop a DDCNet (Deterministic Data Communication Network) based on the established architecture. The DDCNet is composed of 64 nodes and guarantees the transmission bandwidth of 10Mbps and the delay of 10 msec for each node. It turns out that the DDCNet satisfies the aforementioned design criteria and can be adequately utilized for our purpose.

• Nuclear Engineering and Technology
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• 제44권2호
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• pp.113-150
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• 2012

The point we made in this paper is that, although detailed and precise three-dimensional (3D) whole-core transport calculations may be obtained in the future with massively parallel computers, they would have an application to only some of the problems of the nuclear industry, more precisely those regarding multiphysics or for methodology validation or nuclear safety calculations. On the other hand, typical design reactor cycle calculations comprising many one-point core calculations can have very strict constraints in computing time and will not directly benefit from the advances in computations in large scale computers. Consequently, in this paper we review some of the deterministic 3D transport methods which in the very near future may have potential for industrial applications and, even with low-order approximations such as a low resolution in energy, might represent an advantage as compared with present industrial methodology, for which one of the main approximations is due to power reconstruction. These methods comprise the response-matrix method and methods based on the two-dimensional (2D) method of characteristics, such as the fusion method.

• 한국농공학회지
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• 제34권1호
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• pp.100-106
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• 1992

The objective of this study is to develop a real-time runoff forecasting model considering stochastic component. The model is composed of deterministic and stochastic components. Simplified tank model was selected as a deterministic runoff forecasting model. The time series of estimation residual resulting from the tank model simulation was analyzed and was best suited to the second-order autoregressive model. ARTANK model which combined the tank model with the autoregressive process was developed. And it was applied to a BANWEOL basin for validation. The simulation results showed a good agreement with the observed field data.

• 한국지형학회지
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• 제15권2호
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• pp.37-53
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• 2008

본 연구는 장기적인 기상 자료(meterological data)와 하천 유출량 자료(stream run off data)의 획득이 가능한 충청도 병천천 유역에 대해 BROOK90 4.4e 물리 결정 물 수지 모델(physical deterministic water balance model)을 사용하여 '병천천 유역의 물 수지 모델'을 수립한 것이다. 모델 조작 매개변수(model fitting parameter)를 교정(calibration)한 비준 모델(validation model)을 가지고 기상 자료(meterological data)가 있지만 하천 유출량 자료(stream runoff data)는 없는 시기에 대한 장기적인 물 수지를 수립하였다. 연구의 결과는 a priori 모의 단계에서 실측 하천 유출량(measured stream runoff data)과 모의 하천 유출량(simulated stream runoff data)이 유사하게 나옴으로써 물 수지 모의 실험(experiment for water balance modeling)이라는 연구 성격으로서 목표하는 첫 번째 기대 수준에 도달하고 있다. 모델 조작 매개변수(model fitting parameter)를 확정하고 수행한 비준 모의(validated simulation)를 통해 과거 9년(1998년 ~ 2006년)의 물 수지가 복원되었다. 이 유역의 지형(geomophology), 식생(vegetation), 토양(soil), 토지이용(land use) 상황이 변하지 않는다면 기상자료(meterological data)만 가지고서 언제나 하천 유출량(stream runoff amount), 토양수량(siol water amount) 그리고 증발산량(evapotranspiration) 등 다양한 수문기후 자료를 생산할 수 있다. 이 연구는 현재 한국의 물 수지(water balance) 수립은 물론이고 과거의 물 수지 복원(water balance reconstruction) 분야에 또 하나 새로운 지평을 열었다. 이러한 연구 결과는 한반도에서의 기후(climate)와 식생(vegetation)의 변화에 따른 미래 물 수지(water balance) 예측 분야에서도 널리 활용할 수 있을 것이다.

• 대한전기학회논문지:전력기술부문A
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• 제48권4호
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• pp.428-433
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• 1999

The errors of IMU(Inertial Measurement Unit) can be divided into deterministic and random errors. Since the required accuracy of the IMU is very high, the errors must be compensated by using an accurate error mode. In this paper, we present a method to get a more accurate error model in a laboratory test. This was done by considering the setting misalignment in the laboratory test in the IMU error model. We considered here the IMU which consits of DTG(dynamically tuned gyroscope) and pendulum type accelerometer. First, it was shown that the estimation result from the model which does not contain the setting misalignment gives considerable estimation error at the validation test. Second, a new model considering the setting misalignment was derived. Finally, by validation test using the estimation results from new model the validity of it was proved.

• 대한기계학회논문집A
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• 제30권8호
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• pp.923-930
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• 2006

A Kriging model is a sort of approximation model and used as a deterministic model of a computationally expensive analysis or simulation. Although it has various advantages, it is difficult to assess the accuracy of the approximated model. It is generally known that a mean square error (MSE) obtained from the kriging model can't calculate statistically exact error bounds contrary to a response surface method, and a cross validation is mainly used. But the cross validation also has many uncertainties. Moreover, the cross validation can't be used when a maximum error is required in the given region. For solving this problem, we first proposed a modified mean square error which can consider relative errors. Using the modified mean square error, we developed the strategy of adding a new sample to the place that the MSE has the maximum when the MSE is used for the assessment of the kriging model. Finally, we offer guidelines for the use of the MSE which is obtained from the kriging model. Four test problems show that the proposed strategy is a proper method which can assess the accuracy of the kriging model. Based on the results of four test problems, a convergence coefficient of 0.01 is recommended for an exact function approximation.

• Nuclear Engineering and Technology
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• 제56권8호
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• pp.3405-3424
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• 2024

Reactivity worth of fuel rods at the JSI TRIGA research reactor was measured. Differently burned fuel rods were chosen to validate fuel burnup calculations. Two methods of measuring reactivity worth of fuel rods are used, traditional method is compared to newly introduced method using fuel rods swapping. Connection between both methods is described theoretically and the theory is validated experimentally. Fuel rod worth calculated using the newly introduced fuel rod swap method was within 1σ of worth measured using the traditional method. In addition to the recently performed experiments, weekly measurements of reactor core reactivity throughout the operational history are used for validation. The measured data were used to validate the fuel burnup and core criticality calculations. Fuel burnup calculations are performed using three different computer codes: the deterministic TRIGLAV, the Monte Carlo Serpent-2, and the hybrid RAPID. Great agreement was observed for Serpent-2 and RAPID by simulating fuel rod worth and its burnup, indicating that the fuel burnup and criticality calculations are accurate and that reactivity changes due to small burnup differences on the order of 10 pcm can be accurately simulated. In addition it was shown using ex-core detectors and large fission chamber that detector response changes due to fuel swapping are evident for fuel rod burnup differences of 20 MWd/kg. Fuel burnup calculations were further validated on excess reactivity measurements for three mixed TRIGA cores. The calculated burnup reactivity coefficient Δρ_BU using Serpent-2 and RAPID was within 1σ of the measurements, showing both codes are capable of calculating burnup for different TRIGA fuel types.