• Nuclear Engineering and Technology
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• 제32권6호
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• pp.549-558
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• 2000

A transient model for hydrogen generation in molten metal-water interactions was developed with separate models for two stages of coarse mixing and stratification. The model selves the mechanistic equations (heat and mass transfer correlation, heat conduction equation and the concentration diffusion equation) of each stage with non-zero boundary conditions. Using this model, numerical simulations were performed for single droplet experiments in the Argonne National Laboratory tests and for FITS tests that simulated dynamic fragmentation and stratification. The calculation results of hydrogen generation showed better agreement to the experiment data than those of previous works. It was found from the analyses that the steam concentration to be reached at the reaction front might be the main constraint to the extent of the metal droplet oxidized. Also, the hydrogen generation rate in the coarse mixing stage was the higher than that in the stratification stage. The particle size was the most important factor in the coarse mixing stage to predict the amount of hydrogen generation.

• Nuclear Engineering and Technology
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• 제26권1호
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• pp.100-107
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• 1994

가압경수로용 핵연료붕 내압을 계산하는 데 있어 현재의 결정론적 방법에 의한 과다한 보수성을 줄이기 위하여 통계적 계산 방법론을 개발하였다. 개발된 통계적 방법론은 반응표면 분석 방법과 Monte Carlo 계산 방법을 이용하였다. 반응표면 분석 방법을 이용하여 핵연료 제조관련 변수와 성능관련 변수를 고려하여 회귀식을 유도하였으며, 이 식의 검증은 F-test, $R^2$ 및 $C^{p}$-test 방법을 사용하여 수행하였다. 회귀식으로 부터 예측된 봉내압은 결정론적 방법을 사용하여 계산된 값과 잘 일치하였다. Monte Carlo 계산으로 구한 핵연료봉 내압의 분포는 거의 정상분포로 나타났다. 본 연구에서 개발된 통계적 방법론으로 구한 95/95 봉내압과 현재 사용되고 있는 결정론적 방법론의 봉내압과 비교한 결과 결정론적 방법론의 과다한 보수성을 크게 줄일 수 있었다.다.

• Nuclear Engineering and Technology
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• 제48권1호
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• pp.33-42
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• 2016

Spacer grids play an important role in maintaining the proper form of the fuel assembly structure and ensuring the safety of reactor core design. This study applies the Monte Carlo method to the analysis of the neutronics effects of spacer grids in a typical pressurized water reactor (PWR). The core problem used to analyze the neutronics effects of spacer grids is a modified version of Korea Advanced Institute of Science and Technology benchmark problem 1B, based on an Advanced Power Reactor 1400 (APR1400) core model. The spacer grids are modeled and added to this test problem in various ways. Then, by running MCNP5 for all cases of spacer grid modeling, some important numerical results, such as the effective multiplication factor, the spatial distributions of neutron flux, and its energy spectrum are obtained. The numerical results of each case of spacer grid modeling are analyzed and compared to assess which type has more advantages in accuracy of numerical results and effectiveness in terms of geometry building. The conclusion is that the most realistic modeling for Monte Carlo calculation is the "volume-preserving" streamlined heterogeneous spacer grids, but the "banded" dissolution spacer grids modeling is a more practical yet accurate model for routine (deterministic) analysis.

• Nuclear Engineering and Technology
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• 제44권8호
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• pp.961-970
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• 2012

Often, probabilistic fracture mechanics (PFM) approaches have been adopted to quantify the failure probabilities of Ni-base alloy components, especially due to primary water stress corrosion cracking (PWSCC), in a primary piping system of pressurized water reactors. In this paper, the key features of an advanced PFM code, PINEP-PWSCC (Probabilistic INtegrity Evaluation for nuclear Piping-PWSCC) for such purpose, are described. In developing the code, we adopted most recent research results and advanced models in calculation modules such as PWSCC crack initiation and growth models, a performance-based probability of detection (POD) model for Ni-base alloy welds, and so on. To verify the code, the failure probabilities for various Alloy 182 welds locations were evaluated and compared with field experience and other PFM codes. Finally, the effects of pre-existing crack, weld repair, and POD models on failure probability were evaluated to demonstrate the applicability of PINEP-PWSCC.

• Nuclear Engineering and Technology
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• 제53권3호
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• pp.715-731
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• 2021

Uncertainties are calculated for pressurized water reactor (PWR) spent nuclear fuel (SNF) characteristics. The deterministic code STREAM is currently being used as an SNF analysis tool to obtain isotopic inventory, radioactivity, decay heat, neutron and gamma source strengths. The SNF analysis capability of STREAM was recently validated. However, the uncertainty analysis is yet to be conducted. To estimate the uncertainty due to nuclear data, STREAM is used to perturb nuclear cross section (XS) and resonance integral (RI) libraries produced by NJOY99. The perturbation of XS and RI involves the stochastic sampling of ENDF/B-VII.1 covariance data. To estimate the uncertainty due to modeling parameters (fuel design and irradiation history), surrogate models are built based on polynomial chaos expansion (PCE) and variance-based sensitivity indices (i.e., Sobol' indices) are employed to perform global sensitivity analysis (GSA). The calculation results indicate that uncertainty of SNF due to modeling parameters are also very important and as a result can contribute significantly to the difference of uncertainties due to nuclear data and modeling parameters. In addition, the surrogate model offers a computationally efficient approach with significantly reduced computation time, to accurately evaluate uncertainties of SNF integral characteristics.

• 방사성폐기물학회지
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• 제13권1호
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• pp.21-34
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• 2015

원자력발전소 해체를 준비하기 위해서는 해체대상 발전소에 대한 선원항 평가가 선행되어야 한다. 해체전략 수립단계에서 선원항 평가 결과를 토대로 해체 폐기물을 분류하고 비용평가를 수행한다. 본 연구에서는 월성 1호기의 예비 선원항 계산을 수행할 수 있도록 MCNP/ORIGEN-2 모델의 타당성 평가를 수행하였다. 연소도가 다른 핵연료 다발의 악티나이드 계열과 핵분열 생성물의 핵종 수밀도는 싱글 채널 모델을 이용하여 MCNPX 코드로 연소 계산하여 구하였다. 선원항의 정확도에 영향을 미치는 두가지 요인에 대해 조사하였다. 첫번째 요인으로 선원항 계산에 영향을 미치는 중성자 스펙트럼을 MCNP로 계산하여 해당 핵종의 1군 미시 핵단면적에 반영하였다. 중성자 스펙트럼이 반영된 라이브러리로 계산한 선원항과 ORIGEN-2 코드 package에 내장된 library (CANDUNAU.LIB)로 구한 선원항을 비교하였다. 두번째 요인으로 선원항에 대한 출력이력의 영향을 조사하였다. 해체 폐기물의 저준위 폐기물 처분 가능성을 살펴보기 위해, 2010년도 교체된 압력관, 칼란드리아관과 기존 칼란드리아 동체에 대하여 중성자 스펙트럼을 반영한 library를 적용하여 MCNP/ORIGEN-2로 선원항 평가 계산을 수행하였다.

• Nuclear Engineering and Technology
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• 제54권9호
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• pp.3347-3352
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• 2022

In 2013, the International Atomic Energy Agency (IAEA) changed the recommended maximum range of the Emergency Planning Zone (EPZ) to 30 km, and the Kori Nuclear Power Plant in Republic of Korea has also expanded the EPZ to 30 km, following the recommendation. As a result, metropolitan cities with a high population density are contained within the EPZ, and evacuating millions of people should be considered if the 30 km range of evacuation is to take place. This study proposes an evacuation plan using buses (public transportation) to transport people outside of the EPZ, quickly and efficiently. To verify the appropriate mode share ratio of buses that can guarantee the right of vulnerable road users and reduce traffic congestion, a model was built simulating the Kori Nuclear Power Plant in Ulsan Metropolitan City. The scenarios were established by changing the mode share ratio of buses and passenger cars by 10%. Considering a large-scale network analysis at the city level, a cell transmission model was applied to calculate the evacuation time in each scenario. The result shows that the optimal mode share ratio of buses is 40%, with a total evacuation time of 132 min, considering feasible bus fleets in Ulsan Metropolitan City.

• Nuclear Engineering and Technology
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• 제30권1호
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• pp.26-39
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• 1998

The containment pressure and temperature transient analysis computer code CONTEMPT4/MOD5 is modified to incorporate the passive containment cooling models. The correlations are selected from the existing experimental heat transfer correlations to model the natural and mixed convection in annular space between the containment shell and the shield building. The evaporative heat transfer of the water film on the outer shell of the containment is modeled using the correlations derived from the analogy between the heat and mass transfer. The modified code is applied to the Ap600 containment transient analysis for the model verification and the results are compared to the results of GOTHIC calculation done by Westinghouse. Also, d series of parametric sensitivity studies of heat transfer correlations, water film ratio and delay time of the wet cooling on the containment peak pressure and temperature following LOCA are performed for the containment of 1000MWe passive plant, KP1000.

• Nuclear Engineering and Technology
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• 제52권10호
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• pp.2402-2409
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• 2020

Analytical criteria for the onset of fuel fragmentation and Burst Fission Gas Release in fuel rods with ballooned claddings are formulated. On that basis, the GRSW-A model integrated with a fuel behaviour code is updated. After modification, the updated code is successfully applied to simulation of the Halden LOCA test IFA-650.12. Specifically, the calculation with Burst Fission Gas Release during the test resulted in prediction of cladding failure, whereas it could not be predicted at the test planning, before new models were implemented. A good agreement of the current model with experimental data for transient Fission Gas Release in the tests IFA-650.12 and IFA-650.14 is shown, as well.

• Nuclear Engineering and Technology
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• 제36권2호
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• pp.175-183
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• 2004

For the analysis of a dry process fuel cycle, new modules were implemented into the fuel cycle analysis code DYMOND, which was developed by the Argonne National Laboratory. The modifications were made to the energy demand prediction model, a Canada deuterium uranium (CANDU) reactor, direct use of spent pressurized water reactor (PWR) fuel in CANDU reactors (DUPIC) fuel cycle model, the fuel cycle calculation module, and the input/output modules. The performance of the modified DYMOND code was assessed for the postulated once-through fuel cycle models including both the PWR and CANDU reactor. This paper presents modifications of the DYMOND code and the results of sample calculations for the PWR once-though and DUPIC fuel cycles.