• Nuclear Engineering and Technology
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• 제56권9호
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• pp.3950-3960
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• 2024

As a solution to the problem of spent nuclear fuels (SNFs), the disposal of SNF has gained attention from nations using nuclear energy because of hazards posed to the ecosystem. Among many proposed solutions, the most promising method is to dispose of SNF in a deep geological repository (DGR) which utilizes the multi-barrier concept developed by Finland and Sweden. Here, a new fully-coupled Thermal-Hydraulic (TH) code HADES (High-level rAdionuclide Disposal Evaluation Simulator) is developed using the MOOSE framework. This new code suggests basic numerical tools, such as a non-linear solver and finite element discretization, to assess the safety performance of disposal systems. The new TH code considered various TH behavior using Richards' flow approach, assuming gas pressure is constant. The HADES showed promising results when it was compared to various TH codes validated from DECOVAELX-THMC projects. When the single-canister model was utilized to estimate the TH behavior of the Korean Reference disposal System, although it showed significant saturation reduction due to the evaporation of water, the temperature was maintained under the thermal criteria limit, which is 100 ℃. In addition, the new code estimated temperature and degree of saturation of the multi-canisters model, considering two or three canisters, it showed a slightly lower temperature, 5 ℃, than the single-canister model. From these results, the following are concluded: (1) the new TH code contribute to an additional integrity by estimating TH behavior of KRS; (2) however, due to limitations in single-canister simulation, it is recommended to use multi-canisters simulation to estimate TH behavior accurately. Therefore, this model is anticipated not only to help licensing applications and estimation of various multi-physics phenomena and multi-canister at the disposal site.

• 한국연소학회:학술대회논문집
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• 한국연소학회 2001년도 제22회 KOSCI SYMPOSIUM 논문집
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• pp.90-98
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• 2001

Characteristics of high temperature rocket nozzle flow is discussed along with the aspects of computational analysis. Three methods of nozzle flow analysis, frozen-equilibrium, shifting-equilibrium and non-equilibrium approaches, were discussed, those were coupled with the methods of computational fluid dynamics code. A chemical equilibrium code developed for the analysis of general hydrocarbon fuel was coupled with three approaches of nozzle flow analysis. The approaches were used for the performance prediction of KSR-III Rocket, and compared with the theoretical results from NASA CEA (Chemical Equilibrium with Applications) code.

• Nuclear Engineering and Technology
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• 제50권2호
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• pp.268-279
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• 2018

The ALCYONE multidimensional fuel performance code codeveloped by the CEA, EDF, and AREVA NP within the PLEIADES software environment models the behavior of fuel rods during irradiation in commercial pressurized water reactors (PWRs), power ramps in experimental reactors, or accidental conditions such as loss of coolant accidents or reactivity-initiated accidents (RIAs). As regards the latter case of transient in particular, ALCYONE is intended to predictively simulate the response of a fuel rod by taking account of mechanisms in a way that models the physics as closely as possible, encompassing all possible stages of the transient as well as various fuel/cladding material types and irradiation conditions of interest. On the way to complying with these objectives, ALCYONE development and validation shall include tests on $PWR-UO_2$ fuel rods with advanced claddings such as M5(R) under "low pressure-low temperature" or "high pressure-high temperature" water coolant conditions. This article first presents ALCYONE V1.4 RIA-related features and modeling. It especially focuses on recent developments dedicated on the one hand to nonsteady water heat and mass transport and on the other hand to the modeling of grain boundary cracking-induced fission gas release and swelling. This article then compares some simulations of RIA transients performed on $UO_2$-M5(R) fuel rods in flowing sodium or stagnant water coolant conditions to the relevant experimental results gained from tests performed in either the French CABRI or the Japanese NSRR nuclear transient reactor facilities. It shows in particular to what extent ALCYONE-starting from base irradiation conditions it itself computes-is currently able to handle both the first stage of the transient, namely the pellet-cladding mechanical interaction phase, and the second stage of the transient, should a boiling crisis occur. Areas of improvement are finally discussed with a view to simulating and analyzing further tests to be performed under prototypical PWR conditions within the CABRI International Program. M5(R) is a trademark or a registered trademark of AREVA NP in the USA or other countries.

• Nuclear Engineering and Technology
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• 제41권5호
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• pp.715-722
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• 2009

A model to predict failure of designed-to-fail (dtf) fuel particles is discussed. The dtf fuel under study consisted of a uranium oxycarbide kernel coated with a single pyrocarbon seal coat. Coating failure was assumed to be due to fission gas recoil and knockout mechanisms and direct diffusive release of fission gas from the kernel, which acted to increase pressure and stress in the pyrocarbon layer until it ruptured. Predictions of dtf fuel failure using General Atomics' particle fuel performance code for HRB-17/18 and HFR-B1 irradiation tests were reasonably accurate; however, the model could not predict the failure for COMEDIE BD-1. This was most likely due to insufficient information on reported particle fuel failure at the beginning.

• 설비공학논문집
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• 제21권6호
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• pp.355-366
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• 2009

In the present study, fuel cell driven ground source heat pump system is applied to a large community building and performance of the heat pump system is computationally analyzed. Conduction heat transfer between brine pipe and ground is analyzed by TEACH code to predict the performance of heat pump system. Predicted COP of the heat pump system and the energy cost were compared with variation of the location of the objective building the water saturation rate of soil and the driven powers of heat pump system. Significant reduction of energy cost can be accomplished by employing the fuel cell driven heat pump system in comparison with the late-night electricity driven system. It is due to the low electricity production cost of fuel cell system and the application of recovered waste heat generated during electricity production process to the heating of large community building.

• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2578-2590
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• 2023

In this work, a 24-month two-batch fuel management strategy for the APR1400 using LEU + has been investigated, where enrichments of 5.9 and 5.2 w/o are utilized in lieu of the conventional 4-5 w/o UO₂ fuel. In addition, an Accident Tolerant Fuel (ATF) clad based on the swaging technology is applied to APR1400 fuel assemblies. In this special ATF clad design, both outer and inner SS316 layers protect the conventional zircaloy clad. Erbia (Er₂O₃) is introduced as a burnable absorber with two-fold goals to lower the critical boron concentration in the long-cycle LEU + loaded core as well as to handle the LEU + fuel in the existing front-end fuel facilities without renewing the license. Two types of fuel assemblies with different loading of gadolinia (Gd₂O₃) are considered to control both the reactivity and the core radial power distribution. The erbia burnable absorber is uniformly admixed with UO₂ in all fuel pins except for the gadolinia-bearing ones. In this study, two core designs were devised with different erbia loading, and core performance and safety parameters were evaluated for each case in comparison with a core design without any burnable absorbers. The core analysis was done using the two-step method. First, cross-sections are generated by the SERPENT 2 Monte Carlo code, and the 3-D neutronic analysis is performed with an in-house multi-physics nodal code KANT.

• 전기화학회지
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• 제7권2호
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• pp.69-79
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• 2004

전산모사를 이용하여 특성을 정확하게 모사하기 위해서는 전지 내부에서 발생하는 다양한 물리적, 화학적 현상을 고려하여야 한다. 이를 위해, 본 연구에서는 다양한 전지 내부 현상에 대한 변수를 고려할 수 있는 전산유체 상용코드인 CFD-ACE+를 이용하여 평판형 고체산화물 연료전지의 작동 특성을 분석하였다. 단위 스택에서 발생하는 물질전달과 열전달 및 전기화학 반응에 의한 전하이동을 복합적으로 고려하여, 작동조건 하에서 각 공정적, 구조적 변수 변화에 따른 전지특성을 예측하였다. 이러한 전산모사 방법을 통하여 확산과 유동에 의한 전지 내 반응물과 생성물의 mass fraction 분포와 단위 스택의 내부 온도분포 그리고 전지 특성을 나타내는 polarization curve에 의한 고체산화물 연료 전지의 분극 특성을 정성, 정량적으로 제시하였다. 본 연구를 통해 평판형 단위 스택 내에서의 다양한 변수 변화에 따른 전지의 작동 특성에 대한 효율적 예측이 가능하였고, 고체산화물 연료전지 작동 시 발생하는 현상에 대한 전산모사 접근법을 체계적으로 제시할 수 있었다.

• Nuclear Engineering and Technology
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• 제35권5호
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• pp.378-386
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• 2003

Recently failures of nuclear fuel rods in Korean nuclear power plants were reported and their failure causes have been investigated by using PIE techniques. Destructive and physico-chemical examinations reveal that the clad hydriding phenomena had caused the rod failures primarily and secondarily in each case. In this study, the basic mechanisms of the primary and the secondary hydriding failures are reviewed, PIE data such as cladding inner and outer surface oxide thickness and the restructuring of the fuel pellets are analyzed, and they are compared with the predicted behaviors by a fuel performance code. In addition, post-defected fuel behaviors are reviewed and qualitatively analyzed. The results strongly support that the hydriding processes, primary and secondary, played critical roles in the respective fuel rods failures and the secondary hydriding failure can take place even in the fuel rod with low linear heat generation rate.

• 한국가스학회지
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• 제4권1호
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• pp.40-48
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• 2000

연료전지는 일정하게 유지되는 전극-전해질계의 공정에 의해 연료와 산화제의 화학에너지를 전기에너지로 끊임없이 전환시킬 수 있는 전기화학장치이다. 인산형 연료전지는 전해질로 진한 인산염을 사용한다. 연료전지 시스템에서 가장 중요한 부분인 스택은 연료의 산화가 일어나는 anode, 산화물의 환원이 일어나는 cathode, 그리고 anode와 cathode를 분리시키고 이온을 전도시키는 전해질로 이루어져 있다. 연료전지의 성능은 시스템의 환경에 따른 작동 및 디자인 변수들에 의해 좌우된다. 따라서 연료전지의 핵심부분이라 할 수 있는 스택의 성능향상을 위하여 전산유체역학 코드를 이용한 스택에 대한 3차원적 모델링 및 전기화학반응이 포함된 모사를 수행하였다. 이로부터 산화제의 유량변화에 따른 스택 내부에서의 연료, 산화제 및 생성물의 농도, 그리고 반응에 의해 생성된 열의 전달에 의한 스택의 온도 분포 및 변화를 전산유체 코드인 FLUENT를 이용하여 계산하였다.

• Nuclear Engineering and Technology
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• 제35권4호
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• pp.356-368
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• 2003

Material properties such as coolant specific heat, film heat transfer coefficient, cladding thermal conductivity, surface diffusion coefficient of the multi-bubble are improved in MACSIS-Mod1. The axial power and flux profile module was also incorporated with irradiation history. The performance and feasibility of the updated driver fuel pin have been analyzed for nominal parameters based on the conceptual design for the KALIMER breakeven core by MACSIS-MOD1 code. The fuel slug centerline temperature takes the maximum at 700mm from the bottom of the slug in spite of the nearly symmetric axial power distribution. The cladding mid-wall and coolant temperatures take the maximum at the top of the pin. Temperature of the fuel slug surface over the entire irradiation life is much lower than the fuel-clad eutectic reaction temperature. The fission gas release of the driver fuel pin at the end of life is predicted to be $68.61\%$ and plenum pressure is too low to cause cladding yielding. The probability that the fuel pin would fail is estimated to be much less than that allowed in the design criteria. The maximum radial deformation of the fuel pin is $1.93\%$, satisfying the preliminary design criterion ($3\%$) for fuel pin deformation. Therefore the conceptual design parameters of the driver fuel pin for the KALIMER breakeven core are expected to satisfy the preliminary criteria on temperature, fluence limit, deformation limit etc.