• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(2)
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• pp.707-712
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• 1995

Grain boundary diffusion plays significant role in the fission gas release, which is one of the crucial processes dominating nuclear fuel performance. Gaseous fission products such as Xe and Kr generated inside fuel pellet have to diffuse in the lattice and in the grain boundary before they reach open space in the fuel rod. In the mean time, the grains in the fuel pellet grow and shrink according to grain growth kinetics, especially at elevated temperature at which nuclear reactors are operating. Thus the boundary movement ascribed to the grain growth greatly influences the fission gas release rate by lengthening or shortening the lattice diffusion distance, which is the rate limiting step. Sweeping fission gases by the moving boundary contributes to the increment of the fission gas release as well. Lattice and grain boundary diffusion processes in the fission gas release can be studied by 'tracer diffusion' technique, by which grain boundary diffusion can be estimated and used directly for low burn-up fission gas release analysis. However, even for tracer diffusion analysis, taking both the intragranular grain growth and the diffusion processes simultaneously into consideration is not easy. Only a few models accounting for the both processes are available and mostly handle them numerically. Numerical solutions are limited in the practical use. Here in this paper, an approximate analytical solution of the lattice and stationary grain boundary diffusion in a polycrystalline solid is developed for the tracer diffusion techniques. This short closed-form solution is compared to available exact and numerical solutions and turns out to be acceptably accurate. It can be applied to the theoretical modeling and the experimental analysis, especially PIE (post irradiation examination), of low burn up fission. gas release.

• Nuclear Engineering and Technology
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• 제23권4호
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• pp.444-455
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• 1991

본 연구에서는 CANDU-PHWR 형 기존 및 개량 핵연료의 원통형 (soild) 및 환상형 소결체에 대하여, 그 핵연료 전 수명 기간동안, 반경방향 출력분포를 정확하고 신속하게 계산하는 NEDAR 모형을 개발하였다. 본 계산모형에는 핵연료소결체의 직경 범위 8.0-19.5 mm, 농축도 범위 0.71-6.0 wt % U-235이고, 계산 가능 연소도범위가 0-840 Mwh/kgU (35000MWD/T)인 한계내에서, 핵연료 반경방향 출력분포결자식 및 열중성자속감소 계산결과자료가 포함되어 있다. CAN-DU-PHWR 형 원자로 중성자속 스펙트럼을 입력자료로 하여, 로물리 전산코드, CE-HAMMER 를 이용하여 핵연료의 각 설계조건 및 소결체의 환별 국부지점에 대하여, 임의로 설정한 기준 연소시점에서 반경 방향 출력 분포를 계산하였다. 이 계산 결과를 토대로 각 환의 평균출력을 구하는 적분법 및 비선형 곡선희귀계산법에 의하여, Bessel 함수와 지수함수의 다항식으로 구성된 반경방향 출력분포 기본 결과식 및 그 계수들이 산출되었다. 본 연구에서 개발된 NEDAR 모형을 이용하여 산출한 반경방향출력분포값을, 핵연료소결체 표면에서의 값을 기본단위로 환산하여 비교하면, 본 의형에 의한 반경방향 출력분포 결과가 기존 ELESIM 전산코드의 결과에 비교하여 약간 높게 나타났다. 소결체의 반경방향의 출력 및 온도분포는 핵분열기체생성물방출과 밀접한 관계가 있으므로, 본 모형을 기존 ELESIM 전산코트의 반경방향 출력분포 계산 모형과 대체한 전산코트, 즉 KAFEPA-NEDAR에 의한 핵분열기체생 생성물방출량 예측치를 기존 ELESIM 전산코드의 예측치와 비교하였다. 여기서 KAFEPA-NEDAR리 예측치가 실험결과 자료에 보다 더 가깝게 접근하였다. 따라서, 본 연구에서 개 발된 NEDAR모형은 과대한 계산시간의 낭비없이 CANDU-PHWR 형 핵연료소결체의 반경방향출력분포를 효율적이고, 신속/정착하게 계산하는 모형임이 입증되었다.

• Nuclear Engineering and Technology
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• 제26권1호
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• pp.90-99
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• 1994

손상 핵연료에서 발생하는 주요한 현상중의 하나는 수중기의 분해로 갭에 존재하는 산소에 의해 $UO_2$ 가 $UO_{2+}$x/로 산화되고, 이로 인해 결정립내에서의 핵분열기체 확산계수가 증가하여 결과적으로 핵분열 기체의 방출이 증대하는 현상이다. 본 논문은 일반적인 원자로 운전 조건하에서 원자로 및 손상 핵연료의 운전조건을 고려하여 소결체의 산화거동을 모사하고 이를 바탕으로 소결체 산화가 핵분열기체의 방출 중대에 미치는 영향을 분석하였다. 소결체 산화거동은 갭에는 150기압의 포화된 순수한 수증기만이 존재한다는 가정하에 분석하였고, 산화에 의한 핵분열기체의 방출 증대 효과를 정량적으로 분석하기 위해 방출중대 인자를 도입하였다. 실험 치와 비교한 결과 방출증대 인자는 소결체 산화에 의한 핵분열기체의 방출증대 효과를 잘 예측하였다.

• Nuclear Engineering and Technology
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• 제54권4호
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• pp.1464-1470
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• 2022

This paper presents a high performance burnable absorber named as CIMBA (Cylindrically Inserted and Mechanically Separated Burnable Absorber) for the soluble-boron-free SMR. The CIMBA is the cylindrical gadolinia inserted into the annular fuel pellets. Although the CIMBA utilizes the spatial self-shielding effect of the fuel material, a large reactivity upswing occurs when the gadolinia is depleted. To minimize the reactivity swing of the CIMBA-loaded FA, two approaches were investigated. One is controlling the spatial self-shielding effect of the CIMBA as burnup proceeds by a multi-layered structure of the CIMBA (ML-CIMBA) and the other is the mixed-loading of two different types of CIMBA (MIX-CIMBA). Both approaches show promising performances to minimize the reactivity swing, where the MIX-CIMBA is more preferable due to its simpler fabrication process. In conclusion, the MIX-CIMBA is expected to accelerate the commercialization of the CIMBA and can be used to achieve an optimal soluble-boron-free SMR core design.

• 대한기계학회논문집A
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• 제40권5호
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• pp.437-447
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• 2016

원자력 발전소의 반응로에 핵연료 봉으로 이루어진 집합체가 있으며 핵 연료의 연소를 통한 열을 이용하여 발전을 한다. 핵연료 봉은 핵연료와 그를 감싸는 피복관으로 이루어졌으며 연소되는 동안 서로의 상호작용에 대한 해석은 안전성을 평가함에 있어 중요한 사실이다. 본 논문에서는 핵연료와 피복관의 연소 상태에서 기계적 상호작용에 대한 해석 방법에 대하여 제시한다. 온도 해석에 있어서 핵연료와 간극 사이에서의 열전도도가 중요하며 간극 거리와 접촉여부에 따른 접촉 압력이 또한 중요 요소이다. 이에 간극 열전도도는 비결정론적이기 때문에 이를 해결할 수 있는 방법에 대하여 제시했다. 핵 연료의 열팽창에 따른 피복관과의 접촉을 해결하기 위한 계산을 수행하였고 그에 따라 접촉 시 발생하는 응력이 항복함수를 넘어 소성 변형이 일어날 경우 또한 고려하였다. 핵연료의 열팽창에 따라 피복관과 접촉에 의한 소성 변형을 해석하므로 핵연료 봉의 안정성을 평가할 수 있다. 이를 적용하기 위해 3차원 유한요소 모듈을 FORTRAN90을 이용하여 개발하였다. 핵연료와 피복관의 접촉에 의한 탄소성 변형을 주로 다루며 두꺼운 실린더를 통한 간단한 이론 모델을 제시하여 코드에 대해 검증을 실시하였다.

• Nuclear Engineering and Technology
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• 제36권4호
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• pp.338-345
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• 2004

This study examines the sensitivity of several fabrication parameters for dry process fuel, using a random sampling technique. The in-pile performance of dry process fuel with irradiation was calculated by a modified ELESTRES code, which is the CANDU fuel performance code system. The performance of the fuel rod was then analyzed using a Monte Carlo simulation to obtain the uncertainty of the major outputs, such as the fuel centerline temperature, the fission gas pressure, and the plastic strain. It was proved by statistical analysis that for both the dry process fuel and the $UO_2$ fuel, pellet density is one of the most sensitive parameters, but as for the fission gas pressure, the density of the $UO_2$ fuel exhibits insensitive behavior compared to that of the dry process fuel. The grain size of the dry process fuel is insensitive to the fission gas pressure, while the grain size of the $UO_2$ fuel is correlative to the fission gas pressure. From the calculation with a typical CANDU reactor power envelop, the centerline temperature, fission gas pressure, and plastic strain of the dry process fuel are higher than those of the $UO_2$ fuel.

• Nuclear Engineering and Technology
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• 제50권1호
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• pp.80-87
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• 2018

Consideration is given to the possibility to use changes in buoyancy as a negative reactivity feedback mechanism during temperature transients in heavy liquid metal fast reactors. It is shown that by the proper use of heavy pellets in the fuel elements, fuel rods could be endowed with a passive self-ejection mechanism and then with a negative feedback. A first estimate of the feasibility of the mechanism is calculated by using a simplified geometry and model. If in addition, a neutron poison pellet is introduced at the bottom of the fuel, then when the fuel element is displaced upward by buoyancy force, the reactivity will be reduced not only by disassembly of the core but also by introducing the neutron poison from the bottom. The use of induced buoyancy opens up the possibility of introducing greater amounts of actinides into the core, as well as providing a palliative solution to the problem of positive coolant temperature reactivity coefficients that could be featured by the heavy liquid metal fast reactors.

• 한국분말야금학회:학술대회논문집
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• 한국분말야금학회 2005년도 추계학술강연 및 발표대회강연 및 발표논문 초록집
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• pp.25-26
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• 2005

• Nuclear Engineering and Technology
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• 제49권6호
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• pp.1287-1300
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• 2017

Based on the observation that ignoring the angle dependency of multigroup resonance cross sections within a fuel pellet would result in nontrivial underestimation of the spatial self-shielding of flux, a parametrized spectral superhomogenization (SPH) factor library (PSSL) method is developed as a practical means of resolving the problem. Region-wise spectral SPH factors are calculated by the normal and transport corrected SPH iterations after ultrafine group slowing down calculations over various light water reactor pin-cell configurations. The parametrization is done with fuel temperature, U-238 number density, fuel radius, moderator source represented by ${\Sigma}_{mod}V_{mod}$, and the number density ratio of resonance nuclides to that of U-238 in a form of resonance interference correction factors. The parametrization is successful in that the root mean square errors of the interpolated SPH factors over the fuel regions of various pin-cells are within 0.1%. The improvement in reactivity error of the PSSL method is shown to be superior to that by the original SPH method in that the reactivity bias of -200 pcm to -300 pcm vanishes almost completely. It is demonstrated that the environment effect takes only about 4% in the reactivity improvement so that the pin-cell based PSSL method is effective in the assembly problems.

• Nuclear Engineering and Technology
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• 제55권12호
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• pp.4431-4446
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• 2023

A Multiphysics coupling framework, MPCORE, has been developed to analyze safety parameters using the best estimate codes. The framework contains neutron kinetics (NK), thermal hydraulics (TH), and fuel performance (FP) codes to analyze fuel burnup, radial power distribution, and coolant temperature (T_bc). Shuffling and rotation capabilities have been verified on the Watts Bar reactor for three cycles. This study focuses on two coupling approaches for TH and FP modules. The one-way coupling approach involves coupling the FP code with the NK code, providing no data to the TH modules but getting T_bc as boundary condition from TH module. The two-way coupling approach exchanges information from FP to TH modules, so that the simplified heat conduction solver of the TH module is not used. The power profile in both approaches does not differ significantly, but there is an impact on coolant and cladding parameters. The one-way coupling approach tends to over-predict the cladding hydrogen concentration (CHC). This research highlights the difference between one-way and two-way coupling on critical boron concentration, T_bc, CHC, oxide surface temperature, and pellet centerline temperature. Overall, MPCORE framework with two-way coupling provides a more accurate and reliable analysis of safety parameters for nuclear reactors.