• Nuclear Engineering and Technology
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• 제54권8호
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• pp.2771-2782
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• 2022

When assessing the radiological consequences of postulated accident scenarios, it is of primary interest to determine the amount of radioactive fission gas accumulated in the fuel rod free volume. The state-of-the-art semi-empirical approach (ANS 5.4-2010) is reviewed and compared with a mechanistic approach to evaluate the release of radioactive fission gases. At the intra-granular level, the diffusion-decay equation is handled by a spectral diffusion algorithm. At the inter-granular level, a mechanistic description of the grain boundary is considered: bubble growth and coalescence are treated as interrelated phenomena, resulting in the grain-boundary venting as the onset for the release from the fuel pellets. The outcome is a kinetic description of the release of radioactive fission gases, of interest when assessing normal and off-normal conditions. We implement the model in SCIANTIX and reproduce the release of short-lived fission gases, during the CONTACT 1 experiments. The results show a satisfactory agreement with the measurement and with the state-of-the-art methodology, demonstrating the model soundness. A second work will follow, providing integral fuel rod analysis by coupling the code SCIANTIX with the thermo-mechanical code TRANSURANUS.

• Nuclear Engineering and Technology
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• 제55권8호
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• pp.2977-2983
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• 2023

During normal operation, some parts of the fission product in the defective fuel rods can release into the primary loops in PWR and the escape rate coefficients are widely used to assess quantitatively the release behaviors of fission products in the industry. The escape rate coefficients have been standardized and have been validated by some drilling experiments before the 1970s. In the paper, the model to determine the escape rate coefficients of fission products has been established and the typical escape rate coefficients of noble gas and iodine have been deduced based on the measured radiochemical data in one operating PWR. The result shows that the apparent escape rate coefficients vary with the release-to-birth and decay constants for different fission products of the same element. In addition, it is found that the escape rate coefficients from the defective rod with large defects are much higher than the standard escape rate coefficients, i.e., averagely 4.4 times and 1.8 times for noble gas and iodine respectively. The enhanced release of fission products from the severe secondary hydriding of several defective fuel rods in one cycle may lead to the potential risk of the temporary shutdown of the operating reactors.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2395-2407
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• 2022

The sodium-cooled fast reactor is among the innovative nuclear technologies selected in the framework of the development of Generation IV concepts, allowing the irradiation of uranium-plutonium mixed oxide fuels (MOX). A fundamental step for the safety assessment of MOX-fuelled pins for fast reactor applications is the evaluation, by means of fuel performance codes, of the integral thermal-mechanical behaviour under irradiation, involving the fission gas behaviour and release in the fuel-cladding gap. This work is dedicated to the performance analysis of an inner-core fuel pin representative of the ASTRID sodium-cooled concept design, selected as case study for the benchmark between the GERMINAL and TRANSURANUS fuel performance codes. The focus is on fission gas-related mechanisms and integral outcomes as predicted by means of the SCIANTIX module (allowing the physics-based treatment of inert gas behaviour and release) coupled to both fuel performance codes. The benchmark activity involves the application of both GERMINAL and TRANSURANUS in their "pre-INSPYRE" versions, i.e., adopting the state-of-the-art recommended correlations available in the codes, compared with the "post-INSPYRE" code results, obtained by implementing novel models for MOX fuel properties and phenomena (SCIANTIX included) developed in the framework of the INSPYRE H2020 Project. The SCIANTIX modelling includes the consideration of burst releases of the fission gas stored at the grain boundaries occurring during power transients of shutdown and start-up, whose effect on a fast reactor fuel concept is analysed. A clear need to further extend and validate the SCIANTIX module for application to fast reactor MOX emerges from this work; nevertheless, the GERMINAL-TRANSURANUS benchmark on the ASTRID case study highlights the achieved code capabilities for fast reactor conditions and paves the way towards the proper application of fuel performance codes to safety evaluations on Generation IV reactor concepts.

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1929-1934
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• 2022

Besides recent progresses in the physics-based modelling of fission gas and helium behaviour, the scarcity of experimental data concerning their combined behaviour (i.e., cocktail) hinders further model developments. For this reason, in this work, we propose a modelling methodology aimed at providing recommendations for accelerated experimental investigations. By exploring a wide range of annealing temperatures and cocktail compositions with a physics-based modelling approach we identify the most interesting conditions to be targeted by future experiments. To corroborate the recommendations arising from the proposed methodology, we include a sensitivity analysis quantifying the impact of the model parameters on fission gas and helium release, in conditions representative of high and low burnup.

• Nuclear Engineering and Technology
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• 제22권2호
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• pp.116-127
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• 1990

SPEAR-BETA코드에서 사용된 핵분열 기체 방출 모델을 핵연료와 피복관 사이의 갭(gap)과 플레넘(plenum) 사이에서 축방향 핵분열 기체 혼합과 균열된 핵연료에 대한 유효 열전도도를 사용함으로써 개량하여, P$_{max}$와$\Delta$P가 변하는 다양한 출력 감발 조건하에서 핵분열기체 방출 거동을 해석하였다. 핵연료 균열의 영향을 고려한 유효 열전도도는 핵연료의 온도 분포와 내부 기체 압력을 계산하는데 사용되었고, 축방향 기체 유동으로 인한 혼합(mixing)과 회석(dilution)효과는 갭의 폭과 열전도도를 해석하는데에 고려되었다. 축방향 기체 유동 효과를 계산하는데 있어서 계산속도를 빠르게 하기 위하여 유한차분법의 하나인 Crank-Nicholson 방법을 사용하였다. 개량된 모델은 다양한 출력 감발 조건하에서 얻어진 실험 자료들과 SPEAR-BETA와 FEMAXl-IV 코드들에서 사용되는 모델들로부터 얻은 결과들을 비교함으로써 검증하였다. 개량된 모델의 결과는 위의 두 코드로부터 얻은 결과 보다는 실험자료들과 잘 일치하였다. 균열된 핵연료에 대해 유효 열전도도를 사용하여 계산한 핵연료의 중심 온도는 균열되지 않은 핵연료의 경우에 비해 20$0^{\circ}C$ 정도보다 높은 값을 나타냈고, 개량된 핵분열 기체 생성물의 분율은 SPEAR-BETA코드에서 얻은 값보다 평균 6% 정도가 높게 나타났다.평균 6% 정도가 높게 나타났다.다.

• Nuclear Engineering and Technology
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• 제35권1호
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• pp.55-63
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• 2003

Threshold porosity above which fission gas release channels would be formed in the rim egion of high burnup UO$_2$ fuel was estimated by the Monte Carlo method and Hoshen-Kopelman algorithm. With the assumption that both rim pore and rim grain can be represented by cube, pore distribution in the rim was simulated 3-dimensionally by the Monte Carlo method according to porosity and pore size distribution. Then, using the Hoshen-Kopelman algorithm, the fraction of open rim pores interlinked to the outer surface of a fuel pellet was derived as a function of rim porosity. The simulation showed that porosity of 24-25% is the threshold above which the number of rim pores forming release channels increases very rapidly. On the other hand, channels would not be formed if the porosity is less than about 23.5%. This is consistent with the observation that, for porosity less than 23.5%, almost no fission gas is released in the rim. However, once the rim porosity reaches beyond 25%, extensive open paths would be developed and considerable fission gas release would start in the rim.

• 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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• 제43권6호
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• pp.499-508
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• 2011

The paper briefs a fuel performance code, COSMOS, which can be utilized for an analysis of the thermal behavior and fission gas release of fuel, up to a high burnup. Of particular concern are the models for the fuel thermal conductivity, the fission gas release, and the cladding corrosion and creep in $UO_2$ fuel. In addition, the code was developed so as to consider the inhomogeneity of MOX fuel, which requires restructuring the thermal conductivity and fission gas release models. These improvements enhanced COSMOS's precision for predicting the in-pile behavior of MOX fuel. The COSMOS code also extends its applicability to the instrumented fuel test in a research reactor. The various in-pile test results were analyzed and compared with the code's prediction. The database consists of the $UO_2$ irradiation test up to an ultra-high burnup, power ramp test of MOX fuel, and instrumented MOX fuel test in a research reactor after base irradiation in a commercial reactor. The comparisons demonstrated that the COSMOS code predicted the in-pile behaviors well, such as the fuel temperature, rod internal pressure, fission gas release, and cladding properties of MOX and $UO_2$ fuel. This sufficient accuracy reveals that the COSMOS can be utilized by both fuel vendors for fuel design, and license organizations for an understanding of fuel in-pile behaviors.

• 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.

• Nuclear Engineering and Technology
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• 제41권2호
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• pp.155-162
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• 2009

Since the 90's, EDF and AREVA-NP have irradiated, up to very high burnups, lead assemblies housing $M5^{(R)}$ cladded fuels. Post-irradiation examination of high burnup $UO_2$ pellets show an increase in the fission-gas release rate, an increase in fuel swelling, and formation of fission-gas bubbles throughout the pellets. Xenon abundances were quantified, and phenomena leading to this bubble formation were identified. All examinations provided valuable data on the complex state of the fuel during irradiation. They show the good behavior of these fuels, exhibiting various microstructures at very high burnups, none of which is likely to lead to problems during irradiation.