• Nuclear Engineering and Technology
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• 제50권2호
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• pp.259-267
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• 2018

Focused ion beam-scanning electron microscope and electron backscattered diffraction examinations were conducted in the center of a $73\;GWd/t_U\;UO_2$ fuel. They showed the formation of subdomains within the initial grains. The local crystal orientations in these domains were close to that of the original grain. Most of the fission gas bubbles were located on the boundaries. Their shapes were far from spherical and far from lenticular. No interlinked bubble network was found. These observations shed light on previous unexplained observations. They plead for a revision of the classical description of fission gas release mechanisms for the center of high burn-up $UO_2$. Yet, complementary detailed observations are needed to better understand the mechanisms involved.

• 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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• 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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• 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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• 제45권5호
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• pp.581-588
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• 2013

This paper deals with the Safety Analysis for $CANDU^{(R)}$ 6 nuclear reactors as affected by main Heat Transport System (HTS) aging. Operational and aging related changes of the HTS throughout its lifetime may lead to restrictions in certain safety system settings and hence some restriction in performance under certain conditions. A step in confirming safe reactor operation is the tracking of relevant data and their corresponding interpretation by the use of appropriate thermal-hydraulic analytic models. Safety analyses ranging from the assessment of safety limits associated with the prevention of intermittent fuel sheath dryout for a slow Loss of Regulation (LOR) analysis and fission gas release after a fuel failure are summarized. Specifically for fission gas release, the thermal-hydraulic analysis for a fresh core and an 11 Effective Full Power Years (EFPY) aged core was summarized, leading to the most severe stagnation break sizes for the inlet feeder break and the channel failure time. Associated coolant conditions provide the input data for fuel analyses. Based on the thermal-hydraulic data, the fission product inventory under normal operating conditions may be calculated for both fresh and aged cores, and the fission gas release may be evaluated during the transient. This analysis plays a major role in determining possible radiation doses to the public after postulated accidents have occurred.

• 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.

• Nuclear Engineering and Technology
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• 제30권6호
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• pp.541-554
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• 1998

A computer code COSMOS has been developed based on the CARO-D5 for the thermal analysis of LWR UO$_2$ and MOX fuel rod under steady-state and transient operating conditions. The main purpose of the COSMOS, which considers high turnup characteristics such as thermal conductivity degradation with turnup and rim formation at the outer part of fuel pellet, is to calculate temperature profile across fuel pellet and fission gas release up to high burnup. A new mechanistic fission gas release model developed based on physical processes has been incorporated into the code. In addition, the features of MOX fuel such as change in themo-mechanical properties and the effect of microscopic heterogeneity on fission gas release have been also taken into account so that it can be applied to MOX fuel. Another important feature of the COSMOS is that it can analyze fuel segment refabricated from base irradiated fuel rods in commercial reactors. This feature makes it possible to analyze database obtained from international projects such as the MALDEN and RISO, many of which were collected from refabricated fuel segments. The capacity of the COSMOS has been tested with some number of experimental results obtained from the HALDEN, RISO and FIGARO programs. Comparison with the measured data indicates that, although the COSMOS gives reasonable agreement, the current models need to be improved. This work is being performed using database available from the OECD/NEA.

• Nuclear Engineering and Technology
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• 제43권6호
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• pp.477-488
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• 2011

A formulation is proposed for modelling the process of intra-granular diffusion of fission gas during irradiation of $UO_2$ under both normal operating conditions and power transients. The concept represents a simple extension of the formulation of Speight, including an estimation of the contribution of bubble motion to fission gas diffusion. The resulting equation is formally identical to the diffusion equation adopted in most models that are based on the formulation of Speight, therefore retaining the advantages in terms of simplicity of the mathematical-numerical treatment and allowing application in integral fuel performance codes. The development of the new model proposed here relies on results obtained by means of molecular dynamics simulations as well as finite element computations. The formulation is proposed for incorporation in the TRANSURANUS fuel performance code.

• 방사성폐기물학회지
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• 제5권1호
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• pp.39-52
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• 2007

사용후핵연료의 건식 재가공을 위한 핵연료 원격 제조공정중 분말제조를 위한 산화 및 OREOX(산화 환원공정)열처리 공정으로부터 $^{85}Kr$ 및 $^{14}C$ 핵분열기체의 방출거동을 정량적으로 평가하였다. 특히 사용후핵연료의 평균 연소도가 $27,000{\sim}65,000\;MWd/tU$ 범위내에서 연소도 변화에 따른 핵분열기체의 방출 분율은 측정한 실험결과와 ORIGEN 코드로부터 계산된 초기 inventory를 상호 비교하여 구하였다. $500^{\circ}C$ 1차 산화공정(voloxidation)에서 $^{85}Kr$ 및 $^{14}C(^{14}CO_2)$의 시간에 따른 방출거동은 $UO_2$ 핵연료의 $U_3O_8$으로의 분말화 정도와 밀접한 관련이 있는 것으로 보이며, 입계(grain-boundary)에 분포된 핵분열기체가 대부분 방출되는 것으로 여겨진다. 산화분말을 이용한 OREOX 공정으로부터 핵분열기체의 높은 방출율은 $700^{\circ}C$의 환원공정에서 온도 증가에 의한 기체 확산 및 $UO_2$으로의 환원에 의한 U 원자 이동성 증가에 의존하며 주로 inter-grain 및 intra-grain에 분포된 핵분열기체가 방출된 것으로 판단된다. 일차 산화공정시 $^{85}Kr$ 및 $^{14}C$ 핵분열기체의 방출 분율은 핵 연료 연소도가 증가함에 따라 높게 나타났고 방출 분율 범위는 총 inventory의 $6{\sim}12%$정도며, 산화분말의 OREOX 공정처리시 잔류 핵분열기체 대부분이 방출되는 것으로 보인다. 아울러 사용후핵 연료로부터 핵분열기체의 제거를 위해서는 고온 환원분위기보다는 산화에 의한 분말화가 더 효과적인 것으로 여겨진다.

• 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.