Application of the TRANSURANUS code for the fuel pin design process of the ALFRED reactor

Luzzi, L.; Cammi, Antonio; Marcello, V. Di; Lorenzi, Stefano; Pizzocri, D.; Uffelen, P. Van

doi:10.1016/j.nucengdes.2014.06.032

Cited by 25 publications

(11 citation statements)

References 21 publications

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“…As the first safety barrier, the properties and behaviors of the claddings obtain much more attention in fuel performance analysis and design evaluation, so creep, swelling and rupture are essential in the modeling. These related computational expressions are also used in the TRANSURANUS code (Luzzi et al, 2014) and reported below.…”

Section: Claddingmentioning

confidence: 99%

Preliminary Multi-Physics Performance Analysis and Design Evaluation of UO2 Fuel for LBE-Cooled Subcritical Reactor of China Initiative Accelerator Driven System

Wang

Yun

2021

Front. Energy Res.

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The China initiative Accelerator Driven System (CiADS) and the corresponding lead-bismuth eutectic (LBE) cooled subcritical reactor, as the research subject of one of the major national science and technology infrastructure projects, are undertaken by the Institute of Modern Physics-Chinese Academy of Sciences (IMP-CAS). And in the first phase, UO2 fuels will be loaded in the subcritical core to test the coupling technology and achieve a long-term steady operation. A brief description of CiADS subcritical reactor, fuel assembly and fuel element are presented here, and a multi-physics performance analysis and design evaluation of CiADS UO2 fuel are carried out by means of the FUTURE code. FUTURE is a fuel performance analysis code to evaluate the synergy of phenomena occurring in the fuel element and their impact on the fuel design improvement for the liquid metal fast reactor, which was developed jointly by IMP-CAS and Xi’an Jiaotong University (XJTU). In this paper, the FUTURE code was modified and updated focusing on characteristics of CiADS fuels. Relocation and densification models were added. Results of the hottest fuel element, mainly concerning the thermo-mechanical behaviors, are discussed concerning both fuel and cladding performance on the basis of indicative design limits. According to the preliminary design, the CiADS UO2 fuel exhibits good performance, and the main safety parameters are far below the indicative limits. The Fuel Cladding Mechanical Interaction (FCMI) is not very serious, and the permanent cladding strains and Cumulative Damage Fraction (CDF) are small and even negligible thanks to the low level of fuel temperature and corresponding stress. However, some critical issues may still exist, especially on LBE corrosion near the coolant inlet, where protective oxide layers are very thin from BoL to EoL. The modeling is useful for providing feedback to the conceptual design of the CiADS LBE-cooled subcritical reactor and the update of FUTURE code.

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Section: Claddingmentioning

confidence: 99%

Preliminary Multi-Physics Performance Analysis and Design Evaluation of UO2 Fuel for LBE-Cooled Subcritical Reactor of China Initiative Accelerator Driven System

Wang

Yun

2021

Front. Energy Res.

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“…The difference in creep resistance may originate from material properties, for example from the amount of cold work. The amount of cold work was 24% for the 15-15Ti alloy specimen tested in this study, but the amount of cold work for the material used by Filacchioni and Gavoille was not presented in the references (Luzzi et al, 2014;Filacchioni et al, 1990;Gavoille et al, 2013). (Luzzi et al, 2014;Filacchioni et al, 1990;Gavoille et al, 2013) At the second stage, a test with internal pressure and additional axial load was performed for a cold worked E110 specimen with outer diameter of 9.10 mm and a wall thickness of 0.70 mm.…”

Section: Figure 3 Hoop and Axial Strains As A Function Of Time With mentioning

confidence: 82%

“…For creep strain rate evaluation the Többe correlation was adopted to describe the correlation between effective stress and effective creep strain rate (Luzzi et al, 2014). The Többe correlation is based on the Nabarro-Herring equation, which is different from the typical Arrhenius equations used in the creep strain rate modelling (Többe, 1975).…”

Section: Figure 3 Hoop and Axial Strains As A Function Of Time With mentioning

confidence: 99%

Multiaxial Creep Testing Device for Nuclear Fuel Claddings

Pohja¹,

Moilanen²,

Rantala³

et al. 2016

AJS

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Mechanical properties and creep resistance are among the key issues in the selection of fuel cladding materials for future Gen-IV nuclear reactor applications. As for today's nuclear reactors, the Fukushima accident has increased the interest in accident-tolerant fuel cladding materials. Thin-walled fuel cladding tubes may exhibit anisotropic creep properties due to the manufacturing process. A newly developed Pneumatic Loading Apparatus (PLA) is capable of testing the multiaxial creep properties of thin-walled tubular specimens with internal pressure and an additional axial force. Because of its innovative design, the PLA is moveable and can be implemented in testing in extreme environments. The system functionality and accuracy was demonstrated by two biaxial creep tests performed on austenitic stainless steel DIN 1.4970 and E110 zirconium alloy specimens. Both materials showed anisotropic creep behavior and the PLA testing concept and equipment proved its capability to perform reliably and accurately biaxial creep testing of fuel cladding materials.

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“…For each radial zone, Eq. ( 30) is implemented and discretized in Z nodes, specifying the material properties [27] and whether it is a fuel zone or not. In particular, if it is a fuel zone, the power Q 0 0 0 is provided from the MN component through a connector (Fig.…”

Section: Fuelpinht Componentmentioning

confidence: 99%

A Control-Oriented Modeling Approach to Spatial Neutronics Simulation of a Lead-Cooled Fast Reactor

Lorenzi

Cammi

Luzzi

et al. 2015

Journal of Nuclear Engineering and Radiation Science

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The most diffused neutronics modeling approach in control-oriented simulators is pointwise kinetics. In the framework of developing control strategies for innovative reactor concepts, such a simplified description is less effective as it prevents the possibility of exploiting the capabilities of advanced control schemes. In the present work, in order to overcome these limitations, a spatial neutronics description based on the modal method has been considered. This method allows separating the spatial and time dependence of the neutron flux, which can be represented as the sum of the eigenfunctions of the neutron diffusion equation weighted by time-dependent coefficients. In this way, the system dynamic behavior is reduced to the study of these coefficients and can be represented by a set of ordinary differential equations (ODEs), reducing the simulation computational burden. In this paper, a test case involving three fuel pins of an innovative lead-cooled fast reactor has been set up and investigated. Once the eigenfunctions are obtained, the set of ODEs for studying the time-dependent coefficients has been derived and then implemented in the DYMOLA environment, developing an object-oriented component based on the reliable, tested, and well-documented Modelica language. In addition, a heat transfer model for the fuel pin has been developed, still drawing on the principles of the object-oriented modeling. Finally, in order to assess the performance of the\ud developed spatial neutronics component, the outcomes have been compared with the reference results obtained from the multigroup diffusion partial differential equations, achieving a satisfactory agreement

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Application of the TRANSURANUS code for the fuel pin design process of the ALFRED reactor

Cited by 25 publications

References 21 publications

Preliminary Multi-Physics Performance Analysis and Design Evaluation of UO2 Fuel for LBE-Cooled Subcritical Reactor of China Initiative Accelerator Driven System

Preliminary Multi-Physics Performance Analysis and Design Evaluation of UO2 Fuel for LBE-Cooled Subcritical Reactor of China Initiative Accelerator Driven System

Multiaxial Creep Testing Device for Nuclear Fuel Claddings

A Control-Oriented Modeling Approach to Spatial Neutronics Simulation of a Lead-Cooled Fast Reactor

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