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Research and validation on the numerical algorithm of mechanical module in a transient fuel rod performance code for fast reactor

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ANNALS OF NUCLEAR ENERGY 2022年 171卷 108991-108991页

作者： Yang, Guangliang Guo, Zixuan Wang, Peng Wang, Weixiang Liao, Hailong Ding, Tao Chen, Hongli Univ Sci & Technol China Sch Nucl Sci & Technol Hefei 230026 Peoples R China Nucl Power Inst China Sci & Technol Reactor Syst Design Technol Lab Chengdu 610213 Peoples R China

Mechanical performance is an important aspect for the fuel rod performance in LMFR. Operating conditions in LMFR, like large power density, high fast neutron flux and deep burnup, are hash for the fuel rod. The evolution of fuel rod behaviors is complicated and tightly coupled with each other through temperature, stress and burnup. Considering various in-pile behaviors of the fuel rod under fast neutron irradiation, the numerical iterative algorithm based on grid integration was established to deal with the complex nonlinear mechanical problem for the fuel rod in LMFR. Verification and validation have been carried out by comparing results from this algorithm with analytical solutions as well as results from typical code, which are all in good agreement with each other, proving the accuracy and reliability of this algorithm. CO 2022 Elsevier Ltd. All rights reserved.

关键词： In-pile behaviors Numerical algorithm Mechanical analysis fuel rod performance code LMFR

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FUAS: An uncertainty and sensitivity analysis toolkit for fuel rod performance code

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ANNALS OF NUCLEAR ENERGY 2022年 173卷

作者： Wei, Jun Wang, Yang Song, Zi-Fan Wang, Jie Liu, Xin Deng, Yong-Jun China Nucl Power Technol Res Inst Co Ltd Shang Bu Middle Rd Shenzhen 518031 Guangdong Peoples R China

fuel rod performance codes are used to predict the behavior of fuel rods in reactors. The uncertainty and sensitivity analysis are always needed when developing or using these codes. However, fuel rod codes typically have two characteristics that make relevant analysis challenging: First, there are many input parameters for which uncertainty needs to be considered. Second, the structure of the models is complex, non-linear and non-additive. To make the uncertainty and sensitivity analysis of a fuel rod code effectively and conveniently, an assistant toolkit, FUAS, has been developed recently by China Nuclear Power Technology Research Institute Co. Ltd. FUAS considers the fuel rod codes as black boxes and applies sampling-based approaches. Several statistical methods are available in this toolkit. A hypothetical steady-state scenario of the PWR fuel rod was simulated, and uncertainty and sensitivity analysis were performed with the sampling-based approach by coupling the fuel rod code and FUAS toolkit. (C) 2022 Elsevier Ltd. All rights reserved.

关键词： Uncertainty analysis Sensitivity analysis fuel rod performance code Sampling-based method Monte Carlo method

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Under irradiation issues of the CSZ-based inert matrix fuels from IFA-652 Halden experiment

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PROGRESS IN NUCLEAR ENERGY 2010年第4期52卷 352-358页

作者： Calabrese, R. Vettraino, F. Tverberg, T. ENEA Innovat Nucl Reactors & Fuel Cycle Closure Div I-40129 Bologna Italy OECD Halden Reactor Project N-1751 Halden Norway

The renewal of interest in Fast Reactors (FR) deployment as possible answer to the future perspective of natural uranium shortage has led to reconsider the plutonium as an energy source with a consequent radical change in its management. The transition to a closed fuel cycle is estimated to occur in a period ranging from 2030 to 2050. In the near term, Inert Matrix fuels (IMF) are currently viable options, capable to contribute in reducing the proliferation risk associated to nowadays separated plutonium stockpiles in a business-as-usual nuclear energy scenario where a prominent role is still played by Light Water Reactors (LWR) in a once-through fuel cycle strategy. Double-strata strategy with Accelerator Driven Systems (ADS) machines to transmute minor actinides also support the hypothesis of inert matrix heterogeneous fuels. These reasons justify the current interest in IMF A high burning efficiency achieved by preventing new plutonium build-up under irradiation (U-free fuel), a proved high radiation damage and leaching resistance are fundamental requirements when a once-through fuel cycle strategy is planned. Amongst other options, Calcia-Stabilised Zirconia (CSZ) and Yttria-Stabilised Zirconia (YSZ) fulfil these criteria standing as the most promising matrices to host plutonium. ENEA has conceived an in-pile testing (IFA-652 experiment) of viable inert matrices concepts: CSZ and thoria. This experiment was performed in the Halden Heavy Boiling Water Reactor (HBWR), a joint project of the OECD Nuclear Energy Agency. The test-rig is a six-rod bundle loaded with IM, IMT and T innovative fuels. IM and T fuels have, respectively, CSZ and thoria matrix, the fissile phase is High Enriched Uranium (HEU) oxide (UO2 93% U-235 enriched). IMT is a ternary fuel composed of CSZ + thoria matrix and HEU oxide as fissile phase. Thoria is added in IMT fuel mainly to improve the low IM reactivity feedback coefficients. The discharge burnup, under typical LWR conditions, was, for

关键词： Inert Matrix fuel Fission Gas Release fuel rod performance code

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TESTING OF TRANSURANUS code FOR RIA ANALYSIS: THE FK-1 NSRR CASE

TESTING OF TRANSURANUS CODE FOR RIA ANALYSIS: THE FK-1 NSRR ...

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18th International Conference on Nuclear Engineering (ICONE-18)

作者： Calabrese, R. Vettraino, F. Innovat Nucl Reactors & Fuel Cycle Closure Div I-40129 Bologna Italy

ISBN: (纸本)9780791849293

Reactivity Initiated Accident (RIA) leads to an unwanted increase in fission rate and power in a region of the reactor core confined around the position of occurrence. The power excursion due to such events may cause fuel rods failures and a subsequent release of radioactive material into the primary coolant of reactor, in severe cases, this release could damage nearby fuel assemblies. In nuclear power plants, RIAs are due to control system faults, e. g. control elements ejection/insertion, or rapid changes in temperature or pressure of moderator. In Boiling Water Reactors (BWRs), the control rod drop accidents (RDAs) at cold zero power have been deeply investigated, in fact, notwithstanding they are less frequent in comparison with the control rod ejection event in PWRs, in this kind of plant these conditions are the most severe in case of a RIA occurrence. RDA transient, comprised in the design basis events considered in safety analysis, may cause rod failures especially at high burnup. To simulate a RIA, a peaked power pulse is applied to a pre-irradiated and re-instrumented rodlet aiming at investigating the most important phenomena that could lead to the rupture of cladding tubes. This paper is focused on the investigation of the TRANSURANUS fuel performance code capability to predict the thermomechanical state of rodlets subjected to RIA tests. To this purpose the FK-1 test, carried out at the Nuclear Safely Research Reactor (NSRR) of Japan Atomic Energy Research Institute (JAERI), was simulated. This experiment is part of a set of 12 tests performed at the NSRR facility to study the performance under a reactivity initiated accident of BWR rodlets with burnup between 41 and 61 MWd/kg(HM). In the FK-1 test, a STEP I BWR rodlet, previously irradiated in the Fukushima Daiichi Nuclear Power Station (Unit 3) operated by the Tokyo Electric Power COmpany (TEPCO) up to 45 MWd/kg(HM), was subjected to a peak enthalpy insertion of 544 J/g. In this paper the code finding

关键词： Reactivity Initiated Accident fuel rod performance code

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INERT MATRIX fuelS ANALYSIS BY MEANS OF THE TRANSURANUS code: THE HALDEN IFA-652 IN-PILE TEST

INERT MATRIX FUELS ANALYSIS BY MEANS OF THE TRANSURANUS CODE...

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16th International Conference on Nuclear Engineering

作者： Calabrese, R. Vettraino, F. Tverberg, T. ENEA Innovat Nucl Reactors & Fuel Cycle Div I-40129 Bologna Italy

ISBN: (纸本)9780791848159

Inert matrix fuels are a possible option to reduce separated plutonium stockpiles by burning it in LWR fleet. A high burning efficiency targeted by preventing new plutonium build-up under irradiation (U-free fuel), a proved high radiation damage and leaching resistance are fundamental requirements when a once-through fuel cycle strategy is planned. Among other options, both calciastabilised zirconia (csz) and thoria fulfill these criteria standing as the most promising matrices to host plutonium. While several in-pile tests concerning thoria fuels are found, calcia-stabilised zirconia under-irradiation performance is still to be fully assessed, with this regard the thermal conductivity, markedly lower than UOX and MOX cases, plays a fundamental role. For this reason, ENEA has conceived a comparative in-pile testing of three different U-free inert matrix fuel concepts, that have been performed in the OECD Holden HBWR (IFA-652 experiment). The discharge burnup accomplished about 90-97% of the 45 MWd/kgU(eq) target under typical LWR irradiation conditions. The lest-rig is a six-rod bundle loaded with IM, IMT and T innovative fuels. IM and T fuels have, respectively, csz and thoria as matrix, the fissile phase being HEU oxide (UO(2) 93% (235)U enriched). IMT is a ternary fuel composed by csz+thoria matrix and HEU oxide as fissile phase. Thoria is added in IMT fuel to improve the low IM reactivity feedback coefficients. Pins are instrumented providing fuel centerline temperature, pin inner pressure and fuel slack elongation measurements. Our purpose is to investigate the key processes of IMF under irradiation behaviour by means of the TRANSURANUS code. Thermal conductivity and its degradation with burnup, densification-swelling response and FGR are tentatively modelled in the burnup domain of IFA-652. In particular it is pointed out the effects of pellet geometry and fuel microstructures in the IM and IMT cases. The consistency of our results is discussed aiming at under

关键词： Inert Matrix fuel fuel rod performance code

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COMPARATIVE STUDY OF INERT MATRIX fuelS CONCEPTS FROM IFA-652 EXPERIMENT

COMPARATIVE STUDY OF INERT MATRIX FUELS CONCEPTS FROM IFA-65...

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17th International Conference on Nuclear Engineering

作者： Calabrese, R. Vettraino, F. Tverberg, T. ENEA Innovat Nucl Reactors & Fuel Cycle Closure Div I-40129 Bologna Italy

ISBN: (纸本)9780791843550

Notwithstanding Mixed OXide (MOX) fuel is considered the most probable option to reduce the separated plutonium stockpiles, in advance on a significant deployment of Fast Breeder Reactors (FBR), inert matrix fuels (IMF) remain a possible way to accomplish this task in nowadays Light Water Reactors (LWR) fleet. A high burning efficiency achieved by preventing new plutonium build-up under irradiation (U-free fuel), a proved high radiation damage and leaching resistance are fundamental requirements when a once-through fuel cycle strategy is planned. Amongst other options, both calcia-stabilised zirconia (csz) and thoria fulfill these criteria standing as the most promising matrices to host plutonium. The investigation of the under-irradiation performance of the calcia-stabilised zirconia fuels is underway, with this regard the thermal conductivity, lower than Uranium OXide (UOX) and MOX fuels, has proved to play a fundamental role. For this reason, ENEA has conceived a comparative in-pile testing (IFA-652 experiment) of viable U-free inert matrix fuel concepts that has been performed in the Halden Heavy Boiling Water Reactor (HBWR) of the Organisation for Economic Co-operation and Development (OECD). The accumulated burn up, under typical LWR conditions, at the end of the irradiation history, was about 90% of the foreseen 45 MWd/kgU(eq). The test-rig is a six-rod bundle loaded with IM, IMT and T innovative fuels. IM and T fuels have, respectively, csz and thoria matrix, the fissile phase is High Enriched Uranium (HEU) oxide (UO2 93% U-235 enriched). I M is a ternary fuel composed of csz+thoria matrix and HEU oxide as fissile phase. Thoria is added in IMT fuel to improve the low IM reactivity feedback coefficients. Pins are instrumented providing fuel centreline temperature, pin inner pressure and fuel stack elongation measurements. 39 This paper is focused on the csz-based IMF under irradiation performance, especially regarding the Fission Gas Release (FGR) topic. Low

关键词： Inert Matrix fuel FGR fuel rod performance code

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Modelling of stable fission gases diffusion in UO₂ nuclear fuel

Modelling of stable fission gases diffusion in UO<sub>2</sub...

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3rd International Conference on Diffusion in Solids and Liquids

作者： Vallejo, I. Herranz, L. E. CIEMAT Dept Energy Nucl Fiss Div E-28040 Madrid Spain

The current trend in nuclear industry to extend the burnup, limits up to high burnup requires the analysis of the phenomenology related to the nuclear fuel, specifically the diffusion and fission gas release (FGR) related to this phenomenon plays an important role. In this paper, the diffusion of stable fission gases (xenon and krypton) in nuclear fuel UO2 is addressed through the predictions for FGR to rod void volumes obtained with FRAPCON-3 fuel performance code. The theoretical base of diffusion model in FRAPCON-3 code is shown and some modifications are proposed, such as the removing of empirical correlations not related to diffusion phenomenon and fitting parameters included in the model. Besides, the resolution process in the proximities of grain boundaries is considered in a different way, and the grain growth mechanism from a specific temperature threshold is implemented into the code, The database applied for evaluation is presented and the results with the original and modified model are shown.

关键词： diffusion fission gas release nuclear fuel fuel rod performance code

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