Toward UO2 micro/macro machining: a laser processing approach

Linked to experimental data acquisition and to development of improved models, a better detailed description of the behaviour of the nuclear ceramics as regard to the fission gases release during thermal transient representative of nuclear accidents such as RIA (Reactivity Initiated Accident) and or LOCA (LOss of Coolant Accident) requires access to local information within the fuel pellet, and no longer averaged over the whole of the pellet. One of the major challenge in this context is the sample size, which depends on the main objective of the study, typically from the order of a few hundred microns to millimeters. Few techniques allow this dynamic while being compatible with irradiated fuel constraints. Laser micromachining is a high precision non-contact material removal process that would be adapted to this dynamic. We present experimental and numerical studies, carried out in order to evaluate the possibility to apply this process for the preparation of irradiated UO2 samples of various dimensions. First, preliminary experimental and numerical works conduced on graphite, as model material, which have comparable properties (in particular their behaviours under laser irradiation and their melting point) in order to validate the feasibility, will be detailed. Afterwards, based on these results, we present our first results on UO2. The objective is to transfer the technique to non-irradiated UO2 and then to the irradiated material.

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Fuel R&D Needs and Strategy towards a Revision of Acceptance Criteria

Science and Technology of Nuclear Installations ◽

10.1155/2010/646971 ◽

2010 ◽

Vol 2010 ◽

pp. 1-7 ◽

Cited By ~ 4

Author(s):

François Barré ◽

Claude Grandjean ◽

Marc Petit ◽

Jean-Claude Micaelli

Keyword(s):

Loss Of Coolant Accident ◽

Safety Margins ◽

Power Load ◽

Higher Power ◽

Loss Of Coolant ◽

Advanced Fuel ◽

Water Reactors ◽

Improved Performance ◽

Irradiated Fuel ◽

Cladding Material

The study of fuel behaviour under accidental conditions is a major concern in the safety analysis of the Pressurised Water Reactors. The consequences of Design Basis Accidents, such as Loss of Coolant Accident and Reactivity Initiated Accident, have to be quantified in comparison to the safety criteria. Those criteria have been established in the 1970s on the basis of experiments performed with fresh or low irradiated fuel. Starting in the 1990s, the increased industrial competition and constraints led utilities to use fuel in more and more aggressive conditions (higher discharge burnup, higher power, load follow, etc.) and create incentive conditions for the development of advanced fuel designs with improved performance (new fuel types with additives, cladding material with better resistance to corrosion, etc.). These long anticipated developments involved the need for new investigations of irradiated fuel behaviour in order to check the adequacy of the current criteria, evaluate the safety margins, provide new technical bases for modelling and allow an evolution of these criteria. Such an evolution is presently under discussion in France and several other countries, in view of a revision in the next coming years. For this purpose, a R&D strategy has been defined at IRSN.

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Systematic Approach to Transboundary Radioactivity Monitoring for Accidents in External Nuclear Power Plants

Journal of Nuclear Engineering and Radiation Science ◽

10.1115/1.4049279 ◽

2020 ◽

Author(s):

Kampanart Silva ◽

Piyawan Krisanangkura ◽

Krirerk Phungsara ◽

Chatchai Chaiyasaen ◽

Suchin Udomsomporn

Keyword(s):

Emergency Response ◽

Nuclear Power ◽

Power Plants ◽

Atmospheric Dispersion ◽

Nuclear Accidents ◽

Radioactive Materials ◽

Loss Of Coolant Accident ◽

Loss Of Coolant ◽

Radioactivity Monitoring ◽

Dispersion Calculation

Abstract Past nuclear accidents demonstrated that radioactive materials from an accident in a nuclear power station (NPS) can disperse to other countries or even across the globe. This means all countries need to be prepared to respond to a nuclear power emergency even if they have no nuclear power program. This study aims to propose a structured framework for such a country to perform transboundary atmospheric dispersion assessment of an accidental release in an external NPS with limited calculation resources. A trial calculation of a hypothetical release from an interfacing system loss of coolant accident (ISLOCA) in Unit 1 of Fangchenggang NPS during different representative meteorological scenarios is carried out to demonstrate the usability of the proposed framework. It was found that a relatively large release can reach the border of Thailand within 24 hours when the wind along the dispersion pathway is basically in northeast direction with significant amount of rainfall, though it may not be able to trigger the alarm at the radiation monitoring stations. However, it is highly likely that the release that fulfills the aforementioned conditions be detected by one of the stations within 48 hour-timeframe. As the trial calculation could deliver insightful findings with limited calculation resources, the proposed transboundary atmospheric dispersion calculation framework can be used in other non-nuclear power countries to prepare for emergency response to accidents in external NPSs.

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Analysis on Impact of Fuel Thermal Conductivity Degradation (TCD) on Large Break Loss of Coolant Accidents of CAP1000

Volume 6: Thermal-Hydraulics ◽

10.1115/icone25-66283 ◽

2017 ◽

Author(s):

Wang Weiwei ◽

Lu Lu

Keyword(s):

Thermal Conductivity ◽

Nuclear Power ◽

Power Plants ◽

Sensitivity Study ◽

Fuel Pellet ◽

Steady State Condition ◽

Loss Of Coolant Accident ◽

Loss Of Coolant ◽

Fuel Storage ◽

High Burnup

Under high burnup conditions, thermal conductivity of fuel pellet degrades, which is referred to as thermal conductivity degradation (TCD). TCD phenomenon influences fuel average temperature and fuel storage energy under steady state condition before loss of coolant accident (LOCA) and further influences peak cladding temperature (PCT) during large break LOCA process. In this study, sensitivity study on double ended guillotine break of cold leg in CAP1000 at different burnup conditions was performed, using large break LOCA analysis code WCOBRA/TRAC and PCTs under different conditions were obtained. The modified NFI (Nuclear Fuels Institute) TCD model was adopted to model fuel conductivity after degradation in analysis and decrease of peaking factors including FQ and FΔh after 30GWD/MTU was also considered. Sensitivity analysis showed that: after considering the influence of TCD and peaking factor burndown, the PCT limiting case did not occur in low burnup range again, but occurred at burup of about 29GWD/MTU. Compared to other burnup points, the first and second peak values of PCT at that burnup point were all at the highest level. Performing of this study could prefer reference for analysis and estimation of large break LOCA of passive nuclear power plants under high burnup conditions.

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Analyses of loads on reactor pressure vessel internals in a pressurized water reactor due to a loss-of-coolant accident considering fluid-structure interaction

Kerntechnik ◽

10.3139/124.110123 ◽

2010 ◽

Vol 75 (6) ◽

pp. 311-315

Author(s):

P. Akimov ◽

L. Obereisenbuchner

Keyword(s):

Pressure Vessel ◽

Fluid Structure Interaction ◽

Reactor Pressure Vessel ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Water Reactor ◽

Structure Interaction ◽

Loss Of Coolant Accident ◽

Loss Of Coolant ◽

Reactor Pressure

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Axial strength of Zircaloy-4 samples with reduced thickness after a simulated loss of coolant accident

Nuclear Engineering and Technology ◽

10.1016/j.net.2021.01.001 ◽

2021 ◽

Author(s):

Jean Desquines ◽

Tatiana Taurines

Keyword(s):

Axial Strength ◽

Loss Of Coolant Accident ◽

Loss Of Coolant

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Prediction of Ballooning and Burst for Nuclear Fuel Cladding with Anisotropic Creep Modeling During Loss of Coolant Accident (LOCA)

Nuclear Engineering and Technology ◽

10.1016/j.net.2021.04.020 ◽

2021 ◽

Author(s):

Jinsu Kim ◽

Jeong Whan Yoon ◽

Hyochan Kim ◽

Sung-Uk Lee

Keyword(s):

Nuclear Fuel ◽

Fuel Cladding ◽

Loss Of Coolant Accident ◽

Loss Of Coolant ◽

Nuclear Fuel Cladding ◽

Anisotropic Creep

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A Study of Prediction Model Improvement for Air-Oxidation Breakaway in a Postulated Spent Nuclear Fuel Pool Complete Loss of Coolant Accident

Sustainability ◽

10.3390/su13031442 ◽

2021 ◽

Vol 13 (3) ◽

pp. 1442

Author(s):

Sanggil Park ◽

Jaeyoung Lee ◽

Min Bum Park

Keyword(s):

Kinetic Model ◽

Nuclear Fuel ◽

Spent Nuclear Fuel ◽

Complete Loss ◽

Air Oxidation ◽

Loss Of Coolant Accident ◽

Spent Nuclear Fuel Pool ◽

Loss Of Coolant ◽

Model Derivation ◽

Model Improvement

The temperature of zirconium alloy cladding on the postulated spent nuclear fuel pool complete loss of coolant accident is abruptly increased at a certain time and the cladding is almost fully oxidized to weak ZrO2 in the air. This abrupt temperature escalation phenomenon induced by the air-oxidation breakaway is called a zirconium fire. Although an air-oxidation breakaway kinetic model correlated between time and temperature has been implemented in the MELCOR code, it is likely to bring about unexpected large errors because of many limitations of model derivation. This study suggests an improved time–temperature correlated kinetic model using the Johnson–Mehl equation. It is based on that the air-oxidation breakaway is initiated by the phase transformation from the tetragonal to monoclinic ZrO2 at the oxide–metal interface in the cladding. This new model equation is also evaluated with the Zry-4 air-oxidation literature data. This equation resulted in the almost similar air-oxidation breakaway timing to the actual experimental data at 800 °C. However, at 1000 °C, it showed an error of about 8 min. This could be inferred from the influence of the ZrN phase change due to the nitrogen existing in air.

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