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

2010 ◽  
Vol 2010 ◽  
pp. 1-7 ◽  
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.

scholarly journals Safety Evaluation of Silicon Carbide and Zircaloy-4 Cladding during a Large-Break Loss-of-Coolant Accident

Energies ◽  
2018 ◽  
Vol 11 (12) ◽  
pp. 3324 ◽  
Author(s):  
Kwangwon Ahn ◽  
Kyohun Joo ◽  
Sung-Pil Park
Keyword(s):  
Silicon Carbide ◽  
Ultimate Strength ◽  
Safety Margin ◽  
Safety Evaluation ◽  
Loss Of Coolant Accident ◽  
Safety Margins ◽  
Loss Of Coolant

In this study, we aim to conduct structural analyses of cladding materials, such as silicon carbide and zircaloy-4, during a Large-Break Loss-of-Coolant Accident. The safety margin is the key consideration regarding the performance of the cladding materials. Our study shows that, in terms of primary stresses, SiC has a greater safety margin than zircaloy-4 due to SiC having a higher yield and ultimate strength; the cladding outer pressure is not affected by the cladding materials and, thus, the primary stresses of all cladding materials are the same. However, for secondary stresses, zircaloy-4 has the smallest fluctuation and irradiated SiC recorded the largest; secondary stresses and temperature histories are material-dependent. Ultimately, both cladding materials were found to have sufficient safety margins with respect to primary and secondary stresses.

scholarly journals Demonstration of Multi-Physics Evaluation of Fuel Rod Burst Probability With Burnup Extension Under LBLOCA Conditions

2020 ◽  
Author(s):  
Hongbin Zhang ◽  
Cole Blakely ◽  
Jianguo Yu
Keyword(s):  
Operating Time ◽  
Economic Benefits ◽  
Pressurized Water ◽  
Fuel Rod ◽  
Current Limit ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Water Reactors ◽  
Accident Conditions ◽  
Operating Cycles

Abstract Extending the fuel discharge burnup level, e.g., from the current limit of rod averaged discharge burnup limit of 62 GWD/MT to a proposed new limit of 75 GWD/MT, can provide significant economic benefits to the current fleet of operating light water reactors (LWRs). It allows for longer operating cycles and improved resource utilization. The major economic gain of longer operating cycles is attributable to the increased capacity factor resulting from decreased refueling time as a fraction of total operating time, as well as fewer assemblies to be discharged for a given amount of energy produced. The main licensing challenges for higher burnup fuel are to ensure fuel rod safety under design basis accident conditions, especially under large-break loss-of-coolant accident (LBLOCA) and reactivity insertion accident (RIA). In this work, two-year cycle core design for a typical 4-loop pressurized water reactor (PWR) is performed with enrichment increased up to 6% and burnup extended to 75 GWD/MT. The fuel rod burst potential evaluations under large-break loss-of-coolant accident (LBLOCA) conditions are subsequently performed using the multi-physics best estimate plus uncertainty analysis framework LOTUS (LOCA Toolkit for the U.S. LWRs) and the preliminary results are presented.

scholarly journals Toward UO2 micro/macro machining: a laser processing approach

2021 ◽  
Vol 253 ◽  
pp. 07004
Author(s):  
Thomas Doualle ◽  
Matthieu Reymond ◽  
Yves Pontillon ◽  
Laurent Gallais
Keyword(s):  
Material Removal ◽  
Model Material ◽  
Local Information ◽  
Fuel Pellet ◽  
Nuclear Accidents ◽  
Numerical Studies ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
First Results ◽  
Irradiated Fuel

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.

A Preliminary Safety Analysis of the Whole Assembly Seed-Blanket Heterogeneous PWR Core

2004 ◽  
Author(s):  
Jeongik Lee ◽  
Pradip Saha ◽  
Mujid S. Kazimi ◽  
Won-Jae Lee
Keyword(s):  
Temperature Rise ◽  
Safety Analysis ◽  
Complete Loss ◽  
Pressurized Water Reactors ◽  
Primary Flow ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Water Reactors ◽  
Safety Limit

The “Whole Assembly Seed and Blanket” (WASB) design, which utilizes mostly thorium in the blanket, consists of 84 seed and 109 blanket assemblies which may be backfitted into existing Pressurized Water Reactors (PWRs). Since the seed assemblies produce significantly more power than the blanket assemblies, a preliminary safety analysis of this design has been performed. Three accidents/transients (Large Break Loss of Coolant Accident (LBLOCA), Complete Loss of Primary Flow (LOPF) and Loss of Off-site Power (LOSP)), have been analyzed for both the WASB design and a typical all UO2 design for a typical 4-Loop Westinghouse PWR plant. LBLOCA results show that the peak cladding temperature (PCT) for the WASB design is approximately 260 K higher than that for a typical PWR design. However, this higher PCT for the WASB design is still about 200 K lower than the present regulatory safety limit. The response of the WASB and all UO2 core for LOPF and LOSP transients are very similar, and no post-DNB type rapid cladding temperature rise was observed in either of the two calculations.

Closed Model for Calculation of Differential Pressure After a Loss of Coolant Accident in Boiling Water Reactors

2009 ◽  
Author(s):  
Alexander Kratzsch ◽  
Wolfgang Ka¨stner ◽  
Rainer Hampel
Keyword(s):  
Differential Pressure ◽  
Boiling Water ◽  
Boiling Water Reactors ◽  
Insulation Material ◽  
Closed Model ◽  
Loss Of Coolant Accident ◽  
Safety Research ◽  
Loss Of Coolant ◽  
Water Reactors ◽  
Residual Heat

The paper deals with the calculation of differential pressure on sieves after a loss of coolant accident (LOCA) in boiling water reactors. One of the main features in reactor safety research is the safe heat dissipation from the reactor core and the reactor containment of light-water reactors. In the case of loss of coolant accident the possibility of the entry of insulation material into the reactor containment and the building sump of the reactor containment and into the associated systems to the residual heat exhaust is a serious problem. This can lead to a handicap of the system functions. To ensure the residual heat exhaust it is necessary the emergency cooling systems to put in operation which transport the water from the sump to the condensation chamber and directly to the reactor pressure vessel. A high allocation of the sieves with fractionated insulation material, in the sump can lead to a blockage of the sieves, inadmissibly increase of differential pressure, build-up at the sieves and to malfunctioning pumps. Hence, the scaling and retention of fractionated insulation material in the building sump of the reactor containment must be estimated. This allows the potential plant status in case of incidents to be assessed. The differential pressure is the essential parameter for the assessment of allocation of the sieves.

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