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Session Overview
3.04: Fast Neutron Reactors – Computational Method
Wednesday, 18/Mar/2020:
2:00pm - 3:30pm

Session Chair: David SETTIMO, Electricite de France (EDF), France
Location: L-1011

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A Validation Study of a Neutronics Design Methodology for Fast Reactors Using Fuel Reactivity Worth Measurements in the Prototype Fast Reactor Monju

Kazuya Ohgama1, Atsushi Takegoshi2, Hiroki Katagiri1, Taira Hazama1

1Japan Atomic Energy Agency; 2NESI, Inc.

For the reliable design of future fast reactors, it is important to enhance the credibility of design methodologies by conducting verification and validation (V&V), and uncertainty quantification (UQ) study. Thus, Japan Atomic Energy Agency (JAEA) has been developing the experimental database for the validation as well as establishing and verifying design methodologies, establishing V&V and UQ procedure of the neutronics design methodology. As a part of the activities, JAEA currently focuses on developing the experimental database of the prototype fast breeder reactor Monju in order to preserve and utilize valuable experimental data. In Monju, various reactor physics experiments were conducted during its system startup test. Some of the experiments have been intensively evaluated to meet the evaluation standard recommended in the International Reactor Physics Benchmark Experiments (IRPhE) Project and utilized for the validation of the neutronics design methodology. The fuel reactivity worth measurements are a next candidate to be evaluated for the benchmark data. In the core management of fast reactors such as making refueling plans and critical mass prediction of initial cores, evaluation of fuel reactivity worth are important. In the prototype Monju, fuel reactivity worth was measured in several different positions in the core by replacing a fuel subassembly with a dummy one. In this study, reliability and usefulness of the fuel reactivity worth measurements were investigated through a comparison with calculation results by the latest neutronics design methodology developed in JAEA and the Monte Caro simulation code MVP with the nuclear data library JENDL-4.0. Through the comparison, it was confirmed that the both of experimental values and analysis results agreed well within several %. Through this study, these experimental data was found reliable and useful for the validation of fuel reactivity predicted by the neutronics design methodology.

Review of models for the sodium boiling phenomena in SFR subassemblies

Haileyesus Tsige-Tamirat1, Sara Perez-Martin2, Werner Pfrang2, Marine Anderhuber3, Antoine Gerschenfeld3, Laurent Laborde4, Konstantin Mikityuk5, Christophe Peniguel6, Stephane Mimouni6

1European Commission Joint Research Centre; 2Karlsruhe Institute of Technology; 3CEA, Saclay; 4L'Institut de Radioprotection et de Sûreté Nucléaire; 5Paul Scherrer Institut; 6Électricité de France R&D

The Euratom Horizon-2020 project ESFR-SMART aims to enhance the safety and performance of the European Sodium-cooled Fast Reactor (ESFR) considering safety objectives envisaged for Generation-IV reactors and the update of European and international safety frameworks, taking into account the Fukushima accident. Further, the project aims to support the development of the computational tools for each defense-in-depth level in order to support the safety assessments using data produced in the project as well as selected legacy data. Within this activity, the focus is on the further development of computer codes for the analysis of the sodium thermal-hydraulics phenomena in SFR subassemblies under operational and accidental conditions including sodium boiling and transitional convection single-phase and two-phase flows. In support of this activity, it has been agreed to review the sodium boiling models used in the codes participating in the benchmark activity within the project.

The objective of the present paper is to summarize the result of the review which encompasses both the phenomenological and mathematical models implemented in the codes. In particular, the review addresses (i) the geometry representations ranging from one-dimensional single channel to full three-dimensional sub-channel and porous media, (ii) the treatment of heterogeneities, (iii) the physical bases of sodium boiling models and (iv) the analytical models in connection with the numerical implementations in the codes. Moreover, the modelling approaches to the various phenomena are discussed including boiling onset, nucleation, superheat, bubble growth and vapor pressure, two-phase flow regime, liquid vapor interface, bubble collapse re-entry, liquid film properties, dry-out, power, temperature and flow gradients in a sub-assembly.

The final part of the paper aims to summarize the advantages and limitations of the approaches implemented in the codes.

Super FR core design option with high inlet temperature for MA transmutation

Takanari Fukuda, Akifumi Yamaji

Waseda University

 Super Fast Reactor (Super FR) is one of fast reactor concepts of SuperCritical Water-cooled Reactor (SCWR). While the core is cooled by single-phase coolant, the coolant density drops from about 0.8 (g/cc) at the core inlet to about 0.1 (g/cc) near the core outlet. Since minor actinides (MAs) can fission with fast neutrons, decreasing core coolant density by increasing core inlet temperature from conventional 295 (℃) may improve MA transmutation performance, although it may lead to degradation of the plant thermal efficiency. Hence, this study aims to show possible Super FR design options of different core inlet temperature with considerations on its effects on the MA transmutation characteristics and plant thermal efficiency.

 Neutronics and thermal-hydraulics coupled three-dimensional core calculations have been carried out for investigating the MA transmutation and the thermal-hydraulics characteristics with the design criteria of negative void reactivity coefficient, maximum cladding surface temperature ≦ 650 (℃ ) and maximum linear heat generation rate ≦ 39 (kW/m). The plant thermal efficiency has been evaluated through solving heat balance and mass flow conservation in each part of the tentatively determined two-stage reheating and eight-stage regenerative steam cycle.

 The core is arranged with multiple layers of mixed oxide (MOX) fuel and depleted uranium fuel, and MA has been added in the MOX layers. The results show that the MA loading to the core is limited by the negative void reactivity criterion, when Pu enrichment is increased to compensate the reactivity penalty with the addition of MA. Moreover, higher inlet temperature with lower coolant density is found to be favorable from the viewpoint of suppressing such reactivity penalty with MA loading. The estimated plant thermal efficiency is not greatly affected by raising the core inlet temperature, because the core outlet temperature is also increased.

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