scholarly journals Neutronic benchmark of the molten salt fast reactor in the frame of the EVOL and MARS collaborative projects

2019 ◽  
Vol 5 ◽  
pp. 2 ◽  
Author(s):  
Mariya Brovchenko ◽  
Jan-Leen Kloosterman ◽  
Lelio Luzzi ◽  
Elsa Merle ◽  
Daniel Heuer ◽  
...  
Keyword(s):  
Molten Salt ◽  
Fast Reactor ◽  
Fuel Cycle ◽  
Negative Temperature ◽  
Nuclear Data ◽  
Data Set ◽  
The Core ◽  
Waste Production ◽  
Reactivity Coefficients ◽  
Core Characteristics

This paper describes the neutronic benchmarks and the results obtained by the various participants of the FP7 project EVOL and the ROSATOM project MARS. The aim of the benchmarks was two-fold: first to verify and validate each of the code packages of the project partners, adapted for liquid-fueled reactors, and second to check the dependence of the core characteristics to nuclear data set for application on a molten salt fast reactor (MSFR). The MSFR operates with the thorium fuel cycle and can be started with 233U-enriched U and/or TRU elements as initial fissile load. All three compositions were covered by the present benchmark. The calculations have confirmed that the MSFR has very favorable characteristics not present in other Gen4 fast reactors, like strong negative temperature and void reactivity coefficients, a low-fissile inventory, a reduced long-lived waste production and its burning capacities of nuclear waste produced in currently operational reactors.

Experimental Modelling of a Molten Salt Reactor Concept

2013 ◽  
Author(s):  
Bogdán Yamaji ◽  
Attila Aszódi
Keyword(s):  
Molten Salt ◽  
Experimental Model ◽  
Fast Reactor ◽  
Measurement Data ◽  
Internal Heat ◽  
Working Fluid ◽  
Experimental Modelling ◽  
The Core ◽  
Single Region ◽  
Piv Measurement

Based on the MSFR (Molten Salt Fast Reactor) reactor concept proposed within the framework of the EVOL (Evaluation and Viability of Liquid Fuel Fast Reactor System, EU FP7) international research project a scaled and segmented experimental model of the MSFR and first measurement result will be presented in the paper. MSFR is a single region, homogeneous liquid fuelled fast reactor concept. The reactor uses fluoride-based molten salts as fuel and coolant, with fissile uranium and/or thorium and other heavy nuclei content with the purpose of applying the thorium cycle and the burn-up of transuranic elements. The concept has a single region cylindrical core with sixteen radial inlet and outlet nozzles located at the bottom and top of the core. The external circuit (internal heat exchanger, pump, pipes) is broken up in sixteen identical modules distributed around the core. A scaled and segmented experimental model of the MSFR concept was designed and built in order to carry out Particle Image Velocimetry (PIV) measurements. Purpose of the experimental mock-up is to provide measurement data for validation and benchmarking of CFD simulations, and also to study specific problems or phenomena related to the MSFR, such as design of inlet geometry, effects of internal structures, coolant mixing. The experimental model uses water as working fluid with 50 μm polyamide seeding particles added for PIV measurement. Geometrical scaling was applied in order to reduce size and necessary pumping power and the geometry represents a 90 degree segment of the original cylindrical geometry. It was not possible to maintain the nominal value of the Reynolds-number (∼1E+06 for the core) however a highly turbulent flow (Re>1E+05) can be reproduced in the system. Final design of the scaled and segmented plexiglas model will be presented, capabilities and limitations of the measurement assembly will be discussed together with the presentation of first measurements results.

scholarly journals Nuclear Data Uncertainty Propagation to Reactivity Coefficients of a Sodium Fast Reactor

2014 ◽  
Vol 118 ◽  
pp. 535-537
Author(s):  
J.J. Herrero ◽  
R. Ochoa ◽  
J.S. Martínez ◽  
C.J. Díez ◽  
N. García-Herranz ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Uncertainty Propagation ◽  
Nuclear Data ◽  
Data Uncertainty ◽  
Sodium Fast Reactor ◽  
Reactivity Coefficients

scholarly journals IMPACT OF NUCLEAR DATA LIBRARIES ON PREDICTED FAST REACTOR PERFORMANCE

2021 ◽  
Vol 247 ◽  
pp. 09026
Author(s):  
A.G. Nelson ◽  
K.M. Ramey ◽  
F. Heidet
Keyword(s):  
Cross Sections ◽  
Fast Reactor ◽  
Power Distribution ◽  
Evaluation Process ◽  
Nuclear Data ◽  
Reactor Performance ◽  
Data Set ◽  
Experimental Database ◽  
Data Libraries ◽  
Nuclear Data Libraries

The nuclear data evaluation process inherently yields a nuclear data set designed to produce accurate results for the neutron energy spectra corresponding to a specific benchmark suite of experiments. When studying reactors with spectral conditions outside of, or not well represented by, the experimental database used to evaluate the nuclear data, care should be given to the relevance of the nuclear data used. In such cases, larger biases or uncertainties may be present than in a reactor with well-represented spectra. The motivation of this work is to understand the magnitude of differences between recent nuclear data libraries to provide estimates for expected variability in criticality and power distribution results for sodiumcooled, steel-reflected, metal-fueled fast reactor designs. This work was specifically performed by creating a 3D OpenMC model of a sodium-cooled, steel-reflected, metal-fueled fast reactor similar to the FASTER design but without a thermal test region. This OpenMC model was used to compare the differences in eigenvalues, reactivity coefficients, and the spatial and energetic effects on flux and power distributions between the ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0, JEFF-3.2, and JEFF-3.3 nuclear data libraries. These investigations have revealed that reactivity differences between the above libraries can vary by nearly 900 pcm and the fine-group fluxes can vary by up to 18% in individual groups. Results also show a strong variation in the flux and power distributions near the fuel/reflector interface due to the high variability in the 56Fe cross sections in the libraries examined. This indicates that core design efforts of a sodium-cooled, steel-reflected, metalfueled reactor will require the application of relatively large nuclear data uncertainties and/or the development of a representative benchmark-quality experiment.

scholarly journals Heat exchanger design studies for molten salt fast reactor

2019 ◽  
Vol 5 ◽  
pp. 12
Author(s):  
Uğur Köse ◽  
Ufuk Koç ◽  
Latife Berrin Erbay ◽  
Erdem Öğüt ◽  
Hüseyin Ayhan
Keyword(s):  
Heat Exchanger ◽  
Molten Salt ◽  
Fast Reactor ◽  
Heat Exchangers ◽  
Temperature Drop ◽  
Freezing Temperature ◽  
Fuel Temperature ◽  
Design Studies ◽  
The Core ◽  
Heat Exchanger Design

In this study, conceptual design for primary heat exchanger of the Molten Salt Fast Reactor is made. The design was carried out to remove the produced heat from the reactor developed under the SAMOFAR project. Nominal power of the reactor is 3 GWth and it has 16 heat exchangers. There are several requirements related to the heat exchanger. To sustain the steady-state conditions, heat exchangers have to transfer the heat produced in the core and it has to maintain the temperature drop as much as the temperature rise in the core due to the fission. It should do it as fast as possible. It must also ensure that the fuel temperature does not reach the freezing temperature to avoid solidification. In doing so, the fuel volume in the heat exchanger must not exceed the specified limit. Design studies were carried out taking into account all requirements and final geometric configurations were determined. Plate type heat exchanger was adopted in this study. 3D CFD analyses were performed to investigate the thermal-hydraulic behavior of the system. Analyses were made by ANSYS-Fluent commercial code. Results are in a good agreement with limitations and requirements specified for the reactor designed under the SAMOFAR project.

scholarly journals Impact of the Core Minor Actinide Content on Fast Reactor Reactivity Coefficients

2011 ◽  
Vol 48 (4) ◽  
pp. 628-634 ◽  
Author(s):  
Giuseppe PALMIOTTI ◽  
Massimo SALVATORES ◽  
Monchai ASSAWAROONGRUENGCHOT
Keyword(s):  
Fast Reactor ◽  
Minor Actinide ◽  
The Core ◽  
Reactivity Coefficients

Neutronic Analyses of the FREYA Experiments in Support of the ALFRED Lead-Cooled Fast Reactor Core Design and Licensing

2019 ◽  
Vol 6 (1) ◽  
Author(s):  
Massimo Sarotto ◽  
Gabriele Firpo ◽  
Anatoly Kochetkov ◽  
Antonin Krása ◽  
Emil Fridman ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Nuclear Data ◽  
Control Rod ◽  
The Core ◽  
Major Interest ◽  
Reactor Spectrum ◽  
Data Libraries ◽  
Nuclear Data Libraries ◽  
Sensitivity And Uncertainty Analysis

Abstract During the EURATOM FP7 project FREYA, a number of experiments were performed in a critical core assembled in the VENUS-F zero-power reactor able to reproduce the ALFRED lead-cooled fast reactor spectrum in a dedicated island. The experiments dealt with the measurements of integral and local neutronic parameters, such as the core criticality, the control rod and the lead void reactivity worth, the axial distributions of fission rates for the nuclides of major interest in a fast spectrum, the spectral indices of important actinides (238U, 239Pu, 237 Np) with respect to 235U. With the main aim to validate the neutronic codes adopted for the ALFRED core design, the VENUS-F core and its characterization measurements were simulated with both deterministic (ERANOS) and stochastic (MCNP, SERPENT) codes, by adopting different nuclear data libraries (JEFF, ENDF/B, JENDL, TENDL). This paper summarizes the main results obtained by highlighting a general agreement between measurements and simulations, with few discrepancies for some parameters that are discussed here. Additionally, a sensitivity and uncertainty analysis was performed with deterministic methods for the core reactivity: it clearly indicates that the small over-criticality estimated by the different codes/libraries resulted to be lower than the uncertainties due to nuclear data.

Modeling of Reactivity Effects and Transient Behaviour of Large Sodium Fast Reactor

2021 ◽  
Author(s):  
Alexander Ponomarev ◽  
Konstantin Mikityuk
Keyword(s):  
Fast Reactor ◽  
Core Model ◽  
Primary System ◽  
Control Rod ◽  
Transient Behaviour ◽  
The Core ◽  
Sodium Fast Reactor ◽  
Reactivity Coefficients ◽  
Core Conditions ◽  
Point Kinetics

Abstract In the paper the reactivity characteristics of the core of the large sodium fast reactor Superphenix (SPX) were evaluated and compared with available experimental data. The analysis was performed using the TRACE system code modified for the fast reactor applications. The simplified core model was developed aiming to overcome the lack of detailed information on design and realistic core conditions. Point Kinetics neutronic model with all relevant reactivity feedbacks was used to calculate transient power. The paper focuses on challenging issue of modelling of the transient thermal responses of primary system structural elements resulting in reactivity feedbacks specific to such large fast reactor which cannot be neglected. For these effects, the model was equipped with dedicated heat structures to reproduce important feedbacks due to vessel wall, diagrid, strongback, control rod drive lines thermal expansion. Peculiarly, application of the model was considered for a whole range of core conditions from zero power to 100% nominal. The developed core model allowed reproducing satisfactorily the core reactivity balance between zero power at 180?C and full power conditions. Additionally, the reactivity coefficients k, g, h at three power levels were calculated and satisfactory agreement with experimental measurements was also observed. The study demonstrated feasibility of application of relatively simple model with adjusted parameters for analysis of different conditions of very complex system.

Molten salt fast reactor with U–Pu fuel cycle

2015 ◽  
Vol 82 ◽  
pp. 33-36 ◽  
Author(s):  
Alexej Degtyarev ◽  
Andrej Myasnikov ◽  
Leonid Ponomarev
Keyword(s):  
Molten Salt ◽  
Fast Reactor ◽  
Fuel Cycle

Self-Breeding and Radiotoxicity Analysis on Thorium-Based Molten Salt Reactors

2013 ◽  
Author(s):  
Chun-yan Zou ◽  
Jin-gen Chen ◽  
Xiang-zhou Cai ◽  
Cheng-gang Yu ◽  
Da-zhen Jiang ◽  
...  
Keyword(s):  
Molten Salt ◽  
Nuclear Energy ◽  
Fuel Cycle ◽  
Molten Salt Reactor ◽  
Nuclear System ◽  
Energy Sustainability ◽  
Waste Production ◽  
Breeding Ratio ◽  
Fuel Salt ◽  
System Project

As one of the candidates in the Generation IV reactors program., the molten salt reactor (MSR) has the properties of online refueling and fuel salt reprocessing, MSR is especially attractive for the Thorium fuel cycle, which is very ideal for nuclear non-proliferation, radiotoxicity and nuclear energy sustainability. Therefore, the “Thorium-based Molten Salt Reactor (TMSR) nuclear system” project has been proposed as one of the “Strategic Priority Research Program” of Chinese Academy of Science (CAS). In this paper, we mainly investigated the influence on the breeding ratio and waste radiotoxicity with different reprocessing schemes. By considering the key parameters mentioned above, the aim is to choose an efficient reprocessing scheme for TMSR to reach self-breeding with Th/U fuel cycle and minimize the radioactive waste production of the molten salt.

scholarly journals Preliminary proliferation study of the molten salt fast reactor

2020 ◽  
Vol 6 ◽  
pp. 5 ◽  
Author(s):  
Michel Allibert ◽  
Elsa Merle ◽  
Sylvie Delpech ◽  
Delphine Gerardin ◽  
Daniel Heuer ◽  
...  
Keyword(s):  
Molten Salt ◽  
Fast Reactor ◽  
Molten Salts ◽  
Fuel Cycle ◽  
Power Reactor ◽  
Nuclear Material ◽  
Reactor System ◽  
Molten Salt Reactor ◽  
International Forum ◽  
Large Power

The molten salt reactor designs, where fissile and fertile materials are dissolved in molten salts, under consideration in the framework of the Generation IV International Forum, present some unusual characteristics in terms of design, operation, safety and also proliferation resistance issues. This paper has the main objective of presenting some proliferation challenges for the reference version of the Molten Salt Fast Reactor (MSFR), a large power reactor based on the thorium fuel cycle. Preliminary studies of proliferation resistance are presented here, dedicated to the threat of nuclear material diversion in the MSFR, considering both the reactor system itself and the processing units located onsite.

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