Thursday, July 2, 2009

The EURATOM project SUMO

I came across a European power point presentation that listed remaining technology issues before MSR development can begin. I copied the notes from the presentation, although I realize that this might not be for everyone.
ALISIA final meeting, Paris March 2008

Presented by Victor V. Ignatiev

RRC- Kurchatov Institute

Recent years have demonstrated revisiting of interest in nuclear energy systems employing technology of molten salt fluorides for:
Electricity production in Th-U breeder cycle (MSBR…TMSR)

Consumption of TRU while extracting their energy (MOSART)

Pyrochemical reprocessing of advanced fuels

Coolant application in solid fuel FR designs

Process heat applications, H2 production, heat accumulation

Production of isotopes for medical needs (Mo,Sr)

Incineration of organic wastes

Introduction: MSR remaining issues
A modified Ni-base alloy that is immune to Te-attack must be selected and its compatibility with fuel salt demonstrated; means for giving it adequate resistance to radiation damage must be found.

A method of intercepting and isolating tritium to prevent its passage into the steam system must be demonstrated.

Various steps in the processing system must be demonstrated and then be combined in an integrated system.

Graphite elements that are suitable for the MSR should be purchased (radiation behavior, graphite sealing).

On-line chemical analysis devices will be needed for the reactor and processing plant.
Components and systems for reactor must be developed and tested.

MS fuel / coolant circuits salts:
molar compositions, melting temperatures (°C) and PuF3 solubility (mole %) at 600 °C
ISTC #1606:

Focus is placed on experimental & theoretical evaluation of MOlten Salt Actinide Recycler & Transmuter (MOSART) system
#1606 Developments

Experimental consideration included basis properties of fuel salts: phase transition behavior, trifluorides/oxides solubility for An and Ln, viscosity, thermal conductivity, density
Study on different methods of fuel salt clean up in solvents selected (electrodeposition, oxide precipitation, reductive extraction).

Experimental verification of Ni-Mo alloys for fuel circuit in corrosion loops, including PuF3 and Te with redox measurement.

New experimental data received in #1606 studies feed into the conceptual design efforts, which also fit to the need of EU partners.

Physical properties
Corrosion studies
1200 hr Compatibility test between Ni-Mo alloys and molten 15LiF-58NaF-27BeF2 salt in natural convection loop with measurement of redox potential.

400hr Study on PuF3 addition effect in molten 15LiF-58NaF-27BeF2 salt on compatibility with Ni - Mo alloys.

500 hr Te Corrosion study for molten 15LiF-58NaF-27BeF2 salt and Ni-Mo alloys in stressed and unloaded conditions with measurement of redox potential.

Fuel salt clean up

#1606 Deliverables
Six detailed annual technical reports (2002-07)

# 1606 Final project reports - phase 1,2 ( 2004, 2008)

# 3749 Project proposal (2007)

#1606 Training sessions proceedings (2008)

More than 30 papers for magazines:
“Atomnaya Energia”, “Revue Generale Nucleaire”, “Nuclear Technology”, “Transactions of ANS”, “J. of Fluorine Chemistry”
and International Meetings:


The development and commercialization of a fluid fuel MS technology and its use in reactor concepts is a major challenge. However, the interest in advanced reactors for breeding and waste management is creating the incentives to develop this family of technologies.
Preliminary consideration of environmental effects indicate that MOSART system could have attracted performance, good safety features and TRU transmutation efficiency while providing lower total materials inventories and waste compared to prior MSR designs.
While a substantial R&D effort would be required to commercialize MOSART, there are no killing unresolved issues in the needed technology. The major technical uncertainties in the conceptual design are in the area of tritium confinement, fuel salt processing and behavior of some fission products.

The mission of ISTC # 3749 is to test and select molten salts and metallic structural materials, which will operate successfully under the conditions of promising nuclear energy applications.
The work in ISTC # 3749 will be focused on Th-fuelled system to produce energy in a competitive way (in co-operation with FP6 ALISIA project and FP7 SUMO proposal).
Experimental effort will be placed on evaluation of the potential of advanced molten salt fluorides mixtures for pyrochemistry partitioning application (in co-operation with FP7 ACSEPT project).

Expected developments

WP 1: Choice of molten salt compositions for detailed studies; Special attention will be given to the following fuel solvent systems: Li,Th/F, Li,Be,Th/F, Li,Ca,Ba/F, etc.

WP 2: Measurement of key physical and chemical properties for selected salts

WP 3 : Molten salt preparation and clean up

WP 4 : Experimental verification of candidate Ni- base alloys in molten salt corrosion facilities with redox measurement

WP 5: Core physics, thermal hydraulics, safety, fuel cycles and system configuration

WP 6: Final report

WP 2, WP 3
Target: This work group will focus on the experimental study of physical and chemical properties of molten salt mixtures. Starting from LiF-containing systems, composition will be optimised with respect to margin to freezing, An/Ln solubility, adequate transport properties, rare earths removal and An/Ln separation on the basis of possible addition of other components: ThF4 and BeF2 or CaF2 and BaF2. As fissile components UF4 and PuF3 will be considered.

An important requirement for such a program is the availability of experimental facilities for handling of fluoride salts, which is guaranteed by the participation of RRC-KI, IHTE and VNIITF. The chemical and physical properties of the candidate salts will be established based on a variety of experiments and evaluations that will be performed at the participating institutes.

WP 2, WP 3: Salt Properties
Selection of candidate salt compositions

Measurement of viscosity, density, thermal conductivity and heat capacity for salt compositions selected

Measurement of metals trifluorides solubility in salt compositions selected

Tritium permeation studies of clean and oxidized metals

Study on electrochemical properties of metals (conditional standard electrode potentials) and their dissolved trifluorides (reduction kinetics, activity coefficients etc.) in molten solvent systems selected

Study on molten salts clean up processes (reductive extraction, electrodeposition) as applied to conditions of reference designs.

WP 4: Container materials

Target: The main objective of this group is, to select structural materials, which will operate successfully under the required conditions. The major achievements will be: (1) ability to produce and maintain a high level of purity in fuel salt, (2) effective control of the redox potential of the fuel salt in order to minimize corrosion, (3) understanding of basic corrosion mechanisms in systems with temperature gradient, (4) improved materials development and testing as applied to conditions of reference designs.

The long-term behavior of materials will be comparatively investigated by performing exposure tests in fuel salt, defined on the basis of the results obtained in previous section and that should be representative of the nominal operating conditions.

The changes in the alloy microstructure will be studied. The specific corrosion resistance of welds and the influence of the surface state will be examined. The stability of the mechanical properties after exposure will be also verified.

WP 4
Selection of candidate material specimens and corrosion facilities upgrade.
Development of electroanalytical technique for redox potential evaluation in selected salt melts with UF4 additions

Study of Te corrosion of Ni - base alloys in selected salts with different specimens stress loading and evaluation of redox potential

Compatibility tests (>1000 hrs) between selected salt and Ni - base alloys in thermal convection loop with evaluation of redox potential

Detailed examination of nickel- based alloys specimens properties before and after exposure in salt melt

WP 5: System configuration and fuel cycle
New Th-U concept will be evaluated basing on experimental data received within the project. In this task core performance (nuclear data, thermal hydraulics, fuel cycle scenarios, irradiation material damage, tritium production, etc) and core safety related parameters (reactivity feedback coefficients…) are mainly addressed.

Evaluation will include possible unification of Th-U system with MOSART in order to establish sustainable, safe, proliferation resistant symbiotic system with minimal nuclear waste.

#3749 Final report
Confirmation on the salt compositions selection and its key properties data.

Recommendation on the structural material selection.

Optimization of the system configuration and its fuel cycle.

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