Summary of Technical GIF Achievements (GFR, MSR, LFR and ... Documents/12th GIF-IAEA... · –Verification of isothermal saturation solubility measurement technique –Measurements

Summary of TechnicalGIF Achievements

(GFR, MSR, LFR and SCWR Systems)

Alexander Stanculescu

GIF Technical Director

Twelfth GIF – IAEA Interface Meeting

26-27 March 2018

212th GIF – IAEA Interface Meeting, 26-27 March 2018

GFR (1/4)• Fuels and Materials Design

– Exploratory fuel design studies identified promising design options (especially for self-generating cores)

» High density fuels (carbides or nitrides) high actinide content; lower fuel temperatures

» Carbide and nitride fuels both serious candidates, on-going R&D to provide new insights

» Oxide fuel option considered in a transition period before the full development of carbide fuels

» Pin type solid solution selected

– Choice of SiC fiber reinforced SiC matrix (SiCf/SiC) composite cladding; metallic refractory or semi-refractory claddings back-up solution


GFR (2/4)

– Studies confirmed best material choices for reactor structures (sub-assembly structures, CRs, stand pipes, reflectors) to be ceramics (CER) and composite material (e.g. CER/CER, CERf/CER, CER/MET)

• Neutronics

– Neutronics tools validation based on data from the MASURCA ENIGMA and the GCFR-PROTEUS experimental programmes

• Core thermal hydraulics

– Air tests (pressure drop, heat exchange, hot spots risk for various sub-assembly designs)


GFR (3/4)

• Components

– Helium circulator modelling

– Circulator design for previous ALLEGRO (50 MWth) concept

• Helium purification

– Design of purification circuit parallel with the primary circuit

– First stage of trial operation of NRI’s (Řež) experimental High-Temperature Helium Loop (HTHL) ALLEGRO project

– Objective: Development of innovative refractory GFR fuels, GFR-related components and systems (Helium technologies, fuel handling, etc.), and of a specific safety framework applicable to GFRs


GFR (4/4)

– Thermal hydraulics system codes (CATHARE, RELAP, MELCOR) validation

– Safety systems improvements (for safety injection accumulators, increased back-up pressure for the guard containment)

– ALLEGRO demonstrator feasibility studies

» Analysis of power rating (30 – 75 MWth) and density (50 –100 MW/m3) impact on safety limits and the design requirements

» Feasibility of a LEU UOX stainless steel cladded start-up core (to include experimental positions dedicated to the refractory fuel development) as alternative to standard MOX core


MSR (1/6)• Upside/downside: increased interest from private sector/lack of

public funding reflected in MSR system achievements

• Compatibility of molten salts (Li,Be/F, Li,Be,U/F and Li,Be,Th,U/F) with structural materials (US Hastelloy-N, Russian HN80MTY, Czech MONICR, Chinese GH3535, and French EM-721)

– Natural convection loop tests of high nickel alloys and fuel/coolant compatibility, with redox potential measurements

– Study of impact of PuF3 addition on fuel salt/high nickel alloys compatibility

– Tellurium corrosion studies, with redox potential measurements


MSR (2/6)• Instrumentation and control

– Measurement of the redox potential in molten salts ( Li,Be/F, Li,Be,Pu/F, Li,Be,U/F and Li,Be,Th,U/F) using the voltammetric technique good correlation observed for all considered high nickel alloys between the measured redox potential, and the corrosion rates

• Physical-chemical behaviour of fuel/coolant salts (neutronics, thermal hydraulics and chemistry aspects)

– Verification of isothermal saturation solubility measurement technique

– Measurements (550 – 800 °С range) based on the isothermal saturation technique of UF4, PuF3, AmF3 individual solubility in binary LiF-BeF2, as well as joint UF4, PuF3 and CeF3 solubility in ternary LiF–NaF–KF and LiF-ThF4-UF4 melts


MSR (3/6)• Measurements of chemical-physical properties of salts containing

UF4, ThF4 and PuF3

– Melting point behaviour, vapour pressure, heat capacity, phase diagram data and enthalpy of transitions

– Phase diagrams of key systems assessed according to the Calphad method extensive thermodynamic database established (Joint Research Centre Molten Salt Database, JRCMSA)


MSR (4/6)• MSR analyses

– Development and qualification of the Transient Fission Matrix (TFM) neutronics model (able to take into account precursor motion phenomena and to perform coupled transient calculations) to simulate MSRs behavior in normal (e.g. load-following) and incidental (e.g. over-cooling) situations

– Multi-physics code systems developed using rigorous neutronics and CFD calculation methods (the code systems have been benchmarked, validation efforts based on experimental data are planned)

– Preliminary safety assessment performed for the MSFR


MSR (5/6)• Inactive salt testing

– Two CRNS facilities (SWATH-W, SWATH-S) established to study salt heat transfer and phase change phenomena (solidification and fusion processes)

– A forced convection FLIBE loop (materials research, MSR and FHR components testing) built and put into operation at NRI (Řež)

– A versatile facility (LSTL) allowing testing of fluoride salts up to 700 °C established at ORNL in support of the development and demonstration of FHR technology


MSR (6/6)• Demonstrator projects

– Without induced fission

» Planning and design of Mock-ups for the TMSR programme to validate thermal hydraulics and safety analyses

– With induced fission

» Fuel salt irradiations in the HFR Petten to study in-pile corrosion, on-line fission product removal and fission product behavior under accident conditions, as well as the treatment of irradiated fuel waste

» Conceptual design of TMSR-LF1 (<2 MWth, pool type test reactor with LiF-BeF2-ThF4-UF4 fuel salt flowing through channels surrounded by graphite moderator)


LFR (1/4)• Material corrosion and lead chemistry

- Development of silicon doped ferritic-martensitic steel

- Development of Alumina Forming Austenitic (AFA) steels

• Instrumentation

- In-core neutron flux monitoring development

- Development of various measurement techniques based on contactless control of heavy liquid metal flows by magnetic fields

• Fuel handling

- Development of refueling strategies at ambient temperature under full visibility, as well as below the lead level


LFR (2/4)• Advanced modeling and simulation

- Introduction of lead and lead-bismuth eutectic properties and heat transfer correlations in existing LWR system codes and generation of experimental data base for code validation and for the development of new correlations

- Development of 3D approaches for CFD codes and respective validation efforts

- Improvement of neutronics data aiming at reduction of simulation uncertainties

- Studies (based on molecular dynamics analytical tools) of heavy liquid metal – material interactions aiming at improving understanding of corrosion/erosion processes


LFR (3/4)• Fuel

- MOX fuel development efforts: various efforts, cross-cutting for all GIF fast reactor systems

- Nitride fuel

» Manufacturing plant under construction at BREST site

» Fuel testing and qualification initiated (BOR-60, BN-600)

» Preliminary irradiation results support BREST nitride fuel choice

• In-service inspection and repair

- Ultrasonic view systems adapted to heavy liquid metals buoyancy forces and higher inspection temperatures

- Simplified ISI&R strategies maximizing number of components that can be extracted from the primary pool


LFR (4/4)

• Seismic behavior

- Study proving substantially improved seismic response behavior of 600 MWe LFR mounted on seismic isolators

- Development of a new vessel concept using a multilayer structure with relatively small interconnected pools


SCWR (1/6)

• System integration and assessment

– Thermal spectrum reference designs (pressure tubes adopting thorium fuel, and pressure vessel adopting uranium fuel cycle) completed and reviewed by international peers


SCWR (2/6)

• Materials and chemistry

– Materials development approach

» Evaluation of candidate materials with regard to corrosion and stress corrosion cracking, strength, embrittlement and creep resistance, and dimensional and micro-structural stability

» Potential for water chemistry control to minimize impacts as well as rates of deposition on fuel cladding and turbine blades

» In-pile loop measurement of performance data


SCWR (3/6)– Key challenge: fuel cladding material for T>600 °C (Zr-

alloys not applicable)

» Fuel cladding candidate materials identified

» Corrosion characteristics for one of them studied through the Round Robin exercise at supercritical pressures and temperatures

» The weight-change data exhibited considerable discrepancies between laboratories; follow-up Round Robin corrosion exercise initiated

– Combination of ex-situ analytical studies of the oxide film forming processes with modelling approaches used to study various techniques (e.g. coatings and cold working) to improve the oxidation resistance of traditional low alloyed austenitic steels


SCWR (4/6)• Thermal hydraulics and safety

– Evaluation of thermal hydraulics characteristics during normal, off-normal operation, and postulated accidents

» Heat transfer and fluid flow of supercritical water in various geometries

» Critical flow

» Quantification of flow stability and transient behavioursdue to the strong coupling of neutronics and thermal hydraulics phenomena

» Validation of computer codes

» Definition of the safety and licensing approach as distinct from current water reactors, including the spectrum of postulated accidents


SCWR (5/6)

• Extensive databases on supercritical heat transfer compiled for tubes and bundle subassemblies in water and non-aqueous fluids

– Water heat-transfer database for tubes contain over 24’000 data points

– Data applied in validating correlation and developing new prediction methods


SCWR (6/6)

• Test facilities established

– Heat-transfer tests with tubes, annuli and bundle subassemblies in water, carbon dioxide or refrigerant flows

– Critical flow and flow stability experiments at supercritical pressures

– Focus on obtaining data on cladding surface temperatures at bulk temperature close to the pseudo-critical point

– Experiments with water or surrogate-fluid flows through small electric-heated bundle assemblies to investigate deteriorated heat-transfer phenomena and effect of spacers on heat transfer

– A recent blind benchmark exercise on heat transfer in an electrically heated rod bundle in supercritical water confirmed that further work is required to provide reliable predictions