2022 – collaboration with EU JRC-Petten

NRGroup is establishing a new collaboration with the European Commission Joint Research Center (JRC) in Petten (Netherlands).

Research periods (3 to 6 months) to be spent at the JRC-Petten labs.

MSc and PhD thesis work positions are available on the following topics:

Modelling of Small Modular Reactors

1.   Development of an ASTEC model of NUSCALE SMR and its assessment under Design Basis and Design Extension Conditions

To investigate the behaviour of the NUSCALE Small Modular Reactor during design basis and severe accident conditions. To develop a NUSCALE model, using the European reference severe accident analysis code ASTEC, developed by IRSN.

2.   Development of a set of mutually advanced TRACE models of NUSCALE SMR and its assessment under Design Basis Conditions

The present work proposal aims to investigate the behaviour of the NUSCALE Small Modular Reactor during design basis conditions. In order to perform these analyses, it is proposed to develop a series of NUSCALE model, on the basis of the public available information, using the thermalhydraulic   system code TRACE, developed by USNRC.

3. CFD analyses of the SMR Containment (ELSMOR EU project)

R&D activities in support of the work planned in EU ELSMOR project, performing 3-D Computational Fluid Dynamics (CFD) safety analyses on the steel containment / water wall, using the CFD code Ansys CFX.

4.   Safety assessment of the french SMR concept (Nuward)

Model development (ASTEC code) and safety analysis, simulating both Design Basis and Design Extension Conditions.

Nuclear and climate change: Hybrid Energy Systems

5.  Small Modular Reactor (SMR) for Hybrid Energy Systems (HES) with variable renewables and energy storage in different European regions

HES including SMR to be optimized depending on local or regional forms of energy supply and demand. Hydropower, biomass, wind and solar energy plus SMRs and cogeneration needs: electricity, industrial process heat, district heating/cooling, hydrogen and seawater desalination. Different forms of energy storage to be considered (e.g. heat, electricity, pumped hydro, hydrogen, compressed air, and liquefied air). Safety and economic performance to be analysed.

Nuclear materials

6. Modelling of irradiation damage

Modelling of irradiation damage, in particular, the rate kinetics of defect populations (point defects, dislocation loops, precipitates, voids) during irradiation with energetic particles (neutrons and protons).

7.   Experimental studies of corrosion and mechanical properties of steels and alloys in liquid lead environment, in support to safety assessment and development of LFR SMRs

Performance of mechanical and corrosion tests in the JRC LILLA laboratory in high temperature liquid lead environment, complemented by microstructural investigations (mainly SEM) and assessments of the related phenomenology and damage mechanisms.

8. Qualification of new micromechanical test methodologies for mechanical property assessment of proton irradiated materials

Development of methodologies to assess materials performance based on specimens of micrometer size. Micromechanical testing (nanoindentation, microcompression testing, membrane bulge testing and/or microtensile testing) combined with microstructural analyses to correlate mechanical and microstructure properties.

9. Environmental Assisted Fatigue Testing in Light Water Reactor Relevant Conditions

Investigation of various parameters affecting environmental assisted fatigue. Innovative bellows-based loading device within this project. Indirect/direct strain measurement and control, effect of the specimen type and dimensions on environmental assisted fatigue.

10.  Experimental evaluation of Factors Accelerating Initiation of Stress Corrosion Cracking in Light Water Reactors

Acceleration of Stress Corrosion Cracking using supercritical water, to investigate factors affecting the initiation process.

11.   Corrosion behaviour of candidate Accident Tolerant Fuel (ATF) cladding materials in LWR and SCWR conditions

After Fukushima: cladding/coolant interaction during normal operation conditions. Electrochemical, microscopy, spectroscopy techniques and mechanical tests to understand the corrosion behaviour of the ATF candidates.

12.  Long-term degradation of material properties and ductility in support of design and assessment codes

To develop testing and assessment methods to demonstrate that nuclear components can safely operate for more than 60 years under high temperature reactor conditions.

13.   Qualification of a bellows device for testing of fuel cladding tubes

Construction of bellows driven test rig to carry out small punch and ring compression tests of fuel cladding tubes.


For further info and questions: marco.ricotti@polimi.it