1. SKC - Swedish Centre for Nuclear Technology and the MÅBiL project Hans Henriksson
Background to the Material, Ageing and Fuel programme, MÅBiL, within SKC
Status and highlights
Hans Henriksson
Oct 2018

2. Background to MÅBiL
A meeting held in May 2013 outlined three research areas that potentially could gain interest from a majority of the SKC partners:
Accident-tolerant Fuels (ATF):
coating with thin layers and new materials such as SiC.
Structural/fuel material behaviour in severe environments:
radiation, temperature, chemistry and aging.
Research within nuclear physics simulations and measurement technologies.
Hans Henriksson
Oct 2018

3. Original MÅBiL research groups
Material ageing
Klintenberg
UU, Material Theory & Tandem laboratory
Efsing
KTH, Solid Mechanics & Vattenfall
Olsson
KTH, Reactor Physics
Jonsson
KTH, Applied Nuclear Chemistry
Pomp
UU, Applied Nuclear Physics & NRG, Petten
Thuvander, Stiller
Chalmers, Materials Microstructure
Fuel
Ekberg, Retegan
Chalmers, Nuclear Chemstry
Pomp, Sjöstrand, Jacobsson Svärd
UU, Applied Nuclear Physics
Hjörvarsson
UU, Materials Physics
Hans Henriksson
Oct 2018

4. Background to MÅBiL
The new SKC strategy for research funding, led in 2013 to the present funding system with three year commitments for the MÅBiL research project. The aim is:
to create a research collaboration on national level, and
to address three research fields of particular interest for the nuclear industry:
Studies of novel materials with application to Accident Tolerant Fuels (ATF).
Studies of materials and phenomena connected to material ageing.
Studies of nuclear processes under regular and transient conditions.   
The first MÅBiL meeting was held in February 2014 in Uppsala. It was decided that the following “General Assemblies” of MÅBiL were to be held in connection with the Annual SKC Symposium.
Several research activities started late 2014 and two activities started up early 2015
Hans Henriksson
Oct 2018

5. SKC Research within MÅBiL
Time frame
Projects within MÅBiL
Aneta Herman (Sajdova): ”UN fuel manufacturing”, Teodora Retegan
Kristina Lindgren: ”Aging of RPV steel”, Mattias Thuvander, Krystyna Stiller
Elin Toijer: “Irradiation-Assisted SCC”, Pär Olsson, Mats Jonsson
Magnus Boåsen: “Material testing of ageing phenomena”, Pål Efsing
André Tengstrand, Mattia Bergagio: “Failure risk (crack propagation) due to aging of NPP components from thermal and mechanical loads”, Phd project, Pål Efsing, Henryk Anglart
Petter Helgesson: "TMC on fuel and materials", Henrik Sjöstrand
"Fuel Diagnostics at Halden", Peter Andersson, PostDoc
Erki Metsanurk: ”ICEWATER”, Mattias Klintenberg
Marciej Kaplan, ”Amorphous metals for high-temperature applications”, Björgvin Hjörvarsson
Dissertation, June 8,2018
Hans Henriksson
Oct 2018

6. Preparing for dissertation!
Ageing of Reactor Pressure Vessel Steel Welds Kristina Lindgren (and Mattias Thuvander et al.)
Overview
The reactor pressure vessel (RPV) is a life-limiting component of a nuclear reactor. Neutron irradiation during operation decreases the ductility of the RPV steel. The most important mechanism for making the steel brittle is the formation of nanometre-sized particles, consisting of Ni, Mn, Si and Cu.
The aim is to study ageing of RPV steel welds to better understand the mechanisms of ageing and to provide a basis for predicting the lifetime of RPVs.
Atom probe tomography on RPV steels to study clusters formed during ageing
Preparing for dissertation!
Hans Henriksson
Oct 2018

7. Research update by Kristina Lindgren
Clusters cut in half
During 2018:
Post irradiation annealing of irradiated materials (see her presentation 17/ )
Fontevraud 9 conference: comparison of thermal ageing (Ringhals R4 pressuriser) and irradiation on cluster characteristics, in connection to the presentation by Magnus Boåsen (KTH) on mechanical properties of the pressurizer
Cluster formation in in-service thermally aged pressurizer welds, K. Lindgren, M. Boåsen, K. Stiller, P. Efsing, and M. Thuvander, Journal of Nuclear Materials 504, (2018)
Dissertation Dec 11th
Outlook:
If time allows, materials retrieved from Barsebäck RPV will be analysed.
RPV-like material will be ion irradiated and analysed using APT and possibly TEM.
Cooperation with the ICEWATER project is anticipated. Probably APT analysis will be used to ascertain that the irradiation has affected the microstructure.
Thermally aged °C Ni Mn Si Cu
Irradiated surveillance material
Hans Henriksson
Oct 2018

8. Magnus Boåsen
Ageing effects on the mechanical properties of RPV steels
During 2018:
Manuscript for evaluation of true stress-strain of tensile tests based solely on force- displacement data
Paper on cleavage fracture, to be submitted to Engineering Fracture Mechanics,
poster on this symposium!
Continued testing of thermally aged material from the R4 pressurizer,
presentation at this symposium!
Attended the Fontevraud 9 conference
Collaboration 2018:
Continued work with Kristina Lindgren and Mattias Thuvander
Published a paper in Journal of Nuclear Materials on thermal ageing
Published two papers in the proceedings of Fontevraud 9
Oct 2018
Hans Henriksson

9. IASCC – Irradiation Assisted Stress Corrosion Cracking Elin Toijer (Per Olsson, Mats Jonsson, et al.)
Overview
Austenitic stainless steel are commonly used for structural components in nuclear reactors.
IASCC: Premature failure of nuclear reactor core components consisting of stainless steel alloys due to a significant neutron fluence and an aggressive environment.
Irradiation can induce creation and diffusion of point defects. This can lead to:
Formation of solute rich clusters
Defect diffusion to sinks like grain boundaries.
Material degradation, swelling, and loss of ductility.
Thus: The irradiation can have a great impact on the properties of the steel. For this reason the defect interaction with solutes and solvent is important for understanding the microstructure evolution of the material.
Hans Henriksson
Oct 2018

10. ICEWATER, Erki Metsanurk (Mattias Klintenberg)
The aim of ICEWATER project is to construct a test device to study irradiation assisted stress corrosion cracking (IASCC) in different types of austenitic stainless steels - Irradation Corrosion Experiments (ICE) in relevant reactor Water Chemistry.
Aims at studies to assess the synergistic effect of irradiation and chemical environment. IASCC tests performed on neutron-irradiated materials is expensive, time-consuming and requires special handling and machinery due to the high residual activity.
Material damage by protons is similar in nature to that of neutrons, but can be performed in much quicker, cheaper and safer ways and facilitates systematic studies of IASCC.
The Tandem laboratory delivers 10 MeV protons, giving materials 100 dpa
Hans Henriksson
Oct 2018

11. ICEWATER status
First finished version of the component that holds a 50 μm thick 316L stainless steel sheet that is the a barrier between the vacuum in the proton beam line and the high temperature high pressure water environment.
Currently it has been tested at up to 160 bar pressure at room temperature.
Next steps:
estimating the pressure dependent deformation of the thin foil
quantification of the local heating due to the highly energetic and localized proton beam
Hans Henriksson
Oct 2018