Final Report Summary - MATTER (MATerials TEsting and Rules)
Executive Summary:
The MATTER Project was launched in 2010 to support the European initiatives in the field of materials for Gen IV reactors. Since two projects were recognized as the most advanced in Europe, ASTRID and MYRRHA, the major efforts of the European materials community were addressed to both of them. The MATTER project had also the duty to support and to assist from the organizational aspects the start of Joint Program for Nuclear Materials in the frame of EERA. Due to wide interest by European organization to co-operate in the development of materials for Gen IV reactors, 26 institutions from Research institutes, Industry and Academia adhered to the project. The relevant number of objectives induced the coordination to split the project into 4 organizational domains: 1) Materials Characterization, 2) Pre-normative R&D, 3) EERA Joint Program for Nuclear Materials, 4) Management and Education/Training. The first domain included the prioritization among the material properties to be dealt in the project, provided several screening test techniques, reached a consensus within the EU partners on common testing and evaluation procedures, allowed round robin exercises to validate the various testing and evaluation procedures and finally provided analytical and physically-based support to the proposed guidelines.
In the pre-normative domain the design rules were revised and updated in order to answer to some short term needs of the two projects ASTRID & MYRRHA. The validity of design rules such as ratcheting and creep-fatigue was assessed, some materials features such as negligible creep domain, ageing, corrosion were reviewed for update. The best practices for welding processes were thoroughly studied with the support of industrial partners and codification updates were proposed. Finally, characterization tests were performed to assess the properties of grade 91 steel, of improved functional surfaces and to devise weld coefficients.
The main purpose of Domain 3 of MATTER is to support the EERA NM R&D Joint Program. The EERA JPNM is made out of a “Short-term” component addressing structure and clad materials issues in support of ESNII reactors and a “Medium-long term” component addressing structure and clad R&D needs for the FOAK (First of a kind) FNR (Fast Neutrons reactor).The relevant classes of materials considered for the short-term program are Austenitic alloy, F/M steels, hard coatings to replace Stellite. The related studies, in MATTER, are mainly done in domains 1&2. The long-term program includes more innovative materials which are at the development level and have been therefore never used for nuclear systems. These materials are ODS and Refractory Metals and Ceramics. Moreover, in the long-term program, one sub-programme is dedicated to modelling, experiment and validation. This sub-program is considered as cross-cutting and also relevant for the short-term part. MATTER dedicated a relevant effort to ODS supporting several different fabrication paths by the contribution of six European organizations.
The relevant issues of the fourth domain were Relevant issues of Domain 4 were the education & training, the dissemination, the management and the scientific coordination. The education and training was assured by the financial support to nine Positions of PhD which were launched inside the project and by the organization of student courses. The dissemination was performed by the organization of international workshops, the publish of 21 Scientific papers, 11 paper are under revision for the special issue of Journal of Nuclear Material dedicated to MATTER project, and MATTER web site. And ODIN tool for materials database The project management consisted of all the legal, contractual and financial activities and the controlling administrative issues and of the provisions for support of the Governing Council, the Project Co-ordination Committee, and respective meetings. The scientific coordination included the effective conduction of the work programme, the contacts with EERA group, other EC Projects and Gen IV and the provisions for support of the End-Users Advisory Group and of the Scientific Advisory Board.
Project Context and Objectives:
Context
The 2010-2012 implementation plan of the European Sustainable Nuclear Industrial Initiative (ESNII), in the frame of the Sustainable Nuclear Energy Technology Platform (SNE TP), establishes the road map for the start of construction of the European Gen IV prototypes. Namely it has been already decided that the construction of the LFR ETPP (European Technology Pilot Plant) MYRRHA and SFR Prototype ASTRID will be completed in the twenties.
The development of GEN IV fission reactors promises a long lasting, carbon free, power source having sustainable impact on next generations and enhanced safety in terms of resistance to severe nuclear accidents. The achievement of these goals passes also through the development of structural materials having enhanced capabilities to withstand the highly demanding conditions of the fast neutron reactors.
The operational conditions envisaged for the various GEN IV reactor concepts, under consideration in the ESNII initiative, are quite demanding and significantly different from the service conditions of most of the commercial reactors operating today. The combination of severe conditions includes:
- High neutron damage, as an effect of the high neutrons energy (up to few MeV);
- High temperature, as a consequence of the specific thermal cycles required by the coolants properties and by the quest for higher thermodynamic efficiency;
- Chemical and physical compatibility with the aggressive conditions due to contact with the coolants.
The behavior of materials under operation conditions of such systems was not appropriately covered by the available testing and evaluation standards, which therefore required updates of existing standard procedures and sometimes development of new ones. In addition, one of the required features for the success of the Gen IV reactors is the demonstration of their long lifetime: up to 60 years.
It is recognized that the available European standards and rules for material testing and for components design & fabrication do not appropriately and completely cover the specific materials to be used in GEN IV reactors nor the issues related with their relevant operating conditions and environments.
The EERA (European Energy Research Alliance) group for innovative nuclear materials has been established in 2009 regrouping several scientific representatives from research institutions all around Europe with the purpose to foster and perform material studies for nuclear applications.
From this consideration it appeared the necessity to answer to the industrial needs for innovation of the design rules, namely the RCC-MR and R5 standards , which were developed in Europe, capitalising the newly developed knowledge. That was strongly underlined by the Multinational Design Evaluation Program (MDEP) Codes & standards Working Group which has the goal to achieve convergence of practices related to nuclear component design and fabrication.
These topics are in line with a number of European (GETMAT, HeliMnet, GOFASTR, LEADER, ESFR) but also international initiatives (GIF, OECD/NEA) where experimental data were collected.
Scope and objectives
Scope of the MATTER (MATerial TEsting and Rules) project was to contribute to cover the existing gaps by pointing out methodologies, recovering existing experiences, and performing experiments.
Main goals of the MATTER Project were:
• To develop and establish best practice guidelines for testing and evaluation, aimed to screen and characterize nuclear materials for all the ESNII innovative nuclear systems. These harmonized testing and evaluation guidelines will be susceptible to be adopted in the EU laboratories.
• To perform Pre-normative activities in order to address the short term materials needs of the ASTRID & MYRRHA projects with respect to the design and construction of the structural components. In particular, among others, the design rules for 9% Cr steel, material of the ASTRID steam generator, was provided together with the related mechanical design properties and the updated welding and fabrication rules.
• To optimise the effectiveness and efficiency of the EERA Joint Program on structural materials for innovative reactors, on the base of the knowledge produced by the material researchers and of the industrial current practices.
• To carry out the fabrication in Europe, by different methods, of ODS steels, which are under consideration as candidate materials, in the medium-long term, for fuel cladding application.
To achieve the aforementioned objectives, the R&D activities were focused on material testing, analysis of existing data and theoretical research. The relevant advantages of the MATTER approach were the access to a large amount of data, generated by national researches, and the wide comparability of results, thanks to the presence of 26 organizations including industries and steel makers. The large number of participating results and the specific competencies of the several different participating institutes. A relevant advantage of this approach consisted in the final comparability of the experimental data, being produced by consensual procedures, and their immediate availability in view of their pre-normative deployment.
This experience and approach will also be useful for the development of other nuclear systems such as
GFR and LFR, not exclusively MYRRHA and ASTRID. The obtainable results, related to the objectives of the Project, were:
- List of materials properties that are essential for the design of the innovative nuclear systems and selection of the more appropriate ones to be dealt with in the present Project;
- Testing procedures and best practice guidelines for the screening and characterization of the selected materials properties and conditions;
- Materials pre-normative qualification with respect to the selected properties and conditions;
- Contribution to update the design rules in consideration of the needs posed by the new reactors;
- Preparation of the EERA R&D Joint Program and first achievements.
Such a work have improved the interaction between European material research laboratories, established better link between designers of nuclear systems and material scientists and allowed a large dissemination of the obtained knowledge to the benefit of the end–users.
Project Results:
Materials for ASTRID and MYRRHA
The design of ASTRID, which uses sodium as coolant, relies to large extent on the design and operational experience from previous sodium fast reactors, in particular Phénix and Superphénix in France. Stricter requirements on safety and economy and sixty years design life raise a number of design issues described in [D1.1 and D1.3]. The ferritic-martensitic tempered steel Grade 91 (P91/T91) was considered as the reference materials for the steam generator and possibly for intermediate heat exchangers primarily due to its resistance to thermal fatigue. P91 softens under cyclic softening and the ratio between ultimate strength and yield is much lower than for 316 stainless steels. The Design Rules for 316 stainless steels could therefore not be used. Key objectives of WP4, 5 and 7 were therefore the development of Design Rules for ratcheting, creep-fatigue interaction and negligible creep for P91. Welds constitute weak spots in nuclear components and was addressed in WP6 for improved welding procedures and selection of filler materials for both 316 stainless austenitic steel and P91 ferritic-martensitic steel components as well as a proposal for new fatigue weld coefficients for P91. The 60 years design life means that additional design factors need to addressed and then in particular thermal ageing, irradiation and environmental degradation, which was addressed in WP5.
MYRRHA will use lead-bismuth eutectic as coolant, used in Russian nuclear submarines, but not implemented for other nuclear applications. Thus there is basically no operational experience to rely upon and test data is very limited. Moreover, heavy liquid metal coolants (lead or lead-bismuth) will interact with the structural materials and finding materials that sustain this environment is one of the key bottlenecks for the development of the HLM nuclear reactors. In MATTER Work package 2, test procedures were developed both for liquid metal embrittlement and corrosion. A comprehensive test programme was performed to enlarge the database needed for austenitic and ferritic-martensitic steels for the design of HLM reactors. This data provides some guidance on material selection, design rules and mitigation measures to ensure integrity and functioning of the components.
Projects involving design and construction of complex and innovative facilities such as the ESNII prototypes and demonstrators need to be dynamic and adjust to new requirements, changes in available funding and practical as well as more scientific issues that may arise. Hence the needs are a moving target. For ESNII the financial problems in the EU in the last years, in particular the Fukushima accident, have had a large impact on the ESNII projects. As a consequence the a number of modifications have taken place for material selection and component design and the projects are a bit delayed as compared to the plans in 2010.
Test Procedures and LBE degradation (WP2 WP3)
Access to material data representative of the operational conditions is necessary for the design of the ESNII reactors. For innovative reactor systems we can only partially rely on experience from earlier designs and we need to develop test methods and associated procedures to address both new and classical materials. We also need to assess properties and degradation mechanisms for known reference materials in the relevant conditions. This translates into needs for screenings tests of candidate materials as well more challenging tests where specific degradation mechanisms are targeted and assessed with state-of-the-art tools. For screening, miniature tests are of special interests as they are relatively simple to perform and they require small amounts of material. In MATTER the small punch test and nano-indentation were therefore evaluated. The heavy-liquid metal (HLM) coolants provide a corrosive environment and for the design and operation of MYRRHA and ALFRED we need to better understand the compatibility issues between HLM coolant and the structural materials to define material requirements and mitigation procedures. These two aspects were addressed in MATTER Work Package 2 and 3.
Miniature screening tests
Different miniature test methods were evaluated in Work Package 2 of MATTER: in particular small punch test but also macro- and nano-indentation impression creep.
In the small punch test a small clamped disk is subjected to a controlled displacement of force from a puncher. The disk has typically a thickness of 0.5 mm the diameter is typically 10 times larger. Estimates on different uniaxial material properties such as yield stress, ultimate strength, ductile-to-brittle transition temperature, creep rates can then be derived from measured deflection and the force. The deformation is bi-axial and a finite element analysis is generally needed for more detailed evaluation of local strain/stress levels. There are however simplified formulations and correlations for the estimates. The tests are rather simple to perform but the rather complex bending and stretching interaction of the tests complicates data the assessment. The test is clearly suitable for estimates of time to rupture in creep or tensile strength. The SPT has also been shown to be a powerful tool in screening and ranking testing.
In MATTER the SPT has been used to assess how irradiation and liquid metal embrittlement affect selected material properties. One example is where the SP energy is plotted against the temperature data for irradiated and unirradiated specimen. Due to limited data points and scatter, it is difficult to determine the small punch transition temperature, TSP , precisely. For the unirradiated specimens it is about ‐185 °C and ‐145 °C for specimens at 3.49 dpa. The temperature dependent yield stress obtained from the SP tests compared with results from tensile tests in a previous project SPIRE and literature data. The SP based data result in an overestimation of the yield strength in comparison to the values from the tensile test. A clear irradiation effect on the yield stress is observed.
In MATTER SPT was also used to assess the liquid metal embrittlement from LBE on P91 steel. For specimens tested in contact with LBE the energy drops is about 31% at 300 °C and about 22% at 400 °C as compared with specimens tested in air. For the plastic energy the decrease is even more: 35‐40% at 300 °C and 25‐28% at 400 °C. It strongly reflects the LME effect of LBE on the T91 steel in this temperature range. The results indicate also a better repeatability for SPT than slow strain rate tensile tests.
Nano-indentation was used in the MATTER project to assess irradiation effects. There is large interest in using ion irradiation to simulate neutron induced damage in nuclear materials for basic investigations as well as for materials screening. By Ion irradiation neutron activated material is avoided, and there is a tremendous reduction of irradiation times and improved capabilities to vary irradiation conditions. However, the question to what extent results based on ion irradiation can be transferred to of neutron irradiation is still debated. The transferability issue arises not only from different irradiation conditions but also from different testing methods due to the limited penetration depth of the ions. In MATTER the same heat of T91 was exposed to neutron and ion irradiation at the same temperature and the same nominal dose and tested by nano-indentation. In the project it was compared the hardness versus contact depth for un-irradiated, neutron and ion irradiated T91 material. There is very good agreement in the irradiation induced hardness change after neutron and ion irradiation at contact depth < 200 nm, where the substrate effect is negligible, i.e. the mechanical effect in terms of irradiation hardening of neutron and ion irradiation is the same in the present case.
Procedures for small punch test and nano-indentation have been evaluated and improved in MATTER. It has been demonstrated that the small punch test can be used as a screening test to assess irradiation damage and liquid-metal embrittlement and that nano-indentation could be used to assess change in hardness from neutron and ion irradiation and where ion and neutron irradiation give comparable results. These tests can be used for design, but perhaps a more important feature is that both tests require minimal material volume they can be used to monitor the material properties for components subjected to long-term operational conditions.
Test procedures and material degradation in LBE
For the design of MYRRHA, candidate materials need to be screened with respect to their resistance to degradation and failure in LBE environment, and in particular it is necessary to demonstrate sufficient resistance of selected materials against potential degradation mechanisms such as corrosion and liquid metal embrittlement (LME). This is a very complex task as the interaction between the LBE and the material depends on properties of the steel (e.g. chemical composition, microstructure and homogeneity) and the liquid metal (e.g. concentration of dissolved oxygen, temperature and flow velocity). Experimental data in the literature is scattered and often inconsistent. A major reason is that key parameters such as oxygen concentration are not well controlled and that different test procedures have been followed due to lack of test procedures. For design it is very important that generated data are reliable and reproducible. A large effort was therefore dedicated to the development of test procedures through round-robin testing. The test procedures were then used to derive material properties and mapping of degradation mechanisms which can be used as guidance for material selection and boundaries for the operational conditions.
Small strain rate tests (SSRT)
It was demonstrated that the small punch test can be used as a screening test for liquid metal embrittlement. The small strain rate test (SSRT) is probably the most common test for investigation of environmentally assisted cracking of materials. In SSRT a dynamic straining of the specimens is applied in the form of slow, monotonically increasing strain to failure. The dynamic strain requires less testing time in representative environment than static strain and it can therefore provide an early warning to potential problems.
In the project it was compared the measured total elongation versus temperature for T91 steel for different strain rates in both inert environment and LBE. The test show a clear reduction in ductility for tests in LBE, which is affected by both temperature and strain rate. Very clearly pronounced ductility trough is observed for the tests performed at strain rate 5•10-5 s and at 350°C.
For reliable assessment of LME it is necessary that the LBE and the solid are in full contact. The oxygen content in the LBE has a large impact and needs to be controlled. It can also be noted that there is a large scatter in total elongation but the stress-strain curves up to failure are very similar, which is a typical behaviour when elongation is controlled by micro-defects that provide crack initiation sites at the surface.
3.2.2 Fracture toughness tests
Given the reduction in ductility it is also natural to expect that the fracture toughness is affected by liquid metal embrittlement. To this end test procedures for fracture toughness in HLM were developed based on ASTM 1820 and fracture toughness curves were derived for P91 in LBE.
Indeed LME reduces the fracture significantly but to observe this effect requires that the specimen is pre-cracked in LBE. Hence pre-cracking in air will give non-conservative values of the fracture toughness of P91 in LBE.
Corrosion tests
The liquid metal corrosion can be both uniform and localized. As regards uniform corrosion a distinction is made between oxidation, dissolution and erosion. The formation of a thin self-healing oxide layer on the surface of the steel acts as protection against further corrosion, in particular dissolution. In MATTER a thorough investigation of corrosion between LBE and ferritic-martensitic steel (T91) and austenitic steels (316L and 15-15 stabilized steels) was done.
A thorough review of literature data (D3.6A) revealed a large scatter and inconsistencies in reported data. This is to a large extent caused by lack of control of key parameters, in particular the contents of dissolved oxygen. A first requirement for test procedures is therefore that all parameters are well defined and monitored and reported and in the case of dissolved oxygen content ideally also controlled. The MATTER test programme included the round-robin with the aim to develop test procedures as well as dedicated to determine corrosion data for the different steels under for different combinations of parameters (oxygen content, temperature, stagnant vs. flowing liquid metal). Tested samples were investigated by different material characterization methods and the observed corrosion was quantified. Details of the test results are reported in D3.6B and D3.6C. The temperature and the dissolved oxygen content are the two most important parameters. One outcome of MATTER was that to avoid localized dissolution the target dissolved oxygen content, Co , for MYRRHA has been increased from 10-8 to 10-6 wt% and the maximum temperature on the steel surface decreased from 500 to 470°C. In general the MATTER tests give consistent data that provides database for design of HLM cooled reactors. Various gaps to be addressed by further work have also been identified. In the project it was shown test data from literature and MATTER for different values of temperature and oxygen content for ferritic-martensitic steels tested for test period longer than 2000h and 4000h respectively. For the MATTER tests it can be noted that for the data points changed from oxidation to dissolution when the exposure time was increased from 2000h to 4000h.
For HLM cooled reactor systems the degradation of material properties from interaction with the coolant and the associated reduction on integrity and functioning of components is a key issue. Since there is hardly any return of experience it is necessary that material tests to support materials selection and characterization is an urgent need for the design and construction of MYRRHA and ALFRED. The work and the results in the MATTER project will clearly benefit these two systems:
The procedures developed for liquid-metal embrittlement and corrosion allow generation of reliable and consistent data. For liquid-metal embrittlement it is necessary that the surface of the solid is wetted and for corrosion it is very important that the oxygen content is monitored.
Both the fracture tests and slow strain rate tests show that P91 steel has a significant embrittlement in LBE and will therefore not be used for any load bearing components in MYRRHA. Test have been performed for austenitic steels following the same procedures and very limited LME is observed.
For corrosion oxygen content and temperature are the main parameters. Based on the results from MATTER, the target oxygen content has been increased and the maximum temperature has been lowered.
The new data have allowed a much better understanding of LME and corrosion and a first basis for design. Further needs have been clarified, such as long-term corrosion tests, have been identified and provide a basis for further collaborative work between ESNII and EERA JPNM.
Design Rules and Assessment procedures
Development of Design Rules and Assessment Procedures was the second domain in support of the design and construction of the ESNII reactors. Most of the work was devoted to 9%Cr steel grade 91 (P91/T91), whose mechanical behaviour is considerably different from austenitic steels contained in nuclear Codes and Standards (C&S). Thus there is a need to elaborate design rules as a final step before codification. The different 316 grades remain the reference for the vessel and the piping systems and all construction requirements need to be updated according to the current industrial practice and knowledge. Also, a service lifetime of 60 years is envisaged which needs an update of available design properties and an analysis of a possible reconsideration of ageing phenomena in the design.
RCC-MRx has been selected as the Design Code for all ESNII reactor projects and the work in MATTER is therefore specifically addressing revisions of this code. The R5 code developed in the UK is mainly meant as assessment code for components in operations, and was also addressed specifically. The work split into three main areas:
High-temperature degradation (creep, creep-fatigue, ratcheting and negligible creep ( WP 4and 5);
Other long-term degradation mechanisms, thermal ageing, irradiation and coolant compatibility (WP5)
Welding (WP6).
Supporting experiments for all areas were performed in WP7 (D7.1 and D7.4). An in-depth synthesis and recommendations for the MATTER Design Rules are given in Deliverable D4.8.
Design Rules for High-Temperature Assessment (WP4 and 5)
The ESNII reactors will generally operate at elevated temperatures and stresses will mainly result from thermal loads. The design rules for 316SS in RCC-MRx cannot be used for P91 due to the cyclic softening and the different yield to tensile strength ratio. In MATTER proposal for Design Rules for P91 with respect to ratcheting, creep-fatigue interaction and negligible creep have been developed.
Ratcheting Rules for P91
Ratcheting refers to progressive non-recoverable deformation under cyclic loading. Ratcheting is primarily a plastic deformation and occurs at lower temperature, but ratcheting increases at higher temperatures where creep effects also need to be taken into account. At low loads but high temperature we may have pure creep ratcheting. The work in MATTER included the development of visco-plastic constitutive models for more detailed simulation pf ratcheting under more general conditions as well as the development of efficiency diagram in accordance with RCC-MRx approach.
In RCC-MRx there are two basic rules for ratcheting: the 3Sm rule and the efficiency diagram. The 3Sm rule is applicable only under negligible creep and states that the sum of primary stresses and secondary stress range must be below three times the allowable stress.
The efficiency diagram used normalized forms of primary and range of secondary stresses. The actual diagram is purely empirical and gives the relationship between the secondary stress ratio, SR= Q/P, and the "efficiency index", v= P/Peff, where Q is the secondary stress range, P the primary stress and Peff is defined as the primary stress that in absence of cyclic loads using the monotonic stress-strain curve, would give the same cumulative strain as with a superposed cyclic primary and secondary load. A specific Peff is thus linked to a specific cumulative strain value that depends on the stress-strain curve. The efficiency diagram developed shows the data points for P91 from torsion-tension test presented in D7.1 together with the RCC-MRx efficiency diagram for 316L(N) and the proposed revision for P91. The increased ratcheting for P91 which is manifested by the reduction in the efficiency diagram is related to the cyclic softening. This efficiency diagram was also validated for non-isothermal large component tests.
Creep-fatigue interaction for P91
The Creep-fatigue assessment in existing design and assessment codes such as RCC-MRx, ASME III NH and BS-R5, is based on creep-fatigue interaction diagram where separate analyses are done for creep and fatigue. The axes represent usage factors for creep and fatigue respectively and the interaction is taken into account by the bi-linear diagram that defines the allowable region. The main difference in the codes lies the calculation of the creep damage e.g. life exhaustion or different ductility exhaustion. In the MATTER project it is shown that the experimental points of the diagram may also differ but that does not necessarily correspond to difference in safety margins, as it is also a consequence of the difference in how creep damage is modelled. The creep-fatigue interaction diagram has worked well for austenitic steels but for ferritic-martensitic steels there is a large scatter and a failures have been observed in the safe region. This is believed to be caused by cyclic creep softening, where pre-cycling significantly increases the creep rate.
Since the creep-fatigue interaction diagram is based on normalized data it does not distinguish between data with different hold times or number of cycles nor for load sequence effects. The objective of the creep-fatigue part of the MATTER project was therefore to evaluate the interaction diagram for P91 and propose modifications as well as propose and evaluate alternative methods.
One conclusion is that the interaction diagram as proposed in RCC-MRx, seems to provide sufficient conservatism also for P91 for available test data under strain control where all data points are well above the interaction diagram. A major reason is that the design fatigue curve is used, i.e. allowable number of cycles is defined by halving the strain range of the mean LCF curve and by dividing the mean curve number of cycles by a factor of 20. The increased creep rate from cyclic softening is not accounted for but partly compensated by increased stress relaxation. It should be remembered that almost all data are with short hold times and high loads and extrapolation to long hold times is not obvious.
In the MATTER project two alternative simplified were also investigated: the Manson-Halford model and the φ–model based on the assumption that accumulated creep life can be derived from a normalized reference stress defined by strain range, hold time and temperature. These models are not influenced by the material softening (or hardening) and have much fewer fitting parameters than any of the currently used interaction diagram methodologies. If considering the simplicity of use and the clear advantage in total amount of fitting constants needed, it is clear that the simplified methods have an advantage compared to the current code assessment procedures. The simplified models are successfully predicting the JRC creep fatigue test data for temperatures 550 and 600°C. The number of cycles to failure versus the total strain range (creep and fatigue) was evaluated in the porject. The model has been calibrated with available LCF and CF data. Predictions up to 1 h hold time are supported by data, whereas longer hold times are extrapolations. The reduction in number of cycles can be directly inferred from the reduction in fatigue curve.
Negligible creep for P91
The negligible temperature curve (TNEC) is defined as the temperature that causes a "negligible" prescribed creep strain, relaxation or life fraction in a specified time at the specified reference stress. In MATTER a creep strain limit of 0.2% was used. Thus, the negligible creep domain is the limiting temperature/time within which designers do not need to consider creep, i.e. if the service time is below the negligible creep curve, and this simplifies the design analysis significantly. RCC-MRx contains negligible creep curve for austenitic steels but not for P91. Although the concept seems trivial, the approach to determine TNEC is not straightforward. First of all the reference stress and the value of the strain limit is material dependent and may differ between the codes. Moreover, supporting experimental data should be ideally be very long-term and at low temperatures and stresses. In reality we need to rely on extrapolation of creep curves to negligible creep. In MATTER very little low temperature data was available in the public domain or from the project partners so a test programme was initiated and performed at temperatures close to the low temperature limit for time independent design. The stresses were chosen to produce the sought 0.2% strain in about 5000 h. The required stresses are relatively large, situated on both sides of the RCCC-MRx defined reference stress of the ultimate tensile strength divided by a factor of 2.7.
The approach was based on the Wilshire creep model which was calibrated to both generated and public domain strain data to present normalized stress versus temperature compensated time for different life fractions of rupture life or different creep strain values. The Wilshire equations can then be inverted to compute the negligible creep temperature versus time, for a given strain limit and reference stress. The negligible creep curve for 0.2% creep strain together with the time to failure (rupture) curve and the corresponding time to 0.2% strain prediction using the RCC-MRx creep model. The plot also shows the creep rupture curve with a scaling factor of 1000, which almost coincides with the 0.2% creep strain curve. Thus in this case a 0.2% creep strain criterion corresponds to a safety factor of 1000 against creep rupture. It is worth mentioning that the negligible creep curves derived in MATTER assume pure creep. Under cyclic loading the negligible creep curve may be reduced.
Robust Design Rules based on sound mechanical basis are crucial for the design of the ESNII reactors. The proposed Design Rules for ratcheting, creep-fatigue and negligible creep should be reviewed by AFCEN for inclusion as probationary rules in a first stage.
The rules for ratcheting and negligible creep follow the same principles as the Design Rules for austenitic steel that are already included in RCC-MRx so it should be straightforward to include the proposed rules for P91.
For creep-fatigue the situation is a bit different since the interaction diagram, which is established in RCC-MRx and other codes, seems to provide sufficient safety margin for the available data. But the interaction diagram gives very limited insight and includes more ad-hoc assumptions and is more questionable for extrapolation to long hold times.
All rules need also to be validated for loadings more representative for real operational conditions, in particular long hold times and low stresses.
The loads in Generation IV reactors are primarily thermal. Since the creep properties are strongly creep dependent whereas the Design Rules are based on iso-thermal conditions the applicability of the Design Rules in non-isothermal conditions need to be validated. Future work may be needed to develop the Design Rules to non-isothermal conditions.
Design Rules for other long-term degradation mechanisms (WP5)
The designers of Generation IV reactors need to demonstrate that non-replaceable components retain their integrity and functioning for at least 60 years, compared to 40 years in past projects. This requires that the time line of material data and design rules need to be extended and additional slow degradation processes need to be taken into account. In the MATTER project design rules were addressed for three types of processes: thermal ageing, irradiation and environmental effects from heavy liquid metal.
Thermal ageing
Thermal ageing is related to the stability of the material microstructure at service temperature where a variety of processes may occur, including plain grain growth of the material matrix, coarsening of particles, segregation of minor material constituents, dissolution of precipitates and formation of new phases not present in the initial microstructure. As mass transport (diffusion), chemical reactions at interfaces, particle nucleation are involved, the progress of processes constituting thermal ageing strongly depends on time. At LMP of 22 corresponds to 110, 60 and 10 thousand hours at 550, 600 and 650 °C respectively. Simultaneous action of mechanical load and deformation of the material influence thermal ageing and induce additional processes like dynamic recovery or re‐crystallisation. Since tempered ferritic-martensitic steels are in a meta-stable phase it is that thermal ageing may play a larger role than for austenitic stainless steels.
Thermal ageing is difficult to include in design codes due to the complexity of the processes but also since there is no simple way to perform representative tests in reasonable time frames. Tests at elevated temperatures have been proposed but such data will never be truly representative. The only realistic approach is to perform tests on components that have been exposed to operational conditions for a long time. Thermal ageing may reduce the ductility, fracture toughness and increase the creep rate.
In design codes the ageing is therefore handled by reduction factors. In In RCC-MRx provision for thermal ageing coefficient for allowable stress for assessment of P-damages (static load) is included, but actually listed for very few materials and no such factors are for instance provided for 316L(N) or P91 steels. It could be noted that ASME, section III, Division1, subsection NH provides ageing factors. However for 316L(N) welded components fracture toughness curves are provided in RCC-MRx for moderate (e.g. 25,000h at 550°C) and advanced (e.g. 110,000h at 550°C) thermal ageing. Thermal ageing depends to a large extent on the chemical composition and microstructure and the codes, such as RCC-MRx provides tight specifications which should limit thermal ageing.
Given the lack of experimental data mechanistic approaches that incorporate the microstructural features were evaluated in MATTER. Experimental data suggest that creep life can be minimum creep rate (Monkman-Grant relationship). So the effect of thermal ageing on creep may be accounted for if we can determine the minimum creep rate , which potentially can be done using the Dyson mechanistic creep model based on mechanistic models where the parameters can be derived from measured microstructural evolutions and relatively simple physics-based models. However no concrete proposals for inclusion in the design code could be provided in MATTER.
Irradiation effects
The general strategy of considering irradiation effects in design codes and standards for structural components is to first make a classification in terms of the expected significance for the component performance over the service time, i.e. (a) negligible; (b) significant and to be considered; (c) maximum allowable irradiation [38]. Respective thresholds of irradiation fluence or resulting accumulated damage on the atomic scale are material dependent, and may vary with the temperature range under consideration. If a significant impact on strength properties is to be expected, it needs to be either confirmed that design curves provided in the code still provide a sufficient safety margin or further reduced strength properties assumed. No suggestion for revision of irradiation effects were proposed in MATTER.
Heavy Liquid metal effects
As already mentioned the design of nuclear reactors with HLM coolants has very limited return of experience to rely on. The report D4.7 gives an excellent survey and assessment of the limited experience from Soviet and Russian projects and LBE facilities in Europe. Moreover Design Rules for HLM reactors have not been developed. Different approached for environmental factors are provided in the ASME codes as regards e.g. corrosion/erosion allowance, stress corrosion cracking and environmental fatigue. These are more guidance and allowance values rather than design rules. RCC-MRx has even less guidelines on environmental effects, but RCC-MRx is nevertheless considered as the more suitable code for the development and codification of Design Rules as it is based on return of experience from sodium cooled fast reactors and includes rules and guidelines for e.g. high temperature applications and specific nuclear components. The deliverable D4.7 provides a framework for how LME and LMC can be included in a Design Code.
The ensure integrity the designer makes three implicit assumptions (D4.7):
The chosen material, in the chemistry controlled coolant, has the same behaviour as in air except for the depletion of the corrosion allowance;
The ISI schedule will detect flaws, local corrosion, local thinning, before defects become critical, assuming a limited corrosion speed;
The range of composition of coolant stays within the request margin all over the systems
Failure of the chemistry control system that could cause high speed corrosion are detected and the grace time is sufficient to take corrective action.
The first one, in line with the first implicit assumption, is to use only materials that are not susceptible to LME or LMC in heavy liquid metals or where the measured effect is negligible. The challenge for this approach is to define criteria of the negligibility. The advantage of this approach is that no modifications of the design code are required and the experimental program should be done to demonstrate that heavy liquid metals have no significant effect on the design mechanical properties. The disadvantage is that the number of available materials might be very limited since from screening tests it seems that all tested chromium steels as well as high-strength steels are susceptible to LME. Therefore, due to the screening-out of these materials may reduce the operational parameters of the developed installations. From available data 316L could satisfy the requirement but ferritic-martensitic steel would clearly be ruled out.
The second approach is to tolerate susceptibility of the material to LM degradation by imposing additional margins on the mechanical properties used in the design. The advantage of this approach is that it does not exclude susceptible materials from the list of available materials. This approach will require quantification of how the mechanical properties depend on various parameters and to define conservative reduction factors and to validate these from tests representative for the operational conditions. The challenge in the dissolution attach versus exposure times for solution annealed and cold drawn batches of 316L steel is that the standard procedures for such tests are not readily available and their development requires significant collective efforts. The design factors should not be too large and they must be robust. The ferritic-martensitic steel would in practice be ruled out also by this approach.
The third approach would be to develop mitigation strategy to prevent LM degradation. It can be either tailoring the chemical composition of steels or coolant chemistry. The importance of having the oxygen content within a narrow range for both LME and LMC was demonstrated in Section 3. This approach requires very good understanding of the underlying processes for LME and LMC supported by an extensive database. One problem with this approach is to ensure that the coolant chemistry remains within the tight limits in all parts of the system.
Taking account of time dependent and environmental induced degradation mechanisms constitutes the toughest challenge for the design of Generation IV nuclear reactors. This is due to the complexity of the processes, their dependence on operational conditions. Another problem is to replicate the conditions at laboratory scale. These problems are also why Design Rules are not developed or very limited.
The concrete proposals in MATTER for Design Rules for thermal ageing were very limited. The main reason to this was the lack of relevant data. Progress in this requires testing of specimen taken from components that have been exposed to long-term operational conditions. This can then be combined with laboratory test of specimen with simulated ageing conditions. Design curves or design factors could then be derived by extrapolation of laboratory data to operational conditions and time scales using physics based models. Such an approach would require a close integration and collaboration between scientists, designers and operators.
For the design, construction and licensing of HLM cooled reactors it is necessary to demonstrate that components retain their integrity under operational conditions. The development of guidelines, design rules and mitigation approaches is therefore an urgent need. The results in MATTER constitute a very important step as it has demonstrated that P91 is not a suitable material, confirmed that 316L appears to be resistant to HLM degradation, and tightened the limits for oxygen control.
Since there is no return of experience to rely on the design need to a large extent rely on new test data. The MATTER deliverable D4.7 gives proposed a clear way forward. To generate the supporting data and methodology will require a test programme that first confirms the resistance of 316L to HLM degradation. Such a programme should include welded components. In parallel other candidate materials need to be screened and models need to be developed both at the engineering and more fundamental scale.
Design Rules and Manufacturing Rules for welded components (WP6)
Most failures in nuclear components occur in welds due to a combination of geometrical factors, variations in microstructure, micro defects and residual stresses. The production, characterization and integrity assessment of welds is very complex. Manufacturing rules and design rules to ensure that welded joints fulfil the strict safety requirement is crucial for the design and construction of the ESNII reactors. In the MATTER project the work on welds focussed on two areas: i) development of fatigue weld factors for P91 and 316 stainless steels components and different welding techniques and ii) development of welding procedures for P91 and 316L(N) steels.
Fatigue weld factors
In Design Rules and assessment procedures, safety of welded components is generally addressed by weld reduction factors that lumps the various factors mentioned above. In MATTER updates of fatigue weld factors for RCC-MRx and R5 were done by experiments and modelling.
In RCC-MRx the reduction in fatigue life for welds compared to base material is accounted for by the "joint fatigue coefficient, Jf", defined by the ratio of strain variation for base material and weld joint for a given fatigue life (J_f= ∆ε_BM/∆ε_wj). The joint fatigue coefficient depends on the base material, filler material and weld process. For 316L(N) with narrow-TIG weldments Jf is 1.25 but for P91 there are no values in Rcc-MRx. In MATTER joint fatigue factors were derived for P91 with TIG weldments. From tests of P91 welds failure is in the HAZ for low strains but in the base metal for high strains, which also follows from nonlinear FE-analyses. As a consequence, Jf is a function of the strain with proposed values 1.35 for strain ranges below 0.25%, 1.55 for strains above 0.55%, and a linear interpolation for intermediate values.
In MATTER joint fatigue coefficients were also assessed for 316L(N) and P91 with hybrid laser welds. For 316L(N) it was concluded that the weld coefficient for TIG (1.25) was applicable.
The basic approach for fatigue assessment of welded components in R5 is similar to RCC-MRx, but instead of the fatigue joint coefficient they use Fatigue Strength Reduction Factors (FSRF). In MATTER a refined method where the weld factor is separated into two: a Weld Strain Enhancement Factor (WSEF), and Weld Endurance Reduction (WER) derived from the material mis-match. The combination of WSEF and WER reduces the conservatism compared to FSFR for the creep effects. Comparison with tubular weldments of 316H validated that the revised procedure was still conservative.
Welding procedures and code requirements: 316SS and P91
For P91 RCC-MRx have data sheet for chemical composition and mechanical properties for the base material but not for the filler material. In MATTER different filler materials were evaluated in combination with welding methods with the objective to qualify welds for class 1 nuclear components. Three complementary welding technologies were included: fully mechanized GTAW Narrow Gap default method, Hybrid Laser Arc (HLW) and Shielding Metal Arc Welding (SMAW). Although the emphasis was on P91, also 316L(N) weldments were included. It was demonstrated both filler materials for 316L(N) were acceptable. For P91 SMAW and GTAW are promising but elongation is on the low side and softening on the HAZ remain as concerns.
The integrity of welds is a key issue for the design of all ESNII reactors. The development of fatigue weld factors, as well as the assessment of new filler materials and welding procedures, are of direct relevance for ESNII. The proposals form MATTER are should be incorporated into the Design and Assessment codes (RCC-MRx and R5) but some additional validation by experiments is needed in particular with respect to creep and high-cycle fatigue.
Material Database for integration of European materials research
Reliable data on material properties and long-term degradation is necessary for the design and long-term operation of nuclear power plants. The generation of such data is very costly as it requires special facilities, experts and requires very long durations. Moreover it is very important to have detailed information of data (e.g. raw data, test conditions, chemical composition etc) to use the data efficiently as exemplified for test data related to heavy liquid metal degradation. Much data exist in Europe, often generated as part of EURATOM projects, but very little of this data is in a form that it can be shared. It would therefore clearly of great advantage for the nuclear material research community and the reactors designers to build-up a common web enabled database. In fact this would represent a natural step towards the integration of the nuclear material research in the EU on line with the EERA JPNM Vision paper. JRC has proposed to use their web based database MatDB for this purpose and in MATTER it was agreed to test this. At the end of the MATTER project a total of 321 validated data sets for P91 and AISI 316 had been uploaded by 8 project partners to the JRC web enabled database MatDB. The uploaded data included: load and strain controlled low-cycle fatigue, small punch tests, uniaxial creep, uniaxial tensile, creep crack growth and fracture toughness data. The data sets provided by the partners are therefore accessible for the MATTER organizations.
Furthermore, it is only natural that there is a reluctance to share data when the only beneficiary will be a third party that has invested no effort in its generation. To resolve the former, funding agencies need to ensure that data plans are a mandatory part of research proposals and that data sets are recognized project deliverables. For the latter, a mechanism is needed whereby efforts to share data are acknowledged and rewarded. Datacite DOIs offer just such a mechanism, enabling data sets to be published and cited in the same way as conventional publications. The ODIN portal at https://odin.jrc.ec.europa.eu has recently been enabled to use DOIs for data publication. The owners of the data sets in MATTER were therefore offered to have their data sets with DOI.
Data is the key output of a project that can be exploited by project partners for future use. Ensuring reuse of data generated in a project is perhaps the most important output of a project. In MATTER the first attempt was made to have generated so that it could be reused. As a result 321 validated sets from the project are available for the MATTER projects. Depending on the restriction agreed by the MATTER consortium these data can also be by ESNII designers. Data that have been classified as restricted is also available for the broader research community.
Oxide Dispersed Strengthened Steels
The main outcome obtained in the project were related to:
Process Fabrication Characterisation
Development of a production route for ODS steels, which is based on conventional steel making techniques (introducing the oxide particles in the melt). (OCAS)
Manufacture of an ODS steel bar for laboratory testing by powder metallurgy (CSM, CNR-IENI)
Improvement of the powder metallurgical process (including MA) and Optimization of the final thermo-mechanical treatment (KIT)
Elaboration of ODS steels through spark plasma sintering techniques and scale lab heats to an industrial quality production (CEA - INSA)
Production of ODS steels using Spark Plasma Sintering (SPS) and assessment of microstructure and basic properties. (HZDR)
investigated and assess few key points of the fabrication routes of ODS
Develop a fast and simple screening test to check the effect of MA at least in the initial NFA exploratory phase. (ENEA).
Different recommendations were proposed in the project for the development of ODS materials :
The mechanical alloying is still the easiest way to produce ODS materials (CSM, CNR-IENI, ENEA, HZDR, KIT). It is relatively easy to obtain products at a laboratory scale and now the effort should be geared towards large scale production.
The Elaboration of ODS steels through SPS (Spark Plasma Sintering) techniques has been demonstrated and can be compared to HIP Process, large scale production and gains on the productivity have to be evaluated.
It has been shown that the recrystallization is a key issue for the control of the microstructure, the amount of nano-phases and the control of the cold working are the main parameters to reach an isotropic microstructure.
Two deliverables were produced:
Deliverable 10-1, Key "technological" issues for ODS materials
Final report on scientific understanding of ODS production, ranking and recommendations : in progress, should be ready for the end of November
Refractory alloys, Ceramics and Composites
These materials are primarily required in GenIV systems to access operating temperatures well beyond the current inherent limits of the most heat resisting super-alloys. Besides their refractoriness, the ceramic-based families have the additional advantage of chemical inertness and are envisioned as a viable alternative to conventional and advanced steels for their promise of greater resistance to corrosion/erosion in the fast reactor liquid metal coolants. The objective of WP11 of the MATTER project was to support the development of an implementation plan to be run during JPNM first phase 2011-2015, in order to provide the necessary background for the follow-up activities ( i.e. JPNM second phase 2016-2020).
At M12, a list of priority refractory/component systems was identified as well as their nearest-term issues, that were described in D11.1 and were progressively converted, starting from M24, into specific topics to be addressed in the form of separate pilot projects (MS14) in order to prepare the basis for proposal to EC calls. Status at month 30 of SP3 consolidation process was as follows:
1- Nuclear-grade SiCf/SiC composites with improved irradiation resistance and high temperature capability have been developed in the Fusion Materials Programs for the self-cooled DEMO blankets. These materials have most of the attributes required for GFR fuel cladding and they represent the leading option for such an application. The proposed activities cover the most immediate feasibility issues associated with the development of a leak-tight clad made of nuclear grade SiCf/SiC for GFR. The target of this proof-of-concept study is to contribute at progressing current TRL~3 to a stage (~4) at which few optimized SiCf/SiC clad systems could be provided in prototypic format for more quantitative assessments in integrated prototypic GFR environments.
2- MAX-phase are emerging materials that combine properties of metals and ceramics but do not benefit of any feedbacks from previous experience. Owing to their excellent machinability and promising corrosion/erosion resistance in lead alloys, they are explored as an innovative technology for the pump impeller in LFR (current TRL~1-2). A screening work based on a set of well-selected key criteria has been proposed to provide answer about the potential of the MAX phase concept for the given application and guidance for the follow-up.
3- Advanced V-alloys with an increased operating temperature window are under development in the US and in Japan for the fusion reactor blankets. Current candidates with reference composition of V-4Cr-4Ti have also attractive features for application as fuel cladding materials in fast reactors (current TRL~2-4 depending on the fast system end-use). Most pressing feasibility issues to be fully addressed in the scoping phase for maturing these alloys as cladding materials are mainly related to fabrication, processing and coating technology capabilities and performance assessments that are of cross-cutting interest for the fast reactor technologies.
The additional reports produced in the frame of the project are attached to this final report.
Potential Impact:
In the frame of MATTER project the following Workshops were performed:
- Final MATTER International WorkShop, “Rome, Italy, October 20-22 th, 2014. The objective of this WORKSHOP is to supply the final results of the MATTER Project, in order to complete the materials researches, in the frame of the EERA guidelines, with the implementation of pre-normative rules (RCC-MRx, R5, etc.) and new consensually developed testing standards.
- International Workshop on the “Key material properties for Myrrha and Astrid”, Rome, Italy, March 7-9 th, 2012. The first international workshop on the “Key material properties for Myrrha and Astrid” aiming at outlining the most relevant key material properties for both ASTRID and MYRRHA design.
The following Training Course were carried out:
- International School on Materials UNder Extreme Conditions (MUNECO), “Madrid, Spain, June 11-15, 2012. To strengthen the basis of knowledge in nuclear materials science, to promote the transfer of experience on structural materials for GENIV towards technological applications to young generations of scientists and engineers, the GETMAT and MATTER FP7 projects organize MUNECO school.
- International School on " DEsign Rules for gen IV reactos and INnovative reactors" (DERIVIN), INSTN Saclay, French, July 1-5, 2013. The objective of this school is to train European students and PhD applicants on the significance of design rules for nuclear applications, presenting them as living documents which are subjected to review and adapt to the evolving technologies. The attending students, after this Course, will have a better knowledge about the RCC-MRx and R5 rules for design purpose and know in detail some aspects related with the nature of the different materials.
In the frame of MATTER project the following coordination meeting were carried :
- 1st PCC video meeting 1st April, 2011 -
- 2nd PCC video meeting 21st September, 2011 -
- 1st WP2 General meeting 09th-10thMay, 2011 Paul-Scherrer-Institute Villigen (Switzerland)
- 1st WP3 Coordination Meeting 14th-15th September 2011 Headquarters SCK•CEN
- Brussels
- 1st DM2 Technical Meeting- WP4.1-4.2-4.3-6.1-7.1-7.3 03rd-04th May, 2011 Paris
- 2nd DM2 Technical Meeting- WP4.1-4.2-4.3-6.1-7.1-7.3 30th November-1st December, 2011 Paris
- MATTER-DM2-WP 10 Meeting 15th November, 2011 Saclay
- 1st Governing Council meeting 09th February, 2011 Bologna
- MATTER Kick Off meeting 09th -10th February, 2011 Bologna
- DM4 Coordination meeting 24th March, 2011 Casaccia (Rm)
- WP15 Coordination meeting 08th September, 2011 Casaccia (Rm)
- Coordination meeting with D2 leader 02nd December, 2011 Paris
- 3rd PCC meeting 13rd February, 2012 ENEA, Bruxelles
- Rue de Namur, 72-74 - 1000 Bruxelles
- 4th PCC video meeting 6th July, 2012 -
- 5th PCC meeting 6th February, 2013 Centrum výzkumum Řež
- Hlavní 130, 250 68 Husinec – Řež ,Czech Republic
- 1st Plenary meeting 13rd-15th February, 2012 ENEA, Bruxelles
- Rue de Namur, 72-74 - 1000 Bruxelles
- 2nd Plenary meeting 6th -8th February, 2013 Centrum výzkumum Řež
- Hlavní 130, 250 68 Husinec – Řež ,Czech Republic
- 2nd Governing Council meeting 14th February, 2012 ENEA, Bruxelles
- Rue de Namur, 72-74 - 1000 Bruxelles
- 3rd Governing Council meeting 8th February, 2013 Centrum výzkumum Řež
- Hlavní 130, 250 68 Husinec – Řež ,Czech Republic
- DM1 meeting 4th February, 2012
- DM2 meetings 18th and 19th of June 2012 Paris France
- DM2 meetings 19th and 20th of November 2012 Paris France
- DM2 meetings 28th of June 2013 Paris France
- Task 3.1: Fracture Toughness in LBE meeting 5th February 2013 CVR, Řež, Czech Republic
- Task 3.2: Liquid Metal Corrosion meeting 24 th -25 th September, 2012 KIT Campus North, Karlsruhe, Germany
- WP5: Technical meeting with AEKI 6 th -8 th February, 2013 Paris France
- WP5: Technical meeting 2nd April, 2013 video
- Task6.2 and 6.3: technical meeting 22nd May, 2012 Rome
- WP10 meeting 26th of June 2012 Dresden, Germany
- WP10 meeting 6-8th of February, 2013 Prague, Czech Republic
- WP2 – third Technical meeting September 30 ‐ October 1, 2013 Espoo, Finland
- WP2 – fourth Technical meeting September 16-17, 2014 CIEMAT, Madrid (Spain)
- WP3 Technical Meeting November 26-27, 2013 NRG, Petten, The Netherlands
- WP3 Technical Meeting June 26-27, 2014 CIEMAT, Madrid, Spain
- WP3 Technical Meeting November 27-28, 2014 SCK•CEN, Mol, Belgium
- MATTER DM2 FINAL MEETING July 17-18, 2014 Paris, France
- WP7 technicla meeting June 26 – July 7, 2013 Paris, France
- 3rd MATTER Plenary Meeting January 20th – 23rd, 2014 Hotel Freienhof, Thun, Switzerland
- MATTER coordination Meeting November 04, 2014 Rome, Italy
The following paper were produced in the frame of the project:
(1) M. Jong, N.V. Luzginova, H.S. Nolles, T. Bakker, “Liquid metal embrittlement testing in heavy metals; development and first fracture mechanics test results”, Oral Presentation at the IAEA Technical Meeting on Liquid Metal Reactor Concepts: Core Design and Structural Materials, 12-14 June 2012, Vienna, Austria; Paper for the Meeting Proceedings is under review
(2) C. Schroer, O. Wedemeyer, J. Novotny, A. Skrypnik, J. Konys, "Compatibility of ferritic/martensitic steels with flowing lead-bismuth eutectic at 450°–550°C and 10–6 mass% dissolved oxygen", Proceedings of the 20th International Conference on Nuclear Engineering collocated with the ASME 2012 Power Conference (ICONE20-POWER2012), 30 June – 3 August 2012, Anaheim, CA, USA
(3) K. Lambrinou, V. Koch, G. Coen, J. Van den Bosch, C. Schroer, "Characterization of oxide scales resulting from the exposure of various steels to liquid lead-bismuth eutectic", Poster Presentation at the 11th International Workshop on Spallation Materials Technology (IWSMT-11), 5-9 November 2012, Ghent, Belgium
(4) K. Lambrinou, V. Koch, G. Coen, J. Van den Bosch, C. Schroer, "Corrosion scales on various steels after exposure to liquid lead-bismuth eutectic", under review for the Journal of Nuclear Materials
(5) Holmström, S., Pohja, R., Nurmela, A.; Moilanen, P., Auerkari, P, Creep and creep-fatigue of stainless steel 316. 6th International Conference on Creep, Fatigue and Creep-Fatigue Interaction (CF-6), Jan. 22-25, 2012, Mamallapuram, India (2012).
(6) Holmström, S., Auerkari, P. Predicting creep life from small sets of test data and reference material characteristics. 12th International Conference on Creep and Fracture of Engineering Materials and Structures, Creep 2012, May 27 - 31, 2012, Kyoto, Japan (2012)
(7) “Pohja, Rami; Nurmela, Asta; Moilanen, Pekka; Holmström, Stefan; Multifunctional high precision pneumatic loading system (HIPS) for creep-fatigue testing, Procedia Engineering, Elsevier, Vol. 55 (2013), 573 – 577.”
(8) K. Lambrinou, V. Koch, G. Coen, J. Van den Bosch, C. Schroer, Corrosion scales on various steels after exposure to liquid lead-bismuth eutectic", Journal of Nuclear Materials 450 (2014) 244- 255
(9) V. Tsisar, C. Schroer, O. Wedemeyer, A. Skrypnik, J. Konys, “Corrosion behavior of austenitic steels 1.4970 316L and 1.4571 in flowing LBE at 450 and 550°C with 10-7 mass% dissolved oxygen”, Journal of Nuclear Materials 454 (2014) 332–342
(10) V. Tsisar, C. Schroer, O. Wedemeyer, A. Skrypnik, J. Konys, “Compatibility of ferritic/martensitic steel T91 with flowing LBE containing 10-7 mass% dissolved oxygen at 450 and 550°C”, Proc. of the 2014 22nd International Conference on Nuclear Engineering ICONE-22, July 7-11, 2014, Prague, Czech Republic (paper-30399)
(10) V. Tsisar, C. Schroer, O. Wedemeyer, A. Skrypnik, J. Konys, “Corrosion behavior of austenitic steels 1.4970 316L and 1.4571 in flowing LBE at 450 and 550°C with 10-7 mass % dissolved oxygen”, International Conference on Nuclear Engineering ICONE-22, July 7-11, 2014, Prague, Czech Republic; oral presentation
(11) V. Tsisar, C. Schroer, O. Wedemeyer, A. Skrypnik, J. Konys, “Corrosion response of 1.4970 316L and 1.4571 austenitic steels to flowing LBE with 10-7 mass % dissolved oxygen at 400, 450 and 550°C”, 12th International Workshop on Spallation Materials Technology, 19-23 October 2014, Bregenz, Austria; oral presentation
(12) A. Heinzel, A. Weisenburger, G. Müller, “Corrosion behavior of austenitic steels in liquid lead bismuth containing 10-6 wt% and 10-8 wt% oxygen at 400-500°C”, Journal of Nuclear Materials 448 (2014) 163–171
(13) O. Klok, K. Lambrinou, S. Gavrilov, I. De Graeve, “Preliminary results on the temperature dependence of the dissolution corrosion behavior of 316L austenitic stainless steels in contact with static LBE”, Proc. of the SEARCH/MAXSIMA 2014 International Workshop, pp. 645-657, 2014, Editors: J. Pacio, A. Weisenburger, Th. Wetzel; oral presentation
(14) K. Lambrinou, E. Charalampopoulou, R. Delville, T. Van der Donck, “On the dissolution corrosion of austenitic stainless steels in contact with static LBE”, Proc. of the SEARCH/MAXSIMA 2014 International Workshop, pp. 659-664, 2014, Editors: J. Pacio, A. Weisenburger, Th. Wetzel; oral presentation
(15) K. Lambrinou, “Introduction to liquid metal corrosion effects for Gen-IV LFRs”, Spring School on “Modelling of Corrosion for Nuclear Reactors”, Saclay, France, 31 March – 4 April 2014; invited lecture
(10) K. Lambrinou, P. Hosemann, “Factors affecting the dissolution corrosion of 316L austenitic stainless steels in contact with static LBE”, 2015 TMS Annual Meeting and Exhibition, Orlando, Florida, USA, 15-19 March 2015; oral presentation
(16) D. Frazer, M. Popovic, C. Cionea, S. Parker, C. Rose, O. Panova, M. Asta, M. Caro, S.A. Maloy, F. Rubio, K. Lambrinou, E. Stergar, P. Hosemann, “Heavy liquid metals as potential heat transfer fluids in nuclear reactors”, 2015 TMS Annual Meeting and Exhibition, Orlando, Florida, USA, 15-19 March 2015; oral presentation
(17) C. Schroer, “Oxidation and solution of steels in liquid lead–bismuth eutectic (LBE)”, Proc. of the 19th International Corrosion Congress, Jeju, Korea, paper no. TB 1-3 (Corrosion Science Society of Korea (CSSK) 2014)
(18) C. Schroer, O. Wedemeyer, J. Novotny, A. Skrypnik, J. Konys, “Compatibility of ferritic/martensitic steels with flowing lead–bismuth eutectic at 450°–550°C and 10–6 mass% dissolved oxygen”, Proc. of the 20th International Conference on Nuclear Engineering and the ASME 2012 Power Conference, Anaheim, California, paper no. ICONE20Power2012-54277 (ASME 2012)
(19) C. Schroer, O. Wedemeyer, J. Novotny, A. Skrypnik, J. Konys, “Performance of 9% Cr steels in flowing lead–bismuth eutectic at 450 and 550°C, and 10−6 mass% dissolved oxygen”, Nuclear Engineering and Design 280 (2014) 661–672
(20) G. Barbieri, M. Cesaroni, L. Barbieri “SALDATURA SMAW E GTAW DI P91: Qualifica dei materiali d’apporto e processi per la saldatura di componenti per reattori GEN IV in accordo alla normativa RCC MR”, Giornate Nazionali di Saldatura 8.
(21) X. Boulnat, I. Hilger, J. Hofmann, C. Testani, F. Bergner, Y. De Carlan, F. Ferraro, A. Ulbricht , “Comparative study on ODS 14Cr steel fabrication”,
A special issue of journal of nuclear material is under peer review with the main paper presented during Final MATTER Workshop:
(1) S.Gavrilov (SCK-CEN), Slow Strain Rate Test for investigation of materials susceptibility to Liquid Metal Embrittlement
(2) E. Altstadt (HZDR), Critical evaluation of the small punch test as a screening procedure for mechanical properties
(3) M. Serrano (CIEMAT), Liquid metal embrittlement assessment of T91 in contact with LBE by using small punch tests
(4) C. Heintze (HZDR), Ion irradiation and nanoindentation as a screening test for irradiation effects on neutron-irradiated ferritic/martensitic high-Cr steels
(5) C. Schroer (KIT), Investigating steel corrosion caused by liquid lead or lead-bismuth eutectic
(6) C. Schroer (KIT), A. Gessi (ENEA), Round robin test of steel corrosion in oxygen-containing lead-bismuth eutectic
(7) Lida Magielsen (NRG), Fracture toughness testing in Liquid metal
(8) J. Klecka (CVR), Character of liquid metal embrittlement of the steel T91 in contact with lead-bismuth eutectic
(9) Michael Mahler, Jarir Aktaa (KIT), Prediction of fracture toughness based on experiments with sub-size specimens in the ductile and brittle regime
(10) F. Özgün Ersoy (SCK-CEN), Investigating liquid-metal embrittlement of T91 steel by fracture toughness tests
(11) G. Aiello (CEA), Pre-Normative R&D for Codes And Standards
(12) S. Gavrilov (SCK-CEN), Application of codes to LBE environment
(13) Karl-Fredrik Nilsson (JRC), Web based Data base MatDB
(14) Hyeong-Yeon Lee (KAERI), Application of Elevated Temperature Design Rules to High Temperature Components and Design Code Issues
(15) Kuo Zhang, Jarir Aktaa, Characterization and Modeling of the Ratcheting Behavior of the Ferritic-Martensitic Steel P91
(16) M. Angiolini (ENEA), Ratcheting of a P91 Cylindric shell under thermal cycling
(17) P. Matheron, G. Aiello (CEA), Raatcheting rules for P91 componets: improvement of the efficiency diagram method in RCC-MRx
(18) C. Testani (CSM), Study of application of Laser Welding for built GEN IV SS316 components
(19) Attila Kovács (AEKI), Irradiation and Post-Irradiation Experiments for the 316 LN and P91 Materials
(20) G. Barbieri (ENEA), Influence of deposition ratio and PWTH duration time on toughness and ductility of P91 welds
List of Websites:
www.matterfp7.it
contact details: marco.utili@enea.it
The MATTER Project was launched in 2010 to support the European initiatives in the field of materials for Gen IV reactors. Since two projects were recognized as the most advanced in Europe, ASTRID and MYRRHA, the major efforts of the European materials community were addressed to both of them. The MATTER project had also the duty to support and to assist from the organizational aspects the start of Joint Program for Nuclear Materials in the frame of EERA. Due to wide interest by European organization to co-operate in the development of materials for Gen IV reactors, 26 institutions from Research institutes, Industry and Academia adhered to the project. The relevant number of objectives induced the coordination to split the project into 4 organizational domains: 1) Materials Characterization, 2) Pre-normative R&D, 3) EERA Joint Program for Nuclear Materials, 4) Management and Education/Training. The first domain included the prioritization among the material properties to be dealt in the project, provided several screening test techniques, reached a consensus within the EU partners on common testing and evaluation procedures, allowed round robin exercises to validate the various testing and evaluation procedures and finally provided analytical and physically-based support to the proposed guidelines.
In the pre-normative domain the design rules were revised and updated in order to answer to some short term needs of the two projects ASTRID & MYRRHA. The validity of design rules such as ratcheting and creep-fatigue was assessed, some materials features such as negligible creep domain, ageing, corrosion were reviewed for update. The best practices for welding processes were thoroughly studied with the support of industrial partners and codification updates were proposed. Finally, characterization tests were performed to assess the properties of grade 91 steel, of improved functional surfaces and to devise weld coefficients.
The main purpose of Domain 3 of MATTER is to support the EERA NM R&D Joint Program. The EERA JPNM is made out of a “Short-term” component addressing structure and clad materials issues in support of ESNII reactors and a “Medium-long term” component addressing structure and clad R&D needs for the FOAK (First of a kind) FNR (Fast Neutrons reactor).The relevant classes of materials considered for the short-term program are Austenitic alloy, F/M steels, hard coatings to replace Stellite. The related studies, in MATTER, are mainly done in domains 1&2. The long-term program includes more innovative materials which are at the development level and have been therefore never used for nuclear systems. These materials are ODS and Refractory Metals and Ceramics. Moreover, in the long-term program, one sub-programme is dedicated to modelling, experiment and validation. This sub-program is considered as cross-cutting and also relevant for the short-term part. MATTER dedicated a relevant effort to ODS supporting several different fabrication paths by the contribution of six European organizations.
The relevant issues of the fourth domain were Relevant issues of Domain 4 were the education & training, the dissemination, the management and the scientific coordination. The education and training was assured by the financial support to nine Positions of PhD which were launched inside the project and by the organization of student courses. The dissemination was performed by the organization of international workshops, the publish of 21 Scientific papers, 11 paper are under revision for the special issue of Journal of Nuclear Material dedicated to MATTER project, and MATTER web site. And ODIN tool for materials database The project management consisted of all the legal, contractual and financial activities and the controlling administrative issues and of the provisions for support of the Governing Council, the Project Co-ordination Committee, and respective meetings. The scientific coordination included the effective conduction of the work programme, the contacts with EERA group, other EC Projects and Gen IV and the provisions for support of the End-Users Advisory Group and of the Scientific Advisory Board.
Project Context and Objectives:
Context
The 2010-2012 implementation plan of the European Sustainable Nuclear Industrial Initiative (ESNII), in the frame of the Sustainable Nuclear Energy Technology Platform (SNE TP), establishes the road map for the start of construction of the European Gen IV prototypes. Namely it has been already decided that the construction of the LFR ETPP (European Technology Pilot Plant) MYRRHA and SFR Prototype ASTRID will be completed in the twenties.
The development of GEN IV fission reactors promises a long lasting, carbon free, power source having sustainable impact on next generations and enhanced safety in terms of resistance to severe nuclear accidents. The achievement of these goals passes also through the development of structural materials having enhanced capabilities to withstand the highly demanding conditions of the fast neutron reactors.
The operational conditions envisaged for the various GEN IV reactor concepts, under consideration in the ESNII initiative, are quite demanding and significantly different from the service conditions of most of the commercial reactors operating today. The combination of severe conditions includes:
- High neutron damage, as an effect of the high neutrons energy (up to few MeV);
- High temperature, as a consequence of the specific thermal cycles required by the coolants properties and by the quest for higher thermodynamic efficiency;
- Chemical and physical compatibility with the aggressive conditions due to contact with the coolants.
The behavior of materials under operation conditions of such systems was not appropriately covered by the available testing and evaluation standards, which therefore required updates of existing standard procedures and sometimes development of new ones. In addition, one of the required features for the success of the Gen IV reactors is the demonstration of their long lifetime: up to 60 years.
It is recognized that the available European standards and rules for material testing and for components design & fabrication do not appropriately and completely cover the specific materials to be used in GEN IV reactors nor the issues related with their relevant operating conditions and environments.
The EERA (European Energy Research Alliance) group for innovative nuclear materials has been established in 2009 regrouping several scientific representatives from research institutions all around Europe with the purpose to foster and perform material studies for nuclear applications.
From this consideration it appeared the necessity to answer to the industrial needs for innovation of the design rules, namely the RCC-MR and R5 standards , which were developed in Europe, capitalising the newly developed knowledge. That was strongly underlined by the Multinational Design Evaluation Program (MDEP) Codes & standards Working Group which has the goal to achieve convergence of practices related to nuclear component design and fabrication.
These topics are in line with a number of European (GETMAT, HeliMnet, GOFASTR, LEADER, ESFR) but also international initiatives (GIF, OECD/NEA) where experimental data were collected.
Scope and objectives
Scope of the MATTER (MATerial TEsting and Rules) project was to contribute to cover the existing gaps by pointing out methodologies, recovering existing experiences, and performing experiments.
Main goals of the MATTER Project were:
• To develop and establish best practice guidelines for testing and evaluation, aimed to screen and characterize nuclear materials for all the ESNII innovative nuclear systems. These harmonized testing and evaluation guidelines will be susceptible to be adopted in the EU laboratories.
• To perform Pre-normative activities in order to address the short term materials needs of the ASTRID & MYRRHA projects with respect to the design and construction of the structural components. In particular, among others, the design rules for 9% Cr steel, material of the ASTRID steam generator, was provided together with the related mechanical design properties and the updated welding and fabrication rules.
• To optimise the effectiveness and efficiency of the EERA Joint Program on structural materials for innovative reactors, on the base of the knowledge produced by the material researchers and of the industrial current practices.
• To carry out the fabrication in Europe, by different methods, of ODS steels, which are under consideration as candidate materials, in the medium-long term, for fuel cladding application.
To achieve the aforementioned objectives, the R&D activities were focused on material testing, analysis of existing data and theoretical research. The relevant advantages of the MATTER approach were the access to a large amount of data, generated by national researches, and the wide comparability of results, thanks to the presence of 26 organizations including industries and steel makers. The large number of participating results and the specific competencies of the several different participating institutes. A relevant advantage of this approach consisted in the final comparability of the experimental data, being produced by consensual procedures, and their immediate availability in view of their pre-normative deployment.
This experience and approach will also be useful for the development of other nuclear systems such as
GFR and LFR, not exclusively MYRRHA and ASTRID. The obtainable results, related to the objectives of the Project, were:
- List of materials properties that are essential for the design of the innovative nuclear systems and selection of the more appropriate ones to be dealt with in the present Project;
- Testing procedures and best practice guidelines for the screening and characterization of the selected materials properties and conditions;
- Materials pre-normative qualification with respect to the selected properties and conditions;
- Contribution to update the design rules in consideration of the needs posed by the new reactors;
- Preparation of the EERA R&D Joint Program and first achievements.
Such a work have improved the interaction between European material research laboratories, established better link between designers of nuclear systems and material scientists and allowed a large dissemination of the obtained knowledge to the benefit of the end–users.
Project Results:
Materials for ASTRID and MYRRHA
The design of ASTRID, which uses sodium as coolant, relies to large extent on the design and operational experience from previous sodium fast reactors, in particular Phénix and Superphénix in France. Stricter requirements on safety and economy and sixty years design life raise a number of design issues described in [D1.1 and D1.3]. The ferritic-martensitic tempered steel Grade 91 (P91/T91) was considered as the reference materials for the steam generator and possibly for intermediate heat exchangers primarily due to its resistance to thermal fatigue. P91 softens under cyclic softening and the ratio between ultimate strength and yield is much lower than for 316 stainless steels. The Design Rules for 316 stainless steels could therefore not be used. Key objectives of WP4, 5 and 7 were therefore the development of Design Rules for ratcheting, creep-fatigue interaction and negligible creep for P91. Welds constitute weak spots in nuclear components and was addressed in WP6 for improved welding procedures and selection of filler materials for both 316 stainless austenitic steel and P91 ferritic-martensitic steel components as well as a proposal for new fatigue weld coefficients for P91. The 60 years design life means that additional design factors need to addressed and then in particular thermal ageing, irradiation and environmental degradation, which was addressed in WP5.
MYRRHA will use lead-bismuth eutectic as coolant, used in Russian nuclear submarines, but not implemented for other nuclear applications. Thus there is basically no operational experience to rely upon and test data is very limited. Moreover, heavy liquid metal coolants (lead or lead-bismuth) will interact with the structural materials and finding materials that sustain this environment is one of the key bottlenecks for the development of the HLM nuclear reactors. In MATTER Work package 2, test procedures were developed both for liquid metal embrittlement and corrosion. A comprehensive test programme was performed to enlarge the database needed for austenitic and ferritic-martensitic steels for the design of HLM reactors. This data provides some guidance on material selection, design rules and mitigation measures to ensure integrity and functioning of the components.
Projects involving design and construction of complex and innovative facilities such as the ESNII prototypes and demonstrators need to be dynamic and adjust to new requirements, changes in available funding and practical as well as more scientific issues that may arise. Hence the needs are a moving target. For ESNII the financial problems in the EU in the last years, in particular the Fukushima accident, have had a large impact on the ESNII projects. As a consequence the a number of modifications have taken place for material selection and component design and the projects are a bit delayed as compared to the plans in 2010.
Test Procedures and LBE degradation (WP2 WP3)
Access to material data representative of the operational conditions is necessary for the design of the ESNII reactors. For innovative reactor systems we can only partially rely on experience from earlier designs and we need to develop test methods and associated procedures to address both new and classical materials. We also need to assess properties and degradation mechanisms for known reference materials in the relevant conditions. This translates into needs for screenings tests of candidate materials as well more challenging tests where specific degradation mechanisms are targeted and assessed with state-of-the-art tools. For screening, miniature tests are of special interests as they are relatively simple to perform and they require small amounts of material. In MATTER the small punch test and nano-indentation were therefore evaluated. The heavy-liquid metal (HLM) coolants provide a corrosive environment and for the design and operation of MYRRHA and ALFRED we need to better understand the compatibility issues between HLM coolant and the structural materials to define material requirements and mitigation procedures. These two aspects were addressed in MATTER Work Package 2 and 3.
Miniature screening tests
Different miniature test methods were evaluated in Work Package 2 of MATTER: in particular small punch test but also macro- and nano-indentation impression creep.
In the small punch test a small clamped disk is subjected to a controlled displacement of force from a puncher. The disk has typically a thickness of 0.5 mm the diameter is typically 10 times larger. Estimates on different uniaxial material properties such as yield stress, ultimate strength, ductile-to-brittle transition temperature, creep rates can then be derived from measured deflection and the force. The deformation is bi-axial and a finite element analysis is generally needed for more detailed evaluation of local strain/stress levels. There are however simplified formulations and correlations for the estimates. The tests are rather simple to perform but the rather complex bending and stretching interaction of the tests complicates data the assessment. The test is clearly suitable for estimates of time to rupture in creep or tensile strength. The SPT has also been shown to be a powerful tool in screening and ranking testing.
In MATTER the SPT has been used to assess how irradiation and liquid metal embrittlement affect selected material properties. One example is where the SP energy is plotted against the temperature data for irradiated and unirradiated specimen. Due to limited data points and scatter, it is difficult to determine the small punch transition temperature, TSP , precisely. For the unirradiated specimens it is about ‐185 °C and ‐145 °C for specimens at 3.49 dpa. The temperature dependent yield stress obtained from the SP tests compared with results from tensile tests in a previous project SPIRE and literature data. The SP based data result in an overestimation of the yield strength in comparison to the values from the tensile test. A clear irradiation effect on the yield stress is observed.
In MATTER SPT was also used to assess the liquid metal embrittlement from LBE on P91 steel. For specimens tested in contact with LBE the energy drops is about 31% at 300 °C and about 22% at 400 °C as compared with specimens tested in air. For the plastic energy the decrease is even more: 35‐40% at 300 °C and 25‐28% at 400 °C. It strongly reflects the LME effect of LBE on the T91 steel in this temperature range. The results indicate also a better repeatability for SPT than slow strain rate tensile tests.
Nano-indentation was used in the MATTER project to assess irradiation effects. There is large interest in using ion irradiation to simulate neutron induced damage in nuclear materials for basic investigations as well as for materials screening. By Ion irradiation neutron activated material is avoided, and there is a tremendous reduction of irradiation times and improved capabilities to vary irradiation conditions. However, the question to what extent results based on ion irradiation can be transferred to of neutron irradiation is still debated. The transferability issue arises not only from different irradiation conditions but also from different testing methods due to the limited penetration depth of the ions. In MATTER the same heat of T91 was exposed to neutron and ion irradiation at the same temperature and the same nominal dose and tested by nano-indentation. In the project it was compared the hardness versus contact depth for un-irradiated, neutron and ion irradiated T91 material. There is very good agreement in the irradiation induced hardness change after neutron and ion irradiation at contact depth < 200 nm, where the substrate effect is negligible, i.e. the mechanical effect in terms of irradiation hardening of neutron and ion irradiation is the same in the present case.
Procedures for small punch test and nano-indentation have been evaluated and improved in MATTER. It has been demonstrated that the small punch test can be used as a screening test to assess irradiation damage and liquid-metal embrittlement and that nano-indentation could be used to assess change in hardness from neutron and ion irradiation and where ion and neutron irradiation give comparable results. These tests can be used for design, but perhaps a more important feature is that both tests require minimal material volume they can be used to monitor the material properties for components subjected to long-term operational conditions.
Test procedures and material degradation in LBE
For the design of MYRRHA, candidate materials need to be screened with respect to their resistance to degradation and failure in LBE environment, and in particular it is necessary to demonstrate sufficient resistance of selected materials against potential degradation mechanisms such as corrosion and liquid metal embrittlement (LME). This is a very complex task as the interaction between the LBE and the material depends on properties of the steel (e.g. chemical composition, microstructure and homogeneity) and the liquid metal (e.g. concentration of dissolved oxygen, temperature and flow velocity). Experimental data in the literature is scattered and often inconsistent. A major reason is that key parameters such as oxygen concentration are not well controlled and that different test procedures have been followed due to lack of test procedures. For design it is very important that generated data are reliable and reproducible. A large effort was therefore dedicated to the development of test procedures through round-robin testing. The test procedures were then used to derive material properties and mapping of degradation mechanisms which can be used as guidance for material selection and boundaries for the operational conditions.
Small strain rate tests (SSRT)
It was demonstrated that the small punch test can be used as a screening test for liquid metal embrittlement. The small strain rate test (SSRT) is probably the most common test for investigation of environmentally assisted cracking of materials. In SSRT a dynamic straining of the specimens is applied in the form of slow, monotonically increasing strain to failure. The dynamic strain requires less testing time in representative environment than static strain and it can therefore provide an early warning to potential problems.
In the project it was compared the measured total elongation versus temperature for T91 steel for different strain rates in both inert environment and LBE. The test show a clear reduction in ductility for tests in LBE, which is affected by both temperature and strain rate. Very clearly pronounced ductility trough is observed for the tests performed at strain rate 5•10-5 s and at 350°C.
For reliable assessment of LME it is necessary that the LBE and the solid are in full contact. The oxygen content in the LBE has a large impact and needs to be controlled. It can also be noted that there is a large scatter in total elongation but the stress-strain curves up to failure are very similar, which is a typical behaviour when elongation is controlled by micro-defects that provide crack initiation sites at the surface.
3.2.2 Fracture toughness tests
Given the reduction in ductility it is also natural to expect that the fracture toughness is affected by liquid metal embrittlement. To this end test procedures for fracture toughness in HLM were developed based on ASTM 1820 and fracture toughness curves were derived for P91 in LBE.
Indeed LME reduces the fracture significantly but to observe this effect requires that the specimen is pre-cracked in LBE. Hence pre-cracking in air will give non-conservative values of the fracture toughness of P91 in LBE.
Corrosion tests
The liquid metal corrosion can be both uniform and localized. As regards uniform corrosion a distinction is made between oxidation, dissolution and erosion. The formation of a thin self-healing oxide layer on the surface of the steel acts as protection against further corrosion, in particular dissolution. In MATTER a thorough investigation of corrosion between LBE and ferritic-martensitic steel (T91) and austenitic steels (316L and 15-15 stabilized steels) was done.
A thorough review of literature data (D3.6A) revealed a large scatter and inconsistencies in reported data. This is to a large extent caused by lack of control of key parameters, in particular the contents of dissolved oxygen. A first requirement for test procedures is therefore that all parameters are well defined and monitored and reported and in the case of dissolved oxygen content ideally also controlled. The MATTER test programme included the round-robin with the aim to develop test procedures as well as dedicated to determine corrosion data for the different steels under for different combinations of parameters (oxygen content, temperature, stagnant vs. flowing liquid metal). Tested samples were investigated by different material characterization methods and the observed corrosion was quantified. Details of the test results are reported in D3.6B and D3.6C. The temperature and the dissolved oxygen content are the two most important parameters. One outcome of MATTER was that to avoid localized dissolution the target dissolved oxygen content, Co , for MYRRHA has been increased from 10-8 to 10-6 wt% and the maximum temperature on the steel surface decreased from 500 to 470°C. In general the MATTER tests give consistent data that provides database for design of HLM cooled reactors. Various gaps to be addressed by further work have also been identified. In the project it was shown test data from literature and MATTER for different values of temperature and oxygen content for ferritic-martensitic steels tested for test period longer than 2000h and 4000h respectively. For the MATTER tests it can be noted that for the data points changed from oxidation to dissolution when the exposure time was increased from 2000h to 4000h.
For HLM cooled reactor systems the degradation of material properties from interaction with the coolant and the associated reduction on integrity and functioning of components is a key issue. Since there is hardly any return of experience it is necessary that material tests to support materials selection and characterization is an urgent need for the design and construction of MYRRHA and ALFRED. The work and the results in the MATTER project will clearly benefit these two systems:
The procedures developed for liquid-metal embrittlement and corrosion allow generation of reliable and consistent data. For liquid-metal embrittlement it is necessary that the surface of the solid is wetted and for corrosion it is very important that the oxygen content is monitored.
Both the fracture tests and slow strain rate tests show that P91 steel has a significant embrittlement in LBE and will therefore not be used for any load bearing components in MYRRHA. Test have been performed for austenitic steels following the same procedures and very limited LME is observed.
For corrosion oxygen content and temperature are the main parameters. Based on the results from MATTER, the target oxygen content has been increased and the maximum temperature has been lowered.
The new data have allowed a much better understanding of LME and corrosion and a first basis for design. Further needs have been clarified, such as long-term corrosion tests, have been identified and provide a basis for further collaborative work between ESNII and EERA JPNM.
Design Rules and Assessment procedures
Development of Design Rules and Assessment Procedures was the second domain in support of the design and construction of the ESNII reactors. Most of the work was devoted to 9%Cr steel grade 91 (P91/T91), whose mechanical behaviour is considerably different from austenitic steels contained in nuclear Codes and Standards (C&S). Thus there is a need to elaborate design rules as a final step before codification. The different 316 grades remain the reference for the vessel and the piping systems and all construction requirements need to be updated according to the current industrial practice and knowledge. Also, a service lifetime of 60 years is envisaged which needs an update of available design properties and an analysis of a possible reconsideration of ageing phenomena in the design.
RCC-MRx has been selected as the Design Code for all ESNII reactor projects and the work in MATTER is therefore specifically addressing revisions of this code. The R5 code developed in the UK is mainly meant as assessment code for components in operations, and was also addressed specifically. The work split into three main areas:
High-temperature degradation (creep, creep-fatigue, ratcheting and negligible creep ( WP 4and 5);
Other long-term degradation mechanisms, thermal ageing, irradiation and coolant compatibility (WP5)
Welding (WP6).
Supporting experiments for all areas were performed in WP7 (D7.1 and D7.4). An in-depth synthesis and recommendations for the MATTER Design Rules are given in Deliverable D4.8.
Design Rules for High-Temperature Assessment (WP4 and 5)
The ESNII reactors will generally operate at elevated temperatures and stresses will mainly result from thermal loads. The design rules for 316SS in RCC-MRx cannot be used for P91 due to the cyclic softening and the different yield to tensile strength ratio. In MATTER proposal for Design Rules for P91 with respect to ratcheting, creep-fatigue interaction and negligible creep have been developed.
Ratcheting Rules for P91
Ratcheting refers to progressive non-recoverable deformation under cyclic loading. Ratcheting is primarily a plastic deformation and occurs at lower temperature, but ratcheting increases at higher temperatures where creep effects also need to be taken into account. At low loads but high temperature we may have pure creep ratcheting. The work in MATTER included the development of visco-plastic constitutive models for more detailed simulation pf ratcheting under more general conditions as well as the development of efficiency diagram in accordance with RCC-MRx approach.
In RCC-MRx there are two basic rules for ratcheting: the 3Sm rule and the efficiency diagram. The 3Sm rule is applicable only under negligible creep and states that the sum of primary stresses and secondary stress range must be below three times the allowable stress.
The efficiency diagram used normalized forms of primary and range of secondary stresses. The actual diagram is purely empirical and gives the relationship between the secondary stress ratio, SR= Q/P, and the "efficiency index", v= P/Peff, where Q is the secondary stress range, P the primary stress and Peff is defined as the primary stress that in absence of cyclic loads using the monotonic stress-strain curve, would give the same cumulative strain as with a superposed cyclic primary and secondary load. A specific Peff is thus linked to a specific cumulative strain value that depends on the stress-strain curve. The efficiency diagram developed shows the data points for P91 from torsion-tension test presented in D7.1 together with the RCC-MRx efficiency diagram for 316L(N) and the proposed revision for P91. The increased ratcheting for P91 which is manifested by the reduction in the efficiency diagram is related to the cyclic softening. This efficiency diagram was also validated for non-isothermal large component tests.
Creep-fatigue interaction for P91
The Creep-fatigue assessment in existing design and assessment codes such as RCC-MRx, ASME III NH and BS-R5, is based on creep-fatigue interaction diagram where separate analyses are done for creep and fatigue. The axes represent usage factors for creep and fatigue respectively and the interaction is taken into account by the bi-linear diagram that defines the allowable region. The main difference in the codes lies the calculation of the creep damage e.g. life exhaustion or different ductility exhaustion. In the MATTER project it is shown that the experimental points of the diagram may also differ but that does not necessarily correspond to difference in safety margins, as it is also a consequence of the difference in how creep damage is modelled. The creep-fatigue interaction diagram has worked well for austenitic steels but for ferritic-martensitic steels there is a large scatter and a failures have been observed in the safe region. This is believed to be caused by cyclic creep softening, where pre-cycling significantly increases the creep rate.
Since the creep-fatigue interaction diagram is based on normalized data it does not distinguish between data with different hold times or number of cycles nor for load sequence effects. The objective of the creep-fatigue part of the MATTER project was therefore to evaluate the interaction diagram for P91 and propose modifications as well as propose and evaluate alternative methods.
One conclusion is that the interaction diagram as proposed in RCC-MRx, seems to provide sufficient conservatism also for P91 for available test data under strain control where all data points are well above the interaction diagram. A major reason is that the design fatigue curve is used, i.e. allowable number of cycles is defined by halving the strain range of the mean LCF curve and by dividing the mean curve number of cycles by a factor of 20. The increased creep rate from cyclic softening is not accounted for but partly compensated by increased stress relaxation. It should be remembered that almost all data are with short hold times and high loads and extrapolation to long hold times is not obvious.
In the MATTER project two alternative simplified were also investigated: the Manson-Halford model and the φ–model based on the assumption that accumulated creep life can be derived from a normalized reference stress defined by strain range, hold time and temperature. These models are not influenced by the material softening (or hardening) and have much fewer fitting parameters than any of the currently used interaction diagram methodologies. If considering the simplicity of use and the clear advantage in total amount of fitting constants needed, it is clear that the simplified methods have an advantage compared to the current code assessment procedures. The simplified models are successfully predicting the JRC creep fatigue test data for temperatures 550 and 600°C. The number of cycles to failure versus the total strain range (creep and fatigue) was evaluated in the porject. The model has been calibrated with available LCF and CF data. Predictions up to 1 h hold time are supported by data, whereas longer hold times are extrapolations. The reduction in number of cycles can be directly inferred from the reduction in fatigue curve.
Negligible creep for P91
The negligible temperature curve (TNEC) is defined as the temperature that causes a "negligible" prescribed creep strain, relaxation or life fraction in a specified time at the specified reference stress. In MATTER a creep strain limit of 0.2% was used. Thus, the negligible creep domain is the limiting temperature/time within which designers do not need to consider creep, i.e. if the service time is below the negligible creep curve, and this simplifies the design analysis significantly. RCC-MRx contains negligible creep curve for austenitic steels but not for P91. Although the concept seems trivial, the approach to determine TNEC is not straightforward. First of all the reference stress and the value of the strain limit is material dependent and may differ between the codes. Moreover, supporting experimental data should be ideally be very long-term and at low temperatures and stresses. In reality we need to rely on extrapolation of creep curves to negligible creep. In MATTER very little low temperature data was available in the public domain or from the project partners so a test programme was initiated and performed at temperatures close to the low temperature limit for time independent design. The stresses were chosen to produce the sought 0.2% strain in about 5000 h. The required stresses are relatively large, situated on both sides of the RCCC-MRx defined reference stress of the ultimate tensile strength divided by a factor of 2.7.
The approach was based on the Wilshire creep model which was calibrated to both generated and public domain strain data to present normalized stress versus temperature compensated time for different life fractions of rupture life or different creep strain values. The Wilshire equations can then be inverted to compute the negligible creep temperature versus time, for a given strain limit and reference stress. The negligible creep curve for 0.2% creep strain together with the time to failure (rupture) curve and the corresponding time to 0.2% strain prediction using the RCC-MRx creep model. The plot also shows the creep rupture curve with a scaling factor of 1000, which almost coincides with the 0.2% creep strain curve. Thus in this case a 0.2% creep strain criterion corresponds to a safety factor of 1000 against creep rupture. It is worth mentioning that the negligible creep curves derived in MATTER assume pure creep. Under cyclic loading the negligible creep curve may be reduced.
Robust Design Rules based on sound mechanical basis are crucial for the design of the ESNII reactors. The proposed Design Rules for ratcheting, creep-fatigue and negligible creep should be reviewed by AFCEN for inclusion as probationary rules in a first stage.
The rules for ratcheting and negligible creep follow the same principles as the Design Rules for austenitic steel that are already included in RCC-MRx so it should be straightforward to include the proposed rules for P91.
For creep-fatigue the situation is a bit different since the interaction diagram, which is established in RCC-MRx and other codes, seems to provide sufficient safety margin for the available data. But the interaction diagram gives very limited insight and includes more ad-hoc assumptions and is more questionable for extrapolation to long hold times.
All rules need also to be validated for loadings more representative for real operational conditions, in particular long hold times and low stresses.
The loads in Generation IV reactors are primarily thermal. Since the creep properties are strongly creep dependent whereas the Design Rules are based on iso-thermal conditions the applicability of the Design Rules in non-isothermal conditions need to be validated. Future work may be needed to develop the Design Rules to non-isothermal conditions.
Design Rules for other long-term degradation mechanisms (WP5)
The designers of Generation IV reactors need to demonstrate that non-replaceable components retain their integrity and functioning for at least 60 years, compared to 40 years in past projects. This requires that the time line of material data and design rules need to be extended and additional slow degradation processes need to be taken into account. In the MATTER project design rules were addressed for three types of processes: thermal ageing, irradiation and environmental effects from heavy liquid metal.
Thermal ageing
Thermal ageing is related to the stability of the material microstructure at service temperature where a variety of processes may occur, including plain grain growth of the material matrix, coarsening of particles, segregation of minor material constituents, dissolution of precipitates and formation of new phases not present in the initial microstructure. As mass transport (diffusion), chemical reactions at interfaces, particle nucleation are involved, the progress of processes constituting thermal ageing strongly depends on time. At LMP of 22 corresponds to 110, 60 and 10 thousand hours at 550, 600 and 650 °C respectively. Simultaneous action of mechanical load and deformation of the material influence thermal ageing and induce additional processes like dynamic recovery or re‐crystallisation. Since tempered ferritic-martensitic steels are in a meta-stable phase it is that thermal ageing may play a larger role than for austenitic stainless steels.
Thermal ageing is difficult to include in design codes due to the complexity of the processes but also since there is no simple way to perform representative tests in reasonable time frames. Tests at elevated temperatures have been proposed but such data will never be truly representative. The only realistic approach is to perform tests on components that have been exposed to operational conditions for a long time. Thermal ageing may reduce the ductility, fracture toughness and increase the creep rate.
In design codes the ageing is therefore handled by reduction factors. In In RCC-MRx provision for thermal ageing coefficient for allowable stress for assessment of P-damages (static load) is included, but actually listed for very few materials and no such factors are for instance provided for 316L(N) or P91 steels. It could be noted that ASME, section III, Division1, subsection NH provides ageing factors. However for 316L(N) welded components fracture toughness curves are provided in RCC-MRx for moderate (e.g. 25,000h at 550°C) and advanced (e.g. 110,000h at 550°C) thermal ageing. Thermal ageing depends to a large extent on the chemical composition and microstructure and the codes, such as RCC-MRx provides tight specifications which should limit thermal ageing.
Given the lack of experimental data mechanistic approaches that incorporate the microstructural features were evaluated in MATTER. Experimental data suggest that creep life can be minimum creep rate (Monkman-Grant relationship). So the effect of thermal ageing on creep may be accounted for if we can determine the minimum creep rate , which potentially can be done using the Dyson mechanistic creep model based on mechanistic models where the parameters can be derived from measured microstructural evolutions and relatively simple physics-based models. However no concrete proposals for inclusion in the design code could be provided in MATTER.
Irradiation effects
The general strategy of considering irradiation effects in design codes and standards for structural components is to first make a classification in terms of the expected significance for the component performance over the service time, i.e. (a) negligible; (b) significant and to be considered; (c) maximum allowable irradiation [38]. Respective thresholds of irradiation fluence or resulting accumulated damage on the atomic scale are material dependent, and may vary with the temperature range under consideration. If a significant impact on strength properties is to be expected, it needs to be either confirmed that design curves provided in the code still provide a sufficient safety margin or further reduced strength properties assumed. No suggestion for revision of irradiation effects were proposed in MATTER.
Heavy Liquid metal effects
As already mentioned the design of nuclear reactors with HLM coolants has very limited return of experience to rely on. The report D4.7 gives an excellent survey and assessment of the limited experience from Soviet and Russian projects and LBE facilities in Europe. Moreover Design Rules for HLM reactors have not been developed. Different approached for environmental factors are provided in the ASME codes as regards e.g. corrosion/erosion allowance, stress corrosion cracking and environmental fatigue. These are more guidance and allowance values rather than design rules. RCC-MRx has even less guidelines on environmental effects, but RCC-MRx is nevertheless considered as the more suitable code for the development and codification of Design Rules as it is based on return of experience from sodium cooled fast reactors and includes rules and guidelines for e.g. high temperature applications and specific nuclear components. The deliverable D4.7 provides a framework for how LME and LMC can be included in a Design Code.
The ensure integrity the designer makes three implicit assumptions (D4.7):
The chosen material, in the chemistry controlled coolant, has the same behaviour as in air except for the depletion of the corrosion allowance;
The ISI schedule will detect flaws, local corrosion, local thinning, before defects become critical, assuming a limited corrosion speed;
The range of composition of coolant stays within the request margin all over the systems
Failure of the chemistry control system that could cause high speed corrosion are detected and the grace time is sufficient to take corrective action.
The first one, in line with the first implicit assumption, is to use only materials that are not susceptible to LME or LMC in heavy liquid metals or where the measured effect is negligible. The challenge for this approach is to define criteria of the negligibility. The advantage of this approach is that no modifications of the design code are required and the experimental program should be done to demonstrate that heavy liquid metals have no significant effect on the design mechanical properties. The disadvantage is that the number of available materials might be very limited since from screening tests it seems that all tested chromium steels as well as high-strength steels are susceptible to LME. Therefore, due to the screening-out of these materials may reduce the operational parameters of the developed installations. From available data 316L could satisfy the requirement but ferritic-martensitic steel would clearly be ruled out.
The second approach is to tolerate susceptibility of the material to LM degradation by imposing additional margins on the mechanical properties used in the design. The advantage of this approach is that it does not exclude susceptible materials from the list of available materials. This approach will require quantification of how the mechanical properties depend on various parameters and to define conservative reduction factors and to validate these from tests representative for the operational conditions. The challenge in the dissolution attach versus exposure times for solution annealed and cold drawn batches of 316L steel is that the standard procedures for such tests are not readily available and their development requires significant collective efforts. The design factors should not be too large and they must be robust. The ferritic-martensitic steel would in practice be ruled out also by this approach.
The third approach would be to develop mitigation strategy to prevent LM degradation. It can be either tailoring the chemical composition of steels or coolant chemistry. The importance of having the oxygen content within a narrow range for both LME and LMC was demonstrated in Section 3. This approach requires very good understanding of the underlying processes for LME and LMC supported by an extensive database. One problem with this approach is to ensure that the coolant chemistry remains within the tight limits in all parts of the system.
Taking account of time dependent and environmental induced degradation mechanisms constitutes the toughest challenge for the design of Generation IV nuclear reactors. This is due to the complexity of the processes, their dependence on operational conditions. Another problem is to replicate the conditions at laboratory scale. These problems are also why Design Rules are not developed or very limited.
The concrete proposals in MATTER for Design Rules for thermal ageing were very limited. The main reason to this was the lack of relevant data. Progress in this requires testing of specimen taken from components that have been exposed to long-term operational conditions. This can then be combined with laboratory test of specimen with simulated ageing conditions. Design curves or design factors could then be derived by extrapolation of laboratory data to operational conditions and time scales using physics based models. Such an approach would require a close integration and collaboration between scientists, designers and operators.
For the design, construction and licensing of HLM cooled reactors it is necessary to demonstrate that components retain their integrity under operational conditions. The development of guidelines, design rules and mitigation approaches is therefore an urgent need. The results in MATTER constitute a very important step as it has demonstrated that P91 is not a suitable material, confirmed that 316L appears to be resistant to HLM degradation, and tightened the limits for oxygen control.
Since there is no return of experience to rely on the design need to a large extent rely on new test data. The MATTER deliverable D4.7 gives proposed a clear way forward. To generate the supporting data and methodology will require a test programme that first confirms the resistance of 316L to HLM degradation. Such a programme should include welded components. In parallel other candidate materials need to be screened and models need to be developed both at the engineering and more fundamental scale.
Design Rules and Manufacturing Rules for welded components (WP6)
Most failures in nuclear components occur in welds due to a combination of geometrical factors, variations in microstructure, micro defects and residual stresses. The production, characterization and integrity assessment of welds is very complex. Manufacturing rules and design rules to ensure that welded joints fulfil the strict safety requirement is crucial for the design and construction of the ESNII reactors. In the MATTER project the work on welds focussed on two areas: i) development of fatigue weld factors for P91 and 316 stainless steels components and different welding techniques and ii) development of welding procedures for P91 and 316L(N) steels.
Fatigue weld factors
In Design Rules and assessment procedures, safety of welded components is generally addressed by weld reduction factors that lumps the various factors mentioned above. In MATTER updates of fatigue weld factors for RCC-MRx and R5 were done by experiments and modelling.
In RCC-MRx the reduction in fatigue life for welds compared to base material is accounted for by the "joint fatigue coefficient, Jf", defined by the ratio of strain variation for base material and weld joint for a given fatigue life (J_f= ∆ε_BM/∆ε_wj). The joint fatigue coefficient depends on the base material, filler material and weld process. For 316L(N) with narrow-TIG weldments Jf is 1.25 but for P91 there are no values in Rcc-MRx. In MATTER joint fatigue factors were derived for P91 with TIG weldments. From tests of P91 welds failure is in the HAZ for low strains but in the base metal for high strains, which also follows from nonlinear FE-analyses. As a consequence, Jf is a function of the strain with proposed values 1.35 for strain ranges below 0.25%, 1.55 for strains above 0.55%, and a linear interpolation for intermediate values.
In MATTER joint fatigue coefficients were also assessed for 316L(N) and P91 with hybrid laser welds. For 316L(N) it was concluded that the weld coefficient for TIG (1.25) was applicable.
The basic approach for fatigue assessment of welded components in R5 is similar to RCC-MRx, but instead of the fatigue joint coefficient they use Fatigue Strength Reduction Factors (FSRF). In MATTER a refined method where the weld factor is separated into two: a Weld Strain Enhancement Factor (WSEF), and Weld Endurance Reduction (WER) derived from the material mis-match. The combination of WSEF and WER reduces the conservatism compared to FSFR for the creep effects. Comparison with tubular weldments of 316H validated that the revised procedure was still conservative.
Welding procedures and code requirements: 316SS and P91
For P91 RCC-MRx have data sheet for chemical composition and mechanical properties for the base material but not for the filler material. In MATTER different filler materials were evaluated in combination with welding methods with the objective to qualify welds for class 1 nuclear components. Three complementary welding technologies were included: fully mechanized GTAW Narrow Gap default method, Hybrid Laser Arc (HLW) and Shielding Metal Arc Welding (SMAW). Although the emphasis was on P91, also 316L(N) weldments were included. It was demonstrated both filler materials for 316L(N) were acceptable. For P91 SMAW and GTAW are promising but elongation is on the low side and softening on the HAZ remain as concerns.
The integrity of welds is a key issue for the design of all ESNII reactors. The development of fatigue weld factors, as well as the assessment of new filler materials and welding procedures, are of direct relevance for ESNII. The proposals form MATTER are should be incorporated into the Design and Assessment codes (RCC-MRx and R5) but some additional validation by experiments is needed in particular with respect to creep and high-cycle fatigue.
Material Database for integration of European materials research
Reliable data on material properties and long-term degradation is necessary for the design and long-term operation of nuclear power plants. The generation of such data is very costly as it requires special facilities, experts and requires very long durations. Moreover it is very important to have detailed information of data (e.g. raw data, test conditions, chemical composition etc) to use the data efficiently as exemplified for test data related to heavy liquid metal degradation. Much data exist in Europe, often generated as part of EURATOM projects, but very little of this data is in a form that it can be shared. It would therefore clearly of great advantage for the nuclear material research community and the reactors designers to build-up a common web enabled database. In fact this would represent a natural step towards the integration of the nuclear material research in the EU on line with the EERA JPNM Vision paper. JRC has proposed to use their web based database MatDB for this purpose and in MATTER it was agreed to test this. At the end of the MATTER project a total of 321 validated data sets for P91 and AISI 316 had been uploaded by 8 project partners to the JRC web enabled database MatDB. The uploaded data included: load and strain controlled low-cycle fatigue, small punch tests, uniaxial creep, uniaxial tensile, creep crack growth and fracture toughness data. The data sets provided by the partners are therefore accessible for the MATTER organizations.
Furthermore, it is only natural that there is a reluctance to share data when the only beneficiary will be a third party that has invested no effort in its generation. To resolve the former, funding agencies need to ensure that data plans are a mandatory part of research proposals and that data sets are recognized project deliverables. For the latter, a mechanism is needed whereby efforts to share data are acknowledged and rewarded. Datacite DOIs offer just such a mechanism, enabling data sets to be published and cited in the same way as conventional publications. The ODIN portal at https://odin.jrc.ec.europa.eu has recently been enabled to use DOIs for data publication. The owners of the data sets in MATTER were therefore offered to have their data sets with DOI.
Data is the key output of a project that can be exploited by project partners for future use. Ensuring reuse of data generated in a project is perhaps the most important output of a project. In MATTER the first attempt was made to have generated so that it could be reused. As a result 321 validated sets from the project are available for the MATTER projects. Depending on the restriction agreed by the MATTER consortium these data can also be by ESNII designers. Data that have been classified as restricted is also available for the broader research community.
Oxide Dispersed Strengthened Steels
The main outcome obtained in the project were related to:
Process Fabrication Characterisation
Development of a production route for ODS steels, which is based on conventional steel making techniques (introducing the oxide particles in the melt). (OCAS)
Manufacture of an ODS steel bar for laboratory testing by powder metallurgy (CSM, CNR-IENI)
Improvement of the powder metallurgical process (including MA) and Optimization of the final thermo-mechanical treatment (KIT)
Elaboration of ODS steels through spark plasma sintering techniques and scale lab heats to an industrial quality production (CEA - INSA)
Production of ODS steels using Spark Plasma Sintering (SPS) and assessment of microstructure and basic properties. (HZDR)
investigated and assess few key points of the fabrication routes of ODS
Develop a fast and simple screening test to check the effect of MA at least in the initial NFA exploratory phase. (ENEA).
Different recommendations were proposed in the project for the development of ODS materials :
The mechanical alloying is still the easiest way to produce ODS materials (CSM, CNR-IENI, ENEA, HZDR, KIT). It is relatively easy to obtain products at a laboratory scale and now the effort should be geared towards large scale production.
The Elaboration of ODS steels through SPS (Spark Plasma Sintering) techniques has been demonstrated and can be compared to HIP Process, large scale production and gains on the productivity have to be evaluated.
It has been shown that the recrystallization is a key issue for the control of the microstructure, the amount of nano-phases and the control of the cold working are the main parameters to reach an isotropic microstructure.
Two deliverables were produced:
Deliverable 10-1, Key "technological" issues for ODS materials
Final report on scientific understanding of ODS production, ranking and recommendations : in progress, should be ready for the end of November
Refractory alloys, Ceramics and Composites
These materials are primarily required in GenIV systems to access operating temperatures well beyond the current inherent limits of the most heat resisting super-alloys. Besides their refractoriness, the ceramic-based families have the additional advantage of chemical inertness and are envisioned as a viable alternative to conventional and advanced steels for their promise of greater resistance to corrosion/erosion in the fast reactor liquid metal coolants. The objective of WP11 of the MATTER project was to support the development of an implementation plan to be run during JPNM first phase 2011-2015, in order to provide the necessary background for the follow-up activities ( i.e. JPNM second phase 2016-2020).
At M12, a list of priority refractory/component systems was identified as well as their nearest-term issues, that were described in D11.1 and were progressively converted, starting from M24, into specific topics to be addressed in the form of separate pilot projects (MS14) in order to prepare the basis for proposal to EC calls. Status at month 30 of SP3 consolidation process was as follows:
1- Nuclear-grade SiCf/SiC composites with improved irradiation resistance and high temperature capability have been developed in the Fusion Materials Programs for the self-cooled DEMO blankets. These materials have most of the attributes required for GFR fuel cladding and they represent the leading option for such an application. The proposed activities cover the most immediate feasibility issues associated with the development of a leak-tight clad made of nuclear grade SiCf/SiC for GFR. The target of this proof-of-concept study is to contribute at progressing current TRL~3 to a stage (~4) at which few optimized SiCf/SiC clad systems could be provided in prototypic format for more quantitative assessments in integrated prototypic GFR environments.
2- MAX-phase are emerging materials that combine properties of metals and ceramics but do not benefit of any feedbacks from previous experience. Owing to their excellent machinability and promising corrosion/erosion resistance in lead alloys, they are explored as an innovative technology for the pump impeller in LFR (current TRL~1-2). A screening work based on a set of well-selected key criteria has been proposed to provide answer about the potential of the MAX phase concept for the given application and guidance for the follow-up.
3- Advanced V-alloys with an increased operating temperature window are under development in the US and in Japan for the fusion reactor blankets. Current candidates with reference composition of V-4Cr-4Ti have also attractive features for application as fuel cladding materials in fast reactors (current TRL~2-4 depending on the fast system end-use). Most pressing feasibility issues to be fully addressed in the scoping phase for maturing these alloys as cladding materials are mainly related to fabrication, processing and coating technology capabilities and performance assessments that are of cross-cutting interest for the fast reactor technologies.
The additional reports produced in the frame of the project are attached to this final report.
Potential Impact:
In the frame of MATTER project the following Workshops were performed:
- Final MATTER International WorkShop, “Rome, Italy, October 20-22 th, 2014. The objective of this WORKSHOP is to supply the final results of the MATTER Project, in order to complete the materials researches, in the frame of the EERA guidelines, with the implementation of pre-normative rules (RCC-MRx, R5, etc.) and new consensually developed testing standards.
- International Workshop on the “Key material properties for Myrrha and Astrid”, Rome, Italy, March 7-9 th, 2012. The first international workshop on the “Key material properties for Myrrha and Astrid” aiming at outlining the most relevant key material properties for both ASTRID and MYRRHA design.
The following Training Course were carried out:
- International School on Materials UNder Extreme Conditions (MUNECO), “Madrid, Spain, June 11-15, 2012. To strengthen the basis of knowledge in nuclear materials science, to promote the transfer of experience on structural materials for GENIV towards technological applications to young generations of scientists and engineers, the GETMAT and MATTER FP7 projects organize MUNECO school.
- International School on " DEsign Rules for gen IV reactos and INnovative reactors" (DERIVIN), INSTN Saclay, French, July 1-5, 2013. The objective of this school is to train European students and PhD applicants on the significance of design rules for nuclear applications, presenting them as living documents which are subjected to review and adapt to the evolving technologies. The attending students, after this Course, will have a better knowledge about the RCC-MRx and R5 rules for design purpose and know in detail some aspects related with the nature of the different materials.
In the frame of MATTER project the following coordination meeting were carried :
- 1st PCC video meeting 1st April, 2011 -
- 2nd PCC video meeting 21st September, 2011 -
- 1st WP2 General meeting 09th-10thMay, 2011 Paul-Scherrer-Institute Villigen (Switzerland)
- 1st WP3 Coordination Meeting 14th-15th September 2011 Headquarters SCK•CEN
- Brussels
- 1st DM2 Technical Meeting- WP4.1-4.2-4.3-6.1-7.1-7.3 03rd-04th May, 2011 Paris
- 2nd DM2 Technical Meeting- WP4.1-4.2-4.3-6.1-7.1-7.3 30th November-1st December, 2011 Paris
- MATTER-DM2-WP 10 Meeting 15th November, 2011 Saclay
- 1st Governing Council meeting 09th February, 2011 Bologna
- MATTER Kick Off meeting 09th -10th February, 2011 Bologna
- DM4 Coordination meeting 24th March, 2011 Casaccia (Rm)
- WP15 Coordination meeting 08th September, 2011 Casaccia (Rm)
- Coordination meeting with D2 leader 02nd December, 2011 Paris
- 3rd PCC meeting 13rd February, 2012 ENEA, Bruxelles
- Rue de Namur, 72-74 - 1000 Bruxelles
- 4th PCC video meeting 6th July, 2012 -
- 5th PCC meeting 6th February, 2013 Centrum výzkumum Řež
- Hlavní 130, 250 68 Husinec – Řež ,Czech Republic
- 1st Plenary meeting 13rd-15th February, 2012 ENEA, Bruxelles
- Rue de Namur, 72-74 - 1000 Bruxelles
- 2nd Plenary meeting 6th -8th February, 2013 Centrum výzkumum Řež
- Hlavní 130, 250 68 Husinec – Řež ,Czech Republic
- 2nd Governing Council meeting 14th February, 2012 ENEA, Bruxelles
- Rue de Namur, 72-74 - 1000 Bruxelles
- 3rd Governing Council meeting 8th February, 2013 Centrum výzkumum Řež
- Hlavní 130, 250 68 Husinec – Řež ,Czech Republic
- DM1 meeting 4th February, 2012
- DM2 meetings 18th and 19th of June 2012 Paris France
- DM2 meetings 19th and 20th of November 2012 Paris France
- DM2 meetings 28th of June 2013 Paris France
- Task 3.1: Fracture Toughness in LBE meeting 5th February 2013 CVR, Řež, Czech Republic
- Task 3.2: Liquid Metal Corrosion meeting 24 th -25 th September, 2012 KIT Campus North, Karlsruhe, Germany
- WP5: Technical meeting with AEKI 6 th -8 th February, 2013 Paris France
- WP5: Technical meeting 2nd April, 2013 video
- Task6.2 and 6.3: technical meeting 22nd May, 2012 Rome
- WP10 meeting 26th of June 2012 Dresden, Germany
- WP10 meeting 6-8th of February, 2013 Prague, Czech Republic
- WP2 – third Technical meeting September 30 ‐ October 1, 2013 Espoo, Finland
- WP2 – fourth Technical meeting September 16-17, 2014 CIEMAT, Madrid (Spain)
- WP3 Technical Meeting November 26-27, 2013 NRG, Petten, The Netherlands
- WP3 Technical Meeting June 26-27, 2014 CIEMAT, Madrid, Spain
- WP3 Technical Meeting November 27-28, 2014 SCK•CEN, Mol, Belgium
- MATTER DM2 FINAL MEETING July 17-18, 2014 Paris, France
- WP7 technicla meeting June 26 – July 7, 2013 Paris, France
- 3rd MATTER Plenary Meeting January 20th – 23rd, 2014 Hotel Freienhof, Thun, Switzerland
- MATTER coordination Meeting November 04, 2014 Rome, Italy
The following paper were produced in the frame of the project:
(1) M. Jong, N.V. Luzginova, H.S. Nolles, T. Bakker, “Liquid metal embrittlement testing in heavy metals; development and first fracture mechanics test results”, Oral Presentation at the IAEA Technical Meeting on Liquid Metal Reactor Concepts: Core Design and Structural Materials, 12-14 June 2012, Vienna, Austria; Paper for the Meeting Proceedings is under review
(2) C. Schroer, O. Wedemeyer, J. Novotny, A. Skrypnik, J. Konys, "Compatibility of ferritic/martensitic steels with flowing lead-bismuth eutectic at 450°–550°C and 10–6 mass% dissolved oxygen", Proceedings of the 20th International Conference on Nuclear Engineering collocated with the ASME 2012 Power Conference (ICONE20-POWER2012), 30 June – 3 August 2012, Anaheim, CA, USA
(3) K. Lambrinou, V. Koch, G. Coen, J. Van den Bosch, C. Schroer, "Characterization of oxide scales resulting from the exposure of various steels to liquid lead-bismuth eutectic", Poster Presentation at the 11th International Workshop on Spallation Materials Technology (IWSMT-11), 5-9 November 2012, Ghent, Belgium
(4) K. Lambrinou, V. Koch, G. Coen, J. Van den Bosch, C. Schroer, "Corrosion scales on various steels after exposure to liquid lead-bismuth eutectic", under review for the Journal of Nuclear Materials
(5) Holmström, S., Pohja, R., Nurmela, A.; Moilanen, P., Auerkari, P, Creep and creep-fatigue of stainless steel 316. 6th International Conference on Creep, Fatigue and Creep-Fatigue Interaction (CF-6), Jan. 22-25, 2012, Mamallapuram, India (2012).
(6) Holmström, S., Auerkari, P. Predicting creep life from small sets of test data and reference material characteristics. 12th International Conference on Creep and Fracture of Engineering Materials and Structures, Creep 2012, May 27 - 31, 2012, Kyoto, Japan (2012)
(7) “Pohja, Rami; Nurmela, Asta; Moilanen, Pekka; Holmström, Stefan; Multifunctional high precision pneumatic loading system (HIPS) for creep-fatigue testing, Procedia Engineering, Elsevier, Vol. 55 (2013), 573 – 577.”
(8) K. Lambrinou, V. Koch, G. Coen, J. Van den Bosch, C. Schroer, Corrosion scales on various steels after exposure to liquid lead-bismuth eutectic", Journal of Nuclear Materials 450 (2014) 244- 255
(9) V. Tsisar, C. Schroer, O. Wedemeyer, A. Skrypnik, J. Konys, “Corrosion behavior of austenitic steels 1.4970 316L and 1.4571 in flowing LBE at 450 and 550°C with 10-7 mass% dissolved oxygen”, Journal of Nuclear Materials 454 (2014) 332–342
(10) V. Tsisar, C. Schroer, O. Wedemeyer, A. Skrypnik, J. Konys, “Compatibility of ferritic/martensitic steel T91 with flowing LBE containing 10-7 mass% dissolved oxygen at 450 and 550°C”, Proc. of the 2014 22nd International Conference on Nuclear Engineering ICONE-22, July 7-11, 2014, Prague, Czech Republic (paper-30399)
(10) V. Tsisar, C. Schroer, O. Wedemeyer, A. Skrypnik, J. Konys, “Corrosion behavior of austenitic steels 1.4970 316L and 1.4571 in flowing LBE at 450 and 550°C with 10-7 mass % dissolved oxygen”, International Conference on Nuclear Engineering ICONE-22, July 7-11, 2014, Prague, Czech Republic; oral presentation
(11) V. Tsisar, C. Schroer, O. Wedemeyer, A. Skrypnik, J. Konys, “Corrosion response of 1.4970 316L and 1.4571 austenitic steels to flowing LBE with 10-7 mass % dissolved oxygen at 400, 450 and 550°C”, 12th International Workshop on Spallation Materials Technology, 19-23 October 2014, Bregenz, Austria; oral presentation
(12) A. Heinzel, A. Weisenburger, G. Müller, “Corrosion behavior of austenitic steels in liquid lead bismuth containing 10-6 wt% and 10-8 wt% oxygen at 400-500°C”, Journal of Nuclear Materials 448 (2014) 163–171
(13) O. Klok, K. Lambrinou, S. Gavrilov, I. De Graeve, “Preliminary results on the temperature dependence of the dissolution corrosion behavior of 316L austenitic stainless steels in contact with static LBE”, Proc. of the SEARCH/MAXSIMA 2014 International Workshop, pp. 645-657, 2014, Editors: J. Pacio, A. Weisenburger, Th. Wetzel; oral presentation
(14) K. Lambrinou, E. Charalampopoulou, R. Delville, T. Van der Donck, “On the dissolution corrosion of austenitic stainless steels in contact with static LBE”, Proc. of the SEARCH/MAXSIMA 2014 International Workshop, pp. 659-664, 2014, Editors: J. Pacio, A. Weisenburger, Th. Wetzel; oral presentation
(15) K. Lambrinou, “Introduction to liquid metal corrosion effects for Gen-IV LFRs”, Spring School on “Modelling of Corrosion for Nuclear Reactors”, Saclay, France, 31 March – 4 April 2014; invited lecture
(10) K. Lambrinou, P. Hosemann, “Factors affecting the dissolution corrosion of 316L austenitic stainless steels in contact with static LBE”, 2015 TMS Annual Meeting and Exhibition, Orlando, Florida, USA, 15-19 March 2015; oral presentation
(16) D. Frazer, M. Popovic, C. Cionea, S. Parker, C. Rose, O. Panova, M. Asta, M. Caro, S.A. Maloy, F. Rubio, K. Lambrinou, E. Stergar, P. Hosemann, “Heavy liquid metals as potential heat transfer fluids in nuclear reactors”, 2015 TMS Annual Meeting and Exhibition, Orlando, Florida, USA, 15-19 March 2015; oral presentation
(17) C. Schroer, “Oxidation and solution of steels in liquid lead–bismuth eutectic (LBE)”, Proc. of the 19th International Corrosion Congress, Jeju, Korea, paper no. TB 1-3 (Corrosion Science Society of Korea (CSSK) 2014)
(18) C. Schroer, O. Wedemeyer, J. Novotny, A. Skrypnik, J. Konys, “Compatibility of ferritic/martensitic steels with flowing lead–bismuth eutectic at 450°–550°C and 10–6 mass% dissolved oxygen”, Proc. of the 20th International Conference on Nuclear Engineering and the ASME 2012 Power Conference, Anaheim, California, paper no. ICONE20Power2012-54277 (ASME 2012)
(19) C. Schroer, O. Wedemeyer, J. Novotny, A. Skrypnik, J. Konys, “Performance of 9% Cr steels in flowing lead–bismuth eutectic at 450 and 550°C, and 10−6 mass% dissolved oxygen”, Nuclear Engineering and Design 280 (2014) 661–672
(20) G. Barbieri, M. Cesaroni, L. Barbieri “SALDATURA SMAW E GTAW DI P91: Qualifica dei materiali d’apporto e processi per la saldatura di componenti per reattori GEN IV in accordo alla normativa RCC MR”, Giornate Nazionali di Saldatura 8.
(21) X. Boulnat, I. Hilger, J. Hofmann, C. Testani, F. Bergner, Y. De Carlan, F. Ferraro, A. Ulbricht , “Comparative study on ODS 14Cr steel fabrication”,
A special issue of journal of nuclear material is under peer review with the main paper presented during Final MATTER Workshop:
(1) S.Gavrilov (SCK-CEN), Slow Strain Rate Test for investigation of materials susceptibility to Liquid Metal Embrittlement
(2) E. Altstadt (HZDR), Critical evaluation of the small punch test as a screening procedure for mechanical properties
(3) M. Serrano (CIEMAT), Liquid metal embrittlement assessment of T91 in contact with LBE by using small punch tests
(4) C. Heintze (HZDR), Ion irradiation and nanoindentation as a screening test for irradiation effects on neutron-irradiated ferritic/martensitic high-Cr steels
(5) C. Schroer (KIT), Investigating steel corrosion caused by liquid lead or lead-bismuth eutectic
(6) C. Schroer (KIT), A. Gessi (ENEA), Round robin test of steel corrosion in oxygen-containing lead-bismuth eutectic
(7) Lida Magielsen (NRG), Fracture toughness testing in Liquid metal
(8) J. Klecka (CVR), Character of liquid metal embrittlement of the steel T91 in contact with lead-bismuth eutectic
(9) Michael Mahler, Jarir Aktaa (KIT), Prediction of fracture toughness based on experiments with sub-size specimens in the ductile and brittle regime
(10) F. Özgün Ersoy (SCK-CEN), Investigating liquid-metal embrittlement of T91 steel by fracture toughness tests
(11) G. Aiello (CEA), Pre-Normative R&D for Codes And Standards
(12) S. Gavrilov (SCK-CEN), Application of codes to LBE environment
(13) Karl-Fredrik Nilsson (JRC), Web based Data base MatDB
(14) Hyeong-Yeon Lee (KAERI), Application of Elevated Temperature Design Rules to High Temperature Components and Design Code Issues
(15) Kuo Zhang, Jarir Aktaa, Characterization and Modeling of the Ratcheting Behavior of the Ferritic-Martensitic Steel P91
(16) M. Angiolini (ENEA), Ratcheting of a P91 Cylindric shell under thermal cycling
(17) P. Matheron, G. Aiello (CEA), Raatcheting rules for P91 componets: improvement of the efficiency diagram method in RCC-MRx
(18) C. Testani (CSM), Study of application of Laser Welding for built GEN IV SS316 components
(19) Attila Kovács (AEKI), Irradiation and Post-Irradiation Experiments for the 316 LN and P91 Materials
(20) G. Barbieri (ENEA), Influence of deposition ratio and PWTH duration time on toughness and ductility of P91 welds
List of Websites:
www.matterfp7.it
contact details: marco.utili@enea.it