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The current development of accident-tolerant fuel (ATF) claddings is aiming at the improvement of the high temperature integrity of new solutions over the existing zirconium claddings. The characterization of the new proposed cladding materials is essential in order to assess their feasibility in commercial use. The cladding material should tolerate loss of active cooling for a significant period of time without failing. This paper describes the experimental work done at VTT on a candidate cladding solution by using a high temperature steam furnace capable of working at temperatures up to 1600°C. Experiments were performed on CrN coated AISI316L coupon and NiCrAlY coated 800H tube specimens at the temperatures of 1100°C, 1200°C and 1300°C in flowing steam conditions. Post-test analyses are reported in terms of weight change measurements and microstructure observations by using state of the art electron microscopes.

Concerns with greenhouse gas emissions and the uncertainty of long-term supply of fossil fuels have resulted in renewed interest in nuclear energy as an essential part of the energy mix for the future. Many countries worldwide including Canada, China, and EU are currently undertaking the design of generation IV supercritical water-cooled reactor (SCWR) with higher thermodynamic efficiency and considerable plant simplification. The identification of appropriate materials for in-core and out-of-core components to contain the supercritical water (SCW) coolant is one of the major challenges for the design of SCWR. This study is carried out to evaluate the oxidation/corrosion behaviors of bare alloy 214 and NiCrAlY coated 214 under SCW at a temperature of 700°C/25 MPa for 1000 h. The results show that chromium and nickel based oxide forms on the bare surface after exposure in SCW for 1000 h. A dense and adhered oxide layer, consisting of Cr2O3 with spinel (Ni(Cr, Al)2O4), was observed on NiCrAlY surface after 1000 h in SCW.

Small modular reactors (SMRs) are a very attractive option in the near future due to many advantages compared to present day large power plants. One of the major advantages are the improved safety features which are seen as the most important aspect of the innovative reactor designs especially after Fukushima accident in 2011. Another important point of view is the reduced capital cost through design simplification/smaller size in current economic situation. Besides electricity generation, also a wide variety of other applications are available in SMR concepts including process heat production, desalination and hydrogen generation. In addition, SMRs are a potential option for developing countries where grid capacity is not sufficient for large NPPs.For assessing SMR design attractiveness in national and international level, there is a need to have a closer look at the licensing requirements and related safety issues. This project describes some of the safety features of selected SMR designs, as well as their development status in national and international level, which are discussed in terms of selected disciplines available at VTT. Further, the question is answered about the possible know-how needed at VTT in order to better assist our potential customers in the future.Based on the literature, it is shown that SMRs are seen as a very promising reactor design in many countries world-wide. In this work, we were focusing mainly on LWR type SMRs due to the their proven technology and knowledge internationally. Many safety features still need to be studied more closely in order to assess the potential difficulties in licensing an SMR plant in Finland, and possibly suggest appropriate changes in the licensing procedures. The extent of passive safety systems and especially the feature of the control room according to which it is possible to control and monitor several reacors at the same time necessitates the new desing of operator work and, consequently, new design of many human factors engineering related activities. There are also some technical challenges that need to be evaluated and tested before deployment of SMRs. This includes especially challenges related to functionality of passive safety systems, evaluation of possible off-site consequences, control room design as well as manufacturing and non-destructive testing of reactor components. In this report, some suggestions are also made on future directions of SMR research at VTT as well as an internal strategy how to follow international development on this topic.

For higher efficiency and competitive overall performance, it would be an advantage to be able to safely exceed the highest operational values, currently up to about 600-620°C/25-30 MPa in supercritical (SC) boiler plants. Under such operating conditions the oxidation resistance in SC water tends to limit the potential for further improvement of efficiency. The way to increase the oxidation resistance of traditional austenitic boiler tubes e.g. TP 347H is to do additional cold work on the boiler tube inner surface. In the current paper the effect of cold work on the oxidation resistance of TP347H and TP347HFG has been evaluated by shot peened samples with different parameters and subjecting those samples to supercritical oxidation exposure. The results show an improvement in the oxidation resistance of the alloys, especially in the large grained alloy TP347H. Also the uniformity of the deformation layer was seen to have an influence on the oxidation resistance, since the oxide nodules start to grow at locations with the thinnest or no deformation layer.

Slow strain rate tests on several SCWR candidate materials were done over the last 10 years at VTT. The test materials were austenitic stainless steels with 15-23% Cr and 10-15% Ni and oxide dispersion strengthened ferritic/martensitic steels with 12-20% Cr. The tests were performed in supercritical water at 250 bar in the temperature range of 500-650oC with 100-150 ppb dissolved oxygen. With regards to the stress corrosion susceptibility and ductility, the best performer in the tests was AISI 316L steel. On the other hand, large heat to heat variation can be expected: another heat of 316L was very susceptible to intergranular stress corrosion cracking. DL-EPR test revealed that at least some heats of low carbon 316 are subject to fast thermal sensitisation at 550oC in supercritical water.

A number of SSRT tests were performed in order to determine best behaving stainless steel for fuel cladding for supercritical water at 550oC with regards to SCC resistance. The materials were limited to austenitic stainless steels which have a code case for reactor internal applications and have a relatively good general corrosion resistance in supercritical water below 550oC. The initial results indicated that SS 316L has good enough SCC resistance, but further tests on another heat of 316L showed adverse results. The former heat showed no SCC whereas the latter heat showed very high SCC susceptibility in as-received (mill-annealed) condition, shot peened condition and in as-received condition with a TIG weld in the middle of the gauge section of the SSRT specimen. Further electrochemical testing of the degree of sensitisation revealed that the heat giving good enough SCC resistance in SSRT test was also partially sensitised.

The corrosion behavior and oxide layer structure of two 14Cr ODS (Oxide Dispersion Strenghtened) steels with or without alloying element of Al have been investigated in supercritical water (SCW) at 650℃/25MPa for exposure times of 100h, 300h and 1000h. The oxide structure was studied using scanning electron microscope (SEM), energy dispersion spectroscopy (EDS) and X-ray diffraction (XRD). Both ODS alloys showed a three-layer oxide structure after 1000h of exposure in SCW. The 14Cr-Al samples exhibited a structure with Fe2O3 in the outer layer, a mixture of (Cr, Fe)3O4 and (Al,Cr)2O3 in the inner layer, and a mixture of metal and oxide grains Al2O3 in the diffusion layer. The outer layer of Al-free samples consist of Fe2O3, the inner layer and the diffusion layer contain (Fe, Cr)3O4 and Cr2O3, respectively. Cracks, pores and exfoliations could be observed for both alloys after SCW exposure up to 1000h.

Incoloy 800H/HT is one of the promising candidates for the fuel cladding materials in supercritical water-cooled rectors. In the present study, with the help of a specific thermomechanical processing, we studied the effects of grain size and grain boundary character distribution on the oxidation behavior of Incoloy 800H/HT in supercritical water (SCW). The processed samples were exposed to supercritical water at 600 ºC and 25 MPa for 100, 300 and 1000 hours. The results showed that grain size and grain boundaries are important factors that affect the oxidation behaviour of Incoloy 800H/HT in supercritical water. We also found that the thermo-mechanical processing improves the adhesion and integrity of oxide scale.

In the present paper, the kinetic and transport parameters of inner and outer layer growth on two ODS (Oxide Dispersion Strengthened) alloys, PM2000 and MA956, is presented. Parameters are estimated using an upgraded model that assumes that the growth of the outer layer is governed by the transport of cations through the inner layer via an interstitialcy mechanism. The updated model is able to reproduce accurately the depth profiles of individual constituent elements in the inner and outer layers, as well as in the diffusion/transition layer of the alloy situated between the inner layer and the bulk substrate. As a result of the calculations, kinetic and transport parameters of inner and outer layer growth are estimated for oxidation times from 600 to 2000 h at 650 °C. The main conclusion that can be drawn from the calculations is that the most of the rate constants and also the diffusion coefficients tend to decrease with time of oxidation at a constant temperature.

The advantage of using supercritical water (SCW) systems for power generation is based on the increased thermodynamic efficiency when operating at higher temperature and pressure. This has been realised for many years in medium to large size supercritical fossil plants. High efficiency in power generation is not only desirable because of economical reasons but also for enhanced environmental performance meaning reduced quantity of forming ash and emitted pollutants including CO2. Steam oxidation has become an important issue for steam power plants as operating temperatures increase from current to 650°C and even higher. To achieve these higher steam values new materials are needed. This paper presents results of the oxidation performance of potential new alloys (FeCrAlY, NiCrAl and Sanicro 25) in a supercritical water autoclave environment at 650°C/250 bar, with 150 ppb dissolved oxygen.

The main target of the HPLWR Phase 2 follow-up FP7 project “Fuel Qualification Test for Super Critical Water Reactor” is to make significant progress towards the design, analysis and licensing of a fuel assembly cooled with supercritical water in a research reactor which also includes the material research on those in-core materials which could be licensed in near future. In frame of WP4 - Pre-qualification, stress corrosion cracking (SCC) resistance of three selected austenitic stainless steels 08Cr18Ni10Ti (equivalent of AISI 321), AISI 347H and AISI 316L was investigated by conducting slow strain rate tensile tests in an autoclave in 550°C supercritical water. SCC susceptibility was evaluated using three criteria: loss of mechanical properties compared to those in inert environment, fractographic analysis of the fracture surfaces and finally analysis of secondary cracks. Obtained results were compared to those available in the literature and utilization of these results is discussed. Based on the results of SSRT tests and following fractographic analysis, 316L was selected as the best performer.

For improved efficiency and reduced emissions, the future power plants need to operate at high temperatures and pressures, which however are limited by the durability of conventional materials such as ferritic steels. Steam oxidation of a number of coatings (Al slurries, thermal spraying, CVD siliconizing and nickel plating) has demonstrated the feasibility of coatings to improve the oxidation resistance. Al slurry coatings combine good high temperature oxidation resistance through the growth of an Al2O3 layer and the possibility to apply the coating on an industrial scale at moderate cost. This work aimed to test the oxidation performance of coatings and reference alloys in ultrasupercritical (USC, 650ÛC / 250 bar) water. The tested materials included Al slurry coating on ferritic 9%Cr steel and nickel-base A263 substrates, and bulk P92, MARBN and A263 alloys as reference specimens. Oxidation resistance was tested by exposure to flowing supercritical water (SCW) with 125 ppb dissolved oxygen at 650ÛC/250 bar up to 1000 h.

In this study, we evaluated the oxidation resistance of two Fe-Ni-Cr based alloys (stainless steel 316L and incoloy 800H/HT) in air and supercritical water at 600 ºC. Results indicate that Incoloy 800H/HT shows considerably better oxidation resistance in supercritical water, compared to stainless steel 316L, and therefore it is a better candidate as a structural material for SCWR. The difference in chemical composition of the alloys was found to be the main reason for better oxidation resistance of incoloy 800H/HT as it has higher amount of chromium and nickel and, on the contrary, lower amount of iron. This resulted in the formation of a protective layer on the surface which significantly retarded the oxidation in supercritical water environment.

Austenitic stainless steels are attractive candidate materials for the SCWR (Supercritical Water Reactor) concept, with long and successful history in nuclear power generation. Unfortunately their relatively high oxidation rate at estimated peak temperatures of 650 - 850°C (depending on the concept) may severely limit their use as fuel cladding material. However, it has been shown cold working of the exposed surfaces by shot peening can beneficially retard high temperature oxidation of austenitic steels. This report describes the effect of shot peening on the oxidation behavior of three commercially available austenitic steels, 316NG, 347H and 347HFG in SCW at 650°C I 25 MPa for up to 1800 h. Within the tested range, improved oxidation resistance was observed on some samples with certain shot peening parameters.

Performance and efficiency of more advanced reactors like Super Critical Water Reactors (SCWRs) is linked to more severe service conditions. Obviously, functioning under super critical water condition needs construction materials which have superior mechanical properties and oxidation resistance. In this work, we studied effect of thermo-mechanical process on the oxidation resistance of austenitic stainless steel 316L in super critical water. Thermo-mechanical process consisted of multi-step cross-rolling at room temperature followed by annealing in 1050 ºC for 5 minutes per millimeters of thickness. Precursors then were oxidized in super critical water at 600 ºC and 25 MPa for 100, 300 and 1000 hours. Results indicated that texture is very important to oxidation resistance. Samples with {103} texture and {110} texture showed the worst and the best oxidation resistance, respectively. Results also indicated that thermo-mechanical processing improves adhesion between oxide scale and austenite due to reduction of high energy grain boundaries fraction.

To improve the efficiency of fossil fuel power plants the operating temperatures and pressures have a need to be increased. However, at high temperatures the steam side oxidation resistance becomes a critical issue for the steels used especially at the final stages of superheaters and reheaters. Apart from the chemical composition of the material, surface condition is a major factor affecting the oxidation resistance in steam and supercritical water. In this paper, stainless boiler steels (UNS S34710, S31035, S31042, S30942) are investigated in elevated steam oxidation conditions. Tests were conducted in a supercritical water autoclave environment (250 bar, with 125 ppb dissolved oxygen and a pH of 7) at 625°C, 650°C and 675°C for up to 1000h. Materials were tested with different surface finishes shot peened, milled and spark eroded. The results show an influence of surface finish on the early stages of oxidation. Oxides formed on more rough surfaces were more adherent, and thinner than on a ground surface. Tested materials with a chromium over 20 wt-% showed similar oxidation behavior at the lowest temperature, though oxides formed were dependent on the small alloying agents (Cu, Mn). More variation could be seen with increasing temperature.

Plasma coatings, applied by physical vapour deposition, as well as plasma surface treatments could find interesting applications in the nuclear field, both for existing types of reactors as well as for generation IV reactors (such as the supercritical water reactor). A large variety of industrially available coatings exists. Among these, the following coatings (PVD) were investigated: TiAlN, CrN and ZrO2. Small coated samples (AISI 316L and Inconel 600) were exposed to supercritical water (at 650 °C, 250 bar) during about 1000 h at an oxygen content of 150 ppb. Whereas the performance of the TiAlN and the ZrO2 coatings was not satisfactory, the CrN coating was found to provide a stable and protective corrosion barrier. The CrN coated Inconel 600 base metal shows that chromium creates approximately 0.5 µm wide stable natural passivation oxide layer which protects the surface from further corrosion attack during exposure to SCW. Weight gain measurements on the CrN coated samples were performed as well but did not show any significant change after exposure.

At high temperatures the water/steam side oxidation resistance is an important issue for boiler components like superheaters, where the growing internal oxide decreases heat transfer and increases surface temperature. This is increasingly critical for modern plants that aim for high efficiency from elevated levels of operating temperature and pressure. In this paper, a new simple model is presented to describe the steam/water side oxidation rate of common alloys for the high temperature end of the boiler plant. The model is based on the chemical composition of the alloys, applying equivalence expressions for collapsing the multi-variable problem to a more easily manageable setting with a combined parametric oxidation (POX) model. For verification, a range of experimental water/steam oxidation data from numerous tested alloys has been compared with the model predicted oxide growth. The results suggest a surprisingly good agreement with the model and test data. Although the optimised model parameters differ between e.g. steels and nickel alloys, as expected due to difference in the oxides, they remain very similar within a given material class, e.g. for all tested steels.

The materials selection for Gen IV supercritical water reactor in-core components faces major challenges due to severe operating condition. The current Canadian pressure-tube supercritical water (SCW) concept design requires core outlet temperature of 650°C at 25 MPa with peak cladding temperature reaching as high as 850°C. Aside from other challenges, there is currently very limited data on candidate stainless steels' corrosion resistance under this operating condition. Austenite stainless steels that include 316L, 304H, AL6XN and 310S that cover a wide range of Cr contents are promising candidates for either direct in core applications or as base coating materials. However, their corrosion resistance has not been tested systematically under this stringent condition. In this study, these types of stainless steels are tested, and their corrosion resistances are compared.

General corrosion tests and slow strain rate tensile tests (SSRT) were done on 9-20% Cr oxide dispersion strengthened (ODS) steels in supercritical water at 550 and 650oC with 100-150 ppb dissolved oxygen. Based on the results, Al-alloyed high Cr ODS steels PM2000 and MA956 have superior corrosion resistance when compared to 9-14% Cr ODS steels. In SCW at 550oC and above, it is considered that the minimum Cr-content for thin walled components should be higher than 14%. The results indicate that the ODS steels are less susceptible to stress corrosion cracking (SCC) than austenitic stainless steels or Ni-base alloys in supercritical water. However, more extensive studies with different water chemistries and strain rates should be made.

Materials selection for Gen IV supercritical water reactor in-core components faces major challenges due to severe operating condition. Beside high-temperature mechanical properties, SCC and general corrosion resistance of commercially available materials are also major focuses in recent research. Recent reports suggest that certain types of surface modifications are beneficial to SCW corrosion resistance. In this study, the effects of surface modifications are investigated, and possible mechanisms for improved corrosion performance are discussed