General corrosion of iron, nickel and titanium alloys as candidate materials for the fuel claddings of the supercritical-water cooled power reactor

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1 GENES4/ANP2003, Sep , 2003, Kyoto, JAPAN Paper 1132 General corrosion of iron, nickel and titanium alloys as candidate materials for the fuel claddings of the supercritical-water cooled power reactor Shigeki Kasahara 1), Jiro Kuniya 1), Kumiaki Moriya 2) Norihisa Saito 3), Shigenori Shiga 4) 1) Hitachi Research Laboratory, Hitachi, Ltd., Hitachi-shi, Ibaraki, , Japan Tel: , Fax: , 2) Power & Industrial Systems, Hitachi Ltd., Hitachi-shi, Ibaraki, , Japan 3) PIC, Toshiba Corp. Isogo-ku, Yokohama, Japan 4) Isogo Engineering Center, Toshiba Corp. Isogo-ku, Yokohama, Japan The supercritical-water cooled power reactor uses supercritical phase water (more than 374 C, 22.1 MPa) as a coolant in its once-through type circuit. The SCPR offers potential of high thermal efficiencies more than 40% in its power conversion cycle, while the efficiency of the most recent LWR is about 34%. Furthermore, the steam separation systems and the recirculation systems used in BWRs and the steam generators in PWR are eliminated because no phase change occurs in the SCPR. These inherent features facilitate to simplify the system design, and consequently make possible to reduce the operation and maintenance cost, as well as the construction cost of the plants. Under this background, the technical development project was launched to provide technical information essential to demonstrate SCPR technologies in Japan. One of the major technical issues is development of materials for the SCPR fuel claddings and core components. The materials will be used under the high-pressurized water in the temperature range of about 300 C C. Therefore, they should involve good properties of mechanical integrity, corrosion resistance and radiation damage. Technical issues of the material development were clarified from the literature survey of the material technologies for the fields of supercritical pressure fossil fired power plants, supercritical-water waste processing plants, and nuclear power plants. After that, test materials were nominated from austenitic and ferritic stainless steels, Ni-based alloys and Ti-based alloys. Tensile tests at 550 C were conducted to evaluate the mechanical integrity of the materials under high temperature environments. General corrosion tests were carried out under the SCPR core conditions, which were expected from the current results of the SCPR core design. The database will be applied to the screening of the most promising materials for the fuel claddings and to improvement of the SCPR system design. In this paper, the framework of the material development is introduced, and the data from tensile and general corrosion tests are presented. KEYWORDS: supercritical-water cooled power reactor, stainless steels, nickel base alloys, titanium base alloys, mechanical properties, general corrosion, material development, fuel claddings I. Introduction The supercritical-water cooled power reactor (SCPR), which is an once-through type reactor supplying high-temperature pressurized water to the turbine cycle, is the innovative candidate nuclear power system, because it potentially improves economics mainly through three thermodynamic features brought by the adoption of supercritical-water coolant. Primarily, the supercritical turbine cycle can achieve thermal efficiency over 40% according to the low of thermodynamics. This efficiency is high, while advanced BWR system achieves 34% at most. Secondly, thermal components, such as heat exchangers and turbines, can be compact as well as the buildings accommodating them, since the specific volume of supercritical water is small. Thirdly, no phase change in supercritical regime results in elimination of recirculation system and steam generators as well as steam-water separation systems. Compared with conventional LWRs, these features facilitate design simplification and compaction, so that the SCPR potentially improves economics. Because of these expected advantages, the developments of the SCPR started from 1990 when Professor Oka and his colleagues proposed the first conceptual design of SCPR [1], and some R & D projects were organized not only in Japan but also in Europe, Canada, Russian Federation and USA so far. Conceptual designs of the SCPR have been nominated as the water-cooled nuclear power system in next generation, and DOE in 2002 of USA funds feasibility studies, which com-

2 pose the roadmaps of the technical development. In advance of these movements, a joint team consisting of University of Tokyo, Kyushu University, Hokkaido University, Hitachi, Ltd., and Toshiba Corp. being funded by the Institute of Applied Energy (IAE) in Japan since 2000 has launched a SCPR development project. The main objectives of this project are to provide technical information to improve plant conceptual design and thermo-hydraulics, and to develop candidate materials for fuel claddings and core components for the study of the viability of the large-scale experimental or demonstrative reactor. The project consists sub-themes concerning the above three categories. In the material development sub-theme, candidate materials are screened from the viewpoints of mechanical integrities, corrosion resistance, and radiation damage properties through the examinations simulating the SCPR core conditions. In this paper, the interim results of mechanical properties and general corrosion are introduced and discussed as the material screening tests in material development sub-theme. The examinations are carried out on the materials selected from austenitic steels, ferritic steels, nickel base alloys and titanium base alloys in terms of general corrosion under super-critical water conditions and mechanical integrities at room temperature and 550 C. II. Framework of the candidate material development Zirconium alloys (Zircaloy-2 and Zircaloy-4) are widely used for LWR fuel claddings, but it is difficult to use for the SCPR. The main reason is because the tensile strength of Zircaloy is very low over 400 C. Therefore, alternative alloys should be developed for SCPR fuel claddings from the view points of mechanical integrities, corrosion and radiation damage properties under the SCPR core environment. In the first phase of the SCPR development, a technical goal of the material development sub-theme is to find promising materials worth further examining under neutron irradiation field, which is planned after this development phase. The technical information and database obtained from this material development will be able to apply to the material screening for core internals, as well as fuel claddings. Fig. 1 shows the framework of the material development. Prior to the planning, literature survey was carried out to find out technical issues and final goals of the material development, which are attributed to the specifications of the current SCPR core design. Concerning about the literature survey, we started to select the test materials from austenitic stainless steels, high chromium containing ferritic/martensitic steels (ferritic steels), nickel base alloys and titanium base alloys. These are applied to the existing components of supercritical and ultra supercritical (USC) fossil fired power plants [2] and supercritical water oxidation (SCWO) systems for hazardous waste destruction. Austenitic and ferritic steels are mainly applied to tubes, pipes and turbine components in USC power plants. In SCWO systems, the hazardous wastes can be oxidized to acidic products. Such acidic conditions may result in significant corrosion of the process units. For this reason, it is plausible for some kinds of high corrosion resistance materials, such as nickel base alloys and titanium base alloys, to apply to SCWO system components. On the other hand, nuclear materials have been developed to reduce the damage due to neutron irradiation for the core components and structures of light water reactors, fast breeder reactors and fusion reactor. One of the major issues of radiation damage is swelling, so that lots of efforts are dedicated to decrease it. In particular, austenitic steels are highly sensitive to swelling under relatively high temperature irradiation conditions. Special stainless steels have been developed and applied to the fuel claddings and the core components of the fast breeder reactors. The above-mentioned technical information was very important not only to project the selection of test materials and examinations at the beginning of the material screening, but also to minimize time and cost for the development. Therefore, test materials were selected from commercially available materials, with consideration of USC, SCWO and nuclear fields. Considering application for SCPR core environment, the candidate materials are required to involve the reliability in terms of corrosion properties and radiation damage, as well as mechanical properties at high temperatures. Therefore, simulated irradiation test, corrosion test and mechanical test are conducted for the materials to obtain database under the condition of the SCPR core. Using 1 MeV electron irradiation at the temperature range expected for the SCPR condition simulates radiation damage due to neutron irradiation. Tensile tests are performed at the temperature range estimated in SCPR to evaluate high-temperature mechanical integrities. General corrosion and Stress Corrosion Cracking (SCC) susceptibility are very important in terms of corrosion performance under the SCPR water conditions. Database and knowledge from the examinations are used to screen the most promising materials, as well as for setting the materials subjects to be solved at the next phase. The promising material candidates will be proposed for further investigation under neutron radiation field. Meanwhile, these material data will contribute to precise design of fuel assembly and the SCPR systems. Three sub-themes for the SCPR development are connected closely and carried out effectively. III. Experimental procedure Selected materials in this technical development are shown in Table 1. The test materials are categorized into two groups tentatively. Tensile tests for all of the materials have been already finished, and general corrosion of the materials in the first group has been examined. Chemical

3 compositions of the materials are given in Table 2 (1), (2) and (3). Solution heat treatment (SHT) was performed on most of them, and some of them were heat treated under designated conditions after SHT. Details of the conditions were also shown in Table 2. After the heat treatment, they were machined into specimens. Tensile test specimens were carried out at room temperature and 550 C in air. Strain rate was /sec. Tensile strength and total elongation were obtained from analysis of stress-strain curves. General corrosion tests were carried out on the coupon shape specimens, as shown in Fig. 2(1). The specimens were held by the bar, which was covered with sintered alumina tube. Alumina spacers were set between the specimens. The alumina tube and spacers were used as insulators to avoid setting the specimens and bar under metal touch condition. Configuration of specimens and specimen holder is shown in Fig. 2 (2). Four corrosion specimens for each of the materials were immersed, and corrosion behavior was evaluated through their weight change. As shown in Fig. 3, general corrosion tests are being performed in the test section of a supercritical pressurized water loop. The test section and associated loop were assembled to simulate the SCPR condition up to 30 MPa, 600 C. Water chemistry condition is set up for the corrosion tests using the similar way adopted in LWRs: purification and oxygen concentration control. General corrosion test conditions are shown as follows: Temperatures: 290, 380, 550 C Pressure: 25 MPa Dissolved oxygen: 8 ppm Conductivity: less than 0.1 µs/cm Test period: 1,800 ks (500 h) Temperatures of 290 C and 550 C are chosen with concerning the upper and lower temperatures in the SCPR core environment. Nature of water is known to change drastically and accelerate corrosion around the temperature of the critical point (374 C, 22.1 MPa), therefore, corrosion data were obtained at 380 C, which is close to the critical point. IV. Results and discussion All of the specimens were examined to obtain their mechanical properties at room temperature and 550 C. Tensile strength and total elongation, which were analyzed from stress-strain curves, are shown in Fig. 4. Open bars show data at room temperature, and closed bars at 550 C. Generally, tensile stress decreased with increasing the test temperature. From the comparison of alloy groups, tensile stress and total elongation of Ni base alloys are higher than those of stainless steels. Alloy718 showed the highest tensile strength among stainless steels and Ni base alloys. Alloy 718 with ordinary thermal treatment (alloy 718 (ordinary) is one of precipitate hardening alloys, therefore, the tensile strength at 550 C is thought to be attributed to the precipitates in the matrix. Alloy 718 (ordinary) is known to show stress corrosion cracking susceptibility in aerated water at about 300 C, and alloy 718 with modified thermal treatment was developed to decrease SCC susceptibility and keep high temperature strength. However, total elongation of alloy 718 was low compared with other stainless steels and Ni base alloys. Alloy 825, alloy 690 and Hastelloys showed lower tensile strength than that of alloy 718, but their total elongations are larger than that of alloy 718. Materials are usually selected concerning the various conditions of stress and strain. Depending on the ways how the materials apply to the SCPR core components, these alloys are thought to be promising alloys from a viewpoint of mechanical integrities at high temperature conditions. Ti-15Mo-5Zr-3Al showed the highest tensile strength among the tested materials at room temperature, but the strength was almost halved at 550 C. The other titanium alloys also showed higher strength than stainless steels at room temperature, but the tensile strength at 550 C decreased significantly compared with the other alloys. Furthermore, Ti alloys showed low total elongation among the tested materials not only at room temperature but also at 550 C. In the case of Ti alloys, the temperature dependence of the mechanical property change was different from the other alloys significantly. General corrosion test at temperatures of 290, 380 and 550 C was finished on the first group of the test materials, which consisted of SUS304, SUS316L, SUS310S, Alloy 825, Hastelloy C22 (HC22), Alloy 600, Alloy 625, Alloy 718, Alloy 690, and 12Cr-1Mo-1WVNb. Fig. 5 shows the pictures after general corrosion tests on SUS316L, 12Cr-1Mo-1WVNb, Alloy 690. Before corrosion test, surface of the specimens was glossy, but the surface was covered with oxide film after corrosion test. The color of the oxide film was different. It is thought that the different oxide films are formed and the nature of them would be changed depending on test condition. As the first step of analysis of corrosion data, weight change of the specimens was measured. Fig. 6 shows weight change of the specimens. Plotted data were average value of weight gain (loss) of four samples for each of the test materials. The weight gain was observed in SUS304, SUS316L, SUS310S, and 12Cr-1Mo-1WVNb as shown in Fig. 6(1) and (2). Weight gain of 12Cr-1Mo-1WVNb was highest, and that of SUS310S was lowest among the tested stainless steel. On the other hand, weight change of Ni base alloys was smaller than that of stainless steels, as shown in Fig. 6(3) and (4). However, some of Ni base alloys gained their weight, and the others lost them. Different behaviors of weight change between the stainless steels and Ni base alloys are thought to originate from the difference of the nature of the oxide films. Weight change measurement is a preliminary evaluation for corrosion performance of the test materials. The

4 data of weight change, which were obtained just after corrosion tests, is thought gross information which includes weight gain due to oxidation on the surface as well as weight loss due to dissolution of base metals. Precise analysis of corrosion performance requires database about film thickness, morphology, chemical composition and chemical form. Corrosion test will be continued to obtain these data. As of the end of 2002 fiscal year, screening tests of the candidates for the SCPR core components were finished from the viewpoints of mechanical properties and general corrosion performance on a part of the selected test materials. Under the limited interim database of mechanical property and corrosion performance, Ni base alloys are promising in terms of low weight change under general corrosion test and mechanical integrity at 550 C. However, further discussion should be necessary from the viewpoints of the difference of weight change among Ni base alloys and stainless steels. From this point of view, analysis of the oxide films by X-ray diffraction method and net weight loss are undertaken for the corrosion behavior evaluation. Further data accumulation, including corrosion data of Ti base alloys, is required to understand corrosion mechanism of the candidate materials, and the most promising alloys will be proposed from total evaluation of the database at the end of this material development. V. Conclusions General corrosion tests and electron irradiation tests were carried out to screen the candidate materials for the SCPR core components. As of 2002, selection of test materials and the tests on a part of them were carried out. Major results are shown as follows: Test materials were selected from austenitic and ferritic stainless steels, Ni base alloys and Ti base alloys. These materials were applied to existing industrial fields, which are supercitical water fossil fired power systems, supercritical water oxidation (SCWO) systems, and nuclear power systems. The mechanical properties of Ni base alloys were better than the stainless steels and Ti base alloys from the viewpoint of tensile strengh and total elongation. Temperature dependence ot mechanical properties change of Ti base alloys was different from the other alloys significantly. General corrosion tests were started on a part of austenitic and ferritic steels and Ni base alloys. All of the tested stainless steels and some of Ni base alloys gained their weight, but the other of Ni base alloys lost their weight. Acknowledgments This SCPR development project is funded by the Institute of Applied Energy (IAE), Ministry of Economy, Trade and Industry (METI), Japan. References 1) Y. Oka and S. Koshizuka: Design Concept of Once-Through Cycle Supercritical-Pressure Light Water Cooled Reactors Proc. of The First Int. Symp. on Supercritical Water-cooled Reactors, Design and Technology, Nov Univ. of Tokyo, Japan Paper No. 101 (ISBN ) 2) J. Matsuda, N Shimono and K. Tamura, Supercritical Fossil Fired Power Plants Designs and Developments ibid. Paper No ) A. Hishinuma, Y. katano and K. Shiraishi, Swelling and Nickel Segregation around Voids in Electron-irradiated Fe-Cr-Ni alloys J. Nucl. Mater. 103 & (1981)

5 Supercritical Thermal Power Stainless steels, Ni-Alloys (High-Temp. strength, Creep) SCWO Ni Alloys, Ti-Alloys (Corrosion resistance) FBR, Fusion reactor Stainless steels (Irradiation resistance) Screening commercial alloys Viability of Existing Materials Radiation effects Electron Irrad. Tests for Void Swelling Promising Alloys Selection + Alloy Design for Improvement Radiation effects Electron Irrad. Tests for Void Swelling Mechanical integrities High-temp. Tensile Tests Corrosion properties Uniform Corrosion Tests SCC Tests Optimization of Chemical Composition & Microstructure as Candidate Alloys (Plant Design) Fig. 1 Framework of material development for the SCPR core components Table 1 Test materials selected in this project First group Second group Stainless Austenitic SUS304, SUS316L, SUS310S SUS304H, SUS316 steels Ferritic 12Cr-1Mo-1WVNb Mod. 9Cr-1Mo Nickel base alloy Alloy 825, Hastelloy C22, Alloy 600, Alloy 625, Alloy 718, Alloy 690 Alloy 800H, Hastelloy C276 Titanium base alloy Ti-3Al-2.5V, Ti-15V-3Al-3Sn-3Cr, Ti-6Al-4V, Ti-15Mo-5Zr-3Al Table 2 Chemical compositions and thermal treatment of the test materials (wt%) (1) Stainless steels Alloy C Si P Ni Cr Fe Mo Others Thermal Treatment SUS Bal Cx1.8 ks(wq) SUS316L Bal Cx1.8 ks(wq) SUS310S Bal Cx1.8 ks(wq) SUS304H Bal Cx1.8 ks(wq) SUS Bal Cx1.8 ks(wq) Mod. 9Cr-1Mo Bal Nb: A) 12Cr-1Mo-1WVNb Bal Nb: 0.05, W: 0.99, V: 0.25 B) (2)Ti base alloys Alloy H O N C Fe Al V Ti Others Thermal Treatment Ti-6Al-4V Bal. Y< Cx2 h (FC) Ti-3AL-2.5V Bal C APC) Ti-15V-3Al-3Sn-3Cr Bal. Sn 2.95 Cr 3.10 D) Ti-15Mo-5Zr-3Al Bal. Mo:14.7 Zr:4.8 E)

6 Table 2 Chemical compositions and thermal treatment of the test materials (wt%) (Cont.) (3)Ni base alloys Alloy C Si P Ni Cr Fe Mo Others Thermal Treatment Alloy 800H Bal. - Al: 0.51, Ti: 0.57, Cu: Cx1.8 ks(wq) Alloy Bal Al: 0.12, Ti: 0.92, Cu: Cx1.8 ks(wq) Hastelloy C(HC) Co: 0.11, W: 3.92, V: Cx1.8 ks(wq) Hastelloy C(HC) <0.01 Bal C: 0.2, W: 3.0, V: Cx1.8 ks(wq) Alloy Cu: Cx1.8 ks(wq) Alloy Al: 0.19, Ti: 0.30, Nb+Ta: Cx1.8 ks(wq) Alloy 718(Ordinary) Al: 0.53, Ti: 1.13, F) Alloy 718(Mod) Bal Co: 0.02, Cu: 0.01, Nb+Ta: 5.15, G) Alloy 690(SHT) Cu: Cx1.8 ks(wq) A): 1045 C x 1.8 ks C x 5.4 ks (AC) B): 1050 C x 3.6 ks (AC) C x 3.6 ks C): Annealing & Pickling D): SHT 800 C x 1.2 ks (AC) -> 510 C x 50.4 ks (AC) E): SHT 735 C x 3.6 ks (WQ) -> 500 C x 50.4 ks (AC) F): 1010 C x 3.6 ks (WQ) +705 C x 21.6 ks (AC) G): 955 C x 3.6 ks C x 28.8 ks (FC) C x 28.8 ks (AC) (WQ): Water Quench (AC): Air Cooling (FC): Furnace Cooling Bar Spacer Specimen Holder (1) Corrosion specimen (2) Specimen and specimen holder Fig. 2 Specimens and specimen holder of corrosion test N 2 N 2 +O 2 Ion exchange resin DO Control µs Tank P Cooler Sampling line Heat exchanger Test vessel Heater Water Make-up Test Section Control Panel P Water chemistry control section Supercritical water Fig. 3 Overview and loop configuration of corrosion test facility

7 SUS304 SUS304H SUS316 SUS316L SUS310S Mod. 9Cr-1Mo 12Cr-1Mo-1WVNb Alloy 800H 550 C Alloy 825 Alloy 600 Alloy 625 RT Hastelloy C276 Hastelloy C22 Alloy690(SHT) 550 C Alloy718(Ordinary) Alloy718(Mod) Ti-6Al-4V RT Ti-15Mo-5Zr-3Al Ti-15V-3Al-3Sn-3Cr Ti-3Al-2.5V Fig ,000 1,500 Tensile stress (MPa) Total elongation (%) Mechanical properties of test materials at 550 C and room temperature (RT) 290 C 380 C 550 C SUS316L 12Cr-1Mo -1WNb Alloy mm Fig. 5 Typical results of corrosion specimens immersed in super critical pressurized water conditions

8 Weight Change (mg/dm^2) Weight change (mg/dm^2) (a) 12Cr-1Mo-1WVNb Temperature( C) Temperature ( C) Fig. 6 (c) HC22 Alloy 718 Alloy 825 Weight Change (mg/dm^2) Weight change (mg/dm^2) (b) SUS304 SUS310S SUS316L Temperature ( C) Alloy 600 Alloy 625 Alloy Temperature ( C) Temperature dependence of weight change after exposure in Supercritical Pressurized Water (a) and (b): Stainless steels (c) and (d): Ni base alloys (d)

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