SCC and Irradiation Properties of Metals under Supercritical-water Cooled Power Reactor Conditions

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1 SCC and Irradiation Properties of Metals under Supercritical-water Cooled Power Reactor Conditions Y. Tsuchiya*, F. Kano 1, N. Saito 1, A. Shioiri 2, S. Kasahara 3, K. Moriya 3, H. Takahashi 4 1 Power and Industrial Systems R&D Center, Toshiba Corp, Yokohama , Japan 2 Isogo Engineering Center, Toshiba Corp, Yokohama , Japan 3 Hitachi Research Laboratory, Hitachi Ltd., Hitachi , Japan 4 Center for Advanced Research of Energy Technology, Hokkaido Univ., Sapporo , Japan The supercritical-water cooled power reactor (SCPR) is expected to be an innovative nuclear power system with considerably higher thermal efficiency and smaller specific volume than the present light water reactors. SCPR coolant condition is predicted to consist of wide temperature range and high dosage of neutron irradiation. Simulated irradiation tests, corrosion tests, SCC tests, and mechanical tests under SCPR conditions are performed in order to select candidate materials. This paper describes the results of SCC tests with austenitic stainless steels and electron irradiation tests for the assessment of irradiation. The susceptibility to intergranular stress corrosion cracking (IGSCC) was evaluated by slow strain rate tests (SSRTs), which were performed using a supercritical-water (SCW) test loop. The results for sensitized Type 304 SS indicated that the SCC susceptibility decreased with increasing temperature. The upper limit temperature of IGSCC susceptibility for sensitized type 304 SS existed at around 673K, above which IGSCC did not occur in oxygenated supercritical water. Simulated irradiation tests were performed for austenitic stainless steels, nickel base alloys and titanium alloys using a high-energy transmission electron microscope (HVTEM) with the electron beam of 1000 kv. Most of the austenitic stainless steel specimens exhibited significant void formation at 723K and 823K, whereas the nickel base alloys showed high-density dislocation loops and no void formation at 723K and 823K. Titanium alloys exhibited no void formation and few dislocation loops at 563K, 723K and 823K, which results suggest high irradiation under SCPR conditions. KEYWORDS: SCPR, SCC, Irradiation, Austenitic stainless steel, SSRT I. Introduction SCPR coolant condition is predicted to consist of the wide temperature range of 563K 823K and the high pressure of higher than 22MPa. High neutron irradiation dosage is foreseen due to the ineffective shielding by low density water. It is necessary to select structural materials and fuel cladding materials used in SCPR from among materials that have high temperature strength, radiation, corrosion and SCC. Fig.1 shows the materials screening process for SCPR fuel cladding. All test materials exhibit some advantage in other industrial fields, but have not been evaluated for all items. Materials with a fatal disadvantage in some respect and no prospect of improvement were not selected. The materials screening process started with a literature review covering the materials used in other industrial fields e.g., for supercritical fossil power plant, supercritical water oxidation (SCWO) or FBR. As a results of the literature review, the properties of each alloy type for corrosion irradiation and high temperature strength are summarized in Table 1. In this study, we selected candidates *Power and Industrial Systems R&D Center, Toshiba Corp. 8, Shinsugita-cho, Isogo-ku, Yokohama, , Japan Tel: , Fax: , yumiko.tsuchiya@toshiba.co.jp SC fossil power plant Stainless steels, Ni-base alloys : Mech. strength, Creep Candidates Chosen from Commercial Alloys Plant conceptual design Irradiation properties Electron irradiation tests SCW oxidation Ni-base alloys,tialloys : Corrosion Evaluation via literature review Mechanical properties High-temp. tensile tests Nuclear plant Stainless steels : Irradiation Cprrosion General corrosion SCC test Optimization of chemical composition and Microstructure for candidate alloys Plant design fuel assembly design Figure1 Process of the material screening for SCPR fuel claddings from among stainless steels, Ni base alloys and Ti base. Stainless steels and Ni base alloys are favorable in terms of high temperature strength and fabrication for application in SCPR. In particular, in regard to development of ultra supercritical (USC) and supercritical thermal power plants,

2 T able 1 Current status of material development for supercritical water systems and nuclear power system Alloy Type Austenitic Stainless steel Ferritic Corrosion Good (TP, WP) Insufficient (TP, LWR) Property, Data base etc. Irradiation Mechanical properties Swelling Good high temp. (FBR) strength Exist improved Good material hightemp. (FBR) strength Research Subjects Improve irradiation Need corrosion data Ni alloy (High-Ni alloy) Good (WP) Embrittlement (FBR) Good high temp. strength Improve irradiation Ti alloy Excellent (WP) Few data available Insufficient thermal creep Need irradiation property data TP: Supercritical thermal power plant materials, WP: Waste processing,, FBR: Fast Breeder Reactor stainless steels have been improved in terms of high temperature strength. High temperature strength of stainless steels and Ni base alloys deteriorates due to void swelling and helium embrittlement. However, irradiation of stainless steels and Ni base alloys is improved by small amounts of additional elements, solid solution hardening and precipitation hardening. In practice, for FBR, where conditions similar to those of SCPR exit, fuel cladding made of Type316SS has been used. These above-mentioned modifications will be taken into account in alloy development. Ti base alloys are selected as candidate materials in view of their good corrosion in SCWO waste disposal systems in which H 2 SO 4, HCl, etc. Considering the application in the SCPR core environment, the candidates are required to have reliability in terms of corrosion and radiation damage, as well as mechanical properties capable of withstanding high temperatures. Therefore, irradiation test, corrosion test and mechanical test were conducted for the test materials in order to obtain a database on the performance of materials in the simulated SCPR environment. identify the fracture mode. Simulated irradiation tests were performed for austenitic stainless steels, nickel base alloys and titanium alloys at 563K, 723K and 823K up to 5 displacements per atom (dpa) at a dose rate of 2x10-3 dpa/sec using a high voltage transmission electron microscope (HVTEM) with 1000 kv electrons. After the irradiation, the microstructure of each specimen was observed with 200kV TEM to evaluate swelling performance. Table2 Test materials selected from the existing industries Stainless steel Austenitic Ferritic Nickel base alloy Titanium base alloy Type304SS, Type304HSS, Type310SSS, Type316SS, Type316LSS 12Cr-1Mo1WVNb, Mod. 9Cr-1Mo Alloy 690, Alloy 718, Alloy 800H, Alloy 825, Hastelloy C22, Hastelloy C276, Alloy 600, Alloy 625 Ti-3Al-2.5V, Ti-6Al-4V, Ti-15V-3Al-3Sn-3Cr, Ti-15Mo-5Zr-3Al II. Experimental procedure Test materials we re selected from among commercially available materials, which are applied to USC, SCWO and in the nuclear fields. The test materials are listed in Table 2. Chemical compositions of the test materials used in the present work are listed in Table 3. After the solution heat treatment, sensitization heat treatment was carried out for Type304SS and Type316LSS for SCC test. The susceptibility to intergranular stress corrosion cracking (IGSCC) was evaluated by slow strain rate tests (SSRTs) using a test loop at the strain rate of 4X10-7 /s, the temperatures range of K and the pressure of 25MPa in high purity water with 8 ppm dissolved oxygen (DO). After SSRTs, fracture surface of the specimens was inspected with scanning electron microscope (SEM) to III. Results & Discussion 1. SCC property The IGSCC area ratio in the fracture surface was measured to evaluate relative SCC susceptibility. The results of SEM inspection of fracture and side surface for sensitized Type304SS and Type316SS are shown in Fig.2 and Fig.3. IGSCC of sensitized Type304SS was not observed in fracture surface at temperature higher than 673K, even though a considerable number of small cracks were observed in side surface. IGSCC of Type316SS was not observed in fracture surface at 563K and 823K. Small cracks were observed in side surface at 823K. Fig.4 shows stress-strain curve of sensitized Type304SS and Fig.5 shows stress-strain curve of Type316SS. Maximum stress of

3 Table 3 Chemical compositions of test materials (wt%) Alloy C Si P Ni Cr Fe Mo Others Thermal Treatment SUS Bal C 30min(WQ) SUS Bal C 30min(WQ) SUS316L Bal C 30min(WQ) SUS310S Bal C 30min(WQ) Alloy Bal Al:0.12, Ti:0.92, Cu: C 30min(WQ) Alloy Cu: C 30min(WQ) Alloy Al:0.19, Ti:0.30, Nb+Ta: C 30min(WQ) O N C Fe Al V Ti Ti-6Al-4V Bal. 750 C 2h(FC) Ti-3Al-2.5V Bal. 750 CAP(Annealing and Pickling) Ti-15V-3Al-3 Bal. SHT800 C 20min (AC) Cr:2.98, Sn:2.91 Sn-3Cr aging: 510 C 14h(AC) Ti-15Mo-3Zr- Bal. SHT735 C 1h(WQ) Mo:14.9, Zr:4.77 3Al aging: 500 C 14h(AC) WQ: Water Quench AC: Air Cooling FC: Furnace sensitized Type304SS increased with increasing temperature until 673K, which tendency corresponds to the decrease of IGSCC ratio. Meanwhile, the maximum stress of sensitized Type 304 SS decreased at temperature higher than 673k, which corresponds to the decrease of mechanical strength. Stress of sensitized Type 316 SS was higher at temperature 563K than at 823K. Fig.6 shows IGSCC ratio together with water properties. These data indicated that the SCC susceptibility decreased with increasing temperature. It was found that the upper limit of IGSCC susceptibility existed for sensitized Type304SS at around 673K, above which IGSCC susceptibility disappeared. The solubility of inorganic substance and the density of water decrease with increasing temperature. In particular, these values decrease rapidly at around the temperature immediately above the pseudo critical point, which tendency is similar to that of IGSCC susceptibility. According to BWR experience, IGSCC is understood to have the feature of local corrosion, consisting of anodic site at crack tip and cathodic site at open surface. Considerable decrease of the electrolytic property of supercritical water probably brings about the discontinuation of crack propagation even if it initiates. 2. Irradiation property Fig.7 shows the micrographs of TEM images of austenitic stainless steels and nickel base alloys irradiated with 1000kV electrons at 563, 723 and 823K. Most of the austenitic stainless steel specimens exhibited significant void formation at 723K and 823K, whereas the nickel base alloys at 723K and 823K showed high-density dislocation formation and few dislocation loops at 723K and 823K, indicating high irradiation under SCPR conditions. Fig.8 shows swelling rate at 563, 723 and 823K for all test materials. For austenitic stainless steel candidate materials to be selected, their swelling rates should be more restrained, similar to those of fast breeder reactor materials. Ni base alloys and Ti base alloys have high swelling. IV. Conclusion The electron irradiation test and SSRT test were performed for the candidate materials for SCPR fuel cladding. Major results were as follows: Test materials were selected from among austenitic and stainless steels, Ni base alloys and titanium base alloys. These materials are applied in existing industrial field e.g. for supercitical water fossil fired power systems, supercritical water oxidation systems, and nuclear power systems. The results for sensitized Type 304 indicated that the SCC susceptibility decreased with increasing temperature under SCPR simu lated conditions. Ni base alloys and Ti base alloys exhibited less void swelling than austenitic stainless steels at 723K and 823K. Acknowledgment This study is funded by the Institute of Applied Energy (IAE), Ministry of Economy, Trade and Industry (METI), Japan.

4 Temp.563K Temp.653K Temp.673K IGSCC:100% IGSCC:21% IGSCC:0% 500m 500m 500m Temp.723K Temp.823K IGSCC:0% IGSCC:0% 500m 500m Figure 2 SEM image of fracture surface and side surface after SSRT. ( Sensitized Type 304 SS)

5 Temp.563K Temp.823K IGSCC:0% IGSCC:0% 500m 500m Figure 3 SEM image of fracture surface and side surface after SSRTType 316L SS Figure 4 Stress-Strain curve of sensitized Type 304 SS

6 Figure 5 Stress-Strain curve of Type 316LSS Figure 6 IGSCC susceptibility of sensitized Type304SS and Type316LSS under supercritical water condition.

7 563 K 723 K 823 K SUS 316L SUS 310S Alloy 600 Ti nm Figure 7 TEM observation after electron irradiation

8 Swelling Rate(%) The swelling rates of Ni-base alloys (Alloy 600, Alloy 625, Alloy 825) and Ti-base alloys 563K 723K 823K Irradiation Temperature(K) Figure 8 Comparison of swelling rate analyzed from TEM observations 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, (in Japanese) No. 101 (ISBN ) 2) J. Matsuda, N Shimono and K. Tamura, Supercritical Fossil Fired Power Plants Design and Developments, ibid. Paper No. 107.