general corrosion of iron, nickel and titanium alloys …...general corrosion of iron, nickel and...

GENES4ANP2003 Sep 15-19 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 Kasahara1) Jiro Kuniya1) Kumiaki Moriya2) Norihisa Saito3) Shigenori Shiga4)

1) Hitachi Research Laboratory Hitachi Ltd Hitachi-shi Ibaraki 319-1292 Japan Tel +8129-427-5026 Fax +8129-427-5057 Email shigekikgmhrlhitachicojp 2) Power amp Industrial Systems Hitachi Ltd Hitachi-shi Ibaraki 317-8511 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 374degC 221 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 informa-tion essential to demonstrate SCPR technologies in Japan

One of the major technical issues is development of materials for the SCPR fuel claddings and core compo-nents The materials will be used under the high-pressurized water in the temperature range of about 300degC - 550degC 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 550degC were conducted to evaluate the mechanical in-tegrity 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 al-

loys 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 ther-modynamic features brought by the adoption of supercriti-cal-water coolant Primarily the supercritical turbine cycle can achieve thermal efficiency over 40 according to the low of thermodynamics This efficiency is high while ad-vanced 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 re-sults in elimination of recirculation system and steam gen-erators as well as steam-water separation systems Com-pared 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 amp D pro-jects 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 wa-ter-cooled nuclear power system in next generation and DOE in 2002 of USA funds feasibility studies which com-

pose the roadmaps of the technical development In advance of these movements a joint team consisting

of University of Tokyo Kyushu University Hokkaido Uni-versity 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 objec-tives 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 pro-ject consists sub-themes concerning the above three catego-ries In the material development sub-theme candidate ma-terials are screened from the viewpoints of mechanical in-tegrities corrosion resistance and radiation damage proper-ties 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 550degC

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 400degC Therefore alternative al-loys should be developed for SCPR fuel claddings from the view points of mechanical integrities corrosion and radia-tion 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 ma-terial 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 de-velopment 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 fer-riticmartensitic 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 oxida-tion (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 sig-nificant corrosion of the process units For this reason it is plausible for some kinds of high corrosion resistance mate-rials 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 dam-age 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 swell-ing 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 informa-tion was very important not only to project the selection of test materials and examinations at the beginning of the ma-terial 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 per-formed at the temperature range estimated in SCPR to evaluate high-temperature mechanical integrities General corrosion and Stress Corrosion Cracking (SCC) susceptibil-ity 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 set-ting the materials subjects to be solved at the next phase The promising material candidates will be proposed for further investigation under neutron radiation field Mean-while 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 car-ried 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

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 tem-perature and 550degC in air Strain rate was 5times10-3 sec Ten-sile strength and total elongation were obtained from analy-sis 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 alu-mina tube Alumina spacers were set between the speci-mens 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 sec-tion of a supercritical pressurized water loop The test sec-tion and associated loop were assembled to simulate the SCPR condition up to 30 MPa 600degC Water chemistry condition is set up for the corrosion tests using the similar way adopted in LWRs purification and oxygen concentra-tion control General corrosion test conditions are shown as follows Temperatures 290 380 550degC Pressure 25 MPa Dissolved oxygen 8 ppm Conductivity less than 01 microScm Test period 1800 ks (500 h)

Temperatures of 290degC and 550degC are chosen with con-cerning the upper and lower temperatures in the SCPR core environment Nature of water is known to change drasti-cally and accelerate corrosion around the temperature of the critical point (374degC 221 MPa) therefore corrosion data were obtained at 380degC which is close to the critical point

IV Results and discussion

All of the specimens were examined to obtain their me-chanical properties at room temperature and 550degC 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 550degC Gener-ally tensile stress decreased with increasing the test tem-perature 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 ten-sile strength among stainless steels and Ni base alloys Al-loy 718 with ordinary thermal treatment (alloy 718 (ordi-nary) is one of precipitate hardening alloys therefore the

tensile strength at 550degC is thought to be attributed to the precipitates in the matrix Alloy 718 (ordinary) is known to show stress corrosion cracking susceptibility in aerated wa-ter at about 300degC 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 Ma-terials are usually selected concerning the various condi-tions 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 me-chanical 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 550degC The other titanium alloys also showed higher strength than stainless steels at room temperature but the tensile strength at 550degC de-creased significantly compared with the other alloys Fur-thermore Ti alloys showed low total elongation among the tested materials not only at room temperature but also at 550degC 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 550degC 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 sur-face of the specimens was glossy but the surface was cov-ered with oxide film after corrosion test The color of the oxide film was different It is thought that the different ox-ide 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 speci-mens 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 evalua-tion for corrosion performance of the test materials The

data of weight change which were obtained just after cor-rosion 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 analy-sis 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 materi-als Under the limited interim database of mechanical prop-erty and corrosion performance Ni base alloys are promis-ing in terms of low weight change under general corrosion test and mechanical integrity at 550degC 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 ox-ide films by X-ray diffraction method and net weight loss are undertaken for the corrosion behavior evaluation Fur-ther 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 ldquoDesign Concept of

Once-Through Cycle Supercritical-Pressure Light Water Cooled Reactorsrdquo Proc of The First Int Symp on Supercritical Water-cooled Reactors Design and Technology Nov 6 2000 Univ of Tokyo Japan Paper No 101 (ISBN 4-901332-00-4)

2) J Matsuda N Shimono and K Tamura ldquoSupercritical Fossil Fired Power Plants Designs and Develop-mentsrdquo ibid Paper No 107

3) A Hishinuma Y katano and K Shiraishi ldquoSwelling and Nickel Segregation around Voids in Elec-tron-irradiated Fe-Cr-Ni alloysrdquo J Nucl Mater 103 amp 104 1063 (1981)

(Plant Design)

Optimization of Chemical Composition amp Microstructure as Candidate Alloys

Corrosion properties

UniformCorrosion

Tests

SCCTests


UniformCorrosion

Tests

SCCTests

Screening commercial alloysViability of Existing Materials

Promising Alloys Selection +Alloy Design for Improvement

FBR Fusion reactorStainless steels

(Irradiation resistance)



SCWO

Ni Alloys Ti-Alloys(Corrosion resistance)

SCWO


Supercritical Thermal PowerStainless steels Ni-Alloys

(High-Temp strength Creep)



High-tempTensileTests

Mechanicalintegrities




Radiation effectsElectron Irrad Tests for Void Swelling




Fig 1 Framework of material development for the SCPR core components

Table 1 Test materials selected in this project First group Second group

Austenitic SUS304 SUS316L SUS310S SUS304H SUS316 Stainless 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-25V 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

SUS304 004 059 0028 83 1817 Bal - 1050degCx18 ks(WQ) SUS316L 0023 067 0028 1221 1757 Bal 208 1050degCx18 ks(WQ)SUS310S 005 070 0016 1917 2519 Bal - 1100degCx18 ks(WQ)SUS304H 0073 031 0028 901 1845 Bal - - 1100degCx18 ks(WQ) SUS316 004 070 0033 1017 1682 Bal 212 - 1050degCx18 ks(WQ)

Mod 9Cr-1Mo 011 038 0011 013 859 Bal 095 Nb 0084 A) 12Cr-1Mo-1WVNb 011 023 0018 - 1197 Bal 095 Nb 005 W 099 V 025 B)

(2)Ti base alloys

Alloy H O N C Fe Al V Ti Others Thermal TreatmentTi-6Al-4V 00041 018 001 001 023 633 429 Bal Ylt0001 750degCx2 h (FC)

Ti-3AL-25V 00026 011 0005 008 023 305 300 Bal - 750degC APC) Ti-15V-3Al-3Sn-3Cr 0012 010 0009 - 0221 306 1521 Bal Sn 295 Cr 310 D)

Ti-15Mo-5Zr-3Al 001 012 001 - 003 330 - Bal Mo147 Zr48 E)

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 TreatmentAlloy 800H 008 042 0017 3075 1958 Bal - Al 051 Ti 057 Cu 024 1120degCx18 ks(WQ)Alloy 825 001 010 - 4099 2292 Bal 323 Al 012 Ti 092 Cu 197 1100degCx18 ks(WQ)

Hastelloy C(HC)276 0002 005 0008 5884 1562 543 1559 Co 011 W 392 V 001 1120degCx18 ks(WQ)Hastelloy C(HC) 22 0002 001 lt001 Bal 2150 45 134 C 02 W 30 V 002 1120degCx18 ks(WQ)

Alloy 600 007 019 - 7479 1463 977 - Cu 022 1100degCx18 ks(WQ)

Alloy 625 001 010 0009 6121 2140 361 926 Al 019 Ti 030 Nb+Ta 376 1100degCx18 ks(WQ)

Alloy 718(Ordinary) F)

Alloy 718(Mod) 004 008 0003 532 1790 Bal 307

Al 053 Ti 113 Co 002 Cu 001

Nb+Ta 515 G)

Alloy 690(SHT) 002 033 001 5905 2945 103 - Cu 002 1120degCx18 ks(WQ)A) 1045degC x 18 ks + 780degC x 54 ks (AC) B) 1050degC x 36 ks (AC) + 800degC x 36 ks C) Annealing amp Pickling D) SHT 800degC x 12 ks (AC) -gt 510 degC x 504 ks (AC)

E) SHT 735degC x 36 ks (WQ) -gt 500degC x 504 ks (AC) F) 1010degC x 36 ks (WQ) +705degC x 216 ks (AC) G) 955degC x 36 ks + 718degC x 288 ks (FC)

+ 621degC x 288 ks (AC) (WQ) Water Quench (AC) Air Cooling (FC) Furnace Cooling

(1) Corrosion specimen (2) Specimen and specimen holder

Fig 2 Specimens and specimen holder of corrosion test

Supercritical waterSupercritical waterWater chemistry control sectionWater chemistry control section

Sampling line

DOmicroS

P

Heater

Coo

ler

N2+

O2

Heat exchanger

Control Control TankTank

P

N2

Ion exchange resin

Test vessel


Sampling line

DOmicroS

P

Heater

Coo

ler

N2+

O2

Heat exchanger


P

N2

Ion exchange resin

Test vessel

Sampling line

DOmicroS

P

Heater

Coo

ler

N2+

O2

Heat exchanger


P

N2

Ion exchange resin

Test vessel

Control PanelTest SectionWater Make-up Control PanelTest SectionWater Make-up

Fig 3 Overview and loop configuration of corrosion test facility

Specimen

Bar

Spacer Holder

0 500 1000 1500

Ti-3Al-25V

Ti-15V-3Al-3Sn-3Cr

Ti-15Mo-5Zr-3Al

Ti-6Al-4V

Alloy718(Mod)

Alloy718(Ordinary)

Alloy690(SHT)

Hastelloy C22

Hastelloy C276

Alloy 625

Alloy 600

Alloy 825

Alloy 800H

12Cr-1Mo-1WVNb

Mod 9Cr-1Mo

SUS310S

SUS316L

SUS316

SUS304H

SUS304

Tensile stress (MPa)

550degC

RT

0 20 40 60 80

Total elongation ()

550degC

RT

Fig 4 Mechanical properties of test materials at 550degC and room temperature (RT)

290degC 380degC 550degC

SUS316L

12Cr-1Mo -1WNb

Alloy 690

Fig 5 Typical results of corrosion specimens immersed in super critical pressurized water conditions

10mm

0

100

200

300

400

500

250 300 350 400 450 500 550 600

Temperature(degC)

Wei

ght C

hang

e (m

gdm

^2) 12Cr-1Mo-1WVNb

-20

0

20

40

60

80

100

250 300 350 400 450 500 550 600Temperature (degC)

Wei

ght C

hang

e (m

gdm

^2)

SUS304

SUS310S

SUS316L

-20

-10

0

10

20


Wei

ght c

hang

e (m

gdm

^2)

HC22Alloy 718Alloy 825

-20

-10

0

10

20


Wei

ght c

hang

e (m

gdm

^2)

Alloy 600

Alloy 625

Alloy 690

Fig 6 Temperature dependence of weight change after exposure in Supercritical Pressurized Water

(a) and (b) Stainless steels (c) and (d) Ni base alloys

(a) (b)

(c) (d)

pose the roadmaps of the technical development In advance of these movements a joint team consisting

of University of Tokyo Kyushu University Hokkaido Uni-versity 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 objec-tives 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 pro-ject consists sub-themes concerning the above three catego-ries In the material development sub-theme candidate ma-terials are screened from the viewpoints of mechanical in-tegrities corrosion resistance and radiation damage proper-ties 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 550degC

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 400degC Therefore alternative al-loys should be developed for SCPR fuel claddings from the view points of mechanical integrities corrosion and radia-tion 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 ma-terial 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 de-velopment 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 fer-riticmartensitic 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 oxida-tion (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 sig-nificant corrosion of the process units For this reason it is plausible for some kinds of high corrosion resistance mate-rials 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 dam-age 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 swell-ing 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 informa-tion was very important not only to project the selection of test materials and examinations at the beginning of the ma-terial 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 per-formed at the temperature range estimated in SCPR to evaluate high-temperature mechanical integrities General corrosion and Stress Corrosion Cracking (SCC) susceptibil-ity 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 set-ting the materials subjects to be solved at the next phase The promising material candidates will be proposed for further investigation under neutron radiation field Mean-while 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 car-ried 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












V Conclusions





Acknowledgments






(Plant Design)



UniformCorrosion

Tests

SCCTests


UniformCorrosion

Tests

SCCTests







SCWO


SCWO
























(2)Ti base alloys










Alloy 718(Mod) 004 008 0003 532 1790 Bal 307

Al 053 Ti 113 Co 002 Cu 001

Nb+Ta 515 G)







Sampling line

DOmicroS

P

Heater

Coo

ler

N2+

O2

Heat exchanger


P

N2

Ion exchange resin

Test vessel


Sampling line

DOmicroS

P

Heater

Coo

ler

N2+

O2

Heat exchanger


P

N2

Ion exchange resin

Test vessel

Sampling line

DOmicroS

P

Heater

Coo

ler

N2+

O2

Heat exchanger


P

N2

Ion exchange resin

Test vessel



Specimen

Bar

Spacer Holder

0 500 1000 1500

Ti-3Al-25V

Ti-15V-3Al-3Sn-3Cr

Ti-15Mo-5Zr-3Al

Ti-6Al-4V

Alloy718(Mod)

Alloy718(Ordinary)

Alloy690(SHT)

Hastelloy C22

Hastelloy C276

Alloy 625

Alloy 600

Alloy 825

Alloy 800H

12Cr-1Mo-1WVNb

Mod 9Cr-1Mo

SUS310S

SUS316L

SUS316

SUS304H

SUS304


550degC

RT

0 20 40 60 80

Total elongation ()

550degC

RT



SUS316L

12Cr-1Mo -1WNb

Alloy 690


10mm

0

100

200

300

400

500

250 300 350 400 450 500 550 600

Temperature(degC)

Wei

ght C

hang

e (m

gdm

^2) 12Cr-1Mo-1WVNb

-20

0

20

40

60

80

100


Wei

ght C

hang

e (m

gdm

^2)

SUS304

SUS310S

SUS316L

-20

-10

0

10

20


Wei

ght c

hang

e (m

gdm

^2)


-20

-10

0

10

20


Wei

ght c

hang

e (m

gdm

^2)

Alloy 600

Alloy 625

Alloy 690



(a) (b)

(c) (d)












V Conclusions





Acknowledgments






(Plant Design)



UniformCorrosion

Tests

SCCTests


UniformCorrosion

Tests

SCCTests







SCWO


SCWO
























(2)Ti base alloys










Alloy 718(Mod) 004 008 0003 532 1790 Bal 307

Al 053 Ti 113 Co 002 Cu 001

Nb+Ta 515 G)







Sampling line

DOmicroS

P

Heater

Coo

ler

N2+

O2

Heat exchanger


P

N2

Ion exchange resin

Test vessel


Sampling line

DOmicroS

P

Heater

Coo

ler

N2+

O2

Heat exchanger


P

N2

Ion exchange resin

Test vessel

Sampling line

DOmicroS

P

Heater

Coo

ler

N2+

O2

Heat exchanger


P

N2

Ion exchange resin

Test vessel



Specimen

Bar

Spacer Holder

0 500 1000 1500

Ti-3Al-25V

Ti-15V-3Al-3Sn-3Cr

Ti-15Mo-5Zr-3Al

Ti-6Al-4V

Alloy718(Mod)

Alloy718(Ordinary)

Alloy690(SHT)

Hastelloy C22

Hastelloy C276

Alloy 625

Alloy 600

Alloy 825

Alloy 800H

12Cr-1Mo-1WVNb

Mod 9Cr-1Mo

SUS310S

SUS316L

SUS316

SUS304H

SUS304


550degC

RT

0 20 40 60 80

Total elongation ()

550degC

RT



SUS316L

12Cr-1Mo -1WNb

Alloy 690


10mm

0

100

200

300

400

500

250 300 350 400 450 500 550 600

Temperature(degC)

Wei

ght C

hang

e (m

gdm

^2) 12Cr-1Mo-1WVNb

-20

0

20

40

60

80

100


Wei

ght C

hang

e (m

gdm

^2)

SUS304

SUS310S

SUS316L

-20

-10

0

10

20


Wei

ght c

hang

e (m

gdm

^2)


-20

-10

0

10

20


Wei

ght c

hang

e (m

gdm

^2)

Alloy 600

Alloy 625

Alloy 690



(a) (b)

(c) (d)

(Plant Design)



UniformCorrosion

Tests

SCCTests


UniformCorrosion

Tests

SCCTests







SCWO


SCWO
























(2)Ti base alloys










Alloy 718(Mod) 004 008 0003 532 1790 Bal 307

Al 053 Ti 113 Co 002 Cu 001

Nb+Ta 515 G)







Sampling line

DOmicroS

P

Heater

Coo

ler

N2+

O2

Heat exchanger


P

N2

Ion exchange resin

Test vessel


Sampling line

DOmicroS

P

Heater

Coo

ler

N2+

O2

Heat exchanger


P

N2

Ion exchange resin

Test vessel

Sampling line

DOmicroS

P

Heater

Coo

ler

N2+

O2

Heat exchanger


P

N2

Ion exchange resin

Test vessel



Specimen

Bar

Spacer Holder

0 500 1000 1500

Ti-3Al-25V

Ti-15V-3Al-3Sn-3Cr

Ti-15Mo-5Zr-3Al

Ti-6Al-4V

Alloy718(Mod)

Alloy718(Ordinary)

Alloy690(SHT)

Hastelloy C22

Hastelloy C276

Alloy 625

Alloy 600

Alloy 825

Alloy 800H

12Cr-1Mo-1WVNb

Mod 9Cr-1Mo

SUS310S

SUS316L

SUS316

SUS304H

SUS304


550degC

RT

0 20 40 60 80

Total elongation ()

550degC

RT



SUS316L

12Cr-1Mo -1WNb

Alloy 690


10mm

0

100

200

300

400

500

250 300 350 400 450 500 550 600

Temperature(degC)

Wei

ght C

hang

e (m

gdm

^2) 12Cr-1Mo-1WVNb

-20

0

20

40

60

80

100


Wei

ght C

hang

e (m

gdm

^2)

SUS304

SUS310S

SUS316L

-20

-10

0

10

20


Wei

ght c

hang

e (m

gdm

^2)


-20

-10

0

10

20


Wei

ght c

hang

e (m

gdm

^2)

Alloy 600

Alloy 625

Alloy 690



(a) (b)

(c) (d)







Alloy 718(Mod) 004 008 0003 532 1790 Bal 307

Al 053 Ti 113 Co 002 Cu 001

Nb+Ta 515 G)







Sampling line

DOmicroS

P

Heater

Coo

ler

N2+

O2

Heat exchanger


P

N2

Ion exchange resin

Test vessel


Sampling line

DOmicroS

P

Heater

Coo

ler

N2+

O2

Heat exchanger


P

N2

Ion exchange resin

Test vessel

Sampling line

DOmicroS

P

Heater

Coo

ler

N2+

O2

Heat exchanger


P

N2

Ion exchange resin

Test vessel



Specimen

Bar

Spacer Holder

0 500 1000 1500

Ti-3Al-25V

Ti-15V-3Al-3Sn-3Cr

Ti-15Mo-5Zr-3Al

Ti-6Al-4V

Alloy718(Mod)

Alloy718(Ordinary)

Alloy690(SHT)

Hastelloy C22

Hastelloy C276

Alloy 625

Alloy 600

Alloy 825

Alloy 800H

12Cr-1Mo-1WVNb

Mod 9Cr-1Mo

SUS310S

SUS316L

SUS316

SUS304H

SUS304


550degC

RT

0 20 40 60 80

Total elongation ()

550degC

RT



SUS316L

12Cr-1Mo -1WNb

Alloy 690


10mm

0

100

200

300

400

500

250 300 350 400 450 500 550 600

Temperature(degC)

Wei

ght C

hang

e (m

gdm

^2) 12Cr-1Mo-1WVNb

-20

0

20

40

60

80

100


Wei

ght C

hang

e (m

gdm

^2)

SUS304

SUS310S

SUS316L

-20

-10

0

10

20


Wei

ght c

hang

e (m

gdm

^2)


-20

-10

0

10

20


Wei

ght c

hang

e (m

gdm

^2)

Alloy 600

Alloy 625

Alloy 690



(a) (b)

(c) (d)

0 500 1000 1500

Ti-3Al-25V

Ti-15V-3Al-3Sn-3Cr

Ti-15Mo-5Zr-3Al

Ti-6Al-4V

Alloy718(Mod)

Alloy718(Ordinary)

Alloy690(SHT)

Hastelloy C22

Hastelloy C276

Alloy 625

Alloy 600

Alloy 825

Alloy 800H

12Cr-1Mo-1WVNb

Mod 9Cr-1Mo

SUS310S

SUS316L

SUS316

SUS304H

SUS304


550degC

RT

0 20 40 60 80

Total elongation ()

550degC

RT



SUS316L

12Cr-1Mo -1WNb

Alloy 690


10mm

0

100

200

300

400

500

250 300 350 400 450 500 550 600

Temperature(degC)

Wei

ght C

hang

e (m

gdm

^2) 12Cr-1Mo-1WVNb

-20

0

20

40

60

80

100


Wei

ght C

hang

e (m

gdm

^2)

SUS304

SUS310S

SUS316L

-20

-10

0

10

20


Wei

ght c

hang

e (m

gdm

^2)


-20

-10

0

10

20


Wei

ght c

hang

e (m

gdm

^2)

Alloy 600

Alloy 625

Alloy 690



(a) (b)

(c) (d)

0

100

200

300

400

500

250 300 350 400 450 500 550 600

Temperature(degC)

Wei

ght C

hang

e (m

gdm

^2) 12Cr-1Mo-1WVNb

-20

0

20

40

60

80

100


Wei

ght C

hang

e (m

gdm

^2)

SUS304

SUS310S

SUS316L

-20

-10

0

10

20


Wei

ght c

hang

e (m

gdm

^2)


-20

-10

0

10

20


Wei

ght c

hang

e (m

gdm

^2)

Alloy 600

Alloy 625

Alloy 690



(a) (b)

(c) (d)

general corrosion of iron, nickel and titanium alloys …...general corrosion of iron, nickel and...

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