BOOK OF ABSTRACTS SCK•CEN-BA-32
4th international workshop on longterm prediction of corrosion damage in nuclear waste systems.
Bruges, 28.06-02.07.2010
co-sponsored by
SCK•CEN, Boeretang 200, BE-2400 MOL, Belgium Contact:
Bruno Kursten
Tel:
+32 14 33 31 36
Fax:
+32 14 32 35 53
E-mail:
[email protected]
BOOK OF ABSTRACTS SCK•CEN-BA-32
4th international workshop on longterm prediction of corrosion damage in nuclear waste systems.
Bruges, 28.06-02.07.2010
co-sponsored by
SCK•CEN, Boeretang 200, BE-2400 MOL, Belgium Contact:
Bruno Kursten
Tel:
+32 14 33 31 36
Fax:
+32 14 32 35 53
E-mail:
[email protected]
Scientific committee Bruno Kursten - SCK•CEN, Belgium Frank Druyts - SCK•CEN, Belgium Robert Gens - NIRAS/ONDRAF, Belgium Damien Féron - CEA, France Digby D. Macdonald - Pennsylvania State University, USA Tae M. Ahn - NRC, USA Carmen Andrade - IETCC, Spain Hidekazu Asano - RWMC, Japan Gustavo Duffo - CNEA, Argentina En-Hou Han - IMR, China Lawrence Johnson - NAGRA, Switzerland Fraser King - Integrity Corrosion Consulting Ltd., Canada HyukSang Kwon - KAIST, Korea Timo Saario - VTT, Finland Lars Werme - Uppsala University, Sweden Steve Williams - NDA, UK Local organising committee Bruno Kursten - SCK•CEN, Belgium Frank Druyts - SCK•CEN, Belgium Robert Gens - NIRAS/ONDRAF, Belgium Monique Van Geel - SCK•CEN, Belgium
© SCK•CEN Studiecentrum voor Kernenergie Centre d‟Etude de l‟Energie Nucléaire Boeretang 200 BE-2400 MOL Belgium http://www.sckcen.be Status: Unclassified RESTRICTED All property rights and copyright are reserved. Any communication or reproduction of this document, and any communication or use of its content without explicit authorization is prohibited. Any infringement to this rule is illegal and entitles to claim damages from the infringer, without prejudice to any other right in case of granting a patent or registration in the field of intellectual property. Copyright of the individual abstracts and papers rests with the authors and the SCK•CEN takes no responsability on behalf of their content and their further use. SCK•CEN, Studiecentrum voor Kernenergie/Centre d'Etude de l'Energie Nucléaire Stichting van Openbaar Nut – Fondation d'Utilité Publique - Foundation of Public Utility Registered Office: Avenue Herrmann Debroux 40 – BE-1160 BRUSSEL Operational Office: Boeretang 200 – BE-2400 MOL
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Contents Introduction ....................................................................................................................... 7 Programme ....................................................................................................................... 9 Risk-Informed Performance-Based Application of Container/Canister Corrosion in Nuclear Waste Management ........................................................................................................ 15 Review of the Performance of Selected Metals as Canister Materials for UK Spent Fuel and/or HLW ..................................................................................................................... 16 The ONDRAF/NIRAS safety strategy and the strategic choices in the Belgian Supercontainer design .................................................................................................... 18 Review of the corrosion studies of the metallic barrier in geological disposal conditions with respect to the Belgian Supercontainer design .......................................................... 20 Some important issues in the electrochemistry of carbon steel in simulated concrete pore water ............................................................................................................................... 22 Further studies on the effect of irradiation on the corrosion of carbon steel in alkaline media .............................................................................................................................. 23 Propagation behavior of general and localized corrosion of carbon steel in simulated groundwater environment under aerobic condition .......................................................... 25 Application of X-ray Tomography for Characterisation of Localised Corrosion and Stress Corrosion Cracking in Waste Container Materials ........................................................... 27 An experimental approach to study the effect of chloride deposition on the stress corrosion behaviour of 316L stainless steel used for intermediate level radioactive waste containers ....................................................................................................................... 28 Field corrosion experiments in clay in anoxic conditions at high temperature .................. 30 Crevice corrosion testing methods for measuring the repassivation potential of Alloy 22 33 Sulphide Induced Stress Corrosion Cracking of CuOFP in Groundwater ......................... 35 The rate controlling reactions for copper corrosion in anaerobic aqueous sulphide solutions .......................................................................................................................... 36 Further studies of in situ corrosion testing of miniature copper-cast iron nuclear waste canisters.......................................................................................................................... 37 The corrosion rate of pure copper in an oxic bentonite/saline groundwater environment . 40 The Scientific Basis for the Corrosion of Copper in Water and the Implications for Canister Lifetimes.......................................................................................................................... 41 Long-term integrity against corrosion of titanium used for TRU waste container .............. 42 Long-Term Integrity of Overpack Closure Weld for HLW Geological Disposal, (II) Corrosion Property under Anaerobic Condition ............................................................... 47 Corrosion Issues Related to Disposal of 316SS-Zirconium Metal Waste Forms under Simulated Repository Conditions..................................................................................... 49 Passive corrosion of steel in concrete in the context of nuclear waste disposal ............... 50 Electrochemical treatment to condition contaminated electric arc furnace dusts (EAFD) as an addition to the immobilization mortar in low and medium activity repositories ............. 52 Effect of dissolved oxygen on corrosion properties of reinforcement steel ....................... 53 Corrosion Inhibition of Steel in Ageing Nuclear Structures............................................... 55
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Natural analogues and archaeological artefacts for long term prediction in nuclear waste disposal ........................................................................................................................... 57 Studying the mechanisms of long-term atmospheric corrosion of steel: new strategies to understand the role of non-stable phases constitutive of the rust layer ............................ 59 Fluctuation of redox conditions in a waste disposal cell: influence on the corrosion behavior of carbon steel components .............................................................................. 61 Applicability of iron and copper artifacts excavated at a ruin as a natural analog study for long-term corrosion ......................................................................................................... 62 Micro-mechanical characterization of properties of corrosion layer .................................. 64 Computer simulation of local attacks on carbon steel ...................................................... 68 Long-Term Integrity of Overpack Closure Weld for HLW Geological Disposal, (I) Prediction and Evaluation Method for Structural Integrity of the Weld Joint ..................... 70 Pitting corrosion of stainless steel: measuring and modelling pit propagation in support of damage prediction on radioactive waste containers ........................................................ 72 Development of a spinel-structure point defect model for the magnetite film on iron........ 75 Using experimental data from atmospheric corrosion experiments to parameterise models of localised corrosion and stress corrosion cracking ........................................................ 77 Modelling the Long-term Corrosion Behaviour of Copper Canisters due to Sulphide ....... 79 Corrosion of copper in the safety assessment SR-Site .................................................... 80 Modelling the metal interface evolutions: applications to the localised corrosion ............. 82 Effect of temperature on the corrosion parameters and concrete strains in a pilot container in El Cabril repository ...................................................................................................... 84 Long term corrosion of iron in concrete and in atmospheric conditions: From the experience to the modelling............................................................................................. 85
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Introduction Dear participants to the 4th international workshop on long-term prediction of corrosion damage in nuclear waste systems. Welcome to Bruges. The intensive industrialisation worldwide requires an everincreasing production of energy. Currently, almost one fifth of the world‟s total electricity production arises from nuclear power plants. Besides producing electricity for the benefit of society, nuclear power is also leaving a legacy of radioactive residues for which a safe long-term solution has to be found to protect the many generations to come. Among the options considered for dealing with high-level nuclear waste, geological disposal is the one recommended at the international level. Predicting the degradation rate of the metallic barrier in a robust and reliable manner represents one of the greatest scientific and technical challenges currently known to mankind, because its lifetime largely exceeds that of any industrial application. This issue was already explored during three previous sucessful workshops (Cadarache, France, 2001; Nice, France, 2004 and Pennsylvania State University, USA, 2007). The main objective of the workshop is to bring together scientists and engineers from various countries who are developing high-level nuclear waste disposal technologies, with the goal of promoting scientific and technical exchanges concerning long term behaviour of metallic containment materials and engineered barrier systems. In particular, the workshop will compare the approaches that are being developed worldwide for predicting long-term corrosion phenomena, including corrosion strategies for interim storage and geological disposal. This event is a joint organisation of the Belgian Nuclear Research Centre (SCK•CEN) and the Belgian Agency for Radioactive Waste and Enriched Fissile Materials (NIRAS/ONDRAF). The event is also sponsored by the European Federation of Corrosion (EFC): Nuclear Corrosion Working Party (WP4). Moreover, it takes part of the activities of the RILEM Technical Committee “Concrete in the context of the nuclear management” (TC 226-CNM). The workshop addresses the main items related to the long-term corrosion behaviour, with a major emphasis on: Overview of the different (national) programmes o Simularities, common challenges, different approaches, … o Legal issues Experimentation o Laboratory o In situ Modelling o Fundamentals of prediction o Long-term prediction (e.g. localized corrosion after induction time, …) Archaeological artefacts o How can this information help in the long-term prediction? Regulatory issues Corrosion behaviour of steel reinforcements in concrete We wish you an interesting and successful workshop.
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Programme Monday June 28, 2010 17:30 18:30
Registration Welcome reception
Tuesday June 29, 2010 08:00 08:30
Registration and coffee Opening words Philippe Lalieux (Long Term Management Director, ONDRAF/NIRAS, Belgium) Eric van Walle (General Manager, SCK•CEN, Belgium) Damien Féron (Chairman EFC Working Party 4 – Nuclear Corrosion, CEASaclay, France)
SESSION I - Overview of the national programmes Chairman : Ph. Lalieux (ONDRAF/NIRAS, Belgium) Co-chairman : G. Volckaert (SCKCEN, Belgium) 09:00
Risk-informed performance-based application of container/canister corrosion in nuclear waste management (LongTermCorr2010-01) T. Ahn, C. Markley, H. Jung, K. Compton
09:30
Review of the performance of selected metals as canister materials for UK spent fuel and/or HLW (LongTermCorr2010-03) F. King and C. Padovani
10:00
The Ondraf/Niras safety strategy and the strategic choices in the Belgian Supercontainer design (LongTermCorr2010-04) M. Van Geet and R. Gens
10:30
Coffee break
SESSION IIa - Experimentation Chairman : D. Féron (CEA, France) Co-chairman : F. Druyts (SCKCEN, Belgium) 11:00
Review of the corrosion studies of the metallic barrier in geological disposal conditions with respect to the Belgian Supercontainer design (LongTermCorr2010-05) B. Kursten, F. Druyts, D.D. Macdonald, N.R. Smart and R. Gens
11:30
Some important issues in the electrochemistry of carbon steel in simulated concrete pore water (LongTermCorr2010-06) D.D. Macdonald, A. Saleh and O. Rosas-Camacho
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12:00
Prediction of long term corrosion damage to the carbon steel overpack in Belgium's Boom Clay repository (LongTermCorr2010-40) D.D. Macdonald, A. Saleh and O. Rosas-Camacho
12:30
Lunch
14:00
Further studies on the effect of irradiation on the corrosion of carbon steel in alkaline media (LongTermCorr2010-07) N.R. Smart, A.P. Rance, R.J. Winsley, P.A.H. Fennell, B. Reddy, D. Scott, B. Kursten
14:30
Propagation behavior of general and localized corrosion of carbon steel in simulated groundwater environment under aerobic condition (LongTermCorr2010-08) N. Taniguchi, H. Suzuki, M. Kawasaki, M. Naito, M. Kobayashi, R. Takahashi, H. Asano
15:00
An experimental approach to study the effect of chloride deposition on the stress corrosion behaviour of 316L stainless steel used for intermediate level radioactive waste containers (LongTermCorr2010-10) O.E. Albores-Silva, E.A. Charles, C. Padovani
15:30
Application of X-ray tomography for characterisation of localised corrosion and stress corrosion cracking in waste container materials (LongTermCorr2010-09) D.L. Engelberg, N.P.C. Stevens, S.B. Lyon, A.B. Cook, P.J. Withers, T.J. Marrow
16:00
Coffee break
SESSION IIb - Experimentation Chairman : B. Kursten (SCKCEN, Belgium) Co-chairman : H. Asano (RWMC, Japan) 16:30
Field corrosion experiments in clay in anoxic conditions at high temperature (LongTermCorr2010-11) S. Dewonck, C. Bataillon, F. Foct, M.L. Schlegel, Y. Linard, D. Crusset, B. Scwyn, N. Nakayama and G. Kwong
17:00
Crevice corrosion testing methods for measuring the repassivation potential of alloy 22 (LongTermCorr2010-12) C.M. Giordano, M.A. Rodriguez, R.M. Carranza and R. Rebak
17:30
Sulphide induced stress corrosion cracking of CuOFP in groundwater (LongTermCorr2010-13) E. Arilahti, T. Lehtikuusi, T. Saario, P. Varis
18:00
The rate controlling reactions for copper corrosion in anaerobic aqueous sulphide solutions (LongTermCorr2010-14) J. Chen, Z. Qin, D.W. Shoesmith
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18:30
Closure of day 1
Wednesday June 30, 2010 SESSION IIc - Experimentation Chairman : D.D. Macdonald (Pennsylvania state University, USA) Co-chairman : T. Ahn (NRC, USA) 08:30
Further studies of in situ corrosion tesing of miniature copper-cast iron nuclear waste canisters (LongTermCorr2010-15) N.R. Smart, A.P. Rance, B. Reddy, S. Eriksson, K. Pedersen, C. Lilja
09:00
The corrosion rate of pure copper in an oxic bentonite/saline groundwater environment (LongTermCorr2010-16) B. Rosborg, A. Kranjc, V. Kuhar, A. Legat
09:30
The scientific basis for the corrosion of copper in water and the implications for canister lifetimes (LongTermCorr2010-17) F. King, C. Lilja
10:00
Long-term integrity against corrosion of titanium used for TRU waste container (LongTermCorr2010-18) G. Nakayama, Y. Sakakibara, S. Kawakami
10:30
Coffee break
11:00
Effect of material strength on the hydrogen-absorption characteristic of carbon steel weldments (LongTermCorr2010-19) H. Inoue, Y. Itoda
11:30
Long-term integrity of overpack closure weld for HLW geological disposal – II. Corrosion property under anaerobic condition (LongTermCorr2010-20) M. Kobayashi, Y. Yokoyama, R. Takahashi, H. Asano, N. Taniguchi and M. Naito
12:00
Corrosion issues related to disposal of 316SS-zirconium metal waste forms under simulated repository conditions (LongTermCorr2010-21) L.R. Bairi, S. Ningshen, U.K. Mudali and B. Raj
12:30
Lunch
14:00
Social event
18:30
Workshop diner
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Thursday July 1, 2010 Session III – Corrosion behaviour of steel reinforcements in concrete Chairman : R. Gens (ONDRAF/NIRAS, Belgium) Co-chairman : T. Saario (VTT, Finland) 09:00
Passive corrosion of steel in concrete in the context of nuclear waste disposal (LongTermCorr2010-22) V. L'Hostis, E. Amblard
09:30
Apparent activation energy of electrical resistivity in concrete (LongTermCorr2010-24) C. Andrade, P. Zuloaga, I. Martinez, A. Castillo, S. Briz
10:00
Electrochemical treatment to condition contaminated electric arc furnace dusts (EAFD) as an addition to the immobilization mortar in low and medium activity repositories (LongTermCorr2010-26) M. Castellote, C. Andrade, P. Zuloaga, M. Navarro, M. Ordóñez
10:30
Coffee break
11:00
Effect of dissolved oxygen on corrosion properties of reinforcement steel (LongTermCorr2010-27) H. Jung, K.-J. Kwon and E. Lee
11:30
Corrosion inhibition of steel in ageing nuclear structures (LongTermCorr2010-28) M. Balonis and F.P. Glasser
12:00
Effect of temperature on the corrosion parameters and concrete strains in a pilot container in El Cabril repository (LongTermCorr2010-44) C. Andrade, P. Zuloaga, I. Martinez, A. Castillo and S. Briz
12:30
Long term corrosion of iron in concrete and in atmospheric conditions: From the experience to the modelling (LongTermCorr2010-45) E. Burger, A. Millard, S. Perrin, V. L'Hostis, D. Neff, P. Dillmann
13:00
Lunch
Session IV – Archaeological artefacts Chairman : V. L'Hostis (CEA, France) Co-chairman : C. Padovani (NDA, UK) 14:30
Natural analogues and archaeological artefacts for long term prediction in nuclear waste disposal (LongTermCorr2010-29) D. Féron
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15:00
Studying the mechanisms of long-term atmospheric corrosion of steel: new strategies to understand the role of non-stable phases constitutive of the rust layer (LongTermCorr2010-30) E. Burger, H. Faiz, D. Neff, E. Rocca, F. Mirambet, P. Dillmann
15:30
Fluctuation of redox conditions in a waste disposal cell: influence on the corrosion behavior of carbon steel components (LongTermCorr2010-31) M. Saheb, D. Neff, P. Dillmann, F. Marsal, D. Pellegrini
16:00
Coffee break
16:30
Applicability of iron and copper artefacts excavated at a ruin as a natural analog study for long-term corrosion (LongTermCorr2010-32) H. Yoshikawa
17:00
Micro-mechanical characterization of properties of corrosion layer (LongTermCorr2010-33) A. Dehoux, F. Bouchelaghem, Y. Berthaud
17:30
Long term alteration of archaeological slags: an analogue for nuclear waste glasses (LongTermCorr2010-34) A. Michelin, D. Neff, S. Gin, P. Dillmann
Friday July 2, 2010 Session Va – Modelling Chairman : F. King (Integrity Corrosion Consulting Ltd, Canada) Co-chairman : B. Rosborg (Royal Institute of Technology, Sweden) 08:30
Computer simulation of local attacks on carbon steel (LongTermCorr2010-35) S. Tricoit, B. Vuillemin, R. Oltra, D. Crusset
09:00
Pitting corrosion of stainless steel: measuring and modelling pit propagation in support of damage prediction on radioactive waste containers (LongTermCorr2010-37) M. Gahari, A.J. Davenport, T. Rayment, N. Laycock, D. Krouse, C. Padovani, T. Suter, R. Mokso, M. Stampanoni
09:30
Development of a spinel-structure point defect model for the magnetite film on iron (LongTermCorr2010-38) M. Vankeerberghen
10:00
Long-term integrity of overpack closure weld for HLW gelogical disposal – I. Prediction and evaluation method for structural integrity of the weld joint (LongTermCorr2010-36) H. Asano, A. Nakamura, M. Kobayashi
10:30
Coffee break
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Session Vb – Modelling Chairman : C. Andrade (Institute of Construction Science, IETcc, CSIC, Spain) Co-chairman : N.R. Smart (Serco Assurance, UK)
11:00
Using experimental data from atmospheric corrosion experiments to parameterise models of localised corrosion and stress corrosion cracking (LongTermCorr2010-39) A.B. Cook, N.P.C. Stevens, S.B. Lyon
11:30
Modelling the long-term corrosion behaviour of copper canisters due to sulphide (LongTermCorr2010-41) F. King, M. Kolar, M. Vähänen, C. Lilja
12:00
Corrosion of copper in the safety assessment SR-Site (LongTermCorr2010-42) C. Lilja
12:30
Modelling the metal interface evolutions: applications to the localised corrosion (LongTermCorr2010-43) D. Di Caprio, C. Vautrin-Ul,J. Stafiej, J. Saunier, A. Chaussé, D. Féron and J.P. Badiali
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LongTermCor2010-01
Risk-Informed Performance-Based Application of Container/Canister Corrosion in Nuclear Waste Management Tae Ahn1, C. Markley1, H. Jung2 and K. Compton1 1 U.S. Nuclear Regulatory Commission, MS: E2B2, 20555-0001 Washington DC, U.S.A. 2 Center for Nuclear Waste Regulatory Analyses, San Antonio, Texas 78238, U.S.A. +1 301-492-3150 +1 301-492-3357
[email protected] Abstract Engineered barrier systems play an important role in any geologic repository designed for permanent disposal of nuclear waste, whether in unsaturated or saturated environments. The engineered barrier system may consist of a container, canister, waste form (such as spent nuclear fuel or high-level waste reprocessed into glass), and backfill. This paper discusses the corrosion behavior of the container or canister, with respect to the radionuclide release characteristics and consequent dose to the public. The release of radionuclides from the engineered barrier system is often rate-controlled by the container and canister failure rate, dissolution (or other degradation) rate of the waste form, or diffusion (or other transport) rate of the radionuclides through the failed container/canister/backfill design. A risk assessment identifies the topics that dominate dose to the public. Those topics may differ for unsaturated to saturated environments. Examples of such topics in the container (or canister) corrosion include long-term behavior of a passive film, general corrosion rate, initiation time and extent of localized corrosion and stress corrosion cracking in passive and non-passive metals. This presentation describes how the information can be used, in conjunction with appropriate uncertainty analyses, in a safety assessment. Disclaimer The NRC staff views expressed herein are preliminary and do not constitute a final judgment or determination of the matters addressed. This paper presents input by the Center for Nuclear Waste Regulatory Analyses (CNWRA) for NRC under Contract No. NRC-02-07-006. The activities reported here were performed on behalf of the NRC Office of Nuclear Material Safety and Safeguards, Division of High-Level Waste Repository Safety.
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LongTermCor2010-03
Review of the Performance of Selected Metals as Canister Materials for UK Spent Fuel and/or HLW Fraser King1 and Cristiano Padovani2 1 Quintessa Ltd., UK/Integrity Corrosion Consulting Ltd, 3396 Stephenson Point Road, V9T 1K2 Nanaimo, Canada 2 Nuclear Decommissioning Authority, Radioactive Waste Management Division, UK +1 250-571-1125 +1 250-751-1884
[email protected] Abstract A review of the performance of selected canister materials for the disposal of high-level waste (HLW) and spent fuel (SF) in the UK is presented. The canister materials considered are; copper, carbon and stainless steels, titanium alloys, and nickel alloys. A range of generic environmental conditions has been defined for the study. These generic conditions encompass the three UK generic host-rock environments, possible ground water compositions, the maximum canister surface temperature, the evolution of the pore-water composition in contact with the canister, the nature of the backfill material (if any), the external gamma radiation field, the evolution of redox conditions within the geologic disposal facility (GDF), the saturation time, the extent of microbial activity, the mineral content of the host rock, mass transport conditions to and from the canister, the extent of residual stress and external loading, the types and amount of the various wastes, the hydraulic conductivity of the host rock, the depth of the GDF, and the duration of the operational phase. The purpose of the review is to provide a high-level overview of the technical and scientific issues relating to the use of each of these materials for the disposal of HLW/SF in the UK. The advantages and disadvantages of each material are described, as are limiting or “critical” conditions for which the use of a given material is questionable. The candidate canister materials offer different advantages and disadvantages. In terms of the service life of the canister, copper and the titanium- and nickel-based passive materials would be expected to provide longer containment than carbon steel. Copper is possibly the most challenging material from which to fabricate a canister, as not only is there a requirement for an internal structural support but it is also necessary to seal and inspect thick wall sections. In terms of the overall corrosion performance, copper and crevice-corrosion-resistant titanium and nickel alloys perform better than carbon or stainless steels. Stainless steels suffer from localised corrosion in the presence of chloride and thiosulphate ions, the latter a product of the oxidation of pyrite impurities present in many geological settings. Carbon steel is also more likely to have an adverse impact on other barriers in the system. In terms of the flexibility in the design of the geological disposal facility, however, carbon steel is perhaps the most flexible of the materials considered since it is suitable for use with either bentonite or cementitious backfill or with no backfill at all. In terms of the robustness and confidence in the long-term prediction of the canister performance, both copper and carbon steel have advantages over stainless steel, titanium and nickel alloys. Finally, there is greater international experience with the two corrosion allowance materials than with either titanium or nickel. Possible canister materials have been identified for each of five generic disposal scenarios that describe many of the environmental conditions possible in the UK. Copper, carbon steel, and crevice-corrosion-resistant titanium alloys are suitable canister materials
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for a bentonite-backfilled GDF in sedimentary mudrock with high salinity groundwater. In a segregated HLW and SF GDF design in mudrock with low-salinity ground water, titanium alloys and carbon steel have been identified as suitable canister materials. In a bentonitebackfilled GDF in strong rock, either copper or titanium are suitable. In the presence of cementitious backfill, carbon steel may be appropriate, although there would be a concern over localised corrosion due to the relatively high chloride content of UK ground waters. Finally, for a GDF in an evaporite host rock, carbon steel, titanium, or a nickel-based alloy would be suitable.
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LongTermCor2010-04
The ONDRAF/NIRAS safety strategy and the strategic choices in the Belgian Supercontainer design Maarten Van Geet and Robert Gens NIRAS/ONDRAF, Kunstlaan 14, 1210 Brussels, Belgium +32 2 212 10 85 +32 2 218 51 65
[email protected]
Abstract In Belgium, geological disposal of high-level and long-lived radioactive waste in plastic clay is the reference solution for research & development (R&D) in the frame of the longterm management. Boom Clay is currently considered the reference formation for hosting a repository for this type of waste, while Ypresian clays are considered as alternative. The main safety functions that this geological disposal system should fulfil are:
Isolation (I): provided by the Boom Clay and its geological cover to isolate the waste from the biosphere into the far future. Furthermore, the repository is located far enough from underground areas of mineral resources to reduce the likelihood of inadvertent human intrusion, and the self-sealing capacity of the Boom Clay contributes to reducing the possible consequences of such intrusions; Retardation (R): Because of the favourable properties of the Boom Clay and its stability, the host formation, supported by the engineered barriers and the waste forms performs the function of delay and attenuation of the releases over extremely long periods of time; Engineered Containment (C): provided by the engineered barrier system during at least the thermal phase, which last several hundreds to several thousands of years.
According to its safety strategy, ONDRAF/NIRAS has re-evaluated its design since the last formal safety assessment of SAFIR 2 in 2001. Based on the outcome of the latter assessment, a multicriteria analysis and detailed interactions between phenomenologists, technologists and safety assessors the supercontainer design with a high pH buffer made of concrete based on Ordinary Portland Cement has been chosen as the reference design. It is believed that in this way the safety concept has been reinforced as this design should provide:
permanent shielding during the operational phase and facilitated quality control adequately understood engineered containment (corrosion resistance) during the thermal phase.
Moreover this design
is based on proven technologies and widely available, affordable materials has negligible negative impact on the safety functions provided by the most important barrier, the clayey host rock may provide complementary sorption with respect to the clay host rock for radionuclides that are mobile in clay.
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On top of that, it should be kept in mind that in a case of plastic host formation, such as Boom Clay, concrete is difficult to avoid for practicality reasons (e.g. as gallery liner). However, we acknowledge that the use of concrete in this context is still fairly new and although many aspects look very promising, they still need to be scrutinised and results need to be confirmed.
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LongTermCor2010-05
Review of the corrosion studies of the metallic barrier in geological disposal conditions with respect to the Belgian Supercontainer design Bruno Kursten1, Frank Druyts1, Digby D. Macdonald2, Nicholas R. Smart3 and Robert Gens4 1 SCK•CEN, Boeretang 200, 2400 Mol, Belgium 2 Pennsylvania State University, Materials Science and Engineering, PA, United States 3 Serco Assurance, United Kingdom 4 NIRAS/ONDRAF, Kunstlaan 14, 1210 Brussels, Belgium +32 14 33 31 36 +32 14 32 35 53
[email protected]
Abstract The Supercontainer (SC) Design is the preferred Belgian option for the final disposal of vitrified high-level waste (VHLW) and spent fuel (SF) in deep undergound clay layers. The SC comprises a carbon steel overpack, containing two VHLW canisters or four SF assemblies, surrounded by a Portland cement (PC)-based buffer, which in turn, is entirely encased in a stainless steel envelope. The SC Design was developed based on the Contained Environment Concept (CEC), which intention is to fix and preserve a favourable chemical environment in the immediate vicinity of the overpack, so that it will be exposed to essentially unchanged conditions for a long time, at least for the duration of the thermal phase. Under the predicted conditions within the SC (highly alkaline concrete buffer), the carbon steel overpack is expected to remain in a permanent state of passivity, which is believed to result in very low uniform corrosion rates. The backbone of the R&D strategy, which aims at demonstrating and defending that the integrity of the carbon steel overpack can be ensured at least during the thermal phase, is based on proving that the carbon steel overpack will only be prone to uniform corrosion. Therefore, we will have to present wellreasoned arguments to proof that the carbon steel overpack will not be susceptible to localised corrosion phenomena (such as e.g. pitting corrosion, crevice corrosion and stress corrosion cracking) under the high pH conditions prevailing within the Supercontainer (the exclusion principle). These arguments will be compiled from the specialist literature and from dedicated laboratory experiments. An overview of the status of the R&D program related to the corrosion of the overpack is given: • scoping calculations based on a local equilibrium-diffusion transport model indicated that the near-field will likely remain alkaline (pH>12.5) for a time scale of 100,000 years; • radiolysis simulations suggested that the oxygen concentration will remain fairly constant (3.5×10-4 mol/L) over a 300 year period, which could have a significant influence on the duration of the aerobic phase, and hence on the localised corrosion processes; • transport modelling calculations of aggressive species (chloride, sulphide, thiosulphate) towards the overpack have revealed very low concentrations of these species reaching the overpack (e.g. a maximum of 3.1 mM of S2-/HS- will reach the overpack surface after 5,000 years considering very conservative assumptions);
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• the low anaerobic uniform corrosion rates, found in the literature, of carbon steel exposed to cementitious environments have been confirmed for the carbon steel/concrete buffer system representative for the Supercontainer; • gas evolution experiments, performed by Smart and co-workers (Serco Assurance, UK), indicated that the anaerobic uniform corrosion rates tend to converge towards the same very low value in the long term, irregardless of the temperature. The corrosion rates obtained from weight loss measurements are consistent with those derived from gas generation measurements; • irradiation (25 Gy/h) seemed to have a negligible effect on the measured hydrogen gas generation rates compared to the unirradiated situation; • slow strain rate tests (SSRT) showed no susceptibility of carbon steel to stress corrosion cracking (SCC) on the parent material; • the parent carbon steel was found not to suffer from stress corrosion cracking (SCC), under the conditions tested so far; The main remaining uncertainties and future key issues are: • investigate the impact of welding on the corrosion properties of the parent metal; • evaluate the role of sulphur on the corrosion behaviour of the carbon steel overpack; • evaluate the evolution of the properties of the oxide film formed on the carbon steel overpack; • evaluate the impact of pouring a cement-based filler material on a hot steel surface on the corrosion behaviour of the carbon steel overpack;
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LongTermCor2010-06
Some important issues in the electrochemistry of carbon steel in simulated concrete pore water Digby D. Macdonald, Amr Saleh, and Omar Rosas-Camacho Center for Electrochemical Science and Technology, Department of Materials Science and Engineering, Pennsylvania State University, University Park, PA 16801, USA +1 814-360-3859 +1 814-863-4718
[email protected]
Abstract The planned disposal of high level nuclear waste in Belgium will make use of the Super Container concept. In this concept, the vitrified waste is contained within a carbon steel container, which itself is contained within a stainless steel outer shell. The annular gap between the carbon steel container and the outer shell is filled with a cementitious material having a composition close to that of Portland cement concrete. The pore water in contact with the carbon steel container is assumed to contain sulphide ion (S2-), thiosulfate ion (S2O32-), and chloride ion (Cl-) and to have a pH of 13 to 13.5. After an initial oxic period, during which significant oxygen remains in the annulus, the environment becomes anoxic with hydrogen evolution being the principal cathodic reaction. The rate of hydrogen production is determined by the magnitude of the passive current density for carbon steel in contact with the pore water. Using data published in the literature for ipass for carbon steel in alkaline environments led to the prediction that the annulus would become pressurized with hydrogen to an unacceptably high pressure that might risk damage to the stainless steel outer shell. However, it was determined that the literature value for ipass was most likely determined potentiodynamically and not under the steadystate conditions that are deemed to be most appropriate for the super container. Under steady-state conditions, work in the authors‟ laboratory for iron in borate buffer solution indicated that the passive current density under steady-state conditions should be about one hundred times smaller than the potentiodynamic value (10-8 versus 10-6 A/cm2, respectively). If so, modelling of the annulus pressurization indicates that pressurization may not be an issue over the 300 year storage horizon. This paper describes a research program carried out in the authors‟ laboratory, under sponsorship of ONDRAF-NIRAS, to explore the passivity of carbon steel in contact with simulated concrete pore water and to determine the passive current density. The work has made use of potentiostatic polarization, potentiodynamic polarization, steady state polarization, and electrochemical impedance spectroscopy (EIS). This talk will concentrate on the impedance behaviour of carbon steel in concrete pore water that contains chloride ion, sulphide ion, thiosulfate ion, and mixtures of these species at temperatures from 22 oC to 80 oC. The passive film on carbon steel in simulated concrete pore water is found to be an n-type, defect semiconductor with the principal defect being the ferrous metal interstitial. The passive state is well described by the Point Defect Model, which provides a basis for the extrapolation of the data over times that are relevant to HLNW storage. The authors gratefully acknowledge the support of this work by ONDRAF-NIRAS, Brussels, Belgium.
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LongTermCor2010-07
Further studies on the effect of irradiation on the corrosion of carbon steel in alkaline media Nicholas R. Smart1, Andrew P. Rance1, Robert. J. Winsley1, Paul A.H. Fennell1, Bharti. Reddy1, Doug Scott1 and Bruno Kursten2 1 Serco Technical Services, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB, U.K. 2 SCK•CEN, Boeretang 200, B-2400 Mol, Belgium +44 1635-280385 +44 1635-280389
[email protected]
Abstract The reference design for the Belgium Supercontainer for the disposal of high-level nuclear waste and spent fuel consists of a 30 mm thick carbon steel cylindrical vessel („the overpack‟) surrounding 309 grade stainless steel waste canisters. The overpack will be surrounded by a cementitious buffer material, which will be encased in a stainless steel vessel („the envelope'). This paper is concerned with the anaerobic corrosion processes that may affect the carbon steel overpack within the Supercontainer design. The carbon steel will be exposed to alkaline porewater in the cementitious buffer or in the alkaline material which may be used as a filler in an annulus between the cementitious buffer and the carbon steel overpack. After a relatively short period the interface between the steel and alkaline porewater will become anoxic as oxygen is consumed by aerobic corrosion processes and microbial activity within the tunnel. Low concentrations of chloride from the groundwater may penetrate the buffer and reach the surface of the carbon steel overpack. For the purposes of assessing the performance of the Supercontainer design an experimental programme is in progress to provide information about (i) the possible effect of gamma radiation on the anaerobic corrosion rate of carbon steel in cementitious porewaters and on the nature of the corrosion products formed, and (ii) the effect of radiation on the electrochemical and corrosion behaviour of carbon steel corroding anaerobically in simulated concrete porewater conditions. In terms of performance assessment the most important criteria being measured is the rate of generation of hydrogen through anaerobic corrosion. The anaerobic corrosion rates of carbon steel were measured by monitoring hydrogen evolution rates using a manometric gas cell technique. In addition, a few experiments were carried out using autoclaves equipped with pressure gauges and hydrogen sensors. The electrochemical behaviour of anaerobically corroding carbon steel was investigated by measuring open circuit potential, linear polarisation resistance and AC impedance. Most experiments were carried out in a highly alkaline artificial porewater that simulated the cementitious buffer material selected for use in the Supercontainer concept, but a few experiments were also conducted on steel embedded in solid buffer material. The programme took account of γ-irradiation dose rate (0 Grays hr-1 and 25 Grays hr-1), temperature (25 ºC and 80 ºC) and chloride concentration (0 and 100 mg/l). The paper will summarise the results obtained to date, for both gas generation measurements and electrochemical measurements. The gas generation experiments have shown a slow decline in anaerobic corrosion rate, in accordance with previous measurements in other alkaline systems. The results of analyses carried out on
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specimens removed from the experiments will also be described. These have confirmed the formation of magnetite on the surface of the carbon steel during the anaerobic corrosion process. The corrosion rates obtained from weight loss measurements were consistent with those derived from gas generation experiments. It was found that radiation at a dose rate of 25 Grays hr-1 had a negligible effect on gas generation rates compared to the unirradiated situation. The results of the electrochemical measurements will be presented. They are consistent with the occurrence of anaerobic corrosion on the surface of the wires and have shown only a small effect of radiation on the corrosion potential, in contrast to previously reported results. The instantaneous corrosion rates measured electrochemically are consistent with the rates measured using alternative methods. The structure and properties of the corrosion product film will be discussed and the implication of the results for the Supercontainer concept will be outlined. Keywords: carbon steel, anaerobic, corrosion, irradiation, alkali, cement
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LongTermCor2010-08
Propagation behavior of general and localized corrosion of carbon steel in simulated groundwater environment under aerobic condition Naoki Taniguchi1, Hiroyuki Suzuki2, Manabu Kawasaki1, Morimasa Naito, Masato Kobayashi3, Rieko Takahashi3, and Hidekazu Asano3 1 Japan Atomic Energy Agency (JAEA), 4-33 Muramatsu, Tokai-mura, Ibaraki, Japan 2 NESI, 4-33 Muramatsu, Tokai-mura, Ibaraki, Japan 3 Radioactive Waste Management Funding and Research Center (RWMC) 1-15-7, Tsukishima, Chuo-ku, Tokyo, Japan +81-29-282-1111
[email protected]
Abstract Carbon steel has been selected as one of the candidate materials for overpack for geological disposal of high-level radioactive waste (HLW) in Japan. Corrosion of carbon steel is divided into two types; general corrosion and localized corrosion. The latter such as pitting and crevice corrosion is attributed to a local breakdown of passive film in the presence of aggressive ion represented by a chloride ion, while the former generally occurs on carbon steel in many natural water environments. Even in a repository environment for HLW disposal, general corrosion will be the most probable corrosion type, because bentonite clay buffer which surrounds an overpack prevents formation of stable passive film on surface of carbon steel. However, there are some factors that induce the passivation and consequent localized corrosion on carbon steel overpack. One of typical causes of the passivation is the rise of pH in groundwater due to contact with cementitious material for concrete structures such as a tunnel support. Assuming this situation, both general and localized corrosion should be taken into consideration in assessment of lifetime and designing of corrosion allowance for a carbon steel overpack. In this study, propagation behaviors of general and localized corrosions (pitting corrosion and crevice corrosion) were investigated by immersion tests of carbon steel under aerobic condition. The test solutions for general corrosion were saline water such as synthetic seawater and groundwater taken from JAEA‟s Horonobe underground research laboratory, which is now in course of construction. By contrast, the test solutions for localized corrosion were aqueous carbonate/chloride solution and alkalized saline water set up by mixing with powdered concrete. The pH range of the test solutions was 6.1-8.4 for general corrosion and 8.4-13.4 for localized corrosion. In addition to the tests in simple solution, immersion tests in the presence of bentonite-sand mixture were carried out. The test coupons were made from rolled steel (SM400B) and forging steel (SFVC1 and SF340A). Some of the immersion tests were performed using welded coupons which were cut from the samples of a full-scale model of overpack lid welded by different types of method; TIG, MAG, and EBW. The results of the immersion tests showed that the growth rate of corrosion was strongly dependent on the environmental condition, but the upper limit of pitting factor (the ratio of the maximum corrosion depth and the average corrosion depth) was approximately determined by only average corrosion depth as shown in Fig.1. Although preferential corrosion was observed in the weld metals of TIG and MAG, the pitting factors had not exceeded the upper limit of the other test data or literature data. In the weld metal of EBW, no preferential corrosion was observed. Based on these experimental data and literature data, an empirical model that predicts the maximum corrosion depth of overpack
26
from average corrosion depth was developed by applying the extreme value statistical analysis using the Gumbell distribution function. The scale parameter, α and the location parameter, λ, were obtained from the analysis, and then the upper limit of the α(mm) and λ(mm) for both general and localized corrosions were approximated as a function of average corrosion depth, Xm, as follows; General Corrosion: α= 0.45 Xm0.25, λ= Xm+ 1.5 Xm0.25 ------(1) Localized Corrosion: α= 0.54 Xm0.5, λ= Xm+ 1.8 Xm0.5------(2) Herein appointing a confidence ratio, a, the maximum corrosion depth, Pmax, at a is given by, Pmax= λ+ αln T–αln(-ln a) ------(3) where T is the return period given by the ratio of overpack surface area, S, and sample surface area, s (131cm2 for general corrosion and 105cm2 for localized corrosion), as T=S/s. When the S, is 55129cm2 based on the present overpack design specification Natural Water (Soli, Seawater, Lake) General Corrosion 4 and the a is 0.99, the Pmax is given by, Localized Corrosion 10
The Xm can be estimated from the amount of available oxygen trapped in the buffer and backfilling materials. It is possible to make a conservative assessment of the maximum corrosion depth during an oxidizing condition period by adopting larger calculation results of either Eq.(4) or Eq.(5) even if the environmental condition and the corrosion type (general corrosion or localized corrosion) are not specified.
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Pitting Factor
General corrosion: Pmax(mm) = Xm+7.5 Xm0.5------(4) Localized corrosion: Pmax(mm) = Xm+6.4 Xm0.25------(5)
Upper Limit
102 101 100 10-1 10-5 10-4 10-3 10-2 10-1
100
101
Average Corrosion Depth (mm)
Figure 1 Relationship between average corrosion depth and pitting factor of carbon steel
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LongTermCor2010-09
Application of X-ray Tomography for Characterisation of Localised Corrosion and Stress Corrosion Cracking in Waste Container Materials Dirk L. Engelberg, N.P.C.Stevens, S.B.Lyon, A.B.Cook, P.J.Withers, T.J.Marrow Materials Performance Centre, School of Materials, University of Manchester, M13 9PL, Manchester, United Kingdom +44 7855-106770
[email protected]
Abstract The integrity of materials used for nuclear waste containers critically depends on the interaction between the local microstructure and the prevailing environmental conditions. Information to help advance understanding of life-limiting factors, such as the influence of pit-geometry on stress corrosion crack nucleation, can be obtained through nondestructive observations, obtained by 3-dimensional X-ray computed tomography (XCT) and 2-dimensional Digital Image Correlation (DIC). These provide insight into the corrosion and cracking processes. The aim of this paper is to demonstrate the wealth of information that can be obtained through the use of these advanced characterisation techniques, for developing and refining component life-time prediction models. The paper describes the application of X-ray computed tomography to obtain 3dimensional information on corrosion and cracking processes, in austenitic stainless steels relevant to waste containers. The influence of pit geometries, and their effect as stress raisers under mechanical loading for the nucleation of stress corrosion cracking is addressed. Internal pit geometries were characterised at various length scales. The factors that influence the application of tomography for the characterisation and quantification of stress corrosion cracking are addressed, including examples of tomographic characterisations of fracture surfaces of atmospheric induced stress corrosion cracking (AISCC) in stainless steels.
Figure 1: Semi-transparent reconstruction of a stainless steel wire from tomography data. An isosurface of the wire is shown, highlighting the presence of corrosion pits.
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LongTermCor2010-10
An experimental approach to study the effect of chloride deposition on the stress corrosion behaviour of 316L stainless steel used for intermediate level radioactive waste containers O.E. Albores-Silva1, E.A. Charles1, C. Padovani2 1 Newcastle University, School of Chemical Engineering & Advanced Materials, Herschel Building, Newcastle upon Tyne, NE1 7RU, United Kingdom 2 Nuclear Decommissioning Authority, Radioactive Waste Management Directorate, United Kingdom. +44-191-222 3635
[email protected]
Abstract Austenitic stainless steel (SS) type 316L is being used to manufacture waste containers for intermediate level waste (ILW) in the UK. The waste containers are being kept in aboveground storage facilities, and will eventually be transferred to a geological disposal facility (GDF). The GDF, however, is not yet available and by the time the facility is operational, it is likely that waste packages will have been in storage facilities for several decades. If retrieval of the waste from the GDF was needed in the future then this would further increase the duration of the „storage‟ period. These considerations require the containers to remain in a structurally-sound and retrievable condition for long periods of time. It is generally understood that non-sensitized austenitic stainless steels are fairly resistant to stress corrosion cracking even in chloride containing environments, as long as the temperature is below about 50°C and high acidity is avoided. However, reported failure cases and further research showed that key combinations of the deposited chloride salt composition and characteristic relative humidity intervals, could cause atmospherically induced stress corrosion cracking (AISCC) to occur in austenitic stainless steel at temperatures near ambient. Little research is available however, regarding the effect of chloride deposition level on the occurrence of AISCC. This is important, as it is necessary to identify suitable storage strategies to minimise the possibility of developing AISCC on waste packages. The present work describes a laboratory approach being used to study the effect of chloride deposition levels in simulated atmospheric conditions. This is relevant to both surface storage and the operational period of the GDF. Tests were conducted using 316L SS U-bend specimens prepared from real ILW containers, using MgCl2 as contaminant salt with chloride deposition ranging from 10 to 46,000 μg/cm2, at 30 and 50°C. Two values of relative humidity were considered: the equilibrium relative humidity of MgCl2 at the temperatures of interest (32.4% and 30.5% respectively) and 60 % RH, assumed as representative of the lower end of the spectrum of relative humidity observed in storage facilities. This range of humidity is thought to be particularly corrosive towards the material as it is likely to be associated with the development of concentrated electrolytes around the deposited salt. Results showing the occurrence of AISCC as a function of MgCl2 deposition are presented, which indicate that a threshold for AISCC occurrence is found between 10 and 100 μg/cm2 for the most severe conditions tested, as shown in Figures 1 and 2.
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Figure 1. Cross-section showing AISCC of a 316L SS U-bend, under a chloride 2 contamination level of 100 µg/cm , at 50°C and 30.5% RH
Figure 2. Cross-section showing pitting of a 316L SS U-bend, under a chloride 2 contamination level of 10 µg/cm , at 50°C and 30.5% RH
Additionally, the work describes an ongoing long-term atmospheric corrosion test, which has been recently set up in the underground chambers of a hydroelectric power station. 316L SS corrosion coupons and U-bend specimens were installed at three different locations within the facility to examine the behaviour of the material in an environment that may resemble underground vaults of a GDF.
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LongTermCor2010-11
Field corrosion experiments in clay in anoxic conditions at high temperature S. Dewonck1, C. Bataillon2, F. Foct3, M.L. Schlegel4, Y. Linard1, D. Crusset5, B. Schwyn6 and N.Nakayama 7 and G. Kwong8 1 Andra CMHM, RD960, F-55290 Bure, France 2 CEA/DEN/DANS/DPC/SCCME/LECA Bât 458, F-91191, Gif/Yvette, France 3 EDF R&D MMC Site des Renardières, F-77818 Moret sur Loing 4 DEN/DANS/DPC/SCP/LRSI, CEA de Saclay, 91191 Gif sur Yvette, France 5 Andra F-92298 Châtenay-Malabry Cedex, France 6 NAGRA, Hardstrasse 73 CH-5430 Wettingen, Switzerland 7 JAEA, 959-31, Jorinji, Izumi, Tokishi, Gifu 509-5102, Japan 8 NWMO, Toronto Ontario, M4T 2S3 Canada +33-3-29 75 67 36 +33 3 29 75 53 76
[email protected]
Abstract In a radioactive waste disposal facility in deep geological formation, steel will be used for various purposes such as anchorages, reinforced concrete, sleeves, overpacks and canisters. The French concept requires that these structures are safe and reliable in anoxic conditions, to 90°C and for a long period of time. A lot of corrosion studies have been performed in lab in anaerobic conditions at high temperature to predict the behavior of steel in contact of bentonite (MX-80) or CallovoOxfordian clay (COx) by CEA and EDF (Andra Dossier 2005). These studies show very low corrosion rates and allow identifying the corrosion products formed. It‟s important now to validate these results in situ, and for a long period of time. It‟s the reason why several field experiments are carrying out by Andra in the Opalinus clay (IC experiment in the Mont-Terri Rock Laboratory) and in the COX clay (MCO experiment in the Bure URL). The IC experiment started in 2008, consists in monitoring the corrosion rate of various steel (Inconel 690, 316L stainless steel, 2 carbon steels one representative of Andra concept and another of Nagra concept) at 90 °C, in anaerobic condition, in contact with the OPA clay formation. This experiment is representative of what happens at the level of the sleeves. A special design of the experimental setup (Figure 1) was developed to allow optimal interactions between rock and steel samples. The corrosion rate monitoring is based on Electrochemical Impedance Spectroscopy (EIS). This method is not disturbing for the corrosion process i.e. the corrosion rate doesn‟t change during the electrochemical measurement. The main drawback of this method is that the corrosion process must be in stationary or quasi stationary state: EIS can only measure corrosion rates which do not change quickly with time. This method is well adapted for long term corrosion monitoring because long term corrosion rate evolves slowly. A similar experiment will be installed middle of 2010 in Bure URL with the following samples: TiPt (counter electrode), 309S, two carbon steels in contact with the COx clay. The MCO experiment started in summer 2009, is designed to characterize the corrosion of various steel specimens (carbon steels E24, SA516 et P235, welded join, crack) in contact with clay pore water or gas phase floating at 90°C by studying i) the change in mass of the samples, ii) the crystalline structure of the oxides and iii) the phase transformations in the suboxide layer. This experiment is representative of what happens at an earlier step of the radioactive waste repository. A set of steel pieces is placed in
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each of the seven sampling rods. Each rod is removed at different time (after 1 month figure 2, 3 months, 1 year, 3 years, 5 years, 10 years, 20 years). The objective of this contribution is to present the both set up and the first results about the corrosion monitoring. References: [1] Andra dossier 2005. Référentiel des matériaux d‟un stockage de déchets à haute activité et à vie longue. Tome 3 : corrosion des matériaux métalliques
steel samples
Borecore section Figure 1: Experimental setup of IC experiment
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Figure 2: Sampling cell with the steel pieces. right: initial step t0 ; left: after 66 days at ambient temperature. Vcorr in pore water = 3176µm/y Vcorr in gas phase = 1µm/y
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LongTermCor2010-12
Crevice corrosion testing methods for measuring the repassivation potential of Alloy 22 C. M. Giordano1, M. A. Rodríguez1, Ricardo M. Carranza1 and Raul B. Rebak2 1 Depto. Materiales - Comisión Nacional de Energía Atómica, Instituto Sabato - Univ. Nac. de San Martín / CNEA, Av. Gral. Paz 1499, San Martín, B1650KNA Buenos Aires, Argentina 2 GE Global Research, 1 Research Circle, CEB2505, Schenectady, NY 12309, USA +54-11-6772-7270 +54-11-6772-7362
[email protected]
Abstract Alloy 22 (UNS N06022) belongs to the corrosion resistant Ni-Cr-Mo family of nickel based alloys. The required composition and mechanical properties of this family are described in ASTM B575. Alloy 22 has alloying elements for protection against localized corrosion initiation in a variety of environments. Chromium is added for protection against oxidizing conditions while molybdenum protects against reducing acidic conditions. The base element (nickel) protects the alloy against caustic conditions. All three elements, Ni, Cr and Mo act synergistically to provide resistance to environmentally assisted cracking in hot concentrated chloride solutions. Cr and Mo also provide resistance to localized corrosion such as pitting and crevice corrosion in chloride-containing solutions. Some of the Ni-Cr-Mo alloys also contain a small amount of tungsten (W), which may act in a similar way as Mo regarding protection against localized corrosion. Ni-Cr-Mo alloys are practically immune to pitting corrosion. This is why Alloy 22 has been selected for the fabrication of the corrosion-resistant outer shell of the high-level nuclear waste container for the proposed Yucca Mountain repository. However, Alloy 22 may be susceptible to crevice corrosion in aerated hot chloride-containing solutions. The susceptibility of Alloy 22 to crevice corrosion depends on environmental and metallurgical variables including chloride concentration, temperature, presence of inhibitors and applied potential. The susceptibility to crevice corrosion is measured by the value of the repassivation potential in each tested condition. The lower is the value of the repassivation potential the more aggressive is the environment. In the last few years several electrochemical methods have been proposed to measure the value of the repassivation potential, including, the Cyclic Potentiodynamic Polarization (CPP) (ASTM G 61), the Tsujikawa – Hisamatsu Electrochemical (THE) (ASTM G 192), the Potentiostatic (PS), the Mixed Potentiodynamic - Potentiostatic (PD – PS – PD), and the Mixed Potentiodynamic - Galvanostatic (PD – GS –PD) methods. Whenever crevice corrosion readily occurs (that is, when the environment is aggressive), the values of repassivation potential for Alloy 22 seem comparable for the same environmental and metallurgical variables. This finding suggests that the crevice repassivation potential is a property of the alloy in each testing condition and does not depend greatly on the manner it is obtained provided that highly aggressive conditions (such as high chloride concentration and temperatures higher than 75°C) are used. Under less aggressive conditions, when crevice corrosion is more difficult to initiate, the value of repassivation potential may differ substantially from method to method. The objective of this work is to compare four different electrochemical methods (CPP, THE, PD – PS – PD, and PD – GS - PD) used to evaluate the susceptibility of Alloy 22 to
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crevice corrosion in chloride solutions (see Table 1). The effect of crevicing forming material, applied torque and surface roughness was also investigated. The PD – GS – PD technique was found to be the most conservative laboratory technique which gave the lowest repassivation potential in a relatively short testing time. Torque values higher than 3 Nm were found to be necessary in order to obtain reliable repassivation potential values using PTFE wrapped ceramic crevice formers. Surface roughness of the metallic samples did not play a significant role with these crevice formers. As a first step, it is important to determine the lowest possible values of crevice repassivation potentials in order to assess the likelihood that crevice corrosion may initiate in Alloy 22 under naturally polarized conditions, for example, by dissolved oxygen into the electrolyte conditions. Once the most conservative value of repassivation potential under each testing condition is available, crevice corrosion propagation studies should be used to estimate the rate of crevice corrosion propagation or to determine if crevice corrosion eventually stops propagating. Empirically based models may be used to predict the probability that crevice corrosion may or may not initiate in a waste container. Finally, if there is a probability that crevice corrosion would initiate, models may also be used to predict if this form of corrosion attack will eventually perforate the container allowing for water ingress that may spread the stored detrimental radionuclides. Tabel 1. Electrochemical techniques used in the present work for studying the crevice corrosion initiation and repassivation of Alloy 22 in chloride solutions at 90°C
Technique
Repassivation Parameter
CPC
Max. E
Max. ian
Length of the test
VSCE
A/cm
CPC ECO
~1
5m
~5 hours
THE
ERTHE
~ 0.1
2 µ to 20 µ
Min. 24 hours
PD-PS-PD
PD-PS-PD ECO
~ 0.3
20 µ to 0.4 m
~24 hours
PD-GS-PD
PD-GS-PD ECO
~ 0.1
2µ
~4 hours
2
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LongTermCor2010-13
Sulphide Induced Stress Corrosion Cracking of CuOFP in Groundwater Esko Arilahti, Taru Lehtikuusi, Timo Saario and Päivi Varis VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT, Finland +358-T50-3377798
[email protected] Abstract Susceptibility of oxygen free phosphorus alloyed copper (CuOFP) to sulphide induced stress corrosion cracking has been investigated at room temperature with fracture mechanical precracked Compact Tension (CT) specimens. The test type was constant load with stress intensity level of approximately KI = 9 MPam0.5. Experiments were performed in an air-tight stainless steel pressure vessel equipped with a wire leadthrough, a Ag/AgCl/0.05 M KCl –reference electrode and a pull rod sealed with a lip seal and additional flexible silicone. The environment was a saline reference groundwater, with 100 mg/l sulphide added as Na2S. All additions and sample extractions were made through a transferable glove-box under 5N nitrogen atmosphere. Corrosion potential of the CuOFP samples and redox-potential (Pt-sample) were monitored to ensure anoxic conditions. On-line crack growth monitoring was performed with both load train (crack mouth) displacement gauge and reversing potential drop –technique. Crack growth started immediately after loading the specimens, as indicated by the continuous increase of the load train displacement during the max three week exposure time. Post test SEMexamination revealed that the crack growth was 80-100% intergranular stress corrosion cracking (IGSCC). The average crack growth rate was of the order of 1·10-5 mm/s. The results are discussed in the frame work of a diffusion model developed for diffusion/advection scenarios related to the Finnish and Swedish disposal concept.
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LongTermCor2010-14
The rate controlling reactions for copper corrosion in anaerobic aqueous sulphide solutions J.Chen, Z.Qin, and D.W.Shoesmith Department of Chemistry, The University of Western Ontario, London, Ontario, N6A 5B7, Canada +1-519-661-2111(#86333) +1-519-661-3022
[email protected] Abstract Since copper is thermodynamically stable in the anaerobic groundwaters anticipated in a Swedish nuclear waste repository, failure by corrosion of copper nuclear waste containers is extremely unlikely. The only available oxidant appears to be sulphide, present in groundwaters due to either mineral dissolution or microbial production from sulphates. Corrosion would then be controlled by either the diffusive transport of sulphide through the compacted clay buffer surrounding the container or by the protective properties of the copper sulphide layer on the container surface. We have been characterizing the properties of sulphide films on copper surfaces under natural corrosion conditions using corrosion potential measurements, electrochemical impedance spectroscopy (EIS), scanning electron microscopy (SEM) with a focussed ion beam attachment (FIB) used to produce cross-sectioned corroded specimens, X-ray Photoelectron Spectroscopy (XPS), and micro X-ray diffraction (μXRD). Experiments are being conducted in anaerobic sodium chloride solutions (0.1 mol/L) containing various concentrations of sulphide for exposure periods up to 4000 hours (~167 days). The results show that the corrosion product is a single layer Cu2S film, and the film growth kinetics vary with sulphide concentration. When the concentration is low (5 x 10-5 mol/L) the film appears cellular and its thickness increases linearly with immersion time. The film growth process is controlled primarily by HS- diffusion in aqueous solution, implying that the sulphide film is not protective under these conditions up to this exposure time. However, when the sulphide concentration is > 5 x 10-4 mol/L, the sulphide film appears compact and its growth obeys a parabolic law. In this case, film growth is controlled mainly by Cu+ diffusion in the sulphide film indicating the film is protective under these conditions. By fitting EIS spectra to appropriate equivalent circuit models a diffusion coefficient for Cu+ in the sulphide film of 3 to 4 x 10-10 cm2/s was obtained. This value is considerably lower than the anticipated diffusion coefficient for HS- in compacted clay buffer (~ 107cm2/s), and, for these conditions, the overall corrosion rate would be limited by film properties not those of the compacted buffer. Whether or not a similar corrosion limitation would be exerted at lower sulphide concentrations, over waste disposal time frames, remains to be determined. Keywords: Copper, sulphide, corrosion, diffusion, nuclear waste disposal. Acknowledgements; This research is supported by the Swedish Nuclear Fuel and Waste Management Company (SKB), Stockholm, Sweden
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LongTermCor2010-15
Further studies of in situ corrosion testing of miniature coppercast iron nuclear waste canisters Nicholas R. Smart1, Andrew P. Rance1, Bharti. Reddy1, Sara Eriksson2, Karsten Pedersen2 and Christina Lilja3 1 Serco Technical Services, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB, U.K. 2 Microbial Analytics AB, Mölnlycke Fabriker 9, SE-435 35 Mölnlycke, Sweden. 3 Swedish Nuclear Waste Management Co, Box 250 SE-101 24, Stockholm, Sweden. +44 1635-280385 +44 1635-280389
[email protected] Abstract Background To ensure the safe encapsulation of spent nuclear fuel rods for geological disposal, the Swedish Nuclear Waste Management Co is considering using the Copper-Iron Canister, which consists of an outer copper shell and a cast iron insert. The canister will be surrounded by compacted bentonite and placed in a geological repository in a granite rock formation. In recent years a number of corrosion issues associated with the canister have been investigated experimentally, including the possibility of expansion of the outer copper shell as a result of the anaerobic corrosion of the cast iron insert. Experimental work using stacks of specimens to amplify any expansion effect due to anaerobic corrosion did not provide any evidence of corrosion-induced expansion. However, as a further step in developing an understanding of the likely performance of the canister in a repository environment, experiments are in progress using inactive model canisters, in which a leak has been deliberately introduced into the outer copper canister. The experiments are installed in SKB‟s Hard Rock Laboratory at Äspö near Oskarshamn in Sweden, where they are exposed to granitic groundwater. The main aim of the work is to provide information about how the environment inside a copper canister containing a cast iron insert would evolve if failure of the outer copper shell were to occur. This is important because the development of corrosion products in the gap between the copper shell and the cast iron insert could affect the rate of radionuclide release from the canister. In addition the following specific issues are addressed: Does water penetrate through a small defect into the annulus between the cast iron insert and the outer copper canister? How does corrosion product spread around the annulus in relation to the leak point? Does the formation of anaerobic corrosion product in a constricted annulus cause any expansive damage to the copper shell? Is there any detectable corrosion at the copper welds? Are there any deleterious galvanic interactions between copper and cast iron? Does corrosion lead to failure of the lid on the iron insert? Are there any effects of microbial corrosion on the canister? What are the corrosion rates of cast iron and copper in the repository environment? What is the risk of stress corrosion cracking of the copper?
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The corrosion processes are taking place under realistic oxygen-free repository conditions containing microbial populations that are very difficult to reproduce and maintain for long periods of time in the laboratory. Design of experiments Miniature canisters with a diameter of 14.5 cm and length 31.5 cm have been set up in five boreholes with a diameter of 30 cm and a length of 5 m. The model canister design simulates the main features of the SKB canister design. The cast iron insert contains four holes simulating the fuel pin channels, together with a bolted cast iron lid sealed with a Viton O-ring. The copper lid and base are electron beam welded to the cylindrical body. The annulus between the cast iron insert and the outer copper body is
