Tensile and stress corrosion cracking behavior of ferritic–martensitic steels in supercritical water

Ampornrat, Pantip; Gupta, Gaurav; Was, Gary S.

doi:10.1016/j.jnucmat.2009.09.012

Cited by 35 publications

(11 citation statements)

References 8 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…1 These steels have many favorable properties, such as higher thermal conductivity, lower susceptibility to stress corrosion cracking (SCC), and reduced swelling under irradiation. 2 At these temperatures, F-M steels possess higher strength than both zirconium alloys and austenitic stainless steels. 2 F-M steels are also being considered as candidate materials for accident tolerant fuel (ATF) to replace existing zirconium alloy cladding because the reaction with water and steam at high temperatures is less exothermic for F-M steels.…”

Section: Introductionmentioning

confidence: 98%

Stress corrosion cracking of ferritic-maretensitic steels in simulated boiling water reactor environment

Ahmedabadi

Was

2015

Corrosion

View full text Add to dashboard Cite

Stress corrosion cracking of APMT (Advanced Powder Metallurgy Technology) and T91 (ferritic-martensitic) steels were investigated in the as-received and proton-irradiated conditions in simulated boiling water reactor environment (2 ppm O 2 ) using constant extension tensile tests at 288°C at a strain rate of 3×10 -7 s -1 . Significant stress corrosion cracking was not observed in the as-received condition. A few cracks, perpendicular the loading direction, were observed in the proton-irradiated (5 dpa) specimen of T91. No intergranular fracture was observed on the fracture surfaces of as-received and proton-irradiated specimens of T91. No cracking was observed for APMT in un-irradiated and proton-irradiated (5 dpa) conditions. Results indicate that both APMT and T91 are highly resistant to stress corrosion cracking in a reactor environment and that irradiation to 5 dpa does not appreciably increase susceptibility.

show abstract

Section: Introductionmentioning

confidence: 98%

Stress corrosion cracking of ferritic-maretensitic steels in simulated boiling water reactor environment

Ahmedabadi

Was

2015

Corrosion

View full text Add to dashboard Cite

show abstract

“…The major challenge for materials in the SCW environment is the resistance to SCC and irradiation-assisted SCC. Austenitic stainless steels and nickel based alloys suffered IGSCC in deaerated SCW at 400°C (Allen et al, 2012) and ferritic-martensitic steels showed SCC resistance in the same temperature range (Ampornrat et al, 2009). Austenitic stainless steels exhibited extreme embrittlement under neutron irradiation of 40 dpa in SCW; however, ferriticmartensitic steels showed good resistance (Teysseyre et al, 2007).…”

Section: Baldev Raj and U Kamachi Mudalimentioning

confidence: 99%

Materials Science & Technology: Research and Challenges in Nuclear Fission Power

Raja¹

2015

Proceedings of the Indian National Science Academy

View full text Add to dashboard Cite

The performance and integrity of structural and functional materials are key issues in the safety and competitiveness of current and future nuclear fission reactors being developed for sustainable nuclear energy applications. The challenges towards development of nuclear structural materials arise due to the demanding hostile environments with respect to radiation, temperature, stress, etc., and requirements of total reliability and high performance over the long service life (say 60 years). A successful materials science research programme in nuclear industry has to take into account these challenges to improve the performance of materials and components in the emerging scenario of extending life of existing plants and realizing advanced reactors. In this study, material challenges associated with water reactors and fast breeder reactors (FBR) with focus on short-term and long-term strategies for materials development are considered. The materials in the existing and proposed future nuclear fission reactors are summarized along with a description of major material degradation mechanisms in different environments. The priority in the nuclear industry is to extend the life of reactors with robust safety features and sufficient cost-effectiveness, beyond forty to sixty years and from sixty to even one hundred years. The importance of modelling and developing predictive tools to estimate materials behaviour for effective computing of lifetime of nuclear reactor components is fast developing to cut down cost and time and also to enhance safety and confidence in existing and new systems. The future FBR technology relies heavily on advanced waste management and effective proliferation resistance. We review advanced reactor concepts of Generation IV forum and the new materials and technologies. Nuclear energy is not the right option for every country. Careful examination of the energy basket and commitment over a long period with effective mechanisms of safety governance is the key to make a decision to harness large amounts of nuclear energy. China, France, India, Russia, USA,UK, etc. are extremely committed to use large amounts of nuclear energy. Japan after the Fukushima accident, faces a challenge of public acceptance though the country has deep and rich expertise in nuclear technology and it is advantageous to produce low carbon power, on a sustained competitive basis from nuclear energy.

show abstract

“…This has included research on FM steels in a variety of environments including oxygenated or hydrogenated water [19], aqueous environments with a range of pH [20], SCW [21], liquid lead [22], and LBE [23][24][25][26]. All of these research studies have used miniature samples of different geometries under tensile load in the environment desired and evaluated SCC or LME by analyzing change in ductility from the results of the in situ environment to the results from a controlled environment (air or inert gas).…”

Section: Introductionmentioning

confidence: 99%

Evaluation of liquid metal embrittlement susceptibility of oxide dispersion strengthened steel MA956

Baker

Brewer

2014

Journal of Nuclear Materials

View full text Add to dashboard Cite

a b s t r a c tThis research examined the susceptibility of MA956 to liquid metal embrittlement using two experimental approaches. In both approaches, historical data on traditional steels was used to determine likely conditions to promote liquid metal embrittlement in lead and lead-bismuth eutectic environments. U-bend specimens of MA956 were found to be immune to liquid metal embrittlement after prolonged exposure to liquid lead. Similarly, slow strain rate testing of MA956 showed immunity to liquid metal embrittlement for both lead and lead-bismuth at temperatures of 328°C and 150°C respectively corresponding to the melting temperature of each embrittler individually. These results suggest that the same passive protective oxide layers that limit general corrosion and oxidation also prevent liquid metal embrittlement.Published by Elsevier B.V.

show abstract

Tensile and stress corrosion cracking behavior of ferritic–martensitic steels in supercritical water

Cited by 35 publications

References 8 publications

Stress corrosion cracking of ferritic-maretensitic steels in simulated boiling water reactor environment

Stress corrosion cracking of ferritic-maretensitic steels in simulated boiling water reactor environment

Materials Science & Technology: Research and Challenges in Nuclear Fission Power

Evaluation of liquid metal embrittlement susceptibility of oxide dispersion strengthened steel MA956

Contact Info

Product

Resources

About