Effect of the bainitic and martensitic microstructures on the hardening and embrittlement under neutron irradiation of a reactor pressure vessel steel

Energy Technology Data Exchange (ETDEWEB)

Marini, B., E-mail: bernard.marini@cea.fr [Commissariat à l' Energie Atomique et aux Energies Alternatives, DEN/DANS/DMN/SRMA, F-91191 Gif-sur Yvette (France); Averty, X. [Commissariat à l' Energie Atomique et aux Energies Alternatives, DEN/DANS/DMN/SEMI (now DEN/DANS/DM2S/SEMT), F-91191 Gif-sur Yvette (France); Wident, P.; Forget, P.; Barcelo, F. [Commissariat à l' Energie Atomique et aux Energies Alternatives, DEN/DANS/DMN/SRMA, F-91191 Gif-sur Yvette (France)

2015-10-15

The hardening and the embrittlement under neutron irradiation of an A508 type RPV steel considering three different microstructures (bainite, bainite-martensite and martensite)have been investigated These microstructures were obtained by quenching after autenitization at 1100 °C. The irradiation induced hardening appears to depend on microstructure and is correlated to the yield stress before irradiation. The irradiation induced embrittlement shows a more complex dependence. Martensite bearing microstructures are more sensitive to non hardening embrittlement than pure bainite. This enhanced sensitivity is associated with the development of intergranular brittle facture after irradiation; the pure martensite being more affected than the bainite-martensite. It is of interest to note that this mixed microstructure appears to be more embrittled than the pure bainitic or martensitic phases in terms of temperature transition shift. This behaviour which could emerge from the synergy of the embrittlement mechanisms of the two phases needs further investigations. However, the role of microstructure on brittle intergranular fracture development appears to be qualitatively similar under neutron irradiation and thermal ageing.

Correlation methodology for predicting in-service irradiation embrittlement of reactor pressure vessels

International Nuclear Information System (INIS)

Odette, G.R.

1980-01-01

Irradiation embrittlement of reactor pressure steels is the consequence of altered microstructure due to both irradiation and time-at-temperature. Relatively poor characterisation of the initial microstructure and chemistry, and inaccurate dosimetry and temperature control, as well as failure properly to correlate these variables, have all contributed to a very large scatter in the experimental embrittlement data base. This has made improvement of the basic understanding of embrittlement very difficult. Therefore, it is necessary to develop a more realistic approach to utilising the data base. This is discussed, and proposals are made. (author)

Development of neutron irradiation embrittlement correlation of reactor pressure vessel materials of light water reactors

International Nuclear Information System (INIS)

Soneda, Naoki; Dohi, Kenji; Nomoto, Akiyoshi; Nishida, Kenji; Ishino, Shiori

2007-01-01

A large amount of surveillance data of the RPV embrittlement of the Japanese light water reactors have been compiled since the current Japanese embrittlement correlation has been issued in 1991. Understanding on the mechanisms of the embrittlement has also been greatly improved based on both experimental and theoretical studies. CRIEPI and the Japanese electric power utilities have started research project to develop a new embrittlement correlation method, where extensive study of the microstructural analyses of the surveillance specimens irradiated in the Japanese commercial reactors has been conducted. The new findings obtained from the experimental study are that the formation of solute-atom clusters with little or no copper is responsible for the embrittlement in low-copper materials, and that the flux effect exists especially in high-copper materials and this is supported by the difference in the microstructure of the high-copper materials irradiated at different fluxes. Based on these new findings, a new embrittlement correlation method is formulated using rate equations. The new methods has higher prediction capability than the current Japanese embrittlement correlation in terms of smaller standard deviation as well as smaller mean value of the prediction error. (author)

Guidelines for prediction of irradiation embrittlement of operating WWER-440 reactor pressure vessels

International Nuclear Information System (INIS)

2005-06-01

This TECDOC has been developed under an International Atomic Energy Agency Coordinated Research Project (CRP) entitled Evaluation of Radiation Damage of WWER Reactor Pressure Vessels (RPV) using Database on RPV Materials to develop the guidelines for prediction of radiation damage to WWER-440 PRVs. The WWER-440 RPV was designed by OKB Gidropress, Russian Federation, the general designer. Prediction of irradiation embrittlement of RPV materials is usually done in accordance with relevant codes and standards that are based on the large amounts of information from surveillance and research programmes. The existing Russian code (standard for strength calculations of components and piping in NPPs - PNAE G 7-002-86) for the WWER RPV irradiation embrittlement assessment was approved more than twenty years ago and based mostly on the experimental data obtained in research reactors with accelerated irradiation. Nevertheless, it is still in use and generally consistent with new data. The present publication presents the analyses using all available data required for more precise prediction of radiation embrittlement of WWER-440 RPV materials. Based on the fact that it contains a large amount of data from surveillance programmes as well as research programmes, the IAEA International Database on RPV Materials (IDRPVM) is used for the detailed analysis of irradiation embrittlement of WWER RPV materials. Using IDRPVM, the guideline is developed for assessment of irradiation embrittlement of RPV ferritic materials as a result of degradation during operation. Two approaches, i.e. transition temperatures based on Charpy impact notch toughness, as well as based on static fracture toughness tests, are used in RPV integrity evaluation. The objectives of the TECDOC are the analysis of irradiation embrittlement data for WWER- 440 RPV materials using IDRPVM database, evaluation of predictive formulae depending on chemical composition of the material, neutron fluence, flux, and

Surveillance of irradiation embrittlement of nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Najzer, M.

1982-01-01

Surveillance of irradiation embrittlement of nuclear reactor pressure vessels is briefly discussed. The experimental techniques and computer programs available for this work at the J. Stefan Institute are described. (author)

Statistical analysis using the Bayesian nonparametric method for irradiation embrittlement of reactor pressure vessels

Energy Technology Data Exchange (ETDEWEB)

Takamizawa, Hisashi, E-mail: takamizawa.hisashi@jaea.go.jp; Itoh, Hiroto, E-mail: ito.hiroto@jaea.go.jp; Nishiyama, Yutaka, E-mail: nishiyama.yutaka93@jaea.go.jp

2016-10-15

In order to understand neutron irradiation embrittlement in high fluence regions, statistical analysis using the Bayesian nonparametric (BNP) method was performed for the Japanese surveillance and material test reactor irradiation database. The BNP method is essentially expressed as an infinite summation of normal distributions, with input data being subdivided into clusters with identical statistical parameters, such as mean and standard deviation, for each cluster to estimate shifts in ductile-to-brittle transition temperature (DBTT). The clusters typically depend on chemical compositions, irradiation conditions, and the irradiation embrittlement. Specific variables contributing to the irradiation embrittlement include the content of Cu, Ni, P, Si, and Mn in the pressure vessel steels, neutron flux, neutron fluence, and irradiation temperatures. It was found that the measured shifts of DBTT correlated well with the calculated ones. Data associated with the same materials were subdivided into the same clusters even if neutron fluences were increased.

The modelling of irradiation embrittlement in submerged-arc welds

International Nuclear Information System (INIS)

Bolton, C.J.; Buswell, J.T.; Jones, R.B.; Moskovic, R.; Priest, R.H.

1996-01-01

Until very recently, the irradiation embrittlement behavior of submerged-arc welds has been interpreted in terms of two mechanisms, namely a matrix damage component and an additional component due to the irradiation-enhanced production of copper-rich precipitates. However, some of the weld specimens from a recent accelerated re-irradiation experiment have shown high Charpy shifts which exceeded the values expected from the measured shift in yield stress. Microstructural examination has revealed the occurrence of intergranular fracture (IGF) in these specimens, accompanied by grain boundary segregation of phosphorus. Theoretical models were developed to predict the parametric dependence of irradiation-enhanced phosphorus segregation on experimental variables. Using these parametric forms, along with the concept of a critical level of segregation for the onset of IGF instead of cleavage, a three mechanism trend curve has been developed. The form of this trend curve, taking into account IGF as well as matrix and copper embrittlement, is thus mechanistically based. The constants in the equation, however, are obtained by a statistical fit to the actual Charpy shift database

Severe embrittlement of neutron irradiated austenitic steels arising from high void swelling

Energy Technology Data Exchange (ETDEWEB)

Neustroev, V.S. [FSUE ' SSC RF Research Institute of Atomic Reactors' , Dimitrovgrad (Russian Federation)], E-mail: neustroev@niiar.ru; Garner, F.A. [Pacific Northwest National Laboratory, Richland, WA (United States)

2009-04-30

Data are presented from BOR-60 irradiations showing that significant radiation-induced swelling causes severe embrittlement in austenitic stainless steels, reducing the service life of structural components and introducing limitations on low temperature handling especially. It is shown that the degradation is actually a form of quasi-embrittlement arising from intense flow localization with high levels of localized ductility involving micropore coalescence and void-to-void cracking. Voids initially serve as hardening components whose effect is overwhelmed by the void-induced reduction in shear and Young's moduli at high swelling levels. Thus the alloy appears to soften even as the ductility plunges toward zero on a macroscopic level although a large amount of deformation occurs microscopically at the failure site. Thus the failure is better characterized as 'quasi-embrittlement' which is a suppression of uniform deformation. This case should be differentiated from that of real embrittlement which involves the complete suppression of the material's capability for plastic deformation.

A study on the irradiation embrittlement and recovery characteristics of light water reactor pressure vessel steels

International Nuclear Information System (INIS)

Chi, Se Hwan; Hong, Jun Hwa; Lee, Bong Sang; Oh, Jong Myung; Song, Sook Hyang; Milan, Brumovsky

1999-03-01

The neutron irradiation embrittlement phenomenon of light water RPV steels greatly affects the life span for safe operation of a reactor. Reliable evaluation and prediction of the embrittlement of RPV steels, especially of aged reactors, are of importance to the safe operation of a reactor. In addition, the thermal recovery of embrittled RPV has been recognized as an option for life extension. This study aimed to tracer/refine available technologies for embrittlement characterization and prediction, to prepare relevant materials for several domestic RPV steels of the embrittlement and recovery, and to find out possible remedy for steel property betterment. Small specimen test techniques, magnetic measurement techniques, and the Meechan and Brinkmann's recovery curve analysis method were examined/applied as the evaluation techniques. Results revealed a high irradiation sensitivity in YG 3 RPV steel. Further extended study may be urgently needed. Both the small specimen test technique for the direct determination of fracture toughness, and the magnetic measurement technique for embrittlement evaluation appeared to be continued for the technical improvement and data base preparation. Manufacturing process relevant to the heat treatment appeared to be improved in lowering the irradiation sensitivity of the steel. Further study is needed especially in applying the present techniques to the new structural materials under new irradiation environment of advanced reactors. (author)

Beryllium irradiation embrittlement test programme. Material and specimen specification, manufacture and qualification

International Nuclear Information System (INIS)

Harries, D.R.; Dalle Donne, M.

1996-06-01

The report presents the specification, manufacture and qualification of the beryllium specimens to be irradiated in the BR2 reactor in Mol to investigate the effect of the neutron irradiation on the embrittlement as a function of temperature and beryllium oxide content. This work was been performed in the framework of the Nuclear Fusion Project of the Forschungszentrum Karlsruhe and is supported by the European Union within the European Fusion Technology Program. (orig.)

The effect of deformation twinning on irradiation embrittlement in iron single crystals

International Nuclear Information System (INIS)

Kayano, Hideo; Tokutomi, Shoichiro; Yajima, Seishi; Takaku, Hiroshi.

1978-01-01

Single crystals of iron with the [100] crystal orientation were irradiated in JMTR with fast neutrons to a fluence of 8 x 10 18 n/cm 2 (E > 1 MeV). All samples were deformed in tension at temperatures from liquid nitrogen temperature to 200 0 C at different strain rates using an Instron-type tensile testing machine. Scanning electron microscopy of the fractured surfaces revealed that deformation twinning is difficult to occur in irradiated samples, and also that twins formed in both irradiated and unirradiated samples inhibit fracture nucleation and growth. From the results of tensile deformation of the irradiated samples deformed in tension a different strain rates at 159 0 K, it is conceived that twinning suppression is greater in the irradiated than in the unirradiated samples, and that the nucleation and growth of twins are not necessarily related to those of cracks. It is suggested that the irradiation-induced defects impede plastic deformation of the crystals and deformation twinning is suppressed by irradiation, thus causing the irradiation embrittlement. (auth.)

A study on the irradiation embrittlement and recovery characteristics of light water reactor pressure vessel steels

Energy Technology Data Exchange (ETDEWEB)

Chi, Se Hwan; Hong, Jun Hwa; Lee, Bong Sang; Oh, Jong Myung; Song, Sook Hyang; Milan, Brumovsky [NRI Czech (Czech Republic)

1999-03-01

The neutron irradiation embrittlement phenomenon of light water RPV steels greatly affects the life span for safe operation of a reactor. Reliable evaluation and prediction of the embrittlement of RPV steels, especially of aged reactors, are of importance to the safe operation of a reactor. In addition, the thermal recovery of embrittled RPV has been recognized as an option for life extension. This study aimed to tracer/refine available technologies for embrittlement characterization and prediction, to prepare relevant materials for several domestic RPV steels of the embrittlement and recovery, and to find out possible remedy for steel property betterment. Small specimen test techniques, magnetic measurement techniques, and the Meechan and Brinkmann's recovery curve analysis method were examined/applied as the evaluation techniques. Results revealed a high irradiation sensitivity in YG 3 RPV steel. Further extended study may be urgently needed. Both the small specimen test technique for the direct determination of fracture toughness, and the magnetic measurement technique for embrittlement evaluation appeared to be continued for the technical improvement and data base preparation. Manufacturing process relevant to the heat treatment appeared to be improved in lowering the irradiation sensitivity of the steel. Further study is needed especially in applying the present techniques to the new structural materials under new irradiation environment of advanced reactors. (author)

Microstructural design of PCA austenitic stainless steel for improved resistance to helium embrittlement under HFIR irradiation

International Nuclear Information System (INIS)

Maziasz, P.J.; Braski, D.N.

1983-01-01

Several variants of Prime Candidate Alloy (PCA) with different preirradiation thermal-mechanical treatments were irradiated in HFIR and were evaluated for embrittlement resistance via disk-bend tensile testing. Comparison tests were made on two heats of 20%-cold-worked type 316 stainless steel. None of the alloys were brittle after irradiation at 300 to 400 0 C to approx. 44 dpa and helium levels of 3000 to approx.3600 at. ppm. However, all were quite brittle after similar exposure at 600 0 C. Embrittlement varied with alloy and pretreatment for irradiation to 44 dpa at 500 0 C and to 22 dpa at 600 0 C. Better relative embrittlement resistance among PCA variants was found in alloys which contained prior grain boundary MC carbide particles that remained stable under irradiation

Mechanisms of radiation embrittlement of VVER-1000 RPV steel at irradiation temperatures of (50–400)°C

Energy Technology Data Exchange (ETDEWEB)

Kuleshova, E.A., E-mail: evgenia-orm@yandex.ru [National Research Center “Kurchatov Institute”, Kurchatov Sq. 1, Moscow 123182 (Russian Federation); National Research Nuclear University “MEPhI” (Moscow Engineering Physics Institute), Kashirskoe Highway 31, Moscow 115409 (Russian Federation); Gurovich, B.A.; Bukina, Z.V.; Frolov, A.S.; Maltsev, D.A.; Krikun, E.V.; Zhurko, D.A.; Zhuchkov, G.M. [National Research Center “Kurchatov Institute”, Kurchatov Sq. 1, Moscow 123182 (Russian Federation)

2017-07-15

This work summarizes and analyzes our recent research results on the effect of irradiation temperature within the range of (50–400)°C on microstructure and properties of 15Kh2NMFAA class 1 steel (VVER-1000 reactor pressure vessel (RPV) base metal). The paper considers the influence of accelerated irradiation with different temperature up to different fluences on the carbide and irradiation-induced phases, radiation defects, yield strength changes and critical brittleness temperature shift (ΔT{sub K}) as well as on changes of the fraction of brittle intergranular fracture and segregation processes in the steel. Low temperature irradiation resulted solely in formation of radiation defects – dislocation loops of high number density, the latter increased with increase in irradiation temperature while their size decreased. In this regard high embrittlement rate observed at low temperature irradiation is only due to the hardening mechanism of radiation embrittlement. Accelerated irradiation at VVER-1000 RPV operating temperature (∼300 °C) caused formation of radiation-induced precipitates and dislocation loops, as well as some increase in phosphorus grain boundary segregation. The observed ΔT{sub K} shift being within the regulatory curve for VVER-1000 RPV base metal is due to both hardening and non-hardening mechanisms of radiation embrittlement. Irradiation at elevated temperature caused more intense phosphorus grain boundary segregation, but no formation of radiation-induced precipitates or dislocation loops in contrast to irradiation at 300 °C. Carbide transformations observed only after irradiation at 400 °C caused increase in yield strength and, along with a contribution of the non-hardening mechanism, resulted in the lowest ΔT{sub K} shift in the studied range of irradiation temperature and fluence. - Highlights: •Structural elements in RPV steel are studied at different irradiation temperatures. •Highest number density dislocation loops are

Fluence-rate effects on irradiation embrittlement and composition and temperature effects on annealing/reirradiation sensitivity

International Nuclear Information System (INIS)

Hawthorne, J.R.; Hiser, A.L.

1988-01-01

Recent MEA investigation on the effect of neutron fluence rate on radiation-induced embrittlement accrual and the contributions of metallurgical variables to postirradiation annealing and re-irradiation behavior are reviewed. Studies of fluence-rate effects involved experiments in the UBR test reactor and separately, radiation sensitivity determinations for the decommissioned Gundremmingen (KRB-A) vessel material. Annealing-reirradiation studies employed 399 0 C and 454 0 C heat treatments. Material composition is shown to play a major role in postirradiation annealing recovery. Results illustrate effects of variable copper and variable nickel contents on recoveray of steel plate having low phosphorus levels. Composition effects on recovery were also observed for prototypic welds depicting high/low copper and high/low nickel contents and three flux types. The welds, in addition, indicate major differences in re-irradiation sensitivity. The UBR investigations revealed a significant difference in fluence rate sensitivity between the ASTM A 302-B reference plate and a submerged-arc (S/A) Linde 80 weld. Studies of the Gundremmingen reactor vessel, representing a joint USA-FRG-UK undertaking revealed an anomaly in strong vs. weak test orientation radiation sensitivity. (orig./HP)

The flow effect in the irradiation embrittlement in pressure vessel steels of nuclear power plants

International Nuclear Information System (INIS)

Kempf, Rodolfo A.; Cativa Tolosa, Sebastian; Fortis, Ana M.

2009-01-01

This paper deals with the advances in the study of the mechanical behavior of the Reactor Pressure Vessel steels under accelerate irradiations. The objective is to study the effect of lead factors on the interpretation of the mechanisms that induced the embrittlement of the RPV, like those of the reactors Atucha II and CAREM. It is described a device designed to irradiate Charpy specimens with V notch of SA-508 type 3 steel at power reactor temperature, installed in the RA-1 reactor. It is presented also an automatic digital image processing technique for partitioning Charpy fracture surface into regions with a clear physical meaning and appropriate for the work in hot cells. The aim is to obtain the fracture behavior of irradiated specimens with different lead factors in the range of high fluencies and to know the dependence with the composition of the alloy and with the diffusion of other alloy elements. (author)

Irradiation embrittlement of some 15Kh2MFA pressure vessel steels under varying neutron fluence rates

Energy Technology Data Exchange (ETDEWEB)

Valo, M; Bars, B [Technical Research Centre of Finland, Espoo (Finland); Ahlstrand, A [Imatran Voima Oy (IVO), Helsinki (Finland)

1994-12-31

Irradiation sensitivity of two forging materials was measured with Charpy-V and fracture mechanic tests, and with different fluence, fluence rate and irradiation time values. Irradiation sensitivity of the materials was found to be less or equal to the current Russian standard, and appears to be well described by the fluence parameter only. A slight additional effect on embrittlement from a long term low fluence irradiation is noticed, but it stays within the total scatter band of data. 7 refs., 17 figs., 4 tabs.

Temper embrittlement, irradiation induced phosphorus segregation and implications for post-irradiation annealing of reactor pressure vessels

International Nuclear Information System (INIS)

McElroy, R.J.; English, C.A.; Foreman, A.J.; Gage, G.; Hyde, J.M.; Ray, P.H.N.; Vatter, I.A.

1999-01-01

Three steels designated JPB, JPC and JPG from the IAEA Phase 3 Programme containing two copper and phosphorus levels were pre- and post-irradiation Charpy and hardness tested in the as-received (AR), 1200 C/0.5h heat treated (HT) and heat treated and 450 C/2000h aged (HTA) conditions. The HT condition was designed to simulate coarse grained heat-affected zones (HAZ's) and showed a marked sensitivity to thermal ageing in all three alloys. Embrittlement after thermal ageing was greater in the higher phosphorus alloys JPB and JPG. Charpy shifts due to thermal ageing of between 118 and 209 C were observed and accompanied by pronounced intergranular fracture, due to phosphorus segregation. The irradiation embrittlement response was complex. The low copper alloys, JPC and JPB, in the HT and HTA condition exhibited significant irradiation induced Charpy shift but very low or even negative hardness changes indicating non-hardening embrittlement. The higher copper alloy, JPG, also exhibited irradiation hardening in line with its copper content. Fractographic and microchemical studies indicated irradiation induced phosphorus segregation and a transition from cleavage to intergranular failure at grain boundary phosphorus concentrations above a critical level. The enhanced grain boundary phosphorus level increased with dose in agreement with a kinetic segregation model developed at Harwell. The relevance of the thermal ageing studies to RPV Annealing for Plant-Life Extension was identified early in the program. It is of concern that annealing of RPV's has been performed, or is proposed, at temperatures in the range 425--475 C for periods of about 1 week (168h). Much attention has been given to the use of in-situ hardness measurements and machining miniature Charpy and tensile specimens from belt-line plate and weld materials. However, HAZ's, often containing higher phosphorus levels than the present materials, have largely been ignored. A post-irradiation annealing (PIA

Influence of helium embrittlement on post-irradiation creep rupture behaviour of austenitic and martensitic stainless steels

International Nuclear Information System (INIS)

Wassilew, C.

1982-01-01

The author has investigated the influence of helium embrittlement on the creep rupture properties of the austenitic stainless steels 1.4970 and 1.4962 and the martensitic stainless steel 1.4914 after irradiation in the BR-2 reactor in Mol, Belgium. The results show that austenitic steels react much more strongly to the embrittlement effect of the helium than do martensitic steels. The causes of the lower embrittlement tendency of the martensitic than of both austenitic stainless steels were analysed carefully. A new embrittlement model was developed on the basis of data derived from the creep rupture experiments, and reinforced by a simple metallographic investigation of the fracture zone and its immediate environment. This model pays specific attention to the role of the twin planes as the most efficient area of increased vacancy production, and takes into account the ability of the twin boundaries to transport these vacancies with reduced energy and low loss into the high-angle grain boundaries. (author)

Results from Project on Enhancement of Aging Management and Maintenance in Nuclear Power Plants - Irradiation Embrittlement of RPV Steels -

International Nuclear Information System (INIS)

Abe, Hiroaki; Onizawa, Kunio; Katsuyama, Jinya; Murakami, Kenta; Iwai, Takeo; Iwata, Tadao; Katano, Yoshio; Sekimura, Naoto; Nagai, Yasuyoshi; Toyama, Takeshi; Tamura, Satoshi

2012-01-01

As one of the NISA Project on Enhancement of Aging Management and Maintenance in Nuclear Power Plants, we have performed research on the irradiation embrittlement of reactor pressure vessel (RPV) steels, especially focusing on irradiation embrittlement on heat affected zone (HAZ) and on applications of ion beams to deduce fundamental insights irradiation-induced embrittlement. The results obtained from the project are summarized as follows. In order to obtain the technical basis to judge the necessity of surveillance specimens from HAZ, the neutron irradiation program was performed at JRR-3, JAEA. The samples were carefully designed based on the insights from finite element analysis, metallography, 3D atom probe and positron annihilation methods, and were fabricated so as to simulate both heat treatment history and microstructure for typical HAZ from as-fabricated RPV steels which also have variation of impurity levels. The fracture toughness of the unirradiated HAZ specimens was equivalent to or better than that of base metals. Irradiation embrittlement and hardening were roughly identical to those of base metals, while some of the fine-grained HAZ microstructure was susceptible to it. The probabilistic fracture mechanics analysis was applied to the structural integrity assessment taking into account the heterogeneous microstructure as well as susceptibility for irradiation embrittlement of each HAZ microstructure under the variation of welding parameter and PTS condition. It was shown that crack propagation at the fine-grained HAZ, but the discontinuous distribution of the microstructure retards the further propagation. For the precise correlation of irradiation embrittlement of RPV steels for the long term operations, accumulations of high-dose data are required. Ion beam irradiation is one of the solutions for the regime and for mechanism-based descriptions. Another interest of ours was to describe irradiation hardening and embrittlement in terms of

Embrittlement data base, version 1

International Nuclear Information System (INIS)

Wang, J.A.

1997-08-01

The aging and degradation of light-water-reactor (LWR) pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel (RPV) materials depends on many different factors such as flux, fluence, fluence spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Based on embrittlement predictions, decisions must be made concerning operating parameters and issues such as low-leakage-fuel management, possible life extension, and the need for annealing the pressure vessel. Large amounts of data from surveillance capsules and test reactor experiments, comprising many different materials and different irradiation conditions, are needed to develop generally applicable damage prediction models that can be used for industry standards and regulatory guides. Version 1 of the Embrittlement Data Base (EDB) is such a comprehensive collection of data resulting from merging version 2 of the Power Reactor Embrittlement Data Base (PR-EDB). Fracture toughness data were also integrated into Version 1 of the EDB. For power reactor data, the current EDB lists the 1,029 Charpy transition-temperature shift data points, which include 321 from plates, 125 from forgoings, 115 from correlation monitor materials, 246 from welds, and 222 from heat-affected-zone (HAZ) materials that were irradiated in 271 capsules from 101 commercial power reactors. For test reactor data, information is available for 1,308 different irradiated sets (352 from plates, 186 from forgoings, 303 from correlation monitor materials, 396 from welds and 71 from HAZs) and 268 different irradiated plus annealed data sets

A review of formulas for predicting irradiation embrittlement of reactors vessel materials

International Nuclear Information System (INIS)

Petrequin, P.

1995-01-01

Formulas developed in different countries for predicting irradiation embrittlement of reactors vessel materials are presented. Results of predictions were compared with different data sets, from surveillance programmes or studies in test reactors, with different residual elements contents. Figs

Radiation embrittlement of metals and alloys

International Nuclear Information System (INIS)

Wechsler, M.S.

1975-01-01

Three types of radiation embrittlement are identified: (1) radiation embrittlement in nominally ductile metals, (2) radiation embrittlement in metals that undergo a ductile-brittle transition, and (3) high-temperature grain boundary embrittlement. This paper deals with type (1) and, more briefly, type (2) radiation embrittlement. Radiation embrittlement in nominally ductile metals is characterized by the premature onset of plastic instability, which causes a sharp decrease in the macroscopic plastic strain that the material can sustain before necking (uniform strain) and breaking (fracture strain). Dislocation channeling seems to be largely responsible and experimental results are reviewed. The origin of dislocation channeling is discussed. Irradiated metals that exhibit a ductile-brittle transition show an increase in the transition temperature but the nature of the transition (shear to cleavage fracture) does not appear to be greatly altered. A key factor is the temperature dependence of yielding and how it is affected upon irradiation. Impurities exert an influence on the stability of radiation-produced defect clusters and thus can alter the amount of radiation embrittlement experienced upon irradiation at somewhat elevated temperatures. In general, radiation embrittlement appears to stem mostly from changes in plastic properties (particularly in the trend toward more dynamic and inhomogeneous plastic deformation) rather than from changes in the inherent fracture process. 63 references, 10 figures

The Test Reactor Embrittlement Data Base (TR-EDB)

International Nuclear Information System (INIS)

Stallmann, F.W.; Kam, F.B.K.; Wang, J.A.

1993-01-01

The Test Reactor Embrittlement Data Base (TR-EDB) is part of an ongoing program to collect test data from materials irradiations to aid in the research and evaluation of embrittlement prediction models that are used to assure the safety of pressure vessels in power reactors. This program is being funded by the US Nuclear Regulatory Commission (NRC) and has resulted in the publication of the Power Reactor Embrittlement Data Base (PR-EDB) whose second version is currently being released. The TR-EDB is a compatible collection of data from experiments in materials test reactors. These data contain information that is not obtainable from surveillance results, especially, about the effects of annealing after irradiation. Other information that is only available from test reactors is the influence of fluence rates and irradiation temperatures on radiation embrittlement. The first version of the TR-EDB will be released in fall of 1993 and contains published results from laboratories in many countries. Data collection will continue and further updates will be published

TR-EDB: Test Reactor Embrittlement Data Base, Version 1

Energy Technology Data Exchange (ETDEWEB)

Stallmann, F.W.; Wang, J.A.; Kam, F.B.K. [Oak Ridge National Lab., TN (United States)

1994-01-01

The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results.

TR-EDB: Test Reactor Embrittlement Data Base, Version 1

International Nuclear Information System (INIS)

Stallmann, F.W.; Wang, J.A.; Kam, F.B.K.

1994-01-01

The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results

Reactor pressure vessel embrittlement: Insights from neural network modelling

Science.gov (United States)

Mathew, J.; Parfitt, D.; Wilford, K.; Riddle, N.; Alamaniotis, M.; Chroneos, A.; Fitzpatrick, M. E.

2018-04-01

Irradiation embrittlement of steel pressure vessels is an important consideration for the operation of current and future light water nuclear reactors. In this study we employ an ensemble of artificial neural networks in order to provide predictions of the embrittlement using two literature datasets, one based on US surveillance data and the second from the IVAR experiment. We use these networks to examine trends with input variables and to assess various literature models including compositional effects and the role of flux and temperature. Overall, the networks agree with the existing literature models and we comment on their more general use in predicting irradiation embrittlement.

Severe Embrittlement of Neutron Irradiated Austenitic Steels Arising from High Void Swelling

International Nuclear Information System (INIS)

Neustroev, V.S.; Garner, F.

2007-01-01

Full text of publication follows: Data are presented from BOR-60 irradiations showing that significant radiation-induced swelling causes severe embrittlement in austenitic stainless steels, reducing the service life of structural components. Similar loss of ductility is expected when swelling arises in fusion and light water reactor environments. Above 7-16% swelling there is complete loss of ductility, with the onset of ductility loss beginning at lower swelling in ring-pull tensile tests than for flat tensile specimens. For steels that develop extensive precipitation during irradiation, the critical swelling level is even lower. A model is presented to demonstrate the effect of voids acting alone to produce the embrittlement. Although voids are not very effective hardeners, they are very effective to generate stress concentrations between voids. The stress concentration ratio increases strongly when the void diameter exceeds ∼40% of the void-to-void separation distance. When the volume fraction of voids is rather high (about 16 % and higher), a geometric situation develops where it is possible to create an intense field of deformation glide planes residing at an angle of 45 deg. to the void-to-void axis. Significant localized flow then proceeds on these planes for specimen stress levels that are significantly lower than the yield stress. Voids also segregate nickel to their surfaces such that flow localization occurs in the low-nickel inter-void regions to produce strain-induced martensite, which is further accelerated by stress concentrations at the advancing crack tip, leading to catastrophic failure. (authors)

Overview of French activities on neutron radiation embrittlement of pressure vessel steel

Energy Technology Data Exchange (ETDEWEB)

Brillaud, C [Electricite de France (EDF), 37 - Tours (France); Keroulas, F de [Electricite de France (EDF), 93 - Saint-Denis (France); Pichon, C [Electricite de France (EDF), 69 - Villeurbanne (France); Teissier, A [Electricite de France (EDF), 92 - Courbevoie (France). Service Etudes et Projets Thermiques et Nucleaires

1994-12-31

This paper describes recent developments in France`s pressure vessel surveillance program, particularly aimed at assessing the irradiation-caused embrittlement of EDF`s PWRs. The first part presents surveillance program results for base metal, weld metal and heat-affected zones for 74 capsules removed from 34 units. Fluence ranges from 0.3.10{sup 19} n.cm{sup -2} to 5.5.10{sup 19} n.cm{sup -2}. The second part considers research and development activities in this area: these include the metallurgical structure effects of segregated bands on mechanical properties and the embrittlement rate under irradiation, as well as the effect of irradiation parameters such as flux and neutron spectrum on irradiation embrittlement, and more especially to obtain the best damage assessment. (authors). 14 refs., 5 figs., 1 tab.

Investigation of irradiation embrittlement and annealing behaviour of JRQ pressure vessel steel by instrumented impact tests

Energy Technology Data Exchange (ETDEWEB)

Valo, M; Rintamaa, R; Nevalainen, M; Wallin, K; Torronen, K [Technical Research Centre of Finland, Espoo (Finland); Tipping, P [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

1994-12-31

Seven series of A533-B type pressure vessel steel specimens irradiated as well as irradiated - annealed - re-irradiated to different fast neutron fluences (up to 5.10{sup 19}/cm{sup 2}) have been tested with a new type of instrumented impact test machine. The radiation embrittlement and the effect of the intermediate annealing was assessed by using the ductile and cleavage fracture initiation toughness. Although the ductile fracture initiation toughness exhibited scatter, the transition temperature shift corresponding to the dynamic cleavage fracture initiation agreed well with the 41 J Charpy-V shift. The results indicate that annealing is beneficial in restoring mechanical properties in an irradiated nuclear pressure vessel steel. (authors). 8 refs., 11 figs., 1 tab.

The irradiation embrittlement of two pressure vessel steels -Contribution of local approach

Energy Technology Data Exchange (ETDEWEB)

Soulat, P; Marini, B [CEA Centre d` Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France). Service de Recherches Metallurgiques Appliquees; Miannay, D; Horowitz, H [CEA Centre d` Etudes de Fontenay-aux-Roses, 92 (France). Inst. de Protection et de Surete Nucleaire; Schill, R [CEA Centre d` Etudes de Saclay, 91 - Gif-sur-Yvette (France). Dept. de Mecanique et de Technologie

1994-12-31

Within the IAEA Coordinated Research Programme on ``Optimizing the Reactor Pressure Vessel Surveillance Programmes and their Analyses``, the French participation has been focused on the contribution of the local approach to the determination of the sensitivity to radiation embrittlement of two different pressure vessel steels: a low sensitive French forging steel (FFA) and a high sensitive ``monitor`` Japanese plate steel (JRQ) were irradiated to a fluence of 3.10{sup 19} n/cm{sup 2} at 290 C. The irradiation embrittlement of the two steels measured by the shift of Charpy V transition curves is in good agreement with the estimated shifts given by theoretical prediction. The fracture toughness properties were examined at low temperature with brittle fracture, and at service temperature (290 C), with ductile tearing. The values of K{sub 1C} or K{sub JC} for the brittle fracture and J{sub 1C} for the ductile fracture are compared to predictions established using the local approach of cleavage fracture (Weibull analysis) and the critical rate of void growth respectively. 8 refs., 14 figs., 10 tabs.

Evaluation of neutron irradiation embrittlement in the Korean reactor pressure vessel steels (Final Report)

Energy Technology Data Exchange (ETDEWEB)

Hong, J. H.; Lee, B. S.; Chi, S. H.; Kim, J. H.; Oh, Y. J.; Yoon, J. H.; Kwon, S. C.; Park, D. G.; Kang, Y. H.; Choo, K. N.; Oh, J. M.; Park, S. J.; Kim, B. K.; Shin, Y. T.; Cho, M. S.; Sohn, J. M.; Kim, D. S.; Choo, Y. S.; Ahn, S. B.; Oh, W. H. [Korea Atomic Energy Research Institute, Taejeon (Korea)

2001-05-01

Reactor pressure vessel materials, which were produced by a domestic company, Doosan Heavy Industries and construction Co., Ltd., have been evaluated using the neutron irradiation facility HANARO. For this evaluation, instrumented capsules were used for neutron irradiation of various kinds of specimens made of different heats of steels, which are VCD(Y4), VCD+Al(U4), Si+Al(Y5), U4 weld metal, and U4 HAZ, respectively. The fast neutron fluence levels ranged 1 to 5 (x10{sup 19} n/cm{sup 2}, E>1MeV) depending on the specimens and the irradiation temperature was controlled within 290{+-}10 deg C. The test results showed that, in the ranking of the material properties of the base metals, both before and after neutron irradiation, Y5 is the best, U4 the next and Y4 the last. Y4 showed a substantial change by neutron irradiation as well as the properties was worse than others in the unirradiated state. However, Y5, which showed the best properties in unirradiated state, was also the best in the resistance for irradiation embrittlement and one can hardly detect the property change after irradiation. The weldment showed a reasonably good resistance to irradiation embrittlement while the unirradiated properties were worse than base metals. The RPV steels are all expected to meet the screening criteria of the USNRC codes and regulations during the end of plant life. 39 refs., 42 figs., 27 tabs. (Author)

Cooperation modes of the radiation embrittlement

International Nuclear Information System (INIS)

Voevodin, V.N.; Laptev, I.N.; Neklyudov, I.M.; Ozhigov, L.S.; Bryk, V.V.; Parkhomenko, A.A.

2012-01-01

According to the results of experimental and theoretical studies of the structures and properties of irradiated deformed materials with different crystalline structure, the effect of irradiation on mechanisms of radiation embrittlement on all structure levels (from atomic to macrolevel) has been shown. The effects of structural localization, collectivization, long range effects, rotation modes development are described. It was shown that these effects are closely interrelated; they characterized the deformed irradiation material as open dissipative system subjected to the laws of such scientific approach as synergetic.

Lifetime embrittlement of reactor core materials

International Nuclear Information System (INIS)

Kreyns, P.H..; Bourgeois, W.F.; Charpentier, P.L.; Kammenzind, B.F.; Franklin, D.G.; White, C.J.

1994-08-01

Over a core lifetime, the reactor materials Zircaloy-2, Zircaloy-4, and hafnium may become embrittled due to the absorption of corrosion- generated hydrogen and to neutron irradiation damage. Results are presented on the effects of fast fluence on the fracture toughness of wrought Zircaloy-2, Zircaloy-4, and hafnium; Zircaloy-4 to hafnium butt welds; and hydrogen precharged beta treated and weld metal Zircaloy-4 for fluences up to a maximum of approximately 150 x 10 24 n/M 2 (> 1 Mev). While Zircaloy-4 did not exhibit a decrement in K IC due to irradiation, hafnium and butt welds between hafnium and Zircaloy-4 are susceptible to embrittlement with irradiation. The embrittlement can be attributed to irradiation strengthening, which promotes cleavage fracture in hafnium and hafnium-Zircaloy welds, and, in part, to the lower chemical potential of hydrogen in Zircaloy-4 compared to hafnium, which causes hydrogen, over time, to drift from the hafnium end toward the Zircaloy-4 end and to precipitate at the interface between the weld and base-metal interface. Neutron radiation apparently affects the fracture toughness of Zircaloy-2, Zircaloy-4, and hafnium in different ways. Possible explanations for these differences are suggested. It was found that Zircaloy-4 is preferred over Zircaloy-2 in hafnium-to- Zircaloy butt-weld applications due to its absence of a radiation- induced reduction in K IC plus its lower hydrogen absorption characteristics compared with Zircaloy-2

An internal-friction study of reactor-pressure-vessel steel embrittlement

International Nuclear Information System (INIS)

Ouytsel, K. van; Fabry, A.; Batist, R. de; Schaller, R.

1997-01-01

Within an enhanced commercial surveillance strategy, the nuclear-research institute SCK.CEN in Mol, Belgium is investigating, by means of internal friction, the microstructural processes responsible for embrittlement of pressure-vessel steels. The experiments were carried out using a torsion pendulum at the Ecole Polytechnique Federale de Lausanne in Switzerland. Amplitude-independent internal-friction experiments teach us that neutron irradiation induces defects which interact with mobile dislocations. Thermal ageing of JRQ and Doel-IV steel does not cause major embrittlement effects. Amplitude-dependent internal-friction experiments allow us to determine a critical amplitude which corresponds to the yield stress of the material as obtained from static tensile tests. The results also correspond to a three-component model for the yield strength taking into account both hardening and non-hardening embrittlement. Investigations of Doel-I-II weld material in different conditions reveal that embrittlement due to irradiation or thermal ageing can be interpreted in terms of a fine interplay between long- and short-range phenomena. (author)

SCK-CEN Contribution to the IAEA Round Robin Exercise on WWER-440 RPV Weld Metal Irradiation Embrittlement, annealing and Re-Embrittlement. Second Progress Report

International Nuclear Information System (INIS)

Van Walle, E.; Chaouadi, R.; Scibetta, M.; Lucon, E.; Weber, M.

1999-07-01

The report gives the actual status of the contribution of the Belgian Nuclear Research Centre SCK-CEN to the IAEA Round Robin Exercise on WWER-440 RPV Weld Material Irradiation, Annealing and Re-Embrittlement. Results from the reference testing of unirradiated material as well as the results of the CHIVAS-7 experiment are discussed

Models for embrittlement recovery due to annealing of reactor pressure vessel steels

International Nuclear Information System (INIS)

Eason, E.D.; Wright, J.E.; Nelson, E.E.; Odette, G.R.; Mader, E.V.

1995-05-01

The reactor pressure vessel (RPV) surrounding the core of a commercial nuclear power plant is subject to embrittlement due to exposure to high energy neutrons. The effects of irradiation embrittlement can be reduced by thermal annealing at temperatures higher than the normal operating conditions. However, a means of quantitatively assessing the effectiveness of annealing for embrittlement recovery is needed. The objective of this work was to analyze the pertinent data on this issue and develop quantitative models for estimating the recovery in 30 ft-lb (41 J) Charpy transition temperature and Charpy upper shelf energy due to annealing. Data were gathered from the Test Reactor Embrittlement Data Base and from various annealing reports. An analysis data base was developed, reviewed for completeness and accuracy, and documented as part of this work. Independent variables considered in the analysis included material chemistries, annealing time and temperature, irradiation time and temperature, fluence, and flux. To identify important variables and functional forms for predicting embrittlement recovery, advanced statistical techniques, including pattern recognition and transformation analysis, were applied together with current understanding of the mechanisms governing embrittlement and recovery. Models were calibrated using multivariable surface-fitting techniques. Several iterations of model calibration, evaluation with respect to mechanistic and statistical considerations, and comparison with the trends in hardness data produced correlation models for estimating Charpy upper shelf energy and transition temperature after irradiation and annealing. This work provides a clear demonstration that (1) microhardness recovery is generally a very good surrogate for shift recovery, and (2) there is a high level of consistency between the observed annealing trends and fundamental models of embrittlement and recovery processes

Neutron irradiation embrittlement of reactor pressure vessel steels

International Nuclear Information System (INIS)

Steele, L.E.

1975-01-01

The reliability of nuclear power plants depends on the proper functioning of complex components over the whole life on the plant. Particular concern for reliability is directed to the primary pressure boundary. This report focuses on the portion of the primary system exposed to and significantly affected by neutron radiation. Experimental evidence from research programmes and from reactor surveillance programmes has indicated radiation embrittlement of a magnitude sufficient to raise doubts about reactor pressure vessel integrity. The crucial nature of the primary vessel function heightens the need to be alert to this problem, to which, fortunately, there are positive aspects: for example, steels have been developed which are relatively immune to radiation embrittlement. Further, awareness of such embrittlement has led to designs which can accomodate this factor. The nature of nuclear reactors, of the steels used in their construction, and of the procedures for interpreting embrittlement and minimizing the effects are reviewed with reference to the reactors that are expected to play a major role in electric power production from now to about the turn of the century. The report is intended as a manual or guidebook; the aim has been to make each chapter or major sub-division sufficiently comprehensive and self-contained for it to be understood and read independently of the rest of the book. At the same time, it is hoped that the whole is unified enough to make a complete reading useful and interesting to the several classes of reader that are involved with only specific aspects of the topic

Comparative study for the estimation of To shift due to irradiation embrittlement

International Nuclear Information System (INIS)

Lee, Jin Ho; Park, Youn won; Choi, Young Hwan; Kim, Seok Hun; Revka, Volodymyr

2002-01-01

Recently, an approach called the 'Master Curve' method was proposed which has opened a new means to acquire a directly measured material-specific fracture toughness curve. For the entire application of the Master Curve method, several technical issues should be solved. One of them is to utilize existing Charpy impact test data in the evaluation of a fracture transition temperature shift due to irradiation damage. In the U.S. and most Western countries, the Charpy impact test data have been used to estimate the irradiation effects on fracture toughness changes of RPV materials. For the determination of the irradiation shift the indexing energy level of 41 joule is used irrespective of the material yield strength. The Russian Code also requires the Charpy impact test data to determine the extent of radiation embrittlement. Unlike the U.S. Code, however, the Russian approach uses the indexing energy level varying according to the material strength. The objective of this study is to determine a method by which the reference transition temperature shift (ΔT o ) due to irradiation can be estimated. By comparing the irradiation shift estimated according to the U.S. procedure (ΔT 41J ) with that estimated according to the Russian procedure (ΔT F ), it was found that one-to-one relation exists between ΔT o and ΔT F

Investigations of low-temperature neutron embrittlement of ferritic steels

International Nuclear Information System (INIS)

Farrell, K.; Mahmood, S.T.; Stoller, R.E.; Mansur, L.K.

1992-01-01

Investigations were made into reasons for accelerated embrittlement of surveillance specimens of ferritic steels irradiated at 50C at the High Flux Isotope Reactor (HFIR) pressure vessel. Major suspects for the precocious embrittlement were a highly thermalized neutron spectrum,a low displacement rate, and the impurities boron and copper. None of these were found guilty. A dosimetry measurement shows that the spectrum at a major surveillance site is not thermalized. A new model of matrix hardening due to point defect clusters indicates little effect of displacement rate at low irradiation temperature. Boron levels are measured at 1 wt ppM or less, inadequate for embrittlement. Copper at 0.3 wt % and nickel at 0.7 wt % are shown to promote radiation strengthening in iron binary alloys irradiated at 50 to 60C, but no dependence on copper and nickel was found in steels with 0.05 to 0.22% Cu and 0.07 to 3.3% Ni. It is argued that copper impurity is not responsible for the accelerated embrittlement of the HFIR surveillance specimens. The dosimetry experiment has revealed the possibility that the fast fluence for the surveillance specimens may be underestimated because the stainless steel monitors in the surveillance packages do not record an unexpected component of neutrons in the spectrum at energies just below their measurement thresholds of 2 to 3 MeV

Irradiation Embrittlement Monitoring Programs of RPV's in the Slovak Republic NPP's

International Nuclear Information System (INIS)

Kupca, Ludovik

2006-01-01

Four types of surveillance programs were (are) realized in Slovak NPP's: 'Standard Surveillance Specimen Program' (SSSP) was finished in Jaslovske Bohunice V-2 Nuclear Power Plant (NPP) Units 3 and 4, 'Extended Surveillance Specimen Program' (ESSP), was prepared for Jaslovske Bohunice NPP V-2 with aim to validate the SSSP results, For the Mochovce NPP Unit 1 and 2 was prepared completely new surveillance program 'Modern Surveillance Specimen Program' (MSSP), based on the philosophy that the results of MSSP must be available during all NPP service life, For the Bohunice V-1 NPP was finished 'New Surveillance Specimen Program' (NSSP) coordinated by IAEA, which gave arguments for prolongation of service life these units for minimum 20 years, New Advanced Surveillance Specimen Program (ASSP) for Bohunice V-2 NPP (units 3 and 4) and Mochovce NPP (units 1, 2) is approved now. ASSP is dealing with the irradiation embrittlement of heat affected zone (HAZ) and RPV's austenitic cladding, which were not evaluated till this time in surveillance programs. SSSP started in 1979 and was finished in 1990. ESSP program started in 1995 and will be finished in 2007, was prepared with aim of: increasing of neutron fluence measurement accuracy, substantial improvement the irradiation temperature measurement, fixed orientation of samples to the centre of the reactor core, minimum differences of neutron dose for all the Charpy-V notch and COD specimens, the dose rate effect evaluation. In the year 1996 was started the new surveillance specimen program for the Mochovce RPV's unit-1 and 2, based on the fundamental postulate - to provide the irradiation embrittlement monitoring till the end of units operation. The 'New Surveillance Specimen Program' (NSSP) prepared in the year 1999 for the Bohunice V-1 NPP was finished in the year 2004. Main goal of this program was to evaluate the weld material properties degradation due to the irradiation and recovery efficiency by annealing too. The

Neutron-irradiation + helium hardening and embrittlement modeling of 9% Cr-steels in an engineering perspective (HELENA)

International Nuclear Information System (INIS)

Chaouadi, Rachid

2008-01-01

This report provides a physically-based engineering model to estimate the radiation hardening of 9%Cr-steels under both displacement damage (dpa) and helium. The model is essentially based on the dispersed barrier hardening theory and the dynamic re-solution of helium under displacement cascades. However, a number of assumptions and simplifications were considered to obtain a simple description of irradiation hardening and embrittlement primarily relying on the available experimental data. As a result, two components were basically identified, the dpa component that can be associated with black dots and small loops and the He-component accounting for helium bubbles. The dpa component is strongly dependent on the irradiation temperature and its dependence law was based on a first-order annealing kinetics. The damage accumulation law was also modified to take saturation into account. Finally, the global kinetics of the damage accumulation kept defined, its amplitude is fitted to one experimental condition. The model was rationalized on an experimental database that mainly consists of ∝9%Cr-steels irradiated in the technologically important temperature range of 50 to 600 C up do 50 dpa and with a He-content up to ∝5000 appm, including neutron and proton irradiation as well as implantation. The test temperature effect is taken into account through a normalization procedure based on the change of the Young's modulus and the anelastic deformation that occurs at high temperature. Finally, the hardening-to-embrittlement correlation is obtained using the load diagram approach. Despite the large experimental scatter, inherent to the variety of the materials and irradiation as well as testing conditions, the obtained results are very promising. Improvement of the model performance is still possible by including He-hardening saturation and high temperature softening but unfortunately, at this stage, a number of conflicting experimental data reported in literature should

Neutron-irradiation + helium hardening and embrittlement modeling of 9% Cr-steels in an engineering perspective (HELENA)

Energy Technology Data Exchange (ETDEWEB)

Chaouadi, Rachid

2008-07-01

This report provides a physically-based engineering model to estimate the radiation hardening of 9%Cr-steels under both displacement damage (dpa) and helium. The model is essentially based on the dispersed barrier hardening theory and the dynamic re-solution of helium under displacement cascades. However, a number of assumptions and simplifications were considered to obtain a simple description of irradiation hardening and embrittlement primarily relying on the available experimental data. As a result, two components were basically identified, the dpa component that can be associated with black dots and small loops and the He-component accounting for helium bubbles. The dpa component is strongly dependent on the irradiation temperature and its dependence law was based on a first-order annealing kinetics. The damage accumulation law was also modified to take saturation into account. Finally, the global kinetics of the damage accumulation kept defined, its amplitude is fitted to one experimental condition. The model was rationalized on an experimental database that mainly consists of {proportional_to}9%Cr-steels irradiated in the technologically important temperature range of 50 to 600 C up do 50 dpa and with a He-content up to {proportional_to}5000 appm, including neutron and proton irradiation as well as implantation. The test temperature effect is taken into account through a normalization procedure based on the change of the Young's modulus and the anelastic deformation that occurs at high temperature. Finally, the hardening-to-embrittlement correlation is obtained using the load diagram approach. Despite the large experimental scatter, inherent to the variety of the materials and irradiation as well as testing conditions, the obtained results are very promising. Improvement of the model performance is still possible by including He-hardening saturation and high temperature softening but unfortunately, at this stage, a number of conflicting experimental data

The role of phosphorus in the irradiation embrittlement of PWR pressure vessel steels

International Nuclear Information System (INIS)

Jones, R.B.; Buswell, J.T.

1987-02-01

An analysis has been performed of the influence of phosphorus on post-irradiation materials properties and microstructures determined on a variety of PWR steels and variants following exposure to MTR or reactor surveillance irradiations to doses not exceeding 7 x 10 19 n.cm -2 (E>1.0MeV) at 250-290 0 C. The irradiation-induced shifts in impact transition temperature, matrix hardening and the relative small angle neutron scattering response were found to rise most rapidly with increasing phosphorus when the copper content of the steel was 0.03 w/o. The sensitivity of the changes in mechanical properties to phosphorus content decreased as the copper content was increased. At copper levels typical of modern PWR steel manufacture (Cu 3 P) produced by the irradiation induced segregation of phosphorus to defect sinks and the depletion of phosphorus in solid solution as detected by high sensitivity electron microscopy and other analytical techniques. At higher levels of copper (approx. 0.3 w/o) the effect of phosphorus on properties was reduced by a factor of three due to the observed incorporation of phosphorus into the small copper precipitates formed during irradiation. Grain boundary embrittlement by phosphorus under irradiation is not thought to be important but further evidence concerning the post-irradiation fracture mode and the development of the deleterious influence of phosphorus with irradiation dose is required for a comprehensive understanding of its action. Some suggestions for future work are made. (author)

Helium embrittlement model and program plan for weldability of ITER materials

International Nuclear Information System (INIS)

Louthan, M.R. Jr.; Kanne, W.R. Jr.; Tosten, M.H.; Rankin, D.T.; Cross, B.J.

1997-02-01

This report presents a refined model of how helium embrittles irradiated stainless steel during welding. The model was developed based on experimental observations drawn from experience at the Savannah River Site and from an extensive literature search. The model shows how helium content, stress, and temperature interact to produce embrittlement. The model takes into account defect structure, time, and gradients in stress, temperature and composition. The report also proposes an experimental program based on the refined helium embrittlement model. A parametric study of the effect of initial defect density on the resulting helium bubble distribution and weldability of tritium aged material is proposed to demonstrate the roll that defects play in embrittlement. This study should include samples charged using vastly different aging times to obtain equivalent helium contents. Additionally, studies to establish the minimal sample thickness and size are needed for extrapolation to real structural materials. The results of these studies should provide a technical basis for the use of tritium aged materials to predict the weldability of irradiated structures. Use of tritium charged and aged material would provide a cost effective approach to developing weld repair techniques for ITER components

Assessment of the French and US embrittlement trend curves applied to RPV materials irradiated in the BR2 materials test reactor

International Nuclear Information System (INIS)

Chaouadi, R.; Gerard, R.; Boagaerts, A.S.

2011-01-01

The irradiation embrittlement of reactor pressure vessels (RPVs) in monitored through the surveillance programs associated with predictive formulas, the so-called embrittlement trend curves. These formulas are generally empirically derived and contain the major embrittlement-inducing elements such as copper, nickel and phosphorus. There are a number of such trend curves used in various regulatory guides used in the US, France, Germany, Russia and Japan. These trend curves are often supported by surveillance data and regularly assessed in view of updated surveillance databases. With the recent worldwide move towards life extension of existing reactors above their initially-scheduled lifetime of 40 years, adequate and accurate modeling of irradiation embrittlement becomes a concern for long term operation. The aim of this work is to assess the performance of the embrittlement trend curves used in a regulatory perspective. The work presented here is limited to US and French trend curves because the reactor pressure vessels of the Belgian nuclear power plants are either Westinghouse or Framatome design. The chemical composition of the Belgian RPVs being very close to the one of the French 900 MW units, the French trend curve is used except for the Doel 1-2 units for which these curves are not applicable due to the higher copper content of the welds. In this case, the U.S. trend curves are used. The aim of this work is to evaluate the performance of the embrittlement trend curves used in a regulatory perspective to represent the experimental data obtained in the BR2 reactor. In particular, the French (FIM, FIS) and the US (Reg. Guide 1.99 Rev. 2, ASTM E900-02, EWO and EONY) formulas are of prime interest. The results obtained clearly show that the French trend curves tend to over-estimate the actual irradiation hardening while the US curves under-estimate it. Within the long term operation perspective, both over- and under-estimating are undesirable and therefore the

Embrittlement of the nuclear icebreaker Lenin reactor pressure vessel materials reconstruction

International Nuclear Information System (INIS)

Krasikov, E.A.; Nikolaenko, V.A.

2008-01-01

Paper deals with the results of the efforts to examine the radiation damage of the Lenin nuclear-powered ice-breaker decommissioned reactor pressure vessel on the basis of which one has determined the peculiar features of the metal radiation embrittlement. Under 10 10 -10 11 s -1 cm -2 low density neutron flux irradiation one notes the most intensive embrittlement of the metal. Then, as the noxious element content in the metal matrix grows smaller the embrittlement reduces up to the change of sign as to the normal curve plotted at the neutron flux density exceeding 10 13 s -1 cm -2 . One assumes that as a result of the low density neutron flux irradiation the reactor pressure vessel edge spaces at some operation stages may be damaged more severely in contrast to these near the reactor core. The neutron irradiation density is the factor affecting the reactor vessel material embrittlement, that is why, it is important to study the damage mechanism of the materials of the power reactor vessels under design characterized by the low radiation load. The mentioned is important, as well, to evaluate the efficiency of the efforts undertaken to mitigate the effect of the neutron radiation on the reactor vessel [ru

Irradiation embrittlement and mitigation. V. 1. Working material. Proceedings of a specialists meeting held in Espoo, Finland 23-26 October 1995

International Nuclear Information System (INIS)

1995-01-01

The purpose of the meeting was to provide an international forum for discussion on recent results in research and utility experience on radiation damage and its surveillance, annealing and re-embrittlement of PWR, WWER and BWR reactor pressure vessel materials. The scope included: mechanism of radiation damage; effects of operating parameters (flux, temperature, time, etc.); results from surveillance programmes and their analysis; fracture mechanics testing and evaluation; annealing and optimization of the process; re-embrittlement after annealing. Presentations were aimed at better understanding of radiation damage, annealing and re-irradiation behaviour of reactor pressure vessels materials, at providing guidance and recommendations for optimization of annealing and surveillance programmes and directions for further investigations. Refs, figs and tabs

Irradiation embrittlement and mitigation. V. 1. Working material. Proceedings of a specialists meeting held in Espoo, Finland 23-26 October 1995

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-12-31

The purpose of the meeting was to provide an international forum for discussion on recent results in research and utility experience on radiation damage and its surveillance, annealing and re-embrittlement of PWR, WWER and BWR reactor pressure vessel materials. The scope included: mechanism of radiation damage; effects of operating parameters (flux, temperature, time, etc.); results from surveillance programmes and their analysis; fracture mechanics testing and evaluation; annealing and optimization of the process; re-embrittlement after annealing; Presentations were aimed at better understanding of radiation damage, annealing and re-irradiation behaviour of reactor pressure vessels materials, at providing guidance and recommendations for optimization of annealing and surveillance programmes and directions for further investigations. Refs, figs and tabs.

Radiation embrittlement behavior of fine-grained molybdenum alloy with 0.2 wt%TiC addition

Energy Technology Data Exchange (ETDEWEB)

Kitsunai, Y. [Tohoku University (Japan); Kurishita, H. [International Research Center for Nuclear Materials Science, Institute for Materials research (IMR), Tohoku University, Oarai, Ibaraki 311-1313 (Japan)]. E-mail: kurishi@imr.tohoku.ac.jp; Kuwabara, T. [Tohoku University (Japan); Narui, M. [International Research Center for Nuclear Materials Science, Institute for Materials research (IMR), Tohoku University, Oarai, Ibaraki 311-1313 (Japan); Hasegawa, M. [International Research Center for Nuclear Materials Science, Institute for Materials research (IMR), Tohoku University, Oarai, Ibaraki 311-1313 (Japan); Takida, T. [A.L.M.T. TECH Inc., 2 Iwasekoshi-machi, Toyama 931-8543 (Japan); Takebe, K. [A.L.M.T. TECH Inc., 2 Iwasekoshi-machi, Toyama 931-8543 (Japan)

2005-11-15

In order to elucidate the effects of pre-irradiation microstructures and irradiation conditions on radiation embrittlement and radiation-induced ductilization (RIDU), fine-grained Mo-0.2 wt%TiC specimens with high and low reduction rates in plastic working, which are designated as MTC-02H and MTC-02L, respectively, were prepared by powder metallurgical methods. The specimens were neutron irradiated to 0.1-0.15 dpa with controlled 1-cycle and 4-cycle heating between 573 and 773 K, and 473 and 673 K, respectively, in JMTR. Vickers microhardness and three-point bending impact tests and TEM microstructural examinations were made. The degree of radiation embrittlement, assessed by DBTT shift due to irradiation, was strongly dependent on the reduction rate and cycle number. The 4-cycle irradiation suppressed the radiation embrittlement compared with the 1-cycle irradiation, and the suppression was much more significant in MTC-02L than in MTC-02H. The observed behavior is discussed in connection with RIDU and microstructural evolution caused by the 4-cycle irradiation.

Low temperature radiation embrittlement for reactor vessel steels

International Nuclear Information System (INIS)

Ginding, I.A.; Chirkina, L.A.

1978-01-01

General conceptions of cold brittleness of bcc metals are in a review. Considered are experimental data and theoretical representations about the effect of irradiation conditions, chemical composition, phase and structural constitutions, grain size, mechanical and thermomechanical treatments on low-temperature irradiation embrittlement of reactor vessel steels. Presented are the methods for increasing radiation stability of metals (carbon and Cr-Mo steels) used in manufacturing reactor vessels

Neutron irradiation embrittlement of reactor pressure vessel steel 20 MnMoNi55 weld

International Nuclear Information System (INIS)

Ghoneim, M.M.

1987-05-01

The effect of neutron irradiation on the mechanical and fracture properties of an 'improved' 20 MnMoNi 55 Pressure Vessel Steel (PVS) weld was investigated. In addition to very low residual element content, especially Cu (0.035 wt.%), and relatively higher Ni content (0.9 wt.%), this steel has higher strength (30% more) than the steels used currently in nuclear reactor pressure vessels. The material was irradiated to 3.5x10 19 and 7x10 19 n/cm 2 (E > 1 Mev) at 290 0 C and 2.5x10 19 n/cm 2 (E > 1 MeV) at 160 0 C in FRJ-1 and FRJ-2 research reactors at KFA, Juelich, F.R.G. Test methods used in the evaluation included instrumented impact testing of standard and precracked Charpy specimens, tensile, and fracture toughness testing. Instrumented impact testing provided load and energy vs. time (deflection) data in addition to energy absorption data. The results indicated that the investigated high strength improved steel is more resistant to irradiation induced embrittlement than conventional PVSs. (orig./IHOE)

Effect of neutron irradiation on vanadium alloys

International Nuclear Information System (INIS)

Braski, D.N.

1986-01-01

Neutron-irradiated vanadium alloys were evaluated for their susceptibility to irradiation hardening, helium embrittlement, swelling, and residual radioactivity, and the results were compared with those for the austenitic and ferritic stainless steels. The VANSTAR-7 and V-15Cr-5Ti alloys showed the greatest hardening between 400 and 600 0 C while V-3Ti-1Si and V-20Ti had lower values that were comparable to those of ferritic steels. The V-15Cr-5Ti and VANSTAR-7 alloys were susceptible to helium embrittlement caused by the combination of weakened grain boundaries and irradiation-hardened grain matrices. Specimen fractures were entirely intergranular in the most severe instances of embrittlement. The V-3Ti-1Si and V-20Ti alloys were more resistant to helium embrittlement. Except for VANSTAR-7 irradiated to 40 dpa at 520 0 C, all of the vanadium alloys exhibited low swelling that was similar to the ferritic steels. Swelling was greater in specimens that were preimplanted with helium using the tritium trick. The vanadium alloys clearly exhibit lower residual radioactivity after irradiation than the ferrous alloys

Effect of neutron irradiation on vanadium alloys

Energy Technology Data Exchange (ETDEWEB)

Braski, D.N.

1986-01-01

Neutron-irradiated vanadium alloys were evaluated for their susceptibility to irradiation hardening, helium embrittlement, swelling, and residual radioactivity, and the results were compared with those for the austenitic and ferritic stainless steels. The VANSTAR-7 and V-15Cr-5Ti alloys showed the greatest hardening between 400 and 600/sup 0/C while V-3Ti-1Si and V-20Ti had lower values that were comparable to those of ferritic steels. The V-15Cr-5Ti and VANSTAR-7 alloys were susceptible to helium embrittlement caused by the combination of weakened grain boundaries and irradiation-hardened grain matrices. Specimen fractures were entirely intergranular in the most severe instances of embrittlement. The V-3Ti-1Si and V-20Ti alloys were more resistant to helium embrittlement. Except for VANSTAR-7 irradiated to 40 dpa at 520/sup 0/C, all of the vanadium alloys exhibited low swelling that was similar to the ferritic steels. Swelling was greater in specimens that were preimplanted with helium using the tritium trick. The vanadium alloys clearly exhibit lower residual radioactivity after irradiation than the ferrous alloys.

Modeling copper precipitation hardening and embrittlement in a dilute Fe-0.3at.%Cu alloy under neutron irradiation

Science.gov (United States)

Bai, Xian-Ming; Ke, Huibin; Zhang, Yongfeng; Spencer, Benjamin W.

2017-11-01

Neutron irradiation in light water reactors can induce precipitation of nanometer sized Cu clusters in reactor pressure vessel steels. The Cu precipitates impede dislocation gliding, leading to an increase in yield strength (hardening) and an upward shift of ductile-to-brittle transition temperature (embrittlement). In this work, cluster dynamics modeling is used to model the entire Cu precipitation process (nucleation, growth, and coarsening) in a Fe-0.3at.%Cu alloy under neutron irradiation at 300°C based on the homogenous nucleation mechanism. The evolution of the Cu cluster number density and mean radius predicted by the modeling agrees well with experimental data reported in literature for the same alloy under the same irradiation conditions. The predicted precipitation kinetics is used as input for a dispersed barrier hardening model to correlate the microstructural evolution with the radiation hardening and embrittlement in this alloy. The predicted radiation hardening agrees well with the mechanical test results in the literature. Limitations of the model and areas for future improvement are also discussed in this work.

Relationship between irradiation hardening and embrittlement of pressure vessel steels

International Nuclear Information System (INIS)

Odette, G.R.; Lombrozo, P.M.; Wullaert, R.A.

1984-01-01

Based on a large body of test and power reactor data, empirical relationships between irradiation strengthening and embrittlement are derived. It is shown that the Charpy V-notch (C /SUB v/ ) 41-J indexed transition temperature increases and the upper-shelf energy decreases systematically with increases in the yield stress. The transition temperature shifts are related to two mechanisms: increases in the maximum temperature of elastic-cleavage fracture, and decreases in the slope of the C, energy versus test temperature curve associated with reductions in the upper-shelf energy. The cleavage shift contribution, which is usually dominant, can be predicted from the initial temperature of fracture at general yield and the change in ambient temperature static yield stress. In developing this simplified cleavage fracture model, it is shown that: (a) yield stress changes are independent of temperature and strain rate; (b) the increase in yield stress with decreasing temperature is independent of the strain rate, irradiation, and metallurgical state; and (c) the microcleavage fracture stress is independent of irradiation and temperature. A semi-empirical procedure for estimating the shift contribution due to upper-shelf energy decreases and the total temperature shift at 41 J, based on the observation of an approximately constant temperature interval of the transition regime, is proposed, along with a method for forecasting the entire irradiated C, curve

Studies on neutron irradiation effects of iron alloys and nickel-base heat resistant alloys

International Nuclear Information System (INIS)

Watanabe, Katsutoshi

1987-09-01

The present paper describes the results of neutron irradiation effects on iron alloys and nickel-base heat resistant alloys. As for the iron alloys, irradiation hardening and embrittlement were investigated using internal friction measurement, electron microscopy and tensile testings. The role of alloying elements was also investigated to understand the irradiation behavior of iron alloys. The essential factors affecting irradiation hardening and embrittlement were thus clarified. On the other hand, postirradiation tensile and creep properties were measured of Hastelloy X alloy. Irradiation behavior at elevated temperatures is discussed. (author)

Cladding embrittlement during postulated loss-of-coolant accidents.

Energy Technology Data Exchange (ETDEWEB)

Billone, M.; Yan, Y.; Burtseva, T.; Daum, R.; Nuclear Engineering Division

2008-07-31

The effect of fuel burnup on the embrittlement of various cladding alloys was examined with laboratory tests conducted under conditions relevant to loss-of-coolant accidents (LOCAs). The cladding materials tested were Zircaloy-4, Zircaloy-2, ZIRLO, M5, and E110. Tests were performed with specimens sectioned from as-fabricated cladding, from prehydrided (surrogate for high-burnup) cladding, and from high-burnup fuel rods which had been irradiated in commercial reactors. The tests were designed to determine for each cladding material the ductile-to-brittle transition as a function of steam oxidation temperature, weight gain due to oxidation, hydrogen content, pre-transient cladding thickness, and pre-transient corrosion-layer thickness. For short, defueled cladding specimens oxidized at 1000-1200 C, ring compression tests were performed to determine post-quench ductility at {le} 135 C. The effect of breakaway oxidation on embrittlement was also examined for short specimens oxidized at 800-1000 C. Among other findings, embrittlement was found to be sensitive to fabrication processes--especially surface finish--but insensitive to alloy constituents for these dilute zirconium alloys used as cladding materials. It was also demonstrated that burnup effects on embrittlement are largely due to hydrogen that is absorbed in the cladding during normal operation. Some tests were also performed with longer, fueled-and-pressurized cladding segments subjected to LOCA-relevant heating and cooling rates. Recommendations are given for types of tests that would identify LOCA conditions under which embrittlement would occur.

Irradiation embrittlement of reactor vessel steels

International Nuclear Information System (INIS)

Bros, J.

2000-01-01

From the historical decision of closing the Yankee Rowe NPP because of the uncertainties on the level of reactor pressure vessel neutron embrittlement, this paper reviews the technical-scientist bases of the degradation phenomena, and refers to the evolution of reactor pressure vessel radiation surveillance programs. (Author)

Radiation embrittlement of WWER-1000 reactor vessel steels

International Nuclear Information System (INIS)

Nikolaeva, A.V.; Nikolaev, Yu.A.; Kevorkyan, Yu.R.

2001-01-01

Results obtained on the blank samples of materials of the WWER-1000 vessels irradiated by low density neutron flux are discussed. Chemical composition of the materials is characterized by the low content of the impurities (copper and phosphorus) and high content of nickel. Dependence of the radiation embrittlement of the WWER-1000 vessel materials on metallurgic variables and damage dose is treated. The research showed that nickel largely enhanced the radiation embrittlement. New dependences for determination of the radiation embrittlement real rate of the WWER-1000 vessel materials and its conservative estimation were developed [ru

Updated embrittlement trend curve for reactor pressure vessel steels

International Nuclear Information System (INIS)

Kirk, M.; Santos, C.; Eason, E.; Wright, J.; Odette, G.R.

2003-01-01

The reactor pressure vessels of commercial nuclear power plants are subject to embrittlement due to exposure to high energy neutrons from the core. Irradiation embrittlement of RPV belt-line materials is currently evaluated using US Regulatory Guide 1.99 Revision 2 (RG 1.99 Rev 2), which presents methods for estimating the Charpy transition temperature shift (ΔT30) at 30 ft-lb (41 J) and the drop in Charpy upper shelf energy (ΔUSE). A more recent embrittlement model, based on a broader database and more recent research results, is presented in NUREG/CR-6551. The objective of this paper is to describe the most recent update to the embrittlement model in NUREG/CR-6551, based upon additional data and increased understanding of embrittlement mechanisms. The updated ΔT30 and USE models include fluence, copper, nickel, phosphorous content, and product form; the ΔT30 model also includes coolant temperature, irradiation time (or flux), and a long-time term. The models were developed using multi-variable surface fitting techniques, understanding of the ΔT30 mechanisms, and engineering judgment. The updated ΔT30 model reduces scatter significantly relative to RG 1.99 Rev 2 on the currently available database for plates, forgings, and welds. This updated embrittlement trend curve will form the basis of revision 3 to Regulatory Guide 1.99. (author)

Neutron irradiation effects in reactor pressure vessel steels and weldments. Working document

International Nuclear Information System (INIS)

1998-10-01

As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. A separate abstract was prepared for the introduction and for each of the eleven chapters, which are: 1. Reactor Pressure Vessel Design, 2. Reactor Pressure Materials, 3. WWER Pressure Vessels, 4. Determination of Mechanical Properties, 5. Neutron Exposure, 6. Methodology of Irradiation Experiments, 7. Effect of Irradiation on Mechanical Properties, 8. Mechanisms of Irradiation Embrittlement, 9. Modelling of Irradiation Damage, 10. Annealing of Irradiation Damage, 11. Safety Assessment using Surveillance Programmes and Data Bases

Status and task of the study on the hydrogen embrittlement of zirconium alloys

International Nuclear Information System (INIS)

Nagase, Fumihisa; Furuta, Teruo; Seino, Shun; Komatsu, Kazushi.

1995-08-01

As the burnup of the LWR fuel is extended, waterside corrosion and hydrogen pickup increase in the Zircaloy cladding. Hydrogen embrittlement of Zircaloy is one of the main factors which may limit the life of the fuel rod. This report presents a review on the hydrogen embrittlement of zirconium and its alloys including the irradiated materials. Research tasks for the reduction of ductility in the high burnup fuel cladding are also discussed. Many fundamental investigations have been performed on the hydrogen embrittlement of zirconium alloys. However, the embrittlement mechanism of the high burnup fuel cladding is complicated. Especially, a coupled effect of hydrides and radiation defects are expected to be pronounced with neutron dose increase. In order to evaluate the reduction of ductility of the higher burnup fuel cladding properly, it is necessary to investigate the coupled effect of these two factors by systematic examinations. (author) 64 refs

Effect of lead factors on the embrittlement of RPV SA-508 cl 3 steel

Energy Technology Data Exchange (ETDEWEB)

Kempf, Rodolfo, E-mail: kempf@cnea.gov.ar [CNEA, Unidad Actividad Combustibles Nucleares, División Caracterización, Avda. Gral Paz 1499, C.P.B1650KNA, San Martín, Buenos Aires (Argentina); Troiani, Horacio, E-mail: troiani@cab.cnea.gov.ar [Centro Atómico Bariloche (CNEA) e Instituto Balseiro (UNCU), CONICET, Av. Bustillo 9500, CP 8400, Rio Negro (Argentina); Fortis, Ana Maria, E-mail: fortis@cnea.gov.ar [CNEA, Departamento Estructura y Comportamiento, UNSAM, Avda. Gral Paz 1499, C.P.B1650KNA, San Martín, Buenos Aires (Argentina)

2013-03-15

This paper presents a project to study the effect of lead factors on the mechanical behaviour of the SA-508 type 3 Reactor Pressure Vessel (RPV) steel used in the reactor under construction Atucha II in Argentina. Charpy-V notch specimens of this steel were irradiated at the RA1 experimental reactor at a temperature of 275 °C with two lead factors (186 and 93). The neutron flux was 3.71 × 10{sup 15} n m{sup −2} s{sup −1} and 1.85 × 10{sup 15} n m{sup −2} s{sup −1} (E > 1 MeV) respectively. In both cases, the fluence was 6.6 × 10{sup 21} n m{sup −2}, which is equivalent to that received by the PHWR Atucha II RPV in 10 years of full power irradiation. The results of Charpy tests revealed significant embrittlement both in the ΔT = 14 °C and ΔT = 21 °C shifts of the ductile–brittle transition temperatures (DBTT) and in the reduction of the maximum energy absorbed. This result shows that the shift of the DBTT with a lead factor of 93 is larger than that obtained with a lead factor of 186. Then, the results of irradiation in experimental reactors (MTR) with high lead factors may not be conservative with respect to the actual RPV embrittlement.

Status of reactor pressure vessel embrittlement study in Japan

International Nuclear Information System (INIS)

Sasajima, H.

1997-01-01

Since the construction of Japanese first commercial nuclear power plant in 1966, 52 nuclear power plants have been commissioned in Japan to commercial operation. Japanese first nuclear power plant has now been service for 30 years and the aging of nuclear power plants is steadily progressing in general. Under these circumstances, the Japan Power Engineering and Inspection Corporation (JAPEIC) is executing, under consignment by the Ministry of International Trade and Industry (MITI), the development and verification test programs for plant integrity evaluation technology by which nuclear power plant aging can be appropriately handled. This paper shows the outline of study dealing with embrittlement of RPV caused by neutron irradiation, as one of the activity of JAPEIC. The embrittlement of RPV caused by neutron irradiation is manifested as a shift of transition temperature and as a reduction in Upper Shelf Energy (USE). In JAPEIC, the study dealing with a shift of transition temperature was conducted in the ''Reactor Pressure Vessel Pressurized Thermal Shock Test Project (the PTS Project)'', and the study dealing with a reduction in USE has been conducted in the ''Nuclear Power Plant Life Management Technology (the PLIM Project)''. And the reconstitution technology of surveillance test specimen has been conducted in PLIM Project as one of the measures to improve monitoring above material characteristic changes. The integrity evaluation under the Pressurized Thermal Shock (PTS) events including the effect of neutron irradiation embrittlement was initiated in 1983 FY as the PTS Project and was completed in the 1991 FY. The study verified that plant integrity could be assured at not only the end of design life, but also an extended service life even when the severest PTS events were postulated. The PLIM Project, designed to develop and verify the integrity evaluation technology dealing with reduction of USE by neutron irradiation, was started in the 1996 FY as a 10

Effects of irradiation at low temperature on V-4Cr-4Ti

International Nuclear Information System (INIS)

Alexander, D.J.; Snead, L.L.; Zinkle, S.J.

1996-01-01

Irradiation at low temperatures (100 to 275 degrees C) to 0.5 dpa causes significant embrittlement and changes in the subsequent room temperature tensile properties of V-4Cr-4Ti. The yield strength and microhardness at room temperature increase with increasing irradiation temperature. The tensile flow properties at room temperature show large increases in strength and a complete loss of work hardening capacity with no uniform ductility. Embrittlement, as measured by an increase in the ductile-to-brittle transition temperature, increases with increasing irradiation temperature, at least up to 275 degrees C. This embrittlement is not due to pickup of O or other interstitial solutes during the irradiation

Effects of irradiation at low temperature on V-4Cr-4Ti

Energy Technology Data Exchange (ETDEWEB)

Alexander, D.J.; Snead, L.L.; Zinkle, S.J. [Oak Ridge National Lab., TN (United States)] [and others

1996-10-01

Irradiation at low temperatures (100 to 275{degrees}C) to 0.5 dpa causes significant embrittlement and changes in the subsequent room temperature tensile properties of V-4Cr-4Ti. The yield strength and microhardness at room temperature increase with increasing irradiation temperature. The tensile flow properties at room temperature show large increases in strength and a complete loss of work hardening capacity with no uniform ductility. Embrittlement, as measured by an increase in the ductile-to-brittle transition temperature, increases with increasing irradiation temperature, at least up to 275{degrees}C. This embrittlement is not due to pickup of O or other interstitial solutes during the irradiation.

Irradiation embrittlement and optimisation of annealing

International Nuclear Information System (INIS)

1993-01-01

This conference is composed of 30 papers grouped in 6 sessions related to the following themes: neutron irradiation effects in pressure vessel steels and weldments used in PWR, WWER and BWR nuclear plants; results from surveillance programmes (irradiation induced damage and annealing processes); studies on the influence of variations in irradiation conditions and mechanisms, and modelling; mitigation of irradiation effects, especially through thermal annealing; mechanical test procedures and specimen size effects

Irradiation embrittlement and optimisation of annealing

Energy Technology Data Exchange (ETDEWEB)

NONE

1994-12-31

This conference is composed of 30 papers grouped in 6 sessions related to the following themes: neutron irradiation effects in pressure vessel steels and weldments used in PWR, WWER and BWR nuclear plants; results from surveillance programmes (irradiation induced damage and annealing processes); studies on the influence of variations in irradiation conditions and mechanisms, and modelling; mitigation of irradiation effects, especially through thermal annealing; mechanical test procedures and specimen size effects.

U.S. NRC Embrittlement Data Base (EDB)

International Nuclear Information System (INIS)

Pace, J.V.; Rosseel, T.M.; Wang, J.A.

1999-01-01

Large amounts of data obtained from surveillance capsules and test reactor experiments are needed, comprising many different materials and different irradiation conditions, to develop generally applicable damage prediction models that can be used for industry standards and regulatory guides. Version 1 of the Embrittlement Data Base (EDB) [I] is such a comprehensive collection of such data resulting from the merging of the Power Reactor Embrittlement Data Base (PR-EDB) [2] and the Test Reactor Embrittlement Data Base (TR-EDB) [3]. Fracture toughness data were also integrated into Version 1 of the EDB. The EDB data files are in dBASE format and can be accessed with a personal computer using the DOS or WINDOWS operating system. A utility program has been written to investigate radiation embrittlement using this data base. The utility program is used to retrieve and select specific data, manipulate data, display data to the screen or printer, and to tit and plot Charpy impact data

PR-EDB: Power Reactor Embrittlement Database Version 3

International Nuclear Information System (INIS)

Wang, Jy-An John; Subramani, Ranjit

2008-01-01

The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industry standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. 'User-friendly' utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access for

PR-EDB: Power Reactor Embrittlement Database - Version 3

Energy Technology Data Exchange (ETDEWEB)

Wang, Jy-An John [ORNL; Subramani, Ranjit [ORNL

2008-03-01

The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industry standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. "User-friendly" utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access for

A study of the mechanical property changes of irradiation embrittled pressure vessel steels and their response to annealing treatments

International Nuclear Information System (INIS)

Tipping, P.; Waeber, W.B.; Mercier, O.

1991-01-01

Isochronal and isothermal heat treatments have been used to study the recovery of hardness of a neutron irradiated pressure vessel steel forging for the purposes of planning and realizing IAR (Irradiated-Annealed-Reirradiated) experiments. Charpy V notch tests have been performed to assess the toughness of the material irradiated to various fluences up to a maximum of 5 x 10 19 n/cm 2 , E>1 MeV at 290 o C with and without an intermediate annealing treatment at 450 o C x 168 h. The effect of the intermediate annealing was evident. The recovery of the upper shelf energies was strongly enhanced by a thermal ageing effect due to the annealing treatment for all fluence levels investigated compared to the irradiated condition. The transition temperature shifts exhibited a less straightforward behaviour due to the mentioned ageing effect which opposed the recovery process for this property leading to a net shift increase at lower and to a net recovery benefit at higher fluence levels. A phenomenological model description for the IAR embrittlement-recovery path is suggested. For this material and these irradiation conditions a plant life extension (PLEX) may be brought about if a specific annealing treatment is applied at a fluence level that is half the anticipated target fluence F for PLEX. In this case it was found that F>1.6 x 10 19 n/cm 2 . (author)

Evaluation of irradiation damage effect by applying electric properties based techniques

International Nuclear Information System (INIS)

Acosta, B.; Sevini, F.

2004-01-01

The most important effect of the degradation by radiation is the decrease in the ductility of the pressure vessel of the reactor (RPV) ferritic steels. The main way to determine the mechanical behaviour of the RPV steels is tensile and impact tests, from which the ductile to brittle transition temperature (DBTT) and its increase due to neutron irradiation can be calculated. These tests are destructive and regularly applied to surveillance specimens to assess the integrity of RPV. The possibility of applying validated non-destructive ageing monitoring techniques would however facilitate the surveillance of the materials that form the reactor vessel. The JRC-IE has developed two devices, focused on the measurement of the electrical properties to assess non-destructively the embrittlement state of materials. The first technique, called Seebeck and Thomson Effects on Aged Material (STEAM), is based on the measurement of the Seebeck coefficient, characteristic of the material and related to the microstructural changes induced by irradiation embrittlement. With the same aim the second technique, named Resistivity Effects on Aged Material (REAM), measures instead the resistivity of the material. The purpose of this research is to correlate the results of the impact tests, STEAM and REAM measurements with the change in the mechanical properties due to neutron irradiation. These results will make possible the improvement of such techniques based on the measurement of material electrical properties for their application to the irradiation embrittlement assessment

Embrittling effects of residual elements on steels

International Nuclear Information System (INIS)

Brear, J.M.; King, B.L.

1979-01-01

In a review of work related to reheat cracking in nuclear pressure vessel steels, Dhooge et al referred to work of the authors on the relative embrittling parameter for SA533B steels. The poor agreement when these parameters were applied to creep ductility data for SA508 class 2 lead the reviewers to conclude that the relative importance of impurity elements is a function of base alloy composition. The authors briefly describe some of their more recent work which demonstrates that when various mechanical, and other, effects are taken into consideration, the relative effects of the principal residual elements are similar, despite differing base compositions, and that the embrittling parameters derived correlate well with the data for SA Class 2 steel. (U.K.)

Effects of irradiation on mechanical properties

International Nuclear Information System (INIS)

Server, W.L.; Griesbach, T.J.; Dragunov, Y.; Amaev, A.

1998-01-01

As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. The effects of irradiation on the mechanical properties of reactor pressure vessel steels are explained. This chapter provides some background on the critical elements controlling neutron damage effects. Distinction is made between vessels made in the USA and in the former USSR

Neutron irradiation effects on the ductile-brittle transition of ferritic/martensitic steels

Energy Technology Data Exchange (ETDEWEB)

Klueh, R.L.; Alexander, D.J. [Oak Ridge National Lab., TN (United States)

1997-08-01

Ferritic/martensitic steels such as the conventional 9Cr-1MoVNb (Fe-9Cr-1Mo-0.25V-0.06Nb-0.1C) and 12Cr-1MoVW (Fe-12Cr-1Mo-0.25V-0.5W-0.5Ni-0.2C) steels have been considered potential structural materials for future fusion power plants. The major obstacle to their use is embrittlement caused by neutron irradiation. Observations on this irradiation embrittlement is reviewed. Below 425-450{degrees}C, neutron irradiation hardens the steels. Hardening reduces ductility, but the major effect is an increase in the ductile-brittle transition temperature (DBTT) and a decrease in the upper-shelf energy, as measured by a Charpy impact test. After irradiation, DBTT values can increase to well above room temperature, thus increasing the chances of brittle rather than ductile fracture.

Pressure vessel steels: influence of chemical composition on irradiation sensitivity

International Nuclear Information System (INIS)

Ghoniem, M.M.; Hammad, F.H.

1998-01-01

Neutron irradiation of the steels used in the construction of the nuclear reactor pressure vessels can lead to the embrittlement of these materials, increasing the ductile-to-brittle transition temperature and decreasing the fracture energy, which can limit the plant life. The knowledge of irradiation embrittlement and the means for minimizing such degradation is therefore important in the field of assuring the safety of the nuclear power plants. Irradiation embrittlement is quite a complex process. It involves many variables. The most important of these are irradiation temperature, neutron fluence (neutron dose), neutron flux (neutron dose rate), and chemical composition of the irradiated material. This paper is concerned with the effect of chemical composition, the role of residual and alloying elements in the irradiation embrittlement of nuclear reactor pressure vessel steels in light water reactors. It presents a critical review for the published work in this field through the last 25 years

Simulation of He embrittlement at grain boundaries in bcc transition metals

Energy Technology Data Exchange (ETDEWEB)

Suzudo, Tomoaki, E-mail: suzudo.tomoaki@jaea.go.jp; Yamaguchi, Masatake

2015-10-15

To investigate what atomic properties largely determine vulnerability to He embrittlement at grain boundaries (GB) of bcc metals, we introduce a computational model composed of first principles density functional theory and a He segregation rate theory model. Predictive calculations of He embrittlement at the first wall of the future DEMO fusion concept reactor indicate that variation in the He embrittlement originated not only from He production rate related to neutron irradiation, but also from the He segregation energy at the GB that has a systematic trend in the periodic table. - Highlights: • We modeled He grain boundary (GB) segregation of bcc transition metals using first-principles-based rate theory. • We established the quantitative relation between He embrittlement and He segregation using GB cohesive energy. • He embrittlement was strongly dependent on He segregation energy at the GB that has a systematic trend in the periodic table.

Simulation of He embrittlement at grain boundaries in bcc transition metals

International Nuclear Information System (INIS)

Suzudo, Tomoaki; Yamaguchi, Masatake

2015-01-01

To investigate what atomic properties largely determine vulnerability to He embrittlement at grain boundaries (GB) of bcc metals, we introduce a computational model composed of first principles density functional theory and a He segregation rate theory model. Predictive calculations of He embrittlement at the first wall of the future DEMO fusion concept reactor indicate that variation in the He embrittlement originated not only from He production rate related to neutron irradiation, but also from the He segregation energy at the GB that has a systematic trend in the periodic table. - Highlights: • We modeled He grain boundary (GB) segregation of bcc transition metals using first-principles-based rate theory. • We established the quantitative relation between He embrittlement and He segregation using GB cohesive energy. • He embrittlement was strongly dependent on He segregation energy at the GB that has a systematic trend in the periodic table.

Embrittlement of the Shippingport reactor shield tank

International Nuclear Information System (INIS)

Chopra, O.K.; Shack, W.J.

1989-01-01

Surveillance specimens from the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory showed an unexpectedly high degree of embrittlement relative to the data obtained on similar materials in Materials Testing Reactors (MTRs). The results suggest a possible negative flux effect and raise the issue of embrittlement of the pressure vessel support structures of commercial light water reactors. To help resolve this issues, a program was initiated to characterize the irradiation embrittlement of the neutron shield tank (NST) from the decommissioned Shippingport reactor. The Shippingport NST operated at 55 degree C (130 degree F) and was fabricated from A212 Grade B steel, similar to the vessel material in HFIR. The inner wall of the NST was exposed to a total maximum fluence of ∼ 6 x 10 17 n/cm 2 (E > 1 MeV) over a life of 9.25 effective full power years. This corresponds to a fast flux of 2.1 x 10 9 n/cm 2 x s and is comparable to the conditions for the HFIR surveillance specimens. The results indicate that irradiation increases the 15 ft x lb Charpy transition temperature (CTT) by ∼25 degree C (45 degree F) and decreases the upper shelf energy. The shift in CTT is not as severe as that observed in the HFIR surveillance specimens and is consistent with that expected from the MTR data base. However, the actual value of CTT is high, and the toughness at service temperature is low, even when compared with the HFIR data. The increase in yield stress is ∼50 MPa, which is comparable to the HFIR data. The results also indicate a lower impact strength and higher transition temperature for the TL orientation than that for the LT orientation. Some effects of the location across the thickness of the wall are also observed for the LT specimens; CTT is slightly greater for the specimens from the inner region of the wall

High temperature embrittlement of metals by helium

International Nuclear Information System (INIS)

Schroeder, H.

1983-01-01

The present knowledge of the influence of helium on the high temperature mechanical properties of metals to be used as structural materials in fast fission and in future fusion reactors is reviewed. A wealth of experimental data has been obtained by many different experimental techniques, on many different alloys, and on different properties. This review is mostly concentrated on the behaviour of austenitic alloys -especially austenitic stainless steels, for which the data base is by far the largest - and gives only a few examples of special bcc alloys. The effect of the helium embrittlement on the different properties - tensile, fatigue and, with special emphasis, creep - is demonstrated by representative results. A comparison between data obtained from in-pile (-beam) experiments and from post-irradiation (-implantation) experiments, respectively, is presented. Theoretical models to describe the observed phenomena are briefly outlined and some suggestions are made for future work to resolve uncertainties and differences between our experimental knowledge and theoretical understanding of high temperature helium embrittlement. (author)

Hardening and embrittlement mechanisms of reduced activation ferritic/martensitic steels irradiated at 573 K

Energy Technology Data Exchange (ETDEWEB)

Tanigawa, H. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Klueh, R.L. [Oak Ridge Noational Laboratory, TN (United States); Hashimoto, N. [Hokkaido Univ., Materials Science and Engineering Div., Graduate School of Engineering, Sapporo (Japan); Sokolov, M. [Oak Ridge National Laboratory, Materials Science and Technology Div., TN (United States)

2007-07-01

Full text of publication follows: It has been reported that reduced-activation ferritic/martensitic steels (RAFMs), such as F82H, ORNL9Cr-2WVTa, and JLF-1, showed a variety of changes in ductile-brittle transition temperature and yield stress after irradiation at 573 K up to 5 dpa, and those differences could not be interpreted solely by the difference of dislocation microstructure induced by irradiation. To investigate the impact of other microstructural feature, i.e. precipitates, the precipitation behavior of F82H, ORNL 9Cr-2WVTa, and JLF-1 was examined. It was revealed that irradiation-induced precipitation and amorphization of precipitates partly occurred and caused the different precipitation on block, packet and prior austenitic grain boundaries. In addition to these phenomena, irradiation-induced nano-size precipitates were also observed in the matrix. It was also revealed that the chemical compositions of precipitates approached the calculated thermal equilibrium state of M{sub 23}C{sub 6} at an irradiation temperature of 573 K. The calculation also suggests the presence of Laves phase at 573 K, which is usually not observed at this temperature, but the ion irradiation on aged F82H with Laves phase suggests that Laves phase becomes amorphous and could not be stable under irradiation at 573 K. This observation indicates the possibility that the irradiation-induced nano-size precipitation could be the consequence of the conflict between precipitation and amorphization of Laves phase. Over all, these observations suggests that the variety of embrittlement and hardening of RAFMs observed at 573 K irradiation up to 5 dpa might be the consequence of the transition phenomena that occur as the microstructure approaches thermal equilibrium during irradiation at 573 K. (authors)

Modification of the grain boundary microstructure of the austenitic PCA stainless steel to improve helium embrittlement resistance

International Nuclear Information System (INIS)

Maziasz, P.J.; Braski, D.N.

1986-01-01

Grain boundary MC precipitation was produced by a modified thermal-mechanical pretreatment in 25% cold worked (CW) austenitic prime candidate alloy (PCA) stainless steel prior to HFIR irradiation. Postirradiation tensile results and fracture analysis showed that the modified material (B3) resisted helium embrittlement better than either solution annealed (SA) or 25% CW PCA irradiated at 500 to 600 0 C to approx.21 dpa and 1370 at. ppM He. PCA SA and 25% CW were not embrittled at 300 to 400 0 C. Grain boundary MC survives in PCA-B3 during HFIR irradiation at 500 0 C but dissolves at 600 0 C; it does not form in either SA or 25% CW PCA during similar irradiation. The grain boundary MC appears to play an important role in the helium embrittlement resistance of PCA-B3

Controlling RPV embrittlement through wet annealing in support of life attainment and life extension decisions

International Nuclear Information System (INIS)

Krasikov, E. A.

2012-01-01

As a main barrier against radioactivity outlet reactor pressure vessel (RPV) is a key component in terms of Nuclear Power Plant (NPP) safety. Therefore present-day demands in RPV reliability enhance have to be met by all possible actions for RPV in-service embrittlement mitigation. Annealing treatment is known to be the effective measure to restore the RPV metal properties deteriorated by neutron irradiation. Low temperature 'wet' annealing at a maximum coolant temperature which can be obtained using the reactor core or primary circuit pumps, although it cannot be expected to produce complete recovery, is more attractive from the practical point of view especially in cases when the removal of the internals is impossible. As a rule there is no recovery effect up to annealing and irradiation temperature difference of 70 deg. C. It is known, however, that along with radiation embrittlement neutron irradiation may mitigate the radiation damage in metals. Therefore we have tried to test the possibility to use the effect of radiation-induced ductilization in 'wet' annealing technology by means of nuclear heat utilization as heat and neutron irradiation sources at once. In support of the above-mentioned conception the 3-year duration reactor experiment on 15Cr3NiMoV type steel with preliminary irradiation at operating Pressurized Water Reactor (PWR) at 270 deg. C and following extra irradiation (87 h at 330 deg. C) at IR-8 test reactor was fulfilled. In fact, embrittlement was partly suppressed up to value equivalent to 1,5 fold neutron fluence decrease. The degree of recovery in case of radiation enhanced annealing is equal to 27% whereas furnace annealing results in zero effect under existing conditions. Mechanism of the radiation-induced damage mitigation is proposed. It is hoped that 'wet' annealing technology will help provide a better management of the RPV degradation as a factor affecting the lifetime of nuclear power plants which, together with associated

Controlling RPV embrittlement through wet annealing in support of life attainment and life extension decisions

International Nuclear Information System (INIS)

Krasikov, E.A.

2012-01-01

As a main barrier against radioactivity outlet reactor pressure vessel (RPV) is a key component in terms of Nuclear Power Plant (NPP) safety. Therefore present-day demands in RPV reliability enhance have to be met by all possible actions for RPV in-service embrittlement mitigation. Annealing treatment is known to be the effective measure to restore the RPV metal properties deteriorated by neutron irradiation. Low temperature annealing at a maximum coolant temperature which can be obtained using the reactor core or primary circuit pumps, although it cannot be expected to produce complete recovery, is more attractive from the practical point of view especially in cases when the removal of the internals is impossible. As a rule there is no recovery effect up to annealing and irradiation temperature difference of 70 o C. It is known, however, that along with radiation embrittlement neutron irradiation may mitigate the radiation damage in metals. Therefore we have tried to test the possibility to use the effect of radiation-induced ductilization in annealing technology by means of nuclear heat utilization as heat and neutron irradiation sources at once. In support of the above-mentioned conception the 3-year duration reactor experiment on 15Cr3NiMoV type steel with preliminary irradiation at operating Pressurized Water Reactor (PWR) at 270 o C and following extra irradiation (87 h at 330 o C) at IR-8 test reactor was fulfilled. In fact, embrittlement was partly suppressed up to value equivalent to 1,5 fold neutron fluence decrease. The degree of recovery in case of radiation enhanced annealing is equal to 27% whereas furnace annealing results in zero effect under existing conditions. Mechanism of the radiation-induced damage mitigation is proposed. It is hoped that annealing technology will help provide a better management of the RPV degradation as a factor affecting the lifetime of nuclear power plants which, together with associated management methods, will help

Controlling RPV embrittlement through wet annealing in support of life attainment and life extension decisions

Energy Technology Data Exchange (ETDEWEB)

Krasikov, E. A. [National Research Centre Kurchatov Inst., 1, Kurchatov Sq., Moscow, 123182 (Russian Federation)

2012-07-01

As a main barrier against radioactivity outlet reactor pressure vessel (RPV) is a key component in terms of Nuclear Power Plant (NPP) safety. Therefore present-day demands in RPV reliability enhance have to be met by all possible actions for RPV in-service embrittlement mitigation. Annealing treatment is known to be the effective measure to restore the RPV metal properties deteriorated by neutron irradiation. Low temperature 'wet' annealing at a maximum coolant temperature which can be obtained using the reactor core or primary circuit pumps, although it cannot be expected to produce complete recovery, is more attractive from the practical point of view especially in cases when the removal of the internals is impossible. As a rule there is no recovery effect up to annealing and irradiation temperature difference of 70 deg. C. It is known, however, that along with radiation embrittlement neutron irradiation may mitigate the radiation damage in metals. Therefore we have tried to test the possibility to use the effect of radiation-induced ductilization in 'wet' annealing technology by means of nuclear heat utilization as heat and neutron irradiation sources at once. In support of the above-mentioned conception the 3-year duration reactor experiment on 15Cr3NiMoV type steel with preliminary irradiation at operating Pressurized Water Reactor (PWR) at 270 deg. C and following extra irradiation (87 h at 330 deg. C) at IR-8 test reactor was fulfilled. In fact, embrittlement was partly suppressed up to value equivalent to 1,5 fold neutron fluence decrease. The degree of recovery in case of radiation enhanced annealing is equal to 27% whereas furnace annealing results in zero effect under existing conditions. Mechanism of the radiation-induced damage mitigation is proposed. It is hoped that 'wet' annealing technology will help provide a better management of the RPV degradation as a factor affecting the lifetime of nuclear power plants which

Reactor pressure vessel embrittlement management through EPRI-Developed material property databases

International Nuclear Information System (INIS)

Rosinski, S.T.; Server, W.L.; Griesbach, T.J.

1997-01-01

Uncertainties and variability in U.S. reactor pressure vessel (RPV) material properties have caused the U.S. Nuclear Regulatory Commission (NRC) to request information from all nuclear utilities in order to assess the impact of these data scatter and uncertainties on compliance with existing regulatory criteria. Resolving the vessel material uncertainty issues requires compiling all available data into a single integrated database to develop a better understanding of irradiated material property behavior. EPRI has developed two comprehensive databases for utility implementation to compile and evaluate available material property and surveillance data. RPVDATA is a comprehensive reactor vessel materials database and data management program that combines data from many different sources into one common database. Searches of the data can be easily performed to identify plants with similar materials, sort through measured test results, compare the ''best-estimates'' for reported chemistries with licensing basis values, quantify variability in measured weld qualification and test data, identify relevant surveillance results for characterizing embrittlement trends, and resolve uncertainties in vessel material properties. PREP4 has been developed to assist utilities in evaluating existing unirradiated and irradiated data for plant surveillance materials; PREP4 evaluations can be used to assess the accuracy of new trend curve predictions. In addition, searches of the data can be easily performed to identify available Charpy shift and upper shelf data, review surveillance material chemistry and fabrication information, review general capsule irradiation information, and identify applicable source reference information. In support of utility evaluations to consider thermal annealing as a viable embrittlement management option, EPRI is also developing a database to evaluate material response to thermal annealing. Efforts are underway to develop an irradiation

Oak Ridge National Laboratory Embrittlement Data Base (EDB) and Dosimetry Evaluation (DE) program

International Nuclear Information System (INIS)

Pace, J.V. III; Remec, I.; Wang, J.A.; White, J.E.

1996-01-01

The objective of this program is to develop, maintain, and upgrade computerized data bases, calculational procedures, and standards relating to reactor pressure vessel fluence spectra determinations and embrittlement assessments. As part of this program, the information from radiation embrittlement research on nuclear reactor pressure vessel steels and from power reactor surveillance reports is maintained in a data base published on a periodic basis. The Embrittlement Data Base (EDB) effort consists of verifying the quality of the EDB, providing user-friendly software to access and process the data, and exploring and assessing embrittlement prediction models. The Dosimetry Evaluation effort consists of maintaining and upgrading validated neutron and gamma radiation transport procedures, maintaining cross-section libraries with the latest evaluated nuclear data, and maintaining and updating validated dosimetry procedures and data bases. The information available from this program provides data for assisting the Office of Nuclear Reactor Regulation, with support from the Office of Nuclear Regulatory Research, to effectively monitor current procedures and data bases used by vendors, utilities, and service laboratories in the pressure vessel irradiation surveillance program

Microstructure and embrittlement of VVER 440 reactor pressure vessel steels

International Nuclear Information System (INIS)

Hennion, A.

1999-03-01

27 VVER 440 pressurised water reactors operate in former Soviet Union and in Eastern Europe. The pressure vessel, is made of Cr-Mo-V steel. It contains a circumferential arc weld in front of the nuclear core. This weld undergoes a high neutron flux and contains large amounts of copper and phosphorus, elements well known for their embrittlement potency under irradiation. The embrittlement kinetic of the steel is accelerated, reducing the lifetime of the reactor. In order to get informations on the microstructure and mechanical properties of these steels, base metals, HAZ, and weld metals have been characterized. The high amount of phosphorus in weld metals promotes the reverse temper embrittlement that occurs during post-weld heat treatment. The radiation damage structure has been identified by small angle neutron scattering, atomic probe, and transmission electron microscopy. Nanometer-sized clusters of solute atoms, rich in copper with almost the same characteristics as in western pressure vessels steels, and an evolution of the size distribution of vanadium carbides, which are present on dislocation structure, are observed. These defects disappear during post-irradiation tempering. As in western steels, the embrittlement is due to both hardening and reduction of interphase cohesion. The radiation damage specificity of VVER steels arises from their high amount of phosphorus and from their significant density of fine vanadium carbides. (author)

Hydrogen Embrittlement

Science.gov (United States)

Woods, Stephen; Lee, Jonathan A.

2016-01-01

Hydrogen embrittlement (HE) is a process resulting in a decrease in the fracture toughness or ductility of a metal due to the presence of atomic hydrogen. In addition to pure hydrogen gas as a direct source for the absorption of atomic hydrogen, the damaging effect can manifest itself from other hydrogen-containing gas species such as hydrogen sulfide (H2S), hydrogen chloride (HCl), and hydrogen bromide (HBr) environments. It has been known that H2S environment may result in a much more severe condition of embrittlement than pure hydrogen gas (H2) for certain types of alloys at similar conditions of stress and gas pressure. The reduction of fracture loads can occur at levels well below the yield strength of the material. Hydrogen embrittlement is usually manifest in terms of singular sharp cracks, in contrast to the extensive branching observed for stress corrosion cracking. The initial crack openings and the local deformation associated with crack propagation may be so small that they are difficult to detect except in special nondestructive examinations. Cracks due to HE can grow rapidly with little macroscopic evidence of mechanical deformation in materials that are normally quite ductile. This Technical Memorandum presents a comprehensive review of experimental data for the effects of gaseous Hydrogen Environment Embrittlement (HEE) for several types of metallic materials. Common material screening methods are used to rate the hydrogen degradation of mechanical properties that occur while the material is under an applied stress and exposed to gaseous hydrogen as compared to air or helium, under slow strain rates (SSR) testing. Due to the simplicity and accelerated nature of these tests, the results expressed in terms of HEE index are not intended to necessarily represent true hydrogen service environment for long-term exposure, but rather to provide a practical approach for material screening, which is a useful concept to qualitatively evaluate the severity of

Blistering and hydride embrittlement

International Nuclear Information System (INIS)

Louthan, M.R. Jr.

1975-01-01

The effects of hydrogen on the mechanical properties of metals have been categorized into several groups. Two of the groups, hydrogen blistering and hydride embrittlement, are reasonably well understood, and problems relating to their occurrence may be avoided if that understanding is used as a basis for selecting alloys for hydrogen service. Blistering and hydride embrittlement are described along with several techniques of materials selection and used to minimize their adverse effects. (U.S.)

Irradiation, Annealing, and Reirradiation Effects on American and Russian Reactor Pressure Vessel Steels

International Nuclear Information System (INIS)

Chernobaeva, A.A.; Korolev, Y.N.; Nanstad, R.K.; Nikolaev, Y.A.; Sokolov, M.A.

1998-01-01

One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPVs) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. Even though a postirradiation anneal may be deemed successful, a critical aspect of continued RPV operation is the rate of embrittlement upon reirradiation. There are insufficient data available to allow for verification of available models of reirradiation embrittlement or for the development of a reliable predictive methodology. This is especially true in the case of fracture toughness data. Under the U.S.-Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS), Working Group 3 on Radiation Embrittlement, Structural Integrity, and Life Extension of Reactor Vessels and Supports agreed to conduct a comparative study of annealing and reirradiation effects on RPV steels. The Working Group agreed that each side would irradiate, anneal, reirradiate (if feasible ), and test two materials of the other. Charpy V-notch (CVN) and tensile specimens were included. Oak Ridge National Laboratory (ORNL) conducted such a program (irradiation and annealing, including static fracture toughness) with two weld metals representative of VVER-440 and VVER-1000 RPVs, while the Russian Research Center-Kurchatov Institute (RRC-KI) conducted a program (irradiation, annealing, reirradiation, and reannealing) with Heavy-Section Steel Technology (HSST) Program Plate 02 and Heavy-Section Steel Irradiation (HSSI) Program Weld 73W. The results for each material from each laboratory are compared with those from the other laboratory. The ORNL experiments with the VVER welds included irradiation to about 1 x 10 19 n/cm 2 (>1 MeV), while the RRC-KI experiments with the U.S. materials included irradiations from about 2 to 18 x 10 19 n/cm 2 (>l MeV). In both cases, irradiations were conducted at ∼290 C and annealing treatments were conducted at ∼454 C. The ORNL and RRC

Approach for estimating post-annual reirradiation embrittlement of reactor vessel steels

International Nuclear Information System (INIS)

Server, W.L.; Taboada, A.

1985-01-01

Thermal annealing of a commercial nuclear reactor pressure vessel is a possible solution for extending lifetime in situations where excessive radiation embrittlement has taken place or when the original design life is approached. Two difficult facets of thermal annealing are the degree of toughness recovery after annealing and the post-anneal reirradiation embrittlement behavior. These aspects of annealing are evaluated in this paper by using simple models and translation of the initial irradiation damage curve either vertically or laterally at the point of residual shift after annealing. Results using this methodology are compared to limited actual weld metal measurements of annealing behavior. A forthcoming ASTM Guide on in-place annealing uses this methodology to assess annealing recovery and re-embrittlement response

Effect of trapping and temperature on the hydrogen embrittlement susceptibility of alloy 718

Energy Technology Data Exchange (ETDEWEB)

Galliano, Florian; Andrieu, Eric; Blanc, Christine; Cloue, Jean-Marc; Connetable, Damien; Odemer, Gregory, E-mail: gregory.odemer@ensiacet.fr

2014-08-12

Ni-based alloy 718 is widely used to manufacture structural components in the aeronautic and nuclear industries. Numerous studies have shown that alloy 718 may be sensitive to hydrogen embrittlement. In the present study, the susceptibilities of three distinct metallurgical states of alloy 718 to hydrogen embrittlement were investigated to identify both the effect of hydrogen trapping on hydrogen embrittlement and the role of temperature in the hydrogen-trapping mechanism. Cathodic charging in a molten salt bath was used to saturate the different hydrogen traps of each metallurgical state. Tensile tests at different temperatures and different strain rates were carried out to study the effect of hydrogen on mechanical properties and failure modes, in combination with hydrogen content measurements. The results demonstrated that Ni-based superalloy 718 was strongly susceptible to hydrogen embrittlement between 25 °C and 300 °C, and highlighted the dominant roles played by the hydrogen solubility and the hydrogen trapping on mechanical behavior and fracture modes.

Heavy-Section Steel Irradiation Program on irradiation effects in light-water reactor pressure vessel materials

International Nuclear Information System (INIS)

Nanstad, R.K.; Corwin, W.R.; Alexander, D.J.; Haggag, F.M.; Iskander, S.K.; McCabe, D.E.; Sokolov, M.A.; Stoller, R.E.

1995-01-01

The safety of commercial light-water nuclear plants is highly dependent on the structural integrity of the reactor pressure vessel (RPV). In the absence of radiation damage to the RPV, fracture of the vessel is difficult to postulate. Exposure to high energy neutrons can result in embrittlement of radiation-sensitive RPV materials. The Heavy-Section Steel Irradiation (HSSI) Program at Oak Ridge National Laboratory, sponsored by the US Nuclear Regulatory Commission (USNRC), is assessing the effects of neutron irradiation on RPV material behavior, especially fracture toughness. The results of these and other studies are used by the USNRC in the evaluation of RPV integrity and regulation of overall nuclear plant safety. In assessing the effects of irradiation, prototypic RPV materials are characterized in the unirradiated condition and exposed to radiation under varying conditions. Mechanical property tests are conducted to provide data which can be used in the development of guidelines for structural integrity evaluations, while metallurgical examinations and mechanistic modeling are performed to improve understanding of the mechanisms responsible for embrittlement. The results of these investigations, in conjunction with results from commercial reactor surveillance programs, are used to develop a methodology for the prediction of radiation effects on RPV materials. This irradiation-induced degradation of the materials can be mitigated by thermal annealing, i.e., heating the RPV to a temperature above that of normal operation. Thus, thermal annealing and evaluation of reirradiation behavior are major tasks of the HSSI Program. This paper describes the HSSI Program activities by summarizing some past and recent results, as well as current and planned studies. 30 refs., 8 figs., 1 tab

Power Reactor Embrittlement Data Base

International Nuclear Information System (INIS)

Kam, F.B.K.; Stallmann, F.W.; Wang, J.A.

1990-01-01

Regulatory and research evaluations of embrittlement prediction models and of vessel integrity under load can be greatly expedited by the use of a well designed, computerized embrittlement data base. The Power Reactor Embrittlement Data Base (PR-EDB) is a comprehensive collection of data from surveillance reports and other published reports of commercial nuclear reactors. The uses of the data base require that as many different data as available are collected from as many sources as possible with complete references and that subsets of relevant data can be easily retrieved and processed. The objectives of this NRC-sponsored program are the following: (1) to compile and to verify the quality of the PR-EDB; (2) to provide user-friendly software to access and process the data; (3) to explore or confirm embrittlement prediction models; and (4) to interact with standards organizations to provide the technical bases for voluntary consensus standards that can be used in regulatory guides, standard review plans, and codes. To achieve these goals, the data base architecture was designed after much discussion and planning with prospective users, namely, material scientists and members of the research staff. The current compilation of the PR-EDB (Version 1) contains results from surveillance capsule reports of 78 reactors with 381 data points for 110 different irradiated base materials and 161 data points for 79 different welds. Results from heat-affected zone materials are also listed. The time and effort required to process and evaluate different types of data in the PR-EDB have been drastically reduced from previous data bases. The Electric Power Research Institute (EPRI), reactor vendors, and utilities are in the process of providing back-up quality assurance checks of PR-EDB and will be supplementing the data base with additional data and documentation

Thermal annealing of an embrittled reactor pressure vessel

International Nuclear Information System (INIS)

Mager, T.R.; Dragunov, Y.G.; Leitz, C.

1998-01-01

As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. Chapter 11 deals with thermal annealing of an embrittled reactor pressure vessel. Anneal procedures for vessels from both the US and the former USSR are mentioned schematically, wet anneals at lower temperature and dry anneals above RPV design temperatures are investigated. It is shown that heat treatment is a means of recovering mechanical properties which were degraded by neutron radiation exposure, thus assuring reactor pressure vessel compliance with regulatory requirements

Evaluation on the Effect of Composition on Radiation Hardening and Embrittlement in Model FeCrAl Alloys

Energy Technology Data Exchange (ETDEWEB)

Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Briggs, Samuel A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Edmondson, Philip [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hu, Xunxiang [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Littrell, Kenneth C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Parish, Chad M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-09-18

networks acting as defect sinks, resulting in variations in the observed microstructures after irradiation. Dose trends were also observed, with increasing radiation dose promoting changes in the size and number density of the Cr-rich α' precipitates. Based on the microstructural analysis, performed tensile testing, and prior knowledge from FeCr literature it was hypothesized that the formation of the Cr-rich α' precipitates could lead to significant radiation-induced embrittlement in the alloys, and this could be composition dependent, a result which would mirror the trends observed for radiation-induced hardening. Due to the limited database on embrittlement in the FeCrAl alloy class after irradiation, a series of radiation experiments have been implemented. The overarching point of view within this report is the radiation tolerance of FeCrAl is complex, with many mechanisms and factors to be considered at once. Further development of the FeCrAl alloy class for enhanced accident tolerant applications requires detailed, single (or at least limited) variable experiments to fully comprehend and predict the performance of this alloy in LWRs. This report has been submitted as fulfillment of milestone M2FT-15OR0202321 titled, Summary report on the effect of composition on the irradiation embrittlement of Gen 1 ATF FeCrAl for the Department of Energy Office of Nuclear Energy, Advanced Fuel Campaign of the Fuel Cycle R&D program.

Empirical correlation between mechanical and physical parameters of irradiated pressure vessel steels

International Nuclear Information System (INIS)

Tipping, P.; Solt, G.; Waeber, W.

1991-02-01

Neutron irradiation embrittlement of nuclear reactor pressure vessel (PV) steels is one of the best known ageing factors of nuclear power plants. If the safety limits set by the regulators for the PV steel are not satisfied any more, and other measures are too expensive for the economics of the plant, this embrittlement could lead to the closure of the plant. Despite this, the fundamental mechanisms of neutron embrittlement are not yet fully understood, and usually only empirical mathematical models exist to asses neutron fluence effects on embrittlement, as given by the Charpy test for example. In this report, results of a systematic study of a French forging (1.2 MD 07 B), irradiated to several fluences will be reported. Mechanical property measurements (Charpy tensile and Vickers microhardness), and physical property measurements (small angle neutron scattering - SANS), have been done on specimens having the same irradiation or irradiation-annealing-reirradiation treatment histories. Empirical correlations have been established between the temperature shift and the decrease in the upper shelf energy as measured on Charpy specimens and tensile stresses and hardness increases on the one hand, and the size of the copper-rich precipitates formed by the irradiation on the other hand. The effect of copper (as an impurity element) in enhancing the degradation of mechanical properties has been demonstrated; the SANS measurements have shown that the size and amount of precipitates are important. The correlations represent the first step in an effort to develop a description of neutron irradiation induced embrittlement which is based on physical models. (author) 6 figs., 27 refs

Different approaches to estimation of reactor pressure vessel material embrittlement

Directory of Open Access Journals (Sweden)

V. M. Revka

2013-03-01

Full Text Available The surveillance test data for the nuclear power plant which is under operation in Ukraine have been used to estimate WWER-1000 reactor pressure vessel (RPV material embrittlement. The beltline materials (base and weld metal were characterized using Charpy impact and fracture toughness test methods. The fracture toughness test data were analyzed according to the standard ASTM 1921-05. The pre-cracked Charpy specimens were tested to estimate a shift of reference temperature T0 due to neutron irradiation. The maximum shift of reference temperature T0 is 84 °C. A radiation embrittlement rate AF for the RPV material was estimated using fracture toughness test data. In addition the AF factor based on the Charpy curve shift (ΔTF has been evaluated. A comparison of the AF values estimated according to different approaches has shown there is a good agreement between the radiation shift of Charpy impact and fracture toughness curves for weld metal with high nickel content (1,88 % wt. Therefore Charpy impact test data can be successfully applied to estimate the fracture toughness curve shift and therefore embrittlement rate. Furthermore it was revealed that radiation embrittlement rate for weld metal is higher than predicted by a design relationship. The enhanced embrittlement is most probably related to simultaneously high nickel and high manganese content in weld metal.

Radiation hardening and embrittlement of some refractory metals and alloys

International Nuclear Information System (INIS)

Fabritsiev, S.; Pokrovskyb

2007-01-01

Tungsten is proposed for application in the ITER divertor and limiter as plasma facing material. The tungsten operation temperature in the ITER divertor is relatively high. Hence, the ductile properties of tungsten will be controlled by the low temperature radiation embrittlement. The mechanism of radiation hardening and embrittlement under neutron irradiation at low temperature is well studied for FCC metals, in particular for copper. At the same time, low-temperature radiation hardening of BCC materials, in particular for refractory metals, is less studied. This study presents the results of investigation into radiation hardening and embrittlement of pure metals: W, Mo and Nb, and W-Re and Ta-4W alloys. The materials were in the annealed conditions. The specimens were irradiated in the SM-2 reactor to doses of 10 -4 -10 -1 dpa at 80 C and then tested for tension at 80 C. The study of the stress-strain curves of unirradiated specimens revealed a yield drop for W, Mo, Nb, Ta-4W, W-Re. After the yield drop some metals (Mo,Nb) retain their capability for strain hardening and demonstrate a high elongation (20-50%). Radiation hardening is maximum in Mo (∝400MPa) and minimum in Nb (∝100 MPa). In this case the dependence slope for Nb is similar to that for pure copper irradiated in SM-2 under the same conditions. Ii and Ta-4W have a higher slope. Measurement of electrical resistivity of irradiated specimens showed that for all materials it is increased monotonously with an increase in the irradiation dose. A minimum gain in electrical resistivity with a dose was observed for Nb (∝3% at 0.1 dpa). As for Mo it was essentially higher, i.e. ∝ 30%. The gain was maximum for W-Re alloy. Comparison of radiation hardening dose dependencies obtained in this study with the data for FCC metals (Cu) showed that in spite of the quantitative difference the qualitative behavior of these two classes of metals is similar. (orig.)

TAREG 2.01/00 Project, ''Validation of neutron embrittlement for VVER 1000 and 440/213 RPVs, with emphasis on integrity assessment''

International Nuclear Information System (INIS)

Ahlstrand, R.; Margolin, B.; Kostylev, V.; Yurchenko, E.; Akbashev, I.; Piminov, V.; Nikolaev, Y.; Koshkin, V.; Kharshenko, V.; Chyrko, L.; Bukhanov, V.; Comsa, O.

2012-01-01

The irradiation embrittlement and integrity of the VVER reactors has been an important issue in many EC supported TACIS and PHARE projects since 1990. In the EC annual program 2000 two TACIS projects (TAREG 2.01/00 and 2.01/03) were approved on the issue in order to improve the neutron irradiation embrittlement databases, elaborate new trend curves for the embrittlement and to assess the integrity of the RPVs (Reactor Pressure Vessel) by analysing PTS transients (Pressurized Thermal Shock) for some selected Russian and Ukrainian VVER 1000 and 440/213 NPPs. In this paper the TAREG 2.01/00 project is briefly described with some details from the twin project 2.01/03, which served as a materials testing project, providing inputs for the 1st project. As a result of the project new trend curves for neutron irradiation embrittlement were elaborated, based on upgraded and more reliable surveillance results databases. The PTS study shows that the integrity of the selected VVER RPVs can be ensured to the end of RPV design life. (author)

Control of helium effects in irradiated materials based on theory and experiment

International Nuclear Information System (INIS)

Mansur, L.K.; Lee, E.H.; Maziasz, P.J.; Rowcliffe, A.F.

1986-01-01

Helium produced in materials by (n,α) transmutation reactions during neutron irradiations or subjected in ion bombardment experiments causes substantial changes in the response to displacement damage. In particular, swelling, phase transformations and embrittlement are strongly affected. Present understanding of the mechanisms underlying these effects is reviewed. Key theoretical relationships describing helium effects on swelling and helium diffusion are described. Experimental data in the areas of helium effects on swelling and precipitation is reviewed with emphasis on critical experiments that have been designed and evaluated in conjunction with theory. Confirmed principles for alloy design to control irradiation performance are described

Crystal Plasticity Model of Reactor Pressure Vessel Embrittlement in GRIZZLY

Energy Technology Data Exchange (ETDEWEB)

Chakraborty, Pritam [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Biner, Suleyman Bulent [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Zhang, Yongfeng [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Spencer, Benjamin Whiting [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

2015-07-01

The integrity of reactor pressure vessels (RPVs) is of utmost importance to ensure safe operation of nuclear reactors under extended lifetime. Microstructure-scale models at various length and time scales, coupled concurrently or through homogenization methods, can play a crucial role in understanding and quantifying irradiation-induced defect production, growth and their influence on mechanical behavior of RPV steels. A multi-scale approach, involving atomistic, meso- and engineering-scale models, is currently being pursued within the GRIZZLY project to understand and quantify irradiation-induced embrittlement of RPV steels. Within this framework, a dislocation-density based crystal plasticity model has been developed in GRIZZLY that captures the effect of irradiation-induced defects on the flow stress behavior and is presented in this report. The present formulation accounts for the interaction between self-interstitial loops and matrix dislocations. The model predictions have been validated with experiments and dislocation dynamics simulation.

Crystal Plasticity Model of Reactor Pressure Vessel Embrittlement in GRIZZLY

International Nuclear Information System (INIS)

Chakraborty, Pritam; Biner, Suleyman Bulent; Zhang, Yongfeng; Spencer, Benjamin Whiting

2015-01-01

The integrity of reactor pressure vessels (RPVs) is of utmost importance to ensure safe operation of nuclear reactors under extended lifetime. Microstructure-scale models at various length and time scales, coupled concurrently or through homogenization methods, can play a crucial role in understanding and quantifying irradiation-induced defect production, growth and their influence on mechanical behavior of RPV steels. A multi-scale approach, involving atomistic, meso- and engineering-scale models, is currently being pursued within the GRIZZLY project to understand and quantify irradiation-induced embrittlement of RPV steels. Within this framework, a dislocation-density based crystal plasticity model has been developed in GRIZZLY that captures the effect of irradiation-induced defects on the flow stress behavior and is presented in this report. The present formulation accounts for the interaction between self-interstitial loops and matrix dislocations. The model predictions have been validated with experiments and dislocation dynamics simulation.

A wide-range embrittlement trend curve for western RPV steels

International Nuclear Information System (INIS)

Kirk, M.T.

2011-01-01

Embrittlement trend curves (ETCs) are used to estimate neutron irradiation embrittlement as a function of both exposure (fluence, flux, temperature, ...) and composition variables. ETCs provide information needed to assess the structural integrity of operating nuclear reactors, and to determine their suitability for continued safe operation. Past efforts on ETC development in the United States have used data drawn from domestic licensees. While this approach has addressed past needs well, future needs such as power up-rates, license extensions to 60 years and beyond, and the use of low copper materials in new reactors produce future operating conditions for the US reactor fleet that may differ from past experience, suggesting that data from sources other than licensee surveillance programs may be needed. In this paper we draw together embrittlement data expressed in terms of ΔT41J and ΔYS from a wide variety of data sources as a first step in examining future embrittlement trends. We develop a 'wide range' ETC based on a collection of over 2500 data. We assess how well this ETC models the whole database, as well as significant data subsets. Comparisons presented herein indicate that a single algebraic model, denoted WR-C(5), represents reasonably well both the trends evident in the data overall as well as trends exhibited by four special data subsets. The WR-C(5) model indicates the existence of trends in high fluence data (Φ > 2-3*10 19 n/cm 2 , E > 1 MeV) that are not as apparent in the US surveillance data due to the limited quantity of ΔT30 data measured at high fluence in this dataset. Additionally, WR-C(5) models well the trends in both test and power reactor data despite the fact it has not term to account for flux. It is suggested that one appropriate use of the WR-C(5) trend curve may include the design irradiation studies to validate or refute the findings presented herein. Additionally, WR-C(5) could be used, along with other information (e.g., other

Radiation embrittlement of PWR vessel supports

International Nuclear Information System (INIS)

Cheverton, R.D.; Robinson, G.C.; Pennell, W.E.; Nanstad, R.K.

1989-01-01

Several studies pertaining to radiation damage of PWR vessel supports were conducted between 1978 and 1987. During this period, apparently there was no reason to believe that low-temperature (<100 degree C) MTR embrittlement data were not appropriate for evaluating embrittlement of PWR vessel supports. However, late in 1986, data from the High Flux Isotope Reactor (HFIR) vessel surveillance program indicated that the embrittlement rates of the several HFIR vessel materials (A212-B, A350-LF3, A105-II) were substantially greater than anticipated on the basis of MTR data. Further evaluation of the HFIR data suggested that a fluence-rate effect was responsible for the apparent discrepancy, and shortly thereafter it became apparent that this rate effect was applicable to the evaluation of LWR vessel supports. As a result, the Nuclear Regulatory Commission (NRC) requested that the Oak Ridge National Laboratory (ORNL) evaluate the impact of the apparent embrittlement rate effect on the integrity of light-water-reactor (LWR) vessel supports. The purpose of the study was to provide an indication of whether the integrity of reactor vessel supports is likely to be challenged by radiation-induced embrittlement. The scope of the evaluation included correlation of the HFIR data for application to the evaluation of LWR vessel supports; a survey and cursory evaluation of all US LWR vessel support designs, selection of two plants for specific-plant evaluation, and a specific-plant evaluation of both plants to determine critical flaw sizes for their vessel supports. 19 refs., 8 figs., 2 tabs

Design and use of the Embrittlement Data Base (EDB)

International Nuclear Information System (INIS)

Stallmann, F.W.

1987-01-01

The architecture of the Embrittlement Data Base (EDB) is described. This data base contains a comprehensive collection of experimental data related to irradiations of reactor pressure vessel steels in surveillance capsules and test reactors. Software is being developed for easy retrieval and analysis of the data. Data and software will be made available to interested parties on a cooperative basis

Low-temperature embrittlement and fracture of metals with different crystal lattices – Dislocation mechanisms

Directory of Open Access Journals (Sweden)

V.M. Chernov

2016-12-01

Full Text Available The state of a low-temperature embrittlement (cold brittleness and dislocation mechanisms for formation of the temperature of a ductile-brittle transition and brittle fracture of metals (mono- and polycrystals with various crystal lattices (BCC, FCC, HCP are considered. The conditions for their formation connected with a stress-deformed state and strength (low temperature yield strength as well as the fracture breaking stress and mobility of dislocations in the top of a crack of the fractured metal are determined. These conditions can be met for BCC and some HCP metals in the initial state (without irradiation and after a low-temperature damaging (neutron irradiation. These conditions are not met for FCC and many HCP metals. In the process of the damaging (neutron irradiation such conditions are not met also and the state of low-temperature embrittlement of metals is absent (suppressed due to arising various radiation dynamic processes, which increase the mobility of dislocations and worsen the strength characteristics.

Embrittlement recovery due to annealing of reactor pressure vessel steels

International Nuclear Information System (INIS)

Eason, E.D.; Wright, J.E.; Nelson, E.E.; Odette, G.R.; Mader, E.V.

1998-01-01

The irradiation embrittlement of nuclear reactor pressure vessels (RPV) can be reduced by thermal annealing at temperatures higher than the normal operating conditions. The objective of this work was to analyze the pertinent data and develop quantitative models for estimating the recovery in 41 J (30 ft-lb) Charpy transition temperature (TT) and Charpy upper shelf energy (USE) due to annealing. An analysis data base was developed, reviewed for completeness and accuracy, and documented as part of this work. Models were developed based on a combination of statistical techniques, including pattern recognition and transformation analysis, and the current understanding of the mechanisms governing embrittlement and recovery. The quality of models fitted in this project was evaluated by considering both the Charpy annealing data used for fitting and a surrogate hardness data base. This work demonstrates that microhardness recovery is a good surrogate for shift recovery and that there is a high level of consistency between the observed annealing trends and fundamental models of embrittlement and recovery processes. (orig.)

Heavy-section steel irradiation program: Embrittlement issues

International Nuclear Information System (INIS)

Corwin, W.R.

1991-01-01

Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents and the potential for major contamination releases. The RPV is one of only two major safety- related components of the plant for which a duplicate or redundant backup system does not exist. In particular, it is vital to fully understand the degree of irradiation-induced degradation of the RPV's fracture resistance which occurs during service, since without that radiation damage it is virtually impossible to postulate a realistic scenario which would result in RPV failure. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established by the US Nuclear Regulatory Commission (USNRC) to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior, and in particular the fracture toughness properties, of typical pressure vessel steels as they relate to light-water reactor pressure-vessel integrity. Effects of specimen size, material chemistry, product form and microstructure, irradiation fluence, flux, temperature and spectrum, and postirradiation annealing are being examined on a wide range of fracture properties including fracture toughness crack arrest toughness ductile tearing resistance Charpy V-notch impact energy, dropweight nil-ductility temperature and tensile properties. Models based on observations of radiation-induced microstructural changes using the field on microprobe and the high resolution transmission electron microscopy provide improved bases for extrapolating the measured changes in fracture properties to wider ranges of irradiation conditions. The principal materials examined within the HSSI program are high-copper welds since their postirradiation properties are most frequently limiting in the continued safe operation of commercial RPVs

Fracture analysis of HFIR beam tube caused by radiation embrittlement

International Nuclear Information System (INIS)

Chang, S.J.

1994-01-01

With an attempt to estimate the neutron beam tube embrittlement condition for the Oak Ridge High Flux Isotope Reactor (HFIR), fracture mechanics calculations are carried out in this paper. The analysis provides some numerical result on how the tube has been structurally weakened. In this calculation, a lateral impact force is assumed. Numerical result is obtained on how much the critical crack size should be reduced if the beam tube has been subjected to an extended period of irradiation. It is also calculated that buckling strength of the tube is increased, not decreased, with irradiation

Radiation embrittlement of WWER 440 pressure vessel steel and of some improved steels by western producers

International Nuclear Information System (INIS)

Koutsky, J.; Vacek, M.; Stoces, B.; Pav, T.; Otruba, J.; Novosad, P.; Brumovsky, M.

1982-01-01

The resistance was studied of Cr-Mo-V type steel 15Kh2MFA to radiation embrittlement at an irradiation temperature of around 288 degC. Studied was the steel used for the manufacture of the pressure vessel of the Paks nuclear reactor in Hungary. The obtained results of radiation embrittlement and hardening of steel 15Kh2MFA were compared with similar values of Mn-Ni-Mo type steels A 533-B and A 508 manufactured by leading western manufacturers within the international research programme coordinated by the IAEA. It was found that the resistance of steel 15Kh2MFA to radiation embrittlement is comparable with steels A 533-B and A 508 by western manufacturers. (author)

Effect of hydrogen and oxygen content on the embrittlement of Zr alloys

International Nuclear Information System (INIS)

Griger, A.; Hozer, Z.; Matus, L.; Vasaros, L.; Horvath, M.

2001-01-01

An experimental study is carried out in the KFKI Atomic Energy Research Institute in order to clear up the role of oxidation and hydrogen uptake in the embrittlement process. Russian E110 type Zr1%Nb and Zircaloy-4 claddings are used as test materials. The differences between the properties of two alloys are examined. The sample preparation covered the following cases: oxidation in Ar+O 2 atmosphere; hydrogen uptake of as received and pre-oxidised samples (in Ar+O 2 atmosphere); oxidation in steam. The oxidation in Ar+O 2 and the subsequent hydrogen uptake procedure make possible the production of samples with well-characterized hydrogen and oxygen content. Corrosion treated ring samples of 8 mm height are examined in ring compression tests. The force-deformation curves are recorded and the crushing force and deformation are determined. The relative deformation is used for the characterisation of embrittlement level. The results of experiments provide detailed information about the effect of hydrogen and oxygen content on the embrittlement of zirconium alloys. The conclusions are: 1) hydrogen seems to play a more important role in the embrittlement of zirconium alloys than oxygen; 2) the Zircaloy-4 alloy becomes brittle at lower hydrogen content than the Zr1%Nb; 3) under steam oxidation conditions the Zr1%Nb alloy takes up much more hydrogen and becomes more brittle than the Zircaloy-4

Effect of γ-IRRADIATION on the Mechanical Properties of Al-Cu Alloy

Science.gov (United States)

Abo-Elsoud, M.; Ismail, H.; Sobhy, Maged S.

SEM observations and Vickers hardness tests were performed to identify the irradiation effects. γ-irradiation effect during the aging hardening process can be explained depending on the composition of the alloy and is used to derive quantitative information on the kinetics of the transformation precipitates. Increasing the Cu content of an Al-Cu alloy can improve the aging hardness. The present results of the hardness behavior, with SEM observations of surveillance specimens at different doses, suggest that the radiation-induced defects are probably complex valence-solute clusters. These clusters act as nuclei for the precipitation of θ-Al2Cu type. This can be effectively utilized to study the systematics of nucleation of precipitates at vacancy-type defects. γ-irradiation probably plays the key role in defects responsible for material strengthening and embrittlement.

Preliminary irradiation test results from the Yankee Atomic Electric Company reactor vessel test irradiation program

International Nuclear Information System (INIS)

Biemiller, E.C.; Fyfitch, Stephen; Campbell, C.A.

1994-01-01

The Yankee Atomic Electric Company test irradiation program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials and to remove uncertainties in the analysis of existing irradiation data on the Yankee Rowe reactor vessel steel. Plate materials each containing 0.24 w/o copper, but different nickel contents at 0.63 w/o and 0.19 w/o, were heat treated to simulate the Yankee vessel heat treatment (austenitized at 982 o C (1800 o F)) and to simulate Regulatory Guide 1.99 database materials (austenitized at 871 o C (1600 o F)). These heat treatments produced different microstructures so the effect of microstructure on irradiation damage sensitivity could be tested. Because the nickel content of the test plates varied and the copper level was constant, the effect of nickel on irradiation embrittlement was also tested. Correlation monitor material, HSST-02, was included in the program to benchmark the Ford Nuclear Reactor (University of Michigan Test Reactor) which had never been used before for this type of irradiation program. Materials taken from plate surface locations (versus 1/4 T) were included to test whether or not the improved toughness properties of the plate surface layer, resulting from the rapid quench, are maintained after irradiation. If the improved properties are maintained, pressurized thermal shock calculations could utilize this margin. Finally, for one experiment, irradiations were conducted at two irradiation temperatures (260 o C and 288 o C) to determine the effect of irradiation temperature on embrittlement. (Author)

Irradiation, annealing, and reirradiation research in the ORNL heavy-section steel irradiation program

International Nuclear Information System (INIS)

Nanstad, R.K.; Iskander, S.K.; McCabe, D.E.; Sokolov, M.A.

1997-01-01

One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPV) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. This paper summarizes experimental results from work performed as part of the Heavy-Section Steel Irradiation (HSSI) Program managed by Oak Ridge National Laboratory (ORNL) for the U.S. Nuclear Regulatory Commission. The HSSI Program focuses on annealing and re-embrittlement response of materials which are representative of those in commercial RPVs and which are considered to be radiation-sensitive. Experimental studies include (1) the annealing of materials in the existing inventory of previously irradiated materials, (2) reirradiation of previously irradiated/annealed materials in a collaborative program with the University of California, Santa Barbara (UCSB), (3) irradiation/annealing/reirradiation of U.S. and Russian materials in a cooperative program with the Russian Research Center-Kurchatov Institute (RRC-KI), (4) the design and fabrication of an irradiation/anneal/reirradiation capsule and facility for operation at the University of Michigan Ford Reactor, (5) the investigation of potential for irradiation-and/or thermal-induced temper embrittlement in heat-affected zones (HAZs) of RPV steels due to phosphorous segregation at grain boundaries, and (6) investigation of the relationship between Charpy impact toughness and fracture toughness under all conditions of irradiation, annealing, and reirradiation

Estimation of RPV material embrittlement for Ukrainian NPP based on surveillance test data

International Nuclear Information System (INIS)

Revka, V.; Chyrko, L.; Chaikovsky, Yu.; Gulchuk, Yu.

2012-01-01

The WWER-1000 RPV material embrittlement has been evaluated using the surveillance test data for the nuclear power plant which is under operation in Ukraine. The RPV materials after the neutron (E > 0,5 MeV) irradiation up to fluence of 22,9.10 22 m -2 have been studied. Fracture toughness tests were performed using pre-cracked Charpy specimens for the beltline materials (base and weld metal). The maximum shift of T 0 reference temperature is equal to 44 o C. A radiation embrittlement rate, A F , for the RPV materials was estimated using the standard and reconstituted specimens. A comparison of the A F values has shown a good agreement between the specimen sets before and after reconstitution both for base and weld metal. Furthermore it has been revealed there is no nickel effect for the studied materials. In spite of the high nickel content the radiation embrittlement rate for weld metal is not higher than for base metal with low nickel content. Fracture toughness analysis has shown the Master curve shape describes well a temperature dependence of K Jc values. However a higher scatter of K Jc values is observed in comparison to 95 % tolerance bounds. (author)

The study of the irradiation-induced embrittlement of reactor pressure vessels. Analysis of surveillance test specimens of a commercial nuclear reactor pressure vessel studied by three-dimensional atom probe and positron annihilation

International Nuclear Information System (INIS)

Nagai, Yasuyoshi; Toyama, Takeshi; Hasegawa, Masayuki

2007-01-01

The study of embrittlement of nuclear power reactor pressure vessels (RPVs) is of critical importance for the safety assessment in the nuclear industry. Some origins of embrittlement are attributed to fine Cu precipitates, matrix defects, grain boundary segregation of P and late blooming phase. This review article described nanostructural observation by three-dimensional atom probe (3DAP) and positron annihilation spectroscopy (PAS). The density and sizes of Cu-rich nanoprecipitates and grain boundary segregation are sensitively detected by 3DAP, and vacancies are probed by PAS. Element analysis around vacancies and fine microstructural Cu precipitates not containing vacancies are successfully observed by a coincidence doppler broadening method. The nanostructural evolution of irradiation-induced Cu-rich nanoprecipitates (CRNPs) and vacancy clusters in surveillance test specimens of commercial nuclear reactor pressure vessel steel welds of Doel-2 in Belgium were revealed by combining 3DAP and PAS. In both medium (0.13 wt%) and high (0.30 wt%) Cu welds, the CRNPs were found to form readily at the very beginning of the reactor lifetime. On the other hand, small vacancy clusters start appearing after the initial Cu precipitates and accumulate steadily with increasing neutron dose. The CRNPs were also observed at very low dose rate of neutrons in the test specimen of Calder Hall Reactor of Japan Atomic Power Company. The significant enhancement of these Cu precipitates results in the embrittlement in practical RPVs. At very high dose of 2.2x10 18 n/cm 2 by JMTR, the Cu precipitates were scarcely observed, and the irradiation-induced embrittlement was primarily caused from vacancy-impurity complexes and dislocation loops. (author)

Mechanical properties and TEM examination of RAFM steels irradiated up to 70 dpa in BOR-60

Energy Technology Data Exchange (ETDEWEB)

Gaganidze, E., E-mail: Ermile.Gaganidze@kit.edu [Karlsruher Institut fuer Technologie, Institut fuer Angewandte Materialien, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Petersen, C.; Materna-Morris, E.; Dethloff, C.; Weiss, O.J.; Aktaa, J. [Karlsruher Institut fuer Technologie, Institut fuer Angewandte Materialien, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Povstyanko, A.; Fedoseev, A.; Makarov, O.; Prokhorov, V. [Joint Stock Company ' State Scientific Centre Research Institute of Atomic Reactors' , 433510 Dimitrovgrad-10, Ulyanovsk Region (Russian Federation)

2011-10-01

Mechanical properties of Reduced Activation Ferritic/Martensitic (RAFM) steels were studied after irradiation in BOR-60 reactor to a neutron displacement damage of 70 dpa at 330-340 deg. C. Yield stress and Ductile-to-Brittle-Transition-Temperature of EUROFER97 indicate saturation of hardening and embrittlement. The phenomenological models for description of microstructure evolution and resulting irradiation hardening and embrittlement are discussed. The evolution of yield stress with dose is qualitatively understood within a Whapham and Makin model. Dislocation loops examined in TEM are considered a main source for low-temperature irradiation hardening. The analysis of the fatigue data in terms of the inelastic strain reveals comparable fatigue behaviour both for unirradiated and irradiated conditions, which can be described by a common Manson-Coffin relation. The study of helium effects in B-doped model steels indicated progressive material embrittlement with helium content. Post-irradiation annealing of RAFM steels yielded substantial recovery of mechanical properties.

Comparison of embrittlement trend curves to high fluence surveillance results

International Nuclear Information System (INIS)

Bogaert, A.S.; Gerard, R.; Chaouadi, R.

2011-01-01

In the regulatory justification of the integrity of the reactor pressure vessels (RPV) for long term operation, use is made of predictive formulas (also called trend curves) to evaluate the RPV embrittlement (expressed in terms of RTNDT shifts) in function of fluence, chemical composition and in some cases temperature, neutron flux or product form. It has been shown recently that some of the existing or proposed trend curves tend to underpredict high dose embrittlement. Due to the scarcity of representative surveillance data at high dose, some test reactor results were used in these evaluations and raise the issue of representativeness of the accelerated test reactor irradiations (dose rate effects). In Belgium the surveillance capsules withdrawal schedule was modified in the nineties in order to obtain results corresponding to 60 years of operation or more with the initial surveillance program. Some of these results are already available and offer a good opportunity to test the validity of the predictive formulas at high dose. In addition, advanced surveillance methods are used in Belgium like the Master Curve, increased tensile tests, and microstructural investigations. These techniques made it possible to show the conservatism of the regulatory approach and to demonstrate increased margins, especially for the first generation units. In this paper the surveillance results are compared to different predictive formulas, as well as to an engineering hardening model developed at SCK.CEN. Generally accepted property-to-property correlations are critically revisited. Conclusions are made on the reliability and applicability of the embrittlement trend curves. (authors)

Thermal embrittlement of reactor vessel steels

International Nuclear Information System (INIS)

Corwin, W.R.; Nanstad, R.K.; Alexander, D.J.; Stoller, R.E.; Wang, J.A.; Odette, G.R.

1995-01-01

As a result of observations of possible thermal embrittlement from recent studies with welds removed from retired steam generators of the Palisades Nuclear Plant (PNP), an assessment was made of thermal aging of reactor pressure vessel (RPV) steels under nominal reactor operating conditions. Discussions are presented on (1) data from the literature regarding relatively low-temperature thermal embrittlement of RPV steels; (2)relevant data from the US power reactor-embrittlement data base (PR-EDB); and (3)potential mechanisms of thermal embrittlement in low-alloy steels

Neutron irradiation effects on grain-refined W and W-alloys

International Nuclear Information System (INIS)

Hasegawa, A.; Fukuda, M.; Tanno, T.; Nogami, S.; Yabuuchi, K.; Tanaka, T.; Muroga, T.

2014-10-01

Microstructural data of neutron irradiated Tungsten (W) such as size and number density of voids and precipitates obtained by W up to 1.5dpa irradiation in the temperature range of 400-800degC were compiled quantitatively. Nucleation and growth process of these defects were clarified and a qualitative prediction of the damage structure development and hardening of W in fusion reactor environments were made taking into account the solid transmutation effects for the first time. To improve recrystallization behavior and low temperature embrittlement, grain refined-W alloys were fabricated by K- or La-doping method. Rhenium addition to the grain refining process was also examined to improve mechanical properties. Characterizations of unirradiated materials were performed. (author)

Hardening Embrittlement and Non-Hardening Embrittlement of Welding-Heat-Affected Zones in a Cr-Mo Low Alloy Steel

Directory of Open Access Journals (Sweden)

Yu Zhao

2018-06-01

Full Text Available The embrittlement of heat affected zones (HAZs resulting from the welding of a P-doped 2.25Cr-1Mo steel was studied by the analysis of the fracture appearance transition temperatures (FATTs of the HAZs simulated under a heat input of 45 kJ/cm with different peak temperatures. The FATTs of the HAZs both with and without tempering increased with the rise of the peak temperature. However, the FATTs were apparently lower for the tempered HAZs. For the as-welded (untempered HAZs, the FATTs were mainly affected by residual stress, martensite/austenite (M/A islands, and bainite morphology. The observed embrittlement is a hardening embrittlement. On the other hand, the FATTs of the tempered HAZs were mainly affected by phosphorus grain boundary segregation, thereby causing a non-hardening embrittlement. The results demonstrate that the hardening embrittlement of the as-welded HAZs was more severe than the non-hardening embrittlement of the tempered HAZs. Consequently, a post-weld heat treatment should be carried out if possible so as to eliminate the hardening embrittlement.

Lower Length Scale Model Development for Embrittlement of Reactor Presure Vessel Steel

Energy Technology Data Exchange (ETDEWEB)

Zhang, Yongfeng [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schwen, Daniel [Idaho National Lab. (INL), Idaho Falls, ID (United States); Chakraborty, Pritam [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bai, Xianming [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-09-01

This report summarizes the lower-length-scale effort during FY 2016 in developing mesoscale capabilities for microstructure evolution, plasticity and fracture in reactor pressure vessel steels. During operation, reactor pressure vessels are subject to hardening and embrittlement caused by irradiation induced defect accumulation and irradiation enhanced solute precipitation. Both defect production and solute precipitation start from the atomic scale, and manifest their eventual effects as degradation in engineering scale properties. To predict the property degradation, multiscale modeling and simulation are needed to deal with the microstructure evolution, and to link the microstructure feature to material properties. In this report, the development of mesoscale capabilities for defect accumulation and solute precipitation are summarized. A crystal plasticity model to capture defect-dislocation interaction and a damage model for cleavage micro-crack propagation is also provided.

Neutron irradiation effects on mechanical properties in SA508 Gr4N high strength low alloy steel

International Nuclear Information System (INIS)

Kim, Minchul; Lee, Kihyoung; Park, Sanggyu; Choi, Kwonjae; Lee, Bongsang

2012-01-01

The Reactor Pressure Vessel (RPV) is the key component in determining the lifetime of nuclear power plants because it is subject to the significant aging degradation by irradiation and thermal aging, and there is no practical method for replacing that component. Advanced reactors with much larger capacity than current reactor require the usage of higher strength materials inevitably. The SA508 Gr.4N Ni Cr Mo low alloy steel, in which Ni and Cr contents are larger than in conventional RPV steels, could be a promising RPV material offering improved strength and toughness from its tempered martensitic microstructure. For a structural integrity of RPV, the effect of neutron irradiation on the material property is one of the key issues. The RPV materials suffer from the significant degradation of transition properties by the irradiation embrittlement when its strength is increased by a hardening mechanism. Therefore, the potential for application of SA508 Gr.4N steel as the structural components for nuclear power reactors depends on its ability to maintain adequate transition properties against the operating neutron does. However, it is not easy to fine the data on the irradiation effect on the mechanical properties of SA508 Gr.4N steel. In this study, the irradiation embrittlement of SA508 Gr.4N Ni Cr Mo low alloy steel was evaluated by using specimens irradiated in research reactor. For comparison, the variations of mechanical properties by neutron irradiation for commercial SA508 Gr.3 Mn Mo Ni low alloy steel were also evaluated

The effects of normal paraffins mobilizers on irradiated polypropylene

International Nuclear Information System (INIS)

Chen Wenxiu; Gao Ling

1995-01-01

The n-paraffins blended with polypropylene (PP) as mobilizer had been investigated. The effectiveness of mobilizer (n-paraffins) on irradiated polypropylene is dependent on the molecular weight of mobilizer and its content on polypropylene. The n-docosame (n-C 22 ) possesses the best effectiveness of radiation tolerance on PP among the mobilizer paraffins: n-decane (n-C 10 ), n-hexadecane (n-C 16 ), n-docosane (n-C 22 ) and n-hexatriacontane (n-C 36 ). The 2% (w/w) content of a given mobilizer is the most effective at reducing the embrittlement of irradiated PP as evidenced by the elongation at break. The physical properties of polypropylene with mobilizers such as density, Young's modulus, the Fraction of free volume and the weight swelling ratio in p-xylene at room temperature were measured. Above phenomena are related with the constructive of blended PP and demonstrated by its physical properties

Fracture toughness prediction for RPV Steels with various degree of embrittlement

International Nuclear Information System (INIS)

Margolin, B.; Gulenko, A.; Shvetsova, V.

2003-01-01

In the present report, predictions of the temperature dependence of cleavage fracture toughness are performed on the basis of the Master Curve approach and a probabilistic model named now the Prometey model. These predictions are performed for reactor pressure vessel steels in different states, the initial (as-produced), irradiated state with moderate degree of embrittlement and in the highly embrittled state. Calculations of the K IC (T) curves may be performed with both approaches on the basis of fracture toughness test results from pre-cracked Charpy specimens at some (one) temperature. The calculated curves are compared with test results. It is shown that the K IC (T) curves for the initial state calculated with the Master Curve approach and the probabilistic model show good agreement. At the same time, for highly embrittled RPV steel, the K IC (T) curve predicted with the Master Curve approach is not an adequate fit to the experimental data, whereas the agreement of the test results and the K IC (T) curve calculated with the probabilistic model is good. An analysis is performed for a possible variation of the K IC (T) curve shape and the scatter in K IC results. (author)

Low temperature hydrogen embrittlement of niobium. II. Microscopic observations

International Nuclear Information System (INIS)

Grossbeck, M.L.; Birnbaum, H.K.

1977-01-01

The detailed, microscopic processes which occur during the hydrogen embrittlement of pure Nb are examined using in situ SEM crack propagation studies, SEM fractography, electron diffraction and ion probe methods. These results show that the fracture process occurs in a stress induced NbH hydride phase which forms in front of the propagating crack. The experimental results are in good agreement with the stress induced hydride embrittlement mechanism which is discussed. The thermodynamics of precipitation of hydrides under external stress is discussed and calculations are presented for the stress effects on the α-β solvus temperatures. These are related to the embrittlement process and evidence is presented to support the calculated stress effects on the solvus temperature

Effect of heat treatments on the hydrogen embrittlement ...

Indian Academy of Sciences (India)

pipe steel in as received (controlled rolled), normalized, and quenched and tempered conditions. The resistance to hydrogen embrittlement was found in the order of controlled rolled > quenched and tempered > normalized. The fracture mode ...

Intrinsic ductility and environmental embrittlement of binary Ni3Al

International Nuclear Information System (INIS)

George, E.P.; Liu, C.T.; Pope, D.P.

1993-01-01

Polycrystalline, B-free Ni 3 Al (23.4 at.% Al), produced by cold working and recrystallizing a single crystal, exhibits room temperature tensile ductilities of 3-5% in air and 13-16% in oxygen. These ductilities are considerably higher than anything previously reported, and demonstrate that the 'intrinsic' ductility of Ni 3 Al is much higher than previously thought. They also show that the moisture present in ordinary ambient air can severely embrittle Ni 3 Al (ductility decreasing from a high of 16% in oxygen to a low of 3% in air). Fracture is predominantly intergranular in both air and oxygen. This indicates that, while moisture can further embrittle the GBs in Ni 3 Al, they persist as weak links even in the absence of environmental embrittlement. However, they are not 'intrinsically brittle' as once thought, since they can withstand relatively large plastic deformations prior to fracture. Because B essentially eliminates environmental embrittlement in Ni 3 Al - and environmental embrittlement is a major cause of poor ductility in B-free Ni 3 Al - it is concluded that a significant portion of the so-called B effect must be related to suppression of moisture-induced environmental embrittlement. However, since B-doped Ni 3 Al fractures transgranularly, whereas B-free Ni 3 Al fractures predominantly intergranularly, B must have the added effect that it strengthens the GBs. A comparison with the earlier work on Zr-doped Ni 3 Al shows that Zr improves the ductility of Ni 3 Al, both in air and (and even more dramatically) in oxygen. While the exact mechanism of this ductility improvement is not clear at present, Zr appears to have more of an effect on (enhancing) GB strength than on (suppressing) environmental embrittlement

Fractal characteristics of fracture morphology of steels irradiated with high-energy ions

Energy Technology Data Exchange (ETDEWEB)

Xian, Yongqiang; Liu, Juan [Institute of Modern Physics, Chinese Academy of Science, Lanzhou 730000 (China); University of Chinese Academy of Science, Beijing 100049 (China); Zhang, Chonghong, E-mail: c.h.zhang@impcas.ac.cn [Institute of Modern Physics, Chinese Academy of Science, Lanzhou 730000 (China); Chen, Jiachao [Paul Scherrer Institute, Villigen PSI (Switzerland); Yang, Yitao; Zhang, Liqing; Song, Yin [Institute of Modern Physics, Chinese Academy of Science, Lanzhou 730000 (China)

2015-06-15

Highlights: • Fractal dimensions of fracture surfaces of steels before and after irradiation were calculated. • Fractal dimension can effectively describe change of fracture surfaces induced by irradiation. • Correlation of change of fractal dimension with embrittlement of irradiated steels is discussed. - Abstract: A fractal analysis of fracture surfaces of steels (a ferritic/martensitic steel and an oxide-dispersion-strengthened ferritic steel) before and after the irradiation with high-energy ions is presented. Fracture surfaces were acquired from a tensile test and a small-ball punch test (SP). Digital images of the fracture surfaces obtained from scanning electron microscopy (SEM) were used to calculate the fractal dimension (FD) by using the pixel covering method. Boundary of binary image and fractal dimension were determined with a MATLAB program. The results indicate that fractal dimension can be an effective parameter to describe the characteristics of fracture surfaces before and after irradiation. The rougher the fracture surface, the larger the fractal dimension. Correlation of the change of fractal dimension with the embrittlement of the irradiated steels is discussed.

Mechanisms of liquid-metal embrittlement

International Nuclear Information System (INIS)

Popovich, V.V.

1979-01-01

The mechanism of the embrittlement of metals and alloys during deformation in contact with liquid metals are discussed. With 20Kh13 steel in a Pb-Sn melt and polycrystalline Al in the presence of various mercury solutions a.s examples, considered are the three main processes - adsorption, corrosion (dissolution), formation of new phases which cause the disintegration of materials under the action of liquid-metallic media. Presented are data on plastic ductile and strength properties of the above materials in the presence of liquid-metallic media. A model is described that takes into account the effect of the medium upon the plastic deformation and the part the medium plays in liquid-metallic embrittlement

Preliminary irradiation test results from the Yankee Atomic Electric Company reactor vessel test irradiation program

International Nuclear Information System (INIS)

Biemiller, E.C.; Fyfitch, S.; Campbell, C.A.

1993-01-01

The Yankee Atomic Electric Company test irradiation program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials and to remove uncertainties in the analysis of existing irradiation data on the Yankee Rowe reactor vessel steel. Plate materials each containing 0.24 w/o copper, but different nickel contents at 0.63 w/o and 0.19 w/o, were heat treated to simulate the Yankee vessel heat treatment (austenitized at 1800 deg F) and to simulate Regulatory Guide 1.99 database materials (austenitized at 1600 deg. F). These heat treatments produced different microstructures so the effect of microstructure on irradiation damage sensitivity could be tested. Because the nickel content of the test plates varied and the copper level was constant, the effect of nickel on irradiation embrittlement was also tested. Correlation monitor material, HSST-02, was included in the program to benchmark the Ford Nuclear Reactor (U. of Michigan Test Reactor) which had never been used for this type of irradiation program. Materials taken from plate surface locations (vs. 1/4T) were included to test whether or not the improved toughness properties of the plate surface layer, resulting from the rapid quench, is maintained after irradiation. If the improved properties are maintained, pressurized thermal shock calculations could utilize this margin. Finally, for one experiment, irradiations were conducted at two irradiation temperatures (500 deg. F and 550 deg. F) to determine the effect of irradiation temperature on embrittlement. The preliminary results of the irradiation program show an increase in T 30 shift of 69 deg. F for a decrease in irradiation temperature of 50 deg. F. The results suggest that for nickel bearing steels, the superior toughness of plate surface material is maintained after irradiation and for the copper content tested, nickel had no apparent effect on irradiation response. No apparent microstructure

Technical Letter Report on the Cracking of Irradiated Cast Stainless Steels with Low Ferrite Content

Energy Technology Data Exchange (ETDEWEB)

Chen, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Alexandreanu, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, K. [Argonne National Lab. (ANL), Argonne, IL (United States)

2014-11-01

Crack growth rate and fracture toughness J-R curve tests were performed on CF-3 and CF-8 cast austenite stainless steels (CASS) with 13-14% of ferrite. The tests were conducted at ~320°C in either high-purity water with low dissolved oxygen or in simulated PWR water. The cyclic crack growth rates of CF-8 were higher than that of CF-3, and the differences between the aged and unaged specimens were small. No elevated SCC susceptibility was observed among these samples, and the SCC CGRs of these materials were comparable to those of CASS alloys with >23% ferrite. The fracture toughness values of unirradiated CF-3 were similar between unaged and aged specimens, and neutron irradiation decreased the fracture toughness significantly. The fracture toughness of CF-8 was reduced after thermal aging, and declined further after irradiation. It appears that while lowering ferrite content may help reduce the tendency of thermal aging embrittlement, it is not very effective to mitigate irradiation-induced embrittlement. Under a combined condition of thermal aging and irradiation, neutron irradiation plays a dominant role in causing embrittlement in CASS alloys.

Extreme embrittlement of austenitic stainless steel irradiated to 75--81 dpa at 335--360 C

International Nuclear Information System (INIS)

Porollo, S.I.; Vorobjev, A.N.; Konobeev, Yu.V.; Garner, F.A.

1998-01-01

This paper presents the results of an experiment conducted in the BN-350 fast reactor in Kazakhstan that involved the irradiation of argon-pressurized thin-walled tubes (0--2000 MPa hoop stress) constructed from Fe-16Cr-15Ni-3Mo-Nb stabilized steel in contact with the sodium coolant, which enters the reactor at ∼270 C. Tubes in the annealed condition reached 75 dpa at 335 C, and another set in the 20% cold-worked condition reached 81 dpa at 360 C. Upon disassembly all tubes, except those in the stress-free condition, were found to have failed in an extremely brittle fashion. The stress-free tubes exhibited diameter changes that imply swelling levels ranging from 9 to 16%. It is expected that stress-enhancement of swelling induced even larger swelling levels in the stressed tubes. The embrittlement is explained in terms of the sensitivity of the swelling regime to displacement rate and the large, unprecedented levels of swelling reached at 335--360 C at these high neutron fluences. The failure mechanism appears to be identical to that observed at similar swelling levels in other austenitic steels irradiated in US fast reactors at 400--425 C, whereby stress-concentration between voids and nickel segregation at void surfaces predisposes the steel to an epsilon martensite transformation followed by formation of alpha martensite at crack tips. The very slow strain rate inherent in such creep tests and the relatively high helium levels may also contribute to the failure

Power reactor embrittlement data base

International Nuclear Information System (INIS)

Kam, F.B.K.; Stallmann, F.W.; Wang, J.A.

1989-01-01

Regulatory and research evaluations of embrittlement prediction models and of vessel integrity under load can be greatly expedited by the use of a well-designed, computerized embrittlement data base. The Power Reactor Embrittlement Data Base (PR-EDB) is a comprehensive collection of data from surveillance reports and other published reports of commercial nuclear reactors. The uses of the data base require that as many different data as available are collected from as many sources as possible with complete references and that subsets of relevant data can be easily retrieved and processed. The objectives of this NRC-sponsored program are the following: to compile and to verify the quality of the PR-EDB; to provide user-friendly software to access and process the data; to explore or confirm embrittlement prediction models; and to interact with standards organizations to provide the technical bases for voluntary consensus standards that can be used in regulatory guides, standard review plans, and codes. 9 figs

PR-EDB: Power Reactor Embrittlement Data Base, version 1: Program description

International Nuclear Information System (INIS)

Stallmann, F.W.; Kam, F.B.K.; Taylor, B.J.

1990-06-01

Data concerning radiation embrittlement of pressure vessel steels in commercial power reactors have been collected form available surveillance reports. The purpose of this NRC-sponsored program is to provide the technical bases for voluntary consensus standards, regulatory guides, standard review plans, and codes. The data can also be used for the exploration and verification of embrittlement prediction models. The data files are given in dBASE 3 Plus format and can be accessed with any personal computer using the DOS operating system. Menu-driven software is provided for easy access to the data including curve fitting and plotting facilities. This software has drastically reduced the time and effort for data processing and evaluation compared to previous data bases. The current compilation of the Power Reactor Embrittlement Data base (PR-EDB, version 1) contains results from surveillance capsule reports of 78 reactors with 381 data points from 110 different irradiated base materials (plates and forgings) and 161 data points from 79 different welds. Results from heat-affected-zone materials are also listed. Electric Power Research Institute (EPRI), reactor vendors, and utilities are in the process of providing back-up quality assurance checks of the PR-EDB and will be supplementing the data base with additional data and documentation. 2 figs., 28 tabs

Hydrogen embrittlement of steels: study and prevention

International Nuclear Information System (INIS)

Brass, A.M.; Chene, J.; Coudreuse, L.

2000-01-01

Hydrogen embrittlement of steels is one of the important reason of rupture of pieces in the industry (nuclear, of petroleum..). Indeed, there are a lot of situations which can lead to the phenomenon of hydrogen embrittlement: introduction of hydrogen in the material during the elaboration or during transformation or implementation processes (heat treatments, welding); use of steels when hydrogen or hydrogenated gaseous mixtures are present; hydrogen produced by electrolytic reactions (surface treatments, cathodic protection). The hydrogen embrittlement can appear in different forms which depend of a lot of parameters: material (state, composition, microstructure..); surrounding medium (gas, aqueous medium, temperature..); condition of mechanical solicitation (static, dynamic, cyclic..). The industrial phenomena which appear during cases of hydrogen embrittlement are more particularly described here. Several methods of steels studies are proposed as well as some possible ways for the prevention of hydrogen embrittlement risks. (O.M.)

LYRA and other projects on RPV steel embrittlement study and mitigation of the AMES network

International Nuclear Information System (INIS)

Debarberis, L.; Estorff, U. von; Crutzen, S.; Beers, M.; Stamm, H.; Vries, M.I. de; Tjoa, G.L.

1998-01-01

Within the framework of the European Network AMES, Ageing Materials evaluation and Studies, a number of experimental works on RPV materials embrittlement are carried out at the Institute of Advanced Materials (AIM) of the Joint Research Centre (JRC) of the European Commission (EC). The objectives of AMES are mainly the understanding of the property degradation phenomena of RPV western reference steels like JRQ and HSST, eastern RPV steels like 15X2mFA and 15H2X15, and annealing possibilities. In order to conduct a very high quality irradiation rig, LYRA facility, has been designed and developed at the High Flux Reactor (HFR) Petten. An other dedicated rig, named LIMA, has been developed at the HFR Petten in order to irradiate RPV steels, internals and in-core materials under typical BWR/PWR conditions. The samples can be irradiated in pressurised water up to 160 bar, 320 deg. C, and the water chemistry fully controlled. For irradiation of standard or miniaturised LWR related materials samples, another group of well experienced irradiation devices with inert gas or liquid metals environment are employed. These devices are tailored to their various specific applications. This paper is intended to give information about the structure and the objectives of the existing European network AMES, and to present the various AMES main and spin-off projects, including a brief description on he modelling activities related to RPV materials embrittlement. (author)

Effects of surface condition on aqueous corrosion and environmental embrittlement of iron aluminides

Energy Technology Data Exchange (ETDEWEB)

Perrin, R.L.; Buchanan, R.A. [Univ. of Tennessee, Knoxville, TN (United States)

1996-08-01

Effects of retained high-temperature surface oxides, produced during thermomechanical processing and/or heat treatment, on the aqueous-corrosion and environmental-embrittlement characteristics of Fe{sub 3}Al-based iron aluminides (FA-84, FA-129 and FAL-Mo), a FeAl-based iron aluminide (FA-385), and a disordered low-aluminum Fe-Al alloy (FAPY) were evaluated. All tests were conducted at room temperature in a mild acid-chloride solution. In cyclic-anodic-polarization testing for aqueous-corrosion behavior, the surface conditions examined were: as-received (i.e., with the retained high-temperature oxides), mechanically cleaned and chemically cleaned. For all materials, the polarization tests showed the critical pitting potentials to be significantly lower in the as-received condition than in the mechanically-cleaned and chemically-cleaned conditions. These results indicate detrimental effects of the retained high-temperature oxides in terms of increased susceptibilities to localized corrosion. In 200-hour U-bend stress-corrosion-cracking tests for environmental-embrittlement behavior, conducted at open-circuit corrosion potentials and at a hydrogen-charging potential of {minus}1500 mV (SHE), the above materials (except FA-385) were examined with retained oxides and with mechanically cleaned surfaces. At the open-circuit corrosion potentials, none of the materials in either surface condition underwent cracking. At the hydrogen-charging potential, none of the materials with retained oxides underwent cracking, but FA-84, FA-129 and FAL-Mo in the mechanically cleaned condition did undergo cracking. These results suggest beneficial effects of the retained high-temperature oxides in terms of increased resistance to environmental hydrogen embrittlement.

Damage dosimetry and embrittlement monitoring of nuclear pressure vessels in real time by magnetic properties measurement. Final report

International Nuclear Information System (INIS)

Ougouag, A.M.; Stubbins, J.F.; Williams, J.F.; Shong, Wei-Ja.

1995-04-01

This program developed a nondestructive technique for gauging the progress of embrittlement of nuclear pressure vessel steels (PVS) by means of monitoring radiation-induced changes in magnetic properties. The technique was developed by running a series of experiments in reactor on typical nuclear pressure vessel steels and weldment material. Following irradiation, changes in magnetic properties were measured and correlated with irradiation dose and with mechanical properties changes, where possible. The changes in magnetic properties were unique to the irradiation environment, and were much larger than those produce by thermal aging in the absence of irradiation. Special techniques for magnetic properties change measurement were developed and complimented by more standard magnetic properties measurement techniques including SQUID measurements. The results of the experiments revealed that magnetic properties were very sensitive to irradiation. Changes in microstructurally-related magnetic properties of as much as 40% were noted after irradiation exposure of as little as 10 17 n/cm 2 (E > 0.1 MeV). The magnetic properties changes plateaued out after doses of around as 10 18 n/cm 2 (E > 0.1 MeV). It is unclear whether further changes would be noted at higher doses which would also be useful for tracking the embrittlement phenomenon. This is recommended for further study. The work supported here resulted in several publications in the open scientific literature

Nuclear power plant life extension and management aspects; neutron irradiation embrittlement and stress corrosion cracking - two possible degradation mechanisms and methods for their mitigation

International Nuclear Information System (INIS)

Tipping, P.; Ineichen, U.; Cripps, R.C.

1994-01-01

The response of a mock-up low alloy ferritic reactor pressure vessel (RPV) steel and associated weldments to neutron irradiation has been studied using a combination of hardness, tensile, fracture mechanical and toughness tests in combination with annealing treatments. Thermal analysis using isochronal and isothermal techniques has indicated that annealing at a minimum of 440 o C for 168h is needed to mitigate neutron embrittlement received at 290 o C. Rates of re-embrittlement after annealing and reirradiating are no faster than initial rates, even up to neutron fluences as high as 5x10 19 cm -2 (energy E>1 MeV). All mechanical properties measured benefited from annealing. Thus, annealing is indicated as one measure for maintaining mechanical properties in irradiated low alloy steels and welds and should be considered in plant life management strategies. The influence of simulated reactor coolant water chemistry on the stress corrosion cracking propensity of ferritic low alloy steel specimens in autoclave loop experiments has also been studied. The double cantilever bend specimens were fatigue pre-cracked and wedge-loaded to different degrees to induce nominal stress intensity factors between 15-95 MPa.m 1/2 . Other specimens were subjected to stress using a tensile loading device integral with the test autoclave. The importance of close control of the dissolved oxygen content and the conductivity of the water has become evident under these experimental conditions. The RPV material and degree and mode of loading are also important parameters in SCC studies; stress intensity factors above 30 MPa.m 1/2 have been associated with SCC in these studies. (author) 2 figs., 13 refs

Status on the selection and development of an embrittlement trend curve to use in ASTM standard guide E900

International Nuclear Information System (INIS)

Kirk, M.; Brian Hall, J.; Server, W.; Lucon, E.; Erickson, M.; Stoller, R.

2015-01-01

ASTM E900-07, Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials, includes an embrittlement trend curve. The trend curve can be used to predict the effect of neutron irradiation on the embrittlement of ferritic pressure vessel steels, as quantified by the shift in the Charpy V-Notch transition curve at 41 Joules of absorbed energy (ΔT 41J ). The current E900 trend curve was first adopted in the 2002 revision. In 2011 ASTM Subcommittee E10.02 undertook an extensive effort to evaluate the adequacy of the E900 trend curve for continued use. This paper summarizes the current status of this effort, which has produced a trend curve calibrated using a database of over 1800 ΔT 41J values from the light water reactor surveillance programs in thirteen countries. (authors)

Modeling of cavity swelling-induced embrittlement in irradiated austenitic stainless steels

International Nuclear Information System (INIS)

Han, X.

2012-01-01

During long-time neutron irradiation occurred in Pressurized Water Reactors (PWRs), significant changes of the mechanical behavior of materials used in reactor core internals (made of 300 series austenitic stainless steels) are observed, including irradiation induced hardening and softening, loss of ductility and toughness. So far, much effect has been made to identify radiation effects on material microstructure evolution (dislocations, Frank loops, cavities, segregation, etc.). The irradiation-induced cavity swelling, considered as a potential factor limiting the reactor lifetime, could change the mechanical properties of materials (plasticity, toughness, etc.), even lead to a structure distortion because of the dimensional modifications between different components. The principal aim of the present PhD work is to study qualitatively the influence of cavity swelling on the mechanical behaviors of irradiated materials. A micromechanical constitutive model based on dislocation and irradiation defect (Frank loops) density evolution has been developed and implemented into ZeBuLoN and Cast3M finite element codes to adapt the large deformation framework. 3D FE analysis is performed to compute the mechanical properties of a polycrystalline aggregate. Furthermore, homogenization technique is applied to develop a Gurson-type model. Unit cell simulations are used to study the mechanical behavior of porous single crystals, by accounting for various effects of stress triaxiality, of void volume fraction and of crystallographic orientation, in order to study void effect on the irradiated material plasticity and roughness at polycrystalline scale. (author) [fr

Specificity in liquid metal induced embrittlement

CSIR Research Space (South Africa)

Fernandes, PJL

1996-12-01

Full Text Available One of the most intriguing features of liquid metal induced embrittlement (LMIE) is the observation that some liquid metal-solid metal couples are susceptible to embrittlement, while others appear to be immune. This is referred to as the specificity...

Mercury embrittlement of Cu-Al alloys under cyclic loading

Science.gov (United States)

Regan, T. M.; Stoloff, N. S.

1977-01-01

The effect of mercury on the room temperature, high cycle fatigue properties of three alloys: Cu-5.5 pct Al, Cu-7.3 pct Al, and Cu-6.3 pct Al-2.5 pct Fe has been determined. Severe embrittlement under cyclic loading in mercury is associated with rapid crack propagation in the presence of the liquid metal. A pronounced grain size effect is noted under mercury, while fatigue properties in air are insensitive to grain size. The fatigue results are discussed in relation to theories of adsorption-induced liquid metal embrittlement.

Postirradiation examination results for the Irradiation Effects Test IE-5

International Nuclear Information System (INIS)

Cook, T.F.; Ploger, S.A.; Hobbins, R.R.

1978-03-01

The results are presented of the postirradiation examination of four pressurized water reactor type fuel rods which were tested in-pile under a fast power ramp and film boiling operation during Irradiation Effects (IE) Test 5. The major objectives of this test were to evaluate the effects of simulated fission products on fuel rod behavior during a fast power ramp, to determine the effects of high initial internal pressure on a fuel rod during film boiling, and to assess fuel rod property changes that occur during film boiling in a fuel rod with previously irradiated cladding. The overall condition of the rods and changes that occurred in fuel and cladding as a result of the power ramp and film boiling operation, as determined from the postirradiation examination, are reported and analyzed. Effects of the simulated fission products on fuel rod behavior during a power ramp are discussed. The effect of high internal pressure on rod behavior during film boiling is evaluated. Cladding temperatures are estimated at various axial and circumferential locations. Cladding embrittlement by oxidation is also assessed

Microstructure and embrittlement of VVER 440 reactor pressure vessel steels; Microstructure et fragilisation des aciers de cuve des reacteurs nucleaires VVER 440

Energy Technology Data Exchange (ETDEWEB)

Hennion, A

1999-03-15

27 VVER 440 pressurised water reactors operate in former Soviet Union and in Eastern Europe. The pressure vessel, is made of Cr-Mo-V steel. It contains a circumferential arc weld in front of the nuclear core. This weld undergoes a high neutron flux and contains large amounts of copper and phosphorus, elements well known for their embrittlement potency under irradiation. The embrittlement kinetic of the steel is accelerated, reducing the lifetime of the reactor. In order to get informations on the microstructure and mechanical properties of these steels, base metals, HAZ, and weld metals have been characterized. The high amount of phosphorus in weld metals promotes the reverse temper embrittlement that occurs during post-weld heat treatment. The radiation damage structure has been identified by small angle neutron scattering, atomic probe, and transmission electron microscopy. Nanometer-sized clusters of solute atoms, rich in copper with almost the same characteristics as in western pressure vessels steels, and an evolution of the size distribution of vanadium carbides, which are present on dislocation structure, are observed. These defects disappear during post-irradiation tempering. As in western steels, the embrittlement is due to both hardening and reduction of interphase cohesion. The radiation damage specificity of VVER steels arises from their high amount of phosphorus and from their significant density of fine vanadium carbides. (author)

Proton irradiation effects on beryllium – A macroscopic assessment

Energy Technology Data Exchange (ETDEWEB)

Simos, Nikolaos, E-mail: simos@bnl.gov [Nuclear Sciences & Technology Department, Brookhaven National Laboratory, Upton, NY, 11973 (United States); Elbakhshwan, Mohamed [Nuclear Sciences & Technology Department, Brookhaven National Laboratory, Upton, NY, 11973 (United States); Zhong, Zhong [Photon Sciences, NSLS II, Brookhaven National Laboratory, Upton, NY, 11973 (United States); Camino, Fernando [Center for Functional Nanomaterials, Brookhaven National Laboratory, Upton, NY, 11973 (United States)

2016-10-15

Beryllium, due to its excellent neutron multiplication and moderation properties, in conjunction with its good thermal properties, is under consideration for use as plasma facing material in fusion reactors and as a very effective neutron reflector in fission reactors. While it is characterized by unique combination of structural, chemical, atomic number, and neutron absorption cross section it suffers, however, from irradiation generated transmutation gases such as helium and tritium which exhibit low solubility leading to supersaturation of the Be matrix and tend to precipitate into bubbles that coalesce and induce swelling and embrittlement thus degrading the metal and limiting its lifetime. Utilization of beryllium as a pion production low-Z target in high power proton accelerators has been sought both for its low Z and good thermal properties in an effort to mitigate thermos-mechanical shock that is expected to be induced under the multi-MW power demand. To assess irradiation-induced changes in the thermal and mechanical properties of Beryllium, a study focusing on proton irradiation damage effects has been undertaken using 200 MeV protons from the Brookhaven National Laboratory Linac and followed by a multi-faceted post-irradiation analysis that included the thermal and volumetric stability of irradiated beryllium, the stress-strain behavior and its ductility loss as a function of proton fluence and the effects of proton irradiation on the microstructure using synchrotron X-ray diffraction. The mimicking of high temperature irradiation of Beryllium via high temperature annealing schemes has been conducted as part of the post-irradiation study. This paper focuses on the thermal stability and mechanical property changes of the proton irradiated beryllium and presents results of the macroscopic property changes of Beryllium deduced from thermal and mechanical tests.

Effect of neutron irradiation on the properties of the repair welds of the 15Kh2MFA steel

International Nuclear Information System (INIS)

Morozov, A.M.; Khachaturyants, L.V.

1986-01-01

The authors studied the effect of neutron irradiation on the tendency of the metal belonging to the heat affected zone of the weld toward brittle fracture (an increase in the critical temperature of brittleness). For comparison, the authors studied the radiation embrittlement of the original base metal (steel 15Kh2MFA) subjected to the conventional heat treatment of the reactor frames consisting of hardening and high-temperature tempering. Along with these materials, the radiational embrittlement of the base metal in the rehardened condition without tempering was studied. It was concluded that the presence of the regions repaired according to this technology and located in the frame at the level of the reactor core does not pose the problem of decreased resistance to brittle fracture

Grain boundary embrittlement and cohesion enhancement in copper

Energy Technology Data Exchange (ETDEWEB)

Paxton, Anthony; Lozovoi, Alexander [Atomistic Simulation Centre, Queen' s University Belfast, BT7 1NN (United Kingdom); Schweinfest, Rainer [Science+Computing ag, Hagellocher Weg 71-5, 720270 T ubingen (Germany); Finnis, Michael [Imperial College London, Exhibition Road, London SW7 2AZ (United Kingdom)

2008-07-01

There has been a long standing debate surrounding the mechanism of grain boundary embrittlement and cohesion enhancement in metals. Embrittlement can lead to catastrophic failure such as happened in the Hinkley Point disaster, or indeed in the case of the Titanic. This kind of embrittlement is caused by segregation of low solubility impurities to grain boundaries. While the accepted wisdom is that this is a phenomenon driven by electronic or chemical factors, using language such as charge transfer and electronegativity difference; we believe that in copper, at least, both cohesion enhancement and reduction are caused by a simple size effect. We have developed a theory that allows us to separate unambiguously, if not uniquely, chemical and structural factors. We have studied a large number of solutes in copper using first principles atomistic simulation to support this argument, and the results of these calculations are presented here.

Effect of microstructure on the susceptibility of a 533 steel to temper embrittlement

International Nuclear Information System (INIS)

Raoul, S.; Marini, B.; Pineau, A.

1998-01-01

In ferritic steels, brittle fracture usually occurs at low temperature by cleavage. However the segregation of impurities (P, As, Sn etc..) along prior γ grain boundaries can change the brittle fracture mode from transgranular to intergranular. In quenched and tempered steels, this segregation is associated with what is called the temper-embrittlement phenomenon. The main objective of the present study is to investigate the influence of the as-quenched microstructure (lower bainite or martensite) on the susceptibility of a low alloy steel (A533 cl.1) to temper-embrittlement. Dilatometric tests were performed to determine the continous-cooling-transformation (CCT) diagram of the material and to measure the critical cooling rate (V c ) for a martensitic quench. Then subsized Charpy V-notched specimens were given various cooling rates from the austenitization temperature to obtain a wide range of as-quenched microstructures, including martensite and bainite. These specimens were subsequently given a heat treatment to develop temper embrittlement and tested to measure the V-notch fracture toughness at -50 C. The fracture surfaces were examined by SEM. It is shown that martensitic microstructures are more susceptible to intergranular embrittlement than bainitic microstructures. These observed microstructural influences are briefly discussed. (orig.)

Neutron irradiation effect on the strength of jointed Ti-6Al-4V alloy

International Nuclear Information System (INIS)

Ishiyama, Shintaro; Miya, Naoyuki

2002-01-01

In order to investigate applicability of Ti alloy to large scaled structural material for fusion reactors, irradiation effect on the mechanical properties of Ti-6Al-4V alloy and its TIG welded material was investigated after neutron irradiation (temperature: 746-788K, fluence: 2.8 x 10 23 n/m 2 (>0.18 MeV). The following results were obtained. (1) Irradiated Ti alloy shows about 20-30% increase of its tensile strength and large degradation of fracture elongation, comparing with those of unirradiated Ti alloy. (2) TIG welded material behaves as Ti alloy in its tensile test, however, shows 30% increase of area reduction in 373-473K, whereas 1/2 degradation of area reduction over 600K. (3) Irradiated TIG welded material behaves heavier embrittlement than that of irradiated Ti alloy. (4) Charpy impact properties of un- and irradiated Ti alloys shift to ductile from brittle fracture and transition temperature shift, ΔT was estimated as about 100K. (5) Remarkable increase of hardness was found, especially in HAZ of TIG welded material after irradiation. (author)

Influence of sulfur, phosphorus, and antimony segregation on the intergranular hydrogen embrittlement of nickel

International Nuclear Information System (INIS)

Bruemmer, S.M.; Baer, D.R.; Jones, R.H.; Thomas, M.T.

1983-01-01

The effectiveness of sulfur, phosphorus, and antimony in promoting the intergranular embrittlement of nickel was investigated using straining electrode tests in 1N H 2 SO 4 at cathodic potentials. Sulfur was found to be the critical grain boundary segregant due to its large enrichment at grain boundaries (10 4 to 10 5 times the bulk content) and the direct relationship between sulfur coverage and hydrogeninduced intergranular failure. Phosphorus was shown to be significantly less effective than sulfur or antimony in inducing the intergranular hydrogen embrittlement of nickel. The addition of phosphoru to nickel reduced the tendency for intergranular fracture and improved ductility because phosphoru segregated strongly to grain interfaces and limited sulfur enrichment. The hydrogen embrittling potency of antimony was also less than that of sulfur while its segregation propensity was considerably less. It was found that the effectiveness of segregated phosphorus and antimony in prompting inter granular embrittlement vs that of sulfur could be expressed in terms of an equivalent grain boundary sulfur coverage. The relative hydrogen embrittling potencies of sulfur, phosphorus, and antimony are discussed in reference to general mechanisms for the effect of impurity segregation on hydrogeninduced intergranular fracture

Surveillance as a complement to irradiation embrittlement studies: Status and needs

International Nuclear Information System (INIS)

Steele, L.E.

1977-01-01

The history of the study of radiation embrittlement of reactor pressure vessel steels has gone through three stages in the USA. 1) A scientific curiosity. 2) Empirical or laboratory evaluation of typical steels, and 3) Integration of the scientific and empirical to advance status and evolve standard techniques. The current stage is one in which surveillance data compliments the laboratory studies which characterized Stage 3. The early USA surveillance programs were generally analyzed by the same people who were the primary laboratory investigators. An effort must be made to continue this type of collaboration as a useful two-way learning procedure though it will become more and more difficult as nuclear power is broadly commercialized. The current status of both types of USA programs will be presented to encourage the most advantageous use of data from both sources. At this time about 25 USA nuclear power reactors have operated long enough to have provided initial surveillance or dosimetry results. An effort will be made to summarize the general status of these in order to: 1) Provide complimentary data to laboratory studies. 2) Assess directions in handling the problems of radiation embrittlement. 3) Note lessons learned for improving surveillance efforts in the future. 4) Identify possible research tasks for the future to support in-service surveillance and other measures. 5) Justify facts advancing surveillance requirements to status of national codes and standards. 6) Justify facts requiring changes in current national codes and standards. A plan will be presented along with an introduction of each member of the USA delegation for systematic presentation of the status of reactor vessel surveillance in the USA. (author)

Hydrogen embrittlement and galvanic corrosion of titanium alloys

Energy Technology Data Exchange (ETDEWEB)

Soh, Jeong Ryong; Jeong, Y. H.; Choi, B. K.; Baek, J. H.; Hwang, D. Y.; Choi, B. S.; Lee, D. J

2000-06-01

The material properties including the fracture behavior of titanium alloys used as a steam generator tube in SMART can be degraded de to the hydrogen embrittlement and the galvanic corrosion occurring as a result of other materials in contact with titanium alloys in a conducting corrosive environment. In this report the general concepts and trends of hydrogen embrittlement are qualitatively described to adequately understand and expect the fracture behavior from hydrogen within the bulk of materials and under hydrogen containing environments because hydrogen embrittlement may be very complicated process. And the characteristics of galvanic corrosion closely related to hydrogen embrittlement is qualitatively based on wimple electrochemical theory.

Hydrogen embrittlement and galvanic corrosion of titanium alloys

International Nuclear Information System (INIS)

Soh, Jeong Ryong; Jeong, Y. H.; Choi, B. K.; Baek, J. H.; Hwang, D. Y.; Choi, B. S.; Lee, D. J.

2000-06-01

The material properties including the fracture behavior of titanium alloys used as a steam generator tube in SMART can be degraded de to the hydrogen embrittlement and the galvanic corrosion occurring as a result of other materials in contact with titanium alloys in a conducting corrosive environment. In this report the general concepts and trends of hydrogen embrittlement are qualitatively described to adequately understand and expect the fracture behavior from hydrogen within the bulk of materials and under hydrogen containing environments because hydrogen embrittlement may be very complicated process. And the characteristics of galvanic corrosion closely related to hydrogen embrittlement is qualitatively based on wimple electrochemical theory

Effect of microstructure on the susceptibility of a 533 steel to temper embrittlement

Energy Technology Data Exchange (ETDEWEB)

Raoul, S.; Marini, B. [CEA Centre d`Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France). Service de Recherches Metallurgiques Appliquees; Pineau, A. [CNRS, Evry (France). Centre de Materiaux

1998-11-01

In ferritic steels, brittle fracture usually occurs at low temperature by cleavage. However the segregation of impurities (P, As, Sn etc..) along prior {gamma} grain boundaries can change the brittle fracture mode from transgranular to intergranular. In quenched and tempered steels, this segregation is associated with what is called the temper-embrittlement phenomenon. The main objective of the present study is to investigate the influence of the as-quenched microstructure (lower bainite or martensite) on the susceptibility of a low alloy steel (A533 cl.1) to temper-embrittlement. Dilatometric tests were performed to determine the continous-cooling-transformation (CCT) diagram of the material and to measure the critical cooling rate (V{sub c}) for a martensitic quench. Then subsized Charpy V-notched specimens were given various cooling rates from the austenitization temperature to obtain a wide range of as-quenched microstructures, including martensite and bainite. These specimens were subsequently given a heat treatment to develop temper embrittlement and tested to measure the V-notch fracture toughness at -50 C. The fracture surfaces were examined by SEM. It is shown that martensitic microstructures are more susceptible to intergranular embrittlement than bainitic microstructures. These observed microstructural influences are briefly discussed. (orig.) 11 refs.

Effects of strain rate, stress condition and environment on iodine embrittlement of Ziracloy-2

International Nuclear Information System (INIS)

Une, K.

1979-01-01

Iodine stress corrosion cracking (SCC) susceptibility of Zircaloy became higher with decreasing strain rate. Critical strain rate, below which high SCC severity was observed, substantially depended on Zircaloy stress condition. This strain rate (7 x 10 -3 min -1 ) under plane strain condition was about 3.5 times as fast as that (2 x 10 -3 min -1 ) under uniaxial condition. The maximum iodine embrittlement in Zircaloy was found in stress ratio α (axial/tangential stress) range of 0.5 to 0.7. No embrittlement occurred at α = infinity because of its texture effect. The SCC fracture stresses were about 39 kg/mm 2 for unirradiated and stress-relieved material, and about 34 kg/mm 2 for recrystallized material, whose ratios to yield strength of each material were 0.8 and 1.2. Impurity gases of oxygen and moisture in the iodine had the effects of reducing Zircaloy SCC susceptibility. Stress-relieved material was more sensitive to environmental impurities than recrystallized material

High-temperature helium embrittlement (T>=0,45Tsub(M)) of metals

International Nuclear Information System (INIS)

Batfalsky, P.

1984-06-01

High temperature helium embrittlement, swelling and irradiation creep are the main technical problem of fusion reactor materials. The expected helium production will be very high. The helium produced by (n,α)-processes precipitates into helium bubbles because its solubility in solid metals is very low. Under continuous helium production at high temperature and stress the helium bubbles grow and lead to intergranular early failure. Solution annealed foil specimens of austenitic stainless steel AISI 316 were implanted with α-particles: 1. during creep tests at 1023 K (''in-beam'' test) 2. before the creep tests at high temperature (1023 K). The creep tests have been performed within large ranges of test parameter, e.g. applied stress, temperature, helium implantation rate and helium concentration. After the creep tests the microstructure was investigated using scanning (SEM) and transmission (TEM) electron microscopy. All the helium implanted specimens showed high temperature helium embrittlement, i.e. reduction of rupture time tsub(R) and ductility epsilonsub(R) and evidence of intergranular brittle fracture. The ''in-beam'' creep tests showed greater reduction of rupture time tsub(R) and ductility than the preimplanted creep tests. The comparison of this experimentally obtained data with various theoretical models of high temperature helium embrittlement showed that within the investigated parameter ranges the mechanism controlling the life time of the samples is probably the gas driven stable growth of the helium bubbles within the grain boundaries. (orig.)

Multiscale modelling and experimentation of hydrogen embrittlement in aerospace materials

Science.gov (United States)

Jothi, Sathiskumar

Pulse plated nickel and nickel based superalloys have been used extensively in the Ariane 5 space launcher engines. Large structural Ariane 5 space launcher engine components such as combustion chambers with complex microstructures have usually been manufactured using electrodeposited nickel with advanced pulse plating techniques with smaller parts made of nickel based superalloys joined or welded to the structure to fabricate Ariane 5 space launcher engines. One of the major challenges in manufacturing these space launcher components using newly developed materials is a fundamental understanding of how different materials and microstructures react with hydrogen during welding which can lead to hydrogen induced cracking. The main objective of this research has been to examine and interpret the effects of microstructure on hydrogen diffusion and hydrogen embrittlement in (i) nickel based superalloy 718, (ii) established and (iii) newly developed grades of pulse plated nickel used in the Ariane 5 space launcher engine combustion chamber. Also, the effect of microstructures on hydrogen induced hot and cold cracking and weldability of three different grades of pulse plated nickel were investigated. Multiscale modelling and experimental methods have been used throughout. The effect of microstructure on hydrogen embrittlement was explored using an original multiscale numerical model (exploiting synthetic and real microstructures) and a wide range of material characterization techniques including scanning electron microscopy, 2D and 3D electron back scattering diffraction, in-situ and ex-situ hydrogen charged slow strain rate tests, thermal spectroscopy analysis and the Varestraint weldability test. This research shows that combined multiscale modelling and experimentation is required for a fundamental understanding of microstructural effects in hydrogen embrittlement in these materials. Methods to control the susceptibility to hydrogen induced hot and cold cracking and

Hydrogen embrittlement due to hydrogen-inclusion interactions

International Nuclear Information System (INIS)

Yu, H.Y.; Li, J.C.M.

1976-01-01

Plastic flow around inclusions creates elastic misfit which attracts hydrogen towards the regions of positive dilatation. Upon decohesion of the inclusion-matrix interface, the excess hydrogen escapes into the void and can produce sufficient pressure to cause void growth by plastic deformation. This mechanism of hydrogen embrittlement can be used to understand the increase of ductility with temperature, the decrease of ductility with hydrogen content, and the increase of ductility with the ultimate strength of the matrix. An examination of the effect of the shape of spheroid inclusion reveals that rods are more susceptible to hydrogen embrittlement than disks. The size of the inclusion is unimportant while the volume fraction of inclusions plays the usual role

Low dose irradiation effects on DIN 1.4948 mechanical properties

International Nuclear Information System (INIS)

Schaaf, B. van der; Vries, M.I. de

For the SNR 300 the licensing authorities require the determination of the lower boundaries of post-irradiation mechanical properties for DIN 1.4948 parent metal and welded joints. It has been established that with decreasing strain rate the post-irradiation tensile ductility decreases. A transition strain rate has been observed, above which there is no effect of irradiation on ductility. The transition strain rate increases with increasing temperature. Coarse grained heats show lower ultimate tensile strength above 800 K than fine grained heats. There is no significant effect of irradiation on load controlled high cycle fatigue with frequencies of 1 Hz or higher. In low cycle fatigue numbers of cycles to failure decrease with decreasing frequency. Increasing the test temperature reduces the number of cycles to failure even more. The frequency effect is more evident at 823 K. Parent metal has a better fatigue resistance than welded joints in unirradiated and irradiated condition. Creep strength is reduced by irradiation due to loss of ductility. It is shown that with increasing grain size the rupture strength decreases. The ductility of welded joints after irradiation is low, in some cases as low as 0.5% creep strain. After irradiation, tensile, creep and fatigue fracture surfaces show many more intergranular features than in the equivalent unirradiated condition. The promotion of intergranular fracture by irradiation and the consequent degradation of low strain rate mechanical properties is explained by the presence of helium on grain boundaries. Several measures to increase the helium content threshold can be taken, such as grain refinement, homogeneous boron distribution and promotion of helium bubble initiation. In cases where helium embrittlement is encountered, life reduction factors on unirradiated material properties must be applied

Microstructural evolution in reactor pressure vessel steel under neutron irradiation

Energy Technology Data Exchange (ETDEWEB)

Ohno, Katsumi; Fukuya, Koji [Institute of Nuclear Safety System Inc., Seika, Kyoto (Japan)

2000-09-01

Understanding microstructural changes in reactor pressure vessel steels is important in order to evaluate radiation-induced embrittlement, one of the major aging phenomena affecting the extension of plant life. In this study, actual surveillance test specimens and samples of rector vessel low-alloy steel (A533B steel) irradiated in a research reactor were examined using state-of-the-art techniques to clarify the neutron flux effect on the microstructural changes. These techniques included small angle neutron scattering and atom probes. Microstructural changes which are considered to be the main factors affecting embrittlement, including the production of copper-rich precipitates and the segregation of impurity elements, were confirmed by the results of the study. In addition, the mechanical properties were predicted based on the obtained quantitative data such as the diameters of precipitates. Consequently, the hardening due to irradiation was almost simulated. (author)

The low-temperature aging embrittlement in a 2205 duplex stainless steel

International Nuclear Information System (INIS)

Weng, K.L.; Chen, H.R.; Yang, J.R.

2004-01-01

The effect of isothermal treatment (at temperatures ranging between 400 and 500 deg. C) on the embrittlement of a 2205 duplex stainless steel (with 45 ferrite-55 austenite, vol.%) has been investigated. The impact toughness and hardness of the aged specimens were measured, while the corresponding fractography was studied. The results show that the steel is susceptible to severe embrittlement when exposed at 475 deg. C; this aging embrittlement is analogous with that of the ferritic stainless steels, which is ascribed to the degenerated ferrite phase. High-resolution transmission electron microscopy reveals that an isotropic spinodal decomposition occurred during aging at 475 deg. C in the steel studied; the original δ-ferrite decomposed into a nanometer-scaled modulated structure with a complex interconnected network, which contained an iron-rich BCC phase (α) and a chromium-enriched BCC phase (α'). It is suggested that the locking of dislocations in the modulated structure leads to the severe embrittlement

Liquid and Solid Metal Embrittlement.

Science.gov (United States)

1981-09-05

example, embrittlement of AISI 4140 steel begins at T/T, - 0.75 for cadmium, and 0.85 for lead and tin environments (2). In a few cases, e.g. zinc...has recently proposed, however, that liquid zinc can penetrate to very near the tip of a sharp crack in 4140 steel, based upon both direct observation...long could be detected, was observed in delayed failure experi- ments on unnotched 4140 steel, in the quenched and tempered condi- tion, embrittled by

Effect of irradiation on the tensile properties of niobium-base alloys

International Nuclear Information System (INIS)

Grossbeck, M.L.; Heestand, R.L.; Atkin, S.D.

1986-11-01

The alloys Nb-1Zr and PWC-11 (Nb-1Zr-0.1C) were selected as prime candidate alloys for the SP-100 reactor. Since the mechanical properties of niobium alloys irradiated to end-of-life exposure levels of about 2 x 10 26 neutrons/m 2 (E > 0.1 MeV) at temperatures above 1300 K were not available, an irradiation experiment (B-350) in EBR-II was conducted. Irradiation creep, impact properties, bending fatigue, and tensile properties were investigated; however, only tensile properties will be reported in this paper. The tensile properties were studied since they easily reveal the common irradiation phenomena of hardening and embrittlement. Most attention was directed to testing at the irradiation temperature. Further testing was conducted at lower temperatures in order to scope the behavior of the alloys in cooldown conditions

Nanocrystalline Steels’ Resistance to Hydrogen Embrittlement

Directory of Open Access Journals (Sweden)

Skołek E.

2015-04-01

Full Text Available The aim of this study is to determine the susceptibility to hydrogen embrittlement in X37CrMoV5-1 steel with two different microstructures: a nanocrystalline carbide-free bainite and tempered martensite. The nanobainitic structure was obtained by austempering at the bainitic transformation zone. It was found, that after hydrogen charging, both kinds of microstructure exhibit increased yield strength and strong decrease in ductility. It has been however shown that the resistance to hydrogen embrittlement of X37CrMoV5-1 steel with nanobainitic structure is higher as compared to the tempered martensite. After hydrogen charging the ductility of austempered steel is slightly higher than in case of quenched and tempered (Q&T steel. This effect was interpreted as a result of phase composition formed after different heat treatments.

Irradiation environment and materials behavior

International Nuclear Information System (INIS)

Ishino, Shiori

1992-01-01

Irradiation environment is unique for materials used in a nuclear energy system. Material itself as well as irradiation and environmental conditions determine the material behaviour. In this review, general directions of research and development of materials in an irradiation environment together with the role of materials science are discussed first, and then recent materials problems are described for energy systems which are already existing (LWR), under development (FBR) and to be realized in the future (CTR). Topics selected are (1) irradiation embrittlement of pressure vessel steels for LWRs, (2) high fluence performance of cladding and wrapper materials for fuel subassemblies of FBRs and (3) high fluence irradiation effects in the first wall and blanket structural materials of a fusion reactor. Several common topics in those materials issues are selected and discussed. Suggestions are made on some elements of radiation effects which might be purposely utilized in the process of preparing innovative materials. (J.P.N.) 69 refs

Effects of metallurgical variables on hydrgen embrittlement in types 316, 321, and 347 stainless steels

International Nuclear Information System (INIS)

Rozenak, P.; Eliezer, D.

1984-01-01

Hydrogen embrittlement of 316, 321 and 347 types austenitic stainless steels has been studied by charging thin tensile specimens with hydrogen through cathodic polarization. Throughout this study we have compared solution annealed samples having various prior austenitic grain-size with samples given the additional sensitization treatment. The results show that refined grains improves the resistance to hydrogen cracking regardless of the failure mode. The sensitized specimens were predominantly intergranular, while the annealed specimens show massive regions of microvoid coalescence producing ductile rupture. 347 type stainless steel is much more susceptible to hydrogen embrittlement than 321 type steel, and 316 type is the most resistant to hydrogen embrittlement. the practical implication of the experimental conclusions are discussed

Irradiation Effects Test Series: Test IE-3. Test results report

International Nuclear Information System (INIS)

Farrar, L.C.; Allison, C.M.; Croucher, D.W.; Ploger, S.A.

1977-10-01

The objectives of the test reported were to: (a) determine the behavior of irradiated fuel rods subjected to a rapid power increase during which the possibility of a pellet-cladding mechanical interaction failure is enhanced and (b) determine the behavior of these fuel rods during film boiling following this rapid power increase. Test IE-3 used four 0.97-m long pressurized water reactor type fuel rods fabricated from previously irradiated fuel. The fuel rods were subjected to a preconditioning period, followed by a power ramp to 69 kW/m at a coolant mass flux of 4920 kg/s-m 2 . After a flow reduction to 2120 kg/s-m 2 , film boiling occurred on the fuel rods. One rod failed approximately 45 seconds after the reactor was shut down as a result of cladding embrittlement due to extensive cladding oxidation. Data are presented on the behavior of these irradiated fuel rods during steady-state operation, the power ramp, and film boiling operation. The effects of a power ramp and power ramp rates on pellet-cladding interaction are discussed. Test data are compared with FRAP-T3 computer model calculations and data from a previous Irradiation Effects test in which four irradiated fuel rods of a similar design were tested. Test IE-3 results indicate that the irradiated state of the fuel rods did not significantly affect fuel rod behavior during normal, abnormal (power ramp of 20 kW/m per minute), and accident (film boiling) conditions

Role of vanadium carbide traps in reducing the hydrogen embrittlement susceptibility of high strength alloy steels. Final report

Energy Technology Data Exchange (ETDEWEB)

Spencer, G.L.; Duquette, D.J.

1998-08-01

High strength alloy steels typically used for gun steel were investigated to determine their susceptibility to hydrogen embrittlement. Although AISI grade 4340 was quite susceptible to hydrogen embrittlement, ASTM A723 steel, which has identical mechanical properties but slightly different chemistries, was not susceptible to hydrogen embrittlement when exposed to the same conditions. The degree of embrittlement was determined by conducting notched tensile testing on uncharged and cathodically charged specimens. Chemical composition was modified to isolate the effect of alloying elements on hydrogen embrittlement susceptibility. Two steels-Modified A723 (C increased from 0.32% to 0.40%) and Modified 4340 (V increased from 0 to O.12%) were tested. X-ray diffraction identified the presence of vanadium carbide, V{sub 4}C{sub 3}, in A-23 steels, and subsequent hydrogen extraction studies evaluated the trapping effect of vanadium carbide. Based on these tests, it was determined that adding vanadium carbide to 4340 significantly decreased hydrogen embrittlement susceptibility because vanadium carbide traps ties up diffusible hydrogen. The effectiveness of these traps is examined and discussed in this paper.

Investigation of moisture-induced embrittlement of iron aluminides. Final report

Energy Technology Data Exchange (ETDEWEB)

Alven, D.A.; Stoloff, N.S. [Rensselaer Polytechnic Inst., Troy, NY (United States). Materials Engineering Dept.

1997-06-05

Iron-aluminum alloys with 28 at.% Al and 5 at.% Cr were shown to be susceptible to hydrogen embrittlement by exposure to both gaseous hydrogen and water vapor. This study examined the effect of the addition of zirconium and carbon on the moisture-induced hydrogen embrittlement of an Fe{sub 3}Al,Cr alloy through the evaluation of tensile properties and fatigue crack growth resistance in hydrogen gas and moisture-bearing air. Susceptibility to embrittlement was found to vary with the zirconium content while the carbon addition was found to only affect the fracture toughness. Inherent fatigue crack growth resistance and fracture toughness, as measured in an inert environment, was found to increase with the addition of 0.5 at.% Zr. The combined addition of 0.5 at.% Zr and carbon only increased the fracture toughness. The addition of 1 at.% Zr and carbon was found to have no effect on the crack growth rate when compared to the base alloy. Susceptibility to embrittlement in moisture-bearing environments was found to decrease with the addition of 0.5 at.% Zr. In gaseous hydrogen, the threshold value of the Zr-containing alloys was found to increase above that found in the inert environment while the crack growth resistance was much lower. By varying the frequency of fatigue loading, it was shown that the corrosion fatigue component of the fatigue crack growth rate in an embrittling environment displays a frequency dependence. Hydrogen transport in iron aluminides was shown to occur primarily by a dislocation-assisted transport mechanism. This mechanism, in conjunction with fractography, indicates that the zirconium-containing precipitates act as traps for the hydrogen that is carried along by the dislocations through the lattice.

Cracking behavior of thermally aged and irradiated CF-8 cast austenitic stainless steel

Energy Technology Data Exchange (ETDEWEB)

Chen, Y., E-mail: Yiren_Chen@anl.gov [Argonne National Laboratory, 9700 S. Cass Ave, Argonne, IL 60439 (United States); Alexandreanu, B.; Chen, W.-Y.; Natesan, K. [Argonne National Laboratory, 9700 S. Cass Ave, Argonne, IL 60439 (United States); Li, Z.; Yang, Y. [University of Florida, Gainesville, FL 32611 (United States); Rao, A.S. [US Nuclear Regulatory Commission, 11545 Rockville Pike, Rockville, MD 20852 (United States)

2015-11-15

To assess the combined effect of thermal aging and neutron irradiation on the cracking behavior of CF-8 cast austenitic stainless steel, crack growth rate (CGR) and fracture toughness J-R curve tests were carried out on compact-tension specimens in high-purity water with low dissolved oxygen. Both unaged and thermally aged specimens were irradiated at ∼320 °C to 0.08 dpa. Thermal aging at 400 °C for 10,000 h apparently had no effect on the corrosion fatigue and stress corrosion cracking behavior in the test environment. The cracking susceptibility of CF-8 was not elevated significantly by neutron irradiation at 0.08 dpa. Transgranular cleavage-like cracking was the main fracture mode during the CGR tests, and a brittle morphology of delta ferrite was often seen on the fracture surfaces at the end of CGR tests. The fracture toughness J-R curve tests showed that both thermal aging and neutron irradiation can induce significant embrittlement. The loss of fracture toughness due to neutron irradiation was more pronounced in the unaged than aged specimens. After neutron irradiation, the fracture toughness values of the unaged and aged specimens were reduced to a similar level. G-phase precipitates were observed in the aged and irradiated specimens with or without prior aging. The similar microstructural changes resulting from thermal aging and irradiation suggest a common microstructural mechanism of inducing embrittlement in CF-8.

Influence of a cyclic load on the embrittlement kinetics of alloys by the example of the 475 C embrittlement of duplex steel and the dynamic embrittlement of a nickel base alloy; Einfluss einer zyklischen Belastung auf die Versproedungskinetik von Legierungen am Beispiel der 475 C-Versproedung von Duplexstahl und der dynamischen Versproedung einer Nickelbasislegierung

Energy Technology Data Exchange (ETDEWEB)

Wackermann, Ken

2015-07-07

The objective of this study was to investigate the dependence of high temperature embrittlement mechanisms on high temperature fatigue and vice versa. As model embrittlement mechanisms the 475 C Embrittlement of ferritic austenitic duplex stainless steel (1.4462) and the Dynamic Embrittlement of nickel-based superalloys (IN718) were selected. The 475 C Embrittlement is a thermally activated decomposition of the ferritic phase which hardens the material. In contrast to this a cyclic plastic deformation weakens the steel by a deformation-induced dissolution of the decomposition. Fatigue tests with different frequencies, loading amplitudes at room temperature and at 475 C with Duplex Stainless Steel in different states of embrittlement show that the ongoing 475 C Embrittlement and the deformation-induced dissolution are competing mechanisms. It depends on the frequency, the loading amplitude and the temperature which mechanism is dominant. Applying the model of the yield stress distribution function to the hysteresis branches of the fatigue tests allows an analysis of the fatigue behaviour of each phase individually. This analysis shows that the global fatigue behaviour for the test conditions applied in this study is mainly controlled by the ferritic phase. According to the existing understanding of Dynamic Embrittlement it is an oxygen grain boundary diffusion arising by tensile stress at elevated temperatures with the result of a fast intercrystalline crack propagation. In reference tests under vacuum conditions without oxygen grain boundary diffusion, a slow transcrystalline fracture appears. To analyse the Dynamic Embrittlement, the crack propagation was tested at 650 C with different frequencies and superimposed hold times in the fatigue cycle at maximum stress. The results shows that the existing model of Dynamic Embrittlement needs to be adapted to the effects of cyclic plastic deformation. In hold times, the oxygen grain boundary diffusion in front of the

Disk-bend ductility tests for irradiated materials

International Nuclear Information System (INIS)

Klueh, R.L.; Braski, D.N.

1984-01-01

We modified the HEDL disk-bend test machine and are using it to qualitatively screen alloys that are susceptible to embrittlement caused by irradiation. Tests designed to understand the disk-bend test in relation to a uniaxial test are discussed. Selected results of tests of neutron-irradiated material are also presented

Atom probe study of the microstructural evolution induced by irradiation in Fe-Cu ferritic alloys and pressure vessel steels; Etude a la sonde atomique de l`evolution microstructurale sous irradiation d`alliages ferritiques Fe-Cu et d`aciers de cuve REP

Energy Technology Data Exchange (ETDEWEB)

Pareige, P

1996-04-01

Pressure vessel steels used in pressurized water reactors are low alloyed ferritic steels. They may be prone to hardening and embrittlement under neutron irradiation. The changes in mechanical properties are generally supposed to result from the formation of point defects, dislocation loops, voids and/or copper rich clusters. However, the real nature of the irradiation induced-damage in these steels has not been clearly identified yet. In order to improve our vision of this damage, we have characterized the microstructure of several steels and model alloys irradiated with electrons and neutrons. The study was performed with conventional and tomographic atom probes. The well known importance of the effects of copper upon pressure vessel steel embrittlement has led us to study Fe-Cu binary alloys. We have considered chemical aging as well as aging under electron and neutron irradiations. The resulting effects depend on whether electron or neutron irradiations ar used for thus. We carried out both kinds of irradiation concurrently so as to compare their effects. We have more particularly considered alloys with a low copper supersaturation representative of that met with the French vessel alloys (0.1% Cu). Then, we have examined steels used on French nuclear reactor pressure vessels. To characterize the microstructure of CHOOZ A steel and its evolution when exposed to neutrons, we have studied samples from the reactor surveillance program. The results achieved, especially the characterization of neutron-induced defects have been compared with those for another steel from the surveillance program of Dampierre 2. All the experiment results obtained on model and industrial steels have allowed us to consider an explanation of the way how the defects appear and grow, and to propose reasons for their influence upon steel embrittlement. (author). 3 appends.

Correlation between microstructural features and mechanical properties of irradiated LONGLIFE RPV steels

International Nuclear Information System (INIS)

Serrano, M.; Hermandez-Mayoral, E.; Bergner, F.; Viehrig, H.W.; Altstadt, E.; Radiguet, B.; Lim, J.H.; Grovenor, C.R.M.; Meslin, E.; Van Renterghem, W.; Chaouadi, R.; Ortner, S.; Hein, H.; Gillemot, F.; Todeschini, P.; Planman, T.; Wilford, K.; Kocik, J.; Brumovsky, M.; Ruoden, J.

2015-01-01

The possibility of extending the operational life of reactor pressure vessels (RPV) up to 80 years presents the problem of the availability of materials irradiated at high neutron fluence and low neutron flux. The ability of the existing trend curves to predict high fluence embrittlement is a question of debate, and a critical analysis of these curves should be based on a consistent microstructural examination of irradiated materials. Within the LONGLIFE 7FWP, neutron irradiated RPV materials, relevant for long term operation, some of them coming from surveillance programs, have been characterized by means of a combination of microstructural techniques (APT, SANS, TEM) and mechanical tests (hardness, tensile, impact and fracture toughness). In this paper the analysis of the links between microstructural features (solute nano-clusters, dislocation loops and voids) and hardening and embrittlement measurements by mechanical testing, is presented. Current hardening models, based on the contribution of precipitates, or nano-clusters, seem to underestimate irradiation hardening for very high fluences, mainly when matrix damage (dislocation loops) is observed. Regarding chemical composition effects, the predominant role of Ni and the synergism between Ni-Mn and Si are also identified. Low-Cu alloys show a threshold value of radiation induced features to produce an effect on mechanical properties which calls for further in-depth analyses. (authors)

Hydrogen embrittlement of Zr-2.5Nb PT with temperature

International Nuclear Information System (INIS)

Oh, Dong Joon; Ahn, Sang Bok; Kim, Young Suk

2003-01-01

The aim of this study is to investigate the effect of hydrogen embrittlement of Zr-2.5Nb CANDU pressure tube. The tests were performed at three hydrogen contents for transverse tensile and CCT specimens while the test temperatures were changed (RT to 300 .deg. C). The specimens were directly machined from the tube retaining original curvature using electric discharge machine. Both the transverse tensile and the fracture toughness tests showed the hydrogen embrittlement clearly at RT but this phenomenon was disappeared while the test temperature arrived over 250 .deg. C

Effect of Low-Temperature Sensitization on Hydrogen Embrittlement of 301 Stainless Steel

OpenAIRE

Chieh Yu; Ren-Kae Shiue; Chun Chen; Leu-Wen Tsay

2017-01-01

The effect of metastable austenite on the hydrogen embrittlement (HE) of cold-rolled (30% reduction in thickness) 301 stainless steel (SS) was investigated. Cold-rolled (CR) specimens were hydrogen-charged in an autoclave at 300 or 450 °C under a pressure of 10 MPa for 160 h before tensile tests. Both ordinary and notched tensile tests were performed in air to measure the tensile properties of the non-charged and charged specimens. The results indicated that cold rolling caused the transforma...

Post-irradiation annealing of coarse-grained model alloys

Energy Technology Data Exchange (ETDEWEB)

Ray, P H.N.; Wilson, C; McElroy, R J [AEA Reactor Services, Harwell (United Kingdom)

1994-12-31

Thermal ageing and irradiation studies have been carried out on three model alloys (JPC, JPB, JPG) that have identical compositions except for different levels of phosphorus and/or copper. They have been irradiated in three conditions, as-received, heat treated to produce a coarse grained microstructure (similar to heat-affected-zone), and in this condition further aged at 450 C to produce a temper embrittled condition. One of the alloy have been subject to a post-irradiation anneal. The effect of these treatments on mechanical property changes has been characterized by Charpy testing and Vickers hardness measurements; the phosphorus segregation has been studied by a combination of STEM and Auger techniques.

Lessons Learned From Developing Reactor Pressure Vessel Steel Embrittlement Database

Energy Technology Data Exchange (ETDEWEB)

Wang, Jy-An John [ORNL

2010-08-01

Materials behaviors caused by neutron irradiation under fission and/or fusion environments can be little understood without practical examination. Easily accessible material information system with large material database using effective computers is necessary for design of nuclear materials and analyses or simulations of the phenomena. The developed Embrittlement Data Base (EDB) at ORNL is this comprehensive collection of data. EDB database contains power reactor pressure vessel surveillance data, the material test reactor data, foreign reactor data (through bilateral agreements authorized by NRC), and the fracture toughness data. The lessons learned from building EDB program and the associated database management activity regarding Material Database Design Methodology, Architecture and the Embedded QA Protocol are described in this report. The development of IAEA International Database on Reactor Pressure Vessel Materials (IDRPVM) and the comparison of EDB database and IAEA IDRPVM database are provided in the report. The recommended database QA protocol and database infrastructure are also stated in the report.

The role of point defect clusters in reactor pressure vessel embrittlement

International Nuclear Information System (INIS)

Stoller, R.E.

1993-01-01

Radiation-induced point defect clusters (PDC) are a plausible source of matrix hardening in reactor pressure vessel (RPV) steels in addition to copper-rich precipitates. These PDCs can be of either interstitial or vacancy type, and could exist in either 2 or 3-D shapes, e.g. small loops, voids, or stacking fault tetrahedra. Formation and evolution of PDCs are primarily determined by displacement damage rate and irradiation temperature. There is experimental evidence that size distributions of these clusters are also influenced by impurities such as copper. A theoretical model has been developed to investigate potential role of PDCs in RPV embrittlement. The model includes a detailed description of interstitial cluster population; vacancy clusters are treated in a more approximate fashion. The model has been used to examine a broad range of irradiation and material parameters. Results indicate that magnitude of hardening increment due to these clusters can be comparable to that attributed to copper precipitates. Both interstitial and vacancy type defects contribute to this hardening, with their relative importance determined by the specific irradiation conditions

Irradiation Effects Test Series: Test IE-3. Test results report. [PWR

Energy Technology Data Exchange (ETDEWEB)

Farrar, L. C.; Allison, C. M.; Croucher, D. W.; Ploger, S. A.

1977-10-01

The objectives of the test reported were to: (a) determine the behavior of irradiated fuel rods subjected to a rapid power increase during which the possibility of a pellet-cladding mechanical interaction failure is enhanced and (b) determine the behavior of these fuel rods during film boiling following this rapid power increase. Test IE-3 used four 0.97-m long pressurized water reactor type fuel rods fabricated from previously irradiated fuel. The fuel rods were subjected to a preconditioning period, followed by a power ramp to 69 kW/m at a coolant mass flux of 4920 kg/s-m/sup 2/. After a flow reduction to 2120 kg/s-m/sup 2/, film boiling occurred on the fuel rods. One rod failed approximately 45 seconds after the reactor was shut down as a result of cladding embrittlement due to extensive cladding oxidation. Data are presented on the behavior of these irradiated fuel rods during steady-state operation, the power ramp, and film boiling operation. The effects of a power ramp and power ramp rates on pellet-cladding interaction are discussed. Test data are compared with FRAP-T3 computer model calculations and data from a previous Irradiation Effects test in which four irradiated fuel rods of a similar design were tested. Test IE-3 results indicate that the irradiated state of the fuel rods did not significantly affect fuel rod behavior during normal, abnormal (power ramp of 20 kW/m per minute), and accident (film boiling) conditions.

Atom probe study of the microstructural evolution induced by irradiation in Fe-Cu ferritic alloys and pressure vessel steels

International Nuclear Information System (INIS)

Pareige, P.

1996-04-01

Pressure vessel steels used in pressurized water reactors are low alloyed ferritic steels. They may be prone to hardening and embrittlement under neutron irradiation. The changes in mechanical properties are generally supposed to result from the formation of point defects, dislocation loops, voids and/or copper rich clusters. However, the real nature of the irradiation induced-damage in these steels has not been clearly identified yet. In order to improve our vision of this damage, we have characterized the microstructure of several steels and model alloys irradiated with electrons and neutrons. The study was performed with conventional and tomographic atom probes. The well known importance of the effects of copper upon pressure vessel steel embrittlement has led us to study Fe-Cu binary alloys. We have considered chemical aging as well as aging under electron and neutron irradiations. The resulting effects depend on whether electron or neutron irradiations ar used for thus. We carried out both kinds of irradiation concurrently so as to compare their effects. We have more particularly considered alloys with a low copper supersaturation representative of that met with the French vessel alloys (0.1% Cu). Then, we have examined steels used on French nuclear reactor pressure vessels. To characterize the microstructure of CHOOZ A steel and its evolution when exposed to neutrons, we have studied samples from the reactor surveillance program. The results achieved, especially the characterization of neutron-induced defects have been compared with those for another steel from the surveillance program of Dampierre 2. All the experiment results obtained on model and industrial steels have allowed us to consider an explanation of the way how the defects appear and grow, and to propose reasons for their influence upon steel embrittlement. (author). 3 appends

Neutron resistant irradiation alloy and usage thereof

International Nuclear Information System (INIS)

Okada, Osamu; Nakata, Kiyotomo; Kato, Takahiko.

1997-01-01

A neutron irradiation embrittlement-resistant alloy comprising a Ti alloy having an average grain size of 2μm or smaller and containing from 30 to 40wt% of Al is subjected to powder solidification and then to isothermal forging at a forging rate of from 50 to 80% at a temperature range of from 1150 to 1500K. Namely, since the Ti-Al type alloy comprises from 30 to 30wt% of Al, optionally, from 1 to 6% of Mn, from 0.1 to 0.5% of Si, from 4 to 16% of V and the balance of Ti, it has excellent specific strength, high durable temperature and excellent neutron irradiation resistance, and has ductility required as structural materials. Accordingly, if the Ti-Al type alloy excellent in embrittlement resistance to neutron irradiation dimensional stability of materials is applied to constitutional parts of a reactor core of a nuclear reactor and a thermonuclear reactor to be exposed under neutron irradiation, high reliability is provided and the amount of activated materials is reduced by improving the working life of the materials. (N.H.)

Precipitation hardening and hydrogen embrittlement of aluminum ...

Indian Academy of Sciences (India)

Hydrogen susceptibility of alloy AA7020 was evaluated by slow strain-rate tensile ... high pressures because of the embrittling effect of hydrogen. ... The higher the total Zn + Mg content,. ∗ .... dislocations, leading to a local softening of the slip plane, and thus to ... A Vickers hardness testing machine was used to measure the.

The liquid metal embrittlement of iron and ferritic steels in sodium

International Nuclear Information System (INIS)

Hilditch, J.P.; Hurley, J.R.; Tice, D.R.; Skeldon, P.

1995-01-01

The liquid metal embrittlement of iron and A508 III, 21/4Cr-1Mo and 15Mo3 steels in sodium at 200-400 o C has been studied, using dynamic straining at 10 -6 s -1 , in order to investigate the roles of microstructure and composition. The steels comprised bainitic, martensitic, tempered martensitic and ferritic/pearlitic microstructures. All materials were embrittled by sodium, the embrittlement being associated generally with quasicleavage on fracture surfaces. Intergranular cracking was also found with martensitic and ferritic/pearlitic microstructures. The susceptibility to embrittlement was greater in higher strength materials and at higher temperatures. The embrittlement was similar to that encountered previously in 9Cr steel, which depends upon the presence of non-metallic impurities in the sodium. (author)

Further application of the cleavage fracture stress model for estimating the T{sub 0} of highly embrittled ferritic steels

Energy Technology Data Exchange (ETDEWEB)

Sreenivasan, P.R.

2016-02-15

The semi-empirical cleavage fracture stress model (CFS), based on the microscopic cleavage fracture stress, s{sub f}, for estimating the ASTM E1921 reference temperature (T{sub 0}) of ferritic steels from instrumented impact testing of unprecracked Charpy V-notch specimens is further confirmed by test results for additional steels, including steels highly embrittled by thermal aging or irradiation. In addition to the ferrite-pearlite, bainitic or tempered martensitic steels (which was examined earlier), acicular or polygonal ferrite, precipitation-strengthened or additional simulated heat affected zone steels are also evaluated. The upper limit for the applicability of the present CFS model seems to be T{sub 41J} ∝160 to 170 C or T{sub 0} or T{sub Qcfs} (T{sub 0} estimate from the present CFS model) ∝100 to 120 C. This is not a clear-cut boundary, but indicative of an area of caution where generation and evaluation of further data are required. However, the present work demonstrates the applicability of the present CFS model even to substantially embrittled steels. The earlier doubts expressed about T{sub Qcfs} becoming unduly non-conservative for highly embrittled steels has not been fully substantiated and partly arises from the necessity of modifications in the T{sub 0} evaluation itself at high degrees of embrittlement suggested in the literature.

Accelerated irradiation test of Gundremmingen reactor vessel trepan material

Energy Technology Data Exchange (ETDEWEB)

Hawthorne, J.R. [Materials Engineering Associates, Inc., Lanham, MD (United States)

1992-08-01

Initial mechanical properties tests of beltline trepanned from the decommissioned KRB-A pressure vessel and archive material irradiated in the UBR test reactor revealed a major anomaly in relative radiation embrittlement sensitivity. Poor correspondence of material behavior in test vs. power reactor environments was observed for the weak test orientation (ASTL C-L) whereas correspondence was good for the strong orientation (ASTM C-L). To resolve the anomaly directly, Charpy-V specimens from a low (essentially-nil) fluence region of the vessel were irradiated together with archive material at 279{degrees}C in the UBR test reactor. Properties tests before UBR irradiation revealed a significant difference in 41-J transition temperature and upper shelf energy level between the materials. However, the materials exhibited essentially the same radiation embrittlement sensitivity (both orientations), proving that the anomaly is not due to a basic difference in material irradiation resistances. Possible causes of the original anomaly and the significance to NRC Regulatory Guide 1.99 are discussed.

Accelerated irradiation test of gundremmingen reactor vessel trepan material

International Nuclear Information System (INIS)

Hawthorne, J.R.

1992-08-01

Initial mechanical properties tests of beltline trepanned from the decommissioned KRB-A pressure vessel and archive material irradiated in the UBR test reactor revealed a major anomaly in relative radiation embrittlement sensitivity. Poor correspondence of material behavior in test vs. power reactor environments was observed for the weak test orientation (ASTL C-L) whereas correspondence was good for the strong orientation (ASTM C-L). To resolve the anomaly directly, Charpy-V specimens from a low (essentially-nil) fluence region of the vessel were irradiated together with archive material at 279 degrees C in the UBR test reactor. Properties tests before UBR irradiation revealed a significant difference in 41-J transition temperature and upper shelf energy level between the materials. However, the materials exhibited essentially the same radiation embrittlement sensitivity (both orientations), proving that the anomaly is not due to a basic difference in material irradiation resistances. Possible causes of the original anomaly and the significance to NRC Regulatory Guide 1.99 are discussed

Embrittlement and life prediction of aged duplex stainless steel

International Nuclear Information System (INIS)

Kuwano, Hisashi

1996-01-01

The stainless steel, for which the durability for long term in high temperature corrosive environment is demanded, is a complex plural alloy. Cr heightens the oxidation resistance, Ni improves the ductility and impact characteristics, Si improves the fluidity of the melted alloy and heightens the resistance to stress corrosion cracking, and Mo suppresses the pitting due to chlorine ions. These alloy elements are in the state of nonequilibrium solid solution in Fe base at practical temperature, and cause aging phenomena such as segregation, concentration abnormality and precipitation during the use for long term. The characteristics of stainless steel deteriorate due to this. Two-phase stainless cast steel, the example of the embrittlement of the material for an actual machine, the accelerated test of embrittlement, the activation energy for embrittlement, and as the mechanism of aging embrittlement, the spinodal decomposition of ferrite, the precipitation of G phase and the precipitation of carbides and nitrides are described. Also in the welded parts of austenitic stainless steel, delta-ferrite is formed during cooling, therefore, the condition is nearly same as two-phase stainless steel, and the embrittlement due to long term aging occurs. (K.I.)

The effects of composition on the environmental embrittlement of Fe{sub 3}Al alloys

Energy Technology Data Exchange (ETDEWEB)

Alven, D.A.; Stoloff, N.S. [Rensselaer Polytechnic Inst., Troy, NY (United States)

1997-12-01

This paper reviews recent research on embrittlement of iron aluminides at room temperature brought about by exposure to moisture or hydrogen. The tensile and fatigue crack growth behavior of several Fe-28Al-5Cr alloys with small additions of Zr and C are described. It will be shown that fatigue crack growth behavior is dependent on composition, environment, humidity level, and frequency. Environments studied include vacuum, oxygen, hydrogen gas, and moist air. All cases of embrittlement are ultimately traceable to the interaction of hydrogen with the crack tip.

The influences of impurity content, tensile strength, and grain size on in-service temper embrittlement of CrMoV steels

International Nuclear Information System (INIS)

Cheruvu, N.S.; Seth, B.B.

1989-01-01

The influences of impurity levels, grain size, and tensile strength on in-service temper embrittlement of CrMoV steels have been investigated. The samples for this study were taken from steam turbine CrMoV rotors which had operated for 15 to 26 years. The effects of grain size and tensile strength on embrittlement susceptibility were separated by evaluating the embrittlement behavior of two rotor forgings made from the same ingot after an extended step-cooling treatment. Among the residual elements in the steels, only P produces a significant embrittlement. The variation of P and tensile strength has no effect on in-service temper embrittlement susceptibility, as measured by the shift in fracture appearance transition temperature (FATT). However, the prior austenite grain size plays a major role in service embrittlement. The fine grain steels with a grain size of ASTM No. 9 or higher are virtually immune to in-service embrittlement. In steels having duplex grain sizes, embrittlement susceptibility is controlled by the size of coarser grains. For a given steel chemistry, the coarse grain steel is more susceptible to in-service embrittlement, and a decrease in ASTM grain size number from 4 to 0/1 increases the shift in FATT by 61 degrees C (10/10 degrees F). It is demonstrated that long-term service embrittlement can be simulated, except in very coarse grain steels, by using the extended step-cooling treatment. The results of step-cooling studies show that the coarse grain rotor steels take longer time during service to reach a fully embrittled state than the fine grain rotor steels

Effect of irradiation damage and helium on the swelling and structure of vanadium-base alloys

International Nuclear Information System (INIS)

Chung, H.M.; Loomis, B.A.; Smith, D.L.

1993-12-01

Swelling behavior and microstructural evolution of V-Ti, V-Cr-Ti, and V-Ti-Si alloys were investigated after irradiation at 420--600C up to 114 dpa. The alloys exhibited swelling maxima between 30 and 80 dpa and swelling decreased on irradiation to higher dpa. This is in contrast to the monotonically increasing swelling of binary alloys that contain Fe, Ni, Cr, Mo, W, and Si. Precipitation of dense Ti 5 Si 3 promotes good resistance to swelling of the Ti-containing alloys and it was concluded that Ti of >3 wt.% and 400--1000 wppm Si are necessary to effectively suppress swelling. Swelling was minimal in V-4Cr-4Ti, identified as the most promising alloy based on good mechanical properties and superior resistance to irradiation embrittlement. V-20Ti doped with B exhibited somewhat higher swelling because of He generation. Lithium atoms, generated from transmutation of 10 B, formed γ-LiV 2 O 5 precipitates and did not seem to produce undesirable effects on mechanical properties

Strategic Assessment of Causes, Impacts and Mitigation of Radiation Embrittlement of RPV steel in LWRs

International Nuclear Information System (INIS)

Shamim, Jubair Ahmed; Bhowmik, Palash Kumar; Gairola, Abhinav; Suh, Kune Y.

2014-01-01

Nuclear power has been emerged as a proven technology in the present day world to beget electricity after its first successful demonstration in 1942. Due to world's increasing concern over the augmented concentration of 'Greenhouse Gas' emissions primarily caused by burning of fossil fuel, it is not surprising that there will be a galloping demand for nuclear power in near future. As per data of World Nuclear Association, there are currently 435 operable civil nuclear power reactors around the world, with a further 71 under construction, among which the most common type is LWR. Pressure vessel of LWR is the most vital pressure boundary component of Nuclear Steam Supply System (NSSS) as it houses the core under elevated pressure and temperature. It also provides structural support to RPV internals and attempts to protect against possible rupture under all postulated transients that the NSSS may undergo. LWR pressure vessel experiences service at a temperature of 250-320 .deg. C and receives significant level of fast neutron fluence, ranging from about 5*10 22 to 3*10 24 n/m 2 depending on plant design. There are also differences in materials used for various designed reactors. Weldments also vary in type and impurity level. Accordingly, the assessment of degradation of major components such as RPV steel caused by aging and corrosion is a common objective for safe operation of all LWRs. The purpose of this paper is to assess how neutron irradiation contributes to the degradation of mechanical properties of RPV steel and how these effects can be minimized. Since RPV is the only irreplaceable component in NPPs, the degradation of mechanical properties of RPV is the life-limiting feature of LWR nuclear power plant operation. Although there are a number of ways (e.g. thermal neutrons, fast neutrons and gamma-ray irradiation) that may contribute to the displacement of atoms (hence RPV embrittlement and degradation of mechanical properties), most of the

Strategic Assessment of Causes, Impacts and Mitigation of Radiation Embrittlement of RPV steel in LWRs

Energy Technology Data Exchange (ETDEWEB)

Shamim, Jubair Ahmed; Bhowmik, Palash Kumar; Gairola, Abhinav; Suh, Kune Y. [Seoul National Univ., Seoul (Korea, Republic of)

2014-10-15

Nuclear power has been emerged as a proven technology in the present day world to beget electricity after its first successful demonstration in 1942. Due to world's increasing concern over the augmented concentration of 'Greenhouse Gas' emissions primarily caused by burning of fossil fuel, it is not surprising that there will be a galloping demand for nuclear power in near future. As per data of World Nuclear Association, there are currently 435 operable civil nuclear power reactors around the world, with a further 71 under construction, among which the most common type is LWR. Pressure vessel of LWR is the most vital pressure boundary component of Nuclear Steam Supply System (NSSS) as it houses the core under elevated pressure and temperature. It also provides structural support to RPV internals and attempts to protect against possible rupture under all postulated transients that the NSSS may undergo. LWR pressure vessel experiences service at a temperature of 250-320 .deg. C and receives significant level of fast neutron fluence, ranging from about 5*10{sup 22} to 3*10{sup 24} n/m{sup 2} depending on plant design. There are also differences in materials used for various designed reactors. Weldments also vary in type and impurity level. Accordingly, the assessment of degradation of major components such as RPV steel caused by aging and corrosion is a common objective for safe operation of all LWRs. The purpose of this paper is to assess how neutron irradiation contributes to the degradation of mechanical properties of RPV steel and how these effects can be minimized. Since RPV is the only irreplaceable component in NPPs, the degradation of mechanical properties of RPV is the life-limiting feature of LWR nuclear power plant operation. Although there are a number of ways (e.g. thermal neutrons, fast neutrons and gamma-ray irradiation) that may contribute to the displacement of atoms (hence RPV embrittlement and degradation of mechanical properties

Hydrogen environment embrittlement

International Nuclear Information System (INIS)

Donovan, J.A.

1975-01-01

Exposure of many metals to gaseous hydrogen causes losses in elongation, reduction of area, and fracture toughness, and causes increases in slow crack growth rate or fatigue life compared with values obtained in air or vacuum. Hydrogen pressure, temperature, and purity significantly influence deleterious effects. The strength and structural characteristics of the metal influence the degradation of its properties by hydrogen. Several theories have been proposed to explain the loss of properties in hydrogen, but none has gained wide acceptance. The embrittlement mechanism and the role of diffusion are, therefore, open questions and need more quantitative experimental data both to test the proposed theories and to allow the development of realistic preventive measures. (U.S.)

Ni/boride interfaces and environmental embrittlement in Ni-based superalloys: A first-principles study

International Nuclear Information System (INIS)

Sanyal, Suchismita; Waghmare, Umesh V.; Hanlon, Timothy; Hall, Ernest L.

2011-01-01

Highlights: ► Fracture strengths of Ni/boride interfaces through first-principles calculations. ► Fracture strengths of Ni/boride interfaces are higher than Ni/Ni 3 Al and NiΣ5 grain boundaries. ► Ni/boride interfaces have higher resistance to O-embrittlement than Ni/Ni 3 Al and NiΣ5 grain boundaries. ► CrMo-borides are more effective than Cr-borides in resisting O-embrittlement. ► Electronegativity differences between alloying elements correlate with fracture strengths. - Abstract: Motivated by the vital role played by boride precipitates in Ni-based superalloys in improving mechanical properties such as creep rupture strength, fatigue crack growth rates and improved resistance towards environmental embrittlement , we estimate fracture strength of Ni/boride interfaces through determination of their work of separation using first-principles simulations. We find that the fracture strength of Ni/boride interfaces is higher than that of other commonly occurring interfaces in Ni-alloys, such as Ni Σ-5 grain boundaries and coherent Ni/Ni 3 Al interfaces, and is less susceptible to oxygen-induced embrittlement. Our calculations show how the presence of Mo in Ni/M 5 B 3 (M = Cr, Mo) interfaces leads to additional reduction in oxygen-induced embrittlement. Through Electron-Localization-Function based analyses, we identify the electronic origins of effects of alloying elements on fracture strengths of these interfaces and observe that chemical interactions stemming from electronegativity differences between different atomic species are responsible for the trends in calculated strengths. Our findings should be useful towards designing Ni-based alloys with higher interfacial strengths and reduced oxygen-induced embrittlement.

Proposal of guideline for bonding to prevention of hydrogen embrittlement at Ta/Zr bond interface. Hydrogen embrittlement in SUS304ULC/Ta/Zr explosive bonded joint

International Nuclear Information System (INIS)

Saida, Kazuyoshi; Fujimoto, Tetsuya; Nishimoto, Kazutoshi

2010-01-01

The occurrence condition of hydrogen embrittlement cracking at Ta/Zr bond interface was investigated with respect to the hydrogen content and applied stress in order to propose a guideline for the explosive bonding procedure to prevention of hydrogen embrittlement. Hydrogen charging test was conducted for SUS304ULC/Ta/Zr explosive bonded joints applied the different flexural strains. A hydrogen embrittlement crack occurred in the Zr substrate at Ta/Zr bond interface after hydrogen charging, and it was initiated at shorter charging times when the augmented strain was increased. The occurrence condition of hydrogen embrittlement cracking at Ta/Zr bond interface was shifted to lower stress and hydrogen content with an increase in the amount of explosive during bonding. It was suggested that hydrogen embrittlement in Ta/Zr explosive bonded joint could be inhibited by reducing the initial hydrogen content in Ta substrate less than approx. 5 ppm. (author)

Current limitations of trend curve analysis for the prediction of reactor PV embrittlement

International Nuclear Information System (INIS)

Gold, R.; McElroy, W.N.

1986-02-01

In operating light water reactor (LWR) commercial power plants, neutron radiation induces embrittlement of the pressure vessel (PV) and its support structures. As a consequence, LWR-PV integrity is a primary safety consideration. LWR-PV integrity is a significant economic consideration since the PV and its support structures are nonreplaceable power plant components and embrittlement of these components can therefore limit the effective operating lifetime of the plant

Dosimetry, metallurgical and code needs of the U.S. utilities related to radiation embrittlement of nuclear pressure vessels

International Nuclear Information System (INIS)

Rahn, F.J.; Marston, T.U.; Ozer, O.; Stahlkopf, K.

1980-01-01

Codes and regulation guides in the U.S.A., on performance of pressure vessel are examined. Limiting factors in the analysis and prediction of radiation embrittlement in reactor pressure vessels are: accurate measurement of neutron flux and spectrum in-situ, irradiation rate dependence, environmental conditions influence of flaws annealing, analysis of mechanical tests. The establishment of a self-consistent set of irradiated materials properties data taken at realistic flux rates is required, in conjunction with a careful technique in measuring with a careful technique in measuring the fluence and spectrum at the pressure vessel wall and material test specimen positions

Estimation of embrittlement damage risk at neutron embrittled vessel constructions

International Nuclear Information System (INIS)

Staevski, K.; Madzharov, D.; Detistov, P.; Petrova, T.

1998-01-01

In this work a methodology based on Damage mechanics criteria is proposed. This methodology serves for probability assessment of the brittle damage risk for the neutron embrittled vessel elements. The developed methodology is realised in RISK code and has been verified on the base of tough reliability of the pressure vessel, 'Kozloduy' NPP Unit 2. This investigation has been carried out at the given parameters of the possible defects on the vessel's weld 4 taking into account requirements of the western and Russian standards. The obtained values for ductile to brittle transition temperatures, defining the equipment life-time in the presence of maximal defect, are in good consistence with the experimentally determined ones. The analyses of results show that the pressure vessel of 'Kozloduy' NPP Unit 2 has got a high level of reliability from brittle damage risk point of view and that the western standards give more conservative evaluation. On the bases of the results a conclusion is made that the developed methodology enables analysing the influence of possible defects in the neutron embrittled elements on their to reliability and their remained life-time

Long-term embrittlement of cast duplex stainless steels in LWR systems

International Nuclear Information System (INIS)

Chopra, O.K.; Chung, H.M.

1990-08-01

This progress report summarizes work performed by Argonne National Laboratory on long-term embrittlement of cast duplex stainless steels in LWR systems during the six months from April to September 1988. Characteristics of the primary mechanism of aging embrittlement (i.e., spinodal decomposition of ferrite) and synergistic effects of alloying and impurity elements that influence the kinetics of the primary mechanism are discussed. Several secondary metallurgical processes of embrittlement, strongly dependent on the C, N, Ni, Mo, and Si content of various heats, are identified. Information on kinetics and data on impact properties are analyzed and correlated with microstructural characteristics to provide a unified method of extrapolating accelerated-aging data to reactor operating conditions. Fracture toughness data are presented for several heats of cast stainless steel aged at temperatures between 320 and 450 degrees C for times up to 10,000 h. Mechanical property data are analyzed to develop the procedure and correlations or predicting the kinetics and extent of embrittlement of reactor components from known material parameters. The method and examples of estimating the impact strength and fracture toughness of cast components during reactor service are described. The lower-bound values of impact strength and fracture toughness for cast stainless steels at LWR operating temperatures are defined. 42 refs., 14 figs., 6 tabs

NRC data base for power reactor surveillance programs and for irradiation experiments results

International Nuclear Information System (INIS)

Kam, F.B.K.; Stallmann, F.W.

1991-01-01

The radiation damage of pressure vessel materials in nuclear reactors depends on many different factors, primarily fluence, fluence spectrum, fluence rate, irradiation temperature, and chemistry. These factors and, possibly, others such as heat treatment and type of flux used in weldments must be considered to reliably predict the pressure vessel embrittlement and to assure the safe operation of the reactor. Based on embrittlement predictions, decisions must be made concerning operating parameters, low-leakage fuel management, possible life extension, and the need for annealing of the pressure vessel. Large numbers of data obtained from surveillance capsules and test reactor experiments are needed, comprising many different materials and different irradiation conditions, to develop generally applicable damage prediction models that can be used for industry standards and regulatory guides. The US Nuclear Regulatory Agency has, therefore, sponsored a project to construct an Embrittlement Data Base (EDB) for a comprehensive collection of data concerning changes in material properties of pressure vessel steels due to neutron irradiation. A first version containing data from surveillance capsules of commercial power reactors, the Power Reactor Embrittlement Data Base (PR-EDB) Version 1, has been completed and is available to authorized users from the Radiation Shielding Information Center at the Oak Ridge National Laboratory. This document provides a discussion of the features of the current database. 1 fig

Embrittlement of zircaloy cladding due to oxygen uptake (CBRTTL)

International Nuclear Information System (INIS)

Reymann, G.A.

1979-02-01

A model for embrittlement of zircaloy due to oxygen uptake at high temperatures is described. The model defines limits for oxygen content and temperature which, if exceeded, give rise to zircaloy cladding which is sufficiently embrittled to cause failure either on quenching or normal handling following a transient. A significant feature of this model is that the onset of embrittlement is dependent on the cooling rate. A distinction is made between slow and fast cooling, with the boundary at 100 K/s. The material property correlations and computer subcodes described in MATPRO are developed for use in Light Water Reactor (LWR) codes

Experimental data base for assessment of irradiation induced ageing effects in pre-irradiated RPV materials of German PWR

Energy Technology Data Exchange (ETDEWEB)

Hein, H.; Gundermann, A.; Keim, E.; Schnabel, H. [AREVA NP GmbH (Germany); Ganswind, J. [VGB PowerTech e.V (Germany)

2011-07-01

The 5 year research program CARISMA which ended in 2008 has produced a data base to characterize the fracture toughness of pre-irradiated original RPV (Reactor Pressure Vessel) materials being representative for all four German PWR construction lines of former Siemens/KWU company. For this purpose tensile, Charpy-V impact, crack initiation and crack arrest tests have been performed for three base materials and four weld metals irradiated to neutron fluences beyond the designed EoL range. RPV steels with optimized chemical composition and with high copper as well as high nickel content were examined in this study. The RTNDT concept and the Master Curve approach were applied for the assessment of the generated data in order to compare both approaches. A further objective was to clarify in which extent crack arrest curves can be generated for irradiated materials and how crack arrest can be integrated into the Master Curve approach. By the ongoing follow-up project CARINA the experimental data base will be extended by additional representative materials irradiated under different conditions and with respect to the accumulated neutron fluences and specific impact parameters such as neutron flux and manufacturing effects. The irradiation data cover also the long term irradiation behavior of the RPV steels concerned. Moreover, most of the irradiated materials were and will be used for microstructural examinations to get a deeper insight in the irradiation embrittlement mechanisms and their causal relationship to the material property changes. By evaluation of the data base the applicability of the Master Curve approach for both crack initiation and arrest was confirmed to a large extent. Moreover, within both research programs progress was made in the development of crack arrest test techniques and in specific issues of RPV integrity assessment. (authors)

Impact of radiation embrittlement on integrity of pressure vessel supports for two PWR plants

Energy Technology Data Exchange (ETDEWEB)

Cheverton, R.D.; Pennell, W.E.; Robinson, G.C.; Nanstad, R.K.

1989-01-01

Recent data from the HFIR vessel surveillance program indicate a substantial radiation embrittlement rate effect at low irradiation temperatures (/approximately/120/degree/F) for A212-B, A350-LF3, A105-II, and corresponding welds. PWR vessel supports are fabricated of similar materials and are subjected to the same low temperatures and fast neutron fluxes (10/sup 8/ to 10/sup 9/ neutrons/cm/sup 2//center dot/s, E > 1.0 MeV) as those in the HFIR vessel. Thus, the embrittlement rate of these structures may be greater than previously anticipated. A study sponsored by the NRC is under way at ORNL to determine the impact of the rate effect on PWR vessel-support life expectancy. The scope includes the interpretation and application of the HFIR data, a survey of all light-water-reactor vessel support designs, and a structural and fracture-mechanics analysis of the supports for two specific PWR plants of particular interest with regard to a potential for support failure as a result of propagation of flaws. Calculations performed thus far indicate best-estimate critical flaw sizes, corresponding to 32 EFPY, of /approximately/0.2 in. for one plant and /approximately/0.4 in. for the other. These flaw sizes are small enough to be of concern. However, it appears that low-cycle fatigue is not a viable mechanism for creation of flaws of this size, and thus, presumably, such flaws would have to exist at the time of fabrication. 59 refs., 128 figs., 49 tabs.

Irradiation sterilization of semi-crystalline polymers

International Nuclear Information System (INIS)

Williams, J.; Dunn, T.; Stannett, V.

1978-01-01

A semi-crystalline polymer such as polypropylene, is sterilized by high energy irradiation, with the polymer containing a non-crystalline mobilizing additive which increases the free volume of the polymer, to prevent embrittlement of the polymer during and subsequent to the irradiation. The additive has a density of from 0.6 to 1.9 g/cm 3 and a molecular weight from 100 to 10,000 g/mole

The impact of mobile point defect clusters in a kinetic model of pressure vessel embrittlement

International Nuclear Information System (INIS)

Stoller, R.E.

1998-05-01

The results of recent molecular dynamics simulations of displacement cascades in iron indicate that small interstitial clusters may have a very low activation energy for migration, and that their migration is 1-dimensional, rather than 3-dimensional. The mobility of these clusters can have a significant impact on the predictions of radiation damage models, particularly at the relatively low temperatures typical of commercial, light water reactor pressure vessels (RPV) and other out-of-core components. A previously-developed kinetic model used to investigate RPV embrittlement has been modified to permit an evaluation of the mobile interstitial clusters. Sink strengths appropriate to both 1- and 3-dimensional motion of the clusters were evaluated. High cluster mobility leads to a reduction in the amount of predicted embrittlement due to interstitial clusters since they are lost to sinks rather than building up in the microstructure. The sensitivity of the predictions to displacement rate also increases. The magnitude of this effect is somewhat reduced if the migration is 1-dimensional since the corresponding sink strengths are lower than those for 3-dimensional diffusion. The cluster mobility can also affect the evolution of copper-rich precipitates in the model since the radiation-enhanced diffusion coefficient increases due to the lower interstitial cluster sink strength. The overall impact of the modifications to the model is discussed in terms of the major irradiation variables and material parameter uncertainties

Irradiation effects on aluminium and beryllium

International Nuclear Information System (INIS)

Bieth, M.

1992-01-01

ductility of 1.6%. Besides, due to the effects of embrittlement and swelling induced by irradiation, the HFR beryllium reflector elements had to be replaced after more than 25 years of operation. Operational and practical experiences with these reflector elements are commented, as well as main engineering features of the new reflector elements: upper-end fittings of both filler element and insert in stainless steel, no radially drilled holes and no roll pins

Newly developed non-destructive testing method for evaluation of irradiation brittleness of structural materials using ultrasonic

International Nuclear Information System (INIS)

Ishii, Toshimitsu; Ooka, Norikazu; Kato, Yoshiaki; Saito, Junichi; Hoshiya, Taiji; Shibata, Saburo; Kobayashi, Hideo

1999-01-01

Surveillance testing is important to evaluate neutron irradiation embrittlement of reactor pressure vessel material for long life operation. An alternative test method for evaluating the irradiation embrittlement of the pressure vessel material will have to be proposed to support the limited number of surveillance test specimens in order to manage the plant life to be extended. In this study, ultrasonic testing for irradiated A533B-1 steel and weld metal was applied to examine material degradation nondestructively. With increasing the shift of Charpy 41 J transition temperature, ultrasonic velocity decreased and attenuation coefficient of ultrasonic wave increased. Especially, the difference of ultrasonic velocity for 5 MHz shear wave between as-received and irradiated material is corresponding to the shift of transition temperature showing material degradation. (author)

Irreversible traps, their influence on the embrittlement of high strength steel

International Nuclear Information System (INIS)

Mariano, I; Mansilla, G

2012-01-01

Hydrogen (H) can be trapped in lattice defects such as vacancies, dislocations, grain boundaries and interfaces between the matrix and precipitates. The effect on the mechanical properties depends on factors inherent in materials such as the activation energy of irreversible traps (H trapped in Network Places) and its sensitivity to embrittlement. Differential scanning calorimetry (DSC) allows the study of those processes in which enthalpy variation occurs. The purpose is to record the difference in enthalpy change that occurs in the sample as a function of temperature or time. This work represents a study of H embrittlement of high strength steel resulfurized

Irradiation effects on mechanical properties of fuel element cladding from thermal reactors

International Nuclear Information System (INIS)

Chatterjee, S.

2005-01-01

During reactor operation, UO 2 expands more than the cladding tube (Zirconium alloys for thermal reactors), is hotter, cracks and swells. The fuel therefore will interact with the cladding, resulting in straining of the later. To minimize the possibility of rupture of the cladding, ideally it should have good ductility as well as high strength. However, the ductility reduces with increase in fuel element burn-up. Increased burn-up also increases swelling of the fuel, leading to increased contact pressure between the fuel and the cladding tube. This would cause strains to be concentrated over localized regions of the cladding. For fuel elements burnup exceeding 40 GWd/T, the contribution of embrittlement due to hydriding, and the increased possibility of embrittlement due to stress corrosion cracking, also need to be considered. In addition to the tensile properties, the other mechanical properties of interest to the performance of cladding tube in an operating fuel element are creep rate and fatigue endurance. Irradiation is reported to have insignificant effect on high cycle endurance limit, and fatigue from fuel element vibration is most unlikely, to be life limiting. Even though creep rates due to irradiation are reported to increase by an order of magnitude, the cladding creep ductility would be so high that creep type failures in fuel element would be most improbable. Thus, the most important limiting aspect of mechanical performance of fuel element cladding has been recognized as the tensile ductility resulting from the stress conditions experienced by the cladding. Some specific fission products of threshold amount (if) deposited on the cladding, and hydride morphology (e.g. hydride lenses). The presentation will brief about irradiation damage in cladding materials and its significance, background of search for better Zirconium alloys as cladding materials, and elaborate on the types of mechanical tests need to be conducted for the evaluation of claddings

Effect of Microstructure and Alloy Chemistry on Hydrogen Embrittlement of Precipitation-Hardened Ni-Based Alloys

Science.gov (United States)

Obasi, G. C.; Zhang, Z.; Sampath, D.; Morana, Roberto; Akid, R.; Preuss, M.

2018-04-01

The sensitivity to hydrogen embrittlement (HE) has been studied in respect of precipitation size distributions in two nickel-based superalloys: Alloy 718 (UNS N07718) and Alloy 945X (UNS N09946). Quantitative microstructure analysis was carried out by the combination of scanning and transmission electron microscopy and energy dispersive x-ray spectroscopy (EDS). While Alloy 718 is mainly strengthened by γ″, and therefore readily forms intergranular δ phase, Alloy 945X has been designed to avoid δ formation by reducing Nb levels providing high strength through a combination of γ' and γ″. Slow strain rate tensile tests were carried out for different microstructural conditions in air and after cathodic hydrogen (H) charging. HE sensitivity was determined based on loss of elongation due to the H uptake in comparison to elongation to failure in air. Results showed that both alloys exhibited an elevated sensitivity to HE. Fracture surfaces of the H precharged material showed quasi-cleavage and transgranular cracks in the H-affected region, while ductile failure was observed toward the center of the sample. The crack origins observed on the H precharged samples exhibited quasi-cleavage with slip traces at high magnification. The sensitivity is slightly reduced for Alloy 718, by coarsening γ″ and reducing the overall strength of the alloy. However, on further coarsening of γ″, which promotes continuous decoration of grain boundaries with δ phase, the embrittlement index rose again indicating a change of hydrogen embrittlement mechanism from hydrogen-enhanced local plasticity (HELP) to hydrogen-enhanced decohesion embrittlement (HEDE). In contrast, Alloy 945X displayed a strong correlation between strength, based on precipitation size and embrittlement index, due to the absence of any significant formation of δ phase for the investigated microstructures. For the given test parameters, Alloy 945X did not display any reduced sensitivity to HE compared with

Investigation of neutron irradiated reactor vessel steels using post-irradiation annealing techniques

Energy Technology Data Exchange (ETDEWEB)

Nakata, Hayato; Fukuya, Koji [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

2001-09-01

The matrix damage is known to be a major factor that contributes to embrittlement and hardening of irradiated reactor vessel steels, and is assumed to be composed of the point defect clusters. However field emission gun scanning transmission electron microscopy (FEGSTEM) and atom probe (AP) could not detect any evidence of the matrix damage. In this study, post irradiation annealing experiments combining positron annihilation lineshape analysis (PALA) and hardness experiments were applied to an actual surveillance test specimen and a sample of reactor vessel steel irradiated in a material test reactor (MTR), in order to investigate the matrix damage recovery behavior and its contribution to hardening. It was confirmed that higher fluence increased the hardness and the volume fraction of open volume defects and that higher flux decreased the thermal stability of matrix damage and the effect on hardening. The contribution of matrix damage to hardening could be estimated to be below 30%. (author)

Study of intergranular embrittlement in Fe-12Mn alloys

International Nuclear Information System (INIS)

Lee, H.J.

1982-06-01

A high resolution scanning Auger microscopic study has been performed on the intergranular fracture surfaces of Fe-12Mn steels in the as-austenitized condition. Fracture mode below the ductile-brittle transition temperature was intergranular whenever the alloy was quenched from the austenite field. The intergranular fracture surface failed to reveal any consistent segregation of P, S, As, O, or N. The occasional appearance of S or O on the fracture surface was found to be due to a low density precipitation of MnS and MnO 2 along the prior austenite boundaries. An AES study with Ar + ion-sputtering showed no evidence of manganese enrichment along the prior austenite boundaries, but a slight segregation of carbon which does not appear to be implicated in the tendency toward intergranular fracture. Addition of 0.002% B with a 1000 0 C/1h/WQ treatment yielded a high Charpy impact energy at liquid nitrogen temperature, preventing the intergranular fracture. High resolution AES studies showed that 3 at. % B on the prior austenite grain boundaries is most effective in increasing the grain boundary cohesive strength in an Fe-12Mn alloy. Trace additions of Mg, Zr, or V had negligible effects on the intergranular embrittlement. A 450 0 C temper of the boron-modified alloys was found to cause tempered martensite embrittlement, leading to intergranular fracture. The embrittling treatment of the Fe-12Mn alloys with and without boron additions raised the ductile-brittle transition by 150 0 C. This tempered martensite embrittlement was found to be due to the Mn enrichment of the fracture surface to 32 at. % Mn in the boron-modified alloy and 38 at. % Mn in the unmodified alloy. The Mn-enriched region along the prior austenite grain boundaries upon further tempering is believed to cause nucleation of austenite and to change the chemistry of the intergranular fracture surfaces. 61 figures

The influence of second-phase dispersion on environmental embrittlement of Ni3(Si,Ti) alloys

International Nuclear Information System (INIS)

Takasugi, T.; Hanada, S.

1999-01-01

Some quaternary Ni 3 (Si,Ti) alloyed with transition elements V, Nb, Zr and Hf was prepared beyond their maximum solubility limits to investigate the effect of second-phase dispersion on moisture-induced embrittlement. V-added Ni 3 (Si,Ti) alloy contained ductile fcc-type Ni solid solution as the second-phase, while Nb-, Zr- and Hf-added Ni 3 (Si,Ti) alloys contained hard dispersion compounds as the second-phase. V- and Nb-added Ni 3 (Si,Ti) alloys did not display reduced tensile elongation in air, indicating that their second phases have the effect of suppressing the moisture-induced embrittlement. Possible mechanisms for the beneficial effect by the second phase on the moisture-induced embrittlement of V- and Nb-added Ni 3 (Si,Ti) alloys are discussed in association with hydrogen behavior and deformation property in the constituent phases or at matrix/second-phase interface

Role of twinning and transformation in hydrogen embrittlement of austenitic stainless steels

International Nuclear Information System (INIS)

Caskey, G.R. Jr.

1977-01-01

Internal hydrogen embrittlement may be viewed as an extreme form of environmental embrittlement that arises following prolonged exposure to a source of hydrogen. Smooth bar tensile specimens of three stainless steels saturated with deuterium (approximately 200 mol D 2 /m 3 ) were pulled to failure in air at 200 to 400 0 K or in liquid nitrogen at 78 0 K. In Type 304L stainless steel and Tenelon ductility losses are a maximum around 200 to 273 0 K; Type 310 stainless steel is not embrittled at this hydrogen concentration. A distinct change in fracture mode accompanies hydrogen embrittlement, with fracture proceeding along coherent boundaries of pre-existing annealing twins. This fracture path is observed in Tenelon at 78 0 K even when hydrogen is absent. There is also a change in fracture appearance in specimens with no prior exposure to hydrogen if they are pulled to failure in high-pressure hydrogen. The fracture path is not identifiable, however. Magnetic response measurements and changes in the stress-strain curves show that hydrogen suppresses formation of strain-induced α'-martensite at 198 0 K in both Type 304L stainless steel and Tenelon, but there is little effect in Type 304L stainless at 273 0 K

An effective surveillance strategy for reactor pressure vessel assessment in the long term operation perspective

International Nuclear Information System (INIS)

Chaouadi, R.; Gerard, R.

2015-01-01

The reactor pressure vessel (RPV) irradiation embrittlement is monitored by means of surveillance capsules containing the RPV belt-line materials, inserted inside the reactor pressure vessel (RPV) before the start of operation. These capsules are placed at location where they receive a higher neutron flux than the vessel wall, by a factor of the order of 2 to 3. They are regularly retrieved and tested to evaluate the RPV irradiation embrittlement according to specific regulatory procedures and standards, in order to guarantee the safe operation of the RPV throughout its lifetime. These procedures are often relying on empirical but conservative concepts. In parallel, material research reactor (MTR) irradiations are often used to support the surveillance data and to develop a better understanding of irradiation effects, not only qualitatively but also quantitatively. Taking advantage of the increased understanding of irradiation effects, analytical tools were developed to improve the evaluation embrittlement and quality assurance of the RPV embrittlement assessment. In this framework, an alternative but complementary surveillance program assessment was developed in Belgium, the so-called enhanced surveillance, in order to benefit from the latest developments in the area of materials science and irradiation effects. The neutron flux and fracture properties of the surveillance materials can be reliably characterized and correlated to each other using physically-based rather than empirical concepts. The enhanced surveillance approach is complementary to the mandatory regulatory procedure and allows quantifying the conservatism of the regulatory approach. The enhanced surveillance approach that uses the reconstitution technology to fabricate additional small size specimens, appropriate modeling tools and microstructural examination when required, makes it possible to rationalize all available information in a physically-based way

Re-examining reactor vessel embrittlement at Chooz A

International Nuclear Information System (INIS)

Guilleret, J.-C.

1988-01-01

The Chooz A PWR experienced an extended shutdown in 1987/88 following indications that the reactor vessel was embrittling more rapidly than expected. Discrepancies between the expected rate and estimates of the actual rate were not easily explained. The huge body of work done since then to establish safety margins and support restart of the plant should provide a model for the owners of other older PWRs grappling with the embrittlement issue. (author)

Study of Irradiation Effects on the Fracture Properties of A533-Series Ferritic Steels

International Nuclear Information System (INIS)

Lee, Yong Bok; Lee, Gyeong Geun; Kwon, Jun Hyun

2011-01-01

Since the Kori nuclear power plant unit 3 (Kori-3) was founded in 1986, the surveillance tests have been conducted five times. One of the primary objectives of the surveillance test is to determine the effects of irradiation on reactor pressure vessel (RPV) steel embrittlement. The RPV is made out of ferritic steels such as SA533 type B class 1, which were used for early nuclear power plants industry including Kori-2, 3, 4 and Yonggwang-1, 2 units in Korea. The Westinghouse supplied Kori-3 with the RPV steels ASTM A533 grade B class 1, which is equivalent to SA533 type B class 1. The irradiation effects on tensile properties in ASTM A533 grade B class 1 steel had been studied by Steichen and Williams. They experimentally determined the effect of strain rate and temperature on the tensile properties of unirradiated and irradiated A533 grade B steel 1. The effects of neutron irradiation on ferritic steels could be determined from tensile properties, as well as the fracture strength and toughness measurements. Hunter and Williams have reported that the strength and ductility for unirradiated material at a low strain rate increase with decreasing test temperature. Also, neutron irradiation increases strength and decreases ductility. Crosley and Ripling revealed that the yield strength of unirradiated material rapidly increases with the strain rate. Therefore, yield strength for unirradiated and irradiated materials should be determined by test parameters along with strain rate and temperature. In this study we compare ASTM A533 grad B class 1 steel obtained from several papers with SA533 type B class 1 steel taken from the surveillance data of Kori-3 unit, whose mechanical property of unirradiated and irradiated materials was correlated with the rate-temperature parameter

Opening of new field in material science and technology by materials irradiation research

Energy Technology Data Exchange (ETDEWEB)

Kurishita, Hiroaki [Tohoku Univ., Sendai (Japan). Inst. for Materials Research

1998-03-01

It is believed that high energy particle irradiation causes severe degradation of materials, and great efforts have been made to reveal the underlying mechanism of such degradation. However, recent progress of the developments of irradiation rigs performed in the Japan Materials Testing Reactor (JMTR) and materials fabrication techniques has enabled to change our understanding of radiation effects on materials from the above pessimistic one to the very challenging one, i.e., irradiation has the beneficial effect of producing new phenomena and/or innovative materials that will not be available without irradiation. An example to be noted is that irradiation with neutrons in JMTR greatly improved the ductility of less ductile metals. This ductility improvement due to irradiation is directly opposite to irradiation embrittlement and is called radiation induced ductilization (RIDU). In this presentation the significance of RIDU and its mechanism will be stated. (author)

Fracture toughness of irradiated wrought and cast austenitic stainless steels in BWR environment

International Nuclear Information System (INIS)

Chopra, O.K.; Gruber, E.E.; Shack, W.J.

2007-01-01

Experimental data are presented on the fracture toughness of wrought and cast austenitic stainless steels (SSs) that were irradiated to a fluence of ∼ 1.5 x 10 21 n/cm 2 (E > 1 MeV) * (∼ 2.3 dpa) at 296-305 o C. To evaluate the possible effects of test environment and crack morphology on the fracture toughness of these steels, all tests were conducted in normal-water-chemistry boiling water reactor (BWR) environments at ∼ 289 o C. Companion tests were also conducted in air on the same material for comparison. The fracture toughness J-R curves for SS weld heat-affected-zone materials in BWR water were found to be comparable to those in air. However, the results of tests on sensitized Type 304 SS and thermally aged cast CF-8M steel suggested a possible effect of water environment. The available fracture toughness data on irradiated austenitic SSs were reviewed to assess the potential for radiation embrittlement of reactor-core internal components. The synergistic effects of thermal and radiation embrittlement of cast austenitic SS internal components are also discussed. (author)

Power Reactor Embrittlement Data Base

International Nuclear Information System (INIS)

Kam, F.B.K.; Stallmann, F.W.; Wang, J.A.

1990-01-01

Regulatory and research evaluations of embrittlement predication models and of pressure vessel integrity can be greatly expedited by the use of a well-designed, computerized data base. The Power Reactor Embrittlement Data Base (PR-EDB) is such a comprehensive collection of data for US commercial nuclear reactors. The Nuclear Regulatory Commission (NRC) has provided financial support, and the Electric Power Research Institute (EPRI) has provided technical assistance in the quality assurance (QA) of the data to establish an industry-wide data base that will be maintained and updated on a long-term basis. Successful applications of the data base to several of NRC's evaluations have received favorable response and support for its continuation. The future direction of the data base has been designed to include the test reactor and other types of data of interest to the regulators and the researchers. 1 ref

Embrittlement of a 17Cr ferritic steel irradiated in Phenix

International Nuclear Information System (INIS)

Allegraud, G.; Boutard, J.L.; Boyer, J.M.

1987-01-01

Charpy V and tensile tests have been performed with samples made of 17Cr ferritic steel irradiated in Phenix at temperatures between 390 and 540C up to a maximum dose of 83.3 dpaF. All over the temperature and dose ranges, irradiation leads to an increase of the ductile brittle transition temperature (DBTT). The DBTT and hardening are decreasing functions of the irradiation temperature. Fast neutron flux at 390C hardens the material more than a sole thermal ageing does

Assessment of the effects of neutron fluence on Swedish nuclear pressure vessels

International Nuclear Information System (INIS)

Rao, S.

1980-11-01

Nuclear pressure vessels are subject to neutron irradiation during service causing embrittlement. This is one important factor in the overall problem of reactor vessel integrity. At present the irradiation effects are mainly assessed by the Charpy V-notch test. Two measures of embrittlement are defined: the increase of the ductile/brittle transition temperature and the decrease in the upper-shelf energy. The object of the present work is to assess these changes for the Swedish nuclear pressure vessels. On the basis of data from irradiations carried out in other countries and Swedish surveillance programmes, the expected end of life embrittlement is estimated for Swedish vessels. The results show that the embrittlement of most reactor vessels is expected to be quite small. Oskarshamn 1 and PWR-vessels, however, will probably show moderate changes, the former due to the higher copper content, and the latter due to the high end of life fluences. Some of the vessel materials which exhibit marginal properties in the upper-shelf energy, as measured by the Charpy V-notch impact test, are identified. It is recommended that fracture mechanics analyses be applied in these cases. (author)

Behaviour and microstructure of stainless steels irradiated in the french fast breeder reactors

International Nuclear Information System (INIS)

Dubuisson, P.; Gilbon, D.

1991-01-01

The burn-up of Fast Breeder Reactors is limited by the irradiation induced dimensional changes and mechanical properties of structural materials used for replaceable in-core components. This paper describes the behaviour improvements and also the radiation-induced microstructures of the different steels used for fuel pin cladding and wrapper tubes in French reactors. Materials of fuel pin cladding are austenitic steels whose main problem is swelling. Improvements in swelling resistance by cold-working, titanium additions and modifications of matrix (Fe-Cr-Ni) from SA 316 to CW 15-15 Ti are shown. These improvements are correlated with a best stability of microstructure under irradiation. Beneficial effects of phosphorus addition and multistabilisation (NbVTi) on radiation induced microstructure and swelling resistance are also shown. Austenitic steels used for wrapper tubes are limited both by swelling and by void embrittlement. The ferritic F17 (17Cr), ferritic-martensitic EM12 (9Cr-2MoNbV) and martensitic EM10 (9Cr-1Mo) steels present high swelling resistance. Nevertheless radiation-induced embrittlement is observed in EM12 and especially in F17. This embrittlement results from a fine and uniform radiation enhanced precipitation in ferrite grains. By contrast, the microstructure of fully martensitic EM10 steel is mush more stable and its ductile-brizzle transition temperature stays below 0 0 C. 12 figs

On the recovery of neutron irradiation defects of some metals and alloys

International Nuclear Information System (INIS)

Mohamed, H.G.; Matta, M.K.

2001-01-01

This work deals with the recovery of mechanical properties of neutron irradiated material to the pre-irradiating values. Rate of migration of defects responsible for radiation hardening and those inducing radiation embrittlement is analyzed. Role of crystalline structure is also studied. Materials of FCC crystal structure used in these investigations are pure Cu, Cu-5 at. % , Al, Cu-5 at. % Si, some Ni base binary alloys and some austenitic stainless steels mainly of AISI types 304 and 316. Among materials of BCC crystalline structure Fe-6 wt % Cr alloy is used. Alloys with CPH structure used in the present investigations are Zr-l wt. % Nb and Mg - 4.8 wt % Li alloys. History of material is studied such as cold worked state and annealed condition. Character of alloying elements and their amounts were of interest in this study. The result showed that the higher the percentage radiation hardening, the slower is the migration of radiation defects. Irradiated pure metals recovered at a higher temperature than alloys. Cold work accelerated the migration of radiation defects. The amount of alloying elements had little effect on the recovery temperatures. Character of solute alloying elements (substitutional or interstitial) revealed sensitive effect on the migration of radiation defects. Rate of migration of defects causing hardening can be different from those causing embrittlement. (author)

Present status of the disk pressure tests for hydrogen embrittlement

International Nuclear Information System (INIS)

Fidelle, J.P.

1985-05-01

The Disk Pressure Tests (DPT) have been developed considerably theoretically and experimentally for Internal Hydrogen Embrittlement (IHE) e.g. Co, Ti, U alloys, for Environment Embrittlement due to H 2 , hydrogenated media such as water vapor, alcohol, machining fluids or liquid NH 3 . The range has been expanded considerably for pressure up to 300 MPa and temperature (-160 0 C to 1000 0 C). Very low strain rate -longer than a month- tests have been able to evidence embrittlement of FFC alloys where H diffusivity is low. Conversely for very oxidation - sensitive metals (e.g. Nb and Ta) effects may appear only at somewhat high rates. The relationship between dynamic (increasing stress) tests, static (delayed failure) and low-cycle fatigue tests has been determined. In a number of instances, including SCC, other techniques and even fracture mechanics have been compared to the DPT and proved at best equivalent and several times, less sensitive than a well conducted DPT. At extreme they could not reproduce the field service phenomenon whereas the DPT did and could also be applied satisfactorily to low yield stress materials. The main rupture aspects have been analyzed mechanically and organized in a rational and comprehensive chart based on 12,000 + tests over 150 + materials in different conditions. From the tests on a large number of metal systems, a theory of HE has been derived which accounts for the behavior of metals and alloys either embrittled and or hydrited. Finally comparison of HGE tests and service behavior of a large variety of materials and industrial equipments has made possible to specify acceptance criteria for industrial service

Hydrogen embrittlement susceptibility of laser-hardened 4140 steel

Energy Technology Data Exchange (ETDEWEB)

Tsay, L.W.; Lin, Z.W. [Nat. Taiwan Ocean Univ., Keelung (Taiwan). Inst. of Mater. Eng.; Shiue, R.K. [Institute of Materials Sciences and Engineering, National Dong Hwa University, Hualien, Taiwan (Taiwan); Chen, C. [Institute of Materials Sciences and Engineering, National Taiwan University, Taipei, Taiwan (Taiwan)

2000-10-15

Slow strain rate tensile (SSRT) tests were performed to investigate the susceptibility to hydrogen embrittlement of laser-hardened AISI 4140 specimens in air, gaseous hydrogen and saturated H{sub 2}S solution. Experimental results indicated that round bar specimens with two parallel hardened bands on opposite sides along the loading axis (i.e. the PH specimens), exhibited a huge reduction in tensile ductility for all test environments. While circular-hardened (CH) specimens with 1 mm hardened depth and 6 mm wide within the gauge length were resistant to gaseous hydrogen embrittlement. However, fully hardened CH specimens became susceptible to hydrogen embrittlement for testing in air at a lower strain rate. The strength of CH specimens increased with decreasing the depth of hardened zones in a saturated H{sub 2}S solution. The premature failure of hardened zones in a susceptible environment caused the formation of brittle intergranular fracture and the decrease in tensile ductility. (orig.)

Stress corrosion cracking and hydrogen embrittlement of an Al-Zn-Mg-Cu alloy

International Nuclear Information System (INIS)

Song, R.G.; Dietzel, W.; Zhang, B.J.; Liu, W.J.; Tseng, M.K.; Atrens, A.

2004-01-01

The age hardening, stress corrosion cracking (SCC) and hydrogen embrittlement (HE) of an Al-Zn-Mg-Cu 7175 alloy were investigated experimentally. There were two peak-aged states during ageing. For ageing at 413 K, the strength of the second peak-aged state was slightly higher than that of the first one, whereas the SCC susceptibility was lower, indicating that it is possible to heat treat 7175 to high strength and simultaneously to have high SCC resistance. The SCC susceptibility increased with increasing Mg segregation at the grain boundaries. Hydrogen embrittlement (HE) increased with increased hydrogen charging and decreased with increasing ageing time for the same hydrogen charging conditions. Computer simulations were carried out of (a) the Mg grain boundary segregation using the embedded atom method and (b) the effect of Mg and H segregation on the grain boundary strength using a quasi-chemical approach. The simulations showed that (a) Mg grain boundary segregation in Al-Zn-Mg-Cu alloys is spontaneous, (b) Mg segregation decreases the grain boundary strength, and (c) H embrittles the grain boundary more seriously than does Mg. Therefore, the SCC mechanism of Al-Zn-Mg-Cu alloys is attributed to the combination of HE and Mg segregation induced grain boundary embrittlement

Dependence of hydrogen-induced lattice defects and hydrogen embrittlement of cold-drawn pearlitic steels on hydrogen trap state, temperature, strain rate and hydrogen content

International Nuclear Information System (INIS)

Doshida, Tomoki; Takai, Kenichi

2014-01-01

The effects of the hydrogen state, temperature, strain rate and hydrogen content on hydrogen embrittlement susceptibility and hydrogen-induced lattice defects were evaluated for cold-drawn pearlitic steel that absorbed hydrogen in two trapping states. Firstly, tensile tests were carried out under various conditions to evaluate hydrogen embrittlement susceptibility. The results showed that peak 2 hydrogen, desorbed at temperatures above 200 °C as determined by thermal desorption analysis (TDA), had no significant effect on hydrogen embrittlement susceptibility. In contrast, hydrogen embrittlement susceptibility increased in the presence of peak 1 hydrogen, desorbed from room temperature to 200 °C as determined by TDA, at temperatures higher than −30 °C, at lower strain rates and with higher hydrogen content. Next, the same effects on hydrogen-induced lattice defects were also evaluated by TDA using hydrogen as a probe. Peak 2 hydrogen showed no significant effect on either hydrogen-induced lattice defects or hydrogen embrittlement susceptibility. It was found that hydrogen-induced lattice defects formed under the conditions where hydrogen embrittlement susceptibility increased. This relationship indicates that hydrogen embrittlement susceptibility was higher under the conditions where the formation of hydrogen-induced lattice defects tended to be enhanced. Since hydrogen-induced lattice defects formed by the interaction between hydrogen and strain were annihilated by annealing at a temperature of 200 °C, they were presumably vacancies or vacancy clusters. One of the common atomic-level changes that occur in cold-drawn pearlitic steel showing higher hydrogen embrittlement susceptibility is the formation of vacancies and vacancy clusters

The role of pressure vessel embrittlement in the long term operation of nuclear power plants

International Nuclear Information System (INIS)

Ballesteros, A.; Ahlstrand, R.; Bruynooghe, C.; Estorff, U. von; Debarberis, L.

2012-01-01

Highlights: ► Relevant open scientific issues for the long term operation of RPVs are discussed (flux effect, late blooming phases, etc.). ► Several European and American research programmes dealing with these open issues are reviewed. ► A method for consolidation and preservation of knowledge in this field is presented. - Abstract: The lack of new build of plants over the last twenty years has resulted in a switch within the industry from design, construction and development of new systems to the strengthening of safety systems and to the life extension, or long term operation (LTO), of existing reactors. The most relevant component of any nuclear power plan (NPP) is the reactor pressure vessel (RPV). This is because currently the RPV is still considered irreplaceable or prohibitively expensive to replace. This means, that if it degrades sufficiently, it could be the operational life limiting feature of the NPP. A RPV operational life of 60 years is being considered frequently by many utilities in their plant life management programmes. Areas of improvement facing long term operation are the reduction of uncertainties in the embrittlement parameters of irradiated vessels, and the development of embrittlement trend curves at high fluence levels, where surveillance data are scarce. Different techniques can be used to upgrade the surveillance programmes, as the use of miniature or reconstituted specimens and the application of best estimate assessment tools (e.g. Master Curve). Several relevant international research projects are on-going or have been proposed to clarify the material condition of long operated vessels. Knowledge management is a complementary tool, but not for it less important. The general context for LTO of RPVs is presented in this paper. Starting with a review of relevant embrittlement issues still open, followed by presenting the different techniques and tools that can be used to support LTO, and summarising the scopes of relevant European

Preirradiation microstructrual development designed to minimize properties degradation during irradiation in austentic alloys

International Nuclear Information System (INIS)

Maziasz, P.J.; Roche, T.K.

1981-01-01

The first-generation Prime Candidate Alloy (PCA) for the austenitic stainless steel class of alloys for application as a Magnet Fusion Energy (MFE) first-wall material is a 14 Cr-16 Ni-0.25 Ti modification of Type 316 stainless steel. A key parameter for material performance is wall lifetime. The ability of the material to resist swelling and resist embrittlement during irradiation is important to longer wall lifetimes. The microstructure that evolves during irradiation is primarily responsible for both the swelling and embrittlement responses, and helium plays a central role in this microstructural evolution. This paper indicated how preirradiation microstructures that employ control of MC precipitation and dislocation density are designed and produced for fusion application of PCA

Recent evaluation of 'wet' thermal annealing to resolve reactor pressure vessel embrittlement

International Nuclear Information System (INIS)

Server, W.L.; Biemiller, E.C.

1993-01-01

Prior to the decision to close the Yankee Rowe plant in 1992, a great deal of effort was expended in trying to resolve the degree of neutron embrittlement that the reactor pressure vessel had experienced after 30 years of operation. One mitigative measure that was examined in detail was the possibility of performing a relatively low temperature thermal anneal (at approximately 650 deg. F) to partially restore the original design level of mechanical properties of the reactor pressure vessel beltline region which were lost due to the neutron radiation exposure. This low temperature anneal was to involve heating of the primary coolant water using pump heat in a similar manner as that used to anneal the Belgian BR-3 reactor pressure vessel in the early 1980s. This 'wet' anneal was successful in recovering mechanical properties for the BR-3 vessel, but the extent of the recovery, as well as the rate of re-embrittlement after the anneal, were issues that were difficult to quantify since the exact reactor pressure vessel steels were not available for experimental verification. For the case of Yankee Rowe, material was available from past surveillance programs for at least one of the materials in the vessel, as well as materials obtained from various sources which could act as bounding surrogates. An irradiation /annealing/reirradiation program was developed to better quantify the degree of recovery and re-embrittlement for these materials, but this program was halted before significant test results were obtained. Prior to the initiation of the testing program, a review of past annealing data was performed and the data were scrutinized for direct relevance to the annealing response of the Yankee Rowe vessel. This paper discusses the results derived from this review. The results from the critical review of the past annealing data indicated that a 'wet' anneal of the Yankee Rowe vessel may have been successful in reducing the degree of embrittlement to the point that the

Reactor pressure vessel embrittlement of NPP borssele: Design lifetime and lifetime extension

International Nuclear Information System (INIS)

Blom, F.J.

2007-01-01

Embrittlement of the reactor pressure vessel of the Borssele nuclear power plant has been investigated taking account of the design lifetime of 40 years and considering 20 years subsequent lifetime extension. The paper presents the current licensing status based on considerations of material test data and of US nuclear regulatory standards. Embrittlement status is also evaluated against German and French nuclear safety standards. Results from previous fracture toughness and Charpy tests are investigated by means of the Master curve toughness transition approach. Finally, state of the art insights are investigated by means of literature research. Regarding the embrittlement status of the reactor pressure vessel of Borssele nuclear power plant it is concluded that there is a profound basis for the current license up to the original end of the design life in 2013. The embrittlement temperature changes only slightly with respect to the acceptance criterion adopted postulating further operation up to 2033. Continued safe operation and further lifetime extension are therefore not restricted by reactor pressure vessel embrittlement

Oxidation-induced embrittlement and structural changes of Zircaloy-4 tubing in steam at 700-1000 deg. C

Energy Technology Data Exchange (ETDEWEB)

Ali, A E; Huessein, A G; El-Sayed, A A; El Banna, O A [Atomic Energy Authority, Cairo (Egypt); El Raghy, S M [Cairo Univ. (Egypt). Faculty of Engineering

1997-02-01

The oxidation-induced embrittlement and structural changes of Zircaloy-4 (KWU-Type) tubing was investigated under light water reactors (LWR) Loss-of-Coolant. Accident conditions (LOCA) in temperature range 700-1000 deg. C. The effect of hydrogen addition to steam was also investigated in the temperature range 800-1000 deg. C. The oxidation-induced embrittlement was found to be a function of both temperature and time. Fractography investigation of oxidized tubing showed a typical brittle fracture in the stabilized-alpha zone. The microhardness measurements revealed that the alpha-Zr is harder than that near the mid-wall position. The oxidation-induced embrittlement at 900 deg. C was found to be higher than at 1000 deg. C. The results also indicated that the addition of 5% by volume hydrogen to steam resulted in an increase in the degree of embrittlement. (author). 22 refs, 9 figs, 3 tabs.

Grain size effect on the mechanical properties of neutron irradiated niobium

International Nuclear Information System (INIS)

Gusev, M. N.; Maksimkin, O.P.

2000-01-01

Samples for mechanical tests were prepared from niobium of technical purity and have form of plates (10·3.5 ·0.3mm) with grain size from 2 to 100 mcm. Neutron irradiation was carried out at the reactor WWR-K to the fluence of 2·10 22 n/m 2 ( Angstroem >0.1 MeV). Tests on uniaxial tension at 293K were performed at the facility, evolving Calvet's microcalorimeter and miniature rapture machine. The developed technique enabled to record heat effects just during the deformation process. As experimental results the characteristics of strength and ductility were defined, as well as values of the latent energy E s , accumulated in material in the process of its deformation up to the moment of destruction. It was found that irradiation of niobium with large-grain structure by neutrons leads to increasing of strength characteristics (yield strength σ 0 .2 changes from 130 to 210 MPa, time-resistance σ b from 200 to 230 MPa) and decreasing of ductility from 36 to 28%. As this takes place the capability of the material to accumulate and dissipate energy of plastic deformation suffers substantial change. There were revealed some additional effects, for instance, the radiation annealing hardening (RAH) (i.e. additional change of properties of irradiated material at annealing), whose maximum takes place at 473K. Its temperature and kinetic parameters were determined in this work. Decreasing of grain size usually leads to decreasing of strengthening under irradiation and to decreasing of RAH effect intensity at subsequent annealing. At the same time decreasing of radiation embrittlement is observed. Consequently, creation of fine-grain structure for some cases can favored the stability of material's properties under irradiation. The obtained results are discussed in context of views on grain boundaries as a defect sink. The relation 'grain boundary volume - grain matrix volume', its influence on RAH-effect and value of latent energy are considered

Initial assessment of the mechanisms and significance of low-temperature embrittlement of cast stainless steels in LWR systems

International Nuclear Information System (INIS)

Chopra, O.K.; Sather, A.

1990-08-01

This report summarizes work performed by Argonne National Laboratory on long-term embrittlement of cast duplex stainless steels in LWR systems. Metallurgical characterization and mechanical property data from Charpy-impact, tensile, and J-R curve tests are presented for several experimental and commercial heats, as well as for reactor-aged CF-3, CF-8, and CF-8M cast stainless steels. The effects of material variables on the embrittlement of cast stainless steels are evaluated. Chemical composition and ferrite morphology strongly affect the extent and kinetics of embrittlement. In general, the low-carbon CF-3 stainless steels are the most resistant and the molybdenum-containing high-carbon CF-8M stainless steels are most susceptible to embrittlement. The microstructural and mechanical-property data are analyzed to establish the mechanisms of embrittlement. The procedure and correlations for predicting the impact strength and fracture toughness of cast components during reactor service are described. The lower bound values of impact strength and fracture toughness for low-temperature-aged cast stainless steel are defined. 39 refs., 56 figs., 8 tabs

Long-term aging embrittlement of cast duplex stainless steels in LWR systems

International Nuclear Information System (INIS)

Chopra, O.K.; Chung, H.M.

1991-01-01

The primary objectives of this program are to investigate the significance of in-service embrittlement of cast duplex stainless steels in light water reactor (LWR) systems and to evaluate possible remedies for the embrittlement problem in existing and future plants. The scope of the investigation includes three goals: (1) develop a methodology and correlations for predicting the toughness loss suffered by cast stainless steel components during normal and extended life of LWRs, (2) validate the simulation of in-reactor degradation by accelerated aging, and (3) establish the effects of key compositional and metallurgical variables on the kinetics and extent of embrittlement. The emphasis during the current year was on developing a procedure and correlations for predicting fracture toughness J-R curves of aged cast stainless steels from known material information. The present analysis has focused on developing correlations for the fracture properties in terms of material information that can be determined from the certified material test record (CMTR) and on ensuring that the correlations are adequately conservative for structurally weak materials

Evaluation of the current status of hydrogen embrittlement and stress-corrosion cracking in steels

Energy Technology Data Exchange (ETDEWEB)

Moody, N.R.

1981-12-01

A review of recent studies on hydrogen embrittlement and stress-corrosion cracking in steels shows there are several critical areas where data is either ambiguous, contradictory, or non-existent. A relationship exists between impurity segregation and hydrogen embrittlement effects but it is not known if the impurities sensitize a preferred crack path for hydrogen-induced failure or if impurity and hydrogen effects are additive. Furthermore, grain boundary impurities may enhance susceptibility through interactions with some environments. Some studies show that an increase in grain size increases susceptibility; at least one study shows an opposite effect. Recent work also shows that fracture initiates at different locations for external and internal hydrogen environments. How this influences susceptibility is unknown.

Effect of ternary solute interaction on interfacial segregation and grain boundary embrittlement

Czech Academy of Sciences Publication Activity Database

Lejček, Pavel

2013-01-01

Roč. 48, č. 14 (2013), 4965-4972 ISSN 0022-2461 R&D Projects: GA MŠk(CZ) LM2011026; GA ČR GAP108/12/0144 Institutional research plan: CEZ:AV0Z10100520 Keywords : interfacial segregation * intergranular embrittlement * solute interaction * modeling * thermodynamics Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 2.305, year: 2013

Irradiation-assisted stress corrosion cracking of austenitic alloys

International Nuclear Information System (INIS)

Was, G.S.; Atzmon, M.

1991-01-01

An experimental program has been conducted to determine the mechanism of irradiation-assisted stress-corrosion cracking (IASCC) in austenitic stainless steel. High-energy protons have been used to produce grain boundary segregation and microstructural damage in samples of controlled impurity content. The densities of network dislocations and dislocation loops were determined by transmission electron microscopy and found to resemble those for neutron irradiation under LWR conditions. Grain boundary compositions were determined by in situ fracture and Auger spectroscopy, as well as by scanning transmission electron microscopy. Cr depletion and Ni segregation were observed in all irradiated samples, with the degree of segregation depending on the type and amount of impurities present. P, and to a lesser extent P, impurities were observed to segregate to the grain boundaries. Irradiation was found to increase the susceptibility of ultra-high-purity (UHP), and to a much lesser extent of UHP+P and UHP+S, alloys to intergranular SCC in 288 degree C water at 2 ppm O 2 and 0.5 μS/cm. No intergranular fracture was observed in arcon atmosphere, indicating the important role of corrosion in the embrittlement of irradiated samples. The absence of intergranular fracture in 288 degree C argon and room temperature tests also suggest that the embrittlement is not caused by hydrogen introduced by irradiation. Contrary to common belief, the presence of P impurities led to a significant improvement in IASCC over the ultrahigh purity alloy

Reduction of helium embrittlement in stainless steel by finely dispersed TiC precipitates

International Nuclear Information System (INIS)

Kesternich, W.; Rothaut, J.

1982-01-01

The He embrittlement effects in two candidate stainless steels for first wall of fusion reactors were studied in creep tests at 700 0 C simulating the He production by He implantation. Creep rupture life before He implantation and reduction of rupture life due to He were superior by orders of magnitude in 1.4970 steel after pertinent pretreatment compared to 316 steel. The high strength and the low He embrittlement result from a fine dispersion of TiC precipitates in the grain interiors. From microstructural investigations a mechanism explaining the high sink efficiency of TiC for He atom accumulation is suggested. (orig.)

Deformation and Fracture Properties in Neutron Irradiated Pure Mo and Mo Alloys

International Nuclear Information System (INIS)

Byun, T.S.; Snead, L.; Li, M.; Cockeram, B.V.

2007-01-01

Full text of publication follows: The evolution in microstructural and mechanical properties was investigated for molybdenum and molybdenum alloys after high temperature neutron irradiation. Test materials include oxide dispersion-strengthened (ODS) molybdenum alloy, molybdenum- 0.5% titanium-0.1% zirconium (TZM) alloy, and low carbon arc-cast (LCAC) molybdenum. Tensile specimens were irradiated in high flux isotope reactor (HFIR) at temperatures in the range ∼300 - 1000 deg. C to neutron fluences of 2.28 - 24.7 x 10 25 n/m 2 (E>0.1 MeV) or 1.2-13.1 dpa. Tensile tests were performed at temperatures ranging from -150 deg. C to 1000 deg. C. To evaluate irradiation effects, true stress parameters (yield stress, plastic instability stress, and true fracture stress) and ductility parameters (uniform strain, fracture strain, and reduction area) were compared for both irradiated and non-irradiated materials. Fracture toughness was also evaluated from the fracture stress and fracture strain data using a fracture strain model. The fracture strain was used to determine the ductile-to-brittle transition temperature (DBTT). Results indicate that irradiation in the temperature range of 600 - 800 deg. C hardened the materials by up to 70%, while the irradiation hardening outside this temperature range was much lower (<40%). The plastic instability stress was strongly dependent on test temperature; however, it was nearly independent of irradiation dose and temperature. It was also found that the true fracture stress was dependent on test temperature. The true fracture stress was not significantly influenced by irradiation at elevated and high test temperatures; however, it was decreased significantly at sub-zero temperatures after irradiation due to material embrittlement. The DBTT for 600 deg. C irradiated ODS molybdenum alloy was found to be about room temperature or lower, and among the test materials the ODS alloy showed the highest resistance to irradiation embrittlement

Irradiation effects test Series Scoping Test 1: test results report

International Nuclear Information System (INIS)

Quapp, W.J.; Allison, C.M.; Farrar, L.C.

1977-09-01

The report describes the results of the first scoping test in the Irradiation Effects Test Series conducted by the Thermal Fuels Behavior Program, which is part of the Water Reactor Research Program of EG and G Idaho, Inc. The research is sponsored by the United States Nuclear Regulatory Commission. This test used an unirradiated, three-foot-long, PWR-type fuel rod. The objective of this test was to thoroughly evaluate the remote fabrication procedures to be used for irradiated rods in future tests, handling plans, and reactor operations. Additionally, selected fuel behavior data were obtained. The fuel rod was subjected to a series of preconditioning power cycles followed by a power increase which brought the fuel rod power to about 20.4 kW/ft peak linear heat rating at a coolant mass flux of 1.83 x 10 6 lb/hr-ft 2 . Film boiling occurred for a period of 4.8 minutes following flow reductions to 9.6 x 10 5 and 7.5 x 10 5 lb/hr-ft 2 . The test fuel rod failed following reactor shutdown as a result of heavy internal and external cladding oxidation and embrittlement which occurred during the film boiling operation

Irradiation effects of 11 MeV protons on ferritic steels

International Nuclear Information System (INIS)

Hamaguchi, Yoshikazu; Kuwano, Hisashi; Misawa, Toshihei

1985-01-01

It is considered that ferritic/martensitic steels are the candidate of the first wall materials for future fusion reactors. The most serious problem in the candidate materials is the loss of ductility due to the elevation of ductile-brittle transition temperature by the high dpa irradiation of neutrons. 14 MeV neutrons produced by D-T reaction cause high dpa damage and also produce large quantity of helium and hydrogen atoms in first wall materials. Those gas atoms also play an important role in the embrittlement of steels. The main purpose of this work was to simulate the behavior of hydrogen produced by the transmutation in the mechanical properties of ferritic steels when they were irradiated with 11 MeV protons. The experimental procedure and the results of hardness, the broadening of x-ray diffraction lines, Moessbauer spectroscopy and small punch test are reported. High energy protons of 10 - 20 MeV are suitable to the simulation experiment of 14 MeV neutron radiation damage. But the production of the active nuclei emitting high energy gamma ray and having long life, Co-56, is the most serious problem. Another difficulty is the control of irradiation temperature. A small irradiation chamber must be developed. (Kako, I.)

Irradiation experiments on materials for core internals, pressure vessel and fuel cladding

Energy Technology Data Exchange (ETDEWEB)

Tsukada, Takashi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-03-01

Materials degradation due to the aging phenomena is one of the key issues for the life assessment and extension of the light water reactors (LWRs). This presentation introduces JAERI`s activities in the field of LWR material researches which utilize the research and testing reactors for irradiation experiments. The activities are including the material studies for the core internals, pressure vessel and fuel cladding. These materials are exposed to the neutron/gamma radiation and high temperature water environments so that it is worth reviewing their degradation phenomena as the continuum. Three topics are presented; For the core internal materials, the irradiation assisted stress corrosion cracking (IASCC) of austenitic stainless steels is the present major concern. At JAERI the effects of alloying elements on IASCC have been investigated through the post-irradiation stress corrosion cracking tests in high-temperature water. The radiation embrittlement of pressure vessel steels is still a significant issue for LWR safety, and at JAERI some factors affecting the embrittlement behavior such as a dose rate have been investigated. Waterside corrosion of Zircaloy fuel cladding is one of the limiting factors in fuel rod performance and an in-situ measurement of the corrosion rate in high-temperature water was performed in JMTR. To improve the reliability of experiments and to extent the applicability of experimental techniques, a mutual utilization of the technical achievements in those irradiation experiments is desired. (author)

Mechanical properties of Mo and TZM alloy neutron-irradiated at high temperatures

International Nuclear Information System (INIS)

Ueda, Kazukiyo; Satou, Manabu; Hasegawa, Akira; Abe, Katsunori

1997-01-01

This work reports the mechanical properties of irradiated molybdenum (Mo) and its alloy, TZM. Recrystallized and stress-relieved specimens were irradiated at five temperatures between 373 and 800degC in FFTF/MOTA to fluence levels of 6.8 to 34 dpa. Irradiation embrittlement and hardening were evaluated by three-point bend test and Vickers hardness test, respectively. Stress-relieved materials showed the enough ductility even after high fluence irradiation. The role of layered structure of stress-relieved specimen was discussed. (author)

PR-EDB: Power Reactor Embrittlement Data Base, Version 2. Revision 2, Program description

Energy Technology Data Exchange (ETDEWEB)

Stallmann, F.W.; Wang, J.A.; Kam, F.B.K.; Taylor, B.J. [Oak Ridge National Lab., TN (United States)

1994-01-01

Investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes Standard Review Plans (SRP`s) and Guides for license renewal can be greatly expedited by the use of a well-designed computerized data base. Also, such a data base is essential for the validation of embrittlement prediction models by researchers. The Power Reactor Embrittlement Data Base (PR-EDB) is such a comprehensive collection of data for US commercial nuclear reactors. The current version of the PR-EDB contains the Charpy test data that were irradiated in 252 capsules of 96 reactors and consists of 207 data points for heat-affected-zone (HAZ) materials (98 different HAZ), 227 data points for weld materials (105 different welds), 524 data points for base materials (136 different base materials), including 297 plate data points (85 different plates), 119 forging data points (31) different forging), and 108 correlation monitor materials data points (3 different plates). The data files are given in dBASE format and can be accessed with any computer using the DOS operating system. ``User-friendly`` utility programs are used to retrieve and select specific data, manipulate data, display data to the screen or printer, and to fit and plot Charpy impact data. The results of several studies investigated are presented in Appendix D.

Effect of solute interaction on interfacial and grain boundary embrittlement in binary alloys

Czech Academy of Sciences Publication Activity Database

Lejček, Pavel

2013-01-01

Roč. 48, č. 6 (2013), 2574-2580 ISSN 0022-2461 R&D Projects: GA ČR GAP108/12/0144 Institutional research plan: CEZ:AV0Z10100520 Keywords : interfacial segregation * grain boundary embrittlement * binary interaction * modeling * thermodynamics Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 2.305, year: 2013

Microstructure and grain size effects on irradiation hardening of low carbon steel for reactor tanks

International Nuclear Information System (INIS)

Milasin, N.

1964-05-01

Irradiation hardening of steel for reactor pressure vessels has been studied extensively during the past few years. A great number of experimental results concerning the behaviour of these steels in the radiation field and several review papers (1,2) have been published. Most of the papers deal with the effects of specific metallurgical factors or irradiation conditions (temperature, flux) on irradiation hardening and embrittlement. In addition, a number of experiments are performed to give evidence on the mechanism of irradiation hardening of these steels. However, this mechanism is still unknown due to the complexity of steel as a system. Among different methods used in radiation damage studies, the changes of mechanical properties have been mainly investigated. By using Hall-Petch's empirical relation, σ y =σ i +k y d -1/2 between lower yield stress, σ y , and grain size, 2d, the information about the effect of irradiation on the parameters σ i and k y is obtained. Taking as a base interpretation of σ i and k y given by Petch and his co-workers it has been concluded that radiation does not change the stress to start slip but that it increase the friction that opposes the passage of free dislocations across a slip plane. In attempting to apply Hall-Petch's relation to one unirradiated ferritic steel with a carbon content higher than 0.15% some difficulties were encountered. The results obtained indicate that the influence of grain size can not be isolated from other factors introduced by the treatments used to produce different grain sizes. This paper deals with a similar problem in the case of irradiated steel. The results obtained give the changes of the mechanical properties of steel in neutron irradiation field as a function of microstructure and grain size. In addition, the mechanical properties of irradiated steel are measured after annealing at 150 deg C and 450 deg C. On the basis of the experimental results obtained the relative microstructure and

The effect of helium, radiation damage and irradiation temperature on the mechanical properties of beryllium

Energy Technology Data Exchange (ETDEWEB)

Fabritsiev, S.A. [D.V. Efremov Scientific Research Inst., St. Petersburg (Russian Federation); Pokrovsky, A.S.

1998-01-01

In this work different RF beryllium grades were irradiated in the BOR-60 reactor to a dose of {approx}5-10 dpa at irradiation temperatures 350, 420, 500, 800degC. Irradiation at temperatures of 350-400degC is shown to result in Be hardening due to the accumulation of radiation defect complexes. Hardening is accompanied with a sharp drop in plasticity at T{sub test} {<=} 300degC. A strong anisotropy in plasticity has been found at a mechanical testing temperature of 400degC and this parameter may be preferable when the samples are cut crosswise to the pressing direction. High-temperature irradiation (T{sub irr} = 780degC) gives rise to large helium pores over the grain boundaries and smaller pores in the grain body. Fracture is brittle and intercrystallite at T{sub test} {>=} 600degC. Helium embrittlement is accompanied as well with a drop in the Be strength properties. (author)

Effect of hydrogen on the behavior of metals II - Hydrogen embrittlement of titanium alloy TV13CA - effect of oxygen - comparison with non-alloyed titanium

International Nuclear Information System (INIS)

Arditty, Jean-Pierre

1973-01-01

The effect of oxygen on the hydrogen embrittlement of non-alloyed titanium and the metastable β titanium alloy, TV13 CA, was studied during dynamic mechanical tests, the concentrations considered varying from 1000 to 5000 ppm (oxygen) and from 0 to 5000 ppm (hydrogen) respectively. TV13 CA alloy has a very high solubility for hydrogen. The establishment of a temperature range and a rate of deformation region in which the embrittlement of the alloy is maximum leads to the conclusion that an embrittlement mechanism occurs involving the dragging and accumulation of hydrogen by dislocations. This is the case for all annealings effected in the medium temperature range, which, by favoring the re-establishment of the stable two-phase α + β state of the alloy, produce hardening. The same is true for oxygen which, in addition to hardening the alloy by the solid solution effect, tends to increase its instability and, in consequence, favors the decomposition of the β phase. Nevertheless oxygen concentrations of up to 1500 ppm contribute to increasing the mechanical resistance without catastrophically reducing the deformation capacity. In the case of non-alloyed titanium, the hardening effect also leads to an increase in E 0.2p c and R, and to a reduction in the deformation capacity. Nevertheless, hydrogen is only very slightly soluble at room temperature and a distribution of the hydride phase linked to the thermal history of the sample predominates. Thus a fine acicular structure obtained from the β phase by quenching, enables an alloy having a good mechanical resistance to be conserved even when large quantities of hydrogen are present; the deformation capacity remains small. On the other hand, when the hydride phase separates the metallic phase into large grains, a very small elongation leads to a breakdown in mechanical resistance. (author) [fr

Solubility of hydrogen in metals and its effect of pore-formation and embrittlement. Ph.D. Thesis

Science.gov (United States)

Shahani, H. R.

1984-01-01

The effect of alloying elements on hydrogen solubility were determined by evaluating solubility equations and interaction coefficients. The solubility of dry hydrogen at one atmosphere was investigated in liquid aluminum, Al-Ti, Al-Si, Al-Fe, liquid gold, Au-Cu, and Au-Pd. The design of rapid heating and high pressure casting furnaces used in meta foam experiments is discussed as well as the mechanism of precipitation of pores in melts, and the effect of hydrogen on the shrinkage porosity of Al-Cu and Al-Si alloys. Hydrogen embrittlement in iron base alloys is also examined.

Application of magnetomechanical hysteresis modeling to magnetic techniques for monitoring neutron embrittlement and biaxial stress

International Nuclear Information System (INIS)

Sablik, M.J.; Kwun, H.; Rollwitz, W.L.; Cadena, D.

1992-01-01

The objective is to investigate experimentally and theoretically the effects of neutron embrittlement and biaxial stress on magnetic properties in steels, using various magnetic measurement techniques. Interaction between experiment and modeling should suggest efficient magnetic measurement procedures for determining neutron embrittlement biaxial stress. This should ultimately assist in safety monitoring of nuclear power plants and of gas and oil pipelines. In the first six months of this first year study, magnetic measurements were made on steel surveillance specimens from the Indian Point 2 and D.C. Cook 2 reactors. The specimens previously had been characterized by Charpy tests after specified neutron fluences. Measurements now included: (1) hysteresis loop measurement of coercive force, permeability and remanence, (2) Barkhausen noise amplitude; and (3) higher order nonlinear harmonic analysis of a 1 Hz magnetic excitation. Very good correlation of magnetic parameters with fluence and embrittlement was found for specimens from the Indian Point 2 reactor. The D.C. Cook 2 specimens, however showed poor correlation. Possible contributing factors to this are: (1) metallurgical differences between D.C. Cook 2 and Indian Point 2 specimens; (2) statistical variations in embrittlement parameters for individual samples away from the stated men values; and (3) conversion of the D.C. Cook 2 reactor to a low leakage core configuration in the middle of the period of surveillance. Modeling using a magnetomechanical hysteresis model has begun. The modeling will first focus on why Barkhausen noise and nonlinear harmonic amplitudes appear to be better indicators of embrittlement than the hysteresis loop parameters

Hydrogen embrittlement of high strength steel electroplated with zincâ  cobalt allo

OpenAIRE

Hillier, Elizabeth M. K.; Robinson, M. J.

2004-01-01

Slow strain rate tests were performed on quenched and tempered AISI 4340 steel to measure the extent of hydrogen embrittlement caused by electroplating with zincâ  cobalt alloys. The effects of bath composition and pH were studied and compared with results for electrodeposited cadmium and zincâ  10%nickel. It was found that zincâ  1%cobalt alloy coatings caused serious hydrogen embrittlement (EI 0.63); almost as severe as that of cadmium (EI 0.78). Baking cadmium plate...

Experimental tests of irradiation-anneal-reirradiation effects on mechanical properties of RPV plate and weld materials

International Nuclear Information System (INIS)

Hawthorne, J.R.

1996-01-01

The Charpy-V (C V ) notch ductility and tension test properties of three reactor pressure vessel (RPV) steel materials were determined for the 288 degree C (550 degree F) irradiated (I), 288 degree C (550 degree F) irradiated + 454 degree C (850 degree F)-168 h postirradiation annealed (IA), and 288 degree C (550 degree F) reirradiated (IAR) conditions. Total fluences of the I condition and the IAR condition were, respectively, 3.33 x 10 19 n/cm 2 and 4.18 x 10 19 n/cm 2 , E > 1 MeV. The irradiation portion of the IAR condition represents an incremental fluence increase of 1. 05 x 10 19 n/cm 2 , E > 1 MeV, over the I-condition fluence. The materials (specimens) were supplied by the Yankee Atomic Electric Company and represented high and low nickel content plates and a high nickel, high copper content weld deposit prototypical of the Yankee-Rowe reactor vessel. The promise of the IAR method for extending the fluence tolerance of radiation-sensitive steels and welds is clearly shown by the results. The annealing treatment produced full C V upper shelf recovery and full or nearly full recovery in the C V 41 J (30 ft-lb) transition temperature. The C V transition temperature increases produced by the reirradiation exposure were 22% to 43% of the increase produced by the first cycle irradiation exposure. A somewhat greater radiation embrittlement sensitivity and a somewhat greater reirradiation embrittlement sensitivity was exhibited by the low nickel content plate than the high nickel content plate. Its high phosphorus content is believed to be responsible. The IAR-condition properties of the surface vs. interior regions of the low nickel content plate are also compared

Nano-structural changes in the RPV steels irradiated in MTR to high doses. 3D atom probe and positron annihilation study

International Nuclear Information System (INIS)

Dohi, Kenji; Soneda, Naoki; Nomoto, Akiyoshi; Ishino, Shiori

2005-01-01

Reactor pressure vessel (RPV) steels of life-extended light water reactors are to be exposed to higher neutron fluence. The understanding of radiation embrittlement of RPV steels is very important in order to improve prediction of the embrittlement. The radiation embrittlement is mainly cased by copper-enriched cluster (CEC) and matrix damage (MD) due to irradiation. The state-or-the art technique such as three dimensional atom probe (3DAP) and positron annihilation (PA) has enabled to observe these microstructural features. The effect of highly dose irradiation on the formation of clusters in a low copper base metal and a high copper weld metal is investigated by means of the 3DAP and PA observations in this paper. The materials were irradiated to a neutron fluence of 10 20 n/cm 2 at 290 degC in a test reactor. The 3DAP observation shows that high dense CRCs in size of about 2 nm are formed in the high Cu weld metal. The CRCs consist of Si in addition to Fe, Cu, Mn, and Ni. Solute atom clusters below 2 nm are also observed in low Cu base metal, but the clusters include a large amount of Si and free from Cu. These clusters may be peculiar to highly irradiated materials because of no literature reporting such the clusters in the similar steels irradiated at the lower fluence. The data of the positron annihilation coincidence Doppler broadening measurement for both materials also shows the formation of clusters containing Cu, Ni, Mn, and Si. This means the clusters observed by 3DAP are uniformly distributed in the materials. Hardness tests and PA measurement combined with isochronal annealing show that defects, e.g. dislocation loop etc., having a positron lifetime of about 140 psec influence on mechanical properties of the steels. (author)

Nanostructure evolution under irradiation of Fe(C)MnNi model alloys for reactor pressure vessel steels

Energy Technology Data Exchange (ETDEWEB)

Chiapetto, M., E-mail: mchiapet@sckcen.be [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium); Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France); Becquart, C.S. [Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France); Laboratoire commun EDF-CNRS Etude et Modélisation des Microstructures pour le Vieillissement des Matériaux (EM2VM) (France); Domain, C. [EDF R& D, Département Matériaux et Mécanique des Composants, Les Renardières, F-77250 Moret sur Loing (France); Laboratoire commun EDF-CNRS Etude et Modélisation des Microstructures pour le Vieillissement des Matériaux (EM2VM) (France); Malerba, L. [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium)

2015-06-01

Radiation-induced embrittlement of bainitic steels is one of the most important lifetime limiting factors of existing nuclear light water reactor pressure vessels. The primary mechanism of embrittlement is the obstruction of dislocation motion produced by nanometric defect structures that develop in the bulk of the material due to irradiation. The development of models that describe, based on physical mechanisms, the nanostructural changes in these types of materials due to neutron irradiation are expected to help to better understand which features are mainly responsible for embrittlement. The chemical elements that are thought to influence most the response under irradiation of low-Cu RPV steels, especially at high fluence, are Ni and Mn, hence there is an interest in modelling the nanostructure evolution in irradiated FeMnNi alloys. As a first step in this direction, we developed sets of parameters for object kinetic Monte Carlo (OKMC) simulations that allow this to be done, under simplifying assumptions, using a “grey alloy” approach that extends the already existing OKMC model for neutron irradiated Fe–C binary alloys [1]. Our model proved to be able to describe the trend in the buildup of irradiation defect populations at the operational temperature of LWR (∼300 °C), in terms of both density and size distribution of the defect cluster populations, in FeMnNi model alloys as compared to Fe–C. In particular, the reduction of the mobility of point-defect clusters as a consequence of the presence of solutes proves to be key to explain the experimentally observed disappearance of detectable point-defect clusters with increasing solute content.

A small angle neutron study of irradiation induced microstructures in Cr-Mo-V WWER steels

International Nuclear Information System (INIS)

Levit, Vladimir I.; Santos, Ari S.; Louzada, Ana R.R.; Silveira, Cristina M.; Vaniel, Ana Paula H.; Odette, George R.; Mader, Eric

2000-01-01

Small angle neutron scattering (SANS) has proven to be a very effective technique for characterizing the ultrafine (∼1 nm) irradiation induced microstructures which are responsible for hardening and the concomitant embrittlement of reactor pressure vessel steels. SANS measurement were carried out on three irradiated and unirradiated weld materials of WWER- type on 8 m instrument at the National Institute of Standards and Technology, Washington, USA. Small (r m < 1 nm) irradiation induced features were found for all three materials. Were found volume fractions, number densities and ratios of magnetic to nuclear scattering. Some analyses of the irradiation induced precipitation nature and possible chemical composition were made by comparison of the results with other reactor materials SANS and Atom Probe Field Ion Microscopy data. (author)

Effect of microstructure on the impact toughness and temper embrittlement of SA508Gr.4N steel for advanced pressure vessel materials.

Science.gov (United States)

Yang, Zhiqiang; Liu, Zhengdong; He, Xikou; Qiao, Shibin; Xie, Changsheng

2018-01-09

The effect of microstructure on the impact toughness and the temper embrittlement of a SA508Gr.4N steel was investigated. Martensitic and bainitic structures formed in this material were examined via scanning electron microscopy, electron backscatter diffraction, transmission electron microscopy, and Auger electron spectroscopy (AES) analysis. The martensitic structure had a positive effect on both the strength and toughness. Compared with the bainitic structure, this structure consisted of smaller blocks and more high-angle grain boundaries (HAGBs). Changes in the ultimate tensile strength and toughness of the martensitic structure were attributed to an increase in the crack propagation path. This increase resulted from an increased number of HAGBs and refinement of the sub-structure (block). The AES results revealed that sulfur segregation is higher in the martensitic structure than in the bainitic structure. Therefore, the martensitic structure is more susceptible to temper embrittlement than the bainitic structure.

Significance of rate of work hardening in tempered martensite embrittlement

International Nuclear Information System (INIS)

Pietikainen, J.

1995-01-01

The main explanations for tempered martensite embrittlement are based on the effects of impurities and cementite precipitation on the prior austenite grain boundaries. There are some studies where the rate of work hardening is proposed as a potential reason for the brittleness. One steel was studied by means of a specially developed precision torsional testing device. The test steel had a high Si and Ni content so ε carbide and Fe 3 C appear in quite different tempering temperature ranges. The M S temperature is low enough so that self tempering does not occur. With the testing device it was possible to obtain the true stress - true strain curves to very high deformations. The minimum toughness was always associated with the minimum of rate of work hardening. The change of deformed steel volume before the loss of mechanical stability is proposed as at least one reason for tempered martensite embrittlement. The reasons for the minimum of the rate of work hardening are considered. (orig.)

Development of new irradiation facility for BWR safety research

International Nuclear Information System (INIS)

Okada, Yuji; Magome, Hirokatsu; Iida, Kazuhiro; Hanawa, Hiroshi; Ohmi, Masao

2013-01-01

In JAEA (Japan Atomic Energy Agency), about the irradiation embrittlement of the reactor pressure vessel and the stress corrosion cracking of reactor core composition apparatus concerning the long-term use of the light water reactor (BWR), in order to check the influence of the temperature, pressure, and water quality, etc on BWR condition. The water environmental control facility which performs irradiation assisted stress corrosion-cracking (IASCC) evaluation under BWR irradiation environment was fabricated in JMTR (Japan Materials Testing Reactor). This report is described the outline of manufacture of the water environmental control facility for doing an irradiation test using the saturation temperature capsule after JMTR re-operation. (author)

Analytical electron microscopy of neutron-irradiated reactor alloys

International Nuclear Information System (INIS)

Thomas, L.E.

1982-01-01

Exposure to the high neutron fluxes and temperatures from 400 to 650 0 C in the core region of a fast breeder reactor profoundly alters the microstructure and properties of structural steels and superalloys. The development of irradiation-induced voids, dislocations and precipitates, as well as segregation of alloying elements on a microscopic scale has been related to macroscopic swelling, creep, hardening and embrittlement which occur during prolonged exposures in reactor. Microanalytical studies using TEM/STEM methods, primarily energy dispersive x-ray (EDX) microanalysis, have greatly aided understanding of alloy behavior under irradiation. The main uses of analytical electron microscopy in studying irradiated alloys have been the identification of irradiation-induced precipitates and determination of the changes in local composition due to irradiation-induced solute segregation

Heavy-section steel irradiation program. Semiannual progress report, September 1993--March 1994

International Nuclear Information System (INIS)

Corwin, W.R.

1995-04-01

Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. The RPV is the only component in the primary pressure boundary for which, if it should rupture, the engineering safety systems cannot assure protection from core damage. It is therefore imperative to understand and be able to predict the capabilities and limitations of the integrity inherent in the RPV. In particular, ft is vital to fully understand the degree of irradiation-induced degradation of the RPV's fracture resistance that occurs during service. The Heavy-Section Steel (HSS) Irradiation Program has been established; its primary goal is to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior, and in particular the fracture toughness properties of typical pressure-vessel steels, as they relate to light-water RPV integrity. The program includes the direct continuation of irradiation studies previously conducted within the HSS Technology Program augmented by enhanced examinations of the accompanying microstructural changes. During this period, the report on the duplex-type crack-arrest specimen tests from Phase 11 of the K la program was issued, and final preparations for testing the large, irradiated crack-arrest specimens from the Italian Committee for Research and Development of Nuclear Energy and Alternative Energies were completed. Tests on undersize Charpy V-notch (CVN) energy specimens in the irradiated and annealed weld 73W were completed. The results are described in detail in a draft NUREG report. In addition, the ORNL investigation of the embrittlement of the High Flux Isotope RPV indicated that an unusually large ratio of the high-energy gamma-ray flux to fast-neutron flux is most likely responsible for the apparently accelerated embrittlement

The evolution of mechanical property change in irradiated austenitic stainless steels

International Nuclear Information System (INIS)

Lucas, G.E.

1993-01-01

The evolution of mechanical properties in austenitic stainless steels during irradiation is reviewed. Changes in strength, ductility and fracture toughness are strongly related to the evolution of the damage microstructure and microstructurally-based models for strengthening reasonably correlate the data. Irradiation-induced defects promote work softening and flow localization which in turn leads to significant reductions in ductility and fracture toughness beyond about 10 dpa. The effects of irradiation on fatigue appear to be modest except at high temperature where helium embrittlement becomes important. The swelling-independent component of irradiation creep strain increases linearly with dose and is relatively insensitive to material variables and irradiation temperature, except at low temperatures where accelerated creep may occur as a result of low vacancy mobility. Creep rupture life is a strong function of helium content, but is less sensitive to metallurgical conditions. Irradiation-induced stress corrosion cracking appears to be related to the evolution of radiation-induced segregation/depletion at grain boundaries, and hence may not be significant at low irradiation temperatures. (orig.)

The strengthening of embrittled books using gamma radiation

International Nuclear Information System (INIS)

Egan, A.; Mardian, J.; Foot, M.; King, E.; Millington, A.; Nevin, M.; Butler, C.; Barker, J.; Fletcher, D.

1995-01-01

The embrittlement of papers, manufactured through processes introduced in the mid-19th century, has caused many millions of books to become fragile, even to the point of being unusable. During the 1980s the British Library funded a research programme, carried out at the University of Surrey, to develop a technology which could be used to treat brittle books on a large scale, with the goal of greatly extending their useful life. The process developed, known as graft co-polymerization, involves three stages: i) application of a cocktail of monomers to the book's pages; ii) equilibration of these monomers throughout the text block; and iii) a low, slow dose of γ-radiation to effect polymerization. In collaboration with the British Library, Nordion International has designed a full-scale book-strengthening plant capable of processing between 200,000 and 500,000 and 500,000 books per year, with estimated prices to customers in the region of 1 8-10 per volume (US $12-16). In order to test the equipment and procedures that would be involved in such a plant, pilot-scale equipment has been designed and assembled on the premises of Isotron plc, where use is made of a conventional irradiator. This paper gives details of the graft co-polymerization process, and some results of the pilot-scale work, in terms of both efficacy and controllability. It also discusses the technical and economic feasibility of building and running a full-scale plant. (author)

Embrittlement of nickel-, cobalt-, and iron-base superalloys by exposure to hydrogen

Science.gov (United States)

Gray, H. R.

1975-01-01

Five nickel-base alloys (Inconel 718, Udimet 700, Rene 41, Hastelloy X, and TD-NiCr), one cobalt-base alloy (L-605), and an iron-base alloy (A-286) were exposed in hydrogen at 0.1 MN/sq m (15 psi) at several temperatures in the range from 430 to 980 C for as long as 1000 hours. These alloys were embrittled to varying degrees by such exposures in hydrogen. Embrittlement was found to be: (1) sensitive to strain rate, (2) reversible, (3) caused by large concentrations of absorbed hydrogen, and (4) not associated with any detectable microstructural changes in the alloys. These observations are consistent with a mechanism of internal reversible hydrogen embrittlement.

Use of Reactor Pressure Vessel Surveillance Materials for Extended Life Evaluations Using Power and Test Reactor Irradiations

International Nuclear Information System (INIS)

Server, W.L.; Nanstad, R.K.; Odette, G.R.

2012-01-01

The most important component in assuring safety of the nuclear power plant is the reactor pressure (RPV). Surveillance programs have been designed to cover the licensed life of operating nuclear RPVs. The original surveillance programs were designed when the licensed life was 40 years. More than one-half of the operating nuclear plants in the USA have an extended license out to 60 years, and there are plans to continue to operate many plants out to 80 years. Therefore, the surveillance programs have had to be adjusted or enhanced to generate key data for 60 years, and now consideration must be given for 80 or more years. To generate the necessary data to assure safe operation out to these extended license lives, test reactor irradiations have been initiated with key RPV and model alloy steels, which include several steels irradiated in the current power reactor surveillance programs out to relatively high fluence levels. These data are crucial in understanding the radiation embrittlement mechanisms and to enable extrapolation of the irradiation effects on mechanical properties for these extended time periods. This paper describes the potential radiation embrittlement mechanisms and effects when assessing much longer operating times and higher neutron fluence levels. Potential methods for adjusting higher neutron flux test reactor data for use in predicting power reactor vessel conditions are discussed. (author)

The effect of low dose rate irradiation on the swelling of 12% cold-worked 316 stainless steel

International Nuclear Information System (INIS)

Allen, T. R.

1999-01-01

In pressurized water reactors (PWRs), stainless steel components are irradiated at temperatures that may reach 400 C due to gamma heating. If large amounts of swelling (>10%) occur in these reactor internals, significant swelling related embrittlement may occur. Although fast reactor studies indicate that swelling should be insignificant at PWR temperatures, the low dose rate conditions experienced by PWR components may possibly lead to significant swelling. To address these issues, JNC and ANL have collaborated to analyze swelling in 316 stainless steel, irradiated in the EBR-II reactor at temperatures from 376-444 C, at dose rates between 4.9 x 10 -8 and 5.8 x 10 -7 dpa/s, and to doses of 56 dpa. For these irradiation conditions, the swelling decreases markedly at temperatures less than approximately 386 C, with the extrapolated swelling at 100 dpa being around 3%. For temperatures greater than 386 C, the swelling extrapolated to 100 dpa is around 9%. For a factor of two difference in dose rate, no statistically significant effect of dose rate on swelling was seen. For the range of dose rates analyzed, the swelling measurements do not support significant (>10%) swelling of 316 stainless steel in PWRs

Susceptibility of 2 1/4 Cr-1Mo steel to liquid metal induced embrittlement by lithium-lead solutions

International Nuclear Information System (INIS)

Eberhard, B.A.; Edwards, G.R.

1984-08-01

An investigation has been conducted on the liquid metal induced embrittlement susceptibility of 2 1/4Cr-1Mo steel exposed to lithium and 1a/o lead-lithium at temperatures between 190 0 C and 525 0 C. This research was part of an ongoing effort to evaluate the compatibility of liquid lithium solutions with potential fusion reactor containment materials. Of particular interest was the microstructure present in a weld heat-affected zone, a microstructure known to be highly susceptible to corrosive attack by liquid lead-lithium solutions. Embrittlement susceptibility was determined by conducting tension tests on 2 1/4Cr-1Mo steel exposed to an inert environment as well as to a lead-lithium liquid and observing the change in tensile behavior. The 2 1/4Cr-1Mo steel was also given a base plate heat treatment to observe its embrittlement susceptibility to 1a/o lead-lithium. The base plate microstructure was severely embrittled at temperatures less than 500 0 C. Tempering the base plate was effective in restoring adequate ductility to the steel

Power reactor embrittlement data base (PR-EDB): Uses in evaluating radiation embrittlement of reactor vessels

International Nuclear Information System (INIS)

Kam, F.B.K.; Stallmann, F.W.; Wang, J.A.

1992-01-01

Investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current Codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed, computerized data base. Also, such a data is essential for the evaluation of embrittlement prediction models by researchers. The Power Reactor Embrittlement Data Base (PR-EDB) is such a comprehensive collection of data for US commercial nuclear reactors. The current compilation contains data from 92 reactors and consists of 175 data points for weld materials (79 different welds) and 395 data points for base materials (110 different base materials). The different types of data that are implemented or planned for this data base are discussed. ''User-friendly'' utility programs have been written to investigate a list of problems using this data base. The utility programs are also used to add and upgrade data, retrieve and select specific data, manipulate data, display data to the screen or printer, and to fit and plot Charpy impact data. The results of several studies investigated are presented in this paper

Effect of triple ion beam irradiation on mechanical properties of high chromium austenitic stainless steel

International Nuclear Information System (INIS)

Ioka, Ikuo; Futakawa, Masatoshi; Nanjyo, Yoshiyasu; Kiuchi, Kiyoshi; Anegawa, Takefumi

2003-01-01

A high-chromium austenitic stainless steel has been developed for an advanced fuel cladding tube considering waterside corrosion and irradiation embrittlement. The candidate material was irradiated in triple ion (Ni, He, H) beam modes at 573 K up to 50 dpa to simulate irradiation damage by neutron and transmutation product. The change in hardness of the very shallow surface layer of the irradiated specimen was estimated from the slope of load/depth-depth curve which is in direct proportion to the apparent hardness of the specimen. Besides, the Swift's power low constitutive equation (σ=A(ε 0 + ε) n , A: strength coefficient, ε 0 : equivalent strain by cold rolling, n: strain hardening exponent) of the damaged parts was derived from the indentation test combined with an inverse analysis using a finite element method (FEM). For comparison, Type304 stainless steel was investigated as well. Though both Type304SS and candidate material were also hardened by ion irradiation, the increase in apparent hardness of the candidate material was smaller than that of Type304SS. The yield stress and uniform elongation were estimated from the calculated constitutive equation by FEM inverse analysis. The irradiation hardening of the candidate material by irradiation can be expected to be lower than that of Type304SS. (author)

Metal induced embrittlement. Annual report, [March 1, 1987--February 29, 1988

Energy Technology Data Exchange (ETDEWEB)

Hoagland, R.G.

1988-11-01

This program is investigating the causes of embrittlement that occur in certain solid metals when exposed to liquid metals. The degree of embrittlement varies enormously among different solid/liquid pairs as witness, for example, the modest loss of load carrying, ability induced in carbon steels by Pb or the profound embrittlment of aluminum (particularly high strength) alloys by Hg and Ga. The structure of this study involves two types of activities: an experimental fracture mechanics study of the behavior of certain solid metals in liquid metals, and a theoretical study on an atomic scale of the crack tip deformation and extension behavior by means of atomistic simulation. This research, which began March 1, 1987, has completed its 20 month. A brief synopsis is given of performance in each of the areas of activity during the past year.

Development of small punch tests for ductile-brittle transition temperature measurement of temper embrittled Ni-Cr steels

International Nuclear Information System (INIS)

Baik, J.M.; Kameda, J.; Buck, O.

1983-01-01

Small punch tests were developed to determine the ductile-brittle transition temperature of nickel-chromium (Ni-Cr) steels having various degrees of temper embrittlement and various microstructures. It was found that the small punch test clearly shows the ductile-brittle transition behavior of the temper-embrittled steels. The measured values were compared with those obtained from Charpy impact and uniaxial tensile tests. The effects of punch tip shape, a notch, and the strain rate on the ductile-brittle transition behavior were examined. It was found that the combined use of a notch, high strain rates, and a small punch tip strongly affects the ductile-brittle transition behavior. Considerable variations in the data were observed when the small punch tests were performed on coarse-grained steels. Several factors controlling embrittlement measurements of steels are discussed in terms of brittle fracture mechanisms

Status of pressure vessel embrittlement study in Japan

Energy Technology Data Exchange (ETDEWEB)

Kataoka, Shigeki [Japan Power Engineering and Inspection Corp. (JAPEIC), Chiba (Japan)

1997-09-01

The number of nuclear power plants in service for more than 20 years is increasing in Japan. Subsequently, the aging of nuclear power plants will continue to increase and for this reason, the assurance of the safety and reliability of nuclear power plants is becoming more important. Under this circumstances, Japan Government issued a report: ``Specific Concepts in Dealing with Nuclear Power Plant High Aging`` in April, 1996. This report identified that continuous technology development efforts are important to deal with the issues of nuclear power plant aging, and the following items are extracted for important categories to be developed. (1) Aging phenomena evaluation technology. (2) Inspection/monitoring technology (3) Preventive maintenance/repair technology. Japan Power Engineering and Inspection Corporation (JAPEIC) have been implementing various verification test concerning the above items consigned by the Ministry of International Trade and Industry (MITI). This report outlines the Specific Concepts in Dealing with Nuclear Power Plant High Agency and the past achievements and future plans of various verification tests related to irradiation embrittlement of nuclear reactor pressure vessel, mainly related to Pressurized Thermal Shock (PTS). (author). 4 refs, 8 figs, 5 tabs.

Effect of the 718 alloy metallurgical status on hydrogen embrittlement; Effet de l'etat metallurgique de l'alliage 718 sur la fragilisation par l'hydrogene

Energy Technology Data Exchange (ETDEWEB)

Galvano, F.; Andrieu, E.; Blanc, Ch.; Odemer, G.; Ter-Ovanessian, B.; Cocheteau, N.; Holstein, A.; Reboul, Ch. [Universite de Toulouse, CIRIMAT, UPS/CNRS/INPT, 31 - Toulouse (France); Clouez, J.M. [AREVA NP 69 - Lyon (France)

2010-03-15

The Inconel 718 is a nickel superalloy which is widely used in the nuclear industry, but is sensitive to hydrogen embrittlement induced by corrosion and stress corrosion cracking phenomena, and by the presence of dissolved hydrogen in pressurized water reactor environments. As this alloy is hardened by precipitation of different intermetallic phases, it appeared that the presence of these precipitates has a strong influence on the hydrogen embrittlement. The authors report the study of the nature and effect of the different traps (intermetallic phases, carbides or their interfaces) on the hydrogen embrittlement susceptibility of the 718 alloy, and more particularly on the observed failure modes. Experiments are performed on tensile samples in which hydrogen content can be measured. The type and grain size of the observed microstructures are given with respect with the thermal treatment, as well as the mechanical properties with or without hydrogen loading

Modeling of helium bubble nucleation and growth in neutron irradiated boron doped RAFM steels

International Nuclear Information System (INIS)

Dethloff, Christian; Gaganidze, Ermile; Svetukhin, Vyacheslav V.; Aktaa, Jarir

2012-01-01

Reduced activation ferritic/martensitic (RAFM) steels are promising candidates for structural materials in future fusion technology. In addition to other irradiation defects, the transmuted helium is believed to strongly influence material hardening and embrittlement behavior. A phenomenological model based on kinetic rate equations is developed to describe homogeneous nucleation and growth of helium bubbles in neutron irradiated RAFM steels. The model is adapted to different 10 B doped EUROFER97 based heats, which already had been studied in past irradiation experiments. Simulations yield bubble size distributions, whereby effects of helium generation rate, surface energy, helium sinks and helium density are investigated. Peak bubble diameters under different conditions are compared to preliminary microstructural results on irradiated specimens. Helium induced hardening was calculated by applying the Dispersed Barrier Hardening model to simulated cluster size distributions. Quantitative microstructural investigations of unirradiated and irradiated specimens will be used to support and verify the model.

Modeling of helium bubble nucleation and growth in neutron irradiated boron doped RAFM steels

Energy Technology Data Exchange (ETDEWEB)

Dethloff, Christian, E-mail: christian.dethloff@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Gaganidze, Ermile [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Svetukhin, Vyacheslav V. [Ulyanovsk State University, Leo Tolstoy Str. 42, 432970 Ulyanovsk (Russian Federation); Aktaa, Jarir [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

2012-07-15

Reduced activation ferritic/martensitic (RAFM) steels are promising candidates for structural materials in future fusion technology. In addition to other irradiation defects, the transmuted helium is believed to strongly influence material hardening and embrittlement behavior. A phenomenological model based on kinetic rate equations is developed to describe homogeneous nucleation and growth of helium bubbles in neutron irradiated RAFM steels. The model is adapted to different {sup 10}B doped EUROFER97 based heats, which already had been studied in past irradiation experiments. Simulations yield bubble size distributions, whereby effects of helium generation rate, surface energy, helium sinks and helium density are investigated. Peak bubble diameters under different conditions are compared to preliminary microstructural results on irradiated specimens. Helium induced hardening was calculated by applying the Dispersed Barrier Hardening model to simulated cluster size distributions. Quantitative microstructural investigations of unirradiated and irradiated specimens will be used to support and verify the model.

Microstructure and grain size effects on irradiation hardening of low carbon steel for reactor tanks

Energy Technology Data Exchange (ETDEWEB)

Milasin, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1964-05-15

Irradiation hardening of steel for reactor pressure vessels has been studied extensively during the past few years. A great number of experimental results concerning the behaviour of these steels in the radiation field and several review papers (1,2) have been published. Most of the papers deal with the effects of specific metallurgical factors or irradiation conditions (temperature, flux) on irradiation hardening and embrittlement. In addition, a number of experiments are performed to give evidence on the mechanism of irradiation hardening of these steels. However, this mechanism is still unknown due to the complexity of steel as a system. Among different methods used in radiation damage studies, the changes of mechanical properties have been mainly investigated. By using Hall-Petch's empirical relation, {sigma}{sub y}={sigma}{sub i}+k{sub y} d{sup -1/2} between lower yield stress, {sigma}{sub y}, and grain size, 2d, the information about the effect of irradiation on the parameters {sigma}{sub i} and k{sub y} is obtained. Taking as a base interpretation of {sigma}{sub i} and k{sub y} given by Petch and his co-workers it has been concluded that radiation does not change the stress to start slip but that it increase the friction that opposes the passage of free dislocations across a slip plane. In attempting to apply Hall-Petch's relation to one unirradiated ferritic steel with a carbon content higher than 0.15% some difficulties were encountered. The results obtained indicate that the influence of grain size can not be isolated from other factors introduced by the treatments used to produce different grain sizes. This paper deals with a similar problem in the case of irradiated steel. The results obtained give the changes of the mechanical properties of steel in neutron irradiation field as a function of microstructure and grain size. In addition, the mechanical properties of irradiated steel are measured after annealing at 150 deg C and 450 deg C. On the basis of

Heavy-Section Steel Irradiation Program

International Nuclear Information System (INIS)

Anon.

1991-01-01

In FY1990 the Heavy-Section Steel Irradiation (HSSI) Program was arranged into 8 tasks: (1) program management, (2) K Ic curve shift in high-copper welds, (3) K Ia curve shift in high-copper welds, (4) irradiation effects on cladding, (5) K Ic and K Ia curve shifts in low upper-shelf (LUS) welds, (6) irradiation effects in a commercial LUS weld, (7) microstructural analysis of irradiation effects, and (8) in-service aged material evaluations. Of particular interest are the efforts in FY1990 concerning the shifts in fracture toughness and crack arrest toughness in high-copper welds, the unirradiated examination of a LUS weld from the Midland reactor, and the continued investigation into the causes of accelerated low-temperature embrittlement recently observed in RPV support steels. In the Fifth and Sixth Irradiation Series, designed to examine the shifts and possible changes in shape in the ASME K Ic and K Ia curves for two irradiated high-copper welds, it was seen that both the lower bound and mean fracture toughness shifts were greater than those of the associated Charpy-impact energies, whereas the shifts in crack arrest toughness were comparable. The irradiation-shifted fracture toughness data fell slightly below the appropriately indexed ASME K Ic curve even when it was shifted according to Revision 2 of Regulatory Guide 1.99 including its margins. The beltline weld, which was removed from the Midland reactor, fabricated by Babcock and Wilcox, Co. using Linde 80 flux, is being examined in the Tenth Irradiation Series to establish the effects of irradiation on a commercial LUS weld. A wide variation in the unirradiated fracture properties of the Midland weld were measured with values of RT NDT ranging from -22 to 54F through its thickness. In addition, a wide range of copper content from 0.21 to 0.45 wt % was found, compared to the 0.42 wt % previously reported

Recrystallization and embrittlement of sintered tungsten

International Nuclear Information System (INIS)

Bega, N.D.; Babak, A.V.; Uskov, E.I.

1982-01-01

The recrystallization of sintered tungsten with a cellular structure of deformation is studied as related to its embrittlement. It is stated that in case of preliminary recrystallization the sintered tungsten crack resistance does not depend on the testing temperature. The tungsten crack resistance is shown to lower with an increase of the structure tendency to primary recrystallization [ru

A Physically Based Correlation of Irradiation-Induced Transition Temperature Shifts for RPV Steels

Energy Technology Data Exchange (ETDEWEB)

Eason, Ernest D. [Modeling and Computing Services, LLC; Odette, George Robert [UCSB; Nanstad, Randy K [ORNL; Yamamoto, Takuya [ORNL

2007-11-01

The reactor pressure vessels (RPVs) of commercial nuclear power plants are subject to embrittlement due to exposure to high-energy neutrons from the core, which causes changes in material toughness properties that increase with radiation exposure and are affected by many variables. Irradiation embrittlement of RPV beltline materials is currently evaluated using Regulatory Guide 1.99 Revision 2 (RG1.99/2), which presents methods for estimating the shift in Charpy transition temperature at 30 ft-lb (TTS) and the drop in Charpy upper shelf energy (ΔUSE). The purpose of the work reported here is to improve on the TTS correlation model in RG1.99/2 using the broader database now available and current understanding of embrittlement mechanisms. The USE database and models have not been updated since the publication of NUREG/CR-6551 and, therefore, are not discussed in this report. The revised embrittlement shift model is calibrated and validated on a substantially larger, better-balanced database compared to prior models, including over five times the amount of data used to develop RG1.99/2. It also contains about 27% more data than the most recent update to the surveillance shift database, in 2000. The key areas expanded in the current database relative to the database available in 2000 are low-flux, low-copper, and long-time, high-fluence exposures, all areas that were previously relatively sparse. All old and new surveillance data were reviewed for completeness, duplicates, and discrepancies in cooperation with the American Society for Testing and Materials (ASTM) Subcommittee E10.02 on Radiation Effects in Structural Materials. In the present modeling effort, a 10% random sample of data was reserved from the fitting process, and most aspects of the model were validated with that sample as well as other data not used in calibration. The model is a hybrid, incorporating both physically motivated features and empirical calibration to the U.S. power reactor surveillance

Embrittlement phenomenon of Ag core MP35N cable as lead conductor in medical device.

Science.gov (United States)

Wang, Ling; Li, Bernie; Zhang, Haitao

2013-02-01

Ag core MP35N (Ag/MP35N) wire has been used in lead electric conductor wires in the medical device industry for many years. Recently it was noticed that the combination of silver and MP35N restricts its wire drawing process. The annealing temperature in Ag/MP35N has to be lower than the melting temperature of pure Ag (960 °C), which cannot fully anneal MP35N. The lower annealing temperature results in a highly cold worked MP35N, which significantly reduces Ag/MP35N ductility. The embrittlement phenomenon of Ag/MP35N cable was observed in tension and bending deformation. The effect of the embrittlement on the wire flex fatigue life was evaluated using a newly developed flex fatigue testing method. The Ag/MP35N cable fatigue results was analyzed with a Coffin-Manson approach and compared to the MP35N cable fatigue results. The root causes of the Ag/Mp35N embrittlement phenomenon are discussed. Copyright © 2012 Elsevier Ltd. All rights reserved.

Evaluation of temper embrittlement of martensitic and ferritic-martensitic steels by acoustic emission

International Nuclear Information System (INIS)

Lu, Yusho; Takahashi, Hideaki; Shoji, Tetsuo

1987-01-01

Martensitic (HT-9) and ferritic-martensitic steels (9Cr-2Mo) are considered as fusion first wall materials. In this investigation in order to understand the sensitivity of temper embrittlement in these steels under actual service condition, fracture toughness testing was made by use of acoustic emission technique. The temper embrittlement was characterized in terms of fracture toughness. The fracture toughness of these steels under 500 deg C, 100 hrs, and 1000 hrs heat treatment was decreased and their changes in micro-fracture process have been observed. The fracture toughness changes by temper embrittlement was discussed by the characteristic of AE, AE spectrum analysis and fractographic investigation. The relation between micro-fracture processes and AE has been clarified. (author)

Behavior of ferritic/martensitic steels after n-irradiation at 200 and 300 deg. C

International Nuclear Information System (INIS)

Matijasevic, M.; Lucon, E.; Almazouzi, A.

2008-01-01

High chromium ferritic/martensitic (F/M) steels are considered as the most promising structural materials for accelerator driven systems (ADS). One drawback that needs to be quantified is the significant hardening and embrittlement caused by neutron irradiation at low temperatures with production of spallation elements. In this paper irradiation effects on the mechanical properties of F/M steels have been studied and comparisons are provided between two ferritic/martensitic steels, namely T91 and EUROFER97. Both materials have been irradiated in the BR2 reactor of SCK-CEN/Mol at 300 deg. C up to doses ranging from 0.06 to 1.5 dpa. Tensile tests results obtained between -160 deg. C and 300 deg. C clearly show irradiation hardening (increase of yield and ultimate tensile strengths), as well as reduction of uniform and total elongation. Irradiation effects for EUROFER97 starting from 0.6 dpa are more pronounced compared to T91, showing a significant decrease in work hardening. The results are compared to our latest data that were obtained within a previous program (SPIRE), where T91 had also been irradiated in BR2 at 200 deg. C (up to 2.6 dpa), and tested between -170 deg. C and 300 deg. C. Irradiation effects at lower irradiation temperatures are more significant

Behavior of ferritic/martensitic steels after n-irradiation at 200 and 300 °C

Science.gov (United States)

Matijasevic, M.; Lucon, E.; Almazouzi, A.

2008-06-01

High chromium ferritic/martensitic (F/M) steels are considered as the most promising structural materials for accelerator driven systems (ADS). One drawback that needs to be quantified is the significant hardening and embrittlement caused by neutron irradiation at low temperatures with production of spallation elements. In this paper irradiation effects on the mechanical properties of F/M steels have been studied and comparisons are provided between two ferritic/martensitic steels, namely T91 and EUROFER97. Both materials have been irradiated in the BR2 reactor of SCK-CEN/Mol at 300 °C up to doses ranging from 0.06 to 1.5 dpa. Tensile tests results obtained between -160 °C and 300 °C clearly show irradiation hardening (increase of yield and ultimate tensile strengths), as well as reduction of uniform and total elongation. Irradiation effects for EUROFER97 starting from 0.6 dpa are more pronounced compared to T91, showing a significant decrease in work hardening. The results are compared to our latest data that were obtained within a previous program (SPIRE), where T91 had also been irradiated in BR2 at 200 °C (up to 2.6 dpa), and tested between -170 °C and 300 °C. Irradiation effects at lower irradiation temperatures are more significant.

Multiscale Modeling of Hydrogen Embrittlement for Multiphase Material

KAUST Repository

Al-Jabr, Khalid A.

2014-01-01

Hydrogen Embrittlement (HE) is a very common failure mechanism induced crack propagation in materials that are utilized in oil and gas industry structural components and equipment. Considering the prediction of HE behavior, which is suggested

Hydrogen embrittlement of the 22 Cr5 Ni austeno-ferritic stainless steel. Role of the microstructure

International Nuclear Information System (INIS)

Iacoviello, Francesco

1997-01-01

Austenitic-ferritic stainless steels are characterised by very good mechanical properties and by a high corrosion resistance, especially to stress-corrosion and to pitting. However, their duplex structure shows a sensitivity to hydrogen embrittlement. Among duplex stainless steels, the 22 Cr 5 Ni grade has gradually became the most used. In this work the tensile properties and the resistance to fatigue crack propagation of 22 Cr5 Ni duplex stainless steel have been analysed, with and without hydrogen charging, after it had been treated at temperatures ranging between 200-1050 deg. C for varying times. The heat treatment times and temperatures were chosen to characterise completely the effects of the different intermetallic and the carbide and nitride phases and to compare these results with those from the tensile tests and those in the literature. A technique for obtaining the hydrogen diffusion coefficient in the steel was optimised and was shown to be alternative to the permeation technique. Thermal analysis was used to determine the activation energy of the hydrogen traps in the steel. From the results the following conclusions were established: - Grain boundaries and dislocations have very little influence on the process of hydrogen diffusion. - The quantity of hydrogen absorbed depends in that any type of precipitate decrease the absorption. This decrease was probably due to changes in the diffusivity and solubility of hydrogen caused by the precipitation. - The charging with hydrogen caused a large decrease in ε m pc for the steel for all heat treatments temperature, except 1050 deg. C. At this temperature the effect was much less as the dislocation density was very low and the precipitates were now in solution. - Hydrogen charging of the steel did not affect the YS and the decrease in UTS produced depended on the microstructure. Use of the embrittlement index 'F' showed that spinodal decomposition and precipitation of G phase decrease hydrogen embrittlement

Hydrogen embrittlement of titanium tested with fracture mechanics specimens

International Nuclear Information System (INIS)

Aho-Mantila, I.; Rahko, P.

1990-11-01

Titanium is one of the possible canister materials for spent nuclear fuel. The aim of this study is to determine whether the hydrogen embrittlement of titanium could be a possible deterioration mechanism of titanium canisters. This experimental study was preceded by a literature review and an experimental study on crack nucleation. Tests in this study were carried out with hydrogen charged fracture mechanics specimens. The studied hydrogen contents were as received, 100 ppm, 200 ppm, 500 ppm and 700 ppm and the types of the studied titanium were ASTM Grades 2 and 12. Test methods were slow tensile test (0.027 mm/h) and fatigue test (stress ratio 0.7 or 0.8 and frequency 5 Hz). According to the literature titanium may be embrittled by hydrogen at slow strain rates and cracking may occur under sustained load. In this study no evidence of hydrogen embrittlement was noticed in slow strain rate tension with bulk hydrogen contents up to 700 ppm. The fatigue tests of titanium Grades 2 and 12 containing 700 ppm hydrogen showed even slower crack growth compared to the as received condition. Very high hydrogen contents well in eccess of 700 ppm on the surface of titanium can, however, facilitate surface crack nucleation and crack growth, as shown in the previous study

Consequence evaluation of radiation embrittlement of Trojan reactor pressure vessel supports

International Nuclear Information System (INIS)

Lu, S.C.; Sommer, S.C.; Johnson, G.L.; Lambert, H.E.

1990-10-01

This report describes a consequence evaluation to address safety concerns raised by the radiation embrittlement of the reactor pressure vessel (RPV) supports for the Trojan nuclear power plant. The study comprises a structural evaluation and an effects evaluation and assumes that all four reactor vessel supports have completely lost the load carrying capability. By demonstrating that the ASME code requirements governing Level D service limits are satisfied, the structural evaluation concludes that the Trojan reactor coolant loop (RCL) piping is capable of transferring loads to the steam generator (SG) supports and the reactor coolant pump (RCP) supports. A subsequent design margins to accommodate additional loads transferred to them through the RCL piping. The effects evaluation, employing a systems analysis approach, investigates initiating events and the reliability of the engineered safeguard systems as the RPV is subject to movements caused by the RPV support failure. The evaluation identifies a number of areas of additional safety concerns, but further investigation of the above safety concerns, however, concludes that a hypothetical failure of the Trojan RPV supports due to radiation embrittlement will not result in consequences of significant safety concerns

Radiation annealing mechanisms of low-alloy reactor pressure vessel steels dependent on irradiation temperature and neutron fluence

International Nuclear Information System (INIS)

Pachur, D.

1982-01-01

Heat treatment after irradiation of reactor pressure vessel steels showed annealing of irradiation embrittlement. Depending on the irradiation temperature, the embrittlement started to anneal at about 220 0 C and was completely annealed at 500 0 C with 4 h of annealing time. The annealing behavior was normally measured in terms of the Vickers hardness increase produced by irradiation relative to the initial hardness as a function of the annealing temperature. Annealing results of other mechanical properties correspond to hardness results. During annealing, various recovery mechanisms occur in different temperature ranges. These are characterized by activation energies from 1.5 to 2.1 eV. The individual mechanisms were determined by the different time dependencies at various temperatures. The relative contributions of the mechanisms showed a neutron fluence dependence, with the lower activation energy mechanisms being predominant at low fluence and vice versa. In the temperature range where partial annealing of a mechanism took place during irradiation, an increase in activation energy was observed. Trend curves for the increase in transition temperature with irradiation, for the relative increase of Vickers hardness and yield strength, and for the relative decrease of Charpy-V upper shelf energy are interpreted by the behavior of different mechanisms

Stabilization by hals and phenols in γ-irradiated polyproplyene

International Nuclear Information System (INIS)

Carlsson, D.J.; Falicki, S.; Cooke, J.M.; Gosciniak, D.J.

1994-01-01

The γ-radiation initiated oxidation of polypropylene films and test strips has been studied both immediately after irradiation and also during post-irradiation accelerated aging at 60 degrees C. Stabilizers included blocked and unblocked phenols as well as secondary and tertiary hindered amines (HALS) including an oligomeric HALS. Oxidation product formation, yellowing and embrittlement (as measured in an instrumented bend test) have been compared with product formation. A partial correlation between suppression of oxidation during the irradiation step with long term, post-irradiation oven aging at 60 degrees C was found, but complicated by extensive destruction during irradiation of the active phenolic functionality in some additives, essential for peroxyl radical scavenging. Very long lifetimes with barely detectable yellowing were found for combinations of the amines with completely unhindered or only partially hindered phenols

Charles J. McMahon Interfacial Segregation and Embrittlement Symposium

National Research Council Canada - National Science Library

Vitek, Vaclav

2003-01-01

.... McMahon Interfacial Segregation and Embrittlement Symposium: Grain Boundary Segregation and Fracture in Steels was sponsored by ASM International, Materials Science Critical Technology Sector, Structural Materials Division, Materials Processing...

Dissolution of alpha-prime precipitates in thermally embrittled S2205-duplex steels during reversion-heat treatment

Directory of Open Access Journals (Sweden)

V. Shamanth

2015-01-01

Full Text Available Duplex stainless steels offer an attractive combination of strength, corrosion resistance and cost. In annealed condition duplex steels will be in thermodynamically metastable condition but when they are subjected to intermediate homologous temperature of ∼475 °C and below significant embrittlement occurs, which is one of the key material degradation properties that limits its upper service temperature in many applications. Hence the present study is aimed to study the effect of reversion heat treatment and its time on mechanical properties of the thermally embrittled steel. The results showed that 60 min reversion heat treated samples were able to recover the mechanical properties which were very close to annealed properties because when the embrittled samples were reversion heat treated at an elevated temperature of 550 °C which is above the (α + α′ miscibility gap, the ferritic phase was homogenized again. In other words, Fe-rich α and Cr-rich α′ prime precipitates which were formed during ageing become thermodynamically unstable and dissolve inside the ferritic phase.

On the tempered martensite embrittlement in AISI 4140 low alloy steel

Energy Technology Data Exchange (ETDEWEB)

Darwish, F.A. (Dept. of Materials Science and Metallurgy, Catholic Univ., Rio de Janeiro, RJ (Brazil)); Pereira, L.C.; Gatts, C. (Dept. of Metallurgy and Materials Engineering, Federal Univ., Rio de Janeiro, RJ (Brazil)); Graca, M.L. (Materials Div., Technical Aerospace Center, Sao Jose dos Campos, SP (Brazil))

1991-02-01

In the present investigation the Auger electron spectroscopy (AES) technique was used to determine local carbon and phosphorus concentrations on the fracture surfaces of as-quenched and quenched-and-tempered (at 350deg C) AISI 4140 steel specimens austenitized at low and high temperatures. The AES results were rationalized to conclude that, although carbide growth as well as phosphorus segregation are expected to contribute to tempered martensite embrittlement, carbide precipitation on prior austenite grain boundaries during tempering is seen to be the microstructural change directly responsible for the occurrence of the referred embrittlement phenomenon. (orig.).

Atmospheric-Induced Stress Corrosion Cracking of Grade 2205 Duplex Stainless Steel—Effects of 475 °C Embrittlement and Process Orientation

Directory of Open Access Journals (Sweden)

Cem Örnek

2016-07-01

Full Text Available The effect of 475 °C embrittlement and microstructure process orientation on atmospheric-induced stress corrosion cracking (AISCC of grade 2205 duplex stainless steel has been investigated. AISCC tests were carried out under salt-laden, chloride-containing deposits, on U-bend samples manufactured in rolling (RD and transverse directions (TD. The occurrence of selective corrosion and stress corrosion cracking was observed, with samples in TD displaying higher propensity towards AISCC. Strains and tensile stresses were observed in both ferrite and austenite, with similar magnitudes in TD, whereas, larger strains and stresses in austenite in RD. The occurrence of 475 °C embrittlement was related to microstructural changes in the ferrite. Exposure to 475 °C heat treatment for 5 to 10 h resulted in better AISCC resistance, with spinodal decomposition believed to enhance the corrosion properties of the ferrite. The austenite was more susceptible to ageing treatments up to 50 h, with the ferrite becoming more susceptible with ageing in excess of 50 h. Increased susceptibility of the ferrite may be related to the formation of additional precipitates, such as R-phase. The implications of heat treatment at 475 °C and the effect of process orientation are discussed in light of microstructure development and propensity to AISCC.

Low temperature thermal ageing embrittlement of austenitic stainless steel welds and its electrochemical assessment

International Nuclear Information System (INIS)

Chandra, K.; Kain, Vivekanand; Raja, V.S.; Tewari, R.; Dey, G.K.

2012-01-01

Highlights: ► Embrittlement study of austenitic stainless steel welds after ageing up to 20,000 h. ► Spinodal decomposition and G-phase precipitation in ferrite at 400 °C. ► Spinodal decomposition of ferrite at 335 and 365 °C. ► Large decrease in corrosion resistance due to G-phase precipitation. ► Good correlation between electrochemical properties and the degree of embrittlement. - Abstract: The low temperature thermal ageing embrittlement of austenitic stainless steel welds is investigated after ageing up to 20,000 h at 335, 365 and 400 °C. Spinodal decomposition and G-phase precipitation after thermal ageing were identified by transmission electron microscopy. Ageing led to increase in hardness of the ferrite phase while there was no change in the hardness of austenite. The degree of embrittlement was evaluated by non-destructive methods, e.g., double-loop and single-loop electrochemical potentiokinetic reactivation tests. A good correlation was obtained between the electrochemical properties and hardening of the ferrite phase of the aged materials.

The metrological problems of irradiation embrittlement of reactor pressure vessel steel

International Nuclear Information System (INIS)

Vodenicharov, S.; Kamenova, Ts.

1993-01-01

Neutron irradiation of reactor pressure vessel steels increases the T k -values of transition temperature from ductile to brittle fracture. This effect is very important in emergency situations, when the water cooling injection in the reactor results in high thermal gradients. In such cases there is a risk from the appearance of a brittle fracture with catastrophic crack propagation speed at relatively low stresses. That is why the T k -value determination is very important for the safe operation of the reactor systems. Some advanced experimental methods for T k -testing and control have been discussed in the present article and the standards of different countries have been compared. The methods applying subsize specimens and welding-restored specimens have been reviewed. (author)

Comparison of hydrogen gas embrittlement of austenitic and ferritic stainless steels

Science.gov (United States)

Perng, T. P.; Altstetter, C. J.

1987-01-01

Hydrogen-induced slow crack growth (SCG) was compared in austenitic and ferritic stainless steels at 0 to 125 °Cand 11 to 216 kPa of hydrogen gas. No SCG was observed for AISI 310, while AISI 301 was more susceptible to hydrogen embrittlement and had higher cracking velocity than AL 29-4-2 under the same test conditions. The kinetics of crack propagation was modeled in terms of the hydrogen transport in these alloys. This is a function of temperature, microstructure, and stress state in the embrittlement region. The relatively high cracking velocity of AISI 301 was shown to be controlled by the fast transport of hydrogen through the stress-induced α' martensite at the crack tip and low escape rate of hydrogen through the γ phase in the surrounding region. Faster accumulation rates of hydrogen in the embrittlement region were expected for AISI 301, which led to higher cracking velocities. The mechanism of hydrogen-induced SCG was discussed based upon the concept of hydrogen-enhanced plasticity.

Probabilistic approaches applied to damage and embrittlement of structural materials in nuclear power plants

International Nuclear Information System (INIS)

Vincent, L.

2012-01-01

The present study deals with the long-term mechanical behaviour and damage of structural materials in nuclear power plants. An experimental way is first followed to study the thermal fatigue of austenitic stainless steels with a focus on the effects of mean stress and bi-axiality. Furthermore, the measurement of displacement fields by Digital Image Correlation techniques has been successfully used to detect early crack initiation during high cycle fatigue tests. A probabilistic model based on the shielding zones surrounding existing cracks is proposed to describe the development of crack networks. A more numeric way is then followed to study the embrittlement consequences of the irradiation hardening of the bainitic steel constitutive of nuclear pressure vessels. A crystalline plasticity law, developed in agreement with lower scale results (Dislocation Dynamics), is introduced in a Finite Element code in order to run simulations on aggregates and obtain the distributions of the maximum principal stress inside a Representative Volume Element. These distributions are then used to improve the classical Local Approach to Fracture which estimates the probability for a microstructural defect to be loaded up to a critical level. (author) [fr

Experimental study on the resistance to hydrogen embrittlement of NIFS-V4Cr4Ti alloy

International Nuclear Information System (INIS)

Chen Jiming; Xu Zengyu; Den Ying; Muroga, T.

2002-01-01

SWIP (Southwestern Institute of Physics) has joined an international collaboration on the hydrogen embrittlement resistance evaluation of the vanadium alloy. This paper presents some experiments on the tensile properties and Charpy impact properties of the NIFS-V4Cr4Ti alloy with high-level hydrogen concentration. The experiment results show different properties against hydrogen embrittlement in static tension and impact load. The critical hydrogen concentration required to embrittle the alloy was about 215 - 310 mg·kg -1 on static tension load, but less than 130 mg·kg -1 on impact loading

Effect of chemical composition on irradiation embrittlement and annealing in Ni-Cr-Mo-V reactor pressure vessel steel

Energy Technology Data Exchange (ETDEWEB)

Novosad, P [Czech Nuclear Society, Prague (Czech Republic)

1994-12-31

Results concerning copper and phosphorus influence on radiation-induced changes in the Ni-Cr-Mo-V steel mechanical properties, are presented. Correlations between different mechanical properties for steels with different chemical composition, are presented. A comparison of transition temperature shifts obtained for static and dynamic fracture toughness tests and Charpy impact tests, is discussed. Recovery of radiation hardening, measured by hardness test after isochronal annealing of steels with different compositions, is also shown. Copper content strongly affects irradiation-induced changes of mechanical properties, but phosphorus content in connection with variable copper content has only a small effect. (author). 4 refs., 4 figs., 4 tabs.

'In-beam' simulation of high temperature helium embrittlement of DIN 1.4970 austenitic stainless steel

International Nuclear Information System (INIS)

Schroeder, H.; Batfalsky, P.

1982-01-01

This work describes a facility for high temperature creep rupture tests during homogeneous helium implantation. This 'in-beam' creep testing facility is used to simulate helium embrittlement effects which will be very important for first wall materials of future fusion reactors operated at high temperatures. First results for DIN 1.4970 austenitic stainless steel clearly demonstrate differences between samples 'in-beam' tested at 1073 K and those creep tested at the same temperature after room temperature helium implantation. The specimens ruptured 'in-beam' have much shorter lifetimes and lower ductility than the specimens tested after room temperature implantation. There are also differences in the microstructures, concerning helium bubble sizes and densities in matrix and grain boundaries. These microstructural differences may be a key for the understanding of the more severe helium embrittlement effects 'in-beam' as compared to creep tests performed after room temperature implantation. (orig.)

High Temperature Tensile Properties of Unirradiated and Neutron Irradiated 20 Cr-35 Ni Austenitic Steel

Energy Technology Data Exchange (ETDEWEB)

Roy, R B; Solly, B

1966-12-15

The tensile properties of an unirradiated and neutron irradiated (at 40 deg C) 20 % Cr, 35 % Ni austenitic steel have been studied at 650 deg C, 750 deg C and 820 deg C. The tensile elongation and mode of fracture (transgranular) of unirradiated specimens tested at room temperature and 650 deg C are almost identical. At 750 deg C and 820 deg C the elongation decreases considerably and a large part of the total elongation is non-uniform. Furthermore, the mode of fracture at these temperatures is intergranular and microscopic evidence suggests that fracture is caused by formation and linkup of grain boundary cavities. YS and UTS decrease monotonically with temperature. Irradiated specimens show a further decrease in ductility and an increase in the tendency to grain boundary cracking. Irradiation has no significant effect on the YS, but the UTS are reduced. The embrittlement of the irradiated specimens is attributed to the presence of He and Li atoms produced during irradiation and the possible mechanisms are discussed. Prolonged annealing of irradiated and unirradiated specimens at 650 deg C appears to have no significant effect on tensile properties.

Microstructural evolution in neutron irradiated reactor pressure vessel steels

International Nuclear Information System (INIS)

English, C.A.; Phythian, W.J.

1998-01-01

As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. The microstructural evolution in neutron irradiated reactor pressure vessel steels is described. The damage mechanisms are elaborated and techniques for examining the microstructure are suggested. The importance of the initial damage event is analysed, and the microstructural evolution in RPV steels is examined

Electron-microscopic investigation of a pressure vessel steel after neutron irradiation

International Nuclear Information System (INIS)

Klaar, H.J.

1975-01-01

As an introduction, changes in the mechanical properties of pressure vessel steels on neutron irradiation and the causes of radiation embrittlement are discussed. After this, the author describes his own experiments with steel of the composition 0.19% C; 3.88% Ni; 1.57% Cr; 0.51% Mo; 0.2% V. Samples of this material were irradiated in-pile at 300 0 C with various neutron doses. To study the influence of neutron dose, irradiation temperature, and heat treatment on the mechanical properties, tensile tests, notched bar impact bending tests, hardness tests and structural analyses were carried out. The findings are reported. (GSC) [de

Application of advanced irradiation analysis methods to light water reactor pressure vessel test and surveillance programs

International Nuclear Information System (INIS)

Odette, R.; Dudey, N.; McElroy, W.; Wullaert, R.; Fabry, A.

1977-01-01

Inaccurate characterization and inappropriate application of neutron irradiation exposure variables contribute a substantial amount of uncertainty to embrittlement analysis of light water reactor pressure vessels. Damage analysis involves characterization of the irradiation environment (dosimetry), correlation of test and surveillance metallurgical and dosimetry data, and projection of such data to service conditions. Errors in available test and surveillance dosimetry data are estimated to contribute a factor of approximately 2 to the data scatter. Non-physical (empirical) correlation procedures and the need to extrapolate to the vessel may add further error. Substantial reductions in these uncertainties in future programs can be obtained from a more complete application of available damage analysis tools which have been developed for the fast reactor program. An approach to reducing embrittlement analysis errors is described, and specific examples of potential applications are given. The approach is based on damage analysis techniques validated and calibrated in benchmark environments

Fracture toughness of irradiated candidate materials for ITER first wall/blanket structures

International Nuclear Information System (INIS)

Alexander, D.J.; Pawel, J.E.; Grossbeck, M.L.; Rowcliffe, A.F.; Shiba, Kiyoyuki

1994-01-01

Disk compact specimens of candidate materials for first wall/blanket structures in ITER have been irradiated to damage levels of about 3 dpa at nominal irradiation temperatures of either 90 or 250 degrees C. These specimens have been tested over a temperature range from 20 to 250 degrees C to determine J-integral values and tearing moduli. The results show that irradiation at these temperatures reduces the fracture toughness of austenitic stainless steels, but the toughness remains quite high. The toughness decreases as the test temperature increases. Irradiation at 250 degrees C is more damaging than at 90 degrees C, causing larger decreases in the fracture toughness. Ferritic-martensitic steels are embrittled by the irradiation, and show the lowest toughness at room temperature

Effect of Low-Temperature Sensitization on Hydrogen Embrittlement of 301 Stainless Steel

Directory of Open Access Journals (Sweden)

Chieh Yu

2017-02-01

Full Text Available The effect of metastable austenite on the hydrogen embrittlement (HE of cold-rolled (30% reduction in thickness 301 stainless steel (SS was investigated. Cold-rolled (CR specimens were hydrogen-charged in an autoclave at 300 or 450 °C under a pressure of 10 MPa for 160 h before tensile tests. Both ordinary and notched tensile tests were performed in air to measure the tensile properties of the non-charged and charged specimens. The results indicated that cold rolling caused the transformation of austenite into α′ and ε-martensite in the 301 SS. Aging at 450 °C enhanced the precipitation of M23C6 carbides, G, and σ phases in the cold-rolled specimen. In addition, the formation of α′ martensite and M23C6 carbides along the grain boundaries increased the HE susceptibility and low-temperature sensitization of the 450 °C-aged 301 SS. In contrast, the grain boundary α′-martensite and M23C6 carbides were not observed in the as-rolled and 300 °C-aged specimens.

Effect of heat treatment and irradiation temperature on mechanical properties and structure of reduced-activation Cr-W-V steels of bainitic, martensitic, and martensitic-ferritic classes

International Nuclear Information System (INIS)

Gorynin, I.V.; Rybin, V.V.; Kursevich, I.P.; Lapin, A.N.; Nesterova, E.V.; Klepikov, E.Yu.

2000-01-01

Effects of molybdenum replacement by tungsten in steels of the bainitic, martensitic, and martensitic-ferritic classes containing 2.5%, 8% and 11% Cr, respectively, were investigated. The phase composition and structure of the bainitic steels were varied by changing the cooling rates from the austenitization temperature (from values typical for normalization up to V=3.3 x 10 -2 deg. C/s) and then tempering. The steels were irradiated to a fluence of 4x10 23 n/m 2 (≥0.5 MeV) at 270 deg. C and to fluences of 1.3x10 23 and 1.2x10 24 n/m 2 (≥0.5 MeV) at 70 deg. C. The 2.5Cr-1.4WV and 8Cr-1.5WV steels have shown lower values of the shifts in ductile-brittle transition temperature (DBTT) under irradiation in comparison with corresponding Cr-Mo steels. Radiation embrittlement at elevated irradiation temperature was lowest in bainitic 2.5Cr-1.4WV steel and martensitic-ferritic 11Cr-1.5WV steel. The positive effect of molybdenum replacement by tungsten at irradiation temperature ∼300 deg. C is reversed at T irr =70 deg. C

Zinc-induced embrittlement in nickel-base superalloys by simulation and experiment

Science.gov (United States)

Otis, Richard; Waje, Mahesh; Lindwall, Greta; Jefferson, Tiffany; Lange, Jeremy; Liu, Zi-Kui

2017-09-01

The high cost of Re has driven interest in processes for recovering Re from scrap superalloy parts. In this work thermodynamic modelling is used to study Zn-induced embrittlement of a superalloy and to direct experiments. Treating superalloy powder with Zn vapour reduces the average particle size after milling from approximately ?m to 0.5-10 ?m, vs. ?m for untreated powder. Simulations predict the required treatment time to increase with temperature. Agreement between predictions and experiments suggests that an embrittling liquid forms in less than an hour of Zn vapour treatment between 950-1000 ?C and partial pressures of Zn between 14-34 kPa (2-5 psi).

Evaluation of the french test reactors irradiation embrittlement experiments

International Nuclear Information System (INIS)

Miannay, D.; Dussarte, D.; Soulat, P.

1988-07-01

The shifts of CV 41J energy index temperatures due to irradiation measured in France have been stored in a data bank and are analysed. According to a simple physically based model which is here-after verified, correlations are proposed for Base Metal (BM) and Weld Metal (WM). The achemical and phosphorus components of the chemical factor are equivalent. However, nickel and copper play a leading part in BM and WM respectively. The copper nickel interaction is not evident. These correlations are for cleavage fracture and not for intergranular fracture. This work is subject to revision and extension

Microstructural stability of spark-plasma-sintered Wf/W composite with zirconia interface coating under high-heat-flux hydrogen beam irradiation

OpenAIRE

M. Avello de Lama; M. Balden; H. Greuner; T. Höschen; J. Matejicek; J.H. You

2017-01-01

Tungsten is considered as the most suitable material for the plasma-facing armour of future fusion reactors. However, in spite of many advantageous properties, pure tungsten has a major drawback, namely, brittleness at lower temperatures and embrittlement by neutron irradiation. Tungsten fibre-reinforced tungsten (Wf/W) composites are thought to be a promising candidate material for armour owing to the pseudo-toughness effect which is based on controlled cracking of coated interfaces. In this...

Vanadium alloy membranes for high hydrogen permeability and suppressed hydrogen embrittlement

International Nuclear Information System (INIS)

Kim, Kwang Hee; Park, Hyeon Cheol; Lee, Jaeho; Cho, Eunseog; Lee, Sang Mock

2013-01-01

The structural properties and hydrogen permeation characteristics of ternary vanadium–iron–aluminum (V–Fe–Al) alloy were investigated. To achieve not only high hydrogen permeability but also strong resistance to hydrogen embrittlement, the alloy composition was modulated to show high hydrogen diffusivity but reduced hydrogen solubility. We demonstrated that matching the lattice constant to the value of pure V by co-alloying lattice-contracting and lattice-expanding elements was quite effective in maintaining high hydrogen diffusivity of pure V

Influence of LBE long term exposure and simultaneous fast neutron irradiation on the mechanical properties of T91 and 316L

Energy Technology Data Exchange (ETDEWEB)

Stergar, E., E-mail: estergar@sckcen.be [SCK-CEN, Belgian Nuclear Research Centre, Boeretang 200, 2400 Mol (Belgium); Eremin, S.G. [RIAR, Research Institute of Atomic Reactors, Dimitrovgrad (Russian Federation); Gavrilov, S.; Lambrecht, M. [SCK-CEN, Belgian Nuclear Research Centre, Boeretang 200, 2400 Mol (Belgium); Makarov, O.; Iakovlev, V. [RIAR, Research Institute of Atomic Reactors, Dimitrovgrad (Russian Federation)

2016-05-15

The LEXUR–II–LBE irradiation campaign was conducted from 2011 to 2012 and was aimed to investigate the combined influence of irradiation and LBE environment. In this irradiation campaign tensile test samples, pressurized tubes and corrosion samples were irradiated in LBE filled capsules. To separate the effect of exposure to LBE and neutron irradiation a parallel furnace experiment where the samples were exposed to LBE at the irradiation temperature for the corresponding time was conducted. Here we report results of the first extracted capsule which was irradiated about 6 months and dismantled after a cooling phase to decrease activity. The results of SSRT tests for irradiated T91 show that the exposure to LBE at 350 °C for a long time leads to the appearance of liquid metal embrittlement without any pre-treatment which is usually necessary to promote LME. Irradiation increases the effect of LME on the ductility of T91. In contrast to the findings for T91 the gained results also show that tensile tests on irradiated austenitic stainless steel 316L show no influence of LBE environment on the tensile properties.

Meso-scale magnetic signatures for nuclear reactor steel irradiation embrittlement monitoring

Energy Technology Data Exchange (ETDEWEB)

Suter, J. D., E-mail: pradeep.ramuhalli@pnnl.gov; Ramuhalli, P., E-mail: pradeep.ramuhalli@pnnl.gov; Hu, S.; Li, Y.; Jiang, W.; Edwards, D. J.; Schemer-Kohrn, A. L.; Johnson, B. R. [Pacific Northwest National Laboratory, 902 Battelle Blvd, Richland, WA 99352 (United States); McCloy, J. S., E-mail: john.mccloy@wsu.edu; Xu, K., E-mail: john.mccloy@wsu.edu [Washington State University, PO Box 642920, Pullman, WA 99164 (United States)

2015-03-31

Verifying the structural integrity of passive components in light water and advanced reactors will be necessary to ensure safe, long-term operations of the existing U.S. nuclear fleet. This objective can be achieved through nondestructive condition monitoring techniques, which can be integrated with plant operations to quantify the “state of health” of structural materials in real-time. While nondestructive methods for monitoring many classes of degradation (such as fatigue or stress corrosion cracking) are relatively advanced, this is not the case for degradation caused by irradiation. The development of nondestructive evaluation technologies for these types of degradation will require advanced materials characterization techniques and tools that enable comprehensive understanding of nuclear reactor material microstructural and behavioral changes under extreme operating environments. Irradiation-induced degradation of reactor steels causes changes in their microstructure that impacts their micro-magnetic properties. In this paper, we describe preliminary results of integrating advanced material characterization techniques with meso-scale computational models. In the future, this will help to provide an interpretive understanding of the state of degradation in structural materials. Microstructural data are presented from monocrystalline Fe and are correlated with variable-field magnetic force microscopy and micro-magnetic measurements. Ongoing research is focused on extending the measurements and models on thin films to gain insights into the structural state of irradiated materials and the resulting impact on magnetic properties. Preliminary conclusions from these correlations are presented, and next steps described.

Influence of neutron irradiation at 550C on the properties of austenitic stainless steels

International Nuclear Information System (INIS)

Wiffen, F.W.; Maziasz, P.J.

1981-01-01

Types 316 and 316 + 0.23 wt % Ti stainless steels and 16-8-2 weldment were irradiated in HFIR at 55 0 C to fluences up to 1.35 x 10 26 neutrons/m 2 ( 0 C strength properties, with the weldments the weakest of the materials. The ductility of all materials was reduced by the irradiation, the uniform elongation to only 0.4% in the cold-worked material. Tests at temperatures above the irradiation temperature showed an approach to unirradiated properties as the temperature was increased from 200 to 600 0 C. Helium embrittlement at 700 0 C severely reduced elongation

Environmental embrittlement of intermetallic compounds in Fe-Al alloys

Institute of Scientific and Technical Information of China (English)

张建民; 张瑞林; S.H.YU; 余瑞璜

1996-01-01

First,it is proposed that hydrogen atoms occupy the interstitial sites in Fe3Al and FeAl.Then the environmental embrittlement of intermetallic compounds in Fe-Al alloys is studied in the light of calculated valence electron structures and bond energy of Fe3Al and FeAl containing hydrogen atoms.From the analyses it is found that the states of metal atoms will change,in which more lattice electrons will become covalent electrons to bond with hydrogen atoms when the atomic hydrogen diffuses into the intermetallic compounds in Fe-Al alloys,which will result in the decrease of local metallicity in Fe3Al and FeAl.Meanwhile,it is found that the crystal will easily cleave since solute hydrogen bonds with metal atoms and severely anisotropic bonds form.As a conclusion,these factors result in the environmental embrittlement of Fe3Al and FeAl.

HVEM in-situ deformation of neutron irradiated Fe-0.3a/oCu

International Nuclear Information System (INIS)

Gelles, D.S.; Thomas, L.E.; Powell, R.W.

1980-01-01

In an effort to better understand the nature of irradiation embrittlement of low alloy steels, in-situ HVEM deformation tests have been performed on pure iron and iron-0.3% copper ribbon tensile specimens in both the unirradiated and irradiated (2.5 x 10 19 n/cm 2 , E > 1 MeV at 290 0 C) conditions. Microstructural response is described principally in terms of the matrix in which cracks were observed to propagate across the different specimens. Major differences between the irradiation iron-copper alloy and the other material conditions were (1) the plastic zone which developed ahead to propagating cracks was smaller, and (2) dislocations were found to develop a crenulated structure uncharacteristic of the other conditions studied. It is inferred that the crenulations result from the presence of obstacles to dislocation motion and thereby demonstrate the matrix strengthening effect of small, radiation-induced, presumably copper-rich precipitates. However, the obstacles which are most effective in retarding dislocation motion are found to be distributed on a much coarser scale than that of the radiation-induced loops

Effect of solute elements on hardening of thermally-aged RPV model alloys

International Nuclear Information System (INIS)

Nomoto, A.; Nishida, K.; Dohi, K.; Soneda, N.; Liu, L.; Sekimura, N.; Li, Z.

2015-01-01

Embrittlement correlation methods for irradiated reactor pressure vessel (RPV) steels have been developed worldwide to predict the amount of embrittlement during plant operation. The effect of chemical composition on embrittlement is not fully understood, particularly the process of solute atom behavior during solute atom formation. In this series of slides we report the results of thermal ageing experiments of RPV model alloys in order to obtain information on the effect of chemical composition on the hardening process. We can draw the following conclusions. First, the addition of Ni or Si alone to Fe-Cu model alloys does not have clear effect but the addition of Mn to Fe-Cu-Ni alloy accelerates the cluster formation and hardening drastically, the effect of composition on the cluster strength is not clear. Secondly, the hardening process before the hardening peak has linear correlation with APT (Atom Probe Tomography) results and the RSS (Root-Sum-Square)sum model seems to explain the relationship between increase in hardness and APT data in a more consistent manner

Tensile properties and bend ductility of (Fe,Ni)3V long-range-ordered alloys after irradiation in HFIR

International Nuclear Information System (INIS)

Braski, D.N.

1984-01-01

The objective of this work was to determine the effect of neutron irradiation on the tensile properties and bend ductility of (Fe,Ni) 3 V long-range-ordered (LRO) alloys. Several (Fe,Ni) 3 V LRO alloys were irradiated in HFIR-CTR-42 and -43 at 400 to 600 0 C, to approximately 10 dpa and approximately 1000 at. ppm He. Additions of cerium or carbon and the use of cold-worked microstructures did not improve the embrittlement resistance of the LRO alloys. The LRO-37-5RS alloy, with a microstructure produced by rapid solidification, exhibited the highest ductilities, and further study of the RS microstructure is warranted. The correlation between bend ductility and tensile ductility was poor

Development of embrittlement prediction models for U.S. power reactors and the impact of the heat-affected zone to thermal annealing

International Nuclear Information System (INIS)

Wang, J.A.

1998-05-01

The NRC Regulatory Guide 1.99 Revision 2 was based on 177 surveillance data points and the EPRI data base, where 76% of 177 data points and 60% of EPRI data base were from Westinghouse's data. Therefore, other vendors' radiation environment may not be properly characterized by R.G. 1.99's prediction. To minimize scatter from the influences of the irradiation temperature, neutron energy spectrum, displacement rate, and plant operation procedures on embrittlement models, improved embrittlement models based on group data that have similar radiation environments and reactor design and operation criteria are examined. A total of 653 shift data points from the current FR-EDB, including 397 Westinghouse data, 93 B and W data, 37 CE data, and 106 GE data, are used. A nonlinear least squares fitting FORTRAN program, incorporating a Monte Carlo procedure with 35% and 10% uncertainty assigned to the fluence and shift data, respectively, was written for this study. In order to have the same adjusted fluence value for the weld and plate material in the same capsule, the Monte Carlo least squares fitting procedure has the ability to adjust the fluence values while running the weld and plate formula simultaneously. Six chemical components, namely, copper, nickel, phosphorus, sulfur, manganese, and molybdenum, were considered in the development of the new embrittlement models. The overall percentage of reduction of the 2-sigma margins per delta RTNDT predicted by the new embrittlement models, compared to that of R.G. 1.99, for weld and base materials are 42% and 36%, respectively. Currently, the need for thermal annealing is seriously being considered for several A302B type RPVs. From the macroscopic view point, even if base and weld materials were verified from mechanical tests to be fully recovered, the linking heat affected zone (HAZ) material has not been properly characterized. Thus the final overall recovery will still be unknown. The great data scatter of the HAZ metals may

Effect of pre-strain on susceptibility of Indian Reduced Activation Ferritic Martensitic Steel to hydrogen embrittlement

International Nuclear Information System (INIS)

Sonak, Sagar; Tiwari, Abhishek; Jain, Uttam; Keskar, Nachiket; Kumar, Sanjay; Singh, Ram N.; Dey, Gautam K.

2015-01-01

The role of pre-strain on hydrogen embrittlement susceptibility of Indian Reduced Activation Ferritic Martensitic Steel was investigated using constant nominal strain-rate tension test. The samples were pre-strained to different levels of plastic strain and their mechanical behavior and mode of fracture under the influence of hydrogen was studied. The effect of plastic pre-strain in the range of 0.5–2% on the ductility of the samples was prominent. Compared to samples without any pre-straining, effect of hydrogen was more pronounced on pre-strained samples. Prior deformation reduced the material ductility under the influence of hydrogen. Up to 35% reduction in the total strain was observed under the influence of hydrogen in pre-strained samples. Hydrogen charging resulted in increased occurrence of brittle zones on the fracture surface. Hydrogen Enhanced Decohesion (HEDE) was found to be the dominant mechanism of fracture.

Hydrogen Embrittlement Mechanism in Fatigue Behavior of Austenitic and Martensitic Stainless Steels

Directory of Open Access Journals (Sweden)

Sven Brück

2018-05-01

Full Text Available In the present study, the influence of hydrogen on the fatigue behavior of the high strength martensitic stainless steel X3CrNiMo13-4 and the metastable austenitic stainless steels X2Crni19-11 with various nickel contents was examined in the low and high cycle fatigue regime. The focus of the investigations were the changes in the mechanisms of short crack propagation. Experiments in laboratory air with uncharged and precharged specimen and uncharged specimen in pressurized hydrogen were carried out. The aim of the ongoing investigation was to determine and quantitatively describe the predominant processes of hydrogen embrittlement and their influence on the short fatigue crack morphology and crack growth rate. In addition, simulations were carried out on the short fatigue crack growth, in order to develop a detailed insight into the hydrogen embrittlement mechanisms relevant for cyclic loading conditions. It was found that a lower nickel content and a higher martensite content of the samples led to a higher susceptibility to hydrogen embrittlement. In addition, crack propagation and crack path could be simulated well with the simulation model.

Al and Si Influences on Hydrogen Embrittlement of Carbide-Free Bainitic Steel

Directory of Open Access Journals (Sweden)

Yanguo Li

2013-01-01

Full Text Available A first-principle method based on the density functional theory was applied to investigate the Al and Si influences on the hydrogen embrittlement of carbide-free bainitic steel. The hydrogen preference site, binding energy, diffusion behaviour, and electronic structure were calculated. The results showed that hydrogen preferred to be at the tetrahedral site. The binding energy of the cell with Si was the highest and it was decreased to be the worst by adding hydrogen. The diffusion barrier of hydrogen in the cell containing Al was the highest, so it was difficult for hydrogen to diffuse. Thus, hydrogen embrittlement can be reduced by Al rather than Si.

Heavy-section steel irradiation program. Progress report, October 1992--March 1993

International Nuclear Information System (INIS)

Corwin, W.R.

1998-04-01

Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. The RPV is one of only two more safety-related components of the plant for which a duplicate or redundant backup system does not exist. It is therefore imperative to understand and be able to predict the capabilities and limitations of the integrity inherent in the RPV. In particular, it is vital to fully understand the degree of irradiation-induced degradation of the RPV's fracture resistance that occurs during service. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established at Oak Ridge National Laboratory (ORNL) under sponsorship of the Nuclear Regulatory Commission (NRC). The primary goal of this major safety program is to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior (in particular, the fracture toughness properties) of typical pressure-vessel steels as they relate to light-water-reactor pressure-vessel integrity. The program centers on experimental assessments of irradiation-induced embrittlement (including the completion of certain irradiation studies previously conducted by the Heavy-Section Steel Technology Program) augmented by detailed examinations and modeling of the accompanying microstructural changes. Effects of specimen size; material chemistry; product form and microstructure; irradiation fluence, flux, temperature, and spectrum; and postirradiation annealing are being examined on a wide range of fracture properties

Microstructural evolution in low alloy steels under high dose ion irradiation

International Nuclear Information System (INIS)

Fujii, Katsuhiko; Fukuya, Koji; Ohkubo, Tadakatsu; Hono, Kazuhiro

2006-01-01

Radiation hardening and microstructural evolution in low Cu A533B steels (0.03 wt% Cu) irradiated by 3 MeV Ni 2+ ions at 290degC to 1 dpa were investigated by ultra-micro hardness measurement and leaser type three dimensional atom probe analysis. Mn-Ni-Si enriched precipitates were detected in the samples irradiated to 1 dpa by 3DAP analysis. The well-defined precipitates had a size of less than 4 nm, and the number density increased with dose. The formation of the precipitates under high dose rate irradiation suggested that Mn-Ni-Si enriched precipitates were formed by a process such as radiation induced precipitation rather than by thermal equilibrium process. The increase of yield stress calculated by size and number density of the precipitates in 1 dpa irradiated sample using the similar value of hardening efficiency to that of Cu rich precipitates was consistent with that estimated by data of increases of hardness measured by nano-indentation. The result indicates that effects of Mn-Ni-Si enriched precipitates on radiation embrittlement are similar to those of Cu rich precipitates. (author)

Hydrogen embrittlement of ASTM A 203 D nuclear structural steel

International Nuclear Information System (INIS)

Chakravartty, J.K.; Prasad, G.E.; Sinha, T.K.; Asundi, M.K.

1986-01-01

The influence of hydrogen on the mechanical properties of ASTM A 203 D nuclear structural steel has been studied by tension, bend and delayed-failure tests at room temperature. While the tension tests of hydrogen charged unnotched specimens reveal no change in ultimate strength and ductility, the effect of hydrogen is manifested in notched specimens (tensile and bend) as a decrease in ultimate strength (maximum load in bend test) and ductility; the effect increases with increasing hydrogen content. It is observed that for a given hydrogen concentration, the decrease in bend ductility is remarkably large compared to that in tensile ductility. Hydrogen charging does not cause any delayed-failure upto 200 h under an applied tensile stress, 0.85 times the notch tensile strength. However delayed failure occurs in hydrogen charged bend samples in less than 10 h under an applied bending load of about 0.80 times of the uncharged maximum load. Fractographs of hydrogen charged unnotched specimens show ductile dimple fracture, while those of notched tension and bend specimens under hydrogen-charged conditions show a mixture of ductile dimple and quasi-cleavage cracking. The proportion of quasi-cleavage cracking increases with increasing hydrogen content and this fracture mode is more predominant in bend specimens. The changes in tensile properties and fracture modes can reasonably be explained by existing theories of hydrogen embrittlement. An attempt is made to explain the significant difference in the embrittlement susceptibility of bend and tensile specimens in the light of difference in triaxiality and plastic zone size near the notch tip. (orig.)

Effects of neutron irradiation and fatigue on ductility of stainless steel DIN 1.4948

International Nuclear Information System (INIS)

Vries, M.I. de; Schaaf, B. van der; Staal, H.U.; Elen, J.D.

1978-10-01

Test specimens of stainless steel DIN 1.4948, which is similar to AISI type 304, have been irradiated at 723 K and 823 K up to fluences of 1.10 23 neutrons (n).m -2 and 5.10 24 n.m -2 (E > 0.1 MeV). These are representative conditions for the SNR-300 reactor vessel and inner components after 16 years of operation. High temperature (723 K to 1023 K) tensile tests at strain rates (depsilon/dt) from 10 -7 s -1 to 10 s -1 show a considerable decrease of tensile ductility. The extent depends on helium content, test temperature and strain rate. The atomic helium fractions of 3.10 -7 and 7.10 -6 result from the reactions of thermal neutrons with the 14 ppm boron, present in the steel. Helium embrittlement sets in at strain rates below 1 s -1 to 10 s -1 (the range of interest for Bethe-Tait accident analyses). A minimum total elongation value of 6% is shown at 923 K. The post-irradiation fatigue life is reduced by up to about 50% due to intergranular cracking. The combination of irradiation and fatigue causes a decrease of ductility after a smaller number of prior fatigue cycles than in the case of unirradiated material. (Auth.)

Brittle-fracture potential of irradiated Zircaloy-2 pressure tubes

Science.gov (United States)

Huang, F. H.

1993-12-01

Neutron irradiation can degrade the fracture toughness of Zircaloy-2 and may cause highly irradiated reactor components of this material to fail in a brittle manner. The effects of radiation embrittlement on the structural integrity of N Reactor pressure tubes are studied by performing KIc and JIc fracture toughness testing on samples cut from the Zircaloy-2 tubes periodically removed from the reactor. A fluence of 6 × 10 25n/ m2 ( E > 1 MeV) reduced the fracture toughness of the material by 40 to 50%. The fracture toughness values appear to saturate at 260°C with fluences above 3 × 10 25n/ m2 ( E > 1 MeV), but continue to decline with increasing fluence at temperatures below 177°C. Present and previous results obtained from irradiated pressure tubes indicate that the brittle-fracture potential of Zircaloy-2 increases with decreasing temperature and increasing fluence. Fractographic examinations of the fracture surfaces of irradiated samples reveal that circumferential hydride formation significantly influenced fracture morphology by providing sites for easy crack nucleation and leaving deep cracks. However, the deep cracks created at the hydride platelets in specimens containing less than 220 ppm hydrogen are not believed to be the major cause of degradation in postirradiation fracture toughness.

Alloys having improved resistance to hydrogen embrittlement

International Nuclear Information System (INIS)

Kane, R.D.; Greer, J.B.; Jacobs, D.F.; Berkowitz, B.J.

1983-01-01

The invention involves a process of improving the hydrogen embrittlement resistance of a cold-worked high yield strength nickel/cobalt base alloy containing chromium, and molybdenum and/or tungsten and having individual elemental impurity concentrations as measured by Auger spectroscopy at the crystallographic boundaries of up to about 1 Atomic percent. These elemental impurities are capable of becoming active and mobile at a temperature less than the recrystallization temperature of the alloy. The process involves heat treating the alloy at a temperature above 1300 degrees F but below the temperature of recrystallization for a time of from 1/4 to 100 hours. This is sufficient to effect a reduction in the level of the elemental impurities at the crystallographic boundaries to the range of less than 0.5 Atomic percent without causing an appreciable decrease in yield strength

Tensile properties of helium-injected V-15Cr-5Ti after irradiation in EBR-II

International Nuclear Information System (INIS)

Grossbeck, M.L.; Horak, J.A.

1985-01-01

Miniature specimens of V-15Cr-5Ti were prepared in the annealed condition and with 10, 20, and 30% cold work. The annealed specimens were cyclotron injected with helium and irradiated in sodium in EBR-II. The cold-worked specimens were irradiated in EBR-II but not helium injected. The specimens were irradiated at 400, 525, 625, and 700 0 C and received a fluence of 4.1 to 5.5 x 10 26 neutrons/m 2 (E > 0.1 meV). Tensile testing revealed very significant embrittlement as a result of the neutron irradiation but a much smaller change, mostly at 400 0 C, resulting from helium injection. 5 references, 9 figures, 2 tables

Post Irradiation Mechanical Behaviour of Three EUROFER Joints

International Nuclear Information System (INIS)

Lucon, E.; Leenaers, A.; Vandermeulen, W.

2006-01-01

The post-irradiation mechanical properties of three EUROFER joints (two diffusion joints and one TIG weld) have been characterized after irradiation to 1.8 dpa at 300 degrees Celsius in the BR-2 reactor. Tensile, KLST impact and fracture toughness tests have been performed. Based on the results obtained and on the comparison with data from EUROFER base material irradiated under similar conditions, the post-irradiation mechanical behaviour of both diffusion joints (laboratory and mock-up) appears similar to that of the base material. The properties of the TIG joint are affected by the lack of a post-weld heat treatment, which causes the material from the upper part of the weld to be significantly worse than that of the lower region. Thus, specimens from the upper layer exhibit extremely pronounced hardening and embrittlement caused by irradiation. The samples extracted from the lower layer show much better resistance to neutron exposure, although their measured properties do not match those of the diffusion joints. The results presented demonstrate that diffusion joining can be a very promising technique.

Transition temperature of embrittlement of steel 11 474.1 welded joint

International Nuclear Information System (INIS)

Petrikova, A.; Cocher, M.

1987-01-01

The results are presented of tests of notch toughness in dependence on temperature for steel 11 474.1 used for the manufacture of steam separators, in the area of a joint welded using an automatic submerged-arc welding machine with pre-heating at 200 to 250 degC. After welding, the welded joints were annealed for reduced stress for 160 minutes at a temperature of 600 to 650 degC and left to cool off in the furnace. The obtained results show that: (1) critical embrittlement temperature for the welded joint and the given welding technology ranges within -20 and -13 degC; (2) critical embrittlement temperature following heat ageing is shifted to positive temperature values; (3) pressure tests of the steam separator jacket made of steel 11 474.1 may in the process of production be carried out at a minimal wall temperature of 17 degC; (4) in case a pressure test has to be made after the equipment has been in operation for a certain period of time the test will probably have to be made at temperatures higher than 20 degC; (5) further tests will have to be made at temperatures higher than 20 degC in order to determine critical embrittlement temperatures after ageing. (J.B.). 7 figs., 2 tabs., 5 refs

Non-destructive evaluation of thermal aging embrittlement of duplex stainless steels

International Nuclear Information System (INIS)

Yi, Y.S.; Tomobe, T.; Watanabe, Y.; Shoji, T.

1993-01-01

The non-destructive evaluation procedure for detecting thermal aging embrittlement of cast duplex stainless steels has been investigated. As a novel measurement technique for the thermal aging embrittlement, an electrochemical method was used and anodic polarization behaviors were measured on new, service exposed, and laboratory aged materials and then were compared with the results of the mechanical tests and microstructural changes. During the polarization experiments performed in potassium hydroxide solution (KOH), M 23 C 6 carbides on phase boundary were preferentially dissolved, which was comfirmed by the SEM after polarization measurements. The preferential dissolution of M 23 C 6 carbides were obtained. Also, the non-destructive measurement and evaluation method of spinodal decomposition, which has been known as the primary mechanism of embrittlement inferrite phase, was reviewed. When the materials, where spinodal decomposition occurred, were polarized in an acetic acid solution (CH 3 COOH), larger critical anodic current densities were observed than those observed on new materials, and these results were consistent with the result of the microhardness measurement. Concerning these polarization results, a critical electric charge, which was required for stable passive films in passive metals, was defined and the relationship between the microstructural changes and this charge amount was reviewed under various polarization conditions in order to verify the polarization mechanism of the spinodally decomposed ferrite phase

Perspective on present and future alloy development efforts on austenitic stainless steels for fusion application

International Nuclear Information System (INIS)

Maziasz, P.J.

1984-01-01

The purpose of this paper is to address important questions concerning how to effect further alloy development of austenitic stainless steels for resistance, and to what extent the behavior of other properties under irradiation, such as strength/embrittlement, fatigue/irradiation creep, corrosion (under irradiation), and radiation-induced activation must be influenced. To summarize current understanding, helium has been found to have major effects on swelling and embrittlement, but several metallurgical avenues are available for significant improvement relative to type 316 stainless steel. Studies on fatigue and irradiation creep, particularly including helium effects, are preliminary but have yet to reveal engineering problems requiring additional alloy development remedies. The effects of irradiation on corrosion behavior are unknown, but higher alloy nickel contents make thermal corrosion in lithium worse. 67 refs

Evaluation on thermal aging embrittlement of cast stainless steel components in domestic PWRs

Energy Technology Data Exchange (ETDEWEB)

Lee, Bong Sang; Hwa, Hong Jun; Chi, Se Hwan; Ryu, Woo Seog; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1994-06-01

This report reviewed the R and D states of thermal aging embrittlement of cast stainless steel components in PWRs. Cast stainless steel is being widely used in PWRs including primary piping. This material shows the reduction of fracture toughness during operating life due to high temperature. Micromechanisms and kinetics are summarized to improve the materials properties. The reduction of toughness due to thermal embrittlement in domestic reactors are predicted based on each chemical composition until the end of plant life time. Substantial degradation was predicted in some components during plant life time. (Author) 26 refs., 19 figs., 11 tabs.

Hydrogen embrittlement considerations in niobium-base alloys for application in the ITER divertor

International Nuclear Information System (INIS)

Peterson, D.T.; Hull, A.B.; Loomis, B.A.

1991-01-01

The ITER divertor will be subjected to hydrogen from aqueous corrosion by the coolant and by transfer from the plasma. Global hydrogen concentrations are one factor in assessing hydrogen embrittlement but local concentrations affected by source fluxes and thermotransport in thermal gradients are more important considerations. Global hydrogen concentrations is some corrosion- tested alloys will be presented and interpreted. The degradation of mechanical properties of Nb-base alloys due to hydrogen is a complex function of temperature, hydrogen concentration, stresses and alloy composition. The known tendencies for embrittlement and hydride formation in Nb alloys are reviewed

Study and prediction model on low temperature aging embrittlement in duplex stainless steels

International Nuclear Information System (INIS)

Sanchez, L.; Gutierrez-Solana, F.

1997-01-01

Within the framework of a general study on low temperature (280-400 degree centigree) aging embrittlement in duplex stainless steels, a relationship has been obtained between aging, measured from ferrite hardness evolution, and bulk materials embrittlement, determined from fracture toughness and fracture impact tests. The existing correlation between the increase in ferrite hardness and its percentage presence in the fracture path supports this relationship and results in the development of a prediction design model which provides the fracture resistance curves, for any aging level, based on the chemical composition and the steel's properties in an unaged state. (Author)

The effect of segregated sp-impurities on grain-boundary and surface structure, magnetism and embrittlement in nickel

Czech Academy of Sciences Publication Activity Database

Všianská, Monika; Šob, Mojmír

2011-01-01

Roč. 56, č. 6 (2011), s. 817-840 ISSN 0079-6425 R&D Projects: GA AV ČR IAA100100920; GA MŠk(CZ) OC10008; GA ČR GD106/09/H035 Institutional research plan: CEZ:AV0Z20410507 Keywords : grain boundaries * segregation * nickel * embrittlement Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 18.216, year: 2011

Embrittlement of the alloy U 7.5 Nb 2.5 Zr by gaseous oxygen and hydrogen

International Nuclear Information System (INIS)

Lepoutre, D.; Nomine, A.M.; Miannay, D.

1981-04-01

Embrittlement of the alloy uranium 7.5 niobium 2.5 zirconium in gaseous oxygen and hydrogen versus stress intensity, temperature and pressure is studied using rupture mechanics. Cracking speed is determined. In oxygen, only cracks are produced and embrittlement is due to oxidation. In hydrogen at high pressure an hydride is formed and at low pressure cracks are produced but the mechanism is not identified [fr

High temperature service embrittlement of EUROFER´97 steel

Czech Academy of Sciences Publication Activity Database

Stratil, Luděk; Hadraba, Hynek; Dlouhý, Ivo

2010-01-01

Roč. 1, č. 2 (2010), s. 142-145 ISSN 1335-1532. [Fraktografia 2009. Stará Lesná, 08.11.2009-11.11.2009] R&D Projects: GA ČR GA106/08/1397; GA AV ČR 1QS200410502 Institutional research plan: CEZ:AV0Z20410507 Keywords : Eurofer´97 * isothermal ageing * embrittlement * impact properties Subject RIV: JL - Materials Fatigue, Friction Mechanics

Positron Annihilation Study of Ion-irradiated Si

International Nuclear Information System (INIS)

Shin, Jung Ki; Kwon, Jun Hyun; Lee, Jong Yong

2009-01-01

Structural parts like a spaceship, satellite and solar cell are composed of metal alloy or semiconductor materials. Especially, Si is used as a primary candidate alloy. But, manned and robotic missions to the Earth's moon and Mars are exposed to a continuous flux of Galactic Cosmic Rays (GCR) and occasional, but intense, fluxes of Solar Energetic Particles. These natural radiations impose hazards to manned exploration. Irradiation of cosmic particle induces various changes in the mechanical and physical properties of device steels. It is, therefore, important to investigate radiation damage to the component materials in semiconductor. The evolution of radiation-induced defects leads to degradation of the mechanical properties. One of them includes irradiation embrittlement, which can cause a loss of ductility and further increase the probability of a brittle fracture. It can be more dangerous in the space. Positron annihilation lifetime spectroscopy(PALS) have been applied to investigate the production of vacancy-type defects for Ion-irradiated Si wafer penetrated by H, He, O and Fe ions. Then, we carried out a comparison with an un-irradiated Si wafer

Nondestructive characterization of embrittlement in reactor pressure vessel steels -- A feasibility study

International Nuclear Information System (INIS)

McHenry, H.I.; Alers, G.A.

1998-01-01

The Nuclear Regulatory Commission recently initiated a study by NIST to assess the feasibility of using physical-property measurements for evaluating radiation embrittlement in reactor pressure vessel (RPV) steels. Ultrasonic and magnetic measurements provide the most promising approaches for nondestructive characterization of RPV steels because elastic waves and magnetic fields can sense the microstructural changes that embrittle materials. The microstructural changes of particular interest are copper precipitation hardening, which is the likely cause of radiation embrittlement in RPV steels, and the loss of dislocation mobility that is an attribute of the ductile-to-brittle transition. Measurements were made on a 1% copper steel, ASTM grade A710, in the annealed, peak-aged and overaged conditions, and on an RPV steel, ASTM grade A533B. Nonlinear ultrasonic and micromagnetic techniques were the most promising measures of precipitation hardening. Ultrasonic velocity measurements and the magnetic properties associated with hysteresis-loop measurements were not particularly sensitive to either precipitation hardening or the ductile-to-brittle transition. Measurements of internal friction using trapped ultrasonic resonance modes detected energy losses due to the motion of pinned dislocations; however, the ultrasonic attenuation associated with these measurements was small compared to the attenuation caused by beam spreading that would occur in conventional ultrasonic testing of RPVs

Quantitative evaluation of rejuvenators to restore embrittlement temperatures in oxidized asphalt mixtures using acoustic emission

Science.gov (United States)

Sun, Zhe; Farace, Nicholas; Arnold, Jacob; Behnia, Behzad; Buttlar, William G.; Reis, Henrique

2015-03-01

Towards developing a method capable to assess the efficiency of rejuvenators to restore embrittlement temperatures of oxidized asphalt binders towards their original, i.e., unaged values, three gyratory compacted specimens were manufactured with mixtures oven-aged for 36 hours at 135 °C. In addition, one gyratory compacted specimen manufactured using a short-term oven-aged mixture for two hours at 155 °C was used for control to simulate aging during plant production. Each of these four gyratory compacted specimens was then cut into two cylindrical specimen 5 cm thick for a total of six 36-hour oven-aged specimens and two short term aging specimens. Two specimens aged for 36 hours and the two short-term specimens were then tested using an acoustic emission approach to obtain base acoustic emission response of short-term and severely-aged specimens. The remaining four specimens oven-aged for 36 hours were then treated by spreading their top surface with rejuvenator in the amount of 10% of the binder by weight. These four specimens were then tested using the same acoustic emission approach after two, four, six, and eight weeks of dwell time. It was observed that the embrittlement temperatures of the short-term aged and severely oven-aged specimens were -25 °C and - 15 °C, respectively. It was also observed that after four weeks of dwell time, the rejuvenator-treated samples had recuperated the original embrittlement temperatures. In addition, it was also observed that the rejuvenator kept acting upon the binder after four weeks of dwell time; at eight weeks of dwell time, the specimens had an embrittlement temperature about one grade cooler than the embrittlement temperature corresponding to the short-term aged specimen.

Development of PIE techniques for irradiated LWR pressure vessel steels

International Nuclear Information System (INIS)

Nishi, Masahiro; Kizaki, Minoru; Sukegawa, Tomohide

1999-01-01

For the evaluation of safety and integrity of light water reactors (LWRs), various post irradiation examinations (PIEs) of reactor pressure vessel (RPV) steels and fuel claddings have been carried out in the Research Hot Laboratory (RHL). In recent years, the instrumented Charpy impact testing machine was remodeled aiming at the improvement of accuracy and reliability. By this remodeling, absorbed energy and other useful information on impact properties can be delivered from the force-displacement curve for the evaluation of neutron irradiation embrittlement behavior of LWR-RPV steels at one-time striking. In addition, two advanced PIE technologies are now under development. One is the remote machining of mechanical test pieces from actual irradiated pressure vessel steels. The other is development of low-cycle and high-cycle fatigue test technology in order to clarify the post-irradiation fatigue characteristics of structural and fuel cladding materials. (author)

Effect of niobium on the embrittlement of 2.25 Cr and 2.25 Cr-1Mo steels by phosphous

International Nuclear Information System (INIS)

Antunes, J.L.B.

1985-01-01

The influence of niobium on the temper embrittlement of 2.25Cr and 2.25 Cr-1Mo steels doped with phosphorus is evaluated. The transition temperatures of the samples tempered at 650 0 C and aged at different temperatures for niobium steels. (M.J.C.) [pt

Neutron embrittlement of the Kozloduy NPP unit 1 reactor

International Nuclear Information System (INIS)

Vodenicharov, S.; Kamenova, Tz.

1996-01-01

Activities made in the period 1989-1996 according to the Program for metal state monitoring of the Kozloduy NPP Unit 1 are described. Data on P and Cu content in the welded joint 4 are reported. Determination is made by wet chemical analysis of shavings taken out from the inner side of the wall, direct spectral analysis of the vessel itself and spectroscopy of the inner and outer side of 6 templates. The results obtained from 4 different study teams showed a good agreement. The real average P content is 0.046% and tends to diminish in depth. Microstructural investigation does not show any expressed inter-crystalline mechanism of brittle failure at low temperatures. The data on real P and Cu content, as well as the experimental values of the initial critical temperature of embrittlement (Tk o ), the residual part of temperature shift (Tk r ) and the re-embrittlement temperature after annealing at 475 o (Tk) allow to predict the change in Tk o of the joint 4 during the next refueling cycles. The measured low value of Tk after 18-th refueling cycle is considerably lower than that forecasted by lateral re-embrittlement law. This means that the forecasting of Tk for the next cycles is made with big enough conservatisms, and that a second annealing of the vessel until 26-th cycle is not necessary. So according to the most conservative estimate, the Unit 1 can operate safely until the end of the 26-th refueling cycle. It is also concluded, that in terms of radiation degradation of the vessel metal the operation life time of the Unit 1 can reach and exceed the designed one. 2 tab., 7 ref

Investigation of helium-induced embrittlement

International Nuclear Information System (INIS)

Sabelova, V.; Slugen, V.; Krsjak, V.

2014-01-01

In this work, the hardness of Fe-9%(wt.) Cr binary alloy implanted by helium ions up to 1000 nm was investigated. The implantations were performed using linear accelerator at temperatures below 80 grad C. Isochronal annealing up to 700 grad C with the step of 100 grad C was applied on the helium implanted samples in order to investigate helium induced embrittlement of material. Obtained results were compared with theoretical calculations of dpa profiles. Due to the results, the nano-hardness technique results to be an appropriate approach to the hardness determination of thin layers of implanted alloys. Both, experimental and theoretical calculation techniques (SRIM) show significant correlation of measured results of induced defects. (authors)

Metallic materials for the hydrogen energy industry and main gas pipelines: complex physical problems of aging, embrittlement, and failure

International Nuclear Information System (INIS)

Nechaev, Yu S

2008-01-01

The possibilities of effective solutions of relevant technological problems are considered based on the analysis of fundamental physical aspects, elucidation of the micromechanisms and interrelations of aging and hydrogen embrittlement of materials in the hydrogen industry and gas-main industries. The adverse effects these mechanisms and processes have on the service properties and technological lifetime of materials are analyzed. The concomitant fundamental process of formation of carbohydride-like and other nanosegregation structures at dislocations (with the segregation capacity 1 to 1.5 orders of magnitude greater than in the widely used Cottrell 'atmosphere' model) and grain boundaries is discussed, as is the way in which these structures affect technological processes (aging, hydrogen embrittlement, stress corrosion damage, and failure) and the physicomechanical properties of the metallic materials (including the technological lifetimes of pipeline steels). (reviews of topical problems)

Crack-arrest tests on two irradiated high-copper welds

International Nuclear Information System (INIS)

Iskander, S.K.; Corwin, W.R.; Nanstad, R.K.

1994-03-01

The objective of the Heavy-Section Steel Irradiation Program Sixth Irradiation Series is to determine the effect of neutron irradiation on the shift and shape of the lower-bound curve to crack-arrest toughness data. Two submerged-arc welds with copper contents of 0.23 and 0.31 wt % were commercially fabricated in 220-mm-thick plate. Crack-arrest specimens fabricated from these welds were irradiated at a nominal temperature of 288 degrees C to an average fluence of 1.9 x 10 19 neutrons/cm 2 (>1 MeV). This is the second report giving the results of the tests on irradiated duplex-type crack-arrest specimens. A previous report gave results of tests on irradiated weld-embrittled-type specimens. Charpy V-notch (CVN) specimens irradiated in the same capsules as the crack-arrest specimens were also tested, and a 41-J transition temperature shift was determined from these specimens. open-quotes Mean close-quote curves of the same form as the American Society of Mechanical Engineers (ASME) K la curve were fit to the data with only the open-quotes reference temperatureclose quotes as a parameter. The shift between the mean curves agrees well with the 41-J transition temperature shift obtained from the CVN specimen tests. Moreover, the four data points resulting from tests on the duplex crack-arrest specimens of the present study did not make a significant change to mean curve fits to either the previously obtained data or all the data combined

Studies on physical properties and fractography of electron beam irradiated poly(vinyl chloride)/epoxidized natural rubber blend in the presence of trimethylolpropane triacrylate

International Nuclear Information System (INIS)

Chantara Thevy Ratnam; Khairul Zaman Mohd Dahlan; Nasir, M.; Baharin, A.

2000-01-01

The effect of irradiation on the 50/50 poly(vinyl chloride)/epoxidized natural rubber blend was studied in the presence of 3 phr trimethylolpropane triacrylate (TMPTA). The blend was irradiated by using a 3.0 MeV electron beam machine at doses ranging from 20 to 200 kGy in air and room temperature. The tensile properties, resilience and gel fractions of the blends were measured. Electron beam irradiation of the blend in the presence of the TMPTA were found to cause crosslinking which in effect caused an enhancement in modulus and gel fraction together with a concomitant decline in ultimate elongation. The irradiation has resulted in a less hysteretic poly(vinyl chloride)/epoxidized natural rubber blend, with increased rebound resilience. The tensile strength of the blend reached a maximum at 60 kGy followed by a slight decrease at higher doses, implying embrittlement due to the excessive crosslinking. The scanning electron micrographs of the fracture surfaces of the irradiated blends show evidence consistent with the above contention. (Author)

An integrated approach to assessing the fracture safe margins of fusion reactor structures

International Nuclear Information System (INIS)

Odette, G.R.

1996-01-01

Design and operation of fusion reactor structures will require an appropriate data base closely coupled to a reliable failure analysis method to safely manage irradiation embrittlement. However, ongoing irradiation programs will not provide the information on embrittlement necessary to accomplish these objectives. A new engineering approach is proposed based on the concept of a master toughness-temperature curve indexed on an absolute temperature scale using shifts to account for variables such as size scales, crack geometry and loading rates as well as embrittlement. While providing a simple practical engineering expedient, the proposed method can also be greatly enhanced by fundamental mechanism based models of fracture and embrittlement. Indeed, such understanding is required for the effective use of small specimen test methods, which is a integral element in developing the necessary data base

Embrittlement of MISSE 5 Polymers After 13 Months of Space Exposure

Science.gov (United States)

Guo, Aobo; Yi, Grace T.; Ashmead, Claire C.; Mitchell, Gianna G.; deGroh, Kim K.

2012-01-01

Understanding space environment induced degradation of spacecraft materials is essential when designing durable and stable spacecraft components. As a result of space radiation, debris impacts, atomic oxygen interaction, and thermal cycling, the outer surfaces of space materials degrade when exposed to low Earth orbit (LEO). The objective of this study was to measure the embrittlement of 37 thin film polymers after LEO space exposure. The polymers were flown aboard the International Space Station and exposed to the LEO space environment as part of the Materials International Space Station Experiment 5 (MISSE 5). The samples were flown in a nadir-facing position for 13 months and were exposed to thermal cycling along with low doses of atomic oxygen, direct solar radiation and omnidirectional charged particle radiation. The samples were analyzed for space-induced embrittlement using a bend-test procedure in which the strain necessary to induce surface cracking was determined. Bend-testing was conducted using successively smaller mandrels to apply a surface strain to samples placed on a semi-suspended pliable platform. A pristine sample was also tested for each flight sample. Eighteen of the 37 flight samples experienced some degree of surface cracking during bend-testing, while none of the pristine samples experienced any degree of cracking. The results indicate that 49 percent of the MISSE 5 thin film polymers became embrittled in the space environment even though they were exposed to low doses (approx.2.75 krad (Si) dose through 127 mm Kapton) of ionizing radiation.

Hydrogen embrittlement of titanium and its alloys - a literature review

International Nuclear Information System (INIS)

Aho-Mantila, I.; Haemaelaeinen, H.

1986-05-01

Hydrogen embrittlement data of titanium and its alloys is reviewed. Especially the results obtained in spent nuclear fuel repository conditions with commercially pure titanium and TiCode-12 alloy are examined. The results show that the mechanical properties of titanium are not much affected by hydrogen when tested by smooth specimens. Much greater effects can be expected with notched fracture mechanics specimens. However, only limeted data is available. Hydrogen distribution in titanium is affected by stress, alloy composition and temperature gradients. In order to model the hydrogen-induced crack growth in titanium much more mechanistic work is needed especially to understand the behaviour of hydrogen in crack tip stress field. (author)

Gaseous hydrogen embrittlement of an API X80 ferrito-pearlitic steel; Fragilisation par l'hydrogene gazeux d'un acier ferrito-perlitique de grade API X80

Energy Technology Data Exchange (ETDEWEB)

Moro, I.

2009-11-15

This work deals with hydrogen embrittlement, at ambient temperature and under a high pressure gaseous way, of an API X80 high elasticity limit steel used for pipelines construction, and with the understanding of the associated physical mechanisms of the embrittlement. At first has been described a bibliographic study of the adsorption, absorption, diffusion, transport and trapping of hydrogen in the steels. Then has been carried out an experimental and numerical study concerning the implantation in the finite element code CASTEM3M of a hydrogen diffusion model coupled to mechanical fields. The hydrogen influence on the mechanical characteristics of the X80 steel, of a ferrito-pearlitic microstructure has been studied with tensile tests under 300 bar of hydrogen and at ambient temperature. The sensitivity of the X80 steel to hydrogen embrittlement has been analyzed by tensile tests at different deformation velocities and under different hydrogen pressures on axisymmetrical notched test specimens. These studies show that the effect of the hydrogen embrittlement vary effectively with the experimental conditions. Moreover, correlated with the results of the tests simulations, it has been shown too that in these experimental conditions and for that steel, the hydrogen embrittlement is induced by three different hydrogen populations: the hydrogen trapped at the ferrite/perlite interfaces, the hydrogen adsorbed on surface and the reticular hydrogen trapped in the material volume. At last, the tensile and rupture tests of specimens, during which atmosphere changes have been carried out, have shown a strong reversibility of the hydrogen embrittlement, associated with its initiation as soon as hydrogen is introduced in the atmosphere. At last, three hydrogen mechanisms, depending of the different hydrogen populations are presented and discussed. (O.M.)

The effect of microstructural change on the Charpy impact properties of the high-strength ferritic/martensitic steel (PNC-FMS) irradiated in JOYO/MARICO-1

International Nuclear Information System (INIS)

Yano, Yasuhide; Akasaka, Naoaki; Yoshitake, Tsunemitsu; Abe, Yasuhiro

2004-03-01

It is well known that the irradiation embrittlement is one of the most important issues to apply ferritic steels for FBR core materials, although ferritic steels have been considered to be candidate core materials of the commercialized FBR core material because of their superior swelling resistance. In order to evaluate the effects of microstructural changes during irradiation on the Charpy impact properties of the high-strength ferritic/martensitic steel (PNC-FMS), microstructural observations were performed with transmission electron microscopy on ruptured halves of the half-sized Charpy specimens of PNC-FMS irradiated in the JOYO/MARICO-1. The results obtained in this study are as follows: (1) There was remarkable disappearance of the lath of martensite in the samples irradiated at 650degC, although there was no significant change in microstructures, especially the lath of martensite between the samples irradiated at 500degC and unirradiated. The disappearance of martensitic lath in the samples irradiated at 650degC was larger than that of the samples thermally aged at 650degC. (2) The ductile-brittle transition temperature (DBTT) of irradiated PNC-FMS is judged to increase with the disappearance of martensitic lath and to decrease with the recovery in dislocations. (3) The decrease in the upper shelf energy (USE) of irradiated PNC-FMS is significantly accompanied by the change of precipitation behavior. (4) The Charpy impact properties and microstructures of PNC-FMS irradiated at 500degC were superior under these irradiation conditions. In future, it is necessary to establish how to evaluate Charpy impact properties in a high fluence region, based on theoretical methods introduced from the data gained in low fluence experiments, in addition to expanding the data area widely. (author)

Empirical Method to Estimate Hydrogen Embrittlement of Metals as a Function of Hydrogen Gas Pressure at Constant Temperature

Science.gov (United States)

Lee, Jonathan A.

2010-01-01

High pressure Hydrogen (H) gas has been known to have a deleterious effect on the mechanical properties of certain metals, particularly, the notched tensile strength, fracture toughness and ductility. The ratio of these properties in Hydrogen as compared to Helium or Air is called the Hydrogen Environment Embrittlement (HEE) Index, which is a useful method to classify the severity of H embrittlement and to aid in the material screening and selection for safety usage H gas environment. A comprehensive world-wide database compilation, in the past 50 years, has shown that the HEE index is mostly collected at two conveniently high H pressure points of 5 ksi and 10 ksi near room temperature. Since H embrittlement is directly related to pressure, the lack of HEE index at other pressure points has posed a technical problem for the designers to select appropriate materials at a specific H pressure for various applications in aerospace, alternate and renewable energy sectors for an emerging hydrogen economy. Based on the Power-Law mathematical relationship, an empirical method to accurately predict the HEE index, as a function of H pressure at constant temperature, is presented with a brief review on Sievert's law for gas-metal absorption.

Deformation twinning in irradiated ferritic/martensitic steels

Science.gov (United States)

Wang, K.; Dai, Y.; Spätig, P.

2018-04-01

Two different ferritic/martensitic steels were tensile tested to gain insight into the mechanisms of embrittlement induced by the combined effects of displacement damage and helium after proton/neutron irradiation in SINQ, the Swiss spallation neutron source. The irradiation conditions were in the range: 15.8-19.8 dpa (displacement per atom) with 1370-1750 appm He at 245-300 °C. All the samples fractured in brittle mode with intergranular or cleavage fracture surfaces when tested at room temperature (RT) or 300 °C. After tensile test, transmission electron microscopy (TEM) was employed to investigate the deformation microstructures. TEM-lamella samples were extracted directly below the intergranular fracture surfaces or cleavage surfaces by using the focused ion beam technique. Deformation twinning was observed in irradiated specimens at high irradiation dose. Only twins with {112} plane were observed in all of the samples. The average thickness of twins is about 40 nm. Twins initiated at the fracture surface, became gradually thinner with distance away from the fracture surface and finally stopped in the matrix. Novel features such as twin-precipitate interactions, twin-grain boundary and/or twin-lath boundary interactions were observed. Twinning bands were seen to be arrested by grain boundaries or large precipitates, but could penetrate martensitic lath boundaries. Unlike the case of defect free channels, small defect-clusters, dislocation loops and dense small helium bubbles were observed inside twins.

Effect of aluminium concentration and boron dopant on environmental embrittlement in FeAl aluminides

International Nuclear Information System (INIS)

Liu, C.T.; George, E.P.

1991-01-01

This paper reports on the room-temperature tensile properties of FeAl aluminides determined as functions of aluminum concentration (35 to 43 at. % Al), test environment, and surface (oil) coating. The two lower aluminum alloys containing 35 and 36.5% Al are prone to severe environmental embrittlement, while the two higher aluminum alloys with 40 and 43% Al are much less sensitive to change in test environment and surface coating. The reason for the different behavior is that the grain boundaries are intrinsically weak in the higher aluminum alloys, and these weak boundaries dominate the low ductility and brittle fracture behavior of the 40 and 43% Al alloys. When boron is added to the 40% Al alloy as a grain-boundary strengthener, the environmental effect becomes prominent. In this case, the tensile ductility of the boron-doped alloy, just like that of the lower aluminum alloys, can be dramatically improved by control of test environment (e.g. dry oxygen vs air). Strong segregation of boron to the grain boundaries, with a segregation factor of 43, was revealed by Auger analyses

Damage and failure of unirradiated and irradiated fuel rods tested under film boiling conditions

International Nuclear Information System (INIS)

Mehner, A.S.; Hobbins, R.R.; Seiffert, S.L.; MacDonald, P.E.; McCardell, R.K.

1979-01-01

Power-cooling-mismatch experiments are being conducted as part of the Thermal Fuels Behavior Program in the Power Burst Facility at the Idaho National Engineering Laboratory to evaluate the behavior of unirradiated and previously irradiated light water reactor fuel rods tested under stable film boiling conditions. The observed damage that occurs to the fuel rod cladding and the fuel as a result of film boiling operation is reported. Analyses performed as a part of the study on the effects of operating failed fuel rods in film boiling, and rod failure mechanisms due to cladding embrittlement and cladding melting upon being contacted by molten fuel are summarized

Investigation of Liquid Metal Embrittlement of Materials for use in Fusion Reactors

Science.gov (United States)

Kennedy, Daniel; Jaworski, Michael

2014-10-01

Liquid metals can provide a continually replenished material for the first wall and extraction blankets of fusion reactors. However, research has shown that solid metal surfaces will experience embrittlement when exposed to liquid metals under stress. Therefore, it is important to understand the changes in structural strength of the solid metal materials and test different surface treatments that can limit embrittlement. Research was conducted to design and build an apparatus for exposing solid metal samples to liquid metal under high stress and temperature. The apparatus design, results of tensile testing, and surface imaging of fractured samples will be presented. This work was supported in part by the U.S. Department of Energy, Office of Science, Office of Workforce Development for Teachers and Scientists (WDTS) under the Science Undergraduate Laboratory Internships Program (SULI).

An holistic approach to the problem of reactor ageing. [Pressure vessel embrittlement

Energy Technology Data Exchange (ETDEWEB)

Phythian, W.; McElroy, R.; Druce, S.; Kovan, D. (AEA Reactor Services, Harwell (United Kingdom))

1992-12-01

Understanding the process of ageing in reactors is essential to extending their lives beyond original design. To present a sound case -particularly regarding the level of embrittlement in reactor vessels due to radiation damage - an integrated approach using advanced assessment tools is needed. The techniques developed for the purpose involve, on the microscopic level, advanced neutron dosimetry and high resolution measurement techniques (eg advanced electron beam techniques and small angle neutron scattering) with which an analysis can be done of the radiation damage and the microstructural state of the steel test procedures (tensile, fracture toughness and Charpy impact) on standard and sub-sized specimens, the extent of radiation degradation can be characterised. finally, it is possible to predict how the degradation will evolve using physically-based models of embrittlement. (Author).

Hydrogen Embrittlement Mechanism in Fatigue Behaviour of Austenitic and Martensitic Stainless Steels

Directory of Open Access Journals (Sweden)

Brück Sven

2018-01-01

Full Text Available In the present study, the influence of hydrogen on the fatigue behaviour of the high strength martensitic stainless steel X3CrNiMo13-4 and the metastable austenitic stainless steels X2Crni19-11 with various nickel contents was examined in the low and high cycle fatigue regime. The focus of the investigations was the changes in the mechanisms of short crack propagation. The aim of the ongoing investigation is to determine and quantitatively describe the predominant processes of hydrogen embrittlement and their influence on the short fatigue crack morphology and crack growth rate. In addition, simulations were carried out on the short fatigue crack growth, in order to develop a detailed insight into the hydrogen embrittlement mechanisms relevant for cyclic loading conditions.

Initial evaluation of ultrasonic attenuation measurements for estimating fracture toughness of RPV steels

Energy Technology Data Exchange (ETDEWEB)

Hiser, A.L. Jr.; Green, R.E. Jr. [Johns Hopkins Univ., Baltimore, MD (United States). Center for Nondestructive Evaluation

1999-08-01

Neutron bombardment of reactor pressure vessel (RPV) steels causes reductions in fracture toughness in these steels, termed neutron irradiation embrittlement. Currently, there are no accepted methods for nondestructive determination of the extent of the irradiation embrittlement nor the actual fracture toughness of the reactor pressure vessel. This paper provides initial results of an effort addressing the use of ultrasonic attenuation as a suitable parameter for nondestructive determination of irradiation embrittlement in RPV steels. (orig.)

Embrittlement and annealing of reactor pressure vessel steels: comparison of BR3 surveillance and vessel plates to the surrogate plates representative of the Yankee Rowe vessel

International Nuclear Information System (INIS)

Fabry, A.; Chaouadi, T.; Puzzolante, J.L.; Van de Velde, J.; Biemiller, E.C.; Rossinski, S.T.; Carter, R.G.

1996-07-01

The sister pressure vessels at the BR3 and Rowe Yankee PWR plants were operated at a lower-than-usual temperature (260 degrees Celsius) and their plates were austenitized at higher-than-usual temperature (970 degrees Celsius). A heat tratemement leading to a coarser microstructure than typical for the fine grain plates that are considered in development of USNRC Regulatory guide 1.99. This material displayed outlier behaviour charackterized by a 41J CVN-shift significantly larger than predicted by Regulatory Guide 1.99. Because lower radiation temperature and nickell alloying are generally considered detrimental to irradiation sensitivity, there was a major concern that the nickel-modified lower Rowe plate and the nickel-modified BR3 plate may become too embrittled to satisfy the toughness requirements enbodied in the PTS screening criterion. This paper compares three complementary studies undertaken to clarify these uncertainties: 1) the accelerated irradiation and test program launched in 1990 by Yankee Atomic Electric Company using typical vessel plate materials containing 0.24% copper at two nickel levels: YA1, 0.63% (A533-B) and YA9, 0.19% (A302-B). These were heat-treated to produce the coarse and fine grain microstructures representative of the Yankee/BR3 and the Regulatory Guide plates, respectively, 2) the BR3 surveillance and vessel testing program: this vessel was wet-annealed in 1984, relicensed for operation till the plant shutdown in 1987, ANCL was trepanned in early 1995, 3) the accelerated irradiations in the Belgian BR2 test reactor of the Yankee coarse grain plates YA1 and YA9 together with BR3 vessel specimens extracted at nozzle elevation, a location with negligible radiation exposure. It is shown that the PTS screening criterion was never attained by the BR3 and Rowe plates, and that the BR3 vessel anneal was neither necessary nor sufficient. Finally, the sensitivity of embrittlement, annealing and post-annealing reembrittlement to irradiation

Embrittlement and annealing of reactor pressure vessel steels: comparison of BR3 surveillance and vessel plates to the surrogate plates representative of the Yankee Rowe vessel

Energy Technology Data Exchange (ETDEWEB)

Fabry, A.; Chaouadi, T.; Puzzolante, J.L.; Van de Velde, J. [Centre de l``Etude de l``Energie Nucleaire, Mol (Belgium); Biemiller, E.C. [Yankee Atomic Electric Company, Bolton (United States); Rossinski, S.T.; Carter, R.G. [Electric Power Research Institute, Charlotte (United States)

1996-07-01

The sister pressure vessels at the BR3 and Rowe Yankee PWR plants were operated at a lower-than-usual temperature (260 degrees Celsius) and their plates were austenitized at higher-than-usual temperature (970 degrees Celsius). A heat tratemement leading to a coarser microstructure than typical for the fine grain plates that are considered in development of USNRC Regulatory guide 1.99. This material displayed outlier behaviour charackterized by a 41J CVN-shift significantly larger than predicted by Regulatory Guide 1.99. Because lower radiation temperature and nickell alloying are generally considered detrimental to irradiation sensitivity, there was a major concern that the nickel-modified lower Rowe plate and the nickel-modified BR3 plate may become too embrittled to satisfy the toughness requirements enbodied in the PTS screening criterion. This paper compares three complementary studies undertaken to clarify these uncertainties: 1) the accelerated irradiation and test program launched in 1990 by Yankee Atomic Electric Company using typical vessel plate materials containing 0.24% copper at two nickel levels: YA1, 0.63% (A533-B) and YA9, 0.19% (A302-B). These were heat-treated to produce the coarse and fine grain microstructures representative of the Yankee/BR3 and the Regulatory Guide plates, respectively, 2) the BR3 surveillance and vessel testing program: this vessel was wet-annealed in 1984, relicensed for operation till the plant shutdown in 1987, ANCL was trepanned in early 1995, 3) the accelerated irradiations in the Belgian BR2 test reactor of the Yankee coarse grain plates YA1 and YA9 together with BR3 vessel specimens extracted at nozzle elevation, a location with negligible radiation exposure. It is shown that the PTS screening criterion was never attained by the BR3 and Rowe plates, and that the BR3 vessel anneal was neither necessary nor sufficient. Finally, the sensitivity of embrittlement, annealing and post-annealing reembrittlement to irradiation

Diagnostic experimental results on the hydrogen embrittlement of austenitic steels

Energy Technology Data Exchange (ETDEWEB)

Gavriljuk, V.G.; Shivanyuk, V.N.; Foct, J

2003-03-14

Three main available hypotheses of hydrogen embrittlement are analysed in relation to austenitic steels based on the studies of the hydrogen effect on the interatomic bonds, phase transformations and microplastic behaviour. It is shown that hydrogen increases the concentration of free electrons, i.e. enhances the metallic character of atomic interactions, although such a decrease in the interatomic bonding cannot be a reason for brittleness and rather assists an increased plasticity. The hypothesis of the critical role of the hydrogen-induced {epsilon} martensite was tested in the experiment with the hydrogen-charged Si-containing austenitic steel. Both the fraction of the {epsilon} martensite and resistance to hydrogen embrittlement were increased due to Si alloying, which is at variance with the pseudo-hydride hypothesis. The hydrogen-caused early start of the microplastic deformation and an increased mobility of dislocations, which are usually not observed in the common mechanical tests, are revealed by the measurements of the strain-dependent internal friction, which is consistent with the hypothesis of the hydrogen-enhanced localised plasticity. An influence of alloying elements on the enthalpy E{sub H} of hydrogen migration in austenitic steels is studied using the temperature-dependent internal friction and a correlation is found between the values of E{sub H} and hydrogen-caused decrease in plasticity. A mechanism for the transition from the hydrogen-caused microplasticity to the apparent macrobrittle fracture is proposed based on the similarity of the fracture of hydrogenated austenitic steels to that of high nitrogen steels.

Embrittlement recovery due to annealing of reactor pressure vessel steels

International Nuclear Information System (INIS)

Eason, E.D.; Wright, J.E.; Nelson, E.E.; Odette, G.R.; Mader, E.V.

1996-01-01

Embrittlement of reactor pressure vessels (RPVs) can be reduced by thermal annealing at temperatures higher than the normal operating conditions. Although such an annealing process has not been applied to any commercial plants in the United States, one US Army reactor, the BR3 plant in Belgium, and several plants in eastern Europe have been successfully annealed. All available Charpy annealing data were collected and analyzed in this project to develop quantitative models for estimating the recovery in 30 ft-lb (41 J) Charpy transition temperature and Charpy upper shelf energy over a range of potential annealing conditions. Pattern recognition, transformation analysis, residual studies, and the current understanding of the mechanisms involved in the annealing process were used to guide the selection of the most sensitive variables and correlating parameters and to determine the optimal functional forms for fitting the data. The resulting models were fitted by nonlinear least squares. The use of advanced tools, the larger data base now available, and insight from surrogate hardness data produced improved models for quantitative evaluation of the effects of annealing. The quality of models fitted in this project was evaluated by considering both the Charpy annealing data used for fitting and the surrogate hardness data base. The standard errors of the resulting recovery models relative to calibration data are comparable to the uncertainty in unirradiated Charpy data. This work also demonstrates that microhardness recovery is a good surrogate for transition temperature shift recovery and that there is a high level of consistency between the observed annealing trends and fundamental models of embrittlement and recovery processes

Study on the hydrogen embrittlement and corrosion of stainless steels used as NI/MHX battery containers

Energy Technology Data Exchange (ETDEWEB)

Chuang, H.J.; Chan, S.L.I. [National Taiwan University, Taipei (China); Chen, S.Y. [Chung Shan Institute of Science and Technology, Lung-Tan (China)

1998-07-01

Stainless steels are used as the containers for Nickel-metal hydride (Ni/MHx) batteries. In this work stainless steel 304, 304L, 316, 316L, 17-4PH and 430 were selected to study their relative susceptibility to hydrogen embrittlement and alkaline corrosion under battery environments. Comparisons were made by immersion test under different hydrogen pressure over the electrolyte, U-bend tests and slow strain rate tensile test with cathodic H{sub 2} charging. The results showed that high strength 17-4PH suffered severe corrosion after long time immersion in the electrolyte solution and were sensitive to hydrogen embrittlement after hydrogen charging. Ferritic 430 performed better than 17-4PH during immersion test but lost its ductility after hydrogen charging. All the austenitic steels (304, 304L, 316, 316L) were found to be suitable as the materials for Ni/MHx battery container, and the present tests can not discriminate their relative resistance to the corrosion and hydrogen embrittlement in the electrolyte. 5 refs.

Postirradiation examination results for the irradiation effects scoping test 1

International Nuclear Information System (INIS)

Mehner, A.S.; Quapp, W.J.; Goetzmann, O.; Hobbins, R.R.

1976-09-01

A zircaloy-clad UO 2 fuel rod was operated above its critical heat flux within the in-pile test loop of the Power Burst Facility and later examined in the hot cells. The results of the postirradiation examinations are presented in this report. A Zr-UO 2 reaction at the fuel-cladding interface embrittled nearly as much of the cladding wall thickness as the Zr-water reaction on the exterior. Data on both the internal and external reactions, and cladding and fuel microstructures, are presented. Cladding embrittlement and rod failure are compared with several rod fragmentation criteria, and conclusions concerning fuel rod failure propagation in a power reactor system are made

Postirradiation examination results for the irradiation effects scoping test 1

International Nuclear Information System (INIS)

Mehner, A.S.; Quapp, W.J.; Goetzmann, O.; Hobbins, R.R.

1976-09-01

A zircaloy-clad UO 2 fuel rod was operated above its critical heat flux within the in-pile test loop of the Power Burst Facility and later examined in the hot cells. The results of the postirradiation examinations are presented. A Zr-UO 2 reaction at the fuel-cladding interface embrittled nearly as much of the cladding wall thickness as the Zr-water reaction on the exterior. Data on both the internal and external reactions and the cladding and fuel microstructures are presented. Cladding embrittlement and rod failure are compared with several rod fragmentation criteria, and conclusions concerning fuel rod failure propagation in a power reactor system are made

Effects of ATR-2 Irradiation to High Fluence on Nine RPV Surveillance Materials

Energy Technology Data Exchange (ETDEWEB)

Nanstad, Randy K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Odette, George R. [Univ. of California, Santa Barbara, CA (United States); Almirall, Nathan [Univ. of California, Santa Barbara, CA (United States); Robertson, Janet [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Server, W. L. [ATI Consulting, Pinehurst, NC (United States); Yamamoto, T. [Univ. of California, Santa Barbara, CA (United States); Wells, Peter [Univ. of California, Santa Barbara, CA (United States)

2017-05-01

The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations that govern the operation of commercial nuclear power plants require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including pressurized thermal shock (PTS) in pressurized water reactors (PWR). In the irradiated condition, however, the fracture toughness of the RPV may be severely degraded, with the degree of toughness loss dependent on the radiation sensitivity of the materials. The available embrittlement predictive models and our present understanding of radiation damage are not fully quantitative, and do not treat all potentially significant variables and issues, particularly considering extension of operation to 80y.

Reactor pressure vessel embrittlement

International Nuclear Information System (INIS)

1992-07-01

Within the framework of the IAEA extrabudgetary programme on the Safety of WWER-440/230 NPPs, a list of safety issues requiring broad studies of generic interest have been agreed upon by an Advisory Group who met in Vienna in September 1990. The list was later revised in the light of the programme findings. The information on the status of the issues, and on the amount of work already completed and under way in the various countries, needs to be compiled. Moreover, an evaluation of what further work is required to resolve each one of the issues is also necessary. In view of this, the IAEA has started the preparation of a series of status reports on the various issues. This report on the generic safety issue ''Reactor Pressure Vessel Embrittlement'' presents a comprehensive survey of technical information available in the field and identifies those aspects which require further investigation. 39 refs, 21 figs, 4 tabs

Reference manual on the IAEA JRQ correlation monitor steel for irradiation damage studies

International Nuclear Information System (INIS)

2001-07-01

The objective of this report is to provide information on the mechanical properties of the ASTM A533 grade B class 1 steel that was designated as 'JRQ reference steel' and for many years served as a radiation/mechanical property correlation monitor in a number of international and national studies of irradiation embrittlement of reactor pressure vessel steel. This report provides the most comprehensive listing of material test data obtained on the JRQ manufacturing history and material properties in the initial, and as delivered condition during the implementation of two IAEA co-ordinated research projects (CRPs) on behaviour of reactor pressure vessel steels under neutron irradiation

Investigation of structural materials of reactors using high-energy heavy-ion irradiations

International Nuclear Information System (INIS)

Wang Zhiguang

2007-01-01

Radiation damage in structural materials of fission/fusion reactors is mainly attributed to the evolution of intensive atom displacement damage induced by energetic particles (n, α and/or fission fragments) and high-rate helium doping by direct α particle bombardments and/or (n, α) reactions. It can cause severe degradation of reactor structural materials such as surface blistering, bulk void swelling, deformation, fatigue, embrittlement, stress erosion corrosion and so on that will significantly affect the operation safety of reactors. However, up to now, behavior of structural materials at the end of their service can hardly be fully tested in a real reactor. In the present work, damage process in reactor structural materials is briefly introduced, then the advantages of energetic ion implantation/irradiation especially high-energy heavy ion irradiation are discussed, and several typical examples on simulation of radiation effects in reactor candidate structural materials using high-energy heavy ion irradiations are pronounced. Experimental results and theoretical analysis suggested that irradiation with energetic particles especially high-energy heavy ions is very useful technique for simulating the evolution of microstructures and macro-properties of reactor structural materials. Furthermore, an on-going plan of material irradiation experiments using high energy H- and He-ions based on the Heavy Ion Research Facilities in Lanzhou (HIRFL) is also briefly interpreted. (authors)

Radiation effects on reactor pressure vessel supports

International Nuclear Information System (INIS)

Johnson, R.E.

1996-05-01

The purpose of this report is to present the findings from the work done in accordance with the Task Action Plan developed to resolve the Nuclear Regulatory Commission (NRC) Generic Safety Issue No. 15, (GSI-15). GSI-15 was established to evaluate the potential for low-temperature, low-flux-level neutron irradiation to embrittle reactor pressure vessel (RPV) supports to the point of compromising plant safety. An evaluation of surveillance samples from the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) had suggested that some materials used for RPV supports in pressurized-water reactors could exhibit higher than expected embrittlement rates. However, further tests designed to evaluate the applicability of the HFIR data to reactor RPV supports under operating conditions led to the conclusion that RPV supports could be evaluated using traditional method. It was found that the unique HFIR radiation environment allowed the gamma radiation to contribute significantly to the embrittlement. The shielding provided by the thick steel RPV shell ensures that degradation of RPV supports from gamma irradiation is improbable or minimal. The findings reported herein were used, in part, as the basis for technical resolution of the issue

Effect of post weld heat treatments on the resistance to the hydrogen embrittlement of soft martensitic stainless steel

International Nuclear Information System (INIS)

Hazarabedian, Alfredo; Ovejero Garcia, Jose; Bilmes, P.; Llorente, C.

2003-01-01

The effect of external hydrogen on the tensile properties of an all weld sample of a soft martensitic stainless steel was studied. The material was tested in the as weld condition and after tempered conditions modifying the austenite content, and changing the quantity, type and distribution of precipitates. Hydrogen was introduced by cathodic charge or by immersion in an acid brine saturated whit 1 atm hydrogen sulphide, during the mechanical test. The as weld condition showed a good resistance in the hydrogen sulphide, were the tempered samples were embrittled. Under cathodic charge, all samples were susceptible to hydrogen damage. The embritting mechanisms were the same in both environments. When the austenite content, was below 10% the crack path is on the primary austenite grain boundary. At higher austenite content, the crack is transgranular. (author)

Hardening and microstructural evolution of A533b steels irradiated with Fe ions and electrons

Energy Technology Data Exchange (ETDEWEB)

Watanabe, H., E-mail: watanabe@riam.kyushu-u.ac.jp [Research Institute for Applied Mechanics, Kyushu University, 6-1, Kasuga-kouenn, Kasugashi, Fukuoka, 816-8580 (Japan); Arase, S. [Interdisciplinary Graduate School of Kyushu University, 6-1, Kasuga-kouenn, Kasugashi, Fukuoka, 816-8580 (Japan); Yamamoto, T.; Wells, P. [Dept. Chemical Engineering, UCSB Engineering II, RM3357, Santa Barbara, CA, 93106-5080 (United States); Onishi, T. [Interdisciplinary Graduate School of Kyushu University, 6-1, Kasuga-kouenn, Kasugashi, Fukuoka, 816-8580 (Japan); Odette, G.R. [Dept. Chemical Engineering, UCSB Engineering II, RM3357, Santa Barbara, CA, 93106-5080 (United States)

2016-04-01

Radiation hardening and embrittlement of A533B steels is heavily dependent on the Cu content. In this study, to investigate the effect of copper on the microstructural evolution of these materials, A533B steels with different Cu levels were irradiated with 2.4 MeV Fe ions and 1.0 MeV electrons. Ion irradiation was performed from room temperature (RT) to 350 °C with doses up to 1 dpa. At RT and 290 °C, low dose (<0.1 dpa) hardening trend corresponded with ΔH ∝ (dpa){sup n}, with n initially approximately 0.5 and consistent with a barrier hardening mechanism, but saturating at ≈0.1 dpa. At higher dose levels, the radiation-induced hardening exhibited a strong Cu content dependence at 290 °C, but not at 350 °C. Electron irradiation using high-voltage electron microscopy revealed the growth of interstitial-type dislocation loops and enrichment of Ni, Mn, and Si in the vicinities of pre-existing dislocations at doses for which the radiation-induced hardness due to ion irradiation was prominent.

Influence of pre-hydriding on embrittlement of E110 alloy under LOCA conditions

International Nuclear Information System (INIS)

VNIINM, Moscow (Russian Federation))" data-affiliation=" (SC VNIINM, Moscow (Russian Federation))" >Fedotov, P.; VNIINM, Moscow (Russian Federation))" data-affiliation=" (SC VNIINM, Moscow (Russian Federation))" >Kuznetsov, V.; VNIINM, Moscow (Russian Federation))" data-affiliation=" (SC VNIINM, Moscow (Russian Federation))" >Nechaeva, O.; VNIINM, Moscow (Russian Federation))" data-affiliation=" (SC VNIINM, Moscow (Russian Federation))" >Novikov, V.; VNIINM, Moscow (Russian Federation))" data-affiliation=" (SC VNIINM, Moscow (Russian Federation))" >Salatov, A.; Ignatiev, D.; Mokrushin, A.; Soldatkin, D.; Urusov, A.

2015-01-01

The researches presented in this paper were carried out in the framework of TVS-K project developed by JSC “TVEL”. The data on the corrosion and residual ductility of unirradiated and pre-hydrided E110 alloy under LACA conditions at temperature range from 1100 to 1200°C are presented. The hydrogen concentration was varied from 30 (as-received) to 600 wppm. The initial concentration of hydrogen has no effect on the oxidation kinetics, while the oxidation kinetics are parabolic and the breakaway oxidation is not observed. Oxide films on surfaces of claddings are black and shining. There are no cracks, visual spots and peelings. The residual ductility of oxidised samples decrease with hydrogen concentration rise. The residual ductility of claddings oxidized at 1100 °C, generally higher than the same of the claddings oxidized at 1200 °C. E110 alloy has a good residual ductility in comparison to Zry-4, ZIRLO, M5. Joint analysis of the test results allowed us to formulate embrittlement criteria of the E110 alloy under LOCA conditions. This embrittlement criterion is preliminary, because the experimental data base must to be enlarged by results of tests with claddings of another geometry and quench experiments. (author)

Effect of medium and post-irradiation storage on rooting of irradiated onions

International Nuclear Information System (INIS)

Singh, Rita

2000-01-01

Rooting test for detection of irradiation in onion bulbs was studied. Onions were exposed to different dose levels of 30, 60, 90, 120 and 150 Gy. The effects of irradiation dose, cultivar difference, rooting medium and post-irradiation storage on the rooting were investigated. The number and the length of the roots formed in onions were found to decrease on irradiation. The effect was more at higher doses. The effect of irradiation on rooting was also evident after 120 days of storage. (author)

Structural mechanisms of the flux effect for VVER-1000 reactor pressure vessel materials

International Nuclear Information System (INIS)

Gurovich, B.; Kuleshova, E.; Fedotova, S.; Maltsev, D.; Zabusov, O.; Frolov, A.; Erak, D.; Zhurko, D.

2015-01-01

To justify the lifetime extension of VVER-1000 reactor pressure vessels (RPV) up to 60 years and more it is necessary to expand the existing surveillance samples database to beyond design fluence by means of accelerated irradiation in a research reactor. Herewith since the changes in mechanical properties of materials under irradiation are due to occurring structural changes, correct analysis of the data obtained at accelerated irradiation of VVER-1000 RPV materials requires a clear understanding of the structural mechanisms that are responsible for the flux effect in VVER-1000 RPV steels. Two mechanisms are responsible for radiation embrittlement of VVER-1000 RPV steels: the hardening one (radiation hardening due to formation of radiation-induced Ni-based precipitates and radiation defects) and non-hardening one (due to formation of impurities segregations at grain boundaries - reversible temper brittleness). In this context for an adequate interpretation of the mechanical tests results when justifying the lifetime extension of existing units a complex of comparative structural studies (TEM, SEM and AES) of VVER-1000 RPV materials irradiated in different conditions (in research reactor IR-8 and within surveillance samples) was performed. It is shown that the flux effect is observed for materials with high nickel content (weld metals with Ni content > 1.35%) and it is mostly due to the contribution of non-hardening mechanism of radiation embrittlement (the difference in the accumulation kinetics of grain boundary phosphorus segregation) and somewhat contribution of the hardening mechanism (the difference in density of radiation-induced precipitates). Therefore when analyzing the results obtained from the accelerated irradiation of VVER-1000 WM the correction for the flux effect should be made. (authors)

An Atomistic Modeling Study of Alloying Element Impurity Element, and Transmutation Products on the cohesion of A Nickel E5 {001} Twist Grain Boundary

International Nuclear Information System (INIS)

Young, G.A. Jr.; Najafabadi, R.; Strohmayer, W.; Baldrey, D.G.; Hamm, B.; Harris, J.; Sticht, J.; Wimmer, E.

2003-01-01

Atomistic modeling methods were employed to investigate the effects of impurity elements on the metallurgy, irradiation embrittlement, and environmentally assisted cracking of nickel-base alloys exposed to nuclear environments. Calculations were performed via ab initio atomistic modeling methods to ensure the accuracy and reliability of the results. A Griffith-type fracture criterion was used to quantitatively assess the effect of elements or element pairs on the grain boundary cohesive strength. In order of most embrittling to most strengthening, the elements are ranked as: He, Li, S, H, C, Zr, P, Fe, Mn, Nb, Cr, and B. Helium is strongly embrittling (-2.04 eV/atom lowering of the Griffith energy), phosphorus has little effect on the grain boundary (0.1 eV/atom), and boron offers appreciable strengthening (1.03 eV/atom increase in the Griffith energy). Calculations for pairs of elements (H-Li, H-B, H-C, H-P, and H-S) show little interaction on the grain boundary cohesive energy, so that for the conditions studied, linear superposition of elemental effects is a good approximation. These calculations help explain metallurgical effects (e.g. why boron can strengthen grain boundaries), irradiation embrittlement (e.g. how boron transmutation results in grain boundary embrittlement), as well as how grain boundary impurity elements can affect environmentally assisted cracking (i.e. low temperature crack propagation and stress corrosion cracking) of nickel-base alloys

Hydrogen gas embrittlement of stainless steels mainly austenitic steels. Volumes 1 and 2

International Nuclear Information System (INIS)

Azou, P.

1988-01-01

Steel behavior in regard to hydrogen is examined especially austenitic steels. Gamma steels are studied particularly the series 300 with various stabilities and gamma steels with improved elasticity limit for intermetallic phase precipitation and nitrogen additions. A two-phase structure γ + α' is also studied. All the samples are tested for mechanical behavior in gaseous hydrogen. Influence of metallurgical effects and of testing conditions on hydrogen embrittlement are evidenced. Microstructure resulting from mechanical or heat treatments, dislocation motion during plastic deformation and influence of deformation rate are studied in detail [fr

Microstructure and tensile properties of neutron-irradiated (FE061Ni039)3V ordered alloy

International Nuclear Information System (INIS)

Braski, D.N.

1982-01-01

Small tensile specimens of the (Fe 0 61 Ni 0 39 ) 3 V long-range-ordered alloy were irradiated in the ORR to 4 dpa at 523, 623, and 823 K and subsequently tested at the same respective temperatueres. The alloy remained ordered after irradiation at all three temperatures. Irradiation at 523 and 623 K increased the yield strength of the material by producing Frank loops in the microstructure and reduced the total elongation. The low strain hardening observed was attributed to planar slip and the absence of cross slip. Irradiation at 823 K embrittled the alloy. Premature failure was apparently initiated by helium bubbles on sigma phase boundaries which grew rapidly during the test to form microcracks. Fracture occurred after a microcrack propagated across grain boundaries that were weakened by helium and possible sulfur. New LRO alloys without sigma phase should perform better under neutron irradiation

Localization of electromagnetic field on the “Brouwer-island” and liquid metal embrittlement

Energy Technology Data Exchange (ETDEWEB)

Maksimenko, V.V.; Zagaynov, V.A. [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), 31, Kashirskoe shosse, 115409 Moscow (Russian Federation); Karpov Institute of Physical Chemistry, Vorontsovo Pole, 10, 105064 Moscow (Russian Federation); Agranovski, I.E., E-mail: I.Agranovski@griffith.edu.au [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), 31, Kashirskoe shosse, 115409 Moscow (Russian Federation); School of Engineering, Griffith University, Brisbane, 4111 QLD (Australia)

2015-03-01

Liquid metal embrittlement (LME) manifests itself as a sudden destruction of a metal sample if it is covered by a thin liquid film of eutectic mixture of specially selected metals. The proposed theoretical model of this phenomenon is based on an assumption related to the possibility of electromagnetic field localization in folds of interface between the phases or components of eutectic mixture filling cracks in solid metal surface (the typical example is In–Ga eutectic on Al-surface). Based on simultaneous presence of three different components in each space point of eutectic mixture (homogeneous In + Ga melt, solid In, and solid Ga), the system of interface folds could be simulated by the Brouwer surface – well known in topology. This surface separates three different components presented at each of its point. Such fractal surfaces posses by a finite volume. The volume occupied by the surface is defined as a difference between the eutectic mixture volume and the sum of volumes of its components. We investigate localization of external electromagnetic radiation in this system of folds. Due to very large magnitude of effective dielectric permeability of the considered system, at relative small volume change and fractal dimension of interface close to the value 3, the wave length of incident radiation inside the system is considerably decreased and multiscale folds are filled with localized photons. A probability of this process and the life time of the localized photons are calculated. The localized photons play crucial role in destruction of primary cracks in the metal surface. They are capable “to switch of” the Coulomb attraction of charge fluctuations on opposite “banks” of the crack filled with the eutectic. As a result, the crack could break down. - Highlights: • A new theoretical model of liquid metal embrittlement has been developed. • Light localization has a strong influence on liquid metal embrittlement. • Light is localized in folds at

Localization of electromagnetic field on the “Brouwer-island” and liquid metal embrittlement

International Nuclear Information System (INIS)

Maksimenko, V.V.; Zagaynov, V.A.; Agranovski, I.E.

2015-01-01

Liquid metal embrittlement (LME) manifests itself as a sudden destruction of a metal sample if it is covered by a thin liquid film of eutectic mixture of specially selected metals. The proposed theoretical model of this phenomenon is based on an assumption related to the possibility of electromagnetic field localization in folds of interface between the phases or components of eutectic mixture filling cracks in solid metal surface (the typical example is In–Ga eutectic on Al-surface). Based on simultaneous presence of three different components in each space point of eutectic mixture (homogeneous In + Ga melt, solid In, and solid Ga), the system of interface folds could be simulated by the Brouwer surface – well known in topology. This surface separates three different components presented at each of its point. Such fractal surfaces posses by a finite volume. The volume occupied by the surface is defined as a difference between the eutectic mixture volume and the sum of volumes of its components. We investigate localization of external electromagnetic radiation in this system of folds. Due to very large magnitude of effective dielectric permeability of the considered system, at relative small volume change and fractal dimension of interface close to the value 3, the wave length of incident radiation inside the system is considerably decreased and multiscale folds are filled with localized photons. A probability of this process and the life time of the localized photons are calculated. The localized photons play crucial role in destruction of primary cracks in the metal surface. They are capable “to switch of” the Coulomb attraction of charge fluctuations on opposite “banks” of the crack filled with the eutectic. As a result, the crack could break down. - Highlights: • A new theoretical model of liquid metal embrittlement has been developed. • Light localization has a strong influence on liquid metal embrittlement. • Light is localized in folds at

Regulatory aspects of radiation embrittlement of reactor vessel steels

International Nuclear Information System (INIS)

Randall, P.N.

1979-01-01

One purpose of this conference, is to re-examine the conventional wisdom about neutron radiation embrittlement and the methods used to counteract embrittlement in reactor vessels. Perhaps, there have been sufficient advances in fracture mechanics, core physics, dosimetry, and physical metallurgy to permit a forward step in the quantitative treatment of the subject. Certainly this would be consistent with the position of the U.S. Nuclear Regulatory Commission (the NRC) in general. ''There has been a continued evolution toward increased specificity.'' This statement appeared in the response prepared by the staff to a request from the Commission to explain how the staff decides to apply a new requirement and to whom, i.e., to back-fit or forward-fit-only or whatever. Pressure for increased specificity, i.e., for fleshing out general design criteria, comes from ''technical surprises'' in the form of operating experiences or from research information, and from attempts to improve our confidence in the safety of plants, especially new plants. Our goal is to have anticipated and evaluated all possible modes of failure with sufficient quantitativeness that the probability of failure can be estimated with some accuracy. Failing this, regulators demand large margins of safety to cover our ignorance

Postirradiation examination results for the irradiation effects scoping test 1

Energy Technology Data Exchange (ETDEWEB)

Mehner, A.S.; Quapp, W.J.; Goetzmann, O.; Hobbins, R.R.

1976-09-01

A zircaloy-clad UO/sub 2/ fuel rod was operated above its critical heat flux within the in-pile test loop of the Power Burst Facility and later examined in the hot cells. The results of the postirradiation examinations are presented in this report. A Zr-UO/sub 2/ reaction at the fuel-cladding interface embrittled nearly as much of the cladding wall thickness as the Zr-water reaction on the exterior. Data on both the internal and external reactions, and cladding and fuel microstructures, are presented. Cladding embrittlement and rod failure are compared with several rod fragmentation criteria, and conclusions concerning fuel rod failure propagation in a power reactor system are made.

Analysis of WWER-440 and PWR RPV welds surveillance data to compare irradiation damage evolution

Energy Technology Data Exchange (ETDEWEB)

Debarberis, L. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands)]. E-mail: luigi.debarberis@cec.eu.int; Acosta, B. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands)]. E-mail: beatriz.acosta-iborra@jrc.nl; Zeman, A. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands); Sevini, F. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands); Ballesteros, A. [Tecnatom, Avd. Montes de Oca 1, San Sebasitan de los Reyes, E-28709 Madrid (Spain); Kryukov, A. [Russian Research Centre Kurchatov Institute, Kurchatov Square 1, 123182 Moscow (Russian Federation); Gillemot, F. [AEKI Atomic Research Institute, Konkoly Thege M. ut 29-33, 1121 Budapest (Hungary); Brumovsky, M. [NRI, Nuclear Research Institute, Husinec-Rez 130, 25068 Rez (Czech Republic)

2006-04-15

It is known that for Russian-type and Western water reactor pressure vessel steels there is a similar degradation in mechanical properties during equivalent neutron irradiation. Available surveillance results from WWER and PWR vessels are used in this article to compare irradiation damage evolution for the different reactor pressure vessel welds. The analysis is done through the semi-mechanistic model for radiation embrittlement developed by JRC-IE. Consistency analysis with BWR vessel materials and model alloys has also been performed within this study. Globally the two families of studied materials follow similar trends regarding the evolution of irradiation damage. Moreover in the high fluence range typical of operation of WWER the radiation stability of these vessels is greater than the foreseen one for PWR.

Fractography of hydrogen-embrittled iron-chromium-nickel alloys

International Nuclear Information System (INIS)

Caskey, G.R. Jr.

1980-01-01

Tensile specimens of iron-chromium-nickel base alloys were broken in either a hydrogen environment or in air following thermal charging with hydrogen. Fracture surfaces were examined by scanning electron microscopy. Fracture morphology of hydrogen-embrittled specimens was characterized by: changed dimple size, twin-boundary parting, transgranular cleavage, and intergranular separation. The nature and extent of the fracture mode changes induced by hydrogen varied systematically with alloy composition and test temperature. Initial microstructure developed during deformation processing and heat treating had a secondary influence on fracture mode

Accelerated aging embrittlement of cast duplex stainless steel: Activation energy for extrapolation

International Nuclear Information System (INIS)

Chung, H.M.; Chopra, O.K.

1989-05-01

Cast duplex stainless steels, used extensively in LWR systems for primary pressure boundary components such as primary coolant pipes, valves, and pumps, are susceptible to thermal aging embrittlement at reactor operating or higher temperatures. Since a realistic aging embrittlement for end-of-life or life-extension conditions (i.e., 32--50 yr of aging at 280--320 degree C) cannot be produced, it is customary to simulate the metallurgical structure by accelerated aging at ∼400 degree C. Over the past several years, extensive data on accelerated aging have been reported from a number of laboratories. The most important information from these studies is the activation energy, namely, the temperature dependence of the aging kinetics between 280 and 400 degree C, which is used to extrapolate the aging characteristics to reactor operating conditions. The activation energies (in the range of 18--50 kcal/mole) are, in general, sensitive to material grade, chemical composition, and fabrication process, and a few empirical correlations, obtained as a function of bulk chemical composition, have been reported. In this paper, a mechanistic understanding of the activation energy is described on the basis of the results of microstructural characterization of various heats of CF-3, -8, and -8M grades that were used in aging studies at different laboratories. The primary mechanism of aging embrittlement at temperatures between 280 and 400 degree C is the spinodal decomposition of the ferrite phase, and M 23 C 6 carbide precipitation on the ferrite/austenite boundaries is the secondary mechanism for high-carbon CF-8 grade. 20 refs., 10 figs., 3 tabs

Hydrogen embrittlement and stress corrosion cracking in metals

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Suk; Cheong, Yong Mu; Im, Kyung Soo

2004-10-15

The objective of this report is to elucidate the mechanism for hydrogen embrittlement (HE) and stress corrosion cracking (SCC) in metals. To this end, we investigate the common features between delayed hydride cracking (DHC) in zirconium alloys and HE in metals with no precipitation of hydrides including Fe base alloys, Nickel base alloys, Cu alloys and Al alloys. Surprisingly, as with the crack growth pattern for the DHC in zirconium alloy, the metals mentioned above show a discontinuous crack growth, striation lines and a strong dependence of yield strength when exposed to hydrogen internally and externally. This study, for the first time, analyzes the driving force for the HE in metals in viewpoints of Kim's DHC model that a driving force for the DHC in zirconium alloys is a supersaturated hydrogen concentration coming from a hysteresis of the terminal solid solubility of hydrogen, not by the stress gradient, As with the crack growing only along the hydride habit plane during the DHC in zirconium alloys, the metals exposed to hydrogen seem to have the crack growing by invoking the dislocation slip along the preferential planes as a result of some interactions of the dislocations with hydrogen. Therefore, it seems that the hydrogen plays a role in inducing the slip only on the preferential planes so as to cause a strain localization at the crack tip. Sulfur in metals is detrimental in causing a intergranular cracking due to a segregation of the hydrogens at the grain boundaries. In contrast, boron in excess of 500 ppm added to the Ni3Al intermetallic compound is found to be beneficial in suppressing the HE even though further details of the mechanism for the roles of boron and sulfur are required. Carbon, carbides precipitating semi-continuously along the grain boundaries and the CSL (coherent site lattice) boundaries is found to suppress the intergranular stress corrosion cracking (IGSCC) in Alloy 600. The higher the volume fraction of twin boundaries, the

Hydrogen embrittlement and stress corrosion cracking in metals

International Nuclear Information System (INIS)

Kim, Young Suk; Cheong, Yong Mu; Im, Kyung Soo

2004-10-01

The objective of this report is to elucidate the mechanism for hydrogen embrittlement (HE) and stress corrosion cracking (SCC) in metals. To this end, we investigate the common features between delayed hydride cracking (DHC) in zirconium alloys and HE in metals with no precipitation of hydrides including Fe base alloys, Nickel base alloys, Cu alloys and Al alloys. Surprisingly, as with the crack growth pattern for the DHC in zirconium alloy, the metals mentioned above show a discontinuous crack growth, striation lines and a strong dependence of yield strength when exposed to hydrogen internally and externally. This study, for the first time, analyzes the driving force for the HE in metals in viewpoints of Kim's DHC model that a driving force for the DHC in zirconium alloys is a supersaturated hydrogen concentration coming from a hysteresis of the terminal solid solubility of hydrogen, not by the stress gradient, As with the crack growing only along the hydride habit plane during the DHC in zirconium alloys, the metals exposed to hydrogen seem to have the crack growing by invoking the dislocation slip along the preferential planes as a result of some interactions of the dislocations with hydrogen. Therefore, it seems that the hydrogen plays a role in inducing the slip only on the preferential planes so as to cause a strain localization at the crack tip. Sulfur in metals is detrimental in causing a intergranular cracking due to a segregation of the hydrogens at the grain boundaries. In contrast, boron in excess of 500 ppm added to the Ni3Al intermetallic compound is found to be beneficial in suppressing the HE even though further details of the mechanism for the roles of boron and sulfur are required. Carbon, carbides precipitating semi-continuously along the grain boundaries and the CSL (coherent site lattice) boundaries is found to suppress the intergranular stress corrosion cracking (IGSCC) in Alloy 600. The higher the volume fraction of twin boundaries, the more

Gamma-radiation effect on diamond and steel during their irradiation in WWER type reactors

International Nuclear Information System (INIS)

Nikolaenko, V.A.; Karpukhin, V.I.; Amaev, A.D.; Vikhrov, V.I.; Korolev, Yu.N.; Krasikov, E.A.

1996-01-01

A study is made into the influence of reactor gamma radiation on expansion of crystal lattice in diamond. The data obtained are compared to those on radiation embrittlement of reactor vessel steels. The necessity of taking into consideration gamma radiation effects on WWER reactor vessel radiation resistance during long-term operation is shown [ru

Origin of intergranular embrittlement of Al alloys induced by Na and Ca segregation: Grain boundary weakening

International Nuclear Information System (INIS)

Lu Guanghong; Zhang Ying; Deng Shenghua; Wang Tianmin; Kohyama, Masanori; Yamamoto, Ryoichi; Liu Feng; Horikawa, Keitaro; Kanno, Motohiro

2006-01-01

Using a first-principles computational tensile test, we show that the ideal tensile strength of an Al grain boundary (GB) is reduced with both Na and Ca GB segregation. We demonstrate that the fracture occurs in the GB interface, dominated by the break of the interfacial bonds. Experimentally, we further show that the presence of Na or Ca impurity, which causes intergranular fracture, reduces the ultimate tensile strength when embrittlement occurs. These results suggest that the Na/Ca-induced intergranular embrittlement of an Al alloy originates mainly from the GB weakening due to the Na/Ca segregation

Feasibility of and methodology for thermal annealing an embrittled reactor vessel. Volume 2. Detailed technical description of the work. Final report

International Nuclear Information System (INIS)

Mager, T.R.

1982-11-01

Program materials were three weldments fabricated from A533 Grade B class 1 plate material and Mn Mo Ni weld wire. Specimens fabricated from the three submerged arc weldments included Type A Charpy V-notch impact, small size tensile, and 1/2T compact tension specimens. After encapsulation, the specimens were irradiated at the UVAR to two fluence levels, 8 x 10 18 n/cm 2 and 1.5 x 10 19 n/cm 2 (E > 1 MeV). Specimens were subjected to sequences of irradiation and anneals and then tested. Metallurgial/mechanistic analyses were also performed. It was concluded that excellent recovery of all properties could be achieved by annealing at greater than or equal to 850 0 F (454 0 C) for 168 hours. Such an annealing resulted in ductile-brittle transition temperature shift recovery of 80 to 100%, and reirradiation after this annealing indicated that the ductile-brittle transition temperature shift appears to continue at the expected rate. Several drawbacks were identified for wet thermal annealing. A conceptual dry in-situ thermal annealing procedure was developed for thermal annealing embrittled reactor vessels

Hydrogen embrittlement of metals. A bibliography with abstracts. Search period covered: 1964--August 1975

International Nuclear Information System (INIS)

Smith, M.F.

1975-10-01

The research covers the hydrogen embrittlement of both ferrous and nonferrous metals and alloys and includes nuclear technology, aircraft metallurgy, mechanical properties, testing, electroplating, fatigue, corrosion and fracture. Contains 230 abstracts

Gas diffusion and temperature dependence of bubble nucleation during irradiation

DEFF Research Database (Denmark)

Foreman, A. J. E.; Singh, Bachu Narain

1986-01-01

The continuous production of gases at relatively high rates under fusion irradiation conditions may enhance the nucleation of cavities. This can cause dimensional changes and could induce embrittlement arising from gas accumulation on grain boundaries. Computer calculations have been made...... of the diatomic nucleation of helium bubbles, assuming helium to diffuse substitutionally, with radiation-enhanced diffusion at lower temperatures. The calculated temperature dependence of the bubble density shows excellent agreement with that observed in 600 MeV proton irradiations, including a reduction...... in activation energy below Tm/2. The coalescence of diatomic nuclei due to Brownian motion markedly improves the agreement and also provides a well-defined terminal density. Bubble nucleation by this mechanism is sufficiently fast to inhibit any appreciable initial loss of gas to grain boundaries during...

Multiscale Modeling of Grain Boundary Segregation and Embrittlement in Tungsten for Mechanistic Design of Alloys for Coal Fired Plants

Energy Technology Data Exchange (ETDEWEB)

Luo, Jian; Tomar, Vikas; Zhou, Naixie; Lee, Hongsuk

2013-06-30

Based on a recent discovery of premelting-like grain boundary segregation in refractory metals occurring at high temperatures and/or high alloying levels, this project investigated grain boundary segregation and embrittlement in tungsten (W) based alloys. Specifically, new interfacial thermodynamic models have been developed and quantified to predict high-temperature grain boundary segregation in the W-Ni binary alloy and W-Ni-Fe, W-Ni-Ti, W-Ni-Co, W-Ni-Cr, W-Ni-Zr and W-Ni-Nb ternary alloys. The thermodynamic modeling results have been experimentally validated for selected systems. Furthermore, multiscale modeling has been conducted at continuum, atomistic and quantum-mechanical levels to link grain boundary segregation with embrittlement. In summary, this 3-year project has successfully developed a theoretical framework in combination with a multiscale modeling strategy for predicting grain boundary segregation and embrittlement in W based alloys.

Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in BWR Environments

International Nuclear Information System (INIS)

Chen, Y.; Chopra, O. K.; Gruber, Eugene E.; Shack, William J.

2010-01-01

The internal components of light water reactors are exposed to high-energy neutron irradiation and high-temperature reactor coolant. The exposure to neutron irradiation increases the susceptibility of austenitic stainless steels (SSs) to stress corrosion cracking (SCC) because of the elevated corrosion potential of the reactor coolant and the introduction of new embrittlement mechanisms through radiation damage. Various nonsensitized SSs and nickel alloys have been found to be prone to intergranular cracking after extended neutron exposure. Such cracks have been seen in a number of internal components in boiling water reactors (BWRs). The elevated susceptibility to SCC in irradiated materials, commonly referred to as irradiation-assisted stress corrosion cracking (IASCC), is a complex phenomenon that involves simultaneous actions of irradiation, stress, and corrosion. In recent years, as nuclear power plants have aged and irradiation dose increased, IASCC has become an increasingly important issue. Post-irradiation crack growth rate and fracture toughness tests have been performed to provide data and technical support for the NRC to address various issues related to aging degradation of reactor-core internal structures and components. This report summarizes the results of the last group of tests on compact tension specimens from the Halden-II irradiation. The IASCC susceptibility of austenitic SSs and heat-affected-zone (HAZ) materials sectioned from submerged arc and shielded metal arc welds was evaluated by conducting crack growth rate and fracture toughness tests in a simulated BWR environment. The fracture and cracking behavior of HAZ materials, thermally sensitized SSs and grain-boundary engineered SSs was investigated at several doses (3 dpa). These latest results were combined with previous results from Halden-I and II irradiations to analyze the effects of neutron dose, water chemistry, alloy compositions, and welding and processing conditions on IASCC. The

Effect of irradiation on carbohydrates content

International Nuclear Information System (INIS)

Chantharasakul, S.

1971-01-01

Effect of gamma radiation on vitamin C and total acidity contents of Hom Tong banana was described. There was a slight decrease in vitamin C contents in both irradiated and non-irradiated banana during storage. No difference was detected in term of vitamin C contents between irradiated and non-irradiated banana at any storage time. The total acidity of the banana increased with increasing time of storage owing to the ripening effect of the fruit. Higher total acidity content of non-irradiated banana during storage indicated the faster rate of ripening of the fruit

Effects of bonding bakeout thermal cycles on pre- and post irradiation microstructures, physical, and mechanical properties of copper alloys

Energy Technology Data Exchange (ETDEWEB)

Singh, B.N.; Eldrup, M.; Toft, P.; Edwards, D.J. [Pacific Northwest National Laboratory, Richland, WA (United States)

1996-10-01

At present, dispersion strengthened (DS) copper is being considered as the primary candidate material for the ITER first wall and divertor components. Recently, it was agreed among the ITER parties that a backup alloy should be selected from the two well known precipitation hardened copper alloys, CuCrZr and CuNiBe. It was therefore decided to carry out screening experiments to simulate the effect of bonding and bakeout thermal cycles on microstructure, mechanical properties, and electrical resistivity of CuCrZr and CuNiBe alloys. On the basis of the results of these experiments, one of the two alloys will be selected as a backup material. Tensile specimens of CuCrZr and CuNiBe alloys were given various heat treatments corresponding to solution anneal, prime ageing, and bonding thermal cycle followed by reageing and the reactor bakeout treatment at 623K for 100 hours. Tensile specimens of the DS copper were also given the heat treatment corresponding to the bonding thermal cycle. A number of these heat treated specimens of CuCrZr, CuNiBe, and DS copper were neutron irradiated at 523K to a dose level of {approx}0.3 dpa (NRT) in the DR-3 reactor at Riso. Both unirradiated and irradiated specimens with the various heat treatments were tensile tested at 532K. The dislocation, precipitate and void microstructures and electrical resistivity of these specimens were also determined. Results of these investigations will be reported and discussed in terms of thermal and irradiation stability of precipitates and irradiation-induced precipitation and recovery of dislocation microstructure. Results show that the bonding and bakeout thermal cycles are not likely to have any serious deleterious effects on the performance of these alloys. The CuNiBe alloys were found to be susceptible to radiation-induced embrittlement, however, the exact mechanism is not yet known. It is thought that radiation-induced precipitation and segregation of the beryllium may be responsible.

The Role of Hydrogen-Enhanced Strain-Induced Lattice Defects on Hydrogen Embrittlement Susceptibility of X80 Pipeline Steel

Science.gov (United States)

Hattori, M.; Suzuki, H.; Seko, Y.; Takai, K.

2017-08-01

Studies to date have not completely determined the factors influencing hydrogen embrittlement of ferrite/bainite X80 pipeline steel. Hydrogen embrittlement susceptibility was evaluated based on fracture strain in tensile testing. We conducted a thermal desorption analysis to measure the amount of tracer hydrogen corresponding to that of lattice defects. Hydrogen embrittlement susceptibility and the amount of tracer hydrogen significantly increased with decreasing crosshead speed. Additionally, a significant increase in the formation of hydrogen-enhanced strain-induced lattice defects was observed immediately before the final fracture. In contrast to hydrogen-free specimens, the fracture surface of the hydrogen-charged specimens exhibited shallower dimples without nuclei, such as secondary phase particles. These findings indicate that the presence of hydrogen enhanced the formation of lattice defects, particularly just prior to the occurrence of final fracture. This in turn enhanced the formation of shallower dimples, thereby potentially causing premature fracture of X80 pipeline steel at lower crosshead speeds.

Present status of the disk pressure tests for hydrogen embrittlements

International Nuclear Information System (INIS)

Fidelle, J.P.

1988-01-01

The Disk Pressure Tests (DPT) have been developed considerably. Theoretically: a finite elements mechanical analysis shows the build up of a triaxial stress state already at the beginning of the test, which, with other reasons accounts for the very high sensitivity. Experimentally: for Internal Hydrogen Embrittlement (IHE) e.g. Co, Ti, U alloys, for environment embrittlement due to H 2 hydrogenated media such as water vapor, alcohol, machining fluids or liquid NH 3 . The range has been expanded considerably: up to 300 MPa and up to 1000 0 C. Very low strain rate - longer than a month - tests have been able to evidence HGE; of FCC alloys where H diffusivity is low for very oxidation -sensitive metals such as Nb and Ta, effects may appear only at somewhat high rates. The relationship between dynamic tests, static and low-cycle fatigue tests has been determined. In a number of instances, including SCC, other techniques and even fracture mechanics have been compared to the DPT and proved at best equivalent and several times, less sensitive than a well conducted DPT. At extreme they could not reproduce the field service phenomenon whereas the DPT did and could also be applied satisfactorily to low yield stress materials. The main rupture aspects have been analysed mechanically and organized in a rational and comprehensive chart based on 12,000 + tests over 15O + materials in different conditions. Comparison of HGE tests and service behaviour of a large variety of materials and industrial equipments has made possible to specify acceptance criteria for industrial service, which, provided the shape of the stress strain curves is not significantly affected, can be expanded to IHE, HE by other fluids than H 2 , 36 refs

Practical implications for RPV irradiation surveillance under long term operation based on latest research results

International Nuclear Information System (INIS)

Hein, H.; Keim, E.; Barthelmes, J.; Schnabel, H.

2015-01-01

The international programs CARISMA, CARINA and LONGLIFE belong to the research programs which have been performed during the last 10 years to study the irradiation behavior of RPV steels under long term operation of more than 60 years. Some characteristic but different irradiated RPV steels used in Pressurized Water Reactors have been extensively investigated in each of those three programs. Whereas the CARISMA and CARINA programs were mainly focused on material testing to study the irradiation-induced change of material properties in terms of fracture toughness, the main objective of LONGLIFE was to investigate the change of microstructure with various analysis techniques and to understand the mechanisms behind. In this way it was possible to get a comprehensive material characterization in terms of macro-physical properties and micro-structural features for a number of RPV steels which have been studied at different irradiation levels up to 8*10 19 cm -2 (E > 1 MeV). The essential macro-physical and micro-structural results are summarized, in particular regarding the impact of copper and nickel, and the neutron flux on the irradiation behavior and with respect to possible late irradiation effects under long term operation. Moreover, the change of material properties is linked with embrittlement mechanisms such as formation of element specific precipitations, segregations, and matrix defects. Well-known trend curves are also applied to the measured T 41 and T 0 data in order to assess their appropriateness for long term operation. Based on the comprehensive available data base, practical implications for RPV irradiation surveillance programs under long term operation are highlighted with respect to issues like material specific application of reference temperature concepts, data scattering, prediction of high fluence behavior and how to cope with possible late irradiation effects. Finally, best practices for RPV irradiation surveillance programs are suggested from

Anomalous fracture toughness of irradiated Cr-MoV - Reactor pressure vessel steel

Energy Technology Data Exchange (ETDEWEB)

Ahistrand, R [Imatran Voima Oy (IVO), Helsinki (Finland)

1994-12-31

The base metal Crack Opening Displacement (COD) specimens of the irradiation-induced embrittlement surveillance programme in Loviisa 1 revealed an anomalous behaviour of K{sub JC} compared to the Charpy-V results and to expected results according to standards: about 20% of the COD specimens showed an exceptionally low fracture toughness. Abnormal test specimens were analyzed through fractography, metallography and repeated tests using reconstitution technique: the anomalous behaviour appears to be caused by incorrect pre-fatigue cracking of base metal COD specimens. 7 refs., 9 figs.

A novel self-embrittling strippable coating for radioactive decontamination based on silicone modified styrene-acrylic emulsion

Science.gov (United States)

Wang, Jing; Wang, Jianhui; Zheng, Li; Li, Jian; Cui, Can; Lv, Linmei

2017-03-01

Silicone modified styrene-acrylic emulsion and butyl acrylate were used as a main film-forming agent and an additive respectively to synthesize a self-embrittling strippable coating. The doping mass-ratio of butyl acrylate was adjusted at 0, 5%, 10%, 15%, 20%, and the results indicated the optimized doping ratio was 10%. Ca(OH)2 was used to promote the coating film self-embrittling at a moderate doping mass-ratio of 20%. The synthesized coating’s coefficients of α and β decontamination on concrete, marble, glass and stainless steel surfaces were both greater than 85%, which indicated the synthesized coating is a promising cleaner for radioactive decontamination.

Calculational results for radiation embrittlement of WWER pressure vessel at the Kozloduy NPP

Energy Technology Data Exchange (ETDEWEB)

Apostolov, T; Ilieva, K; Petrova, T [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika

1996-12-31

Determination of radiation impact on metal state in the case of WWER-440/230 is made only by calculation methods since a special sample-witness (SW) incorporation had not been implemented. In WWER-1000 reactors such SW are foreseen but their spots are high above the active core. This is why in both reactors the appliance of a calculational procedure for radiation embrittlement determination is compulsory. The authors propose such a procedure accounting for the change in critical temperature of neutron brittleness by the neutron fluence. The neutron fluence and the shift of critical embrittlement temperature have been calculated for the maximum overloaded location and for the weld metal of the Kozloduy-5 and Kozloduy-6 reactors (WWER-1000). The shift of critical temperature in weld 4 of the Units 1-4 (WWER-440) is plotted versus work cycles and compared to experimental values. 4 figs., 5 tabs.

Microstructure and Mechanical Properties of n-irradiated Fe-Cr Model Alloys

International Nuclear Information System (INIS)

Matijasevic, Milena; Al Mazouzi, Abderrahim

2008-01-01

High chromium ( 9-12 wt %) ferritic/martensitic steels are candidate structural materials for future fusion reactors and other advanced systems such as accelerator driven systems (ADS). Their use for these applications requires a careful assessment of their mechanical stability under high energy neutron irradiation and in aggressive environments. In particular, the Cr concentration has been shown to be a key parameter to be optimized in order to guarantee the best corrosion and swelling resistance, together with the least embrittlement. In this work, the characterization of the neutron irradiated Fe-Cr model alloys with different Cr % with respect to microstructure and mechanical tests will be presented. The behavior of Fe-Cr alloys have been studied using tensile tests at different temperature range ( from -160 deg. C to 300 deg. C). Irradiation-induced microstructure changes have been studied by TEM for two different irradiation doses at 300 deg. C. The density and the size distribution of the defects induced have been determined. The tensile test results indicate that Cr content affects the hardening behavior of Fe-Cr binary alloys. Hardening mechanisms are discussed in terms of Orowan type of approach by correlating TEM data to the measured irradiation hardening. (authors)

FP7 Project LONGLIFE: Overview of results and implications

International Nuclear Information System (INIS)

Altstadt, Eberhard; Keim, Elisabeth; Hein, Hieronymus; Serrano, Marta; Bergner, Frank; Viehrig, Hans-Werner; Ballesteros, Antonio; Chaouadi, Rachid; Wilford, Keith

2014-01-01

Highlights: • Radiation effects in reactor pressure vessel steels under long term operation. • Indications of late blooming effects were found in some cases. • Significant flux effect on the size of defect clusters in high-Cu weld materials. • Guideline for monitoring radiation embrittlement during life extension. - Abstract: LONGLIFE (“Treatment of long term irradiation embrittlement effects in RPV safety assessment”) was a collaborative project of the 7th Framework Programme of EURATOM under the umbrella of NULIFE/NUGENIA, aiming at an improved understanding of irradiation effects in reactor pressure vessel steels under conditions representative of long term operation. The LONGLIFE project was completed by the end of January 2014. The paper gives an overview of the main project results and their implications for future research, as discussed at the final project workshop. The microstructural database for neutron-irradiated RPV steels was extended considerably and existing gaps on mechanical property data were closed. Indications of late blooming effects (LBE) were found in some cases, but clear criteria for the occurrence/exclusion in terms of irradiation conditions and chemical composition have still to be developed. The commonly accepted trend, that low flux and low irradiation temperature promotes LBE, is supported. A significant flux effect on the size of defect clusters was observed in two high Cu weld materials, while the changes of mechanical properties are not affected by the neutron flux. The database requires completion in particular for low-Cu RPV steels. The shift of reference temperature T 0 over the thickness location of a VVER-440 welding seam does not follow the prediction Russian code, because of the strong variation of the intrinsic weld bead structure. Therefore, the effect of the initial microstructure and of the heterogeneity on the radiation behaviour has to be addressed in future works. Existing embrittlement trend curves models

Fabrication of poly(methyl methacrylate)-block-poly(methacrylic acid) diblock copolymer as a self-embrittling strippable coating for radioactive decontamination

International Nuclear Information System (INIS)

Liu Renlong; Zhang Huiyan; Li Yintao; Zhou Yuanlin; Zhang Quanping; Zheng Jian; Wang Shanqiang

2016-01-01

The poly(methyl methacrylate)-block-poly(methacrylic acid) diblock copolymer with different monomer compositions was synthesized via reversible addition-fragmentation chain transfer polymerization. Meanwhile, a novel self-embrittling strippable coating was prepared using the diblock copolymers, which is proposed to be used as radioactive decontamination agents without manual operation. Furthermore, the decontamination efficiencies of self-embrittling strippable coatings for radioactive contamination on glass, marble, and stainless steel surfaces were studied. (author)

Imitation and reactor studies of irradiation effect on material mechanic properties

International Nuclear Information System (INIS)

Ozhigov, L.S.

1999-01-01

Processes of low- and high-temperature radiation embrittlement, radiation creeping and their influence on reactor material properties are considered. Role of imitation experiments in these processes is analysed

Investigating liquid-metal embrittlement of T91 steel by fracture toughness tests

Energy Technology Data Exchange (ETDEWEB)

Ersoy, Feyzan, E-mail: fersoy@sckcen.be [SCK-CEN (Belgian Nuclear Research Centre), Boeretang 200, B-2400, Mol (Belgium); Department of Materials Science and Engineering, Ghent University (UGent), Technologiepark 903, B-9052, Ghent (Belgium); Gavrilov, Serguei [SCK-CEN (Belgian Nuclear Research Centre), Boeretang 200, B-2400, Mol (Belgium); Verbeken, Kim [Department of Materials Science and Engineering, Ghent University (UGent), Technologiepark 903, B-9052, Ghent (Belgium)

2016-04-15

Heavy liquid metals such as lead bismuth eutectic (LBE) are chosen as the coolant to innovative Generation IV (Gen IV) reactors where ferritic/martensitic T91 steel is a candidate material for high temperature applications. It is known that LBE has a degrading effect on the mechanical properties of this steel. This degrading effect, which is known as liquid metal embrittlement (LME), has been screened by several tests such as tensile and small punch tests, and was most severe in the temperature range from 300 °C to 425 °C. To meet the design needs, mechanical properties such as fracture toughness should be addressed by corresponding tests. For this reason liquid-metal embrittlement of T91 steel was investigated by fracture toughness tests at 350 °C. Tests were conducted in Ar-5%H{sub 2} and LBE under the same experimental conditions Tests in Ar-5%H{sub 2} were used as reference. The basic procedure in the ASTM E 1820 standard was followed to perform tests and the normalization data reduction (NDR) method was used for the analysis. Comparison of the tests demonstrated that the elastic–plastic fracture toughness (J{sub 1C}) of the material was reduced by a factor in LBE and the fracture mode changed from ductile to quasi-cleavage. It was also shown that the pre-cracking environment played an important role in observing LME of the material since it impacts the contact conditions between LBE and steel at the crack tip. It was demonstrated that when specimens were pre-cracked in air and tested in LBE, wetting of the crack surface by LBE could not be achieved. When specimens were pre-cracked in LBE though, they showed a significant reduction in fracture toughness.