The behaviour of materials in fast reactors

International Nuclear Information System (INIS)

Matthews, J.R.

1977-01-01

Fast neutron damage in fast reactors can limit the life of structural components through the growth voids. The main features of the current theory of point defect production and condensation are surveyed. The role of metallurgical structures and radiation produced extended defects is outlined and used to demonstrate the development of volume swelling and radiation hardening. Mechanisms of radiation creep are described in the context of the preceding treatment of point defect behaviour. Finally, future trends in the field are briefly explored. (author)

In-reactor creep of zirconium-2.5 wt% niobium at 570 K

International Nuclear Information System (INIS)

Coleman, C.E.; Causey, A.R.; Fidleris, V.

1976-01-01

The effect of fast neutron flux at 570 K on the creep rate of specimens of zirconium-2.5 wt% niobium alloy taken from tubes in various metallurgical conditions has been measured using both constant load tensile creep machines and bent-beam stress relaxation. Creep rates calculated from stress relaxation fit on the trend line for the constant load creep data. Between 114 MPa and 450 MPa the creep rate is proportional to neutron flux. The creep rate of specimens from the longitudinal direction is about twice that of specimens from the circumferential direction of a tube. This anisotropy in creep strength is attributed partly to crystallographic texture and partly to deformation substructure. Cold-work is detrimental to in-reactor creep strength; as-extruded material has higher creep strength. In cold-worked material at stresses below 100 MPa the stress exponent, n, is about 1; n gradually increases with stress being about 10 at 525 MPa and about 100 at 660 MPa. In laboratory tests, rupture ductility correlates inversely with n; the lower n the higher the ductility. In-reactor tests support this correlation thus pressure tubes in CANDU reactors, operating at 117 MPa where n approximately 1, should have good ductility. (Auth.)

Creep buckling problems in fast reactor components

International Nuclear Information System (INIS)

Ramesh, R.; Damodaran, S.P.; Chellapandi, P.; Chetal, S.C.; Bhoje, S.B.

1995-01-01

Creep buckling analyses for two important components of 500 M We Prototype Fast Breeder Reactor (PFBR), viz. Intermediate Heat Exchanger (IHX) and Inner Vessel (IV), are reported. The INCA code of CASTEM system is used for the large displacement elasto-plastic-creep analysis of IHX shell. As a first step, INCA is validated for a typical benchmark problem dealing with the creep buckling of a tube under external pressure. Prediction of INCA is also compared with the results obtained using Hoff's theory. For IV, considering the prohibitively high computational cost for the actual analysis, a simplified analysis which involves only large displacement elastoplastic buckling analysis is performed using isochronous stress strain curve approach. From both of these analysis is performed using isochronous stress strain curve approach. From both of these analysis, it has been inferred that creep buckling failure mode is not of great concern in the design of PFBR components. It has also been concluded from the analysis that Creep Cross Over Curve given in RCC-MR is applicable for creep buckling failure mode also. (author). 8 refs., 9 figs., 1 tab

Effects of composition on the in-reactor creep of AISI 316

International Nuclear Information System (INIS)

Bates, J.F.; Gilbert, E.R.

1979-08-01

In-reactor tests designed to provide information on the relationship between compositional variations and irradiation-induced swelling and creep have achieved an exposure of 4.6 x 10 22 n/cm 2 (E > 0.1 MeV) at 450 0 C. Postirradiation diametral measurements of pressurized tube specimens have indicated that irradiation-induced creep of 316 stainless steel can be modified by compositional variations of minor alloying elements. There is a general trend for specimens with higher swelling to exhibit higher creep. Silicon, phosphorus and molybdenum all retard in-reactor creep and inhibit irradiation-induced swelling as well. However, the relationship between creep and swelling is strongly composition dependent. The data suggest that carbon and nitrogen act synergistically the major influence being the nitrogen concentration. The irradiation-induced creep is insensitive to cobalt variations to the fluences investigated

Creep behaviour of porous metal supports for solid oxide fuel cells

DEFF Research Database (Denmark)

Boccaccini, Dino; Frandsen, Henrik Lund; Sudireddy, Bhaskar Reddy

2014-01-01

The creep behaviour of porous ironechromium alloy used as solid oxide fuel cell support was investigated, and the creep parameters are compared with those of dense strips of similar composition under different testing conditions. The creep parameters were determined using a thermo......-mechanical analyser with applied stresses in the range from 1 to 15 MPa and temperatures between 650 and 800 _C. The GibsoneAshby and Mueller models developed for uniaxial creep of open-cell foams were used to analyse the results. The influence of scale formation on creep behaviour was assessed by comparing the creep...... data for the samples tested in reducing and oxidising atmospheres. The influence of preoxidation on creep behaviour was also investigated. In-situ oxidation during creep experiments increases the strain rate while pre-oxidation of samples reduces it. Debonding of scales at high stress regime plays...

Study of creep behaviour in P-doped copper with slow strain rate tensile tests

International Nuclear Information System (INIS)

Xuexing Yao; Sandstroem, Rolf

2000-08-01

Pure copper with addition of phosphorous is planned to be used to construct the canisters for spent nuclear fuel. The copper canisters can be exposed to a creep deformation up to 2-4% at temperatures in services. The ordinary creep strain tests with dead weight loading are generally employed to study the creep behaviour; however, it is reported that an initial plastic deformation of 5-15% takes place when loading the creep specimens at lower temperatures. The slow strain rate tensile test is an alternative to study creep deformation behaviour of materials. Ordinary creep test and slow strain rate tensile test can give the same information in the secondary creep stage. The advantage of the tensile test is that the starting phase is much more controlled than in a creep test. In a tensile test the initial deformation behaviour can be determined and the initial strain of less than 5% can be modelled. In this study slow strain rate tensile tests at strain rate of 10 -4 , 10 -5 , 10 -6 , and 10 -7 /s at 75, 125 and 175 degrees C have been performed on P-doped pure Cu to supplement creep data from conventional creep tests. The deformation behaviour has successfully been modelled. It is shown that the slow strain rate tensile tests can be implemented to study the creep deformation behaviours of pure Cu

Nonlinear Subincremental Method for Determination of Elastic-Plastic-Creep Behaviour

DEFF Research Database (Denmark)

Ottosen, N. Saabye; Gunneskov, O.

1985-01-01

to general elastic-plastic-creep behaviour including problems with a highly nonlinear total strain path caused by the occurrence of creep hardening. This nonlinear method degenerates to the linear approach for elastic-plastic behaviour and when secondary creep is present. It is also linear during step......The frequently used subincremental method has so far been used on a linear interpolation of the total strain path within each main step. This method has proven successful when elastic-plastic behaviour and secondary creep is involved. The authors propose a nonlinear subincremental method applicable...

A regression approach for Zircaloy-2 in-reactor creep constitutive equations

International Nuclear Information System (INIS)

Yung Liu, Y.; Bement, A.L.

1977-01-01

In this paper the methodology of multiple regressions as applied to Zircaloy-2 in-reactor creep data analysis and construction of constitutive equation are illustrated. While the resulting constitutive equation can be used in creep analysis of in-reactor Zircaloy structural components, the methodology itself is entirely general and can be applied to any creep data analysis. The promising aspects of multiple regression creep data analysis are briefly outlined as follows: (1) When there are more than one variable involved, there is no need to make the assumption that each variable affects the response independently. No separate normalizations are required either and the estimation of parameters is obtained by solving many simultaneous equations. The number of simultaneous equations is equal to the number of data sets. (2) Regression statistics such as R 2 - and F-statistics provide measures of the significance of regression creep equation in correlating the overall data. The relative weights of each variable on the response can also be obtained. (3) Special regression techniques such as step-wise, ridge, and robust regressions and residual plots, etc., provide diagnostic tools for model selections. Multiple regression analysis performed on a set of carefully selected Zircaloy-2 in-reactor creep data leads to a model which provides excellent correlations for the data. (Auth.)

Some factors influencing the creep behaviour of alloy 800

International Nuclear Information System (INIS)

Asbury, F.E.; Willoughby, G.

1975-01-01

Studies have been made of the stability of the creep behaviour of two commercial casts of Incoloy 800, one high carbon and the other low carbon. The effects of pre-ageing, of prolonged creep up to 10 4 hours duration, and of grain size were investigated. Three factors were found to excercise a major influence on creep behaviour. Firstly, when the high carbon alloy was heat treated at 1150degC super-saturation effects, ascribed principally to carbon, gave some initial strengthening which would not, however, persist for the duration of service life in nuclear power plant applications above 600degC. Secondly, a gamma-dash type phase precipitated readily at 550 to 600degC, giving a marked increase in creep strength. Nucleation was sluggish at higher temperatures but once established, this form of strengthening could persist up to at least 650degC. Creep under non-isothermal conditions at 600 to 700degC would be complex on account of the behaviour of this phase. The hardening associated with its precipitation was greater in the low carbon alloy. Finally it was demonstrated that, in spite of gamma-dash precipitation, fine grained low carbon material was weak in creep at low stresses and temperatures. This was ascribed to the occurrence of grain boundary diffusion creep. It appears that this source of weakening would persist in service, and severely restrict the maximum temperature of usage for fined grained high tensile material. (author)

Influence of dynamic sodium environment on the creep-fatigue behaviour of Modified 9Cr-1Mo ferritic-martensitic steel

International Nuclear Information System (INIS)

Kannan, R.; Ganesan, V.; Mariappan, K.; Sukumaran, G.; Sandhya, R.; Mathew, M.D.; Bhanu Sankara Rao, K.

2011-01-01

Highlights: → The effects of dynamic sodium on the CFI behaviour of Mod. 9Cr-1Mo steel has investigated. → The cyclic stress response of Mod. 9Cr-1Mo steel under flowing sodium environment is similar to that of air environment. → The creep-fatigue endurance of the alloy is found to decrease with introduction of hold time and with increase in the duration of hold time and the factor of life increase in sodium compared to air environment is reduced with increase in hold time. → In contrast to air environment, tensile holds were found to be more damaging than compression hold in sodium environment. → Design rules based on air environment can be safely applied for the components operating in sodium environment. - Abstract: The use of liquid sodium as a heat transfer medium for sodium-cooled fast reactors (SFRs) necessitates a clear understanding of the effects of dynamic sodium on low cycle fatigue (LCF), creep and creep-fatigue interaction (CFI) behaviour of reactor structural materials. Mod. 9Cr-1Mo ferritic steel is the material of current interest for the steam generator components of sodium cooled fast reactors. The steam generator has a design life of 30-40 years. The effects of dynamic sodium on the LCF and CFI behaviour of Mod. 9Cr-1Mo steel have been investigated at 823 and 873 K. The CFI life of the steel showed marginal increase under flowing sodium environment when compared to air environment. Hence, the design rules for creep-fatigue interaction based on air tests can be safely applied for components operating in sodium environment. This paper attempts to explain the observed LCF and CFI results based on the detailed metallography and fractography conducted on the failed samples.

Creep and low cycles fatigue behaviour of inconel 617 and alloy 800H in the temperature range 1073-1223

International Nuclear Information System (INIS)

Yun, H.M.

1984-01-01

The creep rupture properties of high temperature alloys are being determined as part of the materials programme for the development of the high temperature, gas-cooled reactor (HTGR) as a source of nuclear process heat, especially for the gasification of lignite and coal. INCOLOY 800H AND INCONEL 617 have been tested in the temperature range from 1073 K to 1223 K in air as well as in helium with HTGR specific impurities. The static and dynamic creep behaviour of INCONEL 617 have been determined in constant load creep tests, relaxation tests and stress reduction tests. The results have been interpreted using the internal stress on the applied stress and test temperature was determined. In a few experiments the influence of cold deformation prior to the creep test on the magnitude of the internal stress was also investigated. (Author)

International RILEM Workshop on Creep Behaviour in Cracked Sections of Fibre Reinforced Concrete

CERN Document Server

Llano-Torre, Aitor; Cavalaro, Sergio

2017-01-01

This is the first publication ever focusing strictly on the creep behaviour in cracked sections of Fibre Reinforced Concrete (FRC). These proceedings contain the latest scientific papers about new testing methodologies, results and conclusions of multiple experimental campaigns and recommendations about significant factors of long-term behaviour, experiences from more than ten years of creep testing and some reflections about future perspectives on this topic. This book is an essential reference for all researchers of creep behaviour on FRC. This volume is the result of the efforts of the RILEM TC 261-CCF, that has been working since 2014 to develop standardized methodologies and guidelines to compare results from different laboratories and get a better understanding of the significant parameters related to creep of FRC.

Facility for in-reactor creep testing of fuel cladding

International Nuclear Information System (INIS)

Kohn, E.; Wright, M.G.

1976-11-01

A biaxial stress creep test facility has been designed and developed for operation in the WR-1 reactor. This report outlines the rationale for its design and describes its construction and the operating experience with it. The equipment is optimized for the determination of creep data on CANDU fuel cladding. Typical results from Zr-2.5 wt% Nb fuel cladding are used to illustrate the accuracy and reliability obtained. (author)

Fatigue-creep of martensitic steels containing 9-12% Cr: behaviour and damage

International Nuclear Information System (INIS)

Fournier, B.

2007-09-01

It is in the framework of the research programs on nuclear reactors (generation IV) that the martensitic steels containing 9-12% Cr are studied by the CEA. Most of the structures for which they are considered will be solicited in fatigue-creep at high temperature (550 C). The aim of this work is to understand and model the cyclic behaviour and the damage of these materials. The proposed modelling are based on detailed observations studies (SEM, TEM, EBSD...). The cyclic softening is attributed to the growth of the microstructure. A micro-mechanical model based on the physical parameters is proposed and leads to encouraging results. The damage results of interactions between fatigue, creep and oxidation. Two main types of damage are revealed. A model of anticipation of service time is proposed and gives very satisfying results. The possible extrapolations are discussed. (O.M.)

Structure and creep of Russian reactor steels with a BCC structure

Science.gov (United States)

Sagaradze, V. V.; Kochetkova, T. N.; Kataeva, N. V.; Kozlov, K. A.; Zavalishin, V. A.; Vil'danova, N. F.; Ageev, V. S.; Leont'eva-Smirnova, M. V.; Nikitina, A. A.

2017-05-01

The structural phase transformations have been revealed and the characteristics of the creep and long-term strength at 650, 670, and 700°C and 60-140 MPa have been determined in six Russian reactor steels with a bcc structure after quenching and high-temperature tempering. Creep tests were carried out using specially designed longitudinal and transverse microsamples, which were fabricated from the shells of the fuel elements used in the BN-600 fast neutron reactor. It has been found that the creep rate of the reactor bcc steels is determined by the stability of the lath martensitic and ferritic structures in relation to the diffusion processes of recovery and recrystallization. The highest-temperature oxide-free steel contains the maximum amount of the refractory elements and carbides. The steel strengthened by the thermally stable Y-Ti nanooxides has a record high-temperature strength. The creep rate at 700°C and 100 MPa in the samples of this steel is lower by an order of magnitude and the time to fracture is 100 times greater than that in the oxide-free reactor steels.

On the anisotropy of in-reactor creep of Zr-2.5Nb tubes

International Nuclear Information System (INIS)

Causey, A.R.; Holt, R.A.

1993-06-01

Creep specimens made from cold-worked Zr-2.5Nb tubes, fabricated with two different microstructures and crystallographic textures, were irradiated in the Osiris reactor in France in a fast-neutron flux of about 1.8 x 10 18 n.m -2 .s -1 , E > MeV, at 553 and 585 K. The hoop stresses from internal Fluences, up to 4 x 10 25 n.m -2 , more than double those achieved an any other creep test on cold-worked Zr-2.5Nb in which both axial and transverse strain were measured. Creep rates were obtained from strain versus fluence plots, and creep compliances were obtained from plots of the strain rates against hoop stress for each material at each temperature. The ratio of creep rates at 583 K to those at 553 K was ∼ 1.36, a little higher than that extrapolated from stress relaxation results at temperatures between 523 and 568 K. The ratio of the biaxial creep compliance in the axial direction to that in the transverse directions is different for the two test materials: 0.0 to -0.1 for the fuel sheathing texture and 0.5 to 0.6 for the pressure tube texture. The results were analysed using a self-consistent model developed to account for the contributions to the creep anisotropy of the three microstructure parameters involved and to account for the grain interaction effects. The model, which was normalized to test reactor and power reactor creep data for cold-worked Zr-2.5Nb tubes, predicted the ratio of the creep compliancies to be -0.26 and 0.63, respectively. Thus the creep anisotropy of Zr-2.5Nb tubes with pressure-tube-like crystallographic texture can be adequately predicted. (author). 18 refs., 4 tabs., 13 figs

Creep-rupture-strength and creep-behaviour of stainless steel X6CrNi 1811(DIN 1.4948); Comportamiento a la fluencia lenta del acero X6CrNi 1811 (1.4948)

Energy Technology Data Exchange (ETDEWEB)

Solano, R R; Schirra, M; Rivas, M de la; Seith, B

1977-07-01

The steel X6CrNi 1811 (DIN 1.4948) that will be used as a structure material for the German Fast Breeder Reactor SNR 300 was creep-tested in a temperature range of 550-650 degree centigree under base material condition as well as welded material condition. Tests are foreseen up to 30.000 hours with a continuous measuring of the elongation. The present report describes the test results up to about 4-000 hours. Taking into account the results of other programs carried out with the same material between 550- and 600 degree centigree at similar rupture time, were defined the stresses for the long term tests. The main point of this program (Extrapolation Program) lies in the knowledge of the creep-rupture-strength and creep behaviour of the structure materials up to 3.10{sup 4}h at high temperature in order to extrapolate up to 10{sup 5} h. for reactor operating temperatures. (Author) 14 refs.

Creep behaviour of austenitic stainless steels, base and weld metals used in liquid metal fast breeder reactors, during temperature variations

International Nuclear Information System (INIS)

Felsen, M.F.

1982-07-01

Creep rupture and deformation during temperature variations have been studied for 316 austenitic steel, base and weld metals. Loaded specimens were heated to 900 0 C or 1000 0 C and maintained at this temperature for different durations. The heating rate to these temperatures was between 5 and 50 0 C h -1 , whilst the cooling rate was between 5 and 20 0 C h -1 . The above tests were coupled with short time creep and tensile tests (straining rate 10 -2 h -1 to 10 3 h -1 ) at constant temperature. These tests were used for predicting the creep behaviour of the materials under changing temperature condition. The predictions were in good agreement with the changing temperature and creep experimental results. In addition, a correlation between certains tensile properties, such as the rupture time as a function of stress was observed at high temperature

Helium and its effects on the creep-fatigue behaviour of electron beam welds in the steel AISI-316-L

International Nuclear Information System (INIS)

Paulus, M.

1992-12-01

Within the scope of R and D work for materials development for the NET fusion experiment (Next European Torus) and the International Thermonuclear Experimental Reactor (ITER), the task reported was to examine electron beam welds in the austenitic stainless steel AISI 316 L (NET reference material) for their fatigue behaviour under creep load, and the effects of helium implantation on there mechanical properties. (orig.) [de

Creep behaviour of thin walled composite tubes

International Nuclear Information System (INIS)

Thiebaud, F.; Muzic, B.; Perreux, D.; Varchon, D.; Oytana, C.; Lebras, J.

1993-01-01

Fiber reinforced composites are more and more employed in high performance structure for nuclear power plant, mainly as water piping tubes. The increase of the use of composites is due to the advantages that they give : high stiffness, large ultimate strength, corrosion resistance. This last advantage is sought for the pieces in contact with water, and it's one of the reason why the composite can be preferred to metal. However the mechanical behaviour of composite is actually poorly known. The high anisotropy is the main difficulty to obtain a realistic model of behaviour. This problem implies that the safety factor used in the design of structure is often too large. In this article a general overview of the mechanical behaviour of tube made in glass epoxy material is proposed. We discuss especially the creep behaviour under biaxial loadings. The form of the proposed model presently allows predicting a nonlinearity of the behaviour and provides a good correlation with the experimental data of several tests not described in this paper. It accounts for the change of the Poisson ratio during creep and cyclic tests. However the complete identification requires long time testings and consequently the model must be corrected to take into account the damage which occurs in these cases

Constitutive modelling of creep-ageing behaviour of peak-aged aluminium alloy 7050

Directory of Open Access Journals (Sweden)

Yang Yo-Lun

2015-01-01

Full Text Available The creep-ageing behaviour of a peak-aged aluminium alloy 7050 was investigated under different stress levels at 174 ∘C for up to 8 h. Interrupted creep tests and tensile tests were performed to investigate the influences of creep-ageing time and applied stress on yield strength. The mechanical testing results indicate that the material exhibits an over-ageing behaviour which increases with the applied stress level during creep-ageing. As creep-ageing time approaches 8 h, the material's yield strength under different stress levels gradually converge, which suggests that the difference in mechanical properties under different stress conditions can be minimised. This feature can be advantageous in creep-age forming to the formed components such that uniformed mechanical properties across part area can be achieved. A set of constitutive equations was calibrated using the mechanical test results and the alloy-specific material constants were obtained. A good agreement is observed between the experimental and calibrated results.

Crack Growth Behaviour of P92 Steel Under Creep-fatigue Interaction Conditions

Directory of Open Access Journals (Sweden)

JING Hong-yang

2017-05-01

Full Text Available Creep-fatigue interaction tests of P92 steel at 630℃ under stress-controlled were carried out, and the crack propagation behaviour of P92 steel was studied. The fracture mechanism of crack growth under creep-fatigue interaction and the transition points in a-N curves were analyzed based on the fracture morphology. The results show that the fracture of P92 steel under creep-fatigue interaction is creep ductile fracture and the (Ctavg parameter is employed to demonstrate the crack growth behaviour; in addition, the fracture morphology shows that the crack growth for P92 steel under creep-fatigue interaction is mainly caused by the nucleation and growth of the creep voids and micro-cracks. Furthermore, the transition point of a-lg(Ni/Nf curve corresponds to the turning point of initial crack growth changed into steady crack growth while the transition point of (da/dN-N curve exhibits the turning point of steady creep crack growth changed into the accelerated crack growth.

Design basis for creep of zirconium alloy components in a fast neutron flux

International Nuclear Information System (INIS)

Ross-Ross, P.A.; Fidleris, V.

1975-01-01

The chalk River Nuclear Laboratory's experience with the creep of zirconium alloys in a neutron flux is described. Fast neutron flux changes the creep behaviour of zirconium alloys and new design criteria for in-reactor applications are needed. From experimental results empirical relations describing the effects of neutron flux, stress, temperature, time and anisotropy on creep rate were established. The relations are applied to the design of pressure tubes. (author)

Influence of microstructural parameters on the deformation and failure behaviour of the ODS alloy PM 2000 under creep and creep-fatigue loading

International Nuclear Information System (INIS)

Bothe, K.; Kussmaul, K.; Maile, K.

1999-01-01

The influence of grain size, manufacturing type and specimen direction (anisotropy) with respect to deformation and failure behaviour under creep, fatigue and creep-fatigue load was investigated. Thus, a basis for the correlation between microstructure and mechanical behaviour has been established. The specific damage and failure behaviour could be explained by means of the different microstructures observed. (orig.)

Creep behaviour of modified 9Cr-1Mo ferritic steel

International Nuclear Information System (INIS)

Choudhary, B.K.; Isaac Samuel, E.

2011-01-01

Creep deformation and fracture behaviour of indigenously developed modified 9Cr-1Mo steel for steam generator (SG) tube application has been examined at 823, 848 and 873 K. Creep tests were performed on flat creep specimens machined from normalised and tempered SG tubes at stresses ranging from 125 to 275 MPa. The stress dependence of minimum creep rate obeyed Norton's power law. Similarly, the rupture life dependence on stress obeyed a power law. The fracture mode remained transgranular at all test conditions examined. The analysis of creep data indicated that the steel obey Monkman-Grant and modified Monkman-Grant relationships and display high creep damage tolerance factor. The tertiary creep was examined in terms of the variations of time to onset of tertiary creep with rupture life, and a recently proposed concept of time to reach Monkman-Grant ductility, and its relationship with rupture life that depends only on damage tolerance factor. SG tube steel exhibited creep-rupture strength comparable to those reported in literature and specified in the nuclear design code RCC-MR.

Oxidation and creep behaviour of dense silicon nitride materials with different compositions

International Nuclear Information System (INIS)

Ernstberger, U.

1985-09-01

The study was intended to yield information on the oxidation and creep behaviour of Si 3 N 4 materials of different composition and microstructure, and produced by different processes. The experiments carried out in a vast temperature and load range showed that the chemical grain boundary composition is the key parameter affecting the materials' high-temperature properties. Significant correlations could be established between oxidation and creep behaviour on the one hand, and between microstructure and the behaviour on the other. (orig./IHOE) [de

Material pre-conditioning effects on the creep behaviour of 316H stainless steel

International Nuclear Information System (INIS)

Mehmanparast, A.; Davies, C.M.; Dean, D.W.; Nikbin, K.

2013-01-01

Material pre-conditioning by, for example, pre-strain through component bending and welding is known to alter the creep deformation and creep crack growth (CCG) behaviour of 316H stainless steel. Experimental test data on the creep deformation and crack growth behaviour of 316H weldment compact tension specimens at 550 °C, where the starter defect was introduced into the heat affected zone (HAZ), have been compared to those of obtained from similar specimens manufactured from parent material, which had been subjected to 8% compressive plastic pre-strain at room temperature. Similar degrees of accelerated cracking behaviour compared to parent material, for given values of C*, were exhibited in both 316H HAZ and pre-compressed parent materials. This acceleration has been attributed to the influence of material hardening effects and the reduction of creep ductility in the pre-conditioned materials. These results are discussed in terms of the potential for using material pre-conditioning to assist in predicting the long term cracking behaviour of high temperature 316H stainless steel plant components from shorter term laboratory CCG tests

Elevated temperature design of KALIMER reactor internals accounting for creep and stress-rupture effects

International Nuclear Information System (INIS)

Koo, Gyeong Hoi; Yoo, Bong

2000-01-01

In most LMFBR (Liquid Metal Fast Breed Reactor) design, the operating temperature is very high and the time-dependent creep and stress-rupture effects become so important in reactor structural design. Therefore, unlike with conventional PWR, the normal operating conditions can be basically dominant design loading because the hold time at elevated temperature condition is so long and enough to result in severe total creep ratcheting strains during total service lifetime. In this paper, elevated temperature design of the conceptually designed baffle annulus regions of KALIMER (Korea Advanced Liquid Metal Reactor) reactor internal structures is carried out for normal operating conditions which have the operating temperature 530 deg. C and the total service lifetime of 30 years. For the elevated temperature design of reactor internal structures, the ASME Code Case N-201-4 is used. Using this code, the time-dependent stress limits, the accumulated total inelastic strain during service lifetime, and the creep-fatigue damages are evaluated with the calculation results by the elastic analysis under conservative assumptions. The application procedures of elevated temperature design of the reactor internal structures using ASME code case N-201-4 with the elastic analysis method are described step by step in detail. This paper will be useful guide for actual application of elevated temperature design of various reactor types accounting for creep and stress-rupture effects. (author)

Experimental creep behaviour determination of cladding tube materials under multi-axial loadings

International Nuclear Information System (INIS)

Grosjean, Catherine; Poquillon, Dominique; Salabura, Jean-Claude; Cloue, Jean-Marc

2009-01-01

Cladding tubes are structural parts of nuclear plants, submitted to complex thermomechanical loadings. Thus, it is necessary to know and predict their behaviour to preserve their integrity and to enhance their lifetime. Therefore, a new experimental device has been developed to control the load path under multi-axial load conditions. The apparatus is designed to determine the thermomechanical behaviour of zirconium alloys used for cladding tubes. First results are presented. Creep tests with different biaxial loadings were performed. Results are analysed in terms of thermal expansion and of creep strain. The anisotropy of the material is revealed and iso-creep strain curves are given.

Sub-Surface and Bulk Creep Behaviour of Polyurethane/Clay Nanocomposites.

Science.gov (United States)

Jin, J; Yusoh, K; Zhang, H X; Song, M

2016-03-01

A series of exfoliated and intercalated polyurethane organoclay nanocomposites were prepared by in situ polymerization of polyol/organoclay mixture, chain extender and diisocyanate. The creep behaviour of subsurface and bulk of the polyurethane coatings was investigated by nanoindentation technique and uniaxial conventional creep testing method, respectively. The results showed that the creep resistance of the nanocomposites was significantly improved by incorporation of organoclay. The enhancement of creep resistance was dependent on clay content as well as organoclay structure (exfoliation or intercalation) in the polymer matrix. With 1 wt% organoclay, the creep resistance increased by about 50% for the intercalated organoclay and 6% for the exfoliated organoclay systems, respectively, compared to the pristine polyurethane. Viscoelastic model was employed to investigate the effect of organoclay loadings on the creep performance of the polyurethane. Results showed the model was in good agreement with the experimental data. Incorporation of clay leads to an increase in elastic deformation especially in exfoliated polyurethane nanocomposites and induces a higher initial displacement at the early stage of creep.

Irradiation creep experiments on fusion reactor candidate structural materials

International Nuclear Information System (INIS)

Hausen, H.; Cundy, M.R.; Schuele, W.

1991-01-01

Irradiation creep rates were determined for annealed and cold-worked AMCR- and 316-type steel alloys in the high flux reactor at Petten, for various irradiation temperatures, stresses and for neutron doses up to 4 dpa. Primary creep elongations were found in all annealed materials. A negative creep elongation was found in cold-worked materials for stresses equal to or below about 100 MPa. An increase of the negative creep elongation is found for decreasing irradiation temperatures and decreasing applied stresses. The stress exponent of the irradiation creep rate in annealed and cold-worked AMCR alloys is n = 1.85 and n = 1.1, respectively. The creep rates of cold-worked AMCR alloys are almost temperature independent over the range investigated (573-693 K). The results obtained in the HFR at Petten are compared with those obtained in ORR and EBR II. The smallest creep rates are found for cold-worked materials of AMCR- and US-PCA-type at Petten which are about a factor two smaller than the creep rates obtained of US-316 at Petten or for US-PCA at ORR or for 316L at EBR II. The scatter band factor for US-PCA, 316L, US-316 irradiated in ORR and EBR II is about 1.5 after a temperature and damage rate normalization

A regression approach for zircaloy-2 in-reactor creep constitutive equations

International Nuclear Information System (INIS)

Yung Liu, Y.; Bement, A.L.

1977-01-01

In this paper the methodology of multiple regressions as applied to zircaloy-2 in-reactor creep data analysis and construction of constitutive equation are illustrated. While the resulting constitutive equation can be used in creep analysis of in-reactor zircaloy structural components, the methodology itself is entirely general and can be applied to any creep data analysis. From data analysis and model development point of views, both the assumption of independence and prior committment to specific model forms are unacceptable. One would desire means which can not only estimate the required parameters directly from data but also provide basis for model selections, viz., one model against others. Basic understanding of the physics of deformation is important in choosing the forms of starting physical model equations, but the justifications must rely on their abilities in correlating the overall data. The promising aspects of multiple regression creep data analysis are briefly outlined as follows: (1) when there are more than one variable involved, there is no need to make the assumption that each variable affects the response independently. No separate normalizations are required either and the estimation of parameters is obtained by solving many simultaneous equations. The number of simultaneous equations is equal to the number of data sets, (2) regression statistics such as R 2 - and F-statistics provide measures of the significance of regression creep equation in correlating the overall data. The relative weights of each variable on the response can also be obtained. (3) Special regression techniques such as step-wise, ridge, and robust regressions and residual plots, etc., provide diagnostic tools for model selections

Evaluation of long-term creep behaviour on K-cladding tubes

International Nuclear Information System (INIS)

Bang, J. G.; Jeong, Y. H.; Jeong, Y. H.

2003-01-01

KAERI has developed new zirconium alloys for high burnup fuel cladding. To evaluate the performance of these alloys, various out-pile tests are conducting. At high burnup, the creep resistance as well as corrosion resistance is one of the major factors determining nuclear fuel performance. Long-term creep test was performed at 350 .deg. C and 400 .deg. C and 100, 120, 135, and 150 MPa of applied hoop stress to evaluate the creep properties. The creep resistance was strongly affected by the final heat treatment conditions, while there was no effect of intermediate heat treatment. The creep strain of the recrystallized alloys is higher than that of the stress-relieved alloys by a factor of 3. The alloying elements also influenced the creep behaviour. Increase of Sn content enhanced the creep resistance, while Nb decreased the creep resistance. As a result of texture analysis, basal pole was directed to normal direction, while prism pole was to rolling direction. The development of the deformation texture and the ammealing texture showed almost similar process to Zircaloy cladding

Irradiation-induced creep in fuel compacts for high-temperature reactor applications

Energy Technology Data Exchange (ETDEWEB)

Veringa, H; Blackstone, R [Stichting Energieonderzoek Centrum Nederland, Petten; Loelgen, R

1977-01-01

Restrained shrinkage experiments at neutron fluences up to 3 x 10/sup 21/ n cm/sup -2/ DNE in the temperature range 600 to 1200/sup 0/C were performed on three different dummy coated-particle fuel compacts in the high-flux reactor at Petten. The data were evaluated to obtain the steady-state radiation creep coefficient of the compacts. It was found that, for the materials investigated, the creep coefficient is temperature dependent, but no clear relationship with Young's modulus could be established. Under certain conditions this irradiation-induced plasticity influences the elastic properties, with the concomitant increase of the creep coefficient. This effect coincides with the formation and further opening up of cracks due to stresses caused by irradiation-induced shrinkage of matrix material.

Irradiation-induced creep in fuel compacts for high-temperature reactor applications

International Nuclear Information System (INIS)

Veringa, H.; Blackstone, R.; Loelgen, R.

1977-01-01

Restrained shrinkage experiments at neutron fluences up to 3 x 10 21 n cm -2 DNE in the temperature range 600 to 1200 0 C were performed on three different dummy coated-particle fuel compacts in the high-flux reactor at Petten. The data were evaluated to obtain the steady-state radiation creep coefficient of the compacts. It was found that, for the materials investigated, the creep coefficient is temperature dependent, but no clear relationship with Young's modulus could be established. Under certain conditions this irradiation-induced plasticity influences the elastic properties, with the concomitant increase of the creep coefficient. This effect coincides with the formation and further opening up of cracks due to stresses caused by irradiation-induced shrinkage of matrix material. (author)

In-pile creep behaviour of Zry-4 and ZrNb3Sn1 cladding under uniaxial and biaxial stress

International Nuclear Information System (INIS)

Boehner, G.; Wildhagen, B.; Wilhelm, H.

1987-01-01

An irradiation programme - started in 1977 - was performed at the research reactor FRG-2 at Geesthacht, Germany, as a joint project of GKSS and KWU in order to study the in-pile creep behaviour of zirconium alloy cladding tubes of PWR fuel rods. The test objective was to establish a data base which allows refined modelling of the in-pile creep phenomenon. A wide test matrix was realized in which each of the precisely monitored test conditions (hoop stress, temperature, fast neutron flux) was varied separately. Different cladding materials (Zircaloy-4 and Zirconium-Niob-Tin alloy ZrNb3Sn1) were subjected to those varying test conditions. Cladding tube specimens of 10.75 mm outer diameter were irradiated in test capsules under various stress conditions and levels up to approx. 6000 h, at temperatures ranging from 300 0 C to 400 0 C and fast neutron flux (E > 1 MeV) of approx. 3x10 13 cm -2 .s -1 . Diametrical and/or axial creep deformation of all tubes were measured in the Hot Cells several times in the course of the tests. In order to extract the irradiation induced creep strain some out-pile experiments were carried out under the very same test conditions as the in-pile tests concerned. (orig./GL)

Design and fabrication of a dead weight equipment to perform creep measurements on highly irradiated beryllium specimens

International Nuclear Information System (INIS)

Scibetta, M.; Pellettieri, A.; Wouters, P.; Leenaerts, A.; Verpoucke, G.

2005-01-01

Beryllium is an important material to be used in the blanket of fusion reactors. It acts as a neutron multiplier that allows tritium production. In order to use this material effectively, some data on creep and swelling behaviour are needed. This paper describes preliminary microstructural investigations and the qualification of a creep set-up that will be used to measure creep of highly irradiated beryllium from the BR2 research reactor matrix. (Author)

Effect of simulated sampling disturbance on creep behaviour of rock salt

Science.gov (United States)

Guessous, Z.; Gill, D. E.; Ladanyi, B.

1987-10-01

This article presents the results of an experimental study of creep behaviour of a rock salt under uniaxial compression as a function of prestrain, simulating sampling disturbance. The prestrain was produced by radial compressive loading of the specimens prior to creep testing. The tests were conducted on an artifical salt to avoid excessive scattering of the results. The results obtained from several series of single-stage creep tests show that, at short-term, the creep response of salt is strongly affected by the preloading history of samples. The nature of this effect depends upon the intensity of radial compressive preloading, and its magnitude is a function of the creep stress level. The effect, however, decreases with increasing plastic deformation, indicating that large creep strains may eventually lead to a complete loss of preloading memory.

Temperature-dependence of creep behaviour of dental resin-composites.

Science.gov (United States)

El-Safty, S; Silikas, N; Watts, D C

2013-04-01

To determine the effect of temperature, over a clinically relevant range, on the creep behaviour of a set of conventional and flowable resin-composites including two subgroups having the same resin matrix and varied filler loading. Eight dental resin-composites: four flowable and four conventional were investigated. Stainless steel split moulds (4 mm × 6 mm) were used to prepare cylindrical specimens for creep examination. Specimens were irradiated in the moulds in layers of 2mm thickness (40s each), as well as from the radial direction after removal from the moulds, using a light-curing unit with irradiance of 650 mW/cm(2). A total of 15 specimens from each material were prepared and divided into three groups (n=5) according to the temperature; Group I: (23°C), Group II: (37°C) and Group III: (45°C). Each specimen was loaded (20 MPa) for 2h and unloaded for 2h. Creep was measured continuously over the loading and unloading periods. At higher temperatures greater creep and permanent set were recorded. The lowest mean creep occurred with GS and GH resin-composites. Percentage of creep recovery decreased at higher temperatures. At 23°C, the materials exhibited comparable creep. At 37°C and 45°C, however, there was a greater variation between materials. For all resin-composites, there was a strong linear correlation with temperature for both creep and permanent set. Creep parameters of resin-composites are sensitive to temperature increase from 23 to 45°C, as can occur intra-orally. For a given resin matrix, creep decreased with higher filler loading. Copyright © 2012 Elsevier Ltd. All rights reserved.

Creep crack growth in a reactor pressure vessel steel at 360 deg C

Energy Technology Data Exchange (ETDEWEB)

Wu, Rui; Seitisleam, F; Sandstroem, R [Swedish Institute for Metals Research, Stockholm (Sweden)

1999-12-31

Plain creep (PC) and creep crack growth (CCG) tests at 360 deg C and post metallography were carried out on a low alloy reactor pressure vessel steel (ASTM A508 class 2) with different microstructures. Lives for the CCG tests were shorter than those for the PC tests and this is more pronounced for simulated heat affected zone microstructure than for the parent metal at longer lives. For the CCG tests, after initiation, the cracks grew constantly and intergranularly before they accelerated to approach rupture. The creep crack growth rate is well described by C*. The relations between reference stress, failure time and steady crack growth rate are presented for the CCG tests. It is demonstrated that the failure stress due to CCG is considerably lower than the yield stress at 360 deg C. Consequently, the CCG will control the static strength of a reactor vessel. (orig.) 17 refs.

Creep crack growth in a reactor pressure vessel steel at 360 deg C

Energy Technology Data Exchange (ETDEWEB)

Rui Wu; Seitisleam, F.; Sandstroem, R. [Swedish Institute for Metals Research, Stockholm (Sweden)

1998-12-31

Plain creep (PC) and creep crack growth (CCG) tests at 360 deg C and post metallography were carried out on a low alloy reactor pressure vessel steel (ASTM A508 class 2) with different microstructures. Lives for the CCG tests were shorter than those for the PC tests and this is more pronounced for simulated heat affected zone microstructure than for the parent metal at longer lives. For the CCG tests, after initiation, the cracks grew constantly and intergranularly before they accelerated to approach rupture. The creep crack growth rate is well described by C*. The relations between reference stress, failure time and steady crack growth rate are presented for the CCG tests. It is demonstrated that the failure stress due to CCG is considerably lower than the yield stress at 360 deg C. Consequently, the CCG will control the static strength of a reactor vessel. (orig.) 17 refs.

Anomalous creep behaviour of 316 stainless steel at 550 deg C

International Nuclear Information System (INIS)

Hyde, T.H.

1986-01-01

The results of fifteen constant-load creep tests at 550 0 C, with nominal stresses in the range 200 to 360 MPa and with test durations of up to 14000h, are presented. The usual primary, secondary and tertiary creep behaviour was exhibited for nominal stresses greater than about 330 MPa. At lower stresses, 'renewed' primary and secondary creep regions were observed. The renewed secondary creep strain rates were found to be about an order of magnitude greater than the initial secondary creep strain rates. The results indicate that the occurence of the renewed primary and secondary creep regions is associated with time-dependent exposure to a temperature of 550 0 C. The presence or magnitude of the prior stress level does not appear to have any significant effect. The results are relevant to design procedures because extrapolation of short duration or high stress data to long-term design lifetimes is often required. Unless the possibility of the occurence of renewed primary and secondary creep is taken into account, gross errors in strain predictions could occur. (author)

Critical review of creep FRAPCON-3 model under dry storage conditions

Energy Technology Data Exchange (ETDEWEB)

Feria, F.; Herranz, L.E. [Unit of Nuclear Safety Research, CIEMAT, Avda. Complutense 22, Madrid, Madrid 28040 (Spain)

2009-06-15

There is a general agreement that cladding creep rupture is the most likely and limiting failure mechanism of spent fuel in dry storage compared to other potential mechanisms, like stress corrosion cracking and/or delayed hydride cracking. Nevertheless, occurrence of creep rupture is very improbable since both decay heat and hoop stress tend to decrease throughout dry storage. In spite of this, the current trend to higher burn up levels needs further attention that ensures safe storage of spent fuel irradiated over 45 GWd/MTU. An extensive work has been carried out during the last four decades in the area of in-reactor creep modelling. Unfortunately, the in-reactor conditions are so different from those prevailing under dry storage, that all the experience gained cannot be extrapolated in a straightforward manner. On the other side, as creep tests simulating conditions throughout a 20-40 year dry storage are impractical, post-irradiation cladding creep behaviour has been modelled by means of time-temperature dependent laws developed on the basis of currently available zirconium alloys data. Additionally, some tests have been exploring the effect of irradiation, hydrogen distribution and material composition on the materials creep behaviour. Adaptation of fuel performance codes initially developed for normal and off-normal reactor operation is not an easy task either. Creep modelling is usually dependent of host codes because a good part of its validation and update has been carried out in an integral way, and as a consequence its independent performance assessment is not an easy task. This work examines the current capability of FRAPCON-3 to model creep behaviour under dry storage conditions. To do so, a review of its major fundamentals has been done and its range of applicability discussed. Once its main approximations and drawbacks have been identified, an attempt to overcome some of them has been intended by implementing an alternative expression for creep under

High-Temperature Creep Behaviour and Positive Effect on Straightening Deformation of Q345c Continuous Casting Slab

Science.gov (United States)

Guo, Long; Zhang, Xingzhong

2018-03-01

Mechanical and creep properties of Q345c continuous casting slab subjected to uniaxial tensile tests at high temperature were considered in this paper. The minimum creep strain rate and creep rupture life equations whose parameters are calculated by inverse-estimation using the regression analysis were derived based on experimental data. The minimum creep strain rate under constant stress increases with the increase of the temperature from 1000 °C to 1200 °C. A new casting machine curve with the aim of fully using high-temperature creep behaviour is proposed in this paper. The basic arc segment is cancelled in the new curve so that length of the straightening area can be extended and time of creep behaviour can be increased significantly. For the new casting machine curve, the maximum straightening strain rate at the slab surface is less than the minimum creep strain rate. So slab straightening deformation based on the steel creep behaviour at high temperature can be carried out in the process of Q345c steel continuous casting. The effect of creep property at high temperature on slab straightening deformation is positive. It is helpful for the design of new casting machine and improvement of old casting machine.

Boundary element method for modelling creep behaviour

International Nuclear Information System (INIS)

Zarina Masood; Shah Nor Basri; Abdel Majid Hamouda; Prithvi Raj Arora

2002-01-01

A two dimensional initial strain direct boundary element method is proposed to numerically model the creep behaviour. The boundary of the body is discretized into quadratic element and the domain into quadratic quadrilaterals. The variables are also assumed to have a quadratic variation over the elements. The boundary integral equation is solved for each boundary node and assembled into a matrix. This matrix is solved by Gauss elimination with partial pivoting to obtain the variables on the boundary and in the interior. Due to the time-dependent nature of creep, the solution has to be derived over increments of time. Automatic time incrementation technique and backward Euler method for updating the variables are implemented to assure stability and accuracy of results. A flowchart of the solution strategy is also presented. (Author)

Numerical description of creep of highly creep resistant alloys

International Nuclear Information System (INIS)

Preussler, T.

1991-01-01

Fatigue tests have been performed with a series of highly creep resistant materials for gas turbines and related applications for gaining better creep data up to long-term behaviour. The investigations were performed with selected individual materials in the area of the main applications down to strains and stresses relevant to design, and have attained trial durations of 25000 to 60000 h. In continuing former research, creep equations for a selection of characterizing individual materials have been improved and partly newly developed on the basis of a differentiated evaluation. Concerning the single materials, there are: one melt each of the materials IN-738 LC, IN-939, IN-100, FSX-414 and Inconel 617. The applied differentiated evaluation is based on the elastoplastical behaviour from the hot-drawing test, the creep behaviour from the non interrupted or the interrupted fatigue test, and the contraction behaviour from the annealing test. The creep equations developed describe the high temperature deformation behaviour taking into account primary, secondary and partly the tertiary creep dependent of temperature, stress and time. These equations are valid for the whole application area of the respective material. (orig./MM) [de

Structural evaluation of fast reactor core restraint with irradiation creep-swelling opposition effects

International Nuclear Information System (INIS)

Kalinowski, J.E.

1979-01-01

Irradiation creep and swelling correlations are derived from primary loading in-reactor experiments in which irradiation creep and swelling act in the same direction. When correlation uncertainty bands are applied in core restraint evaluations, significant variability in sub-assembly behavior is predicted. For example, sub-assemblies in the outer core region where neutron flux and duct temperature gradients are significant exhibit bowing responses ranging from a creep dominated outward bow to a swelling dominated inward bow. Furthermore, solutions based on upper bound and lower bound correlation uncertainty combinations are observed to cross-over indicating that such combinations are physically unrealistic in the assessment of creep-swelling opposition effects. In order to obtain realistic upper and lower bound sub-assembly responses, judgement must be applied in the selection of creep-swelling equation uncertainty combinations. Experimental programs have been defined which will provide the needed basic as well as prototypic creep-swelling opposition data for reference and advanced sub-assembly duct alloys. The first of these is an irradiation of cylindrical capsules subjected to a through-wall temperature gradient. This test which is presently underway in the EBR-II reactor will provide the data needed to refine irradiation creep and swelling correlations and their associated uncertainties when applied to core restraint evaluations. Restrained pin and duct bowing experiments in FFTF have also been defined. These will provide the prototypic data necessary to verify irradiated duct bowing methodology. The results of this experimental program are expected to reduce creep and swelling uncertainties and permit better definition of the design window for load plane gaps. (orig.)

Evaluation of results from an in-pile creep test in the Studsvik R2 reactor

Energy Technology Data Exchange (ETDEWEB)

Pettersson, Kjell [Entropy Materials, Stockholm (Sweden)

2002-01-01

An in-pile creep test with bowing of cladding tubes has been performed in a hot water loop in the Studsvik R2 reactor . One test was performed in the core and one outside the core. The out-of-pile sample showed some minor primary creep strain while the in-pile specimen deformed at a steady rate of 5x10{sup -7}/h . However, when the results were compared to a broader data base of Zircaloy in-pile creep it became clear that the creep deformation observed is a primary creep which occurs before the irradiation creep in Zircaloy reaches a constant steady state creep rate. This primary stage is interpreted as a consequence of the development of an irradiation induced microstructure in Zircaloy which does not reach a steady state until a dose of about 10{sup 21} n/cm{sup 2} . At this stage the steady state irradiation creep starts. From this interpretation it is concluded that it is quite feasible to use the test method on pre-irradiated material in which it can be expected that the steady state will be reached already after short irradiation times.

Creep deformation and rupture behaviour of 9Cr–1W–0.2V–0.06Ta Reduced Activation Ferritic–Martensitic steel

International Nuclear Information System (INIS)

Vanaja, J.; Laha, K.; Mythili, R.; Chandravathi, K.S.; Saroja, S.; Mathew, M.D.

2012-01-01

Highlights: ► Creep tests on broad temperature and stress ranges were carried out. ► Microstructural instability on creep and thermal exposures were studied using TEM. ► Creep damage tolerance factor of the material was estimated. - Abstract: This paper presents the creep deformation and rupture behaviour of indigenously produced 9Cr–1W–0.2V–0.06Ta Reduced Activation Ferritic–Martensitic (RAFM) steel for fusion reactor application. Creep studies were carried out at 773, 823 and 873 K over a stress range of 100–300 MPa. The creep deformation of the steel was found to proceed with relatively shorter primary regime followed by an extended tertiary regime with virtually no secondary regime. The variation of minimum creep rate of the material with applied stress followed a power law relation, ε m = Aσ n , with stress exponent value ‘n’ decreasing with increase in temperature. The product of minimum creep rate and creep rupture life was found to obey the modified Monkman–Grant relation. The time to onset of tertiary stage of deformation was directly proportional to rupture life. TEM studies revealed relatively large changes in martensitic sub-structure and coarsening of precipitates in the steel on creep exposure as compared to thermal exposure. Microstructural degradation was considered as the prime cause of extended tertiary stage of creep deformation, which was also reflected in the damage tolerance factor λ with a value more than 2.5. In view of the microstructural instability of the material on creep exposure, the variation of minimum creep rate with stress and temperature did not obey Dorn's equation modified by invoking Lagneborg and Bergman's concepts of back stress.

Creep behaviour of polyurethanes applied in the offshore industry under dynamic service conditions

Energy Technology Data Exchange (ETDEWEB)

Aquino, Fabio G.; Sheldrake, Terry; Clevelario, Judimar; Pires, Fabio S. [Wellstream International S/A - Rio de Janeiro, RJ (Brazil)], e-mail: fabio.aquino@wellstream.com; Souza, Miguel L. [Newtech Ltda, Sao Carlos, SP (Brazil)

2011-07-01

The oil industry commonly uses flexible pipes to convey oil and gas from wells to platforms that move constantly due to weather and tidal conditions. In this scenario, polymeric components are required to transitioning between the flexible material of the pipelines to the rigid material of the platform; polyurethanes are versatile polymers suitable for performing such services. As this material is subjected to constant loading during working conditions, and it its durability is to be maintained for several decades, it is important to determine the material's creep properties that relate to deformation caused by constant loading, which can represent an indirect measurement of the material's lifetime. In this study, creep behaviour data on the polyurethane samples was collected and an asymmetrical and nonlinear behaviour was observed. Additionally the material presented a creep fracture line with points only above 150% of deformation, considerably exceeding maximum values for its service conditions, which is limited to 10% of deformation considering the worst loading case for design premises of the final artifact. (author)

Creep behaviour and microstructure of the ferritic material No. 1-6770 under irradiation

International Nuclear Information System (INIS)

Herschbach, K.; Ehrlich, K.; Materna, E.

Creep behaviour under irradiation of the ferritic steel-DIN-1-6770 is quite different of austenitic steel behaviour, in particular temperature sensitivity is important and response to stress is non linear. The microstructure stays unchanged

Microstructure stability and creep behaviour of advanced high chromium ferritic steels

Czech Academy of Sciences Publication Activity Database

Sklenička, Václav; Kuchařová, Květa; Kudrman, J.; Svoboda, Milan; Kloc, Luboš

43 2005, č. 1 (2005), s. 20-33 ISSN 0023-432X R&D Projects: GA ČR(CZ) GA106/02/0608; GA AV ČR(CZ) IAA2041101; GA AV ČR(CZ) 1QS200410502 Institutional research plan: CEZ:AV0Z20410507 Keywords : 9-12%Cr steels * microstructure stability * creep behaviour * nonsteady creep loading Subject RIV: JG - Metallurgy Impact factor: 0.973, year: 2005

Improvement in the long term creep rupture strength of SUS 316 steel for fast breeder reactors by nitrogen addition

International Nuclear Information System (INIS)

Nakazawa, Takanori; Abo, Hideo; Tanino, Mitsuru; Komatsu, Hazime; Tashimo, Masanori; Nishida, Takashi.

1989-01-01

Improvement of creep fatigue property of structural materials for fast breeder reactors. In order to improve the resistance to creep fatigue of SUS 316 steels, the effects of nitrogen, carbon, and molybdenum on creep properties have been investigated, under the concept that creep fatigue endurance is correspond to creep rupture ductility. Creep rupture tests and slow strain rate tensile tests were conducted at 550degC and extensive microstructural works were performed. The strengthening by nitrogen is much greater than carbon. Moreover, while carbon reduces rupture ductility, nitrogen does not change it. The addition of carbon results in coarse carbide formation on grain boundaries during creep, but with nitrogen very fine Fe 2 Mo particles precipitate on grain boundaries. The difference between the effects of nitrogen and carbon on creep properties is arise from the different morphology of precipitation. Strengthening by molybdenum brings about a slight decrease in rupture ductility. On the basis of these results, 0.01%C-0.07%N-11%Ni-16.5%Cr-2%Mo steel is selected as a promising material for fast breeder reactors. This steel has higher rupture ductility and strength than SUS 316 steel. It is also confirmed that this steel has a higher resistance to creep fatigue. (author)

Investigations on the creep-rupture behaviour of the austenitic stainless steel AISI 316 NET

International Nuclear Information System (INIS)

Schirra, M.; Ritter, B.

1988-12-01

The report describes the creep-rupture tests carried out with a 17/13/2 CrNiMo-steel in the frame of the German-Spanish collaboration (KfK-CIEMAT). The material studied is the austenitic steel AISI 316(L) selected as potential first-wall material for NET (Next European Torus). The test programme on base material with a NET specified batch encompasses until now in the temperature range 500-750 0 C the rupture-time-range till 20 000 h. The results permit statements to the creep- and creep-rupture behaviour and ductility. Metallography examinations give information about fracture behaviour and demonstrate the complex precipitation happening. The results are compared with the literature and own test results from two batches of the Fast-Breeder-Program. (orig.) [de

Analysis of the creep behaviour of die-cast Mg–3Al–1Si alloy

Energy Technology Data Exchange (ETDEWEB)

Zhu, S.M., E-mail: suming.zhu@monash.edu [CAST Cooperative Research Centre, Department of Materials Engineering, Monash University, Victoria 3800 (Australia); ARC Centre of Excellence for Design in Light Metals, Department of Materials Engineering, Monash University, Victoria 3800 (Australia); Easton, M.A. [CAST Cooperative Research Centre, Department of Materials Engineering, Monash University, Victoria 3800 (Australia); Gibson, M.A. [CAST Cooperative Research Centre, CSIRO Process Science and Engineering, Clayton, Victoria 3169 (Australia); Dargusch, M.S. [Centre for Advanced Materials Processing and Manufacturing, School of Mechanical and Mining Engineering, The University of Queensland, Queensland 4075 (Australia); Defence Materials Technology Centre, The University of Queensland, Queensland 4075 (Australia); Nie, J.F. [ARC Centre of Excellence for Design in Light Metals, Department of Materials Engineering, Monash University, Victoria 3800 (Australia)

2013-08-20

The creep behaviour of die-cast Mg–3Al–1Si (AS31) alloy has been studied at 125 °C, 150 °C and 175 °C with stresses ranging 50–110 MPa. The alloy exhibits anomalously high stress exponents, i.e. 14.4 at 125 °C, 11.6 at 150 °C and 9.5 at 175 °C. Contrary to work reported previously, these high stress exponents cannot be rationalised using the threshold stress approach that is commonly adopted in analysing creep behaviour of dispersion strengthened alloys or metal matrix composites. It is shown that the observed high stress exponents are associated with the dominance of power-law breakdown creep in this study, and the stress dependence can be well described by the Garofalo sinh relationship with the natural exponent of 5. Transmission electron microscopy (TEM) observations reveal that cross-slip of 〈a〉 type dislocations is probably the controlling creep mechanism.

Analysis of the creep behaviour of die-cast Mg–3Al–1Si alloy

International Nuclear Information System (INIS)

Zhu, S.M.; Easton, M.A.; Gibson, M.A.; Dargusch, M.S.; Nie, J.F.

2013-01-01

The creep behaviour of die-cast Mg–3Al–1Si (AS31) alloy has been studied at 125 °C, 150 °C and 175 °C with stresses ranging 50–110 MPa. The alloy exhibits anomalously high stress exponents, i.e. 14.4 at 125 °C, 11.6 at 150 °C and 9.5 at 175 °C. Contrary to work reported previously, these high stress exponents cannot be rationalised using the threshold stress approach that is commonly adopted in analysing creep behaviour of dispersion strengthened alloys or metal matrix composites. It is shown that the observed high stress exponents are associated with the dominance of power-law breakdown creep in this study, and the stress dependence can be well described by the Garofalo sinh relationship with the natural exponent of 5. Transmission electron microscopy (TEM) observations reveal that cross-slip of 〈a〉 type dislocations is probably the controlling creep mechanism

Coupled thermo-mechanical creep analysis for boiling water reactor pressure vessel lower head

International Nuclear Information System (INIS)

Villanueva, Walter; Tran, Chi-Thanh; Kudinov, Pavel

2012-01-01

Highlights: ► We consider a severe accident in a BWR with melt pool formation in the lower head. ► We study the influence of pool depth on vessel failure mode with creep analysis. ► There are two modes of failure; ballooning of vessel bottom and a localized creep. ► External vessel cooling can suppress creep and subsequently prevent vessel failure. - Abstract: In this paper we consider a hypothetical severe accident in a Nordic-type boiling water reactor (BWR) at the stage of relocation of molten core materials to the lower head and subsequent debris bed and then melt pool formation. Nordic BWRs rely on reactor cavity flooding as a means for ex-vessel melt coolability and ultimate termination of the accident progression. However, different modes of vessel failure may result in different regimes of melt release from the vessel, which determine initial conditions for melt coolant interaction and eventually coolability of the debris bed. The goal of this study is to define if retention of decay-heated melt inside the reactor pressure vessel is possible and investigate modes of the vessel wall failure otherwise. The mode of failure is contingent upon the ultimate mechanical strength of the vessel structures under given mechanical and thermal loads and applied cooling measures. The influence of pool depth and respective transient thermal loads on the reactor vessel failure mode is studied with coupled thermo-mechanical creep analysis. Efficacy of control rod guide tube (CRGT) cooling and external vessel wall cooling as potential severe accident management measures is investigated. First, only CRGT cooling is considered in simulations revealing two different modes of vessel failure: (i) a ‘ballooning’ of the vessel bottom and (ii) a ‘localized creep’ concentrated within the vicinity of the top surface of the melt pool. Second, possibility of in-vessel retention with CRGT and external vessel cooling is investigated. We found that the external vessel

Creep damage behaviour of modified 9Cr-1Mo steel weld joints

International Nuclear Information System (INIS)

Sakthivel, T.; Laha, K.; Vasudevan, M.; Panneer Selvi, S.

2016-01-01

Creep deformation and rupture behaviour of modified 9Cr-1Mo steel weld joints fabricated by single-pass activated TIG (A-TIG) and shielded metal arc welding (SMAW) processes have been investigated at 923 K over a stress range of 50 to 110 MPa after post weld heat treatment (PWHT). The weld joints exhibited significantly lower creep rupture lives than the base metal at lower applied stresses. Creep rupture location of the weld joints were found to occur in the ICHAZ. An extensive localized creep deformation, coarsening of M 23 C 6 precipitates in the ICHAZ with creep exposure led to the premature type IV failure of the joints. The coarsening of M 23 C 6 precipitates was extensive in the mid-section of the ICHAZ than the sub-surface of the joints, and was more predominant in the SMAW joint. While A-TIG weld joint exhibited reduced creep cavitation and coarsening of M 23 C 6 precipitates due to lower deformation constraints by adjacent regions in the ICHAZ. Hence, A-TIG weld joint exhibited higher creep rupture life than the SMAW joint. (author)

Creep-fatigue behaviour of the titanium alloy IMI 834 at 600 C

International Nuclear Information System (INIS)

Nowack, H.; Kordisch, T.

1998-01-01

In the present study the creep-fatigue behaviour of the titanium alloy IMI 834 at 600 C was investigated. A comparison of the crack initiation life behaviour and of the crack propagation as caused by different types of complex creep-fatigue cycles (with hold times into tension and/or into compression direction and with different loading rates into tension and/or into compression direction) showed, that a slow increase of the loadings into tension reduced the life and increased the crack velocity more than hold times at the maximum load. Furthermore, there existed environmental influences. On the basis of the experimental investigations the prediction capability of convenient crack initiation life prediction methods was evaluated. It turned out that the prediction capability of the strain range partitioning method could be improved if it was frequency modified. The prediction capability of the frequency modification method could also be improved, if mean stresses in the cycles were explicitely accounted for. In the short and long crack stage the propagation behaviour could be correlated well if the effective cyclic J-integral was used. This is of importance for damage tolerance considerations. Because the strains and the stresses at the crack tip are most important for the crack propagation behaviour, they were analysed on the basis of the finite element method. It was found that the strains and stresses differed for different types of creep-fatigue cycles. (orig.)

High temperature graphite irradiation creep experiment in the Dragon Reactor. Dragon Project report

Energy Technology Data Exchange (ETDEWEB)

Manzel, R.; Everett, M. R.; Graham, L. W.

1971-05-15

The irradiation induced creep of pressed Gilsocarbon graphite under constant tensile stress has been investigated in an experiment carried out in FE 317 of the OECD High Temperature Gass Cooled Reactor ''Dragon'' at Winfrith (England). The experiment covered a temperature range of 850 dec C to 1240 deg C and reached a maximum fast neutron dose of 1.19 x 10²¹ n cm^-2 NDE (Nickel Dose DIDO Equivalent). Irradiation induced dimensional changes of a string of unrestrained graphite specimens are compared with the dimensional changes of three strings of restrained graphite specimens stressed to 40%, 58%, and 70% of the initial ultimate tensile strength of pressed Gilsocarbon graphite. Total creep strains ranging from 0.18% to 1.25% have been measured and a linear dependence of creep strain on applied stress was observed. Mechanical property measurements carried out before and after irradiation demonstrate that Gilsocarbon graphite can accommodate significant creep strains without failure or structural deterioration. Total creep strains are in excellent agreement with other data, however the results indicate a relatively large temperature dependent primary creep component which at 1200 deg C approaches a value which is three times larger than the normally assumed initial elastic strain. Secondary creep constants derived from the experiment show a temperature dependence and are in fair agreement with data reported elsewhere. A possible determination of the results is given.

Life prediction of simple structures subject to cyclic primary and secondary loading resulting in creep and platicity

International Nuclear Information System (INIS)

Otter, N.R.; Jones, R.T.

1979-01-01

High temperature reactors are subject to cyclic mechanical and thermal loadings resulting from start up and shut down operations. The design must therefore guard against structural failure resulting from excessive deformation and creep-fatigue damage. Before any simplified inelastic analysis techniques can be applied, their validity needs to be examined under situations representative of the reactor. For this to be carried out it is necessary to determine the behaviour of components, initially geometrically simple, subject to loadings, cyclic primary and secondary in nature, which result in creep and plasticity. Beam-like structures have been investigated on a finite element basis with the aim of determining how cyclic plasticity, creep enhancement and plastic ratchetting vary in relationship with modified shakedown criteria, magnitude of loading and hold time. (orig.)

Preliminary analysis of the creep behaviour of nuclear fuel-waste container materials

International Nuclear Information System (INIS)

Dutton, R.; Leitch, B.W.; Crosthwaite, J.L.; Kasprick, G.R.

1996-12-01

In the Canadian Nuclear Fuel Waste Management Program, it is proposed that nuclear fuel waste be placed in a durable container and disposed of in a deep underground vault. Consideration of various disposal-container designs has identified either titanium or copper as the material suitable for constructing the container shell. As part of the R and D program to examine the structural integrity of the container, creep tests are being conducted on commercially pure titanium and oxygen-free copper. This report presents the preliminary data obtained. It also describes the evaluation of various constitutive equations to represent the creep curves, thus providing the basis for extrapolation of the creep behaviour over the design lifetime of the container. In this regard, a specific focus is placed on equations derived from the 0-Projection Concept. Recognizing that the container lifetime will be determined by the onset of tertiary creep leading to creep rupture, we present the results of the metallographic examination of creep damage. This shows that the tertiary stage in titanium is associated with the formation of transgranular cavities within the region of localized necking of the creep specimens. In contrast, creep damage in copper is in the form of intergranular cavities uniformly distributed throughout the gauge length. These results are analyzed within the context of the extant literature, and their implications for future container design are discussed. (author)

Irradiation creep of various ferritic alloys irradiated at {approximately}400{degrees}C in the PFR and FFTF reactors

Energy Technology Data Exchange (ETDEWEB)

Toloczko, M.B.; Garner, F.A. [Pacific Northwest National Lab., Richland, WA (United States); Eiholzer, C.R. [Westinghouse Hanford Company, Richland, WA (United States)

1997-04-01

Three ferritic alloys were irradiated in two fast reactors to doses of 50 dpa or more at temperatures near 400{degrees}C. One martensitic alloy, HT9, was irradiated in both the FFTF and PFR reactors. PFR is the Prototype Fast Reactor in Dourneay, Scotland, and FFTF is the Fast Flux Test Facility in Richland, WA. D57 is a developmental alloy that was irradiated in PFR only, and MA957 is a Y{sub 2}O{sub 3} dispersion-hardened ferritic alloy that was irradiated only in FFTF. These alloys exhibited little or no void swelling at {approximately}400{degrees}C. Depending on the alloy starting condition, these steels develop a variety of non-creep strains early in the irradiation that are associated with phase changes. Each of these alloys creeps at a rate that is significantly lower than that of austenitic steels irradiated in the same experiments. The creep compliance for ferritic alloys in general appears to be {approximately}0.5 x 10{sup {minus}6} MPa{sup {minus}1} dpa{sup {minus}1}, independent of both composition and starting state. The addition of Y{sub 2}O{sub 3} as a dispersoid does not appear to change the creep behavior.

Numerical investigation of the reactor pressure vessel behaviour under severe accident conditions taking into account the combined processes of the vessel creep and the molten pool natural convection

International Nuclear Information System (INIS)

Loktionov, V.D.; Mukhtarov, E.S.; Yaroshenko, N.I.; Orlov, V.E.

1999-01-01

Analysis of the WWER lower head behaviour and its failure has been performed for several molten pool structures and internal overpressure levels in a reactor pressure vessel (RPV). The different types of the molten pools (homogeneous, conventionally homogeneous, conventionally stratified, stratified) cover the bounding scenarios during a hypothetical severe accident. The parametric investigations of the failure mode and RPV behaviour for various molten pool types, its heights and internal overpressure levels are presented herein. A coupled treatment in this investigation includes: (i) a 2-D thermohydraulic analysis of a molten pool natural convection. Domestic NARAUFEM code has been used in this detailed analysis for prediction of the heat flux from the molten pool to the RPV inner surface; and (ii) a detailed 3-D transient thermal analysis of the RPV lower head. Domestic 3-D ASHTER-VVR finite element code has been used for the numerical simulations of the high temperature creep and failure of the lower head. The effect of an external RPV cooling, temperature-dependent physical properties of the molten pool and vessel steel, the hydrostatic forces and vessel dead-weight were taken into account in this study. The obtained results show that lower head failure occurs as a result of the vessel creep process which is significantly dependent on both an internal overpressure level and the type of molten pool structure. In particular, it was found that there were combinations of 'overpressure-molten pool structure' when the vessel failure started at the 'hot' layers of the vessel. (orig.)

Examination of the creep behaviour of ceramic fuel elements under neutron irradiation

International Nuclear Information System (INIS)

Brucklacher, D.

1978-01-01

This paper examines the creeping of UO 2 , UO 2 -PuO 2 and UN under neutron irradiation. It starts with the experimental results about the relation between the thermal creep rate and the load, the temperature, as well as characteristic material values, stoichiometry, grain size and porosity. These correlation are first qualitatively discussed and then compared with the statements of actual quantitative equations. From the models and theories on which these equations are based a modified Nabarro-Heering-equation results for the correlation between the creep rate of ceramic fuels, stress, temperature and the fission rate. In the experimental part of the examination, length-changes of creep samples of UO 2 , (U,Pu)O 2 and UN were measured in specially developed irradiation creep casings in different reactors. The measuring data were corrected and evaluated considering the thermal expansion effects, irregular temperature distribution and swelling effects in such a way that the dependences of the creep rate of UO 2 , UO 2 -PuO 2 and UN under irradiation on stress, temperature, fission rate, burn-up and porosity is obtained. It shows that creeping of fuels under irradiation at high temperatures is equivalent to thermally activated creeping, while at low temperature the creep rate induced by irradiation is much higher than the condition without irradiation. The increment of oxidic nuclear fuels is greater than in UN, the stress dependence on low burn-up is proportional in both cases, and the influence of temperature is quite small. (orig.) [de

Creep crack growth behaviour of an AISI 316 steel plate for fast reactor structures

International Nuclear Information System (INIS)

D'Angelo, D.; Regis, V.

1985-01-01

The paper presents and analyses creep crack growth data obtained at 550, 600 and 650 0 C in air with SENT and CT specimens on type 316 stainless steel plate for LMFBR applications. Crack initiation and crack growth are tentatively correlated to K, sigmasub(net) and J* taking into account the constraint conditions due to specimen geometry. The validity of these parameters is discussed following the concept of transition time from small scale creep at the crack tip to extensive creep within the ligament. Post exposure microstructural and fractographic investigations do evidence that grain deformation processes are mainly responsible for cavity evolution. (orig.)

Analysis and description of the long-term creep behaviour of high-temperature gas turbine materials

International Nuclear Information System (INIS)

Bartsch, H.

1985-01-01

On a series of standard high-temperature gas turbine materials, creep tests were accomplished with the aim to obtain improved data on the long-term creep behaviour. The tests were carried out in the range of the main application temperatures of the materials and in the range of low stresses and elongations similar to operation conditions. They lasted about 5000 to 16000 h at maximum. At all important temperatures additional annealing tests lasting up to about 10000 h were carried out for the determination of a material-induced structure contraction. Thermal tension tests were effected for the description of elastoplastic short-time behaviour. As typical selection of materials the nickel investment casting alloys IN-738 LC, IN-939 and Udimet 500 for industrial turbine blades, IN-100 for aviation turbine blades and IN-713 C for integrally cast wheels of exhaust gas turbochargers were investigated, and also the nickel forge alloy Inconel 718 for industrial and aviation turbine disks and Nimonic 101 for industrial turbine blades and finally the cobalt alloy FSC 414 for guide blades and heat accumulation segments of industrial gas turbines. The creep tests were started on long-period individual creep testing machines with high strain measuring accuracy and economically continued on long-period multispecimen creep testing machines with long duration of test. The test results of this mixed test method were first subjected to a conventional evaluation in logarithmic time yield and creep diagrams which besides creep strength curves provided creep stress limit curves down to 0.2% residual strain. (orig./MM) [de

Low Temperature (320 deg C and 340 deg C) Creep Crack Growth in Low Alloy Reactor Pressure Vessel Steel

International Nuclear Information System (INIS)

Rui Wu; Sandstroem, Rolf; Seitisleam, Facredin

2004-02-01

Uni-axial creep and creep crack growth (CCG) tests at 320 deg C and 340 deg C as well as post test metallography have been carried out in a low alloy reactor pressure vessel steel (ASTM A508 class 2) having simulated coarse grained heat affected zone microstructure. The CCG behaviour is studied in terms of steady crack growth rate, creep fracture parameter C*, stress intensity factor and reference stress at given testing conditions. It has been found that CCG does occur at both tested temperatures. The lifetimes for the CCG tests are considerably shorter than those for the uni-axial creep tests. This is more pronounced at longer lifetimes or lower stresses. Increasing temperature from 320 deg C to 340 deg C causes a reduction of lifetime by approximately a factor of five and a corresponding increase of steady crack growth rate. For the CCG tests, there are three regions when the crack length is plotted against time. After incubation, the crack grows steadily until it accelerates when rupture is approached. Notable crack growth takes place at later stage of the tests. No creep cavitation is observed and transgranular fracture is dominant for the uni-axial creep specimens. In the CT specimens the cracks propagate intergranularly, independent of temperature and time. Some relations between time to failure, reference stress and steady crack growth rate are found for the CCG tests. A linear extrapolation based on the stress-time results indicates that the reference stress causing failure due to CCG at a given lifetime of 350,000 hours at 320 deg C is clearly lower than both yield and tensile strengths, on which the design stress may have based. Therefore, caution must be taken to prevent premature failure due to low temperature CCG. Both uni-axial and CCG tests on real welded joint at 320 deg C, study of creep damage zone at crack tip as well as numerical simulation are recommended for future work

Creep behaviour of ZrNb1 fuel cans in argon and steam

International Nuclear Information System (INIS)

Adam, E.; Stephan, M.; Wetzel, L.

1988-01-01

The paper is concerned with experimental investigations on the creep behaviour of fuel cans made of the ZrNb1 alloy. The isobaric-isothermal creep tests were performed in the range of temperatures from 990 K to 1290 K and with differential pressures over the can between 1.0 MPa and 2.5 MPa. They were characterized by linear heating of the test cans with 2 K/s until a given temperature was reached, followed by maintaining the cans at a constant temperature (Δ = ± 3 K) and loading it with purified argon produced internal pressure. The experiments were carried out in both an argon atmosphere surrounding the cans from outside and steam. (author)

Evaluation of creep-fatigue crack growth for large-scale FBR reactor vessel and NDE assessment

Energy Technology Data Exchange (ETDEWEB)

Joo, Young Sang; Kim, Jong Bum; Kim, Seok Hun; Yoo, Bong

2001-03-01

Creep fatigue crack growth contributes to the failure of FRB reactor vessels in high temperature condition. In the design stage of reactor vessel, crack growth evaluation is very important to ensure the structural safety and setup the in-service inspection strategy. In this study, creep-fatigue crack growth evaluation has been performed for the semi-elliptical surface cracks subjected to thermal loading. The thermal stress analysis of a large-scale FBR reactor vessel has been carried out for the load conditions. The distributions of axial, radial, hoop, and Von Mises stresses were obtained for the loading conditions. At the maximum point of the axial and hoop stress, the longitudinal and circumferential surface cracks (i.e. PTS crack, NDE short crack and shallow long crack) were postulated. Using the maximum and minimum values of stresses, the creep-fatigue crack growth of the proposed cracks was simulated. The crack growth rate of circumferential cracks becomes greater than that of longitudinal cracks. The total crack growth of the largest PTS crack is very small after 427 cycles. The structural integrity of a large-scale reactor can be maintained for the plant life. The crack depth growth of the shallow long crack is faster than that of the NDE short crack. In the ISI of the large-scale FBR reactor vessel, the ultrasonic inspection is beneficial to detect the shallow circumferential cracks.

Development of Bundle Position-Wise Linear Model for Predicting the Pressure Tube Diametral Creep in CANDU Reactors

International Nuclear Information System (INIS)

Lee, Jae Yong; Na, Man Gyun

2011-01-01

Diametral creep of the pressure tube (PT) is one of the principal aging mechanisms governing the heat transfer and hydraulic degradation of a heat transport system. PT diametral creep leads to diametral expansion that affects the thermal hydraulic characteristics of the coolant channels and the critical heat flux. Therefore, it is essential to predict the PT diametral creep in CANDU reactors, which is caused mainly by fast neutron irradiation, reactor coolant temperature and so forth. The currently used PT diametral creep prediction model considers the complex interactions between the effects of temperature and fast neutron flux on the deformation of PT zirconium alloys. The model assumes that long-term steady-state deformation consists of separable, additive components from thermal creep, irradiation creep and irradiation growth. This is a mechanistic model based on measured data. However, this model has high prediction uncertainty. Recently, a statistical error modeling method was developed using plant inspection data from the Bruce B CANDU reactor. The aim of this study was to develop a bundle position-wise linear model (BPLM) to predict PT diametral creep employing previously measured PT diameters and HTS operating conditions. There are twelve bundles in a fuel channel and for each bundle, a linear model was developed by using the dependent variables, such as the fast neutron fluxes and the bundle temperatures. The training data set was selected using the subtractive clustering method. The data of 39 channels that consist of 80 percent of a total of 49 measured channels from Units 2, 3 and 4 were used to develop the BPLM models. The remaining 10 channels' data were used to test the developed BPLM models. The BPLM was optimized by the maximum likelihood estimation method. The developed BPLM to predict PT diametral creep was verified using the operating data gathered from the Units 2,3 and 4 in Korea. Two error components for the BPLM, which are the epistemic

Creep-rupture-tests on thestainless steel X6 CrNi1811 (DIN 1.4948) in the frame of the ''Extrapolation-Program'' Pt. 2

International Nuclear Information System (INIS)

Solano, R.R.; Barroso, S.; Rivas, M. de las; Schirra, M.; Seith, B.

1979-01-01

The austenitic stainless steel X6 CrNi 1811 (DIN 1.4948) that is used as a structure material for the German Fast Breeder Reactor SNR 300 was creep tested in a temperature range of 550-650 deg under base material condition as well as welded material condition. The main point of this program (''Extrapolation-Program'') lies in the knowledge of the cree-rupture-strength and creep-behaviour up to 3X10 - 4 hours at higher temperatures in order to extrapolate up to (>=)10 5 hours for operating temperatures. In order to study the stress dependency of the minimum creep rate additional tests were carried out over temperature range 550 deg - 750 deg C. The present report describes the state in the running program with test-time up to 35.000 hours. Besides the cree-rupture behaviour it is possible to make a distinct quantitative statement for the creep-behaviour and ductility. Extensive metallographic examinations show the fracture behaviour and changes in structure. (author)

Metallurgical and environmental factors influencing creep behaviour of hastelloy-X

International Nuclear Information System (INIS)

Kiuchi, Kiyoshi; Kondo, Tatsuo

1979-03-01

Creep and rupture behaviours of Hastelloy-X and its modified version were examined with special reference to the effect of different test environments; i.e. air, high vacuum and the simulated HTR helium coolant. The respective environments showed different effects. The vacuum environment of about 10 -8 torr. gave best reproducible behaviour with essentially no surface-to-volume ratio effect. Such size effect was significant in the other two environments. The simulated HTR environment was characterized in its potentiality of both oxidizing selected alloy constituents and carburization. The observed behaviour was attributed to the depletion of strengthning solute elements caused by the surface reactions and the associated solid state reactions. (author)

Irradiation creep in zirconium single crystals

International Nuclear Information System (INIS)

MacEwen, S.R.; Fidleris, V.

1976-07-01

Two identical single crystals of crystal bar zirconium have been creep tested in reactor. Both specimens were preirradiated at low stress to a dose of about 4 x 10 23 n/m 2 (E > 1 MeV), and were then loaded to 25 MPa. The first specimen was loaded with reactor at full power, the second during a shutdown. The loading strain for both crystals was more than an order of magnitude smaller than that observed when an identical unirradiated crystal was loaded to the same stress. Both crystals exhibited periods of primary creep, after which their creep rates reached nearly constant values when the reactor was at power. During shutdowns the creep rates decreased rapidly with time. Electron microscopy revealed that the irradiation damage consisted of prismatic dislocation loops, approximately 13.5 nm in diameter. Cleared channels, identified as lying on (1010) planes, were also observed. The results are discussed in terms of the current theories for flux enhanced creep in the light of the microstructures observed. (author)

In-reactor creep of zirconium alloys by thermal spikes

International Nuclear Information System (INIS)

Ibrahim, E.F.

1975-01-01

The size and duration of thermal spikes from fast neutrons have been calculated for zirconium alloys, showing that spikes up to 1.8 nm radius may exist for 2 x 10 -11 s at greater than melting point, at 570K ambient temperature. Creep rates have been calculated assuming that the elastic strain from the applied stress relaxes in the volume of the spikes (by preferential loop alignment or modification of an existing dislocation network). The calculated rates are consistent with strain rates observed in long term tests-in-reactor, if spike lifetimes are 2 to 2.5 x 10 -11 s. (Auth.)

Creep-fatigue damage rules for advanced fast reactor design. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1997-03-01

The IAEA, following the recommendations of the International Working Group on Fast Reactors, convened a Technical Committee Meeting on Creep-Fatigue Damage Rules to be used in Fast Reactor Design. The objective of the meeting was to review developments in design rules for creep-fatigue conditions and to identify any areas in which further work would be desirable. The meeting was hosted by AEA Technology, Risley, and held in Manchester, United Kingdom, 11-13 June 1996. It was attended by experts from the European Commission, France, India, Japan, the Republic of Korea, the Russian Federation and the United Kingdom. Refs, figs, tabs

Reliability Evaluation on Creep Life Prediction of Alloy 617 for a Very High Temperature Reactor

International Nuclear Information System (INIS)

Kim, Woo-Gon; Hong, Sung-Deok; Kim, Yong-Wan; Park, Jae-Young; Kim, Seon-Jin

2012-01-01

This paper evaluates the reliability of creep rupture life under service conditions of Alloy 617, which is considered as one of the candidate materials for use in a very high temperature reactor (VHTR) system. A Z-parameter, which represents the deviation of creep rupture data from the master curve, was used for the reliability analysis of the creep rupture data of Alloy 617. A Service-condition Creep Rupture Interference (SCRI) model, which can consider both the scattering of the creep rupture data and the fluctuations of temperature and stress under any service conditions, was also used for evaluating the reliability of creep rupture life. The statistical analysis showed that the scattering of creep rupture data based on Z-parameter was supported by normal distribution. The values of reliability decreased rapidly with increasing amplitudes of temperature and stress fluctuations. The results established that the reliability decreased with an increasing service time.

Impression creep behaviour of Mod. 9Cr-1Mo steel weld joints

International Nuclear Information System (INIS)

Ridhin Raj, V.R.; Kottda, Ravi Sankar; Kamaraj, M.; Maduraimuthu, V.M.; Vasudevan, M.

2016-01-01

P91 steel (9Cr-1Mo) steel is extensively used in power plants for super heater coils, headers and steam piping. The aim of the present work is to study the creep behaviour of different zones of A-TIG weld joint using impression creep technique and compare it with that of the TIG weld joint. P91 steel weld joints were made by A-TIG welding without using any filler material and multi-pass TIG welding is done using ER90S-B9 filler rods. Welds were subjected to post-weld heat treatment (PWHT). Impression creep tests were carried out at 650 °C on the base metal, weld metal and HAZ regions. Optical Microscope and TEM were used to correlate microstructures with observed creep rates. The FGHAZ showed significantly higher impression creep rate compared to that of the base metal and weld metal. Fine grain size and relatively coarser M 23 C 6 carbide particles are responsible for higher creep rate. The impression creep rate of A-TIG weld metal and coarse grain HAZ was found to be lower than that of base metal. This is attributed to the higher grain size in weld metal and coarse HAZ attributed to the higher grain size in weld metal and to the higher peak temperature observed during A-TIG welding. (author)

Experimental study and modelling of high temperature creep flow and damage behaviour of 9Cr1Mo-NbV steel weldments

International Nuclear Information System (INIS)

Gaffard, V.

2004-12-01

Chromium martensitic stainless steels are under development since the 70's with the prospect of using them as structural components in thermal and nuclear power plants. The modified 9Cr1Mo-NbV steel is already used, especially in England and Japan, as a material for structural components in thermal power plants where welding is a commonly used joining technique. New generations of chromium martensitic stainless steels with improved mechanical properties for high pressure and temperature use are currently under development. However, observations of several in-service premature failures of welded components in 9Cr1Mo-NbV steel, outline a strong need for understanding the high temperature creep flow and damage behaviour of 9Cr1Mo-NbV steels and weldments. The present study aimed at experimentally determining and then modelling the high temperature creep flow and damage behaviour of both 9Cr1Mo-NbV steels and weldments (typically in the temperature range from 450 C to 650 C). The base metal was first studied as the reference material. It was especially evidenced that tempered chromium martensitic steels exhibit a change in both creep flow and damage behaviour for long term creep exposure. As a consequence, the classically performed extrapolation of 1,000 hours creep data to 100,000 hours creep lifetime predictions might be very hazardous. Based on experimental observations, a new model, integrating and coupling multiple creep flow and damage mechanisms, was developed in the framework of the mechanics of porous media. It was then successfully used to represent creep flow and damage behaviour of the base metal from high to low stress levels even for complex multiaxial loading conditions. Although the high temperature creep properties of the base metal are quite good, the occurrence of premature failure in weldments in high temperature creep conditions largely focused the attention of the scientific community. The lower creep strength of the weld component was also

Influence of pretreatment on creep-rupture-strength and creep-behaviour of a matrix-hardening Ni-base-alloy

International Nuclear Information System (INIS)

Schirra, M.

1982-01-01

The creep and time-to-rupture behaviour of the matrix hardening Nickel base alloy Inconel 625 was investigated in the temperature range 650-800 0 C. Three different thermo-mechanical pretreatment were used: I = Hot rolled finish; II = 870 0 C annealed; III = Sol. treatment 1150 0 C 1 h. The temperature range of this study is for samples which have undergone treatment I and II well above the temperatures normally used. The results show an anomalous stress dependence of creep and time-to-rupture at around 750 0 C. The reason is to be found in the very complex precipitation processes occurring while the stress is applied. The results are explained according to findings about precipitation in this type of alloy. (orig.) [de

Effects of bone damage on creep behaviours of human vertebral trabeculae.

Science.gov (United States)

O'Callaghan, Paul; Szarko, Matthew; Wang, Yue; Luo, Jin

2018-01-01

A subgroup of patients suffering with vertebral fractures can develop progressive spinal deformities over time. The mechanism underlying such clinical observation, however, remains unknown. Previous studies suggested that creep deformation of the vertebral trabeculae may play a role. Using the acoustic emission (AE) technique, this study investigated effects of bone damage (modulus reduction) on creep behaviours of vertebral trabecular bone. Thirty-seven human vertebral trabeculae samples were randomly assigned into five groups (A to E). Bones underwent mechanical tests using similar experimental protocols but varied degree of bone damage was induced. Samples first underwent creep test (static compressive stress of 0.4MPa) for 30min, and then were loaded in compression to a specified strain level (0.4%, 1.0%, 1.5%, 2.5%, and 4% for group A to E, respectively) to induce different degrees of bone damage (0.4%, no damage control; 1.0%, yield strain; 1.5%, beyond yield strain, 2.5% and 4%, post-ultimate strains). Samples were creep loaded (0.4MPa) again for 30min. AE techniques were used to monitor bone damage. Bone damage increased significantly from group A to E (P30% of modulus reduction in group D and E. Before compressive loading, creep deformation was not different among the five groups and AE hits in creep test were rare. After compressive loading, creep deformation was significantly greater in group D and E than those in other groups (Pcreep test were significantly greater in group D and E than in group A, B, and C (Pcreep deformation may occur even when the vertebra was under physiological loads. The boosted creep deformation observed may be attributed to newly created trabecular microfractures. Findings provide a possible explanation as to why some vertebral fracture patients develop progressive spinal deformity over time. Copyright © 2017. Published by Elsevier Inc.

Creep and stress rupture behaviour of zircaloy-2 and Zr-2.5% Nb alloy tubes at 573 K

International Nuclear Information System (INIS)

Laha, K.; Bhanu Sankara Rao, K.; Chandravathi, K.S.; Mannan, S.L.

1992-01-01

Zirconium alloys are extensively used for coolant tubes of pressurised heavy water reactors. The choice of these materials is based on their good corrosion resistance in water, low capture cross section for thermal neutrons and good mechanical properties. In this paper the results of an investigation performed on the creep and rupture behaviour of indigenously produced zircaloy-2 and Zr-2.5% Nb alloy are presented. Samples for creep testing were cut longitudinally from finished pressure tubes. Creep rupture tests were carried out in air under constant load conditions at 300 C employing five stress levels in the range 300-360 MPa. Zr-2.5% Nb alloy displayed higher rupture lives at all stress levels compared to zircaloy-2. Steady state creep rate of Zr-2.5%Nb was lower than that zircaloy-2 at identical stress levels. In the stress range of the experiments, the dependence of the steady state creep rate (ε s ) on applied stress (σ) for both the alloys could be represented by a power law, ε s =A σ n The stress sensitivity (n) for Zr-2.5% Nb was lower than that of zircaloy-2. For both the alloys the time to creep rupture t r was found related to the steady state creep rate through the modified Monkman-Grant relation (ε s ) α . t r = constant. Similar value of α was obtained for both the materials. Zr-2.5%Nb exhibited higher ductility (% elongation to rupture) compared to zircaloy-2 at stress levels ≥ 320 MPa. At lower stresses significant difference in ductility was not noticed. Percentage reduction in area was lower in Zr-2.5%Nb at all stress levels indicating better resistance for necking. The time for onset of tertiary was longer for Zr-2.5% Nb alloy. The proportion of life spent by Zr-2.5% Nb in steady state creep regime was higher compared to that of zircaloy-2. Metallographic investigations on longitudinal sections in both the alloys showed large number of intragranular pores close to the fracture surface. A few number of cracks which are characteristic of

Creep behavior and evolution of microstructure of modified Grade 91 welded joint after short term exposure at 500 deg C

International Nuclear Information System (INIS)

Vivier, F.

2009-03-01

With the increase in worldwide energy demand, the nuclear industry is a way of producing electricity on a large scale and to answer to this need. For the design of a new generation of fission nuclear reactors and among six chosen fission reactor systems, France develops in particularly the Very High Temperature Reactor (VHTR) concept. This implies the use of materials that are more and more resistant to high temperature for long-term exposure. AREVA focuses on materials already used in fossil-fuel power plant, so that the mechanical behaviour of Grade 91 (Fe 9 Cr 1 MoNbV) has to be investigated. This ferritic-martensitic steel is considered to be a potential candidate for welded components. Such structures are combined with welded joints, which have to be studied. Three industrial partners (AREVA, CEA, EDF) have launched a study with the Centre des Materiaux in order to investigate the creep of welded joint of Grade 91. The aim of this work is to complete the available database about the mechanical behaviour of Grade 91, base metal and welded joint, during creep tests performed at 500 C up to 4500 h exposure. Thermal aging tests, tensile tests, and creep tests were performed at 450 C and 500 C using both base metal and cross-weld samples. Several geometries of cross-weld creep specimens were tested. The microstructure has not remarkably changed after tests concerning both nature and size of precipitates, and the characteristic size of the matrix sub-structure. The creep damage is not developed in the ruptured specimens after creep tests. Only little damage by cavity nucleation and growth was found in the creep specimens. Creep fracture at 500 C takes places by viscoplastic flow, contrary to tests performed at 625 C where the creep-induced damage governs the creep rupture at least for long-term lifetime. From creep curves of base metal and cross-weld specimens, a phenomenological model is proposed. The flow rule is a Norton power law with a stress exponent of 19 in

Creep and Recovery Behaviour of Polyolefin-Rubber Nanocomposites Developed for Additive Manufacturing

Directory of Open Access Journals (Sweden)

Fugen Daver

2016-12-01

Full Text Available Nanocomposite application in automotive engineering materials is subject to continual stress fields together with recovery periods, under extremes of temperature variations. The aim is to prepare and characterize polyolefin-rubber nanocomposites developed for additive manufacturing in terms of their time-dependent deformation behaviour as revealed in creep-recovery experiments. The composites consisted of linear low density polyethylene and functionalized rubber particles. Maleic anhydride compatibilizer grafted to polyethylene was used to enhance adhesion between the polyethylene and rubber; and multi-walled carbon nanotubes were introduced to impart electrical conductivity. Various compositions of nanocomposites were tested under constant stress in creep and recovery. A four-element mechanistic Burger model was employed to model the creep phase of the composites, while a Weibull distribution function was employed to model the recovery phase of the composites. Finite element analysis using Abaqus enabled numerical modelling of the creep phase of the composites. Both analytical and numerical solutions were found to be consistent with the experimental results. Creep and recovery were dependent on: (i composite composition; (ii compatibilizers content; (iii carbon nanotubes that formed a percolation network.

In situ monitored in-pile creep testing of zirconium alloys

Science.gov (United States)

Kozar, R. W.; Jaworski, A. W.; Webb, T. W.; Smith, R. W.

2014-01-01

The experiments described herein were designed to investigate the detailed irradiation creep behavior of zirconium based alloys in the HALDEN Reactor spectrum. The HALDEN Test Reactor has the unique capability to control both applied stress and temperature independently and externally for each specimen while the specimen is in-reactor and under fast neutron flux. The ability to monitor in situ the creep rates following a stress and temperature change made possible the characterization of creep behavior over a wide stress-strain-rate-temperature design space for two model experimental heats, Zircaloy-2 and Zircaloy-2 + 1 wt%Nb, with only 12 test specimens in a 100-day in-pile creep test program. Zircaloy-2 specimens with and without 1 wt% Nb additions were tested at irradiation temperatures of 561 K and 616 K and stresses ranging from 69 MPa to 455 MPa. Various steady state creep models were evaluated against the experimental results. The irradiation creep model proposed by Nichols that separates creep behavior into low, intermediate, and high stress regimes was the best model for predicting steady-state creep rates. Dislocation-based primary creep, rather than diffusion-based transient irradiation creep, was identified as the mechanism controlling deformation during the transitional period of evolving creep rate following a step change to different test conditions.

Creep/fatigue damage prediction of fast reactor components using shakedown methods

International Nuclear Information System (INIS)

Buckthorpe, D.E.

1997-01-01

The present status of the shakedown method is reviewed, the application of the shakedown based principles to complex hardening and creep behaviour is described and justified and the prediction of damage against design criteria outlined. Comparisons are made with full inelastic analysis solutions where these are available and against damage assessments using elastic and inelastic design code methods. Current and future developments of the method are described including a summary of the advances made in the development of the post process ADAPT, which has enabled the method to be applied to complex geometry features and loading cases. The paper includes a review of applications of the method to typical Fast Reactor structural example cases within the primary and secondary circuits. For the primary circuit this includes structures such as the large diameter internal shells which are surrounded by hot sodium and subject to slow and rapid thermal transient loadings. One specific case is the damage assessment associated with thermal stratifications within sodium and the effects of moving sodium surfaces arising from reactor trip conditions. Other structures covered are geometric features within components such as the Above Core structure and Intermediate Heat Exchanger. For the secondary circuit the method has been applied to alternative and more complex forms of geometry namely thick section tubeplates of the Steam Generator and a typical secondary circuit piping run. Both of these applications are in an early stage of development but are expected to show significant advantages with respect to creep and fatigue damage estimation compared with existing code methods. The principle application of the method to design has so far been focused on Austenitic Stainless steel components however current work shows some significant benefits may be possible from the application of the method to structures made from Ferritic steels such as Modified 9Cr 1Mo. This aspect is briefly

Predicting diametral creep of the pressure tubes in CANDU reactors using fuzzy neural networks

International Nuclear Information System (INIS)

Lee, Jae Yong; Na, Man Gyun; Park, Jong Ho

2011-01-01

Pressure tube (PT) creep is one of the principal aging mechanisms governing the heat transfer and hydraulic degradation of the heat transport system (HTS) in Canada deuterium uranium reactors. PT diametral creep affects the thermal hydraulic characteristics of coolant channels and the critical heat flux (CHF). CHF is a key parameter in determining the critical channel power, which is used in the trip setpoint calculations of regional overpower protection systems. This paper aims to predict PT diametral creep using the measured signals of the HTS by applying fuzzy neural networks (FNNs) according to operating conditions. The FNN model was optimized in terms of its fuzzy rules and parameters by a genetic algorithm combined with the least-squares method. Informative data that demonstrate the system's characteristic behavior were selected to train the FNN model using the subtractive clustering method. The proposed FNN model for predicting PT diametral creep was verified using the operating data of the Wolsong Unit 1 nuclear power plant in Korea. It was known that the FNN could predict the PT diametral creep accurately. Statistical and analytical uncertainty analysis methods were applied to the models and their uncertainties were evaluated using 60 sampled training and optimization data sets, as well as two fixed test data sets. (author)

Creep-Fatigue Damage Evaluation of a Model Reactor Vessel and Reactor Internals of Sodium Test Facility according to ASME-NH and RCC-MRx Codes

International Nuclear Information System (INIS)

Lim, Dong-Won; Lee, Hyeong-Yeon; Eoh, Jae-Hyuk; Son, Seok-Kwon; Kim, Jong-Bum; Jeong, Ji-Young

2016-01-01

The objective of the STELLA-2 is to support the specific design approval for PGSFR by synthetic reviews of key safety issues and code validations through the integral effect tests. Due to its high temperature operation in SFRs (and in a testing facility) up to 550 °C, thermally induced creep-fatigue damage is very likely in components including a reactor vessel, reactor internals (interior structures), heat exchangers, pipelines, etc. In this study, structural integrity of the components such as reactor vessel and internals in STELLA-2 has been evaluated against creep-fatigue failures at a concept-design step. As 2D analysis yields far conservative results, a realistic 3D simulation is performed by a commercial software. A design integrity guarding against a creep-fatigue damage failure operating at high temperature was evaluated for the reactor vessel with its internal structure of the STELLA-2. Both the high temperature design codes were used for the evaluation, and results were compared. All the results showed the vessel as a whole is safely designed at the given operating conditions, while the ASME-NH gives a conservative evaluation

Creep-Fatigue Damage Evaluation of a Model Reactor Vessel and Reactor Internals of Sodium Test Facility according to ASME-NH and RCC-MRx Codes

Energy Technology Data Exchange (ETDEWEB)

Lim, Dong-Won; Lee, Hyeong-Yeon; Eoh, Jae-Hyuk; Son, Seok-Kwon; Kim, Jong-Bum; Jeong, Ji-Young [KAERI, Daejeon (Korea, Republic of)

2016-05-15

The objective of the STELLA-2 is to support the specific design approval for PGSFR by synthetic reviews of key safety issues and code validations through the integral effect tests. Due to its high temperature operation in SFRs (and in a testing facility) up to 550 °C, thermally induced creep-fatigue damage is very likely in components including a reactor vessel, reactor internals (interior structures), heat exchangers, pipelines, etc. In this study, structural integrity of the components such as reactor vessel and internals in STELLA-2 has been evaluated against creep-fatigue failures at a concept-design step. As 2D analysis yields far conservative results, a realistic 3D simulation is performed by a commercial software. A design integrity guarding against a creep-fatigue damage failure operating at high temperature was evaluated for the reactor vessel with its internal structure of the STELLA-2. Both the high temperature design codes were used for the evaluation, and results were compared. All the results showed the vessel as a whole is safely designed at the given operating conditions, while the ASME-NH gives a conservative evaluation.

Biaxial Creep Specimen Fabrication

Energy Technology Data Exchange (ETDEWEB)

JL Bump; RF Luther

2006-02-09

This report documents the results of the weld development and abbreviated weld qualification efforts performed by Pacific Northwest National Laboratory (PNNL) for refractory metal and superalloy biaxial creep specimens. Biaxial creep specimens were to be assembled, electron beam welded, laser-seal welded, and pressurized at PNNL for both in-pile (JOYO reactor, O-arai, Japan) and out-of-pile creep testing. The objective of this test campaign was to evaluate the creep behavior of primary cladding and structural alloys under consideration for the Prometheus space reactor. PNNL successfully developed electron beam weld parameters for six of these materials prior to the termination of the Naval Reactors program effort to deliver a space reactor for Project Prometheus. These materials were FS-85, ASTAR-811C, T-111, Alloy 617, Haynes 230, and Nirnonic PE16. Early termination of the NR space program precluded the development of laser welding parameters for post-pressurization seal weldments.

Biaxial Creep Specimen Fabrication

International Nuclear Information System (INIS)

JL Bump; RF Luther

2006-01-01

This report documents the results of the weld development and abbreviated weld qualification efforts performed by Pacific Northwest National Laboratory (PNNL) for refractory metal and superalloy biaxial creep specimens. Biaxial creep specimens were to be assembled, electron beam welded, laser-seal welded, and pressurized at PNNL for both in-pile (JOYO reactor, O-arai, Japan) and out-of-pile creep testing. The objective of this test campaign was to evaluate the creep behavior of primary cladding and structural alloys under consideration for the Prometheus space reactor. PNNL successfully developed electron beam weld parameters for six of these materials prior to the termination of the Naval Reactors program effort to deliver a space reactor for Project Prometheus. These materials were FS-85, ASTAR-811C, T-111, Alloy 617, Haynes 230, and Nirnonic PE16. Early termination of the NR space program precluded the development of laser welding parameters for post-pressurization seal weldments

Correlation between microstructure and the creep behaviour at high temperature of Alloy 800 H

International Nuclear Information System (INIS)

Spiradek, K.; Degischer, H.P.; Lahodny, H.

1989-01-01

A systematic metallographic study was performed to identify the nature of the microstructural changes occurring during high temperature creep deformation of Alloy 800 H. Creep tests were carried out at 800 deg. C under constant load conditions corresponding to the initial stresses between 25 and 80 MPa. Some tests were interrupted after certain elongations to provide the samples for electron microscopy. Emphasis was put on the creep periods relevant to design where only a few per cent of deformation are tolerable. The influence of the initial material conditions on the creep behaviour was examined. Variations of the initial microstructures were achieved by different solution treatments (980/1250) deg. C, preageing at 800 deg. C (0/6400) h and cold deformation up to 10% followed by ageing at 800 deg. C. The results of the microstructural examinations were correlated with the creep curves that provide a basis for identification of the creep mechanisms operating at the test conditions. (author). 14 refs, 17 figs

Design project of the experimental device for studying the uranium Creep in the reactor; Predprojekat eksperimentalnog uredjaja za ispitivanje CREEP-a urana u reaktoru

Energy Technology Data Exchange (ETDEWEB)

Pavicevic, M [Institute of Nuclear Sciences Boris Kidric, Odeljenje za reaktorsku eksperimentalnu tehniku, Vinca, Beograd (Serbia and Montenegro)

1963-12-15

The objective of this task was constructing a device for creep testing of uranium, i.e permanent deformation of the irradiated fuel. Deformation results from the influence of thermal neutron flux, temperature, time of irradiation, mechanical stress. This describes two possible technical solutions experimental device for creep testing in the vertical experimental channel and in the horizontal experimental channel of the RA reactor. In addition to the design details, the report covers calculations of heat generation, antireactivity, activation of the materials of the constructed experimental loop, mechanical calculations as well as description of measurements and regulation of the uranium sample temperature.

Thermal stresses and cyclic creep-fatigue in fusion reactor blanket

International Nuclear Information System (INIS)

Liu, K.C.

1977-01-01

Thermal stresses in the first walls of fusion reactor blankets were studied in detail. ORNL multibucket modules are emphasized. Practicality of using the bucket module rather than other blanket designs is examined. The analysis shows that applying intelligent engineering judgment in design can reduce the thermal stresses significantly. Arrangement of coolant flow and distribution of temperature are reviewed. Creep-fatigue property requirements for a first wall are discussed on the basis of existing design rules and criteria. Some major questions are pointed out and experiments needed to resolve basic uncertainties relative to key design decisions are discussed

Creep-rupture-test on the stainless steel X6crni1811 (Din 1.494.8) in the frame of the Extrapolation-Program. (Part III)

International Nuclear Information System (INIS)

Solano, R.; Schirra, M.; Rivas, M. de la; Barroso, S.; Seith, B.

1982-01-01

The austenitic stainless steel X6crni1811 (Din 1.4948) used as a structure material for the German Fast Breeder Reactor SNR 300 was creep tested in a temperature range of 550-650 degree centigree material condition as well as welded material condition. The main point of this program (Extrapolation-Program) lies in the knowledge of the creep-rupture-strength and creep-behaviour up to 3 x 10 4 hours higher temperatures in order to extrapolated up to ≥10 5 hours for operating temperatures. In order to study the stress dependency of the minimum creep rate additional tests were carried out of 550 degree centigree - 750 degree centigree. The present report describes the state in the running program with test-times of 23.000 hours and results from tests up to 55.000 hours belonging to other parallel programs are taken into account. Besides the creep-rupture behaviour it is also made a study of ductility between 550 and 750 degree centigree. Extensive metallographic examinations have been made to study the fracture behaviour and changes in structure. (Author)

Modification of creep and low cycle fatigue behaviour induced by welding

Directory of Open Access Journals (Sweden)

A. Carofalo

2014-10-01

Full Text Available In this work, the mechanical properties of Waspaloy superalloy have been evaluated in case of welded repaired material and compared to base material. Test program considered flat specimens on base and TIG welded material subjected to static, low-cycle fatigue and creep test at different temperatures. Results of uniaxial tensile tests showed that the presence of welded material in the gage length specimen does not have a relevant influence on yield strength and UTS. However, elongation at failure of TIG material was reduced with respect to the base material. Moreover, low-cycle fatigue properties have been determined carrying out tests at different temperature (room temperature RT and 538°C in both base and TIG welded material. Welded material showed an increase of the data scatter and lower fatigue strength, which was anyway not excessive in comparison with base material. During test, all the hysteresis cycles were recorded in order to evaluate the trend of elastic modulus and hysteresis area against the number of cycles. A clear correlation between hysteresis and fatigue life was found. Finally, creep test carried out on a limited number of specimens allowed establishing some changes about the creep rate and time to failure of base and welded material. TIG welded specimen showed a lower time to reach a fixed strain or failure when a low stress level is applied. In all cases, creep behaviour of welded material is characterized by the absence of the tertiary creep.

Collect Available Creep-Fatigue Data and Study Existing Creep-Fatigue Evaluation Procedures for Grade 91 and Hastelloy XR

International Nuclear Information System (INIS)

Asayama, Tai; Tachibana, Yukio

2007-01-01

This report describes the results of investigation on Task 5 of DOE/ASME Materials Project based on a contract between ASME Standards Technology, LLC (ASME ST-LLC) and Japan Atomic Energy Agency (JAEA). Task 5 is to collect available creep-fatigue data and study existing creep-fatigue evaluation procedures for Grade 91 steel and Hastelloy XR. Part I of this report is devoted to Grade 91 steel. Existing creep-fatigue data were collected (Appendix A) and analyzed from the viewpoints of establishing a creep-fatigue procedure for VHTR design. A fair amount of creep-fatigue data has been obtained and creep-fatigue phenomena have been clarified to develop design standards mainly for fast breeder reactors. Following this, existing creep-fatigue procedures were studied and it was clarified that the creep-fatigue evaluation procedure of the ASME-NH has a lot of conservatisms and they were analyzed in detail from the viewpoints of the evaluation of creep damage of material. Based on the above studies, suggestions to improve the ASME-NH procedure along with necessary research and development items were presented. Part II of this report is devoted to Hastelloy XR. Existing creep-fatigue data used for development of the high temperature structural design guideline for High Temperature Gas-cooled Reactor (HTGR) were collected. Creep-fatigue evaluation procedure in the design guideline and its application to design of the intermediate heat exchanger (IHX) for High Temperature Engineering Test Reactor (HTTR) was described. Finally, some necessary research and development items in relation to creep-fatigue evaluation for Gen IV and VHTR reactors were presented.

Orthotropic creep in polyethylene glycol impregnated archaeological oak from the Vasa ship - Results of creep experiments in a museum-like climate

Science.gov (United States)

Vorobyev, Alexey; van Dijk, Nico P.; Kristofer Gamstedt, E.

2018-02-01

Creep in archaeological oak samples and planks from the Vasa ship impregnated with polyethylene glycol (PEG) has been studied in museum-like climate. Creep studies of duration up to three years have been performed in nearly constant relative humidity and temperature of the controlled museum climate. Cubic samples were subjected to compressive creep tests in all orthotropic directions. Additionally, the creep behaviour of planks with and without PEG and of recent oak was tested in four-point bending. The experimental results have been summarised and also compared with reference results from recent oak wood. The effect of variable ambient conditions on creep and mass changes is discussed. The experimental results of creep in the longitudinal direction showed deformations even for the low stresses. There is relatively much more scatter in creep behaviour, and not all samples showed linear viscoelastic response. The creep in radial and tangential directions of the cubes and the plank samples showed a strong dependency on the ambient conditions. Some samples showed expansion for decreasing moisture content, possibly caused by the thermal expansion of the PEG component. For the planks, increasing creep deformation was observed induced by changing ambient conditions. Such behaviour may be related to e.g. oscillations in ambient conditions and presence of PEG in the wood cell wall and cell lumen. The behaviour of PEG archaeological wood depends on the level of deterioration that occurred over centuries. However, although the findings presented here apply to this specific case, they provide a unique view on such wood.

In-situ Creep Testing Capability Development for Advanced Test Reactor

Energy Technology Data Exchange (ETDEWEB)

B. G. Kim; J. L. Rempe; D. L. Knudson; K. G. Condie; B. H. Sencer

2010-08-01

Creep is the slow, time-dependent strain that occurs in a material under a constant strees (or load) at high temperature. High temperature is a relative term, dependent on the materials being evaluated. A typical creep curve is shown in Figure 1-1. In a creep test, a constant load is applied to a tensile specimen maintained at a constant temperature. Strain is then measured over a period of time. The slope of the curve, identified in the figure below, is the strain rate of the test during Stage II or the creep rate of the material. Primary creep, Stage I, is a period of decreasing creep rate due to work hardening of the material. Primary creep is a period of primarily transient creep. During this period, deformation takes place and the resistance to creep increases until Stage II, Secondary creep. Stage II creep is a period with a roughly constant creep rate. Stage II is referred to as steady-state creep because a balance is achieved between the work hardening and annealing (thermal softening) processes. Tertiary creep, Stage III, occurs when there is a reduction in cross sectional area due to necking or effective reduction in area due to internal void formation; that is, the creep rate increases due to necking of the specimen and the associated increase in local stress.

Effects of mechanical-thermal treatments on the creep behaviour of a niobium stabilized stainless steel

International Nuclear Information System (INIS)

Rossi, J.L.

1987-01-01

The influence of microstructural variables controlled by mechanical-thermal treatments on the creep behavior of DIN-Werkstoff Nr. 1,4981 stainless steel a material candidate for use as cladding of fast breeder reactor fuel elements, was studied. The effect of the solution treatment, predeformation, predeformation plus aging and cycles of predeformation-aging, on the creep results obtained at 990 K, for applied stresses in the range 70 MPa - 310 MPa, are analysed. The results show: this material presents a creep strength superior to that show by AISI 316 stainless steel; a transition on the creep behavior is observed at a certain stress; the mechanical-thermal treatments were seen to be ineffective on the improvement of the creep strength; the pre-deformation and pre-deformation plus aging treatments were seen to induce material embrittlement whereas the cyclic treatments induced increased ductility. Transmission electron microscopy, X ray diffraction of extracted precipitates, and microanalysis were use to characterize the microstructure of this material. (author)

Fatigue-creep of martensitic steels containing 9-12% Cr: behaviour and damage; Fatigue-fluage des aciers martensitiques a 9-12% Cr: comportement et endommagement

Energy Technology Data Exchange (ETDEWEB)

Fournier, B

2007-09-15

It is in the framework of the research programs on nuclear reactors (generation IV) that the martensitic steels containing 9-12% Cr are studied by the CEA. Most of the structures for which they are considered will be solicited in fatigue-creep at high temperature (550 C). The aim of this work is to understand and model the cyclic behaviour and the damage of these materials. The proposed modelling are based on detailed observations studies (SEM, TEM, EBSD...). The cyclic softening is attributed to the growth of the microstructure. A micro-mechanical model based on the physical parameters is proposed and leads to encouraging results. The damage results of interactions between fatigue, creep and oxidation. Two main types of damage are revealed. A model of anticipation of service time is proposed and gives very satisfying results. The possible extrapolations are discussed. (O.M.)

Creep-Data Analysis of Alloy 617 for High Temperature Reactor Intermediate Heat Exchanger

International Nuclear Information System (INIS)

Kim, Woo Gon; Ryu, Woo Seog; Kim, Yong Wan; Yin, Song Nan

2006-01-01

The design of the metallic components such as hot gas ducts, intermediate heat exchanger (IHX) tube, and steam reformer tubes of very high temperature reactor (VHTR) is principally determined by the creep properties, because an integrity of the components should be preserved during a design life over 30 year life at the maximum operating temperature up to 1000 .deg. C. For designing the time dependent creep of the components, a material database is needed, and an allowable design stress at temperature should be determined by using the material database. Alloy 617, a nicked based superalloy with chromium, molybdenum and cobalt additions, is considered as a prospective candidate material for the IHX because it has the highest design temperature. The alloy 617 is approved to 982 .deg. C (1800 .deg. F) and other alloys approved to 898 .deg. C (1650 .deg. C), such as alloy 556, alloy 230, alloy HX, alloy 800. Also, the alloy 617 exhibits the highest level of creep strength at high temperatures. Therefore, it is needed to collect the creep data for the alloy 617 and the creep-rupture life at the given conditions of temperature and stress should be predicted for the IHX construction. In this paper, the creep data for the alloy 617 was collected through literature survey. Using the collected data, the creep life for the alloy 617 was predicted based on the Larson-Miller parameter. Creep master curves with standard deviations were presented for a safety design, and failure probability for the alloy 617 was obtained with a time coefficient

The influence of thermomechanical treatment on the creep behaviour of DIN 1.4970 austenitic stainless steel at 973 K

International Nuclear Information System (INIS)

Zahra, A.A.A.; Schroeder, H.

1981-04-01

The creep-rupture behaviour of a Type DIN 1.4970 austenitic stainless steel has been investigated at 973 K (700 0 C) in a high vacuum for three conditions of thermomechanical treatment and a wide range of applied stresses. This type of steel is a candidate for use in the German SNR-300 Fast Breeder Reactor where it shall be used after a 13% cold-working treatment and subsequent aging at 1073 K (800 0 C) for 24 hours ( standard condition ). As an alternative, two other conditions were also investigated, namely aged at 1073 K (800 0 C) for 24 hours before the cold-working (condition 2) and cold worked only (condition 1). Because of various experimental efforts in this laboratory and elsewhere to study helium induced embrittlement effects in α-implanted foil specimens, all tests were performed using foil specimens of 105 μm thickness which were solution annealed at 1373 K (1100 0 C) before the above thermomechanical treatments were applied. The rupture lives and the minimum creep rates were found to be highly dependent on the applied stresses. The treatment of condition 1 material yielded a product as strong as the standard condition 3, while the condition 2 material was less creep resistant. Structural changes as well as fractography were followed using metallographic, transmission and scanning electron microscope techniques. Transgranular ductile fracture was clearly observed in all three conditions. TEM investigations showed that dispersive TiC precipitates were present in the matrix of condition 3 material before creep testing contrary to condition 1 and 2 material. In condition 1 the TiC dispersion was already found after short creep times, while no dispersive TiC precipitates were found in condition 2 material in every test condition. (orig.) [de

A methodology to analyze the creep behaviour of nuclear fuel waste containers

Energy Technology Data Exchange (ETDEWEB)

Dutton, R [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs.

1995-12-01

The concept for the disposal of used-fuel waste from CANDU reeactors operating in Canada comprises a system of natural and engineered barriers surrounding the waste in a mined vault situated at a depth of 500 - 1000 m in plutonic rock of the Canadian Shield. The fuel would be packaged in a highly durable metal container, within a matrix of compacted particulate. The design of the container takes into account that it would be subjected to an external hydrostatic pressure. Consideration of the rate of radioactive decay of the radionuclides contained in the fuel, suggests that the lifetime of the container should be at least 500 years. Consequently, the role of creep deformation, and the possibility of creep rupture of the container shell, must be included in the assessment of time-dependent mechanical integrity. This report describes an analytical approach that can be used to quantify the long-term creep properties of the container material and facilitate the engineering design. The overall objective is to formulate a constitutive creep equation that provides the required input for a finite element computer model being developed to analyze the elastic-plastic behaviour of the container. Alternative forms of such equations are reviewed. It is shown that the capability of many of these equations to extrapolate over long time scales is limited by their empirical nature. Thus, the recommended equation is based on current mechanistic understanding of creep deformation and creep rupture. A criterion for determining the onset of material failure by creep rupture, that could be used in the design of containers with extended structural integrity, is proposed. Interpretation and extrapolation will be supported by the complementary Deformation and Fracture Mechanism Maps. (author) 103 refs., 2 tabs., 54 figs.

A methodology to analyze the creep behaviour of nuclear fuel waste containers

International Nuclear Information System (INIS)

Dutton, R.

1995-12-01

The concept for the disposal of used-fuel waste from CANDU reeactors operating in Canada comprises a system of natural and engineered barriers surrounding the waste in a mined vault situated at a depth of 500 - 1000 m in plutonic rock of the Canadian Shield. The fuel would be packaged in a highly durable metal container, within a matrix of compacted particulate. The design of the container takes into account that it would be subjected to an external hydrostatic pressure. Consideration of the rate of radioactive decay of the radionuclides contained in the fuel, suggests that the lifetime of the container should be at least 500 years. Consequently, the role of creep deformation, and the possibility of creep rupture of the container shell, must be included in the assessment of time-dependent mechanical integrity. This report describes an analytical approach that can be used to quantify the long-term creep properties of the container material and facilitate the engineering design. The overall objective is to formulate a constitutive creep equation that provides the required input for a finite element computer model being developed to analyze the elastic-plastic behaviour of the container. Alternative forms of such equations are reviewed. It is shown that the capability of many of these equations to extrapolate over long time scales is limited by their empirical nature. Thus, the recommended equation is based on current mechanistic understanding of creep deformation and creep rupture. A criterion for determining the onset of material failure by creep rupture, that could be used in the design of containers with extended structural integrity, is proposed. Interpretation and extrapolation will be supported by the complementary Deformation and Fracture Mechanism Maps. (author) 103 refs., 2 tabs., 54 figs

Design project of the experimental facility for testing uranium creep in the reactor; Predprojekat eksperimentalnog uredjaja za ispitivanje CREEP-a urana u reaktoru

Energy Technology Data Exchange (ETDEWEB)

Pavlovic, A [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1963-12-15

This report contains the design for constructing the experimental device for testing metal uranium creep in the RA reactor core under defined neutron flux conditions, temperature, mechanical loads and time of irradiation. This device will be placed in one of the experimental channels in the core. This report contains physical, thermal and mechanical calculations and engineering drawings of the device.

Comparative study of creep behaviour in three Cr Ni 15/15 steel stabilized with Ti and with different contents in W, Mn, Mo and Bor

International Nuclear Information System (INIS)

Solano, R.R.; Rivas, M. de las; Schirra, M.; Seith, B.

1975-01-01

The main difference between the three steels which are tested at temperature range from 650 0 C to 750 0 C is due to the hardening elements pf the matrix and the Boron content: 1. 12R72HV (X10NiCrMoTiB 1515) 2% Mn 1,5% Mo 80 ppmB 2. Vaccutherm (X12CrNiWTi 1613) 3% W 2,5 ppmB 3. RGT 21 (X12CrNiWTi 1613) 3% W 50 ppm B. The investigations of all casts are carried out in two different heat treatments which are suitable for the conditions required for the operation of the reactor. Cond. I: 1150 0 C 30 min, water quenced; 800 0 32 hour, air; 10% cold work. Cond. II: 1150 0 C 30 min, water quenched; 10% cold work. In connection with creep test the condition I irrespective of 3 steels show no remarkable difference. The observation at 750 0 C test temperature and also at condition II above 650 0 C on Boron-free Vaccutherm cast shows an unfavourable behaviour. There is no significant difference in the stress dependence of secondary creep rate and also absolute creep rate. A definite superiority is to be found for 12R72HV when considering the values for time-yield-limit-ratio and ductility compared to the W-steels. The test results shows different fracture behaviour. Transcrystalline fracture is found on cast 12R72HV, whereas RGT 21 and Vaccutherm show transition from transcrystalline to intercrystalline fracture, depending on the rupture time and test temperature. The long term rupture specimens show intercrystalline fracture. (author)

The influence of Boron on creep-rupture behaviour of austenitic unstabilized and Nb-stabilized stainless steel X8CrNi 1613 in unirradiated and irradiated condition

International Nuclear Information System (INIS)

Sen, Susant Kumar.

1976-10-01

The present study deals with influence of boron on creep-rupture behaviour in unirradiated condition at 650 0 C along with precipitation behaviour, heat-treatment and recrystallization of unstabilized and stabilized steel. The results of creep-rupture tests on unirradiated specimens show that boron exerts a beneficial effect on the rupture life and ductility. Boron losses its beneficial effect on creep properties in unstabilized steel by prolong creeping. The magnitude of beneficial effect of Boron on creep properties depends upon the initial boron distribution which influences the number, size and distribution of the precipitates. Boron promotes the precipitation of type M 23 C 6 Carbides in the grain as well as at the grain boundary. Boron segregates in atomic form during slow cooling from austenitizing temperature. The recrystallization will be delayed by the presence of boron. The results of creep tests at 650 0 C shows that boron exerts a beneficial effect on creep life of irradiated steels. (orig./GSC) [de

Mechanical characterization of superalloys for space reactors

International Nuclear Information System (INIS)

Duchesne, J.

1989-01-01

The purpose of this work is the choice of materials usable between 600 and 900 0 C for nuclear space reactor structures. The main criterion of selection for these materials is their good creep behaviour. Consequently, macroscopic theories of creep and several extrapolation methods were described. Superalloys seem the best materials for the studied range of temperatures. Five of them, base nickel, ones unusual in nuclear industry were selected for their good mechanical properties. Three of them are industrial alloys: the first, HAYNES 230 is a recent one, HASTELLOY S and X are more standard materials. The last two, HASTELLOY XR and PYRAD 38 D are issued from special fabrications. Creep tests metallographic investigations, hardness and tensile tests were performed. A contraction of samples was observed during some creep tests under a low stress, 20MPa at 800 0 C, for HAYNES 230 and HASTELLOY X. This could be due to a structural evolution of these materials connected to a decrease of the cristalline parameter. In addition, correlations were observed between certain characteristics determined from slow tensile tests and short duration creep tests. These correlations present a large interest because, at the present time, creep tests cannot be executed on irradiated materials in our laboratories. Consequently creep behaviour of irradiated materials seem may be deduced. Further studies are needed to explain and confirm the behaviour of the most interesting materials under low stresses: HAYNES 230 and HASTELLOY XR to anticipate their behaviour in working conditions [fr

Microstructural characterisation and constitutive behaviour of alloy RR1000 under fatigue and creep-fatigue loading conditions

International Nuclear Information System (INIS)

Stoecker, C.; Zimmermann, M.; Christ, H.-J.; Zhan, Z.-L.; Cornet, C.; Zhao, L.G.; Hardy, M.C.; Tong, J.

2009-01-01

Mechanical behaviour of a nickel-based superalloy, RR1000, has been investigated at 650 deg. C under cyclic and dwell loading conditions. The microstructural characteristics of the alloy have been studied using scanning electron microscopy (SEM) and transmission electron microscopy (TEM), and the distribution patterns of the dislocations and slip planes have been compared between samples tested under fatigue and creep-fatigue loading conditions. Constitutive behaviour of the alloy was described by a unified constitutive model, where both cyclic plastic and viscoplastic strains were represented by one inelastic strain. The results show that the precipitation state is very stable at 650 deg. C and only minor differences exist in the dislocation arrangements formed under pure fatigue and combined creep and fatigue conditions. Hence, a unified constitutive model seems to be justified in describing and predicting the constitutive behaviour in both cases.

Creep behaviour of the alloys NiCr22Co12Mo and 10CrMo9 10 under static and cyclic load conditions

International Nuclear Information System (INIS)

Wolf, H.

1990-01-01

The creep behaviour of NiCr20Co12Mo is investigated under static strain and at 800deg C, with stresses applied ranging from 105 MPa to 370 MPa. The ferritic steel 10CrMo 9 10 is tested for its creep behaviour under static strain and at the temperatures of 600deg C and 550deg C, with stresses applied between 154 MPa and 326 MPa (at 600deg C), or between 250 MPa and 458 MPa (at 550deg C). The experiments are made to determine the effects of changes in strain on the materials' deformation behaviour, placing emphasis on transient creep and elastic or anelastic response. The mean internal stress is determined from changes in strain. Cyclic creep is analysed as a behaviour directly responding to the pattern of change in strain. Effects of certain strain changes not clarified so far are analysed. The cyclic strain experiments are analysed according to the velocity factor concept. The usual models of creep deformation (theta projection concept) are compared with the model of effective strain, which is based on the fundamental equation of plastic deformation by dislocation motion (Orowan equation). (MM) [de

High-temperature transient creep properties of CANDU pressure tubes

International Nuclear Information System (INIS)

Fong, R.W.L.; Chow, C.K.

2002-06-01

During a hypothetical large break loss-of-coolant accident (LOCA), the coolant flow would be reduced in some fuel channels and would stagnate and cause the fuel temperature to rise and overheat the pressure tube. The overheated pressure tube could balloon (creep radially) into contact with its moderator-cooled calandria tube. Upon contact, the stored thermal energy in the pressure tube is transferred to the calandria tube and into the moderator, which acts as a heat sink. For safety analyses, the modelling of fuel channel deformation behaviour during a large LOCA requires a sound knowledge of the high-temperature creep properties of Zr-2.5Nb pressure tubes. To this extent, a ballooning model to predict pressure-tube deformation was developed by Shewfelt et al., based on creep equations derived using uniaxial tensile specimens. It has been recognized, however, that there is an inherent variability in the high-temperature creep properties of CANDU pressure tubes. The variability, can be due to different tube-manufacturing practices, variations in chemical compositions, and changes in microstructure induced by irradiation during service in the reactor. It is important to quantify the variability of high-temperature creep properties so that accurate predictions on pressure-tube creep behaviour can be made. This paper summarizes recent data obtained from high-temperature uniaxial creep tests performed on specimens taken from both unirradiated (offcut) and irradiated pressure tubes, suggesting that the variability is attributed mainly to the initial differences in microstructure (grain size, shape and preferred orientation) and also from tube-to-tube variations in chemical composition, rather than due to irradiation exposure. These data will provide safety analysts with the means to quantify the uncertainties in the prediction of pressure-tube contact temperatures during a postulated large break LOCA. (author)

The influence of mechanical deformation on the irradiation creep of AISI 316 stainless steel irradiated in the EBR-II and FFTF fast reactors

International Nuclear Information System (INIS)

Garner, F.A.; Gilbert, E.R.

2007-01-01

Irradiation creep of stainless steels is thought not to be very responsive to material and environmental variables. To test this perception earlier unpublished experiments conducted in the EBR-II reactor on AISI 316 have been analyzed. While swelling is dependent on the cold-work level at 400-480 o C, the post-transient irradiation creep rate, often called the creep compliance B0, is not dependent on cold-work level. If the tube reaches pressures on reactor start-up that generate above-yield stresses in unirradiated steel, then plastic strains occur prior to significant irradiation, but the post-transient strain rate is identical to that of material that did not exceed the yield stress on start-up. It is shown that both stress-free and stress-affected swelling are isotropic and that the Soderberg relationship is maintained. At temperatures above ∼540 o C thermal creep and stored energy begin to assert themselves, with creep rates accelerating with cold-work and becoming non-linear with stress. These results are in agreement with a similar study on titanium-modified 316 steel in FFTF. (author)

Comparative study of the creep behaviour of single crystals and polycrystals of alpha uranium

International Nuclear Information System (INIS)

Andre, J.P.

1964-03-01

In the first chapter, one describes the creep machine developed to study the deformation of uranium at high temperature in vacuum with a continuous recording. The second chapter presents the results concerning the polycrystals of uranium. The application of the DORN method gives an activation energy for creep of 42 ± 2 Kc, above 550 Celsius degrees, equal to the activation energy for self-diffusion. The study of the variation of the creep rate with the applied stress and the metallographic observations of the deformation induced polygonization allow to conclude that the deformation is controlled by climb of dislocations. In the third chapter, the deformation above 550 Celsius degrees of single crystals of uranium (obtained by β → α change) is studied. The major deformation mode is slip. The preexisting polygonization of these single crystals is very stable and the disorientation between adjacent sub-grains increases with the deformation. The activation energy for creep is higher than that for polycrystals. These results show the influence of the polygonization due to the β → α change on the creep behaviour of α uranium. (authors) [fr

The creep-fatigue crack growth behaviour of a 1CrMoV rotor steel

International Nuclear Information System (INIS)

Priest, R.H.; Miller, D.A.; Gladwin, D.N.; Maguire, J.

1989-01-01

Crack growth rates under simultaneous creep-fatigue conditions have been quantified for a 1CrMoV rotor steel. Measured growth rates were partitioned into cyclic and hold period contributions and these characterized by the relevant fracture mechanics parameters K and C. Cyclic growth rates measured in the creep-fatigue tests were enhanced compared with pure fatigue rates. This observation is compared with the behaviour of other steels and explained by quantitative metallography. The resulting crack growth equation can be used during integrity assessments for plant components containing cracks which are subject to thermal fatigue

Laser interferometer system for the measurement of creep in pressurized tubes

International Nuclear Information System (INIS)

Kirchner, T.L.

1976-07-01

A laser interferometer measurement system was developed to measure the length, diameter, and radius of various pressurized tube specimens. The machine measures and records profilometric data of the pressurized tubes prior to insertion in the reactor and then again after a predetermined fluence has been reached to determine the amount of creep which has occurred. This data provides a statistical basis for the description of steady-state in-reactor creep and creep rupture behavior of the reference fuel cladding and structural materials for the Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor (CRBR). In addition, this data will be used to determine the relative in-reactor creep and creep rupture behavior of candidate alloys for advanced cladding and structural materials. The laser interferometer system, referred to as the Biaxial Creep Measurement Machine (BCMM), was built to meet or exceed design criteria such as: automatic measurement of the five biaxial creep specimens varying in size; complete automation of the machine using a mini-computer; complete specimen loading, unloading, and data processing in less than five minutes; storage of data on magnetic cassette tapes; quick-look data readout and error checking during each run to determine proper machine operation; and remote operation in a radioactive environment

Creep behaviour of heat resistant steels. Pt. 2

International Nuclear Information System (INIS)

Kloos, K.H.; Granacher, J.; Oehl, M.

1993-01-01

Creep data scatter bands of steels 2.25 Cr-1 Mo and 12 Cr-1 Mo-0.3 V were evaluated with the aid of model functions based on time temperature parameters. From the times to reach given strain values, mean isostrain curves in the stress time diagramme were calculated and therefrom, mean creep curves were derived. On this basis, creep equations were established, which include primary-, secondary- and tertiary-creep and are valid in the main range of application of each steel. Further, mean stress strain curves from hot tensile tests were used to describe the initial plastic strain in the creep equations. The values calculated with the established creep equations agreed relatively well with the correspondent original scatter band values from the creep tests. (orig.) [de

A planar model study of creep in metal matrix composites with misaligned short fibres

DEFF Research Database (Denmark)

Sørensen, N.J.

1993-01-01

The effect of fibre misalignment on the creep behaviour of metal matrix composites is modelled, including hardening behaviour (stage 1), dynamic recovery and steady state creep (stage 2) of the matrix material, using an internal variable constitutive model for the creep behaviour of the metal...... matrix. Numerical plane strain results in terms of average properties and detailed local deformation behaviour up to large strains are needed to show effects of fibre misalignment on the development of inelastic strains and the resulting over-all creep resistance of the material. The creep resistance...

Influence of flowing sodium on creep deformation and rupture behaviour of 316L(N) austenitic stainless steel

Energy Technology Data Exchange (ETDEWEB)

Ravi, S., E-mail: sravi@igcar.gov.in [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Laha, K.; Mathew, M.D. [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Vijayaraghavan, S.; Shanmugavel, M.; Rajan, K.K. [Fast Reactor Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Jayakumar, T. [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India)

2012-08-15

The influence of flowing sodium on creep deformation and rupture behaviour of AISI 316L(N) austenitic stainless steel has been investigated at 873 K over a stress range of 235-305 MPa. The results were compared with those obtained from testing in air environment. The steady state creep rates of the material were not influenced appreciably by the testing environments. The time to onset of tertiary stage of creep deformation was delayed in sodium environment. The creep-rupture lives of the material increased in sodium environment, which became more pronounced at lower applied stresses. The increase in rupture life of the material in flowing sodium was accompanied by an increase in rupture ductility. The creep damage on specimen surface as well as inside the specimen was less in specimen tested in sodium. SEM fractographic investigation revealed predominantly transgranular dimple failure for the specimen tested in sodium, whereas predominantly intergranular creep failure was observed in the air tested specimens. Almost no oxidation was observed in the specimens creep tested in the sodium environment. Absence of oxidation and less creep damage cavitation extended the secondary state in liquid sodium tests and lead to increase in creep rupture life and ductility of the material as compared to in air.

Creep and creep rupture properties of cladding tube (type 316) in high temperature sodium

International Nuclear Information System (INIS)

Atsumo, H.

1977-01-01

The thin walled small sized seamless AISI 316 steel tubes, which are designated to be domestically used as the fuel cladding tube for sodium cooled fast breeder reactors in Japan, are irradiated in the following sodium of high temperature in the range of 370 deg. C to 700 deg. C, and receive gradually increased internal pressure caused by the fission produced gas generating from the nuclear fuel burn-up inside the cladding tube. Consequently, the creep behavior of fuel cladding tubes under a high temperature sodium environment is an important problem which must be determined and clarified together with their characteristic features under irradiation and in air. In relation to the creep performance of fuel cladding tubes made of AISI 316 steel and other comparable austenitic stainless steels, hardly any studies are found that are made systematically to examine the effect of sodium with sodium purity as parameter or any comparative studies with in-air data at various different temperatures. The present research work was aimed to obtain certain basic design data relating to in-sodium creep performance of the domestic made fuel cladding tubes for fast breeder reactors, and also to gain further date as considered necessary under several sodium conditions. That is, together with establishment of the technology for tensile creep test and internal pressure creep rupture test in flowing sodium of high temperature, a series of tests and studies were performed on the trial made cladding tubes of AISI Type-316 steel. In the first place, two kinds of purity conditions of sodium, close to the actual reactor-operating condition, (oxygen concentration of 10 ppm and 5 ppm respectively) were established, and then uniaxial tensile creep test and rupture test under various temperatures were performed and the resulting data were compared and evaluated against the in-air data. Then, secondly, an internal pressure creep rupture test was conducted under a single purity sodium environment

Accelerator-Based Irradiation Creep of Pyrolytic Carbon Used in TRISO Fuel Particles for the (VHTR) Very High Temperature Reactors

International Nuclear Information System (INIS)

Wang, Lumin; Was, Gary

2010-01-01

Pyrolytic carbon (PyC) is one of the important structural materials in the TRISO fuel particles which will be used in the next generation of gas-cooled very-high-temperature reactors (VHTR). When the TRISO particles are under irradiation at high temperatures, creep of the PyC layers may cause radial cracking leading to catastrophic particle failure. Therefore, a fundamental understanding of the creep behavior of PyC during irradiation is required to predict the overall fuel performance.

Influence of creep ductility on creep-fatigue behaviour of 20%Cr/25%Ni/Nb stainless steel

International Nuclear Information System (INIS)

Gladwin, D.; Miller, D.A.

1985-01-01

The influence of creep ductility on creep-fatigue endurance of 20%Cr/25%Ni/Nb stainless steel has been examined. In order to induce different creep ductilities in the 20/25/Nb stainless steel, three different thermo-mechanical routes were employed. These resulted in a range of ductilities (3-36%) being obtained at the strain rates of interest. Strain controlled slow-fast creep-fatigue cycles were used with strain rates of 10 -6 s -1 , 10 -7 s -1 in tension and 10 -3 s -1 in compression. It was found that creep ductility strongly influenced the creep-fatigue endurance of the 20/25/Nb stainless steel. When failure was creep dominated endurance was found to be directly proportional to the creep ductility. A ductility exhaustion model has been used to successfully predict creep-fatigue endurance when failure was creep dominated. (author)

Project accent: graphite irradiated creep in a materials test reactor

International Nuclear Information System (INIS)

Brooking, M.

2014-01-01

Atkins manages a pioneering programme of irradiation experiments for EDF Energy. One of these projects is Project ACCENT, designed to obtain evidence of a beneficial physical property of the graphite, which may extend the life of the Advanced Gas-cooled Reactors (AGRs). The project team combines the in-house experience of EDF Energy with two supplier organisations (providing the material test reactors and testing facilities) and supporting consultancies (Atkins and an independent technical expert). This paper describes: - Brief summary of the Project; - Discussion of the challenges faced by the Project; and - Conclusion elaborating on the aims of the Project. These challenging experiments use bespoke technology and both un-irradiated (virgin) and irradiated AGR graphite. The results will help to better understand graphite irradiation-induced creep (or stress modified dimensional change) properties and therefore more accurately determine lifetime and safe operating envelopes of the AGRs. The first round of irradiation has been completed, with a second round about to commence. This is a key step to realising the full lifetime ambition for AGRs, demonstrating the relaxation of stresses within the graphite bricks. (authors)

Effect of titanium on the creep deformation behaviour of 14Cr-15Ni-Ti stainless steel

Science.gov (United States)

Latha, S.; Mathew, M. D.; Parameswaran, P.; Nandagopal, M.; Mannan, S. L.

2011-02-01

14Cr-15Ni-Ti modified stainless steel alloyed with additions of phosphorus and silicon is a potential candidate material for the future cores of Prototype Fast Breeder Reactor. In order to optimise the titanium content in this steel, creep tests have been conducted on the heats with different titanium contents of 0.18, 0.23, 0.25 and 0.36 wt.% at 973 K at various stress levels. The stress exponents indicated that the rate controlling deformation mechanism was dislocation creep. A peak in the variation of rupture life with titanium content was observed around 0.23 wt.% titanium and the peak was more pronounced at lower stresses. The variation in creep strength with titanium content was correlated with transmission electron microscopic investigations. The peak in creep strength exhibited by the material with 0.23 wt.% titanium is attributed to the higher volume fraction of fine secondary titanium carbide (TiC) precipitates.

Effect of titanium on the creep deformation behaviour of 14Cr-15Ni-Ti stainless steel

Energy Technology Data Exchange (ETDEWEB)

Latha, S. [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu 603 102 (India); Mathew, M.D., E-mail: mathew@igcar.gov.in [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu 603 102 (India); Parameswaran, P.; Nandagopal, M. [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu 603 102 (India); Mannan, S.L. [National Engineering College, Kovilpatti, Tamil Nadu 628 503 (India)

2011-02-28

14Cr-15Ni-Ti modified stainless steel alloyed with additions of phosphorus and silicon is a potential candidate material for the future cores of Prototype Fast Breeder Reactor. In order to optimise the titanium content in this steel, creep tests have been conducted on the heats with different titanium contents of 0.18, 0.23, 0.25 and 0.36 wt.% at 973 K at various stress levels. The stress exponents indicated that the rate controlling deformation mechanism was dislocation creep. A peak in the variation of rupture life with titanium content was observed around 0.23 wt.% titanium and the peak was more pronounced at lower stresses. The variation in creep strength with titanium content was correlated with transmission electron microscopic investigations. The peak in creep strength exhibited by the material with 0.23 wt.% titanium is attributed to the higher volume fraction of fine secondary titanium carbide (TiC) precipitates.

Effect of titanium on the creep deformation behaviour of 14Cr-15Ni-Ti stainless steel

International Nuclear Information System (INIS)

Latha, S.; Mathew, M.D.; Parameswaran, P.; Nandagopal, M.; Mannan, S.L.

2011-01-01

14Cr-15Ni-Ti modified stainless steel alloyed with additions of phosphorus and silicon is a potential candidate material for the future cores of Prototype Fast Breeder Reactor. In order to optimise the titanium content in this steel, creep tests have been conducted on the heats with different titanium contents of 0.18, 0.23, 0.25 and 0.36 wt.% at 973 K at various stress levels. The stress exponents indicated that the rate controlling deformation mechanism was dislocation creep. A peak in the variation of rupture life with titanium content was observed around 0.23 wt.% titanium and the peak was more pronounced at lower stresses. The variation in creep strength with titanium content was correlated with transmission electron microscopic investigations. The peak in creep strength exhibited by the material with 0.23 wt.% titanium is attributed to the higher volume fraction of fine secondary titanium carbide (TiC) precipitates.

Effect of carbon activity on the creep behaviour of 21/4Cr, 1Mo steel in sodium

International Nuclear Information System (INIS)

Cordwell, J.E.; Charnock, W.; Nicholson, R.D.

1979-02-01

The creep endurance and creep cracking behaviour of 2 1/4Cr, 1Mo steel in sodium at 475 0 C have been studied at three different sodium carbon activities. Creep endurance was found to increase with increasing carbon activity of the sodium. Tests carried out in high carbon activity sodium were discontinued before fracture. Creep crack initiation displacement at notches decreased with increasing carbon activity, presumably as a result of notch tip carburisation. The plastic zones at the tips of blunt notches in specimens exposed in high carbon activity sodium were preferentially carburised. These observations were similar to those made previously on 9Cr, 1Mo steel. One difference detected metallographically was that in a high carburising environment uniform carburisation was obtained in the 2 1/4Cr, 1Mo steel specimens whereas carburisation gradients were observed in the 9Cr, 1Mo steel. Creep crack propagation rates for given notch opening displacement rates in low and intermediate carbon activity sodium were indistinguishable. However, the strenthening that resulted from the mild carburisation of the specimen in the intermediate carbon activity sodium caused slower notch opening displacement rates and crack propagation rates than in the low carbon activity sodium, when the rates were compared at the same crack length. (author)

Long-term creep test with finite elements

International Nuclear Information System (INIS)

Argyris, J.H.; Szimmat, J.; Willam, K.J.

1975-01-01

Following a presentation of concrete creep, a brief summary of the direct and incremental calculation methods for long-term creep behaviour is given. In addition, a survey on the methods of the inner state variables is given which, on the one hand, gives a uniform framework for the various formulations of concrete creep, and on the other hand leads to a computable calculation method. Two examples on long-term creep behaviour illustrate the application field of the calculation method. (orig./LH) [de

Influence of manufacturing process on the in-reactor creep anisotropy of stress-relieved Zircaloy-2 cladding

International Nuclear Information System (INIS)

Shann, S.H.; Van Swam, L.F.

1995-01-01

A procedure to determine the axial/radial and circumferential/radial contractile strain ratios (the R and P factors respectively in the Backofen-modified von Mises-Hill yield criterion) from post-irradiation dimensional measurements of Zircaloy-2 cladding of BWR fuel rods, tie rods and water rods was developed and has been described previously (S.H. Shann and L.F. van Swam, Creep anisotropy of Zircaloy-2 cladding during irradiation, Trans. SMiRT-11, Vol. C, 1991). The present study employs the procedure to determine the anisotropy factors R and P for textured cold-worked stress-relieved (CWSR) Zircaloy-2 cladding fabricated by various manufacturing processes. The analysis indicates that the cladding manufacturing process can have a pronounced effect on the anisotropy of irradiation-induced creep. Cladding types with identical yield and ultimate tensile strengths but fabricated by different manufacturing processes have different values of R and P during in-reactor creep. ((orig.))

Experimental and analytical studies on creep failure of reactor coolant piping

Energy Technology Data Exchange (ETDEWEB)

Maeda, Akio; Maruyama, Yu; Hashimoto, Kazuichiro; Harada, Yuhei; Shibazaki, Hiroaki; Kudo, Tamotsu; Hidaka, Akihide; Sugimoto, Jun [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Nakamura, N.

1999-07-01

Thermal and structural responses of reactor coolant piping under and elevated internal pressure and temperature are being investigated in WIND project at JAERI. In a recent failure test in which a nuclear grade type 316 stainless steel pipe with an outer diameter of 114.3 mm and a wall thickness of 13.5 mm was used and an internal pressure was kept at approximately 15 MPa. A failure of the piping was observed when the temperature was sustained at 970degC for one hour. In parallel with conducting the tests, post-test analyses were performed. The objective of the analyses is to assess analytical models for the creep deformation and failure of the piping at elevated internal pressure and temperature simulating thermal-hydraulic conditions during a severe accident. The major material properties needed for the analysis were measured at elevated temperatures. Coefficients of a creep constitutive equation including the tertiary stage were determined with the measured creep data and incorporated into ABAQUS code. The analysis reasonably reproduced the time history of the enlargement of the piping diameter, and the wall thickness and the diameter of the piping at the failure. It was also found that the piping failure timing obtained from the analysis agreed well with the test result. (author)

Experimental and analytical studies on creep failure of reactor coolant piping

International Nuclear Information System (INIS)

Maeda, Akio; Maruyama, Yu; Hashimoto, Kazuichiro; Harada, Yuhei; Shibazaki, Hiroaki; Kudo, Tamotsu; Hidaka, Akihide; Sugimoto, Jun; Nakamura, N.

1999-01-01

Thermal and structural responses of reactor coolant piping under and elevated internal pressure and temperature are being investigated in WIND project at JAERI. In a recent failure test in which a nuclear grade type 316 stainless steel pipe with an outer diameter of 114.3 mm and a wall thickness of 13.5 mm was used and an internal pressure was kept at approximately 15 MPa. A failure of the piping was observed when the temperature was sustained at 970degC for one hour. In parallel with conducting the tests, post-test analyses were performed. The objective of the analyses is to assess analytical models for the creep deformation and failure of the piping at elevated internal pressure and temperature simulating thermal-hydraulic conditions during a severe accident. The major material properties needed for the analysis were measured at elevated temperatures. Coefficients of a creep constitutive equation including the tertiary stage were determined with the measured creep data and incorporated into ABAQUS code. The analysis reasonably reproduced the time history of the enlargement of the piping diameter, and the wall thickness and the diameter of the piping at the failure. It was also found that the piping failure timing obtained from the analysis agreed well with the test result. (author)

A phenomenological theory of transient creep

International Nuclear Information System (INIS)

Ajaja, O.; Ardell, A.J.

1979-01-01

A new creep theory is proposed which takes into account the strain generated during the annihilation of dislocations. This contribution is found to be very significant when recovery is appreciable, and is mainly responsible for the decreasing creep rate associated with the normal primary creep of class II materials. The theory provides excellent semiquantitative rationalization for the types of creep curves presented in the preceding paper. In particular, the theory predicts a change in the shape of the primary creep curve from normal to inverted as recovery becomes less important, i.e. as the applied stress and/or temperature decrease(s). It also predicts a minimum creep rate under certain circumstances, hence pseudo-tertiary behaviour. These different types of creep curves are predicted even though the net dislocation density decreases monotonically with time in all cases. Qualitative rationalization is presented for the inverted transient which always follows a stress drop in class II materials, as well as for the inverted primary and sigmoidal creep behaviour of class I solid solutions. (author)

The investigation of expanded polystyrene creep behaviour

Directory of Open Access Journals (Sweden)

Zhukov Aleksey

2017-01-01

Full Text Available The results obtained in long-term testing under constant compressive stress of the cut from the Slabs EPS 50/100 and EPS 150 with the density ranging from 15 to 24 kg/m3, which were manufactured by the same manufacturer by foaming EPS solid granules (beads in closed volume. The creep strain of the above described specimens was used as a criterion for estimating the deformability of the EPS slabs under long-term compressive stress. It was measured using special stands EN 1606, maintaining constant stress during the fixed time interval tn=122 days. Creep strains were determined by the methods described in EN 1606 for constant stress σc=0.35σ10% (compressive stress σ10% was determined in accordance with EN 826:2013. The long-term compressive stress measurement error did not exceed 1 %, while the creep strain measurement error was not larger than 0,005 mm. The tests were conducted at the ambient temperature of (23±2°С and relative humidity of (50±5 %.The long-term constant compressive load σc=0.35σ10%. The method of mathematical and statistical experimental design optimization models taking into account the thickness of specimens is proposed to determine the creep compliance Ic (tn the creep strain εc (tn and predictive point estimate of creep strain εc (T. Graphical interpretation of the abstained models is also presented. It should be noted that the abstained equations may be used in practice for estimating the creep strains at time tn=122 days and predictive estimates of εc (T for the load time of 10 years.

Model-based Approach for Long-term Creep Curves of Alloy 617 for a High Temperature Gas-cooled Reactor

International Nuclear Information System (INIS)

Kim, Woo Gon; Yin, Song Nan; Kim, Yong Wan

2008-01-01

Alloy 617 is a principal candidate alloy for the high temperature gas-cooled reactor (HTGR) components, because of its high creep rupture strength coupled with its good corrosion behavior in simulated HTGR-helium and its sufficient workability. To describe a creep strain-time curve well, various constitutive equations have been proposed by Kachanov-Rabotnov, Andrade, Garofalo, Evans and Maruyama, et al.. Among them, the K-R model has been used frequently, because a secondary creep resulting from a balance between a softening and a hardening of materials and a tertiary creep resulting from an appearance and acceleration of the internal or external damage processes are adequately considered. In the case of nickel-base alloys, it has been reported that a tertiary creep at a low strain range may be generated, and this tertiary stage may govern the total creep deformation. Therefore, a creep curve for nickel-based Alloy 617 will be predicted appropriately by using the K-R model that can reflect a tertiary creep. In this paper, the long-term creep curves for Alloy 617 were predicted by using the nonlinear least square fitting (NLSF) method in the K-R model. The modified K-R model was introduced to fit the full creep curves well. The values for the λ and K parameters in the modified K-R model were obtained with stresses

Creep equations for gas turbine materials

International Nuclear Information System (INIS)

Kloos, K.H.; Granacher, J.; Preussler, T.

1988-01-01

The long-term high-temperature deformation behaviour of typical gas turbine materials can be described on the basis of a differentiated evaluation which takes the results from thermal tension tests, short-term creep tests with continuous extension measurement, long-term creep tests with discontinuous extension measurement as well as annealing tests with contraction measurement into account. By this, especially the 'negative creeping' can be controlled. Equations were developed for individual materials of the type IN-738 LC, IN-939, IN-100 and FSX-414, which describe the high-temperature deformation behaviour with consideration to the primary and secondary creeping and partly the tertiary creeping. The equations are valid in the entire application-relevant range, i.e. up to 100 000 h in the case of industrial turbine materials. (orig.) [de

Creep and swelling of Type 348 stainless steel at temperatures up to 700 K and comparison with fast reactor data

International Nuclear Information System (INIS)

Beeston, J.M.; Thomas, L.E.

1982-01-01

In-reactor creep and swelling of Type 248 stainless steel from ATR SN-5 and ETR H-10 in-pile tube measurements were investigated to identify and characterize their mechanistic relationships at temperatures less than 723 0 K. The principal creep mechanism appears to be diffusion along high conductivity paths related to interstitial loops. The irradiation creep is a function of temperature and is presented as an empirical equation. The swelling in the ATR in-pile tubes is also presented as an empirical equation

Irradiation Creep in Graphite

Energy Technology Data Exchange (ETDEWEB)

Ubic, Rick; Butt, Darryl; Windes, William

2014-03-13

An understanding of the underlying mechanisms of irradiation creep in graphite material is required to correctly interpret experimental data, explain micromechanical modeling results, and predict whole-core behavior. This project will focus on experimental microscopic data to demonstrate the mechanism of irradiation creep. High-resolution transmission electron microscopy should be able to image both the dislocations in graphite and the irradiation-induced interstitial clusters that pin those dislocations. The team will first prepare and characterize nanoscale samples of virgin nuclear graphite in a transmission electron microscope. Additional samples will be irradiated to varying degrees at the Advanced Test Reactor (ATR) facility and similarly characterized. Researchers will record microstructures and crystal defects and suggest a mechanism for irradiation creep based on the results. In addition, the purchase of a tensile holder for a transmission electron microscope will allow, for the first time, in situ observation of creep behavior on the microstructure and crystallographic defects.

Irradiation-induced creep in 316 and 304L stainless steels

International Nuclear Information System (INIS)

Walters, L.C.; McVay, G.L.; Hudman, G.D.

1977-01-01

Recent results are presented from the in-reactor creep experiments that are being conducted by Argonne National Laboratory. The experiments consist of four subassemblies that contain helium-pressurized as well as unstressed capsules of 316 and 304L stainless steels in several metallurgical conditions. Experiments are being irradiated in row 7 of the EBR-II sodium-cooled fast breeder reactor. Three of the subassemblies are being irradiated at temperatures near 400 0 C, and the fourth subassembly is being irradiated at a temperature of 550 0 C. Creep and swelling strains were determined by profilometer measurements on the full length of the capsules after each irradiation cycle. The accumulated neutron dose on the 304L capsules at 385 0 C was 45 dpa; on the 316 capsules at 400 0 C, 40 dpa; and on the 316 capsules at 550 0 C, 25 dpa. It was found that the in-reactor creep rates were linearly dependent on hoop stress, with the exception being capsules of 316 stainless steel that had been given long-term carbide aging treatment and then irradiated at 550 0 C. Those capsules exhibited much higher creep and swelling rates than their unaged counterparts. For the metallurgical conditions where significant swelling was observed (solution-annealed 304L and aged 316 stainless steels), it was found that the in-reactor creep rates were readily fit to a model that related the creep rates to accumulated swelling. Additionally, it was found that the stress-normalized creep rate for 20%-cold-worked 316 stainless steel at a temperature of 550 0 C was 1.6 times that observed at 400 0 C

Fast reactor fuel design and development

International Nuclear Information System (INIS)

Bishop, J.F.W.; Chamberlain, A.; Holmes, J.A.G.

1977-01-01

Fuel design parameters for oxide and carbide fast reactor fuels are reviewed in the context of minimising the total uranium demands for a combined thermal and fast reactor system. The major physical phenomena conditioning fast reactor fuel design, with a target of high burn-up, good breeding and reliable operation, are characterised. These include neutron induced void swelling, irradiation creep, pin failure modes, sub-assembly structural behaviour, behaviour of defect fuel, behaviour of alternative fuel forms. The salient considerations in the commercial scale fabrication and reprocessing of the fuels are reviewed, leading to the delineation of possible routes for the manufacture and reprocessing of Commercial Reactor fuel. From the desiderata and restraints arising from Surveys, Performance and Manufacture, the problems posed to the Designer are considered, and a narrow range of design alternatives is proposed. The paper concludes with a consideration of the development areas and the conceptual problems for fast reactors associated with those areas

Creep behaviour of a polymer-based underground support liner

Science.gov (United States)

Guner, Dogukan; Ozturk, Hasan

2017-09-01

All underground excavations (tunnels, mines, caverns, etc.) need a form of support to ensure that excavations remain safe and stable for the designed service lifetime. In the last decade, a new support material, thin spray-on liner (TSL) has started to take place of traditional underground surface supports of bolts and shotcrete. TSLs are generally cement, latex, polymer-based and also reactive or non-reactive, multi-component materials applied to the rock surface with a layer of few millimeter thickness. They have the advantages of low volume, logistics, rapid application and low operating cost. The majority of current TSLs are two-part products that are mixed on site before spraying onto excavation rock surfaces. Contrary to the traditional brittle supports, the high plastic behaviour of TSLs make them to distribute the loads on larger lining area. In literature, there is a very limited information exist on the creep behavior of TSLs. In this study, the creep behavior of a polymer-based TSL was investigated. For this purpose, 7-day cured dogbone TSL specimens were tested under room temperature and humidity conditions according to ASTM-D2990 creep testing standard. A range of dead weights (80, 60, 40, and 20 % of the tensile strength) were applied up to 1500 hours. As a result of this study, the time-dependent strain behavior of a TSL was presented for different constant load conditions. Moreover, a new equation was derived to estimate tensile failure time of the TSL for a given loading condition. If the tensile stress acting on the TSL is known, the effective permanent support time of the TSL can be estimated by the proposed relationship.

Creep-rupture-test on the stainless steel X6CRNI1811 (DIN 1.4948) in the frame of the ''Extrapolation-Program''. (Part III)

International Nuclear Information System (INIS)

Solano, R.; Las Rivas, M. de; Barroso, S.

1982-01-01

The austenitic stainless steel X6CrNi1811 (DIN 1.4948) used as a structure material for the German Fast Breeder Reactor SNR 300 was creep tested in a temperature range of 550-650 deg under base material condition as well as welded material condition. The main point of this program (''Extrapolation-Program'') lies in the knowledge of the creep-rupture-strength and creepbehaviour up to 3 x 10 4 hours at higher temperatures in order to extrapolate up to >=10 5 hours for operating temperatures. In order to study the stress dependency of the minimum creep rate additional tests were carried out of 550 deg - 750 deg C. The present report describes the state in the running program with test-times of 23.000 hours and results from tests up to 55.000 hours belonging to other parallel programs are taken into account. Besides the creep-rupture behaviour it is also made a study of ductility between 550 and 750 deg C. Extensive metallographic examinations have been made to study the fracture behaviour and changes in structure. (author)

Effect of small cold forming on the creep behaviour of gas turbine blades made of Nimonic 90

International Nuclear Information System (INIS)

Keienburg, K.H.; Krueger, H.; Pickert, U.; Bautz, G.

1987-01-01

In order to obtain information on the material behaviour of Nimonic 90 with and without cold forming at the main temperature of use of 560deg C for large gas turbine blades, creep and relaxation samples were taken from the large volume foot of a gas turbine blade, part of which were tensioned by 3% cold in a tensile test machine. The selected cold forming was obtained as the upper limit from DMS measurements on a gas turbine blade when aligning. The negative effect of cold forming on the creep behaviour known from the literature for other γ hardened nickel base alloys was confirmed. The grain (matrix) is strengthened and the grain boundary is simultaneously weakened by cold forming. The material is also sensitized, so that fine separation occurs in the grain at the sliding bands and at the grain boundaries. Both circumstances contribute to the worsening of the creep behaviour, significantly for stresses below the technical elastic limit in the cold formed state. It follows, relative to large gas turbine blades, that: 1) Aligning operations must be restricted to the absolute minimum necessary and should be avoided completely if possible. 2) Aligned blades should be subjected to renewed solution annealing and separation hardening. 3.) Blades deformed in operation should also be subjected to renewed complete heat treatment. (orig.) [de

Irradiation-induced creep in graphite: a review

International Nuclear Information System (INIS)

Price, R.J.

1981-08-01

Data on irradiation-induced creep in graphite published since 1972 are reviewed. Sources include restrained shrinkage tests conducted at Petten, the Netherlands, tensile creep experiments with continuous strain registration at Petten and Grenoble, France, and controlled load tests with out-of-reactor strain measurement performed at Oak Ridge National Laboratory, Petten, and in the United Kingdom. The data provide reasonable confirmation of the linear viscoelastic creep model with a recoverable transient strain component followed by a steady-state strain component, except that the steady-state creep coefficient must be treated as a function of neutron fluence and is higher for tensile loading than for compressive loading. The total transient creep strain is approximately equal to the preceding elastic strain. No temperature dependence of the transient creep parameters has been demonstrated. The initial steady-state creep coefficient is inversely proportional to the unirradiated Young modulus

Irradiation Creep and Swelling of Russian Ferritic-Martensitic Steels Irradiated to Very High Exposures in the BN-350 Fast Reactor at 305-335 degrees C

International Nuclear Information System (INIS)

Konobeev, Yury V.; Dvoriashin, Alexander M.; Porollo, S.I.; Shulepin, S.V.; Budylkin, N.I.; Mironova, Elena G.; Garner, Francis A.

2003-01-01

Russian ferritic/martensitic (F/M) steels EP-450, EP-852 and EP-823 were irradiated in the BN-350 fast reactor in the form of gas-pressurized creep tubes. The first steel is used in Russia for hexagonal wrappers in fast reactors. The other steels were developed for compatibility with Pb-Bi coolants and serve to enhance our understanding of the general behavior of this class of steels. In an earlier paper we published data on irradiation creep of EP-450 and EP-823 at temperatures between 390 and 520C, with dpa levels ranging from 20 to 60 dpa. In the current paper new data on the irradiation creep and swelling of EP-450 and EP-852 at temperatures between 305 and 335C and doses ranging from 61 to 89 dpa are presented. Where comparisons are possible, it appears that these steels exhibit behavior that is very consistent with that of Western steels. Swelling is relatively low at high neutron exposure and confined to temperatures <420C, but may be camouflaged somewhat by precipitation-related densification. These irradiation creep studies confirm that the creep compliance of F/M steels is about one-half that of austenitic steels.

Recent Advances in Creep Modelling of the Nickel Base Superalloy, Alloy 720Li.

Science.gov (United States)

Harrison, William; Whittaker, Mark; Williams, Steve

2013-03-20

Recent work in the creep field has indicated that the traditional methodologies involving power law equations are not sufficient to describe wide ranging creep behaviour. More recent approaches such as the Wilshire equations however, have shown promise in a wide range of materials, particularly in extrapolation of short term results to long term predictions. In the aerospace industry however, long term creep behaviour is not critical and more focus is required on the prediction of times to specific creep strains. The current paper illustrates the capability of the Wilshire equations to recreate full creep curves in a modern nickel superalloy. Furthermore, a finite-element model based on this method has been shown to accurately predict stress relaxation behaviour allowing more accurate component lifing.

Transitions in creep mechanisms and creep anisotropy in Zr-1Nb-1Sn-0.2Fe sheet

International Nuclear Information System (INIS)

Murty, K.L.; Ravi, J.; Wiratmo

1995-01-01

The creep characteristics of a Zr-1Nb-1Sn-0.2Fe alloy sheet were investigated at temperatures from 773 to 923K and at stresses ranging from 9 to 150MPa along both the rolling and transverse directions. Transitions in creep mechansims are noted, with diffusional viscous creep at low stresses, viscous-glide-controlled microcreep in the intermediate stress regime and the climb of edge dislocations at high stresses. The creep anisotropy decreases with a decrease in the stress exponent and the creep rates differ by only 30% in the viscous creep regime, while an order-of-magnitude difference is noted at high stresses. The solute-strengthening effect of Nb addition is evident in the stress regime where appropriate data are available. These transitions in creep mechansims clearly reveal the dangers in blind extrapolation of short-term high stress data to low stresses and long times relevant to in-reactor conditions. The creep behavior of these materials is similar to that noted in Class I alloys, while the transitions in deformation mechanisms in Zircaloy-4 resemble those found in pure metals or Class II alloys with no viscous glide mechanism. ((orig.))

Effects of composition on the in-reactor creep of AISI 316

International Nuclear Information System (INIS)

Bates, J.F.; Gilbert, E.R.

1980-01-01

Pre- and postirradiation measurements of pressurized tube specimens irradiated at 450/degree/C to 4.6*10/sup 22/ n/cm/sup 2/(E>0.1 MeV) have indicated that increases in the solute concentrations of silicon, phosphorus, and molybdenum retard irradiation creep. The data suggest that carbon and nitrogen act synergistically with the major influence on creep being the nitrogen concentration. Irradiation-induced creep is insensitive to cobalt variations. There is a trend for specimens with higher swelling to exhibit higher creep. As the shear modulus increases, irradiation creep also increases. This shear modulus correlation is opposite to one observed for thermal creep deformation. 8 refs

Creep and Creep-Fatigue Crack Growth at Structural Discontinuities and Welds

Energy Technology Data Exchange (ETDEWEB)

Dr. F. W. Brust; Dr. G. M. Wilkowski; Dr. P. Krishnaswamy; Mr. Keith Wichman

2010-01-27

The subsection ASME NH high temperature design procedure does not admit crack-like defects into the structural components. The US NRC identified the lack of treatment of crack growth within NH as a limitation of the code and thus this effort was undertaken. This effort is broken into two parts. Part 1, summarized here, involved examining all high temperature creep-fatigue crack growth codes being used today and from these, the task objective was to choose a methodology that is appropriate for possible implementation within NH. The second part of this task, which has just started, is to develop design rules for possible implementation within NH. This second part is a challenge since all codes require step-by-step analysis procedures to be undertaken in order to assess the crack growth and life of the component. Simple rules for design do not exist in any code at present. The codes examined in this effort included R5, RCC-MR (A16), BS 7910, API 579, and ATK (and some lesser known codes). There are several reasons that the capability for assessing cracks in high temperature nuclear components is desirable. These include: (1) Some components that are part of GEN IV reactors may have geometries that have sharp corners - which are essentially cracks. Design of these components within the traditional ASME NH procedure is quite challenging. It is natural to ensure adequate life design by modeling these features as cracks within a creep-fatigue crack growth procedure. (2) Workmanship flaws in welds sometimes occur and are accepted in some ASME code sections. It can be convenient to consider these as flaws when making a design life assessment. (3) Non-destructive Evaluation (NDE) and inspection methods after fabrication are limited in the size of the crack or flaw that can be detected. It is often convenient to perform a life assessment using a flaw of a size that represents the maximum size that can elude detection. (4) Flaws that are observed using in-service detection

Creep and Creep-Fatigue Crack Growth at Structural Discontinuities and Welds

International Nuclear Information System (INIS)

Brust, F.W.; Wilkowski, G.M.; Krishnaswamy, P.; Wichman, Keith

2010-01-01

The subsection ASME NH high temperature design procedure does not admit crack-like defects into the structural components. The US NRC identified the lack of treatment of crack growth within NH as a limitation of the code and thus this effort was undertaken. This effort is broken into two parts. Part 1, summarized here, involved examining all high temperature creep-fatigue crack growth codes being used today and from these, the task objective was to choose a methodology that is appropriate for possible implementation within NH. The second part of this task, which has just started, is to develop design rules for possible implementation within NH. This second part is a challenge since all codes require step-by-step analysis procedures to be undertaken in order to assess the crack growth and life of the component. Simple rules for design do not exist in any code at present. The codes examined in this effort included R5, RCC-MR (A16), BS 7910, API 579, and ATK (and some lesser known codes). There are several reasons that the capability for assessing cracks in high temperature nuclear components is desirable. These include: (1) Some components that are part of GEN IV reactors may have geometries that have sharp corners - which are essentially cracks. Design of these components within the traditional ASME NH procedure is quite challenging. It is natural to ensure adequate life design by modeling these features as cracks within a creep-fatigue crack growth procedure. (2) Workmanship flaws in welds sometimes occur and are accepted in some ASME code sections. It can be convenient to consider these as flaws when making a design life assessment. (3) Non-destructive Evaluation (NDE) and inspection methods after fabrication are limited in the size of the crack or flaw that can be detected. It is often convenient to perform a life assessment using a flaw of a size that represents the maximum size that can elude detection. (4) Flaws that are observed using in-service detection

Prediction and Monitoring Systems of Creep-Fracture Behavior of 9Cr-1Mo Steels for Reactor Pressure Vessels

International Nuclear Information System (INIS)

Potirniche, Gabriel; Barlow, Fred D.; Charit, Indrajit; Rink, Karl

2013-01-01

A recent workshop on next-generation nuclear plant (NGNP) topics underscored the need for research studies on the creep fracture behavior of two materials under consideration for reactor pressure vessel (RPV) applications: 9Cr-1Mo and SA-5XX steels. This research project will provide a fundamental understanding of creep fracture behavior of modified 9Cr-1Mo steel welds for through modeling and experimentation and will recommend a design for an RPV structural health monitoring system. Following are the specific objectives of this research project: Characterize metallurgical degradation in welded modified 9Cr-1Mo steel resulting from aging processes and creep service conditions; Perform creep tests and characterize the mechanisms of creep fracture process; Quantify how the microstructure degradation controls the creep strength of welded steel specimens; Perform finite element (FE) simulations using polycrystal plasticity to understand how grain texture affects the creep fracture properties of welds; Develop a microstructure-based creep fracture model to estimate RPVs service life; Manufacture small, prototypic, cylindrical pressure vessels, subject them to degradation by aging, and measure their leak rates; Simulate damage evolution in creep specimens by FE analyses; Develop a model that correlates gas leak rates from welded pressure vessels with the amount of microstructural damage; Perform large-scale FE simulations with a realistic microstructure to evaluate RPV performance at elevated temperatures and creep strength; Develop a fracture model for the structural integrity of RPVs subjected to creep loads; and Develop a plan for a non-destructive structural health monitoring technique and damage detection device for RPVs.

The in-reactor deformation of the PCA alloy

International Nuclear Information System (INIS)

Puigh, R.J.

1986-04-01

The swelling and in-reactor creep behaviors of the PCA alloy have been determined from the irradiation of pressurized tube specimens in the FFTF reactor. These data have been obtained to a peak neutron fluence corresponding to approximately 80 dpa in the FFTF reactor for irradiation temperatures between 400 and 750 0 C. Diametral measurements performed on the unstressed specimens indicate the possible onset of swelling in the PCA alloy for irradiation temperatures between 400 and 550 0 C and at a neutron fluence corresponding to ∼50 dpa. The creep data suggest a non-linear fluence dependence and linear stress dependence (for hoop stresses less than 100 MPa) which is consistent with the in-reactor creep behavior of many cold worked austenitic stainless steels. These PCA creep data are compared to available 316 SS in-reactor creep data. The in-reactor creep strains for PCA are significantly less than observed in 20% cold worked 316 SS over the temperature ranges and fluences investigated

High-throughput design of low-activation, high-strength creep-resistant steels for nuclear-reactor applications

Energy Technology Data Exchange (ETDEWEB)

Lu, Qi; Zwaag, Sybrand van der [Novel Aerospace Materials Group, Faculty of Aerospace Engineering, Delft University of Technology, Kluyverweg 1, 2629 HS, Delft (Netherlands); Xu, Wei, E-mail: xuwei@ral.neu.edu.cn [State Key Laboratory of Rolling and Automation, Northeastern University, 110819, Shenyang (China); Novel Aerospace Materials Group, Faculty of Aerospace Engineering, Delft University of Technology, Kluyverweg 1, 2629 HS, Delft (Netherlands)

2016-02-15

Reduced-activation ferritic/martensitic steels are prime candidate materials for structural applications in nuclear power reactors. However, their creep strength is much lower than that of creep-resistant steel developed for conventional fossil-fired power plants as alloying elements with a high neutron activation cannot be used. To improve the creep strength and to maintain a low activation, a high-throughput computational alloy design model coupling thermodynamics, precipitate-coarsening kinetics and an optimization genetic algorithm, is developed. Twelve relevant alloying elements with either low or high activation are considered simultaneously. The activity levels at 0–10 year after the end of irradiation are taken as optimization parameter. The creep-strength values (after exposure for 10 years at 650 °C) are estimated on the basis of the solid-solution strengthening and the precipitation hardening (taking into account precipitate coarsening). Potential alloy compositions leading to a high austenite fraction or a high percentage of undesirable second phase particles are rejected automatically in the optimization cycle. The newly identified alloys have a much higher precipitation hardening and solid-solution strengthening at the same activity level as existing reduced-activation ferritic/martensitic steels.

Creep and Creep Crack Growth Behaviors for SMAW Weldments of Gr. 91 Steel

International Nuclear Information System (INIS)

Kim, Woo Gon; Yin, Song Nan; Park, Ji Yeon; Hong, Sung Deok; Kim, Yong Wan; Park, Jae Young

2010-01-01

High Cr ferritic resistance steels with tempered martensite microstructures posses enhanced creep strength at the elevated temperatures. Those steels as represented by a modified 9Cr-1Mo steel (ASME Grade 91, hereafter Gr.91) are regarded as main structural materials of sodium-cooled fast reactors (SFR) and reactor pressure vessel materials of very high temperature reactors (VHTR). The SFR and VHTR systems are designed during long-term duration reaching 60 years at elevated temperatures and often subjected to non-uniform stress and temperature distribution during service. These conditions may generate localized creep damage and propagate the cracks and ultimately may cause a fracture. A significant portion of its life is spent in crack propagation. Therefore, a creep crack growth rate (CCGR) due to creep damage should be assessed for both the base metal (BM) and welded metal (WM). Enough CCGR data for them should be provided for assessing their structural integrities. However, their CCGR data for the Gr. 91 steels is still insufficient. In this study, the CCGR for the BM and the WM of the Gr. 91 steel was comparatively investigated. A series of the CCG tests were conducted under different applied loads for the BM and the WM at 600 .deg. C. The CCGR was characterized in terms of the C parameter, and their CCG behavior were compared, respectively

Concrete for PCRVs: strength of concrete under triaxial loading and creep at elevated temperatures

International Nuclear Information System (INIS)

Linse, D.; Aschl, H.; Stoeckl, S.

1975-01-01

To provide detailed information for the calculation of prestressed concrete reactor vessels, investigations of the behaviour of concrete under multiaxial loading and on creep at elevated temperatures were made at the Institut fuer Massivbau of the Technical University of Munich. The strength of concrete under triaxial compression is dependent on the stress ratio. The less the stresses differ from hydrostatic compression the more strength increases. Triaxial compression increases very much the deformability of concrete. Plastic deformations of +-10% and more (all stresses compression, but not equal, strains compression or tension) are possible without large cracks. The creep deformations are considerably dependent on the temperature. Creep at 80 0 C is about three to four times higher than at 20 0 C. The Poisson's ratio of creep at elevated temperature seems to be bigger than at normal temperatures at a rate of loading of 35% and 50% of the ultimate strength. (Auth.)

Mechanisms of transient radiation-induced creep

International Nuclear Information System (INIS)

Pyatiletov, Yu.S.

1981-01-01

Radiation-induced creep at the transient stage is investigated for metals. The situation, when several possible creep mechanisms operate simultaneously is studied. Among them revealed are those which give the main contribution and determine thereby the creep behaviour. The time dependence of creep rate and its relation to the smelling rate is obtained. The results satisfactorily agree with the available experimental data [ru

Stress relaxation and creep of high-temperature gas-cooled reactor core support ceramic materials: a literature search

International Nuclear Information System (INIS)

Selle, J.E.; Tennery, V.J.

1980-05-01

Creep and stress relaxation in structural ceramics are important properties to the high-temperature design and safety analysis of the core support structure of the HTGR. The ability of the support structure to function for the lifetime of the reactor is directly related to the allowable creep strain and the ability of the structure to withstand thermal transients. The thermal-mechanical response of the core support pads to steady-state stresses and potential thermal transients depends on variables, including the ability of the ceramics to undergo some stress relaxation in relatively short times. Creep and stress relaxation phenomena in structural ceramics of interest were examined. Of the materials considered (fused silica, alumina, silicon nitride, and silicon carbide), alumina has been more extensively investigated in creep. Activation energies reported varied between 482 and 837 kJ/mole, and consequently, variations in the assigned mechanisms were noted. Nabarro-Herring creep is considered as the primary creep mechanism and no definite grain size dependence has been identified. Results for silicon nitride are in better agreement with reported activation energies. No creep data were found for fused silica or silicon carbide and no stress relaxation data were found for any of the candidate materials. While creep and stress relaxation are similar and it is theoretically possible to derive the value of one property when the other is known, no explicit demonstrated relationship exists between the two. For a given structural ceramic material, both properties must be experimentally determined to obtain the information necessary for use in high-temperature design and safety analyses

A method of creep damage summation based on accumulated strain for the assessment of creep-fatigue endurance

International Nuclear Information System (INIS)

Hales, R.

1983-01-01

A method of combining long term creep data with relatively short term mechanical behaviour to provide an estimate of creep-fatigue endurance is presented. It is proposed that the creep-fatigue effect in high temperature cyclic deformation is governed by a difference in strain rate around the cycle and the associated variation in ductility with strain rate. (author)

The irradiation creep in reactor graphites for HTR applications

International Nuclear Information System (INIS)

Veringa, H.J.; Blackstone, R.

1976-01-01

A series of restrained shrinkage experiments on a number of graphites in the temperature range 400 to 1400 0 C is described. A description is given of the experimental method and method of data evaluation. The results are compared with data from other sources. Analysis of data confirms that the creep coefficient, which is defined as the radiation induced creep strain per unit stress per unit neutron fluence, is inversely proportional to the pre-irradiation value of the Young's modulus of the material. The radiation creep coefficient increases with temperature in the range 400 to 1400 0 C. It can be represented by the sum of two temperature dependent functions, one of which is inversely proportional to the neutron flux density, the other independent of the neutron flux density. When the data are analysed in this way it is found that the graphites investigated in the present work, although made from widely different starting materials and by different processes, show the same dependence of the irradiation creep coefficient on the temperature and the neutron flux density. (author)

The Effect of Hold Time on Creep-Fatigue in 9Cr-1Mo

Energy Technology Data Exchange (ETDEWEB)

Oh, Tae Young; Kim, Dae Whan; Kim, Yong Wan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Baek, Kyoung Ho [Chungnam National University, Daejeon (Korea, Republic of)

2009-05-15

9Cr-1Mo steel is a candidate material for reactor vessel for VHTR. Because 9Cr-1Mo steel has a good mechanical properties and a lower thermal expansion coefficient than austenitic stainless steel. The reactor vessel of VHTR is operated at about 450 .deg. C. At this temperature, fatigue occurs during start-up and cool-down, and creep occurs during normal operation. Creep-fatigue damage by the interaction between fatigue and creep is an important factor that limits VHTR reactor vessel life. In this study, Effect of hold time on low cycle fatigue behavior of 9Cr-1Mo at 600 .deg. C was investigated in air.

The Effect of Hold Time on Creep-Fatigue in 9Cr-1Mo

International Nuclear Information System (INIS)

Oh, Tae Young; Kim, Dae Whan; Kim, Yong Wan; Baek, Kyoung Ho

2009-01-01

9Cr-1Mo steel is a candidate material for reactor vessel for VHTR. Because 9Cr-1Mo steel has a good mechanical properties and a lower thermal expansion coefficient than austenitic stainless steel. The reactor vessel of VHTR is operated at about 450 .deg. C. At this temperature, fatigue occurs during start-up and cool-down, and creep occurs during normal operation. Creep-fatigue damage by the interaction between fatigue and creep is an important factor that limits VHTR reactor vessel life. In this study, Effect of hold time on low cycle fatigue behavior of 9Cr-1Mo at 600 .deg. C was investigated in air

Fission gas behaviour in water reactor fuels

International Nuclear Information System (INIS)

2002-01-01

During irradiation, nuclear fuel changes volume, primarily through swelling. This swelling is caused by the fission products and in particular by the volatile ones such as krypton and xenon, called fission gas. Fission gas behaviour needs to be reliably predicted in order to make better use of nuclear fuel, a factor which can help to achieve the economic competitiveness required by today's markets. These proceedings communicate the results of an international seminar which reviewed recent progress in the field of fission gas behaviour in light water reactor fuel and sought to improve the models used in computer codes predicting fission gas release. State-of-the-art knowledge is presented for both uranium-oxide and mixed-oxide fuels loaded in water reactors. (author)

Creep fatigue design of FBR components

International Nuclear Information System (INIS)

Bhoje, S.B.; Chellapandi, P.

1997-01-01

This paper deals with the characteristic features of Fast Breeder Reactor (FBR) with reference to creep fatigue, current creep fatigue design approach in compliance with RCCMR (1987) design code, material data, effects of weldments and neutron irradiation, material constitutive models employed, structural analysis and further R and D required for achieving maturity in creep fatigue design of FBR components. For the analysis reported in this paper, material constitutive models developed based on ORNIb (Oak Ridge National Laboratory) and Chaboche viscoplastic theories are employed to demonstrate the potential of FBR components for higher plant temperatures and/or longer life. The results are presented for the studies carried out towards life prediction of Prototype Fast Breeder Reactor (PFBR) components. (author). 24 refs, 8 figs, 5 tabs

Implementation and evaluation of fuel creep using advanced light-water reactor materials in FRAPCON 3.5

Science.gov (United States)

Carroll, Spencer

As current reactors approach the end of their operable lifetime, new reactors are needed if nuclear power is to continue being generated in the United States. Some utilities have already began construction on newer, more advanced LWR reactors, which use the same fuel as current reactors and have a similar but updated design. Others are researching next generation (GEN-IV) reactors which have new designs that utilize alternative fuel, coolants and other reactor materials. Many of these alternative fuels are capable of achieving higher burnups and are designed to be more accident tolerant than the currently used UO2 fuel. However, before these new materials can be used, extensive research must be done in order to obtain a detailed understanding of how the new fuels and other materials will interact. New fuels, such as uranium nitride (UN) and uranium carbide (UC) have several advantages over UO2, such as increased burnup capabilities and higher thermal conductivities. However, there are issues with each that prevent UC and UN from being used as direct replacements for UO2. Both UC and UN swell at a significantly higher rate than UO2 and neither fuel reacts favorably when exposed to water. Due to this, UC and UN are being considered more for GEN-IV reactors that use alternative coolant rather than for current LWRs. In an effort to increase accident tolerance, silicon carbide (SiC) is being considered for use as an alternative cladding. The high strength, high melting point and low oxidation of SiC make it an attractive cladding choice, especially in an accident scenario. However, as a ceramic, SiC is not ductile and will not creep outwards upon pellet-clad mechanical interaction (PCMI) which can cause a large build up in interfacial pressure. In order to understand the interaction between the high swelling fuels and unyielding SiC cladding, data on the properties and behaviors of these materials must be gathered and incorporated into FRAPCON. FRAPCON is a fuel

Effect of heat treatment, with and without mechanical work, on the tensile and creep behaviour at 6000C of austenitic stainless steel stabilised with titanium

International Nuclear Information System (INIS)

Padilha, A.F.

1983-01-01

The effect of various heat treatments, with and without mechanical work, on the microstructure and the tensile and creep behaviour at 600 0 C of the titanium stabilised austenitic stainless steel DIN 1.4970, as well as the effects of aging temperature, pre-strain and small boron additions on the creep behaviour of these steels are discussed. The most probable mechanism is suggested. (Author) [pt

Diffusion creep and its inhibition in a stainless steel

International Nuclear Information System (INIS)

Crossland, I.G.; Clay, B.D.

1977-01-01

The creep of 20% Cr, 25% Ni, Nb stainless steel was examined at low stresses and temperatures around 0.55 T/sub m/. The initial creep behaviour was consistent with the Coble theory of grain boundary diffusion creep; however, steady state creep was not observed and the creep rates quickly fell below the Coble theoretical values although they still remained greater than the Herring--Nabarro predictions. This reduction in creep rate was attributable to an increase in the effective viscosity of the steel rather than to any change in threshold stress. A model is proposed which explains the initial creep rates as being due to Coble creep with elastic accommodation at grain boundary particles. At higher strains grain boundary collapse caused by vacancy sinking is accommodated at precipitate particles by plastic deformation of the adjacent matrix material. 11 figures

Fatigue and Creep Crack Propagation behaviour of Alloy 617 in the Annealed and Aged Conditions

Energy Technology Data Exchange (ETDEWEB)

Julian K. Benz; Richard N. Wright

2013-10-01

The crack propagation behaviour of Alloy 617 was studied under various conditions. Elevated temperature fatigue and creep-fatigue crack growth experiments were conducted at 650 and 800 degrees C under constant stress intensity (triangle K) conditions and triangular or trapezoidal waveforms at various frequencies on as-received, aged, and carburized material. Environmental conditions included both laboratory air and characteristic VHTR impure helium. As-received Alloy 617 displayed an increase in the crack growth rate (da/dN) as the frequency was decreased in air which indicated a time-dependent contribution component in fatigue crack propagation. Material aged at 650°C did not display any influence on the fatigue crack growth rates nor the increasing trend of crack growth rate with decreasing frequency even though significant microstructural evolution, including y’ (Ni3Al) after short times, occurred during aging. In contrast, carburized Alloy 617 showed an increase in crack growth rates at all frequencies tested compared to the material in the standard annealed condition. Crack growth studies under quasi-constant K (i.e. creep) conditions were also completed at 650 degrees C and a stress intensity of K = 40 MPa9 (square root)m. The results indicate that crack growth is primarily intergranular and increased creep crack growth rates exist in the impure helium environment when compared to the results in laboratory air. Furthermore, the propagation rates (da/dt) continually increased for the duration of the creep crack growth either due to material aging or evolution of a crack tip creep zone. Finally, fatigue crack propagation tests at 800 degrees C on annealed Alloy 617 indicated that crack propagation rates were higher in air than impure helium at the largest frequencies and lowest stress intensities. The rates in helium, however, eventually surpass the rates in air as the frequency is reduced and the stress intensity is decreased which was not observed at 650

Survey of creep data on structural materials of fast breeder reactor

International Nuclear Information System (INIS)

Yoshida, S.

1977-11-01

The reactor vessels and other components of fast breeder reactor is affected by high neutron irradiation at elevated temperatures. However, in this regard, related test data on creep property of component materials and welds at elevated temperatures are a few in Japan, and especially, there are no data available on the irradiation effect. It will take 3 to 7 years before the results of currently planned research and development on prototype fast breeder become available. On the other hand, establishment of design base for prototype fast breeder and other needs call for early solution to such problems. The Committee should, therefore, collect from documents the latest data on experiments on structural materials overseas and in our country, and survey and analyze the problems in order to proceed with the future research and development in the most effective way. It was for this purpose that the Fourth Subcommittee at Technical Research Association for Integrity of Structures at Elevated Service Temperatures was commissioned by Power Reactor and Nuclear Fuel Development Corporation to conduct the examination and study of related data by establishing Group 41G. This collection of data is the compilation of the above results. (author)

A Study of the Creep Effect in Loudspeaker Suspension

DEFF Research Database (Denmark)

Agerkvist, Finn T.; Thorborg, Knud; Tinggaard, Carsten

2008-01-01

This paper investigates the creep effect, the visco elastic behaviour of loudspeaker suspension parts, which can be observed as an increase in displacement far below the resonance frequency. The creep effect means that the suspension cannot be modelled as a simple spring. The need for an accurate...... creep model is even larger as the validity of loudspeaker models are now sought extended far into the nonlinear domain of the loudspeaker. Different creep models are investigated and implemented both in simple lumped parameter models as well as time domain non-linear models, the simulation results...

Creep Strength of Nb-1Zr for SP-100 Applications

Science.gov (United States)

Horak, James A.; Egner, Larry K.

1994-07-01

Power systems that are used to provide electrical power in space are designed to optimize conversion of thermal energy to electrical energy and to minimize the mass and volume that must be launched. Only refractory metals and their alloys have sufficient long-term strength for several years of uninterrupted operation at the required temperatures of 1200 K and above. The high power densities and temperatures at which these reactors must operate require the use of liquid-metal coolants. The alloy Nb-1 wt % Zr (Nb-lZr), which exhibits excellent corrosion resistance to alkali liquid-metals at high temperatures, is being considered for the fuel cladding, reactor structural, and heat-transport systems for the SP-100 reactor system. Useful lifetime of this system is limited by creep deformation in the reactor core. Nb-lZr sheet procured to American Society for Testing and Materials (ASTM) specifications for reactor grade and commercial grade has been processed by several different cold work and annealing treatments to attempt to produce the grain structure (size, shape, and distribution of sizes) that provides the maximum creep strength of this alloy at temperatures from 1250 to 1450 K. The effects of grain size, differences in oxygen concentrations, tungsten concentrations, and electron beam and gas tungsten arc weldments on creep strength were studied. Grain size has a large effect on creep strength at 1450 K but only material with a very large grain size (150 μm) exhibits significantly higher creep strength at 1350 K. Differences in oxygen or tungsten concentrations did not affect creep strength, and the creep strengths of weldments were equal to, or greater than, those for base metal.

Thermal and mechanical behaviour of oxygen carrier materials for chemical looping combustion in a packed bed reactor

International Nuclear Information System (INIS)

Jacobs, M.; Van Noyen, J.; Larring, Y.; Mccann, M.; Pishahang, M.; Amini, S.; Ortiz, M.; Galluci, F.; Sint-Annaland, M.V.; Tournigant, D.; Louradour, E.; Snijkers, F.

2015-01-01

Highlights: • Ilmenite-based oxygen carriers were developed for packed-bed chemical looping. • Addition of Mn_2O_3 increased mechanical strength and microstructure of the carriers. • Oxygen carriers were able to withstand creep and thermal cycling up to 1200 °C. • Ilmenite-based granules are a promising shape for packed-bed reactor conditions. - Abstract: Chemical looping combustion (CLC) is a promising carbon capture technology where cyclic reduction and oxidation of a metallic oxide, which acts as a solid oxygen carrier, takes place. With this system, direct contact between air and fuel can be avoided, and so, a concentrated CO_2 stream is generated after condensation of the water in the exit gas stream. An interesting reactor system for CLC is a packed bed reactor as it can have a higher efficiency compared to a fluidized bed concept, but it requires other types of oxygen carrier particles. The particles must be larger to avoid a large pressure drop in the reactor and they must be mechanically strong to withstand the severe reactor conditions. Therefore, oxygen carriers in the shape of granules and based on the mineral ilmenite were subjected to thermal cycling and creep tests. The mechanical strength of the granules before and after testing was investigated by crush tests. In addition, the microstructure of these oxygen particles was studied to understand the relationship between the physical properties and the mechanical performance. It was found that the granules are a promising shape for a packed bed reactor as no severe degradation in strength was noticed upon thermal cycling and creep testing. Especially, the addition of Mn_2O_3 to the ilmenite, which leads to the formation of an iron–manganese oxide, seems to results in stronger granules than the other ilmenite-based granules.

European development of ferritic-martensitic steels for fast reactor wrapper applications

International Nuclear Information System (INIS)

Bagley, K.; Little, E.A.; Levy, V.; Alamo, A.

1987-01-01

9-12%Cr ferritic-martensitic stainless steels are under development in Europe for fast reactor sub-assembly wrapper applications. Within this class of alloys, attention is focussed on three key specifications, viz. FV448 and DIN 1.4914 (both 10-12%CrMoVNb steels) and EM10 (an 8-10%Cr-0.15%C steel), which can be optimized to give acceptably low ductile-brittle transition characteristics. The results of studies on these steels, and earlier choices, covering heat treatment and compositional optimization, evolution of wrapper fabrication routes, pre and post-irradiation mechanical property and fracture toughness behaviour, microstructural stability, void swelling and in-reactor creep characteristics are reviewed. The retention of high void swelling to displacement doses in excess of 100 dpa in reactor irradiations reaffirms the selection of 9-12%Cr steels for on-going wrapper development. Moreover, irradiation-induced changes in mechanical properties (e.g. in-reactor creep and impact behaviour), measured to intermediate doses, do not give cause for concern; however, additional data to higher doses and at the lower irradiation temperatures of 370 0 -400 0 C are needed in order to fully endorse these alloys for high burnup applications in advanced reactor systems

An analysis of irradiation creep in nuclear graphites

International Nuclear Information System (INIS)

Neighbour, G.B.; Hacker, P.J.

2002-01-01

Nuclear graphite under load shows remarkably high creep ductility with neutron irradiation, well in excess of any strain experienced in un-irradiated graphite (and additional to any dimensional changes that would occur without stress). As this behaviour compensates, to some extent, some other irradiation effects such as thermal shutdown stresses, it is an important property. This paper briefly reviews the approach to irradiation creep in the UK, described by the UK Creep Law. It then offers an alternative analysis of irradiation creep applicable to most situations, including HTR systems, using AGR moderator graphite as an example, to high values of neutron fluence, applied stress and radiolytic weight loss. (authors)

SMART - Structure mechanical analysis in reactor technology

International Nuclear Information System (INIS)

Argyris, J.H.; Faust, G.; Szimmat, J.; Warnke, E.P.; Willam, K.J.

1975-01-01

The programme system SMART was developed in the years 1970-75 to calculate prestressed-concrete reactor pressure vessels with finite elements. The present report outlines the course and present state of research and development work. Following the specification of SMART, a brief presentation of the analytical possibilities and of the expansions for investigating creep, ultimate load behaviour and thermodiffusion is given. In conclusion, the fields of application of SMART are illustrated by means of examples. (orig./LH) [de

Irradiation Creep of Ferritic-Martensitic Steels EP-450, EP-823 and EI-852 Irradiated in the BN-350 Reactor over Wide Ranges of Irradiation Temperature and Dose

International Nuclear Information System (INIS)

Porollo, S.I.; Konobeev, Y.V.; Ivanov, A.A.; Shulepin, S.V.; Garner, F.

2007-01-01

Full text of publication follows: Ferritic/martensitic (F/M) steels appear to be the most promising materials for advanced nuclear systems, especially for fusion reactors. Their main advantages are higher resistance to swelling and lower irradiation creep rate as has been repeatedly demonstrated in examinations of these materials after irradiation. Nevertheless, available experimental data on irradiation resistance of F/M steels are insufficient, with the greatest deficiency of data for high doses and for both low and high irradiation temperatures. From the very beginning of operation the BN-350 fast reactor has been used for irradiation of specimens of structural materials, including F/M steels. The most unique feature of BN-350 was its low inlet sodium temperature, allowing irradiation at temperatures over a very wide range of temperatures compared with the range in other fast reactors. In this paper data are presented on swelling and irradiation creep of three Russian F/M steels EP-450, EP-823 and EI-852, irradiated in experimental assemblies of the BN-350 reactor at temperatures in the range of 305-700 deg. C to doses ranging from 20 to 89 dpa. The investigation was performed using gas-pressurized creep tubes with hoop stresses in the range of 0 - 294 MPa. (authors)

High temperature creep of vanadium

International Nuclear Information System (INIS)

Juhasz, A.; Kovacs, I.

1978-01-01

The creep behaviour of polycrystalline vanadium of 99.7% purity has been investigated in the temperature range 790-880 0 C in a high temperature microscope. It was found that the creep properties depend strongly on the history of the sample. To take this fact into account some additional properties such as the dependence of the yield stress and the microhardness on the pre-annealing treatment have also been studied. Samples used in creep measurements were selected on the basis of their microhardness. The activation energy of creep depends on the microhardness and on the creep temperature. In samples annealed at 1250 0 C for one hour (HV=160 kgf mm -2 ) the rate of creep is controlled by vacancy diffusion in the temperature range 820-880 0 C with an activation energy of 78+-8 kcal mol -1 . (Auth.)

Release of fission products and post-pile creep behaviour of irradiated fuel rods stored under dry conditions

International Nuclear Information System (INIS)

Kaspar, G.; Peehs, M.; Bokelmann, R.; Jorde, D.; Schoenfeld, H.; Haas, W.; Bleier, A.; Rutsch, F.

1985-06-01

The release of moisture and fission products (Kr-85, H-3 and I-129) under dry storage conditions has been examined on six fuel rods which have become defective in the reactor. During the examinations, inert conditions prevailed and limited air inlet was allowed temporarily. The storage temperature was 400 0 C. The residual moisture content of the fuel rods was approx. 5 g. At the beginning of the test, the total moisture content and 0,05% (max.) of the fission gas inventory were released. Under inert conditions, fission gas was not released during a prolonged period of time. Under oxidizing conditions, however, fission gas was released in the course of UO 2 oxidation. Post-pile creep of Zircaloy cladding tubes was measured at temperatures between 350 and 395 0 C and interval gauge pressures between 69 and 110 bar. The creep curves indicate that the irradiated cladding tube specimens still bear internal residual stresses which contribute through their relaxation to the post-pile creep. (orig.) [de

Irradiation creep models - an overview

International Nuclear Information System (INIS)

Matthews, J.R.; Finnis, M.W.

1988-01-01

The modelling of irradiation creep is now highly developed but many of the basic processes underlying the models are poorly understood. A brief introduction is given to the theory of cascade interactions, point defect clustering and dislocation climb. The range of simple irradiation creep models is reviewed including: preferred nucleation of interstitial loops; preferred absorption of point defects by dislocations favourably orientated to an applied stress; various climb-enhanced glide and recovery mechanisms, and creep driven by internal stresses produced by irradiation growth. A range of special topics is discussed including: cascade effects; creep transients; structural and induced anisotropy; and the effect of impurities. The interplay between swelling and growth with thermal and irradiation creep is emphasized. A discussion is given on how irradiation creep theory should best be developed to assist the interpretation of irradiation creep observations and the requirements of reactor designers. (orig.)

Prediction of Axial and Radial Creep in CANDU 6 Pressure Tubes

International Nuclear Information System (INIS)

Radu, Vasile S.

2013-01-01

Status and proposals: 1. A review of literature concerning on the in-reactor deformation of PTs has been carried ouţ. 2. A model based on MFNN has been proposed to assess the radial and axial creep of CANDU 6 PTs. 3. Preliminary discussion with Cernavoda NPP (Romania) has been lunched, and now the preparation of official documents (collaboration in providing the inspection data from fuel channel in Unit 1 and 2) are in progress. 4. Further activities: • Improvement MFNN to accommodate complex data base (eventually with many variables) for radial and axial in-reactor deformation PT, and to satisfy the requirements from NPP Cernavoda and hopefully from present CRP database; • To build-up a database by running the creep equations (if the creep constants are provided by AECL); training of MFNN on them and to qualify it as a tool for PT in-reactor deformation prediction

A Model for Creep and Creep Damage in the γ-Titanium Aluminide Ti-45Al-2Mn-2Nb.

Science.gov (United States)

Harrison, William; Abdallah, Zakaria; Whittaker, Mark

2014-03-14

Gamma titanium aluminides (γ-TiAl) display significantly improved high temperature mechanical properties over conventional titanium alloys. Due to their low densities, these alloys are increasingly becoming strong candidates to replace nickel-base superalloys in future gas turbine aeroengine components. To determine the safe operating life of such components, a good understanding of their creep properties is essential. Of particular importance to gas turbine component design is the ability to accurately predict the rate of accumulation of creep strain to ensure that excessive deformation does not occur during the component's service life and to quantify the effects of creep on fatigue life. The theta (θ) projection technique is an illustrative example of a creep curve method which has, in this paper, been utilised to accurately represent the creep behaviour of the γ-TiAl alloy Ti -45Al-2Mn-2Nb. Furthermore, a continuum damage approach based on the θ-projection method has also been used to represent tertiary creep damage and accurately predict creep rupture.

Creep analysis of orthotropic shells

International Nuclear Information System (INIS)

Mehra, V.K.; Ghosh, A.

1975-01-01

A method of creep analysis of orthotropic cylindrical shells subjected to axisymmetric loads has been developed. A general study of creep behaviour of cylindrical shells subjected to a uniform internal pressure has been conducted for a wide range of values of anisotropy coefficients and creep law exponent. Analysis includes determination of stress re-distribution, strain rates, stationary state stresses. Application of reference stress technique has been extended to analysis of shells. (author)

Modelling of degradation processes in creep resistant steels through accelerated creep tests after long-term isothermal ageing

Energy Technology Data Exchange (ETDEWEB)

Sklenicka, V.; Kucharova, K.; Svoboda, M.; Kroupa, A.; Kloc, L. [Academy of Sciences of the Czech Republic, Brno (Czech Republic). Inst. of Physics of Materials; Cmakal, J. [UJP PRAHA a.s., Praha-Zbraslav (Czech Republic)

2010-07-01

Creep behaviour and degradation of creep properties of creep resistant materials are phenomena of major practical relevance, often limiting the lives of components and structures designed to operate for long periods under stress at elevated and/or high temperatures. Since life expectancy is, in reality, based on the ability of the material to retain its high-temperature creep strength for the projected designed life, methods of creep properties assessment based on microstructural evolution in the material during creep rather than simple parametric extrapolation of short-term creep tests are necessary. In this paper we will try to further clarify the creep-strength degradation of selected advanced creep resistant steels. In order to accelerate some microstructural changes and thus to simulate degradation processes in long-term service, isothermal ageing at 650 C for 10 000 h was applied to P91 and P23 steels in their as-received states. The accelerated tensile creep tests were performed at temperature 600 C in argon atmosphere on all steels both in the as-received state and after long-term isothermal ageing, in an effort to obtain a more complete description of the role of microstructural stability in high temperature creep of these steels. Creep tests were followed by microstructural investigations by means of both transmission and scanning electron microscopy and by the thermodynamic calculations. The applicability of the accelerated creep tests was verified by the theoretical modelling of the phase equilibria at different temperatures. It is suggested that under restructed oxidation due to argon atmosphere microstructural instability is the main detrimental process in the long-term degradation of the creep rupture strength of these steels. (orig.)

Dynamic behaviour of CANDU reactor

International Nuclear Information System (INIS)

Subramanian, M.G.; Srikantiah, G.; Pai, M.A.

1976-01-01

Understanding of the dynamic behaviour of a reactor system in a power station is essential for evolving control stragies as well as design modifications. The dynamic behaviour of Rajasthan Atomic Power Station is studied. Mathematical models for the reactor, the steam generator and the steam drum with the natural circulation loop are developed from physical principles like conservation of mass, momentum and energy. Each of these models is then simulated on a digital computer to obtain the characteristics during transients. The models are then combined to yield a dynamic mathematical model of the system comprising the reactor, the steam generator and the steam drum and this results in a nonlinear model. Using this model, responses of the system for various disturbances like step change in the area of the steam valve, step change in the temperature of feed water are obtained and are discussed. These models could be used to devise new control laws using optimal control theory or to evaluate the performance of existing control schemes. (author)

An extension of a high temperature creep model to account for fuel sheath oxidation

International Nuclear Information System (INIS)

Boccolini, G.; Valli, G.

1983-01-01

Starting from the high-temperature creep model for Zircaloy fuel sheathing, the NIRVANA (developed by AECL), a multilayer model, is proposed in this paper: it includes the outer oxide plus alpha retained layers, and the inner core of beta or alpha plus beta material, all constrained to deform with the same creep rate. The model has been incorporated into the SPARA fuel computer code developed for the transient analysis of fuel rod behaviour in the CIRENE prototype reactor, but it is in principle valid for all Zircaloy fuel sheathings. Its predictions are compared with experimental results from burst tests on BWR and PWR type sheaths; the tests were carried out at CNEN under two research contracts with Ansaldo Meccanico Nucleare and Sigen-Sopren, respectively

Creep-rupture-test on the stainless steel X6crni1811 (Din 1.494.8) in the frame of the Extrapolation-Program. (Part III); Ensayos de fluencia lenta en el acero inoxidable X6 Cr Ni 1811 (1.4948) en el marco del Programa Extrapolacion

Energy Technology Data Exchange (ETDEWEB)

Solano, R; Schirra, M; Rivas, M de la; Barroso, S; Seith, B

1982-07-01

The austenitic stainless steel X6crni1811 (Din 1.4948) used as a structure material for the German Fast Breeder Reactor SNR 300 was creep tested in a temperature range of 550-650 degree centigree material condition as well as welded material condition. The main point of this program (Extrapolation-Program) lies in the knowledge of the creep-rupture-strength and creep-behaviour up to 3 x 10{sup 4} hours higher temperatures in order to extrapolated up to {>=}10{sup 5} hours for operating temperatures. In order to study the stress dependency of the minimum creep rate additional tests were carried out of 550 degree centigree - 750 degree centigree. The present report describes the state in the running program with test-times of 23.000 hours and results from tests up to 55.000 hours belonging to other parallel programs are taken into account. Besides the creep-rupture behaviour it is also made a study of ductility between 550 and 750 degree centigree. Extensive metallographic examinations have been made to study the fracture behaviour and changes in structure. (Author)

Novel experiments to characterise creep-fatigue degradation in VHTR alloys

International Nuclear Information System (INIS)

Simpson, J.A.; Wright, J.K.; Wright, R.N.

2015-01-01

It is well known in energy systems that the creep lifetime of high temperature alloys is significantly degraded when a cyclic load is superimposed on components operating in the creep regime. A test method has been developed in an attempt to characterise creep-fatigue behaviour of alloys at high temperature. The test imposes a hold time during the tensile phase of a fully reversed strain-controlled low cycle fatigue test. Stress relaxation occurs during the strain-controlled hold period. This type of fatigue stress relaxation test tends to emphasise the fatigue portion of the total damage and does not necessarily represent the behaviour of a component in-service well. Several different approaches to laboratory testing of creep-fatigue at 950 deg. C have been investigated for Alloy 617, the primary candidate for application in VHTR heat exchangers. The potential for mode switching in a cyclic test from strain control to load control, to allow specimen extension by creep, has been investigated to further emphasise the creep damage. In addition, tests with a lower strain rate during loading have been conducted to examine the influence of creep damage occurring during loading. Very short constant strain hold time tests have also been conducted to examine the influence of the rapid stress relaxation that occurs at the beginning of strain holds. (authors)

Creep constitutive equation of dual phase 9Cr-ODS steel

International Nuclear Information System (INIS)

Sakasegawa, Hideo; Ukai, Shigeharu; Tamura, Manabu; Ohtsuka, Satoshi; Tanigawa, Hiroyasu; Ogiwara, Hiroyuki; Kohyama, Akira; Fujiwara, Masayuki

2008-01-01

9Cr-ODS (oxide dispersion strengthened) steels developed by JAEA (Japan Atomic Energy Agency) have superior creep properties compared with conventional heat resistant steels. The ODS steels can enormously contribute to practical applications of fast breeder reactors and more attractive fusion reactors. Key issues are developments of material processing procedures for mass production and creep life prediction methods in present R and D. In this study, formulation of creep constitutive equation was performed against the backdrop. The 9Cr-ODS steel displaying an excellent creep property is a dual phase steel. The ODS steel is strengthened by the δ ferrite which has a finer dispersion of oxide particles and shows a higher hardness than the α' martensite. The δ ferrite functions as a reinforcement in the dual phase 9Cr-ODS steel. Its creep behavior is very unique and cannot be interpreted by conventional theories of heat resistant steels. Alternative qualitative model of creep mechanism was formulated at the start of this study using the results of microstructural observations. Based on the alternative creep mechanism model, a novel creep constitutive equation was formulated using the exponential type creep equation extended by a law of mixture

Irradiation creep in reactor graphites for HTR applications. [Neutron irradiation

Energy Technology Data Exchange (ETDEWEB)

Veringa, H J; Blackstone, R [Stichting Reactor Centrum Nederland, Petten

1976-01-01

A series of restrained shrinkage experiments on a number of graphites in the temperature range 400 to 1400/sup 0/C is described. A description is given of the experimental method and method of data evaluation. The results are compared with data from other sources. Analysis of data confirms that the creep coefficient, which is defined as the radiation induced creep strain per unit stress per unit neutron fluence, is inversely proportional to the pre-irradiation value of the Young's modulus of the material. The radiation creep coefficient increases with temperature in the range 400 to 1400/sup 0/C. It can be represented by the sum of two temperature dependent functions, one of which is inversely proportional to the neutron flux density, the other independent of the neutron flux density. When the data are analysed in this way it is found that the graphites investigated in the present work, although made from widely different starting materials and by different processes, show the same dependence of the irradiation creep coefficient on the temperature and the neutron flux density.

Experimental verification of creep analyses for prestressed concrete reactor vessels

International Nuclear Information System (INIS)

Aoyagi, Y.; Abe, H.; Ohnuma, H.

1977-01-01

The authors proposed a new method of creep analysis based on the theory of strain hardening, which assumes that accumulated creep at a given time influences the creep after that. This method was applied to calculate step-by-step the behaviors of uniaxial creep of concrete under variable temperatures and stresses, creep in reinforced concrete specimens and the behaviors of prestressed concrete beams under themal gradients. The experimental and calculated results agreed fairly well. Further, this method was incorporated in the finite element creep analysis for the prestressed concrete hollow cylinder and the full scale model. The calculated strain changes with time pursued closely those obtained by experiments. The above led to the conclusion that from the viewpoint of both accuracy and computation time the strain hardening method proposed by the authors may be judged advantageous for practical usages

The influence of long-term annealing at room temperature on creep behaviour of ECAP-processed copper

Czech Academy of Sciences Publication Activity Database

Král, Petr; Dvořák, Jiří; Kvapilová, Marie; Blum, W.; Sklenička, Václav

2017-01-01

Roč. 188, FEB (2017), s. 235-238 ISSN 0167-577X R&D Projects: GA MŠk(CZ) LQ1601 Institutional support: RVO:68081723 Keywords : Equal-channel angular pressing (ECAP) * Ultrafine-grained microstructure * Creep behaviour * Microstructure stability Subject RIV: JG - Metallurgy OBOR OECD: Materials engineering Impact factor: 2.572, year: 2016

Creep of fibrous composite materials

DEFF Research Database (Denmark)

Lilholt, Hans

1985-01-01

Models are presented for the creep behaviour of fibrous composite materials with aligned fibres. The models comprise both cases where the fibres remain rigid in a creeping matrix and cases where the fibres are creeping in a creeping matrix. The treatment allows for several contributions...... to the creep strength of composites. The advantage of combined analyses of several data sets is emphasized and illustrated for some experimental data. The analyses show that it is possible to derive creep equations for the (in situ) properties of the fibres. The experiments treated include model systems...... such as Ni + W-fibres, high temperature materials such as Ni + Ni3Al + Cr3C2-fibres, and medium temperature materials such as Al + SiC-fibres. For the first two systems reasonable consistency is found for the models and the experiments, while for the third system too many unquantified parameters exist...

Creep-Rupture Properties and Corrosion Behaviour of 21/4 Cr-1 Mo Steel and Hastelloy X-Alloys in Simulated HTGR Environment

DEFF Research Database (Denmark)

Lystrup, Aage; Rittenhouse, P. L.; DiStefano, J. R.

Hastelloy X and 2/sup 1///sub 4/ Cr-1 Mo steel are being considered as structural alloys for components of a High-Temperature Gas-Cooled Reactor (HTGR) system. Among other mechanical properties, the creep behavior of these materials in HTGR primary coolant helium must be established to form part...

Stress relaxation analysis and irradiation creep and swelling in pressure tubes

International Nuclear Information System (INIS)

Beeston, J.M.; Burr, T.K.

1979-01-01

An analysis is presented of slit width test information on two pressure tubes that had been irradiated in test reactors. The analysis showed that differential swelling stresses and thermal stresses undergo relaxation. The mechanism responsible for the stress relaxation at temperatures less than 700 K was irradiation creep. Irradiation creep in thermal test reactor pressure tubes is evidently greater than it would be at equivalent conditions in fast reactors. The residual stresses observed in the slit width tests varied between 30 and 257 MPa and would act to reduce the operating stresses, thus allowing for increased service life of the tubes as compared with no stress relaxation

Report On Design And Preliminary Data Of Halden In-Pile Creep Rig

Energy Technology Data Exchange (ETDEWEB)

Terrani, Kurt A [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Karlsen, T. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-09-01

A set of in-pile creep tests is ongoing in the Halden reactor on ORNL’s candidate accident tolerant fuel cladding materials. These tests are meant to provide essential material property information that is needed for an informed analysis of these fuel concepts under normal operating conditions. These tests provide detailed information regarding swelling, thermal creep, and irradiation creep rates of these materials. The results to date have been compared with the limited set of information available in literature that is form irradiation tests in other reactors or out-of-pile tests. Most of the results are in good agreement with prior literature, except for irradiation creep rate of SiC. To elucidate the difference between the HFIR and Halden test results continued testing is necessary. The tests describe in this progress report are ongoing and will continue for at least another year.

Irradiation creep performance of graphite relevant for pebble bed HTRs

International Nuclear Information System (INIS)

Kleist, G.; O'Connor, M.F.

1980-01-01

Irradiation - induced creep in the core reflector component graphite of high temperature reactors is of primary importance to the core designer since it provides a mechanism for the relief of internal stresses arising from differential Wigner shrinkage and thermal expansion. The experimental determination of the extent of this creep for conditions relevant to the reactor is thus imperative

Creep behavior and evolution of microstructure of modified Grade 91 welded joint after short term exposure at 500 deg C; Fluage a 500 deg C d'un joint soude d'un acier 9Cr-1Mo modifie. Evolution de la microstructure et comportement mecanique

Energy Technology Data Exchange (ETDEWEB)

Vivier, F.

2009-03-15

With the increase in worldwide energy demand, the nuclear industry is a way of producing electricity on a large scale and to answer to this need. For the design of a new generation of fission nuclear reactors and among six chosen fission reactor systems, France develops in particularly the Very High Temperature Reactor (VHTR) concept. This implies the use of materials that are more and more resistant to high temperature for long-term exposure. AREVA focuses on materials already used in fossil-fuel power plant, so that the mechanical behaviour of Grade 91 (Fe{sub 9}Cr{sub 1}MoNbV) has to be investigated. This ferritic-martensitic steel is considered to be a potential candidate for welded components. Such structures are combined with welded joints, which have to be studied. Three industrial partners (AREVA, CEA, EDF) have launched a study with the Centre des Materiaux in order to investigate the creep of welded joint of Grade 91. The aim of this work is to complete the available database about the mechanical behaviour of Grade 91, base metal and welded joint, during creep tests performed at 500 C up to 4500 h exposure. Thermal aging tests, tensile tests, and creep tests were performed at 450 C and 500 C using both base metal and cross-weld samples. Several geometries of cross-weld creep specimens were tested. The microstructure has not remarkably changed after tests concerning both nature and size of precipitates, and the characteristic size of the matrix sub-structure. The creep damage is not developed in the ruptured specimens after creep tests. Only little damage by cavity nucleation and growth was found in the creep specimens. Creep fracture at 500 C takes places by viscoplastic flow, contrary to tests performed at 625 C where the creep-induced damage governs the creep rupture at least for long-term lifetime. From creep curves of base metal and cross-weld specimens, a phenomenological model is proposed. The flow rule is a Norton power law with a stress exponent

Experimental investigation of creep behavior of reactor vessel lower head

International Nuclear Information System (INIS)

Chu, T.Y.; Pilch, M.; Bentz, J.H.; Behbahani, A.

1998-03-01

The objective of the USNRC supported Lower Head Failure (LHF) Experiment Program at Sandia National Laboratories is to experimentally investigate and characterize the failure of the reactor pressure vessel (RPV) lower head due to the thermal and pressure loads of a severe accident. The experimental program is complemented by a modeling program focused on the development of a constitutive formulation for use in standard finite element structure mechanics codes. The problem is of importance because: lower head failure defines the initial conditions of all ex-vessel events; the inability of state-of-the-art models to simulate the result of the TMI-II accident (Stickler, et al. 1993); and TMI-II results suggest the possibility of in-vessel cooling, and creep deformation may be a precursor to water ingression leading to in-vessel cooling

IRRADIATION CREEP AND MECHANICAL PROPERTIES OF TWO FERRITIC-MARTENSITIC STEELS IRRADIATED IN THE BN-350 FAST REACTOR

International Nuclear Information System (INIS)

Porollo, S. I.; Konobeev, Yu V.; Dvoriashin, A. M.; Budylkin, N. I.; Mironova, E. G.; Leontyeva-Smirnova, M. V.; Loltukhovsky, A. G.; Bochvar, A. A.; Garner, Francis A.

2002-01-01

Russian ferritic/martensitic steels EP-450 and EP-823 were irradiated to 20-60 dpa in the BN-350 fast reactor in the form of pressurized creep tubes and small rings used for mechanical property tests. Data derived from these steels serves to enhance our understanding of the general behavior of this class of steels. It appears that these steels exhibit behavior that is very consistent with that of Western steels. Swelling is relatively low at high neutron exposure and confined to temperatures less then 420 degrees C, but may be camouflaged somewhat by precipitation-related densification. The irradiation creep studies confirm that the creep compliance of F/M steels is about one-half that of austenitic steels, and that the loss of strength at test temperatures above 500 degrees C is a problem generic to all F/M steels. This conclusion is supported by post-irradiation measurement of short-term mechanical properties. At temperatures below 500 degrees C both steels retain their high strength (yield stress 0.2=550-600 MPa), but at higher test temperatures a sharp decrease of strength properties occurs. However, the irradiated steels still retain high post-irradiation ductility at test temperatures in the range of 20-700 degrees C.

Irradiation creep transients in Ni-4 at.% Si

International Nuclear Information System (INIS)

Nagakawa, J.

1983-01-01

In the course of irradiation creep experiments on Ni-4 at.% Si alloy, two types of creep transients were observed on the termination of irradiation. The short term transient was completed within one minute while the long term transient persisted for nearly ten hours. A change in the temperature distribution was excluded from the possible causes, partly because the stress dependence of the observed transient strains was not linear, and partly because the strain increase expected from the temperature change was much smaller than the observed value. Transient behavior of point defects was examined in conjunction with the climb-glide mechanism and the steady-state irradiation creep data. Calculated creep transient due to excess vacancy flux to dislocations was in good agreement with the observed short term transient. The long term transient appears to be a result of dislocation microstructure change. The present results suggest an enhanced irradiation creep under cyclic irradiation conditions which will be encountered in the early generations of fusion reactors. (orig.)

Creep collapse of TAPS fuel cladding

International Nuclear Information System (INIS)

Chaudhry, S.M.; Anand, A.K.

1975-01-01

Densification of UO 2 can cause axial gaps between fuel pelets and cladding in unsupported (internally) at these regions. An analysis is carried out regarding the possibility of creep collapse in these regions. The analysis is based on Timoshenko's theory of collapse. At various times during the residence of fuel in reactor following parameters are calculated : (1) inelastic collapse of perfectly circular tubes (2) plastic instability in oval tubes (3) effect of creep on ovality. Creep is considered to be a non-linear combination of the following : (a) thermal creep (b) intresenic creep (c) stress aided radiation enhanced (d) stress free growth (4) Critical pressure ratio. The results obtained are compared with G.E. predictions. The results do not predict collapse of TAPS fuel cladding for five year residence time. (author)

Creep properties of welded joints in OFHC copper for nuclear waste containment

International Nuclear Information System (INIS)

Ivarsson, B.; Oesterberg, J.O.

1988-08-01

In Sweden it has been suggested that copper canisters are used for containment of spent nuclear fuel. These canisters will be subjected to temperatures up to 100 degrees C and external pressures up to 15 MPa. Since the material is pure (OFHC) copper, creep properties must be considered when the canisters are dimensioned. The canisters are sealed by electron beam welding which will affect the creep properties. Literature data for copper - especially welded joints - at the temperatures of interest is very scare. Therefore uniaxial creep tests of parent metal, weld metal, and simulated HAZ structures have been performed at 110 degrees C. These tests revealed considerable differences in creep deformation and rupture strength. The weld metal showed creep rates and rupture times ten times higher and ten times shorter, respectively, than those of the parent metal. The simulated HAZ was equally strongen than the parent metal. These differences were to some extent verified by results from creep tests of cross-welded specimens which, however, showed even shorter rupture times. Constitutive equations were derived from the uniaxial test results. To check the applicability of these equations to multiaxial conditions, a few internal pressure creep tests of butt-welded tubes were performed. Attemps were made to simulate their creep behaviour by constitutive equations were used. These calculations failed due to too great differences in creep deformation behaviour across the welded joint. (authors)

Measurement and analysis of pressure tube elongation in the Douglas Point reactor

International Nuclear Information System (INIS)

Causey, A.R.; MacEwan, S.R.; Jamieson, H.C.; Mitchell, A.B.

1980-02-01

Elongations of zirconium alloy pressure tubes in CANDU reactors, which occur as a result of neutron-irradiation-induced creep and growth, have been measured over the past 6 years, and the consequences of thses elongations have recently been analysed. Elongation rates, previously deduced from extensive measurements of elongations of cold-worked Zircaloy-2 pressure tubes in the Pickering reactors, have been modified to apply to the pressure tubes in the Douglas Point (DP) reactor by taking into account measured diffences in texture and dislocation density. Using these elongation rates, and structural data unique to the DP reactor, the analysis predicts elongation behaviour which is in good agreement with pressure tube elongations measured during the ten years of reactor operation. (Auth)

Towards self-healing creep resistant steels

NARCIS (Netherlands)

Van der Zwaag, S.; Zhang, S.; Fang, H.; Bruck, E.; Van Dijk, N.H.

2016-01-01

We report the main findings of our work on the behaviour of binary Fe-Cu and Fe-Au model alloys designed to explore routes to create new creep resistant steels having an in-built ability to autonomously fill creep induced porosity at grain boundaries. The alloying elements were selected on the basis

FE-simulation of the viscoplastic behaviour of different RPV steels in the frame of in-vessel melt retentions scenarios

International Nuclear Information System (INIS)

Altstadt, E.; Willschuetz, H.G.; Mueller, G.

2004-01-01

Assuming the hypothetical scenario of a severe accident with subsequent core meltdown and formation of a melt pool in the reactor pressure vessel (RPV) lower plenum of a Light Water Reactor (LWR) leads to the question about the behavior of the RPV. One accident management strategy could be to stabilize the in-vessel debris configuration in the RPV as one major barrier against uncontrolled release of heat and radio nuclides. To get an improved understanding and knowledge of the melt pool convection and the vessel creep and possible failure processes and modes occurring during the late phase of a core melt down accident the FOREVER-experiments (Failure Of REactor VEssel Retention) have been performed at the Division of Nuclear Power Safety of the Royal Institute of Technology Stockholm. These experiments are simulating the behavior of the lower head of the RPV under the thermal loads of a convecting melt pool with decay heating, and under the pressure loads that the vessel experiences in a depressurization scenario. The geometrical scale of the experiments is 1:10 compared to a common LWR. This paper deals with the experimental, numerical, and metallographical results of the creep failure experiment EC-FOREVER-4, where the American pressure vessel steel SA533B was applied for the lower head. For comparison the results of the experiment EC-FOREVER-3B, build of the French 16MND5 steel, are discussed, too. Emphasis is put on the differences in the viscoplastic behaviour of different heats of the RPV steel. For this purpose, the creep tests in the frame of the LHF/OLHF experiments are reviewed, too. As a hypothesis it is stated that the sulphur content could be responsible for differences in the creep behaviour. (orig.)

Relationship between strain and central deflection in small punch creep specimens

International Nuclear Information System (INIS)

Yang Zhen; Wang Zhiwen

2003-01-01

Acquiring information about creep strain directly from small punch creep tests is difficult because the deformation behaviour of the small punch specimen is complicated. A routine is suggested in the present paper to treat this problem indirectly. Based on a finite element analysis, it is proposed that the relationship of central deflection δ to central creep strain ε c of a specimen subjected to creep can be represented approximately by the relationship of central deflection δ to central (elastic-plastic) strain ε of a specimen not subjected to creep. With this hypothesis, the δ∼ε c relation of the small punch creep specimen is obtained by resorting to a rigid-plastic membrane stretch forming model. Finally, small punch creep test results are used to evaluate creep strain and creep strain rate by taking advantage of this δ∼ε c relation

Creep behavior of 8Cr2WVTa martensitic steel designed for fusion DEMO reactor. An assessment on helium embrittlement resistance

International Nuclear Information System (INIS)

Yamamoto, Norikazu; Murase, Yoshiharu; Nagakawa, Johsei; Shiba, Kiyoyuki

2001-01-01

Mechanical response against transmutational helium production, alternatively susceptibility to helium embrittlement, in a nuclear fusion reactor was examined on 8Cr2WVTa martensitic steel, a prominent structural candidate for advanced fusion systems. In order to simulate DEMO (demonstrative) reactor environments, helium was implanted into the material at 823 K with concentrations up to 1000 appmHe utilizing an α-beam from a cyclotron. Creep rupture properties were subsequently determined at the same temperature and were compared with those of the material without helium. It has been proved that helium caused no meaningful deterioration in terms of both the creep lifetime and rupture elongation. Furthermore, failure occurred completely in a transgranular and ductile manner even after high concentration helium introduction and there was no symptom of grain boundary decohesion which very often arises in helium bearing materials. These facts would mirror preferable resistance of this steel toward helium embrittlement. (author)

Irradiation creep and creep rupture of titanium-modified austenitic stainless steels and their dependence on cold work level

International Nuclear Information System (INIS)

Garner, F.A.; Hamilton, M.L.; Eiholzer, C.R.; Toloczko, M.B.; Kumar, A.S.

1991-11-01

A titanium-modified austenitic type stainless steel was tested at three cold work levels to determine its creep and creep rupture properties under both thermal aging and neutron irradiation conditions. Both the thermal and irradiation creep behavior exhibit a complex non-monotonic relationship with cold work level that reflects the competition between a number of stress-sensitive and temperature-dependent microstructural processes. Increasing the degree of cold work to 30% from the conventional 20% level was detrimental to its performance, especially for applications above 550 degrees c. The 20% cold work level is preferable to the 10% level, in terms of both in-reactor creep rupture response and initial strength

Examination of the creep behaviour of microstructurally unstable ferritic steels

International Nuclear Information System (INIS)

Williams, K.R.

1981-01-01

The inherent microstructural instability of 1/2Cr 1/2Mo 1/4V; 21/4Cr 1Mo and carbon steels creep tested or service exposed at low stresses is demonstrated. Measurements of important dispersion parameters have been made during creep life and have been found to follow normal coarsening kinetics. Using the measured time dependent change of the dispersion parameters, a dislocation source controlled model for recovery creep is used and further developed. The model allows the calculation of the Manson-Haferd plot of log (time to failure) against temperature for unstable steels. In addition, a classification of material stability is proposed, based on the ratio of time to fracture, t(sub f), and time to tertiary creep, tsub(t). This classification enables estimates of remaining creep life to be based either on well established criteria for stable materials or modifications of these criteria for unstable steels. (author)

Dilatational behaviour of ZrNb1 fuel cans of a WWER-type reactor during a loss-of-coolant accident

International Nuclear Information System (INIS)

Adam, E.; Stephan, M.; Wetzel, L.

1987-01-01

Based on an assessment of various factors of influence on the performance of fuel cans during normal operation and imaginable accidents, the necessity of studying creep and burst behaviour of WWER-type fuel cans of ZrNb1 under simulated LOCA conditions has been proved and an experimental facility designed for this purpose is described. Control of fuel can temperature is accomplished through a minicomputer during the creep and bursts experiments. With this, various temperature loading profiles of the fuel cans can be realized. Experimental results on dilatational behaviour of ZrNb1 fuel cans from isothermal creep and burst experiments in air are presented and compared with values for Zircaloy. (author)

Material development for gas-cooled high temperature reactors for the production of nuclear process heat

International Nuclear Information System (INIS)

Nickel, H.

1977-04-01

In the framework of the material development for gas-cooled high temperature reactors, considerable investigations of the materials for the reactor core and the primary cicuit are being conducted. Concerning the core components, the current state-of-the-art and the objectives of the development work on the spherical fuel elements, coated particles and structural graphite are discussed. As an example of the structural graphite, the non-replaceable reflector of the process heat reactor is discussed. The primary circuit will be constructed mainly from metallic materials, although some ceramics are also being considered. Components of interest are hot gas ducts, liners, methane reformer tubes and helium-helium intermediate heat exchangers. The gaseous impurities present in the helium coolant may cause oxidation and carburization of the nickel-base and iron-base alloys envisaged for use in these components, with a possible associated adverse effect on the mechanical properties such as creep and fatigue. Test capacity has therefore been installed to investigate materials behaviour in simulated reactor helium under both constant and alternating stress conditions. The first results on the creep behaviour of several alloys in impure helium are presented and discussed. (orig./GSC) [de

Irradiation creep of dispersion strengthened copper alloy

Energy Technology Data Exchange (ETDEWEB)

Pokrovsky, A.S.; Barabash, V.R.; Fabritsiev, S.A. [and others

1997-04-01

Dispersion strengthened copper alloys are under consideration as reference materials for the ITER plasma facing components. Irradiation creep is one of the parameters which must be assessed because of its importance for the lifetime prediction of these components. In this study the irradiation creep of a dispersion strengthened copper (DS) alloy has been investigated. The alloy selected for evaluation, MAGT-0.2, which contains 0.2 wt.% Al{sub 2}O{sub 3}, is very similar to the GlidCop{trademark} alloy referred to as Al20. Irradiation creep was investigated using HE pressurized tubes. The tubes were machined from rod stock, then stainless steel caps were brazed onto the end of each tube. The creep specimens were pressurized by use of ultra-pure He and the stainless steel caps subsequently sealed by laser welding. These specimens were irradiated in reactor water in the core position of the SM-2 reactors to a fluence level of 4.5-7.1 x 10{sup 21} n/cm{sup 2} (E>0.1 MeV), which corresponds to {approx}3-5 dpa. The irradiation temperature ranged from 60-90{degrees}C, which yielded calculated hoop stresses from 39-117 MPa. A mechanical micrometer system was used to measure the outer diameter of the specimens before and after irradiation, with an accuracy of {+-}0.001 mm. The irradiation creep was calculated based on the change in the diameter. Comparison of pre- and post-irradiation diameter measurements indicates that irradiation induced creep is indeed observed in this alloy at low temperatures, with a creep rate as high as {approx}2 x 10{sup {minus}9}s{sup {minus}1}. These results are compared with available data for irradiation creep for stainless steels, pure copper, and for thermal creep of copper alloys.

Irradiation creep of dispersion strengthened copper alloy

International Nuclear Information System (INIS)

Pokrovsky, A.S.; Barabash, V.R.; Fabritsiev, S.A.

1997-01-01

Dispersion strengthened copper alloys are under consideration as reference materials for the ITER plasma facing components. Irradiation creep is one of the parameters which must be assessed because of its importance for the lifetime prediction of these components. In this study the irradiation creep of a dispersion strengthened copper (DS) alloy has been investigated. The alloy selected for evaluation, MAGT-0.2, which contains 0.2 wt.% Al 2 O 3 , is very similar to the GlidCop trademark alloy referred to as Al20. Irradiation creep was investigated using HE pressurized tubes. The tubes were machined from rod stock, then stainless steel caps were brazed onto the end of each tube. The creep specimens were pressurized by use of ultra-pure He and the stainless steel caps subsequently sealed by laser welding. These specimens were irradiated in reactor water in the core position of the SM-2 reactors to a fluence level of 4.5-7.1 x 10 21 n/cm 2 (E>0.1 MeV), which corresponds to ∼3-5 dpa. The irradiation temperature ranged from 60-90 degrees C, which yielded calculated hoop stresses from 39-117 MPa. A mechanical micrometer system was used to measure the outer diameter of the specimens before and after irradiation, with an accuracy of ±0.001 mm. The irradiation creep was calculated based on the change in the diameter. Comparison of pre- and post-irradiation diameter measurements indicates that irradiation induced creep is indeed observed in this alloy at low temperatures, with a creep rate as high as ∼2 x 10 -9 s -1 . These results are compared with available data for irradiation creep for stainless steels, pure copper, and for thermal creep of copper alloys

Multiaxial creep of tubes of Alloy 800 and Alloy 617 at high temperature

International Nuclear Information System (INIS)

Penkalla, H.J.; Schubert, F.; Nickel, H.

1989-01-01

The deformation behaviour under multiaxial loading at temperature higher than 800 deg. C is strongly controlled by creep. For dimensioning and inelastic analysis the use of v. Mises theory and Norton's creep law for stationary creep are demonstrated for different combination of internal pressure and axial or torsional stress or strains. The experimental results are in satisfactory agreement with the theoretical predicted deformation behaviour if values for the coefficient k and n in Norton's creep law are used, which are close to the real creep resistance in the component. (author). 11 refs, 12 figs, 2 tabs

Irradiation creep lifetime analysis on first wall structure materials for CFETR

Energy Technology Data Exchange (ETDEWEB)

Ma, Bing; Peng, Lei, E-mail: penglei@ustc.edu.cn; Zhang, Xiansheng; Shi, Jingyi; Zhan, Jie

2017-05-15

Fusion reactor first wall services on the conditions of high surface heat flux and intense neutron irradiation. For China Fusion Engineering Test Reactor (CFETR) with high duty time factor, it is important to analyze the irradiation effect on the creep lifetime of the main candidate structure materials for first wall, i.e. ferritic/martensitic steel, austenite steel and oxide dispersion strengthened steel. The allowable irradiation creep lifetime was evaluated with Larson-Miller Parameter (LMP) model and finite element method. The results show that the allowable irradiation creep lifetime decreases with increasing of surface heat flux, first wall thickness and inlet coolant temperature. For the current CFETR conceptual design, the lifetime is not limited by thermal creep or irradiation creep, which indicated the room for design parameters optimization.

Creep and creep fatigue crack behavior of 1Cr- and 9Cr-steels

International Nuclear Information System (INIS)

Maile, K.; Klenk, A.; Schellenberg, G.; Granacher, J.; Tramer, M.

2000-01-01

A large database for creep crack initiation and propagation under constant load conditions is available on conventional power plant steels of types 1%Cr and 12%Cr. Modern plants are often used in the medium and peak load regime, thus the dominant loading situation in high temperature components is creep fatigue. For life assessment data about crack initiation and growth under creep fatigue loading are required. These characteristics can not be substituted by pure fatigue or creep crack data. Therefore, a comprehensive test programme was started to investigate the creep fatigue crack behaviour of a 1%CrMoNiV turbine rotor steel (30CrMoNiV 4 11) at 550 C and a new 9%CrMoVNb pipe steel (type P 9 1) at 600 C. DENT-specimen with 15 and 60 mm thickness as well as side grooved CT-specimen with 25 and 50 mm thickness have been tested to determine possible influences of geometry and thus to check the transferability of the data to components. The creep fatigue crack growth results of tests with dwell times between t H = 0,32h and 10 h lie in the scatterbands given by creep crack growth results. Nevertheless a higher crack growth rate under creep fatigue conditions can be stated. An increase in crack growth rate due to creep fatigue is clearly visible. Loading situations with frequencies higher than 1.10 -4 Hz should be not assessed with pure creep crack results or sufficient safety margins have to be applied. (orig.)

Unaxial stress relaxation and creep behaviour in weldments of the pressure vessel steel A533B between 600 and 640 degree C

International Nuclear Information System (INIS)

Otterberg, R.

1979-10-01

In order to predict the stress reduction during stress relief heat treatment in welded joints of the pressure vessel steel A533B, uniaxial stress relaxation as well as creep tests have been performed. The specimens were isothermally stress relaxed between 600 and 640 degree C from initial stresses corresponding to specimen elongations of 0.25, 0.5 and 0.2 percent. The stress relaxation results are excellently described by a Norton relationship. The magnitude of the initial stress has been found to affect the stress relaxation in the beginning of the tests, but at times longer than one hour the effect is very small. Creep strain data from creep tests in the actual temperature interval was converted to describe stress relaxation behaviour as well. The results will be used in a forthcoming study to predict the multiaxial stress reduction in thick weldments of A533B. (author)

Creep fracture mechanics analysis for through-wall cracked pipes under widespread creep condition

International Nuclear Information System (INIS)

Huh, Nam Su; Kim, Yun Jae; Kim, Young Jin

2003-01-01

This paper compares engineering estimation schemes of C * and creep COD for circumferential and axial through-wall cracked pipes at elevated temperatures with detailed 3-D elastic-creep finite element results. Engineering estimation schemes included the GE/EPRI method, the reference stress method where reference stress is defined based on the plastic limit load and the enhanced reference stress method where the reference stress is defined based on the optimized reference load. Systematic investigations are made not only on the effect of creep-deformation behaviour on C * and creep COD, but also on effects of the crack location, the pipe geometry, the crack length and the loading mode. Comparison of the FE results with engineering estimations provides that for idealized power law creep, estimated C * and COD rate results from the GE/EPRI method agree best with FE results. For general creep-deformation laws where either primary or tertiary creep is important and thus the GE/EPRI method is hard to apply, on the other hand, the enhanced reference stress method provides more accurate and robust estimations for C * and COD rate than the reference stress method

Creep rupture properties of solution annealed and cold worked type 316 stainless steel cladding tubes

International Nuclear Information System (INIS)

Mathew, M.D.; Latha, S.; Mannan, S.L.; Rodriguez, P.

1990-01-01

Austenitic stainless steels (mainly type 316 and its modifications) are used as fuel cladding materials in all current generation fast breeder reactors. For the Fast Breeder Test Reactor (FBTR) at Kalpakkam, modified type 316 stainless steel (SS) was chosen as the material for fuel cladding tubes. In order to evaluate the influence of cold work on the performance of the fuel element, the investigation was carried out on cladding tubes in three metallurgical conditions (solution annealed, ten percent cold worked and twenty percent cold worked). The results indicate that: (i) The creep strength of type 316 SS cladding tube increases with increasing cold work. (ii) The benificial effects of cold work are retained at almost all the test conditions investigated. (iii) The Larson Miller parameter analysis shows a two slope behaviour for 20PCW material suggesting that caution should be exercised in extrapolating the creep rupture life to stresses below 125 MPa. At very low stress levels, the LMP values fall below the values of the 10 PCW material. (author). 6 refs., 19 figs. , 10 tabs

Advances in the assessment of creep data

Energy Technology Data Exchange (ETDEWEB)

Holdsworth, S.R.

2010-07-01

Many of the classical models representing the creep and rupture behaviour of metals were developed prior to and during the 1950s and 1960s, and their subsequent exploitation, in particular for the assessment of large creep property datasets, was initially limited by the capability of the analytical tools available at the time. The formation of ECCC (the European Creep Collaborative Committee) in 1991, with a main objective of providing reliable peer reviewed long-time creep property values for European Design and Product Standards, led to the development of rigorous assessment procedures such as PD6605 and DESA incorporating post assessment tests to verify: physical realism, effectiveness of model-fit within the range of the source experimental data, and extrapolation credibility. The first ECCC assessment recommendations published in 1996 undoubtedly provided a catalyst for others to exploit the availability of low cost, powerful desktop computers to develop rigorous methodologies for the physically realistic analysis of uniaxial and multi-axial data for the reliable and accurate characterisation of creep strain, and rupture strength and ductility properties. More recent improvements in data assessment methodologies have been driven by the need to effectively model the creep deformation and rupture characteristics of the complex new generation alloys and fabrications being designed to cater for the continually evolving requirements of modern advanced power plant. These advances in the assessment of creep data are reviewed. (orig.)

Behaviour of Epoxy Silica Nanocomposites Under Static and Creep Loading

Science.gov (United States)

Constantinescu, Dan Mihai; Picu, Radu Catalin; Sandu, Marin; Apostol, Dragos Alexandru; Sandu, Adriana; Baciu, Florin

2017-12-01

Specific manufacturing technologies were applied for the fabrication of epoxy-based nanocomposites with silica nanoparticles. For dispersing the fillers in the epoxy resin special equipment such as a shear mixer and a high energy sonicator with temperature control were used. Both functionalized and unfunctionalized silica nanoparticles were added in three epoxy resins. The considered filling fraction was in most cases 0.1, 0.3 and 0.5 wt%.. The obtained nanocomposites were subjected to monotonic uniaxial and creep loading at room temperature. The static mechanical properties were not significantly improved regardless the filler percentage and type of epoxy resin. Under creep loading, by increasing the stress level, the nanocomposite with 0.1 wt% silica creeps less than all other materials. Also the creep rate is reduced by adding silica nanofillers.

CANSWEL-2: a computer model of the creep deformation of Zircaloy cladding under loss-of-coolant accident conditions

International Nuclear Information System (INIS)

Haste, T.J.

1982-07-01

The CANSWEL-2 code models cladding creep deformation under conditions relevant to a loss-of-coolant accident (LOCA) in a pressurised water reactor (PWR). It considers in detail the centre rod of a 3 x 3 nominally square array, taking into account azimuthal non-uniformities in cladding thickness and temperature, and the mechanical restraint imposed on contact with neighbouring rods. Any of the rods in the array may assume a non-circular shape. Models are included for primary and secondary creep, dynamic phase change and superplasticity when both alpha- and beta-phase Zircaloy are present. A simple treatment of oxidation strengthening is incorporated. Account is taken of the anisotropic creep behaviour of alpha-phase Zircaloy which leads to cladding bowing. The CANSWEL-2 model is used both as a stand-alone code and also as part of the LOCA analysis code MABEL-2. (author)

Problems of space-time behaviour of nuclear reactors

International Nuclear Information System (INIS)

Obradovic, D.

1966-01-01

This paper covers a review of literature and mathematical methods applied for space-time behaviour of nuclear reactors. The review of literature is limited to unresolved problems and trends of actual research in the field of reactor physics [sr

Creep crack growth verification testing in alloy 800H tubular components

International Nuclear Information System (INIS)

Hunter, C.P.; Hurst, R.C.

1992-01-01

A method for determining the creep crack growth, CCG, and stress rupture behaviour of Alloy 800H tubular components containing longitudinal notches at 800deg C is described. The presence of the notch is found to systematically weaken the tube, the degree of weaking dependent upon the notch length and depth. The creep crack growth rates, determined from a specially adapted potential drop technique are compared with those obtained from conventional compact tension type specimens. Using the stress intensity factor, K 1 , and the C * parameter as the basis of comparison it is found that the latter gives excellent correlation between the specimen and component behaviour. Finally attention is drawn to the potential dangers of predicting the component creep crack growth behaviour from the data obtained using conventional specimens for a structure sensitive material such as Alloy 800H and conversely to the advantages of the component type CCG tests developed in the present work. (orig.)

The thermal fatigue behaviour of creep-resistant Ni-Cr cast steel

Directory of Open Access Journals (Sweden)

B. Piekarski

2007-12-01

Full Text Available The study gives a summary of the results of industrial and laboratory investigations regarding an assessment of the thermal fatigue behaviour of creep-resistant austenitic cast steel. The first part of the study was devoted to the problem of textural stresses forming in castings during service, indicating them as a cause of crack formation and propagation. Stresses are forming in carbides and in matrix surrounding these carbides due to considerable differences in the values of the coefficients of thermal expansion of these phases. The second part of the study shows the results of investigations carried out to assess the effect of carbon, chromium and nickel on crack resistance of austenitic cast steel. As a criterion of assessment the amount and propagation rate of cracks forming in the specimens as a result of rapid heating followed by cooling in running water was adopted. Tests were carried out on specimens made from 11 alloys. The chemical composition of these alloys was comprised in a range of the following values: (wt-%: 18-40 %Ni, 17-30 %Cr, 1.2-1.6%Si and 0.05-0.6 %C. The specimens were subjected to 75 cycles of heating to a temperature of 900oC followed by cooling in running water. After every 15 cycles the number of the cracks was counted and their length was measured. The results of the measurements were mathematically processed. It has been proved that the main factor responsible for an increase in the number of cracks is carbon content in the alloy. In general assessment of the results of investigations, the predominant role of carbon and of chromium in the next place in shaping the crack behaviour of creep-resistant austenitic cast steel should be stressed. Attention was also drawn to the effect of high-temperature corrosion as a factor definitely deteriorating the cast steel resistance to thermal fatigue.

New results in the limit analysis by secondary modified creep

International Nuclear Information System (INIS)

Feijoo, R.A.; Taroco, E.; Zouain, N.

1982-03-01

Two methods for computing upper and lower bounds of colapse loads are proposed by means of generalized creep constitutive relations. The actual material behaviour is rigid-perfectly plastic and the techniques here analized consist in the substitution of this material by a fictitious one which presents steady state creep response. Some analytical examples are also presented. (Author) [pt

Studies in Phebus reactor of fuel behaviour upon LOCA conditions

International Nuclear Information System (INIS)

Manin, A.; Del Negro, R.; Reocreux, M.

1980-09-01

The fuel behaviour upon LOCA conditions is studied in an in-pile loop, in Phebus reactor. This paper presents: a short description of Phebus reactor; the current program (adjusting the thermohydraulic conditions in order to get cladding failure); the program developments (consequences involved by cladding failure); the fuel test conditions determination [fr

Design basis for creep of zirconium alloy components in a fast neutron flux

International Nuclear Information System (INIS)

Ross-Ross, P.A.; Fidleris, V.

1974-01-01

The Chalk River Nuclear Laboratory's experience with the creep of zirconium alloys in a neutron flux is described. Fast neutron flux changes the creep behavior of zirconium alloys and new design criteria for in-reactor applications are needed. From experimental results empirical relations describing the effects of neutron flux, stress, temperature, time, and anisotropy on creep rate were established. The relations are applied to the design of pressure tubes. (author)

Fast reactor fuel pin behaviour modelling in the UK

Energy Technology Data Exchange (ETDEWEB)

Matthews, J R [UKAEA, Harwell, Didcot, Oxon (United Kingdom); Hughes, H [Springfields Nuclear Power Development Laboratories, Springfields, Salwick, Preston (United Kingdom)

1979-12-01

Two fuel behaviour codes have been applied extensively to fast reactor problems; SLEUTH developed at Sprlngfields Nuclear Laboratory and FRUMP at A.E.R.E. Harwell. The SLEUTH fuel pin endurance code was originally developed to define a programme of power cycling and power ramp experiments In Advanced Gas Cooled Reactors (AGRs) where, because of the very soft cladding, pellet clad interaction is severe. The code was required to define accelerated test conditions to generalise from the observed endurance to that under other power histories and to select for investigation the most significant design, material and operational variables. The weak clad and low coolant pressure combine to make fission gas swelling a major contributor to clad deformation while the high clad ductility renders the distribution of strain readily observable. This has led to a detailed study of strain concentrations using the SEER code. SLEUTH and SEER have subsequently been used to specify power cycling and power ramp 112 experiments in water cooled, fast and materials testing reactors with the aim of developing a unified quantitative model of pellet-clad interaction whatever the reactor system. The FRUMP fuel behaviour code was developed specifically for the interpretation of fast reactor fuel pin behaviour. Experience with earlier models was valuable In its development. Originally the model was developed to describe behaviour during normal operation, but subsequently the code has been used extensively in the field of accident studies. Much of the effort in FRUMP development has been devoted to the production of physical models of the various effects of irradiation and the temperature gradients on the structure of the fuel and clad. Each process is modelled as well as is permitted by current knowledge and the limitations of computing costs. Each sub-model has a form which reflects the underlying mechanisms, where quantities are unknown values are assigned semi-empirically, i.e. coefficients

Fast reactor fuel pin behaviour modelling in the UK

International Nuclear Information System (INIS)

Matthews, J.R.; Hughes, H.

1979-01-01

Two fuel behaviour codes have been applied extensively to fast reactor problems; SLEUTH developed at Sprlngfields Nuclear Laboratory and FRUMP at A.E.R.E. Harwell. The SLEUTH fuel pin endurance code was originally developed to define a programme of power cycling and power ramp experiments In Advanced Gas Cooled Reactors (AGRs) where, because of the very soft cladding, pellet clad interaction is severe. The code was required to define accelerated test conditions to generalise from the observed endurance to that under other power histories and to select for investigation the most significant design, material and operational variables. The weak clad and low coolant pressure combine to make fission gas swelling a major contributor to clad deformation while the high clad ductility renders the distribution of strain readily observable. This has led to a detailed study of strain concentrations using the SEER code. SLEUTH and SEER have subsequently been used to specify power cycling and power ramp 112 experiments in water cooled, fast and materials testing reactors with the aim of developing a unified quantitative model of pellet-clad interaction whatever the reactor system. The FRUMP fuel behaviour code was developed specifically for the interpretation of fast reactor fuel pin behaviour. Experience with earlier models was valuable In its development. Originally the model was developed to describe behaviour during normal operation, but subsequently the code has been used extensively in the field of accident studies. Much of the effort in FRUMP development has been devoted to the production of physical models of the various effects of irradiation and the temperature gradients on the structure of the fuel and clad. Each process is modelled as well as is permitted by current knowledge and the limitations of computing costs. Each sub-model has a form which reflects the underlying mechanisms, where quantities are unknown values are assigned semi-empirically, i.e. coefficients

Creep deformation mechanisms in a γ titanium aluminide

Energy Technology Data Exchange (ETDEWEB)

Abdallah, Zakaria [Institute of Structural Materials, College of Engineering, Bay Campus, Swansea University, Swansea SA18EN (United Kingdom); Ding, Rengen [School of Metallurgy and Materials, University of Birmingham, Edgbaston, Birmingham B152TT (United Kingdom); Martin, Nigel; Dixon, Mark [Rolls-Royce plc, P.O. Box 31, Derby DE248BJ (United Kingdom); Bache, Martin [Institute of Structural Materials, College of Engineering, Bay Campus, Swansea University, Swansea SA18EN (United Kingdom)

2016-09-15

Titanium aluminides (TiAl) are considered as potential alternatives to replace nickel-based alloys of greater density for selected components within future gas turbine aero-engines. This is attributed to the high specific strength as well as the good oxidation resistance at elevated temperatures. The gamma (γ) titanium aluminide system Ti-45Al-2Mn-2Nb has previously demonstrated promising performance in terms of its physical and mechanical properties. The main aim of the current study, which is a continuation of a previously published paper, aims at evaluating the performance of this titanium aluminide system under high temperature creep conditions. Of particular interest, the paper is strongly demonstrating the precise capability of the Wilshire Equations technique in predicting the long-term creep behaviour of this alloy. Moreover, it presents a physically meaningful understanding of the various creep mechanisms expected under various testing conditions. To achieve this, two creep specimens, tested under distinctly different stress levels at 700 °C have been extensively examined. Detailed microstructural investigations and supporting transmission electron microscopy (TEM) have explored the differences in creep mechanisms active under the two stress regimes, with the deformation mechanisms correlated to Wilshire creep life prediction curves.

Review of recent irradiation-creep results

International Nuclear Information System (INIS)

Coghlan, W.A.

1982-05-01

Materials deform faster under stress in the presence of irradiation by a process known as irradiation creep. This phenomenon is important to reactor design and has been the subject of a large number of experimental and theoretical investigations. The purpose of this work is to review the recent experimental results to obtain a summary of these results and to determine those research areas that require additional information. The investigations have been classified into four subgroups based on the different experimental methods used. These four are: (1) irradiation creep using stress relaxation methods, (2) creep measurements using pressurized tubes, (3) irradiation creep from constant applied load, and (4) irradiation creep experiments using accelerated particles. The similarity and the differences of the results from these methods are discussed and a summary of important results and suggested areas for research is presented. In brief, the important results relate to the dependence of creep on swelling, temperature, stress state and alloying additions. In each of these areas new results have been presented and new questions have arisen which require further research to answer. 65 references

Creep-fatigue rules in the RCC-MR code

International Nuclear Information System (INIS)

Drubay, B.

1988-01-01

In 1978, CEA, Electricite de France (EDF) and NOVATOME decided to draw up a complete set of design and construction rules for LMFBR components. This RCC-MR code issued in June 1985 and completed in November 1987 was chosen as a sound basis for the next European Fast Reactor (EFR). The purpose of this paper is to describe the present RCC-MR creep-fatigue design rules to be applied with elastic analysis including the modifications adopted in the first addenda. This method is based on a separate evaluation of a fatigue usage fraction V and creep rupture usage fraction W with the common linear summation rule. The fatigue usage fraction is obtained from continuous fatigue curves (without hold times) and from total strain ranges (elastic + plastic + creep). The creep rupture usage fraction W is obtained from stress to rupture curves and a stress σk evaluating the stress generated during the cycle. (author)

Low stress creep of stainless steel

International Nuclear Information System (INIS)

Crossland, I.G.; Clay, B.D.; Baker, C.

1976-06-01

The creep of 20%Cr, 25%Ni, Nb stainless steel has been examined at temperatures from 675 to 775 0 C at sheer stressed below 13 MPa and grain sizes from 6 to 20μm. The results have indicated that the initial creep rates were linearly dependent upon stress but with a threshold stress below which no creep occurred, i.e. Bingham behaviour; in addition, the creep activation energy at small strains was substantially lower than the lattice self-diffusion value and the initial creep rates were approximately related to the grain size through an inverse cube relation. It has been concluded that at low strains (approaching the initial elastic deflection) the creep mechanism was probably that of grain boundary diffusion creep (Coble, 1963) and this is further supported by the close agreement between the observed and theoretically predicted creep rate values. Steady-state creep rates were not observed; initially the creep rates fell rapidly with strain after which a more gradual decrease occurred. Whilst the creep rate - stress relationship continued to be of a Bingham form, the progressive reduction in creep rate with strain was found to be mainly attributable to an increase in the effective viscosity, threshold stress effects being generally of secondary importance. A model has been proposed which explains the initial creep rates as being due to Cable creep with elastic accommodation at grain boundary particles. At higher strains grain boundary collapse caused by vacancy sinking is accommodated at precipitate particles by plastic deformation of the adjacent matrix material. (author)

A comparative study of creep rupture behaviour of modified 316L(N) base metal and 316L(N)/16-8-2 weldment in air and liquid sodium environments

International Nuclear Information System (INIS)

Mishra, M.P.; Mathew, M.D.; Mannan, S.L.; Rodriguez, P.; Borgstedt, H.U.

1997-01-01

Creep rupture behaviour of modified type 316L(N) stainless steel base metal and weldments prepared with 16-8-2 filler wire has been investigated in air and flowing sodium environments at 823 K. No adverse environmental effects have been noticed due to sodium on the creep rupture behaviour of these weldments for tests up to 10 000 h. Rupture lives of the weldment were higher in the sodium environment than those in air. Rupture lives of the weldments were found to be lower than those of the base metal by a factor of two to five in both air and sodium environments. Minimum creep rates were essentially the same for the weldment as well as for the base metal in both the environments, whereas rupture strain was usually lower for the weldment than that of the base metal. The reduction in area of the weldment specimens increased with increase in stress. Failures in the specimens of weldments were in the weld metal region. Microstructural studies carried out on failed weldment specimens after the creep rupture tests revealed extensive cavitation in the weld metal region in air tested specimens predominantly at the austerite/δ-ferrite interphase. However, no cavitation was observed in specimens tested in sodium. (author)

Creep fatigue assessment for EUROFER components

Energy Technology Data Exchange (ETDEWEB)

Özkan, Furkan, E-mail: oezkan.furkan@partner.kit.edu; Aktaa, Jarir

2015-11-15

Highlights: • Design rules for creep fatigue assessment are developed to EUROFER components. • Creep fatigue assessment tool is developed in FORTRAN code with coupling MAPDL. • Durability of the HCPB-TBM design is discussed under typical fusion reactor loads. - Abstract: Creep-fatigue of test blanket module (TBM) components built from EUROFER is evaluated based on the elastic analysis approach in ASME Boiler Pressure Vessel Code (BPVC). The required allowable number of cycles design fatigue curve and stress-to-rupture curve to estimate the creep-fatigue damage are used from the literature. Local stress, strain and temperature inputs for the analysis of creep-fatigue damage are delivered by the finite element code ANSYS utilizing the Mechanical ANSYS Parametric Design Language (MAPDL). A developed external FORTRAN code used as a post processor is coupled with MAPDL. Influences of different pulse durations (hold-times) and irradiation on creep-fatigue damage for the preliminary design of the Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM) are discussed for the First Wall component of the TBM box.

Creep lifetime assessements of ferritic pipeline welds

International Nuclear Information System (INIS)

Ainsworth, R.A.; Goodall, I.W.; Miller, D.A.

1995-01-01

The low alloy ferritic steam pipework in Advanced Gas Cooled reactor (AGR) power stations operates at temperatures in the creep range. An inspection strategy for continued operation of the pipework has been developed based on estimation of the creep rupture life of pipework weldments and fracture mechanics for demonstrating acceptance of defects. This strategy is described in outline. The estimation of creep rupture life is described in more detail. Validation for the approach is illustrated by comparison with pressure vessel tests and with metallographic examination of components removed from service. The fracture mechanics methods are also described. It is shown that the amount of creep crack growth is dependent on the life fraction at which the assessment is made; crack growth being rapid as the creep rupture life is approached. (author). 3 refs., 5 figs., 1 tab

Proceedings of the international conference on irradiation behaviour of metallic materials for fast reactor core components

International Nuclear Information System (INIS)

Poirier, J.; Dupouy, J.M.

Radiation effects on metals or alloys used in fast reactor core components are examined in the papers presented at this conference, the accent being put on swelling and irradiation creep of steels and nickel alloys

High temperature creep-fatigue design

International Nuclear Information System (INIS)

Tavassoli, A. A. F.; Fournier, B.; Sauzay, M.

2010-01-01

Generation IV fission and future fusion reactors envisage development of more efficient high temperature concepts where materials performances are key to their success. This paper examines different types of high temperature creep-fatigue interactions and their implications on design rules for the structural materials retained in both programmes. More precisely, the paper examines current status of design rules for the stainless steel type 316L(N), the conventional Modified 9Cr-1Mo martensitic steel and the low activation Eurofer steel. Results obtained from extensive high temperature creep, fatigue and creep-fatigue tests performed on these materials and their welded joints are presented. These include sequential creep-fatigue and relaxation creep-fatigue tests with hold times in tension, in compression or in both. Effects of larger plastic deformations on fatigue properties are studied through cyclic creep tests or fatigue tests with extended hold time in creep. In most cases, mechanical test results are accompanied with microstructural and fractographic observations. In the case of martensitic steels, the effect of oxidation is examined by performing creep-fatigue tests on identical specimens in vacuum. Results obtained are analyzed and their implications on design allowable and creep-fatigue interaction diagrams are presented. While reasonable confidence is found in predicting creep-fatigue damage through existing code procedures for austenitic stainless steels, effects of cyclic softening and coarsening of microstructure of martensitic steels throughout the fatigue life on materials properties need to be taken into account for more precise damage calculations. In the long-term, development of ferritic/martensitic steels with stable microstructure, such as ODS steels, is proposed. (authors)

High temperature creep-fatigue design

Energy Technology Data Exchange (ETDEWEB)

Tavassoli, A. A. F.; Fournier, B.; Sauzay, M. [CEA Saclay, DEN DMN, F-91191 Gif Sur Yvette (France)

2010-07-01

Generation IV fission and future fusion reactors envisage development of more efficient high temperature concepts where materials performances are key to their success. This paper examines different types of high temperature creep-fatigue interactions and their implications on design rules for the structural materials retained in both programmes. More precisely, the paper examines current status of design rules for the stainless steel type 316L(N), the conventional Modified 9Cr-1Mo martensitic steel and the low activation Eurofer steel. Results obtained from extensive high temperature creep, fatigue and creep-fatigue tests performed on these materials and their welded joints are presented. These include sequential creep-fatigue and relaxation creep-fatigue tests with hold times in tension, in compression or in both. Effects of larger plastic deformations on fatigue properties are studied through cyclic creep tests or fatigue tests with extended hold time in creep. In most cases, mechanical test results are accompanied with microstructural and fractographic observations. In the case of martensitic steels, the effect of oxidation is examined by performing creep-fatigue tests on identical specimens in vacuum. Results obtained are analyzed and their implications on design allowable and creep-fatigue interaction diagrams are presented. While reasonable confidence is found in predicting creep-fatigue damage through existing code procedures for austenitic stainless steels, effects of cyclic softening and coarsening of microstructure of martensitic steels throughout the fatigue life on materials properties need to be taken into account for more precise damage calculations. In the long-term, development of ferritic/martensitic steels with stable microstructure, such as ODS steels, is proposed. (authors)

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)

Secondary creep of porous metal supports for solid oxide fuel cells by a CDM approach

DEFF Research Database (Denmark)

Esposito, L.; Boccaccini, D. N.; Pucillo, G. P.

2017-01-01

The creep behaviour of porous iron-chromium alloy used in solid oxide fuel cells (SOFCs) becomes relevant under SOFC operating temperatures. In this paper, the secondary creep stage of infiltrated and non-infiltrated porous metal supports (MS) was investigated and theoretically modelled...... as function of temperature, determined by the high temperature impulse excitation technique, was directly used to account for the porosity and the related effective stress acting during the creep tests. The proposed creep rate formulation was used to extend the Crofer® 22 APU Monkman-Grant diagram...... in the viscous creep regime. The influence of oxide scale formation on creep behaviour of the porous MS was assessed by comparing the creep data of pre-oxidised samples tested in reducing atmosphere....

High temperature high vacuum creep testing facilities

International Nuclear Information System (INIS)

Matta, M.K.

1985-01-01

Creep is the term used to describe time-dependent plastic flow of metals under conditions of constant load or stress at constant high temperature. Creep has an important considerations for materials operating under stresses at high temperatures for long time such as cladding materials, pressure vessels, steam turbines, boilers,...etc. These two creep machines measures the creep of materials and alloys at high temperature under high vacuum at constant stress. By the two chart recorders attached to the system one could register time and temperature versus strain during the test . This report consists of three chapters, chapter I is the introduction, chapter II is the technical description of the creep machines while chapter III discuss some experimental data on the creep behaviour. Of helium implanted stainless steel. 13 fig., 3 tab

Creep characteristics in thick welded joints and their improvements. 2. Applicability of a simple model for creep analysis of thick welded joints

International Nuclear Information System (INIS)

Nakacho, Keiji; Ueda, Yukio; Kinugawa, Junichi; Yamazaki, Masayoshi

1997-01-01

Reliable predictions of the creep behavior of thick welded joints are very important to secure the safety of elevated temperature vessels like nuclear reactors. Creep behavior is very complex, thus it is difficult to perform the experiment and conduct the theoretical analysis. A simple accurate model for theoretical analysis was developed in the first report. The simple model is constructed of seven one-dimensional finite elements which can analyze not only one-dimensional stress creep behavior but also the three-dimensional situation. The simple model is verified by comparing the analyzed results with the experimental ones in this report. The model is easy to treat, and needs only a little labor and computation time to predict the creep curve and the local strain for a thick welded joint. (author)

Creep properties of base metal and welded joint of Hastelloy XR produced for High-Temperature Engineering Test Reactor in simulated primary coolant helium

International Nuclear Information System (INIS)

Kurata, Yuji; Tsuji, Hirokazu; Shindo, Masami; Suzuki, Tomio; Tanabe, Tatsuhiko; Mutoh, Isao; Hiraga, Kenjiro

1999-01-01

Creep tests of base metal, weld metal and welded joint of Hastelloy XR, which had the same chemical composition as Hastelloy XR produced for an intermediate heat exchanger of the High-Temperature Engineering Test Reactor, were conducted in simulated primary coolant helium. The weld metal and welded joint showed almost equal to or longer rupture time than the base metal of Hastelloy XR at 850 and 900degC, although they gave shorter rupture time at 950degC under low stress and at 1,000degC. The welded joint of Hastelloy XR ruptured at the base metal region at 850 and 900degC. On the other hand, it ruptured at the weld metal region at 950 and 1,000degC. The steady-state creep rate of weld metal of Hastelloy XR was lower than that of base metal at 850, 900 and 950degC. The creep rupture strengths of base metal, weld metal and welded joint of Hastelloy XR obtained in this study were confirmed to be much higher than the design allowable creep-rupture stress (S R ) of the Design Allowable Limits below 950degC. (author)

Microstructure-based assessment of creep rupture behaviour of cast-forged P91 steel

Energy Technology Data Exchange (ETDEWEB)

Pandey, Chandan, E-mail: chandanpy.1989@gmail.com [Department of Mechanical and Industrial Engineering, Indian Institute of Technology Roorkee, Uttrakhand 247667 (India); Mahapatra, M.M. [School of Mechanical Sciences, Indian Institute of Technology Bhubaneswar, Odisha 751013 (India); Kumar, Pradeep; Vidyrathy, R.S. [Department of Mechanical and Industrial Engineering, Indian Institute of Technology Roorkee, Uttrakhand 247667 (India); Srivastava, A. [Senior Engineer, HEEP Section, BHEL Haridwar (India)

2017-05-17

The work presented in this study was performed with the intent to characterize the microstructure evolution for short term creep exposure of cast-forged P91 steel. The short-term creep test was performed at temperature range of 620–650 °C and stresses ranging from 120 to 200 MPa. To characterize the sample after creep exposure, field emission scanning electron microscopy (FESEM) with energy dispersive X-ray spectroscopy (EDS), optical microscope and micro-hardness testing were utilized. Creep tests were performed on round creep specimens. For low temperature service condition, longer creep life was obtained. The fracture surface of creep ruptured specimen were characterized by using the FESEM. The transgranular fracture mode was noticed in all the tests condition. The creep rupture life was found to be decreased with increase in applied stress. The maximum rupture life was measured about to be 3329.28 h for the sample exposed at 620 °C for 120 MPa. A negligible microstructural change was measured in gripping area compared to the gauge area (necking area) of crept sample. The laves phase formation was also noticed along the grain boundaries for creep exposure life of 3329.28 h.

Microstructure-based assessment of creep rupture behaviour of cast-forged P91 steel

International Nuclear Information System (INIS)

Pandey, Chandan; Mahapatra, M.M.; Kumar, Pradeep; Vidyrathy, R.S.; Srivastava, A.

2017-01-01

The work presented in this study was performed with the intent to characterize the microstructure evolution for short term creep exposure of cast-forged P91 steel. The short-term creep test was performed at temperature range of 620–650 °C and stresses ranging from 120 to 200 MPa. To characterize the sample after creep exposure, field emission scanning electron microscopy (FESEM) with energy dispersive X-ray spectroscopy (EDS), optical microscope and micro-hardness testing were utilized. Creep tests were performed on round creep specimens. For low temperature service condition, longer creep life was obtained. The fracture surface of creep ruptured specimen were characterized by using the FESEM. The transgranular fracture mode was noticed in all the tests condition. The creep rupture life was found to be decreased with increase in applied stress. The maximum rupture life was measured about to be 3329.28 h for the sample exposed at 620 °C for 120 MPa. A negligible microstructural change was measured in gripping area compared to the gauge area (necking area) of crept sample. The laves phase formation was also noticed along the grain boundaries for creep exposure life of 3329.28 h.

Drucker-Prager-Cap creep modelling of pebble beds in fusion blankets

International Nuclear Information System (INIS)

Hofer, D.; Kamlah, M.

2005-01-01

Modelling of thermal and mechanical behaviour of pebble beds for fusion blankets is an important issue to understand the interaction of solid breeder and beryllium pebble beds with the surrounding structural material. Especially the differing coefficients of thermal expansion of these materials cause high stresses and strains during irradiation induced volumetric heating. To describe this process, the coupled thermomechanical behaviour of both pebble bed materials has to be modelled. Additionally, creep has to be considered contributing to bed deformations and stress relaxation. Motivated by experiments, we use a continuum mechanical approach called Drucker-Prager/Cap theory to model the macroscopic pebble bed behaviour. The model accounts for pressure dependent shear failure, inelastic hardening, and volumetric creep. The elastic part is described by a nonlinear elasticity law. The model has been implemented by user-defined routines in the commercial finite-element code ABAQUS. To check the numerics, the implementation is compared to an analytical solution. Furthermore, the Drucker-Prager/Cap tool is applied to a single ceramic breeder bed subject to creep under volumetric heating

Study on the creep constitutive equation of Hastelloy X, (1)

International Nuclear Information System (INIS)

Suzuki, Kazuhiko; Mutoh, Yasushi

1983-01-01

In order to carry out the structural design of high temperature pipings, intermediate heat exchangers and isolating valves for a multipurpose high temperature gas-cooled reactor, in which coolant temperature reaches 1000 deg C, the creep characteristics of Hastelloy X used as the heat resistant material must be clarified. In addition to usual creep rupture life and the time to reach a specified creep strain, the dependence of creep strain curves on time, temperature and stress must be determined and expressed with equations. Therefore, using the creep data of Hastelloy X given in the literatures, the creep constitutive equation was made. Since the creep strain curves under the same test condition were different according to heats, the sensitivity analysis of the creep constitutive equation was performed. The form of the creep constitutive equation was determined to be Garofalo type. The result of the sensitivity analysis is reported. (Kako, I.)

Dynamic behaviour of a CAREM type reactor

International Nuclear Information System (INIS)

Abbate, P.; Doval, A.

1990-01-01

As complement to CAREM reactor design studies, behaviour analysis were made in a non-stationary regime, with the aim of developing plant systems and determining process variables variation ranges, characteristic of normal operating conditions, specifying alarm values for different variables, as well as for operating policies. Transient accidental scenes analysis were made, concluding that reactor characteristics provide security, maintaining the core integrity. (Author) [es

Creep properties of 20% cold-worked Hastelloy XR

International Nuclear Information System (INIS)

Kurata, Y.

1996-01-01

The creep properties of Hastelloy XR, in solution-treated and in 20% cold-worked conditions, were studied at 800, 900 and 1000 C. At 800 C, the steady-state creep rate and rupture ductility decrease, while rupture life increases after cold work to 20%. Although the steady-state creep rate and ductility also decrease at 900 C, the beneficial effect on rupture life disappears. Cold work to 20% enhan ces creep resistance of this alloy at 800 and 900 C due to a high density of dislocations introduced by the cold work. Rupture life of the 20% cold-worked alloy becomes shorter and the steady-state creep rate larger at 1000 C during creep of the 20% cold-worked alloy. It is emphasized that these cold work effects should be taken into consideration in design and operation of high-temperature structural components of high-temperature gas-cooled reactors. (orig.)

Effect of dose on creep and recovery of polyethylene

Energy Technology Data Exchange (ETDEWEB)

Novakovic, Lj; Gal, O; Charlesby, A; Stannett, V T

1987-01-01

The effect of high energy radiation on polyethylene is to crosslink it, and connect it into an elastic network above the melting point. In this paper the creep and recovery properties of a stabilized polyethylene subjected to doses from 100 to 870 kGy are measured at 150/sup 0/C. Two cycles are measured - Creep I + Recovery I, and Creep II + Recovery II -mainly over periods of 20 min. The creep or recovery behaviour falls into three steps - immediate, fast and slow, and data are given for these steps together with the time parameter. The first cycle includes a non-recoverable creep which is almost absent in the second cycle.

Effect of dose on creep and recovery of polyethylene

International Nuclear Information System (INIS)

Novakovic, Lj.; Gal, O.; Charlesby, A.; Stannett, V.T.

1987-01-01

The effect of high energy radiation on polyethylene is to crosslink it, and connect it into an elastic network above the melting point. In this paper the creep and recovery properties of a stabilized polyethylene subjected to doses from 100 to 870 kGy are measured at 150 0 C. Two cycles are measured - Creep I + Recovery I, and Creep II + Recovery II -mainly over periods of 20 min. The creep or recovery behaviour falls into three steps - immediate, fast and slow, and data are given for these steps together with the time parameter. The first cycle includes a non-recoverable creep which is almost absent in the second cycle. (author)

Non-isothermal irradiation creep of nickel alloys Inconel 706 and PE-16

International Nuclear Information System (INIS)

Gilbert, E.R.; Chin, B.A.

1984-06-01

The results of in-reactor step temperature change experiments conducted on two nickel alloys, PE-16 and Inconel 706, were evaluated to determine the creep behavior under nonisothermal conditions. The effect of the temperature changes was found to be significantly different for the two alloys. Following a step temperature change, the creep rate of PE-16 adjusted to the rate found in isothermal tests at the new temperature. In contrast for Inconel 706, a reduction in temperature from 540 to 425 0 C produced a 300% increase in creep above that measured at 540 0 C in isothermal tests. The response of in-reactor creep in Inconel 706 to temperature changes was attributed to the dissolution of the gamma double-prime phase and subsequent loss of precipitation-strengthening at temperatures below 500 C

A stochastic approach to anelastic creep

International Nuclear Information System (INIS)

Venkataraman, G.

1976-01-01

Anelastic creep or the time-dependent yielding or a material subjected to external stresses has been found to be of great importantance in technology in the recent years, particularly in engineering structures including nuclear reactors wherein structural members may be under stress. The physics aspects underlying this phenomenon is dealt with in detail. The basics of time-dependent elasticity, constitutive relation, network models, constitutive equation in the frequency domain and its mearurements, and stochastic approach to creep are discussed. (K.B.)

Creep-fatigue evaluation method for modified 9Cr-1Mo steel

International Nuclear Information System (INIS)

Wada, Y.; Aoto, K.

1997-01-01

As creep-fatigue evaluation methods on normalized and tempered Modified 9Cr-1Mo steel for design use, the time fraction rule and the simplified conventional ductility exhaustion rule are investigated for the prediction of tension strain hold creep-fatigue damage of this material. For the above investigation, stress relaxation behaviour during strain hold has to be analyzed using stress-strain-time relation. The initial value of stress relaxation was determined by cyclic stress-strain curves in continuous cycling fatigue tests. Cyclic stress-strain behaviour of Mod.9Cr-1Mo(NT) steel is different from that of austenitic stainless steels, so this effect was considered. Stress relaxation analysis was performed using static creep strain-time relation and conventional hardening rule. The time fraction by using the above stress relaxation analysis results can give good prediction for creep-fatigue life of Mod.9Cr-1Mo(NT) steel. For design use it is practical to be able to estimate creep damages conservatively by both strain behaviour of cyclic plastic (in continuous cycling fatigue tests) and monotonic creep (in standard creep tests). The life reduction by strain hold at the minimum peak of compressive stress in creep-fatigue tests was examined, and this effects can be evaluated by the relationship between the location of oxidation and the effective deformation at crack tip. In an accelerated oxidation environment, for example in high temperature and high pressure steam, a different approach for life reduction should be developed based on the mechanism of growth of oxide and crack growth with oxidation. However, in the creep damage dominant region, its effect is saturated and the effect of cavity growth along grain boundary becomes dominant for long-term strain hold in the high temperature conditions. (author). 6 refs, 6 figs

Analysis of radiation exposure during creep adjustment to the coolant channels at Madras Atomic Power Station

International Nuclear Information System (INIS)

Varadhan, R.S.; Venkataramana, K.; Kannan, R.K.; Sreekumaran Nair, B.; Chudalayandi, K.

1994-01-01

In pressurised heavy water reactors the coolant channels made of zircaloy-2 undergo creep deformation used intense neutron irradiation in the reactor core. In order to measure and provide for the changes in the dimensions, base line data of internal diameters, sag and length of the 306 coolant channels are measured as pre service inspection (PSI) before the reactor is loaded with fuel prior to criticality. Subsequently as part of in service inspection (ISI), axial creep of every channel is measured in every annual shutdown of the reactor and creep adjustment is done on those channels where creep expansion margin for the next one year operation is low. A study was carried out to assess the radiological impact of the job at Madras Atomic Power Station (MAPS). Various measures adopted for reducing the individual and collective doses on the job are discussed in this report. (author). 3 refs., 2 tabs

Prediction of material creep behaviour for strain based life assessment applications

Energy Technology Data Exchange (ETDEWEB)

Rantala, J H; Hurst, R C [EC JRC IAM, Petten (Netherlands); Bregani, F [ENEL, Milan (Italy)

1999-12-31

In this work the idea of using constant load uniaxial creep test results instead of constant stress results for developing a CDM creep model for the P92 material is demonstrated. Due to limited availability of creep test results this work is based on incomplete test data and a general stress rupture line. In spite of these limitations a material creep model was developed for use in a FE analysis. Using P91 material as an example, a method is proposed to account for differences in strain evolution as a function of stress which normally manifests itself as lower strain values at low stresses in a normalised time-strain plot. This allows the CDM model to be used both in FE analysis and in strain-based life assessment engineering calculations. (orig.) 3 refs.

Prediction of material creep behaviour for strain based life assessment applications

Energy Technology Data Exchange (ETDEWEB)

Rantala, J.H.; Hurst, R.C. [EC JRC IAM, Petten (Netherlands); Bregani, F. [ENEL, Milan (Italy)

1998-12-31

In this work the idea of using constant load uniaxial creep test results instead of constant stress results for developing a CDM creep model for the P92 material is demonstrated. Due to limited availability of creep test results this work is based on incomplete test data and a general stress rupture line. In spite of these limitations a material creep model was developed for use in a FE analysis. Using P91 material as an example, a method is proposed to account for differences in strain evolution as a function of stress which normally manifests itself as lower strain values at low stresses in a normalised time-strain plot. This allows the CDM model to be used both in FE analysis and in strain-based life assessment engineering calculations. (orig.) 3 refs.

In-pile Creep Tests of Zircaloy Tubing in the Studsvik R2 Reactor. Final Report

International Nuclear Information System (INIS)

Tomani, Hans; Lindeloew, Ulf

2000-12-01

In this report are presented the findings of a prototype creep test on Zr4 guide tube specimens exposed in-pile and out-of-pile and stressed by constant bending moments. The calculated initial deflection curvature caused by the applied bending moment agrees very well with the measured initial values. Furthermore, the measurement results show excellent consistency. The dominant impact of neutron irradiation is clearly demonstrated. After 3 cycles (∼1300 hours) the irradiation creep is 4 times as large as the thermal creep. This is the case at least when fresh tube material is used. Irradiation creep progresses steadily, but the creep rate is not quite constant during the 3 irradiation cycles. The thermal creep, on the other hand, quickly saturates and there is hardly any further deflection after the second cycle for the specimen situated above the core. A limitation with the rig has been that the tube deflection became limited by the rig carrier body of the rig in the neutron flux (core) that disqualified the results of a fourth irradiation cycle actually performed in the fall of 1998. The test method appears to be suitable for testing the bending creep of different guide tube materials or designs under PWR conditions

The role of particle ripening on the creep acceleration of Nimonic 263 superalloy

Directory of Open Access Journals (Sweden)

Angella Giuliano

2014-01-01

Full Text Available Physically based constitutive equations need to incorporate the most relevant microstructural features of materials to adequately describe their mechanical behaviour. To accurately model the creep behaviour of precipitation hardened alloys, the value and the evolution of strengthening particle size are important parameters to be taken into account. In the present work, creep tests have been run on virgin and overaged (up to 3500 h at 800 ∘C Nimonic 263, a polycrystalline nickel base superalloy used for combustion chambers of gas turbines. The experimental results suggest that the reinforcing particle evolution is not the main reason for the creep acceleration that seems to be better described by a strain correlated damage, such as the accumulation of mobile dislocations or the grain boundary cavitation. The coarsened microstructure, obtained by overageing the alloy at high temperature before creep testing, mainly influences the initial stage of the creep, resulting in a higher minimum creep rate and a corresponding reduction of the creep resistance.

Creep properties of Nb-1Zr and Nb-1Zr-0.1C

International Nuclear Information System (INIS)

Horak, J.A.; Egner, L.K.

1994-12-01

In the early 1980s a compact, lithium cooled, fast-energy spectrum nuclear reactor was selected for space applications requiring prolonged uninterrupted electrical power. This reactor was to be capable of generating up to 100 kilowatts of electricity for times up to seven years in space and thus was given the acronym SP-100. The material selected for the fuel cladding, reactor heat transport systems and structural components was Nb-1 wt % Zr (Nb-1Zr). In addition to commercial Nb-1Zr, modified alloys containing 100--200 wt ppM each of carbon and nitrogen and 900 ± 150 wt ppM carbon were also included, Type B Nb-1Zr and PWC-11, respectively. The SP-100 reactor was designed to operate at temperatures of 1290--1425 K. At these temperatures the principal mode of deformation for Nb-1Zr is creep, and creep strain of the fuel cladding limits the useful reactor lifetime. To develop a creep data base for design, safety and reliability analyses, uniaxial creep testing of Nb-1Zr, Type B Nb-1Zr and PWC-11 was conducted from 1250--1450 K at stresses from 5.0 MPa to 41.4 MPa. Methodology and test results are presented

Development of Zr-2.5Nb pressure tubes for Advanced CANDU Reactor

International Nuclear Information System (INIS)

Bickel, G.A.; Griffiths, M.; Douchant, A.; Douglas, S.; Woo, O.T.; Buyers, A.

2010-01-01

In an Advanced CANDU Reactor (ACR), pressure tubes of cold-worked Zr-2.5Nb materials will be used in the reactor core to contain the fuel bundles and the light water coolant. They will be subjected to higher temperature, pressure and flux than that in a CANDU reactor. In order to ensure that these tubes will perform acceptably over their 30-year design life in such an environment, a manufacturing process has been developed to produce 6.5 mm thick ACR pressure tubes with optimized chemical composition, improved mechanical properties and in-reactor behaviour. The test and examination results show that, when compared with current in-service pressure tubes, the mechanical properties of ACR pressure tubes are significantly improved. Based on previous experience with CANDU reactor pressure tubes an assessment of the grain structure and texture indicates that the in-reactor creep deformation will be improved also. Analysis of the distribution of texture parameters from a trial batch of 26 tubes shows that the variability is reduced relative to tubes fabricated in the past. This reduction in variability together with a shift to a coarser grain structure will result in a reduction in diametral creep design limits and thus a longer economic life for the fuel channels of the advanced CANDU reactor. (author)

Fuel performance and fission product behaviour in gas cooled reactors

International Nuclear Information System (INIS)

1997-11-01

The Co-ordinated Research Programme (CRP) on Validation of Predictive Methods for Fuel and Fission Product Behaviour was organized within the frame of the International Working Group on Gas Cooled Reactors. This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs), and supports the conduct of these activities. The objectives of this CRP were to review and document the status of the experimental data base and of the predictive methods for GCR fuel performance and fission product behaviour; and to verify and validate methodologies for the prediction of fuel performance and fission product transport

Revision of Drucker-Prager cap creep modelling of pebble beds in fusion blankets

International Nuclear Information System (INIS)

Hofer, D.; Kamlah, M.; Hermsmeyer, S.

2004-01-01

A continuum model commonly used in soil mechanics analysis is compiled by use of a finite element software and has been used to simulate the thermomechanical behaviour of pebble beds. The Drucker-Prager Cap theory accounts for inelastic volume change, cap hardening, nonlinear elasticity and pressure dependent shear failure. The hardening mechanism allows for defining the hydrostatic pressure yield stress as a function of the volumetric inelastic strain. Volumetric creep is considered in order to simulate the pebble bed behaviour at high temperatures. Here, the strain hardening option has been used for the consolidation creep mechanism. The model has been calibrated using the fitting curves of the oedometric test given by Reimann et al. The fitted data has been used to calculate a pebble bed with simplified boundary conditions loaded by non-uniform volumetric heating. This calculation demonstrated that the model is capable of representing creep behaviour under volumetric heating conditions. (author)

Creep in rock salt with temperature. Testing methods and results

International Nuclear Information System (INIS)

Charpentier, J.P.; Berest, P.

1985-01-01

The growing interest shown in the delayed behaviour of rocks at elevated temperature has led the Solid Mechanics Laboratory to develop specific equipment designed for creep tests. The design and dimensioning of these units offer the possibility of investigating a wide range of materials. The article describes the test facilities used (uni-axial and tri-axial creep units) and presents the experimental results obtained on samples of Bresse salt [fr

Mechanism-based modeling of solute strengthening: application to thermal creep in Zr alloy

Energy Technology Data Exchange (ETDEWEB)

Tome, Carlos [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Wen, Wei [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Capolungo, Laurent [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-08-01

This report focuses on the development of a physics-based thermal creep model aiming to predict the behavior of Zr alloy under reactor accident condition. The current models used for this kind of simulations are mostly empirical in nature, based generally on fits to the experimental steady-state creep rates under different temperature and stress conditions, which has the following limitations. First, reactor accident conditions, such as RIA and LOCA, usually take place in short times and involve only the primary, not the steady-state creep behavior stage. Moreover, the empirical models cannot cover the conditions from normal operation to accident environments. For example, Kombaiah and Murty [1,2] recently reported a transition between the low (n~4) and high (n~9) power law creep regimes in Zr alloys depending on the applied stress. Capturing such a behavior requires an accurate description of the mechanisms involved in the process. Therefore, a mechanism-based model that accounts for the evolution with time of microstructure is more appropriate and reliable for this kind of simulation.

Estimation of creep life of thick welded joints using a simple model. Creep characteristics in thick welded joint and their improvements. 2

International Nuclear Information System (INIS)

Nakacho, Keiji; Yamazaki, Masayoshi

2001-01-01

The information of the creep behavior of the thick welded joint is very important to secure the safety of the elevated temperature vessels like the nuclear reactors. The creep behavior of the thick welded point is very complex, thence it is difficult to practice the experiment or the theoretical analysis. A simple accurate model for theoretical analysis was developed in the first study. The simple model is constructed of several one-dimensional finite elements which can analyze three-dimensional creep behavior under a assumption. The model is easy to treat, and needs only a little labor and computation time to simulate the creep curve and local strain of the thick welded joint. In this second study, the capability of the model is expanded to estimate the creep life of the thick welded joint. New model can easily estimate the time of the rupture of the thick welded joint. It is verified comparing the result with the experimental one that the model can accurately predict the creep life. The histories of the local strains to the rupture time may be observed in the simulation by using the model. The information will be useful to improve the creep characteristics of the joints. (author)

Effect of reactor irradiation on long-term strength and creep of 0Kh16N15M3B steel under plane stressed state

International Nuclear Information System (INIS)

Khristov, G.P.; Kosov, B.D.

1982-01-01

The paper deals with analysis of results of experimental studies in creep of the austenitic OKh16n15m3b steel with various size of initial-structure grain under conditions of high-intensity reactor irradiation and control tests. It is suggested to consider the material initial structure effect on intensity of minimum creep rates both under ordinary and intrareactor conditions of loading by means of the function grain size effect on the equivalent stress. It is shown that the criterial expression previously suggested by the authors is invariant to the type of stressed and structural states and relative to intensity of minimal creep rates. It is established that the creep rate of the irradiated steel may be calculated from dependence for nonirradiated steel using as an argument a certain reduced equivalent stress which is a function of the acting stress and irradiation parameter

Micromechanical Modeling of Grain Boundaries Damage in a Copper Alloy Under Creep

International Nuclear Information System (INIS)

Voese, Markus

2015-01-01

In order to include the processes on the scale of the grain structure into the description of the creep behaviour of polycrystalline materials, the damage development of a single grain boundary has been initially investigated in the present work. For this purpose, a special simulationmethod has been used, whose resolution procedure based on holomorphic functions. The mechanisms taken into account for the simulations include nucleation, growth by grain boundary diffusion, coalescence and shrinkage until complete sintering of grain boundary cavities. These studies have then been used to develop a simplified cavitation model, which describes the grain boundary damage by two state variables and the time-dependent development by a mechanism-oriented rate formulation. To include the influence of grain boundaries within continuum mechanical considerations of polycrystals, an interface model has been developed, that incorporates both damage according to the simplified cavitation model and grain boundary sliding in dependence of a phenomenological grain boundary viscosity. Furthermore a micromechanical model of a polycrystal has been developed that allows to include a material's grain structure into the simulation of the creep behaviour by means of finite element simulations. Thereby, the deformations of individual grains are expressed by a viscoplastic single crystal model and the grain boundaries are described by the proposed interface model. The grain structure is represented by a finite element model, in which the grain boundaries are modelled by cohesive elements. From the evaluation of experimental creep data, the micromechanical model of a polycrystal has been calibrated for a copper-antimony alloy at a temperature of 823 K. Thereby, the adjustment of the single crystal model has been carried out on the basis of creep rates of pure copper single crystal specimens. The experimental determination of grain boundary sliding and grain boundary porosity for coarse

In-reactor deformation and fracture of austenitic stainless steels

International Nuclear Information System (INIS)

Bloom, E.E.; Wolfer, W.G.

1978-01-01

An experimental technique for determining in-reactor fracture strain was developed and demonstrated. Differential swelling between a sample holder and a test specimen with a lower swelling rate produced uniaxial deformation. In-reactor deformations of 0.7 to 2.1% were achieved in type 304 stainless steel previously irradiated to fluences up to 8.8 x 10 26 n/m 2 without fracture. These strains are significantly higher than found in postirradiation creep-rupture tests on similar samples. From the measured strain values and published irradiation creep data and correlations, the stress levels during the irradiation were calculated. On the basis of previous postirradiation creep-rupture results, many of the samples that did not fail would be predicted to fail. Thus we conclude that the in-reactor rupture life is longer than predicted by postirradiation tests. Strain in a fractured sample was estimated to be less than 3.8%, and the in-reactor fractures were intergranular--the same fracture mode as found in postirradiation tests. Irradiation creep may relax stresses at crack tips and sliding boundaries, thus retarding the initiation and/or growth of cracks and leading to longer rupture lives in-reactor. However, the very high ductility or superplastic behavior predicted by the strain rate sensitivity of irradiation creep is not achieved because of the eventual interruption of the deformation process by grain boundary fracture

Creep damage in zircaloy-4 at LWR temperatures

International Nuclear Information System (INIS)

Keusseyan, R.L.; Hu, C.P.; Li, C.Y.

1978-08-01

The observation of creep damage in the form of grain boundary cavitation in Zircaloy-4 in the temperature range of interest to Light Water Reactor (LWR) applications is reported. The observed damage is shown to reduce the ductility of Zircaloy-4 in a tensile test at LWR temperatures

Creep in sodium

International Nuclear Information System (INIS)

Charnock, W.; Cordwell, J.E.

1978-03-01

Available information on the creep of austenitic, ferritic and Alloy-800 type steels in liquid sodium is critically reviewed. Creep properties of stainless steels can be affected by element transfer and corrosion. At reactor structural component temperatures environmental effects are likely to be less important than changes due to thermal ageing. At high clad temperatures (700 0 C) decarburisation may cause the loss of strength and ductility in unstabilised steels while cavity formation may cause embrittlement in stabilised steels. The properties of Alloy 800 are, in some experiments, found to deteriorate while in others they are enhanced. This may be a consequence of the metallurgical complexity of the material or arise from the nature of the various techniques employed. Low alloy ferritic steels tend to decarburise in sodium at temperatures greater than 500 0 C and this leads to loss of strength and an increase in ductility. High alloy ferritics are immune to this effect and appear to be able to tolerate a degree of carburisation. Although intergranular cracking may be enhanced in liquid sodium the mechanical consequences are not significant and evidence for the existence of an embrittlement effect not associated with element transfer or corrosion is weak. Stress and strain may enhance element transfer at crack tips. However in real cracks the gettering or supply action of the crack faces conditions the chemistry of the cracks in sodium and protects the crack tip from element transfer. Thus creep crack extension rates should be independent of changes in bulk coolant chemistry. (author)

Creep in jointed rock masses. State of knowledge

Energy Technology Data Exchange (ETDEWEB)

Glamheden, Rune (Golder Associates AB (Sweden)); Hoekmark, Harald (Clay Technology AB, Lund (Sweden))

2010-06-15

To describe creep behaviour in hard rock masses in a physically realistic way, elaborate models including various combinations of dash pots, spring elements and sliders would be needed. According to our knowledge, there are at present no numerical tools available that can handle such a creep model. In addition, there are no records over sufficient long time periods of tunnel convergence in crystalline rock that could be used to determine or calibrate values for the model parameters. A possible method to perform bounding estimates of creep movements around openings in a repository may be to use distinct element codes with standard built-in elasto-plastic models. By locally reducing the fracture shear strength near the underground openings a relaxation of fracture shear loads is reached. The accumulated displacements may then represent the maximum possible effects of creep that can take place in a jointed rock mass without reference to the actual time it takes to reach the displacements. Estimates based on results from analyses where all shear stresses are allowed to disappear completely will, however, be over-conservative. To be able to set up and analyse reasonably realistic numerical models with the proposed method, further assumptions regarding the creep movements and the creep region around the opening have to be made. The purpose of this report is to present support for such assumptions as found in the literature.

Creep in jointed rock masses. State of knowledge

International Nuclear Information System (INIS)

Glamheden, Rune; Hoekmark, Harald

2010-06-01

To describe creep behaviour in hard rock masses in a physically realistic way, elaborate models including various combinations of dash pots, spring elements and sliders would be needed. According to our knowledge, there are at present no numerical tools available that can handle such a creep model. In addition, there are no records over sufficient long time periods of tunnel convergence in crystalline rock that could be used to determine or calibrate values for the model parameters. A possible method to perform bounding estimates of creep movements around openings in a repository may be to use distinct element codes with standard built-in elasto-plastic models. By locally reducing the fracture shear strength near the underground openings a relaxation of fracture shear loads is reached. The accumulated displacements may then represent the maximum possible effects of creep that can take place in a jointed rock mass without reference to the actual time it takes to reach the displacements. Estimates based on results from analyses where all shear stresses are allowed to disappear completely will, however, be over-conservative. To be able to set up and analyse reasonably realistic numerical models with the proposed method, further assumptions regarding the creep movements and the creep region around the opening have to be made. The purpose of this report is to present support for such assumptions as found in the literature

The irradiation creep characteristics of graphite to high fluences

International Nuclear Information System (INIS)

Kennedy, C.R.; Cundy, M.; Kleist, G.

1988-01-01

High-temperature gas-cooled reactors (HTGR) have massive blocks of graphite with thermal and neutron-flux gradients causing high internal stresses. Thermal stresses are transient; however, stresses generated by differential growth due to neutron damage continue to increase with time. Fortunately, graphite also experiences creep under irradiation allowing relaxation of stresses to nominally safe levels. Because of complexity of irradiation creep experiments, data demonstrating this phenomenon are generally limited to fairly low fluences compared to the overall fluences expected in most reactors. Notable exceptions have been experiments at 300/degree/C and 500/degree/C run at Petten under tension and compression creep stresses to fluences greater than 4 /times/ 10 26 (E > 50 keV) neutrons/m 2 . This study complements the previous results by extending the irradiation temperature to 900/degree/C. 2 refs., 3 figs

Influenced prior loading on the creep fatigue damage accumulation of heat resistant steels

International Nuclear Information System (INIS)

Kloos, K.H.; Granacher, J.; Scholz, A.

1990-01-01

On two heat resistant power plant steels the influence of prior strain cycling on the creep rupture behaviour and the influence of prior creep loading on the strain cycling behaviour is investigated. These influences concern the number of cycles to failure and the rupture time being the reference values of the generalized damage accumulation rule and they are used for a creep fatigue analysis of the results of long term service-type strain cycling tests. (orig.) [de

Fuel performance and fission product behaviour in gas cooled reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-11-01

The Co-ordinated Research Programme (CRP) on Validation of Predictive Methods for Fuel and Fission Product Behaviour was organized within the frame of the International Working Group on Gas Cooled Reactors. This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs), and supports the conduct of these activities. The objectives of this CRP were to review and document the status of the experimental data base and of the predictive methods for GCR fuel performance and fission product behaviour; and to verify and validate methodologies for the prediction of fuel performance and fission product transport. Refs, figs, tabs.

Analysis of long-term behaviour of nuclear reactor containment

Energy Technology Data Exchange (ETDEWEB)

Hora, Z. [Czech Technical University in Prague, Faculty of Civil Engineering, Department of Mechanics, Thakurova 7, 166 29 Prague 6 (Czech Republic)]. E-mail: Zbynek.Hora@fsv.cvut.cz; Patzak, B. [Czech Technical University in Prague, Faculty of Civil Engineering, Department of Mechanics, Thakurova 7, 166 29 Prague 6 (Czech Republic)

2007-02-15

For assessment of safety and durability of a nuclear power plant (NPP), knowledge of the containment behaviour under various service and extreme conditions is crucial. To perform reliable analysis of such a large-scale structure, a sufficiently realistic but still feasible numerical model must be used, in which the relevant physical phenomena are reflected. Therefore, a constitutive model for concrete including effects of moisture and heat transfer, cement hydration, creep, shrinkage and optionally microcracking of concrete should be chosen. The present paper focuses on the simulation of the service life of NPP containment, aiming to determine the material and model parameters to enable reliable prediction of structural behaviour under various conditions. The purpose of the work is to provide a numerical model calibrated using existing measurements to predict the long-term behaviour reliably. Extensive in situ measurements are used to calibrate the model and to check the validity of the model hypotheses. Moreover, the material model parameters are systematically re-calibrated based on the continuous monitoring of the structure. The structural integrity test is reanalysed numerically to show the model capability of predicting behaviour of the structure under given loading and climate conditions.

Magnetic flux creep in HTSC and Anderson-Kim theory

International Nuclear Information System (INIS)

Lykov, A.N.

2014-01-01

The theoretical and experimental data on flux creep in high-temperature superconductors (HTSC) were analyzed in the review paper. On the one hand, the main attention is paid to the most striking experimental results which have had a significant influence on the investigations of flux creep in HTSC. On the other hand, the analysis of theoretical studies is concentrated on the works, which explain the features of flux creep on the basis of the Anderson-Kim (AK) theory modifications, and received previously unsufficient attention. However, it turned out that the modified AK theory could explain a lot of features of flux creep in HTSC: the scaling behaviour of current-voltage curves of HTSC, the finite rate of flux creep at ultra low temperatures, the logarithmic dependence of effective pinning potential as a function of transport current and its decrease with temperature. The harmonic potential field which is used in this approach makes it possible to solve accurately the both problems: viscous vortex motion and flux creep in this field. Moreover the distribution of pinning potential and the interaction of vortices with each other are taken into account in the approach. Thus, the modification of the AK theory consists, essentially, in its detailed elaboration and approaching to real situations in superconductors

Long term creep strength of silver alloyed copper

International Nuclear Information System (INIS)

Auerkari, P.; Sandlin, S.

1988-12-01

The long term creep strength of silver alloyed copper has been estimated using literature creep data for materials with less than 0.1% Ag. The available data was very limited, and it was necessary to test the differences between various data sets and extrapolation methods. Assuming constant stress level and constant or changing temperature, the creep behaviour has been assessed using mainly Larson-Miller and theta-projection approaches. The calculations indicate that the different extrapolation methods and data sources can yield strongly different life estimates. With the available incomplete data the theta projection method may give the conservative life predictions, whereas the Larson-Miller approach grossly overestimates creep life. It is recommended that supplementary data is acquired to better assess the long term creep properties of canisters in repository conditions

On cyclic yield strength in definition of limits for characterisation of fatigue and creep behaviour

Science.gov (United States)

Gorash, Yevgen; MacKenzie, Donald

2017-06-01

This study proposes cyclic yield strength as a potential characteristic of safe design for structures operating under fatigue and creep conditions. Cyclic yield strength is defined on a cyclic stress-strain curve, while monotonic yield strength is defined on a monotonic curve. Both values of strengths are identified using a two-step procedure of the experimental stress-strain curves fitting with application of Ramberg-Osgood and Chaboche material models. A typical S-N curve in stress-life approach for fatigue analysis has a distinctive minimum stress lower bound, the fatigue endurance limit. Comparison of cyclic strength and fatigue limit reveals that they are approximately equal. Thus, safe fatigue design is guaranteed in the purely elastic domain defined by the cyclic yielding. A typical long-term strength curve in time-to-failure approach for creep analysis has two inflections corresponding to the cyclic and monotonic strengths. These inflections separate three domains on the long-term strength curve, which are characterised by different creep fracture modes and creep deformation mechanisms. Therefore, safe creep design is guaranteed in the linear creep domain with brittle failure mode defined by the cyclic yielding. These assumptions are confirmed using three structural steels for normal and high-temperature applications. The advantage of using cyclic yield strength for characterisation of fatigue and creep strength is a relatively quick experimental identification. The total duration of cyclic tests for a cyclic stress-strain curve identification is much less than the typical durations of fatigue and creep rupture tests at the stress levels around the cyclic yield strength.

Numerically and experimentally analysis of creep

International Nuclear Information System (INIS)

Fontanive, J.A.

1982-11-01

The problems of creep in concrete are analyzed experimentally and numerically, comparing with classical methods and suggesting a numerical procedure for the solution of these problems. Firstly, fundamentals of viscoelasticity and its application to concrete behaviour representation are presented. Then the theories of Dischinger and Arutyunyan are studied, and a computing numerical solutions are compared in several examples. Finally, experiences on creep and relaxation are described, and its result are analyzed. Some coments on possible future developments are included. (Author) [pt

On the derivation of a creep law from isothermal bore hole convergence

International Nuclear Information System (INIS)

Prij, J.; Mengelers, J.H.J.

1981-01-01

Some analytical as well as numerical aspects relevant to the creep behaviour of cavity-like structures in salt domes are presented. Two finite element models are presented for the modelling of the bore hole configuration, both dealing with the problem of a correct choice of the amount of salts which must be taken into account. A numerical procedure is suggested to derive a material creep law from measured bore hole convergence. This procedure is applied on convergence measurement in the ASSE mine (Germany) leading to a secondary creep law (depsilon/dt)sup(c)=8.8x10 -11 sigmasup(5.5) (sigma in MPa, (depsilon/dt)sup(c) in days -1 ) which describes the transient convergence behaviour correctly. Some questions concerning the uniqueness of the derived creep law are discussed

Influence of phosphorus on the creep ductility of copper

International Nuclear Information System (INIS)

Sandström, Rolf; Wu, Rui

2013-01-01

Around 1990 it was discovered that pure copper could have extra low creep ductility in the temperature interval 180–250 °C. The material was intended for use in canisters for nuclear waste disposal. Although extra low creep ductility was not observed much below 180 °C and the temperature in the canister will never exceed 100 °C, it was feared that the creep ductility could reach low values at lower temperatures after long term exposure. If 50 ppm phosphorus was added to the copper the low creep ductility disappeared. A creep cavitation model is presented that can quantitatively describe the cavitation behaviour in uniaxial and multiaxial creep tests as well as the observed creep ductility for copper with and without phosphorus. A so-called double ledge model has been introduced that demonstrates why the nucleation rate of creep cavities is often proportional to the creep rate. The phosphorus agglomerates at the grain boundaries and limits their local deformation and thereby reduces the formation and growth of cavities. This explains why extra low creep ductility does not occur in phosphorus alloyed copper

Transient behaviour study program of research reactors fuel elements at the Hydra Pulse Reactor

International Nuclear Information System (INIS)

Khvostionov, V.E.; Egorenkov, P.M.; Malankin, P.V.

2004-01-01

Program on behavior study of research reactor Fuel Elements (FE) under transient regimes initiated by excessive reactivity insertion is being presented. Program would be realized at HYDRA pulse reactor at Russian Research Center 'Kurchatov Institute' (RRC 'K1'). HYDRA uses aqueous solution of uranyl sulfate (UO 2 SO 4 ) as a fuel. Up to 30 MJ of energy can be released inside the core during the single pulse, effective power pulse width varying from 2 to 10 ms. Reactor facility allows to investigate behaviour of FE consisting of different types of fuel composition, being developed according to Russian RERTR. First part of program is aimed at transient behaviour studying of FE MR, IRT-3M, WWR-M5 types containing meats based on dioxide uranium in aluminum matrix. Mentioned FEs use 90% and 36% enriched uranium. (author)

Micromechanical modelling of fuel viscoplastic behaviour

International Nuclear Information System (INIS)

Masson, R.; Blanc, V.; Gatt, J.M.; Julien, J.; Michel, B.; Largenton, R.

2015-01-01

To identify the effect of microstructural parameters on the viscoplastic behaviour of nuclear fuels, micromechanical (also called homogenisation) approaches are used. These approaches aim at deriving effective properties of heterogeneous material from the properties of their constituents. They stand on full-field computations of representative volume elements of microstructures as well as on mean-field semi-analytical models. For light water reactor fuels, these approaches have been applied to the modelling of the effect of two microstructural parameters: the porosity effects on the thermal creep of dioxide uranium fuels (transient conditions of irradiation) as well as the plutonium content effect on the viscoplastic behaviour (nominal conditions of irradiations) of mixed oxide fuels (MOX). (authors)

A review of the high temperature creep in oxide nuclear fuels (I)

International Nuclear Information System (INIS)

Lee, Young Woo; Na, S. H.; Lee, Y. W.; Kim, H. S.; Kim, S. H.; Joung, C. Y.

1998-06-01

Since the initial stage of fuel developmental until recently, considerable efforts have been extensively directed at studying the creep properties of uranium dioxide and its related phases largely due to the importance of their application to the reactor fuels. In this state-of-the-art report, the creep behavior and mechanisms of UO 2 and its related phases were reviewed and discussed in terms of experimental variables such as applied stress, temperature, microstructure and stoichiometry. The objective of this review is to obtain a complete understanding of the influences of these variables on the creep property and creep mechanism in these materials aiming at devising more proper methods for the improvement of the behavior. The database obtained from the results will be primarily utilized also, as the reference data for studying the creep behavior of UO 2 -based mixed oxide nuclear fuels. (author). 64 refs., 6 tabs., 25 figs

Creep buckling: an experiment, an 'exact' solution and some simple thoughts

International Nuclear Information System (INIS)

Heller, P.; Anderson, R.G.

1986-01-01

The paper presents attempts to analyse and understand a carefully conducted creep buckling experiment. The analysis was conducted using the ABAQUS Finite Element Code coupled to a number of plausible creep laws. The results show good agreement between ABAQUS runs and experimental deflections but it is difficult to reproduce the early loads. A simple model of buckling analysis for n-power creep laws is derived as an aid to understanding the development of the deflections for non-linear creep laws. In particular, the model suggests why deflections develop so rapidly and how the creep deflection development relates to the elastic behaviour. (author)

Long-term behaviour of heat-resistant steels and high-temperature materials

International Nuclear Information System (INIS)

1987-01-01

This book contains 10 lectures with the following subjects: On the effect of thermal pretreatment on the structure and creep behaviour of the alloy 800 H (V. Guttmann, J. Timm); Material properties of heat resistant ferritic and austenitic steels after cold forming (W. Bendick, H. Weber); Investigations for judging the working behaviour of components made of alloy 800 and alloy 617 under creep stress (H.J. Penkalla, F. Schubert); Creep behaviour of gas turbine materials in hot gas (K.H. Kloos et al.); Effect of small cold forming on the creep beahviour of gas turbine blades made of Nimonic 90 (K.H. Keienburg et al.); Investigations on creep fatigue alternating load strength of nickel alloys (G. Raule); Change of structure, creep fatigue behaviour and life of X20 Cr Mo V 12 1 (by G. Eggeler et al.); Investigations on thermal fatigue behaviour (K.H. Mayer et al.); Creep behaviour of similar welds of the steels 13 Cr Mo 4 4, 14 MoV 6 3, 10 Cr Mo 910 and GS-17 Cr Mo V 5 11 (K. Niel et al.); Determining the creep crack behaviour of heat resistant steels with samples of different geometry (K. Maile, R. Tscheuschner). (orig.,/MM) [de

Examination of Experimental Data for Irradiation - Creep in Nuclear Graphite

Science.gov (United States)

Mobasheran, Amir Sassan

The objective of this dissertation was to establish credibility and confidence levels of the observed behavior of nuclear graphite in neutron irradiation environment. Available experimental data associated with the OC-series irradiation -induced creep experiments performed at the Oak Ridge National Laboratory (ORNL) were examined. Pre- and postirradiation measurement data were studied considering "linear" and "nonlinear" creep models. The nonlinear creep model considers the creep coefficient to vary with neutron fluence due to the densification of graphite with neutron irradiation. Within the range of neutron fluence involved (up to 0.53 times 10^{26} neutrons/m ^2, E > 50 KeV), both models were capable of explaining about 96% and 80% of the variation of the irradiation-induced creep strain with neutron fluence at temperatures of 600^circC and 900^circC, respectively. Temperature and reactor power data were analyzed to determine the best estimates for the actual irradiation temperatures. It was determined according to thermocouple readouts that the best estimate values for the irradiation temperatures were well within +/-10 ^circC of the design temperatures of 600^circC and 900 ^circC. The dependence of the secondary creep coefficients (for both linear and nonlinear models) on irradiation temperature was determined assuming that the variation of creep coefficient with temperature, in the temperature range studied, is reasonably linear. It was concluded that the variability in estimate of the creep coefficients is definitely not the results of temperature fluctuations in the experiment. The coefficients for the constitutive equation describing the overall growth of grade H-451 graphite were also studied. It was revealed that the modulus of elasticity and the shear modulus are not affected by creep and that the electrical resistivity is slightly (less than 5%) changed by creep. However, the coefficient of thermal expansion does change with creep. The consistency of

Biaxial creep deformation of Zircaloy-4 in the high alpha phase temperature range

International Nuclear Information System (INIS)

Donaldson, A.T.; Horwood, R.A.; Healey, T.

1983-01-01

The ballooning response of Zircaloy-4 fuel tubes during a postulated loss-of-coolant accident may be calculated from a knowledge of the thermal environment of the rods and the creep deformation characteristics of the cladding. In support of such calculations biaxial creep studies have been performed on fuel tubes supplied by Westinghouse, Wolverine and Sandvik of temperatures in the alpha phase range. This paper presents the results of an investigation of their respective creep behaviour which has resulted in the formulation of equations for use in LOCA fuel ballooning codes. (author)

Sequential creep-fatigue interaction in austenitic stainless steel type 316L-SPH

International Nuclear Information System (INIS)

Tavassoli, A.A.; Mottot, M.; Petrequin, P.

1986-01-01

Influence of a prior creep or fatigue exposure on subsequent fatigue or creep properties of stainless steel type 316 L SPH has been investigated. The results obtained are used to verify the validity of time and cycle fraction rule and to obtain information on the effect of very long intermittent hold times on low cycle fatigue properties, as well as on transitory loads occurring during normal service of some structural components of LMFBR reactors. Creep and fatigue tests have been carried out at 600 0 C and under conditions yielding equal or different fatigue saturation and creep stresses. Prior creep damage levels introduced range from primary to tertiary creep, whilst those of fatigue span from 20 to 70 percent of fatigue life. In both creep-fatigue and fatigue-creep sequences in the absence of a permanent prior damage (cavitation or cracking) the subsequent resistance of 316 L-SPH to fatigue or creep is unchanged, if not improved. Thin foils prepared from the specimens confirmed these observations and showed that the dislocation substructure developed during the first mode of testing is quickly replaced by that of the second mode. Grain boundary cavitation does not occur in 316 L-SPH during creep exposures to well beyond the apparent end of secondary stage and as a result prior creep exposures up to approximately 80% of rupture life do not affect fatigue properties. Conversely, significant surface cracks were found in the prior fatigue tested specimens after above about 50% life. In the presence of such cracks the subsequent creep damage was localized at the tip of the main crack and the remaining creep life was found to be usually proportional to the effective specimen cross section. Creep and fatigue sequential damage are not necessarily additive and this type of loadings are in general less severe than the repeated creep-fatigue cycling. 17 refs.

Behaviour and damage of a superalloy prepared by hot isostatic compression

International Nuclear Information System (INIS)

Dubiez-Le-Goff, Sophie

2003-01-01

This work deals with the behavior and damage of Udimet 720 superalloy prepared by hot isostatic compression. This alloy is considered for manufacturing turbine disks of high temperature reactors (HTR). The material choice for HTR turbine disk depends on the following criteria: a good creep resistance until 700 C, a good behaviour under an helium impure atmosphere, a possible implementation under a disk of 1.5 m diameter. (author) [fr

Creep deformation and rupture behavior of CLAM steel at 823 K and 873 K

Science.gov (United States)

Zhong, Boyu; Huang, Bo; Li, Chunjing; Liu, Shaojun; Xu, Gang; Zhao, Yanyun; Huang, Qunying

2014-12-01

China Low Activation Martensitic (CLAM) steel is selected as the candidate structural material in Fusion Design Study (FDS) series fusion reactor conceptual designs. The creep property of CLAM steel has been studied in this paper. Creep tests have been carried out at 823 K and 873 K over a stress range of 150-230 MPa. The creep curves showed three creep regimes, primary creep, steady-state creep and tertiary creep. The relationship between minimum creep rate (ε˙min) and the applied stress (σ) could be described by Norton power law, and the stress exponent n was decreased with the increase of the creep temperature. The creep mechanism was analyzed with the fractographes of the rupture specimens which were examined by scanning electron microscopy (SEM). The coarsening of precipitates observed with transmission electron microscope (TEM) indicated the microstructural degradation after creep test.

Creep deformation and rupture behavior of CLAM steel at 823 K and 873 K

International Nuclear Information System (INIS)

Zhong, Boyu; Huang, Bo; Li, Chunjing; Liu, Shaojun; Xu, Gang; Zhao, Yanyun; Huang, Qunying

2014-01-01

China Low Activation Martensitic (CLAM) steel is selected as the candidate structural material in Fusion Design Study (FDS) series fusion reactor conceptual designs. The creep property of CLAM steel has been studied in this paper. Creep tests have been carried out at 823 K and 873 K over a stress range of 150–230 MPa. The creep curves showed three creep regimes, primary creep, steady-state creep and tertiary creep. The relationship between minimum creep rate (ε-dot min ) and the applied stress (σ) could be described by Norton power law, and the stress exponent n was decreased with the increase of the creep temperature. The creep mechanism was analyzed with the fractographes of the rupture specimens which were examined by scanning electron microscopy (SEM). The coarsening of precipitates observed with transmission electron microscope (TEM) indicated the microstructural degradation after creep test

A constitutive equation for creep fracture under constant, variable or cyclic positive stress

International Nuclear Information System (INIS)

Snedden, J.D.

1977-01-01

Prediction of creep fracture of metals under variable stress is one of the most difficult problems of applied mechanics. At NEL this problem is under investigation using an approach in which creep is represented by two macroscopic components: an anelastic (reversible) component and a plastic (irreversible) component. Under variable loading conditions, the anelastic component's behaviour will be most important and, if an experimental programme is logically planned, the structural processes responsible will be implicit in the resulting constitutive equation describing the material's behaviour. The present paper deals with the development and application of a constitutive equation for creep fracture of RR58 Aluminium alloy at 180 0 C under variable stress and such a constitutive equation can be extrapolated to cover long-time behaviour just as with conventional constant stress creep fracture equations. Constant stress, in fact, is one of the boundary conditions of the general constitutive equation, representing zero prior damage. The other boundary condition is that of 'cadence loading' in which the stress is completely removed and then re-applied in a cyclic fashion. (Auth.)

Proceedings of the international conference on irradiation behaviour of metallic materials for fast reactor core components

International Nuclear Information System (INIS)

Poirier, J.; Dupouy, J.M.

In this conference are presented papers dealing with swelling of metals and alloys, (and specially ferritic steels), structural evolution and stability under irradiation, modifications of mechanical properties, consequences on the behaviour of fuel elements and the optimization of materials selection, and irradiation creep [fr

Cumulative fatigue and creep-fatigue damage at 3500C on recrystallized zircaloy 4

International Nuclear Information System (INIS)

Brun, G.; Pelchat, J.; Floze, J.C.; Galimberti, M.

1985-06-01

An experimental programme undertaken by C.E.A., E.D.F. and FRAGEMA with the aim of characterizing the fatigue and creep fatigue behaviour of zircaloy-4 following annealing treatments (recrystallized, stress-delived) is in progress. The results given below concern only recrystallized material. Cyclic properties, low-cycle fatigue curves and creep behaviour laws under stresses have been established. Sequential tests of pure fatigue and creep-fatigue were performed. The cumulative life fractions at fracture depend on the sequence of leading, stress history and number of cycles of prestressing. The MINER's rule appears to be conservative with regard to a low-high loading sequence whereas it is not for the reverse high-low loading sequences. Fatigue and creep damage are not interchangeable. Pre-creep improves the fatigue resistance. Pre-fatigue improves the creep strength as long as the beneficial effect of cyclic hardening overcomes the damaging effect of surface cracking. The introduction of a tension hold time into the fatigue cycle slightly increases cyclic hardening and reduces the number of cycles to failure. For hold times of less than one hour, the sum of fatigue and creep life fractions is closed to one

Creep-fatigue deformation behaviour of OFHC-copper and CuCrZr alloy with different heat treatments and with and without neutron irradiation

International Nuclear Information System (INIS)

Singh, B.N.; Johansen, B.S.; Li, M.; Stubbins, J.F.

2005-08-01

The creep-fatigue interaction behaviour of a precipitation hardened CuCrZr alloy was investigated at 295 and 573 K. To determine the effect of irradiation a number of fatigue specimens were irradiated at 333 and 573 K to a dose level in the range of 0.2 - 0.3 dpa and were tested at room temperature and 573 K, respectively. The creep-fatigue deformation behaviour of OFHC-copper was also investigated but only in the unirradiated condition and at room temperature. The creep-fatigue interaction was simulated by applying a certain holdtime on both tension and compression sides of the cyclic loading with a frequency of 0.5 Hz. Holdtimes of up to 1000 seconds were used. Creep-fatigue experiments were carried out using strain, load and extension controlled modes of cyclic loading. In addition, a number of 'interrupted' creep-fatigue tests were performed on the prime aged CuCuZr specimens in the strain controlled mode with a strain amplitude of 0.5% and a holdtime of 10 seconds. The lifetimes in terms of the number of cycles to failure were determined at different strain and load amplitudes at each holdtime. Post-deformation microstructures was investigated using a transmission electron microscopy. The main results of these investigations are presented and their implications are briefly discussed in the present report. The central conclusion emerging from the present work is that the application of holdtime generally reduces the number of cycles to failure. The largest reduction was found to be in the case of OFHC-copper. Surprisingly, the magnitude of this reduction is found to be larger at lower levels of strain or stress amplitudes, particularly when the level of the stress amplitude is below the monotonic yield strength of the material. The reduction in the yield strength due to overaging heat treatments causes a substantial decrease in the number of cycles to failure at all holdtimes investigated. The increase in the yield strength due to neutron irradiation at 333 K

Experimental study of a macrocrack propagation in a concrete specimen subjected to creep loading

Science.gov (United States)

Rossi, P.; Boulay, C.; Tailhan, J.-L.; Martin, E.

2013-07-01

Structures managers need a better prediction of the delayed failure of concrete structures, especially for very important structures like nuclear power plant encasement. Sustained loadings at high level (above 75% of loading capacity of the structure), can lead to structure failure after some time. In this study, a series of 4-point bending tests were performed in order to characterize the creep behaviour of pre-cracked beams under high load level. The specimens were made of normal strength concrete. A power law relationship is observed between the secondary deflection creep rate and the failure time. It is also shown that when crack propagation occurs during the creep loading, the creep deflection rate increases with the creep loading level and with the crack propagation rate.

Structural instabilities of high temperature alloys and their use in advanced high temperature gas cooled reactors

International Nuclear Information System (INIS)

Schuster, H.; Ennis, P.J.; Nickel, H.; Czyrska-Filemonowicz, A.

1989-01-01

High-temperature, iron-nickel and nickel based alloys are the candidate heat exchanger materials for advanced high temperature gas-cooled reactors supplying process heat for coal gasification, where operation temperatures can reach 850-950 deg. C and service lives of more than 100,000 h are necessary. In the present paper, typical examples of structural changes which occur in two representative alloys (Alloy 800 H, Fe-32Ni-20Cr and Alloy 617, Ni-22Cr-12Co-9Mo-1Al) during high temperature exposure will be given and the effects on the creep rupture properties discussed. At service temperatures, precipitation of carbides occurs which has a significant effect on the creep behaviour, especially in the early stages of creep when the precipitate particles are very fine. During coarsening of the carbides, carbides at grain boundaries restrict grain boundary sliding which retards the development of creep damage. In the service environments, enhanced carbide precipitation may occur due to the ingress of carbon from the environment (carburization). Although the creep rate is not adversely affected, the ductility of the carburized material at low and intermediate temperatures is very low. During simulated service exposures, the formation of surface corrosion scales, the precipitation of carbides and the formation of internal oxides below the surface leads to depletion of the matrix in the alloying elements involved in the corrosion processes. In thin-walled tubes the depletion of Cr due to Cr 2 O 3 formation on the surface can lead to a loss of creep strength. An additional depletion effect resulting from environmental-metal reactions is the loss of carbon (decarburization) which may occur in specific environments. The compositions of the cooling gases which decarburize the material have been determined; they are to be avoided during reactor operation

Creep and shrinkage analysis for concrete spent fuel dry storage module

Energy Technology Data Exchange (ETDEWEB)

Zhang, D. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)], E-mail: zhangd@aecl.ca

2009-07-01

CANDU reactors are designed in Canada and are built and operated worldwide to produce electricity economically with no emission of green house gases. This paper presents creep and shrinkage analysis for a concrete spent fuel dry storage module of a CANDU nuclear power plant. Creep and shrinkage analysis was performed using a method outlined in American Concrete Institute (ACI) code, and then the creep and shrinkage strains were analyzed in a finite element model to obtain the structural behavior of the concrete module. This demonstrated that the creep and shrinkage analysis for concrete spent fuel dry storage is reasonable. AECL's spent fuel dry storage module is adequate to resist the time-dependent effects due to creep and shrinkage of concrete. (author)

Creep and shrinkage analysis for concrete spent fuel dry storage module

International Nuclear Information System (INIS)

Zhang, D.

2009-01-01

CANDU reactors are designed in Canada and are built and operated worldwide to produce electricity economically with no emission of green house gases. This paper presents creep and shrinkage analysis for a concrete spent fuel dry storage module of a CANDU nuclear power plant. Creep and shrinkage analysis was performed using a method outlined in American Concrete Institute (ACI) code, and then the creep and shrinkage strains were analyzed in a finite element model to obtain the structural behavior of the concrete module. This demonstrated that the creep and shrinkage analysis for concrete spent fuel dry storage is reasonable. AECL's spent fuel dry storage module is adequate to resist the time-dependent effects due to creep and shrinkage of concrete. (author)

The anisotropic creep behaviour of zircaloy-4 fuel cladding at 1073 K

International Nuclear Information System (INIS)

Rosinger, H.E.; Bowden, J.; Shewfelt, R.S.W.

1982-04-01

The anisotropy coefficients (F, G and H) of Hill's equation, suitably modified for creep deformation, have been determined for Zircaloy-4 fuel cladding from steady-state creep tests at an elevated temperature. Creep specimens were subjected to both uniaxial and biaxial loads (via internal pressure) at 1073 K and the strain measured concurrently in the axial and tangential directions. It has been found that Zircaloy-4 fuel cladding is almost, but not completely, isotropic at 1073 K; the values of F, G and H are 0.57, 0.48 and 0.45 respectively

Contribution of uranium diffusion on creep behaviour of uranium dicarbide

International Nuclear Information System (INIS)

Kurasawa, T.; Kikuchi, T.

1976-01-01

Compressive creep tests of uranium dicarbide (UC 2 ) have been conducted. The general equation best describing the creep rate over the temperature range 1200-1400 0 C and over the stress range 2000-15000psi is represented by the sum of two exponential terms d(epsilon)/dt=A(sigma/E)sup(0.9) exp(-39.6+- 1.0/RT) + B(sigma/E)sup(4.5) exp(-120.6+-1.7/RT), where pre-exponential factors are A(sigma/E)sup(0.9)=12.3/h at low stress region (3000 psi) and B(sigma/E)sup(4.5)=3.17x10 13 /h at high stress region (9000 psi), and the activation energy is given in kcal/mol. Each term of this experimental equation indicates that important processes occurring during the steady state creep are grain-boundary diffusion of the Coble model at low stress region and the Weertman dislocation climb model at high stress region. Both mechanisms are related to migration of uranium vacancies. (Auth.)

Behavior of X 6 CrNi 18 11 under sequential testing of creep and fatigue

Energy Technology Data Exchange (ETDEWEB)

Husslage, W [TNO, Apeldoorn (Netherlands); Breitling, H [INTERATOM, Bergisch Gladbach (Germany)

1977-07-01

The behaviour of the austenitic stainless steel X 6 CrNi 18 11 with about 0.05% C, 18% Cr and 11% Ni was investigated under combined creep and cyclic loading at 550 degrees C. Base metal specimens and specimens containing a weld were tested by: prior cyclic loading followed by creep loading to rupture; prior creep loading followed by cyclic loading to rupture; alternating periods of creep and cyclic loading to rupture. The results were evaluated using the linear cumulative fatigue and creep damage rule. The damage factor D determined on basis of the respective behaviour of base material and welds varied between 0.5 and 1.6 if specimens containing a weld defect were not taken into consideration. Weld defects, which had predominantly an influence on fatigue, lowered the damage factor D up to 0.2. Evaluation of the results on welds with the pure creep and fatigue behaviour of base material shows damage factors between 0.4 and 0.9. By the high margins between allowable creep and fatigue life and life measured with specimens, the cumulative damages of base material and welded joints are much better than the allowable values according to CCI 1592 of the ASME Boiler and Pressure Vessel Code. (author)

Crack propagation behaviour in stainless steel AISI 316L at elevated temperatures under static and cyclic loading

International Nuclear Information System (INIS)

Lange, H.

1991-01-01

Experimental investigations of crack growth under creep and creep-fatigue conditions are presented. The experiments were performed with the austenitic steel AISI 316L, that will be used in fast breeder reactors. A comparison of crack propagation behaviour at temperatures of T = 550deg C and T = 700deg C in common through-thickness cracked specimens and in plates containing surface cracks is carried out by application of several fracture mechanics parameters. The quantitative description of crack initiation times and crack velocities is persued particularly. The propagation rate of one-dimensional cracks under cyclic loading conditions at T = 550deg C is also treated with fracture mechanical methods. The influence of the hold periods on crack speed is discussed. (orig.) [de

Datalogger for the creep laboratory

International Nuclear Information System (INIS)

Sambasivan, S.I.; Karthikeyan, T.V.; Chowdhary, D.M.; Anantharaman, P.N.

1989-01-01

The creep laboratory, MDL/ICGAR is a facility to study the creep properties of materials which are of interest to the fast reactor programme. The creep test is conducted over a few days to several months and years depending on the test variables employed. In these tests the creep strain and creep rate as a function of time are studied while the load and temperature are kept constant. The datalogger automates the process of recording the strain information as a function of time and also monitors the temperature throughout the test. The system handles 126 temperature channels and 42 strain channels from 27 machines. The temperature inputs are from the thermocouples and for cold junction compensation RTD's are used. An extensometer with a linear variable differential transformer (LVDT) or Super Linear Variable Capacitor (SLVC) form the set up to measure strain. The data logger consists of a front end analog input sub-system (AISS), a 8085 based Data Acquisition System (DAS) communicating to a microcomputer with CP/M operating system. The system responds to the user through the console and outputs of a dot matrix printer. The system, running a real time executive, also allows for on line enabling or disabling of a channel, printing of data, examining the current status and value, setting and getting time etc. (author)

Contribution of the Acoustic Emission technique in the understanding and the modelling of the coupling between creep and damage in concrete

International Nuclear Information System (INIS)

Saliba, J.

2012-01-01

In order to design reliable concrete structures, prediction of long term behaviour of concrete is important. In fact, creep deformation can cause mechanical deterioration and cracking, stress redistribution, loss in prestressed members and rarely ruin the structure. The aim of this research is to have a better understanding of the interaction between creep and crack growth in concrete. An experimental investigation on the fracture properties of concrete beams submitted to creep bending tests with high levels of sustained load is reported. The influence of creep on residual capacity and fracture energy of concrete is studied. In parallel, the acoustic emission technique (AE) was used to monitor crack development. The results give wealth information on damage evolution and show a decrease in the width of the fracture process zone (FPZ) characterizing a more brittle behaviour for beams subjected to creep. The AE shows that this may be due to the development of microcracking detected under creep. Based on those experimental results, a mesoscopic numerical study was proposed by coupling a damage model based on the micro-plan theory and a viscoelastic creep model defined by several Kelvin-voigt chains. The numerical results on concrete specimens in tension and in bending confirm the development of microcracks during creep at the mortar-aggregate interface. (author)

Crack growth under combined creep and fatigue conditions in alloy 800

International Nuclear Information System (INIS)

Pfaffelhuber, M.; Roedig, M.; Schubert, F.; Nickel, H.

1989-08-01

To investigate the crack growth behaviour under combined creep-fatigue loading, CT 25 mm-specimens of X10NiCrAlTi 32 20 (Alloy 800) have been tested in experiments with cyclic loadings and hold times, with static loadings and short stress rekief interrupts, with ramp type loadings and with sequences of separate fatigue and creep crack growth periods. The test temperature of 700deg C was selected because only in this temperature range this alloy provides similar amounts of crack growth under creep and fatigue conditions due to equivalent stress levels. For the estimation of crack growth under combined loading conditions a linear accumulation of increase in crack length was proved using the crack growth laws of pure creep and fatigue crack growth. Hold time and ramp loadings lead to a higher crack growth rate compared with pure creep or pure fatigue crack growth tests. In hold time experiments the crack growth rate is higher than ramp tests of the same period time. The results of hold time tests can be fairly enough predicted by linear damage accumulation rules. (orig.) [de

Creep in ceramics

CERN Document Server

Pelleg, Joshua

2017-01-01

This textbook is one of its kind, since there are no other books on Creep in Ceramics. The book consist of two parts: A and B. In part A general knowledge of creep in ceramics is considered, while part B specifies creep in technologically important ceramics. Part B covers creep in oxide ceramics, carnides and nitrides. While covering all relevant information regarding raw materials and characterization of creep in ceramics, the book also summarizes most recent innovations and developments in this field as a result of extensive literature search.

Internat. conference about the radiation behaviour of metallic canning and structure materials for fast breeders in Ajaccio (Korsika)

International Nuclear Information System (INIS)

Anderko, K.; Ehrlich, K.

1979-01-01

The program includes 48 plenary reports as well as 22 contributions in the form of a poster view and has the following structure: - swelling of ferritic steel - structural instability under radiation - theory of swelling - experiments about the swelling of austenitic steels - mechanical properties after radiation - fuel element behaviour and material optimization - radiation creeping. Additional to the items respecting the conference titel some material problems of the fusion reactor were discussed. (orig./RW) [de

An axisymmetric method of creep analysis for primary and secondary creep

International Nuclear Information System (INIS)

Jahed, Hamid; Bidabadi, Jalal

2003-01-01

A general axisymmetric method for elastic-plastic analysis was previously proposed by Jahed and Dubey [ASME J Pressure Vessels Technol 119 (1997) 264]. In the present work the method is extended to the time domain. General rate type governing equations are derived and solved in terms of rate of change of displacement as a function of rate of change in loading. Different types of loading, such as internal and external pressure, centrifugal loading and temperature gradient, are considered. To derive specific equations and employ the proposed formulation, the problem of an inhomogeneous non-uniform rotating disc is worked out. Primary and secondary creep behaviour is predicted using the proposed method and results are compared to FEM results. The problem of creep in pressurized vessels is also solved. Several numerical examples show the effectiveness and robustness of the proposed method

Engineering C-integral estimates for generalised creep behaviour and finite element validation

International Nuclear Information System (INIS)

Kim, Yun-Jae; Kim, Jin-Su; Huh, Nam-Su; Kim, Young-Jin

2002-01-01

This paper proposes an engineering method to estimate the creep C-integral for realistic creep laws to assess defective components operating at elevated temperatures. The proposed estimation method is mainly for the steady-state C * -integral, but a suggestion is also given for estimating the transient C(t)-integral. The reference stress approach is the basis of the proposed equation, but an enhancement in terms of accuracy is made through the definition of the reference stress. The proposed estimation equations are compared with extensive elastic-creep FE results employing various creep-deformation constitutive laws for six different geometries, including two-dimensional, axi-symmetric and three-dimensional geometries. Overall good agreement between the proposed method and the FE results provides confidence in the use of the proposed method for defect assessment of components at elevated temperatures. Moreover, it is shown that for surface cracks the proposed method can be used to estimate C * at any location along the crack front

Elastic-plastic-creep analysis of shells

International Nuclear Information System (INIS)

Pai, D.H.

1979-01-01

This paper presents the recent experience of a designer/fabricator of nuclear heat transport components in the area of elastic-plastic-creep analysis of shell-like structures. A brief historical perspective is first given to highlight the evolution leading to the present industry practice. The ASME elevated temperature design criteria will be discussed followed by examples of actual computations performed to support the design/analysis and fabrication of a breeder reactor component in which a substantial amount of elastic-plastic-creep analysis was performed. Mathematical challenges encountered by the design analyst in these problems will be highlighted. Developmental needs and future trends will then be given

Shearing creep properties of cements with different irregularities on two surfaces

International Nuclear Information System (INIS)

Zhang, Qingzhao; Shen, Mingrong; Ding, Wenqi; Clark, Carl

2012-01-01