Development of Oxide Dispersion Strengthened Steels for FBR Core Application, (II)

Previously manufactured oxide dispersion strengthened (ODS) ferritic steel cladding tubes had inferior internal creep rupture strength in the circumferential hoop direction. This unexpected feature of ODS cladding tubes was substantially ascribed to the needle-like grain structure aligned with the forming direction. In this study, the grain morphology was controlled by using the martensite transformation in ODS martensitic steels to produce an equi-axial grain structure. A major improvement in the strength anisotropy was successfully achieved. The most effective yttria addition was about 1 mass% in improving the strength of the ODS martensitic steels. A simple addition of titanium was particularly effective in increasing the strength level of the ODS martensitic steels to that of ODS ferritic steels.     

Characterization of High Temperature Creep Properties in Recrystallized 12Cr-ODS Ferritic Steel Claddings

The high temperature strengthening mechanism of previously manufactured 12Cr-ODS ferritic steel claddings was clarified. In the recrystallized 12Cr-2W-0.3Ti-0.24Y₂O₃-ODS ferritic steel cladding, αY₂TiO₅ type complex oxide formation was responsible for the drastic reduction of oxide particle size and the resulting shortened distance between particles, which led to superior internal creep rupture strength at 973 K because of the high resistance to gliding dislocation. Internal creep deformation was considered to be controlled by the grain boundary sliding associated with grain morphology: the near Σ11, Σ and Σ19 coincidence boundaries with a (110) common axis.     

Creep behavior of oxide dispersion strengthened 8Cr-2WVTa and 8Cr-1W steels

Microstructures and creep behavior of two martensitic oxide dispersion strengthened (ODS) steels 8%Cr-2%W-0.2%V-0.1%Ta (J1) and 8%Cr-1%W (J2) with finely dispersed Y₂Ti₂O₇ have been investigated. Creep tests have been carried out at 670, 700 and 730 °C. Creep strength of J1 is stronger than that of any other ODS martensitic steels and the hoop strength of the ferritic ODS steel cladding. At the beginning of creep test, shrinkage was frequently observed for J1. This is one of the reasons for high creep strength of J1. The δ-ferrite, which is untransformed to austenite at hot isostatic press and hot rolling temperatures, was elongated along the rolling direction, and volume fraction of δ-ferrite in J1 is larger than J2. Although the elongated δ-ferrite affects the anisotropy of creep behavior, the extent of anisotropy in J1 is not so large as that of the ferritic ODS steel.     

The anisotropy of the high temperature deformation of Zircaloy-cladding tubes for nuclear fuel rods

The anisotropy of the high temperature deformation of Zircaloy-4 cladding tubes for nuclear fuel rods for pressurized water reactors has been investigated. The axial and tangential components of the deformation of internally pressurized tube samples during closed end creep rupture tests in air at 800°C have been measured. An axial contraction of the tube sample is observed. Using Hill's theory of plasticity the axial strain can be described by anisotropy coefficients which depend on the texture of the tube material. The anisotropy coefficients are quantitatively related to the orientation distribution of the basal poles in the radial/tangential plane of the tube sample. For the typical texture of Zircaloy cladding tubes of nuclear fuel rods for pressurized water reactors, an axial contraction has to be expected under the biaxial stress conditions applied.     

Modeling deformation and failure of fast reactor cladding during simulated accident transients

An analysis is made of burst experiments performed on neutron irradiated cladding tubes. This is done by employing a generalized Voce equation to describe the mechanical deformation of type 316 stainless steel, combined with an empirical creep crack growth law, each modified to account for the effects of irradiation matrix hardening, and irradiation induced grain boundary embrittlement, respectively.The results of this analysis indicate that for large initial hoop stress, failure occurs at relatively low temperature and is controlled by the onset of plastic instability. The increase in failure temperature of irradiated material, in low temperature region, is due to irradiation strengthening. Failure in the case of relatively small initial hoop stress occurs at high temperature where the Voce equation reduces to a power law creep formula. The ductility of irradiated material, in this high tem-temperature region, is adequately described through the use of an empirical intergranular crack growth law used in conjunction with the creep law. The effect of neutron irradiation is to reduce the activation energy for crack propagation from the value for creep to some lower value correlated to independent Dorn rupture parameter measurements. The result is a predicted reduced ductility which translates into a reduction in failure temperature at a given hoop stress value for irradiated material.     

Thermal creep modeling of HT9 steel for fast reactor applications

The ferritic martensitic steel HT9 is a primary candidate material for the fuel cladding of liquid-metal-cooled fast reactors (LMFRs) owing to its excellent stability under irradiation. Thermal creep of fuel cladding is a potential life-limiting factor in the long-life fuel design of LMFRs. Using the measured data available in the literature, such as creep strain data, stress rupture data, and steady-state creep rate data, a generalized creep correlation was developed. The new correlation is based on the Garofalo equation and the modified Monkman-Grant equation, and it shows better agreement with the experimental data than existing correlations that use either the theta projection method or the minimum commitment method, making it more appropriate for use in long-life applications.     

The effect of azimuthal temperature distribution on the ballooning and rupture behavior of Zircaloy-4 cladding tube under transient-heating conditions

For the purpose of investigating the effect of azimuthal temperature distribution on the ballooning and rupture behavior of Zircaloy-4 (Zry-4) cladding tube, laboratory-scale experiments on non-irradiated Zry-4 cladding tube specimens were performed under transient-heating conditions which simulate loss-of-coolant accident (LOCA) conditions by using an external heating method, and the data obtained were compared to those from a previous study, where an internal heating method was used. The maximum circumferential strains of the cladding tube specimens were firstly divided by the engineering hoop stress. The divided maximum circumferential strains, ks, of the previous study, which used the internal heating method, were then corrected based on the azimuthal temperature difference (ATD) in the cladding tube specimen. The ks for the external heating method which was used in this study agreed fairly well with the corrected ks obtained in the previous study which employed the internal heating method in the burst temperature range below ～1200 K. Also, the area of rupture opening tended to increase with increasing of the value which is defined as the engineering hoop stress multiplied by the maximum circumferential strain. From the results obtained in this study, it was suggested that the maximum circumferential strain and the size of rupture opening of a cladding tube under LOCA-simulated conditions can be estimated mainly by using the engineering hoop stress, the maximum circumferential strain, and ATD in the cladding tube specimen, irrespective of heating methods.     

Characterization of recrystallization of 12Cr and 15Cr ODS ferritic steels

The recrystallization behavior of 12Cr and 15Cr oxide dispersion-strengthened (ODS) ferritic steels, which are the promising candidate materials for long-life core materials of the advanced fast breeder reactors, was investigated in terms of an intermediate softening heat treatment. It was clarified that keeping recovery structure at the intermediate heat treatment is indispensable for producing recrystallized structure at the final heat treatment. Prevention of repeating recrystallization is owing to the stable {100} 〈110〉 texture formation with less stored strain energy by the cold-rolling of the recrystallized structure. The two-step softening process was proposed to suppress the recrystallization and obtain adequate hardness reduction at the intermediate heat treatment. This process is effective for producing a stable recrystallized structure at the final heat treatment of the manufacturing process of ODS ferritic steel cladding.     

Creep rupture properties of heavy section, type 304 stainless steel forging for LMFBR application

Creep rupture properties of heavy section, Type 304 stainless steel forging were investigated at 550°C. The results were compared with those on the 70 mm thick plate which was made from the same heat and has comparable grain size. The heavy section forging showed significantly higher creep rupture strength than that of the 70 mm thick plate. The reason was attributed to two different phenomena. One is the effect of thermal strain which was generated at the surface region of the heavy section forging during water cooling after solution heat treatment. Another is the effect of grain boundary carbides which were precipitated at the interior of the forging during slow cooling after solution heat treatment. It was also indicated that grain boundary carbides improve creep rupture ductility as well as rupture strength, while thermal strain results in decrease in creep rupture ductility.Effect of grain size was also investigated using the 70 mm thick plates with grain size of No. 0.7 and 2.4. The results showed that, in the range of grain size to be considered for large-sized Type 304 stainless steel forgings, grain refining improves both creep rupture strength and ductility.     

Creep rupture strength of heat affected zone for 9Cr ferritic heat resisting steels

This paper describes creep rupture characteristics of weld heat affected zone, HAZ for 9Cr ferritic steels that are promising materials for nuclear energy uses. In general, creep rupture strength in the heat affected zone of peak temperature between 900 and 1000°C is lower than that in the base metal for ferritic steels. Grain refinement and coagulation of carbides for 9Cr–1Mo steels cause decrease in creep rupture strength of the HAZ. The hardness in the simulated HAZ heated to around 1000°C decreases during creep. This seems to be related to weakening of the HAZ at 1000°C. However, substitution of W for Mo is very effective in enhancement of creep rupture strength of the HAZ due to higher stability of carbides and increase in quantity of precipitated carbides during creep rupture test.     

An approach for evaluating fuel system applied in innovative nuclear energy system

One of the important issues in the study of Innovative Nuclear Energy Systems (INES) is the integrity of the fuel system applied. An approach of evaluating fuel system integrity is discussed here based on the procedure currently used in the integrity evaluation of fast reactor fuel. The fuel failure modes controlling fuel life were reviewed and fuel integrity was analyzed and compared with the failure criteria.Metal and nitride fuels with austenitic and ferritic cladding tubes were examined in this study. For the purpose of representative irradiation behavior analyses of the fuel for INES, the correlations of the cladding characteristics were modeled based on well-known characteristics of austenitic modified 316 SS (PNC316), ferritic-martensitic steel (PNC-FMS) and oxide dispersion strengthen steel (PNC-ODS).The analytical result showed that fuel lifetime is limited by channel fracture which is a nonductile type (brittle) failure associated with a high level of irradiation-induced swelling in the case of austenitic steel cladding. In the case of ferritic steel, on the other hand, the fuel life is controlled by cladding creep rupture. The lifetime evaluated here is no more than 200 GWd/t, which is still lower than the target value 400 GWd/t burnup. Possible measures to extend metal fuel lifetime may be reducing fuel smear density and ventilating fission gas in the plenum.     

Development of Long Life FBR Fuels with Particular Emphasis on Cladding Material Improvement and Fuel Fabrication

The development of FBR fuel systems with high reliability and long in-core residence capability is required to make the fast reactor economically competitive with other electrical energy sources. PNC program of fuels and materials development has been primarily focused on mixed uranium/plutonium oxide (MOX) fuel with cold-worked 316 stainless steel for the past 20 years. Modified 316 stainless steel with excellent swelling resistance and high creep rupture strength was obtained for cladding and duct of the fast prototype reactor MONJU. Advanced austenitic alloys and high strength ferritic alloys are also being investigated for high burnup fuel assemblies of a long life core in large scale FBRs.

In MOX fuel fabrication technology, extensive progress has been achieved during driver fuel fabrication for the experimental reactor JOYO. A new MOX production facility PFPF has been completed with fully automatic and remote handling systems. This facility serves for MONJU core fuel production. The improvement of fuel fabrication technologies promotes cost reduction, safety operation and security from a physical protection standpoint.     

Diametral strain of fast reactor MOX fuel pins with austenitic stainless steel cladding irradiated to high burnup

The C3M irradiation test, which was conducted in the experimental fast reactor, “Joyo”, demonstrated that mixed oxide (MOX) fuel pins with austenitic steel cladding could attain a peak pellet burnup of about 130 GWd/t safely. The test fuel assembly consisted of 61 fuel pins, whose design specifications were similar to those of driver fuel pins of a prototype fast breeder reactor, “Monju”. The irradiated fuel pins exhibited diametral strain due to cladding void swelling and irradiation creep. The cladding irradiation creep strain were due to the pellet-cladding mechanical interaction (PCMI) as well as the internal gas pressure. From the fuel pin ceramographs and ¹³⁷Cs gamma scanning, it was found that the PCMI was associated with the pellet swelling which was enhanced by the rim structure formation or by cesium uranate formation. The PCMI due to cesium uranate, which occurred near the top of the MOX fuel column, significantly affected cladding hoop stress and thermal creep, and the latter effect tended to increase the cumulative damage fraction (CDF) of the cladding though the CDF indicated that the cladding still had some margin to failure due to the creep damage.     

Influence of prior dislocation structure on anisotropy of thermal creep of cold-worked Zr-2.5Nb tubes

Normally, creep anisotropy of hcp metals is thought to be controlled by the crystallographic texture. Here, we show that the creep anisotropy of cold-worked Zr-2.5Nb tubes is also very dependent on the anisotropic dislocation structures introduced by cold-work. The contribution of each slip system to the creep deformation of an individual grain orientation depends upon the activity of that slip system during prior cold-work. This conclusion is reached by comparing the self-consistant visco-plastic polycrystalline models with thermal creep tests performed on internally pressurized thin-wall capsules with different textures under a transverse stress of 300 MPa at 350 °C, where dislocation creep is the dominant operating mechanism. The non-uniform dislocation distributions prior to creep were derived by simulating the cold-work process of Zr-2.5Nb tubes from an Elasto-Plastic Self-Consistent (EPSC) model.     

氧势对快堆不锈钢包壳管腐蚀行为的影响

通过堆外模拟试验，研究了国产３１６－Ｔｉ不锈钢包壳管在４种氧势下对ＦＣＣＩ（燃料包壳化学相互作用）和ＦＰＬＭＥ（裂变产物液态金属脆化效应）的敏感性。结果表明包壳管腐蚀特征与氧势密切相关。当氧势低于Ｃｒ在不锈钢中的氧化阈值，包壳管内壁无明显侵蚀；当氧势超过Ｃｒ在不锈钢中的氧化阈值很多时，包壳管微观侵蚀形貌是基体氧化侵蚀与晶间侵蚀（ＩＧＡ）的混和。随氧势增加，晶间侵蚀深度减小，氧化侵蚀变严重。在与超化学计量混合氧化物燃料对应的高氧势下，同时发生基体氧化侵蚀，晶间侵蚀和剥离性侵蚀，侵蚀区呈“溃疡状”，由富Ｃｒ／贫Ｎｉ和富Ｎｉ／贫Ｃｒ层交替组成。     

Creep behaviour of modified 9Cr-1Mo ferritic steel

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.     

国产快堆包壳材料CW316(Ti)SS高温强度下降的微观机理分析

冷加工316(Ti)不锈钢CW 316(Ti)SS是我国首选的快堆包壳材料,国产材料的常规力学性能与国外数据相当,但高温蠕变和高温持久强度数据却较低.本项研究主要是通过观察、比较国产快堆包壳材料和俄罗斯快堆包壳材料在高温下微观结构的变化情况,并结合对国产材料高温持久断裂试验样品的断口形貌观察结果,分析得出:国产材料长时高温力学性能下降的主要原因是沿晶界的σ相析出.     

The effect of anisotropy on the ballooning of Zircaloy cladding

The effect of creep anisotropy on the ballooning of Zircaloy LWR fuel rod cladding tubes is investigated. A perturbation method for calculating the effect of temperature inhomogenities is developed further. The results are compared with a simple method that is not restricted to small deviations from axisymmetry. The perturbation method is shown to have only limited applicability to the Zircaloy ballooning problem. The other method which assumes that the cladding tubes retain a circular cross-section provides a more useful technique for fuel rod behaviour analysis. Studies of the bending of cladding tubes and the effect of restraint on deformation and failure are presented. Apart from cladding tube bending the effects of creep anisotropy on clad deformation and failure are not large.     

The development of validation methodology of multi-axial creep damage constitutive equations and its application to 0.5Cr0.5Mo0.25V ferritic steel at 590 °C

Research progress on the development of validation methodology for multi-axial creep damage constitutive equations and its specific application to 0.5Cr0.5Mo0.25V ferritic steel at 590 °C is presented. A set of new phenomenological multi-axial creep damage constitutive equations was proposed aiming at overcoming the deficiency of inconsistency between predicted rupture strains and observed ones. Based on these explicit consistent requirements, an improved validation methodology is proposed and applied to 0.5Cr0.5Mo0.25V ferritic steel at 590 °C. It shows that the predictions of this new set of constitutive equations are consistent with experimental observations. It also reveals a significant difference in creep curves between different sets of constitutive equations and the need for experimental data so that the coupling of damage and creep deformation can be further examined.     

The creep rupture properties of beta-phase Zircaloy-2 cladding in the region 1000–1500°C

Accurate creep parameters are essential for the prediction and assessment of fuel cladding behaviour under reactor accident conditions. To resolve uncertainties in the limited data available for beta-phase Zircaloy cladding, an extensive programme of isothermal stress rupture tests has been undertaken for Zircaloy-2 tubes in a non-oxidizing environment at temperatures from 1000 to 1500°C. The secondary creep parameters are calculated from an analysis of the strain histories obtained from each test using a novel photo-electronic technique for the continuous monitoring of tube deformation. The resulting creep equation is compared with recent rupture life data for Zircaloy tubes. The analysis also identifies an anomalous primary stage which significantly retards the early stage of deformation for low stress tests at 1200°C and above and which is found to be extremely sensitive to prior annealing. The influence of this primary component on predictions of cladding behaviour is assessed.