bluuebunny/crossref_metadata_2025 huggingface dataset:
Effects of Electro-Kinetic Coupling on the Measurement of Hydraulic Conductivity
Authors: Yeung AT
Published: March 2009

Hydraulic conductivity is the proportionality constant relating the volume flow rate of a fluid through a porous medium to the imposed hydraulic gradient. Hence, there should not be any gradient or driving force other than a hydraulic gradient imposed on the sample during the measurement of the parameter. However, no precaution is taken to prevent the development of a streaming potential during the experiment. As a result, an electrical gradient is always generated across the sample and the measured hydraulic flow volumes are actually induced by the combined hydraulic and electrical gradients. The effects of electro-kinetic coupling are more pronounced in fine-grained soil. Moreover, the measured conductivity coefficient cannot be used in coupled flow equations to quantify the hydraulic volume flow rate contributed by the hydraulic gradient without appropriate corrections. This paper attempts to present the results of the theoretical analysis and experimental evaluation of the effects of electro-kinetic coupling in the measurement of hydraulic conductivity.
Evaluation of Attenuation Capability of a Micaceous Soil as Determined from Column Leaching Tests
Authors: Yong RN , Tan B-K , Mohamed AMO
Published: March 2009

The attenuation capability of a hydrobiotite-vermiculite (HV) soil from southern Ontario is studied as a pollutant adsorbent landfill liner. The soil has specific surface areas varying from 90–206 m2/g, low cation exchange capacities (CEC) of about 10–20 meq/100 g of dry soil, a maximum dry unit weight of about 1.83 Mg/m3, and low hydraulic conductivities of 10-9 m/sec for the compacted samples. Increase in density with a corresponding decrease in hydraulic conductivity can be obtained by mixing the HV soil with kaolinite or Champlain Sea Clay. Heavy metal retention studies of HV soil and soil admixtures indicate that heavy metals retention can exceed the CEC of the soil, implying that other processes or mechanisms besides cation exchange such as precipitation are involved. While the HV soil alone shows satisfactory values for hydraulic conductivity (k= 10-9 m/sec), pH = 7–8, and high adsorption/retention of cations, especially heavy metals, the addition of Champlain sea clay in particular enhances or improves its performance.
Creep-Fatigue Damage Analysis of Solder Joints
Authors: Ju SH , Kuskowski S , Sandor BI , Plesha ME
Published: March 2009

An anisotropic model of continuum damage mechanics has been developed to predict the creep-fatigue life of solder joints. With the help of the finite element method, the stress, strain, and damage fields of the time-dependent and temperature-dependent solder can be obtained. The main advantages of this model include: (1) It can predict the initial crack location and time and the subsequent crack growth paths; (2) The damage analysis is almost the same as in traditional viscoelastic finite element analysis; (3) It can be applied to a complex structure with any loading; (4) It provides a full-field damage investigation of the structure. This damage theory can be used for various solder joints and also can be applied to analyze the creep-fatigue problems of other ductile and temperature-dependent materials. Extensive experiments including uniaxial creep, uniaxial fatigue, tension-torsion, Moiré, and bimaterial tests were performed to validate the new model. These validations and comparisons indicate that this model can predict adequately crack growth paths and fatigue lives of solder joints.
Creep-Fatigue Interactions in Eutectic Tin-Lead Solder Alloys
Authors: Kuo C-WG , Sastry SML , Jerina KL
Published: March 2009

Due to their high homologous temperature, eutectic tin-lead solder alloys subjected to cyclic loading at room temperature experience creep-fatigue interactions. In this study, superposition of fatigue and creep deformation and damage under isothermal conditions in rapidly solidified 63Sn-37Pb with and without reflow and conventional 63Sn-37Pb solder alloy is investigated. For strain-controlled fatigue with mean strain, damage is considered in terms of stress relaxation and cyclic softening. For stress-controlled fatigue with mean stress, the fatigue life is modeled with a cycle-time fraction damage law determined from creep rupture and fatigue life data. The nature of damage was investigated by metallographic examination, cavitation measurements, and fractographic observations. The extent of individual creep and fatigue contributions to the overall life and nature of superposition of creep and fatigue is discussed.
Thermomechanical and Fatigue Behavior of High-Temperature Lead and Lead-Free Solder Joints
Authors: Pao Y-H , Badgley S , Govila R , Jih E
Published: March 2009

Thermal fatigue tests were performed on six solder alloys, i.e., 90Pb-10Sn, 97Sn-2Cu-0.8Sb-0.2Ag, 97Sn-3Cu, 95.5Sn-4Cu-0.5Ag, 95Bi-5Sb, and 63Sn-37Pb to study their stress/strain hysteresis response and corresponding thermal fatigue behavior. The first five solder alloys, with melting temperature ranging from 250 to 310°C, are regarded as high temperature solders compared to solders with lower melting point, e.g., 63Sn-37Pb (TM = 183°C). A double-beam joint specimen has been developed for this test. The specimen is made by bonding an A12O3 beam and an Al 2024-T4 beam together at their ends with solder. A method based on the measured stress relaxation data at dwell time has been developed to determine the steady-state creep parameters of the solder. Simplified viscoplastic constitutive equations employing the measured creep properties were assumed for the solder and implemented in a finite element program to evaluate their validity in modeling thermal cyclic behavior. During thermal fatigue, the evolution of the stress/strain hysteresis loop was measured, and the fracture surface was examined with SEM to identify the dominant failure mechanism. The results show that some Sn-Cu alloys have a much better thermal fatigue resistance than the high lead one and can be considered as lead-free alternatives. In addition, the eutectic solder also exhibits a surprisingly longer fatigue life than 90Pb-10Sn solder, which implies that using melting temperature as a guideline for solder selection against operation temperature may need to be modified.
A Model for Primary Creep of 63Sn-37Pb Solder
Authors: Schroeder SA , Morris WL , Mitchell MR , James MR
Published: March 2009

A micromechanical model of transient and steady-state creep as it applies to near eutectic Sn-Pb solder is presented. The material is described as a two-phase system having individual material properties, coupled by long-range internal stresses. The constitutive equations permit primary creep and the transition to steady state to be simulated by a numerical analysis in which the global mechanical response is dependent on the local stress state.

The creep equations are calibrated through a series of shear strain measurements of thinwalled tubular solder specimens. Comparisons to predictions include cyclic shear straincontrolled experiments, creep relaxation, and combination ramp-hold shear strain and stress control environments. Application to finite element analysis of surface mount interconnects is described.
Thermal Stresses in Cooled Heat-Releasing Elements of Electronic Devices
Authors: Parnas A
Published: March 2009

Numerical and analytical methods for calculating thermal stresses and deformations in a flat plate connected to a cooled wall with an intermediate layer of solder or adhesive are proposed. In the numerical method, the nonisothermal flow theory for layer material was used. In the analytical method, the layer material was elastic if shear stresses were less than yield limit and ideal-plastic if they exceeded it. The plate and the wall were elastic in both methods. Specific computations were performed for the ferrite plate soldered to the metal wall of a high-power waveguide. Results showed the length of the plastic zone in the layer could be large not only in operation (when heat in the plate was releasing), but also after cooling in a plate-to-wall connecting process. The less accurate but simple analytical method can be useful for estimations of thermal stress and cooling rate, which can provide the necessary temperature and strength of both the plate and the layer. To calculate a deflection of the plate due to the thermal gradient in its cross section, the Bubnov's-Galyorkin's approximate method was used. It was shown that the deflection is small if a layer is made of solder and could be significant if the layer is adhesive.
Usnrc Research on Reactor Pressure Vessel Steel Embrittlement
Authors: Taboada A , Mayfield M
Published: March 2009

Meutron radiation embrittlement of nuclear reactor vessels continues to be a major concern of the USNRC. Research is sponsored for improving understanding of radiation embrittlement effects to predict margins of safety and to improve regulations. Regulatory issues that are affected by neutron embrittlement are examined and related to NRC research. Factors that control radiation embrittlement are examined. A comprehensive collection of radiation embrittlement data from surveillance reports and other published reports has been compiled in a computerized data base (PR-EDB). Fracture toughness data for irradiated pressure vessel steel has provided the basis for crack initiation and crack arrest toughness curves. Work is in place to improve our ability to predict thermal annealing recovery and reembrittlement. Areas where additional data are needed are described.
Radiation Damage Rate Measurements in Nickel
Authors: Schüle W
Published: March 2009

Measurements of the electrical resistivity on almost dislocation free nickel specimens were performed during irradiation with 1.85 MeV electrons. Results obtained for irradiation temperatures varying between -80 and -40°C and for production rates varying between 2.10-11 and 5. 8 ∙ 10-10 dpas -1 are reported.

It was found that the electrical resistivity per Frenkel defect ρ, varies with the measuring temperature yielding a value of ρ = 2.33 ∙10 -4 Ωcm for TM = -60, 0°C.

Interstitials are not mobile for irradiation temperatures below Tirr = -60°C and quasi steady state defect concentrations are obtained in due time for irradiation temperatures higher than Tirr = -45°C. The results are compatible with the finding that the migration activation energy of interstitials decreases with increasing high energy particle flux i.e. from a value EIIM = 1.03 eV without irradiation to EIIM = 0.885 eV for a production rate of K = 3.46∙10 -10 dpas -1.

It was further found that di-interstitials are formed in situ when an existing interstitial encounters a dynamic crowdion. Di-interstitials are very mobile at an irradiation temperature of Tirr = -60°C initiating the formation of interstitial clusters which are enriched in undersized impurity atoms, e.g. in silicon.

In the presence of large interstitial clusters or precipitates the migration activation energy of interstitials decreases further e.g. a value of EIIM ≤ 0.82 eV is found for specimens irradiated for τirr = 2∙106 s.
The Effect of Loading Rate Upon Irradiation Embrittlement Measured by Fracture Mechanical Properties
Authors: Wallin K , Valo M , Rintamaa R , Ahlstrand R
Published: March 2009

Recent investigations have shown that the Charpy-V test is an unreliable indicator of the true irradiation embrittlement. This has increased the need for a direct measurement of the fracture toughness. In the case of surveillance testing, where the specimen size is typically small, especially two types of tests are attractive, static testing based on the J-integral and dynamic testing performed by the instrumented impact hammer. Many factors speak in favor for the dynamic test. However, it is not clear whether the irradiation embrittlement is depicted similarly by static and dynamic tests. In this work the the effect of loading rate upon irradiation embrittlement measured by fracture mechanical properties is investigated. It is shown that the shift in the fracture toughness transition temperature is smaller for dynamic loading than for static loading and a theoretical explanation for the effect is proposed.
Development and Application of an LWR Reactor Pressure Vessel-Specific Flaw Distribution
Authors: Rosinski ST , Kennedy EL , Foulds JR
Published: March 2009

Previous efforts by the U.S. Department of Energy have shown that the PWR reactor vessel integrity predictions performed through probabilistic fracture mechanics analysis for a pressurized thermal shock event are significantly sensitive to the overall flaw distribution input. It has also been shown that modern vessel in-service inspection (ISI) results can be used for development of vessel flaw distribution(s) that are more representative of U.S. vessels. This paper describes the development and application of a methodology to analyze ISI data for the purpose of flaw distribution determination. The resultant methodology considers detection reliability, flaw sizing accuracy, and flaw detection threshold in its application. Application of the methodology was then demonstrated using four recently acquired U.S. PWR vessel inspection data sets. The methodology helped provide original insight into several key inspection performance and vessel integrity prediction practice issues that will impact future vessel integrity evaluation. This paper briefly discusses the development and application of the methodology and the impact to future vessel integrity analyses.
Effects of Irradiation Temperature on Charpy and Tensile Properties of High-Copper, Low Upper-Shelf, Submerged-Arc Welds
Authors: Nanstad RK , Berggren RG
Published: March 2009

The Heavy-Section Steel Irradiation Program conducted the Second and Third Irradiation Series to investigate the effects of irradiation on the ductile fracture toughness of seven commercially fabricated low upper-shelf welds. All seven submerged-arc welds were fabricated with copper-coated wire and Linde 80 flux and have average bulk copper contents ranging from 0.21 to 0.42% with nickel levels of about 0.6%. The preirradiation upper-shelf energies range from 74 to 111 J (55–82 ft-lb). Compact specimens up to 101.6 mm (4 in.) thick, 4TC(T), were irradiated at nominally 288°C in the Bulk Shielding Reactor at Oak Ridge National Laboratory to fluences ranging from about 4 to 12 × 1018 neutrons/cm2 (>1 MeV). Charpy V-notch and tensile specimens were included in the capsules at available locations which were subject to wide variations in irradiation temperature and fluence.

This paper presents analyses of the Charpy impact and tensile test data, including adjustments for irradiation temperature and fluence normalization which make possible comparison of the irradiation sensitivity of the different welds. Analyses revealed a dependence of yield and ultimate strength on irradiation temperature of -1.1 and -0.8 MPa/°C, respectively. Similarly, the Charpy impact energy changes due to irradiation temperature were -0.5°C/°C for transition temperature shift and -0.05 J/°C for the upper-shelf energy decrease. After adjustment to an irradiation temperature of 288°C and normalization to a fluence of 8 × 1018 neutrons/cm2, percentage increases in yield strength due to irradiation ranged from about 21 to 35% while those for ultimate strength ranged from about 13 to 20%. The Charpy transition temperature shifts ranged from 59 to 123°C while the postirradiation upper-shelf energies ranged from 58 to 79 J.
Effect of Specimen Size on the Impact Properties of a Pressure Vessel Weld Material
Authors: Kumar AS , Cannon NS , Hamilton ML
Published: March 2009

A new methodology was developed for correlating the upper shelf energies (USE) of full size and subsize specimens of a pressure vessel weld material A533 Grade B Class I. The methodology worked well in predicting the USE of full size specimens based on half size and third size specimen data. The predicted values of full size USE were within ± 6% of the measured values.
Annealing Experiments on Irradiated NiCrMo Weld Metal
Authors: Leitz C , Klausnitzer EN , Hofmann G
Published: March 2009

Charpy-V impact, tensile, drop weight and fracture mechanics specimens from a NiCrMo weld metal with a relatively high copper content (0.27 %) were irradiated at about 315 °C to neutron fluences between 2 and 7 x 1019 cm-2 (E > 1 MeV). Post-irradiation annealing experiments on all the above mentioned types of specimens were conducted at an annealing temperature of 450 °C, which was selected after recovery experiments on hardness and impact test specimens only, performed at annealing temperatures in the range 350 – 550 °C and with annealing times of up to 168 h. Annealing at 550 °C, a temperature which is not applicable to a reactor pressure vessel (RPV) in situ, yielded impact properties better than the pre-irradiation values. After annealing at 450 °C, which can be applied in annealing of RPVs, residual shifts of transition temperature of 40 to 50 K remained. Another weldment of similar composition irradiated at about 285 °C reached the same post-anneal level of residual shift. Some specimens were re-irradiated after irradiation and subsequent 450 °C/60 h annealing. They reached the same total transition temperature shift after an additional fluence of the same magnitude as the fluence of the primary irradiation. After a second 450 °C annealing the residual shift again was about 40 K.
Degradation of Mechanical Properties of Stainless Steel Cladding Due to Neutron Irradiation and Thermal Aging
Authors: Haggag FM
Published: March 2009

Thermal aging of three-wire series-arc stainless steel weld overlay cladding at 288°C for 1605 h resulted in an appreciable decrease (16%) in the Charpy V-notch (CVN) upper-shelf energy (USE), but the effect on the 41-J transition temperature shift was very small (3°C). The combined effect following neutron irradiation at 288°C to a fluence of 5 x 1019 neutrons/cm2 ( >1 MeV) was a 22% reduction in the USE and a 29°C shift in the 41-J transition temperature. The effect of thermal aging on tensile properties was very small. However, the combined effect of irradiation and aging was an increase in the yield strength (6 to 34% at test temperatures from 288 to -125°C) and no apparent change in ultimate tensile strength or total elongation. Neutron irradiation reduced the initiation fracture toughness (JIc) much more than did thermal aging alone. However, irradiation slightly decreased the tearing modulus but no reduction was caused by thermal aging alone. The effects of long-term thermal exposure times (20,000 and 50,000 h) will be investigated when the specimens become available. Also, long-term thermal exposure of the three-wire cladding as well as type 308 stainless steel weld materials at 343°C is in progress.
Modeling the Influence of Irradiation Temperature and Displacement Rate on Hardening Due to Point Defect Clusters in Ferritic Steels
Authors: Stoller RE
Published: March 2009

The influence of irradiation temperature and displacement rate on point defect cluster formation has been investigated using a kinetic model. The model considers both interstitial and vacancy clustering; the former are treated as Frank loops and the latter as microvoids. Clusters can form directly in displacement cascades or by diffusive encounters between free point defects. The calculations indicate that the assumption of steady state point defect concentrations is not valid for temperatures below about 200°C. At lower temperatures, the time required to reach steady state can approach an operating reactor's lifetime. At higher temperatures, the point defect transient could influence the interpretation of irradiation experiments conducted at accelerated damage rates. The hardening due to point defect clusters was calculated using a simple dislocation barrier model. Both cluster types can give rise to significant hardening; their relative importance is a function of irradiation temperature and displacement rate. The results suggest that the apparently accelerated embrittlement reported for the pressure vessel of the High Flux Isotope Reactor is not the result of a low displacement rate.
The Modelling of Radiation Hardening and Embrittlement in Magnox Mild Steel Submerged-Arc Welds
Authors: Buswell JT , Jones RB
Published: March 2009

This paper presents a modelling analysis of the transition shift data of mild steel submerged-arc welds after surveillance irradiation in a Magnox power station at temperatures of 190°C and 200°C. The data are interpreted in terms of a matrix damage component and a component due tc irradiation-enhanced precipitation. Weld-to-weld variability is assessed in terms of a weld compositional dependence of the precipitate contribution. The variability is consistent with the presence of copper-rich precipitates but there is an additional influence of phosphorus. An increase in the irradiation temperature, or the irradiation damage rate, delays the development of the peak precipitate contribution to higher doses. Once developed, the peak precipitate contribution remains invariant during extended surveillance irradiation at about 200°C.
Irradiation Induced Precipitation in Model Alloys with Systematic Variation of Cu, Ni and P Content: A Small Angle Neutron Scattering Study
Authors: Solt G , Frisius F , Waeber WB , Tipping P
Published: March 2009

Trends in the properties of defects, reflecting the role of Ni and P in the enhanced precipitation of copper during irradiation, have been studied by small angle neutron scattering (SANS) in a set of 8 model RPV steels. The alloys in the set contain different, selected amounts of Cu, Ni and P, while the composition of all the 8 samples concerning the rest of the alloying elements, the ‘RPV steel background’, was invariant throughout the series. The irradiation was done in a test reactor at 290C, at controlled radiation conditions, for two neutron fluences of 5∙1018 and 5∙1019 n/cm2. The SANS data show an approximately quadratic relationship between the precipitated volume fraction fm of damage particles and the nickel content cNi of the sample for the medium-copper cCu = 0.16% alloys, with a nearly zero initial slope of fm; the coefficients in the equation vary substantially with the fluence. A nearly perfect correlation of fm with the increase of microhardness was found for all alloys. The size distribution, as deduced from magnetic scattering, differs systematically from that found by analysis of the nuclear scattering; this difference correlates with both cCu and cNi. For the two low-copper alloys, phosphorus is seen to enhance production of radiation damage centers, whereas for the medium-copper alloys a refinement of the precipitated phase occurs.
An APFIM Survey of Grain Boundary Segregation and Precipitation in Irradiated Pressure Vessel Steels
Authors: Miller MK , Burke MG
Published: March 2009

Atom probe field ion microscopy has revealed that excessive copper levels in pressure vessel steels result in the formation of ultrafine copper-enriched clusters and precipitates during in-service neutron irradiation. In addition to these important clusters and precipitates, this technique has revealed many other fine-scaled microstructural features in these steels. These microstructural features may be divided into two broad categories; 1) intragranular precipitates and clusters and 2) interfacial segregation. The precipitates and clusters include spherical, rod-shaped and disk-shaped molybdenum and vanadium carbides, disk-shaped molybdenum nitrides, and spherical iron oxides. The type of precipitate is dependent on the metallic additions that are made to the various types of steels used in the pressure vessel. Some of these features are present in approximately the same number density and size as the copper clusters and precipitates and will therefore make a contribution to the mechanical properties. The characterization of grain boundaries and precipitate/matrix interfaces has revealed a complex pattern of segregation of various solutes including phosphorus, nickel and manganese.
Influence of Some Alloying Elements on the Formation of Vacancy Defects in Ferritic Steels During Irradiation at 288°C
Authors: Akamatsu M , Dai G , Li X , Moser P , Van Duysen JC , Zacharie G
Published: March 2009

In order to determine if the formation of vacancy defects in pressure vessel steels during neutron irradiation is partly controlled by the chemical composition, we have studied the effects of electron irradiations on the following model alloys : Fe, Fe-0, l%Cu, Fe-0,l%Cu-0,004%C, Fe-0, l%Cu-0,010%C, Fe-0, l%Cu-0, 5%Mo, and Fe-0, l%Cu-l, 2%Mn. The irradiations have been carried out at 288°C and 150°C with fluences up to 4 1023 e-.m-2. After irradiation, some samples underwent 30 min annealing treatments at temperatures ranging up to 380°C.

The mobility of vacancies and their ability to form clusters have been studied by the positron annihilation technique. We have noticed that the clustering of single vacancies in iron is favoured by the presence of molybdenum or carbon. The observed defects are small and are made of 10 to 13 vacancies. In the same conditions, copper and manganese do not seem to have any effect on the formation of vacancy clusters. We have also observed that the recovery of the vacancy-type defects occurs at temperatures close to that of pressure vessel of PWR (288°C). These temperatures depend also on the nature of the alloying elements.
Metallurgical Considerations in the Application of DPA to Reactor Vessel Analysis
Authors: Harbison LS , Lowe AL
Published: March 2009

The displacements per atom (DPA) concept was adopted in Revision 2 to Regulatory Guide 1.99 to account for the through-wall attenuation of radiation damage in the reactor vessel beltline. Models have been developed that are capable of predicting atom displacements, but the overall effect of these models to predict the mechanical properties of the material is not exact. A review of strengthening mechanisms in reactor vessel steels demonstrates that the creation and motion of dislocation is the basis of all mechanisms. Regulatory Guide 1.99, Revision 2, in estimating radiation damage attenuation, does not consider the inherent metallurgical strengthening mechanisms in modelling reactor vessel wall attenuation. Data in the literature shows that experimental, surveillance capsule, and actual pressure vessel data are not modeled well by the use of DPA. At best, DPA is able to model certain stainless steels while not being able to model the more complex low-alloy reactor vessel steels.
Tensile and Fracture Mechanics Properties of Neutron Irradiated Manet I (Martensitic 11Cr Steel Type DIN 1.4914)
Authors: de Vries MI
Published: March 2009

Both flat tensile specimens and small compact-tension specimens of the martensitic 11Cr steel Type DIN 1.4914 have been irradiated to 5 dpa at 525 K and to 10 dpa at 625 K and 725 K. Tensile tests and fracture mechanics experiments have been performed at the irradiation temperatures. Irradiated specimens and thermal control specimens were tested at fully identical conditions. The 0.2 yield stress increased with about 400 MPa to values of about 950 MPa after irradiation at 525 K and 625 K, whereas irradiation at 725 K resulted in minor radiation hardening. The total elongation values were not less than 5-5% strain. The irradiation effect on the subcritical fatigue crack growth rates in the ΔK-domain of 15–40 MPa√m was minor (less than a factor 2). The fracture toughness parameters J0.2 and dJ/da were strongly reduced at all test temperatures.
Influence of Cold Work, Heat Treatment, and Chromium Content on the Fracture Properties of Irradiated Martensitic Steels
Authors: Huang FH
Published: March 2009

Martensitic steels undergo a change in the fracture mode from ductile microvoid coalescence to brittle cleavage as temperature decreases. Neutron exposure can raise the ductile-brittle transition temperature and produce a condition of low toughness during service. This study examines the effects of cold-work, heat treatment, and chromium content on the postirradiation fracture behavior of selected steels to provide a better understanding of fracture resistance in hopes of improving it.

Experimental materials included alloy HT9 containing 12 percent Cr, a vacuum remelt of HT9, modified HT9 containing 10 percent Cr, and 9Cr-1Mo. They were irradiated in the Experimental Breeder Reactor(EBR)-II at 400 to 550°C to fluences of 2.9 and 6 × 1022 n/cm2 (E > 0.1 MeV). Results of fracture toughness tests conducted at temperatures from 93 to 450°C with electric-potential single-specimen techniques were analyzed by means of the J-integral approach.

The initiation fracture toughness of irradiated martensitic steels increased with increasing chromium content and tempering temperature. Remelting of HT9 slightly raised the fracture toughness of the irradiated materials. Cold work did not improve the fracture resistance of irradiated ferritic steels. Initiation fracture toughness decreased slightly while the tearing modulus increased significantly as irradiation temperature increased from 400°C to 500°C. Results also showed that neutron fluence had little effect on the fracture resistance of these steels, over the present range of fluences.
Microstructural Studies in Fast Reactor Irradiated FV448 Ferritic-Martensitic Alloy Plate and Weldment
Authors: Morgan TS , Little EA , Faulkner RG
Published: March 2009

The microstructure of an off-normal 12%Cr FV448-grade ferritic-martensitic weldment with 4.7 wt.% nickel addition has been compared with that of parent plate following fast reactor irradiation at 465°C to damage levels in the range 28–46 dpa. The fully martensitic weld metal exhibits instability under irradiation and transforms to a duplex austenitic/-martensitic structure with high void swelling in the austenite. Irradiation-induced precipitation differs between weldment and parent plate. M6X and G-phase are dominant phases formed in the plate; however, large δ-ferrite grains in the heat-affected zone contain only G-phase. Differences in precipitation in FV448 observed in earlier studies can be rationalized in terms of parameters which influence radiation-induced segregation.
Mechanical Property Degradation of Fast Reactor Fuel Cladding During Thermal Transients
Authors: Shibahara I , Omori T , Sato Y , Onose S , Nomura S
Published: March 2009

Temperature-transient-to-burst tests were performed on irradiated Type 316 stainless steel cladding. Special attention was paid to control the atmosphere for specimen preparation and testing to avoid possible change in the chemical state of the fission products which remained inside the fueled specimens. Under the controlled atmosphere, the burst temperature of the irradiated fueled cladding was not degraded at lower stress level by the ‘fuel adjacency effect (FAE)’. The so-called fuel adjacency effect was suspected to be caused, at least in part, by an atmospheric effect during post-irradiation examination.

The transient burst temperatures of Type 316 stainless steels were fairly insensitive to the material conditions in a controlled testing environment, although the onset of swelling acted to decrease the failure temperature. Results were also presented for irradiated advanced austenitic stainless steel and ferritic/martensitic steel cladding.
Fatigue Crack Growth Properties of Thermal Reactor Irradiated Nimonic PE16
Authors: Douglas JM , Gravenor JG
Published: March 2009

Fatigue specimens were prepared from Nimonic PE16 fuel assembly components irradiated in a Commercial Advanced Gas-Cooled Reactor (CAGR) at temperatures in the range 631 K to 916 K and integrated fast neutron doses up to 21 x 1024 neutrons/m2. The tests were performed in air at temperatures from 623 K to 973 K at a frequency of 10 Hz and R-ratio (minimum/maximum load) of 0.2. Crack growth rates were correlated with the stress intensity factor range using the relation da/dN — CΔKm. The threshold stress intensity factor range, ΔKth) obtained in load-shedding tests was in the range 6 to 9 MN/m3/2 and was not markedly influenced by neutron dose, irradiation temperature or test temperature. Specimens having both uniform fine grained and non-uniform coarser grained microstructures gave similar Δth, so that roughness-induced crack closure was not a major factor, despite wide differences in crack path morphology. In tests at 973 K, somewhat higher values of ΔKth were obtained which were probably a consequence of oxide-induced closure.
Influence of Cold Work on the Irradiation Creep and Creep Rupture of Titanium Modified Austenitic Stainless Steel
Authors: Garner FA , Hamilton ML , Eiholzer CR , Toloczko MB , Kumar AS
Published: March 2009

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°C. The 20% cold work level is preferable to the 10% level, in terms of both in-reactor creep rupture response and initial strength.
Mechanical Properties of 1950's Vintage Type 304 Stainless Steel Weldment Components After Low Temperature Neutron Irradiation
Authors: Sindelar RL , Caskey GR , Thomas JK , Hawthorne JR , Hiser AL , Lott RA , Begley JA , Shogan RP
Published: March 2009

The reactor vessels of the nuclear production reactors at the Savannah River Site (SRS) were constructed in the 1950's from Type 304 stainless steel plates welded with Type 308 stainless steel filler using a multipass metal-inert-gas process. A mechanical properties database for irradiated material has been developed for the vessel with materials from archival primary coolant system piping irradiated at low temperatures (75 to 150°C) in the State University of New York at Buffalo reactor (UBR) and the High Flux Isotope Reactor (HFIR) to doses of 0.065 to 2.1 dpa. Fracture toughness, tensile, and Charpy-V impact properties of the weldment components (base, weld, and weld heat-affected-zone (HAZ)) have been measured at temperatures of 25°C and 125°C in the L-C and C-L orientations for materials in both the irradiated and unirradiated conditions for companion specimens. Fracture toughness and tensile properties of specimens cut from an SRS reactor vessel sidewall with doses of 0.1 and 0.5 dpa were also measured at temperatures of 25 and 125°C. The irradiated materials exhibit hardening with loss of work hardenability and a reduction in toughness relative to the unirradiated materials with a slight sensitivity to exposure. Irradiation increased the yield strength between 22% to 187% with a concomitant tensile strength increase between-9% to 29%. The irradiation-induced decrease in the elastic-plastic fracture toughness (JD at 1 mm crack extension) is between 26% to 64%; the range of JICvalues are 72.8 to 366 kJ/m2 for the irradiated materials. Similarly, Charpy V-notch results show a 38% to 59% decrease in impact absorbed energies. The C-L orientation shows significantly lower absorbed energies and fracture toughness parameters than the L-C orientation for both the base and HAZ components in both the unirradiated and irradiated conditions.
Technique to Eliminate Helium Induced Weld Cracking in Stainless Steels
Authors: Wang C-A , Grossbeck ML , Chin BA
Published: March 2009

Experiments have shown that Type 316 stainless steel is susceptible to heat-affected-zone (HAZ) cracking upon cooling when welded using the gas tungsten arc (GTA) process under lateral constraint. The cracking has been hypothesized to be caused by stress-assisted helium bubble growth and rupture at grain boundaries. This study utilized an experimental welding setup which enabled different compressive stresses to be applied to the plates during welding. Autogenous GTA welds were produced in Type 316 stainless steel doped with 256 appm helium. The application of a compressive stress, 55 MPa, during welding suppressed the previously observed catastrophic cracking. Detailed examinations conducted after welding showed a dramatic change in helium bubble morphology. Grain boundary bubble growth along directions parallel to the weld was suppressed. The results suggest that stress-modified welding techniques may be used to suppress or eliminate helium-induced cracking during joining of irradiated materials.
The Complex Action of Major Solutes on Radiation-Induced Swelling of Fe-Cr-Ni Austenitic Alloys
Authors: Garner FA , Lauritzen T , Mitchell MA
Published: March 2009

The radiation-induced swelling of simple Fe-Cr-Ni austenitic alloys is known to be very sensitive to solute additions. It is shown in this paper that three of the most common solute elements (P,Si,Mo) exert a very complex and often non-monotonic influence on swelling with increasing solute level. The complexity of this influence and its dependence on other variables appears to be the result of a closely balanced competition between two or more roles played by each solute in its interaction with both vacancies and interstitials. This competition yields a variety of different swelling behaviors in response to changes in solute or solvent composition, displacement rate, and irradiation temperature.
Microstructural and Microchemical Evolution During Fast Neutron Irradiation in Nickel and Nickel-Silicon Alloys Doped with Isotopically Controlled Boron
Authors: Muroga T
Published: March 2009

Pure nickel and Ni-2wt%Si doped with several levels of boron- 10 (10B) and natural boron(nB) were irradiated in FFTF at 679, 793 and 873 K and 40 dpa, for the purpose of examining the applicability of. boron addition to the study of helium effects. The comparison of B and nB doped specimens showed that boron addition enhances cavity nucleation and growth via its transmutation effect and suppresses the nucleation via its chemical effect. Generally the transmutation effect was dominant at higher temperatures. In Ni-Si alloys, boron's chemical effect in suppressing the cavity density was stronger than that in pure nickel. Ni3Si precipitates were observed at 679 K but not at 793 and 873 K. Radiation-induced segregation of silicon at grain boundaries was most prominent at 793 K. The influence of boron addition on Ni3Si precipitation or the segregation of silicon is not yet clear.
Void Formation and Induced-Segregation During Electron-Irradiation in Ti and Nb Modified Austenitic Steels
Authors: Ohnuki S , Takahashi H , Kinoshita H , Akasaka N , Onose S , Shibahara I
Published: March 2009

The relationship between void formation and radiationinduced segregation (RIS) has been studied in austenitic steels modified by addition of Ti and Nb after electron-irradiation to 32 dpa at 673-823 K. The void swelling was reduced with increasing Ni content and decreasing Cr content. A remarkable segregation of Ni and Cr was detected at irradiated areas, which was induced by the point defect flow along the radial direction of the electron beam. There is a strong relationship between void swelling and RIS, that is, after the matrix Ni content is reduced by 10%, the swelling increases almost linearly with the increase of segregated Ni fraction. These results suggest that RIS and void formation could progress under the same process of the interaction between solute atoms and point defects, even in such modified steels.
Effects of Applied Stress on Microstructural Evolution in 316SS Under Ion Irradiation
Authors: Katoh Y , Yutaka Y , Kohyama A
Published: March 2009

To obtain fundamental knowledge about the stress effects on microstructural behavior in irradiated materials, we developed a tensile loading apparatus which is to be mounted on the dual-ion irradiation stage of the High-fluence Irradiation Facility of the University of Tokyo and established a corresponding experimental method. In the first part of this paper, a description of the facility and the experimental technique are given.

As an initial experiment using this facility, austenitic Type 316 stainless steel in solution annealed condition was irradiated by 4 MeV nickel ions at 773 or 873K under tensile stresses between 0 and 200 MPa. The irradiation dose was 3 or 50 dpa. Faulted loop microstructure and cavity microstructure were examined with a transmission electron microscope. The effects of applied stress were apparent in both cases, which proved the validity of the facility and the experimental method.

Relative number density of the interstitial type faulted loop appeared to increase when they have such an orientation that relaxes the tensile stress. The degree of such deflection of loop orientation was similar for 50 and 100 MPa of tensile stress. The size distribution of the faulted loops was not clearly affected by external stress.

As for cavity microstructure, an increasing tensile stress increased their mean size but did not clearly influence the number density. The observed stress dependence of cavity size distribution showed that a larger stress advances both the beginning and the end of cavity nucleation. Further microstructural examination suggested a possible effect of stress-enhanced precipitation on cavity nucleation.
Effects of Cascade Damage and Helium Production Under 600 MeV Proton Irradiation of Cu and Cu-Al2O3
Authors: Singh BN , Horsewell A
Published: March 2009

Specimens of pure copper (oxygen free high conductivity grade) and a commercial dispersion strengthened Cu-Al2O3 alloy were irradiated with 600 MeV protons at 523 and 673 K to a dose level of about 0.5 dpa. The irradiation induced microstructure was investigated by transmission electron microscopy (TEM). The irradiation of copper with 600 MeV protons yields an average recoil energy of 2.3 MeV, giving rise to concurrent production of multi displacement cascades and helium (125 atomic part per million/displacement per atom) atoms.

The TEM investigation showed a complete absence of cavity formation in Cu and the Cu-Al2O3 alloy at the irradiation temperature of 523 K. At 673 K, cavities were observed in Cu but not in the Cu- A12O3 alloy. These results are compared with the results of neutron irradiations and α-particle implantations. The implications of these results are discussed particularly in terms of helium generation rate and recoil energy.
The Response of Dispersion-Strengthened Copper Alloys to High Fluence Neutron Irradiation at 415°C
Authors: Edwards DJ , Newkirk JW , Garner FA , Hamilton ML , Nadkarni A , Samal P
Published: March 2009

Various oxide-dispersion-strengthened copper alloys have been irradiated to 150 dpa at 415°C in the Fast Flux Test Facility (FFTF). The Al2O3-strengthened GlidCop™ alloys, followed closely by a HfO2-strengthened alloy, displayed the best swelling resistance, electrical conductivity, and tensile properties. The conductivity of the HfO2-strengthened alloy reached a plateau at the higher levels of irradiation, instead of exhibiting the steady decrease in conductivity observed in the other alloys. A high initial oxygen content resulted in significantly higher swelling for a series of castable oxide-dispersion-strengthened alloys, while a Cr2O3-strengthened alloy showed poor resistance to radiation.
Effects of Hot Implantation of Helium in Copper on Bubble Formation Within Grains and on Grain Boundaries
Authors: Singh BN , Eldrup M , Möslang A
Published: March 2009

Effects of helium implantation rate and temperature on the evolution of the bubble microstructure in copper have been studied by transmission electron microscopy and positron annihilation techniques. Results on the temperature and implantation rate dependencies of bubble density in the grain interiors, helium flux to and bubble density at grain boundaries are described and discussed. The bubble nucleation in the grain interiors and the helium flux to grain boundaries can be understood within the framework of the di-atomic type homogeneous nucleation. The nucleation in the grain interiors appears to be diffusion controlled in the low temperature regime and gas dissociation controlled in the high temperature regime. Although the flux of helium reaching the boundaries remains very similar, the bubble density varies enormously from boundary to boundary. These differences are suggested to arise from the differences in the structure of the grain boundaries.
Modelling the Effect of Texture and Dislocation Structure on Irradiation Creep of Zirconium Alloys
Authors: Christodoulou N , Causey AR , Woo CH , Tome CN , Klassen RJ , Holt RA
Published: March 2009

The effect of texture and dislocation structure on irradiation creep of Zircaloy-2 (irradiated at about 340 K) and Zr-2.5 wt% Nb alloys (irradiated at about 558 K) is studied by means of a self-consistent model. The model relates the creep behaviour of polycrystals to that of single crystals by taking into account the crystallographic texture, dislocation density, grain shape and the intergranular stresses generated due to the crystallographic anisotropy. Three independent creep compliances of the polycrystal obtained from creep tests on a reference material are used to derive the single crystal creep compliances. These are used to calculate the polycrystalline compliances for the remaining materials. At low irradiation temperatures the predicted polycrystalline compliances agree well with the measured values. The observed behaviour can be described by a climb-assisted glide mechanism in which the creep strain is accommodated mainly by prismatic slip with smaller contributions from basal and pyramidal slip systems. At higher irradiation temperatures, the self-consistent approach can also describe well the creep behaviour of Zr-2.5 wt% Nb samples.
The Effect of Irradiation on the Mechanical Properties of 6061-T651 Aluminum
Authors: Alexander DJ
Published: March 2009

Critical components of the Advanced Neutron Source (ANS) reactor, to be built at Oak Ridge National Laboratory (ORNL), will be fabricated from 6061-T651 aluminum alloy. This alloy has been selected for its favorable neutronic, thermal, and mechanical properties. The effect of irradiation on the tensile properties and fracture toughness has been studied to allow the lifetime of these components to be estimated. Irradiations were carried out in the High Flux Isotope Reactor at ORNL at a temperature of approximately 95°C to a fluence of approximately 1026 m-2 (thermal). Testing was conducted from room temperature to 150°C. The yield and ultimate tensile strengths were increased by irradiation, and the total elongation decreased, but the fracture toughness at 26 and 95°C was not degraded by irradiation, and decreased only slightly at 150°C.
Effects of High Thermal Neutron Fluences on Type 6061 Aluminum
Authors: Weeks JR , Czajkowski CJ , Farrell K
Published: March 2009

The control rod drive follower tubes of the High Flux Beam Reactor are constructed from precipitation-hardened 6061-T6 aluminum alloy and they operate in the high thermal neutron flux regions of the core. It is shown that large thermal neutron fluences up to -4 x 1023 n/cm2 at 333K cause large increases in tensile strength and relatively modest decreases in tensile elongation while significantly reducing the notch impact toughness at room temperature. These changes are attributed to the development of a fine distribution of precipitates of amorphous silicon of which about 8% is produced radiogenically. A proposed role of thermal-to-fast neutron flux ratio is discussed.
Mechanical Properties of Vanadium Alloys Doped with Helium Using Various Techniques
Authors: Matsui H , Tanaka M , Yamamoto M , Hasegawa A , Abe K
Published: March 2009

Helium embrittlement of binary vanadium alloys has been studied using three methods of helium doping. Helium doping using the tritium trick method yields severe embrittlement at 800°C for all the alloys studied including unalloyed vanadium. The boron-10 method resulted in slight ductility loss at 800°C. Specimens doped using cyclotron implantation showed virtually no ductility loss, but some effect of helium was observed in the fractured surface. Helium release measured during tension tests has good correlation with the degree of embrittlement. These results are discussed in terms of the different state of helium clusters/ bubbles as a result of different doping methods.