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Microhardness tests for high-energy neutron-source experiments

Description: In a development effort to extract mechanical property information from miniature specimens, standard diamond pyramid microhardness (DPH) tests have been conducted at Hanford Engineering Development Laboratory (HEDL) on specimens irradiated in RTNS-II; and techniques to extend the information available from microhardness tests have been developed at the University of California, Santa Barbara (UCSB). In tests at HEDL, radiation hardening has only been observed in relatively pure materials irradiated to neutron fluences less than 4 x 10/sup 17/ n/cm/sup 2/. In copper, specifically, a proportional increase in the DPH with neutron fluence has been observed, and this microhardness increase has been correlated with an increase in the 0.2 percent offset yield strength. At UCSB it has been found that hardness and microhardness data obtained with spherical indenters can be used to determine the true stress-true plastic strain relationship of the test material; moreover, it has been found that features of the indentation lip geometry can be used to characterize localized flow phenomena like Lueders strain in steel. Consequently, microhardness test techniques appear attractive for small specimen test applications.
Date: August 6, 1981
Creator: Panayotou, N.F. & Lucas, G.E.
Partner: UNT Libraries Government Documents Department

Damage analysis and fundamental studies for fusion reactor materials development for the period March 1, 1991--February 28, 1994. Final report

Description: The philosophy of the program at the University of California Santa Barbara has been to develop a fundamental understanding of both the basic damage processes and microstructural evolution that take place in a material during neutron irradiation and the consequent dimensional and mechanical property changes. This fundamental understanding can be used in conjunction with empirical data obtained from a variety of irradiation facilities to develop physically-based models of neutron irradiation effects in structural materials. The models in turn can be used to guide alloy development and to help extrapolate the irradiation data base to the fusion reactor regime. This philosophy is consistent with that of the national and international programs for developing structural materials for fusion reactors. During this period work has encompassed: (1) analysis of the degradation of the mechanical properties of austenitic stainless steels for the purpose of assessing the feasibility of using these steels in ITER; (2) examining helium effects on radiation damage in austenitic and ferritic stainless steels; (3) development and application of electropotential drop techniques to monitor the growth of cracks in steel specimens for a variety of specimen geometries (4) development of advanced methods of measuring fracture properties; (5) combining micromechanical modeling of fracture with finite element calculations of crack and notch-tip stress and strain fields to predict failure; (6) developing a data base on flow and fracture properties of ferritic steels. Each of these activities is described in more detail below and in greater detail in the attached publications.
Date: January 1, 1995
Creator: Odette, G.R. & Lucas, G.E.
Partner: UNT Libraries Government Documents Department

Damage analysis and fundamental studies for fusion reactor materials development

Description: The philosophy of the program at the University of California Santa Barbara has been to develop a fundamental understanding of both the basic damage processes and microstructural evolution that take place in a material during neutron irradiation and the consequent dimensional and mechanical property changes. This fundamental understanding can be used in conjunction with empirical data obtained from a variety of irradiation facilities to develop physically-based models of neutron irradiation effects in structural materials. The models in turn can be used to guide alloy development and to help extrapolate the irradiation data base (expected to be largely fission reactor based) to the fusion reactor regime. This philosophy is consistent with that of the national and international programs for developing structural materials for fusion reactors.
Date: September 1, 1991
Creator: Odette, G.R. & Lucas, G.E.
Partner: UNT Libraries Government Documents Department

Damage analysis and fundamental studies for fusion reactor materials development

Description: During this period work has encompassed: (a) development of electropotential drop techniques to monitor the growth of cracks in steel specimens for a variety of specimen geometries; (b) micromechanical modeling of fracture using finite element calculations of crack and notch-tip stress and strain fields; (3) examining helium effects on radiation damage in austenitic and ferritic stainless steels; (4) analysis of the degradation of the mechanical properties of austenitic stainless steels for the purpose of assessing the feasibility of using these steels in ITER; (5) development of an integrated approach to integrity assessment; and (6) development of advanced methods of measuring fracture properties.
Date: January 11, 1993
Creator: Odette, G.R. & Lucas, G.E.
Partner: UNT Libraries Government Documents Department

Implications of radiation-induced reductions in ductility to the design of austenitic stainless steel structures

Description: In the dose and temperature range anticipated for ITER, austenitic stainless steels exhibit significant hardening with a concomitant loss in work hardening and uniform elongation. However, significant post-necking ductility may still be retained. When uniform elongation (e{sub u}) is well defined in terms of a plastic instability criterion, e{sub u} is found to sustain reasonably high values out to about 7 dpa in the temperature range 250-350 C, beyond which it decreases to about 0.3% for 316LN. This loss of ductility has significant implications to fracture toughness and the onset of new failure modes associated with hear instability. However, the retention of a significant reduction in area at failure following irradiation indicates a less severe degradation of low-cycle fatigue life in agreement with a limited amount of data obtained to date. Suggestions are made for incorporating these results into design criteria and future testing programs.
Date: December 31, 1995
Creator: Lucas, G.E.; Billone, M.; Pawel, J.E. & Hamilton, M.L.
Partner: UNT Libraries Government Documents Department

Studies of low temperature, low flux radiation embrittlement of nuclear reactor structural materials. Final report

Description: A large matrix of simple alloys and complex commercial type steels was irradiated over a range of fluxes at 60 C up to a fast fluence of about 3 {times} 10{sup 22} n/m{sup 2}. Combined with data in the literature, these results show a negligible effect of flux on irradiation hardening in the range of 2 {times} 10{sup 13} to 5 {times} 10{sup 18} n/m{sup 2}-s. This observation lends indirect support to the proposal that the accelerated embrittlement in the High Flux Isotope Reactor surveillance steels was due to an anomalously high level of damage from gamma rays. A weak dependence of hardening on a number of elements, including copper, nickel, phosphorus, molybdenum and manganese, can be described by a simple empirical chemistry factor. Particular combinations of elements resulted in hardening differences of up to about 60% in the complex commercial type steels and up to about 100% in simple model alloys. Direct effects of microstructure appear to be minimal. Hardening varies with the square root of fluence above a threshold around 4 {times} 10{sup 20} n/m{sup 2}. The results suggest that low temperature hardening is dominated by local intracascade processes leading to the formation of small defect-solute clusters/complexes. The observed hardening corresponds to nominal maximum end-of-life transition temperature shifts in support structure steels of about 120 C.
Date: September 2, 1998
Creator: Odette, G.R. & Lucas, G.E.
Partner: UNT Libraries Government Documents Department

Electron irradiation-induced mechanical property changes in reactor pressure vessel alloys

Description: High-energy electrons were used to study tensile property changes in simple Fe-Cu and Fe-Cu-Mn alloys irradiated at 288C. A comparison was made with neutron irradiation data on the same alloys. An apparent effect of alloy chemistry was observed in which the presence of Mn affected embrittlement differently for electron and neutron irradiation. Comparison of previous experimental studies with the present experimental results indicates that electrons may be more efficient than fast neutrons at producing embrittlement.
Date: November 1, 1995
Creator: Alexander, D.E.; Rehn, L.E.; Odette, G.R. & Lucas, G.E.
Partner: UNT Libraries Government Documents Department

In-Service Design & Performance Prediction of Advanced Fusion Material Systems by Computational Modeling and Simulation

Description: This final report on "In-Service Design & Performance Prediction of Advanced Fusion Material Systems by Computational Modeling and Simulation" (DE-FG03-01ER54632) consists of a series of summaries of work that has been published, or presented at meetings, or both. It briefly describes results on the following topics: 1) A Transport and Fate Model for Helium and Helium Management; 2) Atomistic Studies of Point Defect Energetics, Dynamics and Interactions; 3) Multiscale Modeling of Fracture consisting of: 3a) A Micromechanical Model of the Master Curve (MC) Universal Fracture Toughness-Temperature Curve Relation, KJc(T - To), 3b) An Embrittlement DTo Prediction Model for the Irradiation Hardening Dominated Regime, 3c) Non-hardening Irradiation Assisted Thermal and Helium Embrittlement of 8Cr Tempered Martensitic Steels: Compilation and Analysis of Existing Data, 3d) A Model for the KJc(T) of a High Strength NFA MA957, 3e) Cracked Body Size and Geometry Effects of Measured and Effective Fracture Toughness-Model Based MC and To Evaluations of F82H and Eurofer 97, 3-f) Size and Geometry Effects on the Effective Toughness of Cracked Fusion Structures; 4) Modeling the Multiscale Mechanics of Flow Localization-Ductility Loss in Irradiation Damaged BCC Alloys; and 5) A Universal Relation Between Indentation Hardness and True Stress-Strain Constitutive Behavior. Further details can be found in the cited references or presentations that generally can be accessed on the internet, or provided upon request to the authors. Finally, it is noted that this effort was integrated with our base program in fusion materials, also funded by the DOE OFES.
Date: November 15, 2005
Creator: Odette, G. R. & Lucas, G. E.
Partner: UNT Libraries Government Documents Department

Damage analysis and fundamental studies for fusion reactor materials development. Progress report, December 1, 1986--December 31, 1990

Description: The philosophy of the program at the University of California Santa Barbara has been to develop a fundamental understanding of both the basic damage processes and microstructural evolution that take place in a material during neutron irradiation and the consequent dimensional and mechanical property changes. This fundamental understanding can be used in conjunction with empirical data obtained from a variety of irradiation facilities to develop physically-based models of neutron irradiation effects in structural materials. The models in turn can be used to guide alloy development and to help extrapolate the irradiation data base (expected to be largely fission reactor based) to the fusion reactor regime. This philosophy is consistent with that of the national and international programs for developing structural materials for fusion reactors.
Date: September 1, 1991
Creator: Odette, G. R. & Lucas, G. E.
Partner: UNT Libraries Government Documents Department

Recent progress in shear punch testing

Description: The shear punch test was developed in response to the needs of the materials development community for small-scale mechanical properties tests. Such tests will be of great importance when a fusion neutron simulation device is built, since such a device is expected to have a limited irradiation volume. The shear punch test blanks a circular disk from a fixed sheet metal specimen, specifically a TEM disk. Load-displacement data generated during the test can be related to uniaxial tensile properties such as yield and ultimate strength. Shear punch and tensile tests were performed at room temperature on a number of unirradiated aluminum, copper, vanadium, and stainless steel alloys and on several irradiated aluminum alloys. Recent results discussed here suggest that the relationship between shear punch strength and tensile strength varies with alloy class, although the relationship determined for the unirradiated condition remains valid for the irradiated aluminum alloys.
Date: September 1, 1994
Creator: Hamilton, M. L.; Toloczko, M. B. & Lucas, G. E.
Partner: UNT Libraries Government Documents Department

Damage analysis and fundamental studies for fusion reactor materials development. Technical progredd report, March 1, 1992--January 1, 1993

Description: During this period work has encompassed: (a) development of electropotential drop techniques to monitor the growth of cracks in steel specimens for a variety of specimen geometries; (b) micromechanical modeling of fracture using finite element calculations of crack and notch-tip stress and strain fields; (3) examining helium effects on radiation damage in austenitic and ferritic stainless steels; (4) analysis of the degradation of the mechanical properties of austenitic stainless steels for the purpose of assessing the feasibility of using these steels in ITER; (5) development of an integrated approach to integrity assessment; and (6) development of advanced methods of measuring fracture properties.
Date: January 11, 1993
Creator: Odette, G. R. & Lucas, G. E.
Partner: UNT Libraries Government Documents Department

Correlating radiation exposure with embrittlement: Comparative studies of electron- and neutron-irradiated pressure vessel alloys

Description: Comparative experiments using high energy (10 MeV) electrons and test reactor neutrons have been undertaken to understand the role that primary damage state has on hardening (embrittlement) induced by irradiation at 300 C. Electrons produce displacement damage primarily by low energy atomic recoils, while fast neutrons produce displacements from considerably higher energy recoils. Comparison of changes resulting from neutron irradiation, in which nascent point defect clusters can form in dense cascades, with electron irradiation, where cascade formation is minimized, can provide insight into the role that the in-cascade point defect clusters have on the mechanisms of embrittlement. Tensile property changes induced by 10 MeV electrons or test reactor neutron irradiations of unalloyed iron and an Fe-O.9 wt.% Cu-1.0 wt.% Mn alloy were examined in the damage range of 9.0 x 10{sup {minus}5} dpa to 1.5 x 10{sup {minus}2} dpa. The results show the ternary alloy experienced substantially greater embrittlement in both the electron and neutron irradiate samples relative to unalloyed iron. Despite their disparate nature of defect production similar embrittlement trends with increasing radiation damage were observed for electrons and neutrons in both the ternary and unalloyed iron.
Date: December 22, 1999
Creator: Alexander, D. E.; Rehn, L. E.; Odette, G. R.; Lucas, G. E.; Klingensmith, D. & Gragg, D.
Partner: UNT Libraries Government Documents Department

The characterization of Vicker`s microhardness indentations and pile-up profiles as a strain-hardening microprobe

Description: Microhardness measurements have long been used to examine strength properties and changes in strength properties in metals, for example, as induced by irradiation. Microhardness affords a relatively simple test that can be applied to very small volumes of material. Microhardness is nominally related to the flow stress of the material at a fixed level of plastic strain. Further, the geometry of the pile-up of material around the indentation is related to the strain-hardening behavior of a material; steeper pile-ups correspond to smaller strain-hardening rates. In this study the relationship between pile-up profiles and strain hardening is examined using both experimental and analytical methods. Vickers microhardness tests have been performed on a variety of metal alloys including low alloy, high Cr and austenitic stainless steels. The pile-up topology around the indentations has been quantified using confocal microscopy techniques. In addition, the indentation and pile-up geometry has been simulated using finite element method techniques. These results have been used to develop an improved quantification of the relationship between the pile-up geometry and the strain-hardening constitutive behavior of the test material.
Date: April 1, 1998
Creator: Santos, C. Jr.; Odette, G.R.; Lucas, G.E.; Schroeter, B.; Klinginsmith, D. & Yamamoto, T.
Partner: UNT Libraries Government Documents Department

Proceedings of the Office of Fusion Energy/DOE workshop on ceramic matrix composites for structural applications in fusion reactors

Description: A workshop to assess the potential application of ceramic matrix composites (CMCs) for structural applications in fusion reactors was held on May 21--22, 1990, at University of California, Santa Barbara. Participants included individuals familiar with materials and design requirements in fusion reactors, ceramic composite processing and properties and radiation effects. The primary focus was to list the feasibility issues that might limit the application of these materials in fusion reactors. Clear advantages for the use of CMCs are high-temperature operation, which would allow a high-efficiency Rankine cycle, and low activation. Limitations to their use are material costs, fabrication complexity and costs, lack of familiarity with these materials in design, and the lack of data on radiation stability at relevant temperatures and fluences. Fusion-relevant feasibility issues identified at this workshop include: hermetic and vacuum properties related to effects of matrix porosity and matrix microcracking; chemical compatibility with coolant, tritium, and breeder and multiplier materials, radiation effects on compatibility; radiation stability and integrity; and ability to join CMCs in the shop and at the reactor site, radiation stability and integrity of joints. A summary of ongoing CMC radiation programs is also given. It was suggested that a true feasibility assessment of CMCs for fusion structural applications could not be completed without evaluation of a material tailored'' to fusion conditions or at least to radiation stability. It was suggested that a follow-up workshop be held to design a tailored composite after the results of CMC radiation studies are available and the critical feasibility issues are addressed.
Date: November 1, 1990
Creator: Jones, R.H. (Pacific Northwest Lab., Richland, WA (USA)) & Lucas, G.E. (California Univ., Santa Barbara, CA (USA))
Partner: UNT Libraries Government Documents Department

Effects of helium pre-implantation on the microstructure and mechanical properties of irradiated 316 stainless steel

Description: Transmission electron microscopy (TEM) specimens of a First Core heat of 316 stainless steel, in both the solution annealed and 20% cold worked condition, were irradiated to 46 dpa at 420 C, to 49 dpa at 520 C, and to 34 dpa at 600 C in FFTF/MOTA. Prior to irradiation, about half of the specimens were pre-implanted with approximately 100 appm of helium, and of these, several of the solution annealed and pre-implanted specimens were aged at 800 C for 2 hr. Post-irradiation density measurements showed little difference in density between the unimplanted alloys and their helium implanted counterparts. Microstructural observations on specimens irradiated at 420 C and 520 C showed relatively minor differences in defect distributions between the unimplanted and the helium implanted materials; in all cases the defect sizes and number densities were consistent with data in the literature. Where possible, irradiation hardening of the alloys was experimentally evaluated by microhardness and shear punch; experimentally obtained values were compared to values calculated using a computer model based on barrier hardening and the microstructural data. All methods indicated relatively small effects of helium implantation, and both measured and calculated values were in agreement with the range of values reported in the literature.
Date: November 1, 1994
Creator: Toloczko, M. B.; Tedeski, G. R.; Lucas, G. E.; Odette, G. R.; Stoller, R. E. & Hamilton, M. L.
Partner: UNT Libraries Government Documents Department