Timothy McGreevy

Generation IV Reactors Integrated Materials Technology Program Plan: Focus on Very High Temperature Reactor Materials

Since 2002, the Department of Energys (DOEs) Generation IV Nuclear Energy Systems (Gen IV) Program has addressed the research and development (R&D) necessary to support next-generation nuclear energy systems. The six most promising systems identified for next-generation nuclear energy are described within this roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor-SCWR and the Very High Temperature Reactor-VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor-GFR, the Lead-cooled Fast Reactor-LFR, and the Sodium-cooled Fast Reactor-SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides and may provide an alternative to accelerator-driven systems. At the inception of DOEs Gen IV program, it was decided to significantly pursue five of the six concepts identified in the Gen IV roadmap to determine which of them was most appropriate to meet the needs of future U.S. nuclear power generation. In particular, evaluation of the highly efficient thermal SCWR and VHTR reactors was initiated primarily for energy production, and evaluation of the three fast reactor concepts, SFR, LFR, and GFR, was begun to assess viability for both energy production and their potential contribution to closing the fuel cycle. Within the Gen IV Program itself, only the VHTR class of reactors was selected for continued development. Hence, this document will address the multiple activities under the Gen IV program that contribute to the development of the VHTR. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of the structural materials needed to ensure their safe and reliable operation. The focus of this document will be the overall range of DOEs structural materials research activities being conducted to support VHTR development. By far, the largest portion of materials R&D supporting VHTR development is that being performed directly as part of the Next-Generation Nuclear Plant (NGNP) Project. Supplementary VHTR materials R&D being performed in the DOE program, including university and international research programs and that being performed under direct contracts with the American Society for Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, will also be described. Specific areas of high-priority materials research that will be needed to deploy the NGNP and provide a basis for subsequent VHTRs are described, including the following: (1) Graphite: (a) Extensive unirradiated materials characterization and assessment of irradiation effects on properties must be performed to qualify new grades of graphite for nuclear service, including thermo-physical and mechanical properties and their changes, statistical variations from billot-to-billot and lot-to-lot, creep, and especially, irradiation creep. (b) Predictive models, as well as codification of the requirements and design methods for graphite core supports, must be developed to provide a basis for licensing. (2) Ceramics: Both fibrous and load-bearing ceramics must be qualified for environmental and radiation service as insulating materials. (3) Ceramic Composites: Carbon-carbon and SiC-SiC composites must be qualified for specialized usage in selected high-temperature components, such as core stabilizers, control rods, and insulating covers and ducting. This will require development of component-specific designs and fabrication processes, materials characterization, assessment of environmental and irradiation effects, and establishment of codes and standards for materials testing and design requirements. (4) Pressure Vessel Steels: (a) Qualification of short-term, high-temperature properties of light water reactor steels for anticipated VHTR off-normal conditions must be determined, as well as the effects of aging on tensile, creep, and toughness properties, and on thermal emissivity. (b) Large-scale fabrication process for higher temperature alloys, such as 9Cr-1MoV, including ensuring thick-section and weldment integrity must be developed, as well as improved definitions of creep-fatigue and negligible creep behavior. (5) High-Temperature Alloys: (a) Qualification and codification of materials for the intermediate heat exchanger, such as Alloys 617 or 230, for long-term very high-temperature creep, creep-fatigue, and environmental aging degradation must be done, especially in thin sections for compact designs, for both base metal and weldments. (b) Constitutive models and an improved methodology for high-temperature design must be developed.

Dynamic behavior and creep characteristics of flexible particulate bed fixtures

Abstract Lateral step and impact responses and lateral, vertical, and torsional creep behavior of slender cylindrical workpieces placed in a fluidized particulate bed fixture (FPBF or PBF) were investigated. These experiments were designed, constructed, and performed to determine workpiece rigidity under dynamic loads encountered in manufacturing operations such as machining, milling, drilling, and routing. Results indicate that compaction pressure and rod depth have significant effects on lateral dynamic rigidity; lateral rigidity of the PBF can approach that of a conventional shop vise. Fixture rigidity is reliable and repeatable under lateral dynamic loading. In addition, the PBF sufficiently resists creep behavior due to external lateral, vertical, and torsional loads. Under the loading conditions above, workpieces displacement can be maintained within acceptable tolerances even for close-tolerance machining operations encountered in tight indexing applications such as machining and surface inspection. Results also contribute to a database for applying the PBF in a wider variety of manufacturing processes.

DOE-ASME Generation IV Materials Tasks

The Department of Energy (DOE) and the American Society of Mechanical Engineers (ASME) wish to update and expand appropriate materials, construction and design codes for application in future Generation IV nuclear reactor systems that operate at elevated temperatures. The scope of interest addresses specific materials and design tasks, all of which are tied to the Generation IV Reactors Integrated Materials Technology Program Plan. Many of the tasks are directly applicable to ASME Section III Subsection NH. The tasks are summarized and discussed with respect to Generation IV needs.Copyright

Experimental study of static and dynamic rigidities of flexible particulate bed fixtures under external vertical and torque loads

Abstract Rigidity of particulate bed fixtures under vertical (extraction and downward) loads, both static and dynamic, was investigated experimentally using a slender rod to represent a workpiece. Rigidity under external static and dynamic torque loads was also studied experimentally. Results of rigidity are represented as a function of submergence depth, compaction pressure, and workpiece cross section. Dynamic loads were of two types—step and impact. Loads applied to the workpiece represented typical loads encountered in manufacturing operations, such as in machining, milling, drilling, routing, inspection, welding, and painting. Results show that compaction pressure and rod depth have signifacant effects on bed fixture performance. Fixture rigidity exhibits different trends with load types. Fixture rigidity is reliable and repeatable under static vertical loading; it is also somewhat linear. Under torque loads, rigidity is highly nonlinear. Fixture rigidity is highly nonlinear in the case of dynamic loads as well. Under the loading conditions outlined above, displacement of the workpiece can be maintained within acceptable tolerances even for close tolerance machining operations and tight indexing applications such as surface inspection. Results of rigidity are conservative because the type of workpiece used in the study represents the worst possible geometry for fixturing.

Sensitivity Analysis of Hydro-Rim Deep Drawing of Cylindrical Cups

Extensive nonlinear finite element analyses were conducted to help predict practical test conditions of intelligent hydro-rim deep forming of cylindrical cups under controlled cooled punch and heated blank temperatures, punch speed, chamber and rim pressures, and punch friction. The study focused on finding practical process conditions for maximizing the drawing ratio by variations in blank and punch temperatures, friction, rim pressure, chamber pressure, and punch speed. The study was based on an experimental cell that aimed at using real time control of the mentioned parameters to delay the necking process. The finite element material model considered the plastic behavior to be strain rate and temperature dependent. While conventional deep drawing is limited to a Limit Drawing Ratio (LDR) of about 2, the results show that a parameters listed above. Blank temperature, punch friction, rim pressure, and chamber pressure provide significant influence of various degrees on increasing the cup drawing ratio. Blank heating is very effective, but does not by itself guarantee higher LDR. The presence of punch friction coupled with chamber pressure tends to delay the necking and moves the latter up along the cup wall and away from the cup bottom corner. Rim pressure, while difficult to implement, results in significant improvement of the LDR, since it helps push the material into the die, and in doing so reduces the cup-wall tension that causes the material instability. High rim pressure, on the other hand, increases the blank thickness resulting in increased blank holder loads. Punch temperature does not play as critical a role as the blank temperature in maintaining a high LDR under the conditions investigated. The study revealed that punch speed had to be above a certain critical level for a LDR of 4. However, increased punch speed proved to cause higher variations in the thickness along cup wall. It is important to mention that the results of this study do not necessarily apply to all metals; copper material was used here. Metals with low ductility, for example would react differently, a subject of future studies.Copyright

Applicability of Simplified Methods to Alloy 617 in Excess of 649°C

The Alloy 617 draft Code case does not permit the use of simplified methods to assess ratcheting when temperatures exceed 649°C. This restriction was placed due to an apparent difficulty in distinguishing between creep and plastic deformation at various strain rates for this material. A numerical evaluation of the B-1 and B-2 Tests for a tube under constant pressure and cyclic thermal gradients (linear and nonlinear) was made. Analysis results indicate that simplified methods currently in Appendix T of ASME-NH are applicable for Alloy 617 in excess of 649°C.Copyright

Simplified Inelastic Time Independent (SITI) Method for Predicting Creep

In an effort to address inelastic creep behavior for very high temperature (VHT) applications, a unified state variable material model was used in a time dependent finite element analysis to generate isochronous curves. The resulting isochronous curves were then used in an efficient time-independent plastic analysis to predict the creep behavior of components. This simplified inelastic time-independent (SITI) method can significantly reduce the geometric and load uncertainties, and the over-conservatism in predicting inelastic strain levels. SITI is an effective and computationally efficient approach for predicting inelastic strains of components operating at high and very high temperatures such as the case in the Next Generation Nuclear Plant. This work compares the SITI inelastic strains to those obtained using fully inelastic time-dependent elastic-plastic-creep analysis, and illustrates the effectiveness of the approach in obtaining creep strain predictions without elaborate full inelastic time-dependent simulation.© 2007 ASME

The Effect of Temperature Dependent Yield Strength on Upper Bounds for Creep Ratcheting

The Bree model and the elastic core concept have been used as the foundation for the simplified inelastic design analysis methods in the ASME Code for the design of components at elevated temperature for nearly three decades. The methodology provides upper bounds for creep strain accumulation and a physical basis for ascertaining if a structure under primary and secondary loading will behave elastically, plastically, shakedown, or ratchet. Comparisons of the method with inelastic analysis results have demonstrated its conservatism in stainless steel at temperatures representative of those in LMBR applications. The upper bounds on creep accumulation are revisited for very high temperatures representative of VHTR applications, where the yield strength of the material is strongly dependent upon temperature. The effect of the variation in yield strength on the evolution of the core stress is illustrated, and is shown to extend the shakedown regions, and affects the location of the boundaries between shakedown, ratcheting, and plasticity.Copyright

Coalescence Criterion of Part-Through Wall Cracks in Steam Generator Tubes of Nuclear Power Plants

In scheduling inspection and repair of nuclear power plants, it is important to predict failure pressure of cracked steam generator tubes. Nondestructive evaluation (NDE) of cracks often reveals two neighboring cracks. If two neighboring part-through cracks interact, the tube pressure, under which the ligament between the two cracks fails, could be much different than the critical burst pressure of an individual equivalent part-through crack. The ability to accurately predict the ligament failure pressure, called “coalescence pressure,” is important. The coalescence criterion, established earlier for 100% through cracks using nonlinear finite element analyses [1–3], was extended to two part-through-wall axial collinear and offset cracks cases. The ligament failure is caused by local instability of the radial and axial ligaments. As a result of this local instability, the thickness of both radial and axial ligaments decreases abruptly at a certain tube pressure. Good correlation of finite element analysis with experiments (at Argonne National Laboratory’s Energy Technology Division) was obtained. Correlation revealed that nonlinear FEM analyses are capable of predicting the coalescence pressure accurately for part-through-wall cracks. This failure criterion and FEA work have been extended to axial cracks of varying ligament width, crack length, and cases where cracks are offset by axial or circumferential ligaments. The study revealed that rupture of the radial ligament occurs at a pressure equal to the coalescence pressure in the case of axial ligament with collinear cracks. However, rupture pressure of the radial ligament is different from coalescence pressure in the case of circumferential ligament, and it depends on the length of the ligament relative to crack dimension.Copyright