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EXAMINATION OF Zr AND Ti RECOMBINER LOOP SPECIMENS

Description: Cold-worked specimens of iodide zirconium, Zircaloy-2, iodide titanium, and A-55 titanium were tested in a high-pressure recombiner loop in an attempt to duplicate anomalous results obtained in a prior recombiner loop. Hydrogen analyses and metallographic examinations were made on all specimens. The titanium materials and Zircaloy-2 picked up major amounts of hydrogen in the cell section. None of the materials tested showed appreciable hydrogen absorption in the recombiner section. Complete recrystallization occurred in all cell specimens while only Zircaloy-2, of the recombiner specimens, showed any degree of recrystallization. No explanation for this behavior can be given. A survnnary of the data obtained in previous recombiner loops is compared with the results of this loop. Conclusions were based on the results of three recombiner loops. Primarlly because of the hydrogen absorption data obtained in all three recombiner loops it is recommended that the zirconium and titunium materials tested not be used in environments similar to those encountered in high pressure recombiner loops. (auth)
Date: December 19, 1958
Creator: Rittenhouse, P.L.
Partner: UNT Libraries Government Documents Department

Relationship between carburization and zero-applied-stress creep dilation in Alloy 800H and Hastelloy X. [HTGR]

Description: Typical HTGR candidate alloys can carburize when exposed to simulated service environments. The carbon concentration gradients so formed give rise to internal stresses which could cause dilation. Studies performed with Hastelloy X and Alloy 800H showed that dilations of up to almost 1% can occur at 1000/sup 0/C when carbon pickup is high. Dilation was normally observed only when the carbon increase was >1000 ..mu..g/cm/sup 2/ and ceased when the diffusing carbon reached the center of the specimen.
Date: January 1, 1981
Creator: Inouye, H. & Rittenhouse, P.L.
Partner: UNT Libraries Government Documents Department

EXAMINATION OF HIGH PRESSURE RECOMBINER LOOP SPECIMENS

Description: Speciments of iodide zirconium, Zircaloy-2, Zr-15Nb, iodide titanium, TMCA-45 titanium, A-110AT titanium, and 430 stainless steel were corroded in a highpressure recombiner loop. Analyses were performed to determine the amount of hydrogen pickup. The titanium materials and iodide zirconium showed very high hydrogen pickups, while the zirconium alloys and the 430 stainless steel absorbed smaller amounts of hydrogen Metallographic examination of the specimens showed that recrystallization occurred in all but the Ar-15Nb specimens. There seems to be little difference in the extent of recrystallization and grain growth whether the in the recombiner section at 430 to 500 deg C. Recrystalliplained or correlated in any way with the amount of f hydrogen sion that occured. Since hydrogen is known to seriously embrittle zirconium and titanium, it is recommended that crystal-bar zirconium and titanium alloys not be used as materials of construction in environ ments sinmilar to that of the High Pressure Recombiner Loop. (auth)
Date: August 14, 1958
Creator: Picklesimer, M.L. & Rittenhouse, P.L.
Partner: UNT Libraries Government Documents Department

STATUS OF THE APT MATERIALS HANDBOOK.

Description: The ''Accelerator Production of Tritium (APT) Materials Handbook'' has been developed and prepared by the APT project to provide a controlled source of extensively reviewed and quality-qualified materials data and information for use in all phases of the project, from conceptual and preliminary design through construction and operation. As originally planned, the Handbook was to provide data and information on all materials associated with all APT systems and components. This includes the accelerator and its commissioning beam stops, the Target/Blanket (T/B) system (beam window, target and blanket modules, cavity vessel, vessel internals, and shields), the tritium separation plant, the balance-of-plant (BOP), and the site and buildings. The current version of the Handbook (Revision 1) provides relatively complete coverage for T/B and tritium systems materials, and its next issue will give increased attention to data for materials of the accelerator.
Date: November 1, 2000
Creator: RITTENHOUSE, P.
Partner: UNT Libraries Government Documents Department

Long-term creep testing of 2 1/4 Cr-1 Mo steel in HTGR helium

Description: Long-term creep tests have been conducted on three heats of 2 1/4 Cr-1 Mo ferritic steel, a candidate alloy for a number of applications in gas-cooled reactors. These tests, run at temperatures from 482 to 649/sup 0/C, have reached times of 40,000 h and greater. Essentially no effect of environment, either air or simulated gas-cooled reactor helium, was seen at temperatures up to about 600/sup 0/C. At the highest test temperature the creep behavior of the steel was degraded both by oxidation in air (fall off in creep resistance and lower creep ductility) and decarburization in gas-cooled reactor helium (lower creep resistance). 5 figures.
Date: January 1, 1982
Creator: Rittenhouse, P.L. & McCoy, H.E.
Partner: UNT Libraries Government Documents Department

HTGR structural-materials efforts in the US

Description: The status of ongoing structural materials programs being conducted in the US to support development and deployment of the high-temperature gas-cooled reactor (HTGR) is described. While the total US program includes work in support of all variants of this reactor system, the emphasis of this paper is on the work aimed at support of the steam cycle/cogeneration (SC/C) version of the HTGR. Work described includes activities to develop design and performance prediction data on metals, ceramics, and graphite.
Date: July 1, 1982
Creator: Rittenhouse, P.L. & Roberts, D.I.
Partner: UNT Libraries Government Documents Department

METALLURGY OF ZIRCALOY-2. PART II. THE EFFECTS OF FABRICATION VARIABLES ON THE PREFERRED ORIENTATION AND ANISOTROPY OF STRAIN BEHAVIOR

Description: The preferred orientation and anisotropy of strain behavior of Zircaloy- 2 were studied as functions of fabrication variables. An inverse-pole-figure technique was used for the preferred orientation determinations. Evaluation of the effects of the fabrication variables on the anisotropy of strain behavior was accomplished by a contractile strainaxial strain analysis. An analysis of strain behavior in the normal direction was developed on the basis of theory of plastic flow of anisotropic metals. A simple intuitively derivable relation was found to exist between the strainstrain analysis and the preferred orientation data. Correlations of the strain-strain data with true-stress-truestrain diagrams and mechanical properties were attempted. The preferred orientation of Zircaloy-2 produced by the Oak Ridge National Laboratory-Homogeneous Reactor Project (ORNL- HRP) metallurgy fabrication schedule (ingot breakdown at 1500 to 1900 deg F, major reduction at 1800 to 1900 deg F or 1350 to 1450 deg F, a heat treatment of 30 min at 1800 at 1550 deg F followed by a water quench or rapid air cool to below 1200 deg F, a final reduction of 25 to 40% at 1000 deg F. and a 3O-min anneal at 1400 to 1425 deg F) was weak compared to that of most of the other schedules investigated. Elimination of the beta heat treatment (1800 to 1850 deg F for 30 min) between the major reduction and final reduction steps resulted in a material with a high degree of preferred orienation and with a state of pseudoisotropy in ihe rolling plane. A unique and quite high degree of preferred orientaion was developed when the ORNL-HRP metallurgy fabrication procedure was used, but the ingot axis was in the transverse rather than the rolling direction of the finished plate permitting more contractile sirain to occur in the normal direction than in either the rolling or transverse directions. ...
Date: February 1, 1961
Creator: Rittenhouse, P.L. & Picklesimer, M.L.
Partner: UNT Libraries Government Documents Department

METALLURGY OF ZIRCALOY-2. PART I. THE EFFECTS OF FABRICATION VARIABLES ON THE ANISOTROPY OF MECHANICAL PROPERTIES

Description: The anisotropy of mechanical propertles of Zircaloy-2 was studied as a function of fabricatlon variables. The variatlon in tensile and impact properties with specimen orientation was taken as the measure of the anisotropy of mechanical properties for each material. A qualitative separatlon of the effects of the fabrication variables on the resulting anisotropy of mechanical properties is made, but it is valid only in the rolling plane of the plate. A contractile strain ratio, a ratio of the nataral contractile strain in the rolling plane to that in the direction normal to the rolling plane (measured on the round tensile specimen after testing), is introduced to aid in the interpretation of the tensile data. A Zircaloy-2 fabrication schedule (consisting of, in succession, ingot breakdown at a temperature of 1800 to 1900 gas-cooled F, major reduction at a temperature of 1800 to 1900 or 1350 to 1450 gas-cooled F, a to 1000 deg F. heat treatment of 1800 to 1850 gas- cooled F for 30 min, followed by either a water-quench or a rapid aircool to below 1200 gas-cooled F, a final reduction of 25 to 40% at l000 gas-cooled F, and an anneal at 1400 to 1425 gas-cooled F for 30 min) was found to produce a much more nearly isotropic material than any of the schedules investigated. This material is anisotropic in strain behavior and tensile properties in comparison to the common cubic materials. The elimination of the intermediate to 1000 deg F. heat treatment from the fabrication schedule resulted in the production of a material with tensile properties for all directions in the plane of rolling essentially the same, but which allowed little cortractile strain to occur in the thickness direction of the plate. This indicated that a high degree of three-dimensional anisotropy existed in the material. The ...
Date: November 15, 1960
Creator: Rittenhouse, P.L. & Picklesimer, M.L.
Partner: UNT Libraries Government Documents Department

STRUCTURAL DESIGN CRITERIA FOR TARGET/BLANKET SYSTEM COMPONENT MATERIALS FOR THE ACCELERATOR PRODUCTION OF TRITIUM PROJECT

Description: The design of target/blanket system components for the Accelerator Production of Tritium (APT) plant is dependent on the development of materials properties data specified by the designer. These data are needed to verify that component designs are adequate. The adequacy of the data will be related to safety, performance, and economic considerations, and to other requirements that may be deemed necessary by customers and regulatory bodies. The data required may already be in existence, as in the open technical literature, or may need to be generated, as is often the case for the design of new systems operating under relatively unique conditions. The designers' starting point for design data needs is generally some form of design criteria used in conjunction with a specified set of loading conditions and associated performance requirements. Most criteria are aimed at verifying the structural adequacy of the component, and often take the form of national or international standards such as the ASME Boiler and Pressure Vessel Code (ASME B and PV Code) or the French Nuclear Structural Requirements (RCC-MR). Whether or not there are specific design data needs associated with the use of these design criteria will largely depend on the uniqueness of the conditions of operation of the component. A component designed in accordance with the ASME B and PV Code, where no unusual environmental conditions exist, will utilize well-documented, statistically-evaluated developed in conjunction with the Code, and will not be likely to have any design data needs. On the other hand, a component to be designed to operate under unique APT conditions, is likely to have significant design data needs. Such a component is also likely to require special design criteria for verification of its structural adequacy, specifically accounting for changes in materials properties which may occur during exposure in the service environment. ...
Date: January 1, 2001
Creator: JOHNSON, W.; RYDER, R. & RITTENHOUSE, P.
Partner: UNT Libraries Government Documents Department

High-temperature low-cycle fatigue and tensile properties of Hastelloy X and alloy 617 in air and HTGR-helium

Description: Results of strain controlled fatigue and tensile tests are presented for two nickel base solution hardened alloys which are reference structural alloys for use in several high temperature gas cooled reactor concepts. These alloys, Hastelloy X Inconel 617, were tested at temperatures ranging from room temperature to 871/sup 0/C in air and impure helium. Materials were tested in the solution annealed as well as in the pre-aged condition where aging consisted of isothermal exposure at one of several temperatures for periods of up to 20,000 h. Comparisons are also given between the strain controlled fatigue lives of these alloys and several other commonly used alloys all tested at 538/sup 0/C.
Date: January 1, 1981
Creator: Strizak, J.P.; Brinkman, C.R. & Rittenhouse, P.L.
Partner: UNT Libraries Government Documents Department

Weldability evaluations and weldment properties of Hastelloy X

Description: Studies of weldability and weldment properties were conducted on commerical heats of Hastelloy X. Weldment preparation was done using several combinations of welding techniques and filler metals. Evaluation methods employed included hot cracking susceptibility and tensile and creep properties measured both before and after aging at 593 to 871/sup 0/C for up to 10,000 h.
Date: January 1, 1981
Creator: King, J.F.; McCoy, H.E. & Rittenhouse, P.L.
Partner: UNT Libraries Government Documents Department

Suggested Fabrication Procedures for Zircaloy-2 Mill Products in Ingot Quantities

Description: Suggested fabrication procedures for Zircaloy-2 sheet, plate, rod, and bar are presented. The procedures are based on the physical and mechanical metallurgy of Zircaloy-2 and are designed to produce material with a minimum amount of preferred orientation, anisotropy of mechanical properties, and intermetallic stringers. The recommended procedures cover ingot soaking, fabrication, heat treatment, finish, workmanship, identification, and inspection. A brief discussion of the physical and mechanical metallurgy of Zircaloy-2 is presented. (auth)
Date: January 30, 1961
Creator: Picklesimer, M. L. & Rittenhouse, P. L.
Partner: UNT Libraries Government Documents Department

High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1982

Description: During 1982 the High-Temperature Gas-Cooled Reactor (HTGR) Technology Program at Oak Ridge National Laboratory (ORNL) continued to develop experimental data required for the design and licensing of cogeneration HTGRs. The program involves fuels and materials development (including metals, graphite, ceramic, and concrete materials), HTGR chemistry studies, structural component development and testing, reactor physics and shielding studies, performance testing of the reactor core support structure, and HTGR application and evaluation studies.
Date: June 1, 1983
Creator: Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E. & Sanders, J.P.
Partner: UNT Libraries Government Documents Department

Gas-cooled reactor programs. High-temperature gas-cooled reactor technology development program. Annual progress report, December 31, 1983

Description: ORNL continues to make significant contributions to the national program. In the HTR fuels area, we are providing detailed statistical information on the fission product retention performance of irradiated fuel. Our studies are also providing basic data on the mechanical, physical, and chemical behavior of HTR materials, including metals, ceramics, graphite, and concrete. The ORNL has an important role in the development of improved HTR graphites and in the specification of criteria that need to be met by commercial products. We are also developing improved reactor physics design methods. Our work in component development and testing centers in the Component Flow Test Loop (CFTL), which is being used to evaluate the performance of the HTR core support structure. Other work includes experimental evaluation of the shielding effectiveness of the lower portions of an HTR core. This evaluation is being performed at the ORNL Tower Shielding Facility. Researchers at ORNL are developing welding techniques for attaching steam generator tubing to the tubesheets and are testing ceramic pads on which the core posts rest. They are also performing extensive testing of aggregate materials obtained from potential HTR site areas for possible use in prestressed concrete reactor vessels. During the past year we continued to serve as a peer reviewer of small modular reactor designs being developed by GA and GE with balance-of-plant layouts being developed by Bechtel Group, Inc. We have also evaluated the national need for developing HTRs with emphasis on the longer term applications of the HTRs to fossil conversion processes.
Date: June 1, 1984
Creator: Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E. & Sanders, J.P.
Partner: UNT Libraries Government Documents Department

High-Temperature Gas-Cooled Reactor Technology Development Program: Annual progress report for period ending December 31, 1987

Description: The High-Temperature Gas-Cooled Reactor (HTGR) Program being carried out under the US Department of Energy (DOE) continues to emphasize the development of modular high-temperature gas-cooled reactors (MHTGRs) possessing a high degree of inherent safety. The emphasis at this time is to develop the preliminary design of the reference MHTGR and to develop the associated technology base and licensing infrastructure in support of future reactor deployment. A longer-term objective is to realize the full high-temperature potential of HTGRs in gas turbine and high-temperature, process-heat applications. This document summarizes the activities of the HTGR Technology Development Program for the period ending December 31, 1987.
Date: March 1, 1989
Creator: Jones, J.E.,Jr.; Kasten, P.R.; Rittenhouse, P.L. & Sanders, J.P.
Partner: UNT Libraries Government Documents Department

HRB-22 irradiation phase test data report

Description: Irradiation capsule HRB-22 was a test capsule containing advanced Japanese fuel for the High Temperature Test Reactor (HTTR). Its function was to obtain fuel performance data at HTTR operating temperatures in an accelerated irradiation environment. The irradiation was performed in the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL). The capsule was irradiated for 88.8 effective full power days in position RB-3B of the removable beryllium (RB) facility. The maximum fuel compact temperature was maintained at or below the allowable limit of 1300{degrees}C for a majority of the irradiation. This report presents the data collected during the irradiation test. Included are test thermocouple and gas flow data, the calculated maximum and volume average temperatures based on the measured graphite temperatures, measured gaseous fission product activity in the purge gas, and associated release rate-to-birth rate (R/B) results. Also included are quality assurance data obtained during the test.
Date: March 1, 1995
Creator: Montgomery, F.C.; Acharya, R.T.; Baldwin, C.A.; Rittenhouse, P.L.; Thoms, K.R. & Wallace, R.L.
Partner: UNT Libraries Government Documents Department

Very High Temperature Reactor (VHTR) Survey of Materials Research and Development Needs to Support Early Deployment

Description: The VHTR reference concept is a helium-cooled, graphite moderated, thermal neutron spectrum reactor with an outlet temperature of 1000 C or higher. It is expected that the VHTR will be purchased in the future as either an electricity producing plant with a direct cycle gas turbine or a hydrogen producing (or other process heat application) plant. The process heat version of the VHTR will require that an intermediate heat exchanger (IHX) and primary gas circulator be located in an adjoining power conversion vessel. A third VHTR mission - actinide burning - can be accomplished with either the hydrogen-production or gas turbine designs. The first ''demonstration'' VHTR will produce both electricity and hydrogen using the IHX to transfer the heat to either a hydrogen production plant or the gas turbine. The plant size, reactor thermal power, and core configuration will be designed to assure passive decay heat removal without fuel damage during accidents. The fuel cycle will be a once-through very high burnup low-enriched uranium fuel cycle. The purpose of this report is to identify the materials research and development needs for the VHTR. To do this, we focused on the plant design described in Section 2, which is similar to the GT-MHR plant design (850 C core outlet temperature). For system or component designs that present significant material challenges (or far greater expense) there may be some viable design alternatives or options that can reduce development needs or allow use of available (cheaper) materials. Nevertheless, we were not able to assess those alternatives in the time allotted for this report and, to move forward with this material research and development assessment, the authors of this report felt that it was necessary to use a GT-MHR type design as the baseline design.
Date: January 1, 2003
Creator: Shaber, Eric; Baccaglini, G.; Ball, S.; Burchell, T.; Corwin, B.; Fewell, T. et al.
Partner: UNT Libraries Government Documents Department

Potential effects of gallium on cladding materials

Description: This paper identifies and examines issues concerning the incorporation of gallium in weapons derived plutonium in light water reactor (LWR) MOX fuels. Particular attention is given to the more likely effects of the gallium on the behavior of the cladding material. The chemistry of weapons grade (WG) MOX, including possible consequences of gallium within plutonium agglomerates, was assessed. Based on the calculated oxidation potentials of MOX fuel, the effect that gallium may have on reactions involving fission products and possible impact on cladding performance were postulated. Gallium transport mechanisms are discussed. With an understanding of oxidation potentials and assumptions of mechanisms for gallium transport, possible effects of gallium on corrosion of cladding were evaluated. Potential and unresolved issues and suggested research and development (R and D) required to provide missing information are presented.
Date: October 1, 1997
Creator: Wilson, D.F.; Beahm, E.C.; Besmann, T.M.; DeVan, J.H.; DiStefano, J.R.; Gat, U. et al.
Partner: UNT Libraries Government Documents Department

Next Generation Nuclear Plant Materials Research and Development Program Plan, Revision 4

Description: DOE has selected the High Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 950°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed reactor and use low-enriched uranium, TRISO-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development (R&D) Program is responsible for performing R&D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Some of the general and administrative aspects of the R&D Plan include: • Expand American Society of Mechanical Engineers (ASME) Codes and American Society for Testing and Materials (ASTM) Standards in support of the NGNP Materials R&D Program. • Define and develop inspection needs and the procedures for those inspections. • Support selected university materials related R&D activities that would be of direct benefit to the NGNP Project. • Support international materials related collaboration activities through the DOE sponsored Generation IV International Forum (GIF) Materials and Components (M&C) Project Management Board (PMB). • Support document review activities through the Materials Review Committee (MRC) or other suitable forum.
Date: September 1, 2007
Creator: Hayner, G. O.; Bratton, R. L.; Mizia, R. E.; Windes, W. E.; Corwin, W. R.; Burchell, T. D. et al.
Partner: UNT Libraries Government Documents Department