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Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

Description: The U.S. Department of Energy has selected the High Temperature Gas-cooled Reactor 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 900°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, Tri-Isotopic-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 Program is responsible for performing research and development on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. Studies of potential Reactor Pressure Vessel (RPV) steels have been carried out as part of the pre-conceptual design studies. These design studies generally focus on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Three realistic candidate materials have been identified by this process: conventional light water reactor RPV steels A508/533, 2¼Cr-1Mo in the annealed condition, and modified 9Cr 1Mo ferritic martenistic steel. Based on superior strength and higher temperature limits, the modified 9Cr-1Mo steel has been identified by the majority of design engineers as the preferred choice for the RPV. All of the vendors have concluded, however, that with adequate engineered ...
Date: April 1, 2008
Creator: Wright, J. K. & Wright, R. N.
Partner: UNT Libraries Government Documents Department

Next Generation Nuclear Plant Materials Research and Development Program Plan

Description: The U.S Department of Energy (DOE) has selected the Very High Temperature Reactor (VHTR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production without greenhouse gas emissions. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed, thermal neutron spectrum reactor that will produce electricity and hydrogen in a state-of-the-art thermodynamically efficient manner. The NGNP will use very high burn-up, low-enriched uranium, TRISO-coated fuel and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. 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 Project is envisioned to demonstrate the following: (1) A full-scale prototype VHTR by about 2021; (2) High-temperature Brayton Cycle electric power production at full scale with a focus on economic performance; (3) Nuclear-assisted production of hydrogen (with about 10% of the heat) with a focus on economic performance; and (4) By test, the exceptional safety capabilities of the advanced gas-cooled reactors. Further, the NGNP program will: (1) Obtain a Nuclear Regulatory Commission (NRC) License to construct and operate the NGNP, this process will provide a basis for future performance based, risk-informed licensing; and (2) Support the development, testing, and prototyping of hydrogen infrastructures. 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. The NGNP Materials R&D Program includes the following elements: (1) Developing a specific approach, program plan and other project management tools for managing the R&D program elements; (2) Developing a specific work package for ...
Date: September 1, 2005
Creator: Hayner, G.O.; Bratton, R.L. & Wright, R.N.
Partner: UNT Libraries Government Documents Department

Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

Description: The U.S. Department of Energy (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, with an outlet gas temperature in the range of 750°C, and a 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, Tri-Isotopic (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. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. This technology development plan details the additional research and development (R&D) required to design and license the NGNP RPV, assuming that A 508/A 533 is the material of construction. The majority of additional information that is required is related to long-term aging behavior at NGNP vessel temperatures, which are somewhat above those commonly encountered in the existing database from LWR experience. Additional data are also required for the anticipated NGNP environment. An assessment of required R&D for a Grade 91 vessel has been retained from the first revision of the R&D plan in Appendix B in somewhat less detail. Considerably more development is required for this steel compared to A 508/A 533 including additional irradiation testing for expected NGNP operating temperatures, high-temperature mechanical properties, and extensive studies of long-term microstructural stability.
Date: July 1, 2010
Creator: Wright, J. K. & Wright, R. N.
Partner: UNT Libraries Government Documents Department

FeAl and Mo-Si-B Intermetallic Coatings Prepared by Thermal Spraying

Description: FeAl and Mo-Si-B intermetallic coatings for elevated temperature environmental resistance were prepared using high-velocity oxy-fuel (HVOF) and air plasma spray (APS) techniques. For both coating types, the effect of coating parameters (spray particle velocity and temperature) on the microstructure and physical properties of the coatings was assessed. Fe-24Al (wt.%) coatings were prepared using HVOF thermal spraying at spray particle velocities varying from 540 m/s to 700 m/s. Mo-13.4Si-2.6B coatings were prepared using APS at particle velocities of 180 and 350 m/s. Residual stresses in the HVOF FeAl coatings were compressive, while stresses in the APS Mo-Si-B coatings were tensile. In both cases, residual stresses became more compressive with increasing spray particle velocity due to increased peening imparted by the spray particles. The hardness and elastic moduli of FeAl coatings also increased with increasing particle velocity, again due to an increased peening effect. For Mo-Si-B coatings, plasma spraying at 180 m/s resulted in significant oxidation of the spray particles and conversion of the T1 phase into amorphous silica and {alpha}-Mo. The T1 phase was retained after spraying at 350 m/s.
Date: April 22, 2003
Creator: Totemeier, T.C.; Wright, R.N. & Swank, W.D.
Partner: UNT Libraries Government Documents Department

Processing, properties, and wear resistance of aluminides. [Fe[sub 3]Al; Al[sub 3]Ti]

Description: Fully dense alloys based on Fe[sub 3]Al were produced by reaction synthesis from low cost elemental powders using hot pressing, hot isostatic pressing or Ceracon process. The reaction proceeds by outward spreading of a transient liquid phase from the initial aluminum particle site and precipitation of the compound phase from the liquid. Combustion synthesized material has a very fine grain size that is resistant to coarsening at high temperature because of a high density of fine oxides from the prior particle boundaries. The fine grain size results in approximately twice the yield strength in the reaction synthesized material compared to hot extruded pre-alloyed powder. Combustion synthesis has also been successfully applied to joining Fe[sub 3]Al and to forming coatings on carbon steel substrates. Combustion synthesis has been shown to be viable for fabricating trialuminides from elemental powder compacts. Al[sub 3]Ti, Al[sub 73]Ti[sub 24]Cr[sub 3] and Al[sub 67]Ti[sub 25]Cr[sub 8] were examined. Fully dense, homogeneous materials exhibiting an equiaxed grain structure were produced by conducting reaction and homogenization under pressure, or in a furnace at ambient pressure and subsequently densifying the porous preform by hot consolidation. The tetragonal DO[sub 22] structure was the primary reaction product for all compositions. Most of the Cr remained undissolved after reaction and a homogenization heat treatment at 1200C or above was used to put the Cr into solution and form the desired L1[sub 2] phase.
Date: March 1, 1993
Creator: Wright, R.N.; Rabin, B.H. & Wright, J.K.
Partner: UNT Libraries Government Documents Department

The influence of processing atmosphere on twin-roll melt-spinning of aluminum alloys

Description: Melt-spun samples of Al-2%Fe have been produced in two different processing atmospheres, ambient pressure argon and a high vacuum. High speed video photography and microstructural analysis of the ribbons indicate that the processing pressure influences the interaction of the melt with the copper rolls and thus the thermal history. This results in significant differences in ribbon microstructure.
Date: January 1, 1992
Creator: Sellers, C.H.; Aldrich, K.S.; Cortez, M.M. & Wright, R.N.
Partner: UNT Libraries Government Documents Department

Residual stress determination from a laser-based curvature measurement

Description: Thermally sprayed coating characteristics and mechanical properties are in part a result of the residual stress developed during the fabrication process. The total stress state in a coating/substrate is comprised of the quench stress and the coefficient of thermal expansion (CTE) mismatch stress. The quench stress is developed when molten particles impact the substrate and rapidly cool and solidify. The CTE mismatch stress results from a large difference in the thermal expansion coefficients of the coating and substrate material. It comes into effect when the substrate/coating combination cools from the equilibrated deposit temperature to room temperature. This paper describes a laser-based technique for measuring the curvature of a coated substrate and the analysis required to determine residual stress from curvature measurements. Quench stresses were determined by heating the specimen back to the deposit temperature thus removing the CTE mismatch stress. By subtracting the quench stress from the total residual stress at room temperature, the CTE mismatch stress was estimated. Residual stress measurements for thick (>1mm) spinel coatings with a Ni-Al bond coat on 304 stainless steel substrates were made. It was determined that a significant portion of the residual stress results from the quenching stress of the bond coat and that the spinel coating produces a larger CTE mismatch stress than quench stress.
Date: May 8, 2000
Creator: Swank, W. D.; Gavalya, R. A.; Wright, J. K. & Wright, R. N.
Partner: UNT Libraries Government Documents Department

Novel Experiments to Characterize Creep-Fatigue Degradation in VHTR Alloys

Description: It is well known in energy systems that the creep lifetime of high temperature alloys is significantly degraded when a cyclic load is superimposed on components operating in the creep regime. A test method has been developed in an attempt to characterize creep-fatigue behavior of alloys at high temperature. The test imposes a hold time during the tensile phase of a fully reversed strain-controlled low cycle fatigue test. Stress relaxation occurs during the strain-controlled hold period. This type of fatigue stress relaxation test tends to emphasize the fatigue portion of the total damage and does not necessarily represent the behavior of a component in-service well. Several different approaches to laboratory testing of creep-fatigue at 950°C have been investigated for Alloy 617, the primary candidate for application in VHTR heat exchangers. The potential for mode switching in a cyclic test from strain control to load control, to allow specimen extension by creep, has been investigated to further emphasize the creep damage. In addition, tests with a lower strain rate during loading have been conducted to examine the influence of creep damage occurring during loading. Very short constant strain hold time tests have also been conducted to examine the influence of the rapid stress relaxation that occurs at the beginning of strain holds.
Date: October 1, 2013
Creator: Wright, J. K.; Simpson, J. A.; Carroll, L. J.; Wright, R. N. & Sham, T.-L.
Partner: UNT Libraries Government Documents Department

Characterization of Elevated Temperature Properties of Heat Exchanger and Steam Generator Alloys

Description: The Next Generation Nuclear Plant project is considering Alloy 800H and Alloy 617 for steam generator and intermediate heat exchangers. It is envisioned that a steam generator would operate with reactor outlet temperatures from 750 to 800°C, while an intermediate heat exchanger for primary to secondary helium would operate up to an outlet temperature of 950°C. Although both alloys are of interest due in part to their technical maturity, a number of specific properties require further characterization for design of nuclear components. Strain rate sensitivity of both alloys has been characterized and is found to be significant above 600°C. Both alloys also exhibit dynamic strain aging, characterized by serrated flow, over a wide range of temperatures and strain rates. In general dynamic strain aging is observed to begin at higher temperatures and serrated flow persists to higher temperatures in Alloy 617 compared to Alloy 800H. Dynamic strain aging is a concern for these materials since it is observed to result in reduced ductility for many solid solution alloys. The role of dynamic strain aging in the creep-fatigue behavior of Alloy 617 at temperatures of 800°C and above has also been examined in detail. Serrated flow is found to persist in cyclic stress-strain curves up to nearly the cycle to failure in some temperature and strain regimes. Results of those experiments and implications for creep-fatigue testing protocols will be described.
Date: October 1, 2010
Creator: Wright, J.K.; Carroll, L.J.; Benz, J.K.; Simpson, J.A.; Wright, R.N.; Lloyd, W.R. et al.
Partner: UNT Libraries Government Documents Department

Microstructure and mechanical properties of P/M (powder metallurgy) Fe sub 3 Al alloys

Description: Alloys based on Fe{sub 3}Al have an equilibrium DO{sub 3} structure at low temperatures and transform to a B2 structure above about 550{degree}C. The influence of different rates of quenching from the B2 region to room temperature on the microstructure and mechanical properties of powder metallurgy (P/M) alloys with two different Cr contents has been examined. By optimizing the processing to maximize the amount of B2 order, room temperature ductility approaching 20% has been achieved although the fracture mode is primarily brittle cleavage. The refined microstructure resulting from P/M processing contributes to enhanced yield strength compared to ingot processed materials with similar ductility. Increasing the Cr content from 2 to 5% has little effect on mechanical properties. 8 refs., 12 figs., 2 tabs.
Date: January 1, 1990
Creator: Knibloe, J.R.; Wright, R.N. (EG and G Idaho, Inc., Idaho Falls, ID (USA)) & Sikka, V.K. (Oak Ridge National Lab., TN (USA))
Partner: UNT Libraries Government Documents Department

Irradiation of Microbes from Spent Nuclear Fuel Storage Pool Environments

Description: Microbes have been isolated and identified from spent nuclear fuel storage pools at the Idaho National Engineering and Environmental Laboratory (INEEL). Included among these are Corynebacterium aquaticum, Pseudomonas putida, Comamonas acidovorans, Gluconobacter cerinus, Micrococcus diversus, Rhodococcus rhodochrous, and two strains of sulfate-reducing bacteria (SRB). We examined the sensitivity of these microbes to a variety of total exposures of radiation generated by a 6-MeV linear accelerator (LINAC). The advantage of using a LINAC is that it provides a relatively quick screen of radiation tolerance. In the first set of experiments, we exposed each of the aforementioned microbes along with four additional microbes, pseudomonas aeruginosa, Micrococcus luteus, Escherchia coli, and Deinococcus radiodurans to exposures of 5 x 10{sup 3} and 6 x 10{sup 4} rad. All microbial specimens withstood the lower exposure with little or no reduction in cell population. Upon exposing the microbes to the larger dose of 6 x 10{sup 4} rad, we observed two distinct groupings: microbes that demonstrate resistance to radiation, and microbes that display intolerance through a dramatic reduction from their initial population. Microbes in the radiation tolerant grouping were exposed to 1.1 x 10{sup 5} rad to examine the extent of their resistance. We observe a correlation between radiation resistance and gram stain. The gram-positive species we examined seem to demonstrate a greater radiation resistance.
Date: September 3, 1999
Creator: Breckenridge, C.R.; Watkins, C.S.; Bruhn, D.F.; Roberto, F.F.; Tsang, M.N.; Pinhero, P.J. et al.
Partner: UNT Libraries Government Documents Department