Immobilization and Recovery of Thorium, A Neptunium Surrogate, using Phase-Separated Glasses

1996 ◽  
Vol 465 ◽  
Author(s):  
T. F. Meaker ◽  
D. Karraker ◽  
M. Tosten ◽  
J. M. Pareizs ◽  
W. G. Ramsey
Keyword(s):  
Heat Treatment ◽  
Phase Separation ◽  
Nitric Acid ◽  
Ion Exchange ◽  
National Laboratory ◽  
Target Material ◽  
The United States ◽  
Sodium Borate ◽  
Oak Ridge ◽  
Treatment Conditions

ABSTRACTThe Savannah River Site has the majority of the United States' supply of neptunium currently stored in an acid solution in one of their canyon facilities. A program is being developed that could be utilized to ship this material, as glass, to Oak Ridge National Laboratory where the Np could be leached from the glass, purified by ion exchange and made into target material for the production of Pu-238. Ion exchange purification dictates no material be in the leachate making the isolation of the Np difficult. We have developed a process using thorium as a surrogate for Np that could immobilize the Np into a soda borosilicate glass for shipment. To achieve recovery of the Np, the glass can be phase separated prior to leaching with nitric acid. Phase separation would produces a Np-rich sodium-borate phase and a Si-rich phase similar to a Vycor® glass. The nitric acid selectively attacks the sodium-borate phase allowing high Np recovery in a solution that contains only sodium and boron. These can be easily separated from Np by ion exchange. Essentially all of the silicon which would interfere with ion exchange by precipitation is retained in the Vycor®-type phase. This technology may also be applied to other actinides stored in relatively pure solutions.This paper will report the optimization of variables for maximizing Th (a Np surrogate) recovery while minimizing Si release. Th solubility in glass, heat treatment conditions and leaching parameters will be discussed. Transmission Electron Microscopy (TEM) with energy dispersive spectroscopy (EDS) data will be included to show phase separation after heat treatment.

Building the TeraGrid

2005 ◽  
Vol 363 (1833) ◽  
pp. 1715-1728 ◽  
Author(s):  
Peter H Beckman
Keyword(s):  
High Performance ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
The United States ◽  
Parallel Computer ◽  
Software Environment ◽  
Distributed Resources ◽  
Oak Ridge ◽  
California Institute Of Technology ◽  
Institute Of Technology

On 1 October 2004, the most ambitious high-performance Grid project in the United States—the TeraGrid—became fully operational. Resources at nine sites—the San Diego Supercomputer Center, the California Institute of Technology, the National Center for Supercomputing Applications, the University of Chicago/Argonne National Laboratory, Pittsburgh Supercomputing Center, Texas Advanced Computing Center, Purdue University, Indiana University and Oak Ridge National Laboratory—were joined via an ultra-fast optical network, unified policies and security procedures and a sophisticated distributed computing software environment. Funded by the National Science Foundation, the TeraGrid enables scientists and engineers to combine distributed, multiple data sources with computation at any of the sites or link massively parallel computer simulations to extreme-resolution visualizations at remote sites. A single shared utility lets multiple resources be easily leveraged and provides improved access to advanced computational capabilities. One of the demonstrations of this new model for using distributed resources, Teragyroid, linked the infrastructure of the TeraGrid with computing resources in the United Kingdom via a transatlantic data fibre link. Once connected, the software framework of the RealityGrid project was used to successfully explore lattice-Boltzmann simulations involving lattices of over one billion sites.

Production function for short-rotation woody-crop Populus spp. plantations

1994 ◽  
Vol 24 (1) ◽  
pp. 180-184 ◽  
Author(s):  
David A. Lortz ◽  
David R. Betters ◽  
Lynn L. Wright
Keyword(s):  
Production Function ◽  
Cultural Practices ◽  
National Laboratory ◽  
Development Program ◽  
The United States ◽  
Oak Ridge ◽  
Short Rotation ◽  
Wide Range ◽  
Woody Crop ◽  
Populus Spp

Short-rotation woody-crop Populus spp. plantations have the potential to produce large amounts of biomass in short time periods, typically 4–8 years. A production function equation is shown to predict yields for such plantations. The equation is based, in part, on information from biomass production experiments conducted across the United States. These experimental plots are sponsored by the Biofuels Feedstock Development Program of Oak Ridge National Laboratory. The equation uses nine parameters including both cultural practices and climatic and soil site conditions as independent variables. The equation (R2 = 0.86) is accurate and applicable to a wide range of conditions.

Probabilistic Structural Mechanics Analysis of the Degraded Davis-Besse RPV Head

2006 ◽  
Author(s):  
Paul T. Williams ◽  
Shengjun (Sean) Yin ◽  
B. Richard Bass
Keyword(s):  
National Laboratory ◽  
Damage Model ◽  
The United States ◽  
Structural Mechanics ◽  
Companion Paper ◽  
Experimental Program ◽  
Oak Ridge ◽  
Steel Technology ◽  
Sensitivity Studies ◽  
Technical Basis

The Heavy-Section Steel Technology (HSST) Program at Oak Ridge National Laboratory (ORNL) performed a probabilistic structural mechanics (PSM) analysis of the damaged Davis Besse reactor pressure vessel head in support of the United States Nuclear Regulatory Commission’s ongoing forensic investigations. This paper presents a summary of the results of that PSM analysis, including a description of the Davis-Besse wastage-area damage model, the technical basis for the model, and the results of sensitivity studies based on a cladding capacity analysis (CCA) and an Accident Sequence Precursor (ASP) investigation of the wastage cavity. A companion paper describes the HSST experimental program carried out at ORNL in parallel with the PSM analysis.

Effects of Chemical Content Variation on the Fracture Probability of PWR Pressure Vessel Subjected to PTS Events

2015 ◽  
Author(s):  
Pin-Chiun Huang ◽  
Hsoung-Wei Chou ◽  
Yuh-Ming Ferng
Keyword(s):  
Pressure Vessel ◽  
National Laboratory ◽  
Nickel Content ◽  
The United States ◽  
Fracture Probability ◽  
Pressure Vessels ◽  
Pressurized Water ◽  
Oak Ridge ◽  
Content Variation ◽  
Chemical Content

This paper is to study the effects of copper and nickel content variations on the fracture probability of the pressurized water reactor (PWR) pressure vessel subjected to pressurized-thermal-shock (PTS) transients. The probabilistic fracture mechanics (PFM) code, FAVOR, which was developed by the Oak Ridge National Laboratory in the United States, is employed to perform the analyses. A Taiwan domestic PWR pressure vessel with varied copper and nickel contents of beltline region welds and plates is investigated in the study. Some PTS transients analyzed from Beaver Valley Unit 1 for establishing the USNRC’s new PTS rule are applied as the loading conditions. It is found that the content variation of copper and nickel will significantly affect the radiation embrittlement and the fracture probability of PWR pressure vessels. The results can be regarded as the risk incremental factors for comparison with the safety regulation requirements on vessel degradation as well as a reference for the operation of PWR plants in Taiwan.

scholarly journals Batch-Scale Hydrofluorination of Li27BeF4 to Support Molten Salt Reactor Development

2015 ◽  
Vol 1 (4) ◽  
Author(s):  
Brian C. Kelleher ◽  
Kieran P. Dolan ◽  
Paul Brooks ◽  
Mark H. Anderson ◽  
Kumar Sridharan
Keyword(s):  
Molten Salt ◽  
Nuclear Reactor ◽  
National Laboratory ◽  
The United States ◽  
Hydraulic Testing ◽  
Molten Salt Reactor ◽  
Fluoride Salt ◽  
Oak Ridge ◽  
University Of Wisconsin ◽  
Analysis Process

Li 2 BeF 4 , or flibe, is the primary candidate coolant for the fluoride-salt-cooled high-temperature nuclear reactor (FHR). Kilogram quantities of pure flibe are required for repeatable corrosion tests of modern reactor materials. This paper details fluoride salt purification by the hydrofluorination–hydrogen process, which was used to regenerate 57.4 kg of flibe originating from the secondary loop of the molten salt reactor experiment (MSRE) at Oak Ridge National Laboratory (ORNL). Additionally, it expounds upon necessary handling precautions required to produce high-quality flibe and includes technological advancements which ease the purification and analysis process. Flibe batches produced at the University of Wisconsin are the largest since the MSRE program, enabling new corrosion, radiation, and thermal hydraulic testing around the United States.

Commercialization Status of Ni3Al-Based Alloys

1996 ◽  
Vol 460 ◽  
Author(s):  
V. K. Sikka
Keyword(s):  
National Laboratory ◽  
Melting Process ◽  
Heat Treating ◽  
The United States ◽  
Operating Conditions ◽  
Current Status ◽  
Oak Ridge ◽  
Technical Developments ◽  
Property Data ◽  
Component Manufacturing

ABSTRACTThe Ni3Al-based alloys have been under development at the Oak Ridge National Laboratory (ORNL) and other research institutions in the United States and around the world for the last ten years. The incremental developments of composition, melting process, casting methods, property data, corrosion data, weldability development, and prototype component testing under production-like operating conditions have pushed the ORNL-developed Ni3Al-based alloys closer to commercialization. This paper will present the highlights of incremental technical developments along with the approach and current status of commercialization. It is concluded that cast components are the primary applications of Ni3Al-based alloys, and applications range from heat-treating fixtures to forging dies. It is also concluded that the commercialization process is accelerated when technology is licensed to an organization that can produce the alloy, has component manufacturing capability, and is also a user.

Mechanistic Insights Into Risk-Informed Revision of ASME Section XI–Appendix G

2011 ◽  
Author(s):  
Terry Dickson ◽  
Mark Kirk ◽  
Eric Focht
Keyword(s):  
Structural Integrity ◽  
National Laboratory ◽  
Health And Safety ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Pressure Vessels ◽  
Nuclear Industry ◽  
Operating Life ◽  
Oak Ridge ◽  
Asme Code

The current regulations, as set forth by the United States Nuclear Regulatory Commission (NRC), to insure that light-water nuclear reactor pressure vessels (RPVs) maintain their structural integrity, throughout their operating life, when subjected to planned normal reactor startup (heat-up) and shutdown (cool-down) transients are specified in Appendix G to 10 CFR Part 50, which incorporates by reference Appendix G to Section XI of the American Society of Mechanical Engineers (ASME) Code. The technical basis for these regulations are generally considered to be conservative and some plants are finding it operationally difficult to heat-up and cool-down within the accepted limits. Consequently, the nuclear industry has developed, and submitted to the ASME Code for approval, an alternative risk-informed methodology that reduces the conservatism and is consistent with methods previously used to develop a risk-informed revision to the regulations for accidental transients such as pressurized thermal shock (PTS). The objective of the alternative methodology is to increase operational flexibility while continuing to provide reasonable assurance of adequate protection to public health and safety. The NRC and its contractor at Oak Ridge National Laboratory (ORNL) are reviewing the industry proposed risk-informed methodology. Previous results of this review, have been reported at PVP, and a NRC report summarizing all results is currently in preparation. The objective of this paper is to discuss and illustrate mechanistic insights into trends shown previously associated with normal cool-down.

THE EXTRACTION OF THORIUM AND SOME LOWER LANTHANIDE NITRATES BY DIBUTYL BUTYL PHOSPHONATE

1962 ◽  
Vol 40 (8) ◽  
pp. 1684-1689 ◽  
Author(s):  
P. G. Manning
Keyword(s):  
Nitric Acid ◽  
National Laboratory ◽  
Nitric Acid Concentration ◽  
Distribution Coefficients ◽  
Metal Species ◽  
Trace Metal Concentration ◽  
Oak Ridge ◽  
Separation Factors ◽  
Lanthanide Nitrates ◽  
Aqueous Nitric Acid

The distribution of thorium and a number of lanthanides between nitric acid and solutions of dibutyl butyl phosphonate in odorless kerosene has been examined as a function of the aqueous nitric acid concentration. Experiments were conducted at trace metal concentration using radioisotopes. Separation factors (denoted by S and defined as the ratio of the distribution coefficients, K, for two metal species) have been measured for some lanthanide–lanthanide couples and also for some thorium–lanthanide couples. Results indicate that separation factors between successive lanthanides (given by S = KZ+1/KZ) at the lower end of the rare-earth series are superior to those obtained with either tributyl phosphate (TBP) (D. Scargill et al. J. Inorg. & Nuclear Chem. 4, 304 (1957)) or trioctyl phosphine oxide (TOPO) (J. M. Schmitt. Oak Ridge National Laboratory. Unpublished data), but as Z increases, SDBBP ~STBP > STOPO. For thorium–lanthanide couples, S′DBBP > S′TBP. Measurements over a range of extractant concentrations indicate that the lanthanides are extracted as trisolvates.

Development of 2012 Edition of JSME Code for Design and Construction of Fast Reactors: (3) Development of the Material Strength Standard of Modified 9Cr-1Mo Steel

2013 ◽  
Author(s):  
Takashi Onizawa ◽  
Yuji Nagae ◽  
Shigeru Takaya ◽  
Tai Asayama
Keyword(s):  
Low Cycle Fatigue ◽  
National Laboratory ◽  
The United States ◽  
Material Strength ◽  
Steam Generators ◽  
Fast Reactors ◽  
Oak Ridge ◽  
Environment Effects ◽  
Strain Equation ◽  
Best Fit

This paper describes the material strength standard of Modified 9Cr-1Mo (ASME Gr.91) steel in the design code for fast reactors of 2012 edition published by the Japan Society of Mechanical Engineers. Modified 9Cr-1Mo is to be used for primary and secondary coolant circuits, including intermediate heat exchangers and steam generators for the Japan Sodium Cooled Fast Reactor (JSFR). Modified 9Cr-1Mo steel was developed in Oak Ridge National Laboratory in the United States. Application of Modified 9Cr-1Mo to JSFR needs the material strength standard. Therefore, the authors developed the material strength standard. The material strength standard involved allowable limits such as S0, Sm, Su, Sy, SR and St and so on, environment effects such as sodium effects. In addition, material characteristic equations (Creep rupture equation, creep strain equation and equation of best fit curve for low-cycle fatigue life and so on) necessary for the allowable limits were involved. This paper describes the contents of the material strength standard.

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