Lifetimes of 214Po and 212Po measured with Counting Test Facility at Gran Sasso National Laboratory

2014 ◽  
Vol 138 ◽  
pp. 444-446
Author(s):  
L. Miramonti ◽  
G. Bellini ◽  
J. Benziger ◽  
D. Bick ◽  
G. Bonfini ◽  
...  
Keyword(s):  
National Laboratory ◽  
Test Facility

scholarly journals Development of a Multi-Loop Flow and Heat Transfer Facility for Advanced Nuclear Reactor Thermal Hydraulic and Hybrid Energy System Studies

2014 ◽  
Author(s):  
James E. O’Brien ◽  
Piyush Sabharwall ◽  
SuJong Yoon
Keyword(s):  
High Temperature ◽  
Nuclear Reactor ◽  
National Laboratory ◽  
Energy System ◽  
Operating Conditions ◽  
Test Facility ◽  
Flow Loop ◽  
Hybrid Energy ◽  
Conceptual Design Phase ◽  
Intermediate Heat Exchanger

A new high-temperature multi-fluid, multi-loop test facility for advanced nuclear applications is under development at the Idaho National Laboratory. The facility will include three flow loops: high-temperature helium, molten salt, and steam/water. Molten salts have been identified as excellent candidate heat transport fluids for primary or secondary coolant loops, supporting advanced high temperature and small modular reactors (SMRs). Details of some of the design aspects and challenges of this facility, which is currently in the conceptual design phase, are discussed. A preliminary design configuration will be presented, with the required characteristics of the various components. The loop will utilize advanced high-temperature compact printed-circuit heat exchangers (PCHEs) operating at prototypic intermediate heat exchanger (IHX) conditions. The initial configuration will include a high-temperature (750°C), high-pressure (7 MPa) helium loop thermally integrated with a molten fluoride salt (KF-ZrF4) flow loop operating at low pressure (0.2 MPa) at a temperature of ∼450°C. Experiment design challenges include identification of suitable materials and components that will withstand the required loop operating conditions. Corrosion and high temperature creep behavior are major considerations. The facility will include a thermal energy storage capability designed to support scaled process heat delivery for a variety of hybrid energy systems and grid stabilization strategies. Experimental results obtained from this research will also provide important data for code verification and validation (V&V) related to these systems.

scholarly journals Conceptual Design of a MEDE Treatment System for Sodium Bonded Fuel

2008 ◽  
Author(s):  
Carl E. Baily ◽  
Karen A. Moore ◽  
Collin J. Knight ◽  
Peter B. Wells ◽  
Paul J. Petersen ◽  
...  
Keyword(s):  
Spent Nuclear Fuel ◽  
National Laboratory ◽  
Department Of Energy ◽  
Test Facility ◽  
Evaporation Chamber ◽  
Oak Ridge ◽  
Uranium Fuel ◽  
Fuel Assemblies ◽  
Metal Fuel ◽  
Enriched Uranium

Unirradiated sodium bonded metal fuel and casting scrap material containing highly enriched uranium (HEU) is stored at the Materials and Fuels Complex (MFC) on the Idaho National Laboratory (INL). This material, which includes intact fuel assemblies and elements from the Fast Flux Test Facility (FFTF) and Experimental Breeder Reactor-II (EBR-II) reactors, as well as scrap material from the casting of these fuels, has no current use under the terminated reactor programs for both facilities. The Department of Energy (DOE), under the Sodium-Bonded Spent Nuclear Fuel Treatment Record of Decision (ROD), has determined that this material could be prepared and transferred to an off-site facility for processing and eventual fabrication of fuel for commercial nuclear reactors. A plan is being developed to prepare, package, and transfer this material to the DOE HEU Disposition Program Office (HDPO), located at the Y-12 National Security Complex in Oak Ridge, Tennessee. Disposition of the sodium bonded material will require separating the elemental sodium from the metallic uranium fuel. A sodium distillation process known as MEDE (Melt-Drain-Evaporate), will be used for the separation process. The casting scrap material needs to be sorted to remove any foreign material or fines that are not acceptable to the HDPO program. Once all elements have been cut and loaded into baskets, they are then loaded into an evaporation chamber as the first step in the MEDE process. The chamber will be sealed and the pressure reduced to approximately 200 mtorr. The chamber will then be heated as high as 650 °C, causing the sodium to melt and then vaporize. The vapor phase sodium will be driven into an outlet line where it is condensed and drained into a receiver vessel. Once the evaporation operation is complete, the system is de-energized and returned to atmospheric pressure. This paper describes the MEDE process as well as a general overview of the furnace systems, as necessary, to complete the MEDE process.

scholarly journals Completion of Experimental Breeder Reactor-II Sodium Processing at Argonne National Laboratory

2002 ◽  
Author(s):  
Mary D. McDermott ◽  
Charles D. Griffin ◽  
Daniel K. Baird ◽  
Carl E. Baily ◽  
John A. Michelbacher ◽  
...  
Keyword(s):  
Sodium Hydroxide ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
The United States ◽  
Department Of Energy ◽  
Test Facility ◽  
United States Department ◽  
Safe Condition ◽  
And Control ◽  
Low Level Radioactive Waste

The Experimental Breeder Reactor - II (EBR-II) at Argonne National Laboratory - West (ANL-W) was shutdown in September 1994 as mandated by the United States Department of Energy. Located in eastern Idaho, this sodium-cooled reactor had been in service since 1964, and was a test facility for fuels development, materials irradiation, system and control theory tests, and hardware development. The EBR-II termination activities began in October 1994, with the reactor being maintained in an industrially and radiologically safe condition for decommissioning. With the shutdown of EBR-II, its sodium coolant became a waste necessitating its reaction to a disposal form. A Sodium Process Facility (SPF), designed to convert sodium to 50 wt% sodium hydroxide, existed at the ANL-W site, but had never been operated. The SPF was upgraded to current standards and codes, and then modified in 1998 to convert the sodium to 70 wt% sodium hydroxide, a substance that solidifies at 65°C (150°F) and is acceptable for burial as low level radioactive waste in Idaho. In December 1998, the SPF began operations. Working with sodium and highly concentrated sodium hydroxide presented some unique operating and maintenance conditions. Several lessons were learned throughout the operating period. Processing of the 330 m3 (87,000 gallons) of EBR-II primary sodium, 50 m3 (13,000 gallons) of EBR-II secondary sodium, and 290 m3 (77,000 gallons) of Fermi-1 primary sodium was successfully completed in March 2001, ahead of schedule and within budget.

Creep Strength and Microstructure of AL20-25+Nb Alloy Sheets and Foils for Advanced Microturbine Recuperators

2006 ◽  
Author(s):  
Philip J. Maziasz ◽  
John P. Shingledecker ◽  
Neal D. Evans ◽  
Yukinori Yamamoto ◽  
Karren L. More ◽  
...  
Keyword(s):  
Phase I ◽  
Phase Ii ◽  
Creep Resistance ◽  
National Laboratory ◽  
Rupture Life ◽  
Cost Effective ◽  
Test Facility ◽  
Oak Ridge ◽  
Collaborative Program ◽  
Wide Range

The Oak Ridge National Laboratory (ORNL) and ATI Allegheny-Ludlum began a collaborative program in 2004 to produce a wide range of commercial sheets and foils of the new AL20-25+Nb stainless alloy, specifically designed for advanced microturbine recuperator applications. There is a need for cost-effective sheets/foils with more performance and reliability at 650–750°C than 347 stainless steel, particularly for larger 200–250 kW microturbines. Phase I of this collaborative program produced the sheets and foils needed for manufacturing brazed plated-fin (BPF) aircells, while Phase II provided foils for primary surface (PS) aircells, and modified processing to change the microstructure of sheets and foils for improved creep-resistance. Phase I sheets and foils of AL20-25+Nb have much more creep-resistance than 347 steel at 700–750°C, and foils are slightly stronger than HR120 and HR230. Preliminary results for Phase II show nearly double the creep-rupture life of sheets at 750°C/100 MPa, with the first foils tested approaching the creep resistance of alloy 625 foils. AL20-25+Nb alloy foils are also now being tested in the ORNL Recuperator Test Facility.

Calculation of Critical Heat Flux Using an Inverse Heat Transfer Method to Support TREAT Experiment Analysis

2020 ◽  
Author(s):  
Robert Armstrong ◽  
Charles Folsom ◽  
Connie Hill ◽  
Colby Jensen
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Critical Heat Flux ◽  
Nuclear Reactor ◽  
National Laboratory ◽  
Test Facility ◽  
Inverse Heat Transfer ◽  
Transfer Method ◽  
Nuclear Reactor Safety ◽  
Accident Scenarios

Abstract Heat transfer between cladding and coolant during transient scenarios remains a critical area of uncertainty in understanding nuclear reactor safety. To advance the understanding of transient and accident scenarios involving critical heat flux (CHF), an in-pile experiment for the Transient Reactor Test facility (TREAT) at Idaho National Laboratory (INL) was developed. The experiment, named CHF-Static Environment Rodlet Transient Test Apparatus (CHF-SERTTA), consists of a hollow borated stainless-steel heater rod submerged in a static water pool heated via the (n, α) reaction in boron-10. This paper presents a novel inverse heat transfer method to determine CHF by using the optimization and uncertainty software Dakota to calibrate a RELAP5-3D model of CHF-SERTTA to temperature measurements obtained from a thermocouple welded to the surface of the rod.

scholarly journals Test Results From the Idaho National Laboratory 15kW High Temperature Electrolysis Test Facility

2009 ◽  
Author(s):  
Carl M. Stoots ◽  
Keith G. Condie ◽  
James E. O’Brien ◽  
J. Stephen Herring ◽  
Joseph J. Hartvigsen
Keyword(s):  
High Temperature ◽  
National Laboratory ◽  
The United States ◽  
Solid Oxide ◽  
Test Facility ◽  
United States Department ◽  
Idaho National Laboratory ◽  
Production Of Hydrogen ◽  
Heat Recuperation ◽  
High Temperature Electrolysis

A 15 kW high temperature electrolysis test facility has been developed at the Idaho National Laboratory under the United States Department of Energy Nuclear Hydrogen Initiative. This facility is intended to study the technology readiness of using high temperature solid oxide cells for large scale nuclear powered hydrogen production. It is designed to address larger-scale issues such as thermal management (feedstock heating, high temperature gas handling, heat recuperation), multiple-stack hot zone design, multiple-stack electrical configurations, etc. Heat recuperation and hydrogen recycle are incorporated into the design. The facility was operated for 1080 hours and successfully demonstrated the largest scale high temperature solid-oxide-based production of hydrogen to date.

OBSERVATION OF NONLINEAR THOMSON SCATTERING AT BNL-ATF

2007 ◽  
Vol 21 (03n04) ◽  
pp. 473-480 ◽  
Author(s):  
TETSURO KUMITA ◽  
YOSHIO KAMIYA ◽  
MARCUS BABZIEN ◽  
ILAN BEN-ZVI ◽  
KARL KUSCHE ◽  
...  
Keyword(s):  
High Field ◽  
Electron Beams ◽  
National Laboratory ◽  
Thomson Scattering ◽  
Brookhaven National Laboratory ◽  
Test Facility ◽  
Angular Distributions ◽  
X Rays ◽  
Power Laser ◽  
Good Test

Measurement of X-rays produced in nonlinear Thomson scattering of high power laser and electron beams is a good test of electrodynamics in the high field region. We performed an experiment of head-on collision between sub-terawatt CO 2 laser and 60-MeV electron beams at Brookhaven National Laboratory, Accelerator Test Facility. Angular distributions of X-rays produced in Thomson scattering were measured and contribution of the second order nonlinear interaction was observed by comparing the distributions with Monte Carlo simulations based on the semi-classical theory of photon-electron interactions.

Sign in / Sign up

Export Citation Format

Share Document