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.

Electric Vehicle Battery Simulation: How Electrode Porosity and Thickness Impact Cost and Performance

2021 ◽  
Author(s):  
Yixin Zhao ◽  
Sara Behdad
Keyword(s):  
National Laboratory ◽  
Active Material ◽  
Argonne National Laboratory ◽  
Negative Electrode ◽  
Lithium Ion ◽  
The United States ◽  
Positive Electrode ◽  
Department Of Energy ◽  
United States Department ◽  
And Performance

Abstract Lithium-ion batteries almost exclusively power today’s electric vehicles (EVs). Cutting battery costs is crucial to the promotion of EVs. This paper aims to develop potential solutions to lower the cost and improve battery performance by investigating its design variables: positive electrode porosity and thickness. The open-access lithium-ion battery design and cost model (BatPac) from the Argonne National Laboratory of the United States Department of Energy, has been used for the analyses. Six pouch battery systems with different positive materials are compared in this study (LMO, LFP, NMC 532/LMO, NMC 622, NMC 811, and NCA). Despite their higher positive active material price, nickel-rich batteries (NMC 622, NMC 811, and NCA) present a cheaper total pack cost per kilowatt-hour than other batteries. The higher thickness and lower porosity can reduce the battery cost, enhance the specific energy, lower the battery mass but increase the performance instability. The reliability of the results in this study is proven by comparing estimated and actual commercial EV battery parameters. In addition to the positive electrode thickness and porosity, six other factors that affect the battery’s cost and performance have been discussed. They include energy storage, negative electrode porosity, separator thickness and porosity, and negative and positive current collector thickness.

scholarly journals Shutdown and Closure of the Experimental Breeder Reactor–II

2002 ◽  
Author(s):  
John A. Michelbacher ◽  
Carl E. Baily ◽  
Daniel K. Baird ◽  
S. Paul Henslee ◽  
Collin J. Knight ◽  
...  
Keyword(s):  
Liquid Metal ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Passive Layer ◽  
Department Of Energy ◽  
Primary System ◽  
Secondary System ◽  
Reactor Plant ◽  
Carbon Dioxide Gas ◽  
Safe Condition

The Department of Energy mandated the termination of the Integral Fast Reactor (IFR) Program, effective October 1, 1994. To comply with this decision, Argonne National Laboratory-West (ANL-W) prepared a plan providing detailed requirements to maintain the Experimental Breeder Reactor-II (EBR-II) in a radiologically and industrially safe condition, including removal of all irradiated fuel assemblies from the reactor plant, and removal and stabilization of the primary and secondary sodium, a liquid metal used to transfer heat within the reactor plant. The EBR-II is a pool-type reactor. The primary system contained approximately 325 m3 (86,000 gallons) of sodium and the secondary system contained 50 m3 (13,000 gallons). In order to properly dispose of the sodium in compliance with the Resource Conservation and Recovery Act (RCRA), a facility was built to react the sodium to a solid sodium hydroxide monolith for burial as a low level waste in a land disposal facility. Deactivation of a liquid metal fast breeder reactor (LMFBR) presents unique concerns. Residual amounts of sodium remaining in circuits and components must be passivated, inerted, or removed to preclude future concerns with sodium-air reactions that could generate potentially explosive mixtures of hydrogen and leave corrosive compounds. The passivation process being implemented utilizes a moist carbon dioxide gas that generates a passive layer of sodium carbonate/sodium bicarbonate over any quantities of residual sodium. Tests being conducted will determine the maximum depths of sodium that can be reacted using this method, defining the amount that must be dealt with later to achieve RCRA clean closure. Deactivation of the EBR-II complex is on schedule for a March, 2002, completion. Each system associated with EBR-II has an associated layup plan defining the system end state, as well as instructions for achieving the layup condition. A goal of system-by-system layup is to minimize surveillance and maintenance requirements during the interim period between deactivation and decommissioning. The plans also establish document archival of not only all the closure documents, but also the key plant documents (P&IDs, design bases, characterization data, etc.) in a convenient location to assure the appropriate knowledge base is available for decommissioning, which could occur decades in the future.

Modeling Shock and Vibration on Used Nuclear Fuel During Normal Conditions of Transportation

2019 ◽  
Author(s):  
Nicholas Klymyshyn ◽  
Pavlo Ivanusa ◽  
Kevin Kadooka ◽  
Casey Spitz
Keyword(s):  
Nuclear Fuel ◽  
Numerical Models ◽  
National Laboratory ◽  
Pacific Northwest National Laboratory ◽  
The United States ◽  
Lessons Learned ◽  
Department Of Energy ◽  
United States Department ◽  
Fuel Cladding ◽  
Used Nuclear Fuel

Abstract In 2017, the United States Department of Energy (DOE) collaborated with Spanish and Korean organizations to perform a multimodal transportation test to measure shock and vibration loads imparted to used nuclear fuel (UNF) assemblies. This test used real fuel assembly components containing surrogate fuel mass to approximate the response characteristics of real, irradiated used nuclear fuel. Pacific Northwest National Laboratory was part of the test team and used the data collected during this test to validate numerical models needed to predict the response of real used nuclear fuel in other transportation configurations. This paper summarizes the modeling work and identifies lessons learned related to the modeling and analysis methodology. The modeling includes railcar dynamics using the NUCARS software code and explicit dynamic finite element modeling of used nuclear fuel cladding in LS-DYNA. The NUCARS models were validated against railcar dynamics data collected during captive track testing at the Federal Railroad Administration’s Transportation Technology Center in Pueblo, CO. The LS-DYNA models of the fuel cladding were validated against strain gage data collected throughout the test campaign. One of the key results of this work was an assessment of fuel cladding fatigue, and the methods used to calculate fatigue are detailed in this paper. The validated models and analysis methodologies described in this paper will be applied to evaluate future UNF transportation systems.

Enhanced VVER-1000 Fuel Technology and Performance

2009 ◽  
Author(s):  
H. Shah ◽  
R. Latorre ◽  
G. Raspopin ◽  
J. Sparrow
Keyword(s):  
Pacific Northwest ◽  
Nuclear Power ◽  
Power Plants ◽  
National Laboratory ◽  
Pacific Northwest National Laboratory ◽  
The United States ◽  
Department Of Energy ◽  
United States Department ◽  
Safety Level ◽  
And Performance

The United States Department of Energy, through the Pacific Northwest National Laboratory (PNNL), provides management and technical support for the International Nuclear Safety Program (INSP) to improve the safety level of VVER-1000 nuclear power plants in Central and Eastern Europe.

RELAP5/MOD3.2 Investigation of a VVER-1000 MCP Switching on Problem

2002 ◽  
Author(s):  
Pavlin Groudev ◽  
Malinka Pavlova
Keyword(s):  
Power Plant ◽  
Nuclear Power ◽  
Argonne National Laboratory ◽  
The United States ◽  
Reactor Power ◽  
Department Of Energy ◽  
The Other ◽  
United States Department ◽  
Coolant Pump ◽  
Bulgarian Academy

This paper provides a discussion of various RELAP5 parameters calculated for the investigation of the nuclear power reactor parameter behavior in case of switching on one main coolant pump (MCP) when the other three MCPs are in operation. The reference power plant for this analysis is Unit 6 at the Kozloduy Nuclear Power Plant (NPP) site. Operational data from Kozloduy NPP have been used for the purpose of assessing how the RELAP5 model compares against plant data. During the plant-commissioning phase at Kozloduy NPP Unit 6 a number of experiments have been performed. One of them is switching on MCP when the other three MCPs are in operation. The event is characterized by rapid increase in the flow through the core resulting in a coolant temperature decrease, which leads to insertion of positive reactivity due to the modeled feedback mechanisms. This investigation has been conducted by Bulgarian and Russian specialists on the stage when the reactor power was at 75% of the nominal level. The purpose of the experiment was the complete testing of reliability of all power plant equipment, testing the reliability of the main regulators and defining a jump of the neutron reactor power in case of switching on of one main coolant pump. The Institute for Nuclear Research and Nuclear Energy - Bulgarian Academy of Sciences (INRNE-BAS), Sofia, and Kozloduy NPP have been developing a RELAP5/MOD3.2 model for Kozloduy NPP VVER-1000 for investigation of operational occurrences, abnormal events, and design basis scenarios. This investigation is a process that compares the analytical results obtained by the RELAP5 computer model of the VVER-1000 against the experimental transient data received from the Kozloduy NPP Unit 6. The comparisons between the RELAP5 results and the test data indicate good agreement. This report was possible through the participation of leading specialists from Kozloduy NPP and with the support of Argonne National Laboratory, under the International Nuclear Safety Program (INSP) of the United States Department of Energy.

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.

scholarly journals Risk-Informed Safety Margin Characterization

2009 ◽  
Author(s):  
Stephen M. Hess ◽  
Nam Dinh ◽  
John P. Gaertner ◽  
Ronaldo Szilard
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
National Laboratory ◽  
Safety Margin ◽  
The United States ◽  
Department Of Energy ◽  
United States Department ◽  
Plant Design ◽  
Time Horizons ◽  
Safety Margins

The concept of safety margins has served as a fundamental principle in the design and operation of commercial nuclear power plants (NPPs). Defined as the minimum distance between a system’s “loading” and its “capacity”, plant design and operation is predicated on ensuring an adequate safety margin for safety-significant parameters (e.g., fuel cladding temperature, containment pressure, etc.) is provided over the spectrum of anticipated plant operating, transient and accident conditions. To meet the anticipated challenges associated with extending the operational lifetimes of the current fleet of operating NPPs, the United States Department of Energy (USDOE), the Idaho National Laboratory (INL) and the Electric Power Research Institute (EPRI) have developed a collaboration to conduct coordinated research to identify and address the technological challenges and opportunities that likely would affect the safe and economic operation of the existing NPP fleet over the postulated long-term time horizons. In this paper we describe a framework for developing and implementing a Risk-Informed Safety Margin Characterization (RISMC) approach to evaluate and manage changes in plant safety margins over long time horizons.

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