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Due
to the number of reports, the following
are the categories we present them.
(Note: these reports were cited on Toxline
at Toxnet in November 2005) |
| Fluoride
(all reports except Canada) |
Fluoride:
CANADA |
Fluoride
in the
Nuclear Industry |
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Note:
many of the
Canadian communities
cited in these reports
border the US. |
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Note:
this is a selected
list of reports. |
NTIS
Reports can be ordered by: phone at 1-800-553-NTIS (U.S.
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|
| Order
Number
Source
Number |
Date
Published / Title / Author & Affiliation / Sponsor Agency |
Abstract
/Keywords |
| NTIS/01920131
22 p |
2005.
Efect of Fluoride in NTS Groundwaters on the Aqueous Speciation
of U, Np, Pu, Am and Eu.
Authors:
Bruton CJ, Nimz GJ
Author
Address: Lawrence Livermore National Lab., CA.
Sponsored
by Department of Energy, Washington, DC. |
To address
SNJV concerns that fluoride in Nevada Test site (NTS) groundwaters
may impact radionuclide speciation and transport, NTS water
quality databases were obtained and scanned for analyses with
high fluoride concentrations (> 10 mg/L). The aqueous
speciation of nine representative samples of these groundwaters
with added trace amounts of uranium (U), neptunium (Np), plutonium
(Pu), americium (Am) and europium (Eu) was then calculated with
the computer code EQ3NR assuming a temperature of 25 C, using
currently available thermodynamic data for these species. Under
conditions where U(VI), Np(V), Pu(IV), Am(III) and Eu(III) dominate,
F complexes are insignificant (<1 mole %) for U, Np,
Pu and Am. Eu-F complexes may be significant
in groundwaters that lack bicarbonate, possess pH values less
than about 7 at ambient temperatures, or contain F
in extremely high concentrations (e.g. > 50 mg/L). |
NTIS/00400096
62 p |
2004.
Caustic-Side
Solvent Extraction: Extended Equilibrium Modeling of Cesium
and Potassium Distribution Behavior.
Authors:
Delmau LH
Bostick DA
Haverlock TJ
Moyer BA
Oak Ridge National Lab., TN.
Sponsored
by Department of Energy, Washington,
DC. |
An extension of the model developed in FYOl for predicting equilibrium
distribution ratios in the Caustic-Side Solvent Extraction (CSSX)
process is presented here. Motivation for extending the model
arose from the need to predict extraction performance of the
recently optimized solvent composition and the desire to include
additional waste components. This model involves the extraction
of cesium and potassium from different cesium, potassium, and
sodium media over a large range of concentrations. Those different
media include a large variety of anions such as nitrate, hydroxide,
nitrite, chloride, fluoride, sulfate,
and carbonate. The model was defined based on several hundreds
of experimental data points and predicted satisfactorily the
cesium extraction from five different SRS waste simulants. This
process model encompassed almost exclusively 1:l:l metal:anion:ligand
species. Fluoride, sulfate, and carbonate
species were found to be very little extractable, and their
main impact is reflected through their activity effects.
This model gave a very good cesium and potassium extraction
prediction from sodium salts, which is what is needed when trying
to predict the behavior from actual waste. However, the extraction
from potassium or cesium salts, and the extraction of sodium
could be improved, and some additional effort was devoted to
improve the thermodynamic rigor of the model. Toward this end,
more detailed anion-specific models were developed based on
the cesium, potassium, and sodium distribution ratios obtained
with simple systems containing single anions, but it has not
yet proven possible to combine those models to obtain better
predictions than provided by the process model. |
| NTIS/01920118
20 p |
2003.
Evaporative Concentration of 100x J13 Ground Water at 60%
Relative Humidity and 90 degrees C.
Authors:
Nguyen QA, Hailey P, Sutton M, Staggs K, Alai M
Author
Address: Lawrence Livermore National Lab., CA.
Sponsored by Department of Energy, Washington,
DC. |
In these
experiments we studied the behavior of a synthetic concentrated
J13 solution as it comes in contact with a Ni-Cr-Mo-alloy
selected for waste canisters in the designated high-level
nuclear-waste repository at Yucca Mountain,
Nevada. Concentrated synthetic J13 solution was allowed
to drip slowly onto heated test specimens (90DGC, 60% relative
humidity) where the water moved down the surface of the specimens,
evaporated and minerals precipitated. Mineral separation or
zoning along the evaporation path was not observed. We infer
from solid analyses and geochemcial modeling, that the
most corrosive components (Ca, Mg, and F) are limited
by mineral precipitation. Minerals identified by x-ray diffraction
include thermonatrite, natrite, and trona, all sodium carbonate
minerals, as well as kogarkoite (Na(sub3)SO(sub 4)F), halite
(NaCl), and niter (KNO(sub 3)). Calcite and a magnesium silicate
precipitation are based on chemical analyses of the solids
and geochemical modeling. The most significant finding of
this study is that sulfate and fluoride concentrations are
controlled by the solubility of kogarkoite. Kogarkoite thermodynamic
data are needed in the Yucca Mountain
Project database to predict the corrosiveness of carbonate
brines and to establish the extent to which fluoride is removed
from the brines as a solid. |
NTIS/DE2003-814115
40 p |
2002.
Gamma Radiolysis Study of UO(sub 2)F(sub 2)0.4H(sub 2)O Using
Spent Nuclear Fuelf Elements from the High Flux Istope Reactor.
Authors:
Icenhour AS
Toth LM
Oak
Ridge National Lab., TN.
Sponsored
by Department of Energy, Washington,
DC. |
The development
of a standard for the safe, long-term storage of (sup 233) U-containing
materials resulted in the identification of several needed experimenta1
studies. These studies were largely related to the potential
for the generation of unacceptable pressures or the formation
of deleterious products duriug storage of uranium oxides. The
primary concern was that these conditions could occur as a result
of the radiolysis of residual impurities--specifically fluorides
and water-by the high radiation fields associated with (sup
233)U/(sup 232)U materials.
Keywords:
Fuel elements
*HFIR Reactor
*Uranium oxides
Fluorides
Irradiation
Radiolysis
Helium
Neutron flux
Gamma sources |
| NTIS/01920146
12 p |
2002.
Waste
Stream Generated and Waste Disposal Plans for Molten Salt
Reactor Experiment at Oak Ridge National Laboratory.
Authors:
Haghighi MH, Ford MK, Szozda RM, Jugan MR
Author
Address: Bechtel Jacobs Co. LLC, Oak Ridge, TN.
Operational Management Services, Inc., Kingston, TN.
Department of Energy, Oak Ridge, TN.
Prepared in cooperation with Operational Management Services,
Inc., Kingston, TN. and Department of Energy, Oak Ridge, TN.
Sponsored
by Department of Energy, Washington, DC. |
The
Molten Salt Reactor Experiment (MSRE) site is located in Tennessee,
on the U.S. Department of Energy (DOE) Oak Ridge Reservation
(ORR), south of the Oak Ridge National Laboratory (ORNL) main
plant across Haw Ridge in Melton Valley. The MSRE was run
by ORNL to demonstrate the desirable features of the molten-salt
concept in a practical reactor that could be operated safely
and reliably. It introduced the idea of a homogeneous reactor
using fuel salt media and graphite moderation for power and
breeder reactors. The MSRE reactor and associated components
are located in cells beneath the floor in the high-bay area
of Building 7503. The reactor was operated from June 1965
to December 1969. When the reactor was shut down, fuel salt
was drained from the reactor circuit to two drain tanks. A
'clean' salt was then circulated through the reactor as a
decontamination measure and drained to a third drain tank.
When operations ceased, the fuel and flush salts were allowed
t o cool and solidify in the drain tanks. At shutdown, the
MSRE facility complex was placed in a surveillance and maintenance
program. As a result of the S&M program, it was discovered
in 1994 that gaseous uranium (233U/232U) hexafluoride (UF6)
had moved throughout the MSRE process systems. The
UF6 was generated when radiolysis of the fluorine salts caused
the individual constituents to dissociate to their component
atoms, including free fluorine. Some of the free fluorine
combined with uranium fluorides (UF4) in the salt to form
UF6. UF6 is gaseous at slightly above
ambient temperatures; thus, periodic heating of the fuel salts
(which was intended to remedy the radiolysis problems) and
simple diffusion had allowed the UF6 to move out of the salt
and into the process
systems of MSRE.
|
NTIS/PB2001-103950
28p |
2001.
Research
Needs for the US/Russian Conversion Program.
Los
Alamos National Lab., NM.
Sponsored
by Department of Energy, Oak Ridge,
TN. |
This paper
defines the areas for potential collaboration between Los Alamos
National Laboratory (LANL), The Bochvar Institute of Inorganic
Materials (VNIINM) Moscow, The Research Institute of Atomic
Reactors (NIIAR), Dimitrovgrad, and The Institute of High Temperature
Electrochemistry (IVTE UrO AS), Ekaterinburg. The following
areas were chosen for initial collaboration: 1. Stabilization
of valence states of plutonium in solution. 2.
Dissolution of plutonium and plutonium oxide minimizing the
use of fluorides and 3. Processing of plutonium using
room temperature ionic liquids (RTIL) and moderate temperature
molten salts. |
NTIS/02670617
42p |
2000.
Liquid Hydrofluoric Acid Sorption Using Solid Media. Part
1.
Authors:
Osborne PE
DelCul GD
Mattus CH
Icenhour AS
Oak
Ridge National Lab., TN.
Supporting
Agency: Department of Energy,
Washington, DC. |
The conversion
of the uranium hexafluoride, which is removed from the Molten
Salt Reactor Experiment into a stable oxide for long-term storage
will produce a significant amount of slightly contaminated,
concentrated aqueous hydrofluoric acid. Since
the handling of this HF is complicated and dangerous, it was
decided to transform it into a stable solid fluoride.
Tests have been performed to identify the best media to use
for trapping the HF. These tests are described in this report. |
NTIS/02530217
228p |
2000.
Radiolytic Effects on Fluoride Impurities in a U(sub 3)O(sub
8) Matrix.
Author:
Icenhour AS
Oak
Ridge National Lab., TN. |
The safe
handling and storage of radioactive materials require an understanding
of the effects of radiolysis on those materials. Radiolysis
may result in the production of gases (e.g., corrosives) or
pressures that are deleterious to storage containers. A study
has been performed to address these concerns as they relate
to the radiolysis of residual fluoride compounds in uranium
oxides. |
NTIS/02821209
36p |
1999.
Evaluation
of Fluorine-Trapping Agents for Use During Storage of the
MSRE Fuel Salt.
Authors:
Brynestad J
Williams DF
Oak
Ridge National Lab., TN.
Supporting
Agency: Department of Energy,
Washington, DC. |
A
fundamental characteristic of the room temperature Molten Salt
Reactor Experiment (MSRE) fuel is that the radiation from the
retained fission products and actinides interacts with this
fluoride salt to produce fluorine gas.
The purpose of this investigation was to identify fluorine-trapping
materials for the MSRE fuel salt that can meet both the requirement
of interim storage in a sealed (gastight) container and the
vented condition required for disposal at the Waste Isolation
Pilot Plant (WIPP). Sealed containers will be needed for interim
storage because of the large radon source that remains even
in fuel salt stripped of its uranium content. An experimental
program was undertaken to identify the most promising candidates
for efficient trapping of the radiolytic fluorine generated
by the MSRE fuel salt. Because of the desire to avoid pressurizing
the closed storage containers, an agent that traps fluorine
without the generation of gaseous products was sought. |
NTIS/02410148
50p |
1998.
Characterization Report on Sand, Slag, and Crucible Residues
and on Fluoride Residues.
Author:
Murray AM
Savannah
River Site.
Sponsored
by Department of Energy, Washington,
DC. |
This paper
reports on the chemical characterization of the sand, slag,
and crucible (SS&C) residues and the
fluoride residues that may be shipped from the Rocky Flats Environmental
Technology Site (RFETS) to Savannah River Site (SRS). |
NTIS/DE99050998
580p |
1997.
Tank 241-A-101 cores 154 and 156 analytical results for the
final report.
Author:
Steen FH
Fluor
Daniel Hanford Inc.,
Richland, WA (United States).
Sponsored
by Department of Energy, Washington,
DC. |
This report
contains tables of the analytical results from sampling Tank
241-A-101 for the following: fluorides, chlorides, nitrites,
bromides, nitrates, phosphates, sulfates, and oxalates. This
tank is listed on the Hydrogen Watch List. |
NTIS/DE97060091
506p |
1997.
Sanitary landfill groundwater monitoring report. Fourth quarter
1996 and 1996 summary.
Westinghouse
Savannah River Co., Aiken, SC.
Supporting
Agency: Department of Energy,
Washington, DC. |
A maximum
of eighty-nine wells of the LFW series monitor groundwater quality
in the Steed Pond Aquifer (Water Table) beneath the Sanitary
Landfill at the Savannah River Site (SRS).
These wells are sampled quarterly to comply with the South Carolina
Department of Health and Environmental Control Domestic Waste
Permit DWP-087A and as part of the SRS Groundwater Monitoring
Program. Dichloromethane, a common laboratory contaminant, and
chloroethene (vinyl chloride) were the most widespread constituents
exceeding standards during 1996. Benzene, trichloroethylene,
1,4-dichlorobenzene, 1,1-dichloroethylene, lead (total recoverable),
gross alpha, mercury (total recoverable), tetrachloroethylene,
fluoride, thallium, radium-226,
radium-228, and tritium also exceeded
standards in one or more wells. The groundwater flow
direction in the Steed Pond Aquifer (Water Table) beneath the
Sanitary Landfill was to the southeast (universal transverse
Mercator coordinates). The flow rate in this unit was approximately |
NTIS/01860086
76p |
1997.
Analysis of molten salt separation system for nuclear wastes
transmutation.
Authors:
Hwang IS
Park BG
Kim KB
Kwon OS
Korea
Atomic Energy Research Institute, Taejon (Korea, Republic
of). |
Korean.
Typical molten salt separation is ANL-IFR pyroprocessing and
ORNL-MSRE pyroprocessing. IFR pyroprocessing is based on Chloride
chemistry and electrorefining. MSRE pyroprocessing
is based on fluoride chemistry and reductive extraction.
Major technologies of molten salt separation are electrorefining,
electrowining, reductive extraction, and oxide reduction. Common
characteristics of this technologies is to utilize reduction-oxidation
phenomena in molten salt. Electrorefining process is modeled
on the basis of diffusion layer theory and Butler-Volmor relation.
This model is numerically solved by LSODA package. To acquire
the technology of electrorefining process, 3-electrode electrochemical
cell is developed where electrolyte is 500 degree C LiCl-KCl
eutectic molten salt, working electrodes are Ni and Au, and
reference electrode is Ag/AgCl. We have investigated the stable
potential range using cyclic voltammogram of Ni electrode. We
have measured steady state polarization curve of Ni electrode.
Then corrosion potential of Ni electrode is -0.38V(sub Ag/AgCl)
and corrosion current is 1.23 x 10(sup -4) A/cm(sup 2). 12 refs.,
6 tabs., 24 figs. (author) |
NTIS/DE96011602
557p |
1996.
Plutonium
Finishing Plant (PFP) Stabilization Final
Environmental Impact Statement (EIS), Hanford Site,
Richland, Benton County, Washington.
Department
of Energy, Richland, WA. Richland Operations Office. |
The purpose
of this action is to expeditiously and safely reduce radiation
exposure to workers and the risk to the environment. The preferred
alternative for resolution of the safety issue is removal of
readily retrievable plutonium-bearing material in hold-up at
the PFP Facility and stabilization of these and other plutonium-bearing
materials at the PFP Facility through the following four treatment
processes: (1) ion exchange, vertical calcination, and thermal
stabilization of plutonium-bearing solutions;
(2) thermal stabilization using a continuous furnace for oxides,
fluorides, and process residues; (3) repackaging of metals
and alloys; and (4) pyrolysis of polycubes and combustibles. |
NTIS/DE96006708
16p |
1996.
Disposition
of the fluoride fuel and flush salts from the Molten Salt
Reactor experiment at Oak Ridge National Laboratory.
Author:
Peretz FJ
Oak
Ridge National Lab., TN.
Supporting
Agency: Department of Energy,
Washington, DC. |
The Molten
Salt Reactor Experiment (MSRE) is an 8 MW reactor that was operated
at Oak Ridge National Laboratory (ORNL) from 1965 through 1969.
The reactor used a unique liquid salt fuel, composed of a mixture
of LIF, BeF(sub 2), ZrF(sub 4), and UF(sub 4), and operated
at temperatures above 600(degrees)C. The primary fuel salt circulation
system consisted of the reactor vessel, a single fuel salt pump,
and a single primary heat exchanger. Heat was transferred from
the fuel salt to a coolant salt circuit in the primary heat
exchanger. The coolant salt was similar to the fuel salt, except
that it contains only LiF (66%) and BeF, (34%). The coolant
salt passed from the primary heat exchanger to an air-cooled
radiator and a coolant salt pump, and then returned to the primary
heat exchanger. Each of the salt loops was provided with drain
tanks, located such that the salt could be drained out of either
circuit by gravity. A single drain tank was provided for the
non-radioactive coolant salt. Two drain [abstract truncated] |
NTIS/DE96008049
82p |
1996.
Validation
summary report for the 100-KR-4 Groundwater Round 8.
Bechtel
Hanford, Inc., Richland, WA.
Los Alamos Technical Associates, Inc., NM.
Sponsored
by Department of Energy, Washington,
DC. |
This report
presents a summary of data validation results on groundwater
samples collected for the 100-KR-4 Groundwater Round 8 Project.
The analyses performed for this project were as follows: Metals:
inductively coupled plasma (ICP) metals (filtered and unfiltered);
General chemistry: anions (fluoride,
chloride, nitrate, nitrite, phosphate, and sulfate) and turbidity;
and Radiochemistry: carbon-14, gamma scan, gross alpha, gross
beta, strontium-90, tritium, and uranium-234/235/238. The objectives
of this project were to validate all sample delivery groups
(SDG) at level D as defined in the data validation procedures.
|
NTIS/DE96008045
105p |
1996.
Validation
summary report for the 100-HR-3 Groundwater Round 9 Phase
1 and 2.
Bechtel
Hanford, Inc., Richland, WA.
Los Alamos Technical Associates, Inc., NM.
Sponsored
by Department of Energy, Washington,
DC. |
This report
presents a summary of data validation results on groundwater
samples collected for the 100-HR-3 Groundwater Round 9-Phase
I and II Project. The analyses performed for this project were
as follows: Metals--inductively coupled plasma (ICP) metals
(filtered and unfiltered); General Chemistry--anions
(fluoride, chloride, nitrate, nitrite, phosphate, and
sulfate), turbidity, ammonia, nitrate+nitrite, and sulfide;
and Radiochemistry--gross alpha, gross beta, technetium-99,
tritium, and uranium-234/235/238. The objectives of this project
were to validate sample detection limit as defined in the data
validation procedures (WHC 1993). In addition, this report provides
a summary of the data as defined by laboratory performance criteria
and project-specific data quality objectives. |
NTIS/DE98052067
Product
reproduced from digital image.
21p |
1996.
Atmospheric
dispersion modeling for the worst-case detonation accident
at the proposed Advanced Hydrotest Facility.
Author:
Bowen BM
USDOE,
Washington, DC. |
The Atmospheric
Release Advisory Capability (ARAC) was requested to estimate
credible worst-case dose, air concentration, and deposition
of airborne hazardous materials that would result from a worst-case
detonation accident at the proposed Advanced Hydrotest Facility
(AHT) at the Nevada Test Site (NTS).
Consequences were estimated at the closest onsite facility,
the Device Assembly Facility (DOFF), and offsite location (intersection
of Highway and U.S. 95). The materials
considered in this analysis were weapon-grade plutonium, beryllium,
and hydrogen fluoride which is a combustion product whose concentration
is dependent upon the quantity of high explosives. The
analysis compares the calculated results with action guidelines
published by the Department of Defense in DoD 5100.52-M (Nuclear
Weapon Accident Response Procedures). Results indicate that
based on a one kg release of plutonium the whole body radiation
dose could be as high as 3 Rem at the DOFF. This level approaches
the 5 Rem level for [abstract truncated] |
|
