scholarly journals Technetium Leaching from Cementitious Materials

MRS Advances ◽  
2017 ◽  
Vol 2 (13) ◽  
pp. 717-722
Author(s):  
Steven Simner ◽  
Fanny Coutelot ◽  
Hyunshik Chang ◽  
John Seaman
Keyword(s):  
Cementitious Materials ◽  
Savannah River ◽  
Leaching Behavior ◽  
Flexible Wall ◽  
A New Technique ◽  
Traditional Approaches ◽  
River Site ◽  
Low Activity

ABSTRACTAt the Savannah River Site (SRS) low activity salt solution is stabilized via encapsulation within a grout termed saltstone. Saltstone is emplaced into large (multi-million gallon) concrete storage facilities referred to as Saltstone Disposal Units (SDUs). Technetium-99 (99Tc) is a long-lived radionuclide contained in the low activity salt waste and subsequently incorporated into the grout waste form: it is considered a significant contributor to risk with respect to the long-term radiation exposure of the environment surrounding the SDUs. In the reducing, high pH environment within the grout,99Tc is expected to be relatively immobile since it exists in a reduced Tc(IV) oxidation state in the form of sparingly soluble sulfides (TcSx) or hydrated oxides (TcO2.xH2O). However, in the presence of O2(associated with the future infiltration of air or oxygenated ground waters into the saltstone monolith) it is possible for redox-sensitive Tc(IV) to transition into highly soluble (and mobile) Tc(VII) species, pertechnetate (TcO4-), which is more readily transported to the surrounding environment. Traditional approaches to quantifying the leaching behavior of99Tc from cementitious matrices have involved partitioning experiments using size-reduced (crushed/ground) saltstone samples, and determination of the99Tc fraction immobilized by the cementitious solids. Such experiments create artificially high solid-solution contact areas that likely result in higher99Tc leachate concentrations than would be expected for intact, monolithic samples. In the current study a new technique, termed the Dynamic Leaching Method (DLM), is being used to investigate the99Tc leaching behavior of monolithic saltstone samples. The data derived using this technique is intended to inform the SRS Saltstone Disposal Facility (SDF) Performance Assessment (PA) which models the long-term transport of radionuclides from the SDUs to the environment. The DLM utilizes a flexible-wall permeameter to achieve saturated leaching under an elevated hydraulic gradient in an effort to simulate the transport of groundwater through saltstone. Initial findings indicate that the99Tc concentrations in the leachate are on the order of 1E-08 mol/L which suggests that the saltstone leaching behavior is controlled by the solubility of TcO2.xH2O compounds.

Onsite Packaging and Transport of Tritium Waste

2004 ◽  
Author(s):  
Paul S. Blanton ◽  
T. Kurt Houghtaling
Keyword(s):  
Waste Product ◽  
Health And Safety ◽  
Extraction Process ◽  
Radioactive Material ◽  
Savannah River ◽  
Long Term Storage ◽  
Diameter Pipe ◽  
To Receive ◽  
River Site

Radioactive material packagings designed for out-of-commerce shipments are not necessarily subject to the same regulatory requirements as packagings designed for in-commerce service. For example, DOE Order 460.1B permits application of the notion of Equivalent Safety to out-of-commerce shipping within DOE sites. Equivalent safety can be viewed as a reduction in 10 CFR 71 design conditions without a corresponding loss of public health and safety. This paper presents a packaging design identified as the Tritium Spent Melt Overpack (SMO) that successfully utilized equivalent safety at the Savannah River Site (SRS). Because the spent melt materials are highly radioactive, the container must be loaded and closed remotely. The SMO design is a based on twenty-foot long eighteen-inch diameter pipe, with one end closed by welded plate and the open end closed by a latching plug that incorporates bore seals. The SMO receives a single sixteen-inch diameter by 16-foot long crucible partly filled with the waste product from the tritium extraction process. The loaded overpack is moved from the SRS Tritium Extraction Facility inside a heavily shielded cask. Upon arrival at a waste silo designed to receive the overpack, it is removed from the shielding cask by remote means and placed in the long-term storage silo. This paper provides an overview of the SMO overpack design and its operation.

scholarly journals Modeling Long-Term Plutonium Transport in the Savannah River Site Vadose Zone

2007 ◽  
Vol 6 (2) ◽  
pp. 344-353 ◽  
Author(s):  
Deniz I. Demirkanli ◽  
Fred J. Molz ◽  
Daniel I. Kaplan ◽  
Robert A. Fjeld ◽  
Steven M. Serkiz
Keyword(s):  
Vadose Zone ◽  
Savannah River Site ◽  
Savannah River ◽  
River Site

scholarly journals Thermal Analysis of a 9975 Package in a Facility Fire Accident

2011 ◽  
Author(s):  
Narendra K. Gupta
Keyword(s):  
Department Of Energy ◽  
Savannah River ◽  
Multiple Sources ◽  
Regulatory Requirements ◽  
Bearing Materials ◽  
Hypothetical Accident ◽  
Design Requirements ◽  
Accident Conditions ◽  
River Site

Surplus plutonium bearing materials in the U.S. Department of Energy (DOE) complex are stored in the 3013 containers that are designed to meet the requirements of the DOE standard DOE-STD-3013. The 3013 containers are in turn packaged inside 9975 packages that are designed to meet the NRC 10 CFR Part 71 regulatory requirements for transporting the Type B fissile materials across the DOE complex. The design requirements for the hypothetical accident conditions (HAC) involving a fire are given in 10 CFR 71.73. The 9975 packages are stored at the DOE Savannah River Site in the K-Area Material Storage (KAMS) facility for long term of up to 50 years. The design requirements for safe storage in KAMS facility containing multiple sources of combustible materials are far more challenging than the HAC requirements in 10 CFR 71.73. While the 10 CFR 71.73 postulates an HAC fire of 1475°F and 30 minutes duration, the facility fire calls for a fire of 1500°F and 86 minutes duration. This paper describes a methodology and the analysis results that meet the design limits of the 9975 components and demonstrate the robustness of the 9975 package.

scholarly journals Thermal Analysis of the SAFKEG Package for Long Term Storage

2005 ◽  
Author(s):  
Narendra K. (Nick) Gupta
Keyword(s):  
Thermal Analyses ◽  
Storage Conditions ◽  
Savannah River ◽  
Rocky Flats ◽  
Long Term Storage ◽  
Reactor Building ◽  
River Site ◽  
Plutonium Metal ◽  
Term Storage

Interim plutonium storage for up to 10 years in the K-reactor building is currently being planned at Savannah River Site (SRS). SAFKEG package could be used to store Pu metal and oxide (PuO2) in the K-reactor complex with other packagings like 9975. The SAFKEG is designed for carrying Type-B materials across the DOE complex and meets the 10CFR71 requirements. Thermal analyses were performed to ensure that the temperatures of the SAFKEG components will not exceed their temperature limits under the K-reactor storage conditions. Thermal analyses of the SAFKEG packaging with three content configurations using BNFL 3013 outer container (Rocky Flats, SRS bagless transfer cans, and BNFL inner containers) were performed for storage of PuO2 and plutonium metal.

Chemical Interactions of UOP IONSIV IE-911 (CST) with SRS Waste Simulants

2002 ◽  
Vol 713 ◽  
Author(s):  
May Nyman ◽  
James L. Krumhansl ◽  
Carlos Jove-Colon ◽  
Pengchu Zhang ◽  
Tina M. Nenoff ◽  
...  
Keyword(s):  
Savannah River Site ◽  
High Ionic Strength ◽  
Savannah River ◽  
Waste Solution ◽  
Apparent Decrease ◽  
A Minor ◽  
River Site ◽  
Kinetics Of ◽  
Oxide Precipitate

ABSTRACTIE-911 is a bound form of cystalline silicotitanate (CST) that was extensively tested for removing 137Cs from the Savannah River Site (SRS) tank wastes. In some simulant tests, column plugging incidents were observed, which led to thorough investigations to determine the causes and to develop protocols to avoid future plugging incidents. A related problem was the apparent decrease in Cs scavenging capability in some long-term tests. Our studies revealed that the interaction of IE-911 with the highly basic, high ionic strength, SRS average salt simulant could result in precipitation of; 1) poorly crystalline Nb-oxide, or 2) aluminosilicate zeolitic phases. The source for the Nb-oxide precipitate was determined to be a minor impurity phase that is a byproduct of CST manufacturing. The mechanisms of dissolution and re-precipitation of this phase in column pretreatment solution were investigated, and a protocol to rid IE-911 of this impurity was devised. The source material for the aluminosilicate zeolite precipitate was determined to be predominantly from the waste solution rather than the IE-911. Solubility experiments coupled with a thermodynamic analysis provided a protocol to predict when aluminosilicate precipitation will and will not occur. Finally, it was also established that aluminosilicate precipitation on the surfaces of the IE-911 granules could also account for an apparent decrease in equilibrium Kd and decrease in kinetics of Cs sorption.

Sign in / Sign up

Export Citation Format

Share Document