scholarly journals TREATMENT OF RADIOACTIVE WASTE STREAMS AT THE IDAHO CHEMICAL PROCESSING PLANT BY EVAPORATION AND ION EXCHANGE.

1971 ◽  
Author(s):  
D.W. Rhodes ◽  
G.E. Lohse
Keyword(s):  
Ion Exchange ◽  
Radioactive Waste ◽  
Processing Plant ◽  
Chemical Processing ◽  
Waste Streams

Characteristics of Preliminary Waste Forms for Icpp Low Activity Waste (Law) Fractions after Radionuclide Separations

1995 ◽  
Vol 412 ◽  
Author(s):  
Krishna Vinjamuri
Keyword(s):  
Structural Parameters ◽  
Chemical Durability ◽  
Liquid Sodium ◽  
Processing Plant ◽  
Chemical Processing ◽  
Waste Streams ◽  
Standard Glass ◽  
Radioactive Sodium ◽  
High Level ◽  
Low Activity

AbstractCurrently, at the Idaho Chemical Processing Plant (ICPP) there are about 6800 m3 of liquid sodium-bearing and liquid high-level wastes (HLW), and 3800 m3 of solid calcined HLW. One of the waste processing options under consideration includes separation of the HLW into high activity and low activity (LAW) wastes, followed by immobilization. Preliminary glasses were synthesized for the sodium-bearing, alumina-bearing, and the zirconia-bearing LAW fractions after radionuclide separations. The glasses were formed by crucible melting of a mixture of reagent chemicals representative of the LAW waste streams and frit additives at 1200 °C for 5 hours, followed by overnight annealing at 550 °C and furnace cooling of the melt. These glasses were characterized for density, elastic property, viscosity, chemical durability, structural parameters, and glass phase separation. The results are compared with that of the Hanford's standard glass ARM-i, Savannah River's benchmark glass EA, and the ICPP's grout waste form prepared using the simulated non-radioactive sodium-bearing waste fraction.

scholarly journals Zirconia Solubility in Boroaluminosilicate Glass

1995 ◽  
Vol 412 ◽  
Author(s):  
S. V. Raman ◽  
R. Bopp ◽  
T. A. Batcheller ◽  
Q. Yan
Keyword(s):  
Glass Composition ◽  
Glass Network ◽  
Processing Plant ◽  
Chemical Processing ◽  
Transition Elements ◽  
Waste Streams ◽  
Sintered Ceramic ◽  
Raman Spectroscopic ◽  
Waste Load ◽  
Composition Structure

AbstractIn the Idaho Chemical Processing Plant (ICPP) waste streams, zirconia is often the waste load limiting species. It modifies the glass network, enhances durability, increases viscosity and induces crystallization. The limits of its dissolution in boroaluminosilicate glass, with magnesia and soda additions were experimentally determined. A ternary compositional surface is evolved to present the isothermal regimes of liquid, liquid+zircon, liquid+forsterite, and liquid phase sintered ceramic. The potential of partitioning the transuranics, transition elements and solutes in these regimes is discussed. The visible Raman spectroscopic results are presented to elucidate the dependence among glass composition, structure and chemical durability.

Modelling Approach to LILW-SL Repository Safety Evaluation for Different Waste Packing Options

2007 ◽  
Author(s):  
Janez Perko ◽  
Dirk Mallants ◽  
Geert Volckaert ◽  
George Towler ◽  
Mike Egan ◽  
...  
Keyword(s):  
Ion Exchange ◽  
Radioactive Waste ◽  
Work Performance ◽  
Near Field ◽  
Additional Treatment ◽  
Ion Exchange Resins ◽  
Waste Streams ◽  
Near Surface ◽  
High Integrity ◽  
Exchange Resins

The key objective of the work described here was to support the identification of a preferred disposal concept and packaging option for low and short-lived intermediate level waste (LILW-SL). The emphasis of the assessment, conducted on behalf of the Slovenian radioactive waste management agency (ARAO), was the consideration of several waste treatment and packaging options in an attempt to identify optimised containment characteristics that would result in safe disposal, taking into account the cost-benefit of alternative safety measures. Waste streams for which alternative treatment and packaging solutions were developed and evaluated include decommissioning waste and NPP operational wastes, including drums with unconditioned ion exchange resins in overpacked tube type containers (TTCs). For decommissioning wastes, the disposal options under consideration were either direct disposal of loose pieces grouted into a vault or use of high integrity containers (HIC). In relation to operational wastes, three main options were foreseen. The first is overpacking of resin containing TTCs grouted into high integrity containers, the second option is complete treatment with hydration, neutralization, and cementation of the dry resins into drums grouted into high integrity containers and the third is direct disposal of TTCs into high integrity containers without additional treatment. The long-term safety of radioactive waste repositories is usually demonstrated with the support of a safety assessment. This normally includes modelling of radionuclide release from a multi-barrier near-surface or deep repository to the geosphere and biosphere. For the current work, performance assessment models were developed for each combination of siting option, repository design and waste packaging option. Modelling of releases from the engineered containment system (the ‘near-field’) was undertaken using the AMBER code [1]. Detailed unsaturated water flow modelling was undertaken using the HYDRUS code [2], where the degree of engineered barrier degradation with time is accounted for in each packaging option. Water fluxes relating to each degradation level were then incorporated into the AMBER models for further radionuclide transport calculations appropriate to each packing solution. The approach proved to be highly flexible, transparent and effective in terms of calculation time. Results demonstrate that all waste streams could be accepted at the preferred site with the surface repository option, under the condition that all decommissioning waste would be grouted into high integrity containers. The use of high integrity containers is also recommended for all other waste streams. Results from the detailed analysis further showed that in-drum-dried ion exchange resins in TTCs would be acceptable when grouted into high integrity containers, thereby avoiding the need for complicated processing and repackaging.

scholarly journals Hot Isostatic Press (Hip) Vitrification of Radwaste Concretes

1995 ◽  
Vol 412 ◽  
Author(s):  
Darryl D. Siemer ◽  
Barry E. Scheetz ◽  
Mary Lou Gougar
Keyword(s):  
Potential Application ◽  
Processing Plant ◽  
Chemical Processing ◽  
Waste Forms ◽  
Good Productivity

AbstractProperly prepared cementitious waste forms can be hot-isostatically-pressed into materials that exhibit performance equivalent to typical radwaste-type glasses. The HIPing conditions (temperature/pressure) required to “vitrify” these concretes are quite mild and therefore consistent with both safety and good productivity. This paper describes both the process and its products with reference to potential application to Idaho Chemical Processing Plant (ICPP) reprocessing wastes.

Sign in / Sign up

Export Citation Format

Share Document