Idaho Chemical Processing Plant Research Articles

Power profile analysis of criticality accidents involving fissile solution boiling with considering evaporation

ABSTRACT The total fission energy released in a criticality accident involving fissile solution boiling tends to be high because the relatively high fission power continues during boiling. Simulating fission power change correctly during boiling seems essential to estimate the total fission energy. Fission power during boiling changes depending on fissile concentration and volume as the solution evaporates. In this study, we investigated the effect of concentration and volume change on estimated total fission energy for a long time of boiling. We introduced a model calculating the evaporation of fissile solution into the modified quasi-steady-state method to simulate power change during boiling. Three CRAC experiments and the Idaho Chemical Processing Plant (ICPP) criticality accident in 1959 were analyzed. As a result, the calculated energy considering concentration and volume change during boiling reproduced the measured energy well.

Carbon-14 content in tree and soil samples at the Idaho National Laboratory nuclear site

Idaho National Laboratory (INL) is a nuclear research facility located in southeastern Idaho, USA; over the course of its operational history, INL has operated 52 reactors and 1 reactor fuel reprocessing facility. To determine the extent to which previous nuclear operations at INL have impacted local environmental carbon-14 (14C) concentrations, tree and soil samples from the INL desert and surrounding areas were collected, combusted, and analyzed by Accelerator Mass Spectrometry (AMS). Transport models of the plumes from the Advanced Test Reactor Facility (ATR) and the Idaho Chemical Processing Plant (ICPP) suggest the historic annual integrated plume distributions from each of these source terms was most likely in the northeast-southwest direction, with very little ground contact from stack emissions in the immediate vicinity of the facilities and maximum estimated ground contact arising ∼400–1000 m to the northeast/southwest from each facility. 14C data from annual growth ring data from trees located immediately adjacent to the ATR, INL’s Central Facilities Area (CFA), and Mud Lake, Idaho (ML) are in statistical agreement with regional nuclear weapons testing fallout backgrounds. Surface soil samples taken near a low level radioactive disposal facility and downwind from the ICPP show percent modern carbon (pMC) values ranging from 28 ± 2 to 92 ± 4, suggesting a mixture of aged and modern background carbon containing materials. Taken together, these data suggest that 14C in the INL region is predominantly derived from a mixture of aged and modern background sources (e.g. nuclear weapons testing fallout), with insignificant contributions from INL source terms.

Stainless steel/glass–ceramic interactions under hot isostatic pressing (HIPing) conditions

The interactions between stainless steel (SS) and a glass–ceramic designed for immobilisation of the high-level nuclear waste generated at the Idaho chemical processing plant (ICPP) under hot isostatic pressing (HIPing) conditions (100 MPa in argon at 1200 °C) have been studied and subsequently the effect of such interactions on the chemical durability of the glass–ceramic waste form has been examined. The diffusion of Cr from SS (Cr depletion in SS) through the interaction layer and formation of crystalline Cr/Al oxides in glass dominate the overall interaction process. It appears that the depletion of Cr in SS may reduce the potential of SS as a barrier. However, such interactions have no significant impact on the glass–ceramic and the presence of the interaction layer does not seem to have any detrimental effect on the chemical durability of the glass–ceramic as a waste form.

Crossflow filtration testing of INEEL radioactive and non-radioactive waste slurries

Development of waste treatment processes for the remediation of radioactive wastes is currently under way at the Idaho Nuclear Technology and Engineering Center (INTEC), located at the Idaho National Engineering and Environmental Laboratory (INEEL). INTEC, formerly known as the Idaho Chemical Processing Plant, previously reprocessed nuclear fuel to retrieve fissionable uranium. Liquid waste raffinates resulting from reprocessing were solidified into a calcine material. Waste treatment processes currently being considered include the dissolution of the solidified calcine material and separation of residual undissolved solids (UDS). UDS in solution must be removed prior to downstream processes such as solvent extraction and ion exchange. Filtration experiments were conducted at the INEEL using a crossflow filter apparatus on radioactive and non-radioactive waste slurries [N.R. Mann, T.A. Todd, Evaluation and Testing of the Cells Units Crossflow Filter on INEEL Dissolved Calcine Slurries, INEEL/EXT-98-00749, Idaho National Engineering and Environmental Laboratory, Idaho Falls, ID, 1998]. The purpose of this testing was to evaluate the removal and operational efficiency of crossflow filtration on slurries of various solids loadings. The solids loadings tested were 0.19, 2.44 and 7.94 wt.%, respectively. A matrix of test patterns was used to determine the effects of transmembrane pressure and axial velocity on filtrate flux. Filtrate flux rates for each solids loading displayed a high dependence on transmembrane pressure, indicating that pressure filtration resistance limits filtrate flux. Filtrate flux rates for all solids loading displayed a negative dependency on axial velocity. This would suggest axial velocities tested were efficient at removing filter cake. Prior to testing of actual waste slurries, baseline water runs were performed. Filtrate flowrates observed during baseline water runs exhibited substantial decreases despite numerous backpulses and rinses, suggesting particles that were deeply embedded within the filter membrane as the result of shear-induced deagglomeration

Glassform-Version 1: A Spreadsheet or an Algorithm for Generating Preliminary Glass Formulations for Waste Streams

AbstractGLASSFORM-Version 1, a spreadsheet or an algorithm software package, was developed for generating preliminary glass formulations for waste streams. A spreadsheet program on Microsoft Excel 97 for Windows 95 was also developed for predicting key glass properties for laboratory scale vitrification studies of simulated waste streams. The algorithm version based on Modula 2, which can be run on Windows-95 or NT, was developed to plot the glass component ratios versus the waste loading. At this time the algorithm was not developed for predicting the key glass properties. The spreadsheet or the algorithm version is an effective tool for developing preliminary glasses with a potential to be chemically durable, dense for volume reduction, of low viscosity for glass pouring, redox controlled, and resistant to corrosion of melter components such as Inconel 690. For a given solid waste stream oxide composition in wt.% or in grams or in ppm and waste loading in wt.%, the spreadsheet or the algorithm calculates glass component ratios that provide an empirical indication of the quality of a candidate glass. In addition to the component ratios for glass quality indicators, glass property composition relationships for glass durability and processability were incorporated into the spreadsheet version. These spreadsheet or algorithm versions can also be used for studying the effects of such actions as varying waste loadings or of varying individual waste components on glass properties and glass component ratios, provided specific glass property models are incorporated. As an example, the software was applied to candidate phosphate-containing glasses at the Idaho Nuclear Technology and Engineering Center (INTEC, previously Idaho Chemical Processing Plant) and also the borosilicate glasses such as the Savannah River Laboratory Environmental Assessment (EA) glass, the Hanford standard (ARM-1) glass, and the French R7T7 glass. Application of the software to new waste streams such as at Hanford, Fernald, and other countries is the most useful aspect of the GLASSFORM software during the early stages of glass waste form development.

99Tc, 236U, and 237Np in the Snake River Plain Aquifer at the Idaho National Engineering and Environmental Laboratory, Idaho Falls, Idaho

The Idaho National Engineering and Environmental Laboratory (INEEL) is located on the eastern Snake River Plain in southeastern Idaho; it is a multipurpose complex operated by the U.S. Department of Energy. Among its installations is the Idaho Chemical Processing Plant (ICPP), a facility designed principally to recover highly enriched uranium (≥93% 235U) from different nuclear fuel types used in naval propulsion, research, and test reactors. Starting in 1952 and continuing until 1984, low-level radioactive waste was discharged from the ICPP directly to the Snake River Plain aquifer by means of an injection well and seepage ponds. Over time, a suite of radionuclides have been measured in the aquifer including 3H, 36Cl, 90Sr, 137Cs, 129I, and Pu isotopes. Reported here are the first measurements of the long-lived radionuclides 99Tc, 236U, and 237Np in the aquifer and their downgradient concentration changes during water transport through fractured basalt. Like 36Cl, 99Tc behaves conservatively during transp...

Mercury Extraction By The TRUEX Process Solvent: III. Extractable Species And Stoichiometry†

ABSTRACT Mercury extraction from acidic aqueous solutions by the TRUEX process solvent, 0.2 M n-octyl(phenyl)N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO), 1.4 M tributylphosphate (TBP) in n-dodecane, has not been extensively examined, Research is currently in progress at the Idaho Chemical Processing Plant to evaluate the TRUEX process for actinide removal from several acidic waste streams, including liquid sodium-bearing waste (SBW), which contains significant quantities of mercury. Reactions for two mercury species, HgCl2 and HgCl4 −2, are reported. Classical slope analysis techniques were utilized to evaluate the stoichiometric coefficients of each Hg species independently for both CMPO and TBP. The slope analysis results indicate that even the HgCl4 −2 species extracts as HgCl2. The results also indicate that 1.9 to 2.2 moles of CMPO and 1.5 to 1.81 moles of TBP are required per mole of HgCl2 extracted. A generic equation for mercury extraction into CMPO or TBP has been shown to be: where x = 2 or 4, y = 0 or −2, E = CMPO or TBP, a = the experimentally determined CMPO or TBP stoichiometry, and over lines indicate organic species. Equilibrium constants for HgCl2 species were determined to be 295 when extracted by CMPO and 18.2 when extracted by TBP. Equilibrium constants for the HgCl4 −2 species were found to range between 5.8 and 12.0 for the CMPO reaction and 0.13 to 0.27 with TBP.

AMS measurement of environmental U-236 Preliminary results and perspectives

236U is a 23.4 Ma half-life radioactive isotope whose natural global abundance is expected to be overwhelmed in many places by its production and release during nuclear power generation. The detection of excess 236U in the environment is therefore a method of monitoring the presence of activated uranium. However, 236U has not been widely used for environmental studies due to the difficulties of measuring it at low levels with standard α-spectrometry or thermal ionization mass spectrometry. In principle, accelerator mass spectrometry can overcome some of the difficulties encountered in the traditional methods, making it possible to detect 236U at very low levels so that a large range of environmental samples can be analyzed. In this paper, we report some preliminary results of the 236U measurements of four water samples from the US Idaho Chemical Processing Plant. Limitations with our current AMS system and the perspectives for considerable improvements will be discussed.

Development and Testing of Srex Flowsheets for Treatment of Idaho Chemical Processing Plant Sodium-Bearing Waste Using Centrifugal Contactors

Laboratory experimentation has indicated that the SREX process is effective for partitioning 90Sr from acidic radioactive waste solutions located at the Idaho Chemical Processing Plant. A baseline flowsheet has been proposed for the treatment of sodiumbearing waste (SBW) which includes extraction of strontium from liquid SBW into the SREX solvent (0.15 M 4′,4′ (5′)-di-(tert-butyldicyclohexo)-18-crown-6 and 1.2 M TBP in Isopar L®), a 0.01 M nitric acid strip section to back-extract components from the loaded solvent, and a 2.0 M HNO3 solvent acidification section to equilibrate the solvent with HNO3 prior to recycle to the extraction section. The flowsheet was designed to provide a decontamination factor (DF) of >103 which will reduce the 90Sr activity in the waste solution to below the NRC Class A LLW limit of 0.04 Ci 90Sr/m3. SREX flowsheet testing was performed using sixteen stages of 5.5-cm diameter centrifugal contactors. The behavior of stable Sr and other components which are potentially extracted by the SREX solvent were evaluated. Specifically, the behavior of the matrix components including Pb, K, Hg, Na, Ca, Zr, and Fe was studied. The described flowsheet achieved 99.98% Sr removal (DF=4250) with one cycle of SREX. Potassium and Zr were partially extracted into the SREX solvent with 35% and 21%, respectively, exiting in the strip product. Sodium, Ca, and Fe were essentially inextractable. Lead was determined to extract and accumulate in the SREX solvent and in the strip section. As a result, a Pb precipitate formed in the strip stages of the contactors. Mercury was also determined to extract and accumulate in the SREX solvent.

Geochemistry and stratigraphic correlation of basalt lavas beneath the Idaho Chemical Processing Plant, Idaho National Engineering Laboratory

Thirty-nine samples of basaltic core were collected from wells 121 and 123, located approximately 1.8 km apart north and south of the Idaho Chemical Processing Plant at the Idaho National Engineering Laboratory. Samples were collected from depths ranging from 15 to 221 m below land surface for the purpose of establishing stratigraphic correlations between these two wells. Elemental analyses indicate that the basalts consist of three principal chemical types. Two of these types are each represented by a single basalt flow in each well. The third chemical type is represented by many basalt flows and includes a broad range of chemical compositions that is distinguished from the other two types. Basalt flows within the third type were identified by hierarchical K-cluster analysis of 14 representative elements: Fe, Ca, K, Na, Sc, Co, La, Ce, Sm, Eu, Yb, Hf, Ta, and Th. Cluster analyses indicate correlations of basalt flows between wells 121 and 123 at depths of approximately 38–40 m, 125–128 m, 131–137 m, 149–158 m, and 183–198 m. Probable correlations also are indicated for at least seven other depth intervals. Basalt flows in several depth intervals do not correlate on the basis of chemical compositions, thus reflecting possible flow margins in the sequence between the wells. Multi-element chemical data provide a useful method for determining stratigraphic correlations of basalt in the upper 1–2 km of the eastern Snake River Plain.

Cesium Absorption from Acidic Solutions Using Ammonium Molybdophosphate on a Polyacrylonitrile Support (AMP-PAN)

Recent effort at the Idaho Chemical Processing Plant (ICPP) have included evaluation of cesium removal technologies as applied to ICPP acidic radioactive waste streams. Ammonium molybdophosphate (AMP) immobilized on a polyacrylonitrile support (AMP-PAN) has been studied as an ion exchange agent for cesium removal from acidic waste solutions. Capacities, distribution coefficients, elutability, and kinetics of cesium extraction have been evaluated. Exchange breakthrough curves using small columns have been determined from 1M HNO3 and simulated waste solutions. The theoretical capacity of AMP is 213 g Cs/Kg AMP. The average experimental capacity in batch contacts with various acidic solutions was 150 g Cs/kg AMP. The measured cesium distribution coefficients from actual waste solutions were 3287 mL/g for dissolved zirconia calcines, and 2679 mL/g for sodium-bearing waste. The cesium in the dissolved alumina calcines was analyzed for; however, the concentration was below analytical detectable limits resulting in inconclusive results. The reaction kinetics are very rapid (2-10 minutes). Cesium absorption appears to be independent of acid concentration over the range tested (0.1 M to 5 M HNO3).

Evaluation of the SREX Solvent Extraction Process for the Removal Of 90Sr and Hazardous Metals from Acidic Nuclear Waste Solutions Containing High Concentrations of Interfering Alkali Metal Ions

The SREX process, which was developed at Argonne National Laboratories, has been evaluated for its effectiveness for the decontamination of radioactive liquid waste at the Idaho Chemical Processing Plant located at the Idaho National Engineering Laboratory. The extraction solvent consists of 0.15 M 4′,4′,(5′)-di-(t-butyl-dicyclohexo)-18-crown-6 in either 1.2 M TBP/Isopar L®. Suppressed extraction due to matrix interferences limits DSr values to the range of 3-4 in batch extraction experiments. The SREX solvent has been shown to be effective in the removal of non-radioactive Pb from liquid wastes, in addition to the extraction of 90Sr. The information from this study has been used to develop a proposed flowsheet for the treatment of liquid waste in centrifugal contactors.

Glass-Ceramic Waste Forms for Immobilizing Plutonium

ABSTRACTResults are reported on several new glass and glass-ceramic waste formulations for plutonium disposition. The approach proposed involves employing existing calcined high level waste (HLW) present at the Idaho Chemical Processing Plant (ICPP) as an additive to: 1) aid in the formation of a durable waste form and 2) decrease the attractiveness level of the plutonium from a proliferation viewpoint. The plutonium, PuO2, loadings employed were 15 wt% (glass) and 17 wt% (glass-ceramic). Results in the form of x-ray diffraction patterns, microstructure and durability tests are presented on cerium surrogate and plutonium loaded waste forms using simulated calcined HLW and demonstrate that durable phases, zirconia and zirconolite, contain essentially all the plutonium.

Thermodynamic Modeling of the Extraction of Cadmium Nitrate by the TRUEX-NPH Solvent

Abstract The TRUEX solvent extraction process is being developed to remove low concentrations of transuranic (TRU) elements from high-level and TRU wastes currently stored at US Department of Energy sites. Cadmium is an important constituent of some of these wastes, especially at the Idaho Chemical Processing Plant. Cadmium distribution ratios were measured by using a radioactive Cd tracer under a variety of aqueous-phase compositions to elucidate the mechanism of cadmium extraction by TRUEX-NPH solvent at 25°C. A model for predicting cadmium distribution ratios was developed as a function of the thermodynamic activities of water and nitrate ion. The Generic TRUEX Model (GTM) was used to calculate these activities. The GTMs calculation of activity coefficients is based on a Bromley method.

EFFECT OF THE INTERFERENCE OF ALKALI AND ALKALINE EARTH METAL IONS ON THE EXTRACTION OF90SR FROM ACIDIC NUCLEAR WASTE SOLUTIONS BY 18-CROWN-6 DERIVATIVES

ABSTRACT The extraction of 90Sr from liquid sodium-bearing wastes (SBW) at the Idaho Chemical Processing Plant (ICPP) by 4’, 4’(5’)-di-(t-butyldicyclohexo)-18-crown-6 (DtBuCH18C6) and cis-dicyclohexano-l8-crown-6 (DCH18C6) in 1-octanol has been investigated. It has been shown mat the presence of group I and group II metal ions which are co-extracted by the organic solvent limits the extraction of Sr2+ from nitric acid solutions. For example, extraction experiments which were performed with 0.15 M DtBuCH18C6 in 1-octanol yielded DSr values which ranged from -20 in 3M nitric acid to -5 in a solution of 3M NaNO3 and 3M nitric acid. Furthermore, it has been shown that when the concentration of the Na+ is ≥ l.5M, an increase in the total concentration of NO3 ion has little effect on the value of Dsr The value of DSr has been studied as a function of the concentration of Li+, K+, Na+, Cs+, and Ca2+. Additionally, a comparative study of commercially available 18-crown-6 derivatives has been performed. It has bee...

MERCURY EXTRACTION BY THE TRUEX PROCESS SOLVENT. I. KINETICS, EXTRACTABLE SPECIES, & DEPENDENCE ON NITRIC ACID CONCENTRATION1

ABSTRACT Mercury extraction from acidic aqueous solutions by the TRUEX process solvent (0.2 M CMPO - 1.4 M TBP in n-dodecane) has not extensively been examined. Research at the Idaho Chemical Processing Plant (ICPP) is currently being conducted to evaluate the TRUEX process for actinide removal from several acidic waste streams, including liquid sodium-bearing waste (SBW), which contains significant quantities of mercury. Preliminary experiments were performed involving the extraction of Hg203, added as HgCl2, from 0.01 to 10 M HNO3 solutions. Mercury distribution coefficients (DHg) range between 3 and 60 from 0.01 M to 2 M HNO3. At higher nitric acid concentrations, i.e. 5 M HNO3 or greater, DHg significantly decreases to values less than 1. These results indicate mercury is extracted from acidic solutions ˜5 M HNO3. Experimental results indicate the extractable species is HgCl2 from nitrate media, i.e., chloride must be present in the nitrate feed to e...

MERCURY EXTRACTION BY THE TRUEX PROCESS SOLVENT. II. SELECTIVE PARTITIONING OF MERCURY FROMCO-EXTRACTED ACTINIDES IN A SIMULATED ACIDIC ICPP WASTE STREAM1

ABSTRACT The TRUEX process is being evaluated at the Idaho Chemical Processing Plant (ICPP) as a means to partition the actinides from acidic sodium-bearing waste (SBW). The mercury content of this waste averages 1 g/1. Because the chemistry of mercury has not been extensively evaluated in the TRUEX process, mercury was singled out as an element of interest. Radioactive mercury, 203Hg, was spiked into a simulated solution of SBW containing 1 g/1 mercury. Successive extraction batch contacts with the mercury spiked waste and successive scrubbing and stripping batch contacts of the mercury loaded TRUEX solvent (0.2 M CMPO-1.4 M TBP in dodecane) show that mercury will extract into and strip from the solvent. The extraction distribution coefficient for mercury, as HgCl2, from SBW having a nitric acid concentration of 1.4 M and a chloride concentration of 0.035 M was found to be 3. The stripping distribution coefficient was found to be 0.5 with 5 M HN03M and 0.077 with 0.25 M Na2C03. Because experiments described here show that mercury can be extracted from SBW and stripped from the solvent, a process has been developed to partition mercury from the actinides in SBW.

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

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.

Zirconia Solubility in Boroaluminosilicate Glass

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.

Hot Isostatic Press (Hip) Vitrification of Radwaste Concretes

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.