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All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Materials Data on P2H12N8O3 by Materials Project

P2N7H12NO3 crystallizes in the triclinic P-1 space group. The structure is zero-dimensional and consists of two nitric acid molecules and two P2N7H12 clusters. In each P2N7H12 cluster, there are two inequivalent P5+ sites. In the first P5+ site, P5+ is bonded to four N2- atoms to form corner-sharing PN4 tetrahedra. There are a spread of P–N bond distances ranging from 1.59–1.66 Å. In the second P5+ site, P5+ is bonded to four N2- atoms to form corner-sharing PN4 tetrahedra. There are a spread of P–N bond distances ranging from 1.59–1.66 Å. There are seven inequivalent N2- sites. In the first N2- site, N2- is bonded in a distorted trigonal non-coplanar geometry to one P5+ and two H1+ atoms. There is one shorter (1.02 Å) and one longer (1.03 Å) N–H bond length. In the second N2- site, N2- is bonded in a distorted trigonal non-coplanar geometry to one P5+ and two H1+ atoms. Both N–H bond lengths are 1.02 Å. In the third N2- site, N2- is bonded in a distorted trigonal non-coplanar geometry to one P5+ and two H1+ atoms. There is one shorter (1.02 Å) and one longer (1.03 Å) N–H bond length. In the fourth N2- site, N2- is bonded in a distorted trigonal non-coplanar geometry to one P5+ and two H1+ atoms. There is one shorter (1.02 Å) and one longer (1.03 Å) N–H bond length. In the fifth N2- site, N2- is bonded in a distorted trigonal non-coplanar geometry to one P5+ and two H1+ atoms. There is one shorter (1.02 Å) and one longer (1.03 Å) N–H bond length. In the sixth N2- site, N2- is bonded in a distorted trigonal planar geometry to one P5+ and two H1+ atoms. Both N–H bond lengths are 1.02 Å. In the seventh N2- site, N2- is bonded in a bent 120 degrees geometry to two P5+ atoms. There are twelve inequivalent H1+ sites. In the first H1+ site, H1+ is bonded in a single-bond geometry to one N2- atom. In the second H1+ site, H1+ is bonded in a single-bond geometry to one N2- atom. In the third H1+ site, H1+ is bonded in a single-bond geometry to one N2- atom. In the fourth H1+ site, H1+ is bonded in a single-bond geometry to one N2- atom. In the fifth H1+ site, H1+ is bonded in a single-bond geometry to one N2- atom. In the sixth H1+ site, H1+ is bonded in a single-bond geometry to one N2- atom. In the seventh H1+ site, H1+ is bonded in a single-bond geometry to one N2- atom. In the eighth H1+ site, H1+ is bonded in a single-bond geometry to one N2- atom. In the ninth H1+ site, H1+ is bonded in a single-bond geometry to one N2- atom. In the tenth H1+ site, H1+ is bonded in a single-bond geometry to one N2- atom. In the eleventh H1+ site, H1+ is bonded in a single-bond geometry to one N2- atom. In the twelfth H1+ site, H1+ is bonded in a single-bond geometry to one N2- atom.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on Al(NO6)3 by Materials Project

AlO8NO3N2O5O2 crystallizes in the monoclinic P2_1/c space group. The structure is zero-dimensional and consists of four hydrogen peroxide molecules, four nitric acid molecules, four nitrogen pentoxide molecules, and four AlO8 clusters. In each AlO8 cluster, Al is bonded in a tetrahedral geometry to four O atoms. There are a spread of Al–O bond distances ranging from 1.72–1.89 Å. There are eight inequivalent O sites. In the first O site, O is bonded in a water-like geometry to two O atoms. There is one shorter (1.24 Å) and one longer (1.61 Å) O–O bond length. In the second O site, O is bonded in a water-like geometry to two O atoms. There is one shorter (1.31 Å) and one longer (1.72 Å) O–O bond length. In the third O site, O is bonded in a distorted water-like geometry to one Al and one O atom. The O–O bond length is 1.31 Å. In the fourth O site, O is bonded in a single-bond geometry to one O atom. In the fifth O site, O is bonded in a distorted single-bond geometry to one Al and one O atom. In the sixth O site, O is bonded in a water-like geometry to two O atoms. In the seventh O site, O is bonded in a single-bond geometry to one Al atom. In the eighth O site, O is bonded in a distorted single-bond geometry to one Al and one O atom.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on AgH36S8N11O18 by Materials Project

Ag(S2O3)4(NH4)9(NO3)2 crystallizes in the tetragonal I-42d space group. The structure is zero-dimensional and consists of thirty-six ammonium molecules, eight nitric acid molecules, and four Ag(S2O3)4 clusters. In each Ag(S2O3)4 cluster, Ag1+ is bonded in a tetrahedral geometry to four equivalent S2- atoms. All Ag–S bond lengths are 2.59 Å. There are two inequivalent S2- sites. In the first S2- site, S2- is bonded in a trigonal non-coplanar geometry to one S2- and three O2- atoms. The S–S bond length is 2.04 Å. All S–O bond lengths are 1.49 Å. In the second S2- site, S2- is bonded in a distorted water-like geometry to one Ag1+ and one S2- atom. There are three inequivalent O2- sites. In the first O2- site, O2- is bonded in a single-bond geometry to one S2- atom. In the second O2- site, O2- is bonded in a single-bond geometry to one S2- atom. In the third O2- site, O2- is bonded in a single-bond geometry to one S2- atom.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on UH18C4N10O13 by Materials Project

UC4N8H18O7(NO3)2 crystallizes in the monoclinic P2_1/c space group. The structure is zero-dimensional and consists of eight nitric acid molecules and four UC4N8H18O7 clusters. In each UC4N8H18O7 cluster, U6+ is bonded in a distorted pentagonal bipyramidal geometry to seven O2- atoms. There are a spread of U–O bond distances ranging from 1.81–2.50 Å. There are four inequivalent C4+ sites. In the first C4+ site, C4+ is bonded in a trigonal planar geometry to two N+1.40- and one O2- atom. Both C–N bond lengths are 1.34 Å. The C–O bond length is 1.29 Å. In the second C4+ site, C4+ is bonded in a trigonal planar geometry to two N+1.40- and one O2- atom. Both C–N bond lengths are 1.34 Å. The C–O bond length is 1.28 Å. In the third C4+ site, C4+ is bonded in a trigonal planar geometry to two N+1.40- and one O2- atom. There is one shorter (1.34 Å) and one longer (1.35 Å) C–N bond length. The C–O bond length is 1.28 Å. In the fourth C4+ site, C4+ is bonded in a trigonal planar geometry to two N+1.40- and one O2- atom. Both C–N bond lengths are 1.34 Å. The C–O bond length is 1.28 Å. There are eight inequivalent N+1.40- sites. In the first N+1.40- site, N+1.40- is bonded in a trigonal planar geometry to one C4+ and two H1+ atoms. There is one shorter (1.01 Å) and one longer (1.02 Å) N–H bond length. In the second N+1.40- site, N+1.40- is bonded in a trigonal planar geometry to one C4+ and two H1+ atoms. There is one shorter (1.01 Å) and one longer (1.02 Å) N–H bond length. In the third N+1.40- site, N+1.40- is bonded in a trigonal planar geometry to one C4+ and two H1+ atoms. Both N–H bond lengths are 1.02 Å. In the fourth N+1.40- site, N+1.40- is bonded in a trigonal planar geometry to one C4+ and two H1+ atoms. Both N–H bond lengths are 1.02 Å. In the fifth N+1.40- site, N+1.40- is bonded in a trigonal planar geometry to one C4+ and two H1+ atoms. Both N–H bond lengths are 1.02 Å. In the sixth N+1.40- site, N+1.40- is bonded in a trigonal planar geometry to one C4+ and two H1+ atoms. Both N–H bond lengths are 1.02 Å. In the seventh N+1.40- site, N+1.40- is bonded in a trigonal planar geometry to one C4+ and two H1+ atoms. There is one shorter (1.01 Å) and one longer (1.02 Å) N–H bond length. In the eighth N+1.40- site, N+1.40- is bonded in a trigonal planar geometry to one C4+ and two H1+ atoms. There is one shorter (1.01 Å) and one longer (1.03 Å) N–H bond length. There are eighteen inequivalent H1+ sites. In the first H1+ site, H1+ is bonded in a single-bond geometry to one N+1.40- atom. In the second H1+ site, H1+ is bonded in a single-bond geometry to one N+1.40- atom. In the third H1+ site, H1+ is bonded in a single-bond geometry to one N+1.40- atom. In the fourth H1+ site, H1+ is bonded in a single-bond geometry to one N+1.40- atom. In the fifth H1+ site, H1+ is bonded in a single-bond geometry to one N+1.40- atom. In the sixth H1+ site, H1+ is bonded in a single-bond geometry to one N+1.40- atom. In the seventh H1+ site, H1+ is bonded in a single-bond geometry to one N+1.40- atom. In the eighth H1+ site, H1+ is bonded in a single-bond geometry to one N+1.40- atom. In the ninth H1+ site, H1+ is bonded in a single-bond geometry to one N+1.40- atom. In the tenth H1+ site, H1+ is bonded in a single-bond geometry to one N+1.40- atom. In the eleventh H1+ site, H1+ is bonded in a single-bond geometry to one N+1.40- atom. In the twelfth H1+ site, H1+ is bonded in a single-bond geometry to one N+1.40- atom. In the thirteenth H1+ site, H1+ is bonded in a single-bond geometry to one N+1.40- atom. In the fourteenth H1+ site, H1+ is bonded in a single-bond geometry to one N+1.40- atom. In the fifteenth H1+ site, H1+ is bonded in a single-bond geometry to one N+1.40- atom. In the sixteenth H1+ site, H1+ is bonded in a single-bond geometry to one N+1.40- atom. In the seventeenth H1+ site, H1+ is bonded in a single-bond geometry to one O2- atom. The H–O bond length is 0.99 Å. In the eighteenth H1+ site, H1+ is bonded in a single-bond geometry to one O2- atom. The H–O bond length is 0.98 Å. There are seven inequivalent O2- sites. In the first O2- site, O2- is bonded in a single-bond geometry to one U6+ atom. In the second O2- site, O2- is bonded in a single-bond geometry to one U6+ atom. In the third O2- site, O2- is bonded in a 2-coordinate geometry to one U6+ and one C4+ atom. In the fourth O2- site, O2- is bonded in a distorted single-bond geometry to one U6+ and one C4+ atom. In the fifth O2- site, O2- is bonded in a 1-coordinate geometry to one U6+ and one C4+ atom. In the sixth O2- site, O2- is bonded in a distorted bent 150 degrees geometry to one U6+ and one C4+ atom. In the seventh O2- site, O2- is bonded in a distorted water-like geometry to one U6+ and two H1+ atoms.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on NO3 by Materials Project

NO3 crystallizes in the monoclinic P2_1/c space group. The structure is zero-dimensional and consists of sixteen nitric acid molecules. N is bonded in a trigonal planar geometry to three O atoms. All N–O bond lengths are 1.25 Å. There are three inequivalent O sites. In the first O site, O is bonded in a single-bond geometry to one N atom. In the second O site, O is bonded in a single-bond geometry to one N atom. In the third O site, O is bonded in a single-bond geometry to one N atom.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Recovery of High Purity Rare Earth Elements (REE) from Coal Ash via a Novel Electrowinning Process (Final Report)

The goal of this project was to produce a high purity, separated (>90%) rare earth oxide (REO) product from coal-based sources. This product was achieved using the following three steps: 1. Battelle’s ADP: The ADP process involves pretreatment of ash (milling and caustic leaching), leaching of pretreated ash with nitric acid, roasting of loaded acid solution containing REE, and water leaching of residual from roasting step, resulting in a REE pregnant solution. 2. Solvent Extraction (SX) Upgrading: The SX involves the extraction of REE from a pregnant REE solution using an organic extractant (note after extraction, the aqueous phase was a residual solution and the organic phase contained REE), and stripping of REE from the organic phase product obtained from the extraction using an acid solution (the stripped solution was the final purified product containing REE, organic traces, and traces of other metals for further separation in the electrowinning process). 3. Rare Earth Salts’ (RES) Electrochemical Separation and Purification Process: The staged electrochemical process was separated REO products from the mixed REE solution. Testing by RES was also included in the investigation of options to minimize the solvent extraction steps. This report covers laboratory testing, production of a high purity coal-based REO, process design of the overall REE recovery and purification process, technoeconomic assessment (TEA), and a commercialization plan discussing the overall REE recovery and purification process.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2020

Quantitative Determination of Total Cesium in Savannah River Site (SRS) Tank Waste by Inductively Coupled Plasma Emission Spectroscropy

Utilization of the significant and unique wavelength dynamic range (165 to 1100 nm) available on the Leeman Prodigy Inductively Coupled Plasma Emission Spectroscopy (ICP-ES) allows for quantification of cesium using the emission line 894.347 nm without the potential of Barium interference in nuclear waste samples. Samples are added to a Sodium Peroxide Fusion and Lithium Carbonate buffer in 2% nitric acid solution to increase sensitivity of the easily ionizable (<6 eV) cesium for quantitation to 5 pbb. Through direct comparison of analytical data results, the method is proven to be complementary to the Inductively Coupled Plasma Mass Spectrometry (ICP-MS) analysis (LOQ 100 ppt) in support of the Department of Energy’s (DOE) Nuclear Waste Mission.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Evaluation of the lgnitability Hazards in Remediated Nitrate Salt Waste

The hazards associated with the characteristic of ignitability in remediated nitrate salt (RNS) waste containing nitric acid (HNO 3 ) and metal nitrate salts, currently stored at the Waste Control Specialists (WCS) facility near Andrews, Texas have been evaluated. For these waste drums there are two ignitability hazards that may apply, each separately distinguished from the other and uniquely correlated to one of two competing reaction mechanisms. One ignitability hazard, designated the autocatalytic thermal runway, is uniquely correlated to an acid/base reaction of hydrolysis of a hydrated metal nitrate salt. The other applicable ignitability hazard, designated the oxidizer property, is uniquely correlated to a redox disproportionation reaction involving an anhydrous metal nitrate salt. Which reaction mechanism pathway is accessible to a given metal nitrate, and therefore which ignitability hazard may apply, depends on the acidity of the metal ion and its corresponding capacity to maintain coordination with water.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Reduced Neutralization Feasibility Study for H-Canyon Accelerated Basin De- Inventory (ABD) Program

An alternative approach to Spent Nuclear Fuel (SNF) and Nuclear Material Processing was developed for future H-Canyon (HCAN) and L-Area operations that involves a paradigm shift from current HCAN, Concentrate, Storage, and Transfer Facility (CSTF), and Defense Waste Processing Facility (DWPF) operations. The alternative, referred to as Accelerated Basin De-inventory (ABD), requires that all Domestic and Foreign Research Reactor SNF currently at the Savannah River Site (SRS) will be dissolved, stored, and then transferred to CSTF without recovery of Highly Enriched Uranium (HEU). Concentrated nitric acid is utilized to dissolve aluminum spent nuclear fuel (ASNF) in HCAN. The vessels and piping in HCAN are fabricated from 304L stainless steel and are ideally suited to handle the acidic waste stream. However, as the waste is transferred to the CSTF and DWPF, it will contact the carbon steel waste tanks in CSTF. In order to prevent corrosion of the carbon steel, the acidic waste is neutralized (i.e., pH adjusted over 11) by the addition of sodium hydroxide (NaOH). The NaOH is added until the final solution contains 1.2 M excess -OH. Additionally, if the waste is not neutralized to a pH greater than 11, then aluminum hydroxide (Al(OH) 3 ) would form, and solids may form in the piping as it is transferred to CSTF. This document presents an analysis of the influence of reducing the excess caustic that is added to the neutralization tanks on the corrosion protection scheme primarily for the CSTF waste tanks. The implications to HCAN and DWPF were also assessed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Recovery and Recycle of Irradiated Low-Enriched Uranium from the Production of 99 Mo

Technetium-99m ( 99m Tc), the daughter (decay product) of molybdenum-99 ( 99 Mo), is the most commonly used medical radioisotope in the world. 99 Mo is produced primarily from the fissioning of 235 U. A potential future producer, Eden Radioisotopes, is planning to irradiate low-enriched-uranium metal-foil targets for production of 99 Mo. Their plan is to recover, purify, and recycle the irradiated LEU in a new set of targets inside their facility. This study assesses processes to (1) purify and recycle uranium following 99 Mo recovery and (2) convert the uranium to metal for subsequent foil production and target fabrication. A UREX (uranium recovery by extraction) liquid–liquid extraction flowsheet that utilizes centrifugal contactors was developed using the Argonne model for universal solvent extraction (AMUSE) to recover and purify the uranium. The calculated flowsheet predicted high decontamination from Pu and most fission products and >99.99% recovery of uranium. Suggestions for conversion of the UREX U-product (uranyl nitrate in dilute nitric acid) to U metal were provided based on literature studies and earlier laboratory studies performed at Argonne National Laboratory.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2021

Aluminum Alloy (6061-O, 5052-O, and 1100) Dissolution Rate Testing

The dissolution rates of two aluminum alloys (6061-O and 5052-O) were tested to determine which would be able to replace an aluminum alloy that was difficult to procure (6063-T6). The findings of this memo conclude that the dissolution rates of 6061-O and 5052-O are two orders of magnitude slower than 6063-T6. Therefore, it is unlikely that 6061-O and 5052-O would be adequate replacements for 6063-T6 based on similar dissolution rates in a Hg-catalyzed, nitric acid solution. Another area of interest was how temperature variation would influence the dissolution rate of an aluminum alloy (specifically 1100). This separate study found that the dissolution rate of the aluminum alloy below 86 °C was approximately 0.0014 % of the dissolution rate at 94.5°C and above. Therefore, at or below 86 °C the dissolution rate reduces significantly which translates to much longer dissolution times.

36 MATERIALS SCIENCE↗
DOE OSTI2020

CoDCon Project (Final Report)

The co-decontamination (CoDCon) project was established in FY 2016 with the objectives of (a) evaluating the uncertainty in the uranium (U)/plutonium (Pu) ratio in a mixed U/Pu product from a tributyl phosphate (TBP)–based solvent extraction flowsheet, and (b) developing and demonstrating on-line optical spectroscopy for real-time monitoring of key components (e.g., Pu, U, and HNO 3 concentrations) in the process solutions. We were interested in assessing the accuracy and precision to which a specific uranium-to-plutonium (U/Pu) ratio can be achieved, which for the purposes of this project was set at a U/Pu mass ratio of 7/3. The uncertainty associated with achieving this specific target U/Pu ratio was investigated during five flowsheet tests using laboratory-scale solvent extraction equipment. In addition, optical spectroscopic techniques were incorporated into the CoDCon solvent extraction testing system, allowing real time monitoring of all input and output process streams. Two CoDCon flowsheet tests were performed in FY 2018 using a simple dissolved fuel simulant containing only U (~1 M) and Pu (~15 mM) in nitric acid (HNO 3 ; ~3 M). In FY 2019, two additional flowsheet tests were performed. For the first of these (CoDCon Run 3), the dissolved fuel simulant was similar to that used in the first two tests, with the inclusion of 1 mM neptunium (Np). The second test conducted in FY 2019 (CoDCon Run 4) used a more representative dissolved fuel simulant, including addition of non-radioactive fission product elements. A fifth CoDCon flowsheet test (CoDCon Run 5) was conducted in FY 2020, with the following additional objectives: (1) routing of the technetium (Tc) in the simulated dissolved fuel solution to the solvent extraction raffinate, and (2) routing of the Np in the simulated dissolved fuel solution to the U/Pu product. All tests used a bank of sixteen 2 cm centrifugal contactors. The tests involved first loading the solvent (30 vol% TBP dissolved in n-dodecane) with U and Pu (and Np, for Run 5), then the Pu (and Np) was stripped from the loaded solvent with a U(IV) solution (~50 mM) and the flowsheet conditions were adjusted such that some U partitioned into the Pu-containing product stream. The amount of U accompanying the Pu was monitored in real time using optical spectroscopic techniques coupled with chemometric modeling. Based on the real-time spectroscopic measurement of the U/Pu ratio, adjustments were made to the flowrate of the fresh TBP solvent phase used to scrub U from the aqueous Pu-containing product. This proved to be a very effective way to control the U/Pu mass ratio in the product. This report presents the results of the CoDCon Run 5 test. The flowsheet tested in Run 5 was substantially different than that run in the prior tests, especially the solvent loading section of the flowsheet. Two key changes were made. First, based on the objective to extract all the Np and route it with the U/Pu product, pentavalent vanadium [V(V)] was added to the feed and scrub solutions. The purpose of the V(V) was to convert all the Np to the +6 oxidation state, which is extractable by TBP. Second, a high acid (8 M HNO 3 ) scrub was added to the flowsheet to scrub the Tc from the solvent. This was followed by a low acid scrub (0.05 M HNO 3 ) to reduce the residual HNO 3 concentration in the solvent prior to the Pu stripping step. The output from the low acid scrub was collected separately, rather than routing towards the raffinate. The modifications to the solvent loading part of the flowsheet were only partially successful. The treatment with V(V) was effective at converting the Np to Np(VI). Only 1.3% of the Np remained in the raffinate solution. However, ~40% of the Np stripped out of the solvent in the low acid scrub step; nearly 20% of the Pu also was stripped from the solvent during the low acid scrub. For further development, either modifications to the flowsheet, or concentration and recycle of the low acid stream into the

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Literature Review on Next Generation Solvent Isopar ® L Vapor Pressure Curve and the Partitioning of its Modifier and Extractant

The Next Generation Solvent (NGS) is set to replace the Original Caustic Side Solvent Extractant (CSSX) at the Salt Waste Processing Facility (SWPF). The Savannah River National Laboratory (SRNL) was requested by Savannah River Mission Completion (SRMC), formerly Savannah River Remediation (SRR), to perform a literature review on the following topics to address flammability concerns with the current solvent: Isopar ® L vapor pressure curve for NGS, partitioning ratio for the extractant MaxCalix and the modifier Cs-7SB, and high cesium concentration impacts on NGS radiolysis and potential solvent degradation rates in high cesium concentrations. The following conclusions and recommendations are made based on previous experimental work and literature: (1) Current SWPF flammable gas generation calculations use an Isopar ® L vapor pressure curve based on experimental testing with the Original CSSX solvent. No such testing to date has been performed with NGS. It is suggested that the decrease in Cs-7SB concentration for NGS compared to the Original CSSX solvent would lead to a slightly higher vapor pressure at all temperatures in SWPF vessels. A bounding NGS vapor pressure curve has been provided; it is recommended to see if these values would challenge current flammability controls and to perform testing if needed.(2) The partitioning ratio for Cs-7SB is known in the Original CSSX solvent with dilute nitric acid and caustic solutions. No tests could be found for the partitioning of Cs-7SB to dilute boric acid solutions; however, a similar partitioning ratio is expected. Due to the lipophilic alkyl chains on MaxCalix, it is expected to be even less soluble than BOBCalixC6 in the aqueous phase and should not be considered a significant contributor to the f organic term. Additionally, the reaction rate of N,N’,N’’-Tris(3,7-dimethyloctyl)guanidine (TiDG) or its degradation products with a hydrogen radical should be estimated/determined if they are found to be significant contributors to the f organic term. (3) NGS is expected to see much higher Cs concentrations at SWPF in comparison to its use at the Modular CSSX Unit (MCU). These higher Cs concentrations could influence radiolytic degradation rates of the solvent. NGS appears to be fairly stable to radiolytic degradation based on previous testing and its use at MCU. However, there has not been radiolytic flammable gas generation testing with NGS to date. There is a risk that the continued use of G-values obtained for flammable gases produced from the irradiation of the Original CSSX solvent is not bounding for NGS, but this is considered a very low risk due to the similarities in the composition of the solvents, as well as the conservatisms in the experimental design of the Original CSSX testing.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Elemental Analysis of Bio-Oils by Inductively Coupled Plasma Optical Emission Spectroscopy (ICP-OES). Laboratory Analytical Procedure (LAP), Issue Date: May 13, 2022

Concentrations of inorganic elements is a key quality metric for bio-oils as certain elements impact upgrading processes and product quality. Unless reduced or removed during production and processing, alkali and alkaline metals native to lignocellulosic biomass can carry over into bio-oils contributing to ash content and degraded catalyst performance during upgrading to hydrocarbon fuels or chemical products. Non-metallic elements such as sulfur and phosphorus can also negatively impact upgrading catalysts and product quality. Inductively coupled plasma optical emission spectroscopy (ICP-OES) can be used to measure inorganic elements of interest in bio-oils. This procedure covers the preparation and analysis of fast pyrolysis (FP) and catalytic fast pyrolysis (CFP) bio-oils. The concentrations of these elements can indicate potential bio-oil quality from the perspective of deoxygenation processes. The implications of elemental composition will depend on process parameters such as upgrading catalyst sensitivities. This Laboratory Analytical Procedure (LAP) covers two methods for quantification of inorganic elements by ICP-OES: Procedure A uses microwave assisted digestion with concentrated nitric acid, and Procedure B is an organic ICP-OES method utilizing a diglyme solvent.

09 BIOMASS FUELS↗
DOE OSTI2021

Modeling of ATR fuel in DOE Standard Canisters with Helium Backfilled Condition

Road-ready and final disposition packaging configurations for the advanced test reactor (ATR) fuel dictates storage within helium backfilled sealed DOE standard canisters. These sealed canisters are intended for extened (>50 year) dry storage). The typical packaging configuration for the 15-foot DOE canisters places 10 ATR elements within a Type 1a basket, and three baskets are loaded within each DOE canister. During in-reactor operations and cooling pond storage conditions, oxyhydroxide layers form on the surface of the aluminum clad fuel. These layers produce hydrogen gas over time due to the fuel’s radiation field. As part of the packing procedure, the ATR fuel should be dried to remove any residual physio-/chemi- sorbed water from the surface. As testing to the effectiveness of the drying procedure is still underway, this modeling will include results at fully saturated and fully dried conditions. In previous modeling efforts, the G-value for the production of hydrogen from the oxyhydroxide layers was assumed to be in argon environments as measured by Task 2 - Oxyhydroxide Layer Radiolytic Gas Generation Resolution. Previous experimental testing showed differences in the hydrogen generated based on the gaseous environment. In the associated experimental work, Task 2 - Oxyhydroxide Layer Radiolytic Gas Generation Resolution, additional tests were completed in a helium environment, and updated G-values for the radiolytic production of hydrogen from the oxyhydroxide layers were provided. These values are 28% and 58% higher than values for argon. In addition, a change to the modeling of the oxyhydroxide radiolysis has been made from previous reports. This change assumes the dependency of the dose rate on the overall reaction rate is applied to the total weight of the sample, rather than to just the weight of the oxyhydroxide layer. This decreases the dependency of the radiolytic reaction on the thickness of the oxyhydroxide layer. For a nominal scenario of stored ATR fuel, assuming the chemi-/physio- sorbed water have been fully removed, the internal canister pressure increases to 1.61 atm over a 50 year period, with a hydrogen mole percentage of 21%. As in previous modeling, any oxygen present is in negligible amounts (<1 ppt). If a small amount residual air is present, nitric acid can form up to 1300 ppm. For a scenario with high fuel decay heat, the model shows internal pressure increasing to 2.1 atm, with 39.3 mole percentage of hydrogen. In a scenario where significant chemi-/physio- sorbed water is present within the corrosion layer, the nominal scenario shows a pressure increase to 2.54 atm, with 21.1 mole percent hydrogen. The high decay heat case shows a pressure increase to 3.18 atm with 39.9 mole percent hydrogen.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Extended Modeling of DOE Sealed Canisters with Updated Chemistry Models

Road-ready and final disposition packaging configurations for the advanced test reactor (ATR) fuel currently specifies storage within helium backfilled DOE sealed standard canisters. The aluminum cladding of the ATR fuel contains an oxyhydroxide layer of boehmite/bayerite that generates hydrogen when subjected to irradiation. Understanding the effect of this hydrogen buildup over time to important for long term storage considerations. Previous modeling efforts have built a coupled CFD-chemical model to simulate the temperature gas phase concentrations within the DOE sealed standard canisters. These models have coupled the temperature conditions to both the gas phase radiolysis chemistry and the surface chemistry associated with the oxyhydroxide layer. The previous iteration of the model utilized constant G-values for the hydrogen generation a 50-year period. This new iteration of the model utilizes new experimental data to update the hydrogen generation rate, as well as increase the simulated time to a 200-year period. Given the half-life assumed for the primary Cs-137 isotope responsible for gamma radiation in the ATR spent fuel, the 200-year period decreases the decay heat and dose rate of the spent fuel by a factor of 100, which combined with the updated chemistry substantially decreases new hydrogen generation. Continued experimental work has identified trends for aluminum surrogate samples with oxyhydroxide layers for tests done at higher dose rates. At low initial doses a fast generation rate of hydrogen occurs which starts to roll over to a lower generation as the total dose applied increase. A small-scale chemical model was built to replicate as mini-canister surrogate system at SRNL as well as for the smaller capsule tests performed at INL. A variety of chemical models to capture this effect were tested, and a back reaction of H radical absorbing onto the surface, or inhibition of the reaction by significant H 2 cover gas were both able to fit both the mini-canister and the small capsule test data. Both kinetic fits were used to generate data from the new 200-year simulation. As additional experiments continue, the kinetic fits may be adjusted to adapt to new data. However, updated data shows that hydrogen atmosphere has little effect on actual generation data, so the model was reverted to use a star-stepped G-value for low-dose and high-dose regions. For the undried fuel case, the three models differ significantly with an end concentration of 1% for back reaction, 4.2% for inhibition, and 13.1% for constant G-value for the nominal scenario. For the dried fuel case, the nominal cases showed end concentrations of 0.23% for back reaction, 1.9% for inhibition, and 4.2% for const G-value for the nominal scenario. For the constant G-value case that is less conservative than the other two, the pressure for undried fuel increases to 1.94 atm over 200 years for the nominal case and 2.11 atm for the high decay heat case. The total hydrogen concentration after 200 years is 4.92% for the low decay heat case and 27.4% for the high decay heat case for undried fuel and is 1.4% and 9.6% for low and high decay heat fuel for the dried fuel case. If small amount of residual air is present, the potential for nitric acid formation of 595, 1568, and 2816 ppm for the lower, nominal, and upper fuel decay heat can occur. At long-term timeframes no significant shift in major species present occurs, so no appreciable amount of oxygen is present in the system. The continued rate of hydrogen generation in the canister occurs at an increase of 0.04% by mole over the final 10 years. Increasing the model range out to 1000 years continues to drop the hydrogen generation rate through decreasing dose rate.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

CFD-Chemical Model of One-Third Scale Demonstration of DOE Sealed Canister with ATR Fuel

Road-ready and final disposition packaging configurations for the advanced test reactor (ATR) fuel currently specifies storage within helium backfilled DOE sealed standard canisters. The aluminum cladding of the ATR fuel contains an oxyhydroxide layer of boehmite/bayerite that generates hydrogen when subjected to irradiation. Understanding the effect of this hydrogen buildup over time to important for long term storage considerations. Previous modeling efforts have built a coupled CFD-chemical model to simulate the temperature gas phase concentrations within the DOE sealed standard canisters. A demonstration case for these DOE sealed canisters will be eventually performed with a one-third scale mockup that has been instrumented with thermocouples and gas concentration probes. This study seeks to model the planned canister for validation of the previously developed model, such that confidence in its long-term prediction can be increased. The deployment of the instrumented lid for online monitoring in intended for a period of 10+ years based on previous monitoring of commercial fuel storage; however, the model is still run for the previously used 50-year storage periods. This case is modeled with a G-value using a bi-linear function such that it decreases at higher dose rates to be consistent with experiments. Validation efforts of the model would likely revolve around the results for the 1st year, so the results of this timeframe are also highlighted. For dried fuel of the nominal decay heat (18W), the predicted hydrogen concentration is 0.19% after 1 year, 0.91% after 10 years, and 2.8% after 50 years, with a maximum pressure of 1.26 atm. Consistent with previous modeling, the decay heat of the fuel is the main factor that influences the results. For undried fuel in pure helium, the hydrogen concentration ranges from 0.23 to 1.2% after 1 year, 1.25-6.2% after 10 years and 8.8 to 18.84% after 50 years. For dried fuel in pure helium the hydrogen concentration ranges from 0.06% to 0.34% after 1 year, 0.35-2.11% after 10 years and 0.95 to 6.6% after 50 years. While long-term results in the presence of residual air are mostly the same, early reactions with O 2 can delay significant production of H 2 until it is consumed to form more water vapor, lowering the range to 0.14 to 1.0% after 1 year. The maximum absolute pressure that is reached across any scenario is 2.06 atm. In the event of residual air, the presence of nitric acid is possible in the range of 18-131 ppm after 1 year, 174-1180 ppm after 10 years, and 586-3500 ppm after 50 years. As long as the fuel is sufficiently dried, or of nominal decay heat, a 4% lower flammability limit of hydrogen will not be reached within a 10-year monitoring period.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Direct Extraction of Lanthanide Oxides and Nitrates in Tributyl Phosphate

This work investigated the dissolution rate of lanthanide oxides and nitrates in a 30 vol % TBP-n-paraffin solvent (pre-equilibrated with various nitric acid concentrations) using visible (Vis) spectroscopy over time. Some dissolution mechanics were observed, such as an aqueous layer forming, considerably longer dissolution times for the heavier lanthanides (hours) vs. the lighter lanthanides (minutes), and the impacts of mixing lanthanide oxides or nitrates during dissolution. Neodymium, samarium, holmium, and erbium were selected due to their unique spectroscopic signatures and to represent the lighter (neodymium and samarium) and heavier (holmium and erbium) lanthanides. Even though europium does not have a strong absorbance in the range studied, europium was used in some instances to also represent the lighter lanthanides. Cerium oxide was used to representant dissolution of tetravalent lanthanides.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗