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DOE OSTI2022

Characterization Results for the March 2022 Tank Farm 3H Evaporator Overhead Sample

On an annual basis, Savannah River Mission Completion (SRMC) provides 2H and 3H evaporator overhead samples to Savannah River National Lab (SRNL) to be analyzed per Section 5.2 of the Effluent Treatment Project (ETP) Waste Compliance Plan (WCP) and the Waste Acceptance Criteria (WAC). This report presents characterization results for the March 2022 3H evaporator overhead sample. The sample was clear and colorless with no visible solids. The results provide measurements for cesium-137 ( 137 Cs), strontium-90 ( 90 Sr), and iodine-129 ( 129 I) with the radionuclide concentration limits specified by the WAC. These analyses were performed on a single sample replicate. A summary of the analytical results for the 3H evaporator overhead sample includes the following. The measured cesium-137 activity in the 3H evaporator overhead sample is 4.59E+02 dpm/mL which is above the ETP WAC limit of 3.28E+02 dpm/mL. The strontium-90 activity in the 3H evaporator overhead sample is <2.95E+01 dpm/mL. This is below the ETP WAC limit of 1.76E+02 dpm/mL. The iodine-129 activity in the 3H evaporator overhead sample is <1.92E-02 dpm/mL. This is below the ETP WAC limit of 1.00E+00 dpm/mL.

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DOE OSTI2021

Tank 49H Solids Disturbance Analysis

Tank 49H will serve as the feed tank for the Salt Waste Processing Facility (SWPF). Transfers into Tank 49H may disturb solids that have settled to the bottom of the tank, resulting in feed that may exceed the insoluble solids content limit of 1,200 mg/L from the SWPF Waste Acceptance Criteria (WAC). During a transfer into Tank 49H, material that free falls into Tank 49H through a downcomer and passes through the supernate could potentially disturb the solids on the bottom of the tank and scour or suspend solids from a settled solids layer or turbid region. A previous analysis and report evaluated the potential to disturb solids when transferring into Tank 49H through the B4 riser and recommended a minimum tank level of 120 inches to prevent disturbing any solids in the bottom of the tank. The scope of this task is to perform additional fluid flow analysis to determine the impact of a “plunging jet” through the C3 downcomer on solids in Tank 49H and to determine a minimum liquid level to be maintained in Tank 49H prior to transfers that will minimize disturbing the solids and exceeding the SWPF WAC limits for insoluble solids carryover to SWPF since the C3 riser downcomer is at a lower elevation and farther away from the B5 transfer pump than the B4 downcomer.

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DOE OSTI2021

Tank 49H Solids Disturbance Analysis

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DOE OSTI2021

Acetonitrile Destruction and Fate of Organics in the Reverse Osmosis System at the ETF

The Hanford Site Effluent Treatment Facility (ETF) currently treats aqueous waste streams that include condensates from the 242-A evaporator, leachate from the Environmental Restoration Disposal Facility (ERDF), as well as laboratory wastes and, in the future, will treat liquid effluents from the Hanford Tank Waste Treatment and Immobilization Plant (WTP) and Integrated Disposal Facility (IDF) leachate. Liquid effluents from the WTP will have significant concentrations of acetonitrile. Acetonitrile is formed by reaction of nitrates and sugar in the WTP low activity waste (LAW) melters and is prevalent in the submerged bed scrubber (SBS) and wet electrostatic precipitator (WESP) liquid effluents from WTP off-gas treatment. When these liquids are concentrated in the WTP Effluent Management Facility (EMF) evaporator in the direct feed low activity waste (DFLAW) flow-sheet, testing has shown that the majority of the acetonitrile partitions to the evaporator condensate. Since the evaporator condensate is directed to the ETF, this creates a potential issue with the ETF waste acceptance criteria. Consequently, there is a need to validate flow-sheet assumptions on the fate of acetonitrile and other organics within the ETF. The present plan includes the addition of a steam stripper to the ETF to remove acetonitrile. There is, therefore, also a need to determine a suitable method to destroy acetonitrile in the overhead condensate stream from the new steam stripper. Washington River Protection Solutions, LLC (WRPS) previously contracted with Atkins and the Vitreous State Laboratory (VSL) of The Catholic University of America (CUA) to perform development and testing work to evaluate potential methods for destruction of acetonitrile in WTP secondary liquid effluents. Based on the results of that work, WRPS requested that follow-on testing be conducted to further evaluate acetonitrile destruction in the steam stripper condensate using ultraviolet oxidation (UV/OX) with persulfate. WRPS also requested testing to assess the rejection rate of organics in the reverse osmosis (RO) system installed in the ETF. This report presents the results from testing to address those needs.

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DOE OSTI2022

Tank 49H solids disturbance analysis

Tank 49H serves as the feed tank for the Salt Waste Processing Facility (SWPF). Transfers into Tank 49H may disturb any solids that have settled to the bottom of the tank, resulting in feed that may exceed the insoluble solids content limit of 1,200 mg/L of the SWPF Waste Acceptance Criteria (WAC). During a transfer into Tank 49H, material that free falls into Tank 49H through a downcomer could potentially disturb any solids on the bottom of the tank and scour or suspend solids from a settled solids layer or turbid region. A previous analysis and report evaluated the potential to disturb solids when transferring into Tank 49H through the B4 riser and recommended a minimum tank level of 120 inches to prevent disturbing any solids in the bottom of the tank. The scope of this task is to perform additional fluid flow analysis to determine whether accounting for disturbed particle settling and particle mixing and dispersion during transfer could allow the minimum liquid level to be reduced below 120 inches. The analysis utilized models from the technical literature to calculate the size and shape of the “plunging jet” as a function of input parameters such as initial velocity, initial jet diameter, elevation of the initial jet, liquid level in the tank, and solids depth. The analysis relied on the M-Star® simulations performed for the previous analysis to provide bounding estimates of the amount of solid particles disturbed and used the MStar® software to calculate the dispersion and mixing of the disturbed solids with other liquid in the tank as the solids are transported to the transfer pump. The analysis showed that with a solid particle size of 5 micron or less, a liquid level of 120 inches should be maintained to prevent significant disturbance of the solid layer at the bottom of Tank 49H.

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DOE OSTI2020

Small to Full Height Scale Comparisons of Cesium Ion Exchange Performance with Crystalline Silicotitanate

The U.S. Department of Energy’s (DOE) Hanford Site houses 56 million gallons of high-level radioactive waste generated from plutonium production from 1944 to 1988. The supernatant waste, currently stored in underground tanks, is intended to be vitrified following filtration and 137 Cs removal at the Hanford Waste Treatment and Immobilization Plant (WTP) Low-Activity Waste (LAW) Vitrification Facility. The WTP Pretreatment Facility will not be operational for several years. The Tank Side Cesium Removal (TSCR) system is a technology demonstration that will remove cesium from tank waste supernate to support directly feeding LAW to the vitrification facility. The 137 Cs removal is important to meet the WTP LAW contract specification and ultimately for creating a contact-handled waste form. The waste acceptance criteria (WAC) limit for the WTP LAW Facility is <3.18E-5 Ci 137 Cs per mole Na. The TSCR system is skid mounted and employs two key technologies: dead end filtration for solids removal and ion exchange (IX) for cesium removal. Filtration is necessary to protect the functionality of the IX columns. The IX process utilizes sodium form crystalline silicotitanate (CST) IX media produced by UOP Honeywell, LLC (Des Plaines, IL) under the product name IONSIV R9140-B, 18 x 50 mesh, in a lead-lag-polish column configuration. Each column contains a CST bed height of 92 inches and a volume of approximately 157 gallons of CST IX media. The full-scale TSCR operation will run at a residence time of 1.9 bed volumes (BV)/h, which results in a superficial velocity of 7.3 cm/min and a flowrate of ~5 gal/min. Column testing at small (2.5% of the full bed height), medium (12% of the full bed height), and full-height scales has been previously conducted to evaluate process variables and scale up performance of Cs exchange onto the CST. Cesium load performances from various sieve cuts at the small scale indicated that a 30-mesh sieve cut be tested to determine if it better reflects the 12% and full-height column performances at the small scale. Two process flowrates were tested in the small-column configuration with <30-mesh CST and simulant solutions. These tests were compared to full-height column tests at the same residence times to assess CST particle size effect on column scaling. Table S.1 summarizes the observed column performance determined for the two flowrates juxtaposed to the previous work with 5.6 M Na simulant at higher scales (used as benchmarks). The WAC breakthroughs between the small and full-height tests at 1.3 BV/h differed by ~59 BVs. The WAC breakthroughs for small, medium, and full-height tests at 1.8 BV/h were consistent at 240 BVs. The 50% Cs breakthroughs were nominally equivalent for all column tests. The common transition zones and onsets of Cs breakthrough at the 1.8 BV/h tests indicated that the Cs mass transfers were equivalent and thus the <30-mesh CST Cs load performance at the small scale successfully modeled that of the full-height system. It is recommended that the <30-mesh CST be used in subsequent 10-mL CST bed tests.

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DOE OSTI2020

Tank 49H Solids Disturbance Analysis

Tank 49H will serve as the feed tank for the Salt Waste Processing Facility (SWPF). Transfers into Tank 49H may disturb solids that have settled to the bottom of the tank, resulting in feed that may exceed the insoluble solids content limit of 1200 mg/L from the SWPF Waste Acceptance Criteria (WAC). During a transfer into Tank 49H, material that free falls from the Tank 49H B4 downcomer and passes through the supernate could potentially disturb the solids on the bottom of the tank and scour or suspend solids from a settled solids layer or turbid region. The scope of this task is to perform fluid flow analysis to determine the impact of a “plunging jet” through the B4 downcomer on solids in Tank 49H and to determine a minimum liquid level to be maintained in Tank 49H prior to transfers that will minimize disturbing the solids and exceeding the SWPF WAC limits for insoluble solids carryover to SWPF.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2021

Advanced Low-Temperature Chlorination of Zirconium

Recovery of Zr from nuclear fuel is of high interest because the Zr cladding material comprises up to 50% of the high-level radioactive waste (HLW) that requires disposition. An ideal Zr-recovery process would not only recover the bulk Zr from the fuel or cladding but would also provide decontamination from the components that give rise to the HLW designation. A simplified decontamination pathway providing recovered Zr in compliance with low-level radioactive waste acceptance criteria could significantly reduce the burden of HLW requiring geologic disposition. A newly developed advanced low-temperature chlorination process presents an opportunity to recover a ZrCl4 product that is effectively decontaminated from the elements of concern. This advanced low-temperature chlorination process uses a mixture of the sulfur-containing chlorination compounds sulfur monochloride (S 2 Cl 2 ) and thionyl chloride (SOCl 2 ).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Flowsheet for the Neutralization of Accelerated Basin De-inventory (ABD) Material

Under the Accelerated Basin De-inventory (ABD) program H-Canyon will be dissolving aluminum spent nuclear fuel (ASNF) and then neutralizing that solution without performing head-end strike or uranium recovery operations. After dissolution in H-Canyon the material will be neutralized to a free hydroxide concentration of either 0.6 M or 1.2 M to meet the waste acceptance criteria for transfer to the SRS Concentration, Storage and Transfer Facility (CSTF). In order to ensure the material could be successfully neutralized and transferred, SRNL completed experiments with simulated H-Canyon dissolver solutions. Two bounding simulants were developed based on the expected compositions of ASNF to be dissolved, one containing only Gd as the neutron poison and one containing Fe as an additional poison.

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DOE OSTI2023

Results for the January 2023 Semiannual Tank 50 Salt Solution Sample

In this Technical Report, the chemical and radionuclide contaminant results from the January 2023 Semiannual sample of Tank 50 salt solution are presented in tabulated form. The information from this characterization will be used by Savannah River Mission Completion (SRMC) for the transfer of aqueous waste from Tank 50 to the Saltstone Production Facility (SPF), where the waste will be treated and disposed in the Saltstone Disposal Facility. This Technical Report compares results, where applicable, to SPF Waste Acceptance Criteria (WAC) LIMITS and TARGETS that were established at the time the Tank 50 sample was obtained.1 The chemical and radionuclide contaminant results from the characterization of the January 2023 semiannual sampling of Tank 50 were requested by SRMC personnel via a Task Technical Request (TTR)2 and details of the testing are presented in the Savannah River National Laboratory (SRNL) Task Technical and Quality Assurance Plan (TTQAP).3 This Technical Report is part of Deliverable 2 relating to Task 1 from the SRMC request.2 Data pertaining to the regulatory limits for Resource Conservation and Recovery Act (RCRA) metals per Task 2 from the SRMC request will be obtained semiannually for the January 2023 and July 2023 Tank 50 samples.

Crawford, Charles L.↗

DOE OSTI2023

Evaluation of High Level Waste Sludge Processing Behavior

The U.S. Department of Energy’s (DOE) Hanford Site has 177 underground storage tanks that contain wastes from past nuclear fuel reprocessing and waste-management operations. Over 20% of this waste is in the form of an insoluble sludge that will require slurry modification before its transfer to the Waste Treatment and Immobilization Plant (WTP). Specific WTP acceptance criteria for waste feed delivery describe the physical and chemical characteristics of the waste that must be met before the waste is transferred to the WTP. One challenging requirement relates to the undissolved solids (UDS) composition in a waste feed because the waste contains solid particles that settle, and their concentration and relative proportion can change during the transfer of the waste in individual batches. A key uncertainty is the ability to transfer and mix wastes with large variations in UDS concentrations and resulting settling rates. To address this uncertainty, a number of small scale mixing and settling tests have been conducted to determine the mobilization performance of variable chemistry simulants. Comparison of the size and density of the particulate for each simulant to that of southeast area Hanford sludge was made using metrics for particle mobilization, suspension, settling, and pipeline transfer where dependance on particle size and density may be different, including: 1. Settling velocity, 2. Critical shear stress for erosion, 3. Just-suspended impeller speed, and 4. Pipeline critical transport velocity. Existing high-level waste sludge data has shown the effect that increasing Al concentration has on resulting settled solids. This differential settling of particles in the sludge has the possibility of resulting in solids segregation during feed preparation and uneven particle distribution during pipeline transportation or mixer jet pump operations. Understanding the predictive capabilities of HLW solids settling and transport as well as potential remedies for addressing disparate sludge behaviors can help provide technical guidance during HLW flowsheet planning.

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DOE OSTI2025

Sulfate Conversion of Reillex HPQ Anion Exchange Resin for Disposal (Interim Report)

This report describes preliminary data to validate the Savannah River Plutonium Processing Facility’s (SRPPF) flowsheet for conversion of used Reillex HPQ anion exchange resin from the nitrate form to the sulfate form. The nitrate form is an oxidizer and therefore does not meet acceptance criteria for disposal at the Waste Isolation Pilot Plant (WIPP). The purpose of this study is to develop data to support acceptance for this disposition pathway. Due to the challenges characterizing the nitrate concentration on solid resin, the data developed to date are based upon indirect analysis of the ion exchange column effluent by ion chromatography. These challenges are discussed and two methods for quantification of nitrate directly on the resin are recommended for further development: TGA-MS and permanganate digestion followed by IC. The resin used for this work was provided in the chloride form; this is the form in which resin is supplied by the manufacturer. However, it had to be converted to the nitrate form, which is the form that will be used in SRPPF’s ion exchange process, prior to use in the sulfate conversion experiments. The chloride-form resin was characterized. A lab-scale procedure for the conversion of Reillex HPQ resin from the chloride to nitrate form was validated. The nitrate-form resin was assessed for particle size and chloride concentration to ensure it met SRPPF’s facility specifications. The baseline sulfate conversion flowsheet was tested. However, nitrate was still detectable in the effluent after approximately 10 bed volumes of 1 M sodium sulfate had been passed through the resin bed. Additional experiments were performed to assess the effect of increasing the feed volume, reducing the flowrate, the use of 2 M sulfuric acid instead of sodium sulfate, and the use of irradiated resin. The sulfuric acid test was the only one which provided a nondetectable nitrate concentration (<0.002 M) in the effluent. Detectable nitrate in the column effluent suggests that nitrate is still present on the resin itself.

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DOE OSTI2022

Sludge Batch 10 (SB10) Acceptance Evaluation: Radionuclide Concentrations in Tank 51 Washed Qualification Sample

Savannah River National Laboratory (SRNL) has been tasked with the radionuclide characterization of the washed Sludge Batch 10 (SB10) qualification sample. The washed SB10 qualification sample is based on SRR Engineering guidance and the sample slurry is expected to be similar in composition to Tank 51 slurry after final preparations for transfer to Tank 40. Forty-four radionuclides along with total alpha and beta activity have been reported herein. These radionuclide measurements are required for the Defense Waste Processing Facility (DWPF) Radiological Evaluation Program, DWPF Technical Safety Requirements (TSR)/Waste Acceptance Criteria (WAC) Evaluation, and the DWPF Solid Waste Characterization Program.

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DOE OSTI2022

Physical and Flow Properties of Glass Forming Chemicals (V2O5, SnO, SnO2, Cr2O3, FeCr2O4, and ZrSiO4) and Mixtures

For a sustainable nuclear waste vitrification process at the Hanford Tank Waste Treatment and Immobilization Plant (WTP), proper selection and consistent supply of glass-forming chemicals (GFCs) are crucial. Establishing rigorous acceptance criteria for the characterization of GFCs will be required to operate the vitrification facility and to mitigate any processing issues or failures. Low-activity wastes (LAW) are blended with GFCs to form slurry melter feeds and vitrified in a melter. To enhance properties of waste glasses, new chemicals are being introduced to the current GFC mixture (Vienna et al. 2016; Muller et al. 2017, 2019). In this study, three new GFCs were evaluated for enhanced LAW glass formulations: chromium oxide (Cr 2 O 3 ), vanadium oxide (V 2 O 5 ), and stannic oxide (SnO 2 ). These three oxide components are included in enhanced waste glass (EWG) formulations and GFCs with the appropriate physical and flow properties are needed. As a starting point, single metal oxide GFCs: Cr 2 O 3 , V 2 O 5 , and SnO 2 were sourced and tested. To characterize these new individual GFCs and mixtures of GFCs, the industrial bulk characterization consultant, Jenike and Johanson, was employed to measure physical and flow properties of individual GFCs and their mixtures. Pacific Northwest National Laboratory (PNNL) also measured several selected physical properties for data evaluation as a quality assurance step. In addition, PNNL measured physical and rheological properties of slurry melter feeds containing those GFCs. Subsequent data analyses and verification were conducted. The purpose of this report is to assess the applicability of these GFCs for LAW vitrification based on their properties. This report will help understand measured data and evaluate new GFCs for use. Moreover, this report may give useful insights to help troubleshoot any GFC and melter feed transport and mixing issues that arise during processing, leading to a successful cleanup mission at WTP.

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DOE OSTI2024

STRUCTURAL MODELING TO SUPPORT POST-YIELD ACCEPTANCE CRITERIA FOR SPENT NUCLEAR FUEL CLADDING

Spent nuclear fuel (SNF) is evaluated for structural failure during storage and transportation scenarios. The U.S. Department of Energy’s Spent Fuel and Waste Science and Technology (SFWST) program has sponsored significant research in quantifying mechanical loads on SNF during storage and transportation scenarios using experimental and modeling methods. The SFWST program has also performed significant research on measuring the mechanical behavior of irradiated SNF as defueled cladding segments and cladding with fuel pellets to measure composite behavior. This paper considers some of the key material data from the Sibling Pin testing and uses structural modeling and analysis methods that have been informed by testing to consider post-yield acceptance criteria for SNF cladding structural analysis. Test data published by Oak Ridge National Laboratory (ORNL) and Pacific Northwest National Laboratory (PNNL) are the foundation for informing the material behavior of the models developed in this study. In particular, four-point bend (4PB) tests of fueled and defueled cladding segments provide significant information about the bending failure mode of SNF. ORNL’s 4PB test data is on fueled cladding segments, so the composite behavior of SNF is demonstrated. This paper describes PNNL’s coincident beam model that was developed to approximate the composite behavior of SNF. This paper also presents PNNL’s structural dynamic finite element models of a cask tip-over scenario, which is predicted to cause the strongest mechanical loads on SNF of all postulated storage and transportation scenarios. SNF bending loads predicted in the cask tip-over scenario and cladding acceptance criteria beyond yield are considered, with justification based on the Sibling Pin test data. ASME Boiler and Pressure Vessel code stress intensity limits are also considered. The ultimate goal of this work is to aid in the justification of structural acceptance criteria for SNF cladding beyond the cladding’s irradiated yield strength for use in structural analysis of all storage and transportation scenarios.

Klymyshyn, Nicholas A.↗

NASA NTRS1997

Use of the Aromascan(TM) Instrument for Nonsubjective Evaluation of Rodent Spaceflight Hardware

This report describes the verification and utilization of the AromaScan(TM) (Hollis, NH) instrument for the ground-based evaluation of odor containment by various spaceflight habitats developed at NASA's Ames Research Center (ARC). The AromaScan(TM) instrument is an electronic odor detection system consisting of 32 polymer sensors that respond differentially to 10 different chemical groups present in an air sample. The AromaScan(TM) system also includes neural network software for constructing a database of known odors, against which an unknown odor can be compared. At present, the standard method for characterizing rodent odor containment during the development and testing of spaceflight hardware is the use of a human odor assessment panel. However, this can be a very time consuming and costly process, and the results are inherently subjective. The AromaScan(TM) system should produce more consistent and objective results, as well as a cost savings in the long term. To test and verify the AromaScan(TM) instrument, daily air samples will be collected from the exhaust port of rodent habitats, during experiment development tests, then injected into the instrument and used to create a database of recognizable odors. Human sniff tests will be performed in conjunction with the AromaScan(TM) analysis, and the results will be correlated. We will then teach the neural network to differentiate between an acceptable and an unacceptable odor profile, as defined by the human sniff test, and to be able to accurately identify an odor that would not pass a sniff panel. The results of our efforts will be to verify that the AromaScan(TM) system is a valuable alternative to human sniff panel assessments for the early iterative process of designing and testing rodent waste filters for spaceflight. Acceptance by a human panel will remain one of the final criteria for successful rodent habitat development.

Scribner, K. A.↗

DOE OSTI2020

Test for Reconditioning RA Waste with Simulated Bitumen and Concrete in a 1,2 MW Plasma Test Facility - 20092

The operation and maintenance of nuclear power plants, the non nuclear fuel cycle, etc generate low-level radioactive waste which, along with the historical radioactive waste from past nuclear activities, needs to be treated and stored, awaiting final disposal. Plasma technology offers a very effective way of treating this waste with a high volume reduction factor (VRF), free from organics, liquids and moisture, and meets without a doubt the acceptance criteria for safe storage and disposal. By means of a plasma beam of approximately 5000 deg. C, the inorganic materials are melted into a glassy slag, containing most of the radioactive isotopes while the organic material is gasified, oxidized and purified in an off-gas cleaning system. First the paper describes the new full-scale Plasma Melting Facility (PMF) at the Kozloduy Nuclear Power Plant in Bulgaria which was taken in nuclear operation In May 2018. The plant has a capacity of 250 tons per year and the maximum contact dose rates of the incoming waste is 2 mSv/h. Different mixtures of radioactive waste packed in 200 l drums were successfully treated resulting in a glassy slag free from liquids and organic material with an important volume reduction factor (VRF). The Project was co-financed through a grant by Kozloduy International Decommissioning fund (KIDSF) administrated by the EBRD through Bulgarian national funding. Plasma is a suitable technology for treatment of problematic waste or even reconditioning waste so Belgoprocess was contracted to do plasma tests with simulated conditioned waste types. One can do tests on a laboratory scale on smaller samples and torch capacities of e.g. 50 kW but Belgoprocess wanted to do more realistic and reliable tests. So Belgoprocess contracted Phoenix Solutions Co who has a full-scope test facility equipped with a 1200 kW plasma torch for full-scope treatment of simulated conditioned waste. For a first confidential contract simulated 200 l (55 gallon) bitumen drums were treated. The drums contained different pucks of compacted waste such as rags, used filters, granulates, etc. The pucks were stacked in the 200 l drums and subsequently embedded with bitumen. A total of 6 drums were treated in the plasma facility. For a second contract simulated homogeneous 200 l (55 gallon) concrete drums with on the one hand concentrates and on the other hand spent resins were selected. A total of 6 drums with concrete and spent resins were treated and melted in the plasma testing facility. The paper describe the test facility, volume reduction factor (VRF) of different waste streams and most important parameters. (authors)

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DOE OSTI2023

V31 Test Report

The V31 containment vessel was procured by the US Army Recovered Chemical Materiel Directorate (RCMD) as a third-generation EDS containment vessel. It is the fifth EDS vessel to be fabricated under Code Case 2564 of the 2019 ASME Boiler and Pressure Vessel Code, which provides rules for the design of impulsively loaded vessels. The explosive rating for the vessel, based on the code case, is twenty-four (24) pounds TNT-equivalent for up to 1092 detonations. This report documents the results of explosive tests that were performed on the vessel at Sandia National Laboratories in Albuquerque, New Mexico to qualify the vessel for field operations use. There were three design basis configurations for qualification testing. Qualification test (1) consisted of a simulated M55 rocket motor and warhead assembly of 24lbs of Composition C-4 (30 lb TNT equivalent). This test was considered the maximum load case, based on modeling and simulation methods performed by Sandia prior to the vessel design phase. Qualification test (2) consisted of a regular, right circular cylinder, unitary charge, located central to the vessel interior of 19.2 lb of Composition C-4 (24 lb TNT equivalent). Qualification test (3) consisted of a 12-pack of regular, right circular cylinders of 2 lb each, distributed evenly inside the vessel (totaling 19.2 lb of C-4, or 24 lb TNT equivalent). All vessel acceptance criteria were met .

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