Optimum solar cell shielding for the advanced orbiting solar observatory
Optimum solar cell shielding for Advanced Orbiting Solar Observatory
RADIATION SHIELDING↗
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Alternative Liquid DSA Containers for Sludge Slurry Storage in the SRNL Shielded Cells
A pending revision of the Documented Safety Analysis (DSA) for the Savannah River National Laboratory (SRNL) defines “DSA containers'’ for liquid sludge slurries as containers not susceptible to a flashing spray release in a fire and requires that a list of DSA containers be maintained. High Level Waste (HLW) samples are typically received from the Savannah River Site (SRS) tank farm after retrieval from the storage tanks. Glass bottles are preferred for long-term storage of radioactive liquid sludge slurry samples but are not on the DSA container list. An evaluation has been conducted to identify vessel lids for the glass bottles that would vent prior to pressurization and liquid superheating in a fire event. Low melting Field's Metal eutectic alloy with a melting point of 62°C has been identified as a preferred material of construction for vessel lids for this application because the material is resistant to puncture and is expected to have good chemical compatibility and radiation stability in the shielded cells environment. A test vessel was designed to evaluate the alloy and confirm that it would melt before the water in the vessel reaches the boiling point. The tests confirmed: 1) that the reported melting point of the alloy was accurate, 2) that a top fashioned from this material will melt creating an open atmosphere in the vessel headspace prior to sample boiling under moderate and fast heating rates, and 3) that nonuniform heating of the bottle from the bottom also results in melting of the alloy prior to boiling. Based on the results, neither an engulfing fire with standing or toppled bottles or a fire producing localized heat at the vessel bottom would result in flashing spray conditions. It is recommended that glass vessels of various volumes ranging from 125 mL to 1 gallon with modified tops containing the specified Field's Metal eutectic alloy which are appropriately designed to avoid flashing spray material releases be added to the DSA container list. These vessels will be utilized in the Shielded Cells for the storage of radioactive liquid samples of sludge slurries. Use of these vessels will also be advantageous for other liquid samples containing accountable amounts of nuclear material throughout other SRNL laboratories. Similarly constructed paraffin vessel lids could also potentially be added to the approved list, though additional testing would be required.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
High-Energy Radiation and Solar Cell Shields
High energy electron radiation effect on optical transmission of transparent materials for solar cell shields
TRANSPARENT MATERIAL↗
Sludge Washing And Demonstration Of The DWPF Flowsheet In The SRNL Shielded Cells For Sludge Batch 8 Qualification
The current Waste Solidification Engineering (WSE) practice is to prepare sludge batches in Tank 51 by transferring sludge from other tanks to Tank 51. Tank 51 sludge is washed and transferred to Tank 40, the current Defense Waste Processing Facility (DWPF) feed tank. Prior to transfer of Tank 51 to Tank 40, the Savannah River National Laboratory (SRNL) typically simulates the Tank Farm and DWPF processes using a Tank 51 sample (referred to as the qualification sample). WSE requested the SRNL to perform characterization on a Sludge Batch 8 (SB8) sample and demonstrate the DWPF flowsheet in the SRNL shielded cells for SB8 as the final qualification process required prior to SB8 transfer from Tank 51 to Tank 40. A 3-L sample from Tank 51 (the SB8 qualification sample; Tank Farm sample HTF-51-12-80) was received by SRNL on September 20, 2012. The as-received sample was characterized prior to being washed. The washed material was further characterized and used as the material for the DWPF process simulation including a Sludge Receipt and Adjustment Tank (SRAT) cycle, a Slurry Mix Evaporator (SME) cycle, and glass fabrication and chemical durability measurements.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Alternative Liquid DSA Containers for Sludge Slurry Storage in the SRNL Shielded Cells
A pending revision of the Documented Safety Analysis (DSA) for the Savannah River National Laboratory (SRNL) defines “DSA containers” for liquid sludge slurries as containers not susceptible to a flashing spray release in a fire and requires that a list of DSA containers be maintained. High Level Waste (HLW) samples are typically received from the Savannah River Site (SRS) tank farm after retrieval from the storage tanks. Glass bottles are preferred for long-term storage of radioactive liquid sludge slurry samples but are not on the DSA container list. An evaluation has been conducted to identify vessel lids for the glass bottles that would vent prior to pressurization and liquid superheating in a fire event. Low melting Field’s Metal eutectic alloy with a melting point of 62 °C has been identified as a preferred material of construction for vessel lids for this application because the material is resistant to puncture and is expected to have good chemical compatibility and radiation stability in the shielded cells environment. A test vessel was designed to evaluate the alloy and confirm that it would melt before the water in the vessel reaches the boiling point. The tests confirmed: 1) that the reported melting point of the alloy was accurate, 2) that a top fashioned from this material will melt creating an open atmosphere in the vessel headspace prior to sample boiling under moderate and fast heating rates, and 3) that nonuniform heating of the bottle from the bottom also results in melting of the alloy prior to boiling. Based on the results, neither an engulfing fire with standing or toppled bottles or a fire producing localized heat at the vessel bottom would result in flashing spray conditions.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DNA fragmentation induced by Fe ions in human cells: shielding influence on spatially correlated damage
Outside the magnetic field of the Earth, high energy heavy ions constitute a relevant part of the biologically significant dose to astronauts during the very long travels through space. The typical pattern of energy deposition in the matter by heavy ions on the microscopic scale is believed to produce spatially correlated damage in the DNA which is critical for radiobiological effects. We have investigated the influence of a lucite shielding on the initial production of very small DNA fragments in human fibroblasts irradiated with 1 GeV/u iron (Fe) ions. We also used gamma rays as reference radiation. Our results show: (1) a lower effect per incident ion when the shielding is used; (2) an higher DNA Double Strand Breaks (DSB) induction by Fe ions than by gamma rays in the size range 1-23 kbp; (3) a non-random DNA DSB induction by Fe ions. c2004 COSPAR. Published by Elsevier Ltd. All rights reserved.
NASA Discipline Radiation Health↗
Functional components for a design strategy: Hot cell shielding in the high reliability safeguards methodology
Not Available
Borrelli, R. A.↗
Sludge Washing and Demonstration of the DWPF Nitric/Formic Flowsheet in the SRNL Shielded Cells for Sludge Batch 9 Qualification
Savannah River National Laboratory (SRNL) was requested by Savannah River Remediation (SRR) to qualify the next batch of sludge – Sludge Batch 9 (SB9). Current practice is to prepare sludge batches in Tank 51 by transferring sludge to Tank 51 from other tanks. The sludge is washed and transferred to Tank 40, the current Defense Waste Process Facility (DWPF) feed tank. Prior to sludge transfer from Tank 51 to Tank 40, the Tank 51 sludge must be qualified. SRNL qualifies the sludge in multiple steps. First, a Tank 51 sample is received, then characterized, washed, and again characterized. SRNL then demonstrates the DWPF Chemical Process Cell (CPC) flowsheet with the sludge. The final step of qualification involves chemical durability measurements of glass fabricated in the DWPF CPC demonstrations. In past sludge batches, SRNL had completed the DWPF demonstration with Tank 51 sludge. For SB9, SRNL has been requested to process a blend of Tank 51 and Tank 40 at a targeted ratio of 44% Tank 51 and 56% Tank 40 on an insoluble solids basis.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Shielding‐gas‐controlled atmospheric pressure plasma jets: Optical emission, reactive oxygen species, and the effect on cancer cells
Shielding‐gas‐controlled helium atmospheric pressure plasma jets (APPJs) in contact with conductive or nonconductive liquid are investigated. Two‐dimensional (2D) optical emission intensity maps of plasma species directly above the liquid, the 2D distribution of reactive oxygen species (ROS) on the liquid surface, and the 2D distribution of DNA double‐strand breaks (DSBs) induced in epithelial cancer cells are provided. When conductive liquid is used, the plasma jet develops a glow‐discharge‐like structure at the liquid surface. The ROS distribution is found to be strongly dependent on both the shielding gas composition and the liquid conductivity. The amount of DNA DSBs is found to be largest when using N 2 shielding gas and almost negligible when using O 2 shielding gas.
Kapaldo, James↗
Mark-18A Cold Runs
The Savannah River National Laboratory (SRNL) is tasked by the National Nuclear Security Administration (NNSA) to recover highly valued isotopes from irradiated Mark-18A (Mk-18A) targets. The Savannah River Site (SRS) has sixty-five Mk-18A targets available for the recovery of the high valued materials. The sixty-five Mk-18A targets are currently stored in the L-Area Basin and will be removed one at a time and individually transported to SRNL. Upon receipt at SRNL, the Mk-18A target material will be removed from the confinement, dissolved, chemically separated, and calcined to a stable oxide. The flowsheet is designed to recover the plutonium as well as the trivalent actinides. The remaining unrecovered material will be discarded to the high activity drain (HAD) system in SRNL. A specially designed cask was procured for transport of the targets from L-Area to SRNL. Once received at SRNL, the targets will be loaded into the back of Cell 7 and resized as they enter the cell. The resized targets (1/4 lengths) will then be processed one at a time through the following processes: caustic dissolution and filtration; acidic dissolution and filtration, Reillex anion exchange, diglycolamide (DGA) cation exchange; and DGA calcination. This processing will result in two product streams. The first is an aqueous plutonium solution which will be removed from the shielded cells and taken to a glovebox for further purification and conversion to an oxide. The second is a calcined oxide containing the Am and Cm as well as other lanthanide fission products which will be removed from the shielded cells using a bagless transfer system. Both materials will be packaged for shipment to Oak Ridge National Laboratory (ORNL). All equipment to carry out this process was designed, procured or fabricated, and installed in a mock-up facility (716-4A) at SRS to allow for simulation testing in a non-radioactive area. This equipment was then dismantled and transferred from 716-4A to 773-A and installed in the SRNL Shielded Cells. After installation in the shielded cells facility testing was performed using water followed by surrogates and cold chemicals. Issues were identified during these evaluations, including equipment issues as well as technical challenges. Many of the issues were rectified during performance of the cold runs, and the remaining have a resolution identified. Table ES-1 provides a summary of all issues identified during the cold run operations, as well as the status and identified resolutions to outstanding issues.
07 ISOTOPE AND RADIATION SOURCES↗
Aluminum Metal Dissolution for Mark-18A Target Processing
The Mark-18A Program was established to preserve the unique materials found in Mark-18A targets for future research and development activity within the Department of Energy or other U.S. government agencies. The Savannah River Site has 65 Mark-18A targets to be provided for the recovery of high valued isotopes. The majority of the recovery work will be performed in the Savannah River National Laboratory (SRNL) Shielded Cells due to dose concerns. The recovered product will be transported to the Oak Ridge National Laboratory (ORNL) for future research and development tasks. The Mark-18A target material will be removed from their current configuration, dissolved, chemically separated, and calcined to a stable oxide. The baseline flowsheet is patterned after the flowsheet ORNL has previously used to process both Mark-18A and Mark-42 targets. However, it is necessary for SRNL to adapt its facility and the ORNL flowsheet to identify process improvements to successfully recover the actinide materials within the constraints of the SRNL Shielded Cells. These studies evaluated and demonstrated a chemical method for cladding removal using a mixture of sodium nitrate (NaNO 3 ) and sodium hydroxide (NaOH). Experiments provided better insights into dissolution of the Mark-18A aluminum cladding. In particular, these studies measured the effects of temperature on dissolution rate, flammable gas generation, and nitrite formation. Using the lessons learned from batch dissolution testing at SRNL, a continuous dissolution flowsheet was successfully demonstrated using prototypic-shaped materials and chemically-conservative materials. The SRNL target dissolution flowsheet departs from the baseline ORNL flowsheet in small but significant ways as discussed in the following paragraphs. These flowsheet changes facilitate faster initial reaction rates at a given temperature, enable cladding removal at a lower process temperature, and yield a more uniform dissolution rate throughout cladding removal. Multiple tests indicate that the total time at the dissolution temperatures will be about six hours for one fourth of a target. The shorter dissolution time provides the benefit of a single-shift dissolver operation. Off gas sampling determined that the quantities of flammable hydrogen (H 2 ) and ammonia (NH 3 ) expected in the process off gas will meet allowable flammable gas generation rates for the Shielded Cells Facility. The SRNL approach enables NaOH to be added at a constant rate, which simplifies operations in the shielded cells. In all six continuous dissolution tests, the process was easy to control with minimal operator interface. Analyses of process solutions throughout the experiments provide a better understanding of system behavior. The data show consistent behavior across several similar but different test conditions. Between the batch-test and continuous-test data, if future changes to the flowsheet are required, sufficient data exist to adequately predict how the system will respond to those changes. Future work will focus on design, fabrication, and evaluation of a system for dissolution of one-fourth of a Mark-18A target, which will be full-scale operations for SRNL. Based on these results, the process parameters are adequately understood and the system can be operated safely.
36 MATERIALS SCIENCE↗
Atalante research facility new mass limits in a unit containing plutonium and minor actinides - 14402
ATALANTE, located in Marcoule, is one of the main Nuclear Facilities of the French Alternative Energies and Atomic Energy Commission (CEA). This facility is mainly dedicated to research and development for nuclear energy related to the back end of the fuel cycle: spent fuel and final waste management. It is a large research complex housing chemistry and analytical laboratories and shielded cells (240 glove boxes and 11 shielded cells), equipped for basic studies through to demonstration experiments. Within the facility, one unit can contain plutonium and curium at the same time. In this case, nuclear criticality safety is achieved by exercising control over mass. A subcritical mass limit is defined for plutonium (280 g for {sup 239}Pu) and an exemption mass limit (3 g for {sup 245}Cm) is used for the actinide. In order to increase the maximum mass (in particular {sup 245}Cm) which can be handled in the unit, new mass control limits were studied, based on control over the sum of the mass ratio to the upper subcritical mass, noted m{sub lim}, for each nuclide (350 g for {sup 239}Pu; 34.8 g or 23 g for {sup 245}Cm) as shown in equation (1) below. m({sup 239}Pu)/ m{sub lim}({sup 239}Pu)+ m({sup 245}Cm)/m{sub lim}({sup 245}Cm)≤1 (1) To validate this criticality safety control method, a study was conducted to demonstrate that, for different masses of {sup 239}Pu and {sup 245}Cm meeting equation (1), the calculated effective multiplication factor is always lower than the effective multiplication factor corresponding to a sphere of 350 g of pure {sup 239}Pu, moderated, with 20 cm water reflection. The two pseudo fissile media are: - metallic plutonium (100% {sup 239}Pu, density 19.86), moderated with water, - metallic curium (100% {sup 245}Cm, density 13.566), moderated with water. The calculation model is a sphere of {sup 245}Cm metal positioned in a sphere of {sup 239}Pu metal, with optimum moderation for each fissile material, and a full water reflection (20 cm). The results show that, under certain conditions, the effective multiplication factor of the mixture can be higher than the effective multiplication factor of each pure fissile material when using the upper subcritical mass of 34.8 g for {sup 245}Cm. When using 23 g of {sup 245}Cm, the k{sub eff} of the mixture is always lower than the k{sub eff} of pure {sup 239}Pu. These calculations enable the safety documentation to be changed by introducing a new process operational procedure, based on equation (1). This change in procedure allows the operator to work with a larger quantity of curium in the unit. (authors)
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Mark-18A Target Material Recovery Program: Initial Hot Startup
The Savannah River National Laboratory (SRNL) has been tasked by the National Nuclear Security Administration to recover highly valued isotopes from irradiated Mark-18A (Mk-18A) targets. The Savannah River Site (SRS) has sixty-five Mk-18A targets available for the recovery of the high valued materials. The Mk-18A targets were manufactured with Pu-242 then irradiated under high neutron flux in K-Reactor at the SRS from 1968 to 1978. The sixty-five Mk-18A targets are currently stored in the SRS L-Area Basin and will be removed one at a time and individually transported to SRNL. Upon receipt at SRNL, the Mk-18A target material will be removed from the confinement, dissolved, chemically separated, and calcined to a stable oxide. The flowsheet is designed to recover the plutonium as well as the trivalent actinides. The original targets were manufactured with varying quantities of Pu-242, ranging from 5.2 grams to 121.5 grams. Taking a graded approach to process start up, the lowest loaded target (FT-80-03) was selected as the first target to be received and processed at SRNL. As operational experience and knowledge is gained from processing targets, higher loaded targets will be selected for processing to increase the quantities of valuable isotopes recovered. Due to dose concerns, the receipt and processing of the Mk-18A targets is performed in the SRNL Shielded Cells Facility. The targets are stored in a double J-can confinement in the L-Area Basin. A specially designed cask was procured for transport of the targets from L-Area to SRNL. Once received at SRNL, the targets are loaded into the back of Cell 7 and resized as they enter the cell. The resized targets (1/4 length) are then processed one at a time through the following processes: caustic dissolution and filtration, acidic dissolution and filtration, elutable Reillex anion exchange, diglycolamide (DGA) resin extraction, and DGA calcination. This processing results in two product streams. The first is an aqueous plutonium solution which is removed from the shielded cells and taken to a glovebox for further purification and conversion to an oxide. The second is a calcined oxide product containing the Am and Cm as well as other lanthanide fission products which is removed from the shielded cells using a bagless transfer system. Both materials are packaged for shipment to Oak Ridge National Laboratory (ORNL). This paper will discuss the operating experience, lessons learned, and results from initial process hot startup.
Armstrong, Christopher [Savannah River National La↗
Remanent Activation in the Mini-SHINE Experiments
Argonne National Laboratory is assisting SHINE Medical Technologies in developing a domestic source of the medical isotope 99Mo through the fission of low-enrichment uranium in a uranyl sulfate solution. In Phase 2 of these experiments, electrons from a linear accelerator create neutrons by interacting in a depleted uranium target, and these neutrons are used to irradiate the solution. The resulting neutron and photon radiation activates the target, the solution vessels, and a shielded cell that surrounds the experimental apparatus. When the experimental campaign is complete, the target must be removed into a shielding cask, and the experimental components must be disassembled. The radiation transport code MCNPX and the transmutation code CINDER were used to calculate the radionuclide inventories of the solution, the target assembly, and the shielded cell, and to determine the dose rates and shielding requirements for selected removal scenarios for the target assembly and the solution vessels.
Micklich, Bradley J.↗
Evaluation of Detritiation Strategies for Mitigating Tritium Releases
Tritium confinement is performed using different barriers to minimize releases to the environment. Primary confinement is usually performed by process piping and components, secondary confinement typically by inerted gloveboxes connected to a continuously operating tritium stripper system. Some facility designs employ tertiary tritium confinement by using process room confinement with a tritium stripper system activated after an accident to further mitigate tritium releases to the environment. Secondary confinement of some tritium systems such as radiation-hardened or shielded-cell enclosures are not easily achieved using designs for inerted glovebox. These enclosures can be challenging to seal for secondary tritium confinement due to heating, ventilation, and air conditioning (HVAC) needs for temperature control, as well as feedthroughs for equipment and instrumentation, are potential pathways for tritium leaks or loses. Additional tritium leak paths can be created in secondary confinement enclosures due to a Design Basis Accident (DBA) seismic event. These new, larger tritium leak rates from the DBA event are usually not part of the stripper system design basis so all tritium released to the secondary confinement enclosure is assumed released to the environment. Detritiation strategies for shielded enclosures and/or gloveboxes with significant leak rates are necessary to minimize off-site radiological dose consequences. To address this, the relationship between the confinement system leak rate and stripper system (recirculating) flow rate was evaluated to meet prescribed maximum allowable environmental tritium emissions from the event. The simple analysis described assumes a constant leak rate from the confinement system (i.e. shielded enclosure or glovebox) while a recirculating stripper system strips tritium from the system– a competition between tritium release and tritium recovery. At a high level, the resulting expression summarizes the relationship between system leak rate (F L ) and stripper flow rate (F S ) relative to allowable tritium release (Q A ) and initial tritium release (Q 0 ): $\frac{F_L}{F_s}$ ≈ $\frac{Q_A}{Q_0}$. This report derives the relationship between these parameters, examines the impact of finite stripping times followed by purging of the stripped volume. The analysis provides example results for up to 30 gram tritium releases: the maximum tritium inventory of a Hazard Category III (Department of Energy) Nuclear Facility. The detritiation model presented represents a high-level analysis of tritium recovery versus losses for “leaky” confinement enclosures such as shielded cells after large tritium releases. Key parameters for doing the analyses are the system leak rate, stripper flow rate, initial tritium release, and allowable tritium release values. The analyses show a proportional decrease in releases with reduction in initial tritium release but a non-linear increase in tritium recovery with increased stripper flow rate or reduced leak-rate. The analyses also apply to glovebox confinement systems which could have significant increases in leak rates after a DBA. The analysis recognizes but does not include tritium absorption followed by re-emission from the walls of the confinement volume – an analysis which could be pursued in future studies.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Results of Hg speciation testing on tanks 30, 32, and 37 depth samples
The Savannah River National Laboratory (SRNL) was tasked with preparing and shipping samples for Hg speciation by Eurofins Frontier Global Sciences, Inc. in Seattle, WA on behalf of the Savannah River Remediation (SRR) Mercury Task Team. The twelfth shipment of samples was designated to include 3H evaporator system Tank 30, 32, and 37 depth samples. The Tank 30 depth sample (HTF-30-15-70) was taken at 190 inches from the tank bottom and the Tank 32 depth sample (HTF-32-15-68) was taken at 89 inches from the tank bottom and both were shipped to SRNL on June 29, 2015 in an 80 mL stainless steel dip bottles. The Tank 37 surface sample (HTF-37-15-94) was taken around 253.4 inches from the tank bottom and shipped to SRNL on July 21, 2015 in an 80 mL stainless steel dip bottle. All samples were placed in the SRNL Shielded Cells and left unopened until intermediate dilutions were made on July 24, 2015 using 1.00 mL of sample diluted to 100.00 mL with deionized H 2 O. A 30 mL Teflon® bottle was rinsed twice with the diluted tank sample and then filled leaving as little headspace as possible. It was immediately removed from the Shielded Cells and transferred to refrigerated storage where it remained at 4 °C until final dilutions were made on October 20. A second portion of the cells diluted tank sample was poured into a shielded polyethylene bottle and transferred to Analytical Development for radiochemical analysis data needed for Hazardous Material Transportation calculations.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Provision Of Carbon Nanotube Bucky Paper Cages For Immune Shielding Of Cells, Tissues, and Medical Devices
System and method for enclosing cells and/or tissue, for purposes of growth, cell differentiation, suppression of cell differentiation, biological processing and/or transplantation of cells and tissues (biological inserts), and for secretion, sensing and monitoring of selected chemical substances and activation of gene expression of biological inserts implanted into a human body. Selected cells and/or tissue are enveloped in a "cage" that is primarily carbon nanotube Bucky paper, with a selected thickness and porosity. Optionally, selected functional groups, proteins and/or peptides are attached to the carbon nanotube cage, or included within the cage, to enhance the growth and/or differentiation of the cells and/or tissue, to select for certain cellular sub-populations, to optimize certain functions of the cells and/or tissue and/or to optimize the passage of chemicals across the cage surface(s). A cage system is also used as an immuns shield and to control operation of a nano-device or macroscopic device, located within the cage, to provide or transform a selected chemical and/or a selected signal.
Loftus, David J.↗
Results of Hg Speciation Testing on Tank 38 and 2H Evaporator Overhead Tank (OHT-1) Materials
The Savannah River National Laboratory (SRNL) was tasked with preparing and shipping samples for Hg speciation by Eurofins Frontier Global Sciences, Inc. in Seattle, WA on behalf of the Savannah River Remediation (SRR) Mercury Task Team. The seventh shipment of samples was designated to include two Tank 38 (2H Evaporator System Drop Tank) samples: a surface sample (HTF-38-15-48) and a variable depth sample taken at 265” (HTF-38-15-50). The third sample submitted was from the 2H Evaporator Overhead Tank (OHT) #1. All samples were received by SRNL on May 11, 2015. The OHT-1 was not placed in the SRNL Shielded Cells but was instead placed in a refrigerator in 773-A, B-wing upon receipt and kept at 4°C. The Tank 38 samples were left in their stainless steel dip bottles until they were opened and an aliquot of each diluted 1:10 with Eurofins deionized water on June 1, 2015. Following initial dilution in the Shielded Cells, the diluted Tank 38 samples were immediately moved to refrigeration and kept in the dark until prepared for shipment to Eurofins.
54 ENVIRONMENTAL SCIENCES↗