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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Glycolate Analysis in Tank 22: Developing and Testing Analytical Methods for the Savannah River Site Liquid Waste System

Researchers developed and tested a range of techniques to support a low mg/L Limit of Quantitation (LOQ) of glycolate in radioactive Tank 22 waste solution by Ion Chromatography (IC) and by proton NMR. For the IC method, Dionex OnGuard II cartridges were tested as a means of removing alkali earth and transition metals that can interfere during ion chromatography applications especially when analyzing for low-concentration, chelating analytes. Additionally, concentrations of nitrate in the raw Tank 22 sample (5,000 to 10,000 mg/L) were managed by using reasonable levels of sample dilution. The resulting IC performance quality was enhanced by improving the baseline, increasing sensitivity [Limit of Quantitation (LOQ) of 12 mg/L and Limit of Detection (LOD) of 4.0 mg/L], and resolving analytes into well-defined Gaussian peaks when using the Dionex OnGuard II H⁺ cartridges to remove matrix interferences. High concentrations of nitrate limited the performance of the IC method. To achieve an acceptable baseline, samples with high nitrate content require more dilution, resulting in higher detection limits. Tests on samples with higher concentrations of nitrate (Tank 30 and 32 supernate with approximately 150000 mg/L nitrate) suggested that IC detection limits for glycolate in these samples would be > 500 mg/L. Thus, an LOQ of 12 mg/L is not feasible by this IC method on evaporator feed samples. An alternative method of glycolate analysis, using proton nuclear magnetic resonance (H NMR), was developed by the research team. In initial tests, the H NMR technique provided reasonable quantitation of glycolate in Tank 22 conditions by direct observation of the liquid. The H NMR method may provide improved detection limits for solutions with higher nitrate concentrations. We recommend further development of this analysis for high nitrate LWS samples such as evaporator feed and evaporator drop tank content. A method for pretreatment of samples using crystalline silicotitanate (CST) was developed and tested. The objective of the pretreatment was to facilitate analysis of samples with higher levels of radioactivity. The pretreatment did not influence the glycolate concentration in solution, and the pretreatment is projected to reduce Cs 137 activity in a sample by a factor of 16,200. The double strike CST pretreatment was demonstrated and would allow milliliters of higher activity samples to be transferred from the Shielded Cells and handled in a containment unit for glycolate analysis. The various studies validated IC and H NMR methods for glycolate analysis, defined the range of applicability, and demonstrated key supporting analytical protocols. Based on the results, high quality glycolate analysis of Tank 22 is feasible down to approximately 12 mg/L, with the potential for broader applicability of the methods to other conditions in the Savannah River Site Liquid Waste System (LWS).

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

Characterization of Tank 22 Demister Solids Responsible for Preventing Adequate Purge Ventilation Flow (U)

In December 2020, about 2.5 grams of the solid material from the Tank 22 purge ventilation demister, responsible for preventing adequate purge ventilation flow, were delivered to Savannah River National Laboratory (SRNL) by Savannah River Remediation (SRR) for analysis to identify the chemical nature of the solids and possibly address why fresh water and caustic flushes with 50 wt. % sodium hydroxide solution are essentially ineffective in dissolving or dislodging the accumulated solids from the Tank 22 demister at low ambient temperature conditions.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Analysis of Tank 38H (HTF-38-26-16, -17), Tank 43H (HTF 43-26-18, -19) and Tank 22 (HTF-22-26-20, -21) Samples for Support of the Enrichment Control and Corrosion Control Programs

Savannah River National Laboratory (SRNL) analyzed samples from Tank 38H, Tank 43H and Tank 22H to support the Enrichment Control Program (ECP) and Corrosion Control Program (CCP). The results indicate the concentrations of most soluble species in the Tank 38H surface sample are ~ 76% of the concentrations in the previous Tank 38H surface sample. The current Tank 38H subsurface sample is a clear solution with soluble species that are ~ 70% of the concentrations in the previous Tank 38H subsurface. Significant differences in the concentrations of major components between the current Tank 38H surface and subsurface samples indicate stratification of solution species between these two locations within Tank 38H.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Sludge Batch 11 Assembly: Tank 51

Savannah River Mission Completion (SRMC) Nuclear Safety and Engineering Integration has requested that Savannah River National Laboratory (SRNL) perform Tank 51 characterization analyses in support of Sludge Batch 11 (SB11) assembly. This report provides important characterization of the slurry in Tank 51 after transfers from Tank 22, Tank 35, and Tank 13 (post Tank 15 to Tank 13 transfer) to Tank 51 that demonstrates the sludge concurs with the estimated transfer mass for the SB11 recipe. A total of 3 sets of Tank 51 samples were delivered to SRNL from March 2023 to February 2024. The composite sample was analyzed for the following: density, weight percent solids, chemical composition, radionuclides, and supernate corrosion control analyses. The results of the Tank 51 samples are consistent with and representative of expected sludge projections for Sludge Batch 11.

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

Analysis of Tank 38h (HTF-38-23-19, -20) and Tank 43H (HTF-43-23-21, -22) Samples for Support of the Enrichment Control and Corrosion Control Programs

SRNL analyzed samples from Tank 38H and Tank 43H to support ECP and CCP. The results indicate the concentrations of most soluble species in the Tank 38H surface sample decreased significantly from the previous surface sample. The Tank 38H sub-surface sample shows changes in concentration for soluble species in the solution with some increasing and some decreasing. The current Thank 38H sub-surface sample contains visible sludge solids similar to the previous sample, i.e., less than 1%, based on visual appearance. The significant differences in the concentrations of major components between the Tank 38H surface and sub-surface samples indicate significant stratification of solution species between these two locations within the tank. Savannah River Mission Completion personnel indicate that ~ 150,000 gallons of Tank 22 supernate were received in Tank 38H since its last analysis during which time the 2H evaporator was not operated, so the observed stratification is expected. The Tank 43H surface and sub-surface samples are similar in composition to the previous samples. The similar solution compositions measured in the Tank 43H surface and sub-surface samples indicate a minimal stratification within the tank. The total uranium and plutonium in the current Tank 38H surface sample remains similar to the previous analysis. The Tank 38H sub-surface sample shows an increase in uranium and plutonium concentrations compared to the previous sample likely because of an increase in sludge solids in the current sample. The total uranium concentration in the Tank 43H surface sample is similar to the previous sample results while the plutonium in the Tank 43H sub-surface sample increased relative to the previous analysis. The sum of the major cations versus the sum of the major anions shows a difference of <10% for both samples from Tank 38H sub-surface sample (174 mg/L) increased compared with the previous sample. The Tank 38H surface sample silicon concentrations (27.5 mg/L) is one-half of the previous sample results. The Tank 43H surface sample silicon concentrations compared to the previous sample results indicate Si in the concentration range of 60 to 83 mg/L. Thus, these current samples analyzed from Tanks 38H and 43H show overall silicon concentrations ranging from 27.5 to 174 mg/L.

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

Proton Nuclear Magnetic Resonance (H NMR) of Glycolate in Real Waste: Developing and Testing Analytical Methods for the Savannah River Site Liquid Waste System

In preparation for implementing the NG acid flowsheet for the SRS LWS, analytical methods for determining glycolate at low concentration are desired to support system management and safety. Previous work documented that the IC method performed well for samples with low to moderate ionic strength. That work also included a scoping effort to determine if an alternate analytical strategy using H NMR would complement and extend the capabilities of the IC method. The H NMR scoping results indicated that the method had the potential to expand glycolate analysis in LWS samples to higher ionic strength tanks/solutions. The scope of this work is to develop innovative proton NMR techniques, including demonstrating ion exchange decontamination protocols needed for application of the technique to real waste samples. Part of radioactive waste processing at SRS uses formic acid to reduce oxidized (Hg 2+ ) to more volatile elemental Hg for steam striping, collecting, and disposal. Under acidic conditions found in the Chemical Processing Cell (CPC) at the Defense Waste Processing Facility (DWPF), formic acid has a much higher hydrogen generation rate than an alternative reductant, glycolic acid. Thus, an NG acid flowsheet has been developed utilizing glycolic acid with the benefit of easing the need for headspace monitoring requirements for hydrogen and ammonia at DWPF. Low concentrations of glycolate are conservatively assumed to be in the recycle stream, which will collect in the Recycle Collection Tank (RCT). The DWPF recycle stream collected in the RCT has a distinct pathway to the LWS waste tanks that feed the 2H and 3H Evaporator. This route involves transfer of the DWPF recycle to Tank 22 in the Concentration, Storage, and Transfer Facilities (CSTF) followed by transfer to the LWS tank farm/evaporators.

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

Examining Thermolytic Production of Hydrogen from Lubrication Oil

Savannah River National Laboratory (SRNL) was requested by Savannah River Remediation (SRR) to conduct testing via Technical Task Request (HR) to determine the thermolytic HGR of Mobil SHC™ 630, a lubrication oil. Currently, 35 gallons of contaminated Mobil SHC 630 is proposed for release into the recycle stream from the Defense Waste Processing Facility (DWPF) to Tank 22 and then to the 242-16H (2H) Evaporator system. Inhibited recycle waste in the Recycle Collection Tank (RCT) is transferred to Recycle Pump Tank (RPT) in the Low Point Pump Pit (LPPP) and then to the Concentration, Storage and Transfer Facilities (CSTF) H-area. The lubrication oil would be added directly to the RPT, bypassing the RCT. The current DWPF waste compliance plan for liquid transfers from the RCT to the CSTF limits the concentration of Mobil SHC 630 to <1,100 ppm which is equal to <9.3 gallons of Mobil SHC 630 when considering a 7,500 gal RCT batch with an initial Mobil SHC 630 concentration of 42 mg/L. Mobil SHC 630 is expected to be largely immiscible in the caustic aqueous waste stream. It is a blend of base oils including polyalphaolefin (PAO) base oil and additives such as triphenylphosphate and cresyl diphenyl phosphate at various concentrations (<0.25 wt%). While the base oils are expected to be largely unreactive in CSTF waste, the triarylphosphates additives would be expected to hydrolyze in the caustic waste, forming diarylphosphates and phenol. The tests described herein were governed by a single Run Plan and will determine thermolytic HGR from the caustic aqueous solution, as well as from any organic phases present.

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

Analysis of Glycolate in Radioactive Waste by Ion Chromatography (IC) and Proton Nuclear Magnetic Resonance (H NMR)

Here, in preparation for implementing the Nitric-Glycolic (NG) acid flowsheet for the Savannah River Site (SRS) Liquid Waste System (LWS), analytical methods for determining glycolate at low concentration, below 20 mg/L in radioactive samples, were developed to support system management and safety. To accommodate the wide range of LWS matrix conditions, two alternative methods were developed, refined, and demonstrated for glycolate analysis in radioactive waste samples: ion chromatography (IC) and a proton nuclear magnetic resonance (H NMR). Investigators validated IC and H NMR methods for glycolate analysis, defined the range of applicability, and demonstrated key supporting analytical protocols. The deployed IC method is applicable in low to moderate ionic strength samples and requires sample pretreatment using a Dionex OnGuard II H + cartridge. The deployed H NMR method is more labor intensive but provides options for a broader range of matrices. Based on the results, high quality glycolate analysis of the Defense Waste Processing Facility (DWPF) condensate in Tank 22 is feasible by IC down to approximately 12 mg/L. Using H NMR, glycolate may be determined to 8 mg/L or lower depending on the run time with the potential for broader applicability of the method to higher ionic strength conditions in other tanks of the SRS LWS.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2025

New Tank Mapping Method Improves Waste Removal Process

Waste Tank Mapping Overview • Camera inspections are performed within available tank top risers and used to create waste tank maps – Several camera inspections are performed during waste removal transfers to verify the elevation of the visible salt/sludge mounds against the known elevation of the liquid surface • Tank mappings are used to evaluate the volume and distribution of saltcake or sludge that is present within the waste tank – Allows for refined operating strategies and process safety controls • New tank mapping process creates a standardized approach for accurately defining waste distribution within a waste tank while minimizing the required camera inspection footage – First utilized during the 2023 Tank 22 Sludge Removal Campaign

Mini, Melany [Savannah River Mission Completion (S↗
DOE OSTI2024

Characterization of Precipitate Reactor Feed Tank (PRFT) Batches 15, 22, and 26 from the Defense Waste Processing Facility (DWPF)

The Savannah River Site (SRS) Defense Waste Processing Facility (DWPF) processes a Monosodium Titanate/Sludge Solids (MST/SS) waste stream received from the Salt Waste Processing Facility (SWPF) via the Precipitate Reactor Feed Tank (PRFT). During processing, DWPF is required to provide evidence of compliance with the Waste Acceptance Product Specifications (WAPS) to ensure acceptance of their vitrified high-level waste (HLW) into the Civilian Radioactive Waste Management System. Production Records must document the constituents of the MST/SS material in the PRFT from each salt batch (StB) processed at SWPF. Savannah River Mission Completion (SRMC) has requested Savannah River National Laboratory (SRNL) to analyze PRFT samples representing each SWPF salt batch for thirty-two radionuclides. Additionally, elemental analysis of PRFT slurry and MST/SS solids was performed to aid SRMC in further refinement of the inputs and assumptions used in future frit development and Material Tracking Program calculations. The analyses of PRFT Batches 15, 22, and 26, which corresponds to material from the processing of StB4, StB5, and StB7, respectively, are reported herein. The unwashed dried solids of the PRFT batches were found to be 86-87% MST. Additionally, the total sulfur values are well below the assumed 982 mg of sulfate/kg of PRFT slurry used in Material Tracking Program calculations.

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

Analysis of Defense Waste Processing Facility Sample: Off-Gas Condensate Tank Sample Batch 27862

The Savannah River National Laboratory (SRNL) was requested by Savannah River Remediation (now Savannah River Mission Completion (SRMC)), through a Technical Task Request (X-TTR-H-00107), to characterize the “as-received” Off-Gas Condensate Tank (OGCT) Sample identified as sample batch 27862, which was delivered to SRNL Shielded Cells on April 30, 2021. This OGCT report is the third of three sample characterization reports dealing with this Defense Waste Processing Facility (DWPF) Recycle Diversion Project. The other two recycling stream characterization reports are the Recycle Collection Tank (RCT) and the Slurry Mix Evaporator Condensate Tank (SMECT) sample report.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Mechanistic Modeling of TEG Dehydrator Emissions in Oil and Gas Industry

This work presents a mechanistic modeling approach for simulating methane emissions from triethylene glycol (TEG) dehydrators used in oil & gas (O&G) operations. The model was developed as a modular component of the Mechanistic Air Emissions Simulator (MAES) tool, incorporating species-specific absorption and emission dynamics through two-level, second-order polynomial regression (PR) models trained on ProMax simulation data: (1) species-level regression models that track the transfer rates of individual gas species within the dehydrator unit streams, and (2) outlet flow stream regression models that predict the fraction of inlet gas distributed among the outlet streams of the dehydrator unit. These behaviors were characterized over a range of glycol circulation ratios, wet gas pressures, and temperatures. The model was validated using root mean square error (RMSE) analysis. The species-level PR achieved low root mean square error (RMSE) values (<0.03) for light hydrocarbon species across all dehydrator components, ranging from 0.0009 for methane to 0.029 for normal pentane. Similarly, the outlet-level PR yielded RMSE values below 0.002 for the dry gas fraction, 0.001 for the flash tank fraction, and 0.002 for the still vent fraction, demonstrating strong agreement between predicted and reference ProMax values. When deployed at field facilities, the model significantly improved MAES-simulated dehydrator emissions, revealing that gas-assisted glycol pump emissions are the dominant contributors to both dehydrator-level and site-level methane emissions under uncontrolled conditions. Further analysis of the 154 dehydrator units reported by operators under the AMI 2024 project showed that 54 units (31%) used gas-driven glycol pumps, of which 6 units (11%) operated with uncontrolled flash tanks, and 22 units (40.7%) were identified as potentially oversized. Of the six dehydrator units with uncontrolled gas-assisted pumps, pump emissions accounted for 90.25% of total dehydrator emissions and 63.10% of total site-level emissions. These findings highlight substantial opportunities for emissions mitigation through equipment upgrades.

MAES↗
DOE OSTI2026

Volume Change from Solidification Correlation for Hanford Tank Waste.

The U.S. Department of Energy (DOE), Hanford Field Office’s primary mission is to safely and effectively treat Hanford’s tank waste and deliver environmental remediation. Mixed radioactive waste is stored in the underground tanks at the Hanford Site. It was recently estimated that retrieval of the waste in the 200 West Area underground tanks in the SY, S, SX, and U Tank Farms will result in about 41 million gallons of mixed low-level waste (MLLW) (RPP-RPT-65147, Rev. 1). The current plan is to retrieve at least 22 S, SX, and U Farm tanks and pretreat1 to produce pretreated tank waste (PTW) which will be further treated (including solidification /immobilization) for Resource Conservation and Recovery Act (RCRA) Land Disposal Restriction (LDR) organics and inorganics before being transferred to an offsite out-of-state facility for disposal (RPP-PLAN-66135, Rev. 2). In addition, solidified PTW from the 200 East Area may also be transferred to an offsite out-of-state facility for disposal to ensure availability of critical Double-Shell Tank space, meet retrieval obligations, and optimize 200 East Area operations.

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

Analysis of Respirator Cartridge Performance Testing on Hanford Tank A-101

Washington River Protection Solutions (WRPS) conducted tests on two types of chemical cartridges for use in air purifying respirators (APR) to determine the period of time that the cartridges would provide adequate performance1 to protect workers when exposed to a mixture of Chemicals of Potential Concern (COPCs) from vapors emanating from the headspace of tank A-101 on the Hanford Site. The Occupational Safety and Health Administration (OSHA) identifies cartridge testing as a valid approach for establishing a cartridge change schedules.[3] Testing is commonly applied in situations where mixtures of COPCs exist, and where other approaches, such as manufacturer recommendations and modeling, are less reliable. The tests were designed and conducted to ensure measurement and/or control of the key variables OSHA identified as important to estimate cartridge service life, including temperature, humidity, COPC concentration, breathing rate, and cartridge adsorption capacity. Testing was conducted over a period from July 22-24, 2016, using headspace vapors from Hanford tank A-101 under static conditions2 fed to a respirator cartridge test stand developed by WRPS in collaboration with HiLine Engineering (Richland, Washington). Multipurpose respirator cartridges, SCOTT 7422-SD1 and 7422-SC1 (SCOTT Safety, Monroe, North Carolina) were assessed on separate days with A-101 headspace vapors. Sample media (sorbent tubes) were used to collect samples of the vapor stream entering and exiting the respirator cartridge, and were subsequently analyzed for COPC concentrations. Pacific Northwest National Laboratory was tasked with conducting an independent analysis of the analytical results and making recommendations based on the results for respiratory cartridge performance and service life. Key conclusions from the assessment of the 59 COPCs in this study are described below: • Based on d measured cartridge inlet vapor concentrations from tank A-101 headspace, both ammonia and N-Nitrosodimethylamine (NDMA) exceeded their Occupational Exposure Limits (OEL).3 These measurements were generally consistent with maximum A-101 headspace measurements previously obtained for these two compounds. • Ammonia had respirator cartridge outlet concentrations that exceeded 10% of the OEL for both cartridges tested, indicating breakthrough for each. For the SCOTT 7422-SD1 cartridge, ammonia appeared to breakthrough above 10% of its OEL after 2 hours. For the SCOTT 7422-SC1 cartridge, ammonia breakthrough appeared to occur after 8 hours. 1 “Adequate performance” refers to the breakthrough criteria used in this analysis—that is, sustained cartridge outlet concentrations above 10% of the OEL. For some COPCs, an alternate threshold has been applied when necessary due to higher detection limit (DL)/reporting limit (RL) values for specific compounds. Ultimately, Industrial Hygiene professionals will use these results along with specific hazard assessments to determine service life and cartridge use that provides the necessary performance. 2 These tests were conducted under static conditions absent waste-disturbing activities in the subject tank or tank farm. 3 Occupational Exposure Limits accepted for Hanford Tank Farm use are based on OELs established by a U.S. governmental agency or national professional organization (e.g., OSHA, National Institute for Occupational Safety and Health, American Conference of Governmental Industrial Hygienists), or if no U.S. OEL exists, standard toxicological practices are applied to develop OELs using non-U.S. exposure limits, other established OELs for chemical surrogates when available, or other standard procedures. The OEL for NDMA was established in 2005 based on the MAK (Maximale Arbeitsplatzkonzentration) Commission standard adopted in Europe. iv • Despite respirator inlet measurements for NDMA that were in excess of its OEL, all corresponding outlet measurements, from both respirator cartridges, were below analytical DL 1 for the

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

Empirical characterization of vertical-tube inlets in hot-water storage tanks

Hot-water storage in residential and commercial tanks is of growing interest for peak-shifting the electrical demand associated with hot-water loads. How the water enters the tank can significantly affect the extraction efficiency of the storage. Vertical-tube inlets (VTIs) direct cold water into the tank. The VTI design influences whether large-scale mixing occurs, which can reduce the delivery capacity of the hot-water storage. In this work, the mixing at the VTI discharge was characterized by the eddy diffusivity factor (EDF) in an allometric curve fit to the ratio Re:Ri using inlet superficial velocities. The full experimental campaign tested VTI's with Re’s between 170 and 22,300 and Ri's between 0 and 376. Animated visualization of the experimentally measured data is provided, showing tank temperature evolution during inlet flows. The most predictive correlations for EDF are for VTI Re in the transition zone (i.e. ~2300 < Re < ~10,000) and at a Ri corresponding to a stably stratified tank (i.e. Ri > 5). For vertical-tubes that distribute cold water below the thermocline, increasing the EDF will decrease the stratification in a hot-water tank as EDF quantifies inlet mixing. Overall, the new correlations for VTIs can be used to design and model thermally-stratified storage tanks.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2022

MELCOR Accident Progression and Source Term Demonstration Calculations for a FHR

MELCOR is an integrated thermal hydraulics, accident progression, and source term code for reactor safety analysis that has been developed at Sandia National Laboratories for the United States Nuclear Regulatory Commission (NRC) since the early 1980s. Though MELCOR originated as a light water reactor (LWR) code, development and modernization efforts have expanded its application scope to include non-LWR reactor concepts. Current MELCOR development efforts include providing the NRC with the analytical capabilities to support regulatory readiness for licensing non-LWR technologies under Strategy 2 of the NRC's near- term Implementation Action Plans. Beginning with the Next Generation Nuclear Project (NGNP), MELCOR has undergone a range of enhancements to provide analytical capabilities for modeling the spectrum of advanced non-LWR concepts. This report describes the generic plant model developed to demonstrate MELCOR capabilities to perform fluoride-salt-cooled high-temperature reactor (FHR) safety evaluations. The generic plant model is based on publicly-available FHR design information. For plant aspects (e.g., reactor building leak rate and details of the cover-gas system) that are not described in the FHR references, the analysts made assumptions needed to construct a MELCOR full-plant model. The FHR model uses a TRi-structural ISOtropic (TRISO)-particle fuel pebble-bed reactor with a primary system rejecting heat to two coiled tube air heat ex changers. Three passive direct reactor auxiliary cooling systems provide heat removal to supplement or replace the emergency secondary system heat removal during accident conditions. Surrounding the reactor vessel is a low volume reactor cavity that insulates the reactor with fire bricks and thick concrete walls. A refractory reactor liner system provides water cooling to reduce the concrete wall temperatures. Example calculations are performed to show the plant response and MELCOR capabilities to characterize a range of accident conditions. The accidents selected for evaluation consider a range of degraded and failed modes of operation for key safety functions providing reactivity control, the primary system decay heat removal and also a piping leak of the line to the coolant drain tank.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Characterization results for the 2020 tank farm 2H and 3H evaporator overhead samples

The 2020 annual 2H and 3H Evaporator overhead samples were received at Savannah River National Laboratory (SRNL) on December 22, 2020. The delivered samples were analyzed for concentration of cesium-137 ( 137 Cs), strontium-90 (90 Sr) and iodine-129 ( 129 I) to validate compliance with the Effluent Treatment Project Waste Acceptance Criteria (ETP WAC, Rev. 7). The concentration of all three radionuclides, 137 Cs, 90 Sr, and 129 I, in the 2H Evaporator overhead sample were found to be less than the corresponding ETP WAC limits. Radiochemical analysis of the 3H Evaporator overhead sample indicated a 137 Cs content about an order of magnitude higher than the acceptable threshold value, and 90 Sr and 129 I concentrations within the acceptance limits.

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

Oxygen consumption of sexually mature adult, first-feeding larval, and yearling Pacific Lampreys

Abstract Objective The Pacific Lamprey Entosphenus tridentatus is considered a first food of Native American tribes, such as the Confederated Tribes of the Umatilla Indian Reservation. Populations of this imperiled species have declined such that harvest for traditional use is limited. As a result, propagation and culture of Pacific Lampreys have been initiated to supply animals for restoration and recovery efforts. These efforts require transport and periodic holding of both adults and larvae under static (no-flow) conditions. Hence, guidelines for ensuring an adequate oxygen supply are needed for all life stages of this species. Methods We measured oxygen consumption rates of mature adults, first-feeding larvae, and yearlings under static conditions at 12–15°C. In addition, we recorded indicators of stress during hypoxia for adults and estimated routine respiration rates during and after larval feeding. Result Adults exhibited surprisingly high metabolic rates when at rest in static chambers. At 12°C, a single adult typically reduced dissolved oxygen levels in 10 L to <2 mg/L in just 90 min (oxygen consumption rates of 100–200 mg/kg/h). Adults often started to climb the walls of open chambers when dissolved oxygen levels approached 2 mg/L, allowing them to raise their branchiopores into air. Larvae remained quiescent, even when oxygen levels dropped below 1 mg/L, and costs of feeding increased routine respiration by 22–56%. Conclusion Based on these observations, we recommend that adult Pacific Lampreys always be transported in aerated containers and with access to air at the top of the tank. Although larvae exhibited hypoxia tolerance, care should be taken to ensure adequate oxygen availability, particularly during and immediately after feeding.

Moser, Mary↗