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All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Interoperable Technologies for Advanced Petascale Simulations

Our final report on the accomplishments of ITAPS at Stony Brook during period covered by the research award includes component service, interface service and applications. On the component service, we have designed and implemented a robust functionality for the Lagrangian tracking of dynamic interface. We have migrated the hyperbolic, parabolic and elliptic solver from stage-wise second order toward global second order schemes. We have implemented high order coupling between interface propagation and interior PDE solvers. On the interface service, we have constructed the FronTier application programer's interface (API) and its manual page using doxygen. We installed the FronTier functional interface to conform with the ITAPS specifications, especially the iMesh and iMeshP interfaces. On applications, we have implemented deposition and dissolution models with flow and implemented the two-reactant model for a more realistic precipitation at the pore level and its coupling with Darcy level model. We have continued our support to the study of fluid mixing problem for problems in inertial confinement fusion. We have continued our support to the MHD model and its application to plasma liner implosion in fusion confinement. We have simulated a step in the reprocessing and separation of spent fuels from nuclear power plant fuel rods. We have implemented the fluid-structure interaction for 3D windmill and parachute simulations. We have continued our collaboration with PNNL, BNL, LANL, ORNL, and other SciDAC institutions.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2013

Quantify Water Extraction by TBP/Dodecane via Molecular Dynamics Simulations

The purpose of this project is to quantify the interfacial transport of water into the most prevalent nuclear reprocessing solvent extractant mixture, namely tri-butyl- phosphate (TBP) and dodecane, via massively parallel molecular dynamics simulations on the most powerful machines available for open research. Specifically, we will accomplish this objective by evolving the water/TBP/dodecane system up to 1 ms elapsed time, and validate the simulation results by direct comparison with experimentally measured water solubility in the organic phase. The significance of this effort is to demonstrate for the first time that the combination of emerging simulation tools and state-of-the-art supercomputers can provide quantitative information on par to experimental measurements for solvent extraction systems of relevance to the nuclear fuel cycle. Results: Initially, the isolated single component, and single phase systems were studied followed by the two-phase, multicomponent counterpart. Specifically, the systems we studied were: pure TBP; pure n-dodecane; TBP/n-dodecane mixture; and the complete extraction system: water-TBP/n-dodecane two phase system to gain deep insight into the water extraction process. We have completely achieved our goal of simulating the molecular extraction of water molecules into the TBP/n-dodecane mixture up to the saturation point, and obtained favorable comparison with experimental data. Many insights into fundamental molecular level processes and physics were obtained from the process. Most importantly, we found that the dipole moment of the extracting agent is crucially important in affecting the interface roughness and the extraction rate of water molecules into the organic phase. In addition, we have identified shortcomings in the existing OPLS-AA force field potential for long-chain alkanes. The significance of this force field is that it is supposed to be optimized for molecular liquid simulations. We found that it failed for dodecane and/or longer chains for this particular solvent extraction application. We have proposed a simple way to circumvent the artificial crystallization of the chains at ambient temperature.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

FY12 Final Report for PL10-Mod Separations-PD12: Electrochemically Modulated Separation of Plutonium from Dilute and Concentrated Dissolver Solutions for Analysis by Gamma Spectroscopy

Accurate and timely analysis of plutonium in spent nuclear fuel is critical in nuclear safeguards for detection of both protracted and rapid plutonium diversions. Gamma spectroscopy is a viable method for accurate and timely measurements of plutonium provided that the plutonium is well separated from the interfering fission and activation products present in spent nuclear fuel. Electrochemically modulated separation (EMS) is a method that has been used successfully to isolate picogram amounts of Pu from nitric acid matrices. With EMS, Pu adsorption may be turned “on” and “off” depending on the applied voltage, allowing for collection and stripping of Pu without the addition of chemical reagents. In this work, we have scaled up the EMS process to isolate microgram quantities of Pu from matrices encountered in spent nuclear fuel during reprocessing. Several challenges have been addressed including surface area limitations, radiolysis effects, electrochemical cell performance stability, and chemical interferences. After these challenges were resolved, 6 µg Pu was deposited in the electrochemical cell with approximately an 800-fold reduction of fission and activation product levels from a spent nuclear fuel sample. Modeling showed that these levels of Pu collection and interference reduction may not be sufficient for Pu detection by gamma spectroscopy. The main remaining challenges are to achieve a more complete Pu isolation and to deposit larger quantities of Pu for successful gamma analysis of Pu. If gamma analyses of Pu are successful, EMS will allow for accurate and timely on-site analysis for enhanced Pu safeguards.

Pratt, Sandra H.↗
DOE OSTI2013

Summary of the Graduate Mentoring Assistance Program for PNNL PNNL.12.006 (April 2012 – May 31, 2013)

Summary of Students and Projects: Sarah Bender: The possible improvement of the Multi-Isotope Process (MIP) Monitor by the addition of Compton suppression is investigated for this study. The Compton continuum from the dominant 661.7 keV Cs-137 fission product peak obscures lower energy lines which could be used for spectral and multivariate analysis. A novel Compton suppression system which uses fast, room-temperature materials will be developed using simulation. As opposed to the standard high purity germanium (HPGe) detector suppressed using a thallium-doped sodium iodide (NaI(Tl)) guard detector array, cerium-doped lanthanum bromide (LaBr3) will be used as the primary spectroscopic detector. For Sarah’s Master’s degree, she constructed and validated the unsuppressed detector response of the Pennsylvania State University (PSU)-Compton suppression system. For Sarah’s PhD, she is completing that simulation by including the suppressed operation which will be validated with spent fuel samples. She is also going to use the validation of suppressed operation to predict the detector response of a LaBr3 based Compton suppression system. Richard Clark: Richard Clark completed his study this FY of a new technique known as Intrinsic Dosimetry. Intrinsic Dosimetry is the method of measuring total absorbed dose received by the walls of a container holding radioactive material. By considering the total absorbed dose received by a container in tandem with the physical characteristics of the radioactive material housed within that container, this method has the potential to provide enhanced pathway information regarding the history of the container and its radioactive contents. The latest in a series of experiments designed to validate and demonstrate this newly developed tool are reported. Kenneth Dayman: The MIP Monitor is an online, non-destructive approach to process monitoring in a nuclear fuel reprocessing facility using gamma spectra taken of process streams and multivariate data analysis to detect anomalies in operation. Last year, Mr. Dayman investigated methods to classify spent nuclear fuel according to both reactor type and burn up based on the constituent nuclide activities, which may be used to more accurately calibrate other models in an analysis scheme in order to improve the accuracy and precision of predictions. He is currently writing up his results for publication and has transitioned to another project with Dr. Derrick Haas at PNNL.

Schwantes, Jon M.↗
DOE OSTI2013

Summary of Plutonium-238 Production Alternatives Analysis (Final Report)

The Team implemented a two-phase evaluation process. During the first phase, a wide variety of past and new candidate facilities and processing methods were assessed against the criteria established by DOE for this assessment. Any system or system element selected for consideration as an alternative within the project to reestablish domestic production of Pu-238 must meet the following minimum criteria: Any required source material must be readily available in the United States, without requiring the development of reprocessing technologies or investments in systems to separate material from identified sources. It must be cost, schedule, and risk competitive with existing baseline technology. Any identified facilities required to support the concept must be available to the program for the entire project life cycle (notionally 35 years, unless the concept is so novel as to require a shorter duration). It must present a solution that can generate at least 1.5 Kg of Pu-238 oxide per year, for at least 35 years. It must present a low-risk, near-term solution to the National Aeronautics and Space Administration’s urgent mission need. DOE has implemented this requirement by eliminating from project consideration any alternative with key technologies at less than Technology Readiness Level 5. The Team evaluated the options meeting these criteria using a more detailed assessment of the reasonable facility variations and compared them to the preferred option, which consists of target irradiation at the Advanced Test Reactor (ATR) and the High Flux Isotope Reactor (HFIR), target fabrication and chemical separations processing at the ORNL Radiochemical Engineering Development Center, and neptunium 237 storage at the Materials and Fuels Complex at INL. This preferred option is consistent with the Records of Decision from the earlier National Environmental Policy Act (NEPA) documentation

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Final scientific report for DOE award title: Improving the Representation of Ice Sedimentation Rates in Global Climate Models

It is well known that cirrus clouds play a major role in regulating the earth’s climate, but the details of how this works are just beginning to be understood. This project targeted the main property of cirrus clouds that influence climate processes; the ice fall speed. That is, this project improves the representation of the mass-weighted ice particle fall velocity, V m , in climate models, used to predict future climate on global and regional scales. Prior to 2007, the dominant sizes of ice particles in cirrus clouds were poorly understood, making it virtually impossible to predict how cirrus clouds interact with sunlight and thermal radiation. Due to several studies investigating the performance of optical probes used to measure the ice particle size distribution (PSD), as well as the remote sensing results from our last ARM project, it is now well established that the anomalously high concentrations of small ice crystals often reported prior to 2007 were measurement artifacts. Advances in the design and data processing of optical probes have greatly reduced these ice artifacts that resulted from the shattering of ice particles on the probe tips and/or inlet tube, and PSD measurements from one of these improved probes (the 2-dimensional Stereo or 2D-S probe) are utilized in this project to parameterize V m for climate models. Our original plan in the proposal was to parameterize the ice PSD (in terms of temperature and ice water content) and ice particle mass and projected area (in terms of mass- and area-dimensional power laws or m-D/A-D expressions) since these are the microphysical properties that determine V m , and then proceed to calculate V m from these parameterized properties. But the 2D-S probe directly measures ice particle projected area and indirectly estimates ice particle mass for each size bin. It soon became apparent that the original plan would introduce more uncertainty in the V m calculations than simply using the 2D-S measurements to directly calculate V m . By calculating V m directly from the measured PSD, ice particle projected area and estimated mass, more accurate estimates of V m are obtained. These V m values were then parameterized for climate models by relating them to (1) sampling temperature and ice water content (IWC) and (2) the effective diameter (D e ) of the ice PSD. Parameterization (1) is appropriate for climate models having single-moment microphysical schemes whereas (2) is appropriate for double-moment microphysical schemes and yields more accurate V m estimates. These parameterizations were developed for tropical cirrus clouds, Arctic cirrus, mid-latitude synoptic cirrus and mid-latitude anvil cirrus clouds based on field campaigns in these regions. An important but unexpected result of this research was the discovery of microphysical evidence indicating the mechanisms by which ice crystals are produced in cirrus clouds. This evidence, derived from PSD measurements, indicates that homogeneous freezing ice nucleation dominates in mid-latitude synoptic cirrus clouds, whereas heterogeneous ice nucleation processes dominate in mid-latitude anvil cirrus. Based on these findings, D e was parameterized in terms of temperature (T) for conditions dominated by (1) homo- and (2) heterogeneous ice nucleation. From this, an experiment was designed for global climate models (GCMs). The net radiative forcing from cirrus clouds may be affected by the means ice is produced (homo- or heterogeneously), and this net forcing contributes to climate sensitivity (i.e. the change in mean global surface temperature resulting from a doubling of CO 2 ). The objective of this GCM experiment was to determine how a change in ice nucleation mode affects the predicted global radiation balance. In the first simulation (Run 1), the D e -T relationship for homogeneous nucleation is used at all latitudes, while in the second simulation (Run 2), the D e -T relationship for heterogeneous nucleation is used at all latitudes. For both runs, V m is calculated from D e . Two GCMs were used; the Community Atmosphere Model version 5 (CAM5) and a European GCM known as ECHAM5 (thanks to our European colleagues who collaborated with us). Similar results were obtained from both GCMs in the Northern Hemisphere mid-latitudes, with a net cooling of ~ 1.0 W m -2 due to heterogeneous nucleation, relative to Run 1. The mean global net cooling was 2.4 W m -2 for the ECHAM5 GCM while CAM5 produced a mean global net cooling of about 0.8 W m -2 . This dependence of the radiation balance on nucleation mode is substantial when one considers the direct radiative forcing from a CO 2 doubling is 4 W m -2 . The differences between GCMs in mean global net cooling estimates may demonstrate a need for improving the representation of cirrus clouds in GCMs, including the coupling between microphysical and radiative properties. Unfortunately, after completing this GCM experiment, we learned from the company that provided the 2D-S microphysical data that the data was corrupted due to a computer program coding problem. Therefore the microphysical data had to be reprocessed and reanalyzed, and the GCM experiments were redone under our current ASR project but using an improved experimental design.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2013

An interface tracking model for droplet electrocoalescence.

This report describes an Early Career Laboratory Directed Research and Development (LDRD) project to develop an interface tracking model for droplet electrocoalescence. Many fluid-based technologies rely on electrical fields to control the motion of droplets, e.g. microfluidic devices for high-speed droplet sorting, solution separation for chemical detectors, and purification of biodiesel fuel. Precise control over droplets is crucial to these applications. However, electric fields can induce complex and unpredictable fluid dynamics. Recent experiments (Ristenpart et al. 2009) have demonstrated that oppositely charged droplets bounce rather than coalesce in the presence of strong electric fields. A transient aqueous bridge forms between approaching drops prior to pinch-off. This observation applies to many types of fluids, but neither theory nor experiments have been able to offer a satisfactory explanation. Analytic hydrodynamic approximations for interfaces become invalid near coalescence, and therefore detailed numerical simulations are necessary. This is a computationally challenging problem that involves tracking a moving interface and solving complex multi-physics and multi-scale dynamics, which are beyond the capabilities of most state-of-the-art simulations. An interface-tracking model for electro-coalescence can provide a new perspective to a variety of applications in which interfacial physics are coupled with electrodynamics, including electro-osmosis, fabrication of microelectronics, fuel atomization, oil dehydration, nuclear waste reprocessing and solution separation for chemical detectors. We present a conformal decomposition finite element (CDFEM) interface-tracking method for the electrohydrodynamics of two-phase flow to demonstrate electro-coalescence. CDFEM is a sharp interface method that decomposes elements along fluid-fluid boundaries and uses a level set function to represent the interface.

Erickson, Lindsay Crowl↗
DOE OSTI2013

Safeguards Approaches for Very Long-Term Storage of Spent Fuel

Many States are considering interim very long-term storage (VLTS) strategies for spent fuel storage while final disposition options are developed. Under the VLTS approach, spent fuel will be stored for periods that could exceed 50 to 100 years. The long-term storage of spent fuel — and the increasing stockpile of plutonium contained in this spent fuel — exacerbate the international safeguards and verification needs for spent fuel. In particular, significant delays in any verification of the contents of spent fuel assemblies prior to disposition may cause even further delays in the final disposition (or reprocessing) of spent fuel. As a result, the resources required by the International Atomic Energy Agency (IAEA) to safeguard the increasing stockpile of spent fuel will increase. Safeguarding these interim storage facilities and subsequently, the transportation of long-term stored fuel, will place additional demands on IAEA resources if the current approach, which relies heavily upon inspectors being present at the facility, is to be continued. Current IAEA approaches to safeguarding above ground dry storage sites do not take into account the challenges associated with longer-term storage and are also very dependent on inspector presence. The IAEA is moving toward remote and unattended monitoring systems for safeguards and these systems should be utilized at VLTS facilities as much as possible. The research and development of various monitoring systems should be continued.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Functionalized ultra-porous titania nanofiber membranes as nuclear waste separation and sequestration scaffolds for nuclear fuels recycle.

Advanced nuclear fuel cycle concept is interested in reducing separations to a simplified, one-step process if possible. This will benefit from the development of a one-step universal getter and sequestration material so as a simplified, universal waste form was proposed in this project. We have developed a technique combining a modified sol-gel chemistry and electrospinning for producing ultra-porous ceramic nanofiber membranes with controllable diameters and porous structures as the separation/sequestration materials. These ceramic nanofiber materials have been determined to have high porosity, permeability, loading capacity, and stability in extreme conditions. These porous fiber membranes were functionalized with silver nanoparticles and nanocrystal metal organic frameworks (MOFs) to introduce specific sites to capture gas species that are released during spent nuclear fuel reprocessing. Encapsulation into a durable waste form of ceramic composition was also demonstrated.

Liu, Haiqing↗
DOE OSTI2013

LWR First Recycle of TRU with Thorium Oxide for Transmutation and Cross Sections

Thorium has been considered as an option to uranium-based fuel, based on considerations of resource utilization (thorium is approximately three times more plentiful than uranium) and as a result of concerns about proliferation and waste management (e.g. reduced production of plutonium, etc.). Since the average composition of natural Thorium is dominated (100%) by the fertile isotope Th-232, Thorium is only useful as a resource for breeding new fissile materials, in this case U-233. Consequently a certain amount of fissile material must be present at the start-up of the reactor in order to guarantee its operation. The thorium fuel can be used in both once-through and recycle options, and in both fast and thermal spectrum systems. The present study has been aimed by the necessity of investigating the option of using reprocessed plutonium/TRU, from a once-through reference LEU scenario (50 GWd/ tIHM), mixed with natural thorium and the need of collect data (mass fractions, cross-sections etc.) for this particular fuel cycle scenario. As previously pointed out, the fissile plutonium is needed to guarantee the operation of the reactor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Opportunities for Process Monitoring Techniques at Delayed Access Facilities

Except for specific cases where the International Atomic Energy Agency (IAEA) maintains a continuous presence at a facility (such as the Japanese Rokkasho Reprocessing Plant), there is always a period of time or delay between the moment a State is notified or aware of an upcoming inspection, and the time the inspector actually enters the material balance area or facility. Termed by the authors as “delayed access,” this period of time between inspection notice and inspector entrance to a facility poses a concern. Delayed access also has the potential to reduce the effectiveness of measures applied as part of the Safeguards Approach for a facility (such as short-notice inspections). This report investigates the feasibility of using process monitoring to address safeguards challenges posed by delayed access at a subset of facility types.

Delayed access↗
DOE OSTI2013

Test Report for Permanganate and Cold Strontium Strike for Tank 241-AN-102

Tanks 241-AN-102 and 241-AN-107 supernatants contain soluble Sr-90 and transuranic elements that require removal prior to vitrification to comply with the Waste Treatment and Immobilization Plant immobilized low-activity waste specification (WTP Contract, DE-AC27-01RV 14136, Specification 2.2.2.8, "Radionuclide Concentration Limitations") and the U.S. Nuclear Regulatory Commission provisional agreement on waste incidental to reprocessing (letter, Paperiello, C. J., "Classification of Hanford Low-Activity Tank Waste Fraction"). These two tanks have high concentrations of organics and organic complexants and are referred to as complexant concentrate tanks. A precipitation process using sodium permanganate (NaMnO 4 ) and strontium nitrate (Sr(NO 3 ) 2 ) was developed and tested with tank waste samples to precipitate Sr-90 and transuranic elements from the supernate (PNWD-3141, Optimization of Sr/TRU Removal Conditions with Samples of AN-102 Tank Waste). Testing documented in this report was conducted to further evaluate the use of the strontium nitrate/sodium permanganate process in tank farms with a retention time of up to 12 months. Previous testing was focused on developing a process for deployment in the ultrafiltration vessels in the Waste Treatment and Immobilization Plant. This environment is different from tank farms in two important ways: the waste is diluted in the Waste Treatment and Immobilization Plant to ~5.5 M sodium, whereas the supernate in the tank farms is ~9 M Na. Secondly, while the Waste Treatment and Immobilization Plant allows for a maximum treatment time of hours to days, the in-tank farms treatment of tanks 241-AN102 and 241-AN-107 will result in a retention time of months (perhaps up to12 months) before processing. A comparative compilation of separation processes for Sr/transuranics has been published as RPP-RPT-48340, Evaluation of Alternative Strontium and Transuranic Separation Processes. This report also listed the testing needs for the permanganate precipitation process to be field-deployable. A more comprehensive listing of future testing needs to allow the process to be field deployable are contained in RPP-PLAN-51288, Development Test Plan for Sr/TRU Precipitation Process.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Assessment of Methods to Consolidate Iodine-Loaded Silver-Functionalized Silica Aerogel

The U.S. Department of Energy is currently investigating alternative sorbents for the removal and immobilization of radioiodine from the gas streams in a nuclear fuel reprocessing plant. One of these new sorbents, Ag 0 -functionalized silica aerogels, shows great promise as a potential replacement for Ag-bearing mordenites because of its high selectivity and sorption capacity for iodine. Moreover, a feasible consolidation of iodine-loaded Ag0-functionalized silica aerogels to a durable SiO 2 -based waste form makes this aerogel an attractive choice for sequestering radioiodine. This report provides a preliminary assessment of the methods that can be used to consolidate iodine-loaded Ag 0 -functionalized silica aerogels into a final waste form. In particular, it focuses on experimental investigation of densification of as prepared Ag 0 - functionalized silica aerogels powders, with or without organic moiety and with or without sintering additive (colloidal silica), with three commercially available techniques: 1) hot uniaxial pressing (HUP), 2) hot isostatic pressing (HIP), and 3) spark plasma sintering (SPS). The densified products were evaluated with helium gas pycnometer for apparent density, with the Archimedes method for apparent density and open porosity, and with high-resolution scanning electron microscopy and energy dispersive spectroscopy (SEM-EDS) for the extent of densification and distribution of individual elements. The preliminary investigation of HUP, HIP, and SPS showed that these sintering methods can effectively consolidate powders of Ag 0 -functionalized silica aerogel into products of near-theoretical density. Also, removal of organic moiety and adding 5.6 mass% of colloidal silica to Ag0-functionalized silica aerogel powders before processing provided denser products. Furthermore, the ram travel data for SPS indicated that rapid consolidation of powders can be performed at temperatures below 950°C.

Ag-Functionalized Silica Aerogel↗
DOE OSTI2013

Summary Report on the Fuel Performance Modeling of the AFC-2A, 2B Irradiation Experiments

The primary objective of this work at the Idaho National Laboratory (INL) is to determine the fuel and cladding temperature history during irradiation of the AFC-2A, 2B transmutation metallic fuel alloy irradiation experiments containing transuranic and rare earth elements. Addition of the rare earth elements intends to simulate potential fission product carry-over from pyro-metallurgical reprocessing. Post irradiation examination of the AFC-2A, 2B rodlets revealed breaches in the rodlets and fuel melting which was attributed to the release of the fission gas into the helium gap between the rodlet cladding and the capsule which houses six individually encapsulated rodlets. This release is not anticipated during nominal operation of the AFC irradiation vehicle that features a double encapsulated design in which sodium bonded metallic fuel is separated from the ATR coolant by the cladding and the capsule walls. The modeling effort is focused on assessing effects of this unanticipated event on the fuel and cladding temperature with an objective to compare calculated results with the temperature limits of the fuel and the cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Impact of Fission Products Impurity on the Plutonium Content of Metal- and Oxide- Fuels in Sodium Cooled Fast Reactors

This short report presents the neutronic analysis to evaluate the impact of fission product impurity on the Pu content of Sodium-cooled Fast Reactor (SFR) metal- and oxide- fuel fabrication. The similar work has been previously done for PWR MOX fuel. The analysis will be performed based on the assumption that the separation of the fission products (FP) during the reprocessing of UOX spent nuclear fuel assemblies is not perfect and that, consequently, a certain amount of FP goes into the Pu stream used to fabricate SFR fuels. Only non-gaseous FPs have been considered (see the list of 176 isotopes considered in the calculations in Appendix 1 of Reference 1). Throughout of this report, we define the mixture of Pu and FPs as Pu FP . The main objective of this analysis is to quantify the increase of the Pu content of SFR fuels necessary to maintain the same average burnup at discharge independently of the amount of FP in the Pu stream, i.e. independently of the Pu FP composition. The FP losses are considered element-independent, i.e., for example, 1% of FP losses mean that 1% of all non-gaseous FP leak into the Pu stream.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Fukushima Daiichi Information Repository FY13 Status

The accident at the Fukushima Daiichi nuclear power station in Japan is one of the most serious in commercial nuclear power plant operating history. Much will be learned that may be applicable to the U.S. reactor fleet, nuclear fuel cycle facilities, and supporting systems, and the international reactor fleet. For example, lessons from Fukushima Daiichi may be applied to emergency response planning, reactor operator training, accident scenario modeling, human factors engineering, radiation protection, and accident mitigation; as well as influence U.S. policies towards the nuclear fuel cycle including power generation, and spent fuel storage, reprocessing, and disposal. This document describes the database used to establish a centralized information repository to store and manage the Fukushima data that has been gathered. The data is stored in a secured (password protected and encrypted) repository that is searchable and available to researchers at diverse locations.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2014

Laboratory Evaporation Testing Of Hanford Waste Treatment Plant Low Activity Waste Off-Gas Condensate Simulant

The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream, LAW Off-Gas Condensate, from the off-gas system. The baseline plan for disposition of this stream is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility again. Alternate disposition of this stream would eliminate recycling of problematic components, and would enable de-coupled operation of the LAW melter and the Pretreatment Facilities. Eliminating this stream from recycling within WTP would also decrease the LAW vitrification mission duration and quantity of canistered glass waste forms. This LAW Off-Gas Condensate stream contains components that are volatile at melter temperatures and are problematic for the glass waste form. Because this stream recycles within WTP, these components accumulate in the Condensate stream, exacerbating their impact on the number of LAW glass containers that must be produced. Approximately 32% of the sodium in Supplemental LAW comes from glass formers used to make the extra glass to dilute the halides to be within acceptable concentration ranges in the LAW glass. Diverting the stream reduces the halides in the recycled Condensate and is a key outcome of this work. Additionally, under possible scenarios where the LAW vitrification facility commences operation prior to the WTP Pretreatment facility, identifying a disposition path becomes vitally important. This task examines the impact of potential future disposition of this stream in the Hanford tank farms, and investigates auxiliary evaporation to enable another disposition path. Unless an auxiliary evaporator is used, returning the stream to the tank farms would require evaporation in the 242-A evaporator. This stream is expected to be unusual because it will be very high in corrosive species that are volatile in the melter (chloride, fluoride, sulfur), will have high ammonia, and will contain carryover particulates of glass-former chemicals. These species have potential to cause corrosion of tanks and equipment, precipitation of solids, release of ammonia gas vapors, and scale in the tank farm evaporator. Routing this stream to the tank farms does not permanently divert it from recycling into the WTP, only temporarily stores it prior to reprocessing. Testing is normally performed to demonstrate acceptable conditions and limits for these compounds in wastes sent to the tank farms. The primary parameter of this phase of the test program was measuring the formation of solids during evaporation in order to assess the compatibility of the stream with the evaporator and transfer and storage equipment. The origin of this LAW Off-Gas Condensate stream will be the liquids from the Submerged Bed Scrubber (SBS) and the Wet Electrostatic Precipitator (WESP) from the LAW facility melter offgas system. The stream is expected to be a dilute salt solution with near neutral pH, and will likely contain some insoluble solids from melter carryover. The soluble components are expected to be mostly sodium and ammonium salts of nitrate, chloride, and fluoride. This stream has not been generated yet, and, thus, the composition will not be available until the WTP begins operation, but a simulant has been produced based on models, calculations, and comparison with pilot-scale tests. This report discusses results of evaporation testing of the simulant. Two conditions were tested, one with the simulant at near neutral pH, and a second at alkaline pH. The neutral pH test is comparable to the conditions in the Hanford Effluent Treatment Facility (ETF) evaporator, although that evaporator operates at near atmospheric pressure and tests were done under vacuum. For the alkaline test, the target pH was based on the tank farm corrosion control program requirements, and the test protocol and equipment was comparable to that used for routine evaluation of feed compatibility studies for the 242-A evaporator. One of the radionuclides that is volatile in the melter and expected to be in high concentration in this LAW Off-Gas Condensate stream is Technetium-99 (99Tc). Technetium will not be removed from the aqueous waste in the Hanford WTP, and will primarily end up immobilized in the LAW glass by repeated recycle of the off-gas condensate into the LAW melter. Other radionuclides that are also expected to be in appreciable concentrations in the LAW Off-Gas Condensate are 129I, 90Sr, 137Cs, and 241Am. The concentrations of these radionuclides in this stream will be much lower than in the LAW, but they will still be higher than limits for some of the other disposition pathways currently available. At this time, these scoping tests did not evaluate the partitioning of the radionuclides to the evaporator condensate, since ample data are available separately from other experience in the DOE complex. Results from the evaporation testing show that the neutral SBS simulant first forms turbidity at ~7.5X concentration, while the alkaline-adjusted simulant became turbid at ~3X concentration. The major solid in both cases was Kogarkoite, Na3FSO4. Sodium and lithium fluorides were also detected. Minimal solids were formed in the evaporator bottoms until a substantial fraction of liquid was removed, indicating that evaporation could minimize storage volume issues. Achievable concentration factors without significant insoluble solids were 17X at alkaline pH, and 23X at neutral pH. In both runs, significant ammonia carried over and was captured in the condenser with the water condensate. Results also indicate that with low insoluble solids formation in the initial testing at neutral pH, the use of Reverse Osmosis is a potential alternate method for concentrating the solution, although an evaluation is needed to identify equipment that can tolerate insoluble solids. Most of the ammonia remains in the evaporator bottoms during the neutral pH evaporation, but partitions to the condensate during alkaline evaporation. Disposition of both streams needs to consider the management of ammonia vapor and its release. Since this is an initial phase of testing, additional tasks related to evaporation methods are expected to be identified for development. These tasks likely include evaluation and testing of composition variability testing and evaluations, corrosion and erosion testing, slurry storage and immobilization investigations, and evaporator condensate disposition.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Advanced Seismic Data Analysis Program (The Hot Pot Project), DOE Award: DE-EE0002839 (Phase 1 Report)

A five-line (23 mile) reflection- seismic survey was conducted at the Hot Pot geothermal prospect area in north-central Nevada under the USDOE (United States Department of Energy) Geothermal Technologies Program. The project objective was to utilize innovative seismic data processing, integrated with existing geological, geophysical and geochemical information, to identify high-potential drilling targets and to reduce drilling risk. Data acquisition and interpretation took place between October 2010 and April 2011. The first round of data processing resulted in large areas of relatively poor data, and obvious reflectors known from existing subsurface information either did not appear on the seismic profiles or appeared at the wrong depth. To resolve these issues, the velocity model was adjusted to include geologic input, and the lines were reprocessed. The resulting products were significantly improved, and additional detail was recovered within the high-velocity and in part acoustically isotropic basement. Features visible on the improved seismic images include interpreted low angle thrust faults within the Paleozoic Valmy Formation, which potentially are reactivated in the current stress field. Intermediate-depth wells are currently targeted to test these features. The seismic images also suggest the existence of Paleogene sedimentary and volcanic rocks which potentially may function as a near-surface reservoir, charged by deeper structures in Paleozoic rocks.

15 GEOTHERMAL ENERGY↗