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All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Implementation of flowsheet change to minimize hydrogen and ammonia generation during chemical processing of high level waste in the Defense Waste Processing Facility

Testing was completed to develop a chemical processing flowsheet for the Defense Waste Processing Facility (DWPF), designed to vitrify and stabilize high level radioactive waste. DWPF processing uses a reducing acid (formic acid) and an oxidizing acid (nitric acid) to rheologically thin the slurry and complete the necessary acid base and reduction reactions (primarily mercury and manganese). Formic acid reduces mercuric oxide to elemental mercury, allowing the mercury to be removed during the boiling phase of processing through steam stripping. In runs with active catalysts, formic acid can decompose to hydrogen and nitrate can be reduced to ammonia, both flammable gases, due to rhodium and ruthenium catalysis. Replacement of formic acid with glycolic acid eliminates the generation of rhodium- and ruthenium-catalyzed hydrogen and ammonia. In addition, mercury reduction is still effective with glycolic acid. Hydrogen, ammonia and mercury are discussed in the body of the report. Ten abbreviated tests were completed to develop the operating window for implementation of the flowsheet and determine the impact of changes in acid stoichiometry and the blend of nitric and glycolic acid as it impacts various processing variables over a wide processing region. Three full-length 4-L lab-scale simulations demonstrated the viability of the flowsheet under planned operating conditions. The flowsheet is planned for implementation in early 2017.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Supplemental Thermal-Hydraulic Transient Analyses of BR2 in Support of Conversion to LEU Fuel

Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The RELAP5/Mod 3.3 code has been used to perform transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. A RELAP5 model of BR2 has been validated against select transient BR2 reactor experiments performed in 1963 by showing agreement with measured cladding temperatures. Following the validation, the RELAP5 model was then updated to represent the current use of the reactor; taking into account core configuration, neutronic parameters, trip settings, component changes, etc. Simulations of the 1963 experiments were repeated with this updated model to re-evaluate the boiling risks associated with the currently allowed maximum heat flux limit of 470 W/cm 2 and temporary heat flux limit of 600 W/cm 2 . This document provides analysis of additional transient simulations that are required as part of a modern BR2 safety analysis report (SAR). The additional simulations included in this report are effect of pool temperature, reduced steady-state flow rate, in-pool loss of coolant accidents, and loss of external cooling. The simulations described in this document have been performed for both an HEU- and LEU-fueled core.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

U.S. Commercial Spent Nuclear Fuel Assembly Characteristics - 1968-2013

Activities related to management of spent nuclear fuel (SNF) are increasing in the US and many other countries. Over 240,000 SNF assemblies have been discharged from US commercial reactors since the late 1960s. The enrichment and burnup of SNF have changed significantly over the past 40 years, and fuel assembly designs have also evolved. Understanding the general characteristics of SNF helps regulators and other stakeholders form overall strategies towards the final disposal of US SNF. This report documents a survey of all US commercial SNF assemblies in the GC-859 database and provides reference SNF source terms (e.g., nuclide inventories, decay heat, and neutron/photon emission) at various cooling times up to 200 years after fuel discharge. This study reviews the distribution and evolution of fuel parameters of all SNF assemblies discharged over the past 40 years. Assemblies were categorized into three groups based on discharge year, and the median burnups and enrichments of each group were used to establish representative cases. An extended burnup case was created for boiling water reactor (BWR) fuels, and another was created for the pressurized water reactor (PWR) fuels. Two additional cases were developed to represent the eight mixed oxide (MOX) fuel assemblies in the database. Burnup calculations were performed for each representative case. Realistic parameters for fuel design and operations were used to model the SNF and to provide reference fuel characteristics representative of the current inventory. Burnup calculations were performed using the ORIGEN code, which is part of the SCALE nuclear modeling and simulation code system. Results include total activity, decay heat, photon emission, neutron flux, gamma heat, and plutonium content, as well as concentrations for 115 significant nuclides. These quantities are important in the design, regulation, and operations of SNF storage, transportation, and disposal systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

US Efforts in Support of Examinations at Fukushima Daiichi – 2016 Evaluations

Although it is clear that the accident signatures from each unit at the Fukushima Daiichi Nuclear Power Station (NPS) [Daiichi] differ, much is not known about the end-state of core materials within these units. Some of this uncertainty can be attributed to a lack of information related to cooling system operation and cooling water injection. There is also uncertainty in our understanding of phenomena affecting: a) in-vessel core damage progression during severe accidents in boiling water reactors (BWRs), and b) accident progression after vessel failure (ex-vessel progression) for BWRs and Pressurized Water Reactors (PWRs). These uncertainties arise due to limited full scale prototypic data. Similar to what occurred after the accident at Three Mile Island Unit 2, these Daiichi units offer the international community a means to reduce such uncertainties by obtaining prototypic data from multiple full-scale BWR severe accidents. Information obtained from Daiichi is required to inform Decontamination and Decommissioning activities, improving the ability of the Tokyo Electric Power Company Holdings (TEPCO) to characterize potential hazards and to ensure the safety of workers involved with cleanup activities. This document reports recent results from the US Forensics Effort to use information obtained by TEPCO to enhance the safety of existing and future nuclear power plant designs. This Forensics Effort, which is sponsored by the Reactor Safety Technologies Pathway of the Department of Energy Office of Nuclear Energy Light Water Reactor (LWR) Sustainability Program, consists of a group of US experts in LWR safety and plant operations that have identified examination needs and are evaluating TEPCO information from Daiichi that address these needs. Examples presented in this report demonstrate that significant safety insights are being obtained in the areas of component performance, fission product release and transport, debris end-state location, and combustible gas generation and transport. In addition to reducing uncertainties related to severe accident modeling progression, these insights are being used to update guidance for severe accident prevention, mitigation, and emergency planning. Furthermore, reduced uncertainties in modeling the events at Daiichi will improve the realism of reactor safety evaluations and inform future D&D activities by improving the capability for characterizing potential hazards to workers involved with cleanup activities.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Improved Accident Tolerance of Austenitic Stainless Steel Cladding through Colossal Supersaturation with Interstitial Solutes

We proposed a program-supporting research project in the area of fuel-cycle R&D, specifically on the topic of advanced fuels. Our goal was to investigate whether SECIS (surface engineering by concentrated interstitial solute – carbon, nitrogen) can improve the properties of austenitic stainless steels and related structural alloys such that they can be used for nuclear fuel cladding in LWRs (light-water reactors) and significantly excel currently used alloys with regard to performance, safety, service life, and accident tolerance. We intended to demonstrate that SECIS can be adapted for post-processing of clad tubing to significantly enhance mechanical properties (hardness, wear resistance, and fatigue life), corrosion resistance, resistance to stress–corrosion cracking (hydrogen-induced embrittlement), and – potentially – radiation resistance (against electron-, neutron-, or ion-radiation damage). To test this hypothesis, we measured various relevant properties of the surface-engineered alloys and compared them with corresponding properties of the non–treated, as-received alloys. In particular, we studied the impact of heat exposure corresponding to BWR (boiling-water reactor) working and accident (loss-of-coolant) conditions and the effect of ion irradiation.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Microscopic Fuel Particles Produced by Self-Assembly of Actinide Nanoclusters on Carbon Nanomaterials

Many consider further development of nuclear power to be essential for sustained development of society; however, the fuel forms currently used are expensive to recycle. In this project, we sought to create the knowledge and knowhow that are needed to produce nanocomposite materials by directly depositing uranium nanoclusters on networks of carbon-­ based nanomaterials. The objectives of the proposed work were to (1) determine the control of uranium nanocluster surface chemistry on nanocomposite formation, (2) determine the control of carbon nanomaterial surface chemistry on nanocomposite formation, and (3) develop protocols for synthesizing uranium-­carbon nanomaterials. After examining a wide variety of synthetic methods, we show that synthesizing graphene-­supported UO 2 nanocrystals in polar ethylene glycol compounds by polyol reduction under boiling reflux can enable the use of an inexpensive graphene precursor graphene oxide in the production of uranium-carbon nanocomposites in a one-­pot process. We further show that triethylene glycol is the most suitable solvent for producing nanometer-­sized UO 2 crystals compared to monoethylene glycol, diethylene glycol, and polyethylene glycol. Graphene-­supported UO 2 nanocrystals synthesized with triethylene glycol show evidence of heteroepitaxy, which can be beneficial for facilitating heat transfer in nuclear fuel particles. Furthermore, we show that graphene-supported UO 2 nanocrystals synthesized by polyol reduction can be readily stored in alcohols, preventing oxidation from the prevalent oxygen in air. Together, these methods provide a facile approach for preparing and storing graphene-supported UO nanocrystals for further investigation and development under ambient conditions.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Sludge batch 9 simulant runs using the nitric-glycolic acid flowsheet

Testing was completed to develop a Sludge Batch 9 (SB9) nitric-glycolic acid chemical process flowsheet for the Defense Waste Processing Facility’s (DWPF) Chemical Process Cell (CPC). CPC simulations were completed using SB9 sludge simulant, Strip Effluent Feed Tank (SEFT) simulant and Precipitate Reactor Feed Tank (PRFT) simulant. Ten sludge-only Sludge Receipt and Adjustment Tank (SRAT) cycles and four SRAT/Slurry Mix Evaporator (SME) cycles, and one actual SB9 sludge (SRAT/SME cycle) were completed. As has been demonstrated in over 100 simulations, the replacement of formic acid with glycolic acid virtually eliminates the CPC’s largest flammability hazards, hydrogen and ammonia. Recommended processing conditions are summarized in section 3.5.1. Testing demonstrated that the interim chemistry and Reduction/Oxidation (REDOX) equations are sufficient to predict the composition of DWPF SRAT product and SME product. Additional reports will finalize the chemistry and REDOX equations. Additional testing developed an antifoam strategy to minimize the hexamethyldisiloxane (HMDSO) peak at boiling, while controlling foam based on testing with simulant and actual waste. Implementation of the nitric-glycolic acid flowsheet in DWPF is recommended. This flowsheet not only eliminates the hydrogen and ammonia hazards but will lead to shorter processing times, higher elemental mercury recovery, and more concentrated SRAT and SME products. The steady pH profile is expected to provide flexibility in processing the high volume of strip effluent expected once the Salt Waste Processing Facility starts up.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2014

Advanced Heat/Mass Exchanger Technology for Geothermal and Solar Renewable Energy Systems

Northern Nevada has abundant geothermal and solar energy resources, and these renewable energy sources provide an ample opportunity to produce economically viable power. Heat/mass exchangers are essential components to any energy conversion system. Improvements in the heat/mass exchange process will lead to smaller, less costly (more efficient) systems. There is an emerging heat transfer technology, based on micro/nano/molecular-scale surface science that can be applied to heat/mass exchanger design. The objective is to develop and characterize unique coating materials, surface configurations and membranes capable of accommodating a 10-fold increase in heat/mass exchanger performance via phase change processes (boiling, condensation, etc.) and single phase convective heat/mass transfer.

14 SOLAR ENERGY↗
DOE OSTI2016

BWR Steam Dryer Alternating Stress Assessment Procedures

This report presents an overview of Boiling Water Reactor (BWR) steam dryer design; the fatigue cracking failures that occurred at the Quad Cities (QC) plants and their root causes; a history of BWR Extended Power Uprates (EPUs) in the USA; and a discussion of steam dryer modifications/replacements, alternating stress mechanisms on steam dryers, and structural integrity evaluations (static and alternating stress).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

ESTIMATING WATER FOOTPRINT AND MANAGING BIOREFINERY WASTEWATER IN THE PRODUCTION OF BIO-BASED RENEWABLE DIESEL BLENDSTOCK

This analysis covers the entire biorefinery operation. The study focuses on net water consumed for the production of a unit of biofuel: blue, green, and grey water footprint. Blue water is defined as the water consumed in the biorefinery that is withdrawn from surface and ground water. Blue water footprint includes enzyme cultivation, pretreatment, hydrolysis, bioreactor, cooling system, boiler, fuel upgrading, combustor track, and on-site WWT. Grey water is defined as wastewater generated from the biorefinery and was evaluated based on the wastewater treatment plant design. Green water, defined as rainwater consumed for the production, is not required in the RDB process. Approximately 7–15 gal of water are required to produce a gallon of RDB when corn stover or non-irrigated perennial grasses, switchgrass and Miscanthus x giganteus (Miscanthus), serve as the feedstock in the contiguous United States. Bioelectricity generation from the biorefinery resulted in a net water credit, which reduced the water footprint. The life cycle grey water footprint for nitrogen is primarily from nitrogen in the feedstock production stage because no wastewater is discharged into the environment in the RDB process. Perennial grasses-based RDB production shows a promising grey water footprint, while corn stover-based RDB production has a relatively low green water footprint. Results from the study can help improve our understanding of the water sustainability of advanced biofuel technology under development. Make-up water for cooling and boiling remains a major demand in the biorefinery. The work revealed a key issue or trade-off between achieving zero liquid discharge to maximize water resource use and potentially increasing cost of fuel production. Solid waste disposal was identified as a management issue, and its inverse relationship with wastewater management could affect economic sustainability.

09 BIOMASS FUELS↗
DOE OSTI2015

L3.PHI.CTF.P10.02-rev2 Coupling of Subchannel T/H (CTF) and CRUD Chemistry (MAMBA1D)

The purpose of this milestone is to create a preliminary capability for modeling light water reactor (LWR) thermal-hydraulic (T/H) and CRUD growth using the CTF subchannel code and the subgrid version of the MAMBA CRUD chemistry code, MAMBA1D. In part, this is a follow-on to Milestone L3.PHI.VCS.P9.01, which is documented in Report CASL-U-2014-0188-000, titled "Development of CTF Capability for Modeling Reactor Operating Cycles with Crud Growth". As the title suggests, the previous milestone set up a framework for modeling reactor operation cycles with CTF. The framework also facilitated coupling to a CRUD chemistry capability for modeling CRUD growth throughout the reactor operating cycle. To demonstrate the capability, a simple CRUD \surrogate" tool was developed and coupled to CTF; however, it was noted that CRUD growth predictions by the surrogate were not considered realistic. This milestone builds on L3.PHI.VCS.P9.01 by replacing this simple surrogate tool with the more advanced MAMBA1D CRUD chemistry code. Completing this task involves addressing unresolved tasks from Milestone L3.PHI.VCS.P9.01, setting up an interface to MAMBA1D, and extracting new T/H information from CTF that was not previously required in the simple surrogate tool. Speci c challenges encountered during this milestone include (1) treatment of the CRUD erosion model, which requires local turbulent kinetic energy (TKE) (a value that CTF does not calculate) and (2) treatment of the MAMBA1D CRUD chimney boiling model in the CTF rod heat transfer solution. To demonstrate this new T/H, CRUD modeling capability, two sets of simulations were performed: (1) an 18 month cycle simulation of a quarter symmetry model of Watts Bar and (2) a simulation of Assemblies G69 and G70 from Seabrook Cycle 5. The Watts Bar simulation is merely a demonstration of the capability. The simulation of the Seabrook cycle, which had experienced CRUD-related fuel rod failures, had actual CRUD-scrape data to compare with results. As results show, the initial CTF/MAMBA1D-predicted CRUD thicknesses were about half of their expected values, so further investigation will be required for this simulation.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Investigation of Natural Circulation Instability and Transients in Passively Safe Small Modular Reactors

The NEUP funded project, NEUP-3496, aims to experimentally investigate two-phase natural circulation flow instability that could occur in Small Modular Reactors (SMRs), especially for natural circulation SMRs. The objective has been achieved by systematically performing tests to study the general natural circulation instability characteristics and the natural circulation behavior under start-up or design basis accident conditions. Experimental data sets highlighting the effect of void reactivity feedback as well as the effect of power ramp-up rate and system pressure have been used to develop a comprehensive stability map. The safety analysis code, RELAP5, has been used to evaluate experimental results and models. Improvements to the constitutive relations for flashing have been made in order to develop a reliable analysis tool. This research has been focusing on two generic SMR designs, i.e. a small modular Simplified Boiling Water Reactor (SBWR) like design and a small integral Pressurized Water Reactor (PWR) like design. A BWR-type natural circulation test facility was firstly built based on the three-level scaling analysis of the Purdue Novel Modular Reactor (NMR) with an electric output of 50 MWe, namely NMR-50, which represents a BWR-type SMR with a significantly reduced reactor pressure vessel (RPV) height. The experimental facility was installed with various equipment to measure thermalhydraulic parameters such as pressure, temperature, mass flow rate and void fraction. Characterization tests were performed before the startup transient tests and quasi-steady tests to determine the loop flow resistance. The control system and data acquisition system were programmed with LabVIEW to realize the realtime control and data storage. The thermal-hydraulic and nuclear coupled startup transients were performed to investigate the flow instabilities at low pressure and low power conditions for NMR-50. Two different power ramps were chosen to study the effect of startup power density on the flow instability. The experimental startup transient results showed the existence of three different flow instability mechanisms, i.e., flashing instability, condensation induced flow instability, and density wave oscillations. In addition, the void-reactivity feedback did not have significant effects on the flow instability during the startup transients for NMR-50. ii Several initial startup procedures with different power ramp rates were experimentally investigated to eliminate the flow instabilities observed from the startup transients. Particularly, the very slow startup transient and pressurized startup transient tests were performed and compared. It was found that the very slow startup transients by applying very small power density can eliminate the flashing oscillations in the single-phase natural circulation and stabilize the flow oscillations in the phase of net vapor generation. The initially pressurized startup procedure was tested to eliminate the flashing instability during the startup transients as well. The pressurized startup procedure included the initial pressurization, heat-up, and venting process. The startup transient tests showed that the pressurized startup procedure could eliminate the flow instability during the transition from single-phase flow to two-phase flow at low pressure conditions. The experimental results indicated that both startup procedures were applicable to the initial startup of NMR. However, the pressurized startup procedures might be preferred due to short operating hours required. In order to have a deeper understanding of natural circulation flow instability, the quasi-steady tests were performed using the test facility installed with preheater and subcooler. The effect of system pressure, core inlet subcooling, core power density, inlet flow resistance coefficient, and void reactivity feedback were investigated in the quasi-steady state tests. The experimental stability boundaries were determined between unstable and stable flow conditions in the dimensionless stability plane of inlet subcooling number and Zuber number. To predict the stability boundary theoretically, linear stability analysis in the frequency domain was performed at four sections of the natural circulation test loop. The flashing phenomena in the chimney section was considered as an axially uniform heat source. And the dimensionless characteristic equation of the pressure drop perturbation was obtained by considering the void fraction effect and outlet flow resistance in the core section. The theoretical flashing boundary showed some discrepancies with previous experimental data from the quasi-steady state tests. In the future, thermal non-equilibrium was recommended to improve the accuracy of flashing instability boundary. As another part of the funded research, flow instabilities of a PWR-type SMR under low pressure and low power conditions were investigated experimentally as well. The NuScale reactor design was selected as the prototype for the PWR-type SMR. In order to experimentally study the natural circulation behavior of NuScale iii reactor during accidental scenarios, detailed scaling analyses are necessary to ensure that the scaled phenomena could be obtained in a laboratory test facility. The three-level scaling method is used as well to obtain the scaling ratios derived from various non-dimensional numbers. The design of the ideally scaled facility (ISF) was initially accomplished based on these scaling ratios. Then the engineering scaled facility (ESF) was designed and constructed based on the ISF by considering engineering limitations including laboratory space, pipe size, and pipe connections etc. PWR-type SMR experiments were performed in this well-scaled test facility to investigate the potential thermal hydraulic flow instability during the blowdown events, which might occur during the loss of coolant accident (LOCA) and loss of heat sink accident (LOHS) of the prototype PWR-type SMR. Two kinds of experiments, normal blowdown event and cold blowdown event, were experimentally investigated and compared with code predictions. The normal blowdown event was experimentally simulated since an initial condition where the pressure was lower than the designed pressure of the experiment facility, while the code prediction of blowdown started from the normal operation condition. Important thermal hydraulic parameters including reactor pressure vessel (RPV) pressure, containment pressure, local void fraction and temperature, pressure drop and natural circulation flow rate were measured and analyzed during the blowdown event. The pressure and water level transients are similar to the experimental results published by NuScale [51], which proves the capability of current loop in simulating the thermal hydraulic transient of real PWR-type SMR. During the 20000s blowdown experiment, water level in the core was always above the active fuel assemble during the experiment and proved the safety of natural circulation cooling and water recycling design of PWR-type SMR. Besides, pressure, temperature, and water level transient can be accurately predicted by RELAP5 code. However, the oscillations of natural circulation flow rate, water level and pressure drops were observed during the blowdown transients. This kind of flow oscillations are related to the water level and the location upper plenum, which is a path for coolant flow from chimney to steam generator and down comer. In order to investigate the transients start from the opening of ADS valve in both experimental and numerical way, the cold blow-down experiment is conducted. For the cold blowdown event, different from setting both reactor iv pressure vessel (RPV) and containment at high temperature and pressure, only RPV was heated close to the highest designed pressure and then open the ADS valve, same process was predicted using RELAP5 code. By doing cold blowdown experiment, the entire transients from the opening of ADS can be investigated by code and benchmarked with experimental data. Similar flow instability observed in the cold blowdown experiment. The comparison between code prediction and experiment data showed that the RELAP5 code can successfully predict the pressure void fraction and temperature transient during the cold blowdown event with limited error, but numerical instability exists in predicting natural circulation flow rate. Besides, the code is lack of capability in predicting the water level related flow instability observed in experiments.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Geothermometry Mapping of Deep Hydrothermal Reservoirs in Southeastern Idaho: Final Report

The Eastern Snake River Plain (ESRP) in southern Idaho is a region of high heat flow. Sustained volcanic activities in the wake of the passage of Yellowstone Hotspot have turned this region into an area with great potential for geothermal resources. Numerous hot springs with temperatures up to 75 ºC are scattered along the margins of the plain. Similarly, several hot-water producing wells and few hot springs are also present within the plain. The geothermal reservoirs in the area are likely to be hosted at depth in the felsic volcanic rocks underneath the thick sequences of basalts within the plain and the Paleozoic rocks underneath both basalts and felsic volcanic rocks along the margins. The heat source to these geothermal resources is thought to be the mid-crustal sill complex which sustains high heat flow in the ESRP. Several thermal anomaly areas are believed to be associated with the local thermal perturbation because of the presence of favorable structural settings. However, it is hypothesized that the pervasive presence of an overlying groundwater aquifer in the region effectively masks thermal signatures of deep-seated geothermal resources. The dilution of deeper thermal water and re-equilibration at lower temperatures are significant challenges for the evaluation of potential resource areas in the ESRP. To address this issue, this project, led by the Idaho National Laboratory (INL), aimed at applying advanced geothermometry tools including temperature-dependent mineral and isotopic equilibria with mixing models that account for processes such as boiling and dilution with shallow groundwater that could affect calculated temperatures of underlying deep thermal waters. Over the past several years, we collected approximately 100 water samples from springs/wells for chemical analysis as well as assembled existing water chemistry data from literature. We applied several geothermometric and geochemical modeling tools to the compositions of ESRP water samples. Geothermometric calculations based on principle of multicomponent equilibrium geothermometry with inverse geochemical modeling capability (e.g., Reservoir Temperature Estimator, RTEst) have been useful for evaluation of reservoir temperatures. Similarly, sulfate-water oxygen isotope geothermometry was also applied to several samples in tandem with RTEst. In summary, geothermometric calculations of ESRP thermal water samples indicated numerous potential geothermal areas with elevated reservoir temperatures. Specifically, areas around southern/southwestern side of the Mount Bennet Hills and within the Camas Prairie in the southwestern portion of the ESRP suggest temperatures 140-190 °C. In the northern portion of the ESRP, Lidy Hot Springs, Ashton, Newdale, and areas east of Idaho Falls have expected reservoir temperature ?140 °C. In the southern ERSP, areas near Buhl and Twin Falls are found to have elevated temperatures as high as 160 °C. These areas are likely to host potentially economic geothermal resources; however, further detailed study is warranted to each site to evaluate hydrothermal suitability for economic use.

58 GEOSCIENCES↗
DOE OSTI2017

Impact of Salt Waste Processing Facility Streams on the Nitric-Glycolic Flowsheet in the Chemical Processing Cell

An evaluation of the previous Chemical Processing Cell (CPC) testing was performed to determine whether the planned concurrent operation, or “coupled” operations, of the Defense Waste Processing Facility (DWPF) with the Salt Waste Processing Facility (SWPF) has been adequately covered. Tests with the nitricglycolic acid flowsheet, which were both coupled and uncoupled with salt waste streams, included several tests that required extended boiling times. This report provides the evaluation of previous testing and the testing recommendation requested by Savannah River Remediation. The focus of the evaluation was impact on flammability in CPC vessels (i.e., hydrogen generation rate, SWPF solvent components, antifoam degradation products) and processing impacts (i.e., acid window, melter feed target, rheological properties, antifoam requirements, and chemical composition).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Water NSTF Design, Instrumentation, and Test Planning

The following report serves as a formal introduction to the water-based Natural convection Shutdown heat removal Test Facility (NSTF) program at Argonne. Since 2005, this US Department of Energy (DOE) sponsored program has conducted large scale experimental testing to generate high-quality and traceable validation data for guiding design decisions of the Reactor Cavity Cooling System (RCCS) concept for advanced reactor designs. The most recent facility iteration, and focus of this report, is the operation of a 1/2 scale model of a water-RCCS concept. Several features of the NSTF prototype align with the conceptual design that has been publicly released for the AREVA 625 MWt SC-HTGR. The design of the NSTF also retains all aspects common to a fundamental boiling water thermosiphon, and thus is well poised to provide necessary experimental data to advance basic understanding of natural circulation phenomena and contribute to computer code validation. Overall, the NSTF program operates to support the DOE vision of aiding US vendors in design choices of future reactor concepts, advancing the maturity of codes for licensing, and ultimately developing safe and reliable reactor technologies. In this report, the top-level program objectives, testing requirements, and unique considerations for the water cooled test assembly are discussed, and presented in sufficient depth to support defining the program’s overall scope and purpose. A discussion of the proposed 6-year testing program is then introduced, which outlines the specific strategy and testing plan for facility operations. The proposed testing plan has been developed to meet the toplevel objective of conducting high-quality test operations that span across a broad range of single- and two-phase operating conditions. Details of characterization, baseline test cases, accident scenario, and parametric variations are provided, including discussions of later-stage test cases that examine the influence of geometric variations and off-normal configurations. The facility design follows, including as-built dimensions and specifications of the various mechanical and liquid systems, design choices for the test section, water storage tank, and network piping. Specifications of the instrumentation suite are then presented, along with specific information on performance windows, measurement uncertainties, and installation locations. Finally, descriptions of the control systems and heat removal networks are provided, which have been engineered to support precise quantification of energy balances and facilitate well-controlled test operations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

BISON Fuel Performance Analysis of IFA-796 Rod 3 & 4 and Investigation of the Impact of Fuel Creep

In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace the currently used zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromiumaluminum (FeCrAl) alloys because they exhibit slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and slow cladding consumption in the presence of high temperature steam. These alloys should also exhibit increased “coping time” in the event of an accident scenario by improving the mechanical performance at high temperatures, allowing greater flexibility to achieve core cooling. As a continuation of the development of these alloys, in-reactor irradiation testing of FeCrAl cladded fuel rods has started. In order to provide insight on the possible behavior of these fuel rods as they undergo irradiation in the Halden Boiling Water Reactor, engineering analysis has been performed using FeCrAl material models implemented into the BISON fuel performance code. This milestone report provides an update on the ongoing development of modeling capability to predict FeCrAl cladding fuel performance and to provide an early look at the possible behavior of planned in-reactor FeCrAl cladding experiments. In particular, this report consists of two separate analyses. The first analysis consists of fuel performance simulations of IFA-796 rod 4 and two segments of rod 3. These simulations utilize previously implemented material models for the C35M FeCrAl alloy and UO2 to provide a bounding behavior analysis corresponding to variation of the initial fuel cladding gap thickness within the fuel rod. The second analysis is an assessment of the fuel and cladding stress states after modification of the fuel creep model that is currently implemented in the BISON fuel performance code. Effects from modifying the fuel creep model were identified for the BISON simulations of the IFA-796 rod 4 experiment, but show that varying the creep model (within the range investigated here) only provide a minimal increase in the fuel radius and maximum cladding hoop stress. Continued investigation of fuel behavioral models will include benchmarking the modified fuel creep model against available experimental data, as well as an investigation of the role that fuel cracking will play in the compliance of the fuel. Correctly calculating stress evolution in the fuel is key to assessing fuel behavior up to gap closure and the subsequent deformation of the cladding due to PCMI. The inclusion of frictional contact should also be investigated to determine the axial elongation of the fuel rods for comparison with data from this experiment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

External Cooling of the BWR Mark I and II Drywell Head as a Potential Accident Mitigation Measure – Scoping Assessment

This report documents a scoping assessment of a potential accident mitigation action applicable to the US fleet of boiling water reactors with Mark I and II containments. The mitigation action is to externally flood the primary containment vessel drywell head using portable pumps or other means. A scoping assessment of the potential benefits of this mitigation action was conducted focusing on the ability to (1) passively remove heat from containment, (2) prevent or delay leakage through the drywell head seal (due to high temperatures and/or pressure), and (3) scrub radionuclide releases if the drywell head seal leaks.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Design and Testing for a New Thermosyphon Irradiation Vehicle

The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) requires most materials and all fuel experiments to be placed in a pressure containment vessel to ensure that internal contaminants such as fission products cannot be released into the primary coolant. It also requires that all experiments be capable of withstanding various accident conditions (e.g., loss of coolant) without generating vapor bubbles on the surface of the experiment in the primary coolant. These requirements are intended to artificially increase experiment temperatures by introducing a barrier between the experimental materials and the HFIR coolant, and by reducing heat loads to the HFIR primary coolant, thus ensuring that no boiling can occur. A proposed design for materials irradiation would remove these limitations by providing the required primary containment with an internal cooling flow. This would allow for experiments to be irradiated without concern for coolant contamination (e.g., from cladding failure of advanced fuel pins) or for specimen heat load. This report describes a new materials irradiation experiment design that uses a thermosyphon cooling system to allow experimental materials direct access to a liquid coolant. The new design also increases the range of conditions that can be tested in HFIR. This design will provide a unique capability to validate the performance of current and advanced fuels and materials. Because of limited supporting data for this kind of irradiation vehicle, a test program was initiated to obtain operating data that can be used to (1) qualify the vehicle for operation in HFIR and (2) validate computer models used to perform design- and safety-basis calculations. This report also describes the test facility and experimental data, and it provides a comparison of the experimental data to computer simulations. A total of 51 tests have been completed: four tests with pure steam, 12 tests with argon, and 35 tests with helium. A total of 10 tests were performed at subatmospheric pressure, and four of these were performed with pure steam. One test was conducted at a high power of 92.7 kW, six tests were HFIR startups, and two tests were HFIR loss of offsite power (LOOP). Pressures up to 10 MPa, vapor temperatures up to 583 K (310°C), and heater temperatures above 600 K (327°C) have been reached in these tests. Two computer programs, RELAP5-3D and TRACE, have been used to simulate the tests. The TRACE code has shown good agreement with the test data and has been used to model a variety of tests. This experimental facility has been very useful in demonstrating the viability of this new type of irradiation facility.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗