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At least 955 records · Page 53
All sourcesNASA NTRSDOE OSTI
NASA NTRS2015

Connections Between the TTL and Sea Surface Temperatures: Interannual Variability and Trends

Comprehensive chemistry climate models and satellite data are used to investigate the forcing of variability in the tropical lower stratosphere and upper troposphere. As this region is the origination region for air parcels which enter the stratosphere, it is important to understand variability in this region on timescales ranging from the seasonal to decadal. The warming trend in the tropical upper troposphere over the past 30 years is strongest near the Indo Pacific warm pool, while the warming trend in the western and central Pacific is much weaker. In the lower stratosphere, these trends are reversed: the historical cooling trend is strongest over the Indo Pacific warm pool and is weakest in the western and central Pacific. These zonal variations are stronger than the zonal mean response in boreal winter. Targeted experiments with a chemistry climate model are used to demonstrate that sea surface temperature trends are driving the zonal asymmetry in upper tropospheric and lower stratospheric tropical temperature trends. The anomalous circulation set up by the changing SSTs has led to zonal structure in the ozone and water vapor trends near the tropopause, and subsequently to less water vapor entering the stratosphere. Projected future sea surface temperatures appear to drive a temperature and water vapor response whose zonal structure is similar to the historical response. In the lower stratosphere, the changes in water vapor and temperature due to projected future sea surface temperatures is of similar strength to, though slightly weaker than, that due directly to projected future CO2, ozone, and methane. Finally, targeted experiments with a chemistry climate model are used to demonstrate that seasonality and the location of the peak warming of sea surface temperatures dictate the response of stratospheric water vapor to El Nino. In spring, El Nino events in which sea surface temperature anomalies peak in the eastern Pacific lead to a warming at the tropopause above the warm pool region, and subsequently to more stratospheric water vapor (consistent with previous work). However, in fall and in early winter, and also during El Nino events in which the sea surface temperature anomaly is found mainly in the central Pacific, the response is qualitatively different: temperature changes in the warm pool region are nonuniform and less water vapor enters the stratosphere. The difference in water vapor in the lower stratosphere between the two variants of El Nino approaches 0.3 ppmv, while the difference between the winter and spring responses exceeds 0.5 ppmv.

Tropical Tropopause Layer↗
NASA NTRS2019

Thermal Performance of an Aluminum/Ammonia Heat Pipe Under Reflux Mode

During instrument-level or spacecraft-level ground testing, heat pipes may be placed in reflux mode, with condenser above evaporator. A liquid pool will form at the bottom of the heat pipe. If heat is applied to a site below the surface of the liquid pool in a vertical heat pipe, the heat pipe can work properly under reflux mode. A superheat is required for startup. If heat is applied to a site above the liquid pool, the heat pipe is not expected to work unless additional heat is applied to the liquid pool to provide the needed flow circulation. There are many reason to minimize the additional heater power. An experimental investigation was conducted to study the heat pipe behavior under this configuration.

Thermal control systems↗
NASA NTRSDate not supplied

Does Chirality Influence the Stability of Amino Acid –Cu Complexes in the Salt-Induced Peptide Formation Reaction? Insights from Density Functional Theory Calculations

The polymerization of amino acids into peptides and ultimately proteins is critical to evolution of life. Ribosomally-synthesized proteins are homochiral, composed entirely of L-amino acids. Abiotic syntheses of amino acids, however, produce racemic mixtures(equal amounts of L and D enantiomers), meaning that the pool of monomers available for abiotic protein synthesis would be racemic or nearly so. There are no prebiotically plausible mechanisms known to yield homochiral peptides from racemic pools of amino acids. While some mechanisms capable of producing or amplifying enantiomeric excesses have been observed, these excesses are small or not applicable to all biologically relevant amino acids. As a result, a prebiotically-feasible mechanism to create enantiopure pools of amino acids has not been identified. However, it is not clear that these enantio-enriched or enantiopure pools are necessary to produce homochiral peptides. Steric and electronic effects, or a combination of both, may influence how different enantiomers interact with each other, potentially leading to homochiral peptides from racemic mixtures. Here, we use density functional theory (DFT) calculations to assess the stability of homochiral and heterochiral reactive complexes produced during polymerization via the salt-induced peptide formation (SIPF) reaction. Experimental studies of the SIPF reaction have shown that it enables the formation of peptides under diverse environmental conditions, making it a plausible pathway to peptide formation on early Earth. In the SIPF reaction, NaCl acts as a condensation reagent while a divalent metal cation, primarily Cu2+, facilitates polymerization by complexing and activating amino acids forming a monochlorocuprate complex. To assess if hetero-or homochiral monochlorocuprate complexes were energetically favored, we compared the stability of LL, DD, and LD enantiomers of Cu2+–(alanine)2and Cu2+–(valine) 2 complexes. Gaussian 09 density DFT energy-minimization calculations were made for each Cu –amino acid complex in the cis and trans configuration. Models were energy minimized using the BH and HLYP/6-31++G(d,p) level of theory and free energies were compared to determine the most stable configuration.

A C Fox↗
NASA NTRSDate not supplied

Multilevel Probit Regression for 3-Alternative Forced Choice Audibility Testing

A recent psychoacoustic test at NASA Langley generated a dataset of 3-alternative forced-choice responses for 40 subjects that measured the audibility of a tone complex in a shaped broadband masker. The task was completed by 4 subjects at a time in a small theatre-like environment using predetermined stimuli levels. These data were subject to 4 forms of probit regression: a “complete pooling” analysis in which all data from the test was fit with one curve, two forms of “no pooling” analyses in which subjects’ data were treated individually (using both packaged and custom software), and a “partial pooling” analysis in which multilevel-regression software fit both individual curves as well as population-level parameters at the same time. The results of the analyses are compared in terms of both individual- and population-level parameters. Partial pooling appears to give the most consistent results at both levels, as well as provide the most robustness among the packaged approaches (albeit with added complexity over single-level regression). These results are also contained in a recent NASA technical memorandum entitled “Comparisons of Analysis Methods Applied to Alternative Forced Choice Audibility Data.”

Psychoacoustics↗
NASA NTRS2023

Ocean Carbon From Space: Current Status and Priorities for the Next Decade

The ocean plays a central role in modulating the Earth’s carbon cycle. Monitoring how the ocean carbon cycle is changing is fundamental to managing climate change. Satellite remote sensing is currently our best tool for viewing the ocean surface globally and systematically, at high spatial and temporal resolutions, and the past few decades have seen an exponential growth in studies utilising satellite data for ocean carbon research. Satellite-based observations must be combined with in-situ observations and models, to obtain a comprehensive view of ocean carbon pools and fluxes. To help prioritise future research in this area, a workshop was organised that assembled leading experts working on the topic, from around the world, including remote-sensing scientists, field scientists and modellers, with the goal to articulate a collective view of the current status of ocean carbon research, identify gaps in knowledge, and formulate a scientific roadmap for the next decade, with an emphasis on evaluating where satellite remote sensing may contribute. A total of 449 scientists and stakeholders participated (with balanced gender representation), from North and South America, Europe, Asia, Africa, and Oceania. Sessions targeted both inorganic and organic pools of carbon in the ocean, in both dissolved and particulate form, as well as major fluxes of carbon between reservoirs (e.g., primary production) and at interfaces (e.g., air-sea and land–ocean). Extreme events, blue carbon and carbon budgeting were also key topics discussed. Emerging priorities identified include: expanding the networks and quality of in-situ observations; improved satellite retrievals; improved uncertainty quantification; improved understanding of vertical distributions; integration with models; improved techniques to bridge spatial and temporal scales of the different data sources; and improved fundamental understanding of the ocean carbon cycle, and of the interactions among pools of carbon and light. We also report on priorities for the specific pools and fluxes studied, and highlight issues and concerns that arose during discussions, such as the need to consider the environmental impact of satellites or space activities; the role satellites can play in monitoring ocean carbon dioxide removal approaches; economic valuation of the satellite based information; to consider how satellites can contribute to monitoring cycles of other important climatically-relevant compounds and elements; to promote diversity and inclusivity in ocean carbon research; to bring together communities working on different aspects of planetary carbon; maximising use of international bodies; to follow an open science approach; to explore new and innovative ways to remotely monitor ocean carbon; and to harness quantum computing. Overall, this paper provides a comprehensive scientific roadmap for the next decade on how satellite remote sensing could help monitor the ocean carbon cycle, and its links to the other domains, such as terrestrial and atmosphere.

oceanography↗
DOE OSTI2011

Fluid Flow Phenomena during Welding

MOLTEN WELD POOLS are dynamic. Liquid in the weld pool in acted on by several strong forces, which can result in high-velocity fluid motion. Fluid flow velocities exceeding 1 m/s (3.3 ft/s) have been observed in gas tungsten arc (GTA) welds under ordinary welding conditions, and higher velocities have been measured in submerged arc welds. Fluid flow is important because it affects weld shape and is related to the formation of a variety of weld defects. Moving liquid transports heat and often dominates heat transport in the weld pool. Because heat transport by mass flow depends on the direction and speed of fluid motion, weld pool shape can differ dramatically from that predicted by conductive heat flow. Temperature gradients are also altered by fluid flow, which can affect weld microstructure. A number of defects in GTA welds have been attributed to fluid flow or changes in fluid flow, including lack of penetration, top bead roughness, humped beads, finger penetration, and undercutting. Instabilities in the liquid film around the keyhole in electron beam and laser welds are responsible for the uneven penetration (spiking) characteristic of these types of welds.

36 MATERIALS SCIENCE↗
DOE OSTI2012

Graphical User Interface Software for Gross Defect Detection at the Atucha-I Plant

At the Atucha-I pressurized heavy water reactor in Argentina, fuel assemblies in the spent fuel pools are stored by suspending them in two vertically stacked layers. This introduces the unique problem of verifying the presence of fuel in either layer without physically moving the fuel assemblies. Movement of fuel, especially from the lower layer, would involve a major effort on the part of the operator. Given that the facility uses both natural uranium and slightly enriched uranium at 0.85 w% {sup 235}U, and has been in operation since 1974, a wide range of burnups and cooling times can exist in any given pool. Additionally, while fuel assemblies are grouped together in a uniform fashion, the packing density from group to group can vary within a single pool. A tool called the Spent Fuel Neutron Counter (SFNC) was developed and successfully tested at the site to verify, in an in-situ condition, the presence of fuel up to burnups of 8,000 MWd/t. Since the neutron source term becomes a nonlinear function of burnup beyond this burnup, a new algorithm was developed to predict expected response from the SFNC at measurement locations covering the entire range of burnups, cooling times, and initial enrichments. With the aid of a static database of parameters including intrinsic sources and energy group-wise detector response functions, as well as explicit spent fuel information including burnups, cooling times, enrichment types, and spacing between fuel assemblies, an expected response for any given location can be calculated by summing the contributions from the relevant neighboring fuel assemblies. Thus, the new algorithm maps the expected responses across the various pools providing inspectors with a visual aid in verifying the presence of the spent fuel assemblies. This algorithm has been fully integrated into a standalone application built in LabVIEW. The GUI uses a step-by-step approach to allow the end-user to first calibrate the predicted database against a set of measurements with SFNC at selected locations where spent fuel is present. Once the database is calibrated it can be used to detect gross defects by comparing the measured signal to the one predicted by the database with differences beyond a set tolerance indicating missing fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Gravity-based, non-invasive reactor system and method for coolant inventory monitoring

A gravity-based, non-invasive method of measuring a level of fluid in a container comprises use of at least one gravity meter located as proximate a center of mass of the fluid as possible. In a nuclear reactor system a method for monitoring the level of fluid in a nuclear reactor module, a report of a loss or gain of fluid within a cylindrical module may be generated from capturing a time series of gravity data from a first gravity meter mounted as an upper gravity meter and a second gravity meter mounted as a lower gravity meter, for example, proximate a cylindrical nuclear reactor module so as not to require any invasive conduit through, for example, a containment pressure vessel (CPV) or a reactor pressure vessel (RPV). In one embodiment, the upper and lower gravity meters are mounted on stable mounts as close to the fluid in the module as possible within a coolant pool or a structure containing cooled air. If a coolant pool of water surrounds a nuclear reactor module, the meters may be housed within a dry housing in the coolant pool such that the meters may be accessed from above the coolant pool and are located as close as possible to the reactor module and its contained mass of fluid.

Bloom, Scott Harris↗
DOE OSTI2019

Researching the Effects of Long-Term Dry Storage on Commercial Spent Nuclear Fuel

When the fuel in a nuclear reactor no longer contributes efficiently to power production, it is discharged and stored onsite in a water-filled spent fuel pool. When nuclear power plants in the United States were originally designed, it was expected that onsite storage would be temporary. The used nuclear fuel, also known as spent nuclear fuel (SNF), was to be removed from the storage pools after only a few years of cooling and sent for reprocessing. However, commercial reprocessing was not implemented in the United States, and utility spent fuel pools began running out of room. In order to continue power production, nuclear utilities began expanding onsite fuel storage capacities, first by increasing the density of fuel in the fuel pool racks, and then by developing and implementing dry storage casks. The dry storage casks are stored onsite by each utility at a licensed and secured independent spent fuel storage installation (ISFSI). Dry storage casks are robust systems typically consisting of a sealed metal cylinder surrounded by a combination of steel and concrete to provide radiation shielding. The first dry storage casks were loaded with SNF in 1986. Currently there are over 3,000 loaded dry storage casks, and approximately 200 new casks are being placed into service annually at operating nuclear power plants nationwide. Multiple nuclear power plant sites in the United States have shut down and completed decommissioning activities except for the removal of fuel in dry storage. Eventually, the 75,000+ tons of discharged SNF at more than 70 secured locations will need to be safely and securely transported to a new location(s) for eventual disposition. The continued safety and security of this material must be assured.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Next Generation System Analysis Model: Recently Added Features and Future Plans

To better enable informed decision-making regarding the back-end of the nuclear fuel cycle, the Integrated Waste Management System (IWMS) program within the U.S. Department of Energy, Office of Nuclear Energy (DOE-NE) has been sponsoring the development and application of system analysis tools capable of analyzing various system options for the management of spent nuclear fuel (SNF) and high-level radioactive waste (HLW).With these tools, IWMS architecture analyses are being conducted to support the future deployment of a comprehensive nuclear waste management system that considers all major back-end aspects of the nuclear fuel cycle (i.e., transportation, storage, and disposal). The Next Generation System Analysis Model (NGSAM) is an agent-based simulation software tool expressly designed to be capable of modeling features within various IWMS architectures. NGSAM imports data from Oak Ridge National Laboratory (ORNL)’s unified database (e.g., historic assembly information, thermal profiles for assembly heat, and at-reactor dry storage loadings) to ensure that each simulation initializes with a realistic representation of the state of commercial SNF in the U.S. Recent major enhancements implemented into NGSAM in the period since NGSAM was last presented at the WM2019 conference include: • Tracking of railroad escort car acquisition and buffer car acquisition • The addition of heavy haul truck (HHT) and barge routes for some sites, as well as support for user-defined inter-modal routes • Updates to the logic that checks the transportation cask thermal limit maps prior to package transport • An allocation method that predicts when reactor sites would pack assemblies from their spent fuel pools for dry storage, and prioritizes shipments directly from the pools of those reactor sites in the preceding periods (before the predicted loadings to dry storage), thus reducing the number of casks loaded into dry storage at reactor sites • The addition of “reactor site family” operational limits to restrict the number of loads of SNF taken from the pool and from dry storage at a given reactor site each year • Added support for multiple canister loading map options and packages with multiple compatible transportation overpacks • Updates to the handling of non-commercial SNF, including a new database that contains data to support the updates • Updates to allow analysis of hypothetical scenarios which include repackaging at reactor sites, e.g., for possible comparative analysis with other scenarios • Implementation of additional output reports, or modification of existing ones • The ability to generate and implement user edits via the NGSAM website • The ability to model loading SNF from pool storage at an interim storage facility (ISF) into dry storage at the ISF • The ability to model consolidating SNF from different existing storage containers at a DOE site into the same DOE standard canister • The ability to model transferring SNF casks from one transportation mode to another, e.g., from HHT to rail, referred to as transloading. These new features have improved NGSAM capabilities and users’ experience with the model. Preliminary NGSAM requirements for modeling advanced reactor fuels, reprocessing, treatment, and conditioning were considered, and this paper describes them at a high level. Other nuclear fuel cycle system analysis tools developed under sponsorship of DOE-NE, like the VISION code developed at Idaho National Laboratory, might be better suited for initial high-level analysis of those technologies and advanced fuel cycles. As technologies are developed and system concepts evolve, NGSAM could provide value by providing more detailed modeling of transport, storage, and disposal of spent fuel and wastes from advanced reactors and advanced fuel cycles at the fuel element and waste container level.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Next Generation System Analysis Model Recently Added Features and Future Plans - Abstract

The Nuclear Waste Policy Act of 1982, as amended (NWPA 1982), established the federal government’s responsibility to accept spent nuclear fuel (SNF) and high-level radioactive waste (HLW) from waste owners and generators for ultimate disposition. SNF generated by the current fleet of commercial nuclear reactors is being stored at the reactor sites in spent fuel pools (SFPs) and in dry independent spent fuel storage installations (ISFSIs). The US Department of Energy Office of Nuclear Energy (DOE-NE) is developing an Integrated Waste Management Program (IWMP) comprising a suite of options and supporting analyses to enable future informed choices. The IWMP is applying integrated waste management system architecture analysis, system engineering, and decision analysis principles to inform potential future decisions regarding potential nuclear waste management system architectures. Architecture analyses of the IWM system are being conducted to support the future deployment of a comprehensive system for managing nuclear waste that considers all major aspects of the back end of the nuclear fuel cycle (i.e., transportation, storage, and disposal). The Next Generation System Analysis Model (NGSAM) is an agent-based simulation software tool designed for the express purpose of modeling the IWM system. NGSAM imports data from the Oak Ridge National Laboratory (ORNL) Unified Database (e.g., historic assembly information, thermal profiles for assembly heat, at-reactor dry storage loadings) to ensure that the simulation initializes with a realistic representation of the state of commercial SNF in the United States. Recent major enhancements that have been implemented into NGSAM since NGSAM was last presented at the WM2019 conference include: • Tracking of railroad escort and buffer car acquisition. • Addition of heavy haul and barge routes for some sites, as well as support for user-defined inter-modal routes. • Updates to the logic that checks the thermal maps prior to package transport. • Addition of an allocation method that predicts when reactor sites will pack assemblies from their pools for dry storage and allocates packages to those reactor sites in the preceding periods, favoring direct transport packages and reducing the number of packages that reactor sites pack for dry storage at their ISFSIs. • Addition of reactor site family operational limits, which are used to limit the number of loads from the pool and from dry storage at a given reactor site per year. • Support has been added for multiple canister loading maps and packages having multiple compatible transportation overpacks. • Updates in the handling of non-commercial fuel, including a new database containing data to support the updates. • Support for repackaging at reactor sites. • Implementing additional output reports or modifying existing reports. • User edits can now be created and edited via the NGSAM website. • Ability to load packages for dry storage at ISF pools. • Same-type package blending at DOE sites. • Support for multi-mode transloading at reactor sites. These new features have improved NGSAM capabilities and/or improve the user experience with the model and will be discussed in more detail. The initial NGSAM requirements for advanced reactor fuels, reprocessing, treatment, and conditioning are preliminary and are described at a high level in this paper: analysts will provide more specific requirements to the NGSAM team in the future. Additionally, there are many data needs associated with modeling advanced reactors in NGSAM, but many of the data or plans are still in progress and/or yet to be fully defined. However, this document describes an initial exploration of the data relevant to this program. Advanced reactor data will likely require revision as concepts evolve and new considerations are made. This is a technical paper that does not take into account contractual limitations or obligations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). For example, under the provisions of the Standard Contract, spent nuclear fuel in multi-assembly canisters is not an acceptable waste form, absent a mutually agreed to contract amendment. To the extent discussions or recommendations in this paper conflict with the provisions of the Standard Contract, the Standard Contract governs the obligations of the parties, and this paper in no manner supersedes, overrides, or amends the Standard Contract. This paper reflects technical work which could support future decision making by DOE. No inferences should be drawn from this paper regarding future actions by DOE, which are limited both by the terms of the Standard Contract and Congressional appropriations for the Department to fulfill its obligations under the Nuclear Waste Policy Act including licensing and construction of a spent nuclear fuel repository.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Two-color high speed thermal imaging system for laser-based additive manufacturing process monitoring

Monitoring melt pool temperature in laser powder bed fusion by providing a build laser that produces a laser beam that is directed onto the melt pool and produces an incandescence that emanates from the melt pool, receiving the incandescence and producing a first image having a first spectral band and a second image having a second spectral band, and determining the ratio of said first image having a first spectral band and said second image having a second spectral band to monitor the melt pool temperature.

Calta, Nicholas P.↗
DOE OSTI2012

Validation of CONTAIN-LMR code for accident analysis of sodium-cooled fast reactor containments

CONTAIN-LMR 1 is an analytical tool for the containment performance of sodium cooled fast reactors. In this code, the modelling for the sodium fire is included: the oxygen diffusion model for the sodium pool fire, and the liquid droplet model for the sodium spray fire. CONTAIN-LMR is also able to model the interaction of liquid sodium with concrete structure. It may be applicable to different concrete compositions. Testing and validation of these models will help to qualify the simulation results. Three experiments with sodium performed in the FAUNA facility at FZK have been used for the validation of CONTAIN-LMR. For pool fire tests, calculations have been performed with two models. The first model consists of one gas cell representing the volume of the burn compartment. The volume of the second model is subdivided into 32 coupled gas cells. The agreement between calculations and experimental data is acceptable. The detailed pool fire model shows less deviation from experiments. In the spray fire, the direct heating from the sodium burning in the media is dominant. Therefore, single cell modeling is enough to describe the phenomena. Calculation results have reasonable agreement with experimental data. Limitations of the implemented spray model can cause the overestimation of predicted pressure and temperature in the cell atmosphere. The ability of the CONTAIN-LMR to simulate the sodium pool fire accompanied by sodium-concrete reactions was tested using the experimental study of sodium-concrete interactions for construction concrete as well as for shielding concrete. The model provides a reasonably good representation of chemical processes during sodium-concrete interaction. The comparison of time-temperature profiles of sodium and concrete shows, that the model requires modifications for predictions of the test results. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Preliminary Design of Critical Function Monitoring System of PGSFR

A PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor) is under development at Korea Atomic Energy Research Institute. A critical function monitoring system of the PGSFR is preliminarily studied. The functions of CFMS are to display critical plant variables related to the safety of the plant during normal and accident conditions and guide the operators corrective actions to keep the plant in a safe condition and mitigate the consequences of accidents. The minimal critical functions of the PGSFR are composed of reactivity control, reactor core cooling, reactor coolant system integrity, primary heat transfer system(PHTS) heat removal, sodium water reaction mitigation, radiation control and containment conditions. The variables and alarm legs of each critical function of the PGSFR are as follows; - Reactivity control: The variables of reactivity control function are power range neutron flux instrumentation, intermediate range neutron flux instrumentation, source range neutron flux instrumentation, and control rod bottom contacts. The alarm leg to display the reactivity controls consists of status of control drop malfunction, high post trip power and thermal reactivity addition. - Reactor core cooling: The variables are PHTS sodium level, hot pool temperature of PHTS, subassembly exit temperature, cold pool temperature of the PHTS, PHTS pump current, and PHTS pump breaker status. The alarm leg consists of high core delta temperature, low sodium level of the PHTS, high subassembly exit temperature, and low PHTS pump load. - Reactor coolant system integrity: The variables are PHTS sodium level, cover gas pressure, and safeguard vessel sodium level. The alarm leg is composed of low sodium level of PHTS, high cover gas pressure and high sodium level of the safety guard vessel. - PHTS heat removal: The variables are PHTS sodium level, hot pool temperature of PHTS, core exit temperature, cold pool temperature of the PHTS, flow rate of passive residual heat removal system, flow rate of active residual heat removal system, and temperatures of air heat exchanger temperature of residual heat removal systems. The alarm legs are composed of two legs of a 'passive residual heat removal system not cooling' and 'active residual heat removal system not cooling'. - Sodium water reaction mitigation: The variables are intermediate heat transfer system(IHTS) pressure, pressure and temperature and level of sodium dump tank, the status of rupture disk, hydrogen concentration in IHTS and direct variable of sodium-water-reaction measure. The alarm leg consists of high IHTS pressure, the status of sodium water reaction mitigation system and the indication of direct measure. - Radiation control: The variables are radiation of PHTS, radiation of IHTS, and radiation of containment purge. The alarm leg is composed of high radiation of PHTS and IHTS, and containment purge system. - Containment condition: The variables are containment pressure, containment isolation status, and sodium fire. The alarm leg consists of high containment pressure, status of containment isolation and status of sodium fire. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

A study of transfer of UNF from non-disposable canisters - 15388

A study was performed by AREVA and URS for the Department of Energy (DOE) [1] to establish optimal methods for the opening of dry storage canisters (DSC) and the subsequent repackaging of used nuclear fuel (UNF) from the DSCs into generic standard transportation, aging, and disposal canisters (STADs) in wet or dry environments at reactor, consolidated storage, or repository sites. Eight dry and wet cutting technologies were evaluated: (1) skiving; (2) milling; (3) arc gouging/plasma torch; (4) cutting wheel; (5) diamond cable cutting; (6) laser cutting; (7) sawing; and (8) electrical discharge machining. The evaluated UNF repackaging technologies options included: (1) wet transfer in a pool; (2) dry transfer in a dry transfer system (DTS); (3) dry transfer in a fixed hot cell (FHC); (4) dry transfer in a mobile hot cell (MHC), and (5) a hybrid approach utilizing wet transfer in a pool and dry transfer in a FHC. A systems engineering evaluation was utilized for ranking the eight viable cutting options and the five different repackaging approaches for each of the sites. The systems engineering evaluation took into account six criteria (i.e., life cycle costs, worker dose, operational efficiency, technical maturity, ease of licensing, and waste generation) and performed a qualitative pairwise ranking of their relative importance to one another. Then quantitative or qualitative factors associated with each criterion (e.g., cost in millions of dollars, relative ease of licensing, etc.) for each option at each location (reactor, consolidated storage facility [CSF], repository) were developed based on historical experience, existing design studies, current 'state-of-the art', and/or engineering judgment. Using this input, the systems engineering evaluation identified dry skiving as the preferred cutting technology, with the wheel cutting technology in either a dry or wet environment identified as a promising technology that, with some RD and D, could mature to a better option (depending also on the disposition of the used DSC). For a reactor and a CSF/repository site, the systems engineering evaluation ranked the repackaging option using the skiving dry cutting technique and wet transfer of UNF in a pool as the highest option. However for a CSF and a repository site, the team recommended the hybrid option using the skiving dry cutting technique and both dry and wet transfer of UNF in a pool and FHC as a potentially better option (providing specific benefits such as higher throughput and more flexibility in operation). (authors)

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

Large Cask, Small Building, AREVA's Safe MLLW Disposal Solution Despite Facility Limitations - 16295

In order to begin decommissioning of a 1950's research reactor, Class C Mixed Low Level Waste (MLLW) had to be removed from the facility and disposed of. The AREVA TN RAM cask, designed and licensed for shipping Type B waste packages, was chosen as the most appropriate shipping solution. Due to the building layout, crane capacity and pool size, the standard TN RAM cask loading process could not be utilized. Unique solutions and approaches had to be developed to complete the project. AREVA designed a smaller liner insert that would house the four MLLW components in a single waste package to solve the problems related to crane capacity and physical pool dimensions. AREVA also designed and fabricated a shield bell to provide radiation shielding while moving the liner insert from the reactor pool to the cask that was staged adjacent to the reactor pool. The AREVA team worked with the disposal facility to cover requirements for disposal of Class B MLLW. This included designing spacers filled with inert material to fulfill the requirement for maximum 10% void space in the waste package. This disposal method for Class C waste met the requirements of macro-encapsulation. Therefore, no additional treatment action was required when disposing of Class C MLLW. AREVA's solution provides a model for disposing MLLW requiring macro-encapsulation in a single waste package. This solution also provides a model for how to adapt large shipping casks and liners to smaller aging commercial, DOE or research facilities. (authors)

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

Sensitivity Analysis of an Experimental Breeder Reactor II Fuel Assembly

The Experimental Breeder Reactor II (EBR-II) was selected as a candidate for a reactor physics benchmark evaluation due to its unique features and capabilities. In connection with the benchmark evaluation, the goal of this work was to perform a sensitivity analysis. The EBR-II was a sodium cooled, pool type, metal fueled, fast spectrum reactor. EBR-II operated from 1964 through 1994. It had a maximum heat output of 62.5 MW which could produce 20 MW of electricity. The design of EBR-II consisted of 637 hexagonal shaped, removable assemblies. The assemblies were divided into three regions: the core, an inner blanket, and an outer blanket. The core region contained assemblies with 91 fuel rods in each. The fuel was enriched uranium metal and was clad within stainless steel. The core region also included two safety assemblies and eight control assemblies with the difference being that these assemblies only contained 61 fuel rods. Additional assemblies in the core region included stainless steel dummies, half worth drivers, and experimental/instrumentation assemblies. The inner blanket region initially consisted of depleted uranium for fuel breeding, but was replaced, after proving the concept of breeding, with stainless steel reflectors for the purpose of irradiation. The outer blanket region consisted of depleted uranium. Figure 1 shows a cross section view of the EBR-II core layout. Each assembly was approximately 234 cm long with the uranium fuel section being 33 cm long. Above and below the fuel were stainless steel neutron reflectors. Additionally, orifices at the bottom of the assembly allowed for sodium coolant to flow upward, with larger holes in the subassemblies at the center where the greatest heat was produced. The core of EBR-II was enclosed by a stainless steel reactor vessel which was comprised of a radial shell, a lower grid plenum structure, and an upper reactor vessel cover that served as a neutron shield. Surrounding the reactor vessel radially were layers of graphite and borated graphite shielding. The entire assembly is submerged 3 meters in a pool of sodium. The pool contained up to 337,000 liters of sodium which was forced through the core and blanket regions using two centrifugal pumps that were also submerged within the sodium pool. The design of EBR-II included many safety features. The compatibility between the sodium and stainless steel prevented corrosion and therefore the risk of a radioactive release due to corrosion. The sodium coolant was kept at near atmospheric pressure which made any leak more controllable. The metal fuel, cladding, and coolant allowed for heat to be removed easily. The large volume of sodium allowed for heat to be absorbed without problems occurring. Lastly, the reactor would expand as temperature increased which would result in greater neutron leakage. This extra leakage caused a large negative temperature reactivity feedback which automatically shut the reactor down in case of an accident. EBR-II had unique capabilities which were proven during the Integral Fast Reactor initiative. This program lasted from 1984 to 1994. The goal of this program was to solve many of the perceived problems with nuclear energy through scientific means. The safety features of EBR-II were put to the test in April 1986. The tests simulated loss of reactor coolant flow and loss of a heat sink. For both of these tests, the reactor was at 100% power and was not scrammed. During both tests, the reactor shut itself down and achieved a safe steady state temperature all due to the design of the reactor. It was because of these tests that EBR-II was chosen for a reactor physics benchmark evaluation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Experimental Study of Molten Metallic Fuel Relocation in Sodium-Filled Fuel Structures

The Prototype Generation-IV Sodium-cooled Fast Reactor (PGSFR) is being developed by Korea Atomic Energy Research Institute (KAERI). While their design incorporates inherent safety features demonstrated in the landmark Shutdown Heat Removal Test (SHRT) conducted at EBR-II including a large pool configuration and uranium alloy fuel, it is still necessary to evaluate metal fuel relocation and theirs consequences in case of postulated core disruptive accidents. Extensive data regarding fuel relocation in sodium from both in-pile and out-of-pile tests were obtained for oxide fuels in the 1960's - 1970's, while few studies of metallic fuel behavior were performed because of concerns that metallic fuel could not achieve high burnup due to the irradiation induced swelling problem. Following new metal fuel design development that mitigated the swelling problem, experiments regarding interaction of molten metal fuels and liquid sodium have been conducted by several researchers. Most studies that were conducted using metal simulants in various liquids concluded that the molten metals were heavily fragmented after liquid contact and the dominant factors governing the fragmentation were superheating and latent heat of fusion of the molten metal fuel. Gabor et al. conducted experiments using kilogram quantities of various molten uranium alloys in an open sodium pool configuration. They found that the metal fuel fragments were basically in forms of filaments and sheets with a high bed voidage in the order of 0.9. Their calculation indicated that the debris beds formed by relocated metal fuels in an open sodium pool would be largely coolable by conduction heat transfer; and even if deep beds were to form, convective heat transfer and boiling heat transfer could preclude further melt penetration. Although Gabor's tests are useful in evaluating the coolability of relocated metal fuels during a postulated accident in a SFR, they focused on late phase behavior in which the core material relocates downward through the core support plate into the inlet plenum where the structures are more open and thus consistent with the open pool configuration. There is still a knowledge gap related to nature of metal fuel fragmentation and relocation within the fuel assembly structures. Argonne National Laboratory (ANL) is planning to conduct experiments regarding metal fuel relocation in the core structures which mock up the PGSFR's fuel assembly design. In the present study, preliminary tests for demonstration of melting and injecting the metal fuel were conducted using copper. Metal fuel relocation experiments with single fuel pin test section are planned to be conducted for understanding the flow and freezing behavior of the molten metal fuel in a narrow sodium channel and for evaluating the coolability of the relocated fuel. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗