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DOE OSTI2012

Commercial Building Energy Asset Rating Program -- Market Research

Under contract to Pacific Northwest National Laboratory, HaydenTanner, LLC conducted an in-depth analysis of the potential market value of a commercial building energy asset rating program for the U.S. Department of Energy (DOE) Office of Energy Efficiency and Renewable Energy. The market research objectives were to: (1) Evaluate market interest and need for a program and tool to offer asset rating and rapidly identify potential energy efficiency measures for the commercial building sector. (2) Identify key input variables and asset rating outputs that would facilitate increased investment in energy efficiency. (3) Assess best practices and lessons learned from existing national and international energy rating programs. (4) Identify core messaging to motivate owners, investors, financiers, and others in the real estate sector to adopt a voluntary asset rating program and, as a consequence, deploy high-performance strategies and technologies across new and existing buildings. (5) Identify leverage factors and incentives that facilitate increased investment in these buildings. To meet these objectives, work consisted of a review of the relevant literature, examination of existing and emergent asset and operational rating systems, interviews with industry stakeholders, and an evaluation of the value implication of an asset label on asset valuation. This report documents the analysis methodology and findings, conclusion, and recommendations. Its intent is to support and inform the DOE Office of Energy Efficiency and Renewable Energy on the market need and potential value impacts of an asset labeling and diagnostic tool to encourage high-performance new buildings and building efficiency retrofit projects.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2018

Technical Equivalency Evaluation of Electrically Cooled Detectors for the In Vivo Measurement of Low Energy Photon Emitters in Human Lungs

LLNL maintains equipment, capability, and operations to measure low energy photon emitting radionuclides potentially deposited inside of workers. Ongoing quality improvement includes the maintenance of detection systems capable of measuring radionuclides o f interest to the LLNL Health and Safety program. This document describes the technical equivalency evaluation of upgraded detectors for LLNL’s routine lung measurement program.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

Advanced Sensors and Instrumentation (ASI) Program Plan. Nuclear Energy Enabling Technologies (NEET)

The U.S. Department of Energy Office of Nuclear Energy (DOE-NE) sponsors a crosscutting technology development (CTD) program addressing common instrumentation & controls (I&C) needs in all DOE-NE sponsored programs. In 2012, DOE-NE initiated the Nuclear Energy Enabling Technology (NEET) Program to conduct research, development, and demonstration (RD&D) in crosscutting technologies that directly support and enable the development of existing, new and advanced reactor designs, and fuel cycle technologies. These technologies will advance the state of nuclear technology, improve its competitiveness, and promote continued contribution to meeting our Nation’s energy and environmental challenges. The NEET Program has spurred innovation in the field of sensors and instrumentation, advanced methods for manufacturing, and cyber security. The Advanced Sensors and Instrumentation (ASI) program is one element of the NEET CTD that has spurred innovation in the measurement science field by funding research on sensors, instrumentation and controls, communication, and advanced analytics that measure system parameters, gauge component control input, and provide power plant owners and operators critical awareness so they can plan and safely manage operational evolutions. The NEET ASI Program has carried out research and development (R&D) required to develop and deploy innovative and advanced sensors, and instrumentation and control capabilities for current and future nuclear energy systems, and to enable the advanced technologies essential to NE’s R&D efforts needed to realize mission goals. The concept of technology readiness assessment is helpful for the classification of instrumentation technology as part of the program strategy. A technology readiness assessment evaluates technology maturity using the Technology Readiness Level (TRL) scale and was pioneered by the National Aeronautics and Space Administration in the 1980s for space technology. It has recently been adopted to advance fuels and materials. The NEET ASI program objective is to mature instruments technology to TRL 6, which is “demonstration of the technology in relevant environment.” While there may be instances in which a different strategy is justified, the aim of the program is to respond to DOE program and stakeholder needs for nuclear instrumentation then hand off the technology after demonstration for integration in a more complex system, for example an irradiation test, experimental loop or a plant component such as the core, primary system or heat exchanger. A mechanism to continue some level of support from NEET ASI during and after the adoption (for example, to maintain fabrication and calibration capabilities) will be defined as the first example of technology hand off comes to fruition. In addition, several paths exist within DOE to complement NEET ASI activities to extend the maturity beyond TRL, for example through GAIN, technology commercialization funding, and small-business programs. Role of NEET ASI CINR projects in the TRL space, and integration with direct funded demonstration.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Evaluating Optical Techniques to Characterize Solid State Samples for the Pu-238 Supply Program

This report documents steps that have been taken to establish advanced characterization capabilities for solid-state samples to support the 238 Pu Supply Program. The program has already invested significantly in spectroscopy and online monitoring development for liquid processes. Many of the data, analytical methods, and equipment could also be leveraged for the analysis of Np/Pu oxides and Np pellets to help classify material properties and link them to operations. The techniques described in this work include diffuse reflectance spectroscopy (DRS), Raman spectroscopy, and laser-induced breakdown spectroscopy (LIBS). Each technique is sensitive to various chemical and/or elemental species and can provide complimentary information to more standard methods such as powder x-ray diffraction and scanning electron microscopy. These options are remotely deployable via fiber-optic cables. DRS and Raman spectroscopy are already approved for glove box and hot cell work, but LIBS requires US Department of Energy approval for incorporation. Raman spectroscopy does require laser shielding, which means that DRS is the easiest method to implement. However, each technique provides complimentary data, and the most robust sample characterization could be achieved when all three, among other methods, are harnessed together. This report documents some of the progress made testing these capabilities, some of the challenges, and next steps to optimize the technology and integration strategies.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2021

Evaluation of Triage Methods for Criticality Accidents

Studies indicate that early identification of persons involved in and receiving high doses of radiation in accidents is key to providing life-saving medical treatment. Although the risk of criticality accidents is low the potential impact to workers is significant. For facilities that employ large numbers of workers a key element in the response to a radiological emergency is identifying personnel that received significant, and potentially harmful, doses. Also important is having the ability to screen large numbers of workers to identify persons that did not receive significant exposure so as to reduce the impact on emergency response efforts. At the Y-12 National Security Complex the focus on criticality accident response is the rapid triage of personnel in order to identify persons exposed to large radiation doses and to prioritize those persons receiving the highest exposures. Once identified personnel are transported to local medical facilities including the Radiation Emergency Assistance Center/Training Site (REAC/TS) for medical evaluation and treatment. The Y-12 external dosimetry program utilizes a number of techniques to identify and prioritize workers and these methods were evaluated at a criticality dosimetry intercomparison exercise. Finally, the methods used were shown to perform as intended and other sites may consider incorporating these methods into their accident dosimetry response procedures.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2012

BFS, a Legacy to the International Reactor Physics, Criticality Safety, and Nuclear Data Communities

Two Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) programs, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) have been identifying existing integral experiment data, evaluating those data, and providing integral benchmark specifications for methods and data validation for nearly two decades. The Russian Federation, IPPE, and the BFS laboratories have made substantial contributions to those two projects. Contributions from the BFS facilities are highlighted in this paper.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Conversion from film to image plates for transfer method neutron radiography of nuclear fuel

This paper summarizes efforts to characterize and qualify a computed radiography (CR) system for neutron radiography of irradiated nuclear fuel at Idaho National Laboratory (INL). INL has multiple programs that are actively developing, testing, and evaluating new nuclear fuels. Irradiated fuel experiments are subjected to a number of sequential post-irradiation examination techniques that provide insight into the overall behavior and performance of the fuel. One of the first and most important of these exams is neutron radiography, which provides more comprehensive information about the internal condition of irradiated nuclear fuel than any other non-destructive technique to date. Results from neutron radiography are often the driver for subsequent examinations of the PIE program. Features of interest that can be evaluated using neutron radiography include irradiation-induced swelling, isotopic and fuel-fragment redistribution, plate deformations, and fuel fracturing. The NRAD currently uses the foil-film transfer technique with film for imaging fuel. INL is pursuing multiple efforts to advance its neutron imaging capabilities for evaluating irradiated fuel and other applications, including conversion from film to CR image plates. Neutron CR is the current state-of-the-art for neutron imaging of highly-radioactive objects. Initial neutron radiographs of various types of nuclear fuel indicate that radiographs can be obtained of comparable image quality currently obtained using film. This paper provides neutron radiographs of representative irradiated fuel pins along with neutron radiographs of standards that informed the qualification of the neutron CR system for routine use. Additionally, this paper includes evaluations of some of the CR scanner parameters and their effects on image quality.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Assessing the Optimality of LinDist3Flow for Optimal Tap Selection of Step Voltage Regulators in Unbalanced Distribution Networks

The adoption of distributed energy resources such as photovoltaics (PVs) has increased dramatically during the previous decade. The increased penetration of PVs into distribution networks (DNs) can cause voltage fluctuations that have to be mitigated. One of the key utility assets employed to this end are step-voltage regulators (SVRs). It is desirable to include tap selection of SVRs in optimal power flow (OPF) routines, a task that turns out to be challenging because the resultant OPF problem is nonconvex with added complexities stemming from accurate SVR modeling. While several convex relaxations based on semi-definite programming (SDP) have been presented in the literature for optimal tap selection, SDP based schemes do not scale well and are challenging to implement in large-scale planning or operational frameworks. This paper deals with the optimal tap selection (OPTS) problem for wye-connected SVRs using linear approximations of power flow equations. Specifically, the LinDist3Flow model is adopted and the effective SVR ratio is assumed to be continuous- enabling the formulation of a problem called LinDist3Flow-OPTS, which amounts to a linear program. The scalability and optimality gap of LinDist3Flow-OPTS are evaluated with respect to existing SDP-based and nonlinear programming techniques for optimal tap selection in three standard feeders, namely, the IEEE 13-bus, 123-bus, and 8500-node DNs. For all DNs considered, LinDist3Flow-OPTS achieves an optimality gap of approximately 1% or less while significantly lowering the computational burden.

linear approximations↗
DOE OSTI2022

Assessing the Optimality of LinDist3Flow for Optimal Tap Selection of Step Voltage Regulators in Unbalanced Distribution Networks: Preprint

The adoption of distributed energy resources such as photovoltaics (PVs) has increased dramatically during the previous decade. The increased penetration of PVs into distribution networks (DNs) can cause voltage fluctuations that have to be mitigated. One of the key utility assets employed to this end are step-voltage regulators (SVRs). It is desirable to include tap selection of SVRs in optimal power flow (OPF) routines, a task that turns out to be challenging because the resultant OPF problem is nonconvex with added complexities stemming from accurate SVR modeling. While several convex relaxations based on semi-definite programming (SDP) have been presented in the literature for optimal tap selection, SDP based schemes do not scale well and are challenging to implement in large-scale planning or operational frameworks. This paper deals with the optimal tap selection (OPTS) problem for wye-connected SVRs using linear approximations of power flow equations. Specifically, the LinDist3Flow model is adopted and the effective SVR ratio is assumed to be continuous–enabling the formulation of a problem called LinDist3Flow-OPTS, which amounts to a linear program. The scalability and optimality gap of LinDist3Flow-OPTS are evaluated with respect to existing SDP-based and nonlinear programming techniques for optimal tap selection in three standard feeders, namely, the IEEE 13-bus, 123-bus, and 8500-node DNs. For all DNs considered, LinDist3Flow-OPTS achieves an optimality gap of approximately 1% or less while significantly lowering the computational burden.

linear approximations↗
DOE OSTI2011

Very High-Temperature Reactor (VHTR) Proliferation Resistance and Physical Protection (PR&PP)

This report documents the detailed background information that has been compiled to support the preparation of a much shorter white paper on the design features and fuel cycles of Very High-Temperature Reactors (VHTRs), including the proposed Next-Generation Nuclear Plant (NGNP), to identify the important proliferation resistance and physical protection (PR&PP) aspects of the proposed concepts. The shorter white paper derived from the information in this report was prepared for the Department of Energy Office of Nuclear Science and Technology for the Generation IV International Forum (GIF) VHTR Systems Steering Committee (SSC) as input to the GIF Proliferation Resistance and Physical Protection Working Group (PR&PPWG) (http://www.gen-4.org/Technology/horizontal/proliferation.htm). The short white paper was edited by the GIF VHTR SCC to address their concerns and thus may differ from the information presented in this supporting report. The GIF PR&PPWG will use the derived white paper based on this report along with other white papers on the six alternative Generation IV design concepts (http://www.gen-4.org/Technology/systems/index.htm) to employ an evaluation methodology that can be applied and will evolve from the earliest stages of design. This methodology will guide system designers, program policy makers, and external stakeholders in evaluating the response of each system, to determine each system's resistance to proliferation threats and robustness against sabotage and terrorism threats, and thereby guide future international cooperation on ensuring safeguards in the deployment of the Generation IV systems. The format and content of this report is that specified in a template prepared by the GIF PR&PPWG. Other than the level of detail, the key exception to the specified template format is the addition of Appendix C to document the history and status of coated-particle fuel reprocessing technologies, which fuel reprocessing technologies have yet to be deployed commercially and have only been demonstrated in testing at a laboratory scale.

Moses, David Lewis↗
DOE OSTI2019

Probabilistic Risk Assessment of Flexible Power Operation of a Nuclear Power Plant

Nuclear power plants (NPPs) are becoming cost prohibitive to operate. This is due to historically low natural gas prices and wind and solar prices often being less than the marginal cost of operating baseload coal and NPPs, which sometimes forces NPPs to pay for the electricity they create. Thus, the feasibility of using part of the heat from an NPP to create other commodities such as hydrogen is being evaluated; operating in this way is called flexible power operation (FPO). To change the NPP design and operation, a license review must happen. Part of the review is the Probabilistic Risk Assessment (PRA) which proves the safety of the design and operation by identifying sequences of events that result in plant failure and estimating their likelihood of occurrence. A Level 1 PRA model that evaluates the risk of removing heat from the secondary side of NPP was created. The model was developed using Systems Analysis Program for Hands-On Integrated Reliability Evaluations (SAPHIRE 8) software to calculate core damage frequency (CDF). It is a fault tree that represents the possible pathways that could lead to a loss of steam inventory in the secondary loop. A loss of steam inventory can induce steam generator tube rupture or other events that activates response systems to avoid core damage. This is the main concern for the Nuclear Regulatory Commission (NRC), the group in the United States that licenses NPPs. This model will calculate a total CDF of any of these pathways occurring which will be compared to the baseline CDF in a generic NPP PRA for licensing support.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Technical Evaluation of Accelerated Basin De-Inventory Material Addition to Sludge Batch 11

Savannah River Nuclear Solutions has a need to discard spent nuclear fuel (SNF), currently stored in L Basin, to the Defense Waste Processing Facility (DWPF) for vitrification. The Department of Energy (DOE) has approved the Accelerated Basin De-inventory (ABD) Program for discarding SNF via transfers from H-Canyon to the Savannah River Site (SRS) Liquid Waste (LW) system. The first ABD discards will occur during the preparation of Sludge Batch (SB) 11. An initial impact evaluation of the LW flowsheet was performed by the Savannah River National Laboratory (SRNL) prior to the approval of the ABD Program. This evaluation addressed the LW downstream facilities based on the current H-Canyon flowsheet sequence for the average ABD discard. The flowsheet evaluation only included aluminum-clad SNF, specifically Materials Test Reactor (MTR) fuel and High Flux Isotope Reactor (HFIR) fuel similar to the planned SB11 discard. Following this evaluation, the flowsheet has been slightly altered to address (i) new nuclear criticality safety controls for DWPF that credit a higher amount of gadolinium as a neutron poison for all of the enriched uranium contained in a SB and (ii) potential additions of the H-Canyon neutralized fuel stream prior to the Low Temperature Aluminum Dissolution (LTAD) process in Tank 51. The early introduction of ABD material into the SB assembly process is being investigated to provide flexibility regarding transfer opportunities for H-Canyon to Tank 51 for SB11 and future sludge batches.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

Data Analysis

Analysis of the material protection, control, and accountability (MPC&A) system is necessary to understand the limits and vulnerabilities of the system to internal threats. A self-appraisal helps the facility be prepared to respond to internal threats and reduce the risk of theft or diversion of nuclear material. The material control and accountability (MC&A) system effectiveness tool (MSET) fault tree was developed to depict the failure of the MPC&A system as a result of poor practices and random failures in the MC&A system. It can also be employed as a basis for assessing deliberate threats against a facility. MSET uses fault tree analysis, which is a top-down approach to examining system failure. The analysis starts with identifying a potential undesirable event called a 'top event' and then determining the ways it can occur (e.g., 'Fail To Maintain Nuclear Materials Under The Purview Of The MC&A System'). The analysis proceeds by determining how the top event can be caused by individual or combined lower level faults or failures. These faults, which are the causes of the top event, are 'connected' through logic gates. The MSET model uses AND-gates and OR-gates and propagates the effect of event failure using Boolean algebra. To enable the fault tree analysis calculations, the basic events in the fault tree are populated with probability risk values derived by conversion of questionnaire data to numeric values. The basic events are treated as independent variables. This assumption affects the Boolean algebraic calculations used to calculate results. All the necessary calculations are built into the fault tree codes, but it is often useful to estimate the probabilities manually as a check on code functioning. The probability of failure of a given basic event is the probability that the basic event primary question fails to meet the performance metric for that question. The failure probability is related to how well the facility performs the task identified in that basic event over time (not just one performance or exercise). Fault tree calculations provide a failure probability for the top event in the fault tree. The basic fault tree calculations establish a baseline relative risk value for the system. This probability depicts relative risk, not absolute risk. Subsequent calculations are made to evaluate the change in relative risk that would occur if system performance is improved or degraded. During the development effort of MSET, the fault tree analysis program used was SAPHIRE. SAPHIRE is an acronym for 'Systems Analysis Programs for Hands-on Integrated Reliability Evaluations.' Version 1 of the SAPHIRE code was sponsored by the Nuclear Regulatory Commission in 1987 as an innovative way to draw, edit, and analyze graphical fault trees primarily for safe operation of nuclear power reactors. When the fault tree calculations are performed, the fault tree analysis program will produce several reports that can be used to analyze the MPC&A system. SAPHIRE produces reports showing risk importance factors for all basic events in the operational MC&A system. The risk importance information is used to examine the potential impacts when performance of certain basic events increases or decreases. The initial results produced by the SAPHIRE program are considered relative risk values. None of the results can be interpreted as absolute risk values since the basic event probability values represent estimates of risk associated with the performance of MPC&A tasks throughout the material balance area (MBA). The RRR for a basic event represents the decrease in total system risk that would result from improvement of that one event to a perfect performance level. Improvement of the basic event with the greatest RRR value produces a greater decrease in total system risk than improvement of any other basic event. Basic events with the greatest potential for system risk reduction are assigned performance improvement values, and new fault tree calculations show the improvement in total system risk. The operational impact or cost-effectiveness from implementing the performance improvements can then be evaluated. The improvements being evaluated can be system performance improvements, or they can be potential, or actual, upgrades to the system. The RIR for a basic event represents the increase in total system risk that would result from failure of that one event. Failure of the basic event with the greatest RIR value produces a greater increase in total system risk than failure of any other basic event. Basic events with the greatest potential for system risk increase are assigned failure performance values, and new fault tree calculations show the increase in total system risk. This evaluation shows the importance of preventing performance degradation of the basic events. SAPHIRE identifies combinations of basic events where concurrent failure of the events results in failure of the top event.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2016

SU-F-T-268: A Feasibility Study of Independent Dose Verification for Vero4DRT

Purpose: Vero4DRT (Mitsubishi Heavy Industries Ltd.) has been released for a few years. The treatment planning system (TPS) of Vero4DRT is dedicated, so the measurement is the only method of dose verification. There have been no reports of independent dose verification using Clarksonbased algorithm for Vero4DRT. An independent dose verification software program of the general-purpose linac using a modified Clarkson-based algorithm was modified for Vero4DRT. In this study, we evaluated the accuracy of independent dose verification program and the feasibility of the secondary check for Vero4DRT. Methods: iPlan (Brainlab AG) was used as the TPS. PencilBeam Convolution was used for dose calculation algorithm of IMRT and X-ray Voxel Monte Carlo was used for the others. Simple MU Analysis (SMU, Triangle Products, Japan) was used as the independent dose verification software program in which CT-based dose calculation was performed using a modified Clarkson-based algorithm. In this study, 120 patients’ treatment plans were collected in our institute. The treatments were performed using the conventional irradiation for lung and prostate, SBRT for lung and Step and shoot IMRT for prostate. Comparison in dose between the TPS and the SMU was done and confidence limits (CLs, Mean ± 2SD %) were compared to those from the general-purpose linac. Results: As the results of the CLs, the conventional irradiation (lung, prostate), SBRT (lung) and IMRT (prostate) show 2.2 ± 3.5% (CL of the general-purpose linac: 2.4 ± 5.3%), 1.1 ± 1.7% (−0.3 ± 2.0%), 4.8 ± 3.7% (5.4 ± 5.3%) and −0.5 ± 2.5% (−0.1 ± 3.6%), respectively. The CLs for Vero4DRT show similar results to that for the general-purpose linac. Conclusion: The independent dose verification for the new linac is clinically available as a secondary check and we performed the check with the similar tolerance level of the general-purpose linac. This research is partially supported by Japan Agency for Medical Research and Development (AMED)

60 APPLIED LIFE SCIENCES↗
DOE OSTI2011

EVALUATION OF SOLID SORBENTS AS A RETROFIT TECHNOLOGY FOR CO2 CAPTURE FROM COAL-FIRED POWER PLANTS

Through a U.S. Department of Energy (DOE) National Energy Technology Laboratory (NETL) funded cooperative agreement DE-NT0005649, ADA Environmental Solutions (ADA) has begun evaluating the use of solid sorbents for CO{sub 2} capture. The project objective was to address the viability and accelerate development of a solid-based CO{sub 2} capture technology. To meet this objective, initial evaluations of sorbents and the process/equipment were completed. First the sorbents were evaluated using a temperature swing adsorption process at the laboratory scale in a fixed-bed apparatus. A slipstream reactor designed to treat flue gas produced by coal-fired generation of nominally 1 kWe was designed and constructed, which was used to evaluate the most promising materials on a more meaningful scale using actual flue gas. In a concurrent effort, commercial-scale processes and equipment options were also evaluated for their applicability to sorbent-based CO{sub 2} capture. A cost analysis was completed that can be used to direct future technology development efforts. ADA completed an extensive sorbent screening program funded primarily through this project, DOE NETL cooperative agreement DE-NT0005649, with support from the Electric Power Research Institute (EPRI) and other industry participants. Laboratory screening tests were completed on simulated and actual flue gas using simulated flue gas and an automated fixed bed system. The following types and quantities of sorbents were evaluated: 87 supported amines; 31 carbon based materials; 6 zeolites; 7 supported carbonates (evaluated under separate funding); and 10 hydrotalcites. Sorbent evaluations were conducted to characterize materials and down-select promising candidates for further testing at the slipstream scale. More than half of the materials evaluated during this program were supported amines. Based on the laboratory screening four supported amine sorbents were selected for evaluation at the 1 kW scale at two different field sites. ADA designed and fabricated a slipstream pilot to allow an evaluation of the kinetic behavior of sorbents and provide some flexibility for the physical characteristics of the materials. The design incorporated a transport reactor for the adsorber (co-current reactor) and a fluidized-bed in the regenerator. This combination achieved the sorbent characterization goals and provided an opportunity to evaluate whether the potential cost savings associated with a relatively simple process design could overcome the sacrifices inherent in a co-current separation process. The system was installed at two field sites during the project, Luminant's Martin Lake Steam Electric Station and Xcel Energy's Sherburne County Generating Station (Sherco). Although the system could not maintain continuous 90% CO{sub 2} removal with the sorbents evaluated under this program, it was useful to compare the CO{sub 2} removal properties of several different sorbents on actual flue gas. One of the supported amine materials, sorbent R, was evaluated at both Martin Lake and Sherco. The 1 kWe pilot was operated in continuous mode as well as batch mode. In continuous mode, the sorbent performance could not overcome the limitations of the cocurrent adsorbent design. In batch mode, sorbent R was able to remove up to 90% CO{sub 2} for several cycles. Approximately 50% of the total removal occurred in the first three feet of the adsorption reactor, which was a transport reactor. During continuous testing at Sherco, CO{sub 2} removal decreased to approximately 20% at steady state. The lack of continuous removal was due primarily to the combination of a co-current adsorption system with a fluidized bed for regeneration, a combination which did not provide an adequate driving force to maintain an acceptable working CO{sub 2} capacity. In addition, because sorbent R consisted of a polymeric amine coated on a silica substrate, it was believed that the 50% amine loaded resulted in mass diffusion limitations related to the CO{sub 2} uptake rate. Three additional supported amine materials, sorbents AX, F, and BN, were selected for evaluation using the 1 kW pilot at Sherco. Sorbent AX was operated in batch mode and performed similarly to sorbent R (i.e. could achieve up to 90% removal when given adequate regeneration time). Sorbent BN was not expected to be subject to the same mass diffusion limitations as experienced with sorbent R. When sorbent BN was used in continuous mode the steady state CO{sub 2} removal was approximately double that of sorbent R, which highlighted the importance of sorbents without kinetic limitations.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2014

Overview of the 2014 Edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook)

The International Reactor Physics Experiment Evaluation Project (IRPhEP) is a widely recognized world class program. The work of the IRPhEP is documented in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Integral data from the IRPhEP Handbook is used by reactor safety and design, nuclear data, criticality safety, and analytical methods development specialists, worldwide, to perform necessary validations of their calculational techniques. The IRPhEP Handbook is among the most frequently quoted reference in the nuclear industry and is expected to be a valuable resource for future decades.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

The Office of Groundwater & Soil Remediation Fiscal Year 2011 Research & Development Program

The U.S. Department of Energy's (DOE) Office of Groundwater and Soil Remediation supports technology development and technical assistance for the remediation of environments contaminated by legacy nuclear waste. The core of the program is centered on delivering proactive, responsive expertise and technologies with highly-leveraged, carefully selected investments that maximum impact on life-cycle cleanup costs and risks across the DOE complex. The program currently focuses on four main priorities: improved sampling and characterization strategies, advanced predictive capabilities, enhanced remediation methods, and improved long-term performance evaluation and monitoring. In FY 2010, the program developed a detailed research and development (R and D) plan in support of a larger initiative to integrate R and D efforts across EM. This paper provides an overview of the priority action areas and the program's near-term technical direction.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

High Extraction Phosphors for Solid State Lighting

We have developed high-index, high efficiency bulk luminescent materials and novel nano-sized phosphors for improved solid-state white LED lamps. These advances can potentially contribute to reducing the loss in luminous efficiencies due to scattering, re-absorption, and thermal quenching. The bulk and nanostructured luminescent materials investigated are index matched to GaN and have broad and size-tunable absorption bands, size and impurity tuned emission bands, size-driven elimination of scattering effects, and a separation between absorption and emission bands. These innovations were accomplished through the use of novel synthesis techniques suitable for high volume production for LED lamp applications. The program produced a full-color set of high quantum yield phosphors with high chemical stability. In the bulk phosphor study, the ZnSeS:Cu,Ag phosphor was optimized to achieve >91% efficiency using erbium (Er) and other activators as sensitizers. Detailed analysis of temperature quenching effects on a large number of ZnSeS:Cu,Ag,X and strontium- and calcium-thiogallate phosphors lead to a breakthrough in the understanding of the anti-quenching behavior and a physical bandgap model was developed of this phenomena. In a follow up to this study, optimized phosphor blends for high efficiency and color performance were developed and demonstrated a 2-component phosphor system with good white chromaticity, color temperature, and high color rendering. By extending the protocols of quantum dot synthesis, large nanocrystals, greater than 20 nm in diameter were synthesized and exhibited bulk-like behavior and blue light absorption. The optimization of ZnSe:Mn nanophosphors achieved ~85% QE The limitations of core-shell nanocrystal systems were addressed by investigating alternative deltadoped structures. To address the manufacturability of these systems, a one-pot manufacturing protocol was developed for ZnSe:Mn nanophosphors. To enhance the stability of these material systems, the encapsulation of ZnSeS particle phosphors and ZnSeS screens with Al 2 O 3 and TiO 2 using ALD was shown to improve the stability by >8X and also increased the luminescence efficiency due to improved surface passivation and optical coupling. A large-volume fluidized bed ALD system was designed that can be adapted to a commercial ALD or vapor deposition system. Throughout the program, optical simulations were developed to evaluate and optimize various phosphor mixtures and device configurations. For example, to define the scattering properties of nanophosphors in an LED device or in a stand-off screen geometry. Also this work significantly promoted and assisted in the implementation of realistic phosphor material models into commercial modeling programs.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗