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At least 307 records · Page 17
All sourcesNASA NTRSDOE OSTI
DOE OSTI2010

Progress on the NSTX Center Stack Upgrade

The National Spherical Torus Experiment (NSTX) will be upgraded to provide increased toroidal field, plasma current and pulse length. This involves the replacement of the so-called center stack, including the inner legs of the Toroidal Field (TF) coil, the Ohmic Heating (OH) coil, and the inner Poloidal Field (PF) coils. In addition the increased performance of the upgrade requires qualification of remaining existing components for higher loads. Initial conceptual design efforts were based on worst-case combinations of possible currents that the power supplies could deliver. This proved to be an onerous requirement and caused many of the outer coils support structures to require costly heavy reinforcement. This has led to the planned implementation of a Digital Coil Protection System (DCPS) to reduce design-basis loads to levels that are more realistic and manageable. As a minimum, all components must be qualified for the increase in normal operating loads with headroom. Design features and analysis efforts needed to meet the upgrade loading are discussed. Mission and features of the DCPS are presented.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Overview of LCLS-II Project Status at Fermilab

The superconducting RF Continuous-Wave (CW) Linac for the LCLS-II consists of thirty-five 1.3 GHz and two 3.9 GHz cryomodules that Fermilab and Jefferson Lab are jointly producing in collaboration with SLAC. Fermilab’s scope of work is to build, test, and deliver half the 1.3 GHz and all the 3.9 GHz cryomodules and to design and procure components for the cryogenic distribution system. Fermilab has primary responsibility for delivering a working design. The cryomodule design basis was the European XFEL but several elements evolved to meet CW operation requirements and specifics of the SLAC tunnel. There have been several challenges faced during the design, assembly, testing and transportation of the cryomodules which have required design updates. Success in overcoming these challenges is attributable to the strength of the LCLS-II SRF Collaboration (Fermilab, Jefferson Lab and SLAC with extensive help from DESY and CEA/Saclay). The cryogenic distribution system has progressed relatively well and there are valuable Lessons Learned. An overview of the status, accomplishments, problems encountered, solutions developed, and a summary of Lessons Learned will be presented.

43 PARTICLE ACCELERATORS↗
DOE OSTI2019

Review of Submerged Bed Scrubber Assumptions in the Hanford Waste Treatment and Immobilization Plant Bases and Requirements Document

A review was performed of the assumptions in the Hanford Waste Treatment and Immobilization Plant Bases and Requirements Document (BARD) for the Low Activity Waste (LAW) Submerged Bed Scrubber (SBS). This review was performed to evaluate how the assumptions impact the TOPSim process models that utilize the assumptions for system modeling. The RPP Integrated Flowsheet uses TOPSim to model the entire waste retrieval and treatment process for the Hanford tank waste. This model relies on a number of inputs, including the BARD. The BARD represents the basis for the WTP process models; as such, it contains some assumptions that are valid for the models that will not accurately reflect actual operation. For example, the amount of LAW condensate generated is overstated in the BARD to allow for conservatism in the design of the condensate handling unit operations. The assumptions in the BARD for the LAW SBS were reviewed to evaluate how actual operation could differ from the design basis assumptions. It should be noted that this review did not evaluate nor is it intended to address issues related to design of the WTP facility such that the information in this report is should not be viewed as issues with the design of the WTP facility nor should the information be used without additional evaluation to generate design concerns.

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

U.S. Domestic Molten Salt Reactor: Security-by-Design

U.S. nuclear power facilities face increasing challenges in meeting dynamic security requirements caused by evolving and expanding threats while keeping costs reasonable to make nuclear energy competitive. The past approach has often included implementing security features after a facility has been designed and without attention to optimization, which can lead to cost overruns. Incorporating security in the design process can provide robust, economical, and effective physical protection systems (PPS). The purpose of this work is both to develop a framework for the integration of security into the design phase of a molten salt reactor (MSR) and show how to effectively design a PPS with a reduced staffing headcount. Specifically, this work focuses on integrating PPS design features into a developed facility layout by making minor modifications to building structures. A suite of tools, including Scribe3D©, PathTrace©, and Blender©, were used to model a hypothetical, generic domestic MSR facility. Physical protection elements such as sensors, cameras, barriers, and responders were added into the model based on defending the hypothetical MSR facility against a hypothetical design basis threat (DBT). Multiple outsider sabotage scenarios were examined, with adversary team sizes ranging from 4–8 to determine security system effectiveness. The results of this work will influence PPS designs and facility designs for U.S. domestic MSRs. This work will also demonstrate how a series of experimental and modeling capabilities across the Department of Energy (DOE) complex can impact the design and completion of security-by-design (SeBD) for small modular reactors (SMRs). The conclusions and recommendations in this document may be applicable to all SMR designs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2010

Northern California CO 2 Reduction Project

C6 Resources LLC, a wholly owned subsidiary of Shell Oil Company, worked with the US Department of Energy (DOE) under a Cooperative Agreement to develop the Northern California CO 2 Reduction Project. The objective of the Project is to demonstrate the viability of using Carbon Capture and Sequestration (CCS) to reduce existing greenhouse gas emissions from industrial sources on a large-scale. The Project will capture more than 700,000 metric tonnes of CO 2 per year, which is currently being vented to the atmosphere from the Shell Martinez Refinery in Contra Costa County. The CO 2 will be compressed and dehydrated at the refinery and then transported via pipeline to a sequestration site in a rural area in neighboring Solano County. The CO 2 will be sequestered into a deep saline formation (more than two miles underground) and will be monitored to assure secure, long-term containment. The pipeline will be designed to carry as much as 1,400,000 metric tonnes of CO 2 per year, so additional capacity will be available to accommodate CO 2 captured from other industrial sources. The Project is expected to begin operation in 2015. The Project has two distinct phases. The overall objective of Phase 1 was to develop a fully definitive design basis for the Project. The Cooperative Agreement with the DOE provided cost sharing for Phase 1 and the opportunity to apply for additional DOE cost sharing for Phase 2, comprising the design, construction and operation of the Project. Phase 1 has been completed. DOE co-funding is provided by the American Recovery and Reinvestment Act (ARRA) of 2009. As prescribed by ARRA, the Project will stimulate the local economy by creating manufacturing, transportation, construction, operations, and management jobs while addressing the need to reduce greenhouse gas emissions at an accelerated pace. The Project, which will also assist in meeting the CO 2 reduction requirements set forth in California's Climate Change law, presents a major opportunity for both the environment as well as the region. C6 Resources is conducting the Project in collaboration with federally-funded research centers, such as Lawrence Berkeley National Lab and Lawrence Livermore National Lab. C6 Resources and Shell have identified CCS as one of the critical pathways toward a worldwide goal of providing cleaner energy. C6 Resources, in conjunction with the West Coast Regional Carbon Sequestration Partnership (WESTCARB), has conducted an extensive and ongoing public outreach and CCS education program for local, regional and state-wide stakeholders. As part of a long term relationship, C6 Resources will continue to engage directly with community leaders and residents to ensure public input and transparency. This topical report summarizes the technical work from Phase 1 of the Project in the following areas; Surface Facility Preliminary Engineering: summarizes the preliminary engineering work performed for CO 2 capture, CO 2 compression and dehydration at the refinery, and surface facilities at the sequestration site; Pipeline Preliminary Engineering: summarizes the pipeline routing study and preliminary engineering design; Geologic Sequestration: summarizes the work to characterize, model and evaluate the sequestration site; and Monitoring, Verification and Accounting (MVA): summarizes the MVA plan to assure long-term containment of the sequestered CO 2 .

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2019

Utilizing FLEX Equipment for O&M Cost Reduction in Nuclear Power Plants

There are many ways of reducing the O&M costs; this work presents an innovative framework of reducing O&M costs by utilizing the onsite Flex equipment at NPPs. Flex strategies were postulated by the U.S. Nuclear Regulatory Commission (NRC) in the wake of Fukushima Dai-ichi accident to address beyond-design-basis accidents and improve plant flexibility. Onsite Flex includes equipment such as portable pumps, generators, batteries, compressors, and other supporting equipment or tools, all stored in a dedicated and secure building designed to withstand external hazards. In the past years, several NPPs have invested in procuring and maintaining the onsite Flex asset that stands unutilized most of the time. Recently there have been active efforts to develop strategies through which NPPs can take credit for the Flex equipment. This work focuses on identifying areas where Flex equipment can be utilized during normal plant operation and develop a framework that would aide in reduction of O&M costs without impacting plant safety. Earlier work in the current project, published in INL/EXT 18 51531, had presented the risk- and cost-analysis framework for utilizing FLEX portable equipment during technical specification required shutdown due to component failure. The licensee event report (LER) database of the NRC shows that the commercial NPPs in the US reported 86 technical specification required shutdowns since year 2010. When a component failure or unavailability leads to a technical specification required shutdown, the NPPs suffers both direct costs in terms of loss of revenue arising from the loss of generation and indirect costs in form of reporting and inspection required by the NRC. The earlier work developed the following framework and implemented it on a demonstration probabilistic risk assessment (PRA) model to utilize the portable Flex equipment when a component failure could potentially lead to a technical specification required shutdown: 1. Identify the components, the failure or unavailability of which would result in a 10 CFR 50.73(a)(2)(i)A-reportability requirement, postulated by the NRC for technical specification required shutdown to be reported in NRC’s LER database. 2. Identify the Flex equipment that can be utilized as a standby to the failed component. 3. Develop a probabilistic risk assessment (PRA) model that incorporates the Flex equipment within the current plant PRA model. 4. Perform PRA calculations to determine change in core damage frequency and change in risk-informed allowable outage time. 5. Perform cost-benefit analysis to determine the economic feasibility of implementing the Flex equipment The current work, presented in this report, implement the above framework in a plant PRA model obtained from a US commercial nuclear power plant by prestaging onsite FLEX equipment as a redundancy to the existing auxiliary feedwater system, and pre-coordinating with offsite FLEX centers. The new configuration increased the allowable outage time by about a fifty percent overall for the diesel generator being out of service. The benefits to NPPs of utilizing the Flex equipment during normal operations, and performing the PRA developed in this work, include: 1. A risk-informed plant shutdown alternative to the current technical specification required shutdown. 2. Extension of allowable outage time to initiate technical specification required shutdown. 3. Reducing economic impact of component failure, avoiding plant shutdown, and maximizing generation. 4. Save the direct and indirect costs associated with technical specification required shutdown.

42 ENGINEERING↗
DOE OSTI2023

Semi-Annual Report for Horizontal Compact High Temperature Gas Reactor (HC-HTGR) Development during Performance Period October 2022-March 2023

Horizontal Compact High Temperature Gas Reactor (HC-HTGR) is being designed by a multi-disciplinary team of nuclear, mechanical, and structural engineers under the support of a DOE-NE Advanced Reactor Demonstration Program’s Advanced Reactor Concepts-20 (ARC-20) award. The objective of this ARC-20 project is to deliver a conceptual design for the proposed HC-HTGR in 3 years and support its commercialization as a safe, low-cost HTGR. Argonne National Laboratory (Argonne) is responsible for the design and analysis of the reactor cavity cooling system (RCCS) as a safety system for passive decay heat removal of the reactor concept. Additionally, Argonne is providing analysis of the primary heat transport system to ensure temperatures in the reactor systems, structures and components with significant safety margins during normal operation and design basis accident scenarios. This third semi-annual report summarized the progress made at Argonne on the two tasks during the first half of FY23. As a part of the RCCS design task, a design process for the water panel and the system configuration was performed to improve the thermal performance of the RCCS. The updated water panel design had achieved enhanced thermal performance with 0.96 MWth capability with major design updates made in structural interfaces with the RPV and initial configuration of the water tanks. The loop configuration of the RCCS has been proposed to have two independently working loops for system redundancy. A water panel material study was performed focusing on the use of carbon steel in water systems. Additional modeling strategies of primary system thermal fluids analyses were developed to meet modeling needs that are not well suited for the 1D-3D assembly level approach. The first of these is a reduced order assembly model, in which fuel centered unit cells are used to represent a fuel assembly. The 2D approach used in this model is much more computationally efficient, making this model useful for full core transient scenarios where fuel to coolant heat transfer is still the dominant flow path. The next model is the 3D core conduction model to be used to analyze decay heat removal in loss of primary system flow scenarios. Because these scenarios require a large domain to be modeled, a homogenized core model is being pursued to reduce the required computational costs. To accurately model decay heat scenarios it is necessary to couple a RCCS model to the 3D core conduction model. A simplified case is presented to demonstrate how the coupling methodology will be applied to the full core 3D conduction model.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

Mechanical Design and Fabrication of a New RF Power Amplifier for LANSCE

A Full-scale prototype of a new 201.25 MHz RF Final Power Amplifier (FPA) for Los Alamos Neutron Science Center (LANSCE) has been designed, fabricated, assembled and installed in the test facility. This prototype was successfully tested and met the physics and electronics design criteria. The team faced design and manufacturing challenges, having a goal to produce 2 MW peak power at 13% duty factor, at the elevation of over 2 km in Los Alamos. The mechanical design of the final power amplifier was built around a Thales TH628 Diacrode{sup R}, a state-of-art tetrode power tube. The main structure includes Input circuit, Output circuit, Grid decoupling circuit, Output coupler, Tuning pistons, and a cooling system. Many types of material were utilized to make this new RF amplifier. The fabrication processes of the key components were completed in the Prototype Fabrication Division shop at Los Alamos National Laboratory. The critical plating procedures were achieved by private industry. The FPA mass is nearly 600 kg and installed in a beam structural support stand. In this paper, we summarize the FPA design basis and fabrication, plating, and assembly process steps with necessary lifting and handling fixtures. In addition, to ensure the quality of the FPA support structure a finite element analysis with seismic design forces has also been carried out.

Accelerator Design, Technology, and Operations↗
DOE OSTI2015

Mitigation of Micropitting in Wind Turbine Main Shaft Bearings

In recent years, increasing evidence of failures has been reported from spherical roller main bearings used in three-point mounting (TPM) drivetrains of wind turbines. One of the leading causes has been micropitting, a failure mode that is possibly overlooked by design, selection and life-prediction tools. It remains to be seen if retrofitting problematic spherical roller bearings (SRBs) with improved bearing design solutions can improve their durability. Questions to ask might be: ‘Are the operating conditions of the main bearing well understood?’ and ‘Are the failures caused by deficient design practice or other unidentified external sources within the system?’ These questions fundamentally challenge the underlying design basis and encourage the need for a system analysis approach that is currently being undertaken by researchers from the National Renewable Energy Laboratory (NREL). Specifically, this article discusses a few potential design alternatives and system-level reassessment to circumvent micropitting in main bearings used in TPM drivetrains.

17 WIND ENERGY↗
DOE OSTI2018

Defining and Characterizing Methods, Tools, and Computing Resources to Support Pseudo Exhaustive Testability of Software Based I&C Devices

Under the Department of Energy’s Light Water Reactor Sustainability Program, within the Plant Modernization research pathway, the Digital I&C Qualification Project is identifying new methods that would be beneficial in qualifying digital I&C systems and devices for safety-related usage. One such method that would be useful in qualifying field components such as sensors and actuators is the concept of testability. The Nuclear Regulatory Commission (NRC) considers testability to be one of two design attributes sufficient to eliminate consideration of software-based or software logic-based common cause failure (the other being diversity). The NRC defines acceptable “testability” as follows: Testability – A system is sufficiently simple such that every possible combination of inputs and every possible sequence of device states are tested and all outputs are verified for every case (100% tested). [NUREG 0800, Chapter 7, Branch Technical Position (BTP) 7-19] This qualification method has never proven to be practical in view of the very large number of combinations of inputs and sequences of device states for a typical I&C device. However, many of these combinations are not unique in the sense that they represent the same state space or in that that they represent state space that would not affect the critical design basis functions of the device. Therefore, the state space of interest might possibly be reduced to a manageable dimension through such analysis. This project focuses on a representative I&C device similar in design, function, and complexity to the types of devices that would likely be deployed in nuclear power plants as digital or software based sensors and actuators (e.g. Smart Sensors). Analysis will be conducted to determine the feasibility of testing this device in a manner consistent with the NRC definition. This report describes acceptable test methods, needed tools (existing or new), and computing resources, all based on engineering and computer science principles. This information will then be used in a later phase of this project to develop a test specification for I&C device testability that can be used in a future phase of this project to demonstrate digital qualification with respect to common cause failure. Following that, testing will be conducted and the results analyzed to determine if the methods employed are sufficient to eliminate consideration of software common cause failure in accordance with the concept of testability.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Observations and Lessons Learned from Testing of the Hanford Waste Treatment Plant Standard High Solids Vessel Design - 18635

The Hanford Waste Treatment and Immobilization Plant (WTP) recently completed qualification testing of the mixing system for the standard high solids vessel design the U.S. Department of Energy (DOE) proposed to replace several vessels in the design of the plant Pretreatment Facility. Testing was split into two phases. The first phase exercised the pulse jet mixer control system to ensure that the control software could accommodate fluids across the complete range of expected densities and viscosities within the vessel for all operational modes. The second phase of testing was focused on the mixing performance of the vessel, specifically, whether mixing was adequate to support the transfer, de-inventory, throughput, and sampling functions of the vessel. Key observations and lessons learned from testing include: - Successful transfer of waste within and beyond the design basis slurry solids concentration; - Successful resuspension and transfer of large, heavy particles; - Effective performance of multi-level bubblers for process control and monitoring; - Effective controls performance; - Reduction of sparger flowrate from nominal design flow; - Liquid blending within about 15 minutes; - Insights into sample representativeness and uniformity of transfers to inform process control strategies. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

REALIZATION OF A AUTOMATED T-WAY COMBINATORIAL TESTING APPROACH FOR A SOFTWARE BASED EMBEDDED DIGITAL DEVICE

Under the Department of Energy’s Light Water Reactor Sustainability Program, within the Plant Modernization research pathway, the Digital I&C Qualification Project is identifying new methods that would be beneficial in qualifying digital I&C systems and devices for safety-related usage. One such method that would be useful in qualifying field components such as sensors and actuators is the concept of testability. The Nuclear Regulatory Commission (NRC) considers testability to be one of two design attributes sufficient to eliminate consideration of software-based or software logic-based common cause failure (the other being diversity). The NRC defines acceptable “testability” as follows: Testability – A system is sufficiently simple such that every possible combination of inputs and every possible sequence of device states are tested and all outputs are verified for every case (100% tested). [NUREG 0800, Chapter 7, Branch Technical Position (BTP) 7-19] This qualification method has never proven to be practical in view of the very large number of combinations of inputs and sequences of device states for a typical I&C device. However, many of these combinations are not unique in the sense that they represent the same state space or in that that they represent state space that would not affect the critical design basis functions of the device. Therefore, the state space of interest might possibly be reduced to a manageable dimension through such analysis. This project will focus on a representative I&C device similar in design, function, and complexity to the types of devices that would likely be deployed in nuclear power plants as digital or software based sensors and actuators (e.g. Smart Sensors). Analysis will be conducted to determine the feasibility of testing this device in a manner consistent with the NRC definition. This report describes the development of test process for Bounded Exhaustive Testing with respect to Combinatorial Test (CT) methods. The report describes the candidate Device Under Test SW – the VCU smart sensor, conceptual experimental methods for stated test objectives, description of the process, tools, resources, and computing. This information will be used to fully develop a detailed test plan (based on statistical measure needs) and test environment for conducting an I&C device testability demonstration study. The future planed experimental study of this project is to demonstrate digital qualification via Bounded Exhaustive Testability with respect to Common Cause Failure.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Degradation and Failure Phenomena of Advanced Reactor Fuel Concepts: Sodium-Cooled Fast Reactor Metallic Fuel

The U.S. Nuclear Regulatory Commission (NRC) is anticipating licensing applications and commercial use of new fuel types in advanced nuclear power reactors that would be designed and built in the United States. Pacific Northwest National Laboratory (PNNL) is providing technical assistance to the NRC related to the newly proposed nuclear fuel and cladding designs that would be deployed in these reactors. This report focuses specifically on the metallic fuel that is being considered for sodium-cooled fast reactors and specifically on mechanisms that would cause damage or failure to the fuel under reactor operating conditions and design basis accident conditions. There is historic experience with both metallic and oxide fuels in experimental sodium-cooled fast reactors, but this report will focus solely on metallic fuel. Currently two U.S.-based reactor designers are engaged with the NRC in the application or pre-application review and considering a sodium-cooled fast reactor. TerraPower is engaged with NRC in pre-application review of its Natrium reactor. The current design for this fuel uses a sodium-bonded uranium-10wt% zirconium (U10Zr) fuel clad in HT9 stainless steel. ARC Clean Technology is engaged with NRC in pre-application review of its ARC-100 reactor. The current design for this fuel uses a sodium-bonded uranium-10wt% zirconium (U10Zr) fuel with steel cladding. This report will focus on this fuel system specifically, with broader information given regarding other metallic fuel systems with other stainless steel alloy claddings. To support the NRC’s readiness efforts, this report will identify and discuss degradation and failure modes of metallic fuel concepts for sodium-cooled fast reactors, including fuel performance characteristics that may not be addressed within existing regulatory documents.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Joint probability safety assessment for NPP defense infrastructure against extreme external natural hazards

With the increasing tendency of natural hazards, the typhoon, hurricane and tropical Cyclone induced surge, wave, precipitation, flood and wind as extreme external loads menacing Nuclear Power Plants (NPP) in coastal and inland provinces of China. For all of planned, designed And constructed NPP the National Nuclear Safety Administration of China and IAEA recommended Probable Maximum Hurricane /Typhoon/(PMH/T), Probable Maximum Storm Surge (PMSS), Probable Maximum Flood (PMF), Design Basis Flood (DBF) as safety regulations for NPP defense infrastructures. This paper discusses the joint probability analysis of simultaneous occurrence typhoon induced extreme external hazards and compare with IAEA 2006-2009 recommended safety regulation design criteria for some NPP defense infrastructures along China coast. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Integrated deterministic and probabilistic safety analysis for safety assessment of nuclear power plants

The present special issue contains an overview of the research in the field of Integrated Deterministic and Probabilistic Safety Assessment (IDPSA) of Nuclear Power Plants (NPPs). Traditionally, safety regulation for NPPs design and operation has been based on Deterministic Safety Assessment (DSA) methods to verify criteria that assure plant safety in a number of postulated Design Basis Accident (DBA) scenarios. Referring to such criteria, it is also possible to identify those plant Structures, Systems, and Components (SSCs) and activities that are most important for safety within those postulated scenarios. Then, the design, operation, and maintenance of these “safety-related” SSCs and activities are controlled through regulatory requirements and supported by Probabilistic Safety Assessment (PSA).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

WASTE TREATMENT PLANT (WTP) LIQUID EFFLUENT TREATABILITY EVALUATION (SEPTEMBER 2004)

A forecast of the radioactive, dangerous liquid effluents expected to be produced by the Waste Treatment Plant (WTP) was provided by Bechtel National, Inc. (BNI 2004). The forecast represents the liquid effluents generated from the processing of Tank Farm waste through the end-of-mission for the WTP. The WTP forecast is provided in the Appendices. The WTP liquid effluents will be stored, treated, and disposed of in the Liquid Effluent Retention Facility (LERF) and the Effluent Treatment Facility (ETF). Both facilities are located in the 200 East Area and are operated by Fluor Hanford, Inc. (FH) for the US. Department of Energy (DOE). The treatability of the WTP liquid effluents in the LERF/ETF was evaluated. The evaluation was conducted by comparing the forecast to the LERF/ETF treatability envelope (Aromi 1997), which provides information on the items which determine if a liquid effluent is acceptable for receipt and treatment at the LERF/ETF. The format of the evaluation corresponds directly to the outline of the treatability envelope document. Except where noted, the maximum annual average concentrations over the range of the 27 year forecast was evaluated against the treatability envelope. This is an acceptable approach because the volume capacity in the LERF Basin will equalize the minimum and maximum peaks. Background information on the LERF/ETF design basis is provided in the treatability envelope document.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

DEFE0023863 Final Report, Technology for GHG Emission Reduction and Cost-Competitive Mil-Spec Jet Fuel Production using CTL

Ceramatec Inc, in collaboration with IntraMicron (IM), the Energy & Environmental Research Center (EERC) and Sustainable Energy Solutions, LLC (SES), have completed a three-year research project integrating their respective proprietary technologies in key areas to demonstrate production of a jet fuel from coal and biomass sources. The project goals and objectives were to demonstrate technology capable of producing a “commercially-viable quantity” of jet fuel and make significant progress toward compliance with Section 526 of the Energy Independence and Security Act of 2007 (EISA 2007 §526) lifecycle greenhouse gas (GHG) emissions requirements. The Ceramatec led team completed the demonstration of nominal 2 bbl/day Fischer-Tropsch (FT) synthesis pilot plant design, capable of producing a nominal 1 bbl/day in the Jet-A/JP-8 fraction. This production rate would have a capacity of 1,000 gallons of jet fuel per month and provide the design basis of a 100 bbl/day module producing over 2,000 gallons of jet fuel per day. Co-gasification of coal-biomass blends enables a reduction of lifecycle greenhouse gas emissions from equivalent conventional petroleum derived fuel basis. Due to limits of biomass availability within an economic transportation range, implementation of a significant biomass feed fraction will require smaller plants than current world scale CTL and GTL FT plants. Hence a down-scaleable design is essential. The pilot plant design leverages Intramicron’s MicroFiber Entrapped Catalyst (MFEC) support which increases the catalyst bed thermal conductivity two orders of magnitude, allowing thermal management of the FT reaction exotherm in much larger reactor tubes. In this project, single tube reactors having 4.5 inch outer diameter and multi-tube reactors having 4 inch outer diameters were operated, with productivities as high as 1.5 gallons per day per linear foot of reactor tube. A significant reduction in tube count results from the use of large diameter reactor tubes, with an associated reduction in reactor cost. The pilot plant was designed with provisions for product collection capable of operating with conventional wax producing FT catalysts but was operated with a Chevron hybrid wax-free FT catalyst. Process simplification enabled by elimination of the wax hydrocracking process unit provides economic advantages in scaling to biomass capable plant sizes. Intramicron also provided a sulfur capture system based on their Oxidative Sulfur Removal (OSR) catalyst process. The integrated sulfur removal and FT systems were operated with syngas produced by the Transport Reactor Development Unit (TRDU) gasifier at the University of North Dakota EERC. SES performed modeling of their cryogenic carbon capture process on the energy, cost and CO2 emissions impact of the Coal-biomass synthetic fuel process.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2025

Modularization of Small Modular Reactor Facilities: Physical Protection Recommendations

U.S. nuclear power facilities face increasing challenges in meeting dynamic security requirements caused by evolving and expanding threats while keeping costs reasonable to make nuclear energy competitive. In consideration of the economic viability of small modular reactor (SMR) facilities, many designers and utilities are considering a modular approach to the construction and operation of these facilities. This modular approach considers building and operating a first unit; once the first reactor is in operation, construction will begin on a second reactor. This process would allow the vendor or utility to ensure production of energy and financial income while the second reactor is under construction. This project evaluates the feasibility of modular construction in terms of physical protection and identifies several recommendations for vendors and utilities considering this approach. To conduct this evaluation, a hypothetical three-unit SMR facility was developed, a physical protection system (PPS) was considered for the design, and a hypothetical design basis threat (DBT) was used to evaluate this PPS. Multiple outsider sabotage scenarios were examined, with adversary team sizes ranging from 4-to-8 to determine security system effectiveness. The results of this work will influence PPS designs and facility designs for U.S. domestic SMRs. This work will also demonstrate how a series of experimental and modeling capabilities across the Department of Energy (DOE) complex can impact the design and completion of security-by-design (SeBD) for SMRs considering modular construction. The conclusions and recommendations in this document may be applicable to all SMR designs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗