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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Design of separate-effects In-Pile transient boiling experiments at the $\mathrm{TREAT}$ Facility

The cladding-to-coolant heat transfer during rapid power excursions, such as reactivity-initiated accidents, remains a crucial area of uncertainty in nuclear reactor safety. This uncertainty impacts the ability to accurately predict fuel performance behavior for these conditions. Improving our understanding of transient cladding-to-coolant heat transfer will enhance our ability to model design basis accidents and could increase design and improve safety margins for the current commercial fleet and advanced reactors. To address these issues, the Critical Heat Flux Static Environment Rodlet Transient Test Apparatus (CHF-SERTTA) experiment uses a novel approach with a borated stainless-steel heater rodlet that can replicate cladding temperatures experienced during a design-basis reactivity-initiated accident. The rodlet and experiment is instrumented to provide temperature and thermal-hydraulic conditions throughout the transient. The design of the rodlet allows for separate effect testing that eliminates complexities with a fuel and cladding specimen. We report the experiments have provided data that will be used to improve our understanding of boiling behavior, specifically critical heat flux, under transient conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1983

Theoretical basis for design of thermal-stress-free fasteners

A theoretical basis was developed for the design of fasteners which are free of thermal stress. A fastener can be shaped to eliminate the thermal stress which would otherwise result from differential thermal expansion between dissimilar fastener and sheet materials for many combinations of isotropic and orthotropic materials. The resulting joint remains snug, yet free of thermal stress at any temperature, if the joint is uniform in temperature, if it is frictionless, and if the coefficients of thermal expansion of the materials do not change with temperature. In general, such a fastener has curved sides; however, if both materials have isotropic coefficients of thermal expansion, a conical fastener is free of thermal stress. Equations are presented for thermal stress free shapes at both initial and final temperature, and typical fastener shapes are shown.

Blosser, M. L.↗
DOE OSTI2021

Development of phenomena identification and ranking table for Westinghouse lead fast reactor’s safety

The Westinghouse Lead-cooled Fast Reactor (LFR) is a medium-size, passively safe, economic, Gen-IV nuclear reactor. An important effort within the Westinghouse LFR program is the development of the safety analysis methodology, which comprises computer code development, model development, and experimental testing. A key initial task associated with the development of the safety analysis methodology is the identification of processes and phenomena that affect the plant's capability to meet selected safety performance indicators. This is accomplished through the development of a Phenomena Identification and Ranking Table (PIRT) for selected accident scenarios which, for this specific PIRT effort, included selected postulated design basis accidents and hypothetical beyond design basis accidents in LFRs. This paper describes the role of PIRT in the development of the Westinghouse LFR safety analysis methodology and the process used in the PIRT development. Specifically, the Westinghouse LFR PIRT assessed importance of pertinent phenomena and identified gaps in their knowledge-base by evaluating current modeling capabilities and data available for validation. The ultimate goal was to provide guidance on computer code development and validation efforts and to prioritize testing to support LFR design and licensing. The key phenomena and processes that are deemed highly important for the safety performance indicators, but for which the state of knowledge is low, are presented. The testing program and analyses development are planned to address significant phenomena in the PIRT.

Lead fast reactor↗
DOE OSTI2019

High temperature oxidation and microstructural evolution of cold spray chromium coatings on Zircaloy-4 in steam environments

Oxidation kinetics and microstructural evolution of cold sprayed Cr coatings on Zircaloy-4 at 1130–1310 °C in flowing steam at atmospheric pressure have been studied. The study is aimed at understanding of the response of Cr coated Zr-alloy under a steam environment and high temperatures pertinent to design basis accidents (DBAs) and beyond design basis accidents (BDBAs) in light water reactors (LWRs). Surface morphology, microstructure, and phases of post-oxidation test samples were characterized using Scanning Electron Microscopy (SEM), x-ray diffraction (XRD), and Scanning Transmission Electron Microscopy (STEM). Growth kinetics of the Cr-oxide scale and interdiffusion layers between the Cr coating and the Zr-alloy substrate were quantified from cross-sectional SEM images. Cross-sectional analysis showed that the Cr coatings offered a 50 times reduction in oxidation rate over bare Zircaloy-4 in a 1310 °C steam environment. Oxidation kinetics at 1130 °C followed parabolic law (i.e., n ~ 0.5 in power law kinetics) but at 1230 °C and 1310 °C the value of n was suppressed to below 0.5 possibly due to the volatilization of Cr species at the two highest temperatures. Interdiffusion at coating/substrate interface resulted in formation of a brittle Cr 2 Zr or Zr(Fe, Cr) 2 intermetallic compound layer on the order of micrometer in thickness and scattered Cr-rich precipitates were observed well below the interface within the Zr-alloy substrate after cooling. These experimental results could provide data to LWR system simulation codes for better estimation of coping time in the event of accidents.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Hot Cell Installation and Demonstration of the Severe Accident Test Station

A Severe Accident Test Station (SATS) capable of examining the oxidation kinetics and accident response of irradiated fuel and cladding materials for design basis accident (DBA) and beyond design basis accident (BDBA) scenarios has been successfully installed and demonstrated in the Irradiated Fuels Examination Laboratory (IFEL), a hot cell facility at Oak Ridge National Laboratory. The two test station modules provide various temperature profiles, steam, and the thermal shock conditions necessary for integral loss of coolant accident (LOCA) testing, defueled oxidation quench testing and high temperature BDBA testing. The installation of the SATS system restores the domestic capability to examine postulated and extended LOCA conditions on spent fuel and cladding and provides a platform for evaluation of advanced fuel and accident tolerant fuel (ATF) cladding concepts. This document reports on the successful in-cell demonstration testing of unirradiated Zircaloy-4. It also contains descriptions of the integral test facility capabilities, installation activities, and out-of-cell benchmark testing to calibrate and optimize the system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Baseline Fuel Fabrication Facility

PRO-RR is the research reactor focused program element of the broader Proliferation Resistance Optimization program (PRO-X) under the National Nuclear Safety Administration (NNSA) in the U.S. Department of Energy (DOE). PRO-X provides a framework for integrating proliferation resistance in nuclear system designs to minimize weapons usable nuclear materials (WUNM) production and diversion pathways while optimizing systems performance for peaceful use missions. PRO-RR applies the PRO-X mission objectives to research reactor system design. This document serves as one of the foundational documents for the PRO-RR-Fuel System Design technical team by documenting a baseline fuel fabrication facility to be used for further optimization studies. The PRO-RR-Fuel System Design technical team consists of subject matter experts from Argonne National Laboratory (Argonne) and Savannah River National Laboratory (SRNL). In order to develop specific strategies for fuel fabrication facilities to optimize proliferation resistance, performance, and safety, a baseline fuel fabrication facility design basis was developed. Having a baseline design basis allows for the qualitative and quantitative comparison of design choices in the optimization process. This report describes the baseline fuel fabrication facility and general optimization strategy. Chapter 2 describes the fuel system selected for examination, the fabrication process used as the baseline, a description of the model developed to track uranium utilization, and a generic floorplan of the fabrication facility. Chapter 3 describes the overarching optimization strategy that could be implemented for a fabrication facility.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Integral LOCA fragmentation test on high-burnup fuel

Increasing the fuel burnup limit in light water reactors is sought to enhance fuel cycle economics and requires establishing a technical basis. Experimental observations of severe fuel fragmentation under loss-of-coolant-accident conditions at Halden and Studsvik had raised the need for additional considerations during the development of this technical basis. These test data suggested that the burnup threshold for high burnup fuel fragmentation may be influenced by pre-transient power. Additional loss-of-coolant-accident test data is therefore valuable to complement these tests and enhance the current state of understating. Oak Ridge National Lab has developed the Severe Accident Test Station capable of examining the oxidation kinetics and accident response of irradiated fuel and cladding materials for design basis accident and beyond design basis accident scenarios. Severe Accident Test Station provides various temperature profiles, steam, and the thermal shock conditions necessary for integral loss of coolant accident testing, defueled oxidation quench testing, and high-temperature beyond design basis accident testing. Severe Accident Test Station has been successfully installed and demonstrated in the Irradiated Fuels Examination Laboratory at Oak Ridge National Lab. Furthermore, descriptions of the in-cell re-fabrication capabilities and assembly of the loss-of-coolant-accident test train are provided. Installation of the Severe Accident Test Station system and in-cell re-fabrication restores United States capability to examine postulated and extended loss-of-coolant-accident conditions on spent fuel and cladding and provides a platform for evaluating advanced fuel and accident-tolerant fuel cladding concepts. Lastly, three in-cell integral loss-of-coolant-accident test were performed in the Severe Accident Test Station and compared to the Nuclear Regulatory Commission sponsored loss-of-coolant-accident test as well as the original loss-of-coolant-accident test performed at Argonne National Lab. Finally, the results of these tests as well as all publicly available integral loss-of-coolant-accident test were used to validate a threshold for high burnup fuel fragmentation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Panel discussion: Challenges and improvements in accident dose analysis; Regulatory and industry perspective

Full text of publication follows. Design basis accident (DBA) analyses are one of several ways nuclear power plant licensees demonstrate, and the U.S. Nuclear Regulatory Commission (NRC) staff verifies, that the design and location of a nuclear power plant are appropriate and that the protection of health and safety of the public is maintained. In July 2000, the NRC staff issued Regulatory Guide 1.183, 'Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors' (RG 1.183) [1]. RG 1.183 provides a method acceptable to the NRC staff for complying with the regulatory requirements contained in Title 10 of the Code of Federal Regulation (10 CFR) Section 50.67 (50.67), 'Accident source term' [2]. RG 1.183 also provides assumptions and parameters used to model postulated DBA radiological analyses for light-water reactors (LWR). Since the initial issuance of RG 1.183, the NRC staff and the nuclear power plant licensees have both gained substantial experience with the implementation of 10 CFR 50.67 and RG 1.183. Based on this experience, comments from licensees, the anticipation of licensing advanced LWRs, and new research, the NRC is developing a proposed revision to RG 1.183. This panel discussion will provide a discussion of that proposed revision, its potential impact on existing and future evaluations, and the industry perspective on the revised guidance. References: 1) USNRC, Regulatory Guide 1.183, 'Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors,' Revision 0, U.S. Nuclear Regulatory Commission, Washington, D.C. (2000), Electronic copies are also available in ADAMS (http://www.nrc.gov/reading-rm/adams.html), under Accession No. ML003716792. 2) U.S. Code of Federal Regulations, Title 10 (Energy), Part 50 (Domestic Licensing of Production and Utilization Facilities), Section 67, (Accident source term), Office of the Federal Register, Washington, DC (1999). (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Quarterly Technical Progress Report Piperazine Advanced Stripper (PZAS™) Front End Engineering Design

EXECUTIVE SUMMARY This document summarizes the status of Cooperative Agreement DE-FE0031844, “Piperazine Advanced Stripper (PZAS™) Front-End Engineering Design,” during the reporting period of January 1 through March 31, 2020. The objective of this project is to develop accurate installed costs by conducting a Front-End Engineering Design (FEED) of PZAS™ at Golden Spread Electric Cooperative’s (GSEC) Mustang Station located in Denver City, TX. Complementary benefits include positioning the technology for a commercial project with the 45Q tax credits, qualifying PZAS™ for use on a Natural Gas Combined Cycle (NGCC) Cogen facility, and to provide cost detail to optimize PZAS™ and help guide R&D of second-generation solvent CO2 capture technologies. Results from the FEED will be used to evaluate the economic feasibility of the process at Mustang Station. This project is funded by the U.S. DOE National Energy Technology Laboratory under the aforementioned Cooperative Agreement. Exxon, Total, Chevron, UOP-Honeywell, and the University of Texas (UT) are project co-funders. AECOM and Trimeric are project team members; UT is the prime contractor. Summary of Progress Cooperative Agreement DE-FE0031844 was established in October 2019. The current reporting period, January 1 through March 31, 2020, is the second technical progress reporting period for the project. Several milestones were accomplished during this reporting period, including: • Kickoff Meeting with DOE on February 3, 2020. • Kickoff Meeting with GSEC on March 30, 2020. (Note: Due to Covid-19 travel restrictions and shelter-in-place guidelines, the kickoff meeting was conducted remotely via videoconferencing. See attached notes from that telecon.) • Updated Project Management Plan March 2020, submitted with this quarterly report Other activities during the quarter included progress on contracting and other legal agreements (e.g., non-disclosure agreements), internal kickoff meetings at both AECOM and Trimeric, and development of a Technical Implementation Plan (TIP). The TIP will help the team to make critical, early process decisions and, ultimately, to develop a project and process design basis. Note that all agreements between project team participants are complete as of this submittal, except the vendor agreement with Kiewit (steam cycle modeling). Plans for Next Reporting Period Activities during the next reporting period (April 1, 2020 through June 30, 2020) include: the completion of the Project Design Basis and progress towards the Process Design Basis/Process Design Package (PDP). The Project Design Basis is due as a deliverable and milestone during the next reporting period.

Rochelle, Gary T.↗
DOE OSTI2019

Development of LWR Fuels with Enhanced Accident Tolerance ATF Feasibility Analysis and Final Technical Report Deliverable for the Westinghouse Accident Tolerant Fuel Program (Rev. 2)

The beyond design basis accident at Fukushima Daiichi in March 2011 and the Three Mile Island accident in 1979 imply that severe nuclear accidents may occur at somewhat higher frequencies than previously predicted, and that the financial liabilities of such accidents can cripple a utility. While current fuel is more than adequate for design basis accidents, the potential for harm from beyond design basis accidents (BDBA) has been a driver for Westinghouse development efforts since 2003. After Fukushima, other fuel vendors also began pursuing new fuel materials that can provide significant increases in the time, enabling reactor operators to respond to unforeseen events before significant releases of fuel materials and fission products occur. In order to be commercially successful, accident tolerant fuel (ATF) products must also provide significant fuel cost savings, operating cost improvements as well as safety improvements. This report documents the engineering, design and testing that occurred from October 1, 2016, through September 30, 2018, by the Westinghouse team. In addition, it includes: 1. The results of modeling and simulation work completed to date that supports technical feasibility of the ATF concept in meeting the goals of the enhanced ATF program including a core level analysis predicting how an existing LWR will perform when loaded with ATF. 2. Regulatory Analysis showing path to LTA/LTR acceptance. 3. Business Analysis showing commercial viability of the ATF concept.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Design Load Basis Guidance for Distributed Wind Turbines

Aeroelastic modeling (AM) is the primary methodology for structural and performance assessment of any wind turbine. Nonetheless, the use of AM in the distributed wind (DW) industry sector is limited due to several challenges (Damiani, Davis, & Summerville, 2022). One of these challenges lies in the perceived complexity of generating a proper set of numerical simulations to extract and process the key outputs for component design and verification, and, ultimately, achieve certification. This makes it difficult to reliably predict the structural and performance response of small wind turbines. From the investigation carried out in (Damiani & Davis, 2022), it is apparent that many stakeholders in this sector believe that a comprehensive guide for developing a design load basis (DLB) for distributed wind turbines (DWTs) is necessary.

17 WIND ENERGY↗
DOE OSTI2017

Light Water Reactor Sustainability Program Reactor Safety Technologies Pathway Technical Program Plan

In the aftermath of the March 2011 multi-unit accident at the Fukushima Daiichi nuclear power plant (Fukushima), the nuclear community has been reassessing certain safety assumptions about nuclear reactor plant design, operations and emergency actions, particularly with respect to extreme events that might occur and that are beyond each plant’s current design basis. Because of our significant domestic investment in nuclear reactor technology (99 operating reactors in the fleet of commercial LWRs with five under construction), the United States has been a major leader internationally in these activities. The U.S. nuclear industry is voluntarily pursuing a number of additional safety initiatives. The NRC continues to evaluate and, where deemed appropriate, establish new requirements for ensuring adequate protection of public health and safety in the occurrence of low probability events at nuclear plants; (e.g., mitigation strategies for beyond design basis events initiated by external events like seismic or flooding initiators). The DOE has also played a major role in the U.S. response to the Fukushima accident. Initially, DOE worked with the Japanese and the international community to help develop a more complete understanding of the Fukushima accident progression and its consequences, and to respond to various safety concerns emerging from uncertainties about the nature of and the effects from the accident. DOE R&D activities are focused on providing scientific and technical insights, data, analyses methods that ultimately support industry efforts to enhance safety. These activities are expected to further enhance the safety performance of currently operating U.S. nuclear power plants as well as better characterize the safety performance of future U.S. plants. In pursuing this area of R&D, DOE recognizes that the commercial nuclear industry is ultimately responsible for the safe operation of licensed nuclear facilities. As such, industry is considered the primary “end user” of the results from this DOE-sponsored work. The response to the Fukushima accident has been global, and there is a continuing multinational interest in collaborations to better quantify accident consequences and to incorporate lessons learned from the accident. DOE will continue to seek opportunities to facilitate collaborations that are of value to the U.S. industry, particularly where the collaboration provides access to vital data from the accident or otherwise supports or leverages other important R&D work. The purpose of the Reactor Safety Technology R&D is to improve understanding of beyond design basis events and reduce uncertainty in severe accident progression, phenomenology, and outcomes using existing analytical codes and information gleaned from severe accidents, in particular the Fukushima Daiichi events. This information will be used to aid in developing mitigating strategies and improving severe accident management guidelines for the current light water reactor fleet.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2021

Demonstrate Capability of NEAMS Tools to Generate Reactor Kinetics Parameters for Pebble-Bed HTGRs Transient Modeling

The system analysis of anticipated operating occurrences and design basis accident for pebblebed reactor systems requires knowledge of the neutron kinetic parameters. These parameters take into account various types of feedback from the pebble-bed core and are typically integrated into system analysis tools such as the System Analysis Module (SAM) in the Nuclear Energy Advanced Modeling and Simulation (NEAMS) tool package. These parameters are typically generated by a higher fidelity full-core coupled neutronics/thermal fluids analysis, such as using Griffin and Pronghorn, in 2D or 3D. The delayed neutron fractions (beta) and the neutron lifetimes (lambda) are typically generated by applying the adjoint solution of the neutron flux. Temperature reactivity coefficients are the most important feedback for all anticipated operating occurrences and design basis accident, then depending on the transient other effects might be important as rector control system worth, s-curves, and Xenon generation. This work package will establish a methodology and process for calculating the reactor kinetics parameters for high-temperature gas-cooled reactor and provide them in a suitable form for system level analyses of typical anticipated operating occurrences and design basis accident.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1951

Performance of single-stage compressor designed on basis of constant total enthalpy with symmetrical velocity diagram at all radii and velocity ratio of 0.7 at rotor hub / Jack R. Burtt and Robert J. Jackson

A typical inlet axial-flow compressor inlet stage, which was designed on the basis of constant total enthalpy with symmetrical velocity diagram at all radii, was investigated. At a tip speed of 1126 feet per second, a peak pressure ratio of 1.28 was obtained at an efficiency of 0.76. At a tip speed, the highest practical flow was 28 pounds per second per square foot frontal area with an efficiency of 0.78. Data for a rotor relative inlet Mach number range of from 0.5 to 0.875 indicates that the critical value for any stage radial element is approximately 0.80 for the stage investigated.

FANS↗
DOE OSTI2017

Impacts of common cause failure regulatory requirements on protection system architectures - 304

Regulatory-based design considerations for potential common cause failures (CCF) in digital instrumentation and control (I and C) systems have evolved over time and have impacted the development of protection system I and C architectures. The single failure criterion used to be the main design driver for a protection system I and C architecture and resulted in the familiar four-fold and three-fold system architectures. The single failure criterion has now been replaced by the conservative assumptions associated with digital CCF concerns (i.e., assumed digital CCF coincident with an anticipated operational transient or postulated accident) which are classified as 'beyond design basis events' by regulatory authorities. It is unusual for a beyond design basis event to be the controlling factor for a safety system design but that is now the reality for I and C system designers. The preferred regulatory solution is to introduce diversity into I and C system designs to guard against the general concern of digital CCF. However, the lack of clear criteria on how to define sufficient diversity (i.e., how much, what kind, and whether it is linked to identified digital CCF vulnerabilities) had led to more complex I and C architectures. The trend has been towards lengthy and more difficult reviews for the treatment of digital CCF vulnerabilities and I and C system architectures because of the subjective definition of digital CCF vulnerabilities and the lack of clear acceptance criteria for diversity strategy. Improvement in the treatment of digital CCF is needed to reverse the trend of increased I and C system architecture complexity and longer regulatory reviews. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Reassessing Double-Ended Guillotine Break Requirements: Evidence-Based Analysis of Regulatory Assumptions After Five Decades of Nuclear Operation

After five decades of nuclear power operation encompassing more than 20,000 reactor-years across 35 countries and 647 reactors, zero double-ended guillotine breaks (DEGBs) have been documented in commercial reactor coolant systems—despite DEGB being the fundamental design-basis assumption driving Emergency Core Cooling System (ECCS) sizing, structural protection requirements, and containment design specifications. This report examines the basis for DEGB requirements in nuclear power plant design. The DEGB postulate assumes the instantaneous, complete circumferential severance of the largest diameter pipes in reactor coolant systems, driving major design requirements under 10 Code of Federal Regulations 50.46, General Design Criterion 4 and containment design specifications. The United States (4,880 reactor-years) and France (2,505 reactor-years) contribute the largest operational datasets. Probabilistic assessments estimate direct DEGB occurrence probabilities with extremely low event frequencies, far below the 10-5/reactor-year thresholds typically used to define non-credible events in nuclear-safety analyses; i.e., events with probability this low fall into beyond-design-basis events. Current material-science knowledge demonstrates that the ductile steel materials used in nuclear piping systems exhibit stable crack-growth behavior fundamentally incompatible with instantaneous severance. International regulatory experience, particularly Germany’s comprehensive break-preclusion implementation, and successful leak-before-break (LBB) applications in almost all of U.S. pressurized water reactor units validate that alternatives can maintain safety performance while reducing economic burden. Current DEGB protection systems impose estimated lifetime costs of hundreds of millions of dollars per unit, over the life of a plant across the nuclear industry (including ongoing costs), representing substantial resource allocation toward scenarios with extremely low probability. Although this report acknowledges uncertainties regarding long-term aging effects, potential synergistic degradation mechanisms, and site-specific seismic considerations that warrant continued evaluation as regulatory policy evolves, there remains no documented evidence that a DEGB has occurred as a consequence of the conditions or mechanisms described in this report. This report acknowledges the Nuclear Regulatory Commission’s (NRC’s) recent efforts—outlined in the draft Interim Staff Guidance (ISG) NRC-DSS-ISG-2025-XX (“Treatment of Certain Loss-of-Coolant Accident Locations as Beyond-Design-Basis Accidents Draft Interim Staff Guidance”)—to reduce overly conservative requirements for large-break loss of coolant accidents through technical justifications and exemptions. However, extensive operating experience and validated methodologies—such as LBB and in-service inspection programs—demonstrate that the probability of a DEGB in reactor coolant-loop piping is extremely low, even under seismic conditions. The authors and reviewers of this report recommend that DEGB be removed as a design-basis event through formal rulemaking, rather than case-by-case exemptions, to better reflect credible failure modes, align with current data, and align with modern, risk-informed safety analysis.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Technical Basis for Design Life Extension of the SAVY-4000 Series Nuclear Material Packaging System from Five Years to Fifteen Years

The DOE Manual, M441.1-1, Nuclear Material Packaging, approved March 7, 2008, provides detailed packaging requirements for protecting workers from exposure to nuclear materials stored outside of an approved engineered contamination barrier. The SAVY-4000 Series nuclear material packaging system is a DOE Manual 441.1-1 compliant storage system used at Los Alamos National Laboratory. The purpose of this report is to document the technical basis and DOE approval for the initial 5-year design life of the SAVY-4000 Series and to summarize, with supporting detailed technical reports as references, the technical basis for extending the design life from 5 years to 15 years.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2022

Energetic Basis and Design of Enzyme Function Demonstrated Using GFP, an Excited-State Enzyme

We report the past decades have witnessed an explosion of de novo protein designs with a remarkable range of scaffolds. It remains challenging, however, to design catalytic functions that are competitive with naturally occurring counterparts as well as biomimetic or nonbiological catalysts. Although directed evolution often offers efficient solutions, the fitness landscape remains opaque. Green fluorescent protein (GFP), which has revolutionized biological imaging and assays, is one of the most redesigned proteins. While not an enzyme in the conventional sense, GFPs feature competing excited-state decay pathways with the same steric and electrostatic origins as conventional ground-state catalysts, and they exert exquisite control over multiple reaction outcomes through the same principles. Thus, GFP is an “excited-state enzyme”. Herein we show that rationally designed mutants and hybrids that contain environmental mutations and substituted chromophores provide the basis for a quantitative model and prediction that describes the influence of sterics and electrostatics on excited-state catalysis of GFPs. As both perturbations can selectively bias photoisomerization pathways, GFPs with fluorescence quantum yields (FQYs) and photoswitching characteristics tailored for specific applications could be predicted and then demonstrated. The underlying energetic landscape, readily accessible via spectroscopy for GFPs, offers an important missing link in the design of protein function that is generalizable to catalyst design.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗