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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 343 records · Page 19
All sourcesNASA NTRSDOE OSTI
NASA NTRS1999

GCR Transport in the Brain: Assessment of Self-Shielding, Columnar Damage, and Nuclear Reactions on Cell Inactivation Rates

Radiation shield design is driven by the need to limit radiation risks while optimizing risk reduction with launch mass/expense penalties. Both limitation and optimization objectives require the development of accurate and complete means for evaluating the effectiveness of various shield materials and body-self shielding. For galactic cosmic rays (GCR), biophysical response models indicate that track structure effects lead to substantially different assessments of shielding effectiveness relative to assessments based on LET-dependent quality factors. Methods for assessing risk to the central nervous system (CNS) from heavy ions are poorly understood at this time. High-energy and charge (HZE) ion can produce tissue events resulting in damage to clusters of cells in a columnar fashion, especially for stopping heavy ions. Grahn (1973) and Todd (1986) have discussed a microlesion concept or model of stochastic tissue events in analyzing damage from HZE's. Some tissues, including the CNS, maybe sensitive to microlesion's or stochastic tissue events in a manner not illuminated by either conventional dosimetry or fluence-based risk factors. HZE ions may also produce important lateral damage to adjacent cells. Fluences of high-energy proton and alpha particles in the GCR are many times higher than HZE ions. Behind spacecraft and body self-shielding the ratio of protons, alpha particles, and neutrons to HZE ions increases several-fold from free-space values. Models of GCR damage behind shielding have placed large concern on the role of target fragments produced from tissue atoms. The self-shielding of the brain reduces the number of heavy ions reaching the interior regions by a large amount and the remaining light particle environment (protons, neutrons, deuterons. and alpha particles) may be the greatest concern. Tracks of high-energy proton produce nuclear reactions in tissue, which can deposit doses of more than 1 Gv within 5 - 10 cell layers. Information on rates of cell killing from GCR, including patterns of cell killing from single particle tracks. can provide useful information on expected differences between proton and HZE tracks and clinical experiences with photon irradiation. To model effects on cells in the brain, it is important that transport models accurately describe changes in the GCR due to interactions in the cranium and proximate tissues. We describe calculations of the attenuated GCR particle fluxes at three dose-points in the brain and associated patterns of cell killing using biophysical models. The effects of the brain self-shielding and bone-tissue interface of the skull in modulating the GCR environment are considered. For each brain dose-point, the mass distribution in the surrounding 4(pi) solid angle is characterized using the CAM model to trace 512 rays. The CAM model describes the self-shielding by converting the tissue distribution to mass-equivalent aluminum, and nominal values of spacecraft shielding is considered. Particle transport is performed with the proton, neutron, and heavy-ion transport code HZETRN with the nuclear fragmentation model QMSFRG. The distribution of cells killed along the path of individual GCR ions is modeled using in vitro cell inactivation data for cells with varying sensitivity. Monte Carlo simulations of arrays of inactivated cells are considered for protons and heavy ions and used to describe the absolute number of cell killing events of various magnitude in the brain from the GCR. Included are simulations of positions of inactivated cells from stopping heavy ions and nuclear stars produced by high-energy ions most importantly, protons and neutrons.

Shavers, M. R.↗
DOE OSTI2025

Test and Evaluation of Radiofrequency Tamper Indicating Devices for Remote Monitoring of Advanced/Small Modular Reactors

Advanced and small modular reactors (A/SMRs), due to their versatile nature, are likely to be used in remote locations to provide electrical power or other services in regions that are difficult to access or have limited transportation infrastructure. This will result in limited on-site staff, therefore driving A/SMR vendors to consider remote monitoring as a solution to support nuclear security. Maintaining Continuity of Knowledge (CoK) of nuclear material quantities and locations is vital to nuclear security, and remote monitoring of active tamper indicating devices (TIDs) has been well established as a component of International Atomic Energy Agency (IAEA) Safeguards since the early 2000s. Active TIDs, such as radiofrequency TIDs (RFTIDs), immediately alarm upon unauthorized access attempts, promoting timely detection. In contrast, passive TIDs require a surveillance regime and offer delayed detection. Active TIDs deter insiders and enable prompt detection of malicious acts. They can be used on nuclear material containers and controlled entry points like vaults and toolboxes. Therefore, the implementation of RFTIDs into security programs bolsters overall nuclear material control, and provides a visible deterrent, with primary efficacy in mitigating the insider threat and potentially allowing for Security by Design considerations. They are a strong candidate technology for maintaining nuclear security of A/SMRs but need to be evaluated for feasibility and implementation into the wider physical protection system.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1973

Precise measurement of charged defects in III-V compounds (supplement 2)

Experimental methods and related theory which will permit the measurement of low concentrations of vacancies and other defects in III-V compound semiconductors are discussed. Once the nature of these defects has been determined, this information can be incorporated into a transport theory for devices constructed from these materials, and experiments conducted to test the theory. The vacancies and other defects in the III-V compounds are detected by measurement of the nuclear magnetic resonance (NMR) line width. Most of the III-V compounds have at least one isotope with a nuclear quadrupole moment. In a crystal with a cubic crystal field (characteristic of most III-V compounds) there is no quadrupole splitting of the Zeeman resonance line. However, a defect removes the cubic symmetry locally and causes splitting which result in a change of the NMR width. This change can be used to detect the presence of vacancies.

Soest, J. F.↗
DOE OSTI2020

Radiation Signature Training Device Gamma Equivalent Spheres

Radiation signature training device sources are a continuation of a previous effort to produce a highly enriched uranium equivalent sphere. The original research was designed and motivated to produce a Department of Transportation special form source that emulated the passive gamma signature of a solid metal highly enriched uranium sphere. In addition, the mass reduction (i.e., of the 235 U) minimized logistical costs (security, criticality safety, radiological safety and transportation requirements). Radiation signature training device sources expand on the original concept, redesigning the encapsulation in a modular fashion that allows a variety of loadings. The result is a reconfigurable special nuclear material source set that can vary greatly in the magnitude of the gamma flux and the isotopic distribution. Use of the source to test a variety of standards (N42.34-2015, N42.38-2015, N42.43-2015, N42.53-2018, N42.48-2018, DHS-TCS Backpack Based Radiation Detection Systems, DHS-TCS Handheld Instruments Used for the Detection and Identification of Radionuclides, DHS-TCS Vehicle Mounted Mobile Systems) will be discussed.

Blessinger, Christopher S.↗
DOE OSTI2024

Sister Rod Destructive Examinations (FY23)

As a part of DOE NE High Burnup Spent Fuel Data Project, Oak Ridge National Laboratory (ORNL) is performing destructive examinations (DEs) of high burnup (HBU) (>45 GWd/MTU) spent nuclear fuel (SNF) rods from the North Anna Nuclear Power Station operated by Dominion Energy. The SNF rods, called sister rods or sibling rods, are all HBU and include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin (LT) Zirc-4, ZIRLO, and M5. The DEs are being conducted to obtain a baseline of the HBU rod’s condition before dry storage and are focused on understanding overall SNF rod strength and durability. Composite fuel and defueled cladding will be tested to derive material properties. Although the data generated can be used for multiple purposes, one primary goal for obtaining the post-irradiation examination data and the associated measured mechanical properties is to support SNF dry storage licensing and relicensing activities by (1) addressing identified knowledge gaps and (2) enhancing the technical basis for post-storage transportation, handling, and subsequent disposition. This report provides a cumulative status report of the testing progress, data accumulated, and conclusions reached as a part of the program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Fuel Assembly and Irradiation Parametric Study for Extended-Enrichment and High-Burnup Light-Water Reactor Spent Nuclear Fuel in Dry Storage Casks and Transportation Packages

There is an increased interest in operating commercial light-water reactors (LWRs) in the United States with improved economics that would result from longer fuel cycle lengths, fewer and shorter refueling outages, and fewer fuel assemblies requiring storage at the back end of the fuel cycle. To support this, fuel discharge burnups, as well as initial 235 U enrichments, must be higher than those used in current commercial LWRs. The typical upper limit considered for assembly average burnup in this report is 75 gigawatt-days (GWd) per metric ton of uranium (MTU), as opposed to the current typical upper bound of approximately 62 GWd/MTU. The upper limit considered for initial 235 U enrichment is 8 weight percent (8 wt %), as opposed to the current regulatory limit of 5 wt %. The enrichment range from 5 to 8 wt % is referred to in this report as extended enrichment . To investigate the effect of high burnup and extended enrichment conditions on dose rates and burnup credit for dry storage casks and transportation packages, a fuel assembly and irradiation parametric study was performed. The conclusions from this study will assist U.S. Nuclear Regulatory Commission staff in reviewing applications for dry storage casks and transportation packages that contain high-burnup and extended enrichment fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

EPMA-Based Mass Balance Method for Quantitative Fission Product Distribution Comparison between TRISO Particles

Two irradiated AGR-2 TRISO particles were chosen to demonstrate a recently developed mass balance technique in which EPMA-generated concentration data was used to determine fission product mass on a layer-by-layer basis in TRISO particles. EPMA-calculated fission product masses for most fission products in the two particles were within +/- 20% of their ORIGEN-modelled masses. Results show that Sr, Ba, and Eu accumulate preferentially in the carbon-rich kernel periphery on the particles’ side that lacks a gap between the buffer and IPyC. In addition, the more mobile elements--Cs, Sr, and Pd, accumulate in greater quantity in the outer layers of particle AGR2-223-RS34 compared to particle AGR2-223-RS06, which has relatively more of those elements’ mass located in the kernel and kernel periphery, suggesting enhanced fission product transport in particle AGR2-223-RS34. This model can be used better understand and test fission product transport in TRISO particles.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Speciation and diffusive dynamics in hydrated grain boundaries of complex oxide Gd2Ti2O7

Abstract Grain boundaries in polycrystalline materials significantly affect their properties, such as ionic transport, corrosion, and chemical durability. The pyrochlore compound (Gd 2 Ti 2 O 7 ) is employed as a model for complex oxides and is known for its diverse applications, including nuclear waste immobilization. Density functional theory-based first-principles molecular dynamics simulations were performed at different temperatures on the hydrated grain boundary system. The results show extensive transformations within the grain boundaries among hydrous water species (OH − , H 2 O, and H 3 O + ). The temperature dependence of self-diffusion coefficients follows Arrhenius behavior, with an activation energy of 35.9 kJ/mol for hydrogen and 46.3 kJ/mol for oxygen. The lifetime of OH − is about three to four times longer than that of H 2 O at temperatures from 800 to 2100 K, suggesting the greater stability of OH − over H 2 O, a unique characteristic of the grain boundaries. The estimated lifetime of the hydrous species decreases as the temperature increases, with an activation energy of 9.9 kJ/mol for OH − and 13.4 kJ/mol for H 2 O. While Gd 3 + is more mobile than Ti 4+ , both the Gd 3 + and Ti 4+ cations are orders of magnitude less mobile than the water species. The results suggest that water species are much more mobile within grain boundaries than in the bulk crystal and have the potential to penetrate deep into polycrystalline materials through grain boundaries, leading to grain boundary degradation and dissolution. The different mobilities of cations in complex oxides can lead to leaching of certain cations and incongruent dissolution during the chemical weathering of Earth and industrial materials.

B. Ghosh, Dipta↗
DOE OSTI2020

Panel Session 145: Challenges and Opportunities in Establishing a System for Transportation of SNF/UNF

Issues associated with establishing a SNF Transportation System in the U.S. are complex. In part, this is because in the US when discussing SNF, there are two distinct categories. The two categories represent significant differences in ownership, transportation-regulator, and even the physical characteristics of the SNF. Specifically, the physical differences in the SNF include size of the rods, associated quantities to be shipped, cladding material, and radionuclide content/enrichment. The first is commercial SNF (i.e., that associated with nuclear power plants), owned by utilities, until a permanent repository is built with DOE obligated by law to then assume the responsibility for shipping and ownership of the SNF. Since no-permanent repository currently exists, most commercial SNF is stored near the reactor in which it was installed, hence requiring no shipments. Although current shipments of commercial SNF are very limited, the few that do occur are generally performed to support a utilities SNF consolidation effort utility and for post irradiation examination/testing. However, with approximately 83,000 metric tons currently being stored at more than 70 sites around the U.S., when a permanent repository becomes available, the required number of shipments will be extensive. The second category is DOE and research reactor fuel. These fuel rods are generally much smaller in length than commercial fuel rods, with the quantity of rods transported in a shipment significantly less than the number of rods in a typical commercial fuel assembly. Currently, most of the SNF shipments within the U.S come from this category. DOE is currently managing about 2500 metric tons of heavy metal of SNF. This panel focused on updates and status of transportation of SNF/UNF. Panelists provided updates and information on continuing impacts, risk assessment, equipment (rail cars, handling, securement), possible rail routing, and proposed inspection equipment/routing prior to shipments. Panelists with presentations: Preparing for Nuclear Waste Transportation (Daniel Ogg); Challenges and Opportunities in Establishing a System for Transportation of SNF/UNF (Kathy Langan); UNF Transportation: Challenges and Opportunities (Michael Valenzano)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2003

Benchmark studies of the effectiveness of structural and internal materials as radiation shielding for the international space station

Accelerator-based measurements and model calculations have been used to study the heavy-ion radiation transport properties of materials in use on the International Space Station (ISS). Samples of the ISS aluminum outer hull were augmented with various configurations of internal wall material and polyethylene. The materials were bombarded with high-energy iron ions characteristic of a significant part of the galactic cosmic-ray (GCR) heavy-ion spectrum. Transmitted primary ions and charged fragments produced in nuclear collisions in the materials were measured near the beam axis, and a model was used to extrapolate from the data to lower beam energies and to a lighter ion. For the materials and ions studied, at incident particle energies from 1037 MeV/nucleon down to at least 600 MeV/nucleon, nuclear fragmentation reduces the average dose and dose equivalent per incident ion. At energies below 400 MeV/nucleon, the calculation predicts that as material is added, increased ionization energy loss produces increases in some dosimetric quantities. These limited results suggest that the addition of modest amounts of polyethylene or similar material to the interior of the ISS will reduce the dose to ISS crews from space radiation; however, the radiation transport properties of ISS materials should be evaluated with a realistic space radiation field. Copyright 2003 by Radiation Research Society.

NASA Discipline Radiation Health↗
DOE OSTI2024

Sensitivity Analysis of Irradiated Fueled Experiments using the MOOSE Framework [Slides]

Modeling and simulation (M&S) methods are able to predict uncertainties in experimental parameters (e.g., power and fission density) during irradiation. A shortfall exists in predicting how sensitive some of the parameters will behave during the experimental process. Sensitivity and Uncertainty Quantification (SUQ) is critical in support of qualification and licensing reactor fuels. The application of a method to quantify the uncertainty in these experiments is critical to the prediction of their performance. In this work, we propose the use of a polynomial chaos expansion (PCE) method to quantify the sensitive parameters in these simulations and, in an extension, their experimental surrogates. We propose to perform M&S using PCE uncertainty quantification on a previously irradiated fueled experiment in order to provide a validation case for Griffin and expand its use as a verification and validation (V&V) tool for experiments with a neutronics component. Griffin is an advanced, deterministic neutronics analysis code built using the MOOSE (multiphysics object-oriented simulation environment) framework which can provide state-of-the-art neutronic analysis on M&S of experiments. We will use the stochastic tools module (STM) in MOOSE to provide PCE uncertainty quantification on the proposed experimental setup. Idaho National Laboratory (INL) does not yet have an in-house developed code with V&V approval for experiments performed on-site; this work would provide a necessary addition of support for experiments performed at INL. The Nuclear Regulatory Commission (NRC) has explicitly requested uncertainties in calculated values such as fuel power and burnup, and the development of this capability would benefit the relationship between INL and the NRC.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

ORNL Package Testing Program Overview

When transporting radioactive or hazardous materials, safety and security are top priorities. Packaging and transportation regulations require reliable evidence that containers have passed rigorous performance tests to ensure that the public and the environment are protected from the hazardous nature of the cargo. Because of its long history in energy and security research, Oak Ridge National Laboratory (ORNL) has needed to ship hazardous packages for the past 65 years. Since the 1940s, ORNL has made significant contributions to transportation regulations and has been at the forefront of regulatory testing development. Today, ORNL is a leader in standards development and testing of designs of radioactive material packages. ORNL plays a critical role in ensuring the safe transportation of radioactive materials across the United States by executing rigorous testing campaigns of packages that contain radioactive materials. Early testing activities focused on supporting the development of transportation regulations established by both the International Atomic Energy Agency (IAEA) and the United States. Currently, all package testing activities are performed under the Package Testing Program (PTP) at the Package Evaluation Facility (PEF) located at the National Transportation Research Center (NTRC), which is about 10 miles from the main ORNL campus. The PTP develops and evaluates testing solutions, ensuring that they are safe efficient, and in compliance with regulatory requirements. The vision for the PTP is to be a world-class leader in the evaluation and testing of radioactive and hazardous material packages. The key elements necessary to fulfill this mission are an experienced and professional staff; state-of-the-art facilities, equipment, and instrumentation; and completion of challenging programs and projects important to package transportation. Collaborations with other internal and external organizations play a significant role in building stronger teams and achieving this vision. By adhering to and advancing regulatory standards, ORNL’s PTP supports development of safe and compliant packaging, safeguarding the transportation process from potential risks associated with radioactive material logistics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Verification and Validation of START: A Spent Nuclear Fuel Routing and Decision Support Tool

This is a technical presentation that does not take into account contractual limitations or obligations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). For example, under the provisions of the Standard Contract, spent nuclear fuel in multi-assembly canisters is not an acceptable waste form, absent a mutually agreed to contract amendment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Towards an Era of Low-Temperature Integral Critical Experiments: Surrogate Testing of Low-Temperature TEX Configurations

United States and IAEA regulations for the transportation of radioactive materials require that fissile materials must be subcritical under expected ambient temperatures ranging from -40°C to 38°C. While historical experiments have been conducted at room and elevated temperatures, no criticality integral benchmarks exist for temperatures below room temperature. The Low-Temperature Thermal Epithermal eXperiments (LT-TEX) aim to provide benchmark evaluations down to -40°C. The LT-TEX design incorporates stacked HEU plates interspersed with varying thicknesses of polyethylene disks to provide varying levels of moderation, with five configurations at 20°C having already been benchmarked. A novel vacuum/cryogenic chamber was fabricated to contain the LT-TEX configurations in a thermally isolated environment and cool it to -40°C. Prior to testing with fissile materials, testing with surrogate fuel plates was performed to characterize the thermal properties of the system, ensure no condensation forms on the stack, and to test that the uranium plates would not warp, or fracture during the thermal cycling. Surrogate testing has characterized the cryogenic chamber and verified its viability to perform the benchmark experiments, allowing for the next steps in the LT-TEX experimental campaign. The results of the surrogate testing are discussed and compared to heat transfer calculations performed as part of the experiment design process, to establish confidence in the thermal behavior of the system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Disk transport rates from Ti isotopic signatures of refractory inclusions

Abstract The early solar system was a dynamic period during which the formation of early solids set into motion the process of planet building. Although both astrophysical observations and theoretical modeling demonstrate the presence of widespread transport of material, we lack concrete quantitative constraints on timings, distances, and mechanisms thereof. To trace these transport processes, one needs objects of known early formation times and these objects would need to be distributed throughout parent bodies with known accretion times and distances. Generally, these criteria are met by “regular” (i.e., non–fractionated and unidentified nuclear and excluding hibonite‐rich) Ca‐Al‐rich inclusions (CAIs) as these objects formed very early and close to the young Sun and contain distinctive nucleosynthetic isotope anomalies that permit provenance tracing. However, nucleosynthetic isotopic signatures of such refractory inclusions have so far primarily been analyzed in chondritic meteorites that formed within ~4 AU from the Sun. Here, we investigate Ti isotopic signatures of four refractory inclusions from the ungrouped carbonaceous chondrite WIS 91600 that was previously suggested to have formed beyond ~10 AU from the Sun. We show that these inclusions exhibit correlated excesses in 50 Ti and 46 Ti and lack large Ti isotopic anomalies that would otherwise be indicative of more enigmatic refractory materials with unknown formation ages. Instead, these isotope systematics suggest the inclusions to be genetically related to regular CAIs commonly found in other chondrites that have a broadly known formation region and age. Collectively, this implies that a common population of CAIs was distributed over the inner ~10 AU within ~3.5 Myr, yielding an average (minimum) speed for the transport of millimeter‐scale material in the early solar system of ~1 cm s −1 .

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

Photonuclear Physics in SCALE

Radiation shielding modeling involves a wide range of applications, such as power reactors, critical experiments, portable sources, linear accelerators, and fusion systems. Particle energy ranges vary from thermal energies to fast energy spectra. MAVRIC, as the radiation shielding sequence of the SCALE code system [1] developed and maintained by Oak Ridge National Laboratory, is capable of modeling both neutron and photon interactions in matter and is widely used by the US Nuclear Regulatory Commission, Department of Energy, the National Nuclear Security Administration, and radiation shielding practitioners all around the world. Although most applications don’t involve highly energetic photons that can cause photonuclear reactions and create coupled chains of high energy, ionized electrons and bremsstrahlung radiation within a material medium, when present, these reactions can have a significant impact upon the downstream particle fluxes from incident photons above 20 MeV. For example, linear accelerators used for medical isotope production are examples of applications in which high-energetic photons are generated and used for driving or controlling other components of a complex facility. Radiation shielding analyses of such a facility would require accounting for any and all significant sources of radiation. Unfortunately, MAVRIC included no data and methods to model either photonuclear reactions or the bremsstrahlung photons until this work, which enabled accounting for both. Photonuclear reaction cross sections were imported from the Evaluated Nuclear Data File libraries, and corresponding methods were implemented in the SCALE code system; see Section 2 for how the nuclear data are imported and used. Bremsstrahlung photons were modeled by assuming thick-target approximation whereby all generated electrons and positrons due to photon interactions are assumed to be absorbed within the same material in which they were born, yielding a quick particle transport simulation method by using fast lookup tables for generating subsequent gammas; see Section 4 for details. A suite of verification models was used to test implemented methods and associated data. Tests yielded satisfactory results for the new features. Enhancements will be released as part of the next major SCALE release (v7.0), which is anticipated to be deployed in 2025, with beta releases available in the interim.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Systems Theory Principles and Complex Systems Engineering Concepts for Protection and Resilience in Critical Infrastructure: Lessons from the Nuclear Sector

Part of the Presidential Policy Directive 21 (PPD-21) (PPD 2013) mandate includes evaluating safety, security, and safeguards (or nonproliferation) mechanisms traditionally implemented within the nuclear reactors, materials, and waste sector of critical infrastructure—including a complex, dynamic set of risks and threats within an all-hazards approach. In response, research out of Sandia National Laboratories (Sandia) explores the ability of systems theory principles (hierarchy and emergence) and complex systems engineering concepts (multidomain interdependence) to better understand and address these risks and threats. Herein, this Sandia research explores the safety, safeguards, and security risks of three different nuclear sector-related activities—spent nuclear fuel transportation, small modular reactors, and portable nuclear power reactors—to investigate the complex and dynamic risk related to the PPD-21-mandated all-hazards approach. This research showed that a systems-theoretic approach can better identify inter-dependencies, conflicts, gaps, and leverage points across traditional safety, security, and safeguards hazard mitigation strategies in the nuclear reactors, materials, and waste sector. Resulting from this, mitigation strategies from applying systems theoretic principles and complex systems engineering concepts can be (1) designed to better capture interdependencies, (2) implemented to better align with real-world operational uncertainties, and (3) evaluated as a systems-level whole to better identify, characterize, and manage PPD-21's all hazards strategies.

42 ENGINEERING↗