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At least 307 records · Page 17

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DOE OSTI2020

ProZES: the methodology and software tool for assessment of assigned share of radiation in probability of cancer occurrence

ProZES is a software tool for estimating the probability that a given cancer was caused by preceding exposure to ionising radiation. ProZES calculates this probability, the assigned share, for solid cancers and hematopoietic malignant diseases, in cases of exposures to low-LET radiation, and for lung cancer in cases of exposure to radon. User-specified inputs include birth year, sex, type of diagnosed cancer, age at diagnosis, radiation exposure history and characteristics, and smoking behaviour for lung cancer. Cancer risk models are an essential part of ProZES. Linking disease and exposure to radiation involves several methodological aspects, and assessment of uncertainties received particular attention. ProZES systematically uses the principle of multi-model inference. Models of radiation risk were either newly developed or critically re-evaluated for ProZES, including dedicated models for frequent types of cancer and, for less common diseases, models for groups of functionally similar cancer sites. The low-LET models originate mostly from the study of atomic bomb survivors in Hiroshima and Nagasaki. Risks predicted by these models are adjusted to be applicable to the population of Germany and to different time periods. Adjustment factors for low dose rates and for a reduced risk during the minimum latency time between exposure and cancer are also applied. The development of the methodology and software was initiated and supported by the German Federal Ministry for the Environment, Nature Conservation and Nuclear Safety (BMU) taking up advice by the German Commission on Radiological Protection (SSK, Strahlenschutzkommission). These provide the scientific basis to support decision making on compensation claims regarding malignancies following occupational exposure to radiation in Germany.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2020

Source Term Reduction for Advanced and Small Modular Boiling Water Reactors

The United States Department of Energy (US DOE) is currently supporting the development of various advanced and small modular reactor (SMR) designs. Several of these designs have commenced license application with the US Nuclear Regulatory Commission (US NRC). These reactors have improved safety features that may significantly reduce radiological source terms in the event of design and beyond-design basis accidents. Specifically, some reactors feature a smaller containment volume relative to the available fission product depositional surface area, which supports increased fission product retention in the containment vessel. Pressurized water reactors (PWR) and boiling water reactors (BWR) with this feature include the integral pressurized water reactor (iPWR) and the BWRX-300 design by General Electric. A prior research program supported by the US DOE quantified the source term reduction associated with light water iPWRs and developed iPWR-specific theoretical models for fission product deposition rates. This program included a sequence of research projects that started with a feasibility study, development of theoretical models that predict higher deposition rates, and finally, development of empirical data for verification and validation of the theoretical models. The current project, which is a feasibility study, is the first step in a similar program to quantify the source term reduction associated with small and advanced light water BWRs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Status of EM Pump Modeling Capability for VTR

SAS4A/SASSYS-1 is a simulation tool used to perform deterministic analyses of anticipated events as well as design basis and beyond design basis accidents for advanced liquid-metal-cooled nuclear reactors. Most recently, SAS4A/SASSYS-1 has been selected as the safety analysis software for the Versatile Test Reactor (VTR), a new materials testing Sodium Fast Reactor (SFR) to be built by the Department of Energy. In order to support analysis of the VTR, which will utilize ElectroMagnetic (EM) pumps as the primary coolant pumps, the development of new EM pump modeling capabilities has been initiated for SAS4/SASSYS-1. The new physics-based EM pump model captures systems-level transient performance based on equivalent circuit theory and is highly flexible to model a wide range of electromagnetic pumps of the three-phase linear induction type subject to changes in power supply and plant conditions. Within the equivalent circuit framework, two sub-models are developed to provide options for analysts both with and without a detailed geometric characterization of their pump such that minimal information is required to achieve a realistic working model, but specific details of a particular pump design may be captured if sufficient information is provided. In addition, heat generated from pump operation and its transport through the plant system can be modeled to better capture the behavior of certain transients. This paper will outline the model and derive the governing equations of the new EM pump model. An initial comparison between the simple EM pump model and experimental data is made. This comparison shows that the equivalent circuit model can capture pump performance, with greater accuracy near the rated conditions. In order to demonstrate the transient behavior of the new model, a simple transient with an EM pump replacing the centrifugal pumps in the Advanced Burner Test Reactor is included.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2021

Status of EM Pump Modeling Capability for VTR

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2021

Subcriticality Measurements for a Changing Concentration Uranyl Nitrate Solution Tank by Californium Source-Driven Noise Analysis

Subcritical californium source-driven noise analysis (CSDNA) measurements were performed at the Oak Ridge Critical Experiments Facility in 1983 using a stainless-steel tank with an inside diameter of 76.2 cm and a height of 91.4 cm but only filled with a uranyl (93.16 wt.% 235 U) nitrate solution to a height of 76.2 cm for all measurements. The free acid content of the uranyl nitrate solution was less than 0.01 N to minimize hazards in handling. The total available highly enriched uranium for these measurements was 5,296 g. The concentration of the solution was varied from the highest density of 14.71 grams of uranium per liter (g U/L) initially in 16 steps, lowering the concentration to 0.3492 g U/L, and finally using a tank filled with only water. The tank had a Plexiglas lid to minimize evaporation. The Cf source was contained in a re-entrant Lexan tube which could be in the solution parallel to the axis of the cylindrical tank. The detectors were 3 He proportional counters in the solution in shrink-fit tubing to isolate the counters from the fissile solution and adjacent to the outside of the tank. In some cases, the detectors were scintillators adjacent to the outside of the tank. In other cases, the source was external to the tank. Some data presented in this report are from notes and are not in the logbook. The prompt neutron decay constant was obtained from fitting the date from the CSDNA measurements. The purpose of this report is to document the experimental information for the measurements performed so that later, researchers could perform the required uncertainty and calculational analyses and documentation to use these data for an International Nuclear Criticality Safety Benchmark Program (ICSBEP) or a Nuclear Energy Agency benchmark. The prompt neutron decay measurements could be the basis of as International Reactor Physic Benchmark Program. The data from these measurements are available from the Records Management Services Department of Oak Ridge National Laboratory, and the logbook is also available from the ICSBEP at Idaho National Laboratory. Preparation of the present report is part of a larger cooperative effort between Idaho National Laboratory (INL) and Oak Ridge National Laboratory (ORNL) to document more than 15 undocumented critical and subcritical experiments enumerated in ORNL/TM-2019/18 and performed by ORNL at ORCEF and other USDOE critical experiments facilities using more than 500 operational days of critical facility time.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

NASA NTRSDate not supplied

TPSAS-NF1676L-18539-DND

Technical Challenges (TCs) are aligned to Program goals and Project objectives: - Provide research focus to solving aviation safety problems - Provide consistent framework to focus, direct, plan, execute, manage, and communicate Center-distributed research - Form basis for “contract” between Program and Project and Center

Ersin Ancel↗

DOE OSTI2023

Accident event progression, gaps, and key performance indicators for steam generator tube rupture events in water-cooled SMRs: A review

According to historical records of reactor-related incidents, a steam generator tube rupture (SGTR) is one of the most common occurrences at operating pressurized water reactors (PWRs). Such design-basis accidents (DBAs) could lead to a direct path for radionuclides to be released to the atmosphere via the safety and relief valves, making assessments of radionuclide discharge from operating nuclear power plants (NPPs) into the environment crucial for ensuring safety. In such analyses, the primary focus has been on the extent of radioactive release, not the potential damage to the core. Moreover, significant and timely intervention by operators is needed during the initial stages of SGTRs in order to avert overfilling of the SGs, as well as to restrict the dissemination of radioactive materials. Determining the event sequence and phases that occur during an SGTR incident in an advanced passive (e.g., AP1000) water-cooled nuclear reactor is crucial for implementing effective passive safety systems (PSSs) for a specific water-cooled small modular reactor (SMR) design. Here this study provides a comprehensive review of the accident event progression, associated physical phenomena, knowledge gaps, and key performance indicators that must be evaluated and assessed in thermal-hydraulics models, based on relevant test data, respectively.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

NASA NTRS1983

Space safety and rescue 1979-1981: Worldwide disaster response, rescue and safety employing space-borne systems

Selected papers from the 1979, 1980, and 1981 IAA symposia on space safety and rescue and on worldwide disaster response, safety, and rescue employing spaceborne systems are presented. Available papers published elsewhere and those presented at the 1976, 1977, and 1978 symposia are presented in abstract form. Subjects discussed include man-made space debris, nuclear-waste disposal in space, space-station safety design, psychological training, the introduction of female crewmembers, analysis of the November 23, 1980 earthquake as a design basis for satellite emergency communication, disaster warning using the GOES satellite, and satellite communications for disaster relief operations. Three reviews of the application of space technology to emergency and disaster relief and prevention, given at other symposia in 1981, are presented in an appendix. No individual items are abstracted in this volume

Brown, J. W.↗

NASA NTRS2026

Space Suit Material Compatibility with Space Vehicle Coolant PMS-1.5R

During a survey of 82P on February 11, 2023, a shiny substance was detected on the Progress vehicle. The suspected material was PMS-1.5R, a low volatility silicone oil coolant used in the vehicle. Crew exposure to the coolant was a concern as the low volatility would increase residence time, the time a material will stay on a surface after exposure to vacuum. A review of space suit materials revealed several materials with compatibility concerns that could impact Extravehicular Activity (EVA) operations and safety. To understand the risk, space suit materials were exposed to a simulant and tested. This testing provided the basis for rationale generated to continue with EVA operations for US EVA 86 on April 2, 2023. This testing also served as the basis to assess impacts to ISS EMU hardware due to potential contamination brought inside ISS on Russian space suits during RS EVA 61 on October 25, 2023, following an ISS Russian Multipurpose Lab Module (MLM) radiator leak of the same fluid on October 9, 2023. This paper will discuss the testing performed and analyze test results impacts to EVA operations. Additional work was conducted to determine coolant evaporation rates in space vacuum and demonstrate that no significant unreacted coolant from the MLM leak would remain on ISS surfaces long enough to contaminate the EMU or affect the ISS ECLSS water separator.

Space Suit Materials↗

DOE OSTI2022

Simulation of Peach Bottom turbine trip transients TT1, TT2 and TT3 with SIMULATE-3K

The Peach Bottom 2 Turbine Trip transients tests TT1, TT2 and TT3 are simulated with SIMULATE-3K. This work provides the initial basis for evaluation of the capabilities and accuracy of coupled 3D neutronics/thermal hydraulics best estimate calculations for support of safety analysis. Selected results, obtained with SIMULATE-3K, are compared to available measurements and benchmark results. The comparisons are analyzed and the application of SIMULATE-3K for best estimate analysis approach within the safety analysis methodology is discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2025

Design of Critical Experiments Involving Graphite elements at the IPEN/MB-01 Plate-Type Core

A series of critical experiments were designed involving the use of characterized graphite elements in the new plate-type core of the IPEN/MB-01 reactor. Eleven configurations with different graphite quantities are considered, each producing distinct effects on neutron fluxes and offering opportunities to analyze varying sensitivities to k eff in the system. The analysis shows that the experiments can be performed with acceptably low k eff uncertainties considering an adequate characterization of the graphite elements to be used. The sensitivity results obtained using SCALE TSUNAMI calculations show that the experiments have a notable k eff sensitivity coefficient to graphite. In the configurations tested, the k eff sensitivity to graphite TSL is low. This experiment will be useful for criticality safety validation of applications using uranium fuel enriched to around 19.75 wt % 235 U, light water and graphite. It will also be useful to gain more insight into graphite material properties and their effect on k eff as a result of using the well-characterized graphite elements. The execution of the experiments is planned for the summer of 2025. The 11 designed configurations serve as a basis for the final design, and fewer configurations will be executed. Once executed, these critical experiments will be evaluated and submitted for publication in the International Criticality Safety Benchmark Evaluation Project Handbook, supporting the DOE/NRC Collaboration for Criticality Safety Support for Commercial-Scale HALEU for Fuel Cycles and Transportation project goal.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2023

Structural basis of promiscuous substrate transport by Organic Cation Transporter 1

Organic Cation Transporter 1 (OCT1) plays a crucial role in hepatic metabolism by mediating the uptake of a range of metabolites and drugs. Genetic variations can alter the efficacy and safety of compounds transported by OCT1, such as those used for cardiovascular, oncological, and psychological indications. Despite its importance in drug pharmacokinetics, the substrate selectivity and underlying structural mechanisms of OCT1 remain poorly understood. Here, we present cryo-EM structures of full-length human OCT1 in the inward-open conformation, both ligand-free and drug-bound, indicating the basis for its broad substrate recognition. Comparison of our structures with those of outward-open OCTs provides molecular insight into the alternating access mechanism of OCTs. We observe that hydrophobic gates stabilize the inward-facing conformation, whereas charge neutralization in the binding pocket facilitates the release of cationic substrates. These findings provide a framework for understanding the structural basis of the promiscuity of drug binding and substrate translocation in OCT1.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2024

ANS-8 Nuclear Criticality Safety Consensus Standards - Current Initiatives

The nuclear criticality safety (NCS) consensus standards are developed as using rigorous procedures of the Standards Board of the American Nuclear Society. These procedures have been accredited by the American National Standards Institute, Inc., as meeting the criteria for American National Standards. The Nuclear Criticality Safety Consensus Committee (NCSCC) that approved all 18 NCS consensus standards is balanced to ensure that competent, concerned, and varied interests have had an opportunity to participate. The ANS-8 subcommittee (ANS-8) consists of 17 NCS experts with many years of experience as end users of ANS standards who serve on standard working groups to develop and maintain standards. ANS-8 ensures that the technical content of the standards is adequate for NCS community use. Attempts are made to ensure that ANS-8 consists of NCS professionals with a diverse range of experience such that all standards are applicable to as many sites as possible. ANS-8 is a very active subcommittee, and some active projects in progress are discussed in this paper: basis statement development for all standards, development of a glossary for consistency of definitions across all ANS-8 standards, and Considering the Criticality Safety Support Group (CSSG) Recommendation 2016-04 to the ANS Standards Board for changes in several ANS-8 standards.

96 KNOWLEDGE MANAGEMENT AND PRESERVATION↗

DOE OSTI2024

Methods for Quantitative Thermal Analysis of Lithium Solid-State and Beyond Battery Safety

The use of differential scanning calorimetry (DSC) to measure the thermal behavior of individual components and electrolyte/electrode combinations is common. However, here we focus on DSC tests on an anode, cathode, and electrolyte (ACE) component combination over a temperature range that includes many of the phase transitions and key reactions (i.e., to 500 °C) that contribute to thermal runaway. This method can help quantify the complex reaction network in a full cell, thereby informing potential safety issues. Here, we used DSC heat flow data from a solid-state Li 0.43 CoO 2 +C+PVDF | LLZO | Li metal ACE sample and its components to quantify key factors affecting results. We focused on three areas: (1) ACE sample preparation and assembly in DSC pans, (2) DSC measurement parameters, and (3) heat flow analysis. Key points include the choice of component ratios (e.g., commercially relevant N:P capacity ratio), the importance of conductive carbon and binder, type of pan used, DSC ramp rate, and integration method used when dealing with broad and overlapping exothermic peaks. This work deepens the scientific basis and best practices for obtaining heat flow data from ACE samples for early-stage evaluation of solid-state and beyond battery safety.

25 ENERGY STORAGE↗

NASA NTRS2010

NASA Aerospace Flight Battery Program: Generic Safety, Handling and Qualification Guidelines for Lithium-Ion (Li-Ion) Batteries; Availability of Source Materials for Lithium-Ion (Li-Ion) Batteries; Maintaining Technical Communications Related to Aerospace Batteries (NASA Aerospace Battery Workshop)

This NASA Aerospace Flight Battery Systems Working Group was chartered within the NASA Engineering and Safety Center (NESC). The Battery Working Group was tasked to complete tasks and to propose proactive work to address battery related, agency-wide issues on an annual basis. In its first year of operation, this proactive program addressed various aspects of the validation and verification of aerospace battery systems for NASA missions. Studies were performed, issues were discussed and in many cases, test programs were executed to generate recommendations and guidelines to reduce risk associated with various aspects of implementing battery technology in the aerospace industry. This document contains Part 1 - Volume I: Generic Safety, Handling and Qualification Guidelines for Lithium-Ion (Li-Ion) Batteries, Availability of Source Materials for Lithium-Ion (Li-Ion) Batteries, and Maintaining Technical Communications Related to Aerospace Batteries (NASA Aerospace Battery Workshop).

Manzo, Michelle A.↗

NASA NTRS1996

Feasibility Study of a Rotorcraft Health and Usage Monitoring System (HUMS): Results of Operator's Evaluation

The objective was to evaluate the feasibility of a state-of-the-art health and usage monitoring system (HUMS) to provide monitoring of critical mechanical systems on the helicopter, including motors, drive train, engines, and life-limited components. The implementation of HUMS and cost integration with current maintenance procedures was assessed from the operator's viewpoint in order to achieve expected benefits from these systems, such as enhanced safety, reduced maintenance cost, and increased availability. An operational HUMS that was installed and operated under an independent flight trial program was used as a basis for this study. The HUMS equipment and software were commercially available. Based on the results of the feasibility study, the HUMS used in the flight trial program generally demonstrated a high level of reliability in monitoring the rotor system, engines, drive train, and life-limited components. The system acted as a sentinel to warn of impending failures. A worn tail rotor pitch bearing was detected by HUMS, which had the capability for self testing to diagnose system and sensor faults. Examples of potential payback to the operator with HUMS were identified, including reduced insurance cost through enhanced safety, lower operating costs derived from maintenance credits, increased aircraft availability, and improved operating efficiency. The interfacing of HUMS with current operational procedures was assessed to require only minimal revisions to the operator's maintenance manuals. Finally the success in realizing the potential benefits from HUMS technology was found to depend on the operator, helicopter manufacturer, regulator (FAA), and HUMS supplier working together.

Romero, Raylund↗

NASA NTRS1993

First Lunar Outpost support study

The First Lunar Outpost (FLO) is the first manned step in the accomplishment of the Space Exploration Initiative, the Vice President's directive to NASA on the 20th anniversary of the Apollo moon landing. FLO's broad objectives are the establishment of a permanent human presence on the moon, supporting the utilization of extraterrestrial resources in a long-term, sustained program. The primary objective is to emplace and validate the first elements of a man tended outpost on the lunar surface to provide the basis for: (1) establishing, maintaining and expanding human activities and influence across the surface; (2) establishing, maintaining and enhancing human safety and productivity; (3) accommodating space transportation operations to and from the surface; (4) accommodating production of scientific information; (5) exploiting in-situ resources. Secondary objectives are: (1) to conduct local, small scale science (including life science); (2) In-situ resource utilization (ISRU) demonstrations; (3) engineering and operations tests; (4) to characterize the local environment; and (5) to explore locally. The current work is part of ongoing research at the Sasakawa International Center for Space Architecture supporting NASA's First Lunar Outpost initiative. Research at SICSA supporting the First Lunar Outpost initiative has been funded through the Space Exploration Initiatives office at Johnson Space Center. The objectives of the current study are to further develop a module concept from an evaluation of volumetric and programmatic requirements, and pursue a high fidelity design of this concept, with the intention of providing a high fidelity design mockup to research planetary design issues and evaluate future design concepts.

Bartz, Christopher↗

NASA NTRS2008

High Performance COPVs for In-Space Storage of High Pressure Cryogenic Fuels

Polymeric composite overwrapped pressure vessels (COPVs) provide an attractive material system to support developing commercial launch business and alternate fuel ventures. However to be able to design with these materials, the mechanical behavior of the materials must be understood with regards to processing, performance, damage tolerance, and environment. For the storage of cryogenic propellants, it is important to evaluate the materials performance and impact damage resistance at cryogenic temperatures in order to minimize weight and to ensure safety and reliability. As part of this study, material tests of candidate fiber and resin systems were used as the basis for the selection of the material combinations for evaluation in a COPV at cryogenic conditions. This comprehensive approach has also been expanded to address issues with impact damage tolerance and material degradation due to environmental factors. KEY WORDS: Cryogenic testing, evaluation and applications for pressure vessels, COPVs, tanks, or storage vessels.

Schneider, Judy↗