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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 793 records · Page 44
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Pool scrubbing under jet injection regime: An enhancement of the SPARC90 code

The SPARC90 code was developed to calculate the aerosol pool trapping during vent discharge processes, at low gas velocities. However, there are accident sequences, like SGTR core meltdown sequences, at which particle laden gases reach the aqueous ponds at very high velocities and new particle removal mechanisms become effective right at the inlet. As a result of the shearing off of roll wave water crests, water droplets are entrained in the gas core and sweep out aerosol particles, mainly by inertial impaction and interception. This paper summarizes the update of the SPARC90 code based on state-of-the-art equations for jet hydrodynamics and aerosol removal. Equations for variables like droplets population, size and velocity have been implemented. Based on the anticipated conditions in case of an SGTR severe accident sequence, comparisons of estimates from this new version (SPAR90-Jet) and the original one are set in terms of decontamination factor. Even though further work is still ahead, this work highlights how substantial particle retention at the pool inlet can reach under jet regime and how different aerosol removal mechanisms are with respect to the globule injection regime. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Application of systems engineering techniques to component design - Capturing functionality and linking part 'critical to quality' features to requirements

A systems engineering approach - focusing upon functionality - has predominantly been applied in industry to the design of complex systems with many functional interactions, inputs and outputs, eg the design of a decay heat removal system. This paper presents how systems engineering techniques can be applied to component design, i.e. treating the component as a system in its own right, and using functionality as the 'bridge' between the customer requirements and accepted performance. A pressure relief valve is used as an example to present the techniques of: Functional Modelling to establish the functional requirements and Functional Failure Modes and Effects Analysis to establish any emergent functionality to reduce the risk of adverse behaviour. A key aspect of component design is capturing the design intent and establishing the 'Critical to Quality 'features that can critically affect quality and performance. This paper details 'Quality Function Deployment' being applied to a component to capture such features and to establish a clear link to the overarching performance requirements. This approach is particularly useful in ensuring continuity of design understanding throughout the component life cycle, assessing the effects of any proposed changes to the design and the effects of changes in system or customer requirements, or for using the design in a different application. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Uncertainty methodology for the strongly coupled physical phenomena associated with annular flow

Best-Estimate plus Uncertainty (BEPU) methods are slowly supplanting the use of deterministic analysis methods for thermal-hydraulic analyses. As the uncertainty methodologies evolve it is expected that, where both experimental techniques allow and data are available, there will be a shift to quantifying the uncertainty in increasingly more fundamental parameters. For example, for annular flow in a three-field analysis environment (vapor, liquid film, droplet), the driving parameters would be: a) film interfacial shear stress, b) droplet drag, c) droplet entrainment rate and d) droplet deposition rate. An improved annular flow modeling package was recently developed and implemented in an in-house version of the COBRA-TF best-estimate subchannel analysis tool (Lane, 2009). Significant improvement was observed in the code-to-data predictions of several steam-water annular flow tests following the implementation of this modeling package; however, to apply this model set in formal BEPU analysis requires uncertainty distributions to be determined. The unique aspect of annular flow, and the topic of the present work, is the strong coupling between the interfacial drag, entrainment and deposition phenomena. Ideally the uncertainty in each phenomenon would be isolated; however, the situation is further complicated by an inability to experimentally isolate and measure the individual rate processes (particularly entrainment rate), which results in available experimental data that are inherently integral in nature. This paper presents a methodology for isolating the individual physical phenomena of interest, to the extent that the currently available experimental data allow, and developing the corresponding uncertainty distributions for annular flow. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

DOMINO: A fast 3D cartesian discrete ordinates solver for reference PWR simulations and SPN validation

As part of its activity, EDF R and D is developing a new nuclear core simulation code named COCAGNE. This code relies on DIABOLO, a Simplified PN (SPN) method to compute the neutron flux inside the core for eigenvalue calculations. In order to assess the accuracy of SPN calculations, we have developed DOMINO, a new 3D Cartesian SN solver. The parallel implementation of DOMINO is very efficient and allows to complete an eigenvalue calculation involving around 300 x 10{sup 9} degrees of freedom within a few hours on a single shared-memory supercomputing node. This computation corresponds to a 26-group S{sub 8} 3D PWR core model used to assess the SPN accuracy. At the pin level, the maximal error for the SP{sub 5} DIABOLO fission production rate is lower than 0.2% compared to the S{sub 8} DOMINO reference for this 3D PWR core model. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Cyber-Informed Engineering Guidance—Implementing CIE in Early Systems Engineering Lifecycle Stages

Traditionally, cybersecurity is not considered in the design process. Design engineers typically focus on building safety and reliability into their products and applications. Security against malicious cyber incidents is often an afterthought, resulting in deployment of security solutions during installation or operation. Unfortunately, waiting to consider cybersecurity until later in the systems engineering lifecycle often results in less effective and more expense security. Idaho National Laboratory (INL) developed the concept of Cyber-Informed Engineering (CIE) in 2015 to provide a framework that enables cybersecurity to be built into systems beginning at the conceptual design stage. In addition to ongoing research by INL, the U.S. Department of Energy (DOE) Office of Cybersecurity, Energy Security, and Emergency Response has recently developed a National CIE Strategy document for incorporating CIE into the design and operation of infrastructure systems reliant on digital monitoring or controls. This paper provides a brief review of this National CIE Strategy as well as a roadmap to historical, current, and future CIE research by INL through the U.S. DOE Office of Nuclear Energy (NE) Cybersecurity Crosscutting Technology Development Program. A near-term focus of the DOE-NE’s research and development is to extend the foundational CIE work into detailed guidance for implementation during initial systems engineering stages in nuclear digital instrumentation and control projects and to demonstrate use of the guidance in an integrated energy systems project.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Using Whisper-1.1 to guide improvements to nuclear data evaluations

Along with the latest release of the general purpose Monte Carlo radiation transport code, MCNP6.2, came the first officially released version of Whisper, version 1.1, a statistical analysis code used for nuclear criticality safety (NCS) validation. The Whisper software package is primarily used to calculate the upper subcritical limit (USL) of a particular application case along with components of the margin of subcriticality (MOS). In the calculation of the MOS due to nuclear data uncertainties, Whisper uses the generalized linear-least-squares (GLLS) approach to obtain adjusted nuclear data uncertainties which are then propagated to estimate the nuclear data uncertainties in the k{sub eff} of the application. Due to the fact that Whisper contains the simulated and measured k{sub eff} values for over 1100 benchmarks, nuclear data sensitivity profiles for each benchmark along with the nuclear data covariance matrices, all of which are used in the GLLS method, it is possible to easily utilize these data available to help guide improvements to nuclear data evaluations. In the present work, the Whisper data and the GLLS method are used together to suggest an adjusted nuclear data cross section evaluation specifically for {sup 233}U such that the agreement between the calculated and experimental k{sub eff} measurements of the relevant benchmarks are improved. With this adjusted nuclear data available, feedback to the theoretical and experimental nuclear data community is provided. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Effects of liquid subcooling and initial rod temperature on the minimum film boiling temperature

This paper presents an experimental study on the minimum film boiling temperature (T{sub min}) and its sensitivity to liquid subcooling and initial rod temperature. T{sub min} is measured through a quenching experiment using a vertical stainless steel rod that is quenched in a distilled water bath. The cylindrical stainless steel rod has a flat circular end with diameter of 9.5 mm similar to that of a fuel rod and length of 25 cm. Physically, T{sub min} is used to define the boundary between film boiling and transition boiling, beyond which temperature liquid would not be able to come into physical contact with the solid surface. Results show that the quenching curves are affected by varying the liquid subcooling but are not significantly altered by the initial rod temperature. Additionally, the experiments include visualization of the boiling regimes. The visualization gives insight into the film flow behavior that is found to depend strongly on the liquid subcooling. It also provides an understanding of the phenomenon for quench front axial propagation (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

IEC 62859: A new international standard on the coordination between safety and cybersecurity for nuclear I and C systems - 160

Cybersecurity requirements and controls for digital I and C systems are increasingly added to well-established safety-oriented provisions. Interactions and potential side-effects are possible and must be mastered, taking into account the nuclear I and C system specificities as well as their normative and regulatory contexts. The new international standard IEC 62859 'Nuclear power plants - Instrumentation and control systems - Requirements for coordinating safety and cyber security' aims to optimize the integration of cybersecurity provisions in nuclear I and C architecture and systems, to prevent conflicts between safety and cybersecurity provisions, and to aid the identification and the leveraging of the potential synergies between safety and cybersecurity. The paper presents the structure and the key aspects of this standard. IEC 62859 is structured along three main sections: one dealing with the overall I and C architecture, a second one focusing on the system level, a third one dealing with organizational and procedural issues. The paper also discusses related work in the nuclear industry, and in other sectors. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Assessment of dynamic PRA techniques with industry average component performance data - 208

In the nuclear industry, risk monitors are intended to provide a point-in-time estimate of the system risk given the current plant configuration. Current risk monitors are limited in that they do not properly take into account the deteriorating states of plant equipment, which are unit-specific. Current approaches to computing risk monitors use probabilistic risk assessment (PRA) techniques, but the assessment is typically a snapshot in time. Living PRA models attempt to address limitations of traditional PRA models in a limited sense by including temporary changes in plant and system configurations. However, information on plant component condition is not considered. This often leaves risk monitors using living PRA models incapable of conducting evaluations with dynamic degradation scenarios evolving over time. There is a need to develop enabling approaches to solidify risk monitors to provide time- and condition-dependent risk by integrating traditional PRA models with condition monitoring and prognostic techniques. This paper presents an estimation of system risk evolution over time by integrating plant risk monitoring data with dynamic PRA methods that incorporate aging and degradation. Several online, nondestructive approaches have been developed for diagnosing plant component conditions in the nuclear industry, i.e., using a condition indication index, vibration analysis, current signatures, and operational history . In this work, the component performance measures at U.S. commercial nuclear power plants are incorporated within the dynamic PRA methodologies to provide better estimates of the probability of failures. Aging and degradation are modeled within the Level 1 PRA framework, are applied to several failure modes of pumps, and can be extended to a range of components, namely, valves, generators, batteries, and pipes. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Instrumentation, control, and human system interface contributions to historical severe accidents - 301

In 2015 EPRI completed a study of historical severe accidents. This study was intended to identify the lessons that these events had for instrumentation and control (I and C), and human systems interfaces (HSI) for nuclear power plants. The project studied 19 severe accidents that were identified by a parallel study that was underway for the International Atomic Energy Agency. I and C or HSI issues made significant contributions to 18 of the 19 events studied. This paper will summarize the I and C and HSI deficiencies that contributed to the propagation of the accidents, discuss the likely causes of these deficiencies and describe the role of I and C and HSI in finally terminating the events. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Development of MELCOR 2.1 Guidance Report for LPF Applications

A complex nuclear facility includes several pathways that can transport accidentally released materials throughout the facility and potentially to the exterior environment. To model and estimate material transport and possible release as a source term from a U.S. Department of Energy (DOE) nuclear facility requires proficiency in applying simulation tools, such as the MELCOR code. DOE has designated MELCOR version 1.8.5, an obsolete version, as a DOE Tool Box code in its Central Registry. A leak path factor (LPF) guidance report written in 2004 did not include experimental validation data. Using an obsolete version of MELCOR requires additional verification and validations, which may be cost prohibitive. Without any developer support and lack of experimental data validation, it is difficult to justify to regulators that the calculated source term with the obsolete version of MELCOR from the facility is accurate and defensible. The goal of this work is to replace the obsolete guidance report with MELCOR 2.1 with experimental validations. In this paper, we will present our approach to develop a MELCOR 2.1 guidance report for LPF applications. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Development of CFD Flow Analysis for Nuclear Facility Ventilation Systems

Understanding and modeling of particulate transport has importance related to safety in ventilated environments that contain hot cells, gloveboxes, and fume hoods. In particular, good modeling of the airflow in these spaces is required to understand areas of stagnation where particles drop out of the free stream and accumulate. This accumulation of particulate can cause either contamination and/or higher than acceptable local dose levels. It is known that multiple sites across the DOE complex take credit for hot cells, gloveboxes, and/or hoods in their safety basis for providing a defense-in-depth benefit for both onsite and offsite releases. By providing confinement of radioactive materials, such features serve to reduce direct doses to facility workers and mitigate the consequences to the environment due to an uncontrolled release. Each of these features has access points that interface with the personnel space. Understanding how air flow behaves at these access points is of great interest to those performing hazard analysis. Computational Fluid Dynamics (CFD) can be used to analyze and model the flows in hot cells and glove boxes as a way to confirm operation. While CFD capability continues to advance, there are still important modeling assumptions that are left to the analyst's discretion. These are most notably the choice of the turbulence models, mesh structure, and wall boundary condition assumptions used in the model. In this work, we apply CFD in modeling airflow through an occupiable work area of the Alpha Gamma Hot Cell Facility (AGHCF) at Argonne National Laboratory (ANL). This modeling is then compared to field test data to both inform and verify the modeling assumptions. Based on the results of field tests, modeling assumptions and boundary conditions are refined and the process is repeated until the results are found to be reliable with a high level of confidence.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Development of the MCNP-ORIGEN activation automation tool

This paper introduces the MCNP-ORIGEN activation automation tool for streamlining the calculation of experiment source terms. This tool couples the Monte-Carlo radiation-transport solver, MCNP to the depletion tool, ORIGEN-S. To showcase its current capabilities, this paper presents an activation analysis exercise, which is conducted with Serpent, and the results are com- pared. The experiment consists in a 90%-enriched uranium sphere surrounded by light water. A non-fissile cylinder is placed in the water, representing an irradiation experiment. We conducted simulations for 2 different cylinder materials: iron and cobalt at two different temperatures: 300 K and 900 K. The exercise consists of an irradiation of 50 days at a constant power of 5 MW, followed by a decay of 50 days. These studies highlight the discrepancies between how the different tools handle nuclear data. Finally, the results underscore multiple areas of possible improvement.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Resumption of SFR Overpower Testing and Post-Transient Examination of the THOR-C-2 Irradiation Test

Transient overpower testing for fuel safety research and development has resumed at Idaho National Laboratory. The THOR-C-2 commissioning test has been performed and post-transient examination has been performed revealing the intended failure near the top of the fuel zone. This paves the way for future testing on previously irradiated fuels that will support fuel development and qualification.

advanced reactors↗
DOE OSTI2024

Modernizing the Legacy Fission Wire Measurement System for the Advanced Test Reactor-Critical Facility

Operational lifetime extensions of existing research reactors have emphasized the need for refurbishment, replacements, and upgrades to supporting equipment and instrumentation. The Advanced Test Reactor (ATR) at Idaho National Laboratory (INL), which entered service in 1967, has recently completed the sixth core internals change-out and has scheduled operations until at least 2040. Reactor maintenance and operational risk management is critically important in the research reactor community, however supporting measurement systems sometimes get overlooked when maintenance is planned. The Fission Wire Measurement System (FWMS) is a custom measurement system designed in the 1960s to measure the beta-particle activity of irradiated uranium-aluminum fission wires. This measurement is conducted to determine the fission rate profile of the Advanced Reactor Test Critical (ATR-C) facility. The ATR-C is an open-pool, low-power test reactor that was purpose driven to resemble ATR and is used to qualify experiment configurations and verify core models prior to full-power experiment irradiations in ATR. A power distribution measurement in ATR-C uses uranium-aluminum wires that are distributed throughout the ATR-C core to validate simulation and modeling results. These measurements require 340 to 1500 wires to be irradiated and measured within a 12-hour window. The activity of the wires is measured in the required time with the FWMS, which was put into service in 1965 at the Radiation Measurements Laboratory (RML). The system consists of 4 measurement channels and one reference channel, each with a 2-pi proportional gas flow detector and the measurement channels each have an automated sample changer. This legacy system is crucial to the continued operations of ATR and has undergone some minor hardware upgrades since 1965, however the system presently relies on custom control boards, custom gas ion chambers, analog amplifiers/discriminators, and a user interface (UI) for the system written in outdated code. Much of the equipment and software is custom with no commercial replacements or support and limited documentation. The existing control software requires an operating system that is no longer supported, creating more vulnerabilities to continued operations. A project is underway with a third-party vendor to design, build, and document a new control and data acquisition system (CDAS) for the FWMS. The new upgrade will replace the control system, computer, UI, sample changer motors, and main power supply while maintaining the interface with existing detector hardware. The upgraded system will be operated in parallel with the current hardware and software to conduct validation testing. This equipment upgrade demonstrates the commitment at ATR to ensuring successful operations and potential future research reactors at INL.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗
DOE OSTI2021

Mechanistic Source Term Considerations for Advanced Non-LWRs (Revision 1)

This report is a functional review of the radionuclide containment strategies of fluoride-salt-cooled high temperature reactor (FHR), molten salt reactor (MSR) and high temperature gas reactor (HTGR) systems. This analysis serves as a starting point for further, more in-depth analyses geared towards identifying phenomenological gaps that still exist, hindering the creation of a mechanistic source term for these reactor types. As background information to this review, an overview of how a mechanistic source term is created and used for consequence assessment necessary for licensing is provided. How a mechanistic source term is used within the Licensing Modernization Project (LMP) is also provided. Lastly, the characteristics of non-LWR mechanistic source terms are examined. This report does not assess the viability of any software system for use with advanced reactor designs, but instead covers system function requirements. Future work within the Nuclear Energy Advanced Modeling and Simulations (NEAMS) program will address such gaps. This document is an update of SAND 2020-6730. An additional chapter is included as well as edits to original content.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗