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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 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

System Study: High-Pressure Core Spray 1998–2013

This report presents an unreliability evaluation of the high-pressure core spray (HPCS) at eight U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2013 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10-year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the HPCS results.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

System Study: High-Pressure Core Spray 1998–2012

This report presents an unreliability evaluation of the high-pressure core spray (HPCS) at 8 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2012 for selected components were obtained from the Equipment Performance and Information Exchange (EPIX). The unreliability results are trended for the most recent 10 year period while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the HPCS results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

System Study: High-Pressure Core Spray 1998-2014

This report presents an unreliability evaluation of the high-pressure core spray (HPCS) at eight U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the HPCS results.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

System Study: High-Pressure Core Spray 1998-2020

This report presents an unreliability evaluation of the high-pressure core spray (HPCS) at eight U.S. commercial boiling water reactors. Demand, run hour, and failure data from 1998–2020 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Industry Reporting and Information System (IRIS), formerly the INPO Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the HPCS results.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2023

System Study: High-Pressure Core Spray 1998-2022

This report presents an unreliability evaluation of the high-pressure core spray (HPCS) at eight U.S. commercial operating boiling water reactors. New Standardized Plant Analysis Risk (SPAR) models with the most recent SPAR parameter update results were used in this report. Demand, run hour, and failure data from 1998–2022 for selected components were obtained from the Institute of Nuclear Power Operations Industry Reporting and Information System. The unreliability results are trended for the most recent 10-year period while yearly estimates for system unreliability are provided for the entire active period. Statistically significant increasing trends were identified in both the HPCS system start-only unreliability and 8-hour mission unreliability.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2025

System Study: High-Pressure Core Spray 1998-2024

This report presents an unreliability evaluation of the high-pressure core spray (HPCS) at eight U.S. commercial operating boiling water reactors. New Standardized Plant Analysis Risk (SPAR) models with the most recent SPAR parameter update results were used in this report. Demand, run hour, and failure data from 1998 to 2024 for selected components were obtained from the Institute of Nuclear Power Operations Industry Reporting and Information System. The unreliability results are trended for the most recent 10-year period and yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the HPCS results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

System Study: High-Pressure Core Spray 1998–2018

This report presents an unreliability evaluation of the high-pressure core spray (HPCS) at eight U.S. commercial boiling water reactors. Demand, run hours, and failure data from 1998 through 2018 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Industry Reporting and Information System (IRIS), formerly the INPO Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the HPCS results.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2012

TRACE and TRAC-BF1 benchmark against Leibstadt plant data during the event inadvertent opening of relief valves

In framework of introducing TRACE code to transient analyses system codes for Leibstadt Power Plant (KKL), a conversion process of existing TRAC-BF1 model to TRACE has been started within KKL. In the first step, TRACE thermal-hydraulic model for KKL has been developed based on existing TRAC-BF1 model. In order to assess the code models a simulation of plant transient event is required. In this matter simulations of inadvertent opening of 8 relief valves event have been performed. The event occurs at KKL during normal operation, and it started when 8 relief valves open resulting in depressurization of the Reactor Pressure Vessel (RPV). The reactor was shutdown safely by SCRAM at low level. The high pressure core spray (HPCS) and the reactor core isolation cooling (RCIC) have been started manually in order to compensate the level drop. The remaining water in the feedwater (FW) lines flashes due to saturation conditions originated from RPV depressurization and refills the reactor downcomer. The plant boundary conditions have been used in the simulations and the FW flow rate has been adjusted for better prediction. The simulations reproduce the plant data with good agreement. It can be concluded that the TRAC-BF1 existing model has been used successfully to develop the TRACE model and the results of the calculations have shown good agreement with plant recorded data. Beside the modeling assessment, the TRACE and TRAC-BF1 capabilities to reproduce plant physical behavior during the transient have shown satisfactory results. The first step of developing KKL model for TRACE has been successfully achieved and this model is further developed in order to simulate more complex plant behavior such as Turbine Trip. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

TRACE Code Validation of BWR SVEA Fuel Assembly Spray Cooling

Emergency spray cooling injection is a safety feature implemented in many boiling water reactor designs to reflood a reactor during an accident. Significant experimental work has qualified the efficacy of spray cooling, and ongoing computational modeling efforts strive to more accurately portray the phenomena involved. This work examines the physical phenomena pertaining to emergency spray cooling injection over SVEA-type fuel assemblies. BWR emergency core cooling system (ECCS) designs commonly incorporate a Low-Pressure Core Spray (LPCS) system that introduces coolant through spray nozzles directed to the top of the core, and are located in the upper plenum region. This design is effective for LOCA response, particularly during the initial refill and reflood stages of accident response where the goal is limiting the peak cladding temperature rise in the core. The U.S. NRC thermal-hydraulics code TRACE version 5.0 Patch 4 has been chosen to simulate the separate-effect tests performed by ASEA-ATOM. The computational model was evaluated by performing forward uncertainty quantification using Dakota as the analysis tool and code driver on 31 identified parameters from the input model and TRACE physical constitutive models. This paper discusses the sensitivity of the investigated parameters with respect to spray cooling simulation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Enhanced Component Performance Study: Emergency Diesel Generators 1998–2018

This report presents an enhanced performance evaluation of emergency diesel generators (EDGs) at U.S. commercial nuclear power plants. This report evaluates component performance over time using (1) Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES) data from 1998 through 2018 and (2) maintenance unavailability (UA) performance data from Mitigating Systems Performance Index (MSPI) Basis Document data from 2002 through 2018. The objective is to show estimates of current failure probabilities and rates related to EDGs, trend these data on an annual basis, determine if the current data are consistent with the probability distributions currently recommended for use in NRC probabilistic risk assessments, show how the reliability data differ for different EDG manufacturers and for EDGs with different ratings; and summarize the subcomponents, causes, detection methods, and recovery associated with each EDG failure mode. Engineering analyses were performed with respect to time period and failure mode without regard to the actual number of EDGs at each plant. The factors analyzed are: sub-component, failure cause, detection method, recovery, manufacturer, and EDG rating. A statistically significant increasing trend was identified in the frequency of FTLR demands for emergency power system (EPS) and high pressure core spray (HPCS) EDGs and a statistically significant decreasing trend was identified in the frequency of run > 1H hours for EPS and HPCS EDGs.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2022

Enhanced Component Performance Study: Emergency Diesel Generators 1998-2020

This report presents an enhanced performance evaluation of emergency power system (EPS) and high-pressure core spray (HPCS) emergency diesel generators (EDGs) at U.S. commercial nuclear power plants. This report evaluates component performance over time using (1) Institute of Nuclear Power Operations (INPO) Industry Reporting and Information System (IRIS) data from 1998 through 2020 and (2) maintenance unavailability performance data from Mitigating Systems Performance Index (MSPI) Basis Document data from 2002 through 2020. The objective is to show estimates of current failure probabilities and rates related to EDGs, trend these data on an annual basis, determine if the current data are consistent with the probability distributions currently recommended for use in NRC probabilistic risk assessments, show how the reliability data differ for different EDG manufacturers and for EDGs with different ratings; and summarize the subcomponents, causes, detection methods, and recovery associated with each EDG failure mode. The EDG failure modes considered are fail to start (FTS), fail to load and run (FTLR), and fail to run after one hour of operation (FTR>1H). Engineering analyses were performed with respect to time period and failure mode without regard to the actual number of EDGs at each plant. The factors analyzed are: subcomponent, failure cause, detection method, recovery, manufacturer, and EDG rating. The following increasing trends were identified for EDGs for the most recent 10-year period: • HPCS EDG FTS failure probability • EPS and HPCS EDG frequency of start demands (demands per reactor year) • EPS and HPCS EDG frequency of FTLR demands. The following decreasing trends were identified for EDGs for the most recent 10-year period: • EPS EDG FTS failure probability • EPS EDG FTLR failure probability • EPS EDG unavailability • EPS and HPCS EDG frequency of FTLR events (failures per reactor year).

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2023

Enhanced Component Performance Study: Emergency Diesel Generators 1998–2022

This report presents an enhanced performance evaluation of the emergency power system (EPS) and high-pressure core spray (HPCS) emergency diesel generators (EDGs) at U.S. commercial nuclear power plants. This report evaluates component performance over time using (1) Institute of Nuclear Power Operations (INPO) Industry Reporting and Information System (IRIS) data from 1998 through 2022 and (2) maintenance unavailability performance data from Mitigating Systems Performance Index (MSPI) Basis Document data from 2002 through 2022. The objective is to show estimates of current failure probabilities and rates related to EDGs, trend these data on an annual basis, determine if the current data are consistent with the probability distributions currently recommended for use in Nuclear Regulatory Commission (NRC) probabilistic risk assessments, show how the reliability data differ for different EDG manufacturers and for EDGs with different ratings; and summarize the subcomponents, causes, detection methods, and recovery associated with each EDG failure mode. The EDG failure modes considered are fail to start (FTS), fail to load and run (FTLR), and fail to run after one hour of operation (FTR>1H). Engineering analyses were performed with respect to time-period and failure mode without regard to the actual number of EDGs at each plant. The factors analyzed include subcomponent, failure cause, detection method, recovery, manufacturer, and EDG rating. The following increasing trends were identified for EDGs for the most recent 10-year period: (1) HPCS EDG FTS failure probability, (2) EPS and HPCS EDG frequency of start demands (demands per reactor year), and (3) EPS and HPCS EDG frequency of FTLR demands. The following decreasing trends were identified for EDGs for the most recent 10-year period: (1) EPS EDG FTS failure probability, (2) EPS EDG FTLR failure probability, (3) EPS EDG unavailability, and (4) EPS and HPCS EDG frequency of FTLR events (failures per reactor year).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Enhanced Component Performance Study: Emergency Diesel Generators 1998-2024

This report presents an enhanced performance evaluation of the emergency power system (EPS) and high-pressure core spray (HPCS) emergency diesel generators (EDGs) at U.S. commercial nuclear power plants. This report evaluates component performance over time using (1) Institute of Nuclear Power Operations (INPO) Industry Reporting and Information System (IRIS) data from 1998 through 2024 and (2) maintenance unavailability performance data from Mitigating Systems Performance Index (MSPI) Basis Document data from 2002 through 2024. The objective is to show estimates of current failure probabilities and rates related to EDGs, trend these data on an annual basis, determine if the current data are consistent with the probability distributions currently recommended for use in Nuclear Regulatory Commission (NRC) probabilistic risk assessments, show how the reliability data differ for different EDG manufacturers and for EDGs with different ratings; and summarize the subcomponents, causes, detection methods, and recovery associated with each EDG failure mode. The EDG failure modes considered are fail to start (FTS), fail to load and run (FTLR), and fail to run after one hour of operation (FTR>1H). Engineering analyses were performed with respect to time-period and failure mode without regard to the actual number of EDGs at each plant. The factors analyzed include subcomponent, failure cause, detection method, recovery, manufacturer, and EDG rating. The following increasing trends were identified for EDGs for the most recent 10-year period: • EPS and HPCS EDG frequency of start demands (demands per reactor year) • EPS and HPCS EDG frequency of FTLR demands • EPS and HPCS EDG frequency of run>1H hours. The following decreasing trends were identified for EDGs for the most recent 10-year period: • EPS EDG FTR>1H failure rate • EPS EDG unreliability • EPS and HPCS EDG frequency of FTR>1H events (failures per reactor year).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Detailed assessment of diesel spray atomization models using visible and X-ray extinction measurements

The physical mechanisms characterizing the breakup of a diesel spray into droplets are still unknown. This gap in knowledge has largely been due to the challenges of directly imaging this process or quantitatively measuring the outcomes of spray breakup, such as droplet size. Recent x-ray measurements by Argonne National Laboratory, utilized in this work, provide needed information about the spatial evolution of droplet sizes in selected regions of the spray under a range of injection pressures (50–150 MPa) and ambient densities (7.6–22.8 kg/m3) relevant for diesel operating conditions. Ultra-small angle x-ray scattering (USAXS) measurements performed at the Advanced Photon Source are presented, which quantify Sauter mean diameters (SMD) within optically thick regions of the spray that are inaccessible by conventional droplet sizing measurement techniques, namely in the near-nozzle region, along the spray centerline, and within the core of the spray. To quantify droplet sizes along the periphery of the spray, a complementary technique is proposed and introduced, which leverages the ratio of path-integrated x-ray and visible laser extinction (SAMR) measurements to quantify SMD. The SAMR and USAXS measurements are then utilized to evaluate current spray models used for engine computational fluid dynamic (CFD) simulations. We explore the ability of a carefully calibrated spray model, premised on aerodynamic wave growth theory, to capture the experimentally observed trends of SMD throughout the spray. The spray structure is best predicted with an aerodynamic primary and secondary breakup process that is represented with a slower time constant and larger formed droplet size than conventionally recommended for diesel spray models. Additionally, spray model predictions suggest that droplet collisions may not influence the resultant droplet size distribution along the spray centerline in downstream regions of the spray.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

KIVA Software

KIVA is a family of Fortran-based Computational Fluid Dynamics (CFD) software that predicts complex multi-species multi-phase turbulent fuel and air flows. KIVA models all engine combustion processes with models for spray, soot formation, spark ignition, and reactive chemistry (including pollutantformation processes). Employing KIVA software helps to optimize internal combustion engine processes, including diesel engines, for higher efficiency and lower emissions. KIVA engine modeling routines such as spray and core CFD Arbitrary Lagrangian-Eulerian (ALE) algorithms are now ubiquitous in the industry of engine code simulators. In addition, many engineers and researchers have made modifications to or enhanced our routines to address specific problems in engine designs.

20 FOSSIL-FUELED POWER PLANTS↗
DOE OSTI2017

Final Report: Utilizing Alternative Fuel Ignition Properties to Improve SI and CI Engine Efficiency

Experimental and modeling studies were completed to explore leveraging physical and chemical fuel properties for improved thermal efficiency of internal combustion engines. Fundamental studies of the ignition chemistry of ethanol and iso-octane blends and constant volume spray chamber studies of gasoline and diesel sprays supported the core research effort which used several reciprocating engine platforms. Single cylinder spark ignition (SI) engine studies were carried out to characterize the impact of ethanol/gasoline, syngas (H 2 and CO)/gasoline and other oxygenate/gasoline blends on engine performance. The results of the single-cylinder engine experiments and other data from the literature were used to train a GT Power model and to develop a knock criteria based on reaction chemistry. The models were used to interpret the experimental results and project future performance. Studies were also carried out using a state of the art, direct injection (DI) turbocharged multi- cylinder engine with piezo-actuated fuel injectors to demonstrate the promising spray and spark timing strategies from single-cylinder engine studies on the multi-cylinder engine. Key outcomes and conclusions of the studies were: 1. Efficiency benefits of ethanol and gasoline fuel blends were consistent and substantial (e.g. 5-8% absolute improvement in gross indicated thermal efficiency (GITE)). 2. The best ethanol/gasoline blend (based on maximum thermal efficiency) was determined by the engine hardware and limits based on component protection (e.g. peak in-cylinder pressure or maximum turbocharger inlet temperature) – and not by knock limits. Blends with <50% ethanol delivered significant thermal efficiency gains with conventional SI hardware while maintain good safety integrity to the engine hardware. 3. Other compositions of fuel blends including syngas (H 2 and CO) and other dilution strategies provided significant efficiency gains as well (e.g. 5% absolute improvement in ITE). 4. When the combination of engine and fuel system is not knock limited, multiple fuel injection events maintain thermal efficiency while improving engine-out emissions (e.g. CO, UHC, and particulate number).

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2016

Sensitivity study of PWR upper vessel flow and phenomena using RELAP5-3D

During normal pressurized water reactor (PWR) operation, water enters the reactor pressure vessel (RPV) from the cold leg and exits through the hot leg. While most of this water flows through the downcomer and reactor core, a number of other flow paths exist, including the core baffle/bypass and the upper plenum spray. The upper plenum sprays normally account for a small fraction (less than 10%) of the total inlet flow, and provide cooling to the upper vessel dome and instrumentation/control rod channels. In the event of a loss-of-coolant accident (LOCA), especially one with core blockage, this alternative flow path may play a key role in maintaining core coolability. Because of the possible safety implications of the upper vessel flow, it is important to ensure the models used to simulate this region of the vessel are accurate. In this study, the model was altered to better reflect the internal flow path of the upper head and this was compared to the reference model. The simulation of interest is a hot leg double-ended guillotine (HL DEG) break followed later by a full blockage of the core and core bypass. In this scenario, the coolant can only reach the core through the upper spray bypass, so its impact on core coolability can be clearly seen. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Characterization of Oxy-Coal Swirl Injector

The article presents the design and testing of a coal-slurry swirl-pintle injector for high pressure oxy-coal combustion systems. Pressurized oxy-coal combustion-based systems have the potential to improve efficiency by recovering latent heat of the steam in flue gas and achieve 90% CO2 capture. In oxy-coal combustion, coal is burned in the presence of pure oxygen, which results in higher efficiency and less greenhouse gas in the exhaust. With the advancement of pressurized oxy-coal combustor research, there is a need to investigate injector technology suitable for this unique combustion approach. Effective atomization and mixing of coal-slurry and oxidizer pose a challenge for the efficient burning of coal particles. A swirl-pintle type injector was designed and tested to enhance the atomization of the coal-water slurry spray by adding a swirl to the oxygen flow path. Although pintle type injectors are traditionally used in rocket engines, they present a promising approach in clean coal combustion technology due to higher mixing effectiveness and enhanced performance. The mixing behavior of three pintle injectors with different swirl numbers (S = 0, 0.9, and 1.2) was tested using a high-speed shadow sizing technique. Mixture ratios of 30% (by mass) coal-water slurry were studied. The droplet diameter and dispersion characteristics were determined from shadow images of coal-water spray. The jet breakup length, droplet number, total moment ratio (TMR), and Ohnesorge number were then calculated to compare the performance of different injectors. The paper presents an empirical correlation of the TMR with droplet size and spray regimes. For a pintle injector with S = 1.2 swirler, the droplets diameter was found to be in the range of 1 mm to 12 mm. The jet breakup length was also shorter in comparison to S= 0 and 0.9 injectors. A second atomization zone was also observed at 2 and 5 diameters downstream of injector exist at spray edge and jet core respectively

01 COAL, LIGNITE, AND PEAT↗