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

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At least 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Dismantling of the Bohunice V1 Reactor Coolant System - 20011

About 2.5 kilometers from the village of Jaslovske Bohunice in the Trnava District in western Slovakia at the Bohunice V1 nuclear power plant, Westinghouse and VUJE are working on the dismantling of two VVER-440 units. The consortium was awarded this contract in October 2017 from Jadrova a vyradovacia spolocnost (JAVYS) and has begun onsite work to dismantle the reactor coolant systems of both units that operated at the plant. The Bohunice V1 plant is composed of two V-230 type VVER 440 units (440 MWe each). The first unit ceased operation on December 31, 2006, and the second on December 31, 2008, after 28 years of operation. The project strategy relies on the early release of some plant areas to allow the erection of new facilities to enable a fast and safe dismantling. These up-front activities play a crucial role within the whole project. Since schedule and ALARA are two key elements to be considered, it has been decided to separate the dry cutting activities from the ones occurring underwater. Two Wet Cutting Workshops will allow the segmentation and packaging of the high activity components. To take advantage of the large space available in the turbine building, a Dry Cutting Workshop has been built for segmenting the twelve steam generators of both units. Two more Dry Cutting Workshops (one per unit) have been installed in the reactor building for the fragmentation of the remaining large components and auxiliary systems components. This paper describes all current preparatory works as well as the on-going segmentation and packaging activities. Interesting lessons learned are also provided. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Multi-scale simulation of high burnup UO2 nuclear fuel during loss-of-coolant accident conditions

To improve the economics of commercial nuclear energy generation, U.S. utilities are currently seeking licensing approval to operate UO2 fuel to higher burnups. One significant safety issue that must be addressed to obtain approval is the potential for fine fragmentation/pulverization of the fuel during a loss-of-coolant accident (LOCA). It has been hypothesized that pulverization is caused by the rapid increase of pressure in fission gas bubbles in the high burnup region of the fuel during a LOCA. To better understand this phenomenon, a novel phase-field model of the fission gas bubble microstructure in UO2 has been developed and implemented in Idaho National Laboratory (INL)'s Marmot application for phase-field simulation of nuclear materials. Simulations of the bubble response to steady-state and transient conditions were conducted, and the results were used to inform a mechanistic model of pulverization in BISON, INL’s fuel performance simulation code.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

DEVELOPMENT OF TRANSIENT FISSION GAS RELEASE CAPABILITIES TO STUDY HIGH BURNUP COMMERCIAL FUEL PERFORMANCE UNDER LOSS OF COOLANT ACCIDENT CONDITIONS

The contribution of transient fission gas release to the burst behaviour of high burnup fuel during a loss of coolant accident (LOCA) for commercial light water reactors is a complex phenomenon that is not fully understood. During the temperature ramp associated with a LOCA, there is significant, additional transient fission gas release in excess of steady state fission gas release from normal operation. The timing of this release and the various conditions that impact this release warrant further study. To better characterize this phenomenon, a system was developed to capture and characterize fission gas released from irradiated fuel during simulated LOCAs. This system is an upgrade to the existing Oak Ridge National Laboratory Severe Accident Test Station LOCA test apparatus and detects fission gas by gamma counting cryogenically cooled cold traps and by using an online gas mass spectrometer. The examined fuel can be pressurized or open to the furnace atmosphere. The design and operation of this new capability are discussed. Results from the first tests on high burnup fuel are presented. In the initial test, the LOCA relevant transient fission gas release was 5.3% for an unpressurized segment. Post-test microscopy on the tested samples is presented to demonstrate the likely radial source of transient fission gas release.

Harp, Jason↗
DOE OSTI2024

Computational Investigation of Fuel Dispersal Phenomena during Large-Break Loss of Coolant Accident in Light-Water Reactors

In the event of cladding rupture, which could occur in light water reactor fuel assemblies during a loss-of-coolant accident (LOCA), fuel particles, along with fission gases, can be expelled into the reactor core from the fractured fuel rod. This expulsion of fragmented fuel particles, referred as fuel dispersal, is the subject of investigation to evaluate the safety implications of increasing fuel burnup in light water reactors, with a specific focus on fuel fragmentation, relocation, and dispersal. Particle trajectories and the resulting mass distribution of the settled particles within the reactor pressure vessel can pose a long-term cooling challenge for the reactor core. The fuel dispersal phenomenon is significantly influenced by the ejection characteristics of the fuel fragments, as well as the size and shape of the cladding rupture and fuel rod depressurization history during LOCA transients. In this study, the transport of fuel particles within a scaled 5 × 5 lattice of a pressurized water reactor rod bundle geometry is modeled through a two-fluid Eulerian framework that treats the gas and solid phases as interpenetrating continua. The required boundary conditions are evaluated from the fuel performance code BISON in a postulated large-break LOCA scenario. The modeling framework considers solid fuel particles as granular matter, interacting with the gaseous dry steam phase and fission gases through the governing interfacial momentum and energy exchange between the gas and solid phases. The simulation results provide the volume fraction of the solids settled on the bottom surface of the fuel bundle, quantifying the deposition within the bundle geometry.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Predictive Modeling and Uncertainty Quantification in Condition Monitoring of Active Components: A Reactor Coolant Pump Use Case

This work develops data-driven models for onset of thermal barrier leakage in reactor coolant pumps. It incorporates uncertainty quantification to enhance the reliability and robustness of pre- dictions. Using synthetic data generated by the Generic Pressurized Water Reactor simulator, realistic degradation scenarios were simulated across lifecycle stages—beginning, middle, and end of life. Key variables, including differential pressure, flow rate, vibration, and temperatures, were analyzed using machine learning framework. The fully connected neural network models demonstrated exceptional performance, achieving R2 scores exceeding 0.99 and root mean square errors as low as around 8.23 × 10-2 gallon per minute (gpm) for the three stages of the lifecy- cle. UQ analysis further validated the model’s robustness, with narrow uncertainty bounds during steady-state operations and appropriately wider bounds during transitional phases, reflecting the physical behavior of the system. This work addresses important gaps in real-time condition moni- toring and regulatory compliance by integrating advanced condition monitoring technologies with UQ into IST programs. The ability to detect thermal barrier leakage early and quantify prediction reliability supports optimizing maintenance strategies while ensuring nuclear power plants’ safe and reliable operation.

99 - GENERAL AND MISCELLANEOUS↗
DOE OSTI2023

SPC-70734 Rev 1 MARVEL Primary Coolant System (PCS) Fill and Eutectic NaK Procurement

This specification contains the requirements for the filling the Microreactor Applications Research Validation and Evaluation (MARVEL) reactor Primary Coolant System (PCS) with eutectic sodium-potassium (NaK). The subcontractor will design and fabricate a NaK filling system (referred to herein as the PCS Fill System), the requisite volume of high purity NaK, and perform filling operations in conjunction with the Contractor Operations. The MARVEL reactor system will be operated by the Contractor, and the PCS fill system will be operated by Subcontractor personnel.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2024

ECAR-7813 Rev 0 Secondary Coolant Cover Gas System Stirling Engine Tube Rupture Analysis, Insulation Block Analysis, and Labyrinth Seal Design

The purpose of this ECAR is to document compliance with best engineering practices for the Secondary Coolant Cover Gas System of the MARVEL microreactor which includes a Stirling Engine helium tube rupture analysis, an analysis to verify the strength of the insulation blocks, and the design of the Labyrinth Seal.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2020

Thermal Design Analysis Supporting a Loss-of-Coolant-Accident Experiment at TREAT

The LOCA-SERTTA experiment vehicle along with the power and flexibility of TREAT allow for the ability to create the temperature profile evolution throughout a LOCA that would be seen in a commercial LWR. Presented in this summary are thermal-hydraulic modeling results for LOCA-SERTTA which are then compared to LOCA simulation results from a commercial LWR.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

ECAR-6330 Rev 0 Computational Fluid Dynamics Evaluation of Core Pressure Drop for the Primary Coolant Apparatus Test

PCAT is a thermal-hydraulic test loop for MARVEL reactor. The design gap thickness between the heater rods of PCAT core is 1.0 mm. The heater elements were manufactured within a straightness tolerance for the designed gap thickness, but several elements have been distorted out of tolerance during the welding process. The distortion of heater rods could lead to flow maldistribution and excessive hotspot temperature when the gap between two heater rods is too narrow. Wire-wraps are the spacer devices widely used in the sodium cooled fast reactor system to maintain the design fuel pitch. Generally, a wire spacer can enhance the turbulent heat transfer, but it increases the pressure drop, which is not desirable for a natural-circulation system like PCAT. Therefore, three short wire-wrap spacers per heater have been proposed to contain the core pressure drops and keep the heater-to-heater gap in an acceptable range. The objectives of this ECAR is to evaluate the PCAR core pressure drop with short wire spacers to confirm that the additional pressure drop due to the wire spacer will not disturb the natural circulation of the PCAT.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2025

Bi‐Continuous W‐Rich Refractory High Entropy Alloy‐Cu Composite: Toward Material Innovation of Nuclear Reactor Coolant System

Abstract Refractory high‐entropy alloys (RHEAs) are considered promising candidate materials for next‐generation nuclear reactors due to their superior mechanical strength, irradiation resistance, and thermal stability at high temperatures. However, the significant positive heat of mixing between refractory alloying elements and Cu, commonly used in cooling systems, poses challenges in forming composite structures. This study addresses the issue using a liquid metal dealloying (LMD) process. A precursor alloy (WTaVTi) with a directional dendrite‐interdendrite structure is fabricated and reacted with molten Cu at 1200 °C for 96 h. This approach produced a RHEA‐Cu composite with a stable interface between RHEA (W 31.5 Ta 30.9 V 21.4 Ti 14.3 ) and Cu, featuring a spontaneously formed W‐rich interlayer that enhances interfacial bonding. The composite showed excellent irradiation resistance, with 30% less swelling under α‐ion irradiation than pure W. It also exhibited low thermal conductivity at room temperature, but reached ≈120 W m −1 ·K −1 at ≈650 °C, surpassing pure W. This temperature‐dependent rise in κ, with a positive gradient of +0.075 W m −1 ·K − 2 , is attributed to decreasing diffuse mismatch at elevated temperatures. The large‐scale reaction and stable microstructure achieved through LMD process highlight its industrial potential. This work offers a strategy for developing high‐performance materials by combining RHEA's radiation resistance with Cu's thermal conductivity for extreme environments.

Chemistry↗
DOE OSTI2020

A novel approach for critical heat flux enhancement during severe accident mitigation with removal of radioactive materials from the coolant

In this paper, a new system, which consists of mature and proven technologies, is suggested for simultaneous objectives in nuclear power plants: CHF enhancement and effective decontamination. It consists of multifunctional magnetic nanoparticles and a collector part. CHF enhancement for severe accident mitigation strategies like IVR-ERVC and core catcher is expected along with an efficient decontamination process of target radioactive nuclides, and through the collector part, where external magnetic fields are applied, working fluid can be effectively purified. The multifunctional magnetic nanoparticle possesses potentials to severe accident scenarios based on its distinct characteristics. Also, a nanofluid system is strongly affected by environments and state of the immersed particles; the CHF enhancement can be stably achieved in real applications with proper consideration and optimization.

Kam, Dong Hoon↗
DOE OSTI2021

CFD Evaluation of Pressure Change Along Coolant Passages in Sodium-Cooled Fast Reactor with Nek5000

To support the design efforts of advanced sodium-cooled fast reactors (SFRs), a series of computational fluid dynamics (CFD) simulations are performed to investigate the pressure change along various flow passages in the proposed SFR system. The simulations are carried out with the state-of-the-art spectral element flow solver, Nek5000. Two specific case studies are presented in this paper: the flow exiting the axial neutron reflector channels and the flow entering the fuel pin bundle. Due to the high Reynolds numbers expected, a Reynolds-averaged Navier-Stokes (RANS) approach is necessary to model the turbulence. A newly developed regularized k-omega RANS model is adopted in the related CFD calculations. The first case study explores the effect of Reynolds number on the pressure change when flow exits the reflector channels. The pressure change in this case has two major contributors: the change due to wall friction and the Bernoulli effect. It is noted that the nondimensional pressure loss follows a log-linear trend up to Re = 10 5 , and then the trend is flattened. In the second case study, the advanced NekNek coupling capability is tested where an integral domain can be divided into multiple subdomains with coupling interfaces, which would greatly ease the meshing process of complex engineering geometries and potentially save computational resources. The preliminary results obtained so far confirm the consistency between the NekNek results and those produced by regular Nek5000 simulation. The presented work demonstrates the readiness and flexibility of the related CFD techniques, which is part of the broader effort to leverage cutting-edge CFD to inform the advanced nuclear reactor designs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗