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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 OSTI2019

Assessment of Pump Coastdown and Reactor Protection Scram Delays during Station Blackout Conditions

This report presents an assessment of the impact of pump coastdown time and reactor protection scram delays for the Versatile Test Reactor (VTR) during Protected Station Blackout (PSBO) conditions. While previous simulations for the VTR assumed centrifugal pumps in the primary system, the current pre-conceptual design uses electromagnetic (EM) pumps. Because there is uncertainty as to what the parameters of the EM pump design will be, this report focuses on the impact of the flow coastdown and the reactor protection scram delay, not on the needs of the pump design. PSBO transient simulations were performed using the SAS4A/SASSYS-1 fast reactor safety analysis code (version 5.2.3). The model used in the simulations is based on the FY18 SAS4A/SASSYS-1 model. A change to the model was made to replace the centrifugal pump model with an electromagnetic pump model with a motor generator as a coastdown mechanism. The results of the PSBO transient simulations were assessed against the criteria used in Reference 2 with an additional 50°C for conservatism to account for uncertainties. The results show that as the delay before scramming the rods increases, the moment of inertia of the coastdown mechanism must increase to satisfy the criteria. By increasing the moment of inertia to extend the flow coastdown, temperatures can be kept below the desired limits with reactor protection scram delays between 3 and 10 seconds.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Security Catch and Release Automation Manager (SCRAM) v1.0

This is a security automation tool heavily influenced by the tool written at NCSA https://github.com/ncsa/bhr-site. The tool is a web interface that aims to centralize information on blocking malicious traffic. There is a list of currently blocked traffic, a place for other tools to automatically add traffic they deem malicious, as well as for human users to input blocks. * Logging and auditing, to determine who made what change and when * Automated expiration of changes (e.g. an exception is added for 2 weeks, after which time it will be removed) * Validation of changes and verification that the change has been applied * Fine-grained role-based access

Oehlert, Samuel↗
DOE OSTI2018

CRDM with separate SCRAM latch engagement and locking

A control rod drive mechanism (CRDM) configured to latch onto the lifting rod of a control rod assembly and including separate latch engagement and latch holding mechanisms. A CRDM configured to latch onto the lifting rod of a control rod assembly and including a four-bar linkage closing the latch, wherein the four-bar linkage biases the latch closed under force of gravity.

42 ENGINEERING↗
DOE OSTI2019

CRDM with separate scram latch engagement and locking

A control rod drive mechanism (CRDM) configured to latch onto the lifting rod of a control rod assembly and including separate latch engagement and latch holding mechanisms. A CRDM configured to latch onto the lifting rod of a control rod assembly and including a four-bar linkage closing the latch, wherein the four-bar linkage biases the latch closed under force of gravity.

Dodd, Christopher D.↗
DOE OSTI2019

CRDM designs with separate SCRAM latch engagement and locking

A control rod drive mechanism (CRDM) includes a lifting rod supporting a control rod and a holding mechanism comprising an electromagnetic circuit with magnetic poles drawn together when the electromagnetic circuit is energized to hold the lifting rod. The hold is released upon de-energizing the electromagnetic circuit. A translation mechanism linearly translates the lifting rod held by the holding mechanism. The holding mechanism may include a non-magnetic spacer between the magnetic poles that defines a gap between the drawn together magnetic poles. The translation mechanism may include latches configured to engage an upper end of the lifting rod, and the holding mechanism draws the magnetic poles together to hold the latches engaged with the upper end of the lifting rod. A four-bar cam assembly may be used to cam the latches closed in response to a vertical actuation force applied to the cam bars.

DeSantis, Paul K.↗
DOE OSTI2025

Performance Analysis and Simulaion of the Hydraulic Scram System in TREAT Reactor

The Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) serves a vital role in nuclear fuel safety research, enabling transient experiments that simulate reactivity excursions and accident scenarios. Central to these operations is the transient control rod drive system (TCRDS), which drives rapid motion of the transient control rods such that TREAT can simulate rapid power changes typical of reactor accidents. The reliability and performance of this system are critical for protecting both fuel specimens and reactor infrastructure. This study presents the initial phase of a two-year investigation into the dynamics and reliability of the TREAT hydraulic TCRDS. Conducted in collaboration with INL, the research employs a combined computational and experimental approach to analyze the system's response time, pressure transients, and potential failure modes. Emphasis is placed on understanding how fluid characteristics influence the TCRDS’s ability to achieve both rapid power changes and mechanical stability. The TRDS and the skid that powers it will be analyzed throughout this investigation. Computational modeling using computational fluid dynamics (CFD) will simulate the hydraulic response under varying conditions. In parallel, experimental testing planned at INL will validate these models and capture key performance metrics. This paper outlines the system design, analytical framework, and modeling strategies that form the foundation for later testing. Ultimately, this work aims to support improvements to the TCRDS’s design and reliability, contributing to the broader goal of enhancing nuclear fuel safety and sustaining TREAT’s mission as a premier nuclear fuel test facility.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2019

Analysis of VTR Primary Pump Coastdown

Simulations were performed to evaluate how different flow coastdown rates impact transient results. These simulations were performed with a preliminary SAS4A/SASSYS-1 model of the VTR to assess the advantages and disadvantages of delaying a pump trip following a control rod scram. When the control rods are inserted into the core by the reactor protection system, fission power will drop quickly. Pumps should be tripped soon after to avoid decreasing the power-to-flow ratio too much and preventing thermal shock of components in and above the core. The primary motivation for this report is characterizing the temperatures of these components with various delays between the control rods scramming and the primary pumps tripping. In the simulations with a control rod scram and a delayed pump trip, even if one of the pump coastdown mechanisms fails, a faster pump trip and coastdown is preferable. This can be achieved both by shortening the delay between the scram and trip as well as by decreasing the inertia provided by the coastdown mechanism. Based on the results of these simulations, there is no benefit to delaying when the pumps trip following a control rod scram. In other scenarios, however, a slow flow coastdown rate is preferable. Results from unprotected loss of flow simulations are presented to demonstrate that a slower coastdown reduces how much temperatures increase within the core. An initial flow halving time of between five and seven seconds was predicted to produce peak in-core temperatures with sufficient margins to the limiting criteria. For halving times less than five seconds, peak temperatures significantly reduce margins. For halving times greater than seven seconds in the unprotected loss of flow cases, the resulting peak temperatures are lower, but this leads to unfavorable results for the previous cases with a control rod scram. In order to achieve an acceptable balance between cases with and without a preceding control rod scram, the results presented in this report support an initial flow halving time in the five-to-seven second range, although ongoing PRA analysis could produce results that support a coastdown beyond this range.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Development of a new control rod drive mechanism design for the ISU AGN-201M reactor

The Aerojet General Nucleonics (AGN) model 201-Modified, known as the AGN-201M reactor, plays an essential role in the educational and research activities at Idaho State University (ISU). The ISU AGN-201M's original Control Rod Drive Mechanism (CRDM) has been in operation for more than fifty years with no large-scale redesigns. The CRDM is required to eject the fuel rods within one second during a SCRAM event (also known as a 'reactor trip') and adjust the control rods' insertion speed, and keeps the rod insertion sequence correct. The existing control rod drive mechanisms meet these criteria but experience a few concerns due to aging. Concerns include complex maintenance and costly repairs for old electromechanical components, rod position feedback errors, and the impediment of the plate during a SCRAM due to binding of the lead screws of the existing mechanism. During a binding event, the drive mechanism becomes locked, preventing the control rod's magnetic plate from moving in or out under the reactor's normal and emergency operating conditions. Although, the binding has no effect on the ability of the rod to exit the core during the SCRAM. To counteract the concerns and issues with the current CRDM, a new design has been proposed using newer components and a simplified design. The new design utilizes more advanced electric and mechanical components that are commercially available. The new CRDM system is divided into four main aspects: (1) control rod movement design (motor, lead screw, guide rods), (2) control rod ejection (springs, electromagnet), (3) control rod position and feedback (position transducer, microswitches), and (4) material selection and structural analysis. The new design aims to reduce the overall complexity and probability of failure to improve the reactor's overall reliability. With proper material selection and improved structural design, the new drives are lighter with little to no change in structural integrity. The new control rod drive mechanism eliminates binding scenarios by using a single lead screw and implementing additional guide rods. An advanced linear position sensor and microswitches replace the existing and aging synchro system for accurate rod position feedback resulting in better reactivity control. The new design meets the reactor's operational limits by having an average reactivity insertion of 0.065% Δk/k per second, which corresponds to a total control rod insertion time of 19.23 s, while the control rod's ejection time remains less than one second during a SCRAM event. The new design ensures the reactor's long-term viability for educational and research activities by increasing the reliability and safety of operation for years to come.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Numerical Simulation of Measurements during the Reactor Physical Startup at Unit 3 of Rostov NPP

The results of numerical calculations and measurements of some reactor parameters during the physical startup tests at unit 3 of Rostov NPP are presented. The following parameters are considered: the critical boron acid concentration and the currents from ionization chambers (IC) during the scram system efficiency evaluation. The scram system efficiency was determined using the inverse point kinetics equation with the measured and simulated IC currents. The results of steady-state calculations of relative power distribution and efficiency of the scram system and separate groups of control rods of the control and protection system are also presented. The calculations are performed using several codes, including precision ones.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Analysis of Reverse Flow Restriction Device to Prevent Dryout Fuel Damage during BWR Instability

Boiling Water Reactors (BWRs) are subjected to instabilities due to thermal hydraulics feedback or coupled Neutronics-thermal hydraulics feedback. Feedback is affected by many reactor parameters that could destabilize the core such as distribution of pressure drop across the core, axial power shape, subcooling level at the core inlet, radial power peaking factor, reactivity coefficients, and others. There are three common modes of instability in BWRs: single channel instability, core-wide instability, and regional instability. Each instability mode could damage the fuel if the power oscillations are left without an intervention like reactor scram. However, the instability problem becomes more dangerous if the automatic reactor scram is lost, like in Anticipated Transient without Scram (ATWS) accidents. A reverse flow restriction device (RFRD) would mitigate this problem. Core simulation using TRACE/PARCS code has been used to examine the potential effect of restricting the flow in the downward direction on mitigation of the power and flow oscillations to increase the safety of BWRs. The presented results demonstrate the ability of the device on limiting flow oscillation amplitude in the one bundle where it is used, on an adjacent bundle without it and reducing the power peak when the flow is low and power is high during BWR power oscillations. This exercise demonstrates the potential of previously described hardware device on preventing clad temperature excursion to keep the fuel safe from dryout damage and provide encouraging results ahead of the planned more detailed investigation where the device is to be used in more bundles up to full core. In the future work it is planned to examine the effect of device installed in a fraction of the bundles on the flow rate in other bundles. Also, the effect on the characteristics of dryout and rewetting and the resulting temperature excursions with and without the device will be given special attention. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

The ATWS analysis of the HTR-10GT core induced by the earthquake - 18489

The 10 MW High Temperature Gas Cooled Test Reactor (HTR-10) has been built in Institute of Nuclear and New Energy Technology (INET) and has been operating successfully since the beginning of 2003. The core outlet temperature of HTR-10 is 700 deg. C. To verify the technology of gas-turbine direct cycle, at first INET had a plan to increase its core outlet temperature to 750 deg. C and use a helium gas turbine instead of the steam generator (then the reactor is called HTR-10GT). Though HTR-10 has good intrinsic safety, the design basic accidents and beyond design basis accidents of HTR-10GT must be analyzed according to China s nuclear regulations due to changed operation parameters. THERMIX code system is used to study the Anticipated Transient Without Scram (ATWS) caused by an earthquake. The gravity acceleration of the earthquake will make the porosity of the pebble bed smaller, which will thus induce a positive reactivity. The packing fraction of the HTR-10GT core will increase to 0.64 from 0.61, while a reactivity of 1.24% will be inserted into an initial core and 0.788% for an equilibrium core. Then the reactor power increased and the temperature of the core increased. When the neutron flux of power measuring range exceeded 123% and the reactor period becomes less than 20 seconds, the reactor should scram. It was supposed that all the control rods in the reflectors had been blocked and the reactor could not scram. Thus the accident went on and the core temperature and the system pressure increased but the reactor shutdown at last because of its natural negative temperature reactivity feedback mechanism. The residual heat would be removed out of the core by the cavity cooling system. During the accident sequence the maximum fuel temperature was 1223 deg. C. If a temperature coefficient error of 20% was considered during analyzing, the peak temperature of the fuel during the ATWS will reach 1266 deg. C. It was a little higher than 1230 deg. C --the fuel temperature limitation of HTR-10. Now the sphere fuel used in HTR-10GT will also be used in HTR-PM and the temperature limitation has been raised to 1620 deg. C, so the HTR-10GT is safe during the ATWS caused by an earthquake. The paper also compares the analysis result of HTR10-GT to those of HTR-10. The results shows that the HTR-10GT is still safe during the accident though its operating temperature is higher than HTR-10. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Electric actuator modification on a turbine driven feed pump at Calvert Cliffs nuclear power plant - 296

Feedwater systems are a leading contributor to unit SCRAMs as indicated by the 2015 INPO data, where 9 of the 51 SCRAMs recorded were a result of feedwater issues. Digital feedwater control systems and enhanced feed pump control modifications can greatly improve performance and control of the plant feedwater systems and help drive down the potential for reactivity management events and unit SCRAMs. Westinghouse has a strong background in the implementation of digital feedwater control systems, and in planning for a recently completed feedwater control system upgrade at Calvert Cliffs, the Westinghouse team worked with Calvert Cliffs to design a solution to a number of common operational issues. Primarily, improving feedwater control over the entire operating range, eliminating single points of vulnerability (SPV) in the feedwater/feed pump systems and eliminating operational issues with the feed pump steam admission valve hydraulic control oil system. A robust first-of-a-kind design and implementation program was undertaken to split the existing hydraulic valve gear actuator with separate electric actuators. Westinghouse worked with Calvert Cliffs and Emerson Process Management (EPM) to install and commission General Electric (GE) feed pump turbines with separate electric actuators for the high-pressure (HP) and low-pressure (LP) steam admission valves. This successful mechanical and control modification was a first for both the nuclear and non-nuclear power generation industries. Westinghouse is working with a number of other utilities to implement similar modifications for PWR and BWR feed pump applications. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Evaluating the X2 initial core zero power physics tests with Serpent-Ants

The validation of the Ants nodal neutronics code for VVER applications is started by modelling the zero power physics tests for the initial core of the Khmelnitsky 2 nuclear power plant as described in the X2 benchmark using the Serpent-Ants two step neutronics calculation chain. The Ants prediction compare favorably against an earlier continuous energy Monte-Carlo reference solution as well as the measured data, except for discrepancies in the SCRAM worth between predicted and measured values. Similar discrepancies have been previously reported for VVER-1000 reactors, and the Ants predicted SCRAM worths match well with the Serpent results. Additionally, the assembly power distribution and axial power distribution predicted by Ants for the critical HZP state of the reactor are also compared to a reference Serpent solution with very good results. Lastly, an investigation into the effects of the few-group structure used in the nodal calculations and the use or lack of leakage correction for group constants on the results shows that the best accuracy is reached with seven or eight energy groups without leakage correction, although a reasonable accuracy can also be obtained with two or three energy groups using fundamental mode leakage correction. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

TRACG benchmark study: 2015 Fermi thermal-hydraulic instability event

On March 19, 2015, the Boiling Water Reactor (BWR) Fermi 2 (referred to hereafter as Fermi), located in Michigan, U.S., experienced a scram initiated by the Oscillation Power Range Monitor (OPRM) system. The plant had been operating in Single Loop Operation (SLO) for approximately 10 minutes before the scram occurred. During this time, the plant operating conditions reached a region within the power/flow (P/F) map susceptible to coupled neutronic/thermal-hydraulic (T-H) instabilities. Limited code benchmarking is available in literature for T-H Instability (THI) transients occurring in plants operating at SLO conditions due to scarce plant data for this type of operational condition. This study aims to model and benchmark the THI event at SLO conditions using TRACG, the GE Hitachi (GEH) proprietary best-estimate transient system code. The TRACG model was benchmarked by comparing the results with measured plant data, such as the total reactor power and core flow, from the event. Additionally, to evaluate the inherent stability of the plant at different P/F points, Decay Ratio (DR) calculations were performed. The capability of TRACG for calculating DRs is well documented, and it has also been approved by the U.S. Nuclear Regulatory Commission (NRC) as part of ESBWR TRACG stability licensing topical reports. The results of TRACG DRs were compared to the GEH best-estimate frequency domain code ODYSY, which is the current Fermi licensing basis code for reload analysis DR calculations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Electron temperature measurements inside the ablating plasma of gas-filled hohlraums at the National Ignition Facility

The first measurement of the electron temperature (T{sub e}) inside a National Ignition Facility hohlraum is obtained using temporally resolved K-shell X-ray spectroscopy of a mid-Z tracer dot. Both isoelectronic- and interstage-line ratios are used to calculate the local T{sub e} via the collisional–radiative atomic physics code SCRAM [Hansen et al., High Energy Density Phys 3, 109 (2007)]. The trajectory of the mid-Z dot as it is ablated from the capsule surface and moves toward the laser entrance hole (LEH) is measured using side-on x-ray imaging, characterizing the plasma flow of the ablating capsule. Data show that the measured dot location is farther away from the LEH in comparison to the radiation-hydrodynamics simulation prediction using HYDRA [Marinak et al., Phys. Plasmas 3, 2070 (1996)]. To account for this discrepancy, the predicted simulation T{sub e} is evaluated at the measured dot trajectory. The peak T{sub e}, measured to be 4.2 keV ± 0.2 keV, is ∼0.5 keV hotter than the simulation prediction.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Material Selection and Heat Transfer Model for PTUHS Device

The Path to Ultimate Heat Sink Device (PTUHS) is a passive safety system that utilizes a radiative heat transfer valve to control the rate at which heat is removed from a nuclear reactor pressure vessel (RPV) and dispersed into surrounding soil. The material selection for the PTUHS device is investigated, where the thermophysical properties are either maximized or minimized bases on what section of the PTUHS that material is being used for. Final recommendations of material choice are then presented. The heat transfer model of the PTUHS device is solved in ABAQUS, where a normal operating conditions and SCRAM conditions are solved. These temperature maps and heat flux of the system show the system's ability, where minimal heat is lost during normal operating conditions and system failure in a SCRAM scenario due to temperatures that were reached.

42 - ENGINEERING↗