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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 145 records · Page 8
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Coupled Deterministic and Monte Carlo Neutronics for Vessel Fluence Calculations

Throughout core life, the reactor vessel is continually bombarded with neutron flux embrittling the vessel and limiting its useful life. As a result, it is imperative to adequately predict how much flux has passed through the vessel during all stages of its life, as a means of ensuring it has adequate ductility to preclude a brittle fracture even under design accident conditions. At present many of these predictions are made either by purely deterministic or purely stochastic means, each offering their own advantages and disadvantages. This paper introduces a coupled method for calculating fluence through the VERA framework, which enables modeling incremental accumulations of fluence while simultaneously accounting for several factors as they evolve through core life. The intention of this method is to simplify the calculation process and provide better resource utilization in obtaining the required accuracy.

Henderson, Shane↗
DOE OSTI2017

LOCA Analysis for the NIST Research Reactor

The research reactor (aka NBSR) at the National Institute of Standards and Technology (NIST) is a heavy water cooled, moderated, and reflected tank-type research reactor that operates at atmospheric pressure and a design power of 20 MW. Core power is generated in 30 fuel elements, each with 17 fuel plates in an upper section and a separate lower section. Power is removed using forced convection and heat exchangers in the cooling loop. The licensing analysis of the loss-of-coolant accident (LOCA) in the NBSR was addressed in the current Safety Analysis Report (SAR). The SAR discusses the low probability of a significant pipe break: 'the main piping is located in protected areas, system pressures are low, and flow rates are small so that wear is not an issue.' Therefore, the probability of a large break, including a double-ended guillotine break, is extremely low. For smaller breaks where the operator has time to take action, procedures are in place to 1) mitigate a loss of water by tripping pumps and closing control valves to isolate the reactor vessel after the falling water level in the vessel is detected by instrumentation, and 2) ensure that emergency cooling water continues to flow for as long as needed. The SAR also refers to an analysis of why there is sufficient water without forced cooling to cool the fuel elements (FEs) in a LOCA if the flow channels between two fuel plates (or a fuel plate and an outside plate) remain filled with water. That analysis was based on a flooding-limited critical power but assumed that reactor trip had taken place. A more recent study showed that some scenarios may lead to the draining of water from the coolant channels. In these cases, the cooling on the inside of the fuel element is from emergency cooling water that falls only along the inner surface of one of the side plates in the fuel elements; the remaining surfaces within the fuel element being exposed to gas. The analysis therein was done for both the present high-enriched uranium fuel (HEU) and the proposed low-enriched uranium fuel (LEU). The LOCA analyses done previously assumed that there is sufficient cooling during the short-term, i.e., before the water level reaches the top of the fuel plates and begins to drain from the coolant channels or before decay heat becomes low enough such that the flooding-limited critical power shows that there is no problem. The present study is to determine if that assumption is correct, i.e., to consider the situation immediately after the break, before and after reactor trip when the power is still relatively high.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Mid–height seismic isolation of equipment in nuclear power plants

An innovative seismic isolation solution for designers of safety-class equipment in advanced nuclear power plants is introduced. The test specimen was a tall, slender, carbon steel vessel that could represent a reactor vessel, steam generator, or a heat exchanger in a nuclear power plant: 240 inches tall, outer diameter of 60 inches, and wall thickness of 1 inch. The vessel was supported by three radial mounts at its mid-height, near its center of gravity, on a steel frame. The vessel was subjected to three-component ground motions using a 6DOF earthquake simulator. The specimen was filled with water for testing to indirectly account for the fluid and internal equipment present inside a prototype vessel. Three configurations were tested: non-isolated, isolated using single Friction Pendulum (SFP) bearings, and isolated using triple Friction Pendulum (TFP) bearings. The test results demonstrate that mid-height seismic isolation is practical and enables a significant reduction in horizontal spectral accelerations. Furthermore, these outcomes are not specific to the spherical sliding bearings used in the experiments but are broadly applicable to mid-height, seismically isolated equipment.

42 ENGINEERING↗
DOE OSTI2022

MELCOR Accident Progression and Source Term Demonstration Calculations for a FHR

MELCOR is an integrated thermal hydraulics, accident progression, and source term code for reactor safety analysis that has been developed at Sandia National Laboratories for the United States Nuclear Regulatory Commission (NRC) since the early 1980s. Though MELCOR originated as a light water reactor (LWR) code, development and modernization efforts have expanded its application scope to include non-LWR reactor concepts. Current MELCOR development efforts include providing the NRC with the analytical capabilities to support regulatory readiness for licensing non-LWR technologies under Strategy 2 of the NRC's near- term Implementation Action Plans. Beginning with the Next Generation Nuclear Project (NGNP), MELCOR has undergone a range of enhancements to provide analytical capabilities for modeling the spectrum of advanced non-LWR concepts. This report describes the generic plant model developed to demonstrate MELCOR capabilities to perform fluoride-salt-cooled high-temperature reactor (FHR) safety evaluations. The generic plant model is based on publicly-available FHR design information. For plant aspects (e.g., reactor building leak rate and details of the cover-gas system) that are not described in the FHR references, the analysts made assumptions needed to construct a MELCOR full-plant model. The FHR model uses a TRi-structural ISOtropic (TRISO)-particle fuel pebble-bed reactor with a primary system rejecting heat to two coiled tube air heat ex changers. Three passive direct reactor auxiliary cooling systems provide heat removal to supplement or replace the emergency secondary system heat removal during accident conditions. Surrounding the reactor vessel is a low volume reactor cavity that insulates the reactor with fire bricks and thick concrete walls. A refractory reactor liner system provides water cooling to reduce the concrete wall temperatures. Example calculations are performed to show the plant response and MELCOR capabilities to characterize a range of accident conditions. The accidents selected for evaluation consider a range of degraded and failed modes of operation for key safety functions providing reactivity control, the primary system decay heat removal and also a piping leak of the line to the coolant drain tank.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1975

Materials technology for an advanced space power nuclear reactor concept: Program summary

The results of a materials technology program for a long-life (50,000 hr), high-temperature (950 C coolant outlet), lithium-cooled, nuclear space power reactor concept are reviewed and discussed. Fabrication methods and compatibility and property data were developed for candidate materials for fuel pins and, to a lesser extent, for potential control systems, reflectors, reactor vessel and piping, and other reactor structural materials. The effects of selected materials variables on fuel pin irradiation performance were determined. The most promising materials for fuel pins were found to be 85 percent dense uranium mononitride (UN) fuel clad with tungsten-lined T-111 (Ta-8W-2Hf).

Gluyas, R. E.↗
DOE OSTI2018

CTF_

CTF is a thermal-hydraulic simulation tool designed for modeling the reactor vessel region of light water reactors. The code is capable of modeling both coolant and solid regions and includes a nuclear fuel rod model. Coolant is represented with a 9-equation (two phases plus droplets and non-condensable gas) model with closure relationships being provided for normal and accident conditions. Meshing is done using the subchannel approach, which models the coolant region as a collection of parallel 1-D flow paths that communicate laterally through "gaps". CTF also includes a collection of supporting software tools that include: Xml2ctf - A utility for creating native CTF input decks from a simplified input file; pyCTF - A collection of Python utilities used for post-processing CTF results, manipulating CTF input models, and testing differences between output files; CTFFuel - A separate front-end to the CTF fuel rod solver package; Multistate - A separate driver for CTF that communicates through the code coupling interface; and veraAPI - A c++ interface to the code.

Salko, Robert K↗
DOE OSTI2016

CFD Predictions of Severe Accident Natural Circulation Flows in a Combustion Engineering PWR

Computational fluid dynamics (CFD) is used to predict natural circulation flows between a reactor vessel upper plenum and a steam generator during specific severe accidents in a pressurized-water reactor (PWR). The predictions are based on a model for a simplified Combustion Engineering (CE) design. This work is part of a larger study to investigate the potential for severe accident induced steam generator (SG) tube failures in PWRs being conducted by the U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research. The CFD model builds upon the lessons learned from previous studies of a generic Westinghouse PWR under similar conditions. The CE SG is significantly larger than the Westinghouse design and geometric differences are expected to reduce the mixing and increase the potential thermal challenge to the SG tubes. The CFD predictions are utilized to improve the basis for simplified flow and mixing models applied in one-dimensional severe accident codes. Parameters are determined to support system code modeling of hot leg and SG inlet plenum mixing, hot leg and SG tube bundle flow rates, and tube inlet temperatures. This updated modeling uses boundary conditions obtained from system analysis code predictions of severe accident behavior in a CE PWR. Model refinements include an improved tube bundle representation and improvements to the thermal boundary conditions. Results indicate less mixing and higher tube inlet temperatures for the CE design compared to previous Westinghouse studies. (author)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Non-invasive in-situ imaging of interior of nuclear reactors

Techniques, systems, and devices are disclosed for non-invasive monitoring and imaging of nuclear fuel inside a nuclear reactor using muon detector arrays. In one aspect, these detector arrays are placed outside the reactor vessel or building for investigating the reactors without access to the cores, therefore the imaging process is non-invasive. In some implementation, these detector arrays measure both muon scattering and absorption to enable imaging and characterizing not only the very high-Z fuel materials, but also other materials in the reactor, thereby obtaining a more complete picture of reactor status. When applied to damaged reactors, the disclosed proposed techniques, systems, and devices, through the process of providing an image, can reveal the presence (or absence) of damage to fuel rod assemblies or puddles of molten fuel at the bottom of the containment vessel, thus providing crucial information to guide decisions about remedial actions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Progress in the development of the neutron flux monitoring system of the French GEN-IV SFR: simulations and experimental validations [ANIMMA--2015-IO-98]

The neutron flux monitoring system of the French GEN-IV sodium-cooled fast reactor will rely on high temperature fission chambers installed in the reactor vessel and capable of operating over a wide-range neutron flux. The definition of such a system is presented and the technological solutions are justified with the use of simulation and experimental results. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Experimental validation of an 8 element EMAT phased array probe for longitudinal wave generation

Sodium cooled Fast Reactors (SFR) use liquid sodium as a coolant. Liquid sodium being opaque, optical techniques cannot be applied to reactor vessel inspection. This makes it necessary to develop alternative ways of assessing the state of the structures immersed in the medium. Ultrasonic pressure waves are well suited for inspection tasks in this environment, especially using pulsed electromagnetic acoustic transducers (EMAT) that generate the ultrasound directly in the liquid sodium. The work carried out at CEA LIST is aimed at developing phased array EMAT probes conditioned for reactor use. The present work focuses on the experimental validation of a newly manufactured 8 element probe which was designed for beam forming imaging in a liquid sodium environment. A parametric study is carried out to determine the optimal setup of the magnetic assembly used in this probe. First laboratory tests on an aluminium block show that the probe has the required beam steering capabilities.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Experimental Study for Two Phase Flow Instability in the PCHE Type Steam Generator in SMR

Conventional U-tube type heat exchangers have some limitations when applied to Small Modular Reactor (SMRs). In particular, the size of steam generators is important in the SMR, as it greatly affects the size, and thus the cost, of the reactor vessel.. Therefore, the circuit heat exchanger (PCHE) is suggested as an alternative. PCHE, a compact type heat exchanger manufactured by HEATRIC, has several advantages such as the reduced size, wide operation range until 900? and 900 bar and high power density due to the large heat transfer area over volume ratio. Recently, it is newly suggested as an application of a steam generator. MIT and KAIST have already conceptually suggested PCHE as a steam generator. Under boiling condition, flow instability is one of the critical issues-resulting in unstable mass flow rate, sudden temperature change and system control failure. According to the flow mechanism, there are various types of instabilities in the heat exchanger and each flow tendency is different. Typical examples are Ledinegg instability of the static instability and pressure-drop oscillation and density-wave oscillation of the dynamic instability. The characteristic of Ledinegg instability is that the flow under unstable condition changes into the stable flow. The characteristic of dynamic instability is that the flow fluctuates continuously. Pressure-drop oscillation has the period of tens of seconds and the system pressure also highly oscillates with the period of flow instability. In comparison, the density-wave oscillation has the period of few seconds and the system pressure is relatively stable. Previous studies on flow instability are focused on the single channel flow and helical tube type heat exchanger. There has been very limited experimental research on flow instability using PCHE. In KAIST, the test facility was constructed to study the thermal hydraulics in PCHE. The objectives of our test facility are: (1) to identify the occurrence of instability under two phase and super-heated steam flow in PCHE, and (2) to examine whether it can operate normally using throttling effect. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Plasma nitriding monitoring reactor: A model reactor for studying plasma nitriding processes using an active screen

A laboratory scale plasma nitriding monitoring reactor (PLANIMOR) has been designed to study the basics of active screen plasma nitriding (ASPN) processes. PLANIMOR consists of a tube reactor vessel, made of borosilicate glass, enabling optical emission spectroscopy (OES) and infrared absorption spectroscopy. The linear setup of the electrode system of the reactor has the advantages to apply the diagnostic approaches on each part of the plasma process, separately. Furthermore, possible changes of the electrical field and of the heat generation, as they could appear in down-scaled cylindrical ASPN reactors, are avoided. PLANIMOR has been used for the nitriding of steel samples, achieving similar results as in an industrial scale ASPN reactor. A compact spectrometer using an external cavity quantum cascade laser combined with an optical multi-pass cell has been applied for the detection of molecular reaction products. This allowed the determination of the concentrations of four stable molecular species (CH{sub 4}, C{sub 2}H{sub 2}, HCN, and NH{sub 3}). With the help of OES, the rotational temperature of the screen plasma could be determined.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

The effects of stainless steel radial reflector on core reactivity for small modular reactor

Commercial PWR core is surrounded by a radial reflector, which consists of a baffle and water. Radial reflector is designed to reflect neutron back into the core region to improve the neutron efficiency of the reactor and to protect the reactor vessels from the embrittling effects caused by irradiation during power operation. Reflector also helps to flatten the neutron flux and power distributions in the reactor core. The conceptual nuclear design for boron-free small modular reactor (SMR) under development in Korea requires to have the cycle length of 4∼5 years, rated power of 180 MWth and enrichment less than 5 w/o. The aim of this paper is to analyze the effects of stainless steel radial reflector on the performance of the SMR using UO{sub 2} fuels. Three types of reflectors such as water, water/stainless steel 304 mixture and stainless steel 304 are selected to investigate the effect on core reactivity. Additionally, the thickness of stainless steel and double layer reflector type are also investigated. CASMO-4/SIMULATE-3 code system is used for this analysis. The results of analysis show that single layer stainless steel reflector is the most efficient reflector.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Nanoscale biochemical sample preparation and analysis

Provided herein are methods and systems for biochemical analysis, including compositions and methods for processing and analysis of small cell populations and biological samples (e.g., a robotically controlled chip-based nanodroplet platform). In particular aspects, the methods described herein can reduce total processing volumes from conventional volumes to nanoliter volumes within a single reactor vessel (e.g., within a single droplet reactor) while minimizing losses, such as due to sample evaporation.

Kelly, Ryan T.↗
DOE OSTI2015

Of the final end state for decommissioning of reactors at SRS: collaboration between stakeholders, regulators, and the Federal Government - 15279

The Savannah River Site (SRS) is an 802 square-kilometer (310 square-miles) United States Department of Energy (US DOE) nuclear facility located along the Savannah River near Aiken, South Carolina. Nuclear weapons material production began in the early 1950's, utilizing five (5) production reactors. In the early 1990's all SRS production reactor operations were terminated. The first reactor closure end-state declaration was documented in a Comprehensive Environmental Response and Compensation and Liability Act (CERCLA) Early Action Record of Decision. The decision for the final closure of the 29,543 square-meter (318,000 square-feet) 105-P Reactor was determined to be in situ decommissioning (ISD). ISD is an acceptable and cost-effective alternative to off-site disposal for the reactor building, which will allow for consolidation of remedial action wastes generated from other cleanup activities within the P Area. ISD is considered protective by the regulators, U.S. Environmental Protection Agency (USEPA) and the South Carolina Department of Health and Environmental Control (SCDHEC), public and stakeholders as waste materials are stabilized / immobilized, and radioactivity is allowed to naturally decay, thus preventing future exposure to the environment. Stakeholder buy-in was critical in the up-front planning in order to achieve this monumental final decision. Numerous public meetings and workshops were held in two different states (covering a 321 kilometer (200-mile) radius) with stakeholder and SRS Citizens Advisory Board (CAB) participation. These meetings were conducted over an eight- month period as the end-state decision making progressed. Information provided to the public evolved from workshop to workshop as data became available and public input from the public meetings were gathered. ISD is being considered for the balance of the four (4) SRS reactors and other hardened facilities such as the chemical Separation Facilities (canyons). Closure of the SRS P Reactor Area was precedent setting and sets the stage for closure of other reactors and hardened facilities. The DOE, USEPA, and SCDHEC engaged in active, up-front, and timely involvement with the SRS. The SRS CAB, elected officials, and citizens of South Carolina and Georgia met to discuss the P reactor area closure and associated risks. Early public involvement gave SRS and the regulators assurance that the selected decision was valid and would not be overturned. Public input improved the process and ensured that the important decision on the final end-state of the reactor facilities was not made in isolation. ISD of P Reactor will provide adequate protection of human health and the environment and the lowest cost since short-term risk is minimized to remedial workers from exposure to contaminated equipment and facilities by leaving the reactor vessel in place allowing for radiological decay over time. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Feedback of the Dutch Historical Waste Management Program - 17085

This paper provides a feedback on the Dutch historical waste management program. Over the past decades, the High Flux Reactor located in Petten in the province of North Holland has tested numerous types of materials and fuels irradiation, for both medical and nuclear energy applications. A number of waste containers was accumulated on site, where the storage facility was initially built. Following Dutch political decisions in the 1980's, the contents of the waste storage facility in Petten must be transferred to another location. The consequences of this decision included the transportation of waste containers from Petten to the new location, which is located in the South-West of the Netherlands. This project was initiated and is now running. It includes a complex structure and can be described as follows: each historical-waste container will be opened, its contents segregated in three categories depending on the dose rate emission, characterized and repacked in new containers suitable for transportation and storage at the new site. The approach chosen is based on the final dose rate emitted by the repacked containers, which are sub-divided into 3 categories.. The lowest dose-emitting containers are transported directly to the new storage facility. The 2 other categories are first transported to a third party to be super compacted and subsequently transported to the new storage facility. So far, the segregation method is based on Co-60 gamma emission and nuclide vectors associated to the different materials. A pre-sorting has been established to fit as much as possible IAEA recommendations. Historical waste has then to be classified in families sharing common history. Thus, a set of canisters containing the pieces of the old reactor vessel, which was replaced in 1984 was the first challenge, followed by a set of canisters containing residues of instruments only and pieces coming from the top of the control rods of the reactor which were not directly irradiated in the reactor. So far, the two first sets of canisters are being treated and characterized as well as possible. Reporting and official documents have been filled up to comply to Dutch transport regulations and to define waste acceptance criteria for the storage facility. The first transports of containers with Low Level Waste occurred in October 2015, and another transport took place at the end of 2016. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Demonstration of Passive Phoretic Decontamination by Accident Simulation in a Scaled Reactor Containment

Reactor containment performance is a critical defense-in- depth barrier for the prevention of radioactive release in the event of a nuclear accident. One of the functions of the containment is to assist in the decontamination of aerosolized radioactive particles through both 'passive' or natural processes and 'active' or mechanical processes. Current United States Nuclear Regulatory Commission (NRC) guidance (Reg. guide 1.183) allows the use of aerosol natural deposition correlations to quantify postaccident nuclear containment decontamination factors. Large Light Water Reactor (LLWR) correlations have previously been developed to quantify the contribution of natural phenomena to post-accident aerosol decontamination. Specific correlations must similarly be developed to support the safety analyses of Small Modular Reactors (SMRs). For LLWR plants, decontamination is achieved 'actively' by mechanical systems such as containment sprays and 'passively' via natural phenomena such as gravitational settling, condensation, impaction, thermophoresis, diffusiophoresis and hygroscopic effects. Integrated pressurized water reactor (iPWR) SMR designs do not have containment sprays; therefore, they rely primarily on passive natural phenomena for decontamination in the integrated reactor vessel. Analytical studies and inferences from LLWR test data suggest that the smaller containment vessel volume to surface area ratio, coupled with the thermal hydraulic properties of iPWR plants, would significantly enhance decontamination factors for these smaller designs. The research aims to quantify the amount of decontamination from natural phenomena for the range of iPWR physical and thermal hydraulic characteristics. A preliminary assessment has shown that diffusiophoresis, thermophoresis and hygroscopic effects are the major areas with a research gap for iPWR plants; hence these areas are the foci of the research. These first-of-a-kind experiments will be described and preliminary analytical results will be presented that show the expected effects of SMR-specific correlations on passive decontamination of aerosolized radionuclides.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Power peaking with small modular reactors to support deep penetration of renewable energy sources - 192

As inherently intermittent sources of renewable energy more fully penetrate the US grid, peaking power is largely being supplied by carbon-emitting natural gas turbines. Significant greenhouse gas emissions can be avoided if these plants are replaced with carbon-neutral nuclear facilities to provide peaking power to complement renewable generation and meet overall power demand. There is a great deal of previous work regarding reactor power shaping with control rod movement for both currently operating nuclear power plants and proposed plants, but the literature on load-following to meet less predictable, more rapidly varying power demand is less comprehensive. We have chosen the Westinghouse IRIS integral PWR as our candidate for modeling, simulation, and control studies. A nodal model of the IRIS integral PWR, partially adapted from a MATLAB/Simulink model constructed by prior students at the University of Tennessee in Knoxville, has been constructed in Modelica. Currently the model includes the reactor vessel and steam generator system and simple assumptions for the balance of plant. The control scheme for the load-following operation of an IRIS integral PWR model would ultimately lead to the development of real operational mechanisms and principles in a grid with significant renewables capacity. In the future the model will be integrated with sophisticated grid and hybrid energy systems. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗