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At least 235 records · Page 13
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

High-temperature steam oxidation study of irradiated FeCrAl defueled specimens

Post irradiation examinations (PIE) were performed on irradiated iron-chromium-aluminum (FeCrAl) specimens. These FeCrAl specimens were fabricated at the US Department of Energy's Oak Ridge National Laboratory (ORNL). The experimental setup involved subjecting FeCrAl cladding, along with UO 2 pellets, to irradiation in the Idaho National Laboratory Advanced Test Reactor (ATR). In parallel, the FeCrAl alloy tubing without UO 2 pellets was irradiated at ORNL's High Flux Isotope Reactor (HFIR). After irradiation, the ATR-irradiated rodlet was transported to an ORNL hot cell, where it was sectioned into multiple samples for the PIE and severe-accident testing. The sectioning process revealed that the fuel was not bonded to the cladding and could be easily detached from sectioned cladding slices. Microstructural analysis of the fuel cross sections demonstrated no significant interaction between the fuel and the cladding. Additionally, high-temperature steam oxidation tests on defueled cladding segments showed minimal oxygen uptake even at 1200 °C. Here, the ATR-irradiated specimens began to exhibit signs of enhanced oxidation upon reaching a temperature of 1300 °C. Furthermore, enhanced oxidation was observed on the inner surface of the ATR-irradiated FeCrAl specimen, which had been subjected to 1300 °C for a duration of 1 min. By contrast, high-temperature steam oxidation experiments indicated that the HFIR-irradiated FeCrAl cladding provided good thermal stability when exposed to 1300 °C for up to 4 h. Comparative analysis encompassing the oxidation behavior of the ATR-irradiated fueled FeCrAl, HFIR-irradiated unfueled FeCrAl, and unirradiated FeCrAl suggests that the fuel–cladding interaction, although not visible via standard microscale electron microscopy measurements, may accelerate the deterioration of FeCrAl cladding in beyond-design-basis accident scenarios.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

A Scaling-based Applicability Assessment of the CAER Integrated Systems Test Facility

This paper presents a scaling-based applicability assessment of the Integrated Systems Test (IST) facility located at the Center of Advanced Engineering Research in Forest, Virginia. The assessment draws upon both traditional and contemporary scaling methods, seeking an optimal representation of light water reactor (LWR) and small modular reactor (SMR) characteristics during design-basis events. Based on published LWR loss-of-coolant-accident phenomena identification and ranking tables compiled with a similar assessment completed by BWXT mPower, Inc., for the mPower{sup TM} SMR, the phenomena considered in the IST design include break flow, decay heat, loop thermal-hydraulics, and primary-to-secondary heat transfer. Among these, the single- and two-phase natural circulation and system depressurization, by association to loop thermal-hydraulics, were the principal challenges. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Estimation of LOCA break size using cascaded support vector regression - 334

A failure of emergency core cooling system (ECCS) during a certain small break loss of coolant accident (SBLOCA) can cause a severe core uncovery and fuel failure. The safety injection systems (SISs) might not function properly in case of a SBLOCA, due to the slight change of pressure in the pipes. Early trend of the major parameters has to be observed and the accurate information has to be given to the nuclear power plant (NPP) operators by precisely identifying the break size to effectively manage accidents. In this study, the objective is to provide to operators the information on the break size in a short time by considering the accident situations such as hot-leg LOCA, cold-leg LOCA, steam generator tube rupture (SGTR). A cascaded support vector regression (CSVR) is used in order to estimate the break size. The simulation data set was obtained from the optimized power plant reactor 1000 (OPR1000) using modular accident analysis program (MAAP) code. And genetic algorithm (GA) was used to select the input variables of the CSVR model and optimize its parameters. As a result of this study, the CSVR model estimate very well the break size of LOCAs. If the operators can predict the break size in the LOCA, they can response quickly and properly to LOCA circumstances to prevent the core uncovery and fuel failure. Also, it will be possible to more efficiently manage beyond design basis accidents. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Study on Water-ingress Accident during the Initial Core and the Running-in Phase of HTR-PM - 18429

Water ingress accident is one of the typical accidents with potential hazard to the high temperature gas-cooled reactor (HTGR), which will cause reactivity introduction, as well as the chemical corrosion of the graphite fuel elements and reflector structure material. Besides, increase of the primary pressure may result in the opening of the safety valves, consequently leading the release of radioactive isotopes. Therefore high attention has been paid to water ingress phenomenon during the design of the HTGR to verify its inherent safety characteristics. The online continuous discharge and multi-pass of the fuel elements is another important characteristics for the pebble-bed HTGR which can effectively increase the reactor operation availability. From the initial core, a long period of time, namely the running-in phase, is needed to reach the final long-term stable equilibrium core. During this period of time, the core states are totally different, including the number of the fuel elements, the fuel enrichment, the accumulation of the fission product, the temperature feedback coefficient of the reactivity, and so on. In this paper, based on the design of 250 MW Pebble-bed Module High Temperature gas-cooled Reactor (HTR-PM) and a candidate proposal for its initial core and running-in phase, considering three typical states during the running-in phase, the design basis accident of a double-ended guillotine break of one heating tube has been analyzed with the TINTE code, and the results are compared with that of the equilibrium core. Some important parameters during the transient process, including the power, the temperature of the fuel element, the pressure of the primary coolant, as well as the graphite corrosion rate, are studied. To a certain extent, the results can verity the design of the initial core and the running-in phase, and also further prove the safety feature of the HTR. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Evaluation of Flowing Salt Irradiation Facilities with High Neutron Flux

There is significant current interest in the commercial power industry in developing nuclear reactor concepts using molten salt as a coolant or as a fuel-bearing medium and coolant. The Molten Salt Reactor Experiment (MSRE), which was performed at Oak Ridge National Laboratory from 1964-1969, established the feasibility of this reactor design and performed some of the early work in qualifying appropriate structural materials and coolant salts. However, modern versions of this design often incorporate aspects that are beyond the design basis of the MSRE. For example, some designs intend to operate at a higher temperature, use a different structural material, or a different salt altogether. These new designs will eventually require irradiation testing in a variety of reactors and conditions. This document specifically evaluates the possibilities and requirements associated with the most difficult of these conditions: a flowing salt irradiation under high neutron flux.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Los Alamos National Laboratory SAVY-4000 Field Surveillance Plan Update for 2016

The Packaging Surveillance Program section of the Department of Energy (DOE) Manual 441.1-1, Nuclear Material Packaging Manual (DOE 2008), requires DOE contractors to “ensure that a surveillance program is established and implemented to ensure the nuclear material storage package continues to meet its design criteria.” The Los Alamos National Laboratory (LANL) SAVY-4000 Field Surveillance Plan was first issued in FY 2013 (Kelly et al. 2013). The surveillance plan is reviewed annually and updated as necessary based on SAVY-4000 surveillance and other surveillance findings, as well as results of the lifetime extension studies (Blair et al. 2012, Weis et al. 2015a). The LANL SAVY-4000 Field Surveillance Plan Update was issued in 2014 (Kelly et al. 2014). This 2016 update reflects changes to the surveillance plan resulting from restrictions on handling residue materials greater than 500 g, the addition of specific engineering judgment containers, and 2015 surveillance findings. The SAVY-4000 container has a design life of five years, which was chosen as a conservative estimate of the functional properties of the materials used in the construction of the SAVY 4000 when exposed to the potential insults including temperature, corrosive materials and gases, and radiation. The SAVY-4000 container design basis is described in a safety analysis report (Anderson et al. 2013). In the National Nuclear Security Administration's (NNSA’s) approval of the safety analysis report, it was recommended that the design life clock begin on March 2014 (Nez et al. 2014). However, it is expected that a technical basis can be developed to extend the design life of the SAVY-4000 containers to approximately 40 years (Blair et al. 2012, Weis et al. 2015a). This surveillance plan update covers five years (2015–2019) and is developed to ensure SAVY-4000 containers meet their design criteria over the current five-year design life and to gather data that can be used in developing the technical basis for a 40-year design life. The technical basis for the longer design life will be developed based on the results of this surveillance plan and the concurrent lifetime extension studies (Blair et al. 2012, Weis et al. 2015a). A long-term surveillance plan will be developed by the completion of this plan and will be based on the results of the first several years of this five-year surveillance plan and the lifetime extension studies. In addition, the current Hagan container population is so large (approximately 3,500 LANL-wide, approximately 3,000 in PF-4) that it is anticipated Hagan containers will be in use for many years to come. Therefore, this plan includes surveillance activities for the Hagan container system during the time it takes to replace the Hagan containers with manually-compliant containers such as the SAVY 4000. These Hagan examinations not only provide information to determine if additional Hagan surveillance is needed, but also support SAVY lifetime-extension studies because some of the container components are made from similar material (e.g., stainless steel and Viton).

42 ENGINEERING↗
DOE OSTI2017

Experimental and Computational Investigation of the Emission of Radioactive Products from Irradiated Fuel of Fast Sodium-Cooled Reactor

The key experimental and computational studies of the emission of long-lived radioactive fission products and actinides from uranium dioxide irradiated in BN-600 into sodium and the gas phase in the range 150–800°C, performed on the experimental bench at the Leipunskii Institute of Physics and Power Engineering, are presented. The results of measurements of the properties of aerosols of certain radioactive fission products formed upon heating of irradiated nuclear fuel in sodium are presented. The experimental results are of interest for evaluating the radiation impact of reactor facilities on workers and the public during beyond design basis accidents.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Evaluation of Radiation Impacts of Spent Nuclear Fuel Storage (SNFS-2) of Chernobyl NPP - 13495

Radiation effects are estimated for the operation of a new dry storage facility for spent nuclear fuel (SNFS-2) of Chernobyl NPP RBMK reactors. It is shown that radiation exposure during normal operation, design and beyond design basis accidents are minor and meet the criteria for safe use of radiation and nuclear facilities in Ukraine. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Enhanced Low-Enriched Uranium Fuel Element for the Advanced Test Reactor

Under the current US Department of Energy (DOE) policy and planning scenario, the Advanced Test Reactor (ATR) and its associated critical facility (ATRC) will be reconfigured to operate on low-enriched uranium (LEU) fuel. This effort has produced a conceptual design for an Enhanced LEU Fuel (ELF) element. This fuel features monolithic U-10Mo fuel foils and aluminum cladding separated by a thin zirconium barrier. As with previous iterations of the ELF design, radial power peaking is managed using different U-10Mo foil thicknesses in different plates of the element. The lead fuel element design, ELF Mk1A, features only three fuel meat thicknesses, a reduction from the previous iterations meant to simplify manufacturing. Evaluation of the ELF Mk1A fuel design against reactor performance requirements is ongoing, as are investigations of the impact of manufacturing uncertainty on safety margins. The element design has been evaluated in what are expected to be the most demanding design basis accident scenarios and has met all initial thermal-hydraulic criteria.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Role of Core Exit Axial Reflectors for NTP Full Submersion Criticality Accident Mitigation

One of the only design basis accidents currently codified into law for space nuclear systems is the full-submersion criticality accident. This accident scenario prescribes that the reactor core must remain sub-critical in the event of a full water submersion with the control system in the shut-down position. This accident scenario is of particular interest and relevance for nuclear thermal propulsion (NTP) reactors because of their unique core periphery and internal design considerations. In previous work it has been shown that this accident is particularly sever for low-enriched uranium (LEU) fueled NTP cores due to their moderated spectrum and the minimization of parasitic neutron absorption. These then add to the existing problems associated with the numerous voids within the core and the extensive neutron leakage due to its compact size. In this abstract we examine the role of a core exit axial reflector in minimizing the submersion worth of a graphite composite LEU-NTP core. The neutronics simulations were done using the ENDF/B-VII cross-section library in conjunction with MCNP6.1. The MCNP calculations were done using 20000 particles per cycle with 50 inactive and 300 active cycles. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

The control room radiological habitability assessment of high temperature reactor pebble bed module (HTR-PM) - 18465

The control room radiological habitability is important for staff safety in nuclear power plant accident, which is also a licensing requirement of the High Temperature Reactor Pebble Bed Module (HTR-PM) in China. In this study, a comprehensive evaluation of the control room radiological habitability is performed for three design basis accident scenario of HTR-PM. For each scenario, three exposure pathways are taken into account, which includes the immersion external exposure, the inhalation exposure and the direct exposure from radioactive cloud in the environment. The atmospheric concentration is calculated using the NRC-recommended ARCON96 code, based on the two-year meteorological observation data at HTR-PM s site and the final building layout. The evaluation result demonstrates that the individual dose level of staffs in the control room is far below the requirement of the regulatory, which guarantees the radiological habitability of control room of HTR-PM. The contribution of each radionuclide has been analyzed, which suggests that iodine and metal radionuclides such as Cs-137 and Co-60 are primary contributor of the total dose, which are worthy of particular attention in the control room habitability design. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

An Approach for Assessing Consequences of Potential Supply Chain and Insider Contributed Cyber Attacks on Nuclear Power Plants

The Stuxnet attack at the Natanz facility is an example of a targeted and successful cyber attack on a nuclear facility. Snowden's release of National Security Agency documents demonstrated the consequences of the insider threat. More recently, the United States tried to attack North Korea but failed, South Korea was attempting to attack North Korea, and both applied Stuxnet-like approaches. These sophisticated targeted attacks differ from web-site hacking events that are reported almost daily in the news mainly because targeted attacks require detailed design and operation information of the systems attacked and/or are often carried out by insiders. For instance, in order to minimize disruption of facilities around the world, Stuxnet remained idle until it recognized the specific configuration of the Natanz facility, demonstrating that the attackers possessed extremely detailed information about the facility. Such targeted cyber attacks could become a national-level military weapon and be used in coercion of hostile countries. While U.S. nuclear power plants (NPPs) are well designed and protected and not easily attacked through internet hacking, there have been a few digital-system-related incidents. For example, in 2003, the Slammer worm breached the private network of the Davis Besse plant and disabled a safety monitoring system for almost five hours. The breach did not pose a safety hazard because the plant was offline, but it demonstrates the vulnerability of NPP systems. The worm did not enter the plant systems directly; it began by entering the systems of a Davis Besse contractor, and entered through the T1 line bridging the contractor's computer and Davis Besse's corporate networks. The T1 line was not protected by the plant's firewall. The success of the Stuxnet-Natanz case demonstrates that supply-chain level attack (i.e., physically tampering with digital systems to install undetectable malware for the purpose of bringing harm to a player further down the supply chain network) plays a crucial role in a successful attack. Although attacking an NPP naturally requires more complex strategies and tactics than internet hacking, NPP's have a variety of sophisticated and possibly sensitive systems that may be attacked. Previous studies, for example, Baylon, have identified and confirmed that such attack strategies and tactics can be designed and could produce severe damage to an NPP. Current NPPs were designed according to design basis accidents that predate use of the internet by the general public and did not take into consideration potential cyber attacks that could be carried out by state-sponsored attackers and malicious insiders. To date, no detailed engineering analysis has been performed to realistically examine the consequences these attacks may have on the plants. In this paper, we discuss an approach for assessing the potential consequences of Stuxnet type of attacks. We will examine the engineering criteria used in designing the plants, develop possible attack scenarios in terms of the structures, components, and systems (SSCs) being controlled by digital systems at the NPPs, perform engineering analysis to determine if the attacks would impose conditions that are beyond design bases, and categorize the consequences of the potential accidents. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Metal Matrix Microencapsulated Fuel Technology for LWR Applications

An overview of the metal matrix microencapsulated (M3) fuel concept for the specific LWR application has been provided. Basic fuel properties and characteristics that aim to improve operational reliability, enlarge performance envelope, and enhance safety margins under design-basis accident scenarios are summarized. Fabrication of M3 rodlets with various coated fuel particles over a temperature range of 800-1300 C is discussed. Results from preliminary irradiation testing of LWR M3 rodlets with surrogate coated fuel particles are also reported.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Testing and Versatility of the SAVY 4000 Nuclear Material Container - 15634

The SAVY 4000 container was designed by Nuclear Filter Technology, Inc. to meet DOE M441.1-1 requirements for safe storage of nuclear material, and it is currently being used at Los Alamos National Laboratory, Livermore National Laboratory, Idaho National Laboratory and Nevada National Security Site. The 316L stainless steel container can be manufactured in a range of sizes from 1 quart to 10 gallon capacity, it can be opened and closed easily by hand, it is sealed with a Viton{sup R} O-ring, and the filter retains particulates and resists liquid water entry while allowing hydrogen gas to escape. This innovative design has proven to provide benefits and uses beyond its original intent. For example, the container was tested in a simulated facility design basis accident involving high temperatures, fuel fire, and subsequent drop, and the damage ratio was measured. This has allowed the LANL plutonium facility to take credit for a material-at-risk reduction of approximately two orders of magnitude for material stored in the SAVY 4000 container. In addition, the water resistant filter allows the nuclear facility to credit the container as a safety feature to mitigate the risk of a criticality accident in the event of facility sprinkler system activation. The container has also been approved to transfer small quantities of solid nuclear material by hand carry between Los Alamos facilities, and it has been approved and used as a waste container for transuranic waste destined for the Waste Isolation Pilot Plant. The 1 quart capacity container and the future 2 quart design both fit inside the 9977 shipping container, and this has facilitated the transport of nuclear material from Idaho National Laboratory to Nevada Nuclear Security Site, obviated the need for repackaging the material for storage upon arrival. Finally, Los Alamos has recently contracted with Nuclear Filter Technology, Inc. to perform testing to certify the SAVY 4000 container as a DOT Type A liquid transport container. This paper will provide an overview of the various testing regimes and describe the resulting versatility of the container for nuclear material containment. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

AEROSOL-LM/Na Aided Simulation of Fission Product Production and Transport in the First Loop of a Fast Reactor

The structure of models and a description of the AEROSOL-LM/Na module, which is intended for simulating the transport and behavior of impurities sodium coolant and fission product aerosols in a normal operating regime of a reactor and in beyond design basis accidents, are presented. The characteristic features of the module are: accounting for the basic behavior of dissolved impurities in the sodium coolant, simulation of the dynamics of multi-component and polydisperse aerosols in the gas cavity of the first loop of the reactor, functional coupling with other modules of the code (thermal hydraulic and fuel), and simulation of the behavior of fission products as a function of the running state of the reactor installation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Westinghouse Small Modular Reactor passive safety system response to postulated events

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor. This paper is part of a series of four describing the design and safety features of the Westinghouse SMR. This paper focuses in particular upon the passive safety features and the safety system response of the Westinghouse SMR. The Westinghouse SMR design incorporates many features to minimize the effects of, and in some cases eliminates the possibility of postulated accidents. The small size of the reactor and the low power density limits the potential consequences of an accident relative to a large plant. The integral design eliminates large loop piping, which significantly reduces the flow area of postulated loss of coolant accidents (LOCAs). The Westinghouse SMR containment is a high-pressure, compact design that normally operates at a partial vacuum. This facilitates heat removal from the containment during LOCA events. The containment is submerged in water which also aides the heat removal and provides an additional radionuclide filter. The Westinghouse SMR safety system design is passive, is based largely on the passive safety systems used in the AP1000{sup R} reactor, and provides mitigation of all design basis accidents without the need for AC electrical power for a period of seven days. Frequent faults, such as reactivity insertion events and loss of power events, are protected by first shutting down the nuclear reaction by inserting control rods, then providing cold, borated water through a passive, buoyancy-driven flow. Decay heat removal is provided using a layered approach that includes the passive removal of heat by the steam drum and independent passive heat removal system that transfers heat from the primary system to the environment. Less frequent faults such as loss of coolant accidents are mitigated by passive injection of a large quantity of water that is readily available inside containment. An automatic depressurization system is used to reduce the reactor pressure in a controlled manner to facilitate the passive injection. Long-term decay heat removal is accomplished using the passive heat removal systems augmented by heat transfer through the containment vessel to the environment. The passive injection systems are designed so that the fuel remains covered and effectively cooled throughout the event. Like during the frequent faults, the passive systems provide effective cooling without the need for ac power for seven days following the accident. Connections are available to add additional water to indefinitely cool the plant. The response of the safety systems of the Westinghouse SMR to various initiating faults has been examined. Among them, two accidents; an extended station blackout event, and a LOCA event have been evaluated to demonstrate how the plant will remain safe in the unlikely event that either should occur. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

Bounding Radionuclide Inventory and Accident Consequence Calculation for the 1L Target

A bounding radionuclide inventory for the tungsten of the Los Alamos Neutron Science Center (LANSCE) IL Target is calculated. Based on the bounding inventory, the dose resulting from the maximum credible incident (MCI) is calculated for the maximally exposed offsite individual (MEOl). The design basis accident involves tungsten target oxidation following a loss of cooling accident. Also calculated for the bounding radionuclide inventory is the ratio to the LANSCE inventory threshold for purposes of inventory control as described in the target inventory control policy. A bounding radionuclide inventory calculation for the lL Target was completed using the MCNPX and CINDER'90 codes. Continuous beam delivery at 200 {micro}A to 2500 mA{center_dot}h was assumed. The total calculated activity following this irradiation period is 205,000 Ci. The dose to the MEOI from the MCI is 213 mrem for the bounding inventory. The LANSCE inventory control threshold ratio is 132.

Accelerator Design, Technology, and Operations↗
DOE OSTI2017

Hydride Microstructure at the Metal-Oxide Interface of a Zircaloy-4 Fuel Clad from the H.B. Robinson Nuclear Reactor

During nuclear reactor operation, some of the hydrogen that is produced by the water side corrosion is uptaken by cladding. When the uptaken hydrogen content exceeds its solubility limit in the cladding at a specific temperature, it precipitates as hydride particles in the cladding. This process occurs immediately at the metal-oxide interface forming a hydride rim because of the temperature gradient in the cladding. The presence of hydride rim and its microstructure severely reduces the cladding strength and can induce crack initiation and propagation during design basis accidents, transportation, and wet/dry storage. Because hydrides platelets are one of the potent crack initiation sites, microstructural characterization of hydride rim becomes essential to the safe operation of nuclear reactor and spent nuclear fuel retrievability. Therefore, we investigate the hydride rim microstructure at the metal oxide interface using scanning transmission electron microscopy (S/TEM) with Energy Dispersive X-ray Spectroscopy (EDS) that allows for the microstructural and local chemistry characterization of hydride precipitates to be determined, and enables a relationship to be derived on how the precipitate sites are affected by the oxide and their complex stacking behavior of the precipitates.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗