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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 955 records · Page 53
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Development of 3D pseudo pin-by-pin calculation methodology in ANC

Advanced cores and fuel assembly designs have been developed to improve operational flexibility, economic performance and further enhance safety features of nuclear power plants. The simulation of these new designs, along with strong heterogeneous fuel loading, have brought new challenges to the reactor physics methodologies currently employed in the industrial codes for core analyses. Control rod insertion during normal operation is one operational feature in the AP1000{sup R} plant of Westinghouse next generation Pressurized Water Reactor (PWR) design. This design improves its operational flexibility and efficiency but significantly challenges the conventional reactor physics methods, especially in pin power calculations. The mixture loading of fuel assemblies with significant neutron spectrums causes a strong interaction between different fuel assembly types that is not fully captured with the current core design codes. To overcome the weaknesses of the conventional methods, Westinghouse has developed a state-of-the-art 3D Pin-by-Pin Calculation Methodology (P3C) and successfully implemented in the Westinghouse core design code ANC. The new methodology has been qualified and licensed for pin power prediction. The 3D P3C methodology along with its application and validation will be discussed in the paper. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI1995

Nuclear Safety [Vol. 36, No. 1, January-June 1995]

Nuclear Safety is a journal that covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities. Table of Contents for this issue follows. THE CHORNOBYL ACCIDENT: 1 The Chornobyl Accident Revisited, Part II: The State of the Nuclear Fuel Located Within the Chornobyl Sarcophagus, A A. Borovoi and A. R. Sich; GENERAL SAFETY CONSIDERATIONS: 33 Nuclear Power Safety in Central and Eastern Europe, R. Wilson; 46 Safety of Nuclear Power Reactors in the Former Eastern European Countries, S. Chakraborty; 53 Technical Note: On the Definition of Common-Cause Failures, H. Paula; ACCIDENT ANALYSIS: 58 Modeling and Analysis of Core-Debris Recriticality During Hypothetical Severe Accidents in the Advanced Neutron Source Reactor, S.-H. Kim, V. Georgevich, D. B. Simpson, C. O. Slater, and R. P. Taleyarkhan; 68 Ignitability of Hydrogen/Oxygen/Diluent Mixtures in the Presence of Hot Surfaces, R. K. Kumar and G. W. Koroll; 94 Coupled RELAP5 and CONTAIN Accident Analysis Using PVM, K. A. Smith, A. J. Baratta, and G. E. Robinson; CONTROL AND INSTRUMENTATION: 109 Application of Fuzzy Logic in Nuclear Reactor Control Part I: An Assessment of State-of-the-Art, A. S. Heger, N. K. Alang-Rashid, and M. Jamshidi; DESIGN FEATURES: 122 Twenty-Third DOE/NRC Nuclear Air-Cleaning and Treatment Conference, R. R. Bellamy, J. J. Hayes, and M. W. First; ENVIRONMENTAL EFFECTS: 135 Atmospheric Dispersion and the Radiological Consequences of Normal Airborne Effluents from a Nuclear Power Plant, D. Fang, C. Z. Sun, and L. Yang; 142 Calculation of Distribution Coefficients for Radionuclides in Soils and Sediments, I. Puigdomenech and U. Bergstrom: OPERATING EXPERIENCES: 155 Reactor Shutdown Experience, Compiled by J. W. Cletcher; U.S. NUCLEAR REGULATORY COMMISSION INFORMATION AND ANALYSES: 158 Operating Experience Feedback Report—Reliability of Safety-Related Steam Turbine-Driven Standby Pumps Used in U.S. Commercial Nuclear Power Plants, J. R. Boardman; 166 Turbine Building Hazards, H. L Ornstein; RECENT DEVELOPMENTS: 169 Reports, Standards, and Safety Guides, D. S. Queener; 175 Proposed Rule Changes as of Dec. 31,1994; ANNOUNCEMENTS: 32 Harvard School of Public Health In-Place Filter Testing Workshop; 134 International Conference on Advances in the Operational Safety of Nuclear Power Plants; 193 30th Tennessee Industries Week; 193 DOE Technical Standards Program 1995 Workshop; 194 Multiphase Flow Experiments and Instrumentation; 180 The Authors; 185 Indexes to Nuclear Safety, Volumes 34 and 35.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

TCR Input to NUREG-1537 Process for Advanced Nuclear Technologies Derived from Additive Manufacturing

There has been a renewed interest by several advanced reactor developers to use NUREG-1537 “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors” as a basis for their safety analysis report content and organization. Recently, SHINE Medical Technologies, LLC (SHINE), which is a non-power, Aqueous Homogenous Reactor design radioisotope production facility, received a construction permit based around their NUREG-1537 safety evaluation report (ADAMS No. ML16229A140). Advanced reactor developers are interested in using NUREG-1537 as a basis for their safety analysis report content and organization because of its successful application towards research reactors, graded approach, and simplicity in structure and requirements. However, NUREG-1537 is still largely geared toward light water reactors (LWRs) and many improvements could be made or supported through guidance documents for advanced reactors. For nuclear power to play a role in the future zero-carbon energy portfolio, a supportive regulatory structure is needed to lower regulatory uncertainty and barriers to deployment. At the time of this report, no such document or pathway exists for advanced nuclear technologies, including those derived from nontraditional technology such as advanced manufacturing technology (AMT) and, specifically, additive manufacturing. This report will explore and provide recommendations as to how advanced nuclear technologies derived from additive manufacturing technologies could employ the use of an ISG, other guidance document, or revisions to NUREG-1537 to lower the regulatory uncertainty and barriers for adoption.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Neutronic feasibility analysis of fully ceramic microencapsulated fuel for commercial PWRs

The fully ceramic microencapsulated (FCM) fuel is based on tri-structural isotropic (TRISO) coated particles embedded in silicon carbide (SiC) matrix, which is among the accident tolerant fuel concept options. With the strength of high fission product retention, the FCM fuel provides potential superior safety performance. Compared to the conventional standard UO{sub 2} fuel, the deficiency of FCM fuel is its considerable lower fissile inventory, which results in a reduced reactor power or/and life. Supposing the FCM fuel is used as direct replacement of the fuel in commercial PWRs, the spectrum is softened due to the usage of SiC as matrix, and the moderator temperature coefficient is possible to be positive at the beginning of life, leading to high risk during operation. A neutronic feasibility study of FCM fuel utilized in commercial PWRs is performed in the paper. The fuel enrichment and under-moderated lattice are evaluated by analyzing the typical TRISO particle with UO{sub 2} kernels which is widely used in high temperature gas reactors and that with enlarged UN fuel kernels. It was found that under the mandatory constraint of low enriched uranium, the FCM fuel with UO{sub 2} kernels, although relatively mature, is only capable of small reactor designs, and the assembly lattice needs to be modified. The FCM fuel with UN kernels, when directly loading in the commercial PWRs, can yield comparable reactor power and life; however, a higher fuel enrichment and the fuel design and fabrication need to be optimized. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Enhanced micro-pocket fission detector for high temperature reactor evaluations - 259

A collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Atomic Energy Agency, Commissariat a l'energie Atomique et aux Energies Alternatives, (CEA), has been initiated by the Nuclear Energy Enabling Technologies (NEET) Advanced Sensors and Instrumentation (ASI) program for developing and testing High Temperature Micro-Pocket Fission Detectors (HT MPFD), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package for temperatures up to 800 deg. C. The MPFD technology utilizes small, multi-purpose, robust, in-core fission chambers and can be configured with a thermocouple. The small size, variable sensitivity, and increased accuracy of the MPFD technology represent a revolutionary improvement over current methods used to support irradiations in US Material Test Reactors (MTRs). Previous research conducted through NEET ASI has shown that the MPFD technology could be made robust and was successfully tested in a reactor core. In addition, the accomplishments of this project have attracted independent funding from other Department of Energy Office of Nuclear Energy (DOE-NE) programs for MTR irradiations of the MPFD technology. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

A Reduced-Order Model of a Nuclear Power Plant with Thermal Power Dispatch

This paper presents reduced-order modeling of thermal power dispatch (TPD) from a pressurized water reactor (PWR) for providing heat to nearby heat consuming industrial processes that seek to take advantage of nuclear heat to reduce carbon emissions. The reactor model includes the neutronics of the reactor core, thermal–hydraulics of the primary coolant cycle, and a three-lump model of the steam generator (SG). The secondary coolant cycle is represented with quasi-steady state mass and energy balance equations. The secondary cycle consists of a steam extraction system, high-pressure and low-pressure turbines, moisture separator and reheater, high-pressure and low-pressure feedwater heaters, deaerator, feedwater and condensate pumps, and a condenser. The steam produced by the SG is distributed between the turbines and the extraction steam line (XSL) that delivers steam to nearby industrial processes, such as production of clean hydrogen. The reduced-order simulator is verified by comparing predictions with results from separate validated steady-state and transient full-scope PWR simulators for TPD levels between 0% and 70% of the rated reactor power. All simulators indicate that the flow rate of steam in the main steam line and turbine systems decrease with increasing TPD, which causes a reduction in PWR electric power generation. The results are analyzed to assess the impact of TPD on system efficiency and feedwater flow control. Due to the simplicity of the proposed reduced-order model, it can be scaled to represent a PWR of any size with a few parametric changes. In the future, the proposed reduced-order model will be integrated into a power system model in a digital real-time simulator (DRTS) and physical hardware-in-the-loop simulations.

08 HYDROGEN↗
DOE OSTI2012

Role of research reactors in training of NPP personnel with special focus on training reactor VR-1

Research reactors play an important role in providing key personnel of nuclear power plants a hands-on experience from operation and experiments at nuclear facilities. Training of NPP (Nuclear Power Plant) staff is usually deeply theoretical with an extensive utilisation of simulators and computer visualisation. But a direct sensing of the reactor response to various actions can only improve the personnel awareness of important aspects of reactor operation. Training Reactor VR-1 and its utilization for training of NPP operators and other professionals from Czech Republic and Slovakia is described. Typical experimental exercises and good practices in organization of a training program are demonstrated. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Machine learning for reactor power monitoring with limited labeled data

Real-time reactor power monitoring is critical for a variety of nuclear applications, spanning safety, security, operations, and maintenance. While machine learning methods have shown promise in monitoring reactor power levels, there is limited research on their efficacy in label-starved environments. The goal of this work is to assess the feasibility of classifying nuclear reactor power level using multisource data in scenarios with limited labels. Data were collected using low-resolution multisensors at four nuclear reactor facilities: two large research reactors and two TRIGA reactors. Within each pair, one reactor dataset served as the source and the other as the target in a transfer learning paradigm. Twenty-three supervised models were trained on labeled sequences of magnetic field and acceleration data from each of the target sites. Self-learning and transfer learning methods were applied to the top performing models to assess their classification performance with increasing amounts of labeled data. While reactor power level classification was achieved with a Matthews Correlation Coefficient of up to 0.739 ± 0.003 and 0.622 ± 0.009 with only 400 sequences per power state for the large research reactor and TRIGA target sites, respectively, self-learning and transfer learning leveraging source site data did not improve target classification performance. These findings suggest that alternative methods, such as higher sensitivity sensors, digital twins, or the use of physics-informed models, are required to enable high-performance classification in machine learning approaches to reactor monitoring with a dearth of target ground truth.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Hierarchical Conceptual Model Event and Activity Domains for Forecasting State-Sponsored Civil Nuclear Power Activities

In FY20, the Savannah River National Laboratory and the Sanghani Center for Artificial Intelligence and Data Analytics (Virginia Tech) entered a collaboration funded by the Department of Energy’s Office of Defense Nuclear Nonproliferation Research and Development to develop a machine learning based modeling pipeline to extract proliferation events of interest from open data sources. The prototype modeling pipeline that was developed relies on the use of time dependent word embedding models to identify contextual shifts in key words and phrases that act as indicators of events of interest. The FY20- 21 efforts were focused on a narrow topical domain of forecasting “fissile core fabrication” at the Savannah River Site prior to its official announcement in 2018. In FY22, the research team was funded to continue development of the modeling pipeline by applying it to the problem of forecasting new, and/or significant changes to existing, civil nuclear power reactors around the world. In this effort, the development will focus on proving applicability to a broader topical domain and in data environments that may contain more sparse information, relative to the United States. In this report, the worldwide landscape of civil nuclear reactors is outlined, the timeline of interest is defined, and a hierarchical conceptual model is established to identify the activity domains of interest and the event domains of interest. This preliminary effort will guide the curation of a glossary of key terms for data acquisition, as well as the downstream modeling efforts.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Special Purpose Application Reactors: Systems Integration Decision Support

Microreactors, also known as very-Small Modular Reactors (vSMRs) or Special Purpose Reactors (SPRs), are being considered for use in unique applications where other methods of megawatt level energy production are uneconomical or unavailable. For the purposes of this report, a microreactor is defined as meeting the following criteria: • Factory manufacturable • Easily transportable by truck, plane, train, and/or ship • Produce no more than 20 megawatt thermal (MWth) energy, in order to qualify as Hazard Category 2 under 10CFR830 • Maintains neutronic simplicity, allowing semi- or fully-autonomous operation In the United States (U.S.), there are two general types of microreactors under development: high-temperature gas reactors (HTGRs) and heat pipe reactors. These two reactor types may, due to their differences in working fluid temperatures, be more effectively suited to different use cases. Additionally, they will require different interfacing components depending on their intended use. This report identifies a broad range of use cases for micro-reactors. Additionally, multiple reactor technologies are examined with an emphasis on the reactor sub-systems and interconnections. Using this information, a decision framework is developed to guide future efforts.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Summary on the depressurization from supercritical pressure conditions

When a fluid discharges from a high pressure and temperature system, a 'choking' or critical condition occurs, and the flow rate becomes independent of the downstream pressure. During a postulated loss of coolant accident (LOCA) of a water reactor the break flow will be subject to this condition. An accurate estimation of the critical flow rate is important for the evaluation of the reactor safety, because this flow rate controls the loss of coolant inventory and energy from the system, and thus has a significant effect on the accident consequences[1]. In the design of safety systems for a super critical water reactor (SCWR), postulated LOCA transients are particularly important due to the lower coolant inventory compared to a typical PWR for the same power output. This lower coolant inventory would result in a faster transient response of the SCWR, and hence accurate prediction of the critical discharge is mandatory. Under potential two-phase conditions critical flow is dominated by the vapor content or quality of the vapor, which is closely related with the onset of vaporization and the interfacial interaction between phases [2]. This presents a major challenge for the estimation of the flow rate due to the lack of the knowledge of those processes, especially under the conditions of interest for the SCWR. According to the limited data of supercritical fluids, the critical flows at conditions above the pseudo-critical point seem to be fairly stable and consistent with the subcritical homogeneous equilibrium model (HEM) model predictions, while having a lower flow rate than those in the two-phase region. Thus the major difficulty in the prediction of the depressurization flow rates remains in the region where two phases co-exist at the top of the vapor dome. In this region, the flow rate is strongly affected by the nozzle geometry and tends to be unstable. Various models for this region have been developed with different assumptions, e.g. the HEM and Moody model [3], and the Henry-Fauske non-equilibrium model [4], and are currently used in subcritical pressure reactor safety design[5]. It appears that some of these models could be reasonably extended to above the thermodynamic pseudo-critical point. The more stable and lower discharge flow rates observed in conditions above the pseudo-critical point suggests that even though SCWR's have a smaller coolant inventory, the safety implications of a LOCA and the subsequent depressurization may not be as severe as expected, this however needs to be confirmed by a rigorous evaluation of the particular event and further evaluation of the critical flow rate. This paper will summarize activities on critical flow models, experimental data and numerical modeling during blowdown from supercritical pressure conditions under the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on 'Heat Transfer Behaviour and Thermo-hydraulics Code testing for SCWRs'. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Testing of LWR Surveillance Materials from the High Fluence UCSB ATR-2 Irradiation

The reactor pressure vessel (RPV) in a light water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations that govern the operation of commercial nuclear power plants require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including pressurized thermal shock (PTS) in pressurized water reactors (PWR). In the irradiated condition, however, the fracture toughness of the RPV may be severely degraded, with the degree of toughness loss dependent on the radiation sensitivity of the materials. Of the many significant issues discussed regarding irradiation effects, the issue considered to have the most impact on the current regulatory process is that associated with effects of neutron irradiation on RPV steels at high fluence, for long irradiation times, and as affected by neutron flux. As also stated in a companion paper by Odette, et. al., the primary objective of the DOE-sponsored Light Water Reactor Sustainability (LWRS) Program RPV task is to develop robust predictions of transition temperature shifts (TTS) at high fluence (φt) to at least 10{sup 20} n/cm{sup 2} (>1 MeV) pertinent to plant operation of some pressurized water reactors (PWR) for 80 full power years. While understanding of copper-enriched precipitates (CRPs) have been fully developed, the discovery and experimental verification of 'late blooming' Mn-Ni-enriched precipitates (MNPs) with little to no copper for nucleation has stimulated research efforts to understand the evolution of these phases. The new and existing databases will be combined to support development of physically based models of TTS for high fluence-low flux (φ < 10 {sup 11}n/cm{sup 2}-s) conditions, beyond the existing surveillance database, to neutron fluences of at least 10{sup 20} n/cm{sup 2} (>1 MeV). The RPV task of the LWRS Program has worked with various organizations to obtain archival surveillance materials from commercial nuclear power plants to allow for comparisons of the irradiation-induced microstructural features from reactor surveillance materials with those from similar alloys irradiated under high flux conditions in test reactors, such as the UCSB ATR-2 experiment conducted at the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Development and Verification of MATLAB Functions for Pressure Drop Estimation in Crossflow Tube Bundles

The experimental data on drag coefficients in crossflow tube bundles is severely limited. It is largely contained in nomograms for a small range of tube bundle pitch ratios for both in-line and staggered bundles. These monograms and the corresponding correlations/curve fits are not readily accessible. This lack of experimental data presents a limitation when designing crossflow heat exchangers for nuclear reactors. Specifically, Molten Salt Reactors (MSRs) that utilize cross-flow configurations with turbulent molten salts flowing in the shell side are of interest. Current methods of calculating the drag coefficients are outdated and time consuming when conducting design studies (i.e. picking points off a graph). A computer script could easily run the necessary rapid calculations of the pressure drop for a large possible range of Reynolds numbers, pitch ratios and tube bundle configurations. In an effort to facilitate increased access and accuracy to crossflow pressure drop calculations, a set of MATLAB scripts have been created. These are specifically for the pressure drop and drag coefficients for flow over rows of in-line and staggered tube bundles. For verification of the scripts, the nomograms produced by Zukauskas, of which the scripts are built off, served as a comparison to the plot generated by the MATLAB scripts. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Irradiation Demonstration Element Design Parameters for MURR LEU U-Mo Fuel Conversion

This report contains the results of reactor design and performance calculations for conversion of the University of Missouri Research Reactor (MURR) from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the U.S. Department of Energy National Nuclear Security Administration (NNSA)Office of Material Management and Minimization(M3)Reactor Conversion Program at the Argonne National Laboratory and the MURR Facility.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Preliminary Design of Nuclear Reactor Heat Delivery Systems: Integration with Reference Oil Refinery, Methanol Synthesis, and Hydrogen Production

The Department of Energy’s (DOE) Integrated Energy Systems (IES) program is generating comprehensive analyses validating the opportunity for using nuclear energy in a variety of applications including future clean grids, providing heat for direct use, and providing heat to help reduce emissions in chemical commodities. This work focuses on the preliminary designs of thermal delivery systems that can integrate heat produced from a nuclear core to industrial processes. The key research question that needs to be answered is: what is the prospective method for integrating nuclear generated heat energy into non-electric applications that can facilitate combined heat and power operations by advanced nuclear reactor systems? This research is a composition of case studies showing preliminary conceptual designs for thermal delivery systems integrating advanced nuclear systems with a few industrial systems including high temperature steam electrolysis, a reference oil refinery, and potential future methanol systems that supplant some natural gas use with nuclear energy. Piping and instrumentation diagrams have been developed to show the conceptual integration of nuclear systems with representative industrial systems. Different features of the configurations are dependent on the specific integration requirements including energy source conditions, demand quantity, and require energy application conditions. Design concepts are validated using thermodynamic balance calculations to verify system performance including calculating system losses during transport. Key components: pumps and compressors, heat exchangers, and network piping are reported with key design information and sizing.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Multidimensional simulations of Mckenna-driven flow tube configuration: Investigating non-ideality in NO x formation flow tube experiments

Multidimensional simulations have been conducted to simulate atmospheric pressure, flat-flame/McKenna-burner-driven-flow tube experiments targeted to obtain NO x speciation data for predicting/analyzing syngas combustion emissions. In a prior work, we demonstrated the impacts of multidimensional transport on post flame region prediction departures from those assuming unidimensional flow/transport conditions. In this work, we develop and utilize a multidimensional laminar reacting flow solver to simulate the fully coupled flame and post flame regions to further elucidate the impacts of the earlier unidimensional modeling assumptions on interpreting post flame NO x experimental data. The model is used to simulate a lean, premixed syngas/air flame and its associated post flame regions within a cylindrical flow-tube-like arrangement. The combustion process takes place under atmospheric condition with trace amount of NO x seeding fed into the inlet gas stream. The spatial evolution of NO x species (NO and NO 2 ) in the flame and in the post-combustion zone suggests two distinct regions: 1) a region encompassing the flame structure itself; and 2) a post flame region in which the temperature decays due to both axial and radial transport processes. The predictions show that for the conditions studied, a pulsatile flow field exists due to the formation of an expanding and contracting recirculation zone in the outer periphery of the flow tube. By resolving the nature of the flow, the resulting time-averaged temperature and species concentrations show improved agreement with existing experimental measurements. The flow-field interaction results in radial inhomogeneities in the NO 2 profiles with the maximum concentration offset from the flow centerline. The location of the peak in NO 2 is coupled with radial temperature gradients from wall cooling effects and their significant influence on NO/NO 2 interconversion kinetics, producing notable NO 2 accumulation in regions near the wall. Geometrical configurations capable of suppressing/minimizing the pulsatile nature are also investigated and the results are compared. Other experimental configurations could be considered in parametric simulations to determine the optimal configuration that would minimize non-idealities in the observations. The work shows the value in performing such computations in advance of settling on a particular design for flow tube/flow reactor experiments.

42 ENGINEERING↗
DOE OSTI2017

Corrigendum to “Accelerated materials evaluation for nuclear applications” [J. Nucl. Mater. 488 (2017) 46–62]

The original article addresses the opportunities and complexities of using materials test reactors with high neutron fluxes to perform accelerated studies of material aging in power reactors operating at lower neutron fluxes and with different neutron flux spectra. Radiation damage and gas production in different reactors have been compared using the code, SPECTER. This code provides a common standard from which to compare neutron damage data generated by different research groups using a variety of reactors. This Corrigendum identifies a few typographical errors. Tables 2 and 3 are included in revised form.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Microstructural evaluation of a Fe-12Cr nanostructured ferritic alloy designed for impurity sequestration

Fast reactor fuel cladding candidate materials require proficiency in extreme environments consisting of high temperatures and irradiation doses in excess of 150 displacements per atom (dpa). Nanostructured oxide dispersion strengthened (ODS) alloys have been developed extensively for this purpose due to their notable high temperature strength, creep resistance, and irradiation resistance. However, their properties can deteriorate if interstitial impurities such as C and N are not well controlled during the fabrication process. A new Fe-12Cr nanostructured ODS alloy OFRAC (Oak Ridge Fast Reactor Advanced Fuel Cladding) with solute additions of Mo, Ti, and Nb has been developed to provide the desired properties mentioned above while simultaneously sequestering impurities within the matrix. After extrusion at 850°C, the as-extruded microstructure consists of an average 490 nm grain size and a high number density (6.8 × 10 23 m -3 ) of 2.2 nm diameter (Y,Ti,O) nanoclusters distributed homogeneously in the matrix. Atom probe tomography investigations suggest non-stochiometric compositions for the smallest nanoclusters. In addition, a second population of nanometer scale (Nb,Ti) rich carbonitrides is also present in the microstructure that captures the potentially detrimental C and N impurity atoms present in the matrix. Atom probe tomography results indicate elemental segregation of Cr, Mo, and Nb to grain boundaries in the as-extruded material, consistent with previous investigations of solid solution strengthening by solute additions. In conclusion, the ability of OFRAC to sequester impurities introduced from the powder metallurgical approach to nanostructured ferritic alloy development, compounded with its beneficial mechanical properties, makes this alloy a competitive candidate for fast reactor applications.

36 MATERIALS SCIENCE↗