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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 451 records · Page 25
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

On the novel 3-D neutron transport kinetic tRAPID algorithm and its validation

The Real-time Analysis for Particle-transport and In-situ Detection (RAPID) Code System, based on the Multi-stage Response-function Transport (MRT) methodology, allows for real-time simulation of nuclear systems based on 3-D continuous-energy particle transport. RAPID's steady-state (criticality) neutron transport algorithm is based on the Fission Matrix (FM) method, and has been extensively verified and validated against computational benchmarks and experiments. This paper introduces the novel 3-D time-dependent transport algorithm that has been implemented into the code, tRAPID, and its validation using the JSI TRIGA Mark-II reactor. tRAPID accurately and efficiently calculates neutron kinetics parameters (such as β{sub eff}, l{sub eff} , Λ, α{sub Rossi}) and 3-D time-dependent neutron fission source distribution and neutron importances for both prompt and delayed neutrons. tRAPID is used to simulate a rod insertion experiment performed at the JSI TRIGA Mark-II reactor, during which signals from four fission chambers at four different locations in the core were collected. The results demonstrate how tRAPID is capable of calculating detailed and accurate results with only a minimal use computational resources and time.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

A custom 3D printed paddlewheel improves growth in flat panel photobioreactor

One of the main challenges with using flat panel photobioreactors for algal growth is uneven mixing and settling of cells in corners, especially when bubbling is the only method used for mixing. Here, in order to improve mixing in our flat panel reactor, we designed a custom paddlewheel. Paddlewheels are frequently used in outdoor algae raceway ponds to improve mixing and we are taking advantage of the same principle for mixing in the reactor. The paddlewheel is easily integrated into our PSI FMT150 1‐L flat panel photobioreactor and is printed on a 3D printer using high temperature poly lactic acid (HT‐PLA). With the inclusion of an annealing step, the paddlewheel is autoclavable. Addition of the paddlewheel in the reactor minimized cell settling and improved algal growth, as evidenced by a nearly 40% increase in oxygen production rates. Nutrient dispersion and utilization in the culture was also improved as evidenced by a corresponding 38% decrease in CO 2 concentration. The paddlewheel device presented here is a cost‐effective method for improving algal growth in a flat panel photobioreactor.

09 BIOMASS FUELS↗
DOE OSTI2011

Treatment of municipal landfill leachate by catalytic wet air oxidation: Assessment of the role of operating parameters by factorial design

Highlights: > Landfill leachates can be treated effectively by catalytic wet oxidation. > Addition of H{sub 2}O{sub 2} in the presence of transition metals promotes degradation. > Factorial design evaluates the statistically significant operating conditions. > H{sub 2}O{sub 2}, reaction time and temperature are critical in determining performance. - Abstract: The wet air oxidation (WAO) of municipal landfill leachate catalyzed by cupric ions and promoted by hydrogen peroxide was investigated. The effect of operating conditions such as WAO treatment time (15-30 min), temperature (160-200 deg. C), Cu{sup 2+} concentration (250-750 mg L{sup -1}) and H{sub 2}O{sub 2} concentration (0-1500 mg L{sup -1}) on chemical oxygen demand (COD) removal was investigated by factorial design considering a two-stage, sequential process comprising the heating-up of the reactor and the actual WAO. The leachate, at an initial COD of 4920 mg L{sup -1}, was acidified to pH 3 leading to 31% COD decrease presumably due to the coagulation/precipitation of colloidal and other organic matter. During the 45 min long heating-up period of the WAO reactor under an inert atmosphere, COD removal values up to 35% (based on the initial COD value) were recorded as a result of the catalytic decomposition of H{sub 2}O{sub 2} to reactive hydroxyl radicals. WAO at 2.5 MPa oxygen partial pressure advanced treatment further; for example, 22 min of oxidation at 200 deg. C, 250 mg L{sup -1} Cu{sup 2+} and 0-1500 mg L{sup -1} H{sub 2}O{sub 2} resulted in an overall (i.e. including acidification and heating-up) COD reduction of 78%. Amongst the operating variables in question, temperature had the strongest influence on both the heating-up and WAO stages, while H{sub 2}O{sub 2} concentration strongly affected the former and reaction time the latter. Nonetheless, the effects of temperature and H{sub 2}O{sub 2} concentration were found to depend on the concentration levels of catalyst as suggested by the significance of their 3rd order interaction term.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS1998

Biodegradation of Fresh vs. Oven-Dried Inedible Crop Residue in a Continuously Stirred Tank Reactor

The degradation of soluble organics and mineral recovery from fresh and oven-dried biomass were compared in an Intermediate-Scale Aerobic Bioreactor (8 L working volume) to determine if drying crop residue improves performance in a continuously stirred tank reactor (CSTR). The study was conducted in an Intermediate-Scale Aerobic Bioreactor (ISAB) CSTR with dimensions of 390 mm height x 204 mm diameter. The pH in the bioreactor was controlled at 6.0, temperature at 30 C, and aeration at 7.0 L/min. Gases monitored were CO2 evolution and dissolved oxygen. Homogeneously mixed wheat cultures, used either fresh or oven-dried biomass and were leached, then placed in the ISAB for a 4-day degradation period. Studies found that mineral recovery was greater for leached oven-dried crop residue. However, after activity by the mixed microbial communities in the ISAB CSTR, there were little notable differences in the measured mineral recovery and degradation of soluble organic compounds. Degradation of soluble organic compounds was also shown to improve for leached oven-dried crop residue, but after mixing in the CSTR the degradation of the fresh biomass seemed to be slightly greater. Time for the biomass to turn in the CSTR appeared to be one factor for the experimental differences between the fresh and oven-dried biomass. Other factors, although not as defined, were the differing physical structures in the cell walls and varying microbial components of the fresh and oven-dried treatments due to changes in chemical composition after drying of the biomass.

Crawford, Kamau↗
DOE OSTI2013

Dry Storage of Research Reactor Spent Nuclear Fuel - 13321

Spent fuel from domestic and foreign research reactors is received and stored at the Savannah River Site's L Area Material Storage (L Basin) Facility. This DOE-owned fuel consists primarily of highly enriched uranium in metal, oxide or silicide form with aluminum cladding. Upon receipt, the fuel is unloaded and transferred to basin storage awaiting final disposition. Disposition alternatives include processing via the site's H Canyon facility for uranium recovery, or packaging and shipment of the spent fuel to a waste repository. A program has been developed to provide a phased approach for dry storage of the L Basin fuel. The initial phase of the dry storage program will demonstrate loading, drying, and storage of fuel in twelve instrumented canisters to assess fuel performance. After closure, the loaded canisters are transferred to pad-mounted concrete overpacks, similar to those used for dry storage of commercial fuel. Unlike commercial spent fuel, however, the DOE fuel has high enrichment, very low to high burnup, and low decay heat. The aluminum cladding presents unique challenges due to the presence of an oxide layer that forms on the cladding surface, and corrosion degradation resulting from prolonged wet storage. The removal of free and bound water is essential to the prevention of fuel corrosion and radiolytic generation of hydrogen. The demonstration will validate models predicting pressure, temperature, gas generation, and corrosion performance, provide an engineering scale demonstration of fuel handling, drying, leak testing, and canister backfill operations, and establish 'road-ready' storage of fuel that is suitable for offsite repository shipment or retrievable for onsite processing. Implementation of the Phase I demonstration can be completed within three years. Phases II and III, leading to the de-inventory of L Basin, would require an additional 750 canisters and 6-12 years to complete. Transfer of the fuel from basin storage to dry storage requires integration with current facility operations, and selection of equipment that will allow safe operation within the constraints of existing facility conditions. Examples of such constraints that are evaluated and addressed by the dry storage program include limited basin depth, varying fuel lengths up to 4 m, (13 ft), fissile loading limits, canister closure design, post-load drying and closure of the canisters, instrument selection and installation, and movement of the canisters to storage casks. The initial pilot phase restricts the fuels to shorter length fuels that can be loaded to the canister directly underwater; subsequent phases will require use of a shielded transfer system. Removal of the canister from the basin, followed by drying, inerting, closure of the canister, and transfer of the canister to the storage cask are completed with remotely operated equipment and appropriate shielding to reduce personnel radiation exposure. (authors)

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

Engineering Margin Factors Used in the Design of the VVER Fuel Cycles

The article describes methods for determination of the engineering margin factors currently used to estimate the uncertainties of the VVER reactor design parameters calculated via the KASKAD software package developed at the National Research Center Kurchatov Institute. These margin factors ensure the meeting of the operating (design) limits and a number of other restrictions under normal operating conditions.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Approach to nuclear criticality in IAN-R1 research reactor

The Republic of Colombia is operating a small 100 kW(t) research reactor. The reactor was provided to Colombia under the U.S. Atoms for Peace Program, and which was fueled with MTR HEU fuel enriched nominally to 93% U{sup 235}. With the cooperation of the International Atomic Energy Agency IAEA, it was prepared a new safety analysis report for performing an HEU to LEU conversion of the IAN-R1 reactor and manufacture TRIGA type LEU (19.7%, enriched) fuel to replace the original MTR-HEU fuel plate assemblies. This paper describes the activities which were achieved during the approach to nuclear criticality. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Designing Cladded Components for High Temperature Nuclear Service: Part II—Design Rules

The challenge of using existing ASME Section III, Division 5, class A metallic materials for the construction of structural components of advanced reactors with corrosive coolants could be mitigated by allowing designers to use cladding to protect the base material from corrosion. However, the existing Section III, Division 5 rules provide no guidance on the evaluation of strain accumulation and creep-fatigue damage in cladded components. The availability of design rules for cladded components that do not require long-term clad material testing could promote the application of the cladding approach to accelerate the deployment schedule of these advanced reactor systems. To avoid long-term properties for the clad materials Part I of this work proposes two approximate design analysis methods for two types of clad materials—soft clads that creep much faster and have lower yield stress than the class A base material, and hard clads that creep much slower and have higher yield stress than the class A base material. The proposed analysis methods approximate the response of a soft clad by treating it as perfectly compliant and of a hard clad by treating it as linear elastic. Based on these approximate design analysis strategies this Part II develops a complete set of design rules for class A components cladded with either soft or hard clad materials. In conclusion, Part II discusses the reasoning behind the proposed design rules and uses example finite element analyses of representative reactor components to illustrate the use of these design methods.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

EBR-II Reactor Physics Benchmark Evaluation Report

This report provides a reactor physics benchmark evaluation with associated uncertainty quantification for the critical configuration of the April 1986 Experimental Breeder Reactor II Run 138B core configuration.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI1996

Nuclear Safety [Vol. 37, No. 3, July-September 1996]

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. GENERAL SAFETY CONSIDERATIONS 181 Attitudes and Practices Regarding Disposal of Liquid Nuclear Waste at Clinton Laboratories in the Very Early Years: A Historical Analysis, S. H. Stow; 202 Off-Site Nuclear Emergency Management in Germany Under the Auspices of the Federal Structure, A. Bayer, S. Bittner, and H. Korn; DESIGN FEATURES: 211 Twenty-Fourth DOE/NRC Nuclear Air-Cleaning and Treatment Conference, R. R. Bellamy, J. J. Hayes, R. D. Porco, and M. W. First; OPERATING EXPERIENCES: 222 Explosion in the Tomsk-7 Reprocessing Plant on April 6, 1993, M. L. Hyder, W. G. Lussie, and F. E. Witmer; 234 Reactor Shutdown Experience, Compiled by J. W. Cletcher; U.S. NUCLEAR REGULATORY COMMISSION INFORMATION AND ANALYSES: 237 Assessment of Spent Fuel Cooling, J. G. Ibarra, W. R. Jones, G. F. Lanik, H. L. Ornstein, and S. V. Pullani; RECENT DEVELOPMENTS: 256 Reports, Standards, and Safety Guides, D. S. Queener; 260 Proposed Rule Changes as of June 30, 1996; ANNOUNCEMENTS: 210 Tennessee Industries Week; 221 21st International Symposium on The Scientific Basis for Nuclear Waste Management; 259 IAEA Programme of Scientific Meetings for 1997; 270 PSAM 4 (International Conference on Probabilistic Safety Assessment and Management); 265 The Authors; 268 Errata.

05 NUCLEAR FUELS↗