Assessment of the Back End of Advanced Reactors for Developing Stable Wastes Forms
Explore the source record for details and available documents.
Bonano, Evaristo J.↗
SEARCH · Engineering Papers
Results for “L REACTOR”
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.
Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.
Am-Bearing Blanket Transmutation Strategies in Sodium-cooled Fast Reactors
Explore the source record for details and available documents.
Stauff, Nicolas E.↗
Assessment of the Back End of Advanced Nuclear Reactors for Developing Stable Waste Forms
Explore the source record for details and available documents.
Bonano, Evaristo J.↗
Augmented Monitoring and Condition Assessment Program (AMCAP) Material Test Reactor (MTR) (Fuel Inspection Program Report)
The AMCAP MTR Fuel Inspection Program, a special inspection program comprised of four inspection campaigns, examined a total of ten (10) pre-selected aluminum-clad, aluminum-based fuel core spent nuclear fuel assemblies (ASNF) stored in the L Area Disassembly Basin (L Basin) at the Savannah River Site. A full description of the inspections and the results are reported. The fuel had been stored in bundled-tube storage (Vertical Tube Storage) for periods of 18 to 21 years. The prior service experience of the individual 10 assemblies varies, but all included irradiation followed by wet storage at international research reactor sites prior to shipment to the US and storage in L Basin. The 10 assemblies were expected to be among the “worst"’ in terms of prior corrosion damage of the entire inventory of the direct-bundle-stored ASNF in L Basin. The inventory of MTR ASNF in L Basin will continue to be stored in the bundled-tube configuration or in slug-storage buckets with inserts pending retrieval for ultimate disposition. The MTR fuel inspections focused on collecting information for characterization of the material condition of the ASNF considering various types of aluminum fuel corrosion degradation of its assembled materials in water storage. A custom-designed Fuel Inspection Table was used to stage the fuel for remote, enhanced visual examination (close-up video imaging & recording) with controlled lighting and positioning that enables reproducibility of imaging conditions. The inspections were conducted by fuel subject matter expert staff from Spent Fuel Project Engineering (SFPE) and the Savannah River National Laboratory (SRNL). Stills captured from the video records were used to compare the corrosion evolution from previous records, as available. This evaluation of the inspection results including the comparison to the previous inspection results demonstrate that the water quality and the storage configuration of ASNF in L Basin do not cause aggressive corrosion degradation of the fuel; mitigation of the prior corrosion degradation of the fuel also appears to have been achieved with the good water quality conditions of L Basin. Recommendations are made for future inspection of the fuel to trend corrosion degradation and demonstrate continued safe wet storage of the ASNF in L Basin. The next fuel examination is recommended to be performed in 5 years.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Eddy Current Flow Measurements in the FFTF
The Fast Flux Test Facility (FFTF) is the most recent liquid metal reactor (LMR) to be designed, constructed, and operated by the U.S. Department of Energy (DOE). The 400-MWt sodium-cooled, fast-neutron flux reactor plant was designed for irradiation testing of nuclear reactor fuels and materials for liquid metal fast breeder reactors. Following shut down of the Clinch River Breeder Reactor Plant (CRBRP) project in 1983, FFTF continued to play a key role in providing a test bed for demonstrating performance of advanced fuel designs and demonstrating operation, maintenance, and safety of advanced liquid metal reactors. The FFTF Program provides valuable information for potential follow-on reactor projects in the areas of plant system and component design, component fabrication, fuel design and performance, prototype testing, site construction, and reactor control and operations. This report provides HEDL-TC-1344, “ECFM Flow Measurements in the FFTF Using Phase-Sensitive Detectors”, March 1979.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Beam characterization at the Neutron Radiography Reactor
The quality of a neutron-imaging beam directly impacts the quality of radiographic images produced using that beam. Fully characterizing a neutron beam, including determination of the beam's effective length-to-diameter ratio, neutron flux profile, energy spectrum, potential image quality, and beam divergence, is vital for producing quality radiographic images. This paper provides a characterization of the east neutron imaging beamline at the Idaho National Laboratory Neutron Radiography Reactor (NRAD). The experiments which measured the beam's effective length-to-diameter ratio and potential image quality are based on American Society for Testing and Materials (ASTM) standards. An analysis of the image produced by a calibrated phantom measured the beam divergence. The energy spectrum measurements consist of a series of foil irradiations using a selection of activation foils, compared to the results produced by a Monte Carlo n-Particle (MCNP) model of the beamline. The NRAD has an effective collimation ratio greater than 125, a beam divergence of 0.3 +_ 0.1 degrees, and a gold foil cadmium ratio of 2.7. The flux profile has been quantified and the facility is an ASTM Category 1 radiographic facility. Based on bare and cadmium covered foil activation results, the neutron energy spectrum used in the current MCNP model of the radiography beamline over-samples the thermal region of the neutron energy spectrum.
99 GENERAL AND MISCELLANEOUS↗
Finite element analysis of compliant cladding and base metal systems
This report describes a roadmap towards a design method for high temperature nuclear cladded components that does not require long term testing of clad materials. Molten Salt Reactors (MSRs) will operate using corrosive fuel or coolant salts. The existing Class A metallic materials qualified for ASME Boiler and Pressure Vessel Code Section III, Division 5 design are not optimal for corrosion resistance when exposed to MSR salts. Corrosion could then severely limit the service life of MSRs. One option to alleviate this could be to Code qualify new corrosion resistant materials. However, this is a lengthy and expensive process and MSR developers have a near term need for a design solution that enables reactor designs with long service lives. An alternative would be to use cladded components – overlay some existing Class A material with a thin layer of some corrosion resistant material. This report describes the outlines of a design strategy that would enable this approach. The strategy relies on approximate design analysis methods for certain types of clad/base systems that avoids requiring long-term mechanical properties for the clad materials. This would allow the use of cladded components in the near term, without requiring long testing programs. However, it will limit the types of clad materials that can be used. This report describes the design framework, demonstrates the effectiveness of the approximate analysis methods, and confirms the approach using full scale finite element analysis of prototypical cladded high temperature components. It also describes a roadmap towards a complete design method and surveys available clad materials and bonding methods for suitably, considering only mechanical and not environmental effects.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Data Quality of Salt Property Measurements
Data quality is a key element to assessing technical uncertainties in property values of salt mixtures that are pivotal in molten salt reactor (MSR) design and safety assessments under both normal operating conditions and accident scenarios. Key properties for the salt mixtures being evaluated for use in MSRs have not been measured and the few values for relevant salt compositions available in the literature are inconsistent and not suitable for use in licensing. The applications of established techniques commonly used for property measurements to molten salts are still experimental in nature and vetted consensus standards are not yet available. Work is in progress to develop and standardize methods through DOE and commercially funded projects. The quality of measured property values limits the current ability to reliably design equipment and model normal MSR operation and the effects of operational upsets. Insights from experimental measurements pertinent to reducing technical uncertainty for regulatory development are summarized in this report. Quality controls are recommended for key properties based on the mathematical calculations used to derive property values and methods used to measure experimental values and salt produced for use in the measurements.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Baseline Fuel Fabrication Facility
PRO-RR is the research reactor focused program element of the broader Proliferation Resistance Optimization program (PRO-X) under the National Nuclear Safety Administration (NNSA) in the U.S. Department of Energy (DOE). PRO-X provides a framework for integrating proliferation resistance in nuclear system designs to minimize weapons usable nuclear materials (WUNM) production and diversion pathways while optimizing systems performance for peaceful use missions. PRO-RR applies the PRO-X mission objectives to research reactor system design. This document serves as one of the foundational documents for the PRO-RR-Fuel System Design technical team by documenting a baseline fuel fabrication facility to be used for further optimization studies. The PRO-RR-Fuel System Design technical team consists of subject matter experts from Argonne National Laboratory (Argonne) and Savannah River National Laboratory (SRNL). In order to develop specific strategies for fuel fabrication facilities to optimize proliferation resistance, performance, and safety, a baseline fuel fabrication facility design basis was developed. Having a baseline design basis allows for the qualitative and quantitative comparison of design choices in the optimization process. This report describes the baseline fuel fabrication facility and general optimization strategy. Chapter 2 describes the fuel system selected for examination, the fabrication process used as the baseline, a description of the model developed to track uranium utilization, and a generic floorplan of the fabrication facility. Chapter 3 describes the overarching optimization strategy that could be implemented for a fabrication facility.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Testing of graphite thermal neutron scattering law data in support of treat neutronic analysis
The Transient Reactor Test Facility (TREAT) has historically provided the capability for transient fuel performance and safety testing in the United States. The operation of this facility was suspended in 1994. However, in recent years preparations for the restart of TREAT were initiated. TREAT is a graphite moderated and air cooled reactor. Its core is composed of a 19 x 19 array of fuel assemblies that contain UO{sub 2} (HEU) fuel particles dispersed in a graphite matrix, along with reflector assemblies filled with graphite only. Each fuel assembly is 8 feet tall with a fueled region that is 4 feet tall and upper and lower regions of graphite that are 2 feet tall. In addition, a 2 feet thick radial graphite reflector region is present. Finally, a 5 feet thick concrete biological shield surrounds the core. Previous work showed that the graphite thermal scattering cross sections can introduce variations exceeding 2000 pcm in the predicted eigenvalue of the TREAT core. Recognizing that the graphite in the assembly (and thereby the core) is a porous mixture of crystalline and amorphous carbon (with a lower density in comparison to ideal crystalline graphite), it becomes clear that TREAT specific thermal neutron scattering cross section libraries will be needed (at all temperatures of interest) for use within neutronic codes seeking to simulate TREAT. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
ECAR-8446 Rev 0 MARVEL BPVC Section III Division 5 HBB Fatigue Acceptance Tests
The objective of this analysis is to evaluate the fatigue and creep behavior of 316 SS materials used in the Primary Coolant System (PCS) and Guard Vessel System (GVS) for the first two years of MARVEL reactor operation. The analysis follows the requirements set forth in ASME BPVC Section III, Division 5, HBB-2800, which mandates a fatigue test if operational conditions exceed allowable thresholds. By determining whether the materials meet the HBB-T-1324(a) and HBB-T-1324(b) criteria, this evaluation establishes whether a fatigue test is necessary, as well as the maximum allowable operational hours and temperatures before fatigue testing would be required. The deliverables of this analysis include calculations validating compliance with fatigue acceptance criteria, determination of maximum operational limits for the two-year period, and identification of conditions under which a fatigue test would be necessary. This ensures that the PCS and GVS can operate safely without premature material degradation and align with the design life requirements outlined in SPC-70731. Additionally, this evaluation provides conservative estimates for Service Level B occurrences and their impact on allowable Service Level A hours and temperatures, offering guidance for future operational assessments and potential life extensions beyond the initial two-year period.
21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
Summation Calculations for Reactor Antineutrino Spectra, Decay Heat and Delayed Neutron Fractions Involving New TAGS Data and Evaluated Databases
Fallot, M.; Giot, L.; Guadilla-Gomez, V.; Le Meur, L.; Porta, A.; Algora, A.; Tain, J. -L.; Briz, J. A.; Agramunt, J.; Aysto, J.; Cormon, S.; Cucoanes, A.; Eronen, T.; Fraile, L. M.; Ganogliu, E.; Gelletly, W.; Gorelov, D.; Hakala, J.; Jokinen, A.; Jordan, M. D.; Kankainen, A.; Kolhinen, V. S.; Koponen, J.; Lebois, M.; Martinez, T.; Monserrate, M.; Montaner-Piza, A.; Moore, I.; Nacher, E.; Orrigo, S. E. A.; Penttila, H.; Pohjalainen, I.; Reinikainen, J.; Reponen, M.; Rinta-Antila, S.; Rytkonen, K.; Rubio, B.; Shiba, T.; Sonnenschein, V.; Valencia, E.; Vedia, V.; Voss, A.; Weber, C.; Wilson, J. N.; Zakari-Issoufou, A. -A.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Further work on credible levels of water moderation - 14267
The Research and Test Reactor (RTR) manufacturing operation produces aluminum clad fuel plates that are assembled into 'box type' elements that are used in reactors like the Advanced Test Reactor, University of Missouri, and Massachusetts Institute of Technology. These fuel plates and elements are stored in both horizontal and vertical arrangements during the manufacturing process. Since the presence of water is so important to criticality safety of the operations, a study was undertaken to establish credible levels of water moderation. At the NCSD 2013 Topical meeting, credible levels of interspersed water moderation from multiple sources including firefighting, steam and rain were presented. This covered the initial portions of the research. Subsequent research was conducted on water films resulting from these water sources. This phase of the research involved the formation of water films. These can form from condensation, water sprays or immersion. The thickness of the water films from these sources ranged from 0.04 cm to 0.30 cm for vertical fuel plates. Studies were also conducted on horizontal plates, stacks of plates and assembled elements. For fuel plates, the films ranged from essentially zero for stacked plates to 0.25 cm. For an element, there was sufficient adhesion that all of the channels were filled. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Comparison of Calculated and Experimental Flows and Pressures in the Orifice Flow Test Facility
A new experimental facility for materials irradiation and testing at the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) is being developed. Details of this facility have been presented in previous work. A prototype of this facility, the Thermosyphon Test Loop (TSTL), has been built and experimental data have been obtained and analyzed. Comparison of TSTL experimental data and calculations has been presented previously. The TSTL did not measure any flows (water or vapor) inside the test section because flowmeters could not be installed inside. Instead, two pressure sensors (P1 and P2) were installed at the bottom of the test section to measure the pressure differential ({Delta}P = P1-P2) across the orifices between the bottom of the downcomer and the bottom of the boiler. This {Delta}P is proportional to the water flow, but actual flows were not measured. Thus, another facility, the Orifice Flow Test Facility (OFTF), was built to measure water flow vs {Delta}P. The OFTF duplicates the bottom of the test section of the TSTL. The OFTF models the downcomer and the boiler including three dummy heaters (without power) inside the boiler. This paper presents the experimental data obtained and comparison to computer calculations. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Performance Analysis and Simulaion of the Hydraulic Scram System in TREAT Reactor
The Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) serves a vital role in nuclear fuel safety research, enabling transient experiments that simulate reactivity excursions and accident scenarios. Central to these operations is the transient control rod drive system (TCRDS), which drives rapid motion of the transient control rods such that TREAT can simulate rapid power changes typical of reactor accidents. The reliability and performance of this system are critical for protecting both fuel specimens and reactor infrastructure. This study presents the initial phase of a two-year investigation into the dynamics and reliability of the TREAT hydraulic TCRDS. Conducted in collaboration with INL, the research employs a combined computational and experimental approach to analyze the system's response time, pressure transients, and potential failure modes. Emphasis is placed on understanding how fluid characteristics influence the TCRDS’s ability to achieve both rapid power changes and mechanical stability. The TRDS and the skid that powers it will be analyzed throughout this investigation. Computational modeling using computational fluid dynamics (CFD) will simulate the hydraulic response under varying conditions. In parallel, experimental testing planned at INL will validate these models and capture key performance metrics. This paper outlines the system design, analytical framework, and modeling strategies that form the foundation for later testing. Ultimately, this work aims to support improvements to the TCRDS’s design and reliability, contributing to the broader goal of enhancing nuclear fuel safety and sustaining TREAT’s mission as a premier nuclear fuel test facility.
21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
Delivery of Dynamic Thermal Energy Storage Models and Advanced Reactor Concept Models to the HYBRID Repository
This publication details newly created energy storage and reactor models developed within the HYBRID modeling repository as part of the Department of Energy Office of Nuclear Energy (DOE-NE) Integrated Energy Systems (IES) program, led by Idaho National Laboratory (INL). Model development to-date includes creation of dynamic systems-level models of a pebble bed high temperature gas reactor (HTGR), sodium fast reactor (SFR), compressed air energy storage (CAES), liquid air energy storage (LAES) and Modelica standard library based two-tank sensible heat storage (SHS) in the IES-based HYBRID repository. Models are developed using the latest publicly available data and incorporate the possibility of control strategy inclusion for use with the existing IES modeling, analysis, and optimization toolset. Simulations showcase the abilities of each technology to flexibly operate in ways consistent with IES operation expectations. When these models are available, they can be utilized within different integrated energy park concepts to understand optimal system operation, control, and dispatching. Moreover, given the generic nature of the models, industrial partner technologies can be quickly added to the repository using the existing models as a basis. Additional dynamic models for thermal energy storage concepts can be developed and added to the HYBRID repository as needed. Also detailed in this report are future development goals for the HYBRID repository including adding suites of steady-state models, economic costing information, and reduced-order models. By adding these models in addition to the physical transient models currently existing within HYBRID, HYBRID will be a fully integrable tool for FORCE users.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Room-temperature ballistic energy transport in molecules with repeating units
In materials, energy can propagate by means of two limiting regimes: diffusive and ballistic. Ballistic energy transport can be fast and efficient and often occurs with a constant speed. Using two-dimensional infrared spectroscopy methods, we discovered ballistic energy transport via individual polyethylene chains with a remarkably high speed of 1440 m/s and the mean free path length of 14.6 Å in solution at room temperature. Whereas the transport via the chains occurs ballistically, the mechanism switches to diffusive with the effective transport speed of 130 m/s at the end-groups attached to the chains. A unifying model of the transport in molecules is presented with clear time separation and additivity among the transport along oligomeric fragments, which occurs ballistically, and the transport within the disordered fragments, occurring diffusively. The results open new avenues for making novel elements for molecular electronics, including ultrafast energy transporters, controlled chemical reactors, and sub-wavelength quantum nanoseparators.
37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗