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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 667 records · Page 37
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Development of Helium Vessel Welding Process for SNS PPU Cavities

The Spallation Neutron Source Proton Power Upgrade cavities are produced by Research Instrument with all the cavity processing done at vendor sites with final chemistry applied to the cavity to be electropolishing. Cavities are delivered to Jefferson Lab, ready to be tested. One of the tasks to be completed before the arrival of production-ready PPU cavities is to develop a robust helium vessel welding protocol. We have successfully developed the process and applied it to three six-cell high beta cavities. Here, we present the summary of RF results, welding process development, and post helium vessel RF results.

Dhakal, P.↗
DOE OSTI2021

Evolution of the high-power spallation neutron mercury target at the SNS

The Spallation Neutron Source (SNS) began operation in 2006 and first operated at its full 1.4 MW power in 2013. Targets, which receive the pulsed proton beam, were a limiting factor for reliable full power operation for several years. Reaching reliable target operation at 1.4 MW required not only changes to the target design but also support and coordination across the entire SNS enterprise. The history and some key lessons learned are presented.

Winder, Drew E↗
DOE OSTI2023

Multipacting analysis of the SNS drift tube linac (DTL) RF vacuum window using Spark3d

An ongoing study at the Spallation Neutron Source (SNS) seeks to better understand and address potential multipacting issues associated with the Drift Tube Linac (DTL) RF vacuum windows. An analysis of several failed operational windows showed indications of excessive RF heating on the TiN-coated alumina ceramics. Coupled with vacuum bursts and arcing during conditioning and/or operational periods, these problems have been attributed to electron activity likely caused by multipacting. The status of the study, 3-D electromagnetic simulation results, mitigating techniques and a future experimental plan for studying multipacting in the SNS DTL vacuum windows are presented.

Toby, George↗
DOE OSTI2024

Using a particle-in-cell model for accelerator control room applications

Many accelerator control rooms rely on envelope models to simulate beam dynamics because they are fast and accurate at tracking the beam core. Particle-in-Cell (PIC) models, however, can track particles inside and outside the core and, with the improvements of computers, are now fast enough to be used in control rooms. The Spallation Neutron Source (SNS) at Oak Ridge National Laboratory is currently developing a tool to use a Particle-in-Cell model for control room applications. This report covers the progress so far and the future goals of using PyORBIT, a Particle-in-Cell simulation model, in the SNS control room.

Cathey, Brandon↗
DOE OSTI2024

Proton beam power limits for stationary water-cooled tungsten target with different cladding materials

The proton beam power limit for a solid-tungsten spallation target is largely determined by beam induced thermomechanical structural loads and decay heat power deposition, while its lifetime is limited by radiation damage and fatigue life of the target materials. In this paper, we studied the power limits of a stationary water-cooled solid tungsten target concept. Tantalum clad tungsten was considered as a reference case. Being a low activation material, zircaloy 2 cladding option was studied and its decay heat driven power limit was compared with the reference case. Zirconium alloys have proven operations records in spallation target and nuclear fission environments, supported by materials data obtained from post irradiation examinations. Recent study also demonstrated feasibility of diffusion bonding zirconium to tungsten using vanadium foil inter layer. Particle transport simulations code FLUKA was used to calculate energy deposition and decay heat power deposition in the target, based on the beam parameters technically feasible at the Second Target Station of the Spallation Neutron Source at Oak Ridge National Laboratory. The energy deposition data were used for flow, thermal, and structural analyses to determine the beam intensity limit on the target concept studied. The decay heat deposition data were used to calculate the transient temperature evolution in the tungsten volumes in a loss of coolant accident (LOCA) scenario to determine its beam power limit. For a 1.3 GeV proton beam, the power limit on a stationary target was 400 kW for a tantalum clad target model and 800 kW for a zircaloy 2 clad target model.

Lee, Yong Joong↗
DOE OSTI2024

Proposed muon collider R&D at SNS

Generation of a muon beam at a Muon Collider requires relatively short, high-charge proton bunches. They are produced in a high-average-power proton driver by first accumulating a proton beam from a super-conducting linac, then bunching the beam and finally compressing and combining the bunches into a single high-intensity proton pulse. All of these beam formation stages involve handling of unprecedentedly high beam charges. Validation of these intricate beam manipulations requires better understanding of extreme space-charge effects and experimental demonstration. A facility perhaps most closely resembling the proton driver configuration and beam parameters is the Spallation Neutron Source (SNS) accelerator complex at Oak Ridge National Laboratory (ORNL). Considering the energy scaling of the space-charge parameters, many of the beam formation steps planned for the proton driver can be experimentally checked at the SNS at the relevant space-charge interaction levels. This paper discusses potential proton driver and other muon-collider-related R\&D at the SNS.

43 PARTICLE ACCELERATORS↗
DOE OSTI2022

Cavity Qualification and Production Update for SNS-PPU Cryomodules at Jefferson Lab

The Proton Power Upgrade (PPU) project at Oak Ridge National Lab’s Spallation Neutron Source (SNS) currently being constructed will double the proton beam power capability from 1.4 to 2.8 MW by adding seven cryomodules, each containing four six-cell high-beta (β = 0.81) superconducting radio frequency cavities. Research Instruments, located in Germany, built and processed the cavities at the vendor site, including electropolishing as the final active chemistry step. Twenty-eight cavities for seven cryomodules and an additional four cavities for a spare cryomodules were delivered to Jefferson Lab and first qualification tests were completed on all cavities as received from the vendor. The performance largely exceeded the requirements on quality factor and accelerating gradient. Here we present the status of initial cavity qualification tests, rework on unqualified cavities and final cavity qualification with helium vessel prior to installation in cryomodules. In addition, an update on cryomodule production is presented.

Dhakal, P.↗
DOE OSTI2026

Beam loss modeling and mitigation due to intra-beam stripping

Intra-Beam Stripping (IBS) is a critical beam loss mechanism in high-intensity H- linacs and presents a significant limitation to increasing beam power. This work presents a computational framework to evaluate and mitigate IBS-induced beam loss along the Spallation Neutron Source (SNS) LINAC. Our calculation is based on an analytic theory and involves evaluation of a 9D integral using the Monte-Carlo technique. We first benchmarked our calculations against simplified, analytically solvable cases. We then applied our algorithm to Gaussian bunches with a known probability density function (PDF). We next expanded our algorithm to arbitrary bunch distributions using the Neural Spline Flow (NSF) models trained on PyORBIT tracking data. In the future, we plan to validate our algorithm experimentally and apply it to design IBS mitigation strategies.

Nln, Shivam [ORNL]↗
DOE OSTI2022

Managing Procurements in the Time of Covid-19: SNS-PPU as a Case Study

In early 2020, COVID-19 swept across the world. The accelerator industry, like many others, was impacted by disease, delays, shortages, and new working conditions. All Thomas Jefferson National Accelerator Facility (JLab) employees were sent home in mid-March 2020, with many still working remotely now. At the time, JLab was working on the Proton Power Upgrade (PPU) to the Spallation Neutron Source (SNS) at Oak Ridge National Laboratory (ORNL). Procurements had been placed and were being managed, parts were being received and inspected. This paper details the JLab procurement plan for the SNS PPU project, and the mitigations that were developed to continue to support this project smoothly under the limitations imposed by COVID-19.

Wilson, K. M.↗
DOE OSTI1994

Nuclear Safety [Vol. 35, No. 2, July-December 1994]

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: 179 Consideration of Postaccident Consequences in the Determination of Safety Objectives for Future Nuclear Power Plants in France, D. Queniart, A. Sugier, and J. Lochard; ACCIDENT ANALYSIS: 187 Nuclear Safety Research: The Phebus FP Severe Accident Experimental Program, P. von der Hardt, A. V. Jones, C. Lecomte, and A. Tattegrain; 205 Containment Performance Analysis of the Advanced Neutron Source Reactor at the Oak Ridge National Laboratory, S. H. Kim, R. P. Taleyarkhan, and V. Georgevich; 213 Assessment of Fission Product Deposits in the Reactor Coolant System: The DEVAP Program, G. Le Marois and M. Megnin; 222 Erratum to “A Review of the Available Information on the Triggering Stage of a Steam Explosion," Vol. 35, No. 1, CONTROL AND INSTRUMENTATION: 223 Effects of Normal Aging on Calibration and Response Time of Nuclear Plant Resistance Temperature Detectors and Pressure Sensors, H. M. Hashemian; DESIGN FEATURES: 235 Defense in Depth Against the Hydrogen Risk—A European Research Program, F. Fineschi; ENVIRONMENTAL EFFECTS: 246 Technical Note: A Preliminary Analysis of the Risks to Hong Kong Resulting from Potential Accidents of Daya Bay Nuclear Power Plant, Z. Shi and X. Wei; OPERATING EXPERIENCES: 253 Reactor Shutdown Experience, Compiled by J. W. Cletcher; SPECIAL SECTION ON TMI-2 VESSEL INVESTIGATION PROJECT: 256 Three Mile Island—New Findings 15 Years After the Accident A. M. Rubin and E. Beckjord; 269 Relocation of Molten Material to the TMI-2 Lower Head, J. R. Wolf, D. W. Akers, and L. A. Neimark; 280 Insight Into the TMI-2 Core Material Relocation Through Examination of Instrument Tube Nozzles, L. A. Neimark; 288 Physical and Radiochemical Examinations of Debris from the TMI-2 Lower Head, D. W. Akers and B. K. Schuetz; 301 Results of Metallographic Examinations and Mechanical Tests of Pressure Vessel Samples from the TMI-2 Lower Head, D. R. Diercks and G. E. Korth; 313 Margin-to-Failure Calculations for the TMI-2 Vessel, J. Rempe, L. Stickler, S. Chavez, G. Thinnes, R. Witt, and M. Corradini; U.S. NUCLEAR REGULATORY COMMISSION INFORMATION AND ANALYSES: 328 1993 Accident Sequence Precursor (ASP) Program Results, L. N. Vanden Heuvel, J. W. Cletcher, D. A. Copinger, J. W. Minarick, B. W. Dolan, and P. D. O’Reilley; RECENT DEVELOPMENTS: 339 Reports, Standards, and Safety Guides, D. S. Queener; 345 Proposed Rule Changes as of June 30, 1994; ANNOUNCEMENTS: 234 Thirty-First Annual Meeting of the National Council on Radiation Protection and Measurements; 356 1995 International Incineration Conference; 358 Fifth International Controls and Instrumentation Conference; 359 ANS International Topical Meeting on Safety of Operating Reactors; 359 Fifth International Conference on Nuclear Criticality Safety; 360 International Conference on Probabilistic Safety Assessment Methodology and Applications; 351 The Authors; 357 Reviewers of Nuclear Safety, Vol. 35.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Involute Working Group – Validation of CFD Turbulence Models for Steady-State Safety Analysis

There are three research reactors in the world having fuel plates curved as circle-involute (a spiral generated around a circle): The Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) located in Tennessee, U.S.A. [ORNL, 2018]; The Laue-Langevin Institute (ILL) High Flux Reactor (RHF) located in Grenoble, France [ILL, 2018]; and, The Technical University of Munich (TUM) Research Neutron Source Heinz Maier-Leibnitz (FRM II) located in Garching, Germany [TUM, 2018]. All three reactors are currently using Highly Enriched Uranium (HEU, 235U/U ≥ 20 wt. %) as fuel and all three are actively engaged in activities to convert to Low Enriched Uranium (LEU, 235U/U < 20 wt. %) fuel. More information on these reactors can be found in Section two. For various reasons, these reactors have expressed interest in using Computational Fluid Dynamics (CFD) tools to perform their Steady-State Thermal-Hydraulic (SSTH) safety calculations. Using CFD tools for SSTH analysis represents generally a significant departure from traditional methods and requires thorough verification and validation (V&V) to be accepted by regulators. This is why Argonne National Laboratory (ANL) and the involute reactors formed an informal group to help each other in this endeavor. This so-called Involute Working Group (IWG) aims at supporting the qualification of CFD tools for SSTH safety calculations. Activities include benchmarking, code-to-code comparison, Verification and Validation (V&V) as well as technical support. More information on the IWG and the reasons why involute-plate reactors are interested in CFD can be found in Section three. The present study aims at testing three turbulence models commonly implemented in commercial software (namely K-Epsilon, K-Omega and Reynolds-Stress) and assessing their relevance for involute-plate reactor application. An experiment performed by Gambill and Bundy in 1961 was performed to measure heat transfer coefficients in conditions prototypic of the HFIR reactor [Gambill, 1961]. More details on this experiment are provided in Section four. Based on the information available, CFD models of this experiment have been created and the experimental results have been compared to the ones obtained with these CFD models. CFD models and the results of this comparison are provided in Section five.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Mixed Material Scintillator Systems Particle ID Modelling Report

AMixed-Material Scintillator System (MMSS) is a radiation detector using a scintillator made of a heterogeneous, structured mixture of two or more materials. MMSSs are designed so that the structured mixing of materials encodes properties of the radiation detected in the scintillator. This allows for new radiation detectors with advantages over detectors using traditional homogeneous scintillators. This report contains results from this project’s second task, to model and quantify the promise of the MMSSs invented earlier in the project. Those inventions have been divided into two categories, and this report focuses on the first category: MMSS inventions the use zones of differently-colored scintillators to detect and characterize neutron sources. We call this category particle-ID, or PID, detectors. Inventions using gradients designed to encode the position of radiation interactions using gradients will be discussed in a future report.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2020

Who Invented the Christy Gadget

The Christy Gadget, named after its principal inventor, is the informal name for the device tested at Trinity on July 16th, 1945. In late 1944, Robert Christy proposed a novel invention that altered the design of the plutonium pit in Fat Man. While scientists originally intended to use a hollow sphere of plutonium, there was a substantial risk given a number of asymmetries that resulted from implosion. Christy proposed changing the design of the plutonium core to a solid sphere of plutonium with a modulated neutron source. The new model was adopted and used in the early implosion bombs. While there is little debate about the important role that Christy played in demonstrating its feasibility as a safe design, there is a debate as to who initially proposed the idea of the Christy Device. My project this summer outlines and resolves this dispute. While Christ does deserve the majority of the credit for the invention and design, I also acknowledge the important role Rudolf Peierls and John von Neumann played in its development.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

Removal of High Specific Activity Fission Products from Uranyl Sulfate Waste Solutions

The Savannah River National Laboratory (SRNL) is currently providing support to SHINE Medical Technologies (SHINE) which plans to deploy a low energy, accelerator-based neutron source to fission low enriched U in a uranyl sulfate target solution for 99 Mo production. The 99 Mo is initially separated from the fission products and target solution by an extraction column. Subsequent washing of the column will generate waste solutions containing residual U and fission product activity. A small number of high specific activity fission products (e.g., 90 Sr, 137 Cs, and 144 Ce) in these streams will likely control the classification of the low level waste (LLW). If a sufficient amount of the high specific activity isotopes are separated from the SHINE waste streams and concentrated into a waste form, it would be possible to treat a majority of the wash solutions from the column operations as a lower class of LLW (Class A versus Class B or C or Class B versus Class C). The high specific activity fission product elements could then be dispositioned as a much smaller volume of waste rather than requiring the disposal of the entire waste stream at the higher waste classification. The Savannah River Site (SRS) has experience with using monosodium titanate (MST) and crystalline silicotitanate (CST) to remove Cs and Sr from high salt content waste solutions generated during the reprocessing of nuclear fuels and targets. Both of these materials have worked very well for their intended purposes at the SRS where the fission product elements are removed from highly alkaline waste. On the other hand, SHINE waste streams from the extraction column contain H 2 SO 4 which makes the solution acidic. Additionally, the SRS waste streams do not contain other fission product elements such as transition metals or lanthanides because they precipitate upon neutralization of the SRS waste and are not present in the supernate which is dispositioned as LLW following treatment. As such, there are inherent differences between SHINE and SRS waste treatment strategies. Savannah River National Laboratory was tasked with performing scoping studies to see if MST and CST would remove Sr, Cs, and Ce from an acidic mixed metal simulant solution. Batch contact experiments were performed using MST and two CST type materials. The MST material is a 15 wt % powder in 0.15 M NaOH slurry. The MST showed low adsorption for elements of interest from acidic solution. Furthermore, the powder size makes MST non-ideal for column operations. A CST IE-911 ion exchange material had high Cs adsorption, moderate Sr, and marginal Ce adsorption. Based on adsorption of all species, the ion exchange capacity was found to be 0.032 meq/mL. A bench-top column experiment to measure elemental breakthrough curves was performed using CST IE-911 where chromatographic separations of the mixed simulant were expected to occur. While most elements behaved as expected, the lanthanide series, containing Ce, broke though the column earlier than expected. The second CST material, CST R9120, displayed high adsorption for all elements in the acidic mixed simulant solutions in a batch contact study, and had a calculated loading capacity of 0.091 meq/mL. Future studies to develop a waste treatment flowsheet should focus on CST R9120 to treat SHINE waste solutions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Status Report on Development of a Cask to Enable Pulsed Neutron Characterization of Irradiated Fuel

We present the design of a shielding cask that allows pulsed neutron characterization of irradiated fuel rodlets prior to their destructive examination in hot cells. The goal is to provide complementary and informative measurements that will inform subsequent hot cell PIE examinations by identifying typical and atypical regions with respect to microstructure, tomographic data, or isotope densities. Time-of-flight neutron diffraction has the potential to offer efficient, non-destructive and non-contact microstructural characterization of irradiated fuel pins with spatial resolution of 1 mm3 to 1 cm3 while for energy-resolved neutron imaging (and by extension tomography) a resolution of 100 μm was demonstrated. The potential results include crystallinity vs. amorphous volumes and microstructural information such as phase compositions and textures from the diffraction data as well as distances (e.g. pellet to cladding), cracks, and isotopic distributions of minor actinides, fission products as well as fission gas partial pressures e.g. in the plenum from energy resolved neutron imaging. The capability will identify regions of atypical behavior and provide cost effective bulk characterization of entire samples prior to destructive evaluation in hot cells. This effort is a collaboration between LANL, INL, ORNL, and UC Berkeley. A key facet of the initiative is the ability to cost effectively ship irradiated fuels from INL to the pulsed spallation neutron source at LANSCE. The irradiated fuel would be loaded in the custom designed cask (designated RaMHaM) at INL hot cells which then could be shipped in a BRR Type B shipment container between INL and LANL. No hot cell is required after the loading of the sample at INL in the pathway laid out in this report, greatly simplifying handling at LANL.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Oak Ridge Response to Versatile Test Reactor Environmental Impact Statement Data Request

The Versatile Test Reactor (VTR) is a fast-spectrum test reactor being developed in the United States under the direction of the US Department of Energy Office of Nuclear Energy (DOE-NE). The VTR mission is to enable accelerated testing of advanced reactor fuels and materials required for advanced reactor technologies. The conceptual design of the 300 MWth sodium-cooled metallic-fueled pool-type fast reactor has been led by the US National Laboratories in collaboration with General Electric–Hitachi and Bechtel National, Inc. In support of the VTR project, DOE issued a Notice of Intent (NOI) in the Federal Register on August 5, 2019, announcing the intent to prepare an Environmental Impact Statement (EIS) in accordance with the National Environmental Policy Act (NEPA) and its implementing regulations. The EIS will evaluate alternatives for a versatile reactor–based fast-neutron source facility and associated facilities for the preparation, irradiation, and post-irradiation examination (PIE) of test/experimental fuels and materials. Specifically, the NOI identified two siting alternatives for the VTR reactor facility: Idaho National Laboratory (INL) or Oak Ridge National Laboratory (ORNL). In addition, the NOI also specified two siting alternatives for VTR fuel fabrication: INL and the Savannah River Site (SRS). This report provides information in response to data requests made to ORNL to fill in site-specific knowledge gaps to develop a high-quality EIS. The responses provided are not required to provide full details in every aspect; instead, they adequately bound possible environmental impacts or provide sufficient information to adequately assess likely environmental impacts. This work is being performed under a subcontract from INL to ORNL using DOE-NE funds and is directed by DOE-NE and DOE-ID. Leidos has been contracted by DOE-NE to write the VTR EIS, so most data requests have come from Leidos but were often routed through INL or DOE-ID. DOE-ID is overseeing the NEPA and EIS processes for the VTR project. Leidos will use the information provided in this report to inform the VTR EIS and will also cite this document to establish a clear, publicly available source of the information. Section 2 of this report briefly describes the proposed ORNL VTR Alternative and illustrates the location of the proposed site for the ORNL VTR Alternative. Sections 3 through 7 provide direct responses to data requests received by ORNL. These sections use a tabular format in which data requests are divided into separate items to be addressed; the items are numbered, the data requests are restated with more topical information included, and then the responses are provided. Initial data requests and follow-on requests for additional information (RAIs) are combined under the original data request fields. Finally, Section 8 presents summarized conclusions and describes future work.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Tomographic Image Reconstruction for the Parallel-Slit Ring Collimator Fast Neutron Emission Tomography System

For the past three years, Oak Ridge National Laboratory (ORNL) has been developing a passive fast-neutron emission tomography capability. The goal of this development is the ability to quantify the neutron source strength of individual fuel pins (rods) in spent nuclear fuel assemblies. Such a system could be used to measure the burnup of each fuel pin in a spent fuel assembly in order to take burnup credit when loading dry storage casks or to count individual fuel pins in spent fuel assemblies for safeguards purposes. At present, a laboratory prototype imager is under construction. The purpose of this prototype is to demonstrate imaging capability sufficient to resolve individual fuel pins in spent fuel assemblies. This report documents the development of the iterative reconstruction code used to perform tomographic image reconstruction, the imager response calculation used by the reconstruction code, and the results of reconstructions of simulated tomographic imaging measurements for the prototype imager design.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Fundamental Neutron Physics: Theory and Analysis

The research topic of the project focus on the theoretical study of the possibility to search for manifestations of new physics beyond the Standard model in fundamental neutron physics, with emphasis on fundamental symmetries and "exotic'' interactions. The mail goals concentrate on the search for Time Reversal Invariance Violation (TRIV) in neutron nuclei scattering, and on possible manifestations of new physics in neutron-antineutron oscillations and their interactions with nuclei. Theoretical methods for analysis and interpretations of TRIV parameters in neutron scattering with the possible background interactions, and for unambiguous separation of the effects related to manifestations of new physics have been developed. The obtained results will be applied for the development of experimental programs in fundamental neutron physics at the Spallation Neutron Sources (SNS) in USA, Japan, and Europe.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗