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At least 1,729 records · Page 96

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DOE OSTI2017

DCA_Tools

HPC software for ab-initio, condensed-matter physics, quantum mechanics calculations needs to be built on top of well tested libraries some of which address requirements unique to the programming domain. During the development of the DCA++ code, that we use in our research, we have developed a collection of libraries that may be of use to other computational scientists working in the same or similar domains. The libraries include: a) a pythonic input-language system, b) tensors whose shape is constructed from generalized dimension objects such at time domains. frequency domains, momentum domains, vertex domains et. al. and c) linear algebra operations that resolve to BLA/LAPACK operations when possible. This supports the implementation of Greens functions and operations on them such as are used in condensed matter physics.

Summers, Michael S↗

DOE OSTI2021

Infrasound and Low-Frequency Acoustics MINOS Briefing

Oral presentation intended for Multi-Informatics for Nuclear Operations Scenarios (MINOS) venture teams describing programmed notification algorithm and ground truth collected at the Advanced Test Reactor.

42 ENGINEERING↗

DOE OSTI2022

The Nuclear Fuel Cycle: Safe Management of Spent Nuclear Fuel

The aim for storage of spent nuclear fuel (SNF) either in wet or in dry storage systems is to ensure general safety objective s are met throughout a desired storage period. Staff at the Savannah River National Laboratory (SRNL), in collaborations with partners at other national laboratories, industry research organizations, and the University of South Carolina (UofSC), have performed materials aging testing and analyses, and have established nuclear materials aging management programs to support extended periods of safe storage of research reactor (RR) SNF and of commercial power reactor (PR) SNF pending ultimate disposal. Several example challenges include susceptibility of aluminum SNF from research reactors to corrosion in poor quality water (wet storage). In dry storage, aluminum SNF can release hydrogen via radiolysis of the hydrated oxides on the aluminum cladding. Austenitic stainless steel canisters used for dry storage are susceptible to chloride-induced stress corrosion cracking (outside-in attack) that threaten the confinement boundary provided by the canister. This paper further describes these challenges, among others, and the formulated solutions to support extended safe storage of SNF.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

Andra's Underground Research Laboratory as a Tool to Prepare Monitoring System of Cigeo Project - 18070

Andra, as French radioactive waste management agency, plans to build an underground disposal facility, named Cigeo, dedicated to high and intermediate level, long lived radioactive wastes. To ensure the safety during the disposal operations that includes the principle of reversibility, the design of the associated monitoring system will be based on specific research and development program and feedback obtained over more than 15 years on tests performed in representative conditions. Andra's Underground Research Laboratory (URL), located at 500 m deep, is excavated in a clay-rich formation. It is dedicated to (i) assess the thermo-hydro-mechanical, chemical and radiologic behaviors of the rock and supporting layers (concrete and steel), (ii) test envisaged monitoring system, (iii) provide feedback on operational organization and safety management and (iv) qualify operating methods (digging, concrete pouring, monitoring system design, etc.) to be implemented in the future disposal. Data collected from (i) characteristics of samples retrieved from boreholes, (ii) experts' observations, (iii) sensors installed in situ and (iv) manual measurements, are integrated in a database. This database is dedicated to samples management (more than 73 600 classified samples), data measurements collection, and conversion to physical values and available to plots or exports. Each day, approximately 1.6 million of values are integrated in the database. Finally, a geo-referenced database is linked to the main database in order to allow the spatial localization, both on the surface or in the underground galleries. All these tools give confidence in Andra's capacity to monitor Cigeo and be ready for the future industrial pilot phase. The monitoring system will also allow, step by step, optimizing the disposal design, performing regular safety control, helping in decision making for reversibility and improving long term safety. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

Air-water and steam-water onset of flooding data at variable pressure in a large diameter tube

An experimental investigation to acquire flooding data for steam-water and air-water fluid pairs in a large-diameter vertical tube at above-atmospheric pressures is being conducted. The experimental program presented herein is the first to use a single test section with controlled boundary conditions for these pairs in a large-diameter vertical tube, enabling consistent comparisons of flooding characteristics. Many data exist for air-water flooding in annular flow within a large-diameter vertical tube and at various pressures. Steam-water flows are of interest for reactor applications but insufficient data are available to enable reliable analysis. Evaluation of the effect of the fluid pair is performed to understand the applicability of the air-water data to steam-water applications. The pressure range for the tests reported herein is 0.1-0.4 MPa (3-45 psig) for the air-water experiments and 0.1-0.2 MPa (3-15 psig) for the steam-water experiments. The steam-water tests were conducted with minimal subcooling to reduce the effect of phase change and permit a more level comparison with the air-water data. All data were taken at the onset of flooding, that is, conditions when flow reversal of the liquid phase first occurs. The data at higher pressures suggest that the momentum transfer necessary for the onset of flooding is achievable with lower superficial gas velocities as the gas density increases. Flooding characteristics are presented in terms of the Kutateladze parameter. If condensation is accounted for, the saturated steam-water data at 0.1 and 0.2 MPa (3 and 15 psig) closely trend with the air-water flooding data. (authors)

42 ENGINEERING↗

DOE OSTI2025

Update on R&D progress by DOE

High temperature materials and graphite and composite program review to include: Baseline for unirradiated material properties, ASTM test development, design and construction rules (ASME-based), new fuel matrix mechanical studies, Irradiation damage studies, vender irradiation capsules, graphite supply chain worries, and ceramic composite activities.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Ultrasonic Transducer Irradiation Test Results

Ultrasonic technologies offer the potential for high accuracy and resolution in-pile measurement of a range of parameters, including geometry changes, temperature, crack initiation and growth, gas pressure and composition, and microstructural changes. Many Department of Energy-Office of Nuclear Energy (DOE-NE) programs are exploring the use of ultrasonic technologies to provide enhanced sensors for in-pile instrumentation during irradiation testing. For example, the ability of small diameter ultrasonic thermometers (UTs) to provide a temperature profile in candidate metallic and oxide fuel would provide much needed data for validating new fuel performance models. These efforts are limited by the lack of identified ultrasonic transducer materials capable of long term performance under irradiation test conditions. To address this need, the Pennsylvania State University (PSU) was awarded an Advanced Test Reactor National Scientific User Facility (ATR NSUF) project to evaluate the performance of promising magnetostrictive and piezoelectric transducers in the Massachusetts Institute of Technology Research Reactor (MITR) up to a fast fluence of at least 1021 n/cm2 . A multi-National Laboratory collaboration funded by the Nuclear Energy Enabling Technologies Advanced Sensors and Instrumentation (NEET ASI) program also provided initial support for this effort. This irradiation, which started in February 2014, is an instrumented lead test and real-time transducer performance data are collected along with temperature and neutron and gamma flux data. The irradiation is ongoing and will continue to approximately mid-2015. To date, very encouraging results have been attained as several transducers continue to operate under irradiation.

Daw, Joshua↗

DOE OSTI2025

A Cultural Resource Survey of 661.5 Acres on the Tonopah Test Range (TTR), Nellis Air Force Base, Nye County, Nevada

This report presents the results of a Section 110 cultural resource survey of Pedro Lake and the surrounding area. Sandia National Laboratories’ (Sandia) cultural resources program, in collaboration with personnel from the cultural resource department from the Northern Ranges of the Nevada Test and Training Range (NTTR), Nellis Air Force Base (NAFB) surveyed 267.71 hectares (ha) (661.50 acres [ac]) for the project, located on the Tonopah Test Range (TTR) at Nellis Air Force Base, Nye County, Nevada. The survey resulted in the update of two previously recorded archaeological sites, the documentation of 15 newly discovered archaeological sites, and 18 isolated occurrences. The results of the survey are presented herein.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2018

In Backyards and Bedrooms: Challenges and Successes in Gaining Access to Homes for Radiological Testing - 18204

The Port Hope Area Initiative (PHAI) is a federal environmental clean-up program. Its mandate is the remediation and local, long-term, safe management of approximately 1.7 million cubic metres of historic low-level radioactive waste in the adjacent communities of Port Hope and Port Granby in Southern Ontario, Canada. The PHAI is being undertaken by Canadian Nuclear Laboratories (CNL), on behalf of Atomic Energy of Canada Limited, a federal Crown corporation. It is currently one of the largest environmental remediation projects in Canada The safety of workers, the public and the environment while delivering the PHAI on time and within budget is CNL's number one priority. The PHAI is being carried out as two projects - the Port Hope Project and the Port Granby Project. The Port Hope Project is the larger of the two and includes the cleanup of approximately 1.2 million cubic metres of historic waste from various locations within the community. The Port Hope Project includes a unique and important component - the Property Radiological Survey. This survey involves the interior and exterior investigation of approximately 4,800 small-scale properties - primarily residential - in the Municipality of Port Hope for the presence or absence of historic low-level radioactive waste. The Property Radiological Survey began in 2012 and is being conducted in five campaigns with between 450 and 1,250 properties in each campaign. The survey begins with radon monitoring, followed by interior and exterior gamma radiation tests, and concludes with exterior borehole drilling/soil sampling. The full suite of tests take approximately 18 - 24 months to complete at each property. This paper will describe the communications strategy implemented to obtain consent from thousands of property owners for CNL's contractor to conduct radiological testing in private living spaces such as bathrooms, bedrooms and closets, as well as front yards, backyards and outbuildings. This strategy incorporates tactics used to secure consent from residents who are reluctant to participate in the survey for reasons including privacy issues, family situations, life events and misapprehensions about the survey and/or the PHAI in general. The paper will include information about how residents' concerns are addressed, the stepped approach to securing access from owners who do not respond to repeated attempts at contact, and the importance of close coordination with the contractor to ensure that all components of the investigations are aligned so that work is conducted in a safe, timely and systematic manner. The authors will also describe the process for keeping the regulator and the municipality informed of survey outreach efforts, challenges and successes, and will outline the process for notifying the approximately 800 owners of properties on which historic low-level radioactive waste is expected to be identified and transitioning those owners from the survey to the remediation phase. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Basis of Estimate Summary - Combined Radiation Environments for Survivability Testing (CREST)

The following document is Sandia National Laboratories Facilities Site and Strategic Partnerships methodology and assumptions used in the assembly of programming, planning, and budgeting level cost range for early capital acquisition needs development and communication. This basis of estimate summary is specific to the Combined Radiation Environments for Survivability Testing (CREST) Line Item proposal, and reflects updated estimates provided by the Subject Matter Experts (SME) team as indicated.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2021

OVERVIEW OF RECENT PU-238 PRODUCTION ACTIVITIES AT IDAHO NATIONAL LABORATORY

The Plutonium Fuel Services (PFS) program at Idaho National Laboratory (INL) is active in the qualification of irradiation targets containing Np-237 for irradiation in the Advanced Test Reactor (ATR) to produce Pu-238 for future NASA missions. INL qualified and loaded 7 targets in ATR’s South Flux Trap (SFT) for cycle 169A, which occurred in Spring 2021. This program was reinitiated after two baseline production targets in three positions validated significant production of Pu-238 [ref.1]. The validation model was followed by the PFS-1 experimental test in the ATRC (Critical) facility [ref.2]. This paper outlines the progress and status of the PFS program. The qualification effort, safety analysis, hardware status, and future activities for qualification of an updated target design for use in the ATR will be discussed.

36 MATERIALS SCIENCE↗

DOE OSTI2022

NEAMS Advanced Reactor Model Contributions to the NRIC Virtual Test Bed

The U.S. Department of Energy (DOE) Office of Nuclear Energy’s Advanced Modeling and Simulation (NEAMS) Program develops models of advanced reactor phenomena to demonstrate code applicability to challenging physics problems, drive code development through user assessment, and perform code verification and validation. Meanwhile, the U.S. DOE’s National Reactor Innovation Center (NRIC) hosts an open-source website and associated GitHub repository called the Virtual Test Bed (VTB) on which computational models for advanced reactors are documented and shared with the reactor community. Several NEAMS advanced reactor models (including input files, documentation, and discussion of results) have recently been contributed to the NRIC VTB. The open sharing of these models benefits the reactor community by providing “best practice” examples using NEAMS tools for advanced reactor physics problems. This report summarizes and provides links to these new models, which include modeling phenomena important to liquid metal cooled fast reactors, molten salt reactors, high temperature gas-cooled reactors, and microreactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2018

Design and Analysis of Embedded I&C for a Fully Submerged Magnetically Suspended Impeller Pump

Improving nuclear reactor power system designs and fuel-processing technologies for safer and more efficient operation requires the development of new component designs. In particular, many of the advanced reactor designs such as the molten salt reactors and high-temperature gas-cooled reactors have operating environments beyond the capability of most currently available commercial components. To address this gap, new cross-cutting technologies need to be developed that will enable design, fabrication, and reliable operation of new classes of reactor components. The Advanced Sensor Initiative of the Nuclear Energy Enabling Technologies initiative is investigating advanced sensor and control designs that are capable of operating in these extreme environments. Under this initiative, Oak Ridge National Laboratory (ORNL) has been developing embedded instrumentation and control (I&C) for extreme environments. To develop, test, and validate these new sensing and control techniques, ORNL is building a pump test bed that utilizes submerged magnetic bearings to levitate the shaft. The eventual goal is to apply these techniques to a high-temperature (700°C) canned rotor pump that utilizes active magnetic bearings to eliminate the need for mechanical bearings and seals. The technologies will benefit the Next Generation Power Plant, Advanced Reactor Concepts, and Small Modular Reactor programs. In this paper, we will detail the design and analysis of the embedded I&C test bed with submerged magnetic bearings, focusing on the interplay between the different major systems. Then we will analyze the forces on the shaft and their role in the magnetic bearing design. Next, we will develop the radial and thrust bearing geometries needed to meet the operational requirements of the test bed. In conclusion, we will present some initial system identification results to validate the theoretical models of the test bed dynamics.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

Sister Rod Destructive Examinations (FY20) Appendix E: Mechanical Testing

This report documents work performed under the Spent Fuel and Waste Disposition’s Spent Fuel and Waste Science and Technology program for the US Department of Energy (DOE) Office of Nuclear Energy (NE). This work was performed to fulfill Level 2 Milestone M2SF-21OR010201032, “ORNL High Burnup Confirmatory Demo Sibling Rod Testing Results,” within work package SF-21OR01020103 and is an update to the work reported in M2SF-19ORO010201026 and M2SF-19OR010201028.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Sister Rod Destructive Examinations (FY20), Appendix F: Cyclic Integrated Reversible-Bending Fatigue Tests

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Status of Triso fuel irradiations in the advanced test reactor supporting high-temperature gas-cooled reactor designs - 18468

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven experiments on low-enriched uranium tri-structural isotropic (TRISO) particle fuel. Irradiations and fuel development are being accomplished at Idaho National Laboratory s Advanced Test Reactor to support development of the next generation of reactors in the United States. The experiments are being irradiated over several years to demonstrate and qualify new TRISO-coated particle fuel for use in high-temperature gas reactors, with goals of providing irradiation performance data to support fuel process development, qualifying fuel for normal operating conditions, supporting development and validation of fuel performance and fission product transport models and codes, and providing irradiated fuel and materials for post-irradiation examination and safety testing. The experiments are irradiated in an inert sweep gas atmosphere, with individual online temperature monitoring and control of each capsule and online fission product monitoring of effluent to track individual capsule fuel performance. The first experiment (AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Because the purpose of this experiment was to provide data on fission product migration and retention in the next generation reactor, the design of this experiment was significantly different from the first two experiments. The final experiment (AGR-5/6/7) is scheduled to begin irradiation in early summer 2017. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2018

Radioactivity and neutron damage calculators for research and test reactor neutron-activated materials - 25652

The Idaho National Laboratory Nuclear Science User Facilities program is proposing to launch a new online expert system that will assist users in planning experiments and writing proposals. The Combined Materials Experimentation Toolkit (CoMET) will include two nuclear calculators. The first of these, the Neutron Damage Calculator, will assist users in selecting reactor positions in order to obtain desired material damage in Damage Per Atom (DPA) and provide the required irradiation time to reach the damage levels. The second, the Radioactivity Calculator, will estimate the post-irradiation radioactivity levels of sample materials, which will inform technical staff of examination facility requirements and any potential scheduling delays. Both of these calculators were designed to simplify the information needed from users in order to properly plan and execute awarded proposals. This paper presents the primary purpose and methodologies implemented by these two calculators. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

Equibiaxial Flexure Strength of a Superfine-Grained Nuclear Graphite

The strength of advanced graphite is reported in accordance with ASTM D7846-16 using a two-parameter Weibull distribution for uniaxial strength testing. The rule of thumb for Weibull distributions is to use a minimum of 30 strength measurements to have a high level of confidence in the Weibull characteristic strength and the Weibull modulus. These large sample sets for statistical confidence are easily obtained for as-manufactured graphite, but determination of the Weibull two-factor parameters on graphite that has been exposed to neutron irradiation is nearly impossible due to irradiation testing restraints on specimen size and space limitations in test reactors for the accommodation of specimens. These restrictions have resulted in irradiation programs that use sub-sized specimens to measure uniaxial strength change, but the specimen geometries still hinder the number of replicate specimens for each irradiation condition. In contrast, this work used an advanced ceramic standard, ASTM C1499-05, as the foundation for equibiaxial strength testing of a superfine-grained nuclear graphite to investigate the irradiation-induced strength change and any changes to the two-parameter Weibull distribution. The goal was to investigate the effect of specimen size on the two-parameter Weibull distribution parameters in support of the small specimens and geometry desired for irradiation. A study of the effect of a reduced sample population was also undertaken to investigate the reliability of using the two-parameter Weibull distribution to study changes to the materials properties caused by neutron irradiation damage. The results of this work suggest that these equibiaxial specimens would be ideal for statistically significant studies of the effects of irradiation on the Weibull characteristic strength (but not the modulus) and for providing a method for surveillance specimen campaigns for future operating commercial reactors.

Campbell, Anne↗