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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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All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Light Water Reactor Sustainability Program Status of Silicon Carbide Joining Technology Development

Advanced, accident tolerant nuclear fuel systems are currently being investigated for potential application in currently operating light water reactors (LWR) or in reactors that have attained design certification. Evaluation of potential options for accident tolerant nuclear fuel systems point to the potential benefits of silicon carbide (SiC) relative to Zr-based alloys, including increased corrosion resistance, reduced oxidation and heat of oxidation, and reduced hydrogen generation under steam attack (off-normal conditions). If demonstrated to be applicable in the intended LWR environment, SiC could be used in nuclear fuel cladding or other in-core structural components. Achieving a SiC-SiC joint that resists corrosion with hot, flowing water, is stable under irradiation and retains hermeticity is a significant challenge. This report summarizes the current status of SiC-SiC joint development work supported by the Department of Energy Light Water Reactor Sustainability Program. Significant progress has been made toward SiC-SiC joint development for nuclear service, but additional development and testing work (including irradiation testing) is still required to present a candidate joint for use in nuclear fuel cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Accident Tolerant Fuels Series 1 (ATF-1) Irradiation Testing FY 2017 Status Report

This report contains a summary of irradiation testing of FCRD ATF-1 experiments performed at the INL in FY 2017. ATF-1 irradiation testing work performed in FY 2017 included fabrication of one ATF-1B drop in capsule ATF-1 series experiment and irradiation testing of ATF-1 capsules in the ATR. Additionally, five capsules met the sponsor-specified burnup targets, three of which were shipped from the ATR to the HFEF at INL.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Utilization of the LMP Methodology in Support of the VTR Conceptual Safety Design Report

The Versatile Test Reactor (VTR) is a fast spectrum test reactor currently being developed in the United States under the direction of the US Department of Energy (DOE), Office of Nuclear Energy. The VTR is utilizing a risk-informed performance-based (RIPB) approach for design support and authorization by the DOE, derived from recent efforts by the US industry led Licensing Modernization Project (LMP). This document contains an overview of the implementation of the LMP approach in support of the VTR Conceptual Safety Design Report (CSDR). The work reported here is the result of studies supporting a VTR conceptual design, cost, and schedule estimate for DOE-NE to make a decision on procurement. As such, it is preliminary. The VTR RIPB authorization approach utilizes information from the probabilistic risk assessment (PRA), coupled with deterministic analyses, to aid in decision-making regarding the identification and categorization of safety basis events (SBEs), the classification of structures, systems, and components (SSCs), and the evaluation of defense-in-depth (DID) adequacy. As part of initial reactor design efforts, a VTR conceptual design PRA was developed to support the RIPB process, which focused on at-power internal events, with scoping analyses for seismic and sodium fire hazards. In addition to supporting numerous design studies, preliminary results from the RIPB approach and the VTR conceptual design PRA were utilized as the basis of the VTR CSDR. The initial identification and categorization of SBEs, SSC classification, and DID evaluation were contained within the CSDR, which was submitted to DOE in 2019 as part of the CD-1 submittal package. Following review, DOE approved the CSDR in April 2020 and the CD-1 package in late 2020. Valuable experience was gained through the implementation of the RIPB approach for design and authorization during the VTR conceptual design phase, which is summarized in this document. To the extent possible, this experience has been shared with the advanced reactor industry, through publications and participation in licensing tabletops, in addition to informing DOE:NE advanced reactor regulatory development efforts. Furthermore, the approval of the CSDR by the DOE as part of CD-1 represents a significant milestone in the use of RIPB approaches for advanced reactor licensing.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1976

Plasma core reactor simulations using RF uranium seeded argon discharges

Experimental results are described in which pure uranium hexafluoride was injected into an argon-confined, steady-state, RF-heated plasma to investigate characteristics of plasma core nuclear reactors. The 80 kW (13.56 MHz) and 1.2 MW (5.51 MHz) rf induction heater facilities were used to determine a test chamber flow scheme which offered best uranium confinement with minimum wall coating. The cylindrical fused-silica test chamber walls were 5.7-cm-ID by 10-cm-long. Test conditions included RF powers of 2-85 kW, chamber pressures of 1-12 atm, and uranium hexafluoride mass-flow rates of 0.005-0.13 g/s. Successful techniques were developed for fluid-mechanical confinement of RF-heated plasmas with pure uranium hexafluoride injection.

Roman, W. C.↗
NASA NTRS1982

Study of a heat rejection system for the Nuclear Electric Propulsion (NEP) spacecraft

Two different heat pipe radiator elements, one intended for use with the power conversion subsystem of the NASA funded nuclear electric propulsion (NEP) spacecraft, and one intended for use with the DOE funded space power advanced reactor (SPAR) system were tested and evaluated. The NEP stainless steel/sodium heat pipe was 4.42 meters long and had a 1 cm diameter. Thermal performance testing at 920 K showed a non-limited power level of 3560 watts, well in excess of the design power of 2600 watts. This test verified the applicability of screen arteries for use in long radiator heat pipes. The SPAR titanium/potassium heat pipe was 5.5 meters long and had a semicircular crossection with a 4 cm diameter. Thermal performance testing at 775 K showed a maximum power level of 1.86 kW, somewhat short of the desired 2.6 kW beginning of life design requirement. The reduced performance was shown to be the result of the inability of the evaporator wall wick (shot blasted evaporator wall) to handle the required liquid flow.

Ernest, D. M.↗
DOE OSTI2015

BAYESIAN CALIBRATION OF SAFETY CODES USING DATA FROM SEPARATE-AND INTEGRAL EFFECTS TESTS

Large-scale system codes for simulation of safety performance of nuclear plants may contain parameters whose values are not known very accurately. In order to be able to use the results of these simulation codes with confidence, it is important to learn how the uncertainty on the values of these parameters affects the output of the codes. New information from tests or operating experience is incorporated into safety codes by a process known as calibration, which reduces uncertainty in the output of the safety code, and thereby improves its support for decision-making. Modern analysis capabilities afford very significant improvements on classical ways of doing calibration, and the work reported here implements some of those improvements. The key innovation has come from development of safety code surrogate model (code emulator) construction and prediction algorithms. A surrogate is needed for calibration of plant-scale simulation codes because the multivariate nature of the problem (i.e., the need to adjust multiple uncertain parameters at once to fit multiple pieces of new information) calls for multiple evaluations of performance, which, for a computation-intensive model, makes calibration very computation-intensive. Use of a fast surrogate makes the calibration processes used here with Markov Chain Monte Carlo (MCMC) sampling feasible. Moreover, most traditional surrogates do not provide uncertainty information along with their predictions, but the Gaussian Process (GP) based code surrogates used here do. This improves the soundness of the code calibration process. Results are demonstrated on a simplified scenario with data from Separate and Integral Effect Tests.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Update on the use of Neutron Absorbers for Storage Applications at Y-12

The Y-12 National Security Complex has a high density enriched uranium storage system at the Highly Enriched Uranium Materials Facility (HEUMF) that has been deployed for nearly 10 years. A robust neutron absorber system is provided through the design of Rackable Can Storage Boxes (RCSBs) and their configuration into storage racks. The requirements from ANSI/ANS-8.21, Use of Fixed Neutron Absorbers in Nuclear Facilities outside Reactors, are met with the RCSB and rack system at HEUMF through various design aspects, verification criteria, functional testing, and simulation modeling. Previous publications on a neutron absorber storage system at Y-12 were present at ANS in 2001 and 2002, when the RCSBs were still in a conceptual phase. The publications focused on ongoing studies at the time of what materials worked best for storage applications and how to test those materials, as well the practical design of the unit. The RCSBs have proven to provide a safe, static, and low maintenance storage system, with some minor design issues and several opportunities for future improvement.

42 ENGINEERING↗
DOE OSTI2016

Verification of the Time-dependent Transport Code TDKENO with the Monte Carlo Code KENO-VI

In this paper we focus on the computer code TDKENO that employs a hybrid methodology for solving the time dependent, three-dimensional Boltzmann transport equation. Originally, TDKENO used the Monte Carlo Transport code KENO V.a from SCALE to calculate flux shapes and fixed sources. KENO V.a is quick to compute these quantities but is limited to simple geometries. For this work TDKENO has been modified to support the Monte Carlo Transport code KENO-VI in addition to KENO V.a. KENO-VI allows generalized geometry input not possible in KENO V.a inputs. In order to verify the changes we ran TDKENO using KENO V.a and KENO-VI inputs then compared the results. Three different inputs were analyzed, the 16a1 benchmark,a simple TREAT model, and a transient model (2855) from the M8 calibration tests performed at INL.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

The Physics and Nuclear Nonproliferation Goals of WATCHMAN: A WAter CHerenkov Monitor for ANtineutrinos

Herein, this article describes the physics and nonproliferation goals of WATCHMAN, the WAter Cherenkov Monitor for ANtineutrinos. The baseline WATCHMAN design is a kiloton scale gadolinium-doped (Gd) light water Cherenkov detector, placed 13 kilometers from a civil nuclear reactor in the United States. In its first deployment phase, WATCHMAN will be used to remotely detect a change in the operational status of the reactor, providing a first- ever demonstration of the potential of large Gd-doped water detectors for remote reactor monitoring for future international nuclear nonproliferation applications. During its first phase, the detector will provide a critical large-scale test of the ability to tag neutrons and thus distinguish low energy electron neutrinos and antineutrinos. This would make WATCHMAN the only detector capable of providing both direction and flavor identification of supernova neutrinos. It would also be the third largest supernova detector, and the largest underground in the western hemisphere. In a follow-on phase incorporating the IsoDAR neutrino beam, the detector would have world-class sensitivity to sterile neutrino signatures and to non-standard electroweak interactions (NSI). WATCHMAN will also be a major, U.S. based integration platform for a host of technologies relevant for the Long-Baseline Neutrino Facility (LBNF) and other future large detectors. This white paper describes the WATCHMAN conceptual design,and presents the results of detailed simulations of sensitivity for the project's nonproliferation and physics goals. Furthermore, it describes the advanced technologies to be used in WATCHMAN, including high quantum efficiency photomultipliers, Water-Based Liquid Scintillator (WbLS), picosecond light sensors such as the Large Area Picosecond Photo Detector (LAPPD), and advanced pattern recognition and particle identification methods.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2019

Use of an ISOCS-based Waste Assay System for Decommissioning Waste at Trawsfynydd - 19526

Magnox UK Ltd. are decommissioning the former Magnox nuclear power stations in the UK, a process that generates a significant amount of radioactive waste, much of it in the Low Level or Very Low Level categories. It is therefore necessary to perform radiometric assay of the decommissioning waste in order to determine the disposal route and to identify items of waste that are Out of Scope of regulatory control and can therefore be free-released. They have recently commissioned a Canberra On Site Assay System (OSAS) for this purpose. OSAS is a high-resolution gamma-ray spectrometry based waste assay system, which makes use of the ISOCS methodology in order to perform the necessary efficiency calibration of the system for use when assaying the various different containers for the decommissioning waste employed at Trawsfynydd: 210 litre drums, Dolav boxes, wheeled bins and Pac-Tec bulk bags of heavier waste. This publication describes the approach used by Magnox in the processing of the decommissioning waste at Trawsfynydd. It also presents details of the OSAS system and how it was calibrated along with the results from the proof testing of the new system, involving comparison with the results from the assay of waste previously assayed using other monitors. (authors)

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

Development of a model based assessment process for qualification of embedded digital devices in NPP applications: research approach and current status - 263

The instrumentation and control (I and C) equipment used in currently operating U.S. nuclear power plants (NPPs) is primarily based on mature analog technologies that are progressing towards obsolescence. The continued reliance on this legacy analog technology, which is also being propagated into new NPP designs, imposes performance penalties and maintenance burdens in comparison with modern digital I and C instrumentation. In many instances, currently available I and C equipment contain embedded digital devices (EDDs) such as microprocessors and programmable logic devices. Experience in other industries, such as avionics, has shown that digital I and C equipment containing EDDs can provide significant benefits over analog-based equipment in terms of performance, reliability, and maintainability. In recent years, due to the high demand for digital technologies in the industrial I and C marketplace, it is becoming increasingly difficult for NPPs to acquire instrumentation that is not equipped with an EDD. However, the nuclear power industry has been slow to adopt digital technology, especially in safety-related systems, in large part as a result of regulatory concerns about common-cause failure (CCF) vulnerabilities of equipment with EDDs. Consequently, there is a clear need to develop cost effective qualification methods to contribute to the assessment of CCF vulnerability posed by EDDs in modern instrumentation that could be used in NPPs. This paper describes the research regarding qualification methods for equipment with EDDs that is sponsored by the Nuclear Energy Enabling Technologies (NEET) Advanced Sensors and Instrumentation (ASI) program of the U.S. Department of Energy (DOE). The purpose of the current research is to develop an effective approach employing science-based methods to resolve concerns about CCF vulnerability that serve to inhibit deployment of advanced instrumentation (e.g., sensors, actuators, microcontrollers) with EDDs in nuclear power applications. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Liquid sodium testing of in-house phased array EMAT transducer for L-wave applications

This paper describes the development of an in-house phased array EMAT transducer for longitudinal wave inspection in liquid sodium. The work presented herein is part of an undergoing project aimed at improving in-service inspection techniques for the ASTRID reactor project. The design process of the phased array EMAT probe is briefly explained and followed by a review of experimental test results. We first present test results obtained in the laboratory while the last part of the paper describes the liquid sodium testing and the produced ultrasound images. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Program of Fundamental-Interaction Research for the Ultracold-Neutron Source at the the WWR-M Reactor

The use of ultracold neutrons opens unique possibilities for studying fundamental interactions in particles physics. Searches for the neutron electric dipole moment are aimed at testing models of CP violation. A precise measurement of the neutron lifetime is of paramount importance for cosmology and astrophysics. Considerable advances in these realms can be made with the aid of a new ultracold-neutron (UCN) supersource presently under construction at Petersburg Nuclear Physics Institute. With this source, it would be possible to obtain an UCN density approximately 100 times as high as that at currently the best UCN source at the high-flux reactor of the Institute Laue–Langevin (ILL, Grenoble, France). To date, the design and basic elements of the source have been prepared, tests of a full-scale source model have been performed, and the research program has been developed. It is planned to improve accuracy in measuring the neutron electric dipole moment by one order of magnitude to a level of 10{sup −27} to 10{sup −28}e cm. This is of crucial importance for particle physics. The accuracy in measuring the neutron lifetime can also be improved by one order of magnitude. Finally, experiments that would seek neutron–antineutron oscillations by employing ultracold neutrons will become possible upon reaching an UCN density of 10{sup 3} to 10{sup 4} cm{sup −3}. The current status of the source and the proposed research program are discussed.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2015

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↗
DOE OSTI2015

Status Report on NEAMS System Analysis Module Development

Under the Reactor Product Line (RPL) of DOE-NE’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, an advanced SFR System Analysis Module (SAM) is being developed at Argonne National Laboratory. The goal of the SAM development is to provide fast-running, improved-fidelity, whole-plant transient analyses capabilities. SAM utilizes an object-oriented application framework MOOSE), and its underlying meshing and finite-element library libMesh, as well as linear and non-linear solvers PETSc, to leverage modern advanced software environments and numerical methods. It also incorporates advances in physical and empirical models and seeks closure models based on information from high-fidelity simulations and experiments. This report provides an update on the SAM development, and summarizes the activities performed in FY15 and the first quarter of FY16. The tasks include: (1) implement the support of 2nd-order finite elements in SAM components for improved accuracy and computational efficiency; (2) improve the conjugate heat transfer modeling and develop pseudo 3-D full-core reactor heat transfer capabilities; (3) perform verification and validation tests as well as demonstration simulations; (4) develop the coupling requirements for SAS4A/SASSYS-1 and SAM integration.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗