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At least 1,243 records · Page 69
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Means of Meeting Chapter 15 Requirements for a Small Modular Reactor Cooled by Lead-Bismuth

The mission of the Department of Energy Office of Nuclear Energy (DOE-NE) is to advance nuclear power as a resource capable of meeting the nation’s energy, environmental, and national security needs by resolving technical, cost, safety, proliferation resistance, and security barriers through Research, Development, and Demonstration (RD&D). To assist in accomplishing this, DOE-NE has established the Gateway for Accelerated Innovation in Nuclear (GAIN) to provide the nuclear community with access to the technical, regulatory, and financial support necessary to move innovative nuclear energy technologies toward commercialization. As part of this program, a GAIN Voucher was awarded to the Columbia Basin Consulting Group (CBCG) to develop for the Lead-Bismuth Reactor (LBR): (1) a means of meeting Chapter 15 requirements and (2) to perform an assessment against a gap analysis developed by the U.S. Nuclear Regulatory Commission (NRC) for Chapter 15. These two assessments will determine to what extent the LBR can adapt to or meet the Chapter 15 Accident Analysis requirements, and determine gaps that may still exist and what additional information is needed in order to close them. The CBCG LBR design is being developed with a high level of passive safety. It has the advantages of the classical liquid metal reactor, i.e. small size and passive safety, while having the additional advantage of a chemically inert coolant. The CBCG and the Pacific Northwest National Laboratory (PNNL) collaborated to develop an initial licensing pathway for the CBCG Lead-Bismuth conceptual reactor design. It was found that the reactor is well along the licensing pathway and is licensable although some additional design information is needed to establish a complete licensing basis. The missing information is identified in this report and it is noted that much of the missing information can be obtained from the Fast Flux Test Facility (FFTF), a DOE liquid metal test reactor. It was also found that the traditional Chapter 15 requirements for assessing the response of a reactor to possible accident sequences can be adapted to the LBR concept. The Chapter 15 review examined several classes of initiating events and accidents that are normally covered in Chapter 15: reactivity initiated events, reduction in heat removal events, local fuel failure events, and events initiated on the steam supply side. Although some design information is lacking at present, the LBR can be expected to respond as required by Chapter 15 requirements to almost all of the first three events. For events initiated on the steam side, design information is lacking and so conclusions cannot be reached at present. On the other hand, it should be quite possible to develop the design information for the steam supply side and the impact of events initiated on this side can be assessed in a subsequent study.

Wootan, David W.↗
DOE OSTI2012

A 48-month extended fuel cycle for the B and W mPower{sup TM} small modular nuclear reactor

The B and W mPower{sup TM} reactor is a small, rail-shippable pressurized water reactor (PWR) with an integral once-through steam generator and an electric power output of 150 MW, which is intended to replace aging fossil power plants of similar output. The core is composed of 69 reduced-height, but otherwise standard, PWR assemblies with the familiar 17 x 17 fuel rod array on a 21.5 cm inter-assembly pitch. The B and W mPower core design and cycle management plan, which were performed using the Studsvik core design code suite, follow the pattern of a typical nuclear reactor fuel cycle design and analysis performed by most nuclear fuel management organizations, such as fuel vendors and utilities. However, B and W is offering a core loading and cycle management plan for four years of continuous power operations without refueling and without the hurdles of chemical shim. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

An extended conventional fuel cycle for the B and W mPower{sup TM} small modular nuclear reactor

The B and W mPower{sup TM} reactor is a small pressurized water reactor (PWR) with an integral once-through steam generator and a thermal output of about 500 MW; it is intended to replace aging fossil power plants of similar output. The core is composed of 69 reduced-height PWR assemblies with the familiar 17 x 17 fuel rod array. The Babcock and Wilcox Company (B and W) is offering a core loading and cycle management plan for a four-year cycle based on its presumed attractiveness to potential customers. This option is a once-through fuel cycle in which the entire core is discharged and replaced after four years. In addition, a conventional fuel utilization strategy, employing a periodic partial reload and shuffle, was developed as an alternative to the four-year once-through fuel cycle. This study, which was performed using the Studsvik core design code suite, is a typical multi-cycle projection analysis of the type performed by most fuel management organizations such as fuel vendors and utilities. In the industry, the results of such projections are used by the financial arms of these organizations to assist in making long-term decisions. In the case of the B and W mPower reactor, this analysis demonstrates flexibility for customers who consider the once-through fuel cycle unacceptable from a fuel utilization standpoint. As expected, when compared to the once-through concept, reloads of the B and W mPower reactor will achieve higher batch average discharge exposure, will have adequate shut-down margin, and will have a relatively flat hot excess reactivity trend at the expense of slightly increased peaking. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Modular base for an antenna array

A base configured to be joined with other bases to form a substrate for an antenna array comprises a body, a plurality of male interconnecting features, and a plurality of female interconnecting features. The body includes a front surface and a rear surface and a plurality of edges positioned therebetween. The front surface or the rear surface is configured to retain an antenna. The male interconnecting features of a first base connect with the female interconnecting features of a second base when the first base is joined with the second base to form the substrate or a portion of the substrate.

Furman, Nicholas J.↗
DOE OSTI2023

Modular, disposable 3D printed microfluidic membrane system for separation and purification

The present disclosure relates to a membrane extraction apparatus for extracting a component from a first liquid. The apparatus may incorporate a housing comprised of first and second mating housing halves, with each housing half having an open faced channel formed therein such that the channels at least partially overlay one another when the two housing halves are secured together. A membrane filter is disposed between the two housing halves to overlay the open faced channels. The membrane filter extracts the component from the first liquid and transfers the component into the second liquid as the first and second liquids flow through the first and second housing halves.

Servis, Anna G.↗
DOE OSTI2013

Life Extension Program for the Modular Caustic Side Solvent Extraction Unit at Savannah River Site - 13179

Caustic Side Solvent Extraction (CSSX) is currently used at the U.S. Department of Energy (DOE) Savannah River Site (SRS) for removal of cesium from the high-level salt-wastes stored in underground tanks. Currently, the Actinide Removal Process (ARP) and the CSSX process are deployed in the (ARP)/Modular CSSX Unit (MCU), to process salt waste for permanent disposition. The CSSX technology utilizes a multi-component organic solvent and annular centrifugal contactors to extract cesium from alkaline salt waste. The original plant was permitted for a three year design life; however, given the successful operation of the plant, a life extension program was completed to continue operations. The program included detailed engineering analyses of the life-expectancy of passive and active components, resulting in component replacement and/or maintenance and monitoring program improvements. The program also included a review of the operations and resulted in a series of operational improvements. Since the improvements have been made, an accelerated processing rate has been demonstrated. In addition, plans for instituting a next-generation solvent are in place and will enhance the decontamination factors. (author)

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

Mathematical Calibration of Modular Plastic Scintillators for Various Assay Applications - 17322

Plastic scintillators provide an inexpensive option to assay and characterize radioactive material. Due to the low cost it is generally possible to an array of large numbers of plastic scintillators to achieve high measurement efficiencies. The beta only thin plastic scintillators used in the Argos contamination monitors provide a convenient and reliable scintillator package that can be grouped together to form an assay system. Mathematical calibrations, using MCNP, have been produced for this detector that allow for rapid construction of various custom measurement geometries. The calibrations were constructed by comparing a series of measurements made with a number of beta and beta/gamma emitting nuclides. Characterization of the beta scintillators utilized beta emitting nuclides that included spectra from low average energy, e.g. C-14, to relatively high average energy, e.g. Sr/Y-90. In order to perform the calibration, custom versions of the standard Argos scintillators were produced that provide an MCA output for spectroscopic investigation. The spectra were then compared to detailed MCNP models in an iterative process to energy calibrate and efficiency calibrate each detector type. The validated models and energy calibrations were then tested against a number of volumetric geometries to confirm that the model would translate well to any other geometry. These validated models can then be used to rapidly construct custom measurement geometries for material measurement and screening. Two examples of measurement applications include a beta screening system for dirt on a conveyor belt and a flow through beta liquid assay system. The mathematical calibration process, validation results, and a number of application examples are discussed. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

The hydrogen behavior analysis in the water-ingress accident of the 250 MW pebble-bed modular high temperature gas-cooled reactor - 18639

The demonstration power plant of Chinese 250 MW high temperature gas-cooled reactor pebbled-bed module (HTR-PM) which is being under construction in the Shidao Bay of Shandong province has made breakthrough in all of the critical technologies, and is willing to operate in 2017. HTR-PM does not have the risk of large amount of hydrogen generation resulted from the zirconium water reaction due to its special design of graphite core structure and helium coolant of the primary circuit, however, considering that the HTR-PM has a small containment and does not design the deliberate ignitions which is the main hydrogen management measure in light water reactors, it is still necessary to study the hydrogen behavior in the water-ingress accident in which hydrogen will probably be generated. In this paper, the hydrogen generation amount and its distribution in the HTR-PM containment are analyzed with the TINTE code and GASFLOW code respectively in the water ingress accident caused by the steam generator heating tubes rupture. The calculation results indicate that there are total about 16.6 kg and 21 kg hydrogen produced due to graphite corrosion in the typical design basis water ingress accident and beyond design basis accident. The containment sub-pressure ventilation system makes sure the hydrogen concentration in the containment is less than 0.2% during the accident. There is no hydrogen explosion risk in the HTR-PM. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

MOOSE: A Modular Platform for Fission and Fusion Multiphysics

The Multiphysics Object-Oriented Simulation Environment (MOOSE) Framework, as well as MOOSE-based simulation tools, have accelerated the development of fission energy and advanced reactor technologies through the United States Department of Energy, Office of Nuclear Science, Nuclear Energy Advanced Modeling & Simulation (NEAMS) Program. MOOSE contains a complete platform of multiphysics simulation capabilities, capable of running on massively parallel systems, and is developed in an open-source manner with great attention paid to high-quality software quality assurance practices. This overall approach could greatly benefit the fusion energy community, which requires rapid design iteration and improvement in order to facilitate the successful development of fusion as an alternative energy source to fossil fuels. In the first half of this talk, applications of MOOSE and MOOSE-based tools for advanced reactor designs will be showcased, as well as MOOSE ecosystem infrastructure (such as the NEAMS Virtual Test Bed) that enables and accelerates fission reactor design. In the second half, a discussion of how the MOOSE approach to modeling and simulation is currently being applied internationally in fusion energy research and development at the United Kingdom Atomic Energy Authority will be discussed, and ongoing/future domestic research efforts will be highlighted.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗