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Grandy, C.

Publications and source records attributed to Grandy, C..

50 records · Page 3
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Mechanisms Engineering Test Loop - Phase I Status Report - FY2018 (Update to FY2017 Report)

This report documents the current status of the Mechanisms Engineering Test Loop (METL) as of August 2018. Currently, METL has completed Phase I construction, performed an initial heat-up of the METL piping, vessels, and components, baked out the facility under vacuum (minus the test vessels), addressed a variety of punchlist items, transferred sodium from 15 55-gallon drums to the dump tank, in the final commissioning/operations stage.

42 ENGINEERING↗
DOE OSTI2018

Gear Test Assembly - Status Report FY2018

The gear test assembly completed testing in air and distilled water successfully under a range of loading conditions. An external torque sensor assembly was designed and added to confirm the torque readings provided by the Parker control system. Hand calculations were performed to determine the time range that the heaters could be turned on without exceeding the ASME Boiler and Pressure Vessel code Fatigue Stress Limits for 1,000,000 cycles. Transient computational fluid dynamics analyses were performed to determine the heater cycle on/off time. Steady state computational fluid dynamics analyses were performed to determine the final temperature distribution throughout the insulated assembly using a thinner but more effective insulation configuration. The initial heating tests were completed (without insulation installed) to check the heater control system. Testing of the argon pressurized seal chambers were performed and indicated the need for re-machining of several components to eliminate some leaks. Nondestructive examination of the gears continued successfully using eddy current and ultrasonic inspection techniques.

42 ENGINEERING↗
DOE OSTI2018

Research and Development Roadmaps for Liquid Metal Cooled Fast Reactors – Draft for Public Comment

This report provides the roadmaps for two fast reactor concepts, the Sodium-cooled Fast Reactor (SFR) and the Lead-cooled Fast Reactor (LFR). The roadmaps are generic, i.e., not specific to a particular vendor’s design, but vendor design information may have been used as a representative of the concept family. The SFR technology is mature enough for commercial demonstration by the early 2030s, and the remaining key activities and R&D needs over the next five years are generally related to the completion of qualification of fuel and structural materials, validation of reactor design codes and methods, and support of the licensing frameworks. The LFR’s technology is instead less-mature and will be at the engineering demonstration stage by the early 2030s. Key LFR technology development activities for the acceleration of the engineering demonstration of the LFR will focus on resolving remaining design challenges and demonstrating the viability of systems and components in the integral system, which will be done in parallel with addressing the gaps shared with SFR technology. As discussed in the Advanced Demonstration and Test Reactor Options Study, the R&D approaches and timelines presented here assume that, for the first module demonstration, vendors would pursue a two-step licensing process based on 10CFR Part 50.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

A Summary of the Department of Energy’s Advanced Demonstration and Test Reactor Options Study

An assessment of advanced reactor technology options was conducted to provide a sound comparative technical context for future decisions by the U.S. Department of Energy (DOE) concerning these technologies. Strategic objectives were established that span a wide variety of important missions, and advanced reactor technology needs were identified based on recent DOE and international studies. A broad team of stakeholders from industry, academia, and government was assembled to develop a comprehensive set of goals, criteria, and metrics to evaluate advanced irradiation test and demonstration reactor concepts. Point designs of a select number of concepts were commissioned to provide a deeper technical basis for evaluation. The technology options were compared on the bases of technical readiness and the ability to meet the different strategic objectives. Using the study's evaluation criteria and metrics, an independent group of experts from industry, universities, and national laboratories scored each of the point designs. Pathways to deployment for concepts of varying technical maturities were estimated for the different demonstration systems with regard to cost, schedule, and possible licensing approaches. This study also presents the trade-offs that exist among the different irradiation test reactor options in terms of the ability to conduct irradiations in support of advanced reactor research and development and to serve potential secondary missions. The main findings of the study indicate the following: (1) for industrial process heat supply, a high-temperature gas-cooled reactor is the best choice because of the high outlet temperature of the reactor and its strong passive and inherent safety characteristics; (2) for resource utilization and waste management, a sodium-cooled fast reactor (SFR) is best because of the use of a fast flux to destroy actinides; (3) to realize the advantages of a promising but less-mature technology, a fluoride salt-cooled high-temperature reactor and a lead-cooled fast reactor fare about the same; (4) for fulfilling the needs of a materials test reactor, a SFR is considered best because of its ability to produce high fast flux, incorporate test loops, and provide additional large volumes for testing.

Advanced reactors↗
DOE OSTI2017

Supplement to the NUBOW-3D Manual

The NUBOW-3D code (NUBOW) developed in 1978 is a custom FORTRAN analysis code that models the complex mechanical interaction of the fuel assemblies and the core restraint system in response to the effects of thermal expansion and irradiation swelling and creep. NUBOW-3D models the fuel assembly ducts as beams with springs representing the contact between ducts. The core temperature profile and neutron flux profiles are provided at various times during the planned operation of the reactor. NUBOW then calculates the changing shapes of the fuel assemblies and their equilibrium positions within the core restraints. An optional post-processing step calculates the reactivity changes based on new positions of the fuel assemblies using input displacement reactivity WORTHS. Understanding this reactivity change is a key feature to the inherent safety of a fast reactor core. NUBOW was developed to support the design of limited free bow core restraint systems. The NUBOW software was modified in 1987 but the documentation of the upgrades was not completed due to abrupt funding cuts. In the last few years, the software has been ported to a Windows platform using a current FORTRAN compiler and additional changes have been made including consolidating the code to a single executable. While the original NUBOW manual [1] written in 1978 provides an excellent description of the code and input parameters, it lacks the additional features added in the upgrades and in some cases describes features that are no longer operable. This report attempts to bridge the information gap between the original manual and the current state of the software.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Mechanisms Engineering Test Loop - Phase 1 Status Report

This report documents the current status of the Mechanisms Engineering Test Loop (METL) as of the end of FY2016. Currently, METL is in Phase I of its design and construction. Once operational, the METL facility will test small to intermediate-scale components and systems in order to develop advanced liquid metal technologies. Testing different components in METL is essential for the future of advanced fast reactors as it will provide invaluable performance data and reduce the risk of failures during plant operation.

42 ENGINEERING↗
DOE OSTI2016

FASTER Test Reactor Preconceptual Design Report

The FASTER test reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

FASTER test reactor preconceptual design report summary

The FASTER reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Simulator platform for fast reactor operation and safety technology demonstration

A simulator platform for visualization and demonstration of innovative concepts in fast reactor technology is described. The objective is to make more accessible the workings of fast reactor technology innovations and to do so in a human factors environment that uses state-of-the art visualization technologies. In this work the computer codes in use at Argonne National Laboratory (ANL) for the design of fast reactor systems are being integrated to run on this platform. This includes linking reactor systems codes with mechanical structures codes and using advanced graphics to depict the thermo-hydraulic-structure interactions that give rise to an inherently safe response to upsets. It also includes visualization of mechanical systems operation including advanced concepts that make use of robotics for operations, in-service inspection, and maintenance.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Feasibility study on AFR-100 fuel conversion from uranium-based fuel to thorium-based fuel

Although thorium has long been considered as an alternative to uranium-based fuels, most of the reactors built to-date have been fueled with uranium-based fuel with the exception of a few reactors. The decision to use uranium-based fuels was initially made based on the technology maturity compared to thorium-based fuels. As a result of this experience, lot of knowledge and data have been accumulated for uranium-based fuels that made it the predominant nuclear fuel type for extant nuclear power. However, following the recent concerns about the extent and availability of uranium resources, thorium-based fuels have regained significant interest worldwide.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

A 100 MWe advanced sodium-cooled fast reactor core concept

An Advanced sodium-cooled Fast Reactor core concept (AFR-100) was developed targeting a small electrical grid to be transportable to the plant site and operable for a long time without frequent refueling. The reactor power rating was strategically decided to be 100 MWe, and the core barrel diameter was limited to 3.0 m for transportability. The design parameters were determined by relaxing the peak fast fluence limit and bulk coolant outlet temperature to beyond irradiation experience assuming that advanced cladding and structural materials developed under US-DOE programs would be available when the AFR-100 is deployed. With a de-rated power density and U-Zr binary metallic fuel, the AFR-100 can maintain criticality for 30 years without refueling. The average discharge burnup of 101 MWd/kg is comparable to conventional design values, but the peak discharge fast fluence of {approx}6x10{sup 23} neutrons/cm{sup 2} is beyond the current irradiation experiences with HT-9 cladding. The evaluated reactivity coefficients provide sufficient negative feedbacks and the reactivity control systems provide sufficient shutdown margins. The integral reactivity parameters obtained from quasi-static reactivity balance analysis indicate that the AFR-100 meets the sufficient conditions for acceptable asymptotic core outlet temperature following postulated unprotected accidents. Additionally, the AFR-100 has sufficient thermal margins by grouping the fuel assemblies into eight orifice zones. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Sodium Fast Reactor Safety and Licensing Research Plan – Volume II

Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗