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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 163 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Anode protection system for shutdown of solid oxide fuel cell system

An Anode Protection Systems for a SOFC system, having a Reductant Supply and safety subsystem, a SOFC anode protection subsystem, and a Post Combustion and slip stream control subsystem. The Reductant Supply and safety subsystem includes means for generating a reducing gas or vapor to prevent re-oxidation of the Ni in the anode layer during the course of shut down of the SOFC stack. The underlying ammonia or hydrogen based material used to generate a reducing gas or vapor to prevent the re-oxidation of the Ni can be in either a solid or liquid stored inside a portable container. The SOFC anode protection subsystem provides an internal pressure of 0.2 to 10 kPa to prevent air from entering into the SOFC system. The Post Combustion and slip stream control subsystem provides a catalyst converter configured to treat any residual reducing gas in the slip stream gas exiting from SOFC stack.

30 DIRECT ENERGY CONVERSION↗
DOE OSTI2019

Passive cooling to cold shutdown

A cooling system for a reactor module includes a reactor pressure vessel that houses primary coolant and a steam generator that lowers a temperature of the reactor pressure vessel by transferring heat from the primary coolant to a secondary coolant. The steam generator releases at least a portion of the secondary coolant as steam. Additionally, the cooling system includes a containment vessel that at least partially surrounds the reactor vessel in a containment region. The containment region is dry during normal operation of the reactor module. A controller introduces a source of water into the containment region in response to a non-emergency shut down of the reactor module. The source of water is located external to the containment vessel, and the water is introduced into the containment region after the steam generator has initially lowered the temperature of the reactor pressure vessel in response to releasing the steam.

Keller, Michael↗
DOE OSTI2023

Trepanning Graphite Samples at a Shutdown Magnox Reactor: A US-UK Joint Exercise for Nuclear Verification

In April 2022, a team sponsored by the US DOE Office of Nuclear Verification (ONV) shipped trepanning equipment to the nuclear power station at Trawsfynydd, Wales, UK to collect samples from the graphite moderator of Reactor 1. The sampling campaign to collect 48 graphite samples was completed in five weeks beginning July 11. Half the samples will be analyzed in the US to determine the total energy production of the facility. The other half will be analyzed by Magnox Ltd. to support decisions for options for the ultimate decommissioning path for the facility. This paper provides a description of the equipment and the operations required to trepan, collect, and package the samples while controlling the small amounts of radioactive graphite dust entrained on and within the tool. The most important sample parameter needed is the precise location of the sample, to within ±6 mm in a fuel channel, while the tool is deployed to between 12 and 20 meters below the operations floor; our means of achieving this precision is given.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2023

Validation of Pronghorn’s Subchannel code using the EBR-II shutdown heat removal tests

A Subchannel application (Pronghorn-SC) is developed in MOOSE, which affords the required flow field resolution, while still preserving an engineering-scale approach. This new solver can be natively coupled to Pronghorn and other MOOSE objects to enable full-core, multi-physics, multi-scale engineering studies. This work utilizes the EBR-II SHRT tests to validate the subchannel capabilities. Multi-scale and multi-physics coupling is used to improve the fidelity of the subchannel code calculations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

SCM overview & the EBR-II shutdown heat removal tests validation.

Pronghorn is an engineering-scale, coarse-mesh, thermal-hydraulics tool for supporting reactor-core simulations of advanced nuclear reactors. Most of the current efforts in Pronghorn have been devoted in developing porous finite-volume capabilities and adapting closure correlations for coarse-mesh thermal-hydraulics modeling. However, for liquid-metal reactors (LMRs) with wire-wrapped fuel pin assemblies, a pin-level thermal-hydraulic resolution is required for most safety case studies (pin rupture, channel blockage, etc.). For this purpose, a new Subchannel application is developed in MOOSE, which affords the required flow field resolution, while still preserving an engineering-scale approach. This new solver can be natively coupled to Pronghorn and other MOOSE objects to enable full-core, multi-physics, multi-scale engineering studies. This presentation presents the main features of the SCM code and demonstrates a validation case based on the EBR-II SHRT tests.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2019

Utilizing FLEX Equipment for O&M Cost Reduction in Nuclear Power Plants

There are many ways of reducing the O&M costs; this work presents an innovative framework of reducing O&M costs by utilizing the onsite Flex equipment at NPPs. Flex strategies were postulated by the U.S. Nuclear Regulatory Commission (NRC) in the wake of Fukushima Dai-ichi accident to address beyond-design-basis accidents and improve plant flexibility. Onsite Flex includes equipment such as portable pumps, generators, batteries, compressors, and other supporting equipment or tools, all stored in a dedicated and secure building designed to withstand external hazards. In the past years, several NPPs have invested in procuring and maintaining the onsite Flex asset that stands unutilized most of the time. Recently there have been active efforts to develop strategies through which NPPs can take credit for the Flex equipment. This work focuses on identifying areas where Flex equipment can be utilized during normal plant operation and develop a framework that would aide in reduction of O&M costs without impacting plant safety. Earlier work in the current project, published in INL/EXT 18 51531, had presented the risk- and cost-analysis framework for utilizing FLEX portable equipment during technical specification required shutdown due to component failure. The licensee event report (LER) database of the NRC shows that the commercial NPPs in the US reported 86 technical specification required shutdowns since year 2010. When a component failure or unavailability leads to a technical specification required shutdown, the NPPs suffers both direct costs in terms of loss of revenue arising from the loss of generation and indirect costs in form of reporting and inspection required by the NRC. The earlier work developed the following framework and implemented it on a demonstration probabilistic risk assessment (PRA) model to utilize the portable Flex equipment when a component failure could potentially lead to a technical specification required shutdown: 1. Identify the components, the failure or unavailability of which would result in a 10 CFR 50.73(a)(2)(i)A-reportability requirement, postulated by the NRC for technical specification required shutdown to be reported in NRC’s LER database. 2. Identify the Flex equipment that can be utilized as a standby to the failed component. 3. Develop a probabilistic risk assessment (PRA) model that incorporates the Flex equipment within the current plant PRA model. 4. Perform PRA calculations to determine change in core damage frequency and change in risk-informed allowable outage time. 5. Perform cost-benefit analysis to determine the economic feasibility of implementing the Flex equipment The current work, presented in this report, implement the above framework in a plant PRA model obtained from a US commercial nuclear power plant by prestaging onsite FLEX equipment as a redundancy to the existing auxiliary feedwater system, and pre-coordinating with offsite FLEX centers. The new configuration increased the allowable outage time by about a fifty percent overall for the diesel generator being out of service. The benefits to NPPs of utilizing the Flex equipment during normal operations, and performing the PRA developed in this work, include: 1. A risk-informed plant shutdown alternative to the current technical specification required shutdown. 2. Extension of allowable outage time to initiate technical specification required shutdown. 3. Reducing economic impact of component failure, avoiding plant shutdown, and maximizing generation. 4. Save the direct and indirect costs associated with technical specification required shutdown.

42 ENGINEERING↗
DOE OSTI2017

EPRI Guidance for Transitioning from Operation to Decommissioning at Nuclear Power Plants - 17438

A wide range of key activities are necessary after permanent shutdown of a nuclear power plant and before active dismantlement of the plant can begin. This period between permanent shutdown and the start of active dismantlement is typically referred to as the transition period (or the post-operational phase). In some cases, these transition activities may be prescribed by regulation or performed in the absence of regulations, while in other cases they may be more practically driven or even optional. In either case, it is often unclear what activities are needed and when/how those activities should be conducted to most efficiently transition from operational to decommissioning status while maintaining radiological safety and the protection of the public. The Electric Power Research Institute has prepared a report that presents experiences and guidance for transitioning from an operational to decommissioning status based on US and international decommissioning regulations and plant experiences. Regulations and experiences in a number of countries were considered, including US, Germany, France, Spain, and Switzerland. Best practices and lessons learned related to transitioning to decommissioning are identified, and guidance is presented for developing a plan to transition from normal operations to decommissioning for several scenarios, including: - Transition to DECON (immediate dismantling) for a planned shutdown; - Transition to SAFSTOR (deferred dismantling) for a planned shutdown; - Transition to DECON (immediate dismantling) for an unplanned shutdown; - Transition to SAFSTOR (deferred dismantling) for an unplanned shutdown. Further, the EPRI report contains guidance for currently operating plants that do not anticipate permanent shutdown in the near future to help minimize the cost and impact of an unexpected permanent shutdown if such a scenario were to occur. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Technical Analysis of Installed Micro-Combined Heat and Power Fuel-Cell System

Combined heat and power fuel cell systems (CHP-FCSs) provide consistent electrical power and hot water with greater efficiency and lower emissions than alternative sources. These systems can be used either as baseload, grid-connected, or as off-the-grid power sources. This report presents a technical analysis of 5 kWe CHP-FCSs installed in different locations in the U.S. At some sites as many as five 5 kWe system is used to provide up to 25kWe of power. Systems in this power range are considered “micro”-CHP-FCS. To better assess performance of micro-CHP-FCS and understand their benefits, the U.S. Department of Energy worked with ClearEdge Power to install fifteen 5-kWe PBI high temperature PEM fuel cells (CE5 models) in the commercial markets of California and Oregon. Pacific Northwest National Laboratory evaluated these systems in terms of their economics, operations, and technical performance. These units were monitored from September 2011 until June 2013. During this time, about 190,000 hours of data were collected and more than 17 billion data points were analyzed. Beginning in July 2013, ten of these systems were gradually replaced with ungraded systems (M5 models) containing phosphoric acid fuel cell technology. The new units were monitored until June 2014 until they went offline because ClearEdge was bought by Doosan at the time and the new manufacturer did not continue to support data collection and maintenance of these units. During these two phases, data was collected at once per second and data analysis techniques were applied to understand behavior of these systems. The results of this analysis indicate that systems installed in the second phase of this demonstration performed much better in terms of availability, consistency in generation, and reliability. The average net electrical power output increased from 4.1 to 4.9 kWe, net heat recovery from 4.7 to 5.4 kWth, and system availability improved from 94% to 95%. The average net system electric efficiency, average net heat recovery efficiency, and overall net efficiency of the system increased respectively from 33% to 36%, from 38% to 41%, and from 71% to 76%. The temperature of water sent to sit however reduced by about 16% from 51⁰C to 43 ⁰C. This was a control strategy and the temperature can be controlled depending on building heat demands. More importantly, the number of shutdowns and maintenance events required to keep the systems running at the manufacturer’s rated performance specifications were substantially reduced by about 76% (for 8 to 10 units running over a one-year period). From July 2012 to June 2013, there were eight CE5 units in operation and a total of 134 scheduled and unscheduled shutdowns took place. From July 2013 to June 2014, between two to ten units were in operation and only 32 shutdowns were reported (all unscheduled). In summary, the number of shutdowns reduced from 10 shutdowns per month on average for eight CE5units to an average of 2.7 shutdowns per month for M5 units (between two to ten units).

combined heat and power↗
DOE OSTI2013

Nuclear Industry Input to the Development of Concepts for the Consolidated Storage of Used Nuclear Fuel - 13411

EnergySolutions and its team partners, NAC International, Exelon Nuclear Partners, Talisman International, TerranearPMC, Booz Allen Hamilton and Sargent and Lundy, have carried out a study to develop concepts for a Consolidated Storage Facility (CSF) for the USA's stocks of commercial Used Nuclear Fuel (UNF), and the packaging and transport provisions required to move the UNF to the CSF. The UNF is currently stored at all 65 operating nuclear reactor sites in the US, and at 10 shutdown sites. The study was funded by the US Department of Energy and followed the recommendations of the Blue Ribbon Commission on America's Nuclear Future (BRC), one of which was that the US should make prompt efforts to develop one or more consolidated storage facilities for commercial UNF. The study showed that viable schemes can be devised to move all UNF and store it at a CSF, but that a range of schemes is required to accommodate the present widely varying UNF storage arrangements. Although most UNF that is currently stored at operating reactor sites is in water-filled pools, a significant amount is now dry stored in concrete casks. At the shutdown sites, the UNF is dry stored at all but two of the ten sites. Various types of UNF dry storage configurations are used at the operating sites and shutdown sites that include vertical storage casks that are also licensed for transportation, vertical casks that are licensed for storage only, and horizontally orientated storage modules. The shutdown sites have limited to nonexistent UNF handling infrastructure and several no longer have railroad connections, complicating UNF handling and transport off the site. However four methods were identified that will satisfactorily retrieve the UNF canisters within the storage casks and transport them to the CSF. The study showed that all of the issues associated with the transportation and storage of UNF from all sites in the US can be accommodated by adopting a staged approach to the construction of the CSF. Stage 1 requires only a cask storage pad and railroad interface to be constructed, and the CSF can then receive the UNF that is in transportable storage casks. Stage 2 adds a canister handling facility, a storage cask fabrication facility and an expanded storage pad, and enables the receipt of all canistered UNF from both operating and shutdown sites. Stage 3 provides a repackaging facility with a water-filled pool that provides flexibility for a range of repackaging scenarios. This includes receiving and repackaging 'bare' UNF into suitable canisters that can be placed into interim storage at the CSF, and enables UNF that is being received, or already in storage onsite, to be repackaged into canisters that are suitable for disposal at a geologic repository. The study used the 'Total System Model' (TSM) to analyze a range of CSF capacities and operating scenarios with differing parameters covering UNF pickup orders, one or more CSF sites, CSF start dates, CSF receipt rates and geologic repository start dates. The TSM was originally developed to model movement of UNF to the Yucca Mountain repository and was modified for this study to enable the CSF to become the 'gateway' to a future geologic repository. The TSM analysis enabled costs to be estimated for each scenario and showed how these are influenced by each of the parameters. This information will provide essential underpinning for a future Conceptual Design preparation. (authors)

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

Power Distribution Control Strategy of a Conceptual Soluble Boron Free SMR

In a commercial PWR, the two primary functions of control rods are to provide rapid shutdown and to control power distribution, whereas the initial excess reactivity associated with fuel loading is controlled by the soluble boron in the coolant and burnable absorbers. The required number of total control rods is determined by the shutdown margin of the reactor. The control rod grouping is also required to control power distribution. Therefore, removal of soluble boron requires extra reliance on the control rods to compensate excess reactivity. In a soluble boron free reactor which has highly negative MTC requires an increase in both worth and number of control rod assemblies to ensure adequate shutdown margin. In addition, since control rods are partially inserted into the reactor core, their presence adversely affects the power distribution and thus may make high peaking factors leading to an operating margin reduction. In this paper, the strategy of control rods operation to control excess reactivity and power distribution have been studied. Also, shutdown capability of the control rods was examined. A loading pattern for a Conceptual Soluble Boron Free Small Modular Reactor (CSBF-SMR) has been constructed. Two groups of regulating banks employ Ag-In-Cd to control excess reactivity and power distributions during the cycle operation and 1 group of shutdown bank made of B{sub 4}C were used.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Rigorous Two-Step Activation for Fusion Systems with PyNE - Paper 127

In fusion systems, high energy neutrons activate system components. After shutdown, the biological dose rate from photons emitted from these components must be estimated as a function of position and time for maintenance planning. High resolution results can be obtained using the mesh-based Rigorous Two-Step (R2S) method. A neutron transport simulation provides mesh-based neutron fluxes. Using these fluxes and an appropriate irradiation scenario, a nuclear inventory analysis code is used to determine the photon source density within each mesh volume element at any decay time after shutdown. The photon source density mesh is used as a source for photon transport. Photon fluxes are tallied at positions of interest and flux-to-dose rate conversion factors are used to estimate the biological dose rate. Previous work by the authors involved the creation of R2S-ACT, an open source R2S workflow, using the Direct Accelerated Geometry Monte Carlo (DAGMC) version of MCNP5 (DAG-MCNP5) for both transport steps and the ALARA nuclear inventory analysis code. Various use cases may require different combinations of transport and activation codes, and nearly all of the capabilities of R2S-ACT have broader utility. For these reasons, the individual components of R2S-ACT were reimplemented within PyNE. PyNE is a collaborative, trans-institutional, open source project consisting of Python, Cython, and C++ code pertinent to nuclear simulations and analysis. This work involved additions to the PyNE mcnp, mesh, material, alara, dagmc, and sampling modules. A lightweight r2s module was then added to connect the code from these modules. The r2s and sampling modules, together with DAG-MCNP5 and ALARA form a complete R2S workflow for obtaining shutdown dose rate results. Mesh operations, geometry discretization, physics code input/output parsing, and mesh-based source sampling capabilities have been added to PyNE generically, eliminating the need to duplicate this effort across multiple applications. The PyNE r2s module uses these capabilities to couple MCNP and ALARA to form an R2S workflow, which has been validated with the FNG shutdown dose rate benchmark. Future work will involve benchmarking the workflow using tetrahedral mesh and also non-analog sampling options. added so that they can be used interchangeably. This will allow for comparisons between results obtained by different physics codes with minimal human effort. Work will also be done to create and implement an importance sampling method for biasing neutron transport towards regions of phase space of high importance to photon dose rate tallies. This will improve the efficiency of obtaining shutdown dose rate results in large, highly-attenuating geometries. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗