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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 OSTI2017

Improvements to PHISICS/RELAP5-3D{sup C} Capabilities for Simulating HTGRs

Accurate, high-resolution full core simulations of prismatic high temperature reactors for burnup and transient analysis remain a challenge. The cores are geometrically large, but some features are quite small (tri-structural isotropic (TRISO) particles, burnable poison pellets). The treatment of both neutron scattering in graphite and resonance capture are complex and not adequately captured using the methods traditionally used in High Temperature Gas-cooled Reactors (HTGR) and commonly used in Light Water Reactors (LWR). For transient analysis, temperature feedback is a function of the TRISO fuel form, but full core models cannot resolve phenomena at this scale without averaging over space and energy. For burnup calculations, a second level of heterogeneity must be resolved to capture the local effects of fuel and burnable poisons, while accurately propagating their effects through and between blocks. The HTGR Methods and Simulation group has been utilizing the INL-developed PHISICS/RELAP5-3D{sup C} (P/R), code suite for participation in the Organization for Economic Cooperation and Development (OECD) Modular High Temperature Gas Cooled Reactor (MHTGR) Benchmark, funded by the US Department of Energy (DOE) under the Advanced Reactor Technologies (ART) program. The MHTGR-350 benchmark consists of several lattice, steady-state and transient exercises, and the P/R code has been successfully applied to most of the cases defined. In addition to this code-to-code verification activity, it is planned that P/R will be validated using data from several Loss of Forced Cooling (LOFC) tests performed at the High Temperature Test Reactor (HTTR) in a joint effort with Japan Atomic Energy Agency (JAEA) under the Civil Nuclear Energy Working Group (CNWG) project. These verification and validation simulations will identify code and model discrepancies, sensitivities, and establish the current state-of-the-art in HTGR simulation. The Department of Astronautical, Electrical and Energy Engineering (DIAEE), 'Sapienza' University has been involved in the development of P/R since 2014 through various graduate and PhD student projects. In the framework of this collaboration the P/R optimization for application on the OECD MHTGR-350 and HTTR models will be carried out.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

A Statistical Testing Approach for Quantifying Software Reliability; Application to an Example System

The U.S. Nuclear Regulatory Commission (NRC) encourages the use of probabilistic risk assessment (PRA) technology in all regulatory matters, to the extent supported by the state-of-the-art in PRA methods and data. Although much has been accomplished in the area of risk-informed regulation, risk assessment for digital systems has not been fully developed. The NRC established a plan for research on digital systems to identify and develop methods, analytical tools, and regulatory guidance for (1) including models of digital systems in the PRA's of nuclear power plants (NPPs), and, (2) incorporating digital systems in the NRC's risk-informed licensing and oversight activities. Under NRC's sponsorship, Brookhaven National Laboratory (BNL) explored approaches for addressing the failures of digital instrumentation and control (I and C) systems in the current NPP PRA framework. Specific areas investigated included PRA modeling digital hardware , development of a philosophical basis for defining software failure , and identification of desirable attributes of quantitative software reliability methods . Based on the earlier research, statistical testing is considered a promising method for quantifying software reliability. It is widely recognized that software failures are due to the triggering of pre-existing defects by the software's operational environment. Software defects can arise from errors made in user requirements or coding errors introduced during the developmental process. The software's operational environment includes factors such as the time history of digital system inputs, communication interfaces, the internal state of the digital system, and external conditions. Thus, software reliability is a function of both the number of pre-existing defects and the presence of a triggering condition caused by the manner in which the software is used. In this paper, we describe a statistical software-testing approach for quantifying software reliability and applied it to the loop-operating control system (LOCS) of an experimental loop of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The work involved collaboration between BNL and INL. The objectives of the study include: (1) Development of a statistical testing approach for estimating software failure probability on demand, the results of which are suitable for including in a probabilistic risk assessment (PRA); and, (2) Application of the approach to the LOCS to estimate its failure probability, and obtain insights into the feasibility, practicality, and usefulness of the estimation in models of digital systems for inclusion in nuclear power plants' PRAs. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Mechanical Properties of Irradiated U-10wt%Mo Alloy Degraded by Porosity Development

A plate-type nuclear fuel consisting of a solid monolithic foil of U-10wt%Mo is under development for use in the United States’ high performance research reactors. In support of developing this fuel, the fuel has been fabricated for the first time by a commercial fuel vendor and subsequently irradiated in a test reactor. This provides an opportunity to evaluate post-irradiation mechanical properties of commercially fabricated fuel. Four-point bend testing was conducted on the irradiated U-10Mo fuel and the data produced includes bending strength and Young’s modulus. Although the material behaves in a brittle manner due to the developed porosity, a general trend of strength and modulus reduction are found as fission density increases. The data produced is evaluated using both Weibull statistics and a modulus degradation model with recommendations provided.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Proton radiography, nuclear cross sections and multiple Coulomb scattering

The principles behind proton radiography including multiple Coulomb scattering are discussed for a purely imaginary square well nucleus in the eikonal approximation. It is found that a very crude model can reproduce the angular dependence of the cross sections measured at 24 GeV/c. The largest differences are ~3% for the 4.56 mrad data, and ~4% for the 6.68 mrad data. The prospect of understanding how to model deterministically high-energy proton radiography over a very large range of energies is promising, but it should be tested more thoroughly.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2019

Viability of Data Analytics to Ascertain Component Performance for Additive Manufacturing

The Transformational Challenge Reactor (TCR) program is leveraging additive manufacturing (AM) technologies to produce multiple nuclear components to be assembled into a fully functional microreactor core. AM was selected as the main manufacturing technology for TCR because it has the potential to disrupt the nuclear industry on two fronts: (1) it enables the manufacturing of very complex geometries with optimized and tailored material properties for the intended use of the component, opening up new options for reactor designs, and (2) it allows for a better understanding of the manufacturing process through real-time in situ monitoring, data analytics, and artificial intelligence, which can lead to a streamlined qualification and certification process. Within TCR, the development and deployment of a digital platform aims at addressing the latter opportunity. As part of this effort, the collection of pedigree datasets is vital, should these data be generated before, during or after the manufacturing process. This report focuses on the development of data analytics tools enabling qualitative assessment of the manufacturing processes to automate the identification of flaws in in situ monitoring imagery and to correlate those flaws to the resulting mechanical testing data. Through examples, the report gives an overview of the data collected and presents a path forward for the digital platform.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Implementation and Initial Testing of Advanced Processing and Analysis Algorithms for Correlated Neutron Counting

In order to improve the accuracy and capabilities of neutron multiplicity counting, additional quantifiable information is needed in order to address the assumptions that are present in the point model. Extracting and utilizing higher order moments (Quads and Pents) from the neutron pulse train represents the most direct way of extracting additional information from the measurement data to allow for an improved determination of the physical properties of the item of interest. The extraction of higher order moments from a neutron pulse train required the development of advanced dead time correction algorithms which could correct for dead time effects in all of the measurement moments in a self-consistent manner. In addition, advanced analysis algorithms have been developed to address specific assumptions that are made within the current analysis model, namely that all neutrons are created at a single point within the item of interest, and that all neutrons that are produced within an item are created with the same energy distribution. This report will discuss the current status of implementation and initial testing of the advanced dead time correction and analysis algorithms that have been developed in an attempt to utilize higher order moments to improve the capabilities of correlated neutron measurement techniques.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Strain gauge for testing microreactor hexagonal core blocks in the Single Primary Heat Extraction and Removal Emulator

To support the development and deployment of microreactor technologies, experiments that help verify and validate reactor systems and components are performed at non-nuclear test facilities. The Single Primary Heat Extraction Removal Emulator at Idaho National Laboratory is one of these test facilities and was used to monitor a test article that has a prototypic geometry of a heat-pipe cooled microreactor core block. To collect crucial temperature and strain data during testing, temperature and strain sensing fiber optics were embedded to the surface of the test article using an ultrasonic additive manufacturing technique. To support and provide benchmark strain data for the embedded sensors, a commercial resistive strain gauge was attached. This report will discuss the results from the deployment of the commercial strain gauge which includes the setup/attachment strategies, data acquisition, and analysis of the strain data.

36 MATERIALS SCIENCE↗
DOE OSTI2015

MANTA. An Integral Reactor Physics Experiment to Infer the Neutron Capture Cross Sections of Actinides and Fission Products in Fast and Epithermal Spectra

Neutron cross-sections characterize the way neutrons interact with matter. They are essential to most nuclear engineering projects and, even though theoretical progress has been made as far as the predictability of neutron cross-section models, measurements are still indispensable to meet tight design requirements for reduced uncertainties. Within the field of fission reactor technology, one can identify the following specializations that rely on the availability of accurate neutron cross-sections: (1) fission reactor design, (2) nuclear fuel cycles, (3) nuclear safety, (4) nuclear safeguards, (5) reactor monitoring and neutron fluence determination and (6) waste disposal and transmutation. In particular, the assessment of advanced fuel cycles requires an extensive knowledge of transuranics cross sections. Plutonium isotopes, but also americium, curium and up to californium isotope data are required with a small uncertainty in order to optimize significant features of the fuel cycle that have an impact on feasibility studies (e.g. neutron doses at fuel fabrication, decay heat in a repository, etc.). Different techniques are available to determine neutron cross sections experimentally, with the common denominator that a source of neutrons is necessary. It can either come from an accelerator that produces neutrons as a result of interactions between charged particles and a target, or it can come from a nuclear reactor. When the measurements are performed with an accelerator, they are referred to as differential since the analysis of the data provides the cross-sections for different discrete energies, i.e. σ(E i ), and for the diffusion cross sections for different discrete angles. Another approach is to irradiate a very pure sample in a test reactor such as the Advanced Test Reactor (ATR) at INL and, after a given time, determine the amount of the different transmutation products. The precise characterization of the nuclide densities before and after neutron irradiation allows to infer energy-integrated neutron cross sections, i.e. ∫₀ ∞ σ(E)φ(E)dE, where φ(E) is the neutron flux “seen” by the sample. This approach, which is usually defined and led by reactor physicists, is referred to as integral and is the object of this report. These two sources of information, i.e. differential and integral, are complementary and are used by the nuclear physicists in charge of producing the evaluated nuclear data files used by the nuclear community (ENDF, JEFF…). The generation of accurate nuclear data files requires an iterative process involving reactor physicists and nuclear data evaluators. This experimental program has been funded by the ATR National Scientific User Facility (ATR-NSUF) and by the DOE Office of Science in the framework of the Recovery Act. It has been given the name MANTRA for Measurement of Actinides Neutron TRAnsmutation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Review of fuel/cladding eutectic formation in metallic SFR fuel pins

Sodium-cooled Fast Reactors (SFRs) remain a strong contender amongst the Generation IV reactor concepts. Metallic fuel has been a primary fuel option for SFR designers in the US and was used extensively in the first generation of SFRs. One of the benefits of metallic fuel is its chemical compatibility with the coolant; unfortunately this compatibility does not extend to steel cladding at elevated temperatures. It has been known that uranium, plutonium, and rare earths diffuse with cladding constituents to form a low melting point fuel/cladding eutectic which acts to thin the cladding once the interfacial temperature rises above the system liquidus temperature. Since the 1960's, many experiments have been performed and published to evaluate the rate of fuel/cladding eutectic formation and the temperature above which melting will begin as a function of fuel/cladding interfacial temperature, time at temperature, fuel constituents (uranium, fissium or uranium (plutonium) zirconium), cladding type (stainless steel 316, stainless steel 306, D9 or HT9), beginning of life linear power, plutonium enrichment and burnup. The results of these tests, however, remain scattered across conference and journal papers spanning 50 years. The tests used to collect this data also varied in experimental procedure throughout the years. This paper will consolidate the experimental data into four groups of similar test conditions and expand upon the testing performed for each group in detail. A companion paper in PSA 2011 will discuss predictive correlations formulated from this database. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

A defense in depth approach for nuclear power plant accident management

An initiating event may lead to a severe accident if the plant safety functions have been challenged or operators do not follow the appropriate accident management procedures. Beyond design basis accidents are those corresponding to events of very low occurrence probability but such an accident may lead to significant consequences. The defense in depth approach is important to assure nuclear safety even in a severe accident. Plant Damage States (PDS) can be defined by the combination of the possible values for each of the PDS parameters which are showed on the nuclear power plant simulator. PDS is used to identify what the initiating event is, and can also give the information of safety system's status whether they are bypassed, inoperable or not. Initiating event and safety system's status are used in the construction of Containment Event Tree (CET) to determine containment failure modes by using probabilistic risk assessment (PRA) technique. Different initiating events will correspond to different CETs. With these CETs, the core melt frequency of an initiating event can be found. The use of Plant Damage States (PDS) is a symptom-oriented approach. On the other hand, the use of Containment Event Tree (CET) is an event-oriented approach. In this study, the Taiwan's fourth nuclear power plants, the Lungmen nuclear power station (LNPS), which is an advanced boiling water reactor (ABWR) with fully digitized instrumentation and control (I and C) system is chosen as the target plant. The LNPS full scope engineering simulator is used to generate the testing data for method development. The following common initiating events are considered in this study: loss of coolant accidents (LOCA), total loss of feedwater (TLOFW), loss of offsite power (LOOP), station blackout (SBO). Studies have indicated that the combination of the symptom-oriented approach and the event-oriented approach can be helpful to find mitigation strategies and is useful for the accident management. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Thermal Hydraulic Experimental Test Article: Second Year of Testing with Secondary Sodium System (Fiscal Year 2025 Final Report)

The Thermal Hydraulic Experimental Test Article (THETA) is currently installed in the Mechanisms Engineering Test Loop (METL) 28” test vessel #4. Both the primary and secondary sodium systems remain online to facilitate continued testing. This fiscal year, work was performed using a COMSOL Multiphysics magnetohydrodynamic model to characterize flow more accurately in the secondary electromagnetic flowmeters. Experimental campaigns were then performed to study the thermal hydraulic differences between sodium and water as a surrogate fluid in the THETA geometry as well as a study to better characterize and understand temperature oscillations that exist at the outlet of the core to the hot pool. A peer-reviewed article was published in the ASME Journal of Nuclear Engineering and Radiation Science detailing the THETA facility and providing an overview of a test that was performed with the primary and secondary system online [1]. Work continues to develop a database to house experimental THETA data to better facilitate collaboration with industry and laboratory partners for their use of the data for code benchmarking/validation. THETA remains fully operational and is positioned for continued testing in fiscal year 2026.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Comparison of Oxidation Performance of Graphite Samples in TGA versus Vertical Furnace

Oxidation influences the working life of nuclear graphite and can impact function during or after air (or steam) ingress accidents. Temperature, gas flow, oxygen content, and dimensional constraints at the microstructural and macrostructural levels can substantially influence the oxidation rate. The nature of these constraints must be considered when attempting to extrapolate performance from experimental results to service applications, but must also be evaluated in order to compare data among disparate experimental systems. This work compares oxidation test performance from a vertical furnace system (functioning as described by ASTM D7542) with tests using a thermogravimetric analyzer (TGA) adapted to operate in an analogous manner.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

The development and initial testing of the Ice Pig cleaning method for nuclear reprocessing plants - 15363

There are numerous situations throughout the nuclear industry where it would be advantageous to clean out enclosed systems, e.g. ducts and pipes, to remove sediments and radioactive contamination. However, conventional flushing methods result in the production of significant volumes of contaminated liquor. The wall Shear achieved with water flushing is insufficient to remove and carry dense deposits often found in nuclear facilities. Conventional mechanical pigs can be propelled through the duct by application of a driving medium such as compressed air or water. These are able to achieve very high wall shears, but can also get physically stuck where there is variation in the pipe / duct diameter or small radius bends. The field of application is best suited to straight constant diameter systems with no discontinuities such as Tee pieces, valves and in-pipe instrumentation. Driven by the desire to clean and displace radioactive materials from complex topology ducts and pipes, without the risk of having conventional pigs becoming lodged in a plant and coupled with the need to minimize effluent volumes, the UK Nuclear Industry led by Sellafield Limited (on behalf of the UK Nuclear Decommissioning Authority), has worked to demonstrate and underpin the use of the ice pigging technology for the nuclear industry. It is found that high ice fraction slurries flow as viscous liquids that exhibit a number of mechanical properties that offer much higher shear rates. The 'Ice Pig' exerts shear rates on pipe walls which are typically 1000 times greater than those achieved with water flowing at the same velocity. The 'Ice Pig' is able to negotiate almost any topology, flow through pipe networks of vastly differing diameter, through heat exchangers and in-line instrumentation etc. Experimental data and operational experience from other industries demonstrates that whilst there is high shear, the 'Ice Pig' does not damage the surface of the pipe wall. A further attractive characteristic of the 'Ice Pig' is that it eventually melts rendering it a simple effluent for processing by downstream processes. Due to the special nature of ice slurry, its unique rheological and physical characteristics, together with the complementary underpinning experimental work undertaken by Sellafield Limited enables the Ice Pigging process to move from an 'interesting' technology to one suitable for use by the nuclear industry for operations, Post Operational Clean Out (POCO) and decommissioning. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Methodology for Mechanical Property Testing on Fuel Cladding Using an Expanded Plug Wedge Test

To determine the tensile properties of irradiated fuel cladding in a hot cell, a simple test was developed at ORNL and is described fully in US Patent Application 20060070455, Expanded plug method for developing circumferential mechanical properties of tubular materials. This method is designed for testing fuel rod cladding ductility in a hot cell utilizing an expandable plug to stretch a small ring of irradiated cladding material. The specimen strain is determined using the measured diametrical expansion of the ring. This method removes many complexities associated with specimen preparation and testing. The advantages are the simplicity of measuring the test component assembly in the hot cell and the direct measurement of specimen strain. It was also found that cladding strength could be determined from the test results. The basic approach of this test method is to apply an axial compressive load to a cylindrical plug of polyurethane (or other materials) fitted inside a short ring of the test material to achieve radial expansion of the specimen. The diameter increase of the specimen is used to calculate the circumferential strain accrued during the test. The other two basic measurements are total applied load and amount of plug compression (extension). A simple procedure is used to convert the load circumferential strain data from the ring tests into material pseudo-stress-strain curves. However, several deficiencies exist in this expanded-plug loading ring test, which will impact accuracy of test results and introduce potential shear failure of the specimen due to inherited large axial compressive stress from the expansion plug test. First of all, the highly non-uniform stress and strain distribution resulted in the gage section of the clad. To ensure reliable testing and test repeatability, the potential for highly non-uniform stress distribution or displacement/strain deformation has to be eliminated at the gage section of the specimen. Second, significant compressive stresses were induced by clad bending deformation due to a clad bulging effect (or the barreling effect). The barreling effect caused very large localized shear stress in the clad and left testing material at a high risk of shear failure. The above combined effects will result in highly non-conservative predictions both in strength and ductility of the tested clad, and the associated mechanical properties as well. To overcome/mitigate the mentioned deficiencies associated with the current expansion plug test, systematic studies have been conducted. Through detailed parameter investigation on specific geometry designs, careful filtering of material for the expansion plug, as well as adding newly designed parts to the testing system, a method to reconcile the potential non-conservatism embedded in the expansion plug test system has been discovered. A modified expansion plug testing protocol has been developed based on the method. In order to closely resemble thin-wall theory, a general procedure was also developed to determine the hoop stress $\sigma_\theta$ in the tested ring specimen. A scaling factor called $\chi$-factor is defined to correlate the ring load F ring into hoop stress $\sigma_\theta$, and written as $\sigma_\theta = \chi \frac{F_{ring}} {tl}$. The generated stress-strain curve agrees very well with tensile test data in both the elastic and plastic regions.

expansion plug test↗
DOE OSTI2012

Supporting qualified database for uncertainty evaluation

Uncertainty evaluation constitutes a key feature of BEPU (Best Estimate Plus Uncertainty) process. The uncertainty can be the result of a Monte Carlo type analysis involving input uncertainty parameters or the outcome of a process involving the use of experimental data and connected code calculations. Those uncertainty methods are discussed in several papers and guidelines (IAEA-SRS-52, OECD/NEA BEMUSE reports). The present paper aims at discussing the role and the depth of the analysis required for merging from one side suitable experimental data and on the other side qualified code calculation results. This aspect is mostly connected with the second approach for uncertainty mentioned above, but it can be used also in the framework of the first approach. Namely, the paper discusses the features and structure of the database that includes the following kinds of documents: 1. The' RDS-facility' (Reference Data Set for the selected facility): this includes the description of the facility, the geometrical characterization of any component of the facility, the instrumentations, the data acquisition system, the evaluation of pressure losses, the physical properties of the material and the characterization of pumps, valves and heat losses; 2. The 'RDS-test' (Reference Data Set for the selected test of the facility): this includes the description of the main phenomena investigated during the test, the configuration of the facility for the selected test (possible new evaluation of pressure and heat losses if needed) and the specific boundary and initial conditions; 3. The 'QR' (Qualification Report) of the code calculation results: this includes the description of the nodalization developed following a set of homogeneous techniques, the achievement of the steady state conditions and the qualitative and quantitative analysis of the transient with the characterization of the Relevant Thermal-Hydraulics Aspects (RTA); 4. The EH (Engineering Handbook) of the input nodalization: this includes the rationale adopted for each part of the nodalization, the user choices, and the systematic derivation and justification of any value present in the code input respect to the values as indicated in the RDS-facility and in the RDS-test. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2010

The Statistical Analysis Techniques to Support the NGNP Fuel Performance Experiments

This paper describes the development and application of statistical analysis techniques to support the AGR experimental program on NGNP fuel performance. The experiments conducted in the Idaho National Laboratory’s Advanced Test Reactor employ fuel compacts placed in a graphite cylinder shrouded by a steel capsule. The tests are instrumented with thermocouples embedded in graphite blocks and the target quantity (fuel/graphite temperature) is regulated by the He-Ne gas mixture that fills the gap volume. Three techniques for statistical analysis, namely control charting, correlation analysis, and regression analysis, are implemented in the SAS-based NGNP Data Management and Analysis System (NDMAS) for automated processing and qualification of the AGR measured data. The NDMAS also stores daily neutronic (power) and thermal (heat transfer) code simulation results along with the measurement data, allowing for their combined use and comparative scrutiny. The ultimate objective of this work includes (a) a multi-faceted system for data monitoring and data accuracy testing, (b) identification of possible modes of diagnostics deterioration and changes in experimental conditions, (c) qualification of data for use in code validation, and (d) identification and use of data trends to support effective control of test conditions with respect to the test target. Analysis results and examples given in the paper show the three statistical analysis techniques providing a complementary capability to warn of thermocouple failures. It also suggests that the regression analysis models relating calculated fuel temperatures and thermocouple readings can enable online regulation of experimental parameters (i.e. gas mixture content), to effectively maintain the target quantity (fuel temperature) within a given range.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Extended Storage of High Burn-up Fuel Cask R and D Project - 17145

Under contract by EPRI and DOE, AREVA Federal Services LLC (AFS) is leading a team consisting of AREVA Inc. and Dominion Virginia Power (Dominion) to develop and conduct a full-scale confirmatory Research project with HBU fuel. A test plan was developed using an existing TN Americas (TN) TN-32 bolted storage cask with a modified, specially instrumented lid to be engineered and then loaded with multiple cladding types of HBU pressurized water reactor (PWR) UNF from Dominion's North Anna Power Station (this modified cask is identified as TN-32B). Seven specially designed thermocouple lances, each containing nine detectors, will be inserted into the designated fuel assembly guide tubes through the modified cask lid. The loaded cask with the monitoring system will be placed into dry storage on the existing North Anna Independent Spent Fuel Storage Installation (ISFSI) pad for a period of extended monitoring. In addition, design modifications to the lid have been incorporated to permit periodic gas sampling of the cask internals during the storage period on the ISFSI pad. The fuel selected for this research project ranges in average assembly burnup between 50 GWD/MTU to 67 GWD/MTU. The instrumentation used to collect the thermal data is very similar to in-core thermocouple detectors used at nuclear power plants and is considered very reliable for this application. Prior to cask loading at North Anna, a series of 'fit up' tests will be performed at the fabricator to ensure proper alignment, fit and function of the modified cask and related instrumentation and equipment. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Extended Storage of High Burn-up Fuel Cask R and D Project - 16591

Under contract by EPRI and DOE, AREVA Federal Services LLC (AFS) is leading a team consisting of AREVA Inc. and Dominion Virginia Power (Dominion) to develop and conduct a full-scale confirmatory Research project with HBU fuel. A test plan was developed using an existing AREVA Transnuclear (TN) TN-32 bolted storage cask with a modified, specially instrumented lid to be engineered and then loaded with multiple cladding types of HBU pressurized water reactor (PWR) UNF from Dominion's North Anna Power Station (this modified cask is identified as TN-32B). Seven specially designed thermocouple lances, each containing nine detectors, will be inserted into the designated fuel assembly guide tubes through the modified cask lid. The loaded cask with the monitoring system will be placed into dry storage on the existing North Anna Independent Spent Fuel Storage Installation (ISFSI) pad for a period of extended monitoring. In addition, design modifications to the lid have been incorporated to permit periodic gas sampling of the cask internals during the storage period on the ISFSI pad. The fuel selected for this research project ranges in average assembly burnup between 50 GWD/MTU to 67 GWD/MTU. The instrumentation used to collect the thermal data is very similar to in-core thermocouple detectors used at nuclear power plants and is considered very reliable for this application. Prior to cask loading at North Anna, a series of 'fit up' tests will be performed at the fabricator to ensure proper alignment, fit and function of the modified cask and related instrumentation and equipment. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗