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DOE OSTI2015

Radiation-Hardened Circuitry Using Mask-Programmable Analog Arrays. Final Report

As the recent accident at Fukushima Daiichi so vividly demonstrated, telerobotic technologies capable of withstanding high radiation environments need to be readily available to enable operations, repair, and recovery under severe accident scenarios where human entry is extremely dangerous or not possible. Telerobotic technologies that enable remote operation in high dose rate environments have undergone revolutionary improvement over the past few decades. However, much of this technology cannot be employed in nuclear power environments due the radiation sensitivity of the electronics and the organic insulator materials currently in use. This is the final report of the activities involving the NEET 2 project Radiation Hardened Circuitry Using Mask-Programmable Analog Arrays. We present a detailed functional block diagram of the proposed data acquisition system, the thought process leading to technical decisions, the implemented system, and the tested results from the systems. This system will be capable of monitoring at least three parameters of importance to nuclear reactor monitoring: temperature, radiation level, and pressure.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2020

Investigations on the Thermal-hydraulic Behavior of Accident Tolerant Fuel Cladding Materials

Since the 2011 Fukushima accident, significant research has been devoted to developing accident-tolerant fuel (ATF) cladding materials. These investigations have mostly focused on the ability to resist runaway steam oxidation and retain mechanical strength and structural integrity under thermal shocks. However, much remains unknown about the materials’ thermal-hydraulic behavior, particularly under light-water reactor (LWR) operating conditions. Two phenomena that determine safety margins in both normal and off-normal operating conditions - i.e., the critical heat flux (CHF) and Leidenfrost point (LFP) temperature - have not been investigated thoroughly for ATF materials. Surface properties, such as wettability and roughness, are known to influence pool and flow boiling CHF, as well as the LFP temperature. However, little is known about the surface wettability of ATF materials, particularly at LWR pressure and temperature. Little has been done on flow-boiling CHF and quenching heat transfer, especially for droplet quenching. Nonetheless, it is known that CHF in transient conditions, e.g., an exponentially escalating power transient, can be significantly different from that expected in steady-state operation. Very few studies have investigated transient CHF, either on ATF materials or under LWR pressure and temperature. The thorough understanding of transient CHF under prototypical reactor conditions will benefit not only the deployment of ATF but also the upcoming national efforts to study transient ATF behavior in the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory. In summary, as a part of ATF development efforts, there is an urgent need to understand how and how much these materials may affect two-phase heat transfer phenomena in nuclear reactor conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2019

Report on plasma jet printer for sensor fabrication with process parameters optimized by simulation input

The In-Pile Instrumentation program conducts research to develop instrumentation that provides information regarding test conditions, as well as fuel and materials properties during irradiation. Such instruments must be miniaturized, robust to the nuclear reactor environment, and provide real-time data aligned with user needs. Towards this end, additive manufacturing techniques such as aerosol jet printing and Plasma Jet Printing, have demonstrated significant potential for the development of advanced in-pile instrumentation. Process control during the manufacturing process is critical in order to consistently create reliable sensors. This report details experimental and computational work aimed towards improved process control for robust sensor production using printing techniques.

36 MATERIALS SCIENCE↗

DOE OSTI2022

HYBRID Modeling of DETAIL Experimental Facility

Idaho National Laboratory continues to be at the forefront of advanced reactor systems and integrated energy systems research. Through the Thermal Energy Distribution System (TEDS) and Microreactor Agile Non-nuclear Test Bed (MAGNET) experiments in the Dynamic Energy Transport and Innovation Laboratory (DETAIL) within the Engineering Systems Laboratory, a mix of digital and physical testing systems can explore the operation of thermally and electrically integrated systems. Modeling of these systems has been done using the HYBRID modeling repository of the Integrated Energy Systems program. DETAIL also contains high temperature steam electrolysis (HTSE) units to produce hydrogen and generate the load imposed on MAGNET and TEDS. The dynamic modeling capabilities within the IES program describe the DETAIL components as-built and can be used to evaluate controls, process flows, and overall system conditions.

08 HYDROGEN↗

NASA NTRS1972

NASA-Lewis experiences with multigroup cross sections and shielding calculations

The nuclear reactor shield analysis procedures employed at NASA-Lewis are described. Emphasis is placed on the generation, use, and testing of multigroup cross section data. Although coupled neutron and gamma ray cross section sets are useful in two dimensional Sn transport calculations, much insight has been gained from examination of uncoupled calculations. These have led to experimental and analytic studies of areas deemed to be of first order importance to reactor shield calculations. A discussion is given of problems encountered in using multigroup cross sections in the resolved resonance energy range. The addition to ENDF files of calculated and/or measured neutron-energy-dependent capture gamma ray spectra for shielding calculations is questioned for the resonance region. Anomalies inherent in two dimensional Sn transport calculations which may overwhelm any cross section discrepancies are illustrated.

Lahti, G. P.↗

NASA NTRS1972

NASA-Lewis experiences with multigroup cross sections and shielding calculations.

Description of the nuclear reactor shield analysis procedures employed at NASA-Lewis Research Center. Emphasis is placed on the generation, use, and testing of multigroup cross-section data. Although coupled neutron and gamma ray cross-section sets are useful in two-dimensional Sn transport calculations, much insight has been gained from examination of uncoupled calculations. These have led to experimental and analytic studies of areas deemed to be of first-order importance to reactor shield calculations. A discussion is given of problems encountered in using multigroup cross sections in the resolved resonance energy range. The addition to ENDF files of calculated and/or measured neutron-energy-dependent capture gamma ray spectra for shielding calculations is questioned for the resonance region. Anomalies inherent in two-dimensional Sn transport calculations which may overwhelm any cross-section discrepancies are illustrated.

Lahti, G. P.↗

DOE OSTI2022

The Nuclear Fuel Cycle: Safe Management of Spent Nuclear Fuel

The aim for storage of spent nuclear fuel (SNF) either in wet or in dry storage systems is to ensure general safety objective s are met throughout a desired storage period. Staff at the Savannah River National Laboratory (SRNL), in collaborations with partners at other national laboratories, industry research organizations, and the University of South Carolina (UofSC), have performed materials aging testing and analyses, and have established nuclear materials aging management programs to support extended periods of safe storage of research reactor (RR) SNF and of commercial power reactor (PR) SNF pending ultimate disposal. Several example challenges include susceptibility of aluminum SNF from research reactors to corrosion in poor quality water (wet storage). In dry storage, aluminum SNF can release hydrogen via radiolysis of the hydrated oxides on the aluminum cladding. Austenitic stainless steel canisters used for dry storage are susceptible to chloride-induced stress corrosion cracking (outside-in attack) that threaten the confinement boundary provided by the canister. This paper further describes these challenges, among others, and the formulated solutions to support extended safe storage of SNF.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

Design and Scaling of the Natural Convection Shutdown Heat Removal Test Facility

The Natural convection Shutdown heat removal Test Facility (NSTF) at Argonne National Laboratory (ANL) reflects a 1/2 scale model of one conceptual design for passive safety in advanced reactors. The project was initiated in 2010 primarily to conduct ex-vessel, passive decay heat removal experiments in support of the Advanced Reactor Concepts (ARC), Small Modular Reactor (SMR), and Next Generation Nuclear Plant (NGNP) programs while also generating data for code validation purposes. The facility successfully demonstrated scoping objectives in late 2013, and is expected to begin testing by early 2014. The following paper summarizes some of the key design and scaling considerations used in construction of the experimental facility, along with an overview of the current instrumentation and data acquisition methods. Details of the distributed fiber optic temperature system will be presented, which introduces a level of data density suitable for CFD validation and is a first-of-its-kind for largescale thermal hydraulics facilities.

Lisowski, Darius D.↗

DOE OSTI2016

A Methodology for the Development of a Reliability Database for an Advanced Reactor Probabilistic Risk Assessment

GE Hitachi Nuclear Energy (GEH) and Argonne National Laboratory are currently engaged in a joint effort to modernize and develop probabilistic risk assessment (PRA) techniques for advanced non-light water reactors. At a high level the primary outcome of this project will be the development of next-generation PRA methodologies that will enable risk-informed prioritization of safety- and reliability-focused research and development, while also identifying gaps that may be resolved through additional research. A subset of this effort is the development of a reliability database (RDB) methodology to determine applicable reliability data for inclusion in the quantification of the PRA. The RDB method developed during this project seeks to satisfy the requirements of the Data Analysis element of the ASME/ANS Non-LWR PRA standard. The RDB methodology utilizes a relevancy test to examine reliability data and determine whether it is appropriate to include as part of the reliability database for the PRA. The relevancy test compares three component properties to establish the level of similarity to components examined as part of the PRA. These properties include the component function, the component failure modes, and the environment/boundary conditions of the component. The relevancy test is used to gauge the quality of data found in a variety of sources, such as advanced reactor-specific databases, non-advanced reactor nuclear databases, and non-nuclear databases. The RDB also establishes the integration of expert judgment or separate reliability analysis with past reliability data. This paper provides details on the RDB methodology, and includes an example application of the RDB methodology for determining the reliability of the intermediate heat exchanger of a sodium fast reactor. The example explores a variety of reliability data sources, and assesses their applicability for the PRA of interest through the use of the relevancy test.

Grabaskas, Dave↗

DOE OSTI2015

PIE on Safety-Tested AGR-1 Compact 5-1-1

Post-irradiation examination (PIE) is being performed in support of tristructural isotropic (TRISO) coated particle fuel development and qualification for High-Temperature Gas-cooled Reactors (HTGRs). AGR-1 was the first in a series of TRISO fuel irradiation experiments initiated in 2006 under the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program; this work continues to be funded by the Department of Energy's Office of Nuclear Energy as part of the Advanced Reactor Technologies (ART) initiative. AGR-1 fuel compacts were fabricated at Oak Ridge National Laboratory (ORNL) in 2006 and irradiated for three years in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR) to demonstrate and evaluate fuel performance under HTGR irradiation conditions. PIE is being performed at INL and ORNL to study how the fuel behaved during irradiation, and to examine fuel performance during exposure to elevated temperatures at or above temperatures that could occur during a depressurized conduction cooldown event. This report summarizes safety testing of irradiated AGR-1 Compact 5-1-1 in the ORNL Core Conduction Cooldown Test Facility (CCCTF) and post-safety testing PIE.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Evaluation of the ring tension test (RTT) for robust determination of material strengths

The ring tension test (RTT) is an experimental method for determining mechanical behavior in a material’s circumferential or hoop direction. It is a crucial test for testing anisotropic materials with tube geometry, such as nuclear fuel cladding or irradiated pipes. Several RTT configurations exist, each with their own advantages and disadvantages. However, this test is significantly more complex than traditional tensile testing and can be especially sensitive to small differences and inconsistencies in the test setup and geometry, ultimately affecting the derived mechanical properties. Previous research has focused on method development, and little work has been done on understanding the subtle differences between an ideal test and experiments, specifically when the tests are performed on highly irradiated materials in hot cells. In this work, a finite element-based investigation of the RTT is conducted to fill this gap. The two most used test configurations are investigated, comparing their ability to determine accurate material strengths through plastic deformation. Several non-ideal conditions and uncontrollable effects which are likely to occur during experimental testing such as machining tolerances, variations of specimen geometry from nominal dimensions, rotation of specimens and fixturing, and other test setup discrepancies are studied. The sensitivity of measured strengths to these conditions is presented. A mechanics-based approach to describing and correcting raw data to determine actual strengths is also included for one of the configurations, resulting in a robust correction method with highly accurate material strength measurements. Finally, based on these analyses, the hemicylindrical mandrel configuration is recommended with a gauge region oriented at a 45° angle.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

NASA NTRS1969

Comments on the feasibility of developing gas core nuclear reactors

Recent developments in the fields of gas core hydrodynamics, heat transfer, and neutronics indicate that gas core nuclear rockets may be feasible from the point of view of basic principles. Based on performance predictions using these results, mission analyses indicate that gas core nuclear rockets may have the potential for reducing the initial weight in orbit of manned interplanetary vehicles by a factor of 5 when compared to the best chemical rocket systems. In addition, there is a potential for reducing total trip times from 450 to 500 days for chemical systems to 250 to 300 days for gas core systems. The possibility of demonstrating the feasibility of gas core nuclear rocket engines by means of a logical series of experiments of increasing difficulty that ends with ground tests of full scale gas core reactors is considered.

Rom, F. E.↗

DOE OSTI2016

Experimental validation on the operational and cooling performance of the passive safety system with Atlas

After the Fukushima accident, passive safety systems are considered as the most promising alternatives to reinforce the safety and reliability of an ultimate heat removal system without any operator actions. In Korea, a passive auxiliary feedwater system (PAFS) was developed and adopted to a GEN-III+ nuclear power plant, APR+ to completely replace the conventional auxiliary feedwater system. The integral effect tests were performed to validate the cooling and operational performance of PAFS, coupled with the reactor coolant systems using the ATLAS facility. The representative design basis accidents (DBAs) such as a feedwater line break, a steam line break, and a steam generator tube rupture transients were experimentally simulated in this study. In addition, an integral effect test was performed for simulating a station blackout transient as a limiting beyond DBA. From the integral effect test results, it could be concluded that the APR+ has the capability of coping with the hypothetical DBA and bDBA by adopting PAFS and proper set-points of its operation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2024

AGC-4 Specimen Post-Irradiation Examination Data Interim Report

This interim report documents preliminary results of the post-irradiation examination material property testing from the fourth advanced graphite creep (AGC), AGC 4, capsule specimens. This is the fourth of a series of six irradiation test trains planned as part of the AGC experiment to fully characterize the neutron irradiation effects and radiation creep behavior of current nuclear graphite grades to moderate dose levels (=7 dpa). The AGC 4 capsule was irradiated in the Idaho National Laboratory Advanced Test Reactor at a nominal temperature of 800°C and to a peak dose of 8 dpa. Half of the AGC-4 specimens were subjected to compressive stresses to induce irradiation creep. Post-irradiation testing and measurement results are reported with the exception of thermal testing, which is still in progress, and irradiation mechanical strength testing. Additionally, some specimens initially deemed too hot to be examined in the ART Graphite laboratory may still be measured. The data reported includes specimen dimensions for both stressed and unstressed specimens to establish the irradiation creep rates, mass and dimensional data necessary to derive density, elastic constants (Young?s modulus, shear modulus, and Poisson?s ratio) from ultrasonic time of flight velocity measurements, Young?s modulus from the fundamental frequency of vibration, and electrical resistivity. A more complete evaluation of trends in the material property changes, as well as irradiation-induced creep due to the irradiation environment and applied load on the specimens, will be discussed later in AGC 4 post-irradiation examination analysis reports.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

ADDITIVE MANUFACTURING AND REGULATORY TESTING OF CANISTERS FOR SPENT NUCLEAR FUEL MANAGEMENT

The back end of the fuel cycle focuses on the interim storage, transportation, and final disposition of the spent nuclear fuel from nuclear reactors. Commercial light water nuclear power stations across the United States operate with fuel assemblies being irradiated for up to 6 years (planned) in the reactor pressure vessel. After their planned irradiation, the fuel assemblies are moved to a spent fuel pool within the facility complex. After the spent nuclear fuel is removed from the fuel pool, it is typically inserted into a welded metal canister that can be transferred between overpacks for storage, transportation and possibly disposal. Most dry storage systems being used by industry today use dual purpose canisters (DPCs), designed for use in storage and transportation overpacks, but not specifically designed for disposal. Triple purpose-canisters, designed for disposal in addition to storage and transportation, have also been researched. Traditional manufacturing methods for spent fuel canisters involve fusion welding along the length or circumference of the canister which results in high tensile residual stresses in the joint weld zone (WZ) and heat affected zone (HAZ). In this paper, spent fuel canister designs were printed by wire arc additive manufacturing (AM) using the 316L SS welding wire to demonstrate: 1. Feasibility of spent nuclear fuel canister fabrication using this advanced manufacturing method, 2. Dynamic response of the additive manufacture canister design when subjected to the federally mandated Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions (HAC) physical tests for Type B packages. This paper will focus on the canister printing design and structural tests. The AM 3D printed design, regulatory testing, and post-test evaluation of the canister tested to the 10 CFR 71.71 and 10 CFR 71.73 requirements will be presented. One AM canister design was subjected to the penetration, free drop, and puncture test. Before and after the dynamic structural tests, the AM canister design was scanned with a handheld scanner to capture a 3D CAD geometry to compare to the 3D printed canister design in the deformed shape. The scanned geometry was sectioned in areas with deformation and the cross-section profile was measured to determine accurate and repeated results of the deformed shape of the AM canister design.

Martinez, Oscar↗

DOE OSTI2016

PWR Station Blackout and External Events Analysis for LWRS/RISMC Project

In the framework of the US-DOE Light Water Reactor Sustainability/ Risk-Informed Safety Margin Characterization (LWRS/RISMC) project, methods, data and tools are being developed with the ultimate scope of improving the safety and operation of the current US nuclear fleet. A set of Industrial Applications (IA) concerning the most relevant safety issues is defined for applying and testing the developed tool kits. The ultimate scope of IA no. 2 is a realistic simulation of natural external hazards that impose threat to a Nuclear Power Plant (NPP). This scope requires the development of a methodology and of a qualified set of tools able to perform advanced risk-informed safety analysis. In particular the methodology should be able to combine results from seismic, flooding and thermal-hydraulic (TH) deterministic calculations with dynamic Probabilistic Risk Assessment (PRA). A so-called Risk Informed Margin Management (RIMM) approach is being used to represent meaningful (i.e., realistic facility representation and simulation) event scenarios and consequences by linking probabilistic simulations to external events physics. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2015

Modeling of the Reactor Core Isolation Cooling Response to Beyond Design Basis Operations - Interim Report

Efforts are being pursued to develop and qualify a system-level model of a reactor core isolation (RCIC) steam-turbine-driven pump. The model is being developed with the intent of employing it to inform the design of experimental configurations for full-scale RCIC testing. The model is expected to be especially valuable in sizing equipment needed in the testing. An additional intent is to use the model in understanding more fully how RCIC apparently managed to operate far removed from its design envelope in the Fukushima Daiichi Unit 2 accident. RCIC modeling is proceeding along two avenues that are expected to complement each other well. The first avenue is the continued development of the system-level RCIC model that will serve in simulating a full reactor system or full experimental configuration of which a RCIC system is part. The model reasonably represents a RCIC system today, especially given design operating conditions, but lacks specifics that are likely important in representing the off-design conditions a RCIC system might experience in an emergency situation such as a loss of all electrical power. A known specific lacking in the system model, for example, is the efficiency at which a flashing slug of water (as opposed to a concentrated jet of steam) could propel the rotating drive wheel of a RCIC turbine. To address this specific, the second avenue is being pursued wherein computational fluid dynamics (CFD) analyses of such a jet are being carried out. The results of the CFD analyses will thus complement and inform the system modeling. The system modeling will, in turn, complement the CFD analysis by providing the system information needed to impose appropriate boundary conditions on the CFD simulations. The system model will be used to inform the selection of configurations and equipment best suitable of supporting planned RCIC experimental testing. Preliminary investigations with the RCIC model indicate that liquid water ingestion by the turbine decreases the developed turbine torque; the RCIC speed then slows, and thus the pump flow rate to the RPV decreases. Subsequently, RPV water level decreases due to continued boiling and the liquid fraction flowing to the RCIC decreases, thereby accelerating the RCIC and refilling the RPV. The feedback cycle then repeats itself and/or reaches a quasi-steady equilibrium condition. In other words, the water carry-over is limited by cyclic RCIC performance degradation, and hence the system becomes self-regulating. The indications achieved to date with the system model are more qualitative than quantitative. The avenues being pursued to increase the fidelity of the model are expected to add quantitative realism. The end product will be generic in the sense that the RCIC model will be incorporable within the larger reactor coolant system model of any nuclear power plant or experimental configuration.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2021

Introduction of the PELICAN loop, a Full-Scale Pressure Drop Test Facility

The Versatile Test Reactor (VTR) is a test reactor currently under development by the US Department of Energy. This reactor will rely on fast neutrons enabling novel and wide-ranging experiment to support the development of the various advanced reactor technologies. With the high flux achievable, accelerated testing of fluid and materials will be made possible. To support VTR design efforts [1], an experimental facility has been designed and constructed at Argonne National Laboratory to recreate the hydraulic flow conditions within the VTR’s primary heat transport system (PHTS). This facility, the Pressure drop Experimental Loop for Investigations of Core Assemblies in advanced Nuclear reactors, PELICAN, measures the pressure drop across a full-scale fuel assembly containing prototypic axial reflectors, fuel, and plena components. Here, we first describe the design considerations required to recreate aspects of the VTR. Then we discuss the design and construction of PELICAN to address these design requirements, the design and construction of the test articles placed inside PELICAN’s test section, and finally present some of the first experimental results.

Grannan, A. M.↗