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At least 1,585 records · Page 88

BISON validation to in situ cladding burst test and high-burnup LOCA experiments

The process for developing and qualifying nuclear fuels for commercial nuclear application requires fundamental material development, characterization, and design; out-of-pile testing on unirradiated materials; integral fuel rod irradiations, testing, and postirradiation examinations; and transient analyses. The historical approach depends on the generation of large empirical datasets and series of integral fuel rod irradiations, and this approach ultimately takes ~20 years—or sometimes longer—to acquire data through extensive sequential testing. Thus, the qualification and eventual deployment of new fuel systems constitute a long process. However, recent technological advancements have provided researchers the opportunity to perform out-of-cell, in situ measurements to assess material performance for the duration of the experiment. One such example of this capability is the use of digital image coordination and thermal imaging to assess Zircaloy cladding performance under a simulated loss-of-coolant accident (LOCA) transient condition. In situ measurements generally provide high-fidelity strain, strain rates, and temperature surface maps. This is critical for the US nuclear industry, which is actively developing a technical basis to support extending the peak rod average burnup from 62 to ~75 GWd/tU and the deployment of accident-tolerant fuel. However, the US Nuclear Regulatory Commission (NRC) outlined in its research information letter several technical issues that the industry must address before extending burnup. One topic of specific interest is understanding the cladding balloon and rupture geometry during the LOCA heat-up phase. By leveraging these advanced in situ capabilities, this work used in situ data generated from a simulated LOCA to better understand high-temperature creep and its effect on Zircaloy balloon and rupture performance. Here, this work used the BISON fuel performance code to assess the high-temperature creep model predictions with in situ data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

A correlation-based approach for evaluating mechanical properties of nuclear fuel cladding tubes

The present work offers and evaluates a correlation-based approach to determine the tensile properties of nuclear fuel cladding tubes. A set of model materials with known properties (i.e., austenitic and ferritic steels, fcc- and hcp-alloys) were tested to build a correlation between conventional tensile tests and ring (hoop) and tube (axial) tests. It was shown that close-to-linear relationships exist for the basic mechanical properties (i.e., yield and ultimate stress, uniform and total elongation) between the uniaxial tensile test data and the ring and tube test data. Using the ring and tube specimen geometries, the feasibility of the approach was demonstrated via manufacturing and testing specimens from irradiated Zr and FeCrAl nuclear fuel cladding tubes. Limitations caused by specific deformation mechanisms (e.g., high ductility of 304L steel) were discussed in detail. The approach can be extended to other tubular products with different geometry and dimensions.

36 MATERIALS SCIENCE↗

Neutron Drip Line in the Ca Region from Bayesian Model Averaging

The region of heavy calcium isotopes forms the frontier of experimental and theoretical nuclear structure research where the basic concepts of nuclear physics are put to stringent test. The recent discovery of the extremely neutron-rich nuclei around 60 Ca and the experimental determination of masses for 55-57 Ca provide unique information about the binding energy surface in this region. To assess the impact of these experimental discoveries on the nuclear landscape’s extent, we use global mass models and statistical machine learning to make predictions, with quantified levels of certainty, for bound nuclides between Si and Ti. Using a Bayesian model averaging analysis based on Gaussian-process-based extrapolations we introduce the posterior probability for each nucleus to be bound to neutron emission. We discover that extrapolations for drip-line locations, at which the nuclear binding ends, are consistent across the global mass models used, in spite of significant variations between their raw predictions. In particular, considering the current experimental information and current global mass models, we suggest that 68 Ca has an average posterior probability p ex ≈ 76 % to be bound to two-neutron emission while the nucleus 61 Ca is likely to decay by emitting a neutron (p ex ≈ 46 % ).

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Advanced Sensors and Instrumentation Issue 10

The Advanced Sensors and Instrumentation (ASI) program, on its seventh fiscal year, continues to address key technology gaps for future nuclear energy systems. ASI’s goal is to enable the advanced instrumentation and control (I&C) technologies essential to the Office of Nuclear Energy’s research and development (R&D) efforts needed to realize its mission goals. Information on our projects can be found in our ASI annual awards summaries, annual review webinars, and newsletters posted in the Office of Nuclear Energy website (energy.gov/ne). This year, we have initiated three newly awarded projects in the area of data analytics. We also initiated a project to develop and demonstrate a prototype passive wireless sensor network. This sensor will use advanced manufacturing and printing, and be tested and demonstrated at a U.S. nuclear power plant facility. In addition, we are developing a sensor database to capture advanced sensor and instrumentation technology that is applicable to nuclear power plants. Under the In-Pile Instrumentation program, we continue to develop unique in-pile sensors and instrumentation to provide real-time, accurate, and spatially resolved information regarding the performance of fuels and materials. Due to the closure of the Halden reactor, we are focusing our near-term research in providing equivalent capability for the instrumentation of irradiation experiments in the U.S. that are needed by the nuclear industry. As the ASI program continues to award projects through competitive solicitations, we encourage the I&C community to visit our website for current funding opportunity announcements, especially for the industry-focused solicitation entitled, “U.S. Industry Opportunities for Advanced Nuclear Technology Development.”

99 GENERAL AND MISCELLANEOUS↗

Thermal Modeling of the TN-32B Cask for the High Burnup Spent Fuel Data Project

This report describes thermal modeling done for the High Burnup Dry Storage Research Project that is being performed by the U.S. Department of Energy’s Spent Fuel and Waste Science and Technology (SFWST) research and development (R&D) program. The purpose of this project to investigate the performance of high-burnup spent nuclear fuel in dry storage. Part of this project is a demonstration test being performed with a storage module in the North Anna Nuclear Power Station’s Independent Spent Fuel Storage Installation (ISFSI). The storage module selected for this demonstration is an Orano TN-32B High Burnup cask (Fig. S-1) which has been loaded with fuel assemblies from the North Anna spent fuel pool. The main goals of this test are to provide confirmatory data for model validation and potential improvement, support license renewals and new licenses for ISFSIs, and support transportation licensing for high burnup spent nuclear fuel (EPRI 2014).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Flow tests of a single fuel element coolant channel for a compact fast reactor for space power

Water flow tests were conducted on a single-fuel-element cooling channel for a nuclear concept to be used for space power. The tests established a method for measuring coolant flow rate which is applicable to water flow testing of a complete mockup of the reference reactor. The inlet plenum-to-outlet plenum pressure drop, which approximates the overall core pressure drop, was measured and correlated with flow rate. This information can be used for reactor coolant flow and heat transfer calculations. An analytical study of the flow characteristics was also conducted.

Springborn, R. H.↗

Nuclear physics and cosmology

Nuclear physics has provided one of two critical observational tests of all Big Bang cosmology, namely Big Bang Nucleosynthesis. Furthermore, this same nuclear physics input enables a prediction to be made about one of the most fundamental physics questions of all, the number of elementary particle families. The standard Big Bang Nucleosynthesis arguments are reviewed. The primordial He abundance is inferred from He-C and He-N and He-O correlations. The strengthened Li constraint as well as D-2 plus He-3 are used to limit the baryon density. This limit is the key argument behind the need for non-baryonic dark matter. The allowed number of neutrino families, N(nu), is delineated using the new neutron lifetime value of tau(n) = 890 + or - 4s (tau(1/2) = 10.3 min). The formal statistical result is N(nu) = 2.6 + or - 0.3 (1 sigma), providing a reasonable fit (1.3 sigma) to three families but making a fourth light (m(nu) less than or equal to 10 MeV) neutrino family exceedly unlikely (approx. greater than 4.7 sigma). It is also shown that uncertainties induced by postulating a first-order quark-baryon phase transition do not seriously affect the conclusions.

Schramm, David N.↗

TREAT HODOSCOPE – A Predictive Simulation Model

The Transient Reactor Test Facility (TREAT) reactor is being restarted after over two decades of standby operation. TREAT is designed to test the time-dependent behavior of nuclear fuel during simulated reactor accidents. One of the principle data collection systems of TREAT is the fast neutron hodoscope which is used to observe the motion of the fuel test specimen as it is being damaged, deformed, or melted over the course of the power transient. The hodoscope is a grid of ex-core collimator channels paired with a bank of fast neutron sensitive radiation detectors. These detectors have historically been calibrated using empirical methods utilizing measurement data from smaller power pulses. Given modern computational capabilities this work sets out to establish a predictive computational capability of the fast neutron flux field exiting the hodoscope to support these calibrations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Radiation Hardened Foam Cold Test Plan - Phase II: Foam Characterization Testing and Environmental Chamber Testing of FoamBag Fixative Foam

This document outlines the Phase II test objectives and implementation plan for a down-select foam fixative technology intended to facilitate activities in support of the Savannah River Site (SRS) F/H labs deactivation and decommissioning (D&D) efforts. It is a collaborative effort between Savannah River National Laboratory (SRNL), Florida International University (FIU), and the SRS F/H labs team intended to test and evaluate the potential of a polyurethane resin foam in mitigating the release of contamination during dismantling operations on radioactively contaminated piping in legacy facilities. The cold test plan addresses specific requirements highlighted by site and safety personnel and will be executed at FIU and SRNL test and lab facilities. Results from the cold tests will inform the hot test at F/H labs, which will use the foam fixative to confine and/or isolate residual contamination within a 3-dimensional void space of Hastelloy C-22 piping designated for removal from the area and transported to a designated disposal facility. Phase I testing was previously conducted using Hilti CP620 fixative foam. Results from Phase I testing indicate Hilti CP-620 fixative foam is incompatible with SRS site hot taps as an effective foam delivery method into Hastelloy C-22 piping. Phase II testing will be conducted using FoamBag TM fixative foam as an alternative fixative foam option. Phase II testing will address eight test objectives: (1) evaluation of the adhesion and bonding properties of FoamBag TM in Hastelloy C-22 piping, (2) evaluation of the adhesion of FoamBag TM in piping under various moisture conditions, (3) determination of the heat profile of FoamBag TM during curing, (4) determination of the internal pipe pressure after FoamBag TM deployment and curing, (5) conduct a leak test to determine if FoamBag TM is effective at creating a full seal within piping, (6) headspace testing of FoamBag TM to determine if there are any associated off-gas hazards during FoamBag TM curing, (7) conduct environmental chamber testing of FoamBag TM to understand how environmental parameters impacts FoamBag TM curing, and (8) conduct fire testing to evaluate FoamBag TM fire retardant characteristics.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

STANDARD REFERENCE MATERIAL CHARPY TESTING FROM HIGH FLUENCE SURVEILLANCE CAPSULE

The Palisades Nuclear Generating Station included in its original surveillance program a surveillance capsule, designated A-60. The capsule was removed from its surveillance position in early 1995 and has been resident in the spent fuel pool since that time. It was harvested to perform characterization of surveillance specimens in this capsule in 2023. This capsule was irradiated to a fluence of 1.96x1020 n/cm2 (E> 1MeV) that is equivalent for more than 150 effective full power years for the current US reactor pressure vessel (RPV) fleet. This capsule contained several materials, including Charpy specimens of standard reference material (SRM) from Heavy-Section Steel Technology (HSST) A533-B Plate 01. To perform Charpy testing of this highly irradiated material, the new Charpy specimen transfer system was designed and implemented on Charpy impact testing machine in the hot cell to accommodate remote testing of these specimens. This new transfer system includes integrated environmental chamber to cool or heat Charpy specimens to the desired temperature. Testing of this highly irradiated material revealed very large shift of Charpy transition temperature, 188oC.

Sokolov, Mikhail [ORNL] (ORCID:0000000256373346)↗

Quantifying deformation during Zry-4 burst testing: a comparison of BISON and a combined in-situ digital image correlation and infrared thermography method

The development and verification of nuclear fuel-cladding performance codes require extensive testing to establish empirical correlations, necessitating more accelerated testing methods that can replace large validation studies. Here, in the present work, a technique was developed and implemented to experimentally quantify the relationship between temperature, cladding internal pressure, and strain in-situ during a burst test, with the specific aim of generating data for code validation and model refinement. Digital image correlation was used to measure hoop, axial, and radial strain, and infrared thermography to quantify axial temperature gradients. Several key experimental modifications to traditional LOCA burst testing were necessary, but are shown to effectively have little impact on burst temperatures. The time/temperature dependent pressure and time/spatially dependent thermal gradient data were used with BISON to simulate cladding burst and strain rates. Although the measured DIC strains reach appreciable amounts at slightly lower temperatures than BISON simulated strains, there is good agreement between the measured and simulated strains in terms of magnitude and rates leading up to burst. However, in the few seconds prior to and during burst, the BISON simulated strains are significantly lower than the measured strains, indicating the potential for model improvement. The combined experiment and simulation technique may be applied to any new cladding concept to accelerate fuel qualification.

36 MATERIALS SCIENCE↗

Transient Reactor Test Facility (TREAT) Low Enriched Uranium (LEU) Conversion Progress

The National Nuclear Security Administration Office of Material Management and Minimization (NNSA M3) is working towards converting the Transient Reactor Test Facility (TREAT) located at the Idaho National Laboratory (INL) from its original high-enriched uranium (HEU) core to a low-enriched uranium (LEU) core. TREAT reactor operations were restarted in November of 2017 and data obtained from the initial testing will be used to validate the existing ANL reactor modeling. Fuel block development has progressed with the completion of 2 and 4 inch crack free surrogate fuel blocks. Alternative fabrication methods for the fuel blocks continue to be explored and are showing promising results. Additionally, cladding oxidation studies have been performed and analyzed to initiate the down selection of cladding material to be used for the TREAT LEU fuel elements. This paper describes the current status of the above mentioned activities, and discusses the anticipated path forward for the conversion program as it progresses toward a final fuel design and fuel fabrication process.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

RIA simulation tests using driver tube for ATF cladding

Pellet-cladding mechanical interaction (PCMI) is a potential failure mechanism for accident-tolerant fuel (ATF) cladding candidates during a reactivity-initiated accident (RIA). This report summarizes Fiscal Year (FY) 2017 research activities that were undertaken to evaluate the PCMI-like hoop-strain-driven mechanical response of ATF cladding candidates. To achieve various RIA-like conditions, a modified-burst test (MBT) device was developed to produce different mechanical pulses. The calibration of the MBT instrument was accomplished by performing mechanical tests on unirradiated Generation-I iron-chromium-aluminum (FeCrAl) alloy samples. Shakedown tests were also conducted in both FY 2016 and FY 2017 using unirradiated hydrided ZIRLO™ tube samples. This milestone report focuses on testing of ATF materials, but the benchmark tests with hydrided ZIRLO™ tube samples are documented in a recent journal article.a For the calibration and benchmark tests, the hoop strain was monitored using strain gauges attached to the sample surface in the hoop direction. A novel digital image correlation (DIC) system composed of a single high-speed camera and an array of six mirrors was developed for the MBT instrument to better resolve the failure behavior of samples and to provide useful data for validation of high-fidelity modeling and simulation tools. The DIC system enable a 360° view of a sample’s outer surface. This feature was added to the instrument to determine the precise failure location on a sample’s surface for strain predictions. The DIC system was tested on several silicon carbide fiber/silicon carbide matrix (SiC/SiC) composite tube samples at various pressurization rates of the driver tube (which correspond to the strain rates for the samples). The hoop strains for various loading conditions were determined for the SiC/SiC composite tube samples. Future work is planned to enhance understanding of the failure behavior of the ATF cladding candidates of age-hardened FeCrAl alloys and SiC/SiC composites in detail during RIA conditions informed by the computational studies performed under the US Department of Energy Office of Nuclear Energy Advanced Fuels Campaign. The testing instrument and the new DIC system will be further developed to reach different stress-state conditions and to perform tests at elevated temperatures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Radioactivity Calculator for Research and Test Reactor Neutron-Activated Materials

The Idaho National Laboratory (INL) Nuclear Science User Facility (NSUF) supports nuclear energy-relevant research into the effects of radiation on nuclear fuels and structural materials. The NSUF provides access, at no cost to the researcher, to the Advanced Test Reactor (ATR), the High-Flux Isotope Reactor (HFIR), and other university and national laboratory research and test reactors for radiation effects experiments. Post-irradiation examination of materials poses some difficulties when studying neutron-irradiated materials due to the radioactivity levels of the materials. Inside a test reactor, under high neutron flux, stable materials can activate (capture a neutron) and become unstable or radioactive. These newly created radioactive materials decay, based on their respective half-lives, and emit energetic radiation. Nuclides with short half-lives are able to decay to sufficiently low radiation levels prior to the sample being removed from the reactor. However, nuclides with longer half-lives present a challenge since they must be handled with extreme caution and stored until their radioactivity levels decrease. If immediate examination is required, the examination may take place in a hot cell, which can dramatically increase the cost of the project. The NSUF has access to a variety of hot cells and characterization capabilities at national laboratories, universities, and industrial sites across the United States. In order to estimate the post-irradiation radioactivity levels of materials early in the experiment cycle, we created a radioactivity calculator, hosted on the NSUF web site (nsuf.inl.gov), which will consider three stages of neutron activations and decays for sample materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Test Result of the Short Models MQXFS3 and MQXFS5 for the HL-LHC Upgrade

© 2002-2011 IEEE. In the framework of the High-Luminosity Large Hadron Collider, the installation of a new generation of quadrupole magnets is foreseen on each side of ATLAS and CMS experiments. The new magnets are based on Nb 3 Sn technology and shall be able to reach an ultimate current of 17.9 kA with a peak field of 12.3 T in the coil. In 2016 and 2017, the first two short models, called MQXFS3 and MQXFS5, have been tested at 4.2 and 1.9 K in the two new test benches at the European Organization for Nuclear Research. This paper presents the result of the quench performance of the two models; the first magnet reached nominal but failed to reach ultimate, showing detraining in one coil. MQXFS5 reached ultimate performance without any detraining phenomena, validating the PIT conductor used for the first time in this magnet program.

Bajas, Hugues↗

Review and Experimental Comparison of the Durability of Iodine Waste Forms in Semi-Dynamic Leach Testing

To ensure the safe disposal of radioiodine-containing nuclear wastes, a durable iodine waste form (IWF) is required. Many IWFs have been developed and evaluated for chemical durability using a wide range of test methods, making direct comparisons difficult as highlighted in this work through a review of available IWF corrosion data. To move toward a common dataset, a semi-dynamic leach test suite was experimentally demonstrated using several IWF types, providing standard comparisons. The results highlighted clear differences in corrosion resistance between various IWFs. This test suite is suggested for future studies to generate comparable data for a common dataset.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Preliminary Scaling Analysis of Reactivity Initiated Accidents

The life cycle for new fuel development requires extensive modeling and testing of the fuel concept from benchtop scale experiments to full scale assemblies in prototypical conditions. As a result, the timeline to develop new fuel types can extend up to 20 years and hinders the ability to develop a diverse assortment of fuel types. The licensing of recent reactor designs such as Westinghouse's AP-600/1000 designs utilized a scaling methodology to perform nuclear safety code validation on experimental test data from scaled down thermal hydraulic facilities. The advantage of using a scaling methodology was a significant cost decrease in the testing program because it avoided needing to build full scale prototypes and focused on improving safety code's ability to predict test data. This study investigates how to apply a scaling hierarchy for fuel performance during a Reactivity Initiated Accident that could be implemented with scaling methods such as Hierarchical Two Tiered Scaling (H2TS), Fractional Scaling Analysis (FSA), and Dynamical System Scaling (DSS) .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

High Burnup Transient Fission Gas Release Testing and Post-Test Examinations

The contribution of transient fission gas release (tFGR) to the burst behavior of high burnup fuel during a loss of coolant accident (LOCA) for commercial light water reactors is a complex phenomenon that is not fully understood. During the temperature ramp associated with a LOCA, there is significant additional transient fission gas release in excess of the steady state fission gas release from normal operation. An addition to the Oak Ridge National Laboratory Severe Accident Test Station (SATS) was created to explore the different conditions that impact transient fission gas release. Four tests on commercially irradiated zirconium clad uranium dioxide nuclear fuel were performed to support transient fission gas release study. These experiments and their post-test analyses are discussed. The first test on high burnup fuel revealed a LOCA relevant transient fission gas release of 5.3% for an unpressurised segment. Subsequent tests have measured transient fission gas release at different conditions that span from 4.8% to 14.4 % transient fission gas release.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗