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The purpose of this report is to document a literature and data search for data and information pertaining to the stability of nuclear reactor molten core materials. This includes data and analysis from TMI-2 fuel and INL’s LOFT (Loss of Fluid Test) reactor project and other sources.
Savannah River National Laboratory (SRNL) is accelerating the research of high-level nuclear waste (HLW) glass corrosion using advanced big data techniques available through the Accelerated Leaching and Testing of GLASS (ALTGLASS) database. Developed, maintained and distributed by SRNL, ALTGLASS gives HLW researchers access to chemical durability data for hundreds of glass compositions, allowing scientists to use an “informatics approach” to analyze existing and new experimental data contained in the database to advance glass corrosion science.
We have re-analysed the two breeding blankets experiments performed previously in the frame of the European fusion program on two mock-ups of the European Helium-Cooled-Lithiium Lead (HCLL) and Helium-Cooled-Pebble-Bed (HCPB) test blanket modules for ITER. The tritium production rate and the neutron and photon spectra measured in these mock-ups were compared with calculations using FENDL-3 Starter Library, release 4 and state-of-the-art nuclear data evaluations, JEFF-3.1.2, JENDL-4.0 and ENDF/B-VII.0. The tritium production calculated for the HCPB mock-up underestimates the experimental result by about 10%. The result calculated with FENDL-3/SLIB4 gives slightly smaller tritium production by 2% than the one with FENDL-2.1. The difference attributes to the slight modification of the total and elastic scattering cross section of Be. For the HCLL experiment, all libraries reproduce the experimental results well. FENDL-3/SLIB4 gives better result both for the measured spectra and the tritium production compared to FENDL-2.1.
In this report, data from an electrorefiner (ER) for nuclear fuel reprocessing is evaluated for process monitoring (PM) conclusions. This data comes from tests performed at the Idaho National Laboratory in 2022. Multivariate approaches utilizing methods of Principal Component Analysis (PCA) is applied. This is based off established work in process monitoring for fault detection in industrial facilities. This report will discuss the background, methods, and results of the application and some of the conclusions and applications that can be drawn from them. PCA is applied to two different electrorefiner (ER) operations that occurred at Idaho National Laboratory between August and October 2022. The first operation occurred with little incident while the second had several noted faults in the equipment in operational logs. The data from the first run was used to train the data for “normal” operations and applied to both sets of data to determine when operations were in an “off-normal” condition and identify where the fault occurs through PCA. PCA was able to identify off-normal events and identify the cause for off-normal operations. These identified off-normal events matched with the events and their causes in the operational logs. However, small amounts of variance in the data led to false detection of “off-normal” events. Thus, careful selection of training data and a-posteriori conclusions based off operator assessments will both be required for application of PCA to PM applications. This work demonstrated that multivariate approaches and latent variables are applicable to pyroprocessing PM applications and can be further expanded in future work as quality variables such as salt concentration from sensors and sampling become available.
The presentation summarizes work performed over summer 2015 related to Monte Carlo simulations. A flexible probability table interpolation scheme has been implemented and tested with results comparing favorably to the continuous phase-space on-the-fly approach.
Containment/Surveillance (C/S) measures are critical to any verification regime in order to maintain Continuity of Knowledge (CoK). The Ceramic Seal project is research into the next generation technologies to advance C/S, in particular improving security and efficiency. The Ceramic Seal is a small form factor loop seal with improved tamper-indication including a frangible seal body, tamper planes, external coatings, and electronic monitoring of the seal body integrity. It improves efficiency through a self-securing wire and in-situ verification with a handheld reader. Sandia National Laboratories (SNL) and Savannah River National Laboratory (SRNL), under sponsorship from the U.S. National Nuclear Security Administration (NNSA) Office of Defense Nuclear Nonproliferation Research and Development (DNN R&D), have previously designed and have now fabricated and tested Ceramic Seals. Tests have occurred at both SNL and SRNL, with different types of tests occurring at each facility. This interim report will describe the Ceramic Seal prototype, the design and development of a handheld standalone reader and an interface to a data acquisition system, fabrication of the seals, and results of initial testing.
During the period of underground nuclear testing in the U.S., now ended for over 20 years, the focus of work was to conduct the test and analyze the data as quickly as possible; there was little concern about application of technologies that could be used to detect and characterize the location and depth of the explosion. Thus, there is little experience with applying geophysical methods in places where an underground nuclear explosion (UNE) has taken place. On the other hand, application of geophysical methods to the detection of shallow artifacts and water table depth is commonly applied in the environmental remediation and site characterization realm of commercial geophysical surveys. For nuclear nonproliferation treaty applications which may incorporate both shallow and deep geophysical probing methods (such as on-site inspection (OSI) under the Comprehensive Nuclear Test Ban Treaty [CTBT]), there thus remains a need to better understand detection capabilities of the deeper targets: anomalies related to an explosion cavity or rubble zone that may be located several hundred meters below the ground surface. The purpose of this Coordinated Inspection Technologies (CIT) project was to gain experience in applying active seismic methods and electrical methods at a site in Nevada where an underground nuclear test (UNE) had been conducted in the past to get a better understanding of the capabilities of these methods for detection and characterization of deep subsurface effects of the explosion.
The Central Nevada Test Area, Nevada, was the site of an underground nuclear test in 1968, which resulted in residual contamination at and near the detonation depth of 975 meters (3200 feet). The environmental corrective action strategy for the site was conducted in accordance with the Federal Facility Agreement and Consent Order and all applicable Nevada Division of Environmental Protection (NDEP) policies and regulations. The corrective action strategy initiated in 1999 included developing a conceptual site model, preparing a numerical flow and transport model, calculating contaminant boundaries, negotiating compliance boundaries with NDEP, performing model validation, and monitoring groundwater. This strategy was executed through the Corrective Action Decision Document/Corrective Action Plan phase, when it was determined that the flow and transport model could not be validated because it failed to adequately predict groundwater elevations in the monitoring/validation (MV) wells (MV-1, MV-2, and MV-3) drilled subsequent to the modeling effort. This led to a change in the corrective action strategy that is now designed to validate the compliance boundary through monitoring and institutional controls, rather than relying predominantly on a numerical flow and transport model. This new approach included enhancements to the monitoring well network and additional monitoring to confirm that data were sufficient to proceed to closure. The corrective action activities were completed in 2015, and the Closure Report, which established the long-term postclosure monitoring program, was approved in 2016. The postclosure monitoring program is designed to (1) assess the effectiveness of the compliance boundary by monitoring for the radionuclides of interest and (2) evaluate the effectiveness of monitoring locations within the groundwater flow system by monitoring water elevations to ensure that monitoring wells are located along the potential migration pathways. (authors)
Fuel pins with small defects could allow for the gradual ingress of water into the cladding during normal operation. If these 'waterlogged' fuel pins are subjected to a rapid power change, the internal pressurization could greatly exceed that of a normal fuel pin due to rapid conversion of the water to steam. This could induce either a premature or a more energetic failure during a design basis RIA than in non-defected fuel pins. Experimental studies on this type of fuel are very limited and do not appear to have been fully reported in the open literature. Results from a variety of tests conducted at the Transient Reactor Test Facility (TREAT), Special Power Excursion Reactor Test (SPERT) facility, Nuclear Safety Research Reactor (NSRR), and Power Burst Facility (PBF) are reviewed in this paper. These tests were intended to evaluate both the physical consequences of failure and the threshold and causes for waterlogged fuel failure during RIA. The available data suggests that although waterlogged fuel pins may burst at a lower energy deposition threshold, they do not pose a significant new hazard during design basis accidents. However, data is limited to tests conducted on fresh fuel samples only.
This paper is based on the data processing performance problems of operating command delay and system display not updating real-time for nuclear power plant computerized information and control system. The plant computerized information and control system is responsible for monitoring and controlling the plant operation, and it is a key part in the plant DCS system. Its data processing performance will affect the operation stability of nuclear power plant. The architecture, hardware configuration and data communication for plant computerized information and control system among the different levels in DCS are described. The main reasons affecting data processing performance in different signal channels are analyzed according to the plant load test results. The CPU load curves for real time server and central processing server are displayed, although the detailed cause for operating command delay and system display not updating realtime are not found, central processing server load is heavier than real time server which remind us to analyze the data processing of software resource scheduling deeply. So the resource distribution problem of software execution task is also analyzed according to CPU load test results. This paper provides the optimization analysis for data processing performance, and it is available and effective according to the latest test results. (authors)
This document provides detailed reactivity and fission rate distribution results for MPACT for the Core Physics Benchmark Progression Problems 1-5, including comparisons to initial startup data from Watts Bar Nuclear Unit 1. Also included are computer requirements and runtimes for each type of problem and feedback for issues encountered during the testing. Testing for the benchmark problems that do not require thermal-hydraulic feedback has been very successful. Testing for the coupled problems (6 and 7) is still ongoing and will be documented in a future revision or separate document. A few issues have been identified needing more research/investigation before all of these problems have ideal solutions. These items are included in the Section 8 near the end of this document. MPACT has successfully solved all of the zero power core physics progression problems. The computer resources required and runtimes were all acceptable, though faster cases would be preferable considering there was no thermal-hydraulic feedback. MPACT converged very reliably for all 3D problems, expect in a few isolated studies where very thin axial planes were purposely created to attempt to quantify the errors in the control rod tip homogenization model. Despite the issues listed in Section 8, the accuracy of MPACT compared to the reference solution is considered to be very good.
Tensile test specimens were made from shear drops from three such depleted-uranium fuel plates: CR6064; CS1027; and FP6686. Pieces for test-specimen fabrication originated from each end and along one side of all three plates. Specimen fabrication, testing, and analysis of resulting data are presented.
The mechanical properties of UO{sub 2} fuel are of great importance to the nuclear engineering community for modeling of fuel performance during operation and accident scenarios. Data on the deformation of single crystals at room and elevated temperature can help these efforts by informing multiscale models. Small-scale testing techniques like micro-cantilever bending and nano-indentation have the ability to measure mechanical properties at the right length scale for this, while allowing the observation of the deformation of the material. In our work we perform micro-cantilever testing of UO{sub 2} at room temperature and nano-indentation testing up to 500 deg. C, as well as in-situ TEM micro-cantilever testing.
In this document, we focus mainly on testing libraries written in the legacy ENDF format. As we move to new nuclear data format, we anticipate will need to revisit this document.
The intent of this document is to provide information about best available approaches for performing subsurface soil gas sampling during an On Site Inspection or OSI. This information is based on field sampling experiments, computer simulations and data from the NA-22 Noble Gas Signature Experiment Test Bed at the Nevada Nuclear Security Site (NNSS). The approaches should optimize the gas concentration from the subsurface cavity or chimney regime while simultaneously minimizing the potential for atmospheric radioxenon and near-surface Argon-37 contamination. Where possible, we quantitatively assess differences in sampling practices for the same sets of environmental conditions. We recognize that all sampling scenarios cannot be addressed. However, if this document helps to inform the intuition of the reader about addressing the challenges resulting from the inevitable deviations from the scenario assumed here, it will have achieved its goal.
The on-going AMMON program in EOLE reactor at CEA Cadarache (France) provides experimental results to qualify the HORUS-3D/N neutronics calculation scheme used for the design and safety studies of the new Material Testing Jules Horowitz Reactor (JHR). This paper presents the determination of technological and nuclear data uncertainties on the core reactivity and the propagation of the latter from the AMMON experiment to JHR. The technological uncertainty propagation was performed with a direct perturbation methodology using the 3D French stochastic code TRIPOLI4 and a statistical methodology using the 2D French deterministic code APOLLO2-MOC which leads to a value of 289 pcm (1{sigma}). The Nuclear Data uncertainty propagation relies on a sensitivity study on the main isotopes and the use of a retroactive marginalization method applied to the JEFF 3.1.1 {sup 27}Al evaluation in order to obtain a realistic multi-group covariance matrix associated with the considered evaluation. This nuclear data uncertainty propagation leads to a K{sub eff} uncertainty of 624 pcm for the JHR core and 684 pcm for the AMMON reference configuration core. Finally, transposition and reduction of the prior uncertainty were made using the Representativity method which demonstrates the similarity of the AMMON experiment with JHR (the representativity factor is 0.95). The final impact of JEFF 3.1.1 nuclear data on the Begin Of Life (BOL) JHR reactivity calculated by the HORUS-3D/N V4.0 is a bias of +216 pcm with an associated posterior uncertainty of 304 pcm (1{sigma}). (authors)
This document describes the data quality objectives to select archived samples located at the 222-S Laboratory for Bench-Scale Reforming testing. The type, quantity, and quality of the data required to select the samples for Fluid Bed Steam Reformer testing are discussed. In order to maximize the efficiency and minimize the time to treat Hanford tank waste in the Waste Treatment and Immobilization Plant, additional treatment processes may be required. One of the potential treatment processes is the fluidized bed steam reformer. A determination of the adequacy of the fluidized bed steam reformer process to treat Hanford tank waste is required. The initial step in determining the adequacy of the fluidized bed steam reformer process is to select archived waste samples from the 222-S Laboratory that will be used in a bench scale tests. Analyses of the selected samples will be required to confirm the samples meet the shipping requirements and for comparison to the bench scale reformer (BSR) test sample selection requirements.