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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 595 records · Page 33
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Benchmark Testing of a New 56 Fe Evaluation for Criticality Safety Applications

The SAMMY code was used to evaluate resonance parameters of the 56 Fe cross section in the resolved resonance energy range of 0–2 MeV using transmission data, capture, elastic, inelastic, and double differential elastic cross sections. The resonance analysis was performed with the code SAMMY that fits R-matrix resonance parameters using the generalized least-squares technique (Bayes’ theory). The evaluation yielded a set of resonance parameters that reproduced the experimental data very well, along with a resonance parameter covariance matrix for data uncertainty calculations. Benchmark tests were conducted to assess the evaluation performance in benchmark calculations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Status Report for NEET In-Pile Sensor Irradiation Capabilities at TREAT

The recently-established in-pile sensor development project under the Nuclear Energy Enabling Technology (NEET) program is focused on the development of measurement technologies to enable acquisition of data focused on the performance, properties, and boundary conditions of nuclear fuels/materials under irradiation in nuclear reactors. These sensors can be used in steady state irradiation tests to measure parameters of interest under conditions representing normal nuclear plant operation, or in transient irradiation tests to determine specimen performance under postulated unplanned conditions. In either case, development of suitable sensor technologies must include instrument irradiations to help guide sensor material/design selection, verify instrument performance, and demonstrate deployment viability. The Transient Reactor Test facility (TREAT) while just a part of the overall capabilities portfolio needed to enable this sensor development mission, has unique capabilities and characteristics that make it well poised for a number of instrument irradiation data objectives including sensor survivability screening, response characterization, qualification tests, and in-situ properties measurements. TREAT’s brilliantly-basic design, when married with appropriate irradiation hardware and analytic capabilities, fills a crucial capability gap between out-of-pile testing and irradiation in steady state high power research reactors. Leveraging a recently developed approach to irradiation testing in TREAT, the present work demonstrates that cost-effective and scientifically-impactful instrument irradiation capabilities for TREAT are a well-placed capability in the NEET in-pile sensor development project.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Pre-Shot Simulations of Near-Field and Far-Field Ground Motion for the Source Physics Experiment (SPE) Explosions at the Climax Stock, Nevada National Security Site: SPE3

The Source Physics Experiment (SPE) is designed to improve ground motion prediction capabilities for underground nuclear testing through the development of new physics-based models together with validation via new chemical explosion data. The SPE shots consist of different size and depth chemical explosions in a granite borehole at the Climax Stock on the Nevada National Security Site (NNSS). The third shot in the series (SPE3) will be the same size (1000 kg TNT equivalent) and depth (45 m canister centroid) as the previous shot (SPE2). The main purpose of the SPE3 is to investigate the effects of the damage zone caused by the SPE2 chemical explosion on the SPE3 explosion, keeping all other parameters (e.g. size, depth, explosive type) the same. This SPE3 pre-shot report addresses progress made in simulation of the expected wavefield in the both the near- and the far-field. In this report we consider the far-field to be where seismic wave motion can be characterized by elastic or visco-elastic approximations, in contrast to the near-field where other, often non-linear effects, play a major role. The current simulations build on previous work from the SPE1 and SPE2 shots. The simulations are intended to highlight differences between SPE2 and SPE3 using a variety to approximations intended to mimic the expected effect of the damage zone. The simplest physics of damage zone effects suggest changes in the source region material properties (density, velocity), and in the yield strength (coupling) of the rock, should both impact the far-field signal. We can illustrate this in a simplistic way for P-waves in Figure A using the empirical Mueller-Muphy (1971) far-field explosion model (MM). First if the SPE2 explosion-induced damage reduces velocities and densities without affecting the yield strength or coupling, then by energy conservation the amplitudes for SPE 3 would increase as shown in the top of Figure A. Similarly if the explosion induced damage increased density (say by compacting pore space) the amplitude would decrease. This impedance effect, where the product of the density, wave-velocity, and ground-velocity squared is conserved, is well known. The second and harder effect to quantify, are changes in the energy coupling or yield strength. In the simple MM model this is represented by the coupling factor between cavity radius and yield. If we assume the granite was slightly weakened by SPE2 such that the cavity radius increases slightly then the overall amplitudes of SPE3 decrease relative to SPE2 as shown in the bottom of Figure A. In reality both effects, may exist due to the damage zone, and furthermore we know they are interrelated in complex ways. For S-waves we expect even more complex effects due to motion on joints and faults. For this reason we are developing physics-based near field simulation models coupled to far-field propagation codes to predict the overall effects of changes in the source emplacement region on observed signals. Near-field simulations have been performed using the Eulerian code (GEODYN). An ensemble model, calibrated to match records from past large-scale nuclear events is used. Both 2D axisymmetric (no faults) and fully three-dimensional GEODYN calculations (with /without faults) were conducted in which the damage zones sustained by SPE2 are retained as the initial condition for the SPE3 calculations. The results indicate that the waveform peak amplitudes are slightly decayed since the ensemble model includes no softening mechanism. The pulse widths are reduced, indicating the energy is dissipated more by the damage zone as expected. The dominant motions calculated in all the scenarios are spherically symmetric and non-spherical components are small as seen for the SPE2 calculations. The arrival time of the SPE3 waveform at the boreholes is shorter than SPE2 due to compaction during the previous explosion. As shown in the SPE 1 and 2 observations shear waves are generated, indicating the potential contributions from the induced motion on distributed joints. The effects of joints are evaluated using the Lagrangian hydrocode (GEODYN-L), which was benchmarked against the GEODYN calculations for the ensemble model. A new granite model calibrated for samples from the SPE site is used. The new model includes effects of poroelasticity and water saturation and may be scaled allowing application to both small and large scales. Parametric tests were performed to evaluate the effect of joints within the granite and the results demonstrate that in situ joints introduce anisotropy and significant tangential motions.

58 GEOSCIENCES↗
DOE OSTI2012

Enhancements in SCALE 6.1

The SCALE code system developed at Oak Ridge National Laboratory provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor physics, radiation shielding, and sensitivity and uncertainty analysis. For more than 30 years, regulators, licensees, and research institutions around the world have used SCALE for nuclear safety analysis and design. SCALE provides a 'plug-and-play' framework with 89 computational modules, including three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution. SCALE's graphical user interfaces assist with accurate system modeling, visualization, and convenient access to desired results. SCALE 6.1 builds on the existing capabilities and ease-of-use of SCALE and provides several new features such as enhanced lattice physics capabilities and multigroup Monte Carlo depletion, improved options and capabilities for sensitivity and uncertainty analysis calculations, improved flexibility in shielding and criticality accident alarm system calculations with automated variance reduction, and new options for the definition of group structures for depletion calculations. The SCALE 6.1 development team has focused on improved robustness via substantial additional regression testing and verification for new and existing features, providing improved performance relative to SCALE 6.0, especially in reactor physics calculations and in the nuclear data used for source term characterization and shielding calculations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Silicon Carbide Temperature Monitor Processing Improvements. Status Report

Silicon carbide (SiC) temperature monitors are used as temperature sensors in Advanced Test Reactor (ATR) irradiations at the Idaho National Laboratory (INL). Although thermocouples are typically used to provide real-time temperature indication in instrumented lead tests, other indicators, such as melt wires, are also often included in such tests as an independent technique of detecting peak temperatures incurred during irradiation. In addition, less expensive static capsule tests, which have no leads attached for real-time data transmission, often rely on melt wires as a post-irradiation technique for peak temperature indication. Melt wires are limited in that they can only detect whether a single temperature is or is not exceeded. SiC monitors are advantageous because a single monitor can be used to detect for a range of temperatures that occurred during irradiation. As part of the process initiated to make SiC temperature monitors available at the ATR, post-irradiation evaluations of these monitors have been previously completed at the High Temperature Test Laboratory (HTTL). INL selected the resistance measurement approach for determining irradiation temperature from SiC temperature monitors because it is considered to be the most accurate measurement. The current process involves the repeated annealing of the SiC monitors at incrementally increasing temperature, with resistivity measurements made between annealing steps. The process is time consuming and requires the nearly constant attention of a trained staff member. In addition to the expensive and lengthy post analysis required, the current process adds many potential sources of error in the measurement, as the sensor must be repeatedly moved from furnace to test fixture. This time-consuming post irradiation analysis is a significant portion of the total cost of using these otherwise inexpensive sensors. An additional consideration of this research is that, if the SiC post processing can be automated, it could be performed in an MFC hot cell, further reducing the time and expense of lengthy decontaminations prior to processing. Sections of this report provide a general description of resistivity techniques currently used to infer peak irradiation temperature from silicon carbide temperature monitors along with some representative data, the proposed concepts to improve the process of analyzing irradiated SiC temperature monitors, the completed efforts to prove the proposed concepts, and future activities. The efforts detailed here succeeded in designing and developing a real-time automated SiC resistivity measurement system, and performed two initial test runs. Activities carried out include the assembly and integration of the system hardware; the design and development of a preliminary monitor fixture; the design of a technique to automate the data analysis and processing; the development of the communication, coordination, and user software; and the execution and troubleshooting of test run experiments using the box furnace. Although the automation system performed as required, the designed fixture did not succeed in establishing the needed electrical contacts with the SiC monitor.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

NDMAS

Overview of Current ART-GCR Data: Fuel Fabrication, Irradiation Monitoring (Fuel & Graphite – near real-time for HDG-1), Post-Irradiation Examination (Fuel & Graphite), Graphite Characterization (Baseline and Irradiated), High Temperature Metals Mechanical Tests, Design, Methods, and Validation Data, Japan Atomic Energy Agency’s High Temperature Test Reactor (HTTR), Argonne National Laboratory’s Natural convection Shutdown heat removal Test Facility (NSTF), Oregon State University’s High Temperature Test Facility (HTTF), Generation IV International VHTR Materials Handbook, Additional related data, and Advanced Test Reactor operations (near real-time).

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

NDT System for Metal Components in Concrete Structures (Final Report)

Steel reinforcements embedded within concrete can corrode and crack, leading to weakening of the structures, and existing NDT (non-destructive testing) techniques do not provide the full picture. Condition assessment is critical in evaluating these structures in order to ensure safe, long-term operation of the nuclear power plants. Improved NDT techniques are needed as the nation’s generating stations age and require action based on increasingly critical inspection data. The power industry uses eddy current test (ECT) during in-service inspections to detect and monitor defects such as cracks and corrosion in metals, particularly metal pipes. However, the detection of defects in the reinforcing steel embedded in concrete is far more challenging for numerous reasons. Primary among them is the concrete layer’s thickness, which prevents the ability to have a close proximity between sensor and metallic sample. One solution to this problem is to create a higher sensitivity technique which can inspect at depth, with good spatial resolution and excellent directional sensitivity, to isolate the metal buried deep in the concrete and characterize defects in the metal. The goal of the phase II research program was to study and develop a new reinforced concrete NDT technology based on anisotropic magnetoresistive (AMR) sensors designed specifically for ECT of steel reinforcements in concrete. The first technical objective was to demonstrate the capability of inspecting steel reinforcement structures, including metal rebar and tensioning cable at or below 6” of depth embedded in concrete. Another objective was to be able to confidently identify multiple layers of reinforcements at varying depths. In addition, the technique needed to be able to provide condition assessment of the metal, providing data on the structure and positioning of the steel members, material loss, and detection of the presence of breaks. An important part of the research and development was aimed at achieving superior performance compared to existing technology. The ability to detect corrosion breaks in unbonded multi-wire and multi-strand tendons was another important objective. The final objective that led to a successful program was to build a complete prototype portable scanning system capable of demonstrating the probe performance in the field. All of these objectives were met, which provided data for the system development work going forward. RMD worked with the engineering firm Simpson Gumpertz & Heger to develop a detailed problem definition for concrete structures and the necessary sensor specifications. In addition, we developed finite element simulation models for the eddy current based NDT system through the analysis of the electromagnetic (EM) properties of sample materials and the orientation of the transmitter, sensor array (receivers), metal, and intervening materials like concrete and air. Based on the modeling and specifications, we designed and fabricated prototype sensor modules. We then acquired and fabricated test samples to characterize and optimize the sensors. Finally, we built a complete prototype for testing on actual concrete structures. This new NDT technology has high promise for greatly improving capabilities in the inspection of steel reinforcements within concrete structures, such as rebar, liners, anchors, and prestressing cables. This method could be valuable in the maintenance of nuclear power plants, and many other structures critical to our nation’s infrastructure.

42 ENGINEERING↗
DOE OSTI2012

PHASE II VAULT TESTING OF THE ARGONNE RFID SYSTEM

The U.S. Department of Energy (DOE) (Environmental Management [EM], Office of Packaging and Transportation [EM-45]) Packaging and Certification Program (DOE PCP) has developed a Radio Frequency Identification (RFID) tracking and monitoring system, called ARG-US, for the management of nuclear materials packages during transportation and storage. The performance of the ARG-US RFID equipment and system has been fully tested in two demonstration projects in April 2008 and August 2009. With the strong support of DOE-SR and DOE PCP, a field testing program was completed in Savannah River Site's K-Area Material Storage (KAMS) Facility, an active Category I Plutonium Storage Facility, in 2010. As the next step (Phase II) of continued vault testing for the ARG-US system, the Savannah River Site K Area Material Storage facility has placed the ARG-US RFIDs into the 910B storage vault for operational testing. This latest version (Mark III) of the Argonne RFID system now has the capability to measure radiation dose and dose rate. This paper will report field testing progress of the ARG-US RFID equipment in KAMS, the operability and reliability trend results associated with the applications of the system, and discuss the potential benefits in enhancing safety, security and materials accountability. The purpose of this Phase II K Area test is to verify the accuracy of the radiation monitoring and proper functionality of the ARG-US RFID equipment and system under a realistic environment in the KAMS facility. Deploying the ARG-US RFID system leads to a reduced need for manned surveillance and increased inventory periods by providing real-time access to status and event history traceability, including environmental condition monitoring and radiation monitoring. The successful completion of the testing program will provide field data to support a future development and testing. This will increase Operation efficiency and cost effectiveness for vault operation. As the next step (Phase II) of continued vault testing for the ARG-US system, the Savannah River Site K Area Material Storage facility has placed the ARG-US RFIDs into the 910B storage vault. Deploying the ARG-US RFID system lends to a reduced need for manned surveillance and increased inventory periods by providing real-time access to status and event history traceability, including radiation and environmental monitoring. The successful completion of the testing program will provide field data to support future development and testing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

NBSR reactor control room upgrade - 215

The maintenance and modernization of the National Institute of Standards and Technology (NIST) Center for Neutron Research (NCNR) reactor, known as the NBSR, is an ongoing process to endure until the eventual decommissioning of the reactor. As part of the renovation effort, the reactor control system is to be upgraded. Following some initial outsourcing efforts, the NBSR reactor control system upgrade has been restructured to be implemented by in-house means. The beginning stage of the upgrade involved the development of a simulator environment based on a PC system and digitization of the majority of nuclear reactor monitoring data. The NBSR reactor surveillance data consists of approximately thirteen hundred input/output points, including analog/digital sensor signals, signal conditioners, and relay logic. The aim of the current phase is to implement a new reactor data acquisition and display system (RDS) and lay out the infrastructure for the upgrade of the reactor control system (RCS). Key design criteria are to implement a reliable, redundant, diverse, modular and candid system where any complex digital manipulations are taken outside of the control room. Basically, the data analytics and custom codes are separated from reactor operation control and decision making logic. Furthermore, the new RDS is designed such that it could be maintained internally by NCNR personnel. The RDS is unique in design and modularity, minimizing custom codes and testing, making it ideal for research reactors. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Use of Data Libraries for IAEA Nuclear Security Assessment Methodologies (NUSAM) [section 5.4]

Data libraries are essential for the characterization of the facility and provide the documented input which enables the facility assessment results and subsequent conclusions. Data Libraries are historical, verifiable, quantified, and applicable collections of testing data on different types of barriers, sensors, cameras, procedures, and/or personnel. Data libraries are developed and maintained as part of any assessment program or process. Data is collected during the initial stages of facility characterization to aid in the model and/or simulation development process. Data library values may also be developed through the use of state testing centers and/or site resources by testing different types of barriers, sensors, cameras, procedures, and/or personnel. If no data exists, subject matter expert opinion and manufacturer's specifications/ testing values can be the basis for initially assigning values, but are generally less reliable and lack appropriate confidence measures. The use of existing data libraries that have been developed by a state testing organization reduces the assessment costs by establishing standard delay, detection and assessment values for use by multiple sites or facilities where common barriers and alarms systems exist.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
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 OSTI2016

Evaluating alternate test techniques to characterize mechanical properties in nuclear-grade graphites - 18421

The deployment of graphite as a core structural material in the very high temperature reactor (VHTR) design presents a number of challenges with regard to accurate property and performance assessments. In the United States Advanced Reactor Technologies (ART) program, candidate VHTR graphite grades are being evaluated not only with respect to irradiated property values, but also with a focus on comprehensive data sets that explicate the property variability inherent to nuclear-grade graphites. The large data sets being compiled on baseline physical and mechanical properties provide a means not only to quantify variability within and between major graphite grades, but also to evaluate test techniques utilized on irradiated specimens that are severely limited in size and geometry. Detailed appraisals of emerging mechanical test techniques that either (a) fall outside of the prescribed test size and configuration constraints in order to be assembled into irradiation capsule such as the three-point bend test on sub-sized cylinders or (b) are not yet fully evaluated and standardized by ASTM International such as the Brazilian disc splitting tensile strength test can be more accurately generated when a broad comparison is possible that leverages multiple mechanical tests on full-size specimens using approved standards. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Predicting High Energy Arcing Fault Zones of Influence for Aluminum Using an Arc Flash Modeling Approach: Evaluation of a model bias, uncertainty, parameter sensitivity and zone of influence estimation

This report documents the development of an arc flash hazard model to calculate the incident energy and zone of influence from high energy arcing faults involving aluminum. The NRC has identified the potential for (HEAFs) involving aluminum to increase the damage zone beyond what is currently postulated in fire probabilistic risk assessment (PRA) methodologies. To estimate the hazard from HEAFs involving aluminum an arc flash model was developed. Differences between the initial model and nuclear power plant (NPP) fire PRA scenarios were identified. Modification of the initial model established from existing literature and test data was used to minimize these differences. The developed model was evaluated against NRC datasets to understand the model prediction and relative uncertainties. Finally, a range of fire PRA zone of influences (ZOI) were developed based on the developed model, target fragility estimates and update HEAF PRA methodology. The results were developed to support an NRC LIC-504 evaluation in tandem with other modeling efforts. The report documents the effort and provides a reference for any future advancements in arc flash modeling.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Validation of Thermal Scattering Laws for Light Water at Elevated Temperatures with Diffusion Experiments

Light water is the most important neutron moderator in many reactor applications. Thermal neutron scattering kernels for H bound in H 2 O are represented by ENDF-format thermal scattering laws (TSLs) evaluated at discrete temperatures T . Nuclear data evaluations are conventionally validated utilizing a combination of critical benchmarks and experimental cross section data. Existing public critical benchmarks may be inadequate for testing water TSLs over the full range of T of interest in reactor applications, and experimental thermal scattering cross section data for water is sparse at elevated T . In this work, MC21 is used to simulate the decay of the equilibrium thermal neutron flux in light-water spheres of several radii. The fundamental-mode time eigenvalue $\mathcal{a}$ is calculated at 22 °C and 227 °C (at saturation pressure) for each sphere using three different H-H 2 O TSLs. Fitting $\mathcal{a}$ to a polynomial function of geometric buckling allows calculation of the thermal neutron diffusion length L . Extrapolation lengths are treated as a function of the transport mean free path and geometry. Experimental L data from 25 publications at 49 temperatures (from 10 °C to 295 °C), computed by several different time-dependent and space-dependent decay methods, is used to develop an empirical fit of L vs. T . The MC21-calculated L for the TSLs tested are within ±1% of the predicted values at 22 °C and 227 °C. This validation approach, which may be repeated at arbitrary T , constitutes an integral benchmark specific to and characterizing the detailed physics of the thermal scattering kernel applied.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Sister Rod Destructive Examinations (FY22)

As a part of DOE NE High Burnup Spent Fuel Data Project, Oak Ridge National Laboratory (ORNL) is performing destructive examinations (DEs) of high burnup (HBU) (>45 GWd/MTU) spent nuclear fuel (SNF) rods from the North Anna Nuclear Power Station operated by Dominion Energy. The SNF rods, called sister rods or sibling rods, are all HBU and include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin (LT) Zirc-4, ZIRLO, and M5. This report documents the status of the ORNL Phase 1 DEs of seven sister rods and outlines the DE tasks performed and the data collected to date, as guided by the sister rod test plans.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Burnt Mountain Alaska RTG Extraction, Transport and Disposal Project - 17359

The Air Force Technical Applications Center (AFTAC), headquartered at Patrick Air Force Base (AFB), Florida, removed excess nuclear material from one of the center's seismic arrays on Burnt Mountain, Alaska, after nearly 15 years of un-operability, and transferred it to a the low-level waste disposal facility at the Nevada National Security Site (NNSS). Ten radioisotope thermoelectric generators (RTGs), were used as a power source by AFTAC at its array 97 kilometers (60 miles) north of the Arctic Circle. Originally placed on Burnt Mountain in 1973, the first RTG was installed to power a single seismometer. Twelve years later, due to the increase in communications equipment and demand for additional power, nine more RTGs were placed at the site. For the next 17 years, the RTGs were operated and maintained by AFTAC's Detachment 460 based at Eielson AFB near Fairbanks, Alaska, and were used because of their high reliability and low maintenance requirements to determine if regional seismic activity was caused by nuclear explosions or naturally-occurring events such as earthquakes, volcanic activity, etc. AFTAC's primary mission is to verify compliance with nuclear test ban treaties. In 1992, a tundra fire encroached on the Burnt Mountain site and damaged some of the data cables. Despite the fact the fire did not impact the generators themselves, it raised public concern among nearby inhabitants about the safety of using radioactive material as a power source at the station. The Air Force, in coordination with multiple agencies and mission partners, made the decision to remove the RTGs and relocate them to NNSS for permanent disposal. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Legacy Metallic Fuel U(Pu)Zr Data Qualification

The Metallic Fuels Irradiation and Physics Database (FIPD) is an organized collection of legacy metallic fuel U(Pu)Zr data, measurements, and reactor conditions from the Experimental Breeder Reactor II (EBR-II) and the Fast Flux Test Facility (FFTF). The database provides a wealth of information and is easily accessed and utilized by the U.S. nuclear industry and the Nuclear Regulatory Commission (NRC). The database contains three categories of data. The first category covers fuel pin fabrication specifications, including fuel slug diameter and length, cladding diameter, smeared density, etc. The second consists of operational conditions, which include axial distributions of power, temperature, fluence, burnup, isotopic density, etc. The third category contains post-irradiation examination (PIE) results, largely collected at facilities now part of the Idaho National Laboratory (INL), formerly part of Argonne National Laboratory -West (ANL-W) and at the Alpha-Gamma Hot Cell Facility (AGHCF) at Argonne National Laboratory, comprising fission gas release and chemistry analysis, profilometry measurements, neutron radiography data, etc. All these data have been organized and preserved in FIPD.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Underground Test Area (UGTA) Sampling Plan for Corrective Action Units 101 and 102: Central and Western Pahute Mesa, Nevada National Security Site, Nevada

The purpose of this groundwater sampling plan (referred to herein as “the Plan”) is to describe the approach for collecting and analyzing groundwater samples to meet the objectives of the U.S. Department of Energy (DOE), Environmental Management (EM) Nevada Program’s Underground Test Area (UGTA) Activity. The Plan is focused on the Central and Western Pahute Mesa corrective action units (CAUs 101 and 102). The remaining UGTA CAUs are in the final stage of the UGTA strategy (Closure Report [CR] stage), and monitoring requirements for these CAUs are described in their respective CRs. The Plan supersedes the Nevada National Security Site Integrated Groundwater Sampling Plan and its ensuing Record of Technical Change (ROTC) (DOE/EMNV, 2019b). The Plan is designed to ensure compliance with the UGTA Quality Assurance Plan (QAP) and its ensuing ROTC (DOE/EMNV, 2018), and the Federal Facility Agreement and Consent Order (FFACO) (1996, as amended). The Plan is implemented by the Environmental Program Services (EPS) contractor. Implementation of the Plan provides data to complete corrective actions prescribed by the FFACO to protect the public from groundwater contaminated by historical underground nuclear testing. The Plan’s scope comprises sample collection and analysis requirements for assessing the extent of groundwater contamination from underground nuclear testing. Sampling wells located in the communities surrounding the Nevada National Security Site (NNSS) is performed by other organizations (e.g., management and operating [M&O] contractor) and is therefore not included within the Plan. The results of the community sampling are presented in the annual NNSS environmental reports.

54 ENVIRONMENTAL SCIENCES↗