Search NASA⌕ Search

SEARCH · Search NASA

Results for “Nuclear Data Testing”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 865 records · Page 48
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Prediction of Creep-Induced Strain Using a Symbolic Regression-Based Model

Material creep under high-temperature conditions limits the lifetime and safety of structural systems such as advanced nuclear reactors. Conventional creep testing is slow and often produces inconsistent results across nominally identical experiments, making lifetime prediction uncertain. Here, to address these challenges, this work develops a data-driven symbolic regression (SR) model that consolidates results from duplicate creep tests and predicts the remaining strain-time curve of an ongoing experiment. The method uses piece-wise multi-objective SR with physical constraints to generate analytic, interpretable functions describing transient creep strain. Applied to Inconel Alloy 617 data, the approach achieved relative mean absolute errors of 1.0–9.5%, providing closed-form predictions of strain evolution. These results demonstrate a first step toward reducing the duration and cost of long-term creep testing while retaining physically interpretable model forms.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Spectral X-ray Radiography for Safeguards at Nuclear Fuel Fabrication Facilities: A Feasibility Study

The methods currently used by the International Atomic Energy Agency to account for nuclear materials at fuel fabrication facilities are time consuming and require in-field chemistry and operation by experts. Spectral X-ray radiography, along with advanced inverse algorithms, is an alternative inspection that could be completed noninvasively, without any in-field chemistry, with inspections of tens of seconds. The proposed inspection system and algorithms are presented here. The inverse algorithm uses total variation regularization and adaptive regularization parameter selection with the unbiased predictive risk estimator. Performance of the system is quantified with simulated X-ray inspection data and sensitivity of the output is tested against various inspection system instabilities. Material quantification from a fully-characterized inspection system is shown to be very accurate, with biases on nuclear material estimations of < 0.02%. It is shown that the results are sensitive to variations in the fuel powder sample density and detector pixel gain, which increase biases to 1%. Options to mitigate these inaccuracies are discussed.

Gilbert, Andrew J.↗
DOE OSTI2014

Testing the improved method for calculating the radiation heat generation at the periphery of the BOR-60 reactor core

The application of the improved method for calculating the radiation heat generation in the elements of an experimental device located at the periphery of the BOR-60 reactor core results in a significant reduction in the discrepancies between the calculated and the experimental data. This allows us to conclude that the improved method has an advantage over the one used earlier.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Fission Product Yield Data Adjustment in a Prototype Version of TSURFER

The TSURFER (Tool for Sensitivity/Uncertainty analysis of Response Functionals using Experimental Results) module of Oak Ridge National Laboratory’s (ORNL’s) SCALE code system has been updated to perform nuclear data adjustments for fixed-source irradiation/depletion problems. TSURFER uses a generalized linear least squares (GLLS) approach to consolidate a prior set of measured responses and corresponding calculated values to create the most self-consistent set of nuclear data. Traditionally, TSURFER adjustments have been performed for multigroup nuclear data such as reaction cross sections. In this work, TSURFER is expanded to perform adjustments to independent fission product yields and branching ratios that need equality constraints. To preserve equality constraints after the data adjustment procedure, an updated GLLS formulation includes a new Lagrange multiplier that forces data adjustment to sum to 0 for a given fission yield/branching ratio parent. A test problem illustrates that the newly updated TSURFER module satisfies the required constraint that adjustments for fission yield data sum to 0.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

Application of Nonlinear Seismic Soil-Structure Interaction Analysis for Identification of Seismic Margins at Nuclear Power Plants

Seismic probabilistic risk assessment (SPRA) methods and approaches at nuclear power plants (NPP) were first developed in the 1970s and aspects of them have matured over time as they were applied and incrementally improved. SPRA provides information on risk and risk insights and allows for some accounting for uncertainty and variability. As a result, SPRA is now used as an important basis for risk-informed decision making for both new and operating NPPs in the US and in an increasing number of countries globally. SPRAs are intended to provide best estimates of the various combinations of structural and equipment failures that can lead to a seismic induced core damage event. However, in some instances the current SPRA approach contains large uncertainties, and potentially masks other important events (for instance, it was not the seismic motions that caused the Fukushima core melt events, but the tsunami ingress into the facility). INL has an advanced SPRA research and development (R&D) activity that will identify areas in the calculation process that contain significant uncertainties. One current area of focus is the use of nonlinear soil-structure interaction (NLSSI) analysis methods to accurately capture: 1) nonlinear soil behavior and 2) gapping and sliding between the NPP and soil. The goal of this study is to compare numerical NLSSI analysis results with recorded earthquake ground motions at Fukushima Daichii (Great Tohuku Earthquake) and evaluate the sources of nonlinearity contributing to the observed reduction in peak acceleration. Comparisons are made using recorded data in the free-field (soil column with no structural influence) and recorded data on the NPP basemat (in-structure response). Results presented in this study should identify areas of focus for future R&D activities with the goal of minimizing uncertainty in SPRA calculations. This is not a validation activity since there are too many sources of uncertainty that a numerical analysis would need to consider (variability in soil material properties, structural material properties, etc.). Rather the report will determine if the NLSSI calculations are following similar trends observed in the recorded data (i.e. reductions in maximum acceleration between the free-field and basemat) Numerical NLSSI results presented show maximum accelerations between the free field and basemat were reduced the EW and NS directions. The maximum acceleration in the UD direction increased slightly. The largest reduction in maximum accelerations between the modeled free-field and the NPP basemat resulted in nearly 50% reduction. The observation in reduction of numerical maximum accelerations in the EW and NS directions follows the observed trend in the recorded data. The maximum reductions observed in these NLSSI studies were due to soil nonlinearities, not gapping and sliding (although additional R&D is needed to develop an appropriate approach to model gapping and sliding). This exploratory study highlights the need for additional R&D on developing: (i) improved modeling of soil nonlinearities (soil constitutive models that appropriately capture cyclic soil behavior), (ii) improved modeling of gapping and sliding at the soil-structure interface (to appropriately capture the dissipation of energy at this interface), and (iii) experimental laboratory test data to calibrate the items (i) and (ii).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

AGC-2 Specimen Post Irradiation Data Package Report

This report documents results of the post-irradiation examination material property testing of the creep, control, and piggyback specimens from the irradiation creep capsule Advanced Graphite Creep (AGC)-2 are reported. This is the second 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. The AGC-2 capsule was irradiated in the Idaho National Laboratory Advanced Test Reactor at a nominal temperature of 600°C and to a peak dose of 5 dpa (displacements per atom). One half of the creep specimens were subjected to mechanical stresses (an applied stress of either 13.8, 17.2, or 20.7 MPa) to induce irradiation creep. All post-irradiation testing and measurement results are reported with the exception of the irradiation mechanical strength testing, which is the last destructive testing stage of the irradiation testing program. Material property tests were conducted on specimens from 15 nuclear graphite grades using a similar loading configuration as the first AGC capsule (AGC-1) to provide easy comparison between the two capsules. However, AGC-2 contained an increased number of specimens (i.e., 487 total specimens irradiated) and replaced specimens of the minor grade 2020 with the newer grade 2114. The data reported include specimen dimensions for both stressed and unstressed specimens to establish the irradiation creep rates, mass and volume 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, electrical resistivity, and thermal diffusivity and thermal expansion data from 100–500°C. No data outliers were determined after all measurements were completed. A brief statistical analysis was performed on the irradiated data and a limited comparison between pre- and post-irradiation properties is presented. A more complete evaluation of trends in the material property changes, as well as irradiation-induced creep due to irradiation, temperature, and applied load on specimens will be discussed in later AGC-2 post-irradiation examination analysis reports.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

GAIN Technology Workshops Summary Report

National and global demand for nuclear energy is increasing and United States (U.S.) global leadership is eroding. There is a sense of urgency with respect to the deployment of the innovative nuclear energy technologies. The Gateway for Accelerated Innovation in Nuclear (GAIN) initiative is based on the simultaneous achievement of three strategic goals. The first is maintaining global technology leadership within the U.S. Department of Energy (DOE). The second is enabling global industrial leadership for nuclear vendors and suppliers. The third is focused on utility optimization of nuclear energy within the clean energy portfolio. An effective public-private partnership is required to achieve these goals. DOEs recognizes the recent sense of urgency new developers and investors have in getting their concepts to market. They know that time to market for nuclear technology takes too long and the facilities needed to conduct the necessary research, development and demonstration (RD&D) activities are very expensive to develop and maintain. Early technologies, in the lower technology readiness levels (TRL) need materials testing, analysis, modeling, code development, etc., most of which currently exists in the DOE national laboratory system. However, mature technologies typically need large component testing and demonstration facilities, which are expensive and long-lead efforts. By understanding the needs of advanced nuclear technology developers, GAIN will connect DOE national laboratory capabilities (e.g., facilities, expertise, materials, and data) with industry RD&D needs. In addition, GAIN is working with the Nuclear Regulatory Commission (NRC) to streamline processes and increase understanding of the licensing requirements for advanced reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Evaluation of Historic Nuclear Cloud Measurement for Entrainment Parameter Determination

We evaluated the uncertainties associated with the entrainment parameters computed from nuclear cloud measurements taken during historic weapon testing campaigns. This evaluation found that variations in the way cloud heights and diameters were defined during the analysis of film records result in very large uncertainties in the entrainment parameters, severely limiting the utility of these data. For the measurements of atmospheric conditions at the time of the tests, the main sources of error are associated with the differences in the location and timing of the soundings.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2023

Fission with Exotic Nuclei (Full Technical Report)

Despite its importance for stockpile stewardship or nuclear forensics, data on fission is fragmentary: most experiments have only been performed on stable actinide nuclei and are often incomplete, and theoretical simulations often contain many parameters hard to constrain, which results in large uncertainties. Consequently, nuclear libraries have large gaps for important isotopes. The goal of this project was to prepare to take advantage of the unprecedented yields of radioactive isotopes at the upcoming DOE Facility for Rare Ion Beams and to leverage recent progress in the field of machine learning to develop a comprehensive program of fission studies that could address some of the most pressing problems in nuclear fission over the next several decades. Our project leveraged synergies between experimental nuclear physics, nuclear theory, and data science expertise at LLNL. On the experimental front, our main objective was to develop in-house expertise for inverse kinematics reactions with relativistic beams as well as to field and test new detectors to perform correlated measurements of fission properties. On the theoretical side, the objective was to develop a novel, high-fidelity and scalable approach of fissionfragment calculations based on emulating computationally expensive, quantum-mechanical calculations of nuclear properties with deep neural networks.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2012

Simulated NMIS Imaging Data for an Unknown Object

This report presents simulated Nuclear Materials Identification System (NMIS) imaging data for an unclassified object, whose characteristics are initially unknown to the reader. This data will be used to test various analysis capabilities and was created with a simple deterministic ray-tracing algorithm. NMIS is a time-dependent coincidence counting system that is used to characterize both fissile and non-fissile materials undergoing nondestructive assay. NMIS characterizes materials by interrogating them with neutrons, either from an associated-particle deuterium-tritium (DT) neutron generator, which produces a time and directionally tagged monoenergetic beam of 14.1 MeV neutrons, or a time-tagged spontaneous fission source in an ionization chamber.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2022

Update on the Investigation of Commercial Drying Cycles Using the Advanced Drying Cycle Simulator

The purpose of this report is to document updates on the apparatus to simulate commercial vacuum drying procedures at the Nuclear Energy Work Complex at Sandia National Laboratories. Validation of the extent of water removal in a dry spent nuclear fuel storage system based on drying procedures used at nuclear power plants is needed to close existing technical gaps. Operational conditions leading to incomplete drying may have potential impacts on the fuel, cladding, and other components in the system during subsequent storage and disposal. A general lack of data suitable for model validation of commercial nuclear canister drying processes necessitates well-designed investigations of drying process efficacy and water retention. Scaled tests that incorporate relevant physics and well-controlled boundary conditions are essential to provide insight and guidance to the simulation of prototypic systems undergoing drying processes. This report documents a new test apparatus, the Advanced Drying Cycle Simulator (ADCS). This apparatus was built to simulate commercial drying procedures and quantify the amount of residual water remaining in a pressurized water reactor (PWR) fuel assembly after drying. The ADCS was constructed with a prototypic 17×17 PWR fuel skeleton and waterproof heater rods to simulate decay heat. These waterproof heaters are the next generation design to heater rods developed and tested at Sandia National Laboratories in FY20. This report describes the ADCS vessel build that was completed late in FY22, including the receipt of the prototypic length waterproof heater rods and construction of the fuel basket and the pressure vessel components. In addition, installations of thermocouples, emissivity coupons, pressure and vacuum lines, pressure transducers, and electrical connections were completed. Preliminary power functionality testing was conducted to demonstrate the capabilities of the ADCS. In FY23, a test plan for the ADCS will be developed to implement a drying procedure based on measurements from the process used for the High Burnup Demonstration Project. While applying power to the simulated fuel rods, this procedure is expected to consist of filling the ADCS vessel with water, draining the water with applied pressure and multiple helium blowdowns, evacuating additional water with a vacuum drying sequence at successively lower pressures, and backfilling the vessel with helium. Additional investigations are expected to feature failed fuel rod simulators with engineered cladding defects and guide tubes with obstructed dashpots to challenge the drying system with multiple water retention sites.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Implementation of Distributed Memory Computing in MOSAIC to Enable Large 3D Simulations of Irradiated Concrete

The concrete biological shield (CBS) of light-water reactors protects workers and the surrounding environment by absorbing neutron and gamma irradiation emitted from the reactor core. The radiation dose increases with the CBS’s operational time and, in the long term, becomes significant enough to raise the question of irradiation effects on concrete—and particularly on the structural integrity of the CBS. Irradiation-induced damage has been identified as one of the main degradation mechanisms in the CBS. Neutron radiation causes the swelling of aggregate-forming minerals at different rates and amplitudes depending on the mineral’s nature. Silicate-bearing minerals such as quartz are particularly sensitive to neutron radiation and experience up to 17.8% volumetric expansion. Aggregates comprise several minerals with different orientations and are, therefore, subject to cracking as a result of mismatch strains. Additionally, the swelling of aggregates creates significant stresses in the surrounding cement paste matrix, which also results in crack formation. In parallel with the collection of characterization and irradiation test data, development of modeling and simulation tools for irradiated concrete is ongoing with the support of the US Department of Energy Office of Nuclear Energy’s Light Water Reactor Sustainability (LWRS) program. This effort resulted in the development and application of the fast-Fourier transform (FFT)–based code Microstructure-Oriented Scientific Analysis of Irradiated Concrete (MOSAIC) at Oak Ridge National Laboratory.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2022

Monte Carlo Simulation and Reconstruction: Assessment of Myocardial Perfusion Imaging of Tracer Dynamics With Cardiac Motion Due to Deformation and Respiration Using Gamma Camera With Continuous Acquisition

Purpose: Myocardial perfusion imaging (MPI) with single photon emission computed tomography (SPECT) is routinely used for stress testing in nuclear medicine. Recently, our group extended its potential going from 3D visual qualitative image analysis to 4D spatiotemporal reconstruction of dynamically acquired data to capture the time variation of the radiotracer concentration and the estimated myocardial blood flow (MBF) and coronary flow reserve (CFR). However, the quality of reconstructed image is compromised due to cardiac deformation and respiration. The work presented here develops an algorithm that reconstructs the dynamic sequence of separate respiratory and cardiac phases and evaluates the algorithm with data simulated with a Monte Carlo simulation for the continuous image acquisition and processing with a slowly rotating SPECT camera. Methods: A clinically realistic Monte Carlo (MC) simulation is developed using the 4D Extended Cardiac Torso (XCAT) digital phantom with respiratory and cardiac motion to model continuous data acquisition of dynamic cardiac SPECT with slowly rotating gamma cameras by incorporating deformation and displacement of the myocardium due to cardiac and respiratory motion. We extended our previously developed 4D maximum-likelihood expectation-maximization (MLEM) reconstruction algorithm for a data set binned from a continuous list mode (LM) simulation with cardiac and respiratory information. Our spatiotemporal image reconstruction uses splines to explicitly model the temporal change of the tracer for each cardiac and respiratory gate that delineates the myocardial spatial position as the tracer washes in and out. Unlike in a fully list-mode data acquisition and reconstruction the accumulated photons are binned over a specific but very short time interval corresponding to each cardiac and respiratory gate. Reconstruction results are presented showing the dynamics of the tracer in the myocardium as it continuously deforms. These results are then compared with the conventional 4D spatiotemporal reconstruction method that models only the temporal changes of the tracer activity. Mean Stabilized Activity (MSA), signal to noise ratio (SNR) and Bias for the myocardium activities for three different target-to-background ratios (TBRs) are evaluated. Dynamic quantitative indices such as wash-in (K1) and wash-out (k2) rates at each gate were also estimated. Results: The MSA and SNR are higher with higher TBRs while biases were improved with higher TBRs to less than 10%. The correlation between exhalation-inhalation sequence with the ground truth during respiratory cycle was excellent. Our reconstruction method showed better resolved myocardial walls during diastole to systole as compared to the ungated 4D image. Estimated values of K1 and k2 were also consistent with the ground truth. Conclusion: The continuous image acquisition for dynamic scan using conventional two-head gamma cameras can provide valuable information for MPI. Our study demonstrated the viability of using a continuous image acquisition method on a widely used clinical two-head SPECT system. Our reconstruction method showed better resolved myocardial walls during diastole to systole as compared to the ungated 4D image. Precise implementation of reconstruction algorithms, better segmentation techniques by generating images of different tissue types and background activity would improve the feasibility of the method in real clinical environment.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2014

First negative ion beam measurement by the Short-Time Retractable Instrumented Kalorimeter Experiment (STRIKE)

The Source for Production of Ion of Deuterium Extracted from Rf plasma (SPIDER) test facility is under construction in Padova to optimise the operation of the beam source of ITER neutral beam injectors. The SPIDER beam will be characterised by the instrumented calorimeter STRIKE, whose main components are one-directional carbon-fibre-carbon-composite tiles. A small-scale version of the entire system has been employed in the BAvarian Test MAchine for Negative ions (BATMAN) testbed by arranging two prototype tiles in the vertical direction. The paper presents a description of the mini-STRIKE system and of the data analysis procedures, as well as some results concerning the BATMAN beam under varying operating conditions.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2017

Modeling the Transient Reactor Test Loop in TRACE for PWR Critical Heat Flux Benchmarking

Understanding transient critical heat flux (CHF) resulting from a Reactivity Insertion Accident (RIA) in nuclear systems is an extremely undeveloped field. Though many experiments have been done for CHF, remarkably few have been done for rods at PWR operating temperature and pressure (300 deg. C, 2200 psia) with a short time constant ramp in power. This leads to many concerns regarding understanding how RIA scenarios develop and evolve in commercial nuclear reactors. With a limited understanding of the phenomenology governing transient CHF, there is little hope in being able to accurately model these scenarios using today's thermal hydraulic codes. The Accident Tolerant Fuels (ATF) Program established by the Department of Energy (DOE) is seeking out new fuel and cladding systems that can be used to reduce the probability and consequences of severe accidents. To better understand this phenomenon, the DOE is restarting the Transient Reactor Test (TREAT) Facility at the Idaho National Laboratory (INL). TREAT will be able to provide representative neutronic boundary conditions for testing coupled thermal hydraulics on the Transient Water Environment Recirculating Loop (TWERL) to study transient CHF with the best representation possible. To support this task, Oregon State University (OSU) is designing and building a representative flow loop that will be able to mimic the conditions of TWERL (sans the neutronic boundary conditions) to create a benchmark study for existing lumped parameter thermal-hydraulic codes such as RELAP5 and TRACE. This flow loop, the Transient Reactor Test Facility (TRTL), will support the ATF program by providing much needed NQA-1 data for transient CHF at operating temperature and pressure.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

{gamma}-ray measurements and neutron sensitivity in a fusion environment

A system based on a LaBr{sub 3}(Ce) scintillator and digital data acquisition is discussed in view of {gamma}-ray spectroscopy measurements on a fusion burning plasma. High energy resolution and high rate capability of the system are demonstrated in laboratory tests and in dedicated experiments at nuclear accelerators. Preliminary results on sensitivity of the detector to 14 MeV neutrons are presented.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

AGR-1 Irradiation Test Final As-Run Report, Rev. 3

This document presents the as-run analysis of the AGR-1 irradiation experiment. AGR-1 is the first of eight planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the US Department of Energy (DOE) as part of the Next-Generation Nuclear Plant (NGNP) project. The objectives of the AGR-1 experiment are: 1. To gain experience with multi-capsule test train design, fabrication, and operation with the intent to reduce the probability of capsule or test train failure in subsequent irradiation tests. 2. To irradiate fuel produced in conjunction with the AGR fuel process development effort. 3. To provide data that will support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. In order to achieve the test objectives, the AGR-1 experiment was irradiated in the B-10 position of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) for a total duration of 620 effective full power days of irradiation. Irradiation began on December 24, 2006 and ended on November 6, 2009 spanning 13 ATR cycles and approximately three calendar years. The test contained six independently controlled and monitored capsules. Each capsule contained 12 compacts of a single type, or variant, of the AGR coated fuel. No fuel particles failed during the AGR-1 irradiation. Final burnup values on a per compact basis ranged from 11.5 to 19.6 %FIMA, while fast fluence values ranged from 2.21 to 4.39 x 10 25 n/m 2 (E >0.18 MeV). We’ll say something here about temperatures once thermal recalc is done. Thermocouples performed well, failing at a lower rate than expected. At the end of the irradiation, nine of the originally-planned 19 TCs were considered functional. Fission product release-to-birth (R/B) ratios were quite low. In most capsules, R/B values at the end of the irradiation were at or below 10 -7 with only one capsule significantly exceeding this value. A maximum R/B of around 2 x 10 -7 was reached at the end of the irradiation in Capsule 5. Several shakedown issues were encountered and resolved during the first three cycles. These include the repair of minor gas line leaks; repair of faulty gas line valves; the need to position moisture monitors in regions of low radiation fields for proper functioning; the enforcement of proper on-line data storage and backup, the need to monitor thermocouple performance, correcting for detector spectral gain shift, and a change in the mass flow rate range of the neon flow controllers.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

A Next Generation Digital Counting System For Low-Level Tritium Studies (Project Report)

Since the early seventies, SRNL has pioneered low-level tritium analysis using various nuclear counting technologies and techniques. Since 1999, SRNL has successfully performed routine low-level tritium analyses with counting systems based on digital signal processor (DSP) modules developed in the late 1990s. Each of these counting systems are complex, unique to SRNL, and fully dedicated to performing routine tritium analyses of low-level environmental samples. It is time to modernize these systems due to a variety of issues including (1) age, (2) lack of direct replacement electronics modules and (3) advances in digital signal processing and computer technology. There has been considerable development in many areas associated with the enterprise of performing low-level tritium analyses. The objective of this LDRD project was to design, build, and demonstrate a Next Generation Tritium Counting System (NGTCS), while not disrupting the routine low-level tritium analyses underway in the facility on the legacy counting systems. The work involved (1) developing a test bed for building and testing new counting system hardware that does not interfere with our routine analyses, (2) testing a new counting system based on a modern state of the art DSP module, and (3) evolving the low-level tritium counter design to reflect the state of the science.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗