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At least 451 records · Page 25
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Theoretical and calculable dependent variables and their covariance in nuclear data libraries [Slides]

This presentation begins by defining observables, including theoretical observable, calculable observable, and measured observable. It also provides definitions and examples of experimental effects. Additionally, uncertainty in evaluated libraries and a metric to probe evaluated uncertainty is presented. In conclusion, being related to theoretical quantities such as resonance parameters, the uncertainty in ENDF libraries (e.g., ENDF/B-VIII.0) can be overestimated because it is usually evaluated under the guidance of experimental uncertainty: lim Δσexp.corr.→0 Δσ = Δσ theoretical model . The presentation states that a clear distinction of the uncertainty between nuclear theoretical models and experimental corrections should be revisited and this is important since (e.g., transport) simulations need theoretical quantities convoluted with specific operational parameters and material configurations. It also states that strong coupling between uncertainty quantification methodologies and optimization procedures exists, and it is necessary to develop methodologies to obtain uncertainty on theoretical models from physical and mathematical constraints.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Availability of Critical Benchmark Experiments for the Pebble Tanker Transportation Model for Nuclear Criticality Safety Validation of TRISO Pebbles

This study addresses the need for comprehensive investigations into TRi-structural ISOtropic (TRISO) fuel pebble transportation validation. In this work, an exploratory model, the pebble tanker(PT), was developed with the aim of facilitating the validation of nuclear criticality safety calculations in the context of industrial-scale transportation of TRISO fuel. The PT model was designed to investigate the availability and applicability of critical benchmark experiments crucial for assessing the transportation of these pebbles. This work incorporated sensitivity/uncertainty (S/U) similarity studies to quantify the applicability of critical benchmark experiments and to address nuclear data uncertainties in the context of TRISO transportation. Two container models were investigated: one for the Hermes-type pebble and one for the Pebble Bed Modular Reactor (PBMR)–type pebble. The models were simplified, considering fuel, containment, and either water or air, to enable a focus on the underlying physics of applications involving TRISO fuel pebbles using the PT model. A crucial aspect under consideration was the capacity of the transport package to hold pebbles while ensuring subcriticality in the flooded state. An approach in the criticality validation process involves assessing the similarity between systems through an integral index parameter evaluation. This involves calculating a correlation coefficient (referred to as c k ) based on shared nuclear data–induced uncertainty between a benchmark experiment and the application of the PT model. To facilitate this analysis, the SCALE tools, particularly the CSAS6-Shift, TSUNAMI-3D-Shift, and TSUNAMI-IP sequences, were employed for comprehensive studies in neutronics and S/U analysis. Our findings showed that there are sufficient critical experimental benchmarks to perform this validation of the PT model in the most reactive state, i.e. when the tanker is flooded. This paper provides valuable insights into validating a transport package for Generation IV TRISO fuel pebbles.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

State of Stress Uncertainty Quantification and Geomechanical Risk Analysis for Subsurface Engineering

All subsurface stress measurement methods available for deep subsurface applications are indirect and involve a significant amount of uncertainty. Quantifying these uncertainties is critical for applications where the state of stress is a component of a risk analysis with significant safety, environmental, and economic considerations. A publicly available and open source tool called the State of Stress Analysis Tool (SOSAT) has been developed to aid in performing a Bayesian uncertainty quantification and geomechanical risk analysis using many types of commonly available data. A new feature of this tool is using data from the FutureGen 2.0 site. The new feature of the tool was uses information about the absence or presence of borehole breakouts to constrain the state of stress. This new feature allows the user to specify probability distributions for the relevant parameters such as the drilling mud pressure, temperature, and the formation strength properties. The tool computes the posterior joint probability distribution of the horizontal principal stresses. Using random samples from this posterior distribution the tool then computes the probability of activating a critically oriented fault at a range of pore pressures. This latter feature can be used to aid in the assessment of the risk of induced seismicity and leakage in geologic carbon sequestration and enhanced geothermal systems.

Burghardt, Jeffrey A.↗
DOE OSTI2026

HFIR Steady State Heat Transfer Code (HSSHTC) Statistical Uncertainty Analysis

HSSHTC, the safety basis steady state TH code for HFIR, uses a highly conservative approach in which all input and calculation uncertainties are resolved simultaneously at their most limiting setting. This results in excessive conservatism which does not account for the high unlikelihood of such simultaneous worst-case conditions. The present study explores an alternative approach, BEPU, in which reasonable working assumptions for the probability distribution of each input uncertainty are used to determine a relationship between burnout power margin and core fuel failure probability. This was performed under a philosophy of perturbing uncertainty parameters already defined within the HSSHTC methodology while preserving the HSSHTC calculation approach and solution methodology itself. Based on the assumptions employed in this study, the BEPU approach resulted in a 0.29 increase in burnout power ratio (25 MW increase in burnout power) compared to the latest HSSHTC calculations of C-HFIR-2026-004. The study can be refined in the future by employing fuel fabrication data to provide more realistic input distributions. Future changes to the HSSHTC methodology would potentially allow a more comprehensive treatment of uncertainties which may further increase the burnout power ratio.

Wysocki, Aaron [ORNL] (ORCID:0000000222043779)↗
DOE OSTI2025

Impact of U-10Mo HALEU Fuel Element Tolerances on the Massachusetts Institute of Technology Reactor safety and operational performance – Neutronics

The U.S. is coordinating efforts for the conversion of six U.S. high performance research reactors (USHPRR), including one critical assembly from highly enriched uranium (HEU) to low-enriched uranium (LEU). In order to continue the mission of these reactors, including the Massachusetts Institute of Technology Reactor (MITR), and achieve similar performance, high assay low-enriched uranium (HALEU) with a high-density metallic alloy of uranium with 10 wt% molybdenum (U-10Mo) is being considered. Following the preliminary design of the proposed MITR LEU fuel elements using the U-10Mo monolithic alloy, the impact of the fabrication specification was assessed. This work focuses on the analysis of select neutronics characteristics of the MITR LEU core as a function of the variation of the relevant fuel specification parameters (e.g., U-10Mo composition, fuel plate thickness, etc.). A separate article submitted to this journal addresses the impact on the thermal hydraulic performance. The analyses in these works are performed based on an all-LEU conversion management plan identified in previous work, in which only the proposed elements are utilized for achieving the conversion of MITR. The variations of two main neutronics characteristics are assessed as a function of the variability of the specification parameters resulting from the fabrication process: the MITR LEU core reactivity and the fuel cycle length. The main findings of this work show that the MITR core can meet the operational requirements during the LEU transition plan under the limiting fabrication parameter combinations considered. In addition, the analyses show that the dependency of the core neutronics characteristics on the specification parameters is highly linear within the specification tolerances. The rates of variation are reported in detail for each parameter and can serve as a powerful tool for future MITR fuel management in cases such as when HALEU supply is established that may allow additional cycle length or other operational benefits.

Conversion↗
DOE OSTI2026

Are CO2 batteries ready for the grid? A cross-cutting deployment checklist

Metal-CO2 batteries have attracted considerable research interest in the past decade due to the utilization of the greenhouse CO2 gas. The low cost and abundance of CO2 feedstocks, together with high specific capacity metal anodes, make this technology a promising candidate for electricity-demanding data centers and grids. However, nonstandard data reporting, ambiguous hardware designs, and questionable cycling parameters have emerged as the field has grown. A clear checklist and standardized reporting format for research and development are urgently needed. Here, after carefully re-analyzing 15 Na-CO2 and 9 Li-CO2 studies, we propose a general checklist and a flowchart for CO2 batteries and related gas conversion systems. Specific performance values from a deployment-oriented perspective, based on commercial lithium iron phosphate batteries, are integrated into the list. We frame research and development efforts around grid-relevant outputs, including areal rate and capacity, cyclability, gas utilization, and safety. While illustrated for Na- and Li-CO2 cells, this framework is directly transferable to other CO2 batteries and air/oxygen batteries, helping align benchtop demonstrations with systematic metrics that ultimately govern future deployment.

Wu, Wenda [ORNL] (ORCID:000900033307687X)↗
DOE OSTI2023

Future of the MUSiC Experiment Data

The Measurements of Uranium Subcritical and Critical (MUSiC) experiment was a highly enriched uranium (HEU) experiment performed at the National Criticality Experiments Research Center (NCERC) executed between December 2020 and April 2021. The experiment intended to measure criticality and reactor kinetics parameters in a bare HEU system. The experiment concurrently measured radiation signatures from the system while utilizing different neutron source types. This was an attempt to benchmark both detectors and analysis techniques against one another for identical measurements. The experiment consisted of the Rocky Flats HEU hemi-shells constructed into ten configurations spanning between deeply subcritical (about 14 kgs) to supercritical (about 60 kgs). Two of the ten configurations were supercritical and the other eight were subcritical. The two supercritical configurations, often referred to as the critical configurations, were documented into an International Criticality Safety Benchmark Evaluation Project (ICSBEP) benchmark evaluation and submitted to the technical review group (TRG). The subcritical configurations of the MUSiC experiment were examined using three different detection systems. The systems include: the NoMAD He-3 neutron detector which consists of 15 He-3 tubes surrounded by a polyethylene matrix, a liquid scintillator system called the Rossi-α Measurement Rapid Organic Discriminating Detector (RAM-RODD), and a trans-stilbene organic scintillator array (OSCAR) provided by the University of Michigan. The NoMAD and RAM-RODD data are planned to be evaluated in two separate ICSBEP evaluations utilizing neutron noise methods such as Feynman Variance-to-Mean, Rossi-α and the pulsed neutron source method. Each of the subcritical configurations were examined using three different source types including: a Cf-252 source in the center, with only the intrinsic neutron source in the HEU, and with an external D-T neutron generator.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Internal short circuit and failure mechanisms of lithium-ion pouch cells under mechanical indentation abuse conditions: An experimental study

Electromechanical structural integrity and thermal stability dictate the safety performance of lithium-ion batteries. Progressive deformation and failure across microscopic and macroscopic lengths scales that are responsible for internal short circuit (ISC) in lithium-ion cells under mechanical abuse conditions remains elusive. In this study, a series of indentation tests were conducted on lithium-ion cells with different capacities up to the occurrence of ISC. The external response and internal configuration of these cells were investigated. It is discovered that cells with different capacities and state of charges exhibited different behaviors. Maximum temperature, which is often regarded as the most important parameter related to thermal runaway (TR), varied considerably due to the complicated contact configurations. X-ray computed tomography (XCT) showed that ISC was a collective result of shear band or other strain-localization modes in the electrode assembly, shear offsets in the granular coatings of electrodes, and the accompanying ductile fracture in the metal foils. We believe that the irregular strain-localization modes (kinks, cusps, and buckles), radical mismatches in mechanical properties of different layers, and geometric features of the indenter eventually lead to the tearing/puncture of cell separator at various locations. Furthermore, the results could provide useful guidance for the micromechanical modeling of lithium-ion cells.

25 ENERGY STORAGE↗
DOE OSTI2021

149 Sm Results from DICER Plus DANCE [Slides]

DICER and DANCE can provide detailed and accurate resonance parameters and on smaller samples than any other facility from sub-thermal energies through the RRR. Many additional experiments of interest to NCSP could easily be performed at DICER and DANCE. DANCE also can provide accurate (n,γ) cross sections in the URR.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Equipment Qualification Report Environmental Qualification of GNB Absolyte Valve Regulated Lead Acid (VRLA) 1600 Ah100G33 Battery Rack Assembly (24590-QL-POA-EDB0-00001-11-00002_00A)

Greenberry Environmental Qualification Report 550001.001-35.0.5 provides basis for assignment of qualified life for: GNB Absolyte Valve Regulated Lead Acid (VRLA) 1600 Ah 100G33 Battery Rack Assembly in accordance with the requirements specified 24590-WTP-3PS-G000-T0015 (Rev 2) and Environmental Qualification Plan 550001.001-35.0.1 (Rev. 3). The equipment qualification basis represents the most conservative capability of the equipment. The analysis performed for the qualification is not less conservative than the bounding environmental conditions detailed in contract documents issued to Greenberry in contract 24509-QL-POA-EDB0-00001 Rev.0. The qualified life of 10 years has been established based upon an end-condition objective of the equipment condition indicators that correlate to the ability of equipment to perform its safety function. The VRLA Battery Cell Assembly was aged by 10 years (minimum) in accordance with conditions specified by Bechtel Equipment Qualification Datasheet 24590-LAW-EUQ-UPE 00003 Rev. 3 and the process conditions specified by the Instrument Data Sheet 24590-LAW EUD-UPE-00009 Rev. 2. Greenberry Industrial has contracted with GNB Industrial Battery Co located at 4115 S Zero St, Fort Smith, AR 72908 to perform age conditioning, monitoring, and capacity testing in accordance with Environmental Qualification Plan 550001.001-35.0.1 Rev. 1. The required process at the stated conditions set by the parameters established by the plan were completed satisfactorily. The details of the of the test process observed by Greenberry is detailed in the attached Seismic Test Log 550001.001-7.0.3, including examples of the objective evidence collected during the qualification process.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2021

Progress on the reevaluation and validation of the n+233U neutron cross sections

The set of 233 U resonance parameters of the ENDF/B-VIII.0 nuclear data library was adopted from the previous ENDF/B-VII.1 evaluation using the external levels to update the thermal values. Adoption of IAEA 2017 thermal standards ( σ f = 533.0 ± 2.2 b, σ c = 44.9 ± 0.9 b, and ν ‾ tot = 2.487 ± 0.011 ) and of the IAEA-recommended thermal-neutron induced prompt fission neutron spectrum (PFNS) with average PFNS energy of 2.030 ± 0.013 MeV requires a re-evaluation of 233 U neutron cross sections in the resolved resonance region. A newly produced evaluation is being tested on benchmarks carefully selected from the Handbook of International Criticality Safety Benchmark Experiments (ICSBEP) which are highly sensitive to 233 U data. An important goal of this work was to eliminate the strong negative gradient of the calculated effective multiplication factors with respect to the epithermal fission fraction observed in the validation of the ENDF/B-VIII.0 library for those assemblies. A significant improvement in integral performance of critical 233 U solutions is observed for the newly proposed evaluation. Further work addressing the fast neutron region is needed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Sputter sample preparation for ion beam delivery of radium-223 at ATLAS

A radium-223 ion beam was delivered to an experiment from the electron cyclotron resonance ion source, ECR2, at the Argonne Tandem Linac Accelerator System (ATLAS). Here, the radium-223 material was in a nitrate salt form within a vial, prior to being converted to a usable sputter sample. The sputter sample was produced using a new sample preparation method, where the radium nitrate was dissolved into a solution and pipetted onto pressed aluminum powder. This sample was then allowed to dry, distributing the radium-223 material throughout the sputter sample. Ion source operation using the radium sputter sample is described with the operating parameters listed. The intensity and energy requirements for this ion beam were 1 × 10 6 particles/s and 1.07 GeV, respectively. Because the intensity is relatively low compared to most experiments at ATLAS, previously developed accelerator mass spectrometry methods were used Scott et al. to avoid the need for tuning of the low-intensity beam of interest. Handling of the radium material, as well as loading and unloading of the sputter sample from ECR2, required collaboration with Health Physics. Procedures were used and dry runs were carried out before, during, and after the experiment to ensure the safety of the workers. The processes used and lessons learned are described within.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

Accelerating nuclear fuel development and qualification: Modeling and simulation integrated with separate-effects testing

In this work, an approach to transform and accelerate nuclear fuel development and qualification is outlined. The approach exploits advanced modeling and simulation at the outset to inform constituent and system selection and to enable integral fuel performance analyses. Analyses using these tools identify and prioritize the most important fuel performance parameters and phenomena for subsequent targeted characterization with separate-effects tests. Separate-effects testing spans out-of-pile and in-pile tests and is meant to iterate with and inform engineering-scale integral fuel performance analyses throughout the development process. Exercising this cycle in an agile fashion will increase confidence in the integral fuel performance predictions while reducing uncertainties. This process sets the stage for executing a much more limited set of well-defined integral irradiation tests designed to validate engineering-scale fuel performance codes and to confirm the performance and safety of the fuel system under prototypic conditions. This approach will reduce the time for development and qualification of a new fuel system, and it will also reduce associated costs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Nuclear Data Adjustment for Nonlinear Applications in the OECD/NEA WPNCS SG14 Benchmark—A Bayesian Inverse UQ-Based Approach for Data Assimilation

The Organisation for Economic Co-operation and Development Working Party on Nuclear Criticality Safety has proposed a benchmark exercise to assess the performance of current nuclear data adjustment techniques applied to nonlinear applications and experiments with low correlation to applications. This work introduces Bayesian inverse uncertainty quantification (IUQ) employing scientific machine learning surrogate models as a method for nuclear data adjustments in this benchmark, and compares IUQ to the more traditional methods of generalized linear least squares (GLLS) and Monte Carlo Bayes (MOCABA). Posterior predictions from IUQ showed agreement with GLLS and MOCABA for linear applications. Here, when comparing GLLS, MOCABA, and IUQ posterior predictions to computed model responses using adjusted parameters, we observe that the GLLS predictions failed to replicate the computed response distributions for nonlinear applications, while MOCABA showed near agreement, and IUQ used the computed model responses directly. We also discuss observations on why experiments with low correlation to applications can be informative to nuclear data adjustments and identify some properties useful in selecting experiments for inclusion in nuclear data adjustment. Performance in this benchmark indicates potential for Bayesian IUQ in nuclear data adjustments.

Bayesian calibration↗
DOE OSTI2020

Data-driven analysis of relight variability of jet fuels induced by turbulence

For safety purposes, reliable reignition of aircraft engines in the event of flame blow-out is a critical requirement. Typically, an external ignition source in the form of a spark is used to achieve a stable flame in the combustor. However, such forced turbulent ignition may not always successfully relight the combustor, mainly because the state of the combustor cannot be precisely determined. Uncertainty in the turbulent flow inside the combustor, inflow conditions, and spark discharge characteristics can lead to variability in sparking outcomes even for nominally identical operating conditions. Prior studies have shown that of all the uncertain parameters, turbulence is often dominant and can drastically alter ignition behavior. For instance, even when different fuels have similar ignition delay times, their ignition behavior in practical systems can be completely different. In practical operating conditions, it is challenging to understand why ignition fails and how much variation in outcomes can be expected. The focus of this work is to understand relight variability induced by turbulence for two different aircraft fuels, namely Jet-A and a variant named C1. A detailed, previously developed simulation approach is used to generate a large number of successful and failed ignition events. Using this data, the cause of misfire is evaluated based on a discriminant analysis that delineates the difference between turbulent initial conditions that lead to ignition or failure. From the discriminant analysis, a compressed sensing algorithm is then applied to help pinpoint the locations of relevant turbulent features. Findings from the discriminant analysis are confirmed with the time history of near kernel properties. Next, a clustering strategy is used to identify ignition and misfire modes. With this approach, it was determined that the cause of ignition failure is different for the two fuels. While it was found that Jet-A is influenced by fuel entrainment, C1 was found to be more sensitive to small scale turbulence features. Finally, a larger variability is found in the ignition modes of C1, which can be subject to extreme events induced by kernel breakdown.

42 ENGINEERING↗
DOE OSTI2021

Dose Rate Analysis of the WCS Consolidated Interim Storage Facility

This report describes a confirmatory calculation for a proposed consolidated interim storage facility (CISF) to support the US Nuclear Regulatory Commission’s review of the license application submitted by the Interim Storage Partners. The scope of this report is limited to an assessment of the annual dose from direct radiation associated with Phase 1 of the Waste Control Specialists (WCS) CISF and the determination of the minimum controlled area boundary based on the 10 CFR 72.104 limit of 25 mrem annual dose to the whole body. The confirmatory dose rate calculations used the source term and shielding calculation capabilities of the SCALE 6.2.3 computer code system. The calculation method employed in the confirmatory calculations uses a detailed Monte Carlo radiation transport simulation from source to dose rate. This method differs from the two-step method used in the safety analysis report (SAR), which requires determination of the photon and neutron energy and angular distributions on the cask external surface and then the use of this surface source in a new radiation transport calculation to determine dose rate as a function of distance from the storage facility. A simulation was made using the complete site geometry (all casks present) but with only one cask containing source, and a total of 467 independent calculations were performed to obtain the dose rate maps produced by each storage cask. Basic input data and assumptions used in the SAR (e.g., cask design parameters and design basis assembly characteristics) were used in this confirmatory calculation for consistency with the SAR. Local atmospheric conditions for skyshine calculations were simulated using US standard atmosphere data, which describe average air temperature, pressure, and density as a function of altitude. The determined minimum required distances from the WCS Phase 1 CISF to a contour defined by the 25 mrem annual dose limit were approximately 623 m (0.387 miles) in the N-NE direction, 594 m (0.369 miles) in the SSW direction, and 533 m (0.331 miles) in the E-SE direction and the W-NW direction. The annual dose results of the confirmatory calculation are in relatively good agreement with the annual dose values presented in the WCS SAR. The annual dose produced by the Phase 1 CISF at the nearest site boundary, 0.75 miles away from the CISF, was estimated by this study to be approximately 6.02E-02 ± 2.0E-04 mrem, which is significantly smaller than the 25 mrem annual dose limit provided in 10 CFR 72.104. The SAR annual dose rate value for the nearest site boundary was 7.52E-2 mrem, which included 7.73E-03 mrem due to postulated leakage of the FO-, FC-, and FF-canisters. A comparison with the dose rate values reported in the SAR for 16 detector locations, D1 through D16, on the protected area boundary is also included in this report. This comparison shows that the SAR determined higher annual dose values for detectors D1 through D14 by 12% to 45% and lower annual dose values for detectors D15 and D16 by 18% and 8%, respectively, as compared with this ORNL confirmatory calculation. Sensitivity studies have been performed to support a comprehensive review of the application. In these studies, the effects on dose rate of the modeled air volume, density, and humidity were analyzed. The sensitivity calculations showed that the air volume used in the Monte Carlo simulations was adequate and that a larger air volume would not further increase the dose rate due to skyshine contributions. The sensitivity analysis showed that simulation of local atmospheric conditions is important for the accuracy of dose rate estimates at large distances from the facility. Air humidity was demonstrated to have negligible effects on dose rate values for the simulated local atmospheric conditions.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2022

Open-source release of CGMF 1.1 and Integration into the MCNP6.3 ® Code [Slides]

As a result of a multi-year NA-22 project, CGMF was integrated into MCNP6.2 and publicly released. CGMF was open-sourced and publicly released and MCNP6.3 was updated to include the latest version and is in the process of being publicly released. Current and future plans include global optimization and uncertainty quantification within CGMF, model parameter fitting such that CGMF may be used in ENDF/B evaluations, and improving both standalone and MCNP-integrated CGM (non-fission) simulations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Modernization efforts for the R -Matrix code SAMMY [Abstract]

The R-Matrix code SAMMY is a widely used nuclear data evaluation code focused on the resolved range, which includes corrections for experimental effects. The code is still mostly written in Fortran 77, and uses a memory management system suitable for the time of its initial writing (1984). A modernization effort is under way to bring the code in-line with modern software development practices. A continuous-integration testing framework was added, automating the large existing set of test cases. It is run on every commit. The memory management was updated to current standard practices suitable for modern software analysis tools. The code can be obtained from https://code.ornl.gov/RNSD/SAMMY. The resonance parameters and covariance information are now stored in C++ objects shared by SAMMY and AMPX, the processing code that generates nuclear data libraries for SCALE. This allows for easier maintenance and access to the resonance parameters inside and outside of SAMMY. This feature is already used by accessing and changing parameters in memory in the Bayesian Monte Carlo Evaluation Framework for Cross Sections Nuclear Data and Integral Benchmark Experiments project, Further plans include the switch to the ENDF reading and writing routines in AMPX, as these routines are more robust, easier to maintain, and support more features. Of note here is support for the new GNDS format. Previously it wasn’t easy to share the full covariance matrix for evaluations containing more than one isotope due to limitations on the ENDF format; this is now supported in GNDS. The data are currently available in a binary SAMMY format and can be exported to GNDS to make them more widely available and sharable. The next step will be to use the same resonance processing code at 0K in AMPX and SAMMY as one of the available Reich-Moore R-Matrix formalism. The first step toward this goal is to isolate the reconstruction into a module that takes resonance parameters as its input and does not depend on SAMMY global parameters. This goal has been achieved and it should now be possible to more easily change the resonance formalism and add enhancements as the Phenomenological R-Matrix parameterization of direct, doorway, and compound nuclear reactions discussed elsewhere on this conference. This concerted modernization and enhancement effort provides multiple advantages to the nuclear data community. It will allow parameter optimization using enhanced formalisms, including experimental effects, that better match complex experimental data. Then those evaluated parameters can immediately be passed off to AMPX to be reconstructed with the exact same cross section model and be put into a data library for subsequent testing using SCALE and the Valid Benchmark suite or other suitable benchmark suites.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗