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This presentation looks at past years of NCERC Experiments. With a Look to the future and Integral Experiments and secondary measurements covered.
This report documents the experimental configurations and measurements for IER-532, Thermal/Epithermal eXperiments (TEX) with highly enriched uranium (HEU) fuel and interstitial hafnium (Hf), moderated and reflected by polyethylene. TEX-Hf is a variation of and based on the TEX-HEU (IER-297) design, with the inclusion of hafnium. These configurations provide integral experiments for validation of hafnium in the thermal, intermediate, and fast neutron energy regimes by maximizing the sensitivity in k eff to the hafnium isotope cross sections. The experiment campaign was completed over seven weeks during the end of FY22 and beginning of FY23 at the National Critical Experiments Research Center at the Nevada National Security Site. The campaign produced seven experimental configurations, four reproducibility measurements, and many additional dimensional measurements that will be of use to the future benchmark evaluation of this experiment and other experiments using the same HEU fuel. Table 1 summarizes the TEX-Hf experimental configurations, including their physical parameters, calculated fission fractions, and estimated excess reactivities.
he project PARADIGM (PARallel Approach of Differential and InteGral Measurements) answers this question by selecting via machine learning (ML) an optimal combination of differential and integral experiments to reduce 239 Pu nuclear data uncertainties from 1-600 keV by 50%..
This presentation touches on reducing uncertainties in intermediate-energy actinide nuclear data and this continues to be a high priority for many applications. The goal of PARADIGM (PARallel Approach of Differential and InteGral Measurements): accelerate efforts to reduce biases and uncertainties in nuclear data through improvements to the nuclear data pipeline. This presentation includes integral experiments and a summarization of existing copper nuclear data.
How do we accelerate scientific progress in the Nuclear Data (ND) field? By selecting via Machine Learning (ML) an optimum combination of differential and integral experiments to reduce ND uncertainties.
The low-temperature variant of the TEX HEU (called Low-Temperature TEX or sometimes LT TEX) campaign is a highly anticipated and necessary experimental series by the greater nuclear science community. Fundamentally, the need for low-temperature integral experiments is required to perform validation of cross sections below room temperature. There has been substantial international interest in low-temperature benchmarks to validate below room temperature cross sections, namely talks given at the 2019 International Conference on Nuclear Criticality (ICNC): UK (Watson, 2019), France (Milin, 2019), and UK (Gan & Wilson, 2019). Additionally, NCSP funded thermal scattering laws (TSLs) were produced by North Carolina State University and require low-temperature benchmarks to validate them. Validation of low-temperature cross sections is also necessary for criticality safety applications. One particularly important application is to ensure that during transportation, fissile materials must remain subcritical under normal ambient conditions which is defined as temperatures down to -40°C/°F by the United States 10 CRF 71 as well as a regulation put forward by the International Atomic Energy Agency (IAEA).
Nuclear reaction data libraries make use of experimental data that are differential in reaction and neutron energy for evaluation and data that are integral in reaction and energy for validation. In many cases, integral measurements can provide lower experimental uncertainties than differential measurements, but the impact of individual reactions and energies cannot generally be determined from the single experimental observable. “Quasi-integral” or “quasi-differential” measurements, which are integral in one characteristic (energy or reaction) and differential in the other, can often provide a middle ground of lower experimental uncertainties combined with some ability to unpack the impact of individual reactions or energies. Furthermore, the Flexible Neutron Source (FNS) at the University of Tennessee, Knoxville, is a new experimental facility that can perform integral and quasi-integral measurements by removing components of the integral configuration to determine the impact of specific reactions. Possible flux-averaged cross-section measurements, flux perturbation measurements, and neutron downscattering measurements are simulated here to show the value of the FNS to evaluation and validation of nuclear data.
Abstract not provided.
The Transient Reactor Test Facility (TREAT) loss-of-coolant (LOC) and high-burnup (HBu) experiment series, along with the Severe Accident Test Station (SATS) HBu experiment series, are integral LOC accident (LOCA) experiments planned under the DOE AFC program, which aim to support burnup extension needs by addressing identified R&D priorities in order to achieve an improved understanding of fuel fragmentation, relocation, and dispersal (FFRD) of HBu fuel during LOCA events. The data produced under this plan will be used to further validate and confirm existing models and inform future R&D and model development. The experimental program has been specifically designed to address knowledge gaps and opportunities identified through a detailed review of existing public knowledge on LOCA FFRD. The test program employs a unique combination of in- and out-of-pile experimental approaches and state-of-the-art facilities to provide a clear connection to the existing integral and semi-integral LOCA experiment database. The primary goal of the program is to investigate the impact of prototypic HBu fuel/cladding thermomechanical behaviors under postulated LWR LOCA conditions that have not yet been fully studied. These conditions correspond with prototypic decay-energy heatup (DEH) and stored-energy heatup (SEH) conditions. Importantly, TREAT’s unique capability will enable the first evaluation of the impact of SEH conditions on HBu fuels. The test program will emphasize the development of an improved mechanistic understanding of key experimental phenomena through independent experimental systems, development of a database to support fuel performance modeling tools and employing world-leading advanced materials characterization and in-situ diagnostics to evaluate FFRD. The results of the program will provide novel data to support modeling development and validation and will represent a significant advancement in evaluating prototypic conditions, as well as to inform the technical basis for LOCA-induced FFRD.
Following the shutdown of the CEA Valduc experimental facilities, where, for more than 50 years, IRSN used to perform experiments related to criticality safety, IRSN initiated a new project named PRINCESS (PRoject for IRSN Neutron physics and Criticality Experimental data Supporting Safety). The objective is to continue collecting experimental data necessary for the IRSN missions in nuclear safety. For this purpose, collaborations with various national and international laboratories have been established. The PRINCESS project covers various nuclear physics fields from nuclear data to criticality-safety and reactor physics providing information to both differential and integral data improvements.
In this work, we report here the results of a measurement of the scattered versus unscattered neutron fluence on polyethylene determined via neutron activation of multiple natural indium foils from a deuterium-deuterium (D-D) neutron generator. The neutrons were produced by the High Flux Neutron Generator (HFNG) at the University of California, Berkeley, a specially designed source to maximize neutron flux on a sample while minimizing the total neutron yield. During the experiment, approximately 10 8 n/s were produced with the energies at the indium foils ranging from 2.2 to 2.8 MeV. Both the angle-integrated and the partial angle differential results are consistent with the predictions of the Monte Carlo N-Particle Transport (MCNP) code, using ENDF/B-VII.1. This supports shielding calculations in the fast energy region with high-density polyethylene.
Lead by Lawrence Livermore National Laboratory under the U.S. Department of Energy's Nuclear Criticality Safety Program. The goal of TEX is to provide integral benchmark experiments than span the entire neutron energy spectrum and incorporate high-priority materials. TEX includes two test bed configurations providing a baseline for comparison to better understand the contribution of additional materials
ABSTRACT Nuclear data (ND) are the input data for neutron‐transport simulations to answer questions related to nuclear technologies. Subsets of ND, here > 20,000 data points, are validated with respect to thousands of criticality experiments that represent various applications on a small scale. The aim of validation with these experiments is to find errors in ND or methods. The key challenge here is that several hundreds of ND are used to simulate one integral value. Hence, one cannot clearly identify what ND are leading to bias in criticality measurements. In fact, a mistake in one nuclear‐data observable can be compensated with an error in another, and the predicted criticality value would still be predicted in agreement with experimental data. Random forest (RF) was previously employed to predict bias in criticality measurements using sensitivities of simulated criticality experiments to ND. The SHapley Additive exPlanations (SHAP) metric was then applied to attribute the importance of each ND experiment and observable to bias prediction. This, however, did not highlight what ND were jointly related to predicting bias. This is important as it could inform us about where compensating errors in ND could hide. We tackle this shortcoming here by first decomposing the ND sensitivities to integral‐experiment simulations into principal components. Then we use principal component projections to predict bias via the RF and SHAP. The SHAP values and principal components are employed to reconstruct detailed SHAP values for each ND observable. We demonstrate that these extended SHAP bias predictions are more robust, less noisy, and more efficient. In addition, we show that this approach accounts for covariance in ND sensitivities and automates the identification of where compensating errors could hide in ND.
Abstract not provided.
This report documents the experimental critical configurations and measurements for IER-297, Thermal/Epithermal eXperiments (TEX) using highly enriched uranium (HEU) fuel, moderated and reflected by polyethylene. These experiments establish the baseline TEX-HEU configurations spanning thermal, intermediate, and fast fission energy regimes. The experiment campaign was completed in FY2020 at the National Critical Experiments Research Center at the Nevada National Security Site. The campaign produced six documented critical configurations (0”, ¹/₈” A, ¹/₈” B, ¹/₄”, ¹/₂”, and 1¹/₂”) and one reproducibility measurement (¹/₈” R). Table 1 summarizes these seven experiments along with their period and calculated reactivity, in order of execution. The Comet General Purpose Critical Assembly Machine was used to conduct the TEX-HEU experiments. The HEU fuel consists of U(93+) plates, collectively known as the “Jemima” plates. Each experiment consisted of an alternating stack of these HEU plates with or without polyethylene moderator plates. By varying the thicknesses of the polyethylene plates, the neutron spectrum of the experiment could be fine-tuned to a specific fission energy regime. The experiments were reflected by 1” of polyethylene consisting of a bottom and top plate and outer reflector rings to surround the HEU and polyethylene plates. The fine reactivity control of the system was provided by the thickness of the top reflector plate, which could be varied in increments of ¹/₃₂”.
Washington River Protection Solutions (WRPS) uses MCNP6.2 and the Whisper code for criticality safety analyses of the Hanford Tank Farm. Together the codes derive baseline upper subcritical limits (USLs) for the waste models using experimental benchmarks. Whisper returns higher USLs, i.e. , has less of a conservative penalty, when the neutronic similarity of the experimental benchmarks to the application is high. Unfortunately, few critical benchmarks have high similarity to the Hanford tanks. The waste in the tanks is highly dilute in plutonium and contains large masses of weakly neutron-absorbing elements like iron and manganese. Experimental benchmarks typically have low sensitivity to these absorbers because they are present as structural materials. Lacking similar benchmarks, new Thermal Epithermal eXperiment (TEX) configurations with high Pu content and interstitial iron absorbers have been designed for the criticality safety validation. The features of the design have been iterated upon to maximize the similarity between the experiment and different Hanford waste models. The similarity is quantified with sensitivity analysis and uncertainty quantification using the representativity coefficient, or c k . The representativity calculation requires nuclear data covariances, which may differ between nuclear data libraries and between library versions. Because of these variations, the optimal design may depend on the nuclear data covariances library. A scenario can be envisioned where an experiment is designed, and c k is maximized, with one set of covariance data. However, when the covariance data is changed, say from ENDF/B-VII.1 to ENDF/B-VIII.0, and the benchmark is used in a criticality safety evaluation, the experiment becomes suboptimal with respect to c k . In this paper, we present how the optimal design of the new TEX experiments varied depending on the nuclear data covariances used to calculate c k . We compare ENDF/B-VII.1 and ENDF/B-VIII.0, as if the library had been updated since the design of the experiment. Additionally, we use JEFF3.3 to simulate if the covariance data of a different library had been used. The results show that the covariances do have an important effect on the designs, less so for thermal systems (where the data are more consistent between evaluations) and more so for epithermal systems where more differences exist.
This report presents the final design (CED-2) for three additional mixed-spectra configurations for plutonium Thermal/Epithermal eXperiments (TEX) to target the intermediate energy region (IER-553). The baseline cases of IER-184 (PU-MET-MIXED-002 [2]) spanned the entire fission energy spectrum. Case 3, which had a median fission energy (MFE) of approximately 6E-5 MeV and had a fission fraction of about 42% in the intermediate energy range, resulted in a $k_{eff}$ overestimation of 1.1%. Compared to 749 previous ICSBEP plutonium benchmarks, the baseline cases accurately predicted the experiments in the thermal and fast regions where the majority of benchmarks inhabit. The benchmarks in the intermediate energy region to date are sparse and overestimate $k_{eff}$ with an average C/E between 1.02 and 1.03. The additional proposed configurations span the whole of the intermediate energy region. The experimental design utilizes the plutonium/aluminum metal alloy Zero Power Physics Reactor (ZPPR) Plutonium-Aluminum No-Nickel (PANN) plates with varying polyethylene moderator thicknesses to span the intermediate fission energy region. Each of the cases have varying fractions of thermal, intermediate, and fast fissions. The designs were chosen to maximize the intermediate energy fraction. The experiment will take place on the universal critical assembly machine, Planet. The layers will be split as equally as possible between the lower platen and the upper stationary platform of Planet. The upper half of the experimental configuration will also have an upper reflector of polyethylene of specified thicknesses to achieve criticality when the lower platen is raised fully. The previous IER-184 configurations, specifically Case 3, were used to determine the configurations for the additional experiments and neutronics calculations were used to fine-tune the configurations to ensure criticality. The quadrature sum uncertainty in Δ$k_{eff}$ for Case 3 in PU-MET-MIXED-002 was found to be 0.00219. Section 3.8 gives a detailed description of the uncertainties calculated. The additional configurations, which are based directly on Case 3, are expected to have similar uncertainties. However, it is possible to reduce the overall uncertainty of Δ$k_{eff}$ for the additional configurations using the knowledge obtained from the calculations in the benchmark.
A workable design to perform a CAAS benchmark experiment is detailed herein. Key dimensions, materials, source intensity levels, and detectors are listed in this report. Sensitivity to 21 perturbations was determined to be acceptable. The next step will be for NCSP management to determine whether procurement should occur and if the experiment should proceed. The perturbation study suggests that the room return shield cavity radius and runout should be maintained to within a millimeter, the room return shield should be positioned carefully (perhaps with a laser range finder), and that the detectors should be mounted in a lightweight fixture such as aluminum, so their positioning is assured. Using a 3D scanner or photogrammetry to record part shapes may be beneficial. Further work is also needed to verify source reproducibility.