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DOE OSTI2021

An uncertainty quantification method relevant to material test reactors

Within material test reactor calculations, energy dependent flux and reaction rate uncertainties are typically not quantified when performing as-run analyses to determine the neutron field experienced by the experiment. When high fidelity Monte-Carlo codes are used in such analyses, straight forward methods to calculate output uncertainties are not available, instead expert opinion is used to postulate computational uncertainties. New methods to propagate uncertainties through these high fidelity simulations are available when sufficient computational power is available. A tool is developed here for sampling any part of an MCNP input from random distributions to determine output uncertainties based on those inputs. Another tool is developed to sample nuclear data cross-section in ACE format using multi-group nuclear data covariances. The Total Monte-Carlo Method and Gesellschaft für Anlagen-und Reaktorsicherheit method (GRS) are implemented and compared to one another as well as MCNP sensitivity and uncertainty calculations. The methods were applied to the Godiva critical sphere k-eigenvalue, the UAM pin-cell benchmark energy dependent flux and reaction rates, and the Advanced Test Reactor energy dependent flux within an experimental location. Furthermore, the two methods agree well, with GRS allowing for an order of magnitude speedup for reaction rate uncertainty calculations and several orders of magnitude for eigenvalue uncertainty calculations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2020

The ENDF/B-VIII.0 Compendium Volume I: Neutron Cross Section Plots

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2020

The ENDF/B-VIII.0 Compendium Volume IV: Radioactive Decay Rank-Order Analyses

Version 8 of the Evaluated Nuclear Data File (ENDF/B-VIII.0) contains a wealth of nuclear data representing the most recent collection of evaluated neutron cross section measurements, isotopic radioactive decay, and much more. This Compendium has been prepared as a easily-navigable reference of the contents of the ENDF/B-VIII.0 library, while making no statement regarding the accuracy of those contents. The Compendium is split into four volumes. In Volume IV, radioactive decay properties are presented in tabular form, sorted by isotope, energy, element-then- energy, and probability of emission.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2020

The ENDF/B-VIII.0 Compendium Volume II: Neutron Cross Section Rank-Order Analyses

Version 8 of the Evaluated Nuclear Data File (ENDF/B-VIII.0) contains a wealth of nuclear data representing the most recent collection of evaluated neutron cross section measurements, isotopic radioactive decay, and much more. This Compendium has been prepared as a easily-navigable reference of the contents of the ENDF/B-VIII.0 library, while making no statement regarding the accuracy of those contents. The Compendium is split into four volumes. Volume II presents a series of tables of reaction cross section values for thermal (0.025 eV), fission (1 MeV), and fusion (14 MeV) energies, each sorted by cross section value, isotope, and element-then-isotope to facilitate quick searches.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2016

Preliminary Design of the TEX-MOX Criticality Experimental Program using Optimization Algorithms Implemented in the IRSN PROMETHEE Tool

In the frame of its missions, IRSN is widely involved in design, evaluation and analyses of critical experiments in order to contribute to computer codes and nuclear data validation. The work described in this paper was carried out in the framework of the PRINCESS IRSN research project and is part of the IRSN/LLNL collaboration on the TEX program in the frame of the NCSP. The need for epithermal and intermediate energy range critical benchmarks is an established international criticality safety data need, underpinned by the subsequent use of data assimilation methods like GLLSM to reduce Nuclear Data uncertainties. The goal of the TEX program is to address this need by executing experiments with NCSP fissile assets, which can be used to create critical assemblies that span a wide range of fission energy spectra, from the thermal energy range (below 0.625 eV), through the intermediate energy range (0.625 eV to 100 keV), to the fast one (above 100 keV). The TEX program involves different materials of interest, such as TEX-Ta experiments using plutonium ZPPR plates proposed by LLNL. This paper presents an overview of the studies performed at IRSN using numerical optimization algorithms for the design of critical experiments involving Mixed U and Pu fuel plates, referred to as TEX-MOX program. This program ultimately aims to perform critical experiments to validate the MOX fuel configurations encountered in the MOX fuel fabrication process. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Development of an In-Beam Neutron Flux Monitor for Time of Flight Capture Measurements

Accurate nuclear data are important in several aspects of nuclear engineering, such as reactor simulations, nonproliferation, and criticality safety. One of the nuclear data measurements that requires reduced uncertainties in its related cross-sections is neutron capture. The neutron capture cross-sections have particular deficiencies in the incident neutron energy range from 1 keV to 1 MeV. This is due to the predominance of scattering reactions in this energy region. In order to provide more accurate measurements in this energy region, a new experimental setup has been designed and created at Rensselaer Polytechnic Institute (RPI). This experiment utilizes several C{sub 6}D{sub 6} detectors in a low mass system to minimize the effects of neutron scattering on the experiment. In order to determine the capture yield for the measurements, several additional factors need to be known about the neutron beam including the energy dependence of the neutron flux incident on the sample. A detector made from a thin piece of EJ-204 plastic scintillator was designed to measure the neutron flux, as a function of incident neutron energy, during the capture measurements. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2016

Analysis Capability and Data Needs Identified During the Evaluation of the SILENE CAAS Benchmarks

In October 2010, a series of benchmark experiments was conducted at the Commissariat a l'Energie Atomique et aux Energies Alternatives (CEA) Valduc SILENE facility. These experiments were a joint effort between the US Department of Energy (DOE) and the French CEA to create three benchmarks for verification and validation of radiation transport codes, as well as evaluated nuclear data, used to analyze criticality accident alarm systems (CAASs). Many preliminary details of these experiments and their evaluation have been discussed in previous publications. The first benchmark evaluation was published in 2015, and the second and third evaluations will be published at the end of 2016. During evaluation of the three benchmarks, differences were observed between the measured and calculated benchmark values, which was not surprising. Evaluators spent a great deal of time identifying the sources of these differences. This paper presents two of the most important and clearly identifiable differences, which the authors identify as needs for the criticality safety and shielding communities. The first need is the capability to easily produce and transport delayed neutrons and gammas within existing radiation transport codes without coupling to a separate decay and depletion code. The second need is to fill in missing gamma production data in the evaluated nuclear data libraries. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2015

Towards application of a neutron cross-section uncertainty propagation capability in the criticality safety methodology - 14434

At the Paul Scherrer Institute (PSI) the methodology for criticality safety evaluations (CSE) used for Swiss applications is focused, following common practice, on establishing an upper subcriticality limit (USL) for the effective neutron multiplication factor (k{sub eff}), based on a statistical evaluation of the obtained validation results in terms of the calculated-to-benchmark k{sub eff} values. Recently, a new computational tool referred to as NUSS (Nuclear data Uncertainty Stochastic Sampling) was developed in order to propagate neutron-nuclear data (ND) uncertainties in calculations with Monte Carlo transport codes in combination with general-purpose ND libraries in ACE format. Thereby, this tool should now allow for a quantification of ND data uncertainties in CSE and to introduce these into the CSE criteria definition. This paper concerns an application of the NUSS methodology for criticality validation studies. Calculated k{sub eff} uncertainty estimations were obtained for the current PSI validation suite including in total 149 critical benchmark configurations for which in-house MCNP(X) models were developed, covering both uranium and mixed uranium-plutonium oxide fuels. The MCNPX-2.7.0 code was used together with the ENDF/B-VII.1 ND library while the ND variance-covariance matrixes (VCM) were taken from both SCALE-6.0 44-group libraries and also from ENDF/B-VII.1 files. Based on the newly obtained information on the uncertainties, a re-assessment of the lower tolerance bound (LTB) for k{sub eff} using the total uncertainties as weighting factors is proposed as means for the present PSI CSE methodology upgrade and some possible directions of further methodology enhancements are discussed next. (authors)

42 ENGINEERING↗

DOE OSTI2020

Measurement of the 239 Pu$(n, f$) prompt fission neutron spectrum from 10 keV to 10 MeV induced by neutrons of energy 1–20 MeV

Although the prompt fission neutron spectrum (PFNS) is an essential component of neutron-driven systems that has been measured for decades, there are still multiple glaring unknowns regarding the PFNS of major actinides in the fission neutron incident energy range, specifically with regard to multichance fission and pre-equilibrium neutron emission processes. The only impactful experimental 239 Pu PFNS measurements included in recent nuclear data evaluations were measured over a limited outgoing neutron energy range at thermal and 1.5-MeV average incident neutron energy, while other potentially impactful measurements have been shown to contain errors that resulted in either large uncertainty increases or in complete exclusion from nuclear data evaluation. Here, we report a measurement of the 239 Pu PFNS over a wide range of incident neutron energy (1–20 MeV) and three orders of magnitude in outgoing neutron energy (0.01–10 MeV) resulting from the Chi-Nu experiment at the Los Alamos Neutron Science Center. These results are the combination of separate PFNS measurements in the same experimental area, one using a Li-glass and the other a liquid scintillator detector array. Covariances between all PFNS data points from each detector and within each incident energy range were generated between all other data in both detector arrays and within all other incident neutron energy bins, yielding a single covariance matrix for all 1300 PFNS data points reported here. These covariances are based on a thorough assessment of systematic bias and uncertainties associated with the measurement, PFNS extraction technique, combination of data from each detector type, and other aspects of the analysis. The existence of covariances between PFNS data points in different incident neutron energy ranges yielded covariances between average PFNS energy values at each incident energy to be reported here as well, which allowed for firm statements to be made regarding a shape of a purely experimental mean PFNS energy trend for the first time. Although minor PFNS shape differences exist between the results reported here and recent nuclear data evaluations, the ENDF/B-VIII.0 and JEFF-3.3 PFNS evaluations agree reasonably well with the present results from 1-to 10-MeV incident neutron energy, which spans the well-measured 1.5-MeV incident neutron energy PFNS from Lestone and Shores as well as the onset of second-chance fission. However, while the pre-equilibrium component of the PFNS above 12-MeV incident neutron energy roughly agrees in position and magnitude with ENDF/B-VIII.0 and JEFF-3.3, clear differences relating to the relative magnitude of third-chance fission PFNS features are present in the PFNS shape and in the mean PFNS energy trends.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2020

Selected Uses of TSUNAMI in Critical Experiment Design and Analysis

Validation in criticality safety is performed by comparing the results of critical experiments with the calculated results from models of the experiments using the computational method to be validated. Laboratory critical experiments are controlled systems that achieve a k eff of approximately 1 and enable investigation of the parameters at which such a critical condition is achieved. For the critical experiments used in a validation to capture the biases of the materials and neutron energy spectra of interest, those materials must be included in the experiment such that they influence k eff or another observable parameter with statistical significance. This paper discusses the use of sensitivity uncertainty (S/U) methods to develop critical experiments for various purposes. S/U techniques are useful for understanding the underlying components of nuclear data which affect the k eff or another parameter of a given configuration. S/U calculations are most commonly used to compare existing experiments to applications of interest; however, S/U techniques can also be used to identify, optimize, or assess features of proposed experiments so that they can better test specific portions of nuclear data or match an application of interest. The S/U techniques discussed here are from the TSUNAMI code system. The two primary codes discussed in this work are TSUNAMI-3D, which implements the KENO criticality code to calculate the sensitivity of k eff to nuclear data, and TSAR, which calculates the sensitivity of a reactivity difference between two configurations based on their TSUNAMI-3D generated sensitivity profiles. The methods used in these tools are discussed in more detail in the SCALE manual. This paper is one of a series on the development and use of TSUNAMI tools. The other papers address development of TSUNAMI methods and a review of TSUNAMI applications.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2022

Impact of heavy quark and quarkonium data on nuclear gluon PDFs

A clear understanding of nuclear parton distribution functions (nPDFs) plays a crucial role in the interpretation of collider data taken at the Relativistic Heavy Ion Collider, the Large Hadron Collider (LHC), and in the near future at the Electron-Ion Collider. Even with the recent inclusions of vector boson and light meson production data, the uncertainty of the gluon PDF remains substantial and limits the interpretation of heavy ion collision data. To obtain new constraints on the nuclear gluon PDF, we extend our recent nCTEQ15WZ+SIH analysis to inclusive quarkonium and open heavy-flavor meson production data from the LHC. This vast new data set covers a wide kinematic range and puts strong constraints on the nuclear gluon PDF down to x ≲ 10 -5 . The theoretical predictions for these data sets are obtained from a data-driven approach, where proton-proton data are used to determine effective scattering matrix elements. This approach is validated with detailed comparisons to existing next-to-leading order calculations in nonrelativistic QCD for quarkonia and in the general-mass variable-flavor-number scheme for the open heavy-flavored mesons. In addition, the uncertainties from the data-driven approach are determined using the Hessian method and accounted for in the PDF fits. This extension of our previous analyses represents an important step toward the next generation of PDFs not only by including new data sets, but also by exploring new methods for future analyses.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

DOE OSTI2022

Impact of W and Z Production Data and Compatibility of Neutrino DIS Data in Nuclear Parton Distribution Functions

Vector boson production and neutrino deep-inelastic scattering (DIS) data are crucial for constraining the strange quark parton distribution function (PDF) and more generally for flavor decomposition in PDF extractions. We extend the nCTEQ15 nuclear PDFs (nPDFs) by adding the recent W and Z production data from the LHC in a global nPDF fit. The new nPDF set, referred to as nCTEQ15WZ, is used as a starting point for a follow-up study in which we assess the compatibility of neutrino DIS data with charged lepton DIS data. Specifically, we re-analyze neutrino DIS data from NuTeV, Chorus, and CDHSW, as well as dimuon data from CCFR and NuTeV. To scrutinize the level of compatibility, different kinematic regions of the neutrino data are investigated. Fits to the neutrino data alone and a preliminary global fit are performed and compared to nCTEQ15WZ.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

DOE OSTI2023

ENDF/B-VIII.1 and AMPX [Slides]

This presentation is on ENDF/B-VIII.1 and AMPX. This presentation begins with a brief introduction to the AMPX nuclear data processing code. The lecture then touches on nuclear data libraries that ship with the SCALE code system. Finally, this lecture concludes on preparations for the ENDF/B-VIII.1 libraries.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2018

Sensitivity/uncertainty analyses comparing LR-0 reactor experiments containing FLiBe salt with models for molten-salt-cooled and molten-salt-fueled reactors

Here, critical experiments using an insertion zone with FLiBe salt performed at the LR-0 reactor at Research Centre Rež (RC Rež) have been compared to application models for solid-fueled, fluoride-salt-cooled high-temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts using sensitivity and uncertainty (S/U) analysis techniques. These experiments support FHR and MSR advanced reactor concepts by informing on neutron spectral effects and nuclear data uncertainties related to fluoride salts. The FLiBe salt in the LR-0 experiments is from the Oak Ridge National Laboratory Molten Salt Reactor Experiment and is enriched to greater than 99.99% 7 Li. This work is part of a broader collaboration between the United States and the Czech Republic on MSR and FHR technology. Results from the S/U analyses for FHR and MSR models indicated the most significant potential source of eigenvalue bias due to nuclear data within the FLiBe salt is radiative capture in 7 Li. Other smaller but potentially significant contributions come from 19 F, 6 Li, and other 7 Li reactions. Similarity comparisons of the RC Rež LR-0 experiments and FHR and MSR application models indicate that the LR-0 FLiBe experiments could be useful as candidate benchmarks for low-enriched uranium–fueled application models. However, sensitivity differences observed included both spectral effects, due to the LR-0 reactor being moderated by water instead of graphite and fluoride salt, and sensitivity magnitude effects driven by the amount of FLiBe inserted and its location. New experiments with refined designs to increase the volume and importance of salt in the system could provide improved data. Results also demonstrate that salt systems using thorium and/or 233 U fuel will require additional experiments with relevant driver fuels. Increased contributions from the graphite moderator and 19 F and covariance between reactions such as 233 U fission and 233 U radiative capture indicate higher uncertainty contributions from the salt and significant differences in contributions from the fuel species due to underlying uncertainties in their nuclear data.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Generation of Enrichment-Dependent Thermal Neutron Scattering Data

This work details the generation of enrichment-dependent thermal neutron scattering cross sections for several crucial uranium fuel compounds. The evaluations of the thermal scattering law (TSL) and associated cross sections for uranium dioxide (UO 2 ), uranium carbide (UC), and uranium nitride (UN) were performed using standard ab initio lattice dynamics (AILD) methods. The data for uranium metal was produced using a novel hybrid approach of molecular dynamics combined with lattice dynamics methods. 235 U enrichments of 5%, 10% (LEU+), 19.75% (HALEU), 93% (HEU), and 100% were considered, in addition to natural uranium. The enrichment-dependent masses and free atom cross sections were used in the generation of elastic and inelastic thermal neutron scattering cross sections, while the calculation of the phonon density of states (DOS) and resulting TSL considered only the natural isotopic composition of uranium. The use of an identical DOS for all enrichments is expected to have minimal impact on the final data, as the small change in uranium mass should not significantly affect lattice vibrations. The cross sections are shown to exhibit significant dependence on 235 U enrichment. The submission of this data to the National Nuclear Data Center (NNDC) for release in the ENDF/B-VIII.1 database should support the design of advanced reactor concepts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

CONTINUOUS-ENERGY MONTE CARLO METHODS FOR CALCULATING GENERALIZED RESPONSE SENSITIVITIES USING TSUNAMI-3D

This work introduces a new approach for calculating sensitivity coefficients for generalized neutronic responses to nuclear data uncertainties using continuous-energy Monte Carlo methods. The approach presented in this paper, known as the GEAR-MC method, allows for the calculation of generalized sensitivity coefficients for multiple responses in a single Monte Carlo calculation with no nuclear data perturbations or knowledge of nuclear covariance data. The theory behind the GEAR-MC method is presented here, and proof of principle is demonstrated by using the GEAR-MC method to calculate sensitivity coefficients for responses in several 3D, continuous-energy Monte Carlo applications.

CLUTCH↗

DOE OSTI2020

(U) Jezebel Densities and Uncertainties

The 239 Pu Jezebel benchmark is a keystone for plutonium nuclear data evaluation. At a recent presentation at a nuclear data workshop, Morgan White (LANL P-27) presented a slide about the Jezebel plutonium-alloy part densities and their uncertainties. There are two major errors and one opinion presented as fact.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2022

Nuclear Criticality Safety Program (NA-ESH) [Slides]

This presentation provides an overview of the Nuclear Criticality Safety Program, including its history, mission, vision, organization, measurement and evaluation priorities for FY23, 5-year plan, and cross-cutting nuclear data work.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗