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DOE OSTI2015

Calculating uncertainty on K-effective with MONK10 - 14583

Criticality safety assessments require a demonstration that a particular configuration of fissile material has an adequate sub-critical margin (k-effective sufficiently below unity) to ensure that the risk of criticality under normal operation and accident conditions is acceptable. The required sub-critical margin depends upon the uncertainty in the estimated value of k-effective. The uncertainty in the calculated value of k-effective arises from a number of sources, including: manufacturing tolerances on input data to the code (affecting geometry, compositions and densities), uncertainty in the nuclear data used by the code, stochastic uncertainty resulting from Monte Carlo simulation and modelling approximations/errors, including the inevitable bugs in the software. The ANSWERS Software Service, in collaboration with industrial partners, is developing a number of techniques to better understand and quantify uncertainty on predicted values of k-effective, using MONK. The SPRUCE utility code has been developed to allow uncertainty to be estimated using sampling methods. This can include the sampling of input parameters (including dimensions, compositions and densities) from statistical distributions. It can also include sampling different nuclear data libraries. A set of nuclear data libraries has been generated for this purpose by sampling from statistical distributions that represent the uncertainties in the published nuclear data evaluated files; a set of libraries has been produced for Latin Hypercube Sampling. By varying the input data and nuclear data, separate and combined uncertainties due to manufacturing tolerances and nuclear data can be derived. By performing least squares fitting on the results it is also possible to estimate the contribution of each of the uncertain inputs and a sensitivity method in MONK can break down the nuclear data uncertainty. (authors)

42 ENGINEERING↗

DOE OSTI2019

THE INFLUENCE OF CHANGES IN NUCLEAR COVARIANCE DATA ON THE CALCULATION OF Ck FOR HIGHLY ENRICHED URANIUM SOLUTION SYSTEMS

Since their initial development nearly 20 years ago, sensitivity/uncertainty (S/U) techniques have been increasingly applied to criticality safety validation. These techniques can generally be applied quickly and for a range of purposes after the initial investment of calculating sensitivities for a safety analysis model and a range of potential benchmark experiments. This paper discusses these applications and how they have been affected by changes in covariance data libraries.Critical experiment selection is the most prominent area in which S/U methods impact validation efforts. These methods allow for rigorous calculation of similarity between safety analysis models and potential benchmarks. The most widely used of these methods is the integral parameter ck, although other integral parameters such as E and some non-integral parameters have also been explored. The ck parameter calculates the fraction of nuclear data–induced uncertainty shared between an application and an experiment; this metric is predicated on the assessment that nuclear data errors are the most likely source of computational bias. Therefore, values of ck are inherently dependent on the nuclear covariance library used to calculate them and inevitably change when these data change.Past studies have shown how much the ck values can vary, especially in difficult validation cases like burnup credit. These significant changes are the result of differences in the relative uncertainties of nuclides with large sensitivities, and they imply differences in benchmark applicability to the safety analysis model. This paper demonstrates that more common criticality safety models involving single fissile species have significantly less variability of ck between covariance libraries, thus enhancing confidence in the application of S/U methods for experiment selection in these scenarios. This paper presents case studies examining the impact of these changes on some hypothetical safety analysis systems.

Clarity, Justin↗

DOE OSTI2015

Using Nuclear Theory, Data and Uncertainties in Monte Carlo Transport Applications

These are slides for a presentation on using nuclear theory, data and uncertainties in Monte Carlo transport applications. The following topics are covered: nuclear data (experimental data versus theoretical models, data evaluation and uncertainty quantification), fission multiplicity models (fixed source applications, criticality calculations), uncertainties and their impact (integral quantities, sensitivity analysis, uncertainty propagation).

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2012

Quarterly Technical Report, Fission Time Projection Chamber Project

Reactors, weapons and nucelosynthesis calculations are all dependent on nuclear physics for cross sections and particle kinematics. These applications are very sensitive to the nuclear physics in the fast neutron energy region and therefore have large overlaps in nuclear data needs. High performance computer codes interface the nuclear data through nuclear data libraries, which are a culmination of experimental results and nuclear theory and modeling. Uncertainties in the data contained in those libraries propagate into uncertainties in calculated performance parameters.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2020

ENDF/B-VIII.0 Augmented Covariance Data The first iteration [Slides]

Nuclear data are necessary for reliable modeling and simulation of the next generation of nuclear reactors. However, the variation in the ratio of the computed to experimental values (C/E) for certain types of nuclear systems is much less than predicted by evaluated nuclear data file (ENDF)/B covariances. Figure 1 provides an example of a set of metal-plutonium-fueled, fast-spectrum (PU-MET-FAST) integral experiments from the International Criticality Safety Benchmark Evaluation Project (ICSBEP) in the Oak Ridge National Laboratory (ORNL) VALID database. The variation in the C/E values, shown with one standard deviation error bars, is 100% covered by both the SCALE and ENDB/VIII.0 covariance data. This is due to the comparisons to integral data that are essential during the evaluation process. However, the ENDF evaluations represent uncertainties and correlations in differential data only; they do not reflect the impact of the comparison to integral data in the covariance evaluations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2019

ENDF/B-VIII.0 Augmented Covariance Data The first iteration [Slides]

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2023

Preliminary verification of the MCNP perturbation and fixed-source tally sensitivity tools

Integral benchmark experiments are vital in the adjustment and validation of the nuclear data that govern predictive simulations across the nuclear community. The nuclear data sensitivity capabilities of the Monte Carlo N-Particle (MCNP ®) transport code are currently limited; however, expanding sensitivity capabilities will allow benchmark experiments to be designed to resolve compensating errors and adjust nuclear data where previously prohibitively difficult. This paper provides details of a preliminary verification for the use of (i.) the recently revised perturbation and (ii.) developmental fixed-source sensitivity tools within MCNP to calculate sensitivities of tallied responses (such as current integrated over a surface, F1, and flux averaged over a cell, F4) to nuclear data in fixed-source simulations. Energy-binned and energy-integrated sensitivities calculated with these tools are compared against sensitivities calculated using a central-difference approximation. Here, the verification is completed for four configurations of a benchmarked system using a 4.5-kg plutonium sphere surrounded by varying amounts of copper and/or polyethylene. The results show that sensitivities calculated with the perturbation and fixed-source sensitivity tools agree with the central-difference-based approach.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2022

Preliminary Design of a New Pulsed-Neutron Die--Away Experiments with Absorbers (IER-552 CED-1)

This report presents the preliminary design (CED-1) of IER-552, pulsed-neutron die-away (PNDA) experiments with absorbers, to be conducted by Lawrence Livermore National Laboratory (LLNL). The goal of the PNDA experiments with absorbing materials is to produce high-quality, low-cost integral benchmarks that can be used to validate cross sections that are high priority nuclear data for the Department of Energy’s Nuclear Criticality Safety Program (NCSP). PNDA experiments can be used to generate nuclear data validation benchmarks quickly and cheaply, as they do not require fissile material or a nuclear facility to conduct them. The original PNDA experiments were set up at LLNL under IER-501. Extending the PNDA experiments to include absorbing materials leverages existing facilities and resources at LLNL to validate cross sections, with a basic experiment design that uses a neutron generator to impinge a short, mono-energetic neutron pulse on a poisoned, moderating target sample. The neutron pulse lasts 100 µs, after which the neutron population in the sample reaches thermal equilibrium and later spatial equilibrium. In this fully equilibrized state, the neutron population has a characteristic decay-time eigenvalue. The eigenvalue can be used as an integral parameter to validate nuclear data involved with neutron scattering and absorption. Two observables can be measured: the neutron population and the population of secondary γ-rays produced by inelastic and capture reactions. This report explores the feasibility of both approaches. For γ-ray measurements, this allows for direct measurement of reactions of interest for validation. During the neutron burst time, inelastic (prompt) γ-rays dominate, and these data can be isolated for validation of neutron inelastic scattering data. After the pulse ends, neutron capture (delayed) γ-rays dominate and are produced by many reaction channels as neutrons slow down to thermal energies. Two sets of γ-ray spectra (inelastic, capture) can be produced for validation. Additionally, the time-dependent decay of the capture γ-ray population when the neutron population is thermalized and in its fundamental mode can be used for validation. Two experiments can be imagined where pulsed neutrons are used to validate the nuclear data of absorbing materials. The first integrates well into the existing PNDA testbed at LLNL. The absorber material would be introduced into a moderating target and the same neutron generator, neutron detectors, electronics, and shielding box used for IER-501 would be employed. From the same experimental setup, many absorbing materials can be validated by simply changing the neutron poison in the moderator. The only significant modification would be the introduction of a γ-ray detector. Such an experiment would primarily serve to validate nuclear data in the thermal energy range. The second kind of experiment would validate the fast and epithermal energy range. It would involve pulsing neutron into a bare assembly of the absorber material and measuring the die-away of neutrons. The fast leakage of such an assembly requires nanosecond neutron-generator pulse widths and fast detectors/electronics that could measure ~200 ns die-away curves. This kind of experiment would ideally be performed at a cyclotron, like that available at Lawrence Berkeley National Laboratory.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2021

NEFCD All-Hands Meeting – Overview of Nuclear Criticality, Radiation Transport, and Safety (NCRTS) Section [Slides]

Topics for the meeting were as follows: Safety Minute, High-level Overview of NCRTS, Overview of NCRTS Groups (Radiation Transport, Nuclear Data, and Nuclear Criticality Safety), and Question and Answer.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2023

Release of ENDF/B-VIII.1β1-Based ACE Data Files

On March 1 st , 2023, the National Nuclear Data Center (NNDC) released the ENDF/B-VIII.1β1 nuclear data library, which contained neutron, spontaneous and neutron-induced fission yield, and alpha sublibraries. The library was released in the standard Evaluated Nuclear Data File (ENDF) format. The files can be accessed on Brookhaven National Laboratory’s GitLab. The thermal scattering law (TSL) sublibrary will be released at a later date. The files provided in the neutron sublibrary were processed into A Compact ENDF (ACE)-formatted files, verified, and validated by the XCP-5 Nuclear Data Team. The resulting collection of ACE files will be referred to as the “e81b1” library. This report details the processing of these files and the quality assurance approach taken. This is not intended to be a full validation effort; rather, this library is intended to simply reproduce the released files for further validation testing by the community. The e81b1 ACE library will be released to the NNDC for access through GitLab and hosted locally on LANL computing resources. This library is only intended for analysis to support the future ENDF/B-VIII.1 release and should not be used for production results.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2015

Preface by the CW2014 Organizers-Including Program, Advisory Board, Participants and Photo

This issue of the Nuclear Data Sheets contains the proceedings of the 'International Workshop on Nuclear Data Covariances'. This workshop was the third one in a series that started with the 'Workshop on Neutron Cross Section Covariances' (Port Je erson, USA, 2008) and continued with the 'Second Workshop on Neutron Cross Section Covariances' (Vienna, Austria, 2011). The current workshop returned to the US and took place in the center of the beautiful and historic city of Santa Fe, New Mexico, USA from April 28 to May 1, 2014. The purpose of this workshop was to bring together scientists in the field of nuclear data evaluation, nuclear reaction theory, reactor physics and associated experiments to review recent developments in nuclear data evaluation methodology as well as assess and discuss open questions regarding uncertainty estimates and associated formatting requirements from the point of view of the experimentalist, the theoretician, the evaluator as well as from application side, e.g. in transport calculations. The workshop was open to contributions on a wide variety of nuclear data observables (cross sections, fission yields, energy and angle spectra, etc.), from the resonance range up to the high energy range as well as for light to heavy elements.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2015

NJOY21: Making NJOY Suitable for the 21. Century - 14409

NJOY has a long history of nuclear data processing. NJOY began as a successor to MINX (Multigroup Interpretation of Nuclear X-Sections), a Los Alamos code and was first publicly released in the summer of 1977. The initial version included modules such as RECONR for resonance reconstruction on a union energy grid, BROADR for Doppler broadening plus a new Multigroup averaging program, GROUPR and an associated output module, DTR. Soon thereafter the HEATR, ERRORR, and THERMR modules were added to NJOY. By the mid 1980's, NJOY had a worldwide user community. The current version of NJOY2012 was released in December 2012. NJOY2012 builds on the success of more than 45 years of nuclear data processing at Los Alamos National Laboratory. NJOY2012 was rewritten using Fortran 90/95, utilizing among other things the concepts of public and private variables, eliminating common blocks in lieu of Fortran modules and taking advantage of allocatable arrays. A new version of NJOY is being developed at Los Alamos National Laboratory; NJOY21: NJOY for the 21. Century. NJOY21 uses modern programming standards of quality, reliability, and maintainability. It is being designed from the ground up to utilize modern methods of interacting with computer codes by providing an interactive and scripting (e.g., Python) interface. This will allow scientists the ability to more easily work with the results of the various steps in the processing of nuclear data. NJOY21 will continue the legacy of NJOY by providing a trusted and verified processing code that is fast, flexible, easy to use and customize and capable of handling both modern and established nuclear data formats. This paper introduces NJOY21 by first relating the history of NJOY; the majority of the paper will discuss NJOY21. NJOY21 will take advantage of the many decades of nuclear data processing experience that is contained in NJOY2012 and previous versions of NJOY and extend the capabilities and convenience of processing nuclear data. We will discuss how NJOY21 will remain backwards compatible with NJOY2012 - i.e., how current NJOY2012 input decks will continue to work with NJOY21. We will also discuss the verification and validation of the NJOY21 code base. Finally we will discuss new features in NJOY21 that do not exist in NJOY2012. These features include changes in how a scientist can interact with NJOY to process and verify and validate nuclear data; and the ability to read and write new nuclear data formats. (authors)

42 ENGINEERING↗

DOE OSTI2023

Templates of expected measurement uncertainties

The covariance committee of CSEWG (Cross Section Evaluation Working Group) established templates of expected measurement uncertainties for neutron-induced total, (n,γ), neutron-induced charged-particle, and (n,xn) reaction cross sections as well as prompt fission neutron spectra, average prompt and total fission neutron multiplicities, and fission yields. Templates provide a list of what uncertainty sources are expected for each measurement type and observable, and suggest typical ranges of these uncertainties and correlations based on a survey of experimental data, associated literature, and feedback from experimenters. Information needed to faithfully include the experimental data in the nuclear-data evaluation process is also provided. These templates could assist (a) experimenters and EXFOR compilers in delivering more complete uncertainties and measurement information relevant for evaluations of new experimental data, and (b) evaluators in achieving a more comprehensive uncertainty quantification for evaluation purposes. This effort might ultimately lead to more realistic evaluated covariances for nuclear-data applications. In this topical issue, we cover the templates coming out of this CSEWG effort–typically, one observable per paper. This paper here prefaces this topical issue by introducing the concept and mathematical framework of templates, discussing potential use cases, and giving an example of how they can be applied (estimating missing experimental uncertainties of 235 U(n,f) average prompt fission neutron multiplicities), and their impact on nuclear-data evaluations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2016

Improving Capture-gamma Libraries for Nonproliferation Applications

This report describes the measurement, evaluation and incorporation of new -ray spectroscopic data into the Evaluated Nuclear Data File (ENDF) for nonproliferation applications. Analysis and processing techniques are described along with key deliverables that have been met over the course of this project. A total of nine new ENDF libraries have been submitted to the National Nuclear Data Center at the Brookhaven National Laboratory and are now available in the ENDF/B-VIII.beta2 release. Furthermore, this project has led to more than ten peer-reviewed publications and provided theses for ve graduate students. This project is a component of the NA-22 venture collaboration on \Correlated Nuclear Data in Fission Events" (LA14-V-CorrData-PD2Jb).

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2022

Measurement of the U 235 ( n , f ) prompt fission neutron spectrum from 10 keV to 10 MeV induced by neutrons of energy from 1 MeV to 20 MeV

The characterization of fission-driven nuclear systems primarily relies on calculations of neutron-induced chain reactions, and these calculations require evaluated nuclear data as input. Calculation accuracy heavily depends on input nuclear data evaluation accuracy, and thus high precision on the experimental input to the nuclear data evaluation is essential for fundamental quantities like the energy spectrum of neutrons emitted from neutron-induced fission (i.e., the prompt fission neutron spectrum, PFNS). Despite decades of measurement efforts, prior to the measurements described in this work there were only three literature data sets for the 235 U(n,f) PFNS at incident neutron energies above 1.0 MeV considered reliable for inclusion in nuclear data evaluations and no reliable data sets above 3.0 MeV incident neutron energy. In this work we report on new measurements of the 235 U(n,f) PFNS spanning a grid of 1.0–20.0 MeV in incident neutron energy and 0.01–10.0 MeV in outgoing (PFNS) neutron energy. These measurements were carried out at the Weapons Neutron Research facility at the Los Alamos Neutron Science Center and used a multifoil parallel-plate avalanche counter target with both a Li-glass and a liquid scintillator detector array in separate experiments to span the quoted outgoing neutron energy ranges. The PFNS results are shown in terms of the energy spectra themselves as well as the average PFNS energy $(\langle{E}\rangle)$ and ratios of $\langle{E}\rangle$ at forward and backward angles. Here, the results are compared with literature data and selected nuclear data evaluations. Generally, the data agree with the ENDF/B-VIII.0 evaluation below 5.0-MeV incident neutron energy and more closely with the JEFF-3.3 evaluation above 5.0 MeV, though no evaluations considered for comparison in this work agree with the data across all of the incident and outgoing neutron energies shown, especially in regions where the third-chance fission process becomes available. Additionally, we show a ratio of the present PFNS results for 235 U(n, f) with a recent and highly correlated experiment to measure the 239 Pu(n, f) PFNS at the same experimental facility and with nearly identical equipment and analysis procedures. Many observations reported in this work are the first of their kind and represent significant advancements for knowledge of the 235 U(n, f) PFNS.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2018

Optimization study of B4C filter thickness for shifting to fast spectrum at experimental position in NRAD core

The Japan Atomic Energy Agency (JAEA) has investigated an accelerator-driven system (ADS) to transmute minor actinides (MAs) partitioned from the high level waste. In the nuclear transmutation system such as the ADS, the nuclear data validation of MA is required to reduce the uncertainty caused by the nuclear data of MA. In order to validate the nuclear data, many independent experimental data need to be mutually compared. The Idaho National Laboratory (INL) considers the measurement of fission rate ratio and sample worth with a wide spectral indices by inserting several types of thermal neutron shielding filters in the experimental position in Neutron RADiography (NRAD) which is the thermal neutron reactor. The JAEA has a plan to perform integral experiments with fission chambers in the NRAD because of the expansion of the experimental data contributing to nuclear data validation of MA. In this study, the optimization of the NRAD core configuration, the B4C filter thickness, and the experimental position design was conducted to acquire the high-accuracy and the beneficial experimental data by using MCNP 6.1 and ENDF/B-VII.1. First, by locating the experimental position as close to the center of the core as possible and concentrating the fuel around that position, the neutron flux of the optimized model could be constructed 2.5 times as much as that of the previous model. Second, the optimized B4C filter thickness was determined to be 3mm based on the comprehensive evaluation from the viewpoints of the shielding ability of low energy neutrons, the level of the total neutron flux, and the influence of fabrication accuracy of the B4C filter. Third, we constructed the experimental position model that could further reduce the influence of the low energy neutrons by adopting the aluminum dry box and the bottom filter. Moreover, the neutron energy spectrum of the tentative optimized model constructed by following the above results was compared with that of the fast neutron reactors such as the ADS and the Fast Critical Assembly (FCA). As a result, the tentative optimized model would be able to be obtained the beneficial experimental data for the verification of FCA-IX series data and the validation of MA nuclear data for accuracy improvement of ADS design. Furthermore, the adoption of the tentative optimized model would be acquired the experimental data such as fission reaction rate and sample worth with the wider spectral indices than the previous one because the tentative optimized model would be able to increase the ratio of the fast neutrons by almost completely eliminating the effect of the thermal neutrons.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

A Primer on Nuclear "Recoil" Data

For many years the ENDF-6 format has existed to contain evaluated nuclear data. One primary driver for the format and the data is neutron transport calculations. Evaluated data can be processed by a code like NJOY into either continuous energy form (ACE format) for Monte Carlo codes like MCNP or into multi-group form (NDI tables) for deterministic codes like Partisn. The reaction cross sections are found in the MF 3 section of the ENDF-6 format. If the reaction produces one or more neutrons as outputs, then secondary neutron data must also be given in MF 4,5 or 6 sections of the format. Energy and angular distributions for the output neutrons must be given in one of several available formats. The most general formats are found in MF 6. MF 4 is for angular distributions, MF 5 is for Energy distributions, and MF 6 contains both. In recent years, more interest has developed in the other output particles (i.e., the “recoil” particles) from neutron induced reactions. This has been driven by interest in charged particle transport and in more specialized partial kermas. ( A separate total neutron kerma has been available for a long time.) Partial kermas are a breakdown of the Kinetic Energy Released into the MAterial by output particle or by neutron reaction. The sum of the partial kermas should be equal to the total kerma on a group-wise basis. The ENDF-6 format is general enough for these new data requirements. Ideally, evaluated data would exist for every output particle (including the secondary neutrons) from every neutron-induced reaction. In the MF 6 format section, data for multi-particle outputs may be entered using the LAW =1 option. This option explicitly allows energy and angular output distributions for each particle produced in the reaction. Such information should preserve the balance between partial kermas and the total kerma at the groupwise level as well as the individual particle averaged energies. For simpler 2-body reactions, LAW = 2 is available. This allows the evaluation to specify only enough data to specify the 2-body reaction fully. Such a simplification does not exist for 3 (or more) body breakups. A simplified form of LAW 2, i.e., LAW 3, also exists to generate approximate recoil output distributions in the absence of full data. However, evaluated data does not generally exist at this fine granularity for reactions involving more than 2 output particles ( e.g., the 3-body break-up reaction). When the multi-body detailed output distribution data is not available, LAW=6 may be employed in NJOY. LAW 6 produces approximate output distributions for all the particles produced in the reaction. Just like in the 2-body case, the smaller particles will generally carry off more energy. At any particular energy, the balance between the sum of the partial kermas and the total kerma will not necessarily be preserved. However, the average energy of each output particle from a reaction is preserved.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2018

Release of ENDF/B-VIII.0-Based ACE Data Files

In February 2018, the National Nuclear Data Center released ENDF/B-VIII.0 in the standard Evaluated Nuclear Data Format (ENDF). This represents the advances made in nuclear data since the release of ENDF/B-VII.1 in 2011.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗