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Palmiotti, Giuseppe

Publications and source records attributed to Palmiotti, Giuseppe.

At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Fine Temperature Grid Continuous Energy Cross Section Generation for Monte Carlo Analysis of Xe-100 Design

The standard “A Compact ENDF (ACE)” data libraries used by Monte Carlo based reactor physics codes calculations are provided by Los Alamos National Laboratory (LANL) with a temperature interval mostly of 300 K (e.g. 300 K, 600 K, 900 K) for the cross sections and between 100 K and 200 K for the thermal scattering libraries (TSL). However, some codes such as MCNP lack capability to perform on-the-fly temperature interpolation during simulation both for neutron and TSL cross-sections. To evaluate the impact related to Doppler broadening and spectrum shift associated with TSL changes, this paper explores the potential of adopting a temperature grid finer than the ones contained in the standard data libraries. A 50 K temperature grid was employed to quantify the error in neutronics calculations due to temperature grid resolution. This was achieved by comparing the results of this study (50 K temperature interval) against the results obtained with standard data libraries (>100 K temperature interval). While the adopted grid primarily relies on the ENDF/B-VII.1 library, for neutron cross-sections, it utilizes ENDF/B-VIII.0 library for TSL. The analyses confirmed that the accuracy of neutronics calculations is satisfactory when using a 50 K temperature grid. Notably, adopting a 50 K temperature grid, as opposed to standard libraries or coarser temperature grids, could lead to a difference of no more than a few hundred pcm in dk for both fresh fuel and burnt fuel. The most sensitive reaction type to the temperature grid was as expected identified as the capture cross-section of U-238.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Nuclear data uncertainty propagation applied to the versatile test reactor conceptual design

We report the Versatile Test Reactor (VTR) currently under development is a 300 MWth sodium-cooled fast reactor (SFR) fueled with ternary metal alloy fuel, which aims to accelerate the testing of advanced nuclear fuels, materials, instrumentation, and sensors in high flux environments that are necessary to license the next generation of advanced reactor concepts. To support the VTR design process, uncertainties associated with the nuclear data has been propagated through the reactor core neutronics calculation to global parameters of interest, such as the core multiplication factor, kinetic parameters, and various reactivity feedback coefficients, following the sensitivity based uncertainty propagation approach. By folding the sensitivity coefficients, separately computed by the generalized perturbation theory code PERSENT and Monte Carlo code Serpent 2, with the variance-covariance matrices from COMMARA-2.0, we obtain the reaction-wise, isotope-wise, and overall uncertainties for each response of interest due to nuclear data uncertainty. With Serpent 2, the statistical error of the uncertainty is obtained by propagating the statistical error of the sensitivity coefficients through the same process using a newly developed uncertainty propagation method. From both codes, the overall top uncertainty contributors are found to be the cross section of Fe-56 elastic scattering, Na-23 elastic scattering, and U 238 inelastic scattering. The large contributions of the Fe-56 elastic scattering cross sections to global parameters are due to its relatively large relative uncertainty of 5–10% in nuclear data and the large volume of Fe-containing reflector assemblies in the fairly compact VTR core design. Both codes agreed well for the overall uncertainty estimates of all responses of interest, except the delayed neutron fraction, prompt neutron generation time, and the coolant density feedback coefficient, where Serpent 2 yielded a much larger value than PERSENT due to the large statistical error of sensitivity coefficients. The calculated uncertainties are also compared to those associated with other SFR cores. Another outcome of this study is a variance-covariance matrix of reactivity coefficients, which can be used in the subsequent uncertainty propagation to the system level to investigate the impact of identified uncertainties on system responses in the safety analysis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Assessment of Nuclear Data Needs for Advanced Reactor Demonstrations: Application to the Molten Chloride Reactor Experiment (MCRE)

A number of advanced reactor concepts are planned for near-term demonstrations including microreactors, larger demonstrations and space nuclear systems. These reactor concepts are based on a wide variety of reactor technologies, including sodium, gas, and salt cooling. An overlooked area in the development and ultimate startup of these reactors is addressing nuclear data needs that allow confident prediction of the criticality, safety requirements, and operation of the reactors. In this paper we try to address the problem of assessing nuclear data needs and possible remedies to reduce the existing uncertainties for advanced nuclear reactors. A methodology for defining these needs is described. The case of the Molten Chloride Reactor Experiment (MCRE) has been considered and the related investigation highlights the specific needs for reducing uncertainty on the 235U capture and 35Cl (n,p) reactions. Integral experiments relatively inexpensive are indicated as possible solution for significantly reduce the current associated uncertainties.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Scoping Analysis of Sodium Cooled Fast Spectrum Test Reactor Cores

A team from the national laboratories conducted a study that assessed advanced reactor technology options and the irradiation needs of a broad group of stakeholders. This identified a gap in the fast neutron irradiation capability - the need for a test reactor that can deliver high neutron fluxes of fast neutrons. The study compared fast and thermal neutron irradiation technology options based on technology maturity and the ability to meet strategic objectives, along with pathways to deployment and tradeoffs in mission. It also provided a recommendation on specific technology options to address the current gap in fast neutron irradiation capabilities. In this context, a detailed scope of work was prepared that pertains to a trade study focusing on the core of a fast neutron spectrum test reactor. This report presents the main results obtained by a team of Argonne National Laboratory (ANL) and Idaho National Laboratory (INL) scientists. It aims at developing an overall understanding of how the various areas important to the design and successful operation of such a facility interact with one another. Developing a sense of the relevant sensitivities at a systems level creates an understanding of where and how margins may be available to improve performances (e.g., flux level, capacity factor, and thermal hydraulic limits). Emphasis is given to the reactor (core design options and safety) and its fuel cycle (fuel supply and fabrication, as well as spent fuel management).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Cross section covariances: a user perspective

This paper makes a brief historical review, from the user point of view, of the impact of the different covariance data made available during this period, then look at today current situation in the different evaluated files, using practical examples. In particular, the performance, in terms of C/E, of two of the most recent and reliable evaluate files, CIELO and CIELO-2, is presented for a set of selected integral experiments. Moreover, this performance is compared for consistency against the uncertainty computed with two of the current most used covariance data (COMMARA-2.0 and COMACV1). A limited number of comparisons are also shown relative to the JENDL-4 data. Of specific interest is the observed difference due separately to the diagonal and off-diagonal (correlations) contributions. Finally, some perspective and future needs are discussed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Verification of iterative matrix solutions for multipoint kinetics equations

Multipoint kinetic equation systems have been solved numerically using matrix algebra software and 1st to 4th order implicit schemes based on single-step matrix propagators. These matrix schemes have been validated successfully on demanding point kinetics benchmarks with various prescribed reactivity insertions. Verification tests have also been performed on a simple 3-region fast reactor core model with asymmetric step or ramp reactivity insertions and cross-checked with the multipoint kinetics SACRE code developed at INL and based on existing and validated stiff ODE solver packages. Accurate results can be obtained at a limited time expense. The reason for intriguing results obtained in some transients has been elucidated and linked to the multipoint matrix coefficient interpolation method during the transient.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

New Approaches and Applications for Monte Carlo Perturbation Theory

This paper presents some of the recent and new advancements in the extension of Monte Carlo Perturbation Theory methodologies and application. In particular, the discussed problems involve Brunup calculation, perturbation calculation based on continuous energy functions, and Monte Carlo Perturbation Theory in loosely coupled systems.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Continuous energy cross section adjustment: a new method to generalize nuclear data assimilation for a wider range of applications.

This work presents a new method to perform continuous-energy cross section adjustment. The proposed method is based on recently developed Monte Carlo perturbation capabilities with continuous-energy sensitivity functions implemented in an extended Serpent version. The generation of continuous energy basis functions aseigendecomposition of the covariance matrices and the Generalized Least Squares method applied to these functions are described. The new approach is verified against standard, multi-group adjustment methodologies in a case study involving four different response functions in Jezebel. The results obtained by the two approaches for the selected case study are presented and compared in term of adjusted cross sections, adjusted covariance matrices, and uncertainty reduction.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

New approaches to provide feedback from nuclear and covariance data adjustment for effective improvement of evaluated nuclear data files

A critical examination of the role of uncertainty assessment, target accuracies, role of integral experiment for validation and, consequently, of data adjustments methods is underway since several years at OECD-NEA, the objective being to provide criteria and practical approaches to use effectively the results of sensitivity analyses and cross section adjustments for feedback to evaluators and experimentalists in order to improve without ambiguities the knowledge of neutron cross sections, uncertainties, and correlations to be used in a wide range of applications and to meet new requirements and constraints for innovative reactor and fuel cycle system design. An approach will be described that expands as much as possible the use in the adjustment procedure of selected integral experiments that provide information on “elementary” phenomena, on separated individual physics effects related to specific isotopes or on specific energy ranges. An application to a large experimental data base has been performed and the results are discussed in the perspective of new evaluation projects like the CIELO initiative.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Analysis of a nuclear data experiment: MANTRA-3

To improve the accuracy on the nuclear data, the uncertainties should be kept as low as possible. The major portion of these uncertainties comes from the input data and the assumptions made in the simulation models. For a high profile project like MANTRA, the pre and post irradiation analysis is very important. The current paper discusses the pre and post irradiation analysis of MANTRA-3 that could affect the end results using MCNP5 (a stochastic code) and ECCO-ERANOS (a deterministic code). This analysis was very helpful for understanding the possible uncertainties in MANTRA-3 and for calculating the final results. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

A collision history-based approach to Sensitivity/Perturbation calculations in the continuous energy Monte Carlo code SERPENT

In this work, the implementation of a collision history-based approach to sensitivity/perturbation calculations in the Monte Carlo code SERPENT is discussed. The proposed methods allow the calculation of the eects of nuclear data perturbation on several response functions: the eective multiplication factor, reaction rate ratios and bilinear ratios (e.g., eective kinetics parameters). SERPENT results are compared to ERANOS and TSUNAMI Generalized Perturbation Theory calculations for two fast metallic systems and for a PWR pin-cell benchmark. New methods for the calculation of sensitivities to angular scattering distributions are also presented, which adopts fully continuous (in energy and angle) Monte Carlo estimators.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2015

A-priori and A-posteriori Covariance Data in Nuclear Cross Section Adjustments: Issues and Challenges

In order to provide useful feedback to evaluators a set of criteria are established for assessing the robustness and reliability of the cross section adjustments that make use of integral experiment information. Criteria are also provided for accepting the “a posteriori” cross sections, both as new “nominal” values and as “trends”. Some indications of the use of the “a posteriori” covariance matrix are indicated, even though more investigation is needed to settle this complex subject.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

The Role of Uncertainty Quantification for Reactor Physics

The quantification of uncertainties is a crucial step in design. The comparison of a-priori uncertainties with the target accuracies, allows to define needs and priorities for uncertainty reduction. In view of their impact, the uncertainty analysis requires a reliability assessment of the uncertainty data used. The choice of the appropriate approach and the consistency of different approaches are discussed.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

Comparison of Integral Parameter Estimated Uncertainties in Fast Neutron Systems with COMMARA-2.1 and COMMARA-2.0 Covariance Data

This paper shows the main differences between the COMMARA-2.0 and COMMARA-2.1 evaluated covariance data in the uncertainty estimation of integral parameters of interest for a large number of typical innovative fast neutron systems. The study provides some guidelines to the nuclear data evaluators on which materials and which reaction types need to be evaluated most accurately to achieve reliable results in computational analyses of similar systems. The assessment of the quality and reliability of the covariance data is outside the scope of the study.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2014

Applications of Integral Benchmark Data

The International Reactor Physics Experiment Evaluation Project (IRPhEP) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) provide evaluated integral benchmark data that may be used for validation of reactor physics / nuclear criticality safety analytical methods and data, nuclear data testing, advanced modeling and simulation, and safety analysis licensing activities. The handbooks produced by these programs are used in over 30 countries. Five example applications are presented in this paper: (1) Use of IRPhEP Data in Uncertainty Analyses and Cross Section Adjustment, (2) Uncertainty Evaluation Methods for Reactor Core Design at JAEA Using Reactor Physics Experimental Data, (3) Application of Benchmarking Data to a Broad Range of Criticality Safety Problems, (4) Cross Section Data Testing with ICSBEP Benchmarks, and (5) Use of the International Handbook of Evaluated Reactor Physics Benchmark Experiments to Support the Power Industry.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Physics challenges for advanced fuel cycle assessment

Advanced fuel cycles and associated optimized reactor designs will require substantial improvements in key research area to meet new and more challenging requirements. The present paper reviews challenges and issues in the field of reactor and fuel cycle physics. Typical examples are discussed with, in some cases, original results.

advanced fuel cycles↗
DOE OSTI2014

Working Party on International Nuclear Data Evaluation Cooperation (WPEC)

The OECD Nuclear Energy Agency (NEA) is organizing the cooperation between the major nuclear data evaluation projects in the world. The NEA Working Party on International Nuclear Data Evaluation Cooperation (WPEC) was established to promote the exchange of information on nuclear data evaluation, measurement, nuclear model calculation, validation, and related topics, and to provide a framework for cooperative activities between the participating projects. The working party assesses nuclear data improvement needs and addresses these needs by initiating joint activities in the framework of dedicated WPEC subgroups. Studies recently completed comprise a number of works related to nuclear data covariance and associated processing issues, as well as more specific studies related to the resonance parameter representation in the unresolved resonance region, the gamma production from fission-product capture reactions, the U-235 capture cross-section, the EXFOR database, and the improvement of nuclear data for advanced reactor systems. Ongoing activities focus on the evaluation of Pu-239 in the resonance region, scattering angular distribution in the fast energy range, and reporting/usage of experimental data for evaluation in the resolved resonance region. New activities include two new subgroups on improved fission product yield evaluation methodologies and on modern nuclear database structures. Future activities under discussion include a pilot project of a Collaborative International Evaluated Library (CIELO) and methods to provide feedback from nuclear and covariance data adjustment for improvement of nuclear data. In addition to the above mentioned short-term, task-oriented subgroups, the WPEC also hosts a longer-term subgroup charged with reviewing and compiling the most important nuclear data requirements in a high priority request list (HPRL).

99 GENERAL AND MISCELLANEOUS↗