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DOE OSTI2022

Development of nuclear data processing code FRENDY version 2

Nuclear data processing is an important interface between an evaluated nuclear data library and neutronics calculation codes. JAEA has been developed the new nuclear data processing code FRENDY from 2013. FRENDY version 1 generates the ACE files used for the continuous-energy Monte Carlo codes including PHITS, Solomon, Serpent,and MCNP and it was released as the open-source software under the 2-clause BSD license in 2019.After we released FRENDY version 1, many functions, e.g., the multi-group neutron cross-section library generation, the statistical uncertainty quantification of the probability tables for unresolved resonance cross-section, the 'perturbation' of the ACE file for uncertainty quantification, and the modification of the ENDF-6 formatted nuclear data file, have been developed. We released FRENDY version 2 including these functions. The present paper gives an overview of FRENDY and presents features of new functions implemented in FRENDY version 2. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2018

Transforming State-of-the-Art into Best Practice: A Guide for High-Performance Energy Efficient Buildings in India

India launched the Energy Conservation Building Code (ECBC) in 2007, and a revised version in 2017 as ambitious first steps towards promoting energy efficiency in the building sector. Pioneering early adopters—building owners, A&E firms, and energy consultants—have taken the lead to design customized solutions for their energy-efficient buildings. This Guide offers a synthesizing framework, critical lessons, and guidance to meet and exceed ECBC. Its whole-building lifecycle assurance framework provides a user-friendly methodology to achieve high performance in terms of energy, environmental, and societal impact. Class A offices are selected as a target typology, being a high-growth sector, with significant opportunities for energy savings. The practices may be extrapolated to other commercial building sectors, as well as extended to other regions with similar cultural, climatic, construction, and developmental contexts

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗

NASA NTRS1991

The design/analysis of flows through turbomachinery: A viscous/inviscid approach

The development of a design/analysis flow solver at NASA Lewis Research Center is discussed. The solver is axisymmetric and can be run inviscidly with assumed or calculated blockages, or with the viscous terms computed. The blade forces for each blade row are computed from blade-to-blade solutions, correlated data or force model, or from a full three dimensional solution. Codes currently under development can be separated into three distinct elements: the turbomachinery interactive grid generator energy distribution restart code (TIGGERC), the interactive blade element geometry generator (IBEGG), and the viscous/inviscid multi-blade-row average passage flow solver (VIADAC). Several experimental test cases were run to validate the VIADAC code. The tests, representative of typical axial turbomachinery duct axisymmetric wind tunnel body problems, were conducted on an SR7 Spinner axisymmetric body, a NASA Rotor 67 Fan test bed, and a transonic boatail body. The results show the computations to be in good agreement with test data.

Miller, D. P.↗

DOE OSTI2012

Application of wavelet scaling function expansion continuous-energy resonance calculation method to MOX fuel problem

More and more MOX fuels are used in all over the world in the past several decades. Compared with UO{sub 2} fuel, it contains some new features. For example, the neutron spectrum is harder and more resonance interference effects within the resonance energy range are introduced because of more resonant nuclides contained in the MOX fuel. In this paper, the wavelets scaling function expansion method is applied to study the resonance behavior of plutonium isotopes within MOX fuel. Wavelets scaling function expansion continuous-energy self-shielding method is developed recently. It has been validated and verified by comparison to Monte Carlo calculations. In this method, the continuous-energy cross-sections are utilized within resonance energy, which means that it's capable to solve problems with serious resonance interference effects without iteration calculations. Therefore, this method adapts to treat the MOX fuel resonance calculation problem natively. Furthermore, plutonium isotopes have fierce oscillations of total cross-section within thermal energy range, especially for {sup 240}Pu and {sup 242}Pu. To take thermal resonance effect of plutonium isotopes into consideration the wavelet scaling function expansion continuous-energy resonance calculation code WAVERESON is enhanced by applying the free gas scattering kernel to obtain the continuous-energy scattering source within thermal energy range (2.1 eV to 4.0 eV) contrasting against the resonance energy range in which the elastic scattering kernel is utilized. Finally, all of the calculation results of WAVERESON are compared with MCNP calculation. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2013

Nuclear data processing for energy release and deposition calculations in the MC21 Monte Carlo code

With the recent emphasis in performing multiphysics calculations using Monte Carlo transport codes such as MC21, the need for accurate estimates of the energy deposition-and the subsequent heating - has increased. However, the availability and quality of data necessary to enable accurate neutron and photon energy deposition calculations can be an issue. A comprehensive method for handling the nuclear data required for energy deposition calculations in MC21 has been developed using the NDEX nuclear data processing system and leveraging the capabilities of NJOY. The method provides a collection of data to the MC21 Monte Carlo code supporting the computation of a wide variety of energy release and deposition tallies while also allowing calculations with different levels of fidelity to be performed. Detailed discussions on the usage of the various components of the energy release data are provided to demonstrate novel methods in borrowing photon production data, correcting for negative energy release quantities, and adjusting Q values when necessary to preserve energy balance. Since energy deposition within a reactor is a result of both neutron and photon interactions with materials, a discussion on the photon energy deposition data processing is also provided. (authors)

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2013

Nuclear fragmentation measurements for hadrontherapy and space radiation protection

Nuclear fragmentation measurements are necessary in hadrontherapy and space radiation protection, to predict the effects of the ion nuclear interactions within the human body. Nowadays, a very limited set of carbon fragmentation cross sections has been measured and in particular, to our knowledge, no double differential fragmentation cross sections at intermediate energies are available in literature. We have measured the double differential cross sections and the angular distributions of the secondary fragments produced in the {sup 12}C fragmentation at 62 AMeV on a thin carbon target. The experimental data have been also used to benchmark the prediction capability of the Geant4 Monte Carlo code at intermediate energies, where it was never tested before.

62 RADIOLOGY AND NUCLEAR MEDICINE↗

DOE OSTI2024

Pathways to commercial building plug and process load efficiency and control

Abstract To accomplish net-zero carbon emissions in the built environment by 2050, we must equitably decarbonize commercial buildings, including reducing plug and process loads (PPLs). PPLs are plug-in or hardwired electric and gas loads that are not associated with major building end uses like lighting and HVAC. Research shows PPL energy reduction strategies and control technologies have the potential to save energy. But even when implemented, these savings have rarely been achieved and there has not been widespread uptake in U.S. commercial buildings. We investigate why these technologies and strategies have not seen widespread adoption and identify behavior and technology pathways to increase PPL reduction in U.S. commercial buildings. We examined behaviors of commercial building stakeholders through 44 interviews and cross-referenced qualitative analysis findings with in-depth technical knowledge of existing PPL control technologies and reduction strategies. PPL control implementation must be paired with management strategies, such as occupant engagement and training, to achieve optimal savings, and best practices should be disseminated across the industry. We found that increasing access to cost and energy savings data will promote uptake of PPL control technologies and allow designers to better incorporate PPLs into building design. Improving access to funding for PPL energy efficiency projects and addressing the split-incentive problem will increase adoption of PPL efficiency and control. Code bodies should continue to include PPL monitoring and reduction measures in energy codes. Key building stakeholders, including cybersecurity and information technology teams, should be involved in PPL monitoring and reduction strategy processes for successful implementation.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗

DOE OSTI2019

Doppler Broadening Resonance Correction for Free-gas Scattering in MCNP6.2

The traditional treatment for epithermal neutron scattering in continuous-energy Monte Carlo codes is the free-gas scattering model with the scattering cross-section assumed constant in energy. Ouislomen and Sanchez demonstrated in 1991 that scattering resonances can cause significant departures from the free-gas model, and Becker proposed a correction to the Monte Carlo free-gas scattering model in 2009. Becker’s Doppler Broadening Resonance Correction was tested in mcnp5 in 2011 and recently implemented in mcnp6.2 .

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

NASA NTRS1996

On Kinetics Modeling of Vibrational Energy Transfer

Two models of vibrational energy exchange are compared at equilibrium to the elementary vibrational exchange reaction for a binary mixture. The first model, non-linear in the species vibrational energies, was derived by Schwartz, Slawsky, and Herzfeld (SSH) by considering the detailed kinetics of vibrational energy levels. This model recovers the result demanded at equilibrium by the elementary reaction. The second model is more recent, and is gaining use in certain areas of computational fluid dynamics. This model, linear in the species vibrational energies, is shown not to recover the required equilibrium result. Further, this more recent model is inconsistent with its suggested rate constants in that those rate constants were inferred from measurements by using the SSH model to reduce the data. The non-linear versus linear nature of these two models can lead to significant differences in vibrational energy coupling. Use of the contemporary model may lead to significant misconceptions, especially when integrated in computer codes considering multiple energy coupling mechanisms.

Gilmore, John O.↗

DOE OSTI2012

Temperature Effects of Resonance Scattering for Epithermal Neutrons in MCNP

Epithermal neutron elastic scattering can be significantly affected by the thermal motion of target nuclides. Since the 1950s continuous-energy Monte Carlo codes have generally accounted for the target motion using a free gas scattering model, with the assumption that the scattering cross section is constant in energy. Recent work has shown the importance of resonance scattering, and several methods for an improved free-gas treatment have been developed. We have implemented a rejection-based sampling scheme in the MCNP free-gas treatment to account for cross-section variation. The modified MCNP code was used to investigate a number of practical concerns: results for an LWR Doppler defect benchmark; computational costs; and energy limits for the free gas treatment. Additionally, the impact on a suite of ICSBEP criticality benchmark problems (at room temperature) was determined to be negligible, an important result since such problems are used extensively in testing and evaluating revisions to ENDFIB-VII nuclear data.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2014

SNL-SWAN

Sandia National Laboratories has modified Simulating WAves Nearshore (SWAN), the Delft University of Technology code,, to include functionality to better model wave energy converters (WECs). The code modified by SNL has been named SNL-SWAN. SNL-SWAN includes a frequency dependent WEC Module that builds upon the traditional obstacle module in SWAN. In SNL-SWAN, a frequency dependent transmission coefficient is calculated to model power absorbed by WEC devices. The power transmission coefficients are calculated based on the WEC's power performance, which is defined by the user in the input text files. Two methods of defining WEC power performance have been implemented: a user-specified power matrix, and a user-specified relative capture width. These methods allow the user to define the WEC's frequency dependent energy absorption as a function of period and/or wave height. SWAN test cases were run to verify the compiled code's functionality, and verification of the frequency dependent WEC Module in SNL-SWAN has also been performed. SNL-SWAN code is will be released as open source code and will be made publicly available.

Chartrand, Chris↗

DOE OSTI2015

SNL-SWAN

Source record↗

DOE OSTI2012

Temperature effects of resonance scattering for epithermal neutrons in MCNP

Epithermal neutron elastic scattering can be significantly affected by the thermal motion of target nuclides. Since the 1950's continuous-energy Monte Carlo codes have generally accounted for the target motion using a free gas scattering model, with the assumption that the scattering cross-section is constant in energy. Recent work has shown the importance of resonance scattering, and several methods for an improved free-gas treatment have been developed. We have implemented a rejection-based sampling scheme in the MCNP free-gas treatment to account for cross-section variation. The modified MCNP code was used to investigate a number of practical concerns: results for an LWR Doppler defect benchmark; computational costs; and energy limits for the free-gas treatment. Additionally, the impact on a suite of ICSBEP criticality benchmark problems (at room temperature) was determined to be negligible, an important result since such problems are used extensively in testing and evaluating revisions to ENDF/B-VII nuclear data. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2014

Effects from the Reduction of Air Leakage on Energy and Durability

Buildings are responsible for approximately 40% of the energy used in the US. Codes have been increasing building envelope requirements, and in particular those related to improving airtightness, in order to reduce energy consumption. The main goal of this research was to evaluate the effects from reductions in air leakage on energy loads and material durability. To this end, we focused on the airtightness and thermal resistance criteria set by the 2012 International Energy Conservation Code (IECC).

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗

DOE OSTI2015

Flattening of fuel importance for heterogeneous system to estimate maximal multiplication factor - 14595

In order to quantify risks of re-criticality incidents associated with loss of coolant accidents (LOCA), the maximum reactivity insertion by relocation of damaged fuel should be quantified. As the first step for the quantification, an automated search scheme is studied for an optimal geometry of damaged fuel in the coolant channel where the neutron multiplication factor k{sub eff} takes a maximal value. The scheme is based on the flattening of the fuel importance. The importance can be recognized as the substitution reactivity from fuel to water in heterogeneous systems. The flattening is achieved by substitution of water and the fuel so as to raise the k{sub eff} referring the local substitution reactivities. To calculate the reactivities efficiently, the first order perturbation scheme is adopted. In the scheme, the adjoint flux is calculated with the PARTISN code. The multi group constants used for the calculation and the perturbed reaction rates are estimated with the continuous energy Monte Carlo code MCNP-5. The proposed methods are tested for a quarter pin cell model that is constructed using Cartesian mesh. From a random distribution of a given amount of fragmented fuel, an optimal geometry is found where the fragmented fuels are aggregated. The multiplication factor in the geometry is greater than that in the original random distribution by 1.4%Δk/k. That indicates the significance of this study. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Uranium measurements in the field using high-resolution cadmium zinc telluride detectors

A new generation of cadmium zinc telluride (CZT) detectors has become available and is being evaluated by the International Atomic Energy Agency (IAEA) for safeguards verifications in the field. The new CZT detector, model M400, is a room temperature spectrometer manufactured by H3D, Inc. The M400 demonstrates superior energy resolution, effective isotope identification capabilities, and convenient usability features when tested in a controlled laboratory environment. These characteristics define the M400 as a potential platform for IAEA field detection applications, which could become suitable for nuclear material characterization (e.g., enrichment verification) and nuclear safeguards missions. The capabilities of gamma spectrometry codes including Fixed energy, Response function Analysis with Multiple efficiencies (FRAM) from Los Alamos National Laboratory, CZT for Uranium (CZTU) from Lawrence Livermore National Laboratory, and Gamma Detector Response and Analysis Software (GADRAS) from Sandia National Laboratories were adapted for M400 spectra, and the performance of the codes has been validated. This was reported in a prior work. To further validate the performance of the high-energy resolution CZT detector and the isotopic analysis codes, a field measurement campaign consisting of uranium hexafluoride (UF 6 ) cylinder measurements was conducted at a fuel fabrication facility. A total of 34 Type 30B cylinders containing UF6 were measured using three different M400 CZT detectors. Each detector was outfitted with a custom rectangular collimator and shield made out T-Flex®, a tungsten-impregnated polymer. Measurements were performed at three different locations of the cylinder, ensuring that the measurement geometry satisfied the infinite thickness criterion. The spectra from the M400 CZT were analyzed using the code General Enrichment Meter (GEM). For analyzing the gamma-ray spectra from UF 6 cylinder, the GEM code is the appropriate tool since it relies only on the gamma-ray emissions from 235 U and not from other isotopes. Results from the spectral analysis were compared with the known abundance of 235 U in the cylinders, as well as with the International Target Values 2020 (ITV2020). The suitability of the different underlying techniques used by the various codes for UF 6 analysis is discussed. The challenges of measuring UF 6 contained in cylinders and mitigation strategies are highlighted.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2022

Comparative analysis of energy deposition modes available in Serpent 2 within the framework of the supercritical water reactor - Fuel qualification test reactor physics benchmark

A joint European Canadian Chinese development of a supercritical water-cooled small modular reactor (SCW-SMR) technology is in progress since September 2020 in the framework of a Horizon 2020 project called ECC-SMART. As a main purpose of the project, proper estimates of energy deposition and its spatial distribution are prerequisites for the accurate analysis of safety related parameters of the SCW-SMR concept under development. A supercritical water reactor fuel computational benchmark model, provided by Canadian Nuclear Laboratories, was applied for detailed comparison of different energy deposition calculation options available in the Serpent 2 Monte Carlo code. The effect of energy deposition options on the normalization of the results as well as on the spatial distribution of the energy deposition are discussed. Consistent energy deposition calculation methods are presented between three Monte Carlo codes, viz., Serpent 2, MCNP6 and OpenMC. Although resource-intensive, the use of the coupled neutron-photon transport mode of Serpent 2 is recommended for accurate spatial and quantitative characterization of energy deposition in the SCW-SMR fuel assemblies, accounting for both neutron and photon heating of all the materials. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2021

Review of WEC-Sim Development and Applications

WEC-Sim (Wave Energy Converter Simulator) is an open-source code for simulating wave energy converters that has been actively developed and applied to simulate a wide variety of device archetypes, and has become a popular tool since its release. This paper reviews the development efforts and usage of WEC-Sim. The publications considered in this study have been broken down into six topic areas, namely, feature development, experimental validation, device modelling, control modelling, powertake-off (PTO) and grid modelling, and novel applications, which includes some non-wave energy applications. This review paper also recognizes the contributions of academic researchers and technology developers from around the world toward the broader WEC-Sim development effort. The growing number of external applications of WEC-Sim demonstrates a broader acceptance of the open-source code, and the ways WEC-Sim has been used in certain topic areas also highlight potential future development needs.

applications↗