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DOE OSTI2023

Fuel cycle depletion validation and code-to-code verification studies for High Flux Isotope Reactor highly and low-enriched uranium fuel designs

Here, this paper documents fuel cycle depletion validation and code-to-code verification studies for the High Flux Isotope Reactor (HFIR) highly enriched uranium (HEU) and proposed low-enriched uranium (LEU) fuel designs. In support of HFIR’s world-leading performance, transport and depletion simulations are performed to ensure safe operations, design and qualify irradiation experiments, enhance core components and irradiation facilities, and design and characterize LEU fuel designs. Identifying well-validated, computationally efficient codes is required for the success of these efforts. The HFIR Controller, Shift, and VESTA codes were deployed to simulate HEU uranium–oxide dispersion fuel cycles at 85, 95, and 100 MW operations, as well as LEU fuel cycles operating at 95 MW with uranium–silicide dispersion and uranium–molybdenum monolithic alloy fuel forms. Excellent agreement between the codes and with experimentally obtained 235 U enrichment distributions provides increased confidence in the ability of these codes to model and simulate HFIR’s unique core design.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Validation of general purpose mathematical efficiency modeling with ISOCS - 15579

Gamma spectroscopy is a commonly used method for radioactive waste characterization and activity determination in in-situ objects. The gamma rays emitted by radionuclides in the waste container matrix are highly penetrating and can be measured using one or more gamma ray detectors, thus allowing for a passive nondestructive assay. One problem is that a source based calibration of full energy peak efficiency can be difficult or impractical because of the large size and/or an unusual shape of the sample geometry. One attractive solution is to use mathematical efficiency modeling to calculate the full energy peak efficiency. The In-Situ Object Calibration Software (ISOCS) has successfully been used for more than a decade for mathematical efficiency modeling of template based geometries. Templates include the most common objects in waste characterization and D and D, including but not limited to: boxes, drums, and pipes with hotspots, beams, tanks and walls. Scenarios exist where these templates cannot model the desired measurement geometry. Canberra has developed a general purpose version of ISOCS where primitive objects can be combined into practically any geometry. The primitive objects are right circular cylinder, right elliptical cylinder, right circular cone, right elliptical cone, sphere, spheroid, ellipsoid, box (rectangular parallelepiped), torus, and beaker (any cylindrical symmetric object). Multiple primitive objects can be combined using the priority scheme and cut by planes into almost any desired shape. In addition, multiple objects can be defined in the same scene, for example cylinders next to each other or a pipe and a box. Collimators can be constructed as easily as in template based ISOCS. The ray tracing algorithm in ISOCS has been extended to support the new primitive objects for attenuation calculation between the radioactive source and the detector end cap. The new version uses the same vacuum efficiency response, or characterization, in the energy range 10 - 7000 keV of individual detectors to be able to calculate the efficiency as accurately as possible. MCNP has been shown to be able to calculate detection efficiencies for any geometry and the new version of ISOCS have been validated to MCNP. The vacuum efficiency is based on an MCNP model and this model was the base for constructing an identical model in MCNP as in ISOCS. The calculated efficiencies from the two codes have been compared and the agreement is within 3% except for the lowest energies for one geometry. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2012

FIRST THREE-DIMENSIONAL RECONSTRUCTIONS OF CORONAL LOOPS WITH THE STEREO A+B SPACECRAFT. IV. MAGNETIC MODELING WITH TWISTED FORCE-FREE FIELDS

The three-dimensional coordinates of stereoscopically triangulated loops provide strong constraints for magnetic field models of active regions in the solar corona. Here, we use STEREO/A and B data from some 500 stereoscopically triangulated loops observed in four active regions (2007 April 30, May 9, May 19, and December 11), together with SOHO/MDI line-of-sight magnetograms. We measure the average misalignment angle between the stereoscopic loops and theoretical magnetic field models, finding a mismatch of {mu} = 19 Degree-Sign -46 Degree-Sign for a potential field model, which is reduced to {mu} 14 Degree-Sign -19 Degree-Sign for a non-potential field model parameterized by twist parameters. The residual error is commensurable with stereoscopic measurement errors ({mu}{sub SE} Almost-Equal-To 8 Degree-Sign -12 Degree-Sign ). We developed a potential field code that deconvolves a line-of-sight magnetogram into three magnetic field components (B{sub x} , B{sub y} , B{sub z} ), as well as a non-potential field forward-fitting code that determines the full length of twisted loops (L Almost-Equal-To 50-300 Mm), the number of twist turns (median N{sub twist} = 0.06), the nonlinear force-free {alpha}-parameter (median {alpha} Almost-Equal-To 4 Multiplication-Sign 10{sup -11} cm{sup -1}), and the current density (median j{sub z} Almost-Equal-To 1500 Mx cm{sup -2} s{sup -1}). All twisted loops are found to be far below the critical value for kink instability, and Joule dissipation of their currents is found to be far below the coronal heating requirement. The algorithm developed here, based on an analytical solution of nonlinear force-free fields that is accurate to second order (in the force-free parameter {alpha}), represents the first code that enables fast forward fitting to photospheric magnetograms and stereoscopically triangulated loops in the solar corona.

79 ASTRONOMY AND ASTROPHYSICS↗

DOE OSTI2012

IAEA coordinated research project on thermal-hydraulics of Supercritical Water-Cooled Reactors (SCWRs)

The Supercritical Water-Cooled Reactor (SCWR) is an innovative water-cooled reactor concept, which uses supercritical pressure water as reactor coolant. It has been attracting interest of many researchers in various countries mainly due to its benefits of high thermal efficiency and simple primary systems, resulting in low capital cost. The IAEA started in 2008 a Coordinated Research Project (CRP) on Thermal-Hydraulics of SCWRs as a forum to foster the exchange of technical information and international collaboration in research and development. This paper summarizes the activities and current status of the CRP, as well as major progress achieved to date. At present, 15 institutions closely collaborate in several tasks. Some organizations have been conducting thermal-hydraulics experiments and analysing the data, and others have been participating in code-to-test and/or code-to-code benchmark exercises. The expected outputs of the CRP are also discussed. Finally, the paper introduces several IAEA activities relating to or arising from the CRP. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2012

Benchmark studies of thermal jet mixing in SFRs using a two-jet model

To guide the modeling, simulations and design of Sodium Fast Reactors (SFRs), we explore and compare the predictive capabilities of two numerical solvers COMSOL and OpenFOAM in the thermal jet mixing of two buoyant jets typical of the outlet flow from a SFR tube bundle. This process will help optimize on-going experimental efforts at obtaining high resolution data for V and V of CFD codes as anticipated in next generation nuclear systems. Using the k-{epsilon} turbulence models of both codes as reference, their ability to simulate the turbulence behavior in similar environments was first validated for single jet experimental data reported in literature. This study investigates the thermal mixing of two parallel jets having a temperature difference (hot-to-cold) {Delta}T{sub hc}= 5 deg. C, 10 deg. C and velocity ratios U{sub c}/U{sub h} = 0.5, 1. Results of the computed turbulent quantities due to convective mixing and the variations in flow field along the axial position are presented. In addition, this study also evaluates the effect of spacing ratio between jets in predicting the flow field and jet behavior in near and far fields. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2023

Contributions to MoDELib SOFTWARE

The purpose of the current request is to enable LANL employees to contribute computer source code to the existing public repository of the MoDELib software package. This software implements discrete dislocation dynamics (DDD) and finite element (FEM) methods and is currently a vital component of an ongoing DR project at LANL, in collaboration with its original author and maintainer Giacomo Po. Contributions from LANL employees would aim to enhance the reliability, accuracy, and performance of MoDELib simulations using LANL's high performance computing platforms through bug fixes, algorithmic refinements, and parallelization.

Julian, Nicholas↗

DOE OSTI2011

Heart simulation with surface equations for using on MCNP code

External photon beam radiotherapy is carried out in a way to achieve an 'as low as possible' a dose in healthy tissues surrounding the target. One of these surroundings can be heart as a vital organ of body. As it is impossible to directly determine the absorbed dose by heart, using phantoms is one way to acquire information around it. The other way is Monte Carlo method. In this work we have presented a simulation of heart geometry by introducing of different surfaces in MCNP code. We used 14 surface equations in order to determine human heart modeling. Those surfaces are borders of heart walls and contents.

62 RADIOLOGY AND NUCLEAR MEDICINE↗

DOE OSTI2011

Use of mathematical modeling in nuclear measurements projects

Mathematical modeling of nuclear measurement systems is not a new concept. The response of the measurement system is described using a pre-defined mathematical model that depends on a set of parameters. These parameters are determined using a limited set of experimental measurement points e.g. efficiency curve, dose rates... etc. The model that agrees with the few experimental points is called an experimentally validated model. Once these models have been validated, we use mathematical interpolation to find the parameters of interest. Sometimes, when measurements are not practical or are impossible extrapolation is implemented but with care. CANBERRA has been extensively using mathematical modeling for the design and calibration of large and sophisticated systems to create and optimize designs that would be prohibitively expensive with only experimental tools. The case studies that will be presented here are primarily performed with MCNP, CANBERRA's MERCURAD/PASCALYS and ISOCS (In Situ Object Counting Software). For benchmarking purposes, both Monte Carlo and ray-tracing based codes are inter-compared to show models consistency and add a degree of reliability to modeling results. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2016

A Chemistry Based Criticality Safety Evaluation of the HM-Process Mixer Settlers

The criticality safety of Savannah River Site’s H Canyon two cycle uranium extraction and purification process, called the HM process, is examined. Explicit Monte Carlo models of the multi-stage mixer-settlers are generated for the first time using the KENO-VI module of the SCALE 6.1 code package and spanning normal operating conditions and a wide range of potential upset conditions with stage wise compositions based directly on SEPHIS Mod-4 chemistry models of the same conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

SU-F-R-13: Decoding 18F-FDG Uptake Heterogeneity for Primary and Lymphoma Tumors by Using Texture Analysis in PET Images

Purpose: To explore 18F-FDG uptake heterogeneity of primary tumor and lymphoma tumor by texture features of PET image and quantify the heterogeneity difference between primary tumor and lymphoma tumor. Methods: 18 patients with primary tumor and lymphoma tumor in lung cancer were enrolled. All patients underwent whole-body 18F-FDG PET/CT scans before treatment. Texture features, based on Gray-level Co-occurrence Matrix, second and high order matrices are extracted from code using MATLAB software to quantify 18F-FDG uptake heterogeneity. The relationships of volume between energy, entropy, correlation, homogeneity and contrast were analyzed. Results: For different cases, tumor heterogeneity was not the same. Texture parameters (contrast, entropy, and correlation) of lymphoma were lower than primary tumor. On the contrast, the texture parameters (energy, homogeneity and inverse different moment) of lymphoma were higher than primary tumor. Significantly, correlations were observed between volume and energy (primary, r=−0.194, p=0.441; lymphoma, r=−0.339, p=0.582), homogeneity (primary, r=−0.146, p=0.382; lymphoma, r=−0.193, p=0.44), inverse difference moment (primary, r=−0.14, p=0.374; lymphoma, r=−0.172, p=0.414) and a positive correlation between volume and entropy (primary, r=0.233, p=0.483; lymphoma, r=0.462, p=0.680), contrast (primary, r=0.159, p=0.399; lymphoma, r=0.341, p=0.584), correlation (primary, r=0.027, p=0.165; lymphoma, r=0.046, p=0.215). For the same patient, energy for primary and lymphoma tumor is equal. The volume of lymphoma is smaller than primary tumor, but the homogeneity were higher than primary tumor. Conclusion: This study showed that there were effective heterogeneity differences between primary and lymphoma tumor by FDG-PET image texture analysis.

60 APPLIED LIFE SCIENCES↗

DOE OSTI2013

A BrachyPhantom for verification of dose calculation of HDR brachytherapy planning system

Purpose: To develop a calibration phantom for {sup 192}Ir high dose rate (HDR) brachytherapy units that renders possible the direct measurement of absorbed dose to water and verification of treatment planning system.Methods: A phantom, herein designated BrachyPhantom, consists of a Solid Water™ 8-cm high cylinder with a diameter of 14 cm cavity in its axis that allows the positioning of an A1SL ionization chamber with its reference measuring point at the midheight of the cylinder's axis. Inside the BrachyPhantom, at a 3-cm radial distance from the chamber's reference measuring point, there is a circular channel connected to a cylindrical-guide cavity that allows the insertion of a 6-French flexible plastic catheter from the BrachyPhantom surface. The PENELOPE Monte Carlo code was used to calculate a factor, P{sub sw}{sup lw}, to correct the reading of the ionization chamber to a full scatter condition in liquid water. The verification of dose calculation of a HDR brachytherapy treatment planning system was performed by inserting a catheter with a dummy source in the phantom channel and scanning it with a CT. The CT scan was then transferred to the HDR computer program in which a multiple treatment plan was programmed to deliver a total dose of 150 cGy to the ionization chamber. The instrument reading was then converted to absorbed dose to water using the N{sub gas} formalism and the P{sub sw}{sup lw} factor. Likewise, the absorbed dose to water was calculated using the source strength, S{sub k}, values provided by 15 institutions visited in this work.Results: A value of 1.020 (0.09%, k= 2) was found for P{sub sw}{sup lw}. The expanded uncertainty in the absorbed dose assessed with the BrachyPhantom was found to be 2.12% (k= 1). To an associated S{sub k} of 27.8 cGy m{sup 2} h{sup −1}, the total irradiation time to deliver 150 cGy to the ionization chamber point of reference was 161.0 s. The deviation between the absorbed doses to water assessed with the BrachyPhantom and those calculated by the treatment plans and using the S{sub k} values did not exceed ±3% and ±1.6%, respectively.Conclusions: The BrachyPhantom may be conveniently used for quality assurance and/or verification of HDR planning system with a priori threshold level to spot problems of 2% and ±3%, respectively, and in the long run save time for the medical physicist.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2018

Single-Center Experience Implementing the LOINC-RSNA Radiology Playbook for Adult Abdomen/Pelvis CT and MR Procedures Using a Semi-Automated Method

The LOINC-RSNA Radiology Playbook represents the future direction of standardization for radiology procedure names. We developed a software solution (“RadMatch”) utilizing Python 2.7 and FuzzyWuzzy, an open-source fuzzy string matching algorithm created by SeatGeek, to implement the LOINC-RSNA Radiology Playbook for adult abdomen and pelvis CT and MR procedures performed at our institution. Execution of this semi-automated method resulted in the assignment of appropriate LOINC numbers to 86% of local CT procedures. For local MR procedures, appropriate LOINC numbers were assigned to 75% of these procedures whereas 12.5% of local MR procedures could only be partially mapped. For the standardized local procedures, only 63% of CT and 71% of MR procedures had corresponding RadLex Playbook identifier (RPID) codes in the LOINC-RSNA Radiology Playbook, which limited the utility of RPID codes. RadMatch is a semi-automated open-source software tool that can assist radiology departments seeking to standardize their radiology procedures via implementation of the LOINC-RSNA Radiology Playbook.

62 RADIOLOGY AND NUCLEAR MEDICINE↗

DOE OSTI2014

The Activities of the European Consortium on Nuclear Data Development and Analysis for Fusion

This paper presents an overview of the activities of the European Consortium on Nuclear Data Development and Analysis for Fusion. The Consortium combines available European expertise to provide services for the generation, maintenance, and validation of nuclear data evaluations and data files relevant for ITER, IFMIF and DEMO, as well as codes and software tools required for related nuclear calculations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2010

Astrophysical S-Factors and Reaction Rates of Threshold (p, n)-Reactions on {sup 99-102}Ru

Astrophysical S-factors of (p, n) reactions on {sup 99}Ru, {sup 100}Ru, {sup 101}Ru, and {sup 102}Ru were derived from the sum of experimental isomeric and ground states cross sections measured in the incident proton energy range of 5-9 MeV. They were compared with Hauser-Feshbach statistical model predictions of the NON-SMOKER code. Good agreement was found in the majority of cases. Reaction rates were derived up to 8.7 GK stellar temperature by combining experiment and theory.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2015

COUPLING OF RELAP5-SCDAP MOD4.0 AND NEUTRONIC CODES

High-fidelity and accurate nuclear system codes play a key role in the design and analysis of complex nuclear power plants, which consist of multiple subsystems, such as the reactor core (and its fuel, burnable poisons, control elements, etc.), the reactor internal structures, the vessel, and the energy conversion subsystem and beyond to grid demand. Most commonly the interplay between these various subsystems is modeled using coupled codes, each of which represents one of the subsystems. And the most common direct coupling is that of thermal-hydraulics and neutronics codes. The subject of this paper is the coupling of codes that model not only thermal-hydraulics and neutronics, but also structural components damage. Furthermore, the neutronic component is not limited to the sole core solver. The coupled code system encompasses thermal-hydraulics, material performance of the fuel, neutronic solver, and neutronic data preparation. Thus, this paper presents a framework for coupling RELAP5/SCDAPSIM/MOD4.0 with a suite of neutron kinetics codes that includes NESTLE, DRAGON and a version of the ENDF library. The version of the RELAP5/SCDAPSIM/MOD4.0 code used in this work is one developed by Innovate System Software (ISS) as part of the international SCDAP Development and Training Program (SDTP) for best-estimate analysis to model reactor transients including severe accident phenomena. This RELAP5/SCDAPSIM/MOD4.0 code version is also capable of predicting nuclear fuel performance. It uses nodal power distributions to calculate mechanical and thermal parameters such as heat-up, oxidation and meltdown of fuel rods and control rods, the ballooning and rupture of fuel rod cladding, the release of fission products from fuel rods, and the disintegration of fuel rods into porous debris and molten material. On the neutronics side, this work uses the NESTLE and DRAGON codes. NESTLE is a multi-dimensional static and kinetic neutronic code developed at North Carolina State University. It solves up to four energy groups neutron diffusion equations utilizing the Nodal Expansion Method (NEM) in Cartesian or hexagonal geometry. The DRAGON code, developed at Ecole Polytechnique de Montreal, performs lattice physics calculations based on the neutron transport equation and is capable of using very fine energy group structures. In this work, we have developed a coupling approach to exchange data among the various modules. In the coupling process, the generated nuclear data (in fine multigroup energy structure) are collapsed down into two- or four-group energy structures for use in NESTLE. The neutron kinetics and thermal-hydraulics modules are coupled at each time step by using the cross-section data. The power distribution results of the neutronic calculations are transmitted to the thermal-hydraulics code. The spatial distribution of coolant density and the fuel-moderator temperature, which result from the thermal-hydraulic calculations, are transmitted back to the neutron kinetics codes and then the loop is closed using new neutronics results. Details of the actual data transfers will be described in the full length paper.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2010

Development, analysis, and evaluation of a commercial software framework for the study of Extremely Low Probability of Rupture (xLPR) events at nuclear power plants.

Sandia National Laboratories (SNL) participated in a Pilot Study to examine the process and requirements to create a software system to assess the extremely low probability of pipe rupture (xLPR) in nuclear power plants. This project was tasked to develop a prototype xLPR model leveraging existing fracture mechanics models and codes coupled with a commercial software framework to determine the framework, model, and architecture requirements appropriate for building a modular-based code. The xLPR pilot study was conducted to demonstrate the feasibility of the proposed developmental process and framework for a probabilistic code to address degradation mechanisms in piping system safety assessments. The pilot study includes a demonstration problem to assess the probability of rupture of DM pressurizer surge nozzle welds degraded by primary water stress-corrosion cracking (PWSCC). The pilot study was designed to define and develop the framework and model; then construct a prototype software system based on the proposed model. The second phase of the project will be a longer term program and code development effort focusing on the generic, primary piping integrity issues (xLPR code). The results and recommendations presented in this report will be used to help the U.S. Nuclear Regulatory Commission (NRC) define the requirements for the longer term program.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2015

Postclosure assessment modelling of a deep geologic repository for low and intermediate level radioactive waste in Ontario, Canada - 15145

In 2011, Ontario Power Generation (OPG) submitted an Environmental Impact Statement and a Preliminary Safety Report (PSR) to the federal authorities in Canada to support an application for a site preparation and construction licence for a Deep Geologic Repository (DGR) for low and intermediate level radioactive waste (L and ILW). The DGR, which is proposed to be located close to the existing Western Waste Management Facility (WWMF) at the Bruce nuclear site, would receive L and ILW arising from the operation and refurbishment of OPG's nuclear reactors; the wastes are currently stored safely at the WWMF. The PSR is supported by postclosure safety assessment modelling, including a total-systems level model implemented in the AMBER compartment modelling software that draws directly on detailed groundwater and coupled groundwater and gas modelling using the FRA3DVS-OPG and TOUGH2 codes. A full suite of calculation cases is managed within a single AMBER model, which aided efficient management of the calculation files and quality assurance. Assessment of expected/normal evolution cases results in maximum calculated effective doses to humans that are more than five orders of magnitude below associated criteria. Assessment of conservative disruptive and 'what if' scenarios results in maximum calculated doses that are higher than those for the normal evolution cases, but which, when the low likelihood of occurrence is taken into account, remain significantly below the associated risk criterion. The assessment demonstrates that deep geologic repositories can provide a safe disposal option and provides an example of evaluating the safety of projects on long time scales. It shows how results from detailed groundwater and gas codes can be incorporated into assessment-level codes such as AMBER, thereby providing a single systems model that can be used to efficiently evaluate a wide range of scenarios and associated calculation cases. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

Mathematical Model for Evaluating the Technical Condition and Predicting Collapse of Protective Barriers on Flooded Radiation-Hazardous Sites

The possible consequences of flooding of nuclear and radiation-hazardous sites and objects should be included among the environmental risks arising during the rapid development of the infrastructure in the Arctic. A mathematical model, algorithms, and software codes for evaluating the condition and predicting the collapse of the protective barriers, and determining the emission rate of radionuclides and the possible radioactive contamination of the environment, including long-term prediction and situation analysis, have been developed. The software codes have been accepted into the government registry at Rospatent.

54 ENVIRONMENTAL SCIENCES↗