NASA NDE Working Group Newsletter (Vol. 2, No. 2)
Contents: NASA HQ CODE Q, NASA HQ Code QW Message, New NDE Manager, NNWG Highlight, Current events and Action Items, NASA Centers News and Announcements.
CODE Q Code QW↗
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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.
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Algorithms for software development version control and change detection
Simple computer algorithms for processing source program and text data files in order to extract change detection, version control, version history, and current status information are described. These algorithms presuppose that it is possible to attach to each record of the source files a 6-character code, placed within delimiters that will cause the compiler, or other using program, to ignore this code field. The code contains a 2-character code for a character-by-character position-sensitive checksum of the record, another for the record number in the file, and a third for the data on which the encoding took place. Once the source file has been thus encoded, it is possible to detect the following transactions on the file since the most recent version coding; (1) addition of new records (having no version code), (2) modification of existing records, (3) deletion of a number of records, (4) movement and/or duplication of existing records, and (5) modification and duplication of records. In addition, it is possible to extract a version history of the number of records created or modified by date. A special file listing program is described which prints the file records without showing the version codes, but places a "change bar" at the right margin whenever a change is detected. The program also provides a list of changed pages and a version history.
Tausworthe, R. C.↗
Status of the OECD/NEA Watts Bar unit 1 benchmark and calculations of local reactor core power data of the Zero Power physics tests
The paper aims to provide an update on the status of the Organization for Economic Cooperation and Development (OECD) / the Nuclear Energy Agency (NEA) Tennessee Valley Authority (TVA) Watts Bar 1 (WB1) Multi-Physics Multi-Cycle depletion benchmark and the results associated with Exercise 1. The benchmark relies on the set of benchmark progression problems developed by the Department of Energy (DOE) Consortium for the Advanced Simulation of Light Water Reactors (CASL) for the Virtual Environment for Reactor (VERA), which are based on real plant design and operational data. The OECD/NEA TVA WB1 benchmark is designed for validation of both traditional and novel high-fidelity multi-physics codes to analyze Pressurized Water Reactors (PWR) depletion cycles. The activities are conducted under the Expert Group on Reactor Systems Multi-Physics (EGMUP) at NEA/OECD. In this work, we analyze Exercise 1 of the benchmark: stand-alone Three-Dimensional (3D) neutronics at Start-up Zero Power Physics Test (ZPPT) at Hot Zero Power Conditions (HZP) and the resulting power maps. The code used to model the exercise is the continuous energy Monte Carlo code Serpent 2.1.31 along with the nuclear data library ENDF/B-VII.1. Serpent results are compared with the high-fidelity deterministic code MPACT, which is part of VERA. The paper presents three selected results from power calculations required output. The results compared are the normalized axially integrated radial core power maps, the normalized axial averaged core power shapes, and the normalized core hottest and coldest assemblies radial power maps. The Root Mean Square Deviation (RMSD) between Serpent and MPACT is 0.51 % for the axially integrated radial core power map, 1.41 % for the axial averaged core power shape, 2.42 % for the hottest assembly axial power shape, 0.88 % for the coldest assembly axial integrated hottest assembly radial pin power map, and 0.17 % for the axially integrated coldest assembly radial pin power map. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Assessment of the analysis capability for core-wide PWR pellet-clad interaction screening of Watts Bar Unit 1
This report illustrates the application of the Virtual Environment for Reactor Applications simulation environment to pressurized water reactor core-wide pellet-clad interaction screening, with emphasis on the multi-dimensional analysis capabilities of the BISON fuel performance code. The Virtual Environment for Reactor Applications codes utilized in this effort include the BISON fuel performance code and the VERA-CS sub-element, which combines the MPACT neutron transport code and the COBRA-TF thermal-hydraulics code. The target reactor for this work is Watts Bar Unit 1, cycle 6 and cycle 7, selected on the basis that pellet-clad interaction failures had occurred during power start up in both cycles, despite the moderate power levels. These instances of failure at Watts Bar Unit 1 offered an interesting test of the code’s veracity and a particularly challenging condition for this assessment.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Star Formation Activity of Barred Spiral Galaxies
We study the star formation activity of nearby galaxies with bars using a sample of late-type galaxies at 0.02⩽z⩽0.05489 and M{sub r}<−19.5 from the Sloan Digital Sky Survey. We compare the physical properties of strongly and weakly barred galaxies with those of non-barred galaxies that have stellar mass and redshift distributions similar to barred galaxies. We find that the star formation activity of strongly barred galaxies probed by starburstiness, g−r, NUV−r, and mid-infrared [3.4]–[12] colors is, on average, lower than that of non-barred galaxies. However, weakly barred galaxies do not show such a difference between barred and non-barred galaxies. The amounts of atomic and molecular gas in strongly barred galaxies are smaller than those in non-barred galaxies, and the gas metallicity is higher in strongly barred galaxies than in non-barred galaxies. The gas properties of weakly barred galaxies again show no difference from those of non-barred galaxies. We stack the optical spectra of barred and non-barred galaxies in several mass bins and fit to the stacked spectra with a spectral fitting code, STARLIGHT. We find no significant difference in stellar populations between barred and non-barred galaxies for both strongly and weakly barred galaxies. Our results are consistent with the idea that the star formation activity of barred galaxies was enhanced in the past along with significant gas consumption, and is currently lower than or similar to that of non-barred galaxies. The past star formation enhancement depends on the strength of bars.
79 ASTRONOMY AND ASTROPHYSICS↗
Manufacturing and Hydro Testing of a 10 MWe sCO2 Axial Turbine
Abstract Supercritical CO2 (sCO2) power cycles designed for thermodynamic efficiencies above 50% push component designs into pressure and temperature combinations that approach the limits of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code and ASME Piping Design Codes with Turbine inlet conditions near 250 bar and 715 °C. These design conditions which limit available materials to high nickel alloys are compounded by the high fluid densities of sCO2 at these conditions which result in very compact machinery at the 10 MWe class which are currently being developed for the DOE supercritical transformational electric power (STEP) program. Design of the STEP Turbine is complete with integrated lessons learned from a previous 1 MWth subscale demonstration of a high temperature 10 MWe Turbine designed under the DOE SunShot Program. Fabrication of the STEP Turbine is nearing completion with all housings and pressure containing components in hand. This paper provides a detailed look at design intentions and limitations for the pressurized sections of the main pressure casing and the fabrication of key turbine components. Additional discussion is provided regarding postfabrication inspection and acceptance of the casing for use after issues with casing welds were identified late in the fabrication process.
Engineering↗
H{sub 2}(v = 0,1) + C{sup +}({sup 2} P) {yields} H+CH{sup +} STATE-TO-STATE RATE CONSTANTS FOR CHEMICAL PUMPING MODELS IN ASTROPHYSICAL MEDIA
State-to-state rate constants for the title reaction are calculated using the electronic ground state potential energy surface and an accurate quantum wave-packet method. The calculations are performed for H{sub 2} in different rovibrational states, v = 0, 1 and J = 0 and 1. The simulated reaction cross section for v = 0 shows a rather good agreement with the experimental results of Gerlich et al., both with a threshold of 0.36 eV and within the experimental error of 20%. The total reaction rate coefficients simulated for v = 1 are two times smaller than those estimated by Hierl et al. from cross sections measured at different temperatures and neglecting the contribution from v > 1 with an uncertainty factor of two. Thus, part of the disagreement is attributed to the contributions of v > 1. The computed state-to-state rate coefficients are used in our radiative transfer model code applied to the conditions of the Orion Bar photodissociation region, and leads to an increase of the line fluxes of high-J lines of CH{sup +}. This result partially explains the discrepancies previously found with measurements and demonstrates that CH{sup +} excitation is mostly driven by chemical pumping.
74 ATOMIC AND MOLECULAR PHYSICS↗
Near- and far scrape-off layer transport studies in detached, small-ELM ASDEX Upgrade discharges by means of EMC3-EIRENE
We compare various diagnostics characterizing an ASDEX Upgrade upper single-null discharge to EMC3-EIRENE simulations now including volumetric recombination and main chamber plasma-wall interaction but not yet drifts. The discharge is in a small-ELM regime and is approaching detachment due to a density ramp and nitrogen seeding. Time/ELM averaged upstream density, electron and ion temperature measurements as well as downstream ion saturation current, electron temperature and power flux measurements match the simulation within the error bars before and after reaching detachment, qualifying the code for this regime. Although the Ion Cyclotron Resonance Heating (ICRH) antenna limiters are 35 mm away from the separatrix, i.e. ten times the typical near-SOL power fall-off length $\lambda^{near}_q\ =\ $3.8 mm, they receive power fluxes of several hundreds of kW exceeding the ones to the outer target in the detached phase and strongly contribute to the particle fueling. This is explained by a substantially larger far-SOL power fall-off length $\lambda^{far}_q\ =\ $25 mm possibly caused by the small ELMs in the detached phase and the occurrence of a density shoulder.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
Constitutive modeling for isotropic materials
The objective is to develop a unified constitutive model for finite element structural analysis of turbine engine hot-section components. This effort constitutes a different approach for non-linear finite-element computer codes which have heretofore been based on classical inelastic methods. The unified constitutive theory to be developed will avoid the simplifying assumptions of classical theory and should more accurately represent the behavior of superalloy materials under cyclic loading conditions and high temperature environments. During the first two years of the program, extensive experimental correlations were made with two representative unified models. The experiments were both uniaxial and biaxial at temperatures up to 1093 C (2000 F). In addition, the unified models were adopted to the MARC finite element code and used for stress analysis of notched bar and turbine blade geometries.
Lindholm, Ulric S.↗
Whole core analysis of OECD/NEA TVA-WB1 benchmark exercise 1 using continuous-energy Monte Carlo code MVP with JENDL-4.0u and ENDF/B-VII.0
The TVA WB1 benchmark problem was precisely modeled in accordance with the geometry and material specifications, and benchmark exercise 1 was solved with the MVP code and the nuclear data libraries based on JENDL-4.0u (JENDL-4.0 updated file) and ENDF/B-VII.0. The criticality, the control rod bank worth, the isothermal temperature coefficient (ITC), the integral rod worth of bank D, and differential soluble boron worth (DBW) were obtained. The criticality and the control rod bank worth were compared with the measured data and published calculation results with other codes, Serpent and KENO. The criticality and the control rod bank worth calculated by MVP and the libraries agree with the measured and the published calculation results within the criteria specified in the ANSI/ANS standard. From these results, the calculation model for the MVP and the libraries was verified. The other parameters such as ITC, the integral rod worth, and DBW were also compared between JENDL-4.0u and ENDF/B-VII.0. The comparison of the results shows JENDL-4.0u and ENDF/B-VII.0 give comparable results within the standard deviation, the difference between libraries is small. (author)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Consistent Evaluation of the Prompt-fission Neutron Spectrum and Multiplicity for n+ 235,238 U and n+ 239 Pu
This report was written to satisfy a FY20 NCSP milestone on 235,238 U and 239 Pu. The milestone requires to “finalize a report assessing our methodology to evaluate prompt-fission neutron spectrum (PFNS) and multiplicity consistently”. More specifically, we study whether the code CGMF can reproduce ENDF/B-VIII.0 evaluated PFNS and average prompt-fission neutron multiplicities, $\bar{v}$, for 235,238 U and 239 Pu using one joint parameter set per isotope. If CGMF is shown to be able to reasonably reproduce ENDF/B-VIII.0 within its model-parameter space, this code could be used for future consistent evaluations of PFNS and $\bar{v}$. To answer this question, we explore here the parameter space of CGMF and its impact on calculated values and whether they are close to evaluated and experimental data. We also list experimental data that would enter a future evaluation and statistics method that could be used to obtain evaluated data and covariances. We conclude that values of $\bar{v}$ calculated by CGMF are reasonably close to ENDF/B-VIII.0 data, while more work on modeling the PFNS is needed (parameter optimization and model improvements) to reliably use it for evaluations.
07 ISOTOPE AND RADIATION SOURCES↗
Sub-baud coding
In a spread-spectrum communication system, a data sequence of +1's and -1's is expanded into a sequence of +x's (bottom bar) and -x's (bottom bar) where x (bottom bar) is a length n sequence of +1's and -1's. A sub-baud code is the assignment of a different such baud sequence to each of M users with the objective of reducing interference between users. It is shown that this interference can be described in terms of the even and odd correlation functions between pairs of sequences. A technique for deriving sub-baud codes from cyclic error-correcting codes is given, and a theorem proved which bounds the quality of the resultant code. A specific such cyclic code construction procedure is given which, for any M, yields a sub-baud code for which the mutual interference can be made arbitrarily small by choosing n sufficiently large.
Massey, J. L.↗
MCS solutions for the TVA Watts Bar Unit 1 multi-physics depletion benchmark
The high-fidelity solution for exercises 1-3 of the recently developed TVA Watts Bar Unit 1 multi-physics and multi-cycle depletion benchmark using MCS was coupled with two different thermal-hydraulics (TH) codes: TH1D and COBRA-TF (CTF), and their solutions were compared. The solutions from MCS/CTF and MCS/TH1D for typical Pressurized Water Reactor (PWR) problems are quite close. And to evaluate MCS solutions for the benchmark, their results were also compared against measured data as well as publicly available high-fidelity solutions. The MCS criticality solutions at Hot Zero Power (HZP) deviate within 50 pcm against the reference solutions. While for the critical boron concentration (CBC) search calculations, the difference on the MCS calculated CBC is about 7 ppm, and the root-mean-squared (RMS) errors for the assembly power distribution and outlet coolant temperature are less than 0.5% and 1 Celsius degree respectively. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Erratum to: Measurement of the W boson polarisation in $t\bar{t}$ events from pp collisions at $\sqrt{s} = 8~\hbox {TeV}$ in the lepton + jets channel with ATLAS
A bug in the code used to calculate the total correlation coefficients between the W boson helicity fractions obtained using the leptonic analyser in semileptonic $t\bar{t}$ events in the region with ≥ 2 b -tags was found. It led to quoting incorrect correlation coefficients in the original paper.
72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
Assessment of the Analysis Capability for Core-wide PWR Pellet-Clad Interaction (PCI) Screening
The Level 1 Milestone, L1:CASL.P13.03, “Assess the analysis capability for core-wide PWR Pellet-Clad Interaction (PCI) screening and demonstrate detailed 3-D analysis on selected sub-region.”, has been completed by the Fuels, Materials and Chemistry and Physics Integration focus areas of CASL. This report illustrates the application of VERA simulation to PWR core-wide PCI screening, with emphasis on the multi-dimensional analysis capabilities of BISON fuel performance code. The VERA codes utilized in this effort are: the fuel performance code BISON and the VERA-CS sub-element, which combines the neutron transport code MPACT and the thermal-hydraulics code COBRA-TF (CTF). The target reactor for this work is Watts Bar Unit 1, Cycles 6 and 7, selected on the basis that PCI failures had occurred during power start-up in both cycles, despite the moderate power levels, which offered an interesting test of the code’s veracity and a particularly challenging condition for this Milestone’s assessment. The analysis began by performing BISON-IN-VERA, i.e. VERA-CS + BISON, quarter-core depletion calculations, from which limiting assemblies and limiting fuel rods were selected using a four-parameter selection criterion: high power, high temperature, small gap thickness and high cladding hoop stress. The cladding material being Westinghouse Zirlo, the analysis utilized a new material model developed for this cladding by CASL’s FMC team using open literature sources. The VERA-CS + BISON analysis resulted in identifying 51 limiting rods as candidates for more detailed PCI analysis using stand-alone BISON, employing local 2D and 3D geometric models. This local-effects BISON analysis provided detailed PCI failure assessment, considering the impact of power histories on the two potential sources of PCI failure: the classical SCC-PCI and the missing pellet surface geometry MPS-PCI. The 51-rod analyses of Watts Bar Unit 1 cycle 6 and 7 revealed that the classical SCC-PCI stress levels were well below the failure threshold determined in a previous CASL study, which implicated the other mode of failure, namely MPS-PCI. Considering MPS defects to be the root cause for the failure, the local-effects calculations was repeated using MPS defect sizes equal to and larger than Westinghouse quality assurance standards. Initial results of this new analysis showed a significant increase in the cladding stress, indicating that an MPS may have been present in cycle 6 and 7 for the failure to occur. In order to determine the size of MPS needed for cycle 6 and 7 rods to fail, a sensitivity study using 2-D and 3-D models was performed. The results showed cladding hoop stresses, while remaining below the failure level for an MPS of equal size to the acceptance standard, continued to increase as the size of the MPS increased. This report has demonstrated the use of BISON, in consort with VERA’s simulation methodology, for core-wide PCI failure screening capability. Utilizing this demonstration, and emulating the PCI margin analysis methodology developed in the EPRI PCI guidelines, BISON-IN-VERA simulation can be used very effectively to develop core power histories to minimize PCI failure potential.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
COYOTE
COYOTE is designed for the analysis of steady and transient, multi-dimensional, nonlinear heat conduction problems. The spatial discretization technique is based on the Galerkin finite element method. The code contains a library of linear and quadratic elements and incudes bars and shells. COYOTE Implicit and explicit time integration methods, with adaptive time step control, are available for transient problems and fixed-point methods are used for Conjugate gradient iterative methods are used for the matrix solution. Coupled enclosure radiation with efficient view factor computation is included. Simulation capabilities include material motion, dynamic radiation enclosures, contact boundary conditions, multipoint constraints, material (element) addition, and deletion, bulk nodes, chemically reactive materials, and coupling with other mechanics codes. The code is designed for single or multi-processor computing
Gartling, DavidK.↗
Tensile Properties of Alloy 617 Bar Stock
Alloy 617 is approved for non-nuclear construction in the ASME Boiler and Pressure Vessel Code Section I and Section VIII, Division 1 but is not currently qualified for nuclear use in ASME Code Section III. A draft Code Case was developed in the 1980’s to qualify the alloy for nuclear service but efforts were stopped before the approval process was completed. Renewed interest in high temperature nuclear reactors has resulted in a new effort to qualify Alloy 617 for use in nuclear pressure vessels. Tensile properties of the alloy are needed for several aspects of Code qualification. Minimum yield strength and average tensile strength at temperature are used to set time independent allowable stress, Sm, for structural materials in B&PV Code, Section III, Subsection NH. It must also be demonstrated that there is no minimum ductility between room temperatures and 427°C (the upper temperature limit for Subsection NB) to minimize the number of fatigue tests required for developing the fatigue design curve. The bar properties reported here represent data for an additional heat and product form. A forged and rolled plate, designated as a reference material for the program, has been studied extensively and tensile properties are presented for comparison. Both of the product forms that have been characterized in the current program, plate and bar, were produced using current mill practice that includes additional refining using electro-slag re-melting. This additional processing step was not used for some of the materials in the historical database and it is of interest to demonstrate that the properties of contemporary material are comparable to the historical values.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Initial Implementation of Transient VERA-CS
In this milestone the capabilities of both CTF and MPACT were extended to perform coupled transient calculations. This required several small changes in MPACT to setup the problems correctly, perform the edits correctly, and call the appropriate CTF interfaces in the right order. For CTF, revisions and corrections to the transient timestepping algorithm were made, as well as the addition of a new interface subroutine to allow MPACT to drive CTF at each timestep. With the modifications completed, the initial coupled capability was demonstrated on some problems used for code verification, a hypothetical small mini-core, and a Watts Bar demonstration problem. For each of these cases the results showed good agreement with the previous MPACT internal TH feedback model that relied on a simplified fuel heat conduction model and simplified coolant treatment. After the pulse the results are notably different as expected, where the effects of convection of heat to the coolant can be observed. Areas for future work were discussed, including assessment and development of the CTF dynamic fuel deformation and gap conductance models, addition of suitable transient boiling and CHF models for the rapid heating and cooling rates seen in RIAs, additional validation and demonstration work, and areas for improvement to the code input and output capabilities.
97 MATHEMATICS AND COMPUTING↗