Advanced Reactor Technologies: Gas-Cooled Reactor Third Quarter 2021 Report
Highlights of Advanced Gas Reactor (AGR) fuels development activities during April, May and June 2021.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
SEARCH · Engineering Papers
Results for “validation method”
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.
Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.
Advanced Reactor Technologies: Gas-Cooled Reactor Second Quarter 2021 Report
Highlights and significant accomplishments of Advanced Gas Reactor (AGR) fuels development activities during January, February and March 2021
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
LWR Aging Management for Life Beyond 80
While several decades of materials degradation research enabled the current US fleet of Light Water Reactors (LWR) to plan to produce electrical power through extended operations up to 80 years of plant life, it now seems appropriate to evaluate what issues, methods, and timelines need to be considered to meet the expected electric power demands for life beyond eighty (LBE) years. To address these questions, it will be necessary to review the electrical capacity and the projected LWR fleet capacity including new builds and advanced reactors. This effort should begin with a materials degradation assessment focused on known and possible unknown issues using a reduced Expanded Materials Degradation Assessment [1], which was employed to identify knowledge gaps for a second license renewal (60-80 yrs.). The focus of the LBE assessment should include: Establishing a timeline to initiate assessments of possible materials aging issues and to ensure sufficient time to evaluate materials degradation research gaps; assessing components and materials to be evaluated, e.g., metals (reactor pressure vessels and alloys within and outside the pressure boundary), concrete, cables, as well as mitigation methods, advanced monitoring, and validation with ex-service materials; initiating a LBE research plan focused on developing an expanded mechanistic understanding of materials degradation and refined models through Codes and Standards evaluation for use by the nuclear industry; and maintaining and strengthening the nuclear materials human knowledge base to address new and as yet unknown degradation modes.
Chen, Xiang↗
Thermal Stress Modeling and Analysis of Packed-bed Thermocline Energy Storage Tank for INL Thermal Energy Distribution System (TEDS)
The Thermal Energy Distribution System (TEDS) at Idaho National Laboratory (INL) is a thermal-hydraulic flow loop to support the integration of co-located multiple experimental systems, where a packed-bed thermal energy storage (TES) is installed as a thermal buffer and storage unit for TEDS. The packed-bed TES is adopted in TEDS because of its benefit as a low-cost single-tank storage option compared to the traditional two-tank storage. However, thermal ratcheting is one potential design concern which is caused by the rearrangement of granular filler inside a packed-bed tank during continuous thermal cycling operation of the packed-bed TES tank. If the thermally induced stress exceeds yield strength of the tank wall, it may cause catastrophic consequences like rupture of the thermal storage tank. Thus, it is crucial to understand the phenomenon to ensure the robust operation. Based on the temperature boundary conditions given by transient thermal analyses with computational fluid dynamics (CFD) simulations, the thermal ratcheting analysis is then conducted to evaluate the hoop stress and resultant thermal ratcheting potential of the TES tanks with two different modeling approaches: (1) infinite rigidity model and (2) Drucker-Prager (DP) model. The validity of each modeling method was examined by comparing the numerical simulation with the experimental data obtained from the packed-bed TES tank for Solar One Plant and evaluate the thermal ratcheting potential of the TEDS TES tank.
25 ENERGY STORAGE↗
PWR core and spent fuel pool analysis using scale and nestle
The SCALE nuclear analysis code system [SCALE, 2011], developed and maintained at Oak Ridge National Laboratory (ORNL) is widely recognized as high quality software for analyzing nuclear systems. The SCALE code system is composed of several validated computer codes and methods with standard control sequences, such as the TRITON/NEWT lattice physics sequence, which supplies dependable and accurate analyses for industry, regulators, and academia. Although TRITON generates energy-collapsed and space-homogenized few group cross sections, SCALE does not include a full-core nodal neutron diffusion simulation module within. However, in the past few years, the open-source NESTLE core simulator [NESTLE, 2003], originally developed at North Carolina State Univ. (NCSU), has been updated and upgraded via collaboration between ORNL and the Univ. of Tennessee (UT), so it now has a growingly seamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. This study presents the methodology used to couple lattice physics data between TRITON and NESTLE in order to perform a three-dimensional full-core analysis employing a 'real-life' Duke Energy PWR as the test bed. The focus for this step was to compare the key parameters of core reactivity and radial power distribution versus plant data. Following the core analysis, following a three cycle burn, a spent fuel pool analysis was done using information generated from NESTLE for the discharged bundles and was compared to Duke Energy spent fuel pool models. The KENO control module from SCALE was employed for this latter stage of the project. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
HFE in the design of the socio-technical system of the control room of a new generation reactor: issues faced during multistage validation - 126
This paper deals with the contribution of Human Factors and Ergonomics (HF and E) experts to the design of the control room (CR) of a new generation reactor (NGR), our view of the multistage validation (MSV) process and the issues faced during it. Our approach to MSV was to articulate two points of view: the designers' point of view (test and explore various design options) and the regulatory authorities' point of view (validate the performance of the socio-technical system of the CR). To perform this articulation, both tests and validation shared the same method and goals. The main issues faced during this process were to (1) get a coherent and integrated version of the full-scope simulator when the design of the socio-technical system of the CR was still in progress, (2) determine when to stop MSVs, and (3) define a criterion for the sampling of operational conditions. The first issue led us to the conclusion that specific project milestones synchronising the design of every part of the CR have to be planned early in the project in order to mitigate the risk of delaying WSTs (Whole System Tests). Regarding the second issue, we advocate that no further WSTs are required when the last modifications brought to the CR design have no significant impact on crews' activities. Finally, concerning the last issue, we think that performing WSTs in all classes of situations a crew may face during operating the plant is necessary. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
An in-line burn-up algorithm: proposal and preliminary analysis
A new burn-up algorithm is proposed for coupling neutronics and isotopic depletion. The algorithm is analogous to those proposed for enforcing xenon equilibrium and converging neutronics and thermal-hydraulics. During the corrector step of standard predictor-corrector methods, the corrector nuclide density is updated as neutronics converges. This is expected to improve stability while reducing computational expense compared to other implicit algorithms. The stability of the algorithm is investigated by Fourier analysis for a simple case whose results have been successfully compared with numerical experiments.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
A Brief Survey on High Performance Computing Systems Power Management
This paper provides a survey of software-based power management techniques in High Performance Computing (HPC) systems. Seven existing power management and monitoring tools and frameworks are discussed. These are: Variorum, dynamic energy-performance optimizer (DEPO), Powersched, Bull Dynamic Power Optimizer (BDPO), Energy Aware Runtime (EAR), Global Extensible Open Power Manager (GEOPM), and PoLiMEr. Each of these tools is evaluated based on hardware abstraction, optimization methods, usability, and experimental validation. This survey highlights the diversity of approaches in managing energy efficiency, from vendor-neutral APIs to algorithm-driven power capping, and dynamic frequency adjustments. Given that energy requirements for large computational systems is increasing quickly, the importance of integrating these tools into existing HPC environments and the need for further research in this rapidly evolving field is also discussed.
97 - MATHEMATICS AND COMPUTING↗
Status of the Down-Blending of Irradiated IGR Graphite Fuel in Kazakhstan
Dry processing remains the primary solution for managing the down-blending of irradiated IGR HEU fuel. Several tasks have been initiated and completed to achieve this, including lab modernization, lab-scale and full-scale testing, designing and fabricating an in-paddle mixing drum, and commissioning both a down-blending system and a crushing and milling system. Additionally, a testing and training center was modernized, the equipment was successfully commissioned, and the IGR HEU fuel blocks were repackaged into daily containers. The training of operators and the construction of a new down-blending facility are tasks that will need to be completed in the near future. The progress in these areas continues to validate that the proposed method for down-blending irradiated HEU graphite fuel, followed by cementation of the down-blended material for permanent disposition, appears to be achievable.
12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗
Hadron-Argon Cross-Section Measurements at Protodune
The Deep Underground Neutrino Experiment (DUNE) is a next-generation experiment designed to measure neutrino oscillations with unprecedented precision, specifically seeking to identify CP violation in the lepton sector. DUNE utilizes the Liquid Argon Time Projection Chamber (LArTPC) as its core detecting technology, offering high-resolution imaging to reconstruct neutrino interactions on argon. A primary challenge in this reconstruction is the modeling of hadronic final-state interactions (FSI), where hadrons produced at the initial interaction vertex undergo further scattering before escaping the nucleus. In this thesis, I utilize the ProtoDUNE-SP detector and its 1~GeV/$c$ hadron beam at the CERN Neutrino Platform to present the first measurements of total inelastic $\pi^+$-argon and proton-argon cross sections in an energy regime critical to DUNE. I detail the development and validation of the ``slicing method'', which utilizes a kiloton-scale LArTPC to extract hadron-argon cross sections. These measurements provide an indispensable benchmark for informing DUNE's FSI modeling, and our results show no significant tension with current model predictions. Additionally, I study the impact of FSI modeling on DUNE’s sensitivity to oscillation parameters. This study demonstrates that variations in FSI models may be degenerate with oscillation features, highlighting the necessity of using direct experimental data to refine interaction models. Finally, I provide an outlook on related studies that will further contribute to this effort, supporting the ambitious physics goals of upcoming neutrino oscillation programs including DUNE.
Yin-Rui, Liu [U. Chicago (main)]↗
Increased fissionable material loading of vitrified borosilicate glass in support of efficient disposition of nuclear materials at Savannah River - 15299
The US DOE is vitrifying liquid HLW, converting it into a glass form suitable for long term storage and disposal. The waste includes fissionable isotopes and, as a consequence, may possess an inadvertent criticality hazard. The technical basis for determining when an inadvertent criticality hazard exists is unclear. ANS criticality safety standards do not provide relevant subcritical safety limits and little experimental data exist to support development of such a limit. As a result, the DOE has implemented a conservative limit of 897 grams of fissile nuclides per cubic meter of glass. However, this limit has the potential to impede efficient disposal of various fissionable materials at the Savannah River Site. Work is being conducted to establish an alternate fissionable material limit which would be less restrictive, but still ensure criticality safety. A series of computer models are constructed and analyzed using a general purpose Monte Carlo N-Particle neutron transport code. First, a baseline model is established and analyzed. Next, numerous variations to the baseline model (i.e., perturbations) are analyzed with the code to understand the impact of the change. Based on the results, it is evident the atomic ratio of B-10 to equivalent Pu-239 would be an effective means to ensure subcriticality of HLW vitrified in a borosilicate glass matrix. The proposed limit, its technical basis, and all necessary application constraints are discussed. This proposed limit is expected to permit significantly higher quantities of fissile nuclides in the vitrified waste form. It will also account for several non-fissile, fissionable nuclides which are expected to add reactivity to the glass waste form. The proposed limit embodies large conservatisms intended to provide assurance of safety in spite of the significant uncertainties associated with its development. The single most important uncertainty arises from the inability to validate the calculations which form its technical basis. While they are done with the latest available nuclear data and well established analytical methods, they are, nonetheless, not validated as required for subcritical safety limits. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Validation and Verification of Multi-Material Split Rings
Process induced residual stresses commonly occur in composite structures composed of dissimilar materials. These residual stresses form due to differences in the composite materials coefficients of thermal expansion as well as the shrinkage upon cure exhibited by most thermoset polymer matrix materials. Depending upon the specific geometric details of the composite structure and the materials curing parameters, it is possible that these residual stresses can result in interlaminar delamination and fracture within the composite as well as plastic deformation in the structures metallic materials. It is important to consider potential residual stresses when designing composite parts and their manufacturing processes. However, the experimental determination of residual stresses in prototype parts can be prohibitive, both in terms of financial and temporal costs. As an alternative to physical measurement, it is possible for computational tools to be used to quantify potential residual stresses in composite prototype parts. A simplified method for simulating residual stresses was previously validated with two simple bi-material structures. Continuing on, the objective of this study is to further validate the simplified method for simulating residual stresses for bi-material split rings of different composites and layup variations. The validation process uses uncertainty quantification to develop a distribution of possible simulated residual stress states that are compared to experimentally measured residual stress states of fabricated structures similar to those simulated. The results of the comparisons indicate that the proposed finite element modeling approach is capable of accurately simulating the formation of residual stresses in composite structures and a temperature independent material model is adequate within the composites glassy region.
36 MATERIALS SCIENCE↗
Inversion-based correction of Double-Torsion (DT) subcritical crack growth tests for crack profile geometry
Because of its simplicity and the ability to produce a stable, slow-propagating crack, the Double-Torsion (DT) method has been used widely for investigating the critical and subcritical propagation of a slow-propagating tensile (mode-I) crack. However, to determine the complex relationship between the crack velocity $\mathcal{v_c}$ vs. the strain energy release rate $\mathscr{G}$ (or the stress intensity factor K) from laboratory measurements, several corrections must be made to account for the impact of sample and crack geometry. Particularly, DT test typically produces a crack with a curved edge profile instead of a straight line, causing the local $\mathcal{v_c}$ and $\mathscr{G}$ vary along the crack front. The experimentally measured $\mathcal{v_c}$ and $\mathscr{G}$ data merely reflect collective, averaged behavior of the crack. This makes inversion for the intrinsic, “true” crack growth kinetics necessary, based upon the knowledge of the crack geometry. Simple and effective correction methods have been proposed and validated for the slow, chemical-reaction-controlled part (Region I) of the $\mathcal{v_c}$-$\mathscr{G}$ curve. However, reliable methods for the highly nonlinear, transport-dominated part (Region II) and its sudden transition to the dynamic propagation part (Region III) are still lacking. Here we propose a method for determining the intrinsic $\mathcal{v_c}$-$\mathscr{G}$ relationship across all three Regions based upon DT test data, using a simple model function and its numerical inversion. The performance of this approach is examined and demonstrated using both synthetic and laboratory data for subcritical crack growth in soda lime glass.
42 ENGINEERING↗
TokaMaker: An open-source time-dependent Grad-Shafranov tool for the design and modeling of axisymmetric fusion devices
Here, in this paper, we present a new static and time-dependent MagnetoHydroDynamic (MHD) equilibrium code, TokaMaker, for axisymmetric configurations of magnetized plasmas, based on the well-known Grad-Shafranov equation. This code utilizes finite element methods on an unstructured triangular grid to enable capturing accurate machine geometry and simple mesh generation from engineering-like descriptions of present and future devices. The new code is designed for ease of use without sacrificing capability and speed through a combination of Python, Fortran, and C/C++ components. A detailed description of the numerical methods of the code, including a novel formulation of the boundary conditions for free-boundary equilibria, and validation of the implementation of those methods using both analytic test cases and cross-code validation is shown. Results show expected convergence across tested polynomial degree for analytic and cross-code test cases.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
Experimental depth dose curves of a 67.5 MeV proton beam for benchmarking and validation of Monte Carlo simulation
Purpose: To measure depth dose curves for a 67.5 ± 0.1 MeV proton beam for benchmarking and validation of Monte Carlo simulation. Methods: Depth dose curves were measured in 2 beam lines. Protons in the raw beam line traversed a Ta scattering foil, 0.1016 or 0.381 mm thick, a secondary emission monitor comprised of thin Al foils, and a thin Kapton exit window. The beam energy and peak width and the composition and density of material traversed by the beam were known with sufficient accuracy to permit benchmark quality measurements. Diodes for charged particle dosimetry from two different manufacturers were used to scan the depth dose curves with 0.003 mm depth reproducibility in a water tank placed 300 mm from the exit window. Depth in water was determined with an uncertainty of 0.15 mm, including the uncertainty in the water equivalent depth of the sensitive volume of the detector. Parallel-plate chambers were used to verify the accuracy of the shape of the Bragg peak and the peak-to-plateau ratio measured with the diodes. The uncertainty in the measured peak-to-plateau ratio was 4%. Depth dose curves were also measured with a diode for a Bragg curve and treatment beam spread out Bragg peak (SOBP) on the beam line used for eye treatment. The measurements were compared to Monte Carlo simulation done with GEANT4 using TOPAS. Results: The 80% dose at the distal side of the Bragg peak for the thinner foil was at 37.47 ± 0.11 mm (average of measurement with diodes from two different manufacturers), compared to the simulated value of 37.20 mm. The 80% dose for the thicker foil was at 35.08 ± 0.15 mm, compared to the simulated value of 34.90 mm. The measured peak-to-plateau ratio was within one standard deviation experimental uncertainty of the simulated result for the thinnest foil and two standard deviations for the thickest foil. It was necessary to include the collimation in the simulation, which had a more pronounced effect on the peak-to-plateau ratio for the thicker foil. The treatment beam, being unfocussed, had a broader Bragg peak than the raw beam. A 1.3 ± 0.1 MeV FWHM peak width in the energy distribution was used in the simulation to match the Bragg peak width. An additional 1.3–2.24 mm of water in the water column was required over the nominal values to match the measured depth penetration. Conclusions: The proton Bragg curve measured for the 0.1016 mm thick Ta foil provided the most accurate benchmark, having a low contribution of proton scatter from upstream of the water tank. The accuracy was 0.15% in measured beam energy and 0.3% in measured depth penetration at the Bragg peak. The depth of the distal edge of the Bragg peak in the simulation fell short of measurement, suggesting that the mean ionization potential of water is 2–5 eV higher than the 78 eV used in the stopping power calculation for the simulation. The eye treatment beam line depth dose curves provide validation of Monte Carlo simulation of a Bragg curve and SOBP with 4%/2 mm accuracy.
60 APPLIED LIFE SCIENCES↗
Active multi-mode data analysis to improve fault diagnosis in AHUs
Faults in heating, ventilation and air conditioning systems can lead to increased energy consumption, occupant comfort issues, and reduced equipment lifetime. Commercial fault detection and diagnosis (FDD) tools has been increasingly deployed in U.S. commercial buildings. While they are helping to achieve energy efficiency and operational reliability, there remain gaps in their fault diagnostic capabilities. The diagnostic results often contain multiple distinct candidate root causes (CRCs) or offer no insight into CRCs. This study developed a novel active rule-based multi-mode data analysis method to enhance diagnostic resolution by applying proven rule sets and additional new rules to data from multiple known operational modes. The proposed method was demonstrated using enhanced air handling unit performance assessment rule sets and validated with the simulated data of two air handling units. New metrics, namely, reduced number of CRCs and improvement ratio, were developed to quantify the improvement of fault diagnostic resolution. The validation results showed that the proposed method effectively reduced the number of CRCs in contrast to analyzing data solely for a single mode of operation. It achieved a median improvement ratio of 80% in 19 test cases.
32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
Power Decoupling Method for Voltage Source Inverters Using Grid Voltage Modulated Direct Power Control in Unbalanced System
The grid voltage modulated direct power control (GVM-DPC) is a novel power regulation method for grid-connected voltage source inverters (VSI). However, interactions between real and reactive powers still exist in GVM-DPC under unbalanced grid voltage condition, which may deteriorate its transient performance. Here, this article proposes a power decoupling method for GVM-DPC-based VSI in unbalanced systems based on the dynamic feedforward power compensation strategy, which can reduce the coupling magnitudes of real and reactive powers in the transient stages. First, the power coupling mechanism of positive and negative sequence real and reactive powers of VSI under the unbalanced voltage condition is analyzed. Then, the power coupling magnitudes (PCM) are derived according to relationships among the positive sequence components of real power, reactive power and point of common coupling (PCC) voltages. Next, the PCM are compensated into the GVM-DPC for the better decoupling performance. Furthermore, a stability analysis of the proposed control system is studied based on the impedance model, and a comparison with the traditional power decoupling method based on virtual impedance is conducted to show the superiority of the proposed method. Finally, simulations, hardware-in-loop test and experiment are conducted to validate the effectiveness of the proposed method.
42 ENGINEERING↗
Theoretical investigation of resonant frequencies of unstrapped magnetron with arbitrary side resonators
In this paper, a sector steps approximation method is proposed to investigate the resonant frequencies of magnetrons with arbitrary side resonators. The arbitrary side resonator is substituted with a series of sector steps, in which the spatial harmonics of electromagnetic field are also considered. By using the method of admittance matching between adjacent steps, as well as field continuity conditions between side resonators and interaction regions, the dispersion equation of magnetron with arbitrary side resonators is derived. Resonant frequencies of magnetrons with five common kinds of side resonators are calculated with sector steps approximation method and computer simulation softwares, in which the results have a good agreement. The relative error is less than 2%, which verifies the validity of sector steps approximation method.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗