Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “benchmark testing”

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.

At least 775 records · Page 43
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Outcomes from a Collaborative Approach to a Code Comparison Study for Enhanced Geothermal Systems

A code comparison study, supported by the U.S. Department of Energy, Geothermal Technologies Office, has recently been completed whose stated purpose was to test, diagnose differences, and demonstrate the state of modeling capabilities developed at national laboratories and academic institutes within the United States for modeling enhanced geothermal systems. This three-year study comprised two sets of problems; those classified as benchmark problems and those classified as challenge problems. The benchmark problems were structured to test the ability of the collection of numerical simulators to solve various combinations of coupled thermal, hydrologic, geomechanical, and geochemical processes. This class of problems was strictly defined in terms of properties, driving forces, initial conditions, and boundary conditions. Study participants submitted solutions to problems for which their simulation tools were deemed capable or nearly capable. Some participating codes were originally developed for EGS applications whereas some others were designed for different applications but can simulate processes similar to those in EGS. Solution submissions from both were encouraged. In some cases participants made small incremental changes to their numerical simulation codes to address specific elements of the problem, and in other cases participants submitted solutions with existing simulation tools, acknowledging the limitations of the code. The challenge problems were based on the enhanced geothermal systems research conducted in hot dry rock at Fenton Hill, near Los Alamos, New Mexico, between 1974 and 1995, covering two research stimulation, development and circulation phases in two separate reservoirs. Both challenge problems have specific questions to be answered via numerical simulation in three topical areas: 1) reservoir creation/stimulation, 2) reactive and passive transport, and 3) thermal recovery. Whereas the benchmark class of problems were designed to test capabilities for modeling coupled processes under strictly specified conditions, the stated objective for the challenge class of problems was to demonstrate what new understanding of the Fenton Hill experiments could be realized via the application of modern numerical simulation tools by recognized expert practitioners. Critical observations and data from the Fenton Hill experiments were varied and scattered among a number of individual experiments, and numerical simulation solutions were sought that satisfied as many observations at possible. Achieving agreement in multiple experimental observations from Fenton Hill often required participants to critically think about stimulation mechanisms involving natural and hydraulic fractures and re-evaluate their conceptual models and numerical solution approaches. This process has yielded new insights to Fenton Hill reservoirs and direction for future EGS research. The keys to the success of this study were the high quality numerical simulation tools available, the expertise of the developers and practitioners, and the collaborative approach to reporting preliminary results among the study participants. This paper describes results submitted to the challenge problems and the outcomes of a collaborative approach to conducting a code comparison study.

Enhanced Geothermal Systems, numerical simulation,↗
DOE OSTI2017

Benchmark Credentialing Results for NRG-BR001: The First National Cancer Institute-Sponsored Trial of Stereotactic Body Radiation Therapy for Multiple Metastases

Purpose: The NRG-BR001 trial is the first National Cancer Institute–sponsored trial to treat multiple (range 2-4) extracranial metastases with stereotactic body radiation therapy. Benchmark credentialing is required to ensure adherence to this complex protocol, in particular, for metastases in close proximity. The present report summarizes the dosimetric results and approval rates. Methods and Materials: The benchmark used anonymized data from a patient with bilateral adrenal metastases, separated by <5 cm of normal tissue. Because the planning target volume (PTV) overlaps with organs at risk (OARs), institutions must use the planning priority guidelines to balance PTV coverage (45 Gy in 3 fractions) against OAR sparing. Submitted plans were processed by the Imaging and Radiation Oncology Core and assessed by the protocol co-chairs by comparing the doses to targets, OARs, and conformity metrics using nonparametric tests. Results: Of 63 benchmarks submitted through October 2015, 94% were approved, with 51% approved at the first attempt. Most used volumetric arc therapy (VMAT) (78%), a single plan for both PTVs (90%), and prioritized the PTV over the stomach (75%). The median dose to 95% of the volume was 44.8 ± 1.0 Gy and 44.9 ± 1.0 Gy for the right and left PTV, respectively. The median dose to 0.03 cm{sup 3} was 14.2 ± 2.2 Gy to the spinal cord and 46.5 ± 3.1 Gy to the stomach. Plans that spared the stomach significantly reduced the dose to the left PTV and stomach. Conformity metrics were significantly better for single plans that simultaneously treated both PTVs with VMAT, intensity modulated radiation therapy, or 3-dimensional conformal radiation therapy compared with separate plans. No significant differences existed in the dose at 2 cm from the PTVs. Conclusions: Although most plans used VMAT, the range of conformity and dose falloff was large. The decision to prioritize either OARs or PTV coverage varied considerably, suggesting that the toxicity outcomes in the trial could be affected. Several benchmarks met the dose-volume histogram metrics but produced unacceptable plans owing to low conformity. Dissemination of a frequently-asked-questions document improved the approval rate at the first attempt. Benchmark credentialing was found to be a valuable tool for educating institutions about the protocol requirements.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2013

Verification & Validation of High-Order Short-Characteristics-Based Deterministic Transport Methodology on Unstructured Grids

The research team has developed a practical, high-order, discrete-ordinates, short characteristics neutron transport code for three-dimensional configurations represented on unstructured tetrahedral grids that can be used for realistic reactor physics applications at both the assembly and core levels. This project will perform a comprehensive verification and validation of this new computational tool against both a continuous-energy Monte Carlo simulation (e.g. MCNP) and experimentally measured data, an essential prerequisite for its deployment in reactor core modeling. Verification is divided into three phases. The team will first conduct spatial mesh and expansion order refinement studies to monitor convergence of the numerical solution to reference solutions. This is quantified by convergence rates that are based on integral error norms computed from the cell-by-cell difference between the code’s numerical solution and its reference counterpart. The latter is either analytic or very fine- mesh numerical solutions from independent computational tools. For the second phase, the team will create a suite of code-independent benchmark configurations to enable testing the theoretical order of accuracy of any particular discretization of the discrete ordinates approximation of the transport equation. For each tested case (i.e. mesh and spatial approximation order), researchers will execute the code and compare the resulting numerical solution to the exact solution on a per cell basis to determine the distribution of the numerical error. The final activity comprises a comparison to continuous-energy Monte Carlo solutions for zero-power critical configuration measurements at Idaho National Laboratory’s Advanced Test Reactor (ATR). Results of this comparison will allow the investigators to distinguish between modeling errors and the above-listed discretization errors introduced by the deterministic method, and to separate the sources of uncertainty.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

WE-A-17A-08: Evaluation of the OncentraBrachy Collapsed Cone Convolution Algorithm for Ir-192 Source Using Phantom and Real-Patient Heterogeneous Geometries

Purpose: To evaluate the commercially released Collapsed Cone convolution-based(CCC) dose calculation module of the Elekta OncentraBrachy(OcB) treatment planning system(TPS). Methods: An allwater phantom was used to perform TG43 benchmarks with single source and seventeen sources, separately. Furthermore, four real-patient heterogeneous geometries (chestwall, lung, breast and prostate) were used. They were selected based on their clinical representativity of a class of clinical anatomies that pose clear challenges. The plans were used as is(no modification). For each case, TG43 and CCC calculations were performed in the OcB TPS, with TG186-recommended materials properly assigned to ROIs. For comparison, Monte Carlo simulation was run for each case with the same material scheme and grid mesh as TPS calculations. Both modes of CCC (standard and high quality) were tested. Results: For the benchmark case, the CCC dose, when divided by that of TG43, yields hot-n-cold spots in a radial pattern. The pattern of the high mode is denser than that of the standard mode and is representative of angular dicretization. The total deviation ((hot-cold)/TG43) is 18% for standard mode and 11% for high mode. Seventeen dwell positions help to reduce “ray-effect”, with the total deviation to 6% (standard) and 5% (high), respectively. For the four patient cases, CCC produces, as expected, more realistic dose distributions than TG43. A close agreement was observed between CCC and MC for all isodose lines, from 20% and up; the 10% isodose line of CCC appears shifted compared to that of MC. The DVH plots show dose deviations of CCC from MC in small volume, high dose regions (>100% isodose). For patient cases, the difference between standard and high modes is almost undiscernable. Conclusion: OncentraBrachy CCC algorithm marks a significant dosimetry improvement relative to TG43 in real-patient cases. Further researches are recommended regarding the clinical implications of the above observations. Support provided by a CIHR grant and CCC system provided by Elekta-Nucletron.

60 APPLIED LIFE SCIENCES↗
NASA NTRS2007

The X-43A Flight Research Program: Lessons Learned on the Road to Mach 10

During an aerospace engineer's undergraduate studies, he or she will attend classes in aerodynamics, thermodynamics, structures, stability and control, dynamics, design, propulsion, and computer science, along with the related courses in mathematics, physics, statistics, and chemistry required to understand the material. Upon graduation, the new engineer will have acquired a basic knowledge of how to build an aerospace vehicle. What only comes through experience, however, is the understanding of the inevitable imperfect process through which an aerospace vehicle is built. This is the adventure of turning a basic concept into functional hardware. Engineers working on a project must often deal with ambiguous situations. They are routinely asked by management to provide risk assessments of a project, yet even after careful analysis uncertainties remain. The project must be accomplished within finite limits of time and money. The question an engineer answers is whether the solution to potential problem is worth the cost and schedule delay, or if the solution might actually be worse than the problem it is meant to solve. Review protocols are established to ensure that an unknown has not been overlooked. But these cannot protect against an unknown unknown. Examples of these situations can be found in the history of the X-43A Hyper-X (Hypersonic Experiment) program. In this NASA project, a supersonic combustion ramjet (scramjet) engine was flight tested on a subscale vehicle. The X-43A Hyper-X Research Vehicle (HXRV) was launched from a B-52B mothership, then boosted to the test speed by a modified Pegasus rocket first stage, called the Hyper-X Launch Vehicle (HXLV). Once at the proper speed and altitude, the X-43A separated from the booster, stabilized itself, and then the engine test began. Although wind-tunnel scramjet engine tests had begun in the late 1950s, before the Hyper-X program there had never been an actual in-flight test of such an engine integrated with an appropriate airframe. Thus, while the scramjet had successfully operated in the artificial airflow of wind tunnels, the concept had yet to be proven in real air. These conditions meant changes in density and temperature, as well as changes in angle of attack and sideslip of a free-flying vehicle. A wind tunnel is limited in its ability to simulate these subtle factures, which have a major impact on almost any vehicle, but especially that of a scramjet's performance. The Hyper-X project was to provide a real-world benchmark of the ground test data. The full scale X-43A engine would be operated in the wind tunnel, and then flown, and the data from its operation would then be compared with projections. If these matched, the wind tunnel data would be considered a reliable design tool for future scramjet. If there were significant differences, the reasons for these would have to be identified. Until such information was available, scramjets would lack the technological maturity to be considered for future space launch or high-speed atmospheric flight vehicles.

Peebles, Curtis↗
DOE OSTI2010

ZPR-3 Assembly 11 : A cylindrical sssembly of highly enriched uranium and depleted uranium with an average {sup 235}U enrichment of 12 atom % and a depleted uranium reflector.

Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 11 (ZPR-3/11) was designed as a fast reactor physics benchmark experiment with an average core {sup 235}U enrichment of approximately 12 at.% and a depleted uranium reflector. Approximately 79.7% of the total fissions in this assembly occur above 100 keV, approximately 20.3% occur below 100 keV, and essentially none below 0.625 eV - thus the classification as a 'fast' assembly. This assembly is Fast Reactor Benchmark No. 8 in the Cross Section Evaluation Working Group (CSEWG) Benchmark Specificationsa and has historically been used as a data validation benchmark assembly. Loading of ZPR-3 Assembly 11 began in early January 1958, and the Assembly 11 program ended in late January 1958. The core consisted of highly enriched uranium (HEU) plates and depleted uranium plates loaded into stainless steel drawers, which were inserted into the central square stainless steel tubes of a 31 x 31 matrix on a split table machine. The core unit cell consisted of two columns of 0.125 in.-wide (3.175 mm) HEU plates, six columns of 0.125 in.-wide (3.175 mm) depleted uranium plates and one column of 1.0 in.-wide (25.4 mm) depleted uranium plates. The length of each column was 10 in. (254.0 mm) in each half of the core. The axial blanket consisted of 12 in. (304.8 mm) of depleted uranium behind the core. The thickness of the depleted uranium radial blanket was approximately 14 in. (355.6 mm), and the length of the radial blanket in each half of the matrix was 22 in. (558.8 mm). The assembly geometry approximated a right circular cylinder as closely as the square matrix tubes allowed. According to the logbook and loading records for ZPR-3/11, the reference critical configuration was loading 10 which was critical on January 21, 1958. Subsequent loadings were very similar but less clean for criticality because there were modifications made to accommodate reactor physics measurements other than criticality. Accordingly, ZPR-3/11 loading 10 was selected as the only configuration for this benchmark. As documented below, it was determined to be acceptable as a criticality safety benchmark experiment. A very accurate transformation to a simplified model is needed to make any ZPR assembly a practical criticality-safety benchmark. There is simply too much geometric detail in an exact (as-built) model of a ZPR assembly, even a clean core such as ZPR-3/11 loading 10. The transformation must reduce the detail to a practical level without masking any of the important features of the critical experiment. And it must do this without increasing the total uncertainty far beyond that of the original experiment. Such a transformation is described in Section 3. It was obtained using a pair of continuous-energy Monte Carlo calculations. First, the critical configuration was modeled in full detail - every plate, drawer, matrix tube, and air gap was modeled explicitly. Then the regionwise compositions and volumes from the detailed as-built model were used to construct a homogeneous, two-dimensional (RZ) model of ZPR-3/11 that conserved the mass of each nuclide and volume of each region. The simple cylindrical model is the criticality-safety benchmark model. The difference in the calculated k{sub eff} values between the as-built three-dimensional model and the homogeneous two-dimensional benchmark model was used to adjust the measured excess reactivity of ZPR-3/11 loading 10 to obtain the k{sub eff} for the benchmark model.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2010

ZPR-3 Assembly 12 : A cylindrical assembly of highly enriched uranium, depleted uranium and graphite with an average {sup 235}U enrichment of 21 atom %.

Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 12 (ZPR-3/12) was designed as a fast reactor physics benchmark experiment with an average core {sup 235}U enrichment of approximately 21 at.%. Approximately 68.9% of the total fissions in this assembly occur above 100 keV, approximately 31.1% occur below 100 keV, and essentially none below 0.625 eV - thus the classification as a 'fast' assembly. This assembly is Fast Reactor Benchmark No. 9 in the Cross Section Evaluation Working Group (CSEWG) Benchmark Specifications and has historically been used as a data validation benchmark assembly. Loading of ZPR-3 Assembly 12 began in late Jan. 1958, and the Assembly 12 program ended in Feb. 1958. The core consisted of highly enriched uranium (HEU) plates, depleted uranium plates and graphite plates loaded into stainless steel drawers which were inserted into the central square stainless steel tubes of a 31 x 31 matrix on a split table machine. The core unit cell consisted of two columns of 0.125 in.-wide (3.175 mm) HEU plates, seven columns of 0.125 in.-wide depleted uranium plates and seven columns of 0.125 in.-wide graphite plates. The length of each column was 9 in. (228.6 mm) in each half of the core. The graphite plates were included to produce a softer neutron spectrum that would be more characteristic of a large power reactor. The axial blanket consisted of 12 in. (304.8 mm) of depleted uranium behind the core. The thickness of the radial blanket was approximately 12 in. and the length of the radial blanket in each half of the matrix was 21 in. (533.4 mm). The assembly geometry approximated a right circular cylinder as closely as the square matrix tubes allowed. According to the logbook and loading records for ZPR-3/12, the reference critical configuration was loading 10 which was critical on Feb. 5, 1958. The subsequent loadings were very similar but less clean for criticality because there were modifications made to accommodate reactor physics measurements other than criticality. Accordingly, ZPR-3/12 loading 10 was selected as the only configuration for this benchmark. As documented below, it was determined to be acceptable as a criticality safety benchmark experiment. An accurate transformation to a simplified model is needed to make any ZPR assembly a practical criticality-safety benchmark. There is simply too much geometric detail in an exact (as-built) model of a ZPR assembly, even a clean core such as ZPR-3/12 loading 10. The transformation must reduce the detail to a practical level without masking any of the important features of the critical experiment. And it must do this without increasing the total uncertainty far beyond that of the original experiment. Such a transformation is described in Section 3. It was obtained using a pair of continuous-energy Monte Carlo calculations. First, the critical configuration was modeled in full detail - every plate, drawer, matrix tube, and air gap was modeled explicitly. Then the regionwise compositions and volumes from the detailed as-built model were used to construct a homogeneous, two-dimensional (RZ) model of ZPR-3/12 that conserved the mass of each nuclide and volume of each region. The simple model is the criticality-safety benchmark model. The difference in the calculated k{sub eff} values between the as-built 3-dimensional model and the homogeneous 2-dimensional benchmark model was used to adjust the measured excess reactivity of ZPR-3/12 loading 10 to obtain the k{sub eff} for the benchmark model.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2010

ZPR-3 Assembly 6F : A spherical assembly of highly enriched uranium, depleted uranium, aluminum and steel with an average {sup 235}U enrichment of 47 atom %.

Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 6 consisted of six phases, A through F. In each phase a critical configuration was constructed to simulate a very simple shape such as a slab, cylinder or sphere that could be analyzed with the limited analytical tools available in the 1950s. In each case the configuration consisted of a core region of metal plates surrounded by a thick depleted uranium metal reflector. The average compositions of the core configurations were essentially identical in phases A - F. ZPR-3 Assembly 6F (ZPR-3/6F), the final phase of the Assembly 6 program, simulated a spherical core with a thick depleted uranium reflector. ZPR-3/6F was designed as a fast reactor physics benchmark experiment with an average core {sup 235}U enrichment of approximately 47 at.%. Approximately 81.4% of the total fissions in this assembly occur above 100 keV, approximately 18.6% occur below 100 keV, and essentially none below 0.625 eV - thus the classification as a 'fast' assembly. This assembly is Fast Reactor Benchmark No. 7 in the Cross Section Evaluation Working Group (CSEWG) Benchmark Specifications and has historically been used as a data validation benchmark assembly. Loading of ZPR-3/6F began in late December 1956, and the experimental measurements were performed in January 1957. The core consisted of highly enriched uranium (HEU) plates, depleted uranium plates, perforated aluminum plates and stainless steel plates loaded into aluminum drawers, which were inserted into the central square stainless steel tubes of a 31 x 31 matrix on a split table machine. The core unit cell consisted of three columns of 0.125 in.-wide (3.175 mm) HEU plates, three columns of 0.125 in.-wide depleted uranium plates, nine columns of 0.125 in.-wide perforated aluminum plates and one column of stainless steel plates. The maximum length of each column of core material in a drawer was 9 in. (228.6 mm). Because of the goal to produce an approximately spherical core, core fuel and diluent column lengths generally varied between adjacent drawers and frequently within an individual drawer. The axial reflector consisted of depleted uranium plates and blocks loaded in the available space in the front (core) drawers, with the remainder loaded into back drawers behind the front drawers. The radial reflector consisted of blocks of depleted uranium loaded directly into the matrix tubes. The assembly geometry approximated a reflected sphere as closely as the square matrix tubes, the drawers and the shapes of fuel and diluent plates allowed. According to the logbook and loading records for ZPR-3/6F, the reference critical configuration was loading 5 which was critical on January 4, 1957. The subsequent loadings were very similar but were less clean for criticality because there were modifications made to accommodate reactor physics measurements other than criticality. Accordingly, ZPR-3/6F loading 5 was selected as the only configuration for this benchmark. As documented below, it was determined to be acceptable as a criticality safety benchmark experiment. A very accurate transformation to a simplified model is needed to make any ZPR assembly a practical criticality-safety benchmark. There is simply too much geometric detail in an exact (as-built) model of a ZPR assembly. This is especially true of ZPR-3/6F because of the complex core loading required to approximate a sphere with rectangular plates in a square matrix.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

AGR-1 Depletion Benchmark

The AGR-1 TRISO particle irradiation test was the first Advanced Gas Reactor test irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory. The TRISO particle fuel was 350 μm diameter uranium oxy-carbide (UCO) kernels enriched to 19.75 wt% U-235. The particles were embedded in cylindrical graphite compacts, placed in test capsules, and irradiated over a period of three years or 13 ATR power cycles. The fuel compacts achieved burnups approaching 20% FIMA and accumulated fast fluence up to 4.25E+25 n/m 2 (E n >0.18 MeV). This report describes the AGR-1 test train, capsules, compacts, and TRISO particle fuel in sufficient detail for a physics analyst to model the AGR-1 experiment as part of a larger ATR neutron transport model and perform a depletion calculation. The intent of the data herein is to define a practical numerical TRISO particle depletion benchmark problem for the AGR 1 irradiation test.

42 ENGINEERING↗
DOE OSTI2022

INL Contributions to Draft HTTF Benchmark Specifications

The High-Temperature Test Facility (HTTF) is an integral effects thermal hydraulics test facility at Oregon State University designed as a 1/4 length scale model of the Modular High-Temperature Gas-Cooled Reactor 350 MW core (mHTGR-350). In the spring and summer of 2019, several experiments were conducted at HTTF providing a valuable source of gas-cooled reactor thermal hydraulics data. Idaho National Laboratory (INL), Oregon State University, Argonne National Laboratory, and Canadian Nuclear Laboratories have partnered to use this experimental data to develop a gas-cooled reactor thermal hydraulics benchmark led by INL under the auspices of the Advanced Reactor Technologies (ART) program. This report provides some context on the benchmark, HTTF, and previous Reactor Excursions and Leak Analysis Program (RELAP)5-3D modeling of HTTF. It also provides a draft of the benchmark specifications for the Depressurized Conduction Cooldown problem, which is the INL-led benchmark problem.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Availability of Neutronics Benchmarks in the ICSBEP and IRPhEP Handbooks for Computational Tools Testing

In the past several decades, numerous experiments have been performed worldwide to support reactor operations, measurements, design, and nuclear safety. Those experiments represent an extensive international investment in infrastructure, expertise, and cost, representing significantly valuable resources of data supporting past, current, and future research activities. Those valuable assets represent the basis for recording, development, and validation of our nuclear methods and integral nuclear data [1]. The loss of these experimental data, which has occurred all too much in the recent years, is tragic. The high cost to repeat many of these measurements can be prohibitive, if not impossible, to surmount. Two international projects were developed, and are under the direction of the Organisation for Co-operation and Development Nuclear Energy Agency (OECD NEA) to address the challenges of not just data preservation, but evaluation of the data to determine its merit for modern and future use. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) was established to identify and verify comprehensive critical benchmark data sets; evaluate the data, including quantification of biases and uncertainties; compile the data and calculations in a standardized format; and formally document the effort into a single source of verified benchmark data [2]. Similarly, the International Reactor Physics Experiment Evaluation Project (IRPhEP) was established to preserve integral reactor physics experimental data, including separate or special effects data for nuclear energy and technology applications [3]. Annually, contributors from around the world continue to collaborate in the evaluation and review of select benchmark experiments for preservation and dissemination. The extensively peer-reviewed integral benchmark data can then be utilized to support nuclear design and safety analysts to validate the analytical tools, methods, and data needed for next-generation reactor design, safety analysis requirements, and all other front- and back-end activities contributing to the overall nuclear fuel cycle where quality neutronics calculations are paramount.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Pulse Jet Mixing Tests With Noncohesive Solids (Revised)

This report summarizes results from pulse jet mixing (PJM) tests with noncohesive solids in Newtonian liquid. The tests were conducted during FY 2007 and 2008 to support the design of mixing systems for the Hanford Waste Treatment and Immobilization Plant (WTP). Tests were conducted at three geometric scales using noncohesive simulants, and the test data were used to develop models predicting two measures of mixing performance for full-scale WTP vessels. The models predict the cloud height (the height to which solids will be lifted by the PJM action) and the critical suspension velocity (the minimum velocity needed to ensure all solids are suspended off the floor, though not fully mixed). From the cloud height, the concentration of solids at the pump inlet can be estimated. The predicted critical suspension velocity for lifting all solids is not precisely the same as the mixing requirement for 'disturbing' a sufficient volume of solids, but the values will be similar and closely related. These predictive models were successfully benchmarked against larger scale tests and compared well with results from computational fluid dynamics simulations. The application of the models to assess mixing in WTP vessels is illustrated in examples for 13 distinct designs and selected operational conditions. The values selected for these examples are not final; thus, the estimates of performance should not be interpreted as final conclusions of design adequacy or inadequacy. However, this work does reveal that several vessels may require adjustments to design, operating features, or waste feed properties to ensure confidence in operation. The models described in this report will prove to be valuable engineering tools to evaluate options as designs are finalized for the WTP. Revision 1 refines data sets used for model development and summarizes models developed since the completion of Revision 0.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Lightweight runtime checking of C programs with RTC

The C Programming Language is known for being an efficient language that can be compiled on almost any architecture and operating system. However the absence of dynamic safety checks and a relatively weak type system allows programmer oversights that are hard to spot. In this paper, we present RTC, a runtime monitoring tool that instruments unsafe code and monitors the program execution. RTC is built on top of the ROSE compiler infrastructure. RTC finds memory bugs and arithmetic overflows and underflows, and run-time type violations. Most of the instrumentations are directly added to the source file and only require a minimal runtime system. As a result, the instrumented code remains portable. In tests against known error detection benchmarks, RTC found 98% of all memory related bugs and had zero false positives. Finally, in performance tests conducted with well known algorithms, such as binary search and MD5, we determined that our tool has an average run-time overhead rate of 9.7× and memory overhead rate of 3.5×.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2019

Transmission Expansion Planning Test System for AC/DC Hybrid Grid With High Variable Renewable Energy Penetration

Test systems are important tools and benchmark for power system research. Currently, there is a lack of standard test systems for modern transmission expansion planning (TEP) research, especially under high variable renewable energy (VRE) penetrations. This article describes a 38-bus test system named the HRP-38 system dedicated to TEP. “HRP” stands for high renewable penetration. The objective of establishing such a system is to provide a consistent platform for different TEP methods to be tested and compared. The test system considers high VRE penetration of more than 30% energy share, which will be an important feature in many power systems in the future. Both AC and DC candidate lines are given to provide sufficient transmission planning alternatives. The complexity of the test system is well balanced considering calculation tractability and the ability to test the performance of the TEP planning model. The network topology, generation mix and load characteristics are described in detail. Year-round time series of VRE (wind power and PV) output and load curves for each bus are provided to facilitate a detailed evaluation of the transmission grid configuration under high VRE penetration. Finally, four benchmark results with different optimization settings are provided for the HRP-38 system.

AC/DC hybrid↗
DOE OSTI2022

A Strategy for Low-Mass Dark Matter Searches with Cryogenic Detectors in the SuperCDMS SNOLAB Facility

The SuperCDMS Collaboration is currently building SuperCDMS SNOLAB, a dark matter search focused on nucleon-coupled dark matter in the 1-5 GeV mass range. Looking to the future, the Collaboration has developed a set of experience-based upgrade scenarios, as well as novel directions, to extend the search for dark matter using the SuperCDMS technology in the SNOLAB facility. The experienced-based scenarios are forecasted to probe many square decades of unexplored dark matter parameter space below 5 GeV, covering over 6 decades in mass: 1-100 eV for dark photons and axion-like particles, 1-100 MeV for dark-photon-coupled light dark matter, and 0.05-5 GeV for nucleon-coupled dark matter. They will reach the neutrino fog in the 0.5-5 GeV mass range and test a variety of benchmark models and sharp targets. The novel directions involve greater departures from current SuperCDMS technology but promise even greater reach in the long run, and their development must begin now for them to be available in a timely fashion. The experienced-based upgrade scenarios rely mainly on dramatic improvements in detector performance based on demonstrated scaling laws and reasonable extrapolations of current performance. Importantly, these improvements in detector performance obviate significant reductions in background levels beyond current expectations for the SuperCDMS SNOLAB experiment. Given that the dominant limiting backgrounds for SuperCDMS SNOLAB are cosmogenically created radioisotopes in the detectors, likely amenable only to isotopic purification and an underground detector life-cycle from before crystal growth to detector testing, the potential cost and time savings are enormous and the necessary improvements much easier to prototype.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

A blended continuous–discontinuous finite element method for solving the multi-fluid plasma model

The multi-fluid plasma model represents electrons, multiple ion species, and multiple neutral species as separate fluids that interact through short-range collisions and long-range electromagnetic fields. The model spans a large range of temporal and spatial scales, which renders the model stiff and presents numerical challenges. To address the large range of timescales, a blended continuous and discontinuous Galerkin method is proposed, where the massive ion and neutral species are modeled using an explicit discontinuous Galerkin method while the electrons and electromagnetic fields are modeled using an implicit continuous Galerkin method. This approach is able to capture large-gradient ion and neutral physics like shock formation, while resolving high-frequency electron dynamics in a computationally efficient manner. The details of the Blended Finite Element Method (BFEM) are presented. The numerical method is benchmarked for accuracy and tested using two-fluid one-dimensional soliton problem and electromagnetic shock problem. The results are compared to conventional finite volume and finite element methods, and demonstrate that the BFEM is particularly effective in resolving physics in stiff problems involving realistic physical parameters, including realistic electron mass and speed of light. The benefit is illustrated by computing a three-fluid plasma application that demonstrates species separation in multi-component plasmas.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Detecting Process Equipment Failures Using Acoustic Data and Machine Learning

Nuclear power plant (NPP) process equipment such as fans, motors, valves, and pumps generate frequent or continuous noise, and deviations from the normal operational sounds made by this equipment can indicate potential issues. These deviations can be identified via automated acoustic anomaly detection, which involves using acoustic sensors (i.e., microphones) alongside detection algorithms to continuously monitor for changes in acoustic signatures. This task is made challenging by the substantial background noise that exists, such as operators opening and closing doors, manipulating valves, and conversing—in addition to typical plant noises. In collaboration with a nuclear power utility partner, this effort assessed the efficacy of acoustic anomaly detection when using a specific acoustic sensor that compresses data into a fixed set of features that are transferable over a standard Internet of Things communication protocol, thereby improving usability but potentially degrading detection performance. Two methods of performing automated acoustic anomaly detection were evaluated: one-class support vector machine (OC-SVM) and isolation forest (iForest). To enable the use of high-quality acoustic data encompassing both normal and anomalous conditions, the study utilized the publicly available Malfunctioning Industrial Machine Investigation and Inspection dataset, which includes real measured acoustic sensor data for a range of equipment types, model numbers, and signal-to-noise ratios (SNRs), along with a benchmark set of detection results. Using this dataset, the methods were tested and then compared against the benchmark results. The results indicated that although the specific acoustic sensor did not enable as rich a feature set extraction, the proposed methods with the limited feature set performed just as well. This provides solid justification for both the methods and the use of the proposed acoustic sensor.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗
DOE OSTI2025

Engineering PHL7 for Improved Poly(Ethylene Terephthalate) Depolymerization via Rational Design and Directed Evolution

Enzymatic depolymerization of poly(ethylene terephthalate) (PET) has emerged as a promising approach for polyester recycling, and, to date, many natural and engineered PET hydrolase enzymes have been reported. For industrial use, PET hydrolases must achieve high depolymerization extent and exhibit excellent thermostability. Here, we engineered a natural PET hydrolase, Polyester Hydrolase Leipzig #7 (PHL7), through rational design and directed evolution using a high-throughput screening platform. Four new enzymes were engineered with enhanced properties compared with the parent enzyme, wild-type PHL7 (PHL7-WT), and other benchmark PET hydrolases, under the tested conditions. In bioreactors, the exemplary engineered enzyme, PHL7-Jemez, exhibited improved ability to depolymerize amorphous PET film compared with PHL7-WT at 2.9% and 20% substrate loadings, with 37% and 270% higher hydrolysis, respectively, after 48 h. This study develops several state-of-the-art PET hydrolases and demonstrates a directed evolution platform to engineer high-performance enzymes, which can accelerate enzyme discovery toward improved biocatalytic recycling.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗