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At least 1,333 records · Page 74

VALIDATION AND DETERMINATION OF SIGNIFICANT SIMULATION PARAMETERS USING THE SMOOTHED PARTICLE HYDRODYNAMIC CODE NEUTRINO

In recent years there has been an increase in flooding concern to nuclear power plants; thus, tools to reduce the cost of mitigation efforts are valuable to industry. Simulating flooding events is achieved through different methods and at varied levels of detail. However, these simulations must be able to match application conditions, which requires code validation as well as knowledge of which parameters affect the accuracy of the simulation. To validate and determine the significant parameters for these tools, a sloshing experiment was conducted at the George Washington University and then simulated with the smoothed particle hydrodynamic code Neutrino. The simulation setup was constructed to match the experiment as closely as possible, and the end-wall pressure results were analysed. The 90% and 50% end-wall pressure bounds were compared along with the pressure impulse. The simulation bounds fell mostly within the experimental bounds, with exceptions in insignificant low-pressure areas, such as at the beginning (~40% difference) and the end of the cycle (~40% difference). Another exception was seen at a point in the peak pressure (~75% difference). The simulated pressure impulse matched the experiments well and was within 10%. Additionally, seven parameters were qualitatively assessed and then investigated to determine their significance. The parameters’ significance was determined by sampling five values across the parameters’ range using the Risk Analysis Virtual Environment, then comparing the average pressure for each run. All seven parameters were identified as having some significance. However, the particle size, interaction-radius to particle-size ratio, and fluid settling seem to have a greater effect due to the larger fluctuation of the results. Now that the significant parameters are identified, further research is being done to quantify the significance of each parameter.

97 MATHEMATICS AND COMPUTING↗

BISON High Burnup Structure Modeling Capabilities Validated with a Selection of the Halden IFA-650 Rods

The U.S. Department of Energy (DOE)’s Nuclear Energy Advanced Modeling and Simulation (NEAMS)program aims to develop predictive capabilities using computational methods for the analysis and design of advanced reactor and fuel cycle systems. This program has been supporting the development of BISON, a high-fidelity and high-resolution fuel performance tool at the engineering scale. This document continues analysis and refinement of capabilities added to BISON early this calendar year in regards to the incorporation of capabilities applicable to extended burnups in response to industry interest. Details are provided on high burnup thermal conductivity models, a refitting of the high-burnup structure (HBS) porosity formation model to include additional data, the coupling of the HBS volume fraction model to thermal conductivity and fine fragmentation models, and validation activities. The IFA-650.4 and IFA-650.9 loss of coolant accident (LOCA) analyses are revisited with the latest developments in this report. A new validation case, IFA-650.14 has also been added to the BISON test suite.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Validation of Maui PSCAD Model: Motivation, Methodology, and Lessons Learned: Preprint

As power systems become increasingly dependent on inverter-based resources, the dominant physical and control interactions that determine system stability may no longer be accurately captured in positive sequence dynamic simulation tools; electromagnetic transient (EMT) simulations may be needed. This paper describes the process of creating and validating an EMT model of the entire Maui island transmission system in PSCAD software. The model is validated against the utility's PSSE model and against field data for an event that consisted of a single-phase fault followed by a generation trip. The methodology and lessons learned are summarized here.

41 EE - Solar Energy Technologies Office (EE-4S)↗

Validation Efforts of an Open-Source Aeroacoustics Model for Wind Turbines: Preprint

The open source aeroservoelastic wind turbine solver, OpenFAST, now includes an aeroacoustics model, which is described here and validated against experimental measurements recorded on a GE 1.5-MW wind turbine installed at the Flatirons Campus of the National Renewable Energy Laboratory in Boulder, Colorado. The validation demonstrates satisfactory agreement between numerical predictions and experimental recordings, with discrepancies up to 7 dB in the overall sound pressure levels at low wind speeds and a better agreement around the rated wind speed of the turbine.

49 EE - Wind and Water Power Program - Wind (EE-4W↗

PV Validation Hub

The Validation Hub will be a clearinghouse for the transfer of novel algorithms and software from the PV research community to industry. Potential algorithms tested in the Hub could include the estimation of various PV loss factors and the detection of various operational issues. The primary function of the Hub will be for developers to submit executable code which will run on hosted data sets. Developers will receive private reports on the accuracy and performance (e.g., run-time) of the submitted algorithms, and public high level summaries will be hosted. These summaries will indicate the organization who submitted the algorithm (e.g., links to GitHub pages, documentation websites, etc.), high-level accuracy metrics, and standardized performance metrics. These results will be stored in a publicly available database, accessible through the Hub, with the ability for users to sort and filter the results. In short, the Hub will be presented to public users as a collection of interactive leaderboards, organized around specific analysis tasks pertinent to the PV data science community. These tasks include things such as the estimation of various PV loss factors and the detection of various operational issues. We will present progress on the development of this hub, including preliminary results of comparative validation of PV data science algorithms and progress towards building the platform itself.

algorithm↗

Verification and Validation of the New MCNP6.3 Criticality Features

The MCNP6® code, version 6.3, has been extensively verified and validated for many applications. The use of the same default capabilities existing in the MCNP6.2 code are also used in the verification and validation (V&V) of the MCNP6.3 code. In this paper, selected new features, code enhancements, and bug fixes in the MCNP6.3 code that impact criticality safety applications are described and investigated. More specifically, the changes within the MCNP6.3 code that are studied in this work include both the new fission matrix and the Doppler broadening resonance correction (DBRC) features. For nuclear criticality safety applications, the V&V benchmark problems within the criticality, extended criticality, and Rossi-α suites are used to study the upgrades within the MCNP6.3 code. Some additional investigations into benchmarks at elevated temperatures are used to showcase the impacts of the temperature-specific capabilities.

97 MATHEMATICS AND COMPUTING↗

Testing and Performance Validation of a Shielded Waste Segregation and Clearance Monitor Designed for the Measurement of Low Level Waste-13043

This paper describes the development, testing and validation of a shielded waste segregation and clearance monitor designed for the measurement of low-density low-level waste (LLW). The monitor is made of a measurement chamber surrounded by detectors and a shielded outer frame. The shielded chamber consists of a steel frame, which contains typically 1.5 inches (3.81 cm) of lead and 0.5 inches (1.27 cm) of steel shielding. Inside the shielding are plastic scintillator panels, which serve as gross gamma ray detectors. The detector panels, with embedded photomultipliers, completely surround the internal measurement chamber on all 6 sides. Care has been taken to distribute the plastic scintillator detectors in order to optimise both the efficiency for gamma ray detection and at the same time achieve a volumetric sensitivity, which is as uniform as possible. A common high voltage power supply provides the bias voltage for each of the six photomultipliers. The voltage signals arising from the detectors and photomultipliers are amplified by six sensitive amplifiers. Each amplifier incorporates a single channel analyser with both upper and lower thresholds and the digitised counts from each detector are recorded on six scalars. Operation of the device is by means of a microprocessor from which the scalars are controlled. An internal load cell linked to the microprocessor determines the weight of the waste object, and this information is used to calculate the specific activity of the waste. The monitor makes background measurements when the shielded door is closed and a sample, usually a bag of low-density waste, is not present in the measurement chamber. Measurements of the minimum detectable activity (MDA) of an earlier large volume prototype instrument are reported as part of the development of the Waste Segregation and Clearance Monitor (WSCM) described in the paper. For the optimised WSCM a detection efficiency of greater than 32% was measured using a small Cs-137 source placed in the centre of the measurement chamber. Small sources have also been used to determine the spatial variation of the detection efficiency for various positions within the measurement chamber. The data have been used to establish sentencing limits and different 'fingerprints' for specific waste streams including waste streams containing fission products and others based on other radionuclides including Am-241. Some of the test data that are presented have been used to validate the instrument performance. The monitor is currently in routine use at a nuclear facility for the measurement and sentencing of low-density low activity radioactive waste. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Testing and Performance Validation of a Sensitive Gamma Ray Camera Designed for Radiation Detection and Decommissioning Measurements in Nuclear Facilities-13044

This paper describes the measurements, testing and performance validation of a sensitive gamma ray camera designed for radiation detection and quantification in the environment and decommissioning and hold-up measurements in nuclear facilities. The instrument, which is known as RadSearch, combines a sensitive and highly collimated LaBr{sub 3} scintillation detector with an optical (video) camera with controllable zoom and focus and a laser range finder in one detector head. The LaBr{sub 3} detector has a typical energy resolution of between 2.5% and 3% at the 662 keV energy of Cs-137 compared to that of NaI detectors with a resolution of typically 7% to 8% at the same energy. At this energy the tungsten shielding of the detector provides a shielding ratio of greater than 900:1 in the forward direction and 100:1 on the sides and from the rear. The detector head is mounted on a pan/tile mechanism with a range of motion of ±180 degrees (pan) and ±90 degrees (tilt) equivalent to 4 π steradians. The detector head with pan/tilt is normally mounted on a tripod or wheeled cart. It can also be mounted on vehicles or a mobile robot for access to high dose-rate areas and areas with high levels of contamination. Ethernet connects RadSearch to a ruggedized notebook computer from which it is operated and controlled. Power can be supplied either as 24-volts DC from a battery or as 50 volts DC supplied by a small mains (110 or 230 VAC) power supply unit that is co-located with the controlling notebook computer. In this latter case both power and Ethernet are supplied through a single cable that can be up to 80 metres in length. If a local battery supplies power, the unit can be controlled through wireless Ethernet. Both manual operation and automatic scanning of surfaces and objects is available through the software interface on the notebook computer. For each scan element making up a part of an overall scanned area, the unit measures a gamma ray spectrum. Multiple radionuclides may be selected by the operator and will be identified if present. In scanning operation the unit scans a designated region and superimposes over a video image the distribution of measured radioactivity. For the total scanned area or object RadSearch determines the total activity of operator selected radionuclides present and the gamma dose-rate measured at the detector head. Results of hold-up measurements made in a nuclear facility are presented, as are test measurements of point sources distributed arbitrarily on surfaces. These latter results are compared with the results of benchmarked MCNP Monte Carlo calculations. The use of the device for hold-up and decommissioning measurements is validated. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Validation of numerical simulations of activation by neutron flux - 15038

The knowledge of the radionuclide content of radioactive waste is of utmost importance for safety and waste management reasons. Numerical simulations are used at EDF-CIDEN to anticipate the dismantling and the radioactive waste management. The activation scheme by neutron flux developed at EDF-CIDEN comprises four steps. First of all, a 3 dimensional description of the reactor, the chemical composition of each structure, and the neutron sources at nominal power rating conditions are used as the input data to simulate the neutron propagation and to generate a 3 dimensional multi energy group neutron flux distribution in nominal power rating conditions. Secondly, this output data is used with the chemical compositions (including impurities) and the history of the operating conditions to calculate the radiological inventory for each structure. Thirdly, according to the radioactive inventory of each component or sub-component, a waste classification can be made; the classification criteria are based on the levels of radioactivity and radiotoxicity of 143 radionuclides. Finally, comparisons between the calculated radioactive inventories and the results of radio-chemical analyses can be made. These comparisons are made with regards to the activation of 'standard' chemical elements (meaning radionuclides being perfectly modeled with no uncertainty about the chemical concentrations). The results, linked with 'standard' chemical elements, can be used to validate the calculation scheme, including the neutron propagation, the neutron activation and all the other associated simplified assumptions (geometry, history of irradiation, etc). For steel, activation of 'Fe' is retained, and for stainless steel, activation of 'Fe' 'Ni' are retained. Overall, the feedback from Chooz A (Pressurized Water Reactor) shows that numerical simulations produce a slight overestimation of 'Fe-55' and 'Ni-63'. The results linked with minor chemical elements or impurities allowed us to validate the use of the average measured compositions. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Investigations into validation of plutonium solutions for criticality safety analysis

Plutonium solution chemistry is among the most complex of all the elements in the periodic table. Under acidic conditions up to four different oxidation states can be present in significant amounts in a single solution. From the perspective of nuclear criticality safety, accurate information regarding the concentration of plutonium and other constituents in solution must be well-known in order to correctly model plutonium solution systems for analysis. This requires knowledge of the solution system along with the ability to make accurate assumptions and predictions of the chemical and physical properties of the solution system. If neutron transport methods are used to assess subcriticality in a nuclear criticality safety analysis, the models must be compared with benchmark experiments which are similar and the suitability of the computer code system must be quantified. This paper describes some investigations undertaken to support the validation of plutonium solution systems for nuclear criticality safety analysis and how the Whisper sensitivity/uncertainty (S/U) tool can provide information to support validation of such systems. Case study results for various solution systems such as aqueous plutonium nitrate, aqueous plutonium chloride and hypothetical metal-water mixture are presented. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Results of TRIPOLI-4 version 10 on the TRIPOLI-4 criticality validation suite, thermal spectra

This paper presents the results of the latest version (version 10) of TRIPOLI-4 on its criticality validation suite, thermal spectra part. The criticality validation suite is made of 148 selected ICSBEP benchmarks with various enrichments, fissile materials and spectra. The ICSBEP benchmarks computational models are prepared by the TRIPOLI-4 team and are run in the same conditions for each benchmark: initial source, number of neutrons per cycle, number of inactive and active cycles, k{sub eff} estimator and nuclear data. Comparisons are made with experimental k{sub eff}. On one case (PST-009), we have noted a discrepancy between the computational results and the experiments. This discrepancy is also found with other codes, which might indicate a possible issue with the benchmark. For all the other considered cases, uranium and plutonium solutions and fuel-rods-arrays, the agreement with the experiments is very good.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Implementation and validation of a fully implicit accumulator model in RELAP-7

This paper presents the implementation and validation of an accumulator model in RELAP-7 under the framework of preconditioned Jacobian free Newton Krylov (JFNK) method, based on the similar model used in RELAP5. RELAP-7 is a new nuclear reactor system safety analysis code being developed at the Idaho National Laboratory (INL). RELAP-7 is a fully implicit system code. The JFNK and preconditioning methods used in RELAP-7 is briefly discussed. The slightly modified accumulator model is summarized for completeness. The implemented model was validated with LOFT L3-1 test and benchmarked with RELAP5 results. RELAP-7 and RELAP5 had almost identical results for the accumulator gas pressure and water level, although there were some minor difference in other parameters such as accumulator gas temperature and tank wall temperature. One advantage of the JFNK method is its easiness to maintain and modify models due to fully separation of numerical methods from physical models. It would be straightforward to extend the current RELAP-7 accumulator model to simulate the advanced accumulator design. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Systems usability case in stepwise control room validation - 125

We present a Systems Usability Case (SUC) approach enabling a requirement-based human factors evaluation of complex technical systems. The approach is especially suitable for stepwise verification and validation (V and V) of nuclear power plant (NPP) control room (CR) systems. A SUC is based on the Safety Case approach and on the Systems Usability (SU) construct. A series of V and V test activities carried throughout the design process and constructed according to SUC enables us to monitor how evaluation evidence is aggregated over variety of V and V steps and how the design solution is maturing towards a tool of a good quality and usability. The paper demonstrates an application of the SUC approach to real data from a stepwise CR validation. The results suggest that SUC has some clear advantages. It makes it easier to conduct the V and V activities in a more systematic fashion and it makes the reasoning process more explicit and transparent. Moreover, it enables building a longitudinal view of the progress of the design process. Most importantly, constructing the SUC enables monitoring of the fulfilment of the requirements from the human factors (i.e., SU) perspective. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Use of immersive 3D virtual reality environments in control room validations - 203

Advanced technologies (e.g., virtual and augmented reality) may provide new possibilities to facilitate control room (CR) design and evaluation activities. We studied how immersive 3D environments may augment and advance the evaluation of safety-critical CR systems. The 3D virtual reality (VR) CR was compared with a real-life development simulator environment where hardwired panels were substituted with Virtual panels. A multi-user functionality enables several operators to be and collaborate in the same VR CR environment at the same time. There is also a realistic representation of emergency operating procedures in the VR CR. Spatial audio communication through headsets makes the experience even more realistic. The results section of the paper address both technical and human factors issues associated with the use of immersive 3D VR environments in CR validation tests, e.g.: amount of technical resources required as compared to normal validation in a real-life simulator environment, creation of methodologically new testing opportunities and new opportunities to data registration and analysis. (authors)

47 OTHER INSTRUMENTATION↗

Development of a transient signal validation technique via a modified kernel regression model - 365

The demand for robust and resilient performance has led to the use of online-monitoring and signal validation techniques to monitor the process parameters and equipment conditions using several empirical models. One of these models is kernel regression (KR). Although KR is being used and performed excellently when applied to steady state, it has limitation in time-varying data that has several repetition of the same data point of the signals, especially if those signals are used to infer the other signals. In addition, many situations are related to variation and fluctuation during normal states, such as transient - start-up and shutdown mode. However, accurate estimation of the process signal can lead to the proper understanding of the equipment behaviours as well as enhance the online monitoring applications. In the sight of foregoing, this paper proposed a modified KR model, especially for transient signal validation to resolve the setback of conventional KR by improving the measure of similarities in the conventional KR. The performance of the proposed model is verified using both experimental data and the plant simulation data, and its results are compared with that of conventional KR. The results show that the proposed model has a greater robustness in particular situations compared to conventional KR, and it can be used to improve process and equipment monitoring applications. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Quantitative Analysis Guidance in Safety Software Validation

Quantitative analysis in software development has been a challenging topic over the years in safety software. While qualitative processes for vigorous safety software verification and validation (V and V) have been provided through standard such as IEEE 1012, quantitative analysis for V and V tasks such as correctness, completeness or accuracy evaluation have not been formally defined. In practice, there are quantitative analyses in terms of using software metrics. Many of these software metrics are useful in analyzing the test results. Some of them are used for providing guidance in code evaluation activities e.g. numbers of line of codes not exceeding certain limits, recursive calls are not allowed, etc. However, few can be used for providing quantitative analysis guidance for the V and V tasks such as completeness evaluation. AREVA Federal Services (AFS) has provided services for many nuclear projects in US. Through compliance with IEEE 1012, AFS V and V processes have provided vigorous qualitative processes ensuring quality level of the software products used in the project meeting the required nuclear software level. In addition, quantitative analysis is developed to provide structured validation framework ensuring 'zero defect' in the software products. This paper describes some aspects of the quantitative approaches using a sample safety application. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Validation of Criticality Analysis Capability of SuperMC with ICSBEP

SuperMC is a general purpose, intelligent and multifunctional, continuous energy Monte Carlo program for the design and safety evaluation of nuclear systems. It is designed to perform the comprehensive neutronics calculation, taking the radiation transport as the core and including the depletion, radiation source term/dose/biohazard, material activation and transmutation, etc. The latest version of SuperMC can perform the simulation of neutron, photon and coupled neutron and photon transport, and also equipped with the functions of automatic modeling and visualization. SuperMC has been verified by more than 2000 benchmark models and experiments including the handbook of International Criticality Safety Benchmark Evaluation Project (ICSBEP), the Shielding Integral Benchmark Archive Database (SINBAD), etc. The fusion reactor (ITER benchmark model, FDS Fusion Reactors), fast reactors (BN600, CLEAR-I), PWR reactors (HM, BEAVRS) and cases from the International Reactor Physics handbook Evaluation Program (IRPhEP) were employed for validating the comprehensive capability for reactor applications. In this paper, the validation of SuperMC based on the handbook of ICSBEP will be presented. ICSBEP, released by OECD/NEA, contains criticality safety benchmark specifications derived from experiments that were performed at various nuclear critical facilities around the world, which was widely used to verify neutronic codes and associated nuclear data libraries for criticality safety analysis. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

CFD Verification and Validation of Wire-Wrapped Pin Assemblies

TerraPower is participating in a cooperative project among industry, the national lab system, and academia to perform Verification and Validation (V and V) of Computational Fluid Dynamics (CFD) methods for predicting the flow and heat transfer within liquid-metal-cooled nuclear fuel assemblies with wire-wrapped fuel pins. This project, consisting of both experimental and numerical components, uses surrogate fluids and electrically heated fuel pins to substitute for the liquid metal coolant and the nuclear heat source. The experiments include both unheated assemblies to measure the flow velocities with Laser Doppler Velocimetry (LDV) and Particle Image Velocimetry (PIV), as well as heated assemblies to measure the temperatures of the fluid and the simulated fuel pins with thermocouples. Both sets of experiments also measure the pressure at multiple locations in the system. The numerical component of this work involves high-fidelity Large-Eddy Simulation (LES) modeling and industry-standard Reynolds-Averaged Navier-Stokes (RANS) modeling of the experiments. Furthermore, the experiments and simulations will be conducted on two duct geometries: the first is a non-deformed hexagonal duct that has flat internal duct faces that represents a beginning of life geometry. The other is a deformed duct where the internal hexagonal flow area enlarges then decreases along the axial height to represent the changes that occur to the duct geometry from pressure and irradiation effects in the core. This paper describes the methods and results of the industrial, RANS-based CFD code verification, solution verification, and solution validation. The V and V simulations are being performed for the following experiment test geometries: - Isothermal non-deformed bundle; - Heated non-deformed bundle; - Isothermal deformed bundle; - Heated deformed bundle. These flows have been computed at first with the pre-test design specifications, without the aid of any experimental results. Later, post-test calculations will be performed incorporating the actual test conditions and input from some of the experimental results. These RANS simulations of helically wire-wrapped fuel assemblies employ meshes of bare pins without the wire-wrap in the bundle. The effect of wire-wrap on the flow is accounted for by introducing a momentum source into the governing fluid equations. The momentum source is only applied to cells corresponding to the location of the wire-wrap, and its components in each cell are based on the local flow field. This methodology simplifies the geometry and mesh generation and reduces the cell count, while still capturing the effects of the wire-wrap. The momentum source approach is used for all of the assembly simulations performed in this study. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗