Search NASA⌕ Search

SEARCH · Search NASA

Results for “Reactor Physics 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.

At least 433 records · Page 24
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Verification of Spent Nuclear Fuel in Sealed Dry Storage Casks via Measurements of Cosmic-Ray Muon Scattering

Most of the plutonium in the world resides inside spent nuclear reactor fuel rods. This high-level radioactive waste is commonly held in long-term storage within large, heavily shielded casks. Currently, international nuclear safeguards inspectors have no stand-alone method of verifying the amount of reactor fuel stored within a sealed cask. In this paper, we demonstrate experimentally that measurements of the scattering angles of cosmic-ray muons, which pass through a storage cask, can be used to determine if spent fuel assemblies are missing without opening the cask. Finally, this application of technology and methods commonly used in high-energy particle physics provides a potential solution to this long-standing problem in international nuclear safeguards.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Test based reliability assessment method for a safety critical software in reactor protection system - 284

Digitalization of the instrumentation and control systems in nuclear power plant entails some new features which do not exist in analog systems. When the new features are applied in safety critical systems, the risk stemming from them should be quantified properly to assure the reliability of the entire power plant. Among the new features, software is considered as the most important factor because it can cause common cause failure to many redundant systems. Therefore, in this work, a test based reliability assessment method for a safety-critical software is developed. An output of a software is determined by not only inputs but also the internal state of the software at the time. In the case of safety critical software developed for usage in nuclear power plant, it is possible to directly investigate probable internal states thanks to detailed design specification and programing features, and the probable inputs also can be identified based on the physical linearity of each process parameter and hardware characteristics related to data acquisition. In this work, a development method for practical exhaustive test case consisting of investigation of internal state and probable input sets to a specific internal state are suggested. In addition to this, software logic simulator for execution of software logic test and the reliability quantification method based on the test result are developed. The feasibility of the suggested method is demonstrated via a case study. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Enabling Dynamic Probabilistic Risk Assessment of Physical Security Using EMRALD and MAAP

The optimization of physical security in nuclear power plants requires sophisticated methodologies that integrate operator actions and plant behavior through advanced simulation tools. To address this, Idaho National Laboratory [JL2.1]has developed the Modeling and Analysis for Safety and Security using the Dynamic EMRALD Framework (MASS-DEF) methodology, an approach that integrates force-on-force simulations, dynamic probabilistic risk assessment, and thermal-hydraulics modeling [JL3.1]to enhance security planning while reducing costs. We developed a tool that produces reduced order models using thermal hydraulic simulations from the Modular Accident Analysis Program (MAAP) [1]. These models can quickly evaluate reactor core behavior during attack simulations, and in so doing, address two barriers of traditional methods: (1) MAAP simulations are computationally intensive, and (2) attack scenarios must be run in a secure environment, which complicates analysis and validation. By precomputing scenario outcomes for a small number of modified parameters, the reduced order model significantly decreases the computational cost and enables offsite review of the results.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Combination of thermal and electric properties’ measurement techniques in a single setup suitable for radioactive materials in controlled environments and based on the 3 ω approach

We have designed and developed a new experimental setup, based on the 3 Omega method, to measure thermal conductivity, heat capacity and electrical resistivity of variety of samples in a broad temperature range (2 to 600 K) and under magnetic fields up to 9 T. The validity of this method is tested by measuring various kinds of metallic (copper, platinum and constantan) and insulating (SiO 2 ) materials, which have a wide range of thermal conductivity values (1-400 Wm -1 K -1 ). We have successfully employed this technique for measuring thermal conductivity of two actinide single crystals, uranium dioxide and uranium nitride. Thus, this new experimental approach for studying nuclear materials will help to advance reactor fuel development and understanding. Here, we have also shown that this experimental setup can be adapted to Quantum Design PPMS measurement environment and/or cryocooler systems.

3 Omega Method↗
DOE OSTI2017

Modeling and Preparation for Experimental Testing of Heat Fluxes on W7-X Divertor Scraper Elements

Simulations of heat fluxes to the plasma facing components in the Wendelstein 7-X stellarator will be tested in its next operational phase. The simulations consist of core transport calculations that determine the evolution of the kinetic profiles and the toroidal current, which modifies the fluxes to the divertor, as the magnetic geometry changes. An additional divertor component, the scraper element, was designed to protect the edges of the primary divertor throughout this evolution during certain high-power long-pulse operational scenarios. The effect of unknown parameters of the heat flux calculations, namely, the cross-field thermal diffusivity and the magnetic field structure, is explored. The predicted scaling of the heat flux widths and magnitudes is presented, along with a new method of calculating the 3-D magnetic field structure required to perform the flux calculations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Core Combination Method of Approximation (CCMA) for Neutron Fluence Calculation

Neutron fluence, or accumulative neutron fluence rate, is a key parameter of the neutron irradiation resistance of materials (Wagner, Bergner et al. 2016). So we must try to obtain it during irradiation experiments. Neutron fluence is defined as the time integral of neutron fluence rate. To get the quantity, or to do this integral, we need to know how neutron fluence rate varied with time. The integral might not be difficult when we conduct a material irradiation experiment, because we can put certain kind of neutron detectors in the location of species, to measure neutron fluence rate (Tanaka, Endo et al. 2011). At this circumstance, the reactor power is usually maintained to be constant (Ferroukhi, Hollard et al. 2009; Dupre, Vasiliev et al. 2015). So the neutron fluence is equal to the product of neutron fluence rate and irradiation time. However, it's not the same case when we want to get the neutron fluence of structure materials used in the SPRR- 300 (Swimming Pool Research Reactor -300). This reactor is a light-water-moderated, enriched-uranium-fueled research reactor developed by China. Structural materials, such as graphite, are of great value for research (Asamoah, Gyampo et al. 2012). Graphite was used as reflector in reactor. It is a waste of material and a waste of money to dismantle it simply. However, to reuse post-irradiation graphite assemblies faces some challenges, such as neutron fluence calculation, radiation protection, etc. The former will be discussed in this paper, whereas the latter will remain to be studied for the following time. The problem is how to acquire neutron fluence distribution in structure materials, since specific measurement of neutron fluence rate were not conducted intentionally during historical reactor operation. So we need to figure it out by a method of calculation. Considering a large number of reactor refueling, we propose an approximate method, Core Combination Method of Approximation (CCMA), to calculate accumulative neutron fluence rate of key structure materials during a long period of irradiation time.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Development and Validation of SAM Multi-dimensional Flow Model for Thermal Mixing and Stratification Modeling

Thermal mixing and stratification in large pools or enclosures are very important phenomena that are critical to nuclear reactor safety. Because of the wide ranges of time and length scales associated with such phenomena, accurate modeling and simulation of thermal mixing and stratification remain as the key unresolved, challenging problems for reactor transient analyses. In traditional system analysis codes, simplified zero-dimensional (0-D) models are widely used for their high numerical efficiency, but they generally suffer from very limited prediction accuracies or range of applicability. Like traditional system analysis codes, the current version of the SAM code has implemented such simplified 0-D and one-dimensional mixing models. On the other hand, high-resolution Computational Fluid Dynamics (CFD) tools are often used to model complex thermal mixing and stratification phenomena. They are, however, generally numerically expensive, and they require large amounts of computational resources. It is therefore desirable to implement advanced and efficient thermal mixing and stratification modeling capabilities embedded in a system analysis code. This approach will improve the accuracy of reactor safety analyses when thermal mixing and stratification are involved, and also avoid using the large computational resources needed for high-resolution CFD analysis. Currently, with the support of the U.S. DOE Office of Nuclear Energy’s Nuclear Energy Advanced Modeling and Simulation program, an effort has been launched to develop and implement a multi-dimensional flow model in the system analysis code SAM, and demonstrate its applications to model thermal mixing and stratification phenomena in large enclosures. The main outcomes of this research and development activity are summarized in this report, which presents an attempt to include a built-in advanced multi-dimensional flow model in a system analysis code with the focus on overcoming the simulation challenges of thermal mixing and stratification phenomena. In this report, we start with the introduction of existing SAM code capabilities to simulate thermal mixing and stratification phenomena, which is followed by a short summary of the multi-dimensional model implemented in the SAM code, including both the physical model and the Finite Element Method code implementation. In this study, two options were implemented in the SAM code to model turbulent flows: a relatively simple built-in turbulence model and an interface to accept externally computed turbulent viscosities (e.g., from a high- resolution CFD simulation). Code validation studies on this newly added capability were then carried out to compare SAM simulation results with experimental data from the SUPERCAVNA facility, which was designed to study the complex flow recirculation and thermal stratification phenomena relevant to sodium fast reactor designs. In this study, one transient and two steady-state test cases were used for code validation. Different approaches have been used to model the complex turbulence flow fields in the SUPERCAVNA facility. A highly simplified zero-equation turbulence model was first used, but it was determined that it is too simple to capture the complex turbulence flow fields in these test cases. Subsequently, the code validation continued with the use of turbulent viscosity data from high-resolution STAR-CCM+ CFD simulations to improve the accuracy of the results. Using this approach, the SAM simulation results showed very good agreement with both the SUPERCAVNA experimental data and STAR-CCM+ simulation results. In this report, we demonstrate the development, implementation, and successful validation of a multi-dimensional flow model in the SAM code, which aims to improve the simulation accuracy for complex thermal mixing and stratification phenomena. Lessons have also been learned, including that in cases where the flow fields are not well predicted by the zero-equation model, the iinput of turbulent viscosities from an external source can enhance the overall predictive capabilities needed to accurately capture complex thermal-hydraulic phenomena. Therefore, future research will be needed to further improve the code’s capabilities, such as by developing a more efficient and robust approach to capture the turbulence effects in the SAM code.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Extension of the high-resolution thermal-hydraulics code ESCOT to hexagonal core geometries for multi-physics calculations

The extension of the capabilities of the pin-level nuclear reactor core thermal-hydraulics (T/H) code ESCOT to analyze hexagonal fueled cores and its performance are presented. ESCOT is an accurate yet fast core thermal-hydraulics solution aiming at high-fidelity and high-resolution multi-physics core analysis in the framework of massively parallel computing platforms. Its algorithm solution is based on the four-equation drift-flux model for two-phase calculations, these are numerically solved by applying the Finite Volume Method (FVM) and the Semi-Implicit Method for Pressure-Linked Equation (SIMPLE)-like algorithm in a staggered grid system. Constitutive models such as turbulent mixing, pressure drop, and vapor generation are employed to simulate key phenomena in subchannel-scale analysis. ESCOT is parallelized by a double (radial and axial) domain decomposition that enables its highly parallelized execution. The coupling of the code with the neutronics whole core solver for hexagonal geometries nTRACER is described. The newly implemented ESCOT features are validated by comparing single assembly and full core steady state nTRACER-ESCOT solutions with nTRACER standalone internal one-dimensional T/H solver results. The validation problems are based on the VVER 440 and VVER 1000 cores. ESCOT results show differences within an acceptable range with respect to the simple 1D nTRACER built-in solver. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Coupled Calculations of the NuScale FIV Tests – Pre-test Simulations and Sensitivity Analysis

The NEAMS program aims to develop an integrated multi-physics simulation capability “pellet-to-plant” for the design and analysis of future generations of nuclear power plants. In particular, the Reactor Product Line code suite's multi-resolution hierarchy is being designed to ultimately span the full range of length and time scales present in relevant reactor design and safety analyses, as well as scale from desktop to petaflop computing platforms. Flow-induced vibration (FIV) is widespread problem in energy systems because they rely on fluid movement for energy conversion. Vibrating structures may be damaged as fatigue or wear occurs. Given the importance of reliable components in the nuclear industry, flow-induced vibration has long been a major concern in safety and operation of nuclear reactors. In particular, nuclear fuel rods and steam generators have been known to suffer from flow-induced vibration and related failures. Advanced reactors, such as integral Pressurized Water Reactors (PWRs) considered for Small Modular Reactors (SMR), often rely on innovative component designs to meet cost and safety targets. One component that is the subject of advanced designs is the steam generator, some designs of which forego the usual shell-and-tube architecture in order to fit within the primary vessel. A significant amount of data exists on flow-induced vibration in shell-and-tube heat exchangers, and heuristic methods are available to predict FIV based on a set of given assumptions. In contrast, advanced designs have far less data available. Advanced modeling and simulation based on coupled structural and fluid simulations have the potential to predict flow-induced vibration in a variety of designs, reducing the need for expensive experimental programs, especially at the conceptual design stage. Over the past several years, the Reactor Product Line has developed a FIV capability in SHARP. SHARP contains high-fidelity single-physics codes Diablo for structural mechanics and Nek5000 for fluid mechanics calculations. Both codes are state-of-the-art, highly scalable tools that have been extensively validated. In previous reports we have discussed the implementation and validation of an FIV capability for helical steam generators and fuel assemblies. In this report we discuss one-way coupled calculations performed with Nek5000 and Diablo for the NuScale FIV tests. In the turbulent fretting regime one-way coupling is judged to be sufficient because the pressure loads do not cause substantial displacements. It is also the most common source of vibration in helical steam generators per the low flows expected in integral PWRs. The simulations precede the tests, as experimental data was not available at the delivery of this report. Several assumptions had to be made on the support-tube interaction and multiple tests and sensitivity analysis have been performed. Additional work remains to be done and it is discussed in the future work section.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

A Well-Posed Two Phase Flow Model and its Numerical Solutions for Reactor Thermal-Fluids Analysis

A 7-equation two-phase flow model and its numerical implementation is presented for reactor thermal-fluids applications. The equation system is well-posed and treats both phases as compressible flows. The numerical discretization of the equation system is based on the finite element formalism. The numerical algorithm is implemented in the next generation RELAP-7 code (Idaho National Laboratory (INL)’s thermal-fluids code) built on top of an other INL’s product, the massively parallel multi-implicit multi-physics object oriented code environment (MOOSE). Some preliminary thermal-fluids computations are presented.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Integrated FHR technology development: Tritium management, materials testing, salt chemistry control, thermal hydraulics and neutronics, associated benchmarking and commercial basis

This report summarizes the results of a 3-year Integrated Research Project (IRP) sponsored by the U.S. Department of Energy to address major technical challenges in the development of the Fluoride-salt-cooled High-temperature Reactor (FHR). The IRP universities included the Massachusetts Institute of Technology (MIT), the University of California at Berkeley (UCB), the University of Wisconsin (UW), and the University of New Mexico (UNM). The FHR is a new reactor concept less than 20 years old that combines (1) a clean fluoride salt coolant, (2) the graphite-matrix coated-particle fuel originally developed for High- Temperature Gas-Cooled Reactors and (3) passive decay heat removal systems from sodium fast reactors. The base-line design is a pebble-bed FHR that delivers heat between 600 to 700°C with a 7 Li 2 BeF 4 (flibe) salt coolant to the power cycle or industrial users. The Executive Summary describes what was accomplished and summarizes major conclusions. The main report contains stand-alone sections in each major area of work. Highlights are summarized below. Tritium. The coolant salt produces tritium necessitating understanding and control of tritium. Experimental work included measuring hydrogen uptake on carbon, tritium uptake during irradiation of graphite in 700°C salt and the permeability of tritium through metals with and without coatings. An FHR systems model was developed to predict tritium behavior and enable design of tritium control systems. The leading candidate for tritium removal from the flibe salt is carbon bed outside the reactor core. Thermal hydraulics and Neutronics. A 10-meter high FHR integral effects test facility, using Dowtherm A®, has been built and used to simulate the integral response of a representative FHR to improve understanding of FHR thermal hydraulics and validate thermal hydraulics codes used for the FHR. Neutronic analysis included code-to-code benchmarking, multiphysics modeling, uncertainty analysis for nuclear data uncertainties and shielding analysis. Twisted-tube heat exchangers were examined that may provide significant cost reductions because of their improved heat-transfer performance with liquid salts. The impacts on safety and design caused by uncertainties in physical properties of the salt were assessed with recommendations on where added physical property measurements are needed. Work began on understanding radiative heat transfer that becomes important in salt systems at higher temperatures. Materials. Static corrosion tests were conducted on graphite and materials in 700°C salt -- both in the UW laboratory and in the MIT reactor. In addition to static corrosion testing, a circulating salt corrosion loop was built and is operating. Methods to measure salt properties including redox were developed. The presence of carbon increases corrosion. With good redox control, the evidence indicates stainless steel can be used as a material of construction. Benchmarking. An additional series of workshops where held to enable benchmarking and comparison of results between different investigators. Commercialization. The FHR delivers higher average-temperature (~650°C) heat to the power cycle or industry than other reactors. This enables power cycles with lower-cost heat storage for variable electricity to the grid while the reactor operates at base-load -- providing a major competitive advantage by maximizing revenue from electricity sales. In October, 2016, a new startup company named Kairos Power was formed, and opened its first office in Jack London Square, Oakland, California, in January 2017. Kairos Power exists largely due to the work that was accomplished by this IRP and the previous DOE-funded FHR IRP that performed research on FHR technology from 2012 to 2015. In September, 2018, as this IRP is concluding, Kairos Power was staffed at 60 full time employees, working to develop technology derived from this IRP. Many of these employees worked with the IRP before joining Kairos Power 1 (https://kairospower.com). Other students and researchers from this IRP have joined other startup companies, including the Terrapower Molten Chloride Fast Reactor and the Terrestrial Energy development efforts.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Methods for recovering metals from electronic waste, and related systems

A method of recovering metals from electronic waste comprises providing a powder comprising electronic waste in at least a first reactor and a second reactor and providing an electrolyte comprising at least ferric ions in an electrochemical cell in fluid communication with the first reactor and the second reactor. The method further includes contacting the powders within the first reactor and the second reactor with the electrolyte to dissolve at least one base metal from each reactor into the electrolyte and reduce at least some of the ferric ions to ferrous ions. The ferrous ions are oxidized at an anode of the electrochemical cell to regenerate the ferric ions. The powder within the second reactor comprises a higher weight percent of the at least one base metal than the powder in the first reactor. Additional methods of recovering metals from electronic waste are also described, as well as an apparatus of recovering metals from electronic waste.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Development for Integrated System-Level Analysis Capabilities in SAM for Molten Salt Reactors

In recent years, there has been renewed interest in Molten Salt Reactors (MSRs) for their potential advantages compared to reactors that rely on solid fuel. In response to such interest, many methods and codes have been developed to capture the unique features of MSRs. Among them, the System Analysis Module (SAM) is a modern system analysis tool that provides fast-running, modest-fidelity, whole-plant transient analysis capabilities, essential for fast-turnaround design scoping and engineering analyses of advanced reactor concepts. For liquid-fuel MSRs, the complex physics and chemistry involved in MSR operation—such as reactor kinetics, fluid flow, heat transfer, and salt composition dynamics—pose significant challenges for system-level modeling. Specific modeling capabilities are needed for system-level transient simulation. This paper presents recent advancements in SAM capability enhancements for system-level modeling of MSRs, focusing on improved simulation fidelity, computational efficiency, and multi-physics integration. Key enhancements include the development of species transport, Delayed Neutron Precursor (DNP) drift, modified Point Kinetics Equations (PKE), decay heat modeling, key fission product behavior, salt corrosion, and thermal-hydraulic coupling, as well as code robustness and performance enhancements for MSR applications. The code enhancement allows for better predictive accuracy in safety analysis, transient behavior, and operational optimization, thus supporting the design and licensing of next-generation MSRs. Results from case studies are presented to demonstrate the benefits of these enhancements in accurately capturing key reactor transient behaviors.

Hu, Rui (ORCID:0000000237712920)↗
DOE OSTI2021

Integrating Advanced Modeling and Accelerated Testing for a Modernized Fuel Qualification Paradigm

With the increasing interest in sodium fast reactor technology, as seen by applications to the U.S. Nuclear Regulatory Commission for the OKLO Aurora plant, fuel testing for the TerraPower Traveling Wave Reactor, and the impending construction and startup of the versatile test reactor (VTR), a modernized, accelerated approach to fuel qualification is needed. To guide this effort, a Phenomena Identification Ranking Table–styled analysis was performed for a U-Pu-Zr sodium-free annular fuel system. This analysis evaluated a series of fuel design properties and parameters against their contributions to key fuel performance phenomena. The resulting priority parameters were then reviewed against existing modeling and experimental capabilities to support investigation of the highest-priority parameters. A pathway for qualification was then established using highthroughput, high-volume experiments from MiniFuel and FAST in parallel with advanced physics-based model development. This effort outlines how the first stages of qualification can be reduced from the typical 20+-year development cycle to 5 to 7 years by deploying accelerated irradiation testing platforms. As with any accelerated test, these methods are prototypic in some aspects and less so in others; however, by coupling with advanced fuel performance modeling and simulation capabilities, the larger space of irradiation parameters and material response provided offers advantages for the validation of physics-based models supporting the deployment of novel fuel designs. As a test case, this paper utilizes a proposed Mark II fuel system for the upcoming VTR. Thus, an accelerated qualification method can be tested for the development of MARK II driver fuel so that by the time of VTR startup, lead test assemblies for a Mark II fuel can be initiated.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Making sense of uncertain nuclear data

At its core, and in an imperfect world, the scientific method relies heavily on the concept of uncertain knowledge. Understanding the nature and magnitude of our lack of knowledge on nuclear data has important ramifications on a wide range of nuclear technologies that encompass defense, energy, medicine, astrophysics, among others. Over the span of a long and fruitful carrier, Dr. Massimo Salvatores has made significant contributions to the fields of nuclear reactor physics, nuclear data, and uncertainty quantification and propagation. Advances in machine learning techniques and computational capabilities are enabling a new level of confidence in our ability to predict key nuclear metrics as well as their associated uncertainties across many applications. Here this short article reviews some of those concepts with an eye towards developing a new paradigm to provide nuclear data libraries of benefit to various applications.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Development and Implementation of CFD-Informed Models for the Advanced Subchannel Code CTF

The research described in this PhD thesis contributes to the development of efficient methods for utilization of high-fidelity models and codes to inform low-fidelity models and codes in the area of nuclear reactor core thermal-hydraulics. The objective is to increase the accuracy of predictions of quantities of interests using high-fidelity CFD models while preserving the efficiency of low-fidelity subchannel core calculations. An original methodology named Physics- based Approach for High-to-Low Model Information has been further developed and tested. The overall physical phenomena and corresponding localized effects, which are introduced by the presence of spacer grids in light water reactor (LWR) cores, are dissected in corresponding four building basic processes, and corresponding models are informed using high-fidelity CFD codes. These models are a spacer grid-directed cross-flow model, a grid-enhanced turbulent mixing model, a heat transfer enhancement model, and a spacer grid pressure loss model. The localized CFD-models are developed and tested using the CFD code STAR-CCM+, and the corresponding global model development and testing in sub-channel formulation is performed in the thermal- hydraulic subchannel code CTF. The improved CTF simulations utilize data-files derived from CFD STAR-CCM+ simulation results covering the spacer grid design desired for inclusion in the CTF calculation. The current implementation of these models is examined and possibilities for improvement and further development are suggested. The validation experimental database is extended by including the OECD/NRC PSBT benchmark data. The outcome is an enhanced accuracy of CTF predictions while preserving the computational efficiency of a low-fidelity subchannel code.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Recent Testing of ENDF/B-VIII.0 at ORNL [Slides]

This presentation provides an overview of comparisons of ENDF/B-VII.1 and ENDF/B-VIII.0 and uses VALID suite to test >600 cases with ENDF/B-VII.1 and ENDF/B-VIII.0 in KENO V.a/KENO-VI, although a limited number of cases have been tested with TSUNAMI. It provides a results summary of KENO V.a/KENO-VI with a pooled results summary. The presentation also discusses changes for LCT experiments, MCT experiments, MST experiments, and PST experiments. A few words about graphite are also provided and the presentation states that graphite performance investigated across ENDF/B-VII.0, ENDF/B-VII.1, and ENDF/B-VIII.0 for MSR and HTGR models. A brief summary of the results is presented to make sure the Validation Committee is aware of the work. Also presented is an Eigenvalue comparison with a graphite moderated system, and comparison between an HTR-10 pebble model ENDF/B-VII.1 vs. ENDF/B-VIII.0, which provides different graphite data. In conclusion, ORNL has recently completed a comparison of ENDF/B-VIII.0 and ENDF/B-VII.1 using the VALID library. Work is also being performed for advanced reactor applications. Other efforts on-going and they are also trying to investigate more intermediate spectrum cases.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗