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.

SourceResearch connectionFrom yearThrough year

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 487 records · Page 27

All sources NASA NTRS DOE OSTI

DOE OSTI2015

Status Report on NEAMS System Analysis Module Development

Under the Reactor Product Line (RPL) of DOE-NE’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, an advanced SFR System Analysis Module (SAM) is being developed at Argonne National Laboratory. The goal of the SAM development is to provide fast-running, improved-fidelity, whole-plant transient analyses capabilities. SAM utilizes an object-oriented application framework MOOSE), and its underlying meshing and finite-element library libMesh, as well as linear and non-linear solvers PETSc, to leverage modern advanced software environments and numerical methods. It also incorporates advances in physical and empirical models and seeks closure models based on information from high-fidelity simulations and experiments. This report provides an update on the SAM development, and summarizes the activities performed in FY15 and the first quarter of FY16. The tasks include: (1) implement the support of 2nd-order finite elements in SAM components for improved accuracy and computational efficiency; (2) improve the conjugate heat transfer modeling and develop pseudo 3-D full-core reactor heat transfer capabilities; (3) perform verification and validation tests as well as demonstration simulations; (4) develop the coupling requirements for SAS4A/SASSYS-1 and SAM integration.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Interfacing VPSC with finite element codes. Demonstration of irradiation growth simulation in a cladding tube

This Milestone report shows good progress in interfacing VPSC with the FE codes ABAQUS and MOOSE, to perform component-level simulations of irradiation-induced deformation in Zirconium alloys. In this preliminary application, we have performed an irradiation growth simulation in the quarter geometry of a cladding tube. We have benchmarked VPSC-ABAQUS and VPSC-MOOSE predictions with VPSC-SA predictions to verify the accuracy of the VPSCFE interface. Predictions from the FE simulations are in general agreement with VPSC-SA simulations and also with experimental trends.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Summary Report Of The FY25 Reactor Physics Verification And Validation Exercises In The Advanced Reactor Technologies - Gas-cooled Reactor Program

Valdiation and verification of numerical tools is critical for ensuring reasonable predictions for design scoping, licensing, and safety analsyis. In this report, two reactor physics verification and validation exercises are presented. The first of these exercises focuses on burnup analysis with data from the Advanced Gas Reactor (AGR) program. Simulations are performed with Monte Carlo N-Particle (MCNP) and are compared with the experimental measurements for the AGR 1 and 2 experiments that utilize both UCO and UO2 fuel. The second exercises utilizes data from the HTR-Proteus experiments to perform reactor physics validation. Specifications of the experimental facility are provdied, along with a demonstration of initial modeling efforts in Serpent for one of the determistic packing experiments. Both cases are part of the Generation-IV international forum (GIF) Very High-Temperature Reactor (VHTR) Computational Methods, Validation, and Benchmarking (CMVB) program, an international collaborative organization dedicated to the verification and validation of High-Temperature Gas-Cooled Reactor (HTGR) analysis. Participation in the CMVB allows the US Department of Energy (DOE) to leverage these existing validation activities to provide extra value through benchmarking activities with other CMVB members.

and Benchmarking (CMVB) program↗

DOE OSTI2017

SOLVING CONJUGATE HEAT TRANSFER PROBLEM IN TWO-PHASE FLOW USING NEWTON-KRYLOV METHOD

In this paper, we successfully demonstrated an application of the Jacobian-free Newton Krylov (JFNK) method in solving the conjugated two-phase channel flow and heat conduction problem, in a fully coupled and fully implicit manner. Such a problem represents the most fundamental piece of safety analysis codes, for example the heat transfer between fuel rods and reactor coolant during postulated accidental scenarios. The two-phase flow is modeled with the widely accepted one-dimensional two-fluid two-phase flow model, and the heat conduction is modeled with two-dimensional heat conduction model. The coupling between these two physics is realized through the wall heat transfer term in fluid equations and heat flux boundary conditions of heat conduction model. For spatial discretization, both the first-order and second- order upwind schemes are considered for flow equations, while the standard central differencing method is used for the heat conduction model. For time integration, both the first-order backward Euler (BDF1) and second-order BDF2 schemes are considered. Several treatments were also introduced to enhance the robustness of the JFNK method in solving such a problem that is discontinuous in nature because of the discrete closure correlations. These treatments include the re-introduction of a transition region between the slug and annular mist flow regimes, and a simplification in the interfacial drag term. Numerical results have been extensively validated with experimental data in both vertical pipe and rod bundle geometries with a wide range of thermal-hydraulics conditions, including pressure, inlet subcooling, wall heat flux, outlet void fraction, and etc. Further benchmark with the RELAP5-3D code confirms the correct implementation of wall heat transfer models.

42 ENGINEERING↗

DOE OSTI2016

Neutron Resonance Theory for Nuclear Reactor Applications: Modern Theory and Practices.

The neutron resonance phenomena constitute one of the most fundamental subjects in nuclear physics as well as in reactor physics. It is the area where the concepts of nuclear interaction and the treatment of the neutronic balance in reactor fuel lattices become intertwined. The latter requires the detailed knowledge of resonance structures of many nuclides of practical interest to the development of nuclear energy. The most essential element in reactor physics is to provide an accurate account of the intricate balance between the neutrons produced by the fission process and neutrons lost due to the absorption process as well as those leaking out of the reactor system. The presence of resonance structures in many major nuclides obviously plays an important role in such processes. There has been a great deal of theoretical and practical interest in resonance reactions since Fermi’s discovery of resonance absorption of neutrons as they were slowed down in water. The resonance absorption became the center of attention when the question was raised as to the feasibility of the self-sustaining chain reaction in a natural uranium-fueled system. The threshold of the nuclear era was crossed almost eighty years ago when Fermi and Szilard observed that a substantial reduction in resonance absorption is possible if the uranium was made into the form of lumps instead of a homogeneous mixture with water. In the West, the first practical method for estimating the resonance escape probability in a reactor cell was pioneered by Wigner et al in early forties.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2026

Alchemy: A Model-Based Approach for 2D to 3D Autonomous Nuclear System Design

Engineering design of nuclear power plant (NPP) piping and equipment systems frequently bypasses crucial 2D system planning, instead moving straight to 3D modeling. This often leads to designs that exceed building envelope constraints, forcing expensive and time-consuming redesigns. When 2D modeling is employed, it typically involves labor-intensive manual workflows that convert 2D drawings into 3D models, resulting in inefficiencies and errors across design iterations. These workflows further suffer from poor software interoperability and dependence on proprietary software ecosystems, thereby contributing to schedule delays and cost overruns. This paper presents Alchemy, an autonomous framework that transforms 2D system definitions into Industry Foundation Classes (IFC)-compliant 3D building information models (BIMs) for expediting nuclear facility design at the conceptual preliminary phase. Using a model-based approach, the framework treats the 2D system diagram as the central reference model employed to automatically generate all subsequent outputs, ensuring consistency between the system definition and the resulting physical design. A web-based interface enables engineers to define hierarchical system topologies including associated equipment, geometric properties, and connectivity requirements. A two-phase equipment layout optimization algorithm automatically computes collision-free spatial configurations within predefined building envelopes. An artificial intelligence (AI)-assisted pipe routing module then generates orthogonal, collision-free routing paths, allowing the user to select either an A* search-based method or an Ant Colony Optimization (ACO)-based method. All outputs are authored natively in IFC format, relying on open-source technologies and standardized formats in order to ensure extensibility and eliminate proprietary software dependencies. The proposed framework is validated on two representative pressurized-water reactor (PWR)-based case studies, for which it autonomously generates IFC-compliant 3D models in minutes, drastically reducing workflows that typically require hours of manual effort. The generated model demonstrates topologically correct equipment placement, physically plausible spatial relationships, and collision-free pipe routing consistent with known PWR loop configurations. This work represents a foundational step toward digital engineering for nuclear facility preliminary design, with future ongoing development targeting design code compliance and expanded system complexity.

97 - MATHEMATICS AND COMPUTING↗

DOE OSTI2016

Stability and accuracy of 3D neutron transport simulations using the 2D/1D method in MPACT

A consistent “2D/1D” neutron transport method is derived from the 3D Boltzmann transport equation, to calculate fuel-pin-resolved neutron fluxes for realistic full-core Pressurized Water Reactor (PWR) problems. The 2D/1D method employs the Method of Characteristics to discretize the radial variables and a lower order transport solution to discretize the axial variable. This paper describes the theory of the 2D/1D method and its implementation in the MPACT code, which has become the whole-core deterministic neutron transport solver for the Consortium for Advanced Simulations of Light Water Reactors (CASL) core simulator VERA-CS. Several applications have been performed on both leadership-class and industry-class computing clusters. Results are presented for whole-core solutions of the Watts Bar Nuclear Power Station Unit 1 and compared to both continuous-energy Monte Carlo results and plant data.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗

DOE OSTI2018

State of Electrical Cable Aging in U.S. Nuclear Power Plants

Nuclear energy has supplied one fifth of the electricity generation in the United States for the last thirty years, but headwinds for continued contribution of nuclear to the US energy mix include the aging state of existing plants and the challenge of remaining cost competitive with energy alternative such as natural gas. As an integral component of the power, control and safety operations of a nuclear power plant, electrical cables are a significant factor in both long term and resource-efficient plant operation. The cost and practical challenge of wholesale replacement of the thousands of miles of cables in each reactor make effective management of the existing cable inventory increasingly important as the cables age in service. Total generation costs for US nuclear reactors have increased in the last decade, due in part to safety and security upgrades. Passive, long-lived structures, systems and components may be subject to aging-related degradation, necessitating increase frequency of inspection and repair/replace decisions. Costs associated with cable aging management in long-term operation can be reduced with effective cable condition monitoring technology and sufficient understanding of material aging behavior. Technical basis for confidence in cable status can reduce the frequency of cable system inspections and align cable replacement decisions with physical need. Knowledge of cable remaining useful life can allow longer term cable maintenance planning and fewer unplanned outages. As methods for relating measurable key indicators of aging with cable status develop, industry has begun to demonstrate how informed cable aging management can help to minimize costs necessary for safe, long-term operation.

light water reactors, electrical cables, nuclear p↗

DOE OSTI2015

Uncertainty quantification and propagation in nuclear density functional theory

Nuclear density functional theory (DFT) is one of the main theoretical tools used to study the properties of heavy and superheavy elements, or to describe the structure of nuclei far from stability. While on-going eff orts seek to better root nuclear DFT in the theory of nuclear forces, energy functionals remain semi-phenomenological constructions that depend on a set of parameters adjusted to experimental data in fi nite nuclei. In this study, we review recent eff orts to quantify the related uncertainties, and propagate them to model predictions. In particular, we cover the topics of parameter estimation for inverse problems, statistical analysis of model uncertainties and Bayesian inference methods. Illustrative examples are taken from the literature.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2018

Design and Simulation for Pre-amplifier of ITER Radial X-ray Camera

The radial X-ray camera, which is used to measure magnetohydrodynamic phenomenon and monitor edge localised mode events, is a key diagnostic system for physics studies in the international thermonuclear experimental reactor (ITER) tokamak. Considering the special requirements of physics measurements and spatial resolution, 144 detector channels are required. The board card of the pre-amplifier is designed to improve the integration, on which 16 signal channels have been integrated. In order to suppress the noise and improve the stability of the pre-amplifier, the target impedance analysis method is applied in the design of the power supply, the trans-impedance amplifier and a two-stage active low pass filter are applied in the design. Simulation with Cadence OrCAD and ANSYS SIwave has been adopted to optimize the characteristics of the pre-amplifier. A test of the pre-amplifier shows its good performance for ITER application. After optimization of the pre-amplifier, the band width is 270 kHz, the amplification is 10{sup 7} V/A, and the noise attenuation is − 30 dB, the noise of the pre-amplifier is reduced to 10 mV (noise of the power supply is included).

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2014

LLNL Final Design for PDV Measurements of Godiva for Validation of Multi-Physics Simulation

This document is a Final Design (CED-2) Report for IER-268, “PDV Measurements of Godiva for Validation of Multi-Physics Simulation”. These measurements will measure surface velocities in several locations, the initial α and α-t curve as a function of fission yield using existing LLNL assets including: The portable Photonic Doppler Velocimeter (PDV) detector; The “alpha box”, and; Aluminum-encapsulated 235 U fission foils. These experiments will be simulated using multi-physics methods funded separately under LLNL-AM2. A supercritical benchmark specification will be developed; and, if funding permits, an ICSBEP (or classified equivalent) evaluation will be published.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

Release on the Virtual Test Bed of an MSRE thermal hydraulics model

The nuclear industry is taking leaps in innovations with companies seeking a sustainable energy future through advanced nuclear reactors. The \gls{DOE}’s \gls{neams} program seeks to substantiate and bolster the deployment of advanced reactors through flexible multifidelity, multiphysics simulations of advanced nuclear reactors. Applications like SAM for one-dimensional systems thermal-hydraulics, and Pronghorn for multidimensional coarse mesh thermal-hydraulics, are geared to support innovations in industry by facilitating design, optimization, and licensing of advanced nuclear reactors. Coupling systems thermal-hydraulics and computational fluid dynamics codes can be difficult as the pressure coupling converges slowly; however, it is important to obtain the desired accuracy in each part of the primary loop. The authors of this model created an \gls{odc}~\cite{Mau23} approach to coupling SAM and Pronghorn. Leveraging this coupling technique, a \gls{msre} model was developed and released to the \gls{neams}/\gls{nric} \gls{vtb}. The \gls{msre} was chosen to be modeled because of the wealth of experimental data available and because of the strong physics coupling between the core and primary circuit \cite{doi:10.13182/NT8-2-118}. This document contextualizes the history of the \gls{msre}, describes the thermal hydraulics models used, and detail the implementation of multidimensional thermal-hydraulics and system codes based on the \gls{odc} method~\cite{Penn} for the \gls{msre} model. Finally, this document presents how other modelers could apply the SAM-Pronghorn \gls{odc} for other advanced reactor models. Current item is the set of slides for ANS Winter 23. The release of the model and the ANS summary have already been approved

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

Release on the Virtual Test Bed of an MSRE thermal hydraulics model

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2015

Focusing adaptive-optics for neutron spectroscopy at extreme conditions

Neutron Spectroscopy employing extreme-conditions sample environments is nowadays a crucial tool for the understanding of fundamental scientific questions as well as for the investigation of materials and chemical-physical properties. For all these kinds of studies, an increased neutron flux over a small sample area is needed. The prototype of a focusing neutron guide component, developed and produced completely at the neutron source FRM II in Garching (Germany), has been installed at the time-of-flight (TOF) disc-chopper neutron spectrometer TOFTOF and came into routine-operation. The design is based on the compressed Archimedes' mirror concept for finite-size divergent sources. It represents a unique device combining the supermirror technology with Adaptive Optics, suitable for broad-bandwidth thermal-cold TOF neutron spectroscopy (here optimized for 1.4–10 Å). It is able to squeeze the beam cross section down to a square centimeter, with a more than doubled signal-to-background ratio, increased efficiency at high scattering angles, and improved symmetry of the elastic resolution function. We present a comparison between the simulated and measured beam cross sections, as well as the performance of the instrument within real experiments. This work intends to show the unprecedented opportunities achievable at already existing instruments, along with useful guidelines for the design and construction of next-generation neutron spectrometers.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

DOE OSTI2017

Mongoose Methods and Theory

The Consortium for Advanced Simulation of Light Water Reactors (CASL) has developed a suite of high-fidelity software and methods that is capable of simulating the full operating history of a commercial pressurized water reactor (PWR), including the loading, depletion, shuffling, and discharge of all nuclear fuel assemblies. These tools, collectively known as the Virtual Environment for Reactor Applications, or VERA, include advanced solvers and multi-physics coupling algorithms that provide the most rigorous solutions available today for steady-state nuclear reactor analysis. To demonstrate this capability, VERA has recently been used to simulate over 18 years of operation of Watts Bar Nuclear Unit 1 (WBN1), and the results of these calculations are used to quantify its performance in an industry-grade benchmarking activity.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2019

Risk-informed Physical Security Using Dynamic Force-on-Force Modelling Tools (ABSTRACT)

Physical security requirement of nuclear power plants (NPPs) in the US is regulated in a mainly prescriptive manner. This “one size fits all” approach may consequently overlook plant-specific characteristics which are beneficial for the inherent safety and security of the plant. Although a performance-based security measure has been promulgated in 10 CFR 73.55(r), current evaluation methods generally emphasize static and conservative assumptions on physical security uncertainties. This conservatism contributes to the overall operational and maintenance costs, which account to 70% of total operating expenditures for an NPP. This paper describes an ongoing work at the Idaho National Laboratory within the Light Water Reactor Sustainability pathway, which aims to optimize physical protection system with regards to system effectiveness and dollar costs. To achieve this goal, a risk-informed approach to physical security is proposed. This approach leverages the Event Modeling Risk Assessment using Linked Diagrams (EMRALD) tool to model facility infiltration and Force-on-Force (FoF) engagement in a dynamic manner. The dynamic FoF analysis was conducted by coupling human performance uncertainties modeled in EMRALD to existing commercial 3-dimensional FoF software, i.e. SCRIBE3D, AVERT and Simajin. Meanwhile, risk analysis was accomplished by coupling the attack outcomes to existing plant Probabilistic Risk Assessment (PRA) model with FLEX strategy. This approach reduces the conservatisms in security modeling and further acknowledges the possibility for plant operators to mitigate the attack outcomes. A sabotage attack on a hypothetical NPP facility was devised to demonstrate the proposed methodology. Dynamic FoF rules were implemented in EMRALD coupled to SCRIBE3D. This dynamic evaluation revealed an additional 21% system effectiveness in protecting the target set relative to the static modelling approach. Furthermore, the conditional core damage probability could be reduced by up to 46% by crediting operator actions in mitigating the sabotage outcomes. We evaluated the relative importance of each component of the physical security posture included in the model and assessed tradeoffs between components to identify potential opportunities for cost reduction while maintaining a specific level of system effectiveness. Results of the cost analysis will be presented in the conference. This study is expected to help NPP operators in allocating resources optimally to satisfy the security requirements.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2018

Water Supply System with Light-Water Production Based on a Nuclear Desalination Complex

The aim of this work is to develop an ecologically safe energy efficient system of supplying water with production of light-water on the basis of a nuclear desalination complex with low-power reactors in order to solve the problem of water shortage, reduction of ecological load on the environment, and secure freshwater in regions where it is in short supply. A fundamentally new combined scheme of a water supply system with light-water produced on the basis of nuclear desalination and water circulation (recycling) using membrane and distillation methods is proposed. The system provides for maximum extraction of contaminants before the desalination stage by less expensive methods of purification, which decreases the load on desalination and makes it more efficient.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2016

Pn-FEM Method for Solving the First-Order Neutron Transport Equation

A deterministic method was investigated and developed for the photon-neutron transport calculation for fusion reactor blanket. Firstly, considering the fact that the second order transport equation cannot describe void region accurately due to the appearance of zero total cross section in the denominator, the first order equation was selected as the starting point of the solution. The angular discretization uses the spherical harmonics (Pn) method. Although discrete ordinates (Sn) method has been most commonly applied due to its efficiency, it suffers from the ray effect. To describe unstructured geometry, the finite element method was used to discrete spatial variable. It is well known that Bubnov-Galerkin discretization of the first-order equation is inappropriate which will create oscillations. To obtain robust solutions, Petrov- Galerkin methods have been developed such as the Streamline-Upwind Petrov-Galerkin (SUPG) method and the Generalized Least-Squares (GLS) method. These methods require a stabilization term added to its test functions, which will incur additional computational cost. Another finite element formulation that usually used for first order equation is the Discontinuous-Galerkin (DG) method. In this method, each element has a unique and separate set of basis function. Then the adjacent elements can be connected by using the in-coming flux. Its main drawback is the number of unknowns, much larger than Continuous-Galerkin (CG) method. The Sub-Grid Scale (SGS) finite element method combines the advantages of CG and DG methods together. It inserts the DG method in each element to correct the CG solution, and provides numerically stable solution of first-order equation without increasing the number of unknowns. The boundary conditions and interface conditions for Pn method is a hard nut to crack. We use Riemann method to deal with this problem. It decomposes the neutron angular flux into in-coming and out-going information through an interface, and applies boundary condition to the in-coming one. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗