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At least 739 records · Page 41
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

RELAP5-3D Developer Guidelines and Programming Practices

Our ultimate goal is to create and maintain RELAP5-3D as the best software tool available to analyze nuclear power plants. This begins with writing excellent programming and requires thorough testing. This document covers development of RELAP5-3D software, the behavior of the RELAP5-3D program that must be maintained, and code testing. RELAP5-3D must perform in a manner consistent with previous code versions with backward compatibility for the sake of the users. Thus file operations, code termination, input and output must remain consistent in form and content while adding appropriate new files, input and output as new features are developed. As computer hardware, operating systems, and other software change, RELAP5-3D must adapt and maintain performance. The code must be thoroughly tested to ensure that it continues to perform robustly on the supported platforms. The coding must be written in a consistent manner that makes the program easy to read to reduce the time and cost of development, maintenance and error resolution. The programming guidelines presented her are intended to institutionalize a consistent way of writing FORTRAN code for the RELAP5-3D computer program that will minimize errors and rework. A common format and organization of program units creates a unifying look and feel to the code. This in turn increases readability and reduces time required for maintenance, development and debugging. It also aids new programmers in reading and understanding the program. Therefore, when undertaking development of the RELAP5-3D computer program, the programmer must write computer code that follows these guidelines. This set of programming guidelines creates a framework of good programming practices, such as initialization, structured programming, and vector-friendly coding. It sets out formatting rules for lines of code, such as indentation, capitalization, spacing, etc. It creates limits on program units, such as subprograms, functions, and modules. It establishes documentation guidance on internal comments. The guidelines apply to both existing and new subprograms. They are written for both FORTRAN 77 and FORTRAN 95. The guidelines are not so rigorous as to inhibit a programmer’s unique style, but do restrict the variations in acceptable coding to create sufficient commonality that new readers will find the coding in each new subroutine familiar. It is recognized that this is a “living” document and must be updated as languages, compilers, and computer hardware and software evolve.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Energetic Particle Physics In Fusion Research In Preparation For Burning Plasma Experiments

The area of energetic particle (EP) physics of fusion research has been actively and extensively researched in recent decades. The progress achieved in advancing and understanding EP physics has been substantial since the last comprehensive review on this topic by W.W. Heidbrink and G.J. Sadler. That review coincided with the start of deuterium-tritium (DT) experiments on Tokamak Fusion Test reactor (TFTR) and full scale fusion alphas physics studies. Fusion research in recent years has been influenced by EP physics in many ways including the limitations imposed by the "sea" of Alfven eigenmodes (AE) in particular by the toroidicity-induced AEs (TAE) modes and reversed shear Alfven (RSAE). In present paper we attempt a broad review of EP physics progress in tokamaks and spherical tori since the first DT experiments on TFTR and JET (Joint European Torus) including helical/stellarator devices. Introductory discussions on basic ingredients of EP physics, i.e. particle orbits in STs, fundamental diagnostic techniques of EPs and instabilities, wave particle resonances and others are given to help understanding the advanced topics of EP physics. At the end we cover important and interesting physics issues toward the burning plasma experiments such as ITER (International Thermonuclear Experimental Reactor).

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

Uncertainty Quantification Techniques for Sensor Calibration Monitoring in Nuclear Power Plants

This report describes research towards the development of advanced algorithms for online calibration monitoring. The objective of this research is to develop the next generation of online monitoring technologies for sensor calibration interval extension and signal validation in operating and new reactors. These advances are expected to improve the safety and reliability of current and planned nuclear power systems as a result of higher accuracies and increased reliability of sensors used to monitor key parameters. The focus of this report is on documenting the outcomes of the first phase of R&D under this project, which addressed approaches to uncertainty quantification (UQ) in online monitoring that are data-driven, and can therefore adjust estimates of uncertainty as measurement conditions change. Such data-driven approaches to UQ are necessary to address changing plant conditions, for example, as nuclear power plants experience transients, or as next-generation small modular reactors (SMR) operate in load-following conditions.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Risk-Informed Safety Margin Characterization Methods Development Work

This report summarizes the research activity developed during the Fiscal year 2014 within the Risk Informed Safety Margin and Characterization (RISMC) pathway within the Light Water Reactor Sustainability (LWRS) campaign. This research activity is complementary to the one presented in the INL/EXT-??? report which shows advances Probabilistic Risk Assessment Analysis using RAVEN and RELAP-7 in conjunction to novel flooding simulation tools. Here we present several analyses that prove the values of the RISMC approach in order to assess risk associated to nuclear power plants (NPPs). We focus on simulation based PRA which, in contrast to classical PRA, heavily employs system simulator codes. Firstly we compare, these two types of analyses, classical and RISMC, for a Boiling water reactor (BWR) station black out (SBO) initiating event. Secondly we present an extended BWR SBO analysis using RAVEN and RELAP-5 which address the comments and suggestions received about he original analysis presented in INL/EXT-???. This time we focus more on the stochastic analysis such probability of core damage and on the determination of the most risk-relevant factors. We also show some preliminary results regarding the comparison between RELAP5-3D and the new code RELAP-7 for a simplified Pressurized Water Reactors system. Lastly we present some conceptual ideas regarding the possibility to extended the RISMC capabilities from an off-line tool (i.e., as PRA analysis tool) to an online-tool. In this new configuration, RISMC capabilities can be used to assist and inform reactor operator during real accident scenarios.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2011

BPLU Completion and Verification Report

The Border Profile LU (BPLU) linear equation solver is the default solver for newer versions of RELAP5-3D. It can significantly reduce execution time compared to the previous default solver, MA18. Particularly for 3D cases, it can reduce run time by one to two orders of magnitude over MA18. However, because of some user reported failures, the MA18 solver currently must be used for coupled analyses. Over one dozen User Problems (UP) have been reported between 1999 and 2011 that involve the BPLU solver in RELAP5-3D. These issues can be combined into two categories of problems with the solver: • It fails when running multidimensional components with the nearly-implicit hydrodynamics advancement scheme. • It fails with some input models where the MA18 sparse solver does not fail. The sources of these UP have been found and corrected. The modified coding has been thoroughly tested with over 3000 test cases and on two different compute platforms. The updates are incorporated in RELAP5-3D, version 3.0.2.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Environmental, Transient, Three-Dimensional, Hydrothermal, Mass Transport Code - FLESCOT

The purpose of the project was to modify and apply the transient, three-dimensional FLESCOT code to be able to effectively simulate cesium behavior in Fukushima lakes/dam reservoirs, river mouths, and coastal areas. The ultimate objective of the FLESCOT simulation is to predict future changes of cesium accumulation in Fukushima area reservoirs and costal water. These evaluation results will assist ongoing and future environmental remediation activities and policies in a systematic and comprehensive manner.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Embedded Sensors and Controls to Improve Component Performance and Reliability - System Dynamics Modeling and Control System Design

This report documents the current status of the modeling, control design, and embedded control research for the magnetic bearing canned rotor pump being used as a demonstration platform for deeply integrating instrumentation and controls (I&C) into nuclear power plant components. This pump is a highly inter-connected thermo/electro/mechanical system that requires an active control system to operate. Magnetic bearings are inherently unstable system and without active, moment by moment control, the rotor would contact fixed surfaces in the pump causing physical damage. This report details the modeling of the pump rotordynamics, fluid forces, electromagnetic properties of the protective cans, active magnetic bearings, power electronics, and interactions between different dynamical models. The system stability of the unforced and controlled rotor are investigated analytically. Additionally, controllers are designed using proportional derivative (PD) control, proportional integral derivative (PID) control, voltage control, and linear quadratic regulator (LQR) control. Finally, a design optimization problem that joins the electrical, mechanical, magnetic, and control system design into one problem to balance the opposing needs of various design criteria using the embedded system approach is presented.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Unclassified Data Export from the Containment Database

The attached dataset is an unclassified subset of data copied from the CDB for release to a wider audience. It includes only the clearly unclassified columns from each of the tables in the CDB, and only the rows related to announced tests. Also included is a glossary that lists each of the tables and columns included in the export with its definition, historical labels, unit of measure, and reference information. Our objective is to make this unclassified dataset available for querying on an unclassified server through a repeatable process.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Supplement to a Methodology for Succession Planning for Technical Experts

This report complements A Methodology for Succession Planning for Technical Experts (Ron Cain, Shaheen Dewji, Carla Agreda, Bernadette Kirk, July 2017), which describes a draft methodology for identifying and evaluating the loss of key technical skills at nuclear operations facilities. This report targets the methodology for identifying critical skills, and the methodology is tested through interviews with selected subject matter experts.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Nuclear Power Plant Component Flooding Fragility Research

This report describes progress on Nuclear Power Plant component flooding fragility research conducted at the Component Flooding Evaluation Laboratory (CFEL). The progress includes additional steering committee membership and invitations, performance improvements to the Portal Evaluation Tank (PET), PET based experiments using industrial steel doors, pipe rupture spray pattern characterization, coupling the smoothed particle hydrodynamics (SPH) code Neutrino and the Idaho National Laboratory (INL) developed Risk Analysis Virtual Environment (RAVEN), continued development of component fragility mathematical models, and wave impact simulation device development, testing, and analysis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Consistent Fluid Property Evaluation in RELAP5-3D

Previous investigations have developed and tested a methodology to improve the consistency between used to determine the fluid property derivatives used in the numerical solution scheme and the solution scheme itself. Within a bounding rectangle of the fluid property table for a fluid state point, linear interpolants for both fluid property values and their derivatives are calculated. A weighted average between these interpolants the analytical approximations of the fluid properties and derivatives is formed with a weighting factor between 0.0 and 1.0. This average has shown potential to reduce mass error relating to the fluid properties in RELAP5-3D. This linear interpolation and weighted averaging methodology has been implemented in RELAP5-3D with two additional options that adjust the weighting factor in the average according to the location within the bounding rectangle. The first calculates the weighting factor based only on the position between the left and right internal energy endpoints. The second also factors the position between the upper and lower pressure grid points into the calculation of the weighting factor. The new coding passes verification testing and expanded testing of the various options. Graphs are presented from three kinds of test cases, a two volume basic model, Edward’s O’Brien blowdown, and typical PWR. Some trends of the effect of crossing fluid property table pressure and internal energy grid line are identified and discussed for the simple and Edwards test cases. Preliminary results show that it the consistent fluid methodology can reduce mass error in simple cases, but in more complex input models, there is no clear best.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

U-Mo Mechanical Properties Degradation with Irradiation

What are the U-Mo mechanical properties, and can the degradation of mechanical properties from irradiation be correlated with porosity? Part 1: What are the un-irradiated mechanical properties, and what is the influence of temperature and variations of wrought processing on the mechanical properties Part 2: What are the irradiated mechanical properties and can the degradation be correlated with increasing porosity from irradiation

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Complete Status Report Documenting the Processing, Microstructure and High-Temperature Mechanical Properties of the Nanostructured Ferritic Alloy OFRAC

This report provides a summary of the new nanostructured ferritic alloy, OFRAC (Oak Ridge Fast Reactor Advanced Fuel Cladding), that was developed for the advanced sodium fast reactor fuel cladding and core internals. The composition of OFRAC, which is Fe-12Cr-1Mo-03.Ti-0.3Nb + 0.3Y2O3 (wt. %), was chosen to achieve high temperature strength as well as high temperature creep performance and swelling resistance requirements that past structural materials such as austenitic stainless steels and ferritic/martensitic steels, i.e. HT-9, are not suitable as SFR cladding materials. The mechanical alloying procedures were described for producing OFRAC and the detailed characterization studies revealed uniformly distributed, high concentration of Y-Ti-O enriched oxide particles, or nanoclusters. The Nb addition that was intended to sequester interstitial C and N that are commonly introduced as contamination during ball milling of powder was found to be very effective in forming coarse size Nb carbonitrides along with excess Ti that removes these interstitial elements from the bcc-Fe matrix of OFRAC. The tensile tests on the OFRAC in the extruded condition showed excellent balance in strength and ductility from room temperature to 800°C. Both the microstructure and tensile properties were not significantly degraded after annealing the extruded OFRAC samples for 8 h at 1050°C and 1150°C. The latter temperature is ~76% of the melting point of OFRAC. These results demonstrated the stability of the microstructure and tensile properties to extremely high temperatures. The thermal creep resistance at elevated temperatures and stresses was investigated using strain rate jump testing. The results showed that OFRAC possessed high stress exponents at high stresses and temperatures of 550°C, 600°C, 700ºC and 800°C. This thermal dependence of minimum creep rate indicates that OFRAC will show quite low creep rates at higher temperatures and stresses compared to HT-9 and stainless steels. These significant improvements indicate that the NFA OFRAC alloy is a compelling and viable choice for sodium fast reactor fuel cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Initial High Temperature Inelastic Constitutive Model for Alloy 617

This report describes progress on a Alloy 617 inelastic constitutive model slated for inclusion in the ASME Boiler & Pressure Vessel Code in Code Case N-898 “Use of Alloy 617 (UNS N06617) for Class A Elevated Temperature Construction, Section III, Division 5” as part of an appendix on inelastic modeling. This appendix will provide a reference material model for A617 for use in conjunction with the ASME design by inelastic analysis criteria. This report focuses on the implementation and evaluation of an existing Alloy 617 model developed to represent the data collected at Idaho National Laboratory (INL) and Oak Ridge National Laboratory (ORNL) as part of the Alloy 617 Code Case experimental campaign. While the model accurately captures the cyclic stress/strain behavior of Alloy 617, it fails to capture the long-term creep and lower temperature tensile data. In addition, the model form is complicated and has over 50 configurable parameters. As such, we recommend the development of a simplified model, incorporating key features of the ORNL model but reducing the model complexity.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

FY20 Status Report on Creep Rupture Testing to Support the Development of Alloy 709 Code Case

This report provides an update on the status of the creep rupture testing on the first commercial heat of Alloy 709 in support of the Alloy 709 Code Case development. A creep test laboratory has been set up at ANL that comprises of 15 indirectly loaded test frames capable of testing ASTM-size specimens of the commercial heat of Alloy 709. The systems are equipped with high temperature furnaces capable of operation up to 1000°C in an air environment. During FY18, uniaxial ASTM-sized creep specimens were fabricated from 9 Alloy 709 plates that were processed by AOD, ESR, and ESR-homogenized routes and subsequently solution annealed at 1050, 1100, and 1150°C, and creep-rupture tests were performed at 330MPa and 600°C. The test results led to the down selection of the ESR-1100°C condition to support the code case. In FY19, 15 creep-rupture tests of the ESR-1100°C material, requiring the use of Type K thermocouples, were initiated to support the 100,000-hour Code Case. 8 tests were completed in FY19. In FY20, 6 tests were completed at the time of writing this report. In addition, 1 more test were loaded in FY20, using Type S thermocouples, and was completed. This report presents the data from the completed tests in FY20 and also the microstructural characterization of the fractured specimens. The creep data enabled the preliminary analysis of the creep behavior of this material. The characterization showed that in brittle fractures, the pores were narrow and thin, and elongated in the direction perpendicular to the loading direction, while in ductile fractures, the pores were much rounder. Dense precipitates were observed in all the samples, the size of which grew with test temperature. Some of the precipitates could be identified to be M23C6 based on their cuboid shape.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

An Initial Assessment of the Design Margins of Different ASME Section III, Division 5 Design Rules

This report develops a method for assessing the design margin of the ASME Section III, Division 5 Class A design rules, focusing on a definition of margin as the ratio between the actual, expected component life and the ASME design life. Full inelastic finite element simulations with a complete damage model capturing the available creep, creep-fatigue, fatigue, and tension test failure data produce the expected component lives, which can then be compared to corresponding ASME design calculations. The focus of this particular work is on Alloy 617 and the ASME creep-fatigue design rules, but the approach is general and could be applied to other design limits and to other materials. A margin assessment of a representative Alloy 617 component shows the ASME creep-fatigue design rules applied to this component have a margin of 10 on design life. The report describes several challenges in completing a more comprehensive margin assessment of the ASME rules, in particular challenges in developing probabilistic assessment methods for high temperature structural components. The report details these challenge and discusses the immediate possibilities of the deterministic margin assessment method developed here.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Initial microstructural model for creep-fatigue damage in Grade 91 steel

This report describes an initial Crystal Plasticity Finite Element Method (CPFEM) model for cyclic plasticity and damage in Grade 91 steel. The objective of this work is to develop a framework for modeling creep-fatigue interaction in Grade 91 steel to better predict the onset of damage in high temperature microreactor components. Many microreactor concepts envision low operating pressures but relatively high thermal stresses. Under these conditions, creep-fatigue will likely be the dominant design failure mechanism. Physically based models, like the one under development here, could lead to a better understanding of creep-fatigue mechanisms and the effect of stress multiaxiality and hold time on creep-fatigue damage. In turn, this could lead to more efficient microreactor component designs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗