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At least 919 records · Page 51
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Dakota Uncertainty Quantification Methods Applied to the CFD code Nek5000

This report presents the state of advancement of a Nuclear Energy Advanced Modeling and Simulation (NEAMS) project to characterize the uncertainty of the computational fluid dynamics (CFD) code Nek5000 using the Dakota package for flows encountered in the nuclear engineering industry. Nek5000 is a high-order spectral element CFD code developed at Argonne National Laboratory for high-resolution spectral-filtered large eddy simulations (LESs) and unsteady Reynolds-averaged Navier-Stokes (URANS) simulations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

CZT DTRA final report

The objective of the project is to understand the physical origin of electronic noise injected by the electrical contacts in CZT and CdTe, and moreover to understand how it impacts the current- voltage (IV) relationships of these materials. This understanding is critical to enabling the next crucial enhancement in the performance of CZT radiation detectors, as there have recently been impressive advancements in the growth of CZT crystals, particularly at our commercial partner Redlen Technologies. Redlen scientists have successfully reduced the size of the transport-inhibiting tellurium precipitates to be <3 micrometers, such that, with sufficiently high fields, it is possible to achieve resolution of <1% at 662 keV using suitable electrode geometries. In contrast to the excellent progress in crystal growth, practitioners in the field of radiation detection have been fabricating rather routine contacts on CZT for nearly two decades; there is no basic understanding of the semiconductor physics of the contacts, and consequently no breakthrough progress in this area. Our objective is to resolve this inadequacy in CZT diode fabrication on the basis of a science-based study, such that CZT detectors can achieve their full promise in performance as superior contacts will enable use of higher fields with lower leakage current – thereby enhancing the resolution that is possible while eliminating the well-known “tailing” effect suffered by the photopeak. Our approach is to develop methods that reduce or eliminate leakage currents in CZT devices through “engineering” the surfaces with novel treatments and structures. This includes using high density plasma etching, doping via ion implantation and metal diffusion, rapid thermal annealing, amorphous semiconductor and dielectric films, and controlled oxide growth. Using these methods, sources of injected and generated noise at the surface can be eliminated via plasma etching and film deposition or oxide growth, while advanced junctions (both homo- and heterojunctions) can be created to eliminate current injection from the metal contacts. The work involves the fabrication and characterization of CZT having a variety of contact structures. Tools capable of evaporating and sputtering metals, together with the usual lithographic and chemical processing methods, are employed. The CZT material employed will be of the highest quality available (Redlen Technologies), and we are engaging Prof. Burger of Fisk University, who is internationally recognized as one of the most accomplished materials physicists in the field on radiation detectors. We have also begun collaborating with Prof. Nicholas Kioussis of Cal State Northridge, Srivananthan Labs (Bolingbrook, IL), and Mark Amman of Lawrence Berkeley National Laboratory, all funded separately by DTRA.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Advanced Seismic Probabilistic Risk Assessment Methodology: Development of Beta 1.0 MASTODON Toolset

Design of nuclear power plant (NPP) facilities to resist natural hazards has been a part of the regulatory process from the beginning of the NPP industry in the United States, but has evolved substantially over time. The original set of approaches and methods was entirely deterministic in nature and focused on a traditional engineering margins-based approach. However, over time probabilistic and risk-informed approaches were also developed and implemented in US Nuclear Regulatory Commission guidance and regulation. A defense-in-depth framework has also been incorporated into US regulatory guidance over time. As a result, today, the US regulatory framework incorporates deterministic and probabilistic approaches for a range of different applications and for a range of natural hazard considerations. Although the US regulatory framework has continued to evolve over time, the tools, methods and data available to the US nuclear industry to meet the changing requirements have not kept pace. Notably, there is room for improvement in the tools and methods available for external event probabilistic risk assessment (PRA), which is the principal assessment approach used in risk-informed regulations and risk-informed decision-making applied to natural hazard assessment and design. Development of a new set of tools and methods that incorporate current knowledge, modern best practice, and state-of-the-art computational resources would lead to more reliable assessment of facility risk and risk insights (e.g., the plant elements and accident sequences that are most risk-significant), with less uncertainty and reduced conservatisms. Development of the next generation tools and methods for external events PRA is ongoing under the Risk Informed Safety Margins Characterization (RISMC) technical pathway. The RISMC toolkit development centers on integration of the tools and methods under a common framework, MOOSE. These tools and methods make use of existing and newly developed tools and methods, coupled with the experience and data gained in the past decades, to define and analyze more realistic risk assessment models. Specific focus in this report is on the capability to model seismic risk using advanced SPRA methods. Over the last year, significant capability has been added to Mastodon to simulate 3-D wave passage effects through nonlinear soil. Verification has demonstrated the capability of Mastodon to model wave passage effects in 1-D, 2-D, and 3-D. Methods have been developed to incorporate probabilistic floor response capabilities. Additional capability will be added to Mastodon over the next year to implement a robust gapping and sliding element for cyclic shaking, web-based verification and user manuals, stochastic finite elements, and frequency independent damping. Verification of added capabilities has occured in parallel with code writing activities.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Scan of an Unpublished Report: "Fission Product Behavior During Severe LWR Accidents: Recommendations for the MELCOR Code System"

This document provides a scanned version of a 1987 SAND report that was never formally published. However, this report was referenced within the MELCOR Reference Manual and, therefore, provides historical information and technical basis for the MELCOR code. This document is being made available to permit users of the MELCOR code access to the information. The title page has been edited to prevent any confusion with regards to the possible documentation identifiers, such as the SAND report number or the intended date of publication. Beyond these modifications, a cover, distribution, and back cover are prepended and appended to the document to conform to modern SAND report style guidelines. The first four chapters of this report were updated and released under the title "Fission Product Behavior During Severe LWR Accidents: Recommendations for the MELCOR Code System. Volume I" and were made available by the U.S. NRC through the Adams database under accession number ML19227A327. No prior release of the remaining content of this report has occurred.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Developing Generic Prior Distributions for Common Cause Failure Alpha Factors and Causal Alpha Factors

This report is a revision of the original report, INL/LTD-17-43723. Distribution of the original report was to the Nuclear Regulatory Commission (NRC) only, and the report was not made available to the public. The original report was revised as this report for public distribution. This report presents the latest update of generic prior distributions for common cause failure (CCF) alpha factors, as well as the development of new generic prior distributions for CCF causal alpha factors. The history of CCF treatment and parameter estimations is reviewed. The existing process for developing generic prior distributions is reviewed and used to develop new priors for CCF alpha factors and causal alpha factors. For causal alpha factors, different priors are developed for the five different CCF cause groups: Component (GC), Design (GD), Environment (GE), Human (GH), and Other (GO). These generic prior distributions could be used in the Standardized Plant Analysis Risk (SPAR) models for CCF parameter estimation. The issues and preliminary thoughts regarding prior distribution development are documented. Potential future work is then proposed for improving the process of developing priors.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Property Measurements of the NaCl-PuCl 3 Molten Salt System

Thermal properties of several compositions of the binary NaCl-PuCl 3 salt were measured to confirm the eutectic composition and provide high quality data for use by MSR developers. The PuCl 3 was generated by reducing PuO 2 to plutonium metal in the presence of calcium metal and then chlorinating the metallic plutonium with NH 4 Cl. The resulting PuCl 3 was used to make seven binary NaCl-PuCl 3 compositions with between 59.9 mol % PuCl 3 (Salt 1) and 20.0 mol % PuCl 3 (Salt 7). The eutectic temperature was measured to be 457 ± 4 °C by onset determination in DSC analyses of the seven mixtures, which is consistent with the values in the literature. Additional transitions were observed at approximately 332 and 363 °C in analyses of the PuCl 3 -rich compositions (Salts 1-3) at about 376 and 439 °C in analyses the Na-rich compositions (Salts 4-7). The solid state heat capacity decreased with increasing PuCl 3 content. The liquid state heat capacity was measured for Salt 3 (37.4 mol % PuCl 3 ) and Salt 4 (38.3 mol % PuCl 3 ). A higher heat capacity was measured for Salt 4 and measurements with both salts showed a positive correlation with temperature between 520 and 730 °C. Cells fabricated from nickel and molybdenum for use in DSC measurements at high temperatures could not be adequately sealed because these materials are not sufficiently malleable, even after high temperature annealing. It is recommended that future development focus on corrosion-resistant materials that have mechanical and thermal properties similar to the commercially available gold cells, such as Pt-Rh (80–20) and pure platinum. Those materials are chemically inert, soft, and malleable like gold, but have melting temperatures higher than 1000 °C (Rakhtsaum, 2013). Use of the Pt-Rh alloy should be evaluated first based on superior machinability for making precision parts and its current use in commercial DSC high temperature crucibles. The commercial Pt-Rh cells are not hermetically sealable, but it is expected that cells made from thinner stock can be sealed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Start-up Operation Experience with a Liquid Fluoride Salt Forced Convection Loop

A liquid fluoride salt forced convection test loop was constructed at Oak Ridge National Laboratory. Its unique features include a pebble bed test section, high-temperature instrumentation, a noncontact rotating gas seal for the pump, a silicon carbide flow tube, and a unique inductive heating technique. Initial startup and shake-down testing has been completed. This paper describes how several of the systems performed, highlights those that worked well, and discusses issues that arose during the start-up process. This discussion should be of interest to those who intend to develop molten salt technologies and who are interested in some of the specific techniques used in this experiment.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

CONSIDERATIONS FOR THE TREATMENT OF COMPUTERIZED PROCEDURES IN HUMAN RELIABILITY ANALYSIS

Computerized procedures (CPs) are an emerging technology within nuclear power plant control rooms. While CPs have been implemented internationally in advanced control rooms, to date no US nuclear power plant has implemented CPs in its main control room. Yet, CPs are a reality of new plant builds and are an area of considerable interest to existing plants, which see advantages in terms of easier records management by omitting the need for updating hardcopy procedures. The overall intent of this paper is to provide a characterization of human reliability analysis (HRA) issues for computerized procedures. It is beyond the scope of this document to propose a new HRA approach or to recommend specific methods or refinements to those methods. Rather, this paper serves as a review of current HRA as it may be used for the analysis and review of computerized procedures.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Baseline Evaluations to Support Control Room Modernization at Nuclear Power Plants

For any major control room modernization activity at a commercial nuclear power plant (NPP) in the U.S., a utility should carefully follow the four phases prescribed by the U.S. Nuclear Regulatory Commission in NUREG-0711, Human Factors Engineering Program Review Model. These four phases include Planning and Analysis, Design, Verification and Validation, and Implementation and Operation. While NUREG-0711 is a useful guideline, it is written primarily from the perspective of regulatory review, and it therefore does not provide a nuanced account of many of the steps the utility might undertake as part of control room modernization. The guideline is largely summative—intended to catalog final products—rather than formative—intended to guide the overall modernization process. In this paper, we highlight two crucial formative sub-elements of the Planning and Analysis phase specific to control room modernization that are not covered in NUREG-0711. These two sub-elements are the usability and ergonomics baseline evaluations. A baseline evaluation entails evaluating the system as-built and currently in use. The usability baseline evaluation provides key insights into operator performance using the control system currently in place. The ergonomics baseline evaluation identifies possible deficiencies in the physical configuration of the control system. Both baseline evaluations feed into the design of the replacement system and subsequent summative benchmarking activities that help ensure that control room modernization represents a successful evolution of the control system.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Overview of the U.S. DOE Fast Reactor Fuel Development Program

Research and development activities on fast reactor fuels in the US are focused on their potential use for actinide transmutation in future sodium fast reactors. As part of this application, there is a desire to demonstrate a multifold increase in burnup potential. A number of fuel design innovations are under investigation with a view toward significantly increasing the burnup potential of fuels, since higher discharge burnups equate to lower potential actinide losses during recycle. Promising innovations under investigation include: 1) lowering the fuel smeared density in order to accommodate the additional swelling expected as burnups increase, 2) utilizing an annular fuel geometry for better geometrical stability at low smeared densities, as well as the potential to eliminate the need for a sodium bond, and 3) minor alloy additions to immobilize lanthanide fission products inside the metallic fuel matrix and prevent their transport to the cladding resulting in fuel-cladding chemical interaction. This paper presents results from these efforts to advance fuel and cladding technology in support of high burnup and actinide transmutation objectives. Highlights include examples of fabrication of low smeared density annular metallic fuels, experiments to identify alloy additions effective in immobilizing lanthanide fission products, and early postirradiation examinations of annular metallic fuels having low smeared densities and palladium additions for fission product immobilization. Advanced cladding development activities include the investigation of high dose tolerant steel alloys. An overview of the U.S. Department of Energy fast reactor fuel research and development activities will be provided.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

DOE-ID-INL-20-219 R1

DOE-ID-INL-20-219 R1 for INL-21-219 R1.

(EBR)-II Modifications to Support National Reactor↗
DOE OSTI2026

ECAR-8848 Rev 2 MARVEL Primary Coolant System Lifting Analysis

The objective of this ECAR is to perform a lifting analysis of the Microreactor Applications Research Validation and Evaluation (MARVEL) Primary Coolant System (PCS) vertically, pickup from horizontal, the temporary lifting plate, and spreader.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2026

ECAR-6332 Rev 3 RELAP5-3D Thermal-Hydraulic Analysis of MARVEL Microreactor - Final Design

This document reports the thermal hydraulic analyses results for the MARVEL microreactor, final design, including dry criticality, normal operation, operational transients and a set of very-low probability transients caused by accident conditions. The ultimate scope of this document is to demonstrate the MARVEL microreactor thermal hydraulic performances and its inherent safety. First, the list of the operational and accidental transients with the corresponding acceptance criteria are recalled. Then, details of the final design, the key input parameters, the assumptions, and the methodology used for performing the deterministic safety analyses are presented. Finally, the analyses results are provided, demonstrating the satisfaction of the corresponding acceptance criteria.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2026

SPC-70791 Rev 0 MARVEL Inert Gas Subsystem, Secondary Coolant Management Subsystem, and Secondary Coolant Cover Gas Subsystem Tubing and Valve Specification

This specification covers the requirements and details for the selection, testing, cleaning, marking, shipping, inspection and installation (including welding) of the Inert Gas Subsystem, Secondary Coolant Management Subsystem, and Secondary Cover Gas Management Subsystem from the Valve Panel up to the classification division, the patch panel of the upper confinement. Including NaK, argon, eutectic gallium indium tin (eGaInSn), and helium fill tubing, valves, and hangers. These systems consist of the necessary straight lengths, elbows, reducers, fittings, tube connections, instrument taps, and valves called for by applicable drawings. All components shall be procured as ASME Section III Class 2 unless otherwise modified hereinafter. Pressure relief valves have been sized to prevent overpressure in the event there is full regulator flow. This is documented in ECAR-7365.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2025

SPC-70796 Rev 2 MetCar M-312 Bearings Procurement Specification

A. Idaho National Laboratory (INL) is developing a microreactor to produce electrical power utilizing a small nuclear core under the Microreactor Application Research Validation and Evaluation (MARVEL) program. This procurement specification defines the requirements for fabrication of the MARVEL MetCar M-312 bearings. B. Any conflict between this specification and referenced Codes and Standards, or any supplementary specifications in the procurement documents requires written clarification from the Contractor prior to proceeding with any work. Any deviation from the procurement documents requires approval by the Contractor with the change request process.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗