Data-Driven Integration of High Fidelity Fluid Dynamics into Systems Codes for Design and Safety Applications
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22 GENERAL STUDIES OF NUCLEAR REACTORS↗
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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.
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Use of VERA to Improve Baffle/Reflector Data Generation for the Westinghouse Core Design Code System
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Zhang, Baocheng↗
FORTRAN code-evaluation system
Automated code evaluation system can be used to detect coding errors and unsound coding practices in any ANSI FORTRAN IV source code before they can cause execution-time malfunctions. System concentrates on acceptable FORTRAN code features which are likely to produce undesirable results.
Capps, J. D.↗
Space System Architecture Code
Space System Architecture Code (SSAC) program performs sensitivity studies on operating parameters of space missions. SSAC examines effects of variations in sizes of crews of piloted missions, chemical vs. nuclear electric propulsion, and specific impulse. Written in FORTRAN 77.
Jan, Darrell L.↗
Design and performance of block-coded communication systems.
Block coded frequency multiplexed PM COMMUNICATION system design and performance, considering simplification, cost and weight
COMMUNICATIONS SATELLITE↗
Investigating the Role of Accident Tolerant Cladding on Source Term Reduction for High-Burnup PWRs Using MELCOR
The use of accident tolerant fuel (ATF) cladding can increase coping times during and beyond design basis accidents. While such gains may be incremental, they provide a margin that can potentially be recovered to enable high-burnup (HBU) operation. Realizing such a margin requires demonstrating that the combination of HBU and ATF has not led to an overall increase in source term. This study investigates the influence of cladding technology (Zr-based, Cr-coated Zr, and FeCrAl) and fuel cycle length (18 and 24 months) on radiological dose at the boundary of the exclusion zone for a four-loop pressurized water reactor to investigate whether ATF claddings can provide such benefits. We analyze a recovered large break loss-of-coolant accident scenario to investigate the impact of transient timescale on the benefits of such coping time increases. The simulations have been performed using the MELCOR and MELCOR Accident Consequence Code System codes. For the cases analyzed, increased fuel cycle length did not necessarily increase radionuclide release and hydrogen generation, as these were found to be sensitive to the core power distribution. Similarly, off-site dose consequence is dominated by short-lived radionuclides that tend to saturate earlier in the burnup, so higher burnup operation did not necessarily increase the source term for the phenomena and transients analyzed here. Delays in recovery of the lowpressure safety injection system increase hydrogen production and radionuclide release, especially between 780 s and 1620 s, due to the nonlinear oxidation and core degradation behavior. Results show that Cr-coated Zr enhances safety by delaying heatup and gap release. Here, when uncertainty propagation on oxidation properties is considered, FeCrAl exhibits the lowest overall radionuclide release and off-site dose throughout the spectrum. However, while the considered “base model” performance is superior under delayed injection scenarios, upper-bound cases display hydrogen generation risk comparable to the Zr-based cladding.
Accident Tolerant Fuel↗
Implementation of a Drift Flux Model into SAM with Development of a Verification and Validation Test Suite for Modeling of Noncondensable Gas Mixtures
The advanced thermal-hydraulic system code, System Analysis Module (SAM), was originally developed for the modeling of single-phase flow in advanced reactors. It has since been expanded to include a four-equation drift flux model for the modeling of two-phase flows containing a noncondensable gas. The model was expanded to support the modeling of molten salt reactor (MSR) designs in which the fuel is directly dissolved in the circulating coolant. These designs have shown that circulating gas bubbles can play an important role in the management of fission products and the operational behavior of the reactor. A drift flux model was implemented to more accurately capture the localized behavior of the void in the core and its impact on the mass transfer of fission products. A thorough assessment of the new model was performed by developing a verification and validation test suite. Verification problems were designed to test all major terms in the new governing equations. The new model converged to the correct solution at the expected order of accuracy for all verification cases. The validation cases included a wide range of flow and void conditions in different pipe geometries. Although higher void experiments show a slight underprediction of void by the drift flux model, experiments that aim to reproduce Molten Salt Reactor Experiment (MSRE) experimental conditions show good agreement with the model. The gas transport model was activated for a SAM model of the MSRE to demonstrate that it can be used in a more complex model. Finally, this gas transport model will be used along with an interfacial area transport equation being implemented in SAM for the prediction of mass transport behavior in MSR conditions.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Economic Model for Estimation of GDP Losses in the MACCS Offsite Consequence Analysis Code
The MACCS (MELCOR Accident Consequence Code System) code is the U.S. Nuclear Regulatory Commission (NRC) tool used to perform probabilistic health and economic consequence assessments for atmospheric releases of radionuclides. It is also used by international organizations, both reactor owners and regulators. It is intended and most commonly used for hypothetical accidents that could potentially occur in the future rather than to evaluate past accidents or to provide emergency response during an ongoing accident. It is designed to support probabilistic risk and consequence analyses and is used by the NRC, U.S. nuclear licensees, the Department of Energy, and international vendors, licensees, and regulators. This report describes the modeling framework, implementation, verification, and benchmarking of a GDP-based model for economic losses that has recently been developed as an alternative to the original cost-based economic loss model in MACCS. The GDP-based model has its roots in a code developed by Sandia National Laboratories for the Department of Homeland Security to estimate short-term losses from natural and manmade accidents, called the Regional Economic Accounting analysis tool (REAcct). This model was adapted and modified for MACCS and is now called the Regional Disruption Economic Impact Model (RDEIM). It is based on input-output theory, which is widely used in economic modeling. It accounts for direct losses to a disrupted region affected by an accident, indirect losses to the national economy due to disruption of the supply chain, and induced losses from reduced spending by displaced workers. RDEIM differs from REAcct in its treatment and estimation of indirect loss multipliers, elimination of double counting associated with inter-industry trade in the affected area, and that it is designed to be used to estimate impacts for extended periods that can occur from a major nuclear reactor accident, such as the one that occurred at the Fukushima Daiichi site in Japan. Most input-output models do not account for economic adaptation and recovery, and in this regard RDEIM differs from its parent, REAcct, because it allows for a user-definable national recovery period. Implementation of a recovery period was one of several recommendations made by an independent peer review panel to ensure that RDEIM is state-of-practice. For this and several other reasons, RDEIM differs from REAcct. Both the original and the RDEIM economic loss models account for costs from evacuation and relocation, decontamination, depreciation, and condemnation. Where the original model accounts for an expected rate of return, based on the value of property, that is lost during interdiction, the RDEIM model instead accounts for losses of GDP based on the industrial sectors located within a county. The original model includes costs for disposal of crops and milk that the RDEIM model currently does not, but these costs tend to contribute insignificantly to the overall losses. This document discusses three verification exercises to demonstrate that the RDEIM model is implemented correctly in MACCS. It also describes a benchmark study at five nuclear power plants chosen to represent the spectrum of U.S. commercial sites. The benchmarks provide perspective on the expected differences between the RDEIM and the original cost-based economic loss models. The RDEIM model is shown to consistently predict larger losses than the original model, probably in part because it accounts for national losses by including indirect and induced losses; whereas, the original model only accounts for regional losses. Nonetheless, the RDEIM model predicts losses that are remarkably consistent with the original cost-based model, differing by 16% at most for the five sites combined with three source terms considered in this benchmark.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Systems Improved Numerical Fluids Analysis Code
Systems Improved Numerical Fluids Analysis Code, SINFAC, consists of additional routines added to April, 1983, version of SINDA. Additional routines provide for mathematical modeling of active heat-transfer loops. Simulates steady-state and pseudo-transient operations of 16 different components of heat-transfer loops, including radiators, evaporators, condensers, mechanical pumps, reservoirs, and many types of valves and fittings. Program contains property-analysis routine used to compute thermodynamic properties of 20 different refrigerants. Source code written in FORTRAN 77.
Costello, F. A.↗
AutoBayes/CC: Combining Program Synthesis with Automatic Code Certification: System Description
Code certification is a lightweight approach to formally demonstrate software quality. It concentrates on aspects of software quality that can be defined and formalized via properties, e.g., operator safety or memory safety. Its basic idea is to require code producers to provide formal proofs that their code satisfies these quality properties. The proofs serve as certificates which can be checked independently, by the code consumer or by certification authorities, e.g., the FAA. It is the idea underlying such approaches as proof-carrying code [6]. Code certification can be viewed as a more practical version of traditional Hoare-style program verification. The properties to be verified are fairly simple and regular so that it is often possible to use an automated theorem prover to automatically discharge all emerging proof obligations. Usually, however, the programmer must still splice auxiliary annotations (e.g., loop invariants) into the program to facilitate the proofs. For complex properties or larger programs this quickly becomes the limiting factor for the applicability of current certification approaches.
Whalen, Michael↗
A Class of Codes for Multiple-access Satellite Communication Systems
Codes for multiple-access communication satellite systems using time-frequency multiplexing
COMMUNICATIONS SATELLITE↗
Self-synchronizing outer codes for the TDRSS decoder
In this paper, various aspects of developing a self-synchronizing coding system are discussed for NASA's TDRSS satellite system. The coding system used is a concatenation of a (2, 1, 7) inner convolutional code with a (255,223) Reed-Solomon outer code. The goal is to design the decoder in such a way that both symbol and word synchronization are achieved for the outer code without the need for any separate synchronization patten. The system proposed in this paper uses the properties of the inner (Viterbi) decoder to provide symbol synchronization for the outer code, and maintains word synchronization by making use of a coset code of the Reed-Solomon code.
Lin, S.↗
Need for advanced research reactors for the next-generation reactor physics, analysis tools, and technology
Full text of publication follows. There is an urgent need for design and deployment of advanced research and test reactors in support of design, licensing and operation of advanced power reactors and education of next generation nuclear workforce. Existing research reactors mostly were designed and constructed decades ago with the main objectives of training operators, performing reactor physics experiments, and educating nuclear engineers and scientists. There are already gaps and significant concern about future capabilities for the existing research reactor facilities to address modern instrumentation and/or flexible environments for: performing reactor physics studies for advanced designs which have significantly different core materials forms and compositions, reactor shapes and size; validation of advanced high-fidelity software; development of machine learning algorithms for enhancement of human-machine collaboration in support of reactor monitoring, operation and safeguards; and, effective education of the next-generation workforce. The authors will focus on the need for advanced research reactors to improve and validate fast and accurate simulation tools for high-fidelity modeling and analysis of nuclear reactors in support of their design, optimization, licensing, operation, and monitoring. In the past, simulation tools were limited to relatively coarse models using approximate methodologies that benefited from two main factors: i) allowance for large margins and tolerances; ii) ability to construct prototype (e.g., zero power) reactors for adjustment of approximate methodologies. The next generation reactors have to be designed mainly by using novel high-fidelity computational tools that are accurate and fast, and therefore can be used for parametric studies and uncertainty quantification. To sufficiently demonstrate the accuracy of these tools, advanced research reactors are needed. The authors argue the need for new computational paradigms such as the MRT (Multistage, Response- function Transport) methodology which has resulted in the development of the novel high-fidelity RAPID (Real-time Analysis for Particle-transport and In-situ Detection) code system. Such code systems have to be robust in modeling any complex system, and should be fast and accurate, henceforth their uncertainties can be quantified at reasonable costs. Again, advanced research reactors are needed for the validation of the fidelity and accuracy of new computational tools. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Differential pulse-code modulation
Systems encoding and decoding method allows data to be transmitted with less bandwidth than required for conventional system codes, and is not affected by data-transition density. In addition it requires no direct-current response of transition link and there is little ambiguity in resolution of digital data.
Herman, C. F.↗
Two Serial Data to Pulse Code Modulation System Interfaces
Two pulse code modulation (PCM) system interfaces for asynchronous serial data are described. One interface is for global positioning system (GPS) data on the NASA Dryden Flight Research Center (DFRC) F-15B (McDonnell Douglas Corporation, St. Louis, Missouri) airplane, tail number 836 (F-15B/836). The other is for flight control computer data on the duPont Aerospace (La Jolla, California) DP-1, a 53-percent scale model of the duPont Aerospace DP-2.
Hamory, Phil↗