skip to main content
OSTI.GOV title logo U.S. Department of Energy
Office of Scientific and Technical Information

Title: Assessment of the IVA3 code for multifield flow simulation. Formal report

Abstract

This report presents an assessment of the IVA3 computer code for multifield flow simulation, as applied to the premixing phase of a hypothetical steam explosion in a water-cooled power reactor. The first section of this report reviews the derivation of the basic partial differential equations of multifield modeling, with reference to standard practices in the multiphase flow literature. Basic underlying assumptions and approximations are highlighted, and comparison is made between IVA3 and other codes in current use. Although Kolev`s derivation of these equations is outside the mainstream of the multiphase literature, the basic partial differential equations are in fact nearly equivalent to those in other codes. In the second section, the assumptions and approximations required to pass from generic differential equations to a specific working form are detailed. Some modest improvements to the IVA3 model are suggested. In Section 3, the finite difference approximations to the differential equations are described. The discretization strategy is discussed with reference to numerical stability, accuracy, and the role of various physical phenomena - material convection, sonic propagation, viscous stress, and interfacial exchanges - in the choice of discrete approximations. There is also cause for concern about the approximations of time evolution in some heatmore » transfer terms, which might be adversely affecting numerical accuracy. The fourth section documents the numerical solution method used in IVA3. An explanation for erratic behavior sometimes observed in the first outer iteration is suggested, along with possible remedies. Finally, six recommendations for future assessment and improvement of the IVA3 model and code are made.« less

Authors:
Publication Date:
Research Org.:
Brookhaven National Lab., Upton, NY (United States)
Sponsoring Org.:
USDOE, Washington, DC (United States)
OSTI Identifier:
93743
Report Number(s):
BNL-52473; FZKA-5591
ON: DE95016260; TRN: 95:006274
DOE Contract Number:
AC02-76CH00016
Resource Type:
Technical Report
Resource Relation:
Other Information: PBD: Jul 1995
Country of Publication:
United States
Language:
English
Subject:
22 NUCLEAR REACTOR TECHNOLOGY; 99 MATHEMATICS, COMPUTERS, INFORMATION SCIENCE, MANAGEMENT, LAW, MISCELLANEOUS; WATER COOLED REACTORS; AUTOHYDROLYSIS; FLOW MODELS; EVALUATION; FLUID FLOW; COMPUTERIZED SIMULATION; EXPLOSIONS; I CODES; MATHEMATICAL MODELS

Citation Formats

Stewart, H.B. Assessment of the IVA3 code for multifield flow simulation. Formal report. United States: N. p., 1995. Web. doi:10.2172/93743.
Stewart, H.B. Assessment of the IVA3 code for multifield flow simulation. Formal report. United States. doi:10.2172/93743.
Stewart, H.B. Sat . "Assessment of the IVA3 code for multifield flow simulation. Formal report". United States. doi:10.2172/93743. https://www.osti.gov/servlets/purl/93743.
@article{osti_93743,
title = {Assessment of the IVA3 code for multifield flow simulation. Formal report},
author = {Stewart, H.B.},
abstractNote = {This report presents an assessment of the IVA3 computer code for multifield flow simulation, as applied to the premixing phase of a hypothetical steam explosion in a water-cooled power reactor. The first section of this report reviews the derivation of the basic partial differential equations of multifield modeling, with reference to standard practices in the multiphase flow literature. Basic underlying assumptions and approximations are highlighted, and comparison is made between IVA3 and other codes in current use. Although Kolev`s derivation of these equations is outside the mainstream of the multiphase literature, the basic partial differential equations are in fact nearly equivalent to those in other codes. In the second section, the assumptions and approximations required to pass from generic differential equations to a specific working form are detailed. Some modest improvements to the IVA3 model are suggested. In Section 3, the finite difference approximations to the differential equations are described. The discretization strategy is discussed with reference to numerical stability, accuracy, and the role of various physical phenomena - material convection, sonic propagation, viscous stress, and interfacial exchanges - in the choice of discrete approximations. There is also cause for concern about the approximations of time evolution in some heat transfer terms, which might be adversely affecting numerical accuracy. The fourth section documents the numerical solution method used in IVA3. An explanation for erratic behavior sometimes observed in the first outer iteration is suggested, along with possible remedies. Finally, six recommendations for future assessment and improvement of the IVA3 model and code are made.},
doi = {10.2172/93743},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Sat Jul 01 00:00:00 EDT 1995},
month = {Sat Jul 01 00:00:00 EDT 1995}
}

Technical Report:

Save / Share:
  • Physical models, numerical methods, and program description are presented for SIMMER-I, a computer program which predicts the neutronic and fluid dynamic behavior of an LMFBR during a hypothetical core disruptive accident.
  • During a Loss-of-Coolant Accident, fuel rod cladding may reach temperatures approaching 2200/sup 0/F. At these temperatures, swelling and rupture of the cladding may occur. The resulting flow blockage will affect steam flow and heat transfer in the bundle during the period of reflooding. The COBRA-IV-I subchannel computer code was used to simulate flow redistribution due to sleeve blockages in the FLECHT-SEASET 21-rod bundle and plate blockages in the JAERI Slab Core Test Facility. Sensitivity studies were conducted to determine the effects of spacer grid and blockage interaction, sleeve shape effects, sleeve length effects, blockage magnitude and distribution, thermally induced mixingmore » and bundle average velocity on flow redistribution. Pressure drop due to sleeve blockages was also calculated for several blockage configurations.« less
  • The need to study and assess life-cycle risks of Pu release by nuclear warheads during peace time lead to the development of a code which could model day to day operations involving nuclear weapons and calculate the associated risk involved in these proceedings. The code which accomplishes this risk assessment is called FlowSim. FlowSim is an event driven simulation implemented with the Mathematics computer language. FlowSim uses the paradigm of material flows and processes operating on the material in the flow. The Mathematica programming language, which is especially designed for dealing with symbolic entities, is used to enter the descriptionsmore » of both flows and processes at run time. Risk rate is calculated at the conclusion of each process and flow. Thus, both risk rate and integrated risk are available as outputs for the analyst.« less
  • This report describes a gap analysis performed in the process of developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repositorymore » designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with rigorous verification, validation, and software quality requirements. The gap analyses documented in this report were are performed during an initial gap analysis to identify candidate codes and tools to support the development and integration of the Waste IPSC, and during follow-on activities that delved into more detailed assessments of the various codes that were acquired, studied, and tested. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. The gap analysis indicates that significant capabilities may already exist in the existing THC codes although there is no single code able to fully account for all physical and chemical processes involved in a waste disposal system. Large gaps exist in modeling chemical processes and their couplings with other processes. The coupling of chemical processes with flow transport and mechanical deformation remains challenging. The data for extreme environments (e.g., for elevated temperature and high ionic strength media) that are needed for repository modeling are severely lacking. In addition, most of existing reactive transport codes were developed for non-radioactive contaminants, and they need to be adapted to account for radionuclide decay and in-growth. The accessibility to the source codes is generally limited. Because the problems of interest for the Waste IPSC are likely to result in relatively large computational models, a compact memory-usage footprint and a fast/robust solution procedure will be needed. A robust massively parallel processing (MPP) capability will also be required to provide reasonable turnaround times on the analyses that will be performed with the code. A performance assessment (PA) calculation for a waste disposal system generally requires a large number (hundreds to thousands) of model simulations to quantify the effect of model parameter uncertainties on the predicted repository performance. A set of codes for a PA calculation must be sufficiently robust and fast in terms of code execution. A PA system as a whole must be able to provide multiple alternative models for a specific set of physical/chemical processes, so that the users can choose various levels of modeling complexity based on their modeling needs. This requires PA codes, preferably, to be highly modularized. Most of the existing codes have difficulties meeting these requirements. Based on the gap analysis results, we have made the following recommendations for the code selection and code development for the NEAMS waste IPSC: (1) build fully coupled high-fidelity THCMBR codes using the existing SIERRA codes (e.g., ARIA and ADAGIO) and platform, (2) use DAKOTA to build an enhanced performance assessment system (EPAS), and build a modular code architecture and key code modules for performance assessments. The key chemical calculation modules will be built by expanding the existing CANTERA capabilities as well as by extracting useful components from other existing codes.« less
  • The Fuel Rod Analysis Program - Transient FRAP-T6 was independently assessed through comparisons with experimental data obtained from a Loss-of-Coolant Accident (LOCA) simulation test performed in the National Research Universal (NRU) reactor. A concise computer code description is given and computer code calculations are compared with experimental data from materials deformation test MT-1. Results of these comparison are discussed for different boundary conditions and different mathematical models which describe the physical processes in the fuel rod.