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Title: Containment loads due to direct containment heating and associated hydrogen behavior: Analysis and calculations with the CONTAIN code

Abstract

One of the most important unresolved issues governing risk in many nuclear power plants involves the phenomenon called direct containment heating (DCH), in which it is postulated that molten corium ejected under high pressure from the reactor vessel is dispersed into the containment atmosphere, thereby causing sufficient heating and pressurization to threaten containment integrity. Models for the calculation of potential DCH loads have been developed and incorporated into the CONTAIN code for severe accident analysis. Using CONTAIN, DCH scenarios in PWR plants having three different representative containment types have been analyzed: Surry (subatmospheric large dry containment), Sequoyah (ice condenser containment), and Bellefonte (atmospheric large dry containment). A large number of parameter variation and phenomenological uncertainty studies were performed. Response of DCH loads to these variations was found to be quite complex; often the results differ substantially from what has been previously assumed concerning DCH. Containment compartmentalization offers the potential of greatly mitigating DCH loads relative to what might be calculated using single-cell representations of containments, but the actual degree of mitigation to be expected is sensitive to many uncertainties. Dominant uncertainties include hydrogen combustion phenomena in the extreme environments produced by DCH scenarios, and factors which affect the rate ofmore » transport of DCH energy to the upper containment. In addition, DCH loads can be aggravated by rapid blowdown of the primary system, co-dispersal of moderate quantities of water with the debris, and quenching of de-entrained debris in water; these factors act by increasing steam flows which, in turn, accelerates energy transport. It may be noted that containment-threatening loads were calculated for a substantial portion of the scenarios treated for some of the plants considered.« less

Authors:
; ; ; ; ;
Publication Date:
Research Org.:
Sandia National Labs., Albuquerque, NM (USA); Nuclear Regulatory Commission, Washington, DC (USA). Div. of Reactor Systems Safety
OSTI Identifier:
6438996
Report Number(s):
NUREG/CR-4896; SAND-87-0633
ON: TI87012896
DOE Contract Number:
AC04-76DP00789
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; 99 GENERAL AND MISCELLANEOUS//MATHEMATICS, COMPUTING, AND INFORMATION SCIENCE; CHEMICAL REACTION KINETICS; C CODES; CONTAINMENT; HEATING; HYDROGEN; PWR TYPE REACTORS; BELLEFONTE-1 REACTOR; BELLEFONTE-2 REACTOR; CORIUM; HEAT TRANSFER; NUCLEAR POWER PLANTS; REACTOR VESSELS; RISK ASSESSMENT; SEQUOYAH-1 REACTOR; SEQUOYAH-2 REACTOR; SURRY-1 REACTOR; SURRY-2 REACTOR; SURRY-3 REACTOR; SURRY-4 REACTOR; COMPUTER CODES; CONTAINERS; ELEMENTS; ENERGY TRANSFER; ENRICHED URANIUM REACTORS; KINETICS; NONMETALS; NUCLEAR FACILITIES; POWER PLANTS; POWER REACTORS; REACTION KINETICS; REACTORS; THERMAL POWER PLANTS; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS; 220900* - Nuclear Reactor Technology- Reactor Safety; 210200 - Power Reactors, Nonbreeding, Light-Water Moderated, Nonboiling Water Cooled; 990220 - Computers, Computerized Models, & Computer Programs- (1987-1989)

Citation Formats

Williams, D C, Bergeron, K D, Carroll, D E, Gasser, R D, Tills, J L, and Washington, K E. Containment loads due to direct containment heating and associated hydrogen behavior: Analysis and calculations with the CONTAIN code. United States: N. p., 1987. Web. doi:10.2172/6438996.
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Williams, D C, Bergeron, K D, Carroll, D E, Gasser, R D, Tills, J L, & Washington, K E. Containment loads due to direct containment heating and associated hydrogen behavior: Analysis and calculations with the CONTAIN code. United States. doi:10.2172/6438996.
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Williams, D C, Bergeron, K D, Carroll, D E, Gasser, R D, Tills, J L, and Washington, K E. Fri . "Containment loads due to direct containment heating and associated hydrogen behavior: Analysis and calculations with the CONTAIN code". United States. doi:10.2172/6438996. https://www.osti.gov/servlets/purl/6438996.
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@article{osti_6438996,
title = {Containment loads due to direct containment heating and associated hydrogen behavior: Analysis and calculations with the CONTAIN code},
author = {Williams, D C and Bergeron, K D and Carroll, D E and Gasser, R D and Tills, J L and Washington, K E},
abstractNote = {One of the most important unresolved issues governing risk in many nuclear power plants involves the phenomenon called direct containment heating (DCH), in which it is postulated that molten corium ejected under high pressure from the reactor vessel is dispersed into the containment atmosphere, thereby causing sufficient heating and pressurization to threaten containment integrity. Models for the calculation of potential DCH loads have been developed and incorporated into the CONTAIN code for severe accident analysis. Using CONTAIN, DCH scenarios in PWR plants having three different representative containment types have been analyzed: Surry (subatmospheric large dry containment), Sequoyah (ice condenser containment), and Bellefonte (atmospheric large dry containment). A large number of parameter variation and phenomenological uncertainty studies were performed. Response of DCH loads to these variations was found to be quite complex; often the results differ substantially from what has been previously assumed concerning DCH. Containment compartmentalization offers the potential of greatly mitigating DCH loads relative to what might be calculated using single-cell representations of containments, but the actual degree of mitigation to be expected is sensitive to many uncertainties. Dominant uncertainties include hydrogen combustion phenomena in the extreme environments produced by DCH scenarios, and factors which affect the rate of transport of DCH energy to the upper containment. In addition, DCH loads can be aggravated by rapid blowdown of the primary system, co-dispersal of moderate quantities of water with the debris, and quenching of de-entrained debris in water; these factors act by increasing steam flows which, in turn, accelerates energy transport. It may be noted that containment-threatening loads were calculated for a substantial portion of the scenarios treated for some of the plants considered.},
doi = {10.2172/6438996},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Fri May 01 00:00:00 EDT 1987},
month = {Fri May 01 00:00:00 EDT 1987}
}
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Technical Report:
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DOI:  10.2172/6438996
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  • ;
    The potential exists in a nuclear reactor core melt severe accident for molten core debris to be dispersed under high pressure into the containment building. If this occurs, the set of phenomena that result in the transfer of energy to the containment atmosphere and its surroundings is referred to as direct containment heating (DCH). Because of the potential for DCH to lead to early containment failure, the U.S. Nuclear Regulatory Commission (USNRC) has sponsored an extensive research program consisting of experimental, analytical, and risk integration components. An important element of the analytical research has been the development and assessment ofmore » direct containment heating models in the CONTAIN code. This report documents the DCH models in the CONTAIN code. DCH models in CONTAIN for representing debris transport, trapping, chemical reactions, and heat transfer from debris to the containment atmosphere and surroundings are described. The descriptions include the governing equations and input instructions in CONTAIN unique to performing DCH calculations. Modifications made to the combustion models in CONTAIN for representing the combustion of DCH-produced and pre-existing hydrogen under DCH conditions are also described. Input table options for representing the discharge of debris from the RPV and the entrainment phase of the DCH process are also described. A sample calculation is presented to demonstrate the functionality of the models. The results show that reasonable behavior is obtained when the models are used to predict the sixth Zion geometry integral effects test at 1/10th scale.« less

  • Preliminary experiments have been conducted at Sandia National Laboratories simulating degraded-core accidents. One notable result of these preliminary tests is the observation of the conversion of cesium iodide (CsI) to its elemental form (I/sub 2/) following a hydrogen burn. To evaluate some of the implications of the iodide conversion for the source term, computational simulations of the Surry TML and TMLB' accident sequences using experimental data on near-stoichiometric burns were conducted with the CONTAIN code. CONTAIN is the NRC's general-purpose computer code for modeling containment response to a severe accident. The results provide qualitative insights on a few of themore » more important sensitivities of the source term to the form of radioactive iodine in containment, and can be used to guide further experimental and theoretical developments in assessing the consequences of iodide conversion.« less

  • ; ; ; ...
    The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of themore » input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions.« less

  • ; ; ; ...
    The CONTAIN computer code is a best-estimate, integrated analysis tool for predicting the physical, chemical, and radiological conditions inside a nuclear reactor containment building following the release of core material from the primary system. CONTAIN is supported primarily by the U. S. Nuclear Regulatory Commission (USNRC), and the official code versions produced with this support are intended primarily for the analysis of light water reactors (LWR). The present manual describes CONTAIN LMR/1B-Mod. 1, a code version designed for the analysis of reactors with liquid metal coolant. It is a variant of the official CONTAIN 1.11 LWR code version. Some ofmore » the features of CONTAIN-LMR for treating the behavior of liquid metal coolant are in fact present in the LWR code versions but are discussed here rather than in the User's Manual for the LWR versions. These features include models for sodium pool and spray fires. In addition to these models, new or substantially improved models have been installed in CONTAIN-LMR. The latter include models for treating two condensables (sodium and water) simultaneously, sodium atmosphere and pool chemistry, sodium condensation on aerosols, heat transfer from core-debris beds and to sodium pools, and sodium-concrete interactions. A detailed description of each of the above models is given, along with the code input requirements.« less

  • ; ; ; ...
    The CONTAIN 1.1 computer code is an integrated analysis tool used for predicting the physical, chemical, and radiological conditions inside a containment building following the release of radioactive environment. CONTAIN 1.1 is the US Nuclear Regulatory Commission's principal best-estimate, mechanistic containment analysis code for severe accidents. CONTAIN 1.1 is intended to replace the earlier CONTAIN 1.0, which was released in 1984. The purpose of this User's Manual is to provide a basic understanding of the features and models in CONTAIN 1.1 so that users can prepare reasonable input and understand the output and its significance for particular applications. 148 refs.,more » 63 figs., 10 tabs.« less