Publications

A two-year effort focused on applying ASCI technology developed for the analysis of weapons systems to the state-of-the-art accident analysis of a nuclear reactor system was proposed. The Sandia SIERRA parallel computing platform for ASCI codes includes high-fidelity thermal, fluids, and structural codes whose coupling through SIERRA can be specifically tailored to the particular problem at hand to analyze complex multiphysics problems. Presently, however, the suite lacks several physics modules unique to the analysis of nuclear reactors. The NRC MELCOR code, not presently part of SIERRA, was developed to analyze severe accidents in present-technology reactor systems. We attempted to: (1) evaluate the SIERRA code suite for its current applicability to the analysis of next generation nuclear reactors, and the feasibility of implementing MELCOR models into the SIERRA suite, (2) examine the possibility of augmenting ASCI codes or alternatives by coupling to the MELCOR code, or portions thereof, to address physics particular to nuclear reactor issues, especially those facing next generation reactor designs, and (3) apply the coupled code set to a demonstration problem involving a nuclear reactor system. We were successful in completing the first two in sufficient detail to determine that an extensive demonstration problem was not feasible at this time. In the future, completion of this research would demonstrate the feasibility of performing high fidelity and rapid analyses of safety and design issues needed to support the development of next generation power reactor systems.

Experimental data are compiled and reviewed for aerosol particle releases due to combustion in air of Plutonium (Pu). The aerosol release fraction (ARF), which is the mass of Pu aerosolized, divided by the mass of Pu oxidized, is dependent on whether the oxidizing Pu sample is static (i.e. stationary) or dynamic (i.e. falling in air). ARF data are compiled for sample masses ranging from 30 mg to 1770 g, oxidizing temperatures varying from 113 C to {approx}1000 C, and air flow rates varying from 0.05 m/s to 5.25 m/s. The measured ARFs range over five orders of magnitude. The maximum observed static ARF is 2.4 x 10{sup -3}, and this is the recommended ARF for safety studies of static Pu combustion.

Abstract not provided.

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 the 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.

CONTAIN is a reactor accident simulation code developed by Sandia National Laboratories under US Nuclear Regulatory Commission (USNRC) sponsorship to provide integrated analysis of containment phenomena, including those related to nuclear reactor containment loads and radiological source terms. The recently released CONTAIN 2.0 code version represents a significant advance in CONTAIN modeling capabilities over the last major code release (CONTAIN 1.12V). The new modeling capabilities are discussed here. The principal motivation for many of the recent model improvements has been to allow CONTAIN to model the special features in advanced light water reactor (ALWR) designs. The work done in this area is also summarized. In addition to the ALWR work, the USNRC is currently engaged in an effort to qualify CONTAIN for more general use in licensing, with the intent of supplementing or possibly replacing traditional licensing codes. To qualify the CONTAIN code for licensing applications, studies utilizing CONTAIN 2.0 are in progress. A number of results from this effort are presented in this paper to illustrate the code capabilities. In particular, CONTAIN calculations of the NUPEC M-8-1 and ISP-23 experiments and CVTR test {number_sign}3 are presented to illustrate (1) the ability of CONTAIN to model non-uniform gas density and/or temperature distributions, and (2) the relationship between such gas distributions and containment loads. CONTAIN and CONTEMPT predictions for a large break loss of coolant accident scenario in the San Onofre plant are also compared.

A new gravitational head formulation for the treatment of stratified flows has been developed for CONTAIN, a lumped-parameter code used primarily for the analysis of postulated accidents in nuclear power plants. This new hybrid formulation is discussed and compared in this paper with the old, average-density CONTAIN formulation. In addition, these formulations are assessed against experimental data from three large-scale experiments in which stratified conditions were observed. These are the NUPEC M-8-1, Surtsey ST-3, and the HDR E11.2 experiments.

A new gravitational head formulation for the treatment of stratified conditions has been developed for CONTAIN 1.2, a control volume code used primarily for the analyses of postulated accidents in nuclear power plants. The new CONTAIN formulation of gravitational heads, termed the hybrid formulation, is described. This method of calculating stratified conditions is compared with the old, average-density formulation used in code versions prior to CONTAIN 1.2. Both formulations are assessed in this report with experimental data from three large-scale experiments in which stratified conditions formed by injection of a buoyant gas were observed. In general, the hybrid formulation gives a substantially higher degree of stratification than the old formulation. For stable, fully developed stratifications, the hybrid formulation also gives much better agreement with the measured degree of stratification than the old formulation. In addition, the predicted degree of stratification is robust and not sensitive to nodalization, provided a set of nodalization guidelines are followed. However, for stratification behavior controlled by special physics not modeled in CONTAIN, such as momentum convection, plume entrainment, or bulk molecular diffusion, one should not expect good agreement with experiment unless special measures to accommodate the missing physics are taken.

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 of 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.

The problem of radiative heat transfer through a gray, emitting, absorbing, and scattering medium with uniform optical properties is reduced to one without scattering through two techniques. One uses scaling laws, and the other uses a self-consistent effective gas temperature. The scaling laws are derived via the P1 approximation to the radiative transfer equation and can be applied to multidimensional problems with nonisothermal media. The effective temperature method is presently restricted to isotropic scattering and isothermal media. Both methods are evaluated in the current study as a function of scattering albedo, wall emissivity, and optical thickness for two different geometries, and two sets of wall and gas temperatures. The effects of scattering anisotropy are also assessed for the P1 method.