Novak Zuber

The interrelation between void fraction fluctuations and flow patterns in two-phase flow

A fast response, linearized X-ray void measurement system has been used to obtain statistical measurements in normally fluctuating air-water flow in a rectangular channel. It is demonstrated that the probability density function (PDF) of the fluctuations in void fraction may be used as an objective and quantitative flow pattern discriminator for the three dominant patterns of bubbly, slug, and annular flow. This concept is applied to data over the range of 0.0 to 37 m/sec mixture velocities to show that slug flow is simply a transitional, periodic time combination of bubbly and annular flows. Film thicknesses calculated from the PDF data are similar in magnitude in both slug and annular flows. Calculation of slug length and residence time ratios along with bubble lengths in slug flow are also readily obtainable from the statistical measurements. Spectral density measurements showed bubbly flow to be stochastic while slug and annular flows showed periodicities correlatable in terms of the liquid volume flux.

Quantifying reactor safety margins part 1: An overview of the code scaling, applicability, and uncertainty evaluation methodology☆

Abstract In August 1988, the Nuclear Regulatory Commission (NRC) approved the final version of a revised rule on the acceptance of ECCS entitled “Emergency Core Cooling System; Revision to Acceptance Criteria.” The revised rule states an alternate ECCS performance analysis, based on best-estimate methods, may be used to provide more realistic estimates of plant safety margins, provided the licensee quantifies the uncertainty of the estimates and includes that uncertainty when comparing the calculated results with prescribed acceptance limits. To support the revised ECCS rule, the NRC and its contractors and consultants have developed and demonstrated a method called the Code Scaling, Applicability, and Uncertainty (CSAU) evaluation methodology. The CSAU methodology and an example application, described in this set of six papers, demonstrate that uncertainties in complex phenomena can be quantified. The methodology is structured, traceable, and practical, as is needed in the regulatory arena. It addresses in a comprehensive and systematic manner questions concerned with: (1) code capability to scale-up processes from test facility to full-scale nuclear power plant (NPP). (2) code applicability to safety studies of postulated accident scenario in a specified NPP, and (3) quantifying uncertainties of calculated results. The methodology combines a “top-down” approach to define the dominant phenomena with a “bottom-up” approach to quantify uncertainty. The methedology is able to address both: (1) uncertainties for which bias and distribution are quantifiable, and (2) uncertainties for which only a bounding value is quantifiable. The methodology is general, and therefore applicable to a variety of scenarios, plants, and codes. This paper provides an overview of the CSAU evaluation methodology and its application to a postulated cold-leg, large-break loss-of-coolant accident in a Westinghouse four-loop pressurized water reactor with 17 x 17 fuel. The code selected for this demonstration of the CSAU methodology was TRAC-PF1/MOD1, Version 14.3.

An integrated structure and scaling methodology for severe accident technical issue resolution : Development of methodology

Scaling has been identified as a particularly important element of the Severe Accident Research Program because of its relevance not only to experimentation, but also to analyses based on code calculations or special models. Recognizing the central importance of severe accident scaling issues, the United States Regulatory Commission implemented a Severe Accident Scaling Methodology (SASM) development program involving a lead laboratory contractor and a Technical Program Group to guide the development and to demonstrate its practicality via a challenging application. The Technical Program Group recognized that the Severe Accident Scaling Methodology was an integral part of a larger structure for technical issue resolution and, therefore, found the need to define and document this larger structure, the Integrated Structure for Technical Issue Resolution (ISTIR). The larger part of the efforts have been devoted to the development and demonstration of the Severe Accident Scaling Methodology, which is Component II of the ISTIR. The ISTIR and the SASM have been tested and demonstrated, by their application to a postulated direct containment heating scenario. The ISTIR objectives and process are summarized in this paper, as is its demonstration associated directly with the SASM. The objectives, processes and demonstration for the SASM are also summarized in the paper. The full body of work is referenced.

Quantifying reactor safety margins: Part 2, Characterization of important contributors to uncertainty

Abstract The NRC has issued a revised Emergency Core Coolant System (ECCS) rule which allows the use of best estimate computer codes for safety analysis, providing the uncertainty of the calculations are quantified and compared with acceptance limits contained in 10 CFR Part 50. To support the revised rule, the NRC and its contractors and consultants have developed and demonstrated a methodology to quantify uncertainty called Code Scaling, Applicability and Uncertainty (CSAU). The methodology consists of three primary elements containing 14 steps. The first element, “Requirements and Capabilities”, which contains the first six steps, is described and demonstrated in this paper. The objective of this element is to characterize the important contributors to uncertainty. The objective is accomplished by determining the applicability of a code to analysis of a transient in a Nuclear Power Plant (NPP) through comparison of the scenario- and plant-dictated requirements with the simulation capabilities of the code.

The effects of complexity, of simplicity and of scaling in thermal-hydraulics

This lecture has a twofold purpose. First, we will assess the state of the art and the trends in thermal-hydraulics (T-H) technology, within the context of replicating and non-replicating information systems. Four T-H examples are used to illustrate that an ever-increasing complexity in formulating and analyzing problems leads to inefficiency, obsolescence and evolutionary failure. By contrast, simplicity, which allows for parsimony, synthesis and clarity of information, ensures efficiency, survival and replication. This comparison (complexity versus simplicity) also provides the requirements and guidance for a success path in T-H development. The second objective of this paper is to demonstrate that scaling provides the means to process information in an efficient manner, as required by competitive (and, thereby, replicating) systems. To this end, the lecture summarizes the essential features of the Fractional Change, Scaling and Analysis approach, which offers a general paradigm for quantifying the effects that an agent of change has on a given information system. The paper will further demonstrate that a single concept and a single method may be used to scale and analyze all transport processes in a given field of interest (fluid mechanics, heat transfer, etc.) and/or across fields and disciplines (mechanics, biology, etc.) Therefore, the paradigm: (1) ensures economy and efficiency in addressing and resolving technical or scientific problems; and (2) enables a ‘cultural cross-pollination’ between different information systems (disciplines). By means of a simple example in the Appendix, we shall: (1) demonstrate the efficiency to be gained through scaling; and (2) illustrate the inefficiency and wastefulness of computer-based safety studies as presently conducted.

Quantifying reactor safety margins Part 5: Evaluation of scale-up capabilities of best estimate codes☆

Abstract This paper presents the CSAU procedure and rational for: 1. (1) Evaluating the capability of a best estimate code to scale-up processes from reduced scale test facilities to full scale nuclear power plants, and 2. (2) Quantifying the effects of scale distortions and/or a limited data base, on code uncertainty to calculate a safety parameter of interest (for example peak clad temperature). To this end, the procedure uses and integrates information from test facility design and operation, from scenario processes and phenomena and from code documentation. A flow diagram of the procedure is presented together with the prescribed steps. To present the rationale and need for the procedure, the paper also summarizes the scaling techniques developed and used to design and operate loss of coolant accident related test facilities. The procedure is illustrated by applying it to TRAC-PF1/MOD1 calculations of a large break loss of coolant accident in a four loop Westinghouse pressurized water reactor. The application demonstrates that the procedure is sytematic, traceable and practical.

Quantifying reactor safety margins part 4: Uncertainty evaluation of lbloca analysis based on trac-pf1/mod 1

Abstract The Nuclear Regulatory Commission (NRC) has issued a revised Emergency Core Cooling System (ECCS) rule allowing the use of best estimate computer codes for safety analysis. The rule also requires an estimation of the uncertainty in the calculated system response and a comparison of the resulting bound with the acceptance limits of 10CFR Part 50. To support this revised rule the NRC and its consultants and contractors have developed and demonstrated the Code Scaling, Applicability and Uncertainty Methodology (CSAU). The last of the three elements of the methodology - Uncertainty Evaluation - is described in this paper.

Quantifying reactor safety margins Part 6: A physically based method of estimating pwr large break loss of coolant accident pct

Abstract The Nuclear Regulatory Commission (NRC) has issued a revised Emergency Core Cooling System (ECCS) rule allowing the use of best estimate computer codes for safety analysis. To support this revised rule the NRC and its consultants and contractors have developed and demonstrated the Code Scaling. Applicability and Uncertainty (CSAU) Evaluation Methodology. This effort lead to increased understanding of the phenomena, and their relative dominance, during Large Break Loss of Coolant Accidents (LBLOCAs) in Pressurized Water Reactors (PWRs). Consequently, it became possible, as is done in this paper, to develop a method for establishing clad temperature history by using physically based arguments and engineering correlations. The results from this method are compared with similar uncertainty estimates based on large computer codes. These comparisons provide a rationale, based on physical arguments, for evaluating the large computer code based estimates of uncertainty.