Wolfgang Wulff

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 3: Assessment and ranging of parameters

Abstract Comparisons of results from TRAC-PFl/MOD1 code calculations with measurements from Separate Effects Tests, and published experimental data for modeling parameters have been used to determine the uncertainty ranges of code input and modeling parameters which dominate the uncertainty in predicting the Peak Clad Temperature for a postulated Large Break Loss of Coolant Accident (LBLOCA) in a four-loop Westinghouse Pressurized Water Reactor. The uncertainty ranges are used for a detailed statistical analysis to calculate the probability distribution function for the TRAC code-predicted Peak Clad Temperature, as is described in an attendant paper (Part 4). Measurements from Separate Effects Tests and Integral Effects Tests have been compared with results from corresponding TRAC-PFl/MOD1 code calculations to determine globally the total uncertainty in predicting the Peak Clad Temperature for LBLOCAs. This determination is in support of the detailed statistical analysis mentioned above. The analyses presented here account for uncertainties in input parameters, in modeling and scaling, in computing and in measurements. The analyses are an important part of the work needed to implement the Code Scalability, Applicability and Uncertainty (CSAU) methodology. CSAU is needed to determine the suitability of a computer code for reactor safety analyses and the uncertainty in computer predictions. The results presented here are used to estimate the safety margin of a particular nuclear reactor power plant for a postulated accident. Specifically, this paper describes, first in principle and then through their application, steps 7 through 10 of the fourteen-step CSAU methodology.

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.

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.

Scaling of thermohydraulic systems

Abstract The importance of scaling is demonstrated for experiments in small-size test facilities which are used to simulate the thermohydraulic system response of full-size industrial plants. It is shown how scaling facilitates (1) the quantitative ranking of thermohydraulic processes in their order of priority, (2) the rational selection of a test matrix, and therewith (3) the efficient allocation of resources for resolving technical issues of reactor safety. Thermohydraulic processes in components and the system of components are modeled. 29 Scaling groups are presented for systems with forced flow and natural circulation of single- and two-phase flows. System compliances and the matrices of flow inertia and impedance are introduced and scaled, to obtain the scaling criteria for the global interaction between components of integral systems.

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.

Assessment of RAMONA-3B methodology with oscillatory flow tests

Abstract The instability event at the LaSalle County Plant (GE BWR-5) imposed a new challenge on the computer codes available for reactor transient analysis. While the codes were originally designed to predict non-oscillatory transients, the new requirement on the code is to model limit cycle oscillations with large amplitudes, where feed-back effects from the core and the balance of plant, and the nonlinear effects are significant. Two of the United States Nuclear Regulatory Commissions (USNRC) computer codes, namely RAMONA-3B/MODO [1] and HIPA-BWR of Engineering Plant Analyzer [2] were expected, and are shown in part in this paper, to meet the above demands. The RAMONA-3B/MOD1 has now been upgraded from the RAMONA-3B/MODO. It has a three dimensional neutron kinetics model, coupled to multi-channel nonequilibrium drift-flux formulation, and an explicit integration scheme for the thermal hydraulics. The accuracy of the thermohydraulics in the RAMONA-3B code was assessed for the new application by modelling oscillatory transients in the FR1GG test facilty. Nodalization studies showed that twenty-four axial nodes are sufficient for a converged solution; calculations with twelve axial nodes produce, in comparison to the 24-node calculation, the deviation of 4.4% in the peak gain of the power to flow transfer function. The code predicted correctly the effects of power and inlet subcooling on the transfer function gain and the system resonance frequency. For the six available tests modeled, the code-predicted peak gain differs from the experimentally obtained gain on the average by +7%, with the standard deviation of ±30%. The uncertainty in the experimental data lies between −11% and +12%. The difference between predicted and measured frequency at the peak gain on the average is −6%, with the standard deviation of ±14%.

Computer simulation of two-phase flow in nuclear reactors

Abstract Two-phase flow models dominate the requirements of economic resources for the development and use of computer codes which serve to analyze thermohydraulic transients in nuclear power plants. An attempt is made to reduce the effort of analyzing reactor transients by combining purpose-oriented modeling with advanced computing techniques. Six principles are presented on mathematical modeling and the selection of numerical methods, along with suggestions on programming and machine selection, all aimed at reducing the cost of analysis. Computer simulation is contrasted with traditional computer calculation. The advantages of run-time interactive access operation in a simulation environment are demonstrated. It is explained that the drift-flux model is better suited than the two-fluid model for the analysis of two-phase flow in nuclear reactors, because of the latters closure problems. The advantage of analytical over numerical integration is demonstrated. Modeling and programming techniques are presented which minimize the number of needed arithmetical and logical operations and thereby increase the simulation speed, while decreasing the cost.