A. Petruzzi

Thermal-Hydraulic System Codes in Nulcear Reactor Safety and Qualification Procedures

In the last four decades, large efforts have been undertaken to provide reliable thermal-hydraulic system codes for the analyses of transients and accidents in nuclear power plants. Whereas the first system codes, developed at the beginning of the 1970s, utilized the homogenous equilibrium model with three balance equations to describe the two-phase flow, nowadays the more advanced system codes are based on the so-called “two-fluid model” with separation of the water and vapor phases, resulting in systems with at least six balance equations. The wide experimental campaign, constituted by the integral and separate effect tests, conducted under the umbrella of the OECD/CSNI was at the basis of the development and validation of the thermal-hydraulic system codes by which they have reached the present high degree of maturity. However, notwithstanding the huge amounts of financial and human resources invested, the results predicted by the code are still affected by errors whose origins can be attributed to several reasons as model deficiencies, approximations in the numerical solution, nodalization effects, and imperfect knowledge of boundary and initial conditions. In this context, the existence of qualified procedures for a consistent application of qualified thermal-hydraulic system code is necessary and implies the drawing up of specific criteria through which the code-user, the nodalization, and finally the transient results are qualified.

Best-Estimate Model Calibration and Prediction through Experimental Data Assimilation—II: Application to a Blowdown Benchmark Experiment

Abstract This work presents a paradigm application of a new methodology for simultaneously calibrating (adjusting) model parameters and responses, through assimilation of experimental data, to the benchmark transient thermal-hydraulic experiment IC1, performed at London’s Imperial College. Following the description of the experimental setup, the corresponding mathematical model is developed and solved numerically. The sensitivities of typically important responses (e.g., temperatures, pressures) to model parameters are computed by applying both the forward and the adjoint sensitivity analysis procedures. These sensitivities not only identify the most important model parameters but also propagate, within the data assimilation procedure, parameter uncertainties for obtaining predictive best-estimate quantities, with reduced best-estimate uncertainties (i.e., “smaller” values for the variance-covariance matrices). This assimilation procedure also provides a quantitative indication of the degree of agreement between computations and experiments. In particular, the paradigm application presented in this work indicates the path for validating and calibrating thermal-hydraulic computational models used for reactor safety analyses. The concluding remarks highlight several important open issues, the resolution of which would significantly advance the area of predictive best-estimate modeling, while opening new avenues for applications in nuclear reactor engineering and safety.

Methodology of internal assessment of uncertainty and extension to neutron kinetics/thermal-hydraulics coupled codes

Abstract The best-estimate calculation results from complex system codes are affected by approximations that are unpredictable without the use of computational tools that account for the various sources of uncertainty. The code with (the capability of) internal assessment of uncertainty (CIAU) has been previously proposed by the University of Pisa to realize the integration between a qualified system code and an uncertainty methodology and to supply proper uncertainty bands each time a nuclear power plant (NPP) transient scenario is calculated. The derivation of the methodology and the results achieved by the use of CIAU are discussed to demonstrate the main features and capabilities of the method. In a joint effort between the University of Pisa and The Pennsylvania State University, the CIAU method has been recently extended to evaluate the uncertainty of coupled three-dimensional neutronics/thermal-hydraulics calculations. The result is CIAU-TN. The feasibility of the approach has been demonstrated, and sample results related to the turbine trip transient in the Peach Bottom NPP are shown. Notwithstanding that the full implementation and use of the procedure requires a database of errors not available at the moment, the results give an idea of the errors expected from the present computational tools.

Approaches, Relevant Topics, and Internal Method for Uncertainty Evaluation in Predictions of Thermal-Hydraulic System Codes

The evaluation of uncertainty constitutes the necessary supplement of best-estimate calculations performed to understand accident scenarios in water-cooled nuclear reactors. The needs come from the imperfection of computational tools, on the one side, and the interest in using such a tool to get more precise evaluation of safety margins. The paper reviews the salient features of three independent approaches for estimating uncertainties associated with predictions of complex system codes. Namely, the propagations of code input error and calculation output error constitute the keywords for identifying the methods of current interest for industrial applications, while the adjoint sensitivity-analysis procedure and the global adjoint sensitivity-analysis procedure, extended to performing uncertainty evaluation in conjunction with concepts from data adjustment and assimilation, constitute the innovative approach. Throughout the developed methods, uncertainty bands can be derived (both upper and lower) for any desired quantity of the transient of interest. For one case, the uncertainty method is coupled with the thermal-hydraulic code to get the code with capability of internal assessment of uncertainty, whose features are discussed in more detail.

International Course to Support Nuclear Licensing by User Training in the Areas of Scaling, Uncertainty, and 3D Thermal-Hydraulics/Neutron-Kinetics Coupled Codes: 3D S.UN.COP Seminars

Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers, vendors, and research organizations. The computer code user represents a source of uncertainty that can influence the results of system code calculations. This influence is commonly known as the “user effect” and stems from the limitations embedded in the codes as well as from the limited capability of the analysts to use the codes. Code user training and qualification represent an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In other words, the program aims at contributing towards solving the problem of user effect. In addition, this paper presents the organization and the main features of the 3D S.UN.COP (scaling, uncertainty, and 3D coupled code calculations) seminars during which particular emphasis is given to the areas of the scaling, uncertainty, and 3D coupled code analysis.

The BEMUSE Programme: Results of the first part concerning the LOFT L2-5 test

This paper is aimed at describing results of the first part of the BEMUSE (B est E stimate M ethods – U ncertainty and S ensitivity E valuation) programme. The purpose of BEMUSE is the comparison of best-estimate calculations, followed by the comparison of uncertainty and sensitivity analyses for a Large Break Loss of Coolant Accident (LB-LOCA). The first part of the programme is devoted to the study of the LOFT L2-5 experiment. After a general presentation of the programme, which implies more than ten participants, this paper describes the qualification process and the results of the best-estimate calculations. The results are significantly less dispersed than those of the ISP-13, concerning already LOFT L2-5 more than 20 years ago. Then, it presents extensively the methods and the results of uncertainty and sensitivity analyses. All the participants, apart from the University of Pisa with the CIAU method, use a fully probabilistic approach, based on Wilks’ formula. However, differences appear for the choice of the uncertain input parameters to be considered and for their associated range of variation. Sensitivity analysis is performed with regression techniques, and the results are also compared. As a conclusion, main lessons learnt from BEMUSE and recommendations are presented.Copyright

COMPARISON BETWEEN BEST ESTIMATE-PLUS-UNCERTAINTY METHODS AND CONSERVATIVE TOOLS FOR NUCLEAR POWER PLANT LICENSING

Abstract Computational reactor safety analysis is trending to replace conservative evaluation model calculations with best-estimate analysis complemented by uncertainty evaluation of the code results. In such cases, the evaluation of the margin to acceptance criteria (e.g., the maximum fuel rod clad temperature) is based on the upper limit of the calculated uncertainty range. Uncertainty analysis is compulsory if relevant conclusions are to be obtained from best-estimate thermal-hydraulic code calculations in order to avoid presenting single values of unknown accuracy for comparison with regulatory acceptance limits. This paper, after a thorough introduction of conservative and best-estimate methods and characterization of the main sources of uncertainties affecting best-estimate system codes, applies a best-estimate-plus-uncertainty (BEPU) method to three cases having as reference different nuclear power plants and different types of transients. Finally, the results from the BEPU approach is compared with a conservative approach and a combined approach.

Analyses of instability events in the Peach Bottom-2 BWR using thermal-hydraulic and 3D neutron kinetic coupled codes technique

Boiling water reactor (BWR) instabilities may occur when, starting from a stable operating condition, changes in system parameters bring the reactor towards an unstable region. In order to design more stable and safer core configurations, experimental and theoretical studies about BWR stability have been performed to characterise the phenomenon and to predict the conditions for its occurrence. In this work, contributions to the study of BWR instability phenomena are presented. The RELAP5/MOD3.3 thermal-hydraulic (TH) system code and the PARCS-2.4 3D neutron kinetic (NK) code were coupled to simulate BWR transients. Different algorithms were used to calculate the decay ratio (DR) and the natural frequency (NF) from the power oscillation predicted by the transient calculations as two typical parameters used to provide a quantitative description of instabilities. The validation of the code model set up for the Peach Bottom Unit 2 BWR plant is performed against low-flow stability tests (LFSTs). The four series of LFST have been performed during the first quarter of 1977 at the end of cycle 2 in Pennsylvania. The tests were intended to measure the reactor core stability margins at the limiting conditions used in design and safety analyses.

The BEPU Evaluation Model with RELAP5-3D for the Licensing of the Atucha-II NPP

Abstract Within the licensing process of the Atucha-II pressurized heavy water reactor, the best-estimate plus uncertainty (BEPU) approach has been selected for issuing Chapter 15 of the Final Safety Analysis Report. The RELAP5-3D code developed by Idaho National Laboratory has been adopted as the best-estimate system thermal-hydraulic code to perform the accident analyses. The complexity of a nuclear power plant (NPP) and of the accident scenarios may be a challenge for a conservative analysis and may justify the choice of a BEPU approach in the licensing process. This implies two main needs: (1) the need to adopt and to prove (to the regulatory authority) an adequate quality for the computational tools and (2) the need to account for the uncertainty. The purpose of the present paper is to outline key aspects of the BEPU process aimed at the licensing of the Atucha-II (CNA-II) NPP in Argentina operated by Nucleoeléctrica Argentina (NA-SA). Among the general attributes of a methodology to perform accident analysis of a NPP for licensing purposes, the very first one should be compliance with the established regulatory requirements. A second attribute deals with the adequacy and the completeness of the selected spectrum of events that should consider the combined contributions of deterministic and probabilistic methods. The third attribute is connected to the availability of qualified tools and analytical procedures suitable for the analysis of accident conditions envisaged for the NPP of concern. The execution of the overall analysis and the evaluation of results in relation to slightly fewer than 100 scenarios revealed the wide safety margins available for the NPP of concern, which was licensed on May 29, 2014.

RELAP5-3D Analysis of Pressure Perturbation at the Peach Bottom BWR During Low-Flow Stability Tests

Experimental and theoretical studies about the BWR (Boiling Water Reactor) stability have been performed to design a stable core configuration. BWR instabilities can be caused by inter-dependencies between thermal-hydraulic and reactivity feedback parameters such as the void-coefficient, for example, during a pressure perturbation event. In the present work, the pressure perturbation is considered in order to study in detail this type of transient. To simulate this event, including the strong feedback effects between core neutronic and reactor thermal-hydraulics, and to verify core behavior and evaluate parameters related to safety, RELAP5-3D code has been used in the analyses. The simulation was performed making use of Peach Bottom-2 BWR data to predict the dynamics of a real reactor during this type of event. Stability tests were conducted in the Peach Bottom 2 BWR, in 1977, and were done along the low-flow end of the rated power-flow line, and along the power-flow line corresponding to minimum recirculation pump speed. The calculated results are herein compared against the available experimental data. (authors)