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 &#x201C;two-fluid model&#x201D; 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.

Alessandro Petruzzi

Francesco D&#x27;Auria

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

Since the 1988 amendment of the 10 CFR 50.46 rule in 1988, Westinghouse has been developing and applying realistic or best-estimate methods to perform LOCA safety analyses. A realistic analysis requires the execution of various realistic LOCA transient simulations where the effect of both model and input uncertainties are ranged and propagated throughout the transients. The outcome is typically a range of results with associated probabilities. The thermal/hydraulic code is the engine of the methodology but a procedure is developed to assess the code and determine its biases and uncertainties. In addition, inputs to the simulation are also affected by uncertainty and these uncertainties are incorporated into the process. Several approaches have been proposed and applied in the industry in the framework of best-estimate methods. Most of the implementations, including Westinghouse, follow the Code Scaling, Applicability and Uncertainty (CSAU) methodology. Westinghouse methodology is based on the use of the WCOBRA/TRAC thermal-hydraulic code. The paper starts with an overview of the regulations and its interpretation in the context of realistic analysis. The CSAU roadmap is reviewed in the context of its implementation in the Westinghouse evaluation model. An overview of the code (WCOBRA/TRAC) and methodology is provided. Finally, the recent evolution to nonparametric statistics in the current edition of the W methodology is discussed. Sample results of a typical large break LOCA analysis for a PWR are provided.

Cesare  Frepoli

An Overview of Westinghouse Realistic Large Break LOCA Evaluation Model

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.

Antonella Lombardi Costa

Francesco D'Auria

Walter Ambrosini

Analyses of Instability Events in the Peach Bottom-2 BWR Using Thermal-Hydraulic and 3D Neutron Kinetic Coupled Codes Technique

Thermal-hydraulic analysis tasks aimed at supporting plant operation and control of nuclear power plants are an important issue for the Asociaci&#xF3;n Nuclear Asc&#xF3;-Vandell&#xF2;s (ANAV). ANAV is the consortium that runs the Asc&#xF3; power plants (2 units) and the Vandell&#xF2;s-II power plant. The reactors are Westinghouse-design, 3-loop PWRs with an approximate electrical power of 1000&#x2009;MW. The Technical University of Catalonia (UPC) thermal-hydraulic analysis team has jointly worked together with ANAV engineers at different levels in the analysis and improvement of these reactors. This article is an illustration of the usefulness of computational analysis for operational support. The contents presented were operational between 1985 and 2001 and subsequently changed slightly following various organizational adjustments. The paper has two different parts. In the first part, it describes the specific aspects of thermal-hydraulic analysis tasks related to operation and control and, in the second part, it briefly presents the results of three examples of analyses that were performed. All the presented examples are related to actual situations in which the scenarios were studied by analysts using thermal-hydraulic codes and prepared nodalizations. The paper also includes a qualitative evaluation of the benefits obtained by ANAV through thermal-hydraulic analyses aimed at supporting operation and plant control.

F. Revent&#xF3;s

L. Batet

C. Llopis

C. Pretel

I. Sol

Thermal-Hydraulic Analysis Tasks for ANAV NPPs in Support of Plant Operation and Control

The present paper reviews activates relevant to the boiling water reactor (BWR) stability
                             phenomenon, which has a coupled neutronic and thermal-hydraulic nature, from the viewpoint
                              of model and code developments and their applications to the BWR stability solution
                               methodology in Japan.

Hideaki  Ikeda

Takanori  Fukahori

Yuichiro  Kubo

Hideo  Soneda

Shinya  Mizokami

BWR Stability Issues in Japan

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.

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

The task of regulatory body staff reviewing and assessing a realistic large break
     loss-of-coolant accident evaluation model is discussed, facing the actual regulatory licensing environment 
     related to the acceptance of the analysis of emergency core cooling system performance. Especially, focus is
      directed to the question of how to fulfill the requirement of quantifying the uncertainty in the calculated results
       when they are compared to the acceptance criteria for this system. As it is recognized that the regulation
        governing the loss-of-coolant accident analyses was originally developed by the United States Nuclear 
        Regulatory Commission, a description of its evolution is presented. When using a realistic evaluation model 
        to analyze the loss-of-coolant accident, different approaches have been used in the licensing arena. The 
        Brazilian regulatory body has concluded that, in the current environment, the independent regulatory 
        calculation is recognized as a relevant support for the staff decision within the licensing framework of 
        a realistic analysis.

Maria Regina Galetti

Regulatory Scenario for the Acceptance of Uncertainty Analysis
Methodologies for the LB-LOCA and the Brazilian Approach

About this journal

Science and Technology of Nuclear Installations publishes research on issues related to the nuclear industry, particularly the installations of nuclear technology, and aims to promote development in the area of nuclear sciences and technologies.

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Professor Michael I. Ojovan is the Chief Editor of the journal, and is currently based at the University of Sheffield, UK. He is known for many innovations in nuclear research, including metallic and glass-composite materials for nuclear waste immobilisation.

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Science and Technology of Nuclear Installations publishes research on issues related to the nuclear industry, particularly the installations of nuclear technology, and aims to promote development in the area of nuclear sciences and technologies.

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Science and Technology of Nuclear Installations

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Science and Technology of Nuclear Installations