L. Batet

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

Thermal-hydraulic analysis tasks aimed at supporting plant operation and control of nuclear power plants are an important issue for the Asociacion Nuclear Asco-Vandellos (ANAV). ANAV is the consortium that runs the Asco power plants (2 units) and the Vandellos-II power plant. The reactors are Westinghouse-design, 3-loop PWRs with an approximate electrical power of 1000 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.

Boron transport model with physical diffusion for RELAP5

Rapid boron dilution transients have shown the need for accurate knowledge of the solute particle distribution in pressurized water reactors (PWRs). Small-break loss-of-coolant accidents (SBLOCAs) enable the formation of low-borated slugs in the loop seals. Low-borated water, if driven to the core, could cause a reactivity excursion. Since online boron concentration measurement is impractical in the primary system of PWR plants and quite difficult in test facilities, best-estimate codes should be seen as the most suitable tools. However, transport of a low-borated slug through the primary system requires accuracy of the methods. Several studies have shown high numerical diffusion introduced by the upwind difference schemes habitually used by system codes. Furthermore, most of the boron tracking models implemented in system codes at present do not consider physical diffusion. Nevertheless, to introduce physical diffusion, it is necessary to considerably reduce numerical diffusion. The implicit Godunov scheme, which is available in RELAP5, has proved its capability in almost eradicating numerical diffusion. However, the formulated equation in RELAP5 does not deal with physical diffusion. A new tracking model for the solute field is presented, along with results of its implementation in the RELAP5 code. To evaluate the new model, a numerical test has been performed that demonstrates both the reduction of numerical diffusion and the correct simulation of physical diffusion. Moreover, the UPC model has shown its consistency in experiment F1.1, which is an SBLOCA with boron dilution. The test was part of the OECD-PKL2 program directed by the Organisation for Economic Cooperation and Development (OECD).

Main Results of Phase IV BEMUSE Project: Simulation of LBLOCA in an NPP

Phase IV of BEMUSE Program is a necessary step for a subsequent uncertainty analysis. It includes the simulation of the reference scenario and a sensitivity study. The scenario is a LBLOCA and the reference plant is Zion 1 NPP, a 4 loop PWR unit. Thirteen participants coming from ten different countries have taken part in the exercise. The BEMUSE (Best Estimate Methods plus Uncertainty and Sensitivity Evaluation) Program has been promoted by the Working Group on Accident Management and Analysis (WGAMA) and endorsed by the Committee on the Safety of Nuclear Installations (CSNI). The paper presents the results of the calculations performed by participants and emphasizes its usefulness for future uncertainty evaluation, to be performed in next phase. The objectives of the activity are basically to simulate the LBLOCA reproducing the phenomena associated to the scenario and also to build a common, well-known, basis for the future comparison of uncertainty evaluation results among different methodologies and codes. The sensitivity calculations performed by participants are also presented. They allow studying the influence of different parameters such as material properties or initial and boundary conditions, upon the behaviour of the most relevant parameters related to the scenario.

Comparison of Best-Estimate Plus Uncertainty and Conservative Methodologies for a PWR MSLB Analysis Using a Coupled 3-D Neutron-Kinetics/Thermal-Hydraulic Code

Abstract This paper compares the Best-Estimate Plus Uncertainty (BEPU) methodology with the Conservative Bounding methodology for design-basis-accident analysis. Calculations have been performed with TRACE [for thermal-hydraulic (TH) system calculations] and PARCS [for neutron-kinetics (NK) modeling] under the SNAP platform. DAKOTA is used under the SNAP interface for uncertainty and sensitivity analysis. A simplified three-dimensional (3-D) neutronics model of the Ascó II nuclear power plant is used as the core kinetic model. The TH model is a one-dimensional representation of the primary and secondary systems except for the vessel, which is represented by a 3-D VESSEL component. The design-basis transient selected for the comparison is a main steam line break (MSLB) in a pressurized water reactor. This transient is characterized by space-time effects and requires coupled 3-D kinetics and TH modeling, especially for the recriticality scenario. The comparison methodology is as follows. Once the models are created, a best-estimate base case calculation is performed. The model is modified by selecting the most important parameters and assigning conservative values to them in order to obtain a conservative calculation. Several parameters are modified in this conservative way. These parameters are then perturbed in BEPU calculations. At the end, a comparison is made between results obtained in the conservative calculation and the BEPU methodology, respectively. As a general conclusion the BEPU method has been successfully illustrated in a coupled 3-D kinetics and TH system. Also, the study is an effective test for the adequacy of nodalizations for the neutronic and TH utilized codes. The BEPU methodology gives more margins, which allows for higher operational flexibility of the plant. The results of the BEPU methodology help improve the plant economics while meeting the safety standards. As a conclusion, the BEPU methodology has been successfully tested in NK-TH calculations. Narrow margins between the upper and lower BEPU cases are a consequence of the few perturbed parameters chosen and the transient boundary conditions.