F. D’Auria

University of Pisa, Italy M. Frogheri

University of Genoa, Italy

Abstract

The present paper deals with the Natural Circulation (NC) phenomenon in Pressurized Water nuclear Reactors (PWR). In the first part, data gathered from relevant experiments in PWR simulators are considered. These allowed the establishment of a flow map that has been used for evaluating the NC performance of various reactor concepts. In the second part, a theoretical study has been completed to assess the power removal capability by NC from the core of a PWR having the current geometric configuration. Taking as reference a PWR equipped with U-tubes steam generators, two-phase conditions occur in the core at power levels less than 20% nominal power. Therefore, for core power larger than this value the reactor cannot be classified any more as a PWR. The study shows that from a thermohydraulic point of view, the core can operate at power levels close to the current nominal value without experiencing thermal crisis. Limited consideration has been given to the neutronic design of the core.

1. INTRODUCTION

Natural Circulation (NC) is an important mechanism in several industrial systems and the knowledge of its behaviour is of interest to nuclear reactor design, operation and safety. In the nuclear technology, this is especially true for new reactor concepts that largely exploit the gravity forces for the heat removal capability. Natural circulation in a PWR occurs due to the presence of the heat source (core) and the heat sink constituted by the steam generators. In a gravity environment, with core located at a lower elevation than steam generators, those driving forces generate a flowrate suitable for removing nuclear fission decay power. At present, the NC core power removal capability is only exploited for accident situations, basically to demonstrate the inherent safety features of the plants.

The evaluation of the NC Performance (NCP) in experimental facilities simulating the integral system behaviour of a PWR has been the object of previous activities, e. g. Refs. [1], [2] and [3]. The NC scenarios occurring at different values of the primary system mass inventories were considered. Data have been gathered and analyzed coming from the PWR simulators Semiscale, Spes, Lobi, Bethsy, Pkl and Lstf. The thermohydraulic design of all the considered facilities has been achieved by adopting the same set of scaling laws that can be synthesized as follows: power-to-volume-scaling, full-height, time-preserving. Reference is made to both single phase and two-phase natural circulation. In order to evaluate the NCP of the mentioned facilities, significant information comes from the analysis of the trend of the core inlet mass flowrate and the primary loop mass inventory. The flowrate and the residual masses have been normalised taking into account of the volume of each facility and of the power level (typically n times 1% of the nominal core power, where n ranges between 1 and 5) utilized in the selected experiment. Four main flow patterns were characterized depending upon the value of the mass inventory of the primary loop (see also Ref. [4]):

a) single phase NC with no void in the primary system excluding the pressuriser and the upper head;

b) stable co-current two-phase NC with mass flowrate increasing when decreasing primary system fluid inventory;

c) unstable two-phase NC and occurrence of siphon condensation (Ref. [5]);

d) stable reflux condensation with liquid flowing countercurrent to steam in the hot legs: flowrate is sufficient to remove core power till loop mass inventory achieves values as low as 30-40% of the nominal values.

A Natural Circulation Flow Map (NCFM) has been obtained from the envelope of experimental data: this constituted a significant result of the research. The NCFM has been used for evaluating the NCP of the Pactel and the RD-14M facilities, simulators of the WWER-440 and of the CANDU reactors, respectively. The application of thermalhydraulic system codes allowed the extension of the use of the map to the study of NCP in operating reactors and in reactors under design. Westinghouse PWR (equipped with U-Tubes Steam Generators, UTSG), Babcock & Wilcox PWR (equipped with Once Through Steam Generators, OTSG), WWER-1000 (equipped with Horizontal Tubes Steam Generators, HTSG), EPR (designed by a Siemens-Framatome Consortium, equipped with UTSG), AP — 600 (passive type reactor designed by Westinghouse and equipped with UTSG) and EP-1000 (a passive type reactor under design by Westinghouse equipped with UTSG) have been considered. Significant results are discussed in the first part of the paper, also providing judgements in relation to the NCP of the considered systems.

The second part of the paper deals with the use of the NCFM for the analysis of the NCP of an existing UTSG PWR outside the design and the operating limits: in those conditions the system may not be classified any more as a PWR. In order to analyze the new system performance, a qualified thermalhydraulic system computer code and a qualified nodalisation (as far as possible) are adopted, e. g. Refs. [6] and [7]. The NCFM is used as a reference to confirm the qualification level of the numerical tools and to give an idea of the distance between the considered thermohydraulic configuration and the values of the relevant NC parameters that are consistent with the current design limits. Therefore, the main purposes of the paper can be synthesized as:

• to characterize the NC as a system phenomenon in PWR;

• to give an overview of the NCFM based upon experimental data and of its application;

• to show the possibility for PWR in the current geometrical layout to operate in NC at 100% core power.

The achieved results can be useful in the framework of the design of advanced Nuclear Power Plants.