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

Abstract

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.

Thermal hydraulic simulations of the Angra 2 PWR

“…The steady state parameters calculated are presented in Table 3 and are compared with the nominal values provided in FSAR. The results show good agreement with the reference data and the calculated errors are in correspondence with the usual criteria for quantification of the steady state prediction quality that have been adopted [8,9]. This means that the model reproduces with good approximation the steady state thermal hydraulic behavior of the reactor.…”

Safety Studies and General Simulations of Research Reactors Using Nuclear Codes

“…The model can reproduce a specific part or the whole system to be simulated. However as there is not a fixed rule to perform the nodalization, a large responsibility is passed to the user of the code in order to develop an adequate model scheme which makes best use of the various modules and the prediction capabilities of the specific code (Petruzzi and D'Auria, 2008;D'Auria and Galassi, 1998). Subchannel codes are used to analyse specific processes within the core of the reactor, such as localized flow and heat transfer variables in representative fuel assemblies.…”

BEPU Approach in Licensing Framework, Including 3D NK Applications

“…2. Flow sheet of nodalization qualification procedure [17] steady-state and transients analyses. A chain of Neutronic codes has to be employed, [starting from the Evaluated Nuclear Data Files processing, passing through the neutron transport simulations (both in deterministic and stochastic ways), the few group homogenized cross section libraries calculations and ending in the 3D core-wide NK simulations (eventually coupled with a SYS TH code).…”