W. Giannotti

Development of a code with the capability of internal assessment of uncertainty

The internal assessment of uncertainty is a desirable capability for thermal-hydraulic system codes. This consists of the possibility of obtaining proper uncertainty bands each time a nuclear plant transient scenario is calculated. A methodology suitable for introducing such a capability into a system code is discussed. At the basis of the derivation of the code with (the capability of) internal assessment of uncertainty (CIAU), there is the uncertainty methodology based on the accuracy extrapolation (UMAE), previously proposed by the University of Pisa, although other uncertainty methodologies can be used for the same purpose. The idea of the CIAU is the identification and the characterization of standard plant statuses and the association of uncertainty to each status. One hypercube and one time interval identify the plant status. Quantity and time uncertainties are combined for each plant status. The recently released U.S. Nuclear Regulatory Commission RELAP5/MOD3.2 system code constitutes the CIAU. This is used for showing the applicability of the proposed method. The derivation of the methodology is discussed, and reference results of pressurized water reactor plant transients are shown bounded by the CIAU-calculated uncertainty bands.

Use of the Natural Circulation Flow Map for Natural Circulation Systems Evaluation

The aim of this paper is to collect and resume the work done to build and develop, at the University of Pisa, an engineering tool related to the natural circulation. After a brief description of the different loop flow regimes in single phase and two phase, the derivation of a suitable tool to judge the NC performance in a generic system is presented. Finally, an extensive comparison among the NC performance of various nuclear power plants having different design is done to show a practical application of the NC flow map.

Integrated Approach for Actual Safety Analysis

Actual trend in reactor safety deterministic analysis are evolving toward best estimate approach. Best estimate analyses imply use of best estimate codes and input data. The best estimate concept is not limited to thermal-hydraulics rather in general terms it covers many fields, likewise three dimensional neutron kinetics, structural analysis and containment performance evaluation. The general frame is to put efforts in avoiding conservative assumptions performing analysis adopting the best tool available for each specific topic, all contributing to give an integrated evaluation of the plant response. The needs to adopt an integrated approach in performing safety analysis come from the inherent complexity of a Nuclear Power Plant and from the tight interactions among the subsystems constituting the plant itself. These interactions directly involve the necessity to consider a broad spectrum of disciplines typically coming into play in different not interacting analyses. An example of the integral approach is given in the present document. The integral approach has been pursued for the safety analyses of the ‘post-Chernobyl modernized’ Reactor Bolshoy Moshchnosty Kipyashiy (RBMK) specifically for Smolensk 3. These analyses were performed at the University of Pisa within the framework of a European Commission sponsored activity. The mentioned analyses deal with events occurring in the primary circuit, as well as excluding those events originated from plant status different from the nominal operating conditions. Following the evaluation of the current state of the art in the safety analysis area, targets for the analysis were established together with suitable chains of computational tools. The availability of computational tools, including codes, nodalisations and boundary and initial conditions for the Smolensk 3 Nuclear Power Plant, brought to their application to the prediction of the selected transient evolutions that, however, are not classified as licensing studies. The integrated approach for safety analysis yields to the evaluation of complex scenarios not predictable adopting just a single computational tool. Example is given considering the Multiple Pressure Tube Rupture (MPTR) event which constitute one of the main concern of this kind of plant. The content of this document includes an introduction to the critical issues to be accounted for in the frame of an integral safety analysis approach; the selection of suitable computational tools to proper deal with the scenario subject of the investigation; an

INTERNATIONAL STANDARD PROBLEM 50: THE UNIVERSITY OF PISA CONTRIBUTION

The present paper deals with the participation of the University of Pisa in the last International Standard Problem (ISP) focused on system thermal hydraulic, which was led by the Korean Atomic Energy Research Institution (KAERI). The selected test was a Direct Vessel Injection (DVI) line break carried out at the ATLAS facility. University of Pisa participated, together with other eighteen institutions, in both blind and open phase of the analytical exercise pursuing its methodology for developing and qualifying a nodalization. Qualitative and quantitative analysis of the code results have been performed for both ISP-50 phases, the latter adopting the Fast Fourier Transfer Based Method (FFTBM). The experiment has been characterized by threedimensional behavior in downcomer and core region. Even though an attempt to reproduce these phenomena, by developing a fictitious three-dimensional nodalization has been realized, the obtained results were generally acceptable but not fully satisfactory in replicating 3D behavior.

Addressing Boron Dilution Scenario Through RELAP5/3.3 Analysis of PWR SB LOCA

Reactivity accident scenarios can occur originated by internal boron dilution in the primary system of a nuclear pressurized water reactor type (PWR or VVER). In essence the problem is caused by boron dilution following vaporization and condensation of the primary system coolant in case of decrease of primary system mass inventory, for example during a small-break loss of coolant accident (SB-LOCA) that may include boiling in the core with condensation of steam in the steam generators. When the liquid level in the reactor vessel decreases below the hot leg elevation, steam begins to flow to the steam generators and condenses there. This steam carries no boron and thus boron concentration in the cold leg loop seals begins to decrease. If for some reason this water plug with low boron concentration begins to flow towards the core and enters it without any major mixing with the borated coolant, the result is a positive reactivity insertion. The paper presents an analysis by RELAP5 Mod 3.3 code [1] of a small break LOCA of 20 cm2 area in the lower plenum of a four-loop PWR nuclear reactor. The boundary conditions of the calculations consider the eight accumulator tanks available, two/four low pressure injection systems (LPIS) available, and two of the four high pressure injection systems (HPIS) available. Sensitivity calculations were performed, regarding among other things, the boron concentration in the Emergency Core Cooling Systems (ECCS) and reactor cooling system (RCS) from Design Basis Accident (DBA) to beyond DBA conditions. From the results obtained, in some calculations boron dilution is observed in more than one loop seal. The situation in which the plugs in the loop seals are transported to the core without mixing with other borated water led to a potentially hazardous situation for four calculations in which initial conditions were far from DBA. It is important to emphasize that the present study has not the objective of a safety analysis of the NPP involved, but it should be considered inside research activities regarding the boron dilution issue.Copyright

Investigation of a Possible Emergency Procedure for the VVER 1000 NPP in Case of a Total Loss of Feedwater and a Main Steam Line Break

Accident management procedures associated with nuclear power plant beyond design basis accidents should be developed with the aid of the more recent versions of advanced computational tools (best estimate codes) in order to verify the effectiveness of normal operation systems of the plant to avoid or minimize core damage. A. Madeira showed in her paper “A PWR Recovery Option for a Total Loss of Feedwater Beyond Design Basis Scenario” that it is possible to safely control a total loss of feedwater scenario in the Angra2 nuclear power plant, using two emergency procedures, namely the opening of the steam generator (SG) relief valves, and on the primary side the complete manual opening of all pressurizer relief and safety valves. This paper investigates the effectiveness of the procedure opening of the SG relief valves, followed by primary side feed and bleed for a generic VVER-1000 NPP in case of a total loss of feed water. The results indicate that the procedure is successful in reducing the primary side pressure and temperature to safe conditions, i.e. long term core cooling is achievable.© 2004 ASME