M. Cherubini

Consistent Posttest Calculations for LOCA Scenarios in LOBI Integral Facility

Integral test facilities (ITFs) are one of the main tools for the validation of best estimate thermalhydraulic system codes. The experimental data are also of great value when compared to the experiment-scaled conditions in a full NPP. The LOBI was a single plus a triple-loop (simulated by one loop) test facility electrically heated to simulate a 1300 MWe PWR. The scaling factor was 712 for the core power, volume, and mass flow. Primary and secondary sides contained all main active elements. Tests were performed for the characterization of phenomenologies relevant to large and small break LOCAs and special transients in PWRs. The paper presents the results of three posttest calculations of LOBI experiments. The selected experiments are BL-30, BL-44, and A1-84. They are LOCA scenarios of different break sizes and with different availability of safety injection components. The goal of the analysis is to improve the knowledge of the phenomena occurred in the facility in order to use it in further studies related to qualifying nodalizations of actual plants or to establish accuracy data bases for uncertainty methodologies. An example of procedure of implementing changes in a common nodalization valid for simulating tests occurred in a specific ITF is presented along with its confirmation based on posttests results.

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

Optimizing the Initial Pressure of Accumulators for the Atucha2 NPP Using an Optimization Method

Passive safety systems like hydro accumulators offer high reliability and are therefore, when a choice is possible, often preferred over active safety systems. However, their effectiveness in case of an incident or accident depends on a large number of parameters (like break size in case of a loss of coolant accident, availability of other safety systems, initial pressure in the accumulators) and is in general difficult to predict. This paper presents a study to optimize the initial pressure and the pressure drops in the accumulator line for a intermediate break loss of coolant accident for Atucha 2, a Siemens-KWU, heavy water moderated, channel type pressurized water reactor under construction. An optimization method was applied. The thermal hydraulic system code RELAP5 mod 3.3 was used for the analysis. Three cases have been analyzed. First, the initial pressure and pressure drop in the accumulator line was optimized in case of an intermediate break in cold leg two, assuming safety injection of two of the four trains of safety systems into hot and cold leg. Second, like before, but assuming safety injection into cold leg only. Third, like case two, but grouping the four accumulators in two groups, with different initial pressure and pressure drops in the accumulator lines. The results show that a slight increase of the initial accumulator pressure compared to the design value could be beneficial for the investigated initial event. Further, case three shows that different initial pressure in the accumulators could increase the effectiveness of the intervention for the investigated accident.Copyright

Accident Management in VVER-1000

The present paper deals with the investigation study on accident management in VVER-1000 reactor type conducted in the framework of a European Commission funded project. The mentioned study involved both experimental and computational fields. The purpose of this paper is to summarize the main findings from the execution of a wide-range analysis focused on AM in VVER-1000 with main regard to the qualification of computational tools and the proposal for an optimal AM strategy for this kind of NPP.