Carlo Parisi

BEPU Approach in Licensing Framework, Including 3D NK Applications

During the recent years, a world-wide renewed interest in the exploitation of nuclear energy for electricity production is seen among both the Western and the new industrializing Countries (e.g., China and India). As a result, 61 reactors are now under construction and more than 100 units are planned for the incoming decade. Such impressive development is totally based on Light and Heavy Water Reactor (LWR & HWR) technologies [1], on designs that are an evolution of the robust and reliable Nuclear Power Plants (NPP) designed and built during the seventies-eighties of the last century. At that time, the need to guarantee an high safety level on one side and on the other the limited computational capabilities and the scarce knowledge of some phenomena, drove the main nuclear safety authorities to establish extremely conservative rules. Nowadays, after that tremendous progress has been made into the computational power availability, models accuracy and the knowledge of relevant phenomena, there is the need to go toward more realistic safety analyses and to relax some levels of conservativeness without compromising the always elevated safety level of the nuclear industry. The aim of this Chapter is to give an overview of the current trends in the licensing frameworks for a NPP. International best-practices are presented and discussed and sample applications derived from works of the San Piero a Grado Nuclear Research Group of the University of Pisa (GRNSPG/UNIPI) on existing industrial facilities are also reported.

RELAP5-3D Analysis of OECD-NEA/NRC BFBT Benchmark

OECD-NEA and the NRC organized and sponsored the BWR Fuel Bundle Test (BFBT) Benchmark with the main purpose of assessing sub-channel and Computational Fluid Dynamic codes capabilities in estimating relevant thermal-hydraulics parameters like void fraction and critical power for a BWR boiling channel. The assessment activity is performed comparing the code calculation results with the experimental data at steady state and transient conditions available through the Japanese Nuclear Power Engineering Corporation (NUPEC). In this framework, the San Piero a Grado Nuclear Research Group (GRNSPG) of the University of Pisa (UNIPI) developed a RELAP5-3D© thermal-hydraulic nodalization of the experimental bundle. The main purpose of this activity was the assessment of the capability of the well-known three-dimensional system thermal-hydraulic code RELAP5-3D© for the prediction of relevant parameters at the fuel assembly scale. In order to exploit the large amount of experimental data available, a three dimensional thermal-hydraulic nodalization was developed, simulating all sub-channels with MULTI-D component. The overall activities resulted in challenges for the code and the code users because of the necessary large number of nodes and heat structures used and because of the different solutions that had to be found for performing a typical sub-channel analysis with a system thermal-hydraulic code. Several calculations simulating steady state conditions were performed for different fuel assembly configurations. For each of them the void fraction distributions in all fuel assembly sub-channels and pressure drops along different part of an assembly were compared with the available experimental data. In the next exercises, BWR transients were executed (turbine trip and recirculation pump trip, respectively), calculating again the void distributions and critical power conditions. The results of the activity demonstrated the capability of the RELAP5-3D© code to perform calculations using a sub-channel approach. The code was able to calculate several thermal-hydraulics parameters with high accuracy at “fuel bundle” level of resolution; the results of “sub-channel” level are instead affected by a higher error (e.g. deviation of around 20% in the prediction of void distribution). Sub-channel results showed a better agreement when considering high quality tests compared to the lower quality ones.Copyright

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