Alessandro Del Nevo

Assessment of Transuranus fuel performance code against Studsvik Inter-Ramp BWR database

The fuel matrix and the cladding constitute the first barrier against radioactive fission product release. Therefore a defense in depth concept requires also the comprehensive understanding of fuel rod behavior and accurate prediction of the lifetime in normal operation and in accident condition as well. Investigations of fuel behavior are carried out in close connection with experimental research operation feedback and computational analyses. In this connection, OECD NEA sets up the “public domain database on nuclear fuel performance experiments for the purpose of code development and validation - International Fuel Performance Experiments (IFPE) database”, with the aim of providing a comprehensive and well-qualified database on UO2 fuel with Zr cladding for model development and code validation. This database includes the data set of the Studsvik Inter-Ramp BWR Project. The objectives of the project are to establish the failure-safe operating limits and the failure mechanism and associated phenomena, during power ramp tests, by varying the design parameters (i.e. cladding heat treatment, gap thickness and fuel density). The experimental data are used for the assessment of the Fission Gas Release (FGR) models implemented in the TRANSURANUS code versions “v1m1j07” and “v1m1j08”. The starting point of the activity is the availability of a “new” transient fission gas release model, the “TFGR model”, specifically implemented in the last code version, to cover power ramp conditions. The paper presents the complete set of simulations of all twenty rods irradiated in the R2 research reactor and the corresponding comparisons with the experimental data. Sensitivity calculations are also performed to address the influence of geometric parameters and the choice of the different code options, relevant to model the FGR, on results.© 2009 ASME

Modeling Large Break-LOCA: In Reactor Fuel Bundle Materials Test MT-4 and MT-6A

When creating power from nuclear fission, the fuel matrix and its cladding constitute the first barrier against radioactive fission product release. Therefore a defense in depth concept requires also the comprehensive understanding of fuel rod behavior and accurate prediction of the lifetime in normal operation and in accident condition as well.Investigations of fuel behavior are carried out in close connection with experimental research, operation feedback and computational analyses. In this connection, OECD/NEA sets up the “public domain database on nuclear fuel performance experiments for the purpose of code development and validation (IFPE)”. This database includes the data set of the projects MT-4 and MT-6A analyzed in the current paper.The MT-4 test bundle simulated a 6×6 section of a 17×17 3% enriched, full-length non-irradiated PWR fuel assembly. There were 20 non-pressurized guard fuel rods to isolate the 12 central, pressurized tests rods; the four corner rods were deleted. In the MT-6A test, the 20 guard rods used in the previous tests were replaced with 9 pressurized thus, a total of 21 test rods were in MT-6A. Only limited destructive post irradiation examination was performed on these two tests.The objective of the activity is the validation of TRANSURANUS “v1m1j09” code in predicting fuel and cladding behavior under LOCA conditions using the experimental databases MT-4 and MT-6A. It is pursued assessing the capabilities of the code models in simulating the phenomena and parameters involved, such as: pressure trend in the fuel rod, cladding creep, α to β-phase phase transformation, oxidation, geometry changes and finally failure prediction. The analysis is aimed at having a comprehensive understanding of the applicability and limitations of the code in the conditions of the experiments. Finally, probabilistic calculations are performed to complete the analysis.The objective of the activity is fulfilled addressing the behavior of two equivalent full lengths fuel rods, one for each test., suitable for the assessment of TU code versions “v1m1j09”.Copyright

Methodology for the Analysis of Fuel Behavior During LOCA and RIA for Licensing Purposes

Licensing requirements vary by country in terms of their scope, range of applicability and numerical values and may imply the use of system thermal hydraulic computer codes. Depending on the specific event scenario and on the purpose of the analysis, it might be required the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes, as for burst temperature, burst strain and flow blockage calculations. This may imply the use of a dedicated fuel rod thermo-mechanical computer code, which can be coupled with thermal-hydraulic system and neutron kinetic codes to be used for the safety analysis.This paper describes the development and the application of a methodology for the analysis of the Large Break Loss of Coolant Accident (LB-LOCA) scenario in Atucha-2 Nuclear Power Plant (NPP), focusing on the procedure adopted for the use of the fuel rod thermo-mechanical code and its application for the safety analysis (Chapter 15 Final Safety Analysis Report, FSAR). The methodology implies the application of best estimate thermal-hydraulic, neutron physics and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient.A strong effort has been performed in order to enhance the fuel behaviour code capabilities and to improve the reliability of the code results.Copyright

Post Test Analysis of LB LOCA Transient in PSB-VVER by CATHARE2 Code

The OECD/NEA PSB-VVER project provided unique and useful experimental data from the large-scale PSB-VVER test facility for code validation. This facility represents the scaled down layout of the Russian designed PWR reactors, namely VVER-1000. The fifth experiment is a LB LOCA in cold leg. The objectives of the test are the investigation of the thermal-hydraulic response of the VVER-1000 following a LB LOCA accident as well as the investigation of the phenomena involved in the transient by the availability of experimental data which are useful for the validation of thermal-hydraulic system codes. The paper discusses the achievements of the assessment of CATHARE2 code from the qualitative and the quantitative evaluations of the results. The results are compared with the analysis carried out by RELAP5-3D code. The quantification of the accuracy is based on the Fast Fourier Transform Base Method, developed at University of Pisa, which provides an integral representation of the accuracy quantification in the frequency domain.Copyright

A Procedure to Address the Fuel Rod Failures During LB-LOCA Transient in Atucha-2 NPP

Depending on the specific event scenario and on the purpose of the analysis, the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes might be required. This may imply the use of a dedicated fuel rod thermomechanical computer code. This paper provides an outline of the methodology for the analysis of the 2A LBLOCA accident in Atucha-2 NPP and describes the procedure adopted for the use of the fuel rod thermomechanical code. The methodology implies the application of best estimate thermalhydraulics, neutron physics, and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient. The procedure consists of a deterministic calculation by the fuel performance code of each individual fuel rod during its lifetime and in the subsequent LB-LOCA transient calculations. The boundary and initial conditions are provided by core physics and three-dimensional neutron kinetic coupled thermal-hydraulic system codes calculations. The procedure is completed by the sensitivity calculations and the application of the probabilistic method, which are outside the scope of the current paper.

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