Search

The system code ATHLET (Analysis of THermal-hydraulics of Leaks and Transients) is being developed by the Gesellschaft für Anlagen-und Reaktorsicherheit (GRS) mbH in Garching, Germany. In the paper, an overview of activities performed at Faculty of Electrical Engineering and Computing (FER), University of Zagreb, in application of system code ATHLET in transient analyses for NPP Krško (NEK) is presented. Newly, the NEK input deck for the released ATHLET version (Mod 2.2 Cycle A) has been developed. For that purpose, the NEK data base that has been developed and maintained at FER for the last two decades primarily for development of standard input deck for RELAP5 code was used. The ATHLET model has been validated by analyzing the Rod Withdrawal At Power (RWAP) accident at nominal power. The results for steady state calculation as well as RWAP transient were assessed against the analysis performed by RELAP5/mod 3.3 code. In both ATHLET and RELAP5 calculation, the RWAP accident was simulated by constant reactivity insertion rate equal to 2.4 pcm/sec. For ATHLET analysis, two fluid dynamic options were tested for the primary side: a) base case analysis with 5 conservation equations and mixture level model and b) two-fluid model with separate conservation equations for liquid and vapour phases for all the volumes except for the pressurizer where 5 equations+mixture level model was retained. The Steam Generators (SGs) were built using basic ATHLET elements together with the dedicated separator model. For RELAP5/mod 3.3 analysis, a standard option with thermal and mechanical non-equilibrium (6 equations) was used. The results of the steady state calculation for the ATHLET model have shown a very good agreement with RELAP5 calculation. In the transient analysis very small differences for the main physical parameters between ATHLET and RELAP5 as well as between the two ATHLET models were obtained.

The automatic procedure was developed for fuel assembly decay heat calculation based on PARCS 3D burnup calculation for fuel cycle depletion, and ORIGEN 2.1 calculation during both depletion and fuel cooling. Using appropriate pre-processor and post-processor codes it is possible to calculate fuel assembly decay heat loads for all fuel assemblies discharged from reactor. Simple graphical application is then used to distribute fuel assemblies within fuel pool and to calculate any fuel assembly, SFP rack, or whole pool heat load at arbitrary time. The application can be used for overview of fuel assembly burnups, cooling times or decay heats. Based on given date it is possible to calculate whole pool heat load and time to boiling or time to assembly uncover using simple mass and energy balances. Calculated heat loads can be input to more detailed thermal-hydraulics calculations of spent fuel pool. The demonstration calculation was performed for NPP Krsko spent fuel pool.

The analysis of a Station blackout (SBO) accident in the NPP Krško including thermalhydraulic behaviour of the primary system and the containment, as well as the simulation of the core degradation process, release of molten materials and production of hydrogen and other incondensable gases will be presented in the paper. The calculation model includes the latest plant safety upgrade with addition of Passive Autocatalytic Recombiners (PAR) and the Passive Containment Filter Venting (PCFV) system. The code used is MELCOR, version 1.8.6. MELCOR is an integral severe accident code which means that it can simulate both the primary reactor system, including the core, and the containment. The code is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission. The analysis is conducted in two steps. First, the steady state calculation is performed in order to confirm the applicability of the plant model and to obtain correct initial conditions for the accident analysis. The second step is the calculation of the SBO accident with the leakage of the coolant through the damaged reactor coolant pump seals. Without any active safety systems, the reactor pressure vessel will fail after few hours. The mass and energy releases from the primary system cause the containment pressurization and rise of the temperature. The newly added safety systems, PAR and PCFV, prevent the damage of the containment building by keeping the thermalhydraulic conditions below the design limits. The analysis results confirm the capability of the safety systems to effectively control the containment conditions. Results of the analysis are given with respect to the results of the MAAP 4.0.7 analysis for the same accident scenario. The MAAP and MELCOR codes are the most popular severe accident codes and, therefore, it is reasonable to compare their results. In addition, sensitivity calculations performed by varying most influential parameters, such as the hot leg creep failure, blockage of a pipe connecting the cavity and the sump, inclusion of a radionuclide package in the MELCOR, etc. are done in order to demonstrate correct physical behaviour and the accuracy of the developed NPP Krško MELCOR model.

The baffle jetting phenomenon, intensified by pressure difference across baffle plates, may cause excessive fuel rod vibrations and possible damage. The downward flow direction through the baffle-barrel region opposite to the upward flow through the core in pressurized water reactors affects increase of pressure gradient. Gaps between the baffle plates widen with reactor operating time and water jets become more pronounced. During the 2015 outage the NPP Krsko has undergone the upflow conversion (UFC) modification in order to minimize baffle jetting and prevent fuel rod failure. The modification consisted in altering reactor vessel internals in such a way that the coolant downflow path in the baffle-barrel region was converted to an upflow path. The impact of the UFC modification on the nuclear power plant behaviour was analyzed in the paper. The double ended guillotine break of one direct vessel injection (DVI) line was selected as a representative scenario because the DVI line break is exhibiting the highest peak cladding temperature among the small break LOCA cases. The thermal-hydraulic analysis was carried out with the RELAP5/Mod3.3 code for core average channel conditions, while additional FRAPTRAN runs were performed to evaluate peak cladding temperatures for high power and high burnup fuel assemblies. Distribution of pressure and velocity fields in the reactor pressure vessel baffle-barrel region and change in bypass flow rates due to modification were determined by the ANSYS-Fluent calculation. The accident analysis has been performed for both the downflow configuration (pre-UFC) and the upflow configuration (post-UFC) to assess the influence of the modification. Additional cases with different reactor coolant pump trip times, higher core decay heat and reduced capacity of safety systems were also analyzed to evaluate the impact on reactor safety.

NPP Krško input deck developed at Faculty of Electrical Engineering and Computing (FER) Zagreb, for severe accident code MELCOR 1.8.6 is currently being tested. MELCOR is primarily used for the analyses of severe accidents including in-vessel and ex-vessel core melt progression as well as containment response under severe accident conditions. Accurate modelling of the plant thermal-hydraulic behaviour as well as engineering safety features, e.g., Emergency Core Cooling System, Auxiliary feedwater system and various containment systems (e.g., Passive Autocatalytic Recombiners, Fan Coolers and Containment spray) is necessary to correctly predict the plant response and operator actions. For MELCOR input data verification, the comparison of the results for small break (3 inch) cold leg Loss of Coolant Accident (LOCA) for NPP Krško using MELCOR 1.8.6 and RELAP5/MOD 3.3 was performed. A detailed RELAP5/MOD 3.3 model for NPP Krško has been developed at FER and it has been extensively used for accident and transient analyses. The RELAP5 model has been upgraded and improved along with the plant modernization in the year 2000. and after more recent plant modifications. The results of the steady state calculation (first 1000 seconds) for both MELCOR and RELAP5 were assessed against the referent plant data. In order to test all thermal-hydraulic aspects of developed MELCOR 1.8.6 model the accident was analysed, and comparison to the existing RELAP5 model was performed, with all engineering safety features available. After initial fast pressure drop and accumulator injection for both codes stable conditions were established with heat removal through the break and core inventory maintained by safety injection. Transient was simulated for 10000 seconds and overall good agreement between results obtained with both codes was found.

UpFlow Conversion (UFC) was implemented in NPP Krško during the last outage in order to reduce the pressure differential across baffle plates and the possibility of the fuel damage caused by flow induced vibration. The paper describes the coupled code calculation (RELAP5 and PARCS) of MSLB accident at power for pre and post-UFC configuration of reactor vessel. In the calculation, the split model of the reactor vessel was used to better describe asymmetric conditions in loops. It has been demonstrated that the basic parameters (pressure, temperatures) stayed unchanged and there was little change in the flow rates except in baffle-barrel region of the vessel where both flow direction and amount of flow were changed.

The automatic procedure was developed for fuel assembly decay heat calculation based on PARCS 3D burnup calculation for fuel cycle depletion, and ORIGEN 2.1 calculation during both depletion and fuel cooling. Using appropriate pre-processor and post-processor codes it is possible to calculate fuel assembly decay heat loads for all fuel assemblies discharged from reactor. Simple graphical application is then used to distribute fuel assemblies within fuel pool and to calculate any fuel assembly, SFP rack, or whole pool heat load at arbitrary time. The application can be used for overview of fuel assembly burnups, cooling times or decay heats. Based on given date it is possible to calculate whole pool heat load and time to boiling or time to assembly uncover using simple mass and energy balances. Calculated heat loads can be input to more detailed thermal-hydraulics calculations D. Grgić, S. Šadek, V. Benčik, D. Konjarek, Decay heat calculation for spent fuel pool application, Journal of Energy, vol. 64 (2015) Special Issue, p. 90-101 of spent fuel pool. The demonstration calculation was performed for NPP Krsko spent fuel pool.