NPP KRŠKO CONTAINMENT MODELLING WITH THE ASTEC CODE

ASTEC is an integral computer code jointly developed by Institut de Radioprotection et de Sûreté Nucléaire (IRSN, France) and Gesellschaft für Anlagenund Reaktorsicherheit (GRS, Germany) to assess nuclear power plant behaviour during a severe accident (SA). The ASTEC code was used to model and to simulate NPP behaviour during a postulated station blackout accident in the NPP Krško. The accident analysis was focused on containment behaviour; however the complete integral NPP analysis was carried out in order to provide correct boundary conditions for the containment calculation. During the accident, the containment integrity was challenged by release of reactor system coolant through degraded coolant pump seals, molten corium concrete interaction and direct containment heating mechanisms. Impact of those processes on relevant containment parameters, such as compartments pressures and temperatures, is going to be discussed in the paper

Operation and Performance Analysis of Steam Generators in Nuclear Power Plants

Steam generators are components in which heat produced in the reactor core is transferred to the secondary side, the steam supply system, of the nuclear power plant (NPP). Steam generators (SGs) have to fulfil special nuclear regulatory requirements regarding their size, selection of materials, pressure loads, impact on the NPP safety, etc. The primary-side fluid is liquid water at the high pressure, and the fluid on the secondary side is saturated water-steam mixture at the pressure twice as low. A special attention is given to preserving the boundary between the contaminated water in the primary reactor coolant system and the water-steam mixture in the secondary system. A brief overview of the SG design, its operation and the mathematical correlations used to quantify heat transfer is given in the chapter. Results of the SG transient behaviour obtained by the simulation with the best-estimate computer code RELAP5, developed for safety analyses of NPPs, are also presented. Two types of steam generators are analyzed: the inverted U-tube SG, which is commonly found in the present-day pressurized water reactors and the helical-coil SG that is part of the new-generation reactor designs

NPP Krško Station Blackout Analysis after Safety Upgrade Using MELCOR Code

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

Verification of GOTHIC Multivolume Containment Model during NPP Krško DBA LOCA

New containment multivolume model of NPP Krsko for GOTHIC code is developed. It is based on plant drawings and other available data. It is supported by developed SketchUp 3D containment model. The model is subdivided in volumes following physical boundaries and clearly defined flow paths. All important concrete heat structures are taken into account. Metal heat structures are based on plant’s SAR Chapter 6 licensing model. RCFC (Reactor Containment Fan Cooler) units are explicitly modelled as well as all main ventilation ducts. The model includes two trains of containment spray system. PARs (Passive Autocatalytic Recombiner) and PCFV (Passive Containment Filter Venting) filters added during plant safety upgrade project are part of the model too. It was intention to use model for both DBA (Design Basis Accident) and for DEC (Design Extended Conditions) and BDBA (Beyond Design Basis Accident) calculations. Based on the same discretization and data, and on experience acquired during GOTHIC model development and use, containment models for MELCOR and MAAP integral codes are developed too. As part of initial verification of the GOTHIC model containment DBA LOCA calculation is performed using SAR MER (Mass and Energy Release) data. The influence of different break positions on peak containment atmosphere pressure and temperature was studied. The results were compared against results obtained in single volume containment licensing model. Beside local effects due to different containment subdivision similar results are obtained when comparing containment dome from multivolume and the single volume in licensing model. Special attention was paid to distribution of water in lower part of the containment during recirculation phase. In this case much more valuable information are obtained in multivolume model with explicit volumes for main sump, recirculation sump and sump pit. Another point of interest was influence of containment spray duration on long term pressure and temperature behaviour. The intention was to study consistency of assumed different spray operation times used in safety analyses, EQ analyses and SAMGs and related consequences for plant operation during DBA LOCA

NPP Krško Station Blackout Analysis after Safety Upgrade Using MELCOR Code

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

OPTIMIZATION OF OPDT PROTECTION FOR OVERCOOLING ACCIDENTS

Overcooling accidents are typically resulting in power increase due to negative moderator feedback. There are more protection set points responsible for terminating power increase. OPDT protection set point is typically protection from exceeding fuel centre line temperature due to reactivity and power increase. It is important to actuate reactor trip signal early enough, but to be able to filter out events where actuation is not necessary. Different concepts of coolant temperature compensation as part of OPDT set point protection were studied for decrease of feedwater temperature accident and for small main steam line breaks from full power for NPP Krško. Computer code RELAP5/mod 3.3 was used in calculation. The influence of different assumptions in accident description as well as nuclear core characteristics were described

DECAY HEAT CALCULATION FOR SPENT FUEL POOL APPLICATION

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

NPP Krško 3 inch Cold Leg Break LOCA Calculation using RELAP5/MOD 3.3 and MELCOR 1.8.6 Codes

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

DECAY HEAT CALCULATION FOR SPENT FUEL POOL APPLICATION

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

OPTIMIZATION OF OPDT PROTECTION FOR OVERCOOLING ACCIDENTS

Overcooling accidents are typically resulting in power increase due to negative moderator feedback. There are more protection set points responsible for terminating power increase. OPDT protection set point is typically protection from exceeding fuel centre line temperature due to reactivity and power increase. It is important to actuate reactor trip signal early enough, but to be able to filter out events where actuation is not necessary. Different concepts of coolant temperature compensation as part of OPDT set point protection were studied for decrease of feedwater temperature accident and for small main steam line breaks from full power for NPP Krško. Computer code RELAP5/mod 3.3 was used in calculation. The influence of different assumptions in accident description as well as nuclear core characteristics were described