[en] A Loss of Offsite Power (LOOP) transient scenario is based on a complete loss of non-emergency AC power that results in the loss of all power to the plant auxiliaries, i.e., the Reactor Coolant Pumps (RCPs), condensate pumps, etc. An actual LOOP event would cause a loss of all feedwater, a loss of forced Reactor Coolant System (RCS) flow and a reactor trip within less than 2 seconds as a result of either loss of power to the rod cluster assembly gripper coils or any RCS flow trips. For safety analysis purposes the LOOP event is conservatively modelled as a Loss of Normal Feedwater (LONF) transient with a subsequent loss of offsite power as a result of a reactor trip. The reactor trip followed by RCP trip are delayed until a low-low Steam Generator (SG) level signal is reached. This is a more conservative scenario than the LOOP event because the least amount of SG secondary side water mass available for heat removal and the increased amount of the stored energy in the primary circuit at the time of the loss of RCS flow result. The standard LOOP safety analysis is aimed to demonstrate the natural circulation capability of the RCS to remove residual and decay heat from the core aided by Auxiliary Feedwater in the secondary system. In addition to this goal the presented work is aimed to resolve the potential safety issue resulting from the influence of the Chemical and Volume Control System (CVCS) operation during LOOP event for NPP Krsko. The potential safety concern for the LOOP analysis is that the loss of instrument air system may occur thus leading to the CVCS charging and letdown flow imbalance. A net RCS inventory addition may result with water solid pressurizer condition. Water discharge through the pressurizer relief and safety valves could lead to overpressurization of the Pressurizer Relief Tank (PRT) and rupture of the PRT rupture disks. Additional concern is that pressurizer relief and safety valves may fail to properly reseat when exposed to water relief causing the American Nuclear Society (ANS) condition II to progress to the more severe condition III small break Loss of Coolant Accident (LOCA). To address the pressurizer water-solid concern for NPP Krsko RELAP5/MOD3.3 analyses of the LOOP event for best-estimate and USAR based scenarios have been performed. Different CVCS charging and letdown operation modes including the most conservative case with CVCS charging flow at maximum and letdown flow isolated were analyzed. (author)

[en] The analysis of a Station blackout (SBO) accident in the NPP Krsko including thermal-hydraulic 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 thermal-hydraulic 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 Krsko MELCOR model. (author).

[en] Asymmetric Main Feedwater Isolation (AMFWI) transient in one Steam Generator (SG) for NPP Krsko using RELAP5 standalone code and coupled code RELAP5- QUABOX/CUBBOX (R5QC) was analyzed. In the RELAP5 standalone calculation, a point kinetics model was used, while in the coupled code a three-dimensional (3D) neutronics model of QUABOX with different RELAP5 nodalization schemes of reactor vessel was used. Both code versions use best-estimate thermal-hydraulic system code for all components in the plant and include realistic description of plant protection and control systems. Two different types of calculations were performed: with and without automatic control rod system available. The AMFWI transient causes the great asymmetry of the transferred heat in the SGs and subsequently the asymmetry of the power produced across the core due to different reactivity feedback resulting from the thermal-hydraulic channels assigned to different loops. The work presented in the paper is a part of validation of the 3D coupled code R5QC in the analysis of asymmetric transients.(author)

[en] Results of two large load rejection cases: 100-15 % and 100-50 % for point kinetics (RELAP5 standalone code) and 3D coupled code RELAP5-QUABOX/CUBBOX are analyzed. A coupled code comprises the multidimensional neutron kinetics code model of QUABOX/CUBBOX and thermal hydraulic system code RELAP5. This coupled code version enables best-estimate analyses of dynamic behavior of nuclear power plant (NPP) due to two facts. First, a realistic simulation of the reactor core in three dimensions (3D) is performed. Secondly, a sophisticated thermal hydraulic code RELAP5 provides a respective thermal hydraulic analysis for all components in the plant on one side and on the other side its control variables and trips enable the realistic description of plant protection and control systems. A RELAP5/mod3 input data set for NPP Kriko at uprated conditions (1.06 PN) after steam generator (SG) replacement was extended with realistic feedwater system and all major control systems. (author)

[en] In the paper the results of the RELAP5/MOD 3.3 analysis of the Reactor Coolant Pump (RCP) Trip event at NPP Krsko are presented. The event was initiated by an operator action aimed to prevent the RCP 2 bearing damage. The action consisted of a power reduction, that lasted for 50 minutes, followed by a reactor and a subsequent RCP 2 trip when the reactor power was reduced to 28 %. Two minutes after reactor trip, the Main Steam Isolation Valves (MSIVs) were isolated and the steam dump flow was closed. On the secondary side the Steam Generator (SG) pressure rose until SG 1 Safety Valve (SV) 1 opened. The realistic RELAP5/MOD 3.3 analysis has been performed assuming realistic equipment behavior and operator actions. The comparison of the RELAP5/MOD 3.3 results for the realistic analysis with the measurement has shown small differences for the major parameters (nuclear power, average temperature, secondary pressure) for both power reduction transient (100 - 28 %) and pump trip event. Four additional RELAP5/MOD3.3 analyses with different transient scenarios that contribute to Conditional Core Damage Probability (CCDP) in the pump trip event were performed. (author)

[en] In the paper the results of the RELAP5/MOD 3.3 analysis of the Reactor Coolant Pump (RCP) Trip event at NPP Krsko are presented. The event was initiated by an operator action aimed to prevent the RCP 2 bearing damage. The action consisted of a power reduction, that lasted for 50 minutes, followed by a reactor and a subsequent RCP 2 trip when the reactor power was reduced to 28 %. Two minutes after reactor trip, the Main Steam Isolation Valves (MSIV) were isolated and the steam dump flow was closed. On the secondary side the Steam Generator (SG) pressure rose until SG 1 Safety Valve (SV) 1 opened. The realistic RELAP5/MOD 3.3 analysis has been performed in order to model the particular plant behavior caused by operator actions. The comparison of the RELAP5/MOD 3.3 results with the measurement for the power reduction transient has shown small differences for the major parameters (nuclear power, average temperature, secondary pressure). The main trends and physical phenomena following the RCP Trip event were well reproduced in the analysis. The parameters that have the major influence on transient results have been identified. In the paper the influence of SG 1 relief and SV valves on transient results was investigated more closely. (author)

[en] A simulation of the loss of Residual Heat Removal (RHR) system during midloop operation was performed using the ATHLET Mod 1.1 C system code. The analysis is a posttest calculation of the 6.9 c test performed on the BETHSY facility on April 14, 1992. Test 6.9 c has been analysed within an OECD/NEA/CSNI project as an open exercise named International Standard Problem (ISP)38. BETHSY is an integral test facility which simulates a three loop 900 Mwe (2775 MWt) Framatome PWR. The test consisted in a loss of RHR system at a cold shutdown and midloop condition (i.e. the liquid level was at the axis of the cold legs) and at atmospheric pressure. Along with loss of RHR system, the pressurizer and Steam Generator (SG) 1 outlet plenum manways were open. In the paper the results of the ATHLET calculations were assessed against the measured data. The initial conditions were determined during the first 200 s of the simulation. The transient was simulated for 8800 s. (author)

[en] In 1986 the US NRC published a proposed rule making to tighten the requirements on nuclear plants with regard to their ability to deal safely with a total loss of all a-c electric power, both from external and internal sources of supply. The proposed rule would require all licensees and applicants to: 1. Assess the capability of their plants to cope with a total loss all a-c power(that is, determine the amount of time the plant could maintain core cooling and containment integrity with a-c power unavailable); 2. Have procedures and training to cope with such an event; 3. Make modification, if necessary, to cope with an acceptable minimum duration loss of all a-c power. A total loss of all a-c electric power has been identified as an 'unresolved safety issue' and subjected to considerably study. The article presents an idea to resolve this issue. (author)

[en] In the paper the results of RELAP5/mod3.3 calculations of critical parameters during shutdown for NPP Krsko are presented. Conservative evaluations have been performed at NPP Krsko to determine the minimum configuration of systems required for the safe shutdown operation. Critical parameters in these evaluations are defined as the time to start of the boiling and the time of the core dry-out. In order to have better insight into the available margins, the best estimate code RELAP5/mod3.3 has been used to calculate the same parameters. The analyzed transient is the loss of the Residual Heat Removal (RHR) system, which is used to remove decay heat during shutdown conditions. Several configurations that include open and closed Reactor Coolant System (RCS) were considered in the evaluation. The RELAP5/mod3.3 analysis of the loss of the RHR system has been performed for the following cases: 1) RCS closed and water solid, 2) RCS closed and partially drained, 3) Pressurizer manway open, Steam Generator (SG) U tubes partially drained, 4) Pressurizer and SG manways open, SG U tubes completely drained, 5) Pressurizer manway open, SGs drained, SG nozzle dams installed and 6) SG nozzle dams installed, pressurizer manway open, 1 inch break at RHR pump discharge in the loop with pressurizer. Both RHR trains were assumed in operation prior to start of the transient. The maximum average steady state temperature for all analyzed cases was limited to 333 K. (author)

[en] Best estimate codes and methods for the realistic simulation of operational transients and accidents are being developed in two directions. First, computer codes with models of the interaction between multidimensional neutron kinetic and NPP dynamic behavior enable realistic simulation of transients characterized by strong coupling between neutronics and thermal-hydraulics as well as of transients that result in asymmetrical spatial core power distribution. Coupled codes consisting of a system thermal-hydraulic code and a multidimensional neutronic code are being developed worldwide in order to accomplish that task. Secondly, development of the qualified plant nodalization and of the models of plant protection and control systems is important for the realistic system analysis of operational transients and accidents. Comparison of the coupled code and point kinetic results is important for the validation of the coupled code and to gain more experience in the use of the coupled code in realistic analyses. In this paper the results of two transients for NPP Krsko using the coupled code RELAP5-QUABOX/CUBBOX (R5QC) and RELAP5 stand alone code are discussed. (orig.)