19 research outputs found
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
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