14 research outputs found
Analysis of Rod Withdrawal at Power (RWAP) Accident using ATHLET Mod 2.2 Cycle A and RELAP5/mod 3.3 Codes
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
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 Post-UFC Transient Response during MSLB
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
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