Around the world, countries are increasingly considering carbon-free energy generation
options as the threat of climate change grows. Small modular reactor designs,
promising such carbon-free energy generation, are thriving worldwide with novel and
innovative technologies that improve safety as well as economic performance. Canada
is also considering small modular reactors (SMRs) as a means of achieving net zero
carbon emissions by 2050.
Some of these reactor designs utilize pressurized water for cooling and moderator.
Reactors with pressurized water have been subjected to steam generator tube ruptures
in the past, and a detailed investigation into the possible consequences of such incidents
in SMRs should be conducted.
In this research, a model for one of the newer designs, the NuScale Integrated Small
Modular Reactor, was developed with the RELAP5-3D code for assessing safety related
transients. The NuScale Small Modular Reactor incorporates helical coil steam
generators within its reactor pressure vessel, which are more efficient in terms of heat
transfer than the U-tube steam generators that are widely used in nuclear reactors.
In the first part of the research, a detailed model is created and used to obtain steady
state conditions with parameters collected from NuScale’s Final Safety Analysis Report
(FSAR). The Steam Generator Tube Rupture event is analyzed in the second part
of the work. Slight differences in the broken and intact steam generator pressures as
well as decay heat removal system flow rates are seen in the comparison of reference
values and calculated results. Among the reasons for those differences could be that
the correlations used by the RELAP5-3D code for heat transfer coefficient and pressure
drop in the helical coil steam generators are different than those of the NuScale proprietary
code NRELAP5, with which the analyses have been performed in the FSAR.
Also, post-dryout heat transfer at the exit of helical coil steam generators and evaporator
sections could cause differences in the outlet conditions of the steam, resulting in
different mass flow rates as well.
The final section of the research simulates a comparable but more severe tube rupture
incident without the availability of decay heat removal systems in order to assess
the reactor’s emergency core cooling system reaction. Passive decay heat removal systems
are crucial components for removing heat after reactor shutdown through heat
exchangers that are submerged in the reactor pool and connected to steam generators
by a closed loop. The containment pressures, the containment wall temperatures, and
the peak fuel clad temperatures are considered to be the key design constraints that
must be observed.
Future work on this subject could include modifying the source code, adding specific
correlations for helical coil steam generators, and comparing the results, as well
as quantifying uncertainties in the SGTR event. Main parameters in the quantification
of uncertainties would be reactor power, single phase and two-phase discharge coefficients
from the break, trip signals and delays as well as break size and location. / Thesis / Master of Applied Science (MASc)
| Identifer | oai:union.ndltd.org:mcmaster.ca/oai:macsphere.mcmaster.ca:11375/27539 |
| Date | January 2022 |
| Creators | Alkan, Cahit |
| Contributors | Buijs, Adriaan, Engineering Physics and Nuclear Engineering |
| Source Sets | McMaster University |
| Language | English |
| Detected Language | English |
| Type | Thesis |
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