Abstract: Kuosheng nuclear power plant (NPP) is a BWR/6 plant in Taiwan. There is more concern for the safety of NPPs in Taiwan after Japan Fukushima NPP disaster occurred. Hence, in order to estimate the safety of Kuosheng NPP spent fuel pool (SFP), by using TRACE, MELCOR, and SNAP codes, the safety analysis of Kuosheng NPP SFP was performed. There were two main steps in this research. First, the Kuosheng NPP SFP models were established. Second, the transient analysis of Kuosheng SFP was done by TRACE and MELCOR under the cooling system failure condition (Fukushima-like condition). The results showed that the calculations of MELCOR and TRACE were very similar in this case, and the fuel uncover happened roughly at 4th day after the failure of cooling system. The above results indicated that Kuosheng NPP SFP may be unsafe in the case of long-term SBO situation. In addition, future calculations were needed to be done by the other codes like FRAPTRAN for the cladding calculations.
Abstract: TRACE is developed by U.S. NRC for the nuclear
power plants (NPPs) safety analysis. We focus on the establishment
and application of TRACE/FRAPTRAN/SNAP models for Chinshan
NPP (BWR/4) spent fuel pool in this research. The geometry is 12.17
m × 7.87 m × 11.61 m for the spent fuel pool. In this study, there are
three TRACE/SNAP models: one-channel, two-channel, and
multi-channel TRACE/SNAP model. Additionally, the cooling system
failure of the spent fuel pool was simulated and analyzed by using the
above models. According to the analysis results, the peak cladding
temperature response was more accurate in the multi-channel
TRACE/SNAP model. The results depicted that the uncovered of the
fuels occurred at 2.7 day after the cooling system failed. In order to
estimate the detailed fuel rods performance, FRAPTRAN code was
used in this research. According to the results of FRAPTRAN, the
highest cladding temperature located on the node 21 of the fuel rod
(the highest node at node 23) and the cladding burst roughly after 3.7
day.
Abstract: Kuosheng nuclear power plant (NPP) is a BWR/6 type
NPP and located on the northern coast of Taiwan. First, Kuosheng
NPP TRACE model were developed in this research. In order to assess
the system response of Kuosheng NPP TRACE model, startup tests
data were used to evaluate Kuosheng NPP TRACE model. Second, the
overpressurization transient analysis of Kuosheng NPP TRACE model
was performed. Besides, in order to confirm the mechanical property
and integrity of fuel rods, FRAPTRAN analysis was also performed in
this study.
Abstract: The dry-storage systems of nuclear power plants (NPPs) in Taiwan have become one of the major safety concerns. There are two steps considered in this study. The first step is the verification of the TRACE by using VSC-17 experimental data. The results of TRACE were similar to the VSC-17 data. It indicates that TRACE has the respectable accuracy in the simulation and analysis of the dry-storage systems. The next step is the application of TRACE in the dry-storage system of Kuosheng NPP (BWR/6). Kuosheng NPP is the second BWR NPP of Taiwan Power Company. In order to solve the storage of the spent fuels, Taiwan Power Company developed the new dry-storage system for Kuosheng NPP. In this step, the dry-storage system model of Kuosheng NPP was established by TRACE. Then, the steady state simulation of this model was performed and the results of TRACE were compared with the Kuosheng NPP data. Finally, this model was used to perform the safety analysis of Kuosheng NPP dry-storage system. Besides, FRAPTRAN was used tocalculate the transient performance of fuel rods.
Abstract: Fuel rod analysis program transient (FRAPTRAN)
code was used to study the fuel rod performance during a postulated
large break loss of coolant accident (LBLOCA) in Maanshan nuclear
power plant (NPP). Previous transient results from thermal hydraulic
code, TRACE, with the same LBLOCA scenario, were used as input
boundary conditions for FRAPTRAN. The simulation results showed
that the peak cladding temperatures and the fuel centerline
temperatures were all below the 10CFR50.46 LOCA criteria. In
addition, the maximum hoop stress was 18 MPa and the oxide
thickness was 0.003mm for the present simulation cases, which are all
within the safety operation ranges. The present study confirms that this
analysis method, the FRAPTRAN code combined with TRACE, is an
appropriate approach to predict the fuel integrity under LBLOCA with
operational ECCS.
Abstract: This analysis of Kuosheng nuclear power plant (NPP)
was performed mainly by TRACE, assisted with FRAPTRAN and
FRAPCON. SNAP v2.2.1 and TRACE v5.0p3 are used to develop the
Kuosheng NPP SPU TRACE model which can simulate the turbine
trip without bypass transient. From the analysis of TRACE, the
important parameters such as dome pressure, coolant temperature and
pressure can be determined. Through these parameters, comparing
with the criteria which were formulated by United States Nuclear
Regulatory Commission (U.S. NRC), we can determine whether the
Kuoshengnuclear power plant failed or not in the accident analysis.
However, from the data of TRACE, the fuel rods status cannot be
determined. With the information from TRACE and burn-up analysis
obtained from FRAPCON, FRAPTRAN analyzes more details about
the fuel rods in this transient. Besides, through the SNAP interface, the
data results can be presented as an animation. From the animation, the
TRACE and FRAPTRAN data can be merged together that may be
realized by the readers more easily. In this research, TRACE showed
that the maximum dome pressure of the reactor reaches to 8.32 MPa,
which is lower than the acceptance limit 9.58 MPa. Furthermore,
FRAPTRAN revels that the maximum strain is about 0.00165, which
is below the criteria 0.01. In addition, cladding enthalpy is 52.44 cal/g
which is lower than 170 cal/g specified by the USNRC NUREG-0800
Standard Review Plan.