Abstract: The purpose of the study is to analyze the load rejection transient of ABWR by using TRACE, PARCS, and SNAP codes. This study has some steps. First, using TRACE, PARCS, and SNAP codes establish the model of ABWR. Second, the key parameters are identified to refine the TRACE/PARCS/SNAP model further in the frame of a steady state analysis. Third, the TRACE/PARCS/SNAP model is used to perform the load rejection transient analysis. Finally, the FSAR data are used to compare with the analysis results. The results of TRACE/PARCS are consistent with the FSAR data for the important parameters. It indicates that the TRACE/PARCS/SNAP model of ABWR has a good accuracy in the load rejection transient.
Abstract: To confirm the reactor and containment integrity of the Advanced Boiling Water Reactor (ABWR), we perform the analysis of main steamline break (MSLB) transient by using the TRACE, PARCS, and SNAP codes. The process of the research has four steps. First, the ABWR nuclear power plant (NPP) model is developed by using the above codes. Second, the steady state analysis is performed by using this model. Third, the ABWR model is used to run the analysis of MSLB transient. Fourth, the predictions of TRACE and PARCS are compared with the data of FSAR. The results of TRACE/PARCS and FSAR are similar. According to the TRACE/PARCS results, the reactor and containment integrity of ABWR can be maintained in a safe condition for MSLB.
Abstract: In order to respond the policy decision of non-nuclear homes, Tai Power Company (TPC) will provide the decommissioning project of Kuosheng Nuclear power plant (KSNPP) to meet the regulatory requirement in near future. In this study, the computational fluid dynamics (CFD) methodology has been employed to develop a flow prediction model for boiling water reactor (BWR) with upper pool under decommissioning stage. The model can be utilized to investigate the flow behavior as the vessel combined with upper pool and continuity cooling system. At normal operating condition, different parameters are obtained for the full fluid area, including velocity, mass flow, and mixing phenomenon in the reactor pressure vessel (RPV) and upper pool. Through the efforts of the study, an integrated simulation model will be developed for flow field analysis of decommissioning KSNPP under normal operating condition. It can be expected that a basis result for future analysis application of TPC can be provide from this study.
Abstract: Not only radiation materials, but also the normal chemical material stored in the power plant can cause a risk to the residents. In this research, the ALOHA code was used to perform the concentration analysis under the CO2 storage burst or leakage conditions for Kuosheng nuclear power plant (NPP). The Final Safety Analysis Report (FSAR) and data were used in this study. Additionally, the analysis results of ALOHA code were compared with the R.G. 1.78 failure criteria in order to confirm the control room habitability. The comparison results show that the ALOHA result for burst case was 0.923 g/m3 which was below the criteria. However, the ALOHA results for leakage case was 11.3 g/m3.
Abstract: In this research, the HABIT code was used to estimate the concentration under the CO2 and H2SO4 storage burst conditions for Kuosheng nuclear power plant (NPP). The Final Safety Analysis Report (FSAR) and reports were used in this research. In addition, to evaluate the control room habitability for these cases, the HABIT analysis results were compared with the R.G. 1.78 failure criteria. The comparison results show that the HABIT results are below the criteria. Additionally, some sensitivity studies (stability classification, wind speed and control room intake rate) were performed in this study.
Abstract: Kuosheng nuclear power plant (NPP) is a BWR/6 plant in Taiwan. There is more concern for the safety of Spent Fuel Pools (SFPs) in Taiwan after Fukushima event. In order to estimate the safety of Kuosheng NPP SFP, by using MELCOR2.1 and SNAP, the safety analysis of Kuosheng NPP SFP was performed combined with the mitigation strategy of NEI 06-12 report. There were several steps in this research. First, the Kuosheng NPP SFP models were established by MELCOR2.1/SNAP. Second, the Station Blackout (SBO) analysis of Kuosheng SFP was done by TRACE and MELCOR under the cooling system failure condition. The results showed that the calculations of MELCOR and TRACE were very similar in this case. Second, the mitigation strategy analysis was done with the MELCOR model by following the NEI 06-12 report. The results showed the effectiveness of NEI 06-12 strategy in Kuosheng NPP SFP. Finally, a sensitivity study of SFP quenching was done to check the differences of different water injection time and the phenomena during the quenching. The results showed that if the cladding temperature was over 1600 K, the water injection may have chance to cause the accident more severe with more hydrogen generation. It was because of the oxidation heat and the “Breakaway” effect of the zirconium-water reaction. An animation model built by SNAP was also shown in this study.
Abstract: In this study of ultimate response guideline (URG), Kuosheng BWR/6 nuclear power plant (NPP) TRACE model was established. The reactor depressurization, low pressure water injection, and containment venting are the main actions of URG. This research focuses to evaluate the efficiency of URG under Fukushima-like conditions. Additionally, the sensitivity study of URG was also performed in this research. The analysis results of TRACE present that URG can keep the peak cladding temperature (PCT) below 1088.7 K (the failure criteria) under Fukushima-like conditions. It implied that Kuosheng NPP was at the safe situation.
Abstract: After the measurement uncertainty recapture (MUR) power uprates, Kuosheng nuclear power plant (NPP) was uprated the power from 2894 MWt to 2943 MWt. For power upgrade, several codes (e.g., TRACE, RELAP5, etc.) were applied to assess the safety of Kuosheng NPP. Hence, the main work of this research is to establish a RELAP5/MOD3.3 model of Kuosheng NPP with SNAP interface. The establishment of RELAP5/SNAP model was referred to the FSAR, training documents, and TRACE model which has been developed and verified before. After completing the model establishment, the startup test scenarios would be applied to the RELAP5/SNAP model. With comparing the startup test data and TRACE analysis results, the applicability of RELAP5/SNAP model would be assessed.
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: In this research, TRACE model of Chinshan BWR/4
nuclear power plant (NPP) has been developed for the simulation and
analysis of ultimate response guideline (URG).The main actions of
URG are the depressurization and low pressure water injection of
reactor and containment venting. This research focuses to verify the
URG efficiency under Fukushima-like conditions. TRACE analysis
results show that the URG can keep the PCT below the criteria
1088.7 K under Fukushima-like conditions. It indicated that Chinshan
NPP was safe.
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: In this research, the TRACE/PARCS model of
Lungmen ABWR has been developed for verification of ultimate
response guideline (URG) efficiency. This ultimate measure was
named as DIVing plan, abbreviated from system depressurization,
water injection and containment venting. The simulation initial
condition is 100% rated power/100% rated core flow. This research
focuses on the estimation of the time when the fuel might be damaged
with no water injection by using TRACE/PARCS first. Then, the
effect of the reactor core isolation system (RCIC), control
depressurization and ac-independent water addition system (ACIWA),
which can provide the injection with 950 gpm are also estimated for
the station blackout (SBO) transient.
Abstract: The paper investigates parallel channel instabilities of
natural circulation boiling water reactor. A thermal-hydraulic model
is developed to simulate two-phase flow behavior in the natural circulation boiling water reactor (NCBWR) with the incorporation of
ex-core components and recirculation loop such as steam separator, down-comer, lower-horizontal section and upper-horizontal section
and then, numerical analysis is carried out for parallel channel
instabilities of the reactor undergoing both in-phase and out-of-phase
modes of oscillations. To analyze the relative effect on stability of the reactor due to inclusion of various ex-core components and
recirculation loop, marginal stable point is obtained at a particular inlet enthalpy of the reactor core without the inclusion of ex-core
components and recirculation loop and then with the inclusion of the
same. Numerical simulations are also conducted to determine the
relative dominance between two modes of oscillations i.e. in-phase and out-of-phase. Simulations are also carried out when the channels
are subjected to asymmetric power distribution keeping the inlet enthalpy same.
Abstract: The evaluation of unit cell neutronic parameters and
lifetime for some innovant reactors without on sit-refuling will be
held in this work. the behavior of some small and medium reactors
without on site refueling with triso and cermet fuel. For the FBNR
long life except we propose to change the enrichment of the Cermet
MFE to 9%. For the AFPR reactor we can see that the use of the
Cermet MFE can extend the life of this reactor but to maintain the
same life period for AFPR-SC we most use burnup poison to have the
same slope for Kinf (Burnup). PFPWR50 cell behaves almost in
same way using both fuels Cermet and TRISO. So we can conclude
that PFPWR50 reactor, with CERMET Fuel, is kept among the long
cycle reactors and with the new configuration we avoid subcriticality
at the beginning of cycle. The evaluation of unit cell neutronic
parameters reveals a good agreement with the goal of BWR-PB
concept. It is found out that the Triso fuel assembly lifetime can be
extended for a reasonably long period without being refueled,
approximately up to 48GWd/t burnup. Using coated particles fuels
with the Cermet composition can be more extended the fuel assembly
life time, approximately 52 GWd/t.