Abstract: ALOHA code was used to calculate the concentration under the CO2 storage burst condition for Maanshan nuclear power plant (NPP) in this study. Five main data are input into ALOHA code including location, building, chemical, atmospheric, and source data. The data from Final Safety Analysis Report (FSAR) and some reports were used in this study. The ALOHA results are compared with the failure criteria of R.G. 1.78 to confirm the habitability of control room. The result of comparison presents that the ALOHA result is below the R.G. 1.78 criteria. This implies that the habitability of control room can be maintained in this case. The sensitivity study for atmospheric parameters was performed in this study. The results show that the wind speed has the larger effect in the concentration calculation.
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 this research, TRACE code with the interface code-SNAP was used to simulate and analyze the SBO (station blackout) accident which occurred in Maanshan PWR (pressurized water reactor) nuclear power plant (NPP). There are four main steps in this research. First, the SBO accident data of Maanshan NPP were collected. Second, the TRACE/SNAP model of Maanshan NPP was established by using these data. Third, this TRACE/SNAP model was used to perform the simulation and analysis of SBO accident. Finally, the simulation and analysis of SBO with mitigation equipments was performed. The analysis results of TRACE are consistent with the data of Maanshan NPP. The mitigation equipments of Maanshan can maintain the safety of Maanshan in the SBO according to the TRACE predictions.
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: In case of abnormal situations, the nuclear power plant (NPP) operators must follow written procedures to check the condition of the plant and to classify the type of emergency. In this paper, we proposed a Real Time Expert System in order to improve operator’s performance in case of transient or accident with reactor shutdown. The expert system’s knowledge is based on the sequence of events (SoE) of known accident and two emergency procedures of the Brazilian Pressurized Water Reactor (PWR) NPP and uses two kinds of knowledge representation: rule and logic trees. The results show that the system was able to classify the response of the automatic protection systems, as well as to evaluate the conditions of the plant, diagnosing the type of occurrence, recovery procedure to be followed, indicating the shutdown root cause, and classifying the emergency level.
Abstract: In this study, we focus on the establishment of the analysis model for Maanshan PWR nuclear power plant (NPP) by using RADTRAD and SNAP codes with the FSAR, manuals, and other data. In order to evaluate the cumulative dose at the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) outer boundary, Maanshan NPP RADTRAD/SNAP model was used to perform the analysis of the DBA LOCA case. The analysis results of RADTRAD were similar to FSAR data. These analysis results were lower than the failure criteria of 10 CFR 100.11 (a total radiation dose to the whole body, 250 mSv; a total radiation dose to the thyroid from iodine exposure, 3000 mSv).
Abstract: In this research, the HABIT analysis methodology was established for Maanshan nuclear power plant (NPP). The Final Safety Analysis Report (FSAR), reports, and other data were used in this study. To evaluate the control room habitability under the CO2 storage burst, the HABIT methodology was used to perform this analysis. The HABIT result was below the R.G. 1.78 failure criteria. This indicates that Maanshan NPP habitability can be maintained. Additionally, the sensitivity study of the parameters (wind speed, atmospheric stability classification, air temperature, and control room intake flow rate) was also performed in this research.
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: A Multi-dimensional computational fluid dynamics
(CFD) two-phase model was developed with the aim to simulate
the in-core coolant circuit of a pressurized heavy water reactor
(PHWR) of a commercial nuclear power plant (NPP). Due to the
fact that this PHWR is a Reactor Pressure Vessel type (RPV),
three-dimensional (3D) detailed modelling of the large reservoirs of
the RPV (the upper and lower plenums and the downcomer) were
coupled with an in-house finite volume one-dimensional (1D) code
in order to model the 451 coolant channels housing the nuclear fuel.
Regarding the 1D code, suitable empirical correlations for taking into
account the in-channel distributed (friction losses) and concentrated
(spacer grids, inlet and outlet throttles) pressure losses were used.
A local power distribution at each one of the coolant channels
was also taken into account. The heat transfer between the coolant
and the surrounding moderator was accurately calculated using a
two-dimensional theoretical model. The implementation of subcooled
boiling and condensation models in the 1D code along with the use
of functions for representing the thermal and dynamic properties of
the coolant and moderator (heavy water) allow to have estimations
of the in-core steam generation under nominal flow conditions for a
generic fission power distribution. The in-core mass flow distribution
results for steady state nominal conditions are in agreement with the
expected from design, thus getting a first assessment of the coupled
1/3D model. Results for nominal condition were compared with
those obtained with a previous 1/3D single-phase model getting more
realistic temperature patterns, also allowing visualize low values of
void fraction inside the upper plenum. It must be mentioned that the
current results were obtained by imposing prescribed fission power
functions from literature. Therefore, results are showed with the aim
of point out the potentiality of the developed model.
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: 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.
Abstract: The main features of NPP-2006/MIR-1200 design are
described. Estimation of individual doses for population under
normal operation and accident conditions is performed for
Leningradskaya NPP – 2 as an example. The radiation effect on
population and environment doesn-t exceed the established
normative limit and is as low as reasonably achievable. NPP-
2006/MIR-1200 design meets all Russian and international
requirements for power units under construction.