Open Science Research Excellence

Open Science Index

Commenced in January 2007 Frequency: Monthly Edition: International Paper Count: 9

9
10009905
Using ALOHA Code to Evaluate CO2 Concentration for Maanshan Nuclear Power Plant
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.

8
10009604
Using TRACE and SNAP Codes to Establish the Model of Maanshan PWR for SBO Accident
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.

Keywords:
7
10008606
Design Optimization of the Primary Containment Building of a Pressurized Water Reactor
Abstract:

Primary containment structure is one of the five safety layers of a nuclear facility which is needed to be designed in such a manner that it can withstand the pressure and excessive radioactivity during accidental situations. It is also necessary to ensure minimization of cost with maximum possible safety in order to make the design economically feasible and attractive. This paper attempts to identify the optimum design conditions for primary containment structure considering both mechanical and radiation safety keeping the economic aspects in mind. This work takes advantage of commercial simulation software to identify the suitable conditions without the requirement of costly experiments. Generated data may be helpful for further studies.

6
10007810
A Real Time Expert System for Decision Support in Nuclear Power Plants
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.

5
10007688
Using SNAP and RADTRAD to Establish the Analysis Model for Maanshan PWR Plant
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).
4
10007414
Using HABIT to Establish the Chemicals Analysis Methodology for Maanshan Nuclear Power Plant
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.

3
9997014
The Analysis of TRACE/FRAPTRAN in the Fuel Rods of Maanshan PWR for LBLOCA
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.

 

Keywords:
2
12424
Improved Neutron Leakage Treatment on Nodal Expansion Method for PWR Reactors
Abstract:
For a quick and accurate calculation of spatial neutron distribution in nuclear power reactors 3D nodal codes are usually used aiming at solving the neutron diffusion equation for a given reactor core geometry and material composition. These codes use a second order polynomial to represent the transverse leakage term. In this work, a nodal method based on the well known nodal expansion method (NEM), developed at COPPE, making use of this polynomial expansion was modified to treat the transverse leakage term for the external surfaces of peripheral reflector nodes. The proposed method was implemented into a computational system which, besides solving the diffusion equation, also solves the burnup equations governing the gradual changes in material compositions of the core due to fuel depletion. Results confirm the effectiveness of this modified treatment of peripheral nodes for practical purposes in PWR reactors.
1
1908
Thermo-chemical Characteristics of Powder Fabricated by Oxidation of Spent PWR Fuel
Abstract:
Thermochemcial characteristics of powder fabricated using oxidation treatment of spent PWR fuel and SIMFUEL were evaluated for recycling of spent fuel such as DUPIC process. Especially, the influence of spent fuel burn-ups on the powder fabrication characteristics was experimentally evaluated, ranging from 27,300 to 65,000 MWd/tU. Densities of powder manufactured from an oxidation, OREOX and the milling processes at the same process conditions were compared as a function of the fuel burn-ups respectively. Also, based on chemical analysis results, homogeneity of fissile elements in oxidized powder was confirmed.
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