• Nuclear Engineering and Technology
• /
• 제52권5호
• /
• pp.937-946
• /
• 2020

The aim in the present work is to simulate accident scenarios of AP1000 during the small-break loss-of-coolant accident (SBLOCA) and investigate the performance and behavior of automatic depressurization system (ADS) during accidents by using MIDAC (The Module In-vessel Degradation severe accident Analysis Code). Four types of accidents with different hypothetical conditions were analyzed in this study. The impact on the thermal-hydraulic of the reactor coolant system (RCS), the passive core cooling system and core degradation was researched by comparing these types. The results show that the RCS depressurization becomes faster, the core makeup tanks (CMT) and accumulators (ACC) are activated earlier and the effect of gravity water injection is more obvious along with more ADS valves open. The open of the only ADS1-3 can't stop the core degradation on the basis of the first type of the accident. The open of ADS1-3 has a great impact on the injection time of ACC and CMT. The core can remain intact for a long time and the core degradation can be prevent by the open of ADS-4. The all results are significant and meaningful to understand the performance and behavior of the ADS during the typical SBLOCA.

• Nuclear Engineering and Technology
• /
• 제54권4호
• /
• pp.1508-1515
• /
• 2022

In the present study, the cavity module of the MELCOR code is used for the simulation of molten corium concrete interaction (MCCI) during the late phase of postulated large break loss of coolant (LB-LOCA) accident in the APR1400 reactor design. Using the molten corium composition data from previous MELCOR Simulation of APR1400 under LB-LOCA accident, the ex-vessel phases of the accident sequences with long-term MCCI are recalculated with stand-alone cavity package of the MELCOR code to investigate the impact of water ingression and melt eruption models which were hitherto absent in MELCOR code. Significant changes in the MCCI behaviors in terms of the heat transfer rates, amount of gases released, and maximum cavity ablation depths are observed and reported in this study. Most especially, the incorporation of these models in the new release of MELCOR code has led to the reduction of the maximum ablation depth in radial and axial directions by ~38% and ~32%, respectively. These impacts are substantial enough to change the conclusions earlier reached by researchers who had used the older versions of the MELCOR code for their studies. and it could also impact the estimated cost of the severe accident mitigation system in the APR1400 reactor.

• Nuclear Engineering and Technology
• /
• 제46권3호
• /
• pp.373-380
• /
• 2014

Safety-related parameters are very important for confirming the status of a nuclear power plant. In particular, the reactor vessel water level has a direct impact on the safety fortress by confirming reactor core cooling. In this study, the reactor vessel water level under the condition of a severe accident, where the water level could not be measured, was predicted using a fuzzy neural network (FNN). The prediction model was developed using training data, and validated using independent test data. The data was generated from simulations of the optimized power reactor 1000 (OPR1000) using MAAP4 code. The informative data for training the FNN model was selected using the subtractive clustering method. The prediction performance of the reactor vessel water level was quite satisfactory, but a few large errors were occasionally observed. To check the effect of instrument errors, the prediction model was verified using data containing artificially added errors. The developed FNN model was sufficiently accurate to be used to predict the reactor vessel water level in severe accident situations where the integrity of the reactor vessel water level sensor is compromised. Furthermore, if the developed FNN model can be optimized using a variety of data, it should be possible to predict the reactor vessel water level precisely.

• Nuclear Engineering and Technology
• /
• 제54권1호
• /
• pp.177-185
• /
• 2022

Hydrogen risk in the containment filtered venting system (CFVS) vessel was analyzed, considering operation pressure and modes with the effect of PAR and accident scenarios. The CFVS is to depressurize the containment by venting the containment atmosphere through the filtering system. The CFVS could be subject to hydrogen risk due to the change of atmospheric conditions while the containment atmosphere passes through the CFVS. It was found that hydrogen risk increased as the CFVS opening pressure was set higher because more combustible gases generated by Molten Core Concrete Interaction flowed into the CFVS. Hydrogen risk was independent of operation modes and found only at the early phase of venting both for continuous and cyclic operation modes. With PAR, hydrogen risk appeared only at the 0.9 MPa opening pressure for Station Black-Out accidents. Without PAR, however, hydrogen risk appeared even with the CFVS opening set-point of 0.5 MPa. In a slow accident like SBO, hydrogen risk was more threatening than a fast accident like Large Break Loss-of-Coolant Accident. Through this study, it is recommended to set the CFVS opening pressure lower than 0.9 MPa and to operate it in the cyclic mode to keep the CFVS available as long as possible.

• Nuclear Engineering and Technology
• /
• 제51권3호
• /
• pp.723-730
• /
• 2019

Acquiring instrumentation signals generated from nuclear power plants (NPPs) is essential to maintain nuclear reactor integrity or to mitigate an abnormal state under normal operating conditions or severe accident circumstances. However, various safety-critical instrumentation signals from NPPs cannot be accurately measured on account of instrument degradation or failure under severe accident circumstances. Reactor vessel (RV) water level, which is an accident monitoring variable directly related to reactor cooling and prevention of core exposure, was predicted by applying a few signals to deep neural networks (DNNs) during severe accidents in NPPs. Signal data were obtained by simulating the postulated loss-of-coolant accidents at hot- and cold-legs, and steam generator tube rupture using modular accident analysis program code as actual NPP accidents rarely happen. To optimize the DNN model for RV water level prediction, a genetic algorithm was used to select the numbers of hidden layers and nodes. The proposed DNN model had a small root mean square error for RV water level prediction, and performed better than the cascaded fuzzy neural network model of the previous study. Consequently, the DNN model is considered to perform well enough to provide supporting information on the RV water level to operators.

• Nuclear Engineering and Technology
• /
• 제46권4호
• /
• pp.541-546
• /
• 2014

Probabilistic Safety Assessment (PSA) has been widely used to estimate the overall safety of nuclear power plants (NPP) and it provides base information for risk informed application (RIA) and risk informed regulation (RIR). For the effective and correct use of PSA in RIA/RIR related decision making, the risk estimated by a PSA model should be as realistic as possible. In this work, a best-estimate thermal-hydraulic analysis of loss-of-coolant accidents (LOCAs) for the Hanul Nuclear Units 3&4 is first carried out in a systematic way. That is, the behaviors of peak cladding temperature (PCT) were analyzed with various combinations of break sizes, the operating conditions of safety systems, and the operator's action time for aggressive secondary cooling. Thereafter, the results of the thermal-hydraulic analysis have been reflected in the improvement of the PSA model by changing both accident sequences and success criteria of the event trees for the LOCA scenarios.

• Nuclear Engineering and Technology
• /
• 제34권6호
• /
• pp.596-609
• /
• 2002

Korea Electric Power Research Institute has developed the in-house safety analysis methodologies for non-LOCA(Loss Of Coolant Accident) events based on codes and methodologies of vendors and Electric Power Research Institute . According to the new methodologies, analyses of system responses and calculation of DNBR(Departure from Nucleate Boiling Ratio) during the transient have been carried out with RETRAN code and a sub-channel analysis code, respectively. However, it takes too much time to calculate DNBR for each case using the two codes to search for the limiting case from sensitivity study. To simplify the search for the limiting case, accordingly, RETRAN code has been modified to roughly calculate DNBR using hot channel modeling. The W-3 correlation is already included in RETRAN as one of the auxiliary DNBR models. However, WRB-1 and WRB-2 correlations required to analyze some Westinghouse type fuels are not considered in RETRAN DNBR models. In this paper, the RETRAN DNBR models using the correlations have been developed and the partial and complete loss of forced reactor coolant flow events have been analyzed for Yonggwang units 1 and 2 with the new methodologies to validate the models. The results of the analyses have been compared with those mentioned in the chapter 15 of the Final Safety Analysis Report.

• 한국압력기기공학회 논문집
• /
• 제6권1호
• /
• pp.43-49
• /
• 2010

The piping failure probability of the reactor coolant system in Kori unit 1 was evaluated considering stress corrosion cracking. The P-PIE program (Probabilistic Piping Integrity Evaluation Program) developed in this study was used in the analysis. The effect of some variables such as oxygen concentration during start up and steady state operation, and operating temperature, which are related with stress corrosion cracking, on the piping failure probabilities was investigated. The effects of leak detection capability, the size of big leak, piping loops, and reactor types on the piping failure probability were also investigated. The results show that (1) LOCA (loss of coolant accident) probability of Kori unit 1 is extremely low, (2) leak probability is sensitive to oxygen concentration during steady state operation and operating temperature, while not sensitive to the oxygen concentration during start up, and (3) the piping thickness and operating temperature play important roles in the leak probabilities of the cold leg in 4 reactor types having same inner diameter.

• 한국압력기기공학회 논문집
• /
• 제4권2호
• /
• pp.13-19
• /
• 2008

In order to predict a remaining life of a plant, it is necessary to select the components that are critical to the plant life. The remaining life of those components shall be evaluated by considering the aging effect of materials used as well as numerous factors. However, when evaluating reliability of nuclear structural components, some problems are quite formidable because of lack of information such as operating history, material property change and uncertainty in damage models. Accordingly, if structural integrity and safety are evaluated by the deterministic fracture mechanics approach, it is expected that the results obtained are too conservative to perform a rational evaluation of plant life. The probabilistic fracture mechanics approaches are regarded as appropriate methods to rationally evaluate the plant life since they can consider various uncertainties such as sizes and shapes of cracks and degradation of material strength due to the aging effects. The objective of this study is to evaluate the structural integrity for a reactor pressure vessel under the small break loss of coolant accident by applying the deterministic and probabilistic fracture mechanics. The deterministic fracture mechanics analysis was performed using the three dimensional finite element model. The probabilistic integrity analysis was based on the Monte Carlo simulation. The selected random variables are the neutron fluence on the vessel inside surface, the content of copper, nickel, and phosphorus in the reactor pressure vessel material, and initial RTNDT.

• Nuclear Engineering and Technology
• /
• 제37권4호
• /
• pp.363-374
• /
• 2005

A heterogeneous thorium-based Kyung Hee Thorium Fuel (KTF) assembly design was assessed for application in the APR-1400 to study the feasibility of using thorium fuel in a conventional pressurized water reactor (PWR). Thermal hydraulic safety was examined for the thorium-based APR-1400 core, focusing on the Departure from Nucleate Boiling Ratio (DNBR) and Large Break Loss of Coolant Accident (LBLOCA) analysis. To satisfy the minimum DNBR (MDNBR) safety limit condition, MDNBR>1.3, a new grid design was adopted, that enabled grids in the seed and blanket assemblies to have different loss coefficients to the coolant flow. The fuel radius of the blanket was enlarged to increase the mass flow rate in the seed channel. Under transient conditions, the MDNBR values for the Beginning of Cycle (BOC), Middle of Cycle (MOC), and End of Cycle (EOC) were 1.367, 1.465, and 1.554, respectively, despite the high power tilt across the seed and blanket. Anticipated transient for the DNBR analysis were simulated at conditions of $112\%$ over-power, $95\%$ flow rate, and $2^{\circ}C$ higher inlet temperature. The maximum peak cladding temperature (PCT) was 1,173K for the severe accident condition of the LBLOCA, while the limit condition was 1,477K. The proliferation resistance potential of the thorium-based core was found to be much higher than that of the conventional $UO_2$ fuel core, $25\%$ larger in Bare Critical Mass (BCM), $60\%$ larger in Spontaneous Neutron Source (SNS), and $155\%$ larger in Thermal Generation (TG) rate; however, the radio-toxicity of the spent fuel was higher than that of $UO_2$ fuel, making it more environmentally unfriendly due to its high burnup rate.