Loss Of Feedwater Event Research Articles

Prediction of the Loss of Feed Water Fault Signatures Using Machine Learning Techniques

Fault diagnosis occurrence and its precise prediction in nuclear power plants are extremely important in avoiding disastrous consequences. The inherent limitations of the current fault diagnosis methods make machine learning techniques and their hybrid methodologies possible solutions to remedy this challenge. This study sought to develop, examine, compare, and contrast three robust machine learning methodologies of adaptive neurofuzzy inference system, long short-term memory, and radial basis function network by modeling the loss of feed water event using RELAP5. The performance indices of residual plots, mean absolute percentage error, root mean squared error, and coefficient of determination were used to determine the most suitable algorithms for accurately diagnosing the loss of feed water transient signatures. The study found out that the adaptive neurofuzzy inference system model outperformed the other schemes when predicting the temperature of the steam generator tubes, the radial basis function network scheme was best suited in forecasting the mass flow rate at the core inlet, while the long short-term memory algorithm was best suited for the estimation of the severities of the loss of the feed water fault.

Pilot application of risk informed safety margin characterization to a total loss of feedwater event

In this paper we present the results of application of a risk-informed safety margin characterization (RISMC) approach to the analysis of a loss of feedwater (LOFW) event at a pressurized water reactor (PWR). This application considered a LOFW event with the failure of auxiliary feedwater (AFW) for which feed and bleed cooling would be required to prevent core damage. For this analysis the main parameters which impact core damage for the scenario were identified and probability distributions were constructed to represent the uncertainties associated with the parameter values. These distributions were sampled using a Latin Hypercube Sampling (LHS) technique to generate sets of sample cases to simulate using the MAAP4 code. Simulation results were evaluated to determine the safety margins relative to those obtained using typical probabilistic risk assessment (PRA) modeling (success criteria) assumptions.

원자력 발전소 주제어실 인터페이스 설계를 위한 인적오류 분석 기법의 보완

This study describes a modification of the technique for human error analysis in nuclear power plants (NPPs) which adopts advanced Man-Machine Interface (MMI) features based on computerized working environment, such as LCOs. Flat Panels. Large Wall Board, and computerized procedures. Firstly, the state of the art on human error analysis methods and efforts were briefly reviewed. Human error analysis method applied to NPP design has been THERP and ASEP mainly utilizing Swain`s HRA handbook, which has not been facilitated enough to put the varied characteristics of MMI into HRA process. The basic concepts on human errors and the system safety approach were revisited, and adopted the process of FMEA with the new definition of Error Segment (ESJ. A modified human error analysis process was suggested. Then, the suggested method was applied to the failure of manual pump actuation through LCD touch screen in loss of feed water event in order to verify the applicability of the proposed method in practices. The example showed that the method become more facilitated to consider the concerns of the introduction of advanced MMI devices, and to integrate human error analysis process not only into HRA/PRA but also into the MMI and interface design. Finally, the possible extensions and further efforts required to obtain the applicability of the suggested method were discussed.

Safety Evaluation of Primary-Side Bleed-and-Feed Scenarios by an Experimental Loss of Feedwater Event at the IIST Facility

The current emergency operating procedures (EOPs) for a three-loop pressurized water reactor on a total loss-of-feedwater (LOFW) incident are experimentally evaluated at the Institute of Nuclear Energy Research Integral System Test (IIST) facility. Four experiments were conducted with a different number of pressurizer power-operated relief valves (PORVs) and centrifugal charging pumps (CCPs) engaged in performing the primary-side bleed-and-feed operation. In addition, the effect of opening a pressure vessel venting-path was also evaluated. The focus was on the effectiveness and adequacy of current EOPs and the observation of associated thermal-hydraulic phenomena. The test results indicated that the current EOPs can function well in mitigating the consequences of a LOFW event provided that at least two fully opened PORVs and one actuated CCP are used. The opening of a pressure vessel venting path during the bleed-and-feed operation is helpful in the primary system depressurization but has little effect to increase the pressure vessel coolant inventory. However, by opening one PORV, precaution for the reactor coolant system (RCS) integrity should be taken for system overpressurization caused by insufficient heat removal capability. Furthermore, an effective control of restored feedwater to a low injection rate is favorable to minimize the thermal stress in the RCS pressure boundary.

The development of passive design features for the Korean Next Generation Reactor

Four passive design features, such as a fluidic device, a passive secondary condensing system (PSCS), a passive cavity flooding system, and a passive hydrogen ignitor, are under development as part of the Korean Next Generation Reactor (KNGR), an advanced PWR, to increase its safety. The fluidic device, which is located at the discharge of the safety injection tank, is a system to inject the borated water into the reactor coolant system in a passively regulating way to elongate the allowable start-up time of emergency diesel generators and to enhance performance against the loss of coolant accidents. The PSCS, which supplements the auxiliary feedwater system, secures the heat removal through steam generators in the case of the loss of feedwater event. The PSCS takes inlet flow from the steam line and returns condensate into the feedwater line after condensation through condenser tubes. Fusible plugs have been adopted for passive cavity flooding. If the ambient temperature is high enough, the plugs between the in-containment refueling water storage tank (IRWST) and the reactor cavity melt so that the IRWST water starts to flow into the reactor cavity by gravity. Passive hydrogen ignitors (i.e. catalytic ignitors) have been adopted in addition to the active hydrogen ignitors to maintain containment hydrogen concentrations below a detectable limit of 10 volume percent. Applicability of these passive features on KNGR has been studied following the three-step approach of (1) preliminary analyses using computer codes and small-scale experimental facilities, (2) detailed analyses through large-scale tests, code running, and uncertainty validation, and (3) quality assurance by design verification and analyses on interfaces with other systems.