A Sensor Fault-Tolerant Accident Diagnosis System

Emergency situations in nuclear power plants are accompanied by an automatic reactor shutdown, which gives a big task burden to the plant operators under highly stressful conditions. Diagnosis of the occurred accident is an essential sequence for optimum mitigations; however, it is also a critical source of error because the results of accident identification determine the task flow connected to all subsequent tasks. To support accident identification in nuclear power plants, recurrent neural network (RNN)-based approaches have recently shown outstanding performances. Despite the achievements though, the robustness of RNN models is not promising because wrong inputs have been shown to degrade the performance of RNNs to a greater extent than other methods in some applications. In this research, an accident diagnosis system that is tolerant to sensor faults is developed based on an existing RNN model and tested with anticipated sensor errors. To find the optimum strategy to mitigate sensor error, Missforest, selected from among various imputation methods, and gated recurrent unit with decay (GRUD), developed for multivariate time series imputation based on the RNN model, are compared to examine the extent that they recover the diagnosis accuracies within a given threshold.

Study on Measures During Loss of Normal Feedwater Accident for AP1000 NPP in Lower Power Operation

If an accident occurred in a plant in lower power operation, a series of problems could be caused by low residual heat and excessive removal capability of heat sinks. Taking a real accident as example, this paper discussed how to deal with loss of normal feedwater in AP1000 NPP during lower power operation. The level of Steam Generators (SG) will be rapidly reduced in case of Main Feedwater loss, and Startup Feedwater pump will start. If the reactor is not shut down immediately, the Startup Feedwater will not be enough to prevent SG level from falling, and automatic reactor trip will be inevitable. Therefore, in event of loss of Main Feedwater which cannot be recovered instantly, the reactor should be shut down manually to maintain SG level as high as possible. Appropriate measures should be taken to avoid excessive cooling after trip. Attention should be paid to SG level and feedwater flow to avoid Passive Residual Heat Removal Heat Exchanger (PRHR HX) actuation during regulating Startup Feedwater. Otherwise, PRHR HX may be actuated unexpectedly. With Reactor Coolant Pumps operating, PRHR HX heat removal capacity is strong. The Cold Leg temperature will drop rapidly because of PRHR HX and other heat sinks. In this case, the most effective measure to stabilize temperature of Cold legs is to isolate Streamline. Otherwise, structure integrity of the Reactor Pressure Vessel (RPV) will be challenged under Pressurized Thermal Shock (PTS) due to rapid drop of Cold leg temperature. In order to alleviate PTS, depressurization shall be performed and further cooling shall be suspended. Since PRHR HX is only connected to Loop 1, coolant in Hot Leg of Loop 2 is still kept at high temperature. Therefore, necessary measures shall be taken to prevent coolant flash in Loop 2 during the process of depressurization, which may fill the Pressurizer up with water. The RPV Head Vent valves perform safety-related function by preventing Pressurizer overfill in certain design basis events in AP1000 NPP. However, emergency letdown by opening the valves cannot effectively reduce Pressurizer level when coolant flashed in Loop 2. Because of mitigating PTS, opening RPV Head Vent valves, and steam condensation in Loop 2, the subcooling of the coolant at the core outlet is likely to drop below 6 °C or even 0 °C.

Application of 3D Coupled Code SPARKLE-2 to PWR Anticipated Transient Without Scram

An anticipated transient without scram (ATWS) is an anticipated operational occurrence (AOO) followed by failure of the automatic reactor trip function of the reactor protection system. The failure of the reactor to shut down during the certain AOOs can lead to increase in reactor coolant system (RCS) pressure and decrease in departure from nucleate boiling ratio (DNBR) margin for a pressurized water reactor (PWR). Japanese standard PWRs are equipped with ATWS mitigation system which consists of a diverse mitigation system which is independent from the reactor trip system. The ATWS mitigation system automatically initiates isolation of the main steam line flow and the auxiliary feed water system under condition indicative of an ATWS. Mitsubishi Heavy Industries, Ltd. (MHI) applies 3D coupled code, SPARKLE-2 [1] [2], to the ATWS evaluation. SPARKLE-2 is a 3D coupled code developed by MHI and consist of the PWR system transient analysis code M-RELAP5, the 3D neutron kinetics code COSMO-K [3] and the 3D core thermal-hydraulics code MIDAC [4]. SPARKLE-2 implements the 3D characteristics such as local moderator feedback and change in 3D power distribution during transient. Thanks to gain from the 3D calculation, the analysis results show that the plant transients are effectively mitigated by the ATWS mitigation system and the RCS pressure and the minimum DNBR meet the safety criteria. These results also show that operational margins are increased, which enables more flexible design of the reload core.

The Analysis and Design Technology Research of DAS

Now the entire safety I&C system is based on one kind of software and hardware platform, the Common Cause Failure (CCF) may impact the whole safety I&C system becomes to a potential risk. How to mitigate the effect of CCF in safety system and improve the safety of the nuclear power plant is considered by the system designer. Especially after the Fukushima nuclear accident, the Defense-in-Depth and Diversity (D3) should be more concerned by all designers. The diverse actuation system (DAS) plays a very important role in the D3 system. In this paper, the related codes and standards of DAS are analyzed firstly. Then, this paper expounds the approach to demonstrate the D3 analysis for the digital I&C systems applied to the nuclear power plant in detail. In the D3 analysis, all the safety functions of the digital safety system are assumed to be disabled by a CCF. DAS provides diverse automatic reactor trip and diverse safety injection actuation which are not impaired by the postulated CCF. DAS also provides manual actuation functions and plant parameter monitoring functions which can be used to cope with CCF. Finally, the paper takes the DAS of Fujian Fuqing Nuclear Power Plant as an example, introduce how to design the structure of the DAS and calculate the suitable setpoints.

Digital Systems for Reactor Control: Design, Experience of Operation

Conversion of traditional analog NPP I&C systems to digital systems is a common tendency for many countries. Digital systems for reactor control designed by «Radiy» Company (Kirovograd, Ukraine) are described below. FPGA (Field Programmable Gates Arrays) were used for implementation of control algorithms. An equivalence between FPGA-projects implementation and schemes of control technological algorithms permitted to simplify development and verification processes and decrease the number of development errors. The platform was used for implementation of different safety important systems: reactor protection systems, automatic reactor power control and limitation systems, rods control systems, control safety systems. The main peculiarity of the reactor protection system is different types of diversity (apparatus and program diversity due to different hardware and different languages in main and diverse divisions; functional diversity; difference of CASE-tools). These systems have been used at 10 units. Reliability measures of systems and their components were determined using operational statistical data. Possible uses for these systems: modernization of different types of existing reactors (not only WWER); as full system or as subsystems; not only for Ukraine, but for other countries; for reactors III+ and IV generations.