Influence of Gap Size on Added Mass for Spent Fuel Storage Rack

Free standing spent fuel storage racks are submerged in water contained with spent fuel pool. During a postulated earthquake, the water surrounding the racks is accelerated and the so-called fluid-structure interaction (FSI) is significantly induced between water, racks and the pool walls[1]. The added mass is an important input parameter for the dynamic structural analysis of the spent fuel storage rack under earthquake[2]. The spent fuel storage rack is different even for the same vendors. Some rack are designed as the honeycomb construction, others are designed as the end-tube-connection construction. Therefore, the added mass for those racks have to be measured for the new rack’s design. More importantly, the added mass is influenced by the layout of the rack in the spent fuel pool. In this paper, an experiment is carried out to measure the added mass by free vibration test. The measured fluid force of the rack is analyzed by Fourier analysis to derive its vibration frequency. The added mass is then evaluated by the vibration frequency in the air and water. Moreover, a two dimensional CFD model of the spent fuel rack immersed in the water tank is built. The fluid force is obtained by a transient analysis with the help of dynamics mesh method.

System Safety Analysis Method Based on Real-Time Online Risk Monitoring Technology

In the continuous operation process of Nuclear Power Plant (NPP), its configuration is full of variety over time because of the system’s dynamic characteristics. There is a great need to update the risk/safety analysis models when it becomes necessary to reflect those dynamic characteristics of the system/component. Most of the current methods for risk/safety analysis belong to the scope of safety pre-analyzing, which analyzes the system risk/safety before system being in service. The main purpose of these safety pre-analyzing is to guide system design and optimization, but the real-time operational risk/safety analysis of NPPs is considered little. In order to know well the real-time risk/safety for system, a System Safety Analysis Method based on Real-time Online Risk Monitoring Technology is proposed. The safety risk model is established based on the modular fault tree that is used to represent logic structure of system. The real-time risk/safety is monitored according to the correspondence monitoring signal or data of component/system. Simultaneously the method can account for the change of risks based on the established mapping relationship between the state transition rules and corresponding risk/safety model updating rules. Finally, a case monitoring the safety for the system of two redundant pumps was used to demonstrate the effectiveness of the method.

Application of Level 2 PSA in the Design of Cavity Injection System for Nuclear Power Plant

After Fukushima nuclear power plant accident, severe accident is getting more and more concerns all over the world. In order to mitigate severe accident and improve the safety of nuclear power plant, two different strategies are applied in different plants. One is in-vessel melt retention strategy, and the other is ex-vessel melt retention strategy. Tianwan nuclear power plant is an improved Gen II nuclear power plant and in-vessel melt retention strategy is adopted in the plant. In order to achieve this strategy, cavity injection system is designed for the plant. Probabilistic Safety Analysis is the most commonly used quantitative risk assessment tool for decision-making in selecting the optimal design among alternative options. For this plant, in order to optimize the design of cavity injection system, improve the safety level of nuclear power plant, and meanwhile, improve the engineering implementation and economization, Level 2 PSA was used for this decision-making process. In this paper, the Level 2 PSA for this plant and the application for the design of cavity injection system are introduced.

The Necessity of Independent Configuration of the DC Power Supply System Used for Power Supply and Auxiliary Power Supply in Nuclear Power Plant

At present, in the typical design of China’s nuclear power plant, main generation system is connected to the power grid by 500 kV system. 500 kV system as a priority power source, 220kV system as an auxiliary power source. Independent operation of 500kV and 220kV system, improved the reliability of power supply of nuclear power plant. However, the DC 220V power system used to control the 500kV and 220kV system in the switch station of partial nuclear power project is not independently configured, and the design form of one set of DC system is used in the transformer station. In recent years, there are many accidents that AC power enters into the DC power system, resulted in the loss of power source in the transformer station. The loss of external power source in the whole plant is very significant. In this paper, the influence of AC power entering into DC power system on relay protection device is analyzed, the measures to prevent the AC power into DC power system are discussed, the necessity of independent configuration of DC control power system for the 500kV priority power system and 220kV auxiliary power supply system is analyzed.

Experimental Research on the Fluid Induced Forces of Clearance Flow in Canned Motor Reactor Coolant Pump

Reactor coolant pump is one of the key equipment of the coolant loop in a pressurized water reactor system. Its safety relies on the characteristics of the rotordynamic system. For a canned motor reactor coolant pump, the liquid coolant fills up the clearance between the metal shields of the rotor and stator inside the canned motor, forming a clearance flow. The fluid induced forces of the clearance flow in canned motor reactor coolant pump and their effects on the rotordynamic characteristics of the pump are experimentally analyzed in this work. A vertical experiment rig has been established for the purpose of measuring the fluid induced forces of the clearance. Fluid induced forces of clearance flow with various whirl frequencies and various boundary conditions are obtained through the experiment. Results show that clearance flow brings large mass coefficient into the rotordynamic system and the direct stiffness coefficient is negative under the normal operating condition. The rotordynamic stability of canned motor reactor coolant pump does not deteriorate despite the existence of significant cross-coupled stiffness coefficient from the fluid induced forces of the clearance flow.

Residual Stress Measurement of Sealing Glass Based on Optical Fiber Sensing Technology

Metal-to-glass electrical penetration assemblies (EPA) are highly sophisticated equipment and have been used for electrical connection in containment structures or pressure vessels in nuclear plants because of their high temperature resistance and good hermeticity. One important factor to keep hermeticity and reliability can be attributed to the initial residual stress in sealing glass of metal-to-glass EPA. If the residual stress is too high, small defects easily take place in the sealing materials. An insufficient prestress also cannot meet the requirement of high pressure application. To study the influence of residual stress on hermeticity, we developed a novel method of residual stress measurement in metal-glass sealing based on an embedded optical fiber sensor. The fiber Bragg grating (FBG) sensor was embedded in the glass material during the EPA manufacturing, and the residual stress along the grating could be retrieved via optical fiber sensing technique. Basing on our existing metal-glass sealing technique, the initial residual stress could be modulated by changing the sealing process, then the change of different initial residual stress was measured by the embedded FBG, through which the impact of residual stress on metal-glass sealing hermeticity could be finally revealed. A finite element model was established basing on linear elastic theory, then the localized stress along the FBG and the global stress distribution had been investigated theoretically. Taking the stress measuring by FBG as a breakthrough point, the effect of initial residual stress on sealing hermeticity was studied experimentally. The results showed that the residual stress should be an important assessment indicator to metal-to-glass sealing. This research also provided a new approach to optimize EPA manufacture.

Design and Development of DeRisk: A Fault Tree Analysis Program Package

Fault tree analysis (FTA) has been proven to be a very important tool and has been successfully applied to safety/reliability studies in nuclear, chemical, military, space industries/systems. Hitherto, several useful and popular FTA software/program packages have been developed, like CAFTA+, FAUNET, RiskSpectrum, SAPHIRE, RiskA etc. Minimum Cut Set (MCS) method is the most commonly used traditional FTA method. However, it suffers from low efficiency when solving remarkably large fault trees (FTs). To overcome the shortcomings of the traditional method, several new techniques are proposed such as Binary Decision Diagram (BDD), Zero-suppressed Binary Decision Diagram, (ZBDD) Petri Net (PN), Bayesian Network (BN) and Dynamic Uncertain Causality Graph (DUCG). DUCG is a newly presented Probabilistic Graphic Model to deal with systems with dynamics, uncertainties and logic cycles. DUCG is a good choice to analyze large FTs, in our previous papers, we have proved that any FT can be mapped into a DUCG graph and additional modeling and analytical power can be achieved. DeRisk is a DUCG embedded risk analysis program package written in C# for FTA and is designed as a powerful tool to assist reliability engineers. In this paper, the design schema and the main algorithms of DeRisk are introduced. DeRisk contains five parts: (1) A Graphical User Interface (GUI) Module which interacts with users; (2) A Preprocessing Module which preprocesses FTs (3) An Input Module which allows user to input necessary data by file or by command line; (4) A Calculation Module which offers qualitative/quantitative analysis; (5) An Output Module which outputs the results required by users. Some illustrative examples are used to verify the correctness and effectiveness of DeRisk.

Evaluation of Local Damage to Reinforced Concrete Panels Subjected to Oblique Impact of Rigid and Soft Missiles

Many empirical formulas have been proposed for evaluating local damage to reinforced concrete (RC) structures caused by impacts of rigid missiles. Most of these formulas have been derived based on impact tests normal to the target structures. Up to now, few impact tests oblique to the target structures have been carried out. This study has been conducted with the purpose of proposing a new formula for evaluating the local damage caused by oblique impacts based on previous experimental and simulation results. In this paper, the results of simulation analyses for evaluating the local damage to a RC panel subjected to normal and oblique impacts by rigid and soft missiles, by using the simulation method that was validated using the results of previous impact experiments. Based on the results of these simulation analyses, the effects of the rigidity of the missile as well as the impact angle on the local damage to the target structures are clarified.

Dynamic Fault Tree Analysis Based on Dynamic Uncertain Causality Graph

Dynamic Fault Tree (DFT) has drawn attention from comprehensive industrial systems in recent years. Many analytical approaches are developed to analyze DFT, such as Markov Chain based method, Inclusion-Exclusion Rule based method, and Sum-of-Disjoint-Product theory based method. Novel methods such as Bayesian Network and Petri Net are also used to solve DFT. However, Basic events are usually assumed unrepairable and are restricted to specific probabilistic distributions. And some methods may suffer from combination explosion. This paper applies Dynamic Uncertain Causality Graph (DUCG) to analyze DFT to overcome the aforementioned issues. DUCG is a newly proposed Probabilistic Graphic Model for large complex industrial systems which allows for dynamics, uncertainties and logic cycles. The DUCG based methodology can be summarized as event mapping, logical mapping, and numerical mapping. This paper studies how to map the PAND, FDEP, SEQ AND SPARE sequential logic gates into equivalent representations in DUCG. With the DUCG representation mode, one can analyze DFT with algorithms in DUCG. Future work will be done on benchmark tests and on software development.

Study on PRA Procedure Considering Combination of Multiple Events Using DQFM Methodology

Fukushima Daiichi Nuclear Power Station accident occurred by the Great East Japan Earthquake on March 11, 2011. After that, continuous enhancement of nuclear safety is being required in Japan. The accident of Fukushima was caused by the seismic induced tsunami event, namely, multiple events. The other examples of multiple events due to the seismic event are such as internal fire and internal flooding in the nuclear power plants. In addition, structures, such as a building, and piping might be damaged by the seismic event, which could impact component failure dependently. In order to consider these kinds of events, the development of PRA procedures for multiple events caused by the seismic events will be highly demanded. So, we developed a basic PRA methodology for seismic induced tsunami events using “Direct Quantification of Fault Tree using Monte Carlo simulation (DQFM) methodology”. And we verified its applicability through the evaluation.