Minjun Peng

Small-Break Loss of Coolant Accident Analysis of the Integrated Pressurized Water Reactor

The pressurizer vapour space small-break loss of coolant accident is analyzed based on the nuclear thermal hydraulic transient analysis software RELAP5/SCDAP/MOD3.4 code. Transient characteristics of the Integrated Pressurized Water Reactor (IPWR) system pressure, core temperature, mass flow rate, reactor vessel water level are achieved. The influences of three different break areas are also studied and the conclusion is consistent with local experiment conclusion. The results show that the safety system of the IPWR can make sure the reactor core is under the water after the accident happens and the passive residual heat removal system can remove the core heat.

A Review on Specific Features of Small and Medium Sized Nuclear Power Plants

Due to global warming and high oil prices nuclear power is the most feasible solution for generating electricity. For the fledging nuclear power industry small and medium sized nuclear reactors (SMR’s) are instrumental for the development and demonstration of nuclear reactor technology. Due to the enhanced and outstanding safety features, these reactors have been considered globally. In this paper, first we have summarized the reactor design by considering some of the large nuclear reactor including advanced and theoretical nuclear reactor. Secondly, comparison between large nuclear reactors and SMR’s have been discussed under the criteria led by International Atomic Energy Agency (IAEA). Thirdly, a brief review about the design and safety aspects of some of SMR’s have been carried out. We have considered the specifications and parametric analysis of the reactors like: ABV which is the floating type integral Pressurized water reactor; Long life, Safe, Simple Small Portable Proliferation Resistance Reactor (LSPR) concept; Multi-Application Small Light Water Reactor (MASLWR) concept; Fixed Bed Nuclear Reactor (FBNR); Marine Reactor (MR-X) & Deep Sea Reactor (DR-X); Space Reactor (SP-100); Passive Safe Small Reactor for Distributed energy supply system (PSRD); System integrated Modular Advanced Reactor (SMART); Super, Safe, Small and Simple Reactor (4S); International Reactor Innovative and Secure (IRIS); Nu-Scale Reactor; Next generation nuclear power plant (NGNP); Small, Secure Transportable Autonomous Reactor (SSTAR); Power Reactor Inherently Safe Module (PRISM) and Hyperion Reactor concept. Finally we have point out some challenges that must be resolved in order to play an effective role in Nuclear industry.Copyright

Concept Design of the Multi-Application Integrated Light Water Reactor and Normal Operation Analysis

According to the currently developing status of the integrated light water reactor, a multi-application integrated light water reactor concept design is proposed in this paper. The arc-shaped reactor core fuel elements are adopted, and once- through steam generator is used to produce steam. The pressurizer is located outside the reactor vessel which uses electric heating method. There are pumps in the reactor vessel used to drive the coolant flow to remove the nuclear heat. This paper also designs the startup bypass system of the once through steam generator and passive residual heat removal system. And the system code RELAP5/SCDAPSIM is used to simulate the process during the power change. The normal operation process of the reactor coolant system is introduced in detail.

Design and Development of a Reliability Analysis Tool Based on Multilevel Flow Models

This paper proposes a universal graphical tool for the modeling and reliability analysis of complex industrial process system based on Multilevel Flow Models (MFM). An Extensible Markup Language (XML) is used for structuring the MFM model. An editor is developed and an executor can implement reliability analysis in terms of the established MFM models. The proposed reliability analysis tool is meaningful for further research of the MFM based system reliability analysis and will be useful for more practical applications such as online risk monitoring by integrating the existed algorithms of alarm analysis and fault diagnosis based on MFM.Copyright

Evaluation of a SGTR Accident in the Multi-Application Integrated Pressurized Water Reactor

One of more frequent events in the Pressurized Water Reactor (PWR) is Steam Generator Tube Rupture (SGTR) accident, which is among the main accidents in the field of nuclear safety. This paper studies the SGTR event in the Multi-application Integrated Pressurized Water Reactor (IPWR) using the best-estimate thermal-hydraulic code RELAP5/MOD3.4. In the reactor of IPWR, several Once-Through Steam Generator (OTSG) cassettes are used and located between the core support and the pressure vessel. The tube rupture location is on the top of the tube sheet of a steam generator. Three different tube rupture modeling methods and several different subcooled discharge coefficients in the critical flow model are considered and compared. In the safety analysis, high pressure safety injection system, core makeup system and Passive Residual Heat Removal System (PRHRS) that would affect the accident consequences are considered.Copyright

Strategy Evaluation for Cavity Flooding during an ESBO Initiated Severe Accident

Intentional depressurization and cavity flooding are two important measures in current severe accident management guidelines (SAMGs). An extreme scenario of an extended station blackout (ESBO), when electric power is unavailable for more than 24 hours, actually occurred in the Fukushima Daiichi accident and attracted lots of attention. In an ESBO, the containment spray cannot be activated for condensation, and, thus, cavity flooding will generate a large amount of steam, which, ironically, overpressurizes the containment to failure before the reactor vessel is melted through. Therefore, consideration of these conflicting issues and the ways in which plants operate is crucial for strengthening the strategies outlined in SAMGs. In this paper, the effects of intentional depressurization and cavity flooding in an ESBO for a representative 900 MW second-generation pressurized water reactor (PWR) are simulated with MAAP4 code. Diverse scenarios with different starting times of depressurization and water injection are also compared to summarize the positive and negative impacts for accident mitigation. The phenomena associated with creep ruptures, hydrogen combustion, corium stratification, and cavity boiling are also analyzed in detail to strengthen our understanding of severe accident mechanisms. The results point out the facility limitations of second-generation PWRs which can improve existing SAMGs.

Condition Monitoring of Sensors in a NPP Using Optimized PCA

An optimized principal component analysis (PCA) framework is proposed to implement condition monitoring for sensors in a nuclear power plant (NPP) in this paper. Compared with the common PCA method in previous research, the PCA method in this paper is optimized at different modeling procedures, including data preprocessing stage, modeling parameter selection stage, and fault detection and isolation stage. Then, the model’s performance is greatly improved through these optimizations. Finally, sensor measurements from a real NPP are used to train the optimized PCA model in order to guarantee the credibility and reliability of the simulation results. Meanwhile, artificial faults are sequentially imposed to sensor measurements to estimate the fault detection and isolation ability of the proposed PCA model. Simulation results show that the optimized PCA model is capable of detecting and isolating the sensors regardless of whether they exhibit major or small failures. Meanwhile, the quantitative evaluation results also indicate that better performance can be obtained in the optimized PCA method compared with the common PCA method.

Thermal hydraulic analysis of small nuclear reactor core by using comparative approach of THEATRe and Relap5 code

Thermal analysis of a nuclear reactor is very important in determining its design. It can produce desired thermal power without exceeding limitations on core components, which could lead to fuel failure and radioactive release into the environment. In this paper, thermal hydraulic studies of a small nuclear reactor core have been carried out by using two well-known thermal hydraulic codes, Relap5 and THEATRe. The Relap5 code can deal with plate-type fuel elements but the THEATRe code works only for pin-type fuel elements. For this purpose, the THEATRe code has been modified for the plate-type fuel element and then used in the current research. The results obtained from both codes agree well, confirming the accuracy of simulation.

Numerical Simulation Research of Natural Convection Heat Exchanger

In this paper, computer software is employed to simulate the temperature-field and flow-field of AP1000 PRHR HX, and investigate its heat-transferring and flow characteristic. Through comparative analysis of the distribution of temperature-field and flow-field in different locations at the same time, and the change of temperature-field and flow-field in the same location at different times, heat-transferring process and natural convection situation of PRHR HX are deeply understand. It contributes to analyze the natural circulation capacity of PRHR HX, and provides some references for the effective operation of passive residual heat removal system.

A Review: Activities in the Field of Small and Medium Sized Nuclear Power Plants

Due to increasing demand of energy worldwide there is extensive need of nuclear industry capable of operating under different conditions like sea water desalination processes, floating conditions, natural circulation phenomenon and space applications. A review about the salient features of small and medium sized nuclear reactors (SMRs) have been carried out with all operational and safety features depicting that how SMRs can play a significant role in energy crises under International Atomic Energy Agency (IAEA) criteria and is extremely reliable than large nuclear power plants. A brief introduction of some of SMRs have been included in this paper.