AP1000 Nuclear Power Plant Research Articles

EFFECTS OF FORGING AND HEAT TREATMENTS ON STRESS CORROSION BEHAVIOR OF 316LN STAINLESS STEEL IN HIGH TEMPERATURE CAUSTIC SOLUTION

The reactor coolant piping in the third generation nuclear power plants of AP1000 is manufactured by integrally forging. Therefore, it is of vital importance to investigate the effects of forging and heat treatments on the stress corrosion cracking(SCC) resistance of 316 LN stainless steel(316LNSS), which is the candidate material for the reactor coolant piping in AP1000 nuclear power plants. In this work, electron back scattering diffraction(EBSD) and microhardness measurements(HV) were used to characterize the microstructure and residual strain of the as-received 316 LNSS, the forged and solution anneal treated 316 LNSS and the forged and stress relief treated 316 LNSS, respectively. The average grain size of the as- received 316 LNSS was the largest, and the forged 316 LNSS followed by solution anneal treatment and stress relief treatment showed no obvious differences on grain size. The as-received 316 LNSS exhibited the highest residual strain followed by the forged and stress relief treated 316 LNSS and then solution anneal treated 316 LNSS. Besides, the residual strain in the as- received316 LNSS concentrated on grain boundaries, while the residual strain in the forged and stress relief treated316 LNSS was characterized by a band-like distribution. The U-bend specimens were utilized to investigate the SCC behavior of the 3 kinds of 316 LNSS specimens in high temperature caustic solution. After SCC experiments, the crack morphologies of the 3 kinds of 316 LNSS specimens were examined by SEM. Then the macro and micro fracture morphologies were examined by OM and SEM, respectively. Grain morphology, residual strain and grain boundary character distribution near the SCC crack tip of the forged and stress relief treated 316 LNSS were investigated using EBSD. The results showed that the forged and solution anneal treated 316 LNSS exhibited the lowest SCC sensibility, while the as-received the highest, with the most cracks and the highest growth rate. The as-received and the forged and solution anneal treated 316 LNSS showed obvious intergranular cracking, while the forged and stress relief treated 316 LNSS showed a mixed cracking mode. The larger average grain size and higher residual strain, especially concentrating on the grain boundaries, were considered to be responsible for the highest SCC sensibility of the as-received 316 LNSS. Compared with the forged and stress relief treated 316 LNSS, the higher content of coincidence site lattice boundary(CSLB) and lower residual strain contributed to the lower SCC sensibility of forged and solution anneal treated 316 LNSS. The stress relief treatment failed to eliminate the band-like microstructure effectively, which disadvantaged the SCC resistance.

Research of complete loss of flow accident for AP1000 based on RELAP5 code

Abstract. AP1000 reactor is the typical “third-generation” nuclear power plant in the world at present. The primary system of AP1000 nuclear power plant was modeled using RELAP5/MOD 3.3 code, and the transient thermal hydraulic characteristics were analyzed under complete loss of flow accident (CLOFA). The calculation results by RELAP5 code were compared with those of the LOFTRAN codes. The research results show that the RELAP5 model can accurately simulate the transient thermal hydraulic characteristics of AP1000 under the accident of complete loss of forced reactor coolant flow. Meanwhile, comparison was made between the results of complete loss of reactor coolant flow initiated by loss of voltage and frequency reduction of offsite power, and the analysis indicates that the results of complete loss of flow initiated by complete loss of voltage bound the complete loss of flow initiated by reduction in the reactor coolant pump motor supply frequency with a frequency decay of up to 5 hertz per second.

Calculations for AP1000 nuclear power plant SGTR accident based on RELAP5/mod 3.3 program

AP1000 nuclear power plant model is founded via RELAP5/mod 3.3 program in detail. Refer to SGTR accident basic assumptions in the safety analysis report, calculations for AP1000 nuclear power plant SGTR accident are carried out. And through comparison between the result got and the result from safety analysis report calculated by LOFTTR2 program, the two results are very consistent. It's found that even if the operators do not interfere, AP1000 nuclear power plant protection systems and passive design measures will trigger an automatic response measures, it can terminate the SG heat transfer tube leaking and make RCS be stable in a safe condition and to prevent SG overflowing and ADS actions occurring.

Lumped parameter method application in nuclear engineering – AP1000 DVI line break accident analysis by RELAP5

Based on the PWR best estimate program RELAP5/MOD3.3, the reactor coolant system and the passive core cooling system of AP1000 Nuclear Power Plant were modeled and analyzed. Some key transient parameters were obtained, including RCS pressure, break discharge, etc. The result is in close agreement with the result generated by NOTRUMP code from Westinghouse USA. The calculation results show that safeguard system of APl000 can effectively reduce the pressure of the first loop, also can prevent core uncovered. The safety of AP1000 during a small break Loss of Coolant Accident caused by DVI line breaking is verified.

Development of rapid atmospheric source term estimation system for AP1000 nuclear power plant

The development of rapid atmospheric source term estimation system (RASTES) for AP1000 nuclear power plant (NPP) is described. The system is designed to give rapid estimation of the amount of fission products released to the atmosphere, and to provide necessary input for fission products atmospheric dispersion simulation and public dose calculation, during a nuclear accident. The estimation method in RASTES mainly refers to NUREG-1228. However, some adjustments are made based on an investigation of studies in nuclear emergency response and design features of AP1000 NPP during development of RASTES. These adjustments enable the system to be more flexible and realistic in atmospheric source term estimation, and more applicable to AP1000 NPP. These adjustments include modifications and validations of aerosol removal factors in reactor containment in both natural process and spray process, as well as addition of aerosol removal factors in passive containment cooling process. Besides, three reactor core damage assessment methods are integrated into atmospheric source term estimation. In addition, containment radiation levels for specified fuel damage which are used in core damage assessment are recalculated. The calculation function of RASTES is confirmed through comparison with RASCAL. Atmospheric source terms in different fission product removal processes are also calculated and compared. The result shows that passive containment cooling is able to further reduce aerosol release by about 50% compared with natural process. Development of RASTES will enhance the capability of nuclear emergency response of AP1000 NPP.

Improvement of Radioactive Wastewater Treatment Technology in Inland AP1000 Nuclear Power Plant

In order to meet the standard of effluent radioactive wastewater from inland nuclear power plant, the improvement of radioactive wastewater treatment technology in inland AP1000 nuclear power plant is determined. Based on the origin radioactive wastewater treatment technology of AP1000 nuclear power plant and the advanced radioactive wastewater treatment technology, an Chemical precipitation + CsTreatRR+ Ion exchange technology is proposed and evaluated. The improved technology can provide reference for inland nuclear power plant radioactive wastewater treatment system design.

Application case study of AP1000 automatic depressurization system (ADS) for reliability evaluation by GO-FLOW methodology

AP1000 nuclear power plant (NPP) utilized passive means for the safety systems to ensure its safety in events of transient or severe accidents. One of the unique safety systems of AP1000 to be compared with conventional PWR is the “four stages Automatic Depressurization System (ADS)”, and ADS system originally works as an active safety system. In the present study, authors first discussed the reasons of why four stages ADS system is added in AP1000 plant to be compared with conventional PWR in the aspect of reliability. And then explained the full and partial depressurization of RCS system by four stages ADS in events of transient and loss of coolant accidents (LOCAs). Lastly, the application case study of four stages ADS system of AP1000 has been conducted in the aspect of reliability evaluation of ADS system under postulated conditions of full RCS depressurization during large break loss of a coolant accident (LBLOCA) in one of the RCS cold legs.In this case study, the reliability evaluation is made by GO-FLOW methodology to determinate the influence of ADS system in dynamic reliability of passive core cooling system (PXS) of AP1000, i.e. what will happen if ADS system fails or successfully actuate. The GO-FLOW is success-oriented reliability analysis tool and is capable to evaluating the systems reliability/unavailability alternatively to Fault Tree Analysis (FTA) and Event Tree Analysis (ETA) tools. Under these specific conditions of LBLOCA, the GO-FLOW calculated reliability results indicated that ADS is a key safety system of AP1000 NPP to reduce RCS pressure in controlled fashion and enables the all safety injections for automatic core cooling successfully after the accident. And dynamic reliability of the passive core cooling system of AP1000 plant is increased by the successful actuation of ADS system. Thus, by this application case study, it is pointed out, that high reliability of the passive safety system of AP1000 would depend on the full reliability of ADS system that it would work surely when needed.

Ensemble-based sensitivity analysis of a Best Estimate Thermal Hydraulics model: Application to a Passive Containment Cooling System of an AP1000 Nuclear Power Plant

The reliability analysis of a Nuclear Power Plant (NPP) passive safety system requires to quantify its functional failure probability, using a Best Estimate (BE) Thermal Hydraulic (TH) code to predict the behavior in normal and accidental conditions. To effectively do this, sensitivity analysis (SA) is performed to identify the relevant, important input variables whose variations affect the system functional response (calculated as output of the BE-TH code). In this paper, we propose a novel framework for the sensitivity analysis (SA) of a BE-TH code and apply it to the analysis of the Passive Containment Cooling System (PCCS) of an Advanced Pressurized reactor AP1000 during a Loss Of Coolant Accident (LOCA). We build Finite Mixture Models (FMMs) for approximating the probability density function (pdf) of the pressure reached by the PCCS during the LOCA, on the basis of a limited number of simulations. Then, an ensemble of three alternative SA methods is innovatively set up for identifying the input variables that most affect the TH code output: input saliency, Hellinger distance and Kullback–Leibler divergence. The capability of the ensemble is shown on a numerical case study.

One-dimensional model for containment in AP1000 nuclear power plant based on thermal stratification

Passive containment cooling system operating based on natural circulation is innovatively used in AP1000 nuclear power plant to improve safety. However, the overall heat and mass transfer will decrease and the heat transfer through the containment shell will be slow due to stable stratification which will occur if the forced convection mixing is not sufficiently strong enough to disrupt the stable fluid layers, so it is important for system design and accident analysis to evaluate whether the circulation can establish or not in the containment accurately and efficiently. Many researches indicate that the gradients of such parameters in horizontal direction are so small that can be ignored, in this paper, one-dimensional model is developed for AP1000 containment based on thermal stratification theory, natural circulations in and outside the containment are both included. Based on the results, the thermal stratification and circulation exist simultaneously, the pressure in the containment can be kept at the safe level, and positive pressure gradient in the vertical direction affords the force of circulation. The steam and air with higher temperature concentrating at the top will strengthen the heat and mass transfer because the cooling capacity of top head is higher than of other parts.

Simulation and dose analysis of a hypothetical accident in Sanmen nuclear power plant

In November 2013, an AP1000 nuclear power plant (NPP) will be put into commercial operation. An atmospheric dispersion of radionuclides during a severe hypothetical accident in Sanmen NPP, Zhejiang province, China, is simulated with a Lagrangian particle dispersion model FLEXPART. The accident assumes that a station blackout (SBO) accident occurred on August 25, 2011, 55% core was damaged and 49 radionuclides were released into the atmosphere. Our simulation indicates that, during this dispersion, the radioactive plume will cover the mainland China, Taiwan, Japan, North Korea, South Korea and Russia. The radiation dose levels in Japan, North Korea and Russia are the lightest, usually less than 1mSv. The influenced areas in these countries are 9901km2, 31,736km2 and 2,97,524km2, respectively; dose levels in Taiwan and South Korea are moderate, no more than 20mSv. Information about reducing dose should be given to the public. Total influenced areas in these two countries are 3621km2 and 42,370km2, which take up 100% of the land in Taiwan and 35% of the land in South Korea; the worst situation happens in mainland China. The total influenced area is 3×106km2 and 1,40,000km2 in this area has a dose level higher than 20mSv. Measurement must be taken to reduce the dose. More than 284 thousand residents will face the risk of developing cancer. Furthermore, 96% of this population is mainly concentrated in Zhejiang province, where Sanmen NPP locates.

Challenges of SMR Licensing Practices

This paper aims to increase the understanding of high level Nuclear Power Plant (NPP) licensing processes in Finland, France, the UK, Canada and the USA. These countries have been selected for this study because of their different licensing processes and recent actions in new NPP construction. After discussing their similarities and differences, suitable features for Small Modular Reactor licensing can be emphasized and suggested. Some of the studied licensing processes have elements that are already quite well suited for application to SMRs, but all of these different national processes can benefit from studying and implementing lessons learned from SMR specific licensing needs. The main SMR features to take into account in licensing are standardization of the design, modularity, mass production and serial construction. Modularity can be divided into two different categories: the first category is simply a single unit facility constructed of independently engineered modules (e.g., construction process for Westinghouse AP-1000 NPP) and the second is a facility structure composed of many reactor modules where modules are manufactured in factories and installed into the facility as needed (e.g., NuScale Power SMR design). Short construction schedules will not be fully benefitted from if the long licensing process prolongs the commissioning and approach to full-power operation. The focus area of this study is to better understand the possibility of SMR deployment in small nuclear countries, such as Finland, which currently has four operating NPPs. The licensing process needs to be simple and clear to make SMR deployment feasible from an economical point of view. This paper uses public information and interviews with experts to establish the overview of the different licensing processes and their main steps. A high-level comparison of the licensing steps has been carried out. Certain aspects of the aviation industry licensing process have also been studied and certain practices have been investigated as possibly suitable for use in nuclear licensing. All of the current licensing processes were found to be quite heavy and time-consuming and further streamlining could be possible without compromising safety or the need for public participation in the licensing process. Some examples of the modification possibilities for SMR applications are discussed. A profound discussion on SMR-specific licensing models, and on ways to simplify and harmonize them, will be needed in the near future in Europe too. This would be a natural continuation to the harmonization efforts underway for existing and new large reactors.

Safety Discussion and Precaution Measures about Laser Beam Welding of Pyrotechnic Actuator

Abstract Structure of pyrotechnic actuator used in AP1000 nuclear power plant squib valve was introduced, and laser beam welding danger of pyrotechnic actuator was analyzed. Sulfur-Free Black Powder charged in pyrotechnic actuator was test by Differential Scanning Calorimetry (DSC) method, and temperature at charging position of pyrotechnic actuator was measured during laser beam welding process, results of test and measure show that temperature at charging position is less than autoignition point of powder. To prevent accidents caused by spark splash, laser beam welding workshop was rebuilt, welding fixtures were designed, and safety precaution measures were added in welding technics, these measures can improve safety of laser beam welding process effectively.

Fuzzy uncertainty modeling applied to AP1000 nuclear power plant LOCA

This article presents an uncertainty modeling study using a fuzzy approach applied to the Westinghouse advanced nuclear reactor. The AP1000 Westinghouse Nuclear Power Plant (NPP) is provided of passive safety systems, based on thermo physics phenomenon, that require no operating actions, soon after an incident has been detected. The use of advanced passive safety systems in NPP has limited operational experience. As it occurs in any reliability study, statistically non-significant events report introduces a significant uncertainty level about the failure rates and basic events probabilities used on the fault tree analysis (FTA). In order to model this uncertainty, a fuzzy approach was employed to reliability analysis of the AP1000 large break Loss of Coolant Accident (LOCA). The final results have revealed that the proposed approach may be successfully applied to modeling of uncertainties in safety studies.

Managing construction risks of AP1000 nuclear power plants in China

Large and complex construction projects face risk from various sources and the successful completion of such projects depends on effective risk management. This study investigates the risk faced by Chinese firms participating in constructing AP1000 nuclear power plants in China. AP1000 nuclear reactors are new, Generation III+ reactors designed by Westinghouse and to be built first in China. The semi-structured interview approach is used to elicit information from experts involved in the AP1000 projects in China. Based on the interviews, various sources of risk are identified. In addition to general risks that megaprojects normally face, there are unique risks that arise from various sources such as technological, political, organizational, and individual personnel risks. Risk management strategies are proposed to manage general and unique risks identified in the study. The findings of this study would be helpful for Chinese companies involved in the construction of AP1000 nuclear power plants to mitigate the risks associated with the projects.