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Analytical Method Research of Source Term for Steam Generator Tube Rupture Accident of Small Modular reactor 23rd

Analytical Method Research of Source Term for Steam Generato...

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23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Xia, Ming-ming Wang, Jun-long Zhu, Jian-ping Tian, Chao Liu, Jia-jia Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sicuan Chengdu China

ISBN: (纸本)9789819910229

Based on the full reference to the existing engineering practice and safety review experience, and considering the actual design characteristics of the small modular reactor ACP100, a set of source term analysis method suitable for ACP100 of the steam generator tube rupture (SGTR) accident is proposed, and the source term analysis and consequence evaluation of ACP100 SGTR accident was carried out using this method. The analysis shows that the radiological consequences of the accident source term calculated by this method meet the acceptance criteria of small modular reactor. The analysis results of this article can provide support for the follow-up review of accident source term of ACP100. © 2023, The Author(s).

关键词： Accidents

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Maap5 simulation of pwr high-pressure core melting severe accident due to smallbreak loca

Maap5 simulation of pwr high-pressure core melting severe ac...

引用

2018 International Congress on Advances in Nuclear Power Plants, ICAPP 2018

作者： Wu, Xiaoli Li, Wei Deng, Jian Deng, Chunrui Hou, Liqiang Zhang, Dan Yuan, Hongsheng Science Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9780894487552

MAAP5 is capable of simulating the whole response of a PWR nuclear power plant under severe accident conditions. An analysis of the severe accident scenarios in a PWR by MAAP5 were carried out. It was assumed that the severe accident was initiated by an unusual smallbreak loss-of-coolant accident (SBLOCA) at the reactor pressure vessel (RPV) lower head and the failure of highpressure injection system (HPIS) in a PWR by using MAAP5 were carried out. Different break sizes with and without manual depressurization of the reactor coolant system (RCS) were studied. The results showed that without manually depressurization the RCS, reactor core underwent slow heatup and completely melted and eventually failed the RPV lower head at the primary system pressure of 6.87 MPa for 0.4-inch-break LOCA and 2.71 MPa for 1.0-inch-break LOCA. On the other hand, timely manually opening the pressurizer (PZR) safety valves (SVs) was an effective mitigation measure to recover the core coolabilty by cold-leg accumulator injection system (AIS) and low-pressure injection system (LPIS). Besides, reasonably manually opening the steam generators (SGs) SVs while keeping the auxiliary feedwater work also helped to depressurize the RCS and prevent the severe accident. Both of the two mitigation measures successfully prevented the core from complete melt, but the latter one is preferable to the former one provided no steam generator tube rupture took place. © 2018 American Nuclear Society. All rights reserved.

关键词： Loss of coolant accidents

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Research on Hybrid Modeling Method Based on Mechanism Model and BP Neural Network 4

Research on Hybrid Modeling Method Based on Mechanism Model ...

引用

4th Asia Conference on Information Engineering, ACIE 2024

作者： Chen, Jie Zheng, Yanqiu Chu, Qing Jiang, Guanfu Li, Jing Nuclear Power Institute of China Science and Technology on Reactor System Design Technology Laboratory Chengdu610213 China

ISBN: (纸本)9798350371925

reactor system modeling is the premise of reactor control system design. The common commercial thermal and hydraulic analysis programs like RELAP5, RETRAN have complex structure and slow operational speed, which are not conductive to the transplantation and improvement of reactor system. In order to realize the rapid development of reactor control system, this paper carries out the mechanism modeling of the reactor system in Matlab/Simulink, and then introduce BP neural network to establish the hybrid model, which improves the modeling accuracy to support the design of reactor system. For small pressurized water reactor, we firstly take analytical modeling method to build the core mechanism model and its state space model through three conservation principles and the operating mechanism of core. Based on analyses, we can conclude that the parameter error and structural error of the mechanism model is caused by the assumption and simplification during modeling, and choose the parameter with the most influence (the feedback coefficient of the coolant temperature) as the corrected parameter in the BP neural-based hybrid modeling process. After that, according to the dual constant operation scheme of small pressurized water reactors, and the gradient descent method of BP network, the hybrid model of core is built and the parameter error and structure error of the mechanism modeling are corrected online. The simulation results show that the proposed hybrid modeling method can accurately simulate the reactor operate state. Thus, the proposed hybrid model is accurate and effective. © 2024 IEEE.

关键词： Pressurized water reactors

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Reliability Verification Scheme for Safety Class DCS and Its Implementation in Nuclear Power Plant 5th

Reliability Verification Scheme for Safety Class DCS and Its...

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5th International Symposium on Software Reliability, Industrial Safety, Cyber Security and Physical Protection of Nuclear Power Plant, ISNPP 2020

作者： Liu, Ming-ming Chen, Yang He, Xian-jian Xu, Biao Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9789811634550

system Availability, Rejection rate and Malfunction rate are vital technical indicators for safety class DCS when evaluating the reliability of the control system to meet technical requirements. This trend has attracted great attention of equipment manufacturers, regulatory agencies and users. Since there is no relevant standards or norms to guide the restriction at this stage, there is no unified verification idea in the industry. Although domestic and foreign equipment enterprises have tried relevant verification research, it still remains at the level of theoretical analysis and the confidence is not high. In the research of test method, if the traditional test simulation method is used for index verification, it often requires a lot of manpower, material resources and time. Therefore, this method is also regarded as inoperable, so it has not been able to form a reasonable and mature test of the verification scheme for a long time. Based on the general principle of reliability analysis, this paper makes a comparative analysis of the objects of reliability test, and proposes a reliability analysis verification and implementation scheme of safety class DCS. © 2021, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Reliability analysis

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An Unsupervised Learning-Based Framework for Effective Representation Extraction of reactor Accidents 23rd

An Unsupervised Learning-Based Framework for Effective Repre...

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23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Li, Chengyuan Li, Meifu Qiu, Zhifang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China

ISBN: (纸本)9789811987793

Compared to knowledge-based diagnostic systems, data-based methods tend to perform better in terms of speed and accuracy in diagnosing reactor accidents, and have significant advantages in terms of scalability of models. With the increasing use of high-precision system analysis programs in nuclear engineering, the number of high-fidelity computational data for accident simulation is exploding. Therefore, an algorithm that can achieve both automatic extraction of low-dimensional features from the data and guarantee the validity of the features is needed to improve the performance and confidence of the accident diagnosis system. This study proposes an autoencoder-based autonomous learning framework, namely Padded Auto-Encoder (PAE), which is able to automatically encode accident monitoring data that has been noise-added and with partially missing data into low-dimensional feature vectors via a Vision Transformer-based encoder, and to decode the feature vectors into noise-free and complete reconstructed monitoring data. Thus, the encoder part of the framework is able to automatically infer valid representations from partially missing and noisy monitoring data that reflect the complete and noise-free original data, and the representation vectors can be used for downstream tasks for accident diagnosis or else. In this paper, LOCA of HPR1000 was used as the study object, and the PAE was trained by an unsupervised method using cases with different break locations and sizes as the dataset. The encoder part of the pre-trained PAE was subsequently used as the feature extractor for the monitoring data, and several basic statistical learning algorithms for predicting the break locations and sizes. The results of the study show that the pre-trained diagnostic model with two stages has a better performance in break location and size diagnostic capability with an improvement of 41.62% and 80.86% in the metrics respectively, compared to the diagnostic model with end-to-end model st

关键词： Location

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Study on hydrogen flame acceleration and ddt properties at elevated temperatures and pressures 23

Study on hydrogen flame acceleration and ddt properties at e...

引用

23rd International Conference on Nuclear Engineering: Nuclear Power - Reliable Global Energy, ICONE 2015

作者： Peng, Huanhuan Yu, Hongxing Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China

ISBN: (纸本)9784888982566

Hydrogen security is an important problem of the severe accident analysis in NPP, and the security of hydrogen detonation is more important. The hydrogen risk mainly focused on the risk of flame acceleration (FA) and deflagration-to-detonation transition (DDT) processes. As the processes involved in combustion and detonation are quite complex, it is difficult to present a detailed description of the whole process. Based on the physical phenomena of flame acceleration, the present paper brief introduces two criteria for hydrogen security analysis of detonation, which are FA σ criteria and DDT λ criteria. From the present paper, effect of initial pressures and temperatures on FA and DDT has been discussed. It describes the limit condition of flame acceleration by the quench/re-ignition of flammable gas mixture theory. So as to promote the flame acceleration criteria at elevated temperatures and pressures from the former one. Last the application for hydrogen safety analysis in GASFLOW code has been recommended. Copyright © 2015 by JSME.

关键词： Combustion

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The CFD calculation of a stratification break-up test of light gas in a closed vessel

The CFD calculation of a stratification break-up test of lig...

引用

2017 International Congress on Advances in Nuclear Power Plants: A New Paradigm in Nuclear Power Safety, ICAPP 2017

作者： Yazhe, Lu Yu, Zhang Xu, Ran Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9784890471676

During the course of a hypothetical severe accident in a light water reactor, spray systems could be used in the containment in order to prevent overpressure in case of a steam line break and to enhance the gas mixing in case of the presence of hydrogen. Thus these two effects induced by spray might compete with each other having an influence on the hydrogen risk. Focusing on the dynamic effect of spray, the impact of spray on the breakup of a light gas layer in a closed vessel is simulated and the experimental test evaluated here is TOSQAN 113 test that belongs to the dynamic part of SARNET FP6 benchmarks. ANSYS FLUENT 14.5 is chosen for the simulation tool and spray droplets are tracked with using the Euler-Lagrange approach-Discrete Phase Model (DPM). The code validation and several sensitivity studies are done to sum up the basic method of numerical simulation of spray with CFD code and assess the influence of calculation settings and parameters, such as turbulence model and spray boundary condition. The code validation shows that the results of calculations acceptably agree with the experimental data and the sensitivity studies reveal that spray boundary condition is a rather important factor for the accuracy of simulation. Conclusion gained in from the present work can provide some reference of the numerical analysis of containment.

关键词： Boundary conditions

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A calibration method for raman distributed temperature sensor in nuclear power plant

A calibration method for raman distributed temperature senso...

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2019 Transactions of the American Nuclear Society, ANS 2019

作者： Xiang, Meiqiong Liu, Yanyang Qing, Xianguo Zhu, Jialiang Wu, Qian Zhu, Biwei Xu, Sijie Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

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Research on boundary disposal of pipes decoupling based on PEPs 27

Research on boundary disposal of pipes decoupling based on P...

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27th International Conference on Nuclear Engineering: Nuclear Power Saves the World!, ICONE 2019

作者： Yanli, Yuan Feng, Yuan Xinjun, Wang Xiaoming, Bai Rui, Zhang Liangang, Zheng Xifeng, Lu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China

ISBN: (纸本)9784888983051

Generally, one of the remarkable characteristics is: compared with the main pipe, the sizes of the auxiliary piping are smaller, which makes the auxiliary piping satisfy the criteria to be decoupled from the main piping. So when the auxiliary pipes connected to the main piping are analyzed, there are two methods: one is that the auxiliary pipe is analyzed connected to the main piping, and the other is that the auxiliary pipe is analyzed not to be connected to the main pipe. Because the response spectrum at the location of the connection of the auxiliary pipes and the main piping is lacked, the previous choice is usually adopted. At present, the main software used for analyzing the stress of the pipes in nuclear power station is PEPS, but when the PEPS is adopted to analyze the small pipes, if the small pipes are analyzed connected to the main pipe, the boundaries of the thermal analysis and those of the seismic analysis are different, so the two kinds of analysis cannot be carried out at the only one model. That is, when the small pipe is analyzed connected to the main piping, two models are needed for thermal analysis and the seismic analysis, and then the results of the thermal analysis and the seismic analysis are distilled separately, and the results are combined according to determinate mode, and then the combined results are evaluated according to the criterion. Consequently, not only is the model complex, but also the combination of the seismic stress and the thermal stress is not convenient in the previous method. In the latter method, the auxiliary pipe is not connected to the main piping, accordingly, not only is the modal simple, but also the combination of the seismic stress and the thermal stress is convenient. The only problem which is needed to be solved is the seismic response spectrum and the displacements, and an approach to solve the problem is presented in the article. A response spectrum calculation method is obtained based on the single oscilla

关键词： Thermoanalysis

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Research on Thermal-Hydraulic Model and Characteristics of Lead Cooled Traveling Wave reactor 23rd

Research on Thermal-Hydraulic Model and Characteristics of L...

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23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Zhang, Yaoxiang Zhang, Xue Yu, Hongxing Du, Sijia Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9789819910229

Traveling wave reactor (TWR) is an advanced nuclear power system, which can keep the total amount of fissionable nuclides constant during its lifetime through the transformation of fissionable nuclides in the reactor, so as to maintain the stability of the effective proliferation factor of the reactor core. As a coolant, lead has low neutron moderation ability and capture cross section. At the same time, it also has excellent thermal characteristics. Lead cooled traveling wave reactor is a reactor with high inherent safety and good application prospects. The research on reactor core mainly focuses on the physical calculation of the reactor, and the analysis of its thermal hydraulic model and characteristics is relatively scarce, which limits the design and safety analysis of the advanced reactor. Therefore, it is necessary to investigate the thermal hydraulic model of lead cooled traveling wave reactor. In this paper, a thermal hydraulic model suitable for the core of lead cooled traveling wave reactor is proposed. Considering the axial particularity of heat release from the core of traveling wave reactor, the model considers the axial thermal conductivity of fuel rods ignored by other thermal hydraulic models. Due to the strong thermal conductivity of core coolant, the thermal model also considers the heat transfer of fuel assemblies and the flow distribution of core flow. The research shows that the thermal hydraulic model is suitable for the analysis of thermal hydraulic characteristics of lead cooled traveling wave reactor, the thermal hydraulic characteristics are studied by using the thermal hydraulic model. © 2023, The Author(s).

关键词： Thermal conductivity

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