Study on Typical Initial Event of ATWS for Small Modular Reactor

2016 ◽  
Author(s):  
Zhang Dan ◽  
Ran Xu ◽  
Qiu Zhifang ◽  
Zhou Ke ◽  
Feng Li
Keyword(s):  
Optimum Design ◽  
Safety Analysis ◽  
Accident Analysis ◽  
Pressurized Water ◽  
Control Rod ◽  
Reactor Coolant ◽  
Initial Event ◽  
Small Modular Reactor ◽  
Activation System

The method for ATWS (anticipated transient without scram) analysis was completely developed for commercial pressurized water (PWR) reactor plants, especially for selecting of typical initial events. For accident analysis of ATWS, it is different between PWR and small modular reactor (SMR), as different structures and characters, and it is necessary to study the typical initial events for these reactors. Based on the standard of PWR, the demanding for ATWS analysis was studied and the consequences for typical anticipated transient was calculated using RELAP5/MOD3.2 code, “maintain reactor coolant pressure boundary integrity” was selected as limiting criterion. The results shows for SMR, anticipated transient with the most serious consequence for ATWS are loss of offsite power and inadvertent control rod withdraw event, this conclusion will support to prepare the safety analysis report and optimum design of diversity activation system (DAS) for SMR.

A New Safety Analysis Method of Control Rod Ejection Accident in PWR NPP Based on the Failure of Causal Relationship

2018 ◽  
Author(s):  
Shiyu Yan ◽  
Hua Liu ◽  
Zhaohui Liu ◽  
Xiaohua Yang ◽  
Meng Li ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Safety Analysis ◽  
Pressurized Water ◽  
Control Rod ◽  
Drive Mechanism ◽  
Cause And Effect ◽  
Simulation Calculation ◽  
Causal Logic

In view of control rod ejection accident of the traditional pressurized water reactor, the safety thought of the design phase is to validate reliability and availability of DCS I&C in the severe accidents. Now the most important and effective means is simulation calculation and analysis. It is applied for the imaginary accident of the nuclear power plant by using computer software. The new safety analysis steps based on the analysis of cause-and-effect logic failure: firstly, the composition and working principle of control rod drive mechanism is analyzed; secondly, a list of factors-the dynamics and structure, environmental reasons, the function of the control rod drive mechanism and status analysis-are all taken into account, the initial cause of failure modes with causal logic analysis is carried out; thirdly, based on cause-and-effect logic failure, the prevention and improvement measures of accidents, the new criterion of design are put forward. The advantages of cause-and-effect logic failure safety analysis: 1.be based on causal logic. 2. the system aspects is added compared with the past method that is only based on simulation calculation and analysis of the hypothetical accident, the accident the transient process of the key security parameters as the acceptance criteria. 3. The verification and audit of the lack of safety design criteria, completeness of design content, sufficiency problem are performed before the simulated calculation and analysis. 4. The coverage of safety analysis is expanded. Some good advices are provided for the design, operation and maintenance of nuclear power plant.

scholarly journals Truly-Optimized PWR Lattice for Innovative Soluble-Boron-Free Small Modular Reactor

2021 ◽  
Author(s):  
Xuan Ha Nguyen ◽  
Seongdong Jang ◽  
Yonghee Kim
Keyword(s):  
High Performance ◽  
Local Power ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Control Rod ◽  
Low Leakage ◽  
Small Modular Reactor ◽  
Burnable Absorber ◽  
Safety Parameters ◽  
Atom Core

Abstract A novel re-optimization of fuel assembly (FA) and new innovative burnable absorber (BA) concepts are investigated in this paper to pursue a high-performance soluble-boron-free (SBF) small modular reactor (SMR), named autonomous transportable on-demand reactor module (ATOM). A truly optimized PWR (TOP) lattice concept has been introduced to maximize the neutron economy while enhancing the inherent safety of an SBF pressurized water reactor. For an SBF SMR design, the 3-D centrally-shielded BA (CSBA) design is utilized and another innovative 3-D BA called disk-type BA (DiBA) is proposed in this study. Both CSBA and DiBA designs are investigated in terms of material, spatial self-shielding effects, and thermo-mechanical properties. A low-leakage two-batch fuel management is optimized for both conventional and TOP-based SBF ATOM cores. A combination of CSBA and DiBA is introduced to achieve a very small reactivity swing (<1,000 pcm) as well as a long cycle length and high fuel burnup. For the SBF ATOM core, safety parameters are evaluated and the moderator temperature coefficient is shown to remain sufficiently and similarly negative throughout the whole cycle. It is demonstrated that the small excess reactivity can be well managed by mechanical shim rods with a marginal increase in the local power peaking, and a cold-zero shutdown is possible with a pseudo checker-board control rod pattern. In addition, a thermal-hydraulic-coupled neutronic analysis of the ATOM core is discussed.

scholarly journals Integral PWR-Type Small Modular Reactor Developmental Status, Design Characteristics and Passive Features: A Review

Energies ◽  
2020 ◽  
Vol 13 (11) ◽  
pp. 2898 ◽  
Author(s):  
Chireuding Zeliang ◽  
Yi Mi ◽  
Akira Tokuhiro ◽  
Lixuan Lu ◽  
Aleksey Rezvoi
Keyword(s):  
Technology Development ◽  
Safety Margin ◽  
Heat Removal ◽  
Injection System ◽  
Regulatory Review ◽  
Pressurized Water ◽  
Control Rod ◽  
Energy Applications ◽  
Small Modular Reactor ◽  
Under Construction

In recent years, the trend in small modular reactor (SMR) technology development has been towards the water-cooled integral pressurized water reactor (iPWR) type. The innovative and unique characteristics of iPWR-type SMRs provide an enhanced safety margin, and thus offer the potential to expand the use of safe, clean, and reliable nuclear energy to a broad range of energy applications. Currently in the world, there are about eleven (11) iPWR-type SMRs concepts and designs that are in various phases of development: under construction, licensed or in the licensing review process, the development phase, and conceptual design phase. Lack of national and/or internatonal comparative framework for safety in SMR design, as well as the proprietary nature of designs introduces non-uniformity and uncertainties in regulatory review. That said, the major primary reactor coolant system components, such as the steam generator (SG), pressurizer (PRZ), and control rod drive mechanism (CRDM) are integrated within the reactor pressure vessel (RPV) to inherently eliminate or minimize potential accident initiators, such as LB-loss of coolant accidents (LOCAs). This paper presents the design status, innovative features and characteristics of iPWR-type SMRs. We delineate the common technology trends, and highlight the key features of each design. These reactor concepts exploit natural physical laws such as gravity to achieve the safety functions with high level of margin and reliability. In fact, many SMR designs employ passive safety systems (PSS) to meet the evolving stringent regulatory requirements, and the extended consideration for severe accidents. A generic classification of PSS is provided. We constrain our discussion to the decay heat removal system, safety injection system, reactor depressurization system, and containment system. A review and comparative assessment of these passive features in each iPWR-type SMR design is considered, and we underline how it maybe more advantageous to employ passive systems in SMRs in contrast to conventional reactor designs.

scholarly journals Core Design of a Small Pressurized Water Reactor with AP1000 Fuel Assembly Using SRAC and COBRA-EN Codes

2020 ◽  
Vol 2020 ◽  
pp. 1-7
Author(s):  
Van Khanh Hoang
Keyword(s):  
Safety Margin ◽  
Coolant Temperature ◽  
Pressurized Water Reactor ◽  
Fuel Temperature ◽  
Pressurized Water ◽  
Control Rod ◽  
The Core ◽  
Small Modular Reactor ◽  
Active Core ◽  
And Performance

This paper presents the core design and performance characteristics of a 300 MWt small modular reactor (SMR) with fuel assemblies of the AP1000 reactor. Numerical calculations have been performed to evaluate a proper active core size and core loading pattern using the SRAC code system with the JENDL-4.0 data library and the CORBRA-EN code. The calculated temperature coefficients including fuel temperature, coolant temperature, and isothermal temperature coefficient provide adequate negative reactivity feedbacks. The thermal-hydraulic analysis reveals acceptable radial and axial fuel element temperature profiles with significant safety margin of fuel and clad surface temperature. A safety analysis using the CORBRA-EN code shows that the core will remain covered during the entire transient procedure of the fast transient of remarkably increasing power that would be caused by the ejection of control rod. The analysis results indicate that the core with a cycle length of 2.22 years is achievable while satisfying the operation and safety-related design criteria with sufficient margins.

scholarly journals Truly-optimized PWR lattice for innovative soluble-boron-free small modular reactor

2021 ◽  
Vol 11 (1) ◽  
Author(s):  
Xuan Ha Nguyen ◽  
Seongdong Jang ◽  
Yonghee Kim
Keyword(s):  
High Performance ◽  
Local Power ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Control Rod ◽  
Low Leakage ◽  
Small Modular Reactor ◽  
Burnable Absorber ◽  
Safety Parameters ◽  
Atom Core

AbstractA novel re-optimization of fuel assembly and new innovative burnable absorber (BA) concepts are investigated in this paper to pursue a high-performance soluble-boron-free (SBF) small modular reactor (SMR), named autonomous transportable on-demand reactor module (ATOM). A truly optimized PWR (TOP) lattice concept has been introduced to maximize the neutron economy while enhancing the inherent safety of an SBF pressurized water reactor. For an SBF SMR design, the 3-D centrally-shielded BA (CSBA) design is utilized and another innovative 3-D BA called disk-type BA (DiBA) is proposed in this study. Both CSBA and DiBA designs are investigated in terms of material, spatial self-shielding effects, and thermo-mechanical properties. A low-leakage two-batch fuel management is optimized for both conventional and TOP-based SBF ATOM cores. A combination of CSBA and DiBA is introduced to achieve a very small reactivity swing (< 1000 pcm) as well as a long cycle length and high fuel burnup. For the SBF ATOM core, safety parameters are evaluated and the moderator temperature coefficient is shown to remain sufficiently and similarly negative throughout the whole cycle. It is demonstrated that the small excess reactivity can be well managed by mechanical shim rods with a marginal increase in the local power peaking, and a cold-zero shutdown is possible with a pseudo checker-board control rod pattern. In addition, a thermal–hydraulic-coupled neutronic analysis of the ATOM core is discussed.

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

2019 ◽  
Vol 141 (6) ◽  
Author(s):  
Rui Xu ◽  
Yun Long ◽  
Yaoyu Hu ◽  
Junlian Yin ◽  
Dezhong Wang
Keyword(s):  
Large Mass ◽  
Stiffness Coefficient ◽  
Pressurized Water ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Various Boundary Conditions ◽  
Clearance Flow ◽  
Mass Coefficient ◽  
Cfd Method ◽  
Canned Motor

Reactor coolant pump (RCP) is one of the most important 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 RCP, the liquid coolant fills up the clearance between the metal shields of the rotor and stator inside the canned motor, forming a long clearance flow. The fluid-induced forces of the clearance flow in canned motor RCP and their effects on the rotordynamic characteristics of the pump are numerically and experimentally analyzed in this work. A transient computational fluid dynamics (CFD) method has been used to investigate the fluid-induced force of the clearance. A vertical experiment rig has also been established for the purpose of measuring the fluid-induced forces. Fluid-induced forces of clearance flow with various whirl frequencies and various boundary conditions are obtained through the CFD method and 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 RCP does not deteriorate despite the existence of significant cross-coupled stiffness coefficient from the fluid-induced forces of the clearance flow.

3D Simulation of a Peripheral Adapter J-Groove Attachment Weld in a Vessel Head

2014 ◽  
Author(s):  
Philippe Mourgue ◽  
Vincent Robin ◽  
Philippe Gilles ◽  
Florence Gommez ◽  
Alexandre Brosse ◽  
...  
Keyword(s):  
Complex Geometry ◽  
3D Simulation ◽  
Welding Parameters ◽  
Manufacturing Processes ◽  
Pressurized Water ◽  
Control Rod ◽  
Sequence Modeling ◽  
Component Design ◽  
Water Reactors ◽  
Welding Processes

In Pressurized Water Reactors, most of heavy components and pipes have a large thickness and their manufacturing processes often require multi-pass welding. Despite the stiffness of these components, the distortion issue may be important for operational requirements (e.g. misalignment) or controllability reasons (Non Destructive Examinations have to be achievable, therefore ovalization should be limited). These requirements may be difficult to achieve by simply adjusting welding processes. Indeed because of the complexity of mechanisms involved during a welding operation and the high number of influencing parameters, this process is still essentially based on the experience of the welder. Furthermore the experimental estimation of the stress and distortion level in the component remains a difficult task that is subject to errors even if techniques are currently improved to become more accurate. These are the reasons why AREVA has put a large effort to improve welding numerical simulations, in order to have a better understanding of the involved physical phenomena and also to predict the residual state through the structure. Computational welding mechanics is used to qualify the manufacturing processes in the very early phase of the welded component design. Within the framework of a R&D program whose main objective was to improve tools for the numerical simulation of welding regarding industrial needs, AREVA has decided to validate new methodologies based on 3D computation by comparison with measurements. For this validation task the chosen industrial demonstrator was a Control Rod Drive Mechanism (CRDM) Nozzle with a J-groove attachment weld to the vessel head. For such an application, operations of post-joining straightening have to be limited, if not prohibited, because of their cost or the impossibility to use them in front of a steel giant. The control of distortion during welding operations is a key issue for which simulation can be of great help. Regarding distortion issues, both accurate metal deposit sequence modeling and respect of the real welding parameters are mandatory, especially for multi-pass operation on such a complex geometry. The aim of this paper is to present the simulation of the distortion of a peripheral adapter J-groove attachment weld mock-up. This new full 3D simulation improves the result of the previous one based on lumped pass deposits. It is the result of a fruitful collaboration between AREVA and ESI-Group.

A Design Proposal and Assessment of In-Vessel Retention Strategy for 2nd Generation PWR

2013 ◽  
Author(s):  
Yabing Li ◽  
Lili Tong ◽  
Xuewu Cao
Keyword(s):  
Metal Layer ◽  
Success Probability ◽  
Computer Code ◽  
Injection System ◽  
Accident Analysis ◽  
Fukushima Accident ◽  
Pressurized Water ◽  
Water Storage Tank ◽  
2Nd Generation ◽  
Probability Density Distributions

In order to satisfy the higher safety requirements after the Fukushima Accident, in-vessel retention (IVR) strategy is suggested to be a measure for the improvement for the operating 2nd generation Pressurized Water Reactor (PWR). In this paper, a design of passive cavity flooding system isproposed for the 2nd generation modified PWR with reference to the design of in-containment refueling water storage tank (IRWST)for AP1000, and the IVR assessmentis evaluated with the risk oriented accident analysis methodology (ROAAM). Seven representational accident sequences are selected for the IVR assessment and analyzed by lumped-parameter safety analysis computer code. Accident analysis focuses on four key parameters forthe assessment, that is, decay power, zirconium oxidation fraction, and both the mass of oxidic pool and metal layer. The values of them are obtained through analysis and used for the assessment. The probability density distributions of those parameters are determined by combining the analysis results and engineering judgment. The success probability of IVR from the viewpoint of thermal failure is predicted using a program written by the Matlab code. Furthermore, some sensitivity analysis and parametric studies are investigated to support the assessment. The assessment result shows that the success probability of the cavity injection system is higher than 99%. Detail analysis about system reliability and feasibility is needed for further work.

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