scholarly journals The SPES3 Facility for Testing an Integral Layout SMR: BDBE Simulation Analysis

2011 ◽  
Author(s):  
Roberta Ferri ◽  
Andrea Achilli ◽  
Cinzia Congiu ◽  
Gustavo Cattadori ◽  
Fosco Bianchi ◽  
...  
Keyword(s):  
Simulation Analysis ◽  
Nuclear Fission ◽  
Sensitivity Analyses ◽  
Medium Size ◽  
Design Calculation ◽  
Pressurized Water ◽  
Pressure Equalization ◽  
Plant Safety ◽  
Design Basis ◽  
Safety Features

The SPES3 facility is being built at the SIET laboratories, in the frame of an R&D program on Nuclear Fission, led by ENEA and funded by the Italian Ministry of Economic Development. The facility is based on the IRIS reactor design, an advanced medium size, integral layout, pressurized water reactor, based on the proven technology of PWR with an innovative configuration and safety features suitable to cope with Loss of Coolant Accidents through a dynamic coupling of the primary and containment systems. SPES3 is suitable to test the plant response to postulated Design and Beyond Design Basis Events, providing experimental data for code validation and plant safety analysis. It reproduces the primary, secondary and containment systems of the reactor with 1:100 volume scale, full elevation, prototypical fluid and thermal-hydraulic conditions. A design-calculation feedback process, based on the comparison between IRIS and SPES3 simulations, performed respectively by FER, with GOTHIC and RELAP5 coupled codes, and by SIET, with RELAP5 code, led to reduce the differences in the two plants behaviour, versus a 2-inch equivalent DVI line DEG break, considered the most challenging LOCA for the IRIS plant. Once available the final design of SPES3, further calculations were performed to investigate Beyond Design Basis Events, where the intervention of the Passive Containment Condenser is fundamental for the accident recovery. Sensitivity analyses showed the importance of the PCC actuation time, to limit the containment pressure, to reach an early pressure equalization between the primary and containment systems and to allow passive water transfer from the containment to the RPV, enhanced by the ADS Stage-II opening.

General Safety Design Approach and Main Safety Features of ATMEA1

2009 ◽  
Author(s):  
Eric Mathet ◽  
Philippe Lauret ◽  
Takashi Kanagawa ◽  
Ge´rard Castello ◽  
Yasuhiko Okabe
Keyword(s):  
Joint Venture ◽  
Design Approach ◽  
Severe Accident ◽  
Public Acceptance ◽  
Medium Size ◽  
Pressurized Water ◽  
Safety Design ◽  
Safety Features ◽  
High Level ◽  
General Safety

ATMEA, an AREVA NP and MHI joint venture offers ATMEA1, an innovative and reliable mid-sized evolutionary reactor based on its parent companies’ experience and technology. With a net electrical output between 1100 and 1150 MWe, ATMEA1 is aimed at ensuring competitive power generation. Its operational flexibility and features make it particularly attractive for countries with small or medium size grid. The presentation will focus on and emphasize the general safety design approach and main safety features of this Generation III+ Pressurized Water Reactor (PWR). High level safety, operational objectives, licensability and public acceptance have driven the design: Buildings and all safety systems are being designed in compliance with stringent regulations and worldwide accepted codes and standards, making ATMEA1 a licensable product in any country around the world. Public acceptance is a key element in the nuclear world. Supported by safety features like large commercial airplanes crash protection and resistance to severe accident, ATMEA1 is a robust generation III+ reactor.

Study on Emergency Action Levels of the Modular High Temperature Gas-Cooled Reactor

2017 ◽  
Author(s):  
Xiaochuan Zang ◽  
Tao Liu
Keyword(s):  
High Temperature ◽  
Fission Product ◽  
Operating Mode ◽  
Inherent Safety ◽  
Pressurized Water ◽  
Plant Safety ◽  
Water Reactors ◽  
Safety Features ◽  
Action Levels ◽  
High Temperature Gas

The emergency action level (EAL) scheme for a modular high temperature gas-cooled reactor (HTR) plant refers to the generic EAL development guidance for pressurized water reactors (PWR) with HTR modification due to its design issues. Based on reactor’s accidents analysis and consequence assessment, EAL scheme of HTR is established through the steps of category and classification. Four emergency classes are set for HTR consisting of U (Emergency Standby), A (Facilities Emergency), S (Site Area Emergency) and G (General Emergency). The Recognition Category of Initiating Condition (IC) and EAL contains A - Abnormal Rad Levels / Radiological Effluent, F - Fission Product Barrier, H - Hazards and Other Conditions Affecting Plant Safety, S - System Malfunction. The methodology for development of EALs for HTR on Fission Product Barrier and System Malfunction has some differences from PWR’s due to differences on operating mode, inherent safety features and system characteristics.

scholarly journals SPES3 Facility RELAP5 Sensitivity Analyses on the Containment System for Design Review

2012 ◽  
Vol 2012 ◽  
pp. 1-19 ◽  
Author(s):  
Andrea Achilli ◽  
Cinzia Congiu ◽  
Roberta Ferri ◽  
Fosco Bianchi ◽  
Paride Meloni ◽  
...  
Keyword(s):  
Nuclear Fission ◽  
Sensitivity Analyses ◽  
Medium Size ◽  
Facility Design ◽  
Simulation Design ◽  
Containment System ◽  
Coupled Codes ◽  
Hydraulic Conditions ◽  
Weak Points ◽  
Design And Testing

An Italian MSE R&D programme on Nuclear Fission is funding, through ENEA, the design and testing of SPES3 facility at SIET, for IRIS reactor simulation. IRIS is a modular, medium size, advanced, integral PWR, developed by an international consortium of utilities, industries, research centres and universities. SPES3 simulates the primary, secondary and containment systems of IRIS, with 1:100 volume scale, full elevation and prototypical thermal-hydraulic conditions. The RELAP5 code was extensively used in support to the design of the facility to identify criticalities and weak points in the reactor simulation. FER, at Zagreb University, performed the IRIS reactor analyses with the RELAP5 and GOTHIC coupled codes. The comparison between IRIS and SPES3 simulation results led to a simulation-design feedback process with step-by-step modifications of the facility design, up to the final configuration. For this, a series of sensitivity cases was run to investigate specific aspects affecting the trend of the main parameters of the plant, as the containment pressure and EHRS removed power, to limit fuel clad temperature excursions during accidental transients. This paper summarizes the sensitivity analyses on the containment system that allowed to review the SPES3 facility design and confirm its capability to appropriately simulate the IRIS plant.

Emergency Operating Procedures (EOPs) for the AP1000 Simulator

2006 ◽  
Author(s):  
Yuichi Hayashi ◽  
Gianfranco Saiu ◽  
Richard F. Wright
Keyword(s):  
Pressure Control ◽  
Passive Safety ◽  
Background Information ◽  
Pressurized Water ◽  
Reactor Coolant ◽  
Investment Protection ◽  
Plant Safety ◽  
Safety Features ◽  
Core Cooling ◽  
Emergency Operating

The AP1000 is two-loop 1100 MWe advanced pressurized water reactor (PWR) that uses passive safety features to enhance plant safety and to provide significant and measurable improvements in plant simplification, reliability, investment protection and plant costs. The AP1000 uses proven technology, which builds on over 30 years of operating PWR experience. The AP1000 final design certification was approved by the NRC in December, 2005. A total of 34 Emergency Operating Procedures (EOPs) for operation of the AP1000 simulator have been prepared based on the AP1000 Emergency Response Guidelines (ERGs), background information documents and detailed plant information. These include 28 EOPs at power and 6 EOPs during shutdown. The AP1000 ERGs were developed by using the generic ERGs for the low pressure reference PWR plant as a basis. The AP1000 design differences from the reference plant were reviewed and reflected in the process of developing operational steps in each ERG. The provisions of the AP1000 PRA were also reviewed and incorporated into the ERGs. Although the AP1000 design does not require operator actions for the first 72 hours after accidents, the operator actions with both safety-related and nonsafety-related equipment have an important role to mitigate the consequence of accidents. In the event of a steam generator tube rupture (SGTR), although the AP1000 is designed so that no operator actions are required to recover from the event, there are actions that can be taken by the operator to limit the release of radioactive effluents from the ruptured SG. These actions include isolation of the ruptured SG and depressurization of the reactor coolant system (RCS) to terminate primary-to-secondary leakage, restoring reactor coolant inventory to ensure adequate core cooling and plant pressure control. It is expected that these operator actions should be incorporated into the ERG to reduce the fission product release. To support the development of the AP1000 ERGs, several transient and accident analyses were performed. These include analyses for LOCA, post-LOCA cooldown and depressurization, passive safety system termination, SGTR and faulted SG isolation. These analyses results were incorporated into the ERG background information documents. In the event of SGTR, several cases were analyzed, including consideration of operator recovery actions. These cases were modeled using the best-estimate state-of-art RELAP5 code. The analyses results show that operator recovery actions are effective for SGTR to be placed under operator control.

LOCA Simulation: Analysis of Rarefaction Waves Propagating Through Geometric Singularities

2011 ◽  
Author(s):  
Fabien Crouzet ◽  
Vincent Faucher ◽  
Pascal Galon ◽  
Philippe Piteau ◽  
Patrick Izquierdo
Keyword(s):  
Simulation Analysis ◽  
Rarefaction Waves ◽  
Test Rig ◽  
Transient Wave ◽  
Pressurized Water ◽  
Dissipation Term ◽  
Loss Of Coolant Accident ◽  
Nuclear Plants ◽  
Proposed Model ◽  
Single Pipe

The propagation of a transient wave through an orifice is investigated for applications to Loss Of Coolant Accident in nuclear plants. An analytical model is proposed for the response of an orifice plate and implemented in the EUROPLEXUS fast transient dynamics software. It includes an acoustic inertial effect in addition to a quasi-steady dissipation term. The model is experimentally validated on a test rig consisting in a single pipe filled with pressurized water. The test rig is designed to generate a rapid depressurization of the pipe, by means of a bursting disk. The proposed model gives results which compare favourably with experimental data.

Preliminary LOCA Analysis of Heating-Reactor of Advanced Low-Pressurized and Passive Safety System (HAPPY)

2018 ◽  
Author(s):  
Mian Xing ◽  
Zhaocan Meng ◽  
Xiaotao Liao ◽  
Canhui Sun ◽  
Shuming Zhang ◽  
...  
Keyword(s):  
Conceptual Design ◽  
District Heating ◽  
Central Research Institute ◽  
Safety System ◽  
Passive Safety ◽  
Central Research ◽  
Pressurized Water ◽  
Design And Optimization ◽  
Passive Safety System ◽  
Safety Features

SPICRI (State Power Investment Central Research Institute) is developing a new conceptual design of heating-reactor, named Heating-reactor of Advanced low-Pressurized and Passive safetY system (HAPPY), which is targeted for the district heating, desalination of seawater, and other heat applications. It is a 200MWth two-loop low-pressurized water reactor with low thermal parameters. The whole reactor vessel is deployed inside a shielding and cooling pool with thermal insulation measure. The conceptual design of HAPPY is described in this paper, including the design criteria, safety features, main parameters and main components. A preliminary safety analysis is carried out to provide a reference for the design and optimization of HAPPY. In this paper, four different LOCA analyses are described and compared. The results show that the current design can deal well with all the selected LOCA scenarios and the effectiveness of the safety systems is proved.

Posttest Calculations of PARAMETER-SF4 Experiment With Air Ingress and Water Reflood Using SOCRAT/V3 Computer Modeling Code

2011 ◽  
Author(s):  
Alexander Vasiliev
Keyword(s):  
Experimental Data ◽  
Computer Modeling ◽  
Nuclear Power ◽  
Cooling System ◽  
Thermal Hydraulics ◽  
Pressurized Water ◽  
Vver Fuel ◽  
Design Basis ◽  
Air Ingress ◽  
Modeling Code

The PARAMETER-SF4 test conditions simulated a severe LOCA (Loss of Coolant Accident) NPP (nuclear power plant) sequence in which the overheated up to 1700–2300K core would be reflooded from the bottom in occasion of ECCS (Emergency Core Cooling System) recovery. The test was successfully conducted at the NPO “LUTCH”, Podolsk, Russia, in July 21, 2009, and was the fourth of four experiments of series PARAMETER-SF. PARAMETER facility of NPO “LUTCH” (scientific and industrial association LUTCH), Podolsk, is designed for studies of the VVER fuel assemblies behavior under conditions simulating design basis, beyond design basis and severe accidents. The test bundle was made up of 19 fuel rod simulators. Heating was carried out electrically using tantalum heating elements installed in the center of the rods and surrounded by annular UO2 pellets. The rod cladding was identical to that used in VVER (water-water energetic reactor, Russian type of pressurized water reactor). After the maximum cladding temperature of about 1900K was reached in the bundle during PARAMETER-SF4 test, the bottom flooding was initiated. The important feature of PARAMETER-SF4 test was the air ingress phase during which the air was supplied to the working section of experimental installation. It is known that zirconium oxidation in the air proceeds in a different way in comparison to oxidation in the steam. The thermal hydraulic and SFD (Severe Fuel Damage) best estimate computer modeling code SOCRAT/V3 was used for the calculation of PARAMETER-SF4 experiment. Thermal hydraulics in PARAMETER-SF4 experiment played very important role and its adequate modeling is important for the thermal analysis. The results obtained by the complex SOCRAT/V3 were compared with experimental data concerning different aspects of air ingress phase and thermal hydraulics behavior during the reflood. The temperature experimental data were found to be in a good agreement with calculated results. It is indicative of the adequacy of modeling the complicated thermo-hydraulic behavior in the PARAMETER-SF4 test.

High-Temperature Oxidation Modeling of New Perspective Accident Tolerant Fuel Claddings

2019 ◽  
Author(s):  
Alexander Vasiliev
Keyword(s):  
High Temperature ◽  
High Temperature Oxidation ◽  
Power Plants ◽  
Renewable Energy Sources ◽  
Nuclear Industry ◽  
Safety Level ◽  
Pressurized Water ◽  
Temperature Oxidation ◽  
Design Basis ◽  
New Perspective

Abstract Currently, the comprehension among the specialists and functionaries is getting stronger that the nuclear industry can encounter serious difficulties in development in the case of insufficiently decisive measures to enhance the safety level of nuclear objects. The keen competition with renewable energy sources like wind, solar or geothermal energy takes place presently and is expected to continue in future decades. One of main measures of nuclear safety enhancement could be the drastic renovation of materials used in nuclear industry. The analytical models of high-temperature oxidation of new perspective materials including chromium-nickel-based alloys, zirconium-based cladding with protective chromium coating, FeCrAl alloys and composite claddings on the basis of SiC/SiC in the course of design-basis and beyond-design-basis accidents at nuclear power plants (NPPs) are developed and implemented to severe accident computer running code. The comparison with available experimental data is conducted. The preliminary calculations of nuclear pressurized water reactor loss-of-coolant accidents with new types of claddings demonstrate encouraging results for hydrogen generation rate and integral hydrogen production. It looks optimistic for considerable upgrade of safety level for future generation NPPs using new fuel and cladding materials.

Design and Simulation of Fully Mechanized Hydraulic Support

2014 ◽  
Vol 644-650 ◽  
pp. 199-202
Author(s):  
Pei Qin Wang ◽  
Zeng Shun Xu ◽  
Zuo Feng Sun ◽  
Hui Yuan Jiang
Keyword(s):  
Mechanical Properties ◽  
Mathematical Model ◽  
Mechanical Model ◽  
Simulation Analysis ◽  
Virtual Prototype ◽  
Dynamics Simulation ◽  
Prototype Model ◽  
Design Calculation ◽  
Hydraulic Support ◽  
Element Analysis

Based on theoretical calculation, virtual prototype technology and the method of finite element analysis, the fully mechanized hydraulic support is designed and simulated. Firstly, the four-link mechanism of hydraulic support mechanical model and mathematical model are established, the demission is confirmed by design calculation of structure. Secondly, through the establishment of rigid parameterized virtual prototype model of the system, dynamics simulation analysis and research is finished based on ADAMS on the mechanical properties. Finally, based on FEA, the modal calculation of key components is completed by using ANSYS.

Fracture mechanics in design and service: ‘living with defects’ - Fracture mechanics as an aid to design and operation of nuclear plant

1981 ◽  
Vol 299 (1446) ◽  
pp. 227-236 ◽  
Keyword(s):  
Fracture Mechanics ◽  
Material Properties ◽  
Pressure Vessels ◽  
Nuclear Plant ◽  
Plant Design ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Plant Safety ◽  
Usual Criterion ◽  
Design Protection

Fracture mechanics analyses are an important part of nuclear plant design, supplementing the conventional design protection against failure to cover the possibility of the presence of crack-like defects. The degree of detail and accuracy required for a particular application depends on the possible consequences of a failure and whether the assessment is concerned with plant safety or with aspects of reliability. In the former case, a conservative approach is necessary and the prevention of initiation is the usual criterion. This approach is typified by the safety assessment applied to pressurized water reactor pressure vessels, which is outlined and discussed in relation to elastic plastic approaches and the importance of plant transient conditions, material properties (especially in weldments) and possible defect distributions. Fracture mechanics can help in defining quality control and quality assurance procedures, including both requirements for mechanical property appraisal and nondestructive testing. The latter aspects extend into operation, in respect of monitoring of plant conditions, surveillance of changes in material properties and the use of periodic inspection and plant condition monitoring techniques. A number of examples are quoted and recommendations made to permit improved fracture mechanics assessments.

Sign in / Sign up

Export Citation Format

Share Document