The IRIS Spool-Type Reactor Coolant Pump

2002 ◽  
Author(s):  
J. M. Kujawski ◽  
D. M. Kitch ◽  
L. E. Conway
Keyword(s):  
Reactor Core ◽  
Pressure Vessels ◽  
Reactor Vessel ◽  
Steam Generators ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
The Us ◽  
Loss Of Coolant ◽  
The Individual ◽  
Integral Reactor

IRIS (International Reactor Innovative and Secure) is a light water cooled, 335 MWe power reactor which is being designed by an international consortium as part of the US DOE NERI Program. IRIS features an integral reactor vessel that contains all the major reactor coolant system components including the reactor core, the coolant pumps, the steam generators and the pressurizer. This integral design approach eliminates the large coolant loop piping, and thus eliminates large loss-of-coolant accidents (LOCAs) as well as the individual component pressure vessels and supports. In addition, IRIS is being designed with a long life core and enhanced safety to address the requirements defined by the US DOE for Generation IV reactors. One of the innovative features of the IRIS design is the adoption of a reactor coolant pump (called “spool” pump) which is completely contained inside the reactor vessel. Background, status and future developments of the IRIS spool pump are presented in this paper.

Steam Generator of the International Reactor Innovative and Secure

2002 ◽  
Author(s):  
L. Cinotti ◽  
M. Bruzzone ◽  
N. Meda ◽  
G. Corsini ◽  
C. V. Lombardi ◽  
...  
Keyword(s):  
Steam Generator ◽  
Reactor Core ◽  
Pressure Vessels ◽  
Reactor Vessel ◽  
Design Activity ◽  
Steam Generators ◽  
The Us ◽  
The Subject ◽  
The Individual ◽  
Integral Reactor

IRIS (International Reactor Innovative and Secure) is a light water cooled, 335 MWe power reactor which is being designed by an international consortium as part of the US DOE NERI Program. IRIS features an integral reactor vessel that contains all the main reactor coolant system components including the reactor core, the coolant pumps, the steam generators and the pressurizer. This integral design approach eliminates the large coolant loop piping, and thus eliminates large loss-of-coolant accidents (LOCAs) as well as the individual component pressure vessels and supports. In addition, IRIS is being designed with a long-life core and enhanced safety to address the requirements defined by the US DOE for Generation IV reactors. The design of the steam generators, which are internally contained within the reactor vessel, is a major design effort in the development of the integral IRIS concept. The ongoing design activity about the steam generator is the subject of this paper.

The IRIS General Plant Arrangement

2002 ◽  
Author(s):  
J. Robertson ◽  
J. Love ◽  
R. Morgan ◽  
L. E. Conway
Keyword(s):  
Individual Component ◽  
Reactor Core ◽  
Pressure Vessels ◽  
Steam Generators ◽  
The Us ◽  
Loss Of Coolant ◽  
Coolant Loop ◽  
The Individual ◽  
Integral Reactor ◽  
Coolant System

IRIS (International Reactor Innovative and Secure) is a light water cooled, 335 MWe power reactor which is being designed by an international consortium as part of the US DOE NERI Program. IRIS features an integral reactor vessel that contains all the major reactor coolant system components including the reactor core, the coolant pumps, the steam generators and the pressurizer. This integral design approach eliminates the large coolant loop piping, and thus eliminates large loss-of-coolant accidents (LOCAs) as well as the individual component pressure vessels and supports. In addition, IRIS is being designed with a long life core and enhanced safety to address the requirements defined by the US DOE for Generation IV reactors. Bechtel, with Westinghouse consultation, has performed a layout study of the IRIS plant and this paper will discuss the results of this design effort.

Effect of Support Friction in Thermal Stress of VVER-1000 RCS for NOP Condition

2014 ◽  
Author(s):  
Truong Quang Nguyen ◽  
Ihn Namgung
Keyword(s):  
Thermal Stress ◽  
Yield Stress ◽  
Steam Generator ◽  
Reactor Vessel ◽  
Inside Surface ◽  
Steam Generators ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Hexahedral Elements ◽  
Effect Of Support

The main purpose of this research is to investigate the effect of friction in the thermal stress of Reactor Coolant System (RCS) of VVER-1000. RCS is a large system connecting reactor vessel, steam generators and RC Pumps. During the heat-up of reactor, the RCS expand and during cool-down of reactor, it contracts. Because of the heavy weight of reactor and steam generator, the friction at the support of RCS affects the thermal stress of RCS. In this paper how much support friction contributes to the development of thermal stress is assessed in order to investigate the thermal stress and effect of support friction. A quarter-symmetry model of VVER-1000 RCS is developed in ANSYS and meshed with hexahedral elements to ensure better solution accuracies. The model includes reactor vessel, steam generator and reactor coolant pump. Internals of reactor vessel, steam generators and RCPs are represented by point mass to simplify the model. Temperature of inside surface of hot-leg side of reactor vessel to inlet side of steam generator is assumed same uniform hot-leg temperature, and the temperature of inside surface of outlet side of steam generator to reactor vessel is uniform cold-leg temperature. All outside surface are assumed insulated. The analysis includes neither transient thermal loading nor dynamic loadings. The analysis results show that friction at support brings little effect on the peak thermal stress. The peak thermal stress occurs at hot-leg nozzle of reactor pressure vessel and it approached near yield stress. If load combination is included the localized total stress at hot-leg nozzle could go over the yield stress. This peak stress could affect fatigue life in a long run. A recommendation is made that a detailed fatigue analysis of VVER-1000 RCS is necessary.

Research on Modular Modeling Method of Reactor Coolant System Based on THEATRe

2014 ◽  
Author(s):  
Bo Shi ◽  
Zhao-Fei Tian
Keyword(s):  
Real Time ◽  
Data Exchange ◽  
Reactor Core ◽  
Simulation Software ◽  
Modeling Method ◽  
Time Step ◽  
Coolant Pump ◽  
Modular Modeling ◽  
Reactor Coolant ◽  
Coolant System

At present, research on the reactor coolant system is less yet, though modular modeling method has been widely used in the second-loop system of reactor. This paper takes the reactor coolant system of Qinshan-1 nuclear power plant as the object of study, analyses and researches on modular modeling method of reactor coolant system based on THEATRe, which is a large Thermal-Hydraulic real time simulation software developed by GSE Company and adopts NMNP (Nodal Momentum Nodal Pressure) solving method. This research establishes the modular model of the reactor coolant system equipments (including reactor core, main coolant pump, pressurizer, steam generator) using the THEATRe code. Due to each module is wrote into through different input cards, they can be solved by using their own matrix of velocity-pressure to guarantee the independence of the numerical calculation for different modular modules. THEATRe code does not have its own TDV like relap-5, meanwhile it also needs to ensure the pressurizer module can play a role in the multi-pressure node system. So this paper modifies solving method of the THEATRe source code to get suitable pressure boundary and flux boundary for RCS equipment modular module, and selects reasonable time step and data exchange frequency to achieve the data exchange of boundary pressure, flux and enthalpy among the equipment modules, which lays the foundation of establishing the real-time modular simulation model of the reactor coolant system in the future.

The Research on Core Melting Process: Oxidation

2013 ◽  
Author(s):  
Jun Wang ◽  
Wenxi Tian ◽  
Jianan Lu ◽  
Yingying Ma ◽  
Guanghui Su ◽  
...  
Keyword(s):  
Natural Circulation ◽  
Reactor Core ◽  
Melting Process ◽  
Coolant Pump ◽  
Fuel Rod ◽  
Fuel Pellets ◽  
Design Basis ◽  
Loss Of Coolant ◽  
Design Basis Accident ◽  
Station Blackout

Beyond-design basis accidents in the AP1000 may result in reactor core melting and are therefore termed core melt accidents. The aim of this work is to develop a code to calculate and analyze the oxidation of a single fuel rod with total failures of engineered safeguard systems under a certain beyond-design basis accident such as a gigantic earthquake which can result in station blackout and then total loss of coolant flow. Using the code, the responses of the most dangerous fuel rod in the AP1000 were calculated under the accident. A discussion involving fuel pellets melting, cladding rupture and oxidation, and hydrogen production then was carried out, focused on DNBR during coolant pump coastdown, the cladding intactness under different flow rates in natural circulation, and the delay effect on cladding rupture due to cladding oxidation. By the analysis of calculated results, several suggestions on guaranteeing the security of fuel rods were provided.

The Application of Concurrent Installation of the RCS Cross Over Leg and the Reactor Vessel Internals

2010 ◽  
Author(s):  
K.-W. Park ◽  
J.-H. Bae ◽  
S.-H. Park
Keyword(s):  
Feasibility Study ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Reactor Vessel ◽  
Pressurized Water ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Study Results ◽  
Under Construction

The reactor vessel internals (RVI) of a pressurized water reactor (PWR) must be installed precisely in the reactor vessel (RV) according to the requirements for levelness, orientation and vertical alignments for its proper functions and structural integrity. For the precise installation, deformation of the RV should be controlled during the RVI installation. Traditionally, the RVI has been installed in the RV after the completion of welding work for large bore pipings in the reactor coolant system (RCS). To reduce installation time, the concurrent installation of the RVI and RCS pipings is investigated. This paper describes the feasibility study on the concurrent installation including the Finite Element Method (FEM) analyses of the RV deformation due to the welding and heat treatment of the pipings. Based on the feasibility study results, the optimum schedule of the RVI installation in parallel with the installation of the cross-over leg pipings (reactor coolant pump inlet pipings) and confirmation measurement locations are developed. Thereby the concurrent installation will be applied to the nuclear power plants under construction in Korea, and it is expected to reduce installation period of 2 months compared to the traditional sequential installation method.

CFD Analysis of Thermal Mixing and Mass Flux Distribution in the PWR After MSLB

2017 ◽  
Author(s):  
Hong Xu ◽  
Yiban Xu ◽  
Liping Cao ◽  
Yixing Sung ◽  
Vefa N. Kucukboyaci ◽  
...  
Keyword(s):  
Mass Flux ◽  
Cfd Simulation ◽  
Reactor Core ◽  
Reactor Vessel ◽  
High Flow ◽  
Low Flow ◽  
Cfd Modeling ◽  
Asymmetric Effect ◽  
Reactor Coolant ◽  
Flow Case

During a postulated main steam line break (MSLB) event of a Pressurized Water Reactor (PWR) initiated at the Hot Zero Power (HZP) condition, increased heat removal from the broken steam generator (SG) on the secondary side that significantly reduces the coolant temperature on the primary side, and cold primary coolant enters the reactor vessel through the affected loop resulting in asymmetric temperature and mass flux distributions into the reactor core. A plant safety analysis under the MSLB condition needs to account for the thermal and mass flux asymmetry effects on the reactor core response due to the colder water flowing from the affected SG and the reactor coolant system (RCS) to reactor vessel. High resolution computational fluid dynamics (CFD) methodology with ANSYS CFX (Version 16.1) software was applied to analyze the flow behaviors and thermal-hydraulic phenomena and to study the thermal mixing and asymmetry effects in the downcomer and lower-plenum of a typical Westinghouse design four-loop PWR under the MSLB conditions. Two scenarios were considered for the CFD simulation distinct by reactor coolant pump status: (1) Low-flow case: without offsite power where the reactor core is cooled through natural circulation (2) High-flow case: with offsite power available and the reactor coolant pumps in operation The CFX CFD modeling and simulation were based on the reactor vessel boundary conditions from a system code transient simulation at the limiting time steps with respect to thermal margin of the fuel design. The geometric model included the vessel downcomer and the lower internals up to the reactor core inlet below the fuel assemblies. The results of CFD simulation show the different flow patterns and temperature distributions at the reactor core inlet for the low-flow case and for the high-flow case. Thermal asymmetric effect exists in both cases, but in the low-flow case, cold flow enters into core inlets at the opposite side of faulted loop located, and in the high-flow case cold flow enters into core inlets at the same side of faulted loop located. A mass flux asymmetric effect exists in both cases, but for the low-flow case, the core inlet mass flow distribution is more uniform than that for the high-flow case. The reactor core inlet distributions under the MSLB condition were further evaluated through comparisons with the results from the STAR-CCM+ (Version 10.04.01) CFD modeling and simulation. The evaluation showed that the simulation results are in good agreement with the STAR-CCM+ predictions and consistent with the phenomenon observed in an experiment published in open literature and site engineer judgment based on the available detected data.

Lower Plenum Voiding

1982 ◽  
Vol 104 (3) ◽  
pp. 479-486 ◽  
Author(s):  
D. Bharathan ◽  
G. B. Wallis ◽  
H. J. Richter
Keyword(s):  
Reactor Core ◽  
Reactor Vessel ◽  
Pressurized Water Reactor ◽  
Coolant Circuit ◽  
Two Phase ◽  
Primary Coolant ◽  
Pressurized Water ◽  
The Core ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

One of the phenomena involved in a loss-of-coolant accident in a pressurized water reactor may be lower plenum voiding. This might occur during the blowdown phase after a cold-leg break in the primary coolant circuit. Steam generated in the reactor core may flow out of the bottom of the reactor core, turn in the lower plenum of the vessel, in a direction countercurrent to the emergency core coolant flow, and escape via the break. If its velocity is high enough, this steam may sweep water from the bottom (lower plenum) of the reactor vessel. Emergency coolant added to the vessel may also be carried out by the escaping steam and thus the reflooding of the core would be delayed. This paper describes a study of two-phase hydrodynamics associated with lower plenum voiding. Several geometrical configurations were tested at three different scales, using air to simulate the steam. Comparisons were made with data obtained by other researchers.

scholarly journals Source Term Estimation under the SBLOCA-Induced Severe Accident Condition in the SMART

2021 ◽  
Vol 2021 ◽  
pp. 1-14
Author(s):  
Jaehyun Ham ◽  
Sang Ho Kim ◽  
Sung Il Kim ◽  
Byeonghee Lee ◽  
Jong-Hwa Park ◽  
...  
Keyword(s):  
Source Term ◽  
Thermal Power ◽  
Reactor Core ◽  
Heat Removal ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Injection System ◽  
Spray System ◽  
Loss Of Coolant ◽  
Accident Condition

The SMART is a system-integrated modular reactor in which a nuclear steam supply system with a thermal power of 365 MW is contained inside of the reactor vessel. Although the probability is very low, the reactor core can be damaged during a small break loss-of-coolant accident when both the passive safety injection system and the passive residual heat removal system are completely unavailable. In this work, a total of five cases were analyzed considering the reactor vessel condition and the availability of the radioactivity removal tanks and the ancillary containment spray system as containment condition variables using MELCOR code. It was estimated that there is no containment failure based on pressure, hydrogen mole fraction, and ablation depth, so that the release fractions of the 12 classes of fission products in MELCOR were evaluated considering design leak only for all cases. The overall source term of the case in which the integrity of the reactor vessel is maintained by the early initiation of the cavity flooding system was similar to that of the reactor vessel failure case. While the release fraction of cesium to the environment was analyzed to increase when there is no water in the radioactivity removal tanks, the fraction is small enough at which the radioactivity of the released cesium-137 remains well below 100 TBq, a regulatory limit. Moreover, it was found that the source term can be cut in half if the ancillary containment spray system is available. The results of this study verify the safety performance of the SMART under the small break loss-of-coolant severe accident condition with respect to the source term of interest.

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