Various Design Aspects of the Canadian Supercritical Water-Cooled Reactor Core

2015 ◽  
Vol 2 (1) ◽  
Author(s):  
Metin Yetisir ◽  
Rui Xu ◽  
Michel Gaudet ◽  
Mohammad Movassat ◽  
Holly Hamilton ◽  
...  
Keyword(s):  
Supercritical Water ◽  
Thermal Efficiency ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Operating Conditions ◽  
Stress Distributions ◽  
Light Water Reactors ◽  
Core Components ◽  
Water Reactors ◽  
Supercritical Water Cooled Reactor

The Canadian Supercritical Water-Cooled Reactor (SCWR) is a 1200 MW(e) channel-type nuclear reactor. The reactor core includes 336 vertical pressurized fuel channels immersed in a low-pressure heavy water moderator and calandria vessel containment. The supercritical water (SCW) coolant flows into the fuel channels through a common inlet plenum and exits through a common outlet header. One of the main features of the Canadian SCWR concept is the high-pressure (25 MPa) and high-temperature (350°C at the inlet, 625°C at the outlet) operating conditions that result in an estimated thermal efficiency of 48%. This is significantly higher than the thermal efficiency of the present light water reactors, which is about 33%. This paper presents a description of the Canadian SCWR core design concept; various numerical analyses performed to understand the temperature, flow, and stress distributions of various core components; and how the analyses results provided input for improved concept development.

scholarly journals DEVELOPMENT OF A HYDROGEN MANAGEMENT CONCEPT FOR THE CANADIAN SUPERCRITICAL WATER-COOLED REACTOR

2019 ◽  
Vol 8 (1) ◽  
pp. 1-8
Author(s):  
Lee Brian Gardner ◽  
Don K Ryland ◽  
Sam Suppiah
Keyword(s):  
Supercritical Water ◽  
Nuclear Reactor ◽  
Hydrogen Generation ◽  
Nuclear Reactors ◽  
Operating Conditions ◽  
Severe Accident ◽  
Systematic Assessment ◽  
Management Concept ◽  
Accident Conditions ◽  
Supercritical Water Cooled Reactor

Accidental hydrogen production in nuclear reactors has been a significant focus of nuclear reactor safety for decades. However, since the accident at Fukushima Daiichi nuclear generating station, hydrogen safety in nuclear reactors is a more relevant topic. As new reactor concepts, such as the supercritical water-cooled reactor (SCWR), are designed and developed the risk of unintentional hydrogen generation is not eliminated; however, it can be mitigated in the design. A systematic assessment of the hydrogen risk from both normal and accident conditions in the Canadian SCWR design was performed, in which various techniques to mitigate the hydrogen combustion potential were considered. While the rate of hydrogen generation under normal operating conditions was found to be low when held at supercritical water conditions, conservative estimates suggest that a significant quantity of hydrogen may be produced and released to the containment building in a severe accident. As a result, a hydrogen–oxygen management concept has been proposed to mitigate the hydrogen produced in a severe accident that includes a nitrogen-inerted containment building to reduce the combustion potential of hydrogen and the installation of passive autocatalytic recombiners for oxygen management. This hydrogen–oxygen management concept results in significant design changes and likely significant economic and operational impacts on the Canadian SCWR design.

Fuel Assembly Design for Supercritical Water-Cooled Reactor

2015 ◽  
Vol 2 (1) ◽  
Author(s):  
Feng Linna ◽  
Zhu Fawen
Keyword(s):  
Fuel Assembly ◽  
Supercritical Water ◽  
Nuclear Reactor ◽  
Protective Coatings ◽  
Assembly Design ◽  
Water Reactors ◽  
And Performance ◽  
Materials Used ◽  
Degradation Of Materials ◽  
Supercritical Water Cooled Reactor

The supercritical water-cooled reactor (SWCR) has been selected as one of the most promising reactors for Generation IV nuclear reactors due to its higher thermal efficiency and more simplified structure compared to the state-of-the-art light water reactors (LWRs). However, there are a large number of potential problems that must be addressed, particularly the fuel assembly design of the SCWR. SCWRs are a kind of high-temperature, high-pressure, water-cooled reactor that operates above the thermodynamic critical point of water (374°C, 22.1 MPa). Corrosion and degradation of materials used in supercritical water environments are determined by several environment- and material-dependent factors. In particular, irradiation-induced changes in microstructure and microchemistry are major concerns in a nuclear reactor. Many structural materials including alloys and ceramics have been proposed for use as SCWR components or materials for applying protective coatings in SCWRs. In this paper, the present status of supercritical fuel assembly design at home and abroad is reported. According to the special requirements of supercritical core design, a kind of configuration design of fuel assembly with two-flow core and using SiC as cladding material are proposed. The analysis results have shown that the design basically meets the requirements of fuel assembly design, which has good feasibility and performance.

The Measurement of Light Element Segregation Using EDS and EELS

1998 ◽  
Vol 4 (S2) ◽  
pp. 772-773
Author(s):  
J.T. Busby ◽  
E.A. Kenik ◽  
G.S. Was
Keyword(s):  
Grain Boundaries ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Light Element ◽  
Free Surfaces ◽  
Spatial Redistribution ◽  
Core Components ◽  
Radiation Induced ◽  
Potential Interactions ◽  
Impurity Elements

Radiation-induced segregation (RIS) is the spatial redistribution of elements at defect sinks such as grain boundaries and free surfaces during irradiation. This phenomenon has been studied in a wide variety of alloys and has been linked to irradiation-assisted stress corrosion cracking (IASCC) of nuclear reactor core components. However, several recent studies have shown that Cr and Mo can be enriched to significant levels at grain boundaries prior to irradiation as a result of heat treatment. Segregation of this type may delay the onset of radiation-induced Cr depletion at the grain boundary, thus reducing IASCC susceptibility. Unfortunately, existing models of segregation phenomena do not correctly describe the physical processes and therefore are grossly inaccurate in predicting pre-existing segregation and subsequent redistribution during irradiation. Disagreement between existing models and measurement has been linked to potential interactions between the major alloying elements and lighter impurity elements such as S, P, and B.

scholarly journals Investigation of Corrosion Resistance of Alloys with Potential Application in Supercritical Water-cooled Nuclear Reactors

2019 ◽  
Vol 63 (2) ◽  
pp. 328-332 ◽  
Author(s):  
Ákos Horváth ◽  
Attila R. Imre ◽  
György Jákli
Keyword(s):  
Corrosion Resistance ◽  
Supercritical Water ◽  
Power Plants ◽  
Fuel Cladding ◽  
Pressurized Water ◽  
Water Reactors ◽  
Reactor Types ◽  
Materials Used ◽  
Supercritical Water Cooled Reactor ◽  
Nuclear Applications

The Supercritical Water Cooled Reactor (SCWR) is one of the Generation IV reactor types, which has improved safety and economics, compared to the present fleet of pressurized water reactors. For nuclear applications, most of the traditional materials used for power plants are not applicable, therefore new types of materials have to be developed. For this purpose corrosion tests were designed and performed in a supercritical pressure autoclave in order to get data for the design of an in-pile high temperature and high-pressure corrosion loop. Here, we are presenting some results, related to corrosion resistance of some potential structural and fuel cladding materials.

SIMULASI KARAKTERISTIK ALIRAN DAN SUHU FLUIDA PENDINGIN (H2O) PADA TERAS REAKTOR NUKLIR SMR (SMALL MODULAR REACTOR)

ROTASI ◽  
2013 ◽  
Vol 15 (4) ◽  
pp. 33
Author(s):  
Anwar Ilmar Ramadhan ◽  
Indra Setiawan ◽  
M. Ivan Satryo
Keyword(s):  
Fluid Flow ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Heat Removal ◽  
Axial Direction ◽  
Operating Conditions ◽  
Fluid Temperature ◽  
Wide Range ◽  
Fuel Reactor ◽  
Performance Of Heat Transfer

Safety is an issue that is of considerable concern in the design, operation and development of a nuclear reactor. Therefore, the method of analysis used in all these activities should be thorough and reliable so as to predict a wide range of operating conditions of the reactor, both under normal operating conditions and in the event of an accident. Performance of heat transfer to the cooling of nuclear fuel, reactor safety is key. Poor heat removal performance would threaten the integrity of the fuel cladding which could further impact on the release of radioactive substances into the environment in an uncontrolled manner to endanger the safety of the reactor workers, the general public, and the environment. This study has the objective is to know is profile contour of fluid flow and the temperature distribution pattern of the cooling fluid is water (H2O) in convection in to SMR reactor with fuel sub reed arrangement of hexagonal in forced convection. In this study will be conducted simulations on the SMR reactor core used sub channel hexagonal using CFD (Computational Fluid Dynamics) code. And the results of this simulation look more upward (vector of fluid flow) fluid temperature will be warm because the heat moves from the wall to the fluid heater. Axial direction and also looks more fluid away from the heating element temperature will be lower.

Formation of Local, Transient “Acid Spikes” in the Fast Neutron Radiolysis of Supercritical Water at 400 °C: A Potential Source of Corrosion in Supercritical Water-Cooled Reactors?

2020 ◽  
Vol 6 (3) ◽  
Author(s):  
Md Mohsin Patwary ◽  
Sunuchakan Sanguanmith ◽  
Jintana Meesungnoen ◽  
Jean-Paul Jay-Gerin
Keyword(s):  
Fast Neutron ◽  
Supercritical Water ◽  
Reactor Core ◽  
Radiation Chemistry ◽  
Operating Conditions ◽  
Monte Carlo Simulation Study ◽  
Proper Understanding ◽  
Gen Iv ◽  
Reactor Operating

Abstract The use of supercritical water (SCW) in GEN IV reactors is a logical approach to the ongoing development of nuclear energy. A proper understanding of the radiation chemistry and reactivities of transients in a reactor core under SCW conditions is required to achieve optimal water chemistry control and safety. A Monte Carlo simulation study of the radiolysis of SCW at 400 °C by incident 2 MeV monoenergetic neutrons (taken as representative of a fast neutron flux in a reactor) was carried out as a function of water density between ∼150 and 600 kg/m3. The in situ formation of H3O+ by the generated recoil protons was shown to render the “native” track regions temporarily very acidic (pH ∼ 1). This acidity, though local and transitory (“acid spikes”), raises the question whether it may promote a corrosive environment under proposed SCW-cooled reactor operating conditions that would lead to progressive degradation of reactor components.

Thermal Evaluations for Next Generation Nuclear Plant Fuel Testing

Volume 4 ◽  
2004 ◽  
Author(s):  
Richard G. Ambrosek ◽  
Debbie J. Utterbeck ◽  
Brandon Miller
Keyword(s):  
Nuclear Reactor ◽  
Fuel Cycle ◽  
Nuclear Plant ◽  
Light Water Reactors ◽  
Next Generation ◽  
Preliminary Information ◽  
Advanced Fuel ◽  
Water Reactors ◽  
Test Reactor ◽  
Test Fuel

The DOE Advanced Fuel Cycle Initiative and Generation IV reactor programs are developing new fuel types for use in the current Light Water Reactors and future advanced reactor concepts. The Advanced Gas Reactor program is planning to test fuel to be used in the Next Generation Nuclear Plant (NGNP) nuclear reactor. Preliminary information for assessing performance of the fuel will be obtained from irradiations performed in the Advanced Test Reactor large “B” experimental facility.

Study of the Influence of Microstructure and Intergranular Carbides on the Oxidation Behavior of a Nickel Base Alloy 690 TT in Supercritical Water Nuclear Reactor Conditions

2020 ◽  
Author(s):  
Alberto Sáez-Maderuelo ◽  
María Luisa Ruiz-Lorenzo ◽  
Francisco Javier Perosanz ◽  
Patricie Halodová ◽  
Jan Prochazka ◽  
...  
Keyword(s):  
Supercritical Water ◽  
Nuclear Power ◽  
Power Plants ◽  
Oxidation Behavior ◽  
Nuclear Reactor ◽  
Base Alloy ◽  
Nuclear Industry ◽  
Operating Conditions ◽  
Alloy 690 ◽  
Reactor Operating

Abstract Alloy 690, which was designed as a replacement for the Alloy 600, is widely used in the nuclear industry due to its optimum behavior to stress corrosion cracking (SCC) under nuclear reactor operating conditions. Because of this superior resistance, alloy 690 has been proposed as a candidate structural material for the Supercritical Water Reactor (SCWR), which is one of the designs of the next generation of nuclear power plants (Gen IV). In spite of this, striking results were found [1] when alloy 690 was tested without intergranular carbides. These results showed that, contrary to expectations, the crack growth rate is lower in samples without intergranular carbides than in samples with intergranular carbides. Therefore, the role of the carbides in the corrosion behavior of Alloy 690 is not yet well understood. Considering these observations, the aim of this work is to study the effect of intergranular carbides in the oxidation behavior (as a preliminary stage of degenerative processes SCC) of Alloy 690 in supercritical water (SCW) at two temperatures: 400 °C and 500 °C and 25 MPa. Oxide layers of selected specimens were studied by different techniques like Scanning Electron Microscope (SEM) and Auger Electron Spectroscopy (AES).

Predictive Modeling of Acoustic Signals From Thermoacoustic Power Sensors (TAPS)

2016 ◽  
Author(s):  
Christopher M. Dumm ◽  
Jeffrey S. Vipperman ◽  
Jorge V. Carvajal ◽  
Melissa M. Walter ◽  
Luke Czerniak ◽  
...  
Keyword(s):  
Predictive Modeling ◽  
Acoustic Waves ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Signal Propagation ◽  
Acoustic Signals ◽  
Operating Conditions ◽  
Liquid Sodium ◽  
Thermoacoustic Engines ◽  
Self Powered

Thermoacoustic Power Sensor (TAPS) technology offers the potential for self-powered, wireless measurement of nuclear reactor core operating conditions. TAPS are based on thermoacoustic engines, which harness thermal energy from fission reactions to generate acoustic waves by virtue of gas motion through a porous stack of thermally nonconductive material. TAPS can be placed in the core, where they generate acoustic waves whose frequency and amplitude are proportional to the local temperature and radiation flux, respectively. TAPS acoustic signals are not measured directly at the TAPS; rather, they propagate wirelessly from an individual TAPS through the reactor, and ultimately to a low-power receiver network on the vessel’s exterior. In order to rely on TAPS as primary instrumentation, reactor-specific models which account for geometric/acoustic complexities in the signal propagation environment must be used to predict the amplitude and frequency of TAPS signals at receiver locations. The reactor state may then be derived by comparing receiver signals to the reference levels established by predictive modeling. In this paper, we develop and experimentally benchmark a methodology for predictive modeling of the signals generated by a TAPS system, with the intent of subsequently extending these efforts to modeling of TAPS in a liquid sodium environment.

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