Physical Ageing of The Research Reactor Core Structural Materials Due To Neutron Irradiation Exposure: A Review

A research reactor (RR) is a nuclear reactor that has function to generate and utilize neutron flux and radiation ionization for research purposes and industrial applications. More than 60% of current operating RRs have been operated for 30 years or more. As the time passes, the functional capabilities of structures, systems and components (SSCs) of those RRs deteriorate by physical ageing, which can be caused by neutron irradiation exposure such as irradiation induced dislocation and microstructural changes. To extend the lifetime and/or to avoid unplanned outages, ageing on the safety related SSCs of RRs need to be properly managed. An ageing management is a strategy to engineer, operate, maintenance, and control SSC degradation within acceptablelimits. The purpose of this study is to review physical ageing of the core structural materials of the RRs caused by neutron irradiation exposure. In order to achieve this objective, a wide range of literatures are reviewed. Comprehensive discussions on irradiation behaviors are limited only on reactor vessel and core support structure materials made from zirconium and beryllium as well as their alloys, which are widely used in RRs. It is found that the stability of the mechanical properties of zirconium and beryllium as well as their alloys was mostly affected by the neutron fluences and temperatures.

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HIGH TEMPERATURE HYBRID NANODIAMOND SENSOR SYSTEMS

Additional Conferences (Device Packaging HiTEC HiTEN & CICMT) ◽

10.4071/hitec-ta23 ◽

2014 ◽

Vol 2014 (HITEC) ◽

pp. 000034-000039 ◽

Cited By ~ 1

Author(s):

John R. Fraley ◽

Lauren Kegley ◽

Stephen Minden ◽

Jimmy L. Davidson ◽

David Kerns

Keyword(s):

High Temperature ◽

Nuclear Reactor ◽

Reactor Core ◽

Chemical Species ◽

Chemical Vapor ◽

Smart Sensors ◽

Sensor Systems ◽

High Temperature Electronics ◽

Wireless Telemetry ◽

Wide Range

In recent years, high temperature semiconductors have been utilized in wireless telemetry systems for use in military and commercial applications, wherein a high temperature environment combined with other factors such as rotating machinery or weight-constraints preclude the use of conventional silicon based wireless telemetry or wired sensor solutions. Present systems include those which can measure temperatures, pressures, vibrations, and strains. By combining the advanced electronics developed for these systems with novel sensor elements created using chemical vapor deposition (CVD) nanodiamond technology, a wide range of other high temperature sensing systems can be enabled. The unique properties of the diamond sensors have proven in principle the capability to sense, with quantifiable signal, a wide variety of parameters under extreme conditions including very high temperatures and pressures. It has been clear for some time that diamond would be the ideal material of choice for solid-state sensors, but only in recent years has the advent of CVD diamond (as opposed to natural or HPHT [high pressure, high temperature] formation) opened the door for its practical development into harsh environment sensor systems. By combining these diamond sensor elements with high temperature electronics and high temperature packaging approaches, smart sensors can be developed to measure parameters ranging from gas chemical species on the surface of Venus, to neutron flux rates outside of a nuclear reactor core. The research presented here is centered around the use of hybrid diamond sensors for neutron detection applications in Nuclear Thermal Propulsion systems. The current technology state and development needs for these hybrid high temperature diamond smart sensors will be highlighted to potentially encourage future R&D from the high-temperature electronics community.

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

ROTASI ◽

10.14710/rotasi.15.4.33-40 ◽

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.

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Immobilization of Candida antarctica B (CALB) in Silica Aerogel: Morphological Characteristics and Stability

Biointerface Research in Applied Chemistry ◽

10.33263/briac106.67446756 ◽

2020 ◽

Vol 10 (6) ◽

pp. 6744-6756 ◽

Cited By ~ 1

Keyword(s):

Sol Gel ◽

Morphological Characteristics ◽

Amorphous Structure ◽

Industrial Applications ◽

Candida Antarctica ◽

Type Iv ◽

Wide Range ◽

The Stability ◽

Immobilized Calb ◽

Candida Antarctica B

Enzymes have been extensively used due to their catalytic properties, and immobilization is a promising technique to enhance their catalytic activity and stability. Lipases are enzymes naturally efficient, can be employed for the production of many different molecules, and have a wide range of industrial applications thanks to their broad selectivity. The objective of the present study was to characterize the Candida antarctica B CALB immobilized obtained using the aerogel technique regarding the morphological characteristics of the aerogel silica and its stability. For this purpose, analyzes of XRD, adsorption-desorption isotherms, TGA, SEM, and stability (storage, operational, and thermal) were performed. The supports obtained have an amorphous structure and isotherm type IV. Regarding TGA, two distinct regions were obtained and studied. Aerogels showed an increase in thermal, storage, and operational stability in relation to the free enzyme and demonstrated between 8 and 12 cycles of reuse. The contribution of this work was to present the stability advantages of the immobilized CALB enzyme through the sol-gel technique.

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Reevaluation of Spectral Parameters and Neutron Fluxes in IC-7 Irradiation Channel of TRIGA MARK I IPR-R1 Research Nuclear Reactor

Journal of Nuclear Engineering and Radiation Science ◽

10.1115/1.4051249 ◽

2021 ◽

Author(s):

Radojko Jacimovic ◽

Maria Angela de Barros Correia Menezes

Keyword(s):

Nuclear Reactor ◽

Epithermal Neutron ◽

Research Reactor ◽

Nuclear Research ◽

Irradiation Channel ◽

Spectral Parameters ◽

K0 Method ◽

Nuclear Research Reactor ◽

Research Nuclear Reactor ◽

The Stability

Abstract The core configuration of the TRIGA MARK I IPR-R1 nuclear research reactor, Brazil, has been modified six times since the first criticality and the neutron fluxes have been determined using experimental and semi theoretical methodologies determining the neutron fluxes in different irradiation channels and devices, applying different procedures and materials. This reactor operates at 100 kW, however, after new configuration for 250 kW in 2001, the carousel no longer rotates during irradiations aiming at preserving the rotation mechanism. In 2003, the spectral parameters were determined experimentally by the "Cd-ratio for multi-monitor" in five specific channels aiming at the application of NAA k0-standardized method. The determinations were repeated applying the same procedure in 2016, 2018 and 2019. Values for thermal and epithermal neutron fluxes as well as f and a spectral parameters were determined. The experimental results for CRM BCR-320R were calculated by the k0-method of NAA, using the spectral parameters for irradiation channel IC-7 obtained in 2003, 2016, 2018 and 2019 and evaluated by En-score. The values showed that the differences in the results compared to those in 2003 were lower than 2.5%, inside the uncertainty of the method. It shows that the k0-method installed in CDTN is reliable and useful for various purposes. The results of the spectral parameter f presented small differences, in a period of 16 years, pointing out the stability of operation of the reactor TRIGA MARK I IPR-R1.

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Microstructural Changes of Mechanical Properties of 08Cr18Ni10Ti Austenitic Steel under Neutron Irradiation

Key Engineering Materials ◽

10.4028/www.scientific.net/kem.743.37 ◽

2017 ◽

Vol 743 ◽

pp. 37-40

Author(s):

Daniyar Esimgaliuly Ashimov ◽

Erbolat Taitoleuovich Koyanbayev ◽

Sherzod Rustambekovich Kurbanbekov ◽

Alexander Andreevich Sitnikov ◽

Maznyn Kanapinovich Skakov

Keyword(s):

Mechanical Properties ◽

Radiation Dose ◽

Austenitic Steel ◽

Neutron Irradiation ◽

Research Reactor ◽

Structural Materials ◽

Microstructural Changes ◽

Material Analysis ◽

Hardness Values

This paper reports neutron irradiation induced effects on 08Cr18Ni10Ti austenitic steel, which is one of the structural materials of the IVG.1M research reactor. Chosen samples were stored in the beryllium displacer at two different elevation heights and the radiation dose with fast fluence of 0.4·1019 n/cm2 and 4.3·1019 n/cm2. Thorough material analysis reports changes in hardness values, plasticity and microstructure

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Effect Of U-9mo/Al Fuel Densities On Neutronic And Steady State Thermal Hydraulic Parameters Of Mtr Type Research Reactor

KnE Energy ◽

10.18502/ken.v1i1.456 ◽

2016 ◽

Vol 1 (1) ◽

Author(s):

Surian PINEM

Keyword(s):

Heat Flux ◽

Steady State ◽

Nuclear Reactor ◽

Research Reactor ◽

Research Work ◽

Reactor Core ◽

Hydraulic Parameters ◽

Limit Value ◽

Maximum Temperatures ◽

Coolant Velocity

<p>The objectives of this research work are to carry out a detailed neutronic and steady state thermal hydraulics analysis for a MTR research reactor fuelled with the low enrichment U-9Mo/Al dispersion fuels of various uranium densities. The high density uranium fuel will increase the cycle length of the reactor operation and the heat flux in the reactor core. The increasing heat flux at the fuel will causing increase the temperature of the fuel and cladding so that the coolant velocity has to be increased. However, the coolant velocity in the fuel element has a limit value due to the thermal hydraulic stability considerations in the core. Therefore, the neutronic and the steady state thermal hydraulic analysis are important in the design and operation of nuclear reactor safety. The calculations were performed using WIMS-D5 and MTRDYN codes. The WIMS-D5 code used for generating the group constants of all core materials as well as the neutronic and steady state thermal hydraulic parameters were determined by using the MTRDYN code. The calculation results showed that the excess reactivity increases as the uranium density increases since the mass of fuel in the reactor core is increased. Using the critical velocity concept, the maximum coolant velocity at fuel channel is 11.497 m/s. The maximum temperatures of the coolant, cladding and fuel meat with the uranium density of 3,66 g/cc are 70.85°C, 150.79°C and 153.24°C, respectively. The maximum temperatures are fulfilled the design limit so reactor has a safe operation at the nominal power.</p>

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Comparison of Different Fuel Compositions for a Research Reactor with Thermal Power of up to 10 MW

KnE Engineering ◽

10.18502/keg.v3i3.1604 ◽

2018 ◽

Vol 3 (3) ◽

pp. 34

Author(s):

Yulia Levchenko ◽

Alexander Zevyakin ◽

Yulia Karazhielievskaia ◽

Anna Terekhova

Keyword(s):

Nuclear Power ◽

Nuclear Reactor ◽

Research Reactor ◽

Thermal Power ◽

Reactor Core ◽

Chain Reaction ◽

Extensive Experience ◽

Study Characteristics ◽

Domestic Power ◽

Carbide Fuel

In this article the prospect of using carbide fuel in a research reactor for export to countries with developing nuclear power is consider. The choice of a fuel composition for a research reactor is an important part in substantiating of the neutron-physical and economic characteristics of a reactor facility, and is also an important part of the control-dependent self-sustaining fission chain reaction in a nuclear reactor that affects the specifics of management. For reducing the economic component in the design of this reactor core of the research reactor, structural materials and design solutions are used that have extensive experience in domestic power engineering. In this work UO2-ThO2 and PuO2-ThO2 was selected as the considered fuel compositions. In the course of the study, characteristics were obtained for a burnup of the fuel compositions under study, the initial reserve of reactivity and the duration of the fuel campaign.

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Assessment of the Integrity of the Channel Shelter of the Experimental Nuclear Reactor LVR 15

Volume 3: Design and Analysis ◽

10.1115/pvp2009-77306 ◽

2009 ◽

Author(s):

Miroslav Svrcek ◽

Ivan Krasny ◽

Jan Lestina

Keyword(s):

Nuclear Reactor ◽

Reactor Core ◽

Nuclear Research ◽

Point Of View ◽

Internal Diameter ◽

Neutron Absorption ◽

Nuclear Research Institute ◽

Channel Water ◽

Channel Tube ◽

The Stability

The experimental reactor LVR 15 is operated in Nuclear Research Institute Rez. A pressurized experimental channel forming a field tube is installed in the reactor core. The operating parameters of the channel water are 16MPa and 290°C. From the point of view of operation (neutron absorption) the channel is located in a quadratic shelter 68×68mm with internal diameter 65mm (see Fig. 2). Based on the recommendation of IAEA INSAR mission, the integrity of the shelter must be conserved if a through wall crack in the outer channel tube is postulated, moreover the stability of the through wall crack must be guaranteed. Solutions of both situations are presented in the paper.

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Assessment of radiological consequence of a hypothetical accident at the Ghana Research Reactor-1 facility based on terrorist attack

Science Progress ◽

10.1177/00368504211054986 ◽

2021 ◽

Vol 104 (4) ◽

pp. 003685042110549

Author(s):

Henry K. Obeng ◽

Sylvester A. Birikorang ◽

Kwame Gyamfi ◽

Simon Adu ◽

Andrew Nyamful

Keyword(s):

Nuclear Reactor ◽

Research Reactor ◽

Terrorist Attack ◽

Reactor Core ◽

Nuclear Industry ◽

Air Concentration ◽

Energy Agency ◽

Oak Ridge ◽

Effective Doses ◽

Ground Deposition

The International Atomic Energy Agency defines a nuclear and radiation accident as an occurrence that leads to the release of radiation causing significant consequences to people, the environment, or the facility. During such an event involving a nuclear reactor, the reactor core is a critical component which when damaged, will lead to the release of significant amounts of radionuclides. Assessment of the radiation effect that emanates from reactor accidents is very paramount when it comes to the safety of people and the environment; whether or not the released radiation causes an exposure rate above the recommended threshold nuclear reactor safety. During safety analysis in the nuclear industry, radiological accident analyses are usually carried out based on hypothetical scenarios. Such assessments mostly define the effect associated with the accident and when and how to apply the appropriate safety measures. In this study, a typical radiological assessment was carried out on the Ghana Research Reactor-1. The study considered the available reactor core inventory, released radionuclides, radiation doses and detailed process of achieving all the aforementioned parameters. Oak Ridge isotope generation-2 was used for core inventory calculations and Hotspot 3.01 was also used to model radionuclides dispersion trajectory and calculate the released doses. Some of the radionuclides that were considered include I-131, Sr-90, Cs-137, and Xe-137. Total effective doses equivalent to released radionuclides, the ground deposition activity and the respiratory time-integrated air concentration were estimated. The maximum total effective doses equivalent value of 5.6 × 10−9 Sv was estimated to occur at 0.1 km from the point of release. The maximum ground deposition activity was estimated to be 2.5 × 10−3 kBq/m3 at a distance of 0.1 km from the release point. All the estimated values were found to be far below the annual regulatory limits of 1 mSv for the general public as stated in IAEA BSS GSR part 3.

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