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

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.

Choices

1997 ◽  
Author(s):  
Robert Pool
Keyword(s):  
Nuclear Power ◽  
Nuclear Reactor ◽  
Liquid Metals ◽  
Reactor Core ◽  
Electrical Energy ◽  
Natural Uranium ◽  
Nuclear Industry ◽  
Fuel Oil ◽  
Organic Liquids ◽  
Speed Up

During the 1940s and early 1950s, when atomic energy was new, it was common to hear reactors described as nuclear “furnaces.” They “burned” their nuclear fuel and left behind nuclear “ash.” Technically, of course, none of these terms made sense, since burning is a chemical process and a reactor gets its energy from fission, but journalists liked the terminology because it was easy and quick. One loaded fuel into the reactor, flipped a switch, and things got very hot. If that wasn’t exactly a furnace, it was close enough. And actually, the metaphor was pretty good—up to a point. The basement furnace burns one of several different fuels: natural gas or fuel oil or even, in some ancient models, coal. Nuclear reactors can be built to use plutonium, natural uranium, or uranium that has been enriched to varying degrees. Home furnaces have a “coolant”—the air that is circulated through the furnace and out through the rest of the house, carrying heat away from the fire. Reactors have a coolant, too—the liquid or gas that carries heat away from the reactor core to another part of the plant, where heat energy is transformed into electrical energy. There, however, the metaphor sputters out. In a nuclear reactor, the coolant not only transfers heat to a steam generator or a turbine, but it also keeps the fuel from overheating. The coolant in a furnace does nothing of the sort. And most reactors use a moderator to speed up the fission reaction. The basement burner has nothing similar. But the most importance weakness of the furnace metaphor is that it obscured just how many varieties of reactors were possible—and, consequently, obscured the difficult choice facing the early nuclear industry: Which reactor type should become the basis for commercial nuclear power? The possibilities were practically unlimited. The fuel selection was wide. The coolant could be nearly anything that has good heat-transfer properties: air, carbon dioxide, helium, water, liquid metals, organic liquids, and so on.

FAVOR Version 16.1: A Computer Code for Fracture Mechanics Analyses of Nuclear Reactor Pressure Vessels

2017 ◽  
Author(s):  
B. Richard Bass ◽  
Paul T. Williams ◽  
Terry L. Dickson ◽  
Hilda B. Klasky
Keyword(s):  
Fracture Mechanics ◽  
Nuclear Reactor ◽  
Computer Code ◽  
The United States ◽  
Pressure Vessels ◽  
Nuclear Industry ◽  
Current Status ◽  
Pressurized Water ◽  
Oak Ridge ◽  
Reactor Pressure

This paper describes the current status of the Fracture Analysis of Vessels, Oak Ridge (FAVOR) computer code which has been under development at Oak Ridge National Laboratory (ORNL), with funding by the United States Nuclear Regulatory Commission (NRC), for over twenty-five years. Including this most recent release, v16.1, FAVOR has been applied by analysts from the nuclear industry and regulators at the NRC to perform deterministic and probabilistic fracture mechanics analyses to review / assess / update regulations designed to insure that the structural integrity of aging, and increasingly embrittled, nuclear reactor pressure vessels (RPVs) is maintained throughout the vessel’s operational service life. Early releases of FAVOR were developed primarily to address the pressurized thermal shock (PTS) issue; therefore, they were limited to applications involving pressurized water reactors (PWRs) subjected to cool-down transients with thermal and pressure loading applied to the inner surface of the RPV wall. These early versions of FAVOR were applied in the PTS Re-evaluation Project to successfully establish a technical foundation that served to better inform the basis of the then-existent PTS regulations to the original PTS Rule (Title 10 of the Code of Federal Regulations, Chapter I, Part 50, Section 50.61, 10CFR 50.61). A later version of FAVOR resulting from this project (version 06.1 - released in 2006) played a major role in the development of the Alternative PTS Rule (10 CFR 50.61.a). This paper describes recent ORNL developments of the FAVOR code; a brief history of verification studies of the code is also included. The 12.1 version (released in 2012) of FAVOR represented a significant generalization over previous releases insofar as it included the ability to encompass a broader range of transients (heat-up and cool-down) and vessel geometries, addressing both PWR and boiling water reactor (BWR) RPVs. The most recent public release of FAVOR, v16.1, includes improvements in the consistency and accuracy of the calculation of fracture mechanics stress-intensity factors for internal surface-breaking flaws; special attention was given to the analysis of shallow flaws. Those improvements were realized in part through implementation of the ASME Section XI, Appendix A, A-3000 curve fits into FAVOR; an overview of the implementation of those ASME curve fits is provided herein. Recent results from an extensive verification benchmarking project are presented that focus on comparisons of solutions from FAVOR versions 16.1 and 12.1 referenced to baseline solutions generated with the commercial ABAQUS code. The verifications studies presented herein indicate that solutions from FAVOR v16.1 exhibit an improvement in predictive accuracy relative to FAVOR v12.1, particularly for shallow flaws.

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

KnE Energy ◽  
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>

scholarly journals Powering Down

2003 ◽  
Vol 125 (04) ◽  
pp. 46-48
Author(s):  
Harry Hutchinson
Keyword(s):  
Heat Transfer ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
The United States ◽  
Nuclear Industry ◽  
Significant Feature ◽  
Pressurized Water ◽  
Safety Testing ◽  
Fuel Assemblies ◽  
Enriched Uranium

This article reviews that after a half century of safety testing for the nuclear industry, a key heat-transfer lab is losing its home. Columbia University’s Heat Transfer Research Facility has been the only place to go for key safety testing. Since the days of the Atoms for Peace program during the Eisenhower years, the lab has tested generations of nuclear reactor fuel assemblies. The lab’s clients over the years have included all the designers of pressurized water reactors in the United States and others from much of the world. The tests are primarily concerned with one small, but significant feature of a reactor core. A core contains as many as 3000 fuel assemblies, bundles of long, slender rods containing enriched uranium. Controlled fission among the bundles heats water to begin the series of heat-transfer cycles that send steam to the turbines that will drive generators.

scholarly journals AI-based design of a nuclear reactor core

2021 ◽  
Vol 11 (1) ◽  
Author(s):  
Vladimir Sobes ◽  
Briana Hiscox ◽  
Emilian Popov ◽  
Rick Archibald ◽  
Cory Hauck ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Rapid Development ◽  
Reactor Core ◽  
Systems Design ◽  
Axial Direction ◽  
Arbitrary Geometry ◽  
Oak Ridge ◽  
Geometry Design ◽  
Nuclear Systems ◽  
Nuclear Reactor Core

AbstractThe authors developed an artificial intelligence (AI)-based algorithm for the design and optimization of a nuclear reactor core based on a flexible geometry and demonstrated a 3× improvement in the selected performance metric: temperature peaking factor. The rapid development of advanced, and specifically, additive manufacturing (3-D printing) and its introduction into advanced nuclear core design through the Transformational Challenge Reactor program have presented the opportunity to explore the arbitrary geometry design of nuclear-heated structures. The primary challenge is that the arbitrary geometry design space is vast and requires the computational evaluation of many candidate designs, and the multiphysics simulation of nuclear systems is very time-intensive. Therefore, the authors developed a machine learning-based multiphysics emulator and evaluated thousands of candidate geometries on Summit, Oak Ridge National Laboratory’s leadership class supercomputer. The results presented in this work demonstrate temperature distribution smoothing in a nuclear reactor core through the manipulation of the geometry, which is traditionally achieved in light water reactors through variable assembly loading in the axial direction and fuel shuffling during refueling in the radial direction. The conclusions discuss the future implications for nuclear systems design with arbitrary geometry and the potential for AI-based autonomous design algorithms.

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

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.

A Review on Specific Features of Small and Medium Sized Nuclear Power Plants

2010 ◽  
Author(s):  
Salah Ud-din Khan ◽  
Minjun Peng ◽  
Muhammad Zubair ◽  
Shaowu Wang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Nuclear Reactors ◽  
Fixed Bed ◽  
Nuclear Industry ◽  
Energy Agency ◽  
Pressurized Water ◽  
Water Reactor ◽  
Energy Supply System

Due to global warming and high oil prices nuclear power is the most feasible solution for generating electricity. For the fledging nuclear power industry small and medium sized nuclear reactors (SMR’s) are instrumental for the development and demonstration of nuclear reactor technology. Due to the enhanced and outstanding safety features, these reactors have been considered globally. In this paper, first we have summarized the reactor design by considering some of the large nuclear reactor including advanced and theoretical nuclear reactor. Secondly, comparison between large nuclear reactors and SMR’s have been discussed under the criteria led by International Atomic Energy Agency (IAEA). Thirdly, a brief review about the design and safety aspects of some of SMR’s have been carried out. We have considered the specifications and parametric analysis of the reactors like: ABV which is the floating type integral Pressurized water reactor; Long life, Safe, Simple Small Portable Proliferation Resistance Reactor (LSPR) concept; Multi-Application Small Light Water Reactor (MASLWR) concept; Fixed Bed Nuclear Reactor (FBNR); Marine Reactor (MR-X) & Deep Sea Reactor (DR-X); Space Reactor (SP-100); Passive Safe Small Reactor for Distributed energy supply system (PSRD); System integrated Modular Advanced Reactor (SMART); Super, Safe, Small and Simple Reactor (4S); International Reactor Innovative and Secure (IRIS); Nu-Scale Reactor; Next generation nuclear power plant (NGNP); Small, Secure Transportable Autonomous Reactor (SSTAR); Power Reactor Inherently Safe Module (PRISM) and Hyperion Reactor concept. Finally we have point out some challenges that must be resolved in order to play an effective role in Nuclear industry.

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

2017 ◽  
Vol 18 (2) ◽  
pp. 93
Author(s):  
Julwan Hendry Purba
Keyword(s):  
Neutron Irradiation ◽  
Nuclear Reactor ◽  
Research Reactor ◽  
Reactor Core ◽  
Structural Materials ◽  
Industrial Applications ◽  
Physical Ageing ◽  
Functional Capabilities ◽  
Wide Range ◽  
The Stability

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.

Preliminary Evaluation of the Fixed Bed Nuclear Reactor Concept Using IAEA-INPRO Methodology

2004 ◽  
Author(s):  
Farhang Sefidvash
Keyword(s):  
Nuclear Reactor ◽  
Nuclear Reactors ◽  
Fixed Bed ◽  
Reactor Core ◽  
Primary Circuit ◽  
Preliminary Evaluation ◽  
Spent Fuel ◽  
Energy Agency ◽  
Reactor Technology ◽  
Fuel Elements

The Atomic Energy Agency (IAEA) through its INPRO Project has developed a methodology to evaluate the innovative nuclear reactors. The main objectives of INPRO are to help to ensure that nuclear energy is available in the 21st century in a sustainable manner; and bring together both technology holders and technology users to achieve desired innovations in nuclear reactors and nuclear fuel cycles which are to be acceptable to the public because they are economic, safe, proliferation resistant, sustainable, and having reduced environmental impact. Here is a preliminary application of this methodology (IAEA-TECDOC-1362) to evaluate the Fixed Bed Nuclear Reactor Concept (FBNR). Some of the characteristics of the proposed reactor are: The FBNR is based on pressurized light water reactor technology. It is a small, modular, and integrated primary circuit reactor. The fuel elements of FBNR are 8 mm diameter spherical uranium dioxide pellets cladded by zircaloy or made of compacted TRISO type fuel particles. The reactor core is suspended by the flow of water coolant. The stop in flow causes the fuel elements leave the reactor core by the force of gravity and fall into a passively cooled fuel chamber or even leave the reactor completely and become deposited in the spent fuel pool. It is an inherently safe and passively cooled reactor concept. FBNR in its advanced versions can use supercritical steam or helium gas as coolant, and utilize MOX or thorium fuel.

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