scholarly journals Neutronic Analysis of The SMART Modular Reactor Fuel Using SRAC 2006

2021 ◽  
Vol 4 (2) ◽  
pp. 60
Author(s):  
N. Nailatussaadah ◽  
I. Irsyad
Keyword(s):  
Fuel Assembly ◽  
Nuclear Power ◽  
Power Plants ◽  
Electrical Energy ◽  
Nuclear Data ◽  
Conversion Ratio ◽  
Reactor Fuel ◽  
Advanced Technology ◽  
Operating Cycle ◽  
Nuclear Data Library

Neutronic analysis of The SMART modular reactor fuel using SRAC 2006 has been carried out. Electrical energy is important today because the need is increasing along with the increase in human population, advanced technology and the economy. On the other hand, there are demands from the community for the clean, efficient and consistent energy. This is the reason why nuclear power plants are considered as one of the candidates for electrical energy suppliers in Indonesia in particular. This study evaluates a SMART reactor with Gadolinium as the burnable absorber material. The two kinds of fuel assembly were analyzed using the SRAC 2006 code system with the JENDL 4.0 as nuclear data library. This study aims to observe the neutronic characteristics of the fuel assembly designs according to the reference used. The results of the study show that of all types of fuel assemblies used can reach criticality at the beginning of the operating cycle and last up to 3 till 5 years when it finally reaches subcritical condition. Another parameter observed is the conversion ratio value, which from this study is in accordance with the characteristics of the conversion ratio for thermal reactors.

Power Cycles of Generation III and III+ Nuclear Power Plants

2014 ◽  
Author(s):  
Alexey Dragunov ◽  
Eugene Saltanov ◽  
Igor Pioro ◽  
Pavel Kirillov ◽  
Romney Duffey
Keyword(s):  
Renewable Energy ◽  
Natural Gas ◽  
Thermal Efficiency ◽  
Nuclear Power ◽  
Power Plants ◽  
Renewable Energy Sources ◽  
Thermal Power ◽  
Electrical Energy ◽  
Energy Sources ◽  
Thermal Power Plants

It is well known that the electrical-power generation is the key factor for advances in any other industries, agriculture and level of living. In general, electrical energy can be generated by: 1) non-renewable-energy sources such as coal, natural gas, oil, and nuclear; and 2) renewable-energy sources such as hydro, wind, solar, biomass, geothermal and marine. However, the main sources for electrical-energy generation are: 1) thermal - primary coal and secondary natural gas; 2) “large” hydro and 3) nuclear. The rest of the energy sources might have visible impact just in some countries. Modern advanced thermal power plants have reached very high thermal efficiencies (55–62%). In spite of that they are still the largest emitters of carbon dioxide into atmosphere. Due to that, reliable non-fossil-fuel energy generation, such as nuclear power, becomes more and more attractive. However, current Nuclear Power Plants (NPPs) are way behind by thermal efficiency (30–42%) compared to that of advanced thermal power plants. Therefore, it is important to consider various ways to enhance thermal efficiency of NPPs. The paper presents comparison of thermodynamic cycles and layouts of modern NPPs and discusses ways to improve their thermal efficiencies.

scholarly journals Effects of different nuclear evaluation data on the RMC keff calculation

2020 ◽  
Vol 239 ◽  
pp. 19005
Author(s):  
Zhang Wenxin ◽  
Qiang shenglong ◽  
Yin qiang ◽  
Cui Xiantao
Keyword(s):  
Cross Section ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Neutron Cross Section ◽  
Nuclear Data ◽  
Section Data ◽  
Calculation Results ◽  
The Impact ◽  
Physical Calculation

Neutron cross section data is the basis of nuclear reactor physical calculation and has a decisive influence on the accuracy of calculation results. AFA3Gassemble is widely used in nuclear power plants. CENACE is an ACE format multiple-temperature continuous energy cross section library that developed by China Nuclear Data Centre. In this paper, we calculated the AFA3G assemble by RMC.We respectively used ENDF6.8/, ENDF/7 and CENACE data for calculation. The impact of nuclear data on RMC calculation is studied by comparing the results of different nuclear data.

Radiation Protection Technician Two-Year Associates of Applied Science Curriculum for National Implementation

2008 ◽  
Author(s):  
William H. Miller ◽  
David Jonassen ◽  
Rose Marra ◽  
Matthew Schmidt ◽  
Matthew Easter ◽  
...  
Keyword(s):  
Community College ◽  
Radiation Protection ◽  
Nuclear Power ◽  
Power Plants ◽  
Science Curriculum ◽  
Nuclear Industry ◽  
Advanced Technology ◽  
State Technical ◽  
University Of Missouri ◽  
National Implementation

The U.S. Department of Labor awarded a $2.3 million grant to the University of Missouri-Columbia (MU) in 2006 in response to the need for well-trained Radiation Protection Technicians (RPTs). The RPT curriculum initiative resulted from significant collaborations facilitated by MU with community colleges, nuclear power plants, professional organizations, and other nuclear industry stakeholders. The objective of the DOL project is to help increase the pool of well-qualified RPTs to enter the nuclear workforce. Our work is designed to address the nuclear industry’s well-documented, increasingly significant need for RPTs. In response to this need, MU and AmerenUE’s Callaway Nuclear Power Plant first partnered with Linn State Technical College’s Advanced Technology Center (LSTC/ATC) to initiate a two-year RPT degree program. The success of this program (enrollments have been increasing over the past four years to a Fall 2007 enrollment of 23) enabled the successful proposal to the DOL to expand this program nationwide. DOL participants include the following partners: Linn State Technical College with AmerenUE – Callaway; Central Virginia Community College with AREVA; Estrella Mountain Community College with Arizona Public Service – Palo Verde; MiraCosta Community College with Southern California Edison – San Onofre; and Hill College with Texas Utilities – Comanche Peak. The new DOL grant has allowed redevelopment of the LSTC/ATC curriculum using a web-based, scenario driven format, benchmarked against industry training standards. This curriculum will be disseminated to all partners. Integral in this curriculum is a paid, three to four month internship at a nuclear facility. Two of the six new RPT courses have been developed as of the end of 2007. Four of five partner schools are accepting students into this new program starting in the winter 2008 term. We expect that these institutions will graduate 100 new RPTs per year to help alleviate the personnel shortage in this critical area of need.

Analysis of Non-Uniformly Degraded Pipe Sections

2003 ◽  
Author(s):  
Amy J. Smith ◽  
Keshab K. Dwivedy
Keyword(s):  
Wall Thickness ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Operating Cycle ◽  
Moment Capacity ◽  
Asme Code ◽  
Uniform Cross Section ◽  
Uniform Wall ◽  
Set Up

The management of flow assisted corrosion (FAC) has been a part of the maintenance of piping in nuclear power plants for more than 15 years. Programs have been set up to identify vulnerable locations, perform inspections, characterize the degraded configurations, and evaluate the structural integrity of the degraded sections. The section of the pipe is repaired or replaced if the structural integrity cannot be established for the projected degraded section at the next outage. During the past 15 years, significant improvements have been made to every aspect of the program including structural integrity evaluation. Simplified methods and rules are established in ASME Section XI code and in several code cases for verifying structural integrity. The evaluation of structural integrity is performed during the plant outage prior to a decision for repair or replacement. Any improvement in structural integrity evaluation to extend the life of a component by one additional operating cycle can help in performance of repair/replacement of component in a planned manner. Simplified methods and rules provided in the code can be easily used for analysis of pipe sections with degraded area with uniform wall thickness and for non-uniformly degraded sections, provided the degraded portions are modeled with uniform wall thickness equal to the lowest thickness of the section. The representation of a non-uniformly degraded section in this manner is necessarily conservative. The purpose of this paper is to develop methodology to analyze the non-uniformly degraded sections subjected to pressure and moment loading by modeling it in a manner that accounts for the non-uniform cross-section. The formulation developed here is more realistic than the code methodology and is still conservative. The results are presented in form of charts comparing the limit moment capacity of the degraded sections calculated by the formulation in this paper with that using ASME code formulation. The paper concludes that the proposed formulation can be used to supplement the ASME Code method to extend the remaining life of FAC degraded components.

Supercritical Water-Cooled Nuclear Reactors (SCWRs): Current and Future Concepts — Steam Cycle Options

2008 ◽  
Author(s):  
R. B. Duffey ◽  
I. Pioro ◽  
X. Zhou ◽  
U. Zirn ◽  
S. Kuran ◽  
...  
Keyword(s):  
Steam Turbine ◽  
Supercritical Water ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Hydrogen Generation ◽  
Electrical Energy ◽  
Reactor Outlet ◽  
Cycle Efficiency ◽  
Direct Cycle

One of the six Generation IV nuclear reactor concepts is a SuperCritical Water-cooled nuclear Reactor (SCWR), which is currently under development. The main objectives for developing and utilizing SCWRs are to increase the thermal efficiency of Nuclear Power Plants (NPPs), to decrease electrical energy costs, and possibility for co-generation, including hydrogen generation. Atomic Energy of Canada Limited (AECL) and Research and Development Institute of Power Engineering (RDIPE or NIKIET in Russian abbreviations) are currently developing pressure-tube SCWR concepts. The targeted steam parameters at the reactor outlet are approximately 25 MPa and 625°C. This paper presents a survey on modern SuperCritical (SC) steam turbine technology and a study on potential steam cycles for the SCWR plants. The survey reveals that by the time the Gen IV SCWRs are market-ready, the required steam turbine technology will be well proven. Three potential steam cycles in an SCWR plant are presented: a dual-cycle with steam reheat, a direct cycle with steam reheat, and a direct cycle with a Moisture Separator and Reheater (MSR). System thermal-performance simulations have been performed to determine the overall cycle efficiency of the proposed cycles. The results show that the direct cycle with steam reheat has the highest efficiency. The direct cycle with MSR is an alternative option, which will simplify the reactor design at the penalty of a slightly lower cycle efficiency.

Design Advancement, Parameters, and Modes of High Power Turbo Generators

2017 ◽  
Author(s):  
Коган ◽  
Feliks Kogan
Keyword(s):  
Power Systems ◽  
High Power ◽  
Nuclear Power ◽  
Power Plants ◽  
Electrical Energy ◽  
Local Heat ◽  
Specific Power ◽  
System Control ◽  
Modes Of Operation ◽  
Power System Control

Turbo generators installed at the thermal and nuclear power plants produce about 85% of the electrical energy generated in the world. High power turbo generators have higher specific power normalized to the volume, and, as a result, higher tensions in the machine structure and increased sensitivity to switching between the modes of operation. Disruptions in turbo generator standard modes of operation cause increase in electrical currents, significant local heat increase as well as large electrodynamic stresses. The book discusses the improvement of design of Russian turbo generators and how their parameters impact the machine reliability; the book provides analysis of specifics of the abnormal modes of operation, and their limits and the risk of miscounting of these limits in the power system control; the book also discusses the modeling of turbo generators. The book is recommended as a textbook for college and university students, who study Power and Electrical Engineering, engineering and technical staff of power plants and power systems as well as professionals involved in the design and testing of turbine generators.

Detailed CFD Analysis of Coolant Mixing in VVER-440 Fuel Assembly Heads Performed With the Code CFX-5.5

2004 ◽  
Author(s):  
Ga´bor Le´gra´di ◽  
Attila Aszo´di
Keyword(s):  
Fuel Assembly ◽  
Nuclear Power ◽  
Power Plants ◽  
Sensitivity Study ◽  
Temperature Fields ◽  
Head Model ◽  
Detailed Knowledge ◽  
Mixing Processes ◽  
Significant Information ◽  
Fuel Assemblies

3D modeling of the thermal hydraulical processes in a fuel assembly head means a great challenge for the CFD technique due to the complexity of its structure and the flow domain. On the other hand, this field is of great importance since detailed knowledge on mixing processes in the assembly heads and calculations on the signals of the thermocouples positioned just above the heads would give very significant information for the safety analyses connected to the power upgrading of nuclear power plants. Therefore development of a complex fuel assembly model was started in the Institute of Nuclear Techniques of the Budapest University of Technology and Economics in the near past. In this paper, the fuel assembly head model, the sensitivity study of it, calculations and results are presented. The main goal of our work is investigating the signal of the thermocouples which are placed just above the fuel assemblies. The calculations were performed with consideration of four kinds of different fuel assemblies. The inlet velocity and temperature fields were calculated by the COBRA subchannel code of the Paks Nuclear Power Plant of Hungary. With all kind of fuel assemblies, calculations were performed with assumptions of normal symmetrical and highly asymmetrical heat source profiles of inner assemblies and assemblies positioned beside absorber elements.

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