FCC-NG: The AREVA Safe Fresh Fuel Shipping Cask to Deliver the Future EPR™ Reactors

2010 ◽  
Author(s):  
M. Doucet ◽  
P. Faye ◽  
Ch. Faignet ◽  
R. Babut ◽  
M. Landrieu ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Nuclear Power ◽  
Power Plants ◽  
Neutron Absorber ◽  
Balsa Wood ◽  
Array Configuration ◽  
Safety Margins ◽  
Safety Case ◽  
Calculation Results ◽  
Set Up

AREVA as a worldwide PWR fuel provider has to have a fleet of fresh UO2 shipping casks being agreed within a lot of countries and to accommodate foreseen EPR™ Nuclear Power Plants fuel buildings. To reach this target the AREVA NP Fuel Sector together with TN International (a subsidiary of AREVA NC) decided to develop an up to date shipping cask gathering experience feedback of the today fleet and an improved safety allowing the design to comply with international regulations (NRC and IAEA) and other local foreign Safety Authorities requirements. Based on pre design features a safety case was set up to highlight safety margins. Criticality hypothetical accidental assumptions were defined: • Preferential flooding, • Fuel rod array expansion (so called “bird caging” effect), • Fuel sliding, • Neutron absorber penalty, • …. The French criticality code package CRISTAL is used to check several configurations reactivity and derived safety margins. The CRISTAL code package relies on a wide validation process helping defining uncertainties as required by regulations in force. The shipping cask, containing two fuel assemblies, is designed to maximize fuel isolation inside the cask and with neighboring ones even for large array configuration cases. Few and proven industrial products are used: • Stainless steel for the structural frame, • Balsa wood for impact limiters, • BORA® resin as neutron absorber. The cask is designed to handle mainly the EPR™ fuel assembly type and may be extended to other contents such as APWR fuel assembly type. After a brief presentation of the computer codes and the description of the shipping cask, the CRISTAL calculation results as well as the applied uncertainties will be discussed.

scholarly journals Structural Modeling and Dynamic Analysis of a Nuclear Reactor Building

2020 ◽  
Author(s):  
Evrim Oyguc ◽  
Abdul Hayır ◽  
Resat Oyguc
Keyword(s):  
Nonlinear Dynamic ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Energy Sources ◽  
Finite Element Program ◽  
Design Spectrum ◽  
Reactor Building ◽  
Set Up

Increasing energy demand urge the developing countries to consider different types of energy sources. Owing the fact that the energy production capacity of renewable energy sources is lower than a nuclear power plant, developed countries like US, France, Japan, Russia and China lead to construct nuclear power plants. These countries compensate 80% of their energy need from nuclear power plants. Further, they periodically conduct tests in order to assess the safety of the existing nuclear power plants by applying impact type loads to the structures. In this study, a sample third-generation nuclear reactor building has been selected to assess its seismic behavior and to observe the crack propagations of the prestressed outer containment. First, a 3D model has been set up using ABAQUS finite element program. Afterwards, modal analysis is conducted to determine the mode shapes. Nonlinear dynamic time history analyses are then followed using an artificial strong ground motion which is compatible with the mean design spectrum of the previously selected ground motions that are scaled to Eurocode 8 Soil type B design spectrum. Results of the conducted nonlinear dynamic analyses are considered in terms of stress distributions and crack propagations.

Qualification Requirements Design for Neutron Absorber Plate for Spent Fuel Racks in Domestic Nuclear Power Plants

2019 ◽  
Vol 5 (3) ◽  
Author(s):  
You Shi ◽  
Dong Ning ◽  
Yi-zhong Yang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Aluminum Matrix ◽  
Practical Significance ◽  
Spent Fuel ◽  
Absorber Plate ◽  
Neutron Absorber ◽  
Fuel Storage ◽  
B4c Particle

Boron carbide (B4C) particle-reinforced aluminum matrix composite is the key material for use as neutron absorber plate in fuel storage applications for Generation III advanced passive nuclear power plants in China. This material has once depended upon importing with various restrictions so that it has meaningful practical significance to realize the localized manufacturing for this material in China. More importantly, since it is the first time for this material to be used in domestic plant, particular care should be taken to assure the formal supplied products exhibit high stabilized and reliable service in domestic nuclear engineering. This paper initiates and proposes a principle design framework from technical view in qualification requirements for this material so as to guide the practical engineering application. Aiming at neutron absorber materials supplied under practical manufacturing condition in engineering delivery, the qualification requirements define B4C content, matrix chemistry, 10B isotope, bulk density, 10B areal density, mechanical property, and microstructure as key criteria for material performance. The uniformity assessment as to different locations of this material is also required from at least three lots of material. Only qualified material meeting all of the qualification requirements should proceed to be verified by lifetime testing such as irradiation, corrosion, and thermal aging testing. Systematic and comprehensive performance assessments and verification for process stabilization could be achieved through the above qualification. The long-term service for this neutron absorber material in reliable and safe way could be convincingly expected in spent fuel storage application in China.

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.

Effect of Opening on the Response of Containment Model to Pressure Loading

2006 ◽  
Author(s):  
M. K. Agrawal ◽  
A. Ravi Kiran ◽  
A. K. Ghosh ◽  
H. S. Kushwaha
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Dead Weight ◽  
Modeling Methodology ◽  
Containment Model ◽  
Scale Ratio ◽  
Heavy Water Reactor ◽  
Set Up ◽  
Accident Conditions ◽  
Design Pressure

The Containment Studies Facility (CSF) is being set up in BARC for studying various containment related thermal hydraulic and other phenomena which occur during simulated accident conditions in Nuclear power Plants. The facility consists of a concrete containment model having a volumetric scale ratio of 200:1 with respect to the actual containment of Indian Pressurized Heavy Water Reactor. The structure is designed for pressure of 1.73 Kg/cm2 for specified leak tightness. Adequacy to withstand design pressure is checked by test as well as numerical analysis before commissioning of the facility. Accordingly Containment building model has been analyzed by finite element method for internal design pressure and dead weight. Analysis has been carried out for the structure with and without the opening in the containment. Effect of opening on the response of containment has been studied. The paper includes the modeling methodology, maximum deflection and stress amplification around the opening for various models.

Proposed Improvements of the ASME B&PV Code and Future Directions of CIWG

2013 ◽  
Author(s):  
Jinquan Yan ◽  
Yinbiao He ◽  
Gang Li ◽  
Hao Yu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Third Generation ◽  
Future Directions ◽  
The Past ◽  
Development Direction ◽  
Asme Code ◽  
Set Up ◽  
Near Future ◽  
First Time

The ASME B&PV Code, Section III, is being used as the design acceptance criteria in the construction of China’s third generation AP1000 nuclear power plants. This is the first time that the ASME Code was fully accepted in Chinese nuclear power industry. In the past 6 years, a few improvements of the Code were found to be necessary to satisfy the various requirements originated from these new power plant (NPP) constructions. These improvements are originated from a) the stress-strain curves needed in elastic-plastic analysis, b) the environmental fatigue issue, c) the perplexity generated from the examination requirements after hydrostatic test and d) the safe end welding problems. In this paper, the necessities of these proposed improvements on the ASME B&PV code are further explained and discussed case by case. Hopefully, through these efforts, the near future development direction and assignment of the ASME B&PV-III China International Working Group can be set up.

AFCEN RCC-F: A New Standard for the Fire Protection Design of New Built Light Water Nuclear Power Plants

2018 ◽  
Author(s):  
Bernard Gautier ◽  
Mickael Cesbron ◽  
Richard Tulinski
Keyword(s):  
Fire Protection ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fire Hazard ◽  
Pressurized Water ◽  
Light Water ◽  
Protection Scheme ◽  
Safety Case ◽  
Under Construction

Fire hazard is an important issue for the safety of nuclear power plants: the main internal hazard in terms of frequency, and probably one the most significant with regards to the design costs. AFCEN is publishing in 2018 a new code for fire protection of new built PWR nuclear plants, so-called RCC-F. This code is an evolution of the former ETC-F code which has been applied to different EPR plants under construction (Flamanville 3 (FA3, France), Hinkley Point C (HPC, United Kingdom), Taïshan (TSN, China)). The RCC-F code presents significant enhancement and evolutions resulting from eight years of work by the AFCEN dedicated sub-committee, involving a panel of contributors from the nuclear field. It is now opened to any type of PWR (Pressurized Water Reactor) type of nuclear power plants and not any longer limited to EPR (European Pressurized Reactor) plants. It can potentially be adapted to other light water concepts. Its objective is to help engineers design the fire prevention and protection scheme, systems and equipment with regards to the safety case and the defense in depth taking into account the French and European experience in the field. It deals also with the national regulations, with two appendices dedicated to French and British regulations respectively. The presentation gives an overview of the code specifications and focuses on the significant improvements.

Experimental Study on the Interaction of Molten Sn With Water

2018 ◽  
Author(s):  
Longkun He ◽  
Pengfei Liu ◽  
Xisi Zhang ◽  
Wenjun Hu ◽  
Bo Kuang ◽  
...  
Keyword(s):  
Water Temperature ◽  
High Speed ◽  
Nuclear Power ◽  
Power Plants ◽  
Dynamic Pressure ◽  
Ambient Pressure ◽  
Influence Factors ◽  
Video Camera ◽  
Metallic Oxide ◽  
Set Up

In nuclear power plants, fuel-coolant interaction (FCI) often accompanied with core melt accidents, which may escalate to steam explosion destroying the integrity of structural components and even the containment under certain conditions. In the present study, a new facility for intermediate-scaled experiments named ‘Test for Interaction of MELt with Coolant’ (TIMELCO) has been set up to study FCI phenomena and thermal-hydraulic influence factors in metal or metallic oxide/water mixtures with melt at maximum 2750°C. The first series of tests was performed using 3kg of Sn which was heated to 800°Cand jetted into a column of 1m water depth (300mm in diameter) under 0.1MPa ambient pressure. The main changing parameter was water temperature, at 60 °C and 72 °C respectively. From the high-speed video camera, violent explosion phenomenon occurred at water temperature of 60°C, while no evident explosion observed at 72°C. The size of melt debris at 60°C is smaller than this at 72°C.On the contrary, the dynamic pressure at 60°C is larger. The results indicate that water temperature has an important effect on FCI and decreasing the temperature of the coolant is advantageous to the explosion.

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.

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