Primary coolant activity of 54Mn, 59Fe, 58Co, 60Co, and 51Cr in system integrated small and modular reactor

2021 ◽  
Vol 160 ◽  
pp. 108356
Author(s):  
Khurram Mehboob ◽  
Yahya A. Al-Zahrani ◽  
Mohammad S. Aljohani ◽  
Abdulsalam Alhawsawi
Keyword(s):  
Primary Coolant ◽  
Coolant Activity

Thermodynamic considerations and prediction of the primary coolant activity of 99Tc

2005 ◽  
Vol 340 (1) ◽  
pp. 69-82 ◽  
Author(s):  
B.J. Lewis ◽  
W.T. Thompson ◽  
F. Akbari ◽  
C. Morrison ◽  
A. Husain
Keyword(s):  
Primary Coolant ◽  
Coolant Activity

JP-8+100: The Development of High-Thermal-Stability Jet Fuel

1996 ◽  
Vol 118 (3) ◽  
pp. 170-179 ◽  
Author(s):  
S. P. Heneghan ◽  
S. Zabarnick ◽  
D. R. Ballal ◽  
W. E. Harrison
Keyword(s):  
Thermal Stability ◽  
Jet Fuel ◽  
High Thermal Stability ◽  
Maximum Temperature ◽  
Additive Package ◽  
Primary Coolant ◽  
Free Radical Theory ◽  
Radical Theory ◽  
Current Availability ◽  
Strategic Factors

Jet fuel requirements have evolved over the years as a balance of the demands placed by advanced aircraft performance (technological need), fuel cost (economic factors), and fuel availability (strategic factors). In a modern aircraft, the jet fuel not only provides the propulsive energy for flight, but also is the primary coolant for aircraft and engine subsystems. To meet the evolving challenge of improving the cooling potential of jet fuel while maintaining the current availability at a minimal price increase, the U.S. Air Force, industry, and academia have teamed to develop an additive package for JP-8 fuels. This paper describes the development of an additive package for JP-8, to produce “JP-8+100.” This new fuel offers a 55°C (100°F) increase in the bulk maximum temperature (from 325°F to 425°F) and improves the heat sink capability by 50 percent. Major advances made during the development of JP-8+100 fuel include the development of several new quantitative fuel analysis tests, a free radical theory of autooxidation, adaptation of new chemistry models to computational fluid dynamics programs, and a nonparametric statistical analysis to evaluate thermal stability. Hundreds of additives were tested for effectiveness, and a package of additives was then formulated for JP-8 fuel. This package has been tested for fuel system materials compatibility and general fuel applicability. To date, the flight testing has shown an improvement in thermal stability of JP-8 fuel. This improvement has resulted in a significant reduction in fuel-related maintenance costs and a threefold increase in mean time between fuel-related failures. In this manner, a novel high-thermal-stability jet fuel for the 21st century has been successfully developed.

Stand-Alone Containment Analysis of the PHÉBUS FPT-1 Test With the ASTEC and the MELCOR Codes

2016 ◽  
Author(s):  
Bruno Gonfiotti ◽  
Sandro Paci
Keyword(s):  
State Of The Art ◽  
External Environment ◽  
Fission Products ◽  
Severe Accident ◽  
Sensitivity Analyses ◽  
Complex Task ◽  
Primary Coolant ◽  
Control Volumes ◽  
Coolant System

The estimation of Fission Products (FPs) release from the containment system of a nuclear plant to the external environment during a Severe Accident (SA) is a quite complex task. In the last 30–40 years several efforts were made to understand and to investigate the different phenomena occurring in such a kind of accidents in the primary coolant system and in the containment. These researches moved along two tracks: understanding of involved phenomenologies through the execution of different experiments, and creation of numerical codes capable to simulate such phenomena. These codes are continuously developed to reflect the actual SA state-of-the-art, but it is necessary to continuously check that modifications and improvements are able to increase the quality of the obtained results. For this purpose, a continuous verification and validation work should be carried out. Therefore, the aim of the present work is to re-analyze the Phébus FPT-1 test employing the ASTEC (F) and MELCOR (USA) codes. The analysis focuses on the stand-alone containment aspects of the test, and three different modellisations of the containment vessel have been developed showing that at least 15/20 Control Volumes (CVs) are necessary for the spatial schematization to correctly predict thermal-hydraulics and the aerosol behavior. Furthermore, the paper summarizes the main thermal-hydraulic results, and presents different sensitivity analyses carried out on the aerosols and FPs behavior.

FEM Simulation of Waves Excited by Array Probe Propagating in Centrifugally Cast Stainless Steel With Hexagonal Column

2021 ◽  
Author(s):  
Shan Lin ◽  
Yasushi Ikegami
Keyword(s):  
Stainless Steel ◽  
Fem Simulation ◽  
Transversely Isotropic ◽  
Primary Coolant ◽  
Centrifugally Cast ◽  
Anisotropic Property ◽  
Array Probe ◽  
Columnar Grains ◽  
Promising Solution ◽  
Cast Stainless Steel

Abstract Centrifugally cast stainless steel (CCSS) is widely used in PWR primary coolant systems. However, ultrasonic testing for such material is very challenging because its coarse grains and anisotropic property. The phased array ultrasonic technology (PAUT) is considered the most promising solution to the problem mentioned. To improve the accuracy of PAUT for CCSS with columnar grains, we used the voxel-based finite element method to perform simulation of wave propagation in CCSS, where waves were excited by a linear array. We modeled columnar grains in CCSS with hexagonal columns and introduced a side-drilled hole. It was easily to have different inclined columnar grains by rotating the crystal axes. All column crystals were considered cubic crystals while CCSS with columnar grains was macroscopically transversely isotropic. Wave propagations were computed for different focal laws and their results were compared. Waves exactly propagated toward and focused at the targeted SDH when focal laws were calculated according to the anisotropic property of CCSS, but deviated the target for focal laws based on isotropy.

Construction Aspects of the Doel 1 and 2 SGR Projects

2009 ◽  
Author(s):  
Bert Kroes ◽  
Edmond Gobert ◽  
Xavier Delhaye ◽  
Peter Devolder ◽  
Michel Sonville
Keyword(s):  
Nuclear Power ◽  
Steam Generators ◽  
Primary Coolant ◽  
Worker Exposure ◽  
Power Stations ◽  
Exchange Operation ◽  
Radiation Worker ◽  
Opening And Closing ◽  
Leak Before Break ◽  
Specific Construction

The Doel 1 and 2 PWR Nuclear Power Stations are the oldest commercially operating units in Belgium and the last to replace their two Steam Generators. The Doel 2 Steam Generators were replaced in 2004 and those of Doel 1 will be replaced late 2009. The replacement poses a particular challenge as these are the only stations in Belgium requiring the creation of primary and secondary containment opening for the SG exchange operation. Other construction challenges result from the a-typical SG support configuration which dates from the period well before the more or less standardized support configuration as used for later PWR units. The current paper discusses the construction approaches selected to facilitate the exchange operation and to minimize the outage duration and radiation worker exposure. The main particularities of the construction effort concern the secondary containment opening and closing using a structural formwork assembly, the use of containment platforms hanging inside the primary containment allowing for parallel primary and secondary containment reconstruction and the de-activation of some of the primary coolant piping and SG restraints following the licensing acceptance of the Leak Before Break concept for the primary piping. The specific construction options that made the Doel 2 replacement a success will be presented in this paper.

Development of Laser Cleaning for Metallic Equipment

2018 ◽  
Author(s):  
L. Carvalho ◽  
W. Pacquentin ◽  
M. Tabarant ◽  
J. Lambert ◽  
A. Semerok ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Laser Scanning ◽  
Surface Defects ◽  
Complex Geometry ◽  
Laser Cleaning ◽  
304L Stainless Steel ◽  
Primary Coolant ◽  
Aisi 304L ◽  
Weight Percentage ◽  
Aisi 304L Stainless Steel

Laser cleaning study was performed on prepared samples using a nanosecond pulsed ytterbium fiber laser. To prepare samples, AISI 304L stainless steel samples were oxidized and implemented with non-radioactive contaminants in a controlled manner. In order to validate the cleaning process for metallic equipment polluted in nuclear installations, two types of contamination with europium (Eu) and with cobalt (Co) were studied. Eu was used as a simulator-product resulting from uranium fission, while Co — as an activation-product of nickel, which is a composing element of a primary coolant system of a reactor. The oxide layers have suffered laser scanning which was followed by the furnace treatment to obtain thicknesses in the range of 100 nm to 1 μm depending on the oxidation parameters [1] with a mean weight percentage of 1% of Eu and 1 % of Co in the volume of the oxide layer. Glow Discharge Optical Emission (GD-OES) and Mass Spectrometry (GD-MS) analyses have been performed to assess the efficiency of the cleaning treatment and to follow the distribution of residual contamination with a detection limit of 0.1mg/kg of Eu and Co. Decontamination rates up to 95.5 % were obtained. One of the identified reasons for this limitation is that the radionuclides are trapped in surface defects like micro cracks [2, 3]. Therefore, cleaning treatments have been applied on surface defects with controlled geometry and a micrometric aperture obtained by laser engraving and juxtaposition of polished sheets of AISI 304L stainless steel. The goal of this study is surface decontamination without either welding or inducing penetration of contamination into the cracks. GD-MS analysis and Scanning Electron Microscopy (SEM) were performed to analyze the efficiency of the treatment and the diffusion of contaminants in this complex geometry.

Root Cause Analysis of SI Nozzle Thermal Sleeve Breakaway Failures Occurring at PWR Plants

2008 ◽  
Vol 131 (1) ◽  
Author(s):  
Jong Chull Jo ◽  
Myung Jo Jhung ◽  
Seon Oh Yu ◽  
Hho Jung Kim ◽  
Young Gill Yune
Keyword(s):  
Cold Water ◽  
Operation Mode ◽  
Branch Pipe ◽  
Thin Wall ◽  
Reactor Operation ◽  
Support Condition ◽  
Primary Coolant ◽  
Root Cause ◽  
Main Pipe ◽  
Flow Situation

At conventional pressurized water reactors (PWRs), cold water stored in the refueling water tank of emergency core cooling system is injected into the primary coolant system through a safety injection (SI) line, which is connected to each cold leg pipe between the main coolant pump and the reactor vessel during the SI operation, which begins on the receipt of a loss of coolant accident signal. In normal reactor power operation mode, the wall of SI line nozzle maintains at high temperature because it is the junction part connected to the cold leg pipe through which the hot main coolant flows. To prevent and relieve excessive transient thermal stress in the nozzle wall, which may be caused by the direct contact of cold water in the SI operation mode, a thermal sleeve in the shape of thin wall cylinder is set in the nozzle part of each SI line. Recently, mechanical failures that the sleeves are separated from the SI branch pipe and fall into the connected cold leg main pipe occurred in sequence at some typical PWR plants in Korea. To find out the root cause of thermal sleeve breakaway failures, the flow situation in the junction of primary coolant main pipe-SI branch pipe and the vibration modal characteristics of the thermal sleeve are investigated in detail by using both computational fluid dynamics code and structure analysis finite element code. As a result, the transient response in fluid pressure exerting on the local part of thermal sleeve wall surface to the primary coolant flow through the pipe junction area during the normal reactor operation mode shows oscillatory characteristics with the frequencies ranging from 15Hzto18Hz. These frequencies coincide with the lower mode natural frequencies of thermal sleeve, which has a pinned support condition on the outer surface with the circumferential prominence set into the circumferential groove on the inner surface of SI nozzle at the midheight of thermal sleeve. In addition, the variation of pressure on the thermal sleeve surface yields alternating forces and torques in the directions of two rectangular axes perpendicular to the longitudinal axis of cylindrical thermal sleeve, which causes both rolling and pitching motions of the thermal sleeve. Consequently, it is seen that this flow situation surrounding the thermal sleeve during the normal reactor operation can induce resonant vibrations accompanying the shaking motion of the thermal sleeve at the pinned support condition, which finally leads to the failures of thermal sleeve breakaway from the SI nozzle.

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