scholarly journals Steel–Concrete Composite Pressure Vessels for Hydrogen Storage at High Pressures

2019 ◽  
Vol 142 (2) ◽  
Author(s):  
Maan Jawad ◽  
Yanli Wang ◽  
Zhili Feng
Keyword(s):  
Internal Pressure ◽  
High Pressures ◽  
Solid Wall ◽  
National Laboratory ◽  
Large Diameter ◽  
Pressure Vessels ◽  
Steel Shell ◽  
Theoretical Formulation ◽  
Oak Ridge ◽  
Concrete Shell

Abstract The need to store large quantities of hydrogen in large diameter steel vessels under high pressures results in shell thicknesses that are too large to produce by most steel mills and not practical to fabricate. Accordingly, a research program was undertaken by Oak Ridge National Laboratory to develop a new concept of combining steel with concrete to construct such vessels economically and practically. The concept is to fabricate vessels where the steel shell thickness is approximately one half that required to resist the hoop forces due to internal pressure. As such, the steel shell is designed to carry the full amount of the longitudinal forces in the vessel but only one half of the hoop loads due to internal pressure. The other half of the hoop loads is carried by a prestressed and reinforced concrete shell. In large diameter vessels, the cost of the shell can further be reduced by using layered steel shell construction rather than solid-wall construction. Such shell construction has also the added advantage of easily venting the hydrogen that permeates through the steel shell directly to the atmosphere through vent holes. This mechanism prevents the hydrogen from damaging the steel shell. The theoretical formulation of the steel concrete shell design is presented in this paper. In addition, details of a full-scale mock up vessel designed, fabricated, and tested to prove the proposed methodology are given.

ORNL Evaluation of Safety Cases for Two Belgian Reactor Pressure Vessels Containing Quasi-Laminar Defects

2017 ◽  
Author(s):  
Hilda B. Klasky ◽  
B. Richard Bass ◽  
Terry L. Dickson ◽  
Sarma B. Gorti ◽  
Randy K. Nanstad ◽  
...  
Keyword(s):  
New York ◽  
Finite Element ◽  
Structural Integrity ◽  
National Laboratory ◽  
Relevant Literature ◽  
Pressure Vessels ◽  
Oak Ridge ◽  
Safety Margins ◽  
Safety Cases ◽  
Reactor Pressure

The Oak Ridge National Laboratory (ORNL) performed a detailed technical review of the 2015 Electrabel (EBL) Safety Cases prepared for the Belgium reactor pressure vessels (RPVs) at Doel 3 and Tihange 2 (D3/T2). The Federal Agency for Nuclear Control (FANC) in Belgium commissioned ORNL to provide a thorough assessment of the existing safety margins against cracking of the RPVs due to the presence of almost laminar flaws found in each RPV. Initial efforts focused on surveying relevant literature that provided necessary background knowledge on the issues related to the quasi-laminar flaws observed in D3/T2 reactors. Next, ORNL proceeded to develop an independent quantitative assessment of the entire flaw population in the two Belgian reactors according to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, Appendix G, “Fracture Toughness Criteria for Protection Against Failure,” New York (both 1992 and 2004 versions). That screening assessment of the EBL-characterized flaws in D3/T2 used ORNL tools, methodologies, and the ASME Code Case N-848, “Alternative Characterization Rules for Quasi-Laminar Flaws”. Results and conclusions derived from comparisons of the ORNL flaw acceptance assessments of D3/T2 with those from the 2015 EBL Safety Cases are presented in the paper. The ORNL screening analyses identified fewer flaws than EBL that were not compliant with the ASME Section XI (1992) criterion; the EBL criterion imposed additional conservatisms not included in ASME Section XI. Furthermore, ORNL’s application of the updated ASME Section XI (2004) criterion produced only four non-compliant flaws, all due to design-basis loss-of-coolant loading transients. Among the latter, only one flaw remained non-compliant when analyzed using the warm-prestress (WPS) cleavage fracture model typically applied in USA flaw assessments. ORNL’s independent refined analysis of that flaw (#1660, which was also non-compliant in the EBL screening assessments) rendered it compliant when modeled as a more realistic individual quasi-laminar flaw using a 3-dimensional XFEM (eXtended Finite Element Method) approach available in the ABAQUS© finite element code. Taken as a whole, the ORNL-specific results and conclusions confirmed the structural integrity of Doel 3 and Tihange 2 under all design transients with ample margin in the presence of the 16,196 detected flaws.

Experiments and Simulations With Large Gas Bubbles in Mercury Towards Establishing a Gas Layer to Mitigate Cavitation Damage

2006 ◽  
Author(s):  
Mark Wendel ◽  
David Felde ◽  
Thomas Karnowski ◽  
Bernard Riemer ◽  
Arthur Ruggles
Keyword(s):  
Liquid Metal ◽  
Solid Wall ◽  
National Laboratory ◽  
Large Bubble ◽  
Cavitation Damage ◽  
Oak Ridge ◽  
Beam Experiment ◽  
Order Of Magnitude ◽  
Protective Gas ◽  
Gas Layer

One option that shows promise for protecting solid surfaces from cavitation damage in liquid metal spallation targets involves introducing an interstitial gas layer between the liquid metal and the containment vessel wall. Several approaches toward establishing such a protective gas layer are being investigated at the Oak Ridge National Laboratory including large bubble injection and methods that involve stabilization of the layer by surface modifications to enhance gas hold-up on the wall or by inserting a porous media. It has previously been reported that using a gas layer configuration in a test target showed an order-of-magnitude decrease in damage for an in-beam experiment. Video images that were taken of the successful gas/mercury flow configuration have been analyzed and correlated. The results show that the success was obtained under conditions where only 60% of the solid wall was covered with gas. Such a result implies that this mitigation scheme may have much more potential. Additional experiments with gas injection into water are underway. Multi-component flow simulations are also being used to provide direction for these new experiments. These simulations have been used to size the gas layer and position multiple inlet nozzles.

NESC-VII: Fracture Mechanics Analyses of WPS Experiments on Large-Scale Cruciform Specimen

2011 ◽  
Author(s):  
Shengjun Yin ◽  
Paul T. Williams ◽  
B. Richard Bass
Keyword(s):  
Fracture Mechanics ◽  
Large Scale ◽  
Structural Integrity ◽  
National Laboratory ◽  
Pressure Vessels ◽  
Cooperative Action ◽  
Oak Ridge ◽  
Integrity Assessment ◽  
Feasibility Test ◽  
Reactor Pressure

This paper describes numerical analyses performed to simulate warm pre-stress (WPS) experiments conducted with large-scale cruciform specimens within the Network for Evaluation of Structural Components (NESC-VII) project. NESC-VII is a European cooperative action in support of WPS application in reactor pressure vessel (RPV) integrity assessment. The project aims in evaluation of the influence of WPS when assessing the structural integrity of RPVs. Advanced fracture mechanics models will be developed and performed to validate experiments concerning the effect of different WPS scenarios on RPV components. The Oak Ridge National Laboratory (ORNL), USA contributes to the Work Package-2 (Analyses of WPS experiments) within the NESC-VII network. A series of WPS type experiments on large-scale cruciform specimens have been conducted at CEA Saclay, France, within the framework of NESC VII project. This paper first describes NESC-VII feasibility test analyses conducted at ORNL. Very good agreement was achieved between AREVA NP SAS and ORNL. Further analyses were conducted to evaluate the NESC-VII WPS tests conducted under Load-Cool-Transient-Fracture (LCTF) and Load-Cool-Fracture (LCF) conditions. This objective of this work is to provide a definitive quantification of WPS effects when assessing the structural integrity of reactor pressure vessels. This information will be utilized to further validate, refine, and improve the WPS models that are being used in probabilistic fracture mechanics computer codes now in use by the NRC staff in their effort to develop risk-informed updates to Title 10 of the U.S. Code of Federal Regulations (CFR), Part 50, Appendix G.

Probabilistic Fracture Mechanics Evaluations That Consider Nozzles in the Extended Beltline Region of Reactor Pressure Vessels

2015 ◽  
Author(s):  
Gary L. Stevens ◽  
Mark T. Kirk ◽  
Terry Dickson
Keyword(s):  
Fracture Mechanics ◽  
National Laboratory ◽  
Computer Code ◽  
Nuclear Regulatory Commission ◽  
Pressure Vessels ◽  
Stress Concentrations ◽  
Oak Ridge ◽  
Probabilistic Fracture Mechanics ◽  
Regulatory Issue ◽  
Reactor Pressure

For many years, ASME Section XI committees have discussed the assessment of nozzle penetrations in various flaw evaluations for reactor pressure vessels (RPVs). As summarized in Reference [1], linear elastic fracture mechanics (LEFM) solutions for nozzle penetrations have been in use since the 1970s. In 2013, one of these solutions was adopted into ASME Code, Section XI, Nonmandatory Appendix G (ASME App. G) [2] for use in developing RPV pressure-temperature (P-T) operating limits. That change to ASME App. G was based on compilation of past work [3] and additional evaluations of fracture driving force [4][5]. To establish the P-T limits on RPV operation, consideration should be given to both the RPV shell material with the highest reference temperature as well as regions of the RPV (e.g., nozzles, flange) that contain structural discontinuities. In October 2014, the U.S. Nuclear Regulatory Commission (NRC) highlighted these requirements in Regulatory Issue Summary (RIS) 2014-11 [6]. Probabilistic fracture mechanics (PFM) analyses performed to support pressurized thermal shock (PTS) evaluations using the Fracture Analysis Vessels Oak Ridge (FAVOR) computer code [7] currently evaluate only the RPV beltline shell regions. These evaluations are based on the assumption that the PFM results are controlled by the higher embrittlement characteristic of the shell region rather than the stress concentration characteristic of the nozzle, which does not experience nearly the embrittlement of the shell due to its greater distance from the core. To evaluate this assumption, the NRC and the Oak Ridge National Laboratory (ORNL) performed PFM analyses to quantify the effect of these stress concentrations on the results of the RPV PFM analyses. This paper summarizes the methods and evaluates the results of these analyses.

Effects of Chemical Content Variation on the Fracture Probability of PWR Pressure Vessel Subjected to PTS Events

2015 ◽  
Author(s):  
Pin-Chiun Huang ◽  
Hsoung-Wei Chou ◽  
Yuh-Ming Ferng
Keyword(s):  
Pressure Vessel ◽  
National Laboratory ◽  
Nickel Content ◽  
The United States ◽  
Fracture Probability ◽  
Pressure Vessels ◽  
Pressurized Water ◽  
Oak Ridge ◽  
Content Variation ◽  
Chemical Content

This paper is to study the effects of copper and nickel content variations on the fracture probability of the pressurized water reactor (PWR) pressure vessel subjected to pressurized-thermal-shock (PTS) transients. The probabilistic fracture mechanics (PFM) code, FAVOR, which was developed by the Oak Ridge National Laboratory in the United States, is employed to perform the analyses. A Taiwan domestic PWR pressure vessel with varied copper and nickel contents of beltline region welds and plates is investigated in the study. Some PTS transients analyzed from Beaver Valley Unit 1 for establishing the USNRC’s new PTS rule are applied as the loading conditions. It is found that the content variation of copper and nickel will significantly affect the radiation embrittlement and the fracture probability of PWR pressure vessels. The results can be regarded as the risk incremental factors for comparison with the safety regulation requirements on vessel degradation as well as a reference for the operation of PWR plants in Taiwan.

Mechanistic Insights Into Risk-Informed Revision of ASME Section XI–Appendix G

2011 ◽  
Author(s):  
Terry Dickson ◽  
Mark Kirk ◽  
Eric Focht
Keyword(s):  
Structural Integrity ◽  
National Laboratory ◽  
Health And Safety ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Pressure Vessels ◽  
Nuclear Industry ◽  
Operating Life ◽  
Oak Ridge ◽  
Asme Code

The current regulations, as set forth by the United States Nuclear Regulatory Commission (NRC), to insure that light-water nuclear reactor pressure vessels (RPVs) maintain their structural integrity, throughout their operating life, when subjected to planned normal reactor startup (heat-up) and shutdown (cool-down) transients are specified in Appendix G to 10 CFR Part 50, which incorporates by reference Appendix G to Section XI of the American Society of Mechanical Engineers (ASME) Code. The technical basis for these regulations are generally considered to be conservative and some plants are finding it operationally difficult to heat-up and cool-down within the accepted limits. Consequently, the nuclear industry has developed, and submitted to the ASME Code for approval, an alternative risk-informed methodology that reduces the conservatism and is consistent with methods previously used to develop a risk-informed revision to the regulations for accidental transients such as pressurized thermal shock (PTS). The objective of the alternative methodology is to increase operational flexibility while continuing to provide reasonable assurance of adequate protection to public health and safety. The NRC and its contractor at Oak Ridge National Laboratory (ORNL) are reviewing the industry proposed risk-informed methodology. Previous results of this review, have been reported at PVP, and a NRC report summarizing all results is currently in preparation. The objective of this paper is to discuss and illustrate mechanistic insights into trends shown previously associated with normal cool-down.

Failure Analysis of Large-Diameter Coiled Tubing Based on Diameter Growth

2019 ◽  
Vol 142 (3) ◽  
Author(s):  
Le Zhao ◽  
Hong Zhang ◽  
Qingquan Duan ◽  
Guoping Tang
Keyword(s):  
Internal Pressure ◽  
Wall Thickness ◽  
High Pressures ◽  
Fatigue Behavior ◽  
Large Diameter ◽  
Fatigue Tests ◽  
Diameter Growth ◽  
Finite Element Software ◽  
Coiled Tubing ◽  
Bending Radius

Abstract Fatigue tests were conducted to analyze the fatigue behavior and diameter growth of large-diameter coiled tubing (CT) under the combined loads of bending and internal pressure. The experimental results reveal that mechanical limitations on the allowable diameter growth mean that the effective working life of CT at high pressures is only a fraction of the available fatigue life. The finite element software abaqus is used to further research the changes in diameter growth and to analyze the sensitivity of CT diameter growth to the main influencing factors, including internal pressure, tubing outside diameter (OD), wall thickness, yield strength, and bending radius. For CT with a diameter larger than 2 in., the diameter growth is sensitive to the above factors. As the bending and straightening cycles increase, the OD of the CT increases in association with obvious ovalization deformation, and the increase in the OD is closely related to the internal pressure load. The redistribution of material causes the wall thickness of the CT to become universally thinner. The ovality of the CT and the uneven decrease in wall thickness reduce the resistance to external extrusion. Therefore, it is becoming increasingly necessary to account for diameter growth as one of the key elements when predicting CT life or determining when to retire a string from service.

Review of Proposed Methodology for Risk-Informed Relaxation to ASME Section XI: Appendix G

2010 ◽  
Author(s):  
Terry Dickson ◽  
Eric Focht ◽  
Mark Kirk
Keyword(s):  
Structural Integrity ◽  
National Laboratory ◽  
The United States ◽  
Pressure Vessels ◽  
Reactor Vessel ◽  
Nuclear Industry ◽  
Oak Ridge ◽  
Wide Range ◽  
Asme Code ◽  
Reactor Pressure

The current regulations, as set forth by the United States Nuclear Regulatory Commission (NRC), to insure that light-water nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to planned normal reactor startup (heat-up) and shut-down (cool-down) transients are specified in Appendix G to 10 CFR Part 50, which incorporates by reference Appendix G to Section XI of the American Society of Mechanical Engineers (ASME) Code. The technical basis for these regulations are now recognized by the technical community as being conservative and some plants are finding it increasingly difficult to comply with the current regulations. Consequently, the nuclear industry has developed, and submitted to the ASME Code for approval, an alternative risk-informed methodology that reduces the conservatism and is consistent with the methods previously used to develop a risk-informed revision to the regulations for accidental transients such as pressurized thermal shock (PTS). The objective of the alternative methodology is to provide a relaxation to the current regulations which will provide more operational flexibility, particularly for reactor pressure vessels with relatively high irradiation levels and radiation sensitive materials, while continuing to provide reasonable assurance of adequate protection to public health and safety. The NRC and its contractor at Oak Ridge National Laboratory (ORNL) have recently performed an independent review of the industry proposed methodology. The NRC / ORNL review consisted of performing probabilistic fracture mechanics (PFM) analyses for a matrix of cool-down and heat-up rates, permutated over various reactor geometries and characteristics, each at multiple levels of embrittlement, including 60 effective full power years (EFPY) and beyond, for various postulated flaw characterizations. The objective of this review is to quantify the risk of a reactor vessel experiencing non-ductile fracture, and possible subsequent failure, over a wide range of normal transient conditions, when the maximum allowable thermal-hydraulic boundary conditions, derived from both the current ASME code and the industry proposed methodology, are imposed on the inner surface of the reactor vessel. This paper discusses the results of the NRC/ORNL review of the industry proposal including the matrices of PFM analyses, results, insights, and conclusions derived from these analyses.

Plug-in scripts for EFTEM automation

1996 ◽  
Vol 54 ◽  
pp. 546-547
Author(s):  
N. D. Evans ◽  
M. K. Kundmann
Keyword(s):  
Electron Microscopy ◽  
National Laboratory ◽  
Oak Ridge ◽  
Multiple User ◽  
Transmission Electron ◽  
Short Period ◽  
User Facility ◽  
Instrument Control ◽  
And Control ◽  
Research Equipment

Post-column energy-filtered transmission electron microscopy (EFTEM) is inherently challenging as it requires the researcher to setup, align, and control both the microscope and the energy-filter. The software behind an EFTEM system is therefore critical to efficient, day-to-day application of this technique. This is particularly the case in a multiple-user environment such as at the Shared Research Equipment (SHaRE) User Facility at Oak Ridge National Laboratory. Here, visiting researchers, who may oe unfamiliar with the details of EFTEM, need to accomplish as much as possible in a relatively short period of time.We describe here our work in extending the base software of a commercially available EFTEM system in order to automate and streamline particular EFTEM tasks. The EFTEM system used is a Philips CM30 fitted with a Gatan Imaging Filter (GIF). The base software supplied with this system consists primarily of two Macintosh programs and a collection of add-ons (plug-ins) which provide instrument control, imaging, and data analysis facilities needed to perform EFTEM.

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