Non-elastic stresses in composite plates under transient thermal conditions

1968 ◽  
Vol 3 (2) ◽  
pp. 103-108
Author(s):  
D J Lewis ◽  
J S Campbell
Keyword(s):  
Thermal Conditions ◽  
Composite Plates ◽  
Pressure Vessels ◽  
Stress System ◽  
Direct Solution ◽  
Plate Material ◽  
Thermal Transients ◽  
Elastic Stresses ◽  
Water Reactors ◽  
Transient Thermal

The successful use of stainless-steel cladding in the pressure vessels of nuclear water reactors is likely to stimulate interest in the applications of composite-plate material. The high cost of using austenitic materials in corrosive conditions could well make the use of clad materials attractive for certain components. A direct solution has been developed for finding the stresses induced by thermal transients in composite plates for equal biaxial stresses. A computer programme has been written and specimen results are shown. Non-elastic effects are included and residual stresses can be found. The programme can be used in any situation where a shell component has material properties varying through the shell thickness, provided that the stress system is equi-biaxial.

Failure Frequency Estimates for Boiling Water Reactor Pressure Vessels

2010 ◽  
Author(s):  
F. A. Simonen ◽  
S. R. Gosselin ◽  
J. E. Rhoads ◽  
A. T. Chiang
Keyword(s):  
Pressure Vessels ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Thermal Transients ◽  
Core Damage ◽  
Water Reactors ◽  
Vessel Rupture ◽  
Reactor Pressure

This paper reviews estimates of rupture frequencies for reactor pressure vessels (RPV) at boiling water reactor (BWR) nuclear power plants as reported in the literature. Results permit improved probabilistic risk assessments (PRA) for severe accidents that could cause core damage and/or challenge the capabilities of BWR reactor containment systems. Current and historical estimates of failure frequencies are considered for light water reactors in general and more specifically for BWR plants. The focus is on large ruptures that could give flow rates exceeding the rates associated with double ended breaks of large diameter recirculation piping. Rupture frequencies for BWR vessels as used for PRA evaluations have historically been assigned low values (e.g. 10−7 to 10−6 per vessel per year). The objective of the present work was to establish possible technical bases for more realistic values of rupture frequency (i.e. 10−8). Historical estimates from the early WASH-1400 reactor safety study were first reviewed and used as a point-of-reference. More recent estimates came from various sources such as a U.S. Nuclear Regulatory Commission expert elicitation process that estimated Loss-of-Coolant Accident (LOCA) frequencies. Other studies both by industry and by the USNRC have addressed rupture frequencies for BWR vessels subject to low-temperature-over-pressure (LTOP) events. On the other hand, recent comprehensive evaluations have focused mostly on RPV failure frequencies for pressurized water reactors (PWRs) caused by pressurized thermal shock events. An important consideration was that rupture frequencies for BWR vessels are believed to be lower than those for PWR vessels, because BWR vessels are less embrittled than PWR vessels and are subject to less severe thermal transients. The review concludes that prior studies support an estimate of 10−8 or less for BWR vessel rupture frequencies. Probabilistic fracture mechanics calculations for individual vessels accounting for plant specific conditions are recommended to support even lower estimated frequencies. Use of more realistic vessel rupture frequencies in a plant’s PRA provides an improvement in not only the perceived plants risk of core damage, but also provides better decision making for plant operation and maintenance activities in that a conservative initiating event treatment within a PRA can mask other initiating events of higher importance.

Extension of PASCAL4 Code for Probabilistic Fracture Mechanics Analysis of Reactor Pressure Vessel in Boiling Water Reactor

2020 ◽  
Author(s):  
Kai Lu ◽  
Jinya Katsuyama ◽  
Yinsheng Li
Keyword(s):  
Fracture Mechanics ◽  
Outer Surface ◽  
Surface Crack ◽  
Pressure Vessels ◽  
Boiling Water ◽  
Thermal Transients ◽  
Probabilistic Fracture Mechanics ◽  
Water Reactors ◽  
Embedded Crack ◽  
Reactor Pressure

Abstract In Japan, Japan Atomic Energy Agency has developed a probabilistic fracture mechanics (PFM) analysis code, PASCAL4, for probabilistic evaluation of reactor pressure vessels (RPVs) in pressurized water reactors (PWRs) considering neutron irradiation embrittlement and pressurized thermal shock (PTS) events. Besides severe PTS events, however, transients associated with normal operations, such as the cooldown and heatup transients associated with reactor shutdown and startup, respectively, should also be considered in the integrity assessment of RPVs in both PWRs and boiling water reactors (BWRs). With regard to a heatup transient, because temperature is at its minimum, and tensile stress at its maximum on the RPV outer surface, outer surface crack and embedded crack near the RPV outer surface should be taken into account. To extend the applicability of PASCAL4, we improved the code to include analysis functions for these cracks. The improved PASCAL4 can be used to run PFM analyses of RPVs subjected to both cooldown (including PTS) and heatup transients. In this paper, improvements made to PASCAL4 are firstly described, including the incorporated stress intensity factor solutions and the corresponding calculation methods for vessel outer surface crack and embedded crack near the outer surface. Using the improved PASCAL4, PFM analysis examples for a Japanese BWR-type model RPV subjected to thermal transients including a low temperature overpressure event and a heatup transient are presented.

Strength Loss in Composite Cylinders Under Impact

1988 ◽  
Vol 110 (2) ◽  
pp. 180-184 ◽  
Author(s):  
A. P. Christoforou ◽  
S. R. Swanson
Keyword(s):  
Composite Structures ◽  
Composite Plates ◽  
Strength Loss ◽  
Pressure Vessels ◽  
Fourier Series Expansion ◽  
Lateral Impact ◽  
Low Velocity ◽  
Expansion Procedure ◽  
The Impact ◽  
Composite Cylinders

The problem of strength loss in composite structures due to impact appears to be important due to the sensitivity of advanced composites to these loadings. Although a number of studies have been carried out on impact of flat composite plates, relatively little work has been done on tubular geometries such as pressure vessels despite the usage in applications. We have addressed the problem of calculating strength loss due to low velocity, lateral impact of composite cylinders. In our model we use an existing Fourier Series expansion procedure to calculate ply stresses and strains, compare these values with allowables to predict fiber breakage during the impact, and finally use fracture mechanics to predict the strength loss due to the impact. Although the model is quite simplified, the general trends of experiments appear to be represented.

scholarly journals Concrete strains under transient thermal conditions: A state-of-the-art review

2016 ◽  
Vol 127 ◽  
pp. 172-188 ◽  
Author(s):  
Giacomo Torelli ◽  
Parthasarathi Mandal ◽  
Martin Gillie ◽  
Van-Xuan Tran
Keyword(s):  
State Of The Art ◽  
Thermal Conditions ◽  
Transient Thermal

A Risk-Informed Methodology for ASME Section XI, Appendix G

2012 ◽  
Vol 134 (3) ◽  
Author(s):  
Ronald Gamble ◽  
William Server ◽  
Bruce Bishop ◽  
Nathan Palm ◽  
Carol Heinecke
Keyword(s):  
Nuclear Regulatory Commission ◽  
The United States ◽  
Service Level ◽  
Pressure Vessels ◽  
Operational Flexibility ◽  
Pressurized Water ◽  
Leak Testing ◽  
Alternative Methodology ◽  
Water Reactors ◽  
Regulatory Commission

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code [1], Section XI, Appendix G provides a deterministic procedure for defining Service Level A and B pressure–temperature limits for ferritic components in the reactor coolant pressure boundary. An alternative risk-informed methodology has been developed for ASME Section XI, Appendix G. This alternative methodology provides easy to use procedures to define risk-informed pressure–temperature limits for Service Level A and B events, including leak testing and reactor start-up and shut-down. Risk-informed pressure–temperature limits provide more operational flexibility, particularly for reactor pressure vessels with relatively high irradiation levels and radiation sensitive materials. This work evaluated selected plants spanning the population of pressurized water reactors (PWRs) and boiling water reactors (BWRs). The evaluation included determining appropriate material properties, reviewing operating history and system operational constraints, and performing probabilistic fracture mechanics (PFM) analyses. The analysis results were used to define risk-informed pressure–temperature relationships that comply with safety goals defined by the United States (U.S.) Nuclear Regulatory Commission (NRC). This alternative methodology will provide greater operational flexibility, especially for Service Level A and B events that may adversely affect efficient and safe plant operation, such as low-temperature-over-pressurization for PWRs and system leak testing for BWRs. Overall, application of this methodology can result in increased plant efficiency and increased plant and personnel safety.

SPH Analysis for Transient Thermal Stress of FGMs with Temperature-Dependent Properties

2015 ◽  
Vol 1096 ◽  
pp. 297-301
Author(s):  
Gui Ming Rong ◽  
Hiroyuki Kisu
Keyword(s):  
Thermal Stress ◽  
Thermal Conditions ◽  
Functionally Graded ◽  
Temperature Dependent ◽  
Graded Materials ◽  
Initial Stage ◽  
Particle Hydrodynamics ◽  
Smoothed Particle ◽  
Temperature Dependent Properties ◽  
Transient Thermal

A formulation using the deviatoric stress and the continuity equation is extended to the analysis of the dynamic response of functionally graded materials (FGMs) subjected to a thermal shock by smoothed particle hydrodynamics (SPH), in which temperature dependent properties of materials are considered. Several dynamic thermal stress problems are analyzed to investigate the fluctuation of thermal stress at the initial stage under three types of thermal conditions, with the addition of two kinds of mechanical boundary conditions.

Advanced Structural Materials and Cladding

MRS Bulletin ◽  
2009 ◽  
Vol 34 (1) ◽  
pp. 20-27 ◽  
Author(s):  
T. Allen ◽  
H. Burlet ◽  
R.K. Nanstad ◽  
M. Samaras ◽  
S. Ukai
Keyword(s):  
High Temperature ◽  
Structural Materials ◽  
Ceramic Composites ◽  
Pressure Vessels ◽  
Austenitic Steels ◽  
High Temperature Strength ◽  
Light Water Reactors ◽  
Modeling Tools ◽  
Light Water ◽  
Water Reactors

AbstractAdvanced nuclear energy systems, both fission- and fusion-based, aim to operate at higher temperatures and greater radiation exposure levels than experienced in current light water reactors. Additionally, they are envisioned to operate in coolants such as helium and sodium that allow for higher operating temperatures. Because of these unique environments, different requirements and challenges are presented for both structural materials and fuel cladding. For core and cladding applications in intermediate-temperature reactors (400–650°C), the primary candidates are 9–12Cr ferritic–martensitic steels (where the numbers represent the weight percentage of Cr in the material, i.e., 9–12 wt%) and advanced austenitic steels, adapted to maximize high-temperature strength without compromising lower temperature toughness. For very high temperature reactors (>650°C), strength and oxidation resistance are more critical. In such conditions, high-temperature metals as well as ceramics and ceramic composites are candidates. For all advanced systems operating at high pressures, performance of the pressure boundary materials (i.e., those components responsible for containing the high-pressure liquids or gases that cool the reactor) is critical to reactor safety. For some reactors, pressure vessels are anticipated to be significantly larger and thicker than those used in light water reactors. The properties through the entire thickness of these components, including the effects of radiation damage as a function of damage rate, are important. For all of these advanced systems, optimizing the microstructures of candidate materials will allow for improved radiation and high-temperature performance in nuclear applications, and advanced modeling tools provide a basis for developing optimized microstructures.

Neutronics Studies on a Generic Pressure-Channel Reactor With Nuclear Steam Superheat

2014 ◽  
Author(s):  
Matthew Baldock ◽  
Wargha Peiman ◽  
Andrei Vincze ◽  
Rand Abdullah ◽  
Khalil Sidawi ◽  
...  
Keyword(s):  
Power Distribution ◽  
Power Plants ◽  
Saturation Pressure ◽  
Pressure Vessels ◽  
Limiting Factor ◽  
Outlet Temperature ◽  
Pressurized Water ◽  
Water Reactors ◽  
1960S And 1970S ◽  
Pressure Channel

In order to increase the thermal efficiency of steam-cycle power plants it is necessary to achieve steam temperatures as high as possible. Current limiting factor for Nuclear Power Plants (NPPs) in achieving higher operating temperatures and, therefore, thermal efficiencies is pressures at which they can operate. From basic thermodynamics it is known that to increase further an outlet temperature in water-cooled reactors a pressure must also be increased. Current level of pressures in Pressurized Water Reactors (PWRs) is about 15–16 MPa. Therefore, next stage should be supercritical pressures, at least 23.5–25 MPa. However, such supercritical-water reactors with pressure vessels of 45–50 cm thickness don’t exist yet. One way around larger pressure vessels as well as the limit of temperature of the coolant on the saturation pressure is to employ a Pressure Channel (PCh) design with Superheated Steam channels (SHS). PCh reactors allow for different coolants and bundle configurations in one reactor core, in this case, steam would be a secondary coolant. In the 1960s and 1970s the USA and Soviet Union tested reactors using pressure channels to super-heat steam in-core to achieve outlet temperatures greater than what is currently possible with convention reactors. Nuclear materials are carefully chosen based on their neutron interaction properties in addition to their strength and resistance to corrosion. Introducing steam channels will not only change the neutronics behavior of the coolant, but require different fuel cladding and pressure-channel materials, specifically, stainless steels or Inconels, to withstand high-temperature steam. This paper will investigate the affect that steam, SS-304 and Inconel will have on neutron economy when introduced into a reactor design as well as required changes to fuel enrichment. It will also be necessary to investigate the effects of these material changes on power distribution inside a reactor. Pressure-channel design requires methods of fine control to maintain a balanced core-power distribution, the introduction of non-uniform coolant and reactor materials will further complicate maintaining uniform reactor power. The degree to which SHS channels will affect the power distribution is investigated in this paper.

Transient Thermal Behaviors of SBO Accident for a 200MW OFNP Under Heaving Motion Conditions

2018 ◽  
Author(s):  
Qiqi Yan ◽  
Simin Luo ◽  
Yapei Zhang ◽  
Limin Liu ◽  
Guanghui Su ◽  
...  
Keyword(s):  
System Analysis ◽  
Nuclear Reactor ◽  
Analysis Model ◽  
Pressurized Water ◽  
Thermal Behaviors ◽  
Severe Accidents ◽  
Nuclear Plants ◽  
Station Blackout ◽  
Water Reactors ◽  
Transient Thermal

For some Pressurized Water Reactors (PWR) operated on automobiles, boats or deep sea vessels, system characteristics is important for understanding their safety during severe accidents. The development of an analysis code and the transient thermal beaviors of a floating nuclear reactor under heaving motion are described in this paper. By modifying the control equations based on the mathematical models of ocean conditions, an ocean condition available system analysis code named RELAP5/GR was developed from RELAP5 MOD3.2 to simulate the transient thermal-hydraulic response of the nuclear reactor systems to the motion conditions in accidents, which is an advanced and independent node programming code. Using the code, the analysis model was established for a small 200MW offshore floating nuclear plants (OFNP). The transient thermal behaviors of the whole system were analyzed in the cases of the station blackout accident under heaving motion conditons. The analysis shows that all the results can be reasonably explained and the code development is successful at this stage.

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