Evaluation and Prediction of Critical Overhang Length Under Pipe Whip Accident

1986 ◽  
Vol 108 (2) ◽  
pp. 182-187
Author(s):  
T. Yano ◽  
N. Miyazaki ◽  
S. Miyazono
Keyword(s):  
Estimation Method ◽  
Plastic Collapse ◽  
Light Water Reactors ◽  
Pipe Length ◽  
Loss Of Coolant Accident ◽  
Overhang Length ◽  
Water Reactors ◽  
The Moment ◽  
Accident Conditions ◽  
Good Agreement

When the pipe length between break exit and restraint is long in the pipe whip accident, the pipe will undergo a plastic collapse as the moment increases. The length at which plastic collapse may occur is called the critical overhang length, (OH)cr. The experimental results of (OH)cr show good agreement with the prediction by a static simplified estimation method for (OH)cr although the pipe whipping is a dynamic phenomenon. The diagrams of (OH)cr are also described for a range of sizes of stainless steel pipe under the loss of coolant accident conditions of light water reactors.

scholarly journals Steam Oxidation of Silicon Carbide at High Temperatures for the Application as Accident Tolerant Fuel Cladding, an Overview

Thermo ◽  
2021 ◽  
Vol 1 (2) ◽  
pp. 151-167
Author(s):  
Hai V. Pham ◽  
Masaki Kurata ◽  
Martin Steinbrueck
Keyword(s):  
Silicon Carbide ◽  
High Temperatures ◽  
Nuclear Power ◽  
Steam Oxidation ◽  
Severe Accident ◽  
Light Water Reactors ◽  
Severe Accidents ◽  
Potential Benefits ◽  
Water Reactors ◽  
Accident Conditions

Since the nuclear accident at Fukushima Daiichi Nuclear Power Station in 2011, a considerable number of studies have been conducted to develop accident tolerant fuel (ATF) claddings for safety enhancement of light water reactors. Among many potential ATF claddings, silicon carbide is one of the most promising candidates with many superior features suitable for nuclear applications. In spite of many potential benefits of SiC cladding, there are some concerns over the oxidation/corrosion resistance of the cladding, especially at extreme temperatures (up to 2000 °C) in severe accidents. However, the study of SiC steam oxidation in conventional test facilities in water vapor atmospheres at temperatures above 1600 °C is very challenging. In recent years, several efforts have been made to modify existing or to develop new advanced test facilities to perform material oxidation tests in steam environments typical of severe accident conditions. In this article, the authors outline the features of SiC oxidation/corrosion at high temperatures, as well as the developments of advanced test facilities in their laboratories, and, finally, give some of the current advances in understanding based on recent data obtained from those advanced test facilities.

Computational Study for Inventory Estimation of Se-79, Tc-99, Sn-126, and Cs-135 in High-Level Radioactive Wastes From Spent Nuclear Fuels of Light Water Reactors

2011 ◽  
Author(s):  
Keisuke Okumura ◽  
Shiho Asai ◽  
Yukiko Hanzawa ◽  
Tsutomu Okamoto ◽  
Hideya Suzuki ◽  
...  
Keyword(s):  
Parametric Study ◽  
Computational Study ◽  
Estimation Method ◽  
Fission Products ◽  
Radioactive Wastes ◽  
Light Water Reactors ◽  
Nuclear Fuels ◽  
Light Water ◽  
Water Reactors ◽  
High Level

Inventory estimation of long-lived fission products (LLFPs) in high-level radioactive wastes (HLW) from spent nuclear fuels of light water reactors is important for a safety assessment of their disposal. In order to develop an inventory estimation method of difficult-to-measure LLFPs (Se-79, Tc-99, Sn-126, and Cs-135), a parametric study was carried out by using a sophisticated burnup calculation code and data. In the parametric study, fuel specifications and irradiation conditions are changed in the conceivable range. The considered parameters are fuel assembly types (PWR / BWR), U-235 enrichment, moderator temperature, void fraction, power density, and so on. From the calculated results, we clarify the burnup characteristics of the target LLFPs and their possible ranges of generations. Finally, candidates of the key nuclide are proposed for the scaling factor method of HLW.

Stratified Flow Experiments Pertaining to Advanced LWR Reactor Passive Safety System Designs

2004 ◽  
Author(s):  
Jim C. P. Liou ◽  
Alan G. Stephens ◽  
Richard R. Schultz
Keyword(s):  
Flow Behavior ◽  
Stratified Flow ◽  
Water Hammer ◽  
Test Facility ◽  
Light Water Reactors ◽  
Loss Of Coolant Accident ◽  
Passive Safety System ◽  
Water Reactors ◽  
Lower Temperature ◽  
System Designs

During a loss-of-coolant-accident in advanced light water reactors, outside coolant enters the cold leg by gravity to cool the core. This coolant is at a substantially lower temperature and thus is heavier than the liquid in and from the reactor. Consequently, stratified flow may occur. A stratified flow may cause condensation-induced water hammer, and will influence the coolant flow behavior. Two sets of experiments are in progress to better understand stratified flow conditions that lead to water hammer, and the density stratification behavior. The first set uses air-oil-water as the test media. Its purposes are to conduct exploratory tests and to provide instruction an apparatus for education purposes. The second set of tests will use steam and water and, later, the refrigerant R123. This paper describes the exploratory test facility, gives a brief description of the facility that will be used for the steam-water and refrigerant tests, describes the overall test plan, and finally gives some preliminary results on the intrusion of a lighter liquid into a pipe against flow.

Iodine-steel reactions under severe accident conditions in light-water reactors

1996 ◽  
Vol 166 (3) ◽  
pp. 357-365 ◽  
Author(s):  
F. Funke ◽  
G.-U. Greger ◽  
S. Hellmann ◽  
A. Bleier ◽  
W. Morell
Keyword(s):  
Severe Accident ◽  
Light Water Reactors ◽  
Light Water ◽  
Water Reactors ◽  
Accident Conditions

scholarly journals Demonstration of Multi-Physics Evaluation of Fuel Rod Burst Probability With Burnup Extension Under LBLOCA Conditions

2020 ◽  
Author(s):  
Hongbin Zhang ◽  
Cole Blakely ◽  
Jianguo Yu
Keyword(s):  
Operating Time ◽  
Economic Benefits ◽  
Pressurized Water ◽  
Fuel Rod ◽  
Current Limit ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Water Reactors ◽  
Accident Conditions ◽  
Operating Cycles

Abstract Extending the fuel discharge burnup level, e.g., from the current limit of rod averaged discharge burnup limit of 62 GWD/MT to a proposed new limit of 75 GWD/MT, can provide significant economic benefits to the current fleet of operating light water reactors (LWRs). It allows for longer operating cycles and improved resource utilization. The major economic gain of longer operating cycles is attributable to the increased capacity factor resulting from decreased refueling time as a fraction of total operating time, as well as fewer assemblies to be discharged for a given amount of energy produced. The main licensing challenges for higher burnup fuel are to ensure fuel rod safety under design basis accident conditions, especially under large-break loss-of-coolant accident (LBLOCA) and reactivity insertion accident (RIA). In this work, two-year cycle core design for a typical 4-loop pressurized water reactor (PWR) is performed with enrichment increased up to 6% and burnup extended to 75 GWD/MT. The fuel rod burst potential evaluations under large-break loss-of-coolant accident (LBLOCA) conditions are subsequently performed using the multi-physics best estimate plus uncertainty analysis framework LOTUS (LOCA Toolkit for the U.S. LWRs) and the preliminary results are presented.

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