Design criteria for the thermal insulation of reactor coolant system main equipments and piping of PWR nuclear power plants

Ultimate behaviors of cable trays, used in nuclear plants, have not been well studied since cable trays are designed based on conservative design criteria. In this study, by employing a large shaker table, an ultimate strength test was conducted for cable trays used in nuclear power plants. This report describes the results of shaker table test. The following results were obtained: 1) In an S2* earthquake, the damping ratio was so large — more than 30% due to the rubbing of cables — that a large response was not present and the strains in the support were within the elastic limits. 2) The support was strong enough to sustain the cable trays even when the strain in the support was 20 times larger than elastic strain.

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Fatigue Effect of RCS Branch Line by Thermal Stratification

Problems Involving Thermal Hydraulics, Liquid Sloshing, and Extreme Loads on Structures ◽

10.1115/pvp2003-1817 ◽

2003 ◽

Author(s):

Hag-Ki Youm ◽

Kwang-Chu Kim ◽

Man-Heung Park ◽

Tea-Eun Jin ◽

Sun-Ki Lee ◽

...

Keyword(s):

Thermal Stratification ◽

Nuclear Power ◽

Power Plants ◽

Evaluation Methodology ◽

Element Analysis ◽

Branch Line ◽

Reactor Coolant ◽

Stratification Effect ◽

Fatigue Evaluation ◽

Coolant System

Recent events reported at a number of nuclear power plants worldwide have shown that thermal stratification, cycling, and striping in piping can cause excessive thermal stress and fatigue on the piping material. These phenomena are diverse and complicated because of the wide variety of geometry and thermal hydraulic conditions encountered in reactor coolant system. Thermal stratification effect of re-branched lines is not yet considered in the fatigue evaluation. To evaluate the thermal load due to turbulent penetration, this paper presents a fatigue evaluation methodology for a branch line of reactor coolant system with the re-branch line. The locations of fatigue monitoring and supplemented inspections are discussed as a result of fatigue evaluations by Interim Fatigue Management Guideline (ITFMG) and detail finite element analysis. Although the revised CUF was increased less than 50 %, the CUF values for some locations was greater than the ASME Code limits.

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Nuclear power plants. Instrumentation important to safety. Acoustic monitoring systems for detection of loose parts: Characteristics, design criteria and operational procedures

10.3403/30135885u ◽

2015 ◽

Keyword(s):

Nuclear Power ◽

Power Plants ◽

Nuclear Power Plants ◽

Acoustic Monitoring ◽

Monitoring Systems ◽

Design Criteria ◽

Loose Parts

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