Calculation of Local Stress and Fatigue Resistance due to Thermal Stratification on Pressurized Surge Line Pipe

Journal Article · Tue Jun 22 00:00:00 EDT 2010 · AIP Conference Proceedings

DOI:https://doi.org/10.1063/1.4757165· OSTI ID:21410488

Bandriyana, B ^[1]; Utaja, ^[2]

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1. Center for Nuclear Industry Materials- National Nuclear Energy Agency, Serpong (Indonesia)
2. Center for Nuclear Engineering Development-BATAN, Serpong (Indonesia)

Thermal stratification introduces thermal shock effect which results in local stress and fatigue problems that must be considered in the design of nuclear power plant components. Local stress and fatigue calculation were performed on the Pressurize Surge Line piping system of the Pressurize Water Reactor of the Nuclear Power Plant. Analysis was done on the operating temperature between 177 to 343 deg. C and the operating pressure of 16 MPa (160 Bar). The stagnant and transient condition with two kinds of stratification model has been evaluated by the two dimensional finite elements method using the ANSYS program. Evaluation of fatigue resistance is developed based on the maximum local stress using the ASME standard Code formula. Maximum stress of 427 MPa occurred at the upper side of the top half of hot fluid pipe stratification model in the transient case condition. The evaluation of the fatigue resistance is performed on 500 operating cycles in the life time of 40 years and giving the usage value of 0,64 which met to the design requirement for class 1 of nuclear component. The out surge transient were the most significant case in the localized effects due to thermal stratification.

• OSTI ID: 21410488
• Journal Information: AIP Conference Proceedings, Vol. 1244, Issue 1; Conference: ICANSE 2009: 2. international conference on advances in nuclear science and engineering 2009, Bandung (Indonesia), 3-4 Nov 2009; Other Information: DOI:; (c) 2010 American Institute of Physics; ISSN 0094-243X
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND TECHNOLOGY
COMPUTERIZED SIMULATION
DESIGN
EVALUATION
FATIGUE
FINITE ELEMENT METHOD
FLUIDS
NUCLEAR POWER PLANTS
PRESSURE RANGE MEGA PA
PWR TYPE REACTORS
STRATIFICATION
STRESSES
SURGES
THERMAL SHOCK
TRANSIENTS
TWO-DIMENSIONAL CALCULATIONS
WATER
CALCULATION METHODS
ENRICHED URANIUM REACTORS
HYDROGEN COMPOUNDS
MATHEMATICAL SOLUTIONS
MECHANICAL PROPERTIES
NUCLEAR FACILITIES
NUMERICAL SOLUTION
OXYGEN COMPOUNDS
POWER PLANTS
POWER REACTORS
PRESSURE RANGE
REACTORS
SIMULATION
THERMAL POWER PLANTS
THERMAL REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS