scholarly journals Design/Installation and Structural Integrity Assessment of the Bethel Valley Low-Level Waste Collection and Transfer System Upgrade for Building 3544 (Process Waste Treatment Plant) at Oak Ridge National Laboratory, Oak Ridge, Tennessee

1996 ◽  
Author(s):  
Keyword(s):  
Structural Integrity ◽  
Waste Treatment ◽  
National Laboratory ◽  
Treatment Plant ◽  
Transfer System ◽  
Waste Collection ◽  
Oak Ridge ◽  
Integrity Assessment ◽  
Process Waste ◽  
Waste Treatment Plant

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.

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