00/02548 Global rupture of a nuclear reactor lower head with elevated temperature due to severe accidents

2000 ◽  
Vol 41 (5) ◽  
pp. 286
Author(s):  
Qiqi Yan ◽  
Simin Luo ◽  
Yapei Zhang ◽  
Limin Liu ◽  
Guanghui Su ◽  
...  

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.


2018 ◽  
Vol 4 (0) ◽  
pp. 18-00038-18-00038 ◽  
Author(s):  
Hiroshi MADOKORO ◽  
Alexei MIASSOEDOV ◽  
Thomas SCHULENBERG

Author(s):  
K. H. Deng ◽  
Y. Zhang ◽  
C. L. Wang ◽  
Y. P. Zhang ◽  
W. X. Tian ◽  
...  

After the severe accident inside a nuclear reactor, the IVR (In-vessel retention) management strategy is an effective way to keep the integrity of pressure vessel and reduce risk of radioactive leakage by holding the damaged core materials through External Reactor Vessel Cooling (ERVS). The damaged core materials aggregate in the lower head of pressure vessel and releasing heat to the lower head. Therefore, it is very important to remove heat timely to keep the integrity of pressure vessel by ERVS. The shape of lower head is hemispherical and the local Critical Heat Flux (CHF) of different parts changed with latitude. In this paper, influence of orientation angles, area and length-width ratio on CHF of plate heating surface for saturated pool boiling is investigate experimentally. The results show that CHF increases with increasing orientation angles and decreasing area, meanwhile, length-width ratio has a significantly effect on CHF.


Author(s):  
Hiroshi Ogawa ◽  
Hideo Machida ◽  
Naoto Kasahara

As the important lessons learned from Fukushima-nuclear power plant accident, mitigation of failure consequences and prevention of catastrophic failure were strongly recognized against severe accidents (SA) and excessive earthquake conditions. To improve mitigation measures and accident management, clarification of failure behaviors with locations is premised under design extension conditions (DEC) such as severe accidents and earthquakes. Design extension conditions induce some different failure modes from design conditions. Furthermore, the best estimation for these failure modes is required for preparing countermeasures and management. Therefore, this study focused on identification failure modes under design extension conditions. To realize best estimation, it is prerequisite to clarify failure modes with ultimate structural strength under extreme loadings such as very high temperature, pressure and great earthquakes. The authors attempt to clarify unclear failure mechanisms by extreme loadings under DEC using numerical simulations. In this paper, relations between failure modes and extreme loadings were investigated by the numerical simulation using the cylindrical model which is a typical structure of nuclear reactor structures (for example, Formed Head, Nozzle, Instrument Tube, Guide Tube, Support Skirt, etc.). Moreover, it was shown that failure modes change with an effect of structural discontinuities. Local failure dominates than ductile fracture at locally constraint portions where stress triaxiality becomes high.


Author(s):  
Young J. Oh ◽  
Kwang J. Jeong ◽  
Byung G. Park ◽  
Il S. Hwang

Most past studies for the creep rupture of a nuclear reactor pressure vessel (RPV) lower head under severe accident conditions, have focused on global deformation and rupture modes. Limited efforts were made on local failure modes associated with penetration nozzles as a part of TMI-2 Vessel Investigation Project (TMI-2 VIP) in 1990’s. However, it was based on an excessively simplified shear deformation model. In the present study, the mode of nozzle failures is investigated using data and nozzle materials from Sandia National Laboratory’s Lower Head Failure Experiment (SNL-LHF). Crack-like separations were revealed at the nozzle weld metal to RPV interfaces indicating the importance of normal stress component rather than the shear stress in the creep rupture. Creep rupture tests were conducted for nozzle and weld metal materials, respectively, at various temperature and stress levels. Stress distribution in the nozzle region is calculated using elastic-viscoplastic finite element analysis (FEA) using the measured properties. Calculation results are compared with earlier results based on the pure shear model of TMI-2 VIE It has been concluded from both LHF-4 nozzle examination and FEA that normal stress at the nozzle/lower head interface is the dominant driving force for the local failure with its likelihood significantly greater than previously assumed.


1976 ◽  
Vol 98 (2) ◽  
pp. 105-110
Author(s):  
J. M. Steichen ◽  
R. L. Knecht

The elevated temperature mechanical properties of large diameter (28 in.) seamless pipe produced by roll extrusion for use as primary piping for sodium coolant in the Fast Flux Test Facility (FFTF) have been characterized. The three heats of type 316H stainless steel piping material used in this study exhibited very consistent mechanical properties and chemical compositions. Tensile and creep-rupture properties exceed values on which the allowable stresses for ASME Code Case 1592 on Nuclear Components in Elevated Temperature Service were based. Tensile strength and ductility were essentially unchanged by aging in static sodium at 1050°F (566°C) for times to 10,000 hr. High strain rate tensile tests showed that tensile properties were insensitive to strain rate at temperatures to 900°F (482°C) and that for temperatures of 1050°F (566°C) and above both strength and ductility significantly increased with increasing strain rate. Fatigue-crack propagation properties were comparable to results obtained on plate material and no differences in crack propagation were found between axial and circumferential orientations.


2018 ◽  
Vol 123 ◽  
pp. 372-384 ◽  
Author(s):  
Md Saifur Rahman ◽  
Jie Ding ◽  
Ali Beheshti ◽  
Xinghang Zhang ◽  
Andreas A. Polycarpou

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