scholarly journals MODAL CHARACTERISTIC ANALYSIS OF THE APR1400 NUCLEAR REACTOR INTERNALS FOR SEISMIC ANALYSIS

2014 ◽  
Vol 46 (5) ◽  
pp. 689-698 ◽  
Author(s):  
JONG-BEOM PARK ◽  
YOUNGIN CHOI ◽  
SANG-JEONG LEE ◽  
NO-CHEOL PARK ◽  
KYOUNG-SU PARK ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Seismic Analysis ◽  
Characteristic Analysis ◽  
Modal Characteristic ◽  
Reactor Internals

Ultrasound and Nonlinear Resonant Testing of Nuclear Reactor Internals Bolts

2011 ◽  
Author(s):  
S. W. Glass ◽  
B. Thigpen ◽  
J. Renshaw
Keyword(s):  
Phased Array ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Cost Effective ◽  
Life Extension ◽  
Safe Operation ◽  
Plant Life ◽  
Nuclear Plants ◽  
Bolt Failure ◽  
Reactor Internals

As many nuclear plants approach the end of their initial 40 year license period, inspection or replacement of their reactor internals bolts must be considered. This is consistent with the Materials Reliability Program (MRP 227/228) guideline for plant life extension [1,2]. Assurance of the internals structural integrity is essential for continued safe operation of these plants. If there is no suspicion or indication of bolt failure, simple inspection is normally more cost-effective than replacement. Inspection vendors have inspected thousands of internals bolts with conventional and Phased Array UT but different head configurations and bolt capture mechanisms mandate specific qualifications for each bolt type. In some cases, complex bolt and head geometries coupled with counter-bore and locking bar interferences render classical UT inspections difficult or impossible. A range of solutions to inspect reactor internals including these difficult-to-inspect-by-conventional-UT baffle bolts has been developed by several vendors [3]. This presentation references developments to make bolt inspection a relatively quick and easy task through adaptations to the SUSI submarine inspection platform, the extensive UT qualification work suitable for conventional UT plus more recent advanced nonlinear resonant techniques to distinguish between flawed or loose, vs. acceptable bolts where conventional UT cannot be applied. Initial evaluations show that these advanced techniques may have the ability to reliably detect smaller flaws than previously possible with conventional techniques as well as provide information on bolt tightness.

Seismic Analysis Approach Applied to a Small Size Next Generation Nuclear Reactor

2008 ◽  
Author(s):  
G. Forasassi ◽  
R. Lo Frano
Keyword(s):  
Nuclear Power ◽  
Nuclear Reactor ◽  
Seismic Analysis ◽  
Methodological Approach ◽  
Free Field ◽  
Time History ◽  
Future Generation ◽  
Energy System ◽  
Ground Acceleration ◽  
Input Motion

The aim of the paper is to evaluate the behaviour of a Near Term nuclear energy system example with reference to IRIS (International Reactor Innovative and Safety) project. As it is well known the development of new and future-generation nuclear power plant (Gen IV NPP) is strictly related to the sustainability, safety and reliability as well as to the proliferation resistance. In this paper, the safety aspects related to the effects of a severe earthquake (Safe Shutdown Earthquake) as well as to the induced loads are treated by means the Substructure and Time History Approaches, assuming a free field Peak Ground Acceleration equal to 0.3 g as input motion. The analyses and upgrading of the geometry structures with highest probability of criticality are performed on rather complex and detailed 3D finite element (FE) models. The main goals were: the evaluation of the dynamic characteristics of each considered structure, the verification of the load bearing structures in order to obtain a preliminary assessment of the adopted methodological approach and structural models. The analyses results and dynamic response of internal components (e.g. Nuclear Buildings, etc.) seem to confirm the possibility to upgrade the geometry and the performances of the proposed design choices.

Modal Test of Scaled-Down Reactor Internals in SMART Considering the Fluid-Structure Interaction

2011 ◽  
Author(s):  
Seungho Lim ◽  
Kyungrok Ha ◽  
Kyoung-Su Park ◽  
No-Cheol Park ◽  
Young-Pil Park ◽  
...  
Keyword(s):  
Dynamic Characteristics ◽  
Seismic Analysis ◽  
Fluid Structure Interaction ◽  
Mass Effect ◽  
Mode Shapes ◽  
Structural Dynamic ◽  
Fluid Structure ◽  
Structure Interaction ◽  
Standard Design ◽  
Reactor Internals

The System-integrated Modular Advanced ReacTor (SMART) is a small modular integral-type reactor for the seawater desalination and small-scaled power generation under development in Korea. Although the SMART is innovative reactor with a sensible mixture of the proven technology and advanced design features aimed at enhanced safety, there is no valid prototype which can specify the structural dynamic characteristics of reactor internals. Thus, extensive research for the technology verification and standard design approval are in progress. One of them is to perform the dynamic characteristics identification of reactor internals. Especially, it is focused on the added mass effect caused by the fluid-structure interaction because the reactor internals is submerged in the reactor coolant. The extracted dynamic characteristics such as the natural frequencies and the vibratory mode shapes can be used as the basis on further dynamic analysis, for example, seismic analysis and a postulated pipe break analysis.

Seismic Analysis of a Nuclear Reactor With Fluid-Structure Interaction

2005 ◽  
Author(s):  
Jean-Franc¸ois Sigrist ◽  
Daniel Broc ◽  
Christian Laine
Keyword(s):  
Internal Structure ◽  
Pressure Vessel ◽  
Nuclear Reactor ◽  
Seismic Analysis ◽  
Mass Operator ◽  
Fluid Structure Interaction ◽  
Added Mass ◽  
Operating Conditions ◽  
Fluid Structure ◽  
Structure Interaction

The present paper is related to a seismic analysis of a naval propulsion ground prototype nuclear reactor with fluid-structure interaction modeling. Many numerical methods have been proposed over the past years to take fluid/structure phenomenon into account [14] in various engineering domains, among which nuclear engineering in seismic analysis [15]. The purpose of the present study is to apply general methods on a global approach of the nuclear reactor. A simplified design of the pressure vessel and the internal structure is presented; fluid-structure interaction is characterized by the following effects: • added mass effects are highlighted with the calculation of an added mass operator, obtained from a finite element discretisation of the coupled problem. The numerical model is developed within the CASTEM code using an axi-symmetric model of the industrial structure; • coupling effects between the external and internal structure via the confined inner fluid are also illustrated and numerically described with the added mass operator; • added stiffness effects are taken into account with an added stiffness matrix describing pre-stress effects due to a static pressure loading simulating the actual operating conditions of the reactor. The expected fluid-structure interaction effects on the nuclear pressure vessel and their numerical modeling leads to the definition of a global coupled model which can be used to perform a seismic analysis. A modal analysis is first performed and classical linear methods (static, spectral and temporal) are then applied on the studied structure with taking fluid-structure into account.

Laser Peening Technology as the SCC Mitigation for Reactor Internals

2011 ◽  
Author(s):  
Hiroyuki Miyasaka ◽  
Masaki Yoda ◽  
Itaru Chida ◽  
Tatsuya Ishigaki
Keyword(s):  
Nuclear Reactor ◽  
Nuclear Reactors ◽  
Austenitic Stainless Steels ◽  
Laser Pulses ◽  
Nuclear Industry ◽  
Impulsive Effect ◽  
Laser Peening ◽  
Bottom Part ◽  
New Process ◽  
Reactor Internals

Stress Corrosion Cracking (SCC) is one of the major concerns for aged nuclear reactors. SCC-susceptible materials have been employed in a wide variety of applications in the nuclear industry. Laser Peening (LP) is a method for the SCC mitigation that eliminates surface tensile stress using impulsive effect of high-pressure plasma induced by irradiation with high-power laser pulses in the water. To apply LP to nuclear reactor internals, Toshiba has developed a new process which needs no protective coating on the materials and optimizes the conditions for the laser irradiation. Its effects for stress improvement and SCC-mitigation of laser-peened materials were confirmed through SCC tests for austenitic stainless steels and nickel-based alloys. Also integrity of the laser-peened materials was confirmed through various examinations. Toshiba developed the LP system for the core shroud and the reactor bottom part of BWRs, and has been applying it to actual Japanese nuclear reactors since 1999. For PWRs, Toshiba developed the system for Bottom-Mounted Instrumentation (BMI) nozzles and other Reactor Vessel (RV) nozzles, and has been applying it to Japanese PWRs since 2004. So far Toshiba has already completed LP operations for 2 PWR plants and 8 BWR plants in Japan. In consideration of extending the LP system to BWRs and PWRs overseas, a portable LP system equipped with a small size laser oscillator has been developed. We confirm the possibility that the portable LP system makes the outage period shorter.

LASER-ULTRASONIC TESTING AND ITS APPLICATIONS TO NUCLEAR REACTOR INTERNALS

2008 ◽  
Author(s):  
M. Ochiai ◽  
T. Miura ◽  
S. Yamamoto ◽  
Donald O. Thompson ◽  
Dale E. Chimenti
Keyword(s):  
Ultrasonic Testing ◽  
Nuclear Reactor ◽  
Laser Ultrasonic ◽  
Reactor Internals

Simplified Thermal Analysis for Embedded Piping

1992 ◽  
Author(s):  
Pirooz M. H. Joodi
Keyword(s):  
Cross Section ◽  
Pressure Vessel ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Seismic Analysis ◽  
Elementary Theory ◽  
Thermal Characteristics ◽  
Soil Parameters ◽  
Tubular Structures ◽  
Hollow Beam

Abstract The growing use of tubular structures in nuclear reactor facilities such as pipes, conduits and ducts that are buried underground, requires more detailed stress analysis to demonstrate structural integrity as required by Section III of the Boiler Pressure Vessel Code and other applicable industry codes. Thermal behavior of the pipe and soil interference can be conservatively evaluated by implementing the thermal characteristics and properties of the pipe into the expressions, as deduced by previous studies, which are made for the seismic analysis of buried piping. This paper presents procedures to evaluate the different pipe/soil parameters to be applied in those expressions, and explains these equations from designers perspectives and, finally, suggests an approach to combine various pipe stresses to check against ASME Boiler and Pressure Vessel Code Section III. The analysis assumes that the soil is linearly elastic and homogenous and the structure is a straight slender solid or hollow beam with a uniform, symmetrical cross section that satisfies the conditions of the elementary theory of beams on elastic foundations.

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