OS1429 Development of Water Jet Peening(1) : Development and Field application of Water Jet Peening to Nuclear Reactor Internals

2009 ◽  
Vol 2009 (0) ◽  
pp. 328-330
Author(s):  
Fujio Yoshikubo ◽  
Kouichi Kurosawa ◽  
Akihiro Kanno ◽  
Takashi Inada ◽  
Noboru Saitou ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Field Application ◽  
Water Jet ◽  
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.

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

Helium Effects on 316L Austenitic Stainless Steel Fracture Mechanism

2016 ◽  
Vol 713 ◽  
pp. 228-231 ◽  
Author(s):  
I. Villacampa ◽  
Jia Chao Chen ◽  
Philippe Spätig ◽  
Hans Peter Seifert ◽  
F. Duval
Keyword(s):  
Fracture Mechanism ◽  
Nuclear Reactor ◽  
Pressurized Water ◽  
Term Operation ◽  
316L Austenitic Stainless Steel ◽  
Radiation Induced ◽  
Water Reactors ◽  
High Doses ◽  
Reactor Internals

The most common fracture mechanism of nuclear reactor internals is irradiation-assisted stress corrosion cracking (IASCC). Its susceptibility at relatively low dose is dominated by conventional mechanisms such as radiation-induced segregation and radiation hardening. However, the aging of the nuclear fleet combined with the increase of their life-span reveals other mechanisms that could play an important role on IASCC susceptibility. A large amount of helium (He) can be accumulated in reactor internal components of pressurized water reactors (PWR) after long term operation. This occurrence could significantly increase (or even dominate) the IASCC susceptibility at high doses. He has been homogeneously implanted in an especially designed miniaturized specimen at 300°C up to 1000 appm. Slow strain rate tests (SSRT) results in high temperature air and in simulated PWR conditions indicate that homogenized, as-implanted He does not have a significant effect on IASCC up to 1000 appm under these test conditions.

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

Development of Technologies on Innovative-Simplified Nuclear Power Plant Using High-Efficiency Steam Injectors: Part 12—Evaluations of Spatial Distributions of Flow and Heat Transfer in Steam Injector

2006 ◽  
Author(s):  
Yutaka Abe ◽  
Yujiro Kawamoto ◽  
Chikako Iwaki ◽  
Tadashi Narabayashi ◽  
Michitsugu Mori ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Heat Exchanger ◽  
Flow Rate ◽  
Direct Contact ◽  
Nuclear Reactor ◽  
High Efficiency ◽  
Flow Behavior ◽  
Water Jet ◽  
Next Generation ◽  
Contact Condensation

Next-generation nuclear reactor systems have been under development aiming at simplified system and improvement of safety and credibility. One of the innovative technologies is the supersonic steam injector, which has been investigated as one of the most important component of the next-generation nuclear reactor. The steam injector has functions of a passive pump without large motor or turbo-machinery and a high efficiency heat exchanger. The performances of the supersonic steam injector as a pump and a heat exchanger are dependent on direct contact condensation phenomena between a supersonic steam and a sub-cooled water jet. In previous studies of the steam injector, there are studies about the operating characteristics of steam injector and about the direct contact condensation between static water pool and steam in atmosphere. However, there is a little study about the turbulent heat transfer and flow behavior under the great shear stress. In order to examine the heat transfer and flow behavior in supersonic steam injector, it is necessary to measure the spatial temperature distribution and velocity in detail. The present study, visible transparent supersonic steam injector is used to obtain the axial pressure distributions in the supersonic steam injector, as well as high speed visual observation of water jet and steam interface. The experiments are conducted with and without non-condensable gas. The experimental results of the interfacial flow behavior between steam and water jet are obtained. It is experimentally clarified that an entrainment exists on the water jet surface. It is also clarified that discharge pressure is depended on the steam supply pressure, the inlet water flow rate, the throat diameter and non-condensable flow rate. Finally a heat flux is estimated about 19MW/m2 without non-condensable gas condition in steam.

The Effects of Fluid Structure Interactions on the Dynamic Characteristics of the Reactor Internals Structure of SMART

2015 ◽  
Vol 752-753 ◽  
pp. 851-858
Author(s):  
Jang Won Lee ◽  
Young Shin Lee
Keyword(s):  
Nuclear Reactor ◽  
Lanczos Method ◽  
Fluid Structure Interactions ◽  
Energy Research ◽  
Environment Friendly ◽  
Smart System ◽  
Solid Models ◽  
Block Lanczos Method ◽  
Operational Time ◽  
Reactor Internals

The Korea Atomic Energy Research Institutes has been developing the SMART (System integrated Modular Advanced ReacTor), an environment-friendly nuclear reactor, for the generation of electricity, and to perform desalination. SMART reactors can be exposed to various external and internal loads caused by seismic and coolant flows. Reactor structures need to be maintained for the reactor’s safety and integrity against these loads during the operational time of the SMART. This paper presents two FE-models, 3-D solid models of the reactor internals in the air and in the coolant, and then compares the results of the dynamic characteristic for the two FE-models using a commercial Finite Element tool, ANSYS V12. A solver was selected by the Block Lanczos method. These FE-models are looking forward to being executed in various researches concerning the SMART in further studies.

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