AMPP Store - Products tagged with 'environmental degradation 2022'

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Characterization Of Materials Showing Large Differences In IASCC Response

Product Number: ED22-18360-SG

Author: Anders Jenssen, Daniel Jädernäs, Peter Chou, Kristina Lindgren, Mattias Thuvander

Publication Date: 2022

$20.00

Irradiation-assisted stress corrosion cracking (IASCC) is a term used to describe cracking of austenitic materials in components subjected to a relatively high fast neutron flux. Like intergranular stress corrosion Paper No.18360 cracking (IGSCC), IASCC appears as intergranular cracks, but thermal sensitization of the grain boundaries is not required for the material to become susceptible to cracking. Service failures caused by IASCC have occurred in components such as core shrouds and top guides in boiling water reactors (BWRs) and baffle bolts in pressurized water reactors (PWRs).

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Chemistry Change Of CRUD With Various Metal Ion Concentration Conditions In PWR Fuel Cladding

Product Number: ED22-17313-SG

Author: Yunju Lee, Seung Chang Yoo, Junhyuk Ham, Daehyeon Park, Ji Hyun Kim

Publication Date: 2022

$20.00

Since 1960s, CRUD induced problems become an issue in PWR power plant. To enhance economic efficiency of PWR, the burn-ups of nuclear fuel have been increased. In the reactor core, some unexpected power shifts, which shows negative axial offset from prediction, have been observed. This phenomenon is called by AOA (Axial offset Anomaly) and it seems to be caused by deposits on the upper side of fuel cladding surface which is called CRUD.

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Cold Spray for Repair of Nuclear Reactor System Components

Product Number: ED22-17267-SG

Author: Catherine Cmar, Michael Ickes, Arash Parsi

Publication Date: 2022

$20.00

Supersonic particle deposition also known as cold dynamic gas spray, or “Cold Spray” is a materials deposition process. During the cold spray process a gas stream, typically helium or nitrogen, is split into two streams where 90% of the gas is sent to an electric heater and 10% is sent to a powder feeder. The powder feeder contains powder composed of small metallic particles, or blends of metallic and non-metallic particles, ranging from 5 to 100 μm in size.

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Comparing Stress Corrosion Cracking Initiation Times of Sensitized 304 Stainless Steel in KOH And LiOH Primary Water Chemistries with High Levels Of Dissolved Oxygen

Product Number: ED22-18341-SG

Author: Stuart Medway, Peter Chou, David Tice

Publication Date: 2022

$20.00

The testing described in this paper is part of a wider initiative by the Electric Power Research Institute (EPRI) to perform a due-diligence assessment to support possible application to plants and demonstrate the use of potassium hydroxide (KOH) in western Pressurized Water Reactors (PWRs). Lithium hydroxide (LiOH) is used in the primary coolant loop of PWRs to modify the pH of the coolant water. LiOH is most commonly used as the alkalizing substance as 7Li is already present in the water
chemistry as a by-product of the neutron reaction with boron (10B). To reduce the risk of accelerated corrosion of the Zircaloy fuel cladding material, there is an upper limit of 3.5 ppm for 7Li, although standard starting chemistry is typically 2 ppm. The amount of lithium is subsequently reduced during operation as the required level of boron is reduced due to fuel burnup through the fuel lifecycle.

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Contribution Of Cathodic Reaction Inside Crevice On Propagation Of Crevice Corrosion Of 304L SS In Chloride Solution

Product Number: ED22-17321-SG

Author: Kazuki Yakata, Yutaka Watanabe, Yuichi Fukaya

Publication Date: 2022

$20.00

On March 2011 at Fukushima Daiichi NPS, seawater was injected into spent fuel pools just after the accident for emergency cooling. The temperature of the water in a pool raised up to 93 ℃, and the chloride ion concentration raised up to 1,944 ppm (maximum) after seawater injection. In this high temperature and high chloride ion concentration environment, localized corrosion including crevice corrosion may have occurred on components made of passive metals such as stainless steels. The environment is assumed to be susceptible to crevice corrosion for 304 SS based on laboratory experiments and the concept of ER, CREV. There is a low possibility of initiating localized corrosion after the water was purified and deoxygenated. However, it is not certain whether localized corrosion, once initiated and propagated under the severe condition, will repassivate after the bulk water is purified. It is necessary to examine the continuity of crevice corrosion propagation when the bulk water is purified and deaerated, which means the cathodic reactions outside the crevice will no longer contribute to the propagation of the corrosion.

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Corrosion Behavior Of Welded Austenitic Stainless Steels In Coastal Atmospheric Environment

Product Number: ED22-18359-SG

Author: Mei-Lin Chen, Tsung-Kuang Yeh, Mei-Ya Wang

Publication Date: 2022

$20.00

Austenitic stainless steels (SS) are the candidate materials for dry storage canisters. In general, high corrosion resistance is one of the advantageous properties of austenitic stainless steels. However, austenitic stainless steels are susceptible to chloride induced stress corrosion cracking (CISCC) when three criteria are satisfied: susceptible materials, tensile stress and corrosive environments. Dry storage canisters are located at a coastal site in Taiwan, as a result, canisters are exposed to high concentration of moisture and chloride salts under atmospheric environments.

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Crack Growth Behaviour Of Cobalt Based Alloy In High Temperature Water

Product Number: ED22-17294-SG

Publication Date: 2022

$20.00

Co-based alloy (30Cr-4W-Bal.Co), such as Stellite TM grade 6 (UNS No. ERCOCr-A) valve seat, is used extensively in applications where superior resistance to wear corrosion are required. But there are only rare data about crack growth behavior of the alloy in the high temperature water. It is difficult to extract specimen from as welded Stellite-6, because of the valve sheet did not have enough volume. So, to evaluate the crack growth behavior and its temperature dependence, crack growth rate measurements were performed using forged Co-based alloy bar (Stellite TM grade 6B) (UNS No. R30016) in a simulated PWR primary water at temperature ranging from 250 to 320°C using half-inch size compact tension specimens (1/2TCT).

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Crack Growth Rate Testing Of Alloy 82 In BWR Environments

Product Number: ED22-17295-SG

Author: Johan Stjärnsäter, Katsuhiko Kumagai

Publication Date: 2022

$20.00

The nickel base weld metal Alloy 82 is used in various applications in boiling water reactors (BWRs). Applications that are vital from a safety point of view are e.g., welds between core shroud support legs and the reactor pressure vessel (RPV), and feedwater nozzle to safe end welds. Laboratory testing and service history have shown that Alloy 82 is susceptible to stress corrosion cracking (SCC) in BWR environments. However, in comparison with Alloy 182, fewer failure cases have been reported, which could be related to the higher Cr content in Alloy 82 (~ 15 vs. ~ 20 %). It is also possible that the higher frequency of SCC in Alloy 182 is related to the wider use of this weld metal, and the larger surface area exposed to reactor water. Given the lower frequency of failures in Alloy 82, the database regarding SCC in BWR environments is much larger for Alloy 182.

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Crack Initiation Testing of Neutron-irradiated Type 347 Stainless Steel in Simulated PWR with LiOH or KOH

Product Number: ED22-17414-SG

Author: Michael Ickes, Lijin Dong, Abdullah Sinjlawi, Gary Was

Publication Date: 2022

$20.00

The immediate objective of this experiment is to investigate the IASCC initiation behavior of Type 347 stainless steel in lithium hydroxide and potassium hydroxide water chemistries across a range of irradiation damage and stress levels. A further objective is to provide data supporting improved predictive capabilities of IASCC failures by assessing the radiation dose dependence of IASCC initiation. In power plant components like the baffle-former bolts, the crack initiation step of IASCC is the rate limiting step, taking much longer than crack propagation as a fraction of time to failure. The results of this study will also be directly beneficial to the U.S. nuclear industry by providing an understanding of IASCC susceptibility in potassium hydroxide water chemistry, which may provide cost savings and more secure supply chains to nuclear power plants.

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Cracking Behavior Of Decommissioned Baffle Plates In Light Water Reactor Environment

Product Number: ED22-17301-SG

Author: Yiren Chen, Bogdan Alexandreanu, Appajosula S. Rao

Publication Date: 2022

$20.00

Most of the core internal components of light water reactors (LWRs) are made of austenitic stainless steels (SSs). Exposed to fast neutron irradiation, reactor core internal components are vulnerable to irradiation-induced or irradiation-enhanced degradations. At LWR-relevant temperatures and irradiation doses, significant microstructural and microchemical changes can take place in SSs, leading to deteriorated mechanical and corrosion properties.

Dealloying And Its Relationship To SCC Of Alloy 800 And Ni-Fe-Cr Alloys In Boiling Caustic Solutions

Product Number: ED22-17138-SG

Author: Hooman Gholamzadeh, Adil Shaik, Bander Alsekhan, Kevin Daub, Matthew Topping, Mark R. Daymond, Suraj Persaud

Publication Date: 2022

$20.00

Ni-based alloys and stainless steels have superior mechanical properties and good resistance to general and localized corrosion, mainly due to the formation of a passive film. Due to their properties, Ni-based
alloys and stainless steels have been historically used in applications where an aggressive environment is involved. For example, Ni- and Fe- based alloys have been extensively used in the nuclear power
industry. Despite their good corrosion performance, these materials have been shown to suffer from environmentally assisted cracking (EAC) in certain environments.

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Determination of Activation Energies for Nuclear Electrical Cable Insulation Using Density and Total Color Difference

Product Number: ED22-17336-SG

Author: D.Li, Y.Ni, A.Zwoster, M.R. Pallaka, M.Spencer, L.S. Fifield

Publication Date: 2022

$20.00

Electrical cables are critical components of nuclear power plants (NPPs) that ensure their safe operation. In use they are exposed to a wide range of environmental stresses, such as elevated temperature, gamma radiation, moisture, and electrical stress. Among the different environmental stresses, elevated temperature leads to the thermo-oxidative degradation of electrical cable insulation, which may cause early failure of electrical cables and subsequently lead to unplanned plant shutdowns, electrical transients, and/or a loss of safety redundancy.

Products tagged with 'environmental degradation 2022'

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