Effect of BR-5-reactor operating conditions on the development of vacancy porosity in 1Kh18N9T steel

1977 ◽  
Vol 43 (3) ◽  
pp. 834-835 ◽  
Author(s):  
S. I. Porollo ◽  
V. I. Shcherbak ◽  
N. N. Aristarkhov ◽  
V. N. Bykov ◽  
V. D. Dmitriev ◽  
...  
Keyword(s):  
Operating Conditions ◽  
1Kh18n9t Steel ◽  
Reactor Operating

Formation of Local, Transient “Acid Spikes” in the Fast Neutron Radiolysis of Supercritical Water at 400 °C: A Potential Source of Corrosion in Supercritical Water-Cooled Reactors?

2020 ◽  
Vol 6 (3) ◽  
Author(s):  
Md Mohsin Patwary ◽  
Sunuchakan Sanguanmith ◽  
Jintana Meesungnoen ◽  
Jean-Paul Jay-Gerin
Keyword(s):  
Fast Neutron ◽  
Supercritical Water ◽  
Reactor Core ◽  
Radiation Chemistry ◽  
Operating Conditions ◽  
Monte Carlo Simulation Study ◽  
Proper Understanding ◽  
Gen Iv ◽  
Reactor Operating

Abstract The use of supercritical water (SCW) in GEN IV reactors is a logical approach to the ongoing development of nuclear energy. A proper understanding of the radiation chemistry and reactivities of transients in a reactor core under SCW conditions is required to achieve optimal water chemistry control and safety. A Monte Carlo simulation study of the radiolysis of SCW at 400 °C by incident 2 MeV monoenergetic neutrons (taken as representative of a fast neutron flux in a reactor) was carried out as a function of water density between ∼150 and 600 kg/m3. The in situ formation of H3O+ by the generated recoil protons was shown to render the “native” track regions temporarily very acidic (pH ∼ 1). This acidity, though local and transitory (“acid spikes”), raises the question whether it may promote a corrosive environment under proposed SCW-cooled reactor operating conditions that would lead to progressive degradation of reactor components.

Devices to Mitigate Gas Entrainment in the Surge Tank of PFBR

2006 ◽  
Author(s):  
D. Ramdasu ◽  
N. S. Shivakumar ◽  
G. Padmakumar ◽  
C. Anand Babu ◽  
G. Vaidyanathan
Keyword(s):  
Porous Plate ◽  
Operating Conditions ◽  
Liquid Column ◽  
Surge Tank ◽  
Secondary Circuit ◽  
Scaled Model ◽  
Gas Entrainment ◽  
Minimum Height ◽  
Continuous Feed ◽  
Reactor Operating

Surge tank is one of the important components in the secondary circuit of a typical sodium cooled fast breeder reactor, provided to take care of pressure surges in case of a sodium water reaction in Steam Generators (SG). The blanket of argon cover gas at the top of the tank acts as a cushion for the surges. The argon gas above the free surface of sodium in the tank is a source of entrainment into the sodium which is undesirable from the consideration of effective heat transfer in Inter mediate Heat Exchanger (IHX) and SG cavitation in pumps and operational problems of continuous feed and bleed of cover gas, thus leading to unfavourable reactor operating conditions. To investigate the phenomenon of gas entrainment in surge tank, hydraulic experiments were conducted in water using 1/12 scaled model. The minimum height of liquid column in the tank when gas entrainment is completely avoided was established. Different methods to mitigate gas entrainment were tested in the model and a combination of porous plate and stiffener ring was found to be optimum in reducing the liquid column required to mitigate gas entrainment in surge tank.

Scoping Analyses of Parameterized Cool-Down Transients Associated With Reactor Shutdown

2007 ◽  
Author(s):  
T. L. Dickson ◽  
M. T. EricksonKirk
Keyword(s):  
Nuclear Reactor ◽  
Structural Integrity ◽  
Operating Temperature ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Pressure Vessels ◽  
Operating Conditions ◽  
Temperature And Pressure ◽  
Reactor Operating ◽  
Technical Basis

The current regulations, as set forth by the United States Nuclear Regulatory Commission (NRC), to insure that light-water nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to planned startup (heat-up) and shutdown (cool-down) transients are specified in Appendix G to 10 CFR Part 50, which incorporates by reference Appendix G to Section XI of the ASME Code. The technical basis for these regulations contains many aspects that are now broadly recognized by the technical community as being unnecessarily conservative. During the past decade, the NRC conducted the interdisciplinary Pressurized Thermal Shock (PTS) Re-evaluation Project that established a technical basis to support a risk-informed revision to current PTS regulations (10CFR Part 50.61). Once the results of the PTS reevaluation are incorporated into a revision of the 10 CFR 50 guidance on PTS, it is anticipated that the regulatory requirements for the fracture toughness of the RPV required to withstand a PTS event (accidental loading) will in some cases be less restrictive than the current requirements of Appendix G to 10 CFR Part 50, which apply to normal operating conditions. This logical inconsistency occurs because the new PTS guidelines will be based on realistic models and inputs whereas existing Appendix G requirements contain known and substantial conservatisms. Consequently, a goal of current NRC research is to derive a technical basis for a risk-informed revision to the current requirements of Appendix G to 10 CFR Part 50 in a manner that is consistent with that used to develop the risk-informed revision to the PTS regulations. Scoping probabilistic fracture mechanics (PFM) analyses have been performed for several hundred parameterized cool-down transients to (1) obtain insights regarding the interaction of operating temperature and pressure parameters on the conditional probability of crack initiation and vessel failure and (2) determine the limits on the permissible combinations of operating temperature and pressure within which the reactor may be brought into or out of an operational condition that remains below the acceptance criteria adopted for PTS of 1 × 10−6 failed RPVs per reactor operating year. This paper discusses the modeling assumptions, results, and implications of these scoping analyses.

Study of the Influence of Microstructure and Intergranular Carbides on the Oxidation Behavior of a Nickel Base Alloy 690 TT in Supercritical Water Nuclear Reactor Conditions

2020 ◽  
Author(s):  
Alberto Sáez-Maderuelo ◽  
María Luisa Ruiz-Lorenzo ◽  
Francisco Javier Perosanz ◽  
Patricie Halodová ◽  
Jan Prochazka ◽  
...  
Keyword(s):  
Supercritical Water ◽  
Nuclear Power ◽  
Power Plants ◽  
Oxidation Behavior ◽  
Nuclear Reactor ◽  
Base Alloy ◽  
Nuclear Industry ◽  
Operating Conditions ◽  
Alloy 690 ◽  
Reactor Operating

Abstract Alloy 690, which was designed as a replacement for the Alloy 600, is widely used in the nuclear industry due to its optimum behavior to stress corrosion cracking (SCC) under nuclear reactor operating conditions. Because of this superior resistance, alloy 690 has been proposed as a candidate structural material for the Supercritical Water Reactor (SCWR), which is one of the designs of the next generation of nuclear power plants (Gen IV). In spite of this, striking results were found [1] when alloy 690 was tested without intergranular carbides. These results showed that, contrary to expectations, the crack growth rate is lower in samples without intergranular carbides than in samples with intergranular carbides. Therefore, the role of the carbides in the corrosion behavior of Alloy 690 is not yet well understood. Considering these observations, the aim of this work is to study the effect of intergranular carbides in the oxidation behavior (as a preliminary stage of degenerative processes SCC) of Alloy 690 in supercritical water (SCW) at two temperatures: 400 °C and 500 °C and 25 MPa. Oxide layers of selected specimens were studied by different techniques like Scanning Electron Microscope (SEM) and Auger Electron Spectroscopy (AES).

Ag/YBa2Cu3O7-x Powder Generation by Aerosol Decomposition

1989 ◽  
Vol 169 ◽  
Author(s):  
Toivo T. Kodas ◽  
Altaf H. Carim ◽  
Kevin C. Ott
Keyword(s):  
Melting Point ◽  
Aerosol Particles ◽  
Nitrate Solution ◽  
Metal Nitrate ◽  
Operating Conditions ◽  
Grain Sizes ◽  
Reactor Operating ◽  
Uniform Composition

AbstractYBa2Cu3O7-x (123) powders containing silver have been prepared by aerosol decomposition. Metal nitrate solution droplets were decomposed at temperatures above and below the melting point of the Ag-O eutectic. In both cases, the Ag was present as a separate grain attached to YBa2Cu3O7-x. Individual aerosol particles had dimensions of 50 – 1000 nm. Grain sizes of Ag and 123 crystallites within these particles were 10 to 100 nm. Larger 123 grain sizes could be obtained by varying the reactor operating conditions. The powders provide a source of material for generation of YBa2Cu3O7-x/Ag ceramics with smaller Ag and 123 grain sizes and more uniform composition than can be obtained by other methods.

Optimizing pyrolysis reactor operating conditions to increase nicotine recovery from tobacco leaves

2015 ◽  
Vol 112 ◽  
pp. 80-87 ◽  
Author(s):  
Mohammad M. Hossain ◽  
Ian M. Scott ◽  
Franco Berruti ◽  
Cedric Briens
Keyword(s):  
Operating Conditions ◽  
Tobacco Leaves ◽  
Pyrolysis Reactor ◽  
Reactor Operating

scholarly journals Temperature Dependence of Primary Species G(values) Formed from Radiolysis of Water by Interaction of Tritium β-Particles

2017 ◽  
Vol 19 (1) ◽  
pp. 17
Author(s):  
Sofia Loren Butarbutar ◽  
Sriyono Sriyono ◽  
Geni Rina Sunaryo
Keyword(s):  
Temperature Dependence ◽  
Nuclear Power ◽  
Reactor Core ◽  
Cooling Water ◽  
Radiation Chemistry ◽  
Operating Conditions ◽  
Water Radiolysis ◽  
Direct Measurements ◽  
Temperature Ranges ◽  
Reactor Operating

TEMPERATURE DEPENDENCE OF PRIMARY SPECIES G(VALUES) FORMED FROM RADIOLYSIS OF WATER BY INTERACTION OF TRITIUM β-PARTICLES. G(values) are important to understand the effect of radiolysis of Nuclear Power Plant (NPP) cooling water. Since direct measurements are difficult, hence modeling and computer simulation were carried out to predict radiation chemistry in and around reactor core. G(values) are required to calculate the radiation chemistry. Monte Carlo simulations were used to calculate the G(values) of primary species , H•, H2, •OH dan H2O2 formed from the radiolysis of tritium β low energy electron. These radiolytic products can degrade the reactor components and cause corrosion under the reactor operating conditions. G(values) prediction can indirectly contribute to maintain the material reliability. G(values) were calculated at 10-8, 10-7, 10-6 and 10-5 s after ionization at temperature ranges. The calculation were compared with the G(values) of g-ray 60Co. The work aimed to understand temperature effect on the water radiolysis mechanism by the tritium β electron. The results show that the trend similarity was found on the temperature dependence of G(values) of tritium β electron and g-ray 60Co. For tritium β electron, G(values) for free radical were lower than g-ray 60Co, but higher for molecular products as temperature raise at 10-8 and 10-7. The significant differences for these two type of radiations were on G(H2), G(•OH) and G(H•) at 10-6and 10-5 s above 200 oC.

Delayed Hydride Cracking Initiation at Simulated Debris Fretting Flaws in Zr-2.5Nb Alloys

2006 ◽  
Author(s):  
Jun Cui ◽  
Gordon K. Shek ◽  
Douglas A. Scarth ◽  
Zhirui Wang
Keyword(s):  
Hydrogen Diffusion ◽  
Notch Root ◽  
Threshold Value ◽  
Operating Conditions ◽  
Notch Root Radius ◽  
Pressure Tubes ◽  
Delayed Hydride Cracking ◽  
Program Test ◽  
Reactor Operating ◽  
Hydride Cracking

Flaws in Zr-2.5Nb alloy pressure tubes in CANDU nuclear reactors are susceptible to a crack initiation and growth mechanism known as Delayed Hydride Cracking (DHC), which is a repetitive process that involves hydrogen diffusion, hydride precipitation, growth and fracture of the hydrided region at the flaw-tip. In-service flaw evaluation requires an analysis to demonstrate DHC will not initiate from the flaw. The work presented in this paper examines DHC initiation behavior from simulated debris fretting flaws. Groups of cantilever beam specimens containing V-notches with root radii of 15, 30 and 100 μm were prepared from two unirradiated pressure tubes hydrided to a nominal hydrogen concentration of 57 wt. ppm. The specimens were loaded to different stress levels that straddled the threshold value predicted by an engineering model, and subjected to multiple thermal cycles relevant to reactor operating conditions to form hydrides at the flaw-tip. Threshold conditions for DHC initiation were established for the flaw geometries and thermal cycling conditions used in this program. Test results indicate that the susceptibility to DHC initiation was affected by material variability and notch root radius. The results are also compared with model predictions.

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