First private nuclear reactor operating

1956 ◽  
Vol 262 (3) ◽  
pp. 243-245
Keyword(s):  
Nuclear Reactor ◽  
Reactor Operating

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).

scholarly journals Patchy nuclear chain reactions

2021 ◽  
Vol 4 (1) ◽  
Author(s):  
Eric Dumonteil ◽  
Rian Bahran ◽  
Theresa Cutler ◽  
Benjamin Dechenaux ◽  
Travis Grove ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Reactor Core ◽  
Rensselaer Polytechnic Institute ◽  
Operating Conditions ◽  
Chain Reactions ◽  
Branching Random Walks ◽  
Stochastic Fluctuations ◽  
Nuclear Chain ◽  
Reactor Operating ◽  
Experimental Findings

AbstractStochastic fluctuations of the neutron population within a nuclear reactor are typically prevented by operating the core at a sufficient power, since a deterministic (i.e., exactly predictable) behavior of the neutron population is required by automatic safety systems to detect unwanted power excursions. In order to characterize the reactor operating conditions at which the fluctuations vanish, an experiment was designed and took place in 2017 at the Rensselaer Polytechnic Institute Reactor Critical Facility. This experiment however revealed persisting fluctuations and striking patchy spatial patterns in neutron spatial distributions. Here we report these experimental findings, interpret them by a stochastic modeling based on branching random walks, and extend them using a “numerical twin” of the reactor core. Consequences on nuclear safety will be discussed.

scholarly journals A Basis for Applying Elastic Perfectly-Plastic Design Methods to Cyclic Softening Materials

2018 ◽  
Author(s):  
M. C. Messner ◽  
Yanli Wang ◽  
R. I. Jetter ◽  
T.-L. Sham
Keyword(s):  
Yield Stress ◽  
Nuclear Reactor ◽  
Cyclic Hardening ◽  
Cyclic Softening ◽  
Perfectly Plastic ◽  
Ratcheting Strain ◽  
Material State ◽  
Operating Temperatures ◽  
Elastic Perfectly Plastic ◽  
Reactor Operating

Design approaches using elastic perfectly-plastic (EPP) analysis have recently been approved as Code Cases for the Section III, Division 5 design of high-temperature nuclear reactor components made from austenitic stainless steel. These methods bound the ratcheting strain and creep-fatigue damage accumulated over the life of a component with a simplified, elastic-perfectly plastic analysis using a special pseudo-yield stress — often not equal to the true material yield stress. The austenitic materials specified in the existing Code cases are cyclic-hardening for all allowable operating temperatures. However, other Section III, Division 5 materials, such as Grade 91 steel, are cyclic softening at expected advanced reactor operating temperatures. This work describes the extension of EPP methods to cyclic softening materials through the use of a postulated saturated material state and softening factors to be applied to the pseudo yield stress. We demonstrate the conservatism of the modified EPP method against a series of inelastic simulations of two bar tests, using a constitutive model that captures work and cyclic softening.

scholarly journals CONCEPTUAL DESIGN OF AN ORGANIC-COOLED SMALL NUCLEAR REACTOR TO SUPPORT ENERGY DEMANDS IN REMOTE LOCATIONS IN NORTHERN CANADA

2020 ◽  
Vol 9 (1) ◽  
pp. 39-44 ◽  
Author(s):  
Colin Shannon ◽  
Paul Chan ◽  
H.W. Bonin
Keyword(s):  
Conceptual Design ◽  
Nuclear Reactor ◽  
Electrical Power ◽  
District Heating ◽  
Northern Canada ◽  
Remote Locations ◽  
Energy Demands ◽  
Reactivity Coefficients ◽  
Enriched Uranium ◽  
Reactor Operating

Small nuclear reactors can offer safe, reliable, and long-lasting district heating and electrical power generation to remote locations in northern Canada. A conceptual design of an organic-cooled and moderated reactor based upon the SLOWPOKE-2 research reactor is proposed for potential employment in northern Canada. For viability, this design extends the SLOWPOKE-2’s power to 1 MWth. An added pump circulates the organic coolant, a partially hydrogenated terphenyl mixture known as HB-40, to facilitate greater heat transfer. The reactor incorporates the same low-enriched uranium dioxide fuel as the SLOWPOKE-2. Reactor control is accomplished through hafnium absorber rods and a movable beryllium reflector. The reactor neutronics are simulated using the deterministic code, WIMS-AECL, and the probabilistic code, MCNP 6. The service life of fuel in this reactor operating at full power exceeds 11 years. The conceptual design has demonstrated negative reactivity coefficients indicating strong potential for inherent safety.

scholarly journals Changes of the Neutron Flux of the Nuclear Reactor Triga Mark III Since the Conversion from High to Low 235U Enrichment

2021 ◽  
Vol 8 (2) ◽  
pp. 149-153
Author(s):  
C. Vázquez-López ◽  
O. Del Ángel-Gómez ◽  
R. Raya-Arredondo ◽  
S. S. Cruz-Galindo ◽  
J. I. Golzarri-Moreno ◽  
...  
Keyword(s):  
Neutron Flux ◽  
Nuclear Reactor ◽  
Research Reactor ◽  
Nuclear Research ◽  
Reactor Power ◽  
Uranium Fuel ◽  
Acrylic Sheet ◽  
Enriched Uranium ◽  
Reactor Operating ◽  
Fuel Elements

The neutron flux of the Triga Mark III research reactor was studied using nuclear track detectors. The facility of the National Institute for Nuclear Research (ININ), operates with a new core load of 85 LEU 30/20 (Low Enriched Uranium) fuel elements. The reactor provides a neutron flux around 2 × 1012 n cm-2s-1 at the irradiation channel. In this channel, CR-39 (allyl diglycol policarbonate) Landauer® detectors were exposed to neutrons; the detectors were covered with a 3 mm acrylic sheet for (n, p) reaction. Results show a linear response between the reactor power in the range 0.1 - 7 kW, and the average nuclear track density with data reproducibility and relatively low uncertainty (±5%). The method is a simple technique, fast and reliable procedure to monitor the research reactor operating power levels.

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