The method of successive linearization in problems of optimizing nuclear reactor operating conditions

1975 ◽  
Vol 39 (5) ◽  
pp. 1005-1006
Author(s):  
V. V. Khromov ◽  
A. A. Kashutin
Author(s):  
T. L. Dickson ◽  
M. T. EricksonKirk

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.


Author(s):  
Alberto Sáez-Maderuelo ◽  
María Luisa Ruiz-Lorenzo ◽  
Francisco Javier Perosanz ◽  
Patricie Halodová ◽  
Jan Prochazka ◽  
...  

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).


2021 ◽  
Vol 4 (1) ◽  
Author(s):  
Eric Dumonteil ◽  
Rian Bahran ◽  
Theresa Cutler ◽  
Benjamin Dechenaux ◽  
Travis Grove ◽  
...  

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.


2017 ◽  
Vol 15 (6) ◽  
pp. 367
Author(s):  
Edward Waller, PhD ◽  
George Bereznai, PhD ◽  
John Shaw, BSc ◽  
Joseph Chaput, MASc ◽  
Jean-Francois Lafortune, PhD

Training offsite emergency response personnel basic awareness of onsite control room operations during nuclear power plant emergency conditions was the primary objective of a week-long workshop conducted on a CANDU® virtual nuclear reactor simulator available at the University of Ontario Institute of Technology, Oshawa, Canada. The workshop was designed to examine both normal and abnormal reactor operating conditions, and to observe the conditions in the control room that may have impact on the subsequent offsite emergency response. The workshop was attended by participants from a number of countries encompassing diverse job functions related to nuclear emergency response. Objectives of the workshop were to provide opportunities for participants to act in the roles of control room personnel under different reactor operating scenarios, providing a unique experience for participants to interact with the simulator in realtime, and providing increased awareness of control room operations during accident conditions. The ability to “pause” the simulator during exercises allowed the instructors to evaluate and critique the performance of participants, and to provide context with respect to potential offsite emergency actions. Feedback from the participants highlighted (i) advantages of observing and participating “hands-on” with operational exercises, (ii) their general unfamiliarity with control room operational procedures and arrangements prior to the workshop, (iii) awareness of the vast quantity of detailed control room procedures for both normal and transient conditions, and (iv) appreciation of the increased workload for the operators in the control room during a transient from normal operations. Based upon participant feedback, it was determined that the objectives of the training had been met, and that future workshops should be conducted.


Author(s):  
Charles Forsberg

A combined-cycle power plant is proposed that uses heat from a high-temperature nuclear reactor and hydrogen produced by the high-temperature reactor to meet base-load and peak-load electrical demands. For base-load electricity production, air is compressed; flows through a heat exchanger, where it is heated to between 700 and 900°C; and exits through a high-temperature gas turbine to produce electricity. The heat, via an intermediate heat-transport loop, is provided by a high-temperature reactor. The hot exhaust from the Brayton-cycle turbine is then fed to a heat recovery steam generator that provides steam to a steam turbine for added electrical power production. To meet peak electricity demand, after nuclear heating of the compressed air, hydrogen is injected into the combustion chamber, combusts, and heats the air to 1300°C—the operating conditions for a standard natural-gas-fired combined-cycle plant. This process increases the plant efficiency and power output. Hydrogen is produced at night by electrolysis or other methods using energy from the nuclear reactor and is stored until needed. Therefore, the electricity output to the electric grid can vary from zero (i.e., when hydrogen is being produced) to the maximum peak power while the nuclear reactor operates at constant load. Because nuclear heat raises air temperatures above the auto-ignition temperatures of the hydrogen and powers the air compressor, the power output can be varied rapidly (compared with the capabilities of fossil-fired turbines) to meet spinning reserve requirements and stabilize the grid.


2019 ◽  
Vol 34 (3) ◽  
pp. 238-242
Author(s):  
Rex Abrefah ◽  
Prince Atsu ◽  
Robert Sogbadji

In pursuance of sufficient, stable and clean energy to solve the ever-looming power crisis in Ghana, the Nuclear Power Institute of the Ghana Atomic Energy Commission has on the agenda to advise the government on the nuclear power to include in the country's energy mix. After consideration of several proposed nuclear reactor technologies, the Nuclear Power Institute considered a high pressure reactor or vodo-vodyanoi energetichesky reactor as the nuclear power technologies for Ghana's first nuclear power plant. As part of technology assessments, neutronic safety parameters of both reactors are investigated. The MCNP neutronic code was employed as a computational tool to analyze the reactivity temperature coefficients, moderator void coefficient, criticality and neutron behavior at various operating conditions. The high pressure reactor which is still under construction and theoretical safety analysis, showed good inherent safety features which are comparable to the already existing European pressurized reactor technology.


ROTASI ◽  
2013 ◽  
Vol 15 (4) ◽  
pp. 33
Author(s):  
Anwar Ilmar Ramadhan ◽  
Indra Setiawan ◽  
M. Ivan Satryo

Safety is an issue that is of considerable concern in the design, operation and development of a nuclear reactor. Therefore, the method of analysis used in all these activities should be thorough and reliable so as to predict a wide range of operating conditions of the reactor, both under normal operating conditions and in the event of an accident. Performance of heat transfer to the cooling of nuclear fuel, reactor safety is key. Poor heat removal performance would threaten the integrity of the fuel cladding which could further impact on the release of radioactive substances into the environment in an uncontrolled manner to endanger the safety of the reactor workers, the general public, and the environment. This study has the objective is to know is profile contour of fluid flow and the temperature distribution pattern of the cooling fluid is water (H2O) in convection in to SMR reactor with fuel sub reed arrangement of hexagonal in forced convection. In this study will be conducted simulations on the SMR reactor core used sub channel hexagonal using CFD (Computational Fluid Dynamics) code. And the results of this simulation look more upward (vector of fluid flow) fluid temperature will be warm because the heat moves from the wall to the fluid heater. Axial direction and also looks more fluid away from the heating element temperature will be lower.


2019 ◽  
Vol 283 ◽  
pp. 07011
Author(s):  
Didier Flotté ◽  
David Macel ◽  
Abd Ennour Bouzenad ◽  
Frédéric Navacchia

Monitoring the operation of the latest-generation nuclear reactor requires ultrasonic transducers able to operate at very high temperatures (> 600°C). To achieve this, CEA has requested from “Institut de Soudure” to help developing a new technology for these transducers compared to the one previously developed. This began with the development of a reliable assembly technique between a lithium niobate piezoelectric disc whose Curie temperature exceeds 1100°C and stainless steel discs. The chosen solution was to braze the niobate disc between two stainless steel discs. Parallel to this development, it was also necessary to develop a NDE procedure to verify the quality of the brazing assemblies. This development began with a simulation of immersion ultrasonic testing of the assemblies. The constraints were to be able to control the two brazed interfaces from the same access face, with the possibility of detecting and dimensioning defects with an equivalent diameter of 0.25 mm. This phase is important to define the optimal transducer with the associated operating conditions. The first assemblies validated the preliminary choices. To exploit the cartographies obtained, a signal processing procedure was developed. This enabled an automatic characterization of the indications observed. However, the analysis of the signals observed proved to be more complex than the one predicted by the simulation. Once the origin of the various observed signals was identified it was then possible to define windows allowing the construction of the cartographies to analyze. In case of a good quality assembly, it was possible to qualify the generated beam and to image it in the focal plane but with an observed signal having a very low damping. These first encouraging results, however, show that there is still some validation and development work to increase the sensitivity of the developed translator and its damping.


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