Power Maximization of Fully Passive Lead-Bismuth Eutectic Cooled Small Modular Reactor

2014 ◽  
Author(s):  
Jaehyun Cho ◽  
Yong-Hoon Shin ◽  
Il Soon Hwang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Natural Circulation ◽  
Reactor Core ◽  
Design Stage ◽  
Pressurized Water ◽  
Small Modular Reactor ◽  
Power Maximization ◽  
Lead Bismuth Eutectic ◽  
Water Reactors

Although the current Pressurized Water Reactors (PWRs) have significantly contributed to the global energy supply, PWRs have not been considered as a trustworthy energy solution owing to its several problems; spent nuclear fuels (SNFs), nuclear safety, and nuclear economy. In order to overcome these problems, lead-bismuth eutectic (LBE) fully passive cooling Small Modular Reactor (SMR) system is suggested. It is possible to not only provide the solution of the problem of SNFs through the transmutation feature of LBE coolant, but also increase the safety and economy through the concepts of the natural circulation cooling SMRs. It is necessary to maximize the advantages (safety and economy) of this type of Nuclear Power Plants for several applications in future. Accordingly, objective of the study is to maximize the reactor core power while the limitations of shipping size, materials endurance, long-burning criticality as well as safety under Beyond Design Basis Events must be satisfied. Design limitations of natural circulating LBE-cooling SMRs are researched and power maximization method is developed based on obtained design limitations. It is expected that the results are contributed to reactor design stage with providing several insights to designers as well as the methods for design optimization of other type of SMRs.

scholarly journals INTELLIGENT TECHNIQUES FOR ANOMALY DETECTION IN NUCLEAR REACTORS

2021 ◽  
Vol 247 ◽  
pp. 21011
Author(s):  
George Ioannou ◽  
Thanos Tagaris ◽  
Georgios Alexandridis ◽  
Andreas Stafylopatis
Keyword(s):  
Signal Processing ◽  
Neutron Flux ◽  
Nuclear Power ◽  
Power Plants ◽  
Reactor Core ◽  
Machine Learning Techniques ◽  
Full Potential ◽  
Pressurized Water ◽  
Learning Techniques ◽  
Water Reactors

The safe operation of nuclear power plants is highly dependent on the ability of quickly and accurately identifying possible anomalies and perturbations in the reactor. Operational defects are primarily diagnosed by detectors that capture changes in the neutron flux, placed at various points inside and outside of the core. Neutron flux signals are subsequently analyzed with signal processing techniques in an effort to be better described (have their higher-order characteristics uncovered, locate transient events, etc). To this end, the application of intelligent techniques may be extremely beneficial, as it may assist and extend the current level of analysis. Besides, the combination of signal processing methodologies and machine learning techniques in the framework of nuclear power plant data is an emerging topic that has yet to show its full potential. In this context, the current contribution attempts at introducing intelligent approaches and more specifically, deep learning techniques, in neutron flux signal analysis for the identification of perturbations and other anomalies in the reactor core that may affect its operational capabilities. The obtained results of an initial stage of analysis on neutron flux signals captured at pressurized water reactors are encouraging, underlying the robustness and the potential of the proposed approach.

PWR Nuclear Power Plants Fuel Management Optimization

2009 ◽  
Author(s):  
Ahmad Zolfaghari ◽  
Hamid Minuchehr ◽  
Ali Noroozy ◽  
Peymaan Makarachi ◽  
F. Koshahval
Keyword(s):  
Genetic Algorithm ◽  
Objective Function ◽  
Nuclear Power ◽  
Power Plants ◽  
Reactor Core ◽  
Complex Problem ◽  
Pressurized Water ◽  
Water Reactors ◽  
Management Optimization ◽  
Optimum Solution

The objective of this paper is to develop a new hybrid mutation in genetic algorithm (GA) for designing the loading pattern (LP) in pressurized water reactors. Because of huge number of possible combinations for the fuel assemblies (FA’s) loading in a core, finding the optimum solution is truly a complex problem. In common genetic algorithm the mutation and crossover techniques are used to optimize an objective function but in this paper a new hybrid mutation is presented. In this study flattening of power inside a reactor core is chosen as an objective function. To obtain optimal FA arrangement, a core reload package code, MAKNOGA, based on well established MAKGA code is developed. This code is applicable for all types of PWR core having different geometries and designs with an unlimited number of FA types. The result is well improved in comparison with pattern proposed by designer.

scholarly journals Reactor Vessel Internals Segmentation Experience using Mechanical Cutting Tools

2013 ◽  
Vol 10 (2) ◽  
pp. 6-10 ◽  
Author(s):  
Petr Pospíšil
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cutting Tools ◽  
Reactor Vessel ◽  
Boiling Water Reactors ◽  
Abrasive Waterjet ◽  
Pressurized Water ◽  
Water Reactors ◽  
Mechanical Cutting ◽  
Reactor Internals

Abstract Some commercial nuclear power plants have been permanently shut down to date and decommissioned using dismantling methods. Other operating plants have decided to undergo an upgrade process that includes replacement of reactor internals. In both cases, there is a need to perform a segmentation of the reactor vessel internals with proven methods for long term waste disposal. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive waterjet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Mechanical cutting has been used by Westinghouse since 1999 for both Pressurized Water Reactors (PWR’s) and Boiling Water Reactors (BWR’s) and its process has been continuously improved over the years. The complexity of the work requires well designed and reliable tools. Different band saws, disc saws, tube cutters and shearing tools have been developed to cut the reactor internals. All of those equipments are hydraulically driven which is very suitable for submerged applications. Westinghouse experience in mechanical cutting has demonstrated that it is an excellent technique for segmentation of internals. In summary, the purpose of this paper will be to provide an overview of the Westinghouse mechanical segmentation process, based on actual experience from the work that has been completed to date.

scholarly journals Plant Performance History of an Innovative Gate Valve in Critical Service Applications

2017 ◽  
Author(s):  
M. S. Kalsi ◽  
Patricio Alvarez ◽  
Thomas White ◽  
Micheal Green
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Gate Valve ◽  
Plant Performance ◽  
Flow Loop ◽  
Performance History ◽  
Pressurized Water ◽  
Maintenance Costs ◽  
History Of ◽  
Water Reactors

A previous paper [1] describes the key features of an innovative gate valve design that was developed to overcome seat leakage problems, high maintenance costs as well as issues identified in the Nuclear Regulatory Commission (NRC) Generic Letters 89-10, 95-07 and 96-05 with conventional gate valves [2,3,4]. The earlier paper was published within a year after the new design valves were installed at the Pilgrim Nuclear Plant — the plant that took the initiative to form a teaming arrangement as described in [1] which facilitated this innovative development. The current paper documents the successful performance history of 22 years at the Pilgrim plant, as well as performance history at several other nuclear power plants where these valves have been installed for many years in containment isolation service that requires operation under pipe rupture conditions and require tight shut-off in both Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). The performance history of the new valve has shown to provide significant performance advantage by eliminating the chronic leakage problems and high maintenance costs in these critical service applications. This paper includes a summary of the design, analysis and separate effects testing described in detail in the earlier paper. Flow loop testing was performed on these valves under normal plant operation, various thermal binding and pressure locking scenarios, and accident/pipe rupture conditions. The valve was designed, analyzed and tested to satisfy the requirements of ANSI B16.41 [9]; it also satisfies the requirements of ASME QME 1-2012 [10]. The results of the long-term performance history including any degradation observed and its root cause are summarized in the paper. Paper published with permission.

Underground Siting of Nuclear Power Plants: Insights From Fukushima

2012 ◽  
Author(s):  
Jay F. Kunze ◽  
James M. Mahar ◽  
Kellen M. Giraud ◽  
C. W. Myers
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Light Water Reactors ◽  
Light Water ◽  
Waste Storage ◽  
Small Modular Reactor ◽  
Fukushima Daiichi ◽  
Water Reactors ◽  
High Level

Siting of nuclear power plants in an underground nuclear park has been proposed by the authors in many previous publications, first focusing on how the present 1200 to 1600 MW-electric light water reactors could be sited underground, then including reprocessing and fuel manufacturing facilities, as well as high level permanent waste storage. Recently the focus has been on siting multiple small modular reactor systems. The recent incident at the Fukushima Daiichi site has prompted the authors to consider what the effects of a natural disaster such as the Japan earthquake and subsequent tsunami would have had if these reactors had been located underground. This paper addresses how the reactors might have remained operable — assuming the designs we previously proposed — and what lessons from the Fukushima incident can be learned for underground nuclear power plant designs.

Background on Low Temperature Overpressure Protection System Setpoints for Pressure-Temperature Curves

2020 ◽  
Author(s):  
Jeffrey C. Poehler ◽  
Gary L. Stevens ◽  
Anees A. Udyawar ◽  
Amy Freed
Keyword(s):  
Low Temperature ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Pressure Vessels ◽  
Pressurized Water Reactors ◽  
Documentation System ◽  
Pressurized Water ◽  
System Pressure ◽  
Water Reactors

Abstract ASME Code, Section XI, Nonmandatory Appendix G (ASME-G) provides a methodology for determining pressure and temperature (P-T) limits to prevent non-ductile failure of nuclear reactor pressure vessels (RPVs). Low-Temperature Overpressure Protection (LTOP) refers to systems in nuclear power plants that are designed to prevent inadvertent challenges to the established P-T limits due to operational events such as unexpected mass or temperature additions to the reactor coolant system (RCS). These systems were generally added to commercial nuclear power plants in the 1970s and 1980s to address regulatory concerns related to LTOP events. LTOP systems typically limit the allowable system pressure to below a certain value during plant operation below the LTOP system enabling temperature. Major overpressurization of the RCS, if combined with a critical size crack, could result in a brittle failure of the RPV. Failure of the RPV could make it impossible to provide adequate coolant to the reactor core and result in a major core damage or core melt accident. This issue affected the design and operation of all pressurized water reactors (PWRs). This paper provides a description of an investigation and technical evaluation regarding LTOP setpoints that was performed to review the basis of ASME-G, Paragraph G-2215, “Allowable Pressure,” which includes provisions to address pressure and temperature limitations in the development of P-T curves that incorporate LTOP limits. First, high-level summaries of the LTOP issue and its resolution are provided. LTOP was a significant issue for pressurized water reactors (PWRs) starting in the 1970s, and there are many reports available within the U.S. Nuclear Regulatory Commission’s (NRC’s) documentation system for this topic, including Information Notices, Generic Letters, and NUREGs. Second, a particular aspect of LTOP as related to ASME-G requirements for LTOP is discussed. Lastly, a basis is provided to update Appendix G-2215 to state that LTOP setpoints are based on isothermal (steady-state) conditions. This paper was developed as part of a larger effort to document the technical bases behind ASME-G.

Coordinated Research Projects at IAEA Concerning RPV and Structural Integrity

2006 ◽  
Author(s):  
William Server ◽  
Timothy Hardin ◽  
Milan Brumovsky´
Keyword(s):  
Ferritic Steel ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Current Status ◽  
Research Projects ◽  
Energy Agency ◽  
Pressurized Water ◽  
Toughness Testing ◽  
Water Reactors

The International Atomic Energy Agency (IAEA) has had a series of reactor pressure vessel (RPV) structural integrity programs that started back in the 1970s. These Coordinated Research Projects most recently have focused on use of the Master Curve fracture toughness testing approach for RPV and other ferritic steel components and on the issue of pressurized thermal shock (PTS) in operating pressurized water reactors. This paper will provide the current status for these projects and discuss the implications for improved safety of key ferritic steel components in nuclear power plants (NPPs).

Comparison of Nuclear Fitness for Service Codes: ASME Section XI and RSE-M AFCEN Codes

2014 ◽  
Author(s):  
Claude Faidy
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Light Water Reactors ◽  
Pressurized Water ◽  
Term Operation ◽  
Mechanical Components ◽  
Water Reactors ◽  
Test Requirements ◽  
Service Inspection

Two major Codes are used for Fitness for Service of Nuclear Power Plants: one is the ASME B&PV Code Section XI and the other one is the French RSE-M Code. Both of them are largely used in many countries, partially or totally. The last 2013 RSE-M covers “Mechanical Components of Pressurized Water Reactors (PWRs): - Pre-service and In-service inspection - Surveillance in operation or during shutdown - Flaw evaluation - Repairs-Replacements parts for plant in operation - Pressure tests The last 2013 ASME Section XI covers “Mechanical components and containment of Light Water Reactors (LWRs)” and has a larger scope with similar topics: more types of plants (PWR and Boiling Water Reactor-BWR), other components like metallic and concrete containments… The paper is a first comparison covering the scope, the jurisdiction, the general organization of each section, the major principles to develop In Service Inspection, Repair-Replacement activities, the flaw evaluation rules, the pressure test requirements, the surveillance procedures (monitoring…) and the connections with Design Codes… These Codes are extremely important for In-service inspection programs in particular and essential tools to justify long term operation of Nuclear Power Plants.

The Next EDF Tracking System

2009 ◽  
Author(s):  
Emmanuelle Julli ◽  
Bertrand Lantes
Keyword(s):  
Radioactive Waste ◽  
Nuclear Power ◽  
Power Plants ◽  
Tracking System ◽  
Treatment Plant ◽  
Computer Application ◽  
Pressurized Water ◽  
Waste Package ◽  
Additional Processing ◽  
Water Reactors

EDF’s network of nuclear power plants (NPP) comprises 58 pressurized water reactors. Solid waste arising during plant operation (mainly VLLW, LLW and ILW) are conditioned and sent either to interim storage, an off site treatment plant for additional processing (e.g. the Centraco incinerator or the melting facilities of SOCODEI) or directly to one of the two final repositories operated by ANDRA, the French national radioactive waste management agency. The tracking system allows: - the checking of waste package characteristics against acceptance criteria of the final disposal facilities or off site treatment facilities; and - the transmission of the waste package data to ANDRA and SOCODEI. Since 1992, the EDF computer application DRA has been run on networked computers at EDF and ANDRA, and more recently at SOCODEI. DRA is also a key element in the management of radioactive waste. It allows a large range of inter comparisons to be made between the NPPs in operation and is thus the principal tool used optimize technical and economic performance. After 15 years of use, DRA was technically obsolete and could no longer be successfully developed to meet evolving regulatory requirements. It was therefore decided to completely replace the DRA system and in so doing to introduce new functionality.

Extensions to SCDAP/RELAP5/ATHENA for Analysis of HTGRs

2004 ◽  
Author(s):  
Edwin A. Harvego ◽  
Larry J. Siefken
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Molecular Diffusion ◽  
Natural Circulation ◽  
Reactor Core ◽  
Heat Sinks ◽  
Block Type ◽  
Flow Losses ◽  
Decay Heat ◽  
Extended Code

The SCDAP/RELAP5/ATHENA code was extended to enable the code to perform transient analyses of two leading candidates for the next generation of nuclear power plants, namely the Pebble Bed and Block-Type (prismatic) High Temperature Gas Reactors (PB-HTGRs and BT-HTGRs). Models in five areas of reactor behavior were implemented into the code. First, models were implemented to calculate the transfer of decay heat by conduction and radiation from the center of the heterogeneous cores of these reactors to their outer surfaces. Second, models were implemented to calculate the transfer of decay heat from the outer surfaces of these reactor cores to ultimate heat sinks such as the atmosphere. Third, models were implemented to calculate the flow losses and convective heat transfer in the core of a PB-HTGR. Fourth, models were implemented to calculate the oxidation of graphite due to the ingress of air. Fifth, a model was implemented to calculate the ingress of air into the reactor core by molecular diffusion from the point of break in the coolant system. The extended code was applied to the analysis of conduction cooldown accident in a BT-HTGR. The analysis indicated that the new models in SCDAP/RELAP5/ATHENA are properly interfaced with the previously developed and assessed fluid behavior models in the code. Preliminary results indicate that post-shutdown decay heat can be adequately removed from BT-HTGRs by natural circulation of air from the atmosphere.

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