scholarly journals Anthropogenic Radiocarbon: Past, Present, and Future

Radiocarbon ◽  
1986 ◽  
Vol 28 (2A) ◽  
pp. 668-672 ◽  
Author(s):  
Pavel Povinec ◽  
Martin Chudý ◽  
Alexander Šivo
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Nuclear Reactors ◽  
Pressurized Water Reactors ◽  
Anthropogenic Radionuclides ◽  
Pressurized Water ◽  
Heavy Water Reactor ◽  
Weapon Testing ◽  
Water Reactors

14C is one of the most important anthropogenic radionuclides released to the environment by human activities. Weapon testing raised the 14C concentration in the atmosphere and biosphere to +100% above the natural level. This excess of atmospheric C at present decreases with a half-life of ca 7 years. Recently, a new source of artificially produced 14C in nuclear reactors has become important. Since 1967, the Bratislava 14C laboratory has been measuring 14C in atmospheric 14CO2 and in a variety of biospheric samples in densely populated areas and in areas close to nuclear power plants. We have been able to identify a heavy-water reactor and the pressurized water reactors as sources of anthropogenic 14C. 14C concentrations show typical seasonal variations. These data are supported by measurements of 3H and 85Kr in the same locations. Results of calculations of future levels of anthropogenic 14C in the environment due to increasing nuclear reactor installations are presented.

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.

A Review on Specific Features of Small and Medium Sized Nuclear Power Plants

2010 ◽  
Author(s):  
Salah Ud-din Khan ◽  
Minjun Peng ◽  
Muhammad Zubair ◽  
Shaowu Wang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Nuclear Reactors ◽  
Fixed Bed ◽  
Nuclear Industry ◽  
Energy Agency ◽  
Pressurized Water ◽  
Water Reactor ◽  
Energy Supply System

Due to global warming and high oil prices nuclear power is the most feasible solution for generating electricity. For the fledging nuclear power industry small and medium sized nuclear reactors (SMR’s) are instrumental for the development and demonstration of nuclear reactor technology. Due to the enhanced and outstanding safety features, these reactors have been considered globally. In this paper, first we have summarized the reactor design by considering some of the large nuclear reactor including advanced and theoretical nuclear reactor. Secondly, comparison between large nuclear reactors and SMR’s have been discussed under the criteria led by International Atomic Energy Agency (IAEA). Thirdly, a brief review about the design and safety aspects of some of SMR’s have been carried out. We have considered the specifications and parametric analysis of the reactors like: ABV which is the floating type integral Pressurized water reactor; Long life, Safe, Simple Small Portable Proliferation Resistance Reactor (LSPR) concept; Multi-Application Small Light Water Reactor (MASLWR) concept; Fixed Bed Nuclear Reactor (FBNR); Marine Reactor (MR-X) & Deep Sea Reactor (DR-X); Space Reactor (SP-100); Passive Safe Small Reactor for Distributed energy supply system (PSRD); System integrated Modular Advanced Reactor (SMART); Super, Safe, Small and Simple Reactor (4S); International Reactor Innovative and Secure (IRIS); Nu-Scale Reactor; Next generation nuclear power plant (NGNP); Small, Secure Transportable Autonomous Reactor (SSTAR); Power Reactor Inherently Safe Module (PRISM) and Hyperion Reactor concept. Finally we have point out some challenges that must be resolved in order to play an effective role in Nuclear industry.

Nuclear Power Plant Fires and Explosions: Part I — Plant Designs and Hydrogen Ignition

2017 ◽  
Author(s):  
Robert A. Leishear
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Hydrogen Generation ◽  
Reactor Power ◽  
Pressurized Water ◽  
Hydrogen Ignition ◽  
History Of ◽  
Water Reactors

Requiring further investigation, hydrogen explosions and fires have occurred in several operating nuclear reactor power plants. Major accidents that were affected by hydrogen fires and explosions included Chernobyl, Three Mile Island, and Fukushima Daiichi. Smaller piping explosions have occurred at Hamaoka and Brunsbüttel Nuclear Power Plants. This paper is the first paper in a series of publications to discuss this issue. In particular, the different types of reactors that have a history of fires and explosions are discussed here, along with a discussion of hydrogen generation in commercial reactors, which provides the fuel for fires and explosions in nuclear power plants. Overall, this paper is a review of pertinent information on reactor designs that is of particular importance to this multi-part discussion of hydrogen fires and explosions. Without a review of reactor designs and hydrogen generation, the ensuing technical discussions are inadequately backgrounded. Consequently, the basic designs of pressurized water reactors (PWR’s), boiling water reactors (BWR’s), and pressure-tube graphite reactors (RBMK) are discussed in adequate detail. Of particular interest, the Three Mile Island design for a PWR is presented in some detail.

Auxiliary Ventilation Systems Design Basis and Analysis Approach

2010 ◽  
Author(s):  
Stuart R. Douglass
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Systems Design ◽  
Infiltration Rate ◽  
Operating Conditions ◽  
Pressurized Water Reactors ◽  
Pressurized Water ◽  
Air Infiltration ◽  
Water Reactors ◽  
Heat Loads

Auxiliary systems supporting pressurized water reactors (PWR) within commercial nuclear power plants are enclosed within a special ventilation (SV) zone that is isolated post-accident. Air within the SV zone is recirculated through carbon adsorbers, and discharged at a rate equal to the SV zone air infiltration rate. The SV zone relies on safety-related fan coil units (FCUs) to remove heat since air infiltration is kept to a minimum in order to reduce the spread of contamination. This paper discusses efforts undertaken to quantify area heat loads and FCU operating conditions within the SV zone, and transient analyses performed for loss of FCUs using the GOTHIC code.

scholarly journals Methodology for determination of alarm and warning set-points for radioactive effluent monitors in Korean pressurized water reactors

2018 ◽  
Vol 33 (4) ◽  
pp. 406-410 ◽  
Author(s):  
Tae Kong ◽  
Siyoung Kim ◽  
Youngju Lee ◽  
Jung Son
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Active Management ◽  
Pressurized Water Reactors ◽  
Regulatory Body ◽  
Pressurized Water ◽  
Radioactive Effluent ◽  
Regulatory Limits ◽  
Water Reactors

All radioactive gaseous and liquid effluents discharged from Korean nuclear power plants are monitored by effluent monitors to prevent effluent releases to the environment under uncontrolled conditions. This paper provides the methodology and parameters used in the calculation of alarm (high) and warning (low) set-points for gaseous and liquid effluent monitors in Korean pressurized water reactors. Alarm set-points are determined to assure compliance with the Korean regulatory limits of concentration of radioactive effluents. Even though warning set-points are not required by the regulatory body, Korean pressurized water reactors determine the warning set-points of effluent monitors not only to take an active management of effluent discharge but also to keep radiation doses to members of the public living around nuclear power plants as low as reasonably achievable.

scholarly journals Characteristics of Radioactive Effluent Releases from Pressurized Water Reactors after Permanent Shutdown

Energies ◽  
2020 ◽  
Vol 13 (10) ◽  
pp. 2436
Author(s):  
Ji Su Kang ◽  
Jae Hak Cheong
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Source Term ◽  
Monitoring Program ◽  
Pressurized Water Reactors ◽  
Pressurized Water ◽  
Radioactive Effluent ◽  
Theoretical Reasoning ◽  
Discharge Monitoring ◽  
Water Reactors

In order to expand our understanding of the characteristics of radioactive effluent from nuclear power plants under decommissioning, which have not been systematically investigated, a series of source term models of radioactive effluent after permanent shutdown has been established based upon theoretical reasoning on the design and operation features of plants and derived in terms of fifteen arguments. Comprehensive radioactive effluent data have been collected and profiled from twenty-eight decommissioning pressurized water reactors, and annual trends of effluent from each plant have been quantitatively analyzed using Mann-Kendall statistical test. In addition, the characteristics of collected effluent data have been qualitatively interpreted based upon arguments newly proposed in this study. Furthermore, potential decreasing of dilution factor for liquid effluent and its safety implications are identified. The source term models and verified characteristics of radioactive effluent after permanent shutdown developed in this study can be used for establishing more efficient discharge monitoring program for decommissioning authorization.

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