scholarly journals Robustness Study of Electro-Nuclear Scenario under Disruption

2021 ◽  
Vol 2 (1) ◽  
pp. 1-8
Author(s):  
Jiali Liang ◽  
Marc Ernoult ◽  
Xavier Doligez ◽  
Sylvain David ◽  
Léa Tillard ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Pressurized Water Reactors ◽  
Nuclear Wastes ◽  
Fast Reactors ◽  
Pressurized Water ◽  
Static Strategy ◽  
The Future ◽  
Political Choices ◽  
Water Reactors ◽  
Development Objective

As the future of nuclear power is uncertain, only choosing one development objective for the coming decades can be risky; while trying to achieve several possible objectives at the same time may lead to a deadlock due to contradiction among them. In this work, we study a simple scenario to illustrate the newly developed method of robustness study, which considers possible change of objectives. Starting from the current French fleet, two objectives are considered regarding the possible political choices for the future of nuclear power: A. Complete substitution of Pressurized Water Reactors by Sodium-cooled Fast Reactors in 2180; B. Minimization of all potential nuclear wastes without SFR deployment in 2180. To study the robustness of strategies, the disruption of objective is considered: the objective to be pursued is possibly changed abruptly from A into B at unknown time. To minimize the consequence of such uncertainty, the first option is to identify a robust static strategy, which shows the best performance for both objectives A and B in the predisruption situation. The second option is to adapt a trajectory which pursues initially objective A, for objective B in case of the disruption. To identify and to analyze the adaptively robust strategies, outcomes of possible adaptations upon a given trajectory are compared with the robust static optimum. The temporality of adaptive robustness is analyzed by investigating different adaptation times.

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 Comparative Study on Severe Accident Phenomena Related to Melt Progression in Sodium Fast Reactors and Pressurized Water Reactors

2020 ◽  
Vol 7 (3) ◽  
Author(s):  
Andrea Bachrata ◽  
Fréderic Bertrand ◽  
Nathalie Marie ◽  
Fréderic Serre
Keyword(s):  
Mitigation Strategies ◽  
Severe Accident ◽  
Pressurized Water Reactors ◽  
Physical Parameters ◽  
Fast Reactors ◽  
Pressurized Water ◽  
The Core ◽  
Sodium Fast Reactors ◽  
Water Reactors ◽  
Accident Scenarios

Abstract The nuclear safety approach has to cover accident sequences involving core degradation in order to develop reliable mitigation strategies for both existing and future reactors. In particular, the long-term stabilization of the degraded core materials and their coolability has to be ensured after a severe accident. This paper focuses on severe accident phenomena in pressurized water reactors (PWR) compared to those potentially occurring in future GenIV-type sodium fast reactors (SFR). First, the two considered reactor concepts are introduced by focusing on safety aspects. The severe accident scenarios leading to core melting are presented and the initiating events are highlighted. This paper focuses on in-vessel severe accident phenomena, including the chronology of core damage, major changes in the core configuration and molten core progression. Regarding the mitigation means, the in-vessel retention phenomena and the core catcher characteristics are reviewed for these different nuclear generation concepts (II, III, and IV). A comparison between the PWR and SFR severe accident evolution is provided as well as the relation between governing physical parameters and the adopted mitigation provisions for each reactor concept. Finally, it is highlighted how the robustness of the safety demonstration is established by means of a combined probabilistic and deterministic approach.

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.

Evaluation of single heated channel and subchannel modeling of a nuclear once through steam generator (OTSG)

Kerntechnik ◽  
2020 ◽  
Vol 85 (1) ◽  
pp. 54-67
Author(s):  
A. Hamedani ◽  
O. Noori-Kalkhoran ◽  
R. Ahangari ◽  
M. Gei
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Axial Direction ◽  
Pressurized Water Reactors ◽  
Steam Generators ◽  
Pressurized Water ◽  
Heated Channel ◽  
Water Reactors ◽  
Secondary Side

Abstract Steam generators are one of the most important components of pressurized-water reactors. This component plays the role of heat transfer and pressure boundary between primary and secondary side fluids. The Once Through Steam Generator (OTSG) is an essential component of the integrated nuclear power system. In this paper, steady-state analysis of primary and secondary fluids in the Integral Economizer Once Through Steam Generator (IEOTSG) have been presented by Single Heated Channel (SHC) and subchannel modelling. Models have been programmed by MATLAB and FORTRAN. First, SHC model has been used for this purpose (changes are considered only in the axial direction in this model). Second, the subchannel approach that considers changes in the axial and also radial directions has been applied. Results have been compared with Babcock and Wilcox (B&W) 19- tube once through steam generator experimental data. Thermal- hydraulic profiles have been presented for steam generator using both of models. Accuracy and simplicity of SHC model and importance of localization of thermal-hydraulic profiles in subchannel approach have been proved.

scholarly journals Role of Passive Safety Systems in Chinese Nuclear Power Development

2009 ◽  
Vol 2009 ◽  
pp. 1-7 ◽  
Author(s):  
X. Cheng ◽  
Y. H. Yang ◽  
Y. Ouyang ◽  
H. X. Miao
Keyword(s):  
Nuclear Power ◽  
Passive Safety ◽  
Pressurized Water Reactors ◽  
Chinese Government ◽  
Power Development ◽  
Pressurized Water ◽  
Passive Safety Systems ◽  
Nuclear Power Development ◽  
Water Reactors ◽  
Safety Systems

Passive safety systems have been widely applied to advanced water-cooled reactors, to enhance the safety of nuclear power plants. The ambitious program of the nuclear power development in China requires reactor concepts with high safety level. For the near-term and medium-term, the Chinese government decided for advanced pressurized water reactors with an extensive usage of passive safety systems. This paper describes some important criteria and the development program of the Chinese large-scale pressurized water reactors. An overview on representative research activities and results achieved so far on passive safety systems in various institutions is presented.

Thermotransport of Hydrogen in the Modified Zircaloy-4 as a Function of Hydrogen and Oxygen Concentrations

2002 ◽  
Vol 713 ◽  
Author(s):  
Hyun Sook Kim ◽  
Kyung Sub Lee ◽  
Seon Jin Kim
Keyword(s):  
Hydrogen Concentration ◽  
Nuclear Power ◽  
Pressurized Water Reactors ◽  
Chemical Compositions ◽  
Fuel Cladding ◽  
Pressurized Water ◽  
Nuclear Fuel Cladding ◽  
Water Reactors ◽  
Cladding Material ◽  
Oxygen Concentrations

ABSTRACTThe hydrogen redistribution induced by the thermotransport in the modified Zircaloy-4 at temperatures likely to be encountered in nuclear power reactors (300-340°C) was investigated by means of steady state techniques. The modified Zircaloy-4 was prepared by changing the chemical compositions of Zircaloy-4, which is used widely as a nuclear fuel cladding material in pressurized water reactors. The change of Q for hydrogen, which describes the direction and magnitude of the thermotransport, with increasing hydrogen and oxygen concentrations was investigated in the modified Zircaloy-4. The value of Q for hydrogen in the modified Zircloy-4 alloys was found to be about 7 kcal/mol and it was not affected by hydrogen concentration in the hydrogen concentration range from 63.3 ppm to 91.7 ppm. While the value of Q for hydrogen decreased from 6.8 kcal/mol to 4.5 kcal/mol with increasing oxygen concentration from 0.2 wt% to 1.0 wt% and it was considered to be due to the trapping of hydrogen by oxygen. In addition, the hydrogen redistribution and Q in Zircaloy-4 was also investigated in order to compare the characteristics of thermotransport of hydrogen between Zircaloy-4 and modified Zircaloy-4. The hydrogen redistribution and Q in Zircaloy-4 showed the same results to those of the modified Zircaloy-4.

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