scholarly journals Radioactivity control strategy for the JUNO detector

2021 ◽  
Vol 2021 (11) ◽  
Author(s):  
◽  
Angel Abusleme ◽  
Thomas Adam ◽  
Shakeel Ahmad ◽  
Rizwan Ahmed ◽  
...  
Keyword(s):  
Natural Radioactivity ◽  
Nuclear Power ◽  
Power Plants ◽  
Liquid Scintillator ◽  
Neutrino Flux ◽  
Neutrino Interaction ◽  
Background Count ◽  
Background Count Rate ◽  
Scientific Goal ◽  
The Impact

Abstract JUNO is a massive liquid scintillator detector with a primary scientific goal of determining the neutrino mass ordering by studying the oscillated anti-neutrino flux coming from two nuclear power plants at 53 km distance. The expected signal anti-neutrino interaction rate is only 60 counts per day (cpd), therefore a careful control of the background sources due to radioactivity is critical. In particular, natural radioactivity present in all materials and in the environment represents a serious issue that could impair the sensitivity of the experiment if appropriate countermeasures were not foreseen. In this paper we discuss the background reduction strategies undertaken by the JUNO collaboration to reduce at minimum the impact of natural radioactivity. We describe our efforts for an optimized experimental design, a careful material screening and accurate detector production handling, and a constant control of the expected results through a meticulous Monte Carlo simulation program. We show that all these actions should allow us to keep the background count rate safely below the target value of 10 Hz (i.e. ∼1 cpd accidental background) in the default fiducial volume, above an energy threshold of 0.7 MeV.

scholarly journals Comparison and Evaluation of Domestic and Foreign Radiation Environmental Standards for Nuclear Power Plants

2021 ◽  
Vol 2083 (2) ◽  
pp. 022020
Author(s):  
Jiahuan Yu ◽  
Xiaofeng Zhang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
The United States ◽  
Discharge System ◽  
Radioactive Effluent ◽  
Effluent Discharge ◽  
The Impact

Abstract With the development of the nuclear energy industry and the increasing demand for environmental protection, the impact of nuclear power plant radiation on the environment has gradually entered the public view. This article combs the nuclear power plant radiation environmental management systems of several countries, takes the domestic and foreign management of radioactive effluent discharge from nuclear power plants as a starting point, analyses and compares the laws and standards related to radioactive effluents from nuclear power plants in France, the United States, China, and South Korea. In this paper, the management improvement of radioactive effluent discharge system of Chinese nuclear power plants has been discussed.

Root Cause of Thermal Sleeve Loosening in a Korean Standard Nuclear Power Plant

2006 ◽  
Author(s):  
Sang-Nyung Kim ◽  
Sang-Gyu Lim
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Measurement Data ◽  
Significant Loss ◽  
Design Modification ◽  
Root Cause ◽  
Explosive Expansion ◽  
The Impact

The safety injection (SI) nozzle of a 1000MWe-class Korean standard nuclear power plant (KSNP) is fitted with thermal sleeves (T/S) to alleviate thermal fatigue. Thermal sleeves in KSNP #3 & #4 in Yeonggwang (YG) & Ulchin (UC) are manufactured out of In-600 and fitted solidly without any problem, whereas KSNP #5 & #6 in the same nuclear power plants, also fitted with thermal sleeves made of In-690 for increased corrosion resistance, experienced a loosening of thermal sleeves in all reactors except KSNP YG #5-1A, resulting in significant loss of generation availability. An investigation into the cause of the loosening of the thermal sleeves only found out that the thermal sleeves were subject to severe vibration and rotation, failing to uncover the root cause and mechanism of the loosening. In an effort to identify the root cause of T/S loosening, three suspected causes were analyzed: (1) the impact force of flow on the T/S when the safety SI nozzle was in operation, (2) the differences between In-600 and In-690 in terms of physical and chemical properties (notably the thermal expansion coefficient), and (3) the positioning error after explosive expansion of the T/S as well as the asymmetric expansion of T/S. It was confirmed that none of the three suspected causes could be considered as the root cause. However, after reviewing design changes applied to the Palo Verde nuclear plant predating KSNP YG #3 & #4 to KSNP #5 & #6, it was realized that the second design modification (in terms of groove depth & material) had required an additional explosive energy by 150% in aggregate, but the amount of gunpowder and the explosive expansion method were the same as before, resulting in insufficient explosive force that led to poor thermal sleeve expansion. T/S measurement data and rubbing copies also support this conclusion. In addition, it is our judgment that the acceptance criteria applicable to T/S fitting was not strict enough, failing to single out thermal sleeves that were not expanded sufficiently. Furthermore, the T/S loosening was also attributable to lenient quality control before and after fitting the T/S that resulted in significant uncertainty. Lastly, in a flow-induced vibration test planned to account for the flow mechanism that had a direct impact upon the loosening of the thermal sleeves that were not fitted completely, it was discovered that the T/S loosening was attributable to RCS main flow. In addition, it was proven theoretically that the rotation of the T/S was induced by vibration.

Nuclear power generation, renewable resources and endogenous growth

2012 ◽  
pp. 65-93
Author(s):  
Riccardo Costantini
Keyword(s):  
Radioactive Waste ◽  
Endogenous Growth ◽  
Nuclear Power ◽  
Power Plants ◽  
Preference Relation ◽  
Learning By Doing ◽  
Environmental Literature ◽  
Long Run ◽  
Simplifying Assumptions ◽  
The Impact

The author develops an endogenous growth framework in which energy production is based on a learning by doing technology exploiting renewable reproducible capital and nuclear power plants. Consumption activities generates radioactive waste according to an exogenous factor reflecting the economy energy mix, while an abatement technology, reducing the impact of solid waste accumulation on welfare, is explicitly taken into account. Differently from traditional growth and environmental literature, the author includes an explicit preference for the technology mix by postulating a non separable utility in consumption, radioactive waste and stock of renewable capital. Within this framework the author derives conditions on preferences under which sustained growth is attainable without imposing, ex ante, neither compensation nor a distaste effect characterizing utility. Finally, introducing simplifying assumptions on the preference relation, an investigation of the dynamic property of the equilibrium is provided. The results obtained suggest a high complementarity of renewable capital and nuclear technology exploitation in determining potential long run growth.

The Impact of Nuclear Power Plants on the Cooling Pond

2021 ◽  
pp. 309-314
Author(s):  
Irma Martyn ◽  
Yaroslav Petrov ◽  
Sergey Stepanov ◽  
Artem Sidorenko
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Cooling Pond ◽  
The Impact

scholarly journals Effects of different nuclear evaluation data on the RMC keff calculation

2020 ◽  
Vol 239 ◽  
pp. 19005
Author(s):  
Zhang Wenxin ◽  
Qiang shenglong ◽  
Yin qiang ◽  
Cui Xiantao
Keyword(s):  
Cross Section ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Neutron Cross Section ◽  
Nuclear Data ◽  
Section Data ◽  
Calculation Results ◽  
The Impact ◽  
Physical Calculation

Neutron cross section data is the basis of nuclear reactor physical calculation and has a decisive influence on the accuracy of calculation results. AFA3Gassemble is widely used in nuclear power plants. CENACE is an ACE format multiple-temperature continuous energy cross section library that developed by China Nuclear Data Centre. In this paper, we calculated the AFA3G assemble by RMC.We respectively used ENDF6.8/, ENDF/7 and CENACE data for calculation. The impact of nuclear data on RMC calculation is studied by comparing the results of different nuclear data.

scholarly journals Blowout Accident Impact Analysis Method for the Siting of Offshore Floating Nuclear Power Plant in Offshore Oil Fields

2019 ◽  
Vol 2019 ◽  
pp. 1-7
Author(s):  
Zhigang Lan
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Impact Analysis ◽  
Oil Spills ◽  
Oil Field ◽  
Oil Fields ◽  
Offshore Oil ◽  
The Impact

Focused on the utilization of nuclear energy in offshore oil fields, the correspondence between various hazards caused by blowout accidents (including associated, secondary, and derivative hazards) and the initiating events that may lead to accidents of offshore floating nuclear power plant (OFNPP) is established. The risk source, risk characteristics, risk evolution, and risk action mode of blowout accidents in offshore oil fields are summarized and analyzed. The impacts of blowout accident in offshore oil field on OFNPP are comprehensively analyzed, including injection combustion and spilled oil combustion induced by well blowout, drifting and explosion of deflagration vapor clouds formed by well blowouts, seawater pollution caused by blowout oil spills, the toxic gas cloud caused by well blowout, and the impact of mobile fire source formed by a burning oil spill on OFNPP at sea. The preliminary analysis methods and corresponding procedures are established for the impact of blowout accidents on offshore floating nuclear power plants in offshore oil fields, and a calculation example is given in order to further illustrate the methods.

Thermal Ageing Embrittlement of Casting Duplex Stainless Steels for Nuclear Power Plant

2010 ◽  
Author(s):  
Yuhong Yao ◽  
Jianfeng Wei ◽  
Jiangnan Liu ◽  
Zhengpin Wang ◽  
Yu Wang
Keyword(s):  
Stainless Steels ◽  
Thermal Aging ◽  
Nuclear Power ◽  
Power Plants ◽  
Ageing Time ◽  
Thermal Ageing ◽  
Duplex Stainless Steels ◽  
Impact Properties ◽  
Α Phase ◽  
The Impact

Cast duplex stainless steels (CSS) used for PWR pipes are degraded due to thermal ageing embrittlement during long-term service at 288 °C to 327 °C. Z3CN20-09M Cast duplex Stainless Steels (CSS) made in France for domestic nuclear power plants were thermally aged at 400 °C for 100 h, 300 h, 1000 h, 3000 h and 10000 h. The tensile properties and the impact properties at different thermal aging duration were measured and the effects of the thermal aging time on the microscopic structures and substructures of Z3CN20-09M were respectively investigated by optical microscopy and transmission electron microscopy. The results showed that the tensile strengths of Z3CN20-09M CSS increased gradually with the increment of the thermal ageing time, whereas the impact properties decreased with the prolonging of the thermal ageing time. After long thermal ageing time the dislocation configurations were greatly changed in austenite, and there were precipitates along the austenite-ferrite interface. Moreover, the iron-rich α phase and the chromium-rich α phase precipitated in ferrite aged for 10000h by nucleation and growth rather than the spinodal decomposition. All of above revealed that Z3CN20-09M CSS became brittle during thermal ageing.

Methodology to Evaluate Nuclear Power Plant Vulnerability to Events in the Transmission Grid

2014 ◽  
Author(s):  
Alberto Del Rosso ◽  
Jean-François Roy ◽  
Frank Rahn ◽  
Alejandro Capara
Keyword(s):  
Risk Assessment ◽  
Nuclear Power ◽  
Power Plants ◽  
Assessment Tool ◽  
Transmission Grid ◽  
Simulation Tools ◽  
Nuclear Plants ◽  
Western Electricity Coordinating Council ◽  
The Impact ◽  
The U.S

This paper presents a general approach to evaluate the risk of trip or Loss of Off-site Power (LOOP) events in nuclear power plants due to contingencies in the power grid. The proposed methodology is based on the Zone of Vulnerability concept for nuclear plants introduced by EPRI in previous work. The proposed methodology is intended to be part of an integrated probabilistic risk assessment tool that is being developed under ongoing EPRI R&D programs. A detailed analysis of many events occurred in actual nuclear plants has been performed in order to identify, classify and characterize the various vulnerability and type of failures that may affect a nuclear plant. Based the outcome of that analysis, a methodology for evaluating the impact of off-site transmission system events on nuclear plants has been outlined. It includes description of the type of contingencies and conditions that need to be included in the analysis, as well as provisions regarding the simulation tools and models that should be used in each case. The methodology is illustrated in a simplified representation of the Western Electricity Coordinating Council (WECC) system in the U.S.

Study on the Impact Limiter Design in Spent Fuel Transfer Cask in Nuclear Power Plants

2020 ◽  
Author(s):  
Yuchen Hao ◽  
Yue Li ◽  
Jinhua Wang ◽  
Bin Wu ◽  
Haitao Wang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Theoretical Method ◽  
Structure Design ◽  
Design Approach ◽  
Spent Fuel ◽  
Typical Structure ◽  
Explicit Finite Element ◽  
The Impact

Abstract In nuclear power plants, the amount of spent fuel stored on-site is limited. Therefore, it is necessary to be shipped to off-site storage or disposal facilities regularly. The key risk in the transfer of spent fuel involves a release of radiation that could cause harmful effects to people and the environment. Transfer casks with impact limiters on both ends are always employed to ensure safe containment of radioactive materials, which should be verified by the 9 meters drop test onto an unyielding surface according to IAEA SSR-6. In this paper, we focus on the influence of the impact-limiter parameters, including geometry dimensions and mechanical properties, on the results of drop events to achieve an optimized approach for design. The typical structure of impact limiter is bulk energy-absorbed material wrapped by thin stainless-steel shells. Compared to traditional wood, foam has advantages of isotropy and steady quality. In this paper, theoretical and numerical methods are both adopted to investigate the influence of impact limiters during hypothetical accidental conditions for optimizing buffer influence and protecting the internal fuel components. First of all, a series of polyurethane foam is selected according to the theoretical method, because its mechanical property is related to density. Therefore, using explicit finite element method to investigate the influence of parameters of foam in impact limiter. These discrete points from the above result can be utilized to establish damage curves by date fitting. Finally, a design approach for spent fuel transfer cask is summarized, to provide a convenient formula to predict the damage and optimize structure design in drop condition. Furthermore, this design approach can be applied in the multi-module shared system of SNF, which can contain different fuel assemblies.

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