Nuclear installation licensing and democratic decision making in Finland: a case study regarding the Olkiluoto 3 nuclear power plant unit and the final disposal repository for spent nuclear fuel

New methods and approaches to development of nuclear power plant unit imitation models development. An imitation model is proposed for nuclear power plant unit developed on the basis of oriented graph applied to unit flow diagram description logic-numerical operators for calculation of parameters and characteristics determining its operation efficiency, reliability and safety. Automatic system is described developed on the basis of imitation model for decision-making support, applied to analysis of VVER nuclear power plant unit operation efficiency with due consideration of equipment operation reliability and safety indices.

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Estimate of the equilibrium radioactivity of fission products in spent nuclear fuel from the smolensk nuclear power plant

Atomic Energy ◽

10.1007/bf02415467 ◽

1997 ◽

Vol 82 (1) ◽

pp. 78-79 ◽

Cited By ~ 1

Author(s):

D. K. Ershov

Keyword(s):

Power Plant ◽

Nuclear Power Plant ◽

Nuclear Fuel ◽

Nuclear Power ◽

Spent Nuclear Fuel ◽

Fission Products

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Thorium MSR as a Small-Scale Energy Source: Opportunities for Japan

ASME 2011 Small Modular Reactors Symposium ◽

10.1115/smr2011-6655 ◽

2011 ◽

Author(s):

Takashi Kamei

Keyword(s):

Power Plant ◽

Nuclear Power Plant ◽

Nuclear Fuel ◽

Nuclear Power ◽

Power Plants ◽

Large Scale ◽

Spent Nuclear Fuel ◽

Small Scale ◽

Fissile Material ◽

Implementation Capacity

Even after the huge impact of Fukushima Daiichi nuclear power plant accident, Japan has to establish its energy supply system satisfying requirements of both global warming and resistibility of natural disaster. Nuclear power has a potential to reduce carbon emission but large-scale and centralized nuclear power plant may lose large volume of electricity supply at once. Small-scale nuclear power plants will bring solution in Japan. Thorium molten-salt reactor (MSR) is selected to simulate implementation capacity of small reactors in Japan. In order to use thorium as nuclear fuel, fissionable isotope is indispensable since natural thorium does not include fissile material. Japan owns plutonium in spent nuclear fuel of uranium usage. Quantitative evaluation of implementing capacity of thorium MSR in Japan by using plutonium accumulated in Japan. Implementation capacity of thorium MSR will be about 38 GWe and 11.2 GWe in the maximum and minimum cases at 2050, respectively.

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Using CFD to Improve the Thermal Characteristic Analysis Methodology of Dry Storage System for Chinshan Nuclear Power Plant

ASME 2010 Power Conference ◽

10.1115/power2010-27302 ◽

2010 ◽

Author(s):

Yung-Shin Tseng ◽

Jong-Rong Wang ◽

Chi-Hung Lin ◽

Chunkuan Shin ◽

F. Peter Tsai

Keyword(s):

Power Plant ◽

Nuclear Power Plant ◽

Nuclear Fuel ◽

Nuclear Power ◽

Spent Nuclear Fuel ◽

Storage System ◽

Thermal Characteristic ◽

Life Extension ◽

Fuel Management ◽

Dry Storage

Chinshan Nuclear Power Plant (CSNPP) is a two-unit BWR4 plant with 1804MWt power per unit. Taipower Co., the owner of the plant is preparing the life extension procedure to extend the CSNPP operation time. In order to meet the life extension requirement, many issues need to be solved before life extension licensing, such as the spent nuclear fuel management, structure aging etc. For the spent nuclear fuel management, ROC Atomic Energy Council (ROCAEC) certified method is employed to analyze the thermal behaviors of Dry Storage System (DSS). This method uses ANSYS coupled with RELAP5-3D to solve the thermal characteristic and successfully accomplish the licensing procedure of the Chinshan Nuclear Dry Storage Project. However, further validation results demonstrate that the coupled method still exists uncertainty and deficiency. In this study, a new Computational Fluid Dynamics (CFD) numerical model for spend nuclear fuel (NSF) dry storage system (DSS) has been developed to improve the accuracy of DSS thermal analysis results. Its accuracy has been validated by comparing the temperature predictions with the experimental results of VSC-17 DSS. It has been found that the thermal behaviors and physical phenomena in the DSS could be predicted with good agreement for the measurements. Moreover, the uncertainty and reasonableness of results in previous method can be improved by the new thermal analyses methodology.

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The impact of Ignalina Nuclear Power Plant on Carbon-14 content in Lake Druksiai, Lithuania

10.5194/egusphere-egu2020-18812 ◽

2020 ◽

Author(s):

Laurynas Butkus ◽

Rūta Barisevičiūtė ◽

Žilvinas Ežerinskis ◽

Justina Šapolaitė ◽

Evaldas Maceika ◽

...

Keyword(s):

Power Plant ◽

Nuclear Power Plant ◽

Nuclear Fuel ◽

Sediment Core ◽

Nuclear Power ◽

Power Plants ◽

Spent Nuclear Fuel ◽

Carbon 14 ◽

Fish Samples ◽

The Impact

Nuclear Power Plants (NPPs) and nuclear fuel reprocessing sites are main producers of anthropogenic radiocarbon. Anthropogenic 14C can be released into the environment in gaseous forms, with liquid effluents or with spent nuclear fuel [1]. During photosynthesis radiocarbon can be easily assimilated into the plants. As a result, carbon-14 can be transported through the food chain and accumulate in a human body. Therefore, radiocarbon is considered a primary source of increased human radiation dose from industrial nuclear activities [2].Main goal of this research was to evaluate the influence Ignalina NPP on carbon-14 content in the Lake Druksiai. The sediment core was collected from the Lake Druksiai. The ages of sediment layers were estimated using 137Cs and 210Pb dating methods. ABA (acid-base-acid) chemical pretreatment procedure was used to extract humin (HM) and humic acid (HA) fractions from the sediments. Chemically pretreated samples were graphitized with the Automated Graphitization Equipment AGE 3 (IonPlus AG). Carbon-14 measurements in prepared samples were performed using the single stage accelerator mass spectrometer (SSAMS, NEC, USA).Radiocarbon content was measured in the sediment core which covers all phases of the NPP exploitation (commissioning, operation and decommissioning). These measurements in HM and HA fractions showed that after the start of the operation of the Ignalina NPP in 1983, the 14C concentration in these organic fractions increased by 4 pMC and 3 pMC, respectively. In addition, a sharp increase of radiocarbon content (concentration almost doubled) in HA fraction was observed in the year 1999. Similar increase in 14C activity in fish samples from Lake Druksiai was measured. In HM fraction such drastic changes in radiocarbon concentration were not observed. These results suggest that 14C enriched effluents were released from the Ignalina NPP in 1999.[1] Z. Ezerinskis et al., Annual Variations of 14C Concentration in the Tree Rings in the Vicinity of Ignalina Nuclear Power Plant, Radiocarbon 60, 1227&#8211;1236 (2018).[2] IAEA, Generic Models for Use in Assessing the Impact of Discharges of Radioactive Substances to the Environment (2001).

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Assessment of activity resources by nomenclature of main radiation-hazardous radionuclides during the long-term management of spent nuclear fuel of the Belarusian NPP

Proceedings of the National Academy of Sciences of Belarus Physical-Technical Series ◽

10.29235/1561-8358-2020-65-2-243-251 ◽

2020 ◽

Vol 65 (2) ◽

pp. 243-251

Author(s):

J. A. Korchova ◽

N. V. Harbachova ◽

N. D. Kuzmina ◽

N. V. Kulich

Keyword(s):

Power Plant ◽

Nuclear Power Plant ◽

Nuclear Fuel ◽

Management Strategy ◽

Nuclear Power ◽

Spent Nuclear Fuel ◽

Fission Products ◽

Fuel Management ◽

Probabilistic Forecasting ◽

The Government

The purpose of the study is calculation research of the radiation characteristics of fission products and actinides at different phases of spent nuclear fuel (SNF) management of the Belarusian Nuclear Power Plant (NPP). The study is aimed at the scientific support of the government decision as determined by the “On approval of the spent nuclear fuel management Strategy of the Belarusian nuclear power plant”. А probabilistic forecasting model and an effective code CUB for the spent nuclear fuel radioactivity inventory assessment were developed by the authors. Radionuclides activities as function of nuclear fuel burnup for nuclear fuel with the initial enrichment on the 235U equals to 4.81 % on the base of approximation relations of Regulation RB-093-14 (Moskow, 2014) have been calculated. Basic relations between specific activities of the main hazardous fission products and actinide, the SNF burnup and initial degree of fuel enrichment were analyzed. The rates of decrease of individual and total fission products and actinides activities of the Units №1 and 2 of the Belarusian NPP were obtained depending on the specific phase of spent SNF management. The results are of value for decision-making on ecology acceptable SNF management option introduced by Spent Nuclear Fuel Management Strategy of the Belarusian NPP.

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