The significance of thyroid cancer in reactor safety assessment

2010 ◽  
Vol 7 (3) ◽  
pp. 217
Author(s):  
D.J. Higson
Keyword(s):  
Thyroid Cancer ◽  
Safety Assessment ◽  
Reactor Safety

Heat Transfer Problems in Nuclear Reactor Safety

1976 ◽  
Vol 190 (1) ◽  
pp. 215-224
Author(s):  
W. B. Hall
Keyword(s):  
Heat Transfer ◽  
Safety Assessment ◽  
Nuclear Reactor ◽  
Reactor Safety ◽  
Operating Characteristics ◽  
Nuclear Reactor Safety ◽  
Thermal Explosions ◽  
Transient Boiling ◽  
Transfer Problems

Reactor safety assessment is a highly specialized topic which, in many of its aspects, depends heavily on a satisfactory understanding of a wide variety of heat transfer phenomena. It is the aim of the paper to air some of these problems outside the ranks of the reactor safety specialists. Typical liquid-cooled reactors, their operating characteristics, and some heat transfer aspects of their safety assessment are discussed: for example, transient boiling, quenching of hot surfaces and thermal explosions.

Radiation safety assessment of caregivers of thyroid cancer patients treated with 131I in Taiwan

2020 ◽  
Vol 172 ◽  
pp. 108781
Author(s):  
Li-Yen Chen ◽  
Yu-Yi Huang ◽  
Pei-Ing Lee ◽  
Sheng-Chieh Chan ◽  
Ku-Hung Lin ◽  
...  
Keyword(s):  
Thyroid Cancer ◽  
Cancer Patients ◽  
Safety Assessment ◽  
Radiation Safety

NPP

Gamification ◽  
2015 ◽  
pp. 735-769
Author(s):  
Eugene Brezhnev ◽  
Vyacheslav Kharchenko
Keyword(s):  
Nuclear Power ◽  
Input Data ◽  
Safety Assessment ◽  
Hybrid Approach ◽  
Bayesian Belief Network ◽  
Complex Nature ◽  
Reactor Safety ◽  
Integration Framework ◽  
Related Information ◽  
Different Types

The problem of the safe interaction between a Nuclear Power Plant (NPP) and a Power Grid (PG), considering the Fukushima nuclear accident, is becoming topical. There are a lot different types of influences between NPPs and PG, which stipulate NPPs' safety levels. To evaluate the influences, two metrics are proposed: linguistic and numerical. The approach to the NPP-PG safety assessment is based on the application of Bayesian Belief Network (BBN), where nodes represent different PG systems and links are stipulated by different types of influences (physical, informational, geographic, etc). It is suggested to evaluate criticality of the PG system considering the change of criticalities of all connected systems. The total criticality of each node in BBN is assessed considering particular criticalities caused by different types of influence. The complex nature of NPP and PG mutual interaction calls for the need for integration of different methods that use input data of different qualimetric nature (deterministic, stochastic, linguistic). Application of one specified group of risk methods might lead to loss and/or disregard of a part of safety-related information. BBN and Fuzzy Logic (FL) represent a basis for development of the hybrid approach to capture all information required for safety assessment of NPP – PG under uncertainties. Integration of FL-based methods and BBNs allows decreasing the amount of input information (measurements) required for safety assessment, when these methods are used independently outside from the proposed integration framework. An illustrative example for the NPP reactor safety assessment is considered in this chapter.

Discussion on the Application of Fire Probability Safety Assessment to Sodium Cooled Fast Reactor

2013 ◽  
Author(s):  
Wei Song ◽  
Hongyi Yang ◽  
Chunming Zhang ◽  
Jiaxu Zuo
Keyword(s):  
Risk Assessment ◽  
Chemical Reactions ◽  
Fast Reactor ◽  
Safety Assessment ◽  
Fire Risk ◽  
Liquid Sodium ◽  
Reactor Safety ◽  
Probabilistic Safety Assessment ◽  
China Experimental Fast Reactor ◽  
Probabilistic Safety

As sodium cooled fast reactor (SFR) uses liquid sodium as coolant, the risk of sodium fire is brought to reactor safety, which is different from conventional fire. Sodium is very chemically active, and violent chemical reactions can happen when sodium is exposed to air or water. Sodium fire is a typical and distinctive hazard in sodium cooled fast reactors, which is probably one of the main contributors to the total reactor risk. In this paper, the methodology of fast reactor sodium fire risk assessment is studied, and the principles and procedure of sodium fire probabilistic safety assessment (PSA) are given. The application of this technology in China Experimental Fast Reactor (CEFR) is explored, and several key problems which need more research in the future during the process of sodium fire probabilistic safety assessment are discussed.

scholarly journals Special topics reports for the reference tandem mirror fusion breeder. Volume 2. Reactor safety assessment

1984 ◽  
Author(s):  
I. Maya ◽  
C.G. Hoot ◽  
C.P.C. Wong ◽  
K.R. Schultz ◽  
J.K. Garner ◽  
...  
Keyword(s):  
Safety Assessment ◽  
Reactor Safety ◽  
Fusion Breeder ◽  
Tandem Mirror

scholarly journals CALCULATION OF THE MODERATOR TEMPERATURE COEFFICIENT OF REACTIVITY FOR MINIATURE NEUTRON SOURCE REACTORS

2012 ◽  
Vol 15 (3) ◽  
pp. 26-35
Author(s):  
Binh Quang Do ◽  
Hai Hoang Nguyen
Keyword(s):  
Temperature Coefficient ◽  
Neutron Source ◽  
Safety Assessment ◽  
Analytical Representation ◽  
Reactor Safety ◽  
Reactor Kinetics ◽  
Reactor Calculation ◽  
Coefficient Of Reactivity

This paper presents results of the evaluated group constants for fuel and other important materials of the Miniature Neutron Source Reactor (MNSR) and the moderator temperature coefficient of reactivity through global reactor calculation. In this study, the group constants were calculated with the WIMSD code and the global reactor calculation is accomplished by the CITATION code. This work also presents a method for evaluation of the moderator temperatures directly through the values of moderator temperature for MNSRs. This method provides simple analytical representation convenient for reactor kinetics calculation and reactor safety assessment.

scholarly journals VVUQ of a thermal-hydraulic multi-scale tool on unprotected loss of flow accident in SFR reactor

2021 ◽  
Vol 7 ◽  
pp. 3
Author(s):  
Nathalie Marie ◽  
Simon Li ◽  
Amandine Marrel ◽  
Michel Marquès ◽  
Sophie Bajard ◽  
...  
Keyword(s):  
Safety Assessment ◽  
Current Practice ◽  
Local Scale ◽  
Assessment Methodology ◽  
Reactor Safety ◽  
Regulatory Requirement ◽  
Multi Scale ◽  
Safety Studies ◽  
Station Blackout ◽  
Nuclear Safety Authority

Within the framework of the French 4th-generation Sodium-cooled Fast Reactor safety assessment, methodology on VVUQ (Verification, Validation, Uncertainty Quantification) is conducted to demonstrate that the CEA's thermal-hydraulic Scientific Computation Tools (SCTs) are effective and operational for design and safety studies purposes on this type of reactor. This VVUQ-based qualification is a regulatory requirement from the French Nuclear Safety Authority (NSA). In this paper, the current practice of VVUQ approach application for a SFR accidental transient is described with regard to the NSA requirements. It constitutes the first practical, progressively improvable approach. As the SCT is qualified for a given version on a given scenario, the transient related to a total unprotected station blackout has been selected. As it is a very complex multi-scale transient, the SCT MATHYS (which is a coupling of the CATHARE2 tool at system scale, TrioMC tool at component scale and TrioCFD tool at local scale) is used. This paper presents the preliminary VVUQ application to the qualification of this tool on this selected transient. In addition, this work underlines some feedback on design and R&D aspects that should be addressed in the future to improve the SCT.

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