Agenda

Date

September 23-24, 2019

Location

Paris, France

Conference Agenda

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Keynote Session:

Meetings International -  Conference Keynote Speaker G.S.Rothwel photo

G.S.Rothwel

OECD, France

Title: Scale Economies In Extendedsnf Storage

Biography:

G.S.Rothwel​ is the Chief Consulting Economist for Turner|Harris, specializing in all aspects of the economics of nuclearpower. G.S.Rothwel​has written extensively on energy economics and electricity markets. His book, “Economics of Nuclear Power,” was published by Routledge in 2016. 

Abstract:

Fuel is periodically replaced in nuclear power plants (NPPs). Irradiated or Spent Nuclear Fuel (SNF, where SNF could be used nuclear fuel if reprocessing facilities are available) cools in suitable facilities, where the type and the length of time depend on plans for the ultimate disposition of the SNF, for example, reprocessing or permanent long-term storage (“extended” implies storage longer than 50 years). The paper attempts to calculate the relationships between the costs and the sizes of on-site wet and on-site/off-site dry storage facilities. This is done by estimating reduced-form equations based on publicly available data, which can be modified with more recent, detailed, or proprietary data to update or extend the analysis: the values reported here should not be considered as the only possible outcomes; they are used here to understand relative NPP SNF owner economic incentives. The paper finds that once the NPP has been decommissioned, and only the on-site dry storage remains, there might not be a cost reason (from the point of view of the NPP owner/operator) to move the SNF to consolidated facilities. However, there is a consensus that consolidated facilities (a) would be more safe and secure than dispersed on-site storage locations, (b) would facilitate final disposal, and (c) can reduce the risks perceived by local communities near SNF storage facilities.

Meetings International -  Conference Keynote Speaker Yanhua Zheng photo

Yanhua Zheng

Tsinghua University, China

Title: Inherent safety and ATWS analysis of High Temperature Gas-cooled Reactor

Biography:

Yanhua Zheng has her expertise in nuclear reactor thermal-hydraulic design and accident analysis. She is responsible for the thermal-hydraulic design and accident analysis of the Chinese 200 MWe HTR-PM project. She has comprehensive and in-depth study on typical accidents and key phenomena of the pebble-bed HTGR, especially on the water-ingress accident, air-ingress accident, uncertainty analysis and so on. Her research covers the software development and V&V, accident management procedure development, and BDBA mitigation method design of HTGR. She is also the leading researcher or main participant of several National Science & Technology Major Projects and National Natural Science Foundation Projects of China.

 

Abstract:

The modular high temperature gas-cooled reactor (HTGR), recognized as a candidate for the Generation IV nuclear energy system technology, has well-known inherent safety features. A commercial-scale 200 MWe Pebble-bed Modular High Temperature gas-cooled Reactor (HTR-PM) has been designed and constructed in Shandong Province, China. Most of the construction and installation work have been finished and the connection to the electric grid will be expected in 2019 or the first half of 2020.In this paper, the design and the inherent safety feature of the HTR-PM has been introduced. Several Anticipated Transient Without Scram (ATWS) accidents, a type of Beyond Design Basis Accident (BDBA) receiving high attention especially in Pressurized Water Reactor (PWR) analysis, have been studied, including the reactivity introduction ATWS, loss of off-site power ATWS, depressurized loss of coolant ATWS.Calculation results prove that, even in such kind of BDBAs with very low probability, the inherent safety design of HTR-PM can guarantee the reactor shut-down itself by negative temperature feedback. During the accidents, the decay heat of the reactor can be transferred to the environment safely by heat conduction, natural convection and radiation, and the fuel temperature and the reactor pressure vessel (RPV) temperature would never exceed the limitation. The large release of the fission products would not happen.

Meetings International -  Conference Keynote Speaker Yan Wang photo

Yan Wang

Institute of Nuclear and New Energy Technology, China

Title: Some Preliminary Thermal-Hydraulic Safety Analysis on the NHR-200II Reactor

Biography:

Yan Wang is working on Institute of Nuclear and New Energy Technology, China.

Abstract:

The NHR-200II as a new 200 MWth nuclear heating reactor, which adopts natural circulation reactor with high passive safety could serve as a safe and economic energy source for combined heat and power of city. The thermal-hydraulic transient response of reactor system during the postulated loss of coolant accident have a significant impact on the containment design and the evaluation on passive safety of reactor. In this research, based on the preliminary design of the NHR-200 II at the current stage, the sub-channel calculation of reactor core and the analysis on some typical loss of coolant accidents were carried on by using COBRA and PCNHR code which was a validated transient analysis programs developed by Tsinghua University. The calculation results show that the NHR-200II has enough safety margin during the normal operation. Under the postulated loss of coolant accident, the reactor core of the NHR-200 II is able to keep being coved by the residual coolant in the reactor pressure vessel during the transient and the relative safety parameters are effectively controlled under the design limitation values without special safety injection like general PWR, which indicates the good safety feature of the NHR-200 II design.

 

 

Meetings International -  Conference Keynote Speaker Ramandeep Singh Sidhu photo

Ramandeep Singh Sidhu

Bathinda Colleage of Law

Title: Right of recourse under nuclear civil liability law of India

Biography:

Ramandeep Singh Sidhu has Joined Law College as Assistant Professor in January 2017  Working as Officiating Principal in the college since August2018. 

Abstract:

The Civil Liability for Nuclear Damage Act, 2010 follows global practice in the field of nuclear civil liability and has a specific provision that enables a nuclear operator to exercise right of recourse against a supplier. The Civil Liability for Nuclear Damage Rules, 2011 provides explanation regarding the provisions enshrined in the Act. The Rules explains about the Supplier which has been formulated based on the industry practices of nuclear sector. The paper analyses the Civil Liability for Nuclear Damage Act and Rules on the issue of compatibility of right of recourse with the international nuclear civil liability principles and conventions, particularly with the Convention on Supplementary Compensation for Nuclear Damage, 1997. This paper also discusses the issue of right of recourse by nuclear operator against the supplier in the light of explanation about a supplier as provided under the Act and Rules.

 

Meetings International -  Conference Keynote Speaker Irena Kratochvílová photo

Irena Kratochvílová

1Institute of Physics of the Czech Academy of Sciences, ,

Title: Zr alloy protection against high-temperature oxidation: double-layered coating with active and passive functional properties

Biography:

Irena Kratochvilova obtained very important results which were published in prestigious journals and were cited by scientific community. She has published 85 papers in impacted journals.  She has been working in the Institute of Physics AS CR and teaching at the Faculty of Nuclear Engineering Czech Technical University.

 

Abstract:

In this work we investigated the hot steam oxidation of ZIRLO fuel cladding coated with a double layer consisting of 500 nm nanocrystalline diamond (NCD) as the bottom layer and 2 mm chromium-aluminum-silicon nitride (CrAlSiN) as the upper layer. Coated and noncoated ZIRLO samples were exposed for 4 days at 400 °C in an autoclave (working water-cooled nuclear reactor temperature) and for 60 minutes at 1000 °C (nuclear reactor accident temperature) in a hot steam furnace. We have shown that the NCD coating protects the Zr alloy surface against oxidation in an active way: carbon from NCD layer enters the Zr alloy surface and, by changing the physical and chemical properties of the Zr cladding tube surface, limits the Zr oxidation process. In contrast, the passive CrAlSiN coating prevents the Zr cladding tube surface from coming into physical contact with the hot steam. The advantages of the double layer were demonstrated, particularly in terms of hot (accident-temperature) oxidation kinetics: in the initial stage, CrAlSiN acts as an impermeable barrier, but after a longer time, the protection by CrAlSiN decreases as an increasing number of cracks in the carbon of NCD penetrate the Zr cladding surface and worsen conditions for Zr oxidation. For the double-layer coating, the underlying NCD layer mitigates thermal expansion, reducing cracks and defects in upper layer CrAlSiN.