Published 2021 | Version v1
Book

First approach to analyse control rod withdrawal transient in SFR with a multi-physics methodology

  • 1. CEA - Centre de Cadarache, DES, DER, F-13108 Saint-Paul-lez-Durance (France)
  • 2. CEA - Centre de Cadarache, DES, DEC, F-13108 Saint-Paul-lez-Durance (France)
  • 3. LPSC, 38000 Grenoble (France)

Description

Control Rod Withdrawal (CRW) accident in Sodium-Cooled Fast reactors (SFR) is considered, from a safety point of view, as a potential trigger of a core melting accident. It occurs when a Control Rod (CR) is accidentally extracted from the core. This causes a power increase locally around the faulty Control Rod and globally in the core, followed by fuel and sodium temperature increases. Temperature variations induce neutron feedback effects, expansion of materials and mechanical stresses on clads. Then, high temperatures and stresses can respectively lead to fuel melting and clad failures. The potential consequences of this accident requires safety studies to be made, at an early design stage, to ensure manageable behaviour of reactor. Such studies rely on fast calculation schemes, which allow performing probabilistic safety assessments. As high fidelity coupling schemes are time and resource consuming, conservative approach based on single-physics (neutronics, thermal-hydraulics and thermomechanics) discretization are often preferred at this stage. It allows treating each physics domain according to the optimal time and space scales but requires introducing important approximations to take into account the multi-physics feedbacks and couplings. A local multi-physics approach is therefore necessary to analyse CRW transient more accurately. In order to improve both the understanding and the modelling of this transient, this paper will describe the multi-physics phenomenology involved in a CRW and presents a simplified multi-physics coupling scheme compatible with the possibility of performing probabilistic safety assessments. A first implemented coupling scheme, MORPHEE based on improved point kinetics, core thermo-hydraulics and in pin fuel thermo-mechanical models is presented and tested on ASTRID core. A comparison to GERMINAL-V2 shows relatively good agreement on thermal resolution. Further development will focus on transient behaviour of thermal properties and mechanical strains to allow comparison with current safety criteria

Availability note (English)

Available from the American Nuclear Society, 555 North Kensington Avenue, La Grange Park, Illinois 60526 (US)
Part of:
Proceedings of the international conference on mathematics and computational methods applied to nuclear science and engineering - M and C 2021

Additional details

Publishing Information

Publisher
ANS - American Nuclear Society
Imprint Place
La Grange Park (United States)
Imprint Title
Proceedings of the international conference on mathematics and computational methods applied to nuclear science and engineering - M and C 2021
Imprint Pagination
2418 p.
Journal Page Range
p. 2257-2266

Conference

Title
International conference on mathematics and computational methods applied to nuclear science and engineering
Acronym
M and C 2021
Dates
3-7 Oct 2021
Place
Raleigh, NC (United States)

Optional Information

Notes
8 refs.; Virtual meeting