Institute for Neutron Physics and Reactor Technology (INR)

Nuclear Safety

Nuclear spent fuel, especially the long living fission products as well as the minor actinides requires special handling. A promising way to reduce the amount of long-living waste is to convert it into short-living waste. This can be achieved in fast critical or subcritical reactors.  The institute contributes to respective projects of the EU Framework Programs and to industrial partner funded projects, concentrating on the safety evaluation of innovative reactor concepts, dedicated to transmutation.

For assessing the safe response of those fast neutron system under accidental conditions (Design Basis Accidents or Beyond Design Basis Accidents), fast running in-house codes like SIM-family code are used. For Beyond Design Basis Accidents, the system code SAS-SFR, further developed and maintained in our institute, CONTAIN-LMR and ASTEC-Na are applied, among others. The integral approach of the safety analysis includes all main physical topics involved: fuel pin thermal-mechanics, coolant thermal-hydraulics (both single- and two-phase flow conditions), neutron physics, thermal-chemical fuel-coolant reactions and coolant-air and ‑water interactions (i.e. Sodium fires).

Our institute is supporting the European Projects of the H2020 Framework Program dealing with the design of systems and measurements that prevent and mitigate the consequences of accidental conditions in innovative systems.

The data banks are fed by the experimental tests held by German programs (NKS, KNS, FAUNA, NALA, etc.) and by international collaborative programs (CABRI, SCARABEE, etc.). Today, such data banks are still of great value to support the validation of computational codes.

Recent EU projects are: JASMIN, ESNII+ and ESFR-SMART.

Other Gen IV systems are Lead cooled reactors (LFR) (ELSY, LEADER) and GFR Gas Fast Reactors (GCFR, GoFastR).

 

Publications

Transmutation

Perez-Martin, S, Ponomarev, A., Krüssmann, R., Pfrang, W. Importance of Fuel Thermo-mechanical Properties in an ULOF Transient of a Sodium-cooled Fast Reactor Used for Minor Actinides Transmutation. Proceedings of the ICAPP-2013, Jeju Island, Korea, April 2013.

S. Perez-Martin, A. Ponomarev, R. Kruessmann, D. Struwe, et al. Safety Analysis of a Sodium-Cooled Fast Reactor with Transmutation Capabilities. European Nuclear Conf. (ENC 2012), Manchester, GB, December 9-12, 2012 Transactions Advanced Reactors Bruxelles: European Nuclear Society 2012 www.euronuclear.org/events/enc/enc2012/transactions.htm

D. De Bruyn, R. Fernandez, L. Mansani, A. Woaye-Hune, M. Sarroto, E. Bubelis. The Fast-spectrum Transmutation Experimental Facility FASTEF: Main Design Achievements (Part 1: Core & Primary System) Within The FP7-CDT Collaborative Project Of The European Commission. Proceedings of the 2012 International Congress on Advances in Nuclear Power Plants – ICAPP 2012, 24-28 June, 2012. – Chicago, USA. 2012. P. 1-9.

Experiments and Codes

J. Pacio, M. Daubner, F. Fellmoser, W. Hering, W. Jäger, R. Stieglitz, T. Wetzel. 3.3 - Construction of experimental liquid-metal facilities. Thermal Hydraulics Aspects of Liquid Metal Cooled Nuclear Reactor 2019, Pages 107-125 doi.org/10.1016/B978-0-08-101980-1.00013-2

W. Hering, A. Onea, A. Jianu, J. Reiser, S. Ulrich, R. Stieglitz. Liquid metals, materials and safety measures to progress to CSP 2.0 AIP Conference Proceedings 2126, 080002 (2019) doi.org/10.1063/1.5117597

LFR

G. Grasso, F. Lodi, P. Romojaro, N. Garcia-Herranz, F. Alvarez-Velarde, D. Lopez, E. Bubelis, G. Bandini.  Stress-testing the ALFRED design - Part I: Impact of nuclear data uncertainties on Design Extension Conditions transients. Progress in Nuclear Energy (2018) 106 372-386 doi.org/10.1016/j.pnucene.2018.03.013

Grasso, G., Lodi, F., Bubelis, E., et al. Stress-testing the ALFRED design – Part I: Impact of nuclear data uncertainties on Design Extension Conditions transients. Progress in Nuclear Energy Volume (2018) 106 372-386 doi.org/10.1016/j.pnucene.2018.03.013

E. Bubelis, M. Schikorr, L. Mansani, G. Bandini, K. Mikityuk, Y. Zhang, G. Geffraye. Safety analysis results of the DBC transients performed for the ALFRED reactor. Proceedings of the IAEA International Conference on Fast Reactors and Related Fuel Cycles: Safe Technologies and Sustainable Scenarios (FR13), 4-7 March, 2013. – Paris, France, 2013. P. 1-10.

Alemberti A, Carlsson J, Malambu E, et al. European lead fast reactor – ELSY. Nuclear Engineering and Design (2011) 241(9) 3470-3480 doi.org/10.1016/j.nucengdes.2011.03.029

GFR

Chen X.-N., Andriolo L., Rineiski A., Bubelis E., Mayer G., Bentivoglio F.  Extension and validation of SIMMER III code for gas cooled fast reactor. Annals of Nuclear Energy (2015) 81 320-331 doi.org/10.1016/j.anucene.2015.03.007

ADS

X-N. Chen, R. Li, F. Belloni, F. Gabrielli, et al. Safety studies for the MYRRHA critical core with the SIMMER-III code. Annals of Nuclear Energy (2017) 110 1030-1042 doi.org/10.1016/j.anucene.2017.08.021

Bubelis E., Jaeger W., Bandini G., Alemberti A., Palmero M. Assessment of the enhanced DHRS configuration for MYRRHA reactor. Nuclear Engineering and Design (2016) 307 181-187 doi.org/10.1016/j.nucengdes.2016.07.017

Bubelis, E., Jaeger, W., et al. Assessment of the enhanced DHRS configuration for MYRRHA reactor. Nuclear Engineering and Design (2016) 307 181-187 doi.org/10.1016/j.nucengdes.2016.07.017

Dagan, R., Jianu, A., Rimpault, G., Weisenburger, A., Schikorr, M. The consequences of a sharp temperature change in the fuel pins of an accelerator-driven subcritical system. Nuclear Technology Volume 184, Issue 2, November 2013, Pages 210-216 doi.org/10.13182/NT13-A22316

Schikorr W, Struwe D. On the importance of reactivity feedback coefficients in ADS systems. International Meeting on Nuclear Applications of Accelerator Technology: Accelerator Application in a Nuclear Renaissance (2003) 684-686

Dagan R, Brooders C, Struwe D. Source trip effects on transient ADS behaviour. International Meeting on Nuclear Applications of Accelerator Technology: Accelerator Application in a Nuclear Renaissance (2003) 467-470