Safety Analysis of ADS

Accelerator Driven System (ADS)

The ADS is a coupled system of a subcritical reactor and an external proton accelerator..ADS may utilize fuels with large fractions of transuranic elements (TRUs) because of ADS large safety margins related to sub-criticality. Therefore ADS was proposed for transmutation.

The proton beam enters the center of the reactor core through a vacuum tube and hits the lead target where neutrons are emitted.

The pool type reactor is cooled by liquid lead or lead-bismuth. In the PDS-XADS design, a bubble lift pump supports natural convection and transports thermal energy to the heat exchangers.  The following schematic  illustrates the mechanism of an accelerator driven system.

 

 

Safety Aspects and Analyses

Compared to critical reactors, the subcritical ADS has a larger safety margin for reactivity insertion. Nevertheless safety analyses have to take into account many new and innovative features of such systems:

  • new system of subcritical reactor and external neutron source
  • new fuels: dedicated transmutation fuel
  • new coolant: lead, lead-bismuth eutectic (LBE)

 

The sub-criticality is an essential advantage for ADS, as mentioned before. There is an additional safety effect:  in the case of a loss of power (station black-out) the accelerator beam providing the external neutron source is shut off automatically and the chain reaction in the reactor stops.

Due to the external source, a new class of transients /accidents related to the accelerator side have to be analysed. These transients are due to possible change of the source strength.

Other transients, related to the reactor side, are possible due to the same initiators as in critical reactors. They result in smaller power variations as follows from analyses of transients such as loss of flow (ULOF), transient over power (UTOP), blockage accident (BA), etc.

 

Recent European Projects

Recently we were involved in CDT, SEARCH and MAXSIMA projects in 7th framework of EU programme and performed safety analyses for MYRRHA with SIMMER code. The principal partner is SCK-CEN. For details of the projects and the MYRRHA reactor we refer readers to following links:  

CDT: https://cordis.europa.eu/project/id/232527/reporting

SEARCH: https://cordis.europa.eu/project/id/295736

MAXSIMA: https://cordis.europa.eu/project/id/323312

MYRRHA: https://www.sckcen.be/en/projects/myrrha

Besides we performed typical unprotected over power for the critical mode of MYRRHA, we developed the so-called fine mesh approach for both SIMMER-III (2-D) and SIMMER-IV (3-D), so that we could solve reactor transient problems with SIMMER in a resolution of sub-channel scale.

Below there is an example of simulation of a central fuel assembly blockage accident, i.e. active core entrance blockage with all sub-channels blocked [Chen et al. 2017]. The figure shows the development of the FA damage with material distributions. At t = 9 s after the blockage (Time = 69 s), the upper part of central pins starts to melt and the fuel becomes particles locally. At t = 33 s after the blockage (Time = 93 s), the whole FA pins are molten and the fuel becomes ‘‘particles” totally in this FA. 2 s later, i.e. at t = 35 s after the blockage the inter-wrapper can-wall in the upper part of the active core is molten. The fuel particles and molten steel can go through the can-wall breakup into the inter-wrapper gap, where the flow is small but normal. Then the fuel particles will escape from the core to the upper plenum. Consequently the reactivity and power will decrease abruptly, as can be seen in the next associated figure.

 

 

 

The following figure illustrates the 3-D sub-channel approach. This approach was validated in steady-state by a code-to-code comparison and by a code-to-experiment comparison, using KALLA Lab results.  For details see the following journal papers:

  1. X.-N. Chen , R. Li, A. Rineiski, W. Jäger, Macroscopic pin bundle model and its blockage simulations. Energy Conversion and Management 91 (2015) 93–100.
  2. R. Li, X.-N. Chen, A. Rineiski, V. Moreau, Studies of fuel dispersion after pin failure: Analysis of assumed blockage accidents for the MYRRHA–FASTEF critical core. Annals of Nuclear Energy 79 (2015) 31–42.
  3. X.-N. Chen et al., Safety studies for the MYRRHA critical core with the SIMMER-III code. Annals of Nuclear Energy 110 (2017) 1030–1042.
  4. R. Li, X.-N. Chen, L. Andriolo, A. Rineiski, 3D numerical study of LBE-cooled fuel assembly in MYRRHA using SIMMER-IV code. Annals of Nuclear Energy 104 (2017) 42–52.