SharePoint
Aide
IRSN, Institut de radioprotection et de sûreté nucléaire

Search our site :

ok

Contact us :

ok
En Fr

Enhancing Nuclear Safety


Research

Computer codes

SCANAIR computer code


​Last update on April 2015


The SCANAIR (French acronym for “system of codes for analyzing reactivity-initiated accidents”) computer code was developed by the IRSN in the early 1990s. It is designed to simulate the thermo-mechanical behavior of fuel rods during a reactivity-initiated accident (RIA) in a pressurized water reactor (PWR).



Background and objectives


In a pressurized water reactor, the typical reactivity-initiated accident is a rod ejection accident. It is caused by the failure in the pressure casing housing a control rod drive mechanism, which is an assembly of rods made of elements that absorb neutrons and thus regulate the nuclear reaction. In these accidents, the control rod is ejected due to the pressure differential, of around 150 bar, between the reactor's core and the containment building. The local nuclear reactions subsequently go out of control very quickly (in tens of milliseconds), causing a rapid and violent generation of energy within the fuel. However, this runaway process is limited thanks to an opposing force, the Doppler effect, where the increase in fuel temperature means more neutrons are absorbed, thus limiting nuclear fission.


Scanair-Schema-barres-REP.jpg

Diagram of control rods in a PWR (from top to bottom: spider assembly, control rod, retaining spring, top nozzle, top grid, guide tube, fuel rod, mixing grid, bottom grid, bottom nozzle)

© IRSN


The dramatic energy increase causes transformations in the fuel. The first is that it expands, for two reasons: the expansion of the fuel material itself, and that of the gases produced by nuclear fission in normal operations (essentially krypton and xenon) that are trapped in the fuel. This expansion exerts mechanical pressure on the cladding. Secondly, some of the fission gases are released by the fuel and spread into the existing interstices between the fuel pellets and cladding, again exerting pressure on the latter. These phenomena are even more sensitive due to the fact that a large amount of energy is injected into the rod and the burn-up is high.


An important issue that needs to be considered during the RIA study is an estimation of the conditions that could damage the integrity of the cladding, the first confinement barrier against the fuel and fission products, which could then lead to radioactive materials being released into the primary circuit and to interaction between the hot fuel fragments and the coolant, further exacerbating the accident.


Scanair-Gaine-RIA.jpg

Failed cladding during a RIA test in the CABRI test reactor © IRSN


In the years that followed the Chernobyl disaster (1986), research programs were carried out in Japan and France that included series of tests conducted in the CABRI test reactor in Cadarache. One of the chief aims of these programs was to further our understanding of the physical phenomena that could affect the rod cladding's seal-tightness and result in fuel being ejected into the primary circuit.


At the start of the 90s, while these research programs were still ongoing, the SCANAIR software program was developed by IRSN in order to simulate the behavior of a fuel rod in a RIA, from the moment power starts to be inserted up until the failure of the cladding, paying particular attention to the specific effects linked to the irradiated fuel. The main aims of this software are to:

  • help interpret the results from test programs;

  • transpose the test results onto a power reactor;

  • check the validity of safety criteria for different types of fuel and cladding offered by operators;

  • carry out studies to assess the safety margins for the different types of fuel rods used;

  • contribute towards the critical analysis of new safety criteria.



Models


The SCANAIR software is composed of three main modules that are closely interlinked:

  • The first module is dedicated to modeling the thermal behavior of the fuel rod and the thermo-hydraulic behavior of the coolant. It calculates the radial heat conduction in the fuel and in the cladding, as well as the heat transfer in the pellet-cladding gap and between the cladding and the coolant, taking into account the various boiling regimes.
  • The second module is dedicated to modeling the behavior of the gases in the rod. In particular, it calculates the expansion of fission gas bubbles, the rupture of the fuel grain boundaries, and the release and flow of gases towards the rod's free volume.

  • The third module is dedicated to modeling the mechanical behavior of the fuel rod and cladding. It calculates the contribution of different kinds of deformation (heat, elastic, plastic, viscoplastic, caused by the existence of cracks, and due to swelling caused by pressurization of fission gases) to the overall deformation of the rod. Accounting for the initial extent of corrosion and hydriding in the cladding, modeling the thermo-mechanical behavior of the cladding during an accidental power transient makes it possible to predict if there will be a failure of the cladding.


Scanair-Schema-Flux.png

Overview diagram of date flow between the different SCANAIR modules © IRSN


Scanair-Calculs.png

Evolution of temperature and deformation field of a test-rod, as tested in the CABRI reactor calculated by SCANAIR (x-axis: radial coordinates, y-axis: axial coordinates) © IRSN



Validation and outlook


The mechanical properties of the zirconium alloy cladding materials used by SCANAIR are mostly taken from the PROMETRA experimental program. The heat exchange correlations between the cladding and coolant during rapid transient have been validated using separate effect tests, under conditions representative of those found in a PWR in the French Atomic Energy Commission's PATRICIA loop, and during the NSRR program in standing water conditions. The SCANAIR software has been validated for pellet-cladding mechanical interaction (PCMI) in 12 experiments in the CABRI-PWR-Na program, as well as the first two tests of the CIP carried out in the CABRI reactor's sodium test loop. The validation has been further complemented by tests conducted in the Japanese NSRR reactor.


Other tests have been scheduled as part of the CIP (CABRI International Program), led by the OECD. These tests will be performed in a new water test loop that will make it possible to reach the coolant boiling crisis. The results of these tests, carried out in conditions that are most similar to typical reactor functioning, can be used to improve and validate modeling of the post-boiling crisis phase (post-DNB), likely to damage the cladding's integrity due to excessive swelling.


The results of the FGD (Fission Gas Dynamics) test program, to be conducted in the Japanese NSRR reactor, will provide data that will help improve modeling the behavior of fission gases during a power transient.



Collaboration


Development of the SCANAIR software benefits from the support of the French operator EDF, which uses the software to support its safety analyses. As part of the software agreement with the IRSN, a number of international safety bodies (SSM in Sweden, VTT in Finland, CSN and CIEMAT in Spain, and JAEA in Japan) are permitted to conduct studies using SCANAIR, or to help develop new models.


Send Print

Involved IRSN laboratory

Presentation sheet

Publications

Contact


Close

Send to a friend

The information you provide in this page are single use only and will not be saved.
* Required fields

Recipient's email:*  

Sign with your name:* 

Type your email address:*   

Add a message :

Do you want to receive a copy of this email?

Send

Cancel

Close

WP_IMPRIMER_TITLE

WP_IMPRIMER_MESSAGE

Back

Ok