Sodium void effect

In the event of coolant loss (by boiling or gas intrusion) traditional large LMFR cores show a significant reactivity increase. Some experts consider that in the LMFR, an increase of the coolant temperature should initiate an expansion of the absorber rod guide structure, of the fuel in the axial direction and of the core grid plate in the radial direction, resulting in negative reactivity coefficients counteracting the positive sodium void coefficient in large LMFR. However, taking into account that the cooling disruptions and sodium boiling might be on a time scale much shorter than the time scale of the passive negative feedbacks, there is a strong incentive to reduce the positive sodium void coefficient in large LMFR cores. The idea of the IAEA/CEC benchmark calculation 1990/94 [10] was to investigate the capability

of reducing the sodium voiding feedback reactivity of an axially heterogeneous core where a sodium plenum is introduced above the core instead of the upper axial blanket.

The benchmark has shown that the overall void reactivity effect of the reference 2 100 MW(th) MOX-fuelled reactor core with postulated voiding configuration might be close to zero. Future investigations were required to determine the differences in severe accident response, in order to estimate the improvement in overall safety that could be achieved from a reduction in the sodium void worth for reactor cores. Knowledge of in-core sodium boiling and voiding physical phenomena was necessary for the determination of the reactivity insertion rate.

There is a feature of contemporary fast reactors which have attracted particular attention on the grounds of safety: positive reactivity transients which may be exacerbated by the fact that the coolant void reactivity coefficient is under some circumstances positive.

An IAEA/EC 1994/98 project assessed the capability of reducing the sodium void feedback reactivity of the homogeneous core by introducing a volume of sodium or plenum immediately above the core, as was proposed by Russian specialists in advanced reactors.

It has been established by analyses done in various countries in the framework of the Coordinated Research Project (CRP) that, in the event of an accident in which the sodium is overheated, boiling in the core generates vapour which expands rapidly and voids not only the core where it causes a positive reactivity change, but also the plenum where it causes a negative change so that the overall change is less positive or even negative. This approach has been adopted in the BN-800 fuel assembly design (Fig. 1) [11].

Main advantages of the innovative BN-800 type core design are to be seen in providing an additional inherently activated safety margin of preventing fuel pin failure or local boiling in the domain of operational and severe transients to be considered in the design basis. These features complement well the large margin to fuel pin failure achieved already with the hollow pellet fuel pin design and a clad material providing ductility even under high dose loads. In the beyond design basis accident, domain some clear advantages of the innovative core design have been identified:

─ Unprotected reactivity initiated accidents most probably lead to an early reactor shutdown either due to pre-failure in-pin fuel relocation and/or due to a rapid fuel dispersal after fuel pin failure in a few subassembly groups. Linear ratings at failure conditions are most probably high, i.e. at about 1000 W/cm and more. Evaluations of the long term coolability of the established core configuration after reactivity initiated accidents were not part of this comparative exercise. They need careful consideration to evaluate potential consequences of a thermally induced subassembly to subassembly propagation.

─ In case of unprotected loss-of-flow (ULOF) accidents, the main advantage of the as specified innovative core design is that it is hardly possible to approach or exceed prompt criticality in the initiating phase of the transient. At the end of most of the calculated event sequences, core configurations that needed transition phase analyses were established.

Release of thermal and/or mechanical energy cannot be predicted without performing appropriate analyses taking representative results of the initiation phase as initial conditions.

FIG. 1. BN-800 core fuel assembly with MOX-fuel (dimensions in mm, from the top:

boron screen (50), sodium plenum (300), core (880), blanket (350), gas (670) [11].

Conclusions drawn from this comparative exercise hold for the specified case set-up. They need to be reviewed when some of the design features are changed or when more detailed evaluations lead to different input data like:

— Magnitude and/or spatial distributions of reactivity feedback coefficients of core materials;

— Reactivity feedback effects due to radial core expansion;

— Fuel pin mechanical properties; and

— If more rapidly developing consequences of control rod drive line expansion could be demonstrated.

It is recognized that there are possibilities for improvement of the analyses and/or for optimization, especially when a more realistic core design would be considered. However, the comparative exercise has shown as well that consequences of these type of modifications need to be analyzed carefully and in detail on a case to case basis. The use of more sophisticated and experimentally validated theoretical models would be helpful to improve the reliability of results. Evaluation of the impact of the as specified core design features on the core behaviour during operational transients was not part of this exercise as well as stability analyses. This would have needed other theoretical approaches to evaluate the potentially involved problems.

1.2.1.1. Methods and codes for transient analysis

Results of this comparative exercise have shown as well that theoretical approaches chosen by India with their PREDIS code package provides comprehensive results for single phase analyses but they use simplified approaches for two-phase flow. Fuel pin mechanics is not yet modelled in transient.

The Russian GRIF-SM code package with the complementary CANDLE-code package provides results for ULOF-type transients up to molten clad relocation. However, it is strongly recommended to couple a transient fuel pin mechanics code package to the system, to develop fuel pin failure criteria considering special features of the BN-800 fuel pin design and to extend the capabilities of the code system to describe core material relocation phenomena after fuel pin failure or break-up. The different code versions of the SAS4A-code family available in Japan, France and Germany allow evaluation consequences of accident initiators leading to core destruction along all stages of the initiation phase up to hexcan melting on the basis of experimentally qualified models. Even these code systems undergo continuous improvement.

In France, the pre-failure in-pin fuel relocation model EJECT is approaching completion with qualification and in Japan coupling to space time kinetics methods is far advanced. Thus, more improved analysis tools will become available in future which provide the possibility of re-evaluating the current results and to follow continuously the impact of new and/or optimized design features of innovative core designs on the results of accident analyses to be considered in the beyond design basis accident domain [12]. The IAEA/EC program on evaluation of benchmark calculations on a fast reactor core with near zero sodium void effect was one very good example of international cooperation. The participation included Germany, India, Italy, Japan 1 (PNC, Mitsubishi, Hitachi, Toshiba), Japan 2 (Osaka University), Russian Federation, U.K., and the USA.

1.2.2. Intercomparison of LMFR seismic analysis codes: comparison of experimental