Elsevier

Nuclear Engineering and Design

Volume 262, September 2013, Pages 404-408
Nuclear Engineering and Design

Calculation of the neutron importance and weighted neutron generation time using MCNIC method in accelerator driven subcritical reactors

https://doi.org/10.1016/j.nucengdes.2013.04.037Get rights and content

Highlights

  • All reactor kinetic parameters are importance weighted quantities.

  • MCNIC method has been developed for calculating neutron importance in ADSRs.

  • Mean generation time has been calculated in spallation driven systems.

Abstract

The difference between non-weighted neutron generation time (Λ) and the weighted one (Λ) can be quite significant depending on the type of the system. In the present work, we will focus on developing MCNIC method for calculation of the neutron importance (Φ) and importance weighted neutron generation time (Λ) in accelerator driven systems (ADS). Two hypothetic bare and graphite reflected spallation source driven system have been considered as illustrative examples for this means. The results of this method have been compared with those obtained by MCNPX code. According to the results, the relative difference between Λ and Λ is within 36% and 24,840% in bare and reflected illustrative examples respectively. The difference is quite significant in reflected systems and increases with reflector thickness. In Conclusion, this method may be used for better estimation of kinetic parameters rather than the MCNPX code because of using neutron importance function.

Introduction

Accelerator driven subcritical reactors (ADSRs) are modern nuclear reactor types that an extensive research is conducted to make the reactor as an industrial technology. Furthermore, ADSRs are called hybrid reactors that make the possibility of producing energy and transmutation of radioactive wastes in a clean and safe manner.

Calculating accurate point kinetics parameters such as Λ plays a key role in analyzing dynamics of reactor behaviour especially in fast transients (Snoj et al., 2008, Eriksson et al., 2005). For modern nuclear reactor types such as ADSRs, Monte Carlo MCNP code (Briesmeister, 2000) is widely used for neutronic calculations. Despite its considerable advantages in solving problems with complex geometries, MCNP fails in predicting the accurate value of neutron generation time (i.e. importance-weighted) especially in highly reflected reactors (Feghhi et al., 2008). This drawback originates from the fact that all quantities simulated in this code are non-importance weighted.

The difference between the non-weighted neutron generation time and the importance-weighted neutron generation time as an important kinetic parameter can be quite significant depending on the type of the system. This difference is very substantial especially in highly reflected reactors (Verboomen et al., 2006).

Multi-group adjoint method can estimate kinetics parameters more realistic at the expense of complexities arise from multi-group particle transport consideration of the problem. The application of weighting function method in calculation of these parameters is recommended by many authors to obtain a good estimation of the problem (Eriksson et al., 2005, Becker, 1968, Henry, 1958).

As the physics of ADSRs is different from ordinary reactor types, some modifications are needed for weighting function method especially in systems that are highly related to external neutron source (Rineiski and Maschek, 2005, Keisuke, 2005).

Feghhi et al. (2007) has proposed a useful method, called Monte Carlo neutron importance calculation (MCNIC), in which the neutron importance is also considered. In addition, the applicability of MCNIC method had been extended for the calculation of Λ in fissionable assemblies of complex geometries (Feghhi et al., 2008).

In what follows we will attempt to focus on developing MCNIC method for calculation of Φ and Λ in spallation source-driven systems. The methodology has been applied to MCNPX for calculations in two hypothetic bare and graphite reflected spallation source driven systems. The results are compared with MCNPX code calculations (Hughes et al., 2002).

Section snippets

MCNIC method principal

Neutron importance is interpreted in different ways depending on the object of interest. If a number of neutrons are inserted into a subcritical system, the neutron population will increase depending on the location, energy and direction of initial neutrons. Injection of initial neutron having population of N0, in a subcritical reactor will change the neutron population to N0 + ΔN. In this subcritical system, the variation of neutron population per inserted neutron is called “neutron importance”

Prompt neutron energy spectrum in ADSRs

For calculation of the neutron importance function in ADSRs using MCNIC method, χp(E) in conventional critical reactors should be replaced by ξ(E) in an ADSR. As shown in Fig. 2, the prompt neutron energy spectrum in an ADSR, calculated using MCNPX code contains prompt neutron spectrum from high energy fission as well as spallation reactions in the hypothetic ADSR.

Each MCNPX run is done for 2,000,000 neutron histories that lead to an eigenvalue statistical error less than 130 pcm. As shown in

Conclusion

Neutron generation time as an important kinetic parameter, in all nuclear power systems has a key role in the analysis of fast transients. MCNPX code despite its considerable advantages in solving problems with complex geometries and physics, fail to predict the accurate value of Λ, especially in highly reflected reactors. This drawback originates from the fact that all quantities simulated in this code are nonimportance-weighted.

The difference between Λ and Λ, can be quite significant

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