Elsevier

Annals of Nuclear Energy

Volume 35, Issue 8, August 2008, Pages 1519-1534
Annals of Nuclear Energy

Impact of spread in minor actinide data from ENDF/B-VII.0, ENDF/B-VI.8, JENDL-3.3 and JEFF-3.0 on an IAEA-CRP FBR benchmark for MA incineration

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Abstract

Under an IAEA Coordinated Research Project (CRP), an FBR Model has been designed to study feasibility of incineration of long-lived minor actinides (MA). The predictions depend on the accuracy of the nuclear data used. There are several evaluations for actinide nuclear data, produced based on state-of-the-art procedures, but substantial deviations persist among them. The effect of spread in the MA nuclear data over a few recent evaluations, on the predicted material and Doppler worths of the FBR model, has been estimated and presented in this paper.

Introduction

With the nuclear energy production, as well as the public concern about its potential hazards, showing substantial growth around the world, management of the nuclear waste deserves maximum feasible attention. Long-term radio-toxicity of the spent-fuel is dominated primarily by the transuranic (TRU) nuclides including minor actinides (MA) (viz. isotopes of Np, Am, and Cm), and certain long-lived fission products (LLFP). Appendix A gives the paths of isotopic evolution of MA due to various decay modes, and (n, γ) reaction.

Though the option of long term storage of these wastes is strongly in consideration, recent attention has been focused on reducing the storage time considerably by means of (i) incineration, wherein a toxic actinide is fissioned into fission products (FP) with, generally, much shorter half-lives and (ii) transmutation, wherein a toxic nuclide is converted into another with shorter half-life. These are found feasible in the newly emerging systems such as the Accelerator Driven Subcritical Systems (ADSS). The TRUs may in principle be incinerated in fast reactors and hence are also under study. Presently a lot of studies are on in designing systems for actinide incineration and to calculate/predict the extent of incineration. In the once-through cycle, since the TRUs in the discharge are considered waste, their decay properties only are important. But, when their incineration is considered, they become part of the fuel fed, and their interaction behaviour with neutrons also becomes important. Thus such calculational efforts rely heavily on the nuclear data, such as decay data, and the cross-sections for reactions like neutron capture, fission, etc. When these data are available from several equally reliable sources, the spread in the prediction due to the spread in the nuclear data must be known. This is the motivation behind this paper.

IAEA has been piloting a Coordinated Research Project (CRP) on “Studies of Advanced Reactor Technology Options for Effective Incineration of Radioactive Waste” (Stanculescu et al., 2002). Its objective is towards demonstrating the waste incineration capacities of ADSS, fast reactors, etc. Participating in the CRP, the Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, has designed an FBR model, having features similar to its Prototype Fast Breeder Reactor (PFBR) (Mohanakrishnan, 2004), presently under construction, with modifications done as necessitated (a) for inclusion of 5% MA in the fresh fuel, and (b) for replacing UO2 radial blanket by ThO2. The initial composition of MA corresponds to that of the Indian Pressurised Heavy Water Reactors (PHWR) discharge fuel. This FBR model (Devan et al., 2004, Harish et al., 2005) shows a potential to incinerate nearly 10% of the MA fed, during one equilibrium cycle, with satisfactory safety parameters. This is the system under study in this paper, and is called hereinafter as ‘FBR-MA’ for ease of reference.

Taking FBR-MA as a benchmark, the effect of recent revisions in MA nuclear data are studied in the present work. Different nuclear data libraries, viz. ENDF/B-VII.0 (Chadwick, 2006), ENDF/B-VI.8 (Rose, 1991), JENDL-3.3 (Shibata et al., 1990) and JEFF-3.0 (Koning et al., 2004) are used for MA nuclear data, while a common multigroup set is used for all other materials involved, so that the observed spread in the computed reactor parameters could be ascribed to the spread in the MA data over these evaluations. To compute the neutronic parameters of FBR-MA, the codes, ALCIALMI (2-D diffusion theory code) (Byard et al., 1965, Narayanan and John, 2000), ALEX (code for calculating breeding ratio, power distribution, reaction rates, etc.) (Giacometti, 1969) and NEWPERT (perturbation code for predicting Doppler and material reactivity worths) (John, 1984) are used. For nuclide evolutions with respect to irradiation or cooling, the code ORIGEN2 (Croff, 1983) has been used.

Salient features of the different evaluations of the MA nuclides studied (viz. 237Np, 241Am, 242mAm, 243Am, 242Cm, 243Cm, and 244Cm) data are presented in Section 2. The multigroup data used for the present work is briefed in Section 3. The features of FBR-MA are given in Section 4. The results of the present study are discussed in Section 5.

Section snippets

MA data in different evaluations

Table 1 gives salient details of the MA evaluations. It may be noted that each evaluation is not necessarily independent, and there is partial or full adaptation from other libraries or previous versions. In Table 1, ‘new’ means a substantially new evaluation, and a reference to a different library indicates the data source for adaptation. The respective evaluation document or the general information records (MF1, MT451) of the particular evaluation may be seen for full details. JENDL-3.3 has

Multigroup cross-sections

A “modified 25-group Cadarache Version 2 cross-section set” (called CV2M) (Ravier and Chaumont, 1966, Barre et al., 1969, Devan et al., 1992) is used for the neutronics calculations. This is the set used for the core design of PFBR. In the present work, CV2M is considered as the base set, in the sense that all major materials, including fuel, structural, control and coolant, are taken from CV2M, while 25 group data for the MA are obtained from different evaluations. The basic MA data from an

Salient features of FBR-MA

The FBR-MA uses a mixed oxide fuel, and its total power is 1150 MW. Its inner and outer cores respectively have 85 and 102 subassemblies (SA) with enrichments 19.5% and 27.1%. The axial blankets are depleted UO2. There are 180 ThO2 radial blanket SA in three rows, and 72 SA of steel reflector in one row beyond the radial blanket. There are nine Control and Safety Rods (CSR) for the usual controls, and three Diverse Safety Rods (DSR) only to SCRAM the reactor. B4C pellets, 65% enriched in 10B,

k-eff, breeding ratio, and β-eff

The code ALCIALMI (2-D diffusion theory) is used for k-eff calculations, and ALEX for breeding ratio, power distribution, reaction rates, etc. Table 6 gives the k-eff, the loss of reactivity with burnup, and the effective delayed neutron fractions (β-eff), for the FBR-MA, corresponding to the four evaluations. The spread given in pcm stands for the difference between the maximum and the minimum predictions. In this sense, the predicted k-effs agree within a spread of 450 pcm. The β-eff shows

Conclusion

The effect of spread in the minor actinide (MA) nuclear data over a few recent evaluations, on the predicted material and Doppler worths of an FBR model, designed under an IAEA Coordinated Research Project, for studying feasibility of MA incineration, has been estimated. The nuclear data libraries, viz. ENDF/B-VII.0, ENDF/B-VI.8, JENDL-3.3 and JEFF-3.0 are used for the MA, while a common multigroup set is used for all other materials involved, so that the observed spread in the computed reactor

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