Removal of C-14 from irradiated graphite

SCK•CEN Mentor

Druyts Frank, fdruyts@sckcen.be, +32 (0)14 33 31 37

Expert group

R&D Waste Packages

SCK•CEN Co-mentor

Malambu Mbala Edouard, emalambu@sckcen.be, +32 (0)14 33 22 83

Introduction

The management of irradiated graphite (i-graphite) is a worldwide concern. On a global scale, more than 250000 tons of i-graphite has now accumulated. At the same time, progress toward ultimate disposal solutions remains slow, leaving out a vast domain of research work. In Belgium, the main source of i-graphite is the BR1 reactor at SCK•CEN, which is an air-cooled, graphite-moderated reactor with aluminium clad natural uranium as fuel. The moderator consists of 14000 individual blocks of graphite, leading to a total weight of 492 ton. At the current regime, the BR1 operation is destined to continue for a few more decades, but ultimately the reactor will be decommissioned and the graphite blocks removed for treatment and disposal as radioactive waste. When considering irradiated graphite (or i‑graphite) as waste, from a safety assessment point of view, the main concern are the mobile isotopes (in the geo- and biosphere), notably 3H, 14C, and 36Cl, with the latter two isotopes being long-lived. The fate of 14C is currently studied in the EC project CAST (CArbon-14 Source Term), which focuses on the 14C release and speciation from, among other materials, graphite. From recent studies, several potential decontamination methods, aimed at removing 14C from i-graphite, emerged, including thermal oxidation and treatment with molten salts. In particular, thermal oxidation, applied on a laboratory scale to British and German i-graphite, seems a promising technique. The effect of thermal oxidation is based on its removal of a thin layer from the graphite's surface, which is reported, at least for British and German i-graphite, to contain an important fraction of the 14C inventory. The purpose of the current Master thesis proposal is to investigate whether a 14C-enriched surface layer is also present on the BR1 i-graphite. This finding can form the basis for further research on the decontamination of BR1 graphite. The ultimate goal of the study is to find a method to decontaminate the i-graphite in such a way that geological disposal is not needed.

Objective

The thesis proposal consists of the following actions:

  • Review of decontamination methods, including thermal oxidation and chemical decontamination: what are the current/future decontamination methods for i-graphite, specifically aimed at the removal of critical nuclides? The review should also look at incineration as an alternative treatment method.

  • Investigation of the production pathways of 14C. There are three main routes for the production of 14C: through the neutron activation of 13C, 14N and 17O. These isotopes each have different neutron adsorption cross section. Although 13C is much more abundant than 14N in graphite, the neutron adsorption for 14N is much higher, meaning that at relatively low concentrations 14N can become the main precursor of 14C. The student will calculate at what 14N concentration the nitrogen pathway becomes prevalent over the 13C pathway, using the Monte Carlo burn-up code ALEPH.

Calculation of the 14C content of irradiated graphite as a function of depth, using the ALEPH code, in order to predict whether a surface 14C-enriched layer is present . The basis of this calculation will be a depth profile of 14N, obtained by X-ray Photo-electron Spectroscopy (XPS).

The minimum diploma level of the candidate needs to be

Academic bachelor

The candidate needs to have a background in

Chemistry