Development of scaling factors for the activated concrete of the KRR-2

Abstract

The biological shielding concrete of KRR-2 was activated by a thermal neutron reaction during the operation of the reactor, thus a variety of radionuclides were generated in the concrete. In order to verify the radioactivity for the final disposal of waste and to achieve a more efficient cutting of the concrete, the radioactivity inventories and distributions of the activated concrete were evaluated. The activity of gamma-emitting radionuclides was measured by using an HPGe detector. The beta-emitting radionuclides were measured by an oxidation/combustion method for ³H and ¹⁴C and a combined method of an extraction chromatography and a liquid scintillation for ⁵⁵Fe and ⁶³Ni.

The dominant radioactive nuclides in the activated concrete were ³H, ¹⁴C, ⁵⁵Fe and ⁶⁰Co, and the maximum gamma activity was 105 Bq/g at the surface around the thermal column. The specific activities of all the nuclides were found to decrease almost linearly on a logarithmic scale along the depth from the inner surface of the concrete. Equations for scaling factors were obtained by a linear regression of logarithms from the radioactivity data of ³H/⁶⁰Co, ¹⁴C/⁶⁰Co and ⁵⁵Fe/⁶⁰Co nuclide pairs of the activated concrete. The scaling factors can be utilized for the estimation of beta radioactivity without the time consuming separation processes of the nuclides.

Introduction

In 1996, it was concluded that Korean Research Reactor-1 (KRR-1) and KRR-2 would be shut down and dismantled. A decommissioning project was launched in January 1997 with the goal of completion by 2008. The decommissioning project included the dismantlement of all the facilities and the removal of all the radioactive materials from the reactor site. At the end of 2006, the dismantlement of KRR-2 was completed and all the radioactive waste from the KRR-2 was packed into 200 L drums and 4 m³ containers which are stored in the KRR-2 reactor hall (Lee et al., 2006). One of the most important objects in the dismantlement was the biological shielding concrete of the reactor because its lower and inner parts were radioactive by neutron activation during the reactor operation. The radionuclide distributions of the activated part were evaluated to minimize the work amounts in cutting the shielding concrete by an optimal design of cutting lines. And its inventories were also evaluated to reduce the volume of waste by an exact classification of wastes and to verify the lower radioactivity than the clearance criteria for the final disposal of waste.

The biological shielding concrete was activated by thermal and intermediate-energy neutron reactions during the operation of the reactor, thus a variety of radionuclides could be generated in the activated concrete, including ³H, ¹⁴C, ⁵⁵Fe, ⁶⁰Co, ⁶³Ni, ¹³⁴Cs, ¹³⁷Cs, ¹⁵²Eu and ¹⁵⁴Eu. The concentration of the gamma emitting nuclides can be easily determined by a gamma spectrometry. However, for the determination of the relevant beta emitting nuclides, a complete separation of each radionuclide is required due to a poor energy resolution of beta spectrometry and therefore much time and cost were required. The scaling factor method was based on developing a correlation between easily measurable gamma emitting nuclides and difficult-to-measure (DTM) nuclides. A scaling factor method was applied for verification of the radioactivity inventories of the beta emitting nuclides in the concrete of the KRR-2.

In the course of activation of the concrete around the reactor core, most of ¹⁴C comes from neutron activation reactions of ¹⁴N (n,p) ¹⁴C and ¹⁷O (n,α) ¹⁴C, and the later reaction has a slightly higher contribution due to the higher concentration of oxygen in the concrete. The sources of ³H in the concrete are the neutron activation reactions of ⁶Li(n,α)³H , ²H(n,γ)³H and ³He(n,p)³H. The main contribution to ³H amount is the neutron activation reaction of ⁶Li, while only a very small amount is probably generated from ²H and ³He because ⁶Li has a high neutron activation cross section. In concrete, ¹⁴C exists in the form of carbonates or elementary carbon, while ³H exists as HT or HTO in the pores of the concrete. ³H and ¹⁴C are pure beta emitters and the energies of their beta particles are relatively low (E_β max ³H=18.6 keV, E_β max ¹⁴C=156.5 keV), thus the concrete should be completely decomposed to separate ³H and ¹⁴C from the concrete matrix. Their radioactivity in the concrete was measured from the separated solutions by using a liquid scintillation counter (LSC), due to its low energy (Xiaolin, 2005; Krasznai, 1993).

⁵⁵Fe is produced by neutron activation reactions (⁵⁴Fe(n,γ)⁵⁵Fe and ⁵⁶Fe(n,2n)⁵⁵Fe) of two major stable iron isotopes. A low energy gamma and an X-ray detector can be used for measuring the level of ⁵⁵Fe, but their counting efficiencies are usually very low. The most common and sensitive technique is LSC. Due to very low energy of ⁵⁵Fe, iron has to be completely separated from the samples (radionuclides) before counting. ⁶³Ni is produced by two neutron reactions of Ni and Cu, that is, ⁶²Ni(n,γ)⁶³Ni and ⁶³Cu(n,p)⁶³Ni. ⁶³Ni exists mainly in steel materials while only a tracer amount is contained in other reactor materials such as graphite, concrete, lead, and Al alloy. It is also a pure beta emitting radionuclide with the maximum beta energy of 66.95 keV and a half-life of 100 years. For its measurement, a LSC method is used instead of a windowless gas flow proportional counter or an implanted silicon detector which have low counting efficiencies. A chemical separation procedure for nickel is also required before counting (Hou et al., 2005; Warwick and Croudace, 2006).