During the treatment campaigns, carried out in ITREC (Treatment and Remaking Fuel Elements Plant), near the Trisaia locality, Italy, liquid wastes of low and higher activity have been produced and stored in the plant site. A process based on the cementation of liquid wastes by addition to dry cement in a drum was applied to immobilize them in an encapsulated and shielded solid form. It was performed in the SIRTE-MOWA plant (Unit System to prepare, to transfer and to treat radioactive liquid waste in Mobile Waste). Liquid wastes having a Cs-137 specific activity of about 2.77 GBq/1 were cemented and stored in about 300 drums, each of them characterised by a mean total activity of around 2100 GBq, of which 555 GBq due to Cs-137. Each conditioning product is composed of a metal drum containing the radioactive solution (about 200 litres), solidified by approximately 400 kg of cement, and a shielding shell made of steel and lead. Dose rates external to the cemented drum containers were measured during and after the drum production process for working staff radioprotection purposes and as a support in the fulfilment of the safety rules and procedures that' have to be guaranteed for drums transfer and storage in their final disposal. The paper presents the approach and the results obtained to estimate the doses outside a shielded drum containing an homogeneously mixed radioactive material, by knowing only the geometry of the drum and its radio-isotopic content (in terms of activity per each isotope). The ANITA-2000-D activation code package has been used to obtain the decay gamma sources related to each one of the analysed drum containers. Calculations of the dose rates were performed using three different and independent schemes based: a) on the 1D-Sn deterministic approach SCALENEA-1 calculation sequence, b) on the Monte Carlo approach, with the MCNP-4C2 code, and c) on the 2-D Sn deterministic approach, using the DORT code. A cross check between the different calculation schemes and the experimental dose rate measures represent a significant step in the quality assurance process related to the management of radioactive wastes, that could have a positive impact on the public opinion.
Calculated-experiment cross-check for dose rates outside drums containing cemented liquid radioactive wastes
2005-07-01
Abstract
During the treatment campaigns, carried out in ITREC (Treatment and Remaking Fuel Elements Plant), near the Trisaia locality, Italy, liquid wastes of low and higher activity have been produced and stored in the plant site. A process based on the cementation of liquid wastes by addition to dry cement in a drum was applied to immobilize them in an encapsulated and shielded solid form. It was performed in the SIRTE-MOWA plant (Unit System to prepare, to transfer and to treat radioactive liquid waste in Mobile Waste). Liquid wastes having a Cs-137 specific activity of about 2.77 GBq/1 were cemented and stored in about 300 drums, each of them characterised by a mean total activity of around 2100 GBq, of which 555 GBq due to Cs-137. Each conditioning product is composed of a metal drum containing the radioactive solution (about 200 litres), solidified by approximately 400 kg of cement, and a shielding shell made of steel and lead. Dose rates external to the cemented drum containers were measured during and after the drum production process for working staff radioprotection purposes and as a support in the fulfilment of the safety rules and procedures that' have to be guaranteed for drums transfer and storage in their final disposal. The paper presents the approach and the results obtained to estimate the doses outside a shielded drum containing an homogeneously mixed radioactive material, by knowing only the geometry of the drum and its radio-isotopic content (in terms of activity per each isotope). The ANITA-2000-D activation code package has been used to obtain the decay gamma sources related to each one of the analysed drum containers. Calculations of the dose rates were performed using three different and independent schemes based: a) on the 1D-Sn deterministic approach SCALENEA-1 calculation sequence, b) on the Monte Carlo approach, with the MCNP-4C2 code, and c) on the 2-D Sn deterministic approach, using the DORT code. A cross check between the different calculation schemes and the experimental dose rate measures represent a significant step in the quality assurance process related to the management of radioactive wastes, that could have a positive impact on the public opinion.I documenti in IRIS sono protetti da copyright e tutti i diritti sono riservati, salvo diversa indicazione.
