During the treatment campaigns, carried out in ITREC (Treatment and Remaking Fuel Elements Plant), near the Trisaia locality, Italy, some liquid wastes of low and higher activity have been produced and stored in the plant site. A chemical process based on the cementation of liquid wastes by addition to dry cement in a drum was studied and applied in order 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/l were cemented and stored (between April 1999 and November 2000 ) in 337 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. Other 433 drums were produced (between May 1995 and November 1997) using liquid wastes at lower Cs-137 specific activity. Their total activity (per drum) is approximately 30 GBq, of which 15 GBq are due to Cs-137. Exposure rate measurements were made on some conditioning products (three positions at the external surface and one meter distant on the equatorial plane), for working staff radioprotection purposes. All the drums are ready to be stored in the future Storage Site. 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 present paper present the approach and the results obtained to estimate the contact dose and the dose 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 activation code package has been used to obtain the decay gamma source related to each one of the analysed drum containers. Therefore the calculated dose rates were obtained 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-4C code, and c) on the 2-D Sn deterministic approach, using the DORT code. The approach has been applied to each one of the drum for which dose rate measurement are available. 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

CAMBI, GILIO;
2005

Abstract

During the treatment campaigns, carried out in ITREC (Treatment and Remaking Fuel Elements Plant), near the Trisaia locality, Italy, some liquid wastes of low and higher activity have been produced and stored in the plant site. A chemical process based on the cementation of liquid wastes by addition to dry cement in a drum was studied and applied in order 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/l were cemented and stored (between April 1999 and November 2000 ) in 337 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. Other 433 drums were produced (between May 1995 and November 1997) using liquid wastes at lower Cs-137 specific activity. Their total activity (per drum) is approximately 30 GBq, of which 15 GBq are due to Cs-137. Exposure rate measurements were made on some conditioning products (three positions at the external surface and one meter distant on the equatorial plane), for working staff radioprotection purposes. All the drums are ready to be stored in the future Storage Site. 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 present paper present the approach and the results obtained to estimate the contact dose and the dose 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 activation code package has been used to obtain the decay gamma source related to each one of the analysed drum containers. Therefore the calculated dose rates were obtained 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-4C code, and c) on the 2-D Sn deterministic approach, using the DORT code. The approach has been applied to each one of the drum for which dose rate measurement are available. 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.
Proceedings of the ICEM ‘05: The 10th International Conference on Environmental Remediation and Radioactive Waste Management
G. Cambi; D.G. Cepraga; M. Frisoni; L. Di Pace
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Utilizza questo identificativo per citare o creare un link a questo documento: https://hdl.handle.net/11585/6036
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