Validation of Dose Calculation Codes for Clearance

Abstract

Various international and national bodies such as the International Atomic Energy Agency, the European Commission, the US Nuclear Regulatory Commission have put forward proposals or guidance documents to regulate the "clearance" from regulatory control of very low level radioactive material, in order to allow its recycling as a material management practice. All these proposals are based on predicted scenarios for subsequent utilisation of the released materials. The calculation models used in these scenarios tend to utilise conservative data regarding exposure times and dose uptake as well as other assumptions as a safeguard against uncertainties.None of these models has ever been validated by comparison with the actual real life practice of recycling. An international project was organised in order to validate some of the assumptions made in these calculation models, and, thereby, better assess the radiological consequences of recycling on a practical large scale.The validation was proposed to be carried out by comparing the results of dose measurements during a chain of recycling operations to dose values calculated for the same operations using the (US) RESRAD-RECYCLE and the (French) CERISE programmes. The operations were to cover all recycling activities, including the receipt of contaminated scrap at the radiologically controlled melting facility, its segmentation and melting, transport of released ingots to a manufacturing industry for use with other scrap as feed material and production of industrial products (rolls).The project was initiated by the Swedish Radiation Protection Institute and was a co-operation between authorities, research institutes and commercial companies from Sweden, France, USA and Belgium.A first phase of melting of contaminated scrap at Studsvik, release of ingots and transport to Åkers was carried out. The ingots were re-melted along with other (uncontaminated) scrap at Åkers to be used for manufacturing rolls. The doses to workers were measured at Studsvik, Åkers and during ingot transport. Dose calculations were made in parallel with these operations using the RESRAD-RECYCLE and CERISE programmes. However, the results of these calculations could not be compared with the corresponding values of doses taken by workers, because all of the doses were below the limit of detection.Due to this fact, a second phase was executed involving the segmenting and melting of a 3.4 t stainless steel fuel rack with an estimated activity concentration of over 150 Bq/g, mostly Co-60. The fuel rack was melted for volume reduction in the Studsvik facility in the middle of January 2001, in the presence of project team including the dose modellers, who then made code calculations to estimate the dose uptake of the workers.All personnel involved in the project operations were equipped with electronic (display) dosimeters. The measurements showed that segmenting was the work operation that gave the highest dose, almost 65 % of the total dose incurred, while melting itself accounted for only abou...