TY - JOUR

T1 - Calculation of dose rates at the surface of storage containers for high-level radioactive waste

AU - Pönitz, Erik

AU - Walther, Clemens

N1 - Funding information: This work was financially supported by the German Federal Ministry of Education and Research (Bundesministerium für Bildung und Forschung, BMBF) under the research grant 15S9082A.

PY - 2017/5/2

Y1 - 2017/5/2

N2 - In several countries, the high-level radioactive waste that will be disposed of in deep geological formations has to be retrievable for a certain time. Since 2010, retrievability is required for the operation phase of a repository also in Germany. Depending on the effort and the feasibility of remote handling, a certain exposure of the involved employees to ionising radiation is caused. The estimation of the exposure requires the knowledge of the inventory of radionuclides and the radiation field around the storage containers. This paper focuses on German concepts for final storage casks for the drift emplacement in rock salt for both spent fuel rods and high-active waste (HAW) coquilles. Calculations of dose rates at the surfaces of the casks are presented. The calculations show that the gamma radiation of fission and activation products can be efficiently shielded by materials like cast iron or low-alloyed steel. Ductile cast iron, however, has a positive influence on the neutron moderation because of the high carbon content. The shielding of the neutron radiation strongly depends on the quantity and position of the polyethylene (PE) rods which are used as neutron moderator. PE has unfavourable features at elevated temperatures that can be reached in a repository for high-level waste. The simulations show that TiH2 is a promising alternative material for the shielding of neutron radiation. The current assumption that a final storage cask for HAW coquilles has the same outer dimensions as the cask for spent fuel rods will lead to unacceptable dose rates at the side wall. An increase of the diameter is necessary to provide sufficient shielding. Graphite insets in the cask interior lead to a considerably lowered dose rate.

AB - In several countries, the high-level radioactive waste that will be disposed of in deep geological formations has to be retrievable for a certain time. Since 2010, retrievability is required for the operation phase of a repository also in Germany. Depending on the effort and the feasibility of remote handling, a certain exposure of the involved employees to ionising radiation is caused. The estimation of the exposure requires the knowledge of the inventory of radionuclides and the radiation field around the storage containers. This paper focuses on German concepts for final storage casks for the drift emplacement in rock salt for both spent fuel rods and high-active waste (HAW) coquilles. Calculations of dose rates at the surfaces of the casks are presented. The calculations show that the gamma radiation of fission and activation products can be efficiently shielded by materials like cast iron or low-alloyed steel. Ductile cast iron, however, has a positive influence on the neutron moderation because of the high carbon content. The shielding of the neutron radiation strongly depends on the quantity and position of the polyethylene (PE) rods which are used as neutron moderator. PE has unfavourable features at elevated temperatures that can be reached in a repository for high-level waste. The simulations show that TiH2 is a promising alternative material for the shielding of neutron radiation. The current assumption that a final storage cask for HAW coquilles has the same outer dimensions as the cask for spent fuel rods will lead to unacceptable dose rates at the side wall. An increase of the diameter is necessary to provide sufficient shielding. Graphite insets in the cask interior lead to a considerably lowered dose rate.

UR - http://www.scopus.com/inward/record.url?scp=85040241372&partnerID=8YFLogxK

U2 - 10.1093/rpd/ncx054

DO - 10.1093/rpd/ncx054

M3 - Article

C2 - 28472493

AN - SCOPUS:85040241372

VL - 177

SP - 361

EP - 372

JO - Radiation Protection Dosimetry

JF - Radiation Protection Dosimetry

SN - 0144-8420

IS - 4

ER -