This paper aims to investigate the main modeling aspects used to reproduce as accurately as possible the measurements of gamma dose rates of a spent fuel assembly performed on an irradiated UO2 fuel assembly by thermoluminescence dosimeters during the course of nondestructive assay tests in 1981. This activity, that follows up that carried out in the context of the Benchmark on dose rate calculations proposed by the Advanced Fuel Cycle Scenarios Expert Group of the OECD Nuclear Energy Agency, was developed in order to determine the real self-protection properties of fuel assemblies against the risk of diversion and theft of nuclear material for illegal use. Calculations were divided into three phases. The first two – depletion, decay and gamma source calculation – were performed with ORIGEN-S, whereas for the last one – gamma transport and gamma dose rate calculations – MCNP6 was used. Even if the ANSI/ANS-6.1.1–1977 flux-to-dose-rate conversion factor provides generally better results than the ANSI/ANS-6.1.1–1991 conversion factor, the relative differences between calculated and measured dose rates never exceed −54% thereby confirming the validity of the strategy adopted in determining the effective self-protection of irradiated FAs.
Gamma dose rates from a spent UO2 fuel Assembly: Calculations vs measurements
Pergreffi R.;Rocchi F.;Guglielmelli A.;Ferrari P.
2022-01-01
Abstract
This paper aims to investigate the main modeling aspects used to reproduce as accurately as possible the measurements of gamma dose rates of a spent fuel assembly performed on an irradiated UO2 fuel assembly by thermoluminescence dosimeters during the course of nondestructive assay tests in 1981. This activity, that follows up that carried out in the context of the Benchmark on dose rate calculations proposed by the Advanced Fuel Cycle Scenarios Expert Group of the OECD Nuclear Energy Agency, was developed in order to determine the real self-protection properties of fuel assemblies against the risk of diversion and theft of nuclear material for illegal use. Calculations were divided into three phases. The first two – depletion, decay and gamma source calculation – were performed with ORIGEN-S, whereas for the last one – gamma transport and gamma dose rate calculations – MCNP6 was used. Even if the ANSI/ANS-6.1.1–1977 flux-to-dose-rate conversion factor provides generally better results than the ANSI/ANS-6.1.1–1991 conversion factor, the relative differences between calculated and measured dose rates never exceed −54% thereby confirming the validity of the strategy adopted in determining the effective self-protection of irradiated FAs.I documenti in IRIS sono protetti da copyright e tutti i diritti sono riservati, salvo diversa indicazione.
