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The safety cases for the UKs Advanced Gas-Cooled Reactors require a robust understanding of the structural integrity of their constituent graphite bricks. Both neutron damage and graphite weight loss have significant effects upon material properties. Graphite weight loss occurs via radiolytic oxidation, initiated by neutron and gamma-ray dose. This includes contributions from fast neutrons, gamma-rays from the fuel, and secondary gamma-rays. An accurate knowledge of the dosimetry history of trepanned graphite samples is important, as well as the ability to make forward predictions for any brick within the core. It is therefore necessary to calculate both these dosimetry quantities with a high degree of spatial accuracy, taking into account the time-varying nature of local fuel channel powers as well as the way in which weight loss itself affects the dosimetry data. The Monte Carlo method is the most accurate available for the calculation of radiation dosimetry data. However, it is impractical to undertake full scale Monte Carlo calculations for a whole reactor core and to represent every required period of time and graphite weight loss, which will also vary in time and space. An alternative approach has been developed by Jacobs which exploits the high accuracy afforded by the Monte Carlo method, but is also flexible and practical enough to provide 3D dosimetry data for any graphite brick within the active cores, at any point in time and at any required graphite weight loss. The Power Partitioning Method uses MCBEND-generated dosimetry “Contribution Matrices”, coupled with PANTHER “core follow” fuel powers, to calculate dosimetry data anywhere within selected fuel channels or interstitial bricks.

Following an initial Round Robin inter-comparison of gamma spectrometry measurements reported in 2014, this paper presents results from the first part of a second Round Robin inter-comparison commissioned by the European Working Group on Reactor Dosimetry in 2018. Measurements were performed by thirteen European organisations on a set of irradiated neutron activation detectors representative of those commonly used by the Reactor Dosimetry community to measure neutron fluence using gamma spectrometry methods. The radionuclides measured were 110mAg, 58Co, 60Co, 54Mn, and 46Sc. The purpose of the exercise was to demonstrate the level of consistency between participating organisations in blind tests of measurements. The samples used were disks of iron, nickel, titanium, and two standard alloys of aluminium, one of 1% cobalt and the other 1% silver. They were irradiated in the MARIA reactor operated by National Centre for Nuclear Research, Poland. Participants provided their results to an independent referee who collated and compared the data. The results are presented in an anonymised form together with discussion and conclusions which may be drawn from the exercise. Good agreement was obtained with standard deviations for individual measurements between 2.0% (54Mn) and 5.6% (60Co). Overall, the results obtained from the latter Round Robin show less consistency than those from the first Round Robin, which is attributed to the participation of a wider pool of organisations.

Following on from an initial Round Robin inter-comparison of gamma spectrometry measurements reported in 2014, this paper presents the results of the second part of a further Round Robin inter-comparison commissioned in 2018 by the European Working Group on Reactor Dosimetry. The purpose of the exercise was to demonstrate the level of consistency between different organisations’ measurements of the 93Nb(n,n’)93mNb reaction, which plays a key role in underwriting reactor dosimetry assessments. To achieve this, measurements of 93mNb activity were performed by twelve European organisations on six sets of near identical niobium samples, each having its own geometry and tantalum concentration. The samples were provided by CEA, France and irradiated in the MARIA reactor, operated by National Centre for Nuclear Research, Poland. Participants provided their results to an independent referee who collated and compared the data. The inter-comparison has demonstrated agreement to within standard deviations ranging from ±2.2% to ±7.9%, with a tendency for some organisations to measure elevated values. The results of the inter-comparison are presented in an anonymised form together with discussion and conclusions which may be drawn from the exercise.

This paper presents a summary of the methods used in the first ∼40 years of AGR neutron dosimetry and nuclear heating calculations, and the influence of the earlier Magnox reactor dosimetry programme. While the current state-of-the-art Monte Carlo methods are extremely powerful they still require very careful consideration of the quality of the input data, nuclear data validation and variance reduction techniques; in particular, this paper examines the difficulties in assuring the adequate convergence of calculations when Monte Carlo acceleration is applied in the presence of significant streaming paths through attenuating or scattering media.