Abstract. This paper presents the results of a round robin exercise carried out to compare specific activity measurements performed by eight European organisations on a set of ten neutron activation detectors containing the radio-nuclides 110m Ag, 60 Co, 54 Mn, 46 Sc and 94 Nb. The purpose of the exercise was to demonstrate the level of consistency between the participating organisations in blind tests of measurements relevant to reactor metrology. The samples used were selected from a stock of pre-existing irradiated material held at SCK•CEN. Taking turns over a period of approximately 9 months, the participating organisations received the samples, measured them and provided their results to an independent referee who collated and compared the data. The inter-comparison has demonstrated good agreement between the participants with standard deviations for each dosimeter varying between 1.6% and 3.1%. The paper provides results of the EWGRD Round Robin in an anonymised form together with discussion and conclusions which may be drawn from the exercise.

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The core restraints of advanced gas-cooled reactors are important structural components that are required to maintain the geometric integrity of the cores. A review of neutron dosimetry for the sister stations Hunterston B and Hinkley Point B identified that earlier conservative assessments predicted high thermal neutron dose rates to key components of the restraint structure (the restraint rod welds), with the implication that some of them may be predicted to fail during a seismic event. A revised assessment was therefore undertaken [Thornton, D. A., Allen, D. A., Tyrrell, R. J., Meese, T. C., Huggon, A.P., Whiley, G. S., and Mossop, J. R., “A Dosimetry Assessment for the Core Restraint of an Advanced Gas Cooled Reactor,” Proceedings of the 13th International Symposium on Reactor Dosimetry (ISRD-13, May 2008), World Scientific, River Edge, NJ, 2009, W. Voorbraak, L. Debarberis, and P. D’hondt, Eds., pp. 679–687] using a detailed 3D model and a Monte Carlo radiation transport program, mcbend. This reassessment resulted in more realistic fast and thermal neutron dose recommendations, the latter in particular being much lower than had been thought previously. It is now desirable to improve confidence in these predictions by providing direct validation of the mcbend model through the use of neutron flux measurements. This paper describes the programme of work being undertaken to deploy two neutron flux measurement “stringers” within the side-core region of one of the Hunterston B reactors for the purpose of validating the mcbend model. The design of the stringers and the determination of the preferred deployment locations have been informed by the use of detailed mcbend flux calculations. These computational studies represent a rare opportunity to design a flux measurement beforehand, with the clear intention of minimising the anticipated uncertainties and obtaining measurements that are known to be representative of the neutron fields to which the vulnerable steel restraint components are exposed.

Reactor pressure vessel (RPV) dose assessments for UK PWR plant have been performed using the AEA Technology Monte Carlo radiation transport code MCBEND. Monte Carlo models of the plant are validated against fast neutron activation measurements performed on flux monitor wires from surveillance capsules removed at the end of each cycle. These monitors have a range of time and spectrum dependencies. A computationally efficient method has been developed that calculates dosimeter activation values in a single calculation per reaction. This utilizes MCBEND's advanced acceleration techniques as well as an innovative treatment of neutron source data. The method employs plant validated, three-dimensionsal (3D) time-dependent neutron source data obtained from the UK reactor physics code PANTHER. Mean ratios of calculation (C) to experiment (E) obtained for the surveillance capsules removed to date are 1.02, 0.99 and 1.04. RPV dose distributions in axial, azimuthal and radial detail are obtained with a single calculation per reactor octant. Normalization to mean C/E provides best estimate dose data for the RPV inner surface and quarter thickness. A similar calculation provides dose data for the RPV top weld.