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.

This paper describes the validation of fast neutron transport calculations for Magnox RPVs operated by Nuclear Electric. The work compares calculated data, derived using the Monte Carlo code McBEND, with extensive plant measurements. Agreement is typically within ∼20%. This level of agreement has been achieved using individual models for the top, side and bottom of each reactor in which almost explicit representation of all the major components was required. The modelling detail included the geometric effects of temperature and irradiation on reactor components as well as source data derived from reactor measurements and operating histories. Such effort is justified because the reduced uncertainty in the pressure vessel dose can be translated into increased operating margins and extended plant lifetimes.

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This paper describes the calculation of the attenuation of neutron dose through the steel pressure vessels of Magnox power plant operated by BNFL, Magnox Generation Division. These data are required for the assessment of the structural integrity of the plant. Detailed three-dimensional neutron transport models of the reactors, pressure vessels and surrounding bioshields have been developed using the AEA Technology Monte-Carlo code MCBEND. These have been used to calculate neutron fluxes through the pressure vessel in both sub-core and side-core regions of pressure vessels, the two bounding extremes of the incident neutron spectra. The effect of postulated cracks in the RPV is shown not to influence the predicted neutron doses. A comparison with neutron fluxes measured from through wall samples removed from a decommissioned reactor show the calculation route to be accurate.