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This document provides a procedure for the evaluation of irradiation data in the region between the reactor core and the inside surface of the containment vessel, through the pressure vessel and the reactor cavity. NOTE These irradiation data could be neutron fluence or displacements per atom (dpa), and Helium production. The evaluation employs both neutron flux computations and measurement data from in-vessel and cavity dosimetry, as appropriate. This document applies to pressurized water reactors (PWRs), boiling water reactors (BWRs), and pressurized heavy water reactors (PHWRs). This document also provides a procedure for evaluating neutron damage properties at the reactor pressure vessel and internal components of PWRs, BWRs, and PHWRs. Damage properties are focused on atomic displacement damage caused by direct displacements of atoms due to collisions with neutrons and indirect damage caused by gas production, both of which are strongly dependent on the neutron energy spectrum. Therefore, for a given neutron fluence and neutron energy spectrum, calculations of the total accumulated number of atomic displacements are important data to be used for reactor life management.

This document is applicable to fuel fabrication. It gives guidelines on the determination of the specific surface area of as-fabricated uranium dioxide and plutonium dioxide powders by volumetric or gravimetric determination of the amount of nitrogen adsorbed on the powder. The measurement of other uranium oxide powders refers to uranium dioxide, such as UO3 and U3O8. The measurement of MOX(UO2-PuO2) powders refers to plutonium dioxide. When conditions described are fulfilled, modifications using other adsorbing gases are included.

The method is relevant as long as the expected value is in the range from 1 m2/g to 10 m2/g for uranium dioxide powders, in the range from 0.1 m2/g to 45 m2/g for plutonium dioxide powders.

This International Standard establishes an evaluation methodology for nuclear criticality safety with burnup credit. It identifies important parameters and specifies requirements, recommendations, and precautions to be taken into account in the evaluations. It also highlights the main important technical fields to ensure that the fuel composition or history considered in calculations provides a bounding value of the effective neutron multiplication factor, keff. A more practical approach is also presented.

This International Standard is applicable to transport, storage, disposal or reprocessing units implying irradiated fissile material from pressurized water reactor (PWR) fuels that initially contain enriched uranium oxide (UOX). Uranium could originate either from natural uranium or recycled uranium.

Fuels irradiated in other reactors (e.g. boiling water reactors) and fuels that initially contain mixed uranium-plutonium oxide are not covered in this International Standard.

This International Standard does not specify requirements related to overall criticality safety evaluation or eventual implementation of burnup credit.