ISO/TC 85/SC 6/WG 1 - Power reactor analyses and measurements

This document applies to the reactor physics tests that are performed following a refuelling or other core alteration of a PWR for which nuclear design calculations are required. This document does not address the physics test program for the initial core of a commercial PWR. This document specifies the minimum acceptable startup reactor physics test program to determine if the operating characteristics of the core are consistent with the design predictions, which provides assurance that the core can be operated as designed. This document does not address surveillance of reactor physics parameters during operation or other required tests such as mechanical tests of system components (for example, the rod drop time test), visual verification requirements for fuel assembly loading, or the calibration of instrumentation or control systems (even though these tests are an integral part of an overall program to ensure that the core behaves as designed).

This document provides guidance in the preparation, verification, and validation of group-averaged neutron and gamma-ray cross sections for the energy range and materials of importance in radiation protection and shielding calculations for nuclear reactors[1], see also Annex A. [1] This edition is based on ANSI/ANS-6.1.2-2013[1].

This document provides the basis for calculating the decay heat power of non-recycled nuclear fuel of light water reactors. For this purpose the following components are considered: — the contribution of the fission products from nuclear fission; — the contribution of the actinides; — the contribution of isotopes resulting from neutron capture in fission products. This document applies to light water reactors (pressurized water and boiling water reactors) loaded with a nuclear fuel mixture consisting of 235U and 238U. Application of the fission product contribution to decay heat developed using this document to other thermal reactor designs, including heavy water reactors, is permissible provided that the other contributions from actinides and neutron capture are determined for the specific reactor type. Its application to recycled nuclear fuel, like mixed-oxide or reprocessed uranium, is not permissible. The calculation procedures apply to decay heat periods from 0 s to 109 s.

This document specifies an analytical method for determining heavy water isotopic purity by Fourier transform infrared spectroscopy (FTIR). It is applicable to the determination of the whole range of heavy water concentration. The method is devoted to process controls at the different steps of the process systems in heavy water reactor power plant or any other related areas. The method can be applied for heavy water isotopic purity measurements in a heavy water reactor power plant or research reactor, heavy water production factory and heavy water related areas.

ISO 18075:2018 provides guidance for performing and validating the sequence of steady-state calculations leading to prediction, in all types of operating UO2-fuel commercial nuclear reactors, of: - reaction-rate spatial distributions; - reactivity; - change of nuclide compositions with time. ISO 18075:2018 provides: a) guidance for the selection of computational methods; b) criteria for verification and validation of calculation methods used by reactor core analysts; c) criteria for evaluation of accuracy and range of applicability of data and methods; d) requirements for documentation of the preceding.

ISO 19226:2017 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, between the ends of active fuel assemblies, given the neutron source in the core. 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). ISO 19226:2017 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.

ISO 18077:2018 applies to the reactor physics tests that are performed following a refuelling or other core alteration of a PWR for which nuclear design calculations are required. This document does not address the physics test program for the initial core of a commercial PWR[1]. ISO 18077:2018 specifies the minimum acceptable startup reactor physics test program to determine if the operating characteristics of the core are consistent with the design predictions, which provides assurance that the core can be operated as designed. This document does not address surveillance of reactor physics parameters during operation or other required tests such as mechanical tests of system components (for example the rod drop time test), visual verification requirements for fuel assembly loading, or the calibration of instrumentation or control systems (even though these tests are an integral part of an overall program to ensure that the core behaves as designed). ISO 18077:2018 assumes that the same previously accepted analytical methods are used for both the design of the reactor core and the startup test predictions. It also assumes that the expected operation of the core will fall within the historical database established for the plant and/or sister plants. When major changes are made in the core design, the test program should be reviewed to determine if more extensive testing is needed. Typical changes that might fall in this category include the initial use of novel fuel cycle designs, significant changes in fuel enrichments, fuel assembly design changes, burnable absorber design changes, and cores resulting from unplanned short cycles. Changes such as these may lead to operation in regions outside of the plant's experience database and therefore may necessitate expanding the test program. [1] The good practices discussed in this document should be considered for use in the physics test program for the initial core of a commercial PWR. One test that provides useful information (without additional test time) is the hot-zero-power to hot-full-power reactivity measurement.