ISO 27468:2011 臨界安全性— PWR UOX燃料を含むシステムの評価—限界燃焼クレジットアプローチ | ページ 6

3.1

actinide

element with atomic number in the range from 90 to 103

Note 1 to entry: Many actinides are produced during the irradiation due to neutron capture on other actinides and/or decay of other actinides and/or by (n,2n) reactions, etc. The corresponding nuclides are all neutron producers and some are net (considering neutron production and absorption) neutron producers in a slow neutron energy spectrum.

3.2

axial burnup profile

real or modelled axial distribution of the burnup in the fuel assembly

Note 1 to entry: The axial distribution of the burnup is caused by axial neutron leakage, axial variations in the fuel enrichment, moderator temperature rise through the core, non-full length burnable poison and partial insertion of control rods.

3.3

burnable poison

nuclide neutron absorber added to the fuel assembly to control reactor reactivity and power distribution

Note 1 to entry: As the reactor operation progresses, the amount of neutron absorbing material is depleted, or ‘‘burned’’. Then, if the presence of burnable poisons (fixed or removable) is considered in a criticality safety evaluation, the most reactive condition may not be for the fresh fuel.

Note 2 to entry: See also ISO 921:1997, entry 135.

3.4

burnup

average energy released by a defined region of the fuel during its irradiation

Note 1 to entry: This region could be a complete fuel assembly or some part of the assembly. Burnup is commonly expressed as energy released per mass of Initial fissionable actinides (uranium only for this International Standard). Units commonly used are expressed in megawatt day per metric tonne of initial uranium (MWd/t) or gigawatt day per metric tonne of initial uranium (GWd/t).

Note 2 to entry: See also ISO 921:1997, entry 1156.

3.5

burnup credit

margin of reduced k_eff for an evaluated system, due to the irradiation of fuel in a reactor, as determined with the use of a structured evaluation process

3.6

cooling time

time following the final irradiation of the fuel in a reactor

Note 1 to entry: During this period, the radioactive decay results in changes in the fuel nuclide concentrations.

3.7

depletion calculation

calculation performed to determine the concentrations of individual nuclides in the fuel at the end of irradiation in a reactor; that is a cooling time equal to zero

Note 1 to entry: Other fuel properties can usually be determined by depletion calculations (e.g. flux-weighted macroscopic cross-sections or lattice cell k_∞).

Note 2 to entry: Radioactive decay between reactor irradiation periods and after final shutdown is usually included in the same calculation procedure.

3.8

end effect

impact on k_eff of the less irradiated parts of the fuel assembly (upper and lower ends of the assembly)

3.9

fission product

nuclide produced from nuclear fission

Note 1 to entry: During this reaction two or more fission products are produced together with neutrons and radiations (gamma, etc.). The fission products can be a direct result of the fissions or can be created after the decay of (or neutron absorption with) other fission products. Often only a selection of fission products is accounted for as neutron absorbers in burnup credit, but consideration of all fission products absorption is required to simulate fuel irradiation during reactor operation.

Note 2 to entry: See also ISO 921:1997, entry 478.

3.10

loosely coupled system

system in which two or more regions with high “local” values of k_eff are separated by regions with low k_eff importance

Note 1 to entry: Convergence problems can occur when a Monte Carlo method is used for the k_eff calculation of such systems where neutron interaction between the highly fissile regions is weak.

3.11

validation

documented determination that the combination of models, methods and data as embodied in a computer code methodology is an appropriate representation of the process or system for which it is intended

Note 1 to entry: This documented determination is accomplished by comparing code results to benchmark experimental results to define code bias and areas of applicability of a calculation method.