Method for calculating a PCI margin associated with a loading pattern of a nuclear reactor, associated system, computer program and medium

An electronic system for calculating a pellet-cladding interaction margin associated with a loading pattern of a nuclear reactor comprising a core in which fuel assemblies are loaded according to the loading pattern, the fuel assemblies comprising fuel rods each including nuclear fuel pellets and a cladding surrounding the pellets, the electronic system comprising:

• a first calculating module configured to calculate a reference principal pellet-cladding interaction margin for a reference loading pattern of the fuel assemblies in the core;
• a second calculating module configured to calculate, on the one hand, a reference secondary pellet-cladding interaction margin for the reference loading pattern, and on the other hand, a modified secondary margin for a modified loading pattern of the fuel assemblies in the core;
• and
• a comparison module configured to compare the modified secondary pellet-cladding interaction margin with the reference secondary pellet-cladding interaction margin,
• the comparison module further being configured to calculate a modified principal pellet-cladding interaction margin for the modified loading pattern, depending on the comparison of the modified secondary pellet-cladding interaction margin with the reference secondary pellet-cladding interaction margin,
• the electronic system configured for controlling the state of the power balance of the nuclear reactor by using the calculated principal pellet-cladding interaction margin for a considered loading pattern of the fuel assemblies in the core to avoid rupture by pellet-cladding interaction of the claddings present in the core,
• wherein neutronic calculations and thermodynamic calculations are done to calculate each pellet-cladding interaction margin, and
• wherein the neutronic calculations and the thermomechanical calculations are coupled to calculate a corresponding principal pellet-cladding interaction margin, the thermomechanical calculations being uncoupled from the neutronic calculations to calculate a corresponding secondary pellet-cladding interaction margin.

A method for calculating a pellet-cladding interaction margin associated with a loading pattern of a nuclear reactor comprising a core in which fuel assemblies are loaded according to the loading pattern, the fuel assemblies comprising fuel rods each including nuclear fuel pellets and a cladding surrounding the pellets,

• the method being implemented by an electronic system and comprising the following steps:
• a) determining a rupture value of the physical quantity characterizing a rupture of the cladding,
• b) calculating a reference principal pellet-cladding interaction margin for a reference loading pattern of the fuel assemblies in the core,
• c) calculating a reference secondary pellet-cladding interaction margin for the reference loading pattern,
• d) calculating a modified secondary pellet-cladding interaction margin for a modified loading pattern of the fuel assemblies in the core,
• e) calculating a modified principal pellet-cladding interaction margin for the modified loading pattern, depending on a comparison of the modified secondary pellet-cladding interaction margin with the reference secondary pellet-cladding interaction margin;
• the method further comprising controlling the state of the power balance of the nuclear reactor by using the calculated principal pellet-cladding interaction margin for a considered loading pattern of the fuel assemblies in the core to avoid rupture by pellet-cladding interaction of the claddings present in the core;
• wherein step b) comprises the following sub-steps:
• b1) simulating at least one operating transient of the nuclear reactor,
• b2) calculating the value reached by at least one physical quantity during the operating transient in at least part of a cladding of a fuel rod, and
• b3) determining, as reference principal pellet-cladding interaction margin, the deviation between the maximum value reached by the value calculated in sub-step b2) during the transient and a technological limit of the fuel rod;
• wherein step a) includes:
• subjecting previously irradiated fuel rods to experimental nuclear power ramps,
• calculating the values reached by the physical quantity in at least one cladding broken during a power ramp, and
• selecting the rupture value as being the minimum value from among the calculated values reached;
• wherein each of steps c) and d) includes, for each fuel assembly, the following sub-steps:
• i) simulating an evolution of the operation of the nuclear reactor by applying, to the fuel rods, a nuclear power ramp from the nil power,
• ii) calculating the values reached by a physical quantity in the claddings of the fuel rods,
• iii) comparing the values reached to the rupture value,
• iv) determining a power at break equal to:
• I) the power associated with the rupture value, if a minimum value from among the values reached calculated in sub-step ii) is equal to the rupture value, or
• II) infinity, if no value, from among the values reached calculated in sub-step ii), is equal to the rupture value,
• v) evaluating a power margin by difference between the power at break determined in sub-step iv) and an estimated maximum power in the fuel assembly,
• the corresponding secondary pellet-cladding interaction margin, calculated during each of steps c) and d), being equal to the minimum margin from among the power margins evaluated for the fuel assemblies in sub-step v).