## Abstract

A statistical fuel failure analysis procedure has been developed, and now applied in full scale to a large break loss-of-coolant accident (LBLOCA) scenario in EPR-type nuclear power plant. The goal is to statistically determine the number of failing fuel rods. It is also important to bring out the underlying causes for rod

failures, and therefore a sensitivity analysis procedure for the preceding complex calculation chain is outlined and adopted.

In order to produce a statistically reliable estimation on the number of failing rods, a large number of simulations are required. In the analysis, single rods are simulated with the coupled fuel performance - thermal hydraulics code

FRAPTRAN-GENFLO. The statistically varied factors are divided into global and local by their range of influence. Transient thermal hydraulic and power history boundary conditions and model parameters are global and affect all the rods. Fuel rod manufacturing parameters and other rod position related factors are local and are related to a certain rod. The number of global scenarios is 59 as dictated by the Wilks’ formula, and 1000 randomly sampled rods are simulated in each scenario. The thermal hydraulic and transient power boundary conditions for the fuel performance code are calculated with a statistical version of the system code APROS. The steady-state irradiation histories are simulated with FRAPCON, and the steady-state power histories used as boundary conditions for

FRAPCON are obtained from SIMULATE calculations. Thus, a multistage calculation chain is required to consolidate the procedure. As an outcome, in the worst global scenario, 1.4% of the simulated rods failed. It can be concluded that the Finnish safety regulations, i.e. max. 10% of the rods allowed to fail, are met.

In order to find out which input parameters have significance to the outcome of the analysis, a sensitivity analysis is done. At the moment, only the worst global scenario is considered but later on it is possible to include also the sensitivities of the global factors. Due to complexity of the existing data, first the relevant input parameters for the sensitivity analysis have to be specified. Data visualization with a cobweb graph is used for the screening. Then, selected sensitivity measures are calculated between the chosen input and output parameters. The sensitivity indices calculated are the Borgonovo’s delta measure, the first order Sobol’ sensitivity index, and squared Pearson correlation coefficients. The first mentioned is a novelty in this context. As an outcome, the most relevant

parameters with respect to the cladding integrity were determined to be the decay heat power during the transient, the thermal hydraulic conditions in the rod’s location in the reactor, and the steady-state irradiation history of the rod

as represented in this analysis by the rod burnup.

failures, and therefore a sensitivity analysis procedure for the preceding complex calculation chain is outlined and adopted.

In order to produce a statistically reliable estimation on the number of failing rods, a large number of simulations are required. In the analysis, single rods are simulated with the coupled fuel performance - thermal hydraulics code

FRAPTRAN-GENFLO. The statistically varied factors are divided into global and local by their range of influence. Transient thermal hydraulic and power history boundary conditions and model parameters are global and affect all the rods. Fuel rod manufacturing parameters and other rod position related factors are local and are related to a certain rod. The number of global scenarios is 59 as dictated by the Wilks’ formula, and 1000 randomly sampled rods are simulated in each scenario. The thermal hydraulic and transient power boundary conditions for the fuel performance code are calculated with a statistical version of the system code APROS. The steady-state irradiation histories are simulated with FRAPCON, and the steady-state power histories used as boundary conditions for

FRAPCON are obtained from SIMULATE calculations. Thus, a multistage calculation chain is required to consolidate the procedure. As an outcome, in the worst global scenario, 1.4% of the simulated rods failed. It can be concluded that the Finnish safety regulations, i.e. max. 10% of the rods allowed to fail, are met.

In order to find out which input parameters have significance to the outcome of the analysis, a sensitivity analysis is done. At the moment, only the worst global scenario is considered but later on it is possible to include also the sensitivities of the global factors. Due to complexity of the existing data, first the relevant input parameters for the sensitivity analysis have to be specified. Data visualization with a cobweb graph is used for the screening. Then, selected sensitivity measures are calculated between the chosen input and output parameters. The sensitivity indices calculated are the Borgonovo’s delta measure, the first order Sobol’ sensitivity index, and squared Pearson correlation coefficients. The first mentioned is a novelty in this context. As an outcome, the most relevant

parameters with respect to the cladding integrity were determined to be the decay heat power during the transient, the thermal hydraulic conditions in the rod’s location in the reactor, and the steady-state irradiation history of the rod

as represented in this analysis by the rod burnup.

Original language | English |
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Title of host publication | Top Fuel 2016 |

Subtitle of host publication | LWR Fuels with Enhanced Safety and Performance |

Publisher | American Nuclear Society (ANS) |

Pages | 1115-1124 |

ISBN (Electronic) | 978-0-89448-730-9 |

ISBN (Print) | 978-0-89448-734-7 |

Publication status | Published - 2016 |

MoE publication type | A4 Article in a conference publication |

Event | Top Fuel 2016 - Boise, United States Duration: 11 Sept 2016 → 15 Sept 2016 |

### Conference

Conference | Top Fuel 2016 |
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Country/Territory | United States |

City | Boise |

Period | 11/09/16 → 15/09/16 |

Other | Top Fuel, also called Light Water Reactor (LWR) Fuel Performance Meeting, is an annual topical meeting organized by the American Nuclear Society, the European Nuclear Society, the Atomic Energy Society of Japan, the Chinese Nuclear Society, the Korean Nuclear Society, and the International Atomic Energy Agency. |

## Keywords

- loss-of-coolant accident
- statistical fuel failure analysis
- sensitivity analysis
- FRAPTRAN-GENFLO
- EPR