"Burn-up credit will play a major role in future licensing evaluations for spent fuel storage, transportation and dissolution." –Raap, B., Nomura, Y. and Sartori, E., 2001. Overview of the burnup credit activities at OECD/NEA/NSC.
Criticality safety must be guaranteed based on reliable methods in the fuel manufacture, storage, transportation and reprocessing and disposition. However, it is specifically difficult in the case of spent nuclear fuels to estimate the reactivity of spent fuels due to large uncertainties. Limited quantification of spent fuels elemental composition and spatial distribution exists fuels, even if structural/elemental information of fresh fuels and operation history are available. Therefore, to avoid criticality accidents under any considerable conditions, relatively large safety margins are required, knowing that they have significant conservatisms that impose enormous economic burden on stakeholders.
Burn-up credit is a safety approach that accounts for the reduction in the reactivity of configurations with spent nuclear fuel due to the change in their composition after irradiation. Criticality is the state of a nuclear reactor when there is a self-sustaining chain reaction of fissionable material.
The reactivity of nuclear reactor fuels typically decreases as the fuel burnup proceeds. Essentially, the decrease in reactivity comes from a reduction in concentration of fissile nuclides and an increase in concentration of fission products that absorb neutrons. Burn-up credit, which is currently of wide interest in the field of nuclear criticality safety research, generally involves taking into account this reactivity decrease for criticality safety assessments and control of spent fuel by crediting burn up of fuel.
To provide information of more accurate safety analysis, Expert Group on Burn-up Credit Criticality Safety (EGBUC) was formed in 1991 and published several benchmark reports. The NEA's activities had examined burn-up credit as applied to criticality safety in the transportation, storage and treatment of spent fuel for a wide range of fuel types and reactors, including UOX and MOX fuels for PWR, BWR and VVER.
| Benchmarks on burn-up credit | |||
|---|---|---|---|
| Benchmark | Application | Description | Status |
| I-A | PWR UOX | Multiplication factor, spectra and reaction rates calculations for an infinite PWR fuel rod lattice with varying compositions (different enrichments, burnups, cooling-times, presence or absence of actinides -major and minor- and of fission products - major and minor- | |
| I-B | PWR UOX | Depletion calculations for a simple infinite PWR pin-cell lattice. | Report published in January 1994 |
| II-A | PWR UOX | Multiplication factor calculations of an infinite array of PWR fuel with finite axial height. The aim being to study the effect of axial burn-up profile on criticality calculations of PWR fuel storage. | Report published in February 1996 |
| II-B | PWR UOX | Multiplication factor and spatial fission distribution calculations of a realistic PWR spent fuel transport cask including accidental situations. The aim being to further study the effect of axial burn-up profile on criticality calculations. | Report published in May 1998 |
| II-C | PWR UOX | Multiplication factor and fission distribution calculations of a realistic PWR fuel transport cask. The aim being to study the sensitivity to the axial burnup shape. | Report published in September 2008 |
| II-D | PWR UOX | Multiplication factor and fission distribution calculations of a realistic PWR fuel transport cask. The aim being to study control rods effects on spent fuel composition. | Report published in December 2006 |
| II-E | PWR UOX | Study on the impact of changes in the isotopic inventory due to control rod insertions in PWR UO2 fuel assemblies during irradiation on the end effect | Report published in June 2015 |
| III-A | BWR UOX | Criticality calculations of an infinite array of BWR spent fuel assemblies with emphasis on axial burnup and void profiles. | Report published in September 2000 |
| III-B | BWR UOX | Depletion calculations of an array of BWR fuel. | Report published in February 2002 |
| III-C | BWR UOX | Compositions of Spent Fuel Assembly for BUC and Criticality control of damaged fuel. | Report published in March 2016 |
| III-D | BWR UOX | Gadolinium- Bearing Fuel Rods in Boiling Water Reactor Assemblies for Storage and Transportation |
Report published in August 2023 |
| IV-A | PWR MOX | Reactivity Prediction Calculations for Infinite Arrays of PWR MOX Fuel Pin Cells | Report published in May 2003 |
| IV-B | PWR MOX | Inventory MOX Fuel Depletion Calculations | Report published in May 2003 |
| VI | PWR UOX | Burn-up profile in a VVER-440 assembly | Analysis of results ongoing |
| VII | PWR UOX | Study of spent fuel compositions for long-term disposal | Report Published February 2012 |
| VIII | PWR UOX | Numerical benchmark for the analysis of small-sample reactivity experiments |
Report published August 2017 |
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