MOX Fuel Performance Experiment Specified for Japanese PWR Utilisation in the HBWR

Abstract

In order to obtain high burn-up MOX fuel irradiation performance data, SBR and MIMAS MOX fuel rods with Pufissile enrichment of about 6 wt% have been irradiated in the HBWR. In-pile performance data of MOX have been obtained, and the peak burn-up of MOX pellet have reached to 66 GWd/tM as of October 2004. MOX fuel temperature is confirmed to have no significant difference compared to UO 2 , if taking into account adequately for thermal conductivity degradation due to PuO 2 addition and burn-up development, and measured fuel temperature agrees well with HB-FINE code calculation up to high burn-up region. Fission gas release of MOX is possibly larger than UO 2 based on temperature and pressure assessment. No significant difference is confirmed between SBR and MIMAS MOX on FGR behaviour. MOX fuel swelling rate agrees well with solid swelling rate. Cladding elongation data shows onset of PCMI in high power region. Ramp test data from other experiment programs with various types of MOX fabrication route confirms superior PCI resistance of MOX compared to UO 2 , due to enhanced creep rate of MOX. The irradiation is expected to continue until achieving of 70 GWd/tM (MOX pellet peak).

Recent activities in the field of nuclear materials and nuclear fuels

“…To explain the difference in the thermal conductivity degradation between mixed oxide (MOX) fuel and UO 2 fuel after burnup [16], the quasi-two phase material model was proposed [17]. The numerical result showed the gradual increase of the difference with burnup which was consistent with the results from the irradiation tests.…”

Thermal conductivity degradation analyses of LWR MOX fuel by the quasi-two phase material model

“…The irradiation test results of the in situ measurements were analyzed and reported by Fujii et al [4]. In this irradiation tests, the fuel center temperatures measured during irradiation of the MOX fuel were comparable or slightly lower in comparison with those of the reference UO 2 fuel at the same power as shown in Figure 1.…”

Thermal Energy Transport in Oxide Nuclear Fuel

“…Recommended values of A0 and B for pristine samples of UO2, ThO2, and CeO2 are summarized in Table 1 and representative conductivity profiles as a function of temperature are shown in Figure 24. The simplified formulation, and minor variations thereof, has been used to describe the thermal conductivity of fresh and irradiated UO2 fuel [538,539], mixed uranium-plutonium dioxide (MOX) fuel [540,541,542], and rare-earth doped UO2 [543].…”

“…In particular, A can be represented as A = A 0 + ∑ i x i A i ′ , where the summation is over defects i , x i is the concentration of defect i , A i ′ is the scattering parameter for defect i , and A 0 is the residual value for a pristine material. , Recommended values of A 0 and B for pristine samples of UO 2 , ThO 2 , and CeO 2 are summarized in Table , and representative conductivity profiles as functions of temperature are shown in Figure . The simplified formulation and minor variations thereof have been used to describe the thermal conductivity of fresh and irradiated UO 2 fuel, , mixed uranium–plutonium dioxide (MOX) fuel, − and rare-earth-doped UO 2 …”