Bundle Degradation Research Articles

Analysis of Core Degradation and Fission Products Release in Phebus FPT1 Test at IRSN by Detailed Severe Accidents Analysis Code, IMPACT/SAMPSON

The Phebus FPT1 test at IRSN, selected to be ISP-46, has been analyzed by the IMPACT/SAMPSON code, a detailed severe accidents (SAs) analysis code for an LWR. The results for the Bundle phase, involving an examination of bundle degradation and fission products release during SA conditions, were reported. Conclusions obtained from the analyses are as follows: (1) Temperature changes of fuel, cladding, and control rod were well predicted until about 12,000 s when almost all thermocouples measuring them failed. (2) Accumulated hydrogen generation due to Zr/steam reaction differed only about 3% from the test result. (3) Overall, good agreement was obtained for the fuel relocation and an accumulation of debris just below original spacer position was well predicted in the analysis. (4) Analysis of enhanced diffusion due to degraded fuel by Lewis et al.'s method enabled simulation of release behaviors of Xe, Cs, I, and Te within the uncertainties in the test when the effective surface to volume ratio was changed in the evaluation of the UO2 oxidation. (5) Diffusion analysis through single-crystal grain could trace the release behavior of Mo, Sb, Tc, Ru, and Ba observed in the Phebus FPT1 test after optimizing their diffusion coefficients with the effective surface to volume ratio determined above.

Analysis of Core Degradation and Fission Products Release in Phebus FPT1 Test at IRSN by Detailed Severe Accidents Analysis Code, IMPACT/SAMPSON

The Phebus FPT1 test at IRSN, selected to be ISP-46, has been analyzed by the IMPACT/SAMPSON code, a detailed severe accidents (SAs) analysis code for an LWR. The results for the Bundle phase, involving an examination of bundle degradation and fission products release during SA conditions, were reported. Conclusions obtained from the analyses are as follows: (1) Temperature changes of fuel, cladding, and control rod were well predicted until about 12,000 s when almost all thermocouples measuring them failed. (2) Accumulated hydrogen generation due to Zr/steam reaction differed only about 3% from the test result. (3) Overall, good agreement was obtained for the fuel relocation and an accumulation of debris just below original spacer position was well predicted in the analysis. (4) Analysis of enhanced diffusion due to degraded fuel by Lewis et al.'s method enabled simulation of release behaviors of Xe, Cs, I, and Te within the uncertainties in the test when the effective surface to volume ratio was changed in the evaluation of the UO2 oxidation. (5) Diffusion analysis through single-crystal grain could trace the release behavior of Mo, Sb, Tc, Ru, and Ba observed in the Phebus FPT1 test after optimizing their diffusion coefficients with the effective surface to volume ratio determined above.

B4C/Zircaloy Reaction at Temperatures from 1,173 to 1,953 K

With a view to obtaining the basic information on the influence of the BWR type control rod failure on the bundle degradation during a severe accident, the chemical interactions between B4C neutron absorber material and Zircaloy were investigated in the temperature range 1,173 to 1,953 K. Reaction layers were formed at the interface between two materials. Growth of the reaction layer and decrease in the Zircaloy thickness were measured as a function of temperature and time, thereby the reaction kinetics was evaluated. The overall reaction generally obeyed a parabolic rate law and the reaction rate constant at each examined temperature was determined. A sudden increase of the reaction rate was found at a temperature between 1,823 and 1,873 K. It could be caused by liquid phase formation of the zirconium-boron carbide system at the corresponding temperature, which occurred locally at the reaction front. Accordingly, the temperature dependence of the reaction rate constants was estimated separately for the two temperature ranges of 1,173 to 1,823 K and 1,873 to 1,953 K.

Material Distribution in Light Water Reactor-Type Bundles Tested Under Severe Accident Conditions

Severe fuel damage experiments simulating small-break loss-of-coolant accidents have been carried out in the CORA out-of-pile test facility at Forschungszentrum Karlsruhe. Rod bundles with electrically heated fuel rod simulators containing annular UO2 pellets, UO2 full pellet rods, and absorber rods of two kinds (Ag/In/Cd to represent pressurized water reactor conditions and B4C to represent boiling water reactor and VVER-1000fuel elements) were subjected to temperature transients up to 2300 K. A special method was applied to determine the axial mass distribution of bundle materials.The low-temperature melt formation by various interactions between zirconium and components of absorber and spacer grids strongly influences the bundle degradation and material relocation. Absorber materials can separate from the fuel by a noncoherent relocation of the materials at different temperatures. The distributions of solidified materials in the different test bundles show a clear dependence on the axial temperature profile. Coolant channel blockages are observed mainly at the lower end of the bundle, i.e., near the lowest elevation at which an oxidation excursion resulting from the highly exothermic zirconium-steam reaction had been experienced. This elevation corresponds with a steep axial temperature gradient in the maximum temperature attained. Oxide layers on Zircaloy result in reduced melt formation.

B4C/Zircaloy Reaction at Temperatures from 1,173 to 1,953K.

With a view to obtaining the basic information on the influence of the BWR type control rod failure on the bundle degradation during a severe accident, the chemical interactions between B4C neutron absorber material and Zircaloy were investigated in the temperature range 1,173 to 1,953 K. Reaction layers were formed at the interface between two materials. Growth of the reaction layer and decrease in the Zircaloy thickness were measured as a function of temperature and time, thereby the reaction kinetics was evaluated. The overall reaction generally obeyed a parabolic rate law and the reaction rate constant at each examined temperature was determined. A sudden increase of the reaction rate was found at a temperature between 1,823 and 1,873 K. It could be caused by liquid phase formation of the zirconium-boron carbide system at the corresponding temperature, which occurred locally at the reaction front. Accordingly, the temperature dependence of the reaction rate constants was estimated separately for the two temperature ranges of 1,173 to 1,823 K and 1,873 to 1,953 K.

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The international Phébus Fission Product Programme investigated key phenomena occurring in light water reactor core meltdown accidents in a series of five in-pile experiments. Four of these tests focused on the degradation of fuel rod bundles, containing a central control rod, and on the resulting release of fission products, structural materials and actinides from the fuel rod bundle, their transport in the reactor coolant system (RCS) and their subsequent behaviour in the containment vessel. Various steam contents were used in the RCS, from highly oxidising conditions (in FPT0 and FPT1) to more reducing ones (in FPT2 and FPT3).The “early degradation phase” took place at the beginning of the Phébus driver core and fuel bundle heat-up phase, with a quasi-intact fuel bundle geometry. During this phase, the degradation of the control rod and the oxidation runaway due to the fast oxidation of the Zircaloy claddings of the fuel rods, were two major events which took place.The oxidation runaway locally increased the temperatures much above the temperatures resulting from the Phébus driver core heat transfer to the bundle and yielded a large hydrogen release, which amounted to 70–80% of the whole hydrogen production during the tests. The maximum hydrogen flow rates increased with increasing steam flow rates injected at the fuel bundle inlet.The failure mechanisms of silver–indium–cadmium (used in three tests) and boron carbide (used in one test) control rods involve eutectic interactions amongst the components of these control rods. Mechanical deformations of the control rod stainless steel cladding against the Zircaloy control rod guide tube are the main presumed mechanisms for the beginning of these eutectic formations. However, different post-failure scenarios can be postulated for the effect of control rod degradation on fuel bundle degradation for both types of control rods. The exothermic oxidation of the exposed boron carbide pellets led to the release of carbonaceous species (CO, CO2) as well as of additional hydrogen, but no significant methane release could be detected above the limits of detection.Overall, the results confirmed existing knowledge concerning early phase degradation phenomenology found in previous integral experiments such as CORA and QUENCH (Karlsruhe Institute of Technology) and Phébus SFD (IRSN Cadarache), and formed a sound basis for analysis of the late phase degradation subsequently observed. Quantitative analysis of boron carbide control rod degradation in FPT3 pointed to a need for improved modelling of chemical reactions involving this material, particularly its oxidation in steam; this has been studied in the BECARRE experiments conducted by IRSN in the International Source Term Programme, leading to better quantitative understanding and improved modelling in codes such as the European reference severe accident analysis code ASTEC.