Christian Hellwig

Cermet sphere-pac concept for inert matrix fuel

In the inert matrix fuel concept, plutonium reprocessed from spent fuel is burned in an inert matrix, e.g. yttria-stabilized zirconia. Coming from wet reprocessing, the internal gelation can perform an easy micro-spheres production. Utilization of these particles in a sphere-pac realizes a direct fuel production. Besides being economical, this direct usage offers an almost dustless fabrication. One disadvantage of yttria-stabilized zirconia as matrix is its low thermal conductivity. A further reduction by the macroscopic structure of a sphere bed seems unacceptable. This can be eluded by the insertion of a highly conducting phase. Similar to the cermet concept with the embedment of ceramic fuel into metal, the infiltration of a fine metal fraction into a coarse ceramic fuel fraction is studied here. The initial thermal conductivity shows much higher calculated values and the sintering behaviour is also clearly enhanced compared to the pure ceramic bed.

Cladding Tube Deformation Test for Stress Reorientation of Hydrides

The phenomenon of stress reorientation of hydrides in fuel cladding tubes was studied with the help of a test technique named as the “cladding tube deformation test” (CTDT). Unirradiated Zircaloy-2 (Zry-2) cladding tube specimens charged with 250 ppm of hydrogen were tested. The test consisted of heating specimens to 400°C and then cooling them down to room temperature at cooling rates of 0.7°C/min and 2°C/min. During the cooling phase, a constant tensile load was applied to specimens with the help of two inner cylinder halves. The stress-strain fields developed in the tube specimen were calculated with the help of the finite element method (FEM). Post-testing metallographic observations revealed considerable amount of hydrides oriented radially as a result of loading. Lengths and relative position of all hydride bands were determined through a semi-automatic image processing technique. Mapping of FEM calculated stress field on the metallographic section helped to determine a stress threshold value of 72 MPa. Tests performed with a cooling rate 2°C/min showed a considerably lower extent of hydride reorientation. The above test technique was validated by performing internal pressurization tests using pre-hydrided Zry-2 cladding tubes. The external diameter of the tube was tapered by fine turning in order to exert different uniform hoop stress values on different cross sections of the tube. Internal pressurization tests were performed with a maximum temperature of 400°C and a cooling rate of 0.5°C/min. Post-testing metallographic observations on several cross sections of the specimen revealed a stress reorientation threshold of 65 MPa. Percentage of radially oriented hydrides as well as their lengths increased with the stress level until it reached a plateau. Further tests were carried out with the CTDT technique on specimens with H2 content ranging from 560 ppm to 750 ppm. These results revealed a reorientation stress threshold of around 120 MPa for higher hydrogen content.

Interpretation of experimental results from moderate-power in-pile testing of a Pu-Er-Zr-oxide inert matrix fuel

Pu-Er-Zr oxide as an inert matrix fuel (IMF) could be an attractive option for a once-through LWR strategy aimed at reducing the currently growing plutonium stockpiles. A basic question related to the practical introduction of such an IMF into a current-day LWR, without affecting any significant change in fuel element design and core performance, is the irradiation behaviour of the proposed IMF. The present paper reports first results from a corresponding validation experiment which has been launched recently, viz. the comparative in-pile testing of specialty fabricated IMF and PuO2/UO2 mixed-oxide (MOX) fuel rodlets in the Halden reactor. The maximum linear power rating was limited to about 25 kW/m during the currently reported initial testing phase, as compared to the design value of 35 kW/m to be aimed at during subsequent cycles at Halden. While fuel centre-line temperatures were found to be within the expected range for both fuel types, the variation of pressure in the rodlets has clearly indicated significant densification in the IMF specimens, with fission gas release being non-detectable at the low burnups achieved to date. A possible sintering model for the IMF is proposed to explain the experimental results obtained during this moderate-power testing phase, viz. the apparent geometrical stability of the IMF in spite of its strong initial densification

FUJI, an Initial Sintering Comparison Test for Pelletized-, Sphere-Pac- and Vipac-Fast Breeder Reactor Mixed Oxide Fuel

Options for fuel cycle technology improvement have strongly regained attention lately with the revival of nuclear energy production interests and plans for next generation nuclear systems. Various fuel forms, geometries and production paths are being looked at. Within the FUJI collaboration program among Japan Atomic Energy Agency (JAEA, former JNC), Paul Scherrer Institute (PSI, Switzerland) and Nuclear Research and Consultancy Group (NRG, the Netherlands) the production paths of plutonium and neptunium mixed oxide- (sphere-pac- and vipac-) particle fuels (20 wt% Pu, 5 wt% Np and 75 wt% U) are tested as well as the initial sintering and power-to-melt behavior under simulated Fast Breeder Reactor (FBR) conditions. The various fuel forms were produced at PSI under the support of JNC, the irradiations were accomplished at High Flux Reactor (HFR) in Petten, the post irradiation examinations are being achieved mainly at NRG and the fuel modelling being performed at JNC and PSI. The present paper reviews mainly the project planning, fuel behaviour-pre-calculations and the fuel- and fuel segment-production. A short overview of the irradiation conditions and ceramographic post irradiation examination analyses is also given.

Development of Sphere-Pac Nuclear Fuel Behavior Analysis Code

In the feasibility study of a FBR cycle system, alternative nuclear fuels are considered. One option is the sphere-pac fuel, which consists of a spherical particle bed vibro-packed into the cladding tube. Potential advantages of this concept are an almost powder less production and lower mechanical interaction between fuel and cladding. JNC has developed a new fuel irradiation behavior analysis code for sphere-pac fuel. It is based on a pellet type fuel code because of its analogous irradiation behavior. However, the macrostructure of the sphere-pac fuel affects some properties. Therefore, some modifications have to be applied. The PSI SPACON model is used to calculate the effective thermal conductivity. This theoretical calculation model takes into account the discontinuous structure with voids between the particles. It also accounts the increased thermal conductivity due to the sintering of the bed. A mechanistic sintering model addresses this structural change. Finally, the mechanical interaction between particle bed and cladding is studied. A three-dimensional Distinct Element Method is applied to investigate mechanical properties of the bed. Together with the theory of radial expansion pressure, these results then built a new model for the sphere^ac mechanical interaction.