Thermo-mechanical analysis of supercritical pressure light water-cooled fast reactor fuel rod by FEMAXI-6 code

Thermo-mechanical behaviors of supercritical pressure light water cooled fast reactor (SWFR) fuel rod and cladding have been investigated by FEMAXI-6 (Ver.1) code with high enriched MOX fuel at elevated operating condition of high coolant system pressure (25 MPa) and high temperature (500 °C in core average outlet temperature). Fuel rod failure modes and associated fuel rod design criteria that is expected to be limiting in SWFR operating condition have been investigated in this fuel rod design study. Fuel centerline temperature is evaluated to be 1853 °C and fission gas release fraction is about 45% including helium production. Cumulative damage fraction is evaluated by linear life fraction rule with time-to-rupture correlation of advanced austenitic stainless steel. In a viewpoint of mechanical strength of fuel cladding against creep rupture and cladding collapse at high operation temperature, currently available stainless steels or being developed has a potential for application to SWFR. Admissible design range in terms of initial gas plenum pressure and its volume ratio are suggested for fuel rod design The stress ranges suggested by this study could be used as a preliminary target value of cladding material development for SWFR application.     

Impact of thicker cladding on the nuclear parameters of the NPP Krško fuel

To make fuel rods more resistant to grid-to-rod fretting or other cladding penetration failures, the cladding thickness could be increased or strengthened. Implementation of thicker fuel rod cladding was evaluated for the NPP Krško that uses 16 × 16 fuel design. Cladding thickness of the Westinghouse standard fuel design (STD) and optimized fuel design (OFA) is increased. The reactivity effect during the fuel burnup is determined. To obtain a complete realistic view of the fuel behaviour a typical, near equilibrium, 18-month fuel cycle is investigated. The most important nuclear core parameters such as critical boron concentrations, isothermal temperature coefficient and rod worth are determined and compared.     

CFD analyses in tight-lattice subchannels and seven-rod bundle geometries of a Super Fast Reactor

This paper presents CFD analyses in heat unsymmetric subchannels and heat symmetric seven-rod bundle geometries of a Super Fast Reactor (Super FR) fuel assembly using STAR-CD. The purpose of CFD analyses in heat unsymmetric subchannels is to evaluate the effect of the power differences on the heat transfer in subchannels of the Super Fast Reactor. For heat symmetric seven-rod bundles, the effects of the gap clearance between the fuel rod and the assembly wall and the displacement of the fuel rod on the circumferential temperature distributions and Maximum Cladding Surface Temperature (MCST) are analyzed. The results show that larger power difference between fuel rods gives larger circumferential temperature difference of the hottest fuel rods. Considering cross flow between edge and ordinary subchannels, 1 mm gap between the fuel rod and the assembly wall is better for small MCST although the circumferential temperature difference in edge subchannel is large. MCST increases exponentially with the displacement. The relative error of displacement should be less than 1% if the allowable increment of MCST due to displacement is less than 6 °C.     

Fuel performance analysis for PWR cores

Attainable discharge burnups for oxide and hydride fuels in PWR cores were investigated using the TRANSURANUS fuel performance code. Allowable average linear heat rates and coolant mass fluxes for a set of fuel designs with different fuel rod diameters and pitch-to-diameter ratios were obtained by VIPRE and adopted in the fuel code as boundary conditions. TRANSURANUS yielded the maximum rod discharge burnups of the several design combinations, under the condition that specific thermal-mechanical fuel rod constraints were not violated. The study shows that independent of the fuel form (oxide or hydride) rods with (a) small diameters and moderate P/Ds or (b) large diameters and small P/Ds give the highest permissible burnups limited by the rod thermal-mechanical constraints. TRANSURANUS predicts that burnups of ∼74 MWd/kg U and ∼163 MWd/kg U (or ∼65.2 MWd/kg U oxide-equivalent) could be achieved for UO₂ and UZrH_x fuels, respectively. Furthermore, for each fuel type, changing the enrichment has only a negligible effect on the permissible burnup. The oxide rod performance is limited by internal pressure due to fission gas release, while the hydride fuel can be limited by excessive clad deformation in tension due to fuel swelling, unless the fuel rods will be designed to have a wider liquid metal filled gap. The analysis also indicates that designs featuring a relatively large number of fuel rods of relatively small diameters can achieve maximum burnup and provide maximum core power density because they allow the fuel rods to operate at moderate to low linear heat rates.     

Principle of rationalizing the criteria for abnormal transients of the Super LWR with fuel rod analyses

A detailed fuel rod design is carried out for the first time in the development of Supercritical-pressure Light Water Reactor (Super LWR). The fuel rod design is similar to that of LWR, consisting of UO₂ pellets, a gas plenum and a Stainless Steel Cladding. The principle of rationalizing the criteria for abnormal transients of the Super LWR is developed. The fuel rod integrities can be assured by preventing plastic strains on the cladding, preventing the cladding buckling collapse, and keeping the pellet centerline temperature below its melting point. The FEMAXI-6 fuel analysis code is used to evaluate the fuel rod integrities in abnormal transient conditions. Detailed analyses have shown that allowable limits to the maximum fuel rod power and maximum cladding temperature can be determined to assure the fuel integrities. These limits may be useful in the plant safety analyses to confirm the fuel integrities during abnormal transients.     

Design of a fuel element for a lead-cooled fast reactor

The options of a lead-cooled fast reactor (LFR) of the fourth generation (GEN-IV) reactor with the electric power of 600 MW are investigated in the ELSY Project. The fuel selection, design and optimization are important steps of the project. Three types of fuel are considered as candidates: highly enriched Pu-U mixed oxide (MOX) fuel for the first core, the MOX containing between 2.5% and 5.0% of the minor actinides (MA) for next core and Pu-U-MA nitride fuel as an advanced option. Reference fuel rods with claddings made of T91 ferrite-martensitic steel and two alternative fuel assembly designs (one uses a closed hexagonal wrapper and the other is an open square variant without wrapper) have been assessed. This study focuses on the core variant with the closed hexagonal fuel assemblies. Based on the neutronic parameters provided by Monte-Carlo modeling with MCNP5 and ALEPH codes, simulations have been carried out to assess the long-term thermal-mechanical behaviour of the hottest fuel rods. A modified version of the fuel performance code FEMAXI-SCK-1, adapted for fast neutron spectrum, new fuels, cladding materials and coolant, was utilized for these calculations. The obtained results show that the fuel rods can withstand more than four effective full power years under the normal operation conditions without pellet-cladding mechanical interaction (PCMI). In a variant with solid fuel pellets, a mild PCMI can appear during the fifth year, however, it remains at an acceptable level up to the end of operation when the peak fuel pellet burnup ∼80 MW d kg⁻¹ of heavy metal (HM) and the maximum clad damage of about 82 displacements per atom (dpa) are reached. Annular pellets permit to delay PCMI for about 1 year. Based on the results of this simulation, further steps are envisioned for the optimization of the fuel rod design, aiming at achieving the fuel burnup of 100 MW d kg⁻¹ of HM.     

Fuel design study and optimization for PEACER development

In order to increase the transmutation capability for the Pb-Bi cooled burner, PEACER, metallic fuel rods (60U–30TRU–10Zr with Pb-bond in HT-9 clad) having a short (50 cm) active length with a large gas plenum have been designed with a peak design TRU burnup of 15%. A 17 × 17 square-lattice with relatively high pitch-to-diameter ratio was employed to reduce the actinide production and pumping load associated with the high-density coolant. Fuel rod failure modes are identified and fuel design criteria are established. A fuel rod design model, named as RODSIS, has been obtained by incorporating Pb properties and a cladding oxidation rate equation. A thermal analysis has been conducted for a fuel rod having peak-power based on a predicted power distribution and history during an equilibrium cycle. Taking into account the high coolant density, all fuel rods are fastened in the assembly using a stiff middle grid structure and softer end grids made of HT-9. Based on fuel rod thermal analysis results, a finite element analysis (FEA) has been conducted for both thermal and mechanical analyses of the middle grid structure. Furthermore, a fuel assembly static analysis has been conducted to determine the consequences of the axial loading caused by buoyancy and flow. The PEACER fuel system design was visualized by using a three-dimensional design and visualization software.     

Some insights into the role of axial gas flow in fuel rod behaviour during the LOCA based on Halden tests and calculations with the FALCON-PSI code

A model for axial gas flow in a fuel rod during the LOCA is integrated into the FRELAX model that deals with the thermal behaviour and fuel relocation in the fuel rods of the Halden LOCA test series. The first verification was carried out using the experimental data for the inner pressure during the gas outflow after cladding rupture in tests 3, 4 and 5. Furthermore, the modified FRELAX model is implicitly coupled to the FALCON fuel behaviour code.The analysis with the new methodology shows that the dynamics of axial gas-flow along the rod and through the cladding rupture can have a strong influence on the fuel rod behaviour. Specifically, a delayed axial gas redistribution during the heat-up phase of the LOCA can result in a drop of local pressure in the ballooned area, which is eventually able to affect the cladding burst. The results of the new model seem to be useful when analysing some of the Halden LOCA tests (showing considerable fuel relocation) and selected cases of LOCA in full-length fuel rods. While the short rods used in the Halden tests only show a very small effect of the delayed gas redistribution during the clad ballooning, such an effect is predicted to be significant in the full-scale rods - with a power peak located sufficiently away from the plenum - resulting in a considerable delay of the predicted moment of cladding rupture.     

Steady state thermal-hydraulic analysis of hydride-fueled grid-supported BWRs

A steady state thermal-hydraulic analysis was performed to estimate the power density attainable with hydride-fueled boiling water reactor (BWR) cores with respect to that of an existing oxide BWR core chosen as reference. The power-limiting constraints taken into account were the minimum critical power ratio (MCPR), core pressure drop, fuel average and centerline temperature, cladding outer temperature, flow-induced vibrations and power/flow ratio.The study consisted of two independent analyses: a whole core analysis and a single bundle analysis. The whole core analysis was performed, with a fixed core volume, for both hydride and oxide fuel over hundreds of combinations of rod diameter-rod pitch, referred to as “geometries”, in the ranges 0.6 ≤ D ≤ 1.6 cm and 1.1 ≤ P/D ≤ 1.6. For each geometry, the maximum achievable steady state core power was calculated. Preliminary neutronics results derived from a companion neutronic study were then overlaid on the whole-core thermal-hydraulic results to estimate the reduction in maximum achievable power caused by the application of neutronic constraints. The single bundle analysis was performed to compare in greater detail the thermal-hydraulic performance of a limited number of hydride and oxide fuel bundles having D and P values similar to those of the reference oxide bundle, and for which the compliance with neutronic constraints was demonstrated in a companion neutronic study.The study concluded that, if the core pressure drop is not allowed to increase above the reference core value, the power density increase attainable with hydride fuel is estimated to be in the range 0-15%. If the pressure drop is allowed to increase up to a value 50% higher than the reference core value, the power density increase is estimated to be in the range 25-45%. These power density increases, which are defined with respect to the reference oxide core, decrease about 10% if the comparison is made with respect to oxide designs resembling the most recent commercial high-performance oxide cores.The power gain capability of hydride fuel is primarily due to the possibility of: (1) replacing volumes occupied by water rods and water gaps in oxide fuel cores with fuel rods, thus increasing the heat transfer area per core volume, and (2) flattening the bundle pin-by-pin power distribution.The actual achievement of the above-mentioned power density increase is however conditioned to the compliance of hydride-fueled cores to safety requirements related to core behavior during transients, hydrodynamic stability and steam dryer performance, which are fields of study not addressed in this work. A potential 25-45% power density increase justifies however interest for further investigation on this alternative fuel.     

Experimental and calculation results of the integral reflood test QUENCH-15 with ZIRLO™ cladding tubes in comparison with results of previous QUENCH tests

The QUENCH-15 experiment investigated the effect of ZIRLO™¹ cladding material on bundle oxidation and core reflood, in comparison with tests QUENCH-06 (standard Zircaloy-4), QUENCH-12 (VVER, E110), and QUENCH-14 (M5^®). The QUENCH-15 bundle cross-section corresponded to a Westinghouse PWR core design and consisted of 24 heated rods (internal tungsten heaters between 0 and 1024 mm axial elevation, cladding oxidised region between −470 and 1500 mm), six corner rods made of Zircaloy-4, two corner rods made of E110, and a Zirconium 702 shroud. The test was conducted in principle with the same protocol as QUENCH-06, -12 and -14, so that the effects of the change of cladding material and bundle geometry could be more easily observed. The test protocol involved pre-oxidation to a maximum of about 150 μm oxide thickness at a temperature of about 1473 K over a period of about 3000s. The power was then ramped at a rate of 0.25 W/s/rod to cause a temperature increase until the desired maximum bundle temperature of 2153 K was reached. The maximum oxide layer thickness observed was 380 μm. Then reflood with 1.3 g/s/rod water at room temperature was initiated. The electrical power was reduced to 175 W/rod during the reflood phase, approximating effective decay heat level. The post-test metallography of the bundle showed neither noticeable breakaway oxidation of the cladding nor melt release into space between rods. The average outer oxide layer thickness at hottest elevation of 950 mm was 620 μm (QUENCH-06: 630 μm). The molten cladding metal at hottest elevation was localised between the outer and inner oxide layers. The thickness of inner oxide layer reaches 20% of that of the outer oxide layer. The measured hydrogen release during the QUENCH-15 test was 41 g in the pre-oxidation and transient phases and 7 g in the quench phase which are comparable with those in QUENCH-06, i.e. 32 g and 4 g, respectively. Post-test calculations were performed using a version of SCDAP/RELAP5/MOD3.2. The calculation results support the heuristic observation that there was no major difference between the influence of Zircaloy-4, M5^® or ZIRLO™ for the beyond-design basis accident present conditions here studied.     

The study on a statistical methodology for PWR fuel rod internal pressure evaluation

The most limiting design criteria for high Burnup PWR fuel are known to be rod internal pressure and cladding oxidation. Some fuel vendors have been increasing the design margin of rod internal pressure by increasing fuel rod plenum volume or optimizing fuel pellet grain size. In this study, a sophisticated statistical methodology that employs the response surface method and Monte Carlo simulation has been proposed to increase the design margin of rod internal pressure and subsequently a simplified statistical methodology has been developed to simplify the sophisticated statistical methodology. The simplified statistical methodology utilizes the system moment method combined with a deterministic approach for calculating a maximum variance of rod internal pressure. This simplified statistical methodology may be more efficient in the reload core fuel rod performance analyses than the sophisticated statistical methodology since the former eliminates numerous calculations needed for the evaluation of power history-dependent variances. It is found that this simplified methodology also generates more conservative rod internal pressure than the typical statistical methodology.     

International Solvent Extraction Conference ISEC ′90

The supercritical-water-cooled power reactor (SCPR) is expected to reduce power costs compared with those of current LWRs because of its high thermal efficiency and simple reactor system. The high thermal efficiency is obtained by supercritical pressure water cooling. The fuel cladding surface temperature increases locally due to a synergistic effect from the increased coolant temperature, the expanded flow deflection due to coolant density change and the decreased heat transfer coefficient, if the coolant flow distribution is non-uniform in the fuel assembly. Therefore, the SCPR fuel assembly is designed using a subchannel analysis code based on the SILFEED code for BWRs.

The SCPR fuel assembly has many square-shaped water rods. The fuel rods are arranged around these water rods. The fuel rod pitch and diameter are 11.2 mm and 10.2 mm, respectively. Since coolant flow distribution in the fuel assembly strongly depends on the gap width between the fuel rod and the water rod, the proper gap width is examined. Subchannel analysis shows that the coolant flow distribution becomes uniform when the gap width is 1.0 mm. The maximum fuel cladding surface temperature is lower than 600°C and the temperature margin of the fuel cladding is increased in the design.     

Inter fuel-assembly thermal-hydraulics for the ELSY square open reactor core design

The lead-cooled reactor is one of the six proposed innovative reactor types by the Generation IV International Forum (GIF). In Europe, the lead-cooled reactor design is known as the European Lead-cooled System (ELSY), which is a 600 MWe medium size fast reactor. The reference design of the ELSY core foresees square open (wrapper-less) fuel-assemblies with a staggered arrangement. In this design, the fuel rods in a fuel-assembly are separated by 3.4 mm. The gap between fuel rods of neighboring fuel-assemblies is 5.5 mm. In other words, the reference gap size between fuel-assemblies is larger than the gap between fuel rods within a fuel-assembly.This article discusses the involved inter fuel-assembly thermal-hydraulics between neighboring fuel-assemblies in the ELSY core. For this purpose as a starting point a validated Reynolds Averaged Navier Stokes (RANS)-based Computational Fluid Dynamics (CFD) approach is adopted. Moreover, bare fuel rods are considered in the present analyses that serve as a step towards inclusion of a spacer grid when its design is fixed. As the next step, the fuel-assemblies are numerically arranged with different gap sizes of 2.1 mm and 3.4 mm in order to analyze the influence of gap size on the inter fuel-assembly thermal-hydraulics. As a final step, analyses on the influence of different power levels of neighboring fuel-assemblies in the ELSY core are presented based on the reference ELSY core design. These inter-fuel assembly thermal hydraulic analyses lead to a conservative Nusselt number correlation for calculating maximum surface temperature of bare fuel rods that are located in the gap region between neighboring fuel-assemblies having different power levels. Such correlations, when implemented, will improve the applicability of system codes.     

Design and analysis of HYPER

KAERI (Korea Atomic Energy Research Institute) has been developing an accelerator driven transmutation system called HYPER (hybrid power extraction reactor). It is designed to transmute long-lived TRU and fission products such as Tc-99 and I-129. HYPER is a 1000 MW_th system with k_eff = 0.98 which requires 17 mA proton beam for an operation at EOC (end of cycle). Pb–Bi is used as the coolant and target material at the same time. HYPER core has 186 ductless hexagonal fuel assemblies. The fuel blanket is divided into three TRU (transuranic elements) enrichment zones to flatten the radial power distribution. The core height of HYPER was compromised at 150 cm, and the power density was determined such that the average coolant speed could be about 1.64 m/s. The inlet and exit coolant temperatures are 340 and 490 °C, respectively, in the core. The cylindrical beam tube and spherical window is adopted as the basic window design of HYPER. We have also introduced an Lead–Bismuth eutectic injection tube to maximize the allowable proton beam current. A metallic alloy of U-TRU-Zr is considered as the HYPER fuel, in which pure lead is used as the bonding material. As a result, a large gas plenum is placed above the active core. TRU transmutation rate is 282 kg/yr. In the case of a FP transmutation, 28.0 kg of Tc-99 and 7.0 kg of I-129 are incinerated per year. The MACSIS-H (metal fuel performance analysis code for simulating the in-reactor behavior under steady-state conditions-HYPER) for an metallic fuel was developed as the steady-state performance computer code. The MATRA (multichannel analyzer for transient and steady-state in rod array) code was used to perform the thermal-hydraulic analysis of HYPER core.     

Calculation of the power distribution in the fuel rods of the low power research reactor using the MCNP4C code

The Monte Carlo method, using the MCNP4C code, was used in this paper to calculate the power distribution in 3-D geometry in the fuel rods of the Syrian Miniature Neutron Source Reactor (MNSR). To normalize the MCNP4C result to the steady state nominal thermal power, the appropriate scaling factor was defined to calculate the power distribution precisely. The maximum power of the individual rod was found in the fuel ring number 2 and was found to be 105 W. The minimum power was found in the fuel ring number 9 and was 79.9 W. The total power in the total fuel rods was 30.9 kW. This result agrees very well with nominal power reported in the reactor safety analysis report which equals 30 kW. Finally, the peak power factors, which are defined as the ratios between the maximum to the average and the maximum to the minimum powers were calculated to be 1.18 and 1.31 respectively.     

Failure of high burnup fuels under reactivity-initiated accident conditions

Pulse irradiation experiments of high burnup light-water-reactor fuels were performed to assess the fuel failure limit in a postulated reactivity-initiated accident (RIA). A BWR-UO₂ rod at a burnup of 69 GW d/t failed due to pellet-cladding mechanical interaction (PCMI) in the test LS-1. The fuel enthalpy at which fuel failure occurred was comparable to those for PWR-UO₂ rods of 71 to 77 GW d/t with more corroded cladding. Comparison of cladding metallographs between the BWR and PWR fuel rods suggested that the morphology of hydride precipitation, which depends on the cladding texture, affects the fuel failure limit. The tests BZ-1 and BZ-2 with PWR-MOX rods of 48 and 59 GW d/t, respectively, also resulted in PCMI failure. The fuel enthalpies at failure were consistent with a tendency formed by the previous test results with UO₂ fuel rods, if the failure enthalpy is plotted as a function of the cladding outer oxide thickness. Therefore, the PCMI failure limit under RIA conditions depends on the cladding corrosion states including oxidation and hydride precipitation, and the same failure limit is applicable to UO₂ and MOX fuels below 59 GW d/t.     

Oxide thickness-dependent transient cladding hoop stress

A few thrice-burned Zry-4 fuel assemblies which were loaded in one of the PWRs operating in Korea were found to be failed due to PCI during a power ramp following a rector trip, while thrice-burned Zr–Nb fuel assemblies and twice-burned Zry-4 ones were intact. To investigate the effect of fuel rod oxide thickness on power ramp-induced cladding hoop stress, three intact fuel rods were selected, which include an intact twice-burned Zry-4 fuel rod, an intact thrice-burned Zr-4 fuel rod and an intact thrice-burned Zr–Nb fuel rod. With the use of a fuel performance analysis code, burnup-dependent steady-state cladding stress and ramp power-dependent cladding stresses at the power-ramped burnup were predicted for the three intact fuel rods. It was found that the cladding oxide thickness has a considerable effect on the ramp power-dependent cladding hoop stresses. In addition, the cladding maximum stress of the thrice-burned Zry-4 fuel rod with 125 μm oxide thickness exceeded an ultimate cladding tensile strength of the Zry-4 cladding when the pellet–clad friction coefficient-dependent cladding stress concentration ratio was considered. However, the thrice-burned Zr–Nb fuel rod with 50 μm oxide thickness was evaluated to have a considerable margin against the power ramp-induced PCI failure.     

Monte Carlo simulation of a research reactor with nominal power of 7 MW to design new control safety rods

The Monte Carlo simulation has been established for a research reactor with nominal power of 7 MW. A detailed model of the reactor core was employed including standard and control fuel elements, reflectors, irradiation channels, control rods, reactor pool and thermal column. The following physical parameters of reactor core were calculated for the present LEU core: core reactivity (ρ), control rod (CR) worth, thermal and epithermal neutron flux distributions, shutdown margin and delayed neutron fraction. Reduction of unfavorable effects of blockage probability of control safety rod (CSR)s in their interiors because of not enough space in their sites, and lack of suitable capabilities to fabricate very thin plates for CSR cladding, is the main aim of the present study. Making the absorber rod thinner and CSR cladding thicker by introducing a better blackness absorbing material and a new stainless steel alloy, respectively, are two studied ways to reduce the effects of mentioned problems.     

Possible design improvements and a high power density fuel design for integral type small modular pressurized water reactors

The study evaluates potential weaknesses and possible improvements for integral type small modular pressurized water reactor designs. By taking International Reactor Innovative and Secure (IRIS) as the reference design and keeping the power output as the same, a new fuel and reactor design were proposed. The proposed design relocates the primary coolant pumps and the pressurizer outside the reactor pressure vessel (RPV). Three recirculation lines and jet pumps/centrifugal pumps are introduced to provide the coolant circulation similar to Boiling Water Reactor designs. The pressurizer component is expected to be similar to the AP600 design. It is located at one of the recirculation lines. The new fuel assembly adopts 264 solid cylindrical fuel pins with 10 mm diameter and 2.3 m height, arranged at a hexagonal tight lattice configuration. Large water rods are introduced to preserve the moderating power and to accommodate finger type control rods. The resulting fuel can operate with 104.5 kW/l power density while having substantially higher margin for boiling crisis compared to typical large PWRs. Full core neutronic analysis shows that 24-month cycle length and 50 MWd/kg burnup is achievable with a two-batch refueling scheme. Furthermore, the fuel behavior study shows that the new fuel with M5 type Zircaloy cladding show fairly acceptable steady state performance. A preliminary Loss of Coolant analysis shows that the new design could be advantageous over IRIS due to its low ratio of the water inventory below the top of the active fuel to total RPV water inventory. The proposed reactor pressure vessel height and the containment volume are 30% lower than the reference IRIS design.     

Fast flux irradiation tests performed at high temperature

The first gas-cooled fast breeder reactor (GCFR) fast flux irradiation experiment [F-1(X094)] consists of seven fuel rods clad in 20% cold-worked 316 stainless steel. The rods are individually encapsuled, with sodium filling the gaps within the capsule walls. The rods are fueled with (15% Pu, 85% U)O₂ and have depleted UO₂ lower and upper axial blankets and charcoal to trap volatile fission products. The cladding i.d. temperature range covered by these rods is 570–760°C (1055–1400°F).The in-reactor performance of the fuel rods in the F-1 high-temperature experiment, which achieved a burnup of 121 MWd/kg (13.0 at.%) on the lead rod, is described. All rods in the experiment have remained intact. The results of interim examinations [at 25 and 50 MWd/kg (2.7 and 5.4 at.%)] of fuel and fission product behavior and transport and comparisons of observed results with LIFE-III code predictions are described.The F-3 experiment, which consists of ten encapsulated GCFR fuel rods with surface-roughened (ribbed) cladding, shares a nineteen capsule subassembly with Argonne National Laboratory. Temperatures are controlled over the range 675°C (1250°F) to 750°C (1380°F). Irradiation is in the core region of the EBR-II and thus permits achievement of a higher fluence-to-burnup ratio than that obtained in the F-1 experiment.Preliminary results of a planned interim examination at an exposure of 46 MWd/kg (4.9 at.%) burnup and a fluence of 5.2 × 10²² n/cm² show that cladding failures occurred in nine of the ten rods. Preliminary indications are that the failures are due to defects in the sodium bond between the fuel rod and the capsule.The tests completed and currently under way have been scoping in nature, and irradiation in EBR-II of GCFR prototypical fuel (pressure equalized) rods with ribbed cladding is required to provide the information needed for reactor design on effects of exposure to high fluence and burnup and on design reliability for a statistically significant number of rods. The design and the operating conditions for the F-5 experiment being prepared for this purpose are described.