Determination of Heats of Formation for Solid and Liquid Lithium Metasilicates by Vaporization Method

In-reactor experiments were performed in Nuclear Safety Research Reactor of Japan Atomic Energy Research Institute to study the failure behavior of stainless steel clad fuel rods under a simulated reactivity initiated accident (RIA) condition. A single test fuel rod with stainless steel cladding was contained in a capsule filled with water at room temperature and atmospheric pressure and irradiated by pulsing power simulating an RIA. It was revealed through the experiments that the failure mechanism of the stainless steel clad fuel rod was cladding melting, which was different from oxygen-induced embrittlement observed in the Zircaloy clad fuel rod in the same test condition, and the failure threshold energy was determined to be about 240cal/g·UO₂ (–1,000 kJ/kg·UO₂), which was about 20 cal/g·UO₂ (–85 kJ/kg·UO₂) lower than that of the Zircaloy clad fuel rod. It was also found that the mechanical energy was generated by explosive vaporization of coolant due to molten fuel-coolant interaction as a consequence of the fuel rod failure accompanying fuel pellet fragmentation at an energy deposition of nearly 380 cal/g·UO₂ (–1,600 kJ/kg·UO₂) or more.     

Effects of Fuel Oxidation and Dissolution on Volatile Fission Product Release under Severe Accident Conditions

The release of volatile fission products from high-burnup UO₂ fuel was examined in a steam atmosphere under severe accident conditions as a part of the VEGA program. The effects of fuel oxidation and dissolution were totally evaluated, by comparing the results with those from previous inert, hydrogen and steam atmosphere tests. It was shown that the oxidation of UO₂ to UO_2+x by steam generally enhances Cs and Kr release. However, the enhancement becomes smaller above the melting temperature of Zircaloy, about 2030 K, likely due to reduction of UO_2+x by molten Zircaloy. The burst release of Cs occurs above about 2300K in the hydrogen atmosphere, while the release rate does not increase so significantly for the examined temperature range (<2800 K) in the steam atmosphere. Analysis of the hydrogen atmosphere test showed that fuel dissolution is apparently connected with the burst release and that a large fraction of Cs is quickly released from the dissolved fuel above 2300 K. It is considered that the fuel dissolution rate in the steam atmosphere is about 1/1000 of that in the hydrogen atmosphere.     

Molten fuel-coolant interaction during a reactivity initiated accident experiment

The results of a reactivity-initiated accident experiment, designated RIA-ST-4, are discussed and analyzed with regard to molten fuel-coolant interaction (MFCI). In this experiment, extensive amounts of molten UO₂ fuel and zircaloy cladding were produced and fragmented upon mixing with the coolant. Coolant pressurization up to 35 MPa and coolant overheating in excess of 940 K occurred after fuel rod failure. The initial coolant conditions were similar to those in boiling water reactors during a hot startup (that is, coolant pressure of 6.45 MPa, coolant temperature of 538 K, and coolant flow rate of 85 cm³/s). It is concluded that the high coolant pressure recorded in the RIA-ST-4 experiment was caused by an MFCI and was not due to gas release from the test rod at failure, Zr/water reaction, of UO₂ fuel vapor pressure. The high coolant temperature indicated the presence of superheated steam, which may have formed during the expansion of the working fluid back to the initial coolant pressure; yet, the thermal-to-mechanical energy conversion ratio is estimated to be only about 0.3%.     

Synopsis and outcome of the QUENCH experimental program

The paper gives an overview of the main outcome of the QUENCH program launched in 1997 at the Karlsruhe Institute of Technology (KIT), formerly Karlsruhe Research Center (FZK). The research program comprises bundle experiments as well as complementary separate-effects tests. The focus of the experiments performed from 1997 to 2009 was on scenarios of severe accidents whereas that of the future test program will be on large-break loss-of-coolant accidents (LOCA) in the frame of design-basis accidents, and debris coolability, in the frame of severe accidents. The major objective of the program is to deliver experimental and analytical data to support the development and validation of quench and quench-related models as used in code systems that model severe accident progression in light water reactors.So far, 15 integral bundle QUENCH experiments with 21-31 electrically heated fuel rod simulators of 2.5 m length have been conducted. The following parameters and their influence on bundle degradation and reflood have been investigated: degree of pre-oxidation, temperature at initiation of reflood, flooding rate, influence of neutron absorber materials (B₄C, AgInCd), air ingress, and influence of the type of cladding alloy.In six tests, reflooding of the bundle led to a temporary temperature excursion driven by runaway oxidation of zirconium alloy components and resulting in release of a significant amount of hydrogen, typically two orders of magnitude greater than in those tests with “successful” quenching in which cool-down was rapidly achieved. Considerable formation, relocation, and oxidation of melt were observed in all tests with escalation. The temperature boundary between rapid cool-down and temperature escalation was typically in the range of 2100-2200 K in the “normal” quench tests, i.e. in tests without absorber and/or steam starvation. Tests with absorber and/or steam starvation were found to lead to temperature escalations at lower temperatures.All phenomena occurring in the bundle tests have been investigated additionally in parametric and more systematic separate-effects tests. Oxidation kinetics of various cladding alloys, including advanced ones, have been determined over a wide temperature range (873-1773 K) in different atmospheres (steam, oxygen, air, and their mixtures). Hydrogen absorption by different zirconium alloys was investigated in detail, recently also using neutron radiography as non-destructive method for determination of hydrogen distribution in claddings. Furthermore, degradation mechanisms of absorber rods including B₄C and AgInCd as well as the oxidation of the resulting low-temperature melts have been studied. Steam starvation was found to cause deterioration of the protective oxide scale by thinning and chemical reduction.The most recent topic of the QUENCH program has been investigation of the behavior of advanced cladding materials (ACM) in comparison with the classical Zircaloy-4. Although separate-effects tests have shown some differences in oxidation kinetics, the influence of the various cladding alloys on the integral bundle behavior during oxidation and reflooding was only limited.     

Reflooding experiments with LWR-type fuel rod simulators in the QUENCH facility

The QUENCH experiments at the Karlsruhe Research Center are carried out to investigate the hydrogen generated during reflooding of an uncovered Light Water Reactor (LWR) core. The QUENCH test bundle is made up of 21 fuel rod simulators approximately 2.5 m long. The Zircaloy-4 rod cladding is identical to that used in PWRs (Pressurized Water Reactors) with respect to material and dimensions. Pellets are made of zirconia to simulate UO₂. After a transient phase with a heating rate of 0.5–1 K s⁻¹ water of approx. 395 K is admitted from the bottom when the test bundle has reached its pre-determined temperature. Except for the flooding (quenching) phase, the QUENCH test phases are conducted in an argon/steam atmosphere at 3 g s⁻¹ each. The results of the first two experiments, QUENCH-01 (with pre-oxidation of 300 μm oxide layer thickness at the cladding outside surface) and QUENCH-02 (reference test without pre-oxidation), are compared in the paper. The pre-oxidized LWR fuel rod simulators of QUENCH-01 were quenched from a maximum temperature of 1870 K. In the second bundle experiment, QUENCH-02, quenching started at 2500 K. Pre-oxidation apparently prevented a temperature escalation in the QUENCH-01 test bundle, while the QUENCH-02 test bundle experienced a temperature excursion which started in the transient phase and lasted throughout the flooding phase. The different behavior of the two experiments is also reflected in hydrogen generation. While the bulk of H₂ was produced during pre-oxidation of test QUENCH-01 (30 g), the largest amount, i.e. 170 g, of hydrogen was generated during the reflooding phase of test QUENCH-02, at a maximum production rate of 2.5 g s⁻¹ as compared to 0.08 g s⁻¹ in test QUENCH-01. Similarities between the two experiments exist in the thermo-hydraulics during the quench phase, e.g. in the cooling behavior, the quench temperatures, and quench velocities.     

Pellet cladding interaction — Evaluation of lubrication by graphite

Graphite CANLUB interlayers between UO₂ pellets and Zircaloy cladding increase the tolerance of fuel pins to power ramping defects. This paper evaluates the effectiveness of graphite as a lubricant in dry, low oxygen potential environments such as exist at the UO₂ — Zircaloy interface of an operating fuel pin. Two types of laboratory tests were performed at 573 ± 5 K in moist air and also dry inert gas. In the first test series, graphite-coated Zircaloy cladding was expanded by the outward movement of a 4-segment bronze mandrel. Depleting the oxygen and water vapour reduced circumferential failure strains from about 19% to about 12%, but there remained a large margin of improvement over the 4% failure strain attained with bare Zircaloy. In the second series of laboratory tests a disc of UO₂ was clamped between Zircaloy plates, and the force required to cause it to slide was measured as a function of normal loading with and without graphite lubrication. The coefficient of friction μ for graphite lubricating a Zircaloy-UO₂ interface rose only slightly from 0.19 to 0.24 on changing the gaseous environment from laboratory air to dry deoxygenated helium. These values were low compared with μ ~ 0.9 for bare Zircaloy-UO₂ interfaces in air or inert gas. An indirect evaluation was made of whether or not graphite acts as a lubricant in-service by counting peripheral radial UO₂ cracks in metallographic cross-sections of irradiated fuel pins. The number of peripheral cracks in highly powered fuel depends on the interfacial shear stress at the pellet-clad interface, and hence on the friction coefficient. In fuel pins ramped to a linear power of 55 kW/m there were, on average, 17 peripheral UO₂ cracks in graphite lubricated CANLUB fuel compared with 30 in bare Zircaloy-clad fuel. We deduce that the ratio of friction coefficients in-service (CANLUB graphite : bare Zircaloy) had been 0.57 : 1.0. Finally, we discuss theoretical limitations of CAN-LUB coatings and their implications regarding fuel design and performance.     

Zircaloy-U02 and -Water Reactions and Cladding Temperature Estimation for Rapidly-Heated Fuel Rods under an RIA Condition

Zircaloy cladding chemical reactions with coolant water and UO₂ fuel at elevated temperatures under a reactivity initiated accident (RIA) condition were studied from a metallurgical point of view on the basis of the nuclear safety research reactor (NSRR) experiments. The cladding-fuel chemical reaction was extensively analyzed and found to be explainable from equilibrium phase diagrams. The systematic estimation methods of maximum cladding temperature were proposed and examined from metallographies. Maximum cladding temperature can be estimated from measured oxidation thicknesses in the temperature range of 1,000~1,600°C, from melting microstructures in the range of 1,600~1,950°C and also from the volume fraction of the precipitates, (U, Zr)0_2-x, in once-molten oxygen-stabilized α-zircaloy in 1,950~2,400°C. The estimation by the method proposed in the paper is more valid than thermocouple indications at high temperatures, since thermocouples perturb the temperatures they are measuring or fail at the extremely high temperatures. The results are thought to be applicable also to understand general fuel rod behavior under hypothetical accident conditions which cause severe fuel damage.     

Effects of UO2/Zircaloy-4 Mole Ratios on Reaction of UO2 Fuel with Molten Zircaloy-4

Dissolution of UO₂ crucibles by molten Zircaloy-4 (Zry) was investigated in the temperature range of 2,223-2,373 K and for specimens having UO₂/Zry mole ratios between 7 and 18.2. The uranium concentration in the Zry melt rapidly increased during initial reaction time and approached saturated values, depending on reaction temperature and UO₂/Zry mole ratio. Kinetics of uranium concentration increase in the melt was analyzed based on a natural convection mass transfer model that takes into account the change of contact surface area/melt volume ratio with reaction time. The saturated uranium concentration in the Zry melt was inversely proportional to the U0₂/Zry mole ratio. An empirical correlation of saturated uranium concentration in the Zry melt was obtained as a function of UO₂/Zry mole ratios and reaction temperature. This study of the empirical correlation was intended to estimate maximum UO₂ fuel dissolution by molten Zry cladding during severe fuel damage accidents for three different reactor type fuels.     

AgInCd control rod failure in the QUENCH-13 bundle test

The QUENCH off-pile experiments performed at the Karlsruhe Research Center are to investigate the high-temperature behavior of Light Water Reactor (LWR) core materials under transient conditions and in particular the hydrogen source term resulting from the water injection into an uncovered LWR core. The typical LWR-type QUENCH test bundle, which is electrically heated, consists of 21 fuel rod simulators with a total length of approximately 2.5 m. The Zircaloy-4 rod claddings and the grid spacers are identical to those used in Pressurized Water Reactors (PWR) whereas the fuel is represented by ZrO₂ pellets. In the QUENCH-13 experiment the single unheated fuel rod simulator in the center of the test bundle was replaced by a PWR-type control rod. The QUENCH-13 experiment consisting of pre-oxidation, transient, and quench water injection at the bottom of the test section investigated the effect of an AgInCd/stainless steel/Zircaloy-4 control rod assembly on early-phase bundle degradation and on reflood behavior. Furthermore, in the frame of the EU 6th Framework Network of Excellence SARNET, release and transport of aerosols of a failed absorber rod were to be studied in QUENCH-13, which was accomplished with help of aerosol measurements performed by PSI–Switzerland and AEKI–Hungary.Control rod failure was initiated by eutectic interaction of steel cladding and Zircaloy-4 guide tube and was indicated at about 1415 K by axial peak absorber and bundle temperature responses and additionally by the on-line aerosol monitoring system. Significant releases of aerosols and melt relocation from the control rod were observed at an axial peak bundle temperature of 1650 K. At a maximum bundle temperature of 1820 K reflood from the bottom was initiated with cold water at a flooding rate of 52 g/s. There was no noticeable temperature escalation during quenching. This corresponds to the small amount of about 1 g in hydrogen production during the quench phase (compared to 42 g of H₂ during the pre-reflood phases). Posttest examinations of bundle structures revealed the presence of only little relocated AgInCd melt in the form of rivulets, mainly in the coolant channels surrounding the control rod simulator and at axial elevations between the third (0.55 m) and first spacer grids (−0.1 m).Results of QUENCH-13 on the onset of absorber rod failure are in agreement with CORA results of nine experiments each containing one or more AgInCd/stainless steel/Zircaloy-4 control rod assemblies. Bundle degradation triggered by early melt formation was, however, more pronounced in the CORA experiments with maximum bundle temperatures of 2300 K (compared to 1800 K in QUENCH-13). Consequently, QUENCH-13 allowed studying the initiation of absorber rod failure by eutectic reactions of SS-Zr, and later on of AgInCd-Zr, as well as the redistribution of the absorber material within the test bundle. Furthermore, input data for modeling of aerosol release during severe accidents are considered as benefits of the experiment.     

Zircaloy-4 Cladding Embrittlement due to Inner Surface Oxidation under Simulated Loss-of-Coolant Condition

Embrittlement of Zircaloy-4 cladding by oxidation of the inner surface occurring in an LWR loss-of-coolant accident was studied using simulated fuel containing of A1₂O₃ pellets sheathed in Zircaloy-4 specimen cladding, filled with Ar gas, and sealed. This simulated fuel rod was heated from outside until the isothermal oxidation temperature between 880 and 1,167°C was obtained after the cladding burst. This exposed the inner surface of the cladding to the environmental atmosphere, provided by steam flowing at a constant rate in the range of 0.13–1.6 g/cm²-min.

The embrittlement of the specimen due to inner surface oxidation is influenced primarily by the amount of hydrogen absorbed by the Zircaloy-4. Ring compression tests conducted at 100°C on test pieces constituted of sliced sections of oxidized specimen showed that Zircaloy containing more than 200–300 wt.ppm of absorbed hydrogen became brittle when oxidized at temperatures above 1,000°C. In the range of oxidation temperature 932 to 972°C, brittleness did not appear below 500–750 wt.ppm absorbed hydrogen.

Hydrogen absorbed by the Zircaloy precipitated in the form of fine hydride crystals formed along previous β-phase grain boundaries. Peaks were found in the distribution of hydrogen absorbed on the inner surface, at a distance of 15–45 mm upward and downward of the rupture opening. Within this range, the distance was influenced by the oxidation temperature and steam flow rate.     

Multiphysics coupled modeling of light water reactor fuel performance

A fuel performance code for light water reactors called CityU Advanced Multiphysics Nuclear Fuels Performance with User-defined Simulations (CAMPUS) was developed. The CAMPUS code considers heat generation and conduction, oxygen diffusion, thermal expansion, elastic strain, densification, fission product swelling, grain growth, fission gas production and release, gap heat transfer, mechanical contact, gap/plenum pressure with plenum volume, fuel thermal and irradiation creep, cladding thermal and irradiation creep and oxidation. All the equations are implemented into the COMSOL Multiphysics finite-element platform with a 2D axisymmetric geometry of a fuel pellet with cladding. Comparisons of critical fuel performance parameters for UO₂ fuel using CAMPUS are similar to those obtained from BISON, ABAQUS and FRAPCON. Additional comparisons of beryllium doped fuel (UO₂-10%volBeO) with silicon carbide, instead of Zircaloy as cladding, also indicate good agreement. The capabilities of the CAMPUS code were further demonstrated by simulating the performance of oxide (UO₂), composite (UO₂-10%volBeO), silicide (U₃Si₂) and mixed oxide ((Th_0.9,U_0.1)O₂) fuel types under normal operation conditions. Compared to UO₂, it was found that the UO₂-10%volBeO fuel experiences lower temperatures and fission gas release while producing similar cladding strain. The U₃Si₂ fuel has the earliest gap closure and induces the highest cladding hoop stress. Finally, the (Th_0.9,U_0.1)O₂ fuel is predicted to produce the lowest fission gas release and a lower fuel centerline temperature when compared with the UO₂ fuel. These tests demonstrate that CAMPUS (using the COMSOL platform) is a practical tool for modeling LWR fuel performance.     

Effect of Fuel Pellet Eccentricity with Cladding on Fuel Rod Thermal Behavior under Reactivity Initiated Accident Condition

Fuel rod behavior under Reactivity Initiated Accident (RIA) conditions has been studied in the Nuclear Safety Research Reactor (NSRR), JAERI. In the experiments, cladding thermal behavior was observed to be influenced by the fuel pellet eccentricity to produce large azimuthal temperature variation in the cladding. The maximum azimuthal cladding temperature difference was measured to be as large as 150°C by thermocouples attached to opposite sides of the cladding around the circumference, though the thermocouples did not always detect the maximum temperature difference around the circumference. The actual temperature differences in the fuel rods subjected to less than 290 cal/g?UO₂ were estimated to be 350°C at maximum based on metallographies. A simple calculation considering gap conductance variations also showed that the maximum temperature difference became 350°C under fully eccentrical condition in the fuel rod subjected to 260 cal/g?UO₂. Moreover, as the rod damage such as cladding deformation, melting and failure occurs unevenly around the circumference due to the fuel pellet eccentricity in general, the fuel pellet eccentricity should influence the fuel rod failure under RIA conditions.     

Rod Overpressure/Lift-off Testing at Halden In-pile Data and Analysis

The cladding lift-off experiments at Halden yield direct data for the maximum pressure to which a rod can be operated without causing a lasting fuel temperature increase. UO₂ or MOX fuel segments irradiated to high burnup in light water reactors are equipped with a fuel thermocouple and a cladding extensometer. Gas lines attached to the end plugs are connected to a high pressure system for pressurisation with argon and a low pressure system for hydraulic diameter measurements to study cladding outward deformation and axial gas communication within the fuel rod.

The first experiment of the test series utilised a UO₂ fuel segment irradiated in an LWR to 52 MWd/kgUO₂. The test was operated for 4,400 h PWR conditions (155 bar, 310°C) provided by a loop system. The rod was pressurised starting at 205 bar and increasing to 455 bar in steps of 50 bar, while recording fuel centreline temperature and cladding elongation. The hold times at the different pressure levels were long enough to assess temperature trends.

The measured rates of fuel temperature increase suggest that the necessary overpressure to cause a discernible lasting temperature change was 130–145 bar, equivalent to a cladding hoop stress of 70–77 MPa.     

Candu type fuel behavior during rapid overpower transients

Results obtained in the pulse irradiation tests performed on segmented fuel elements in the Romanian Annular Core Pulse Reactor (ACPR) are discussed below. Tests included the effects of initial element internal pressure and a wide range of energy deposition on the fuel element behavior. All tests were conducted in stagnant water at room temperature and atmospheric pressure inside the capsule. The fuel elements were instrumented with thermocouples for cladding surface temperature measurement. Transient histories of reactor power, cooling water pressure, fuel element internal pressure and cladding temperature were recorded during the tests. The fuel elements were subjected to total energy depositions from 70 to 265 cal g⁻¹ UO₂. Cladding failure mechanism and the failure threshold have been established. The fuel failure mechanism is a burst type and is very similar to LOCA failure mechanism even though the rate energy deposition is higher in the ACPR tests. At higher energy deposition brittle cladding fracture near endcap weld region can be produced. The failure threshold is situated between 190 and 200 cal g⁻¹ UO₂ for standard fuel rod (0.2–0.3 MPa internal pressure) and less than 160 cal g⁻¹ UO₂ for pressurized fuel rods (internal pressure between 1 and 3.0 MPa). Pre-pressurization could be an important factor to control the failure threshold energy. The experimental program is still in progress and new experiments are foreseen to be performed in the following period.     

Materials chemistry and transport modeling for severe accident analyses in light-water reactors II: Gap processes and heat release

Several consequences of steam starvation of the gas filling the internals of the core of a light-water reactor in the fuel-uncovery phase of a severe accident up to cladding melting are analysed. Emphasis is placed on processes that occur in the H₂-rich gas external to the fuel rod cladding; absorption of oxygen and hydrogen by the cladding; the composition and flow rates of gas in the fuel-cladding gap; and the response of the fuel to these conditions. The transport processes and chemical reactions in the cladding, and the fuel controlled by the behavior of the gas in the gap are modeled for a simple temperature transient characteristic of a severe fuel damage accident in a light-water reactor. Cladding burst is assumed to occur at 1273 K at the midplane elevation of the fuel rod, permitting the gas in the gap to come into contact with that external to the fuel rod. The results of the analysis include the following. Steam ingress is restricted to a few centimeters from the failure site by the gettering action of the metal-water reaction on the cladding inner wall. Hydrogen moves axially into the gap only a few times further than steam by diffusion in the Xe-He mixture. The chief process restricting H₂ ingress is the backflow resulting from thermal expansion of the gas in the fuel rod as the temperature rises. When the protective ZrO₂ scale on the outer surface of the cladding disappears by dissolution in the metal, hydrogen permeation through the cladding wall rapidly replaces the inert gas in the gap with H₂. Hydrogen uptake by the cladding draws gas into the core region from the upper plenum and augments the heat release by the metal-water reaction. Exposure of the fuel to this H₂-rich gas results in minor fuel reduction and accompanying cladding oxidation.     

Molten fuel radial motion and cladding melting during a PCM event in LWRs

The thermal behavior of the fuel and cladding during off-normal operating conditions, generally termed power-cooling-mismatch (PCM), are of great interest to light water reactor (LWR) safety analysis. During a power-cooling-mismatch event, fuel melting may begin at the center of the rods propagating radially outward. The induced pressure at the center of the rod due to fuel melting, fission gas release, and UO₂ fuel vapor would tend to force such molten fuel to flow through radially open cracks in the outer unmelted portion of the pellet and relocate in the fuel-cladding gap. The zircaloy cladding, which is at high temperature during film boiling, may undergo melting at its inside surface upon being contacted by the extruded molten fuel, eventually resulting in a thermal failure of the cladding.Three topics of interest are analyzed in this paper. First, fuel conditions during a hypothesized PCM accident are assessed with regard to pellet cracking and central fuel melting. Secondly, the transient freezing of a superheated liquid penetrating an initially empty crack, maintained at constant subfreezing temperatures, is studied analytically. The analysis is presented in a dimensionless form, illustrating the effect of the governing parameters, namely the driving pressure, crack shape (that is, a divergent, a parallel wall, or a convergent crack), density ratio, Stefan number for freezing, and steady state crust thickness. The calculational results are used to assess the radial extrusion of molten UO₂ fuel observed in some in-pile tests, in which PCM conditions in a pressurized water reactor were simulated. Thirdly, conditions for potential melting of zircaloy cladding upon being contacted by the extruded molten fuel are investigated analytically. The analytical predictions were consistent with the experimental results from PCM in-pile tests.     

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.     

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.     

Effects of Rod Pre-Pressurization on Light Water Reactor Fuel Behavior during Reactivity Initiated Accident Conditions

Previously pressurized (pre-pressurized) fuel rod tests recently performed in the Nuclear Safety Research Reactor (NSRR) investigate the effects of initial internal pressure on fuel rod behavior during reactivity initiated accident (RIA) conditions. A single PWR type fuel rod was contained within a waterfilled, ambient temperature and ambient pressure capsule. The fuel rod was then heated by the pulsing operation of the NSRR.

Results from the tests show that the effect of pre-pressurization was significant for the fuel rods with initial internal pressure of 0.8 MPa and above, and fuel rod failure occurred from rupture of the cladding with lower threshold energy deposition for failure as the initial internal pressure was increased. The cladding rupture was governed mainly by the cladding temperature rise, not by the rod internal pressure rise during the transient. The relationships between cladding burst pressure and cladding burst temperature and between cladding strain and cladding temperature at cladding rupture obtained in the present study under an RIA condition agree with the results obtained from various in- and ex-reactor experiments under a LOCA condition, although the obtained time-averaged strain rate of the Zircaloy cladding was much greater than that in a LOCA condition.     

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.