Fuel Rod Cladding Research Articles

Spent LWR fuel dry storage in large transport and storage casks after extended burnup

Dry spent LWR fuel storage is licensed for single fuel assemblies with rod burnup to 65 GWd/tHM. This allows dry spent fuel storage of reloads with a batch average up to 55 GWd/tHM. The leading defect mechanism for spent fuel rods in dry storage is hoop strain. Fuel rod degradation can be prevented by limiting creep. Post-pile creep of fuel rod cladding can be described conservatively by the creep of unirradiated cladding. In order to extend the database, internally pressurized creep samples were investigated for time intervals up to 10 000 h. Test temperatures were between 250 and 400°C, and the hoop stresses applied ranged from 80 to 150 N/mm 2. The resulting data were described mathematically by an interpolation formula. Based on the fuel assemblies end-of-life data the maximum CASTOR V cask storage temperature was calculated to be between 348°C and 358°C at the beginning.

Estimating the tensile properties of an austenitic cladding tube steel following corrosion by uranium dioxide, simultated fission products and neutron exposure

As there seems to be a general lack of systematic data about mechanical properties of fuel rod cladding tubes following in-pile operations, the tensile properties (tensile strength, 0.2% offset yield strength, ultimate elongation and uniform elongation) were measured in specimens which had either only been corroded or irradiated, been both corroded and irradiated or were in the initial state. The tensile properties Z of the specimens which had been both corroded and irradiated were calculated from the tensile properties of the specimens which had only been corroded, those which had only been irradiated and those kept in their initial state, in accordance with this formula: Z( corroded + irradiated) = Z( corroded) · Z( irradiated) Z( initial state) The flat specimens (0.7 × 4 mm) were made of 1.4970 type austenitic cladding tube steel. As with breeder fuel rods, the corrosive medium used was a mix of hyperstoichiometric uranium dioxide and simulated fission products (Cs, I, Te). The strain rate, test temperature, depth of corrosion, simulated burnup, and state of the material parameters were varied systematically. The results of calculations were, on average, in good agreement with the measured results, although some measured values showed considerable variation due to corrosion.

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 H 2-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 H 2 ingress is the backflow resulting from thermal expansion of the gas in the fuel rod as the temperature rises. When the protective ZrO 2 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 H 2. 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 H 2-rich gas results in minor fuel reduction and accompanying cladding oxidation.

EFFECT OF BUOYANCY ON FORCED CONVECTION IN A CUSPED DUCT

ABSTRACT In the event of a loss of coolant accident in a pressurized water reactor, swelling of the fuel rod cladding will lead to reduction of the subchannel flow area and worsening of the core heat transfer in the region of the blockage. The four-cusped duct is an ideal geometry for the simulation of such a channel blockage. Understanding the characteristics of flow and heat transfer in the cusped duct is essential for better design of the emergency core cooling system. Thus, in this paper, combined natural and forced convection in a vertical cusped duct has been investigated in the region of both hydrodynamically and thermally fully developed flow. The thermal boundary condition imposed on the cusped duct is the axial uniform heat flux with peripheral uniform temperature. The results indicate that the fluid flow and heal transfer in the comer region of the cusped duct are improved because of the influence of natural convection. As the Rayleigh number increases, the friction factor and Nusselt number in...

Investigation of in-pile formed corrosion films on Zircaloy fuel rod claddings by impedance spectroscopy and galvanostatic anodization

Abstract Hot-cell investigations have been executed to study the corrosion behaviour of irradiated Zircaloy fuel rod claddings by impedance spectroscopy and galvanostatic anodization. The thickness of the compact oxide at the metal-oxide interface and the thickness of the minimum barrier oxide have been determined at different positions along the claddings. As shown by analysis, both quantities first increase and then decrease with increasing thickness of the total oxide.

Assessment of the residual time to rupture of fuel pins after reactor core disturbances using the Lebensanteil rule

Abstract An important aspect of disturbances in the reactor core is the way in which they affect the service life of fuel rod cladding tubes. This factor also determines whether and how long the reactor core can be continued in operation, i.e., matters of safety and economy are involved. Potential disturbances of the reactor core affect the fuel rod cladding tubes as increases in temperature and, sometimes, as mechanical stresses for limited periods of time. As thermomechanical stresses acting on a cladding tube also give rise to creep events which may limit the service life of fuel elements, it is important to know how much creep life or time to rupture is consumed in the course of a core disturbance, and what the residual life is. For this purpose, the times to rupture before and during the accident must be added up and the balance calculated. As a rule of computation, the Lebensanteil rule is used in its form expressing the time to rupture of creeping solids. The simulation of accidents with unirradiated cladding tubes revealed a drastic decrease of the residual time to rupture in those cases in which the cladding material had recrystallized. On the other hand, because of its structural stability, irradiated material turned out to be almost insensitive even under extremely severe accident conditions. The materials data so far available are sufficient for useful estimates. Presuming one of the damage accumulating processes of the creeping cladding material is predominant, there are no further validity limiting ranges concerning kind of accident, loading condition, cladding material and so on.

Kinetics of liquid phase formation due to solid/solid chemical interactions and its modeling; application to the Zircaloy/Inconel system

Abstract According to recent results of LWR bundle meltdown experiments the formation of liquid phases was observed as a consequence of chemical interactions between the various bundle components below their melting points. The reaction kinetics of the different material combinations in a LWR fuel rod bundle was studied. In this paper the chemical interactions between Inconel spacer grids and as-received and preoxidized Zircaloy (Zry) fuel rod claddings, respectively, are described. The experimental results between 1000 and 1200°C show that the interactions obey parabolic rate laws. The fundamental difference of the experiments with preoxidized Zry (oxide layer thickness ≦ 100 μm) compared to as-received Zry is the delay time for the start of the interactions between Inconel and Zircaloy. A model is presented to describe the experimental results of the solid Inconel/solid Zircaloy interaction kinetics. The simple relations obtained can be used as modules in SFD (severe fuel damage) code systems to describe the behavior of a LWR core during high temperature transients.

Eddy currents determine defects in fuel rod cladding : Atom, no. 365, pp. 13–15 (Mar. 1987)

Eddy currents determine defects in fuel rod cladding : Atom, no. 365, pp. 13–15 (Mar. 1987)

Eddy currents determine defects in fuel rod cladding Atom, no. 365, pp. 13–15 (Mar. 1987)

Eddy currents determine defects in fuel rod cladding Atom, no. 365, pp. 13–15 (Mar. 1987)

Examination of the destructive forces in the chernobyl accident based on NSRR experiments

The possible causes of the destruction of the Chernobyl reactor core were examined by making use of the Nuclear Safety Research Reactor (NSRR) experimental results concerning the destructive forces generated by a fuel failure. A complementary experiment with Chernobyl reactor conditions was performed in order to observe the fuel failure behavior and the resultant vessel pressure rise, etc. Also, generation of hydrogen from the fuel rod cladding and the consequent system pressure rise were estimated based on the experiments. These examinations led to the conclusion that the most probable cause of the core pressure tube rupture in the accident was a static pressure rise due to rapid energy release from fragmented fuel. Other phenomena such as the hydrogen generation and molten fuel contact to the tube wall might have contributed to the tube rupture. The water hammer force is also estimated to have been large enough to break tubes even using conservative assumptions.

Experimental Investigations on the Reflooding and Deformation Behavior of an Advanced Pressurized Water Reactor Tight-Lattice Fuel Rod Bundle in a Loss-of-Coolant Accident

The work performed in the FLORESTAN program at the Karlsruhe Nuclear Research Center on the reflooding and deformation behavior of a tight-lattice fuel rod bundle in a loss-of-coolant accident of an advanced pressurized water reactor (APWR) is described.The present forced-feed reflooding tests in an extremely tight bundle with a pitch-to-diameter ratio p/d = 1.06 show a very different thermal-hydraulic behavior compared to a standard pressurized water reactor. Blind code predictions have shown that most thermal-hydraulic computer codes do not adequately predict the reflooding behavior of this type of bundle.Deformation tests on stainless steel cladding tubes have shown that those with integral helical fins limit the burst strains and have high potential for APWR fuel rod cladding.

Pressurized water reactor blockage—prediction of laminar flow and temperature distributions following a loss of coolant accident

During the unlikely occurrence of a loss of coolant accident (LOCA) in a pressurized water reactor, it has been postulated that the fuel rod cladding may swell due to the combination of a pressure differential across the cladding and the increased temperature levels in the core. In this event, adjacent fuel rods may ‘balloon’ until they make contact with their neighbours, leading to a reduction in subchannel flow area, subchannels of highly noncircular cross-section, and worsening heat transfer in the blocked region of the core. A particularly important measure in predicting an upper limit on the severity of this core blockage is the peripheral variation of the wall temperature for a representative blocked subchannel. This paper is concerned with the prediction of the temperature variations in the duct wall and the fluid for fully developed laminar flow. A single subchannel is modelled as a four-cusped duct, bounded by a conducting wall of constant thickness, which is subject to uniform heat flux from the fuel. Results for this idealized problem are presented for different values of the thickness and thermal conductivity of the cladding. The important wall temperature distributions have been calculated for superheated steam to cover fluid flow conditions which might be envisaged during a LOCA. Here it must be observed that two-phase flow effects are unlikely to lead to worse heat transfer than can be predicted for single-phase steam cooling. Thus, the predicted temperature variations represent an upper bound for the low Reynolds number end of the Reflood phase in the LOCA.

The iodine-induced strain rate sensitivity of Zircaloy fuel rod cladding

The strain rate sensitivities and failure times characteristic of iodine-induced stress corrosion cracking of Zircaloy fuel rod cladding are important because they can be related to certain power ramping rates which may increase fuel rod failure probabilities. The CCSCC model developed in this paper approximates these failure characteristics by simulating the transition from the slower (less observable) non-corrosive creep cracking (CC) regime to the faster (more observable) stress corrosion cracking (SCC) regime. Components of the CCSCC model include: ZrI 4 production by chemical reaction between Zircaloy and iodine; diffusion of the ZrI 4 to the crack tip; chemisorption, embrittlement, and damage accumulation at the crack tip; and crack initiation times and growth rates. Results indicate that the Zr-I 2 SCC process is dominated by competition between chemical reaction and material creep rate phenomena, rather than by stress thresholds or by the requirement that a complete monolayer of ZrI 4 form on the exposed Zircaloy surface at the crack tip. Failure times are dominated by the time required to initiate active crack growth. The SCC process is apparently not limited by diffusion kinetics on the time scale of laboratory experiments. Conflicting results were found concerning the chemical reaction rate at the crack tip.

Optical techniques for nuclear reactor inspection

This paper reports an experimental setup developed to detect and quantify the deformations of the cladding surface of nuclear fuel pellets submitted to analytical transient conditions. It consists of an optical instrument based on the speckle interferometry technique, able to provide non-destructive measurements on scattering surfaces such as nuclear fuel rod claddings. Here are presented first results performed on-line and in situ, with micrometric resolution. Then, limitations of the system as well as further developments are discussed.

A method of predicting the temperature response of ballooning fuel rod cladding for PWR LOCA conditions

A method is described for calculating fuel rod cladding temperatures in a blockage formed by a group of ballooned fuel rods in a larger rod array, for heat transfer conditions appropriate to the reflooding phase of a postulated PWR LOCA. The model is suitable for describing the extreme case of co-planar axially extended balloons, where steam superheating and skin friction effects are believed to have an important effect on blockage heat removal. Attention is restricted to the constricted zone within the blockage. Reasonable agreement is shown with available heat transfer data from partially ballooned rod arrays, for conditions of steam cooling, steam-and-droplet cooling and reflood cooling. The model is also able to describe flow velocity distribution data from partially blocked rod bundles with reasonable accuracy. Parametric calculations for typical PWR LOCA heat transfer conditions suggest that blockage length has a strong effect on fuel coolability, mainly as a result of extra superheating of the steam within the blockage. However calculations also indicate that the presence of entrained water droplets has a powerful effect in reducing the clad temperatures attained.

Influence of Partial Blockage of a BWR Bundle on Heat Transfer, Cladding Temperature, and Quenching During Bottom Flooding or Top Spraying Under Simulated LOCA Conditions

In a test facility with two parallel boiling water reactor fuel assemblies, experiments were carried out with top spray and bottom flooding, simulating loss-of-coolant accident (LOCA) conditions. The flow area restriction, caused by the ballooning of fuel rod cladding within one of the bundles, was provided by blockage plates, which had reductions of 37% in one case and in a second series 70% of the flow area. Test parameters were system pressure (1, 5, and 10 bars), spray (0.68 and 1.02 m/sup 3//h) and flooding rates (1.5,2, and 3.3 cm/s), power input (520 and 614 kW), and the initial cladding temperature (600 and 800/sup 0/C at midplane) of the heaters. The test results showed no significant variations from those without blockage, except in the blocked region. An enhancement of heat transfer was observed in a close region downstream from the blockage in cases such as bottom flooding and top spray tests. The results will serve the purpose of code verification for reactor LOCA analysis.

Quenching and Rewetting of Nuclear Fuel Rods

Many postulated nuclear reactor accidents result in dryout or film boiling within the nuclear core. To minimize fuel rod damage and potential rod failure, safe or lower cladding fuel temperatures must be reestablished by encouraging coolant-cladding contact. This process is commonly referred to as quenching or rewetting, and these terms are often incorrectly assumed to be synonymous.Quench and rewet are distinctly different phenomena. Quench is the rapid cooling of a hot solid surface (fuel rod cladding) resulting from enhanced heat transfer conditions and does not necessitate liquid-solid contact. Rewet, however, implies direct liquid-solid contact and the establishment of a liquid-solid-vapor triple interface. The rewet temperature is normally lower than the quench temperature. Estimation of quench and rewet temperatures appears to be possible from first principles.

Three-Dimensional Thermal-Hydraulic Analysis of Wire-Wrapped Rods in Liquid-Metal Fast Breeder Reactor Core Assemblies

A study is presented to determine the detailed coolant velocity and temperature profile around the entire rod circumference in liquid-metal fast breeder reactor (LMFBR) core assemblies as well as the detailed radial and circumferential temperature profile in the rod. The digital computer code FATHOM-360 developed to perform the above calculations is described. Fuel, radial blanket, and control assembly rods (both wire-trapped and bare) can be analyzed. Coolant, cladding, and fuel (or absorber) temperature profiles are calculated for uniform and nonuniform heat generation (i.e., accounting for power skew across the pellet) in the rod. Temperature distributions can be calculated for both concentric and eccentric positions of the pellet with respect to the fuel rod cladding. Typical examples of the calculational capabilities of the code are presented. Such capabilities are needed for a reliable design of LMFBR core assemblies and rods to provide detailed cladding temperature profiles and accurately calculate the cladding strain on which the fuel rod lifetime and allowable burnup depend. Overall, a more realistic core thermofluids design is possible by implementing the study presented here.

Photomask array placement on slices in LSI processing : A. F. Tasch, Jr. and V. R. Porter. Microelectron & Reliab.12, 539 (1973)

The effects of coolant injection into a reactor vessel by the passive actuation of SITs (Safety Injection Tanks) on the in-vessel core melt progression under a severe accident have been evaluated in an APR (Advanced Power Reactor) 1400. A best estimate simulation from initiating events of 2-in. and 3-in. break SBLOCAs (Small Break Loss Coolant Accidents) without SI (Safety Injection) to a reactor vessel failure has been carried out using the SCDAP/RELAP5 computer code in conditions with and without the passive actuation of the SITs. The SCDAP/RELAP5 results have shown that the passive coolant injection into the reactor vessel by the actuation of the SITs leads to postpone the reactor vessel failure time by 4–5 h in the SBLOCA without SI. At the time of the reactor vessel failure, 44.5% and 42.8% of the fuel rod cladding were oxidized in the 2-in. and 3-in. SBLOCAs with the actuation of the SITs, respectively. However, 37.5% and 34.6% of the fuel rod cladding were oxidized in the 2-in. and the 3-in. SBLOCAs without the actuation of the SITs, respectively. Even though the SITs are designed for a large break LOCA, their actuation is very effective on the delay of a reactor vessel failure in spite of more hydrogen generation in the SBLOCA without SI.

Calculational procedure for determining creep collapse of LWR fuel rods

Observed collapses in pressurized water reactor fuel rods have been attributed to the radiation enhanced creep of Zircaloy cladding into regions where separations in the fuel pellet stack have occurred. A computer code, COLAPX, has been written to determine the growth of ovality and the ultimate collapse of fuel rod cladding under reactor operating conditions. This paper describes the theoretical bases of this code, the finite element formulation used, the constitutive relations between the displacement fields and the element forces, and the radiation, temperature and stress dependent material model for creep of Zircaloy tubing. Comparisons of the creep rate predictions and of the ovality predictions with data from irradiated tubes and fuel cladding are presented.