The performance of defected spent LWR fuel rods in inert and dry air storage atmospheres

A testing program using eight commercial PWR and BWR spent fuel rods was conducted to investigate their long-term stability under a variety of possible dry storage conditions. The objective of this project is to provide the Nuclear Regulatory Commission (NRC) with the information to confirm or establish spent-fuel, dry storage licensing positions regarding long-term, low-temperature ( <523 K) spent fuel rod behavior during dry storage, and for radioactive contamination arising from spallation of cladding crud. Until now, the testing program has included three interim nondestructive examinations and one destructive examination. This paper presents the results of the third examination conducted to determine any degradation in eight fuel rods after being subjected to 13168 h at temperature. During this examination, visual observations, diametrical measurements, and isotopic analysis of smears were used to assess the fuel rod behavior and particulate release.     

Thermal-fluid flow analysis and demonstration test of a spent fuel storage system

Thermal-fluid flow analysis and demonstration test were performed for a spent fuel storage system. The commercial computational fluid dynamics (CFD) code, FLUENT was used for the numerical analysis. Effective thermal conductivities of a spent fuel assembly and a fuel basket were derived to optimize a thermal analysis model. Also, a porous model, which can simplify a complex configuration of a fuel assembly, was used in the thermal analysis. Demonstration test were performed to verify the thermal analysis method and procedure using a half scaled-down model and an electrically heated dummy fuel. The numerical analysis results were compared with the experimental data. Thermal analyses of the storage system were carried out for normal and off-normal conditions by using the verified analysis method.     

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.     

Power ramp induced iodine and cesium redistribution in LWR fuel rods

Volatile fission product migration in LWR fuel rods which are power ramped above a certain threshold beyond the envelope of their previous power history, plays an important role in stress corrosion cracking of Zircaloy. This may cause fuel rods to fail already at stresses below the yield strength. In the HFR, Petten, many power ramp experiments have been performed with subsequent examination of the ramped rods for fission product distribution. This study describes the measurement of iodine and cesium distribution using γ-spectroscopy of I-131 and Cs-137. An evaluation method is presented which makes the determination of absolute amounts of I/Cs feasible. It is shown that a threshold for I/Cs redistribution exists beyond which it depends strongly on local fuel rod power and fuel type.     

Evaluation of LWR fuel rod behavior under operational transient conditions

To evaluate the effects of fission gas flow and diffusion in the fuel-cladding gap on fuel rod thermal and mechanical behaviors in light water reactor (LWR) fuel rods under operational transient conditions, computer sub-programs which can calculate the gas flow and diffusion have been developed and integrated into the LWR fuel rod performance code BEAF. This integrated code also calculates transient temperature distribution in the fuel-pellet and cladding.The integrated code was applied to an analysis of Inter Ramp Project data, which showed that by taking into account the gas flow and diffusion effects, the calculated cladding damage indices predicted for the failed rods in the ramp test were consistent with iodine-SCC (Stress Corrosion Cracking) failure conditions which were obtained from out-of-reactor pressurized tube experiments with irradiated Zircaloy claddings. This consistency was not seen if the gas flow and diffusion effects were neglected. Evaluation were also made for the BWR 8 × 8 RJ fuel rod temperatures under power ramp conditions.     

Modeling of gap conductance for LWR fuel rods applied in the BISON code

ABSTRACT

A new gap conductance model is proposed in this study as a combination of Toptan’s model and the Ross-Stoute model. A variance-based sensitivity analysis is performed to understand how simulation results depend on all input parameters of the proposed model. Additionally, new modeling options (e.g. fill gas thermal conductivity, temperature jump distance, thermal accommodation coefficient, etc.) are added into the nuclear fuel performance code, BISON. The need for further investigation of the gap heat transfer between fuel and cladding in BISON motivated this study to evaluate its impact on the code’s predictions. New gap conductance modeling is proposed. A series of integral-effects validation tests is performed: (1) to demonstrate the impact of the proposed model on the code’s fuel temperature predictions at the beginning of life and through the reactor’s life; (2) to ensure that the proposed model is capable of accurately modeling gap heat transfer characteristics in real-world problems; and (3) to investigate the impact of the estimation of fission gas release on the fuel temperature predictions with the proposed model. The results indicate that the proposed gap conductance model improves BISON’s predictions.     

Fuel oxidation and thermal conductivity model for operating defective fuel rods

A model has been developed to describe the fuel oxidation behaviour, and its influence on the fuel thermal conductivity, in operating defective nuclear fuel rods. The fuel-oxidation model is derived from adsorption theory and considers the influence of the high-pressure environment that results from coolant entry into the fuel-to-clad gap. This model is in agreement with the fuel-oxidation kinetics observed in high-temperature annealing experiments conducted at 1473-1623 K in steam over a range of pressure from 0.001 to 0.1 MPa. Using a Freundlich adsorption isotherm, the current model is also consistent with recent experiments conducted at a higher pressure of 7 MPa. The model also considers radiolytic effects as a consequence of fission fragment bombardment in the fuel-to-clad gap. This treatment suggests that radiolysis-assisted oxidation is insignificant in operating defective rods (as compared to thermal effects), as supported by limited in-reactor data. The effects of diffusion of the interstitial oxygen ions in the solid in the operating rod is further discussed.     

Escape of fission products from defective fuel rods of light water reactors

The escape behaviour of various fission product isotopes from defective fuel rods in PWRs and BWRs is analyzed.Diffusion in the UO₂ is the rate controlling step for the release of noble gases from defective fuel rods. The escape of fission iodine from defective fuel rods is controlled by a mechanism which includes migration and additional delay steps, probably in the nature of a chemical reaction.The inferred effective diffusion constants for fission gases are noticeably higher for defective fuel rods than for intact fuel rods. The difference is about two orders of magnitude. The enhancement of diffusion in defective fuel rods is believed to be due to the increase in the -ratio of the UO₂ in the defective fuel rods.     

Delayed hydride cracking of spent fuel rods in dry storage

Failures of zirconium alloy cladding tubes during a long-term storage at room temperature were first reported by Simpson and Ells in 1974, which remains unresolved by the old delayed hydride cracking (DHC) models. Using our new DHC model, we examined failures of cladding tubes after their storage at room temperature. Stress-induced hydride phase transformation from γ to δ at a crack tip creates a difference in hydrogen concentration between the bulk region and the crack tip due to a higher hydrogen solubility of the γ-hydride, which is a driving force for DHC at low temperatures. Accounting for our new DHC model and the failures of zirconium alloy cladding tubes during long-term storage at room temperature, we suggest that the spent fuel rods to be stored either in an isothermal condition or in a slow cooling condition would fail by DHC during their dry storage upon cooling to below 180 °C. Further works are recommended to establish DHC failure criterion for the spent fuel rods that are being stored in dry storage.     

Evaluation of implementation of thorium fuel cycle with LWR and MSR

In order to construct a sustainable society, it is necessary to consider fairness beyond generations and between countries. It is expected that Asian countries continue growing their economy and will result consuming more energy. More CO₂ emission is not acceptable.Nuclear power has many advantages for reducing CO₂ emission. However, it still has concerns of nuclear proliferation, radioactive waste and safety. It is necessary to overcome these concerns if nuclear power is expanded to Asian countries. Thorium utilization as nuclear fuel will be an opening key of these difficulties because thorium produces less plutonium, less radioactive waste. Safety will also be enhanced. The use of molten-salt reactor (MSR) triggered by plutonium supply from ordinary light water reactor (LWR) with uranium fuel will allow implementation of thorium fuel cycle with electricity capacity of about 446 GWe around at 2050.The other important sector in a view of sustainability is transportation. Transportation is essential for economy growth. Therefore it is inevitable to reduce CO₂ emission from transportation sector. Electric vehicle (EV) will be used as a major mobility instead of gasoline engine cars. Rare-earth materials such as neodymium and dysprosium are necessary for producing EV. These materials are expected to be mined from Asian countries. It is often obtained with thorium as by-product. Thorium has not been used as nuclear fuel because it is not good for nuclear weapon and it does not have fissionable isotopes. Recent global trend of nuclear disarmament and accumulation of plutonium from uranium fuel cycle can support starting the use of thorium.Thorium utilization will help both to provide clean energy and to produce rare-earth for clean vehicle. These will create new industries in developing Asian countries. An international collaborative framework can be established by supplying resource from developing countries and supplying technology from developed countries. “THE Bank (THorium Energy Bank)” is proposed here as one part of such a framework.     

Effects gap heat transfer on LWR fuel behavior during an RIA transient: In-pile experimental results with helium and xenon fillet rods

Gap heat transfer characteristics and their effects on LWR fuel behavior during an RIA have been studied through the in-pile experiment with UO₂ pellet fuel rods. The report describes the experimental results obtained in the NSRR tests in which PWR type test fuel rods of helium and xenon filled as the gap gas have been irradiated in the pulse reactor, NSRR, to simulate the prompt heat up of RIAs. The relation between the cladding temperature history and the gap heat transfer conditions, and the effects of gap gas composition on fuel behavior and on the fuel failure threshold are discussed based on the in-pile experimental data.     

Radiological aspects and behavior of spent fuel considering long-term storage

For spent nuclear fuel management in Germany, the concept of dry interim storage in dual purpose casks before direct disposal is applied. Current operation licenses for storage facilities have been granted for a storage time of 40 years. Due to the current delay in site selection, an extension of the storage time seems inevitable. In consideration of this issue, GRS performed burnup calculations, thermal and mechanical analyses as well as particle transport and shielding calculations for UO₂ and MOX fuels stored in a cask to investigate long-term behavior of the spent fuel related parameters and the radiological consequences. It is shown that at the beginning of the dry storage period, cladding hoop stress levels sufficient to cause hydride reorientation could be present in fuel rods with a burnup higher than 55 GWd/tHM. The long-term behavior of the cladding temperatures indicates the possibility of reaching the ductile-to-brittle transition temperature during extended storage scenarios. Surface dose rates are 3 times higher when a cask is partially loaded with 4 MOX fuel assemblies. Due to radioactive decay, long-term storage will have a positive impact on the radiological environment around the cask.     

Void coefficient analysis of an LWR lattice experiment with high-enrichment MOX fuel

A series of physics experiments performed between 1985 and 1990 in the PROTEUS reactor at the Paul Scherrer Institute in Switzerland included the investigation of a Pu-fuelled light water reactor (LWR) lattice with a moderator-to-fuel volume ratio of 2.07 and an effective enrichment of about 8%. The analysis of the measurements in this test lattice and in the tighter light water high conversion reactor (LWHCR) lattices investigated previously permits the determination of the k_∞ void coefficient of the LWR lattice for cases of partial and total voiding. A comparison of the measured changes of k_∞ with values calculated using the cell codes WIMS and KAPER4 shows a satisfactory prediction of the partial void coefficient in the range from 0–54% voidage. Discrepancies increase up to twice the estimated experimental error in cases of further voiding to 100% void. The total void coefficient (0–100% void) results from large compensating effects from individual reaction rate ratios. Its accurate prediction by cell calculations appears to be fortuitous. Improved nuclear data and refined calculational methods are thus required for a more accurate calculation of the void coefficient in high-enrichment MOX-LWRs.     

Mathematical modeling and analysis of fuel rods

This paper presents a comprehensive method of analysis for nuclear fuel rods by first developing appropriate constitutive relations for the fuel and the clad materials that take into account elastic-plastic behavior, primary as well as secondary creep, mechanical densification (hot pressing), fuel swelling, fuel cracking and healing, and fuel redistribution. The geometric problem is treated on the basis of a two-dimensional, axisymmetric, and plane model within the framework of the finite element method. General power histories of the transient, cyclic, and steady-state types are considered. Analysis results uncovering the importance of certain deformation processes, which heretofore have been neglected, and comparisons with experiments are given.     

SCALE calculation for evaluation of spent fuel condition during long-term storage

Experiences with an advanced spent nuclear fuel management in Slovakia are presented in this paper. The evaluation and monitoring procedures are based on practices at the Slovak wet interim spent fuel storage facility in NPP Jaslovské Bohunice. Since 1999, leak testing of WWER-440 fuel assemblies were completed using a special leak tightness detection system developed by Framatome-anp, “Sipping in Pool”. This system utilized external heating for the precise defects determination.Optimal methods for spent fuel disposal and monitoring were designed. A new conservative factor for specifying of spent fuel leak tightness is introduced in the paper. Limit values of leak tightness were established from the combination of SCALE4.4a (ORIGEN-ARP) calculations and measurements from the “Sipping in Pool” system. These limit values are: limiting fuel cladding leak tightness coefficient for tight fuel assembly – k_FCT(T) = 3 × 10⁻¹⁰, limiting fuel cladding leak tightness coefficient for fuel assembly with leakage – k_FCT(L) = 8 × 10⁻⁷.     

UO2/Zry-4 chemical interaction layers for intact and leak PWR fuel rods

In this study, the UO₂ pellet-Zry-4 cladding interfaces of intact and leak PWR fuel rods were examined with the help of an optical microscope and a scanning electron microscope to investigate typical chemical interaction layers formed at the pellet-cladding interface during the normal reactor operations. The two intact and two leak fuel rods with the burnup of between 35,000 and 53,000 MWD/MTU were selected to evaluate the effects of gap-gas compositions and fuel burnup on the chemical interaction layer formation. Based on the optical and scanning electron micrographs, it is found that the intact fuel rod generates apparently one interaction layer of (U,Zr)O_2−x at the interface, whereas the leak fuel rod generates apparently two interaction layers of ZrO_2−x and (U,Zr)O_2−x. These interaction layers for the intact and leak fuel rods were predicted by several diffusion paths drawn on a U-Zr-O ternary phase diagram. The variations of chemical element compositions around the interface of one intact rod were generated by an electron probe micro-analyzer to confirm the interaction layers at the pellet-cladding interface. The interaction layer growth rates of the ZrO_2−x and (U,Zr)O_2−x phases were estimated, using the layer thicknesses and the reaction times.