Accident Tolerant Fuel Concepts Research Articles

Hydrothermal corrosion of PVD and cold spray Cr-coatings on Zircaloy-4 in hydrogenated and oxygenated LWR coolant environments

Accident tolerant fuel (ATF) are currently being developed as part of an international effort to modify or replace zirconium-based cladding materials in light water reactors. The main objective of ATF cladding materials is to improve fuel reliability and safety during accident scenarios. Thus, these materials should perform comparable to or better than the current zirconium alloy. One of the ATF concepts explores modifying standard Zircaloy cladding by coating their outer surface. In this study, surface-modified Zr-4 is produced by depositing a protective coating of chromium through two different coating techniques, Physical Vapor Deposition (PVD) and Cold Spray (CS). The coated specimens were tested at ∼300 °C in out-of-pile autoclaves simulating Boiling Water Reactor (BWR) conditions. Results show that after one month of immersion, Cr-coated Zr-4 evidenced a small mass gain in hydrogenated environments. However, both types of coated specimens showed significant coating degradation and mass loss in oxygenated environments.

Effect of the redox potential on the general corrosion behavior of industrial nuclear alloys

During the last decade, the international nuclear materials community has been investigating a robust fuel version for Light Water Reactors called Accident Tolerant Fuel (ATF). One of the ATF concepts is to replace the current Zr tubing with an iron-chromium-aluminum (FeCrAl) alloy or IronClad. Extensive out-of-pile autoclave testing showed that the dissolution (corrosion) of the FeCrAl alloys in ∼300 °C water is highly dependent on the chemistry of the alloy and on the chemistry of the water. The dissolution rate of FeCrAl and other industrial Cr containing alloys is higher in hydrogen environments than in oxygen environments, since under oxygen conditions they form a more protective surface oxide. Alloys with higher Cr content (e.g., APMT) are more resistant to dissolution than lower Cr alloys. Studies also showed that the presence of a few parts per billion Zn considerably reduced the dissolution rate of FeCrAl.

Effect of ATF Cr-coated-Zircaloy on BWR In-vessel Accident Progression during a Station Blackout

The deposition of protective coatings on nuclear fuel cladding has been considered as a near-term Accident Tolerant Fuel (ATF) concept that will reduce the high-temperature oxidation rate and enhance accident tolerance of the cladding while providing additional benefits during normal and transient. In this study, the performance of the proposed ATF concept of Cr-coated-Zircaloy is assessed using a generic Boiling Water Reactor MELCOR plant model considering a Short-term Station Blackout (STSBO) scenario. Simulation results indicate that the use of Cr-coated-Zircaloy as cladding and canister material mitigates the core degradation process as compared to the traditional Zircaloy cladding and canister design. The onset of fuel degradation and collapse is delayed by over thirty minutes, and the extent of fuel degradation is reduced. Specifically, the gross in-vessel hydrogen generation decreased by almost a factor of three. Although the eutectic reaction between Cr-coating and Zircaloy could cause an early failure of the coating, the improvement in the delay of fuel degradation is still notable. Additionally, a thicker coating is found helpful to obtain additional coping time and to decrease hydrogen generation. In addition to the eutectic formation that may compromise Cr-coated Zr, a different failure mode is identified for the Cr-coated-Zr when compared to Zircaloy; i.e., a complete melt of base material leads to component collapse before the coating is oxidized and consumed. These findings can help the industry focus on productive areas of research and development for accident-tolerant fuel concepts and enhancement of core safety margins.

UN microspheres embedded in UO2 matrix: An innovative accident tolerant fuel

Uranium nitride (UN)-uranium dioxide (UO2) composite fuels are being considered as an accident tolerant fuel (ATF) option for light water reactors. However, the complexity related to the chemical interactions between UN and UO2 during sintering is still an open problem. Moreover, there is a lack of knowledge regarding the influence of the sintering parameters on the amount and morphology of the α-U2N3 phase formed. In this study, a detailed investigation of the interaction between UN and UO2 is provided and a formation mechanism for the resulting α-U2N3 phase is proposed. Coupled with these analyses, an innovative ATF concept was investigated: UN microspheres and UO2.13 powder were mixed and subsequently sintered by spark plasma sintering. Different temperatures, pressures, times and cooling rates were evaluated. The pellets were characterised by complementary techniques, including XRD, DSC, and SEM-EDS/WDS/EBSD. The UN and UO2 interaction is driven by O diffusion into the UN phase and N diffusion in the opposite direction, forming a long-range solid solution in the UO2 matrix, that can be described as UO2-xNx. The cooling process decreases the N solubility in UO2-xNx, causing then N redistribution and precipitation as α-U2N3 phase along and inside the UO2 grains. This precipitation mechanism occurs at temperatures between 1273 K and 973 K on cooling, following specific crystallographic grain orientation patterns.

Quantification of the effect of Cr-coated-Zircaloy cladding during a short term station black out

Accident Tolerant Fuel (ATF) has the potential to improve the inherent safety of current reactors. In particular, an ATF Cr-coated Zirconium alloy (Cr-coated-Zr) cladding can be an effective short-term approach to improve the cladding oxidation resistance without changing the cladding base materials. In order to appropriately assess the effect of such an ATF concept it is important to quantify the performance of such a cladding improvement not only during normal operation but also during postulated accidents. In this work, we examine the effect of Cr-coated-Zr cladding (Cr-Zr) during a short-term station blackout (STSBO) accident as well as quantify the uncertainty in the use of ATF Cr-coated-Zr cladding. This is accomplished using a coated cladding model developed for the MELCOR severe accident systems code; i.e., MELCOR-1.8.6-UDGC model. First, we quantify the effect of the coated cladding using a set of base case simulations that show the improved oxidation resistance of the Cr-coated-Zr cladding with the UDGC model parameters. Then, we quantify the uncertainty in these simulations by a Monte-Carlo analysis that examines the effect of the ATF clad input parameters (e.g., oxidation rate and failure temperature) on key output parameters; i.e., timing of rapid hydrogen generation, clad temperatures and hot leg creep rupture. Finally, we use regression analysis methods to estimate the importance of these input parameters. These analyses show that the Cr-coated-Zr cladding can delay rapid clad oxidation and generation of hydrogen by 3600 s to 7200 s with associated delays in rapid clad temperature increases. This additional time can be of benefit for compensatory operator actions. In contrast there is little effect on the timing of hot leg creep rupture since this failure is mainly controlled by core decay heat. We also find that these results are not significantly affected by the uncertainty in input parameters, and based on the regression analysis, the high temperature Zr oxidation rate has the major influence on the selected output parameters for the Cr-coated-Zr cladding.

Integrating a multi-layer deformation model in FRAPTRAN for accident Tolerant fuel analysis

Fuel performance analysis capability is needed to assess proposed Accident Tolerant Fuel (ATF) designs. In this work, a deformation model was implemented in FRAPTRAN 2.0 to model a multi-layer ATF concept during both reactivity-initiated accident (RIA) and loss-of-coolant accident (LOCA). The modified FRAPTRAN (M−FRAPTRAN) performance was assessed against limited UO2-Zircaloy validation database and the performance was found to be comparable to FRAPTRAN 2.0 and experimental data. Finally, in order to demonstrate the added capabilities in M−FRAPTRAN, two-layer SiC clad design consisting of SiC/SiC ceramic matrix composite (CMC) and monolithic SiC (mSiC) was modeled. During RIA, the onset of pellet clad mechanical interaction can significantly raise the mSiC principal stresses. During LOCA, the SiC potential annealing behavior (recovery of swelling strain) has a significant influence on cladding stress distribution and outer region of mSiC cladding may undergo tensile stress, which can threaten the leak tightness of the cladding during LOCA.

Reactor physics analysis of a new Accident Tolerant Fuel called Fuel-in-Fibers

Fuel-in-Fibers (F-in-F) is an Accident Tolerant Fuel (ATF) concept aimed at significantly increasing the fission product retention capability of nuclear fuel at high temperatures, similar to another ATF concept, Fully Ceramic Microencapsulated Fuel (FCM). FCM, however, when implemented in a conventional Pressurized Water Reactor (PWR) fuel geometry, is not able to maintain an 18-month fuel cycle length and remain below the proliferation enrichment limit of 20 w/o 235U using the standard cycle reload strategy. The F-in-F enrichment is decreased by putting the fuel in a cylindrical fiber instead of a spherical particle in order to increase the volume fraction of fuel in the pellet. The fibers are created with a direct manufacturing process that allows for the minimization of the volume occupied by the SiC matrix and the maximization of uranium loading via direct deposition of high density fuels like uranium nitride (UN). The full core reactor physics simulation showed that when the F-in-F design used UN as a fuel, it only required enrichments of 6% to maintain an 18-month fuel cycle length. Based on a full core analysis when compared to the current reference fuel for a standard 4-loop Westinghouse PWR, F-in-F has a similar steady state thermal hydraulic and neutronic performance.

Assessment of Neutronic Characteristics of Accident-Tolerant Fuel and Claddings for CANDU Reactors

The objective of this study was to evaluate accident-tolerant fuel (ATF) concepts being considered for CANDU reactors. Several concepts, including uranium dioxide/silicon carbide (UO2-SiC) composite fuel, dense fuels, microencapsulated fuels, and ATF cladding, were modelled in Serpent 2 to obtain reactor physics parameters, including important feedback parameters such as coolant void reactivity and fuel temperature coefficient. In addition, fuel heat transfer was modelled, and a simple accident model was tested on several ATF cases to compare with UO2. Overall, several concepts would require enrichment of uranium to avoid significant burnup penalties, particularly uranium-molybdenum (U-Mo) and fully ceramic microencapsulated (FCM) fuels. In addition, none of the fuel types have a significant advantage over UO2 in terms of overall accident response or coping time, though U-9Mo fuel melts significantly sooner due to its low melting point. Instead, the different ATF concepts appear to have more modest advantages, such as reduced fission product release upon cladding failure, or reduced hydrogen generation, though a proper risk assessment would be required to determine the magnitude of these advantages to weigh against economic disadvantages. The use of uranium nitride (UN) enriched in N15 would increase exit burnup for natural uranium, providing a possible economic advantage depending on fuel manufacturing costs.

Metrics for the Technical Performance Evaluation of Light Water Reactor Accident-Tolerant Fuel

The safe, reliable, and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry’s success. Enhancing the accident tolerance of light water reactors (LWRs) became a topic of serious discussion following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal for the development of accident-tolerant fuel (ATF) for LWRs is to identify alternative fuel system technologies to further enhance the safety, competitiveness, and economics of commercial nuclear power. Designed for use in the current fleet of commercial LWRs or in reactor concepts with design certifications (GEN-III+), fuels with enhanced accident tolerance would endure loss of active cooling in the reactor core for a considerably longer period of time than the current fuel system while maintaining or improving performance during normal operations. The complex multiphysics behavior of LWR nuclear fuel in the integrated reactor system makes defining specific material or design improvements difficult; as such, establishing desirable performance attributes is critical in guiding the design and development of fuels and cladding with enhanced accident tolerance. Research and development of ATF in the United States is conducted under the U.S. Department of Energy (DOE) Fuel Cycle Research and Development Advanced Fuels Campaign. The DOE is sponsoring multiple teams to develop ATF concepts within multiple national laboratories, universities, and the nuclear industry. Concepts under investigation offer both evolutionary and revolutionary changes to the current nuclear fuel system. This paper summarizes the technical evaluation methodology proposed in the United States to aid in the optimization and prioritization of candidate ATF designs.

3 차원 프린팅 기술을 이용한 신개념 경수로 핵연료 기술 개발에 관한 연구

To enhance the safety of nuclear reactors after the Fukushima accident, researchers are developing various types of accident tolerant fuel (ATF) to increase the coping time and reduce the generation of hydrogen by oxidation. Coated cladding, an ATF concept, can be a promising technology in view of its commercialization. We applied 3D printing technology to the fabrication of coated cladding as well as of coated pellets. Direct metal tooling (DMT) in 3D printing technologies can create a coated layer on the tubular cladding surface, which maintains stability during corrosion, creep, and wear in the reactor. A 3D laser coating apparatus was built, and parameter studies were carried out. To coat pellets with erbium using this apparatus, we undertook preliminary experiments involving metal pellets. The adhesion test showed that the coated layer can be maintained at near fracture strength.

Critical processes and parameters in the development of accident tolerant fuels drop-in capsule irradiation tests

The goal of the Accident Tolerant Fuel (ATF) program is to develop the next generation of Light Water Reactor (LWR) fuels with improved performance, reliability, and safety characteristics during normal operations and accident conditions and with reduced waste generation. An irradiation test series has been defined to assess the performance of proposed ATF concepts under normal LWR operating conditions. The Phase I ATF irradiation test series is planned to be performed as a series of drop-in capsule tests to be irradiated in the Advanced Test Reactor (ATR) operated by the Idaho National Laboratory (INL).Design, analysis, and fabrication processes for ATR drop-in capsule experiment preparation are presented in this paper to demonstrate the importance of special design considerations, parameter sensitivity analysis, and precise fabrication and inspection techniques for figure innovative materials used in ATF experiment assemblies. A Taylor Series Method sensitivity analysis approach was used to identify the most critical variables in cladding and rodlet stress, temperature, and pressure calculations for design analyses. The results showed that internal rodlet pressure calculations are most sensitive to the fission gas release rate uncertainty while temperature calculations are most sensitive to cladding I.D. and O.D. dimensional uncertainty. The analysis showed that stress calculations are most sensitive to rodlet internal pressure uncertainties, however the results also indicated that the inside radius, outside radius, and internal pressure were all magnified as they propagate through the stress equation.This paper demonstrates the importance for ATF concept development teams to provide the fabricators as much information as possible about the material properties and behavior observed in prototype testing, mock-up fabrication and assembly, and chemical and mechanical testing of the materials that may have been performed in the concept development phase. Special handling, machining, welding, and inspection of materials, if known, should also be communicated to the experiment fabrication and inspection team.