Commercial Light Water Reactors Research Articles

Simulation of nuclear fuel clad high-temperature ballooning under loss-of-coolant accident conditions considering anisotropic creep

Zirconium (Zr) alloys are standard clad material in commercial light water reactors. However, it is vulnerable to ballooning and burst under high-temperature conditions such as LOCA (Loss-Of-Coolant Accident), due to large internal/external pressure difference. Considering that clad ballooning is detrimental in terms of nuclear reactor safety especially for high-burnup fuel rod, high-fidelity numerical simulation of Zircaloy clad ballooning is desirable. This paper presented a finite-element based implementation for the simulation of Zircaloy clad ballooning. The developed simulation capability predicts progressive clad ballooning up to failure given empirical burst limits. The material models coupled together are phase transformation and high-temperature creep. Especially, anisotropic creep was considered for α phase Zircaloy. The implemented algorithm for stress update is implicit in time integration, which allows relatively large time step size to efficiently simulate the process that can last for hundreds of seconds. An analytic benchmark, the PUZRY ballooning and burst test, and the Hardy stress-rupture test were utilized to verify and validate the simulation capability. The anisotropic creep calculation conforms to the analytical solution. The comparison with PUZRY and Hardy test data shows that considering anisotropic creep in the α phase region is essential. 3D ballooning was also simulated by considering circumferential temperature difference for both conventional and novel annular clad designs. Interestingly, it is the first time to note that the risk of buckling failure is predicted for the double-cooled annular nuclear fuel clad during the accident condition, only after considering the anisotropic creep algorithm.

Transport equivalent full core neutronics calculations via finite volume method and optimized diffusion parameters

In this study, three-dimensional neutronics analysis of the hot zero power (HZP) situation of the BEAVRS full core benchmark problem is performed by employing a finite volume method (FVM), Krylov subspace methods, and transport equivalent diffusion model (TEDM). TEDM is used for generating transport-corrected diffusion parameters and we use a genetic algorithm for the correction. The study shows that a transport equivalent and accurate neutronics analysis of a highly detailed commercial light water reactor (LWR) core is feasible using only the two-group diffusion theory. The core model in the benchmark is represented by 165 different diffusion parameter sets since the model is highly heterogeneous. The reaction rates, flux, and current distributions obtained from Serpent code, the fuel assembly-based FVM diffusion theory solutions, and a genetic algorithm are used to generate the transport equivalent diffusion parameters. The task of the genetic algorithm used is to minimize the difference between Serpent and diffusion theory solutions by estimating the two group diffusion coefficients and the scattering matrix. The transport equivalent diffusion parameters are then used for the full core neutronics analysis by FVM and Krylov subspace methods. The results of the analysis are compared with the measured data provided by the BEAVRS benchmark and the results of Serpent code simulations and found to be in complete agreement with the Serpent code results and satisfactorily consistent with the measured data provided by the BEAVRS benchmark.

Formation and characterization of hydride rim structures in Zircaloy-4 nuclear fuel cladding tubes

Zirconium alloy tubes used as nuclear fuel cladding are subject to oxidation and subsequent hydrogen pickup during their long service in commercial light water reactors. The hydrogen picked up in the cladding can precipitate as a brittle hydride rim feature on the cladding outer surface. To better understand the effect of hydride rims on the fracture behavior of stress-relieved zirconium alloy cladding tubes, a procedure to produce these rim-like structures has been developed and is described herein. Extensive characterization of the ‘as hydrided’ tubes is performed.The hydrogen charging apparatus consists of a tube furnace with a quartz chamber that is connected to a vacuum pump as well as a bottle of pure hydrogen gas. The charging station uses a static charge of hydrogen as opposed to a flowing gas. The chief advantage of this approach is the ability to monitor the pressure drop in the hydriding chamber and correlate this pressure drop to a known rate of hydrogen pickup in the cladding tube. It was found that the hydrogen partial pressure, metal temperature, and surface treatment all clearly played a role in whether a hydride rim was formed. Extensive characterization of the hydride rims shows they consist of needle like platelets of δ phase hydrides (ZrH1.66) oriented in the circumferential direction with a radial spacing of several microns in a sandwich-like structure.

Nondestructive Characterization and Monitoring of Cracking in Pressurized Piping Using Surface Strain Measurements

Primary loop piping integrity is a regulatory and safety concern for the long-term operation (LTO) commercial light water reactor (LWR) nuclear power plants (NPPs). Typically, primary loop piping in existing LWRs was manufactured from cast or wrought austenitic stainless steel alloy material and operates at a pressure of ~2250 psi (16MPa) for a pressurized water reactor (PWR) design type and ~1000 psi (7 MPa) for a boiling water reactor (BWR). The objective of this research was to determine if measurement of surface strains can be used to characterize inner diameter flaws. For this feasibility study, wrought carbon steel Alloy A106 Grade B material was selected in lieu of stainless steel. This research was performed on pre-notched, 10 in. (25 cm) nominal outer diameter (OD) Schedule 120 (0.844 in. [21.4 mm) wall thickness) pipe samples of 24 in. (61 cm) length. The notches were made using electrical discharge machining (EDM) on the inner diameter (ID) of the pipe samples, with notch features two-dimensional (2D) and planar in nature to simulate cracking. The surface strains in the samples were evaluated under increasing hydrostatic pressure, up to a maximum of 2300 psi (16MPa) to determine their response to flaws with variable depth. Reusable end caps were utilized in lieu of welded caps for the pressure testing. This article describes the test procedure for the pressure testing and results of the feasibility study, which is the first phase of research with the end goal of the development of online crack monitoring capability for use at operating NPPs.

Coupled module of radioactive waste analysis for CANDU reactor based on the Monte Carlo method

When decommissioning nuclear power plants, it is necessary that various types of radioactive wastes are to be estimated and be analyzed accurately. The first decommissioning process of the commercial light water reactor (LWR) is expected to start soon in South Korea, however, lots of processes are still under development for CANada Deuterium Uranium (CANDU) reactor. Unlike the LWR type reactor, the CANDU has complicated and different shapes of components in the reactor, which makes it difficult to carry out activation analysis considering region-wise different irradiation histories. In this study, a coupled module has been developed with Monte Carlo full core analysis and the region-wise depletion analysis by equipped automatic input generation. Furthermore, the levels of radioactive wastes are determined for transportation including estimation of cask contents based on the coupled module results.

Preliminary Neutronics Study of an Accelerator-Driven Molten Spallation Target–Molten Lithium Source of Tritium

Current tritium production might not be enough for all future fusion research reactors. Different approaches for tritium production have been studied in the past, one of which was tritium production using the accelerator-driven subcritical systems. This idea was dismissed in the 1990s as uneconomical when compared to using existing commercial light water reactors. This paper presents changes to the basic idea, mainly the use of a molten spallation target and molten lithium breeding volume. This advanced design is described, optimized for tritium yield using the MCNP 6.2.0 code, and compared between different accelerators. The optimized configuration consists of a 1-GeV, 200-mA proton accelerator, a molten Pb-Bi eutectic spallation target with a length of 60 cm and a diameter of 75 cm, and molten lithium breeding volume with dimensions of 500 cm in length and 900 cm in diameter. As calculated, the annual production of the proposed accelerator-driven tritium production system could be as high as 350 g of tritium with the optimized configuration.

Deformation mechanism in nanolaminate FeCrAl alloys by in situ micromechanical strain rate jump tests at elevated temperatures

FeCrAl alloys have attracted extensive attention as one of the potential cladding materials for the commercial light water reactors. Although the creep behaviors of coarse-grain FeCrAl alloys have been investigated at elevated temperatures, little is known on the deformation mechanism of fine-grain FeCrAl alloys. Here, we performed surface mechanical grinding treatment to fabricate nanolaminated ferritic FeCrAl alloys. In situ micropillar compression tests were applied to determine the underlying deformation mechanism of nanolaminated FeCrAl alloys at elevated temperatures. The strain rate sensitivity and activation energy for the dominant deformation mechanism were estimated considering the presence of a threshold stress.

Seismic classification of high temperature engineering test reactor

In the late 1980s, the seismic classification of the high temperature engineering test reactor (HTTR) was formulated during its design stage. Because of the HTTR’s lack of operation experiences to sufficiently understand the safety characteristics of high temperature gas cooled reactors (HTGRs) at that time, the seismic classification of commercial light water reactors (LWRs) was applied to HTTR. However, subsequent operation experiences and test results using HTTR made it clear that the seismic classification of commercial LWRs was somewhat too conservative for the HTGRs. The seismic classification of HTTR was reevaluated using the following approach. Class S facilities with high safety requirements ensured that they maintained their function even when the safety functions of Class B and Class C facilities were lost during earthquake-triggered accidents, thereby safely controlling the accident without exposing the members of the public in the vicinity of the reactor to excessive radiation exposure. Consequently, Class S facilities were downgraded compared with the commercial LWRs. Moreover, an evaluation was conducted to verify the validity of the new seismic classification under the conditions where the functions of Class B and Class C facilities are lost. It was confirmed that the reactor was stable and that the evaluated public dose was below the authorized limit. In June 2020, the Nuclear Regulatory Authority in Japan confirmed that the results of the seismic classification conformed to the standard rules of the reactor installation change.

Phase-field simulations of fission gas bubbles in high burnup UO2 during steady-state and LOCA transient conditions

To improve the economics of commercial light water reactors, increased understanding of UO2 nuclear fuel with the high burnup structure (HBS) is required in both steady-state and transient conditions. Here, a phase-field model of the fission gas bubble microstructure in nuclear fuel is developed based on the Kim-Kim-Suzuki (KKS) formulation and implemented in Idaho National Laboratory’s Marmot application for phase-field simulation of nuclear materials. The model includes the effects of gas pressure and the surface tension of the bubble-fuel matrix interface for arbitrary interfacial curvature. Simulations of bubble growth in the HBS region during steady-state conditions showed that initially overpressurized bubbles decreased in pressure during growth, but still remained above equilibrium pressure. During a loss-of-coolant accident (LOCA) transient, simulations of bubbles in the HBS region showed that bubble size did not change significantly. The bubble pressure in response to the LOCA transient was calculated for a variety of bubble sizes, initial pressures, and external restraint pressures.

Improvement of the BISON U3Si2 modeling capabilities based on multiscale developments to modeling fission gas behavior

• The recommended models for analyzing U 3 Si 2 fuel behavior with BISON are provided. • Lower-length scale methods have been used to reduce uncertainty in fission gas modeling of U 3 Si 2 . • Comparisons to PIE measurements of U 3 Si 2 fueled rodlets are acceptable. Uranium silicide ( U 3 Si 2 ) is a concept explored as a potential alternative to UO 2 fuel used in light water reactors (LWRs) since it may improve accident tolerance and economics due to its higher thermal conductivity and increased uranium density. U 3 Si 2 has been previously used in research reactors in the form of dispersion fuel which operates at lower temperatures than commercial LWRs. The research reactor data illustrated that significant gaseous swelling occurs as the fuel burnup increases. Therefore, it is imperative to understand the fission gas behavior of U 3 Si 2 under higher temperature LWR operating conditions. In this work, molecular dynamics and phase-field modeling techniques are used to reduce the uncertainty in select modeling assumptions made in developing the fission gas behavior model for U 3 Si 2 in the BISON fuel performance code. To support the implementation of a U 3 Si 2 fission gas model in BISON, cluster dynamics simulations of irradiation enhanced Xe diffusion have been carried out. Similarly, MD simulations were used to predict the athermal contribution due to atomic mixing during ballistic damage cascades. By combining our results with literature DFT data for thermal equilibrium diffusion, Xe diffusivity has been described over a wide range of temperatures for in-reactor conditions. These lower length scale informed models are then utilized in the assessment of BISON U 3 Si 2 modeling capabilities by simulating the ATF-1 experiments irradiated in the Advanced Test Reactor (ATR). Sensitivity analysis (SA) and uncertainty quantification (UQ) are included as part of the assessment process to identify where further experiments and lower length scale modeling would be beneficial. The multiscale modeling approach utilized in this work can be applied to new fuel concepts being explored for both LWRs and advanced reactors (e.g., uranium nitride , uranium carbide).

The thermal and elastic properties of U3Si5 and their variations induced by incorporated aluminum

Uranium silicide compounds have attracted intensive attention as candidate alternative fuels in commercial light water reactors (LWRs). In this work, the electronic, thermal, mechanical and elastic properties of U3Si5 are comprehensively investigated based on first-principles density functional calculations and semi-classical Boltzmann transport theory. U3Si5 is determined to be a brittle and magnetic metal, and thermal conductivity is dominated by its electronic contribution at high temperature. The ratio between the bulk and shear moduli is 1.205. In addition, the elastic ideal strength is strongly anisotropic, with the minimum value only 6.831 GPa at a uniaxial tensile strain of 0.07. This low strength is mainly caused by an emerged structural transition forming silicon-silicon bonds to absorb strain energy. Moreover, the influences of aluminum incorporation on the structural and elastic properties, and thermal conductivities of U3Si5 are further studied. The aluminum atom prefers to replace the silicon atom at the vertex shared by the silicon triangles and pentagons. The incorporated atom impedes the structural transition, and enhances the toughness of U3Si5. The results from this work may provide useful clue for the improvement in the application of U3Si5.

Instability of U3Si2 in pressurized water media at elevated temperatures

Following the Fukushima Daiichi accident, significant efforts from industry and the scientific community have been directed towards the development of alternative nuclear reactor fuels with enhanced accident tolerance. Among the proposed materials for such fuels is a uranium silicide compound (U3Si2), which has been selected for its enhanced thermal conductivity and high density of uranium compared to the reference commercial light water reactor (LWR) nuclear fuel, uranium oxide (UO2). To be a viable candidate LWR fuel, however, U3Si2 must also demonstrate that, in the event of this fuel coming in contact with aqueous media, it will not degrade rapidly. In this contribution, we report the results of experiments investigating the stability of U3Si2 in pressurized water at elevated temperatures and identify the mechanisms that control the interaction of U3Si2 under these conditions. Our data indicate that the stability of this material is primarily controlled by the formation of a layer of USiO4 (the mineral, coffinite) at the surface of U3Si2. The results also show that these layers are destabilized at T > 300 °C, leading to the complete decomposition of U3Si2 and its pulverization due to its full oxidation to UO2.

Accommodation and diffusion of Nd in uranium silicide - U3Si2

Uranium silicide, U3Si2, is considered as an advanced nuclear fuel for commercial light water reactors with improved accident tolerance as well as competitive economics. Nd is employed as a local burnup indicator for conventional oxide fuels due, among other reasons, to its low mobility in the UO2 fuel matrix and its high fission product yield. As part of the studies necessary to determine whether Nd can be considered as a candidate burnup indicator in the U3Si2 concept fuel, we investigate the mobility of Nd in U3Si2. In this work, density functional theory (DFT) calculations are performed to predict the most stable accommodation sites of Nd in U3Si2, found to be within the uranium sublattice. Based on DFT calculations of binding energies and migration activation energies, we investigate Nd diffusion by computing the transport coefficients within the framework of the self-consistent mean-field method. Our calculations predict that the diffusion ratio of Nd to U is smaller in U3Si2 than in UO2. Moreover, at the individual maximum centerline temperature of the fuel, the diffusion of Nd in U3Si2 is much slower than in UO2. From this perspective, Nd represents a good candidate burnup indicator, in similarity to that in UO2.

Process Engineering Challenges for the Development of Electrolytic Reduction of Uranium Oxide in Molten LiCl-Li2O

Electrolytic uranium oxide reduction has the potential to be a key process for recycling spent fuel from commercial light water reactors to advanced nuclear reactors—including molten salt reactors and metal fueled reactors. However, several problems with the process have been identified that need to be addressed to support efficient, cost-effective commercial implementation. Process optimization requires attention to metal corrosion, anode stability, cell efficiency, and cathode product purity. Generation of oxygen bubbles at the anode combined with high temperature (650oC) and molten chloride salt (LiCl + 1 wt% Li2O) create a highly oxidizing environment for metals needed for salt containment and shrouding of the anode. The current standard choice for the inert anode is platinum metal, which has fairly good stability in the system. But given the high cost of platinum, alternative anode materials need to be discovered. Efficient removal of oxygen bubbles requires use of a shroud around the anode. The easy fabricability of metals makes them a reasonable choice for the shroud material, but corrosion due to interaction with the salt and oxygen needs to be minimized. Cell efficiency is affected by hydroxide impurity in the salt. Methods for removing water and hydroxides from the salt will be reported that improve cell efficiency. Cell efficiency can also be affected by controlling the reduction mechanism. Uranium oxide can either be reduced chemically by reaction with electrochemically generated lithium atoms or electrolytically with direct reduction of UO2. The reduction mechanism also affects the buildup of entrained lithium oxide in the cathode product. Entrained lithium oxide is problematic as it gets carried on to subsequent processing steps such as electrorefining and molten salt oxidation.

Process Engineering Challenges for the Development of Electrolytic Reduction of Uranium Oxide in Molten LiCl-Li2O

Electrolytic uranium oxide reduction has the potential to be used for recycling spent fuel to a range of nuclear reactors from commercial light water reactors to advanced nuclear reactors—including molten salt reactors and sodium cooled reactors. However, several process engineering-related problems have been identified that need to be addressed to support efficient, cost-effective commercial implementation. Two keys are cathode product purity and cell current efficiency. High product purity relies upon the effective removal of LiCl-Li2O from the cathode basket and reduced uranium. Removal of the salt from the basket can be achieved via high speed spinning at 650oC. But entrained Li2O will remain and should be minimized during the reduction process. Methods for removing hydroxide impurities from the salt are reported that improve cell-current efficiency. The extent of this reduction can be determined via three distinct techniques (thermogravimetric analysis, elemental oxygen analysis, and bromine dissolution) and is key in helping determine the efficiency of the electrochemical cell.

Irradiation performance of nonfertile (Pu-MA-Zr) fast reactor metal fuels

This work was part of a program begun in 2001 to develop advanced nuclear fuels, originally as carriers for plutonium and minor actinides (neptunium, curium, and americium) taken from spent commercial light-water reactors (LWR) so that the plutonium and minor actinides could be ‘burned’ or transmuted in an accelerator or a fast nuclear reactor. A central part of these experiment programs has been the development of advanced fast reactor fuels, because a fast reactor was considered the most efficient vehicle to transmute the actinide waste products, and metallic fuels is a central focus of these tests. An experiment design was developed in which a thermal test reactor, the Advanced Test Reactor (ATR), was used to test small fuel pin prototypes, by creating areas in the core shielded by cadmium filters to produce a largely epithermal and fast neutron spectrum environment in which the pins could be irradiated. The results of non-fertile metallic fuel (no uranium) tests are presented here.Pu-Am-Np-Zr fuels were irradiated to fission densities up to 33 × 1020 fission/cm3 and Pu-239 depletions of up to 39%. The depletions were created by roughly 2/3 by fission and 1/3 by transmutation neutron capture. Up to five fuel ‘rodlets’ were irradiated in three sealed capsules stacked axially in the core, and the peak cladding temperatures ranged from 300°C to 500°C, depending on axial location as those near the core centerline are operating hotter and to higher fission densities. Several post-irradiation examinations (precision gamma scanning and fission gas release) were similar to other historical metal fuel experiments in fast reactors. However, optical metallography indicated that two of the rodlets had breached. The exact reasons are unclear. Due to the design of this irradiation experiment a rodlet breach could have increased the temperature in others in the same capsule by contaminating the thermal gap helium with heavier and less conductive fission product gases. Some of those rodlets showed high amounts of fuel/cladding chemical interaction (FCCI).

Coupled Neutronics–Thermal-Hydraulic Simulation of BEAVRS Cycle 1 Depletion by the MCS/CTF Code System

The coupled neutronics–thermal-hydraulic simulation of the Benchmark for Evaluation and Validation of Reactor Simulations (BEAVRS) Cycle 1 depletion has been performed by the Monte Carlo–based multiphysics coupling code system MCS/CTF. MCS/CTF is a cyclewise pi-card iteration-based inner-coupling code system that couples the subchannel thermal-hydraulic code CTF as a thermal-hydraulic solver in the Monte Carlo neutron transport code MCS. MCS has been developed by the Computational Reactor Physics and Experiment Lab group at the Ulsan National Institute of Science and Technology for the full-core analysis of large-scale commercial light water reactors with high fidelity at the engineering level. With the high-fidelity performance of MCS, the quarter-core pinwise depletion simulation for the BEAVRS Cycle 1 benchmark has been conducted with thermal-hydraulic feedback including fuel temperature, coolant temperature, and coolant density. Moreover, the MCS internal one-dimensional thermal-hydraulic solver TH1D (MCS/TH1D) has been utilized as the reference. On one hand, the simulated results of the criticality boron concentration and axially integrated assemblywise detector signals were compared with measured data. On the other hand, the comparisons of power, fuel temperature, coolant temperature, and density are also presented in this paper.

Fuel performance analysis of BEAVRS benchmark Cycle 1 depletion with MCS/FRAPCON coupled system

A fuel performance (FP) analysis of the BEAVRS (Benchmark for Evaluation and Validation of Reactor Simulations) benchmark Cycle 1 depletion is performed using the MCS/FRAPCON coupled code system. MCS/FRAPCON is a cycle-wise Picard-iteration inner-coupling code system. It is based on the Monte Carlo neutron-transport code MCS and employs the steady-state fuel behavior prediction code FRAPCON as a thermal-hydraulic (T/H) and FP solver. MCS is developed by the Computational Reactor Physics and Experiment Lab of Ulsan National Institute of Science and Technology for the high-fidelity full-core analysis of large-scale commercial light water reactors. Results of power, fuel temperature, coolant temperature and coolant density distributions are presented and analyzed for a quarter-core pin-wise depletion simulation of the BEAVRS Cycle 1 benchmark with T/H and FP feedback (10 axial meshes per pin, 180,870 depletion cells in total). For code-code comparison purpose, the depletion simulation is also conducted with the MCS/TH1D (internal one-dimensional T/H solver) and MCS/CTF (external sub-channel 3D T/H solver) coupled code systems. The dependence to the burnup of the power, fuel temperature, coolant temperature, and coolant density distributions is analyzed by comparison between the three coupled systems. Validation is performed against BEAVRS measured data for the calculated boron letdown curve and calculated distributions of axially-integrated assembly-wise detector signal. Finally, unique distributions of parameters that can only be obtained by FP analysis are examined to illustrate the advanced analysis capability of MCS/FRAPCON.

A simple treatment of increased gap due to fuel assembly bowing through correction of cross sections

ABSTRACT Increased fuel assembly gap due to bowing in commercial light-water reactors (LWRs) has an impact on local pin-power distribution due to increased local moderation. In order to consider the effect of increased assembly gap without explicit consideration of increased gap width, a correction method of cross sections in the gap region is proposed. In this cross section correction method, the average chord length of gap region is preserved to capture the effect of increased gap width. The validity of the present method is confirmed by verification calculations in a single assembly and 5 × 5 assembly geometries using the GENESIS code, which is a transport code based on the method of characteristics.

Investigation of sol-gel feedstock additions and process variables on the density and microstructure of UN microspheres

The kernel of fully ceramic microencapsulated (FCM) fuel requires a material with high fissile density. For that reason, among others, UN is the appropriate chemical state for low enriched U fuel. The UN kernels are spheres ∼800 μm in diameter made using a sol-gel process. The effect of additives on the chemistry and density of the UN microspheres are investigated in this work. Gadolinium nitrate hexahydrate, Gd2O3, B, SiO2 and SiC were incorporated into the sol-gel broth in varying concentrations. It was found that Gd can serve as both a sintering aid and burnable poison when added to the sol-gel broth as a Gd(NO3)36H2O. However, even with the increased theoretical density of the UN microspheres, the U density was still too low for the FCM design that replaces UO2 pellets in a commercial light water reactor. Silicon carbide and B were also successfully added but produced a lower density final product. Other sol-gel processing variables were investigated. The pour density of the sol-gel feedstock was found to influence the final converted UN kernel density.