Light Water Reactor Research Articles

New RMC Energy Deposition Treatment and its Application in Multi-Physics Simulation

Accurate estimation of energy deposition is important in core physics and severe accident analyses for design optimizations. In this study, a new energy deposition treatment is implemented in the Reactor Monte Carlo (RMC) code, offering multiple modes with varying levels of fidelity and computational requirements. The most precise mode is utilized in coupling simulations between RMC and the subchannel thermal-hydraulic analysis code SUBCHAN, incorporating an explicit moderator heating fraction in the coupling interface. The new treatment is verified against references from MCNP, Serpent, and OpenMC for three light water reactor (LWR) assembly cases, and great agreement is achieved. Energy deposition in different materials and components is emphasized in Kilowatt Reactor Using Stirling TechnologY (KRUSTY) modeling, and the results obtained using different modes are compared. The RMC-SUBCHAN coupling calculations for the three LWR assembly cases, employing the most accurate model, reveal a maximum increase of 94.6 K in the control rod centerline temperature, with a normalized energy deposition of 35.9% in the control rod regions. In the assembly case with gadolinium (Gd) burnable poison, a temperature increase of 7.3 K is observed in the Gd rod centerline, while the coolant outlet temperature decreases by 1.6 K due to the reduced explicit moderator heating fraction of 2.1%, compared to the constant 2.6% in the previous coupling scheme.

Role of loading direction on high-temperature tensile behavior of Fe-13Cr-4Al-2Mo-0.65Nb alloy

The Fe-13Cr-4Al-2Mo-0.65Nb alloy is being considered as a promising accident-tolerant cladding fuel material for light water reactors design. The present work investigated the mechanical properties and serration behavior of the hot-rolled FeCrAl alloy in the rolling direction (RD) and transverse direction (TD) under various temperatures. The presence of the high-density dislocation bands parallel to RD and the smaller grain size in TD are recognized as leading causes of the lower ductility of the TD specimens. Additionally, both RD and TD specimens show identical serrated flow behaviors at each temperature, suggesting the serration behavior is independent of the loading direction.

Evolution of extended defects in UO2 during high temperature annealing

The evolution of thermophysical characteristics of uranium oxide (UO2) under irradiation is a critical aspect of nuclear fuel performance of light water reactors. This study examines the effect of high-temperature annealing at 1000 °C - 1600 °C on defect evolution in krypton irradiated UO2 at room temperature. Transmission electron microscopy analysis reveals that the size of irradiation-induced bubbles and dislocation loops increases but their density decreases with annealing temperatures. Perfect dislocation loops with b⇀ = <110> dominate, regardless of the annealing temperature. A rate theory model for extended defect evolution is applied to conclude that current understanding of dynamical loop and bubble evolution at high-temperature annealing is governed by defect coalescence.

The main structural-phase states of interaction between model corium of a nuclear reactor and a sacrificial material based on Al2O3 and a lead layer

This article presents the results of post-experimental studies of the interaction of model light water reactor corium with a candidate sacrificial material (SM) based on Al2O3 (corundum) and a lead layer. The purpose of the experiments was to study the features of the interaction of the sacrificial material with corium in the conditions of simulating a severe accident with core meltdown. An active interaction of corium with the sacrificial material was established as a result of the research. This is confirmed by the fact that part of the sacrificial material entered into physical and chemical interaction with the components of corium with the formation of liquid-phase reaction products and, as a result, creating different conditions for the crystallization of corium. The study of the phase composition showed that the microstructure of solidified corium in various areas after interaction with the sacrificial material is represented by a phase of the (U,Zr, …)O2±x type with different types of crystal lattice. At the same time, the phase analysis showed a close correspondence of the composition of the solid solutions to the initial ratio of uranium and zirconium in the model corium charge. This allows us to conclude that liquid zirconium interacted with aluminum oxide during the experiment. Thus, it was concluded that the proposed sacrificial material is promising based on the results obtained and the identified features of the interaction of aluminum oxide with the lead layer and the corium of a nuclear reactor.

A Framework to Assess Advanced Reactor Spent Fuel Management Facility Deployment

Advanced nuclear reactors offer various operational advantages over existing light water reactors but could produce types of spent nuclear fuel (SNF) with a wide variety of forms and characteristics depending on how many different concepts are deployed. Each advanced reactor SNF type potentially poses unique management challenges. New planning efforts will be necessary to anticipate how the management requirements of advanced reactor SNF will affect the deployment of an integrated waste management system. This paper applies a framework of high-level facility deployment milestones to a generic SNF management system, reviewing them together with the advanced reactor SNF characteristics and management requirements. This allows for the investigation of factors that influence facility and system deployment, and ultimately, the identification of challenges facing the deployment of different kinds of SNF management facilities. The back end of the once-through fuel cycle is examined for four advanced reactor system technology types: sodium-cooled fast reactors, high-temperature gas-cooled reactors, liquid-fuel molten salt reactors, and lead-cooled fast reactors. It is observed that milestones earlier in the facility deployment process (e.g., siting and facility design) are more impacted by the uniqueness of advanced reactor SNF characteristics than others (e.g., construction and testing). Ultimately, none of the differences are seen as fundamentally disqualifying in a technical sense; however, they should be considered early, potentially as part of reactor design, to avoid issues in the future.

Analytic Error Analysis of the Partial Derivatives Cross-Section Model—II: Numerical Results

Accurately predicting errors incurred in a cross-section model for two-step reactor analysis enables the development of optimal case matrices and more efficient cross-section models. In a companion paper, we developed a systematic methodology for the partial derivatives cross-section model through rigorous analytic error analysis. In this paper, we verify our methodology against the conventional “brute force” numerical approach using a typical pressurized water reactor (PWR) lattice. By successfully reproducing known results, we gain confidence in our methodology’s application to advanced reactor environments, where optimal case matrices are generally not known. Our error methodology relies on accurately estimating bounds on the derivatives of the cross-section functions, a task we achieve through an order of convergence study. We use these derivative bounds in derived error expressions to obtain pointwise and setwise cross-section error bounds and verify these results with reference solutions of various two-group cross sections. We then propagate the cross-section error bounds to reactivity error using first-order perturbation theory and analyze their combined effect. Application of this approach to our test problem corroborates our prior qualitative findings with quantitative evidence and reveals the relative magnitudes of interpolation and model form error sources across diverse PWR cross-section functionalizations. Our results suggest systematic pathways for improving case matrix construction to minimize the overall error. These findings also confirm what is well known to the light water reactor design community, which is that interpolation error of cross sections for standard interpolation procedures and case matrix structures is on the order of 10 pcm or less. Future work includes exploring different lattice types and functionalizations, extending reactivity bounds to multi-lattice problems, and investigating historical effects within the macroscopic depletion model.

Cradle to grave: the importance of the fuel cycle to molten salt reactor sustainability

Advanced reactor technologies are being considered for the next-generation of nuclear power plants. These plants are designed to have a smaller footprint, run more efficiently at higher temperatures, have the flexibility to meet specific power or heating needs, and have lower construction costs. This paper offers a perspective on molten salt reactors, promoted as having a flexible fuel cycle and close-to-ambient pressure operation. A complexity introduced by reducing the reactor footprint is that it may require low-enriched fuel for efficient operation, available from enrichment of the feed salt or by reusing actinides from existing used nuclear fuel (UNF). Recycling UNF has the potential to reduce high-level waste, if done correctly. Release limits from UNF processing are stringent, and processes for waste reduction, fission gas trapping, and stable waste-form generation are not yet ready for commercial deployment. These complex processes are expensive to develop and troubleshoot because the feed is highly radioactive. Thus, fuel production and supply chain development must keep abreast of reactor technology development. Another aspect of reactor sustainability is the non-fuel waste streams that will be generated during operation and decommissioning. Some molten salt reactor designs are projected to have much shorter operational lifetimes than light-water reactors: less than a decade. A goal of the reactor sustainability effort is to divert these materials from a high-level waste repository. However, processing of reactor components should only be undertaken if it reduces waste. Economic and environmental aspects of sustainability are also important, but are not included in this perspective.

Oxide Layers in Ni-doped FeCrAl Alloy in 320°C Radioactive Hydrogenated Water

Iron-chromium-aluminum (FeCrAl) alloys have been the interest of the international nuclear energy industry as an accident tolerant fuel cladding replacement to the Zircaloy cladding alloys due to their enhanced corrosion resistance in both normal operation and accident conditions in light water reactors. It is important to understand the corrosion behaviors of FeCrAl before deploying it to nuclear power plants. This study aims to investigate if the addition of a small amount of nickel (Ni) could stabilize the oxide layer by forming the Ni-bearing spinel oxide, such as NiFe2O4, and increase the stability of the oxide layer under irradiation conditions. Current results show that the outer oxide crystals formed on the irradiated FeCrAl-Ni sample surface became larger than that on the FeCrAl sample, which indicates that the Ni addition facilitates the growth of oxide layers and that the FeCrAl-Ni sample exhibited a protective inner oxide layer in the harshest irradiation environment, indicating that Ni improves the corrosion resistance of FeCrAl in radioactive hydrogenated water.

Effects of pressurized water reactor environment and cyclic loading parameters on the low cycle fatigue behavior of 304L stainless steel

Austenitic stainless steels used in light water reactor coolant environments are susceptible to environmentally assisted fatigue due to non-monotonic loading conditions, primarily associated with load-follow, thermal transients, or intermittent plant shutdowns and start-ups. This study investigates the effects of a high-temperature pressurized water reactor (PWR) water environment and cyclic loading parameters on the low cycle fatigue behavior of austenitic 304L stainless steel. Prolonged exposure to a PWR environment and cyclic loading conditions such as a lower strain rate or a higher fraction of slow strain rate enhances the initiation and accelerates the crack growth rate of fatigue cracks, resulting in decreased fatigue life. The deformation-induced α'-martensite is observed in proximity to fatigue crack tips primarily in specimens tested in simulated PWR primary water, while cellular dislocation structures are more frequently observed near crack tips in specimens tested in high-temperature air. The deformation-induced martensitic transformation from γ-austenite to α'-martensite, occurring via the precursor ε-martensite phase, contributes to the accelerated fatigue crack growth rate in a PWR environment with hydrogen.

Investigating Hydrogen in Zirconium Alloys by Means of Neutron Imaging.

Neutrons interact with the magnetic moment of the atomic shell of an atom, as is common for X-rays, but mainly they interact directly with the nucleus. Therefore, the atomic number and the related number of electrons does not play a role in the strength of an interaction. Instead, hydrogen that is nearly invisible for X-rays has a higher attenuation for neutrons than most of the metals, e.g., zirconium, and thus would be visible through dark contrast in neutron images. Consequently, neutron imaging is a precise, non-destructive method to quantify the amount of hydrogen in materials with low attenuation. Because nuclear fuel cladding tubes of light water reactors are made of zirconium (98%), the hydrogen amount and distribution in metallic claddings can be investigated. Even hydrogen concentrations smaller than 10 wt.ppm can be determined locally with a spatial resolution of less than 10 μm (with a high-resolution neutron microscope). All in all, neutron imaging is a very fast and precise method for several applications. This article explains the basics of neutron imaging and provides samples of investigation possibilities, e.g., for hydrogen in zirconium alloy cladding tubes or in situ investigations of hydrogen diffusion in metals.

High-performance Cr2AlC MAX phase coatings for ATF application: Interface design and oxidation mechanism

Surface-modified Zr-based alloy (ZIRLO) claddings with advanced ceramic coatings are increasingly required for accident-tolerant fuel (ATF) systems in light-water reactors. Cr2AlC MAX phase coatings are promising for this purpose owing to their remarkable properties combining radiation/oxidation/corrosion resistance. However they are suffering from weak interface compatibility to ZIRLO substrate and poor structural densities for long-term services. Herein, we fabricated densely high-purity Cr2AlC MAX phase coatings with uniquely designed Cr/CrCx interfacial layers. The oxidation behavior of the coatings was focused under steam environments at 1000–1200 °C. Results showed that Cr2AlC coatings exhibited an oxidation mass gain of 8.9 mg/cm2 and an oxide thickness of 680 nm after oxidation at 1200 °C for 30 min, which were about 10% and 0.5% of ZIRLO substrate, respectively. Based on microstructural evolutions, the embedded interfacial layers significantly suppressed the rapid diffusion of Al in Cr2AlC coatings to the substrate and the premature delamination of oxidized coatings. Particularly, the formed oxides were identified as dense yet pure α-Al2O3, which endowed the protection against further oxidation and excellent resistance to high-temperature steam corrosion.

Corrosion mechanism and performance of Cr-coated Zr-4 alloy in 360 °C water and 1300 °C steam

A dense and uniform Cr coating is prepared on the Zr-4 alloy using multi-arc ion plating technology, with the aim of serving as an accident-tolerant fuel cladding for light water reactors. The corrosion performance of Cr-coated Zr-4 alloy is investigated after exposure to 360 °C and 18.6 MPa corrosion conditions for up to 5040 h, and its corrosion mechanism is discussed. Furthermore, the oxidation behavior of the Cr-coated Zr-4 alloy in steam at 1300 °C is mainly examined. It is observed that the ZrO2 channel within the Cr coating acts as an oxygen diffusion path during steam oxidation and facilitates the reduction of Cr2O3 by providing Zr. When the ZrO2 channel diffuses to the Cr2O3/Cr interface, Cr2O3 is reduced to form ZrO2 and Cr. When the thickness of the Cr2O3 layer is minimized, the Cr coating fails. The in-situ oxidation of the Cr2Zr layer results in the formation of ZrO2 and Cr2O3, while the substrate beneath the Cr coating is oxidized to form a ZrO2 layer. The whole oxidation process of the Cr-coated Zr-4 alloy up to 1300 °C is presented in detail.

Assessing the master curve reference temperature T0 estimated from the Charpy impact transition curve and consequences on the correlation between Charpy DBTT- and fracture toughness T0-shifts

Integrity assessment of reactor pressure vessels of light water reactors follows well established procedures relying on the surveillance program and defined by codes, standards and nuclear regulatory requirements. These procedures were developed in times where scientific knowledge was limited increasing therefore conservative safety margins to overcome the scientific shortcomings.Taking profit from the development of a number of test techniques and evaluation, it is possible nowadays to propose an alternative interpretation of the test results that provides a complementary information to the standard regulatory approach. In this work, the focus is on the Charpy impact test as it plays a critical role in surveillance programs monitoring reactor vessel degradation induced by neutron irradiation. The development of methodologies allowing experimental determination of fracture toughness from small size specimens triggered alternative approaches of extracting fracture toughness from the Charpy impact test. One of those approaches was recently successfully developed using the instrumented Charpy impact test from which the dynamic and static fracture toughness can be estimated with a reasonably good accuracy. This paper aims to provide an equally performant but simplified procedure allowing estimation of the master curve reference temperature, T0, from the Charpy impact transition curve applied to a large variety of reactor materials and irradiated conditions.

Contribution of IAEA coordinated research projects to light water reactors advanced technology fuel testing and simulation

The Fukushima-Daiichi accident in 2011 highlighted the pressing need for enhanced design and safety analysis of nuclear fuels, especially under accident conditions in nuclear power plants (NPPs). To address this need, the international nuclear fuel community has developed several concepts of Accident Tolerant and Advanced Technology Fuels (ATFs) for light water reactors. The International Atomic Energy Agency (IAEA), at the request of its Member States, has initiated a series of three Coordinated Research Projects (CRPs) to aid in the development and testing of these ATFs. These projects, known as FUMAC, ACTOF, and ATF-TS, focus on the experimentation and simulation of ATFs under various accident conditions, including design basis accidents and design extension conditions. This paper provides a comprehensive overview of the main activities, as well as the achieved or anticipated results of these three CRPs. The insights and advancements gained from these projects will enable IAEA Member States to further refine nuclear fission technology, thereby contributing to efforts aimed at mitigating climate change.

In-situ determination of strain during transient burst testing and the temperature dependence of Zircaloy-4 claddings

Understanding fuel system behavior during postulated loss-of-coolant accidents is pertinent for continued safe and efficient operation of light water reactors, particularly as higher burnups are being pursued and safety margins re-evaluated. Conventional mechanical models for the incumbent Zr alloys typically rely on the assumption that steady-state creep is the dominant fuel cladding response during transient accident conditions. To investigate this assumption, simulated accident burst testing was performed on Zircaloy-4 claddings with balloon behavior measured in-situ. Two distinct loading conditions were utilized during burst testing: (1) constant-gas-inventory where pressure was allowed to increase with temperature and (2) constant pressure. In-situ strains and strain rates were measured via 2-dimensional digital image correlation techniques and synchronized with temperature to determine deformation dependencies. The temperature dependence of strain rate was characterized by a two segment Arrhenius relationship, with a distinct transition between the high and low temperature/strain regimes. The average activation energy of the lower temperature/strain regime was 328 ± 25 kJ/mol, in agreement with the ∼320 kJ/mol used for conventional LOCA models. However, the higher temperature/strain segment, which encompassed most of ballooning, showed increased activation energies as well as a dependence on whether the burst region was in view. For tests that burst away from the camera view, the average high temperature/strain segment activation energy was 635 ± 150 kJ/mol. For samples where the rupture opening formed in view, the average activation energy was 1015 ± 179 kJ/mol. This observed shift in temperature dependence indicates a transition in deformation mechanism at the end of life, possibly to time independent failure mechanisms, which has not yet been visualized in the literature for Zr alloys. Parameters at the transition points were analyzed to determine thresholds for this change in behavior, which occurred at an average hoop strain of 6.9 ± 2.1 %.

Fuel depletion study of the molten salt demonstration reactor

In this study, a molten salt reactor model and a pressurized water reactor model were run in SCALE for comparison. The parameters of interest in the analysis were the concentrations of different nuclides in a molten salt reactor, specifically compared with concentrations of these same nuclides in a pressurized water reactor. This paper discusses this analysis, as well as discussing and comparing the burnup in both models. The results suggest that using a mechanistic source term for radionuclides instead of a bounding, conservative source term enables more realistic reactor modeling, making it a great option for licensing both advanced reactors and light-water reactors.

Rapid repair of thermal aging embrittlement in 17-4PH martensitic stainless steel using pulsed electric current

The 17-4PH stainless steel serves as the main steam isolation valve in a light water reactor due to excellent comprehensive mechanical properties and corrosion resistance. Nonetheless, the thermal aging resulting from prolonged service poses a threat to the safe operation of nuclear power plants. The mechanism of thermal-aging embrittlement of 17-4PH stainless steel and its repairing behavior under the action of the pulsed electric current were studied using transmission electron microscopy and 3D atom probe tomography. The ductile to brittle transition occurred following the spinodal decomposition and G phase formation during the thermal aging of 5000 h at 673 K, resulting in a 100 HV increase in hardness of the 17-4PH stainless steel, of which 94.4 HV was attributed to the spinodal decomposition. Under the action of a pulsed electric current at 733 K, the hardness and impact energy can be restored to its original state, whereas heat treatment at the same temperature is unable to restore these properties. This is because the pulsed electric current treatment decreased the thermodynamic barrier to dissolution for the spinodal decomposition and G phases. The Kolmogorov-Johnson-Mehl-Avrami model was used to calculate the kinetics of spinodal decomposition dissolution. The results suggest that the pulsed electric current could increase the diffusion pre-exponential factor of Cr element and accelerate the elimination of spinodal decomposition. This process ultimately facilitates the quick regeneration of the 17-4PH stainless steel's performance.

Initial Verification and Validation of a New CASMO5 JENDL-5 Nuclear Data Library for Typical LWR Applications

A new nuclear data library for the advanced lattice physics code CASMO5 has been prepared based on JENDL-5. In JENDL-5, the range of data, such as the number of nuclides, has been dramatically expanded for general-purpose applications. At the same time, many essential nuclides for conventional light water reactor (LWR) analysis have also been modified based on state-of-the-art evaluations. The new JENDL-5–based CASMO5 library was prepared by replacing as much of the nuclear data of the current CASMO5 ENDF/B-VII.1–based library as possible with JENDL-5. This study performs initial verification and validation of the new library under typical LWR conditions. Verifications were performed based on the Organisation for Economic Co-operation and Development/Nuclear Energy Agency burnup credit criticality safety benchmark phase III-C, and the calculated kinf and fuel compositions of the boiling water reactor fuel assembly were compared with reported benchmark results. Comparison with the MCNP6.2 result was also performed using the same benchmark model. In addition, a tank-type critical assembly critical experiment and Takahama-3 postirradiation experiment were used for validation. The results indicate that the new library performs well and is comparable to the ENDF/B-VII.1–based library in predictions of reactivity and fuel compositions for typical LWR systems.

Exploring the effects of Al and Si dopants on the accident tolerant fuels of UO2 pellets for light water reactor

This study investigates the potential of Al and Si dopants to enhance the performance of UO2 pellets used in accident tolerant fuel for light water reactors. These dopants are anticipated to improve the uranium loading level, fission product retention, and thermal conductivity of UO2 pellets, while preserving their neutronic performance. Utilizing first-principles calculations, this study analyzes the kinetics of UO2-Al and UO2-Si interactions, predicting the resulting compound types, compositions, properties, and doping techniques. The calculations consider the reaction products and their thermodynamic properties. The study proposes the use of uranium aluminide, uranium silicide, or uranium aluminum silicide (e.g., UAl3, USi, U3Si2, or U3Al2Si3) to fill the UO2 grain boundaries, as opposed to Al2O3 and SiO2. This could be achieved by synthesizing uranium aluminide and uranium silicide separately, and then mixing them with UO2.

TRU utilization and MA transmutation in thorium-based fluorinated molten salt fast reactor

In this paper, a thorium-based fluorinated molten salt fast neutron reactor is taken as the research object, and the utilization of thorium resources, reuse of transuranic waste (TRU), and transmutation of minor actinide (MA) in fast reactor are the research purpose. The fast reactor adopts a multi-layer core structure, in which the fuel is divided into an inner fission fuel zone and an outer breeding fuel layer, respectively loaded with fluorinated molten salts LiF+ThF4+XF4 (where X is 233U, TRU, or 233U+MA) and LiF+ThF4. In addition to serving as nuclear fuel, these two molten salts also act as coolants in the primary circuits.Through the cross sections of nuclear reactions, transmutation chain of actinide nuclides, and the particle population density formula of actinide nuclides in nuclear reaction equilibrium, the possible nuclear reactions of actinide nuclides in thorium-based fast reactors are analyzed. The calculated results show that the thorium-based fast reactor has the following characteristics: fast neutron energy spectrum, negative temperature reactivity coefficient, effective utilization of thorium resources, the ability to breed fissle fuel such as 233U, production of high-purity radioisotope 238Pu material, and the ability to incinerate Pu and MAs in the TRU spent fuel.In the 233U fuel scheme, the 233U fuel is breeded, breeding ratio of fissile fuels is 1.21, the reaction products contain very little Np, and almost no Pu, Am, and Cm long-lived radionuclides. In the TRU fuel scheme, the annual capacity to incinerate MAs and Pu is 1072.6 kg, while producing a large amount of 233U and 233Pa fuel (together 1162.5 kg/year).In the 233U+2.0 wt% TRU-MAs fuel scheme, 127.35 kg of MA waste can be incinerated annually, and their transmutation rate is 15.95%, which is equivalent to the annual production of MA nuclear waste from 4.9 conventional 1000 MWe light water reactors. Here, TRU-MAs is the mixtured MA fuels after separation of Pu from the TRU fuel. At the same time, radioactive fuel 238Pu, which can be used for space nuclear power, is produced with an annual yield of 87.89 kg and an isotopic abundance of 87.57%. If the added TRU-MAs fuel is replaced with 2.0 wt% 237Np fluorinated salt, the annual yield and isotope abundance of 238Pu are further increased to 130.24 kg and 95.4%.