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Increasing the number of pins within a pressurized water reactor (PWR) assembly reduces pin temperature for a given assembly power. In conjunction with a core retrofit, this presents a potential route to PWR uprate, which is of growing interest given recent increases in electricity prices. However, most PWRs utilize regular lattice designs with fixed guide tube positions, such as the very common 17 × 17 lattice design with 25 guide/instrumentation tubes. These tubes are aligned with penetrations in the reactor pressure vessel, which presents a prohibitive obstacle to retrofit, and more widely, may "lock" many PWRs to this particular fuel configuration. In this paper, an irregular PWR fuel assembly is proposed. It is shown that a backward-compatible lattice with 324 fuel pins per assembly (BL324), uniform enrichment, and the same hydrogen-to–heavy metal ratio as a reference 17 × 17 assembly with 264 fuel pins can achieve within-assembly power peaking within 2% of the reference assembly under equivalent conditions while fixing the guide tube positions. Power peaking can be further reduced to reach that of the existing fuel assembly by reducing the enrichment of 36 of the pins by 0.2 wt%. The fuel assembly could potentially either support a significant uprate of up to ~20% in conjunction with low-enriched uranium plus (LEU+) fuel or a more aggressive cycle design, and hence, improved discharge burnup at the same power and batch strategy. A subchannel analysis shows that the coolant heat-up distribution is comparable to the reference assembly. However, the pressure drop is estimated to be 4% higher, which would challenge the performance of transition cores containing both 17 × 17s and BL324s. Further incremental changes to BL324 may be attractive, either to improve manufacturability or to slightly improve performance through formal optimization. [ABSTRACT FROM AUTHOR]

AbstractThis research presents the development of a one-dimensional analytical model to investigate the impact of pressure variations in the primary loop on natural circulation (NC). The model takes into account a sinusoidal input heat distribution and derives equations for the parameters of NC. The model covers a broad spectrum of NC patterns, spanning from fully single-phase to fully two-phase flow. The research demonstrates a smooth and continuous transition between various kinds of NC. Moreover, the research demonstrates that NC is capable of efficiently dissipating the decay heat generated inside the core of a pressurized water reactor, encompassing a range from 100% to 60% of the total inventory present within the primary loop. The findings of this study are compared to prior research outcomes and demonstrate a reasonable level of consistency. Additionally, comparisons are made with uniform input power distribution to demonstrate that there are no significant differences in the NC parameters when using sinusoidal heat input. [ABSTRACT FROM AUTHOR]

This paper presents the results of the Power To Melt and Maneuverability (P2M) Simulation Exercise on past fuel melting irradiation experiments, organized within the Organisation for Economic Co-operation and Development/Nuclear Energy Agency Framework for IrraDiation ExperimentS (FIDES) framework by the Core Group (CEA, EDF, and SCK‧CEN) and open to all FIDES members. The exercise consisted in simulating two past power ramps where fuel melting was detected: (1) the xM3 staircase power transient [ramp terminal level (RTL) 70 kW‧m−1, average burnup 27 GWd‧tU−1], carried out in 2005 in the R2 reactor at Studsvik (Sweden), where the rodlet maintained its integrity, and (2) the HBC4 fast power transient (RTL 66 kW‧m−1, average burnup 48 GWd‧tU−1), carried out in 1987 in the BR2 reactor at SCK‧CEN (Belgium), where the cladding failed during the experiment. The exercise was joined by 13 organizations from 9 countries using 11 different fuel performance codes. In this paper, the main results of the Simulation Exercise are presented and compared to available postirradiation examinations (PIE) or on-line measurements during the power ramps (fuel and clad diameters, rod elongation, pellet-clad gap, and fission gas release). Since the focus of the Simulation Exercise is on fuel melting assessment, determination of the boundary between melted/nonmelted fuel and the consequent definition of a melting radius from PIE are first discussed. During the HBC4 ramp, fuel melting was predicted by most of the codes despite differences in the melting models. Higher discrepancies were observed for the xM3 rod that can be attributed partly to power uncertainty and partly to the limited capability of the models to describe partial melting of the fuel during this ramp. Finally, possible code developments to improve simulation results are presented. [ABSTRACT FROM AUTHOR]

This paper presents 2D(r, $\it \theta $ θ) simulations of the HBC-4 power-to-melt experiment performed with the fuel performance code ALCYONE. The HBC-4 experiment is one of the two test cases selected for the simulation exercise on past fuel melting experiments of the Power to Melt and Maneuverability (P2M) project. The ramp terminal level (RTL) at peak power node (PPN) has been estimated at 66 kW·m−1 by gamma scanning and 70 kW·m−1 based on online measurements of thermal fluxes. The fuel burnup at PPN was close to 60 GWd/tU−1. The cladding failed during the short holding time at a RTL of 40 s. Fuel melting took place at the pellet center, and in particular, in front of clad cracks. In this paper, simulations of the HBC-4 power-to-melt experiment are performed using an updated version of the 2D(r, $\it \theta $ θ) scheme of ALCYONE where half of the fuel pellet is described. This configuration allows for the modeling of clad failure by iodine stress corrosion cracking and of its consequences on fuel pellet deformation. The modeling of fuel melting relies on thermochemical equilibrium calculations performed with the OpenCalphad Gibbs Energy Minimizer and the Thermodynamics of Advanced Fuels International Database. The simulation without clad failure indicates that the solidus is reached during the HBC-4 experiment but not the liquidus. The simulation with clad failure leads to a small increase in the fuel temperature that is sufficient to reach the liquidus at the pellet center, in agreement with postirradiation examination (PIE). The impact of water ingress in the rod and vaporization at the pellet surface is discussed, showing that it could explain the pronounced swelling of the fuel pellet reported from the PIE. [ABSTRACT FROM AUTHOR]

The extension of the operating domain of PreussenElektra's Konvoi-type pressurized water reactors (PWRs) beyond the natural end of cycle is known as stretch-out operation. A range of possibilities exists to increase nuclear fuel utilization to continue operation after the boron concentration reaches its dilution limit. The most basic option is to continue operation with constant average moderator temperature, which results in a relatively fast decrease in reactor power. From a fuel utilization point of view, this is the least optimal procedure. In PreussenElektra's PWR fleet, an enhanced operation mode is adopted, leading to a comparatively modest decrease in reactor power and very high utilization of nuclear fuel. Initially, the stretch-out mode provided an option to gain flexibility regarding outage planning. More recently, the stretch-out method has served as a practical approach to optimizing electricity generation costs during the last cycles before the final shutdown as stipulated by law, as operators can extend the cycle length in a range of 30 to 60 days after the natural end of cycle. This paper describes the licensing rationale, the feasibility of this type of operation, and the operating requirements and experience. The system parameters affected by stretch-out operation are discussed. Adjustments of set points of thermal-hydraulic variables in the primary and secondary systems are explained. Licensing requirements for safe reactor operation in stretch-out mode are reviewed. Furthermore, aspects of neutronic and thermal-hydraulic core surveillance are included. After more than 35 years and counting, the methods of increasing fuel utilization are not new, and an evaluation of experience and effectiveness is in order. [ABSTRACT FROM AUTHOR]

Corrosion products in pressurized water reactors are challenging to study in situ, yet understanding their properties is key to improving reactor performance and radiation reduction. In this study, a hydrothermal synthesis technique was used to produce nickel ferrite (NiFe2O4) particles from goethite (α-FeOOH) and nickel nitrate hexahydrate [Ni(NO3)2 6H2O] in the presence of sodium hydroxide (NaOH). X-ray diffraction was used for phase identification, with scanning electron microscopy used for particle shape and size analysis. By varying the [Ni]:[Fe] ratio of the precursors and synthesis temperature between 100°C to 250°C, a phase diagram was developed to determine the stability field in both composition and temperature for obtaining a single-phase, nonstoichiometric nickel ferrite product. The compositional boundaries of the single-phase region of the diagram are a function of temperature, consistent with the increased solubility and reaction rates at temperatures above 125°C. The single-phase nickel ferrite encompasses [Ni]:[Fe] ratios in a very narrow range at 150°C, only 0.35 to 0.375, but widens as a function of temperature and reaches its greatest breadth at 250°C. At this temperature, a single-phase product is obtained for a range of starting compositions from 0.30 to 0.425. Outside of this window, additional nanoparticles are obtained whose identity and composition vary with both temperature and starting mixture. On the lower nickel content side of the single-phase region, the mixture contains either unreacted goethite (for temperatures below 200°C) or hematite (α-Fe2O3) at 200°C or higher. On the Ni-rich side of the single-phase region, theophrastite [β-Ni (OH)2] was obtained along with the nickel ferrite, at all temperatures studied. The single-phase window was widest at 250°C, resulting in nickel ferrites with a Ni mole fraction between 0.23 and 0.31. [ABSTRACT FROM AUTHOR]

Fully ceramic microencapsulated (FCM) fuel consists of TRISO (tristructural-isotropic) fuel particles embedded in a ceramic matrix (SiC) to form fuel pellets and rods and offers improved fission product retention and lower operating temperature with expected superior performance in normal and off-normal conditions compared to conventional fuel. When coupled with SiC cladding, FCM fuel eliminates zirconium altogether and is expected to drastically reduce hydrogen generation during a beyond-design- basis accident. In order to be deployed in current or future pressurized water reactors (PWRs), FCM fuel must meet or exceed the neutronic performance of conventional fuel. Limited by low heavy metal loading, an FCM fuel assembly requires increased enrichment and large fuel rods to match the cycle length of a conventional fuel assembly. This study investigated the core design, neutronics, and thermal hydraulics of a PWR loaded with FCM fuel and sought to optimize the assembly design to minimize the enrichment required to reach fuel performance similar to that of conventional fuel. It was found that the implementation of FCMfuel in a 17 X 17 assembly requires close to 20% enrichment and large fuel rods. Such design performs comparably to conventional fuel (4.5% enrichment) in terms of cycle length, reactivity coefficients, intra-assembly power peaking factor, burnable poison penalty, and control rod worth but requires an increase of pumping power. A parametric analysis spanned a large design space varying fuel outer diameter and pitch-to- diameter ratio (P/D) and downselected two alternate assembly designs: 11 X 11 (1.65-cm outer diameter and 1.18 P/D) and 9 X 9 (2.12-cm outer diameter and 1.12 P/D). These designs meet the cycle length requirement with 18.6% and 16.2% enrichments, respectively, but feature a smaller minimum departure from nucleate boiling ratio (MDNBR) compared to a reference assembly. It was estimated that a slight increase in rod outer diameter increases MDNBR to the desired level and implies a pressure drop increase of 10%.

Radioactive corrosion product materials collected from the control rod drive mechanism (CRDM) housing in a pressurized water reactor (PWR, HANBIT-1 KNPP) were analyzed using an electron probe micro analyzer (EPMA). It is challenging to analyze the composition of radioactive corrosion products using an EPMA due to the rough surface shape and size, and even more so when the products are stacked in the form of small grains. The purpose of this study is to determine whether the corrosive products found inside the CRDM housing are stuck in contact with primary coolant or just oxide. In this study, not only was the surface condition of the samples very rough, but the samples that were quantitatively analyzed using a normal method had extremely low electrical conductivity using a normal method. We therefore tested a new semiquantitative analysis method using X-ray image mapping. In this technical note, we propose a method for collecting and analyzing corrosion products adsorbed in the CRDM. Reference papers on radioactive corrosion products collected from the CRDM could not be found. It is consequently difficult to argue that the method of collecting samples and performing the quantitative analysis suggested in this study is the best, but it can be said that it is an appropriate analysis method. Finally, the usefulness of the semiquantitative analysis is reviewed by verifying the analysis results of radioactive corrosion products collected from the CRDM housing in a PWR. [ABSTRACT FROM AUTHOR]

This paper proposes the application of a pattern recognition–based technique to enhance the process of control rod position identification in pressurized water reactors (PWRs). The proposed technique employs a multivariant analysis technique, namely, principal component analysis (PCA) and clustering analysis (CA) to identify the position of the PWR control rod using its impact on the core radial thermal neutron flux along the axial track of motion. The results of these investigations have shown that the proposed technique successfully removed the limitation on the data size and any limitations imposed by outlier samples, extracted the noise, and provided near-instantaneous analytical and visual ways for position identification process with excellent generalization fitting and prediction efficiencies. In the context of this paper, multiple in-depth simulations are conducted to ascertain the efficiency of the proposed technique in identifying the control rod positions. These simulations have been conducted using a Westinghouse 2772-MW(thermal) PWR benchmark at 100% thermal power generation, where a three-dimensional TRITON FORTRAN-code has been utilized to simulate the radial thermal neutron flux of the PWR core. The PCA model is developed, tested, and generalized using the SIMCA software package. In addition, CA is also performed via the Minitab statistics software package in order to confirm the efficiency of the proposed technique. [ABSTRACT FROM AUTHOR]

All licensable critical heat flux (CHF) correlations/regressions models must determine and demonstrate a "design limit" that bounds the CHF correlation predicted/measured residuals via a 95/95 tolerance limit. This is a quick and straightforward calculation when the residuals are well behaved, exhibiting no trends and no heteroscedasticity. However, as models become increasingly complex and as required parameter ranges become more extended, the likelihood of nonconservative subregions increases. A suggested solution from the open literature is the overly conservative approach of basing the design limit on the subregion with the largest variance. This approach unavoidably overly constrains the overall regression model and often is too conservative for subregions due to a loss in degrees of freedom. Quantile regressions alleviate these issues by smoothly varying the design limit based on covariates and adapting to each subregion. Thus, a quantile regression achieves the objective of appropriately bounding all subregions without overly biasing the overall regression model. [ABSTRACT FROM AUTHOR]

Fuel-coolant interaction (FCI) is an important issue for the assessment of severe accident safety for both sodium-cooled fast reactors (SFRs) and pressurized water reactors (PWRs). For the ASTRID SFR demonstrator, FCI is a key phenomenon affecting the relocation of molten fuel in engineered discharge tubes between the core region and the core catcher plenum. FCI controls jet fragmentation and debris bed formation and raises the issue of potentially energetic vapor explosions in the ASTRID lower head. In this frame, experimental data will be necessary to validate SCONE, the fuel-sodium interaction code under development at CEA. For PWRs, one of the configurations of interest lies within the residual case where in-vessel retention would fail. In this case, it is expected that a light metallic layer would be the first to interact with water, before a heavier oxide melt discharge. Here, steam explosion and debris bed formation are the two major points of interest. Based on the experimental expertise gained from the KROTOS facility and its X-ray radioscopic imaging system, new test facilities have been designed to carry out prototypic (depleted uranium-containing) corium interactions with either sodium or water in PLINIUS2, the CEA future large-mass experimental platform dealing with masses above 100 kg. Some test sections have been specially designed to ensure proper visualization of the fuel, liquid coolant, and vapor phases by an improved X-Ray imaging system. This paper presents the future PLINIUS 2 platform as well as the experimental programs foreseen to study both water-corium and sodium-corium interactions. [ABSTRACT FROM AUTHOR]

The lifetime of prompt neutrons is a basic characteristic of reactors since it determines the neutron kinetics of the reactor in all transient processes. This paper focuses on calculation of the prompt neutron lifetime for pressurized water reactors (PWRs). The calculation was performed using two independent methods. The first method uses the fundamental definition of the neutron lifetime with adjoint weighting that has recently been included in MCNPX. The second method is the l/v absorber insertion method, where a l/v absorber such as, 0B is placed uniformly throughout a nuclear reactor and the change in reactivity is calculated. This prompt neutron lifetime is then extracted from the changes in the reactivity as the 10B concentration approaches zero. The results of the two methods are compared together at two points in the operation cycle [at beginning of cycle (BOC) and at end of cycle (EOC)]. The values of the prompt neutron lifetime as calculated with MCNPX are compared to values calculated with another independent method, and the results are in reasonable agreement with each other. Also, these results compared with the PWR final safety analysis report show good agreement. In the two methods of calculation, the prompt neutron lifetime was determined to be longer at EOC when compared to that at BOC. [ABSTRACT FROM AUTHOR]

New pressurized water reactor (PWR) cross-section libraries were generated for use with the ORIGEN-ARP depletion sequence in the SCALE nuclear analysis code system. These libraries are based on ENDF/B-VII.O nuclear data and were generated using the two-dimensional depletion sequence, TRITON/NEWT, in SCALE 6.1. The libraries contain multiple burnup-dependent cross sections for seven PWR fuel designs, with enrichments ranging from 1.5 to 6 wt% 235U and burnups from 0 to 90 GW(d)/tonne U. The computational methodology and studies performed to establish an optimal depletion model or cross-section library generation are discussed in this paper. Verification against detailed TRITON simulations or the considered assembly designs showed that depletion calculations performed in ORIGEN-ARP with the pregenerated libraries provide results similar to those obtained with direct TRITON depletion while greatly reducing the computation time. Validation of the libraries, carried out using radiochemical assay measurements and decay heat measurements for PWR spent fuel, showed good agreement between calculated and experimental data. [ABSTRACT FROM AUTHOR]

A novel machine learning approach for nuclear power plant modeling and state identification is presented together with its test results using data from the Loss-of- Fluid Test experimental facility. The approach exploits Gaussian processes whose principal function is to tackle the temporal problem of forecasting the actual system state in the varying environment of a nuclear reactor facility that undergoes successive overcooling tramisients. The approach fuses independent Gaussian process expert predictions to provide a single recommendation to the plant operators in a form that is suitable to appear on a decision support system screen. A variety of test cases are developed to explore the validity and relevance of Gaussian processes. The proposed implementation is examined with various predictor variables under different conditions, and the results obtained are in accordance with model expectations. [ABSTRACT FROM AUTHOR]

In the case of large burnup, a control rod (CR) guide tube in the pressurized water reactor of a commercial nuclear power plant might bend. As a consequence, a CR drop experiment may indicate an event of a CR partially inserted and whether the CR should be deemed inoperable. Early prevention of such an event can be achieved by measuring two friction coefficients: the hydraulic coefficient and the sliding coefficient. The hydraulic coefficient hardly changes, so that the curvature of the guide tube can only be detected thanks to a variation of the sliding coefficient. A simple model for the CR drop is established and validated with CR drop experiments. If tmx denotes the instant of CR maximum velocity, a linear relationship between (tmx)-2 and the sliding coefficient is found. [ABSTRACT FROM AUTHOR]

To have adequate confidence in the proposed design of the internally and externally (I&E) cooled annular fuel it is important to identify the fuel operational constraints from a materials performance perspective. To accomplish this goal a capability for modeling I&E cooled annular fuel has been developed for two manufacturing approaches: (a) the sintered and pressed pellet approach and (b) the vibrationally compacted (VIPAC) particle approach. New models for the burnup and power radial distribution, fuel thermal and irradiation dimensional changes as well as fuel-cladding interaction mechanisms for annular fuels have been developed and incorporated into a modified version of the FRAPCON code. Fission gas release from the sintered fuel is found to be lower for the same burnup than the traditional solid fuel but slightly higher for the VIPAC fuel. The VIPAC internal rod pressure, however, remains below that of the solid fuel for much higher burnup. The power density constraints and design limits are studied, as well as sensitivity of the annular fuel design to fabrication and operation uncertainties. It is concluded that such fuel can be operated at 30 to 50% higher core power density than the current operating light water reactors, and to a burnup of 80 to 100 MWd/kg U. The major issue for the pellet fuel rod design is the asymmetry in heat transfer that might develop when the outer gap is closed early in the irradiation due to the outward thermal expansion of the fuel. Solutions to this issue via smaller initial inner gap, small roughness and tolerances on fuel and clad surfaces, or the addition of a highly porous ZrO2 layer on the outer pellet surface are evaluated. The main issue for the VIPAC fuel is selection of the particle sizes, which control both the effective density of the fuel and the fission gas release. [ABSTRACT FROM AUTHOR]

Nuclear energy technology has to evolve continuously toward two goals: reduction of the cost of power plants and increasing their safety margins. One approach that can enhance both the economics and safety performance is the adoption of new fuel geometries that increase the fuel surface-to-volume ratio in the core. This will facilitate heat transfer to coolant at lower fuel temperatures. This paper reviews the methodology applied to develop an appropriate design of an innovative internally and externally cooled annular fuel for high-powerdensity pressurized water reactors (PWRs). It provides an overview of key results obtained during a U.S. Department of Energy-sponsored project that addressed thermal hydraulics, reactor physics, fuel manufacturing, and fuel performance characterization. Using this fuel, it is shown that power density uprates up to 150% can be achieved in PWRs while maintaining or improving the safety margins. If this is applied to future reactors, significant economic advantages can be realized. Even when the power level is maintained close to what it is in existing power reactors, the annular fuel will greatly enhance safety and facilitate higher burnup of the fuel. [ABSTRACT FROM AUTHOR]

Reactor coolant system (RCS) injection is an important strategy for severe accident management of a pressurized water reactor (PWR) system. Maanshan is a typical Westinghouse PWR nuclear power plant (NPP) with large, dry containment. The severe accident management guideline (SAMG ) of Maanshan NPP is developed based on the Westinghouse Owners Group (WOG) SAMG. The purpose of this work is to analyze the RCS injection strategy of PWR system in an overheated core condition. Power is assumed recovered as the vessel water level drops to the bottom of active fuel. The Modular Accident Analysis Program version 4.0.4 (MAAP4) code is chosen as a tool for analysis. A postulated station blackout sequence for Maanshan NPP is cited as a reference case for this analysis. The hot leg creep rupture occurs during the mitigation action with immediate injection after power recovery according to WOG SAMG, which is not desired. This phenomenon is not considered while developing the WOG SAMG. Two other RCS injection methods are analyzed by using MAAP4. The RCS injection strategy is modified in the Maanshan SAMG. These results can be applied for typical PWR NPPs. [ABSTRACT FROM AUTHOR]