Off-normal Conditions Research Articles

Identifying a probe to visualize the variability of operating teams for supporting the human reliability analysis of nuclear power plants: An explanatory study

Operating teams consisting of several team members still play a critical role in coping with off-normal conditions in socio-technical systems. Thus, various kinds of human reliability analysis methods have been suggested based on the consideration of diverse performance shaping factors that can affect the performance of team members. Unfortunately, since multiple performance shaping factors can vary across operating teams (i.e., crew-to-crew variability), it is crucial to figure out how to visualize this variability in a systematic way. In this regard, comparing the cultural characteristics of operating teams with their performance would be a good starting point. This study investigates how cultural characteristics can be correlated with the occurrence of unsafe acts based on empirical data collected from operating teams working in the main control room of Korean domestic nuclear power plants. The cultural characteristics of the operating teams were visualized using five Hofstede’s cultural indices and compared with the number of unsafe acts observed from simulated off-normal conditions. As a result, a statistically significant correlation is found between the occurrence of unsafe acts and one of the Hofstede’s indices. From this finding, it is expected that a relevant probe to scrutinize crew-to-crew variability could be soundly determined in future works.

ALCYONE: the fuel performance code of the PLEIADES platform dedicated to PWR fuel rods behavior

The aim of this contribution is to deeply introduce the ALCYONE Fuel Performance Code. ALCYONE is being implemented within the PLEIADES numerical framework and is dedicated to the multiphysics behavior of fuel rods in pressurized water reactors (PWRs). It has been a joint development between CEA, EDF and Framatome since 2005 and has been used since then for numerous studies covering normal, off-normal and accidental conditions. However, a detailed description of ALCYONE’s physical and numerical foundations has never been published. In this sense, this paper is intended to be the reference presentation paper for ALCYONE. On the first hand, it provides a comprehensive description of the modeling features and associated multiphysics and multiscale computational schemes. On the other hand, focuses are made on advanced modeling features showing enhanced capabilities of simulation of ALCYONE.

Startup and Multiphysics Analysis of a Compact kW-Class Thermal Spectrum Microreactor

Self-regulating compact nuclear microreactor concepts are being developed for use in space and at terrestrial remote sites. The current emphasis is on the development of high-assay low-enriched uranium–fueled reactors that rely on metal hydrides to achieve size, specific weight, and power parity with the legacy highly enriched uranium systems. In the case of terrestrial applications, metal hydride–moderated systems can also improve economic feasibility. The neutronic design of metal hydride–moderated high-temperature reactor cores is complicated by the need to optimize and synchronize delayed spatial and temporal reactivity feedback from outer reflectors. Zebra is a thermal spectrum proof-of-principle core designed for demonstrating the reactor dynamics of hydride-moderated spectrum reactors, which is studied in this work. The reactor, including the fuel, yttrium hydride moderator, heat pipes, central control rod, structural supports, and the beryllium reflector, weighs ~425 kg. The design shares several features common with nuclear criticality test geometries routinely used in National Criticality Experiments Research Center (NCERC) experiments such that a prototype of the reactor can be readily assembled and tested. Additional features, such as heat pipes and hydrogen barrier clads, were added to minimize the potential for large thermal gradients that in turn could induce hydrogen loss or structural deformation during extended periods of operation at ~1000 K. In this work, normal and off-normal state performance of the Zebra rector core was analyzed using MCNP and Abaqus-based Reactor Multiphysics (MARM) software. Up-to-date nuclear, thermal, and mechanical performance data were used to characterize the performance of the yttrium hydride moderator. Steady-state analyses established that at a postulated power between 20 and 50 kW(thermal), the reactor core is nearly isothermal irrespective of quality of conduction coupling between heat pipes and fuel plates. Additionally, heat pipe failure modes, including simultaneous failure of all heat pipes in a quadrant of a reactor, were examined to bound the maximum credible temperature spike in the reactor core for extreme off-normal operating conditions. Finally, we detail the startup design challenges for the hydride-moderated thermal core, and analyze load-following cases to achieve “self-regulation” using the Dynamic Analysis of Reactor Transients module of MARM. The reactor is self-regulating with a reactivity temperature coefficient of ~−1 pcm/K at the operating point based solely on nuclear cross-section feedback. It can be further strengthened using additional spectral shift neutron absorbers and incorporating design features that enhance core expansion. This work captures that hydride-moderated systems are feasible for various self-regulating applications once systematic checks are verified in order to achieve a well-engineered core design.

COMPARATIVE ANALYSIS OF SOFTWARE TOOLS TO PREDICT THE RADIOECOLOGICAL CONSEQUENCES AFTER ROUTINE AND ACCIDENTAL EMMISIONS BY NPP FOR THE POPULATION

This article provides an overview and comparison of software tools to predict the public radiation effect under normal and off-normal NPP operating conditions. The main characteristics, functional features, advantages and disadvantages of the software packages RESRAD, CROM, PC-CREAM 08®, ESTE Annual Impacts, РОМ, ESTE, NOSTRADAMUS, RECASS, RODOS were addressed. Criteria for assessing the quality and efficiency of software tools were highlighted. The optimal computational toolware applicable to the NPP normal (CROM package) and off-normal (RODOS system) operating conditions was substantiated.

On the behavior of a concrete-based dry cask for spent nuclear fuel in off-normal and accidental conditions

The safe storage of Spent Nuclear Fuel (SNF) within a dry cask system must be guaranteed for any operating condition. In particular, the Peak Cladding Temperature (PCT) is currently employed as a measure of the fuel integrity in normal conditions. The same fuel integrity has to be ensured also in off-normal and accidental conditions by keeping the PCT below the regulatory limits. In this regard, the work proposed in this paper aims at evaluating the performance of a concrete-based dry cask in off-normal and accidental conditions. More specifically, the influence of a partial to complete blockage of the air inlet openings on the PCT is assessed by means of a MELCOR model of the HI-STORM 100S cask. Attention is also paid to the temperature of the external concrete cask in order to check the integrity of the concrete itself.

Amplified spontaneous emission at the band edges of Ag-coated Al nanocone array

Surface lattice resonances (SLRs) with ultra-narrow linewidth (high quality factor) can enhance light–matter interactions at the nanoscale and modulate the propagating light from the emission wavelength direction to efficiency by photonic band engineering. Therefore, SLRs can serve as an excited candidate to enhance and, more importantly, modulate amplified spontaneous emission (ASE) with more optical parameters. Here, this work presents a system of two-dimensional Ag-coated Al nanocone array (Ag-NCA) packaged with Nile red, and a normal ASE with 15-fold enhancement is observed under external driving light. This enhancement fades away, obviously, in the case of the off-normal condition, as the optical feedback evolves from the band edge steady state to the propagating state. The ASE of this hybrid plasmonic system expands the possibilities of interaction between light and matter and has great promise for applications in nanolasing, super-resolution imaging, and photonic integration circuits.

Concept of accident tolerant fuel in nuclear reactors

The excellent behavior of zirconium alloy under normal operating state as well as significant advantage of neutron economy have resulted in the use of zirconium alloy as the cladding material. However, extensive metal-water reaction in the event of severe accidents and subsequent hydrogen generation can be considered to be a big disadvantage for zirconium alloy. The Fukushima Daiichi Nuclear Plant accident has brought challenges to the nuclear industry and renewed attention about the idea of accident tolerant fuels (ATFs). Concept of ATFs is to mitigate the devastating interaction of steam at high temperature with fuel clad in off-normal condition of the reactor. These materials have higher resistance to core degradation, lower oxidation kinetics and can withstand against high temperature for a longer duration of time. Being in the direct contact with the coolant the accident tolerant fuel cladding plays an important role at high temperature in off-normal conditions. Many studies have been initiated worldwide to find an advanced oxidation-resistant alternative cladding material to enhance fuel performance and safety, specifically under accident conditions. One of the candidate materials for fuel cladding in LWRs is oxidation-resistant iron alloys. The stainless steel technology and the material science have been studied extensively and several new generations of increasingly higher performance steels are now commercially available that may offer significant performance improvements over the relatively simple austenitic steels that were utilized as cladding in some early commercial fission reactors. One study has been performed to find the reactivity impact of SiC and stainless steel as clad material on various parameters. This paper presents the results of a study that has been made to understand the reactor physics implications of using SiC and stainless steel as the cladding material.

Multi-parameter optimization of gamma emission tomography instruments for irradiated nuclear fuel examination

Material test reactors have an extended use in irradiation testing of novel nuclear fuel materials and the fuel behavior in off-normal conditions. The performance of the nuclear fuel is examined in in-pile and out-of-pile post-irradiation examinations (PIEs), e.g., using Gamma Emission Tomography (GET). GET is a nondestructive assay that images the internal spatial distribution of gamma-emitting nuclides built up in the fuel due to irradiation. Since GET can be performed close to the reactor and without intrusion in the fuel object, it can potentially speed up the data generation from PIE in irradiation testing.The performance metrics of GET devices can be identified regarding time requirements, noise in the reconstructed image, signal-to-background ratio, and spatial resolution. However, these are complicated to determine, partly due to inherent trade-offs between the metrics themselves, partly because they depend on the fuel activity and its spectrum (i.e., object dependent), and, finally, on the GET setup and its configuration.This work proposes a structured methodology for optimizing the collimator design for a new generation of GET tomography setups, intending to improve spatial resolution by one order of magnitude: from the millimeter scale to the hundred-micron scale. The conflicting performance metrics are determined based on the controllable parameters of the GET setup and the uncontrollable parameters of an anticipated fuel object, able to provide a signal-to-background ratio above a customized limit, chosen based on the specific application. The trade-off between the performance remaining metrics is then visualized by a Pareto approach, where dominated solutions are rejected. Finally, constraints on noise level and measurement time are used to find the optimal spatial resolution, without applying noise suppression filters.Two GET setups are presented using the outlined method. Firstly, to upgrade the tomography test bench BETTAN at Uppsala University, a new segmented HPGe detector is planned to be used with low-activity fuel rod mock-ups. Secondly, a GET system for investigating high-activity nuclear fuel rods of representative burnup. For a nuclear fuel inspection, the results showed that a spatial resolution of about 300 μm is possible with reasonable noise and measurement time constraints.

Numerical study of the impact of blockages on the Thermal-Hydraulic performance of 19 and 61-pin Wire-Wrapped bundles with conjugate heat transfer using RANS

Liquid metal fast reactors assemblies with wire spacers present a compact rod pitch to diameter ratio. Therefore, a typical off-normal condition that may occur is the blockage accident, caused by the accumulation of corrosion products or due to the detachment of broken wires. Blockage accidents may threaten the safety performance, leading to fuel rod damage. This work presents a thermal–hydraulic analysis using the RANS k-ε Realizable model to assess the impact of blockages. The model was validated on a blocked 19-pin wire-wrapped assembly comparing cladding and coolant temperatures and pressure drop with experimental data. Then, a numerical comparison between nominal and blocked conditions on a 61-pin wire-wrapped bundle was performed, analyzing solid and porous blockages. The greatest peak temperatures were obtained for the solid blockage case. Both types of blockages resulted in a similar velocity prediction upstream of the obstruction, while major differences were found in the recirculation region.

High power laser heating of nuclear ceramics for the generation of controlled spatiotemporal gradients

The use of high-power lasers to reach the high temperatures typical for off-normal conditions in nuclear reactor fuel or to determine its properties at high temperature is a relevant and proven technique in the nuclear research community. Conducting studies at high temperature with full control in space and time of the thermal gradients is, however, a challenge. In this paper, we present a combination of methodologies, experiments, and models to drive nuclear fuel samples in thermal conditions relevant to the study of nuclear fuel materials. Based on the use of high-power (kW) lasers with beam shaping and temporal laser control, our experimental approach allows one to heat depleted UO2 samples and obtain precise control of the temporal gradients, in the ms range, and spatial gradients, in the 100 μm range. The coupling with numerical simulations allows one to determine the temperature distribution in the depth of the sample and to develop a thermo-mechanical interpretation of the results. We will present the methodology allowing one to properly compare experimental and simulation results by taking into account the optical response of the instrumentation, the laser–UO2 interaction, and the resulting heat source term. We will show that we can express the heat source term resulting from the laser loading either by an analytical description or statistically based on Monte-Carlo simulation. A fairly good agreement between the recorded temperature and the simulation results is shown. The presented methodology can be extended to other materials with a proper choice of laser wavelength and instrumentation based on the optical properties of the investigated material.

Effects of elevated temperature exposure on the residual stress state and microstructure of PVD Cr coatings on SiC investigated via in situ X-ray diffraction and transmission electron microscopy

Pure Cr coatings are being considered as corrosion-resistant layers for SiC-SiCf composite fuel cladding in light water reactors. The residual stress state and interdiffusion behavior of Cr coatings deposited on SiC substrates using bipolar high-power impulse magnetron sputtering (B-HiPIMS) and direct current impulse magnetron sputtering (DCMS) were studied using in situ X-ray diffraction experiments at 300 °C, 700 °C, and 1200 °C. After elevated temperature exposure, evolution of the coating microstructure and coating/substrate interdiffusion was characterized by transmission electron microscopy. The compressive residual stress of the dense B-HiPIMS coating at room temperature was gradually relieved at 300 °C, while the stress transitioned to a tensile state at 700 °C. Grain growth of the Cr coating and a Cr–SiC interdiffusion layer consisting of newly formed Cr–C and Cr–Si phases were observed in the B-HiPIMS coating after exposure at 700 °C. The phases in the interdiffusion layer were identified to be Cr7C3, Cr3Si, and Cr5Si3Cx experimentally and by thermodynamic predictions. For the DCMS coating, however, changes in residual stress and microstructure at the elevated temperatures were limited by the inherent porosity and inhomogeneous structure of the as-deposited coating. At 1200 °C, both coating types interacted rapidly with the underlying SiC substrate in the early stages of the experiment. The results from this study suggest that the characteristics of the Cr coatings do not significantly change at the reactor operating temperature of 300 °C, while changes in microstructure and stress state are expected in off-normal conditions of 700 °C and 1200 °C. The results from this study suggest the B-HiPIMS coating is best-suited as a protective coating for SiC fuel cladding.

Microstructural analysis of tristructural isotropic particles in high-temperature steam mixed gas atmospheres

High-temperature gas-cooled reactors (HTGRs) use tristructural isotropic (TRISO) particles embedded in a graphitic matrix material to form the integral fuel element. Potential off-normal reactor conditions for HTGRs include steam ingress with temperatures above 1,000 °C. Fuel element exposure to steam can cause the graphitic matrix material to evolve, forming an atmosphere composed of oxidants and oxidation products and potentially exposing the TRISO particles to these conditions. Investigating the oxidation response of TRISO particles exposed to a mixed gas atmosphere will provide insight into the stability under off-normal conditions. In this study, surrogate TRISO particles were exposed to high temperatures (T = 1,200 °C) in flowing steam (3% < pH2O < 21%) and CO (pCO < 1%) to determine the oxidation behavior of the SiC layer when exposed to various mixed gas atmospheres. Scanning electron microscopy, x-ray diffraction, and focused ion beam milling was used to determine the impact of CO and steam on the oxidation behavior of the SiC layer. The data presented demonstrates how the SiC layer showed strong oxidation resistance due to limited SiO2 growth and maintained its structural integrity under these off-normal conditions.

Multiphysics simulation of VVER-1200 fuel performance during normal operating conditions

Nuclear fuel performance modeling and simulation are critical tasks for nuclear fuel design optimization and safety analysis under normal and transient conditions. Fuel performance is a complicated phenomenon that involves thermal, mechanical, and irradiation mechanisms and requires special multiphysics modules. In this study, a fuel performance model was developed using the COMSOL Multiphysics platform. The modeling was performed for a 2D axis-symmetric geometry of a UO2 fuel pellet in the E110 clad for VVER-1200 fuel. The modeling considers all relevant phenomena, including heat generation and conduction, gap heat transfer, elastic strain, mechanical contact, thermal expansion, grain growth, densification, fission gas generation and release, fission product swelling, gap/plenum pressure, and cladding thermal and irradiation creep. The model was validated using a code-to-code evaluation of the fuel pellet centerline and surface temperatures in the case of constant power, in addition to validation of fission gas release (FGR) predictions. This prediction proved that the model could perform according to previously published VVER nuclear fuel performance parameters. A sensitivity study was also conducted to assess the effects of uncertainty on some of the model parameters. The model was then used to predict the VVER-1200 fuel performance parameters as a function of burnup, including the temperature profiles, gap width, fission gas release, and plenum pressure. A compilation of related material and thermomechanical models was conducted and included in the modeling to allow the user to investigate different material/performance models. Although the model was developed for normal operating conditions, it can be modified to include off-normal operating conditions.

Simulation of a TRISO MiniFuel irradiation experiment with data-informed uncertainty quantification

An irradiation experiment using tristructural isotropic (TRISO) fuel particles and the miniature fuel (MiniFuel) irradiation vehicle was performed in Oak Ridge National Laboratory’s High Flux Isotope Reactor (HFIR) to support development of the Kairos Power fluoride salt–cooled, high-temperature reactor (KP-FHR). This paper describes modeling predictions of temperatures and fuel burnup for the as-built experiment. An uncertainty quantification (UQ) analysis was performed to determine the effect of TRISO particle volume and position on the temperature predictions at various fuel heat generation rates (HGRs). This UQ study utilized fuel kernel position and volume measurements previously collected using X-ray computed tomography (XCT) techniques and Monte Carlo sampling methods to generate fuel compact cases that were then analyzed using a finite element thermal model. The UQ analysis indicated that uncertainty in calculated temperatures caused by varying TRISO particle arrangement is relatively small, even at high fuel HGR. Final predictions of particle temperatures throughout the irradiation are shown to be relevant to KP-FHR normal and off-normal operating conditions and to previous TRISO irradiation experiments. The combination of XCT with UQ analyses will inform post-irradiation examination (PIE) of the irradiated fuel compacts, and these analyses can be used to develop fuel performance models for coated particle fuel forms. Both PIE of separate-effects irradiation data and enhanced fuel performance modeling support accelerated qualification of TRISO fuels for a broad range of advanced reactor applications. The novel approach demonstrated here of measuring TRISO particle configurations with XCT methods and generating representative fuel compacts for finite element modeling and UQ analysis could be leveraged by the broader particle fuel community in the development of other TRISO fuel experiments in which these variables may have a significant impact on key outcomes.

Thermal Analysis of Vertical Dry Storage Cask for Nuclear Spent Fuel in Off-Normal Conditions

One of the key aspects for safety assessment of spent fuel dry storage systems is evaluation of the temperature profile for spent fuel in various storage conditions. In this effort, a three-dimensional model of a vertical dry storage HI‑STORM 190 UA system for VVER-1000 fuel is developed for ANSYS CFX code. This model is then used for the analysis of spent fuel thermal state in normal conditions and in a hypothetical case of a loss of the fuel canister integrity. Canister depressurization with helium leakage and its substitution with air lead to degradation of heat removal and increase of spent nuclear fuel temperature.&#x0D; The results presented in this article demonstrate that the maximum cladding temperature of spent fuel does not exceed the design limit in normal storage conditions. In the case of loss of the fuel canister integrity, the maximum cladding temperature exceeds this design limit for normal storage conditions, but remains below the limit for short-term off-normal conditions. The results also demonstrate that loss of the fuel canister integrity leads to insignificant change in vents outlet temperature and cannot be reliably identified by temperature monitoring according to the existing procedures.

Validation of the TWOPORFLOW code for the core analysis of liquid metal-cooled reactor with selected experiments

TWOPORFLOW (TPF) is a thermal–hydraulic code that utilizes a porous medium approach in three-dimensional Cartesian coordinates. It can simulate the flow phenomenon in fuel assemblies or reactor cores by solving three conservation equations (single-phase) or six conservation equations in a two-fluid liquid–vapor approach. Many constitutive correlations e.g. pressure loss, interfacial friction, and heat transfer (wall, liquid–vapor interface), are included to close the system of equations. The code was recently extended to simulate pre-Critical Heat Flux (CHF) and CHF heat transfer relevant for Boiling Water Reactor (BWR). In this paper, the code's simulation capability for liquid metal-cooled reactors core is validated using three experimental data sets: pressure drop of the 19-pin rod in THEADES LBE loop, the BREST-type reactor benchmark problem, and the NACIE-UP Block Fuel Pin Simulator (BFPS) test. The results show that TPF can be used not only for square rod bundle arrangements but also for hexagonal arrangements; the code can reasonably predict the thermal–hydraulic behavior of liquid metal-cooled reactors under normal and off-normal conditions.

Towards grain-scale modelling of the release of radioactive fission gas from oxide fuel. Part I: SCIANTIX

When assessing the radiological consequences of postulated accident scenarios, it is of primary interest to determine the amount of radioactive fission gas accumulated in the fuel rod free volume. The state-of-the-art semi-empirical approach (ANS 5.4–2010) is reviewed and compared with a mechanistic approach to evaluate the release of radioactive fission gases. At the intra-granular level, the diffusion-decay equation is handled by a spectral diffusion algorithm. At the inter-granular level, a mechanistic description of the grain boundary is considered: bubble growth and coalescence are treated as interrelated phenomena, resulting in the grain-boundary venting as the onset for the release from the fuel pellets. The outcome is a kinetic description of the release of radioactive fission gases, of interest when assessing normal and off-normal conditions. We implement the model in SCIANTIX and reproduce the release of short-lived fission gases, during the CONTACT 1 experiments. The results show a satisfactory agreement with the measurement and with the state-of-the-art methodology, demonstrating the model soundness. A second work will follow, providing integral fuel rod analysis by coupling the code SCIANTIX with the thermo-mechanical code TRANSURANUS.

Uncertainty and Sensitivity Analysis of a Dry Cask for Spent Nuclear Fuel

Nuclear safety relies to a good extent on thoroughly validated codes. However, code predictions are affected by uncertainties that need to be quantified for a more accurate evaluation of safety margins. In this regard, the present paper proposes a preliminary uncertainty and sensitivity analysis of the thermal behavior of a concrete-based dry cask for spent nuclear fuel storage, employing the MELCOR code and a series of MATLAB scripts. As thermal behavior is of utmost importance for the fulfillment of United States Nuclear Regulatory Commission (USNRC) safety requirements, the Peak Cladding Temperature (PCT) has been addressed as the key Figure of Merit (FOM). Variables related to the main heat transfer mechanisms have been selected as input parameters for the uncertainty quantification, whereas heat source and heat sink, namely decay power and external air temperature, have been dealt with in a separate sensitivity analysis. The results show that the selected parameters have a weak influence on the PCT, whereas it is strongly related to the decay power and external air temperature values. In any case, PCT stays below the regulatory threshold even under the considered off-normal conditions.

An analytic approach to the design of passively safe lead-cooled reactors

A methodology to assist the design of liquid metal reactors, passively cooled by natural circulation during off-normal conditions, is derived from first principle physics. Based on this methodology, a preliminary design of a small LFR is accomplished and presented with accompanying neutronic and reactor dynamic characterizations. The benefit of using this methodology for reactor design compared to other available methods is discussed.

3D-FOX—A 3D Transient Electromagnetic Code for Eddy Currents Computation in Superconducting Magnet Structures: DTT TF Fast Current Discharge Analysis

Bulky metallic structures are needed in the toroidal field (TF) superconducting magnets for fusion applications to withstand the large Lorentz forces acting on the winding. The pulsed coil operation during a plasma scenario and the fast current discharge, or a plasma disruption event in off-normal operating conditions, cause transient magnetic fields, inducing eddy currents in the TF structures. The eddy currents generate heat in the structures heating in turn the winding pack, eroding the temperature margin: such power deposition is a key input for thermal-hydraulic (TH) analyses. However, the computation of eddy currents in fusion magnets is a challenging topic since a transient, fully 3D electromagnetic (EM) model is required. The EM problem is solved here by means of the finite element (FE) open source code FreeFEM++. First, the correct implementation of the EM problem is verified by means of suitable benchmarks against both simple analytical cases and the results obtained with state-of-the-art FE commercial codes on the DTT TF coil, used as a reference geometry. Then the EM code is applied to the evaluation of the magnetic fields and eddy currents induced in the same reference coil during the normal (static) and off-normal (transient) operation; the output of the EM analysis is used as input to the TH analysis carried out with the 4C code, aimed at computing the temperature margin evolution during the transient.