Power Excursion Research Articles

Understanding and Reducing HVAC Power Consumption Post-Evacuation Events in Commercial Buildings

Buildings are required to follow standard operational procedures during emergency evacuation. In addition to people evacuating the building, one of the recommended steps during a fire evacuation is to shut down the air handling units (AHUs) of the heating, ventilation, and air conditioning (HVAC) system to prevent smoke from spreading in the building via the air ducts. Shutting down the AHU will inevitably cut-off cooling, resulting in internal temperatures rising steeply particularly on hot days. This phenomenon imposes considerable power demand on the HVAC to rapidly cool the building down during reoccupation. In this article, we study the energy implications of post-evacuation scenarios. Our contributions are threefold: 1) we quantify power excursion caused in 43 evacuation events across 14 buildings of a university campus using a data-driven building thermal model. We show evacuations during summer season can result in power consumption up to 150% above the power demand threshold; 2) we develop a method to reschedule planned evacuations in order to eliminate the power excursions while adhering to building evacuation standards; and 3) we develop a formal optimization framework to minimize the energy costs during planned and emergency evacuations without compromising the desired thermal comfort temperatures by intelligently cooling the building post evacuation. This is the first study to understand and reduce the HVAC power consumption associated with building evacuation events.

A study of the behavior of the floating absorber for safety at transient (FAST) in an innovative sodium-cooled fast reactor (iSFR)

Negative reactivity feedback is one of the most important inherent safety characteristics of reactors for suppressing changes in thermal power and component temperatures during accidents. The floating absorber for safety at transient (FAST) has been suggested to improve the coolant temperature coefficient (CTC) in a sodium-cooled fast reactor (SFR) for a better self-stabilization of the reactor system. Previous studies on the FAST application showed that the FAST causes power excursion and temperature oscillations under anticipated transient without scram (ATWS) conditions. This study explains the variables affecting the FAST behavior and oscillations with mathematical expressions. In addition, a damping FAST is proposed to passively remove oscillations without any mechanical system. The damping FAST increases the damping ratio by reducing the pin diameter at a height where the difference between the FAST bottom elevation and the equilibrium position is maximum. The performance evaluation for the damping FAST is conducted in an innovative sodium-cooled fast reactor (iSFR) under ATWS conditions. The 7-mm damping FAST unilaterally inserts the negative reactivity in unprotected transient overpower event, dramatically lowering the core power without oscillations. While the previous FAST shows oscillatory behavior repeatedly approaching the coolant boiling temperature, the damping FAST stably keeps the coolant temperature below 800 °C. Moreover, the damping FAST lowers the maximum coolant temperatures for other ATWS events. These findings show the potential of the FAST system to significantly improve the inherent safety of SFRs.

Experimental and numerical investigations into molten-pool sloshing motion for severe accident analysis of sodium-cooled fast reactors: A review

Safety issues are particularly crucial for sodium-cooled fast reactors (SFRs). Data obtained from SFR safety analyses over recent years have shown that a specific type of sloshing motion probably occurs in the molten pool during core disruptive accidents (CDAs) of SFRs due to local neutronic power excursion or pressure developments, thereby significantly influencing recriticality. Recognizing the importance of improving the evaluation of CDAs of SFRs, extensive knowledge about this phenomenon has been garnered through experimental studies of their thermal-hydraulic mechanism and characteristics. Based on these studies, simulations using various numerical approaches, such as SIMMER code, the finite volume particle method, and the smoothed particle hydrodynamic method, have attempted to reproduce the sloshing motion under various experimental conditions to verify their reasonability and applicability, thereby promoting the development of SFR safety analysis. To provide useful references for future SFR safety analyses and assessments, we have systematically reviewed and summarized these experimental and numerical investigations into the thermal-hydraulic aspect of molten-pool sloshing motion. In addition, to enhance deeper and more comprehensive research into sloshing motion, we have also discussed future prospects. Knowledge gained from experimental and numerical investigations into molten-pool sloshing motion is valuable not only for improving and verifying SFR safety analysis codes but also for providing reference data for studies of sloshing motion in other fields of engineering.

Design of separate-effects In-Pile transient boiling experiments at the TREAT Facility

The cladding-to-coolant heat transfer during rapid power excursions, such as reactivity-initiated accidents, remains a crucial area of uncertainty in nuclear reactor safety. This uncertainty impacts the ability to accurately predict fuel performance behavior for these conditions. Improving our understanding of transient cladding-to-coolant heat transfer will enhance our ability to model design-basis accidents and could increase design and safety margins for the current commercial fleet and advanced reactors. To address these issues, the Critical Heat Flux Static Environment Rodlet Transient Test Apparatus experiment was designed and uses a novel approach with a borated stainless-steel heater rodlet that can replicate cladding temperatures experienced during a design-basis reactivity-initiated accident. The design of the rodlet allows for separate-effect testing that eliminates complexities with a fuel and cladding specimen. The rodlet and experiment is instrumented to provide temperature and thermal–hydraulic conditions throughout the transient. The experiment results show that the current correlations in RELAP5-3D are conservative and predict much higher temperatures than measured during the experiments. The data from the experiments will be used to improve our models and understanding of boiling behavior, specifically critical heat flux, under transient conditions.

Spectral and Spatial Power Evolution Design With Machine Learning-Enabled Raman Amplification

We present a machine learning (ML) framework for designing desired signal power profiles over the spectral and spatial domains in the fiber span. The proposed framework adjusts the Raman pump power values to obtain the desired two-dimensional (2D) profiles using a convolutional neural network (CNN) followed by the differential evolution (DE) technique. The CNN learns the mapping between the 2D profiles and their corresponding pump power values using a data-set generated by exciting the amplification setup. Nonetheless, its performance is not accurate for designing 2D profiles of practical interest, such as a 2D flat or a 2D symmetric (with respect to the middle point in distance). To adjust the pump power values more accurately, the DE fine-tunes the power values initialized by the CNN to design the proposed 2D profile with a lower cost value. In the fine-tuning process, the DE employs the direct amplification model which consists of 8 bidirectional propagating pumps, including 2 s-order and 6 first order, in an 80 km fiber span. We evaluate the framework to design broadband 2D flat and symmetric power profiles, as two goals for wavelength division multiplexing (WDM) system performing over the whole C-band. Results indicate the framework’s ability to achieve maximum power excursion of 2.81 dB for a 2D flat, and maximum asymmetry of 14% for a 2D symmetric profile.

Timing of propulsion-related biomechanical variables is impaired in individuals with post-stroke hemiparesis

BackgroundIn individuals with post-stroke hemiparesis, reduced paretic leg propulsion, measured through anterior ground reaction forces (AGRF), is a common and functionally-relevant gait impairment. Deficits in other biomechanical variables such as plantarflexor moment, ankle power, and ankle excursion contribute to reduced propulsion. While reduction in the magnitude of propulsion post-stroke is well studied, here, our objective was to compare the timing of propulsion-related biomechanical variables. Research questionAre there differences in the timing of propulsion and propulsion-related biomechanical variables between able-bodied individuals, the paretic leg, and non-paretic leg of post-stroke individuals? MethodsNine able-bodied and 13 post-stroke individuals completed a gait analysis session comprising treadmill walking trials at each participant’s self-selected speed. Two planned independent sample t-tests were conducted to detect differences in the timing of dependent variables between the paretic versus non-paretic leg post-stroke and paretic leg versus the dominant leg of able-bodied individuals. ResultsPost-stroke individuals demonstrated significantly earlier timing of peak AGRF of their paretic leg versus their non-paretic leg and able-bodied individuals. Post-stroke participants displayed earlier timing of peak power of their paretic leg versus their non-paretic leg and able-bodied individuals, and earlier timing of peak ankle moment of the paretic leg versus able-bodied. No significant differences were detected in the timing of peak ankle angle. SignificanceThe earlier onset of peak AGRF, peak ankle power, and peak ankle moment may be an important, under-studied biomechanical factor underlying stroke gait impairments, and a potential therapeutic target for stroke gait retraining. Future investigations can explore the use of interventions such as gait biofeedback to normalize the timing of these peaks, thereby improving propulsion and walking function post-stroke.

An accelerator–driven nuclear reactor

Abstract The basic parameters of the accelerator facility for a proton beam of 100 mA and 1 GeV using a superconducting separate orbit cyclotron, are given. By means of such an accelerator system one can activate a subcritical 3 GW thermal power nuclear reactor of the well known RMBK–type 1000. The system is stable against any “Chernobyl–type” power excursion. When this accelerator system will be coupled to existing RMBK–type 1000 power stations, the resulting electricity production would become substantially more safe. Associated problems and design parameters are discussed.

Mathematical modeling of point kinetic equations with temperature feedback for reactivity transient analysis in MTR

Abstract The behavior of the nuclear reactor in response to any sudden change in reactivity is very important for reactor control. Positive reactivity insertions causes power excursion and could have a destructive impact on the reactor core. The aim of the study is to investigate the safety features of a material test reactor (MTR) during reactivity transient with emphasis on the capability of the mathematical modeling using programming language. Therefore a mathematical model using Python3.6; high-level programming language is developed to solve the point kinetic equations taking into account Doppler and moderator feedback effects. The model is validated with AIREKMOD_RR; point kinetic computer code for reactivity transient analysis in nuclear research reactors. The results of the Python model demonstrate the inherent safety features of the MTR reactor. Also, there is good agreement between the results of the Python model and AIREKMOD_RR code, illustrating the efficiency of the Python model in simulating the behavior of the reactor core under reactivity transient.

Reinforcement Learning for Compensating Power Excursions in Amplified WDM Systems

Wavelength-dependent power excursions in gain-controlled erbium doped fiber amplifiers (EDFA) is a challenging issue in optical networks. We investigate a launch channel power control method using reinforcement learning (RL) to mitigate the power excursions of EDFA systems. A machine learning engine is developed, trained and evaluated with four different policy-gradient RL algorithms that are compared according to two main criteria: achieved power excursion reduction and learning time. Different scenarios are considered with 12-, 24-, 40- active channels at fixed wavelengths and with variable number of active channels (between 12 and 64) assigned randomly at different wavelengths during RL process. We show 62% power excursion reduction in the 40-channel scenario and 28% in the variable scenario, which demonstrates the promising role of online RL approach for controlling power excursion in EDFA systems.

Study on High-Fidelity Thermal-Neutronic Coupling Method Based on the Unified Geometry Modeling and its Application in Experimental Reactor Core Calculation for SPERT

To solve the problem that the thermal-neutronic grid mapping relation is complicated and cannot be preset in a centralized manner due to the existence of irregular geometry in performing the thermal-neutronic coupled simulation calculation for various small power reactors and experimental reactors, this paper studies the thermal-neutronic coupling method based on the unified geometric modeling, using the high-fidelity numerical code for reactor neutronics calculation, NECP-X. This study establishes the mapping relation for the thermal-neutronic coupling on the basis of the neutronics model, and enables the direct transient calculation of the experimental reactor core for the special power excursion reactor test (SPERT) via combination with the transient calculation method in NECP-X. Then, this study calculates the steady-state case for the experimental reactors of SPERT, and compares the calculation results with the results gained from the Monte Carlo code. On this basis, this study conducts transient calculation and analysis for these experimental reactors and compares the corresponding results with the experimental results. The final results show that the eigenvalues from the neutronics calculation by the NECP-X and the rod power distribution calculation results are of high accuracy; that the grid mapping method based on the unified geometric modeling allows a easy and fast thermal-neutronic coupled calculation of the PWRs of complex geometry; and that compared with the experimental values, the curve of change in the total power and reactivity gained from transient calculation with time is more accurate and can provide refined distributions of power and temperature.

Dynamical System Scaling Application to Zircaloy Cladding Thermal Response During Reactivity-Initiated Accident Experiment

New fuel design and development currently require 20 to 25 years to be qualified for use by the nuclear power industry. The thermal-hydraulics community has taken advantage of scaling theory to design reduced-scale experiments that correctly preserve dominant key phenomena while quantifying distorted phenomena. These techniques can be leveraged in the design and analysis of fuel performance experiments to help reduce the timeline associated with fuel design and development. This study uses the Dynamical System Scaling (DSS) method to analyze cladding temperature data from the recent SETH-C experiment in the Transient Reactor Test Facility (TREAT) and accompanying BISON simulations to assess dynamic distortions occurring throughout the fast power excursion transient. The DSS analysis revealed that on the cooldown from peak cladding temperature, the fuel radial power profile is the most sensitive modeling parameter, with a heterogeneous radial peaking factor corresponding to the lowest distortion compared to a uniform energy deposition. For the heatup to PCT, the heterogeneous radial power profile corresponded to the shortest process action. Last, for the heatup to PCT, the gap conductance model sensitivity was quantified using process actionsm and showed that the default light water reactor gap conductance model corresponded to the longest process action.

Recent developments in PSI BEPU analysis for SPERT-III RIA transient: ND uncertainty breakdown and kinetic parameters uncertainty assessment

The current research summarizes recent developments performed within the PSI BEPU methodology and the corresponding results on the SPERT-III RIA experiments. In this regard, three main studies have been carried out. First, an uncertainty breakdown study has been performed in order to identify the dominant nuclear data and to quantify their respective contributions to the overall uncertainties of the parameters of interest. On the other hand, the second study was devoted to assess the impact of considering the effect of cross section perturbation (limited for the time being to U-238 capture) in the resonance self-shielding calculation, known as the implicit effect, on different parameters’ uncertainties in SPERT-III RIA transient. Finally, the third study was conducted with the main goal to assess the impact of kinetic parameter uncertainties, such as delayed neutron yields and associated decay constants, on the overall uncertainties.The uncertainty quantification results of the first study, based on the ENDF/B-VII.1 library, showed that, for steady-state, the first dominant contributor to the overall uncertainties of the static reactivity worth and k-eff is U-235 nubar, followed by U-238 then U-235 capture cross sections. However, for the transient, the first dominant contributor to the overall uncertainties, in e.g. total power and reactivity, is the U-238 inelastic scattering. More interestingly, it was found that, the second dominant contribution to the overall power uncertainty, during the initial excursion phase, is due to the U-235 nubar, while during the power reversal phase, the U-238 capture becomes the second dominant contribution, and even the first one by the end of the transient. Moreover, results of the second study demonstrated a clear systematic decrease of the uncertainty of all parameters due to the inclusion of the implicit effect. In steady state, due to the impact of the implicit effect, the Doppler coefficient relative uncertainty could be reduced up to 30%, which is reflected in the transient, especially in the reversal phase of the power excursion where Doppler effect plays a key role, by a systematic decrease of the relative uncertainty of total power and other parameters. Finally, in the third study, the estimated uncertainties, regarding both steady-state and transient parameters, have shown systematic larger values due to kinetic parameter uncertainties compared to those due to the other nuclear data.

Patchy nuclear chain reactions

Stochastic fluctuations of the neutron population within a nuclear reactor are typically prevented by operating the core at a sufficient power, since a deterministic (i.e., exactly predictable) behavior of the neutron population is required by automatic safety systems to detect unwanted power excursions. In order to characterize the reactor operating conditions at which the fluctuations vanish, an experiment was designed and took place in 2017 at the Rensselaer Polytechnic Institute Reactor Critical Facility. This experiment however revealed persisting fluctuations and striking patchy spatial patterns in neutron spatial distributions. Here we report these experimental findings, interpret them by a stochastic modeling based on branching random walks, and extend them using a “numerical twin” of the reactor core. Consequences on nuclear safety will be discussed.

Direct whole core modeling and simulation of the SPERT III E-Core experiments by nTRACER

The SPERT III E-core experiments are modeled and simulated by a direct whole core calculation code nTRACER as an effort to validate its dynamics calculation capability. The control rod ejection experiments performed for the E-core of the Special Power Excursion Reactor are faithfully modeled by explicitly representing all the complex components such as fuel cans and the cruciform transient rod. The unknown experiment conditions such as the initial position of the ejected rod is estimated rigorously to yield the proper rod worth. The simulations are performed in the one-step manner such that the entire calculation for rod ejection is carried out with a single input deck without any prior calculation. All the kinetics data such as the fractions and decay constants of the delayed neutron precursors are directly obtained from the suitable evaluated nuclear data files. Doppler feedback is modeled by the direct estimation of the fuel temperature profiles inside each pellet. The transient results for the five ejection tests are compared with the measured data such as the peak time and magnitude of the power pulse, the net power release up to the power peak, and the post-burst power level. The good agreement between the nTRACER results and measured data demonstrates the efficacy of direct whole core dynamics calculation as well as the accuracy of nTRACER. In addition, a comparison is made with a two-step calculation to demonstrate the advantage of the direct whole core transient calculation.

Thermal-hydraulic two-phase modeling of reactivity-initiated transients with CATHARE2 – Application to SPERT-IV simulation

Reactivity-Initiated Accidents (RIA) in nuclear reactor cores are very complex multiphysic transients. They consist in a very fast power excursion in the core, leading to wall temperature excursions. Many experiments have shown that the heat exchanges coefficients during fast transients differ from those in steady-states. The thermal-hydraulic phenomenology of such transients remains complex and some experimental studies allowed to better understand and quantify the heat exchanges between the wall and the fluid, for both single and two-phase flows.The current study lies in a continuous effort of developing and validating a coupled neutronic and thermal-hydraulic code for the analysis of protected and unprotected transient behavior of reactors. More specifically, this study deals with the extension of the thermal hydraulic model of the CATHARE2 code to fast transient configurations. The extended CATHARE2 model is validated against experimental results from SPERT-IV. This reactor was driven mainly by the coolant density reactivity feedback. Thanks to that, many flow regimes can be observed in the core during the transient. That allows a more accurate evaluation of the adequacy of available transient two-phase flow models and correlations. The results, that represent the state of modeling, allow concluding that introduced models in CATHARE2 are able to simulate RIA tests such as SPERT-IV involving fast transient boiling.

Performance Evaluation of Digital Back-Propagation in the Presence of Power Excursion

An accurate power map is mandatory to evaluate the performance of optical fiber transmissions in the nonlinear transmission region. In this letter, we discuss the performance variation in dispersion unmanaged fiber links due to power excursions, overarching the fiber nonlinearity that imposes a fundamental limit on the capacity of optical fiber transmissions. After using digital back-propagation (DBP) in optical fiber transmissions, the performance degradation due to power excursions shows different characteristics, depending on the location of the power excursion. Further, due to the cascading effect of the nonlinear signal-noise interaction (NSNI), the power excursion on the latter spans exerts a greater impact than that on the front spans. We implement a polarization-multiplexed 16-QAM over 12 spans of standard single-mode fiber (SSMF) experiment. The experimental results verify our theoretical analyses that the received SNRs after DBP can be sensitive to the power excursion and its location.

Modified Unipolar Latissimus Transfer to Restore Elbow Flexion in Musculocutaneous Nerve Palsy

Modified Unipolar Latissimus Transfer to Restore Elbow Flexion in Musculocutaneous Nerve Palsy

Power Evolution Modeling and Optimization of Fiber Optic Communication Systems With EDFA Repeaters

In this article, machine learning is used to create a differentiable model for the input-output power spectral profile relations of C-band erbium-doped fiber amplifiers (EDFAs). The EDFA model is demonstrated to generalize to multiple physical devices of the same make while only trained on experimental data from a single unit. The model is combined with a differentiable model for simulating stimulated Raman scattering (SRS) effects during propagation through the the optical fiber to create a differentiable model for a multi-span system with an arbitrary configuration of number of spans, length per span and launch power per span. The cascade system model is used to predict and optimize the power profile output of several such experimental configurations of up to three spans with an arbitrary target power profile. A flat target profile is exemplified experimentally, achieving <; 3 dB of power excursions for EDFAs exhibiting $>$10 dB of excursion per device in the cascade. The experimental data used to create the EDFA model is made public and available online.

A Nodal Model for Tokamak Burning Plasma Space-Time Dynamics

Fusion alpha heating introduces new phenomena into plasma dynamics and control. On the worrisome side is the well-known fact that the dependence of the predominantly central fusion heating mechanism, coupled with the less appreciated fact of the predominantly plasma edge location of bremsstrahlung and impurity line and recombination radiation cooling and of ion orbit loss cooling, suggests the possibility of a thermal runaway fusion power excursion in the plasma core. On the encouraging side is the fact that the fusion alpha energy is transferred first to heat the core electrons and produce electron cyclotron radiation that is transferred instantaneously, predominantly to outer plasma regions and the surrounding material wall, reducing its availability for further heating of core plasma ions. This paper discusses the temporal and spatial dependence of the various heating and cooling mechanisms involved in the burn dynamics of a fusion plasma, introduces a spatially coarse nodal space-time calculation model (suitable for dynamics and controller calculations) for the analysis of burning plasmas, and identifies the research needed to fully evaluate the parameters of such a model.

ANALYSIS OF REACTIVITY INSERTION AS A FUNCTION OF THE RSG-GAS FUEL BURN-UP

Analysis of the control rod insertion is important as it is closely related to reactor safety. Previously, the analysis has been carried out in RSG-GAS during static condition, not as a function of the fuel fraction. The RSG-GAS reactor in one cycle is a function of the fuel burn-up. It is necessary to analyze RSG-GAS core reactivity insertion as a function of the fuel burn-up to determine the behavior of the reactor, especially in uncontrolled operations such as continuous pulling of control rods. This analysis is carried out by the computer simulation method using WIMSD-5B and MTR-DYN codes, by observing power behavior as a function of time due to neutron chain reactions in the reactor core. Calculations are performed using point kinetics equation, and the feedback effect will be evaluated using static power coefficient and fuel burn-up function. Analyzes were performed for the core configuration of the core no. 99, by lifting the control rod or inserting positive reactivity to the core. The calculation results show that with the reactivity insertion of 0.5% Δk/k at start-up power of 1 W and 1 MW, safety limit is not exceeded either at the beginning, middle, or end of the cycle. The maximum temperature of the fuel is 135°C while the safety limit is 180°C. The margin from the safety limit is large, and therefore fuel damage is not possible when power excursion were to occur.