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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Plenum plate standoff spools for separating a pair of plenum plates in a nuclear reactor

A nuclear reactor has first and second plenum plates disposed in a pressure vessel. Both plenum plates have a plurality of apertures. The second plenum plate is parallel to the first plenum plate. A fuel element includes a fuel element coolant flow tube which extends through aligned apertures of the parallel plenum plates. A fuel element standoff spool is disposed about a portion of the fuel element coolant flow tube which is located between the plenum plates. The nuclear reactor is also usable in nuclear thermal propulsion.

Inman, James B.↗
DOE OSTI2018

Experimental and Computational Investigations of Plenum-to-Plenum Heat Transfer and Gas Dynamics under Natural Circulation in a Prismatic Very High Temperature Reactor

All the goals and the objectives set for the project were successfully executed and achieved and all the milestones have been successfully completed. The results that have been obtained for the first time advance the scientific and engineering knowledge and understanding of the plenum-to plenum natural convection of prismatic block nuclear reactors that is encountered during accident or abnormal operation. These have been accomplished by developing and implementing for the first time unique and flexible scaled-down separate and integrated effects experimental plenumto- plenum facility (P2PF) with dual channels at this time that has been equipped with sophisticated measurement techniques integrated in a novel way on the heated and cooled channels. The unique facility is an asset now that can be extended to research multiple channels and to study the effects of hot plumes in the plena for future projects if funding will be available. It can also be modified to research natural convection of pebble bed reactors. Hence, it complement the HTTF at Oregon State University. However, in this study, heat transfer coefficients from the inner wall surface to the flowing gas (both helium and air were used) and the radial temperature and gas velocity profiles have been measured and investigated along the height of the heated and cooled channels using in house developed wall flush mounted heat transfer probes, thermocouple with in house developed adjuster for radial movement with 1 mm increment inside the channel and hot wire anemometry with also in house developed adjuster for 1 mm radial movement inside the channel, respectively. Also advanced tracer technique has been developed to quantify also for the first time the dispersion of the gas dynamics of the hot and cold channels. The research has provided new knowledge and new benchmarking data that can be used to validate computational fluid dynamics (CFD) codes with conjugate heat transfer. The work and its results that have been performed within the budget have demonstrated their superior technical effectiveness and high economic feasibility to perform needed studies for safety analysis and assessment at least cost for these types of gas cooled very high temperature 4th generation nuclear reactors. Accordingly, the results obtained in this project and the unique facility and techniques that have been developed will benefit greatly the public by advancing the technology of the prismatic block nuclear reactors toward commercialization and to ensure they will be designed and operated safely by utilizing the obtained knowledge and having well validated CFD simulations integrated with heat transfer computations

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Buoyancy-driven air flow within plenum-to-plenum facility down-comer channel

Reliable measurements of thermal and velocity fields were experimentally investigated under different natural circulation intensities in a dual channel facility designed and developed with a representative geometry of prismatic modular reactor core. Experiments were conducted under steady state conditions to obtain statistically stationary time series signals of temperature and velocity. Insignificant radial temperature variation was observed along the down-comer channel implying that the majority of supplied heat is removed through the upper plenum. Velocity results obtained emphasize the delicate nature of natural circulation phenomena in terms of flow destabilization, and recirculation penetration length. Observed air velocity distribution reveals that downward velocity is proportional to the extent of cooling applied to the setup. The calculated values of the dimensionless group Froude number ( Fr ) in the range of 2 to 10 indicate that air is dominated by momentum and is exiting the channel as a jet. Quantification of turbulent intensities implies that flow destabilization is noticeable along the down-comer channel particularly for the case of high cooling intensity. For 5 °C cooling water temperature, turbulent intensity ($v$ rms / $v$) peaks at the mid-channel (z/L = 0.5) reaching a value of 0.5. This value of turbulent intensity is much higher than the inlet (z/L = 0.96) and outlet (z/L = 0.04) turbulent intensities of 0.2 and 0.06 respectively. This destabilization is because of the flow reversal and heat conduction effects through stainless steel flanges connecting both channels. In conclusion, current results provide detailed velocity and temperature variation which can be useful for validating computational fluid dynamics codes.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2023

Investigation of thermal hydraulic behavior of the High Temperature Test Facility's lower plenum via large eddy simulation

A high-fidelity computational fluid dynamics (CFD) analysis was performed using the Large Eddy Simulation (LES) model for the lower plenum of the High–Temperature Test Facility (HTTF), a ¼ scale test facility of the modular high temperature gas-cooled reactor (MHTGR) managed by Oregon State University. In most next–generation nuclear reactors, thermal stress due to thermal striping is one of the risks to be curiously considered. This is also true for HTGRs, especially since the exhaust helium gas temperature is high. In order to evaluate these risks and performance, organizations in the United States led by the OECD NEA are conducting a thermal hydraulic code benchmark for HTGR, and the test facility used for this benchmark is HTTF. HTTF can perform experiments in both normal and accident situations and provide high-quality experimental data. However, it is difficult to provide sufficient data for benchmarking through experiments, and there is a problem with the reliability of CFD analysis results based on Reynolds–averaged Navier–Stokes to analyze thermal hydraulic behavior without verification. To solve this problem, high-fidelity 3-D CFD analysis was performed using the LES model for HTTF. It was also verified that the LES model can properly simulate this jet mixing phenomenon via a unit cell test that provides experimental information. As a result of CFD analysis, the lower the dependency of the sub-grid scale model, the closer to the actual analysis result. In the case of unit cell test CFD analysis and HTTF CFD analysis, the volume-averaged sub-grid scale model dependency was calculated to be 13.0% and 9.16%, respectively. As a result of HTTF analysis, quantitative data of the fluid inside the HTTF lower plenum was provided in this paper. As a result of qualitative analysis, the temperature was highest at the center of the lower plenum, while the temperature fluctuation was highest near the edge of the lower plenum wall. The power spectral density of temperature was analyzed via fast Fourier transform (FFT) for specific points on the center and side of the lower plenum. FFT results did not reveal specific frequency-dominant temperature fluctuations in the center part. It was confirmed that the temperature power spectral density (PSD) at the top increased from the center to the wake. The vortex was visualized using the well-known scalar Q-criterion, and as a result, the closer to the outlet duct, the greater the influence of the mainstream, so that the inflow jet vortex was dissipated and mixed at the top of the lower plenum. Additionally, FFT analysis was performed on the support structure near the corner of the lower plenum with large temperature fluctuations, and as a result, it was confirmed that the temperature fluctuation of the flow did not have a significant effect near the corner wall. In addition, the vortices generated from the lower plenum to the outlet duct were identified in this paper. It is considered that the quantitative and qualitative results presented in this paper will serve as reference data for the benchmark.

97 MATHEMATICS AND COMPUTING↗
NASA NTRS1961

Effects of Geometric Variations on Lift Augmentation of Simple-plenum-chamber Ground-effect Models

Considerable interest has been shown during recent years in ground-effect vehicles. Of the various types proposed, the simple-plenum-chamber vehicle has indicated promise because, although the lift augmentation obtainable appears to be less than that of an annular jet, it may be somewhat less complicated structurally. The present investigation was undertaken to study the effects of some geometric variations upon lift augmentation of a simple plenum chamber within ground proximity. The variables included the ratio inlet area to exit area, plenum-chamber depth, and entrance configuration. An optimum plenum-chamber depth appeared to be between 3 and 10 percent of the plenum-chamber diameter with a ratio of inlet diameter to plenum-chamber diameter of 0.15 for the range of plenum-chamber depths investigated. The most important effect of multiple inlets was the elimination of negative lift augmentation, which was experienced with single sharp-edged inlets, at intermediate heights. Installation of a flared inlet and a turning-vane assembly improved lift augmentation of a single-inlet configuration at intermediate heights.

Davenport, Edwin E.↗
DOE OSTI2019

Design of an experimental facility with a unit cell test section for studies of the lower plenum in prismatic high temperature gas reactors

A proposed design for a Generation IV reactor is the very-high temperature reactor (VHTR), which employs helium as the primary coolant. Although many advantages exist with this design, a number of challenges remain. One such area of concern is the turbulent mixing of non-isothermal flows in the lower plenum, where an array of coolant jets collectively mix and traverse around a series of structural support posts before exiting the core. Neighboring jets in the lower plenum could experience temperature differences as high as 300–400 K, causing potential hot streaking in the support posts. Additional problems could occur further downstream if the level of non-uniformity in the temperatures is too high (e.g, inducing thermal stresses on gas turbine blades). Therefore, it is important that adequate tools exist to predict the turbulent mixing within the VHTR lower plenum. This task remains difficult to achieve due to the complex geometry and flow physics that dictate the mixing. Various modeling approaches can only be assessed using data acquired from a carefully designed experimental facility. This paper presents details of such a facility, where high-fidelity measurements can be acquired for a scaled portion of the lower plenum defined herein as a ‘‘unit cell”, which consists of a hexagonal array of jets whose flow is directed into a test section with support posts positioned between neighboring jets, and subjected to a crossflow. Through adjusting velocity and temperature conditions at the jet inlets, different locations within the lower plenum can be experimentally simulated. System response quantities (SRQs) include both temperature and velocity measurements, captured via multiple thermocouples embedded in each post, and particle image velocimetry (PIV), respectively. These SRQs provide assessment metrics for both the momentum and energy models present in current computational solvers, and therefore the experimental facility is a valuable source for validation data. Another benefit is a more comprehensive understanding of the thermal-hydraulic loading conditions and flow physics present in the VHTR lower plenum.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Counterflow heat exchanger with core and plenums at both ends

Here, this paper illustrates the morphing of flow architecture toward greater performance in a counterflow heat exchanger. The architecture consists of two plenums with a core of counterflow channels between them. Each stream enters one plenum and then flows in a channel that travels the core and crosses the second plenum. The volume of the heat exchanger is fixed while the volume fraction occupied by each plenum is variable. Performance is driven by two objectives, simultaneously: low flow resistance and low thermal resistance. The analytical and numerical results show that the overall flow resistance is the lowest when the core is absent, and each plenum occupies half of the available volume and is oriented in counterflow with the other plenum. In this configuration, the thermal resistance also reaches its lowest value. These conclusions hold for fully developed laminar flow and turbulent flow through the core. The curve for effectiveness vs number of heat transfer units (N tu ) is steeper (when N tu < 1) than the classical curves for counterflow and crossflow.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
NASA NTRS1980

Plenum response to simulated disturbances of the model and fan inlet guide vanes in a transonic tunnel

In order to aid in the design of the National Transonic Facility (NTF) control system, test section/plenum response studies were carried out in a 0.186 scale model of the NTF high speed duct. Two types of disturbances, those induced by the model and those induced by the compressor inlet guide vanes were simulated. Some observations with regard to the test section/plenum response tests are summarized as follows. A resonance frequency for the test section/plenum area of the tunnel of approximately 50 Hz was observed for Mach numbers from 0.40 to 0.90. However, since the plenum is 3.1 times (based on volume) too large for the scaled size of the test section, care must be taken in extrapolating these data to NTF conditions. The plenum pressure data indicate the existence of pressure gradients in the plenum. The test results indicate that the difference between test section static pressure and plenum pressure is dependent on test section flow conditions. Plenum response to inlet guide vane type disturbances appears to be slower than plenum response to test section disturbances.

Gloss, B. B.↗
NASA NTRS2010

Emission Spectroscopic Measurements in the Plenum Region of the NASA IHF Arc Jet Facility

A newly designed segment with optical access was installed in the plenum chamber of the 60 MW Interaction Heating arcjet Facility at NASA Ames Research Center. This special segment has ports located off axis, and the optical fibers can be inserted into these ports. The special segment allows for optical examination of the arc-heated flow as it enters the plenum, and thus assists in determining estimates of the thermodynamic state of the inflow to the convergent section of the nozzle. In the present work, optical emission measurements have been made in VIS-NIR region (wavelengths between 500 nm to 900 nm) for two settings of the arc heater - a 6000 A condition (high condition) with the minimum amount of radial injection of cold air in the plenum, and a 3300 A condition (low condition) with significant amount of cold air injection to reduce the enthalpy of the arc-heated stream. The results presented here were obtained using an Acton SP300i spectrometer coupled to a Princeton Instruments PI-max intensified camera. In addition to the optical emission measurements, computations were performed for the flow in the plenum and radiation along lines of sight corresponding to the optical ports. Along the centerline, i.e., the longest line of sight across the plenum cross-section, there is good agreement between computations and measurements for the high enthalpy condition, although the off-axis radial profiles show some differences. For the low enthalpy condition, there are significant differences between computations and measurements. The current working hypothesis is that the computational model does not capture details of the mixing process in the plenum.

Winter, Michael W.↗
DOE OSTI2019

Calculation of Neutron Damage to Sodium Coolant Inlet Plenum

The coolant inlet plenum is a life limiting component of a Sodium-cooled Fast Reactor (SFR). The inlet plenum is a fixed reactor component which receives sodium from the cold side of the sodium pool redan and directs it into the inlet nozzles of the fuel assemblies. Neutron leakage from the reactor core causes irradiation damage to the steel components of the inlet plenum leading to a reduction in strength. The life of the reactor is essentially tied to the acceptable strength degradation of the inlet plenum. Therefore, typical SFR designs employ a neutron reflector/shield as an integral feature of the fuel assembly. The role of the axial reflector/shield is to: 1) direct fast neutrons back to the core, and 2) attenuate the neutron flux to minimize neutron interaction with the inlet plenum. The longer the reflector and the greater the share of steel within the reflector design, the greater the attenuation. The axial reflector is heated by gamma rays from the core, and must have sufficient coolant penetrations to ensure adequate cool-ability. Also, the sodium coolant must pass through the reflector before reaching the fuel rods above it. The present work investigates varying reflector geometries in support of the Versatile Test Reactor (VTR), a SFR research reactor planned to restart fast reactor irradiation capabilities in the U.S. DOE laboratory complex. This work will demonstrate that the neutron damage to the inlet plenum is relatively insensitive to the number and arrangements of coolant channel penetrations through the reflector.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

Effect of cold cap coverage and emissivity on the plenum temperature in a pilot-scale waste vitrification melter

Integrated computational fluid dynamics (CFD) models are being developed to model the complex physics occurring within the high-level waste melter for vitrification of legacy tank waste at the Hanford site. Here, we present a validation of the integrated CFD model by using data from two experimental runs in a pilot-scale melter. While the model uses several simplifying assumptions (such as constant heat sinks from a cooling jacket and inleakage of ambient air, steady state feed-to-batch conversion heat, and a cold cap model with a simplified shape), it closely predicts the molten glass (1150°C and 1175°C) and plenum temperatures (550°C) obtained from thermocouples during two pilot-scale tests, with an average cold cap coverage of 80%. Additional simulations were performed to explore the sensitivity of the predicted plenum temperatures to variations in cold cap coverage (fraction of melt surface covered by the glass batch) and batch emissivity. The plenum temperature was found to be in the range of 606°C when cold cap coverage decreased from 95% to 70%. Cold cap emissivity had a smaller effect, increasing the plenum temperature by as much as 179°C when cold cap emissivity increased from 0.2 to 0.8. Maintaining a high cold cap coverage without overfeeding is important for a sustained melter operation with high glass throughput. This work provides a tool for achieving that goal in terms of correlating the plenum temperature with the cold cap coverage.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2000

Effects of Hole Length, Supply Plenum Geometry, and Freestream Turbulence on Film Cooling Performance

Experimental measurements are presented in this report to document the sensitivity of film cooling performance to the hole length and coolant delivery plenum geometry. Measurements with hot-wire anemometry detail velocity, local turbulence, and spectral distributions over the exit plane of film cooling holes and downstream of injection in the coolant-freestream interaction zone. Measurements of discharge coefficients and adiabatic effectiveness are also provided. Coolant is supplied to the film cooling holes by means of a large, open plenum and through plenums which force the coolant to approach the holes either co-current or counter-current to the freestream. A single row of film cooling holes with 35 degree-inclined streamwise at two coolant-to-freestream velocity ratios, 0.5 and 1.0, is investigated. The coolant-to-freestream density ratio is maintained in the range 0.96 to 1.0. Measurements were taken under high-freestream (FSTI = 12%) and low-freestream turbulence intensity (FSTI = 0.5%) conditions. The results document the effects of the hole L/D, coolant supply plenum geometry, velocity ratio, and FSTI. In general, hole L/D and the supply plenum geometry play influential roles in the film cooling performance. Hole L/D effects, however, are more pronounced. Film cooling performance is also dependent upon the velocity ratio and FSTI.

Burd, Steven W.↗
DOE OSTI2023

Performance Characteristics of a Rotating Detonation Combustor Exiting Into a Pressurized Plenum to Simulate Gas Turbine Inlet

Abstract The present study aims to experimentally characterize the performance of a rotating detonation combustion (RDC) system integrated with a pressurized downstream plenum to simulate the high-pressure inlet conditions of power-generating gas turbines. A thorough understanding of the operational behavior including wave mode behavior, static pressure profile along the combustor length, and dynamic features of pressure fluctuations is crucial for successful integration of RDC with the turbine. In this study, two RDC configurations are investigated, RDC with a constant area annulus and RDC with a converging nozzle. In both cases, the RDC flow exited into a plenum chamber kept at pressures varying from 155 kPa to 330 kPa. RDC was operated on methane and oxygen-enriched air to represent reactants used in land-based power generation. Experiments were conducted for the two RDCs configurations operated at three reactant mass flow rates (0.23, 0.32, and 0.46 kg/s). The RDC performance is characterized by time-averaged static pressure measurements, and wave velocity determined by ionization probes. In addition, dynamic pressure measurements were recorded both inside and near the exit of RDC channel to investigate wave interactions between RDC and downstream plenum. Results show that the RDC with the converging nozzle achieved superior performance while minimizing detrimental interactions with the reflected shock and/or acoustic waves from the downstream plenum.

Engineering↗
DOE OSTI2020

3D CFD Model Validation Using Benchmark Data of 1/16th Scaled VHTR Upper Plenum and Development of Wall Heat-Transfer Correlation For Laminar Flow

With support from the U.S. Department of Energy-Office of Nuclear Energy’s (DOE-NE’s) Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, an effort has been pursued to support high-temperature gas-cooled reactor (HTGR) technology development and its modeling and simulation needs. There is a particular need for advanced modeling and simulation tools to predict thermal-fluid behavior in the nuclear reactor primary system, especially in the core and the lower and upper plena, during safety-related transients. In this report, two main such activities are presented relevant to the HTGRs: (1) three-dimensional (3D) computational fluid dynamics (CFD) validation using benchmark data from the upper plenum of Texas A&M University’s 1/16th scaled very-high-temperature gas-cooled reactor (VHTR), and (2) development of wall heat-transfer correlation for laminar flow in a wall-heated pipe. The CFD tool validation exercises can be helpful to choose the models and CFD tools to simulate and design specific components of the HTRGs such as upper plenum where jet mixing is a complex phenomenon. In a loss of forced circulation event, the laminar flow can be observed during the development of natural circulation flow. This work includes the development and validation of heat transfer correlations for laminar flow using the Nek5000 CFD code due to limited available experimental data for laminar flow conditions to guide low-order models (1D). In this report, the flow characteristics of a single isothermal jet discharging into the upper plenum was investigated using the Nek5000 Large-Eddy Simulation (LES) CFD tool. Several numerical simulations were performed for various jet-discharged Reynolds numbers ranging from 3,413 to 12,819. A grid-independent study was performed. The numerical results of mean velocity, root-mean-square fluctuating velocity, and Reynolds stress were compared against the benchmark data. Good agreement was obtained between simulated and measured data for axial mean velocities, except near the upper plenum hemisphere. The maximum predicted errors for axial mean velocities at various normalized coolant channel diameter heights of 1, 5, and 10 are 1.56%, 1.88%, and 3.82%, respectively. In addition, the predicted root-mean-square fluctuating velocity and Reynolds stress are qualitatively in agreement with the experimental data. The Nek5000 code was used to develop wall-heat transfer correlation for laminar flow in a cylindrical tube. Several simulations were performed for various Reynolds flow and wall-heat fluxes. A new heat transfer correlation was developed using data from Nek5000 simulation results and regression functions in Matlab. The developed heat transfer correlation is valid for various Reynolds flows from 200 to 2000. The predicted R² value for model fit was 0.875, which ensures that 87.5% of the model data lies on the Nek5000 data. Moreover, a machine learning (ML) tool was used to train and test the Nek5000 data. A good fit of the ML-based model was observed with the test data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS2012

Composite with In Situ Plenums

A document describes a high-performance thermal distribution panel (TDP) concept using high-conductivity (greater than 800 W/mK) macro composite skin with in situ heat pipes. The processing technologies proposed to build such a panel result in a one-piece, inseparable assembly with high conductance in both the X and Y planes. The TDP configuration can also be used to produce panels with high structural stiffness. The one-piece construction of the TDP eliminates the thermal interface between the cooling plenums and the heat spreader base, and obviates the need for bulky mounting flanges and thick heat spreaders used on baseline designs. The conductivity of the TDP can be configured to exceed 800 W/mK with a mass density below 2.5 grams per cubic centimeter. This material can provide efficient conductive heat transfer between the in situ heat plenums, permitting the use of thinner panel thicknesses. The plenums may be used as heat pipes, loop heat pipes, or liquid cooling channels. The panel technology used in the TDP is a macro-composite comprised of aluminum-encapsulated annealed pyrolytic graphite (APG). APG is highly aligned crystalline graphite with an in-plane thermal conductivity of 1,700 W/mK. APG has low shear strength and does not constrain the encapsulating material. The proposed concept has no thermal interfaces between the heat pipes and the spreader plate, further improving the overall conductance of the system. The in situ plenums can also be used for liquid cooling applications. The process can be used to fabricate structural panels by adding a second thin sheet.

Montesano, Mark↗
DOE OSTI2023

Turbine airfoil with leading edge cooling passage(s) coupled via plenum to film cooling holes, and related method

A turbine airfoil includes a body including a wall defining pressure and suction sides, and a leading edge extending between the pressure and suction sides. A cooling circuit inside the wall of the body includes at least one of: a) a suction side to pressure side cooling sub-circuit including a first cooling passage(s) extending from the suction side to the pressure side around the leading edge to a first plenum, and a plurality of first film cooling holes communicating with the first plenum and extending through the wall on the pressure side; and b) a pressure side to suction side cooling sub-circuit including second cooling passage(s) extending from the pressure side to the suction side around the leading edge to a second plenum, and a plurality of second film cooling holes communicating with the second plenum and extending through the wall on the suction side.

Lacy, Benjamin Paul↗
DOE OSTI2019

Computational study of full-scale VHTR lower plenum for turbulent mixing assessment

Next-generation nuclear reactors are poised to efficiently provide reliable power at enhanced safety levels for many years to come. Among the fundamental designs for these reactors is the very-hightemperature reactor (VHTR), which employs helium as the primary coolant and has the potential of reaching elevated temperatures such that chemical processing (e.g., electrolysis for hydrogen production) is viable alongside traditional electricity generation. This and other next-generation reactors also represent a strategic bridge for ultimately transitioning from fossil fuel-dominated energy portfolios to one where renewable options abound. For the VHTR, the ultimate goal of constructing a new plant will be preceded by additional fundamental research aimed at generating trusted models for behavior prediction under normal and accident scenarios. This paper presents a detailed computational fluid dynamics simulation of the lower plenum, where hot coolant from the core mixes together in a turbulent fashion before traveling to power conversion equipment. Because the flow is expected to enter the lower plenum across a wide range of temperatures and velocities, concerns exist when the mixing is incomplete. The potential hot spots on cylindrical support pedestals and the uniformity of the main outlet are the primary metrics of interest in this study, along with flow velocities and temperatures at different locations within the lower plenum. Three locations are probed in detail and suggest a large range exists for approach velocities and temperatures. A large degree of stratification is also seen on the surfaces of the support pedestals, suggesting a facility should accommodate testing for such behavior. The characterizations presented provide the valuable data needed in order to design appropriate experimental testbeds, where scaled modeling can be carried out in a manner meaningful in predicting the full-scale behavior.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗