Axial Power Density Research Articles

Methods to study power density distribution in the IVG.1M research reactor after conversion

In the Republic of Kazakhstan, within the framework of international initiatives on the non-proliferation of nuclear weapons, the IVG.1M research reactor is being converted to low enriched fuel. Due to the change in the fuel composition, it becomes necessary to assess the power density distribution in the fuel assemblies and in the core altogether. Close attention to this parameter is explained by the fact that the non-uniformity of the radial and axial power density fields determines the operating parameters of the reactor, its power, the core resource and the efficiency of fuel utilization. The purpose of this work is approbation of the activation gamma-spectrometry method for determining the power density distribution in the fuel of the IVG.1M research reactor after conversion.Additionally, for a preliminary assessment of the experimental outcomes, the paper presents the MCNP6 calculation results for the power distribution in fuel assemblies and the peaking factors.

Applicability of the Krško nuclear power plant core Monte Carlo model for the determination of the neutron source term

A detailed geometrical model of a Krško reactor core was developed using a Monte Carlo neutron transport code MCNP. The main goal of developing an MCNP core model is for it to be used in future research focused on ex-core calculations. A script called McCord was developed to generate MCNP input for an arbitrary fuel cycle configuration from the diffusion based core design package CORD-2, taking advantage of already available material and temperature data obtained in the nuclear core design process. The core model was used to calculate 3D power density profile inside the core. The applicability of the calculated power density distributions was tested by comparison to the CORD-2 calculations, which is regularly used for the nuclear core design calculation verification of the Krško core. For the hot zero power and hot full power states differences between MCNP and CORD-2 in the radial power density profile were <3%. When studying axial power density profiles the differences in axial offset were less than 2.3% for hot full power condition. To further confirm the applicability of the developed model, the measurements with in-core neutron detectors were compared to the calculations, where differences of 5% were observed.

Comparative thermohydraulic analyses of VVER 1000 active core for two different construction types of assemblies

Abstract The OECD/NEA Benchmark on NPP Kalinin Unit 3 “Switching-off of one of the four operating main circulation pumps at nominal power” based on measurements is at its end phase. A new benchmark is starting based on measurements from the NPP Rostov Unit 2 “Reactivity compensation with diluted boron by stepwise insertion of control rod cluster into the VVER-1000 core”. The Rostov-2 core is loaded with assemblies of a new modern type “TVS-2M” which differs in construction in comparison with the “TVSA”-type implemented in the Kalinin-3 core. The goal of the performed study is to determine the thermohydraulic differences, during steady-state operation, between the two cores based on loads with different assembly types. Results of steady-state simulations are compared for Kalinin-3 and Rostov-2 cores, taken from the respective Benchmarks specifications. To ensure that the observed differences are coming only from the fuel assembly construction, the same axial power density is used for all cases. Calculations are done with an under-development version of the GRS system code ATHLET, which has sub-channel modeling capabilities. Thus, in addition to the typical assembly-wise models, calculations with seven assemblies modeled at the sub-channel (pin-by-pin) level were also performed and analyzed. This work will help the Benchmarks' Rostov-2 participants to set up more accurate core thermohydraulic models.

Different techniques for reducing DLOFC fuel temperatures in a PBMR-DPP-400 core

Abstract This article addresses the topic of modifying the fuel cycles of Pebble Bed Reactors in order to reduce their maximum fuel temperatures during Depressurised Loss of Forced Coolant (DLOFC) accidents, in order to prevent unacceptable levels of release of radioactive fission products from the fuel into the environment. The principle strategies used for reducing the maximum DLOFC temperatures were (a) flattening the peaks in the axial power density profile, in order to increase the surface areas over which effective evacuation of decay heat takes place. This reduces the resulting maximum heat fluxes and temperatures in the hotspots; and (b) “pushing” the radial profiles of the equilibrium power density outward towards the external reflector, thereby decreasing the distance, and thus the thermal resistance, over which the decay heat has to be evacuated towards the external reflector. Easier radial evacuation of decay heat reduces the maximum DLOFC temperatures, which always occur in the inner layers of the fuel core. These strategies were applied for both 6-pass recirculation fuelling schemes and Once Through Then Out (OTTO) fuelling schemes by (a) flattening of the peaks in the axial profiles of the equilibrium power density by adding thorium to the LEU fuel and (b) placing purposely-designed distributions of neutron poison in the central reflector. This strategy was further implemented by creating asymmetric cores in which the enrichment of the fuel in the outer fuel flow channels was higher than in the inner ones. The result was a reduction in the maximum DLOFC temperatures from 1536 °C to 1298 °C for the multi-pass and from 2273 °C to 1448 °C for the OTTO. The use of neutron poison in the central reflector to flatten the peaks in the axial profiles of the maximum DLOFC temperatures reduced the maximum DLOFC temperature much more effectively than any of the other techniques.

A specialized code for operation transient analysis and its application in fluoride salt-cooled high-temperature reactors

Fluoride salt-cooled high-temperature reactors (FHRs) include many attractive features, such as high temperature, large heat capacity, low pressure and strong inherent safety. Transient characteristics of FHR are particularly important for evaluating its operation performance. Thus, a specialized code OCFHR (operation and control analysis code of FHR) issued to study an experimental FHR’s operation behaviors. The geometric modeling of OCFHR is based on one-dimensional lumped parameter method, and some simplifications are taken into consideration during simulation due to the existence of complex structures such as pebble bed, intermediate heat exchanger (IHX), air radiator (AR) and multiply channels. A point neutron kinetics model is developed, and neutron physics calculation is needed to provide some key inputs including axial power density distribution, reactivity coefficients and parameters about delayed neutron precursors. For analyzing the operational performance, five disturbed transients are simulated, involving reactivity step insertion, variations of coolant mass flow rate of primary loop and intermediate loop, adjustment of air inlet temperature and mass flow rate of air cooling system. Simulation results indicate that inherent self-stability of FHR restrains severe consequences under above transients, and some dynamic features are observed, such as large negative temperature feedbacks, remarkable thermal inertia and high response delay.

Shaping of the axial power density distribution in the core to minimize the vapor volume fraction at the outlet of the VVER-1200 fuel assemblies

The possibility of decreasing the vapor fraction at the VVER-1200 fuel assembly outlet by shaping the axial power density field is considered. The power density field was shaped by axial redistribution of the concentration of the burnable gadolinium poison in the Gd-containing fuel rods. The mathematical modeling of the VVER-1200 core was performed using the NOSTRA computer code.

Benchmarking of the WIMS9/PARCS/TRACE code system for neutronic calculations of the Westinghouse AP1000™ reactor

The objective of this paper is to assess the accuracy of the WIMS9/PARCS/TRACE code system for power density calculations of the Westinghouse AP1000™ nuclear reactor, as a representative of modern pressurized water reactors (Gen III+). The cross section libraries were generated using the lattice physics code WIMS9 (the commercial version of the legacy lattice code WIMSD). Nine different fuel assembly types were analyzed in WIMS9 to generate the two-group cross sections required by the PARCS core simulator. The nine fuel assembly types were identified based on the distribution of Pyrex discrete burnable absorber (Borosilicate glass) and integral fuel burnable absorber (IFBA) rods in each fuel assembly. The generated cross sections were passed to the coupled core simulator PARCS/TRACE which performed 3-D, full-core diffusion calculations from within the US NRC Symbolic Nuclear Analysis Package (SNAP) interface. The results which included: assembly power distribution, effective multiplication factor (keff), radial and axial power density, and whole core depletion were compared to reference Monte Carlo results and to a published reactor data available in the AP1000 Design Control Document (DCD). The results of the study show acceptable accuracy of the WIMS9/PARCS/TRACE code in predicting the power density of the AP1000 core and, hence, establish its adequacy in the evaluation of the neutronics parameters of modern PWRs of similar designs. The work reported here is new in that it uses, for the first time, the combination of WIMS9/PARCS/TRACE codes to perform neutronics calculation for the Westinghouse’ AP1000™ reactor, as a representative of modern PWRs, with its challenging core configuration.

Simplified model to predict the maximum temperature locations in PWR core

Multi-group diffusion depletion 3D code TRITON was used to calculate the power density profile for a 2775 MW PWR core. TRITON was run for different control rod axial positions at full power: (a) all the control rods are outside the core and (b) control rod group D position at 220 cm with length 213 cm in core. Depending on the axial power density shape, a suitable function was chosen to fit the power density. This function was used to evaluate the temperature distribution in the core. A MATLAB script was used to predict the temperature distribution. The maximum temperatures for the coolant, clad and fuel centre were determined for each control rod position. The maximum temperature locations, when all the control rods were outside the core, were found to exist in assembly (F6), and the vicinity of the centre of the core in assemblies that are burnable absorber free. When control rod group D was inserted in the core in steps, a significant number of high temperatures were found on the same axial position or became closer to the centre of the core.

Analysis and modeling of a high-power Yb:fiber laser beam profile

Recently the fiber and the disk laser have been developed as advanced high power lasers of continuous wave type. Their beams are fiber-guided. Their low beam parameter product determines the focused beam as a high power density tool for laser materials processing. The lateral and axial power density distribution is crucial for the process. The measurement of a focused Yb:fiber laser beam is compared with its theoretical profile. While a Gauss beam describes the real beam in the far field, in the vicinity of the focus the beam is similar to a top-hat profile. In particular, the peak power density is lower around the focus than for a Gaussian beam. This provides a flatter, less varying optical tool along a certain axial range. A suitable model of the focused laser beam is achieved by superposition of a Gauss beam with a second but negative Gauss beam, each with matched parameters for power, spot diameter, and Rayleigh length. Owing to its much better accuracy with less than a 20% error, the model beam is applicable, e.g., for simulation of the welding process.

Analysis of optimum power extraction in a MHD generator with spatially varying electrical conductivity

Analytical expressions for optimum power extracted in an MHD generator, with constant channel width, uniform magnetic field and constant electrical conductivity have been proposed by Neuringer. In this paper, a numerical approach to study optimum power extraction in a MHD generator for an arbitrary geometry, magnetic field and electrically conductivity has been developed. This is accomplished by coupling a flow solver to an optimization code. Using this approach, Neuringer's analysis has been extended to study MHD channels with spatially varying electrical conductivity. An Euler flow solver coupled to an optimization code is used to predict fluid-dynamical variables, optimum power extracted and the voltage drop in the external load corresponding to optimum power extraction. Neuringer's results showed that in an MHD uniform channel with constant conductivity, most of the power delivered to the external load takes place near the extremities. Using our coupled flow/optimizer approach, we show that a more uniform axial power density distribution can be obtained with a spatially varying electrical conductivity distribution, thus making it a useful tool in design/analysis of practical MHD generators.

Response functions for calculating axial power-density profiles in fuel rods using in-core neutron detectors

Micro-pocket fission detectors (MPFDs) have recently been fabricated and successfully tested as in-core flux monitoring devices in the nuclear reactor in Kansas State University. These detectors are sufficiently small to be inserted between fuel elements. Data from an MPFD array can be converted into a 3D power-density map of the reactor core for real-time flux monitoring. In the present study, the necessary mathematical models are developed that relate the power-density profiles in fuel rods to the MPFDs' response. The applicability of an inversion algorithm proposed for obtaining power-density profiles is also verified in this study.

Searching for full power control rod patterns in a boiling water reactor using genetic algorithms

One of the most important questions related to both safety and economic aspects in a nuclear power reactor operation, is without any doubt its reactivity control. During normal operation of a boiling water reactor, the reactivity control of its core is strongly determined by control rods patterns efficiency. In this paper, GACRP system is proposed based on the concepts of genetic algorithms for full power control rod patterns search. This system was carried out using LVNPP transition cycle characteristics, being applied too to an equilibrium cycle. Several operation scenarios, including core water flow variation throughout the cycle and different target axial power distributions, are considered. Genetic algorithm fitness function includes reactor security parameters, such as MLHGR, MCPR, reactor k eff and axial power density.

A 633 nm iodine-stabilized diode-laser frequency standard

An extended-cavity diode laser has been stabilized to hyperfine components of the 6-3, P(33) transition of the B-X system of 127I2 at 633 nm, and its performance evaluated by comparison with an iodine-stabilized He-Ne laser. The Allan standard deviation follows a slope of 3.8 × 10-11 τ-1/2, reaching a minimum of 1.7 × 10-12 at 500 s, and the reproducibility of the diode laser locked to component b21 has been determined to be 7 kHz (standard uncertainty) corresponding to a fractional uncertainty of 1.5 × 10-11. In addition, the influence of modulation depth, iodine pressure and axial power density on laser frequency has been investigated. Complete hyperfine interval sets and absolute frequency determinations of component b21 are presented for modulation depths of 2 MHz and 6 MHz.

Aspherical telescopes for laser amplifiers in the theory of optical inverse problems

An approach is developed to provide a framework in which the task of searching for and selecting the configuration of a two-mirror telescope is formulated as an optical inverse problem: the telescope configuration is found from a given amplitude and/or phase distribution of the field in the plane of the output mirror aperture. Integral equations are obtained for determination of the shape of the telescope mirrors. The solutions of this inverse problem are studied in the geometric optics approximation. A numerical investigation is made of the output radiation power and of the axial radiation intensity in the far-field zone as a function of the configuration of a telescope amplifier and of the input beam characteristics. The use of an aspherical telescope makes it possible to increase, compared with the spherical configuration, the output radiation power and the axial power density of the focused beam by optimisation of the number of trips the radiation makes inside the telescope and of the amplitude distribution of the output radiation.

Focal region characteristics of focused array antennas

The effects of focused aperture amplitude tapering, both direct and inverse, on axial forelobes and aftlobes are investigated. Taylor tapers give low sidelobes but high forelobes and aftlobes, plus a modest gain degradation. Inverse tapers give low forelobes and aftlobes but high sidelobes and large gain degradation. Uniform excitation gives comparable axial and transverse subsidiary lobes, and is probably the best choice. Focal shift, where the peak axial power density occurs between the aperture and the phase (geometric) focus, due to the 1/R^{2} factor, is accurately calculated for a uniform square aperture. Since peak axial power density available from a focused aperture depends only on aperture size in wavelengths and on distance, these focal shift results allow system trade-offs to be made.

Nearfield pressures and surface intensity for cylindrical vibrators

A numerical approach is presented to evaluate the acoustic nearfield of cylindrical vibrators with specified harmonic radial velocity distributions. The approach is based on the use of a combined Green’s function and FFT method. The acoustic field is first represented in terms of a Fourier transform of the specified radial velocity. A normalized axial power density at the surface of the vibrator is then introduced. This quantity provides a useful measure of the local power transfer into the fluid per unit length of the vibrator. The resulting spectral representation of the acoustic field and power density can be rapidly evaluated via the use of FFT methods to obtain the spatial characteristics of the acoustic field and the local power transfer into the fluid. Numerical results are presented to illustrate the spatial characteristics of the acoustic fields and power density of finite length vibrators with axisymmetric and nonaxisymmetric distributions.

Comparison of the Power Distribution in Uranium and Plutonium Fuel Elements in the KWO Core

The introduction of plutonium-charged fuel elements was investigated with regard to a change in the axial power density distribution. An attempt was made to gather information regarding the influence of coolant temperature change on the local power production. The state of the reactor during the investigations was supposed to correspond to realistic operating conditions. During stretch-out operation, as a result of the reduction of the mean coolant temperature and the reactor power, the fall in coolant outlet temperature is greater than the rise in inlet temperature. Hence, the greatest coolant density change occurs at the coolant outlet. In this manner, the relative power density distribution is displaced toward the upper core half. In a special experiment, the influence of turbine load changes on the axial power density distribution in a noncontrolled reactor was investigated. A power reduction at the turbine causes a rise in the mean coolant temperature of the reactor. Owing to local coolant temperature differences, the power density was found to displace toward the upper core half in a noncontrolled reactor, this being more so the case for plutonium fuel elements. The investigations cited in this report and the good agreement between the theoretical predictions and the experimentsmore » permitted the recycling of the self-bred plutonium at KWO without restrictions on the operation of the reactor.« less