Reactor Core Parameters Research Articles

The influence of droplet breakup model on the prediction of reactor core parameters during reflood conditions

During reflood conditions, spacer grids affect the complex flow dynamics inside the reactor core. They can breakup the dispersed droplets, thereby significantly reducing their diameters. This process enhances local heat and mass interfacial transfer, consequently reducing the peak cladding temperature (PCT). Therefore, accurately predicting droplet transport and size within a reactor core is very important for reactor safety analysis. Recently, the US. NRC TRACE code has been modified by extending its governing equations to incorporate a three-field approach that captures the dynamic behavior of droplet entrainment, transport, and their interactions with structures within the reactor core. The new TRACE (v5.0 patch 8) three-field framework incorporates droplet models accounting for entrainment from liquid pools, liquid films on vertical surfaces, and droplet breakup. This paper focuses on the influence of the spacer grid droplet breakup model on the behavior of reactor core during reflood transient. The selected droplet breakup model effectively captures the enhancement in interfacial heat transfer observed in experimental data. We evaluate the influence of the droplet breakup model on the three-field version of TRACE by comparing it to recent Rod Bundle Heat Transfer (RBHT) experiments. These experiments had previously indicated that TRACE tended to overpredict the PCT for various initial and boundary conditions. Subsequently, we conducted a detailed comparison between TRACE v5.0 patch 7 and the three-field TRACE (v5.0 patch 8) with and without the spacer grid droplet breakup model. Our findings demonstrate a significant improvement in predictive accuracy with the new breakup model compared to the base capability of TRACE v5.0 patch 7. This improvement results in more accurate predictions of PCTs and quenching times for the assessed conditions.

Radial Basis Function Surrogate Model-Based Optimization of Small Light-Weight Lead-Bismuth Reactor Core

Owing to their high miniaturization potential and strong natural circulation capability, lead-bismuth reactors have been widely employed for nuclear energy utilization as well as military and civilian applications. However, there is a need to optimize the design of these reactors. To solve the complex high-dimensional nonlinear single-/multiobjective optimization problem of lead-bismuth reactors, we proposed a single-/multiobjective optimization method (termed as RBF-OLHS-NGA/NSGA II method) for optimizing the reactor core, which is based on the radial basis function (RBF) surrogate model (for prediction), orthogonal Latin hypercube sampling, niche genetic algorithm (for single-objective optimization), and nondominated sorting genetic algorithm (for multiobjective optimization). In view of this approach, the design optimization procedure of the lead-bismuth reactor based on RBF (DOPPLER-R) was developed, which combined the reactor Monte Carlo (RMC) code and the steady-state thermo-hydraulic analysis calculation (STAC) code to sample, predict, and optimize reactor core parameters. Furthermore, the single-/multiobjective optimization approach was verified by considering the core fuel loading and active zone volume of SPALLER-4 as the optimization objectives. The results show that the RBF surrogate model can accurately and rapidly predict the core characteristic parameters of lead-bismuth reactors. When compared to the value calculated using the RMC code, the relative error associated with the predicted effective multiplication factor ( <math xmlns="http://www.w3.org/1998/Math/MathML" id="M1"> <msub> <mrow> <mi>k</mi> </mrow> <mrow> <mi>e</mi> <mi>f</mi> <mi>f</mi> </mrow> </msub> </math> ) is within ±0.1%. Compared to the unoptimized values, the single-objective optimized fuel loading reduced by 400 kg, multiobjective optimized fuel loading decreased by 455–493 kg (optimization ratio of 78%–84%), and active zone volume reduced by 166362–182888 cm3 (optimization ratio of 72%–79%). These results prove that the proposed optimization method is feasible, exhibits high efficacy, and can provide new technical ideas for single-/multiobjective optimization of multiphysics, multivariable, and multiconstraint lead-bismuth reactors.

A Study of Gamma Heating, Neutron Flux, and Gamma Flux Under Several Reactor Core Conditions in the McMaster Nuclear Reactor

Changes in the thermal power of a nuclear research reactor will lead to changes in experimental, irradiation, and testing conditions. Consequently, reactor core parameters are inevitably susceptible to changes. One such parameter is gamma heating (GH), which results from gamma interaction with materials. In this work, a gamma thermometer was used to measure GH over the course of 7 operational days and nights. In addition, the Monte Carlo reactor physics code Serpent-2 was used to evaluate the sensitivity of common detection methods for monitoring reactor core parameters such as neutron fluxes, GH, and gamma flux under the following conditions: reactor core power variation, reactor core fuel shuffling, and detector vicinity fuel assembly shuffling. The GH values obtained through measurements and calculations were linearly proportional to the reactor power. In addition, the Serpent-2 code for the McMaster nuclear reactor showed that despite maintaining the reactor power core at the same level, the fuel burnup distribution could alter the studied parameters.

Uncertainty quantification of the power control system of a small PWR with coolant temperature perturbation

The coolant temperature feedback coefficient is an important parameter of reactor core power control system. To study the coolant temperature feedback coefficient influence on the core power control system of small PWR, the core power control system is built with the nonlinear model and fuzzy control theory. Then, the uncertainty quantification method of reactor core parameters is established based on the Latin hypercube sampling method and the Bootstrap method. Finally, under the conditions of reactivity step perturbation and coolant inlet temperature step perturbation, uncertainty analysis for two cases is carried out. The result shows that with fuzzy controller and fuzzy PID controller, the uncertainty of the coolant temperature feedback coefficient affects the core power control system, and the maximum uncertainties of core relative power, coolant temperature deviation, fuel temperature deviation and total reactivity are acceptable.

VALIDATION OF SRAC CODE SYSTEM FOR NEUTRONIC PARAMETERS CALCULATION OF THE PWR MOX/UO2 CORE BENCHMARK

VALIDATION OF SRAC CODE SYSTEM FOR NEUTRONIC PARAMETERS CALCULATION OF THE PWR MOX/UO2 CORE BENCHMARK. Determination of neutronic parameter value is an important part in determining reactor safety, so accurate calculation results can be obtained. This study is focused on the validation of SRAC code system in the calculation of neutronic parameters value of a PWR (Pressurized Water Reactor) reactor core. MOX/UO2 Core Benchmark was choosed because it is used by several researchers as a reference core for code validation in the determination of neutronic parameters of a reactor core. The neutronic parameters calculated include critical boron concentration, delayed neutron fraction, and Power Peaking Factor (PPF) and its distribution in axial and radial directions. When compared with reference data, the calculation results of the critical boron concentration value show that there is a difference of 22.5 ppm on SRAC code system. Meanwhile, differences in power per fuel element (assembly power error) value of power-weighted error (PWE) and error-weighted error (EWE) is 2.93% and 3.94%, respectively. Maximum difference between PPF value in axial direction with reference reaches a value of 4.57%. SRAC calculation results also show consistency with the calculation results of other program packages or code. Results of this study indicate that SRAC code system is still quite accurate for the calculation of neutronic parameters of PWR reactor core benchmark. Therefore, SRAC code system can be used to calculate neutronic parameters of PWR reactor core, especially when using MOX (mixed oxide) fuel.Keywords: Neutronic parameter, critical boron concentration, power peaking factor, SRAC code system.

Effects of various types of reflectors on the 99Mo production in the VVER-Ts reactor targets

The effects from introducing various types of reflectors in the VVR-Ts reactor core on the 99Мо production were analyzed. Earlier the effects of only the beryllium reflector on the VVR-Ts reactor core characteristics, such as reactivity margin, neutron flux in experimental channels, and activity of the accumulated 99Мо, were calculated. The calculations are based on a generated precision model of the core which comprises one experimental channel where targets are irradiated for the 99Мо production. The model was built using the SCALE code. The code allows a fairly broad range of calculations to be performed, from criticality estimation to radiological assessment tasks. As the result of the computational analysis of the model, such characteristics were obtained as the effective multiplication factor, the power density in the 99Мо production targets, the neutron flux in the target raw material, and the quantity of the produced 99Мо after 120 hours of irradiation. The data was compared with the results of similar calculations of the VVR-Ts reactor core parameters. Further, the list of the materials used extensively as the reactor core reflector or moderator was formed based on reference literature. A number of models were obtained and analyzed on its basis, in which the water space on the core periphery was substituted for the investigated materials.

Transient Studies on Low-Enriched-Uranium Core of Ghana Research Reactor–1 (GHARR-1)

Following the core conversion of Ghana’s miniature neutron source reactor (MNSR) from highly enriched uranium (HEU) to low-enriched uranium (LEU), there has been a change in the fuel composition, fuel, clad, and other reactor core parameters. Since the allowable core power in a nuclear reactor is limited by thermal considerations, this study presents transient analysis of the LEU core of Ghana Research Reactor−1 (GHARR-1). The transient study has been carried out using the Monte Carlo N-Particle code version 5 (MCNP5) and the Program for the Analysis of Reactor Transients (PARET)/Argonne National Laboratory (ANL) computational tools. The behavior of the reactor core at normal and accident conditions of large reactivity insertions was studied. Transient results obtained for accidental large reactivity insertions of 6.71 mk indicated that boiling might occur in the coolant because under such large reactivity insertions, the coolant temperature was close to the saturation temperature of the coolant. The results show that boiling will not occur in the core for other reactivity insertions of 1.94, 2.1, 2.99, 3.87, and 4.0 mk considering that the outlet coolant temperatures obtained are far below the saturation temperature of 100°C at a pressure of 1 atm. The clad and fuel meat temperatures obtained for all the reactivity insertions are far below the melting points of Zircaloy-4 clad material and UO2 fuel. The results of the power profiles obtained show that the reactor is inherently safe even under large reactivity insertion conditions. The results obtained were found to agree well with the available experimental results. Comparison of the results of the LEU core with the previous HEU core has shown that temperature rise in the LEU core is lower than that in the HEU core under reactor transient conditions.

Deterministic Analysis of the Low Enriched Uranium SLOWPOKE-2 Research Reactor Using DRAGON-5 and DONJON-5 Codes System

This paper discusses the capability of the deterministic code system DRAGON-5 and DONJON-5 to reliably simulate the SLOWPOKE-2 research reactor. The characteristics of complex geometry, highly heterogeneity and large leakage of the SLOWPOKE-2 core, poses a challenge for classical neutronics calculation tools and schemes for power reactors. The simulation model is presented (geometry, cross section library, assumption, etc.), and the results obtained by DRAGON-5/DONJON-5 codes and the evaluated nuclear data from WLUP project were compared to the calculations performed with Monte Carlo code MCNP 4A using a detailed geometrical model of the reactor and the experimental data. Criticality calculations have been performed to verify and validate the model based on the multiplication factor comparisons for 4 energy groups with two different leakage models. The results indicate that the DRAGON-5 and DONJON-5 codes system agrees well with the MCNP 4A continuous-energy Monte Carlo simulation and experiment. Since created model properly describes the reactor core it can be used for calculations of reactor core parameters and for optimization of research reactor application.

Neutron inelastic scattering of 232Th: measurements and beyond

Gamma production cross sections have been obtained for 81 transitions in 232Th from (n,n’γ), 11 in 231Th from (n,2nγ) and 7 in 230Th from (n,3nγ) reactions using prompt gamma spectroscopy. Incident neutron energies were determined using the neutron time-of-flight technique. Sources of uncertainty have been examined and their correlations have been computed. Total uncertainty on cross sections ranges from 4 to 20%. Obtained cross sections are in agreement with prior experiments, but are not well reproduced by the TALYS 1.8 reaction code using default parameters. During analysis, discrepancies between our findings and the Evaluated Nuclear Structure Data File (ENSDF) were noted. Future work related to the present experiment includes: improving theoretical models, quantifying the influence of the 232Th inelastic neutron cross section on reactor core parameters, and conducting additional measurements.

Validation of the MCNP computational model for neutron flux distribution with the neutron activation analysis measurement

The objective of this work is to demonstrate the method for validating the predication of the calculation methods for neutron flux distribution in the irradiation tubes of TRIGA research reactor (TRR-1/M1) using the MCNP computer code model. The reaction rate using in the experiment includes 27Al(n, α)24Na and 197Au(n, γ)198Au reactions. Aluminium (99.9 wt%) and gold (0.1 wt%) foils and the gold foils covered with cadmium were irradiated in 9 locations in the core referred to as CT, C8, C12, F3, F12, F22, F29, G5, and G33. The experimental results were compared to the calculations performed using MCNP which consisted of the detailed geometrical model of the reactor core. The results from the experimental and calculated normalized reaction rates in the reactor core are in good agreement for both reactions showing that the material and geometrical properties of the reactor core are modelled very well. The results indicated that the difference between the experimental measurements and the calculation of the reactor core using the MCNP geometrical model was below 10%. In conclusion the MCNP computational model which was used to calculate the neutron flux and reaction rate distribution in the reactor core can be used for others reactor core parameters including neutron spectra calculation, dose rate calculation, power peaking factors calculation and optimization of research reactor utilization in the future with the confidence in the accuracy and reliability of the calculation.

Optimizing the water chemistry of the Novovoronezh NPP Units 3 and 4 to prevent growth of coolant pressure difference across the reactor

For a few fuel campaigns in the first years of their extended service life, the Novovoronezh NPP Units 3 and 4 had to operate with a limited reactor thermal power output due to constraints imposed on certain parameters of the reactor core. The decrease of thermal power output during a campaign was accompanied by a growth of pressure difference across the reactor caused by accumulation of deposits in the fuel assemblies, which in turn led to a growth in the number of leaky fuel rods. Compensating measures were implemented based on the results obtained from the performed investigations and modeling of corrosion product mass transfer processes, which in the final analysis made it possible to minimize the accumulation of deposits in fuel assemblies and loss of fuel rod tightness, and to exclude the need to limit the reactor thermal power as demanded by constraints imposed on the reactor core parameters.

Multi-group formulation of the temperature-dependent resonance scattering model and its impact on reactor core parameters

A multi-group formulation for the exact neutron elastic scattering kernel is developed. It incorporates the neutron up-scattering effects stemming from lattice atoms thermal motion and it accounts for them within the resulting effective nuclear cross-section data. The effects pertain essentially to resonant scattering off of heavy nuclei. The formulation, implemented into a standalone code, produces effective nuclear scattering data that are then supplied directly into the DRAGON lattice physics code where the effects on Doppler reactivity and neutron flux are demonstrated. The correct accounting for the crystal lattice effects influences the estimated values for the probability of neutron absorption and scattering, which in turn affect the estimation of core reactivity and burnup characteristics. The results show an increase in values of Doppler temperature feedback coefficients up to −10% for UOX and MOX LWR fuels compared to the corresponding values derived using the traditional asymptotic elastic scattering kernel. This paper also summarizes research performed to date on this topic.

Neutronic calculations of AFPR-100 reactor based on Spherical Cermet Fuel particles

The Atoms For Peace Reactor (AFPR-100), as a 100MW(e) without the need of on-site refueling, was originally based on UO2 TRISO fuel coated particles embedded in a carbon matrix directly cooled by light water. AFPR-100 is considered as a small nuclear reactor without open-vessel refueling which is proposed by Pacific Northwest National Laboratory (PNNL). An account of significant irradiation swelling in the silicon carbide fission product barrier coating layer of TRISO fuel element, a Spherical Cermet Fuel element has been proposed. Indeed, the new fuel concept, which was developed by PNNL, consists of changing the pyro-carbon and ceramic coatings that are incompatible with low temperature by Zirconium. The latter was chosen to avoid any potential Wigner energy effect issues in the TRISO fuel element. Actually, the purpose of this study is to assess the goal of AFPR-100 concept using the Cermet fuel; undeniably, the fuel core lifetime prediction may be extended for reasonably long period without on-site refueling. In fact, we investigated some neutronic parameters of reactor core by the calculation code SRAC95. The results suggest that the core fuel lifetime beyond 12 equivalent full power years (EFPYs) is possible. Hence, the adoption of Cermet fuel concept shows a core lifetime decrease of about 3.1 EFPY.

Visual numerical steering in 3D AGENT code system for advanced nuclear reactor modeling and design

The AGENT simulation system is used for detailed three-dimensional modeling of neutron transport and corresponding properties of nuclear reactors of any design. Numerical solution to the neutron transport equation in the AGENT system is based on the Method of Characteristics (MOCs) and the theory of R-functions. The latter of which is used for accurately describing current and future heterogeneous lattices of reactor core configurations. The AGENT code has been extensively verified to assure a high degree of accuracy for predicting neutron three-dimensional point-wise flux spatial distributions, power peaking factors, reaction rates, and eigenvalues. In this paper, a new AGENT code feature, a computational steering, is presented. This new feature provides a novel way for using deterministic codes for fast evaluation of reactor core parameters, at no loss to accuracy. The computational steering framework as developed at the Technische Universität München is smoothly integrated into the AGENT solver. This framework allows for an arbitrary interruption of AGENT simulation, allowing the solver to restart with updated parameters. One possible use of this is to accelerate the convergence of the final values resulting in significantly reduced simulation times. Using this computational steering in the AGENT system, coarse MOC resolution parameters can initially be selected and later update them – while the simulation is actively running – into fine resolution parameters. The utility of the steering framework is demonstrated using the geometry of a research reactor at the University of Utah: this new approach provides a savings in CPU time on the order of 50%.

Knowledge-directed characterization of nuclear power plant reactor core parameters

A major life-limiting factor of the UK's Advanced Gas-Cooled Reactors (AGRs) is the condition of the graphite core. Installation of new measurement equipment is difficult and expensive, therefore maximizing the information gained from existing equipment is highly desirable. The main approach to determining the health of an AGR core is through periodic inspections undertaken during planned outages. However, there is the desire to supplement this inspection activity through the analysis of data gathered as part of routine plant operation. One such source of data is measurements taken during refueling and this paper describes knowledge-directed characterization of this refueling data, both spatially across the reactor core and temporally across the operational lifetime of the core. Characterization provides information relating to the current condition of the reactor core and allows suspected ageing trends to be visualized and confirmed. A standard approach for characterizing reactor core data is presented and applied to a variety of different reactor core parameters. The benefit of this approach is that it allows engineers to distill large volumes of refueling data into a readily understandable format in a short period of time. It also allows hypothesized trends relating to the ageing process within the core to be tested and provides supporting evidence for these hypotheses. The trending data is also valuable as it can form the basis of a predictive model of ageing of the reactor core. The ageing process of nuclear graphite is understood from theoretical and experimental viewpoints and this empirical data, gathered from operating reactors, further supports this understanding. This paper represents the initial exploration of using refueling data to construct a predictive model of AGR reactor core ageing.

Performance evaluation/analysis of Pakistan Research Reactor-1 (PARR-1) current core configuration

Neutronic and thermal hydraulic analyses have been carried out for current core of Pakistan Research Reactor-1 (PARR-1). Comparison was made between calculated and measured key neutronic parameters. Reactor core parameters important for reactor operation and safety have been calculated. Calculated neutronic parameters include: excess reactivity, shut down margin, control rod worth, peak power density location, criticality position, peaking factors, neutron flux in fuel elements and neutron flux at irradiation sites in the core. Calculated thermal hydraulic parameters include: steady-state temperatures and peak temperatures at fuel centerline, clad surface and in water coolant. In order to determine safety margins, heat fluxes at Onset of Nucleate Boiling (ONB), Onset of Flow Instability (OFI) and Departure from Nucleate Boiling (DNB) were determined using standard correlations. After assembling the core, performance of the core was also evaluated by experimentation. The core was assembled and some of the core parameters namely: excess reactivity, shut down margin, control rod worth and flux profile at in-core irradiation sites have been measured. On comparison with experimental data, reasonable agreement has been found between the calculated and the measured parameters.

Analysis of neutron flux distribution for the validation of computational methods for the optimization of research reactor utilization

In order to verify and validate the computational methods for neutron flux calculation in TRIGA research reactor calculations, a series of experiments has been performed. The neutron activation method was used to verify the calculated neutron flux distribution in the TRIGA reactor. Aluminium (99.9 wt%)-Gold (0.1 wt%) foils (disks of 5mm diameter and 0.2mm thick) were irradiated in 33 locations; 6 in the core and 27 in the carrousel facility in the reflector. The experimental results were compared to the calculations performed with Monte Carlo code MCNP using detailed geometrical model of the reactor. The calculated and experimental normalized reaction rates in the core are in very good agreement for both isotopes indicating that the material and geometrical properties of the reactor core are modelled well. In conclusion one can state that our computational model describes very well the neutron flux and reaction rate distribution in the reactor core. In the reflector however, the accuracy of the epithermal and thermal neutron flux distribution and attenuation is lower, mainly due to lack of information about the material properties of the graphite reflector surrounding the core, but the differences between measurements and calculations are within 10%. Since our computational model properly describes the reactor core it can be used for calculations of reactor core parameters and for optimization of research reactor utilization.

Monte Carlo simulation of a research reactor with nominal power of 7 MW to design new control safety rods

The Monte Carlo simulation has been established for a research reactor with nominal power of 7 MW. A detailed model of the reactor core was employed including standard and control fuel elements, reflectors, irradiation channels, control rods, reactor pool and thermal column. The following physical parameters of reactor core were calculated for the present LEU core: core reactivity ( ρ), control rod (CR) worth, thermal and epithermal neutron flux distributions, shutdown margin and delayed neutron fraction. Reduction of unfavorable effects of blockage probability of control safety rod (CSR)s in their interiors because of not enough space in their sites, and lack of suitable capabilities to fabricate very thin plates for CSR cladding, is the main aim of the present study. Making the absorber rod thinner and CSR cladding thicker by introducing a better blackness absorbing material and a new stainless steel alloy, respectively, are two studied ways to reduce the effects of mentioned problems.

A network-based system of simulation, control and online assistance for HTR-10

A network-based computer system has been developed for HTR-10. This system integrates three subsystems: the simulation subsystem (SIMUSUB), the visualized control designed subsystem (VCDSUB) and the online assistance subsystem (OASUB). The SIMUSUB consists of four functional elements: the simulation calculating server (SCS), the main control client (MCC), the data disposal client (DDC) and the results graphic display client (RGDC), all of which can communicate with each other via network. It is intended to analyze and calculate physical processes of the reactor core, the main loop system and the steam generator, etc., as well as to simulate the normal operational and transient accidents. The result data can be dynamically displayed through the RGDC. The VCDSUB provides a platform for control system modeling where the control flow systems can be automatically generated and graphically simulated. Based on the data from the field bus, the OASUB provides some of the reactor core parameters, which are difficult to measure. This integrated system can be used as an educational tool to understand the design and operational characteristics of the HTR-10, and can also provide online support for operators in the main control room, or as a convenient powerful tool for the control system design.

Response of incore γ-detectors to parametric fluctuations in a reactor core

Adjoint transport theory is applied to coupled neutron- and γ-transport in order to analyse the detection properties of γ-detectors with respect to fluctuating reactor core parameters. In a three-group analysis with two neutron-groups and one γ-group it is shown that a small incore γ-detector acts like a spatially smeared-out neutron-detector, concentrated mainly in the surrounding fuel rods and with a relatively high epithermal sensitivity.