Reactivity Of Reactor Research Articles

A unifying bubble-layer-based theoretical model for flow boiling and application to a rod bundle

Current analyses of flow boiling primarily utilize two−/three-dimensional numerical simulations or one-dimensional empirical correlations/models. To develop a quasi-two-dimensional theoretical model, a unifying bubble-layer-based theoretical model is proposed, which can be applicable to the entire flow boiling process, from the single-phase liquid flow to the dryout point. This model considers bubble behaviors in each region and heat and mass transfer at the interfaces of two regions, consisting of conservation equations of mass, momentum, and energy, supplemented by a few empirical correlations. A sensitivity analysis of these correlations is conducted, revealing that the slip ratio correlation and bubble condensation rate correlation have the most significant impacts. Using experimental data on void fraction, the most accurate combination of empirical correlations is identified, with a total absolute error of 27.9 %. The verification ranges are 0.1-8Mpa for pressure, 182-1700 kW/m2 for heat flux, 240-2200 kg/m2s for mass flux, 0.83–1.76 for Prandtl number, and 4.4e3–2.98e5 for Reynold number. Finally, this model is applied to the analysis of the NHR200-II rod bundle, revealing the variations of several two-phase parameters under micro-boiling conditions and the influence of operating conditions on the length proportion of different flow sections, serving as important references for analyzing the impact of boiling on reactor reactivity.

Searching for an optimal fuel for a supercritical water reactor (SCWR) to dispose of minor actinides resulting from the nuclear waste

This work investigates the optimal treatment of Minor Actinides (MAs) produced in the spent fuel of nuclear reactors operating around the world. This process is very important in addressing the challenges associated with nuclear waste and reducing environmental impact. A three-dimensional model of a supercritical water reactor (SCWR) has been designed using MCNPX to make a comprehensive neutronic analysis. MAs with a concentration of 1 % have been added to UO2, (Th, U)O2, (Th, 233U)O2 and (Th, rgPu)O2. These fuels have been uploaded in the SCWR assembly and burned in a separate fuel cycle. The infinity multiplication factor (kinf) of the suggested has been investigated with and without adding the minor actinides to analyze the effect of MAs on the reactor reactivity. The fuel constituents, plutonium concentration, MAs concentration and transmutation rate have been tracked with fuel burnup. The reactivity temperature coefficients have been calculated for the suggested cases to ensure the validity of the suggested fuels. The power peaking factor (PPF) and radial power distribution have been calculated for the suggested fuels. The neutronic analysis confirms the suitability of the suggested fuel in burning a significant amount of MAs.

Reactor core design with enriched gadolinia burnable absorbers for soluble Boron-Free operation in the innovative SMR

Effective strategies for maintaining reactor reactivity at a consistent level are indispensable to enable soluble boron-free (SBF) operation in innovative small modular reactor (i-SMR) operation. This paper presents a reactor core design tailored for SBF operation in i-SMR, utilizing concentrated gadolinia burnable absorber rods. This investigation evaluates the concentration range of 155,157Gd and its reactivity characteristics, which are crucial for sustaining excess reactivity. The viability of employing enriched 155,157Gd for the i-SMR SBF operation is established through comprehensive analysis and simulations. Moreover, integrating 155,157Gd with integral gadolinia burnable absorber rods enhances the reactivity control and operational adaptability. Leveraging the spatial self-shielding effect of gadolinia extends the reactivity management capabilities and ensures the operational performance of the i-SMR as optimizing the cycle length. Core design considerations encompassing FA configurations and control rod patterns are investigated to ensure safety and performance across the initial and reload cycles. The feasibility of the SBF operation is verified by assessing compliance with the top-tier requirements of the i-SMR, thus underlining its potential for practical implementation.

Burnup and neutronic parameter analysis of GAMA molten salt reactor (GAMA-MSR)

The GAMA molten salt reactor (GAMA-MSR) is a small modular MSR design developed by the Nuclear Reactor Research Team of the Department of Nuclear Engineering and Engineering Physics, Universitas Gadjah Mada. In the GAMA-MSR design, the reactor core is placed above the primary heat exchanger. The primary heat exchanger is partially filled with fuel in normal operation, allowing room to be filled with the fuel drained from the reactor. Thus, in addition to control rods, the design of GAMA-MSR provides unique reactivity control through the balance of fuel salt fuel from the reactor core to the primary heat exchanger vice versa by adjusting fuel pump capacity. This design allows to shutdown the reactor by switching off the fuel pump. The reactor is initially fueled with 7LiF-ThF4-UF4 with 75: 20.1: 4.9 mol composition. The uranium has 19.75 % U-235 mol fraction. The reactor criticality and nuclide composition in the reactor fuel are calculated using OpenMC code during the reactor operation lifetime, with periodic adjustments to the fuel composition to simulate the effects of the fissile and fertile additions to the fuel and the extraction of actinides and fission products from the fuel. The calculated reactor thermal power is 60 MWth in this paper. The reactor is operated until it achieves a burnup value of 34,000 MWd/tHM. At this burnup value, the U-235 inventory is consumed and the amount is reduced from 794 kg to 250 kg. The U-233, Pu-239 and Pu-241 are produced with the amount are 246 kg, 24 kg and 8 kg respectively at the end of reactor operation. The effects of temperature, fuel density change, and fission product neutron poisons are also calculated. The average temperature reactivity coefficient value is approximately -4.4395×10-5Δk/k per °C. The effect of thermal expansion has been included in the calculation of this temperature reactivity coefficient value. The average value of the void reactivity coefficient is +1.2606×10-3Δk/k per % of void fraction for the void fraction value range up to 50 %. However, molten salt fuel suffers only small changes in density during operation. The negative value of the temperature reactivity coefficient will be dominant; thus, the GAMA-MSR becomes inherently safe. The GAMA-MSR is designed for a 30-year operation time. The effect of the fission product to the reactor reactivity is -1.4597×10-2Δk/k. Full draining of the fuel to primary heat exchanger brings the reactor to shutdown condition with the reactivity of −0.20525. Results showed that this reactor has an inherently safe characteristic and has an effective shutdown system, especially due to the fuel draining system.

Optimized FOPID controller for nuclear research reactor using enhanced planet optimization algorithm

Nuclear reactor control is pivotal for the safe and efficient operation of nuclear power plants, facilitating the regulation of reactor reactivity. This study introduces an optimized fractional-order proportional-integral-derivative (FOPID) controller tailored for maintaining reactivity levels in nuclear power plants, particularly during load-following operations. The controller adjusts the position of control rod to regulate power output effectively. We enhance FOPID controller's performance using a modification of Planet Optimization Algorithm (POA-M), leveraging the strengths of the Arithmetic Optimization Algorithm (AOA) to improve its exploitation capabilities. We evaluate the efficacy of POA-M-FOPID controller against traditional techniques, including POA, AOA, and Particle Swarm Optimization (PSO). We assess its performance using the Egyptian Testing Research Reactor (ETRR-2) as a case study. Our results demonstrate that the POA-M-FOPID controller outperforms alternative algorithms across various control metrics, exhibiting superior resilience and efficiency. Notably, the utilization of the POA-M-FOPID controller yields remarkable improvements in reactor power performance, achieving significantly reduced settling time (25.27 sec) and maximum overshoot (0.67 %) compared to conventional FOPID controllers incorporating POA, AOA, and PSO methods. These findings underscore the effectiveness of POA-M-FOPID in enhancing nuclear reactor control systems, offering potential benefits for broader nuclear power industry in terms of safety, stability, and operational efficiency.

Burnable absorber element (Gd and Er) use in the WWER-type reactor to increase the refueling time

In a WWER-type pressurized water reactor, high-neutron-absorbing materials can compensate for the excess reactivity. Theoretical and computational analyses were conducted in this article to use burnable absorbers placed in a fuel rod to reduce the reactivity margin for extending refueling time for WWER-type reactors. To investigate the reduction of the reactivity margin, the fuel burnup calculation was performed without a burnable absorber and with a combination of a variety of burnable absorbers, namely natural Gadolinium (Gd) and Erbium (Er), applying a simplified GETERA program. The calculation found that the variation of the quantity of the burnable absorber (Gd, Er) inside the fuel assemblies governs the reactivity margin for the fuel burnup and increases the efficiency of Uranium fuel (UO2). The combined use of Gd and Er leads to a smoother decrease in reactor reactivity due to a smaller Er absorption cross section, which makes it possible to reduce the total mass of Gd in the fuel element and reduce the blocking effect. In the calculation, Gd concentrations employed in the computation at 1.5% and 3%, while Er concentrations were used within the range of 0.1%–0.6%.

意外事故下控制鼓和安全鼓对空间核反应堆反应性影响研究

意外事故下控制鼓和安全鼓对空间核反应堆反应性影响研究

Estimation of the buildup of He-3 and Li-6 poisoning, He-4 and tritium activity in beryllium oxide reflector block in APSARA-U reactor

The Apsara-U research reactor, situated at the Bhabha Atomic Research Center, began its operation in 2018. This facility employs a fuel core configuration employing low-enriched uranium and reflector assemblies made of beryllium oxide. In neutron-rich environment of a nuclear reactor, these beryllium oxide reflector blocks participate in nuclear reactions identified as 9Be(n,α) and 9Be(n,2n). The 9Be(n,α) reaction initiates the formation of isotopes such as 4He, 6Li, 3He, and 3H in the process. The accumulation of gasses like 3He, 4He, and 3H causes the beryllium oxide reflector to swell, ultimately restricting how long it can effectively serve as a reflector or moderator material within the reactor. Extended periods of reactor shutdown result in a significant increase in the concentrations of 3He, mainly due to the radioactive decay of 3H. 6Li and 3He nuclei have significant neutron capture cross-sections, leading to a reduction in the reactor's reactivity, a phenomenon termed "poisoning". This effect varies depending on factors like neutron flux, neutron flux spectrum, and the reactor's operational history. Thorough investigations have been carried out to gain insights into the accumulation of 4He, 3He, 6Li, and 3H within beryllium oxide and their effects on reactivity considering the diverse levels of neutron exposure it encounters.

A spatially auto-corrected reactimeter algorithm

Reactimeter algorithms enable real-time evaluation of the reactivity of a near-critical reactor from the signal of a neutron detector. They only depend on a few parameters defined by the inverse point kinetic equations, and which can be calculated by most neutron transport codes. However, the neutron population measured locally by a detector may not always lead to a good estimate of the reactivity. For example, this would be the case for an in-core detector placed just next to a moving control rod. In such special cases – or when measurement precision is an issue – costly simulations are needed to provide correction factors that account for variations in the spatial distribution of the neutron flux. This paper aims to provide an auto-corrected reactimeter algorithm (ACRA) in the special cases where rapid perturbations are applied to a core close to criticality. Such perturbations were successfully performed in the MINERVE reactor using a dedicated control rod to alternatively insert and withdraw a small sample of material under study. ACRA has been validated against 3D Monte Carlo neutron transport calculations using the signals of four in-core detectors located just next to the moving sample.

A Digital Controller for Reactivity Monitoring and Power Control

This paper introduces a controller unit for reactivity monitoring and automatic power control that was designed and constructed for the 500 kW Dalat Nuclear Research Reactor (DNRR). For power control and reactivity calculations, frequency signals from neutron measurement channels of starting and working ranges of the reactor are used. Two abovementioned independent functions were combined in an Artix-7 FPGA board for determining reactivity values by solving the point reactor kinetics equations with six delayed neutron groups and for stabilizing the reactor power at preset levels by determining the unbalance voltage signal to control the automatic control rod. With real-time calculations, the newly developed controller can monitor the reactor reactivity and control the reactor power online. The developed controller unit’s reactivity measuring and power stabilizing capabilities were assessed using the DNRR in normal operation and assumed emergency conditions and compared with those of the preexisting imported BNO-102R1 module of the DNRR control and protection system, known as ASUZ-14R. The results of the experiments show that the produced FPGA-based unit and the BNO-102R1 unit have the same technical characteristics and features, with the disparities being less than 5% and 1%, respectively, in reactivity measurement and power stabilization. The experimental data of reactivity measurements by the FPGA-based unit and the calculation results were also compared and found that the relative deviations between those are also less than 10%. The developed controller unit is capable of carrying out a variety of training and operational experiments on the DNRR.

Reactor reactivity calculations with simplified-P3 and perturbation theories

ABSTRACT A new method to calculate reactivity of nuclear reactors through combining the simplified-P3 (SP3) and perturbation theories is proposed. The focus of this study is the sodium void reactivity calculations in the fast spectrum reactors. Verification of this new method was conducted with an OECD/NEA benchmark, which contains four sodium-cooled fast reactors differing in fuel type and core size. Sodium void reactivity attribution analysis indicates that more accurate prediction of the scattering and leakage components of reactivity can be obtained with the new method compared to the diffusion-perturbation method, and that the computation time is reduced compared with the SN-perturbation method. A term having an unclear physical meaning in the S P 3 -perturbation method was investigated using a transformed S P 3 equation set which is different from the widely used form. Results suggested the equation derivation and code implementation are successful, and the new method shows obvious advantage in component-wise reactivity calculations.

Concept design and neutronics analysis of a heat pipe cooled nuclear reactor with CERMET fuel

Special off-grid power generations require the energy system to be reliable and compact. The heat pipe cooled reactor (HPR) is passively safe and highly modular, and is well suitable for energy demands ranging from a few kW to several MW. This study proposes a compact HPR design in which the ceramic metal composite (CERMET) fuel is used (HPR-CF). The CERMET fuel has advantages such as high temperature resistance and thermal conductivity. The fuel assembly using the CERMET fuel contains 6 heat pipes and fuel blocks with the hexagonal prism shape. The HPR-CF consists of 36 CERMET fuel assemblies containing 216 lithium heat pipes, 12 control drums, 1 safety control rod and peripheral beryllium dioxide reflector, all of which are mounted in a reactor vessel. The Monte Carlo code RMC is used to calculate the neutronics characteristics. The excess reactivity of the HPR-CF reactor is sufficient for operating at high temperature and 3MWt for over 3000 days. The power distribution and the criticality safety are calculated and discussed. The results show the neutronics availability of the HPR-CF design.

Study of the thermal expansion effects of a space nuclear reactor with an integrated honeycomb core design using OpenMC and ANSYS

The negative reactivity feedback in the nuclear fuel plays an important role in maintaining the stable operation of nuclear reactors. For heat pipe cooled space reactors using highly enriched U, the fuel temperature reactivity feedback mainly due to Doppler effect and the thermal expansion of solid fuel. In this paper, based on an integrated heat pipe space reactor design, a one-way coupling of the Monte Carlo code OpenMC and finite element method ANSYS is used to evaluate the reactor expansion and the impacts to the reactor reactivity. The average expansion was 0.126 cm for the fuel block from cold state to full power operation state. The fuel thermal expansion introduces a negative reactivity feedback of −59 pcm, and the reflector expansion introduces another −675 pcm, resulting a temperature coefficient of −0.638 pcm/K.

Key nuclear data for non-LWR reactivity analysis

An assessment of nuclear data performance for non-light-water reactor (non-LWR) reactivity calculations was performed at Oak Ridge National Laboratory that involved a thorough literature review to collect related observations made across different research institutions, an interrogation of the latest ENDF/B evaluated nuclear data libraries, and propagation of nuclear data uncertainties to key figures of merit associated with reactor safety for six non-LWR benchmarks. The outcome of this comprehensive study was published in a technical report issued by the US Nuclear Regulatory Commission. This paper provides a summary of the study’s key observations and conclusions and demonstrates with two examples how the various methods available in the SCALE code system were used to identify key cross section uncertainties for non-LWR reactivity analyses.

Inversion Point of the Isothermal Reactivity Coefficient with Increasing System Temperature in a Heterogeneous Core Lattice of a Light Water Reactor

As part of efforts to develop coupled multiphysics experiments for the benchmark of modern multiphysics reactor simulators, a low-power and open-pool type of light water reactor at the Walthousen Reactor Critical Facility (RCF) was reconfigured with additional equipment, and its neutronic characteristics were fully surveyed. A water loop system was designed and installed to pass through the central region of the reactor core, making the central region overmoderated. The overmoderation would lead to a positive temperature reactivity feedback in the modified reactor configuration. This phenomenon is observed when the system temperature is between 10.69°C and 28.70°C. The inversion point of the isothermal reactivity coefficient is at 28.70°C ± 1.07°C. At this temperature, competition between the negative and positive thermal effects on reactivity compensate each other, and the isothermal reactivity coefficient becomes negative at temperatures higher than the inversion point. This paper presents the experimental determination of the isothermal reactivity and reactivity coefficient at different temperatures as well as the inversion point in the modified RCF reactor configuration. To obtain the best-quality results possible, special attention is given to the choice and adaptation of all the available methods for data postprocessing of experiment measurements. Neutron flux denoising is performed with multivariate wavelet transforms and principal component analysis. The Inverse Kinetics Method is applied to derive reactivity from the neutron flux measurements. To provide accurate and high-fidelity experiment benchmark data for modern code validation, in-depth experimental uncertainty quantification is developed. The results of the experiments show the mixed effects of system temperature on reactor reactivity due to the combined effects of Doppler broadening in the fuel, S(α,β) thermal scattering physics, and change in water density and can be used to validate previously developed cross-section interpolation models in the low-temperature range and positive isothermal reactivity coefficient conditions.

Impact of Boron-Containing Lubricant Additive on Three-Way Catalyst Reactivity and Physicochemical Properties

Boron-containing compounds are one of the lubricant additive options due to their suitable properties for additives and have been used as commercial lubricant additives. In the present study, the impact of a boron-containing lubricant oil additive, AR9100 (BR), on Pd/Rh-based three-way catalyst (TWC) performance is investigated, and the results are compared with the baseline no-additive (NA) case and the industry standard zinc dialkyl-dithiophosphate (ZDDP) results. Accelerated engine aging is performed using a genset to expose the catalysts to lubricant additives at high temperatures. All aged TWC samples are investigated for reactivity in a bench-flow reactor and characterized using a variety of analytical techniques. Compared with the no-additive case, the temperatures of 90% conversion (T90) of NO, CO, C3H6, and C3H8 for the ZDDP-aged TWC sample increased by 34, 30, 37, and 48 °C. However, the T90 of all gas species for the BR-aged TWC sample are similar to the NA-aged TWC sample. Additionally, a significant decrease in water–gas shift reactivity and oxygen storage capacity is observed in the ZDDP-aged sample, but not in the BR-aged sample. Inductively coupled plasma-optical emission spectrometry (ICP-OES) analysis and electron probe microanalysis (EPMA) maps of accelerated engine aging samples show the presence of phosphorus and boron in ZDDP- and BR-aged TWC samples, respectively. However, no boron-related peaks are observed in the X-ray diffraction (XRD) pattern of the BR-aged TWC sample, which may exist in the form of an amorphous phase.

PLIF experiment and verification of boron mixed diffusion model driven by turbulence in nuclear reactor

10B has a large neutron absorption cross section and is commonly used as a neutron poison in the light water reactor (LWR) to adjust the reactivity of nuclear reactors. In this paper, for the first time, a new planar laser-induced fluorescence (PLIF) method with sodium fluorescein as the tracer is used to measure the concentration field of boric acid. In this paper, the injection process is abstracted as a T-tube. The accuracy of LIF is verified by wire mesh sensor. The CFD method is used to simulate the mixing process to come up with proper parameters to consider the influence of turbulence on the flow. The branch velocity of T-tube has a great influence on the diffusion process. The mixing process in the pipeline is mainly driven by turbulence of the side jet. Boric acid is evenly mixed in the pipe within 8 ∼ 10 times of the pipe diameter.

Study on an open fuel cycle of IVG.1M research reactor operating with LEU-fuel

The fuel cycle characteristics of the IVG.1M reactor were studied within the framework of the research reactor conversion program to modernize the IVG.1M reactor. Optimum use of the nuclear fuel and reactor was achieved through routine methods which included partial fuel reloading combined with scheduled maintenance operations. Since, the additional problem in planning the fuel cycle of the IVG.1M reactor was the poisoning of the beryllium parts of the core, reflector, and control system. An assessment of the residual power and composition of spent fuel is necessary for the selection and justification of the technology for its subsequent management. Computational studies were performed using the MCNP6.1 program and the neutronics model of the IVG.1M reactor. The proposed scheme of annual partial fuel reloading allows for maintaining a high reactor reactivity margin, stabilizing it within 2–4 βeff for 20 years, and achieving a burnup of 9.9–10.8 MW × day/kg U in the steady state mode of fuel reloading. Spent fuel immediately after unloading from the reactor can be placed in a transport packaging cask for shipping or safely stored in dry storage at the research reactor site.

Модель реактивнісної аварії на РБМК-1000

The reactor model was used to study the accident that destroyed the RBMK-1000 reactor at Unit 4 of the Chornobyl nuclear power plant on 26 April 1996. The model of reactivity accident of the RBMK-1000 reactor is based on equations of nuclear reactor kinetics, taking into account feedback in reactor reactivity. Reactivity changes as a result of both external influences – the movement of regulatory organs, changes in the reactor inlet coolant temperature, – and as a result of feedback by core parameters – changes in fuel temperature, coolant density, and 135Хе concentration. The model takes into account steam generation in the reactor core, which corresponds to the real physics of processes at the RBMK reactor, and allows obtaining simulation results that best match the recorded data and the consequences of the accident process. The study of reactivity accident on RBMK-1000 reactor is carried out for different combinations of values of control rods efficiency; reactivity coefficients by fuel temperature and coolant density; changes in the reactor inlet coolant temperature; the emergency protection time, as well as the reactor power level before closing the turbine generator stop valve. Different reactivity accident scenarios at RBMK-1000 reactor allow us to determine the most unfavorable combinations of external influences on the course of reactivity accident, namely: start time of main coolant pump rundown, time of activation of emergency protection, power level before the closing of turbine generator stop valves. In most reactivity accident scenarios, first of all, the critical values of fuel enthalpy are reached, at which the process of fuel destruction in the fuel element, destruction of the fuel assembly, and assembly channel start. Important results of studies are 1 – determination of the fact that time of activation of emergency protection after the closing of stop valves of turbine generator significantly affects the value of the maximum neutron power that is achieved during a reactivity accident; 2 – determination of the effect of reactor power before the closing of turbine generator stop valves on the course of the accident; 3 – it is not necessary to achieve supercritical on instantaneous neutrons, supercritical on delayed neutrons is enough to start fuel destruction.

Void effect research on MOX-fueled lead-cooled fast reactor

Liquid metal-cooled reactor (LMR), including lead-cooled fast reactor (LFR), may have strong positive void reactivity coefficient due to the hard neutron spectrum, which is unfavorable for reactor reactivity control. This paper presents a detailed research on the lead void effect for deeper understanding on the causes. Through analytical analysis and numerical simulation, the lead void effect was decomposed into several components, and the spatial distribution of lead void worth was also calculated. The calculation results have shown that spectral hardening effect, which will cause an increase in fertile fission (mainly 238U) and decrease in the capture rate of all heavy nuclides, accounts for the high positive lead void reactivity. Meanwhile, it has been further concluded that high positive lead void reactivity mainly exists in active region. Finally, comparison between two different lead coolant lost scenarios were carried out to further prove these conclusions.