Delayed Neutron Fraction Research Articles

SERPENT CODE POINT KINETIC PARAMETERS ANALYSIS: A NEXUS FOR NIGERIA RESEARCH REACTOR – 1 CORE CONVERSION USING ENRICHED URANIUM DIOXIDE FUEL

The International Atomic Energy Agency (IAEA) requires that all test and research reactors operating on Higher Enriched Uranium (HEU) should be converted to Low Enriched Uranium (LEU) for safety and security purposes. Nigeria having a Miniature Neutron Source Reactor (MNSR) has been long interested in fuel technological research not just to develop the area but also to meet with resolution on the nuclear treaty set out by the global nuclear regulatory body. In this study, reactor kinetic parameters such as effective delayed neutron fractions, prompt neutron lifetime as well as mean neutron generation time were analysed for Nigerian Research Reactor-1 (NIRR-1). Serpent Monte Carlo code 1.17 is used in the analysis. For delayed neutron parameters determination, we used fission probability iteration under one averaged generation time and neutron population rate.The calculated values for delayed neutron were recorded as analogue prompt and implicit prompt neutron lifetime, reproduction time and emission time are in the order of 3×10-7 (s), in agreement with the calculated data from the nuclear data libraries and some literature.The result for delay neutron fraction and other time-based parameters support the fuel core conversion for NIRR-1.The computational and pictorial results obtained from Serpent code simulation described well the transient behavior of the delayed neutron in this reactor.The analytical results also spelled out the relevance and compatibility of low enriched uranium dioxide fuel over higher enriched type.The result of this study conforms with other results obtained from similar reactors but with different Monte Carlo codes and with higher enriched uranium

Comparative analysis between homogeneous and heterogeneous models of gas cooled fast reactor core (GFR-2400)

Abstract The main objective of the presented work is to investigate the effectiveness of homogenization of the GFR-2400 core in simulating its neutronic characteristics. To explore and evaluate the neutronic behavior of the large scale Gas cooled Fast Reactor GFR-2400 core, two computer models (homogeneous and heterogeneous) were designed using the MCNPX code. The designed heterogeneous model has been validated by comparing its results with a previously published paper. The results of both models were compared with each other to study the effect of fuel homogeneity on the radial flux and power distribution through the core. As part of a safety analysis of the reactor core for both designs, the reactivity worth of control rods, Effective delayed neutron fraction (ß eff), Doppler constant and Depressurization effect have been analyzed and compared. The variations of the effective multiplication factor (k eff), the minor actinides concentration and the most important fission products as a function of operation time have been investigated for the presented designs.

Nuclear data uncertainty propagation applied to the versatile test reactor conceptual design

The Versatile Test Reactor (VTR) currently under development is a 300 MWth sodium-cooled fast reactor (SFR) fueled with ternary metal alloy fuel, which aims to accelerate the testing of advanced nuclear fuels, materials, instrumentation, and sensors in high flux environments that are necessary to license the next generation of advanced reactor concepts. To support the VTR design process, uncertainties associated with the nuclear data has been propagated through the reactor core neutronics calculation to global parameters of interest, such as the core multiplication factor, kinetic parameters, and various reactivity feedback coefficients, following the sensitivity based uncertainty propagation approach. By folding the sensitivity coefficients, separately computed by the generalized perturbation theory code PERSENT and Monte Carlo code Serpent 2, with the variance–covariance matrices from COMMARA-2.0, we obtain the reaction-wise, isotope-wise, and overall uncertainties for each response of interest due to nuclear data uncertainty. With Serpent 2, the statistical error of the uncertainty is obtained by propagating the statistical error of the sensitivity coefficients through the same process using a newly developed uncertainty propagation method. From both codes, the overall top uncertainty contributors are found to be the cross section of Fe-56 elastic scattering, Na-23 elastic scattering, and U-238 inelastic scattering. The large contributions of the Fe-56 elastic scattering cross sections to global parameters are due to its relatively large relative uncertainty of 5–10% in nuclear data and the large volume of Fe-containing reflector assemblies in the fairly compact VTR core design. Both codes agreed well for the overall uncertainty estimates of all responses of interest, except the delayed neutron fraction, prompt neutron generation time, and the coolant density feedback coefficient, where Serpent 2 yielded a much larger value than PERSENT due to the large statistical error of sensitivity coefficients. The calculated uncertainties are also compared to those associated with other SFR cores. Another outcome of this study is a variance–covariance matrix of reactivity coefficients, which can be used in the subsequent uncertainty propagation to the system level to investigate the impact of identified uncertainties on system responses in the safety analysis.

Evaluating the variations of point kinetics parameters in pressurized heavy water reactor analyses

Since the pressurized heavy water reactor design utilizes natural uranium bundles as its fuel, there is a need to keep replenishing the reactor core with unirradiated fuel bundles in order to keep the reactor critical. This practice causes the fuel inventory in the core to be very dynamic since the changes are not solely due to the depletion of the fuels due to irradiation but also due to the presence of new fuel bundles in the core. An average behaviour of the core after operating in an equilibrium stage for a long time can be captured through the time-average approach. This approach has been extensively used in the pressurized heavy water reactor fuel management, especially during the design phase where the actual data from the operating reactor are not readily available. The time-average core is utilized for many purposes such as to determine the average fueling rate, to determine the average worth of various reactivity control devices, as well as to determine the point kinetics parameters for transient analyses. The objective of this paper is to quantify potential variations in the point kinetics parameters (namely the decay constants and the delayed neutron fractions) as a function of core inventories with respect to the time-average-based values. In particular, 100 core configurations representing various snapshots from an equilibrium operation of the core have been utilized to quantify these potential variations. It should also be noted that the point kinetics equation (PKE) is utilized in this study, instead of full-core space–time model, since the objective of this paper is to bring about the awareness of potential variations and the PKE should be sufficient. The results of this study indicate that, for a point kinetics equations system involving six precursor groups, the variations in precursor decay constants and delayed neutron fractions could be as high as ±0.45% and ±4.85%, respectively. Using the point kinetics parameters in a simple transient involving an insertion of positive reactivity, it is demonstrated that the variation in the trip time for the transient could vary up to approximately ±3.55%, depending on the amount of positive reactivity insertion. These potential variations should be acknowledged since it is an important part of safety analysis. However, it is also important to note that, based on the amount of positive reactivity insertion evaluated in this study, the higher variation corresponds to a lower level of positive reactivity insertion which means that the transient will be slowly developing, could be observed more readily, and, therefore, could be easily terminated.

Application of dynamic mode decomposition to Rossi-α method in a critical state using file-by-file moving block bootstrap method

ABSTRACT Prompt neutron decay constant in a critical state is useful information to validate the numerically predicted ratio of the point kinetics parameters , where and are the effective delayed neutron fraction and prompt neutron lifetime, respectively. To directly measure in a target critical system, this study proposes the application of the dynamic mode decomposition (DMD) to the reactor noise analysis based on the Rossi- method. The DMD-based Rossi- method enables us to robustly estimate the fundamental mode component of from the Rossi- histograms measured using multiple neutron detectors. Furthermore, the file-by-file moving block bootstrap method is newly proposed for the statistical uncertainty quantification of to prevent huge memory usage when the neutron count rate is high and/or the total measurement time is long. A critical experiment has been conducted at Kyoto University Critical Assembly to demonstrate the proposed method. As a result, the proposed method can uniquely determine the value of which the statistical uncertainty is smallest. By utilizing this experimental result of , numerical results of ratio using the continuous energy Monte Carlo code MCNP6.2 with recent nuclear data libraries, which are processed by the nuclear data processing code FRENDY, are validated.

EFFECTS OF FUEL DENSITY ON REACTIVITY COEFFICIENTS AND KINETIC PARAMETERS OF PEBBLE BED REACTOR

EFFECTS OF FUEL DENSITY ON REACTIVITY COEFFICIENTS AND KINETIC PARAMETERS OF PEBBLE BED REACTOR. Few decades ago a large number of nuclear reactors were designed to use HEU as the fuel. But the use of HEU is being discouraged since it can be used as a nuclear explosive material which makes its proliferation potential. Most of the HEU-fueled nuclear reactors in the world are either closed or converted into other types that have LEU fuel with 235U enrichment below 20%. To extend lifetime, LEU fuel with high density is developed. The change in fuel density from low to high will also change core neutronics and thermal-hydraulics of the reactor, and as a result, it affects the transient response of the reactor. This paper studies the effects of fuel density on reactivity coefficients and kinetic parameters of pebble bed reactor through several calculations with MCNP6 code combine with ENDF/B-VII.1 continuous energy cross-section nuclear data library. The overall calculation results show that the Doppler temperature coefficient (DTC) increases with increasing fuel density, but the moderator temperature coefficient (MTC) decreases due to hardening of neutron spectrum. Kinetic parameters such as effective delayed neutron fraction (βeff), prompt neutron lifetime(ℓ) and neutron generation time (Ʌ) which significantly reduced with increasing fuel density will strongly affect the reactor control and safety. The results of this work conclude that the selection of 9–15 g/cm3 fuel density should be considered carefully given that its effect on reactor controllability.Keywords: Fuel density, reactivity coefficients, kinetic parameters, MCNP6, ENDF/B-VII.1.

Analysis of Fuel Burnup and Transmutations at High Burnup of Sodium Fast Breeder Reactor

In this paper, the Monte Carlo N-Particle extended computer code (MCNP) were used to design a model of the European Sodium-cooled Fast Reactor. The multiplication factor, conversion factor, delayed neutrons fraction, doppler constant, control rod worth, sodium void worth, masses for major heavy nuclei, radial and axial power distribution at high burnup are studied. The results show that the reactor breeds fissile isotopes with a conversion ratio of 0.994 at fuel burnup 70 (GWd/T), and minor actinides are buildup inside the reactor core. The study aims to check the efficiency of the model on the calculation of the neutronic parameters of the core at high burnup.

Determination of Reactor Parameters for Different Subcritical Configurations of the Philippine Research Reactor-1 TRIGA Nuclear Fuel

The Philippine Research Reactor-1 (PRR-1) will be revived as a subcritical assembly for training, education, and research (SATER) in the field of nuclear science and technology. SATER will utilize the existing slightly irradiated nuclear fuel rods that have been maintained in the PRR-1 facility for more than 30 years. A subcritical arrangement for the fuel rods was chosen considering the inherent safety of this configuration. In this paper, we calculate relevant reactor parameters that characterize different annular and hexagonal subcritical configurations of 44 PRR-1 fuel rods. These parameters include the neutron multiplication factor ( ), the effective delayed neutron fraction ( ), and the mean neutron generation time ( ), which are essential quantities to describe reactor behavior. Calculations were performed using the well-validated Monte Carlo radiation transport code MCNP5 v.1.6 together with the ENDF/B- VII.1 evaluated nuclear data library. The maximum value is at 4.0 cm pitch for the chosen annular arrangement, while the maximum value is at 4.3 cm pitch for the chosen hexagonal arrangement. For these configurations, the reactor kinetic parameters were and for the annular arrangement, while and for the hexagonal arrangement. Results demonstrate that with 44 fuel rods, different fuel arrangements remain subcritical with a subcriticality margin that is at least or maximum of 0.97. The key reactor performance characteristics determined in this study can aid in the analysis of transient behavior and safety assessment of subcritical core configurations with TRIGA fuel rods. Our results provide support in expanding the utilization options for irradiated TRIGA fuel rods, even for other TRIGA facilities

Neutronic feasibility study of ([formula omitted] fuel in a modelled TRIGA research reactor

This paper presents a comparative analysis when transitioning from fuel meat based on uranium zirconium hydride (U-ZrH1.65) to one based on thorium-uranium-233 zirconium hydride ((Th-233U)-ZrH1.65) as an alternative option for TRIGA reactors. The aim of this study was to investigate the neutronic characteristics of (Th-233U)-ZrH1.65 fuels with two different Th-233U weight contents (i.e., 8.5 wt% and 12 wt%) and various U-233 enrichment levels. Using MCNP6.2 computer code based on the Monte Carlo method, a three-dimensional (3D) hypothetical TRIGA model was developed to investigate various neutronic parameters, such as the effects of U-233 enrichment and different (Th-233U) weight contents on the effective multiplication factor, burnup behavior, spent fuel composition, neutron flux, and fission product poison concentration. Since fuel temperature coefficients of reactivity and effective delayed neutron fractions play important roles in the reactor’s kinetics, these were also investigated, and the results were compared to the model with U-ZrH1.65 fuel. The obtained results revealed that the presence of Th-232 in the fuels did not drastically reduce the fuel temperature coefficients due to the zirconium hydride within the fuels providing most of the feedback in a TRIGA reactor. In contrast, the presence of U-233 in the (Th-233U)-ZrH1.65 fuels did reduce the effective delayed neutron fraction. Despite this, the lower transuranic generation in cores fueled by (Th-233U)-ZrH1.65 make this fuel advantageous in terms of achieving longer cycle lengths and less-hazardous nuclear waste for disposal.

Investigation of a new approach for regulating the reactivity and achieving economic feasibility using thorium in a blanket‐seed assembly of pressurized water reactors

The traditional methods for controlling reactor excess reactivity are expensive and have defects during reactor operation. By altering the fuel structure and fissile material distribution, this work presents a new technique for managing excess reactivity at the start of the fuel life. MCNPX code edition 2.7 was used to investigate the neutronic characteristics of the suggested blanket-seed assembly models and compare them with the reference PWR fuel assembly. Instead of U-238, Th-232 was suggested as a fertile nuclide. The impact of the proposed models on the infinite multiplication factor, fuel component concentrations, reactor-grade plutonium, minor actinides, radial flux of thermal neutrons, and the radial distribution of power has been analyzed. The safety parameters, including the control rods worth, Doppler effect, moderator temperature reactivity coefficients, and the effective delayed neutron fraction, have been studied to reach the most optimum fuel structure.

An adjoint sensitivity analysis of the Telegrapher’s neutron kinetic equations

The work presented herein is an approach to the kinetic uncertainty analysis of the Telegrapher’s neutron kinetic equations. The adjoint sensitivity analysis framework has been resorted to meet this end. Burnup induced uncertainty in the value of the delayed neutron fraction is addressed and its impact on the neutron population (as an observable quantity by the way of flux detectors) is investigated. Results compared against those obtained through an exact forward analysis approach confirm quite satisfactory accuracy of the proposed variational method. The effect of relaxation time on the adjoint system dynamics is moreover explored.

STUDY ON CRITICALITY AND NEUTRONIC SAFETY PARAMETERS OF NUSCALE FUEL ASSEMBLY

STUDY ON CRITICALITY AND NEUTRONIC SAFETY PARAMETERS OF NUSCALE FUEL ASSEMBLY. NuScale, a typical Pressurized Water Reactor (PWR) Small Modular Reactors (SMRs), offers a new opportunity for the future of nuclear industry. With 160 MW thermal power, NuScale has several advantages such as flexibility due to its modularity in construction. This work is focused on the study of criticality and neutronic safety parameters of NuScale fuel assembly using MCNP6 code and ENDF/B-VII library. The calculation results shows that criticality of fuel assembly type D is the highest among other assembly types because it has a fuel pin with pure UO2 without Gd2O3 concentration. The Doppler temperature coefficient (DTC) of fuel assembly type C is the most negative among other assemblies due to Doppler broadening effect on resonance region of capture cross section of 238U which is the highest concentration. The moderator temperature coefficient (MTC) of fuel assembly type D is the most negative among the other assembly types. The effective delayed neutron fraction (βeff) does not reflect a consistent trend among fuel assembly types. Fuel assembly type D shows the highest prompt neutron lifetime (ℓ) while the highest neutron generation time (Ʌ) is shown in assembly type C. It can be concluded that this study provides adequate results that can be used as a first step to carry out the neutronic computation and analysis of the NuScale full core.Keywords: Criticality, safety parameters, NuScale fuel assembly, MCNP6, ENDF/B-VII.

Calculation of kinetic parameters βeff and Λ with modified open source Monte Carlo code OpenMC(TD)

This work presents the methodology used to expand the capabilities of the Monte Carlo code OpenMC for the calculation of reactor kinetic parameters: effective delayed neutron fraction βeff and neutron generation time Λ. The modified code, OpenMC(Time-Dependent) or OpenMC(TD), was then used to calculate the effective delayed neutron fraction by using the prompt method, while the neutron generation time was estimated using the pulsed method, fitting Λ to the decay of the neutron population. OpenMC(TD) is intended to serve as an alternative for the estimation of kinetic parameters when licensed codes are not available. The results obtained are compared to experimental data and MCNP calculated values for 18 benchmark configurations.

Investigation of the safety features of advanced PWR assembly using SiC, Zr, FeCrAl and SS-310 as cladding materials

In this work, SiC (Silicon carbide), FeCrAl (ferritic), SS-310 (stainless steel 310) and Zirconium are simulated by MCNPX (Monte Carlo N‐Particle eXtended) code as cladding materials in advanced PWR (Pressurized Water Reactor) assembly. A number of reactor safety parameters are evaluated for the candidate cladding materials as reactivity, cycle length, radial power distribution of fuel pellet, reactivity coefficients, spectral hardening, peaking factor, thermal neutron fraction and delayed neutron fraction. The neutron economy presented by Zr and SiC models is analyzed through the burnup calculations on the unit cell and assembly levels. The study also provided the geometric conditions of all cladding materials under consideration in terms of the relation between fuel enrichment and cladding thickness from the viewpoint to achieve the same discharge burnup as the Zircaloy cladding. It was found that the SiC model participated in extending the life cycle by 2.23% compared to Zr. The materials other than SiC largely decreased discharge burnup in comparison with Zircaloy. Furthermore, the claddings with lower capture cross-sections (SiC and Zr) exhibit higher relative fission power at the pellet periphery. The simulation also showed that using SiC with a thickness of 571.15 μm and 4.83% U-235 can satisfy the EOL irradiation value as Zr. For reactivity coefficient, the higher absorbing materials (SS-310 and FeCrAl) exhibit more negative FTCs, MTCs and VRCs at the BOL But, at the intermediate stages of burnup Zr and SiC have a strong trend of negative reactivity coefficients. Finally, the delayed neutron fraction of SiC and Zr models is the highest among all the four models.

Comparison of probabilistic and deterministic methods for calculation of kinetic parameters of HWZPR

The accurate calculation of kinetic parameters, including the effective delayed neutron fraction and the neutron mean generation time, is very important in the study of time dependent neutron population behavior. To calculate these parameters, it is necessary to calculate the flux and adjoint flux in the phase space. In this paper, the kinetic parameters of the new core configuration of the heavy water zero power reactor (HWZPR) are calculated using two deterministic and probabilistic methods. The quantitative and qualitative discussion is presented to illustrate the advantages and disadvantages of each method. The maximum difference between the two accounting methods was less than 3%. The effect of different energy structures on the criticality and kinetic parameters has been investigated in the deterministic approach. Additionally, the calculated kinetic parameters of the HWZPR validated experimentally with a reactivity inject response. Good consistency has been observed between different methods of computation and experimental values.

Burnup and neutronic analysis for UOX and MOX fuel assemblies in VVER reactor

The present work examined the feasibility of utilizing Mixed Uranium–Plutonium Oxide (MOX) fuel in the VVER-1000 reactor core. In this study, about 30% of Uranium Oxide (UOX) fuel assemblies of the original core are replaced by MOX fuel assemblies. The effect of MOX fuel on the safety parameters is studied by using the MCNP6 code and the nuclear data library (ENDF/B)-VII.1. The safety parameters examined included the effective multiplication factor, temperature coefficients, void coefficient, boron coefficient, shutdown margin and the kinetic parameters. The excess reactivity, neutron flux and power density distributions and power peaking factor are also examined. All estimated values of reactivity coefficients analyzed for the core contains 30% MOX fuel confirmed to be negative, which guarantees the reactor safety and power stability in a normal state. Also, the effective delayed neutron fraction was lower than that of the full UOX core as well as the neutron lifetime was shorter, this shows that the MOX core responds faster than the UOX core. The maximum power factor for the MOX core was equal to the design limit. The shutdown margin satisfied the design requirement. The results confirmed that the VVER-1000 is able to meet the safety requirement for core designs with MOX loadings of up to 30%.

Explore the possible advantages of using thorium-based fuel in a pressurized water reactor (PWR) Part 1: Neutronic analysis

This study discusses the effect of using 232Th instead of 238U on the neutronic characteristics and the main operating parameters of the pressurized water reactor (PWR). MCNPX version 2.7 was used to compare the neutronic characteristics of UO2 with (Th, 235U)O2 and (Th, 233U) O2. Firstly, the infinity multiplication factor (Kinf), thermal neutron flux, and power distribution have been studied for the investigated fuel types. Secondly, the effect of Gd2O3 and Er2O3 on the Kinf and on the radial thermal neutron flux and thermal power has been investigated to distinguish which of them is more suitable than the other in reactivity management. Thirdly, to illustrate the effectiveness of 232Th in decreasing the inventory of both the actinides and non-actinides, the concentration of plutonium (Pu) isotopes and minor actinides (MAs) has been simulated with the fuel burnup. Besides, due to their large thermal neutron absorption cross-section, the concentrations of 135Xe, 149Sm, and 151Sm with the fuel burnup have been investigated. Finally, the main safety parameters such as the reactivity worth of the control rods (ρCR), the effective delayed neutron fraction βeff, and the Doppler reactivity coefficient (DRC) were calculated to determine to which extent these fuel types achieve the acceptable limits.

Development and verification of fast reactor burnup calculation module FRBurner in code system CBZ

ABSTRACT CBZ is a general-purpose reactor physics analysis code system, and FRBurner, which focuses on fast reactor burnup calculations, was developed recently with diverse combinations of available methodologies. Verification of this module is conducted with the OECD/NEA fast rector benchmark since this benchmark provides various types of fast reactors. Four key reactor physics parameters, effective neutron multiplication factor k eff, effective delayed neutron fraction β eff, sodium void reactivity ∆ρ void, and Doppler reactivity ∆ρ Doppler are the focus and compared to two references provided by JAEA and CEA, respectively. The biases between the results from FRBurner and the JAEA and CEA references on each of the above key parameters are less than 0.5%, 1%, 3% and 7%, and less than 1.0%, 4%, 12%, and 12%, respectively. The comparison indicates that the FRBurner module would provide acceptable results for general-type fast reactor physics analysis in research. As one innovation, the detailed burnup chain model, which is significantly different from a generally used pseudo fission product model in fast reactor neutronic analysis, is applied in FRBurner. The detailed burnup chain model helps FRBurner explicitly provide information about the inventory of fission products for nuclear waste management and spent fuel reprocessing.

Reactivity determination in a subcritical reactor: Computational, analytical and experimental methods

In this paper, an analysis of reactivity determination methods at Jordan Subcritical Assembly (JSA), which is driven by an external neutron source, is presented. Results from previous computational analysis estimate the effective multiplication factor, keff, to be 0.96553 ± 0.00004. Calculations of reactivity in a subcritical system predicts the source multiplication factor, ks, to be 0.9637 ± 0.0007, which is very close to the computed value of keff. The experimental values of the prompt neutron decay constant (α) using Rossi-α and Feynman-α methods, were 551.5 ± 8.0 (s−1) and 555.5 ± 9.8 (s−1), respectively. Using the values for effective delayed neutron fraction, βeff, and the mean neutron generation time, Λ, from the computational analysis, give estimations for keff by Rossi-α and Feynman-α to be 0.9548 ± 0.0009 and 0.9544 ± 0.0011, respectively, which are lower than the computed value by 1.1%.

Neutronic Analysis of LEU-started Molten Chloride Fast Reactor without Fuel Reprocessing

One of the rarely explored molten salt reactor (MSR) designs is the molten chloride fast reactor (MCFR). This MSR design employs chloride salt instead of fluoride and operated in a fast spectrum. MCFR brings all the advantages of an MSR including breeding whilst being able to burn plutonium and minor actinides efficiently. Since not many countries have access to civilian plutonium, MCFR can also be started using low-enriched uranium (LEU). This study is an initial neutronic analysis of an MCFR using LEU as its startup fuel. Parameters analyzed are conversion ratio (CR) and its neutronic safety, namely effective delayed neutron fraction (βeff), temperature coefficient of reactivity (TCR), and void coefficient of reactivity (VCR). The core is divided into Core Zone and Blanket Zone. The fuel composition of NaCl-UCl3 with a molar fraction ratio of 60:40 and 50:50 is used in Core Zone and Blanket Zone, respectively. The neutronic calculation is performed using MCNP6 code with ENDF/B-VII library. For reference geometry, CR is valued at 0.9298, βeff at 0.00731, TCR at -19.8 pcm/°C, and average VCR at -154.31 pcm/void%. Thereby, the MCFR fulfills inherent safety criteria. Although its value is remarkably high, CR can be further optimized by modifying the separator and reflector material.