Effective Delayed Neutron Fraction Research Articles

MCNP Simulation of Physics Parameters of Dispersion Fuels for Conversion of NIRR-1 to LEU

The Nigeria Research Reactor-1 (NIRR-1) is one of the Commercial Miniature Neutron Source Reactors (MNSRs) sited outside China and scheduled for conversion under the auspices of Reduced Enrichment for Research and Test Reactors (RERTR) program. Since 2006, the reduction in the fuel enrichment of MSNR facilities from greater than 90% HEU cores to less than 20% LEU cores has been embarked upon. Consequently in this work, the physics parameters of three dispersion LEU fuels, which include U3Si, U3Si2, and U9Mo enriched to 19.75% were determined by the MCNP code to investigate their suitability for the conversion of NIRR-1 to LEU. The following reactor core physics parameters were computed for the LEU fuel options: clean cold core excess reactivity (ρex), control rod (CR) worth, shut down margin (SDM), neutron flux distributions in the irradiation channels and kinetics data (i.e. effective delayed neutron fraction, βeff and prompt neutron lifetime, lf). Results are compared with experimental and calculated data of the current HEU core and indicate that it would be feasible to use any of the LEU options for the conversion of commercial MNSR in general and NIRR-1 in particular from HEU to LEU.

Sensitivity and uncertainty analysis of βeff for MYRRHA using a Monte Carlo technique

This paper presents a nuclear data sensitivity and uncertainty analysis of the effective delayed neutron fraction βeff for critical and subcritical cores of the MYRRHA reactor using the continuous-energy Monte Carlo N-Particle transport code MCNP. The βeff sensitivities are calculated by the modified k-ratio method proposed by Chiba. Comparing the βeff sensitivities obtained with different scaling factors a introduced by Chiba shows that a value of a = 20 is the most suitable for the uncertainty quantification of βeff. Using the calculated βeff sensitivities and the JENDL-4.0u covariance data, the βeff uncertainties for the critical and subcritical cores are determined to be 2.2 ± 0.2% and 2.0 ± 0.2%, respectively, which are dominated by delayed neutron yield of 239Pu and 238U.

Current Mode Neutron Noise Measurements in the Zero Power Reactor CROCUS

The present article is an overview of developments and results regarding neutron noise measurements in current mode at the CROCUS zero power facility. Neutron noise measurements offer a non-invasive method to determine kinetic reactor parameters such as the prompt decay constant at criticality α = βeff / λ, the effective delayed neutron fraction βeff, and the mean generation time λ for code validation efforts. At higher detection rates, i.e. above 2×104 cps in the used configuration at 0.1 W, the previously employed pulse charge amplification electronics with BF3 detectors yielded erroneous results due to dead time effects. Future experimental needs call for higher sensitivity in detectors, higher detection rates or higher reactor powers, and thus a generally more versatile measurement system. We, therefore, explored detectors operated with current mode acquisition electronics to accommodate the need. We approached the matter in two ways: 1) By using the two compensated 10B-coated ionization chambers available in CROCUS as operational monitors. The compensated current signal of these chambers was extracted from coremonitoring output channels. 2) By developing a new current mode amplification station to be used with other available detectors in core. Characteristics and first noise measurements of the new current system are presented. We implemented post-processing of the current signals from 1)and 2) with the APSD/CPSD method to determine α. At two critical states (0.5 and 1.5 W), using the 10B ionization chambers and their CPSD estimate, the prompt decay constant was measured after 1.5 hours to be α=(156.9 ± 4.3) s-1 (1σ). This result is within 1σ of statistical uncertainties of previous experiments and MCNPv5-1.6 predictions using the ENDF/B-7.1 library. The newsystem connected to a CFUL01 fission chamber using the APSDestimate at 100 mW after 33 min yielded α = (160.8 ± 6.3) s-1, also within 1σ agreement. The improvements to previous neutron noise measurementsinclude shorter measurement durations that can achievecomparable statistical uncertainties and measurements at higherdetection rates.

Fuel cycle analysis of molten salt reactors based on coupled neutronics and thermal-hydraulics calculations

A new fuel cycle analysis code named FAMOS was developed for molten salt reactor (MSR) applications by extending the DIF3D code to model the transport of delayed neutron precursors and by coupling the extended DIF3D code with an in-house multi-channel thermal-hydraulics solver and an in-house nuclide depletion solver for the simulation of fuel depletion, reprocessing and feeding. FAMOS can also calculate the effective delayed neutron fraction accounting for the motion of fuel salt, the neutron generation time and the reactivity feedback coefficients. In order to verify the FAMOS code, the molten salt fast reactor (MSFR) proposed in the EVOL (Evaluation and Viability of Liquid Fuel Fast Reactor System) project was analyzed using FAMOS and the calculation results were compared with the reported results. The numerical results indicate that FAMOS is capable of simulating the fuel cycle process of MSRs with good accuracy. FAMOS was applied to the fuel cycle analysis of the molten salt breeder reactor (MSBR) as well. The effects of the temperature distribution and the drift of delayed neutron precursors on the neutronics characteristics and fuel cycle performances of MSBR were evaluated using three different calculation models. The drift of delayed neutron precursors and the temperature distribution shift the axial power distribution upward and downward, respectively, and the latter effect is stronger than the former one. However, the impacts of the temperature distribution and the drift of delayed neutron precursors on the depletion and buildup behaviors of actinides and fission products in MSRs are negligible. It is also possible to achieve a 233U conversion ratio larger than 1.0 without protactinium removal and thus to operate the MSBR by only feeding 232Th, except for the initial start-up stage.

Beta-effective sensitivity and uncertainty analysis of MYRRHA reactor for possible use in nuclear data validation and improvement

New reactor concept studies, such as those of the MYRRHA Accelerator Driven System (ADS), with their extended nuclear data needs relaunched the interest in sensitivity and uncertainty (S/U) studies. S/U analysis play an essential role in assuring the accuracy and reliability of nuclear reactor computations and provide an insight in the physical phenomena involved in the neutron transport. This paper presents the study of the MYRRHA critical core, which is designed to operate at 100 MW thermal power. The previously presented analysis of the MYRRHA effective multiplication factor (keff) was extended to study the sensitivity and uncertainty in kinetic parameters, such as the effective delayed neutron fraction (βeff). The sensitivity and uncertainty computations were performed by means of the SUSD3D code based on forward and adjoint-flux first-order perturbation theory. The sensitivity coefficients of βeff with respect to the basic nuclear data were calculated by the k-ratio derivation method proposed in 2011. The main components of the uncertainties in βeff were estimated using the JENDL-4.0u covariance data. The energy dependent sensitivity profiles of keff and βeff with respect to nuclear data are compared in view of exploiting the differences among them for a physically more complete, comprehensive and consistent nuclear data validation and improvement procedure. The accuracy of the keff and βeff measurements and calculations has to be carefully evaluated in order to conclude on the nuclear data status and possible improvements.

Effective point kinetic parameters calculation in Tehran research reactor using deterministic and probabilistic methods

The exact calculation of point kinetic parameters is very important in nuclear reactor safety assessment, and most sophisticated safety codes such as RELAP5, PARCS, DYN3D, and PARET are using these parameters in their dynamic models. These parameters include effective delayed neutron fractions as well as mean generation time. These parameters are adjoint-weighted, and adjoint flux is employed as a weighting function in their evaluation. Adjoint flux calculation is an easy task for most of deterministic codes, but its evaluation is cumbersome for Monte Carlo codes. However, in recent years, some sophisticated techniques have been proposed for Monte Carlo-based point kinetic parameters calculation without any need of adjoint flux. The most straightforward scheme is known as the “prompt method” and has been used widely in literature. The main objective of this article is dedicated to point kinetic parameters calculation in Tehran research reactor (TRR) using deterministic as well as probabilistic techniques. WIMS-D5B and CITATION codes have been used in deterministic calculation of forward and adjoint fluxes in the TRR core. On the other hand, the MCNP Monte Carlo code has been employed in the “prompt method” scheme for effective delayed neutron fraction evaluation. Deterministic results have been cross-checked with probabilistic ones and validated with SAR and experimental data. In comparison with experimental results, the relative differences of deterministic as well as probabilistic methods are 7.6 and 3.2%, respectively. These quantities are 10.7 and 6.4%, respectively, in comparison with SAR report.

Analysis of thorium fuel feasibility in large scale gas cooled fast reactor using MCNPX code

In this paper thorium fuel feasibility in large scale Gas Cooled Fast Reactor (GCFR) is investigated. The neutronics benchmark used in this study, GFR2400, corresponds to a 2400MWth GFR concept proposed by the French CEA. MCNPX computational code is used to design a 3D heterogeneous model of the GFR2400 core. A detailed feasibility analysis of the performance of thorium fuel cycle is performed by using thorium as an alternative fertile fuel for the natural uranium vector of the reference core design. The most essential neutronic parameters characterizing the core are determined both for beginning of life (BOL) conditions as well as during burnup. Also, a three-dimensional core cycle-by-cycle simulations is performed to allow explicit characterization of the core behavior and safety-related parameters during both open and equilibrium cycles. The thorium-based core shows favorable neutronic characteristics with an acceptable control and safety parameters for both BOL and open cycle states. The depressurization reactivity effect and core expansion coefficients (axial and radial) show improvement compared to the uranium-based core. However, this improvement is compensated by the deterioration in the effective delayed neutron fraction (β-eff) and the Doppler reactivity Effect. The results of isotopic transmutation and fuel burnup confirm the capability of the core to work in both open and closed cycles and to self-recycle its own MA vector and plutonium of LWRs. However-as is the case in other fast reactors including the uranium-based GFR2400 core, the fuel cycle closure causes safety related parameters to degrade.

Neutron multiplication in random media: Reactivity and kinetics parameters

Eigenvalue problems for neutron transport in random geometries are key for many applications, ranging from reactor design to criticality safety. In this work we examine the behaviour of the reactivity and of the kinetics parameters (the effective delayed neutron fraction and the effective neutron generation time) for three-dimensional UOX and MOX assembly configurations where a portion of the fuel pins has been randomly fragmented by using various mixing statistics. For this purpose, we have selected stochastic tessellations of the Poisson, Voronoi and Box type, which provide convenient models for the random partitioning of space, and we have generated an ensemble of assembly realizations; for each geometry realization, criticality calculations have been performed by using the Monte Carlo code Tripoli-4®, developed at CEA. We have then examined the evolution of the ensemble-averaged observables of interest as a function of the average chord length of the random geometries, which is roughly proportional to the correlation length of the fuel fragmentation. The methodology proposed in this work is fairly general and could be applied, e.g., to the assessment of re-criticality probability following severe accidents.

A method for predicting fuel maintenance in once-through MSRs

Liquid fuel molten salt reactors allow reactivity control by material addition. This paper presents a method to adjust material flows in a molten salt reactor to keep the core critical, and to maintain desired reduction-oxidation potential in the core salt melt. The method is aimed at low-enriched uranium fueled thermal systems. It is developed as a Python library and uses Serpent2 Monte-Carlo transport and depletion code. A toy 300MW(th) reactor with a FLiBe carrier salt is employed to demonstrate the performance of the method over 10 full power years. Results of the calculation are presented, including material flows, conversion ratio, effective delayed neutron fraction, and expected limits on trifluoride concentrations and graphite lifetime are investigated. This method lays a foundation for future studies including fuel cycle performance of molten salt reactors and dynamic behavior of the core during depletion.

Design and optimization of the new LEU MNSR for neutron radiography using thermal column to upgrade thermal flux

In this article, the preliminary feasibility of using the new miniature neutron source reactor (MNSR) with unique thermal column (TC), as a neutron source for neutron radiography (NR), is investigated. The new reactor core is designed based on the Isfahan MNSR with high density low-enriched uranium (LEU) UO2 fuel instead of high-enriched uranium (HEU) UAl4-Al fuel. The safety of the LEU reactor is studied by the calculation of important neutronics parameters such as core excess reactivity, central control rod worth, shut-down margin (SDM) and effective delayed neutron fraction. The vertical collimator dimensions, the materials to be employed in the collimator, and the collimator position are optimized using MCNP Monte Carlo code. The results showed that presence of the graphite thermal column increases thermal neutrons flux and reduces gamma-ray dose that can satisfy the requirements for neutron radiography. Therefore there is no need for extra gamma filters, fast neutron filters, and neutron moderators at the collimator. We emphasize here that the current designed LEU MNSR can be applied for NR and neutron activation analysis (NAA).

Calculation of the fuel temperature coefficient of reactivity considering non-uniform radial temperature distribution in the fuel rod

Abstract The safe operation of a reactor is based on feedback models. In this paper we attempted to discuss the influence of a non-uniform radial temperature distribution on the fuel rod temperature coefficient of reactivity. The paper demonstrates that the neutron properties of a reactor core is based on effective temperature of the fuel to obtain the correct fuel temperature feedback. The value of volume-averaged temperature being used in the calculations of neutron physics with feedbacks would result in underestimating the probable event. In the calculation it is necessary to use the effective temperature of the fuel in order to provide correct accounting of the fuel temperature feedback. Fuel temperature changes in different zones of the core and consequently reactivity coefficient change are an important parameter for analysis of transient conditions. The restricting factor that compensates the inserted reactivity is the temperature reactivity coefficient and effective delayed neutron fraction.

Coupled neutronics and thermal-hydraulics simulation of molten salt reactors based on OpenMC/TANSY

An OpenMC/TANSY code system has been developed in this paper for the coupled neutronics and thermal-hydraulics simulations of MSRs. The homogenized cross section data library is generated using the continuous-energy Monte-Carlo code OpenMC which provides significant modeling flexibility compared against the traditional deterministic lattice transport codes. The few-group cross sections generated by OpenMC are provided to TANSY which is based on OpenFOAM to perform the full-core coupled neutronics and thermal-hydraulics simulations. In order to verify the OpenMC/TANSY code system, the Molten Salt Fast Reactor benchmark problem was calculated and both the neutronics results and thermal-hydraulics results were compared with those obtained by other researchers. For application of the OpenMC/TANSY codes sequence, the simulation of a representative molten salt reactor core MOSART has been performed. First, to verify the generation of the few-group cross sections, the neutronics results obtained by the “two-step” scheme were compared with those obtained by full-core Monte-Carlo solution. Good agreement can be observed for the multiplication factor as well as the power distributions. Then the full-core coupled neutronics and thermal-hydraulics simulation was performed. The distribution of the important neutronics and thermal-hydraulics parameters are presented and analyzed in detailed in this paper. For the further study of the characteristics of MSRs, several effects like the external-loop transit time, inlet velocity and inlet temperature on the effective delayed neutron fraction and critical fuel concentration have been analyzed. The numerical results indicated that the TANSY code with the cross section library generated by OpenMC has the capability for the steady-state analysis and reactor core design of MSRs.

Investigation of neutronic and safety parameters variation in 5 MW research reactor due to U3O8Al fuel conversion to ThO2 + U3O8Al

Abstract Thorium-based fuels could comprise several advantages and are being investigated as a potentially competitive option with uranium-based fuels for research or power reactors. The present study investigates computationally the application of two different thorium-based fuels in a research reactor. Void and temperature reactivity coefficients, safety factor, power peaking factor, neutron generation time, effective delayed neutron fraction and 135Xe worth parameter were investigated for the fuel conversions. The results showed both the investigated fuels would not significantly disturb neutronic and safety parameters of the modeled core in comparison with its routine fuel loading. However, 235-enriched thorium based fuel concluded in noticeably reduction of High Level Waste (HLW) but 233-enriched type could be taken in attention because of its longer fuel cycle (∼15 %) and integrated neutron flux (∼23 %).

Calculation of adjoint-weighted kinetic parameters with the reactor Monte Carlo code RMC

In this work, the capability of computing adjoint-weighted kinetic parameters, including effective delayed neutron fraction and neutron generation time, was implemented in the Reactor Monte Carlo (RMC) code based on the iterated fission probability (IFP) method. Three algorithms, namely, the Non-Overlapping Blocks (NOB) algorithm, the Multiple Overlapping Blocks (MOB) algorithm and the superhistory algorithm, were implemented in RMC to investigate their accuracy, computational efficiency and estimation of variance. The algorithms and capability of computing kinetic parameters in RMC were verified and validated by comparison with MCNP6 as well as experimental results through a set of multi-group problems and continuous-energy problems.

Sensitivity of power spectral density techniques to numerical parameters in analyzing neutron noise experiments

Power spectral density (PSD) methods are well-known and widely used for the analysis of neutron noise experiments and obtaining the reactor's integral kinetic parameters, i.e., the effective delayed neutron fraction βeff and the prompt neutron generation time Λ. Many uncertainties are usually associated with PSD methods, e.g., statistical fluctuations in the neutron flux, power drifts, uncertainties in the Diven factor, the integral fission rate, and in the reactivity value. However, the uncertainty associated with the numerical parameters used in the power spectrum calculation procedure is hardly discussed in the literature and generally overlooked.The aim of this paper is to study the uncertainties in the kinetic parameters of a reactor core, obtained by PSD methods, which are associated with the numerical parameters of the method. A comprehensive estimation of the kinetic parameters, including all other uncertainties, is not pursued. In this paper, PSD methods are implemented to analyze critical and subcritical configurations of the MINERVE zero power reactor in order to measure its integral kinetic parameters βeff and Λ. Both cross and auto power spectral densities are calculated and the kinetic parameters are obtained via Lorentzian curve fitting over the calculated PSD. The sensitivity of the obtained kinetic parameters to the choice of numerical parameters used for spectrum calculations is studied and found to be significant with respect to other uncertainties. A novel methodology is proposed for analyzing the kinetic parameters' sensitivity to the PSD calculations and for quantifying the associated uncertainties.

Evaluation of [formula omitted] measurements from BERENICE programme with Tripoli4® and uncertainty qualification

The effective delayed neutron fraction (βeff) is an important characteristic of nuclear reactors since it affects transients significantly. It is therefore important to characterise it correctly. The use of best-estimate codes and data together with an evaluation of the uncertainties is required not only for their use in safety studies but also to assess reactivity effects which are being measured relative to the effective delayed neutron fraction (βeff) in $.The use of the Iterated Fission Probability method in the Monte Carlo code Tripoli4® confirms results obtained with deterministic codes such as ERANOS for calculating βeff. The asset of Tripoli4® is the possibility to get a better representation of experimental cores. Its use for evaluating the calculated parts of the “experimental” βeff of the BERENICE experimental programme has led to significant improvements of the C/E ratios, especially the R2 experimental core.This work is in support of the ASTRID project which developed the concept of a Sodium cooled Fast Reactor (SFR). In order to get a reduced uncertainty it is recommended to repeat the βeff measurement within the future GENESIS experimental programme in the refurbished zero power reactor (ZPR) MASURCA with an improved noise measurement technique.The nuclear data uncertainty propagation has led to a 2.8% uncertainty for U-Pu core and 2.6% for enriched uranium cores (with JEFF-3.2) with main contributors being the delayed neutron fission yield and the fission cross section of U238.

Effective delayed neutron fraction by Rossi-α method in accelerator-driven system experiments with 100 MeV protons at kyoto university critical assembly

ABSTRACTIn the accelerator-driven system (ADS), the effective delayed neutron fraction (βeff) is a requisite for converting the subcriticality from dollar units to pcm units. To evaluate the accurate βeff value in ADS, measurement of βeff complements its calculation methodology and the nuclear data on delayed neutrons. Subcriticality measurements are carried out by the pulsed neutron source method in the Kyoto University Critical Assembly, and the neutron noise analyses are conducted by the Rossi-α method with the pulsed shape of the spallation neutron source. The value of βeff is deduced with the combined use of measured subcriticality in dollar units and correction parameters by MCNP6.1 together with JENDL-4.0 and JENDL/HE-2007. A comparison between the calculated and the measured βeff represents the acceptable accuracy within the subcriticality range of around keff = 0.93 in the ADS operations. Here, the applicability of the measurement methodology based on the Rossi-α method is demonstrated by varying the subcriticality.

Effect of Kinetic Parameters on Simultaneous Ramp Reactivity Insertion Plus Beam Tube Flooding Accident in a Typical Low Enriched U 3 Si 2 -Al Fuel-Based Material Testing Reactor-Type Research Reactor

Abstract This work looks at the effect of changes in kinetic parameters on simultaneous reactivity insertions and beam tube flooding in a typical material testing reactor-type research reactor with low enriched high density (U3Si2-Al) fuel. Using a modified PARET code, various ramp reactivity insertions (from $0.1/0.5 s to $1.3/0.5 s) plus beam tube flooding ($0.5/0.25 s) accidents under uncontrolled conditions were analyzed to find their effects on peak power, net reactivity, and temperature. Then, the effects of changes in kinetic parameters including the Doppler coefficient, prompt neutron lifetime, and delayed neutron fractions on simultaneous reactivity insertion and beam tube flooding accidents were analyzed. Results show that the power peak values are significantly sensitive to the Doppler coefficient of the system in coupled accidents. The material testing reactor-type system under such a coupled accident is not very sensitive to changes in the prompt neutron life time; the core under such a coupled transient is not very sensitive to changes in the effective delayed neutron fraction.

Understanding kinetic behaviour of TRIGA reactor core

The kinetic parameters govern the transient behaviour of a nuclear reactor. Estimation of these parameters has great importance for the safe design and operation of a nuclear reactor. In order to understand the kinetic behaviour of TRIGA core, the influence of graphite dummy elements, absorber materials and beam ports on kinetic parameters has been studied. To do so, the effective delayed neutron fraction (βeff), prompt removal lifetime (e) and mean neutron generation time (Λ) have been calculated for an infinite system of LEU fuel cells. The investigation has been extended to four different cases of TRIGA reactor core. Monte Carlo simulation has been carried out to compute the kinetic parameters by Monte Carlo code MCNP5. The βeff value remains unchanged but the e and Λ values are found to be sensitive to the local perturbations of core components.

Combined use of k-effective and beta-effective measurements for nuclear data validation and improvement

Nuclear cross-sections are being re-evaluated time and again since decades, and in the recent years they are “tuned” to match better in particular the k eff measurements. However, the k eff parameter as a very global parameter provides simply too many possible combinations and variations, all more or less in agreement with the differential cross section data measurements and the associated uncertainties. Adding other types of integral measurements, such as shielding and kinetics benchmarks, on the cross-section evaluation menu is expected to provide a complementary and a more complete view on the problem. This paper discusses the advantages of exploiting the effective delayed neutron fraction (β eff ) measurements for nuclear data (ND) evaluation and testing. The cross-section sensitivity and uncertainty analysis revealed some crucial differences in the k eff and β eff sensitivity profiles which are favorable for ND validation.