Reactor Kinetic Parameters Research Articles

Analysis of kinetic parameters of 3 MW TRIGA Mark-II research reactor using the SRAC2006 code system

The aim of this study is to evaluate the kinetic parameters of 3MW TRIGA Mark-II research reactor at AERE, Savar, Dhaka, Bangladesh from the viewpoint of reactor safety. The most important kinetic parameters of nuclear reactors are the effective delayed neutron fraction (βeff), the effective decay constant for ith family of delayed neutron precursor (λeff,i), the prompt neutron lifetime (lp) and the mean neutron generation time (Λ). These parameters are calculated using the 3-D diffusion code SRAC-CITATION of the comprehensive neutronics calculation code system SRAC2006 based on the evaluated nuclear data libraries JENDL-3.3 and ENDF/B-VII.0 in both cases. The calculated results of reactor kinetic parameters are compared to the available safety analysis report (SAR) values of 3MW TRIGA Mark-II reactor by General Atomic as well as the MCNP5 values (numerically benchmark) based on the evaluated nuclear data library ENDL/B-VII.0. It was found that in most cases, the calculated results of kinetic parameters demonstrate a good agreement between the JENDL-3.3 and the ENDF/B-VII libraries as well as the SAR and the MCNP5 values respectively. Therefore, this study will be essential to improve the basic nuclear data of reactor kinetic parameters for safe operation of 3MW TRIGA Mark-II research reactor.

Adjoint-Weighted Tallies for k-Eigenvalue Calculations with Continuous-Energy Monte Carlo

A Monte Carlo method is developed that performs adjoint-weighted tallies in continuous-energy k-eigenvalue calculations. Each contribution to a tally score is weighted by an estimate of the relative magnitude of the fundamental adjoint mode, by way of the iterated fission probability, at the phase-space location of the contribution. The method is designed around the power iteration method such that no additional random walks are necessary, resulting in a minimal increase in computational time. The method is implemented in the Monte Carlo N-Particle (MCNP) code. These adjoint-weighted tallies are used to calculate adjoint-weighted fluxes, point reactor kinetics parameters, and reactivity changes from first-order perturbation theory. The results are benchmarked against discrete ordinates calculations, experimental measurements, and direct Monte Carlo calculations.

Development of Calculation Technique for Iterated Fission Probability and Reactor Kinetic Parameters Using Continuous-Energy Monte Carlo Method

New algorithms and techniques are developed for calculating the iterated fission probability (IFP ) using a Monte Carlo eigenvalue calculation scheme. The proportionality of the calculated IFP to the adjoint flux is confirmed numerically. This IFP is then used with the MCNP code and its point-wise cross section data libraries to calculate kinetic parameters weighted by the adjoint flux. Experimental data for critical cores validate these theoretical estimates.

Development of Calculation Technique for Iterated Fission Probability and Reactor Kinetic Parameters Using Continuous-Energy Monte Carlo Method

Development of Calculation Technique for Iterated Fission Probability and Reactor Kinetic Parameters Using Continuous-Energy Monte Carlo Method

Calculation of kinetic parameters for mixed TRIGA cores with Monte Carlo

Abstract Modern Monte Carlo transport codes in combination with fast computer clusters enable very accurate calculations of the most important reactor kinetic parameters. Such are the effective delayed neutron fraction, β eff , and mean neutron generation time, Λ . We calculate β eff and Λ for various realistic and hypothetical annular TRIGA Mark II cores with different types and amount of fuel. It is observed that the effective delayed neutron fraction strongly depends on the number of fuel elements in the core or on the core size. β eff varies for 12 wt.% uranium standard fuel with 20% enrichment from 0.0080 for a small core (43 fuel rods) to 0.0070 for a full core (90 fuel rods). It is found that calculated value of β eff strongly depends also on the nuclear data set used in calculations. The prompt neutron lifetime mainly depends on the amount (due to either content or enrichment) of 235 U in the fuel as it is approximately inversely proportional to the average absorption cross-section. It varies from 28 μs for 30 wt.% uranium content fuelled core to 48 μs for 8.5 wt.% uranium content LEU fuelled core. Description of the calculation method and detailed results are presented in the paper.

Calculation of the importance-weighted neutron generation time using MCNIC method

In advanced nuclear power systems, such as ADS, the need for reliable kinetics parameters is of considerable importance because of the lower value for β eff due to the large amount of transuranic elements loaded in the core of those systems. All reactor kinetic parameters are weighted quantities. In other words each neutron with a given position and energy is weighted with its importance. Neutron generation time as an important kinetic parameter, in all nuclear power systems has a significant role in the analysis of fast transients. The difference between non-weighted neutron generation time; Λ ; standard in most Monte Carlo codes; and the weighted one Λ † can be quite significant depending on the type of the system. In previous work, based on the physical concept of neutron importance, a new method; MCNIC; using the MCNP code has been introduced for the calculation of neutron importance in fissionable assemblies for all criticality states. In the present work the applicability of MCNIC method has been extended for the calculation of the importance-weighted neutron generation time. The influence of reflector thickness on importance-weighted neutron generation time has been investigated by the development of an auxiliary code, IWLA, for a hypothetic assembly. The results of these calculations were compared with the non-weighted neutron generation times calculated using the Monte Carlo code MCNP. The difference between the importance-weighted and non-weighted quantity is more significant in a reflected system and increases with reflector thickness.

The Behavior of ThO2-Based Fuel Rods during Normal Operation and Transient Events in LWRs

The thermal, mechanical, and chemical behavior of both thorium and uranium dioxide (ThO2-UO2) and thorium and plutonium dioxide (ThO2-PuO2)-based fuels during in-service and hypothetical accident conditions in light water reactors (LWRs) is described. These fuels offer the possibility for increased proliferation resistance and a reduction in the stockpile of weapons-grade and reactor-grade PuO2 as well as being a more stable waste form. The behavior is described for three different designs of ThO2-based fuels: a homogeneous mixture of ThO2-UO2, a microheterogeneous arrangement of the ThO2 and UO2, and a homogeneous mixture of ThO2-PuO2. The behavior was calculated with widely known LWR analysis tools extended for ThO2-based fuels: (a) MATPRO for calculating material properties, (b) FRAPCON-3 for calculating in-service fuel temperature and fission-gas release, (c) VIPRE-01 for calculating the possibility for departure from nucleate boiling, (d) HEATING7 for calculating in-service two-dimensional temperature distributions in microheterogeneous fuel, (e) SCDAP/RELAP5-3D for calculating the transient reactor system behavior and fuel behavior during loss-of-coolant accidents, and (f) FRAP-T6 for calculating the vulnerability of the cladding to cracking due to swelling of the fuel during hypothetical reactivity-initiated accidents.The analytical tools accounted for the following differences in ThO2-based fuels relative to 100% UO2 fuel: (a) higher thermal conductivity, lower density and volumetric heat capacity, less thermal expansion, and higher melting point; (b) higher fission-gas production for 233U fission than 235U fission, but a lower gas diffusion coefficient in the ThO2 than in the UO2; (c) less plutonium accumulation at the rim of the fuel pellets; (d) greater decay heat; (e) microheterogeneous arrangement of fuel; and (f) more-negative moderator temperature and Doppler coefficients and a smaller delayed-neutron fraction. The newly developed models for ThO2 were checked against data from the light water breeder reactor program. Calculations by these analytical tools indicate that the in-service and transient performance of homogeneous ThO2-UO2-based fuels with respect to safety is generally equal to or better than that of 100% UO2 fuel. The in-service and transient temperatures in the most promising neutronic design of microheterogeneous ThO2–UO2–based fuel are greater than the temperatures in 100% UO2 fuel but are still within normal LWR safety limits. The reactor kinetics parameters for ThO2-PuO2–based fuel cause a higher transient reactor power for some postulated accidents, but in general, the margin of safety for ThO2-PuO2 fuels is equal to or greater than that in 100% UO2 fuels.

Reactor noise in accelerator driven systems

Reactor noise methods have long been used for measuring reactor kinetics parameters in low power critical and sub critical reactors. The recent interest in accelerator driven sub critical systems (ADS) and the necessity of monitoring their degree of sub criticality has created a renewed interest in these methods. In this paper we derive formulae for interpreting some of the classical noise methods such as Rossi alpha, the variance to mean, auto correlation and power spectral density, as well as some recently proposed methods, such as the cross power spectral density between the proton current and the neutron detector signals for systems driven by periodically pulsed sources having a large multiplicity. It is shown that the rather high contribution due to the large multiplicity of the accelerator driven source to the variance (or any other measure of the noise) is significantly different for a periodically pulsed source as compared to that for Poisson source events.

Reactor Noise Experiments by Using Acquisition System for Time Series Data of Pulse Train

An acquisition system for the time series data of a pulse train was developed to perform the reactor noise experiments. Using this acquisition system, some important reactor kinetic parameters can be obtained by using various data analysis methods for the reactor noise experiments, which facilitates to achieve new experimental knowledge about characteristics of them proposed so far through direct comparison among the results of various methods. In the present study, the prompt neutron decay constant αpat two near delayed critical states of a core constructed in the Kyoto University Critical Assembly were measured by four well-known data analysis methods; the Feynman-α, the Rossi-α (Type-I and Type-II) and the frequency analysis methods. The results showed that ap values measured by the Type-I Rossi-α and the frequency analysis methods agreed with those by the Feynman-α method considering both effects of the delayed neutrons and the counting loss of neutron counter, whereas it was impossible to obtain the adequate αp values by the Type-II Rossi-α method. Moreover, it was demonstrated that one cannot neglect both effects for the measurement of αp by the Feynman-a method, whereas those by the Type-I Rossi-a and the frequency analysis methods hardly suffered.

Feynman-α correlation analysis by prompt-photon detection

Two-detector Feynman-α measurements were carried out using the UTR-KINKI reactor, a light-water-moderated and graphite-reflected reactor, by detecting high-energy, prompt gamma rays. For comparison, the conventional measurements by detecting neutrons were also performed. These measurements were carried out in the subcriticality range from 0 to $1.8. The gate-time dependence of the variance-and covariance-to-mean ratios measured by gamma-ray detection were nearly identical with those obtained using standard neutron-detection techniques. Consequently, the prompt-neutron decay constants inferred from the gamma-ray correlation data agreed with those from the neutron data. Furthermore, the correlated-to-uncorrelated amplitude ratios obtained by gamma-ray detection significantly depended on the low-energy discriminator level of the single-channel analyzer. The disriminator level was determined as optimum for obtaining a maximum value of the amplitude ratio. The maximum amplitude ratio was much larger than that obtained by neutron detection. The subcriticality dependence of the decay constant obtained by gamma-ray detection was consistent with that obtained by neutron detection and followed the linear relation based on the one-point kinetic model in the vicinity of delayed critical. These experimental results suggest that the gamma-ray correlation technique can be applied to measure reactor kinetic parameters more efficiently.

Reactor noise analysis of the experimental fast reactor JOYO

Experimental fast reactor JOYO achieved its first criticality in 1977 spring. Since then a series of reactor noise analysis has been carried out. During zero power testing, the reactor kinetic parameters, i.e. β/l and subcriticality, were obtained, applying band-pass filter and polarity correlation methods. As the high power operation started from 1978, the reactor noise analysis based on neutron flux spectra has been proceeded. This paper presents study of noise analysis for rather lower frequency region at high power operation. At 50 Mwt power level operation, noises of neutron flux signals, reactor inlet temperatures and subassembly outlet temperatures were measured, and then auto power spectral densities and coherence functions were calculated. The frequency range for the noise analysis is for 10 −3 − 10 −1 Hz. In the APSD of neutron flux, a low peak was observed at 2.5×10 −2 Hz. For neutron noise, it was estimated that the noises in the frequency region lower than 1.5 × 10 −2 Hz is a core inherent temperature noises, while that in the frequency region higher than 1.5 × 10 −2 Hz is a core inherent neutron noise. The coherence function of neutron fluxes, which are signals of ex-core detectors located at opposite positions against the core center, indicated that some space dependent phenomena of neutron noise might exist in 5×10 −3 − 6×10 −2 Hz. The space dependency in 5×10 −3 − 1.5×10 −2 Hz were estimated to be produced by the sodium temperature noise due to the insufficient mixing of two primary coolant loops. In 1.5×10 −2 − 6×10 −2 Hz, the peak existed at 2.5×10 −2 Hz and reverse phases were observed in the coherence of two neutron signals at opposite side of the core. With the investigation of phase relations between neutron flux and subassembly outlet temperature in this frequency region, the phenomenon was estimated to be due to mechanical vibration of some reactor core components. When the static gain of reactivity-to-power transfer function increased, the changes of reactor noise characteristics were observed. Since the change of the gain indicates a change of the reactor core characteristics, the observation of reactor noise characteristics has a potential for the surveillance of the reactor core.

The determination of reactor kinetic parameters in a two-core reactor

The kinetic parameters, α the coupling coefficient and \\ ̄ gt the mean neutron transit time have been determined using a reactor oscillator on the coupled-core of the Queen Mary College research reactor. By using correlation techniques it has proved possible to use detectors small enough to be inserted in the fuel tanks. It is shown that the simplified Baldwin model with one-group diffusion theory is inadequate to describe the kinetic behaviour and the experimentally-determined parameters are dependent upon the positioning of the detectors.

Upper-Bound Estimation of Bilinear Functionals in Reactor Theory

Expressions for estimating upper bounds to integral parameters of the form (W,K psi)/(W, psi) are derived using functional analysis techniques. The resulting formulations are applied for estimating upper bounds to different types of integral parameters in reactor theory, including direct eigenvalues of the Boltzmann equation, reactivity, and effective reactor kinetics parameters. The upper-bound estimation method is compared with perturbation theory techniques for estimating changes in the total collision eigenvalue.

Investigations of the two-detector covariance method for the measurement of coupled reactor kinetics parameters

For the measurement of point kinetics parameters of zero power reactors many different methods have been widely used and described in the literature. These include Rossi-α, probability distribution, variance-over-mean, frequency analysis, and polarity correlation of neutron detector signals. Up to now kinetics parameters of a coupled two-point reactor model have been measured only by two noise analysis techniques: Rossi-α (correlation functions) and frequency analysis (power spectral density, coherence functions) using two detectors to obtain space-dependent neutron signals. This paper describes for the first time investigations of two-detector covariance measurements and their application to the determination of coupled kinetics parameters. This technique is compared with the Rossi-α method by theoretical considerations and measurements using identical detector signals in both analysis techniques. Results were also compared to those measured previously at the same reactor by frequency analysis. Theoretical formulas for the Rossi-α experiment and variance and covariance measurements in a symmetrical two-point reactor model were derived. The material properties and neutron lifetimes for each reactor zone were assumed equivalent. The transport time of neutrons from one zone to the other was neglected. The parameters of the model include the decay constant of prompt neutrons and the coupling reactivity of the two core zones. These parameters were determined by least-squares fitting the theoretical curves to the experimental data. The measurements have been performed at the Argonaut Reactor Karlsruhe (ARK) with a symmetrical two-slab core loading for different subcritical levels. Variances and covariances of neutron counts from detectors placed in both core regions were measured in two different ways. In the first method the two-dimensional probability distribution of the number of counts from the detectors was measured using a small digital computer. The computer was used to control the gating times of two specially designed counting registers and to calculate the first and second moments. From that the covariances were calculated in real time and displayed on a CRT as a function of the counting time interval. In the second method the variance and the covariance of the detector pulses were directly measured without using the probability distributions. For this purpose a new analyser based on a special calculational algorithm was developed. This analyser can be built easily with low hardware expense. The results obtained by the Rossi-α and variance method agreed quite well. In contrast to point reactor parameter determination, where Rossi-α and variance techniques are equivalent, the variance and covariance method appear superior to the Rossi-α method in applications to coupled reactors. Kinetics parameters of a two-node reactor model can be determined more accurately by the variance technique. The effect of dead-time losses in the signal channels or in the analyser on the measured variances was also studied. The variance technique was found to be more sensitive to dead-time than the Rossi-α technique. Therefore variance measurements at fast reactors yield incorrect results except special fast electronics is used in the pulse channels.

Interpretation of Coherence Function Measurements in Zero-Power Coupled-Core Reactors

New techniques for inferring several reactor kinetic parameters from frequency-domain noise analysis measurements in coupled-core reactors are developed and utilized. These parameters are the eigenvalue separation, delayed-neutron fraction, and, independently, the prompt-neutron lifetime. Reactor characteristics and detector placement criteria necessary for obtaining the parameters are discussed. Comparison with other time-domain noise techniques is undertaken and the merits and shortcomings of each are reviewed.

Reactor Kinetic Parameters of Lattices of Plutonium and Uranium in Water

The ratio βeff/l has been deduced from reactor noise measurements in several uniform light-water lattices in the Plutonium Recycle Critical Facility. These lattices included one in which the fissil...

Determination of Reactor Kinetic Parameters by Photon Observation

"Determination of Reactor Kinetic Parameters by Photon Observation." Nuclear Science and Engineering, 36(1), p. 132This article refers to:The Application of a Gamma-Ray Cascade Model to the Calculation of Neutron Energy-Dependent Capture Gamma-Ray Production Cross Sections

Determination of Reactor Kinetic Parameters by Photon Observation

Fission rate fluctuations at low power in a reactor with a large fission-product inventory have been observed in the pool-type Ford Nuclear Reactor. A gaseous Cerenkov detector was used to sense th...

A Method for Measuring Absolute Reactor Power through Neutron Fluctuation

A new principle is presented for obtaining absolute reactor power by processing the random fluctuation of neutron flux based on the stochastic nature of nuclear reactions. The required combination of instruments to carry out experiments is described, and experimental results obtained in a swimming pool reactor are reported. The power spectral density of the output current of an ion chamber located near the reactor core is determined by reactor kinetic parameters such as delayed neutron yield, life time, ν (mean number of neutrons generated per fission) and counter efficiency as well as by the total number of neutrons in the core, which is a measure of absolute power. Using either logarithmic amplifier or reactivity meter, absolute reactor power can be measured without any information about detector efficiency. This method has such merits as easiness and simplicity in operation, ability to measure absolute power in the range 0.01∼100 W where other methods are inapplicable, and negligible effect of changes in core configuration or in detector position. The results of actual reactor experiments with this method proved to agree fairly well with those of absolute measurement by gold foil activation.

Distribution of the number of counts on a neutron detector placed in a reactor

A discussion is given in the stochastic process of neutron multiplication and detection in a nuclear reactor. The use of physically justifiable assumptions makes it possible to simplify the mathematically complicated problem and find the productive probability function for the number of counts in a definite interval of time. The distribution found has been confirmed experimentally. Some applications of the theory are indicated for determining the reactor kinetic parameters.