External Neutron Source Research Articles

Newton’s second law analogy for the traveling wave of nuclear burning

We consider a model of neutron-nuclear wave burning. The traveling wave of nuclear burning of the medium is initiated by an external neutron source and is the basis for the new generation reactors the so-called “traveling-wave reactors”. We develop a model of nuclear traveling wave burning, for which it is possible to draw a Newton’s second law analogy with a mechanical dissipative system. On the basis of this analogy, we find that the wave velocity has a continuous spectrum bounded below. Within the framework of the new model, we show the autowave to be possible for certain neutron energies only. Also we find that two burning modes are possible depending on the control parameters: a traveling autowave and a wave driven by an external neutron source.

Neutronic analysis of DECY-13 cyclotron target system as a neutron source for SAMOP

The Centre for Accelerator Science and Technology (PSTA) has developed a subcritical reactor for the molybdenum-99 (99Mo) production (SAMOP). This device requires an external neutron source. For experimental purposes, SAMOP is still driven by an external neutron source from a critical reactor, i.e. from the radial beam-port of the Kartini reactor. PSTA is also developing a 13 MeV cyclotron (DECY-13) for the production of fluorine-18 (18F) which can generate neutron as by-products. This cyclotron has an opportunity to be used as an external neutron source for the SAMOP. In this work, particle transport simulations have been carried out to determine the characteristics of the neutron produced by the DECY-13. The simulation results show that DECY-13 produces an average neutron flux of 2.6347 × 109 particles/cm2s and around 97% of them are the fast neutron. Based on the analysis, it concluded that the DECY 13 could be used as an external neutron source for the SAMOP by the addition of a neutron collimator.

Third-Order Adjoint Sensitivity Analysis of an OECD/NEA Reactor Physics Benchmark: I. Mathematical Framework

This work extends to third-order previously published work on developing the adjoint sensitivity and uncertainty analysis of the numerical model of a polyethylene-reflected plutonium (acronym: PERP) OECD/NEA reactor physics benchmark. The PERP benchmark comprises 21,976 imprecisely known (uncertain) model parameters. Previous works have used the adjoint sensitivity analysis methodology to compute exactly and efficiently all of the 21,976 first-order and (21,976)2 second-order sensitivities of the PERP benchmark’s leakage response to all of the benchmark’s uncertain parameters, showing that the largest and most consequential 1st- and 2nd-order response sensitivities are with respect to the total microscopic cross sections. These results have motivated extending the previous adjoint-based derivations to third-order, leading to the derivation, in this work, of the exact mathematical expressions of the (180)3 third-order sensitivities of the PERP leakage response with respect to these total microscopic cross sections. The formulas derived in this work are valid not only for the PERP benchmark but can also be used for computing the 3rd-order sensitivities of the leakage response of any nuclear system involving fissionable material and internal or external neutron sources. Subsequent works will use the adjoint-based mathematical expressions obtained in this work to compute exactly and efficiently the numerical values of these (180)3 third-order sensitivities (which turned out to be very large and consequential) and use them for a third-order uncertainty analysis of the PERP benchmark’s leakage response.

Computational study on distribution feasibility of radioactive waste transmutation in accelerator driven system

In this paper, the feasibility of high-level radioactive waste transmutation in accelerator driven system sub-critical reactor assembly, has been studied for two zone's model and with three different core configurations. The inner zone has a fast neutron spectrum and the outer one has a thermal neutron spectrum. The subcritical core is coupled with external neutron source of energy 14 MeV (D-T source). The effects of high level waste isotopes sample (238Pu, 239Pu, 240Pu, 241Pu, 242Pu, 241Am, 243Am, 244Cm, and 245Cm) distribution on the neutron spectrum and burnup performance in the inner zone have been investigated and discussed, by proposed three core configurations non-uniform, uniform, and spiral. The burnup calculations have been performed for one-year operation cycle for all the all proposed models. This work shows that one can effectively transmute most of the actual minor actinides isotopes in the inner fast spectrum zone of the proposed system, with optimal distribution of these isotopes.

РЕЗУЛЬТАТЫ МОДЕЛИРОВАНИЯ ВЫХОДА РУ БН-800 НА МКУ ПО ЗАВЕРШЕНИИ 5 МИКРОКАМПАНИИ

The paper presents the simulation results of the BN-800 reactor being placed in a critical state upon completion of 5 microcampaign when detectors are placed inside the reactor building in the vicinity of discharge elevators with a demonstration of signal levels in off-threshold (based on 235U) and threshold (based on 238U) detectors. The calculated data are supplemented by experimental signals from non-threshold detectors located in the same region of the reactor. The computational simulation of the reactor exit to the critical state was carried out using three algorithms: the method of point kinetics; by solving a non-uniform stationary equation with an external neutron source; by solving a non-stationary equation with a space-time dependence of the neutron flux density, taking into account the presence of an external neutron source in the reactor. In addition to directly evaluating the calculated behavior of the reactor power in the simulated process, the features of the used solution algorithms are analyzed.

Numerical Investigation of Th-229 Production in an Accelerator Driven System Reactor

The expected increase in radionuclide demand for treatment of various tumor diseases has led to the investigation of alternative production routes to provide sufficient amount of medical isotopes [1]. In Targeted Alpha Therapy (TAT), there is a present and future need for the Ac-225 and Bi-213, as the most promising alpha emitting isotopes, one of the limited ways to provide these radioisotopes is by production of Th-229 in reactor route [2][3]. The present study conducted to investigate the production of Th-229 to expand the availability of Ac-225 and Bi-213 in innovative nuclear reactor concept like Accelerator Driven System reactor (ADS)[4], which is a subcritical reactor and currently in development that predict to play an important role in the transmutation process of heavy elements and isotopes production. The possibility of producing Th-229 from neutron transmutation of Ra-226 are numerically investigated for simple model of ADS reactor consist of two zones, inner region with fast neutron spectra and outer region with thermal neutron spectra, and the subcritical core coupled with external neutron source. The calculations of transmutation behavior and mass ratio for the produced isotopes are conducted by using Monte Carlo N-particle Transport (MCNPX) code. This research was supported by International Center for Theoretical Physics and The Institute of International Education Scholar Rescue Fund. My deep thanks to colleagues from Institute for Safety Problems of Nuclear Power Plants, Ukraine who provided expertise that assisted the research. [1]E. R. Birnbaum, L. J. Bitteker, A. Couture, and M. E. Fassbender, “Ac and Ra-223 Production via 800 MeV Proton Irradiation of Natural Thorium Targets,” vol. 836.[2]J. R. Griswold et al., “Large scale accelerator production of Ac-225: Effective cross sections for 78–192 MeV protons incident on Th-232 targets,” Appl. Radiat. Isot., 2016.[3]S. Hogle et al., “Reactor production of Thorium-229,” Appl. Radiat. Isot., vol. 114, pp. 19–27, 2016.[4]H. Aït Abderrahim, P. Baeten, D. De Bruyn, and R. Fernandez, “MYRRHA - A multi-purpose fast spectrum research reactor,” in Energy Conversion and Management, 2012, vol. 63, pp. 4–10.

Analysis of Uranyl Nitrate Hexahydrate composition for optimum neutron multiplication factor of SAMOP

Subcritical Assembly for Molybdenum Production (SAMOP) is a concept of nuclear reactor where the neutron multiplication factor (keff) less than one (keff ∼ 0.99). Therefore, SAMOP reactor need an external neutron source for its operation. Uranyl Nitrate Hexahydrate, UO2(NO3)2 or UN is used as fuel as well as target material for 99Mo production. The 99Mo isotope is used to generate 99mTc which is the most widely used radioisotopes for diagnostic in nuclear medical fields. The composition of Uranyl Nitrate Hexahydrate (UNH) should be analyzed to achieve the required subcritical level in the SAMOP core i.e. k-eff ∼ 0.99. Methodology for this analysis is an elemental composition calculation of UNH, with enrichment of uranium of 19.75% and 2.81 gr/cc of density. The analysis result shows that the elemental composition of 235U, 238U, N, O, and H, are 4.9715%, 20.2009%, 5.3448%, 37.4138%, and 32.069% respectively, and therefore, the composition of U and N-O-H are 25.1724%, and 74.8276% respectively. The composition is then used as input for the neutronic analysis using the WIMS-D/4 computer code for keff calculation of SAMOP. The result shows that the optimum keff of 0.99 is achieved with the above elemental compositions.

Pulse shape measurement and experimental validation of noise theory developed under consideration of finite pulse width in D-T neutron generator driven subcritical system

Degweker and Rana (2007) had postulated a noise theory assuming the external neutron source as exponentially correlated Gaussian character for finite pulse width comparable to prompt neutron decay constant of subcritical system. They considered various conditions of source correlation time (i.e. the decay time of correlated neutrons), accelerator pulse width and pulse period. This noise theory has been validated experimentally for the first time in D-T neutron generator coupled BRAHMMA subcritical system. The present work has been carried out under the consideration that the source correlation time is small compared to the pulse width and rectangular pulse shape. The pulse shape of the neutron generator was characterised through an AC Current Transformer (ACCT) placed in the D+ beam path of the neutron generator. Variations of source correlation time with pulse widths were studied; it has been observed that source correlation time decreases with smaller pulse width. In validation experiments, data have been analysed using Auto Covariance, Variance to Mean ratio, and Auto Power Spectral Density method. The measured prompt neutron decay constant was found to be in good agreement with the theoretical estimate of the system. Furthermore, the present research also discusses the modulation of the correlated and uncorrelated part in variance to mean ratio method for a wide range of accelerator frequency at fixed source strength. The present research findings also lay the criteria for choosing the proper accelerator frequency for reactivity measurement in ADS.

Instrumentation & control system for “SAMOP” reactor experimental facility

The instrumentation and control system (ICS) for a subcritical assembly for molly-99 production (SAMOP) reactor experimental facility is presented in this paper. The system is designed based on the defense-in-depth concept. SAMOP reactor is fuelled by uranyl nitrate and driven by an external neutron source from Kartini reactor beam-port. The ICS was designed using 2 neutron detectors of fission chamber and compensated ionization chamber types and provided with a safety rod made from boral used for shutting-down the reactor in emergency condition. The method used for reactor protection system is based on hardwire with a fail-safe principle. The safety rod drive mechanism is designed by using rack and pinion, and provided by the safety button of key release and electromagnetic holder status.

Physical and thermal coupling calculation for accelerator driven subcritical core

The lead Anode (LBE) cooling ADS core is taken as the research object, and the ADS core axial one-dimensional physical thermal coupling model is established. The deterministic method is used to calculate the subcritical flux neutron flux distribution based on Monte Carlo. The method calculates the accelerator beam and the external neutron source generated by the spallation target. The ADS core thermal hydraulic analysis is carried out by a single-channel model, and the fuel elements are respectively divided into nodes in the axial direction and the radial direction, and the influence of thermal feedback on the neutronics characteristics of the subcritical core is considered. The one-dimensional physical thermal coupling model established in this paper can accurately calculate the spatial distribution of key parameters such as neutron flux density, fuel and cladding temperature in ADS core.

Experimental analyses of βeff/Λ in accelerator-driven system at Kyoto University Critical Assembly

ABSTRACTThe capability of βeff /Λ obtained by λ-mode and ω-mode calculations is examined for target subcriticality in the accelerator-driven system through a comparison with that obtained in the pulsed-neutron source (PNS) experiments at the Kyoto University Critical Assembly. Directly measured results of βeff /Λ, α and ρ$ in the PNS experiments are validated by varying the experimental conditions: the external neutron source, detector position, detector type and subcriticality ranging between 500 and 7500 pcm. The numerical analyses of βeff /Λ are conducted by using MCNP6.1 together with ENDF/B-VII.1 for λ-mode calculations and PARTISN (SCALE6.2.2 with ENDF-VII.1 for the effective cross sections) for both the λ-mode and the ω-mode calculations. The comparison between calculated and measured βeff /Λ with varying subcriticality shows good agreement between the experiments and the ω-mode calculations, although a difference is observed between the experiments and the λ-mode calculations.

Paralyzable and non-paralyzable dead-time corrections for the neutron detectors of the KUCA facility driven by external neutron sources

ABSTRACT This study examines dead-time corrections for the neutron detectors of the Kyoto University Critical Assembly (KUCA) driven by a stationary or pulsed external neutron sources. For the stationary (californium) external neutron source, the MCNP simulations producing the detector timestamps and the MATLAB scripts correcting for the dead-time effect have been validated with analytical formulae from literature. For the pulsed external neutron source (produced by a proton accelerator), the experimental detector timestamps have been corrected for the dead-time effect using the aforementioned MATLAB scripts utilizing the backward extrapolation method and the paralyzable or non-paralyzable dead-time model. Then, the keff obtained from the dead-time corrected timestamps has been corrected for the detector position and type. Finally, the corrected experimental keff has been compared with the one obtained from MCNP criticality simulations.

Calculation of the prompt neutron decay constant for the KUCA facility driven by a stationary or pulsed external neutron source

ABSTRACT In this study, the I-4 configuration of the KUCA facility, as one of the subcritical experimental benchmarks, was examined using computer simulations and experiments. This configuration has all control rods withdrawn. Two different external neutron sources have been analyzed: a pulsed neutron source generated from protons interactions with target materials and a californium neutron source. The two neutron sources have similar spatial effects because the californium neutron source has been located at the position of the proton beam target. Different methods have been applied to calculate the prompt neutron decay constant (alpha) including: the count rate as a function of time fitting, the Rossi, the Feynman, the Cross-Power Spectral Density (CPSD), the Auto-Power Spectral Density (APSD), and the Maρta methods. The input data for these methods consist of the detector timestamps, which can be obtained experimentally using neutron detectors operating in pulse mode or by MCNP fixed-source simulations. The CPSD, the APSD, the count rate fitting, and Maρta methods do not require the detector timestamps since these methods use the average number of neutron counts over some time bins. These averages can be obtained experimentally using detectors operating in the current mode or by MCNP fixed-source simulations.

Experimental validation of noise theory developed by considering the source as an exponentially correlated process in D-T neutron generator driven subcritical system

Degweker and Rana (2007) have developed a noise theory considering the external neutron source as exponentially correlated Gaussian character for Accelerator Driven Subcritical system. First time experimental validation of this noise theory has been carried out in BRAHMMA subcritical system using an external neutron source. To understand the source statistics, the Purnima Neutron Generator has been used in D-T mode. A series of experiments have been carried out to determine the source correlation factor. It was observed that time correlation between source emissions, or in other term source correlation factor, decay much faster than the decay of prompt neutron chain (i.e. prompt neutron decay constant) in subcritical system. In validation experiments of the noise theory, the time stamped data have been analysed using variance to mean ratio, Auto Covariance and Auto Power Spectral Density method. The measured prompt neutron decay constant using the postulate are in good agreement with the theoretical value. The work also offers qualitative discussion on space dependent noise theory addressed by several authors.

Radiation shielding design of neutron source from Kartini reactor’s beam-port for SAMOP test facility

SAMOP is a subcritical assembly for 99Mo production fueled with uranyl nitrate solution which is designed to be operated by using external neutron source from a neutron generator. The most common radioisotope used in diagnosis for medical imaging in nuclear medicine is 99mTc which is daughter isotope of 99Mo. The SAMOP experimental or test facility is a facility to be used for performance testing of the SAMOP system by using neutron source from radial beam port of Kartini reactor instead of using neutron generator. This paper presents the results of the radiation shielding design and calculation for external neutron source of SAMOP experimental facility. The method used is a modeling and calculation by using MCNPX computer code. Based on the assumed source terms, the materials being used, and the geometrical arrangements, it is concluded that by using paraffin of 60 cm thickness for the beam catcher and 50 cm for the concrete of the outer shield would be sufficient to reduce the radiation dose below the maximum recommended limit. The presence of beam catcher can significantly reduce the contribution of neutrons and secondary particles to the radiation dose.

Commissioning Preparation of a Subcritical Experimental Facility For 99Mo Production

A sub-critical reactor experimental facility for 99Mo isotope production (SAMOP) has been designed and constructed at the Centre for Accelerator Science and Technology (CAST), Nuclear Energy Agency (BATAN). The facility is fueled by uranyl nitrate and will be driven by an external neutron source from the radial beam-port of Kartini reactor. The aim of the paper is to report the status, progress and readiness for commissioning of the subcritical reactor experimental facility. The safety analysis results and integration of all components and systems are also discussed. The commissioning method is referring to the IAEA document on Safety of Research Reactors, Specific Safety Requirements No. SSR-3, 2016. All main components such as reactor core and reflector assembly, fuel tubes, and control rod have been constructed and tested. In overall the SAMOP experimental facility is ready for fuel loading/commissioning and operation performance test.

Regulatory Assessment on a New utilization of SAMOP Test Facility: Determination on Fission Power

SAMOP test facility is a subcritical solution of uranyl nitrate to produce Mo99 that is connected with external neutron source from one of Kartini reactor beam ports. Determining the SAMOP fission power is a critical step toward thermal hydraulic calculation ensuring the adequacy of cooling and to calculate the fission product inventory without modelling the Kartini reactor geometry. As part of regulatory assessment, the analyses were intended to determine the power level of SAMOP and its distribution in its solution tubes using MCNP5 code. Two cases of Kartini reactor power 100 % and 110 % of nominal power were simulated. The results showed that SAMOP can produce fission power ranging from 563.2 up to 860.8 watt. Taking uncertainties into account, it is recommended to conduct thermal hydraulic calculations on SAMOP test facility based on minimum total power of 1000 W.

РАСЧЕТ ПАРАМЕТРОВ СТОХАСТИЧЕСКОЙ КИНЕТИКИ НЕЙТРОНОВ В ЯДЕРНОМ РЕАКТОРЕ НУЛЕВОЙ МОЩНОСТИ

The majority of neutron-physical problems of nuclear power plants design can be solved on the basis of various approximations to the Boltzmann transport equation in terms of the averaged characteristics of the reactor: the effective multiplication factor, neutron flux, average neutron lifetime, etc. However, the neutron chain reaction itself is always stochastic. There are situations in which the stochastic nature of the chain reaction cannot be ignored. This is the so-called “blind” start-up problem with a weak external neutron source, the work of physical assemblies of “zero” power, the analysis of the reactivity noise of such assemblies, etc. Despite the well-developed theoretical basis for the stochastic description of the behavior of neutrons in a nuclear reactor, there are still not enough calculation algorithms and programs for stochastic kinetics analysis. The paper presents two computational algorithms for point reactor model, which are developed on the basis of the theory of Markov branching random processes. The first one is based on the balance equation for the probabilities of death and birth of prompt and delayed neutrons in a reactor, the second one is based on calculating the first and second moments of neutron number distributions and using the assumption that these distributions can be approximated by gamma distributions with sufficient accuracy. On the basis of these algorithms, programs have been created that allow one to calculate various scenarios for introducing reactivity and an external source into the system. The calculation of pulsed experiments on prompt and delayed neutrons to determine the waiting time of the threshold neutron pulse at the Godiva II installation showed a qualitative agreement of the results for both the two computational algorithms and the experimental data.

An analytical approach to bifurcations and stability in simplified mathematical models of nuclear reactors

Asymptotic analytical methods are applied to study some problems related with bifurcations (both local and global) and stability in three simple mathematical models of nuclear reactors.The first case is a reactor which is subcritical at rest and driven by a distributed external neutron source. Under suitable conditions in the temperature feedback reactivity, the mathematical model predicts a static global bifurcation with three critical states, two stable and one unstable, with the possibility of a runaway. The dynamics of this system is studied, in a framework of slow manifold theory, by methods of restricted nonlinear modal analysis, and the results of a digital simulation are summarized. The model could be modified and extended to study the space time dynamics of sub-critical multiplying systems driven by external neutron sources (neutron beams produced by accelerators).The second case is related with in phase and out of phase xenon oscillations in large thermal reactors. The known subcritical Hopf bifurcation that appears in the context of global mode oscillations is revisited. After applying a nonlinear modal analysis to the mathematical model, a normal form is derived by an averaging method. Approximate analytical formulae for the radii of the unstable limit cycles and for the trajectories of the state variables in a neighborhood of the bifurcation point are obtained.The third case is a non-trivial modification and development of a simple mathematical model intended to describe certain mechanical kinetic effects stemming from a possible coupling of nuclear, thermal and mechanical vibration processes. We show that, under suitable conditions, a dynamic supercritical Hopf bifurcation in the reactor power appears in the framework of our modified model. A normal form is derived by an averaging method. Analytical formulae for the radii of the stable limit cycles and the trajectories of the state variables in a neighborhood of the bifurcation are given.

Numerical Investigation of Th-229 Production in an Accelerator Driven System Reactor

The expected increase in radionuclide demand for treatment of various tumor diseases has led to the investigation of alternative production routes to provide sufficient amount of medical isotopes [1]. In Targeted Alpha Therapy (TAT), there is a present and future need for the Ac-225 and Bi-213, as the most promising alpha emitting isotopes, one of the limited ways to provide these radioisotopes is by production of Th-229 in reactor route [2][3]. The present study conducted to investigate the production of Th-229 to expand the availability of Ac-225 and Bi-213 in innovative nuclear reactor concept like Accelerator Driven System reactor (ADS)[4], which is a subcritical reactor and currently in development that predict to play an important role in the transmutation process of heavy elements and isotopes production. The possibility of producing Th-229 from neutron transmutation of Ra-226 are numerically investigated for simple model of ADS reactor consist of two zones, inner region with fast neutron spectra and outer region with thermal neutron spectra, and the subcritical core coupled with external neutron source. The calculations of transmutation behavior and mass ratio for the produced isotopes are conducted by using Monte Carlo N-particle Transport (MCNPX) code. This research was supported by International Center for Theoretical Physics and The Institute of International Education Scholar Rescue Fund. My deep thanks to colleagues from Institute for Safety Problems of Nuclear Power Plants, Ukraine who provided expertise that assisted the research. [1]E. R. Birnbaum, L. J. Bitteker, A. Couture, and M. E. Fassbender, “Ac and Ra-223 Production via 800 MeV Proton Irradiation of Natural Thorium Targets,” vol. 836.[2]J. R. Griswold et al., “Large scale accelerator production of Ac-225: Effective cross sections for 78–192 MeV protons incident on Th-232 targets,” Appl. Radiat. Isot., 2016.[3]S. Hogle et al., “Reactor production of Thorium-229,” Appl. Radiat. Isot., vol. 114, pp. 19–27, 2016.[4]H. Aït Abderrahim, P. Baeten, D. De Bruyn, and R. Fernandez, “MYRRHA - A multi-purpose fast spectrum research reactor,” in Energy Conversion and Management, 2012, vol. 63, pp. 4–10.