External Neutron Research Articles

A direct calculation method for subcritical multiplication factor in Reactor Monte Carlo code RMC

The subcritical multiplication factor ks with the existence of external source neutrons is important for the evaluation of the neutron multiplication M, especially in the accelerator-driven system (ADS) performance assessment. The effective multiplication factor keff calculated from the traditional source iteration method in Monte Carlo codes can’t fully describe the subcritical system, acquiring the spurious neutron flux distribution. A direct Monte Carlo method called modified source iteration method by external source is introduced to calculate the subcritical multiplication factor ks directly, acquiring the real neutron flux at the same time. Compared with Monte Carlo fixed source calculation, the modified source iteration method has a great advantage in calculating the subcritical multiplication factor ks and its statistical error directly, especially for systems with complex geometry and a variety of fissile materials where much post-processing work on the tally results from fixed source calculation is needed. Moreover, much calculation time can be saved through this method when obtaining the subcritical multiplication factor ks with similar precision. This method has already been implemented in Reactor Monte Carlo code RMC including parallel calculation capability and can be effectively applied to the analysis of accelerator driven subcritical system (ADS).

Energy Dependence of Total Cross Sections for Reactions with 4, 6he, 6, 7, 9li Nuclei

The paper presents the results of measurement of the total cross sections for reactions 4,6He + Si and 6,7,9Li + Si in the beam energy range 5−50 A⋅MeV. The enhancements of the total cross sections for reaction 6He + Si compared with reaction 4He + Si, and 9Li + Si compared with reactions 6,7Li + Si have been observed. The performed microscopic analysis of total cross sections for reactions 6He + Si and 9Li + Si based on numerical solution of the time-dependent Schrödinger equation for external neutrons of projectile nuclei 6He and 9Li yielded good agreement with experimental data.

Dose coefficients for children and young adolescents exposed to external neutron fields

The risks associated with exposure to external fields of ionising radiation are important to quantify in order to provide guidance towards public and worker protection. In Publication 116 of 2010, the International Commission on Radiological Protection (ICRP) published adult male and female fluence-to-dose coefficients (henceforth referred to as dose coefficients) for external exposures to six types of idealised neutron fields. However, ICRP 116 dose coefficients are not appropriate for applications involving children due to their smaller body weight and stature. Our current work details dose coefficient calculations for children and young adolescents using the UF-NCI pediatric hybrid phantoms at all neutron energies considered in ICRP 116 (0.001 eV to 10 GeV); those dose coefficients with energy up to 150 MeV are discussed. The hybrid UF-NCI phantoms are divided into five separate age groups: newborn, 1, 5, 10, and 15 years. For these phantoms, we calculated dose coefficients for the six idealised neutron fields for 28 organs, two bone tissues, and the overall whole body effective dose. All calculations were performed using the MCNP6 radiation transport code. To validate our methodology, we first calculated dose coefficients for the ICRP adult male and female phantoms and confirmed our ability to reproduce the values published within ICRP 116. The same methodology was then applied to calculate dose coefficients for the UF-NCI pediatric phantoms. Energy-dependent trends were observed in the neutron dose coefficients for pediatric phantoms: below 100 keV, dose increases with phantom age (a proxy for body size); however, above 100 keV, the opposite trend was observed. Comparisons between field geometries showed varying trends depending on the location of the organ within the body. Explanations for these trends are also explored within. Our results are the first comprehensive set of neutron dose coefficients derived for children and young adolescents using the newest generation of hybrid phantoms. The primary application of the pediatric neutron dose coefficients presented in this work will be for a planned effort to update the dosimetry for the Japanese atomic bomb survivors.

Transient safety analysis for simplified accelerator driven system with gas-lift pump

Owing to remarkable advantage for transmutation of the long-lived nuclear waste, Accelerator Driven System (ADS) is one of the most promising systems in nuclear power field. In this study, the transient analysis of a simplified lead bismuth alloy cooling accelerator driven system with gas-lift pump has been numerically simulated. By using point reactor neutron kinetics equation and related thermal-hydraulic models, TASAR (Transient Analysis of Simplified ADS Reactor) code is developed. The five kinds of transient cases are simulated: Starting gas-lift pump (SG), Unprotected loss of flow (ULOF), Beam interruption (BI), Unprotected transient over current (UTOC), Unprotected transient overpower (UTOP). Meanwhile, the calculation results of TASAR code are validated by the experimental data of NACIE (Natural Circulation Experiment) facility. The analysis is as following: In SG and ULOF conditions, the introduction of gas-lift pump has a remarkable influence of loop natural circulation capability. In BT and UTOC conditions, the short-time beam cutting off and a 100% increase of the external neutron source would not lead to the immediate fuel melting. In UTOP condition, a reactivity jump of 2$ would not lead to the damage of fuel and the cladding. The above numerical results show that, the accelerator and gas-lift pump influence the loop mass flow rate and core temperatures of research system in this paper greatly, and the simplified accelerator driven system with gas-lift pump has a very good safety performance.

Preliminary analysis of reactivity reconstruction capability based on inverse kinetics method under different initial reactivity states

Online reactivity monitoring plays an important role in operation and safety analyses of fission reactor systems. The inverse kinetics method, which is based on a point kinetics model, is the most widely used method for reactivity reconstruction of critical water reactors. However, this method is seldom applied to the reactivity reconstruction of subcritical reactors. In this study, an inverse kinetics method was employed for the reactivity reconstruction of a lead-based reactor under different initial reactivity states (ρ0 = 0, − 2786, − 5486, − 8367, and − 12,371 pcm). The results showed that the deviation in the reactivity of the lead-based subcritical reactor was greater when ρ0 became smaller. The reactivity reconstructed using the inverse kinetics method was globally underestimated. At a given reactivity perturbation, the relative and absolute errors increased with the decrease in the initial reactivity. At a given initial reactivity, with the increase in the reactivity perturbation, the absolute error increased, whereas the relative error remained the same. This deviation is due to the variation in the external neutron source, spatial-spectral effects, and sub-diffusive effects, which require further study.

Fast neutron detection at near-core location of a research reactor with a SiC detector

The measurable charged-particle produced from the fast neutron interactions with the Si and C nucleuses can make a wide bandgap silicon carbide (SiC) sensor intrinsically sensitive to neutrons. The 4H-SiC Schottky detectors have been fabricated and tested at up to 500 °C, presenting only a slightly degraded energy resolution. The response spectrum of the SiC detectors were also obtained by exposing the detectors to external neutron beam irradiation and at a near-core location where gamma-ray field is intense. The fast neutron flux of these two locations are ∼4.8×104cm−2⋅s−1 and ∼2.2×107cm−2⋅s−1, respectively. At the external beam location, a Si detector was irradiated side-by-side with SiC detector to disjoin the neutron response from Si atoms. The contribution of gamma ray, neutron scattering, and charged-particles producing reactions in the SiC was discussed. The fast neutron detection efficiencies were determined to be 6.43×10−4 for the external fast neutron beam irradiation and 6.13×10−6 for the near-core fast neutron irradiation.

Neutronic Study of an Innovative Thorium-Uranium–Based Fusion-Fission Hybrid Energy Reactor with 233U Breeding Enhancement by Using Dual-Coolant System

In this technical note, an innovative thorium-uranium–fueled fusion-fission hybrid reactor (FFHR) design that employs a dual-coolant system to enhance 233U breeding and is based on a three-dimensional engineering model is presented. The reactor consists of two kinds of modules: a water-cooled, thermal spectrum power generation natural uranium–fueled module and helium-cooled, fast spectrum fissile-breeding natural thorium–fueled modules, which are arranged alternately in the poloidal direction of the blanket. An interesting and important neutronic characteristic of the FFHR is found in this technical note: Energy multiplication is primarily determined by the uranium module parameters and is almost independent of the thorium module parameter. Uranium module design should first consider improving energy production. The 232Th neutron capture rate is primarily determined by the thorium module parameters. The uranium module parameter has almost no influence on the 232Th neutron capture rate in the thorium module. The uranium and thorium modules have weak coupling in neutronic behavior. However, with the fixed design parameters of the uranium and thorium modules, the most important influencing factor on energy multiplication factor M (the ratio of total blanket energy output and the fusion energy) and the 233U breeding rate is the fraction of the external fusion neutron source irradiated on the uranium or thorium module or the blanket coverage rate of the uranium or thorium modules. Based on this characteristic, an innovative hybrid reactor design that employs a dual-coolant system is proposed in this technical note. Uranium modules still use water as the coolant to maintain a high energy multiplication factor, whereas helium is used as the coolant for the thorium module to obtain a fast neutron spectrum to enhance the 233U breeding. The simulation results show that the helium-cooled thorium module is 2.5 times more efficient in 233U breeding compared to the original water-cooled thorium module design. Approximately 10 tons of 233U is produced after 20 years of operation for the helium-cooled thorium module design.

Time-dependent microscopic description of total reaction cross sections for collisions of 6He and 9Li nuclei with 28Si nuclei at low and intermediate energies

A microscopic description is given for the collisions of the 9Li and 6He nuclei with the 28Si nucleus. It is based on numerical solution of the time-dependent Schrodinger equation for the external neutrons of the projectile nuclei. The corrections to the nucleus–nucleus potential associated with the redistribution of external neutrons were determined for energies of 5 to 60 MeV/nucleon using the time-dependent folding potential. The probability of leaving the elastic channel was calculated using the reflectionless boundary condition for partial waves. The calculated cross sections are in agreement with the experimental data on the total cross sections for the above reactions.

Application of the backward extrapolation method to pulsed neutron sources

Particle detectors operated in pulse mode are subjected to the dead-time effect. When the average of the detector counts is constant over time, correcting for the dead-time effect is simple and can be accomplished by analytical formulas. However, when the average of the detector counts changes over time it is more difficult to take into account the dead-time effect. When a subcritical nuclear assembly is driven by a pulsed neutron source, simple analytical formulas cannot be applied to the measured detector counts to correct for the dead-time effect because of the sharp change of the detector counts over time. This work addresses this issue by using the backward extrapolation method. The latter can be applied not only to a continuous (e.g. californium) external neutron source but also to a pulsed external neutron source (e.g. by a particle accelerator) driving a subcritical nuclear assembly. The backward extrapolation method allows to obtain from the measured detector counts both the dead-time value and the real detector counts.

Mass differences and neutron pairing in Ca, Sn and Pb isotopes

Various estimates of the even-odd effect of the mass shell of atomic nuclei are considered. Based on the experimental mass values of the Ca, Sn, and Pb isotopes, the dependence of the energy gap on the neutron number is traced and the relationship of this characteristic to the properties of external neutron subshells is shown. In nuclei with closed proton shells, effects directly related to neutron pairing and effects of nucleon shells are discussed.

Study of the neutron multiplication effect in an active neutron method

ABSTRACTThe neutron multiplication effect appears when an item contains large amounts of nuclear material. The neutron multiplication effect in this paper means the effect of subsequent fission reactions which are caused by fission neutrons produced by interrogation neutrons from a neutron generator. The previous active neutron method could not distinguish between first-fission and subsequent-fission neutrons and might overestimate the amount of nuclear material. However, the neutron multiplication effect in the active neutron method has not been adequately investigated. We discuss the evaluation method of the multiplication effect in the fast neutron direct interrogation method, one of the active neutron methods, using simulations with the Monte Carlo code MVP and experiments involving uranium waste drums. The first-generation neutrons from an external neutron source generate fission neutrons called second-generation neutrons, the second-generation neutrons generate third-generation neutrons, and so on. This study supposes that the neutron multiplication effect is mainly caused by the third-generation neutrons under the condition that the fourth-generation neutrons are much fewer. This paper proposes a correction method for the neutron multiplication effect in the measured data.

A closed nuclear energy system by accelerator-driven ceramic reactor and extend AIROX reprocessing

For the future energy system, we propose a new closed nuclear energy cycle system, which consists of an accelerator-driven external neutron source, a ceramic reactor and an extend AIROX reprocessing. The attractive features of this system are as follows. (1) The operating mode of the reactor is a combination of subcritical mode and critical mode. Initially, the reactor would be driven by the accelerator external neutron source in subcritical mode. A few years later, the reactor would reach the critical mode, and then would operate for a long time. (2) Nuclear fuels, coolants, and structure materials in the ceramic reactor core are made up of ceramic with excellent thermodynamics properties and neutron performance. Therefore, the ceramic reactor has extremely inherent safety, good breeding performance and high power generation efficiency. (3) Fuel reprocessing uses an extend AIROX reprocessing, which is a simple high-temperature dry process and rarely involved in chemical process. In this reprocessing, only most of fission products are separated. Other isotopes, including uranium isotopes, transuranic nuclides and long-lived fission products, would re-enter the reactor as new fuels. Therefore, this closed nuclear energy system could be known as ADANES, short for Accelerator-Driven Advanced Nuclear Energy System, which can greatly improve the utilization rate of nuclear fuels, enhance the nuclear safety, reduce the nuclear proliferation and become a sustainable and low-carbon energy supply for thousands of years.

Numerical study on cooling characteristics in granular and liquid spallation targets

Accelerator Driven System (ADS) has been developed to provide external neutron source for transmutation and multiplication to handle issues of short nuclear fuel supply and increased nuclear waste. With heavy nuclides bombarded by high-energy proton beam, neutron yields in spallation material which is accompanied by formidable energy deposition. Due to little survival of solid target from high power density and many unresolved issues in liquid target, granular target is motivated with higher sustainable power density. Although many investigations into cooling capability in spallation targets have been conducted, little literature is available on granular target which merits further study.In the present study, investigation is focused on cooling characteristics of granular target compared with liquid target based on the same target model. Due to combination of granules and fluid, coupled CFD-DEM method is employed to simulate movement and heat exchange for granular target, but only CFD technique for liquid target. Directed against same studied domain, comparison and analysis are carried out for two types of targets in the present study. Our research indicates that the granular target owns higher safety margin to withstand severe conditions especially under high energy deposition. The present findings may help to clarify and confirm the safe and reliable spallation target for some specific applications.

Dose conversion coefficients for Chinese reference adult male and female voxel phantoms from idealized neutron exposures

ABSTRACTA whole-body voxel phantom called CRAF representing the Chinese adult female was constructed in this study from color photographs. The CRAF has a height of 160 cm and a weight of 54 kg with 80 organs and tissues, including almost all organs required in the ICRP Publication 103. Masses of all organs in CRAF are consistent with the Chinese reference data within 1%. Our research group has previously developed a Chinese reference adult male phantom called CRAM, which has a height of 170 cm and a weight of 60 kg with 80 tissues and organs. A new set of fluence-to-dose conversion coefficients based on the Chinese reference adult voxel phantoms CRAM and CRAF are presented for six idealized external neutron exposures from 10−8 MeV to 20 MeV. The calculation of organ-absorbed doses and effective doses were performed with the Monte Carlo transport code MCNPX 2.4. The resulting dose conversion coefficients were compared with those published in ICRP Publication 116, which represents the reference Caucasian. It was observed that the two sets of organ-absorbed dose conversion coefficients agreed well for most organs in ROT and ISO geometry; however, considerable deviations were found in some organs in AP, PA, RLAT and LLAT exposures. Organ-absorbed dose coefficients difference showed good agreement with the difference of organ depth distribution between Chinese and ICRP phantoms. For neutrons with energies above 2 MeV, the effective dose conversion coefficients of Chinese reference adult are almost identical to those of ICRP Publication 116 in AP, PA, ROT and ISO geometries, while are generally 15% higher than those of ICRP Publication 116 from lateral geometries. When energies range from 10−8 MeV to 1 MeV, differences are within 10% in AP (–5%), PA (–8%) and ROT (4%) geometries; however, relatively large discrepancies are shown in lateral and ISO geometries, with differences of 15% for LLAT, 20% for RLAT and 12% for ISO, respectively.

Detection of Combined n/γ Fission Signatures Induced by an Epithermal Neutron Source

In this paper, a method is presented for the detection of special nuclear materials (SNMs) in shielded containers, which is both sensitive and applicable under field conditions. The method uses an external pulsed neutron source to induce fission in SNM and subsequent detection of the fast prompt fission neutrons. The detectors surrounding the container under investigation are liquid scintillation detectors able to distinguish gamma rays from fast neutrons by means of pulse shape discrimination method (PSD). One advantage of these detectors, besides the ability for PSD analysis, is that the analog signal from a detection event is of very short duration (typically few tens of nanoseconds). This allows the use of very short coincidence gates for the detection of the prompt fission neutrons in multiple detectors, while benefiting from a low background coincidence rate, yielding a low detection limit. Another principle advantage of this method derives from the fact that the external neutron source is pulsed. By proper time gating, the interrogation can be conducted by epithermal source neutrons only. These neutrons do not appear in the fast neutron signal following the PSD analysis, thus providing a fundamental method for separating the interrogating source neutrons from the sample response in the form of fast fission neutrons. This paper describes laboratory tests with a configuration of eight detectors in the Pulsed Neutron Interrogation Test Assembly (PUNITA). Both the photon and neutron signature for induced fission is observed, and the methods used to isolate these signatures are described and demonstrated.

Support to the identification of anomalies in an external neutron source using Hurst Exponents

A new methodology is proposed here to identify anomalies in the neutron flux as a result neutron production trips in an Accelerator Driven System - ADS type reactor. This methodology is based on the calculation of Hurst exponents, where the neutron flux as monitored in the reactor core is treated as a temporal series. In several recent articles, related to fractional diffusion, the Hurst exponent is used as an estimate for the order of the fractional derivative. Our object of study considered a reactor based on the Myrrha simulated with the Serpent Monte Carlo code for two kinds of trips occurred in the production of neutrons, as follows: a Peak of Production (PP), the Unprotected* Accelerator Beam Overpower (UABO), and the spurious Beam Trip (BT). In order to estimate the Hurst exponent it was used two different methods, namely the Rescaled Range Analysis (R/S) and the Detrended Fluctuation Analysis Method (DFA). The results obtained showed that the R/S methodology had some advantages in a comparison with the DFA in indicating the occurrence of those anomalies. Also being able to provide a scale for the assessment of the intensity of the trip occurred, showing to be an useful tool to support anomaly identification in the neutron flux of ADS reactors.

Influence of reflector materials and core coolant on the characteristics of accelerator driven systems

In this paper we simulated the behavior of a simple ADS model, based on MOX fuel embedded in solid lead, in terms of multiplication coefficient keff, thermal power and absolute neutron spectra. Our focus on a subcritical, accelerator-driven system is motivated by the need to guarantee enhanced safety based on the distance from criticality and the dependance of the chain reaction on the external neutron source provided by the accelerator. The studies presented here are meant to expand the scientific reach of a previously proposed low-power ADS as the core apparatus of a research and training center in nuclear technologies, such as to increase the flexibility of the system in terms of accessible neutron spectra. In the first part of the paper, we report on the results obtained when modifying the reflector surrounding the fission core, by replacing pure lead with a layered graphite/lead structure. We found that, by appropriately choosing position and thickness of the graphite and lead layers, it is possible to obtain a “hybrid” system where the neutron spectrum in the core still exhibits a fast character, while the spectrum in the graphite layer is considerably softer, becoming thermal in the most peripheral positions. In order to obtain such a modulation of the neutron spectra from the center of the system to the periphery, a careful choice of the materials has to be made in order to avoid large variations of the local power at the core boundary. However, the smoothness of the power distribution is obtained at the expense of lower values of keff and the total power of the system. In the second part of the paper, we explored the option of adopting light water as coolant, instead of the helium gas assumed in the initial design. We found that this produces an increase in keff and thermal power, without significantly perturbing the fast character of the system and without introducing spatial power excursions in any place within the core. The characteristics obtained may allow to design a system where fast, mixed and thermal spectra can be used to expand the scope of the ADS as an irradiation facility.

Inverse kinetics for subcritical systems with external neutron source

Nuclear reactor reactivity is one of the most important properties since it is directly related to the reactor control during the power operation. This reactivity is influenced by the neutron behavior in the reactor core. The time-dependent neutrons behavior in response to any change in material composition is important for the reactor operation safety. Transient changes may occur during the reactor startup or shutdown and due to accidental disturbances of the reactor operation. Therefore, it is very important to predict the time-dependent neutron behavior population induced by changes in neutron multiplication.Reactivity determination in subcritical systems driven by an external neutron source can be obtained through the solution of the inverse kinetics equation for subcritical nuclear reactors. The main purpose of this paper is to find the solution of the inverse kinetics equation the main purpose of this paper is to device the inverse kinetics equations for subcritical systems based in a previous paper published by the authors (Gonçalves et al., 2015) and by (Gandini and Salvatores, 2002; Dulla et al., 2006). The solutions of those equations were also obtained. Formulations presented in this paper were tested for seven different values of keff with external neutrons source constant in time and for a powers ratio varying exponentially over time.

PO-0815: External neutron spectra measurements for a single room compact proton system

PO-0815: External neutron spectra measurements for a single room compact proton system

The Analysis of the External Neutron Monitor Responses in a Simplified JET-Like Tokamak Using ADVANTG

The application of the Automated Variance Reduction Generator (ADVANTG) code to accelerate MCNP neutron transport calculations in fusion-relevant geometries is presented. The ADVANTG code generates variance-reduction parameters using the Consistent Adjoint Driven Importance Sampling (CADIS) and Forward-Weighted Consistent Adjoint Driven Importance Sampling (FW-CADIS) methods based on deterministic transport calculations performed by the discrete ordinates code Denovo. The aim of ADVANTG is to reduce the MCNP computational time by automating the process of variance-reduction parameter generation. ADVANTG was tested on a simplified model of a JET-like tokamak that in spite of its simplicity retains all the major characteristics of such a tokamak. The performance of the nuclear data libraries provided with ADVANTG and of various other ADVANTG/Denovo settings on variance-reduction efficiency was tested. Several cases using deuterium-deuterium or deuterium-tritium (D-T) volumetric (plasma) sources and 252Cf or D-T point neutron sources were analyzed to find guidelines for successful use of the code for fusion applications. Additionally, the use of ADVANTG as a tool to identify major neutron pathways from the neutron source to the detector is demonstrated.