Neutron Source Strength Research Articles

Deterministic and stochastic computational models for the analysis of Nuclear Power Plant (NPP) start-up operations

The aim of this paper is to present an overview of both the deterministic and stochastic aspects of nuclear power plant (NPP) start-up and start-up accident scenarios. The neutron kinetics and corresponding nuclear reactor control mechanisms associated with deterministic NPP start-up are explained. The stochastic behaviour of neutron kinetics is explained and the procedures required to safely start-up an NPP due to these stochastic phenomena are outlined. Specific attention is given to the stochastic behaviour of neutrons in a low neutron source regime and the corresponding mathematical methods required for modelling such phenomena during NPP start-up accidents. The Pál-Bell equations are compared against the forward generating function equations utilising the gamma approximation in their ability to model NPP start-up accidents. The Pál-Bell equations are shown to be necessary to accurately calculate the maturity time for systems with either a very low neutron source strength or a very rapid reactivity insertion profile. Both the Hurwitz calculation route and MacMillan calculation route for low neutron source systems are explained. Full deterministic and stochastic calculations are performed for multiple nuclear reactor systems to illustrate the necessary analyses required for low neutron source NPP start-up accident scenarios.

A Method to Calculate the Maturity Time for Low Neutron Source Nuclear Reactor Startup Simulations

An approximate method for determining the maturity time is presented for applications in low neutron source nuclear reactor startup simulations. This new method relies only on the calculation of the mean neutron density and does not require the additional calculation of the variance in the neutron density as the traditional method does. The most accurate method for determining the safe neutron source strength, required to sufficiently mitigate the probability of a rogue transient during nuclear reactor startup, uses the Pál-Bell equations. However, as space and energy dependencies are included, the numerical computation become computationally demanding. Therefore, approximate methods that significantly reduce the computation time and improve the computational efficiency of the simulation while remaining very accurate are extremely useful. The approximate method for determining the maturity time presented in this study has shown excellent agreement with traditional methods while offering an order of magnitude reduction in computation time.

Study on the shielding performance of bismuth oxide as a spent fuel dry storage container based on Monte Carlo simulation

For traditional spent fuel shielding materials, due to physical and chemical defects and cost constraints, they have been unable to meet the needs. Therefore, this paper carries out the first discussion on the application and performance of bismuth in neutron shielding by establishing Monte Carlo simulation on the neutron flux model of shielded spent fuel. Firstly, functional fillers such as bismuth oxide, lead oxide, boron oxide, gadolinium oxide and tungsten oxide are added to the matrices to compare the shielding rates of aluminum alloy matrix and silicone rubber matrix. The shielding rate of silicone rubber mixture is higher than aluminum alloy mixture, reaching more than 56%. The optimal addition proportion of bismuth oxide and lead oxide is 30%, and the neutron radiation protection efficiency reaches 60%. Then, the mass attenuation coefficients of bismuth oxide, lead oxide, boron oxide, gadolinium oxide and tungsten oxide in silicone rubber matrix are simulated with the change of functional fillers proportion and neutron energy. This simulation result shows that the mixture with functional fillers has good shielding performance for low energy neutrons, but poor shielding effect for high energy neutrons. Finally, in order to further evaluate the possibility of replacing lead oxide with bismuth oxide as shielding material, the half-value layers and various properties of bismuth oxide and lead oxide are compared. The results show that the shielding properties of bismuth oxide and lead oxide are basically the same, and the mechanical properties, heat resistance, radiation resistance and environmental protection of bismuth oxide are better than that of lead oxide. Therefore, in the case of neutron source strengths in the range of 0.01–6 MeV and secondary gamma rays produced below 2.5 MeV, bismuth can replace lead in neutron shielding applications.

Polarization of Compton-Scattered Prompt Gamma-Ray Technique for HEU Detection at 186 keV

Prompt gamma-ray polarization is a practical method for detecting highly enriched uranium (HEU) in concealed sources. It also provides information on their geometry, magnetic fields, and radiation mechanisms. However, prompt gamma-ray polarization measurements have rarely been applied in nuclear nonproliferation areas to detect HEU. In this study, the feasibility of detecting the characteristic energy peak of 186 keV, which is associated with the asymmetry of the activation mechanism and the detection of energy-dependent polarization from concealed HEU sources, was evaluated using the Compton scattering approach. A Monte Carlo N-particle transport code simulation was used to realize the activation mechanism of HEU via two 1.4-mm strips of converter material [i.e., cesium lead tribromide (CsPbBr3)], transported by secondary scattered gamma rays during the three-stage process of Compton scattering, polarization, and detection. This paper presents the mathematical model, the physics of Compton scattering, and the polarization mechanism for the detection technique. In this case, the physics is relevant to both processes in which the emitted secondary scattered gamma rays undergo initial orthogonal polarization. Specifically, to meet the objective of testing the technical protocol for the enhanced detection of energy peaks associated with HEU, particularly 186 keV, simulations were conducted to quantify the HEU volume and neutron source strength in the MCNP data card to perform error analysis. The detector system had the potential to acquire good resolved photopeak with a 4.5% relative error or less, with a 1 Ci source activity, and a peak-to-background ratio of 1.15. This resolution took 163 s for high-purity germanium detection, which is comparable to current methods used for material detection placed within 100 to 900 s to completion. The small error difference was due to the attributes of the phenomenal enhancement properties of cesium tribromide and polarimetry. The identified photo peaks included K-shell X-rays from 235U, 61 keV from fission, 511-keV annihilation, and the peak of interest at 186 keV. The result from spectral analyses showed clear signatures related to pure and adulterated HEU. HEU detection with the low neutron yield and the easiness of shielding the yield of the HEU sample showed that the HEU characterization was feasible when shielded, with the highest success rate under both enhancement approaches. The optimization and scale-up of this technique are expected to enhance its applications in a large-scale HEU detection design.

Neutron source strength verification via reaction rate measurement and Monte Carlo simulation

Determining the neutron source strength is essential to ensure optimal use of the source and to implement adequate radiation protection measures. In this study, the source strength of two (α,n) neutron sources – a newly acquired 185 GBq 241AmBe and a legacy source 74 GBq 239PuBe – were verified. Source strength verification was performed by measuring the induced activities in irradiated 115In foils and comparing the results to MCNP calculated reaction rates. The results for 241AmBe agree well with a maximum and average deviations of 8.1% and 4.1%, respectively. While for 239PuBe the maximum deviation is 11.7% and the average deviation is 6.3%. Results of this study provide a reliable reference for the application of the sources, in particular, for the 241AmBe which is used as the external neutron source of the PRR1-Subcritical Assembly for Training, Education, and Research (PRR1-SATER), and for the design of a neutron irradiator that will repurpose the 239PuBe.

Feasibility of neutron coincidence counting for spent TRISO fuel

High-temperature gas reactors rely on TRIstructural-ISOtropic (TRISO) fuel for enhanced fission product retention. Accurate fuel characterization would improve monitoring of efficient fuel usage and accountability. We developed a new neutron multiplicity counter (NMC) based on boron coated straw (BCS) detectors and used it in coincidence mode for 235U assay in TRISO fuel. In this work, we demonstrate that a high-efficiency version of the NMC encompassing 396 straws is able to estimate the 235U in used TRISO-fueled pebbles or compacts with a relative uncertainty below 2.5% in 100s. We performed neutronics and fuel depletion calculation of the HTR-10 pebble bed reactor to estimate the neutron and gamma-ray source strengths of used TRISO-fueled pebbles with burnup between 9 and 90GWd/t. Then, we measured a gamma-ray intrinsic efficiency of 10−12 at an exposure rate of 340.87R/h. The low gamma-ray sensitivity and high neutron detection efficiency enable the inspection of used fuel.

Estimation of neutron source strength for external source subassembly in PFBR using GEANT4

Proper estimation of neutron source strength from external source subassembly of PFBR is very important. The present work uses the radiation transport simulation toolkit GEANT4 to do the same by modelling full geometric details of the source. Simulations have been carried out to first determine the gamma source strength from activated Antimony isotopes in the subassembly. This is followed by transport of these Antimony decay gamma photons to estimate photo-neutron production yield. The calculated results are in good agreement with earlier calculations, indicating feasibility of GEANT4 code for use in FBR calculations.

Improved S factor of the C12(p,γ)N13 reaction at E=320–620 keV and the 422 keV resonance

The 12C(p,{\gamma})13N reaction is the onset process of both the CNO and Hot CNO cycles that drive massive star, Red and Asymptotic Giant Branch star and novae nucleosynthesis. The 12C(p,{\gamma})13N rate affects the final abundances of the stable 12,13C nuclides, with ramifications for meteoritic carbon isotopic abundances and the s-process neutron source strength. Here, a new underground measurement of the 12C(p,{\gamma})13N cross-section is reported. The present data, obtained at the Felsenkeller shallow-underground laboratory in Dresden (Germany), encompass the 320-620 keV center of mass energy range to include the wide and poorly constrained E = 422 keV resonance that dominates the rate at high temperatures. This work S-factor results, lower than literature by 25%, are included in a new comprehensive R-matrix fit, and the energy of the 1+ first excited state of 13N is found to be 2369.6(4) keV, with radiative and proton width of 0.49(3) eV and 34.9(2) keV respectively. A new reaction rate, based on present R-matrix fit and extrapolation, is suggested.

Capture γ-Ray Dose Equivalent at Double-Bend Maze Entrance; Monte Carlo Simulation and Analytical Methods and Measurements

Purpose: The production of the secondary neutron in the high-energy megavoltage medical accelerator machines has been extensively studied. In this study, MCNP5 MC code and two analytical methods, the proposed method and International Atomic Energy Agency (IAEA) 47 proposed method were used to capture γ-ray dose equivalent calculation.
 Materials and Methods: MCNP5 code of the MC simulation method was used for code calculation in this study. The main components of a Varian 2100Clinac were simulated as well as a 30×30×30 cm3 water phantom, in a Source to Surface Distance (SSD) of 100cm. Apparent neutron source strength (QN) was obtained using F1, *F8 tallies, and a small scoring cell at the isocenter with a mass equal to 0.625g.
 Results: QN was obtained as 1.25 n/Gy X for the simulated Linear Accelerators )linac( head and was used in the other calculations. In the simulated double-bend maze treatment room with first and second lengths of the maze as 7m and 3m, the proposed method calculated capture γ-ray dose with 6.2% and 60% differences compared with MC simulation and IAEA 47 methods, respectively.
 Conclusion: We concluded that Ghiasi and Mesbahi's proposed method performed better in capturing γ-ray dose equivalent calculation compared to IAEA 47 report. The proposed method reduced the difference from 60% to 6.2%.

Nuclear Analyses for the Assessment of the Loads on the ITER Radial Neutron Camera In-Port System and Evaluation of Its Measurement Performances

The radial neutron camera (RNC) is a key ITER diagnostic system designed to measure the uncollided 14- and 2.5-MeV neutrons from deuterium–tritium (DT) and deuterium–deuterium (DD) fusion reactions, through an array of detectors covering a full poloidal plasma section along collimated lines of sight (LoS). Its main objective is the assessment of the neutron emissivity/ <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$\alpha $ </tex-math></inline-formula> source profile and the total neutron source strength, providing spatially resolved measurements of several parameters needed for fusion power estimation, plasma control, and plasma physics studies. The present RNC layout is composed of two fan-shaped collimating structures viewing the plasma radially through vertical slots in the diagnostic shielding module (DSM) of ITER Equatorial Port 1 (EP01): the ex-port subsystem and the in-port one. The ex-port subsystem, devoted to the plasma core coverage, extends from the Port Interspace to the Bioshield Plug: it consists of a massive shielding unit hosting two sets of collimators lying on different toroidal planes, leading to a total of 16 interleaved LoS. The in-port system consists of a cassette, integrated inside the port plug DSM, containing two detectors per each of the six LoS looking at the plasma edges. The in-port system must guarantee the required measurement performances in critical operating conditions in terms of high radiation levels, given its proximity to the plasma neutron source. This article presents an updated neutronic analysis based on the latest design of the in-port system and port plug. It has been performed by means of the Monte Carlo MCNP code and provides nuclear loads on the in-port RNC during normal operating conditions (NOC) and inputs for the measurement performance analysis.

Fuel Cell Imaging with a Wolter Optics Neutron Microscope

Wolter optics promise to radically transform neutron imaging by improving the time resolution of the method by at least a factor 1,000 while providing spatial resolution of order a few micrometers [1]. These optics will enable neutron tomography of operating fuel cells with a time resolution of about a few minutes, which will provide unique insight into the water transport phenomena in membrane electrode assemblies of fuel cells and electrolyzers.Neutron imaging has played a critical role in understanding water transport in proton exchange membrane fuel cells and electrolyzers [2]. The primary limitation of the method is the inherent neutron source strength, which is about a billion times weaker than X-ray synchrotron sources. Additionally, neutrons are difficult to focus using refractive lenses so that neutron imaging beamlines heretofore can be described as pinhole cameras. In a pinhole camera, the best achievable spatial resolution (λg) is determined by the diameter of the pinhole (D ~1 cm), the pinhole-to-detector distance (L ~10 m), and the sample-to-detector distance (Z ~1 cm):λg = Z D / (L – Z) ~ Z D / LReturning to the source strength, the neutron fluence rate at the sample position scales as (D / L)2. This means that higher spatial resolution for finite sized objects requires sacrificing neutron intensity and incurring long exposure times. Typical measurements require about 20 minutes for a single two-dimension image of the water content in fuel cell for a spatial resolution of about 10 µm. Such time requirements preclude the use of tomography, which is required to understand the water content in real fuel cell components since the MEA cannot be considered to be planar at the few micrometer level.We will describe Wolter optics, and the optical system that will be first available in 2022. After the NCNR cold source upgrade is anticipated to yield a time resolution gain of about 10,000 compared to a pinhole camera. We will discuss our expected image acquisition and analysis procedures guided by ray tracing simulations.Figure 1. Proof of concept Wolter optics image of a fuel cell test section.[1] D. S. Hussey et al., “Design of a neutron microscope based on Wolter mirrors,” Nucl. Instruments Methods Phys. Res. Sect. A Accel. Spectrometers, Detect. Assoc. Equip., vol. 987, no. October 2020, p. 164813, 2021.[2] T. A. Trabold, J. P. Owejan, J. J. Gagliardo, D. L. Jacobson, D. S. Hussey, and M. Arif, “Use of neutron imaging for proton exchange membrane fuel cell (PEMFC) performance analysis and design,” Handb. fuel cells, vol. 6, no. Table 1, 2010. Figure 1

MCCALISO: a californium neutron source strength prediction software based on the Monte Carlo method of propagating probability density functions

MCCALISO: a californium neutron source strength prediction software based on the Monte Carlo method of propagating probability density functions

Reactivity estimation based on the linear equation of characteristic time profile of power in subcritical quasi-steady state

ABSTRACT The reactivity was estimated from a time profile of neutron count rate or simulated data in a quasi-steady state (QSS) after a sudden status change of reactivity or external neutron source strength. The estimation was based on the equation of power in subcritical QSS. The purpose of the study is to develop the method of timely reactivity estimation from complicated time profile of neutron count rate. The developed method was applied to the data simulating neutron count rate created by using one-point kinetics code, AGNES, and Poisson-distributed random noise. It was also applied to the transient subcritical experiment data measured by using transient experiment criticality facility. The result shows that the difference of the estimated and reference values was within about 5% or less for for simulated data and within about 7% or less for ρ$ ≃1.4 and for the experimental data. The possibility of the reactivity estimation several 10 seconds after the status change was also shown.

Fuel Cell Imaging with a Neutron Microscope Based on Wolter Optics

Wolter optics promise to radically transform neutron imaging by improving the time resolution of the method by at least a factor 1,000 while providing spatial resolution of order a few micrometers [1]. These optics will enable neutron tomography of operating fuel cells with a time resolution of about a few minutes, which will provide unique insight into the water transport phenomena in membrane electrode assemblies of fuel cells and electrolyzers.Neutron imaging has played a critical role in understanding water transport in proton exchange membrane fuel cells and electrolyzers [2]. The primary limitation of the method is the inherent neutron source strength, which is about a billion times weaker than X-ray synchrotron sources. Additionally, neutrons are difficult to focus using refractive lenses so that neutron imaging beamlines heretofore can be described as pinhole cameras. In a pinhole camera, the best achievable spatial resolution (λg) is determined by the diameter of the pinhole (D ~1 cm), the pinhole-to-detector distance (L ~10 m), and the sample-to-detector distance (Z ~1 cm):λg = Z D / (L – Z) ~ Z D / LReturning to the source strength, the neutron fluence rate at the sample position scales as (D / L)2. This means that higher spatial resolution for finite sized objects requires sacrificing neutron intensity and incurring long exposure times. Typical measurements require about 20 minutes for a single two-dimension image of the water content in fuel cell for a spatial resolution of about 10 µm. Such time requirements preclude the use of tomography, which is required to understand the water content in real fuel cell components since the MEA cannot be considered to be planar at the few micrometer level.We will describe Wolter optics, and the optical system that will be first available in 2022. After the NCNR cold source upgrade is anticipated to yield a time resolution gain of about 10,000 compared to a pinhole camera. We will discuss our expected image acquisition and analysis procedures guided by ray tracing simulations.Figure 1: Proof of concept Wolter optics image of a fuel cell test section.

Design, fabrication and characterization of neutron irradiation facility using accelerator-based photoneutron source

In this paper we describe the design of a neutron generator assembly using an electron accelerator. We have carried out design calculations for this assembly by optimizing neutron source strength and maximizing output neutron flux at a reasonable L/D ratio keeping in view the radiography applications. For this purpose, detailed Monte Carlo calculation using EGS-4 code was carried out to estimate photon flux. The results of EGS4 calculations are used in source definition card of MCNP simulation to compute neutron source strength from Beryllium in order to arrive at an optimum design. The parameters such as neutron source strength; collimation ratio etc. were optimized. The thermal neutron flux at each position of this assembly including those at the end of collimator, Cd ratio, neutron-to-gamma ratio has been estimated. Experimental measurements of these parameters are in good agreement with simulation results.

Comparison of the response of handheld neutron detectors in differing deployment environments: Measurements, calculations and practical implications

Authorities responding to nuclear security events have to be prepared for a situation in which a neutron source or nuclear material is out of regulatory control. In such a situation, handheld neutron detectors are used to search for neutron sources or nuclear material, during either an overt or a covert search. These detectors are also used to support the initial threat assessment, radiation protection at the scene and the initial characterisation of the source. Long range neutron detection, i.e. detection several tens of meters from the source, referred to here as “stand-off” detection, is of importance to nuclear security. An estimate of the neutron source strength is especially important for the initial threat assessment. However, the neutron dose rate and the neutron response are strongly dependent on the distance between the detector and the source and on the surrounding environment. These influences should be considered when estimating the neutron source strength from initial measurements. In this paper the variation of the neutron dose rate and the neutron count rate with distance from a neutron source in different, representative environments was determined. Monte-Carlo simulations were carried out in order to gain more understanding of the results. In addition, the variation of the neutron dose rate with distance from a neutron source in different surroundings is summarized in order to assist the initial threat assessment at the scene of a nuclear security event.

Evaluation and validation of external neutron source strength of operating Fast Breeder Test Reactor

Fast Breeder Test Reactor is one of the high flux operating test facility in India. The study aims to find out the source strength of the neutron source subassembly during its residence time in the reactor core and its contribution to the observed count rates in the startup neutronic channels which are ex-core 235U coated fission detectors. The entire power history during various fuel irradiation campaigns was considered in the calculation of neutron source strength. The study is experimentally validated by shifting the external neutron source subassembly to an elevated position in the subcritical core. The experiment is devised in such a way that inherent neutron source strength is not amplified and the count rates obtained in the detectors will be solely from the external neutron source. The count rates observed during the experiment at the ex-core detectors are compared with the simulated count rates during the lifted position of the source subassembly and found to be in good agreement.

Core power control of the ADS based on genetic algorithm tuning PID controller

The accelerator driven subcritical system (ADS) reactor, an accelerator driven transmutation research facility relies on control rod and external neutron source to realize core power regulation. This paper researches the dynamic and control characteristics of the ADS core under the external neutron source disturbance. The state space model of the ADS core with multi-input multi-output is established by linearizing the core nonlinear model with small disturbance. The response characteristics of the ADS core power control under external neutron source strength regulation are simulated by using genetic algorithm tuning proportional-integral-derivative (PID) controller and traditional PID controller. The results show that under different external neutron source relative strength, the effect of genetic algorithm tuning PID controller is better than that of traditional PID controller. The ADS core power can be well controlled by using genetic algorithm tuning PID controller.

Uncertainty quantification of PWR spent fuel due to nuclear data and modeling parameters

Uncertainties are calculated for pressurized water reactor (PWR) spent nuclear fuel (SNF) characteristics. The deterministic code STREAM is currently being used as an SNF analysis tool to obtain isotopic inventory, radioactivity, decay heat, neutron and gamma source strengths. The SNF analysis capability of STREAM was recently validated. However, the uncertainty analysis is yet to be conducted. To estimate the uncertainty due to nuclear data, STREAM is used to perturb nuclear cross section (XS) and resonance integral (RI) libraries produced by NJOY99. The perturbation of XS and RI involves the stochastic sampling of ENDF/B-VII.1 covariance data. To estimate the uncertainty due to modeling parameters (fuel design and irradiation history), surrogate models are built based on polynomial chaos expansion (PCE) and variance-based sensitivity indices (i.e., Sobol’ indices) are employed to perform global sensitivity analysis (GSA). The calculation results indicate that uncertainty of SNF due to modeling parameters are also very important and as a result can contribute significantly to the difference of uncertainties due to nuclear data and modeling parameters. In addition, the surrogate model offers a computationally efficient approach with significantly reduced computation time, to accurately evaluate uncertainties of SNF integral characteristics.

Shielding for transporting an 241Am-Be source for industrial applications

Paraffin, water and water-extended polyester (WEP) were used as main moderator to design a mobile shielding for a 666 GBq 241Am–Be source used in oil industry. The shielding performance was estimated using Monte Carlo methods where the γ-rays induced by the neutron interaction with the shielding materials were also included. The spectra of neutrons and γ-rays around the shielding were estimated, as well as the total neutron and g-ray total fluences per history. The neutron source strength was used to calculate ambient dose equivalent rates, aiming to satisfy the international recommendation (2 mSv/h at the shielding surface). Moderators modify the neutron spectrum of the source reducing the amount of fast neutrons (0.5–11 MeV) and producing epithermal and thermal neutrons. During neutron transport in the moderator neutron capture and inelastic scattering produce gamma-rays. Paraffin has the best shielding performance however it is not suitable because can be affected by extreme temperatures. WEP and water have similar shielding performance, however water do not satisfy the international recommendations because at one site the neutron and γ-ray doses rates are larger than 2 mSv/h. WEP is the best option because fulfill the international recommendations, it is stable and temperatures where paraffin fails. The WEP-based shielding weights 66 kg.