Neutron Fluence Distribution Research Articles

Estimation of neutron fluence distribution within German PWR components for decommissioning studies

On 15 April 2023, the last German NPPs were shut down. The final shutdown is followed by a post-operational phase in which measures can be carried out to prepare for the NPPs dismantling and decommissioning. One of the tasks in preparation for the dismantling is to acquire accurate knowledge of the activation distribution within the reactor components. The activation levels depend on the neutron fluence within the reactor. Therefore, to achieve this task, a 3D detailed Monte Carlo model of a German PWR has been developed to calculate the neutron fluence characteristics (spectrum and distribution) within the components of such nuclear reactors type. The reliability of the calculations was validated via metal foil-activation measurements carried out in two active PWRs. This paper presents the Monte Carlo model and describes the calculations and measurement procedures.

Measurement of energy and directional distribution of neutron ambient dose equivalent for the [formula omitted]Li(p,n)[formula omitted]Be reaction

Double differential neutron fluence distributions were measured in the 7Li(p,n)7Be reaction for proton beam energies 7, 9 and 12 MeV. Seven liquid scintillator based detectors were employed to measure neutron fluence distributions using the Time of Flight technique. Neutron ambient dose equivalents were determined from the measured fluence distribution using ICRP (International Commission on Radiological Protection) recommended fluence to dose equivalent conversion coefficients. Neutron dose equivalents were also measured using a conventional BF3 detector based REM counter. Ambient dose equivalent measured by the REM counter is found to be in agreement with that determined from the neutron fluence spectra within their uncertainties. Angular distributions of the ambient dose equivalents were also determined from the measured fluence distributions at different angles.

A dose calculation algorithm for boron neutron capture therapy using convolution/superposition method

The convolution/superposition (C/S) method originally designed for photon dose calculation was first applied for developing a treatment planning system for boron neutron capture therapy. The original concept of TEGMA (total energy generated per unit mass) was proposed to represent distinctive dose components from neutron reactions with the elements in the patient's tissue. First, neutron fluence distributions in a homogeneous brain phantom irradiated with an energy-groupwise pencil beam of 2.5 × 2.5 mm2 were calculated using the MCNP6.2 code. Then, a library of energy-groupwise TEGMA and KERMA were generated and stored in the developed C/S code. As a benchmark, dose distributions in a cuboid phantom and a human head phantom were calculated using the developed C/S and PHITS Monte Carlo codes. A neutron beam having a continuous epithermal spectrum and a square field of 22.5 × 22.5 mm2 or a circle field of 22.5 mm in diameter was assumed to be incident on the phantoms. The human head phantom was created by the pre-processing including the voxelization and transformation of test DICOM CT images. The differences in boron doses between C/S and MC ranged from 2% to 6%. In nitrogen doses, the differences were from 4% to 9%. A large discrepancy observed in hydrogen lateral dose profiles could be explained by the differences in cross-section data and recoil-proton transport algorithms of MCNP6.2 and PHITS. With isodose curves normalized at the center of the tumor in the human head phantom, they were almost identical in the range of 60%–110% for both cases. The C/S have underestimated the backscattering neutron and showed a larger absorbed dose gradient around 40% region. The calculation time of C/S using Intel i7-10700 processor was less than 1 min for both phantoms. The calculation time of PHITS using three Intel Xeon E5-2640 v4 processors was 15.5 min for the cuboid phantom and ∼380 min for the human head phantom. The proposed algorithm has the advantages of high speed while promising fair accuracy in BNCT dose calculations.

Multisignal ionization chamber with a B4C coating as an active neutron beam spectrometer: Monte Carlo simulations.

In this study, the influence of additional 10B4C polarizing electrodes coating on KW-1 multisignal ionization chamber performance was investigated. Using the FLUKA Monte Carlo code energy deposition, neutron fluence and neutron track-length distributions were simulated. Further calculations allow us to obtain response functions for monoenergetic neutron beams in the energy range from 1 meV to 30MeV. Results indicate further upgrades that can be adapted for moderator-based gaseous detectors for the neutron beam spectrometry approach.

Optimization of shadow shield and analysis of radiation characteristics for nuclear powered spacecraft

The layout of nuclear-powered spacecraft is very different from the layout of nuclear reactors used on Earth, so the design principle is also different. Taking nuclear-powered spacecraft as the object, this paper studies the coverage of the shadow area formed by the shield and distribution of neutron fluence and photon dose in the shadow area. In this paper, a nuclear power spacecraft model with a width of 6 m and a length of 25 m is built. The Monte Carlo method is used to calculate the transport of neutrons and photons, and the distribution of neutrons and photons in space is obtained. The coverage of the shadow area formed by shields with different widths is obtained, and the main factors affecting the width and angle of the shadow area are summarized. When the diameter of the shield decreases, the width of the shadow area also decreases, and the angle of the shadow area decreases, which greatly reduces the area of the shadow area. The distribution of neutron fluence and photon dose in the shadow area is obtained. At the end of the truss, the maximum value of the photon dose is 1.03 × 10-4 Gy∙s-1, and the average value is 6.42 × 10-5 Gy∙s-1. The maximum value of the neutron fluence rate is 1.01 × 104 n∙cm-2∙s-1, and the average value is 6.18 × 103 n∙cm-2∙s-1. And the influence of spacecraft structure on radiation is analyzed. The truss and droplet emitters have a greater impact on photons, and the heat sink and droplet emitters have a greater impact on neutrons. Based on the above work, a local shield scheme is proposed to reduce the mass. After being blocked by local shield, the photon dose is reduced to 1/10 and the neutron fluence rate is reduced to 1/6.

Peripheral Organ Equivalent Dose Estimation Procedure in Proton Therapy.

The aim of this work is to present a reproducible methodology for the evaluation of total equivalent doses in organs during proton therapy facilities. The methodology is based on measuring the dose equivalent in representative locations inside an anthropomorphic phantom where photon and neutron dosimeters were inserted. The Monte Carlo simulation was needed for obtaining neutron energy distribution inside the phantom. The methodology was implemented for a head irradiation case in the passive proton beam of iThemba Labs (South Africa). Thermoluminescent dosimeter (TLD)-600 and TLD-700 pairs were used as dosimeters inside the phantom and GEANT code for simulations. In addition, Bonner sphere spectrometry was performed inside the treatment room to obtain the neutron spectra, some relevant neutron dosimetric quantities per treatment Gy, and a percentual distribution of neutron fluence and ambient dose equivalent in four energy groups, at two locations. The neutron spectrum at one of those locations was also simulated so that a reasonable agreement between simulation and measurement allowed a validation of the simulation. Results showed that the total out-of-field dose equivalent inside the phantom ranged from 1.4 to 0.28 mSv/Gy, mainly due to the neutron contribution and with a small contribution from photons, 10% on average. The order of magnitude of the equivalent dose in organs was similar, displaying a slow reduction in values as the organ is farther from the target volume. These values were in agreement with those found by other authors in other passive beam facilities under similar irradiation and measurement conditions.

Computational Investigation of Neutron Field modified with Various Materials

Medical accelerators, such as cyclotrons for Positron Emission Tomography (PET) and linear accelerators (LINAC), generate neutrons as a byproduct during operation. Neutron fields in these facilities tend to vary depending on the facility design as well as the presence of various materials in the facility. Unfortunately, these variations in the neutron field cannot be accounted for by neutron survey meters that are typically deployed in the workplace for radiation monitoring. In this work, the researchers used Monte-Carlo-N-Particle (MCNP) 6.1 to investigate the extent of field modification with materials that are commonly found in accelerator facilities. A computational investigation was performed using the field of the 252Cf source in the Philippine Nuclear Research Institute Neutron Laboratory (PNL) that was modified using polyethylene, lead, concrete and water. Variations in the spatial and energy distribution of neutron fluence as a result of introducing each modifier were presented. It was found that the relative deviation in the neutron dose due to polyethylene, lead, concrete and water modifiers were 96, 47, 83 and 94%, respectively. The result of this work was used as a basis for experiments performed at the PNL to investigate the limitations of commercial neutron survey meters when conducting dose measurements in different neutron fields.

Radiological characterization of a German pressurized water reactor based on a highly resolved method for activity analysis and dose rate calculation

Abstract. The amendment to the atomic act in 2011 results to phase out nuclear energy in Germany until the end of 2022. Subsequently, the licensee of the nuclear power plant is responsible for decommissioning and dismantling. During operation, activation of structures near the core of the reactor occur which govern the amount of radioactive waste, the dose rate distribution and dismantling strategies. Thus, a detailed radiological characterization of in-core and out-core structures is required to optimize decommissioning processes regarding the quantification and minimization of radioactive waste, radiation protection and reducing radiation exposure. These objectives are achieved using an innovative and efficient method developed and applied at the Chair of Repository Safety (Lehrstuhl für Endlagersicherheit, ELS) RWTH Aachen University. Within the framework of the joint project „Development of a methodology for activity analysis and dose rate estimation“, funded by the Federal ministry of Education and Research, approaches the objective to develop a standardized and highly resolved method to calculate time-dependent activity of components and structures near the reactor core based on operating history of the nuclear power plant and neutron fluence distribution. The approach requires the development of a detailed model for Monte-Carlo simulations which provides the basis to neutron fluence, neutron spectra and radiation transport simulations. To calculate the nuclide specific 3-Dimensional (3D) activity distribution of the entire facility, a facility-dependent activation cross section library is produced which focuses on recent nuclear databases (ENDF/B-VIII.0). A highly resolved and space-dependent 3D activity distribution of the entire facility is obtained using a modular program package, developed at ELS, including the activation code ORIGEN2. The results are produced in the form of detailed 3D activity maps. The source terms are generated on the basis of the space-dependent 3D activity distribution using an additional module of the program package. The combination of recent nuclear databases focusing on ENDF/B-VII.1 and complemented by JEFF-3.3 ensures a comprehensive characterisation of source terms. Subsequently, source terms are prepared for 3D radiation transport simulation using the Monte-Carlo method and the computer code MCNP. The simulations are conducted separately for each individual component obtaining the partial contribution of all in-core and out-core structures as well as the dose rate distribution of the entire facility. Similar to the activity calculation, the simulation results are used to generate 3D gamma flux and dose rate maps using the graphic module of the whole program system. On the basis of the radiological characterisation and in view of a high-level radiation protection these maps allow the optimum planning and realisation of the decommissioning and dismantling process of the nuclear power plant.

Modeling long-term distribution of fast and thermal neutron fluence in degraded concrete biological shielding walls

Since neutron radiation can deteriorate mechanical properties of the concrete materials, it’s critical to obtain accurate neutron radiation levels in concrete structures during their service live. A two-group neutron diffusion model for fast and thermal neutron transport is developed in this study which considers the kinetic energy of the neutron more realistically. Based on the cross-property correlation theory, the variation of transport properties of concrete can be estimated using the degradation of mechanical properties of concrete, then, the thermal and fast neutron radiation fields with two representative energy levels as well as the thermal field in a concrete bioshield can be calculated. Coupled radio-thermo analysis of a section of simplified concrete bioshield was performed to 80 years as a case study, and the results were discussed and compared with the results from the one-speed model. The degradation of concrete induced by neutron radiation and elevated temperature can lead to considerable increases of thermal and fast neutron flux/fluence in the concrete. The profiles of thermal neutron flux/fluence are very different from the profiles of fast neutron flux/fluence. A proper treatment of the neutron energy dependence is needed to predict long-term neutron radiation levels in concrete biological shields accurately.

Neutron fluence distribution in experiments with 3 MA deuterium gas-puff z-pinch

Deuterium gas-puff z-pinches are very efficient laboratory sources of neutron pulses. Using a novel hybrid gas-puff load on the GIT-12 generator, a significant increase in the neutron yields up to 5.6×1012 is reached. At the same time, a very broad neutron energy spectrum up to energies on the order of tens of MeV is observed. In these experiments, the neutrons are produced not only by the D(d,n)3He (DD) nuclear reaction but also by reactions of multi-MeV deuterons with the experimental hardware. The angular distribution of the neutron fluence and dependence of the fluence on the distance from the z-pinch are measured using neutron activation diagnostics. By this method, a number of produced non-DD neutrons and anisotropy of DD neutron emission are evaluated.

Experimental evaluation of the Geant4-calculated response functions of a Bonner sphere spectrometer on monoenergetic neutron sources

The response function of a Bonner sphere spectrometer (BSS) plays a crucial role in deriving reliable energy distributions of neutron fluence from the experimental count of each sphere by using unfolding techniques. When using Monte Carlo codes to calculate the responses of a BSS, the factors of Bonner spheres descripted in the codes, e.g., material density and geometrical parameters, will influence the accuracy of the calculated responses. Therefore, experimental evaluation of the calculations using neutron sources is essential for establishing the response matrix. A Monte Carlo code Geant4 was used to calculate the responses of a BSS based on a 3He gas counter. The uncertainties of the calculated responses were analyzed after considering some factors, including the air gap between the gas counter and spherical moderators, the mass density of moderators, and the 3He gas pressure in the counter. Good agreement was found between the calculated responses (within uncertainties) and the experimental results on six ISO recommended monoenergetic neutron sources with energies from 144 keV to 14 MeV. To evaluate the capability of the Geant4 code in establishing the response matrix of a BSS, the calculated responses before and after experimental evaluation were illustrated and compared with the results obtained by using the Monte Carlo code MCNP5 for the BSS response calculation. These comparison results provide primary support of the effectiveness of the Geant4 code in establishing the response matrix of a BSS.

Evaluation of mechanical properties distribution for irradiated stainless steels simulated by utilizing the gradient of the cold working ratio

It is known that irradiation damage can cause a reduction in fracture toughness and an increase in the strength of austenitic stainless steels used for the internal structures of nuclear power plant reactors. A change in the mechanical properties of material, such as fracture toughness and strength, affects the fracture behavior of the components. Actual reactor internal structures, such as the top guides and core shrouds in BWR plants, are irradiated from the core in a single direction. This causes the attenuation of neutron flux, which results in the distribution of neutron fluence in the components, meaning that the distribution of mechanical properties is in the through-wall direction of the components. It is therefore possible for the fracture mode of the components to change in the through-wall direction. It would be valuable for more accurate assessment of the structural integrity of the nuclear components to make the effect of the mechanical properties distribution on the fracture mode clear. However, it is difficult to prepare large-scale test specimens with the mechanical properties distribution induced by uni-directional neutron irradiation. In this study, the mechanical properties distribution was simulated by applying the gradient of the cold working ratio using Type 316L stainless steel plate with an uneven thickness. The targeted range of the cold working ratio was designed by assuming BWR reactor internal structures, such as the top guides. The mechanical properties distribution was evaluated based on the results of hardness, tensile and fracture toughness tests. It was found that the macroscopic mechanical properties of irradiated stainless steels can be simulated by using the cold working and the macroscopic distribution of the mechanical properties of irradiated stainless steels can also be simulated by using the cold working ratio with a gradient. Furthermore, fracture tests were performed using test plates with the mechanical properties distribution assuming irradiated components with a crack. In some cases, the maximum loads obtained by the tests were almost the same value in spite of the different cold working ratio in the ligament. Based on the results of the fracture tests, it was considered that the dominant mechanical properties for the fracture load were those at the tip of the slit, not those in the ligament.

Numerical analysis of VVER-440/213 concrete biological shield under normal operation

The numerical approach for the coupled damage-creep modeling of concrete biological shield, which combines the current and the past knowledge regarding the effects of irradiation and temperature on concrete with the real measured and calculated neutron fluence and temperature distribution for VVER-440/213 reactors, is described in detail in this study. The proposed approach takes into account the real structural geometry as well as the real neutron fluence and temperature distribution and the latest knowledge about the effect of irradiation and temperature on concrete strength and stiffness. The radiation induced volumetric expansion and the thermal expansion of concrete are modeled. According to the results of the numerical simulation, the analyzed structure reaches critical damage within the time interval from 10.00 to 35.25 years of normal operation. The damage of the concrete biological shield of the VVER reactor will not affect the load-bearing function of the containment building, since the biological shield is self-bearing. The shielding properties of the biological shield may be reduced due to the appearance of the radial cracks, however, the concrete wall, which is situated right behind the biological shield, will ensure the necessary shielding. Therefore, the concrete biological shield of the VVER reactors can be considered as sacrificial structure and can be damaged without significant consequences. However, this study implies the importance of capability to predict the behavior of those PWR reactor biological shields which serve both the load-bearing and shielding purposes.

Mini Subcritical Nuclear Reactor

A mini subcritical nuclear reactor was designed using Monte Carlo methods. The reactor has light water as moderator, natural uranium as fuel, and a 239PuBe neutron source. In the design uranium fuel was modeled in an arrangement of concentric rings: 8.5, 14.5, 20.5 26.5, 32.5 cm-inner radius, 3 cm-thick, and 36 cm-high. Different models were made from a single ring of natural uranium to five rings. For each case, the neutron spectra, the neutron fluence distribution, the effective multiplication factor, the amplification factor, and the reactor power were estimated. The ambient dose equivalent rate outside the mini reactor was also estimated. The maximum value for the keff (0.78) was obtained when five rings of fuel were used; this value is close to 0.86 which belongs to a Nuclear Chicago subcritical reactor which requires almost twice the amount of uranium than the mini subcritical reactor.

Long-term neutron radiation levels in distressed concrete biological shielding walls.

Neutron radiation can deteriorate mechanical properties of the concrete materials, and thus it is questionable that neutron transport properties of concrete can remain unchanged during the life span of biological shielding walls. one-speed neutron diffusion equation and heat conduction equation were used as governing equations for prediction of neutron radiation and thermal field in concrete, respectively. The potential variations of transport properties due to neutron radiation and elevated temperature were estimated. A simplified example of a typical concrete biological shielding wall was analyzed up to 80 years, and the results were discussed. The radiation damage and radiation heating lead to minor changes of the temperature profile in the concrete. However, neutron radiation and elevated temperature can result in considerable increases of neutron flux and fluence in the concrete. The damage of concrete induced by neutron radiation and elevated temperature can considerably accelerate the penetration of neutron radiation into the concrete. This work is the first attempt to deal with the degradation of neutron and heat transport properties of concrete and its effect on neutron fluence distribution in concrete, and provides a possible way to determine the long-term neutron and thermal fields in concrete biological shielding walls.

NEUTRON WORKPLACE SPECTROMETRY (ENERGY AND DIRECTION) USING TL DETECTORS: FIRST APPROACH AND RESPONSE FUNCTIONS.

The response of albedo dosemeters depends on the energy and angle of the incident neutron radiation. For their use as personal dosemeters, a field-calibration factor has to be applied. The presently used single sphere method for field calibration can be extended and optimised by putting five albedo dosemeters on the surface of a polyethylene sphere and two TL cards in the centre. To investigate the potential of this extension, reference measurements and Monte Carlo calculations were performed and the fluence response of the detectors at different positions on and within the sphere was determined. Calculated response functions demonstrate that information on the energy and directional distribution of neutron fluence can be extracted with this simple set-up for unknown neutron fields.

Displacement damage cross section and mechanical properties calculation of an Es-Salam research reactor aluminum vessel

Nuclear facility aging is one of the biggest problems encountered in nuclear engineering. Radiation damage is among one of the aging causes. This kind of damage is an important factor of mechanical properties deterioration. The interest of this study is on the Es-Salam research reactor aluminum vessel aging due to neutron radiation. Monte Carlo (MC) simulations were performed by MCNP6 and SRIM codes to estimate the defects created by neutrons in the vessel. MC simulations by MCNP6 have been performed to determine the distribution of neutron fluence and primary knock-on atom (PKA) creation. Considering our boundary conditions of the calculations, the helium and hydrogen gas production in the model at a normalized total neutron flux of 6.62 × 1012 n/cm2 s were determined to be 2.86 × 108 and 1.33 × 109 atoms/cm3 s, respectively. The SRIM code was used for the simulation of defects creation (vacancies, voids) in the aluminum alloy of the Es-Salam vessel (EsAl) by helium and hydrogen with an approximate energy of 11 MeV each. The coupling between the two codes is based upon post-processing of the particle track (PTRAC) output file generated by the MCNP6. A small program based on the MatLab language is performed to condition the output file MCNP6 in the format of a SRIM input file. The concentration of silicon was determined for the vessel by the calculation of the total rate of 27Al(n,γ)28Si reaction. The DPA (displacement per atom) was calculated in SRIM according to R.E. Stoller recommendations; the calculated value is 0.02 at a fast neutron fluence 1.89 × 1019 n/cm2. RCC-MRx standard for 6061-T6 aluminum was used for the simulation of the evolution of mechanical properties for high fluence. The calculated values of nuclear parameters and DPA obtained were in agreement with the experimental results from the Oak Ridge High Flux Isotope Reactor (HFIR) reported by Farrell and coworkers.

The spatial dose rate and energy characterization of a P 385 D-D Neutron Generator using a nested neutron spectrometer and a tissue equivalent proportional counter

A nested neutron spectrometer (NNS) and a tissue equivalent proportional counter (TEPC) were used to measure the spatial dose equivalent and energy neutron fluence distribution of a P 385 D-D Thermo Fisher Scientific Neutron Generator located in the mixed field dosimetry facility at the University of Ontario Institute of Technology. The P 385 Neutron Generator uses the D-D nuclear reaction to generate neutrons with an expected average energy of 2.5 MeV. The NNS uses a He-3 proportional counter to measure the neutron capture count rate as a function of nested cylindrical moderator thickness. Using the NNS and associated unfolding software, the energy neutron fluence distribution at various locations from the P 385 neutron generator were determined along with other radiation protection quantities of interest. Measurements of the dose equivalent rate were also conducted at the same locations using a 5-inch spherical TEPC and compared to the NNS results. The TEPC was further used to measure the dose equivalent as a function of radial distance from the neutron generator target and also for different P 385 operating conditions.

Development of fuel temperature calculation code for HTGRs

In order to ensure the thermal integrity of fuel in High Temperature Gas-cooled Reactors (HTGRs), it is necessary that the maximum fuel temperature in normal operation is to be lower than a thermal design target. In the core thermal-hydraulic design of block-type HTGRs, the maximum fuel temperature should be evaluated considering data such as thermal power, core geometry, power density and neutron fluence distributions, and core coolant flow distribution. The fuel temperature calculation code used in the design stage of the High Temperature engineering Test Reactor (HTTR) presupposes to run on UNIX systems, and its operation and execution procedure are complicated and are not user-friendly. Therefore, a new fuel temperature calculation code named FTCC which has a user-friendly system such as a simple and easy operation and execution procedure, was developed. This paper describes calculation objects and models, basic equations, improvement points from the HTTR design code in FTCC, and the result of a validation calculation with FTCC. The calculation result obtained by FTCC provides good agreement with that of the HTTR design code. In addition, the effect of cooling forms on fuel is investigated by using FTCC.

Neutron Energy Reconstruction and Fluence Determination at 27 keV With the LNE-IRSN-MIMAC MicroTPC Recoil Detector

The aim is to characterize the energy distribution of neutron fluence in the energy range 8 keV–5 MeV based on a primary standard: the LNE-IRSN/MIMAC microTPC. The microTPC is a time projection chamber. Time projection chambers are gaseous detectors able to measure charged particles energy and to reconstruct their track. The gas is used as a (n, p) converter in order to detect neutrons down to few keV. The neutron energy is reconstructed event by event thanks to proton scattering angle and proton ionization energy measurements. The scattering angle is deduced from the 3-D track. The proton energy is obtained by charge collection measurements, knowing the ionization quenching factor. The fluence is reconstructed thanks to the detected events number and the simulation of the detector response. The microTPC is a new reliable detector able to measure energy distribution of the neutron fluence without unfolding procedure or prior neutron calibration contrary to usual gaseous counters. The microTPC is characterized at the AMANDE facility, with neutron energies going from 8 keV to 565 keV. This work shows the first direct reconstruction of neutron energy and fluence, simultaneously, at 27.2 keV in a continuous irradiation mode.