Fast Neutron Fluence Research Articles

Forecasting System Requirements to the Materials of the Shell of Fuel Elements of Innovative Fast Reactors

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The effect of fast neutron irradiation on the superconducting properties of REBCO coated conductors with and without artificial pinning centers

Superconductors are essential components of future fusion power plants. The magnet coils responsible for producing the field required for confining the fusion plasma are exposed to considerable neutron radiation. This makes irradiation studies necessary for understanding the radiation response of the superconductor. High temperature superconductors are promising candidates as magnet coil materials. YBCO and GdBCO tapes of several manufacturers were irradiated to fast neutron fluences of up to 3.9 × 1022 m−2 in the research reactor at the Atominstitut. Low energy neutrons contribute to the fission reactor spectrum but not to the expected spectrum at the fusion magnets. Low energy neutrons have to be shielded in irradiation experiments to avoid their substantial effect on the superconducting properties of tapes containing gadolinium. The critical current (Ic) of the tapes in this study was examined at fields of up to 15 T and down to a temperature of 30 K. Ic first increases upon irradiation and reaches a maximum at a certain fluence, which depends highly on temperature, being highest at low temperature. Ic declines at high fluences and eventually degrades with respect to its initial value. Tapes with artificial pinning centers (APCs) degrade at lower fluences than tapes without them. The n-values decrease in all types of tapes after irradiation even when the critical currents are increased. The field dependence of the volume pinning force differs in pristine tapes with and without APCs but shows the same behavior after irradiation.

On the characteristics of the flow and heat transfer in the core bypass region of a PWR

The development of analysis models for the Swiss reactors is a key objective of the STARS project at the Paul Scherrer Institut (PSI). Within this context there is a need for the development of computational fluid dynamics (CFD) models of the Swiss reactors in support of future high fidelity investigations of steady-state and transient scenarios. This article presents initial results for the CFD analysis of a Siemens KWU PWR with a focus on the flow behaviour and heat transfer in the gap between the core shroud and core barrel. Temperatures and densities in this region of the reactor are important, for example, for accurate estimations of fast neutron fluence and activation in the steel structures of the core shroud, core barrel and reactor pressure vessel. The flow behaviour in this region may also be relevant for better understanding of ex-core detector responses. The flow conditions in the core bypass region were found to be in the transition-to-turbulence regime, with vortex shedding taking place downstream of the core formers as a result of flow instabilities. The non-stationary nature of the flow presented a challenge in terms of obtaining a solution within a reasonable time period. Two approaches were proposed to address this challenge: time-averaging of the flow-field information before solving the conjugate heat transfer problem; time-averaging of surface heat fluxes in order to derive detailed surface heat transfer coefficients. Both approaches yielded similar results with similar computational effort. Several characteristics and features of the core bypass flow are discussed. Updated Monte Carlo simulation results show that the influence of the core bypass temperatures on the neutron fluence predictions is non-negligible. This highlights the importance of including accurate bypass temperatures in future Monte Carlo simulations focused on ex-core regions.

Neutron Fluence And DPA Rate Analysis In Pebble-Bed HTR Reactor Vessel Using MCNP

In the Pebble-bed HTR reactor, the distance between the core and the reactor vessel is very close and the media inside are carbon and He gas. Neutron moderation capability of graphite material is theoretically lower than that of water-moderated reactors. Thus, it is estimated much more the fast neutrons will reach the reactor vessel. The fast neutron collisions with the atoms in the reactor vessel will result in radiation damage and could be reducing the vessel life. The purpose of this study was to obtain the magnitude of neutron fluence in the Pebble-bed HTR reactor vessel. Neutron fluence calculations in the pebble-bed HTR reactor vessel were performed using the MCNP computer program. By determining the tally position, it can be calculated flux, spectrum and neutron fluence in the position of Pebble-bed HTR reactor vessel. The calculations results of total neutron flux and fast neutron flux in the reactor vessel of 1.82x108 n/cm2/s and 1.79x108 n/cm2/s respectively. The fast neutron fluence in the reactor vessel is 3.4x1017 n/cm2 for 60 years reactor operation. Radiation damage in stainless steel material caused by high-energy neutrons (> 1.0 MeV) will occur when it has reached the neutron flux level of 1.0x1024 n/cm2. The neutron fluence results show that there is no radiation damage in the Pebble-bed HTR reactor vessel, so it is predicted that it will be safe to operate at least for 60 years.

Test of Fibre Bragg Gratings samples under High Fast Neutrons Fluence

Optical fibre sensors (OFS) are worthy of interest for measurements in nuclear reactor thanks to their unique features, particularly compact size and remote multi-point sensing for some of them. But besides non negligible constraints associated with the high temperature environment of the experiments of interest, it is well known that the performances of OFS can be severely affected by high level of radiations. The Radiation Induced Attenuation (RIA) in the fibre is probably most known effect, which can be to some extent circumvented by using rad hard fibres to limit the dynamic loss. However, when the fast neutron fluence reaches 1018 to 1019 n/cm2, the density and index variations associated to structural changes may deteriorate drastically the performances of OFS even if they are based on rad hard fibres, by causing direct errors in the measurements of temperature and/or strain changes. The aim of the present study is to access the effect of nuclear radiations on the Fabry Perot (FP) and of Fibre Bragg Grating (FBG) sensors through the comparison of measurements made on these OFS - or part of them - before and after irradiation [1]. In the context of development of OFS for high irradiation environment and especially for Material Testing Reactors (MTRs), Sake 2 experiment consists in an irradiation campaign at high level of gamma and neutron fluxes conducted on samples of fibre optics – bare or functionalised with FBG. The irradiation was performed at two levels of fast neutron fluence: 1 and 3.1019 n/cm2 (E>1MeV), at 250°± 25°C, in the SCK•CEN BR2 reactor (Mol Belgium). An irradiation capsule was designed to allow irradiation at the specified temperature without active control. The neutron fluence was measured with activation dosimeters and the results were compared with MCPN computations. Investigation of bare samples gives information on the density changes, while for the FBGs both density and refractive index perturbation are involved. Some results for bare fibres were reported recently. In this paper, we will focus on the measurements made on FBGs that have been manufactured by different laboratories on SMF 28 fibers: CEA, University of St-Etienne and University of Mons. Tested gratings have been written using various conditions (type of fibre, of laser, writing wavelength, power density, post writing thermal annealing,…), leading to various behaviours after Sake 2 irradiation. Bragg wavelength and reflectivity have been measured before and after irradiation thanks to a special mounting at the same temperature. It appears that a change in the shape after irradiation of the Bragg peak disturb the retrieval of the Bragg wavelength. The measurements show that for nearly all gratings the Bragg peak remains visible after the irradiation, and that Radiation Induced Bragg Wavelength Shifts (RI-BWSs) vary from few pm (equivalent to an error of less than 1°C for a temperature sensor) to nearly 1 nm (equivalent to 100°C) depending of the FBG types. High RI-BWSs could indeed be expected when considering the huge refractive index variation and compaction of the bare fibre samples that have been measured by other techniques. Post writing thermal annealing is confirmed as a key parameter in order to obtain a more radiation tolerant FBG. Our results show that specific annealing regimes allow making FGBs suitable to perform temperature measurements in a MTR experiment.

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.

Study on variance reduction technique based on adjoint Discrete Ordinate method

For deep-penetration shielding calculation, Monte Carlo (MC) method requires simulating a great number of particles to obtain reliable results, thus huge amount of computation time is the main problem of the MC method. Source biasing and weight window technique effectively decrease the tally error of deep-penetration problem. This paper studies variance reduction techniques (VRTs) based on adjoint fluence rate from a Discrete Ordinate (SN) code JSNT, generates source biasing factors and weight window parameters for a MC code JMCT, develops source sampling subroutine for JMCT, and verifies the VRT on measurements, finally applies the VRT to reactor pressure vessel fast neutron fluence rate and cavity neutron and photon dose rate calculation. Numerical results show that the VRT increases calculation efficiency by 1–2 orders for deep-penetration shielding calculation with high precision compared with unbiased MC method.

A 4H silicon carbide based fast-Neutron detection system with a neutron threshold of 0.4 MeV

Special requirements in pulsed fast-neutron detection of prompt fission and fusion devices call for a detection detector or system which could detect fast-neutrons in a severe radiation field, and thus should necessarily have a flat energy response, a high signal-noise-ratio, proper threshold energy, and a good radiation resistance. In this paper, a fast-neutron detection system based on a thin silicon carbide (SiC) detector and a 237Np fission target was developed and tested on Xi’an Pulsed Reactor, which was with a neutron threshold of 0.4 MeV, a flat energy response to neutrons in the range of 1MeV to 20MeV and got good response spectra to fission fragments. The SiC detector used here is a 4H-SiC Schottky diode detector with the dimension of 20mm×20mm×20μm, a dark current of lower than 5nA within 600V, a best energy resolution to alpha particles of about 1.78% and a high radiation resistance under alpha particle irradiation with a max incident number of 1.03×1010. The detection system shows advantaged prospects in fast-neutron fluence monitoring in nuclear fission and fusion equipments.

Measuring fluence of fast neutrons with planar silicon detectors

The results of measurements of 1-MeV (Si) equivalent fast neutron fluence with silicon planar detectors are reported. The measurement method is based on the linear dependence of the reverse detector current increment on the neutron fluence: ΔI = α I × Φ × V. This technique provides an opportunity to measure the equivalent fluence in a wide dynamic range from 108 to 1016 cm–2 with an unknown neutron energy spectrum and without detector calibration. The proposed method was used for monitoring in radiation resistance tests of different detector types at channel no. 3 of IBR-2 and for determining the fluence of fission and leakage neutrons at the KVINTA setup.

Thermal Predictions of the AGR-3/4 Experiment With Post Irradiation Examination Measured Time-Varying Gas Gaps

A thermal analysis was performed for the advanced gas reactor test experiment (AGR-3/4) with post irradiation examination (PIE) measured time (fast neutron fluence) varying gas gaps. The experiment was irradiated at the advanced test reactor (ATR) at the Idaho National Laboratory (INL). Several fuel irradiation experiments are planned for the AGR Fuel Development and Qualification Program, which supports the development of the very high-temperature gas-cooled reactor under the advanced reactor technologies project. The AGR-3/4 test was designed primarily to assess fission product transport through various graphite materials. Irradiation in the ATR started in December 2011 and finished in April 2014. Forty-eight (48) tristructural-isotropic-fueled compacts were inserted into 12 separate capsules for the experiment. The purpose of this analysis was to calculate the temperatures of each compact and graphite layer to obtain daily average temperatures using PIE-measured time (fast neutron fluence) varying gas gaps and compare with experimentally measured thermocouple (TC) data. PIE-measured experimental data were used for the graphite shrinkage versus fast neutron fluence. PIE dimensional measurements were taken on all the fuel compacts, graphite holders, and all of the graphite rings used. Heat rates were input from a detailed physics analysis for each day during the experiment. Individual heat rates for each nonfuel component were input as well. A steady-state thermal analysis was performed for each daily calculation. A finite element model was created for each capsule.

The effect of neutron irradiation on hydride reorientation and mechanical property degradation of zirconium alloy cladding

Abstract Zirconium alloy cladding tube specimens were irradiated at 380°C up to a fast neutron fluence of 7.5 × 10 24 n/m 2 in a research reactor to investigate the effect of neutron irradiation on hydride reorientation and mechanical property degradation. Cool-down tests from 400°C to 200°C under 150 MPa tensile hoop stress were performed. These tests indicate that the irradiated specimens generated a smaller radial hydride fraction than did the unirradiated specimens and that higher hydrogen content generated a smaller radial hydride fraction. The irradiated specimens of 500 ppm-H showed smaller ultimate tensile strength and plastic strain than those characteristics of the 250 ppm-H specimens. This mechanical property degradation caused by neutron irradiation can be explained by tensile hoop stress-induced microcrack formation on the hydrides in the irradiation-damaged matrix and subsequent microcrack propagation along the hydrides and/or through the matrix.

Effect of Neutron Irradiation on Epitaxial 4H-SiC PiN UV<sup></sup>-Photodiodes

4H-SiC UV-photodetectors based on full-epitaxial p +p-n+ multilayer structures werefabricated. The diodes were irradiated with fast neutrons up to the fluence of 1·1014 cm-2 . Current-voltage characteristics, life time of non-equilibrium charge carriers as well as photosensitivityspectra of the diodes before and after irradiation were investigated. The studies showed that PiNUV-photodiodes with base doping below 1·1015 cm-3 retain their performance up to the fluence of5·1012 neutrons per cm2 . The further increase in fast neutron fluence stimulates the creation ofexcessive deep recombination centers. This leads to degradation of I-V-characteristics, reduction ofcarrier life time and, consequently, to degradation of the photosensitivity of devices.

Development of indigenous insulation material for superconducting magnets and study of its characteristics under influence of intense neutron irradiation

Epoxy based glass fiber reinforced composites are the main insulation system for the superconducting magnets of fusion machines. 14MeV neutrons are generated during the DT fusion process, however the energy spectra and flux gets modified to a great extent when they reach the superconducting magnets. Mechanical properties of the GFRP insulation material is reported to degrade up to 30%. As a part of R & D activity, a joint collaboration with IGCAR, Kalpakkam has been established. The indigenous insulation material is subjected to fast neutron fluence of 1014 - 1019 n/m2 (E>0.1 MeV) in FBTR and KAMINI Reactor, India. TRIM software has been used to simulate similar kind of damage produced by neutrons by ion irradiation with 5 MeV Al ions and 3 MeV protons. Fluence of the ions was adjusted to get the same dpa. We present the test experiment of neutron irradiation of the composite material (E-glass, S-glass fiber boron free and DGEBA epoxy). The test results of tensile, inter laminar shear and electrical breakdown strength as per ASTM standards, assessment of micro-structure surface degradation before and after irradiation will be presented. MCNP simulations are carried out for neutron flux, dose and damages produced in the insulation material.

Evaluation of PWR pressure vessel fast neutron fluence benchmarks from NUREG/CR-6115 with ares transport code

An accurate evaluation of PWR pressure vessel fast neutron fluence is essential to ensure pressure vessel integrity over the design lifetime. The discrete ordinates method is one of the main methods to treat such problems. In this paper, evaluations have been performed for three PWR benchmarks described in NUREG/CR-6115 using ARES transport code. The calculated results were compared to the reference values and a satisfactory agreement was obtained. In addition, the effects of SN numeric and source distribution modeling for pressure vessel fast neutron fluence calculation are investigated. Based on the fine enough grids adopted, the different spatial and angular discretization introduces derivations less than 3 %, and fix-up for negative scattering source causes no noticeable effects when calculating pressure vessel fast neutron fluence. However, the discrepancy of assembly-wise and pin-wise source modeling for peripheral assemblies reaches ~20 %, which indicates that pin-wise modeling for peripheral assemblies is essential. These results provide guidelines for pressure vessel fast neutron fluence calculation and demonstrate that the ARES transport code is capable of performing neutron transport calculations for evaluating PWR pressure vessel fast neutron fluence.

Neutron-gamma flux and dose calculations for feasibility study of DISCOMS instrumentation in case of severe accident in a GEN 3 reactor

The present paper presents the study carried out in the frame of the DISCOMS project, which stands for “DIstributed Sensing for COrium Monitoring and Safety”. This study concerns the calculation of the neutron and gamma radiations received by the considered instrumentation during the normal reactor operation as well as in case of a severe accident for the EPR reactor, outside the reactor pressure vessel and in the containment basemat. This paper summarizes the methods and hypotheses used for the particle transport simulation outside the vessel during normal reactor operation. The results of the simulations are then presented including the responses for distributed Optical Fiber Sensors (OFS), such as the gamma dose and the fast neutron fluence, and for Self Powered Neutron Detectors (SPNDs), namely the neutron and gamma spectra. Same responses are also evaluated for severe accident situations in order to design the SPNDs being sensitive to the both types of received neutron-gamma radiation. By contrast, fibers, involved as transducers in distributed OFS have to resist to the total radiation gamma dose and neutron fluence received during normal operation and the severe accident.

The Activation Detector Activity Calculations Using the Effective Source Method and Measurement

In the paper the application of effective source to the solution of activation detector activities in the reactor pressure vessel cavity of the VVER-1000 reactor is presented. The effective source method applies the Boltzmann transport operator to time integrated source data to obtain detector activities. Weighting the source data by time dependent depletion of the detector activity, the result of the calculation is the detector activity. The approach works because of the inherent linearity of radiation transport in non-multiplying time-invariant media. Integrated in this way, the source data are referred to as the ‘effective source’. The effective source method thereby enables the analyst to replace numerous intensive transport calculations with a single transport calculation in which the time dependence and magnitude of the source are correctly represented. Detailed description of the effective source method is presented in previous works. First, there were performed neutron-physical calculations of few real VVER-1000 cycles using MOBY-DICK macrocode. Second, there follows 3-D transport calculation using the deterministic code TORT and the cross section library BUGLE-B7 and obtained results are presented. These calculation results of activation detector activities in the reactor cavity are compared with relevant activation detectors results of the ex-vessel measurement. The comparison between calculation and measurement of activation detectors activity in the reactor cavity is necessary to the calculation quality verifying for further fast neutron fluence onto the reactor pressure vessel credible calculation. The activation detectors positions are evident from Figs 1, 2, 3.

Comparison of two fast neutron fluence measurement methods based on Np-237 fission-to-capture ratio measurement (spectral index) and a reverse dark current measurement in a planar silicon detector

The idea of the first method is to search the neutron energy for the ratio of fission cross section to capture cross section of the selected actinide isotope from the nuclear data base that is equal to the measured ratio of the fissioned and captured actinide isotope Np-237.The idea of the second method consists in the measurement of the reverse dark current increase, which is linearly proportional to neutron fluence, induced by the fast neutron irradiation in planar silicon detectors.Np-237 samples and planar silicon detectors were placed inside a subcritical assembly (the Quinta assembly at the Joint Institute for Nuclear Research, Russia) very close to each other assuming that in both samples the same neutron fluence should pass.We concluded that minor actinide samples and planar silicon detectors can be used as neutron fluence detectors especially in the high neutron energy range, where measurements are difficult.Considering the importance of high energy neutron measurement in the ADS (Accelerator Driven System), actinide and silicon detectors could be a very useful tool.

Presence of neutrons in the low-level background environment estimated by the analysis of the 595.8 keV gamma peak

In order to explore possible improvements of the existing techniques developed to estimate the neutron fluence in low-background Ge-spectroscopy systems, gamma spectra were collected by a HPGe detector in the presence of the 252Cf spontaneous fission neutron source. The spectra were taken with and without a Cd envelope on the detector dipstick, with different thicknesses of plastic used to slow down neutrons. We have analyzed the complex 595.8keV gamma peak, as well as several more gamma peaks following the neutron interactions in the detector itself and surroundings materials. The investigation shows that some changes of the initial neutron spectra can be monitored by the analysis of the 595.8keV gamma peak. We have found good agreement in the intensity changes between the long-tail component of the 595.8keV and the 691keV gamma peak (72Ge(n,n′)72Ge reaction), usually used for the estimation of the fast neutron fluence. Results also suggest that the thermal neutrons can have a stronger influence on creation of the Gaussian-like part of 595.8keV peak, than on the 139keV one following 74Ge(n,γ)75mGe reaction and used in the standard methods (Škoro et al., 1992) [8] for determination of the thermal neutron flux.

Irradiation Campaign in the EOLE Critical Facility of Fiber Optic Bragg Gratings Dedicated to the Online Temperature Measurement in Zero Power Research Reactors

The control of temperature during operation of zero power research reactors participates to the overall control of experimentation conditions and reveals itself of a major importance when measuring neutronic parameters such as multiplication factors. Within the framework of the refurbishment of the MASURCA facility, the development of a new temperature measurement system based on the optical fiber Bragg grating (FBG) technology is under consideration. In a first step, a series of FBGs is irradiated in the EOLE critical facility with the aim to select the most appropriate. Online temperature measurements are performed during a set of irradiations that should allow reaching a fast neutron fluence of nearly $5\cdot 10^{14}$ n.cm $^{-2}$ and a total gamma dose lower than 4 kGy. The results obtained, especially the Bragg wavelength shifts during the irradiation campaign, are discussed in this paper and compared to data from standard PT100 temperature sensors to highlight possible radiation effects on sensor performances. Work to be conducted during the second step of the project, aiming to a feasibility demonstration using a MASURCA assembly, is also presented.

Polyethylene terephthalate degradation under reactor neutron irradiation

This paper is devoted to study the defects generated by reactor neutron in polyethylene terephthalate (PET) films. The explored fast neutron fluence ranges from 2.02×1016 to 2.07×1018ncm−2. The induced damages were investigated using ultraviolet-visible spectrophotometry (UV-vis), Fourier Transform Infrared spectrometry (FTIR) and X-ray diffraction (XRD). The UV-vis spectra show important changes indicating the degradation of the chemical structure and the creation of new chromophores. FTIR spectra reveal that the intensities of the different absorption bands decrease linearly under fast neutron irradiation. The internal reference band at 1410cm−1 is used to follow the overall damage during irradiation. The 1342cm−1 band corresponding to CH2 wagging of trans conformation of crystalline phase show a sharpe linear decrease as the fast neutrons fluence goes up. The creation of the monosubstituted benzene, investigated using the 1610cm−1 band. It shows a linear increase with fast neutron fluence. It is found from XRD analysis that the diffraction peak (100) intensity is drastically reduced after irradiation at 2.02×1016ncm−2.