Neutron radiation effects on metal oxide semiconductor (MOS) devices

The main purpose of this study is to provide the knowledge and data on Deuterium-Tritium (D-T) fusion neutron induced damage in MOS devices. Silicon metal oxide semiconductor (MOS) devices are currently the cornerstone of the modern microelectronics industry. However, when a MOS device is exposed to a flux of energetic radiation or particles, the resulting effects from this radiation can cause several degradation of the device performance and of its operating life. The part of MOS structure (metal oxide semiconductor) most sensitive to neutron radiation is the oxide insulating layer (SiO₂). When ionizing radiation passes through the oxide, the energy deposited creates electron-hole pairs. These electron-hole pairs have been seriously hazardous to the performance of these electronic components. The degradation of the current gain of the dual n-channel depletion mode MOS caused by neutron displacement defects, was measured using in situ method during neutron irradiation. The average degradation of the gain of the current is about 35 mA, and the change in channel current gain increased proportionally with neutron fluence. The total fusion neutron displacement damage was found to be 4.8 × 10⁻²¹ dpa per n/cm², while the average fraction of damage in the crystal of silicon was found to be 1.24 × 10⁻¹². All the MOS devices tested were found to be controllable after neutron irradiation and no permanent damage was caused by neutron fluence irradiation below 10¹⁰n/cm². The calculation results shows that (n,α) reaction induced soft-error cross-section about 8.7 × 10⁻¹⁴ cm², and for recoil atoms about 2.9 × 10⁻¹⁵ cm², respectively.     

Neutron induced damage in reactor pressure vessel steel: An X-ray absorption fine structure study

The radiation damage produced in reactor pressure vessel (RPV) steels during neutron irradiation is a long-standing problem of considerable practical interest. In this study, an extended X-ray absorption fine structure (EXAFS) spectroscopy has been applied at Cu, Ni and Mn K-edges to systematically investigate neutron induced radiation damage to the metal-site bcc structure of RPV steels, irradiated with neutrons in the fluence range from 0.85 to 5.0 × 10¹⁹ cm⁻². An overall similarity of Cu, Ni and Mn atomic environment in the iron matrix is observed. The radial distribution functions (RDFs), derived from EXAFS data have been found to evolve continuously as a function of neutron fluence describing the atomic-scale structural modifications in RPVs by neutron irradiations. From the pristine data, long range order beyond the first- and second-shell is apparent in the RDF spectra. In the irradiated specimens, all near-neighbour peaks are greatly reduced in magnitude, typical of damaged material. Prolonged annealing leads annihilation of point defects to give rise to an increase in the coordination numbers of near-neighbour atomic shells approaching values close to that of non-irradiated material, but does not suppress the formation of nano-sized Cu and/or Ni-rich-precipitates. Total amount of radiation damage under a given irradiation condition has been determined. The average structural parameters estimated from the EXAFS data are presented and discussed.     

Fluence and dose measurements for an accelerator neutron beam

The 3 MV Van de Graaff accelerator at McMaster University accelerator laboratory is extended to a neutron irradiation facility for low-dose bystander effects research. A long counter and an Anderson-Braun type neutron monitor have been used as monitors for the determination of the total fluence. Activation foils were used to determine the thermal neutron fluence rate (around 10⁶ neutrons s⁻¹). Meanwhile, the interactions of neutrons with the monitors have been simulated using a Monte Carlo N Particle (MCNP) code. Bystander effects, i.e. damage occurring in cells that were not traversed by radiation but were in the same radiation environment, have been well observed following both alpha and gamma irradiation of many cell lines. Since neutron radiation involves mixed field (including gamma and neutron radiations), we need to differentiate the doses for the bystander effects from the two radiations. A tissue equivalent proportional counter (TEPC) filled with propane based tissue equivalent gas simulating a 2 μm diameter tissue sphere has been investigated to estimate the neutron and gamma absorbed doses. A photon dose contamination of the neutron beam is less than 3%. The axial dose distribution follows the inverse square law and lateral and vertical dose distributions are relatively uniform over the irradiation area required by the biological study.     

Present and future trends for neutron source calibrations at the National Institute of Standards and Technology

The neutron source calibration facility operated by the United States National Institute of Standards and Technology (NIST) is a world-class calibration laboratory providing neutron source calibration services for radioisotopic sources with neutron emission rates ranging from 5 × 10⁵ to 1 × 10¹⁰ s⁻¹. Calibrations are performed using the manganous sulfate bath technique with a relative expanded uncertainty of approximately 3.5% (2σ). Recently, an improvement to the calibration procedure has been implemented whereby sources are regularly cross-calibrated against the national standard neutron source as well as one of three (international) standard sources formerly maintained by the Bureau International des Poids et Mesures. This feature helps ensure that the fidelity of NIST neutron source calibrations is maintained at the highest level. In addition to the Institute’s external customers, NIST’s neutron source calibration facility also provides important contributions to other neutron irradiation and calibration services provided by the Institute, as well as to NIST’s intramural research programs in neutron metrology, nuclear reactor pressure-vessel dosimetry, and fundamental neutron physics.     

Simulation of e–γ–n targets by FLUKA and measurement of neutron flux at various angles for accelerator based neutron source

A 6 MeV Race track Microtron (an electron accelerator) based pulsed neutron source has been designed specifically for the elemental analysis of short lived activation products where the low neutron flux requirement is desirable. The bremsstrahlung radiation emitted by impinging 6 MeV electron on the e–γ primary target, was made to fall on the γ–n secondary target to produce neutrons. The optimisation of bremsstrahlung and neutron producing target along with their spectra were estimated using FLUKA code. The measurement of neutron flux was carried out by activation of vanadium and the measured fluxes were 1.1878 × 10⁵, 0.9403 × 10⁵, 0.7428 × 10⁵, 0.6274 × 10⁵, 0.5659 × 10⁵, 0.5210 × 10⁵ n/cm²/s at 0°, 30°, 60°, 90°, 115°, 140° respectively. The results indicate that the neutron flux was found to be decreased as increase in the angle and in good agreement with the FLUKA simulation.     

In-Core SCC Growth Behavior of Type 304 Stainless Steel in BWR Simulated High-Temperature Water at JMTR

Irradiation-assisted stress corrosion cracking (IASCC) is one of the critical concerns when stainless steel components have been in service in light water reactors for a long period. In-core IASCC growth tests have been carried out using the compact tension-type specimens of type 304 stainless steel that had been pre-irradiated up to a neutron fluence level around 1 × 10²⁵ n/m² under a pure water simulated boiling water reactor (BWR) coolant condition at the Japan Materials Testing Reactor (JMTR). In order to investigate the effect of synergy of neutron/gamma radiation and stress/water environment on SCC growth rate, we performed ex-core IASCC tests on irradiated specimens at several dissolved oxygen contents under the same electrochemical potential condition. In this paper, results of the in-core SCC growth tests are discussed and compared with the results obtained by ex-core tests from a viewpoint of the synergistic effects on IASCC. From results of in-core and ex-core tests using pre-irradiated specimens, the effect of synergy of neutron/gamma radiation and stress/water environment on SCC growth rate was considered to be small, because the in-core data under the same ECP condition were similar to the ex-core data under the DO = 32 ppm condition.     

Measurement of thermal neutron cross-section and resonance integral for 186W(n,γ)187W reaction by the activation method using a single monitor

The thermal neutron cross-section (σ₀) and the resonance integral (I₀) of the reaction ¹⁸⁶W(n,γ)¹⁸⁷W were measured by the activation method using ⁵⁵Mn as a single comparator. The diluted MnO₂ and WO₃ samples within and without a cylindrical Cd shield case were irradiated in an isotropic neutron field of the ²⁴¹Am–Be neutron source. The γ-ray spectra from the irradiated samples were measured by high resolution γ-ray spectrometry with a calibrated high purity Ge detector. The necessary correction factors for gamma ray attenuation, thermal and resonance neutron self-shielding effects, and the shape factor (α) for epithermal neutron spectrum were taken into account in the determinations. The thermal neutron cross-section for ¹⁸⁶W(n,γ)¹⁸⁷W reaction has been determined to be 39.5±2.3 b at 0.025 eV. This result has been obtained relative to the reference thermal neutron cross-section value of 13.3±0.1 b for the ⁵⁵Mn(n,γ)⁵⁶Mn reaction. The present value of 39.5±2.3 b for ¹⁸⁶W(n,γ)¹⁸⁷W, in general is in good agreement with most of experimental data and evaluated ones in ENDF/B-VI and JENDL-3.2 within the limits of error. The resonance integral has also been measured relative to the reference value of 14.0±0.3 b for the ⁵⁵Mn(n,γ)⁵⁶Mn monitor reaction using a 1/E^1+α epithermal neutron spectrum of the ²⁴¹Am–Be neutron source. By defining Cd cut-off energy 0.55 eV, the resonance integral obtained was 493±40 b. The existing experimental and evaluated data for the resonance integral are distributed from 290 to 534 b. The present resonance integral value agrees with some previously reported values.     

Development of an asymmetric multiple-position neutron source (AMPNS) method to monitor the criticality of a degraded reactor core

Tne analytical/experimental method has been developed to to monitor the subcritical reactivity and unfold the k_∞ distribution of a degraded reactor core. The method uses several fixed neutron detectors and a ²⁵²Cfneutron source placed sequentially in multiple positions in the core. Therefore, it is called the asymmetric multiple-position neutron source (AMPNS) method. The AMPNS method employs the nucleonic codes to analyze in two dimensions the neutron multiplication of a ²⁵²Cf neutron source. An optimization program, GPM, was utilized to unfold the k_∞ distribution of the degraded core, in which the desired performance measure minimizes the error between the calculated and the measured count rates of the degraded reactor core. The analytical/experimental approach is validated by performing experiments using the Penn. State Breazeale TRIGA reactor (PSBR). A significant result of this study has been to provide a means to plan the source and detector placements and assign core cells to the damaged TMI-2 core as well as to monitor the criticality during the recovery period.     

Improved Los Alamos model applied to the neutron induced fission of 239Pu and 240Pu and to the spontaneous fission of Pu isotopes

The Los Alamos model with multiple fission chances upgraded with (a) the linear relation between the average prompt gamma ray energy and the average prompt neutron multiplicity and (b) the dependence of the average fission fragment kinetic energy on the incident neutron energy, is used for the n+²³⁹Pu and n+²⁴⁰Pu reactions, and also for the spontaneous fission of ^237–241Pu isotopes. In the case of ²⁴⁰Pu fissioning nucleus the variation of the average energy released versus the incident neutron energy is also taken into account. The calculated prompt fission neutron spectra and average prompt neutron multiplicity well represent the experimental data, proving a better predictive power of the improved Los Alamos model.     

Trace elements in reactor steels: implications for decommissioning

Trace elements in stainless steel have been systematically examined for the production of long-lived radioisotopes through neutron activation in reactors. Niobium-94 has been identified as the most important impurity. It is a long-lived ( yr) gamma ray emitter (0.7 and 0.87 MeV), which is produced by the neutron capture reaction ⁹³Nb(n, γ)⁹⁴Nb. Through X-ray fluorescence Nb concentrations of 160 ± 20 ppm have been found in type 304 stainless steel which agrees with earlier published values. At this concentration, the gamma radiation dose rate inside the pressure vessel of a reactor would remain close to 1 rem/hr for thousands of years after the ⁶⁰Co activity has decayed. This could be important for the delayed dismantling option considered for reactors.Nitrogen as an impurity in stainless steel has been shown to result in the buildup of 100 and 1000 Ci of carbon-14 over the lifetime of a BWR and a PWR, respectively. Although ¹⁴C is only a β⁻-emitter, its long half-life (5730 yr) and the crucial role of carbon in the biosphere may be important in deciding on the ultimate disposal method of the radioactive reactor components.     

Radiation-induced phenomena related to electrical properties of hydrogen-doped strontium-cerium-ytterbium oxides under reactor irradiation

The effects of radiation on the electrical properties of hydrogen-doped (H-doped) strontium-cerium-ytterbium oxide (SrCe_0.95Yb_0.05O_3−δ), a perovskite-type ceramic, were investigated by irradiating specimens with thermal and fast neutrons and gamma rays in a fission reactor. The electrical conductivities of the H-doped SrCe_0.95Yb_0.05O_3−δ, which were measured at thermal and fast neutron fluxes of 4.1 × 10¹⁷ and 2.7 × 10¹⁶ n/m²s and an ionizing dose rate of 0.5 kGy/s, were approximately two orders of magnitude higher than the base conductivity in the absence of radiation and slightly higher compared to those of the non-doped SrCe_0.95Yb_0.05O_3−δ. The radiation-induced phenomena on the electrical properties can allow radiation-enhanced diffusion of H as well as electronic excitation, which is caused by ionization effects. It was observed that the radiation-enhanced diffusion of H significantly depended on the irradiation temperatures in the range 384-519 K, whereas it was not affected by radiation-induced defects produced with a fast neutron fluence of approximately 1.3 × 10²³ n/m² under the present experimental conditions.     

Studies on new neutron-sensitive dosimeters using an optically stimulated luminescence technique

The neutron response of detectors prepared using α-Al₂O₃:C phosphor developed using a melt processing technique and mixed with neutron converters was studied in monoenergetic neutron fields. The detector pellets were arranged in two different pairs: α-Al₂O₃:C + ⁶LiF/α-Al₂O₃:C + ⁷LiF and α-Al₂O₃:C + high-density polyethylene/α-Al₂O₃:C + Teflon, for neutron dosimetry using albedo and recoil proton techniques. The optically stimulated luminescence response of the Al₂O₃:C + ^6,7LiF dosimeter to radiation from a ²⁵²Cf source was 0.21, in terms of personal dose equivalent H_p(10) and relative to radiation from a ¹³⁷Cs source. This was comparable to results obtained with similar detectors prepared using commercially available α-Al₂O₃:C phosphor. The H_p(10) response of the α-Al₂O₃:C + ^6,7LiF dosimeters was found to decrease by more than two orders of magnitude with increasing neutron energy, as expected for albedo dosimeters. The response of the α-Al₂O₃:C + high-density polyethylene/α-Al₂O₃:C + Teflon dosimeters was small, of the order of 1% to 2% in terms of Hp(10) and relative to radiation from a ¹³⁷Cs source, for neutron energies greater than 1 MeV.     

Contribution of thermal neutrons to radiation hardening of pure copper

The paper presents the results of an experiment the aim of which was to estimate directly the effect of the thermal neutron fluence on pure copper hardening. Identical specimens were irradiated in two reactors (SM-2 and RBT-6) in the dose range 10⁻³-10⁻¹ dpa at T_irr=80 °C under substantially different, by a factor of 5, thermal neutron fluences, with other irradiation parameters being close. The results show that the elevated thermal fluence in the SM-2 reactor increases the radiation hardening of pure copper by 50% at a dose of about 10⁻³ dpa as compared with specimens irradiated in the RBT-6 reactor. The contribution of thermal neutrons proved to be much more considerable than the theoretical estimates.     

Uncertainty evaluation of calculated and measured kinetics parameters of Tehran Research Reactor

Effective delayed neutron fraction β_eff and neutron generation time Λ are important factors in reactor physics calculation and transient analysis. In the first stage of this research, these kinetics parameters have been calculated for two states of Tehran Research Reactor (TRR), i.e. cold (fuel, clad and coolant temperature 20 °C) and hot (fuel, clad and coolant temperature 65, 49 and 44 °C, respectively) states using MTR_PC computer code. The ratio of (β_eff)_i/(β_eff)_core plays an important role in reactivity accident analysis codes. This parameter and its contribution to effective delayed neutron fraction from each nucleus have been calculated in cold and hot reactor states. Uncertainty of effective delayed neutron fraction is evaluated in terms of following four quantities; basic delayed neutron constants, delayed neutron spectra, energy dependence of delayed neutron yield (ν_d) and fission cross-section of ²³⁵U and ²³⁸U. In the second stage, these parameters have been measured with an experimental method based on Inhour equation. The calculated and measured values are in good agreement. Relative Percent Errors (RPEs) are 2.8% for β_eff and 5.7% for Λ in the cold state.     

Minimization of an initial fast reactor uranium-plutonium load by using enriched lead-208 as a coolant

Long-term scenarios of nuclear energy evolution over the world scale predict deployment of fast reactors (FRs) from 2020 to 2030 and achievement on 2050 the world installed capacity equal to 1500 GW_e with essential increasing the FRs number. For several countries (i.e. Russia, Japan) whose policies are based on a sharp increase of nuclear production, at the stage near 2030-2040 when plutonium, Pu, from the PWR spent nuclear fuel is consumed, the Pu lack will stimulate minimization of its load in FRs. The period of Pu deficiency will be prolonged till the years when breeding gain (BG) equal to 0.2-0.3 in fast breeding reactors (FBRs) is obtained which corresponds to Pu inventory doubling time of 44-24 years.In this paper one of opportunities to minimize fuel loading is considered: it is related to using a low neutron capturing lead isotope, ²⁰⁸Pb, as a FR coolant. It is known, that natural lead, ^natPb, contains a stable lead isotope, ²⁰⁸Pb, having a small cross-section of neutron capture via (n, γ) reaction. In the paper it is shown that the macroscopic cross-sections 〈σ_n,γ〉 of radiation neutron capture by the lead isotope ²⁰⁸Pb averaged on the ADS core neutron spectra are by ∼3.7-4.5 times less than the corresponding macroscopic cross-sections for a natural mix of lead isotopes ^natPb. This circumstance allows minimizing load of a lead fast reactor (LFR) core for achievement its criticality, as well as the load of an accelerator-driven system (ADS) subcritical core—for achievement of its small subcriticality. In using ²⁰⁸Pb instead of ^natPb in the ADS blanket, the multiplication factor of the subsritical core, K_eff, could be increased from the initial value K_eff = 0.953 up to the value of K_eff = 0.970. To achieve this higher value of K_eff in the same core cooled by ^natPb an additional amount of 20-30% of U-Pu fuel will be needed.The isotope ²⁰⁸Pb content in the natural mix of isotopes, ^natPb, is high enough, above 52%, and its separation in large amounts (several tens’ and hundreds’ of tonnes) is expensive but really solvable technical task. In the project (ISTC #2573, 2005), developed with authors’ participation, it is shown that a new laser photochemical technique of lead isotope separation, being developed in future, permits to obtain large quantities of ²⁰⁸Pb under its acceptable price, of close to $200 kg⁻¹.     

Measurement of thermal neutron cross-section and resonance integral for the Ho(n,γ)Ho reaction using electron linac-based neutron source

The thermal neutron cross-section and the resonance integral of the ¹⁶⁵Ho(n,γ)^166gHo reaction have been measured by the activation method using a ¹⁹⁷Au(n,γ)¹⁹⁸Au monitor reaction as a single comparator. The high-purity natural Ho and Au foils with and without a cadmium shield case of 0.5 mm thickness were irradiated in a neutron field of the Pohang neutron facility. The induced activities in the activated foils were measured with a calibrated p-type high-purity Ge detector. The correction factors for the γ-ray attenuation (F_g), the thermal neutron self-shielding (G_th), the resonance neutron self-shielding (G_epi) effects, and the epithermal neutron spectrum shape factor (α) were taken into account. The thermal neutron cross-section for the ¹⁶⁵Ho(n,γ)^166gHo reaction has been determined to be 59.7 ± 2.5 barn, relative to the reference value of 98.65 ± 0.09 barn for the ¹⁹⁷Au(n,γ)¹⁹⁸Au reaction. By assuming the cadmium cut-off energy of 0.55 eV, the resonance integral for the ¹⁶⁵Ho(n,γ)^166gHo reaction is 671 ± 47 barn, which is determined relative to the reference value of 1550 ± 28 barn for the ¹⁹⁷Au(n,γ)¹⁹⁸Au reaction. The present results are, in general, good agreement with most of the previously reported data within uncertainty limits.     

Measurement of thermal neutron cross section and resonance integral for (n,γ) reaction in Sm

This study implies that ⁵⁵Mn(n,γ)⁵⁵Mn monitor reaction may be a convenient alternative comparator for the activation method and thus, it was used for the determination of thermal neutron cross section (TNX) and the resonance integral (RI) of the reaction ¹⁵²Sm(n,γ)¹⁵³Sm. The samples of MnO₂ and Sm₂O₃ diluted with Al₂O₃ powder were irradiated within and without a cylindrical 1 mm-Cd shield case in an isotropic neutron field obtained from the ²⁴¹Am–Be neutron sources. The γ-ray spectra from the irradiated samples were measured by high resolution γ-ray spectrometry with a calibrated n-type Ge detector. The correction factors for γ-ray attenuation, thermal neutron and resonance neutron self-shielding effects and epithermal neutron spectrum shape factor (α) were taken into account in the determinations. The thermal neutron cross section for ¹⁵²Sm(n,γ)¹⁵³Sm reaction has been determined to be 204.8 ± 7.9 b at 0.025 eV. This result has been obtained relative to the reference thermal neutron cross section value of 13.3 ± 0.1 b for the ⁵⁵Mn(n,γ)⁵⁶Mn reaction. For the TNX, most of the experimental data and evaluated one in JEFF-3.1, ENDF/B-VI, JENDL 3.3 and BROND 2.0, in general, agree well with the present result. The RI value for ¹⁵²Sm(n,γ)¹⁵³Sm reaction has also been determined to be 3038 ± 214 b, relative to the reference value of 14.0 ± 0.3 b for the ⁵⁵Mn(n,γ)⁵⁶Mn monitor reaction, using a 1/E^1+α epithermal neutron spectrum and assuming Cd cut-off energy of 0.55 eV. In surveying literature, the existing experimental and evaluated data for the RI values are distributed from 1715 to 3462 b. However, when the Cd cut-off energy is defined as 0.55 eV, the present RI value agrees with some previously reported RI values, 3020 ± 163 b by Simonits et al., 3141 ± 157 by Van Der Linden et al., and 2962 ± 54 b by Kafala et al., within the limits of error.     

Measurement of the 239Pu(n,f) cross section from thermal up to 30 keV neutron energy

At GELINA measurements of the ²³⁹Pu fission cross-section were performed covering the neutron energy region from thermal up to 30 keV. Fission fragment as well as fission neutron detection techniques were used. Also for the neutron flux determination different methods were applied. From the σ_f-data, several fission integrals were calculated and compared with other results.     

Alpha Emission Spectra of 27Al, 50,52Cr, 55Mn, 54,56Fe, 58,60Ni Nucleus for Neutron Induced Reaction

The design of novel nuclear facilities, fusion as well as fission reactors, requires the knowledge of all properties of relevant materials, including the nuclear differential cross sections for a careful selection. The nuclear cross sections data for gas production via particle (neutron, proton, alpha, etc.) induced reactions are great importance in the domain in the fusion reactor technology, particularly in the calculation of nuclear transmutation rates, nuclear heating and radiation damage due to gas formation. In fusion reactor structures, a serious damage mechanism has been gas production in the metallic resulting from diverse nuclear reactions, mainly through (n, p) and (n, α), (n, d), (n, t). In the present study, by using equilibrium reaction mechanisms, the (n, xα) reaction alpha emission spectra for ²⁷Al, ^50,52Cr, ⁵⁵Mn, ^54,56Fe, ^58,60Ni isotopes were investigated from 9 to 15 MeV incident neutron energy. The equilibrium results have been calculated by using the hybrid model, the geometry dependent hybrid model. Calculation results have been also compared with the available measurements in literature.