Displacement Per Atom Research Articles

Calculation and analysis of DPA in the main components of CSNS target station

The damage energy cross sections and displacement cross sections of W, SS316 and Al-6061, which will be used in the CSNS (China Spalllation Neutron Source) target station, were calculated with the high energy transport code MCNPX 2.5.0. The DPA (Displacement Per Atom) distributions were also calculated and anlyzed for the main components in the CSNS target station, such as the W target, the target vessel, the moderators and reflector vessels. Based on the DPA caused by neutrons and protons, the lifetime of these components were estimated. The influence of the proton beam profile on the peak value of DPA for W target and target vessel were also presented. These calculations and analyses are very important for the under going CSNS project.

Method for the Calculation of DPA in the Reactor Pressure Vessel of Atucha II

One of the limiting factors of the life of a nuclear power plant (NPP) is the state of the reactor pressure vessel (RPV). Embrittlement is the most important effect affecting RPV aging. The irradiation with neutrons, especially fast neutrons, is the primary cause of this embrittlement. NPP safe operation requires to ensure RPV integrity over its lifetime, threatened by the neutron radiation‐induced embrittlement. In this paper, we identify the areas where the RPV neutron radiation is maximum and perform calculations of the displacement‐per‐atom (DPA) rate in those areas using the MCNP5 code. It was determined that the maximum DPA rate in the RPV wall with fresh fuel element (FE) is 3.76(3) × 10-12 s-1, it takes place in front of FEs BA42 and BL43, and it is symmetrical about the central channel, LG04, and LH03.

Neutron Dosimetry and 3-D Neutron Transport Calculations as Toolkit at HFR Petten for Detailed Neutron Monitoring and Damage Analysis

Dosimetry at the High Flux Reactor in Petten, the Netherlands, serves a variety of purposes ranging from demonstrating compliance with the license for operating the reactor to validating the predictions of displacements per atom (DPA) in material irradiations. It is also used in the validation of the 3-D neutron transport calculations during post-irradiation analysis to normalize the spectrum obtained from these calculations. In this paper, a thorough examination of irradiated steel specimens is carried out using the measurements performed with the neutron activation technique as well as the 3-D neutron transport calculations to check the consistency in the resulting DPA values.

Measurement and Evaluation of the Irradiation Test Parameters for a Specimen in a HANARO Material Irradiation Capsule

A material capsule system has been developed for irradiation tests of nuclear materials in the core region of the high-flux advanced neutron application reactor (HANARO) and has been actively utilized for the various material irradiation tests requested by numerous users. For an irradiation test in the reactor, nuclear characteristics and irradiation temperature of the capsule parts are essential for the safety analysis of the capsule and for the detailed capsule design. The irradiation temperature and neutron fluence of the specimens are key parameters required by a user to analyze any neutron irradiation damage effects. Based on a basic design of a capsule, the reactivity effect, neutron flux/fluence, and gamma heating of specimens are calculated by the MCNP code. To compare the calculated neutron fluence of the specimens, thermal and fast fluence monitors (F/Ms) are installed in a capsule. After an irradiation, the F/Ms are dismantled and analyzed to obtain a neutron spectrum. The experimental fast neutron fluences (E > 1.0nMeV) derived by using the SAND II code were found to lie within about 20% of the predicted values calculated by the MCNP code. The irradiation temperature of the specimens is analyzed preliminary by using the GENGTC or ANSYS codes based on the basic design of a capsule and gamma heating. During an irradiation test, the temperatures of the specimens in a capsule are measured and monitored by thermocouples. The displacement per atom (DPA) and activation of the irradiated specimens are also evaluated by using the SPECTER and ORIGEN2 codes, respectively, at HANARO.

Radiation damage/activity calculation for CSNS target station

The radiation damages have been performed for Chinese spallation neutron source (CSNS) target center components that relies on Monte Carlo simulation code MCNPX. During the calculation, Bertini intranuclear cascade model, three level-density formulation GCCI, and multistage pre-equilibrium model MPM on which are provided within MCNPX are employed. We calculate the displacement per atom (DPA) and afterheat of the tungsten target, the stainless steel target vessel window and the aluminum alloy moderator vessel. As a hundred kW-level source, these spallation center components have the lifetime more than 5 year. We also give the activity for the T0 chopper of the beam line HIPD to get the primary data for making out a maintenance scenario.

Facile synthesis of highly stablea-Siby ion implantation of low-keV H isotopes

It is experimentally shown that silicon is 'easily' amorphized by low-keV H ions at the relatively high temperature of 150 K and for an ion fluence equivalent to <1 DPA (displacement per atom). The a-Si layer is much more stable against recrystallization than a-Si produced by other ions and more stable against chemical modification than c-Si that is H-implanted at room temperature. These results are unexplained by the current atomic collision theory, including molecular-dynamics simulations, but they demonstrate the stabilizing effect of dangling bond passivation by H atoms in postulated, metastable, amorphous droplets.

Radiation effect on the photoluminescence properties of Si/SiO2 thin films

Silicon ions were implanted into SiO2 thin films with various doses and energies. For the films implanted with various ion doses the photoluminescence (PL) intensity of 470nm firstly increased with the increase of Si ion dose, which is similar to the variation trend of displacement per atom (DPA) number during ion radiation. Further increasing Si ion dose the PL intensity of 470nm decreased gradually since the neutral oxygen vacancy centers were destroyed. For the samples implanted with different energy the variation trend of PL intensity for 470nm peak is similar to the result of DPA under different radiation energy according to SRIM2006 simulation. With the increase of radiation energy a new PL peak at 550nm appeared because of the variation of defect type. Combining with the simulation results and PL spectra the radiation effect on Si/SiO2 thin films were proposed.

Monte Carlo calculation for various enrichment lithium coolant using different data libraries in a hybrid reactor

The main objective of this study is to compare the effect of the natural lithium and lithium with different enrichments between 10% and 90% on neutronic parameters; such as tritium breeding ratio (TBR), displacement per atom (DPA) and gas production using the different data libraries (ENDF/B.V, ENDF/B.VI and CLAW IV). Therefore, the natural lithium and different enrichment lithium were used as a moderator in an experimental hybrid reactor for the calculation of the nuclear parameters. Neutronic calculations were performed by recent Monte Carlo Neutron-Particle Transport code MCNP5 version 1.40 for a 14.1 MeV (D, T) fusion driver under a neutron wall load of 2.25 MW/m 2 (10 14 n/s). TBR values in the blanket for all investigated cases were obtained greater than the minimum requirement (TBR > 1.05). Considering radiation damage limits (100 DPA and 500 appm/FPY) for structural materials, the FW replacement will be needed every 2.1 and 3.5 years for DPA and He-production, respectively.

Monte Carlo calculation of radiation damage in first wall of an experimental hybrid reactor

This work was focused on the neutronic calculation of the nuclear parameters (neutron spectrum, displacement per atom (DPA), gas production, tritium breeding ratio (TBR), nuclear heating) for structural materials in the first wall (FW) and fuel clad (made of ferritic/martensitic steels, vanadium alloy, silicon carbide, copper alloy, and stainless steel) of an experimental hybrid reactor using the most current Monte Carlo Neutron-Particle Transport code MCNP5 1.4. Neutronic calculations were performed using a (DT) fusion driver hybrid reactor under a neutron wall loud of 2.25 MW/m 2 by full reactor power for one year. Obtained results were compared with three different data libraries (ENDF/B-V, ENDF/B-VI and CLAW-IV). TBR values in the reactor blanket for all investigated materials became greater than the minimum requirement (TBR > 1.05). Nuclear parameters like DPA, He-production and nuclear heating were considered as radiation damage limits for structural materials, copper alloy (Cu0.5Cr0.3Zr) showed better performance than all investigated materials.

Feasibility Study on a Simple Method of Retrospective Neutron Dosimetry for Reactor Internals and Reactor Vessel

Abstract Nondestructive and simple retrospective dosimetry method was developed. Mercuric iodide or cadmium zinc telluride detector was used as a detector to measure 54Mn, 58Co, and 60Co that were created in reactor internals and vessels. Detector is surrounded by tungsten collimator which shield background gamma rays and detect gamma rays from desired measuring point. Photopeak count rate of this detector were calculated using the mcnp-4c code. This response was calculated for each reactor internal geometry that was irradiated by unit neutron flux which energy spectrum was calculated by the tort code. Neutron fluence is calculated using precalculated response, measuring time, decay time and reactor power history. We can also get Displacement per Atom (DPA), hydrogen and helium production simultaneously using neutron spectrum precalculated in reactor vessel. Because the time and cost of this method are much less than the sampling method, we can measure at many positions and even get a contour map.

Effects of ion irradiation on cobalt nanocomposite

The effects of ion irradiation of 3-dimensional arrays of Co nanoparticles were investigated. Arrays were obtained by electron beam deposition of 15 Co/SiO2 bilayers 0.5 and 20nm thick, respectively. The Co layers consist of Co nanoparticles 3.2nm in diameter with a standard deviation of 16%. Irradiation was carried out using combinations of 150kV Ar2+ and 90kV Ar+ ion bombardments with a Ar+:Ar2+ fluence ratio of 1:4. The effects of ion irradiation were followed by Rutherford backscattering, measurements of hysteresis loops at 5K and of temperature-dependent field-cooled (FC) and zero-field-cooled (ZFC) measurements of the magnetic susceptibility. A decrease of the peak temperature in the ZFC curve for displacements per atom (DPA) up to 1.1 was observed. Irradiation also induces progressively lower coercivity values. The Co particles showed a remarkable high resistance to ion irradiation, surviving damage up to 33DPA.

Dependence of the structural relaxation of amorphous silicon on implantation temperature

The structural relaxation of amorphous silicon, created by ion implantation, was investigated by in situ differential scanning nanocalorimetry. Nanocalorimetry provided the possibility to measure the heat released by relaxation during annealing, for a wide range of implantation fluences and beginning at cryogenic temperatures. Ion implantation was first carried out for fluences between ${10}^{\ensuremath{-}5}$ and 0.8 displacements per atom (DPA) at 133 K and 297 K, and then for temperatures ranging from 118 K to 463 K for fluences of 0.0185 and 0.37 DPA. A heat release saturation occurred above 0.1 DPA, and was found to depend on implantation temperature. The saturation level was extrapolated to 0 K, leading to an estimate of $28\ifmmode\pm\else\textpm\fi{}3\phantom{\rule{0.3em}{0ex}}\mathrm{kJ}∕\mathrm{mol}$ for the maximum enthalpy that can be stored in $a\text{\ensuremath{-}}\mathrm{Si}$, relative to crystalline Si.

The effects of ion irradiation on porous silicon photoluminescence

The effects of ion irradiation on porous Si (po-Si) photoluminescence (PL) are investigated to gain insight into the emission mechanism. Po-Si was obtained by standard electrochemical etching of p-doped Si(100) and the samples were aged for several weeks to achieve stability of the PL intensity. Specimens were progressively irradiated with H+, He+, or Ne++ ions followed by PL measurements. PL spectra were obtained as a function of the displacement per atom (DPA) parameter up to a DPA level of ∼10−2, which was sufficient to nearly extinguish PL. The quenching behavior showed a strong dependence on the chemical nature of the implanted species. Within experimental error, the quenching efficiency was equivalent for He+ and Ne++ irradiations, but was considerably higher for H+ irradiation. Channeling spectrometry showed the efficiency to be correlated with the self-recovery of defects generated during irradiation. The observed PL quenching is associated with the creation of nonradiative recombination sites within the band gap, but not with Si nanostructure amorphization and∕or surface oxygen removal. Significant recovery of quenched PL occurs over a period of 120days for specimens that are stored in air, but not for samples that are stored in vacuum. These results demonstrate the importance and complexity of surface oxidation on the po-Si luminescence mechanism.

Displacement damage in silicon carbide irradiated in fission reactors

Calculations are performed for displacement damage in SiC due to irradiation in the neutron environments of various types of nuclear reactors using the best available models and nuclear data. The displacement damage calculations use recently developed damage functions for SiC that are based on extensive molecular dynamics simulations of displacement events. Displacements per atom (DPA) cross sections for SiC have been calculated as a function of neutron energy, and they are presented here in tabular form to facilitate their use as the standard measure of displacement damage for irradiated SiC. DPA cross sections averaged over the neutron energy spectrum are calculated for neutron spectra in the cores of typical commercial reactors and in the test sample irradiation regions of several materials test reactors used in both past and present irradiation testing. Particular attention is focused on a next-generation high-temperature gas-cooled pebble bed reactor, for which the high-temperature properties of silicon carbide fiber-reinforced silicon carbide composites are well suited. Calculated transmutations and activation levels in a pebble bed reactor are compared to those in other reactors.

Radiation Effects on the Long-term Ageing of Spent Fuel Storage Containers

The properties of a long-term interim storage low-carbon steel container could be changed by radiation-induced damage. In this paper, the gamma and neutron flux in the container are estimated over 300 years for pressurised water reactor spent fuels. Gamma radiation is the main contributor to the damage flux during the whole storage period (from 3 × 10−8 dpa y−1 after 2 years of storage to 4 ×10−11 dpa y−1 after 300 years, where dpa is displacements peratom). A cluster dynamics model is used to calculate the point defect concentrations and their effect on the copper precipitation kinetic and point defect clustering for different storage conditions. The shear stress increases in the first years of storage, from 23 to 50 MPa for a Cu content varying from 0.05 to 0.25%. The temperature or the damage rate has an effect on the kinetic only at the beginning of the storage. The hardening is entirely caused by copper precipitation enhanced by vacancy supersaturation.

Radiation Damage Study on Various Structural Refractory Alloys of a Multi-Purpose Reactor

Radiation damage properties of structural materials play a key role in design of a fusion–fission (hybrid) reactor. Refractory alloys offer a significant advantage of high neutron wall load capability under fusion neutron environment. In this study, main radiation damage parameters (displacement per atom (DPA) and helium production) on three different refractory alloys, namely W-5Re, TZM (Mo alloy) and Nb–1Zr used as structural material in a hybrid reactor were found. Neutron transport calculations were conducted with the aid of SCALE4.3 System by solving the Boltzmann transport equation with code XSDRNPM. The lowest radiation damage values were obtained for W-5Re alloy. Moreover, all investigated materials will require to be replaced frequently due to their radiation damage values during reactor life (~ 30 years).

Displacement damage cross sections for neutron-irradiated silicon carbide

Displacements per atom (DPA) is a widely used damage unit for displacement damage in nuclear materials. Calculating the DPA for SiC irradiated in a particular facility requires a knowledge of the neutron spectrum as well as specific information about displacement damage in that material. In recent years significant improvements in displacement damage information for SiC have been generated, especially the energy required to displace an atom in an irradiation event and the models used to describe electronic and nuclear stopping. Using this information, numerical solutions for the displacement functions in SiC have been determined from coupled integro-differential equations for displacements in polyatomic materials and applied in calculations of spectral-averaged displacement cross sections for SiC. This procedure has been used to generate spectrally averaged displacement cross sections for SiC in a number of reactors used for radiation damage testing of fusion materials, as well as the ARIES-IV conceptual fusion device.

Neutronic and photonic analysis of the water-cooled Pb17Li test blanket module for ITER-FEAT

Within the European Fusion Technology Program, the Water-Cooled Lithium Lead (WCLL) DEMO breeding blanket line was selected in 1995 as one of the two EU lines to be developed in the next decade, in particular with the aim of manufacturing a Test Blanket Module (TBM) to be implemented in ITER. This specific goal has been maintained also in ITER-FEAT program even if the general design parameters of the TBMs have reported some changes. This paper is focused on the investigation of the WCLL-TBM nuclear response in ITER-FEAT through detailed 3D-Monte Carlo neutronic and photonic analyses. A 3D heterogeneous model of the most recent design of the WCLL-TBM has been set-up simulating realistically its new lay out and taking into account 9% Cr martensitic steel as structural material. It has been inserted into an existing 3D semi-heterogeneous ITER-FEAT model accounting for a proper D–T neutron source. The analyses have been performed by means of MCNP-4C code running on a cluster of four workstations through the implementation of a parallel virtual machine. The main WCLL-TBM nuclear responses have been determined focusing the attention on power deposition density, material damage through displacement per atom (DPA) and He and H production rate, daily tritium production and tritium production rate radial distribution in the module. Moreover, the impact of using lithium at various Li 6 enrichment on the TBM nuclear response has been investigated. The results obtained are herewith presented and critically discussed.

A Preliminary Shielding Design of 150MWe Liquid Metal Reactor

A preliminary shielding design of the KALIMER (Korea Advanced Liquid Metal Reactor) was established by two-dimensional discrete ordinate radiation transport analyses. The shielding design has tentatively adopted to use two limits on the fast neutron fluence and the DPA (Displacement Per Atom) simultaneously as a base for the exposure limit for neutron damages to reactor structures. In addition, activities in the PSDRS air effluent and the IHX secondary sodium were also examined.The DORT two-dimensional transport code was used to evaluate the KALIMER shielding design. The reactor system was represented by four axial zones, each of which was modeled in the R-Z geometry. The KAFAX-F22 library was used in the analyses, which was generated from the JEF-2.2 of OECD/NEA for LMR applications by KAERI (Korea Atomic Energy Research Institute), and consists of 80 neutron and 24 gamma energy groups.The performance of the shielding design is compared against the shielding design criteria. The results indicate that the support barrel, upper grid plate, and other reactor structures meet both the maximum neutron fluence and DPA limits established in the shielding design criteria. Activities of the air effluent in the PSDRS were also evaluated and are shown to satisfy the effluent concentration limits in 10 CFR Part 20.It is found that maximum DPAs show larger margins than the case when the neutron fluence limit is used as a design criterion. Therefore, the use of DPA as a shielding design criterion allows for a more flexibility in the LMR shielding design.

Radiation Damage in Liquid-Protected First-Wall Materials for IFE-Reactors

Material damage through displacements per atom (DPA) and helium gas production, as well as the tritium breeding and energy absorption in an IFE (Inertial Fusion Energy) reactor chamber have been in...