The energy dependence of neutron damage in Cu and Nb

Primary recoil spectra and specific damage energies have been computed for neutron interactions in Cu and Nb at neutron energies up to 32 MeV. The calculations are based on theoretical neutron cross sections and are in good agreement with recent radiation damage experiments using high energy neutrons from the Be(d,n) reaction. The results are particularly relevant to the understanding of radiation effects from high energy deuteron-breakup neutron sources.     

基于D-T中子源的板状铌样品的评价数据基准检验实验

利用中国原子能科学研究院核数据国家重点实验室的脉冲化氘氚聚变中子源产生的145 MeV单能中子,通过飞行时间法,测量了5、10、15 cm厚度板状铌（Nb）样品在与60°和120°两个方向上的泄漏中子飞行时间谱。利用蒙特卡罗模拟软件MCNP 4C进行了泄漏中子飞行时间谱的模拟计算,分别获得了CENDL 31、ENDF/B Ⅷ0和JENDL 40 3个数据库中Nb评价数据的模拟结果。通过各数据库不同能区的模拟结果与实验结果的比值（C/E）,对3个数据库中93Nb与145 MeV中子作用的角分布和双微分截面等相关评价数据进行了检验,重点分析了CENDL 31库的数据。结果表明,CENDL 31数据库的模拟结果在弹性散射能区、非弹性散射能区以及(n,2n)反应能区与实验结果均存在一定的偏差。而JENDL 40数据库除在120°弹性散射能区有高估现象,其他能区的模拟结果与实验结果均符合较好。ENDF/B Ⅷ0数据库的模拟结果除在60°方向弹性散射峰偏低外,其他能量范围的模拟结果均高于实验。     

The characteristics of 15 MeV and fission neutron damage in niobium

Displacement damage by 15 MeV (d-Be source) and fission neutrons at 30°C in high purity niobium single crystals has been studied by transmission electron microscopy. The fluence of the 15 MeV neutrons was 1.8 × 10¹⁷n/cm² and the fluence of the fission neutrons (5 × 10¹⁷ n/cm²) was chosen so that samples from both types of irradiations had approximately the same damage energy. In both 15 MeV and fission neutron irradiated specimens, the loops were observed to be about $\text{2}\text{3}$ interstitial and $\text{1}\text{3}$ vacancy type. The analysis of Burgers vectors of the dislocation loops showed that more than $\text{2}\text{3}$ of the loops were perfect $\text{a}\text{2}\text{〈111〉}$ and that the rest were $\text{a}\text{2}\text{〈110〉}$ faulted. It is concluded that at equal damage energies, the detailed nature of the damage is the same for 15 MeV and fission neutron irradiated niobium.     

Calculated neutron cross sections for Cu and Nb up to 32 MeV for neutron damage analysis

Cross sections for neutron interaction with Cu and Nb, with emphasis on spectra of light particles from binary reactions, are calculated for neutron energies from 4 to 32 MeV for estimating recoil probability densities for the analysis of damage experiments with a Be (d, n) neutron source. Nuclear model parameters were adjusted to reproduce the available cross-section data around 14 MeV. Helium production cross sections were also calculated for ⁶³Cu for neutrons below 20 MeV, as an illustration of the Hauser-Feshbach method for calculating tertiary reaction cross sections.     

Shielding design and analyses of KIPT neutron source facility

Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of a neutron source facility. An electron accelerator drives a sub-critical facility (ADS) is used for generating the neutron source. The facility will be utilized for performing basic and applied nuclear researches, producing medical isotopes, and training young nuclear specialists. Monte Carlo code MCNPX has been utilized as the major design tool for the design, due to its capability to transport electrons, photons, and neutrons at high energies. However the ADS shielding calculations with MCNPX need enormous computational resources and the small neutron yield per electron makes sampling difficulty for the Monte Carlo calculations. The high energy electrons (E > 100 MeV) generate very high energy neutrons and these neutrons dominant the total radiation dose outside the shield. The radiation dose caused by high energy neutrons is ∼3-4 orders of magnitude higher than that of the photons. However, the high energy neutron fraction within the total generated neutrons is very small, which increases the sampling difficulty and the required computational time. To solve these difficulties, the user subroutines of MCNPX are utilized to generate a neutron source file, which record the generated neutrons from the photonuclear reactions caused by electrons. This neutron source file is utilized many times in the following MCNPX calculations for weight windows (importance function) generation and radiation dose calculations. In addition, the neutron source file can be sampled multiple times to improve the statistics of the calculated results. In this way the expensive electron transport calculations can be performed once with good statistics for the different ADS shielding problems. This paper presents the method of generating and utilizing the neutron source file by MCNPX for the ADS shielding calculation and similar accelerator facilities, and the accurate radiation dose analyses outside the shield using modest computational resources.     

活化法测量197Au等核素的14MeV中子反应截面

采用活化法测量了197Au等核素的14MeV中子反应截面.当入射中子能量为14.75MeV时,197Au(n,2n)196Au反应截面测量结果为(2175±76)×10-31m2,并与其他测量结果和ENDF/B-6评价数据库数据进行了比较.对于结果的一些不确定度因素,采用MCNP程序进行了分析.     

Theoretical and Experimental Determinations of Neutron Energy Deposition in Silicon

Theoretical calculations have been made of the energy deposited in silicon in ionization and elastic interactions by neutrons in the energy range of 60 keV to 15 MeV. In contrast to earlier determinations, care was taken to calculate accurately the effects of atomic recoils. These are of primary importance for permanent effects in silicon at all neutron energies and account for about 30% of the transient ionization at 14 MeV. To test the calculation of the energy deposited by 14 MeV neutrons, experiments were performed with a pulsed d-t generator. The experimental values in rads per unit fluence were (11 ± 2) × 10-10 rad-cm2 for special p-n junctions and (8 ± 2) × 10-10 rad-cm2 for power transistors. Both experimental values agree, withi jerror, with the calculated value of (8 ± 1) × 10-10 rad-cm2. The calculated variation of radiation damage in silicon with neutron energy was compared with existing experimental data. There is general agreement of theory with experiment from 200 keV to 14 MeV.     

Neutron dose equivalent rate for heavy ion bombardment

The fluence rate distribution of neutrons in the reactionsof 50MeV/u ^18O-ion on thick Be,Cu and Au targets have been measured with an activation method of threshold detectors andthe neutron dose equivalent rate distributions at 1m from the tqrgets in intermediate energy heavy ion target area are obtained by using the conversion factors from neutron fluence rate to neutron doseequivalent rate.     

加速器中子源大厅内散射中子分布的模拟

本文针对加速器中子源可在较宽能量区间产生单能中子的特点,采用MCNP5对0.2～20 MeV的源中子在加速器中子源大厅内的散射情况进行模拟计算和分析。结果表明,直射中子通量随离源距离的增大呈平方反比衰减,散射中子通量则随离源距离的增大而几乎保持不变;大厅内的散射中子主要来自墙壁的贡献,离墙壁越近散射率越高。能量为0.4 MeV和1 MeV的源中子散射率最高,10 MeV和15 MeV的源中子散射率最低。用中子的宏观散射截面可较好解释散射率模拟结果,中子的弹性散射截面远大于非弹性散射截面,因此弹性散射起主导作用。中子能量大于1 MeV后,散射截面随中子能量增加而减小直至进入一段坪区,散射率也随之降低并进入坪区。结合待测位置处直射、散射中子通量和不同能量的散射中子份额的计算,能解释能量较高的源中子散射率较低的现象。通过在墙壁表面附上一层中子慢化吸收材料的方法可有效减弱中子散射,如5 cm的含硼聚乙烯（10%B4C）可降低散射率约40%。     

Preliminary Experiments on Accelerator-Driven Subcritical Reactor with Pulsed Neutron Generator in Kyoto University Critical Assembly

A series of preliminary experiments on an accelerator-driven subcritical reactor (ADSR) with 14 MeV neutrons were conducted at Kyoto University Critical Assembly (KUCA) with the prospect of establishing a new neutron source for research. A critical assembly of a solid-moderated and -reflected core was combined with a Cockcroft-Walton-type accelerator. A neutron shield and a beam duct were installed in the reflector region for directing as large a number as possible of the high-energy 14MeV neutrons generated by deuteron-tritium (D-T) reactions to the fuel region, since the tritium target is located outside the core. And then, neutrons (14MeV) were injected into a subcritical system through a polyethylene reflector. The objectives of this paper are to investigate the neutron design accuracy of the ADSR with 14MeV neutrons and to examine experimentally the neutronic properties of the ADSR with 14MeV neutrons at KUCA. The reaction rate distribution and the neutron spectrum were measured by the foil activation method for investigating the neutronic properties of the ADSR with 14 MeV neutrons. The eigenvalue and fixed-source calculations were executed using a continuous-energy Monte Carlo calculation code MCNP-4C3 with ENDF/B-VI.2 for the subcriticality and the reaction rate distribution, respectively; the unfolding calculation was done using the SAND-II code coupled with JENDL Activation Cross Section File 96 for the neutron spectrum. The values of the calculated subcriticality and the reaction rate distribution were in good agreement with those of the experiments. The results of the experiments and the calculations demonstrated that the installation of the neutron shield and the beam duct was experimentally valid and that the MCNP-4C3 calculations were accurately carried out for analyzing the neutronic properties of the ADSR with 14MeV neutrons at KUCA.     

Measurements of neutron induced activation of concrete at 64.5 MeV

The present paper presents the measurement of neutron induced activations on concrete using the 64.5 MeV quasimonoenergetic neutrons produced at the intense ⁷Li(p, n) neutron source at Cyclotron and Radioisotope Center, Tohoku Univeristy (CYRIC). The data were corrected for the effect of continuous neutrons in the source. The neutron energy, neutron yields and the spectrum of continuous neutrons were confirmed with the neutron time-of-flight method and the neutron activation measurement of the ²⁰⁹Bi(n, Xn) reactions having various threshold energy values. The nuclides produced by thermalized source neutrons are negligible. New data were obtained for concrete activation.     

Measurement of ~(232)Th(n,2n)~(231)Th reaction cross-sections at neutron energies of 14.1 MeV and 14.8 MeV using neutron activation method

In this study, the activation cross-sections were measured for ~(232)Th(n,2n)~(231)Th reactions at neutron energies of 14.1 and 14.8 MeV, which were produced by a neutron generator through a T(d,n)~4He reaction. Induced gamma-ray activities were measured using a low background gamma ray spectrometer equipped with a high resolution HPGe detector. In the cross-section calculations, corrections were made regarding the effects of gamma-ray attenuation, dead-time, fluctuation of the neutron flux, and low energy neutrons. The measured cross-sections were compared with the literature data, evaluation data(ENDF-B/VII.1, JENDL-4.0 and CENDL-3.1), and the results of the model calculation(TALYS1.6).     

Proton simulation of displacement effects induced in metals by 14 mev neutrons

In designing a D-T fusion reactor, one must know the effect of a high flux of 14 MeV neutrons on structural materials. Available laboratory sources of 14 MeV neutrons are not intense enough to expose samples to the expected flux. Bombardment with other particles is one way of simulating the anticipated neutron environment. The energy spectrum of atoms recoiling from collisions with bombarding particles can be calculated from elastic-scattering and nonelastic-reaction data for the incident species. This analysis shows that 16 MeV protons closely simulate the displacement effects caused by 14 MeV neutrons. In niobium the average atom recoiling from a 14 MeV neutron interaction has 65 keV of damage energy. The mean damage energy deposited per cm³ of niobium by a fluence of one 14 MeV neutron per cm² is 14 keV. The equivalent quantity for 16 MeV protons incident on niobium is 33 keV.     

Measurement of Gamma-Ray Production Cross Sections of Iron for Incident Neutron Energies between 6 and 33 MeV

Measurement of differential γ-ray production cross sections, i.e. (n, x γ) cross sections, of Fe was made for neutron energies from 6 to 33 MeV. Neutrons used in the experiment were white neutrons produced with (p, n) reactions by 35 MeV protons using a thick Be target. The neutron energy was analyzed by the time-of-flight method and bunched into 3 MeV wide energy bins, for each of which the spectrum of secondary γ-rays produced in an Fe sample was measured by a BGO scintillator at an angle of 144° to the neutron beam direction.

The obtained (n, xγ) cross sections agreed well with other data and the evaluated data file of ENDF/B-IV at neutron energies below 15 MeV where data were existing. The JENDL-3 file overestimated the γ-ray spectra at γ-ray energies of 3 to 7 MeV. The present work newly provided the data in the neutron energy range above 20 MeV. The GNASH calculation made by Young reproduced the measured data fairly well even at these higher energies.     

A function to provide neutron spectrum produced from the 9Be + p reaction with protons of energy below 12 MeV

A function to give the total neutron production cross section, angular distribution, and energy spectrum via the ⁹Be + p reaction has been created by fitting experimental data to characterize compact neutron sources with thick Be targets bombarded by protons with energy below 12 MeV. To examine the suitability of the function, calculations of the angle-dependent neutron energy spectra produced in thick Be targets with 4- and 12-MeV protons using the function were compared with corresponding experiments and calculations using the nuclear data libraries of ENDF/B-VII.0 and JENDL4.0/HE. The function was in better agreement with the experiments than the calculations using the libraries except for at backward angles. The ¹¹⁵In(n,n’)^115mIn reaction rates calculated using GEANT4 with source neutrons given by both the function and ENDF/B-VII.0 were compared with that measured at the RIKEN Accelerator-Driven Compact Neutron Source to evaluate the neutron spectrum above 1 MeV. The function slightly overestimated the measurement by 14% and the calculation with ENDF/B-VII.0 underestimated by 35%. It was concluded that the function can be applied in compact neutron source designs.     

Ratio tests of dosimetry cross-sections for Nb(n,2n)Nb and Au(n,2n)Au to Al(n,α)Na reactions

Cross-section ratios of reaction ⁹³Nb(n,2n)^92mNb and ¹⁹⁷Au(n,2n)¹⁹⁶Au to the standard reaction ²⁷Al(n,α)²⁴Na have been measured in order to test the recently compiled dosimetry files: JENDL Dosimetry File and International Reactor Dosimetry File 1990 (IRDF-90). The experimental results for both reactions were consistent with the calculated ones based on IRDF-90 except for the ⁹³Nb(n,2n)^92mNb above 19 MeV.Both reactions, especially ⁹³Nb(n,2n)^92mNb, were recommended as new neutron monitors above 12 MeV because of their favorable characteristics from the decay-property and cross-section viewpoints.     

Calculations of the transport of neutrons and secondary gamma rays through concrete for incident neutrons in the energy range 15 to 75 MeV

Earlier work by Alsmiller et al. considered coupled neutron and secondary-gamma-ray transport through a thick shield of silicon dioxide with 5% water by weight for neutron sources with energies of 50, 100, 200, 300 and 400 MeV. In that work, the approximation was made that gamma rays were produced only by neutron capture. In the present work, coupled neutron and secondary-gamma-ray transport through a thick shield of concrete for neutron sources with energies of 15, 25 and 75 MeV is considered. In this study, gamma-ray production for all interactions involving neutrons with energies up to 15 MeV was included; i.e., the approximation made here is that gamma-ray production can be neglected for interactions by neutrons with energies > 15 MeV.For incident neutron energies of 15, 25, 50, and 75 MeV, results of total and gamma-ray dose equivalents are given as a function of depth into the slab. For the 50- and 75-MeV incident neutron energies, the gamma-ray dose equivalent was found to be no more than 5% of the total dose equivalent at all depths considered ( 1500 g/cm²). For the 15- and 25-MeV incident neutron energies, however, the gamma-ray dose equivalent dominates at greater depths into the slab. A conservative estimate of the effect of including gamma rays produced in interactions with neutrons of energies > 15 MeV indicates that the calculated total dose equivalent would increase by no more than 5%.     

<Superscript>239</Superscript>Pu prompt fission neutron spectra

Calculations of ²³⁹Pu(n, F) prompt fission neutron spectra have been performed for neutron energy up to 20 MeV. The exclusive spectra of pre-fission neutron reactions (n, xnf) were calculated on the basis of the Hauser-Feshbach model simultaneously with the cross sections of (n, F) and (n, 2n) reactions. The spectra of neutrons emitted by fission fragments were approximated by a sum of two Watt distributions. The components of the prompt fission neutron spectra due to pre-fission neutrons are manifested in the prompt fission neutron spectra and the average neutron energy. A correlation is established between this effect in the contribution of emissive fission (n, xnf) in the fission cross-section of ²³⁹Pu(n, F) and ²³⁵U(n, F). It is shown that the ²³⁹Pu(n, F) prompt fission neutron spectra used in applied calculations do not correspond to the experimental differential data and the systematic regularities in the spectra and their average energy found for the most carefully studied nuclei ^235,238U and ²³²Th. __________ Translated from Atomnaya énergiya, Vol. 103, No. 2, pp. 119–124, August, 2007.     

Measurements of the Y(n,γ)Y cross-section in the neutron energy range of 13.5-14.6 MeV

The ⁸⁹Y(n,γ)^90mY cross-section has been measured at three neutron energy points between 13.5 and 14.6 MeV using the activation technique and a coaxial HPGe γ-ray detector. The data for the ⁸⁹Y(n,γ)^90mY cross-sections are reported to be 0.39 ± 0.02, 0.43 ± 0.02, and 0.38 ± 0.02 mb at 13.5 ± 0.2, 14.1 ± 0.1, and 14.6 ± 0.2 MeV incident neutron energies, respectively. The first data for the ⁸⁹Y(n,γ)^90mY reaction at neutron energy points of 13.5 and 14.1 MeV are presented. The natural high-purity Y₂O₃ powder was used as target material. The fast neutrons were produced by the T(d,n)⁴He reaction. Neutron energies were determined by the method of making cross-section ratios of ⁹⁰Zr(n,2n)^89m+gZr and ⁹³Nb(n,2n)^92mNb reactions, and the neutron fluencies were determined using the monitor reaction ⁹³Nb(n,2n)^92mNb. The results obtained are compared with existing data.     

Cross-section measurements for short-lived isotopes of 46Ti, 75As and 92Mo at the neutron energies from 13.6 to 14.9 MeV

Cross-sections were measured at neutron energies from 13.6 to 14.9 MeV for the reactions ⁴⁶Ti(n,p) ^46mSc, ⁷⁵As(n,p) ^75mGe and ⁹²Mo(n,2n) ^91mMo leading to short-lived products. Corrections were made for the effects of gamma ray attenuation, coincidence summing, pulse pile-up, dead time, neutron flux fluctuations and scattered low energy neutrons. Statistical model calculations taking into account precompound effects were also performed.