可燃毒物铒的辐照及其同位素热电离质谱法测定

铒(Er)是一种适用于轻水堆(LWR)的长效可燃毒物,Er_2O_3的堆内辐照是研究中子毒物Er辐照性能的基本手段。Er_2O_3在高通量工程试验堆(HFETR)中子注量率约为2×10~(14)/(cm~2·s)的辐照孔道中辐照91.3h。采用热电离质谱法(TIMS)测定辐照前、后样品中Er同位素丰度,跳峰模式测定6个Er同位素,低丰度~(162)Er由法拉第杯检测。测定时准确控制升温测量电流,将蒸发带和电离带电流控制在1 400、5 500mA以下,可减小~(168)Yb、~(170)Yb对~(168)Er、~(170)Er的同量异位素影响。结果表明,经中子辐照后~(166)Er、~(167)Er、~(168)Er同位素丰度变化较大,丰度变化与中子吸收截面大小密切相关。     

Determination of critical mass and neutron flux distribution by means of physical models

The design of nuclear reactors, especially new reactors, requires experimental measurements in order to obtain accurate values of the pertinent parameters. In the present paper we present a new method for the preliminary determination of the critical mass of a reactor and the neutron flux distribution; this method is based on the use of physical models. In carrying out these experiments use is made of a model of the reactor which does not contain fissionable material. The working channels in the model are filled with a neutron absorber whose cross section simulates the absorption cross section for neutrons in the fissionable material. The production of fast fission neutrons is simulated by means of a neutron source which is moved along the channels. The distribution of thermal neutrons is measured by means of detectors which are sensitive to thermal neutrons. If the source strength and the absolute value of the neutron flux are known, it is possible to find the critical mass of the reactor.This method has been checked in a reactor with uranium hexafluoride. The value of the critical mass found experimentally was found to be in good agreement with the value obtained when the reactor was started up.The proposed method can also be useful in preliminary investigations of reactor designs, the choice of optimum lattice parameters, etc. The technique is extremely simple and does not require fissionable material or high neutron fluxes.     

Cross section for thermal-neutron absorption and resonance absorption integral in Pu240

Using the oscillating reactor method the cross section for the absorption of thermal neutrons neutrons and the resonance absorption integral in the isotope Pu²⁴⁰ have been measured. The measurements were performed at two positions inside the heavy-water reactor of the Academy of Sciences, USSR; in the lattice and in the thermal column. Plutonium samples with three different abundances of the isotopes Pu²³⁹, Pu²⁴⁰ and Pu²⁴¹ were used. The oscillations in the power level produced by excursions of the samples were recorded with a recording potentiometer. To avoid self-shielding effects, the dependence of sample effect on plutoniuni weight was investigated. To determine the spectral composition of the neutron ilux in the lattice and in the column the cadmium ratio for indium was measured. The values for the cross sections in Pu²⁴⁰ are as follows in the column — (560 ± 35) barns, in the lattice — (1010 ± 120) barns. The cross section for absorption of thermal neutrons in Pu²⁴⁰ was found to be T = (460 ± 45) barns and the resonance absorption integral — = (10,000 ± 2,800) barns. The value obtained for the thermal neutron cross section agrees with a contribution in the thermal cross section from a resonance level at E₀ 1 ev. This indicates the absence of strong levels in Pu²⁴⁰ in the energy region from 1 ev down to negative values and also indicates that in the thermal-neutron region the cross section in Pu²⁴⁰ varies according to a 1/v law.The author wishes to take this opportunity to express his gratitude to Academician A. I. Alikhanov for his continued interest in the work and for hii valuable advice.The author also wishes to thank F. S. Laptev, V. L. Ioffe, and G. M. Kukavadze, who performed a number of measurements and calculations, and A. Ya. Diament, N. C. Khokhulin, V. F. Belkin, A. K. Dubasov, and Yu. V. Oriov for participating in the construction of the apparatus and the measurements.After this paper was submitted for publication a short report [13] appeared on the measurement of the Pu²⁴⁰ constants by a mass-spectrometer method. The data on the resonance integral is in agreement with the data of the present work and [7]; the data on the cross section for thermal neutron absorption, however, does not agree with the data of the present work. It should be noted, however, that the determination of the cross section for thermal neutrons by a mass-spectrometer method has a number of shortcomings.     

Measurement of the effective cross section in Th232 for thermal neutrons and the resonance integral for neutron absorption

The effective cross section in thorium for thermal neutrons (_ther = 7.31 ± 0.10 barns) and the resonance integral for thorium have been-measured in a heavy water reactor. The measurements were made by the activation method. Gold, indium, and uranium were used as comparison standards. The precise value of the effective cross section for indium for thermal neutrons is (_ther = 162 ± 10 barns) and the resonance integral in indium is (RI=2340 ± 200 barns).     

Neutron yield in inelastic collisions of 14-Mev neutrons with nuclei and cross section for the (n, 2n) reaction

A measurement has been made of , the number of neutrons produced in one inelastic scattering event between a neutron and a number of elements of natural isotopic composition: Fe, Cu, Mo, Cd, Sn, Sb, Hg, Pb, Bi and U. The measurements were performed by determining the relative change in the total neutron flux and the attenuation of the primary neutrons after passage through samples of the materials being investigated. It was also possible to obtain data on the cross section _in for inelastic collisions of neutrons with the above-mentioned nuclei. The values of and _in, in conjunction with the known cross sections for neutron capture, were then used to compute the cross section for the (n, 2n) reaction (averaged over the isotopic composition) in nonfissile nuclei.This work was completed in 1952.The authors wish to thank A.A. Malinkin for comparing the neutron yields from sources used to measure the dependence of counter sensitivity on neutron energy.     

Measurement of thermal neutron and resonance integral cross sections of the reaction V(n,γ)V using a 20 Ci Am–Be isotopic neutron source

The thermal neutron capture cross section (σ_o) and the resonance integral (I_o) of the ⁵¹V(n,γ)⁵²V reaction were measured with an activation method to provide fundamental data for reactor calculation, activation analysis, and other theoretical and experimental uses concerning the interaction of neutron with matter. The vanadium and manganese samples were irradiated within and without a Cd shield case using a 20 Ci Am–Be neutron source. The activities of the samples were measured using gamma-ray spectroscopy. The thermal neutron capture cross section and the resonance integral were determined relative to the reference reaction ⁵⁵Mn(n,γ)⁵⁶Mn and the values obtained are 5.16 ± 0.19 barns and 2.53 ± 0.1 barns respectively. The previous measurements of the σ_o and I_o of the reaction ⁵¹V(n,γ)⁵²V were reviewed and the difference between the present values and the previous results were discussed.     

Application of stable boron isotopes

This article presents a brief survey of the basic applications of stable boron isotopes as materials with an altered isotope composition in cases where the differences between the nuclear properties of boron isotopes are used directly. The neutron flux transformation into heavy ionizing particles by means of the B¹⁰ (n, )Li⁷ reaction and the large value of the effective thermal neutron cross section of this reaction make it possible to use B¹⁰ in medicine, nuclear studies, and radiation chemistry. The differences between the neutron cross sections of B¹⁰ and B¹¹ make It possible to use these isotopes in reactor construction and, in particular, to use B¹⁰ in materials for control rods and reactor shields.     

Determination of Activation Cross Section by Gamma-Ray Spectrometry

A quick and easy way of determining activation cross sections by the analysis of γ-ray spectra is described. To determine cross sections by this method, it is not necessary to measure the neutron flux density, to undertake any chemical separation process, nor to use enriched target substances. To ascertain the validity of the method, the activation cross section of ¹⁰²Ru(n,γ) ¹⁰²Ru reaction was obtained, resulting in the value of 1.37±0.132 barns, which compares well with published data of 1.44±0.16 barns, determined by another method.     

Measurement of neutron spectra in the thermal column of the reactor of a power station

Knowledge of neutron spectra In nuclear reactors allows comparison of various theories of the slowing down of neutrons with experiment, and also allows carrying out reactor calculations which are based on actual neutron distributions therein. In this paper is described a neutron intensity monochtomator Intended for the measurement of neutron spectra in the energy interval 0 to 0.5 ev.Results are given for measurements for neutron spectra in the thermal column of the reactor of an atomic power station. Discontinuities in the neutron flux were discovered at neutron velocities of 600, 1000 and 1650 m/sec; an analysis is given of the causes of discontinuities of the neutron flux; an evaluation is made of the inelastic scattering cross section for neutrons in graphite. By the method of least squares, the temperature of the neutron gas was found, being equal to 354 ° K at a graphite temperature of 304 ° K.In conclusion we consider it our duty to express gratitude to A. K. Krasin and B. G. Dubovskii for interest and help in the work and F. L. shapiro for valuable interpretation of previous results.     

~(176)Hf(n,2n)~(175)Hf反应截面测量

采用相对测量技术,以活化法对13.4～14.8MeV范围内的176Hf(n,2n)175Hf反应截面进行了测量。样品固定在距离D-T中子源20cm处的圆环的不同位置上进行中子辐照,采用93Nb(n,2n)92Nbm作为监测反应,活化产物采用高纯锗探测器进行了测量,所得14MeV附近的176Hf(n,2n)175Hf反应截面实验值为(2100±85)mb,对实验结果与公开文献值和ENDF/B6.8评价库数据进行了比对。     

Fast neutron fluence calculations as support for a BWR pressure vessel and internals surveillance program

Materials surveillance programs are required to detect and prevent degradation of safety-related structures and components of a nuclear power reactor. In this work, following the directions in the Regulatory Guide 1.190, a calculational methodology is implemented as additional support for a reactor pressure vessel and internals surveillance program for a BWR. The choice of the neutronic methods employed was based on the premise of being able of performing all the expected future survey calculations in relatively short times, but without compromising accuracy. First, a geometrical model of a typical BWR was developed, from the core to the primary containment, including jet pumps and all other structures. The methodology uses the Synthesis Method to compute the three-dimensional neutron flux distribution. In the methodology, the code CORE-MASTER-PRESTO is used as the three-dimensional core simulator; SCALE is used to generate the fine-group flux spectra of the components of the model and also used to generate a 47 energy-groups job cross section library, collapsed from the 199-fine-group master library VITAMIN-B6; ORIGEN2 was used to compute the isotopic densities of uranium and plutonium; and, finally, DORT was used to calculate the two-dimensional and one-dimensional neutron flux distributions required to compute the synthesized three-dimensional neutron flux. Then, the calculation of fast neutron fluence was performed using the effective full power time periods through six operational fuel cycles of two BWR Units and until the 13th cycle for Unit 1.The results showed a maximum relative difference between the calculated-by-synthesis fast neutron fluxes and fluences and those measured by Fe, Cu and Ni dosimeters less than 7%. The dosimeters were originally located adjacent to the pressure vessel wall, as part of the surveillance program. Results from the computations of peak fast fluence on pressure vessel wall and specific weld locations on the core shroud are also presented.     

Measurements of Neutron Capture Cross Section of 237Np for Fast Neutrons

The neutron capture cross section of ²³⁷Np has been measured for fast neutrons supplied at the center of the core in the Yayoi reactor. The activation method was used for the measurement, in which the amount of the product ²³⁸Np was determined by γ-ray spectroscopy using a Ge detector. The neutron flux at the center of the core calculated by the Monte Carlo simulation code MCNP was renormalized by using the activity of a gold activation foil irradiated simultaneously. The new convention is proposed in this paper to make possible a definite comparison of the integral measurement by the activation method using fast reactor neutrons with differential measurements using accelerator-based neutrons. “Representative neutron energy” is defined in the convention at which the cross section deduced by the activation measurement has a high sensitivity. The capture cross section of ²³⁷Np corresponding to the representative neutron energy was deduced as 0:80 ± 0:04b at 214 ± 9 keV from the measured reaction rate and the energy dependence of the cross section in the nuclear data library ENDF/B-VII.0. The deduced cross section of ²³⁷Np at the representative neutron energy agrees with the evaluated data of ENDF/B-VII.0, but is 15% higher than that of JENDL-3.3 and 13% higher than that of JENDL/AC-2008.     

基于启明星Ⅱ号的铅基堆核数据宏观检验

为提高铅基堆中子学模拟的可靠性,基于启明星Ⅱ号铅基零功率反应堆,开展铅基堆相关核数据的入堆宏观基准检验研究。采用周期法测量堆芯反应性,进而获得有效增殖因数keff为1001 14±0000 07。采用MCNP程序对铅基堆进行精细化建模,结合不同数据库内的中子评价核数据,计算实验燃料棒装载下的铅基堆芯的keff。比较结果可知,4种截面库计算的铅基堆keff模拟结果与实验结果吻合较好,最大相对偏差小于1%,其中,ENDF/B Ⅶ.1库的模拟结果与实验结果吻合最好,相对偏差和绝对偏差分别为025%和251 pcm。通过计算关键材料元素核数据引起keff的变化量,可知铅元素核数据引起的堆芯keff结果的波动量最大,在CENDL 31和JENDL 40中的铅元素引起keff的波动值分别为219 pcm和166 pcm。     

Effective Average Integrated γ-Ray Flux Striking Radiation Defects during Irradiation in a Reactor

The results of experiments on the irradiation of diamond which attest to the fruitfulness of the zone theory of damage to materials are presented. The theory is used to derive an expression for the effective average integrated -ray flux striking radiation defects during irradiation in a reactor. It is shown by statistical analysis of the results on the critical neutron fluence for reactor graphite that when the average -ray flux is taken into account, other conditions being equal, the statistical error in the estimate of the critical neutron fluence for graphite decreases by a factor of 1.5. This shows that the effective average integrated -ray flux is another important parameter of the irradiation conditions, and this expands the physical meaning of the radiation composition factor. 3 figures, 1 table, 9 references.     

Pu-breeding feasibility in PWR

This study addresses the issue of alternative pathways for breeding plutonium in a 900 MWe three loop thermal pressurized water reactor (PWR), either fueled with uranium fuel (3.5% U-235) or with mixed fuel (20% MOX). During the operation of a nuclear reactor the in-core neutron flux and the ex-core neutron flux are monitored with flux detectors. At the places where those detectors operate, the guide thimbles and the vessel wall, respectively, the neutron flux can be used to irradiate material samples. This paper investigates whether it would be possible to produce plutonium by breeding it at the walls of a PWR vessel and/or in the guide thimbles. The neutron flux in the reactor and the corresponding multi-group spectra are estimated with Monte Carlo simulations for different positions at the vessel wall of a PWR operating with either UO₂ or MOX. Then the irradiation of fresh uranium samples at the vessel wall and in the guide thimbles are calculated and the isotopic composition of the irradiated samples are determined. The minimum irradiation period and the necessary minimum amount of fresh uranium to breed different grades of plutonium are derived.     

14.8 MeV中子诱发78Kr(n,2n)77Kr反应截面的测量

快中子诱发(n,2n)反应截面的测量在核反应机制研究和核技术应用等方面有着广泛的应用价值。本文在中国原子能科学研究院的高压倍加器上,基于活化法实验测量了78Kr(n,2n)77Kr在148 MeV能点的反应截面。样品靶为高纯78Kr气体样品,用十万分之一天平称重得到78Kr的质量,将两片高纯93Nb薄片分别固定在样品靶两侧以监测中子注量率。利用T(d,n)4He反应产生148 MeV中子,轰击距中子源约10 cm的样品靶大于4 h后,用准确刻度过效率的HPGe探测器测量活化产物 77Kr和92Nbm的活度。利用蒙特卡罗程序计算中子注量率修正、样品自吸收修正、样品几何修正等因子,得到了78Kr(n,2n)77Kr的反应截面,并将结果与文献值和评价数据库进行了比较。研究结果有助于提高78Kr(n, 2n)77Kr反应截面测量和评价的水平。     

铀水栅零功率反应堆的功率刻度

反应堆功率的测量,在堆功率高时一般用热工方法,功率低时,可用各种堆物理方法,如中子源引进法、中子统计法和全堆总裂变率法。 中子源引进法误差较大,中子统计法需知探测器在堆内的效率和堆的β_(aff)值,此二者都较难测量。全堆总裂变率法是由测量堆的总裂变率来求得堆功率,它可避免前面两种方法的缺点,但需依赖裂变率相对分布的     

离散纵标输运计算方法在压水堆核电厂41Ar活化源项分析中的应用

压水堆核电厂功率运行期间,反应堆压力容器外的环形空腔空气中所含的⁴⁰Ar被中子活化,形成具有放射性的⁴¹Ar。文章采用二维离散纵标输运计算程序DORT分析了反应堆堆腔区域的中子注量率分布情况,采用NJOY评价核数据处理程序,根据DORT分析得到的通量作为权重通量,利用基础评价核数据库ENDF/B-Ⅶ.0制作⁴⁰Ar中子俘获反应的微观截面,在此基础上,分析了百万千瓦级压水堆核电厂每台机组反应堆堆腔空气中⁴⁰Ar中子活化生成⁴¹Ar的生成率以及电厂⁴¹Ar的环境排放源项。文章给出的⁴¹Ar源项分析方法可作为压水堆核电厂设计中确定⁴¹Ar源项的最佳估算值的参考。     

辐照条件下高温锂热管不凝性气体产生特性研究

在新型热管冷却反应堆中,高温金属热管会受到持续的中子辐照。锂在热中子区的中子反应微观截面很大,会产生一定量的氦气,氦气作为不凝性气体将影响高温热管的正常运行。本文分析了堆内中子辐照条件对高温锂金属热管中不凝性气体产生特性的影响。首先对稳态标准算例进行了产氦量分析,并转换得到了不凝性气体体积份额。此外,得到了不凝性气体产量随热管充液量、金属锂富集度、中子通量密度、热管工作温度等因素的变化关系。不凝性气体产量随热管充液量、锂富集度的增大而增加。控制转鼓位于不同角度时,中子通量密度改变有限,对产氦量影响不大,由于高温锂热管工作温度很高,高温下中子反应微观截面差距很小,因此热管工作温度对产氦量影响也有限。本研究可为热管冷却反应堆内高温锂热管中锂富集度设计提供借鉴。     

用钴活化法测定反应堆中热中子积分通量

本文叙述了用钴活化法测定高通量堆中热中子积分通量的方法。测得的热中子积分通量值与计算值作了比较。本法适于测定在高通量堆中长期辐照的较高热中子积分通量。