Research of Radioisotope Production with Fast Neutrons, (VII)

Production of radioisotopes of high specific activity was studied in the JRR-1 reactor using several (n,p) and (n,α) reactions, such as ²⁴Mg(n,p)²⁴Na, ²⁷A1(n,α)²⁴N_a, ³⁵Cl(n,p)³⁵S, ³⁵C1(n,α)³²P, ⁵⁸Ni(n,p)⁵⁸Co, ⁶⁴Z_n (n,p) ⁶⁴C_u and ⁶⁷Z_n(n,p) ⁶⁷Cu. The target materials for these reactions were irradiated in several experimental holes of JRR-1 and the radioisotopes formed in the target materials were separated. The amount of the radioisotopes produced and the specific activity were determined, and the possibility of producing high specific activity radioisotopes by these reactions was investigated. The specific activity of the radioisotopes produced by the (n,p) and (n,γ) reactions was more than several hundreds times higher than when produced by the corresponding (n,γ)reactions. Although the yield of the radioisotopes by the former two reactions was fairly small, practical production of high specific activity radioisotopes by this method was thought to be possible, at least for elements of lower atomic number such as those studied in the present work.

For each experimental hole, the thermal and the fast neutron fluxes were determined respectively by the reactions ¹⁹⁷Au(n,γ)¹⁹⁸Au and ⁵⁸Ni(n,p)⁵⁸Co. In order to apply these (n,p) and (n,α) reactions effectively to radioisotope production, such basic informations as the dependence of the reactions on neutron energy and the effect of irradiation position on the reaction yield were studied on the basis of the neutron flux distribution, and the cross section of the reactions for fast neutrons in JRR-1 was estimated.     

Research of Radioisotope Production with Fast Neutrons, (VI)

Copper-67, the longest-lived radioisotope of copper, was prepared by using the reaction ⁶⁷Zn(n p,)⁶⁷Cu. Zinc enriched in ⁶⁷Zn up to 93.4% was irradiated in JRR-1 reactor for 5hr with a fast neutron flux of 1.5×10¹²n/cm²/sec. Copper-67 formed in the zinc target was separated by anion exchange separation using the chloro-complex of the metals. The yield of ⁶⁷Cu and radiochemical impurity were investigated. Seventeen millimicrocuries of ⁶⁷Cu was obtained from 1mg of zinc. No radioactive impurities other than ⁶⁴Cu were not found. Even with the use of enriched target, a more ^MCu was produced than ⁶⁷Cu. The proportion of ⁶⁷Cu to total copper radioactivity was 26% immediately after the irradiation.     

Detection of Fast Neutrons with Nal(Tl) Crystal

An attempt has been made of using Nal(Tl) crystal as fast neutron detector in conventional Time-of-flight apparatus. Typical pulse hight distributions, in the γ-ray energy range below 1 MeV, produced by the incidence of fast neutrons are presented. The characteristics of this neutron detection system have been studied, using pulsed neutrons from the T(p, n) and D(d, n) reactions. Neutron detection efficiency curves are presented, covering both whole γ-rays and those from a few low excited levels of ¹²⁷I nucleus. The detection bias can be stably set at the first excitation energy (58 keV), which is lower than possible for an organic scintillator.     

利用中国实验快堆生产放射性同位素的可行性研究

中国实验快堆（CEFR）不仅能进行各种燃料、材料辐照实验,也是放射性同位素生产的优良平台。本文对CEFR的辐照性能进行了描述,并利用计算程序对适宜在CEFR上生产的同位素³²P、³³P、³⁵S、⁸⁹Sr、¹⁴C、⁶⁰Co进行理论计算,得到了产量和比活度等参数。计算结果表明,在CEFR堆芯辐照可得到纯度极高的³²P、³³P、³⁵S,利用快中子的（n,p）反应可得到无载体的⁸⁹Sr,在CEFR反射层布置慢化材料可得到比活度较高的¹⁴C、⁶⁰Co。以上结果表明,在CEFR上生产同位素是可行的。     

研究试验堆用于同位素生产的物理分析

本文对研究试验堆开展同位素生产进行了物理分析。分析了控制棒提棒顺序对同位素产量的影响,提出了提棒因子的概念。依据点堆模型和反应性-燃耗线性公式,得到了同位素的转换比和产量公式。最后根据这些公式,分析了高通量工程试验堆(HFETR)在高浓铀和低浓铀堆芯装载下,堆芯炉的运行寿期、燃料元件装载数量、燃料元件初始平均燃耗和堆芯功率对同位素转换比和产量的影响。结果显示,从小到大提棒、增加堆芯燃料组件盒数和功率水平均会增加堆芯同位素产量,而全年运行段数（运行段间检修时间不变）和堆芯平均初始燃耗增加则起到相反的作用。这些结果已经用于指导反应堆的堆芯装载设计。     

Void Formation in Magnesium Aluminate Spinel Heavily Irradiated with Fast Neutrons

Single crystals of magnesium aluminate spinel (MgAl₂O₄) were heavily irradiated to fission neutron fluences from ?1 × 10²⁶ to ?2×10²⁷ n/m² (E>0.1 MeV) at 658 and 1,023 K in FFTF (Fast Flux Test Facility) to investigate their microstructural evolution under heavy neutron irradiation and to confirm their previously observed resistance to void swelling. At 658 K no voids were observed up to 2.29 × 10²⁷ n/m², but 1/4[110] interstitial loops were formed. These loops changed their habit planes from (111) to (110) with increasing neutron fluence. At 1,023 K the 1/4[110] type of interstitial loops grew to form stacking fault netuorks composed of stacking faults on each of the six equivalent (110) planes. Tiny voids were also observed to form preferentially on or near stacking faults after 1.37×10²⁷ n/m². A limited number of very small cavities, contributing to a volumetric swelling of only 0.07%, were also observed in the crystal matrix after 2.17×10²⁷ n/m². The possible mechanisms of suppression of void formation in MgAl₂O₄ are discussed.     

Response of RadFET Dosimeters to High Fluences of Fast Neutrons

We present irradiation experiments carried out on RadFETs in the high-intensity T2 neutron beam at the CRC-UCL in Belgium. The aim of the test was to characterize the neutron response of RadFETs in view of their use as an integrated part of a radiation-monitoring sensor for the CERN Large Hadron Collider (LHC) experiments. Two types of RadFETs were investigated up to a total neutron fluence of$3times10^14 hbox cm^-2$corresponding to a deposited dose of 744 Gy in silicon. The responses of bare devices to neutrons are compared to the commonly used reference measurements with gamma rays. It is found that the gamma ray calibration cannot directly be adopted to convert the RadFET signals into neutron dose. In a second experiment, the influence of a plastic packaging, simulated by polyethylene slabs of different thicknesses, was tested in the neutron beam and compared to GEANT4 Monte Carlo simulations. An increase of the RadFET neutron sensitivity by a factor of up to 7 due to the packaging is found. The influence of these findings on the conception of the radiation-monitoring sensor is discussed.     

Generalized Synthetic Kernel Approximation for Elastic Moderation of Fast Neutrons

A method of synthetic kernel approximation is examined in some detail with a view to simplifying the treatment of the elastic moderation of fast neutrons. A sequence of unified kernel {f _N } is introduced, which is then divided into two subsequences {W _n } and {G_n} according to whether N is odd (W _n =f _2n-1, n=1, 2,…) or even (G _n =f _2n , n=0,1, …). The W ₁ and G ₁ kernels correspond to the usual Wigner and GG kernels, respectively, and the W _n and G _n kernels for n ≧ represent generalizations thereof. It is shown that the W _n kernel solution with a relatively small n (≧2) is superior on the whole to the G_n kernel solution for the same index n, while both converge to the exact values with increasing n. To evaluate the collision density numerically and rapidly, a simple recurrence formula is derived. In the asymptotic region (except near resonances), this recurrence formula allows calculation with a relatively coarse mesh width whenever h_a ≧0.05 at least. For calculations in the transient lethargy region, a mesh width of order ?/10 is small enough to evaluate the approximate collision density φ _N with an accuracy comparable to that obtained analytically. It is shown that, with the present method, an order of approximation of about n=7 should yield a practically correct solution diviating not more than 1% in collision density.     

Formation Cross Sections of Zirconium-90m and Lead-207m with Fast Neutrons

Abstract

The cross sections for the formation of the metastable states of ⁹⁰Zr, (σ zr), for neutron energy of 14.8 MeV, and that of ²⁰⁷Pb, (σ pb,), for neutron energies from 4.1 to 5.2 MeV have been measured and the results are reported. The experimentally determined value of the cross section for the former is 700±94mb, while for the latter, the cross section increases from 316 + 44 mb to 788±110 mb. The cross section ratio, σpb/σZr, is also studied, which is found to vary from 1.87 to 2.18 for neutron energies from 14.8 to 18.2 MeV respectively.

The equality in the half-lives of the metastable states of ⁹⁰Zr and ²⁰⁷Pb is exploited in the present measurements, which enabled comparative evaluation of the cross sections.

Natural samples of zirconium and lead were irradiated simultaneously and the x-rays were measured off-line by a 55 cm³ Ge(Li) detector. Cyclic activation process was adopted to improve the counting statistics. The overall accuracy achieved in the measurements is about 10%.     

Study of the Effects of Fast Neutrons on Silicon Controlled Rectifiers

Medium power silicon controlled rectifiers have been exposed to a maximum integrated fastneutron flux of 4 x 1013 nvt. An increase in the current required to turn on thie controlled rectifier and an increase in the break-over voltage were observed after irradiation. The saturation voltage increased as a result of irradiation, exceeding 200 volts at an integrated fast neutron flux of 4 x 1013 nvt. These changes are attributed to a decrease of the minority carrier lifetime as reflected by a decrease in alpha.     

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.     

Neutronics Experiments of D-T Fast Neutrons by Application of Associated Particle Method

A probabilistic method is proposed that will permit in the design stage to estimate quantitatively the likelihood with which any or all design criteria applicable to a nuclear reactor are actually satisfied after its construction. The method is trially applied to the core reactivity balance problem of the experimental Very High Temperature Reactor, and calculations are performed on the probability with which a design study core will, upon construction, satisfy design criteria concerning (a) one rod stuck and (b) startup margin. The method should prove useful in making engineering judgments before approving reactor core design.     

含反射中子的快脉冲堆中子动力学方程研究

提出了含有反射中子的快脉冲堆中子动力学方程,并据此开展了快脉冲堆在超瞬发临界状态下产生脉冲的数值模拟计算。结果表明,反射中子增加了反应性和脉冲产额,由于反射中子使脉冲波形展宽,波形也不再对称。     

Dieaway Measurements of Fast Neutrons in an Iron Assembly using Californium-252

The fast transient behavior of the fast neutron population in an iron assembly immediately following the burst was measured by changing the energy response of the neutron detector with use made of ²⁵²Cf as a randomly pulsed source.

Comparison between experiment and calculation proved that this type of transient experiment would be useful for evaluating the method of calculation adopted in solving space-time-dependent problems.     

Fission Product Yields of U, U, U and Pu in Fields of Thermal Neutrons, Fission Neutrons and 14.7-MeV Neutrons

The yields of more than fifteen fission products have been carefully measured using radiochemical techniques, for ²³⁵U(n,f), ²³⁹Pu(n,f) in a thermal spectrum, for ²³³U(n,f), ²³⁵U(n,f), and ²³⁹Pu(n,f) reactions in a fission neutron spectrum, and for ²³³U(n,f), ²³⁵U(n,f), ²³⁸U(n,f), and ²³⁹Pu(n,f) for 14.7 MeV monoenergetic neutrons. Irradiations were performed at the EL3 reactor, at the Caliban and Prospero critical assemblies, and at the Lancelot electrostatic accelerator in CEA-Valduc. Fissions were counted in thin deposits using fission ionization chambers. The number of fission products of each species were measured by gamma spectrometry of co-located thick deposits.     

反应堆生产放射性同位素热源材料及其应用

可用于深空探测等领域的放射性同位素温差发电器利用半导体热电元件直接将放射性同位素的衰变热转化为电能.目前所使用的热源材料--放射性同位素~(238)Pu具有半衰期适中、热功率密度较高、α衰变易于辐射防护等特点,已获得较为广泛的应用.本文简要介绍利用反应堆生产~(238)Pu的途径以及分离、纯化过程.     

Studies on Fundamental Properties of Alumina Adsorbent for Radioisotope Production

Some fundamental properties of chromatographic alumina, used for the preparation of radioisotope generators and for the separation of fission products, were examined in connection with pretreatment procedures prescribed for the alumina to improve its properties in respect of alumina dissolution by certain eluants, its adsorption capacity for certain ions, and the effects of differences in the mode of storage.

The commercially available brand of chromatographic alumina used in this work showed an increase in its adsorption capacity and a decrease in alumina dissolution, when it was ignited for 10 hr at 400°C and treated thereafter with boiling dilute hydrochloric acid. When treated by the above procedure, the properties of the alumina in respect of dissolution and adsorption capacity were found to be unaffected by extended periods of storage under conditions corresponding to normal operation.

The pretreated alumina was made into column form. The dissolution of alumina with various kinds of eluants and the adsorption capacity for chloride, sulfate, fluoride, tellurite, tellurate, and molybdate ions were examined. The dissolution of alumina differed according to eluant, the sequence from larger to smaller dissolution being sodium hydroxide, hydrochloric acid, nitric acid, ammonium hydroxide and sodium chloride. It was also observed that the adsorption capacity and the elution conditions for tellurium and molybdenum were greatly affected by the pH of the solution and their valency.