Application of BGO Scintillators to Absolute Measurement of Neutron Capture Cross Sections between 0.01 eV and 10 eV

Applying a total energy absorption γ-ray detector composed of 12 bricks (5 × 5 cm², 7.5 cm thick) of BGO scintillators, the absolute measurement of capture cross sections for Au and Sb has been made in an energy region between 0.01 and 10eV using the linac time-of-flight method. Incident thermal neutron flux was absolutely determined by using the BGO detection system with a Sm sample. To extend the neutron flux measurement from the thermal neutron region to higher neutron energies, the ¹⁰B(n, αγ) reaction was applied. Absolute capture yield for the relevant capture sample was obtained by the saturated capture yield at a large resonance of the sample.

Gold was selected to investigate the application of the BGO detection system to the absolute measurement of the capture cross sections, since the ¹⁹⁷Au(n, γ)¹⁹⁸ Au reaction cross section is a well known standard one. The result of the ¹⁹⁷Au(n, γ)¹⁹⁸ Au reaction cross section showed good agreement with the evaluated data in JENDL Dosimetry File and ENDF/B-VI. Then, the detection system was applied to the Sb(n, γ) cross section measurement. Antimony has a large scattering-to-capture cross section ratio comparing to that of gold. The result showed good agreement with the evaluated data in JENDL-3.2 and ENDF/B-VI.     

Simplified Method for Calculation of Neutron Capture Transformation Effects of Fission Products on Decay Power

A simplified method is proposed for the calculation of the effects of neutron capture transformations of fission products (FPs) on the decay power of FPs. The decay power of FPs after shutdown changes by the neutron capture transformations of FP nuclides during reactor operation. It is proposed to calculate the neutron capture transformation effects considering the production of the following 7 nuclides ¹⁰³Ru, ¹³⁴Cs, ¹³⁶Cs, ^148mPm, ¹⁴⁸Pm, ¹⁵⁴Eu and ¹⁵⁶Eu by the neutron capture reaction of the direct mother nuclide alone giving a cumulative fission yield for the mother nuclide. The present method was assessed by com-paring the calculation results with the rigorous calculation results for the thermal-neutron fission of ²³⁵U irradiated between 1 and 5 yr in a light water reactor with thermal-nentron flux between 3 x 10¹³ and 6 x 10¹³ n/cm²·s and for the fast-neutron fission of ²³⁹Pu irradiated between 1 and 5 yr in a fast breeder reactor with total neutron flux between 3 x 10¹⁵ and 6 x 10¹⁵ n/cm²·s. It has been clarified that the present method can calculate the neutron capture transformation effects within the accuracy of ±1% of the decay power for the irradiation of 1yr and cooling time less than 10⁹s irrespective of fission type and neutron flux. The accuracy varies little with neutron flux but considerably with irradiation time. For a irradiation of 5 yr the present method can calculate the capture effect within the accuracy of +1% and -5% of the decay power. The accuracy can be improved to ±1% of the decay power with the simple correction factors.     

Particle Release from Vanadium Metal Surface by Neutrons

Sputtering yields from vanadium metal surface due to neutron irradiation were studied. A carefully prepared Pyrex glass tube, containing a vanadium foil as target and a polyethylene film pasted on a nickel plate as catcher, was sealed after evacuation, irradiated in a reactor, disassembled to take up the film, and the ⁵²V activity on it was counted for estimating the thermal neutron sputtering yield due to the recoil by(n, γ) reaction. The reactivation of the film gave the fast neutron sputtering yield. These values were found to be 2.3×10^?9 and 2.1×10^?1 respectively.     

Determination of thermal neutron cross-section and resonance integral for the As (n, γ)As reaction by activation method using Mn (n, γ)Mn as a monitor

Nuclear constants for use in reactor activation analysis especially (n, γ) cross-sections and absolute gamma intensities, are known to show a rather large scatter in literature. Thermal and resonance cross-sections for the ⁷⁵As (n, γ)⁷⁶As reaction is determined by the method of foil activation using ⁵⁵Mn (n, γ)⁵⁶Mn as a reference reaction. The experimental sample with and without a cadmium cover of 1-mm wall thickness was irradiated in the isotropic neutron field of the outer irradiation sites 7 of Ghana Research Reactor-1 facility which is a miniature neutron source reactor designed by the Chinese. The irradiation channel used has a neutron spectral parameter (α) found to be (0.037 ± 0.001). The induced activity in the sample was measured by gamma ray spectrometry with a high purity germanium detector. A standard solution of Arsenic was used for the analysis. The necessary correction for gamma attenuation, thermal neutrons and resonance neutron self-shielding effects were not taken into account during the experimental analysis because they were negligible. By defining cadmium cut-off energy of 0.55 eV, the result for ⁷⁵As (n, γ)⁷⁶As reaction was found to be: thermal neutron cross-section σ₀ = (4.28 ± 0.19) b and resonance integral I₀ = (61.88 ± 1.07) b.     

Beam Hole Configuration and Analysis of Medical Reactor for Boron Neutron Capture Therapy

This report presents an investigation of beam holes to be provided in a medical reactor for Boron Neutron Capture Therapy. The principal requirement for the beam holes is to deliver the therapeutic doses of thermal and epithermal neutrons in a modest time (30 to 60min) with minimal fast neutron and γ-contaminants. Characteristics of the beam holes have been evaluated by 2-dim. n-γ coupling S _N transport calculations. Reexaminations and revisions of the beam hole design have brought improvements of the characteristics, especially an increase of the thermal neutron flux at the horizontal thermal neutron beam port and a decrease of the fast neutron flux at the vertical epithermal neutron beam port. The design objectives for the beam holes set up in this study may be achievable even if the thermal power of the reactor is reduced from 2 to 1MW.     

Measurement of Thermal Neutron Capture Cross Section and Resonance Integral of the 129I(n, γ) 130I Reaction

To obtain fundamental data for research on the transmutation of nuclear wastes, the thermal neutron cross section and the resonance integral of the ¹²⁹I(n, γ)¹³⁰I reaction have been measured using an activation method. The neutron cross sections for the formation of the ground (5⁺) state and the isomeric (2⁺) state of ¹³⁰I were measured separately.

Six ¹²⁹I targets were irradiated for 10 min with thermal reactor neutrons; three of them containing 2.55- 2.61 kBq of ¹²⁹I were irradiated within a Cd capsule, and the other three targets containing 259–261 Bq of ¹²⁹I were irradiated without it. The Co/Al and Au/Al alloy wires were used to monitor the neutron flux and the fraction of the epithermal part (Westcott's epithermal index). The gamma-ray spectra from the irradiated samples were measured with a Ge detector.

The thermal neutron capture cross section (the 2,200 m/s neutron cross section) and the resonance integral of the ¹²⁹I(n, 7)¹³⁰I reaction were determined to be 12.5±0.5b and 15.6±0.7b for the formation of the ground state ^130gI(5+), 17.8±0.7b and 18.2±0.8b for the formation of the isomeric state ^130mI(2⁺), and 30.3±1.2b and 33.8±1.4b for the formation of ¹³⁰I (the sum of the 2⁺ and the 5⁺ states), respectively. The sum of the thermal neutron capture cross sections forming the 2⁺ and the 5⁺ states was 12% larger than the evaluated values of JENDL-3.2 and ENDF/B-VI and that reported by Roy et al. This discrepancy is explained by the population of the isomeric level.     

Measurement of thick-target gamma-ray production yields of the 7Li(p,p′)7Li and 7Li(p, γ)8Be reactions in the near-threshold energy region for the 7Li(p,n)7Be reaction

The gamma-ray production reactions, ⁷Li(p, p′)⁷Li and ⁷Li(p, γ)⁸Be, occur along with the neutron production reaction ⁷Li(p, n)⁷Be in a p-Li neutron source. These gamma-ray production reactions contribute to a patient's absorbed dose in boron neutron capture therapy (BNCT) when using a neutron beam from the ⁷Li(p, n)⁷Be reaction. The present work experimentally determined the thick-target gamma-ray production yields of the ⁷Li(p, p′)⁷Li and ⁷Li(p, γ)⁸Be reactions at incident proton energies of 1.670 and 1.870 MeV. The present results were compared with previous measurements. The gamma-ray production yield of ⁷Li(p, p′)⁷Li was measured to be 30%–50% smaller than as reported by previous studies. For the ⁷Li(p, γ)⁸Be reaction, the present thick-target yield is 30% smaller than one estimated from cross-section data measured in previous studies. The results must be included in future dose evaluation for BNCT using a p–Li neutron source.     

Preparation of Uranium-233

Thorium dioxide amounting to 1.3 kg was irradiated for 241 days at an average neutron flux of 6×10¹⁰n/cm²'sec, and allowed to stand for 1.2yr. This cooling lowered the γ-activity of the irradiated ThO₂ sufficiently to permit handling without special shielding. Nine milligrams of ²³³U were extracted with TBP in dodecane, and determined by α-ray counting. The mass ratio of the uranium isotopes obtained was determined by mass spectrometry. It was indicated that the ²³³U was contaminated with natural uranium.     

Neutron Capture Cross Section Measurement of Praseodymium in the Energy Region from 0.003 eV to 140 keV

The neutron capture cross section of praseodymium (¹⁴¹Pr) has been measured relative to the ¹⁰B(n,αγ) standard cross section in the energy region from 0.003 eV to 140 keV by the neutron time-of-flight (TOF) method with a 46-MeV electron linear accelerator (linac) of the Research Reactor Institute, Kyoto University (KURRI). An assembly of Bi₄Ge₃O₁₂ (BGO) scintillators was used for the capture cross section measurement. In addition, the thermal neutron cross section (2,200 m/s value) of the ¹⁴¹Pr(n, γ)¹⁴²Pr reaction has been also measured by an activation method at the heavy water thermal neutron facility of the Kyoto University Reactor (KUR). The thermal neutron flux was monitored with the ¹⁹⁷Au(n, γ)¹⁹⁸Au standard cross section. The above TOF measurement has been normalized to the current activation data (11.6±1.3 b) at 0.0253 eV.

The evaluated data in JENDL-3.3, ENDF/B-VI, and JEF-2.2 have been in general agreement with the current result, except that the JENDL-3.3 and the JEF-2.2 values are clearly lower than the measurement in the cross section minimum region from about 10 to 500 eV.     

Neutron Irradiation Effects on Superconducting Nb-Ti Wires for Fusion Reactor

A systematic study has been carried out on Nb-Ti wire of two different kinds of metallurgical structure, to examine the effects of neutron irradiation on the critical density J_c. The samples used were Nb-47.6 a/0 Ti (sample #A) and Nb-59.8a/0 Ti (sample #B), which were aged at 380°C for 0~10⁴min and irradiated to 1.3×10¹⁸n/cm² (E _n0.1 MeV). The sample temperature during the irradiation is believed to have been below 70°C. The values of J_c of both #A and #B aged up to 50 min were found to increase with irradiation. But when aged beyond 100 min, #B had its value of J_c lowered by the irradiation. The presence of Ti enriched precipitates such as α and ω phases in the samples was surmised from the behavior shown by the critical temperature T_c. The T_c of f #A and #B changed little by irradiation when aged not longer than 100 min, but with aging beyond 500 min, #B showed a decrease in its value of T_c. This decrease indicates that the Ti concentration in the matrix may have increased through radiation-induced breakup of the above-mentioned precipitates, which, in turn, would have brought about the reduction observed in J_c upon irradiation. It is concluded that superconducting Nb-Ti wire with J_c endanced by precipitation does not appear very resistant to neutron irradiation. This underlines the importance of the choice of superconducting materials to be used in fusion reactor magnets.     

Separation of Antimony-125 in Fission Products

Most of the known radioactive nuclides of antimony produced by neutron irradiation of uranium have fission yields below 1% and have half-lives below 60 days. An exception is ¹²⁵Sb with a half-life of 2.7 yr, which raise its relative importance among the fission products with lapse of time after irradiation, and after 1 yr of cooling, its radioactivity is no longer negligible. This circumstance has led to its being separated from such sources as fall-out. No studies have so far been reported on using the nitrate system for this separation, though it is utilized in the reprocessing of spent fuel and in the dissolution of uranium samples. The present work describes a method of separating ¹²⁵Sb from fission products with use made of silica gel—nitric acid system, and an example of its application to the separation of ¹²⁵Sb from the spent fuel of JPDR-1. The fuel was irradiated from Oct. 1963 to Sep. 1969. The amount of ¹²⁵Sb measured after separation was (1.7± O.19)×10^?1Ci/gU at June 1972.     

Production of the Isomeric State of 138Cs in the Thermal Neutron Capture Reaction 137Cs(n, γ)138Cs

In order to obtain precise data of the neutron capture cross section of the reaction ¹³⁷Cs(n, γ)¹³⁸Cs, the production probability of isomer state ^138mCs was measured in this work. Targets of about 0.37MBq ¹³⁷Cs were irradiated for 3 min in. the pneumatic tube facility (Pn-3) of Kyoto University Reactor (KUR). The 1,436 keV gamma;-ray emitted from both of ^138gCs and ^138mCs was measured. A ratio of the production probability between ^138gCs and ^138mCs was deduced from time dependence of peak counts of 1,436 keV γ-ray by making use of difference of half-lives of ^138gCs (33.41 min) and ^138mCs (2.91 min). The production probability of ^138mCs was obtained as 0.75plusmn;0.18 and this value revised the effective cross section upwards by 9plusmn;2percnt;. The effective cross section ô and the thermal neutron capture cross section σo were obtained as ô=0.29±0.02 b and σ=0.27±0.03b with taking into account the production of ^138mCs.     

Correction of the thermal neutron capture cross section of 241Am obtained by the Westcott convention

There is large discrepancy among the reported experimental data of the thermal neutron capture cross section of ²⁴¹Am, where the activation measurements provided larger cross sections than those in the time-of-flight ones. The Westcott convention has been widely used for the derivation of the thermal neutron capture cross section in the activation measurements. We have estimated that this large discrepancy is due to the existence of the resonances below the cadmium cut-off energy (E_Cd ～ 0.5 eV). By reviewing the Westcott convention, we developed the correction method taking account of the contribution of the resonances near or below E_Cd. The correction term was evaluated using the JENDL-4.0. Application of the present method successfully improved the existing discrepancy of the thermal capture cross section of ²⁴¹Am.     

Activation Analysis of Gadolinium Impurity in High Purity Europium Oxide

Abstract

An analytical method for determining Gd impurity in high purity Eu₂O₃ is proposed, which makes use of neutron activation and cation-exchange separation to examine its suitability as target material for the production of ^{152 m, 152, 154}Eu.

Long-term irradiation of an Eu sample resulted in ¹⁵³Gd activity amounting to 1.8 times that of the same nuclide produced from an equal quantity of Gd by (n, γ) reaction. This experimental value is quite consistent with that obtained by calculation under the assumption that the ¹⁶³Gd results from nuclear reaction on ¹⁵¹Eu induced by secondary neutrons.

For the accurate determination of Gd, Gd impurity was separated from the Eu sample prior to neutron irradiation in order to reduce the self-shielding effect in the Eu sample. Separation by cation-exchange with α-hydroxyisobutyrate (0.33 M, pH 3.77) made it possible to reduce the content of Eu in the Gd fraction below 7×10^?3%. This sufficed to assure that the ¹⁵⁹Gd content in ^152mEu was smaller than 1×10^?3%. For the determination of Gd content below 10⁴ ppm, however, should necessitate further purification of the irradiated Gd fraction.     

Radiolysis of Dicarboxylic Acid Lithium Salt with Neutron Irradiation

A series of dicarboxylic acids of Li—(CH₂)_n (COOLi)₂, n= 0–5, i.e., Li-oxalate, malonate, succinate, -glutalate, -adipate and -pimelate, was irradiated with mixed radiations in a nuclear reactor. The subject of study was the decarboxylation and the reaction of labelling with T atoms brought about by the exposure of the dicarboxylic acid lithium salts to neutrons.

The T-labelled hydrocarbons were determined by radiogaschromatography, while gaschro- matography was used to measure macroscopic amounts of compounds. The gaseous products of radiolysis produced from the neutron irradiation of dicarboxylic acid lithium salt were CO₂, CO and hydrocarbons.

The yield of CO₂ from the different dicarboxylic acid lithium salts was higher than obtained from γ-ray irradiation, and decreased with increasing number of methylene groups in the original salt. Also, while with γ-radiolysis, CO is only generated in negligibly small amounts, it was found quite significant in the present case of neutron irradiation. The yield of tritiated hydrocarbon was proportional to the number of methylene groups in the salt molecule. Hydrocarbon was mainly produced by the decarboxylation of the original compounds. Several kinds of degradated hydrocarbon were also observed from the mass peaks as also from activity peaks.     

Research of Radioisotope Production with Fast Neutrons, (V)

The preparation of carrier-free ⁶⁵Ni was studied using the reaction ⁶⁵CU(n,/p)⁶⁵.Ni in a nuclear reactor. High purity copper metal was irradiated for 20 min in the JRR-2 reactor. Nickel-65 formed in the copper target was separated by anion exchange method. About 1 μ ⁶⁵Ni per gram of copper was obtained by irradiation with a fast neutron flux of 9×10¹¹ n/cm=²/sec and thermal neutron flux of 3.6 ×10¹³ n/cm²/sec. The specific activity was greater than 3 mc/mg Ni which was over 600 times more than that produced by n, γ) reaction on nickel irradiated at the same time. However, the reaction yield of ⁶⁵Ni per gram of target by the (n, p) reaction was only 1/5,000 of that by the (n, γ) reaction, so that it proved that the method would be unsuitable for practical application.

The cross section for the reaction ⁶⁵Cu(n/p)⁶⁵Ni has been estimated semi-theoretically to be 0.6 mb, an experimental value of 0.2 mb was calculated from the results of the present work.     

Measurement of Thermal Neutron Capture Cross Section and Resonance Integral of the Reaction 133Cs(n, γ)134m,134gCs

The measurements of the thermal neutron (2,200 m/s neutron) cross section (σ₀) and the resonance integral (I ₀) of the ¹³³Cs(n, γ;) reaction were performed by an activation method to obtain fundamental data for research on the transmutation of nuclear wastes. The cross sections for the formations of the isomeric state ^134mCs and the ground state ^134mCs were measured respectively by following the behavior of the γ-ray counting rate after the irradiation.

The thermal neutron capture cross sections and the resonance integrals of the ¹³³Cs(n, γ) reaction were determined to be 2.70±0.13 b and 23.2±1.8 b for the formation of the isomeric state ^134mCs, and 26.3±1.0 b and 275±16 b for the formation of the ground state of ^134gCs. The results for the reaction ¹³³Cs(n, γ)^134m+gCs were 29.0±l.0 b and 298±16 b, respectively. As for the thermal neutron capture cross section for the formation of ^134m+gCs, the evaluated value (29.00 b) of JENDL-3.2 agreed with the present result. The reported value by Baerg et al. was in good agreement with the present result within the limits of error on the thermal neutron capture cross section for ^134mCs. On the other hand, the resonance integral for ^134m+g Cs was 32% smaller than the experimental value by Steinnes et al.     

Verification Experiment of Threshold Detector 107Ag(n,n′)107mAg on “YAYOI” Standard Neutron Field

The validity of the ¹⁰⁷Ag(n, n′)^107mAg reaction for reactor neutron dosimetry has been examined through an irradiation experiment using the standard neutron field in the “YAYOI”, a fast neutron source reactor. The test foil used was 99.0% ¹⁰⁷Ag enriched silver, which was irradiated along with five reference foils already well evaluated. The reaction rate of 107Ag was determined by K X-ray counting with a high-purity Ge detector. The cross section for the ¹⁰⁷Ag(n, n′) ^107mAg reaction was compiled from the excitation cross section data of ¹⁰⁷Ag published in ENDF/B-IV, in combination with the decay scheme of ^107mAg listed in the “Table of Isotopes“. Using the neutron spectrum which had previously been precisely measured, it was found that the compiled activity of ^107mAg was underestimated by 12.8±5%. This error is attributed to significant underestimation involved in the cross section of the ¹⁰⁷Ag(n, n′)^107mAg reaction.     

Measurement of Effective Capture Cross Section of Americium-243 for Thermal Neutrons

The effective capture cross section of ²⁴³Am for thermal neutrons was measured with an activation method. A sample of ²⁴³Am was irradiated for 10 hrs at Kyoto University Reactor, KUR. After the irradiation, the sample was cooled for one month. In the cooling time, ^244mAm and ^244gAm produced by the irradiation decayed out to ²⁴⁴Cm. The α rays emitted from ²⁴³Am and ²⁴⁴Cm were measured with a silicon surface barrier detector. The neutron flux at the irradiation position was monitored using Au/Al and Co/Al wires. The effective capture cross section was deduced as 174.5±5.3b from the α-ray counts and the neutron flux. The quantity r√T/T₀ in Westcott's convention was 0.037±0.004. The present result was compared with the effective capture cross sections from the JENDL-3.3 and the Mughabghab evaluations.