Application of Nuclear Recoil to Radioisotope Enrichment of Copper-64

A new target material has been proposed for nuclear recoil enrichment of ⁶⁴Cu. This radionuclide was hitherto produced by tedious chemical processing using concentrated sulfuric acid solution of neutron irradiated copper phthalocyanine. Simplified chemical processes using a water-soluble copper phthalocyanine derivative, octacarboxyphthalocyanato copper (II), have been adopted in this work. Two methods of separation involving solvent ex-traction and precipitation have been tested. Higher specific activity of ⁶⁴Cuhas been obtained for the extraction method than for the precipitation method. Feasibility of the extraction method in ⁶⁴Cu enrichment is discussed.     

Characteristics of Fabricated Neutron Detectors Based on a SiC Semiconductor

Wide-band-gap semiconductors such as SiC, AlN, and GaN are promising materials for harsh environment applications due to their high-temperature operation capability. Two types of PIN-type semiconductor neutron detectors based on SiC were designed and fabricated for nuclear power plant (NPP) applications such an in-core reactor neutron flux monitoring and safeguarding nuclear materials. One is for fast neutron detection and the other, which was evaporated with ⁶LiF, is for thermal neutron detection. In this study, preliminary tests, such as the determination of I-V and alpha responses, were performed. Reaction probabilities with respect to neutron energies were also calculated by using an MCNPX code for comparison with the experimental results. Responses of the neutrons were measured at the Ex-core Neutron irradiation Facility (ENF) of the High-flux Advanced Neutron Application Reactor (HANARO) research reactor at the Korea Atomic Energy Research Institute (KAERI). Pulse height spectra and count rates were measured with respect to the neutron fluxes from 1:6 × 10⁶ n/cm²·s to 1:9 × 10⁷ n/cm²·s. Also, a 0.99 root-mean-square value of linearity against the fluxes to the count rates was obtained with the fabricated neutron detectors. For a thermal neutron detector, a 3.3% detection efficiency was obtained.     

Stopping power measurements for O, F and Si ions in Mylar by a transmission technique

Electronic energy loss of charged particles in materials is a fundamental process responsible for the unique response of materials in applications of advanced nuclear power, radiation detectors and advanced processing of electronic devices. In this study, stopping powers of ¹⁶O, ¹⁹F and ²⁸Si heavy ions crossing thin Mylar foils have been determined in transmission geometry. The energy loss was measured over a continuous range of energies from 1.6 to 5.5 MeV/n (MeV per nucleon) using the data that was tagged by a surface barrier detector (SBD) with and without stopping foils. We have compared the obtained stopping values to those predicted by SRIM-2008 computer code, ICRU-73 stopping data tables and MSTAR calculations. The effective charge values of these heavy ions have been also deduced from the experimental set of data.     

Remarkable improvement of the signal-to-noise ratio of Mn/Fe in-beam Mössbauer spectroscopy

In-beam Mössbauer spectroscopy utilizing unstable ⁵⁷Mn beams is a powerful method to extract physical and chemical properties at the atomic scale. A parallel plate avalanche counter (PPAC), optimized to detect conversion electrons generated by the Mössbauer effect, can be employed to suppress higher-energy background γ rays. However, β rays are emitted by the ⁵⁷Mn parent nuclei of ⁵⁷Fe, which can significantly degrade the spectrum quality. In the present work, we have developed a new anti-coincidence-detection system with a thin plastic scintillation counter (0.5 mmt), which can be used to detect β rays and reject them from the recorded PPAC events. To demonstrate the anti-coincidence system, we carried out Mössbauer spectroscopy utilizing ⁵⁷Mn nuclei that were implanted into a non-magnetic aluminum metal plate at room temperature. Using the anti-coincidence method, we obtained a typical Mössbauer spectrum of high quality, despite a very low number of implanted ⁵⁷Mn atoms, of ∼5 × 10⁹. The signal to noise ratio of the obtained spectrum was increased remarkably, and the relative peak height above the baseline increased from 10% to 220% using the anti-coincidence method. The developed detection system is applicable to investigation of in situ properties, and avoids the potentially problematic agglomeration of probes in a sample.     

Evaluation of 17O(n,a)14C Cross Section

The ¹⁷O(n, a)¹⁴C cross section has been evaluated for incident neutron energies from 10^?5eV to 20MeV for accurate calculation on the ¹⁴C production in nuclear reactors. Evaluation was based on the single-level Breit-Wigner formula for the thermal to resonance energy regions. In the higher energy region, a multi-step evaporation model code PEGASUS was used, and the results were normalized using the ¹⁶O(n, a) cross section ratio of JENDL-3 evaluation to PEGASUS calculation. Results are given in tabular and graphical forms, and also as one-group cross sections using typical BWR, PWR and FBR spectra of ORIGEN-2.     

Accelerator mass spectrometry of U at low energies

The performance of the compact ETH-TANDY system for accelerator mass spectrometry measurements of ²³⁶U is presented. Despite the low ion energies of around 1.2 MeV we can demonstrate a background level that is comparable to larger facilities. The careful ion-optical design of the high-energy spectrometer leads to a high suppression of neighboring isotopes sufficient to measure samples with isotopic ratios of ²³⁶U/²³⁸U > 10^-11 the ion chamber only, as demonstrated by systematic investigations with different slit settings and time-of-flight measurements. Additionally, a high overall efficiency is achieved due to a high transmission through the accelerator.     

Recent advances in AMS of Cl with a 3-MV-tandem

Accelerator mass spectrometry (AMS) of ³⁶Cl (t_1/2 = 0.30 Ma) at natural isotopic concentrations requires high particle energies for the separation from the stable isobar ³⁶S and was so far the exclusive domain of tandem accelerators with at least 5 MV terminal voltage. Using terminal foil stripping and a detection setup consisting of a split-anode ionization chamber and an additional energy signal from a silicon strip detector, a ³⁶S suppression of >10⁴ at 3 MV terminal voltage was achieved. To further increase the ³⁶S suppression energy loss straggling in various counter gases (C₄H₁₀, Ar-CH₄ and C₄H₁₀-Ar) and the effect of “energy focusing” below the maximum of the Bragg curve was investigated. The comparison of experimental data with simulations and published data yielded interesting insights into the physics underlying the detectors. Energy loss, energy straggling and angular scattering determine the ³⁶S suppression. In addition, we improved ion source conditions, target backing materials and the cathode design with respect to sulfur output and cross contamination. These changes allow higher currents during measurement (³⁵Cl⁻ current ≈ 5 μA) and also increased the reproducibility. An injector to detector efficiency for ³⁶Cl ions of 8% (16% stripping yield for the 7+ charge state in the accelerator, 50% ³⁶Cl detection efficiency) was achieved, which can favorably be compared to other facilities. The memory effect in our ion source was also thoroughly investigated. Currently our measured blank value is ³⁶Cl/Cl ≈ 3 × 10⁻¹⁵ when samples with a ratio of 10⁻¹¹ are used in the same sample wheel and ³⁶Cl/Cl ≈ 5 × 10⁻¹⁶ if measured together with samples with a ratio of 10⁻¹² or below. This is in good agreement with the lowest so far published isotope ratios around 5 × 10⁻¹⁶ and demonstrates that 3 MV tandems can achieve the same sensitivity for ³⁶Cl as larger machines.     

贫铀/聚乙烯交替球壳中裂变反应率的测量与分析

为校验次临界能源堆的概念设计,采用活化法在贫铀/聚乙烯球壳交替装置上开展14 MeV中子学积分实验。用HPGe探测器测量²³⁸U(n,f)及²³⁵U(n,f)反应的裂变碎片¹⁴³Ce衰变产生的293.3 keV特征γ射线,得到装置中与入射D粒子束成90°方向上的²³⁸U(n,f)及²³⁵U(n,f)反应率分布,相对不确定度为5.1%~6.9%。采用MCNP5程序在ENDF/B-Ⅵ库下进行模拟计算,计算结果较实验结果高约5%。     

Production of high-purity medical radio isotope 64Cu with accelerator-based neutrons generated with 9 and 12 MeV deuterons

We conducted a feasibility study for producing a high-purity medical radioisotope ⁶⁴Cu from natural zinc with accelerator-based neutrons. ⁶⁴Cu isotopes were produced via the ⁶⁴Zn(n,p) reaction. The accelerator-based neutrons were generated via the C(d,n) reaction using low-energy deuterons of 9 and 12 MeV on a 1-mm-thick carbon target. First, the production purity was estimated using the evaluated nuclear data library JENDL-4.0 and our previously measured thick target neutron yield. We found that even when natural zinc was used as the starting material, significantly high-purity ⁶⁴Cu could be obtained. Next, irradiation experiments for producing ⁶⁴Cu using natural zinc were conducted at Kyushu University Tandem Laboratory, with the amounts of ⁶⁴Cu isotopes and other gamma-emission nuclides measured by a high-purity germanium detector. As a result, high-purity ⁶⁴Cu isotopes of 1.11(49) × 10⁰ and 3.70 (17) × 10⁰ Bq/g/μC were produced with incident deuteron energies of 9 and 12 MeV, respectively.     

Isotope dilution inductively coupled plasma mass spectrometry for determination of 126Sn content in spent nuclear fuel sample

The ¹²⁶Sn content in a spent nuclear fuel solution was determined by isotope dilution inductively coupled plasma mass spectrometry (ID-ICP-MS) for its inventory estimation in high-level radioactive waste. A well-characterized irradiated UO₂ fuel sample dissolved in a hot cell was used as a sample to evaluate the reliability of the methodology. Prior to the ICP-MS measurement, Sn was separated from Te (¹²⁶Te), which causes major isobaric interference in the determination of ¹²⁶Sn content, along with highly radioactive coexisting elements, such as Sr (⁹⁰Sr), Y (⁹⁰Y), Cs (¹³⁷Cs) and Ba (^137m Ba), using an anion-exchange column. The absence of counts attributed to Te at m/z = 125, 128, and 130 in the Sn-containing effluent (Sn fraction) indicates that Te was completely removed from the anion-exchange column. After washing, Sn retained on the column was readily eluted with 1 M HNO₃ accompanied with approximately 80% of the Cd and 0.03% of the U in the initial sample. Owing to the presences of Cd and U in Sn fraction, the measurements of ¹¹⁶Sn and ¹¹⁹Sn were affected by the isobaric ¹¹⁶Cd and the doubly charged ²³⁸U²⁺ion, resulting in the positive bias of the determined values. With the exception of the isotopic ratios including ¹¹⁶Sn and ¹¹⁹Sn, ¹¹⁷Sn/¹²⁶Sn, ¹¹⁸Sn/¹²⁶Sn, ¹²⁰Sn/¹²⁶Sn, ¹²²Sn/¹²⁶Sn and ¹²⁴Sn/¹²⁶Sn were successfully determined and showed good agreement with those obtained through ORIGEN2 calculations. The measured concentration of ¹²⁶Sn in the spent nuclear fuel sample solution was 0.74 ± 0.14 ng/g, which corresponds to 23.0 ± 4.5 ng per gram of the irradiated UO₂ fuel (excluding the presence of ¹²⁶Sn in the insoluble residue). The results reported in this paper are the first experimental values of ¹²⁶Sn content and isotope ratios in the spent nuclear fuel solution originating in spent nuclear fuel irradiated at a nuclear power plant in Japan.     

Determination of Tehran Research Reactor power by N gamma detection

The radioisotope ¹⁶N is produced by the interaction of fast neutrons with ¹⁶O in water reactor coolant. This radioisotope emits at the two major gamma ray energies of 6.13 MeV and 7.1 MeV. Exploiting the linear relation between the number of gamma particles versus the reactor power change, the reactor power is determined by detecting and counting the emitted gammas. In this work, for the detection of gammas to measure the reactor power, two different methods are employed. First, by NaI(Tl) scintillator detector and second, by assembly of ten GM detectors. The obtained results confirm that the number of emitted gammas is proportional to the change in reactor power as shown by different monitoring systems such as UIC, CIC, FC, Cherenkov and thermal power. Both of the applied methods are shown to give reliable results for reactor power above 20 kW. Both systems, having been calibrated, are being used as monitoring systems of power in Tehran Research Reactor. These systems are usable in other research reactors and possibly in power reactors as well.     

Determination of the 129I transmutation rate and 129I(n, 2n)128I correction using gas mass spectrometry

Two groups of ¹²⁹I and ¹²⁷I targets were analyzed using a gas quadrupole mass spectrometer (QMS) to determine the transmutation rates via the melting method. Sodium iodide was chosen to make the target. The iodine composition in the ¹²⁹I targets is 82.7% ¹²⁹I and 17.3% ¹²⁷I. The transmutation rate of the ¹²⁹I(n, γ)¹³⁰I reaction was determined by measuring the ¹³⁰Xe with QMS. An equivalent corrective method was brought out to correct the ¹²⁹I(n, 2n)¹²⁸I branch which is interfered with by the ¹²⁷I(n, γ)¹²⁸I reaction. And the correction formula was deduced in theory. For very little ¹²⁸Xe from the ¹²⁹I(n, 2n)¹²⁸I reaction, the equivalent corrective method could not be suitable here. However, it is suitable for the mass of ¹²⁸Xe from ¹²⁹I(n,2n)¹²⁸I reaction that reaches the accurately detective level of the mass spectrometry.     

Precise Intensity Measurements in the 14N(n, γ)15N Reaction as a γ-Ray Intensity Standard up to 11MeV

The main γ-ray intensities in the ¹⁴N(n, γ)¹⁵N reaction were determined at an accuracy of 0.3–1.0% based on the intensity balance principle. Measurements were performed at the supermirror neutron guide tube at the Kyoto University reactor (KUR). A liquid nitrogen target and a deuterated melamine (C₃D₆N₆) one were used. We prepared a full-energy-peak efficiency curve by combining the measured and calculated efficiency curves. The previous values widely used as intensity standards agreed with the present results within 4–5% in the 2–11 MeV region, but the ratio to the present values showed a monotonic decrease with the increase in _-ray energy. The results reported by Belgya in 2006 agreed with the present one within 2–3% in the 2–8MeV region; nevertheless, the ratio showed a small oscillation versusγ-ray energy.     

Activation cross sections of the Yb(p,xn)Lu reaction for indirect production of the therapeutic radionuclide Yb

Activation cross sections of the ^natYb(p,xn)¹⁶⁹Lu reaction have been measured for the first time up to 70 MeV to investigate the production possibility of the radionuclide ¹⁶⁹Yb through decay of its parent ¹⁶⁹Lu. The cross sections were measured using the stacked foil irradiation technique and gamma spectrometry. The experimental data were compared with the results of the ALICE-IPPE theoretical model code. Different production routes were compared for the internal radiotherapy related radioisotope ¹⁶⁹Yb. Above 30 MeV proton energy the integral yield of the ^natYb(p,xn)¹⁶⁹Lu reaction is higher than that of the earlier investigated ¹⁶⁹Tm(p,n)¹⁶⁹Yb, ¹⁶⁹Tm(d,2n)¹⁶⁹Yb, ^natEr(α,xn)¹⁶⁹Yb, ^natYb(α,x)¹⁶⁹Lu and ^natHf(p,x)¹⁶⁹Lu reactions at the equivalent particle energies.     

Feedback on neutron capture cross sections of 238Pu and 241Am from analyses of measured isotopic compositions of irradiated LWR fuels and MOX core physics experiments

The atomic fractions of ²³⁸Pu and ²⁴¹Am in MOX fuels recycled in light water reactors are 1% to 2% and not significant compared with those of major Pu isotopes. On the other hand, recent evaluated nuclear data libraries, such as JENDL-4.0 and JEFF-3.2, give noticeably different thermal and epithermal neutron capture cross sections for ²³⁸Pu and ²⁴¹Am. The thermal neutron capture cross sections of ²³⁸Pu and ²⁴¹Am in JEFF-3.2 are 31% and 9% larger than those of JENDL-4.0, respectively. This paper shows the effect of the differences in the neutron cross sections on analysis results of two different integral experiments. The first is the isotopic compositions of ²³⁸Pu on UO₂ and MOX fuels irradiated in BWR and PWR, and the second is the critical experiments of the water moderated cores fully loaded with MOX fuels. The former was analyzed by using the continuous energy Monte Carlo burnup calculation code MVP-BURN and the latter by the continuous energy Monte Carlo calculation code MVP. The comparisons between the calculated and measured results indicate that the most likely thermal and epithermal neutron capture cross sections of ²³⁸Pu and ²⁴¹Am should be around at the middle between those of JEFF-3.2 and JENDL-4.0.     

Measuring the Photoneutrons Originating from D(γ, n)H Reaction after the Shutdown of an Operational BWR

The purpose of this study is to measure the photoneutron emission rate after the shutdown of an operational BWR. The photoneutrons originate from the D(γ, n)H reaction in the moderator region, while the high-energy gamma rays are generated from ¹⁴⁰Ba-¹⁴⁰La transient equilibrium of fission products in irradiated fuel. The photoneutron emission rate is measured by means of the photoneutron signal ratio of the Start-up Range Neutron Monitor (SRNM). The ratio is defined as a ratio of the photoneutrons to neutrons originating from spontaneous fission and the oxygen (γ, n) reaction of actinides (²⁴²Cm, ²⁴⁴Cm, etc.) in irradiated fuel. The principle of the measurement of the photoneutron signal ratio is the large difference between the decay constants of the ¹⁴⁰Ba-¹⁴⁰La transient equilibrium and those of the actinides. The measurement of the SRNM signal was continuously carried out over several months, and the photoneutron signal ratio was evaluated by using the least-squares method to fit a theoretical model to SRNM signal data. The measurements were performed in the middle of the cycle at three BWR cores. Comparisons of the measured photoneutron signal ratio and the calculated one showed reasonable agreement. This demonstrates the validity and usefulness of the measurement. The absolute value of photoneutrons in the SRNM signal ranged from approximately 1 to 35 counts per second during a five-day cooling period after shutdown. Converting the absolute value to the relative fraction of photoneutrons in the SRNM signal results in a range from approximately 2 to 50%.     

Excitation functions of (p,x) reactions on natural nickel up to 40 MeV

Production cross-sections of the residual radionuclides for the ^natNi(p,x)^{55,56,57,58 m + g}Co, ^56,57Ni nuclear reactions were measured up to 40 MeV by using a stacked-foil activation technique at the MC-50 cyclotron of the Korea Institute of Radiological and Medical Sciences. The results were compared with the reported experimental data as well as the theoretical calculations based on the TALYS and the ALICE-IPPE codes. The present results are in general good agreement with the other experimental data and calculated results. The integral yields for thick target were also deduced from the measured excitation functions of the produced radionuclides. The present experimental results will play an important role in enrichment of the literature data base for proton-induced reactions on natural nickel leading to various applications.     

Calcium isotope separation by band chromatography using 18-crown-6-ether resin

To develop the ⁴⁸Ca enrichment process, a feasibility study on a band chromatography was made using 9 M HCl solution and crown ether resin synthesized in porous silica beads. Prior to the chromatographic experiments, distribution coefficients, Kd, of Ca²⁺ and Sr²⁺ were measured at different concentrations of these ionic species. The frontal boundary of the chromatography was made by a usual manner of the breakthrough mode of calcium feeding, and the rear boundary was made by introducing strontium as a following ion on the basis of the observed Kd values. It was confirmed that the heavy isotope ⁴⁸Ca was depleted in the rear boundary region, while ⁴⁸Ca was enriched in the front boundary region. The values of separation coefficient ε (= α – 1) in three chromatographic operations at different temperatures were observed as 2 × 10^?3 ~ 3 × 10^?3. The separation coefficients observed in the front boundary regions, where ⁴⁸Ca was enriched, agreed with those observed in the rear boundary regions, where ⁴⁰Ca was enriched.