Estimation of Covariance Matrices for Nuclear Data of 237Np, 241Am and 243Am

Covariance matrices were estimated for the fission and capture cross sections and the numbers of neutrons per fission of ²³⁷Np, ²⁴¹Am and ²⁴³Am given in JENDL-3.3. GMA and KALMAN codes were applied to estimate them for the fission and capture cross sections, respectively. In the low energy region, the errors of resonance parameters were given. The covariance matrices for the numbers of prompt neutrons per fission (V_p ) were evaluated by assuming a linear equation. For the delayed neutrons (v_d ), only their standard deviations were estimated. The results were compiled in the ENDF-6 format and merged with JENDL-3.3.     

Radiochemical Determination of Neutron Capture Cross Sections of 241Am

The thermal neutron capture cross sections and the neutron capture resonance integrals of ²⁴¹Am leading to the production of the isomer ²⁴²Am and the ground-state ^242gAm were measured radiochemically by the Cd-ratio technique with neutron flux monitors of Co/Al and Au/Al alloy. Highly-purified ²⁴¹Am targets were irradiated in an aluminum capsule by using JMTR. The neutron fluxes and their epithermal neutron fractions were determined by measuring γ-rays of ⁶⁰Co and ¹⁹⁸Au. The yields of ^242mAm and ^242gAm were decided by analyzing growth and decay curves of the α-ray activity ratios ²⁴²Cm/²⁴¹Am. The resultant thermal neutron capture cross sections are 85.7 ± 6.3 b and 768 ± 58 b for ^242mAm and ^242gAm, and the resonance integrals 114±7 b and 1,694±146 b, respectively. The differences between the present results and the evaluated values by Mughabghab are 38–59%. The isomeric ratios, g/(m+g), of 0.90±0.09 for thermal neutrons and 0.94±0.11 for epithermal neutrons are, however, almost consistent with evaluated values.     

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.     

Rapid Transmutation of High-Level Nuclear Wastes in a Catalyzed Fusion-Driven System

The aim of this study is to investigate the high-level waste (HLW) transmutation potential of fusion-driven transmuter (FDT) based on catalyzed D–D fusion plasma for various fuel fractions. The Minor actinide (MA) (²³⁷Np, ²⁴¹Am, ²⁴³Am and ²⁴⁴Cm) and long-lived fission product (LLFP) (⁹⁹Tc, ¹²⁹I and ¹³⁵Cs) nuclides discharged from high burn-up pressured water reactor-mixed oxide spent fuel are considered as the HLW. The volume fractions of the MA and LLFP are raised from 10 to 20% stepped by 2% and 10 to 80% stepped by 5%, respectively. The transmutation analyses have been performed for an operation period (OP) of up to 6 years by 75% plant factor (η) under a first-wall neutron load (P) of 5 MW/m² by using two different computer codes, the XSDRNPM/SCALE4.4a neutron transport code and the MCNP4B Monte Carlo code. The numerical results bring out that the considered FDT has a high neutronic performance for an effective and rapid transmutation of MA and LLFP as well as the energy generation along the OP.     

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.     

Physical investigation for neutron consumption and multiplication in fusion–fission hybrid test blanket module

Inelastic scattering of high energy fusion neutrons does affect the performance of fusion blanket based on the choice of different materials. It will also affect the behavior of source neutrons in a subcritical fusion fission hybrid blanket and consequently the transmutation and tritium breeding performance. A fusion fission hybrid test blanket module (HTBM) is designed which is presumed to be tested in a large sized tokamak and plasma neutron source is similar to ITER. In this preliminary design of HTBM the neutron source and loss factors are computed for the detailed neutronic performance analysis. The neutronic analysis of hybrid blanket module is performed for five different TRU fuel types: TRU-Zr, TRU-Mo, TRU-Oxide, TRU-Carbide and TRU-Nitride. In this module design, it is aimed to burn and transmute the TRU nuclides from high-level radioactive waste of PWR spent fuel. The effect of TiC reflector on transmutation and tritium breeding performance of HTBM is also quantified. MCNPX is used for neutronic computations. Neutron spectrum, capture to fission ratio and waste transmutation ratio of each fuel type are compared to evaluate their waste transmutation performance. Tritium breeding ratio is also compared for two coolant options: Li and LiPb eutectic.     

Wide Range Wave-Number Spectra of Place-to-Place Variations in Terrestrial Gamma-Ray Exposure Rate

The precipitation behavior of Pu. Np and Am during the denitration of high-level radioactive liquid waste (HLW) by formic acid was studied using a simulated HLW. The dissolution of the precipitate formed in denitrated HLW was also studied using oxalic acid to recover transuranium (TRU) elements from the precipitate. In the denitration, the precipitated fractions of TRU elements increased with decreasing acidity of the denitrated HLW. In the denitration at [HCOOH]/[HNO₃]=1.5, which was adopted in the partitioning process developed in JAERI, the precipitated fractions of Np and Am were only 0.6% and 0.06%, respectively, whereas that of Pu was 90 %. The precipitation fractions of Pu and Np did not depend on their concentrations in the range of 6x10^?5–6x10^?4 M for Pu and 10^?5–10^?3 M for Np. Plutonium was not precipitated itself by polymerization or hydrolysis but coprecipitated with other elements such as Mo and Zr. It was found that the precipitate formed during the denitration of 1 l of HLW could be dissolved in a 800 ml of 0.5 M oxalic acid solution.     

Neutron Excess Generation by Fusion Neutron Source for Self-Consistency of Nuclear Energy System

The present day fission energy technology faces with the problem of transmutation of dangerous radionuclides that requires neutron excess generation. Nuclear energy system based on fission reactors needs fuel breeding and, therefore, suffers from lack of neutron excess to apply large-scale transmutation option including elimination of fission products. Fusion neutron source (FNS) was proposed to improve neutron balance in the nuclear energy system. Energy associated with the performance of FNS should be small enough to keep the position of neutron excess generator, thus, leaving the role of dominant energy producers to fission reactors. The present paper deals with development of general methodology to estimate the effect of neutron excess generation by FNS on the performance of nuclear energy system as a whole. Multiplication of fusion neutrons in both non-fissionable and fissionable multipliers was considered. Based on the present methodology it was concluded that neutron self-consistency with respect to fuel breeding and transmutation of fission products can be attained with small fraction of energy associated with innovated fusion facilities.     

First nuclear transmutation of 237Np and 241Am by accelerator-driven system at Kyoto University Critical Assembly

This study demonstrates, for the first time, the principle of nuclear transmutation of minor actinide (MA) by the accelerator-driven system (ADS) through the injection of high-energy neutrons into the subcritical core at the Kyoto University Critical Assembly. The main objective of the experiments is to confirm fission reactions of neptunium-237 (²³⁷Np) and americium-241 (²⁴¹Am), and capture reactions of ²³⁷Np. Subcritical irradiation of ²³⁷Np and ²⁴¹Am foils is conducted in a hard spectrum core with the use of the back-to-back fission chamber that obtains simultaneously two signals from specially installed test (²³⁷Np or ²⁴¹Am) and reference (uranium-235) foils. The first nuclear transmutation of ²³⁷Np and ²⁴¹Am by ADS soundly implemented by combining the subcritical core and the 100 MeV proton accelerator, and the use of a lead-bismuth target, is conclusively demonstrated through the experimental results of fission and capture reaction events.     

Evaluation of Neutron Nuclear Data for Z>88 Minor Nuclides

Neutron nuclear data for 15 minor nuclides (Z>88) have been evaluated in the energy range of 10^?5 eV–20 MeV. Since only few experimental data are available, the present evaluation was mainly based on the systematics of the data from neighboring nuclides and also optical and statistical model calculations. The evaluations have been carried out for neutron cross sections of total, elastic scattering, inelastic scattering, (n, 2n), (n, 3n), (n, 4n), fission and capture reactions. In addition, angular and energy distributions of the emitted neutrons and average number of the emitted neutrons per fission were also evaluated. The results were compiled in the ENDF/B-V format and stored in the JENDL-3.     

Fission Spectrum Averaged Cross Sections of Some Threshold Reactions Measured with Fast Reactor “YAYOI”

Abstract

Fission spectrum averaged cross sections of twenty one threshold reactions were measured in the core center of YAYOI which was a fast neutron source reactor. Fast neutron spectrum in the core was experimentally determined by using a set of activation foils and micro-fission counters, prior to the cross section measurement. It was found that the shape of the fast neutron spectrum was approximately the same as that of fission neutrons above about 2MeV. This fact was also supported by theoretical calculation.

Since this neutron field has scarce thermal and epithermal neutrons, measurement of nuclei produced by threshold reactions is not affected by (n, γ) reactions which are induced by thermal and epithermal neutrons. Moreover, considerably high fast neutron flux (about 5 x 10¹¹n/cm²·sec) enables to measure cross sections of small values.

The results in general agreed with the previous values obtained in a reactor core or with a fission plate within an experimental error, while they were systematically smaller by about 10% than those recommended by Fabry. The measured values are also compared with the results calculated by Pearlstein based on a statistical model.     

活化法测量散裂靶中子能谱的实验验证

散裂靶中子的能谱对加速器驱动次临界系统的倍增因数和嬗变率等影响很大,计算表明散裂靶中子谱在MeV能区与裂变中子谱相近。本文利用活化法测量临界装置的泄漏中子谱和中子注量率,提出了用In、Al、Mg、Ti、Au、Zn、Ni、Rh、Fe和Co等活化箔测量散裂靶中子能谱和中子注量率的方案。结果表明,将活化箔在散裂靶中子场中辐照5h,中子注量最高达5×1014 cm-2量级,辐照后1h内取出活化箔,根据半衰期的长短安排测量顺序,可测量散裂靶的中子能谱和中子注量率。     

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.     

Improvement of Neutron Source Introduction Method for Absolute Measurements of Low Reactor Power

The neutron source introduction method was applied to absolute measurements of low reactor power at the Static Experiment Critical Facility STACY. To obtain the effective neutron source intensity more accurately, which is a key parameter for the source introduction method, the neutron source is newly defined as fission neutrons from the first fission reaction caused by neutrons emitted from the external neutron source. To obtain the newly defined effective neutron source intensity, the probability that a neutron from the external neutron source causes a fission reaction is calculated using the Monte Carlo code MCNP. This calculation took into consideration the three-dimensional complicated core structures. Furthermore, the fission reaction distribution, fundamental mode forward and adjoint flux distribution in a critical state were calculated using the three-dimensional transport code THREEDANT. Following the principle of the neutron source introduction method, an external neutron source was inserted near the STACY core tank and the reactor power was measured. The reactor powers by the neutron source introduction method were in good agreement with the ones from the analyses of the FP activity generated by high power operation.     

Long-Lived-Radionuclide Transmutation Scenarios

The results of multigroup calculations of continuous irradiation of Np, Am, and Cm in VVÉR-1000, PHWR-880, Superphoenix-1200, BREST-1000, and ÉLYaU-800 reactors are used to compare transmutation efficiency. The sources of continuous replenishment for the transmuters were Np, Am, and Cm extracted after a 3-yr holding period from the VVÉR and Superphoenix spent fuel. It is shown that the most effective transmuter is a subcritical liquid-fuel ÉLYaU system with an average thermal-neutron flux in the blanket 2·10¹⁵ sec^–1·cm^–2. For solid-fuel reactors, the continuous-irradiation model makes it possible to describe approximately the multiple transmutation regime. In the foreseeable future, one-time transmutation of Np, Am, and Cm in a solid-fuel reactor followed by storage in a long-term storage facility is feasible. The results of different computational variants for such regimes show that for transmutation in 10 yr in PHWR the radiotoxicity of Np, Am, and Cm accumulated in long-term storage reaches an equilibrium in no longer than 100 yr.     

Measurement and Analysis of Fission Rate Distributions of 237Np and 238U in a Polyethylene System with a 65 MeV Neutron Source

Fission rates of ²³⁷Np and ²³⁸U in a polyethylene (CH₂) system were measured with a 65MeV quasi monoenergetic neutron source. Relative fission rate distributions dependent on polyethylene thickness up to about 70 cm were obtained for both nuclides with the experimental error within 7%. The present experiment was analyzed by the NMTC/JAERI code that has been employed for designing accelerator-driven transmutation systems. The fission rates of both ²³⁷Np and ²³⁸U calculated by the NMTC/JAERI did not agree with the experimental ones. The C/E values for both were about 2.0 at 71.8cm of polyethylene thickness when both experimental and calculated values were normalized to 1.0 at 0.0 cm of polyethylene thickness. A sensitivity analysis of the NMTC/JAERI was performed by changing cross sections and angular distributions of hydrogen and carbon and by employing three options of the intra-nuclear cascade/evaporation calculation of the NMTC/JAERI. The disagreement of the NMTC/JAERI calculation with the experimental values was partially improved by increasing the nonelastic-scattering cross section of carbon and by broadening the elastic-scattering angular distribution of carbon.     

Construction of Neutron Guide Tubes in Upgraded JRR-3

Five neutron guide tubes have been installed in the upgraded JRR-3 (Japan Research Reactor No. 3). Two of them are for thermal neutrons and the other three are for cold ones. The characteristic wavelength of the thermal neutron guide tubes is 2 Å, and those of the cold neutron guide tubes are 4 and 6 Å. The longest guide tube is 59.9 m long and the total length of guide tubes is 232.1 m.

The beam sizes are 2 cm × 20 cm for the thermal neutron beams and 2 cm × 12 cm for the cold neutron beams. A curved part of the neutron guide is assembled by a polygonal approximation with use of 85 cm long straight units. The neutron mirrors of these units are made of natural Ni deposited borosilicate glasses. The Ni layer is about 2,000 Å in thickness.

The mean fabrication error of guide tube units is 4 μm. The mean installation errors are 8 μm for the positional abutment error and 5 × 10^?6 rad for the angular error. The neutron losses by these errors will be about 5%, and the neutron fluxes at the exits of the neutron guides are estimated to be about 2 × 10⁸ n/cm²·s.     

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.     

Inherent Protection of Plutonium by Doping Minor Actinide in Thermal Neutron Spectra

The present study focuses on the exploration of the effect of minor actinide (MA) addition into uranium oxide fuels of different enrichment (5% ²³⁵U and 20% ²³⁵U) as ways of increasing fraction of even-mass-number plutonium isotopes. Among plutonium isotopes, ²³⁸Pu, ²⁴⁰Pu and ²⁴²Pu have the characteristics of relatively high decay heat and spontaneous fission neutron rate that can improve proliferation-resistant properties of a plutonium composition. Two doping options were proposed, i.e. doping of all MA elements (Np, Am and Cm) and doping of only Np to observe their effect on plutonium proliferation-resistant properties. Pressurized water reactor geometry has been chosen for fuels irradiation environment where irradiation has been extended beyond critical to explore the subcritical system potential. Results indicate that a large amount of MA doping within subcritical operation highly improves the proliferation-resistant properties of the plutonium with high total plutonium production. Doping of 1% MA or Np into 5% ²³⁵U enriched uranium fuel appears possible for critical operation of the current commercial light water reactor with reasonable improvement in the plutonium proliferation-resistant properties.     

Neutron cross section sensitivity analysis of UO2 and MOX core physics experiments on light water reactors

The perturbation theory based on the transport calculation has been applied to study sensitivity of neutron multiplication factors (k_eff's) to neutron cross sections used for the reactivity analysis of UO₂ and MOX core physics experiments on light water reactors. The studied cross sections were neutron capture, fission and elastic scattering cross sections, and a number of fission neutrons, ν. The obtained sensitivities were multiplied to relative differences in the cross sections between JENDL-4.0 and JENDL-3.3 in order to estimate the reactivity effects. The results show that the increase in k_eff, 0.3%Δk/kk′, from JENDL-3.3 to JENDL-4.0 for the UO₂ core is mainly attributed to the decreases in the capture cross sections of ²³⁸U. On the other hand, there are various contributions from the differences in the cross sections of U, Pu, and Am isotopes for the MOX cores. The major contributions to increase in k_eff are decreases in the capture cross sections of ²³⁸U,²³⁸Pu, ²³⁹Pu, and those to decrease in k_eff are decreases in ν of ²³⁹Pu and increases in the capture cross sections of²⁴¹Am. They compensate each other, and the difference in k_eff between JENDL-3.3 and JENDL-4.0 is less than 0.1%Δk/kk′ and relatively small.