Transmutation of Cesium-137 Using Proton Accelerator

Abstract

Two transmutation methods of ¹³⁷Cs using a proton accelerator were evaluated in terms of the effective half life and the transmutation energy. One was the proton method which mainly used high energy proton spallation reaction for transmutation, and the other was the spallation neutron method which mainly used thermal neutron capture reaction. The transmutation energies and the effective half lives for the two transmutation methods were calculated by Monte Carlo simulation codes for particle transport, the NMTC/JAERI code and the MCNP code. The calculated transmutation energies were 510 MeV and 570 MeV for the spallation neutron method and the proton method, respectively, for an effective half life of 2 yr for ¹³⁷Cs.     

External Radiation Conversion Coefficients using Radiation Weighting Factor and Quality Factor for Neutron and Proton from 20 MeV to 10 GeV

Conversion coefficients from neutron and proton fluences to effective dose are calculated in the range of incident neutron energy from 20 MeV to 10 GeV. Two different versions of effective dose are treated, respectively calculated using: (a) the radiation weighting factor wR, and (b) the Q-L relationship given in ICRP 60. Monte Carlo calculations are performed applying the HETC-3STEP and the MORSE-CG/KFA in the HERMES code system. The calculations are based on a modified MIRDS anthropomorphic phantom, which is irradiated in anterior-posterior and posterior-anterior directions of beam. For effective dose calculation using the Q-L relationship, a database was compiled and added to the HETC-3STEP, to derive the average quality factor. The effective dose derived using wR proved to overestimate that obtained with the Q-L relationship by about 80% at 10 GeV incident neutron energy in the case of conversion from neutron fluence. For proton fluence, the corresponding overestimation reaches a maximum factor of 4.     

Evaluation of Computational Models for Fission and Spallation Reactions Used in Accelerator Breeding and Transmutation Analysis Code

Computational models for spallation and fission reactions used in an accelerator breeding and transmutation code have been evaluated by performing calculations for thin targets of Bi, Pb, Th and U in the energy range of 50–1,000 MeV. Proton and neutron non-elastic and fission cross sections have been derived from the counts of real collisions and fission events in the targets.

Several fission models in combination with a spallation model are compared with the experimental data due to Steiner et al. and Schimmerling et al. A good agreement has been obtained for a model with the level density parameters a_n = A/10 and a_f/a_n fitted to the data due to Il'inov et al. The mass dependence of the non-elastic cross sections has been calculated also with the use of the best fit model. A good agreement with the experimental data was obtained over a wide range of nuclear masses.     

Evaluation of Neutron- and Proton-induced Nuclear Data on 27Al up to 2 GeV

We have evaluated neutron and proton nuclear data of ²⁷ Al for energies up to 2 GeV. The best set of optical model parameters were obtained from 20MeV for neutrons and from reaction threshold for protons up to 250 MeV with the phenomenological non-relativistic potential forms incorporating effects of the dispersion relationship and results of the Dirac phenomenology. The transmission coefficients for neutrons and protons derived from the optical models were fed into the GNASH code system to calculate angle-energy correlated emission spectra for light ejectiles and gamma rays. For energies between 250 MeV and 2 GeV, the total, reaction and elastic scattering cross sections were evaluated by an empirical fit and recent systematics. Emitted nucleon and pion spectra were estimated by use of QMD+SDM (Quantum Molecular Dynamics+Statistical Decay Model).     

Measurement and Analysis of Neutron Spectrum in Spherical Pile of Thoria

The energy spectrum of neutrons in a pile of thoria was measured from several keV to several MeV by the time-of-flight method with an electron linear accelerator and the results were compared with the theoretical ones predicted with three group constants for thorium, (1) constants of Abagyan et al.'s (ABBN), (2) newly produced constants of J AERI-FAST type (NNM) and (3) constants produced from ENDF/B-H in ORNL (DLC-2D).

General agreement between measurements and calculations was seen for all cases. However, (1) the predicted spectra with ABBN and DLC-2D gave less flux by about 30% than the measured in the energy below about 100 keV, (2) similar disagreement was observed only above 10 keV for the spectrum with NNM and (3) the measured spectrum became about 20% lower than all of the predicted above 1 MeV.

In parallel with the spectrum measurement, neutron spatial distribution was measured by the activation method and the results showed spherical symmetry around a photoneutron target at the center of the pile. From this fact adequancy of using one-dimensional transport code (ANISN or DTF-IV) with spherical symmetry for analysis could be verified.     

Comparison of Gamma-Ray Point Isotropic Buildup Factors Including Fluorescence and Bremsstrahlung in Lead Using Discrete Ordinates and Point Monte Carlo Methods

Exposure buildup factors and energy spectra of γ-rays, including fluorescence or bremsstrahlung radiations, in Pb for a point isotropic source have been calculated in the vicinity of the K edge and at 10 MeV using a discrete ordinate code, PALLAS. Comparisons of PALLAS results with those by the point Monte Carlo code, EGS4 showed good agreement for exposure buildup factors and energy spectra in two energy regions. Exposure buildup factors calculated with another discrete ordinate code, ASFIT, are also compared with those calculated by the PALLAS and EGS4 codes. The values obtained by all three codes showed good agreement. The effects of the cross section data on the buildup factors were investigated by using the PALLAS code.     

Measurement of Fast Neutron Induced Fission Cross Section Ratios of Pu-240 and Pu-242 Relative to U-235

Fission cross section ratios of ²⁴⁰Pu and ²⁴²Pu relative to ²³⁵U were measured by using the 4.5 MV Dynamitron accelerator of Tohoku University. The measurement using mono-energetic neutrons was performed in the neutron energy range of 0.6–7 MeV with the time-of-flight method. Prior to the measurement, a fast timing back-to-back fission chamber was developed with good time resolution to reduce the backgrounds due to α-particles and spontaneous fissions. Furthermore, we took account of the effect of the nonuniformity of fission sample thickness for accurate determination of fission cross section ratio. The uncertainty was estimated by analyzing the correlation between the error sources. The correlation matrix between the measured data was given. The overall uncertainty of the present results is about 2%. For both nuclides, the present results agree well with those by Meadows and by Kuprijanov et al. The JENDL-3 evaluation generally has good agreement with the present results. However, the evaluated data are slightly higher around 1 MeV and lower above 6 MeV than the present results.     

C 1s Binding-Energy Shift of Pyrolytic Graphite Bombarded by keV-Deuterium Ions

As a basic study of plasma-wall interactions, chemical state of graphite basal face after deuterium bombardment are studied by means of X-ray photoelectron spectroscopy. As the fluence of deuterium ions with energy of 1 or 4keV is increased from 10¹⁴ to 10¹⁸ ions/cm², the C Is line is observed to shift towards the lower binding energy at first, down to the minimum by about 0.2eV and then towards the higher energy up to an asymptotic value of about 0.2eV higher than the initial position. The first negative shift is due to lattice displacement at the target surface, and the subsequent positive shift is due to the deuterium trapping at the graphite surface. The fluence dependence of the observed C 1s shift is explained by a simple saturation model.     

Calculation of Effective Doses for External Neutrons

Monte Carlo calculations of the effective dose, on the basis of 1CRP Publication 60, were performed for external neutrons from thermal energy to 18.3 MeV for five irradiation geometries: AP, PA, RLAT, ROT and ISO. A unisex anthropomorphic phantom and the MORSE-CG code were used in conjunction with a nuclear data set based on the JENDL-3 library. The effective dose was found to be superior to the effective dose equivalent, the former quantity, for neutrons below about 1 MeV and inferior above this energy for all the geometries. The ambient dose equivalent based on the new Q-L relationship proposed in the Publication was found not necessarily to give a conservative estimate of the effective dose for the AP and PA geometries. The results obtained here were in good agreement with those calculated with a different computer code and a different nuclear data set.     

Measurements and Calculations of Neutron Energy Spectra Behind Polyethylene Shields Bombarded by 40- and 65-MeV Quasi-Monoenergetic Neutron Sources

Measurements of neutron energy spectra behind 30.5-, 61.0-, 122.0-, 183.0-cm-thick polyethylene shields bombarded by 40- and 65-MeV quasi-monoenergetic neutrons are performed at the 90-MeV AVF cyclotron of the TIARA (Takasaki Ion Accelerator for Advanced Radiation Application) at JAERI (Japan Atomic Energy Research Institute). Source neutrons are produced at 3.6- and 5.2-mm-thick⁷ Li targets bombarded by 43- and 68-MeV protons, respectively. A BC501A organic liquid scintillator and multi-moderator spectrometer with a ³He counter (Bonner ball) are used for spectrometry of transmitted neutrons and their energy spectra are obtained with the unfolding technique. The energy spectra from a few MeV up to a peak energy are obtained by the BC501A scintillator measurement and those below a few MeV down to thermal energy are obtained by the Bonner ball measurement. The measurements are performed on the neutron beam axis and at off-center positions, and attenuation profiles of neutron fluxes along the beam axis are obtained. The MORSE Monte Carlo calculations are performed with the DLC119/HIL086 multi-group cross section library for comparison with the measured data. The calculation generally gives a little overestimated fluxes, and a few % longer attenuation lengths of peak flux and dose equivalent.