Ken Nakajima

First Injection of Spallation Neutrons Generated by High-Energy Protons into the Kyoto University Critical Assembly

At the Kyoto University Research Reactor Institute, the world’s first injection of spallation neutrons generated by high-energy proton beams into a reactor core was successfully accomplished on March 4, 2009. By combining the fixed field alternating gradient (FFAG) accelerator with the A-core (Fig. 1) of the Kyoto University Critical Assembly (KUCA), a series of accelerator-driven system (ADS) experiments were carried out by supplying spallation neutrons to a subcritical core through the injection of 100MeV protons onto a tungsten target of 80mm diameter and 10mm thickness. In these experiments, the proton beams from the FFAG accelerator were generated at 30Hz repetition rate and 10 pA current. The neutron intensity generated at the tungsten target was around 1 10 s . The objective of these experiments was to conduct a feasibility study on ADS from the viewpoint of reactor physics, in order to develop an innovative nuclear reactor for a high-performance transmutation system with a capability of power generation or for a new neutron source for scientific research. The A-core employed in the ADS experiments was essentially a thermal neutron system composed of a highly enriched uranium fuel and a polyethylene moderator/reflector. In the fuel region, a unit cell is composed of a 93% enriched uranium fuel plate 1/1600 thick and polyethylene plates 1/400 and 1/800 thick. In these ADS experiments, three types of fuel rods designated as the normal, partial, and special fuels were employed. From the reason of the safety regulation for KUCA, the tungsten target was located not at the center of the core but outside the critical assembly, and the outside location was similar to that in previous experiments using 14MeV neutrons. As in previous ADS experiments with 14MeV neutrons, the introduction of a neutron guide and a beam duct is requisite to lead the high-energy neutrons generated from the tungsten target to the center of the core as much as possible. The detailed composition of the normal, partial, and special fuel rods, the polyethylene rod, the neutron guide, and the beam duct was described in Refs. 3–5). To obtain the information on the detector position dependence of the prompt neutron decay measurement, neutron detectors were set at three positions shown in Fig. 1: near the tungsten target (position (17, D); 1=200 BF3 detector) and around the core (positions (18, M) and (17, R); 100 He detectors). The prompt and delayed neutron behaviors (Fig. 2), which were an exponential decay behavior and a slowly decreasing behavior, respectively, were experimentally confirmed by observing the time evolution of neutron density in ADS. These behaviors clearly indicated that neutron multiplication was caused by an external source: sustainable nuclear chain reactions were induced in the subcritical core by spallation neutrons through the interaction of the tungsten target and the proton beams from the FFAG accelerator. In these kinetic experiments, subcriticality was deduced from the prompt neutron decay constant by the extrapolated area ratio method. The difference between the measured results of 0.74% k=k and 0.61% k=k at the positions (17, R) and (18, M) in Fig. 1, respectively, from the experimental evaluation of 0.77% k=k, which was deduced from the combination of both the control rod worth by the rod drop method and its calibration curve by the positive period method, was within 20%. Note that the subcritical state was attained by the full insertion of C1, C2, and C3 control rods into the core. The thermal neutron flux distribution was estimated through the horizontal measurement of the In(n, )In Atomic Energy Society of Japan Corresponding author, E-mail: pyeon@rri.kyoto-u.ac.jp Present address: SR Center, Ritsumeikan University, 1-1-1 Nojihigashi, Kusatsu-shi, Shiga 527-8577, Japan Present address: Institute of Nuclear Safety System, Incorporated, 64, Sata, Mihama-cho, Mikata-gun, Fukui 919-1205, Japan Journal of NUCLEAR SCIENCE and TECHNOLOGY, Vol. 46, No. 12, p. 1091–1093 (2009)

Measurements of the Modified Conversion Ratio by Gamma-Ray Spectrometry of Fuel Rods for Water-Moderated UO2 Cores

The modified conversion ratio is defined as the ratio of 238U captures to total fission. Gamma-ray spectrometry of irradiated fuel rods has been introduced to measure this quantity in two types of ...

Validation of Pb nuclear data by Monte Carlo analyses of sample reactivity experiments at Kyoto University Critical Assembly

Sample reactivity experiments on the uncertainty analyses of Pb nuclear data are carried out by substituting Al plates for Pb ones at the Kyoto University Critical Assembly, as part of basic research on Pb–Bi for the coolant. Numerical simulations of sample reactivity experiments are performed with the Monte Carlo calculation code MCNP6.1 together with four nuclear data libraries JENDL-3.3, JENDL-4.0, ENDF/B-VII.0 and JEFF-3.1, to examine the accuracy of cross-section uncertainties of Pb isotopes by comparing measured and calculated sample reactivities. A library update from JENDL-3.3 to JENDL-4.0 is demonstrated by the fact that the difference between Pb isotopes of the two JENDL libraries is dominant in the comparative study, through the experimental analyses of sample reactivity by the MCNP approach. In addition, JENDL-4.0 reveals a slight difference from ENDF/B-VII.0 in all Pb isotopes and 27Al, and from JEFF-3.1 in 238U and 27Al. Based on these results, further experiments are needed to investigate the uncertainties of Bi isotopes with the use of the Pb–Bi and Bi plates.

Study on effective average (γ, n) Cross Section for 89Y, 90Zr, 93Nb, and 133Cs and (γ, 3n) cross section for 99Tc

A nondestructive detection technique was proposed for the easy assessment of long-lived radionuclides by the use of bremsstrahlung photons. The nuclide of 99Tc was considered for the assessment over an effective average 99Tc (γ, 3n) 96Tc cross section. For validating the experimental method on 99Tc, photonuclear (γ, n) cross sections of 89Y, 90Zr, 93Nb, and 133Cs were measured. Continuous-energy bremsstralung photons were generated from a platinum target bombarded by an electron beam of 32/36MeV from an electron linac. The photonuclear (γ, n) cross sections were previously measured by Saclay (France) and Livermore (USA) laboratories for 89Y, 90Zr, 93Nb, and 133Cs nuclides. For 89Y, 90Zr, and 133Cs, the present results were in good agreement, within 9% deviation with Saclay and an acceptable deviation of 14–36% from Livermore. In the case of 93Nb, the contribution for 93Nb (γ, n) 92Nb* in the metastable state was 55.2% of the total (γ, n) cross section of Saclay in an averaged form. The present experimental method was thus confirmed to show a good accuracy. The effective average cross section of 99Tc (γ, 3n) 96Tc was obtained as 2.30 mb in the energy range of 25.723–36 MeV.

Modified conversion ratio measurement in light water-moderated UO2 lattices

To investigate the accuracy of the neutronic calculations in various neutron spectra, the modified conversion ratios [(MCR): ratio of {sup 238}U capture rate-to-total fission rate] of four kinds of light water-moderated UO{sub 2} fuel lattices were measured. In the measurements, the relative reaction rates of {sup 238}U capture and total fission were obtained from the nondestructive gamma-ray spectrometry of {sup 239}Np and {sup 143}Ce, respectively, which accumulated in the fuel rod irradiated at the Tank-Type Critical Assembly. The moderator-to-fuel volume ratios V{sub m}/V{sub f} of the measured lattices were 1.50 (undermoderate) to 3.00 (overmoderate). The measured MCRs were 0.477 {+-} 0.014(V{sub m}/V{sub f} = 1.50), 0.434 {+-} 0.013(1.83), 0.383 {+-} 0.011(2.48), and 0.356 {+-} 0.011(3.00), respectively. The Monte Carlo calculation employing the JENDL-3 library showed good agreement with the experiments for all the cores, although they showed a tendency of overestimation for larger values of MCR, as shown in the case of UO{sub 2} tight lattices. Therefore, it was concluded that, for the cores investigated, the accuracy of the neutronic calculation method currently used was very good.

Determination of the modified conversion ratio of light-water-moderated uranium-plutonium mixed-oxide-fuel lattice

The modified conversion ratio (MCR) (the ratio of the {sup 238}U capture rate to the total fission rate) in a light-water-moderated uranium-plutonium mixed-oxide (MOX-) fuel lattice was measured for four types of lattices with different plutonium enrichment. In the current method, the relative reaction rates of {sup 238}U capture and total fission were obtained from nondestructive gamma-ray spectrometry of {sup 239}Np and fission products, respectively, which accumulated in the fuel rod irradiated at the Tank-Type Critical Assembly. The measured results of the fission rates derived from two different fission products agreed well with each other, and the measured MCRs showed good agreement with the results of the Monte Carlo calculation with the whole-core model. Therefore, the current nondestructive method is applicable to the MCR measurement of MOX fuel.

Sensitivity and Uncertainty Analyses of Lead Sample Reactivity Experiments at Kyoto University Critical Assembly

Abstract Sensitivity and uncertainty analyses of lead (Pb) isotope cross sections are conducted with the use of sample reactivity experiments at the Kyoto University Critical Assembly (KUCA). With the combined use of the SRAC2006 and MARBLE code systems, attempts are made to precisely examine the contributions of the reactions and energy regions of Pb isotope cross sections to reactivity based on the covariance data of JENDL-4.0. Moreover, the effect of decreasing uncertainty is discussed in terms of the accuracy of sample reactivity by applying the cross-section adjustment method to the uncertainty analyses. From the results of the sensitivity and uncertainty analyses, the reliability of Pb isotope cross sections, such as the Pb isotope covariance data of JENDL-4.0, is compared with the JENDL-3.3, ENDF/B-VII.0, and JEFF-3.1 libraries. Additionally, the numerical results reveal the applicability of the sensitivity and uncertainty analyses to the thermal neutron spectrum cores, such as the KUCA core, and demonstrate the improvement in the calculation results generated by the cross-section adjustment.

Measurement of the photonuclear (γ, n) reaction cross section for 129I using bremsstrahlung photons

Abstract Nuclear waste contains a significant amount of long-lived non-gamma-emitting nuclei such as 129I and 14C. A method of nondestructive detection for monitoring long-lived waste products is proposed as an application of the (γ,n) reaction. This method is useful for surveying long-lived “difficult-to-measure” nuclides, e.g., 129I. Iodine-128 produced from the reaction of 129I(γ,n)128I emits gamma rays that can easily be measured by a gamma-ray counter. We measured the inclusive photonuclear 129I(γ,n)128I reaction cross section induced by bremsstrahlung photons. The photons were produced at a Ta target bombarded by 30-MeV electrons from a linear accelerator. The intensity of the slow neutrons was considered in the reactions of 127I(n, γ)128I and 129I(n, γ)130I. The activity of 128I was measured by a high-purity germanium spectrometer. The gamma-ray flux and the neutron flux were calculated using the EGS and MCNP codes, respectively. The average activation cross section of the 129I(γ,n)128I reaction had a 12% deviation from the evaluated International Atomic Energy Agency photonuclear data.

Reaction rate analyses of accelerator-driven system experiments with 100 MeV protons at Kyoto University Critical Assembly

ABSTRACT At the Kyoto University Critical Assembly, a series of reaction rate experiments is conducted on the accelerator-driven system (ADS) with spallation neutrons generated by the combined use of 100 MeV protons and a lead–bismuth target in the subcritical state. The reaction rates are measured by the foil activation method to obtain neutron spectrum information on ADS. Numerical calculations are performed with MCNP6.1 and JENDL/HE-2007 for high-energy protons and spallation process, JENDL-4.0 for transport and JENDL/D-99 for reaction rates. That the reaction rates depend on subcriticality is revealed by the accuracy of the C/E (calculation/experiment) values. Nonetheless, the accuracy of the reaction rates at high-energy thresholds remains an important issue in the fixed-source calculations. From reaction rate analyses, the indium ratio is newly defined as another spectrum index with the combined use of 115In(n, γ)116mIn and 115In(n, n′)115mIn reaction rates, and considered useful in examining the neutron spectrum information on ADS with spallation neutrons.

Neutron capture cross section measurements of 151,153Eu using a pair of C6D6 detectors

ABSTRACT We have measured the neutron capture cross sections of 151Eu and 153Eu by the time-of-flight (TOF) method in the range from 0.005 eV to keV region using the Kyoto University Research Reactor Institute - Linear Accelerator (KURRI-LINAC). We employed a pair of C6D6 liquid scintillators for the prompt capture γ-ray measurement. The pulse-height weighting technique was employed to obtain the capture yields from the γ-ray spectra of 151,153Eu. The obtained thermal cross sections at 0.0253 eV are 9051 ± 683 b for 151Eu and 364 ± 44 b for 153Eu, respectively. The resonance integrals have been derived as 3490 ± 162 b for 151Eu and 1538 ± 106 b for 153Eu. The obtained capture cross sections were compared with the previously reported experimental data and the evaluated data. The evaluated data in JENDL-4.0 and JEFF-3.2 show good agreement with the present experiment results of 151Eu, however, the evaluated data in ENDF/B-VII.1 are larger than the present experiment results of 151Eu about 10% to 20% in the energy region from 0.03 to 0.2 eV. For the neutron capture cross sections of 153Eu, the evaluated data in ENDF/B-VII.1 and Widders data are in good agreement with the present results in the energy region below 0.35 eV.