Shigetaka Maeda

Numerical analysis of irradiated Am samples in experimental fast reactor Joyo

Americium is a key element to design the FBR based nuclear fuel cycle, because of its long-term high radiological toxicity as well as a resource of even-mass-number plutonium by its transmutation in reactors, which contributes the enhancement of proliferation resistance. The present paper deals with the numerical analysis of the Am sample irradiation in Joyo to examine the transmutation performance of pure isotope in fast neutron environment during the irradiation, and deals with the comparison with the experimental result to evaluate the accuracy of current available numerical tool. In 241Am pure isotope sample, the burn-up calculation of Am transmutation ratio and principal nuclides accumulation are agreed with the measured data within 1-σ uncertainty caused of cross-section covariance. Isomeric ratio of 242Am in total 241Am capture reaction were calculated as 0.852±0.016 in the core and 0.85±0.025 in the axial and radial reactors. The current data and recently reported data by Koyama et. al 2008 support the latest version of nuclear data sets in ENDFB-VII and JENDL/AC-2008. From the view point of proliferation resistance, it was confirmed 241Amp reduces un-attractive Pu to abuse from the beginning to the end of irradiation, and it would have important role to denature Pu in future FBR based nuclear fuel cycle.

CHARACTERIZATION OF NEUTRON FIELDS IN THE EXPERIMENTAL FAST REACTOR JOYO MK-III CORE

In 2003, Joyo MK-III core was upgraded to increase the irradiation testing capability. This paper describes the details of distributions of neutron flux and reaction rate in the MK-III core that was measured by characterization tests during the first two operating cycles. The calculation accuracy of the core management codes HESTIA, TORT and MCNP, was also evaluated by the measured data. The calculated fission rates of U by HESTIA agreed well with the measured one within approximately 4% in the fuel region. MCNP could simulate within 6% in the central non-fuel irradiation test subassembly and the radial reflector region, while large discrepancies were obtained in TORT results. Hence, the precise geometry model was effective in evaluating the neutron spectrum and the flux at such locations.

Lumped Group Constants of FP Nuclides for Fast Reactor Shielding Calculation Based on JENDL-3.2

Fission Products were not considered in conventional fast shielding analyses that were predominantly developed in clean core experiments. However, in power reactors with high burn-up, the accumulation of FP nuclides affects the neutron balance mainly due to the absorption reaction and it cannot be neglected when calculating the neutron flux of a high burn-up reactor. In this study, the lumped group constants of FP nuclides used for a fast reactor shielding calculation were computed with the JENDL-3.2 library and compiled to the JSDJ2/JFTJ2 set. The effect of considering FP nuclides on the neutron flux calculations was evaluated in the JOYO experimental fast reactor. These tests showed conventional calculations that ignored FP nuclides overestimated neutron flux by about 2%. The effect on reaction rate calculation of the spent fuel in IVS (in-vessel storage rack) is a maximum of 10%.

Development of Neutron Measurement in Intense Gamma Field Using New Type of Nuclear Emulsion

To precisely measure the neutron emissions from a spent fuel assembly of a fast breeder reactor, we formed nuclear emulsions based on a non-sensitized Oscillation Project with Emulsion tRacking Apparatus (OPERA) film with AgBr grain sizes of 60, 90, and 160 nm. The efficiency for 252Cf neutron detection of the new emulsion was calculated to be 0.7 × 10−4, which corresponded to an energy range from 0.3 to 2 MeV and was consistent with a preliminary estimate based on experimental results. The sensitivity of the new emulsion was also experimentally estimated by irradiating with 565 keV and 14 MeV neutrons. The emulsion with an AgBr grain size of 60 nm had the lowest sensitivity among the above three emulsions but was still sensitive enough to detect protons. Furthermore, the experimental data suggested that there was a threshold linear energy transfer of 15 keV/μm for the new emulsion, below which no silver clusters developed. Further development of nuclear emulsion with an AgBr grain size of a few tens of nanometers will be the next stage of the present study.

Decay Heat of Fast Reactor Spent Fuel

Decay heat from JOYO Mk-II spent fuel subassemblies was measured to obtain experimental data and to improve the accuracy of related calculations. The measurement was taken in the JOYO spent fuel storage pond. The fuel burn-up was approximately 66 GWd/t and the cooling time was between 40 and 385 days. The decay heat was calculated with the ORIGEN2 code using the JENDL-3.2 cross section library and the JNDC-V2 decay data and fission yield data library. The fuel power used as an input to ORIGEN2 was calculated by the MAGI core management code system. The ratios between calculated and experimental values were between 0.94 and 0.89 and decreased with a longer cooling time. This systematic discrepancy is not fully understood, but the change with cooling time appears to be due to the actinide decay heat uncertainty. This indicated that cross sections of actinides are important to evaluate decay heat accurately.

Core Modification to Improve Irradiation Efficiency of the Experimental Fast Reactor Joyo

Core modification has been investigated to further increase the core burnup and to improve the irradiation efficiency of the experimental fast reactor Joyo. This modification enables the core to accommodate more irradiation test subassemblies that have lower fissile material contents compared with the driver fuel. The design calculations showed that the replacement of the radial reflector elements made of stainless steel with those made of zirconium alloy or nickel-based alloy is effective in improving neutron efficiency. The irradiation test capacity can be increased by reducing the number of control rods based on a reevaluation of the design margin in the control rod worth calculation. The design calculation results show that these modifications, without any change in fuel specification, will be useful for conserving driver fuels and enhancing the irradiation capability of Joyo.

Characterization of Neutron Fields Using MCNP in the Experimental Fast Reactor JOYO

An extensive set of neutron dosimeters ranging from the core center to beyond the reactor vessel were irradiated during the first two operating cycles of the MK-III core to allow a detailed characterization of the neutron spectra and flux distributions for this new core configuration. New analysis methods for predicting the reaction rates for comparison with the dosimetry measurements based on the MCNP code were developed. Analysis of previous MK-II cycle 34-35 dosimetry tests was used to verify the methods. Core models were developed for the different types and locations of dosimetry test subassemblies and biasing schemes were developed. MCNP optimization techniques and the C/E differences were explored. Most of the important parameters that affect the reaction rate calculations and measurements were investigated. In most cases, MCNP provides more accurate estimations of reaction rates and neutron flux than DORT. This indicates that treatment of heterogeneous structure is needed in the predictions for the irradiation test subassembly with no fuel and the irradiation field with large flux gradients.

A new unfolding code combining maximum entropy and maximum likelihood for neutron spectrum measurement

We present a new spectrum unfolding code, the Maximum Entropy and Maximum Likelihood Unfolding Code (MEALU), based on the maximum likelihood method combined with the maximum entropy method, which can determine a neutron spectrum without requiring an initial guess spectrum. The Normal or Poisson distributions can be used for the statistical distribution. MEALU can treat full covariance data for a measured detector response and response function. The algorithm was verified through an analysis of mock-up data and its performance was checked by applying it to measured data. The results for measured data from the Joyo experimental fast reactor were also compared with those obtained by the conventional J-log method for neutron spectrum adjustment. It was found that MEALU has potential advantages over conventional methods with regard to preparation of a priori information and uncertainty estimation.

Student Training Course Using the Experimental Fast Reactor Joyo and Related Facilities

The student training courses using the experimental fast reactor Joyo and related facilities of the Japan Atomic Energy Agency (JAEA) have been initiated to utilize the nuclear facilities and their engineering staffs for the education purpose. The development of the student training course was also strongly supported by the faculty of nuclear engineering of domestic universities whose curriculum has recently been reduced. The program covers the reactor physics test analysis of Joyo core or experiments using the Joyo full-scope training simulator, neutron dosimetry, trace amount of noble gas measurement and chemical analysis of sodium, and the program has started after check and review by the specialists in university education. It is expected to promote the human resource development for the younger generation in nuclear industry, and to strengthen the relation between JAEA and universities in research area.Copyright

ENHANCEMENT OF IRRADIATION CAPABILITY OF THE EXPERIMENTAL FAST REACTOR JOYO

The experimental fast reactor Joyo is the first sodium-cooled fast reactor in Japan. One of its primary missions is to perform irradiation tests of fuel and structural materials to support the development of fast reactors. The MK-III high performance core upgrade to enhance the irradiation testing capabilities was completed in 2003. In order to expand Joyos capabilities for innovative irradiation testing applications, neutron spectrum tailoring, lower irradiation temperature, movable sample devices and fast neutron beam holes are being considered. This program responds to existing irradiation needs and aims to further expand capabilities for a variety of irradiation tests.