Mikio Uematsu

Studies on Limestone Concrete as a Low-Activation Structural Material for Nuclear Power Plants

Because of low content of Li, Co and Eu, the target nuclides of activation reaction, limestone concrete is considered to be effective in reducing the decommissioning cost of nuclear plants. Induced activity calculation and structural strength test were performed for limestone concrete and the results were compared with the data obtained for sandstone concrete, which is generally used in nuclear plants. Minor elements, which are important from the viewpoint of activation, were measured with elementary analysis for limestone samples from three different quarries in Japan. Induced activity in biological shield walls (BSW) of Boiling Water Reactor (BWR) plants was calculated with the isotope generation code ORIGEN-79 using neutron flux data obtained with the one-dimensional Sn transport code ANISN and MGCL137-group activation cross section library based on JENDL-3. Estimated total radioactivity accumulated in limestone concrete BSW was 5 times lower than that in the sandstone concrete BSW. Structural strength were compared between limestone concrete and sandstone concrete, and limestone concrete was found to have enough compressive strength and tensile strength.

Development of Low-Activation Design Method for Reduction of Radioactive Waste Below Clearance Level

Design methodology for reinforced concrete of nuclear power plants to reduce radioactive wastes in decommission phase has been developed. To realize this purpose, (1) development of raw materials database of cements, aggregates and steel bars on concentration of radioactive target elements, (2) trial production of low activation cements and steel bars based on the material database developed in (1), and (3) development of tools for estimation and prediction of the amount of radioactive elements in reactor shielding walls have been carried out. Radioactive analysis showed that Co and Eu were the major target elements which decide the radioactivity level of reinforced concrete from wide survey of raw materials for concrete (typically aggregates and cements). Material database for the contents of Co and Eu was developed based on the chemical analysis and radioactivation analysis. Upon the above survey and execution expreiment of concrete, six types of low-activation concrete are proposed for various radioactive portion in the plant. These concrete have a 1/10 – 1/300 rasioactivity compare to the ordinary concrete, which are assumed the concrete with Andesite aggregate and ordinary Portland cement. Baed on the above data base, it was clarified that the low activation cement would be successfully manufactured by adequate selection of raw materials. The prospect to produce the low-heat portland cement which would have a 1/3 radioactivity in comparison with conventioanl cements obtained by means of selection of limestone and natural gypsum. An attempte was carried out to produece low activation heavy-mortar which would have radioactivity below the clearance level when using at the radiation shielding wall of BWR. Characterization and optimization of consturction conditions with new additives have also been carried out. These two new raw materials for low-activation concrete are conducted in pre-manufacture size, and over the laboratry level. Boron added low-activation concrete are also carried out as extreamly high performance low-activation concrete. It was claryfied that the accurcy of calculation results of the radioactivity evaluation was very high compared to available benchmark calculation for the JPDR and commercial light water reactor. The specification of the mapping system for judging the activation classification was also developed by using the general-purpose radio activation calculation tool. This work is supported by a grant-in-aid of Innovative and Viable Nuclear Technology (IVNET) development project of Ministry of Economy, Trade and Industry, Japan.Copyright

DEVELOPMENT OF LOW-ACTIVATION REINFORCED CONCRETE DESIGN METHODOLOGY—I: MANUFACTURE OF LOW-ACTIVATION CONCRETE

Abstract Screening tests using several reactors were performed to select low-activation raw materials. The number of samples was about 1500. Detailed data were obtained on the concentrations of Co and Eu in low-activation aggregates, low-activation cements, low-activation additives, and low-activation B4C sands. After that, we manufactured various types (1/10, 1/20, 1/30, 1/50, 1/100, 1/300) of low-activation concrete. The term “1/10 low-activation” concrete denotes that the activity reduction rate to ordinary concrete is designed to be 1/10. By admixing with a boron content of ~1 × 1021/cm3, the total residual radioactivity reduction rates of low-activation concrete to ordinary concrete, in units of ΣDi/Ci (Di: concentration of radionuclide i, Ci: clearance level of radionuclide i cited from IAEA-RS-G-1.7), are estimated to range from ~1/300 to 1/10 000. It was concluded that most of the shielding concrete around the advanced boiling water reactor (ABWR) or the advanced pressurized water reactor (APWR) are classified below the clearance level of decommissioning by adopting some suitable types of low-activation concrete.

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.

Development of “SKYSHINE-CG” Code : A Line-Beam Method Code Equipped with Combinatorial Geometry Routine

A boiling water reactor (BWR) plant has a single loop coolant system, in which main steam generated in the reactor core proceeds directly into turbines. Consequently, radioactive 16N (6.2MeV photon emitter) contained in the steam contributes to gamma-ray skyshine dose in the vicinity of the BWR plant. The skyshine dose analysis is generally performed with the line-beam method code SKYSHINE, in which calculational geometry consists of a rectangular turbine building and a set of isotropic point sources corresponding to an actual distribution of 16N sources. For the purpose of upgrading calculational accuracy, the SKYSHINE-CG code has been developed by incorporating the combinatorial geometry (CG) routine into the SKYSHINE code, so that shielding effect of in-building equipment can be properly considered using a three-dimensional model composed of boxes, cylinders, spheres, etc. Skyshine dose rate around a 500 MWe BWR plant was calculated with both SKYSHINE and SKYSHINE-CG codes, and the calculated results were compared with measured data obtained with a Nal(Tl) scintillation detector. The C/E values for SKYSHINE-CG calculation were scattered around 4.0, whereas the ones for SKYSHINE calculation were as large as 6.0. Calculational error was found to be reduced by adopting three-dimensional model based on the combinatorial geometry method.

Progress and prospects of calculation methods for radiation shielding

Progress in calculation methods for radiation shielding are reviewed based on the activities of research committees related to radiation shielding fields established in the Atomic Energy Society of Japan. A technological roadmap for the field of radiation shielding; progress and prospects for specific shielding calculation methods such as the Monte Carlo, discrete ordinate Sn transport, and simplified methods; and shielding experiments used to validate calculation methods are presented in this paper.

Development of Low-Activation Reinforced Concrete Design Methodology - II: Concrete Activation Analyses of BWR/PWR

Abstract A precise method for estimating residual radioactivity and decommissioning cost is indispensable when deciding whether to adopt low-activation material. For this precise estimation, accurate estimation of both the thermal neutron flux and the activation cross section of the structural material is necessary. We developed a new groupwise cross-section library that has ten thermal groups for SN transport calculation and activation calculation. These libraries are tested and used for advanced boiling water reactor (ABWR) and advanced pressurized water reactor (APWR) activation analyses.

Measurement and Analysis of Structural Activation in a Boiling Water Reactor

Induced radioactivity of structural materials of a nuclear power plant introduces the possibility of exposure of workers. In order to assess evaluation accuracy of the induced radioactivity, measurements and calculations were performed for gamma-ray dose inside an irradiated reactor pressure vessel of a boiling water reactor. Neutron flux was calculated with two-dimensional Sn transport code DOT3.5 with RZ and RΘ models. Induced radioactivity was calculated with the ORIGEN-79 code, in which three-group activation cross section was produced considering neutron spectrum instead of the original ORIGEN-79 three-group constants. Calculated dose rate by DOT3.5 agreed well with the measured value, and calculational accuracy was improved by taking account of Θ dependence of neutron flux distribution and precise neutron spectrum in activation calculation compared to the calculation with a simplified method such as a single RZ model calculation of neutron flux and activity calculation with the three-group constants built-in the ORIGEN-79 code.

Measurement of (γ, n) reaction cross section for long-lived β-emitting radionuclide129I by using bremsstrahlung photons

Many long-lived non-gamma emitting radioactive nuclei such as 129I, 14C and 93Zr are produced as wastes from nuclear fuel cycle facilities. They are called “difficult-to-measure” radio nuclei. Among them 129I is a long-lived β-emitting isotope with a half-life of 1.57 × 107 years. Iodine compounds are mobile in the vadose zone and groundwater and increase a significant long-term risk. Transmutation of 129I is a challenging issue in nuclear waste management and disposal. If 129I is transmuted into 128I (half life; 25 minutes), it can easily be measured by a Ge detector. However, the 129I(γ, n) reaction cross section has not been measured so far. In this study we have measured the inclusive 129I(y, n)128I cross section by using bremsstrahlung photons. The bremsstrahlung photons were produced from a 30-MeV electron linac. Measured average activation cross section agrees with 12% deviation from the evaluated one in the IAEA photonuclear data library. Gamma and neutron fluxes for the (γ, n) and the (n, γ) reaction were also calculated by the EGS and the MCNP codes.