Degradation prediction of a γ-ray radiation dosimeter using InGaP solar cells in a primary containment vessel of the Fukushima Daiichi Nuclear Power Station

ABSTRACT

Indium gallium phosphide (InGaP) solar cell with a superior high-radiation resistance is expected to be a powerful candidate for a dosimeter under a high-radiation dose rate environment. In this study, in order to predict the lifetime as the dosimeter using the InGaP solar cell, we clarify the effect of minority-carrier diffusion length (L) on a radiation-induced current as a dose signal in the InGaP solar cell by irradiation tests and empirical calculations. In the irradiation tests, the short-circuit current density (J_sc) as a function of the γ-ray dose rate is measured to estimate the L for the InGaP solar cell by irradiation tests. The operational lifetime as a detector using the InGaP solar cell under various dose rates is estimated by using the empirical calculations based on the relation between the L and absorbed dose. The results suggest that the dosimeter using InGaP solar cell is able to be used during more than 10 h in the primary containment vessel of the Fukushima Daiichi Nuclear Power Plant and it has a high potential of being a radiation-resistant dosimeter that would contribute to the decommissioning.     

Development of radiation-resistant optical fiber for application to observation and laser spectroscopy under high radiation dose

In the Fukushima Daiichi Nuclear Power Station, it is necessary to survey the locations and conditions of fuel debris inside reactor pressure vessels or primary containment vessels under water and radiation environment in preparation for removing fuel debris. An optical fiber is well known for features such as signal transmission, light weight, superior insulation performance, water resistance and electromagnetic noise resistance. These features allow the optical fiber to simplify the instrumentation systems for in-vessel inspection, as long as provide that the optical fiber can be used under high radiation dose environment. The radiation resistance of an optical fiber was improved by increasing the amount of hydroxyl up to 1000 ppm in pure silica fiber. The improved optical fibers were irradiated with γ-ray up to 1 × 10⁶ Gy using a ⁶⁰Co source. They indicated a large peak around 600 nm and a peak tail from ultraviolet region, but no large absorption in infrared region except a hydroxyl absorption peak of 945 nm. We have confirmed that the optical fiber containing 1000 ppm hydroxyl has enough radiation resistance for radiation-induced transmission losses, and the infrared imaging is effective for observation under high radiation doses.     

Development of a fiber-coupled laser-induced breakdown spectroscopy instrument for analysis of underwater debris in a nuclear reactor core

To inspect the post-accident nuclear core reactor of the TEPCO Fukushima Daiichi nuclear power plant (F1-NPP), a transportable fiber-coupled laser-induced breakdown spectroscopy (LIBS) instrument has been developed. The developed LIBS instrument was designed to analyze underwater samples in a high-radiation field by single-pulse breakdown with gas flow or double-pulse breakdown. To check the feasibility of the assembled fiber-coupled LIBS instrument for the analysis of debris material (mixture of the fuel core, fuel cladding, construction material and so on) in the F1-NPP, we investigated the influence of the radiation dose on the optical transmittance of the laser delivery fiber, compared data quality among various LIBS techniques for an underwater sample and studied the feasibility of the fiber-coupled LIBS system in an analysis of the underwater sample of the simulated debris in F1-NPP. In a feasible study conducted by using simulated debris, which was a mixture of CeO₂ (surrogate of UO₂), ZrO₂ and Fe, we selected atomic lines suitable for the analysis of materials, and prepared calibration curves for the component elements. The feasible study has guaranteed that the developed fiber-coupled LIBS system is applicable for analyzing the debris materials in the F1-NPP.     

Thermophysical and mechanical properties of CrB and FeB

ABSTRACT

Fuel debris in the Fukushima Daiichi nuclear power plant is thought to contain borides that are generated from B₄C control material. In order to successfully devise methods to effectively remove the debris material, characterization of the properties of the fuel debris is essential. This paper presents some of the thermophysical and mechanical properties of two types of borides, FeB and CrB, and compares them with those of Fe₂B and ZrB₂. These are the representative borides that might have been generated in the Fukushima Daiichi (1F) nuclear power plant accident. We observed that the thermal conductivity values of both CrB and FeB are much lower than that of ZrB₂, and are similar to that of Fe₂B. Moreover, the Vickers hardness and fracture toughness of FeB and CrB are almost identical to each other, and similar to those of ZrB₂ and Fe₂B.     

Uranium Dioxide Fuel Pin Elongation Behavior under Irradiation

The modified damage constant, KL = (1/Li² – 1/L_O ²).φ^?1, is obtained from the drop rate of the photo- current at 1.0 μm monochromatic light irradiation, where the minority carrier diffusion length in the bulk region before γ-irradiation, L₀ , is estimated by fitting the observed spectral response curve to those calculated with the diffusion length as parameter. The values Li and φ represent the minority carrier diffusion length after γ-irradiation, and the total flux of incident γ-ray photons, respectively. When N/P-type cells are contaminated with Cu, the values of KL by about one order of magnitude compared with non-doped cells, while those of the P/N-type cell contaminated with either Cu or Ni ane only slightly smaller than when not doped. The KL values of the pulled (C.Z) bulk P/N-type cell are about 1/2 those of the floating zone (F.Z) bulk cell. The curves of KL of the non-doped F-Z and C-Z bulk P/N-type cell vs. total dosage begin to decrease from a point below 10¹⁶ photons/cm² total dose. Also, the damage constant, K=(γi-¹ –TO^?1)φ^?1, of n-type floating zone (F-Z) and pulled (C-Z) bulk P/N-type cells and of p-type C-Z bulk N/P-type cells, both γ-irradiated, is seen to increase with dopant concentration in the bulk region.     

A study on the electric properties of single-junction GaAs solar cells under the combined radiation of low-energy protons and electrons

Displacement damage induced by charged particle radiation is the main cause of degradation of orbital-service solar cells, while the radiation-induced ionization shows no permanent damage effect on their electrical properties. It is reported that in single crystal silicon solar cells, low-energy electron radiation does not exert permanent degradation of their properties, but the fluence of electron radiation exerts an influence on the damage magnitude under the combined radiation of protons and electrons. The electrical properties of the single-junction GaAs/Ge solar cells were investigated after irradiation by sequential and synchronous electron and proton beams. Low-energy electron radiation showed no effects on the change of the solar cell properties during sequential or synchronous irradiation, implying ionization during particle radiation could not exert influence on the displacement damage process to the solar cells under the experimental conditions.     

辐照剂量增量对泥石流物质热释光等效剂量的影响

等效剂量的准确测定是进行地质样品热释光年龄测定的关键。采用不同辐照间隔增量测量泥石流物质标样,测得的等效剂量值是不同的,说明辐照剂量增量对泥石流物质样品等效剂量值存在影响。测试选用已知等效剂量为30 Gy的泥石流标样,按不同的辐照剂量增量进行等效剂量测定。2.5Gy/min的β源(90Sr-90Y)辐照剂量增量分别选用1、2、4、6 min时,测得的等效剂量分别为19.3、22.3、26.4、23.0 Gy;8 Gy/min的β源(90Sr-90Y)辐照剂量增量分别选用0.5、1、1.5、2 min时,测得的等效剂量分别为17.6、19.6、25.3、15.6 Gy。研究表明:对于30 Gy的泥石流标样,剂量率为2.5 Gy/s的β源(90Sr-90Y)测定泥石流物质时最适合的辐照剂量增量为4 min。剂量率为8 Gy/s的β源(90Sr-90Y)测定泥石流物质时最适合的辐照剂量增量为1.5 min。     

F-40的辐射交联及其交联度的XPS表征

本文首次用X—光电子能谱(XPS)研究了F—40的辐射交联机理以及交联度的表征。用XPS方法获得的F—40凝胶化剂量为4.1Mrad。并发现在特定的剂量范围内,F—40的交联程度最大。     

Chemical forms of uranium evaluated by thermodynamic calculation associated with distribution of core materials in the damaged reactor pressure vessel

ABSTRACT

To suggest efficient process of the fuel debris treatment after the retrieval from the Fukushima Daiichi Nuclear Power Plant (1F), thorough investigation is indispensable on potential source of U in the fuel debris. The present study aims to estimate chemical forms of U in the in-vessel fuel debris, especially in the minor phases such as metallic phases, by performing the thermodynamic calculation under various conditions considering material relocation and changing environment during the accident progression in the 1F Unit 2. Input conditions for the thermodynamic calculation such as composition, temperature, and oxygen amount were set mainly based on the transient change in the core material distribution which had been calculated with severe accident analysis codes such as MAAP and MELCOR. The result showed that chemical form of U varied depending on the local amount of Fe and O. In regions of low steel content, the U-containing metallic phase was dominated by α-(Zr,U)(O), while regions of high steel content were dominated by Fe₂(Zr,U) (Laves phase). A few percent of U transferred to the metallic phases were highly expected under reducing conditions. Therefore, those metallic phases should be one of the potential sources of U.     

Calcium Sulfate Activated by Lead and Manganese for Thermoluminescence Dosimetry

CaSO₄: Pb, Mn has been found to be free from the serious disadvantage of rapid fading possessed by CaSO₄: Mn, the most sensitive thermoluminescence phosphor available so far. A study has been made on the effect of lead and manganese content on CaSO₄, and it is concluded that lead in CaSO₄ produces new traps for radiation energy, resulting in improvement of the properties of the phosphors for dosimetry. The optimum content of such activator additives was found to be 0.2 mol/₀ of lead and 3 mol/₀ of manganese.

The improved phosphor thus obtained produces glow peaks at 160° and 190°C, and the energy yield of the thermoluminescence is about which is twice that of CaSO₄: Mn. The more significant properties of this phosphor from the viewpoint of application to radiation dosimetry include:

(1) Linear responce from 50μR to 10⁴R

(2) Minimum detectable dose of 50μR±25% by experimental reader

(3) Fading rate of 5% in a week     

Chemical Dosimeters for Mixed Radiations in a Nuclear Reactor

The dose rates due to mixed reactor radiations were measured by five gaseous chemical dosimeters – nitrous oxide (natural), ¹⁵N-enriched nitrous oxide, ethylene, ethane and carbon dioxide. The observed dose rates for these gases at the same irradiation position in a nuclear reactor were, 1.8×10⁸, 1.5×l0⁸, 1.9×10⁸, 2.5×l0⁸ and 1.0×10⁸ rad/hr, respectively. These values were compared with those calculated from the mass stopping power of the gases for secondary electrons produced by γ-rays and those from thermal and fast neutron fluxes. No contradiction was found among them. A method of analysis of the reactor radiation dose rates into γ, thermal and fast neutron components is proposed, which is based solely on chemical dosimetry.     

Criticality configuration design methodology applied to the design of fuel debris experiment in the new STACY

ABSTRACT

The new critical assembly STACY will be able to contribute to the validation of criticality calculations related to the fuel debris. The experimental core designs are in progress in the frame of JAEA/IRSN collaboration. This paper presents the method applied to optimize the design of the new STACY core to measure the criticality characteristics of pseudo fuel debris that simulated Molten Core Concrete Interaction (MCCI) of the fuel debris. To ensure that a core configuration is relevant for code validation, it is important to evaluate the reactivity worth of the main isotopes of interest and their k_eff sensitivity to their cross sections. In the case of the fuel debris described in this study, especially for the concrete composition, silicon is the nucleus with the highest k_eff sensitivity to the cross section. For this purpose, some parameters of the core configuration, were adjusted using optimization algorithm to find efficiently the optimal core configurations to obtain high sensitivity of silicon capture cross section. Based on these results, realistic series of experiments for fuel debris in the new STACY could be defined to obtain an interesting feedback for the MCCI. This methodology is useful to design other experimental conditions of the new STACY.     

Formation of radioactive cesium microparticles originating from the Fukushima Daiichi Nuclear Power Plant accident: characteristics and perspectives

ABSTRACT

Following identification of radioactive Cs microparticles (CsMPs) in aerosol samples from the Fukushima Daiichi Nuclear Power Plant (FDNPP), numerous reports on CsMPs have been published. This paper reviews recent progress in the measurement and characterization of CsMPs by advanced analytical techniques, including advanced transmission electron microscopy (TEM) and synchrotron X-ray analysis. These analyses revealed that the CsMPs contained Si, Fe, Zn, Cs, and minor quantities of U together with some fission products. Uranium in the CsMPs was identified as being in the form of uraninite and (U,Zr)O₂. Detailed advanced TEM analysis has clarified some of the processes resulting in the alteration of constituents of the nuclear fuels and containment vessel materials during this severe accident. In addition, a detailed report on the elemental compositions and structures of the fuel debris fragments collected inside and outside of the primary containment vessel²⁵ highlighted the fact that the fuel debris fragments contained nanoparticles with the U and Zr components having similar structures to that of the CsMPs. This similarity in structures has stimulated further research on the structure and elemental constituents, especially for U and Zr in the CsMPs, and has opened up new avenues for studying the chemical characteristics of the fuel debris.     

Effect of quenching on molten core-concrete interaction product

ABSTRACT

Characterization of fuel debris is required to develop fuel debris removal tools for decommissioning Fukushima Daiichi nuclear power plant (1F). Especially, knowledge about the characteristics of molten core-concrete interaction (MCCI) product is needed because of the limited information available at present. Samples from a large-scale MCCI test performed under quenching conditions, VULCANO VW-U1 were analyzed to evaluate the characteristics of the surface of MCCI product. Four samples were selected from test sections at different locations. As a result, the characteristics of the samples were found to be similar. Several corium phases, such as cubic-(U,Zr)O₂ and tetragonal ZrO₂, were detected by X-ray diffraction (XRD), but concrete-based phases, such as the crystalline SiO₂ phase, were not detected by XRD because the quantity of the SiO₂ phase was too small to be measured. The Vickers hardness of each phase in these samples was higher than that of previously analyzed samples in another VULCANO test campaign, VBS-U4. Based on a comparison between MCCI product generated under quenching condition, such as VW-U1, and gently cooled MCCI product, such as VBS-U4, the MCCI product generated under quenching condition is more homogeneous, and its hardness is higher than that of the gently cooled MCCI product.     

Gamma irradiation effects on ZnO-based scintillating glasses containing CeO2 and/or TiO2

Irradiation resistance of as-studied scintillating glass samples were tested under the ⁶⁰Co γ-ray radiation at doses between 60 and 700 Gy. Photoluminescence properties of glasses with and without cerium ions were investigated to show the effect of cerium (III) on the luminescence of ZnO. Ultraviolet and visible optical transmittance spectra were compared before and after irradiation treatment. The so-called radiation-induced absorption coefficient (RIAC) was introduced to compare more effectively the radiation damage on glass samples. The much reduced transmittance change and decreased RIAC value in UV-Vis region indicate that the density of electron centers and hole centers caused by radiation decreased, which helps to confirm that the reduction and oxidation reaction of cerium ions took place in radiation process. From RIAC curves, it can be seen that TiO₂ enhances the irradiation resistance of sample in UV region. However, high TiO₂ content has negative effect on visible transmittance of glasses after the higher dose irradiation (700 Gy).     

稳定剂对PVC辐射变色性能影响的研究

采用^60Coγ射线对聚氯乙烯（Polyvinyl chloride,PVC）／稳定剂复合体系进行辐照。研究了不同稳定剂对PVC辐射变色．陛能的影响;并采用紫外光谱仪（Ultraviolet spectrometer,UV）和傅立叶变换红外光谱仪（Fouriertransform infrared spectrometer,FT—IR）对其结构进行了表征。结果表明,硬脂酸钙（Calcium stearate,CaSt）,硬脂酸锌（Zinc stearate,ZnSt）复合稳定剂、环氧大豆油、亚磷酸酯和光稳定剂的加入可提高PVC耐辐射变色性能。亚磷酸酯的加入可提高PVC的耐辐射氧化性能。在相同吸收剂量下,添加硬脂酸钙／硬脂酸锌复合稳定剂的PVC具有优异的耐辐射变色性能。在吸收剂量为30kGy时,CaSt：Znst为4：1的PVC黄变指数变化率为8．3％。PVC经辐照后主要形成3至4个共轭双键的不饱和酮或共轭烯烃结构。     

Gamma-Ray Irradiation Effect on Loss Increase of Single Mode Optical Fibers, (II)

In order to examine the reliability of optical fiber systems in nuclear power plants, we estimated the induced losses of pure silica core and GeO₂ added core single mode optical fibers (Si-SMF and Ge-SMF) exposed over a long period to low dose rate γ-rays (less than 100 R/h, 40 yr). Also γ-ray induced losses under a high dose rate and high temperature condition such as 10⁶R/h and 200°C were estimated in view of an accident.

In both cases, the induced losses were concluded to depend mainly on total dose from the results of low dose rate irradiation experiments and a study on temperature dependences of induced losses

It was ascertained from the estimation that the Si-SMF had less radiation sensitivity than the Ge-SMF in the radiation environment of nuclear power plants The induced loss of the Si-SMF exposed for 40 yr to 1 R/h γ-rays was estimated to be about 2 dB/km and almost five times lower than that of the Ge-SMF.     

Calculation of gamma and neutron emission characteristics emitted from fuel debris of Fukushima Daiichi Nuclear Power Station

ABSTRACT

Determination of fuel debris location and distribution inside the primary containment vessel of Fukushima Daiichi Nuclear Power Station is important to decide further decommissioning step and strategy. We calculate neutron and photon spectra including the contribution of secondary particles in the primary containment vessel of Fukushima Daiichi Nuclear Power Station. The Calculated Neutron and photon spectra can be used as the base for determination suitable spectrometer system or detector for searching, localizing and treatment of fuel debris.     

柔性倒置赝型三结太阳电池高能质子辐射效应研究

为考察柔性薄膜GaInP/GaAs/InGaAs倒赝型三结（IMM3J）太阳电池的抗辐照性能,本文对其进行了1、3、5 MeV高能质子辐照。SRIM模拟结果表明,1、3、5 MeV质子辐照在IMM3J电池中造成均匀的位移损伤。光特性(LIV)结果表明,开路电压(Voc)、短路电流(Isc)和最大输出功率(Pmax)与质子注量呈对数退化规律。通过非电离能量损失（NIEL）将不同能量质子的注量转化为位移损伤剂量（DDD）,结果显示,Voc和Pmax与DDD呈对数退化规律,而Isc遵循两种不同的退化规律。光谱响应测试证明,GaInP子电池具有优异的抗辐照性能,3个子电池中InGaAs(10 eV)子电池的抗辐照性能最差。     

Optimization of the radiation shielding mass for the magnet coils of the VISTA spacecraft

An attempt has been made for the optimisation of the radiation shielding of a spacecraft design concept with inertial fusion energy propulsion for manned or heavy cargo deep space missions beyond earth orbit. Rocket propulsion is provided by fusion power deposited in the inertial confined fuel pellet debris, and with the help of a magnetic nozzle. The allowable nuclear heating in the super conducting magnet coils (up to 5 mW/cm³) is the crucial criterion for the dimensioning of the radiation shielding structure of the spacecraft. The optimized design reduced the shield mass from 600 tons to 93 and 88 tons with natural and enriched lithium, respectively. The space craft mass was 6000 tons. Total peak nuclear power density in the coils is calculated to be 5.0 mW/cm³ for a fusion power of 17,500 MW. Peak neutron heating density is 2.6 mW/cm³ and peak γ-ray heating density is 2.9 mW/cm³ (all on different points). However, volume averaged heat generation in the coils is much lower, namely 0.30, 0.73 and 1.03 mW/cm³ for neutron, γ-ray and total nuclear heating, respectively.