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Hirouchi, Jun; Tokashiki, Yuji*; Takahara, Shogo; Manabe, Kentaro
JAEA-Research 2021-001, 284 Pages, 2021/03
JAEA-Research-2021-001.pdf:4.23MB
Doses to the public are calculated with internal dose coefficients based on the publications of the International Commission on Radiological Protection (ICRP) in OSCAAR, which is a level 3 Probabilistic Risk Assessment code developed by Japan Atomic Energy Agency (JAEA). The gastrointestinal absorption fraction, 
, which is one of parameters of internal dose coefficient, is given the recommended value. However, although it has been reported that has uncertainty, the uncertainty analysis of has been performed on few radionuclides. In this report, to evaluate the influence of uncertainty of on the internal dose, we calculated the internal dose coefficient with various , and derive the relationship between the coefficient and . As a result, we indicate that the relationships are expressed by a linear function for radionuclides with a half-life of more than 0.5 days and are expressed by a cubic function for radionuclides with a half-life of less than 0.5 days.
Aoki, Jo; Matsueda, Makoto; Koarai, Kazuma; Terashima, Motoki; Fujiwara, Kenso; Abe, Hironobu
JAEA-Research 2021-002, 20 Pages, 2021/05
JAEA-Research-2021-002.pdf:2.75MB
In order to analyze extremely low concentrations of 
I in environmental samples by ICP-MS with high sensitivity and rapidity, it is necessary to remove interfering elements (Na, Mg, K, Ca, Mo, Cd and In) using a pretreatment method with Solid-phase Extraction Resin. Anion Exchange Resins with amino groups have been widely used as Solid-phase Extraction Resins, while Ag+ Supported Resins have also been widely used in recent years. It is necessary to optimize the pretreatment technique based on characteristics of the resins. In this study, we compared in terms of separation of I from matrix elements (Na, Mg, K, Ca, Mo, Cd and In) for DOWEX1-X8, AG 1-X8 and CL Resin, and investigated their suitability for ICP-MS analysis of I in environmental samples. The results of adsorption and elution experiments showed that all resins examined uptake and elute I quantitatively. CL Resin showed the highest removal performance of interfering elements, with 3.1% of Mo remaining, but other interfering elements were removed below the detection limit of ICP-MS. However, the Mo remained after the CL Resin treatment could interfere the ICP-MS measurement of I, based on the consideration of ratio of I and Mo. The eluate from CL Resin was treated with a Cation exchange resin (DOWEX 50WX8). As a result, Mo in the eluate was removed by up to 98% and the interference from Mo was reduced to measurable level. Therefore, the pretreatment method using CL Resin in combination with DOWEX 50WX8 is effective for ICP-MS analysis of I at extremely low concentrations (background level).
Sakamoto, Atsushi; Kibe, Satoshi*; Kawanobe, Kazunori*; Fujisaku, Kazuhiko*; Sano, Yuichi; Takeuchi, Masayuki; Suzuki, Hideya*; Tsubata, Yasuhiro; Ban, Yasutoshi; Matsumura, Tatsuro
JAEA-Research 2021-003, 30 Pages, 2021/06
JAEA-Research-2021-003.pdf:1.81MB
Japan Atomic Energy Agency has been developing a solvent extraction process called SELECT to recover minor actinides (MA) from spent nuclear fuel. In the SELECT process, TDdDGA, HONTA, and ADAAM are used as the extractants for MA + Ln corecovery, MA/Ln separation and Am/Cm separation, respectively. These extractants do not contain phosphorus (P), and consist of carbon (C), hydrogen (H), oxygen (O), and nitrogen (N). In this study, in order to give beneficial information for designing flowsheet, the mass transfer coefficients of Ln between HNO
solution and TDdDGA or HONTA / n-dodecane solvent were evaluated by the single drop technique. Prior to the evaluation of mass transfer coefficient, we had optimized the structure of the single drop apparatus to improve accuracy of the measurement. Based on the mass transfer coefficients obtained in HNO / TDdDGA-n-dodecane system, Ln behaviors in the counter-current extraction and back-extraction using mixer-settlers and centrifugal contactors were estimated by simple calculation, and they had a good agreement with our previous experimental results. We also confirmed the mass transfer coefficients of Ln in HNO / HONTA - n-dodecane system are under 10
m/s.
Misono, Toshiharu; Nakanishi, Takahiro; Sanada, Yukihisa; Shiribiki, Takehiko; Urabe, Yoshimi*; Tsuruta, Tadahiko
JAEA-Research 2021-004, 214 Pages, 2021/11
JAEA-Research-2021-004.pdf:12.76MB
After the accident of TEPCO's Fukushima Daiichi Nuclear Power Station (1F), the project of marine monitoring survey on radioactive substances have been conducted by the contract research from the Nuclear Regulatory Agency in FY2020. Results obtain in the project are presented in this report. Based on the monitoring results of radioactive substances of seawater and sediments, we suggested the evaluation method for optimizing the survey points and frequency, and examined the proper monitoring method on marine monitoring. In addition, core samples were collected at 70 points at the coast of Fukushima Prefecture to reveal the accumulation of radioactive Cs in the sediments. Furthermore, in order to evaluate the inflow of radioactive Cs from the river, sediment traps were installed at the coastal area to collect sinking sediment, and the horizontal distribution of the radioactive Cs concentration on the surface sediment in front of the rivers was measured. We carried out the revaluation of the towed radiation monitoring data conducted from 2013 to 2018, taking into account the natural radionuclides, and improved the radioactive Cs distribution map in the coastal sediments. In addition, a seabed topography and sediments distribution survey was conducted in the silt band area off the 1F to understand the topography and sediment distribution. From these results, we estimated the distribution and its dynamics of radioactive Cs in the sediments in the front area on the 1F.
Yoshida, Kazuo; Tamaki, Hitoshi; Hiyama, Mina*
JAEA-Research 2021-005, 25 Pages, 2021/08
JAEA-Research-2021-005.pdf:2.91MB
An accident of evaporation to dryness by boiling of high level liquid waste (HLLW) is postulated as one of the severe accidents caused by the loss of cooling function at a fuel reprocessing plant. In this case, volatile radioactive materials, such as ruthenium (Ru) are released from the tanks with water and nitric-acid mixed vapor into atmosphere. Accurate quantitative estimation of released Ru is one of the important issues for risk assessment of those facilities. To resolve this issue, an empirical correlation equation of Ru mass transfer coefficient across the vapor-liquid surface, which can be useful for quantitative simulation of Ru mitigating behavior, has been obtained from data analyses of small-scale experiments conducted to clarify gaseous Ru migrating behavior under steam-condensing condition. A simulation study has been also carried out with a hypothetical typical facility building successfully to demonstrate the feasibility of quantitative estimation of amount of Ru migrating in the facility using the obtained correlation equation implemented in SCHERN computer code which simulates chemical behaviors of nitrogen oxide based on the condition also simulated thermal-hydraulic computer code.
Yoshikawa, Shinji; Yamaji, Akifumi*
JAEA-Research 2021-006, 52 Pages, 2021/09
JAEA-Research-2021-006.pdf:3.89MB
In Fukushima Daiichi Nuclear Power Station (referred to as "FDNPS" hereafter) unit2 and unit3, failure of the reactor pressure vessel (RPV) and relocation of some core materials (CRD piping elements and upper tie plate, etc.) to the pedestal region have been confirmed. In boiling water reactors (BWRs), complicated core support structures and control rod drive mechanisms are installed in the RPV lower head and its upper and lower regions, so that the relocation behavior of core materials to pedestal region is expected to be also complicated. The Moving Particle Semi-implicit (MPS) method is expected to be effective in overviewing the relocation behavior of core materials in complicated RPV lower structure of BWRs, because of its Lagrangian nature in tracking complex interfaces. In this study, for the purpose of RPV ablation analysis of FDNPS unit2 and unit3, rigid body model, parallelization method and improved calculation time step control method were developed in FY 2019 and improvement of pressure boundary condition treatment, stabilization of rigid body model, and calculation cost reduction of debris bed melting simulation were achieved in FY2020. These improvements enabled sensitivity analyses of melting, relocation and re-distribution behavior of deposited solid debris in RPV lower head on various cases, within practical calculation cost. As a result of the analyses of FDNPS unit2 and unit3, it was revealed that aspect (particles/ingots) and distribution (degree of stratification) of solidified debris in lower plenum have a great impact on the elapsed time of the following debris reheat and partial melting and on molten pool formation process, further influencing RPV lower head failure behavior and fuel debris discharging behavior.
Ishimaru, Tsuneari; Ogata, Nobuhisa; Kokubu, Yoko; Shimada, Koji; Niwa, Masakazu; Shimada, Akiomi; Watanabe, Takahiro; Sueoka, Shigeru; Yokoyama, Tatsunori; Fujita, Natsuko; et al.
JAEA-Research 2021-007, 65 Pages, 2021/10
JAEA-Research-2021-007.pdf:4.21MB
This annual report documents the progress of research and development (R&D) in the 6th fiscal year during the JAEA 3rd Mid- and Long-term Plan (fiscal years 2015-2021) to provide the scientific base for assessing geosphere stability for long-term isolation of the high-level radioactive waste. The planned framework is structured into the following categories: (1) Development and systematization of investigation techniques, (2) Development of models for long-term estimation and effective assessment, (3) Development of dating techniques. The current status of R&D activities with previous scientific and technological progress is summarized.
Sugawara, Takanori; Moriguchi, Daisuke*; Ban, Yasutoshi; Tsubata, Yasuhiro; Takano, Masahide; Nishihara, Kenji
JAEA-Research 2021-008, 63 Pages, 2021/10
JAEA-Research-2021-008.pdf:4.43MB
This study aims to perform the neutronics calculations for accelerator-driven system (ADS) with a new fuel composition based on the SELECT process developed by Japan Atomic Energy Agency because the previous studies had used the ideal MA (minor actinide) fuel composition without uranium and rare earth elements. Through the neutronics calculations, it is shown that two calculation cases, with/without neptunium, satisfy the design criteria. Although the new fuel composition includes uranium and rare earth elements, the ADS core with the new fuel composition is feasible and consistent with the partitioning and transmutation (P&T) cycle. Based on the new fuel composition, the heat removal during fuel powder storage and fuel assembly assembling is evaluated. For the fuel powder storage, it is found that a cylindrical tube container with a length of 500 [mm] and a diameter of 11 - 21 [mm] should be stored under water. For the fuel assembly assembling, CFD analysis indicates that the cladding tube temperature would satisfy the criterion if the inlet velocity of air is larger than 0.5 [m/s]. Through these studies, the new fuel composition which is consistent with the P&T cycle is obtained and the heat removal with the latest conditions is investigated. It is also shown that the new fuel composition can be practically handled with respect to heat generation, which is one of the most difficult points in handling MA fuel.
Hiyama, Tomoyuki; Aizawa, Kosuke; Nishimura, Masahiro; Kurihara, Akikazu
JAEA-Research 2021-009, 29 Pages, 2021/11
JAEA-Research-2021-009.pdf:2.25MB
In sodium-cooled fast reactors, high burnup of fuel is required for practical use. It is important to predict and evaluate the flow behavior in a fuel assembly because there is a concern that the heat removal capacity of the fuel assembly with high burnup will be locally reduced due to swirling and thermal deformation of the fuel rods. In this study, flow field measurement tests were conducted using a 3-pin bundle system test specimen for the purpose of elucidating the phenomenon and constructing a verification database for thermal hydraulics analysis code. The viewpoints of the experiment for elucidating the phenomenon are as follows; (1) Overall flow behavior in the subchannel including near the wrapping wire, (2) Relationship between Reynolds number including laminar flow region and flow field, and (3) Evaluation of the effect of the presence or absence of wrapping wire on the flow field. As a result, detailed flow field data in the subchannel was obtained by PIV measurement. It was found that when the wrapping wire crossed the subchannel, the flow occurred toward adjacent subchannel and the flow occurred that follows the winding direction of the wrapping wire. It was confirmed that the tendency of the flow velocity distribution of the Reynolds number in the laminar flow region is significantly different from that of the transition region and the turbulent region under the condition. The test was conducted using a same 3-pin bundle system without the wrapping wire, and it was confirmed that mixing by the wrapping wire occurred even in the laminar flow region.
Nagai, Takayuki; Okamoto, Yoshihiro; Yamagishi, Hirona*; Ota, Toshiaki*; Kojima, Kazuo*; Inose, Takehiko*; Sato, Seiichi*; Hatakeyama, Kiyoshi*
JAEA-Research 2021-010, 62 Pages, 2022/01
JAEA-Research-2021-010.pdf:6.05MB
The local structure of glass-forming elements and waste elements in borosilicate glasses varies with its chemical composition. In this study, borosilicate glass frit and simulated waste glass samples were prepared and the local structure and chemical state regarding boron(B), oxygen(O), silicon(Si) and waste elements of iron(Fe), cesium(Cs) were estimated by using XAFS measurement in soft X-ray region. Following results were obtained by XAFS measurements of prepared glass frit and simulated waste glass samples: (1) The effect of Na
O concentration on B-O coordination structure is greater than that of the waste elements concentration. (2) The height of pre-edge by O K-edge spectrum depends on the concentration of first transition elements such as Fe in glass samples. Following results were obtained by XAFS measurements of simulated waste glass samples after immersion test to investigate the long chemical stability. (1) A new compound was formed on the sample surface after the immersion test, and changes in the surface state were confirmed by Raman spectroscopy. (2) Cs on the sample surface after immersion test dissolves into the leaching solution. The Si K-edge XANES spectra of borosilicate glass frits and simulated waste glass samples included lanthanides oxide were measured, and following was confirmed. (1) As the NaO concentration increases in borosilicate glass frit, the K-edge peak of Si shifts to the low energy side. (2) The peak height of the K-edge of Si differs depending on the kind of lanthanide.
Yoshida, Naoki; Ono, Takuya; Yoshida, Ryoichiro; Amano, Yuki; Abe, Hitoshi
JAEA-Research 2021-011, 12 Pages, 2022/01
JAEA-Research-2021-011.pdf:1.49MB
In boiling and drying accidents involving high-level liquid waste in fuel reprocessing plants, emphasis is placed on the behavior of ruthenium (Ru). Ru would form volatile species, such as ruthenium tetroxide (RuO
), and could be released to the environment with coexisting gases, including nitric acid, water, or nitrogen oxides. In this study, to contribute toward safety evaluations of these types of accidents, the migration behavior of gaseous Ru into the liquid phase has been experimentally measured by simulating the condensate during an accident. The gas absorption of RuO was enhanced by increasing the nitrous acid (HNO) concentration in the liquid phase, indicating the occurrence of chemical absorption. In control experiments without HNO, the lower the temperature, the greater was the Ru recovery ratio in the liquid phase. Conversely, in experiments with HNO, the higher the temperature, the higher the recovery ratio, suggesting that the reaction involved in chemical absorption was activated at higher temperatures.
Iwamoto, Hiroki; Meigo, Shinichiro; Nakano, Keita; Yee-Rendon, B.; Katano, Ryota; Sugawara, Takanori; Nishihara, Kenji; Sasa, Toshinobu; Maekawa, Fujio
JAEA-Research 2021-012, 58 Pages, 2022/01
JAEA-Research-2021-012.pdf:7.23MB
A radiation shielding analysis was performed for the structure located above the spallation target of an accelerator-driven system (ADS), assuming one cycle of an 800 MW thermal and 30 MW beam power operation. In this analysis, the Monte Carlo particle transport code PHITS and the activation analysis code DCHAIN-PHITS were used. The structures to be analyzed are a beam duct above the target, a beam transport room located above the ADS reactor vessel, beam transport equipment, and the room ceiling. For each structure, the radiation doses and radioactivities during and after the operation were estimated. Furthermore, the shielding structure of the ceiling was determined. As a result, it was found that the radiation dose at the site boundary would be sufficiently lower than the legal limit by applying the determined shielding structure. Moreover, under the condition of this study, it was shown that the effective dose rate around the beam transport equipment positioned above the target after the operation exceeded 10 mSv/h, and that the maintenance and replacement of the equipment in the room would require remote handling.
Yoshida, Kazuo; Tamaki, Hitoshi; Hiyama, Mina*
JAEA-Research 2021-013, 20 Pages, 2022/01
JAEA-Research-2021-013.pdf:2.35MB
An accident of evaporation to dryness by boiling of high level liquid waste (HLLW) is postulated as one of the severe accidents caused by the loss of cooling function at a fuel reprocessing plant. In this case, volatile radioactive materials, such as ruthenium (Ru) are released from the tanks with water and nitric-acid mixed vapor into atmosphere. An idea has been proposed to implement a steam condenser as an accident countermeasure. This measure is expected to prevent nitric acid steam diffusing in facility building and to increase gaseous Ru trapping ratio into condensed water. A simulation study has been carried out with a hypothetical typical facility building to analyze the efficiency of steam condenser. In this study, SCHERN computer code simulates chemical behaviors of Ru in nitrogen oxide, nitric acid and water mixed vapor based on the conditions obtained from simulation with thermal-hydraulic computer code MELCOR. The effectiveness of steam condenser has been analyzed quantitively in preventing mixed vapor diffusion and gaseous Ru trapping effect. Some issues to be solved in analytical model has been also clarified in this study.
Nakano, Keita; Matsuda, Hiroki*; Meigo, Shinichiro; Iwamoto, Hiroki; Takeshita, Hayato*; Maekawa, Fujio
JAEA-Research 2021-014, 25 Pages, 2022/03
JAEA-Research-2021-014.pdf:2.1MB
For the development of accelerator-driven transmutation system (ADS), measurement of nuclide production cross-sections in proton-induced reactions on 
Be, C, 
Al, 
Sc, and V have been performed. The measured data are compared with the calculations by the latest nuclear reaction models and with the nuclear data library to investigate the reproducibilities.
Shielding Integral Test Working Group, JENDL Committee
JAEA-Research 2021-015, 86 Pages, 2022/03
JAEA-Research-2021-015.pdf:6.15MB
The major revised version of Japanese Evaluated Nuclear Data Library (JENDL), JENDL-4.0, was released in May, 2010. Thus Shielding Integral Test Working Group in JENDL Committee has performed benchmark tests of JENDL-4.0 for the shielding and fusion neutronics fields in order to verify accuracy of JENDL-4.0. We adopted many benchmark experiments: 1) the in-situ and Time-of-Flight (TOF) experiments with DT neutrons at JAEA/FNS, 2) the TOF experiments with DT neutrons at Osaka Univ./OKTAVIAN, 3) the aluminum, iron, nickel and lead experiments with DT neutrons at IPPE, Russia, 4) the iron and stainless steel experiments at ORNL/TSF, US, 5) the ASPIS iron experiment at Winfrith, England, 6) the iron experiments with 
Cf neutron source at KfK, Germany and 7) the iron experiments with Cf neutron source at NIST, US. These experiments were analyzed by using the Monte Carlo code MCNP4 or MCNP5 and/or Sn codes ANISN, DOT3.5 and DORT with mainly JENDL-3.3 and JENDL-4.0. The calculated results demonstrated that JENDL-4.0 led to better agreements with the measured data due to revisions of JENDL-3.3 problems.
Katano, Ryota; Nishihara, Kenji; Kondo, Yasuhiro; Meigo, Shinichiro
JAEA-Research 2021-016, 16 Pages, 2022/03
JAEA-Research-2021-016.pdf:1.65MB
It has to be confirmed that the accelerator-driven system (ADS), which is dedicated to transmuting minor actinides, is subcritical in any state by measurements. In the previous research, we have proposed a procedure in which the core safely and efficiently approaches the target subcriticality before the operation. In this procedure, the reference value of the subcriticality at the initial state is measured by the area ratio method capable of the absolute value measurement. The area ratio method uses a pulsed neutron source. However, specific and practical parameters of the accelerator for the area ratio method have not been determined. In this study, we determined the accelerator parameters with the consideration of the uncertainties derived by the dead-time of the detector and the statistical error of the count ratio. In addition, we estimate the coating amount of the sample nuclide in the assumption of the use of the fission chambers.
Choi, B.; Nishida, Akemi; Kawata, Manabu; Shiomi, Tadahiko; Li, Y.
JAEA-Research 2021-017, 174 Pages, 2022/03
JAEA-Research-2021-017.pdf:9.33MB
Standard methods such as lumped mass models have been used in the assessment of seismic safety and the design of building structures in nuclear facilities. Recent advances in computer capabilities allow the use of three-dimensional finite element (3D FE) models to account for the 3D behavior of buildings, material nonlinearity, and the nonlinear soil-structure interaction effect. Since the 3D FE model enables more complex and high-level treatment than ever before, it is necessary to ensure the reliability of the analytical results generated by the 3D FE model. Guidelines for assuring the dependability of modeling techniques and the treatment of nonlinear aspects of material properties have already been created and technical certifications have been awarded in domains other than nuclear engineering. The International Atomic Energy Agency performed an international benchmark study in nuclear engineering. Multiple organizations reported on the results of seismic response studies using the 3D FE model based on recordings from the Niigata-ken Chuetsuoki Earthquake in 2007. The variation in their analytical results was significant, indicating an urgent need to improve the reliability of the analytical results by standardization of the analytical methods using 3D FE models. Additionally, it has been pointed out that it is necessary to understand the 3D behavior in the seismic fragility assessment of buildings and equipment, which requires evaluating the realistic nonlinear behavior of building facilities when assessing their seismic fragility. In view of these considerations, a standard guideline for the seismic response analysis method using a 3D FE model was produced by incorporating the latest knowledge and findings in this area. The purpose of the guideline is to improve the reliability of the seismic response analysis method using 3D FE model of reactor buildings. The guideline consists of a main body, commentaries, and appendixes; it also provides standard procedures
Nakano, Keita; Iwamoto, Hiroki; Nishihara, Kenji; Meigo, Shinichiro; Sugawara, Takanori; Iwamoto, Yosuke; Takeshita, Hayato*; Maekawa, Fujio
JAEA-Research 2021-018, 41 Pages, 2022/03
JAEA-Research-2021-018.pdf:2.93MB
Neutronic analysis of beam window of the Accelerator-Driven System (ADS) proposed by Japan Atomic Energy Agency (JAEA) has been conducted using PHITS and DCHAIN-PHITS codes. We investigate gas production of hydrogen and helium isotopes in the beam window, displacement per atom of beam window material, and heat generation in the beam window. In addition, distributions of produced nuclides, heat density, and activity are derived. It was found that at the maximum 12500 appm H production, 1800 appm He production, and damage of 62.1 DPA occurred in the beam window by the ADS operation. On the other hand, the maximum heat generation in the beam window was 374 W/cm
. In the analysis of LBE, 
Bi and 
Po were found to be the dominant nuclides in decay heat and radioactivity. Furthermore, the heat generation in the LBE by the proton beam was maximum around 5 cm downstream of the beam window, which was 945 W/cm.
Komatsu, Atsushi
JAEA-Research 2021-019, 24 Pages, 2022/05
JAEA-Research-2021-019.pdf:1.53MB
In order to reduce the corrosion rate of materials in molten lead bismuth eutectic (LBE), it is important to adjust the oxygen concentration, and past reports show that the oxygen concentration is often adjusted to about 10
to 10
wt%. However, it is not clearly stated what concentration is optimal, and there are some reports of severe corrosion even within this concentration range. In this study, a corrosion model considering diffusion in oxide and LBE was developed for 9Cr-1Mo steel, and the corrosion control method estimated from the corrosion model were investigated. We also tried to calculate the optimum oxygen concentration to prevent the flow blockage at the low temperature of loop environment while reducing the corrosion of 9Cr-1Mo steel in molten LBE. As a result, it was expected that the corrosion mode of 9Cr-1Mo steel in LBE could be classified into three types, dense film formation, precipitation film formation, and film dissolution, depending on the ratio of oxide film thickness to diffusion layer thickness, iron concentration in LBE, and temperature. In order to inhibit corrosion, it is important to adjust the oxygen concentration so that the conditions for dense film formation can be maintained. For this purpose, it was expected that a pre-oxidized film of more than 10m should be applied before immersion in LBE. The oxygen concentration of about 10 to 10wt% is the appropriate oxygen concentration when the oxide film has grown to some extent, and a higher oxygen concentration was expected to be required when the film is thin.