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JAEA Reports

Progress report on Nuclear Safety Research Center; JFY 2015 - 2017

Nuclear Safety Research Center, Sector of Nuclear Safety Research and Emergency Preparedness

JAEA-Review 2018-022, 201 Pages, 2019/01

JAEA-Review-2018-022.pdf:20.61MB

Nuclear Safety Research Center (NSRC), Sector of Nuclear Safety Research and Emergency Preparedness, Japan Atomic Energy Agency (JAEA) is conducting technical support to nuclear safety regulation and safety research based on the Mid-Long Term Target determined by Japanese government. This report summarizes the research structure of NSRC and the cooperative research activities with domestic and international organizations as well as the nuclear safety research activities and results in the period from JFY 2015 to 2017 on the nine research fields in NSRC; (1) severe accident analysis, (2) radiation risk analysis, (3) safety of nuclear fuels in light water reactors (LWRs), (4) thermohydraulic behavior under severe accident in LWRs, (5) materials degradation and structural integrity, (6) safety of nuclear fuel cycle facilities, (7) safety management on criticality, (8) safety of radioactive waste management, and (9) nuclear safeguards.

JAEA Reports

Data survey and compilation of material property tables of irradiated stainless steel for evaluation of radiation effects on structural material properties of core internals in pressurized water reactors (Contract research)

Kasahara, Shigeki; Fukuya, Koji*; Fujimoto, Koji*; Fujii, Katsuhiko*; Chimi, Yasuhiro

JAEA-Review 2018-013, 171 Pages, 2019/01

JAEA-Review-2018-013.pdf:6.89MB

For structural integrity assessment of reactor internals of light water reactors, it is important to evaluate and predict the property changes of structural materials, based on existing data obtained from austenitic stainless steel irradiated with neutrons. Compilation of the data into tables is valuable for discussing the representative or the most probable values of the properties applied to the assessment. When the data compilation, the data must be distinguished clearly in consideration of different service conditions of core internals of pressurized water reactors (PWR) and boiling water reactors. Main objective of this work is to provide material property tables of irradiated austenitic stainless steel which will be applicable for assessment of structural integrity of core internals of PWRs. To compile the table, published literature reporting irradiated stainless steel data were surveyed and screened by considering the service conditions of PWRs. In addition to the data, various parameters for the data evaluation, e.g. chemical compositions and pre-treatments of the materials, irradiation and examination conditions, were extracted from the literature, and listed into tables.

JAEA Reports

Development of fuel performance code FEMAXI-8; Model improvements for light water reactor fuel analysis and systematic validation

Udagawa, Yutaka; Yamauchi, Akihiro*; Kitano, Koji*; Amaya, Masaki

JAEA-Data/Code 2018-016, 79 Pages, 2019/01

JAEA-Data-Code-2018-016.pdf:2.75MB

FEMAXI-8 is the latest version of the fuel performance code FEMAXI developed by JAEA. A systematic validation work has been achieved against 144 irradiation test cases, after many efforts have been made, in development of new models, improvements in existing models and the code structure, bug-fixes, construction of irradiation-tests database and other infrastructures.

Journal Articles

Study on restricted use of contaminated rubble on Fukushima Daiichi NPS site, 2; Validation of reference radiocesium concentration for recycling materials

Miwa, Kazuji; Shimada, Taro; Takeda, Seiji

Progress in Nuclear Science and Technology (Internet), 6, p.166 - 170, 2019/01

In this study, in order to validate the restricted use of recycling material at the reference radiocesium concentration (determined in series report (1)), we evaluated worker annual doses, air dose rate at the site boundary and impact of migrated radiocesium into groundwater. Firstly, we evaluated the additional annual dose for workers, on the assumption that typical workers coming in contact with the source after construction (Road: 1.2 mSv/y, Building: 1.3 mSv/y). Secondly, we evaluated the air dose rates by distance from road and building including recycling material, and investigated the distance for not exceeding 1 mSv/y (including additional dose rate by recycling and background dose rate of 0.6 mSv/y) at the site boundary (Road: 25 m, Building: 1 m). Thirdly, we evaluated the Cs migration in groundwater, and investigated the distance required for satisfying the operation target value ($$^{134}$$Cs: 1 Bq/L, $$^{137}$$Cs: 1 Bq/L) at the boundary (coastal line) (Road: 10 m, Building: 10 m).

Journal Articles

Study on restricted use of contaminated rubble on Fukushima Daiichi NPS site, 1; Estimation of reference radiocesium concentration for recycling materials

Shimada, Taro; Miwa, Kazuji; Takeda, Seiji

Progress in Nuclear Science and Technology (Internet), 6, p.203 - 207, 2019/01

Rubbles less than 5 $$mu$$Sv/h of surface dose rate, which are stored outdoor in the Fukushima Daiichi NPS (1F) site, will be recycled and applied in a restricted reuse only within 1F site in the future. In this study, we suggested a concept for establishing the reference radioactive concentration of recycling material for the restricted use in the 1F site. Reference radiocesium concentration is calculated so that increased dose rate by restricted reuse does not exceed 1 $$mu$$Sv/h which is the minimum value of dose rate map in the 1F entire site. In order to justify the restricted reuse under the reference concentration calculated, additional occupational dose, dose rate at the site boundary and groundwater concentration at the outlet to the ocean are evaluated and confirmed that the values are below 2 mSv/y, 1 mSv/y and 1 Bq/cm$$^{3}$$ of $$^{134}$$Cs and $$^{137}$$Cs, respectively. And then calculated the reference radiocesium concentrations of the recycling material used for paved roads and the bases of concrete building.

Journal Articles

A Probabilistic Approach to Assess External Doses to the Public Considering Spatial Variability of Radioactive Contamination and Interpopulation Differences in Behavior Pattern

Takahara, Shogo; Iijima, Masashi*; Yoneda, Minoru*; Shimada, Yoko*

Risk Analysis, 39(1), p.212 - 224, 2019/01

A dose assessment model was developed based on measurements and surveys of individual doses and relevant contributors in Fukushima City for four population groups: Fukushima City Office staff, Senior Citizens' Club, Contractors' Association, and AgriculturalCooperative. In addition, probabilistic assessments were performed for these population groups by considering the spatial variability of contamination and interpopulation differencesresulting from behavior patterns. As a result of comparison with the actual measurements, the assessment results for participants from the Fukushima City Office, Senior Citizens' Club and the Agricultural Cooperative agreed with the measured values. By contrast, the measurements obtained for the participants from the Contractors' Association were not reproduced well in the present study. To assess the doses to this group, further investigations of association members' work activities and the related dose reduction effects are needed.

Journal Articles

Influence evaluation of loading conditions during pressurized thermal shock transients based on thermal-hydraulics and structural analyses

Katsuyama, Jinya; Uno, Shumpei*; Watanabe, Tadashi*; Li, Y.

Frontiers of Mechanical Engineering, 13(4), p.563 - 570, 2018/12

For the structural integrity assessments on reactor pressure vessels (RPVs) under pressurized thermal shock (PTS) events, thermal hydraulic (TH) behavior of coolant water is one of the most important influence factors. Configuration of plant equipment and their dimensions, and operator action have large influences on TH behavior. In this study, to investigate the influence of operator action on TH behavior during a PTS event, we developed an analysis model for a typical Japanese plant, and performed TH and structural analyses. Two different operator action times were assumed based on the Japanese and US' rules. From the analysis results, it was clarified that differences in operator action times have a significant effect on TH behavior and loading conditions, that is, following the Japanese rule may lead to lower stresses compared to that when following the US rule because earlier operator action caused lower pressure in the RPV.

JAEA Reports

Annual report of Nuclear Emergency Assistance and Training Center; April 1, 2017 to March 31, 2018

Nuclear Emergency Assistance and Training Center

JAEA-Review 2018-015, 78 Pages, 2018/11

JAEA-Review-2018-015.pdf:3.85MB

Since JAEA is a designated public institution, an agency dealing with an emergency situation in cooperation with the national government under the Disaster Countermeasures Basic Act, it has the responsibility of providing technical assistance to the national government in case of a nuclear emergency. The Nuclear Emergency Assistance and Training Center, NEAT, is the main center of the technical assistance in case of emergency, and dispatches experts of JAEA and supplies equipment and materials to the national government for emergency. In normal time, the NEAT provides the technical assistance such as the exercises and training courses concerning nuclear preparedness and response to emergency responders including the national government officers in addition to JAEA staff members. This report introduces the results of activities in FY2017 conducted by NEAT.

JAEA Reports

Data survey and compilation of material property tables of irradiated stainless steel for evaluation of radiation effects on structural material properties of core internals in boiling water reactors (Contract Research)

Kasahara, Shigeki; Fukuya, Koji*; Koshiishi, Masato*; Fujii, Katsuhiko*; Chimi, Yasuhiro

JAEA-Review 2018-012, 180 Pages, 2018/11

JAEA-Review-2018-012.pdf:10.71MB

For structural integrity assessment of reactor internals of light water reactors, it is important to evaluate and predict the property changes of structural materials, based on existing data obtained from austenitic stainless steel irradiated with neutrons. Compilation of the data into tables is valuable for discussing the representative or the most probable values of the properties applied to the assessment. In the process of the data compilation, the data must be distinguished clearly in consideration of different service conditions of core internals of boiling water reactors (BWR) and pressurized water reactors. Main objective of this work is to provide material property tables of irradiated austenitic stainless steel which will be applicable for assessment of structural integrity of core internals of BWRs. To compile the table, published literature reporting irradiated stainless steel data were surveyed and screened by considering the service conditions of BWRs. In addition to the data, various parameters for the data evaluation, e.g. chemical compositions and pre-treatments of the materials, irradiation and examination conditions, were extracted from the literature, and listed into the tables.

JAEA Reports

Mechanical properties database of reactor pressure vessel steels related to fracture toughness evaluation

Tobita, Toru; Nishiyama, Yutaka; Onizawa, Kunio

JAEA-Data/Code 2018-013, 60 Pages, 2018/11

JAEA-Data-Code-2018-013.pdf:1.67MB

Mechanical properties of materials including fracture toughness are extremely important for evaluating the structural integrity of reactor pressure vessels (RPVs). In this report, the published data of mechanical properties of nuclear RPVs steels, including neutron irradiated materials, acquired by the Japan Atomic Energy Agency (JAEA), specifically tensile test data, Charpy impact test data, drop-weight test data, and fracture toughness test data, are summarized. There are five types of RPVs steels with different toughness levels equivalent to JIS SQV2A (ASTM A533B Class 1) containing impurities in the range corresponding to the early plant to the latest plant. In addition to the base material of RPVs, the mechanical property data of the two types of stainless overlay cladding materials used as the lining of the RPV are summarized as well. These mechanical property data are organized graphically for each material and listed in tabular form to facilitate easy utilization of data.

Journal Articles

Development and application of a method for discriminating the influence of radon progenies in air from aerial radiation monitoring data

Hirouchi, Jun; Nishizawa, Yukiyasu*; Urabe, Yoshimi*; Shimada, Kazumasa; Sanada, Yukihisa; Munakata, Masahiro

Applied Radiation and Isotopes, 141, p.122 - 129, 2018/11

The influence of $$gamma$$-rays from natural nuclides (particularly the radon progenies, $$^{214}$$Pb and $$^{214}$$Bi) must be excluded from aerial radiation monitoring (ARM) data to accurately estimate the deposition amount of artificial radionuclides. A method for discriminating the influence of $$^{214}$$Pb and $$^{214}$$Bi in air from the ARM data was developed. The influence of the radon progenies in air was excluded using the relation between the count rates of six NaI (Tl) detectors and a LaBr$$_{3}$$ detector. The discrimination method was applied to the ARM data obtained from around the Sendai Nuclear Power Station. To verify the validity of the discrimination method, the dose rate estimated from the ARM data was compared with the dose rate measured using a NaI survey meter at a height of 1 m above the ground. The application of the discrimination method improved the dose rate estimation, showing the validity of the discrimination method.

Journal Articles

Failure behavior analyses of piping system under dynamic seismic loading

Udagawa, Makoto; Li, Y.; Nishida, Akemi; Nakamura, Izumi*

International Journal of Pressure Vessels and Piping, 167, p.2 - 10, 2018/11

It is important to assure the structural Integrity of piping systems under severe earthquakes because those systems comprise the pressure boundary for coolant with high pressure and temperature. In this study, we examine the seismic safety capacity of piping systems under severe dynamic seismic loading using a series of dynamic-elastic-plastic analyses focusing on dynamic excitation experiments of 3D piping systems which was tested by NIED. Analytical results were consistent with experimental data in terms of natural frequency, natural vibration mode, response accelerations, elbow opening-closing displacements, strain histories, failure position, and low-cycle fatigue failure lives. Based on these results, we concluded that the analytical model used in the study can be applied to failure behavior evaluation for piping systems under severe dynamic seismic loading.

Journal Articles

Analysis of transport behaviors of cesium and iodine in VERDON-2 experiment for chemical model validation

Shiotsu, Hiroyuki; Ito, Hiroto*; Ishikawa, Jun; Sugiyama, Tomoyuki; Maruyama, Yu

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 6 Pages, 2018/11

The VERDON-2 experiment for FPs transport in steam environment was analyzed with the mechanistic FPs transport code incorporating thermodynamic chemical equilibrium model in order to assess its predictive capability for transport behavior of key FPs, especially for highly volatile FPs such as Cs and I. The present analysis reproduced well the Cs deposition profile obtained from the experiment, which revealed that Cs was transported as CsOH in early phase of FP release from fuel, and then formed Cs$$_{2}$$MoO$$_{4}$$ after increasing Mo release. On the other hand, the deposition peak of I was predicted to appear at 720 K, which was significantly higher than the experimental result at 600 K. This discrepancy was potentially caused by the following two points: lack of the other stable species in thermodynamics database for thermodynamic chemical equilibrium model, or failure of chemical equilibrium assumption for iodide species.

Journal Articles

Computational fluid dynamics analysis for hydrogen deflagration tests at ENACCEF2 facility

Trianti, N.; Sato, Masatoshi*; Sugiyama, Tomoyuki; Maruyama, Yu

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 7 Pages, 2018/11

Simulation techniques have been developed to analyze the deflagration behavior of hydrogen generated during a hypothetical severe accident in nuclear power plants. The CFD analysis was carried out on the hydrogen deflagration experiment performed at the ENACCEF2 facility composed mainly of a vertical cylindrical tube filled with hydrogen-air mixture and nine annular obstacles were placed in the lower part of the tube. The simulation was carried out by the reactingFoam solver of OpenFOAM 3.0, an open source software for the CFD analysis. The RNG (Renormalization group) k-$$varepsilon$$ model was applied for turbulent flow. The interaction of the chemical reaction with the turbulent flow was considered using PaSR (Partial Stirred Reactor) model with 19 elementary reactions for the hydrogen combustion. The analysis result showed the characteristic of flame acceleration by the obstacle region was qualitatively reproduced even though has discrepancy with the experiment.

Journal Articles

Study on criticality in natural barrier for disposal of fuel debris from Fukushima Daiichi NPS

Shimada, Taro; Takubo, Kazuya*; Takeda, Seiji; Yamaguchi, Tetsuji

Progress in Nuclear Science and Technology (Internet), 5, p.183 - 187, 2018/11

After fuel debris is removed from the reactor containment vessel at Fukushima Daiichi NPS (1F) and collected in waste containers in the future, the waste containers will be disposed at a deep geological repository. The uranium inventory and uranium-235 ($$^{235}$$U) enrichment of the fuel debris are larger than those of high-level vitrified wastes which are produced from liquid waste during reprocessing of spent nuclear fuels. Therefore, there is a possibility not to be excluded that a criticality occurs in the geological media where the uranium precipitates at the far-field from the repository, after the uranium located in the repository is dissolved by groundwater. In this study, we calculated the quantity of uranium precipitated at the natural barrier, and studied dimension of uranium deposited in the natural barrier and carried out the criticality analysis.

JAEA Reports

Activities on predictions of atmospheric dispersion of radionuclides for nuclear tests by North Korea

Ishizaki, Shuhei; Hayakawa, Tsuyoshi; Tsuzuki, Katsunori; Terada, Hiroaki; Togawa, Orihiko

JAEA-Technology 2018-007, 43 Pages, 2018/10

JAEA-Technology-2018-007.pdf:5.67MB

When North Korea has carried out a nuclear test, by a request from Nuclear Regulation Authority (NRA), Nuclear Emergency Assistance and Training Center (NEAT) predicts atmospheric dispersion of radionuclides by WSPEEDI-II system in cooperation with Nuclear Science and Engineering Center (NSEC), and submits the predicted results to NRA as the activity to assist responses by the Japanese Government. This report explains frameworks of the Japanese Government and Japan Atomic Energy Agency (JAEA) to cope with nuclear tests by North Korea, and describes a series of activities by NEAT regarding predictions of atmospheric dispersion of radionuclides in response to the 5th and 6th nuclear tests carried out by North Korea in September 2016 and September 2017. Future plans and issues to be solved for responses to nuclear tests are also described in this report, together with an outline of a computer program system used in the predictions.

JAEA Reports

CHEMKEq; Evaluation code for chemical composition based on partial mixed model with Chemical Equilibrium and Reaction Kinetics (Contract research)

Ito, Hiroto*; Shiotsu, Hiroyuki; Tanaka, Yoichi*; Nishihara, Satomichi*; Sugiyama, Tomoyuki; Maruyama, Yu

JAEA-Data/Code 2018-012, 42 Pages, 2018/10

JAEA-Data-Code-2018-012.pdf:4.93MB

Chemical composition of fission products transported in nuclear facilities in severe accidents is controlled by slower chemical reaction rates, therefore, it could be different from that evaluated on the chemical equilibrium assumption. Hence, it is necessary to evaluate the chemical composition with reaction kinetics. On the other hand, databases applicable to the analysis of nuclear facilities have not been constructed because knowledge of reaction rates of complex chemical reactions in severe accidents is currently limited. Accordingly, we have developed the CHEMKEq code based on a partial mixed model with chemical equilibrium and reaction kinetics to decrease uncertainties of the chemical composition caused by the reaction rate. The CHEMKEq code, under mass conservation law, firstly evaluates chemical species obeying the chemical equilibrium model, and then, relatively slow reactions are solved by the reaction kinetics model. Moreover, the CHEMKEq code has a multiplicity of use in evaluations of chemical composition because general chemical equilibrium and reaction kinetics models are also available and databases required to calculation are external file formats. This report is the user's guide of the CHEMKEq code, showing models, solution methods, structure of the code and calculation examples. And information to run the CHEMKEq code is summarized in appendixes.

Journal Articles

Effects of ballooning and rupture on the fracture resistance of Zircaloy-4 fuel cladding tube after LOCA-simulated experiments

Yumura, Takanori; Amaya, Masaki

Annals of Nuclear Energy, 120, p.798 - 804, 2018/10

To investigate the relationship between the fracture resistance of a cladding tube and the amount of deformation of the cladding tube due to ballooning and rupture during a loss-of-coolant accident (LOCA), four-point-bending tests were performed using non-irradiated Zircaloy-4 cladding tubes which experienced a LOCA-simulated sequence (ballooning, rupture, high temperature oxidation and quench). According to the obtained results, it was found that the maximum bending stress of the cladding tube after the LOCA-simulated sequence, which was defined as the fracture resistance, correlated to the average thickness of prior-$$beta$$ layer in the cladding tube. Based on the average thickness of prior-$$beta$$ layer, the fracture resistance of the cladding tube with ballooning and rupture was expressed as functions of isothermal oxidation time and temperature and the maximum circumferential strain on the cladding tube.

Journal Articles

Trace analysis for nuclear non-proliferation and security

Esaka, Fumitaka

Bunseki, 2018(10), p.408 - 411, 2018/10

Recent development of analytical techniques for identification of particles containing nuclear materials, isotope ratio analysis of uranium and plutonium using mass spectrometry, and age determination is described in this paper. These techniques are successfully applied to the trace analysis for nuclear non-proliferation.

Journal Articles

Monte Carlo criticality analysis of random media under bounded fluctuation driven by normal noise

Ueki, Taro

Journal of Nuclear Science and Technology, 55(10), p.1180 - 1192, 2018/10

In Monte Carlo criticality analysis under material distribution uncertainty, it is necessary to evaluate the response of neutron effective multiplication factor ($$K_{rm eff}$$) to the space-dependent random fluctuation of volume fractions within a prescribed bounded range. Normal random variables, however, cannot be used in a straightforward manner since the normal distribution has infinite tails. To overcome this issue, a methodology has been developed via forward-backward-superposed reflection Brownian motion (FBSRBM). Here, the forward-backward superposition makes the variance of fluctuation spatially constant and the reflection Brownian motion confines the fluctuation driven by normal noise in a bounded range. FBSRBM was implemented using Karhunen-Loeve expansion and applied to the fluctuation of volume fractions in a model of UO$$_{2}$$-concrete media with stainless steel.

2934 (Records 1-20 displayed on this page)