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.
Nuclear Emergency Assistance and Training Center
JAEA-Review 2018-015, 78 Pages, 2018/11
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.
Kasahara, Shigeki; Fukuya, Koji*; Koshiishi, Masato*; Fujii, Katsuhiko*; Chimi, Yasuhiro
JAEA-Review 2018-012, 180 Pages, 2018/11
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.
Tobita, Toru; Nishiyama, Yutaka; Onizawa, Kunio
JAEA-Data/Code 2018-013, 60 Pages, 2018/11
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.
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 -rays from natural nuclides (particularly the radon progenies, Pb and 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 Pb and 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 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.
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.
Ishizaki, Shuhei; Hayakawa, Tsuyoshi; Tsuzuki, Katsunori; Terada, Hiroaki; Togawa, Orihiko
JAEA-Technology 2018-007, 43 Pages, 2018/10
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.
Ito, Hiroto*; Shiotsu, Hiroyuki; Tanaka, Yoichi*; Nishihara, Satomichi*; Sugiyama, Tomoyuki; Maruyama, Yu
JAEA-Data/Code 2018-012, 42 Pages, 2018/10
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.
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- layer in the cladding tube. Based on the average thickness of prior- 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.
Bunseki, (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.
Hasegawa, Kunio*; Strnadel, B.*; Li, Y.; Lacroix, V.*
Journal of Pressure Vessel Technology, 140(5), p.051204_1 - 051204_7, 2018/10
Subsurface flaws are sometimes found as blowholes near free surfaces of structural components. It can be easily expected that the stress intensity factor at the tip of the subsurface flaw increases with decreasing the ligament distance. Fitness-for-service (FFS) codes provide flaw-to-surface proximity rules which are transformation from subsurface to surface flaw. Although the concept of the proximity rules of the FFS codes are the same, the specific criteria for the rules on transforming subsurface flaws to surface flaws are different amongst FFS codes. This study demonstrates the proximity criteria provided by the FFS codes and indicates that the increment of the stress intensity factors before and after the transformation from subsurface to surface flaws. In addition, it is shown that remaining fatigue lives for pipes with flaws are strongly affected by the location at the transformation from subsurface to surface flaws.
Ha, Yoosung; Tobita, Toru; Otsu, Takuyo; Takamizawa, Hisashi; Nishiyama, Yutaka
Journal of Pressure Vessel Technology, 140(5), p.051402_1 - 051402_6, 2018/10
The applicability of miniature compact tension (Mini-C(T)) specimens to fracture toughness evaluation of neutron-irradiated reactor pressure vessel (RPV) steels was investigated. Three types of RPV steels neutron-irradiated to a high-fluence region were prepared and manufactured as Mini-C(T) specimens according to Japan Electric Association Code (JEAC) 4216-2015. Through careful selection of the test temperature by considering previously obtained mechanical properties data, valid fracture toughness, and reference temperature T was obtained with a relatively small number of specimens. Comparing the fracture toughness and T values determined using other larger specimens with those determined using the Mini-C(T) specimens, T values of both unirradiated and irradiated Mini-C(T) specimens were found to be the acceptable margin of error. The scatter of 1T-equivalent fracture toughness values of both unirradiated and irradiated materials obtained using Mini-C(T) specimens did not differ significantly from the values obtained using larger specimens. The correlation between the Charpy 41J transition temperature (T) and the T values agreed very well with that of the data in the literature, regardless of specimen size and fracture toughness of the materials before irradiation. Based on these findings, it was concluded that Mini-C(T) specimens can be applied to fracture toughness evaluation of neutron-irradiated materials without significant specimen size dependence.
Watanabe, Tadashi*; Ishigaki, Masahiro*; Katsuyama, Jinya
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 9 Pages, 2018/10
The analyses of LSTF experiment and PWR plant for 5% cold-leg break LOCA are performed using the RELAP5/MOD3.3 code. The discharge coefficient of critical flow model is determined so as to obtain the agreement of pressure transient between the LSTF experiment and the experimental analysis, and used for the PWR analysis. The characteristics of thermal-hydraulic phenomena in the experiment are shown to be simulated well by the two analyses. The decrease in core differential pressure during the loop-seal clearing is, however, underestimated by the two analyses, and the core heat up is not predicted. The loop flow rates are also underestimated by the two analyses. Although the duration of core heat up during the boil-off period is longer in the experimental analysis, the results of two analyses agree well, and the effect of scaling is found to be small between the experimental analysis and the PWR analysis.
Dostl, M.*; Rossiter, G.*; Dethioux, A.*; Zhang, J.*; Amaya, Masaki; Rozzia, D.*; Williamson, R.*; Kozlowski, T.*; Hill, I.*; Martin, J.-F.*
Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2018) (Internet), 10 Pages, 2018/10
The benchmark on PCMI was initiated by OECD/NEA Expert Group on Reactor Fuel Performance (EGRFP) in June 2015 and is currently in the latter stages of compiling results and preparing the final report. The aim of the benchmark is to improve understanding and modelling of PCMI amongst NEA member organisations. This is being achieved by comparing PCMI predictions of different fuel performance codes for a number of cases. Two of these cases are hypothetical cases aiming to facilitate understanding of the effects of code-to-code differences in fuel performance models. The two remaining cases are actual irradiations, where code predictions are compared with measured data. During analysis of participants' results of the hypothetical cases, the assumptions for number of radial pellet cracks and the pellet-clad friction coefficient (which can be zero, finite or infinite) were identified to be important factors in explaining differences between predictions once pellet-cladding contact occurs. However, these parameters varied in the models and codes used originally by the participants. This fact led to the extension of the benchmark by inclusion of two additional cases, where the number of radial pellet cracks and three different values of the friction coefficient were prescribed in the case definition. Seven calculations from six organisations contributed results were compared and analysed in this paper.
Hirouchi, Jun; Takahara, Shogo; Komagamine, Hiroshi*; Munakata, Masahiro
Proceedings of Asian Symposium on Risk Assessment and Management 2018 (ASRAM 2018) (USB Flash Drive), 8 Pages, 2018/10
Narukawa, Takafumi; Yamaguchi, Akira*; Jang, S.*; Amaya, Masaki
Proceedings of 14th International Conference on Probabilistic Safety Assessment and Management (PSAM-14) (USB Flash Drive), 10 Pages, 2018/09
The reduction of epistemic uncertainty for safety-related events that rarely occur or require high experimental costs is a key concern for researchers worldwide. In this study, we develop a new framework to effectively reduce parameter uncertainty, which is one of the epistemic uncertainties, by using the Bayesian optimal experimental design. In the experimental design, we used a decision theory that minimizes the Bayes generalization loss. For this purpose, we used the functional variance, which is a component of widely applicable information criterion, as a decision criterion for selecting informative data points. Then, we conducted a case study to apply the proposed framework to reduce the parameter uncertainty in the fracture boundary of a non-irradiated, pre-hydrided Zircaloy-4 cladding tube specimen under loss-of-coolant accident (LOCA) conditions. The results of our case study proved that the proposed framework greatly reduced the Bayes generalization loss with minimal sample size compared with the case in which experimental data were randomly obtained. Thus, the proposed framework is useful for effectively reducing the parameter uncertainty of safety-related events that rarely occur or require high experimental costs.
Kaku Nenryo, 53(2), P. 5, 2018/08
Takeda, Takeshi; Otsu, Iwao
Mechanical Engineering Journal (Internet), 5(4), p.18-00077_1 - 18-00077_14, 2018/08
We conducted an experiment focusing on nitrogen gas behavior during reflux condensation in PWR with ROSA/LSTF. The primary pressure was lower than 1 MPa under constant core power of 0.7% of volumetric-scaled (1/48) PWR nominal power. Steam generator (SG) secondary-side collapsed liquid level was maintained at certain liquid level above SG U-tube height. Nitrogen gas was injected stepwise into each SG inlet plenum at certain constant amount. The primary pressure and degree of subcooling of SG U-tubes were largely dependent on amount of nitrogen gas accumulated in SG U-tubes. Nitrogen gas accumulated from outlet towards inlet of SG U-tubes. Non-uniform flow behavior was observed among SG U-tubes with nitrogen gas ingress. The RELAP5/MOD3.3 code indicated remaining problems in predictions of the primary pressure and degree of subcooling of SG U-tubes depending on number of nitrogen gas injection. We studied further non-uniform flow behavior through sensitivity analyses.
Nishida, Akemi; Nagai, Minoru*; Tsubota, Haruji; Li, Y.
Mechanical Engineering Journal (Internet), 5(5), p.18-00087_1 - 18-00087_21, 2018/08
To date, oblique impact has not been studied and few experimental data on the local damage of reinforced concrete (RC) panels exist for oblique impact of deformable projectiles. The final purpose of this study is to propose a new formula for evaluating the local damage to reinforced concrete structures caused by oblique impact based on past experimental results and simulation results. As the first step of this final purpose, we validate the analytical method by comparison with the experimental results and simulate the damage caused by oblique impact using the validated method. First, we analyze and simulate the local damage of RC panel caused by a deformable projectile owing to an impact test normal to the target structure to verify the validity of the simulation analysis. Next, we perform simulation analyses for evaluating the perforation of RC panel due to oblique impact by the deformable projectile and present the results. Various response characteristics and perforation mechanisms to be the basis of examination of oblique impact evaluation were clarified in this paper.
Takeda, Takeshi; Otsu, Iwao
Nuclear Engineering and Technology, 50(6), p.829 - 841, 2018/08
An experiment was conducted for OECD/NEA ROSA-2 Project using LSTF, which simulated 17% hot leg intermediate-break LOCA in PWR. Core uncovery started simultaneously with liquid level drop in crossover leg downflow-side before loop seal clearing, and water remaining occurred on upper core plate. Results of uncertainty analysis with RELAP5/MOD3.3 code clarified influences of combination of multiple uncertain parameters on peak cladding temperature within defined uncertain ranges. An experiment was performed for OECD/NEA PKL-3 Project with PKL. The LSTF test simulated PWR 1% hot leg small-break LOCA with steam generator secondary-side depressurization as accident management measure and nitrogen gas inflow. Some discrepancies appeared between the LSTF and PKL test results for primary pressure, core collapsed liquid level, and cladding surface temperature probably due to effects of differences between LSTF and PKL in configuration, geometry, and volumetric size.