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JAEA Reports

Annual report of Nuclear Emergency Assistance and Training Center (April 1, 2013 - March 31, 2014)

Sato, Takeshi; Muto, Shigeo; Akiyama, Kiyomitsu; Aoki, Kazufumi; Okamoto, Akiko; Kawakami, Takeshi; Kume, Nobuhide; Nakanishi, Chika; Koie, Masahiro; Kawamata, Hiroyuki; et al.

JAEA-Review 2014-048, 69 Pages, 2015/02

JAEA-Review-2014-048.pdf:13.91MB

JAEA was assigned as a designated public institution under the Disaster Countermeasures Basic Act and under the Armed Attack Situations Response Act. Based on these Acts, the JAEA has the responsibility of providing technical support to the national government and/or local governments in case of disaster responses or response in the event of a military attack, etc. In order to fulfill the tasks, the JAEA has established the Emergency Action Plan and the Civil Protection Action Plan. In case of a nuclear emergency, NEAT dispatches specialists of JAEA, supplies the national government and local governments with emergency equipment and materials, and gives technical advice and information. In normal time, NEAT provides various exercises and training courses concerning nuclear disaster prevention to those personnel taking an active part in emergency response institutions of the national and local governments, police, fire fighters, self-defense forces, etc. in addition to the JAEA itself. The NEAT also researches nuclear disaster preparedness and response, and cooperates with international organizations. In the FY2013, the NEAT accomplished the following tasks: (1) Technical support activities as a designated public institution in cooperation with the national and local governments, etc. (2) Human resource development, exercise and training of nuclear emergency response personnel for the national and local governments, etc. (3) Researches on nuclear disaster preparedness and response, and sending useful information (4) International contributions to Asian countries on nuclear disaster preparedness and response in collaboration with the international organizations

JAEA Reports

Annual report of Nuclear Emergency Assistance and Training Center (April 1, 2012 - March 31, 2013)

Sato, Takeshi; Muto, Shigeo; Okuno, Hiroshi; Katagiri, Hiromi; Akiyama, Kiyomitsu; Okamoto, Akiko; Koie, Masahiro; Ikeda, Takeshi; Nemotochi, Toshimasa; Saito, Toru; et al.

JAEA-Review 2013-046, 65 Pages, 2014/02

JAEA-Review-2013-046.pdf:11.18MB

When a nuclear emergency occurs in Japan, the Japan Atomic Energy Agency (JAEA) has the responsibility of providing technical support to the National government, local governments, police, fire stations and nuclear operators etc., because the JAEA has been designated as the Designated Public Institution under the Basic Act on Disaster Control Measures and the Act on Response to Armed Attack Situations, etc.. The Nuclear Emergency Assistance and Training Center (NEAT) of JAEA provides a comprehensive range of technical support activities to an Off-Site Center in case of a nuclear emergency. Specifically, NEAT gives technical advice and information, dispatches specialists as required, and supplies the National Government and local governments with emergency equipments and materials. NEAT provides various exercise and training courses concerning nuclear disaster prevention to those personnel taking an active part in emergency response organizations at normal times. The tasks of NEAT, with its past experiences as a designated public institution including the responses to TEPCO's Fukushima Accident, have been shifted to technical supports to the national government for strengthening its abilities to emergency responses; the NEAT therefore focused on maintenance and operation of its functions, and strengthening its response abilities in cooperation with the national government. This annual report summarized these activities of JAEA/NEAT in the fiscal year 2012.

JAEA Reports

Annual report of Nuclear Emergency Assistance and Training Center (April 1, 2011 - March 31, 2012)

Katagiri, Hiromi; Okuno, Hiroshi; Okamoto, Akiko; Ikeda, Takeshi; Tamura, Kenichi; Nagakura, Tomohiro; Nakanishi, Chika; Yamamoto, Kazuya; Abe, Minako; Sato, Sohei; et al.

JAEA-Review 2012-033, 70 Pages, 2012/08

JAEA-Review-2012-033.pdf:6.38MB

When a nuclear emergency occurs in Japan, JAEA has the responsibility of providing technical support to the National government, local governments, etc., by the Basic Law on Emergency Preparedness and the Basic Plan for Disaster Countermeasures. NEAT of JAEA gives technical advice and information, dispatch specialists as required, supplies with the National Government and local governments emergency equipment and materials. NEAT provides various lectures and training courses concerning nuclear disaster prevention for emergency response organizations at normal time. Concerning the assistance to the Accident of Fukushima No.1 Nuclear Power Station caused by the Great East Japan Earthquake on 11 March, 2011, JAEA assisted activities including environmental radiation monitoring, environmental radioactivity analyses, resident public consulting etc., with its the utmost effort. This annual report summarized these activities of NEAT in the fiscal year 2011.

JAEA Reports

Annual report of Nuclear Emergency Assistance and Training Center (April 1, 2010 - March 31, 2011)

Katagiri, Hiromi; Okuno, Hiroshi; Sawahata, Masayoshi; Ikeda, Takeshi; Sato, Sohei; Terakado, Naoya; Nagakura, Tomohiro; Nakanishi, Chika; Fukumoto, Masahiro; Yamamoto, Kazuya; et al.

JAEA-Review 2011-037, 66 Pages, 2011/12

JAEA-Review-2011-037.pdf:4.52MB

When a Nuclear emergency occurs, Nuclear Emergency Assistance & Training Center (NEAT) of JAEA gives technical advice and information, dispatch specialists as required, supplies emergency equipment and materials to the National Government and local governments. NEAT provides various lectures and training courses concerning nuclear disaster prevention for those personnel taking an active part in emergency response organizations at normal time. NEAT also researches on nuclear disaster prevention and cooperates with international organizations. Concerning about the assistance to the Accident of Fukushima No.1 Nuclear Power Station caused by the Great East Japan Earthquake at 11 March, 2011, JAEA assisted activities including environmental radiation monitoring, environmental radioactivity analyses, resident public consulting etc., with its full scale effort. NEAT served as the center of these supporting activities of JAEA.

JAEA Reports

Annual report of Nuclear Emergency Assistance and Training Center (April 1, 2009 - March 31, 2010)

Kanamori, Masashi; Shirakawa, Yusuke; Yamashita, Toshiyuki; Okuno, Hiroshi; Terunuma, Hiroshi; Ikeda, Takeshi; Sato, Sohei; Terakado, Naoya; Nagakura, Tomohiro; Fukumoto, Masahiro; et al.

JAEA-Review 2010-037, 60 Pages, 2010/09

JAEA-Review-2010-037.pdf:3.11MB

When a nuclear emergency occurs in Japan, the Japan Atomic Energy Agency (JAEA) provides technical support to the National government, local governments, police, fire station and license holder etc. They are designated public organizations conforming to the basic law on emergency preparedness and the basic plan for disaster countermeasures. The Nuclear Emergency Assistance & Training Center (NEAT) of JAEA provides a comprehensive range of technical support activities to an off-site center in case of a nuclear emergency. Specifically, NEAT gives technical advice and information, provides for the dispatch of specialist as required, supplies emergency equipments and materials to the national government and municipal office. NEAT provide various lectures and training course concerning nuclear disaster prevention for those personnel taking an active part in emergency response organizations at normal time. And NEAT researches on nuclear disaster prevention and also cooperate with international organizations. This annual report summarized the activities of JAEA/NEAT in the fiscal year 2009.

JAEA Reports

Annual report of Nuclear Emergency Assistance and Training Center (April 1, 2008 - March 31, 2009)

Kanamori, Masashi; Hashimoto, Kazuichiro; Terunuma, Hiroshi; Ikeda, Takeshi; Omura, Akiko; Terakado, Naoya; Nagakura, Tomohiro; Fukumoto, Masahiro; Watanabe, Fumitaka; Yamamoto, Kazuya; et al.

JAEA-Review 2009-023, 61 Pages, 2009/09

JAEA-Review-2009-023.pdf:8.49MB

When a nuclear emergency occurs in Japan, the Japan Atomic Energy Agency (JAEA) provides technical support to the National government, local governments, police, fire station and license holder etc. They are Designated Public Organizations conforming to the Basic Law on Emergency Preparedness and the Basic Plan for Disaster Countermeasures. The Nuclear Emergency Assistance and Training Center (NEAT) of JAEA provides a comprehensive range of technical support activities to an Off-Site Center in case of a nuclear emergency. Specifically, NEAT gives technical advice and information, provides for the dispatch of specialist as required, supplies emergency equipments and materials to the Joint Council of Nuclear Disaster Countermeasures, which meets at the Off-Site Center. NEAT provide various lectures and training course concerning nuclear disaster prevention for those personnel taking an active part in emergency response organizations at normal time. And NEAT researches on nuclear disaster prevention and also cooperate with international organizations. This annual report summarized the activities of JAEA/NEAT in the fiscal year 2008.

JAEA Reports

Current status and future plan for thermaI striping investigations at JNC

; kasahara, Naoto; ; ; Kamide, Hideki

JNC TN9400 2000-010, 168 Pages, 2000/02

JNC-TN9400-2000-010.pdf:8.78MB

Thermal striping is significant issue of the structural integrity, where the hot and cold fluids give high cycle fatigue to the structure through the thermal stress resulted from the time change of temperatur distibution in the structure. In the sodium cooled fast reactor, temperature change in fluid easily transfers to the structure because of the high thermal conductivity of the sodium. It means that we have to take care of thermal striping, The thermal striping is complex phenomena between the fluid and structure engineering fields. The investigations of thermal striping are not enough to evaluate the integrity directly. That is the fluctuation intensity at the structure surface is assumed to be temperature difference between source fluids (upstream to the mixing region) as the maximum value in the design. 0therwise, the design conditions are defined by using a mockup experiment and large margin of temperature fluctuation intensity. Furthermore, such evaluation manners have not yet been considered as a design rule. Transfer mechanism of temperature fluctuation from fluid to structure has been investigated by the authors on the view points of the fluid and structure. Attenuation of temperature fluctuation was recognized as a dominant factor of thermal fatigue. We have devdoped a numerical analysis system which can evaluate thermal fatigue and crack growth with consideration of the attenuation of temperature fluctuation in fluid, heat transfer, and structure. This system was applied to a real reactor and the applicability was confirmed. Further verification is planned to generalize the system. For the higher cost performance of the fast reactor, an evaluation rule is needed, which can estimate thermal loading with attenuation and can be applied to the design. An idea of the rule is proposed here. Two methods should be prepared; one is a precise evaluation method where mechanism of attenuation is modeled, and the other is simple evaluation method where ...

Journal Articles

Proposal of a Strain Concentration Model of Welded Joints for Creep-Fatigue Evaluation of Welded Structures

Kasahara, Naoto;

JSME International Journal, Series A, 40(3), p.247 - 254, 1997/00

None

JAEA Reports

Thermal fatigue failure test of Mod.9Cr-1Mo piping specimen containing circumferential weldments, 2; Elastic analysis and damage estimation

; Wakai, Takashi; ; ; ;

PNC TN9410 93-220, 112 Pages, 1993/09

PNC-TN9410-93-220.pdf:3.26MB

This report describes elastic thermal stress analysis and creep-fatigue damage estimation results of Mod.9Cr-1Mo piping specimen which consists of three different thickness portions (20, 15, 10mm), and contains six circumferential weldments. Thermal fatigue failure test on the specimen has been conducted in the sodium test loop named SPTT to clarify fatigue crack initiation behavior under thermal bending stress conditions, which is caused by temperature gradient arisen in the wall thickness of the specimen. The thermal transient test is now underway, with the condition where sodium of 550$$^{circ}$$C and 300$$^{circ}$$C sodium alternately flow into the specimen for 5 minutes in one cycle. The test is scheduled to complete after loading 9,000 cycles of thermal transients. Crack inspection by using penetration test tequnique is planned after the test. For an analytical study, heat transfer analysis using measured temperature data and elastic thermal stress analysis were carried out with the finite element method, and the analysis results were utilized to develop a candidate creep-fatigue damage evaluation method for Mod.9Cr-1Mo steels based on elastic analysis. Some analysis and damage estimation results of the specimen are presented here, and crack initiation after 9,000 cycles of thermal transients are predicted based on the calculated creep-fatigue damage.

JAEA Reports

Thermal creep-fatigue failure test of SUS316FR piping specimen containing circumferential weldment; (2) Elastic analysis and damage evaluation

; Wakai, Takashi; ; ; ;

PNC TN9410 93-209, 115 Pages, 1993/09

PNC-TN9410-93-209.pdf:4.24MB

This report describes elastic thermal stress analysis and creep-fatiguedamage evaluation results of an SUS316FR piping specimen which contains circumferential weldment in the middle portion of it. Thermal creep-fatigue failure test on the specimen was conducted in a sodiumtest loop named STST to clarify creep-fatigue crack initiation behavior of SUS316FR base metal and weldment under thermal bending stress conditions, which is caused by temperature gradient arisen in the wall thickness or the specimen. Thermal transient test was conducted under the condition that 550$$^{circ}$$C and 300$$^{circ}$$C sodium alternately flow into the specimen for 5 hours and 1 hour, respectively in one cycle. The test had been completed after loading 1,600 cycles or thermal transients. After the test, crack inspection by PT was performed, and cracks were observed successfully in both the base metal and weldment. For an analytical study, heat transfer analysis using measured temperature data and elastic thermal stress analysis were carried out with the finite element method, and the analysis results were utilized to develop a candidate creep-fatigue damage evaluation method for SUS316FR steel based on elasticanalysis. Some analysis results and damage evaluation results of the specimen arepresented here, and crack initiation after 1,600 cycles of thermal transients is predicted based on the calculated creep-fatigue damage, which demonstrates a good agreements with the crack distribution on the specimen inner surface.

JAEA Reports

Thermal transient strength test of a welded vessel model, Vol.6; Thermal transient test

Umeda, Hisao;

PNC TN9410 93-156, 147 Pages, 1993/06

PNC-TN9410-93-156.pdf:4.62MB

(1)[0BJECT] The object of this test is to obtain structural failure data of a typical structural model representing FBR components subjected to thermal loadings in order to develop more rationalized structural strength evaluation methods. This report provides the experimental results. (2)[TEST MODEL] The Welded Vessel Model has noteworthy typical shapes and stress distributions as can be found in the structural design of Large Type Fast Breeder Reactor, and the modified austenitic stainless steel, SUS316FR, which is hopeful for LFBR is incorporated. (3)[THERMAL TRANSIENT TEST CONDITION] Thermal creep-fatigue test was conducted with the Thermal Transient Test Facility for Structures (TTS). The test model was subjected to cyclic thermal transient of 250 $$^{circ}$$C - 600 $$^{circ}$$C by sodium. The cycle time of one thermal transient was 3 hrs, in which 250$$^{circ}$$C sodium flowed into the model for 1 hr and 600 $$^{circ}$$C sodium for 2 hrs. The thermal transient was as severe as the by temperature change rate of 40 $$^{circ}$$C/sec. (4)[TEST RESULTS] The Thermal Transient Test was completed at 1055 cycle due to crack observation on the upper Y-junction. (5)[TEMPERATURE TRANSIENT] The temperature data was obtained which was necessitated for the subsequest thermal stress analysis. The observed temperature transients were steeper than analyses. (6)[CRACK OBSERVASION] Observation method with use of "STRAW SCOPE" during the test was effective for identification of crack initiation.

Journal Articles

Creep-fatigue failure test and analysis of a vessel-type structure subjected to cyclic thermal transients

Umeda, Hisao; Tanaka, Nobuyuki; Watashi, Katsumi; Kikuchi, Masayuki; Iwata, Koji

Nuclear Engineering and Design, 140, p.349 - 372, 1993/06

None

JAEA Reports

Thermal transient strength test of a welded vessel model; No.5 Fatigue and creep test data of the model materials

; ;

PNC TN9410 92-308, 159 Pages, 1992/08

PNC-TN9410-92-308.pdf:4.36MB

The fatigue and creep tests were carried out for the Welded Vessel Model constructing materials to investigate the fatigue and creep strength of them. Test pieces were made of SUS304 and FBR grade SUS316, and these were cut out from model materials in advance. The fatigue and creep characteristics of these materials show the average properties of the material strength standard. For detailed stress analyses or creep-fatigue evaluations which are carried out after thermal transient strength test, the fatigue and creep test data which were obtained from these material tests are summarized in this report.

JAEA Reports

Creep-fatigue failure test of nozzle models under thermal transient loadings; (3)Inelastic analysis and damage evaluation

; ; ; ;

PNC TN9410 92-284, 229 Pages, 1992/05

PNC-TN9410-92-284.pdf:7.08MB

This report describes the inelastic thermal stress analysis and creep-fatigue damage evaluation results of the three kinds of nozzle-like configuration models made of SUS304. Crack initiation test on the models has been conducted in Thermal Shock Test Sodium Loop to clarify crack initiation behavior in thermal bending stress conditions. The models have been loaded with two types of thermal bending stresses; one is caused by throughtout temperature distribution arisen in the nozzle-like configuration and the other is caused by temperature gradient arisen in the wall thickness of the models. Thermal transient test has been conducted under the condition that 550$$^{circ}$$C and 300$$^{circ}$$C sodium flow into the models for 5 hrs and 1 hr, respectively in one cycle, and finished after loading 1,700 cycles of thermal transient. After the test, crack inspection test by PT has been performed, and crack initiation at cross-sections interested have been observed successfully. For the analytical study, heat transfer analysis using the measured temperature data and inelastic thermal stress analysis were carried out by using cyclic stress-strain curves obtained from two kinds of strain rate conditions as the constitutive stress-strain relationships. Analysis results were utilized mainly for two objectives. One was to present the prospect of damage evaluation method based on inelastic analysis. And the other was to advance the present creep-fatigue damage evaluation method based on elastic analysis. Many analysis results and damage evaluation results of the test models are presented here, and discussed focusing on the application of inelastic analysis results to rational creep-fatigue failure predicting method.

JAEA Reports

Thermal transient strength test of a welded vessel model; No.4 Mechanical properties of the test model materials

; ;

PNC TN9410 92-202, 56 Pages, 1992/05

PNC-TN9410-92-202.pdf:1.09MB

The fatigue and creep test were carried out for the Welded Vessel Model constructing materials to investigate the fatigue and creep strength of them. Test pieces were made of SUS304 and FBR grade SUS316, and these were cut out from model materials in advance. The observations were compared with calculations of fatigue equation, creep strain equation, creep rupture equation and dynamic stress strain equation which regulated in material strength standard. The results shows the fatigue and creep strength of the model materials have average properties which calculated with equations of the material strength standard. There is no big difference between observation and calculation value of dynamic stress strain relation and creep rate. Hence, the equations regulated in material strength standard can be applied to elasto-plastic stress analysis and creep-fatigue strength evaluation of the model.

JAEA Reports

Thermal transient strength test of a filletted vessel thermal bending model, 1; Design and fabrication of the model

; Umeda, Hisao; ; ;

PNC TN9410 92-116, 174 Pages, 1992/01

PNC-TN9410-92-116.pdf:5.89MB

The design and fabrication of the Filletted vessel Thermal Bending Model which is to be tested at TTS are described in this report. The objective of this model is to obtain the thermal transient strength of biaxial stress condition, which appears at Y-piece structures or the portions in the vicinity of sodium surface, and fillet welding which are noteworthy portions from the structural integrity viewpoint of Fast Breeder Reactor (FBR) components. As the testing portion, this model has three kinds of skirt structure, a thick walled cylinder and four stabilizer plates which are attatched on inner shell by fillet welding. The objective of the skirt structures and the thick walled cylinder is to investigate the relations between stress ratio, hoop and axial stress, and thermal transient strength, and they are designed in order to have same maximum principal stress range and large variation in the ratio of hoop to axial stress. The model has a straw-bag-like body which constituted with filleted cylinders, a thick walled cylinder and upper/lower trisphere dished plate. The model is supported with support cone at lower trisphere dished plate. The model has inner shell which strive for sodium flow stability.

JAEA Reports

Creep fatigue test of thermal stress mitigation structure model (2) under thermal transient loadings; Vol.2 Thermal transient test and crack observasion

Umeda, Hisao; ; ;

PNC TN9410 91-253, 221 Pages, 1991/01

PNC-TN9410-91-253.pdf:11.84MB

The objective of this test is to obtain structural failure data of a typical structural model represening FBR components subjected to thermal loadings, and thereby to develope structural strength evaluation methods. This report provides the test results and crack observasion on the Stress Mitigation Structure Model(2). Thermal creep-fatigue test was conducted with the Thermal transient Test Facility for Structures(TTS). The test was successfuly performed, and we observed creep-fatigue cracks at all expected portions. The flow straightner which is a tested portion for confirmation of function appeared to be safety. Cracks were found on the surfaces of perforated plate and also inside of the tube-to-perforated plate weldments. There was no corrosion althogh sodium adhered to the joining face. The Performance of a thermal insulator could not be confirmed because sodium flowed inside it. The ultrasonic examination before inspection was effective for fine cracks even though it was impossible to catch exact depth. The temperature data and creep-fatigue cracks, that can be utilized to develop the design methods of the FBR components, could be obtained.

JAEA Reports

Creep fatigue test of thermal stress mitigation structure model (2) under thermal transient loadings; Vo1.1 Design and fabrication of the model

*; *; *; Kasahara, Naoto; *; ; Imazu, Akira

PNC TN9410 89-088, 187 Pages, 1989/06

PNC-TN9410-89-088.pdf:14.67MB

Thermal transient strength tests of structure models are carried out to develop the design method of the fast breeder reactor components under thermal loadings. The fifth testing model for Thermal Transient, Test Facility for Structures (TTS), "Thermal stress mitigation structure model (2)", have been designed and fabricated. The purpose of this model is to get the thermal transient strength data for the typical shape of FBR components and to confirm the function of specific structures under thermal loading. This testing model is a vertical type cylindrical vessel supported by a skirt. It has seven testing portion for failure test, such as two types of nozzle, a Y-junction, two types of skirt and a plate to shell junction. And it has three testing portion for confirmation of function, such as a thermal insulator, a flow straightener and tube to perforated plate weldments. In designing the model, thermo-hydraulic analysis, heat transfer analysis, thermal Stress analysis were performed. Testing portions were evaluated using the design guide for TTS exclusive use. Material and welding method are basically comparable to the prototype reactor internals.

Journal Articles

None

Kasahara, Naoto;

ASME PVP-Vol.313-2, , 

None

Journal Articles

None

Tanaka, Nobuyuki; Watashi, Katsumi; Umeda, Hisao; Kikuchi, Masayuki; Iwata, Koji

Int Symp on Structral Mechanics in Reactor Technology, , 

None

24 (Records 1-20 displayed on this page)