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Journal Articles

10.2.1 Global trends in improvement of light water reactor

Hidaka, Akihide

Genshiryoku No Ima To Ashita, p.264 - 265, 2019/03

no abstracts in English

Journal Articles

Investigation on the plasticity correction of stress intensity factor calculations for underclad cracks in reactor vessels

Lu, K.; Katsuyama, Jinya; Li, Y.

Nippon Kikai Gakkai M&M 2016 Zairyo Rikigaku Kanfarensu Koen Rombunshu (Internet), p.499 - 501, 2016/10

When conducting structural integrity assessments for reactor pressure vessels (RPVs) subjected to pressurized thermal shock (PTS) events, the stress intensity factor (SIF) is evaluated for a postulated surface crack in the inner surface of RPVs. It is known that the cladding made of a stainless steel is a ductile material which is overlay-welded on the inner surface, therefore, the plasticity of cladding should be considered in SIF calculations for a postulated underclad crack to ensure a conservation evaluation. Recently, the authors performed three-dimensional (3D) elastic and elastic-plastic FEAs for Japanese three-loop RPVs and proposed a rational evaluation method on SIFs of underclad cracks. In this paper, further studies were conducted to discuss the applicability of the proposed plasticity correction method. The effect of neutron irradiation was considered. In addition, different Japanese RPV geometries such as two-loop and four-loop RPVs were also investigated.

JAEA Reports

Inspection and repair techniques in the reactor vessel of the experimental fast reactor Joyo; Observation technical development in a reactor vessel of the fast reactor, 3

Okuda, Eiji; Sasaki, Jun; Suzuki, Nobuhiro; Takamatsu, Misao; Nagai, Akinori

JAEA-Technology 2016-017, 20 Pages, 2016/07

JAEA-Technology-2016-017.pdf:5.75MB

In-Vessel Observation (IVO) techniques for Sodium Cooled Fast Reactors (SFRs) in service are important for confirming their safety and integrity. Since IVO equipment for an SFR has to be designed to tolerate the severe conditions (high temperature, high radiation dose and limited access route), fiberscopes used to be used in previous IVO for SFRs. However, in order to attain an IVO with higher quality and resolution, IVO using a radiation resistant camera was conducted in the fast experimental reactor Joyo and obtained some results. The demonstration results provided valuable insights for use in further improving and verifying IVO techniques in SFRs.

JAEA Reports

Inspection and repair techniques in the reactor vessel of the experimental fast reactor Joyo; Development of cover gas recycling system with precise pressure control

Ushiki, Hiroshi*; Okuda, Eiji; Suzuki, Nobuhiro; Takamatsu, Misao; Nagai, Akinori

JAEA-Technology 2015-042, 37 Pages, 2016/02

JAEA-Technology-2015-042.pdf:16.51MB

The reactor vessel of a sodium-cooled fast reactor (SFR) is filled with sodium coolant and cover gas (argon gas). In case of a cover gas boundary open (ie., in-vessel repair), installation of a temporary cover gas boundary and controlling the cover gas pressure slightly positive are required to prevent the cover gas release and the contamination of impurities, and during upper core structure (UCS) replacement in the experimental SFR Joyo from March to December 2014, a vinyl bag was installed as a part of the temporary cover gas boundary. However, because it has inferior thermal resistance, supply a cooling gas too much was required to maintain proper temperature for two months. On the basis of this requirement, a cover gas recycling system with precise pressure control was developed and adopted for UCS replacement. The system has a good pressure controllability and recyclability. The successful results of this system contributed to the certain promotion of UCS replacement. In addition, the insights and the experience gathered in this development are expected to improve the in-vessel repair techniques in sodium-cooled fast reactors.

JAEA Reports

Inspection and repair techniques in the reactor vessel of the experimental fast reactor Joyo; Observation techniques development in a reactor vessel of the fast reactor, 2

Okuda, Eiji; Sasaki, Jun; Suzuki, Nobuhiro; Takamatsu, Misao; Nagai, Akinori

JAEA-Technology 2015-005, 36 Pages, 2015/03

JAEA-Technology-2015-005.pdf:44.42MB

In-Vessel Observations (IVO) techniques for Sodium cooled Fast Reactors (SFRs) are important in confirming its safety and integrity. In order to secure the reliability of IVO techniques, it was necessary to demonstrate the performance under the actual reactor environment with high temperature, high radiation dose and remained sodium. The IVO equipment for the Upper Core Structure (UCS) fitting area was specifically developed in the experimental fast reactor "Joyo". And the IVO was successfully completed as shown below. (1) Improvement of picture quality and resolution. The IVO of UCS fitting area with the gap of 5mm in minimum was achieved using the IVO equipment with video-scope under the actual reactor environment. The picture quality and resolution could be improved comparing with the radiation resistant fiberscope which was used in past IVO. (2) Prevention of video-scope hypofunction by high temperature / radiation dose. Since video-scope is inferior in thermal and radiation resistance, the IVO equipment was designed to be able to withdraw and insert video-scopes with cooling gas. This measure could achieve the observation in short radiation time with available temperature under the actual reactor environment. The IVO equipment for UCS fitting area provided useful information on UCS replacement. In addition, the experience provided valuable insights into further improvements for IVO techniques in SFRs.

JAEA Reports

Review of incident involving reactor pressure vessel head degradation at U.S. PWR

Watanabe, Norio

JAERI-Review 2004-015, 144 Pages, 2004/07

JAERI-Review-2004-015.pdf:25.58MB

In March 6, 2002, significant degradation to the reactor vessel head material adjacent to a control rod drive nozzle was discovered at a U. S. pressurized water reactor (PWR), Davis Besse. Based on the documents issued by the U.S. Nuclear Regulatory Commission (USNRC) and the licensee, this report provides a brief description on circumstances related to discovery of the reactor vessel head degradation, and describes the degraded condition and causes, the USNRC's responses to this event and so on. In attachments, given are the contents of the generic communications (Bulletins, Generic Letters and Information Notices) issued by the USNRC for the events involving boric acid corrosion and degradation of reactor coolant pressure boundary components and the summary of the investigation reports issued by the USNRC's lessons-learned task force.

Journal Articles

Evaluation of neutron irradiation embrittlement in structural materials for reactor pressure vessel

Ooka, Norikazu*; Ishii, Toshimitsu

Hihakai Kensa, 52(5), p.235 - 239, 2003/05

no abstracts in English

JAEA Reports

Irradiation tests report of the 35th cycle in "JOYO"

*

JNC-TN9440 2000-008, 79 Pages, 2000/08

JNC-TN9440-2000-008.pdf:2.33MB

This report summarizes the operating and irradiatlon data of the experimental reactor "JOYO" 35th cycle. Irradiation tests in the 35th cycle are as follows: (1)C-type irradiation rig (C4F) (a)High burnup performance test of advanced austenitic stainless steel cladding fuel pins (in collaboration with France) (2)C-type irradiation rig (C6D) (a)Large diameter fuel pins irradiation tests (3)Core Materials Irradiation Rig (CMIR-5) (a)Cladding tube materials irradiation tests for "MONJU" (4)Structure Materials Irradiation Rigs (SMIR) (a)Decision of material design base standard of structure materials for prototype reactor and large scale reactor (5)Upper core structure irradiation Plug Rig (UPR-1-5) (a)Upper core neutron spectrum effect and accelerated irradiation effect (6)SurVeillance un-instrument Irradiation Rig (SVIR) (a)Confimation of surveillance irradiation condition for "JOYO" (b)Material irradiation tests (based on a contract with universities) The maximum burnup driver assembly "PFD253" reached 67,600 MWd/t (pin average).

JAEA Reports

Study on sodium coolant loop-type reactor; Parametric study on maximum thermal stress depending on routing dimension of piping system

Tsukimori, Kazuyuki; Furuhashi, Ichiro*

JNC-TN9400 2000-049, 93 Pages, 2000/03

JNC-TN9400-2000-049.pdf:2.82MB

lt is one of the important key points to reduce thermal stress of the primary piping system in the design of sodium coolant loop-type FBR plants. The objectives of this study are to understand the characteristics of the thermal stresses in the simple S-shaped hot leg piping systems which run from the outlet nozzle of the reactor vessel (R/V) to the inlet nozzle of the intermediate heat exchanger (IHX), and to propose some recommendable routings of piping systems. Results are summarized as follows. (1)Generally, the thermal stresses in elbows are severer than those at nozzles. The tendency was observed that the stress in elbow decreases with the increase of the distance between the outlet nozzle of R/V and the inlet nozzle of IHX and also the distance between the outlet nozzle of R/V and the liquid surface level. (2)lt is expected to reduce thermal stresses in elbow to big extent by adopting super 90 degree elbows. Therefore, in these cases the dimension region which satisfies the allowable stress is broad compared with that in the case of the conventional 90 degree elbow. (3)The stress estimations in elbow based on 'MITl notice No.501' become excessively large compared with the results by FEA using shell elements, when the maximum stress occurs at the end of elbow. ln these cases, the estimation can be rationalized by replacing the maximum stress by the mean of stresses at the end and at the middle of the elbow. (4)Two routings with 105 degree elbows are recommended. 0ne has the advantage from the view point of reduction of length of pipe and the other does from the view point of reduction of thermal stresses, compared with the routing with 90 degree elbows.

JAEA Reports

Microstructural assessment of damaged materials in FBR assessment of creep damage in weldment

Momma, Yoshio*; *; ; ; ; Aoto, Kazumi

JNC-TN9400 2000-044, 22 Pages, 2000/03

JNC-TN9400-2000-044.pdf:1.37MB

ln the past the microstructural observation was mostly applied to understand the materials behavior qualitatively in R&D of the new materials and the life prediction for the fast breeder reactor components. However, the correlation between the changes in properties and microstrutures must be clarified to ensure the structural integrity. Particularly we are interested in the method to correlate the long-term properties and microstructural changes at high temperatures. The current research is to quantify the changes in microstructure of the weld metal for the welded structure of the reactor vessel. ln this research we have conducted creep testing of the weld metals at 823 and 873K up to 37,000h. Two types of the weld metals (16Cr-8Ni-2Mo and 18Cr-12Ni-Mo) were subjected to the creep testing. Based on the areas of the precipitates, the microstructural characterization with time and creep damage was attempted. The creep strength of the 16Cr-8Ni-2Mo weld metal is lower than that of the 18Cr-12Ni-Mo one at higher stresses, shorter times. But there is a trend toward to become similar strength with lower stresses and increasing times. The creep-rupture ductility of the 16Cr-8Ni-2Mo weld metal is superior to that of the 18Cr-12Ni-Mo one. The creep-rupture takes place at the interface of the sigma ($$sigma$$) phases precipitated in the delta ($$delta$$) ferrites at 823K lower stresses and 873K. The amount of precipitates in the 16Cr-8Ni-2Mo weld metal is smaller than that in the 18Cr-12Ni-Mo one at each temperature and stress. Also it is apparent that the amount of the precipitates is primarily responsible to the decomposition of the $$delta$$ phase, because the amount of the residual $$delta$$ ferrites measured by the Magne-Gauge reduces with times. Using the Larson-Miller parameter it was possible to correlate the amount of the precipitates linearly with the LMP values.

JAEA Reports

JASPER Experimental data book (VII); Gap streaming Experiment

Takemura, Morio*

JNC-TJ9450 2000-002, 112 Pages, 2000/03

JNC-TJ9450-2000-002.pdf:2.55MB

This report is intended to make it easier to apply the measured data obtained from the Gap Streaming Experiment, which was conducted at the Oak Ridge National Laboratory (ORNL) during about two months beginning at the start of March, 1992 as the sixth one of a series of eight experiments planned for the Japanese-American Shielding Program for Experimental Research (JASPER) which was started in 1986. For this reason. the information presented includes specifications and measurement data for all configurations, compositions of all materials, characteristics of the measurement system. and daily-basis records of measurements. The Gap Streaming Experiment was planned to obtain the data of neutron streaming characteristics in the inclosure system above the core of an advanced fast reactor for verification and improvement of the analysis method to be applied to the shielding design. A iron-lined solid or slit concrete assembly was placed, with or without a spectrum modifier forming soft incident neutron spectrum, behind the TSR-II reactor of Tower Shielding Facility. Inserting central cylinders and cylindrical sleeves gave various gap width and offset in the slit concrete assembly. Neutron flux was measured behind the configurations with various types of detectors. The integral neutron flux in wide energy region was measured on radial traverse and on the axis behind the concrete assembly in almost all configurations. Neutron spectrum and fine radial distribution in high energy region was measured further in case of hard incident neutron spectrum, Information presented in this report is based mainly on a report issued by ORNL (ORNL/TM-12140. "Measurements for the JASPER Program Gap Streaming Experiment"). Additional information reported by the assignee is utilized also.

JAEA Reports

JASPER Experimental data book (VI); Special materials experiment

Mori, Tomoaki*; Takemura, Morio*

JNC-TJ9450 2000-001, 96 Pages, 2000/03

JNC-TJ9450-2000-001.pdf:2.04MB

This report is intended to make it easier to apply the measured data obtained from the Special Materials Experiment, which was conducted at the Oak Ridge National Laboratory (ORNL) during about a month beginning at the end of June, 1992 as the last one of a series of eight experiments planned for the Japanese-American Shielding Program for Experimental Research (JASPER) which was started in 1986. For this reason. the information presented includes specifications and measurement data for all configurations, compositions of all materials, characteristics of the measurement system. and daily-basis records of measurements. The Special Materials Experiment was planned to obtain the data of neutron attenuation characteristics of selected shielding materials for use in advanced fast reactors. The material of particular interest for the experiment was zirconium hydride that is rich in hydrogen. The mockup slabs for the special materials were preceded by the spectrum modifier behind the TSR-II reactor of Tower Shielding Facility. The layer of zirconium hydride was simulated with a combination of zirconium and polyethylene slabs. The thick layer of polyethylene with no zirconium was installed in some configurations.Neutron flux was measured behind the configurations with various types of detectors. The integral neutron flux in wide energy region was measured in eight configurations and neutron spectrum in high energy region was measured also in almost all configurations. Information presented in this report is based mainly on a report issued by ORNL (ORNL/TM-12277. "Measurements for the JASPER Program Special Materials Experiment"). Additional information reported by the assignee is utilized also.

JAEA Reports

Revise of a basic data base for shielding design

*; Takemura, Morio*

JNC-TJ9440 2000-005, 157 Pages, 2000/03

JNC-TJ9440-2000-005.pdf:3.7MB

With use of the two-dimensional discrete ordinates code DORT and the standard groupwise shielding design library JSSTDL produced from the latest evaluated nuclear data library JENDL-3.2, experimental analyses for the representative configurations in the Radial Shield Attenuation Experiment of the JASPER were performed. The results were compared with those obtained with use of traditional method DOT3.5/JSDJ2 for the previous JASPER experimetal analyses. In general, the change of the cross section library gives higher results and the change of the transport code gives lower results. Finally the new analysis method gives better agreement with the experimental results and also less deviations of calculational errors between various detectors. Experimental analyses for the thick concrete configulation in the Gap Streaming Experiment of the JASPER was also performed with the new analysis method, after solving the poor agreement found in last year with the original JASPER experimental analyses. The same tendency due to the library change was confirmed with the above mentioned analyses of the Radial Shield Attenuation Experiment. Compilation of the input data necessary for future reanalyses of important configurations in JASPER experiments were continued through the above-mentioned experimental analyses and related informations were added for repletion of the database preserved in a computer disk holding previously accumulated data. Input data descriptions were made for auxiliary routines needed for the experimental analyses and their sample data were compiled and stored in the database.

JAEA Reports

A Design study of high electric power for fast reactor cooled by super critical light water

Koshizuka, Seiichi*

JNC-TJ9400 2000-011, 102 Pages, 2000/03

JNC-TJ9400-2000-011.pdf:2.71MB

In order to evaluate the possibility to achieve high electric power by a fast reactor with supercritical light water, the design study was carried out on a large fast reactor core with high coolant outlet temperature (SCFR-H). Since the reactor coolant circuit uses once-through direct cycle where all feedwater flows through the core to the turbine at supercritical pressure, it is possible to design much simpler and more compact reactor systems and to achieve higher thermal efficiency than those of current light water reactors. The once-through direct cycle system is employed in current fossil-fired power plants. In the present study, three types of core were designed. The first is SCFR-H with blankets cooled by ascending flow, the second is SCFR-H with blankets cooled by descending flow and the third is SCFR-H with high thermal power. Every core was designed to achieve the thermal efficiency over 43%, positive coolant density reactivity coefficient and electric power over 1600MW. Core characteristics of SCFR-Hs were compared with those of SCLWR-H (electric power: 1212MW), which is a thermal neutron spectrum reactor cooled and moderated by supercritical light water, with the same diameter of the reactor pressure vessel. It was shown that SCFR-H could increase the electric power about l.7 times maximally. From the standpoint of the increase of a reactor thermal power, a fast reactor has advantages as compared with a thermal neutron reactor, because it can increase the power density by adopting tight fuel lattices and eliminating the moderator region. Thus, it was concluded that a reactor cooled by supercritical light water could further improve the cost competitiveness by using a fast neutron spectrum and achieving a higher thermal power.

JAEA Reports

None

*; *; *

JNC-TJ8420 2000-003, 99 Pages, 2000/03

JNC-TJ8420-2000-003.pdf:5.47MB

no abstracts in English

JAEA Reports

Post processing system for multi-dimensionaI thermal-hydraulic analyses

Miyake, Yasuhiro*; *; ; Kimura, Nobuyuki

JNC-TN9400 2000-016, 40 Pages, 1999/12

JNC-TN9400-2000-016.pdf:3.71MB

ln the conventional visualization system for the computational results, only Japanese (Nihongo) Line Printer (NLP) was available to print two dimensional cross sectional plots of vector and scalar fields. To evaluate the phenomena, an analyst had to print many plots on the NLP. This task makes difficult to check the computational results immediately after the calculation. Recently, as the visualization tools, we introduced Micro AVS and Field View which are utilized widely in the scientific and the industrial fields. ln order to show the numerical results on the visualization software, we constructed a post processing system which convert the results of the numerical code to "lntermediate files" which can be read by the visualization tools. As using this system, the examination of the numerical results can be executed on the display of the personal computer. Furthermore, the persuasive report and paper with high quality can be produced due to the color printing. As for the transient calculation, the change of the phenomena can be visually evaluated by using the animation function.

JAEA Reports

None

*; *; *

JNC-TY9400 2000-003, 99 Pages, 1999/08

JNC-TY9400-2000-003.pdf:6.48MB

no abstracts in English

Journal Articles

Fracture toughness evaluation by precracked Charpy specimens in the transition temperature range of RPV steels

Onizawa, Kunio; Tobita, Toru; Suzuki, Masahide

Proceedings of 15th International Conference on Structural Mechanics in Reactor Technology (SMiRT-15), 4, p.137 - 144, 1999/00

no abstracts in English

JAEA Reports

Research on the usage of a deep sea fast reactor

;

PNC-TN9420 97-006, 70 Pages, 1997/09

PNC-TN9420-97-006.pdf:2.49MB

Many new types of fast reactors have been studied in PNC. A deep sea fast reactor has the highest realization probability of the reactors studied because its development is desired by many specialists of oceanography, meteorology, deep sea bottom oil field, seismology and so on and because the development does not cost big budget and few technical problems remain to be solved. This report explains the outline and the usage of the reactor of 40 kWe and 200 to 400 kWe. The reactor can be used as a power source at an unmanned base for long term climate prediction and the earth science and an oil production base in a deep sea region. On the other hand, it is used for heat and electric power supply to a lavoratory in the polar region. In future, it will be used in the space. At the present time, a large FBR development plan does not proceed successfully and a realization goal time of FBR has gone later and later. We think that it is the most important to develop the reactor as fast as possible and to plant a fast reactor technique in our present society.

JAEA Reports

Development of seismic isolation system in vertical direction; Coupled response analysis of reactor building and common deck isolation system

PNC-TN9410 97-032, 126 Pages, 1997/04

PNC-TN9410-97-032.pdf:4.62MB

Study was made on the effect of dynamic interaction between reactor building and common deck isolation structure. Dynamic response analyses were made with a coupled and a decoupled model of the reactor building and the isolation structure. The effect of the dynamic interaction was evaluated by comparing these analysis results. The results of the study can be summarized as below; (1)Dynamic coupling effect between reactor building and the common deck isolation structure is not significant. This can be attributed to the fact that the isolation frequency is sufficiently lower than the fundamental frequency of the soil-structure system. (2)The decoupled analysis model was found to give response results with practically sufficient accuracy and slight conservatism. (3)The effect of the response of common deck isolation structure on the response of the reactor building is also negligible.

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