A Rational identification of creep design area using negligible creep curve

Sukekawa, Masayuki*; Isobe, Nobuhiro*; Shibamoto, Hiroshi; Tanaka, Yoshihiko*; Kasahara, Naoto

Proceedings of 2006 ASME Pressure Vessels and Piping Division Conference (PVP 2006)/International Council on Pressure Vessel Technology (ICPVT-11) (CD-ROM), 5 Pages, 2006/00

For expansion of non-creep design area and simplification of design procedures, a rational identification method of creep design area by negligible creep (NC) curves was studied. NC curves of six kinds of stainless and ferrite steels for fast reactors were determined at 1.5Sm (Sm: design stress intensity). These NC curves are based on domestic material data. NC curves provide the relation between temperature and time that does not induce meaningful creep strain under the constant primary stress. As for 316FR steel, which is used for reactor vessel in Japanese fast reactor, non-creep design area is identified with comparing the highest temperature and 425C (constant upper limit for austenite stainless steal) by existing Japanese Guides. However, this temperature limit can be enhanced by NC curve concept when operating (thermal transient) time is long. NC curves under higher primary stress, and the curves under secondary stress were also studied. However, at the present stage, NC curves for stress level 1.5Sm were adopted to identify creep design area. The concept of NC curve was introduced into the interim FDS (fast reactor design standard for commercialized fast reactors in Japan) to simplify the creep design of fast reactor systems. Utilizing these curves, design becomes easier for components which are employed at comparatively lower temperature under normal condition and short holding time at high temperature.

Research and development issues for fast reactor structural design standard (FDS)

kasahara, Naoto; Ando, Masanori; Ito, Kei; Tanaka, Yoshihiko; Shibamoto, Hiroshi; Inoue, Kazuhiko

ASME PVP-Vol.472, p.25-32, p.25 - 32, 2004/07

For the realization of safe and economical fast reactor (FR) plants, the Japan Nuclear Cycle Development Institute (JNC) and Japan Atomic Power Company (JAPC) are cooperating on a research project titled "Feasibility Study on Commercialized FR Cycle Systems. To certify the design concepts and validate their structural integrity,the research and development of the "Fast Reactor Structural Design Standard (FDS)" is recognized as begin an essential theme.

Feasibility Study on Commercialization of Fast Breeder Reactor Cycle Systems Interim Report of Phase II; Technical Study Report for Reactor Plant Systems

Konomura, Mamoru; Ogawa, Takashi; Okano, Yasushi; Yamaguchi, Hiroyuki; Murakami, Tsutomu; Takaki, Naoyuki; Nishiguchi, Youhei; Sugino, Kazuteru; Naganuma, Masayuki; Hishida, Masahiko; et al.

JNC TN9400 2004-035, 2071 Pages, 2004/06

JNC-TN9400-2004-035.pdf:76.42MB

The attractive concepts for Sodium-, lead-bismuth-, helium- and water-cooled FBRs have been created through using typical plant features and employing advanced technologies. Efforts on evaluating technological prospects of feasibility have been paid for these concepts. Also, it was comfirmed if these concepts satisfy design requierments of capability and performance presumed in the feasibilty study on commertialization of Fast Breeder Reactor Systems. As results, it was concluded that the selection of sodium-cooled reactor was most rational for practical use of FBR technologies in 2015.

Study on inelastic analysis method for structural design,1; Estimation method of loadig histry effect

Tanaka, Yoshihiko; Kasahara, Naoto

JNC TN9400 2003-037, 95 Pages, 2003/05

JNC-TN9400-2003-037.pdf:4.11MB

The advanced loop-type reactor system, one of the promising concepts in the Feasibility study of the FBR Cycle, adopts many innovative ideas to meet the challenging requirements for safety and economy. As a results, it seems that the structures of the reactor system would be subjected to severer loads than the predecessors. 0ne of the countermeasures to them is the design by inelastic analysis. In the past, many studies showed that structural design by inelastic analysis is much more reasonable than one by conservative elastic analysis. However, inelastic analysis has hardly been adopted in nuclear design so far. One of the reasons is that inelastic analysis has loading history effect, that is, the analysis result would differ depending on the order of loads. It seems to be difficult to find the general solution for the loading history effect. Consequently, inelastic analysis output from the four deferent thermal load histories which consists of the thermal load cycle including the severest cold shock ("C")and the one including the severest hot shock ("H") were compared with each other. From this comparison, it was revealed that the thermal load history with evenly distributed "H"s among "C" s tend to give the most conservative damage estimation derived from inelastic analysis output. Therefore, such thermal load history pattern is proposed for the structural design by inelastic analysis.

R&D issues in Structural Design Standard for commercialized Fast Rreactor Components

Shibamoto, Hiroshi; Tanaka, Yoshihiko; kasahara, Naoto; Ito, Kei; Inoue, Kazuhiko

GENES4/ANP2003, 120 Pages, 2003/00

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Extension of applicability of green function method to thermal transient stress analysis (2); Responsive stress to two thermal fluids of varying flow-rate

Tanaka, Yoshihiko; kasahara, Naoto

JNC TN9400 2002-038, 95 Pages, 2002/06

JNC-TN9400-2002-038.pdf:2.56MB

PARTS, Program for Arbitrary Real Time Simulation is being developed: it is cxpected to make great contribution to fast reactor components' design by enabling integration of thermal hydraulic and structural analysis. At this moment, the Green function method is mainly used as a stress analysis method for PARTS. The Green function is a description of the relationship between input and response of a system. Strictly, the response depends only on the natures of the input and the system. However if a function precisely simulating their natures is established based on the response to elemental inputs (pulse wave, step wave, etc.,), it becomes possible to find approximate responses to random and/or complicated inputs. This procedure is called "Green function method". This method is applicable to structural design of the fast reactors. Green function method finds thermal transient stress arising in structures in the form of convolute integration corresponding to coolant fluids' step-changes of temperature. It is expected to calculate faster than Finite Elemental Method (FEM) that solves innumerable balance equations of stress and strain at every time step. In order to apply the Green function method to actual plant design in near future, it is necessary to prove that the method gives appropriate results even under the conditions assumed in plant design works. The authors have successfully developed the Green function method which had been applicable only to a cylinder contacting with sole fluid under constant thermal transfer rate into the one being applicable to a cylinder with primary and secondary fluids under step-changing thermal transfer rates. In this report, applicability of Grecn function method to structural design of a components of complicated shape cxposed to thermal transients of the two independent coolant systems under changing heat transfer rate. As an example, the internal components of the intermediate heat exchanger (IHX) of the advanced loop ...

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; Aoto, Kazumi; kasahara, Naoto; ; Sagayama, Yutaka*; *; *

JNC TY9400 2001-026, 978 Pages, 2002/01

JNC-TY9400-2001-026.pdf:28.0MB

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Heavy liquid Metal cooled fast breeder reactor; Results in 1999

; ; ; ;

JNC TN9400 2000-079, 189 Pages, 2000/07

JNC-TN9400-2000-079.pdf:5.94MB

Based on the medium and long-term program of JNC, the feasibility study for fast breeder reactors (FBRs) including related nuclear fuel cycles has been started from the 1999 fiscal year. Various options of FBR plant systems have been studied and a concept of Heavy Liquid Metal cooled FBRs is one of these options. The purpose of this paper is to research and evaluate Heavy Liquid Metal cooled FBRs on the basis of literatures. First, we selected four types of plant concepts listed below. Concept 1: Large-scale pond type reactor with Pb cooled. Concept 2: Large-scale loop type reactor with Pb cooled. Concept 3: Medium-scale module tank type reactor with Pb cooled. Concept 4: Small scale module tank type reactor with Pb-Bi cooled. Concept l and 2 are selected to seek for scale merit on economical aspect. ln Concept 3 and 4, we tried to reduce the inventory of HLMC and to ease the load conditions on structures and seek for competitiveness with module effect such as mass production and learning effect. Through a preliminary design study, we identified some technical features of each concept and roughly evaluated economical competitiveness based on total weight of the NSSSs. From this study, we concluded (1)lngeneral, the large-scale type concepts have little economical advantage because of its huge amount of material needed for its severe load conditions. (Concept 1&2) (2)Even for the large-scale pond type reactor, the conclusion seems to be the same. Total amount of the thermal shielding material became huge. Aseismatic structure makes the amount of material increase under the Japanese seismic condition. (Concept1) (3)For the large-scale loop type reactor, we selected side entry and dual walled piping concept with slide-joint inner wall to cope with thermal expansion of piping system. However, there seemed to be difficulty with compatibility between slide-joint and oxide film corrosion prevention measures. (Concept2) (4)The medium and small modular type ...

A Conceptual Design Study on Various Type of HLMC Fast Reactor Plant

; Tanaka, Yoshihiko;

IAEA-AGM on Design and Perfomanceof Reactor and Su, 0 Pages, 2000/00

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