This series of papers has been issued to give general views on recent Japanese activities related to nuclear codes and standards. Part I of the series includes an overview, a discussion of administrative and design aspects related to new methods of evaluating the integrity of components, that is, piping seismic rules based on ratcheting fatigue, a fracture toughness equation for vessel materials, an extension of the Ke factor based on elastic follow-up, a discussion of high cycle thermal fatigue, and a three-dimensional finite element method elastic-plastic analysis procedure.

Issue Section:
Research Papers
Keywords:
nuclear facility regulation, nuclear power stations, fatigue, fracture toughness, elastoplasticity, thermal stress cracking, finite element analysis, pressure vessels, fission reactor design, fission reactor materials
Topics:
Cycles, Design, Engineering standards, Evaluation methods, Fatigue, Pipes, Stress, Nuclear power stations, Earthquake resistant design, Fracture toughness, Japan Society of Mechanical Engineers, Earthquakes
1.
Japan Electric Association
, 1984, “
Technical Guidelines for Aseismic Design of Nuclear Power Plants: Supplement Allowable Stress, Classification
,” JEAG 4601: Supplement.
2.
Japan Electric Association
, 1987, “
Technical Guidelines for Aseismic Design of Nuclear Power Plants
,” JEAG 4601.
3.
Japan Electric Association
, 1991, “
Technical Guidelines for Aseismic Design of Nuclear Power Plants: Supplement
,” JEAG 4601, Supplement edition.
4.
Park
,
Y. J.
, and
Hofmayer
,
C. H.
, NUREG/CR-6241 (BNL-NUREG-52422) 1994, “
Technical Guidelines for Aseismic Design of Nuclear Power Plants
,” Translation of JEAG 4601-1987.
5.
EPRI
, 1993, “
Piping and Fitting Reliability Program, Vol. 1, Project Summary
.”
6.
EPRI
, 1992, “
Piping and Fitting Reliability Program, Vol. 2, Component Test Report
.”
7.
EPRI
, 1992, “
Piping and Fitting Reliability Program, Vol. 3, System Test Report
.”
8.
EPRI
, 1992, “
Piping and Fitting Reliability Program, Vol. 4, Specimen Fatigue Ratcheting Test Report
.”
9.
EPRI
, 1989, “
Piping and Fitting Reliability Program, Vol. 5, Piping Design Rules Revisions
.”
10.
ASME
, “
Boiler and Pressure Vessel Code, Sec. 3, Division 1
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11.
ASME
, 2001, “
Boiler and Pressure Vessel Code, Sec. 3, Division 1
.”
12.
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, 1998, “
Seismic Analysis of Piping: Final Program Report
,” NUREG/CR-5361.
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,
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, 1985, “
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,”
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, II 2.2, 1985, pp.
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,
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,
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,
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,
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,
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, and
Asada
,
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, 1993, “
Life Prediction Technique for Ratcheting Fatigue
,”
ASME PVP
,
266
, pp.
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,
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,
Kobayashi
,
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,
Isogai
,
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, and
Asada
,
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, 2000, “
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,”
ASME PVP
,
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, pp.
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.
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,
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,
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,
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,
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,
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,
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,
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,
Fujiwaka
,
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,
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,
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, and
Suzuki
,
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, 2000, “
Study on Seismic Design of Nuclear Power Plant Piping in Japan. Part 1: Overview of the Study
,”
ASME PVP
,
407
, pp.
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–
123
.
17.
Yokota
,
H.
,
Endou
,
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,
Kawabata
,
M.
,
Sakakida
,
T.
,
Fujiwaka
,
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,
Asada
,
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, and
Suzuki
,
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, 2000, “
Study on Seismic Design of Nuclear Power Plant Piping in Japan. Part 2: Material Tests Results
,”
ASME PVP
,
407
, pp.
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–
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.
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Yoshino
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,
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,
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,
Fujiwaka
,
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,
Asada
,
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, and
Suzuki
,
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, 2000, “
Study on Seismic Design of Nuclear Power Plant Piping in Japan. Part 3: Component Tests Results
,”
ASME PVP
,
407
, pp.
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–
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.
19.
Sakakida
,
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,
Endou
,
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,
Kawabata
,
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,
Yokota
,
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,
Fujiwaka
,
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,
Asada
,
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, and
Suzuki
,
K.
, 2000, “
Study on Seismic Design of Nuclear Power Plant Piping in Japan. Part 4: Analytical Evaluation of Piping Component Tests
,”
ASME PVP
,
407
, pp.
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–
146
.
20.
Suzuki
,
K.
,
Namita
,
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,
Abe
,
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,
Ichihashi
,
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,
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,
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,
Ishiwata
,
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,
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,
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, and
Yokota
,
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Seismic Proving Test of Ultimate Strength (Current Status of Preliminary Tests-II)
,”
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, ICONE-22225.
21.
Suzuki
,
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,
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,
Ishiwata
,
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,
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,
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, and
Tai
,
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, 2002, “
Seismic Proving Test of Ultimate Strength (Results on Piping Component and Simplified Piping System)
,”
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,
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,
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,
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,
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,
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, and
Tai
,
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, 2003, “
Seismic Proving Test of Ultimate Strength (Simulation Analysis of Simplified Piping System Test)
,”
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,
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, pp.
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23.
Suzuki
,
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,
Namita
,
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,
Abe
,
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,
Ichihashi
,
I.
,
Suzuki
,
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,
Sakakida
,
T.
,
Sato
,
T.
, and
Yokota
,
H.
, 2003, “
Seismic Proving Test of Ultimate Strength (Status of Design Method Confirmation Test)
,”
Proceedings of SMiRT-17
,
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-
2
.
24.
Suzuki
,
K.
, and
Suzuki
,
K.
, 2004, “
Seismic Proving Test of Ultimate Strength (Design Method Confirmation Test)
,”
Proceedings of ASME-PVP
,
486
, pp.
187
–
194
.
25.
Suzuki
,
K.
, and
Suzuki
,
K.
, 2004, “
Seismic Proving Test of Ultimate Strength (Ultimate Strength Test)
,”
Proceedings of ASME-PVP
,
486
, pp.
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–
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.
26.
Nuclear Power Engineering Corporation
, 1999, “
Report on Demonstration Test on Seismic Reliability of Nuclear Power Plants Part 3, FY 1998: Piping-System Ultimate Strength Demonstration Test
.”
27.
Nuclear Power Engineering Corporation
, 2000, “
Report on Demonstration Test on Seismic Reliability of Nuclear Power Plants Part 3, FY 1998: Piping-System Ultimate Strength Demonstration Test
.”
28.
Nuclear Power Engineering Corporation
, 2001, “
Report on Demonstration Test on Seismic Reliability of Nuclear Power Plants Part 2, FY 1998: Piping-System Ultimate Strength Demonstration Test
.”
29.
Nuclear Power Engineering Corporation
, 2002, “
Report on Demonstration Test on Seismic Reliability of Nuclear Power Plants Part 2, FY 1998: Piping-System Ultimate Strength
.”
30.
Nuclear Power Engineering Corporation
, 2003, “
Report on Demonstration Test on Seismic Reliability of Nuclear Power Plants, Part 1, FY 1998: Piping-System Ultimate Strength Demonstration Test
.”
31.
Nuclear Power Engineering Corporation
, 2003, “
Report on Demonstration Test on Seismic Reliability of Nuclear Power Plants Part 1, FY 1998: Piping-System Ultimate Strength
.”
32.
Maekawa
,
O.
,
Matsushita
,
A.
,
Iida
,
M.
,
Murakami
,
E.
,
Iida
,
K.
, and
Asada
,
Y.
, 1997, “
Establishment of a New KIR Curve Based on TNDT Obtained by the One Pass Procedure, Pressure Vessel and Piping Codes and Standards
,”
PVP
,
353
, ASME 1997, pp.
351
–
358
.
33.
Asada
,
S.
,
Nakamura
,
T.
, and
Asada
,
Y.
, 2001, “
Evaluation of Conservativeness in the Simplified Elastic-Plastic Analysis Using Analytical Results Part 2: Proposal of a New Ke Function
,” Pressure Vessel and Piping Codes and Standards,
PVP
,
419
, The American Society of Mechanical Engineers, pp.
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–
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.
34.
Tagart
,
S. W.
, Jr., 1968, “
Plastic Fatigue Analysis of Pressure Components
,” ASME 68-PVP-3.
35.
Kawamura
,
T.
,
Shiina
,
K.
,
Ohtsuka
,
M.
,
Mizuno
,
T.
,
Kurosaki
,
M.
,
Wakamatsu
,
M.
,
Tanimoto
,
K.
,
Fyukuda
,
T.
,
Minami
,
Y.
,
Moriya
,
S.
, and
Madarame
,
H.
, 2003, “
Study on High-Cycle Fatigue Evaluation for Thermal Striping in Mixing Tees With Hot and Cold Water (2)
,”
Proceedings of the 11th International Conference on Nuclear Engineering
(ICONE-11), No. 36182.
36.
Noguchi
,
H.
,
Tanimoto
,
K.
,
Kondo
,
Y.
,
Ishiga
,
H.
,
Ogura
,
K.
,
Shiina
,
K.
,
Minami
,
Y.
,
Fukuda
,
T.
,
Moriya
,
S.
, and
Madarame
,
H.
, 2003, “
Study on High-Cycle Fatigue Evaluation for Thermal Striping in Mixing Tees With Hot and Cold Water (3)
,”
Proceedings of the 11th International Conference on Nuclear Engineering
(ICONE-11), No. 36376.
37.
Kondo
,
Y.
,
Tanimoto
,
K.
,
Shiraishi
,
T.
,
Suzuki
,
S.
,
Hirayama
,
H.
,
Shiina
,
K.
,
Minami
,
Y.
,
Isaka
,
H.
,
Fukuda
,
T.
,
Mizutani
,
J.
,
Moriya
,
S.
, and
Madarame
,
H.
, “
Study on Relationship Between Thermal Stratification and Cavity Flow in a Branch Pipe With a Closed End
,” 2003,
Proceedings of the 11th International Conference on Nuclear Engineering
(ICONE-11), No. 36214.
38.
Kawamura
,
T.
,
Shiina
,
K.
,
Ohtsuka
,
M.
,
Tanaka
,
I.
,
Hirayama
,
H.
,
Tanimoto
,
K.
,
Fukuda
,
T.
,
Sakashita
,
A.
,
Mizutani
,
J.
,
Minami
,
Y.
,
Moriya
,
S.
, and
Madarame
,
H.
, 2002, “
Experimental Study on Thermal Striping in Mixing Tees with Hot and Cold Water
,”
Proceedings of the 10th International Conference on Nuclear Engineering
(ICONE-10).
39.
Tanimoto
,
K.
,
Shiraishi
,
T.
,
Suzuki
,
S.
,
Hirayama
,
H.
,
Shiina
,
K.
,
Fukuda
,
T.
,
Mizutani
,
J.
,
Minami
,
Y.
,
Isaka
,
H.
,
Moriya
,
S.
, and
Madarame
,
H.
, 2002, “
Study on Relationship between Thermal Stratification and Cavity Flow in a Branch Pipe with a Closed End
,”
Proceedings of the 10th International Conference on Nuclear Engineering
(ICONE-10), No. 23340.
40.
Fukuda
,
T.
,
Sakashita
,
A.
,
Mizutani
,
J.
,
Matsunaga
,
T.
,
Ogura
,
K.
,
Shiina
,
K.
,
Tanimoto
,
K.
,
Moriya
,
S.
, and
Madarame
,
H.
, 2003, “
Current Effort to Establish a JSME Code for the Evaluation of High-Cycle Thermal Fatigue
,”
Proceedings of the 11th International Conference on Nuclear Engineering
(ICONE-11), No. 36439.
41.
Wakamatsu
,
M.
,
Hirayama
,
H.
,
Kimura
,
K.
,
Ogura
,
K.
,
Shiina
,
K.
,
Tanimoto
,
K.
,
Mizutani
,
J.
,
Minami
,
Y.
,
Moriya
,
S.
, and
Madarame
,
H.
, 2003, “
Study on High-Cycle Fatigue Evaluation for Thermal Striping in Mixing Tees with Hot and Cold Water (1)
,”
Proceedings of the 11th International Conference on Nuclear Engineering
(ICONE-11), No. 36208.
42.
Okamoto
,
A.
,
Nishiguchi
,
I.
, and
Aoki
,
M.
, “
Recent Advancement on the Draft or Alternative Stress Evaluation Criteria in Japan Based on Partial Inelastic Analysis
,” Pressure Vessel and Piping Codes and Standards,
PVP
,
419
, The American Society of Mechanical Engineers, pp.
17
–
24
.
43.
Asada
,
S.
,
Okamoto
,
A.
,
Nishiguchi
,
I.
,
Aoki
,
M.
, and
Asada
,
Y.
, 2004, “
Technical Bases for Alternative Stress Evaluation Criteria in Japan Based on Partial Inelastic Analysis
,” Pressure Vessel and Piping Codes and Standards,
PVP
,
472
, The American Society of Mechanical Engineers, pp.
85
–
92
.
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 by American Society of Mechanical Engineers
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