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UNDERSTANDING EARTHQUAKE DESIGN CRITERIA USED IN JAPAN Charles H. Hofmayer*, Young J. Park* and James F. Costello"
SUMMARY
This paper summarizes the current earthquake design criteria used in Japan for nuclear power plants. Information is presented on the codes and standards and seismic requirements for reactor buildings and containment structures. The most interesting features of the earthquake design criteria used in Japan, in the light of those used in the United States, are summarized.
INTRODUCTION
The U.S. Nuclear Regulatory Commission (USNRC) has undertaken various efforts to understand Japanese earthquake engineering practice.
In 1984, the first Seismic Information Exchange Meeting sponsored by the USNRC and the Ministry of International Trade and Industry (MOT) was held in Palo Alto, California. The aim of this meeting was to improve understanding of the seismic research being conducted in Japan and the United States and to identify areas which could be the basis for future cooperation (Ref. [1]). A similar meeting was held in Tokyo, Japan in 1988.
One of the outcomes of the first information exchange meeting was the establishment of a cooperative agreement between MITI/NUPEC (Nuclear Power Engineering Corporation) and USNRC/BNL (Brookhaven National Laboratory) to perform high level vibration tests and analyses of nuclear power piping. The resulting test program provided extensive non-linear dynamic response data of piping in the elastic-plastic region (Ref. [2]). Currently, a collaborative program between MJTI/NUPEC and USNRC/BNL is underway which involves the Main Steam and Feedwater System Seismic Proving Test at NUPEC's Tadotsu Engineering Laboratory.
In order to further understand Japanese practice, the USNRC sponsored an effort by BNL to translate into English a Japan Electric Association (JEA) publication entitled JEAG 4601-1987, "Technical Guidelines for Aseismic Design of Nuclear Power Plants" (Ref. [3]). The translation has been published as NUREG/CR-6241 (Ref. [4]).
"Brookhaven National Laboratory, Upton, NY 11973-5000
"U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-00j
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DISTRIBUTION OP THIS DOCUMENT IS UNUM&TEG
tf
JEAG 4601-1987, which contains approximately 900 pages of technical material, is a comprehensive handbook for the structural analysis and design of nuclear power plants. It contains chapters dealing with: (a) the selection of earthquake ground motions for a site, (b) procedures to be used to investigate foundation and bedrock conditions, (c) criteria for the evaluation of slope stability and the effects of ground movement on buried piping and structures, and (d) procedures for the analysis and design of structures, equipment and distribution systems (piping, electrical raceways, instrumentation, tubing and HVAC ducting). There are also Supplements to this document published by JEA in 1984 and 1991 which contain additional information. Ref. [5] provides an excellent background on the development of these documents.
In addition to the JEAG documents, a large number of references published by MITI and the Architectural Institute of Japan (AD) have been reviewed. This paper summarizes the current earthquake design criteria used in Japan for nuclear power plants. Additional detailed information on reinforced concrete shear wall structures is presented in a companion paper (Ref. [6]).
CODES AND STANDARDS
Many of the design formulas for reactor buildings and containment structures have been adopted from non-nuclear standards with or without modifications. In the following, both non-nuclear and nuclear design codes and standards related to the seismic design of nuclear power plants are outlined.
The basic design requirements for non-nuclear building structures are defined in orders by the Ministry of Construction (MOC) (Ref. [7]). The detailed design requirements, similar to the ASME and ACI codes, are provided by the following series of AD standards:
• Standard for Structural Calculation of Reinforced Concrete Structures, 1991 also known as "RC-Standard" (Ref. [8]) Design Standard for Steel Structures, 1973 also known as "Steel-Standard" (Ref. [9])
• Standard for Structural Calculation of Steel Reinforced Concrete Structures, 1987 also known as "SRC-Standard" (Ref. [10])
• Design Guidelines for Foundation Structures, 198 8 also known as "Foundation Guidelines" (Ref. [11]).
The above AU standards are used primarily for the traditional allowable stress design required for the seismic design against moderate earthquakes. For the ultimate strength design, or for relatively new design approaches, such as base isolation systems, a series of AD "Guidelines" and "Recommendations" are available, e.g.,
• Design Guidelines for Earthquake Resistant Reinforced Concrete Buildings Based on Ultimate Strength Concept, 1990 (Ref. [12])
• Design Guidelines for Base-Isolated Buildings, 1989 (Ref. [13]).
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DISCLAIMER
This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, make any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.
DISCLAIMER
Portions of this document may be illegible in electronic image products. Images are produced from the best available original document.
For the design/construction standards of nuclear facilities, the Ministry of International Trade and Industry (Mill) is the responsible governmental body. The design of nuclear facilities/equipment is based on the following M1T1 Orders and Notifications.
MITI Order No. 62, "Technical Standards for Nuclear Power Plant Facilities", 1989 (Ref. [14]).
• MITI Notification No. 501, "Technical Standards for Structural Design of Nuclear Power Plant Equipment", 1992 (Ref. [15]).
• M m Notification No. 452, "Technical Standards for Structural Design of Concrete Containment Structures of Nuclear Power Plant Facilities", 1990 (Ref [16]).
In addition, the basic design requirements for the seismic design of nuclear facilities are defined in the following guide:
• "Regulatory Guide for Aseismic Design of Nuclear Power Reactor Facilities", 1981, Japan Atomic Energy Commission (Ref. [17]).
The AIJ has also prepared the following recommendations: • Recommendations for Structural Design of Reactor Buildings, 1988 (Ref. [18]). • Recommendations for Structural Design of Nuclear Reactor Containment Structures,
1978 (Ref. [19]).
SEISMIC REQUIREMENTS BY MOC
One significant feature of the Japanese seismic codes by the MOC is the dual requirement that buildings should be evaluated both for moderate and severe earthquakes, as summarized below. For a moderate earthquake (basic seismic coefficient is C=0.2), conventional allowable stress design is performed, which is similar to the UBC design requirements. However, for a severe earthquake (seismic force is l.Og in terms of 5% damped linear response), ultimate strength design should be performed which requires some type of nonlinear analysis to identify the failure mechanism of a building and the ductility requirements of components.
Earthquake Base Coefficient Analysis Component Evaluation
Moderate Severe
C = 0.2 C = 1.0
Linear Nonlinear
Short-Term Allowable* Ultimate Strength
* 2/3 f c for concrete, and yielding for steel.
hi addition to the above basic seismic load requirements, limitations on the eccentricity and story drift (less than 1/200 for C=0.20) are imposed on each story. The lateral seismic force, Qi9 is defined as,
Qt-Da.Fm.*t*At.fC*Wt (1)
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where D,
F„ Rt A, Z C
= reduction factor due to ductility (~ l /y2 | i - l ; 0.3 for ductile frame, 0.5 for shear wall construction).
= penalty factor due to eccentricity (1.0 for no eccentricity). = spectral shape (function of building vibration frequency). = lateral shear distribution factor (function of story number, i). = Zone coefficient. = 0.2 for moderate earthquake, = 1.0 for severe earthquake. = weight of building above the i-th story.
SEISMIC REQUIREMENTS FOR REACTOR BUILDINGS
The seismic requirements of nuclear power facilities are determined according to the importance classification, As, A, B and C, as listed in Table 1 (Ref. [4]).
In Table 1, the Si earthquake is probably somewhat higher than the OBE of U.S. practice, and the S2 earthquake is roughly equivalent to the SSE. Most parts of reactor buildings and containment structures are classified as As facilities. In general, static and linear dynamic analyses are performed for the Si earthquake, and nonlinear dynamic analyses for the S2 earthquake. Examples of the aseismic importance classification are listed in Table 2 for both BWR and PWR facilities (Ref. [4]).
Table 1. Seismic Requirements
Aseismic Importance Classification Required Analysis
Design Earthquake Aseismic Importance Classification Required Analysis
Horizontal Vertical
Building &
Structures
As Dynamic S 2 1/2 Sj Building
& Structures
As, A Dynamic Static 3.0 C,
1/2 Si c v
Building &
Structures B Static 1.5 Ci
Building &
Structures
C Static Q
Equipment &
Piping
As Dynamic s 2 1/2 S 2
Equipment &
Piping As, A Dynamic
Static s,
3.6 Q 1/2 Si 1.2 C v
Equipment &
Piping B Static 1.8 C,
Equipment &
Piping
C Static 1.2 C, Note: S2 = extreme design earthquake
S, = maximum design earthquake Cx = static seismic coefficient (=0.2)
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The AIJ recommendations for the design of reactor buildings (Ref. [18]), which are also described in JEAG 4601-1987 (Ref. [4]), are based largely on the foregoing AIJ RC-Standard (Ref. [8]). In the recommendations, the concept of "allowable state" has been introduced, which is similar to the classification of levels A, B, C and D limits (i.e., normal, upset, emergency and faulted) in the ASME code. The classification of the allowable states, I, JJ and HI, are summarized briefly in Table 3 based on the information in Ref. [18]. One of the unique features of this recommendation is that for the allowable state III, i.e., for an S2 earthquake or an accident condition, the thermal stress can be neglected.
Table 2. Examples of Aseismic Importance Classification (Ref. [4])
Aseismic importance
Major equipment Aseismic importance BWR PWR
Class As
(i) Nuclear reactor pressure vessel; vessels, pipes, pumps, and valves within the nuclear reactor coolant pressure boundary
(ii) Spent fuel pool (iii) Control rods, control rod driving mecha
nism, control rod driving hydraulic system (scram function)
(iv) Residual heat removal system (cooling mode in shutdown state)
(v) Nuclear reactor containment vessel; piping and valves within the boundary of the nuclear reactor containment vessel
(i) Nuclear reactor pressure vessel; vessels, pipes, pumps, and valves within the nuclear reactor coolant pressure boundary
(ii) Spent fuel pool (iii) Control rods, control rod driving mecha
nism, control rod driving hydraulic system (scram function)
(iv) Residual heat removal system (v) Nuclear reactor containment vessel; piping
and valves within the boundary of the nuclear reactor containment vessel
Class A
(i) Emergency nuclear reactor core cooling system
(ii) Standby gas treatment system (iii) Reactor internal structures
(i) Safety injecting system (ii) Annular air cleaning equipment (iii) Reactor internal structures
Class B (i) Waste disposal system (ii) Steam turbine, condenser, feedwater heater (iii) Fuel pool cooling system
(i) Waste disposal system (ii) Spent fuel pit water cleaning system
Class C
(i) Sample collecting system, floor drainage system, etc.
(ii) Main generator/transformer
(i) Sample collecting system, floor drainage system, etc.
(ii) Turbine equipment, main generator/transformer
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Table 3. Allowable States Allowable
States Plant Condition Allowable Concrete
Compression Stress Other Allowable
Stresses Thermal Stiffness Reduction Factor
I Normal operation y 3 F c Long-term allowable
V2
n S x EQ, storm, snow
%F C 1.5 times above y 3
m S 2 EQ, accident 0.85 F c (e=0.3%) Material strength neglect
DESIGN REQUIREMENTS FOR CONTAINMENT STRUCTURES
The seismic design of concrete containment structures is performed based on MITI Notification No. 452 (Ref. [16]). This document, and in particular the background information upon which this standard is based, may be useful as the test results of large-scale containment structures are extensively utilized. Some unique features are highlighted herein, in the light of the U.S. practice for the seismic design of containment structures.
Loading State According to Notification No. 452, the structural design of containment structures is performed based on the "Loading States" shown in Table 4. Compared to the AIJ recommendations for reactor buildings (see Table 3), the basic design requirements may be considered to be similar.
Table 4. Loading State for Containment Structures
Loading States
Plant Condition
Allowable Concrete Compressive Stress
Other Allowable Stresses
Thermal Stiffness Reduction Factor
I Normal V 3 F C Long-term allowable
(RC standard) Vz n Relief
valve/test
V 3 F C Long-term allowable (RC standard)
Vz
m S,EQ % F C Short-term allowable (RC standard)
y 3
rv S2 EQ/accident Strain limit 0.3% for concrete 0.5% for steel
Neglect
Thermal Stress The evaluation of the thermal stresses is performed according to the following procedure.
• Reduce the elastic stiffness (i.e., Young's modulus) by a factor of 14 for Loading States -1 and II, and Vb for Loading State - III, and calculate the thermal stresses.
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Calculate stresses for other loads using the original elastic stiffness. Combine the above stresses. For the Loading State - IV (S2 earthquake and accident), the thermal stress is neglected.
P t = 0 . 5 0 {%) P t = 0 . 7 1 {%) P t « 0 . 9 7 {%) P t = Steel Ratio
2000 3000 4000 5000
Stress in Rebars (kg/cm2)
Figure 1. Reduction of Stiffness and Stress Level in Rebars (Ref. [18]).
Figure 1 shows the basis for the stiflhess reduction factors of Vi and Vs shown in Tables 3 and 4. This figure is based on the results of tests of reinforced concrete components subjected to lateral loading under elevated temperature conditions. The stress level of the component is represented by the stress in rebars and the stiflhess ratio is the secant stiflhess at each stress level divided by the initial stiffness. The allowable stress for Loading States I & II roughly corresponds to a stress of 2,000 kg/cm2 (28.5 ksi), and for Loading State HI, it is about 3,500 kg/cm2 (49.8 ksi). Tables 3 and 4 also indicate that the effects of thermal stresses can be neglected for the S 2 earthquake and accidental loadings both for containment structures and reactor buildings. This is based on the observation that the thermal stresses do not alter the ultimate strength of shear wall structures (Ref. [16]).
Further information on the seismic design of reactor buildings and containment structures is provided in Reference [6].
SUMMARY OF INTERESTING FEATURES
The most interesting features of the earthquake design criteria used in Japan, in the light of that used in the United States are summarized below.
For the seismic analysis and design of structures: • In principle, reactor buildings should be supported on rock. • Structures are classified into four classes, As, A, B and C. Different seismic
requirements and different structural analyses are applied to each class.
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Nonlinear analyses are used for dynamic response evaluations, including the uplifting of foundations. Dynamic effects are not considered in evaluating the seismic force due to vertical motions. Lower damping values are used. See Table 5 for a comparison between Japanese and U.S. damping values. The absolute sum is used to combine the responses of the vertical component and one horizontal component. Static analyses (in addition to dynamic analyses) are required for class As and A structures. Reduced component stiffnesses are used to evaluate the thermal stresses. Extensive experimental results are available in Japan on the nonlinear behavior of structural components, e.g., shear walls and steel bracing systems, and large scale containment structures.
Table 5. Comparison of Damping (%) Values
JAPAN U.S.
S„ S 2 OBE SSE
Concrete Structures: (a) Reinforced (b) Prestressed (PCCV)
5.0 3.0
4.0 2.0
7.0 5.0
Welded Steel Structures 1.0 2.0 4.0
Bolted Structures 2.0 4.0 7.0
Piping Regulatorv Guide 1.61
(a) Large Diameter > 12.0 in D 0 . (b) Small Diameter < 12.0 in D 0 .
ASME Code Case N-411 (c) Function of frequency for
response spectrum analysis.
JEAG4601 (d) Function of type and number
of supports, with and without thermal insulation.
0.5 to 2.5
2.0 1.0
3.0 2.0
Piping Regulatorv Guide 1.61
(a) Large Diameter > 12.0 in D 0 . (b) Small Diameter < 12.0 in D 0 .
ASME Code Case N-411 (c) Function of frequency for
response spectrum analysis.
JEAG4601 (d) Function of type and number
of supports, with and without thermal insulation.
0.5 to 2.5
5.0 (s 10 Hz) 2.0 (* 20 Hz)
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For the seismic analysis and design of equipment and piping: • In principle, equipment/piping systems should be designed within the "rigid zone"
(natural frequency higher than 20 Hz or the dominant frequency of the related floor response spectrum).
• Criteria specify four seismic classes with different seismic input requirements. • The transient portion of LOCA loads are not combined with seismic loads. • In terms of the allowable stresses for reactor vessels and primary piping, the criteria
for St earthquakes corresponds to the level-C limit in ASME code, and the criteria for S 2 corresponds to the level-D limit. See Table 6.
• The use of nonlinear analysis is recommended for equipment and piping systems. • Criteria requires both static and dynamic analysis for As and A class equipment. • Japanese practice prefers simple modeling of equipment, such as a stick model. • ± 10% broadening of floor spectra is recommended. • Lower damping values are used for equipment and piping. • A higher allowable strength is used for expansion anchor bolts.
Table 6. Comp arison of Allowable Stresses Under Seismic Loads (a) ASME Section IE
LOADING CONDITIONS REACTOR VESSEL CLASS 1 PIPING
Level B Limit, Upset (OBE) 1.5 SL 1.8 S„„ 1.5 S y
Level C Limit, Emergency 1.8 S m , 1.5 Sy 2.25 S m , 1.8 S y
Level D Limit, Faulted (SSE) 3.6 S m , S„ 3 S m , 2 S v
(b)JEAG4601
LOADING CONDITIONS REACTOR VESSEL PRIMARY PTPING
S x Earthquake
S 2 Earthquake
1-5 Sy, S„
s. 2.25 S m
3 S m
CONCLUSIONS
There is a major continuing research effort in Japan on the seismic safety of nuclear power plants. A number of differences between the U.S. and Japanese practices have been observed, as well as some interesting approaches. U.S. engineers are encouraged to obtain Japanese literature to better understand the bases for Japanese seismic practice and, where appropriate, incorporate this knowledge into U.S. practice.
ACKNOWLEDGEMENTS
The work presented in this paper was performed under the auspices of the U.S. Nuclear Regulatory Commission. The permission granted by the Japan Electric Association (JEA) to publish the
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translation of JEAG 4601-1987 is greatly appreciated, as well as the permission by the Architectural Institute of Japan to use the figures and tables from ALTs publications.
The authors also appreciate the assistance provided by Professor Heki Shibata and the members of his JEA Code Division Committee, Dr. Muneaki Kato, Mr. Rokuro Endo and others, who reviewed and commented on the English Translation of JEAG 4601-1987. Special thanks is also extended to Mr. Nagatoshi Tanaka of NUPEC for his many years of effort in furthering good cooperation between the United States and Japan in the seismic area.
The translated text of JEAG 4601-1987 was reviewed by an expert panel consisting of the following individuals:
Dr. Robert Cloud, Robert L. Cloud & Associates, Inc. Dr. Carl Costantino, City College of New York Dr. I.M. Idriss, University of California, Davis Dr. Robert Kennedy, Structural Mechanics Consulting, Inc. Dr. Joseph Penzien, International Civil Engineering Consultants, Inc. Dr. Paul Pomeroy, Rondout Associates Dr. John Stevenson, Stevenson & Associates
The authors appreciate the expert panel's comments and encouragement to publish the translation, as well as the assistance in identifying the most interesting features of the criteria used in Japan, in the light of that used in the United States.
DISCLAIMER
The findings and opinions expressed in this paper are those of the authors, and do not necessarily reflect the views of the U.S. Nuclear Regulatory Commission or Brookhaven National Laboratory.
REFERENCES
[1] "Proceedings of the MITI-USNRC Seismic Information Exchange Meeting," Palo Alto, California, July 18-20, 1984, NUREG/CP-0059, 1985.
[2] Y.J. Park, J.R. Curreri and C.H. Hofmayer, "The High Level Vibration Test Program," NUREG/CR-5585, May 1991.
[3] "Technical Guidelines for Aseismic Design of Nuclear Power Plants," JEAG 4601-1987, Japan Electric Association, 1987.
[4] Y.J. Park and C.H. Hofinayer, "Technical Guidetines for Aseismic Design of Nuclear Power Plants - Translation of JEAG 4601-1987," NUREG/CR-6241, June 1994.
[5] H. Shibata, "Recent Development of Aseismic Design Practice of Equipment and Piping System in Japan," Transactions of the ASME, Journal of Pressure Vessel Technology, Vol. 115, May 1993.
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[6] Y.J. Park and C.H. Hofinayer, "Shear Wall Experiments and Design in Japan," 5th Symposium on Current Issues Related to Nuclear Power Plant Structures, Equipment and Piping, Lake Buena Vista, Florida, Dec. 1994.
[7] "Guidelines and Commentary on Structural Design Calculations," Japan Architectural Center, 1981.
[8] "Standard for Structural Calculation of Reinforced Concrete Structures," AIJ, 1991. [9] "Design Standard for Steel Structures," AIJ, 1973. [10] "Standard for Structural Calculation of Steel Reinforced Concrete Structures," AIJ, 1987. [11] "Design Guidelines for Foundation Structures," AIJ, 1988. [12] "Design Guidelines for Earthquake Resistant Reinforced Concrete Buildings Based on
Ultimate Strength Concept," AD", 1990. [13] "Design Guidelines for Base-Isolated Buildings," ALT, 1989. [14] "Technical Standards for Nuclear Power Plant Facilities," MITI Order No. 62, ANRE/MITI,
1989. [15] "Technical Standards for Structural Design of Nuclear Power Plant Equipment," MITI
Notification No. 501, ANRE/MJ.TL 1992. [ 16] "Technical Standards for Structural Design of Concrete Containment Structures of Nuclear
Power Plant Facilities," MITI Notification No. 452, ANRE/Mm, 1990. [17] "Regulatory Guide for Aseismic Design of Nuclear Power Reactor Facilities," Japan Atomic
Energy Commission, 1981. [18] "Recommendations for Structural Design of Reactor Buildings," AIJ, 1988. [19] "Recommendations for Structural Design of Nuclear Reactor Containment Structures," AIJ,
1978.
DISCLAIMER
This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.
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