UNDERSTANDING SEISMIC DESIGN CRITERIA FOR JAPANESE NUCLEAR POWER PLANTS

Transcription

1 BNL-NUREG UNDERSTANDING SEISMIC DESIGN CRITERIA FOR JAPANESE NUCLEAR POWER PLANTS Y.J. Park and C.H. Hofmayer Brookhaven Natonal Laboratory Upton, Long Island, New York J.F. Costello U.S. Nuclear Regulatory Commsson Washngton, D.C ABSTRACT Ths paper summarzes the results of recent survey studes on the sesmc desgn practce for nuclear power plants n Japan. The sesmc desgn codes and standards for both nuclear as well as nonnuclear structures have been revewed and summarzed. Some key documents for understandng Japanese sesmc desgn crtera are also lsted wth bref descrptons. The paper hghlghts the desgn crtera to determne the sesmc demand and component capacty n comparson wth U.S. crtera, the background studes whch have led to the current Japanese desgn crtera, and a survey of current research actvtes. More detaled techncal descrptons are presented on the development of Japanese shear wall equatons, desgn requrements for contanment structures, and ductlty requrements. INTRODUCTION As part of the USNRC efforts to understand Japanese earthquake engneerng practce, a publcaton by the Japan Electrc Assocaton enttled JEAG (Ref. [1]), "Techncal Gudelnes for Asesmc Desgn of Nuclear Power Plants", was translated and recently publshed as NUREG/CR (Ref. [2]). The gudelnes, whch contan approxmately 900 pages of techncal materal, provde a detaled descrpton of the current Japanese sesmc desgn methods, and can be consdered to be a key document n understandng the sesmc desgn crtera and practce for Japanese nuclear power plants. In addton to the JEAG document, a large number of related documents publshed by the Mnstry of Internatonal Trade and Industry (MITT) and the Archtectural Insttute of Japan (AD) have been revewed to understand the sesmc desgn crtera for varous parts of nuclear power plants n Japan. As a result of the survey studes, t was found that the Japanese desgn practce generally follows the ASME codes n desgnng components made of metallc materals, e.g., ppng and vessels. However, dfferences between U.S. and Japanese practces have been found regardng the sesmc desgn of concrete structures, partcularly n shear wall desgn crtera. DISTRIBUTION OF THIS DOCUMENT SS UNLIMITED LP

2 DISCLAIMER Portons of ths document may be llegble n electronc mage products, mages are produced from the best avalable orgnal document.

3 Ths paper s ntended to provde U.S. engneers wth an overvew of the current sesmc desgn practce of nuclear power plants n Japan n some detal. The background studes for key desgn crtera for reactor buldngs and contanment structures are descrbed based on the nformaton collected n a recent survey. CODES AND STANDARDS Many of the desgn formulas for reactor buldngs and contanment structures have been adopted from non-nuclear standards wth or wthout modfcatons. In the followng, both non-nuclear and nuclear desgn codes and standards related to the sesmc desgn of nuclear power plants are outlned. The basc desgn requrements for non-nuclear buldng structures are defned n orders by the Mnstry of Constructon (MOC) (Ref [3]). The detaled desgn requrements, smlar to the ASME and ACI codes, are provded by the followng seres of AIJ standards: Standard for Structural Calculaton of Renforced Concrete Structures, 1991 also known as "RC-Standard M (Ref. [4]). Desgn Standard for Steel Structures, 1973 also known as "Steel-Standard" (Ref. [5])- Standard for Structural Calculaton of Steel Renforced Concrete Structures, 1987 also known as "SRC-Standard" (Ref. [6]). Desgn Gudelnes for Foundaton Structures, 1988 also known as "Foundaton Gudelnes" (Ref. [7]). The above AIJ standards are used prmarly for the tradtonal allowable stress desgn requred for the sesmc desgn aganst moderate earthquakes. For the ultmate strength desgn, or for relatvely new desgn approaches, such as base solaton systems, a seres of AD "Gudelnes" and "Recommendatons" are avalable, e.g., Desgn Gudelnes for Earthquake Resstant Renforced Concrete Buldngs Based on Ultmate Strength Concept, 1990 (Ref. [8]). Desgn Gudelnes for Base-Isolated Buldngs, 1989 (Ref. [9]). For the desgn/constructon standards of nuclear facltes, the Mnstry of Internatonal Trade and Industry (Mll) s the responsble governmental body. The desgn of nuclear facltes/equpment s based on the followng MTTI Orders and Notfcatons: MITI Order No. 62, "Techncal Standards for Nuclear Power Plant Facltes", 1989 (Ref. [10]). MITI Notfcaton No. 501, "Techncal Standards for Structural Desgn of Nuclear Power Plant Equpment", 1992 (Ref. [11]). MTTI Notfcaton No. 452, "Techncal Standards for Structural Desgn of Concrete Contanment Structures of Nuclear Power Plant Facltes", 1990 (Ref. [12]).

4 In addton, the basc desgn requrements for the sesmc desgn of nuclear facltes are defned n the followng gude: "Regulatory Gude for Asesmc Desgn of Nuclear Power Reactor Facltes", 1981, Japan Atomc Energy Commsson (Ref. [13]). The AJJ has also prepared the followng recommendatons: Recommendatons for Structural Desgn of Reactor Buldngs, 1988 (Ref. [14]). Recommendatons for Structural Desgn of Nuclear Reactor Contanment Structures, 1978 (Ref. [15]). SEISMIC REQUIREMENTS BY MOC One sgnfcant feature of the Japanese sesmc codes by the MOC s the dual requrement that buldngs should be evaluated both for moderate and severe earthquakes, as summarzed below. For a moderate earthquake (basc sesmc coeffcent s C=0.2), conventonal allowable stress desgn s performed, whch s smlar to the UBC desgn requrements. However, for a severe earthquake (sesmc force s l.og n terms of 5% damped lnear response), ultmate strength desgn should be performed whch requres some type of nonlnear analyss to dentfy the falure mechansm of a buldng and the ductlty requrements of components. Earthquake Base Coeffcent Analyss Component Evaluaton Moderate C = 0.2 Severe C=1.0 * 2/3 f c for concrete, and yeldng for steel. Lnear Nonlnear Short-Term Allowable* Ultmate Strength In addton to the above basc sesmc load requrements, lmtatons on the eccentrcty and story drft (less than 1/200 for C=0.20) are mposed on each story. The lateral sesmc force, Q;, s defned as, Q t -D r F M.R t -A r Z.C.W t (1) where D s = reducton factor due to ductlty ( s 1A/2J» - l; 0.3 for ductleframe,0.5 for shear wall constructon). F^, = penalty factor due to eccentrcty (1.0 for no eccentrcty). Rt = spectral shape (functon of buldng vbraton frequency). Aj = lateral shear dstrbuton factor (functon of story number, ). Z = Zone coeffcent. C =0.2 for moderate earthquake, = 1.0 for severe earthquake. W; = weght of buldng above the -th story.

5 OVERVIEW OF DESIGN CRITERIA FOR NPP The sesmc requrements of nuclear power facltes are determned accordng to the mportance classfcaton, As, A, B and C, as lsted n Table 1 (Ref. [2]). Table 1. Sesmc Requrements Asesmc Importance Requred Analyss Classfcaton Desgn Earthquake Horzontal Vertcal Buldng & Structures As Dynamc S 2 1/2 S 2 As, A Dynamc Statc S 3.0 C B Statc 1.5 CI 1/2 S, c v C Statc Q Equpment & Ppng As Dynamc s 2 1/2 S 2 As, A Dynamc Statc S 3.6 Q B Statc 1.8 C C Statc 1.2 Q Note: S 2 = extreme desgn earthquake Sj = maxmum desgn earthquake Q = statc sesmc coeffcent (=0.2) 1/2 S x 1.2 C v Table 2 compares U.S. and Japanese desgn allowable stresses for reactor vessels and Class 1 ppng. The Japanese S x desgn earthquake, whch represents a moderate earthquake, s somewhat hgher than the OBE of U.S. practce, and the S 2 earthquake, whch represents a severe earthquake, s roughly equvalent to the SSE. Table 3 lsts some desgn dampng values used n both countres. In general, the dampng values used n Japan are lower than those used n the U.S. It should be noted that, snce nonlnear analyses are requred for the S 2 earthquake, an addtonal hysteretc dampng s also accounted for n the Japanese sesmc desgn.

6 Table 2. Comparson of Allowable Stresses Under Sesmc Loads (a) ASME Secton HI [LOADING CONDITIONS REACTOR VESSEL CLASS 1 PIPING Level B Lmt, Upset (OBE) 1.5 S m 1.8 S,, 1.5 S, Level C Lmt, Emergency 1.8 S^ 1.5 S y 2.25 S,,, 1.8 S y 1 Level D Lmt, Faulted (SSE) 3.6 S. S 3 S, 2 S (b)jeag4601 LOADING CONDITIONS REACTOR VESSEL PRIMARY PIPING St Earthquake S^ Earthquake 1-5 Sy, S 2.25 S m 3S m Table 3. Comparson of Dampng (%) Values JAPAN U.S. Su S 2 OBE SSE Concrete Structures: (a) Renforced (b) Prestressed (PCCV) Welded Steel Structures Bolted Structures Ppng Regulators Gude 1.61 (a) Large Dameter > 12.0 n D 0. (b) Small Dameter < 12.0 n D (c) ASME Code Case N-411 Functon of frequency for response spectrum analyss. 5.0 (< 10 Hz) 2.0 (s 20 Hz) (d) JE AG 4601 Functon of type and number of supports, wth and wthout thermal nsulaton. 0.5 to 2.5

7 DESIGN REQUIREMENTS FOR SHEAR WALL STRUCTURES The AU standard for RC structures (Ref [4]), whch can be consdered as the equvalent of ACI-318, defnes the allowable stresses for concrete structures as lsted n Table 4. Fgure 1 compares the short-term allowable shear stress from Table 4 wth avalable test data. The upper and lower lmts of ACI-318 (3.5^ and 1.9JZ ), are also shown n the fgure. Table 4. Allowable Stresses for Concrete (unt: kg/cm 2 ) Long-term allowable (D.L. & L.L.) Short-term allowable (Sesmc Load) Compresson Shear, v c Compresson Shear, v e V 3 F' C mn I \ F F K 30~' * loo" t %F' e mm ( ^, F ' ) 0 :oo 20c 2:4 30a 360 *oo oc kg em* Concrete Strength (kg/cm 3 ) Fgure 1. Shear Crackng Stress of Shear Walls Under Lateral Loads (Ref. [4]).

8 The AIJ recommendatons for the desgn of reactor buldngs, whch are also descrbed n JEAG (Ref. [1], [2]), are based largely on the foregong AD RC-Standard (Ref. [4]). In the recommendatons, the concept of "allowable state" has been ntroduced, whch s smlar to the classfcaton of levels A, B, C and D (.e., normal, upset, emergency and faulted) n the ASME code. DEVELOPMENT OF SHEAR WALL EQUATIONS In the sesmc desgn of reactor buldngs, nonlnear dynamc analyses are generally requred when consderng the response to the S 2 earthquake. Therefore, the shear wall equatons are used not only for the component strength evaluaton, but also to determne the restorng force characterstcs of the nonlnear structural models. The latter requrement necesstates the development of shear wall equatons whch can predct the ultmate capacty and the nonlnear deformaton propertes wthout bas or excessve conservatsm. Currently, Hrosawa's equaton s extensvely used n the sesmc desgn of reactor buldngs. Hrosawa's equaton, whch s also extensvely used for non-nuclear buldngs, s also called the modfed Arakawa's equaton, as t was developed by modfyng the shear strength equaton for beams and columns by Arakawa (Ref. [4]). Some benchmark studes, whch led to the current Japanese shear wall equatons, are descrbed below. Arakawa's Study (1960) The current Japanese desgn crtera on the shear capacty of R.C. components are based on the studes by Arakawa (Ref. [16], [17]). Fgure 2 shows a part of the correlaton study to develop the followng shear strength equaton for renforced concrete beams, whch now s called the orgnal Arakawa's equaton:, 0.12(180 *F_),» (*/«V r. K p - ^ ^ 2.7 ^ 7 r (2) n whch K = reducton factor for scale effect (=0.72 when d>40 cm) K = 0.82 p, 0-23 p t = tenson steel rato n percent M/Vd = shear span rato (replaced by 3 when larger than 3) p w = strrup rato The contrbuton of the renforcement to the shear strength, v s or the second term of Eq. 2, was obtaned emprcally as llustrated n Fgure 3. A total of 219 pars of specmens, for whch the shear renforcement was the only parameter (.e., one beam s shear renforced and the other s not), were used to drectly determne the contrbuton by the shear renforcement as follows: (Contrbuton by Shear Renforcement) =(Shear Strength of Renforced Beam) - (Shear Strength of Unrenforced Beam). The results presented n Fgure 3 ndcate that the truss theory approxmaton, on whch the current ACI equatons are based, largely overestmates the shear contrbuton by steel. The data shown n Fgures 2 and 3, n fact, had a major mpact on the subsequent desgn code development n Japan.

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10 TRUSS ANALOGY Fgure 3. Contrbuton by Rebars to Shear Capacty (replotted based on the data n Ref. [16], components wth p w f t <20 are excluded). 30 2S 20-»=138»=0.9S0 <r=» =0.083 IS oj- :on»=70 =1.IS»= =0.»2 ELLX 0.S testq /e*le<2» (a) AIJ equaton (Eq. 1) t \ : '! l. (b) ACI equaton (Eq. 2) Fg. 4. Hstograms for the Ratos of Tests to Calculated Ultmate Shear Strength of R.C. Beams (Ref. [18]).

11 The correspondng ACI equaton s, Vjjbs). V c to f y bd V e (Ibs) - < > 1.9 ^f e 2500 p w bd 3.5,fjbd & V M Fgure 4 shows the results for correlaton studes of the above shear strength equatons (Ref. [18]). The ACI equatons show a much larger scatter compared wth the AU equaton. Hrosawa's Study on Shear Walls (1975) Hrosawa's equaton s extensvely used n Japan to calculate the ultmate shear strength of shear walls both for non-nuclear and nuclear power structures. The emprcal equaton was obtaned by modfyng the above Arakawa's equaton (Ref. [19]): VAKg) p t (F e * 180)., 2.7Jf P y t 0.1a \JMIVD * 0.12 bj (4) where b e = effectve thckness of wall (when a wall hasflanges, *b e ' s calculated from a unformthckness cross-secton wth equal area); j = 0.83 D (D = total length of wall) p t (%) =100 A,/(b e j), tenson axal steel rato consderng a wall as a column (A, = total axal steel area n a flange); MAD = shear span rato; p w = horzontal steel rato usng the effectve thckness b e ; F c = concrete strength (kg/cm 2 ) f^, = steel yeld stress (kg/cm 2 ) o 0 = average axal stress (kg/cm 2 ) Table 5 shows part of the correlaton studes on ths equaton. DESIGN REQUIREMENTS FOR CONTAINMENT STRUCTURES The sesmc desgn of concrete contanment structures s performed based on MITI Notfcaton No. 452 (Ref. [12]). Ths document, and n partcular the background nformaton upon whch ths standard s based, may be useful as the test results of large-scale contanment structures are extensvely utlzed. Some unque features are hghlghted heren, n the lght of the U.S. practce for the sesmc desgn of contanment structures. Loadng State Accordng to Notfcaton No. 452, the structural desgn of contanment structures s performed based on the "Loadng States" shown n Table 6.

12 Table 5. Correlaton of Hrosawa's Equaton (Ref. [19]). Range of Number of Range Mean Standard Parameters Shear Walls Vtest/Vcal Vtest/Vcal Devaton 0<M/VD* <M/VD* <M/VD <t/L< <t/L< t/l= Pw= <p w < <p w < <p w < l-2* Pw Total M/VD = shear span rato bjd = thckness rato = horzontal shear renforcement rato p w Loadng States I n Plant Condton Normal Relef valve/test Table 6. Loadng State for Contanment Structures Allowable Concrete Compressve Stress V 3 F C Other Allowable Stresses Long-term allowable (RC standard) Thermal Stffness Reducton Factor y 2 m S,EQ %F C Short-term allowable (RC standard) y 3 rv S 2 EQ/accdent Stran lmt 0.3% for concrete Neglect 0.5% for steel Thermal Stress The evaluaton of the thermal stresses s performed accordng to the followng procedure. Reduce the elastc stffness (.e., Young's modulus) by a factor of Vz for Loadng States -1 and n, and Vz for Loadng State - HI, and calculate the thermal stresses. Calculate stresses for other loads usng the orgnal elastc stffness. Combne the above stresses. For the Loadng State - TV (S 2 earthquake and accdent), the thermal stress s neglected.

13 Q >».» Stress n Rebars (kg/cur) Fgure 5. Reducton of Stffness and Stress Level n Rebars (Ref. [14]). Fgure 5 shows the bass for the stffness reducton factors of Vz and Vz shown n Table 6. Ths fgure s based on the results of tests of renforced concrete components subjected to lateral loadng under elevated temperature condtons. The stress level of the component s represented by the stress n rebars and the stffness rato s the secant stffness at each stress level dvded by the ntal stffness. The allowable stress for Loadng States I & II roughly corresponds to a stress of 2,000 kg/cm 2 (28.5 ks), and for Loadng State HI, t s about 3,500 kg/cm 2 (49.8 ks). Table 6 also ndcates that the effects of thermal stresses can be neglected for the S 2 earthquake and accdental loadngs both for contanment structures and reactor buldngs. Ths s based on the observaton that the thermal stresses do not alter the ultmate strength of shear wall structures (Ref. [12]). DUCTTLTrY CAPACITY/REQUIREMENTS The allowable shear deformaton for both reactor buldngs and contanment structures are defned to be 0.2% n radans (Ref. [1], [2], [12]). Based on a statstcal analyss of avalable box-shaped and cylndrcal shear wall structures, the mnmum shear deformaton capacty was estmated to be 0.4% n radans (Ref. [20]). Therefore, a safety factor of 2.0 s used to account for the large scatter assocated wth the shear deformaton capacty. Fgure 6 shows the results of a study of avalable data on box-shaped and cylndrcal shear walls by the authors. In the fgure, the ultmate shear deformaton capacty s plotted aganst the maxmum shear stress. For hgh strength shear walls, the deformaton capacty tends to decrease to about 0.4% n radans. However, for shear walls wth less shear strength, a much hgher ductlty can be expected. The results of a smlar study on shear deformaton capacty are presented for both RCCV and PCCV structures n Ref. [12], n whch the ultmate shear deformaton capacty was plotted aganst the normalzed steel rato. Accordng to ths study, the ductlty of contanment structures tends to decrease as the steel rato (therefore shear strength) ncreases. The mnmum shear deformaton

14 2.0 o o "3 a LOT 1 I C ^«S^JOL-g. Y = 0.4% 0-5 LO L5 (ks) Shear Strength (Test) Fgure 6. Shear Deformaton Capacty of Box-Shaped and Cylndrcal Shear Walls. capacty for heavly renforced contanment structures converges to about 0.4% n radan. A more detaled statstcal study on the shear deformaton capacty of ordnary RC shear walls s also descrbed n Ref. [18]. SUMMARY AND CONCLUSIONS The current Japanese practce for sesmc desgn, partcularly that related to the sesmc desgn of reactor buldngs and contanment structures, was revewed n some detal as well as related research actvtes. There have been three decades of extensve expermental work, and judgng from publcatons such as the transactons of the Annual Meetng of AJJ the expermental studes on components are contnung as the emphass s beng shfted more to newer desgn approaches, such as base solaton devces and pre-fabrcated components. The value of the wealth of accumulated nformaton n Japan has been recognzed by many promnent U.S. engneers. In fact, the test results of large-scale shear wall structures and pre-stressed components have been utlzed n the fraglty evaluaton of nuclear reactor facltes, as they are the only avalable source of test data for these types of structures. The extent to whch Japanese test data have been used by U.S. engneers, however, has been serously lmted due to the language barrer. Further efforts should be made to utlze the avalable nformaton. The crtcal evaluaton of exstng nuclear facltes and the development of new desgn concepts requre much more than structural analyss codes. The accumulated test results, mostly performed usng modern testng facltes and earthquake-lke loadng condtons, should also be revewed and evaluated n detal.

15 ACKNOWLEDGEMENTS The authors wsh to thank K. Akno, S. Kawakam, T. Tara and N. Tanaka of NUPEC, S. Yoshzak of Tase Co., M. Kanechka of Kajma Co., and R. Shohara and Y. Takeuch of Shmzu Co. for the nformaton and cooperaton n ths survey study. The permsson by the Archtectural Insttute of Japan to use the fgures and tables from AITs publcatons s greatly apprecated. The work presented n ths paper was performed under the auspces of the U.S. Nuclear Regulatory Commsson. Thefndngsand opnons expressed n ths paper are those of the authors, and do not necessarly reflect the vews of the U.S. Nuclear Regulatory Commsson or Brookhaven Natonal Laboratory. REFERENCES [I] "Techncal Gudelnes for Asesmc Desgn of Nuclear Power Plants, JEAG , Japan Electrc Assocaton, [2] Y.J. Park and C.H. Hofnayer, "Techncal Gudelnes for Asesmc Desgn of Nuclear Power Plants - Translaton of JEAG , "NUREG/CR-6241, June [3] "Gudelnes and Commentary on Structural Desgn Calculatons," Japan Archtectural Center, [4] "Standard for Structural Calculaton of Renforced Concrete Structures," AU, [5] "Desgn Standard for Steel Structures," AH, [6] "Standard for Structural Calculaton of Steel Renforced Concrete Structures," AIJ, [7] "Desgn Gudelnes for Foundaton Structures,"AIJ, [8] "Desgn Gudelnes for Earthquake Resstant Renforced Concrete Buldngs Based on Ultmate Strength Concept," AD, [9] "Desgn Gudelnes for Base-Isolated Buldngs," AIJ, [10] "Techncal Standards for Nuclear Power Plant Facltes," METI Order No. 62, ANRE/MITL [II] "Techncal Standards for Structural Desgn of Nuclear Power Plant Equpment," MITI Notfcaton No. 501, ANRE/MITI, [12] "Techncal Standards for Structural Desgn of Concrete Contanment Structures of Nuclear Power Plant Facltes," MITI Notfcaton No. 452, ANRE/MITI, [13] "Regulatory Gude for Asesmc Desgn of Nuclear Power Reactor Facltes," Japan Atomc Energy Commsson, [14] "Recommendatons for Structural Desgn of Reactor Buldngs," AIJ, [15] "Recommendatons for Structural Desgn of Nuclear Reactor Contanment Structures," ALT, [16] T. Arakawa, "Allowable Shear Stress and Shear Renforcement of RC Beams," Concrete Journal, Japan, Vol. 8, No. 7, July [17] K. Ohno and T. Arakawa, "Study on Shear Resstance of Renforced Concrete Beams," Transacton of AIJ, No. 66, October 1960.

16 [18] "Data for Ultmate Strength Desgn of Renforced Concrete Structures," AU, [19] M. Hrosawa, "Past Expermental Results on Renforced Concrete Shear Walls and Analyss on Them," Buldng Research Insttute, Mnstry of Constructon of Japan, March [20] T. Setogawa, "Allowable Lmt of Shear Walls n Reactor Buldngs," Annual Meetng of AU, pp , Oct DISCLAIMER Ths report was prepared as an account of work sponsored by an agency of the Unted States Government. Nether the Unted States Government nor any agency thereof, nor any of ther employees, makes any warranty, express or mpled, or assumes any legal lablty or responsblty for the accuracy, completeness, or usefulness of any nformaton, apparatus, product, or process dsclosed, or represents that ts use would not nfrnge prvately owned rghts. Reference heren to any specfc commercal product, process, or servce by trade name, trademark, manufacturer, or otherwse does not necessarly consttute or mply ts endorsement, recommendaton, or favorng by the Unted States Government or any agency thereof. The vews and opnons of authors expressed heren do not necessarly state or reflect those of the Unted States Government or any agency thereof.