ML20044B702

Partially Withheld Commission Paper Informing of Denial of Request for Seismic Reanalysis of All Operating Plants

ML20044B702
Person / Time
Issue date:	01/28/1980
From:	Malsch M NRC OFFICE OF THE GENERAL COUNSEL (OGC)
To:
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References
FOIA-92-436 SECY-A-80-013, SECY-A-80-13, NUDOCS 9303030147
Download: ML20044B702 (51)

Text

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UNITED ETATES NUCLEAR REGULATORY COMMISSION stcy. A.e0.u LADJUDICATORY ITEM

% ts, m 0 COMMISSIONER ACTIO N Fer:

The Cc==issicners Frc=:

P.artin G. Maisch Deputy General Counsel DIRECTOR'S DEN ~AL OF 2.206 RELIEF (IN THE MiTTER

Subject:

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To inform the Ccc=issien of the denial of a Purcese:

request for seismic reanalysis of all operat-q u" '

ing plants p'hich,j l

Review Tire Expires:

TebrJary 8, 19R0 (as extended).

Discussien:

Cn March 28, 1979, the Unicn of Concerned Scien-tists ("UCS") petitioned the Cc==1ssion to require all pla.nts with operating licenses to perform a seismic reanalysis within~120 days.

[ Attach =ent A).

The petition was prompted by the previcusly aracunced discovery of errers in the ec=puter code used to analyce seismic stress on piping systens in five operating plants.

UCS views this errcne-tr,fs c2 2, d m.s :::d u n a. o, ous seistic analysis as cnly cne example of brcad-in 2::0 dan:e wdh the fm:? m cf 'nforO2C0 ranging deficiencies in the URC's seistic analysis of Operating plants.

Accordingly, UCS requested Act, exem;m seistic reanalysis of six.seistic factors for f4 #/3g cperating plants. If The petition was referred

• 0M-c to the Directer of Nuclear Reactor Ee ulation pursuant to 10 CFR 2.206. 2/

Cn Januiry 10, 19c0, the Director, NE.R denied the petition.

freerield O

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These are: =agnitude of the Safe Shutdown Earthquake; C

g cund =ction at the site; structure motien during a seistic event; =ction of plant equiptent supported by site structures;

' cads en structures, syste=s, and compenents in appre-e 5m d'a priate combinations with other icads, and the corresponding allow.

sa o

y3' able leadings; and ccnformance of the "as-built" plant to the

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design specificatiens.

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0-m UCS cited this provisien as a casis for Cctrissicn jurisdiction.

2f raised Ecwever; the petition was referred to staff because it L*

~ ~ 0-On December 10, 1979, UCS moved for S O@$

many technical issues.

mission reconsideration of the decisicn to refer the petit' F

staff, noting in particular staff's failure to act prc=

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[ Attach =ent C].

New that the petitien has been denie'

$$o i. % asis for UCS' =ction is teot.

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t UCS contends that two circumstances support its petition: staff reliance on industry computer codes, and the development of improved techniques of seismic analysis.

First, the NRC staff does not examina or verify the complex conputer codes used to model the effects of earthquakes.

UCS believes that the use of these codes presents " obvious problems" because different computational methods are used by individual architect-engineers.

Thus, UCS contends that the NRC cannot be reasonably con-fident of the adequacy of codes used in the past or those which are considered acceptable today.

Second, the NRC has not used improved seismic analysis methods to reconsider the adequacy of earlier licensing decisions.

UCS believes that current knowledge must be applied to existing l

plants to check whether their. earthquake protec-tion is adequate to permit continued operation.

In addition, UCS notes that the NRC has not ade-quately explained why older plants can continue to operate even though they were subjected to seismic reviews considered incomplete by today's i

standards.

t In his denial, the Director discussed four major the ongoing NRC seismically oriented programs:

Systematic Evaluation Program (SEP) for eleven older plants; Task Action Plan A 40 (TAP A 40);

)

the Seismic Safety Margins Research Program (SSMRP); and the Code Verification Program (CVP).

j These programs are in addition to the sei.smic reviews of operating plants for the items identi-fied in recent Inspection and Enforcement Bulle-l tins which in part address areas of reanalysis requested by the UCS petition.

The Director' believes that these programs, a planned NRC review of seismic design of plants not included in the SEP, the conservatisms built into old and new seismic criteria, and the inherent

3 seismic resistance of nuclear power plants render unnecessary the seismic review requested by the UCS.

r

' '[ b' CF The Director has agreed that seismic reantif s is required, and has referred to several NRC pro-However, the Director grams for that purpose.

did not find that the current situation warrants a crash program of review.

Moreover, the NRC seismic review program cannot be accomplished in the short time period suggested because as data is developed it will be used to direct further studies.

The Director believes that this sequential program is the proper approach to seismic reanalysis.

Several factors support the Director's decision.

the almost complete review of Dresden 2,

First, of the SEP, suggests that plants performed as part of this vintage possess adequate seismic safety for a few margins in an overall sense, except minor areas which have been identified for further evaluation.

Second, although initial review of the older facilities in the SEP indicates that some seismic strengthening may be necessary, staff believes that several factors provide inherent seismic safety margins which are adequate to permit continued plant These

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operation until the SEP is completed.

factors are: low seismicity at the plant site; historical records which suggests that older i

large industrial facilities have not been'signi-ficantly damaged by larger than anticipated-earth '"

quakes, and unquantifiable inherent resistance to Third, certain design assumptions" seismic stress.

for older plants were more conservative than those used for later plants.

Smaller damping values were used for piping and steel and concrete struc-tures, and some damping values were limited to ten of the critical damping value.

Consequent:

percent differences in seismic capacity between older and newer plants are less than indicated by consider-ing only changes in the calculation of earthquake induced ground motion.

Finally, as a result of five plant shutdown, NRC has verified the adequacy of many industry computer codes for

4 piping analysis.

Generally, all codes reviewed have been found adequate.

Staff believes that this experience provides a basis for reasonable confidence in the computational adequacy of other computer codes used for seismic analysis and plant design.

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i Recommendation:

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Martin G. Malsch Deputy General Counsel Attachments as stated I

c Comissioners' comments should be provided directly to the Office of the Secretary by c.o.b. Thursday, February 7, 1980.

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Comission Staff Office comments, if any, should be submitted to the Comissioners NLT February 1, 1980, with an information copy to the Office of the Secretary.

If the paper is of such a nature that it requires additional tirne for analytical review and coment, the Comissioners and the Secretariat should be apprised of when coments i

may be expected.

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DISTRIBUTION Comissioners Comission Staff Offices Secretariat l

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W I LLI A m AD MITTCC em utCwsC AN CN*

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th March 25, 1979 em.,,

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Joseph M. Hendrie, Chairman L

John Ahearne, Commissioner j

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v' Victor Gilinsky, Co=lssioner D

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Richard T. Kennedy, Cc=.is sioner

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Peter A.

Bradford, Commissioner U.S. Nuclear Regulatory Commission 4

g Washington, D.C.

20555 l

l Gentlemen:

Enclosed is a copy of a petition which the Union of Concerned Scientists (UCS) has filec this day asking the Co=lssion to order an immediate re-analysis of cperating plants to dete=ine whether they are adequately protected against earthquakes.

In our view, the events surrounding the Commission's shutdown of 5 operating plants because of an error in one l

coraputer code have revealed a serious and nore widespread Preblem.

The methods for analyzing earthquake ha:ards have changed enormously over the years.

Operatine clants were esproved on tha ha d e M methods whFch are unaccentable i

today._ Yet, there has apparently been no ef fort made to I

look back at the operating plants, using today's sophisticated methods, to dete=ine whether they are suf ficiently saf e.

The fortuitous comparison of the old and new Stone and Webster pipe stress codes in connection with the Beaver Valley incident has shown that the difference in results depending on which code is used can be critical.

This is compounded by the fact that the staff has never systemati-callv reviewej the vendors' co= outer codes, but has pe=it-ted their use on the basis of only cursory comparisen with other co_ des.

The petition asks the Commission to impose some order and rationality by identifying current state-of-the-art methods of seismic analysis and reassessing all operating plants according to these methods in order to deter =ine whether the plants are safe or whether modifications are necessary to make them safe.

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Cc==issioners March 28, 1979 Page 2 We will be available at your convenience to assist you in considering the matters presented by this petition.

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Leonard Eickwit, Escuire 2

General Counsel Docketing & Service f

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UNION OF CONCERNED SCIENTISTS' D1ERGENCY PETITION FOR FIANALYSIS OF THE CAPACI'"Y OF OPE?ATING U.

S. NUCLEAR ?OWER FIANTS TO WITHSTAND EARIEQUAKES s

i INTRODUCTION This petition to the Nuclear Regulatory Cc= mission (NRC) f J

is brought by the Union of Concerned Scientists (UCS).

It seeks immediate reanalysis of all operating nuclear power plants, I

using the best analytical methods presently available, to

)

dete mine their ability to withstand seismic stresses.

This l

l petition is prompted by disclosures subsecuent to the Commission's I

action of March 13, 1979, shutting down -five operating plants..

a In summary, it is now clear that, while regulatory standards inad-for seismic analysis have changed greatly over the years,

~I equate review has been undertaken to determine whethe plants' l

i licensed using methods now unacceptable are safe.

The events j

surrounding the Eeaver Valley plant show that the differences l

in results between old and new analytical methods can be critical.

i DESCRIPTION OF THE PETITIONER The Union of Concerned Scientists is a non-profit, public I

e f

I corporation which conducts scientific and technical research i

concerning advanced technologies.

The orga.ni:ation grew out j

cf an informal faculty group at the Massachusetts' Institute l

of Technology in'the last 1960s.

It has grc.r. into a coalition

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s of scientists, engineers and other professionals concerned pri-t i

marilv. with the health, s a f e tv., environmental and national t

I securitv. e. roblems posed by the develcpment of nuclear technolo=v l

in this country and abroad.

UCS has published nany technical

.i f

UCS maintains re.corts on various aspect of nuclear technology.

professional staff in Carbridge, Massachusetts, and k*ashington, 1

D.~C., and has a current public merbership of some 70,000 l

1

s. censors who have made financial centributiens to its work.

l I

JURISDICTICN s

This petition is brought before the Conmissien pursuant i

7 t

4 to the authority granted to it in 4 2 USC 5223 3 (d), 2236 (a),

j 2237 and 10 CFR 552.204,

2. 2 0 6 (c) (1), 50.54, 50.100 and 50.109.

t 4

1 Turthermore, this petition invokes the inherent supervisory authority of the Cc::rnissica to oversee all aspects of the regu-i la ory and licensing process and its " overriding responsibility i

for assuring public health and safety in the operation cf nuclear j

1 l

ower facilities."

In the Matter of Consolidated Edisen Co.

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of N.

Y.,

Inc.

(Indian Point, Units 1, 2 and 3).

C*,I-75-B, I

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NRCI 7516, 173, 1975.

The inherent authority of the Cc= mission has been exercised l

r on a nurber of occasions, desc. ite the absence of express cro-cedural authorization for Cc:c.ission oversight er review in the e

I regulations.

Petition for Emergency and Remedial Acticn, CH-73-6, i

i 7 NRC 400(1978); see also, U.

5. Energy Research and Development Adrinistration (Clinch River 3reeder Reacter Project), CLI-76-13,,

NRC, 76/B, 67, 75-76: Censumers Pcwer Co.

(Midland Units 1 and 2), CI. -7 5 - 2 5, F.AI-7 3-12, 1054.

This authcrity is necessary

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its ndssion to see that "public for the Commission to carry out safety is the first, last, and a permanent consideration in any decisica on the issuance of a construction perrit or a license i

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b to operate,a nuclear facility."

Pcwer Reactor Develoc. ment Coro.

International Union, 367 U.S.

396,402(1961).

t The Commission's inherent authoritv. is exclicitiv recoc.nized t

i in 10 C.ra 52. 206 (c) (1).

10 CFR 52. 206 (a) and 52.206 (b) provide l

a mechanism for petitions recuesting show cause orders to be filed with the Directer cf Nuclear Reacter Eeculation or the i

Direc cr of Inspection and Enforcement, as app:cpriate, and reviewed sua sponte by the Commission.

'devev er, 52. 206 (c) (1) 1 4

i i

states:

i his reviewan

. cower coes not.3imit in any way either the Commission's sur. erviser.v ocwer over t

delegated Staff actions or the Cc==ission's powe-i to consult with the Staf f on a formal or informa' yasis recardinc the institution or croceecings unter this sect en.

i is necessary for the Cc= ission itself i

In this case, it action because the e tition raises bread-ranc.ine. issues e

to take It l

of policy which are not appropriately decided by the. Staff.

I i

is questionable that the Staff would have the authority to grant a

I the relief recuested.

Certain1v the Commissioners would in any I

case have to review and approve the course of action which UCS J

l requests.

In addition, it is the Staff's conduct which has i

largely caused the problems complained of in this petitien.

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3riefly stated, the Co==issien's regulations require a f

i demonstration that nuclear plant systems inportant to safety l

severa histcrical can.cithstand the effects of the =cs:

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Criterion 2; 10 CFR Part 100, Appendix A.

A variety of Regula-t e r v. Guides s.cecifv in c.reater detail acceptable methods for a

Y In many cases, demonstrating compliance with these regulations.

is used to.credict the resc.onse of the nuclear cc. cuter modeline.

plant structures to an earthcuake and to determine the resultant A

stresses imposed on plant structures, s.vster.s and com.ocnents.

In orders to Show Cause issued on March 13, 1979, the i

This action Ccamission directed five plants to cease operation.

was ccmpelled by the discoven of an enc =ous discrepancy be-l Stone & Webster ccmo. uter code tween the results of a cresent h

fcr calculating seisnic stress on piping systems and the results 2

i of a ccde used prior to 1972.

Briefly, the pre-1972 code l

erreneously predicted loads only 1/3 to 1/6 cf the magnitude of a

code.

This cbviousiv.

the stresses predicted by the current 4

casts serious doubt on the ability of those plants which were designed according to the erronecus code to withstand the l

effects of an earthcuake.

In shutting do n the five affected t

riants.cendine. reanalv. sis of all yotentially a f fec.ted.cloine.

systems important to safetv, the Cc=ission took the on1v.

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h prudent and responsible action in the circumstances.

r The facts surrounding the disecvery of this discrec.ancv.

e are still exceedingly murkv., but it ac..cears that a completelv.

f ortuitous secuence of events resulted in its disclosure at Although the Cc=ission has acted swif tly to address l

this time.

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the relatively narrow problem posed by the five plants which used the pre-1972 Stone & Webster code, there has been no the Commission recognizes the larger i=plicatiens indicatien that involved.

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?.to crucial f acts about the regulation of nuclear power i

clants were revealed as a result of the Cc:mision orders and l;

the subsequent Congressional hearings.

First, the NFC Staff f

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f systematic wa.v exard.ne or verify the complex n

1 does not in any J

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connuter codon med in modeline earthquake impacts.

Second, i

as seismic analvsis methods " evolved," no real ef fert has i

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apparently been made to reconfirm earlier licensing decisions usine. the newer and c.resumably improved methods.

Taken to-t facts call ccmpellingly fer a syste.atic reviev j

gether, these i

a of the capacity of operating plants to withstand earthquakes.

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i Of all the testimony that has been presented by the NF.C Webster before Congress on this matter, the Staff and Stone &

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interesting relates to the "evolutien" of analysis nethods i

most i

i for determining nuclear power safety in the event of earthquakes.

w The Staf f described the process as one of increasing the l

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sophistication of the analyses, i.e.,.croc.ressine from hand calculation and static analysis to increasing 1v sc.ohisticated ceneuter c.vnamic mocels.

What this re resents, in essence, is i

a transition from " seat-of-the eants" enc.ineerine. estimates 4

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l to reliance on successive 1v more complex conputer models in-t i

corporating an arrav of changing assu=ptions and simplifications.

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s astounding admission about the uses of computer The most P

l analyses in assuring nuclear plant seismic safety was presented to the Ccmmission by Harold Denton, Directer of Regulation, c

and himself a principal.EC/NRC analyst on seismic issues.

I Webster said l

Centen told the Cc=-.ission on March 13th, " Stone &

we will use this code, we said fine, that s:unds like an a

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We c...ic nc:, co,.,.

se n.n n :.t to get into I

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this area [ issues relating to the technical validity ef the l

tcode) in any of these types of discussions."

s This is compounded by the fact, revealed in statements by Stone & Webster Chairman William Allen before Mr. Udall's Subconmittee on March 19, 1979, that twenty or more computer models have been used by Stone & Webster over the years in doing i

pipe stress analyses of nuclear power plants.

Given the failure of the Staff of the AIC/NRC to check the industry's computations l

s of plant earthquake protection, and given the obvious problems p-.c i

associated with the use of man %v different methc63)%bv individua' t

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the(NRCcannotreasonablybeconfidenNof l

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architect-engineers, khe adequacy of ccdes used in the past or those which are censidered acceo. table tcdav..

even if one were to assume that the presentiv-used

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means of performing seismic analysis are adequate, an enormous problem remains which was dramatically h i c. h l i c. h t e d b v. the past week's activities.

During the " evolutionary" process of i

developing regulatory standards, the NRC has kept its ga:e focussed resolutely forward.

New analysis methods were required for use on new plants; but for already licensed plants, except I

for the very limited work being done in the systematic eval-no systematic checking has been done to deter- _

nation program, f

mine, in light of current knowledge, whether the earthquake

.or:rection of existing plants is adequate.

the evolution has not been limited to the codes

Moreover, To the con-used for determining stresses on piping systems.

rrary, the same sort of development has taken place in virtually all c:her aspects of seismic analysis.

Ter example, current 1

i analysis of the intensity of potential earthcuakes at proposed i

nuclear plant sites varies considerably fre the seismic anal.ysis of a decade or more ac.o.- / The net result of this is t

i that an inconsistent and technically suspect set cf croc.edures

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has been used to determine the degree of earthcuake protection g

I necessary for various nuclear facilities.

It is critical, given the revelations cf the past week, i

that a searching re-examination be carried out to determine whether other nuclear plants licensed by NRC are fit er un it

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for continued operation.

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.l The Staff has consistently avoided coming to grips with I

this fundamental issue.

In the Staff's May 1976 report on the i

unrescived generic issue relating to seismic design criteria (Task A-40), the Staff confidently states, without reference to any technical justification, their conclusion that despite seismic analysis, l

.cotential non-conservative aspects of current "it is believed that the overall [ seismic anal.vsis) is adec.uatel.v conservative."

It must be emphasized that this " bel'ief" has

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not been subject to any convincing confirmation.

That, of course, e

is the purpose of the Task Action Plan.

But even more glaringly absent than a cogent defense of r*andards is an explanation of the basis for allowing present l

licensed under presently unaccectable standards, existing plants, Consultants to the Advisory Cc=mittee on Reactor Safeguards

• /

"have recommence tnat, in view c:. t.ne uncertainties c:. knowl-1 edge concerning the sources of earthcuakes in the Eastern (SSI) cf United States, a minimum saf e shutdown earthcuake O.2; acceleration should be utilized fer new plants f or which applications are submitted in the future." '

ccnstruction permit

(~~etter to Marcus A. Rowden, NRC Chairman, from ", 3 ender, ACRS Chairman, January 17, 1977, p.3).

By way of contrast, considerf

,-- vMeh received its ::eee ine license in,

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to centinue in co.eratien.

Althouc.h the description of Task A-40,

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The Staff does not even venture a "' relief" with respect to the level of safety of all other w e -.=.4...' v c... c. 4 + u *. e a... a-4 c

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available seismic analysis nethods, ac. olv them to existin-v and deternine whether the plants neer what is nucle ar plants, e,a #e as .e...4.7.4.",. d e _e.d -. s c.da-ds

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changed and refined in future years, it would neverthe-may be

" state-of-the-guidance based on the current less provide scme art" for deter.ininc. whether or not a seisnic safety crisis exists in any nere than the five plants that MC'has shut down.

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calculations _ werA_.nc4--subject tQiled technical review by 1

the Staff.

Indeed, the Staff has not claimed to have even superficially reviewed any other aspect cf the methods used to I

9 jude.e adec.cate earthquake protection.

The Staff nerelv. took h

them to be satisfactory because they " sounded like" accredited I

methods.

Moreover, the effect of an earthquake on a nuclear l

c pcwer reactor extends f ar beyond the ef fect on the piping l

4 The electrical components of the safety system would j

system.

I also be affected bv the earthquake.

Acccrding to the Cc =ission's i

I internal files which we have reviewed, this aspect of nuclear 1

plant safety protecticn is very much in doubt.

Another aspect of plant respense to earthquakes that has been larc.eiv. cvericcked e

in the discussion of these five plants is the elenentary question cf equipn. ant cperabilitv..

ECCS performance, fer example, is not assured merely by the fact that the ICCS pipes renain intact e

following the earthquake; whether the ECCS pumo.s and valv.es will remain operable in bent or deformed piping is also an l

1 l

issue o., carc..i n a., amportance.

.-mE: REQ 1.

T D ass z The Cc= mission has been encouraged by the Staff to focus its attention on a very narrow questien.

However, the window nas caen opened on a complex an, breac-ranging pres.,em.

In crier to assu-e the required degree cf public protection, it is necessary to bring some order and rationality Oc the issue of seismic analysis.

We propose that the Cc= mission require elants with an cperating license, wichin 120 days of the 1

t..

I

.Ccr.-ission order, to perform a seismic reanalysis.

The Cor:aission should specify the methods of seismic analysis to be used.

The Ccamission should base its selectica of ac.crocriate methods in tems of what the Commission staff determines to be the present i

" state-of-the-art."

The reanalysis should include, as a minimum, all cf the steps outlined in the Standard Review Plan, Sections 2.5, 3.7, 3.9, and 3.10.

To summarize the majer ccmponents of this reanalysis, the followine, should be covered:

+

i 1.

Define the magnitude er intensitv of the earthcuake i

wnich W1J.1.croduce the maximum virratcrv c.round motion at the site (the saf e shutdown earthcuake or SSI).

l 2.

Determine the free-field ground motien at the site that would result if the SSE cccurred.

t I

3.

rete mine the motion of site structures bv modif.v-t ing the free-field motion te account fer the interaction of the site structures with the underlying foundation soil.

}

4.

Dete mine the motion of the plant ec.ti.onent succorted i

d by the site structures.

5.

Ccmpare the seismic 1 cads, in appropriate ccabinatien with other loads, on structures, systems, and components important to safetv, with the allowable loacs.

i t

a 6.

Inspect the plant to determine whether the "as built" plant conforms to design specifications.

Submitted by the Union of j

Concerned Scientists, i

i tv -ke:- A'+

v J

~

~

cll.vn n.

Weiss Sheldon, Harmon,.Roisman & Weiss 1025 15th Street, N.W.

Washingten, D.

C.

20005 s

(202) 533-9070 l

I hereby affi m that the facts alleged herein are true and correc:

i s

a i

my.newlec.ge anc cel.g;.

the best o:.

r.

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DGEisenhut ARosenthal, ASLAB f2 RHVo11mer ASLBP Ellyn R. Weiss, Esq.

JFelton(3)

Sheldon, Hamon, Roisman & Weiss BGrimes 102515th Street, It.W.

LCShao stRn J

TCarter ACRS(16)

Suite 500 CHofmayer VNoonan Washington, D. C. 20005 JShepaker KHerring I:

k CMiles, OPA JLieberman

Dear Hs. Weiss:

Central Files RMattson

[

Attached to your March 28, 1979 letter to the Comissioners was a petition to the Comission filed by the Union of Concerned Scientists (UCS).

Their peti-T tion has been referred to me by the Ccanission for response in accordance with

[

10 CFR 2.206.

The petition proposed that the !!RC require all plants with an

[

operating license to perfom a seismic reanalysis within a 120-day time period.

[l Upon review of UCS' petition and other relevant infomation, I have determined F

not to order a seis::ic reanalysis of all operating power reactors as requested i

for the reasons given in my en::losed decision.

I' A copy of this decision will be placed in the Com.ission's Public Document P.oom

~

at 1717 H Street, it.W., Washington, D.C. 20555.

A copy will also be filed with the Secretary for the Comission's review in accordance with 10 CFR 2.206(c) cf f

the Co nission's regulations.

)

Sincerely, c.ti. :' T-4 ti.

l

' H. ~.. -.::: l 1

Harold R. Denton, Director Office of fluclear Reactor Regulation

Enclosures:

Director's Denial Under 10 CFR 2.206

otice cf Issuance I

AD: DSS

• f4RR*

D:DDR?

in,

JKnight FSCHROEDER DEISEHHUT

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( HDEt. A 1/ /SO 1/

/80 1/ /30 1/ /b /00

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-SEE PREVIOUS Y(LLONS FOR C0!!CURRENCE y <l'\\

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NUCLN1R32bbLkibRh C0dil5510N '

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'L 0FFICE OF NUCLEAR PIACTOR REGULAT10N IIe A

A HARDLD R. DENTON, DIFICTOR

~

4 peccEo g

ususo In the Matter of i

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N4' g G80 >

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cieit[ittS$*"fp PETITION FIQUT. STING SEISMIC

)

(10 CFR 2.206)

Ddt t

REA*MLYSIS

)

G DIRECTOR'S DENIAL UNDER 10 CFR 2.206 g

N I

~

I.

Cn March 28, 1979, the Union of Concerned Scientists (UCS) petitioned the Cc=issicn to re;; ire all piants with an cperating license to perfom a seismic reanalysis within a 120 day time period.

The major features of the requested reanalysis would involve:

(1) the magnitude of the Safe Shutdown Earthquake (SSE); (2) the freefield ground motion at the site; (3) the motions of the struc;ure during a seismic event; (4) the motion of the plant equ'pment supported by the site structures; (5) the seismic leads on structures, systems and componen9 in a::repriate combinations with other loads, and the corresponding allowable leadir.gs; and (6) the confor ance of the "as-built" plant to the design specifi-ine petitien was referred to the Director of the Office of Nuclear catic.s.

Reactor Regulation for response in accordance with 10 CFR 2.206 and noticed in 16,1979 (44 FR 28737).E the Federal Recister on May Orior to the date of this petition, on March 13, 1979, the U.S. Nuclear 1/

segu;atory Comnissier. (SRC) issued imediately effective orders suspending Beaver Valley Unit 1,

erations of five ruclear power reactors, namely:

Surry Units 1 ar.d ".aine Yankee and Fit: gerald.

In each case the licensees

<.ere crdered to s?>: cause:

',1 ) why they shoulc not reanalyze facility oiping systems for seismic icads on ali octentially affected safety systems using an accrocriate piping analysis commuter code which does not combine loads algebraical]

',2 )

why they sho:.;ld ne: make any modifications to the facility piping sys-te.ms indicated by such reanalysis to be necessary; and

3) eny facili y creration should not be suspended pending such reanalysis.

and comple:i:. of any recuired modifications.

\\

i i

II.

ESIGN RE0'JIREME'iTS A.

Current Seisnic Design Re:uirements Currently acceptable seismic design requirements for nuclear power plants are generally delineated in 10 CFR Part 50.55a, Appendices A and B of Part 50, See_ also, U.S. N.R.C. Standard Review Plan Sections 2.4 i

Appendix A of Pa.rt 103.

i and 2.5, and 3 (excludin; Sections 3.3, 3.4 and 3.5), with their associated Reculitcry Guides (e.3.,

.e,. -u1d.s 1.12,1.26

• 1.28,1.29,1.38,1.48,

1. 5',1. 5^, 1. 61, 1. 7 2, 1. ~ ~, *.100, 1.122, 1.124, 1.142, etc.) and the re-i todes and s:Inda-cs (e.g., ASME, ANSI ACI, IEEE, AISC, etc.).

These l

f eren:e:

seisr'; design recuirerer.s dell with the entire seisnic analysis / design chain 30.- the defini;i:r

• e seisr.ic hazard at a site through the j

analyIis, desien and :or.s r<.::'on/f tbri:Etion cf safe:y related structures, i

s.vsters, e;uipment an: c en: :ne.:s.

These requirements are briefly summari-l

ed otIOW.

.e seismic ha:ard (f.E., the eartho.uake induced ground motions at the site) f i

i l eviden:e.

I:

i

etermined c- :hi : as's of historical and geolog ca is #i-::

i c ar;htuake levels; namely, the Operating Basis i

is ce e: i n :eres c# :w:

which could be reasonably expe:ted to affe::

Ezr:h:_ake (DEE) whic is : a:

site during t'r.e ::-eriting life of the plant, and the Safe Shut-l the :'.ar 1

a ;rauake (SSE) whi:t is based u:.on an ev.Ivation of the maxirum ear:n-d:wr c.aks :; e.tici for ; e s' e.

Earthquake hazards are nonmally expressed as a 2/

fun::':

apnitu:e a-

5:1.:e #rer the source or intensity at the si;s. --

s 3; ' ;de is inti: :i.e :# he inerny release asse:iated with the earth-

t

.e ir.:e s :y is indicative of the local damage

3. s 1

e s:ur:e,..

ass::'! 5: with the ei. :.1 :1.

i s.acnitude of an ear:r.:;ake is cer=:nly defined in terms of the Ricr. er i

ar. earth vake is cormonly defined in terms cf 2/

E:g.e and the in sr.si:3 :#

~

bjective "::i'ie: "s-:Ei".i S cal e.

-3

~--v

[

\\

x i

? ess.t day requirements for determining the SSE can be found in Appendix A l

t: I D F?. Far: 103.

In it the required regional geological and seismologicti l

t ir.vestigatiens are described.

When known earthquake generators such as capable j

r fau!:s are identified, the regulatiens require that the Safe Shutdown Earthquake t

be de:arti.ed : r.sidering both historic and geologic history.

b' hen earthquakes t

canr.o be :::Tilated with faults or tectonic structures the Safe Shutdown Earth-l 1

q;aP.e is citarnir.ed assuaing that the largest historic earthquake in the sine te:::ri: : : rirce could re:ur at the site.

A te::enic province is a large geo-gra:h': ri;i:n of sicilar ge:logi: structure.

Although these regulations te:Eme effe::ive ir. ;+:emier 1973 they were to a large part based on the practice prior to thIt di e.

During that tice safe shutd:en etrin:vaie (or " design earth:uake") ;

desi; gr:;nd rotien were adopted based upon geological and seismological recc -

l nenca:icts :1 the U.S. Geological Survey ar.d the U.S. Coast and Leodetic Survey, l

}

and e ;irserin; re:camendatiens from prominent earthquake engineers such as Dr.

l Nit.r.a-M r.a rk ar.d Dr. John 51ume.

i

-.s same sarthevake r.acnitude, the detailed nature lof the crcund I.a - ;

"I :.":e cif ferent fr:n or.e earthquake to ar.0 her.

There are s ut s ir. i!

v!-iations in su:h parareters asse:iated with the cround i

~

i r:: :

as : cat. a::eleration, peak velocity, peak displacement, duration, l

i r.a!. i at: E t ;y :cnter.t at various frequen:ies.

Due to these uncertain-

' E s, - e ; :. :

i:n a: a site is defined by a snoothed, free. field

-es :: It ::t::.

.itr a share inten:ed to have c. :lification f actors for i ;

i

51-3::tiera-icn :orresoond'ng ic a near. :ius one standard evia.

,.,. g 3.

l

.z..

I s

o

l

-(

(_

J/

i In evaluating a plant for a given definition of the ground motion, a e

c'etailed er.gineerine evaluation is conducted considering three directional

(

l Ercurd rction, foundatien-structure interaction, structural response, piping system response, equipment response and componen: response.

The

" uncertainties in the various steps of the overall analysis and design lead to conservative assumptions,being made in each step regarding such i

c. arace:ers as load cercinations, material croperties, allowable stresses an: :ampin;.

For tne :.o levels of eartnquake, :ne cesign and analysis l

arare:ers are specifiec sucn inat, in general, stractures, systems and cong:nen s are cesignec to remain in the linear rar;e, well Delow yield, j

i fer ne C3E, and near er semewhat toove tr.i linear ran;e and yield, yet l

)

suos antia',1y belcw :neir ultimate capability, for :ne SSE such that the l

capa:ili:y to shutcesn :ne plant and to maintain the plant in a safe shut-J co<n cenci ion is ensurec.

i It has been our experience in evaluating some-of ine older. seismic designs

na wnile :ne geological anc tectenic analyses have ne changec racically

ners r. ave Deen larger enanges in the way we cnaracterize the grounc motien This is due asso:iatec with an eartnquake of a S ven magnituce er intensity.

i to t.e availability cf core data, anc greater in ceptn systematic analysis Presently, practice woule usually result in ci s:ren; notion recores.

s rc ger assumed motion than previously stipula ec for earlier plants.

how-evaluating these design mo-ions all the engineering assu :ti!

ever, in addition ::

ust be a ;en into account in evaluating.the overall..sei.smic, design.

Certain det asse.:-i:n a_ssoc.iated.with_these earlier-plants wert_more_conserva-ive so that.

differen:ss between trer and present day plants are less than the se inalesis a-:ve would indicate.

I 1

t 5

.L CHRONOLOGY OF SASIC SEISMIC DESIG$ REQUlREMENTS S.

The basic seismic design requirements have undergone many changes over Prior to 1950, there were no specific approximately the past 25 years.

Since requirements other than those containec in local building codes.

that time, the development of the basic seismic design practices can be generally summarized as follows:

i PRIOR TO 1950 - Uniform Builcing Code Requirements

- Static seismic coefficient applied to structures

- Grounc motion cescrioed by Housner's averaged 1960 - 1964 grounc response spectra.

l

- Sincie cegree of f reecom systems were used for tne'evalua: ion of seismic, responses.

- Horizontal anc vertical eartnquake responses were no: ccmbi ne a.

1965 - 1967

- Ground motion cescribed by Housner's averaged l

grounc response spectra (in some cases housner mace revisions from the previous spectra).

~

- Multi-modal two ciEensional mocels were used The for tne evaluation of seismic responses.

response spectrum approach was usec most often.-

Time history was usec occasionally.

- Damping values were taken as 0.5% for piping.

1% 1/25 for steel structures, anc 41 1/2; I

f or concrete structures.

- Compliance (flexibility) for p1 Ant foundation l

medium was considerec.

- Sum of the EOsolute value of the responses arts 1ng t

from the largest horizontal anc the vertical ear r.-

quake was generally usec for response ceterminatien.

i h

4 D

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- (~

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- Ground m: tion described by Housner's averaged 1967 - 1971 ground response spectra modified, especially in short perieds, using t eur. ark criteria (known as rodified Newcark spe:tra, 1967 - 1969).

t

- Soil structure interaction effects were con-sidered using discrete soil springs and in some cases assucing material damping.

l

- Floor response spectra generated and used in l

i the evaluation of equipment and piping.

1971 - 1973

- F.odel danpine values.for the soil-structure system to represent contributions from both material ar.: raciation campir.o limited to 105 of critical camping.

1973 - 1977

- Reg. Guides 1.fC and 1.61 we e in:roduced te define grount res ense s;e::ra, and darping values (for stru ures; sipir.;, equipe.ent and components), res:ettively.

- Camping fer smati and large piping was raised to 2% ar.d 2%, respectively.

- Soil dam:ing determinations were required to j

account for the nonlinear stress _ strain relationships fer tne foundation nedium.

- Finite ele..ent pr :edures were reauired in the calcul ation Of soil-structure interaction I

for deeply enbetted structures.

- Three con;cnents of earthquake retion were required to be censidered by taking the SRSS of the responses to each corponent (Reg. Guide 1.92).

- Floor response s:ectra generated per Reg. Guide 1.122.

t AFTER 1977

- Layered se'is a:: unted for in ar. elastic half space s:il-s ructure.intera: tion analyses.

- The limit of 1 t of critical car;ing on modal danping values '

soil-s:ru:ture _ interaction analyses sas re..:ved.

e

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Equipment qualification oer Reg. Guide 1.100 Comparison of elastic half-space and finite element soil-structure interaction analyses results.

C.

CDNSERVATISM5 INHERENT IN THE SEISMIC DESIGN REQUIREMENTS In today's approach many conservatisms are introduced in the various stages ;

i of the seismic design process.

These conservatisms are briefly itemized as f

follows and would be applicable to different vintage plants, including the older nucieer pov:er piants, in varying degrees:

l 1.

Conservatisrs associated witr : e seis::':

f :ne design event.

l w'-' ::nse vative ar;1ifica: ion a.

Wide band ground response s:e: r:

factors.

i The ground response see::ra used as '.:ut are smoothed, and broad The spe:tra for a eil ear:n: ake are jagged in nature,

+

banded.

producing less response in :ertair. fretuen:y (or period) ranges of the 5;e:tra than in adja:er. fre:ve :y (or period) ranges. 'ite '

spectral arolification f t ::rs are :E*.ernined #r0m considerati:ns of :ne see:tra for a set :f real it.F: vites.

In the case of the devel c:rer. of R. G.1.6', -he an:'. * .:a icn f ac ors at each frecuen:y were based on ::rsi: era-i:" ;f a:out an E4 percent con-fidence level that the res;:nse a. 2 :ir-i:ular frequen:y wou',d not be exceeded.

i b.

Envelccing synthetic time histories.

In the cevelo; ment of seisr' res:enses for the design of str :-

i I

tures, systers, equipment a-d cer:: er s, syr.tnetic earthquakt tice hist: ries are devele:et with *es;;rse see ra that essen-izily envel:: :te ground desigr s:e: ra.

L I

i

I i

1 c.

Conservative OBE.

Seismic design criteria are such that the 03E, rather than the SSE, i

can control the design of certain strue:ures, systers, equipment Those iters for which design is controlled by the and corponents.

OBE have a capability to resist an SSE with targins greater than those intended in the SSE design criteria.

Conservatisms associated with the methodologies for seismic analysis and j

2 design.

t Conservatists for structures, systers, and ccmponents.

.a.

1.

Dynamic analysis.

i Elastic cynanic analyses are perf;rmed using low danping values and time-history or rescense spe::rur analysis methods.

In modal nodes are combined by response spectruc analy ses, cicsely 5:2:e:

absolute summation.

i 2.

Soil sited structures evaluaticr..

Soil site structures a-e evalut et usir; :enservative seistic inputs into soil-str.::ure intere: i:r ar.alyses.

3.

Three input corpenen s.

Three input corponen s of an earthcuzie (2' horizontal and 1 vertical) are cor.sideret.

Bcth rori: r:a1_earthquate l

components are assurid ic be ecaal.

4.

Loading comb' nations.

i Loading combinations c:nsider other icadings (e.g., cead weight, load-live loads, ;ressure 1: ads, etc.) in aedition to the seismi:

ings.

Seismic leadir.g is only a par: of the total loading and l

i in fact, other loadings besides seismic tay in cases govern design.

I b.

Ef fect of inelas-ic beha ti:r.

In reality, *: ell encinet e: s ructures, Core:nents and systens are capable of sustainin; ica:s which are bey:-d these which would tring then to their elastic li '

withce; sustai-ir.g car.29e.

For sna:1 i

excursions inte -he ine15stic range, seisric inertial loads are reduced as e fun:: ion of : e amoun cf ineitstic action in cor:arison,

i r

i t

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L with those calculated elastically.

This phenomenon can be con-sidered by the use of a ductility factor which is equal to unity for purely elastic behavior and increases with increasing inelastic behavior.

For example, a ductility cf 1.5 would have the effect of reducing accelerations of elastically. calculated response spectra by as much as 1/3.

here ductility is defined as the ratio of dis-placement level in the nonlinear range to the displacement associated with the yield point for an elastic / perfectly plastic resistan:e vs.

dispiacement fun: tion.

Conservatists for' ele:trical and mechanical equipment.

c.

1.

Peak widening of floor response spe::ra.

uken the 90:r response spectra are :eveloped for the cesign of

nese com:enents located at differen: locations in the structure, to account for un:ertainties in the Enalysis the peaks in the individual floor reso:nse spe::ra are Orcidened in order

predict conservative equipment responses.

2.

Use of maximum response spe :ra for n.ltiple supported systers.

'Jhere the system has multiple su:per:s, the maximum response spe:tra are generally ap;1iet c all su; port points so that con-servative seismic loads are generated for design purposes.

3.

Multiple applications of damoing values.

In calculating the seisni: loads for these components, darting values are a;;1ied tv: ice (first, t:

ajor stru:tures ar.d tren to the equipment).

The cultiple applications of the c:nservati.ely low damping values compounds the corservatisms in the seisr.ic responses which these items are designed to resist.

4.

System P,edundancy Even identically designed redundant systers may not always experie similar seismic excitation due to different mounting 1::ations, wi different structural filtering effe:ts.

Thus a single loss f re.

dundancy may not mean a loss of fun: tion for the syste..

This :r:

vices additional assuran:e that a plant will safety wi;hstant a sa event.

9

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l.

l l

i i

l d.

Conserva:ists in the qualification of electrical and mechanical equipment 1.

Required response spe:tra.

I The required test input is norra11y defined as the envelop cf i

floor response spectra cbtained using structural analysis methods. This ensures that the recuired response spectra are conservative.

2.

Test response spectra, i

The test spectra nust envele: the eeuired response spectra.

j 3.

iest for nul:1-plant appli:ation.

j The equipment suppliers cenerally test the ecuipr.ent fer cc -i-plant application.

Censicera:le rir; irs are added to ne tis:

res;onse spectra so ina:

ey are i::14:able to many plants with differing seismic receireren:s.

6 4.

Multi-axis testing.

The tes: input motions should :e a:; lied to the vertical an:

l the horizontal axes simultaneously unless decoupling of res:anses along two dire:: ions is justifiable.

i S.

Tes: for DEE and SSE.

1 A number of CEE tes:s are :erf:rne: ;rier to the SSE test.

The number cf CEE tests is conservatively selected to re: resent l

the u;;er bound icr a :T an; site.

Inis provides an adcitio al i

margin in tne consideration of cyciic loading effects.

3.

Conservatists in the stru:tural and techanical resistance.

t a.

Allowable stress limits.

Engineering codes spe:ify *:oce r.ininur. strength" for materials.

These codes tinitur; strengths are in turn s;ecified by the app 1':an:

when the tz erials are creered; any r.a:erial found to be under : a; streng:n is ref e::ed.

The esult is :te the caterial su:;iier :rovices!

ta:erial of higner streng:h.

• ls:,

a gins exist between allcw1:le stresses and ultirate strerg:hs.

b 4

e i

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h

l.

({

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28 day con: rete strencth (structural only).

b.

f Designs are usually based upon the 28 day design strength of Concrete continues to gain strength with in:reasing concrete.

Additicially, the strength at 28 days often l

tire beyond 28 days.

exceeds that called for design strength.

t Static strength vs. dynamic resistance.

c.

Since Code material strencths are based upon static load tests.

dynamic loads contain a limited amount of energy and are a: Plied at a faster rate, the r.argin between stress limits and failure for dynamic loads is greater than that-for static loads.

d.

Standard size structural nerbers and pipes.

The design of the s*ructural elements is such that their capacities Much usually exceed the re:virenents :211ed for by the analyses.

by the availability of the actual structural design is contr:11e:

of standard stru:: ural ned:ers such as beans and piping se:: ions, so that larger si:es than are neeced are eften used.

Redundancy in indeterminite stru::ures anc cenponents allows for e.

redistribution of loads.

Frem the standpoint of function, r.ajor structures and cor::nents can tolerate ruch deformation, and typically failure of numerous struc-Inis defonration and loss of structural renaers can tural members.

l be sustained be:ause of reduncancy, (i.e., r: e than one path available to carry icads) whi:h z' lows f:r redistribution of loads formerly carried by f ailed rerpers.

f.

Du::ility to f ailure.

In deforming to failure, beyond the elastic limit, the inelastic behavior of well engineered concrete and steel structures, components and systers provides for energy absorption not norra11y ce;nted on in -

The effe: s of this are discussed in detail in item IV.B.b.

1 design.

Minor attachments absorb energy.

g.

Nonstructural ele. ents which are not considered to carry ary 1: ads in design, de absor: energy irr ugh inelastic benavior or :ella:se

~

during a seismic event.

h.

Nuclear quality assurance (IT. :rograr..

han nost found throuch The nuclear QA procedures a-e rcre strincen

~ is cr:vides additional safe:y for ou: the constru: ier industry.

n nuclear plants bey:nd tha: 00.sidered c:ce;;aole for mos: nennu:l ear e

1

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(..

i facilities designed using r.any of the same practices as J

used for nuclear plants.

These conservatisms are difficult to quantify; however, the extent of these structural and mechanical conservatisms for plants designed using current standards has been estimated by studies made by Newmark, and Corne A r.edian factor of safety for structures, equipment and piping has been estimated to be within the range of 4 to 8.

For older facilities, it is rs:conized that these factors of safety would be somewhat less. Ongoine seismic p.0yecins, which are discussed later in this document, will provide bs:ter insight as to what these factors are likely to be.

D.

OTHER CONSIDEFAT]DNS Fer comparison, hospitals, schools, a;artment complexes and similar l

eisential facilities are designed by curren: r.cn-nuclear criteria that for the same earthquake exposure in tents of ground acceleration result in designs i

several times less conservative overall thar. current nuclear plant criteria w:uld cictate.

[dhitional substantiation of the inherent seismic capability of structures, systems, equipment and components is found through the examination of the l

This inherent capability is performance of structures in past earthquakes.

nit always due to a conservative seismic design, but to the fact that the design l fcr loadings other than seismic (e.g., wind, pressure, etc.) leads to ar.

irriicit level of seismic resistante.

Explicit consic:eration of seismi:

l i

"Or. the Seismic F.eliability of S; clear power Plants," C..A. Corneli If anc M. M. Sewmark, May 1978.

j

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t t

Specific examples of the performance loadings increases this resistance.

i of industrial and fossil power facilities in response to real earthquakes i

to illustrate these points are cited below.

l The oil fired Kern County Steam Station ih California (designed and built I

in 1947-8) had structures designed for 0.29 static coefficient with stress l

i limits increased by 33% for combined dead, liYe, and earthquake loadings.

l l

Piping systems were designed using static coefficient hand calculations i

S/ smoothed response spectrum (narrew ar.d heavily techniques and the Biot f

damped compared to those used for nuclear plants) with peak accelerations l

of 0.1 9 at the ground level varyin; iinearly 2: higher levels of 0.39 at Ecuipmen anchorages were reviewed for lateral the top of the structure.

The plant operated through the July 21, 1952 Kern County load resistance.

The peak ground earthquake (Magnitude 7.7) with no significant damage.

acceleration at the site was estimated to be about 0.25g. 5/

l l

During the 1971 San Fernando Earthquake,'the fossil fueled Valley Power Plant, which was cesigned to 0.2; - 0.25g, was not damaged althouch accelera-Other nearby l

tions at the site were estimated to be in excess of 0.259 1

power plants which were not as close to the epicenter as the Valley plant l

were aise undamaged. 6/

1 Siot, N.

A., Analytical and Experimental Methods in Engineering Seismology,j

--4/

Trans ASCE 105 Pg. 365-40S,19a2.

" Seismic Capability of Nuclear Piping " Robert L. Cloud, May 197g, 5/

(" Report on the Reanalysis of Safety Related Piping Systems - Surry Power Sta: ion, Unit 1 - Virginia Electric and Power Co.." Appendix F, June 5,1979:

Ibid.

However, the San Fernanc: Power Plant did experience a structural i

f ailure which led to a penstock failure, however, it was' built in 1921.

5/

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i

An ESSO refinery was subject to measured peak grcund accelerations of e

t 0.39g E-W and 0.34g N-S in the December 25, 1972, Mana;ua, Nicaragua earthquake (Magnitude 7.5). The design of the refinery met provisions of the Unifem j

Building Code for Zone 2.

There was almost no damage te the refinery which i

resumed operation 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after it was shutdown for inspection.

Also, the fossil-fueled power plant in Managua, imediately adja:ent to the causative fault, and for which the design basis is reported to teve been 0.lg, probably experiented accelerations on the order of 0.6g' and su#fered some damage, yet was cne :f th: fir-t ir.'ustrial facilities to return to operation following the earthquake.

Many of the problems were caused by a:sent or inadequate

~

anchors. 7/ b/

l The Chugach Power Plant in Anchorage, Aiasda was s.: ject to accelerations of approximately 0.2; at the site during the 1964 Alas <an earthquake of Magnitude E.4 The design of the plant was based on a C.lg static coefficient, c

yet there were nc power piping failures. L, On June 7,1975, the Humboldt Bay Nu: lear pcvier f.cnt 'experien:ed an earthquake with peak measured accelerations in plant structures of up te 0.35g. l The duration was short, therefore, the energy was limited in comparison to that which is implied by anchoring a design spectrum at this valve. However, the i

damage to the facility was insignificant. The plant was shutdown for refueling ;

4 at the time and there was no damage to safety systems.

Review of fossil power plants that were shaken by the Alaskan eart.puake, i

and of fossil and nuclear plants shaken by earthcuake.s in Japan during -.e recen

_7/

Ibid.

l 8/

"Managua, Nicaragua Earthquake of December 23, 1972 " Earthquake En:ineeringl Research Institute Reconnaissance Report, May,19'

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i 9/

" Nuclear Plant Siesmic Margin," Robert L. Cloud, Ju.e 8,1979.

.(

l l

experience in Fukushi..a where the nuclear power plant operated right throug i

r the strong motion, fr-ther demonstrates the point that carefully engineered structures, piping and equipment of the types found in the nuclear and the fossil pwer generation and the petrochemical industries, typically possess i

high resistance to seismic forces.

\\

l E.

CURREhi IE BtT_LETI!3 REMRDING SEISMIC ISSUES FOR OPERAT 1

Recently, several IE Bulletins regarding seisr.ic issues have been issued These in par:~ acd-ess a-eas of reentlysis to all licensed power rea: tors.

Tne subjects of these bulletins are su..trized T

re;uested ty the UC Atition.

as follows:

This Ev'le:'n recui-ed each licensee.

IEE 79-02 for the s;;;cr: base plates which are (issued 3/8/79, anchore: using -he :encrete expansion revised 6/21/79, type an: hor bol s, to verify tha; a supplemented 5/2?/75 proper factor of safety on cesien loads and revised 11/E/79) exists considering the flexibility of th base plates and tne cyclic nature of the loadings.

• dditionally, a :est procram requiret to verify :he adeqcacy of the i situ installation e' the an: hor bolts.

requiret nodifica-ions cust de cede.

This Bulletin reovired each licensee to IES 79-04 verify that the current weights for cer$

(issued 3/30/79; Velan swing che:k valves were used in t#

l seismic analyses and design of piping k'here dis:repancies are founde l

systems.

the affected piping systens must be re-

~

evaluated and any modifications performs ihis Bulletin recuired each licensee to IES 79-0/

cetermine if the seismic analysis cf an (issuec4/11/7?:

safety related pioin5 syste~s wire base upon the inappropriate algeorai: :omoir tion of responses to differen; etrtn:va components.

For any that were, the sys were re:uired to be reanaly:ec csing aa appropriate computer code wnich,could b verified by the NRC.

Any re:vired ::di cations cast be performed.

-(.

L.

I This Bulletin was issued as a result of IE3 79-14 the finding during the review of (issued 7/2/79, revised 7/18/79 responses to IEB 79-07 and the re-and supplemented analyses of the five plants which were initially shutdown by Order that certain 8/15/79)

I piping system and support as-built configurations differed from that assumed in the analyses and the designs.-

This could result in substantial changes in piping system responses, and piping and support stresses.

Therefore, this Bulletin req ired each licensee to ver-ify that the.as-built piping system, including supports, are essentially the sar.e as assumed in their seismic I

analysis and design.

L'here significar.t l

discrepancies are noted, the effect on I

the anlaysis and the design rust be evalue:ed and any necessary modificati:ns must be perf ormed.

As the reviews cf the responses to these bulletins proceed, the liRC will take such actions as may be necessary to assure the public health and safety.

The reviews of responses-to-date indicate that some installation and design deficiencies exist in the-creas addressed by these Bulletins.

Tnese dcfici.encies are beir.; resolved in a timely, prudent manner.

Affected licenseeg are co :-itted to takinc appropriate remedial action.

If necessary, the staff will take enforcement action to ensure changes are made.

1 IIJ. OtiG0ItG tiRC SEI5M:C ORIENTED PROGRI?.S There are currently four major ongoing seismically oriented programs withis

-he tiRC; namely, the Systir.atic Evaluation Program (SEP)

Task Action Plan A-40

', TAD A 40), the Seismic Safety Margins Research Program (SSMRP) and the Code

'lerification Program (CYP).

These programs are in addition to the reviews of c:erating plants with regard to the items identified in the recently issued IE I lletins.

. (. _

(

i F

A.

Systematic Evaluation Program (SEP) Review i

l The SEP was conceived by the NRC in 1975, a plan for it was defined in 1977, and it was implemented in 1978.

A major effort of the SEP is an evaluation of the seismic design adequacy of the eleven older nuclear i

r

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The SEP facilities f

power plant facilities under review in the program.

i received construction permits between 1956 and 1957.

Seismic design

[

i procedures evolved sienificantly during this period and through publication l

t of the Standard Review Plan (SRP) in 1975.

As a result, the seismic design t

I bases of the SEP fa:ilities vary in degree from Uniform Building Coce considerations (static analysis) up inrcugh and at: roaching current l

i

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L standards (dynamit analysis).

)

Recognizing this evolution, the NRC nas fosnd it necessary to take _ an assessment of the seismic design safety of the SEF facllities relative to I

those designed under current standards, criteria and procedures and to I

make an integrated evaluation to verify that these facilities possess l

~

acceptable levels of seismic resistance capability.

e I

Tc reach these findings the SEP seismic review must utilize te:hnical j 1

approaches thoucht more realistic in light of current knowledge rather than those dictated by current requirements which are felt to yield l

+

conservative desiens when considered in an integrated manner but 1

do not necessarily produce an accurate representation of the true seismic response.

':aving re:ocnized an: considered in more' detail the 4

inherent capabilities of these f acilities, a ce:ision will be made

)

regarding the nee: to retrefit.

I must be emchasized that if su:h i

l ne:essarily imply that the an eventual decision is made, it coes i

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r i

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k-t existing facilitie are unsafe but rather that syistantial benefit to the public health and safety can be attained through such actions in accordance with 10 CFR 50.109.

If however, during the course of review significant deficiancies are found, appropriate action will be taken by the <

NRC.

The SEP facilities fall into two groups based upon the degree to which The li:ensees of the earlier seismic design was cri;inal-ly considered.

facilities

.ase en:arked en seisnic re-evaluz:ien programs of their l

SE) for the most part far

• 5 to su;;1e.e.: :he existing data base which is These programs less rigere.s:y di/elc:ed the, would te ex:e::ed :: day.

4 are being cesel: ped su:h that iney are cen:reher.sive enough to provide assessment of the v

r.e staff wi

. su'ficien; data to enable an : veri:'

seismic safe:/ of these fa:ilities.

The NRC staff is currently reviewing the original seismic design docu-l nentation of -he 'ater f acilities.

In sene cases, Fe existine inforr.ationi J

4 nas :een sc:: ese :ed :y NO.C skudies to veri #y stiff'judgements.

All of i

J tise plants na ve eer visite: :o date by specia'ly s:cffed seismic review hand knowledge of facility geenetry and to visually

sans to gair fir-identify any obvices an:calies.

One such review of the Dresden ? facility is nearing completion. This review has : :w ided v11 cable insight into the seisr.ic designs of sicilar v'r. age faci'i ies, itsed u:Or. initial juc;ener. an: an extrapolation the o-i- f a:U.i-f es, it would a:oear that -he later SEP f acilities

. r:..: cotior. input has n:: es:a'.z:ed significantly, e ere the se' sri:

ssess, ir 1- :t i-al' sense, a dequate seisri: tar: ins w.ith possibly a

1. et:.inor er:er i:-s.

- this coint the exce::i: s refer to areas that

'. (

i have been identified for further evaluation and do not necessarily iroly It is anticipated that minor modifications will be re aired deficiencies.

by the fiRC staff in areas where substantial additional protection to the In other words, the change.

p :bli: health and safety can be attained.

of seinic design criteria over the years can be accomodated by utilizing realistic evaluation techniques and the intent of current criteria as a

'sta-iari provided there are no significant changes in the stated seismic hEurd isst:ned for design at the site.

4 Ths SE? program also has provisions for re-evaluating the desigr, seismi@

gre;r.d r.otion input for each site utilizing site specific informatier. to ar:ive at which is intended to be a realist : estimation of the seisric This information will be incorpc-a ed into the structural /mE:hanical ha za rd.

per-ions of the review as it be:cmes available.

Additiona sly, the SED licensees have initiated a program of their own to re-evaluate the seismic 1

inp;t design bases of their facilities.

i e

i l

.n'.tial review of the early SEP f acilities indicates that a cer a n t=o of -e:r:fitin; may be required especially in providing additional pi;ing In certain cases structural modifications miy also ar.d eouipment supports.

Tja fiP1_slaff recognizes tha. these older facilities co t

be.e:essary.

- - ~ ~ - -

l

. n:t ;csiss the sate seismic margins as facilities being designed undsr

._ -.._.. ~

c.-- s. standa-ts; however it is the belief of the staff that these 4:#~ ':'es cessess adecuate inherent seisri:.argins to continue coera-j

'- :ne i e-i 1 ur.til the SEP seismic ecaluation is complete.

T:is a

E-:~..s'ni ased upon (1) tne f act that -hese fa:ilities are site: in J

re:3 'ctiy low seismic regions, (2) historical data which suggests t.at la-;5 .:astrial f acilities have not been significantly damaged unde-seis-i:::' gs and (3' consideration of the inherent and in many cases unc: anti-4

f i

l fiable seismic resistance capabilities of these facilities.

l It is anticipated that topics ray be identified within SEP which potentially could impact other operating reactors or new plant licensing.

A feedback mechanism has been established to relay the information in an expeditious manner to others on the NRC staff, licensees and applicants to assure that appropriate actions are taken in a timely manner.

1 B.

Task Act Plan A-40 (TA? A-40)

L Task Action Flan A-40 is a short range program which was instituted in i

conserva: isms in :ne calculatec resp:nses of structures, systems an:

I components, ine'uting -he consiceration of elasto-plastic seismic ar.alyses, site spectra (as opposec to site incepen:ent spectra such as that cescrioeo I

in 7.eg. Guice 1.5e), n:nlinear strutural cynamic analyses, and soil /stru:ture';

interaction. Phase 11 consists of an evaluatier, of :ne conservatists in 1

the seismic input cefinition, inclucing the study of earthquake source i

mocaling anc -he anaysis of nearfiele ground motion.

Results of the various,

tasts in this pre; ram to case have su:stantia:ec ine existence of c:nser-vatisas in the curren; seismic cesign me nocology.

C. Seismic Safety Margins Research Program (SSMRP)

The SSMRp is a long range research program (apprcximately 61/2 yetrs)!

whicn is aimec at impr:ving the seismic cesign methocologies. The cojectivesi l

of :nis progra: are to:

I es;imate tne c:nservatisms in the Stancarc Review Plan seismic a) cesign req,irere-:s, c) cevelop im:r:vec re:uirements, ano

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c) develop meth:dologies that realistically estimate the behavior of nuclear p:wer plants when subjected to earthquakes.

This program will build upon and extend the results from Task Action Plan A-?D.

D.

IE Bulletin Reviews The scopes of ths recently issued IE Bulletins regarding seismic i

1ssues were surna.-ized in the section II.E.

Many of the issues addressed by these bulletin; have already been resolved for many operating nuclear power plants. The cocpletion of the efforts involved in satisfying these bulletins has givsn and will give added confidence in the adequacy of ~the seismic designs of the operating nuclear power plants.

As the review of the respenses pre:eeds, the NRC will take appropriate actions based upon our assessments cf the responses.

i E.

Code Verification Program (Piping)

This pregrar was instituted in March 1979 and has as its objective the verification :f computer codes used by the industry for the seismic analyses of pipin; systems.

It is related to an older program entitled,

" Piping Eenchnark Problems", which has the goal of generating sets of piping problems f:r banchnarking computer codes used for both static and dynamic pipin; system analyses.

IV.

PLAN FOR RE57J.-'.3N OF SEISMIC ISSUES i

K'hile the staff a;rees that further seismic evaluation is necessary, as j

exclained above:

a) many conserta: isms exist in the seismic, design methodologies employed ir t e design of both old and new nuclear plants, l

I

I b) structures and systems have an inherent level of seismic 4

resistance, even if no explicit seismic design requirements are considered, and c) many investigations are currently in progress which are aimed (1) evaluating the seismic capabilities of older plants, at:

(2) cualifying the conservatisms in the current seismic design requirements, and (3) developing improved, more realistic seismic design criteria.

Eased upon these f acts and nsiderations of a plant's seismic design as a whole the NRC staff does not believe it is necessary te recuire all nuclear power plants with cperating licenses to be seismi.: ally reanalyzed from the seismi: input definition through the evaluation cf the designs of structures, systems, and equipment and components as demanded in the UC5 Petition; In a sense, a " seismic reevaluation program" is continually being conducted which 1

integrates the lessons learned frc: cast experientes and the results of engoin; i

seismic programs (SEP, TAP A-40, tne SSMRP and the IE Eulletin reviews). These basic evaluations are addressing tnese issues whi:h are identified as being most important to public health and safety.

The IE Sull'etins which have been issued thus far, and any corrective actier.s

~

deemed necessary, have provided and will provide an additional level of tssura.:ei that the as-built configurations of oiping systems and their supports indeed have sufficient safety margins. Far; of the efforts has been to verify :.e I

acequacy of many computer codes which are used for the analysis of piping in i

the nuclear industry.

Generally, all computer codes reviewed have been f:und to be adequate. Therefore, this in:reases the level cf confidence that can be placed.on the ccmputational adequa:y of piping analyses once the intende:

methodology is confirmed.

This confiden:e can also be extrapolated t: a :ertain

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i decree to the computational adequacy of ce=puter codes used throughout the l

seismic analysis and design of the plants.

The results of the SEP will also provide a general data base regarding the If adequacy cf the seistic capability of operating nuclear power plants.

i a.y cencerns are idertified in the review of the licensees' responses to 7

trase Bulletins cr in the SEP reviews, appropriate actions will be taken by the NRC.

i i

The NRC staff is embarking on three parallel efforts which will aid in re-t assuring *.he ade:uacy cf the overall seiscit designs of the operating plants w.".ich are outsic'e thE s::pe of 15e~$brrent SE? review.

The first of these efforts involves a detailed study of the criteria used f:r tr.e desien ;f ea:h c;erating p' an*.

c sis:s cf determining for ea:h plan: the seisci: 'n;ut used for tre ;1er.: cesigr (peak ground acceleration, ground spectra, carsing values, etc), the analy*ical techniques, the load c:=binations and

~.e allenable leadines used for the design of structures, ard totponents, and any Other significant parareters which systers, e:vipr.e-t are ir.:ce; crated i- :he cesign c' the everall ;lant.

The se:end of these efforts involves the reassessment of the seismic ha:ard at each of the plant sites.

This effort will then progress to a detailed evaluation of the seistic risk at any plant site where concerns arise as t: the adequacy :# t"e ground nctions specified for the original seismic Where i y significant discre;ancies between the originally ce:er-tesigrs.

'ned seismic ris a-d that deterrined throuch this reevaluation are ncted, a: ore:riate actit s sculd be tahEr by the NRC,cCnsicering the information

-r s effor: '

r.jun: tion with that cbtained in the firs effort.

e e :.

- 44

. f _.

i The third effort involves the development of capabilities for the verifica-These codes are j

tion of the computer codes beyond the existing requirements.

used not only for piping analysis and design, but also for the analysis and i

This involves the design of all structures, systems, equipment and components.

develcpment of sets of benchmark problems which would verify the computational 1

methodology of the computer codes.

These benc{. mark problems must therefore t

be of a sufficiently complex and diversified nature which would generally be t

a

~

Once a l

t.eyond the scope of existing closed form solutions to the problems.

ccmprehans! e program is established, it will be implemented as necessary.

This is a sizable effort and will take a f airly lon; time period to complete, i

As these efforts progress, their findings, the fir. dings of engoing and future seismic reviews of operating plan s, and the findings of TAP.A-40, the SSMRp and foreign data will be continually assessed and factored into any decisions and/or the initiation of additional studies and programs. We feel this is a responsible and intellicent approach for the resolution of seismic ~-

issues, and overall involves an effort far beyond that which can be accomplished in the"short term.

On the basis of the assessment of past, ongoing, and future seismic relatedi studies, the conservatisms built into both the old and the new seismic criteria {

an'd the inherent seismic resistance of nuclear power plants, I have determined ;

that the efforts delineated in the UCS Petition are unnecessary in the succeste;,

depth and time frame.

I believe that the direction in which we are proceeding,;

with evaluation and resolution of any seismic issues which may have a dele:eric!

impact on~ public health and safety, will not only address the concerns raised in the USC petition, but will lead to more appropriate and realistic seismic design requirements than are dictated by even current criteria.

Accordingly, f i

I have detemined not to issue an order requiring seismic reanalysis. The request of UCS is denied.

t I

A copy of the decision in this matter is available for inspection in the Cc=ission's Public Document Room,1717 H Street, N.W., Washington, D. C. 20555.;

A copy of this de:ision will also be filed with the Secretary for the C mission's review in accordance with 10 CFR 2.205(c) of the Comission's I

regulations.

i As provided in 10 CFR 2.205(c) this decision will constitute the final i

a:: ion cf the Comission 20 days after the date of issuance of the decision, l'

unless the Comission on its own motion institutes a review of this decision

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l

. /W h 5

Haroic R. Denton. Director Office of Nuclear Rez: tor hegulation Deted at Bethesda, Maryland this 10 rcay of January,1930.

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3 8'1. g50 P Le UNITED STATES OF AMERICA E

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NUCLEAR REGULATORY CO:'J41SSION Uy'.., d ct StcT '

D..gg & StN -

p,,

t 3.,,.c UNION OF CONCERNED SCIENTISTS; DIRECTOR'S DENIAL N.

N 0F REOUEST FOR SEISMIC REANALYSIS

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8 By petition dated March 23, 1979, the Union of Concerned Scientists (UCS) requested that the Comission order al1 licensees of operating reactors to perform seismic reanalysis within 120 days of the Comissien's order. This request w2s referred to the Director, Nuclear Reacter Eegulaticn tc be treated under 10 CFR 2.206 of the Cc. mission's regulations.

Notice of receipt of UCS' petition was published in the Federal Recister. [44 FR 28737 (May 15,1979))

~

Upon review of the material submitted by UCS, and upon censideration of other relevant information. I have detemined not to issue an order equiring seismic reanlaysis.

Accordingly the request of UCS is denied.

A copy of the decision in this matter is available for inspection in the Comission's Public Document Roem,1717 H Street, N.

W.~, i.'ashington, D. C. 205S5.

l A copy of this decision will also be filed with the Secretary for the Comission's review in accordance with 10 CFF. 2.205(c) of the Comission's regulations.

As provided in 10 CFR 2.205(c), this decision wil' constitute the final action of the Comission 20 days after the date of issuance of the decision, unless the tcmission on its own motion institutes a review of this decisio that time.

< 410/

Harold R. Denton, Director Office of Nuclear Reactor Regulation Dated at Bethesda, Maryland, this 1Crth day of January, 1980.

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UNITED. STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE TEE CO M SSION

_______________ _______________ ___x-UNION OF CONCERNED SCIENTISTS EMERGENCY PETITION FOR REANALYSIS OF THE CAPACITY OF OPERATING U.S. NUCLEAR :

POWER PLANTS TO WITHSTAND EARTEQUAKES

__x MOTION FOR RECONSIDERATION OF THE CCMMISSION'S DECISION TO REFER THE UNION OF CONCERNED SCIENTISTS SEISMIC REANALYSIS PETITION TO THE STAFF On March 28, 1979 the Union of Concerned Scientists filed an Energency Petition for Reanalysis of the Capacity of Operating U.S. Nuclear Power Plants'to Withstdnd' Earth' quakes.

On 5pril 25,

' 1979 the Commission referred this matter to the s'taff for treat-

=ent as a petition filed under 10 CFR 2.206, requested the staff to "promptly respond to the petition," and finally directed the staff to provide it a schedule for response to the petition with-i.n one week.

On June 28, 19 79, a public meeting of the Commission was held to hear a staff Briefing on Seismic Design Capability of 0perating Reactors and Responses to OIE Bulletins on Seismic Analysis.

At this briefing, the Co==ission was informed, inter alia, that many operating plants differed.substantially from the drawings.

At this meeting, the staff informed the Commission that

o

~ 1 the UCS petition was one of the factors they were " folding into" their seismic review program.

In addition, Mr. Eisenhut was directed by Chairman Hendric to have the response to UCS's peti-tion " fairly soon."

It is nmi 8 1/2 months after the filing of the petition and 5 1/2 months since Mr. Eisenhut undertook to respond

" fairly soon."

UCS has had no response whatsoever from the staff, nor have we even been inforced of the progress of the staff's seismic evaluation program, which bears directly on the issues raised by DCS.

This situation represents an unconscionable violation of the most fundamental requirements of fairness and due process of law.

The staff has ignored both the Commission and the Petitioner.

This; is either a manifestation of gross inconpetence or purposeful evasion.

Under either interpretation, UCS is now clearly en-titled to the relief it originally requested.

The Cc= mission now is compelled to assume jurisdiction of this matter.

It cannot countenance the staff's continued course of nonaction.

UCS argued in the petition that the Commission should exercise its inherent supervisory authority.

One of the grounds we alleged to justify this was that the problems complainedi of in the petition are largely the result of the staff's conduct.

The staff's continued f ailure to respond to the petition makes a renewal of this request absolutely necessary.

9 p

9 3

Therefore, UCS hereby requests the Commission to reconsider its decision referring this matter to the staff and to i=nediately take jurisdiction of the UCS petition.

Respectfully submitted, e

^M Ellyn R. gi;ss Sheldon, Harmon & Weiss 1725 I Street, NW, Suite 506 Washington, DC 20006 (2021 833-9070 General Counsel for Union of Concerned Scientists Decertber 10, 1979 w

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