ML20101E552

Summary of Reviews & Analyses Performed for Part 21 Evaluation Re Floor Response Spectra Peak Broadening for Ei Hatch Nuclear Plant Units 1 & 2

ML20101E552
Person / Time
Site:	Hatch
Issue date:	12/14/1984
From:	GEORGIA POWER CO.
To:
Shared Package
ML20101E541	List: ML20101E538 ML20101E552
References
NUDOCS 8412260291
Download: ML20101E552 (47)

Text

.

!f 4 Enclosure to hED-84-622 December 14, 1984

SUMMARY

OF REVIEWS AND ANALYSES PERFORMED FOR THE PART 21 EVALUATION CONCERNING FLOOR RESPONSE SPECTRA PEAK BROADENING FOR s E. I. HATCH NUCLEAR PLANT UNITS 1 AND 2 DECEMBER 14, 1984 8412260291 841214 PDR ADOCK 05000321 S PDR

me ;_ o. .

,ll ..?  ? Q_ -

lx

~

is- INDEX -

' !- I . - ' INTRODUCTION

~

l A. . Scope B.~sBackground Information .

'C. ' Organization forLPart 21 D. Summary of Engineering Metho'd ology E. -Engineering Criteria-

I I '.

ENGINEERING METHODOLOGY AND RESULTS

, LA. Generation of New Floor Response Spectra

~1. Modeling Enhancements

2. New. Synthetic Time Histories

3. Soil-Damping for Unit 2 Analyses B. Comparison of FRS and Results for Equipment and Structural: Subsystems

1. Results of ERS Comparison

~2. Results of the Equipment Sample and Evaluation-

3. Results.of-the Structural Subsystem .

Sample and Evaluation-

-C. Comparison of FRS and Results for Piping '

. Systems . -

D. . Comparison of FRS and Results for Cable

-Tray Supports

,-^ 'III.

SUMMARY

IV. CONCLUSION V. REFERENCES o VI. TABLES AND FIGURES t

, ii

-- -e --

-sv w g-w-r- n m,- - - -n,,- , - , - - ,,, .----,-

p;;; 3, WW

,L,

n~ ~

4 I. _ INTRODUCTION

Jn, '

q; ,.

g'" _ ' . .

A. ': Scope 1r ,-= . og

37. ;This report presents the' basic methodology:and result's.

L

'of a'Part 21-sa'fetyLevaluation concerning the peak I b'roadening of the floor. response spectra (FRS) for.

~

Edwin I.. Hatch Nuclear Plant Units.1 and 2.

= B; Background'Information

~

,Al discrepancy _in'the Final Safety Analysis Report (FSAR) commitments;for. peak broadening of the seismic i- ~

floor responsefspectra curves was bro ~ught to the ,

attention of Georgia Power Company (GPC) on-December

^ '

, f20)'1983, ~in the course of an analysis performed for

_ the' recirculation pipe replacement on-Unit 2. Section

~'

'3.7A.2.8'of the Unit 2 FSAR states that the computed FRS were smoothed,_and peaks. associated with the structural frequencies.were widened by "+ 15 percent."

s It'was discovered that the floor response curves

, generated-for'the Unit 2 Reactor Building were widened by + 1'O percent.

Based on research for the Preliminary Safety Analysis

~ Report-(PSAR)'and-the original seismic analysis, GPC

-has concluded that the intent and appropriate

' commitment was to broaden the FRS by + 10 percent.

The commitment to a + 10 perc'ent b'roadening was accepted by the Nuclear Regulatory Commission's (NRC's) consultants in their review of the PSAR for Unit-2 and by the NRC in the Safety' Evaluation Reports ffor'the Unit 2 PSAR and the Unit'l FSAR.

c _

In question A-5 of the Unit 2 PSAR the NRC requested the following: List all'Categ,ory I (specified as i

1 h

9 *w rev --. - - - -Nei - v

1

$p :y,. _ y y,

il? .

Class'l'in PSAR).. structures,-systems, and components, including-reactor internals,. piping, cable hanger trays.and mounted equipment, and the method 6fiseismic

" analysis (modal analysis response spectra, modal' analysis time. history, equivalent static' load, etc.)

=or empirical (tests) analyses which will be employed in the. design, including _ applicable stress or-

-deformation criteria. Provide a brief-description of

.all methods that are used'for seismic analysis.

In_ Amendment 9, page A.S.2 under Equ.*pment, Georgia t: -

' T-Power Company responded:

The floor response spectra will be sm'oothed such that the response curve will-be an upper. bound envelope of

,-all the acceleration points. Whenever the response

. curve comes to a peak, the curve will be made flat in a region + 10% of that peak frequency.

.In their letter to the AEC dated November 2, 1971, Mr.

N. M. Newmark, W. J. Hall,, and A. J. Hendron stated:

"After our review of.the PSAR, including Amendments 1-though 9, it11's believed that

~

the design of Edwin I. Hatch Nuclear Plant-Unit 2 can be considered adequate in terms of provision for safe shutdown for a Design Basis Earthquake of 0.12g "(note

~

. typo: should be_O.lSg)"' maximum horizontal ground acceleration, and capable otherwise of withstanding the effects of an Operating Basis Earthquake of half this intensity."

m 2

- a

,e

+

3

-s

.r Further, 'it was' stated that: -

i "We believe th'e procedures used in the design

^

and analysis-are in accord with the.

L state-of-the-art. It:is our conclusion that-

-the design' incorporates:an. acceptable range Lof margins of safety forfthe. hazards consideration."-

Signed: W .- J . Hall Section 1.3.2 of the: Unit 2 FSAR lists all the' signifi' cant changes that have been.made to the plant since submittal of.the PSAR. There is no reference to a1 change to increase the peak broadening to 115%.

This change would have been extremely;significant since it would necessitate changes in the purchase-specification of all' equipment for Unit 2. ItLis based upon this review that GPC contends the

' appropriate peak. broadening commitment for Unit 2 was 1 10%. ,

A complete : review was subsequently performed for all floor, response spectra generated for both Unit 1 and Unit 2. The results of this' review showed that not all-FRS were broadened fully 1 10 percent.

Georgia Power Company (GPC), an operating company of the Southern Company, directed the Nuclear Safety and Fuel' Group of its engineering service company, Southern Company Services, Inc. (SCSI), to perform a Part 10 CFR'21 evaluation regarding the significance

.of-this discrepancy. This evaluation was based on the intended commitment to broaden the FRS by 1 10 percent.

, 3 k- J

V'_. '

j@ i <

h ,

On January 6, 1984, :in letter NED-84-008,- GPC informed '

"~

the' United States Nuclear Regulatory Commission, Region II,.that discrepancies were found to' exist-between seismic.FRS broadening commitments:as stated

'in the FSARs and those actually.used in the seismic analysis of the plant.. In subsequent; letters NED-84-066 dated February 10, 1984, and.NED-84-274

" dated May 24,.1984,'GPC informed the NRC of the results.available to those dates of the Part 10 CFR 21 evaluation being conducted on this subject. On September 12, 1984, in letter NED-84-484, GPC informed ,

the NRC,1 Region II, and the Director of Nuclear Reactor Regulation that a reportable condition in accordance with the criteria of Part 10 CFR 21 did not

-exist for the discrepancies in the FRS for Edwin I.

Hatch Nuclear Plant Units l'and 2.

s C. Organization for Part 21

-Figure l' illustrates the organization for the Part 21-

evaluation. In addition to the' evaluation requested

-of SCSI, GPC retained the services of Dr. Robert P.

Kennedy of Structural Mechanics Associates and Mr.

Donald F.. Landers of Teledyne Engineering Services to

! provide' engineering and technical assistance to its architect-engineers, SCSI and Bechtel Power ,

Corporation. These consultants attended meetings,

_ assisted in developing engineering methodology, reviewed and approved major engineering decisions, reviewed the comparison of the new FRS to the original FRS, and concurred with the decisions regarding which new FRS warranted an evaluation of subsystems.

The SCSI Nuclear Plant Support Department - Hatch

, coordinated the detailed implementation of the Part 21 evaluation. The engineering work on the evaluation 4

e s ,

4

$ g -

_ - , _, e.

wasperformed]ointlyJbyBechtel. Power.Corporationand

.SCSI:DesignLEngineering. Bechtel Power Corporation a responsible for the Unit 1. Reactor Building, the-w's gq q , Unit 12 Reactor' Building, and the Main Stack, which is a -- shared struc.ture. SCSI Design Engineering was l responsible.for,the-Control Building, Diesel Generator

' Building, and the. Intake Structure. These three-

' structures areishare'd by Unit 1 and Unit 2 and are Ltherefore' analyzed in accordance with the respective

' seismic criteria.for both' Unit'l and Unit 2. The' Y- .' division of responsibility noted above is-consistent

,- with that which was in effect during the original 1 -

plant design.

-D. Summary-of Engineering Methodology *

-Figure 2:is a $ist of steps taken to' perform this Part-3: , . :21 - evaluation. First, the. original'FRS, time histories, models, and analytical. methods were 4

. reviewed. This review provided information on 7- conser9atisms; that . existed in .the analysis. beyond those committed to in the FSAR,-and which changes, if any,:were required so that the models_ accurately

-reflect existing conditions. These changes were.made t because they represent more accurate data than was available when the original analyses were performed; they are consistent with all appropriate 1-icensing commitments.

, .TheLsecond step was to develop-engineering criteria-which were used throughout this evaluation to maintain control.and-uniformity.between engineering

. organizations. These criteria are discussed in section E.

+

r 4

5 o

_t

- _ . _-.-n-,-

gq ,

Qg --

/

_ -u 4s: '

^

2 w

-The, third: step (wassto develop new FRS, that metJall:

' intended l'icensing commitments. These new1FRS were.

~~ developed atLthe;same= locations;and directions:as:the original FR'.S 1These new FRS were.-broadenedLto'the c' ,

Lintended. licensing commitment-of +'10. percent.

+

.Q' : '

Fourth l these new FRS;were-compared'to the original:

c -

!FRS'at'a specified' damping value. eat sample:of subsystems at a;particular location was evaluated lfor i i$ ;

the' impact of~the new FRS:o'n the seismic qualification g) , . of :tph subsystems 'irr the sample if the changes in' FRS ,

at that. location were judged to warrant'such an

' evaluation.

• ~

LThe consultants participated-throughout this fevaluation. :In~ particular, they were. involved in-

-developing the criteria, in making major engineering decisionsf in reviewing FRS comparisons, and in making s! 'judgements concerning which FRS warranted an ' '

Levaluation of subsystems.

cE. Engineering Criteria '

After a thorough investigation of the original FRS, b

'the. mathematical.models and. time histories used to

~

generate these FRS,.and the PSAR and FSNR' commitments and related documenta, criteria were developed to

?A( perform'the Part.21' evaluation. These criteria not only-established-a consistent approach between oorganizations,-but they also enabled overall control

-in performing a very complicated evaluation. The consultants participated.throughout the development of Ethe criteria.

J V

o

, 6 u_

n ; q- ~

. Figure 3 is a l'ist of the. major-items. covered by the criteria. In general,.the. criteria addressed the

' generation of new FRS,.the~ comparison of'the'new FRS to?the original FRS, and.the selection and evaluation of different cubsystems. The FRS that were' generated-consistent with the criteria met all-licensing commitments.

i.

In particular,fthe " input parameters for reanalysis" reflected-the changes.which were made to.the original-

-mathematical.models and time histories to improve the accuracy of-the; evaluation. These changes included, for1 example, reduced soil damping over FSAR allowed damping for Unit 2 analyses,.use of l'ter a and more accurate. soil-data for the Unit 1 Diesel. Generator Building seismic analysis, new synthetic time histories, and changes to the mass and stiffness of the mathematical models derived from any significant changes to the structure since the original analyses were done. All. changes met or were more conservative than the original licensing commitments.

The " criteria for reanalysis of structures and generation'offFRS" provided integration time steps for the time history analyses, spectra generation frequency steps, the peak broadening requirement of

+ 10 percent, and the requirement to vary the soil properties for the Diesel Generator Building and envelop the associated FRS to account for uncertainty 4

.n the soil properties for this building. Again, all work was done in accordance-with licensing commitments.

.The " criteria for. comparison of floor response spectra" provided. damping values at which the new FRS were compared to the original FRS and the critoria 7

y y; -

b ,

b J

r ~

used to determine whether the. change.in-FRS. warranted' an evaluation of the-impact.of this change on the-seismic (qualification'of subsystems.

p.

The' rest of;the. criteria primarily provided procedures-sfor selecting.and evaluating subsystems-if such evaluations were judged necessary.

c ,

. II. l ENGINEERING METHODOLOGY AND RESULTS A. Generation of New Floor Response Spectra L.

'An evaluation was made of the original FRS, the original models, and the time histories that were used

.to.-generate these FRS; consequently, it was decided

.that the most appropriate method to perform the safety

,. , evaluation was to generate new FRS. The generation of

-new FRS was undertaken for two reasons: .first, there was conservatism in some of the original FRS beyond Et hat contained.in the licensing commitments; second, the original FRS were based on models that, in some c a s e s', do'not' reflect existing conditions.

.The'new FRS were developed only for the Design Basis Earthquake (DBE) since this is the largest design earthquake and would have the greatest impact on plant

. safety.

The following discussion explains major. changes to the

-models, soil damping, and time histories. These

~

., changes reflect either as-built conditions or improved n

input parameters. These changes all meet or-are more

., conservative than the intended licensing commitments.

7.

8 t

y ;ic Q. M

.i.

4

1. Model-Changes -

~

The original models were used unless the-existing

.1 s 1 conditions ~ warranted a change. For example,Ethe-o - .

-Unit liseismic analysis of the Control Building' only considered the orig'inally' intended three bay

~

structure for Unit:1. The building was actually constructed as a.four bay structure'and is'used as-a' shared Control. Building for Unit 1 and Unit 2:.

'The Unit 1 model.was'therefore upgraded to.the original ~ Unit 2 model that reflected the fourth.

bay and contained a more detailed mass calculation. This caused a downward shift of-the calculated fundamental. frequency of the Control Building of approximately 0.8 Hz from that which-was calculated using the original Unit 1 model.

JAimodeling change was also made to the soil shear modulus values (G3) for the Unit 1 analysis of the Diesel Generator Building. The Diesel Generator Building is founded at grade, and there

~

y is a. degree of uncertainty to its soil shear mc'ulus d that does not exist for the other Seismic Category I structures which are founded in deeper soil. Accordingly, the soil shear modulus values were varied to account for this uncertainty. The original- Unit 2 ' analysis of : the Diesel Generator

~ Building reflected a more recent estimate of the

, mean G and its variation, and this mean value S

was used for both Unit 1 and Unit 2 reanatyses.

The value was varied consistent with the appropriate licensing coevitments to account for uncertainties. All avail a soil data was reviewed to ensure the mean value used in this analysis was indeed appropriate. Three separate FRS were generated for each unit to account for 9

ra.'

43

.possible. variations.in soil shear modulus. The three separate FRS-were then enveloped-to.l form.the new FRS-for'the-particular. unit.

.A review of the Intake Structure also indicated

'that some changes'were warranted. The soil springsLwere recalculated to more accurately L

,  ; reflect the non-rectangula'r base slab and'the significant~ degree of embedment. Embedment was s

accounted for by using procedures similar to those-contained in BC-TOP-4.'"' In addition,

. revisions to mass and structural stiffness were made to reflect existing conditions.

'All modeling changes reflected more complete

~

information than.was available when the original analyses were performed. It was the intent of this-part 21 evaluation to use the original models and the premises on which they were developed Tunless the existing conditions and/or assumptions of material properties were found to be significantly different. '

2 . - -New Synthetic Time Histories The original Unit 1 FRS are based on the North-South componentlof the 1940 El Centro earthquake scaled to the appropriate maximum ground acceleration. The FRS generated from this

. time history were then multiplied by a

. normalization ratio to scale the Zero Period 1 Amplitude (Z?A) of the FRS to'be equal to the maximum floor acceleration obtained from the response spectrum analysis using the ground design response spectrum. The original Unit 2 FRS are based on a modified Taft 1952 earthquake f.

10 c.. .

ll- e.

ow,

} .r l s

b

' horizontal component.

, The seismic design basis t '

r zfor.the plant is'the ground-design response-t L spectra-for each unit. The time histories were

! ~

tised .only as a tool to obtain FRS. Currently,.

'better techniques are available to obtain synthetic ~ time histories:that more reasonably

' envelop-the ground design response. spectra. It was therefore decided to develop a synthetic time l

. history for' Unit 1 and another synthetic time a;

-history.for Unit.2, both of which more closely  !

envelop the appropriate design response spectra ,

and do not use'a normalization ratio. 'Each of-

~

these new : synthetic time histories is at least 20 seconds.in duration, and both wer'e reviewed by the consultants.

~

-3. Soil Damping for Unit 2 Analyses Originally, for most of the Unit 2 Category I structures, the FRS were developed-using 5 percent of critical damping for soil for the Design Basis Earthquake (DBE) case. The FSAR allows the use of.

equivalent radiation soil damping in accordance with Table 3.7A-2 which, in almost all cases, will give higher damping than was originally ~used.

.I New FRS were generated for the Unit 2 structures using equivalent radiation soil damping. However, a value somewhat more conservative than that allowed by the Unit 2 FSAR was used. The

. following procedures were used to obtain soil damping values for the Unit 2 analyses for the Part-21 evaluation.

p i

, 11 )

L.

i l

\

' ~ ~- -- ~

-~~ ~ ~ ~ ~

r a=

4  : q. .

A

- a <

].< ,

g'j !For all' structures otherithan the Diesel Generator,

,.I j Building:

=

For translational~_ soil damping, the smaller of i ,

100 percent of. Table 3.7A-2 or.75 percent of-Tablef3.7A-2 plus.6 percent soil material

~ damping was used.

'For rocking soil damping, 100 percent of 1

Table 3.7A-2 was.used.

For the' Diesel-Generator Building to account for soil variations:

For all soil damping, the smaller of 30 percent of-Table 3.7A-2 plas 6; percent soil material. damping or 20 percent of critical damping was used.

In every case where-it was applied, 75 percent of Table 3.7A-2 plus 6 percent soil material damping

- controlled. For the Diesel Generator Building, 20 percent of critical damping controlled.

-B. Comparison of FRS and the,Results for Equipment and Structural Subsystems e

This section covers the initial comparison of the new FRS to the original FRS. It also covers the evaluation of the impact of the.new FRS on equipment seismic qualification and on structural subsystems except for piping systems and cable tray supports.

The initial comparison of the new FRS to the original FRS'was made using the five percent damped spectra-curves. The value of five percent damping was chosen

.since most types of construction have a specified damping of five percent for DBE loading. In addition, comparisons at five percent damping reasonably sindicate which new FRS exceed the original FRS 12 e- --, e - -- .ve, n .-.- - v - -

4 Wi  ::

~ o , , i 1. , sufficiently. to; warrant ~ art evaluation of the impact of: -the-new1FRS on equipment and structural subsystems. . , This ' comparison ~was made' at :every location and ~ Idirection where.the orig'inal FRS'were' developed for ' Seismic Category I buildings.

1. Results of_FRS Comparison Th'e results of the' Unit 1 and Unit 2 FRS comparisons are displayed in Tables 1 and 2,

.respectively. A~ total of'60 FRS comparisons were r made for Unit 1~and a total of 103 FRS comparisons. were made for Unit 2. Each FRS comparison was evaluated to determine if-the changetin spectra warranted a further evaluation of equipment and structural subsystems-at that particular locatien. In general, this evaluation included reviewing the degree, if any, < .by_which the new FRS exceeded the original FRS, - the frequency range or ranges at which these excesses occurred, and the magnitude of the -spectral accelerations. Based on-this evaluation and on-the concurrence cf the consultants, it was decided that only the FRS in the.8 locations . indicated in Tables 3 and 4 exceeded the original FRS'by a degree which warranted evaluations of equipment and' structural subsystems. The three Unit 1 Control Building FRS comparisons that were judged to warrant evaluations of the impact of the new FRS on equipment and structural subsystems are shown in Figures 4, 5 and 6. The dashed line is the original five percent FRS and the solid line is the new five percent FRS. The 13 1 g; +;e .

v .

' ~ ' primary reason for-these excesses is the ~ enhancements _made1to theJUnit 1 model of the . Control' Building. .. T I. + -Thet'wo' Unit 1 Diesel-Generator:Buik~dingFRS comparisons that-were judged to warrant an ^es _ r zevaluationfof the impact of.the new FRS on ~ . equipment and structural subsystems are'sh'wn o in ,~_ -Figures 17nand 8. cThe excesses are due'to the- -shiftLin the funaamental' frequency of the building . caused by the use of:the updated-estimate of the mean soil 1 shear modulus. ~

The three UnitL2 Intake Structure'.FRS comparisons that were judged to warrant an evaluation of the

~ impa'ct-of the new FRS on equipment and structural subsystems.are shown'in Figures 9, 10 and 11. .In. conclusion, only 8 new FRSiwere judged to i contain sufficient changes to warrant an

evaluation offtheir~iq. pact'on the seismic

' qualification of equipment and structural subsystems. The consultants concurred with this. conclusion. : Tables 3 and 4 provide _a summary of the locations at which evaluations of the impact of new FRS on subsystems were made for Unit 1~and. Unit 2..respectively. -2. Results of the Equipment Sample and Evaluation A sample of equipment was selected for each of .these-locations listed in Tables 3 and 4. The Tsample was formed by choosing a wide variety of I types of se'smic i class 1E equipment from the o Equipment Location Index (ELI). The sample was ' enhanced by adding items of class 1E equipment 14

y.  ;- - ,

3J . ?i? chosen during a'walkdown of the equipment'at these .eight: locations. The purpose of the walkdown was to_ choose equipment that appeared to1be-particularly' vulnerable to the effects of an earthquake. This biased the sample _toward equipment that would be most affected by. changes ' in the FRS and therefore would be more.likely-to' indicate problems with.the seismic qualification of equipment cau3ed>by the change in FRS. Dr. [ Robert Kennedy. participated in the walkdown. His -experience was.used to ensure that particularly vulnerable equipment was included in the sample. Table 5 presents, at each locatio'n, a summary of the' total number of pieces of safety related r -equipment, the number of items chosen from the ELI,-the numberi of items added due to the walkdown, and the total number of items included ' in the sample. The table also provides the percentage of the total number of pieces of 4 equipment at each location which were chosen for review. As indicated in Table 5, approximately 30 ' percent or.moreaof all seismic class 1E equipment at each location was evaluated. U- In all, a total.of 187 items of equipment were evaluated. .This represents approximately 38 percent of all~ safety related equipment located at these eight locations. Based on the size of the sample, the inclusion of a wide variety of equipment, and the bias toward the most vulnerable types of. equipment, the sample is an adequate basis on which to make a determination regarding the impact of_the new FRS on the seismic qualification of equipment. 15 = - - - - - -- - - - - - = , - ---~- - = - --- - -- ----- -- - - - - - - - - - - -- - 3ra' fvw _ - , ?_v ' , .4 2 _ ~ LThefeguipmenttseismic qualificationJreports were. 9 :- ' ' ' m ~ .-

Levaluated1using;the' appropriate set of new FRS. '

~ . ,  ;(N-S,c.E-W,pand? Vertical).to.-determine.the effect-s _ ~ .c lof the-new FRS'on,the equipment's seismic ' ' ' ', ~

qual'ification.' iThe engineering criteria provided sa.setiof; forms-for~this evaluation. LA separate 1 ' form was, developed for-qualification-by~the 1following. methods: . testing,. dynamic. analysis, static)coef icient. method, Land qualification'by~

X . Lother'means. These. forms-provided.a consistent. method'for: performing and documenting the '~ 2 evaluations'. Bechtel Power Corporation's Equipment; Qualification Group performed each evaluation. '? , . Table.6 lists all equipment that.was evaluated'as'- well as-the evaluation results. -This table-

provides the Master Parts-List-(MPL) number, a.

s

. description'of the item,-the vendor,'and the location of each item chosen. As the information

demonstrates, no cases were.found where the new p- #

'FRS had a detrimentalJeffect.on' equipment seismicL ' Equalification. .There were a few cases in which ! insufficient information precluded a decision , , regarding theJimpact of the new FRS on seismic 4 . qualification. LHowever, the equipment sample --g , remains.sufficiently large to' evaluate safety. ju ' Based'on.the results of an extensive evaluation of of the 177 items of equipment for which a . determination could be-made, it was concluded that the equipment seismic qualification was not adversely affected.by the new FRS. ~.- 16 . - . . _ - - - - ~ - g3 . CG  ?.. . , c

13. Results'of;the Structural Subsystem Sample and

, Evaluation-L .At the eight locations where equipment was evaluated, a' sample.of structuralisubsystems other than-piping systems and~ cable tray supports was also chosen to evaluate the effect of the new FRS on the structural integrity of such systems. The selection of the structural subsystems (except-concrete masonry walls) was based on a walkdown 'since this was determined to be the most effective . method of identifying subsystems that should be evaluated. The purpose of the walkdown was, a)_ visual identification of potential vulnerable -subsystems, i.e.,-flexible subsystems that would-tend to amplify-floor motion, and b) selection of representative examples of different types of subsystems, biasing the selection toward the more -heavily loaded cases. Dr. Robert Kennedy participated in this walkdown. His experience was used to ensure that the selection of the' ' structural subsystems met these criteria. Concrete masonry walls were selected based'on the most severely stressed conditions as determined during the NRC IE Bulletin 80-11 evaluation. All concrete masonry walls at Plant Hatch that are in proximity to safety related equipment are reinforced both horizontally and vertically. The subsystems were evaluated using the complete set of_the new FRS (i.e., N-S, E-W and Vertical) at the appropriate FSAR damping values. The effect of the change in FRS on the structural integrity of these subsystems was evaluated. 17 P hn- ~ ..l~ ;)' ___ .

{s 4

Table 7 tabulatesnthe results of the evaluation of theceffect-of theinew FRS on' structural. subsystems. >The table ~providesHthe classification: ~ of'theJsubsystems.and;the: number.of items

. evaluated as.well as the results of'the

< , ' evaluation. The evalua' tion of equipment; supports considered; both anchorage of equipment and' intermediate supports for' equipment. During the'walkdown, ( ~ . instances were identified in which equipment p , ' mounted directly;to the-floor or_aiwall appeared' to'have questionable anchorage, and such cases -were.therefore(evaluated using th'e_new FRS. These items-included,-for example, a control panel, some battery chargers, a transformer, and.a battery , s rack. : Intermediate equipment supports were also evaluated. For. example, these~ included instrument- . racks, panel' supports, and supports for tanks and !Neaters. Some-non-seismic Category I subsystems were also ' chosen for evaluation since their collapse could t.

f ' damage seismic Category I equipment. Those items are listed in the footnote to Table 7. Concrete block walls were evaluated using the same

, ' procedures-that were used during the evaluation .. a for NRC IE-Bulletin 80-11. The conduit supports and HVAC_ supports that were evaluated were chosen-s as representative of the most vulnerable components of these subsystems. All the subsystems shown in Table 7 except those 4 in' Category VI " Cable Tray Supports" are located at-one of'the eight locations where the new FRS ~ -were judged to warrant an evaluation of the effect 18 E + r- + . $ J -of~the new FRS on. subsystems. There-are a number of seismic class 1E-cables in the east cableway of ~ ~ the Unit 1 and-Unit 2-Turbine Buildings. The- ^ turbine buildings are not seismic Category I T. structures, but a time history analysis was l performed and FRS were generated during their . design phase. It should be.noted that the ' buildings have,been designed not to collapse due to a DBE. The only safety related components in the buildings are cable' tray' supports for the class 1E cables an'd these components occupy only a y -

very small portion of the buildings. Because of this, rather.than regenerating these FRS, the original FRS were broadened to th'e intended licensing commitments of + 10 percent. These 4

revised spectra' curves were used to evaluate a sample of cable tray supports in the Unit 1 and - Unit.2 Turbine Building east cableway. The -selection of the cable tray supports was again chosen to represent the most vulnerable components of these subsystems. , Damping in accordance with the FSAR was.used in this evaluation. ' The evaluations of all of-these subsystems are based on dynamic analyses and stress evaluations. l The effect of the change in the FRS on the structural integrity of the subsystems was evaluated, and no instance was found in which the 'new FRS had a significant impact. Based on this . extensive evaluation, it was concluded that the Le ' new FRS do not adversely affect the existing ability of the subsystem to withstand the DBE. 19 c- 1 'T w ms;; _ ,by .. b , . x s.- ' st - + -m !? ' , , e, . y :C; " Comparison ofi FRS and, Results7 for : Piping -Systems . hit *" ' . . ,fADseparate; comparison of theln'ew FRS-to the~-original ~ FRS . was. performsd ?for piping: systems- 'since . it L has .been - 3s: determined;th'at higheridamping'is jus'tified than.was' originally (used-for the d'esign'of the Plant: Hatch j ~ tpiping' systems.- For this Part'21: evaluation thelone fpercent. damped FRS'that/was originally used to define the DBE' loading for piping was compared to=the new~FRS; ~ at the: Pressure-Ves'sel ResearchLCommittee:(PVRC) 'recomme'ded damping value.' Figure 12.shows PVRC' ~ n 1 recommended' damping for se'ismic~ analysis of piping . systems. :The-FRSocomparisons were'made at the same . eidhtibuilding' locations for which eq'uipment and ' y - structural 1 subsystems were evaluated. , TheselFRS comparisons for piping systems showed that .for most cases'the original one percent damped FRS enveloped the new'FRS at-the PVRC damping.at all

frequencies. For those few cases where-thes new ERS-exceeded the original FRS, the. consultants judged the increase to be insignificant. Therefore, no *

< evaluations of piping systems were' required for the

.Part 21 evaluation. -Tables-3Jand 4 summarize'the results of this evaluation. The conclusion of this evaluation was that the new FRS did.not affect plant safety regarding piping systems. D.'. Comparison of FRSiand Results for Cable Tray Supports -A1 separate comparison of the new FRS to the original 'FRS was also made for cable tray supports since it has 'been'shown that higher damping is justified than was ~ , originally.used for the design of the Plant Hatch cable: tray supports. For this Part 21 evaluation, the 20 4

• n , _ -

T y L.ip original-FRS that were used for-the DBE design.of cable tray-supports were compared to the new~seven pescent? damped FRS. The higher damping of seven- ~ i percent'used for~this safety' evaluation.is

conservatively based cut the results of tests.'"

These tests'were' performed by ANCO Engineers, Incorporated,-in collaboration with Bechtel Power. Corporation. " Georgia Power Company was one of the sponsoring! utilities of this test program. Figure 13 shows:the.results of this test program. This figure .is a plot -of lower bound damping . for cable . tray supports.as a function of-input ZPA. A review was , made of the test. program, the percent damping licensed at other nuclear plants, and the. cab 1'e tray support parameters ~for Plant Hatch-that affect damping. Based on this review, damping for cable tray supports at Plant Hatch should~be seven percent'or higher. The FRS comparisons for cable tray supports were made at 'the same eight building locations for which equipment and. structural subsystems.were' evaluated. The FRS~ comparisons for cable tray supports showed the new 7 percent damped FRS was enveloped by, or only .slightly exceeded, the original FRS used to design the cable tray supports for DBE loading. The consultants concurred that no evaluation of-any cable tray supports was required for this Part 21 evaluation. Tables 3,and 4 summarize the results of this ~ ' evaluation. The conclusion of this evaluation was that the new FRS did not affect plant safety regarding cable tray supports. t 21 mi Mo E-III.

SUMMARY

The following is a_ summary of_the major points.and results of this extensive-and complex Part 21 evaluation of the discrepancies in the original-floor response spectra.

'1. New FRS were-generated'using as-built models, soil

~

damping, etc. which meet or are.more conservative than I FSAR requirements.

~2. New FRS were broadened to + 10 percent which was the original intended percent broadening commitments.

3. The Design Basis Earthquake was used 'to evaluate b ' safety.

4. Detailed criteria were developed for FRS generation,

'FRS. comparison, and subsystem evaluation that provided consistency between engineering organizations.

5. Highly. qualified, independent consultants provided comments and. reviews for this safety evaluation.

6. Only_a small percent of'the new FRS were found to show

~ sufficient-deviation to warrant an evaluation of equipment and structural subsystems. The primary reason that most of these new FRS exceeded the original.FRS was modeling changes that-more accurately reflect existing conditions.

7. No equipment or-structural subsystems were found to be adversely affected by the changes in the FRS.

22

'} }

e

.e  : e

18. ' Comparison of FRS'for piping systems showed no-

, significant increase in. spectral accelerations,

@d :'

3 -therefore, there'was no. adverse.effection piping .

3 csystems.

-9. Comparison of FRS for-cable tray supports showed no-significant~inc,rease in spectral accelerations; g (therefore, there was no adverse effect on cable tray

-supports.

IV. -CONCLUSIONS-1

.The-conclusion'of this Part 21 evaluation' was.that no reportable-condition ~in accordance with the criteria of

, ,Part.10 CFR 21-existed-for the floor, response spectra for-

>Edwin I. H tch Nuclear Plant --Units 1 and 2. This conclusionsvas' based on a very' extensive safety evaluation with comment ~and review by. independent. consultants.

4

/ .

ll. REFERENCES.

< 1. - Seismic Analysis;of Structures and: Equipment for-Nuclear Power Plants, BC-TOP-4, Revision 4.

.A

2. - PVEC Technical Committee on Piping Systems of the Pressure Vessel Research Committee, Progress Report on

, Damping-Values,,1983.

3. " Cable Tray and Conduit Raceway Seismic Test Program",

1

. prepared for,-and in collaboration with, Bechtel Power

' Corporation, Los Angeles Power Division, ANCO

- Engineers, INC., December 15, 1978.

t 23

, , , ,. , s,

., 41 Ir

[.

~

1.

g 7.

.L

.i 4

VI TABLES AND FIGURES h!-

k-5 f

h t

i L.

4-r I

I e

f T

i t'

}

T' z

I k

) .,

si i

f i

1-f.-

h l

[: 0364n

{

$ 24 i -

I t, _m_ s

. i-a.,

o i4 PLANT HATCH UNIT 1 RESULTS OF COMPARISONS OF NEW FRS TO ORIGINAL FRS -

FOR 5 PERCENT-DAMPED DBE SPECTRA

- NUMBER OF NEW FRS THAT WERE

, NUMBER DETERMINED TO HAVE STRUCTURE OF FRS SIGNIFICANT EXCEEDANCES l REACTOR BUILDING 42 0

CONTROL BUILDING 10 3 INTAKE STRUCTURE 6 0 DIESEL GENERATOR BUILDING 2(2) 2

! STACK (1) -

~

i

! TOTAL 60 5

. NOTES:

, 1. ALL MAIN STACK FRS WERE DEVELOPED USING THE UNIT 2 SEISMIC CRITERIA WHICH IS j MORE SEVERE THAN THE UNIT 1 CRITERIA FOR THE MAIN STACK. *

2. THE NORTH-SOUTH AND EAST-WEST FRS WERE ENVELOPED TO FORM A HORIZONTAL FRS FOR THE DIESEL GENERATOR BUILDING. - -

l

m. e, TABLE 1

~

x a

PLANT HATCH UNIT 2 RESULTS OF COMPARISONS OF NEW FRS' TO ORIGINAL FRS FOR 5 PERCENT DAMPED DBE SPECTRA NUMBER OF NEW FRS THAT WERE NUMBER DETERMINED TO HAVE OF FRS SIGNIFICANT EXCEEDANCES REACTOR BUILDING 69 0 CONTROL BUILDING 15 0 INTAKE STRUCTURE 9 3 DIESEL GENERATOR BUILDING ' 4(1) 'O STACK 6(2) o_

TOTAL 103 3 NOTES:

1. THE N' ORTH-SOUTH AND EAST-WEST FRS WERE ENVELOPED TO FORM A HORIZONTAL FRS FOR THE DIESEL GENERATOR BUILDING.

2. COfAPARISONS FOR THE MAIN STACK WERE MACE FOR 3% DAMPlNG SINCE THIS WAS THE HIGHEST DBE DAMPING VALUE FOR WHICH FRS WERE ORIGINALLY GENERATED.

TABLE 2

i

. PLANT HATCH UNIT 1

SUMMARY

OF i EVALUATIONS PERFORMED BY LOCATION i

1 l

1 DID EQUIP. AND DID PIPING DID CABLE SUBSYSTEMS HAVE TRAY SUPP.

HAVE TO BE TO BE HAVE TO BE STRUCTURE /EL REVIEWED? REVIEWED? REVIEWED?

CONTROL BUILDING l EL.112 YES NO NO 4 EL.130 YES NO NO l EL.180 YES NO NO DIESEL GEN. BUILDING j . EL.130 YES NO NO -

l EL.150 YES NO NO m ,.oc. ,.

TABLE 3 l . --

..t

'~

PLANT HATCH UNIT 2

SUMMARY

OF EVALUATIONS PERFORMED BY

! LOCATION i

i

! , DID EQUIP. AND DID PIPING DID CABLE SUBSYSTEMS HAVE TRAY SUPP.

l HAVE TO BE TO BE HAVE TO BE STRUCTURE /EL REVIEWED? REVIEWED? REVIEWED?

l INTAKE STRUCTURE EL. 88.75 YES NO NO EL.111 YES NO NO ,

EL.128 YES NO NO T163146AC006-t t TABLE 4

k i ..

i PLANT HATCH UNITS 1 AND 2 PERCENTAGES OF SAFETY RELATED' EQUIPMENT LOCATED AT FLOORS OF INTEREST WHICH WERE EVALUATED FOR THE IMPACT OF THE NEW '

FRS ON SEISMIC QUALIFICATION OF THE EQUIPMENT NO. OF SAFETY NO. OF SAFETY NO. OF SAFETY RELATED TOTAL NO. OF RELATED ITEMS RELATED ITEMS CHOSEN ITEMS ADDED SAFETY RELATED

BUILDING /EL. AT THAT ELEV. BEFORE WALKDOWN - DURING WALKDOWN ITEMS REVIEWED NO.  %

UNIT #1

CONTROL BLDG.

EL.112 30 8 2 10 33

, EL.130 43 9 10 19 44 j EL.180 70 27 1 28 40 1

UNIT #1 DIESEL BLDG.

EL.130 111 36 12 48 43 EL.150 54 27 0 27 50 i UNIT #2

INTAKE STR.

EL. 88.75 30 9 0 9 30 EL.111 155 35 9 44 28 1

EL.128 4 2 0 2 E0 l

se v.sas se TABLE 5 i

.m -

y My R E

a v

q. ~

1 r- ,

j T" -

Page 1 of 2

-TABLE 6 PLANT HATCH LMITS 1 AND 2

RESULTS OF EVALUATION CF THE IMPACT OF NEW

'FRS ON EQUIPMENT SEISMIC QUALIFICATION

1. -LOCATION' OIO THE NEW FRS IMPACT 3:c

MPL: (See THE SEIS:11C QUALIFICATION

! No.. EQUIPMENT /VENCOR ' Note 2) 0F EQUIPMENT 7

- 1 Z41'-N019 ' Exhaust Flow Switch /Ericksons IC112 See Note _1

.1C71-N003A. B Root Valves /Hancock . IC112 No ..

IPS2-F102A. B Gate Valve /P-W Industries - IC112 No

1752-F201 Control Valve / Fisher Centrols . . 1C112 No-

1R42-5001A. B 125/250v Station Battery /ESB. Inc. 1C112 No 1 Z41-C014.15 Station Bat. Room Exhaust' Fan /Ericksons IC112 : No

.IC71-PC03A-F. R.P.S. Panel /ANCO , IC130 30 1H21-P246. 600v Switchgear Panel / Reliance IC120 No i!H21'-P248 - 250v DC Switchgear Panel / Reliance IC130 NO

!R11-5041' Essential Transformer /ZINSCO IC130 No IR21-5042 600v Bus /GE- .

. IC130 No IR22-5017 250v DC Bettery Switchgear/GE IC130 .No 1R23-5004 Station '$ervice Switchgear. Transformer /GE . IC130 No IR25-5001 125v OC Cabinet /ITE Imerial 1C130 No 1R25-5071 - 277/480v AC Cabinet /ITE Irperial -IC130 No 7' IR42-5029: 125v Battery Charger /ES3, Inc. 1C130 No

.;242-5030 125v Battery Charger /ESS. : Inc. 1C130 No

?!Z41-N01EA-C Fan 7-Discharge Flow Switch /Ericksons 'IC130 See Note 1

H21-7271 Centrol Rocm HVAC Panel / Reliance '1C130 30 x1741-F42CA. 8 Globe Valve /P-W Industries 1C120 No

-1741-N520 CP Switch /ITT-Barten = . IC180 NO iR24-SC02 600/208v FCC/Allis-Chalters ICIE0 No No iZ41-AC01B Centrol Room Air Accumulator /RECO 1C180

.1Z41'-8003A-c Centrol. Room AHU/Ericksons - - .

ICIE0 No 4 IZ41'-BC02A-C control Recm Cendensing Unit /Ericksens IC120 No

.1Z41aB010A-C Centrol Rcom Elect. Heating Coll /Ericksons : IC180 NO IZ41-C012A. B Centrol Rcom Supply Fans /Ericksons IC180 No 21Z41;CC04A. B Centrol Room Filter Train /Farr IC120 No

' ? 41-F012A, 8 Actuator /Settis; A.0 Dacer/Erickson '

1C150 No 1Z41-F100 - Cteck Valve /Reckwell IC180 No 1241-N017A-C Fan 8 Disc. Flow Switch /Ericksons 1C120 See Note 1 1:41-4022A. 3 Chlorine t'enitor/Wallace-T8ernan ICISO NO

-! 41-V0 F0iG ' Control Recm Sucaly valuee Dameer/Ericksons ICISO No

'1H21-7202- Diesel Generater Relay Panel / Reliance 10130 No lH21-7232 Diesel Generator Relay Panel / Reliance 10130 NO

. ;R22-5005 4150v Sta. Ser. IE Switchgear/Westingnouse- 10130 ' No 1224-5025 600/208v FCC 1A/A111s-Chaleers 10130 No IR24-5027 E00/208v FCC/Allis Chaleers 10130 No 1R25-5029- . AC Cabinet /ITE toerial 10130 No 1R34-5004A-C Neutral Resister /GE . 10130 See Note 1

, 1R42-5002A-C' 12Sv Ofesel System Battery /ESS. Inc. 10130 No

+ ,

- -w,.

w ,

R O I

,g .

t 4

Table 6 Page 2 of 2

,PJant Hatch Units 1 and 2 Results of Evaluation of the Impact of New

- FRS on IDuipment Seismic Cualification 1

IDCATICN DID THE NEW FRS IMPACT MI L ' -(See 'INE SEIS4IC CUALIFICATICt

' NC.. - ECUIINENT/\TEDOR Note 2). . CF EQUIINENF?

4 II 42-S032A-F - 125v Battery Charger /ISB., Inc. ID130 - No _

c1241-BC01A-H,J Unit Heater Mark H-1/Chrcoolox 1D130 No e : D 41-B002A-H,Ja Unit Heater Mark H-2/Chrcmolox 1D120 - No '

l>41-C027A-C Iouver/Constru: tion Spec. 1D130 No .

1>41-N004A-C - Fan Hermostat/Foneywell '

1D130 No

. D 41-N016A-C Fan Firestat/Honeywell 1D130 No if43-0000A-C Diesel Generators /Fairbanks Morse 1D130 No

.~1141-CC02A-F Ecot Fan /Inoustrial Air 1D150 No D 41-CCO3A-C ' - Roof Fan / Industrial Air ID150  ::o .

1)41-CC04A-F Root Fan /Inoustrial Air ID150 No as . D 41-CCC6A-F . Roor~ Fan / Industrial Air 1D150 No D 41-CCCEA-F- Scot Fan /Inoustrial Air 10150  ::o 2141-DC01A,B Service Water Strainers /Adans 2I088 No 2141-F310 . Butterfly Valves /Pratt 2IC68 No

-2141-F842A,B' Globe Valves /Yarway 2IC88 ro 2141-N322A-D Futps Disc Flow Elecents/BIF 2IC88 24 2)41-20C6A,B Unit Heater /Chrm.olox 2Illl No

2) 41-BCG5 B , C Unit Heaters /Chrm olox 2Illi !b

D 41-N002 Fan 'Iternostat/Eeneywell 2Illi No 1)41-N022A Fan Firestats/Eonep ell 21111 *b 3 ll-CC01A-D - Service Water Ptrp anc Motor /Jchnston 2I111 ho 21ll-F012D - Cate Valve /Kalworth - 2Illi in 2111-Fl:6A, B . Press. Ccntrol Valve / Fisher , -

2Illi S:o 2I11-F203A-D. Chect Valve /Rockwell 2Illi 24 2111-F207D -Press. control Valve / Fisher 2Illi No 3 11-F209A, F ' Air Release Valve /Johnston 2Illl to 2Ill-F222A-D Cate Valve /Yarway 2I111 tb 2141-CCCIA-D Serv. Water Futp & Motor /Johnston 2Illl No 2141-F30.:A-D Butterfly Valve /Fratt 2Illi No 2I 41-F31' A-D Solenoid Valve / Target Pock 2Illi to 2141-F334A, B Control Valve / Fisher 2Illl No 2141-F850 Globe valve /Yarway 2Illi to 2I41-N328 .0003 CPS /ITT-Batten 2Illl No

2I41-N286 ' Leve1 Transn1tter/Ross ount 2Illi b

'2I41-N387 Press. Transnitter/Fose ount 2Illl  !:o 2124-SCC 9 600/208v NCC/Allis-Chabers 2Illi !b 2124-5010 600/2C8v MCC/Allis-Chalmers 2Illl  ::o

_1 D 41-CC09B, C Roof Fan /Incustrial Air 2Il28 No t<IIS:

1. Not enough infor: nation availacle to cecide the i:rpact of new tRS on setante qualitication.

2. The first digit identifies the unit, the second digit identifies the building (C = centre Building, D ' = Diesel Generator Building, ano I = Intake Structure), are the last thre

- digits identify the floor elevation.

9

. i, k

i l PLANT HATCH UNITS 1 AND 2 ' .

RESULTS OF EVALUATION OF THE IMPACT OF .

NEW FRS ON CIVIL SUBSYSTEMS i NUMBER OF. DID THE NEW FRS IMPACT '

ITEMS THE EXISTING SEISMIC ADEQUACY OF

CLASSIFICATION EVALUATED ANY OF THE CIVIL SUBSYSTEMS

1. EQUIPMENT SUPPORTS i

A. ANCHORAGE EVALUATED 6 NO -

l B. SUPPORTS AND ANCHORAGE 12 NO i , EVALUATED -

i i

11. NON SEISMIC CATEGORY I 8 NO ,

SUBSYSTEMS 0) 111. CONCRETE BLOCK WALLS 6 NO l IV. CONDUlT SUPPORTS 4 NO V. HVAC SUPPORTS 3 NO VI. CABLE TRAY SUPPORTS 6 NO l

NOTE:

J

1. THESE SUBSYSTEMS WERE EVALUATED SINCE THEIR COLLAPSE COULD DAMAGE SEISMIC CATEGORY I EQUIPMENT. THESE ITEMS INCLUDE SPRAY BARRIERS, HOIST, PAGE SPEAKER, CONTROL ROOM CEILING, CATWALKS, AND GYPSUM BOARD WALL.

e i TABLE 7 I

1

4 .g

, l. O*5_f 8  ; N,  ;

Lt;:  ;

,a.  ;

. .< +s  :-l '

-l S.;. . ,{

?

1 4

I v A!

!I. '

., j ,

ORGANIZATON FOR PART 21. ,'

opco . y '. ,

.z

'i

. ,h SCS l NUCLEAR SAFETY & FUEL g

CONSULTANTS

! , R.P.KENNEDV D. F. LANDERS "*

i SCS' NPS . HATCH 4

• 1 1

1 I J

i - BECHTEL CRITERIA SCS

-POWEP +------ -------- .+

7 CORP,

+

?

i . .

9

• REACTOR CONTROL SLDG UNIT s1 BLDG I I

) .>.

4 DIESEL REACTOR CENERATOR

BLDG UNIT 82 BLDG

}

ASAIN INT AKE '

s STACK $TRUCTURE

~

l

)

i FIGURE 1, 4 s,' .

e t . .-

~

.i r ~.

i . > -

~'

i

SUMMARY

OF ENGINEERING "

l METHODOLOGY FOR SAFETY EVALUATION i

f 1. REVIEWED ORIGINAL CALCULATIONS ,

i 2. MADE IMPROVEMENTS (ENHANCEMENTS TO MODELS

! AND INPUT) -

l 3. DEVELOPED CRITERIA THAT WAS FOLLOWED BY BOTH i ORGANIZATIONS

' 4. GENERATED NEW FRS MEETING INTENDED LICENSING ,

COMMITMENT (l.E. 10%)

5. COMPARED NEW FRS TO ORIGINAL FRS l

6. EVALUATED THE EFFECT OF NEW FRS ON j SUBSYSTEMS l 7. TECHNICAL INPUT WAS PROVIDED BY CONSULTANTS l

b FIGURE 2 i

i t

l .

2 j

i j

l ENGINEERING ~ CRITERIA INPUT PARAMETERS FOR REANALYSIS l CRITERIA FOR REANALYSIS OF STRUCTURES AND

! GENERATION OF FRS 1'

CRITERIA FOR COMPARISON OF FLOOR RESPONSE l SPECTRA

, CRITERIA FOR SAMPLING AND EVALUATING EQUlPMENT CRITERIA FOR SELECTING AND EVALUATING

) STRUCTURAL SUBSYSTEMS CRITERIA FOR SELECTING AND EVALUATING PIPING CRITERIA FOR SELECTING AND EVALUATING CABLE

,l TRAY SUPPORTS f

e L ... .

FIGURE 3

~- --

. ]

e o' q@ >

SOUTHERN COMPANY SERVICES HATCH NUCLEAR PLANT UNIT 1 1 92 1.60 CONTROL BUILDING - ...-. UNIT 1 ORIGlhAL DESIGldENVE LOPE SEISMIC RESPOriSE SFECT.4A UNIT 1 NEW DI SIGN ENVELOPE NITH MASS PCINT NO 1 10% BR3ADENIN(iFACT4>R FLOORI LEVATION 1 12.0 U T.

DESIGN BA91S EARTHOUAKE RESPON LE DIRECTIOff E AILT49 CS

D AMPING RATIO 5%

?

c E

z 9 @.96 7

5 8

8.64 --

1 1 *

/~ %

/ 's 0 32

, -- y \s

,// e 6 7-h _

e 0.0 10'3 2 3 4 5 6 7 8 91 2 3 4 5 6 7 8 9 10 20 30 40 50 50 70 80 90 100 FREQUENCY (Hal

, FIGURE 4 l

c.

6' SOUTHERN COMPANY SERVICES HATCH NUCLEAR PLANT UNIT 1 3.75 3.15 CONTROL, BUILDING . . - UNIT 1 ORIGINnL DESIGN ENVELOPE SEISMIC llESPONSE SPI CTHA UNIT 1 NEW DElilGN ENVE LOPEVPITH MASS POINT NO. 2 10% BRCADENING FACTOR FLOORI .EVATIDN130.0 FT DEStGN 13 ASIS EARTHOUAp.E R ESPONSE DIREC TION E As r W EST DAMPIN(1 RATIO S%

?

O 3

z 9 1.89 7

5 d

8 1.26 m)

/

)  %

\

/ \

1

/ ,

0.63 / '-

/

j

\

-- ~

\ \-

/ -

.J .

Q- ____

C2 13'I 2 3 4 5 6 7 8 91 2 3 4 5 6 7 8 9 10 20 30 -40 50 60 70 80 90 100-FREQUENCY (Hal FIGURE 5 .

a

.,. o hc 0

aO 0 1

- 0 9

0 8

0 7

- 0 5

0 5

EH PT OI R 0 l W O 4 E ET VPC -

NOA ELF .

i

. 0 3

V G I

h GI N N I SEN E

. D GNE D _

LI A, B A 0 O 2 N E R I D B G

IRE W%

s'NN ON1 0

1 1 TT I INN

,Q \

UU 0 1

. \

1l 9

. ) -

/ 8 j 7 S1 /

E /I 6

CI T .i I

/

VN r f 5 RU /

ET l a

SN 4 H YA l

NLP Y C

8

- N E A -- E R PR 3 U U MA s Q E

GI OE - R F

CL F C -

NU , 2 R

E N ,

HH TC UT "s*

OA

• SH 1

• 9 b

T

• 8 L

O

• L

, . 7 H

A ET a H 1

>H 6

T FAO O

C L OUN P

I 8 N

[I 5 1 H O GS N T NEE 7 ) RT I %

S _

I 4

DN . I A C O LO I

PNA OTEE S RT I

U ES T VI I

A B N E.S A:D R _

I fl I l 3 8 SG ORICPR OE N NN TMSO NSS OSSP M G OI I

P OE A LE E A I

CSMFDRD 2

I

- '0 1

4 6 7 8 9 0 7 1 3 5 7 0 2 1 0 0 4 3 .

.o* 297 5 8 .

+.

r .

o:

SOUTHERN COMPANY SERVICES HATCH NUCLEAR PLANT UNIT 1

. e20 DIESEL GENERAL OR81JtLDlNG SEISMIC NESPON5 E SPECTRA UNIT 1 ORIGINAL DESIGP ENVEB OPE-UNIT 1 NEW DEEIGN ENVI; LOPE MASS POliiT NO. 3 FLOOR Et EVATION 1300F" DESIGN 8 GIS EARTHCUAKE

• R ESPONSI' DIRECTION HOR 12 01JTA L DAMPlNG R ATIO 5%

2.30 O

3 2

0 2.13 4

L:

U O

W 3.40

- s

1. ,s'. \

\

/

C.70 ,! '

F j

l  %

l  %~

C00 - --

-1 to 2 3 4 5 6 7 3 gy 4 5 6 7 8910 20 30 40 50 60 70 80 90 100 FREQUENCY (Hal

. FIGURE 1 .

=

-en 6'.

SOUTHERN COMPANY SERVICES HATCH NUCLEAR PLANT UNIT 1 c.25 C375 .-- -

DIESEL GENERA'rOR BUILE'INC -..-- UNIT 1 ORIGIN 4LL DESIGN ENVELOPE SEISMIC IIESPONliE SPI CTRA UNIT 1 NEW DE ilGN ENVI LOPE MASS PO6MT NO. 2 FLOOR El.EVATI )N 1510 FT DEssGe4 aASIS ELRTHOUAD E R ESPO'JS E DIRECTION HOF 82 04T;iL D AMPING RATIOS %

3.50 O

E 2

0 2.625 4

S 8

1.75

. M

- \

" , " \

\

CS75 ,-

f f f \

/  %

,I marg _ ____

"~ ~ -

_..-C/

0.0 ---

10'I 2 3 4 5 6 7 8 91 2 3 4 5 6 7 s 9 10 20 30 40 50 60 70 8090 100 FREQUENCY (Hal FIGURE 8 * .

a.

C; o

o 41 a

SOUTHERN COMPANY SERVICES HATCH NUCLEAR PLANT UNIT 2 '

c.is INTAKE 5TRUCT JRE ..- - UNIT 2 ORIGIN, LL DESIGh ENVEL OPE SEISMIC llESPONsE SPI:CTRA MASS POs NT NO. 3 UNIT 2 NEW DE ilGN ENVI! LOPE tflTH FLOOR E . EV ATlDN 8815 FT 10% BRt ADENING FACTOR DESIGN EASIS EARTHOUA> E R ESPONSE DIREl; TION E AS T-Wl:ST i

D AMPING R ATIOS%

0.44

?

o E

z

' C.0s j

U d

8 2.12

~ ~

I l i 1 1 36 1 1

f

' \

- - [ \ w ._ _ . . ,

C00 --

10' ' 2 3 4 5 6 7 ae1 2 3 4 5 s 7 s e ts 20 30 40 50 to 70 3090100 FREOUENCYIH4 FIGUR E 9

-w c,

o, 3

SOUTHERN COMPANY SERVICES HATCH NUCLEAR PLANT UNIT 2 8.14 C.80 INTAKE 5TRUCT'JRE - ... UNIT 2 ORIGIN.LL DESich ENVEL OPE SEISMIC llESPONsE SPl!CTHA UNIT 2 NEW DE ElGN ENVI!LOPEHITH MASS POINT NO. 4 10% SRilADENING FACTOR FLOOR El.EVATION 111.0 FT DESIGN HASIS EJ,RTHOUAD:E R ESPONSE DIRE ' TION EASr-WliST DAMPINC R ATIO S%

5.44

?

U f

z 9 COS 7

Li 8

f" 7 2.72 , g i i i I 1

/D i

us4

\ ,

j t a '

% ~ ~ -- ...

g - -

10' ' 2 3 4 5 6 7 8 91 2 3 4 5 s 7 s 9 to 20 30 . 40 50 60 70 80 to 100 FREQUENCY (Hal FIGURE 10

wj ,

O.O%'

0 0

1 0

0 0

8 0

7 0

5 0

5 EH PT OI R P 0

L VO 4 E _

ET VPC _

NOA ELF 0 NIV G ,. 3 GI N N SEN E

I DG NE D  %

L R A.

h ~ 0 J

C ~ 2 N E R I D B G

I W%

RE ON1 0

22 TT I

NNI UU'

\ W 0 1

9

- ('

- 8 i ii1h1

- ~ 7 S

E 2 " ;iI l I1I CI T I 's 6 I

VN 5 RU /

ET

' s.

l a

SN 4 H YA (

NLP f C Y 1 1

E A N E R PR 3 U U MA Q G OE I

E F R

CL F C -

N R

U -

2 E N HH TC UT OA SH s_-

- 1 9

8 T

S E

7 W

A T;T E -

F F l S 6 T AA C 0.U E 5

2 O N E 5 P 1 H O ES N T RE RT 5%

I JE 5O I AC O 4 TN .

CO UPNOTEE A S RT I RS T VISD A I

T E R NLE AE R 5 I I E SC 3 EC O I PR E N NN K

A MSOIGOI P

P T I SSOSSM A LEEA NE I SMFDRD 2

I

'0 1

0 0 0 0 0 0 2 5 m 1 5

4 7 1

0 0

0 8 E . 3 1

?O 3 z9 <$::8<

1  : 'I d ,

'e g ..

l f-~b e

-t ,

A.

f DAMPlNG VALUE FOR SEISMIC ANALYSIS OF PIPING (Applicable to both OBE & SSE, independent of Pipe Diameter) i I

. l .

2 4 -

I s

l

- i

$3 -

l E -

l 0 '

g 2 -

,I ,

ji i I

<t 1 I o 3

_ l i I I I I I I

' ' ~

O 0 10 20 30 FREQUENCY (Hz)

FIGURE 12 1

.l l .

6.i

- 4.

4 i

.l ,

1 LOWER BOUND DAMPlNG AS A FUNCTION OF INPUT ZPA .

i .FOR CABLE TRAY SUPPORTS j

l~

. i i i i i 50% TO FULLY LOADED TRAY j

^

,20 3-TESTED LIC5NSEbf~

i b '

[ Y 5 0 16 1 r l

1

[ 15%

1E 12

/

g 8 p_ ___.__ _____ __ __ . _ _ _ _ .__.

l

/

0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 l

! INPUT FLOOR SPECTRUM ZPA -

! NOTE: 1. VALUE KNOWN TO BE USED FOR OTHER NUCLEAR POWER PLANTS.

l FIGURE 13 4

i

, .-- - ,