ML20127N686

Forwards Request for Addl Info Re Seismic & Structural Design of Plant to Aid in Review of Application for Provisional OL

ML20127N686
Person / Time
Site:	Monticello
Issue date:	05/20/1969
From:	Morris P US ATOMIC ENERGY COMMISSION (AEC)
To:	Mcelroy D NORTHERN STATES POWER CO.
References
NUDOCS 9212010377
Download: ML20127N686 (16)

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i May 20, 1969 Docket No. 50-263 w

I Nor:n.rn States Power Company 414.acollet Mall Mirce. : polis, Minnesota $5401 At: r.: ion:

Mr. D. F. McElroy Gen:lemen:

To continue our review of your application for a provisional operating license for the Monticello plant, we need additional information in regard to the seismic and structural design of the plant. This matter was discussed with representatives of your company and the General Electric Company at meetings held on April 1 and 2, 1969.

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The structural and seismic design information presented in the Fincl Safety Analysis Report (FSAR) is insufficient to permit a de:ermination of the adequacy of the Monticello plant to withstand seismic loadings.

In many respects the information presented in the FSAR is less informative than that presented in the earlier Preliminary Safety Analysis Report (PSAR).

However, your letter of transmittal applying for an operating license states that the FSAR "... supersedes in its entirety the application dated July 25, 1966."

The FSAR does not state in sufficient detail-how the structural and seismic design objectives presented in the PSAR and amend-ments were translated into the final design..Accordingly, we request that the appropriate sections of the FSAR related to the seismic design of Class I structures, equipment, piping, 8.

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instrumentation and controls be = revised.. We beliave thatit

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require to continue.our review of your application.

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To assist.you, we have attached a summary of the type of informa--

tion that should be included in the appropriate revisions to the i

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Please contact us if you have any questions regarding this requast.

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Sincerely, (Original signed by Peter A. Morris)

Peter A. Morrisi Director Diviinion of Reactor Licensing i

Enclosure:

Sunnary Info. for

.aAR Revision Distribution:

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RPB-1 Reading C. K. Beck.

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SUMMARY

OF TYPES OF INFORMATION TO BE INCLUDED IN REVISED SECTIONS OF FSAR RELATING TO STRUCTURAL AND SEISMIC DESIGN OF MONTICELLO PLANT A.

SEISMIC DESIGN An indication of the criteria defining the functibnal adequacy for each type of Class I structure, ecuipment, piping, instrumentation, and i

controls, as related to information presented in the FSAR that Class I

,l structures and equipment are designed in such a way that for a ground acceleration of 0.12g a safe shutdown can be achieved.

-- With respect to the methods used in the seismic analysis for Class I structures and equipment, the f ollowing saould be provided:

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- Verification that the earthquake record employed in the seismic analysis (i.e., using as indicated in the FSAR a time history approach

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using the Taft, July 21, 1952, N 69' W earthquake record appropriately scaled) leads to a spectra of the type presented in Plate 3 of the t

I seismic criteria (first portion of ppendix A), which was the basis of the specified criteria.

Also, e discussion as to whether slight changes in the time history input, or alternatively slight changes in the method of modeling the structure could lead to any significant changes in the design values arrived at, and whether an approximate i

check of the results obtained by the time history approach was made.

1 An indication for each Class I structure and equipment of whether the response spectrum method or the time history method of analysis was used for seismic design.

If a modal analysis has been used, an indication of how many modes have been considered and a description of how the damping was evaluated for each mode.

-- A comprehensive discussion of the loading combinations, and the applicable stress and deformation limits employed in the design of Class I struc-tures, equipment, piping, and instrumentation and control systems. Also, a listing of the specifications and/or codes related to the information presented in the FSAR that these structures and equipment are designed in such a way that for the combination of normal loads plus design earthquake, the stresses are within code allowable. Also, an indication as to whether for some elements, a stress increase has been used, as permitted by the codes.

In the PSAR (Section 5.3.1.2) it was noted for the maximum earthquake that ".

. where calculations indicate that a structure or piece of equipment is stressed beyond the yield point, an analysis is made to i.__

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determine its energy absorption capacity.

In addition, the design is reviewed to assure that any resulting deflections or distortions will not

,r, prevent the proper functioning of the structure or piece of equipment."

Additional information is needed to indicate whether such conditions were encountered at any point in the desigo and if so, what limits on stress and deformation criteria were adopted to ensure the adequacy of the design to meet the intended design criteria.

A description of the mathematical models used for the seismic design of each of the Class 1 structures, equipment, piping, instrumentation and controls systems, and an explanation of how the elasticity of the structures, and the damping have been evaluated.

Also, a discussion of how closely the mathematical models represent the actual conditions, especially the effect of non-linear behavior of the

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actual structures, piping and equipment; ef fect of appendages (small masses elastically attached to large masses) such as vent pipes and header, equipment hatch, and personnel lock; ef f ect of clearances (gap) at equipment restraints and supports; and effect of variable f riction.

-- Justification of the assumption used in the seismic analysis of the primary containment structure that the drywell is completely fixed on l

its foundation.

-- A detailed discussion of the seismic design of the plant stack, torus, ring header and its supports.

The seismic design of these features was omitted f rom the FS AR.

-- Justification for subdividing the Class I structures and equipment into the three categories; rigid, resonant, and flexible, and an explana-tion of the manner in which these categories are used in the final design.

An explanation of how the interaction between soil and the reactor building has been provided for in the seismic analysis and the design of the building and whether non-linear behavior of soils has been considered; e.g., details as to how spring constants, such as those designated K3 and K4, which represent the foundation stiffness and lateral resistance (soil-structure interaction), were obtained for use in most of the analyses presented in Appendix A of the FSAR.

-- Details on the foundation design and construction; e.g., the conditions that were encountered and the a tual procedures that were employed in constructing the foundations for the plant.

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4 listing of the amplification factors resulting from seismic analysis, 4

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ts compared with the ground motion for the reactor, recirculating pumps, Ulass I piping, and spent fuel pool.

Clarification as to the damping values used it the final design.

In Appendix A it is noted that damping values of 10 percent of critical _are to be employed for ground rocking modes of vibration. On the other hand, our records indicate that in reply to Question 8.8 of Amendment 6 of the PSAR, Northern States Power stated that a value of 5 percent

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vould be used.

-- Clarification as to whether as noted on page 12.2.2 of the FSAR that I

Class II structures were designated by the Uniform Building Code for Zone 1 conditions.

This appears to be at variance with the material presented in answer to Question 8.10 of Amendment 6 of the PSAR wherein Northern States Power indicated that a seismic coefficient of 0.05 would be used I,

for Class II structures and equipment.

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Information to show that at points where structures and/or equipment

-l are interconnected, the dynamic deformations are compatible; e.g.,

(a) for horizontal restraints of the reactor at elevation 994'-2", and of the drywell and the shield at elevation 992'-5-1/2",

(b) for the drywell, at the shear lugs between the drywell and the reactor building at elevation 992'-4-13/16".

The design criteria supporting the statement that parts of Class II structures covering or supporting Class I equipment have been designed as Class I structures.

4 B.

STRUCTURAL DESIGN

-- For the drywell, clarification of the following:

- Method of evaluating the jet forces and the area subjected to their effect.

- How the maximum metal temperature of 300*F was established for jet impacted steel plate and an evaluation of the corresponding thermal stresses in the shell.

- Why the temperature of the steel plates was reduced to 150*F when jet action is considered with design internal pressure, and what this pressure is.

- Why local yielding has been permitted when the shell is backed up by concrete.

This criterion is not a code criterion. How was it established that a rupture will not occur? Where the shell is not n

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to go u; to 0.9 of the yield point; this is not allowed by the code.

Note oli of the " loads which were combined when this criterion was used.

-- For the design of the torus, consideration of the followingt A justification of the value of 21 kips for the jet force at each downcomer pipe in the torus.

- A description of the stress criteria and design methods used for the derign of the torus for jet forces. A discussion of whether other jet forces exist in the torus in addition to downcomer pipe jet forces.

- A description of how the flooding of the drywell and the torus has been considered and combined with seismic loads. Also, an indication of the corresponding critical stresses.

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-- With respect to the penetrations in the drywell and torus, the following:

- A discussion of the applicable design criteria and the load combina-tions used in the design.

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- A description of the stress analysis methods used to evaluate the stresses in the shell; penetration sleeves, bellows and guard pipes;

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and process piping at penetrations, their anchors and supports.

The significance of the statement in the FSAR that ".

. the contain-ment vessel was code stamped for the design pressure and design tempera-ture";

i.e., does this mean that it is not a code vessel for other loads such as seismic loads, jet forces and equipment loads?

-- An evaluation of the capability of the f acility, including the stack, to withstand a tornado with 300 mph rotational velocity, 60 mph trans-l lational velocity, and a 3 psi pressure drop in 3 seconds. Also, an indication of whether stack failure can endanger any Class I systems or structures required for safe shutdown.

-- An explanation of how the gap between the drywell and the shielding con-crete outside of the drywell is drained and vented, since it appears that strips of polyurethane foam used in construction have been left in the gap at specific elevations.

Indicate the temperature stresses in the concrete walls of the spent fuel pool under normal operating conditions, and the provisions made to limit cracking of the concrete.

-- A discussion of whether means will be available to monitor possible settlement of Class I structures and equipment.

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REACTOR. PIPING AND EQUIPMENT DESIGN

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-- A detailed explanation and supporting analysesitof show how ' the 'of fecijf P

a seismic disturbance on the Class II part of 'the' main steam lineslhas been' taken care of in the design of the Class I part, especially, for:the

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anchors. and the valves.

A comprehensive discussion and details in regard to the design of'the-

; H reactor internals and primary system piping including but not '11mited-to i the following:

- The extent and findings of the analyses by which the conclusions stated in Section 3.6.3 of the FSAR are reached; i.e., it-is stated the reactor vessel internals - are designed to maintain a reflooding capability following a loss-of-coolant accident and that the :

internals are also designed to preclude a failure mode which would result in any part being discharged through the main steam line in -

i the event of a steam line break.

In your discussion pertaining to the directly added simultaneous

- peak loads resulting f rom normal operation plus the worst loss-of-coolant accident plus the design basis earthquake, 'and for the combi-nation of the normal. operating loads plus the peak loads from the worst loss-of-coolant accident include:

(a)

The methods of analysis (elastic, elastic-plastic, limit).

(b) The limits to which the critical components were - evaluated (stress, strain, deflection, buckling) with numerical values l

for several critical items.

(c)

For each method of analysis, relate the possible errors in the -

analytical method and the possible errors-in loads used'for the' stress calculations to the margin of safety which is being

_ p rovided.

Information as to the frequencies and mode. shapes.for the " coupled' system" related to the design analysis'of the reactor pressure vessel noted in Section 4 of Appendix A.

Informar. ion showing how the seismic-design of Class I tunnels. and.under-ground piping and cables entering or leaving a structure were handled, since the seismic response within and outside the structure is quite different.

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4y Nathods of analysis'used.

' Strass' limits to which the. piping s'ystems were designed.

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<,79 Supporting systems, snubber locations, etc.

Seismic factors, amplification factors and other pertinent factors-used.

Accommodation of pipe whip.

_j Method used to support the recirculation pumps. Also, an indication of the design criteria, materials, and design methods used for these supports.

Indicate the potential for pumps to become' missiles under the combined action of earthquake and jet forces.

A justification of the adequacy of the se_ismic design of the battery racks.

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q Northern States Power onpany 414 Nicollet Mall Minneapolis, Minnesota.5401 Attention:

Mr. D. F. McEl oy Gca:lemen:

To con:inue our review'of your application for a provisional oparating license for the Monti ello plant, we need additional

'information in regard to the sei mic and structural design of the plant.

This matter was discu sed with representatives of your company and_the General-Elect ic Company at meetings heldt on April 1 and 2, 1969.

The structural and seismic design int rm,r4nn nvacanted in the Final Safety Analysis Repo'rt (FSAR) is totally inadequa?thto permit a determination of the adequacy 1 the Monticello plant to withstand seismic loadings.

.In many espects the.informa-tion presented in the FSAR is less info tive than that pre-z sented in the earlier Preliminary Safety-talysis Report'(PSAR).

However, your letter of transmittal applyi

-for.an operating:

license states that the FSAR "..... superse s in-its) entirety the application dated July 25, 1966."

b rincipal concern is thaDhe, FSAR. does no state in suf ficient detail how the structural and seismi design objec-tives presented in the PSAR and amendments were t anslated'

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into the final 1 esign. - Accordingly, we request th t.. the-d appropriate sections of the FSAR related.co the sei ic design t_

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ThE of Class I s ructures, equipment, piping,. instrumentation -

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.and controls, e extensively revised. We believe that this-.

- is the most de trable.and ' expeditious manner in which to.

- update, augment and present in. sufficient detail tha infor,

mation we requir - to continue.our review of your application.

To assist you, we. ave attached a summary of the itype of

'l informa: ion that sh id be included in the: appropriate:-

revisions to the FSA.

Please contact us if you have any; questions regarding ti's request..

Sincerely,.

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Peter A. Morris, Director 7

ivision of Reactor Licensing' c

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Enclosure:

i Summary info. for j

FSAR revision

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Dis t ribution :

AEC ?ub Doc Rm Docket File

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SUMMARY

OF TYPEhF INFORMATION TO BE INCLUDED IN' REVISED SECTIONS 0F FSAR TING TO STRUCTURAL AND SEISMIC DESIGN-OF

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- HONTICELLO PLANT _

Sk A detailed explanation o( the means by_ which functionalladeq'uacy will bei 3

analyzed for Class I.strd tures, equipment -piping,. instrumentation land; controls which, as stated n the FSAR "... are designed in such a way that for a ground accelera on of 0.12g a safe shutdown can be achieved."

A listing of the codes used. gr Class I structures, equipment, piping.1 instrumentation and controls ry" lated to the statement in the FSAR that -

L these structures and equipment

. are designed in such a'way that for-9 the. combination of. normal loads plus design earthquake the stresses: ara -

elements,-astressincreasehasb(e vithin' code allowable." Also, a indication: as : to whether for some j

used, as ' permitted by ;the codes.

A description of the mathematical mgdels used for the seismic design of -

each of the Class'I structures, equig' ment,: piping, instrumentation and controls. systems, and an explanation f how the elasticity of the

_i structures, and the damp'ing have been valuated.-

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An explanation of whether the response s ectrum method-or the time' history method of analysis was used for.s ismic design. IfLa modal'._

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analysis has been used then for every im rtant structure, piping' system,.

or equipment, an' indication of how many mo a have been considered and

-j a description of. how the damping was evalua d' for each mode. -- Also, an indication of the degree to which-the true re ponse of Class I structures y

l and equipment is underestimated'by the use of mooth response spectra.-

A discussionLof how closely the mathematical' mod s represent the~ actual' conditions, especially the ef fect 'of non-linear b avior of.the actual--

structures,-piping and equipment; ' effect ofiappend esy(small masses <

elastically. attached to large masses)1such as vent-pes:and header, equipment hatch and personnel lock; effect 'of ' clearan. es (gape) at equip-

. ment' restraints and supports; and effect of variable f etion.

Sufficient information-to show that at points where struc ures and/oru eq,uipment are interconnected, the-dynamic deformations are ompatible:-

For instance l(a) for horizontal restraints of. the reactor'

-elevation' 994'-2" and of the-drywell and~the shield.at elevation 992'-

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(b) for the drywell, at the shear lugs between the drywell an the f

reactor building at elevation 992'-4-13/16".

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'-- Justification for subdividing the Class I structures and equipment into the three cate oriest rigid, resonant, and flexible,'and'an~ explanation of the manner 1 which these categor'ies are used in the final design.

4 An explanation o how the ' interaction between soil and the reactor building has been rovided for in the seismic analysis and the design of the building and wh ther non-linear behavior of soils has be'en considered.

A listing of the ampi ication factors resultin'g f rom' seismic analysis, as compared with the g und motion for the reactor, recirculating pumps,-

Class I-piping, and spe t-fuel pool.

l Information as to whether nstallation of strong motion seismographs is

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planned for the facility (t e number and type) and how determination will be made that the response o the structure and primary equipment is l

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P within. allowable design limit The data obtained from seismographs

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would be helpful in evaluating. ost-earthquake damage to the facility.

I A justification of the adequacy the seismic design cf the battery racks.

Supporting information to show how he seismic design of Class I tunnels and underground piping and cables en ering or leaving a structure were handled, since the seismic response w thin and outside the structure -is

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quite dif ferent.

A discussion of-whether means will be av lable to monitor possible

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4 settlements of Class I structures and equ ment.

Provisions made to limit the gradual-increa of the leakage rate of the reactor building due to gradual deterioratio of the structure, such as-j Od

_ increased cracking, aging of caulked joints. a gaskets.

i-A detailed explanation and supporting analyses show how the effect or-a' seismic disturbance on the Class II part of th main steam lines has been taken care of in the design of the Class I p t, especially for the anchors and the valves.

An explanation to support the statement that parts o Class'II structures

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covering or supporting Class I equipment have been de gned as Class.I structures. All such equipment in this category shoul be included in the discussion.

An evaluation of the capability :of the f ac111ty, includin the, stack,~

to withstand a tornado with 300 mph rotational velocity, 6 mph trans-lational velocity, and a ) psi pressure drop in 3 seconds. - 1so,'an ;I indication of whether stack failure can endanger any Class I ystema oi structures required for safe shutdown.

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e A description on how the gap between the drywell and the shield,ing con-

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crete outside f the drywell is drained and vented, since it appears that strips of olyurethane foam used in construction have been left in the gap at spec ic elevations.

-- Method used to suport the recirculation pumps -- design criteria, materials used, anc design methods used for thepe supports -- potential for pua.ps to become.issiles.

-- With respect to the sp nt fuel pool:.

An evaluation of -he temperature stresses in the walls of the spent fuel pool an ' provisions made to limit cracking and prevent leakage.

w The provisions to protict the pool from the loss of water tr 8"*4 tornado generated missi'es and such things as an inadvertent

'mn ing of a spent fuel shipping cask.

j The design procedures employed for *he Class I piping including as a

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minimum the following:

Methods of analysis used.

Stress limits to which the piping systems were 3esigned.

- Supporting systems, snubber locatio

, etc.

Seismic factors, amplification factors and other pertinent f actors used.

j Accommodation of pipe whip.

r U m m details on the foundation design. J e'only mention oP j

i sunoa m a.acaign in the FSAR consists oI a brief des iption of 'the foundation investigation in Section 2.5.5 and several ief sentences j

in Section'12.2.1.5.

The latter discussion indicates ti t "... and.

other structures are supported on undisturbed soils or co acted W

k l ected backfill." ' Without some discussion of the conditVns,that were encounterea and the actual procuuureT that were employed in b onstructing the foundations for t'he Monticello plant, 6 cannot assess t % adequacy'

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or tne foundation desi @

g Verification-that the rev,ised seismic criteria developed in the n-struction permit review of.the PSAR and amendments were actually ployed

'in the final design of the facility; e.g.,

one of the seismic repor -

included in Appendix A of the FSAR pre-dates any fin'al dated documents relating to the construction permit review.

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Verification t at the earthquake record employed in the seismic analysis

' (1.e., using as indicated in the FSAR a time history ' approach using1 the

'j Taft, July 21,-1 52, N 69* R earthquake record appropriately scaled) leads to a spectr of the type presented in Plate 3 of the seismic i

criteria (first po tion of Appendix A), which was-the basis of the:

specified criteria.

Also, a discussion as to whether slight changes in.

the time history inp

, or alternatively slight - changes in' the method of l N

c.odeling the structure could lead to any significant changes in the design calues arrived at, and hether an approximate check of the results' obtained by the time his cry approach was made.

Sufficient information'as - the frequencies and mode shapes for the

" coupled system" related to he design analysis of the reactor pressure j

vessel noted in Section '4 of 'ppendix A.

-- Additional details as to how sp -ing constants such as those designated.

K3 and K4, which represent the i undation stiffness and lateral resistance l

(soil structure interaction), we obtained for' use in most of ;the '

analyses presented in Appendix A.

Clarification as to the damping valu s used in the final design.

In' Appendix A it is-noted that damping lues of-10 percent of crit,1 cal.are to be employed fo'r ground rocking mode of vibration.. On the-other hand, our records indicate that in reply to Q estion 8.8 of Amendment 6 of the ~

PSAR, Northern States Power stated that-value of 5 percent would be_used.

In the PSAR it was noted in Section 5.3.1.

that for the maximum earth-quake in cases in which the stress combinat ns exceeded' yield's' tress :

analyses would be made to determine the ener absorption capacity.and l

ro review that the resulting deflections;or d tortions.would not prevent.

proper functioning of the structure-or piece o equipment.. Additional s

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information is needed to indicate whether such c nditions were encountered '

i at any point in the. design and if so, what limit on stress land deforma-

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tion criteria were adopted to insure-the adequacy.-

the design to' meet the intended design criteria.-

-- A comprehensive discussion of the loading combinations employed in the :

design, the applicable stress and deformation limits em oyed in the' design, and any other information of this type - to provide a basi for evaluating-the adequacy of the f acility seismic design. The limited ata;of this=

type which appeared in the = PSAR apparently. were deleted fro the FSAR,.-

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Clarification as t.o whether as noted on page 12.2.2 of lthe"FS that i.

Class II structures verb designed by the. Uniform Building Code Zone'l' j-conditions. This appears to be at -variance with tha material pre ented :

in answer to Question 8.10 of Amendment 6 of the PSAR wherein: Northern e

States Power-indicated that a seismic coef ficient of 0.05 would be used f or Class II. structures and' equipment.

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A comprehe ive discussion and details in' regards co the-design of-the' reactor' int nals and primary system piping including but' not. limited f ro the followin :-

- The ex nt and findings of the analyses by which the conclusions

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stated n Section 3.6.3 of the FSAR are reached; i.e., it is-stated t t the reactor vessel internals are. designed to maintain a reflood ng capability following a loss-of-coolant acci'ent 'and-4 that the i ternals are also designed _to preclude a failure mode

'j which would esult in any part being discharged through the main-1 steam line i the event of a steam line break.

I In your discuss on pertaining to the directly added simultaneous..

peak loads resul ing from normal operation plus the worst loss-of-!

coolant accident us the design basis earthquake,.and for the.

combination of the ormal operating loads plus the peak loads I

from the worst loss f-coolant accident include:

1 (a) The methods of an ysis (elastic, elastic plastic, limit),

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(b). The limits' to which he critical components.were evaluated (stress, strain, def ction, buckling, wit' numerical h

values for several cri cal items.

(c)

For each method of analy a relate the possible err 6rs in-6:

the analytical method and he possible errors in loads used-for the stress calculations to the margin of safety which is being provided.

For the drywell, clarification of the. follow ng:

1 Method of evaluating the jet forces _an the area subjected to their effect.

How the maximum me,tal temperature of 300*E. - s established >for-jet impacted steel. plate and an evaluation o the corresponding thermal stresses in the shell.

Why the temperature of the steel plates was redue to 150*FLwhen jet action is' considered with design ' internal pres re', and-what this pressure is.

Why local yielding has been permitted when the shell is acked up by concrete., This criterion is not a code criterion.- low:',

was it established. that a ' rupture will not occur? Where thg shell is not backed up by uncrete, the primary. membrane-stresses are f S

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permitted to go up to 0.9.of:the yield point: this.is not allowed-

, by he_ code. Note all of the loads which were combined whent this criterion was used.

For the design f the torus, consideration-of the following A justific ' tion of the value of 21 kips for. th'e jet force at 4

s each downco, er pipe in the torus. -

A description -

the stress criteria and design methods used for the design of t'.- torus.for jet forces...- A discussion of whether u

other jet forces sist in the torus in addition to downcomer l

pipe jet forces.

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- A description'of how e flooding of the drywell and the torus f

has been considered an combined with seismic loads. Also, an indication of the corres nding critical stresses..

With respect to the penetrations in the drywell and torus', the following:'

A discussion-of the applicabic design criteria and the load:

combinations'used in the design' A description of the stress analy s methods lused to evaluate t

the stresses in the shell;. penetra on sleeves, bellows and guard pipes; and process piping at-p netrations, their anchors and-supports.

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'The significance of the statement in the-FSAR-th-t ".

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ll ment vessel was code stamped for the design press re and design Ltempera-ture";

i.e., does this mean that'It is not a code ssel for.other loads =

such as aeismic loads, jet forces and. equipment. loa

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Just1fication of. the assumption used in the seismic an ysis'of the-primary containment structuge that the drywell is comp 1 ely fiied on'its foundation.

- A detailed discussion of the seismic design of the plant s ek, torus',-

ring header and its supports. The seismic design of these.

atures were completely. omit.ted from the FSAR.

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