ML20148J629
| ML20148J629 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 10/15/1975 |
| From: | Purple R Office of Nuclear Reactor Regulation |
| To: | Groce R YANKEE ATOMIC ELECTRIC CO. |
| References | |
| NUDOCS 8011250129 | |
| Download: ML20148J629 (7) | |
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.i rocket rc. Et-29 OCT 151975 P
Yerk6e Atcric Flectric Ccerrnv - '
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Certlerer:
FF: Yarkee-Teke Atcric Ecwcr Ftetien
- pr The nurocce of this letter is tc inferr ycu of a cotentici cafety cucctier which her been reisc<* recorFine the desien cf reecter precrurc vessel cuptcrt systerr for prcccurited veter recctcrr (WE's) .
Cn Pay 7,1975 the FFC was infor:ted by a licensee that certain trenrient Ictds I en the reactor vessel seppert rerberc that would rerelt frce a ecstulated I
=. reactor ecolant cire rupture irrediately edjacent to the reacter vessel hed j E been enderectivated in their cricinal deelen enalyser.
It ir the !!FC steff's cpinion that the cuestien teleted to tre treatrert of
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trencient 1cade in the derion cf reacter verrel currert syrters rcy cpply to other FFF facilitics, especially there fer which the desien entlyses were cerferred scre tire ecc. Fe beve therefere initiated a systcretic review cf this ratter to deterrine bcw tl.ese Iceds were teken into acccent cn l other FF facilitice, end wbet, if cny, ccrrective recsures ra'/ be recuired l fer roccific facilitiec. l The results cf licenree strdier rerorted to dete indicete that, e]thcuch the rarcins of scfety rey te less tbtn cricinelly intended, tbc reecter vessel suctort syster would retain sufficient structural intecrity to succcrt
- .3 tre vessel and that the.ultirete cenrecuerces of this postuleted eccident which eculd effect the ccreral public cre no werte then cricinally etcted.
Fe have not ec:micted cur independent evaluation cf there studicc. Pcwever, based on the recults of our eveluaticn of this pbencrenen to date crd in E rececnition of the ]cw prebebility cf the carticuler rice ructure t<hich could Iced to e/diticnal trersient Icede cn the currort cyrteer, we cercles that centinued reacter creratien erd ccotinued licersirr cf fccilitier fer cceratien are ecceptable thi3e te corduct cur ceretic review.
Fe recuert that ycu revicu the 6erien bares fer tbc recctcr verre) cercert syster fer your fccilitv to Feterrire ubetter the trerricrt 3cefs
' de egi gt r c crure were te . intc' crecent acprcpriate]" in t e
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- YcrPea Atcric Electric Ccerzr.v -
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..s desicr. Ficare irferr us cf the restitc cf yeer reviev within 2C Ccyr. I Che attachrerts te the ecclerure t.tc frcvited to irdicctc rec irferrcticr thct ceuld be rcec'c6, thou]6 m feterrire, en the Ecsir cf ycer l review, thtt e rceracterent cf the wcsel rur. pert decicr is rcceired.
Ve cre ccntinuire te cvelerte trC review tbc rethc6ciccy fer calcu3etir.c tPc rtFeccler ble,(cwr Ic /r with trc recierr etccr everr- rere]icrs.
Ycu cheulf cer.tret yeer reeleer ctecr rystcc curnller fer irictrrtier l remerdire there cnicel-tirn: I' rccerrcr'. re cc m 1cte yccr r vics.
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Sircerely, Oris;inal signed by
- n. A. Purple Fchert A. Fure.lc, Chicf Creratire reecterr rrcrch fl F.ivisier of Fcceter Iiccrriro
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Yankee Atomic Electric Company October 15, 1975 cc: Mr. Donald G. Allen, President Yankee Atomic Elcetric Company 20 Turnpike Road Westboro, Fbssachusetts 01581 Greenfield Public Library 402 Main Street Greenfield, Massachusetts 01581 l
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- '"d ENCLOSURE -
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STATEMENT OF THE PFOBLEM In the unlikely event of a WR primary coolant system pipe rupture in the irrediate vicinity of the reactor vessel, transient loads'oriainatina from three principal causes will be exerted on the reactor vessel support system.
These are:
Aq 1. Blowdown jet forces at the location of the rupture (reaction forces),
- 2. Transient differential pressures in the annular reglen between the vessel and the shield, and
- 3. Transient-differential pressures across the core berrel within the reactor vessel.
3 The blowdown jet forces are adequately understood and desion procedures are
- .j available to account for them. Both of the " differential pressure" forces,
-; bowever, are three-dinensional and tim dependent and recuire sophisticated
.a analytical procedures to translate them into loads actino on the reactor
- j vessel support system. All of the loads are resisted by the inertie and
- by the support members and restraints of other components of the primary
- coolant system including the reactor pressure vessel supports.
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The transient differential pressure actino externally on the reactor vessel is a- result of the flow 'of the blowdown effluent in the reactor cavity. %e maanitude and the time dependence of the resultina forces depends on the
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nature and the size of the pipe rupture, the clearance between. the vessel
_ and the shield and the size and location of the vent openings leading from
= the cavity to the containment as a whole. For some time refined analytical methods have been available for calculating these transient differential
- 3 pressures (multi-node analyses) . %e results of such analyses indicate that the consecuent loads on the vessel support system calculated by less sophisticated nethods ray not be es conservative as originally intended for
." earlier desiens. Attachment 1 to this enclosure provides for your information a list of information requests for which responses could te needed for a proper assessment of the impact of the cavity differential pressure on the se "+ design adecuacy of the vessel support system for a power plant.
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The controlling loads for desion purposes, however, appear in typical cases to be those associated with the internal differential pressures across the core barrel. - The internally generated loads are due -to a momentary differential pressure which is calculated to exist across the core barrel when the pressure in the reactor annular recion between the core barrel and vessel wall in the vicinity of the ruptured pipe is assumed to rapidly decrease to the saturation pressure of the primary coolant due to the outflow of water. Although the depressurization wave travels rapidly around the
- core barrel, there is a finite period of tire during which the pressure in
, the annular region opposite the break location is assumed to remain at, or near, the original reactor operating pressure. Thus, transient asynwetrical-forces are exerted on the core barrel and the. vessel wall which ultimately result in transient Icads en the support systems. These are the. loads which were underestimated by the licensee originally reporting this problem and which may be underestimated in other cases. They are therefore of generic concern to the staff. Attachment 2 to this enclosure provides for your information a list of information recuests for which responses would be needed
, for a proper assessment of the impact that the vessel internal. differential pressure, in conjunction with the other concurrent Icads, could have en the design adeauacy of the support system.
In that there are considerable differences in the reactor support system designs for various facilities and probably in the design margins provided Dj
't by the desioners of older facilities, the underestimation of these " differ-ential pressure" loads may or may not result in a determination that the !
adequacy of the vessel support system for a specific facility is guestion- ;
able. Since local failures in the vessel supports (such as plast'ic deformation) do not necessarily lead to the failure of the supports as an integral system,
^ there may be some limited reactor vessel notion provided that no further j sionificant consecuences would ensue and the emergency core cooliry systems I (ECCS) would be able to perform their design functions. '
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q ATTAClCIE;lT 1 CONTAll!MEi!T SYSTEMS BRANCH REQUEST FOR ADDITIONAL IllFORMATION In the unlikely event of a pipe rupture.inside major component subcompartments, the initial blowdown transient would lead to non-uniform pressure loadings on both the structures and enclosed components. To assure the integrity of these design features, we request that you perform a compartment multi-node pressure response analysis to provide the follrwing information:
(a) fhe results of analyses of the differential pressures resulting from hot leg and cold leg (pump suction and discharge) reactor coolant system pipe ruptures within the reactor cavity and pipe penetrations.
1 (b) Describe the nodalization sensitivity study performed to determine a
j the minimum number of volume nodes required to. conservatively g
predict the maximum pressure within the . reactor cavity. The nodalization sensitivity study should include consideration of spatial pressure variation; e.g., pressure variations circumferentially, axially and radially within the rea'ctor cavity.
W ij (c) Provide a schematic drawing showing the nodalization of the reactor 1
cav'ty. Provide a tabulation of the nodal net free volumes and intercor.necting flow path areas.
l (d) Provide sufficiently, detailed plan and section drawings for several y
views showing the arrangement of the reactor cavity structure,
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reactor ve sel, piping, and other major obstructions, and vent areas, to permit verification of the reactor cavity nodalization and vent locations.
(e) Provide and justify the break type and area used in each analysis.
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p (f) Provide and justify values of vent loss coefficients and/or friction ur E , factors .used to calculate flow between nodal volumes. When a loss E
coefficient; consists of more than one component, identify each component, its value and the flow area at which the loss coefficient l2.. applies.
(g) Discuss the manner in which movable obstructions to vent flow (suchas' insulation,' ducting, plugs,andseals)weretreated. Provide e
analytical justification for the removal of'such items to obtain vent area. Provide justification that vent areas will not be partially or completely plugged by displaced objects. .
(h) Provide a table of blowdown mass flow rate and energy release rate as a function of time for the reactor cavity ciesign basis accident.
- (i) Graphically show the pressure (psia) and differential pressure (psi) g=q responses as functions of time fo[ each node. Discuss the basis for
- .:::::g establishing the differential pressures.
(j) Provide the peak calculated differential pressure and time of peak pressure for each node, and the design differential pressure (s) for the q reactor cavity. Discuss whether the design differential pressure is
- .a uniformly applied to the reactor cavity or whether it is spatially i i
varied. (Standard Rev~:ew Plen 6.2.1.2, Subcompartment Analysis attached, l l
provides additional guidance in establishing acceptable design values, , '
for determining the acceptability of the calculated results.)
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U.S. NUCLEAR REGULATORY COMMISSION February, 1975 STANDARD REVIEW PLAN OFFICE OF NUCLEAR REACTOR REGULATION
. SECTION 6.2.1.2 SUBCOMPARTMENT ANALYSIS REVIEW RESPONSIBILITIES Primary - Containment Systems Branch (CSB)
Secondary - Mechanical Engineering Branch (MEB)
Core Performance Branch (CPB)
Auxiliary and Power Conversion Systems Branch (APCSB) .
- 1. AREAS OF REVIEW The CSB reviews the information presented by the applicant in the safety analysis report concerning the determination of the design differential pressure values for containment sub-compartments. A subcompartment is defined as any fully or partially enclosed volume within the primary containment that houses high energy piping and would limit the flow of fluid to 4
the main containm'ent volume in the event of a postulated pipe rupture within this volume.
A short-term pressure pulse would exist inside a containment subcompartment f ollowing a pipe rupture within this elume. This pressure transient produces a pressure differential
, across the walls of the subcompartment which reacnes a maximum value generally within the
, first second af ter blowdown begins. The magnitude of the peak value is a function of Y. several parameters, which include blowdown mass and energy release rates, subcompartment i
t volume, vent area, and vent flow behavior. A transient dif ferential pressure response analysis should be provided for each subcompartment or group of subcompartments that meets 9
g the above definition. ,
The CSB review includes the manner in which the mass and energy release rate into the break i:.j compartment were determined, nodalization of subcompartments, subcompartment vent flow j
behavior. and subcompartment design pressure margins. This includes a coordinated review effort with the CPB. The CPB is responsible for the adequacy of the blowdown model.
The CSB review of the mass and energy release rates includes the basis for the selection of the pipe break size and location within each subcompartment containing a high energy line and the analytical procedure for predicting the short-term mass and energy release rates.
The CSB review of the subcompartment model includes the basis for the nodalization within each subcompartment, the initial thermodynamic conditions within each subcompartment, the nature of each vent flow path considered, and the extent of entrainment assumed in the vent M: j flow mixture. The review may also include an analysis of the dynamic characteristics of l b components, such as doors, blowout panels, or sand plugs. that must open or be removed to N
USNRC STAND ARD REVIEW PLAN
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provide a vent' flow path, and the methods.and results of components tests performed to demonstrate.the validity of these analyses. The analytical procedure to determine the loss
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coefficients for each vent flow path and to predict the vent mass flow rates, including
, ;;;g:; j flow correlations used to compute sonic and subsonic flow conditions within a vent, is re- J viewed. LThe design pressure chosen for each subcompartment is also reviewed.
On triuest l
from the' APCSB, the CSB evaluates or performs pressure. response analyses for subcompartments 1 outside containment.
The MEB is responsible for reviewin the acceptability of the break' locations chosen and of the design criteria and provision. methods er loyed to justify limited pipe motion l
_for breaks postulated to occur within subccmpartments (See Standard Review Plan 3.5.2). l I I .~ ACCEPTANCE CRITERIA The subcompartment analysis should incorporate the following assumptions: f
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~a. Break locations and types should be chosen according to Regulatory Guide 1.46 for subcompartments inside containment and to Branch Technical Position MEB 3-1
- (attached to Standard Review Plan 3.6.2) for subcompartments outside containment.
An acceptable alternate procedure is to postulate a circumferential double-ended rupture of each high pressure system pipe in the subcompartment,
- b. Of several breaks postulated on the basis of a, above, the break selected as the reference case for subcompartment analysis should yield the highest mass and energy release rates, consistent with the criteria for establishing the break location and area,
- c. The initial' plant operating conditions, such as pressure, temperature, water inventory, and power level, should be selected to yield the maximum blowdown
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condit.ons. The selected operating conditions will be acceptable if it can be
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' shown that a change of each parameter would result in a less severe blowdown profile.
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J. The analytical approach used to compute the mass and energy release profile will be E accepted if both the computer program and volume noding of the piping system are
" -U similar to those 'of an approved emergency core cooling system (ECCS) ' analysis. The h computer programs that are currently acceptable . include SATAN-VI (Ref. 24), CRAFT (Ref. 23),' CE FLASH-4 (Ref. 25), and RELAP3 (Ref. 21), when a flow nultiplier of F 1.0 is used with the applicable choked flow correlation. 'An alternate a'pproach,
.E which is also acceptable, is to assume a constant blowdown profile using the initial '
conditions with an acceptable choked flow correlation. When RELAP-4 is accepted by lll;.*;;
the staff as an operational ECCS blowdown code, it will be acceptable for subcompart- ,
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. ment analyses.
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iE;;;:;r :::.5 3. The initial atmospheric conditions within a subcompartment should be selected to max-imize the resultant dif ferential pressure. An acceptable model would be to assume air
~~ at the maximum allowable temperature, minimum absolute pressure, an'd zero percent rel-ative humidity. If the assumed initial atmospheric conditions differ from these, the selected values should be justified.
6.2.1.2-2
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Another model that is also acceptable, for a restricted class of subcompartments, in-volves simplifying the air model outlined above. For this model, the initial atmos-phere within the subcompartment is modeled as a homogeneous water-steam mixture with an average density equivalent to the dry air model. This approach should be limited to subcompartments tnat have choked flow withiri the vents. However, the adequacy of this simplified model for subcompartments having primarily subsonic flow through the vents has not been established. .
4 Subcompartment nodalization schemes should be chosen such that there is no substantial pressure gradient within a node, i.e. , the nodalization scheme should be verified by a sensitivity study that includes increasing the number of nodes until the peak cal-culated pressures converge to small resultant changes.
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- 5. If vent flow paths are used which are not immediately available at tt' time of pipe rupture, the following criteria apply:
a, The vent area and resistance as a function of time af ter the . _ be based on a dynamic analysis of the subcompartment pressure response to pipe ruptures.
- b. The validity of the analysis should be supported by experimental data or a testing program should be proposed at the construction permit stage that will support this analysis,
- c. The ef fects of missiles that may be generated during the transient should be considered in the safety analysis.
, 6. The vent flow behavior through all flow paths within the nodalized compartment model should be based on a homogeneous mixture in thermal equilibrium, with the assumption of 100% water entrainment. In addition, the selected vent critical flow correlation should be conservathe with respect to available experimental data. Currently accept-able vent critical flow correlations are the " frictionless Moody" with a multiplier of 0.6 for water-steam mixtures, and the thermal homogeneous equilibrium model for air-steam-water mixtures.
- 7. At the construction permit stage, a factor of 1.4 should be applied to the peak e dif ferential pressure calculated in a manner found acceptable to the CSB for the subcompartment. The calculated pressure multiplied by 1.4 should be considered the 0"
design pressure. At the operating license stage, the peak calculated dif ferential pressure should not exceed the design pressure. It is expected that the peak calcu-lated differential pressure will not be substantially different from that of the construction permit stage. However, improvements in the analytical models or changes in the as-built subcompartment may affect the available margin. ,
Ill. REVIEW PROCEDURES The procedures descri a below are followed for the subcompartment analysis review. The reviewer selects and emphasizes material from these procedures as may be appropriate for 6 2 1 2-2
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-a particular case. Portions of. the review may be' carried out on a generic basis or by adopting the results of previous reviews of plants with essentially the same subcompartment 5 and high_ pressure piping design.
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.?E The CSB reviews the initial conditions selected for determining the mass and energy release
-rate to the subcompartments. These values are compared to the spectrum of-allowable opera-ting conditions for the plant. The CBS will ascertain the adequacy of the assumed conditions M based on this review.
The CSB confirms with the MEB the validity of the applicant's analysis of subcompartments containing high energy lines and postulated pipe break locations, using elevation and plan drawings of the containment showing the routing of lines containing high energy, fluids. The CSB determines that an appropriate reference case for subcompartment analysis has been identified. In the event a pipe break other than a double. ended pipe rupture is postulated by the applicant, the MEB will evaluate the applicant's justification for assuming a limited displacement pipe break.
- The CSB may perform confirmatory analyses of the blowdown mass and energy profiles within a subcompartment. The analysis is done using the RELAP3 computer program (See Reference 21 for a description of this code). The purpose of the analysis is to confirm the predic.
- tions of the mass and energy release rates appearing in the safety analysis report, and to confirm that an appropriate break location has been considered in this analysis. The use of RELAP3 will continue until the RELAP4 computer code h6s been approved by the staff as an acceptable blowdown code. At that time, the CSB will replace RELAP3 with RELAP4 for all subsequent analyses.
The CSB determines the adequacy of the information in the safety analysis report regarding subcompartment volumes, vent areas, and vent resistances. If a subcompartment must rely on doors, blowout panels, or equivalent devices to increase vent areas, the CSB reviews the analyses and testing programs that substantiate their use, bEri. The CSB reviews the nodalization of each subcompartment to determine the adequacy of the calculational.model. As necessary, CSB performs iterative nodalf ration studies for sub-compartments to confirm that sufficient nodes have been included in the model.
=y EE The CSB compares the initial subcompartment air pressure, temperature, and humidity condi-tions to the criteria of II, above, to assure that conservative conditions were selected.
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._:::e The CSB reviews the bases, correlations, and computer codes used to predict subsonic and sonic vent flow behavior and the capability of the code to model compressible and un- (
compressible flow. The bases should include comparisons of the correlations to both experimental data and recognized alternate correlations that have been accepted by the staff.
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' Using the nodalization of each subcompartment as specified in the safety analysis report. -
' the CSB performs analyses using one of several available computer programs te, detennine -
'EE the adequacy of the calculated peak differential pressure. The computer program used will -
. depend upon the subcompartment under review as well as.the flow regime. At the present time, the two programs used by the CSB are RELAP3 (Ref. 21) and CONTEMPT-LT.(Refs 7, 8, and 9). A multi-volume computer code is currently under development. l
,== 'At the construction permit stage, the CSB will ascertain that the subcompartment design l
pressures include appropriate margins above the calculated values, as given in II, above. I
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3;,. IV. EVALUATION FINDINGS The conclusions reached on completion of the review of this section are presented in
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Standard Review Plan 6.2.1. -
V. REFERENCES The references for this plan are those listed in Standard Review Plan 6.2.1, together with
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la. Regulatory Guide 1.46. " Protection Against Pipe Whip Inside Containment." 2a. Standard Review Plan 3.6.2, " Determination of Break Locations and Dynamic Effects issociated with the Postulated Rupture of Piping," and attached Branch Technical Position MEB 3-1, " Postulated Break and Leakage Locations in Fluid System Piping Outside Containment." Fe"
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4 s a ATTACHMENT 2 . MECHANICAL ENGINEERI'NG BRANCH
- =r RE0 VEST FOR ADDITIONAL INFORMATION Z ,
Recent analyses have shown th'at reactor pressure vessel supports may be subjected to previously underestimated lateral loads under the conditions
- that would exist if an instantaneous double ended break is postulated in the reactor vessel cold leg' pipe at the vessel-nozzle. It is therefor,e necessary to reassess the capability of the reactor coolant system supports
%w to-limit the calculated motion of the reactor vessel during a postulated cold leg break within bounds necessary to assure a high probability that the reactor could be brought safely to a cold shutdown condition. i The following information is required'for purposes of making the necessary reassessment of the reactor vessel supports: [ 1. Provide engineering drawings of the reactor support system sufficient to show the geometry of all principle . elements and materials 6f con-
, struction.
g 2,'. Specify the detail design loads used ,in the original design analyses of
, the reactor supports giving magnitude, direction of application and the basis for each load. Also provide the calculated maximum stress in each principle element of the support system and the corresponding allowable stresses.
2 3. Provide the information requested in 2 above for the RV supports con-x g P sidering a postulated break at the cold leg nozzle. Include a summary W of the. analytical methods emploved and specifically state the effects of g[ 'short term pressure differentials across the core barrel in combination, j
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b541 - j]- .with all external loadings calculated to result from.the required Eiifi. ?$t Sj ,pos tul ate. This analysis should consider:
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(b)- consideration of fluid structure interaction ~ (c) use of actual time dep.endent forcing function
'(d) reactor support stiffness. .
- 4. If the' results of the analyses required by 3 above indicates loads
- leading to inelastic action in the reactor supports or displacements
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. exceeding previous design limits provide an evaluation of the following:
J70 (a) Yield behavior (ef,fects of possible strain energy buildup) of the l - ~' material used in the reactor support design and the effect on the loads mj transmitted to the reactor coolant system and the backup structures to'which the reactor coolant system supports are attached.
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(b) The adequacy of the reactor coolant system piping, control rod drives, steam generator and pump supports, structures surrounding the reactor coolant system, reactor internals and ECCS piping to assure that the reactor can be safely brought to cold shutdown.
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