ML19319D377
| ML19319D377 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 11/14/1975 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | Rodgers J FLORIDA POWER CORP. |
| Shared Package | |
| ML19319D378 | List: |
| References | |
| RTR-NUREG-75-087, RTR-NUREG-75-87 NUDOCS 8003160212 | |
| Download: ML19319D377 (5) | |
Text
.
Docket Filt. '
recket re:
.MOV.14 975 LWR 2-3 File TTC VMoore RWKlecker FIcrida Fever Cerncratien MWilliams m:
Fr. J. T. Fedcers LEngle Vice President, Facilitier EGoulbourne t,ITelear Project Panacer ACRS (16)
- p. c. Box 14042 ELD St. Tetersburc, Flcride 33733 IE (3)
Centleren:
The purpose of this letter is to infore you cf a pctential cafety cuestien which has been raised recardino the cesicn of reacter cressure veccel cucpc-t syste~s for pressurized water reectcrs (n T's).
Cn t'av 7,1975 the IUC was inferred by a licensee that certain transient 1ceds en the reacter wssel succert verbers that would reruit free e rectuletec reacter ccclant pire' rupture irrediately adjacent to the reacter versel bed teen underesti:reted in their cricinal decien enalyser.
It is the ITC staff's ocinicn that the cuestien related tc the treatrent of trensient Iceds in the desien cf reactor rescel succert systers ray accly to cther MP facilities, errecially there for which the desien analyres were perferred ccre tine aco. W have therefere initiate 6 a svrteretic review cf this reatter to deterrine hcw these Icacs were teken into acccunt en etter FFF facilities, ard stat, if any, ccrrcetive wesurer rav te receired fer crccific facilities.
The results of licencee studies reperted to date irdicate that, cItheucp the cercirs cf safety rey be less than cricinally interded, the reacter veccel suppcrt systep would retain sufficient structurel intecrity te cunrcrt the versel and that the ultirate correcuences cf this postulated accidert which could affect the ceneral public are nc verce than cricinally stated.
Ce haw not empleted ' cur inderendent evaluatien of there studies. Fcvever, -
bared en the reruits cf cur evaluatien of thir thercreren to date crd ir recceritien of tre Icw probability of the particular cire ructure which ceuld Irad to edditicnal trenrient IceFs cn the remrt syrters, we cerclude that centirned reacter cceratien end centinued licercim of f?cilitics fer ee rstien are eccertsble while we cenduct cur gereric revicw.
Fe rectest that ycu review the Fesien beses fer the reacter versel ruccert everer fer ycur fccility(ice) te detcrrine sterber +c tracciert Ice /s (crcribed in the erclcsure were te!<en into acccunt arrrepriately in the devien. Pierse inferr us cf the resulte cf ycur review withir 20 days.
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v i i We' attacbrents to the enciesure are crovided to indicate the inferwatien ther cculd be needed, sbculd we &tcrrine, en the basis of ycur l
review, that e reassessrent cf the vescel curcort desien is recuired.
Fe are centinuina to evaluote and review the retbcdolocy for calculatine the subcooled blewdcwn Iceds with the nuclear steem syster sucpliers.
Ycu sbculd centact your nuclear.stenir Evctem surplier for inferretien recordine these calculations if neccesary to eckplete ycur review.
Ois recuest for ceneric inferratien was apprcved by GK ur. der a b1cnket elecrance nurber E-1E0225 (P0072). This cicarance expires July 33, 1977.
9 Sincerely, Original Signed by A. Schwencer A. Sch'aencer, Chief Light Enter Feactors Franch 2-2 Elvisice of Peacter Licensinc i
rnclosure:
Staterent of the Preblem cc w/ enc 1:
i Pr. S. A. Frandirere Vice President & Cencral Councel P. C. Ecy 14042 St. Fetersburg, riorida 33732 RL:LW 2-3.._ _RL:
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Form AEC.318 (Rev. 9 53) AECM h240 ' 9h
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1 EDCLCSCPE STATEPINI CF THE PFCELEN In the unlikely event of a PWR primary coolant system pipe rupture in the innediate vicinity of the reactor vessel, transient loads oricinatino from three principal causes will be exerted on the reactor vessel support system.
These are:
1.
Blowdown jet forces at the location of'the rupture (reaction forces), -
2.
Transient differential pressures in the annular region between the vessel and the shield, and 3.
Transient differential pressures across.the core barrel within the reacter vessel.
D e blowdown jet forces are adecuately understood and desian precedures are i
available to account for them. Both of the " differential pressure" forces, bowever, are three-dimensional and tine dependent and require sophisticated analytical procedures to translate them into loads actino on the reactor vessel support system. All of the loads are resisted by the inertia and by the support members and restraints of other components of the primary coolant system includina the reactor pressure vessel supports.
D e transient differential pretsure actina externally on the reactor vessel is a result of the flow of the blowdown effluent in the reactor cavity. %e macnitude and the tine dependence of the resultino forces depends on the nature and the size of the pipe rupture, the clearance between the vessel and the shield and the size and locaticm cf the vent openings leading from the cavity to the containment as a whole.
For some time refined anal'ytical methods have been available for calculating these transient differential pressures (multi-node analyses). He results of such analyses indicate that the consecuent loads on the vessel support system calculated by less sephisticated methods may not te as conservative as criainally intended for earlier desiens. Attachment 1 to this enclosure provides for your information a list of information recuests for which resconses could to needed for a proper assessment of the impact of the cavity differential pressure ~ co the desian adecuacy of the vessel support system for a power plant.
. Tr;e contro111nc loads for desian purposes, however, appear in typical cases to be those associated with the internal differential pressures across the core barrel. The internally cenerated loads are due to a rcrentary 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 assured to rapidly decrease to the satur' tion pressure of the primary coolant due to the outflow of water. Although ti.e depressurization wave travels rapidly around the core barrel, there is a finite period of tire during which the pressure in the annular recion cpposite the break location is assumed to remain at, or near, the original reactor operating pressure. Thus, transient asymetrical forces are exerted on the core barrel and the vessel wall which ultirately result in transient 1 cads on the support systers. These are the loads which were underestirated-by the Tfcensee or~i~gfFilly ~reportfro this probler and whidi may be underestimated in other cases. They are therefore of generic concern to'the staff. Attachnent 2 to this enclosure provides for your information a list of information reauests.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 on the desian adecuacy of the support syster.
In that there are considerable differences in the reactor support ryster desions for various facilities and probably in the design rargins provided by the desioners of older facilities, the underestimatial of these " differ-ential pressure" loads may or may not result in a determination that the adecruacy of the vdssel support syster for a specific facility is cuestion-able. Since local failures in the vessel supports (such as plastic deformation) do not necessarily lead to the failure of the supports as an inteoral syster, there ray be serm limited reactor vessel notion provided that no further sienificant consecuences would ensue and the energency core cooling systers (ECCS) would be able to perform their design functions.
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ATTAC;CIENT 1
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v CONTAINMENT SYSTEMS BRANCH I
RE0 VEST FOR ADDITIONAL INFORMATION 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 following information:
(a) The results of analyses of the differential pressures resulting from hot leg and cold leg (oump suction and discharge) reactor coolant system pipe ruptures within the reactor cavity and pipe penetrations.
(b) Describe the nodalization sensitivity study performed to determine
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the minimum number of volume nodes required to. conservatively predict the maximum pressure within the reactor cavity.
The nodalization sensitivity study should include consideration of spatial pressure variation; e.g., pressure variations circumferential1y,-
axially and ra,dially within the rea~ctor cavity.
(c) Provide a schematic drawing showing the nodalization of the reactor cavity.
Provide a tabulation of the nodal net free volumes and interconnecting flow path areas.
(d) Provide sufficiently detailed plan and section drawings for several views showing the arrangement of the reactor cavity structure, reactor vessel, 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.
y
. (f) Provide and justify values of vent loss coefficients'and/or friction
, factors used to cali:ulate flow between nodal volumes. When a loss coefficient consists of more than one component, identify each component, its value and the flow area at which the loss coefficient applies.
(g) Discuss the manner in which movable obstructions to vent flow (suchasinsulation, ducting, plugs,andseals)weretreated. Provide analytical justification for the removal of such items to obtain vent Provide justification that vent areas will not be partially or area.
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 design basis accident.
(i) Graphically show the pressure (psia) and differential pressure (psi) responses as functions of time for each node.
Discuss the basis for
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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 reactor cavity. Discuss whether the design differential pressure is uniformly applied.to the reactor cavity or whether it is spatially (Standard Review Plan 6.2.1.2, Subcompartment Analysis attached, y ried.
provides additional guidance in establishing acceptable design values, for determining the acceptability of the calculated results.)
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