ML19262A323
| ML19262A323 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 10/15/1975 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | Arnold R METROPOLITAN EDISON CO. |
| Shared Package | |
| ML19262A324 | List: |
| References | |
| NUDOCS 7910260650 | |
| Download: ML19262A323 (9) | |
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.m u.x_A n.a.w m b a _ m ugs Gec>ct I:c. cc-2ES OCT 1 i 1975 Netrerclitan Eciser Ceretru
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Fr. P. C. Arec]d Vice Frcsident - Cereratict F. C. Ecx 542 Feedire, Ferreylvania 19CC3 Certleren:
FE: "hree File Isicrd "Pe currese cf this letter is to inferr vcu cf a cetential refety cuestier Wich has Leen raised reccrdire the desicn cf reacter trescure vessel surrert systers fer pressurized water reccters (D.E's).
Cn Pay 7,1975 the FK was inferred by a licerste trat certain transient Iceds en the reacter wesel cup;crt certcrs that wculd result frer a pesteIctc(
reccter ecclant pipe recture irudietely edjacert to the reacter veccel trd teen underesticated in their cricinal desien analyrer.
It is the NPC staff's cpinico that the cuestien related te the treatrer.t of trersiert Iceds in the decien of reacter vessel curycet syrters rey cycly to other FvP facilities, especially tFCse fcr which the desien anelvrec were perferred scre tire coc. We have therefere initiated a systcratic review cf thic ratter to deterrine hcw these 1 cads were tcken into ccccunt en ether 97 facilities, and what, if any, corrective reasures rev N recuired fer crecific facilitier.
S.e results cf licensic studict rererted to date irdicate thet, altretch the rarcins of safety way be less than cricinalJy interdec, tre reecter vessel rurrert syster wculd retzir sufficient ctructural intecrity to sortert the vesse) end that the ultirate ccrsecuercer cf this ecstulated acci/crt which eculd effect the cereral cutlic are rc verse t.*.cr cricirally stcted.
Ve ham ret cercleted cur indecerdert evaluetice cf trere studies.
- Fevever, bered cn the results cf cur evalecticn cf this thercrerer to date cnd ir recceriticn of the Icv prehability cf the particuler rire rupture which eculd lecd te a(ditieral trarsient Icads en the surpcrt svsters, we cerclude
- Pat centinue? reactcr cecretico crd cortirecd licersire cf facilitier fcr cccraticr etc ecceptthJe while w cerdect cur cencric review.
ve reuest that ycu reviw the decian bares fer tFr reectcr vessel surrert syster fer veer facilitv te deterrire @cther the trerciert ] cede described in the ercicsure were t&en into acccurt accrecriatelv in the r
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et'rcrc]iten FFicer Cerrrnv 6ccien. Ficare in:crr us cf the results cf vcur revicw within 2f favr.
The attachrents te the er.c1csure are creviccc te indicate tie inferraticn thzt ceuld be ree?ed, shculd we detereirc', en the basic cf veer
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review, that a reacreercent cf the veccel sutrert fesien ir receired.
Fe cre centiruine te evalucte erd review tre rcthcic1cev fer calcu'etire the cutcceled blevdc>r Icede with tr:e r.uch.or eter-tvrter succlictr.
Yeu thculd. centret ycur nucleFr steer' system rucclicr fcr inferrctier recardinc thcce calculaticns if neccesary te cer:1ere ycur revicw.
Sir recuest fer cercric infer:raticn war agrevcd by CIC erder a tier.ket clearance nurter E-180225 (PCC72). 7his clearccce expirce July 31, 1977.
Sincerely,
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Pebert F. Feld, Chief Ccoratino Fecctors Eranch 34 Division of Fezeter Licensinc Etelcsure:
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- Form AEC 518 (Rev. 9 53) AECM 0240 W u. e.oovsan esar paintine orricassere.eae.ies
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Metropolitan Edison Company October 15, 1975 cc:
G. F. Trowbridge, Esquire Shaw, Pittman, Potts, Trowbridge & Madden Barr Building 910 17th Street, N. W.
Washington, D. C.
20006 GPU Service Corporation Richard W. Heward, Project Manager Thomas M. Crimmins, Jr., Safety and Licensing Manager 260 Cherry Hill Road Parsippany, New Jersey 07054 Pennsylvania Electric Company Vice President, Tecanical 1001 Broad Strcet Johnstown, Pennsylvania 15907 Mr. Weldon B. Archart, Chairman Board of Supervisors of Londonberry Township 2143 Foxiana ' cad Middletown, Pennsylvania 17057 Miss Mary V. Southard, Chairman Citi: ens for a Safe Environment P. O. Box 405 Harrisburg, Pennsylvania 17108 Government Publications Section State Library of Pennsylv.ria Box 1601 (Education Building)
Harrisburg, Pennsylvania 17126
1 ENCLCSCEE STATEMEUI CF THE PFCELEM In the unlikely event of a WE primary coolant syster pipe rupture in the irrediate vicinity of the reacter vessel, transient loads cricirating from three principal causes will be exerted on the reactor vessel support syster.
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.
Trancient differential pressures across the core berrel within the reacter vessel.
The blowdown iet forces are edecuately understood and desion precedures are available to account for them. Both cf the " differential pressure" forces, however, are three-dimensienal and tine dependent and recuire scphisticated analytical precedures to translate ther into loeds acting on the reactor vessel sucport syster. All of the loads are resisted by the inertia and by the support members and restraints of other corpenents of the primary coolant syster including the reactor pressure vessel supports.
The transient differential pressure acting externally on the reacter vessel is a result of the flew of the blowdown effluent in the reactor cavity. The racnitude and tbc tire dependence of the resultino forces depends on the nature end the size of the cipe rupture, the clearance between the vessel and the shield and the size and lccation cf the vent openings leading from the cavity to the centsinnent as a whcle.
For scre tire refined analytical rethods have been available for calculatina these transient differential pressures (rulti-node analyses). 'The results of such analyses indicate that the consecuent loads on tre vessel support syster calculated by less sephisticated cethods may nct be es conservative as criainally intended for earlier desiens. Attachnent 1 to this enclosure provides for your information a list cf informatien recues'.s for which respcoses could be needed for a proper assessrent of the impact of the cavity dif ferential oressure en the design adecuacy of the vessel support syster for a power plant.
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The centrollinc loads for desicn purocse - however, appear in typical cases, to be these associated with the internal differential pressurec acrces the core barrel. The internally cenerated loads are due to a acrentary differentia? pressure which is calculated to exist acrcss the core barrel when the pressure in the reactor annular recion between the cere barrel and vesse) wall in the vicinity of the ruptured pipe is assured to rapidly decrease to the saturation pressure of the primary coolant due to the cutflow of water. Although the depressurization wave travels rapidly around the core barrel, there is a finite period cf tire during which the pressure in the annular recien cpposite the break Iccation is assured to remain at, er near, the original reactor coeretinc pressure. Thus, trarsiert asy:nretrical forces are exerted on the core barrel and the vessel well which ultimately result in transient leads en the support systers. These are the Icads which were underesti: rated by the licens e cricirally reportire this probler and which may he underesti: rated in other cases. They are therefore of ceneric concern to the staff. Attachrent 2 to this ercicsure provides for ycur information a list of infor: ration recuests for which resperses would te needed for a procer essessrent of the impact that the vessel internal differential pressure, in conjunction with the other cercurrent Icads, could have en the desien adecuccy of the support syster.
In that there are considerable differences in the reactor support systerr desians for various facilities and crobably in the desien trarcirs provided by the desicners of older facilities, the underestireticn of these " differ-ential pressure" loads rey or ray not result in a deterrinaticn that the adectuacy of the vessel support syster for a specific facility is cuestion-acle. Since local failurcs in the vessel supports (such as plastic deforration) do not necessarily lead to the failure of the supports as an intecral syster, there ray be scre li'ited reactor vessel roticn provided that oc further sionificant consecuences would ensue and the energency core ecolinc syste:rs (ECCS) wculd be able to perforr their design functions.
O70
ATTACli:!E:1T 1 C0tlTAlf!MEf1T SYSTEMS BRA?lCH REQUEST FOR ADDITIO !AL If1FOR"ATI0t; In the unlikely event of a pipe rupture.inside major component subcompartments, the initial blowdown transient would lead to non-uniform pressure loaa'ngs 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 (pump suction and discharge) reactor ccolant system pipe ruptures within the reactor cavity and pipe penetrations.
(b) Describe the nodalization sensitivity study performed to determine the minimum number of volume nodes required to conservatively predict the maximum pressure within the reactor cavity.
The nodalization sensitivity s tudy should include consideration of spatial pressure variation; e.g., pressure variations circumferentially, axially and radially within the reactor 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.
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(f) Provide and justify values of vent loss ccefficients and/or friction factors used to calculate 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 (such as insulation, ducting, plugs, and seals) were treated.
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 rato and energy release rate as a function of time for the reactor cavity d'esign b.
's accident.
(i) Graphically show the pressure (psia) and differential pressure (psi) responses as functions of time for each node. Discuss the basis for 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 cevity or whether it is spatially varied.
(Standard Review Plan 5.2.1.2, Subcompal ' ment Analysis attached, provides additional guidance in establishing acceptable design values, for determining the acceptability of the calculated results.)
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ATTACH!!ENT 2 MECHANICAL ENGINEERING BRANCH REQUEST FOR ADDITIONAL INFORMATION Recent analyses have shown that reactor pressure vessei 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 therefore necessary to reassess the capability of the reactor coolant system supports 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.
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 sterials 6f con-struction.
2 '.
Specify -the detail design loads used in tne 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.
3.
Provide the information requested in 2 above for the RV supports con-sidering a postulated break at the cold leg nozzle.
Include a summary of the analytical methods employed and specifically state the effects of short term pressure differentials across the core barrel in combination 1485 073
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with all external loadings calculated to result from the required
,pos tul ate.
This analysis should consider:
(a) limited displacement break areas where applicable (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 exceeding previous design limits provide an evaluation of the following:
(a)
Yield behavior (effects of possible strain energy buildup) of the material used in the reactor support design and the effect on the loads transmitted to the reactor coolant system and the backup structures to which the reactor coolant system supports are attached.
(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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