ML19260A048
| ML19260A048 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 06/09/1976 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | Arnold R METROPOLITAN EDISON CO. |
| References | |
| NUDOCS 7910290678 | |
| Download: ML19260A048 (8) | |
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DISTRIETION:
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EC PDR Dross L PDR CTrammell Docket No. 50-289 Docket Fi ORB #4 Rdg VStello AR 9 1978 KRcoller TJCarter Metropolitan 2dison Company OELD ATTN:
Mr. R. C. Arnold OI&E (3)
Vice President - Generation P. O. Box 542 RReid Readinp, Pennsylvania 1%C3 CNelson RIngram Centleren:
DEisenhut JRBuchanan RE: THREE MILE ISLAUD l' NIT NO. 1 TBAbernathy On October 15, 1975, we informed you of a potential safety question which has been raised regarding the design of reactor pressure vessel support sy s t e-'s. Pe recuested that you review the desien bases for the reactor vessel support system for your f acility to deternine whether the transient loads described in the enclosure to our letter were appropriately taken into account in the desipn.
Your reply of Hovenber 21, 1975, indicated that the transient differential pressures in the annular region between the reactor vessel and the cavity shield wall and across the core barrel were not considered in the support design.
In our letter of October 15, 1975, we indicated that on the basic of your initial review, a reassessnent of the vessel support design might be required. We have now determined that such a reassessrent is required.
As you are probably aware, we have been discussing with the PWR vendors and various architect / engineer firms the peneric aspects of this problen. Should you contenplate utilizing organizations other than your PWR vendor for calculation of the sub-cooled internal loads, we surgest you contact us for the benefit of a brief review of our generic discussions to date. We will continue these ceneric discussicns with the vendors and architect /eneineers, but such discussions are not intended to pace your evaluation of this concern nor to eliminate the oossibility that we may have addit ional questions reccrdin:; your evaluat ion e.f ter sub.ittal.
While the e phasis aiven in this letter deals with the reactor vessel cavity, for your information and 2uidance our generic review mcy consider other areas in the nuclear steam supply systen and further evaluation may be required.
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39troeolit an E ilson Ccrpany Please inform us within 30 days af ter receipt of this letter of your schedule for providing us your evaluation of the adecuacy of the pressure vessel supports when the sub cooled loads are calculated and taken into account in a manner which you determine best represents these pheno ~ena.
Your evaluation should include the answers to the attached request for additional in fo rmat ion.
This request for generic information was approved by CAO blanket clearance nuebe-- B-130223 (R0072). This clearance expires July 31, 1977.
Sincerely, crtginq lM Robert W. Peid, Chief Operating Reactors Branch e4 Division of Operating Feactors
Enclosure:
Request for Additional Information cc w/ encl:
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F[j r E?; '
o G. F. Trowbridge, Esq.
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g Shaw, Pittman, Potts, & Trowbridge ba s
1300 M Street, ii. W.
Washint;on, D. C.
20036 CPU Service Corporation Richard W. Heward, Project Manager Thomas N. Crimmins, Jr., Safety and Licensing Mananer 260 Cherry Hill Road Parsippany, New Jersey 07054 Pennsylvania Electric Co pany ifr. K. W. Conrad Vice Pre:ident, Ge ne ra t ion 1001 Broad Street Johns town, Pennsylvania 15907 1479 078
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etronalitan Edison Co?.psny 3
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fr. We ldon 3. A reh art, Chairan Board of Supervisors of Londonberry Township 2148 Foxiana Road Middletovn, Pennsylvania 17057
- iss Mary V. Southard, Chairman Citizent for a Safe Environ.ent m
P. O. Ec.x 405 I!arrisburg, Pennsylvania 17105 Covernment Publications Section State Library of Pennsylvania Eox 1601 (Education Suilding)
Harrisburg, Pennsylvanic 17126 j
1479 079
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Form AEC.318 (Rev. 9 53) AECM 0240 W u. s. soVERNMENT PRINTl. G QFFiggg 1974 5 SelGG
RE00EST FOR ADDITIO::AL If!FOR"ATI0fl Recent analyses have shown that reactor pressure vessel supports may be subjected to previously underestimated lateral loads under the conditions that result from the postulation of design basis ruptures of the reactor coolant piping at the reactor vessel nozzles.
It is therefore necessary to reassess the capability of the reactor coolant system supports to assure that the calculated motion of the reactor vessel'under the most severe design basis pipe rupture condition will be within the bounds necessary to assure a high probability that the reactor can be brought safely to a cold shutdown condition.
The following information should be included in your reassessment of the reactor vessel supports and reactor cavity structure.
1.
Provide engineering drawings of the reactor support system sufficient to show the geometry of all principle elements and materials of construction.
2.
Specify the detail design loads used in the original design analyses of the reactor supports giving magnitude, direction of applic.ation and the basis for each load.
Also provide the calculated maximuu stress in eacr. nrinciple element of the support system and the corresponding allowable stresses.
3.
Provide the information requested in 2 above considering a postulated break at the design basis location that results in the most severe loading condition for the reactor pressure vessel supccrts.
Include 1479 080
. a surmary of the analytical methods employed and specifically state the effects of asymmetric pressure differentials across the core barrel in combination with all external loadings including asymmetric cavity pressurization calculated to result from the required postulate.
This analysis shou.ld consider:
(a) limited displacement break areas where applicable (b) consideration of fluid structure interaction (c) use of actual time dependent 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)
Inelastic behavior (including strain hardening) of the material used in the reactor support design and the effect on the load transmitted to the reactor coolant system and the backup structures to which the reactor coolant system supports are attached.
5.
Address the adequacy of the reactor coolant system piping, control rod drives, steam generator and pump supports, structures surrounding the reactor coolant system, [ core support structures, fuel assemblies, other reactor internals
....] and ECCS piping for both the elastic and/or inelastic analyses to assure that the reactor can be safely brought to cold shutdown.
For each item include the method of 1479 081
,. analysis, the structural and hydraulic computer codes employed, drawings of the models employed and comparisons of the calculated to allowable stresses and strains or deflections with a basis for the allowable values.
The ccmpartment multi-node pressure response analysis should include the following information:
6.
The results of analyses of the differential pressures resulting frca hot leg and cold leg (pump suction and discharge) reactor coolant system pipe ruptures within the reactor cavity and pipe penetrations.
7.
Describe the nodalization sensitivity study performed to determine the mi;1imum number of volume nodes required to conservatively predict the maximum pressure within the reactor cavity.
The nodalization sensitivity study should include considerat!an of spatial pressure variation; e.g., pressure variations circumferential1y, axially and radially within the reactor cavity.
8.
Provide a schematic drawing showing the nadalization of the reactor cavity.
Provide a tabulation of the nodal net free volumes and interconnecting flow path areas.
9.
Provide sufficiently detailed plan and section drawings for several views showing the arrangement of the pactor cavity structure, reactor vessel, piping, and other major obstructions, and vent areas, to permit verification of the reactor cavity nodalization and tent i
locations.
1479 082
... 10.
provide and justify the. break type and area used in each analysis.
11.
Provide and justify values of vent loss coefficients and/or friction factors used to calculate ficw between nadal volumas.
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.
12.
Discuss the manner in which movable obstructions to vent ficw (such as insulation, ducting, plugs, and seals) were treated.
Provide analytical justification for the removal of such ite.ms to obtain vent area.
Provide ~ justification that vent areas will not be partially or completely plugged by displaced objects.
- 13. Provide a table of blowdownmass ficw rate and energy release rate as a function of time for the reactor cavity design basis accident.
14.
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.
15.
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 thr: design differential pressure is uniforgly applied to the reactor cavity or whether it is spatially varied.
In order to review the methods employed to ccmpute the asyn=etrical pressure differences across the core support barrel during the subcooled portion of the blowdown analy-is, the following information is requested:
- 16. A ccmplete description of the hydraulic code (s) used including the 1479 083
.c
. development of the equations being solved, the assumptions and simplifications used to solve the equations, the limitations resulting from these a~sumptions and simplifications and the s
numerical methods used to solve the final set of equations.
17.
In support of the hydraulic code (s) used provide comparisons with the code (s) to applicabie experimental tests, including the following:
(a). CSE tests B-63 and B-75 (b). LOFT test L1-2 (c). Semiscale tests S-02-6 and S-02-3 The models develcped should be based on the assumptions proposed for the analysis of a PWR.
18.
Provide a detailed description of the model proposed for your plant and include a listing of the input data used and a time zero edit.
Identify the assumptions used in developing the model, specifically the treatment of area, length and volume.
19.
Typically the current generation of hydraulic subecoled blowdown analysis codes solve the one-dimensional conservation equations.
However, they are used to model the multi-dimensional aspects of the reactor system (i.e. the downcomer annulus region).
Provide justification for the use of the code (s) to model multi-dimensional regions, including the equivalent representation of the region as modelled by the code (s).
1479 084
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