ML19220C314

Requests Review of Design Bases for Reactor Vessel Support Sys to Determine If Transient Loads Were Appropriately Accounted for in Design

ML19220C314
Person / Time
Site:	Crane
Issue date:	06/30/1976
From:	Vassallo D Office of Nuclear Reactor Regulation
To:	Arnold R METROPOLITAN EDISON CO.
References
NUDOCS 7904300170
Download: ML19220C314 (8)

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Metropolitan Edison Ccmpany George F. Trewbridge, Esq.

Shaw, Pittman, Potts & Trewbridge 910 17th S treet, N.

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Washington, D. C.

20006 Chauncey R. Kepford, Esq.

Chairman York Ccmnittee for a Safe Environment 2586 Broad Street York, Pennsylvania 17404 Mr. Richard W. Heward Project Manager GPU Service Corporation 260 Cherry Hill Road Parsippany, New Jersey 07054 s

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ENCLOSURE I REQUEST FOR ADDITIONAL INFOR."ATIGN 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 sever e 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 conditicn.

The foilcwing information shculd be included in your reassessment of the reactor vessel supports and reactor cavity structure.

3.89 Provide engineering drawings of the reactor support system sufficient to show the gecmetry of all pri alements and materials of construction.

3.90 Specify the 2 tail 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 r.:aximun st each principle element of the support system and the corresponding allcwable stresses.

3.91 Provide the information requested in 2 above considering a postulated break at the dcsign basis location that resolts in the most severe loading condition for the reactor pressure vessel supports.

Include 09 2[_ $

._ a summary 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 should 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.

3. 9'd 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.

3.33 Address the adequacy of the reactor coolant systen piping, control rod drives, steam generator and pump supports, structures surrounding the reactor coolant s stem, [ core support structures, fuel assemolies.

s other reactor internals

....] and ECCS piping for coth the elastic and/or inelastic analyses to assure that the reactor can be safely brought to cold shutdown.

For each item include the trethod of omE o g!

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. 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 valeas.

e The compartment m. it:-node press ire response analysis should include the following inf" mation.

3.94 TI.e results of ana h ;n r# '* e differential pressures resulting i

frcm het leg and con (ptmp suct.on and discharge) reactor coolan'. s."si. m Dire ruptures within the reactor cavity and pipe penetrations.

3.95 Describe the nodalization sensitivity study performed to detennine the minimum number of volume nodes required to conservatively predict the maximum pressure within the reactor cavity.

The nadalization sensitivity study should include consideration of spatial pressure variation; e.g., pressure variations circumferentially, axially and radially within the reactor cavity.

3.9c Provide a schematic drawing showing the nadalization of the reactor cavity.

Provide a tabulation of the ncdal net free volumes and interconnecting flow path areas.

3.97 Provide sufficiently detailed plan and section drawings for several views showing the arrangement of the reactor cavity structure, reactor vessel, piping, and other r/ajor obstructions, and vent areas, to permit verification of the reactor cavity nodalization and vent locations.

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_4-3.98 Provide and justify the break type and area used in each analysis.

3.99 Provide and justify values of vent loss coefficients and/or friction factors used to calculate flow between nadal volumes.

When a loss coefficient consists of more than one ccmponent, identify each component, its value and the flow area at which the loss coefficient applies.

3.1';0 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 area.

Provide justification that vent areas will not be partially or ccmpletely pluggt. by displaced objects.

3.101 Provide a table of blowdownmass flow rate and energy release rate as a function of time for the reactor cavity design basis accident.

3.102 Graphically show the pressare (psia) and differential pressure (psi) responses as functions of time for each node.

Discuss the basis for establishing the differential pressures.

3.103 Prcvide 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 reacter cavity or whether it is spatially varied.

In order to review the methcds employed to compute the asymmetrical pressure differences across the core support barrel during the subccaled portion of the bicwdown analysis, the following information is requested:

2.ic-A complete description of the hydraulic code (s) used including the m7 n

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development of the equations being solved, the assumptions and simplifications used to solve the equations, the limitaticos resulting from these assumptions and simplifications and the numerical methods used to solve the final set of equations.

3.105 In support of the hydraulic code (s) used provide comparisons with the ccde(s) to applicable 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-02-8 The models developed should be based on the assumptions proposed fcr the analysis of a PWR.

3.iO6 Provide a detailed description of the model proposea 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.

3.107 Typically the current generation of hydraulic subcooled bicwdown 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).

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