ML19316A067

Forwards Request for Addl Info Re Reassessment of Reactor Pressure Vessel Supports & Reactor Cavity Structure

ML19316A067
Person / Time
Site:	Oconee
Issue date:	06/08/1976
From:	Schwencer A Office of Nuclear Reactor Regulation
To:	Parker W DUKE POWER CO.
References
NUDOCS 7911210637
Download: ML19316A067 (8)

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Duke Power Company ATT":

tir. Willias 0. Parkerg Jr.

Vice President Steas Production

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422 South Church Street Charlotte, North Carolina 28242

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on October 15, 1975, Je' informed you of a potential' safety question'

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which has been raised regarding theidesign of reactor pressere vessel'.

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support systems. We requested that 'you review the design bases for?

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the reactor vessel support systes for your facility to determine whether the transient ' loads described in the enclosure to cor 1etter'

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Your reply of Novecher 14, 1975, indicated that'the transient e

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differential pressores in the annular region between the reactor'

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vessel and the cavity shield vall and across the core barrel were- [.Z MS ~ l

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not considered in the support design..

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s In our letter of October 15, 1975, we indicated that.on the bas!s

- 9 of your initial review,.a resasessment of the vessel su.oport design might be roonired. We have noa deternined'that such a reassessment-is required.

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As you are probably aware, we have Men discussing with the. PSR (

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other than your PWE venderfforfcalg'alationlof the sub-cooled

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internal' loads, weTsuggestoy4s "centact'us for the knefit. of Ja.

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brief review'of eer "generie'~dNeises' lens 'to date., We will costinue *M these generic discuarises 41th theItendors and' architect / engineers,. ~'

i but such discussions areleet inteedeJ 4e' pace your evaluation of.

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l additional'qars tlose regardlag your. evaluation af ter submittal.

_,a While the emphasisgiven in this letter; deals with th.a reactor vesset ^ (( ~ ', -

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Please iaform us iritbin:30fdaysiafterJruc4,tJ ofl Nile letter

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2 of your schedule for providlag us year evalwellen of the I

ade wacy of the pressere.vesse!~ supports when the sub-cooled l'esde are cateulated and taken into account in a sanperwhick you determine best represents these pheneser.4. Your evaluaties shovid include the saavers to the attached request for additiona!

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. June 8, 1976 Duks Power Conpany cc: : r. k'illias L. Porter

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Duke Power Company P. O. Cox 2175 422 South Church Street Charlotte, ;;crth Carolina 28242

!c. Troy B. Conner Connor G Knotts 1747 Pennsylvania Avenue,17.1 1.'ashington, D. C.

20005 Oconee Public Library 201 South Spring Strcot 1lalhalla, South C..rolina 29691 O

4

REQUEST FOR ADDITIONAL INFORMATION Recent analyses have shown that reactor pressure' vessel supports may be subject ed to previously underestimated lateral loads under the condition: that result from the postulation of design basis ruptures of the reacto coolant piping at the reactor vessel nozzles.

It is therefore recessary to reassess the capability of the reactor coolant system supp)rts to assure that the calculated motion of the reactor vessel unde, 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.

Speci'y the detail design loads used'in the original design analyses of t.e re c tor supports giving magnitude, direction of application and the basis for each load.

Also provide the calculated maximum stress in each p: inciple element of the support system and the corresponding allowable stresses.

3.

Provide the information requestod in 2 above considering a postulated break at the design basis lact. tion that results in the most severe loading condition for the reactor pressure vessel supports.

Include a sumary of the analytical methods employed and specifically state the effects of asymetric pressure differentials across the core barrel in combination with all external loadings including asymetric 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.

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

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- 1 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 compartment multi-node pressure response analysis should include the following infomation:

6.

The results of analyses of the differential pressures resulting from hot leg and cold leg (pump suc-in and discharge) reactor coolant system pipe ruptures within che reactor cavity and pipe penetrations.

7.

Describe the nadalization sensitivity study performed to detemine 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 circumferentially, axially and radially within the reactor cavity.

8.

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.

9.

Provide sufficiently detailed plan and section drawings for several views showing the arrangement of the reactor cavity structure, l

reactor vessel, piping, and other major obstructions, and vent areas, to permit verification of the reactor cavity nodalization and vent i

l locations.

. 10.

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

Provide and justify values of vent loss coefficients and/or friction 11.

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.

12.

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 completely plugged by displaced objects.

13.

Provide a table of blowdownnass flow 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.

Ofscuss 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 the design differential pressure is uniformly applied to the reactor cavity or whether it is spatially varied.

In order to review the methods employed to compute the asymmetrical pressure differences across the core support barrel during the subcooled

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portion of the blowdown analysis, the following information is requested:

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16. A complete description of the hydraulic code (s) used including the

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.. development of the equations being solved, the assumptions and simplifications used to solve the equations, the limitations resulting from these assumptions and simplifications and the nwnerical 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 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 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 carrent generation of hydraulic subcooled 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).

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