ML20028F613

Seismic Analysis of Reactor Vessel Internals.

ML20028F613
Person / Time
Site:	Oyster Creek
Issue date:	08/31/1982
From:	Oates J, Pyron J, Shabazian A GENERAL ELECTRIC CO.
To:
Shared Package
ML20028F611	List: ML20028F610 ML20028F613
References
TASK-03-06, TASK-3-6, TASK-RR B11-00220, B11-220, NSED-55-0682, NSED-55-682, NUDOCS 8302020311
Download: ML20028F613 (10)

Text

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NSE0-55-0682 DRF #Bil-00220 1

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OYSTER CREEK SEISMIC ANALYSIS OF REACTOR VESSEL INTERNALS

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.i J.W. Pyron A. Shabazian 0 Approved by: A J.H(/0ates, Manager Plant Systems & Structural Analysis l

i August 1982 I

General Electric Company Nuclear Services Department 8302020311 830124 PDR ADOCK 05000219 P PDR

OYSTER CREEK SEISMIC ANALYSIS

1.0 INTRODUCTION

This report presents an estimate of the maximum seismic loadings and resultant stresses calculated to occur in the Oyster Creek nuclear reactor pressure vessel f nternals when nubjected to a " Maximum Credible Earthquake plus Normal Operating Loads" event. Although previous seismic analyses have been performed for the Oyster Creek reactor building and pressure vessel in References 1 and 2, detailed modeling and subsequent response of the vessel internals was not included. In the present report, the vessel internals seismic loads predicted for more recent General Electric reactors, which included detailed internals modeling, are utilized to calculate the maximum Oyster Creek stresses expected f rom the above described loading event. Hand ca lculation techniques were used to evaluate the vessel internal components previously selected for analysis by the Jersey Central Power and Light Company in Reference 6. No acceptance criterion is specified for any of the stresses calculated in this report.

2.0 RESULTS Table 1 summarizes the maximum loads and accelerations predicted for reactor vessel internal components in seismic analyses of BWR-4, 5 and 6 General Electric product line reactors. The final stress calculation for a particular internal component was made by applying l

the largest loading listed together with normal operating (or greater) loads to the appropriate Oyster Creek vessel internal counterpart.

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Table 2 summarizes the maximum stress and "g" loading values i

i calculated to occur in the Oyster Creek internals analyzed for

" Maximum Credible Earthquake and Normal Operating Loads" condition.

The applicable faulted condition allowables are also tabulated for l

comparison along with the percent allowable stress or "g" loading reached. It is seen that all values are below allowable with the highest stress region being the shroud lugs which are stressed to l approximately 68% of the allowable value for the faulted loading l

l condition analyzed.

3.0 ANALYSIS 3.1 Seismic Loading A reasonable estimate can be made of Oyster Creek vessel internals seismic loads by using the maximum values predicted for a dynamically similar reactor system (including soil, building, etc.) in which current state-of-the-art analysis methods are applied. In the present analysis, a conservative approach wa.s taken by using the largest seismic loads predicted for three typical General Electric product line reactors. Pressure vessel internals loads were extracted from seismic analyses of representative plants in the BWR-4, 5 and 6 product lines. ,

The SSE (Safe Shutdown Earthquake) design spectra and resultant loads of these BWR 4, 5 and 6 reactors were normalized, where appropriate, to the Oyster Creek zero period acceleration design value originally recommended in Reference 3 for " Maximum Credible Earthquake". The

> resultant BWR-4, 5 and 6 " free field" design spectra completely envelope the Oyster Creek design spectrum thereby providing an initial degree of conservatism. Ideally, in order to assure that this conservatism carries through the soil and reactor building to the presse.re vessel and internals, a similar enveloping of the Oyster l Creek Reference 2 dynamic model nodal spectra is desirable. Since the. Reference 2 analysis did not calculate model nodal response spectra, it is not possible to assure consistency with regards to frequency / amplification content. However, it is judged that the conservatism inherent with using the largest loads resulting from l

seismic analysis of three different General Electric product line reactors would tend to offset the above concerns of frequency / amplification content enveloping at vessel support locations.

3.2 Normal Operating Loads Internal pressure loading (steady-state pressure differences) at rated conditions were obtained from the Reference 4 Reload Coolant Hydraulics Data Book for Oyster Creek. Pressure loads not listed in Reference 4 were estimated based on data from a similar plant in Reference 5.

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3.3 Stress Calculations The analytical methods used in this analysis are limited to " hand calculation" techniques as specified in Reference 6. The reactor pressure vessel internals analyzed were:

o Shroud Head (Bolts & Lugs) o Shroud o . Shroud Support o Top Guide o Fuel Channel o Core Plate o Control Rod Drive Guide Tube (CRD) 3.3.1 Shroud Head Region Shroud Head Bolts The maximum Oyster Creek bolt loading calculated is considerably

> below the maximum loading listed in Reference 7 for faulted SSE:

Oyster Creek Max. Cr. Reference Test  % of Earthquake + Nor. Loads Faulted SSE Allow.

Axial Load 12,200 lb/ bolt 53,800 lb/ bolt 23%

Shear Load 7,700 lb/ bolt 15,200 lb/ bolt 51%

, 3.3.2 Shroud Head Lugs.

4 The Maximum Stress Intensity calculated for the shroud head lugs is also found to be below the appropriate faulted allowable value: the maximum lug / weld stress intensity was found to be 34,700 psi which is 68% of the 3 Sm stress intensity faulted allowable of 50,700 psi.

3.3.3 Shroud Buckling The maximum axial shroud membrane stress was calculated to be 4,058 psi which is 24% of the allowable faulted critical stress value of 17,100 psi.

3.3.4 Shroud Support Region The shroud support region was analyzed in Reference 8 for various pressure loadings combined with an overturning ' earthquake load of 0

approximate 1y' 48 x 10 in-lb. - The stresses f rom' that analysis were adjusted-to reflect the effect of the maximum BWR-4 & 5 moment of 153 x 100in-lb with the steam _line break ' pressure condition defined therein (this load case is beyond that specified for the present analysis but was used for expediency).

The resultant maximum stress intensities are calculated to be :

Shroud Support-Cone Shroud Above Core Support Ring Si ,,,= 40,980 psi Si ,,,= 29,200 psi 1 Allowable: Allowable :

3Sm = 69,900 psi 3Sm ='50,700 psi i

~(59% of allowable) (58% of allowable) 3.3.5 Top Guide Calculctions for the top guide show the weight of the top guide in water is greater than the combined upward force due to vertical seismic acceleration and normal pressure. Therefore, the top guide will not lift during the postulated seismic' event.

Stresses due to lateral earthquake loading were not examined as this calculation would entail finite element modeling. However, a similar analysis result is offered from a finite element analysis previously performed on a similar size BWR-3 top guide with a quasi static earthquake loading of 0.6g. Those results indicate the Oyster Creek top guide would not be overstressed.

3.3.6 Fuel Channels Maximum estimated Oyster Creek fuel channel loads are compared to the design basis earthquake loads extracted from Reference 9 in the table below:

s-

Calculated Design Basis i

' Horizontal Shear at Top (1b) 226 892 BWR-5 Horizontal Shear at Bottom (1b) 175 871

. LMaximum Bending Moment (in-lb) 12,100 43,200 BWR-4 Horizontal Acceleration (g's) 1.19 30 Vertical Acceleration (g's) 0.32 0.75 3 3.7 Core Plate Based on past analysis, the core plate structure is limited by a beam buckling type collapse due to vertical loading. The maximum stress calculated to develop from vertical seismic loading and normal pressure is 5180 psi which is 42% of the critical faulted allowable stress,a = 12,402 psi.

crit.

3.3.8 CRD Guide Tube Similar guide tubes were analyzed in Reference 10 using a vertical-load of 1.0g. The resulting stresses were considerably below faulted allowable stresses. Since the maximum vertical g load for Oyster Creek is estimated to be 0 31g the guide tube structure is deemed to be substantia 11y'below faulted limits.

4.0 CONCLUSION

S Based on the assumptions inherent with the contained analysis, the following conclusions are made:

- The most highly stressed region in the Oyster Creek pressure vessel internals is predicted to occur in the shroud head lugs.

The maximum stress is calculated to reach approximately 68% of the allowable stress for the " Maximum Credible Earthquake and Normal Loads" f aulted condition analyzed.

- The foregoing analysis represents a best estimate of the maximum stresses /g loads calculated to occur under the postulated seismic loading without resorting to more expensive detailed finite element modeling techniques.

5.0 REFERENCES

1. " Seismic Analysis of' Reactor Building Revised", Jersey Central Reactor Project, by John A. Blume & Associates, Engineers, June 18, 1965.

2. " Earthquake Analysis: Reactor Pressure Vessel", Jersey Central Reactor Project, by John A. Blume & Associates, Enginecr, March 16, 1966.

3. " Recommended Earthquake Design Criteria, Jersey Central Nuclear Power Plant", by George W. Housner, March 4, 1964.

4. " Reload Core Coolant Hydraulics Data Book," for Oyster Creek Cycle 9, 458EA23, Rev. 1.

5. " Reload Core Coolant Hydraulics," NMP-1 Cycle 9, 459HA497.

6. " Reactor Vessel Internals Seismic Load-Calculations", General Electric EWA EAF 93-BV, by E. H. Ehrlich/R. Panek, February 10, 1982.

7. "0yster Creek Replacement Shroud Head Bolts," Design Report, by D. F. Holland, NSE-46-0981, September 30, 1981.

8. "0yster Creek Lower Shroud Stress Analysis," by J. W. Pyron, Document #257HA761, January 24, 1969.

9. "BWR/2-5 Loadings on Fuel Assemblies," Data Information transmittal, to F. E. Cooke from T. G. Dunlap, #M25776027, February 12, 1976.

10. " Control Rod Guide Tube-Stress Report," 22A4678, Rev. 3.

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s TABLE 8 OTSTER CREEK WESSEL INTERNALS MARIMUM STRESSES & ACCELERATIONS CALCULATED FRON EWR-4, 5 OR 6 SEISMIC AN4LTSES: ACCELERATION SPECTRA. g S = 0.22 g's ('7A)

MAX. ALLOWABLE VALUES MAXIMUM CALCULATED OTSTER CREEE ,

PAULTED CONDITION INTERNAL STRESSES & ACCEL's.

RPV INTERNAL IAADINC STRESS ACC.(g's) ACC.(g's) STRESS HOR. ACC. VERT. ACC. Z OF COMPONENT CRITERIA (PSI) HOR. VERT. (PSI) (g's) (g's) ALLOWABLE SHROUD HEAD LUCS STRESS 35 ,= - -

34.700 - - 68%

INTENSITT 50,700 SHROUD BUCKLING 1s,100 - -

4058 - -

24%

SNROUD SUPPORT STRESS 3S,= - - 40,980 - - 59%

REGION ITTENSITT 69.900 TOP CUIDE NO LIFT - - - - - -

NO LIFT FUEL CHANNEL STRESS / - 3.0 0.75 - 1 19 .32 43%

g LOAD CORE PLATE COLLAPSE 12,402 - -

5,190 - .30 42% .

CRD CUIDE TUBE g LOADINC - NOT 1.0 - - .31 31%

LIMITING Noter ZPA denotes "rero period acceleration" value.

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