ML20077A708

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Rev 3 to NSSS Equipment & Piping, Vol Viii to Seismic Margin Review
ML20077A708
Person / Time
Site: Midland
Issue date: 07/31/1983
From: Campbell R, Kennedy R, Kipp T
STRUCTURAL MECHANICS ASSOCIATES
To:
Shared Package
ML20077A687 List:
References
SMA-13701.05, SMA-13701.05R003-V08, SMA-13701.05R3-V8, NUDOCS 8307220411
Download: ML20077A708 (35)
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Text

_ _ _ _ _ _ _ _ _ _

SMA 13701.05-R003 (VOLUME VIII) l SEISMIC MARGIN REVIEW MIDLAND ENERGY CENTER PROJECT l

VOLUME VIII NSSS EQUIPMENT AND PIPING by T. R. Kipp R. D. Campbell Approved:

Approved: [,h. IM

/

R. P. Kennedy R. B. Narver President Acting Manager of Quality Assurance prepared for CONSUMERS POWER COMPANY Jackson, Michigan July, 1983 g g STRUCTURAL mECHRnlCS 830722o412 sao7 3 DR ADOCK 05000 RSSOCIRTES A

WM A C alif Corp.

516oBachStreet NewportBeach,Cakf.9266o (714)833 7552

1 REVISIONS l

Document Number SMA 13701.05-R003 (VOLUME VIII)

Title Seismic Margin Review Midland Energy Center Project NSSS Equipment and Piping Rev.

Description QA Project Manager 4/83 Draft issued for review fd, h. M g g sp Hgs /s3 04

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SEISMIC MARGIN REVIEW MIDLAND ENERGY CENTER PROJECT TABLE OF CONTENTS VOLUME NO.

TITLE I

METHODOLOGY AND CRITERIA II REACTOR CONTAINMENT BUILDING III AUXILIARY BUILDING IV SERVICE WATER PUMP STRUCTURE V

DIESEL GENERATOR BUILDING VI BORATED WATER STORAGE TANK VII ELECTRICAL, CONTROL, INSTRUMENTATION AND MECHANICAL EQUIPMENT

- VIII NSSS EQUIPMENT AND PIPING IX BALANCE-OF-PLANT CLASS 1, 2 AND 3 PIPING, PIPE SUPPORTS AND VALVES X

MISCELLANE0US SUBSYSTEMS AND COMPONENTS 6

TABLE OF CONTENTS (VOLUME VIII)

Section Title Page LIST OF TABLES..................

ii LIST OF FIGURES iii 1

INTRODUCTION...................

1-1 2

CRITICAL REGION SELECTION 2-1 2.1 NSSS Piping and Equipment Supports 2-1 2.2 Reactor Vessel Internals 2-2 3

SEISMIC MARGIN METHODOLOGY,...........

3-1 4

SUMMARY

AND CONCLUSIONS 4-1 5

EVALUATION OF NSSS PIPING AND EQUIPMENT SUPPORT 5-1 6

EVALUATION OF REACTOR VESSEL INTERNALS......

6-1 REFERENCES l

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VIII-i l

.. _,. _.. - _ _.. - _.. _. _. - _ _. - ~

i LIST OF TABLES Table Title Page VIII-2-1 Selected Critical NSSS Piping and Equipment Support Locations..............

VIII-2-3 VIII-2-2 Selected Critical Reactor Vessel Internals Locations..................

VIII-2-4 VIII-4-1 Summary r-Seismic Margins for Selected NSSS Piping and Equipment Supports VIII-4-2 VIII-4-2 Summary of Seismic Margins for Selected Reactor Vessel Internals VIII-4-3 VIII-5-1 Comparison of NSSS Piping and Equipment Support Faulted Condition pynamic Load Level Responses VIII-5-2 VIII-6-1 Comparison of Reactor Vessel Internal Faulted Condition pynamic Stress Responses VIII-6-2 4

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LIST OF FIGURES Figure Title Page VIII-2-1 Midland Loop Math Model (View 1)......

VIII-2-5 VIII-2-2 Midland Loop Math Model (View 2)......

VIII-2-6 VIII-2-3 Midland Loop Math Model (View 3)......

VIII-2-7 VIII-2-4 Midland Plan View D-Ring 2.........

VIII-2-8 VIII-2-5 Midland Loop Mathematical Model Isometric

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D-Rings 1, 2, 3 & 4 VIII-2-9 VIII-2-6 Midland RV Isolated Math Model Elevation View....................

VIII-2-10 VIII-2-7 Reactor Pressure Vessel Support System...

VIII-2-11 VIII-2-8 RPV Upper Lateral Support Arrangement VIII-2-12 VIII-2-9 RPV Upper Lateral Support Bracket Detail..

VIII-2-13 VIII-2-10 Unit 1 RPV Anchor Stud Arrangement.....

VIII-2-14 VIII-2-11 RPV Anchor Stud Detail...........

VIII-2-15 VIII-iii

I 1.

INTRODUCTION Seismic analyses of the critical regions of the Nuclear Steam Supply System (NSSS) for the Midland Plant have been reviewed to determine the margin which exists between code allowable response levels and response levels associated with the Seismic ' Margin Earthquake (SME) event.

Areas of the NSS system reviewed included piping, nozzles, equipment supports and the reactor internals.

The dynamic analyses of the system components for the SME as well as for the qualification design basis SSE + LOCA faulted event were conducted by Babcock and Wilcox, the reactor vendor. SME dynamic inputs to the NSSS were provided in terms of in-structure translational and rotational time history records and in-structure response spectra at the NSSS-structure interfaces for the three soil conditions described in Volume I.

The Babcock and Wilcox effort pertaining to the Seismic Margin Study is documented in Reference 3.

For the critical regions, the results of the dynamic analyses were then compared on a load level basis to determine if response to the SME event exceeded that for the design basis event and margins against the SSE were selectively computed.

The Seismic Margin Review (SMR) of the Midland plant was limited to seismic Category 1 equipment required to achieve and maintain safe shutdown from full power operation. Since the entire primary coolant system must function properly to achieve and maintain safe shutdown, all areas of the NSS system were carefully investigated.

In accordance with the specifications of the Safety Evaluation Report (Reference 4), the critical locations within the NSSS are evaluated for te combination of dead load, live load and seismic load. As-a result, the primary stress levels are computed for an event consisting of VIII-1-1

normal static loads (pressure, dead loads, live loads and certain thermal effects) and the dynamic SME load. This load combination has been defined as a faulted event and resulting responses are compared with faulted condition allowable response values.

Chapter 2 of this report describes the criteria and considera-tions for selecting critical regions within the NSS system that are evalu-ated in this study. Chapter 3 defines the methodology utilized in developing margins against the SME while Chapter 4 summarizes the results of the review of the critical regions. The evaluation of the NSSS piping and equipment supports is presented in Chapter 5 while Chapter 6 presents the evaluation of the reactor vessel internals.

All regions of the NSS are shown to exhibit margins against the SME which exceed the margins against the design basis faulted event and to meet acceptance criteria formulated for the SME. The general accept-ance criteria, for which the NSSS components were demonstrated to comply, were presented in Chapter 9 of Volume I and are reiterated herein.

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2.

CRITICAL REGION SELECTION j

Throughout the plant, equipment most vital to achieve and main-tain a safe shutdown of the reactor were selected for independent evalua-tion of response resulting from the SME. The selection process considered criticality of function, location in the structure, and review of the FSAR to identify equipment components whose reported stress levels are particu-larly sensitive to seismic loading.

Because of its importance, all areas of the NSS System were examined in the SMR for effects of the SME combined with normal operating loads and for compliance with stated code acceptance criteria. The SME NSS System dynamic analysis was conducted by Babcock and Wilcox (B&W) for SME base mat input motion determined from the overall soil-structure interaction model of the reactor building (Volume II). The NSSS response was then computed by B&W using their existing NSSS Loop Model and their existing Reactor Vessel Isolated Model. The computer codes and mathemati-cal models enployed by B&W in their dynamic analyses of the NSS system i

are described in Sections 3.9.1.2 and 3.7.3.1 of the Midland FSAR, I

respectively. Figures VIII-2-1 through VIII-2-5 present details of the primary loop system mathematical model used in the dynamic analyses. The system model includes the reactor vessel, steam generator, primary coolant pumps, pressurizer, all associated primary piping, integral component supports, and non-integral snubbers and bumpers. Similarly.

Figure VIII-2-6 presents details of the reactor vessel isolated model showing the mathematical model of the internals. The results from these two analysis models provided force and moment data for comparison with design faulted condition loadings.

VIII-2-1

2.1 N_SSS PIPING AND NSSS EQUIPMENT SUPPORTS Fourteen locations whose stress levels are judged to be sensitive to dynamic loadings were selected as being representative of the NSS System. Table VIII-2-1 lists the selected NSSS piping and equipment support locations together with the model, figure number, and model joint identifier representing the selected location. The selected locations are also indicated on the specified figures.

Of particular interest in this analysis is the evaluation of the modified reactor vessel support system consisting of the anchor studs and the upper lateral supports as shown in Figure VIII-2-7. The arrangement of the 12 upper lateral supports and a detail of the support bracket are shown in Figures VIII-2-8 and VIII-2-9, respectively, while the arrange-ment of the inner and outer anchor stud rings and a detail of the anchor studs is presented in Figures VIII-2-10 and VIII-2-11, respectively.

Figure VIII-2-10 also identifies the three Unit 1 anchor studs which failed prior to modification of the support system. The margin against the SME presgnted in this report for the anchor studs is based upon the Unit 1 configuration.

2.2 REACTOR VESSEL INTERNALS The selection of the most highly stressed locations associated with the reactor internals was accomplished by Babcock & Wilcox as des-cribed in Reference 3.

Since normal loads are small and even negligible for most reactor vessel internals locations, the primary loads on these structures are associated with seismic and hydrodynamic events. The fifteen locations selected exhibited a ratio of calculated to allowable stress for the design basis faulted condition greater than 0.40 and were the " worst case" location for a given major internals component. The selected internals locations are listed in Table VIII-2-2 together with the Reactor Vessel Isolated model element and joint identifiers defining the region of interest. The selected locations are indicated in Figure VIII-2-6.

VIII-2-2

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TABLE VIII-2-1 l

SELECTED CRITICAL NSSS PIPING AND EQUIPMENT SUPPORT LOCATIONS Description Model*

Figure Joint l

1.

RPV Support Skirt / Base Interface RVIM VIII-2-6 50 i

2.

RPV Upper Support RVIM VIII-2-6 166 i

3.

OTSG Support Skirt / Base Mat Interface Loop VIII-2-1 23 3

4.

OTSG Upper Support Loop VIII-2-5 51, 34 5.

Pressurizer Support Lug / Support Structure Interface Loop VIII-2-4 115 6.

Pressurizer Upper Support Loop VIII-2-5 240, 241 7.

RPV 36" Hot Leg Outlet Nozzle Loop VIII-2-1 200 8.

RPV 28" Cold Leg Inlet Nozzle Loop VIII-2-2 149 9.

OTSG 36" Hot Leg Inlet Nozzle Loop VIII-2-1 15

10. OTSG 28" Cold Leg Outlet Nozzle Loop VIII-2-2 21
11. RCP 28" Cold Leg Inlet Nozzle Loop VIII-2-3 46
12. RCP 28" Cold Leg Outlet Nozzle Loop VIII-2-2 47
13. CRD Housing /RPV Interface RVIM VIII-2-6 120 l

I4. RCP Snubbers Loop VIII-2-4 76

  • RVIM = Reactor Vessel Isolated Model Loop = NSSS Half System Loop Model l

l TABLE VIII-2-2 SELECTED CRITICAL REACTOR VESSEL INTERNALS LOCATIONS i

Description Figure Bar Joint i

i 1.

Plenum Cover VIII-2-6 101 53 2.

Upper Grid Assembly - Rib Section VIII-2-6 115 14 3.

Upper Grid Pad Joint VIII-2-6 109 20 4.

Core Support Shield - Lower End VIII-2-6 136 44 5.

Core Support Shield - Upper Flange VIII-2-6 139 60 l

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Thermal Shield - Upper End VIII-2-6 130 9

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Thennal Shield / Lower Grid Shell Bolted Joint VIII-2-6 121 13 8.

Thermal Shield Upper Restraint Flange VIII-2-6 130 9

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Core Barrel Assembly - Upper End VIII-2-6 131 9

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10. Core Barrel /Former Bolted Joint VIII-2-6 142 18 j
11. Lower Grid Assembly - Top Rib Section VIII-2-6 104 13

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12. Lower Grid Assembly - Top Rib Section/Shell Forging Bolted Joint VIII-2-6 120 16 l
13. Lower Grid Assembly - Support Post / Support Forging j

Welded doint VIII-2-6 120 16 i

14. Control Rod Guide Tubes - Slotted Region VIII-2-6 115 23
15. Plenum Cylinder - Upper End VIII-2-6 113 38 4

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3.

SEISMIC MARGIN METHODOLOGY Dynamic analyes were conducted for the Seismic Margin Earthquake (SME) event using the previously described Loop Model and Reactor Vessel Isolated Model. Results from these analyses were conpared on a load level basis with the design basis faulted conditions loading consisting of SSE +

LOCA.

Throughout this report, the margin against the SME event is denoted as FSME and is defined as the factor by which the SME would have to be increased for the combined normal plus seismic response quantity to reach code allowable level. The FSME provides a concise measure of the sensitivity of the component to the SME loading. Since the critical components of the NSS System were qualified by analysis, the seismic margin is defined as:

SME, 'A'"N (1) p SME where A

Allowable stress from governing code

=

N Stress due to normal operating loads

=

SME =

Stress due to the SME o

For the cases where stress contribution due to normal operating loads is not separated from design dynamic load induced stresses, the margin is reported as being greater than the margin against the design fdOTted condition.

DESIGN FSME >

R (2)

D

(

SME)

VIII-3-1

where:

'D + N + "SSE + LOCA bESIGN = Response associated with design dynamic event RSME = Response associated with SME event 4

This definition of safety margin is conservative for all cases where the SME event dynamic response (i.e.,

SME) is less than or equal to the design faulted condition dynamic response (i.e.'

SSE + LOCA) and where a positive margin exists for the design faulted condition ( A'"D)*

Since the LOCA event tends to be the controlling faulted condition dynamic event for all locations within the primary cooling loop and the reactor internals, it is found that oSME is always less than OSSE +

such that the above definition of seismic margin is always conservative.

The dynamic response may be any characteristic variable (i.e., force, moment, stress).

The normal operating loads for the reactor internals are generally negligible and therefore, the margin against the SME event for the reactor internals becomes:

"A F

i SME "

(3) oSME In some cases, the design faulted condition response level for piping nozzles, snubbers, or supports is not given and is only stated to be less than code allowable values. For these cases, it is conservatively assumed that the design faulted condition response equals the code allowable value and therefore the margin against the SME is calculated to be-greater than the ratio of faulted condition dynamic response v'ariables (i.e.,

resultant moments for pipe nozzles).

1 MNDESIGN F

SME M

(4)

RSME VIII-3-2

1 This is again conservative since load level dynamic response to the SME event is always less than that for the design faulted event.

At 100% power, a gap of 0.121 inches exists between the Reactor Vessel and each of the 12 RV Upper Lateral Supports (ULS).

Based on established criteria given in Reference 5, the gap has been determined to ensure that the RV will only contact the ULS in the event of the design basis LOCA at 100% power. The anchor studs alone are sufficient to resist seismic overturning moments. Therefore, the margin for the RPV Upper Support is calculated as:

0.121 FSME " a (5) 3g where ASME equals the relative displacement between the RPV and the Primary Shield Wall as the result of the SME event based upon time history analysis. It should be noted that this does not in actuality constitute a failure margin since activating the upper lateral supports is a viable means of carrying additional load. Ho-Jer, based upon the above mentioned operating criteria, the quoted margin is a margin against gap closure for seismic accelerations and is stipulated as such in Table VIII-4-1.

1 The determination of the margin for the Reactor Pressure Vessel anchor studs is calculated in a manner similar to Equation (3) and is based upon the evaluation of the Unit I configuration. The anchor stud allowable stresses are taken from Section 3.8.3.4.1 of the Midland FSAR which is based upon the work reported in Reference 5.

The maximum stress in the studs associated with the SME are determined from recent analyses of the Unit I anchor configuration.

VIII-3-3

^

4.

StM4ARY AND CONCLUSIONS Minimum conservatively computed seismic margins against th are presented in Tables VIII-4-1 and VIII-4-2 for all selected locations within the Nuclear Steam Supply System.

In all cases, the margins relative to code allowable response levels are greater than 1.0 a fore the primary cooling loop and the reactor internals are cons be qualified for the SME.

o The lowest margin against the SME for the NSSS piping and equipment supports is greater than 2.34 while th for the reactor internals is 14.4.

Since the design faulted condition dynamic loadings include bot SSE and LOCA, and since LOCA is the predominant dynamic loadin NSS System, it is clear that significant margins against the SME sho exist for all regions where response to the SSE and SME excitations similar.

The component supports for major equipment items tend to be more sensitive to seismic response.

The lower margins are associated with equipment supports whose response to the LOCA loading is dominant.

The margins against the SME are, in all cases, greater than the code margins against the design faulted condition VIII-4-1

1 TABLE VIII-4-1 SUPf1ARY OF SEISMIC MARGINS FOR SELECTED NSSS P Minimum Margin Description FSME 1.

RPV Support Skirt / Base Interface (Vessel Skirt)

>8.10 1

2.

RPY Upper Support (RPV Anchor Studs) i l

31.0 3.

OTSG Support Skirt / Base Mat Interface 3.54*

{

]

(Skirt) l 6.43 j

i 4.

OTSG Upper Support (OTSGAnchorStuds)

>4.65 j

5.

Pressurizer Lug / Support Structure Interface

>4.75 6.

i Pressurizer Upper Support 8.26 7

[

RPV 36" Hot leg Outlet Nozzle

>3.82 8.

RPV 28" Cold leg Inlet Nozzle 9.98 9.

OTSG 36" Hot Leg Inlet Nozzle 5.83 I

10. OTSG 28" Cold Leg Outlet Nozzle 12.99 l
11. RCP 28" Cold Leg Inlet Nozzle 9.87
12. RCP 28" Cold Leg Outlet Nozzle

>4.51 i

13.' CRD Housing /RPV Interface

>6.65 14 RCP Snubbers (PIA 1 Upper Horizontal Support) 8.94

>2.34 Margin Against Gap Closure

TABLE VIII-4-2

SUMMARY

OF SEISMIC MARGINS FOR SELECTED REACTOR VESSEL INTERNALS Description Minimum Margin FSME 1.

Plenum Cover 2.

Upper Grid Assembly - Rib Section 26.2 3.

Upper Grid Pad Joint 25.0 4.

Core Support Shield - Lower End 14.4 5.

Core Support Shield - Upper Flange 37.7 6.

Thermal Shield - Upper End 22.7 7.

Ther.nal Shield / Lower Grid Shell Bolted Joint 107.3 8.

Thermal Shield Upper Restraint Flange 63.1 9.

Core Barrel Assembly - Upper End 67.9 10.

Core Barrel /Former Bolted Joint 31.5 11.

Lower Grid Assembly - Top Rib Section 21.7 12.

Lower Grid Assembly - Top Rib Section/Shell Forging 73.9 Bolted Joint 13.

Lower Grid Assembly - Support Post / Support Forging 101.8 Welded Joint 14.

Control Rod Guide Tubes - Slotted Region 145.5 15.

Plenum Cylinder - Upper End 203.5 80.8 9

VIII-4-3

5.

EVALUATION OF NSSS PIPING AND EQUIPMENT SUPPORTS Although minimum margins against the SME excitation are shown in 1

Section 4 for the NSSS piping and equipment supports, the bash for the evaluation of these critical locations is a load level comparison of the dynamic response for the SME event and the design basis faulted event dynamic response used for licensing purposes. These load level responses are obtained from dynamic analyses of the NSS System employing the previously mentioned half loop and reactor vessel isolated models.

Where the design basis faulted condition dynamic response exceeds that for the SME, no further margin evaluation is in fact required.

Since the LOCA load response is so predominant for the NSS System components, it is clear that the design faulted condition dynamic loading which includes both SSE and LOCA will induce response which is generally well in excess of that generated by the SME faulted event.

Table VIII-5-1 presents a load level response comparison of the forces and moments at the selected locations which result from the SME and the design faulted condition excitations. It can be seen from this table that the loads resulting from the SME are less than those for the design basis faulted i

dynamic loads for all locations.

The Reactor Vessel Upper Supports are designed to limit the vessel displacement during a LOCA. The clearance between the reactor vessel and the upper support bumpers is not taken up during seismic events and therefore the upper support forces are zero for the SME event.

The calculated margins for piping and equipment supports given in Table VIII-4-1 are primarily established based upon a load level comparison of design faulted condition and SME dynamic response. 'There-fore, so long as the design faulted condition stresses can be shown to meet appropriate code criteria, the margins against the SME will be at least as great as those tabulated.

VIII-5-1 l

1

Table VIII-5-1 COMPARISON OF NSSS PIPING AND EQUI FAULTED CONDITION DYNAMIC LOAD LEVE Location Loading F,

RPV Support Skirt /Sese Interface F,

F, n

1.

g n

p 5ME 280.8 Des 19n 163.2 213.8 6329 241.8 8078 402.3 2.

RPV Upper Support 4032 3857 82.497 709.3 10,158 SME 0

0 Design 2442 0

3.

075G Support Skirt /Sese Met 0

1727 Interface SME 1 73.6 259.1 Design 192.7 2624 765.9 2538 649.8 6446 1486 15,676 1492 4

075G Upper Support SME 305.5 6153 Design 1452 311.0 5.

Pressuciter Lug /5upport Structure 3613 SME 25.0 10.3 21.0 Detign 151.9 6.

Pressuriser Upper Support 42.4 154.1 SME 67.5 Design 340.6 77.1 7.

RPV 36" Hot Leg Outlet 294.9 SME 43.6 28.6 Design 245.3 40.3 277.7 317.5

  • 187.3 8.

RPV 28' Cold Leg Inlet 214.5 152.5 1653.7 1948 SME 651.4 47.3 35.9 32.1 231.0 352.4 300.1 Design 138.8 9.

OTSG 36" Not teg Inlet 231.3 365.2 808.9 1877.6 1143.3 SME 59.2 38.6 43.4 119.7 36 0.0 187.8 Design 279.1 10.

075G 28" Cold Leg Outlet 719.6 279.5 1549 1170 1695 SME 77.5 71.7 14.2 150.9 185.8 184.5 Design 707.4 11.

RCP 28' Cold Leg Inlet 673.5 230.8 846.2 928.4 2049 SME 36.7 68.9 9.4 117.1 48.8 1304 Design 258.2 12.

RCP 28' Cold Leg Outlet 619.4 341.6 4458 846.3 3791 1

l SME

~

34.2 26.7 35.0 405.8 280.2 550.3 i

Design 566.8 13 CRD Housing /RPV Interface 140.1 465.5 1706 4183 1936 SME 0.8 0.1 Desfgn 2.3 4.4

14. RCP Snubbers Asial Force 19.6 m0 (P141 Upper Horizontal Support -

$st 356.7 pt 76)

Design 834.3

  • Forces in kips, Moments in ft. kips

==

1 VIII-5-2

.r ww'

l 6.

EVALUATION OF REACTOR VESSEL INTERNALS In a manner similar to that for the NSSS Piping and Equip Supports, the seismic margin earthquake evaluation of the reacto internals involved taking the forces and moments obtained from t r vessel l

dynamic analysis of the reactor internals using the Reactor Vess Isolated Model, converting these into appropriate forces and mo e

the selected internal components and bolted or welded joints on faulted condition dynamic excitations whic

, and, purposes.

Again, LOCA is the predominant loading for the design faulted censing condition and the combined design dynamic induced respon i

well in excess of that generated by the SME event.

Therefore, very large margins against the SME are reported for the reactor vessel inter The design basis faulted dynamic load component forces and moments were developed by B&W using a generally conservative load ratioing technique described in Reference 3.

Information was also available describing the calculation of critical stress parameters f the component forces and moments.

Therefore, the reactor internal j

components stress levels were computed for the SME loadings and l

margins against the SME were calculated.

i 1

Due to the large margins against the SME which exist as a res of the LOCA being included in the definition of the design basis fau condition, it is again of interest to assess the relative magnitude of the reactor vessel internals responses to the SME and SSE events VIII-6-1 presents a comparison of the critical stress parameter re Table

't the selected locations resulting from the SME and the design faulted e

condition dynamic loads.

It can be seen from this table that, similar to thii'NSSS Piping and Equipment Supports, stress responses associa i

the SME are significantly less than stresses resulting from th with basis loads, e design VIII-6-1

'l l

l TABLE VIII-6-1 COMPARISON OF REACTOR VESSEL INTERNAL FAULTED CO 4

location Critical 1

Parameter SME Design 1.

Plenum Cover Bearing Stress (psi) 3044 16,059 2.

Upper Grid Assembly - Rib Section P

3.

Upper Grid Pad Joint t+Pb (psi) 2352 25,505 Bolt Stress (psi) 2731 24,677 4.

Core Support Shield - Lower End 5.

CSS - Upper Flange P,(psi) 690 14,172

5 6.

Thennal Shield - Upper End Limit Load (lbs/ Rad) 1.270 (10 )

2.637(10) 0 6

7.

TS/ Lower Grid Shell Joint P,(psi) 356 10,319

{

P,(psi) 1443 43,970 4

8.

TS Upper Restraint Flange Shear Stress (psi) 281 8,145 j

9.

Core Barrel Assembly - Upper End

10. Core Barre 1/Former Joint Buckling Stress (psi) 659 8,572 P,(psi) 1759 22,505 j

11.

Lower Grid Assembly - Top Rib Section

(

12. LGA - Top Rib Section/Shell Joint t+Pb (psi) 775 33,901 P

P,(psi) 375 20,187 l

13.

LGA - Support Post / Forging Joint P,(psi) 210 17,226 14.

Control Rod Guide Tubes - Slotted Region i

15. Plenum Cylinder - Upper End P,(psi) 193 4,709 Buckling Stress (psi) 254 7,418 i

1

{

REFERENCES 1.

NUREG-0800, Standard Review Plan, Nuclear Regulatory Commission, July,1981.

2.

Final Safety Analysis Report (FSAR), Midland Plant, Units 1 and 2, Consumers Power Company.

3.

" Consumers Seismic Margin Study Report for Midland Units 1 and 2",

Document 51-1140965-00, Babcock and Wilcox, March, 1983.

4.

NUREG-0793, Supplement No. 2, Safety Evaluation Report Related to tne Operation of the Midland Plant, Units 1 and 2, Nuclear Regulatory Commission, October, 1982.

5.

CPCo Report, " Reactor Pressure Vessel Support Modification for Midland Nuclear Power Plant, Midland, Michigan, Report No. 3",

Revision 1, dated December, 1981.

VIII-R-1

-.

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