ML20247H130

Rev 0 to Structural Analysis & Evaluation of Sequoyah Reactor Coolant Pump Support Columns

ML20247H130
Person / Time
Site:	Sequoyah
Issue date:	11/30/1988
From:	Himler J, Ott H, Palm S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
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ML20247D009	List: ML20247D003 ML20247H130
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. STRUCTURAL ANALYSIS AND EVALUATION OF THE SEQUOYAH REACTOR COOLANT PUMP SUPPORT COLUMNS NaOV 301888 Revision 0 September 1988 Author: 7 JN4K Himler, Sr. Engineer Structural Analysis Reviewed:

- H.L. Ott, Sr. Engineer Structural Analysis Approved: a) .

S.A. Palm, Manager Structural Analysis Westinghouse Electric Corporation Power Systems Divis1 9n PO Box 2728 Pittsburgh, Pa. 15230-2728 l

8904040392 890324 PDR ADOCK 05000327  !

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STRUCTURAL ~ ANALYSIS AND EVALUATION OF THE SEQUOYAH' .k REACTOR COOLANT PUMP SUPPORT. COLUMNS Table of. Contents p Section Description Pa

________________________________________________________ge ___

Executive Summary 1 1.0 Introduction 2 2.0 Background i

2.1' Westinghouse Analysis Methodologies 2-2.2 Specific Sequoyah Analyses 3

3. 0, SD-119 Sequoyah RCP Column Analysis- 4-4.0 Revised-Sequoyah RCP Column Analysis 4.1 Revised Column Seismic Loads 4 4.2 Column, Reevaluation 5 4.3' Evaluation Results 6 5.0 Generic Implications 7 6.0 Conclusion 8 7.0 References 9 11 r

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'Listiof Figures' Figure No. Title 4

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1 Sequoyah Nulcear Plant 10 original RCP Column Model 2 Sequoyah Nuclear, Plant / 11 Refined RCP. Column Model List of' Tables.

Table No. Title .Pa

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___________.____________________ge 1 Sequoyah Nuclear Plant 12 RCP Support Column Seismic Loads 2 Sequoyah Nuclear Plant 12 RCP Support Column Seismic Loads Modal Technique 3 Sequoyah Nuclear Plant 13 L

Design Basis Load Combinations 4 Sequoyah Nuclear Plant 13 Design Basis Criteria 5 Sequoyah Nuclear Plant 7 14 l

RCP Column Evaluation Results Design Basis Criteria (ABS of DBE and~LOCA) l l

" 6 Sequoyah Nuclear Plant 14 RCP Column Evaluation Results I Alternste Criteria (SRSS of DBE and LOCA) 111

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TEXECUTIVE

SUMMARY

During the US Nuclear Regulatory Commission's'(NRC) Integrated Design. Inspection (IDI) designLreview of Sequoyah Unit 2,1

.specifically the review associated with the Reactor Coolant Pump (RCP) support column embedments, questions arose withJregard to the support column seismic; loads:provided in. Westinghouse stress report SD-119. The validity of the seismic loads was questioned because .the Operating Basis Earthquake (OBE) Lloads . ware . greater .

' than the Design; Basis L Earthquake - (DBE) loads.

The difference in-the. seismic loads was initially attributed to:

OBE enveloped building: response spectra being . greater 'than the DBE response spectra; and, since the spectra were not of the same shape, this resulted in different RCP moment vector. orientations-which then resulted in different: column loads.-

During further review however, it.was: determined that the RCP support column seismic' loads, provided in SD-119, were developed with appropriate techniques but.did not consider a complete set of support; plane load (SPL). sign. combinations. -Consequently, RCP column seismic loads,have been developed with a complete-set.of sign combinations resulting.in column-load increases.

The new loads were provided to Tennessee Valley Authority (TVA) for. inclusion;in their:embedment analysis and were included in a revised' column analysis performed by Westinghouse. The results of the revised column. analysis show a maximum overstress of 19%

in'one of the three columns based on sequoyah design basis criteria. However, applying criteria consistent with both the Sequoyah Unit 2 evaluation criteria for the RCP column embedments Plan criteria,and with section'3.9.3 of the NRC's Standard Review (i.e., square root sum of the squares (SRSS)

^ combination of seismic and pipe rupture loads) structural integrity of the RCP support columns has been demonstrated.

A detailed review of the design documentation for the remaining Sequoyah Reactor Coolant System'(RCS) equipment supports, the

. Watts Bar RCS equipment supports,.and other Westinghouse plants was performed to determine if incomplete support plane load sign combinations were used. This review indicates that the inccmplete Sequoyah RCP support column sign combination assumptions is an isolated condition and that the remaining Sequoyah RCS equipment supports, the Watts Bar RCS equipment supports, and other Westinghouse plants were evaluated considering all unpropriate sign combinations.

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1.0 INTRODUCTION

As a result ~of the US NRC's Sequoyah Unit 2 IDI, questions.were

" raised relative to the validity: of the RCP support column seismic loads. This report summarizes the design. basis analysis,

' identifies-the reasons for~the discrepancy in the column seismic ~

. loads,.and documents additional analyses performed to. correct the invalid assumption used in.the development of the RCP. support column ~ seismic. loads.

In addition, information is provided that serves as a basis for justification.for the' operation.of both-Sequoyah units.-; This report demonstrates that the structural. integrity of the RCP' support columns is not adversely affected and that this condition is limited to an isolated. assumption by the analyst.

This report also confirms that this is an isolated' incident, applicable other only to Sequoyah, Westinghouse' plant. and has no generic impact on any The calculations which form the basis for this report were performed under the. Westinghouse-Quality Assurance-Program,.

reference-1 and are filed in Westinghouse central file-folder TVA-954/13C.

2.0 BACKGROUND

l To adequately summarize the RCP support column load issue, a-brief description of Westinghouse seismic analysis methodologies and a history of the specific RCS analyses performed by Westinghouse for Sequoyah are provided in~this section.

2.1 WESTINGHOUSE ANALYSIS METHODOLOGIES' Westinghouse RCS seismic analysis of the sequoyah vintage (early 1970's) consists of a one loop, fixed reactor vessel linear.

response spectrum analysis. A model of the RCS consisting of mathematical representations of the primary equipment and the interconnected piping was analyzed using simplified primary equipment' supports system stiffness matricos. Primarily due to computer size and software limitations during this time period, indiv developed independently and used to generate 6x6' stiffness matrices. These stiffness matrices represent the restraints that the support system provide for each component of the RCS.

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RCS analysis model includes the effects of the' primary equipment'supportsLby. including the 6x6 stiffness matrices . .

at the: attachment: point to the primary equipment.. The attachment

point to.the primary equipment is defined as the support: plane.

The RCS model is then analyzed with a response spectrum analysis.

technique-using'anveloped building response spectra..:The results Lof the analysis consists of a set of equipment, nozzle-, piping, and. support plane seismic;1oads for!which the RCS evaluations and qualifications are performed.

The seismicisupport~ plane loads consist of a set of six loads, three forces and three moments, which result from the displacements support plane.

and rotations of the equipment centerline at the These displacements and rotations are multiplied by the plane loads. support structure stiffness natrix to produce support j SPLs are developed for.each individual. mode of the i system. . Total seismic SPLs are determined as the square' root sum of the squares.of'the SPLs corresponding to each of the individual system modes.

Total seismic SPLs, taken-from the RCS analysis, are unsigned and are used to perform independent support evaluations _and qualifications. 'The: seism 3cLSPLs are applied to the same support models used to. develop the stiffness matrices. To. account for~' 1 the. unsigned nature.of the SPLs, the analyst considers all possible sign combinations of the critical loading terms of ,

vertical force and overturning moment. Since the RCP support- (

I columns are pinned at each end,.SPLs which produce axial force-in the column (primarily Fy,-Mx, and Mz) are the critical loading.

terms. . The~considere. tion of.all possible. sign combinations results'in the evaluation of eight sets of support plane leads-each-with the same magnitude but varying-signs. The structural j

analysis member seismic of the models loads. produces eight cases of individual support The maximum individual support member l seismic loads are selected from these eight cases.

The maximum individual support member seismic loads are combined with member loadings from the other p. ant design loading conditions such as dead weight, thermal expansion, system pressure, and Loss of Coolant Accidents (LOCAs). The total member loads for each loading combination are then evaluated and the resultant stressas are compared to the design basis allowable stress limits.

2.2 SPECIFIC SEQUOYAH ANALYSES The original RCS design basis analyses for the. Sequoyah Nuclear Plant were performed by Westinghouse in 1972 and documented in stress report SD-105, reference-2. This effort included the analysis, evaluation and qualification of the RCS primary I

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equipment, piping, an,d primary equipment supports for plant design pipe l rupturebasis loads. loading conditions of operating, earthquake, and i l

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-In 1975 the design basis' criteria of seismic response spectra,: -

reference-3, and pipe rupture criteria (WCAP-8082, reference-4)'

-were changed. As a result, the design basis analyses were .

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. updated.T Results of these-analyses were documented /in stress a report SD-119, reference-5.

3.O SD-119 SEQUOYAH_RCP-COLUMN ANALYSIS 1

The Sequoyah RCP column seismic loads, reported in stress report- 1 SD-119, were' developed using the standard analytical approach described above. However,-a recent review of these loads show that'they were developed with the conside.ation of an incomplete set of SPL sign combinations. Only two of the possible eight sign combinations for.the Fy,HKx, and Mz loading terms were considered. .The two sign combinations used result in an L

approximate maximum axial load for column 5 but not for columns 4 E.6. The sign combinations which produce the~ maximum loads for ]

column 4, 5 & 6.were inadvertently omitted. The effect of the 1 '

support plane 7ead sign combinations on individual column loads is shown in Taols-1.

As shown in Table-1 a complete set.of sign combinations.results  !

in; increased column seismic loads compared to those reported'in~

stress-report SD-119.

1 4.0 REVISED.SEQUOYAH RCP COLUMN ANALYSIS *i Since the design basis stress margins for the faulted loading. .j' condition are small, a more realistic approach.for developing seismic loads was used.

4.1~ REVISED COLUMN SEISMIC LOADS

1 The use of'all eight sign combinations is a simplistic approach

- l in determining column seismic loads since some of the eight sign combinations mayLnot occur during the actual response of the RCP. d A more precise method is'to calculate column loads for each mode of the RCP. _ Individual modes of the RCP have a specific set of signs associated with the six components of loading. Column seismic loads developed on a mode by mode basis are totaled consistent with the Sequoyah FSAR-method of combining modal affects and represent the overall column seismic load.

Table-2 provides the'overall sequoyah RCP support column seismic loads developed with the mode by mode technique applying the 1974, reference-6, enveloped building response. spectra. These are the seismic loads for the Sequoyah RCP support columns.

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iTha column seismic. loads, Table-2,.are the-final design' basis-column seismic loads..'These loads were provided to;TVA, 3 reference-7,:for inclusion in-the RCP' support column concrete embedment analysis,and qualification.

.i Based upon these new seismic loads of Table-2 a' reevaluation of the support columns was performed.

4.2' SUPPORT COLUMN REEVALUATION A refined model of.the RCP support structure was developed.for-the reevaluation of the support columns. This model.provides l- a distribution of loading throughout the column assembly and permits the accurate analysis of each individual element. For comparison, Figure-1 shows the original RCP support column model1 1 and. Figure-2 shows the refined RCP support column model.  :

W The three support columns were modeled with elements representing the column foot bolt,.the top adaptor block plates, the top and bottom clavis plates, the top and bottom shear pins, the top and bottom column tangs, the column pipe section, and the. column base assembly.

SPLs for the normal operating conditions, the revised-seismic cases, the RCP. analysis and the existing pipe rupture cases were used as input to Individual member loads for each element of the model were combined, as shown in Table-3, into .the Sequoyah J design faulted, loads.

basis combinations of normal, upset, emergency, and i

Total combined loads were used.to develop actual element '

Etresses. Actual element stresses were computed based on an

- elastic section for the normal, upset, and emergency loading j cenditions. An elastic section employs section properties, such {

i as the elastic section modulus and the effective shear area. For the faulted condition the full strength of the cross section is l accounted shear area. for by using.the plastic section modulus and the gross Allowable stress limits were established in accordance with the FSAR and promulgated by the design basis criteria. The design basis criteria are the working stress limits of AISC 1969, used 1

' for the normal condition and are multiplied by increase factors for the remaining loading conditions. For some structural

-elements, actual yield and actual ultimate stress, as documented by certified mill test reports (CMTRs), were used in the calculation of the allowable stresses for the faulted condition.

A summary of the design basis criteria is provided in Table-4.

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g-lActudl 01cannt otracecs were compared to'al owable stressL11mits.

Dividing the actual element stress by the corresponding allowable element stress. produces a stress. interaction ratio,-(IR).

(Actual ~ Stress)

IRi = ----------------1 -

i = Element No.

(Allowable Stress)i o

Members qualified. with interaction ratios'less than 1.0 are considered 4.3 EVALUATION RESULTS Table-5 provides a summary of the controlling interaction ratios for each of the three RCP columns.

Using the design basis criteria.results in the bottom column clavis tang plate of Column _4 being overstressed by 19% in tension and bending and the bottom shear pin of Column 6 by 11%

in shear.- However,.by applying more realistic criteria or by performing. additional _ analyses, listed below, structural integrity can be demonstrated.

Criteria / Analysis Options:

1. SRSS method-of combination of DBE and LOCA loads.

2. Utilize Leak-Before-Break (LBB) results to reduce LOCA loads.,

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Support Limit analysis and/or a RCS plastic analysis L accounting for the effect of strain rate on the material.

The preferred option for demonstrating structural integrity is to use the SRSS method of combination of DBE and LOCA loads. This approach was used for the evaluation of the RCP embedments for Unit 2 restart.

NRC, reference-8. That evaluation was reviewed and accepted by the It is consistent with the NRC Standard Review Plan ' (SRP) , section 3.9.3, and has been used extensively on numerous plants as the method for load combinations for mechanical equipment. Tab 39-6 provides the. controlling RCP support column interactic- ratios when the SRSS method of combination is used for the faulted condition. Since the maximum qualified. ratio from all three columns is 0.97, the columns are interaction 4

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The second option involves the reduction in pipe rupture loads.

The Leak Before. Break approach eliminates the large breaks in the RCS-loop piping. Since the faulted RCP-support column loads are

predominantly loop pipe rupture loads, eliminating these loads produces significant reduction in total load well below the capacity of the column. The. Leak-Before-Break analysis for Sequoyah are favorable.

is currently being performed and the preliminary results The other analysis option involves performing.a limit analysis for the support-column and/or a RCS plastic analysis.- This analytical approach has a high probability of success considering

.the level of column overstress (19%).

'5.0' GENERIC IMPLICATIONS A detailed ' review was performed to determine what impact, .if any, the Sequoyah RCP' sign combination assumption has on the remaining-

.Sequoyah RCS equipment supports, the Watts Bar RCS equipment supports, and the RCS equipments supports for other-Westinghouse plants.

The: remaining'Sequoyah RCS equipment' supports which could have been affected by the sign combination assumption are the Steam-Generator'(SG) support. columns. A detailed review of the,SG support' column seismic loads reported in stress report SD-119 was performed.

This review indicates that a complete set of sign combinations was used in the SG column analysis.

The' analysis documentation for the Watts Bar SG and RCP support column evaluation calculations were reviewed. Based on the results of this review it was determined that the column seismic loads were developed using the mode by mode technique discussed previously. This technique does not require the analyst to make assumptions with regard to SPL sign combinations. Therefore the invalid assumption addressed by this report is not applicable to the Watts Bar plant, reference-9.

A review of the remain,ing RCS equipment supports for Westinghouse plants shows that either a~ full set of SPL sign combinations were used or the mode'by mode technique was used.

In summary, the Sequoyah RCP seismic SPL sign combination assumption was not duplicated on the remaining Sequoyah RCS equipment supports, the Watts Bar RCS equipment supports, nor on the_RCS equipment supports for other Westinghouse plants of the Sequoyah vintage.

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6.0 CONCLUSION

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L In conclusion, the' questions with regard to the sequoyah RCP-column seismic loads reported in stress report SD-119, identified an isolated, invalid assumption in the SPL sign combinations. To

' correct this: condition, new RCP. support column seismic loads were developed. .The new seismic loads-were combined with other design basis loadings and were evaluated. The results show a, slight overstress using the conservative sequoyah design basis criteria.

- The use of evaluation criteria, consistent with that used for,the

- evaluation of the:Sequoyah RCP embedments (which permits the SRSS combination of DBE and LOCA), show all RCP support columns-to havs interaction ratios less than 1.0.

- Therefore, the invalid assumption.in the original RCP column seismic loads was isolated to the Sequoyah-RCP columns, and.this.

condition has been corrected.and the RCP columns for both units have been shown to be structurally adequate.

The plan.for the long-term resolution is to complete the ongoing LBB work, and to then apply the results to the RCP embedments and.

the Westinghouse scope RCS equipment supports, thus assuring compliance with Sequoyah design basis criteria and the FSAR. '.

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, ,.7.0: REFERENCES'

1. Westinghouse WCAP-9565, "NTSD/GTSD Quality Assurance Program Manual for Nuclear Basic Components" 2.'" Structural Analysis of Reactor Coolant! Loop / Supports System for TVA (Sequoyah)' Nuclear Power Plant, Units 1 and 2", Report SD-105, K.M. Vashi and W.L. Patrick, August 1972, Westinghouse'

, Electric Corporation h

3. TVA Contract.68C60-91934,- Letter No. 3271, " Reactor Building Interior. Concrete Structure Acceleration - Time History Records and Floor Accelerations N2H-2-12 & 19", October 2, 1973, D.R.

Patterson to M.E. Wright

4. Westinghouse WCAP-8082-P-A, " Pipe Breaks for the LOCA Analysis-i of the Westinghouse Primary Coolant Loop", January 1975

5. " Addendum to Structural Analysis of Reactor Coolant-Loop / Supports System for TVA (Sequoyah) Nuclear Power Plant, Units 1 and 2", Report SD-119, Volume II - Supports System, H.L.

Ott, February 1976, Westinghouse Electric Corporation

-6.

TVA Civil Engineering Branch report CEB-80-23-C-R1, "Sequoyah Nuclear Plant, Reactor Building - Dynamic Earthquake Analysis of the Internal Concrete Structure end Response Spectra for Attached-Equipment"

7. Westinghouse project letter TVA-88-760, August 24, 1988,- .

" Reanalysis of RCP Support Columns", T.A; Lordi to P.G. Trudel

8. "NRC Inspection Report Nos. 50-327/88-13 and.50-328/88-13 Integrated Design Inspection Follow-Up", May 26, 1988 (A02 880607 006)

9. Westinghouse project letter TVA-88-696, June 29, 1988, "Sequoyah and Watts Bar Nuclear Plants RCP Column Seismic Loads",

T.A. Lordi to D.W. Wilson-l Page 9

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Sequoyah Nuclear Plant original RCP Column Model i

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Refined RCP Column Model  !

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. . Table-1

', Sequoyah Nuclear Plant-RCP: Support Column '

Seismic Loads Seismic Loads [1]

2 Sign Comb.[2] 8 Sign' Comb.

Column No.[3] Load- ~ OBE DBE OBE DBE L

Axial o6'7 . 3 10.7 613.4 493.0 4 Shear 0.0' O.0 0.0 0.0 Moment 6i4.1 80.2 243.7 211.8 Axial ~509.8 391.1 509.8 391.1 5 Shear' 2.0 1.9 2.0 'l.9-

' Moment 226.4 274.7 732.8 633.3 Axial 290.5 282.0 378.2 350.9-6 Shear 2.0 1.9. 2.0 1.9-Moment 62.0 11.7 354.3

_____ ____. .___________________._______________ __________283.2 ___ _

Notes: [1] Units, kips and inch-kips.

[2] Original SD-119, reference-4

[3]'See Figures 1 and 2

, , , Table-2 Sequoyah Nuclear Plant RCP Support. Column Seismic loads Modal Technique [1]

Column No. Load Column Loads [2]

OBE DBE Axial 430. 449.

4 Shear 4.12 4.34 Moment. 533. 560.

Axial 341. 341.

5 Shear 5.02 S.26 Moment 684. 716.

Axial 402. 421.

Shear 6 4.01 4.19 Moment 535. 559.

Notes: [1] See reference-6

[2] Units are kips and inch-kips I

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' . Table-3 L jc Sequoyah Nuclear Plant i

1 Design Basis'Icad Combinations -)

--_-_------_ . ..-- .....----- i Loading Condition Combination

__..--_ . --__-___'_ .....------._____---__----[.1] ..--

Normal DW+THM+PR Upset Normal + OBE LEmergency Normal + DBE Faulted DW+PR + DBE + LOCA Notes: [1] DW = Dead Weight-THM =' System Thermal Expansion PR = System Pressure OBE = OBE Seismic-DBE = DBE Seismic LOCA = Loss of Coolant Accident Table-4 Sequoyah Nuclear Plant Design Basis Criteria Londing Condition Stress Limit Normal Working Stress Limits AISC 1969 Upset 1.3 x Normal Stress Limits Emergency 1.5 x Normal Stress Limits Faulted IF x Normal Stress Limits [1][2]

Notes: [1] IF = Increase Factor, where IF = Lesser of 1.2(Fy)/Ft or 0.7(Fu)/Ft Membrane stresses are limited 0.7Fu and

[2] Per Shear stresses Sequoyah are Table FSAR, limited to 0.42Fu.[*]

5.2-5, support members must be within yield stress after load redistribution.

[*] Except for the column shear pins, for which the t

faulted yield, allowable shear stress is limited to shear (Fvy=0.5BFy).  !

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Table-5 Sequoyah Nuclear Plant RCP Column Evaluation Results g Design Basis Criteria (ABS of DBE and LOCA) 3 t

controlling-Interaction Ratios Loading Column No.

Condition 4 5 6 Criteria Summ

______________.._______________________.._________..___ary ___ -

Normal 0.55 0.19 0.35 -AISC-69 Working Stress Limits l Upset 0.83 0.71 0.88 -1.3 x Working Stress Limits Emergency 0.76 0.68 0.84 -1.5 x Working Stress Limits Faulted 1.19 0.77 1.11 -IF x Working Stress Limits but-less than 0,7Fu and 0.42Fu, based on the use of CMTRs.[*]

Table-6 Sequoyah Nuclear Plant RCP Column Evaluation Results l Alternate Crlteria (SRSS of DBE and LOCA)

Controlling Interaction Ratios Loading . Column No.

Condition 4 5 6 Criteria Summar

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Normal 0.55 0.19 0.35 -AISC-69 Working Stress Limits Upset 0.83 0.71 0.88 -1.3 x Working Stress Limits Emergency 0.76 0.68 0.84 -1.5 x Working Stress Limits Faulted 0.97 0.60 0.90 -IF x Working l Stress Limits but less than 0.7Fu and 0.42Fu, based on the use I[*] See footnote, Page 13 Page 14 l

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