Text

Rev 0 to MPL-09-100, Grand Gulf In-Plant Safety Relief Valve Test Fatigue Evaluation Final Rept
	ML20235X624
Person / Time
Site:	Grand Gulf
Issue date:	01/31/1986
From:	Mcconaghy W, Mcinnes I, Taylor M; MISSISSIPPI POWER & LIGHT CO., NUTECH ENGINEERS, INC.
To:	;
Shared Package
ML20235X520	List: AECM-87-0127, Forwards Nutech Rept MPL-09-100, Grand Gulf In-Plant Safety Relief Valve Test Fatigue Evaluation Final Rept, in Response to NRC Request for Addl Info Re Util on Safety Relief Valve Fatigue Evaluation; L-09-100, Rev 0 to MPL-09-100, Grand Gulf In-Plant Safety Relief Valve Test Fatigue Evaluation Final Rept;
References
	MPL-09-100, MPL-09-100-R-00, MPL-9-100, MPL-9-100-R, NUDOCS 8707240263
	Download: ML20235X624 (84)
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4 li4 s- s: r s.m j GRAND GULF IN-PLANT f SAFETY RELIEF VALVE TEST i

FATIGUE EVALUATION n

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Prepared for Mississippi Power & Light Company Prepared by NUTECH Engineers, Inc.

San Jose, California Prepared by: Approved by:

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.A I. D. McInns , v.'E. M. Tayl ot', P.E. 1 Project Engineer Project Manager

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. McConagny' R . 'H . Smith W. P .

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8707240263 870717 3 PDR ADOCK 05000416 P PDR Q&.Q-L__ _ d

s REVISION CONTROL SHEET TITLE: Grand Gulf In-Plant DOCdMENT FILE NUMBER: MPL-09-100 Safety Relief Valve Test Fatigue Evaluation f Final Report 1

Ian D. McInnes, P.E., Project Engineer II*M N AME / Tl?LE INITIALS i

Nathaniel G. Cofie, Princioal Consultant Nb NGC NAME/ TITLE INITIALS Mo.ses Taylor, Jr.,.P.E., Supervising Engineer /h[

NAME / TITLE INITIALS NAME / TITLE INITIALS AFFECTED DOC PREPARED ACOUAACY . CRITERIA REMARKS PAGEIS) REV BY / DATE CHECK SY / DATE CHfCK Sif / DATE i-ix 0 \)M th!6b /)'$//n/aI',

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1 ABSTRACT A Safety Relief Valve (SRV) test program was conducted at the Grand Gulf Nuclear Power Station, Unit 1 on April 23 through 25, 1985. The purpose of the Grand Gulf test program was to confirm that the SRV hydrodynamic loads on the Grand G.lf plant and their effect on structure and equipment ere: (1) less than design, and (2) consistent with the loads and effects predicted by accepted analytical techniques.

The secondary purpose of the Grand Gulf SRV test was to collect strain data on two typical fatigue sensitive components to demon-strate that the induced strains during SRV actuations are small, i

and can be ignored from design considerations. This report des-cribes the instrumentation, addresses the measured strains and demonstrates that a fatigue program to meet Section 2.C(10),

Part A of the Grand Gulf Nuclear Station Unit 1 Facility Oper-ating License NPF29 (Reference 1) is not required.

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, TABLE OF CONTENTS Pace ABSTRACT iii LIST OF TABLES y LIST OF FIGURES r vi 1.0 INTRODUCT."ON 1.1 1.1 Test Program 1.1 l ' .2 Test Objectives 1.2 l 1.3 Summary of Test Results 1.3 2.0 TEST SEQUENCE AND EVENTS 2.1 3.0 INSTRUMENTATION

SUMMARY

3.1

/r 3.1 Accuracy of Instrumentation 3.3 3.2 Fa!, led Sensors 3.3 4.0 . DATA REDUCTION 4.1 l

5.0 DISCUSSION OF RESULTS 5.1 5.1 Measured Strain Data 5.1 5.2 Measured Acceleration Data 5.2 6.0 EXTRAPOLATION FROti TEST TO DESIGN CONDITIONS 6.1 C.1 Effect of Power Leeve.'. 6.1 6.2 Number of Valves Actuated 6.2 6.3 Number of Significant Stress Cycles 6.4 l 6.4 Expected ShV Fatigue Life 6.5 7.0 EXTRAPOLATION TO ALL EQUIPMENT 7.1

8.0 CONCLUSION

S 8.1

9.0 REFERENCES

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LIST OF TABLES a

Numcer Title Page I I

l 2.1 Test Matrix 2.3 3.1 Location of Accelerometers - Structure 3.4 3.2 Location of Accelerometers - Equipment 3.5 3.3 Location of Accelerometers - Piping 3.6 5.1 Fatigue Evaluation Peak Strain 5.5 l Results )

l 5.2 Fatigue Evaluation HCU Module Peak l Stresses 5.6 3.3 Peak Measured Accelerations 5.7 i

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LIST OF FIGURES Figure Title Page 3.1 Hydraulic Control Unit Module 3.7 General Arrangement 3.2 HCU Fatigue Evaluation Strain Gauge 3.8 Locations 3.3 Hydrogen Recombiner Fatigue 3.10 Evaluation Strain Gauge Locations l 3.4 Plan /iew Showing Test Quencher vs. 3.11 Equipment Locations 3.5 Accelerometer Locations - Structure 3.12 (Elevation View - Accelerometers Rotated into View) i 3.6 Pipe Mounted Accelerometers A38, 3.13 A39, A40

, 3.7 Pipe Mounted Accelerometers A41, 3.14 l A42, A43 3.8 Pipe Mounted Accelerometers A44, 3.15 A45, A46 3.9 Pipe Mounted Accelerometers A47, 3.16 A48, A49 3.10 Pipe Mounted Accelerometers A50, 3.17 l A51, A52 5.1 Typical SVA Strain Time History 5.9 5.2 Typical SVA Strain Time History 5.10 l 5.3 Typical SVA Stress Time History 5.11 5.4 Typical SVA Stress Time History 5.12 5.5 Typical CVA Strain Time History 5.13 5.6 Typical CVA Strain Time History 5.14 l

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LIST OF FIGURES (Continued)

Figure Title Page 5.7 Typical CVA Stress Time History 5.15 5.8 Typical CVA Stress Time History 5.16 i

5.9 Typical MVA Strain Time History 5.17 5.10 Typical MVA Strain Time History 5.18 5.11 Typical MVA Stress Time History 5.19 5.12 Typical MVA Stress Time History 5.20 5.13 Envelope Respo se Spectra 5.21 Accelerometer A32 (Radial)

I j 5.14 Envelope Response Spectra 5.22 Accelerometer A33 (Vertical) 5.15 Envelope Response Spectra 5.23 Accelerometer A34 (Tangential) 6.1 Design Fatigue Curves for Carbon, 6.7 Low Alloy and.High Tensile Steels 6.2 Design Fatigue Curve for Austenitic 6.8 l Steels l

l 7.1 Envelope Response Spectra 7.4 Accelerometer Al, Containment Base Mat Elevacion 93'-0", Horizontal 7.2 Envelope Response Spectra 7.5 Accelerometer A3, A5, A6, Containment Elev. 109'-l 1/2", Horizontal l

7.3 Envelope Response Spectra 7.6 I Accelerometer A4, Containment Elev. 109'-l 1/2", Vertical ;

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LIST OF FIGURES (Continued) l l

Figura_ Title Page ;

j 7.4 Envelope Response Spectra 7.7 j Accelerometer A7, Containment !

Elev. 147'-7", Horizontal j i

7.5 Envelope Response Spectra 7.8 !

Accelerometer A8, Containment l Elev. 147'-7", Vertical 7.6 Envelope Response Spectra 7.9 l Accelerometer A9, A10, Containment !

Elev. 147'-7", Horizontal 7.7 Envelope Response Spectra 7.10 Accelerometer A13, Containment Dome Elev. 302'-3", Vertical l 7.8 Envelope Response Spectra 7.11 l Accelerometer A14, Containment Dome '

Elev. 302'-3", Horizontal 7.9 Envelope Response Spectra 7.12 Accelerometer A15, A17, A18, Drywell Elev. 120'-10", Horizontal 7.10 Envelope Response Spectra 7.13 Accelerometer A16, Drywell Elev. 120'-10", Vertical i 7.11 Envelope Response Spectra 7.14 l Accelerometer A19, Drywell Clev. 147'-6", Horizontal 7.12 Envelope Response Spectra 7.15 Accelerometer h20, Drywell Eler. 147'-7", Vertical 7.13 Envelope Response Spectra 7.16 Accelerometer A21, Drywell Elev. 184'-6", Horizontal l

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t LIST OF FIGURES (Concluded)

Figure Title Page 7.14 Envelope Response Spectra 7.17 Accelerometer A22, Drywell Elev. 184'-6", Tangential (Horizontal) 7.15 Envelope Response Spectra 7.18 Accelerometer A25, A27, A28, RPV Pedestal Elev. 100'-9", Horizontal 7.16 Envelope Response Spectra 7.19 Accelerometer A26, RPV Pedestal Elev. 100'-9", Vertical l

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1.0 INTRODUCTION

1.1 Test Program A major portion of the planned Safety / Relief Valve (SRV)

Discharge test for the Grand Gulf Nuclear Station was conducted April 23 through April 25, 1985. This program was formulated to meet Mississippi Power and Light Company's (MP&L's) commitments as outlined in Reference 2 and modified by Reference 3. This test program provided measurements of the suppression pool hydro-dynamic loads, submerged structure strains and response of the reactor building structures and equipment. In addition, strain data was collected for key fatigue sensitive plant components to meet the requirements of Section 2.C(10), Part A of the Grand Gulf Operating License (Reference 1). This strain data has been used to perform a fatigue evaluation of safety related equipment when subjected to building response induced by SRV discharge loads.

The SRV test instrumentation and actual test matrix are fully described in the SRV Test Final Report, Reference 4 and the test procedure provided in References 5 and

6. The strain gauges added to the SRV test program for j the fatigue evaluation are described in Section 3.0 of this report.

I For the tests completed in the period April 23 to April 25, 1985, data was collected for one shakedown test, I

three-initial actuations of a single valve (SVA) fol- 4 I

lowed by a consecutive valve actuation (CVA) of the same

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1 valve with an elevated pipe temperature, and one four-valve multiple valve actuation (MVA). The tect program was suspended at this point because all test valves were leaking and additional testing was not practical until ,

cold pipe discharge lines could be established.

1.2 Test Obiectives The primary objective of the test program was to verif y the design adequacy of the plant piping, structures and equipment for the hydrodynamic loads imposed during SRV actuation.

The secondary objective of the test was to collect data necessary to evaluate equipment / component fatigue I effects due to stresses induced during actuation of the plant S RV s . A program to address hydrodynamic load fatigue effects is required to satisfy Section 2.C(10),

Part A of the Grand Gulf Nuclear Station Unit 1, Facility Operating License NPF-29 (Reference 1).

i Section 2.C(10), Part A, of Reference 1, states that j there are three requirements that must be met before the fatigue evaluation issue is formally closed. These are:

1. Document the occurrence of every safety / relief valve discharge into the suppression pool.

2. Determine the cumulative damage factor for selected pieces of equipment, due to SRV actuation, and MP L- 0 9-10 0 1.2 Revision 0 rititeac:ti ENG4NLERS

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3. Report to the NRC if malfunction of equipment i should occur or be likely to occur because of safety /rejief valve discharge.

I 1.3 Summary of Test Results l The test objectives, described in Section 1.2, were j satisfied by measuring strain data on two critical l pieces of equipment, and accelerometer data en both the structure and equipment inside the reactor building.

Measured test results were conservatively factored to j l

design conditions and extrapolated tc show that these

! results are representative for all ecuipment subject to SRV induced dynamic loads.

This report provides a detailed discussion of the measured SRV equipment strain data and demonstrates that SRV discharge load fatigue effects do not limit the life of individual pieces of equipment. A complete discus-sion of the test results for the primary test objectives is contained in the final report (Reference 4).

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2.0 TEST SEQUENCE AND EVENTS 1 l

i The test program consisted of one shakedown test, three SVA/CVA tests and one MVA test as shown by Table 2.1.

Single actuation tests were conducted on the main test valve (V-12), as well as both backup valves (V-10 and V-11), with the MVA test performed on the 4 planned valves (V-2, V-7, V-12 and V-17). As noted in Table 2.1, valve V-12 was leiking prior to the MVA test, so data l collected by the V-12 line and adjacent pressure sensors can be considered as indicative of typical results for a l leaking valve. ;

I The data collected, on the data acquisition system (DAS), for the shakedown test (SDl) was lost due to a malfunction in the tape recorder. However, the real time oscillograph data recorded demonstrated that the l instrumentation was functioning correctly and the signal l

conditioning was set at the correct gain settings.

Therefore, it was concluded that the test had served its function and testing could proceed to MT10/MT11. l On completion of test MT70, all principal test valves i demonstrated some degree of leakage, and the program was i suspended. Following a detai. led review of the data, it was determined that the SRV hydrodynamic loads were bounded by expected pressures, and the measured strains, and building and equipment responses were small compared to expected values. Results of the initial data reduc-tion were submitted to the NRC, References 7, 8, and 9, l MPL-09-100 2.1 Revision 0 l

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i for their review and concurrence in terminating the tests. The NRC has responded, Reference 10, that, sub-ject to review of the final test report ( Re f e rence 4 ) ,

sufficient test data has been collected to satisfy Grand Gulf FSAR and licensing commitments and no further j testing is required.

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1 3.0 INSTRUMENTATION.

SUMMARY

Pressure transducers, strain gauges and accelerometers were installed throughout the Reactor and Auxiliary Buildings to. provide measurements of' suppression pool hydrodynamic loads, submerged structure strains, and; building and equipment response to the-SRV discharge loads. The . locations for the. primary instrumentation,-

reasons for selection, and inst'rument sensitivities are' presented in Reference 4.

In addition to the primary. SRV test instrumentation, twenty strain gauges were installed on the: Hydrogen. ]

Recombiner, at elevation.208'-10",'and. Hydraulic Control j Unit (HCU) number 3257 at elevation 135'-4", to measure induced strains during an SRV actuation.

As shown in Figures 3.1 and 3.2,-sixteen strain gauges were installed on HCU module' number 3257 at azimuth 64*

and elevation 135'-4".. Strain gauges were located as follows:

- SG39 to SG4 2 on pipe f rom manif old block to scram inlet valve

- SG4 3 to SG4 5 on vertical leg of HCU module support frame ,

- SG4 6, SG53 .and SG54 on hexagonal nut at top of' nitrogen bottle g

- SG4 7 to SG50 on pipe from inlet rcram va'lve to j accumulator

- SG51 and SG52 on pipe from nitrogen bottle to-accumulator instrument block l l

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1 These locations were selected after discussions with the i HCU module manufacturer, The General Electric Co., as j being representative of the most highly stressed points. These locations represented the weak points found during seismic qualification of the HCU modules.

As shown in Figures 3.3 and 3.4 four strain gauges were also installed on the Westinghouse Electric Co. Model B Hydrogen Recombiner located at elevation 208'-10" and l azimuth 130 . The four strain gauges are located on the cabinet external cover plates adjacent to the attachment l welds which transfer the vertical load to the lower frame and anchor bolts of the unit. These locations were chosen as they represent theoretically high stress points. Two sets of triaxial accelerometers, one set at l

the base and one set at the top, were also mounted on the cabinet to measure the acceleration response time history to an SRV actuation. The locations of these accelerometers are shown on Figures 3.3 through 3.5.

The foil strain gauges used for the fatigue evaluation are model no. CEA-09-250UW-350 manufactured by Micromea-surements. The signals from the strain gauges were processed through the appropriate signal conditioning equipment, filtered at 200 Hz a.7d atored in digital for-mat on magnetic tape. Details of the data acquisition system are given in Reference 4.

Table 3.1 tarough 3.3 and Figures 3.5 through 3.10 provide the locations for the 56 accelerometers installed for the SRV test program. Reference 4 provides additional information on the accelerometers.

MPL-09-100 3.2 Revision 0 nut _ech.

3.1 ,

Accuracy of Instrumentation i

A detailed diccussion of instrumentation accuracy and the accuracy of the measured data is provided in l

Reference 4. The SRV fatigue eve.luation strain gauges, l SG35 to SG54, have an overall maximum error of 5% of ,

the highest strain (20 u in/in) measured during the test

-J program.

3.2 Failed Sensors i

There were no failed sensors in the strain gauges used j for the fatigue evaluation program.

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Table 3.1 LOCATION OF ACCELEROMETERS - STRUCTURE.

Sensor Location.

Radius Environ-(2) ment kirecIII I.D. Azimuth Elev. lon Reactor Buildina_ Sensors:

Al 32' 93.0' 65'0" E2 R A2 32* 93.0' 65'0" E2 V A3 32* 109.12' 65'0" E2 R A4 32* 109.12' 65'0" E2 V A5 302* 109.12' 65'0" E2 'R A6 302* 109.12' 65'0" E2 T A7 0* 147.58' 62'0" E2 R A8 0* 147.58' 62'0" E2 V A9 270* 147.58' 62'0" E2 R A10 270* 147.58' 62'0" E2 T All- 32* 237.0" 62'0" E2 JR A12- 32* 237.0" 62'0" E2 Vs A13 32' 302.25' 0'0" E2 V A14 32* 302.25' 0'0" E2 R A15 32* 120.83' 41'5" E2 R l A16 32* 120.83' 41'6" E2 V A17- 302* 120.83' 41'6" E2 R A18 302* 120.83* 41'6" E2 T A19 0* 147.58' 41'6" E2 R' A20 0* 147.58' 41'6" E2 V A21 302* 184.5' 41'0" E2 R A22 302* 184. 5' 41'0" .E2 T A23 32* 208.83' 41'6" E2 R A24 32* 208.83' 41'6" E2 V A25 0* 100.75' 10'7"- E2 R A26 0' 100.75' 10'7" E2~ V A27 270* 100.75' 10'7" E2 R -

A28 270* 100.75' 10'7" E2 T s

Auxiliarv Builaina Sensors:

A53 32* 93.0' 68'0" E2 R-AS4 32' 93.0' 68'0" E2 V A55 32* 184.5' 66'0" E2 V A56 32* 184.5' 66'0" E2 R. ,

NOTES: (1) R = Horizontal, radial V = Vertical T = Horizontal, tangential (2) E2 = See Table 3.2 MPL-09-100 3.4 Revision 0 nutgch !

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Table 3.2 {

LOCATION OF ACCELEROMETERS - EQUIPMENT f

Sensor Equipment Incation Direction Canments ID Description Azimuth Elevation l

Polar R (0*-180*) Crane parked on A29 A30 Crane 0* 237'0" V N-S (122*-302*)

A31 Girder T (90-270') azimuth during tests A32 Base of R-A33 Hydrogen 130* 208'10" V A34 Recambiner T A35 Top of R A36 Hydrogen 130* 208'10" V A37 'Pacombiner T NOTE: Environmental conditions are all E2 Pressure 42.2 psia Relative Humidity 100%

Temperature 135'F I

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INLET SCRAM VALVE I

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'"CU FATIGUE EVALUATION ST. RAIN GAUGE LOCATIONS (CONCLUDED)

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1 A35, 36, 37 i

SG 36 SG 36 SG 37 (B ACK) A32, 33; 34 SG 38 (BACXI

-l SUPPORT SASE ELEV. 208' 10" .

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Figure 3.3 HYDROGEN RECOMBINER FATIGUE EVALUATION STRAIN GAUGE LOCATIONS MPL-09-100 3.10 Revision 0 l

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64 Azimuth 1200 Azimuth

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, '/. - -

SG 38 A32 to A34 160 Azimuth h ,. A35 to A37 ' l SG 36

.v.11 .

k 00 Admuth + -

v 10 v2 2000 Azimuth

\ - y,7 3120 Azimuth 3 l

l-1 l

l l

Figure 3.4 PLAN VIEW SHOWING TEST QUENCHER VS. EQUIPMENT LOCATIONS MPL-09-100 3.11: Furt.as.o142 Revision 0 ;

i l 1 nutg,gh :

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- p 3 Figure 3.5 ACCELEROMETER LOCATIONS - STRUCTURE (ELEVATION VIEW - ACCELEROMETERS ROTATED INTO VIEW)

MPL-09-100 Revision 0 nutpsh

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Figure 3.6 PIPE MOUNTED ACCELEROMETERS'A38, A39, A40 MPL-09-100 3.13 Revision 0 nutgqh

-['* u .

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A 41, A41 i A45 E L, . I 4 9 ' - S '/2.*

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. PIPE MOUNTED ACCELEROMETERS A41, A42, A43 MPL-09-100 3.14 Revision 0 nutgsj]

L____.____.___._________.___._.___._ _ . . _ . . _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ . _ _ _ _ _ _

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MPL-09-100 ,

3.15 Revision 0 nutggh

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1 MPL-09-100 3.1G i Revision 0 I

nutp_qh 4 :

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> MPL-09-1;D0 3.17 l' Revision 0 .s l-l

~

, nutggh

l

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4.0 DATA REDUCTION Section 5.0 of the final SRV test report, Reference 4, l t

provides an overall description of the data reduction j program. The data from the twenty fatigue evaluation j strain gauges was reduced simultaneously with data from the primary SRV test instrumentation using the software package j REDUCE, Version 1.2.0. The data reduction was performed on a CYBER-730 system.

l l

i 1

l I

I MPL-09-100 4.1 Revision 0 nutech ENGtPd&ERS

5.0 DISCUSSION OF RESULTS This section provides the measured strain results, the maximum stresses calculated using the strain time histories and measured acceleration data.

5.1 Measured Strain Data The peak measured straia data for each sensor is tabulated, by test, in Table 5.1. Review of this table shows that the peak strain measured for any test is 20 uin/in for test MT70.

The measured strain time histories were converted to stress time histories using the FPOST subroutine from the REDUCE program. FPOST calculates the axial, bending and combined stress time histories, and peak stresses, for each group of orthogonal gauges. Table 5.2 presents the peak calculated stress for each group of gauges located on the HCU module. Strain gauges SG46, 53 and 56 mounted on the nitrogen bottle connecting nut were installed at 120' ]

intervals and do not meet the input requirements of FPOST. Therefore, the stresses reported in Table 5.2 for this group of gauges were conservatively derived from the i peak measured stresses using hand calculations.

1 Figures 5.1 to 5.12 provide typical strain and stress time histories for the SVA, CVA and MVA tests.

As shown in Table 5.1, the four strain gauges on the Hydrogen Recombiner (SG35 to 38) recorded very small l

l MPL-09-100 5.1 l Revision 0 nut.e_c..h

_ _ _ _ _ _ _ _ w

1 l

magnitudes of strain (8 pin /in peak) which are almost indistinguishable f rom the background noise. The cal- !

i culated peak stress for the Hydrogen Recombiner is 0.23 I ksi.

r i 5.2 Measured Acceleration Data Table 5.3 provides a tabulation of the peak measured I

acceleration for each accelerometer, for each test. In addition, the table provides average SVA and CVA l l

accelerations, and design and expected values. The structural acceleration design values were taken from the i Grand Gulf design response spectra, and are equal to the zero period acceleration (zpa) for the single valve consecutive actuation (SRVone) case. The design )

accelerations for valve actuators and the hydrogen recombiner are taken directly from the appropriate analyses. The expected values are 80% of the design values. This ratio was selected based on the ratio of peak predicted pool pressure from test to design conditions.

The 4 valve SRV discharge is not a plant design case and therefore no calculated responses are available. The 4 valve acceleration acceptance criteria were developed using ;

i

, existing analytical data. There are two limiting cases 1 ,

considered. The calculated CVA design and predicted i responses were used for all horizontal accelerations. The ADS (8 valve) design spectra vertical accelerations ratioed by the theoretical decrease in total vertical force from the ADS to four valve case were used for all vertical accelerations.

A review of Table 5.3 shows that the majority of peak measured structural accelerations are considerably less MPL-09-100 5.2 Revision 0 nut.ec.h.

_____a

i i

than 50% of the predicted value. The measured equipment responses (A29 to A52) are generally an order of magnitude less than the predicted values and show that the high f re-quency content of the SRV time histories are greatly attenuated by the attached piping systems and floors. l

)

l l

Based on the very low levels of acceleration measured {

during SDl, many of the accelerometers were set to maximum i sensitivity to try and read the very small induced vibra-tions. The result of this was that the measured signal is i

in many cases equal to, or less than, the background i i

noise. In general, as can be seen from Table 5.3, the !

I measured acceleration is only a small fraction of the l expected value. {

I Some measured acceleration time histories contained I

(

anomalies such as D.C. offset, wild points and an offset j due to charge amplifier saturation. The data reduction program, REDUCE, was used to remove such anomalies from the measured acceleration time histories for A4, A25, and A26, for all SVA tests, and from A7 and A8, for all tests.

l l

l Other anomalies in the measured acceleration time histories were:

1 1

o A2 is a vertical accelerometer located on the con-l tainment base mat at elevation 93'-0" and azimuth 312*

at radius 65'-0", outside the suppression pool. This I

accelerometer recorded apparent maximum vertical accelerations an order of magnitude higher than those measured by all other vertical accelerometers on the containment, drywell, RPV pedestal and equipraent.

Therefore, it is concluded that this accelerometer, or the signal conditioning, is malfunctioning and A2 1

MPL-09-100 5.3 l Revision 0 nut _ec._h.

results have been omitted from Table 5.3 and the calculated spectra. l 0 Accelerometers All and A12 were mounted at mid-span on I

(

the bottom flange of the polar crane rail girder at elevation 237'-0". Accelerometer All measured radial response and A12 the vertical response. The radial response for All is approximately an order of magnitude higher than the containment response at elevation 145'-7" (A7, A9, A10) and at the top of the dome (elevation 302'-3", A14). This is a local response and has no effect on the containment or polar crane design. This is demonstrated by the very low !

accelerations measured at mid-span of the polar crane l

l girder (A29 to A31) where the peak measured accelera-tion for all tests was 0.02g.

1 o A48 results are also omitted from Table 5.3 as the accelerometer was classified as a " failed sensor" for all tests.

Accelerometers A32 to A34 were mounted at the base of the j Hydrogen Recombiner. Figures 5.13 to 5.15 provide envelope response spectra for the SVA, CVA and MVA generated from the measured acceleration time histories. l l

l l

l 1

1 MPL-09-100 5.4 Revision 0 nut _ec_h.

Table 5.1 FATIGUE EVALUATION - PEAK STRAIN RESULTS (u in/in) 1

~SVA Tests III MVA Test CVA Tests Sensor MT10 MT20 MT30 MT70 MTll MT21 MT31 l

S35 3 6 3 4 3 5 4 S36 3 3 5 8 3 4 7 S37 6 6 5 6 4 5 5 S38 1 3 2 7 2 3 5-S39 3 4 4 7. 3 3 4 S40 3 5 5 10 2 3 4 S41 5 7 5 8 4 3 4 S42 4 6 6 8 4 6 4

'S43 2 4 3 4 2 3 3 S44 1 2 2 7 2 2 3 S45 2 3 3 4 .1 3 4 S46 9 14 10 20 8 7 7 S47 3 5 5 8 3 3 6 S48 4 6 6 9 3 4 4 S49 4 3 4 9 3 3 3 S50 4 7 5 8 3 -5 4 551 5 7 8 12 6 3 6 SS2 4 6 4 7. 3 4 5 S53 3 4 5 10. 5 2 6 S54 12 14 11 16 12 5 7

)'

NOTE:

l. SD1 results are not available due to DAS malfunction.

MPL-09-100 5.5 Revision 0 ritstgic:li- l l

L_ _ ______

l Table 5.2 FATIGUE EVALUATION - HCU MODULE PEAK STRESSES (KSI)

Group of(1) SVA Tests -l MVA Test CVA Tests Sensors MT10 MT20 MT30 l MT70 MTil MT21 MT31 l SG 39 to 42 0.16 0.23 0.18 0.29 0.15 0.17 0.14 l t

SG 43 to 45 0.07 0.13 0.10 0.23 0.07 0.10 0.12 SG 47 to 50 0.14 0.23 0.19 0.27 0.12 0.13 0.17 .

SG 51 & 52 0.15 0.22 0.24 0.36 0.18 0.11 0.18-1 SG46, 53,(2) 0.34 0.40 0.31 0.56 0.43 0.20 0.20

. 54 l

MOTES:

l

1. For location of strain gauges see Figures 3.1 and 3.2.

1

2. Strain gauge group located on hexagonal nut at top of nitrogen accumulator (SG4 6, 53 and 54) not included in REDUCE program.

Peak values given are derived from uniaxial strains given in Table 5.1.

1 4

MPL-09-100 5.6 )

Revision 0 nutgch

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MPL-09-100 5.8 Revision 0 nut.ech 1

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i MPL-09-100 5.10 Revision 0 . i l

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l MPL-09-100 5.11 j Revision 0 .j nutggh !

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00'09'I 00*0'8 00'O .00'09'- 0 0 ' 0 9 I' C (ISd3 SS2815 02NISWOJ Figure 5.4 TYPICAL SVA STRESS TIME HISTORY MPL-09-100 5.12 Revision 0 nut.e&.h

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- O 00'ON 0 0 ' 0 9' t 00'0'9 00'0 00'09- 00 ' 0 91' C (ISd) SS2'd 15 02NIEWOJ Figure 5.7 TYPICAL CVA STRESS TIME HISTORY MPL-09-100 5.15 Revision 0 nut.e:_ch

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MPL-09-100 i Revision 0 5.17 l nutp_qj]

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MPL-09-100 5.21 Revision 0 nutech ENGaNEESIS

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I (VERTICAL) I i

MPL-09-100 5.22 4 Revision 0 l t

nutech ENOWEERS

D Pj Damping = .010 av t 1

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MPL-09-100 -

Revision 0 a.23 nutggh

6.0 EXTRAPOLATION FROM TEST TO DESIGN CONDITIONS Strain and acceleration time histories were recorded during each of the seven SRV matrix tests for single valve, consecutive valve and multiple valve actuations. This data has been reduced to develop peak strains, peak (zero period) accelerations ( pa's), strain time histories, and the maximum stresses associated with the measured strains, and envelope acceleration response spectra. The purpose of this section is to describe the methodology used to compare the measured data to the plant design basis, and to calculate the maximum number of SRV actuations each piece of equipment can withstand. Section 7.0 extends these conclusions to all equipment subject to SRV actuation induced loads.

6.1 Effect of Power Level The original Grand Gulf plant analysis was performed using a 95-95% pressure data base, i.e., there is 95% confidence that 95% of all pressures will be less than that used in the design. This confidence level was chosen to ensure that the calculated structure and equipment response would be conservative. The GESSAR pressure correlation, published in Appendix 6D of the Grand Gulf FSAR, uses the quantity MPPOV in calculating the peak pressure design values. The correlation provides a method to calculate suppression pool pressures for various plant conditions such as different power levels, reactor pressure and pool I

\

l MPL-09-100 6.1 !

Revision 0 l 11 tat _cac.:_F I

l temperature. This quantity contains the bubble correlation '

for predicted values and modifies it te, allow for the i

statistical variations in the basic data. ,

i Table 3.1 of Reference 11 provides a tabulation of MPPDV for both design and test conditions. The calculated ratio, {

between design and measured test values, for each type of j i

SRV test, is:

1.39 for SVA I 1

1.49 for CVA j 1.34 for MVA l l

For conservatism, the pcak measured strain values are i multiplied by a factor of 1.5 to determine an equivalent strain at design conditions.

i 6.2 Number of Valves Actuated l l

As described in Section 1.0, the S RV test series performed at Grand Gulf included the initial actuation of a single valve (SVA), subsequent actuation of a single valve (CVA),

j and simultaneous actuation of four valves (MVA). The plant design criteria described in Appendix 6D of the Grand Gulf FSAR considers the actuation of:

l One valve - first and subsequent actuations Two adjacent valves - initial actuation Eight valves (ADS) -

initial actuation Twenty valves - Vessel pressure >;1123 psi (all valves) first actuation MPL-09-100 6.2 Revision 0 nut _ec_h.

These cases were investigated during the plant design phase and design acceleration response spectra produced tor the three controlling (one valve subsequent actuatiom TOS and all valve) cases.

Avh.lable test results for the CVA case, when factored to design conditions, are directly comparable to the one valve subsequent actuation design response spectra. However, in order to be able to extrapolate results to predict the effect of the eight valve (ADS) and all valve cases it is necessary to establish a ratio for the measured test results.

The simplest, conservative, method to establish such a ratio is to compare the zero period accelerations (zpa's) t for all three design cases, at consistent locations

)

throughout the reactor building. The analytical approach l used to calculate the design response spectra used linear 1 I

l elastic theory and, therefore, provides a direct ratio of j 1

the total SRV input vertical load and overturning moment I for each design case. j i

Comparison of zpu's for each of the controlling design cases shows that the ratio of one valve to eight, or all valve cases varies from 1.0 to 4.3. Therefore, to conservatively establish the effect of the eight valve, and all valve design cases, all measured stresses were multiplied by a factor of 5.

MPL-09-100 6.3 Revision 0 nut _ec_h..

~ - - _ - _ _ _ _ _ _ _

~ ^

v, Eg 6.3 Number of Significant Stress Cycles To determine a cumulative damage factor due to SRV dis-charge loads, it is necessary to define the number of significant stress cycles associated with each event. As described in Reference 12, three cycle counting methods are available. These are the range-pair, rainflow and j racetrack methods. Each of these methods produces essentially - same resul t if all cycles are counted, ;

however, the latter two methods provide approaches which

- simplify the cycle counting while still providing an ;

accurate estimate of the number of significant stress cycles. i i

l Materials used in the manufacture of the HCUs and Hydrogen Recombiner are typically austenitic stainless steel with a j

, small amount of carbon steel. Figures 6.1 and 6. 2 a re I taken from the'1983 Edition of the ASME Code and provide j typical S-N curves for these materials. The endurance limit for both types of steels is approximately 7 ksi at 10 12 cycles. ThereLore, using the racetrack method of calculating significant cycles, it is reasonable to take ene width of the track as 7 ksi.

Examination of Figuros 5.1 through 5.12 and the calculated stresses shows that the peak stresses are an order of magnitude less than this value and thus there are no significant stress cycles for any SRV actuation. l Table 3.9-1 of the Grand Gult FSAR, provides the design number of SRV actuations during the plant life time. In total, 1820 actuations are considered; 220 are all valve actuations and 1600 are single valve actuations. Table MPL-09-100 6.4 Revision O nut _ec._h.

t,

3.9-1 states that each actuation is accompanied by 7 cycles due to air bubble oscillation. For piping desiqn, the ;

first 3 of 7 cycles are considered at full pressure range with the remaining 4 cycles at half range due to decaying pressures i.e., 5460 cycles at full range plus 7280 cycles at half range. For conservatism in calculating a cumulative damage usage factor, 20 cycles per actuation are l I

assumed for this calculation. i l

6.4 Expected SRV Fatigue Life l

l !

l To establish a very conservative SRV cumulative fatigue ;

usage factor for the HCU modules and Hydrogen Recombiner, the peak measured stress for all tests (0.56 ksi from MT70) l was applied to all cycles. The value of 0.56 ksi was {

multiplied by the largest factor described in Snction 6.1 )

(1.5) to allow for differences in power level from test to design for single valve actuations. For multiple valve I

actuations the 0.56 ksi was also multiplied by five as ;

1 described in Section 6.2. In addition to these very conservative factors, a fatigue strength reduction factor of two has been added to the measured stresses to account l

for small geometry and material flaws. Therefore, the peak stress for a CVA is 0.56 x 1.5 x 2 or 1.7 ksi and for an MVA is 0.56 x 1.5 x 5 x 2 or B.4 ksi.

I I

l The expected number of cycles is established in Section 6.3 i as 1,820 x 20 or 36,400 cycles. This is more conservative than the 5460 + 7280 cycles used for piping design or the 18,000 cycles used for the quencher design for fatigue l evaluation.

l l

l MPL-09-100 6.5 Revision 0 nutec. h.

As shown by examination of Figures 6.1 and 6.2, there are no practical fatigue limits for such small stresses. j However, assuming that the maximum permissible number of cycles are 106, then the cumulative fatigue usage far: tor for SRV actuations is 36,400/10 6 or 0.036, l

l 1

This very conservative number is insignificant when con-sidering the overall tatigue life of a piece of equip- i i

ment and it is clear that there are no SRV fatigue I related problems with the HCU modules or Hydrogen Recombiners at Grand Gulf.

1 j

1 l

l l 1

l 1

MPL-09-100 6.6 Revision 0 rit!t_eac_:l1

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'3 .n 's,eemea 6.7 MPL-09-100 Revision B nute,g ;
l l
i This is clearly shown by the minimal measured responses !
of the HCU module and Hydrogen Recombiner where measured data can be factored by an order of magnitude and still not approach any fatigue limits. Also, the strain measurements reported for the containment and quencher base, Reference 4, showed very small magnitudes compared i to predicted values. The Grand Gulf measured accelera-tions and strains are consistent with those reported from j the Kuosheng tests reported in Reference 15 and verify l l
~
that the small displacement high frequency accelerations produce little actual stress.
l Reference 14 showed that the typical SRV time history with its high frequency exceedences are of no concern for Grand Gulf piping and equipment. Modal participation factors, and the percent relative contribution of each l
mode, for five critical piping systems were evaluated. j The studies showed that between 50% to 90% of the total system response is captured by modes with frequencies i less than 30 Hz, and less than 10% of the total system l response comes from frequencies above 60 Hz. Therefore, as much as an order of magnitude reduction in system response could be expected if the test spectra were used to generate SRV discharge load stresses.
Also, generic studies conducted by GE, and reported in NEDE-25250, show that at frequencies above 60 Hz high accelerations are of na concern. For equipment qualified by test, the actual test response spectra are generally far above the predicted hign frequency exceedences.
In summary, the variations in measured acceleration response spectra are consistent with the reactor building MPL-09-100 7.2 Revision 0 nut.e_c_h.
design spectra. Therefore, the studies performed to show conservatism in calculated stresses are applicable to all reactor building systems and equipment. The conserva-tisms in the calculated stresses are also demonstrated by the low level of measured strain in the Hydrogen Recom-biner, HCU module and strain gauges mounted on the con-tainment and quencher base; the measured strains can be increased by an order of magnitade and still produce neglible fatigue effects in equipment or systems. The SRV induced strains are small when compared to material endurance limits and the SRV discharge loads are not of concern for the reactor building piping or equipment.
l l
I l
MPL-09-100 7.3 Revision 0 ritttenc:l1 kPwGerwEEtte u___.___________________
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CONTAINMENT BASE MAT ELEV. 93'-0", HORIZONTAL MPL-09-100 7*4 l Revision 0 l l
nutgch !
w-___---___-_ _ - - _ _ _ - _ - _ _ - _ _ _ _ _ _ _ - _ - _ . .-_ .- - _ - _ _ _ . __ _
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EINELOPE RESPONSE SPECTRA ACCELEROMETER A3, A5, A6, CONTAINMENT ELEV. 109'-1 1/2", HORIZONTAL I MPL-09-100 Revision 0 7.5 l I nutech 1
1
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l MPL-09-100 7.6 l Fevision 0 l
l nutech -- -- -
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l ENVELOPE RESPONSE SPECTRA ACCELEROMETER A7, CONTAINMENT ELEV. 147'-7", HORIZONTAL i MPL-09-100 Revision 0 l i nutech l
_ - - - - - - - - - i
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l !
MFL-09-100 Revision 0 7,g nutech
2.14 2,15
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Revision 0 l
l 7.9
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/ SRVone . [y ji 'iy I,
< ; di i ! ./ '.i. e t 0.16 - .l d'.\ \
f 'I
.j 0 cf
-f,; n.g y\.
l
'. J I 1
j Il N 'k
/ [
.,/ ,(l{W '
.. d '* N 'h, A ,'
0.C8 -
/
) ,1.$%
,s -
h\ i j 'y,.
)
w-.
- Js' -
0.00 l
'100 2 3 4 5 6 7 8 9 '101 2 3 4 5 6 7 8 9'102 FREQUENCY (CPS)
Figure 7.7 ENVELOPE RESPONSE SPECTRA ACCELEROMETER A13 CONTAINMENT DOME ELEV. 302'-3", VERT 1 CAL MPL-09-100.
Revision 0 7.10 nutgch
2.06 -
2.06 1.ss f f~ - N \ / \
/
" # 0.63 0.62
,/
0 30 .15 0.20 ,__*'7
- ' O.20 0.10 - O CV A I o SVA
+ MyA /
I I
g 0.08 - ggy ,i I i, I
SRVone .k 'i '
Q l l x .
l "j 0.06 - / l lqft w . ,i 1,l - -
O
) '
I j!
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/ !3 I
0.04 - / [; , j )\};'q \,k Il\jl
/ \n i "# i! Q *
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i % .l i $ '.![ l!(
/ '
N
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j e
s p x ,,
, / m,f(i s.q ..'-.' ,,
_) ' 'I ' .] i
'=~ = - Dr !
=.. --
0.00 g 0 3 4 5 6 7 8 9 '101 2 3 4 5 6 7 8 9 102 !
10 2 FREQUENCY (CPS)'
Figure 7.8 l
ENVFLOPE RESPONSE SPECTRA ACCELEROMETER A14 CONTAINMENT DOME ELEV. 302'-3", HORIZONTAL l MPL-09-100 Revision 0 7.11 ,
nutggj] !
l
.i
2.10 2.oS f-- m '
/.
g 2.00 -
/ i .e 1 l \u
/ '%
/
'3 's N o.s 0.50 - O C/A / \o.6
/
s SVA / os .}
+ MVA j
' \
/ I t
g 0.40 - t-g (-
- t J)
SRVail l f
fI
\{ 4 f
i j kl t 5
j 0.30 - l 1, ;$ !
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0.20 - / -
,i ,
,' / ,
) l l\ \ ,
0.10 - / ,O '
)
i ,- s, ,;/ . , ;
___/
0.00 "
100 2 3 4 5 6 7 8 9 '101 2 3 4 5 6 7 8 9 102 FREQUENCY (CPS)
Figure 7.9 .]
i ENVELOPE RESPONSE SPECTPA ACCELEROMETER A15, A17, A18 DRYWELL ELEV. 120'-10", HORIZONTAL i
i MPL-09-100 Revision 0 7.12 nutgch i
.72 0.7
/-- - ~"X l / \
/ \
.68 - f g
/ \
, / \ o.32
-m 1 0.24
3
/ :os \-
-g 0.20 - 0 CVA a SVA / '
+MVA j .
\
l \
\ *
/
. g 0,16 - \
.;)/. l.l f \ -
I j
z o j i \
\ -! / , i! '
q l r < '
< ./ \i s ,
e ,\
Lu 0.12 - / .' l' SRVan ,', /\lM. '
u
' :' il \ \ i f
./ .l .\I j,y f,"1
. d. i
! SRVone 1 'i ,
$I Y 0.08 -
I' f,
b
/I ,3 ,4 q d
/
l v. .
I 0.04 -
4- ./ 4 !
/ l / , v .~
i
' s ,/ / . m ?~
s 7 ; /
,_' 'W "
0.00 I100 2 3 4 5
7 o 9 101 2 3 4 5 6 7 8 9'102 '
FREQUENCY (CPS)
Figure 7.10 ENVELOPE RESPONSE SPECTRA ACCELEROMETER A16 DRYWELL ELEV. 120'-10", VERTICAL- a 1
l 1
MPL-09-100 Revision 0 7.13 nutgch
1.52 1.5
/ 1.4 s
/ ..g..%
1.48 - / \ 0.90 ' ,
p ',O.20 \ ,N
,# [ 0.30
.00 - f 0^0 ^'
c CV A /
" ,' s sV A - /
7 -MVA /
l l
@ 3.20 - / i z sav.a / s
'l 4) i 2 l \
l-
!U i > :
7 '/
= / 1 . t . li A ,
b 2 #0 ~ l SRVone f f } ,\ ; l l l\
N I 1 :j \, f' l',)) M
'j\' i
< / '
l . ,
/ '
e 's :t j A lV '
] !,/ If i j lg
, : I <
t 1.60 - / / ,p I / ,
i' <
M [ij :
l ' '
s
/ '
\s
~
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'\
U \
/ /
y .\ ,
'e
~
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0.80 - f 3
, i ,' 5 p
.A r. - .' A lI ,r e /h* 1
/ / Q <
,~ ,- ;
- \
^fO 0.00 i
'100 2 3 4 5 6 7 8 9 '101 2 3 4 5 6 7 8 9 '102 FREQUENCY (CPS)
Figure 7.11' EtWELOPE RESPONSE SPECTRA ACCELEROMETER A19 DRYWELL ELEV. 147'-6", HORIZONTAL MPL-09-100 Revision 0 7.14 nutg,gh
o
.32 0.81
/~ ~ n
/ 's
/ s oas
30 ,
/ 0.19 ua ~~m s 3 j
/ 0.1 *j U I \ '%
/ '% ~
0.12 7 ,
. 0.10 - I c CVA I a SVA l
MV A C
~
l Z 0.08 - ,
1 I
? I 3 I d
u l ;
y 0.06 - / ; 'p r
/ ,I
/ -
!:4 ,
!j ((1 f
SRVais j ; p 0.04 -
/ SRVone F If, '
/ $ +
ii l ')
\: il
/ / , /\, *
/ / i s (G
i
/ / \
0.02 -
/ ,-
'l \) ,' ,. ,s, fI g )d :
]y _:
m\A , v % v , %w ' v. ,.~sy ,j y ' >.'. , *'%,
+. , 'k,
'Y,y ' y ,.
A 17/ .__
x A -
0.00
'100 2 3 4 5 6- 7 8 9 '101 2 3 4 5 6 7 8 9 102 FREQUENCY (CPS)
Figure 7.12 ENVELOPE RESPONSE SPECTRA ACCELEROMETER A20
~
DRYWELL ELEV. 147'-7", VERTICAL MPL-09-100 Revision 0 7.15 .
nutgch
- - - - _ - _ - - - - - - - - - - - -- a
I.. j I i 1 i i
]
1 l
1 0.20 -
'O CVA d
a SVA
+ MV A 0.17
.- - - h g 0.16 - ;
z ;
9
~
l I
l
< I j
$ 0.12 - ---
}
W #
y / L 0.11 ,
n' 3 4 ,-- - / 0.10
\ [
t I \ i\ , .I I
\ \
SRVail - \ 0.085
\
0.08 - / \ .f
\ \ , i. j i SRyane ,
\ ,~, \ > 2 l
+ - t
,As

f
/
1
-i ,
,\ -
. N
/ N 4 ',)\l !
/ \ I/ .
4 i Ai
'k) }]}
1 a 0.04 - , ,
s sj k[' . \, .
- # s f_- l.% , ,.
\ '.
/ ,.-- ? / a
/ ., y) * %. A
,; ~ ~ 0 0.00 4.
00 2 3 4 5 6 7 8 9 '101 2 3 4 5 6 7 8 9 '102 FRECUENCY (CPS)
Figure 7.13 ENVELOPE RESPONSE SPECTRA ACCELEROMETER A21 DRYWELL ELEV. 184'-6", HORIZONTAL i
MPL-09-100 Revision 0 l 7.16 l nut.e_ch l
h .
.185
-p_ _ _s i
I I
.165 - (
. \
.ny - - > ,, ;
.115 ," -106 --
0.10 - e cy, j a SVA j g
. vvA , g i \ ,
i
_ i- s g t .C8 - f 2 l .
5 I b \
n I
\ i ll\ \
d 0 06 - SRVail I 1 i fI' \ ' ;"l f
5 l j' '\i .) \
\\\
l\
\ \
I \
'>\\ \N, t , i '\ \
,' , Savone '! \
('
c
) } , \ l\}y > lY{
/ ,'
A '.
. I d- !
i: h .;n l\' l ,4\s.ie,-
, ,, y ; l
/ j 4 <
l lMj \ \ lp b "' ]
0.02 - l ' \4 )> 'l
.n I . /\ ll
.)
i-j d fY!
100 2 3 4 5 6 7 8 9 '101 2 3 4 5 6 7 8 9'102 i
FREQUENCY (CPS) j Figure 7.14 ENVELOPE RESPONSE SPECTRA ACCELEROMETER A22 DRYWELL ELEV. 184'-6", TANGENTIAL 1 1
(HORIZONTAL)
MPL-09-100 7.17 !
Revision 0 !
nutsch !
.364- __,
/ \
/ \
/ \
l \
.356 - / \
/ (o.::s
~'
, / o. tis 0.14 O'25
's 0.12 4.00 - N o CV A / g_
a SVA o + MVA I J j 3.20 -
.. .G j z jl I 'N 3
i i
\
g 2.40 - SRVone / ; ,
/
ji j 8 /
s ]i i i L
< n
/
i I.l j
j 1 l
1.60 - / SRVall h !.W
/ '
O.80 - -
I 100 2 3 4 5 6 7 8 9 '101 2 3 4 5 6 7 8 9 102 FREQUENCY (CPS) i Figure 7.15 ENVELOPE RESPONSE SPECTRA ACCELEROMETER A25, A27, A28 ;
RPV PEDESTAL ELEV. 100'-9", HORIZONTAL j l
l MPL-09-100 l Revision 0 7.18 .
l nutgch l 1
E __ __ - - - - - - - - - - - - _ _ _ _ _ _ _ _ _ - _ - - - - - - - - - - - - - -
81.5 -
0.81
/"~~ ~ %
/ \ o,23 l 81.0 - , -- 0.21
/ o 22 ,
04 a SV A [
l N + MVA j
/
@ 2.00 - /
{
=
9
/ 60 Hz '
Noise
/
SRVone
-j 1.50 j .
U
< /
SRVail 1.00 - !
l 0.50 - /
0.00 i
'100 2 A.
3 4 5 6 7 8 9 '101 2 3 4 5 6 7 s9 102 8
FREQUENCY (CPS) l Figure 7.16 ENVELOPE RESPONSE SPECTRA ACCELEROMETER A26 RPV PEDESTAL ELEV. 100'-9", VERTICAL MPL-09-100 Revision 0 7,19 nutggb
i
8.0 CONCLUSION
S The stated test objectives of Section 1.2 have been !
I l
satisfied by:
f
- The strain and acceleration data collected during the SRV test program clearly establishes that the l piping and equipment response are considerably less I than predicted.
i
- Using the most conservative assumptions, the ;
'I measured SRV induced stresses were factored to show that the cumulative damage f actor due to SRV loads is effectively zero. Thi.s satisfies item 2 of Reference 1.
]
i 1
- The actual induced SRV stresses f or all equipment i are very small compared to the material endurance limits and SRV discharge loads are not of concern for any of the piping or equipment.
4 I
l I
MPL-09-100 8.1 Revision 0 nut _ec_h.
._ ___._____._ _ _ _ m
9.0 REFERENCES
I
1. Grand Gulf Nuclear Station Unit 1 Facility Operating l 1
License NPF-29,
2. Letter, L.F. Dale (MP&L) to H.R. Denton (NRC) AECM-l 82/150, Dated April 13, 1982.
3. Letter, L.F. Dale (MP&L) to H.R. Denton (NRC) AECM-85/0076, Dated March 11, 1985.
4. " Grand Gulf In-Plant Safety Relief Valve Test - Final Report", NUTECH Document No. MPL-01-220, Revision 0. l
5. " Grand Gulf In-Plant Safety Relief Valve Test - Shakedown Tests - Grand Gulf Startup Test Procedure 1-M62-SU-78-3 (Supplement 1)", NUTECH Document No. MPL-01-010, Revision 5.
6. " Grand Gulf In-Plant Safety Relief Valve Test - Matrix j Tests - Grand Gulf Startup Test Procedure 1 '9 ?-S U-7 8-3 (Supplement 2)", JUTECH Document No. MPL-01 . , Revision ;
5.
J
7. Letter, L. F . Dale (MP&L) to H.R. Denton (NRC)
AECM-85/0179, DatJd June 6, 1985. !
l
8. Letter L.F. Dale (MP&L) to H.R. Denton (NRC) I 1
l AECM-85/0196, Dated June 18, 1985. l 1
9. Letter L.F. Dale (MP&L) to H.R. Denton (NRC) l AECM-85/0212, Dated July 3, 1985.
I l
l MP L- 0 9- 10 0 9.1 Revision 0 nut.e_c_h.
i i
?
I
10. - Letter, T.M. Novak (NRC) to J.B. Richard (MP&L) Docket i No. 50-416, " Grand Gulf Nuclear Station Unit 1 - Safety j Relief Valve In-Plant Tests," Dated July 23, 1985.
11. " Grand Gulf In-Plant Safety Relief Valve Test -
I Acceptance Criteria for Real Time Pressure Measurements",
NUTECH Document No. MPL-01-035, Revision 2.
12. H.O. Fuchs and R.I. Stephens, " Metal Fatigue in Engineering," John Wiley and Sons , New York, 1980.
l 13. Requalification of Westinghouse Model B Electric Hydrogen !
I l
Recombiner for Grand Gulf Nuclear Power Station, Rev. O, I
t October, 1980.
14. Letter, L.F. Dale (MP&L) to H.R. Denton (NRC) AECM-82/79, Dated March 15, 1982.
15. " Final Test Report Safety Relief Valve Discharge Test -
Kuosheng Nuclear Power Station Unit No. 1", NUTECH Docume nt No. 2TP-06-310, Revision 0.
1 I
1 i
l i
I I
MPL-09-100 9.2 j Revision 0 rititeac:t1 !
-.

L-09-100, Rev 0 to MPL-09-100, Grand Gulf In-Plant Safety Relief Valve Test Fatigue Evaluation Final Rept

Contents

Text

SUMMARY

8.0 CONCLUSION

9.0 REFERENCES

1.0 INTRODUCTION

SUMMARY

8.0 CONCLUSION

9.0 REFERENCES

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L-09-100, Rev 0 to MPL-09-100, Grand Gulf In-Plant Safety Relief Valve Test Fatigue Evaluation Final Rept

Text

SUMMARY

8.0 CONCLUSION

9.0 REFERENCES

1.0 INTRODUCTION

SUMMARY

8.0 CONCLUSION

9.0 REFERENCES

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