ML20244D230

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Topical Rept Evaluation of G&W Fsd 2538, Nuclear MSIV Sys. Use of Seismic Qualification Procedure as Ref in Licensing Actions Acceptable
ML20244D230
Person / Time
Issue date: 03/31/1984
From:
Office of Nuclear Reactor Regulation
To:
Shared Package
ML20238A944 List:
References
FOIA-87-438 NUDOCS 8502270603
Download: ML20244D230 (24)
v • d • e


Text

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Enclosum Safety Evaluation Report on Gulf and Western Topical  !

Report G&W FSD 2538

" Nuclear Main Steam Isolation Valve Sys tens" l

Prepared by the Equipment Qualification Branch March, 1984 )

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ABSTRACT A comprehensive review of Gulf and Western Manufacturing Company's i Topical Report submittal for seismic qualification of its nuclear main steam isolation valve systems is made to determine compliance with current Nuclear Regulatory Commission (NRC) seismic qualification requirements for safety-related equipment. The seismic qualification procedures submitted by Gulf and Western are determined to be adequate per current NRC requirements. However, each specific plant application of the valve system must be reviewed to insure that the plant specific requirements do not exceed the Topical Report qualification levels.

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SUMMARY

- A. detailed technical evaluation of Topical Report GW-FSD-2538 by Gulf and Western Manufacturing Company was performed. The scope of the review was limited to an evaluation of the dynamic qualification m'ethocology Hydrodynamic loads were not considered in the qualification procedure which

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would be a concern for a b, oiling water reactor plant. The procedures are acceptable. - However, each application of the valve system will have to be reviewed in the light of specific plant requirements.

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, CONTENTS ABSTRACT .............................................................. 11

SUMMARY

.........................................................l...... iii-

1. . INTRODUCTION ,

.................................................... 1

2.

SUMMARY

OF G&W FSD'S MAIN STEAM ISOLATION VALVE SYSTEM!S SEISMIC QUALIFICATION PROGRAM ........................... 3 .

2.1 Deuble Trunnion Ball Valve ................................. 3' 2.2 Valve Actuator ............................................. 4 2.3 Logic Cabinet .............................................. 9 3.

EVALUATION ....................................................... 14 3.1 Evaluation of Valve Seismic Qualification .................. 14 3.L Evaluation of Valve Actuator Seismic Qualification ......... 15 3.3 Evaluation of Logic Cabinet Seismic Qualification .......... 16 3.4 'Open Items ................................................. 17

4. CONCLUSIONS ...................................................... -18
5. REFERENCES ....................................................... 19 APPENDIX A--REVIEW QUESTIONS .......................................... A-1

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1. INTRODUCTION The staff has completed a technical review of the topical report

" Nuclear Hain Steam Isolation Valve Systens" Report No. G&W-FSD-2538 submitted by Gulf and Western Manufacturing Company. The report was reviewed for compliance with current NRC dynamic qualification mquirenents for safety-related equipment including functional integrity. This safety evaluation report (SER) provides documentation of the resuits of the review.

Equipment in nuclear power plants that is used to perform a necessary safety function must be capable of maintaining functional operability ender all service conditions postulated to occur during the installed life for which it is required to operate. This requirement, which is enbodied in the Code of Federal Regulations 10 CFR Part 50, is applicable to safety-related equipment located inside as well as outside containment.

Safety-related structures , systems, and components are those that are relied upon to remain functional during and following design basis events to ensure (a) the integrity of the reactor coolant pressure boundary.

(b) the capability to shut down the reactor and maintain it in a safe shutdown condition, and (c) the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures conparable to the guidelines of 10 CFR Part 100. Design basis events are defined as conditions of normal operation,' including anticipated operation occurrences ; design basis accidents; external events; and natural phenomena for which the plant must be designed to ensure functions (a) through (c) above.

Revision 2 of NUREG-0800, NRC Standard Review Plan (SRP), Section 3.10 and the applicable portions of Section 3.9 describe seismic qualification requirements for nuclear power plant safety related equipment. Industrial Standard IEEE-344-1975, as supplemented by NRC Regulatory Guide 1.100 also contains seismic qualification requirements for nuclear po"er plant safety-mlated equipment. The seismic qualification requirements contained in these documents provide the technical basis for performining this review.

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1 l Gulf and Western (G&W) Manufacturing Company, Energy Pmducts Group, l Fluid System Division (FSD) has developed a main steam isolation valve

! system for installation in light water mactor power stations. This valve system consists of a large double-trunnion ball valve with an electro-mechanical fail-dafe rotary actuator and an electrical control panel . G&W FSD submittad a topical report (Ref.1) describing the engineering, construction, quality control, and testing programs undertaken and ongoing in the development of this valve system. G&W FSD also submitted seismic qualification analyses and test reports (Ref. 2-10) for this system. These reports were reviewed for conpliance with the current seismic qualification requirements discussed above. Initial reviews generated some questions which were forwarded to G&W. This final SER includes the evaluation of G&W msponses (Ref.11-14) to the initial review questions.

The G&W 24-in. main steam isolation valve was designed to be supported solely by the pip M g system. The valve will function mechanically in any position, however, th? recommended installation position, due to the valve system's basic design is in a horizontal run of piping with the valve mounted in the vertical and upright position and the actuator mounted on top of the valve in the plance of the piping run. Since the valve support and orientation may vary from plant to plant and since seismic, BWR plant hydrodynamic and valve opening and closing transient loadings am plant specific with mgard to both magnitude and frequency content, the scope of this review is limited to an evaluation of the sSmitted dynamic qualification methodology. Site specific applications must be evaluated on a case by case basis to assum that the dynamic qualification requirements for the specific application of the main steam isolation valve system conponents do not exceed the qualification levels for these components.

The first part of this report gives a sumary of G&W's qualification program. The second part has the evaluation of valve, actuator, logic cabinet and open itens. The subsequent part give our conclusio'n followed by references.

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- - 1 E.

SUMMARY

OF G&W FSD'S MAI STEAM ISOLATION VALVE SYSTEM'S SEISMIC QUALIFICATION PROGRAM 1

l- G&W's main steam isolation valve system consists of three components.

They are the double trunnion ball valve, the valve electro-mechanical actuator, and the actuator logic cabinet. Seismic qualification for each of these components is discussed separately in the following three sections.

2.1 Doutsle Trunnion Ball Valve The seismic qualification of G&W's 24 in. double trunnion ball valve was performed by analyses, using two and three dimensional finite element computer models of the valve. These analyses were performed using the ANSYS computer code. The analysis evaluated the valve stresses and I operability for combined normal operating loads and postulated seismic loading. Stress levels were evaluated based on ASME Boiler and Pressure Vessel Section III requirements for Class I valves. Operability of the valv.e was assured by extensive evaluation of critical clearances and alignment between valve parts to assure that no areas of interference or binding would occur. The valve loading consisted of design pressure i l loading, pipe collapse moment nozzle loads, and a static equivalent seismic loading of 7.5 g's applied in the three orthogonal model axes simultaneously.

The adequacy 6f the 24 in ball valve computer models was demonstrated by a developmer.t program consisting of testing and analysis of a prototypical valve system. A heavily instrumented 8 in, test valve was subjected to pressure, nozzle loads, vibration, and flow tests. Detailed analytical models of th'e 8 in. test valve were developed and analyses were performed for test loading conditions. A detailed comparison ci test and analysis results was performed to confirm the validity of the computer j model representation of the valve. l l

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1 2.2 . Valve Actuator .

The G&W 24.in. valve actuator was qualified for dynamic loading by testing performed by Wyle Laboratories and documented in Referenc'e 9. The qualification testing consisted of 5 OBE and 2 SSE random motion, biaxial

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tests in each test direction; lateral / vertical and longitudinal / vertical.

The valve and actuator were operated during each test series. The base input motion required response spectra (RRS) was as shown in Figures 1 '

through 4 The actuator was oriented in a vertical position on top of the valve in the plane of the piping. The test input motion response spectra enveloped these RRS for all frequencies above 4 Hz. Additional biaxial single frequency tests were performed at one-third octave frequency intervals over the frequency range of 1 to 4 Hz. The duration of the tests was approximately 30 s. The sine dwell test levels were as shown by the following:

a Run ~ Frequency Number Test Orientation (Hz) HCA* VCA 30 Lateral / vertical 1.0 9.0 9.0 31 Lateral / vertical 1.25 6.5 6.5 32 -

Lateral / vertical 1.6 12.0 13.0 34 Lateral / vertical 2.5 24.0 19.0 35 Lateral / vertical 3.2 34.0 24.0 36 Longitudinal / vertical 1.0 4.4 5.2 40 Longitudinal / vertical 1.0 15.0 16.0 41 Longitudinal / vertical 2.0 20.0 25.0 42 Longitudinal / vertical 2.5 28.0 23.0 43 Longitudinal / vertical 3.2 23.0 29.0 The valve actuator possessed sufficient integrity to withs'tand, '

without compromise of structure or isolation function, the prescribed simulated seismic environment. ~

The peak dynamic stress measured during the random multifrequency testing was determined to be 9750 psi,

a. HCA and VCA g levels are input g's analyzed at 1% damping.

HCA--Horizontal control accelerometer. VCA--Vertical control accelerometer.

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2.3 Logic Cabinet i loading The-G&W 24 in. valve logic cabinet was qualified di Reference 10.for seism c by testing performed by Wyle Laboratories dand documente _ n motion The qualification testing consisted _of 5 OBEd and 1 SSE ran om ,

The biaxial tests in each test direction, lateral / vertical, a longitudinal / vertical. diameter Grade 5 logic cabinet was attached to the testi table using s closely as 3/8 in, base bolts, simulating the actual in-service configurat f the test practical.

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axis was colinear with the longi The SSE test response spectra was as shown in in the horizontal plane. functionality of the logic cabinet was verified Figures 5 through 8. The lowest natural frequencies were determined during and after the tests. l nd greater than 33 Hz to be 30.1 Hz longitudinal, 30.9 Hz latera , a vertical.

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l 3. EVALUATION '

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This technical evaluation of G&W FSD's main steam isolation valve {

system's dynamic qualification program was performed using current NRC

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' qualification requirements for safety r, elated equipment. .A review of the l initial submittals (Ref.1-10) resulted in questions submitted by the '

staff. This evaluation is of the initial submittals and subsequent

, i submittals (Ref.11-14), responses to the questions. In general, the {

t qualification procedures are consistent with current NRC qualification requirements for safety related equipment. However, approval of these  !

procedures does not constitute a blanket approval of use of these valve systems in any nuclear plant. Since seismic BWR plant hydrodynamic and valve operating transient loading requirements (magnitude and frequency content) as well as valve mounting and orientation are plant specific, each plant application must be evaluated on a case by case basis to assure that the seismic qualification requirements with regard to magnitude and frequency content for the specific application of the valve system and corrpenents do not exceed the qualification levels.

3.1 Evaluation of Valve Seismic Qualification G&W's submittal on qualification procedures for their 24 in. double

- trunnion ball valve were reviewed with respect to current NRC seismic qualification requirements. for safety related equ'ipment.. The ability of  !

their analytical methods to predict the response of the valve to dynamic ,

and static loading was well demonstrated by their development program consisting of both testing and analysis of a prototypical valve. The 24 in, ball valve was analyzed for normal design static loading in .

combination with seismic loading consisting of pipe collapse nozzle loads and a static acceleration loading of 7.5 g's applied in the three orthogonal model axes simultaneously. Valve body stresses were below the material yield strength and ASME code allowable stresses for this loading.

Operability of the valve was as'sured based on a detailed' evaluation of clearances and alignment of valve parts to assure that no areas of interference or binding would occur.

14

It was determined that seismic qual'ification procedures for the 24 in, nuclear main steam isolation valve were adiquate per NRC current requirements.

Plant specific applications of the valve must be reviewed to assure that the specific seismic requirements do not exceed the valve l seismic qualification levels.

  • In addition to dynamic testing, considerable testing was performed on i the valve to develop assurance that the leakage requirements of 10 CFR 50 k

/,ppendix J and Criterion 54 of Appendix A are met. It was the feeling of 4 G&W that this assurance would negate the installation of the leakage l control system described in Regulatory Guide 1.96. Per Regulatory Guide 1.96, such a proposal must be reviewed by the NRC staff for f acceptability or a leakage control system must be provided. 1 3.2 Evaluation of Valve Actuator Seismic Qualification G&W's procedure for seismic qualification of their valve actuator (EFC0 600) was reviewed for comp 1.iance with current NRC qualification procedures for safety related equipment. The testing of this actuator was performed by a combination of single frequency biaxial and multifrequency biaxial tests. The actuator was bolted to a test table using the same bolting arrangement as is used in mounting the actuator to a valve. The

^ actuator was then subjected to a series of 5 ODE and 2 SSE tests with phase incoherent biaxial input (longitudinal / vertical and lateral / vertical).

Actuator solenoid valves were tested in both energized and de-energized conditions. The single frequency biaxial tests were performed to complement the multifrequency tests since the multifrequency test motion response spectra did not envelope the RRS below 4 Hz. Operability of the actuator was verified during and after the tests. No str'uctural damage to {

l the actuator was observed to occur as a result of the tests. The maximum l

stress level calculated from stress gauge readings during the tests was 9750 psi.

In addition to the valve actuator test report (Ref. 9), G&W also submitted a dynamic analysis of their series 600 actuator frame. Other than as a design guide, this report has little bearing on the total seistric qualification effort of the actuator. This report should not be 15

i id si,nce this analysis was of t

interpreted to mean that the actuator is r g the frame only. The mass of the actuator functional de included in this analysis. h mechanism or el.ectrical internal safety related devices such h as the latcgross structural natural '

components must be considered as well as t e frequency of the actuator. ,

t be considered in Further, for a BWR plant, hydrodynamic loading musGen combination with seismic loading. 33 Hz upper range of seismic idered rigid in a hydrodynamic loading is considerably above th motion.

BWR plant, it typically must have a fundamenta 50-60 Hz range.

f the seismic Based on the above discussion, the evaluation olimited to the s qualification of EFCO-600 actuator wasIt was determined that the seismic report submitted, Reference 9. h ctuator is in accordance with  ;

qualification testing performed for t eta for safety related equipment current NRC seismic qualification requiremen sHowever, each plant .

and industrial standard IEEE-344-1975. iewed to insure the seismic  ;

application of these actuators should be revtion do not exceed the i requirements of the specific applicaThe seismic input motion a qualification levels. d and compared to the acceleration fied as shown by Figures 1-4.

attachment location must be determine levels to which the operator has been quali i

3.3 Evaluation of Logic Cabinet Seismic Qualificat on .

isolation valve system's l

The qualification reports for the main d steam with current NRC logic cabinet were determined to be in accor ancef safety relatl loading by random i requirements for seismic qualification oThe testing.

motion phase incoherent biaxial testsTest mounting of the cabin .

longitudinal / vertical). Operability of the cabinet and binet.

recommended field mounting of the ca h tests. Tests demonstrated its contents was verified during and after t e

'--w--__.,.,___,___

I that the logic cabinet possessed sufficient integrity to withstand, without compromise of structures or electrical functions, the simulated seismic environment.

G&W's seismic qualification of the main steam isolation valve system's logic cabinet was performed adequately in accordance with current NRC seismic qualification requirements. However, each specific plant application of the logic cabinet should be reviewed to insure that the seismic qualification requirements for the specific plant do not exceed the qualification levels of the logic cabinet as shown by the test response spectra on Figures 5-8. .

3.4 Open Items The G&W submittals made no mention of BWR plant hydrodynamic loading.

Current seismic qualification requirements state that the required response spectra be representative of the seismic environment postulated for the

~

safety V'e lated equipment and other expected concurrent loadings, such as valve operating transients and hydrodynamic loading for BWRs. This has no bearing on the seismic qualification methodology just reviewed, except that for plant specific applications it must be shown that the RRS for a combination of seismic and hydrodynamic loading is enveloped by the

__ qualification test response spectra.

Since the G&W 24-in, valve is not provided with a leakage control system, a staff review of the applicants proposed approach to implementing General Design Criterion 54 with regard to the control or limitation of leakage past the main steam isolation valves must be made to determine its acceptability.

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4. CONCLl!SIONS I

The seismic qualification procedure submitted by G&W FSD for their main steam isolation valve systems has been reviewed. The . finite element model used in the analysis is capable of predicting responses from seismic and other dynamic loads. The qualification procedure is adequate per current NRC seismic qualification requirements for safety related equipmen t. This evaluation, nowever, does not constitute a generic approval of use of these valve systens.in any nuclear plant. It is an approval of the seismic qualification procedure only for referencing in licensing actions. Each specific plant application of the valve system must be reviewed to insure that the specific plant seismic alog with other dynamic requirements have been net by the systens seismic qualification levels.

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5. REFERENCES

{

1. Gulf and Western Fluid Systems Division (Warwick, Rhode Island)

Report, " Topical Report. Nuclear Main Steam Isolation Valve Systems,"

Report No. G&W-FSD-2538, January 1979.

2. Basic Technology Incorporated (Pittsburgh, Pennsylvania) Report,

" Validated Analysis of an 8 Inch EPG Double Trunnion Ball Valve,"

Report No. BTI-78006-'1, Vo_lume 1, May 15, 1978, Revision 1.

3. Basic Technology Inco'rporated (Pittsburgh, Pennsylvania) Report,

'" Structural Functional Performance of a 24 Inch Nuclear Double-Trunnion Ball Valve," Report No. BTI-78006-2, October 1978.

4. Wyle Laboratories (Huntsville, Alabama) Test Report, " Seismic Simulation Test Program on an 8" Ball Valve and Operator," Report (

Number 43144-3, August 9, 1976. 1

5. Wyle Laboratories (Huntsville, Alabama) Test Report, "Stt Valve Test on One Eight-Inch Ball Valve and Operator," Report Number 93144-4, August 23, 1976.
6. Wyle Laboratories (Huntsville, Alabama) Certification Test Report on One Eight-Inch Ball Valve, Report Number 43144-01, June 9, 1976.
7. Gulf and Western Fluid Systems Division (Warwick, Rhode Island)

Report, " Dynamic Evaluation Tests of an EFC0 600 Actuator," Report Number G&W FSD 1798N, September 1979.

8. Gulf and Western Fluid Systems Division (Warwick, Rhode Island)

Report, " Dynamic Evaluation Tests of an EFC0 Actuator,60-150 Hz,"

Report Number G&W 1798-2, March 1980.

)

, , . 9. WyleLaboratories(Huntsville,diabama)TestReport," Seismic Simulation Test Program on an EFC0 600 -Actuator," Report j Number 44473-1, June 6, 1979.

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10. Wyle Laboratories (Huntsville, Alabama) Test Report, " Seismic Simulation Test Program on a EFC0 600 Actuator Logic Cabinet," Report Number 44473-3, June 15, 1979.
11. D. A. Weisz, letter to C. O. Thomas, Response to Reouest Number 1 for Additional Information on GEW-FSD 2538, FSD File No. 83-11, January 13, 1983.

i

12. D. A. Weisz, letter to C. O. Thomas, Response to Requist Number 1 for Additional Information on G&W-FSD 2538, FSD File No. 83-21, January 20, 1983.

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13. D. A. Weisz, letter to C. 0 Thomas, Response to Request Number 1 for l Additional Information on G&W-FSD 2538, FSD' File No. 83-28, l January 26, 1983.
14. D. A. Weisz, letter to C. O. Thomas, Response to Request Number 1 for Additional Information on G&W-FSD 2538, FSD File No. 83-43, l February 8, 1983. l i

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