ML20027A439
| ML20027A439 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 11/15/1978 |
| From: | Rosa F Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20027A437 | List: |
| References | |
| NUDOCS 7812040327 | |
| Download: ML20027A439 (12) | |
Text
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- ? e TESTIMONY OF FAUST ROSA
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.: 1 i Diablo Canyon Testimony of John Knox and Faust Rosa Seismic Qualification of
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Class 1E Equipment Introduction A detailed description of our evaluation of the seismic qualification of Class 1E equipment for Diablo Canyon is contained in Section 3.10 of the Safety Evaluation Report and its Supplements 7 and 8.
This des-t' cription includes identification of the Class 1E equipment and the applicable seismic criteria, and a discussion of how these criteria were applied in evaluating the seismic qualification that was performed.
This testimony will augment this description with emphasis on the elec-trical aspects of the seismic evaluatitm, particularly the ames iden-tified in the Intervenor's Response to Applicant's Interrogatories Dated September 27, 1978. A summary status of the seismic evaluation of Class 1E equipment as of December 1,1978 is also included.
General As stated in Section 3.10.2 of SEF. Supplement 7, the majority of the safety-related electrical instrumentation and centrol equipment was qualified by testing. The balance was qualified by analysis, or a combination of test and analysis. This equipment was previously quali-fied in accordance with IEEE Standard 344-1971, "IEEE Guide for Seis-mic Qualification of Class I Electrical Equipment for Nuclear Power Generating Statiers," to the level of the double design earthquake approved for the construction permit or higher. Where the original
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2-qualification level does not envelope the required seismic inputs to equipment for the Hosgri event, we have required the applicant to requalify the equipment for the Hosgri required response spectra. This has been done, principally by retesting using the required response spectra.
In the requalification process the applicant employed seismic qualifi-cation methods that conform to our current criteria (Regulatory Guide 0
1.100, Revision 1, " Seismic Qualification of Electrical Equipment fo-Nuclear Power Plants," August 1977, and IEEE Standard 344-1975, "IEEE Recomended Practice for Seismic Qualification of Class lE Equipment for Nuclear Power Generating Stations").
This updating to current criteria applies to the seismic qualification methods including shake testing methods 'and the type and severity of shak-ing employed.
It did not, however, include the sequential aging require-ments and other general environmental requalification recomendations that are reflected in our current positions for new plants and are re-ferenced in Regulatory Guide 1.100. That is, the sequential aging requirement prior to seismic testing is not included in the qualification i
criteria for plants of the Diablo Canyon vintage. Our current criteria i
for environmental qualification for new plants are described in Regula-tory Guide 1.89, " Qualification of Class lE Equipment for Nuclear Power Plants," November 1974, ar.d IEEE Standard 323-1974, "IEEE Standard for Qualifying Class lE Equipment for Nuclear Power Generating Stations,"
February 1974.
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- 's 3-Seismic Evaluation Summary For Diablo Canyon, the seismic qualification of Class IE equipment must (1) demonstrate that the equipment can withstand the effects of five Operating Basis Earthquakes, and following this, (2) demonstrate the equipment's ability to perform its required function during and after the time it is subjected to the forces resulting from one Safe Shutdown Earthquake.
Our evaluation includes review of test data and other supporting analy-ses and documentation to ascertain the adequacy of the required demonstration of seismic capability. More specifically, since quali-fication of most equipment is based on seismic (shake) testing, our review has or will establish that the equipment performance monitoring performed during testing provides a valid demonstration of functional-ability during and following a seismic (Hosgri) event.
The following tabulation provides a summary of the seismic qualification including:
(1) a list of the Class lE equipment, (2) the location of the corresponding seismic documentation, and (3) the basis for accep-tability and present status of our evaluation. A detailed description
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of our evaluation for specific equipments is contained in Supplements 7 and 8 of the Safety Evaluation Report.
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Basis and Status Catecory of Seismic Qualification of Class lE Equipment A.
Original qualifiestion per IEEE Std 344-1971 enveloped the Hosgri event and is acceptable.
B.
Requalification was required to envelope Hosgri event. This was performed per Regulatory Guide 1.100, Rev. I and IEEE Std 344-1975 (except aging), and was found acceptable.
C.
Requalification to envelope Hosgri was required and performed.
This was found acceptable subject to submission of additional confirmatory justification or test results.
D.
Requalification to envelope Hosgri was required and performed.
Additional testing required to confirm electrical functionability will be performed. Found acceptable subject to successful con-firmatory testing.
E.
Further seismic evaluation is required; if evaluation of the quali-fication performed is not acceptable, then additional testing, addi-tional justification, design modificat1ons, or replacement of equipment will be required.
SEISMIC QUAL.I-CLASS lE FSAR AMENDMENT 50 FICATION BASIS EQUIPMENT SECTION NO.
AND STATUS Nuclear Stean Supply System Equipment 1.
Auxiliary Safeguards Cabinet 10.3.2 A
2.
Static Inverter 10.3.10 A
3.
Nuclear Instrumentation System 10.3.16 A
4.
Pressure and Differential Pressure Transmitters 10.3.17 E
5.
Process Control and Prc.tection Equipment 10.3.19 A
6.
Reactor Trip Switchgear 10.3.20 A
7.
Solid State Protection System 10.3.22 A
8.
Resistance Temperature Detectors 10.3.27 E
9.
Safeguards Test Cabinet 10.3.28 A
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- 10. Fan Cooler Motor 3.10.2.16 (FSAR)
A Balance of Plant Ecuipment 1.
Battery Char:ar 10.3.1 0
2.
Station Battery 10.3.4 C
3.
DC 125/250 VDC Motor Control Center 10.3.5.1 D
4.
125 VDC Distribution Panel 10.3.5.2 D
5.
Diesel Generator Excitation l
Cubicle 10.3.6 0
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. SEISMIC QUALI-CLASS 1E FSAR AMENDMENT 50 FICATION BASIS EQUIPMENT SECTION NO.
AND STATUS Balance of Plant Eouipment 6.
Diesel Generator Control Cabinet and Subpanel 10.3.6 0
7.
Fire Pump Controller 10.3.7 0
8.
Emergency Light Battery Pack Later B
9.
Hot shutdown panel 10.3.9 (a) Indicating Meters C
(b) Switches B
(c) Fisher Cortroller B
10.
Instrumentation Pov.er AC Panel-boards 10.3.11 E
11.
Instrumer.t Panels PIA, PIB and PIC 10.3.12 A
- 12. Local Instrument Panels 10.3.13 A
- 13. Local starters 10.3.14 E
- 14. Main Control Board (a) Indicating Meters C
(b) Switches B
- 15. Pressure and Differential Pressure transmitters 10.3.18 A
- 16. Safeguards Relay Board 10.3.21 C
- 17. Ventilating Control Logic Cabinet 10.3.23 E
- 18. Ventilating Control Relay 10.3.24 E
Cabinet
- 19. Vital Load Centers 10.3.25 E
- 20. Vital Load Center Auxiliary Relay Panel 10.3.25A E
- 21. Fan Cooler Motor Controller 10.3.25B E
- 22. 4160-volt Switchgear 10.3.26 D
- 23. Limotorque Valve Operator with Gear and Stem Mounted Limit Switches 10.3.30 C
- 24. Diesel Generators 10.3.6 A
- 25. Cable Trays 10.3.29 A.
- 26. Penetrations 10.3.7 A
s The seismic qualification of the equipment in status categories A and B has been found acceptable on the basis indicated. The equipment in categories C and D are also considered to be acceptably qualified; however, additional justification or testing is required to resolve any questionable monitoring of functionalability during prior testing. We will review the additional justification and confirmatory testing to verify the adequacy of qualification in this regard. Our evaluation of the qualification of the equipment in status Category E is incomplete; we will require that this equipment be acceptably qualified by the methods indicated prior to completion of our review.
Non inclusion of 4cino in Seismic DuaMfication As stated above the acceptance criteria for the qualification of Class lE equipment for plants of the Diablo Canyon vintage did not include the aging consideration specified in IEEE-Standard 323-1974 and Regula-tory Guide 1.89 (which endorses IEEE-323-1974).
In 1974, during the deliberations of the NRC's Regulatory Requirements Review Committee on the implementation of Regulatory Guide 1.89, con-sideration was given to the incremental improvements to safety it afforded in comparison of the then current staff review practice.
The Committee recommended thi.t the guide be applied only to future CP applications; i.e., it should not be backfitted. The decision was l
based on the Staff's judgment that the incremental imorovements were i
j not significant to safety and that full impler.entation of IEEE-323-1974 required the further development of other ancillary standards to pro-vide guidance on specific safety-related equipment and components, i
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l m Subsequent public comments and review by the ACRS did not alter the recommendation concerning implementation of Regulatory Guide 1.89.
We recognize that additional guidance is needed in the area of accelerated aging techniques used to establish a qualified life for electrical equip-ment and assemblies. Our Category A technictl activity on equipment qualification (Task Action Plan A-24) and an NRC extensive research i
program being carried cut at Sandia Laboratires are intended to provide additional guidance for the development of test methods and licensing review procedures on aging. These programs will also allow us to make informal judgements regarding the effects of aging.
In addition, as
- p. art of the Staff's Systematic Evaluation Program (SEPO, the Staff is assessing the surveillance and maintenance records of the eleven SEP plants for equipment inside and outside of containment. Since this equipment has been effectively " aged", the assessment of these records should provide additional infomation on the effects of aging.
Following completion of these ongoing activities -- the Task Action Plan A-24, the NRC research program, and the SEP effort -- we will reconsider our position on the need for backfitting the aging requirements. At I-that time, should we deem it necessary, we will take appropriate steps to ensure that acing effects are considered in assessing the adequacy of Class lE equipment ured in the Diablo Canycn plant.
It is our judgment that tne natural aging that the Class lE equipment will undergo in the period prior to this reassessment will have little effect on its seismic capability.
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. 1 Installation of Seismically Tested Eauictent Some of the Class 1E equipment which has been shake tested to seis-mically qualify it for the Hosgri event will be installed in the plant.
In all such cases only one of a redundant set of equipment will have been tested. ThE tested equipment Will have demonstrated electrical 1:
functiontbility during and following the testing; and it will be carefully inspected after testing to assure that its structural and electrical integrity has not been impaired, and that it remains fully capable of withstanding a Hosgri event. Therefore, we conclude that the installation of tested equipment is acceptable.
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FAUST ROSA
' PROFESSIONAL QUALIFICATIONS PO7ER SYSTEMS BRANCH DIVISION OF SYSTEMS SAFETY I have been employed by the Nuclear Regulatory Commission since January 1971. From January 1977 to the present time, I have been Chief, Power Systems Branch, Division of Systems Safety. Prior to my present assign-s t
ment I served as a Section Chief in the Electrical, Instrumentation and Control Systems Branch, Division of Systems Safety, and in the Plant Systems Branch, Division of Operating Reactors. I have participated in the review of instrumentation, control and electrical systems of numerous nuclear power stations and in the formulation of related standards and Regulatory Guides.
The Power Systems Branch performs an in-depth technical review of the design, fabrication, qualification and operation of nuclear power plant electrical power systems important to safety and the related instrumentation and con-trols. The area of branch review responsibility also includes that portion of the steam system downstream of the main steam isolation valves. This review includes a comprehensive assessment of these systems for all power reactors for adherence to appropriate codes and standards and encompasses complete evaluation of applicant's safety analysis reports, generic reports, and other ralated system design information. Further, the Branch develops the bases for Regulatory acceptance criteria for electrical power systems designs; evaluates experience obtained during the construction and operation
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Paust Rosa 2
t of nuclear power plants and relates this information to future evaluations
.l and acceptance criteria; and participates in the development of Regulatory Guides and regulations pertaining to electrical power systems and other systems in the branch area of responsibility.
I hold a Bachelor of Electrital Engineering degree from the University of Pittsburgh, Pittsburgh, Pennsylvania. In addition, I have taken courses s i
in Mathematics, Theoretical Physics, Nuclear Physics and Engineering, and Radiation Shielding at the University of Pittsburgh and at the Reactor School of the Bettis Atomic Power Laboratory, Westinghouse Electric Corpora-tion.
9ty wlear engineering experience background derives from my employment at I
the Bettis Atomic Power Laboratory of Westinghouse Electric Corporation, West Mifflin, Pennsylvania, from May 1955 to September 1962; and from my employment at the Bechtel Corporation, Vernon, California, from September 1969 to January 1971. At Bettis Laboratory I was a lead engineer in the nuclear submarine power plant group with technical responsibility for nuclear instru-mentation, rod control, and reactor protection systems. Work involved com-4 l
ponent and system design, installation, testing, modification and documenta-I
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tion. I also served as Bettis representative during full-scale tests con-i ducted by the Navy. At Bechtel I conducted engineering studies and prepared reports and specifications relating to the design and construction of the Rancho Seco Nuclear Power Station. This work was primarily in the areas of safety-related electrical power, instrumentation and control systems.
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A Faust Rosa 3
My non-nuclear engineering background derives primarily from my employment in the Construction Engineering Department of the National Tube Company, United States Steel Corporation, Lorain, Ohio, from June 1947 to April 1955; and from my employment at the Rocketdyne Division of North American Rockwell Corpor-ation, Canoga Park, California, from October 1962 to March 1968. At National Tube I served as a Senior Engineer engaged in design and development of electrical power and control systems for new pipe mills from cr.,nceptual des,ign i
e through 4ecail design, procurement, installation, and initial operation.
This work extended through completion of two major pipe mill construction projects. At Rocketdyne I was a Research Specialist engaged in design and development of controls and instrumentation for a dual turbo-pur.p liquid hydrogen feed system for a nuclear rocket engine. My primary responsibility was for control system integration extending from conceptual design through procurement, installation, and completion of the. test program.
I am a member of the Institute of Electrical and Electronic Engineers and have participated in the nuclear st.andards development work of this organ-ization since 1972.
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