ML20196E024
| ML20196E024 | |
| Person / Time | |
|---|---|
| Issue date: | 02/16/1988 |
| From: | Jordan E Committee To Review Generic Requirements |
| To: | Bernero R, Martin T, Ross D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I), NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS), NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| Shared Package | |
| ML20153G441 | List: |
| References | |
| REF-GTECI-A-40, REF-GTECI-SC, RTR-REGGD-01.100, RTR-REGGD-1.100, TASK-A-40, TASK-OR NUDOCS 8802250361 | |
| Download: ML20196E024 (9) | |
Text
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February 16, 1988 MEMORANDUM FOR:
Robert M. Bernero, NMSS Thomas T. Martin, RI Denwood F. Ross, RES Joseph Scinto, OGC James H. Snlezek, NRR FROM:
Edward L. Jordan, Chairman Committee to Review Generic Requirement
SUBJECT:
CRGR MEETING NO. 129 The Committee to Review Generic Requirements (CRGR) will meet in the afternoon on Tuesday, February 23, 1988, in Room 6507 MNBB.
The agenda is as follows:
1-2 p.m.
G. Arlotto will present for CRGR review proposed final Regulatory Guide 1.100, Revision 2, "Seismic Qualification of Electric and Mechanical Equipment for Nuclear Power Plants."
(Review package previously distributed.
Clean copy of Reg. Guide enclosed.)
2-3 p.m.
C.E. Rossi will present for CRGR review a proposed NRC Bulletin on "Inadequate Latch Engagement in GE HFA Type Latching Relays."
(Review package is enclosed.)
3-4 p.m.
L. Shao will present for CRGR review a proposed Generic Letter on "Boric Acid Corrosion of Carbon Steel Reactor Vessel Compon-ents in PWR Plants." (Review package is enclosed.)
NOTE:
CRGR review of the proposed resolution for USI A-40, "Seismic Design,"
previously scheduled for review at this meeting, has been postponed at the request of the sponsoring office.
That item has now been tentatively rescheduled to Meeting 130; CRGR members should retain the review packages already distributed for that item.
If a CRGR member cannot attend the meeting, it is his responsibility to assure that an alternate, who is approved by the CRGR Cnairman, attends the meeting.
Persons making presentations to the CRGR are responsible for (1) assuring that the information required for CRGR review is provided to the Committee (CRGR Charter - IV.B), (2) coordinating and presenting views of other offices, (3) as appropriate, assuring that other offices are represented during the presenta-tion, and (4) assuring that agenda modifications are coordinated with the CRGR contact (J. Conran X29855) and others involved with the presentation.
Division Directors or higher management should attend meetings addressing agenda items under their purview.
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In accordance with the E00's March 29, 1984 memorandum to the Commission con-cerning "Forwarding of CRGR Documents to the Public Document Room (POR)," the enclosure, which contains predecisional information, will not be released to the PDR until the NRC has considered (in a public forum) or decided the matter addressed by the infurmation.
Orio aal 89=d 4 n
L C:)wihn Edward L. Jordan, Chairman Committee to Review Generic Requirements
Enclosures:
As stated cc w/ enclosures:
SECY V. Stello, Jr.
cc w/o enclosures:
Commission (5)
J. Lieberman S. Ebneter W. Mcdonald Regional Administrators W. Parler G. Arlotto Distribution:
E. Jordan J. Heltemes J. Conran C. Sakencs R. Fraley B. Doolittle (w/o enc.)
CRGR SF S. Treby (w/o enc.)
T. Rehm (w/o ene.)
J. Johnson, OE00 l
POR (!!RC/CRGR) (w/o enc.)
Central File (w/o enc.)
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[ August] December 1987 Division 1 Task EE 108-5
Contact:
S. K. Aggarwal (301) [443-7840] 492-3829
[PROPOSEB] REVISION 2 TO REGULATORY GUIDE 1.100 SEISMIC QUALIFICATION OF ELECTRIC AND MECHANICAL EQUIPMENT FOR NUCLEAR POWER PLANTS A.
INTRODUCTION The Commission's regulations in 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities," require that certain structures, systems, and components in a nuclear power plant be designed to withstand the effects of natural phenomena such as earthquakes, and that design control measures such as testing be used to check the adequacy of design.
This general requirement is contained in Appendix A, "General Design Criteria for Nuclear Power Plants,"
to Part 50; in Criterion III, "Design Control," and Criterion XVII, "Quality Assurance Records," of Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to Part 50; and in Appendix A, "Seismic and Geolog;c Siting Criteria for Nuclear Power Plants," to Part 100, "Reactor Site Criteria."
In Appendix A to 10 CFR Part 100, Section VI, "Application to Engineering Design, requires that the nuclear power plant be designed so that, if the safe shutdown earthquake occurs, certain structures, systems, and components will I
remain functional.
These safety-related structures, systems, and components are those necessary to ensure (1) the integrity of the reactor coolant pressure boundary, (2) the capability to shut down the reactor and maintain it in a safe condition, or (3) the capability to prevent or mitigate the consequences of accidents that could result in offsite exposures comparable to the Part 100 guidelines.
In Appendix A to Part 100,Section VI(a)(2) requires that l
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structures, systems, and components of the nuclear power plant necessary for continued operation without undue risk to the health and safety of the public-be designed to remain functional and within applicable stress and deformation limits when subjected to the effects of the vibratory motion of an operating basis earthquake in combination with normal operating loads.
The engineering method used to ensure that the required safety functions are maintained during and after the vibratory ground motion associated with the safe shutdown earthquake or the operating basis earthquake must involve the use of either a suitable dynamic analysis or a suitable qualification test to demonstrate that structures, systems, and components can withstand the seismic and other con-cur ent loads.
This regulatory guide describes a method acceptable to the NRC staff for complying with NRC's regulations with respect to seismic qualification of i
electric and mechanical equipment.
Any information collection activities mentioned in this [ draft) regulatory guide are contained as requirements in 10 CFR Parts 50 or 100, which provide 4
the regulatory basis for this guide.
The information collection requirements in 10 CFR Parts 50 and 100 have been cleared under OMB Clearance Nos. 3150-0011 and 3150-0093, respectively.
B.
DISCUSSION IEEE Std 344-1987, "Recommended Practice for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations,"* was prepared by
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Working Group 2.5 (Seismic Qualification) of Subcommittee 2 (Equipment Quali-fication) of the Institute of Electrical and Electronics Engineers (IEEE)
Nuclear Power Engineering Committee, and was subsequently approved by the IEEE Standards Board on June 11, 1987.
The IEEE Standard includes principles, procedures, and methods of seismic 4
qualification that, when satisfied, will confirm the adequacy of the equipment design for the performance of safety functions before, during, and after the i
time the safety-related equipment is subjected to high stresses resulting from "Copies may be obtained from the Institute of Electrical and Electronics Engineers, IEEE Service Center, 445 Hoes Lane, P.O. Box 1331, Piscataway, NJ 08855.
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a design basis events.
For this guide, the design basis events are the loadings imposed by seismic events:
the operating basis earthquake (0BE) and the safe shutdown earthquake (SSE).
It is also necessary to combine other dynamic or vibratory loads as part of seismic qualification.
It is recognized that hydrodynamic loads have their primary energy content in a frequency range greater than that of seismic vibrations; however, they are a part of the in-plant equipment aging process, along with other nonseismic vibration loaos, and therefore should be considered in seismic testing.
Revision 1 of this guide was issued in August 1977.
Since then, several new technical issues have arisen, such as treatment of hydrodynamic loads, the lieits of generic testing, the treatment of ra+tling, methods of qualifying line-mounted devices, and the use of actual seismic experience data bases to qualify identical or similar equipment.
These issue are covered by IEEE Std 344-1987, which reflects the state-of-the-art technology.
Further, the NRC has extended the application of this standard to the qualiiication of mechanical equipment on an interim basis.
In extending the application of IEEE Std 344-1987 to mechanical equipment, the NRC staff recognizes that there are differences in qualification methods for electric and mechanical equip-ment.
Specifically, qualification of mechanical equipment by analysis is per-mitted when such equipment can be modeled to adequately predict its response.
The American Society of Mechanical Engineers is currently developing a standard for seismic qualification of mechanical equipment.
Upon publication of this standard, the NRC staff will review it for suitability for endorsement by a revisfon to this regulatory guide.
This regulatory guide covers two categories of equipment:
(1) safety-related electric (Class IE) equipment and safety-related mechanical equipment, and (2) non safety-related equipment whose failure can prevent the satisfactory accomplishment of safety functions.
Examples of mechanical equipment (and equipment-sepperts] within the scope of this guide are valves, valve operators, pumps, compressors, chillers, air handlers, fans, blowers, [ pipe-supports-snebberst-restraints--hangers-] fuel rod assemblies, and control rod drive mechanisms.
IEEE Std 344-1987 references other standards that contain valuable informa-tion.
Those referenced standards not endorsed by a regulatory guide or incor-porated into the regulations, if used, are to be used in a manner consistent with current regulations.
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C.
REGULATORY POSITION The procedures described by IEEE Std 344-1987, "Recommended Practice for Seismic Qualification of Class IE Equipment for Nuclear Power Generating Stations," are acceptable to the NRC staff for satisfying the Commission's regulations pertaining to seismic qualification of electric and mechanical equipment subject to the following:
1.
For mechanical equipnent, thermal distortion effects on operability should be considered, and loads imposed by the attached piping should also be accounted for.
If dynamic testing of a pump or a valve assembly is impracticable, static testing of the assembly is acceptable provided that (1) the end loadings are I
applied and are equal to or greater than postulated event leads, (2) all dynamic amplification effects are accounted for, (3) the component is in the operating mode during and after the application of loads, and (4) an adequate analysis is made to show the validity of the static application of loads.
2.
Section 9 of IEEE Std 344-1987 recognizes the use of experience data as a method for seismic qualification.
This method of qualification should be appropriately justified so it can be evaluated by the NRC staff on a case-by-case basis.
D.
IMPLEMENTATION The purpose of this section is to provide information to applicants and licensees regarding the NRC staff's plans for using this regulatory guide.
Except in those cases in which the applicant or licensee proposes an acceptable alternative method for complying with specified portions of the Commission's regulations, the methods described herein will be used in the evaluation of seismic r **11fication of electric and mechanical equipment for nuclear power plants as follows:
1.
Plarits for which the construction permit is issued after the issue date of the final guide, 4
2.
Plants fcc. which the operating license application is docketed for 6 months or more after the issue date of the final guide, 3.
Plants for which the applicant or licensee voluntarily cornmits to the provisions of this guide.
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[BRAFT] VALUE/ IMPACT STATEMENT BACKGROUND IEEE Std 344-1975, "Recommended Practice for Seismic Qualification of Class IE Equipment for Nuclear Power Generating Stations," was approved by the IEEE in January 1975.
In August 1977, the NRC staff issued Revision 1 to Regulatory Guide 1.100, which endorsed IEEE Std 344-1975, subject to four excep-tions.
Since then the staff has worked with the IEEE in developing IEEE Std 344-1987.
As a result of these efforts, the. exceptions to IEEE Std 344-1975 have been satisfactorily resolved.
IEEE Std 344-1987 also addresses several recent technical issues, for example, treatment of hydrodynamic loads, the limits of generic testing, the treatment of rattling, methods of qualifying line-mounted devices, and the use of actual seismic experience data bases to qualify identical or similar equip-ment.
IEEE Std 344-1987 thus reflects the state-of-the-art technology.
Issuance of this Proposed Revision 2 is consistent with the NRC policy of evaluating the latest versions of national standards in terms of their suit-ability for endorsement by regulatory guides.
SUBSTANTIVE CHANGES IEEE Std 344-1987 applies to seismic and dynamic qualification of Class IE (safety-related electric) equipment.
The nuclear industry has used this stan-dard for seismic qualification of mechanical equipment as well.
The NRC staff recognizes this fact and intends to extend the application of this standard to seismic qualification of mechanical equipment by this regulatory guide.
Spe-cifically, this regulatory guide covers two categories of equipment:
(1) safety-i related electric (Class 1E) equipment and safety-related mechanical equipment, and (2) non safety-related equipment whose failure can prevent the satisfactory accomplishment cf safety functions.
Regulatory Position C.1 provides guidance for qualification of mechanical equipment that is consistent with current NRC practice.
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Regulatory Position C.2 recognizes the use of experier a data as a method of seismic qualification, ibis method should be appropriately justified and will be evaluated by the NRC staff on a case-by-case basis.
Regulatory Positions C.1 to C.4 in Revision 1 are not included in this Proposed Revision 2 because they have been incorporated in IEEE Std 344-1987 as follows:
Regulatory Position in IEEE Std 344-1987 Rev. 1 of this Guide Section Number
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C.1 6.3 C. 2 7.6.2.1 C.3 7.6.2.5 C.4 10.3.2(6)
VALUE This guide endorses the latest version of a national standard and reflects the current state-of-the art technology.
The guide should also enhance the licensing process.
IMPACT Although the scope of this revision has been extended to include seismic qualification of mechanical equipment, the requirements are consistent with NRC current licensing practice.
Thus, this regulatory guide does not impose any new requirements or costs on licensees or applicants.
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