ML20198E306

Summarizes 921110 NRC-NUMARC Meeting in Rockville,Md,To Discuss first-of-a-kind Engineering Planned for Alwr Plants

ML20198E306
Person / Time
Site:	05200001
Issue date:	11/19/1992
From:	Bagchi G Office of Nuclear Reactor Regulation
To:	Richardson J Office of Nuclear Reactor Regulation
References
NUDOCS 9212070012
Download: ML20198E306 (15)

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s ,i NUCLEAR REGULATORY COMMISSION WASHINoToN. D.C. 20565 1 ...e AIGV 19 gg MEMORANDUM FOR: James E. Richardson,' Director Division of Engineering Office of Nuclear Reactor Regulation FROM: Goutam Bagchi, Chief Civil Engineering and Geosetences Branch Division of Engineering

SUMMARY

OF PUBLIC MEETING ON FIRST-0F-A-KIND ENGINEERING WITH NUMARC On November 10, 1992, the NRC staff met with the Nuclear Management Resources Council (NUMARC) and its representatives, in the NRC offices in Rockville, Maryland to discuss the first-of-a-kind engineering (F0AKE) planned for standardized advanced light water reactor (ALWR)~ plants. NUMARC described the formation of the Advanced Reactor Corporation (ARC) - a dedicated project organization that is being established to manage the implementation of the Nuclear Power Oversight Committee (NP0C) Strategic Plan. The F0AKE is one of the core technical blocks of the NP0C plan. The purposes of the meeting were to provide an information exchange on the F0AKE approach and to discuss the potential issues arising under the F0AKE process that might need to be considered early. The two specific issues selected for discussion with the NRC staff for this meeting were (1) piping design improvements, and (2) seismic equipment qualification. A list of meeting attendees is provided-in Enclosure 1. The slides presented by the NUMARC and its representatives are provided-in Enclosure 2.

NUMARC presented an overview of the F0AKE plan and schedule. The goals of the F0AKE program are to complete the engineering of certified designs of standard ALWR plants in sufficient detail to' define the cost estimates and prepare for construction as well as to define the process to achieve commercial standardization. The program is divided into-tnree phases: In Phase 1,-the ARC will define the F0AKE scope and perform a feasibility study. In-Phase 2, the ARC will evaluate and select the. design (s). In' Phase 3 the design and engineering will be completed for the selected design (s). The F0AKE: program is currently in Phase 1. A report on=the der.onstration of feasibility of the F0AKE process is expected to be completed by December 1992. NUMARC committed to provide-the NRC staff a copy of the report when it becomes available.

NUMARC discussed the need for piping design improvements --in ALWR ' plants since it believer that the ASME Boiler and Pressure Vessel Code (ASME Code),Section III design requirements for piping might be overly conservative and do not reflect the actual failure mode of piping systems under seismic loadings.

NUMARC also stated that the NRC regulations plus the current ASME Code

- requirements cause unnecessarily'high costs for piping design both in design and construction and in operation and maintenance.

9212070012 921119 i PDR ADOCK-05200001  ;^

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9 James E. Richardson WW 3 9 g The NRC staff noted that several changes to the NRC regulations and guidelines for piping design have already_ been proposed for implementation'in the ALWR lead plant piping design. These changes include the elimination of the-operating basis earthquake from design and_an increase in the functional capability stress limits for piping based on the results of dynamic testing performed by the EPRI and NRC in the mid-1980's. The NRC staff believes that these changes, in conjunction with other recent changes in piping design criteria (e.g., higher damping values), provide substantial relief from the

" overly-conservative" assumptions used in the design methods employed in currently operating nuclear plants while maintaining a sufficient safety margin. A copy of the NRC staff's proposed criteria for the ALWR lead plant is provided in Enclosure 3. The staff requested that the NUMARC assess the reasonableness of the total package of piping design criteria proposed for the ALWR lead plant and its impact on piping system design before it proposes to reduce the safety margin further. NUMARC committed to conduct such an assessment and provide the results to the NRC staff.

The NRC staff further pointed out that the piping methods used in the.F0AKE must be consistent with the methods described in the standard plant design certification. If changes to the ASME Code alping criteria are proposed for F0AKE, then it might be a long time before tiose changes are finally adoptea by the Code and endorsed by the NRC. The time factor could be a detriment in the incorporation of-the proposed changes into the design certification process for the ALWR evolutionary plants.

NUMARC discussed -its proposed approach for the seismic qualifiation of safety-related equipment. It intends to develop guidelines for equipment seismic qualification that would encompass all qualification methodologies (i.e., analyses, tests, and experience). The proposed approach for equipment seismic qualification would use traditional seismic qualification methods on equipment types with the highest seismic uncertainty but allow for insights gained from actual earthquake experience. With respect to an experience-based approach, NUMARC noted that there is a need for more. definitive guidance on how this_ approach may be used on a case-by-case basis.

The NRC staff'noted that the proposed approach might not be a viable approach for F0AKE at this time based on the information currently submitted by the ALWR vendors for design certification. The staff stated that it would re-review the design certification-applications to assess whether the description of the equipment seismic qualification methods would encompass such an approach. The NUMARC also committed to discuss this aparoach with the ALWR vendors to confirm whether they believe such an approac1 is encompassed by their submittal.

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James E. Richardson gy 19 jg The fiUMARC concluded that it plans to consider the NRC staff's comments and proposed to discuss further with the NRC staff any improvements in its F0AKE approach in mid-January 1993.

L h-@N8 h Goutam Bagchi, Chief Civil Engineering and Geosciences Branch Division of Engineering

Enclosures:

as stated cc: T. Murley Meeting Attendees 1

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Enclosure 1 ATTENDEES OF NOVEMBER 10, 1992 NUMARC/NRC MEETING ON F0AKE NAME AFFIllATION PHONE #

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William T. Russell NRC/NRR/ADT 301-504-1274 James' Richardson NRC/DE 301-504-2722 B.D. Liaw NRC/NRR/DE 301-504-3298 David Terao NRC/NRR/DE/ECGB 301-504-3317 Larry Shao NRC/RES/DE 301-492-3800 Armand Langmo Bechtel Power 301-417-4815 Goutam Bagchi hRC/NRR/DE/ECGB -301-504-2733

Y.C. (Renee) Li NRC/NRR/DE/EMEB 301-504-2772  ;

301-492-3894 i John O'Brien. NRC/RES/SSEB 4 Jerry Wilson NRC/ADAR

John Chambers GE Nuclear Energy 301-770-9650 4 David Stellfox McGraw-Hill 202-383-2162 i Andrew Murphy NRC/RES 301-492-3860 t

!. Charles Brinkman ABB/CE 301-881-7040 Russ Bell NUMARC .

202-872-1280

, Gary Vine EPRI (DC Office) 202-293-6347 i

James Norberg NRC/NRR/DE/EMEB _301-504-3286 Dale Thatcher NRC/NRR/DE/EELB 301-504-3260 i C.W. Fay Advanced Reactor Corp. 414-377-5788 Don Landers Teledyne Engineering Services 617-932-9000

John Knox NRC/NRR/DE/EELB 301-504-2763_

j Robert Rothman NRC/NRR/DE/ECGB 301-504-3306

R. Borchardt NRC/PDST 301-504-1193 Robert Pierson NRC/PDST 301-504-1118

! Tom Boyce _NRR/ADAR/PDST 301-504-1130 i Robert P. Kassawara EPRI 415-855-2775 William. R. Schmidt MPR Associates .703-519-0200 5

Ray Ng NUMARC '202-872-1280

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i David-L. Rehn Sam W. Tagart, Jr.

Duke Power Company EPRI 704-382 4685 415-855-2793- 1 l

4 Joseph Santucci EPRI/ ARC 415-812-8103 3

Terence L. Chan NRC/NRR/DE/EMEB 301-504-2169 l

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. ARC i

ADVANCED REACTOR CORPORATION i

j First-Of A-Kind Engineering Design Independent Activities ,

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l Presentation to NRC i

[ November 10,1992 hMC 6mg New to. itsi pg i i

ARC

Meeting Agenda

!. . Introductons and meeting objectwes (D. Rehn) i

. Task plan schedule overview (J. Santucci):

• ASME/ Piping Design Criteria -

- Project Overview (D. Rehn) i iTechnical Approach (D Landers) _

. Seismic Equipment Qualification j _ Project Overview (D. Rehn)-

._ - Technical Approach (W. Schmidt)

Resolution of Generic Technical Areos (D.'Rehn) -

• Summary (D. Rehn)

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arc Background and Introduction D.Rehn Nuclear Power Oversight Committee Strategic Plan For New Nuclear Power Plants

• Primary Objective Be able to order an ALWR with confidence in the mid 1990's.

Represents the technical and instrtutional enabling conditions necessary to permit utilrties to order, construct and begin operation of en ALWR on or about the turn of the century,

• A living document that is reviewed on an annual basis to make

- appropriate changes, assess progress and identify critical conditions and actions necessary for successful completior.

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4 ARC First-Of A-Kind Engineering Program Goals Complete engineering on certified designs in sutticient detad to define firm cost estimates and prepare for construction of ALW9 plants, and to support commercial standardization

. Ensure that an institutional infrastructure is in place to provide resources and manage completion of detailed designs

. Define the process to E:hieve commercial standardization

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ARC NPOC Strategic Plan / Core Technical Blocks Regulatory Stabilization . Utility Requirements NRC Design Certification Site Identification First Of A-Kind Engineering Enhanced Standardization Beyond Design 1

ARC implementation Approach

- + Phase 1 Define the FOAK Engineering scope and gather information Perform design independent support activities

) . Phase 2 Evaluate and select designs l

Phase 3

! - Complete FOAK Engineering for selected designs i

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. Designindependent Activities 4

) Develop products that can result in design simpidication and reducton of FOAK Engineering costs and plant ide<ycle costs

• Piping Design improvements

. Seismic Equipment Qualification

+ information Management System

. Commercial Grade Procurement Guidelines .

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e arc Task Plan Schedule J. Santucci o w 0. ., .

j arc Schedule Stan End Phase 1:

Develop Information 3/92 6/92 Genenc Desgn Tasks 3/92 8/93 Phase 2: 6/92 2/93 Develop Cnteria Issue Requests for Proposal Form Utihty Sponsor Groups Negotutte Contracts -

- Phase 3: 3/93 9/96 tem. Meg P.v 10.1#2 pg to

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! ARC l Phase 1 Activities

Tasks A E b

. Task A: FOAKE Scope Definnion (Complete)

. Task B

FOAKE Design Standardization (In-Progress)

! . Task C: Design Uncertainties (Complete)

. Task D: Selection Process (Complete)

. Task E-1

Piping Design Critena (in Progress)

• Task E 2: Seismic Equipment Qualifcation (in Progress)

. Task E 3:IMS (in Progress) l

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Remaining Phase 1 Schedule

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. Task E.1

- Design-by-rule work plan 12/92

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, Margins evaluation 12/92

! Design enteria 2/93

. . Task E.2 Phase I Demonstration of Feasibikty 12/92

! Phase ll - Production 8/93 (Final Guidelines for Design & Procurement)

. Task E.3 1 - Neutral File Specification 1/93

- Product Demonstration Specification - 3/93 l - Project Completion 4/93 l

• Tentative k

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ARC ASME Piping Design Criteria Project Overview D.Rehn NRC lag Nov 10.1962 pg 13 l

l ARC Philosophy ALWR Piping should be:

Reliable and safe -

• Meet NRC regulations

. Meet ASME code.

NRC net her 10. I9b2 pg 64 1

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Current Nuclear Piping Design Requirements Are Not Cost-Effective Because:

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. ASME seismic piping requirements are out of date, unrealistic, overty conservative, and not reflective of current technical knowledge

. NRC regulations plus current ASME code drives nuclear industry to produce unnecessanly costly piping designs (both capital and O&M) i l ARC a

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- Project Objective i.- To f acilitate establishment of appropriate changes in l ASME Code erneria and design practice to improve safety and reliability,' and to reduce cost and construction -

time for ALWR piping and support systems.

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l ASME/ Piping Design Criteria Technical Approach D. Landers i

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Pressure Vessel Research Committee j Technical and Steering Committee on i Piping Systems i

Formed task groups starting in 1982

• Welding Research Council Bulletins 300,316 addressed:

- Industry practice

! - Damping values i -

Spectral broadening _

. Piping installation tolerances

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a Dynamic stress allowable task group considered:

- Modifying allowable code stress Recategorizing the loading Insufficient experimental, analytical or experience 4

~ data were available to reach conclusions i

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i ARC EPRl/NRC Sponsored Piping and Fitting Dynamic Reliability Program

+ Developed in response to dynamic stress allowable task group I Conclusons

. Principal objective: Develop realistic and defensible set of design rules

!- to evaluate dynamic loading for nuclear piping

- Program demonstrated the pipe failure mode due to earthquake type loading

- Fatigue or combinaton of fatigue and ratcheting l

. Current code rules protect against a collapse f ailure mode

• Presented proposed code changes to Section 111 subcommittee in 1989 i

j ARC Section 111 Review of Proposed Code Changes j

• Formed specialtask group under working group on piping design

- Reviewed entire EPRl/NRC sponsored program contained in i

five volumes j - Meetings held over a two-year period l - Did not bring forward a recommendation

. . Subgroup on design then formed a special task group f - Responsible for establishing a basis for evaluating any proposed changes to piping dynamic criteria

- Oticipate ecmpletion of efforts early 1993 MAC tag toev 10.1882 39 M i

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ARC Technical Core Group (TCG)

. Formed in mid 1992 by ARC

. Composed of following membership

- Don Landers, Teledyne Chairman

- Ernie Branch, Sament & Lundy

- John Stevenson, Stevenson & Associates

- Sam Tagart, EPRI

- Peter Yanov, EOE Debbie Ramsey, Duke Power - Coordinator

. Objectives

- Develop technical framework and strategic plan Enhance approval of new piping and pipe support rules

- Optimize piping system design

- Integrate results and prepare design specification Technical framework and strategic plan

. Collect and review relevant data Priornize techncal issues Support industry interf ace with NRC and code bodies Technical Core Group (Continued)

Enhance approval of new piping and pipe support rules Margin comparison of piping criteria Evaluate margin on seismic loading Correlate world wide research on piping

- Review and evaluate earthquake experience of piping Perform technical tasks required to resolve concems Integrate results and prepare design specification

~ Prepare recommendations for code changes Interface with ASME code bodies to explain recommendations Develop a technical specification for optimum piping system design l-

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TCG - Major Findings to Date i

4 . Piping Failure Mode

- Fatigue or combination of fatigue and ratcheting l

- Not collapso

+ Earthquake experience supports expenmental and analytical data i

• Collapse type f ailures have not occurred, and we understand why t

l = Current level D (SSE) intended minimum margin on collapse

fallure: 1.3 to t.5 t

l . Proposed TCG minimum margin on iatigue/ratcheting: 2.0 to 2.5

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i-I ARC Major TCG interface and i Code Approval Process

• Major interface I

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{= - Subgroups Subgroup on Design Subgroups

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> a Code Approval Process I -- Recommendations must satisfy Wichman task group criteria 3

- Subgroup design

- Subcommittee 111

- - Main Committee ASME Boller and pressure vessel committee -

- Board on Nuclear Codes and Standards j - Anticipate TCG interface, as required, at all levels in process t

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arc Seismic Equipment Qualification Project Overview-D.Rehn arc Project Scope Develop guidelines for seismic equipment qualification wnich:

9 encompass all applicable qualification methodologies Incorporate recent technology improvements gained through seismic experience provide guidance to equipment designers and procurers

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ARC Project Objectives

. Focus traditional SEQ methods on equipment types with highest seismic uncertainty

. Incorporate lessons learned f rom actual earthquake experience

. Improve cost benefit of SEO process for ALWR plants

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. ARC Topics to be Discussed 4

Project Overview

. Technical Approach

. Project Technicalinterf aces

> - Preliminary Findings i

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l Background Requirements for Seismic Equipment Quahfication (SEO) addressed in:

- -10CFR100, Appendix A ,

IEEE 344 1 - Reg. Guide 1.100 SRP Standards provide general guidance and several methods l - Test Analysis Experience (case basis) j Combinations

. There is need for more definitive guidance on application

.of experience-based approach 4

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Background

(Continued) i

. Experience based SEQ Methodology / Data provide potential advantages:

j- - lessons learned from strong earthquakes and tests I - known seismic sensitivities addressed explicitly

' - focus test / analysis SEO effort in areas of highest

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cost effectiveness I

. Not applicable to all equipment classes

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ARC Extensive Data Base Exists on Equipment Performance in Earthquakes and Tests l

- 20 classes of power plant equipment

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Wide diversity of type _s, models, vintages

. Certain classes show consistent high ruggedness

. Specific vulnerabilities have been identified

. Equipment seismic capacity levels and inclusion rules, exclusions, caveats have been developed .

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Purpose Of This Project i

Develop SEO Guidelines which consider:

• Seismic ruggedness 4ragility of equipment classes

. Anticipated design variabilrty/ standardization of ALWR equipment classes

. Limrtations of SEO methods

. Experience from earthquakes and prior testing

. Safety and cost benefits

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Application Of Experience-Based Qualification To ALWR Utiltty Requirements Document (URD) states intention to use experience data where appropriate and justified (Chapter 1, Section 4.8.1)

NSSS vendors have indicated their intent to use the completed methodology -

NRC SER on URD limits experience based SEQ to case-by-case

. ' Guidelines for use of experience data will be refined, refocused for ALWR equipment. Will augment, not replace traditional approaches

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Technical Approach 1

• Estabhsh applicability of experience data to 1 ALWR equipment

- equipment vintage -

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- equipment design

- expanded / extended database i

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• Categonze equipment classes i- inherently rugged, standard power plant equipment -

j Inherently or intermediately rugged; . design variability, I new features, sensitive subcomponents i

- high degree of design variability / complexity, possible seismic sensitivities

' - Develop guidelines for SEQ for each category

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4 arc Main Technicalissues to be Addressed in ALWR SEQ Project l

. Are ALWR equipment types represented in experience data?

. Does experience data base contain modern vintage equipment?

• Are ALWR teismic demand levels compatible with experience data? (ALWR SSE is 0.3g R. G.1.60)

• ls experience-based SEQ coet-effective when ALWR seismic and equipment standardization requirements are considered?

1 ARC Current Effort includes Demonstration of SEQ Principle

• Four equipment classes selected for demonstration:

- transformer battery horizontal pump

- motor operated valve Determine for these classes

- seismic capacity demonstrated by experience data related to ALWR demand applicability of experience data to ALWR equipment inclusion rules, specification attributes J

Review Groups Utility Advisory Panel (UAP) USNRC N. P. Srnnh Comm. Ed. Internal J. Thomas Duke J. O'Brien RES H. Hanneman Wisconsin Electric D. Terao NRR D. Moore So, Co. Services P Y.Chen NRR J. Reynolds Southern Nuclear T.Akos GPUN Contractor R.P. Kennedy A. Schiff D.Kana K.Bandyopadhyay

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ARC Preliminary Findings

. ALWR equipment representation significant traction of ALWR equipment .

represented in experience data base

- Equipment Vintage

- current experience data base contains modem vintage equipment (19801988)

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! Preliminary Findings (Continued) t 1

- Experience-based seismic capacity vs. ALWR seismic demand

. preliminary ALWR demand spectra under evaluation

! realistic ALWR seismic demand levels compatible with i experience-based capacrties for lower plant elevations

- ALWR arrangements place much equipment at low elevatons

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l Generic Technical Areas -

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{ Resolution Approach

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[ FOAKE Generic Technical Areas Resolution Approach

• Design Certification applications and reviews proceed

• NRC interactions on advanced FOAKE design / engineering analyses methods

- Piping design criteria

' Seismic equipment qualifcation 4

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. ,I, NUCLEAR HEGULATORY COMMISSION l W AMINoToN. D.C. NEE 5

• • • • • September 11, 1992  :

i Docket No. ',F . l Mr. Patrick W. Harriott, Manager '

Licensing & Consulting Services GE Nuclear Energy  ;

175 Curtner Avenue San Jose. California 95125 [

SUBJECT:

STAFF GUIDANCE FOR THE 'jSE OF A SINGLE-EARTHQUAKE DESIGN FOR SYSTEMS, STRUCTURES,-AND COMPONENTS IN THE ADVANCED BOILING WATER '

REACTOR (ABWR)

Recently, GE Nuclear Energy (GE) indicated that it is planning to take advantage of the option to decou) 1e the level of the operating basis earth-quaka (OBE) ground motion from tsat of the safe-shutdown eartl5 quake (SSE) in its design of systems, structures Inand components (SSCs in the A8WR. In a meating held on August 26, 1992,- San Jose. Californi),

a with GE staff and repre:rttatives frem the Electric Power Research Institute, the staff provided-preliminary guidance to GE concerning what types of analyses and information would be required in the Standard Safety Analysis Report (SSAR) for the staff >

l to approve design of SSCs without the OBE.

During the meeting several-recommendations were-made by the participants.- The-staff has evaluated them and has incorporated-them into the enclosed guidance document to be used oy GE to conduct the set of required analyses, and to amend the SSAR to reflect decoupling of the DBE. Please note that it has been

.- prepared in safety evaluation report (SER) format for ultimate use in the-l preparation of the final SER for the ABWR on this subject.

It is expected tht GE will boilin its work on this design basis chanfle and will be prepared to discuss thts subject at the next management meet'ng I scheduled for the week of October 5.1992. If you have any immediate ques -

tions, please contact Chet.Posiusny on (301) 504-1132.

. Sincerely,3 2

h (. N Robert C. Pierson, Director Standardization Project Directorate Associate Directorate for Advanced Reactor:

and License Ranewal

• Office of Nuclear Reactor Regu h tion i

'Enchsure:-

As stated-cc w/ enclosure:

See.<. ext page hE0$5'.lOS$

I SAFETY EVALUATION ON THE USE OF A  !

SINGLE-EARTHQUAKE DESIGN FOR '

SYSTEMS, STRUCTURES, AND COMPONENTS IN THE ABWR STANDARD PLANT i

A. INTRODUCTION Appendix A to 10 CFR Pa-t 100 requires, in part, t% t all structures, syston ,

and components of the nuclear power plant necessary for continued operation without undue risk to the health and safety of the public shall be designed to remain functional and within appilcable stress and deformation limits when subject to an operating basis earthquake (0BE). Changes to Appendix A to Part-100 are bel g proposed to redefine the OBE to a level such that the function of the OBE can be satisfied without the need to perform explicit responses analyses.

The Nrpose of this safety evaluation'is to identify the necessary chwges to existing seismic design criteria that are acceptable to the NRC staff for implementing the proposed rule enan a e as it pertains to the design of s:;fety-related systems, structures, and components in the General Electric Advanced Boiling Water Reactor (ABWR). These criteria apply only to the ABWR standard plant design and are not intended to replace the seismic design criteria approved by the Commission in the licensing bases of currently operating facilities. The guidelines provided herein are proposed for use as a pilot program for implementing the proposed rule change specifically for the ABWR.

L. BACKGROUND In SECY-90-016. " Evolutionary Light Water Reactor (LWR) Certification Issues and Their Relationship to Current Regulatory Re p rements,? the staff requested the Commission's approval to decouple the level of the OBE groand  ;

motion from that of the safe-shutdown earthquake (SSE). The Commission approved the staff;s position in its Staff Requirements hemorandum (SRM) of June 26, 1990.

In the draft Commission paper, " Issues Pertaining to Evol'utionary and Passin Light Water Reactors and Their Relationship to Current Regulatory Requirements," the staff further requested the Commission-to approve eliminating the OBE from the design of systems,_ structures, and components in both evolutionary and passive advanced reactors designs. The proposed amendment tc 10 CFR Part 100, Appendix A would . allow, as an option, that the OBE be eliminated from design certification when the OBE is established at less than or equal tc one-third the SSE. In this manner, the 0BE serves the function as an inspection level earthquake:below which the effect on the health and safety of the public would be insignificant and above which the licensee would be_ required to shut down the plant and insped for damage. The elimination of the OBE frvm design was requested by the Electric-Power:

Research Institute (EPRI).and also recommended by the Advisory Committee on

! Reactor Safety (ACRS) in its_ letter of' April 26, 1990.

In the draft Commission paper, " Design Certification and Licensing Policy

. -9 f<09 M o oL 2 -

Issues Pertaining to Passive and Evolutionary Advanced Light Water Reactor Designs," the staff examined the safety impact of eliminating the OBE as it pertains to civil structures, piping systems, and equi) ment seismic qualification. Several recommendations were made by tie staff to ensure that eliminating the OBE would not result in a significant decrease in the overall l plant safety margin. The following sections of this safety evaluation contain the specific actions needed for the ABWR st adard plant design to ensure that adequate safety margins are maintained when the OBE is eliminated from the design. The sections identify those actions needed for: (1) piping systems, j2) concrete and steel structures, (3) equipment seismic qualification, and s4) pre-earthquake planning and post-earthquake operator actions.

C. ASME CODE CLASS 1, 2, AND 3 COMPONENTS AND CORE SUPPORT STRUCTVRES The dynamic analysis methods to be used for saismic analyses of ASME Code Class 1, 2, and 3 components and core support structures in the ABWR shall use those methods described in the ABWR SSAR as approved by the NRC staff in its Final Safety Evaluation Report (FSER). The loads and load combinations to be used for evaluating ASME Code Class 1, 2, and 3 components and core support structures are provided in the ABWR SSAR and discussed in the staff'r FSER.

The OBE may be eliminated from the applicable design load combination when the following supplemental c:lteria are used.

1. Fation l In order to ensure adequate design considerations for the fatigue effects of i earthquake cycles, it is necessary to establish a bounding load definition and

) number of earthquake cycles to account for the more frequent occurrences of

!. lesser earthquakes and their aftershocks. For the ABWR, an acceptable cyclic L load basis.for fatigue analysis of earthquake loading for AWiE Code Class 1, j 2, and 3 components and core support structures is two SSE events with 10

maximum itress cycles per event (20 full cycles of the maximum SSE stress range). his is equivalent to the cyclic load basis of one SSE and five OBE i

events as currently recommended in tne Standard Review Plan (NUREG-0800)

. Section 3.9.2. - Alternatively, an equivalent number of bactional vibratory cycles to that of 20 full SSE vibratory cycles may be used (but with an-I amplitude not less than one-third of the maximum SSE amplitude) when derived j in accordance with Appendix 0 of IEEE Stendard 344-1987.

2. Seismic Anchor Motion (SAM)

) For the ABWR, the effects of displacement-limited, seismic anchor motions

- (SAM) due to-a safe-shutdown earthquake should be evaluated for safety-related

ASME Code Class 1, 2,'and 3 components and component supports to ensure their

functionality during and following an SSE. The-SAM effects should include

(but are not limited to) relative displacements of piping between building i floors and slabs, at equipment nozzles, at piping penetrations, and at connections of small-diameter piping to large-diaceter. piping.

4

For piping systems, the effects of seismic anchor motions due to a safe-l- shutdown earthquake-should be combined with the effects of other normal 2

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5 I l I operational loadings that might occur concurrently as specified in Section i C.3.1 and C.3.2 of this safety evalu: tion.  ;

1 j 3. Pipino Stress Limits

! For ASME Code Class 1, 2, and 3 piping, the design requirements in the 1989 Edition of the ASME Boiler and Pressure Vessel Code,Section III, Subsections j NB, NC, and ND shall be met. In addition, the following changes and additions 4 to paragraphs NB-3650, NC-3650, and ND-3650 are necessary and shall be 1 satisfied for piping systems when the OBE is eliminated from the design. l 4 3.1 ASME Code Class 1 Piping Stress Li!siis

(a) For primary stress evaluation (NB-3654.2), earthquake loads are not

{_ required to be evaluated for consideration of Level B Service Limits for

Eq.(9).

L i (b) For satisfaction of primary plus secedary stress intensity range (NB-3653.1), in Eq. (10), M the resultant range of all loads considering one-half- thb shall be either (' range of the safe-shutdown earthquake o i resultant range of moment due to the full range of the safe-shutdown ,

! earthquake alone, whichever is greater. The use of the safe-shutdown

{ earthquake is intended to provide a bounding design for the cumulative effects of earthquakes of a lesser magnitude and is therefore to be included in

! consideration -of Level B Service limits for Eq.(10). A reduced range (with an equivalent number of fractional vibratory peak cycles) of the safe-shutdown

- earthquake moment may be used for-consideration of Level B Service Limits (but with a range not less than one-third of the maximum SSE moment range).

. (c) For satisfaction of peak stress intensity (NB-3653.2), the load sets

develnped in NR-3653.1 based on the above Position C.3.1(b) should be used in

cakulating the peak stress intensity, S , and the alternating stress l intensity, Salt, for evaluating the fatihue effects and cumulative damage.

l (d) for simplified elastic-plastic discontinuity analysis _ (NB-3653.6), if

. Eq. (10) cannot be satisfied for all pairs of load sets, then the alternative

! analysis as de.'.ribed in NB-3653.6 should be followed. In addition, the j following condition shall be satisfied:

l-D 3 . ,,

7T (Mj + Mg _) 1 6.0 S , _Eq. (12a)

Ssam " T2 where: S i s the nominal value of seismic anchor motion stress l_

y sam Hj i s the same as H g in Eq. (12)

M,*is the same as M j in Eq. (10), except that it includes ~

! o ly moments due to seismic anchor motion. displacements 3

i 1-S Y

e r,r--- , cy--,- - , y e-,-ve, c, #-, =e +~w-- ,w, ,.,-y..,,-,- gn,--.,-m y - ,, sy , .h,-w we se e ,w y ,* er-=,--e=wy w-w.vwe v - vem , v

I i

y caused by a safe-shutdown earthquake 1

i l The combined moment range (M4+M4 ) shall be either (1) the resultant  !

i range of thermal expansion and thermal anchor movements plus one-half '

i the range of the safe-shutdown earthquake anchor motion or (2) the resultant range of moment due to the full range of the safe-shutdown '

j earthquake anchor motion alone, whichever is greater.

3.2 ASME Code Class 2 and 3 Piping Stress Limits s

I (a) For consideration of occasional loads. (NC/ND-3653.1), earthquake loads l (i.e., inertia and seismic anchor motion) are not required for satisfying l

Level B Service Limits for Eq.(9).

(b) For consideration of thermal expansion or secondary stress'es (NC/ND- ,

- 3653.2),M in Eq.-(10) is not required to include the moment effects of j

seismican6hormotionsduetoanearthquake.

(c) For consideration of secondary stresses in Level D Service Limit (NC/ND-3655), the following condition ~should be satisfied:

l

• I Mc+M c Eq. (10b) i S 3

- i g - s 3.0 Sh

! where: M is the range of moments due to seismic anchor motions

d6etoasafe-shutdownearthquake f Mc is the range of moments due to thermal expansion

! 4. Pipe Break Postulation Without OBE i It-is recognized that-pipe rupture is:a rare event which might only occur i

under unanticipated conditions, such as those which might be caused by-2 possible design, construction, or operational errors; unanticipated loads or

unanticipated corrosive environments. The staff's observation of actual piping failures have found that they generally occur. at high stress and i fatigue locations, such as at tha terminal ends of a piping system at its i connection to component nozzles. Currently, in accordance with Standard L Review Plan (NUREG-0800) Section 3.6.2, Revision 2 dated June 1987,_ pipe 3

breaks are postulated in high energy piping at locations of high stress and high fatigue usage factor. The load combination used in calculating the high stress and usage factor includes normal and upset load conditions (i.e.,

- pressure, weight, thermal, OBE, and other operational transient loadings),

{

j From a historical-viewpoint, the criteria for postulating high energy breaks -

at specified _ locations were first introduced in the early -1970;.- - The basis for the mechanistic approach for selecting pipe break locations wa> derived a

from the premise that although pipe breaks could result from random e,'ents

. induced by unanticipated conditions,:the failure, mechanism and the ex)ec'.ed j- location of failure would likely be caused by local conditions of hig1 stess 4

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s- re ,- w r-v y ,- w ws. e-e . + , ,. -,r---rw,+ e,+v-y-r.w. .4 ,+ e o w w~ , w e r ver- o ,, m , w ,,r. vc s., w iv g -r y. v v w -*-a--9 w---er--wor **ww--r-r* t'e

or high fatigue in the piping. In order to insure that a sufficient number of pipe breaks would be postulated, breaks were recommended to be postulated for a wide spectrum of events to envelope the uncertainties of unanticipated failure mechanisms. Breaks were postulated at terminal ends of the piping, at high stress and high fatigue locations, and as a minimum at two additional intermediate locations when the stresses were below the high stress threshold limit. The resulting criteria which were incorporated in Standard Review Plan Section 3.6.2 resulted in many postulated pipe break locations and caused the installation of numerous pipe rupture mitigation devices in nuclear plants, in the mid-1980s, the NRC's Exe:utive Director for Operations initiated a comprehensive review of nuclear power plant piping to identify areas where changes to the piping requirements could improve the licensing process as well as the safety and reliability of nuclear power plants. The NRC's Piping Review Committee (PRC) in an integrated effort with the nuclear industry under the Pressure Vessel Research Council conducted a comprehensive-study of alping criteria including the mechanistic pipe break postulation guidelines. Tie PRC found that when an excessive number of pipe rupture mitigation devices (i.e.,

pipe whip restraints and jet impingement shields) are installed on high energy piping systems, the potential exists for piping systems to be overly constrained. This condition was found in several nuclear plants in which massive pipe restraints adversely affected the ability of the high temperature piping to freely expand during normal plant operation. The PRC also found through numerous dynamic tests and field observations of non-seismically designed piping systems that had undergone high seismic loadings that butt-welded piping possesses an inherent ability to withstand large seismic inertial loadinas without failure.

As a result of the PRC's effort, the NRC staff recognized that the mechanistic pipe rupture criteria for selecting locations of pipe breaks resulted in an excessive number of pipe rupture mitigation devices that could hinder the t normal operation of the plant and might not contribute significantly to the l overall safety of the plant. Accordingly, the Standard Review Plan was revised to reduce the number of postulated pipe breaks by (1) eliminating the need to postulate pipe breaks at the two arbitrary intermediate-locations and (2) providing a leak-before-break approach in lieu of postulating pipe breaks when the system and material specific information is adequate to justify its application.

Based on recent dynamic pipe tests conducted by the EPRI and NRC, it has been demonstrated that the piping can withstand seismic inertial loadings higher than an SSE without rupturing. Thus, the staff believes the likelihood of a pipe break in a seismically-designed piping system due to an earthquake magnitude of one-third SSE is remote. Openting experience has shown that pipe breaks are more likely to occur under conditions caused by normal l

operation (e.g., erosion-corrosion, thermal constraint, fatigue, and operational transients).

On the basis of the above discussion, the staff concludes that no replacement l earthquake loading should be used to establish postulated pipe rupture l locations. Instead, the criteria for sostulating pipe breaks in scismically-designed, high energy piping systems siould be based on factors attributed to 5

normal and operational transients only. The staff's revised criteria for pipe break postulation are provided below. The revised criteria are intended to ensure that breaks are postulated to occur at the most likely locations and to reduce the number of pipe rupture mitigation devices (e.g., pipe whip restraints and jet impingement shields) that might hinder plant operation without providing a compensatory level of safety.

The elimination of earthquake loads in the revised pipe break criteria below is justified, in part, on the fact that the equipment environmental qualification and compartment pressurization analyses for the ABWR are based on a worst-case break assumption in each compartment and are not post 6 ated at mechanistic break locations, in addition, GE should commit to a monitoring progrt.m for erosion-corrosion that provides assurances that procedures or administrative controls are in place to assure that the NUMARC program (or another equally effective program) is implemented and the structural integrity of all high-energy (two-phase as well as single-phase) carbon-steel systems is maintained as discussed in Generic letter 89-08 and NUREG-1344,

" Erosion / Corrosion-Induced Pipe Wall Thinning in U.S. Nuclear Power Plants,"

April 1089.

Consistent with the above staff finding, the guidelines provided in SRP Section 3.6.2, Branch Technical Position MEB 3-1, " Postulated Rupture Locations in Fluid Systea Piping Inside and Outside Containment," may be revised as follows:

B.I.b.(1).(a): Footnote 2 should read, "For thosr loads and conditions in which Level A and Level B stress limits have been specified in the Design i specification (excluding earthquake loads)."

1 4

B.I.b.(1).(d): "The maximum stress as calculated by the sum of Eqs. (9) and (10) in Paragraph NC-3652, ASME Code,Section III, considering those loads und conditions thereof for which level A and level B stress limits have been j specified in the system's Design Specification (i.e., sustained loads,

occasional loads, and thermal expansion) excluding earthquake loads should not exceed 0.8(1.8 Sh + 3 A )'"

(

I D. CONCRETE AND STEEL STRUCTURES

1. SSE Relative Disolacements Between Structures As discussed in Appendix 3G to the SSAR, the seismic response (building i displacements, structural member forces, floor response spectra, etc.) of the reactor building could be significantly underestimated, when the through-soil,

. structure-to-structure' interaction effect is not considered. GE did not

! consider this effect in the anclyses of certain ABWR structures such as the control building, ultimate heat sink pump house, radwaste building, and turbine building. This effect might be more pronounced for these other buildings because they are lighter than the reactor building and the energy feedback frcn the reactor building during an earthquake could significantly affect the seismic responses of these buildings.

! s i

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Therefore, the staff concludes that the effects of through-soil, structure-to-structure interaction under SSE loadings for all structures housing seismically-designed piping should be determined under SSE loadings to i establish the relative displacements between buildings (seismic anchor movement for-piping systems). ,

2. Seismic Instrumentation  ;

j GE should ensure that adequate design provisions allow for the placement of l

seismic instrumentation in the free field so that the control room operator-can be immediately informed through the event indicators when the response i spectra level and the cumulative absolute velocity (CAV) experienced at this i location exceeds the shutdown level and can take the necessary actions. The j details of the instrumentation requirements are discussed in Section F of this

, safety evaluation.

i

3. Use of RG 1.143 and 1.27 j

The staff guidelines in RG 1.143, " Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water-cooled

Nuclear Power Plants," and in RG-1.27, " Ultimate Heat Sink for Nuclear Power ,

! Plants " provide for a seismic design of radwaste buildings and ultimate heat

! sink features based on the operating basis earthquake. .With the elimination

of the OBE, the staff finds that these structures and features should be designed to withstand the safe-shutdown earthquake. The structural design

, criteria using the SSE loading should use the appropriate loads and load l combinations provided in Standard Review Plan Section 3.8.4.

I l E. EQUIPMENT SEISMIC QUALIFICATION

! When equipment qualification for seismic loadings is performed by analysis, i testing, or a combination of both, the staff recommends a use of the IEEE l Standard 344-1987 as endorsed in Regulatory Guide 1.100, Revision-2. This-i standard has detailed requirements for performing seismic qualification using

! five OBE events followed by an SSE event. With the elimination of the OBE, it is necessary to qualify equipment with-the equivalent of five OBE events followed by one SSE event. Therefore, the staff concludes that equipment j should be qualified with five %SSE events followed by one full SSE event.

Alternatively, a number of fractional peak cycles equivalent to the maximum peak cycles for five %SSE events may be used in accordance with Appendix 0 of IEEE Standard 344-1987 when followed by one full SSE.

F. PRE-EARTHQUAKE PLANNING AND POST-EARTHQUAKE OPERATOR ACTICMS Th'e design certification of-the ABWR using a single-earthquake (SSE) desig is predicated on the adequacy of pre-earthquake planning and post-earthquake inspections for damage that are to be . implemented by the COL applicant.

The COL applicant shall submit to the NRC staff as a part of.its application

, .the procedures it plans to'use for pre-earthquake planning and post-earthquake actions. For the ABWR, the NRC staff finds acceptable the criteria developed 7

= .- _ - .- - - - . . . . . . _ - -,. - . -._ - . - - . , . . . - - - _ , -

by the Electric Power Research Institute (EPRI) in EPRI Reports EPRI NP-5930.

EPRI NP-6695, and EPRI 1R-100082 together with the amendments, additions, and changes outlined below for evaluating the need to shut down the plant following an earthquake.

IERI NP-5930 The EPRI Report NP-5930 shall be used with the following exceptions:

1. A free field instrument must be used for determining the CAV and the spectral acceleration level.

2. The response spectrum check is as follows:

The 5% damped ground response spectrum for the earthquake motion at the site exceeds (1) the corresponding OBE response spectral acceleration

! between 2 and 10 Hz, or it exceeds an acceleration of 0.20g between 2 it exceeds the corresponding OBE

] and response spectral velocity between 1 an d 2 Hz or a velocity of 6 inches 10 Hz whichever is greater, or (2) per second between 1 and 2 Hz, whichever is greater.

3. The licensee shall consider as sufficient evidence to shut down the plant the simultaneous exceedance of the 5% damped ground response spectrum enumerated in item 2 and the CAV exceedsnce of 0.16 g-sec for any one frequency on any one component of the free field ground motion. The CAV shall be determined in accordance with EPRI Report EPRI NP-100082. Also, any evidence of significant damage observed during the plant walkdown in accordance with the EPRI Report NP-6695 recommendations shall be sufficient cause for plant shutdown.

i 4. The instrumentation installed at the nuclear aower plant shall be

capable of on-line digital recording of all taree components of the

! ground motion and of converting the recorded (digital) signal into the j standardized CAV and the 5 percent damped response spectrum. The i

digitizing rate of the time history of the ground motions shall be at

! least 200 samples per second and the band-width shall be at least from l 0.20 Hz to 50 Hz. The pre-event mer.y of the instrument shall be sufficient to record the onset of tne earthquake, f' 5. The system must be capable of routinely calibrating the response spectrum check of 0.20g. Also, the CAV of 0.169-sec should be

! calibrated with a copy of the October, 1987 Whittier, California i

earthquake or an equivalent calibration record provided for this purpose

! by the manufacturer of the instrumentation. In the event that an actual

earthquake has been recorded at the plant site, the above calibration shall be performed to demonstrate that the system was functioning properly at the time of the earthquake.

l EPRI NP-6695 The EPRI Report NP-6695 shall be used with the following exceptions:

l 8

i

Section 3.1. Short-Term Actions Item 3. " Evaluation of Ground Motion Records" There is a time limitation of four hours within which the licensee shall determine if the shutdown criterion has been exceeded. After an earthquake has been recorded at the site, the licensee shall provide a response spectrum calibration record and CAV calibration record to demonstrate that the system was functioning properly.

Iten 4. " Decision on Shutdown" Exceedance of the EPRI criterion as amended by the NRC or observed evidence of significant damage as defined by EPRI NP-6695 shall constitute a condition for mandatory shutdown unless conditions prevent the licensee from accomplishing an orderly shutdown without jeopardizing the health and safety of the public.

Add item 7. " Documentation" The licensee shall record the chronology of events and control room problems while the earthquake evaluation is in progress.

Section 4.3. " Guidelines" (c. 4-32 Because earthquake-induced vibration of the reactor vessel could lead to changes in neutron fluxes a prompt check of the neutron flux monitoring instruments shall be made to indicate if the reactor is stable. Therefore, this check should be added to the checks listed in this section.

Section 4.3.4.1. " Safe Shutdown Eauioment" (c. 4-7):

In addition to the safe shutdown systems on this list containment integrity must be maintained folloviing an earthquake. Since the containment isolation valves may have malfunctioned during the earthquake, inspection of the containment-isolation system is necessary to assure continued containment integrity.

Section 4.3.4. " Pre-Shutdown Inspection" Exceeding the EPRI criterion or evidence of significant damage should constitute a condition for mandatory plant shutdown, as the staff stated in its recommendation for Section 3.1, item 4, " Decision on Shutdown."

G. CONCl.USIONS On the basis of the changes to the existing seismic design criteria discussed above, the staff concludes that eliminating the operating basis earthquake from the design of systems, structures, and components in the ABWR standard plant will not reduce the level of safety provided in current regulatory guidelines for seismic design. On the contrary, the staff finds that the changes provide an enhancement to safety by. refocusing current design requirements to emphasize those areas where failure nudes are more likely to 9

- _ _ - - - _ _ _ _ _ _ _ _ \

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h- occur and by precluding the need for seismic design requirements that do not significantly contribute to the overall safety of the plant.

J Contingent upon GE submitting a revision to its SSAR (including the appropriate changes to its Tier 1 Design Certification Material) reflecting

the above criteria, the staff concludes that the elimination of the OBE from

- the design of systems, structures, and components, in the ABWR standard plant j is acceptable.

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