ML17309A222

Forwards Draft Evaluation Re SEP Topics III-6,seismic Design Consideration & III-11,component Integrity. Evaluation Will Contribute to Basic Input to Integrated Safety Assessment

ML17309A222
Person / Time
Site:	Ginna
Issue date:	01/29/1982
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:	Maier J ROCHESTER GAS & ELECTRIC CORP.
Shared Package
ML17258A195	List: ML17258A194 ML17258A518 ML17258A520 ML17309A222
References
TASK-03-06, TASK-03-11, TASK-3-11, TASK-3-6, TASK-RR LS5-82-1-70, LSO5-82-01-070, LSO5-82-1-70, NUDOCS 8202090358
Download: ML17309A222 (70)

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4**~4 Docket No. 50-244 LS05 01-070 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 January 29, 1982 Mr. John E. Maier Vice President Electric and Steam Production Rochester Gas

& Electric Corp.

89 East Avenue Rochester, Hew York 14649

SUBJECT:

SEP SAFETY TOPICS III-6 SEISMIC DESIGH CONSIDERATION AHD III-ll,COMPONENT INTEGRITY - GINNA NUCLEAR POWER PLANT We have completed our seismic review of Ginna Nuclear Power Plant.

Enclosed is a copy of our draft combined evaluation report of the two subject topics.

As'discussed in this draft report, four items are required to be upgraded to meet SEP requirements for the postulated SSE:

(1) steel bracing at north-'ast corner of auxiliary building, (2) the support system of component cool-ing surge tank, (3) refueling water storage tank; and (4) essential service water pumps.

Six items still remain open due to lack of design information.

According to mutual agreement between the staff and your representative, the responses to these items are scheduled by January 31, 1 982.

A supplement to this report will be issued after the review of your responses for the six open items are completed.

.This evaluation will be a basic input to the integrated safety assessment for your facility unless you identify changes needed to reflect the as-built conditions at your facility.

With respect to the potential modifications outlined in the conclusion of this report, a determination of the need to actually implement these changes will be made during the same integrated as-sessment.

This topic assessment may be revised in the future if. your facil-ity design is changed or if NRC criteria relating to. this topic are modified before the integrated assessment is completed.

Your response is requested within 30 days of receipt of this letter. If no response is received within that time, we will assume that you have no comments or corrections.

Sincerely, Dennis M. Crutchfi eld, C

f Operating.Reactors Branc No.

5 Division of Licensing

Enclosure:

As stated cc w/enclosure:

See next page

Hr. John E. Haier CC Harry H. Yoigt, Esquire LeBoeuf, Lamb, Leiby and MacRae 1333 Hew Hampshire

Avenue, N.

W.

Suite 1100 Washington, D. C.

20036 Nr. Michael Slade 12 Trailwood Circle Rochester, New York 14618 Ezra Bialik Assistant Attorney General Environmental Protection Bureau Mew York State Department of Law 2 World Trade. Center New York, Hew York 10047 U. S. Environmental Protection Agency Region II Office ATTN:

Regional Radiation Representative 26 Federal Plaza Hew York New York 10007 Herbert Grossman, Esq. Chairman Atomic Safety and Licensing Board U. S. Huclear Regulatory Co~ission Washington, 'D. C.

20555 Resident Inspector R. E. Ginna Plant c/o U. S.

NRC 1503 Lake Road

Ontario, Hew York 14519 Director, Bureau of Nuclear Operations State of New York Energy Office Agency Building 2 Empire State Plaza Albany, New York 12223 Rochester Public Library 115 South Avenue Rochester, Hew York 14604 Supervisor of the Town of Ontario 107 Ridge Road West

Ontario, Hew York 14519 Dr. Emmeth A. Luebke Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.

20555 Dr. Richard F. Cole Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.

20555

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SEP SAFETY TOPIC EVALUATION GINNA NUCLEAR POWER PLANT TOPICS:

III-6, SEISMIC DESIGN CONSIOERA'TION III-ll, COMPONENT INTEGRITY IHTROOUCTIOH Oc The nuclear power plant facilities under review in the SEP received construc-tion permits between 1 956 and 1 967.

Seismic design procedures evolved

'ignificantly during and after this period.

The Standard Review Plan (SRP) first issued in 1975, along with the Regulations 10 CFR Part 50, Appendix A and 10 CFR Part 100, Appendix A constitute current'licensing criteria for seismic design reviews.

As a result, the original seismic design of the SEP. facilities vary in degree from the Uniform Building Code up through and approaching current standards.

Recognizing this evolution, the staff found that it is necessary to make a reassessment of the seismic safety of these plants.

Under SEP seismic reevaluation, these eleven plan'ts were categorized into two groups based upon the original seismic design and the availability of seismic design documentation.

Oifferent approaches were used to review the plant facilities in each group.

The approaches were:

Group I:

Oetailed HRC review of existing seismic design documents with limited reevaluation of the existing facili.ty to confirm judg-ments on the adequacy of original design with respect to cur-rent requi~ements:;.

0 Group II:

Licensees were required to reanalyze their facilities and to upgrade, if necessary, the seismic capacity of their facili=

ty.

The staff will review the licensee's reanalysis

methods, scope, and results.

Limited independent NRC analysis will be performed to confirm the adequacy of the licensee's method and results.

Based upon the staff's assessment of the original seismic design; the Ginna plant was placed in Group I for review.

The Ginna plant, a pressurized light-water moderated and cooled nuclear re-'

actor, is located on the south shore of Lake Ontario, about 16 miles east of Rochester, Hew York.

Westinghouse Electric Corporation was the prime con-tractor for the plant.

The Westinghouse engaged Gilbert, Associates,. Inc. as the architect-engineer responsible for the plant design and construction specifications.

Bechtel Power Corporation was the general contractor for con-struction.

The plant received its Construction Permit on April 25, 1966 and Provisional Operating License on September 19, 1969.

Rochester Gas and Elec-tric Corporation (RG8E), the owner, filed its application for a Full-term Operating License on August 9, 1972.

The Ginna plant was originally designed for an operating basis earthquake (OBE) with a peak ground acceleration (PGA) of 0.08g and reviewed for a safe shutdown earthquake (SSE) with a PGA of 0.2g.

Housner ground response spectra scaled to the specified PGA's were used as seismic input for the analyses and design.

The vertical component of ground motion was assumed to be the same as the hori-zontal components.

For the anlayses of most seismic Class I structures (con- '

tainment shell, containment internal structures, auxiliary building; and di'easel generator building); the buildings were modelled as lumped. mass-spring systems with fixed bases for calculating the fundamental frequency of each building; then, the corresponding spectral accelerations were used for performing the equivalent static analysis and seismic design.

For the control building and intermediate building, only the seismic resisting mechanisms (shear walls and steel bracings) were checked to determine if they were capable of resist-ing the equivalent seismic loads.

The same approach used for Class I structures.was applied for the analysis and design of the seismic Class I piping systems and equipment with the Housner ground response spectra used as input.

The damping ratios recommended by Housner were used for struc-tural and system analyses.

Chapter 3 of NRC NUREG/CR-1821 report, "Seismic Review of the Robert E. Ginna Nuclear Power Plant as Part of the Systematic Evaluation Program" (ref. 1) summarizes the details of the original analysis and design.

The SEP seismic review of Ginna facilities addressed only the Safe Shutdown Earthquake, since it represents the most severe event that must be consider-ed in the plant design.

The scope of the review included three major areas:

the integrity of the reactor coolant pressure boundary; the integrity of fluid and electrical distribution systems related to safe shutdown; and the integrity and functionability of mechanical and electrical equipment and engineered safety features systems (including containment).

A detailed re-view of the facilities was not conducted by the staff; rather our evaluations relied upon sampling representative structures,

systems, and components.

>>4-Confirmatory analyses using a conservative seismic input wel.e performed for the sampled structures, systems,:and components.

The results of these anal-yses served as the principal input for our evaluation of the seismic capacity of the facility.

REVIEW CRITERIA Since the SEP plants were not designed to current codes, standards, and NRC requir ements, it was necessary to perform "more. realistic" or "best estimate" assessments of the seismic capacity of the facility and to consider the con-..;

servatisms associated with original analysis methods and design criteria.

A set of review criteria and guidelines was developed for the SEP plants.

These review criteria and guidelines are described in the following documents:

1.

NUREG/CR-0098,

" Development of Criteria for Seismic'eview of Selected Nuclear Power Plants",

by H.

M. Hewmark and W. J. Hall, May 1978.

2.

"SEP Guidelines for Soil-Structure Interaction Review",

by SEP Senior Seismic Review Team, December 8, 1980.

0-For the cases that are not covered by the criteria stated

above, the following SRPs.'and Regulatory Guides were used for tge review:

1.

Standard Review Plan, Sections 2.5, 3.7, 3.8, 3.9 and 3.1 0 2.

Regulatory Guides 1.26, 1.29, 1.60, 1.61, 1.92, 1.100 and 1.T22.

RELATED TOPICS AND IHTERfACES The related SEP topics to the review of Seismic design considerations and component integrity are II-4, II-4.A, II-4.8, and II-4.C.

These topic's relate to specification of seismic hazard at the site, i.e. site specific ground response spectrum for the Ginna site.

The seismic input selected for the confirmatory analysis of Ginna facility, namely the Regulatory Guide 1.60 spectrum scaled to 0.'2g peak ground acceleration, envelopes the Ginna site specific gound response as shown in Fig. 1, therefore the results for these four safety topic eval-uation will not affect the review of seismic design considerations and component integrity.

EVALUATION A.

GENERAL APPROACH The seismic reevaluation of Ginna Nuclear Power Plant was initiated by conducting a detailed review of the plant seismic documentation.

The results of this review are summarized in the draft report, "Seismic Review of Ginna Nuclear Power Plant - Phase I Report".

Then, the staff and our consultants conducted a site-visit.

The purposes of this site visit were:

(1) to observe the as-built plant specific: features rela-tive to the seismic design of the faciltiy, (2) to obtain seismic design information which wa's not availabl.'e to the staff in the docket, (3) to

discuss, wi'th the licensee, seismic design information that the staff and our consultants had reviewed, and (4) based on the results of this field inspection, experience and judgement, to identify sample struc-

turess, systems, and components for which the confirmatory analyses (or audit analyses) would be performed.

The results of these

analyses, then, served as the basis for safety assessment of the plant facility.

When a structure was evaluated, it was judged adequately designed if the results from the structural analysis

'met one of.the following three cri-teria; 1.

The loads generated from confirmatory analysis were less than orig-i,nal loads; 2.

The seismic stresses from confirmatory analysis were low compared to the yield stress of steel or the compressive strength of concrete; or.'.

>>6-3.

The seismic stresses from confirmatory analysis exceeded the steel yield stress or the concrete compressive

strength, but,estimated re-served capacity (5r ductility) of the structure was such that in-."

elastic deformation without. failure 'would be expected.

If one of the above criteria were not satisfied, a more comprehensive reanalysis was required to demonstrate its design adequacy.

For piping reevaluation, the results from the audit analysis of each of the sampled piping systems were compared with ASHE Code requirements for Class 2 piping systems at appropriate service conditions.

This compar-ison provided the basis for reevaluating the structural adequacy of piping systems.

Because limited documentation exists regarding the original specifica-tions applicable to procurement of equipment, as weil as for the qualification of the equipment, the seismic review of equipment was based on expert experience and judgement.

Two levels of qualification were performed, structural integrity and functionability.

The results of this reevaluation of equipment served as the basis for modifications or reanalysis to be undertaken by the licensee.

B.

CONFIRMATORY ANALYSIS In order to provide independent analytical results for the reevaluation, a relatively complete seismic confirmatory analysis, which started with a definition of seismic input ground motion and ended with responses of the safety related structures and selected systems and components, during the postulated earthquake

event, was performed.

The analysis procedures

. and results are briefly discussed on the following sections.

1.

SEISMIC INPUT When seismic review of'Ginna plant started in mid 1979, the site specific gound response spectra were not avilable.

In order to per-0 form the review on a sampling basis that could be applied with-confidence, a more conservative ground motion, namely Regulatory Guide 1.60 horizontal ground response spectrum (R.

G. 1.60 spectra) scaled to 0.2g, the original design peak ground acceleration (PGA),

was used as the horizontal 'component o'f postuIated ground motion for analysis.

The input motion in the vertical direction was taken as 2/3 of the value in horizontal direction across the entire frequency range.

Recently, the site specific spectra development program was completed, arid the spectrum generated for.the Ginna site was::issued to the licensee on June 17, 1981 (ref. 2) for any future work that may be required.

The basis for the development of site. specific spectra was documented in NRC NUREG/CR-1582 report, "Seismic Hazard Analysis" (ref. 3).

This site specific spectrum is appropriate for assessing the actual safety margins present for any structures,

systems, and components that have been identified. as open items.

In Figure 1, a

comparison is made for the ground response spectra that were used for the original plant design and for SEP seismic reevaluation (Reg.

Guide 1.60 spectrum and the site specific spectra).

2.

ACCEPTANCE CRITERIA ANO SCOPE The specific SEP reevaluation criteria are documented in NUREG/CR-0098 and SEP Guidelines for Soil-Structures Interaction Review.

These documents provide guidance for:

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a) selection of the earthquake hazard; b) design seismic loadings; c) soil-structure interaction.;

d) damping and energy absorption; e) methods of dynamic analysis; f) review analysis and design procedures; and g) special topics such as under ground piping, tanks and vaults, equipment qualification, etc.

These criteria are felt to more accurately represent the actual stress level in structures, systems and components during a postulated earth-quake event and consider, to certain extent, nonlinear behavior of the systems.

The SEP shismic reevalUation of Gihna facility was a limited review

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'enteripg on:

o Assessment of the general integrity of the reactor coolant pressur e boundary.

o Evaluation of the capability of essential structures,

systems, and components required to shutdown the reactor safely and to maintain it in a safe shutdown condition (including the capability for removal of*residual'eat) during and after a postulated seismic event.

A total of two (2) structures, two (2) piping systems, seventeen (17) equipment components (mechanical and electrical) were fully evlauated.

They were:

o Structures - Containment building (containment shell and internal structures) and the interconnected auxiliary, turbine, intermediate, control, service, and diesel generator building complex.

o Piping Systems - Portions of residual heat removal line and safety injection line.

o Equipment - 12 mechanical items and 5 electrical items.

Additjonal samples will be selected if any open items cannot be resolved by analysis.

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ANALYSIS OF STRUCTURES Analytical procedures and methods conforming with the current state of the art were used.

These procedures and methods considered the three-dimension dynamic response effects of buildings, interaction between C

buildings; equipment masses

, structural damping in accordance with calculated stress

levels, and so forth.

A.

ANALYSIS OF CONTAINMENT BUILOING The containment building is a Vertical, cylindrical concrete struc-ture with a flat base mat and a hemispherical dome.

The building is 99 ft. high (from base mat to spring line) and has a 105 ft.

inside diameter.

Thy concrete wall, which is prestressed vertically and reinforced horizontally, is 3.5 ft. thick.

The thickness of the reinforced concrete dome and base mat are 2.5 ft. and 2 ft. respec-A tively.

Housed by containment shell, the internal reinforced con-crete structures'are supported by the same base mat which is founded on bedrock. by means of post-tensioned rock anchors.

A hybrid computer model* was used for the containment building (containment shell, internal structures, and base Sat).

The contain-ment shell was modelled as a fixed-base lumped mass-spr ing system and the internal structures were modelled as a fixed-base three-dimension-al finite element model.

These two models are coupled through the crane structure and the NSSS.

Because the building is founded on rock; soil-structure interaction effect corrections are not required.

The detailed discussion of modelling techniques and the final dynamic model used for the confirmatory analysis are found in NRC NUREG/CR-1821 report.

"The mode was originally developed by the licensee and their consult-anC (Gilbert Assoc., Inc.) and reviewed by the staff.

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In order t'o generate the building responses (dynamic moments,

shears, and axial forces) for the structural. evaluation, the model was anal-yzed through the response spectrum analysis method with R.'G. 1.60 spectrum scaled to 0.2g as seismic input.

The time-history analysis approach together with an artificial time history record (acceler-'tion)."scaled to the same PGA, namely 0.2g, was used for generating in-structure (or floor) response spectra.

After the peaks were broadened

+15K of corresponding frequency in accordance with R.

G.

1.122, the smoothed response spectra were used as input motions for the evaluation of piping systems and equipment.

All in-structure response spectra were summarized in Chapter 4 of NUREG/CR-1821 report.

The results of structural evaluation showed that containment building is capable of withstanding the postulated.

SSE event.

B.

ANALYSIS OF INTERCONNECTED AUXILIARY, INTERMEDIATE, TURBINE, CONTROL,

SERVICE, AND DIESEL GENERATOR BUILDING COMPLEX As shown in the plot plan (Fig.

2 of NUREG/CR-1821 report), the auxiliary, intermediate, control, and diesel generator buildings were classified as Class I structures.and the turbine service buildings Class III structures.

Most of these buildings are steel frame struc-tures with reinforced concrete basements that are structurally con-nected together.

Since the staff and its consultants believed that the coupling between all these buildings would effect the dynamic response of structures, systems and components, the buildings were modeled as a U-shape three dimensional space frame model with a fixed base to simulate the rock foundation.

The same approaches, ap-plied for the containment building analysis, were used here for 0'

generating 'the building responses (dynami c moments, shears, member,

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forces, etc.)

and in-structure response spectra. that were used as input for the evaluation of the piping systems and equipment.

The details of modelling techniques, analysis procedures and an-alysis results are found in Chapter 4 of Ginna NUREG report.

The results of evaluation showed that the buildings have sufficient capacity to withstand the postulated SSE event.

However, four.

I sets of steel bracing (bracing at northeast corner of auxiliary building and bracings in the south, north, and west walls. of tur-

. bine building) were found to exceed the allowable stress level for the postulated SSE.

The licensee provided additional infor-mation for review on October 28, 1981 and November 13, 1981 (Ref.

4 8 5).

This open item is expected to be resolved by January 31, 1982 and will be addressed in a supplement to this Safety Evalua-tion Report.

4.

ANALYSIS OF PIPING SYSTEMS As a result of SEP preliminary seismic r eview of Ginna plant, NRC IE Bulletin 79-14, and other NRC Seismic requirements, the licensee initiated a seismic upgrade program after the completion of piping support modifications required by IE Bulletin 79-14.

In order to conservktdvely respond to the SEP seismic r'eview and possible future NRC seismic requirements, a set.of analysis procedures and criteria that conform with current NRC review criteria (namely, R.G. 1.60 Spectrum, R. G. l. 61 damping, SRP criteria, etc. ) were used for the piping arral.-

ysis

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. To 'date;:the.analysis of all safety related piping systems inside containment has been completed.

The overall upgrade program is scheduled for the completion by 1984 refueling outage.

As discussed in the section 8.2 of this report, two pipe lines from those piping systems completed to date were selected and analyzed independently to verify the adequacy of the as-built design and con-firm the upgr ade analysis results.

The pipe lines selected were portions of residua1 heat removal (PHR) and safety injection (SI).

system piping.

Audit analyses which incorporated current ASME Code and Regulatory Guide Criteria and used the floor response spectra as input motion were performed for each portion of piping system se-lected.

The results from these analyses were compared to ASME Code:.

r equirments for Class 2 piping systems at the appropriate service conditions.

This comparison provided the bases for assessing the structural adequacy of the piping under the postulated seismic load-ing condition.

Assumptions made for the analysis, methodology em-ployed and. analysis results are found in the INEL'eport (Ref. 9).

The results from. the confirmatory analysis showed that the sampled piping systems are capable of withstanding the postulated SSE seismic input.

5.

ANALYSES OF SELECTED MECHANICAL AND ELECTRICAL E UIPMENT:

The evaluation of equipment was done on

a. sampling. basis.

Safety're-

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- 1ated components required for safe

shutdown, the primary pressure

boundary, and engineering safeguard features were categorized as ac-tive or passive and as rigid or flexible according.to the criteria in R.

G. 1.45 and SRP 3. 9.3.

A representative sample (or samples) from each group was selected and evaluated to determine the seismic design margin or adequacy of each group.

In this way, groups of similar components were evaluated without the need for detailed re-evaluations of al I individual components.

The licensee was asked to provide seismic qualifications data for each sampled component including design drawings, specifications, and design calculations.

After a detailed evaluation of each com-ponent was completed, conclusions were drawn as to the overall seis-mic capacity of the safety related equipment at the Ginna facility.

The description of selected components; analytical procedures and evaluations are found in Chapter 5 of the Ginna NUREG report.

As discussed in the NUREG report, a total of 13 open items (struc-r tural and/or functional integrity) out of 18 sampled equipment were I

addressed as a result of the evaluation.

Some of these 13 items remain open due to lack of design information.

After the review and incorp-oration of additional information submitted by the licensee (Ref. 10-15), the results are summarized below:

o 3 Mechanical equipment items and one electrical item were found to be adequately designed.

o The component cooling surge tank support system was found to re-quire upgrading.

The staff accepted licensee's design criteria and analysis results.

o Refueling Water Storage Tank (RRST) was found to require upgrading.

This item will be resolved as part of the integrated assessment.

o Reactor Coolant Pumps were left open (structural integrity) due to lack of design information.

The licensee agreed to provide addi-tional information by January 29, 1982.

i o

The licensee's structural integrity evaluation of motor operated valves (both valves and piping) larger than 2"

under their seismic upgrade program was consi'dered to be adequate.

The licen-see included the reanalyses of small pipe line (2" in diameter and smaller), to which motor operated valves are attached,-:.in the on-going seismic upgrade program.

A separ ate evaluation will be per-formed 'to determine the effect of valve eccentricity on the pipe stresses when the analysis res'ults become available.

The licensee has demonstrated that the functional integrity of motor operated valves will be maintained under the postulated SSE.

The existing essential service water pumps were determined to be not qualified (structural, and functional integrity) due to the lack of support near the suction of the pumps, resulting in over stress in the pump casing support.

These pumps are unique to the service water system.

o The modifiedanchorage and support systems for safety related electrical equipment as well as the evaluations and modifications of internally mounted elements of safety related electrical equip-ment are found to be adequate.

o Motor Control Centers and Switchgears

- The structural design ade-quacy of the load path between an internally mounted component or device through the panel frame and. bracing to the anchorage system was not evaluated due to lack of design information.

The licensee agreed to provide this infonaation by January 27, 1982.

This item is expected to be closed out by January 31, 1982.

Control Room Panels - In order to demonstrate the structural inte-giity (load path from a internally mounted element to anchorage and support system) of panels, the licensee agreed to conduct a

low impedance test for a sample panel to determine the dynamic char-acteristics of the panels and to perform seismic analysis to demon-strate the design adequacy

'in the near future.

o The functionality of all safety related electrical equipment as well as the structural integrity of internal components of all safety related electrical equipemtn is being evaluated through SEP Owner Group program.

This program is scheduled for the completion by the end of 1982.

o gualification of electrical cable trays is being evaluated by test-ing through SEP Owners Group program.

This program is scheduled for completion by June of 1982.

CONCLUSION Based on the review of the original design analyses, the results of con-I firmatory analyses performed by the staff and its consultants, and the licensee's responses to the SEP seismic related safety issues, the follow-ing conclusions can be drawn:

Structure - All safety related structures and structural elements of the Ginna facility are adequately designed to resist the postulated seismic event.

However, four (4) sets of steel bracing system were found to exceed the allowable stress level for the postulated SSE.

The licensee provided additional analysis information for review on October 28, 1981 and November 13, 1981.

This open item is expected'o be resolved by January 31, 1982.

Piping Systems - According to the results of SEP piping audit analysis performed for the sampled piping systems (Ref. 9), the piping systems have been found to be capable of withstanding the postulated SSE.

Mechanical Equipment - A total of 12 mechanical equipment items were sampled.

From the 12 items, 7 have been determined to be adequate and two were determined to be inadequate.

Generally, the remaining open items are due to lack of design information.

This does not necessarily imply that safety deficiencies exist.

Rather, it is the staff's judge-ment that documentation of the adequacy of these open items can be ac.-

complished by February 28, 1982 and will be addressed in a supplement to this evaluation (Attachment 1).

However, our evaluation on tht ee (3) sampled safety related tanks (namely, component cooling surge tank, boric acid storage

tank, and refuel.jng water storage tank} showed that the sup-prot of component cooling surge tank needs to be upgraded and the refuel-ing water storage tank requires both with regard to support and structural integrity.

Since two of the sampled tanks were found to require upgrading, the seismic review of safety related tanks should be performed by the licensee to demonstrate the design adequacy of the remaining safety re-lated tanks (volume control tank and HaOH spray additive tank}.

Electrical Equipment - As a result of SEP seismic review, three (3) ac-tivities have been or are being completed by the licensee:

a) upgrading of anchorage and support of all safety related electrical equipment re-.

quired by HRC letters dated January 1,

and July 28 of 1980 (Refs.

16 8 17) has been completed, and found to be adequately designed (Attachment 1),

(b) a program has been initiated for the documentation of seismic quali-fication (functionality of the equipment and structural integrity of internal components) of all safety related electrical equipment, namely the SEP Owners Group program, and (c) a program for seismic qualification of electrical cable trays based upon testing by the SEP Owners has been implemented.

These latter two programs are intended to confirm the ade-quacy of existing designs and equipment.

Recently, HRC has initiated a generic program to develop criteria for the seismic qualifications of equipment in operating plant; Unresolved Safety Issue (USI) A-46.

This program is scheduled for the completion in'arch 1983.

Under this program, an explicit set of guidelines (or criteria) that could be used to judge the adequacy of the seismic qualifications

]both functional capability and structural integrity) of safety related mechan-

/

ical and electrical equipment at all operating plants will be developed.

Considering that:

(1)

All safety related electrical equipment has been properly anchored; (2)

Past experience and testing results (from both nuclear and nonnuclear-facilities) indicate in general that electrical equipment will con-

~ tinue to operate under dynamic loading conditions with only limited transient behavior, if the equipment is adequately anchored; and (3) the SEP Owners Group programs from which a set of general analytical methodologies is being developed for the seismic qualifications of cable trays and for documentation of other safety related electrical equipment (functionability);

it is our judgement that for the interim period until a technical resolu-tion of USI A-46 is reached regarding methods for assessing seismic qualification of equipment in operating plants, the safety related elec-trical equipment at Ginna plant will function during and after an earth-quake up to and including the postulated SSE.

If additional requirements are imposed.

as a result of USI A-46, regarding functional capability of safety related electrical equipment, the Ginna facility will be required to address these new requirements along with other operating reactors.

Furthermore, since the ground response spectrum.(0.2g R.

G. 1.60 spectrum) used for Ginha seismic reevaluation envelopes the Ginna site specific ground response

spectrum, additional safety margins'n the structures, sys-

tems, and components do exist for resisting seismic loadings.

Thus, the staff concludes that Ginna plant has an adequate seismic capacity to resist a, postulated

SSE, and therefore, there is reasonable assurance that the operation of the facility will not be inimical to health and safety of the public.

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J REFERENCES 1.

HRC HUREG/CR-1821

Report, "Seismic Review of the Robert E. Ginna Nuclear Power Plant a Part of 'the Systematic Evaluation Program",

December 1980.

2.

Letter;;from NRC 'to RG&E dated Jiine:17,.1981.

3.

HRC HUREG/CR-1582 Report, "Seismic Hazard Analysis", Vol. 4, October 1981.

4.

Letter from RG&E to NRC dated October 28, 1981.

5.

Letter from RG&E to HRC dated November 13, 1981.

6.. Letter from RG&E to HRC dated February 27, 1 981.

7.

Letter from NRC to RG&E dated February 20, 1 981.

8.

Letter from RG&E to NRC dated April 1, 1981.

9, EGG-EA-5513 Report, "Summary of the R.

E. Ginna Piping Calculations Per-formed for the Systematic Evaluation Program", July 1981.

10.

Letter from HRC to RG&E dated January 7, 1981.

ll.

Letter from RG&E to HRC dated February 6, 1981.

12.

Letter from RG&E to NRC dated May 26, 1981.

13.

Letter from RG&E to HRC dated September 24, 1981.

14.

Summary of September 9,

1981 meeting held at Rochester,-Hew York dated December 14, 1981.

15.

Summary of Integrated Assessment held on December 1,. 1981 at Hethesda, Maryland (to be issued in the near future).

16.

Letter from HRC to RG&E dated January 1, 1980.

17.

Letter from NRC to RG&E dated July 28, 1980.

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Lawrence Livermore National Laborat ry October 30, 1981 SM 81-290 Mr. William T. Russell, Branch Chief Systematic Evaluation Program Branch Division of Licensing Office of Nuclear Reactor Reg.

Washington, O.C.

20555

Dear Bill:

I have enclosed a progress report on open items for Ginna.

Of the portion assigned to SMA, Newport Beach, the CRO support has been completed and the one remaining item is the primary coolant oump casing.

The report on this item will be forwarded when it becomes available.

Sincerely, TAN/mg 0184m Thomas A. Nelson Project Manager Structural Mechanics Group Nuclear Test Engineering Division I'nclosure

STRUCTURAL mECHAA1CS ASSOCIATE S A conf. Coro.

5160 Birch Street, Newport Beach. Calif. 92660 (114) 633-7552 SMA 12205.20 October 15, 1981 Mr. Thomas A. Nelson (L-90)

Nuclear Test Engineering Division Lawrence Livermore Hational Laboratory P.O.

Sox 808 Livermore, California 94550

Subject:

Resolution of 0 en Items on Ginna Equipment

References:

1)

SMA Letter, R.

Campbell (SMA), to T. A. Nelson (LLNL) 27 August, 1981.

2)

RGE CROM Seismic Analysis, Westinghouse PWR Systems Division, May 22, 1979.

Dear Tom:

The reference letter summarized the status of open items for Ginna HSSS equipment.

At that time we were wafting further documentation on the control rod

. drives, control rod drive support structure and for the primary coolant pump casing at nozzle penetrations.

We received a Westinghouse analysis of the CRD housing,Ref.

2, for a'.8g static coefficient and verified that the loadings'.us'ed in the CRD housing seismic support analysis are correct and that stresses are within acceptable levels.

Therefore, the only remaining item is the primaty coolant pump casing at. nozzle penetrati'ons.

In our last con-versation with Westinghouse, they were comparing the Ginna pump casing and nozzle loading to the San Onofre units for which a detailed finite element stress ana'lysis has been conducted.

That comparison will be reviewed when received.

Very truly your s, STRUCTURAL MECH'AHICS ASSOCIATES, IHC.

Project Manager RDC:mw

Mr. William T. Russell, Branch Chief Systematic Evaluation Program Branch Division of Licensing Office of Nuclear Reactor Reg.

Washington O.C.

20555

Dear Bill:

September 30, 1981 SM81 259/012lb p/"'o FIN A0415 jog(

g I have enclosed a copy of a report addressing resolution of open items for the Ginna plant as a result of the September 9,

1981 meeting at the RG h E offices.

Sincerely, TAN/tlm enclosure Thomas A. Nelson Structural Mechanics Group Nuclear Test Engineering Division Sii00604i6 810930 CF ADQCH 05000244 CF pd

STRUCTURAL.

mECHAnICS ASSOCIATES 3645 Warrensville Center Road Cleveland, Ohio 44122 (216) 991-8842 16 September 1981 Mr. T. A. Nelson Program Manager, SEP Seismic Peview Nuclear Test Engineering Division Lawrence Livermore Laboratory P. 0. Box 808 Livermore, California.94550 k

Dear Tom:

Attached hereto please find my comments regarding the meeting held at RGB offices on 9/9/81 to address the unresolved or open items in the mechanical electrical equipment SEP concerning seismic integrity.

We have also updated Chapter 5 of t+REG/CR-1821 to reflect the resolution of items discussed at the meetina where appropriate.

Please advise if you require any clarification.

Sincerely, JDS:clj Enclosure John D. Stevenson Yice President and General Manager c.c. Or.

Tom Cheng

I

'i1

RESULTS OF MEETING BEiVlEEN RGKE THEIR CONSULTANTS ANO THE NRC-SEP STAFF f~ THEIR CONSULTANTS The following agenda items were discussed with resolution or comments as follows:

A.

Component Coolino Suroe Tank The review corments contain d in Section 5.3.1.3 of NUR G/CR-1821 were discussed by GAI personnel acting as consultants to RGhE.

Basically, they confirmed that no positive anchorage aoainst slidino currently existed in the longitudinal direction and they provided a stress

summary, Attachment A which indicated that the horizontal base plate and anchor bolts ar over stressed.

They have proposed a solution of adding two additional bolts to each of the two support legs.

This has the effect of reducing bending stresses in the horizontal base plate and helping to carry shear stress.

It was pointed out.during the discussion by Or. Stevenson that the addition of the two n w bolts in each saddle would induce loads from restraint of free end.displacement Cue to thermal gradients that currently are not identified in the design of supoorts and tank.

GAI representatives stated that they had reviewed the effect of the restraint of thermal expansion in the tank and attached pipe and stated that the resultant stresses were qu'te low.

They have not completed the analysis of their proposed fix but assur d that the acceptable stress limits presented in Attachment A would be met.

Or. Stevenson stated that if the stress limit criteria of Attachment A were met the zesultant design should demonstrate seisnic design adequacy.

However, he further stated.

a personal preference that, if at all possible, the modified support system should not provide thermal restraint.

B.

Refuelino Water Storage Tank GAI'has not finished the analysis which is scheduled now for a 1 Oecember 1981 completion.

Or. Stevenson reported the concern and conclusions reached in Section 5.3.6.6 of MREG/CR-1821, that if the potential amplified response of the tank under impulsive load was considezed instead of the assumption o

tank rigidity used in the original design, linear elastic analysis would indicate that the tank shell would buckle and the anchor bolts fail.

Results of the GAI analysis.,should be available for zeview by 1 December 1981.

C.

Auxilia Buildino Bzaci Bracing evaluation of the auxiliary building is scheduled. for completion by 1 November 1981.

O.

Anchorage of Electrical ouioment and Internall Mounted Components Or. Stevenson reviewed typical design fixes suoplied by RGhE in response to IE~E Bulletin BO-21 concerning anchorage of electrical equipment.

The criteria used in modifying the anchorage as expressed in "Final Report Anchorage and Seismic Support of Safety Related El ctrical Equipment" RGB Project No.

ERR-2831 dtd. 12/31/80 appeared quite conservative in that a factor of 1.5 times the peak of the aoplicable floor spectra was'sed for the design modification.

The RGhE ana ysis also considered the effect of bolt prying in their reevaluation and redesign.

In general, they used the expedient of providing new anchorag in the form of stick welded angles to the cabinet plate at the base which was then expansion bolt anchored to the concrete slab rather than evaluatino the existing anchorage design and installation integrity.

All internally mounted comoonents and devices weiching more than 25 pounds were analyzed as separate assemblies.

Attachment of all internal devices and components were surveyed to assure all indicated attachments in the form of bolts, screws, clips, etc. were installed.

Or. Stevenson concurred that the electrical equipment anchorage design and internal mounted devices and components evaluations and modification appeared quite adequate.

However, he'expr ssed a concern that the load path structuzal design adequacy between an electrical component or device through the panel frame and bracing to the equipment anchorage had not been adequately demonstrated as required (se ).

This was a

notable 'concern in Ginna as compared to Oresden-2 in that Dresden-2 provided upper lateral suppozts as well as new base supports to the cabinets thereby effectively halving the r action forces and reducing bending moments by a factor of four.

In addition, cabinet fundamental frequencies are increased by a factor or 3 as a result of the upper lateral restraint which in this case should also reduc the inertia loads.

Or. Stevenson suggested that RGhE should structurally evaluate, on p sample basis, electrical panelboards, cabinets and racks to demonstrate their structural design adequacy to the requirements of the AISC Code as modified by the SRP Section 3.8.4 for the load combination which included the SSE.

The battery racks are essentially the same as the Gould racks used on Oresden-2.

Oetailed structural analysis of the Gould racks for Oresden-2 indicate the only area of potential failure is in the wooden battens.

In Ginna the existing racks have been stiffened by an external structural steel bracing system which is independently expansion anchored to the floor.

In Or. Stevenson's opinion the design modification to the racks is obviously capable of carrying currently d fined seismic loads.

G.

Valve Qoerators I

In general, RGKE has made evaluation of Seismic Category I motor operated valves larger than 2" part of their seismic upgrade program where stresses in piping including the effect of eccentricity aze determined to be within code allowables.

The valve assembly is modeled for analysis as an equivalent tee section.

Dr. Stevenson expzessed a concern that it is the smaller diameter piping that is part'cularly sensitive to eccentric valve laods.

RGbE agreed to review its Seismic Cateaory I 2" and under lines to identify any MOV.

A separate calculation would be performed to evaluate the effect of valve eccentricity on the piping stresses.

'o date no additional evaluation of valve operability has been supplied.

See Section 5.3.1.7 of NUREG/CR-1821.

Essential Service Hat " Pumps No additional information has been supplied.

Oe.ionstration of functionality during a seismic disturbanc is still an unresolved issue.

WN

AhALYSIS R"-PORT Component Coolina Mater Surge Tank (CMST) Supports Analysis Basis As Hddels The CCOST oas considered co 'be an idealised single degree-of freedom rigid body supported by tvo saddle, supports.

The saddle supports vere considered to be fixed at the top at the veld

')oint cozneccing them to the rank body, and pin connecced at the base at che anchor bolts connecting them to the supporting structural steel beams.

B-Loads The three orthogonal components of SSF. seismic loads vere de>>

tecmined by 1) considering the support system (combined saddle and beam) frequency in each direction, respectively,

2) using damping equal to 3X of critical da~ping, and 3) interpolating between floor response. curves ac. elevations 271'-0" and 315'-0".

Pressure and te=perature loads vere determined considering the cank 'design condirions (section 7.0 of the Design Criteria) and che lateral stiffness of the supporting structural 'sreel beams.

C.

Stresses Scresses vere calculated by hand using conventional formulas for stress and strain-II.

Analysis Results Nomenclature is consistent vi.th the definitions given in section 8.0 of the Design Criteria (unless noted).

Only maximum Actual Stresses resulting from the load combinations specified in the Design Criteria are presented belov.

Also, only controling Scress Limits are defined.

Component /Location Actual Stress KSI Stress Limit KSI A.

Saddle 1.

All Vertical Plates 2.

Corner of Outside Vertical Flange Plate 3.

Shear Stress in Vert-ical Flange Plates

1. 16

<,+~, - 32.23 M 2.30 1.5S M 21.75

~2.25S M 32.63 loot defined by A~ code for Class III plate and shell structures,. consid-ered acceptable

Co=oonent/Location 4.

Shear Stress in Vert-ical Feb Plate Actual Stress KSI

0. 39 Stress Limit (KSI) lt 5.

Horizontal Base Plate 6.

Shear Stress in Melds joining Saddle Plat,es f,+ K~

76.73 33-17 2.25S M 32.63 F

2.25S - 36.90 B.

Shear Stress in Meld joining Tank and Saddle

~ 16.14 P~

2. 25S

36. 90 C.

Anchor Bolts 1.

Shear Stress 15.90 P~

~ 1.6S' 16.00 2.

Tension Stress E

21. 31 7t

1. 6 (26) -1. 8f

12. 97

@~A C HMEST 8

, ATTACHMENT I ANCHORAGE At<0 SUPPORT OF SAF:"T7 RELATE) ELECTRICAL, EQUIPMENT POINTS TQ BE AOORESSEO BY SEP

~ LICENSEES IH OECB'BER 31, 1980 5USHITTAL 1..In ormation should b. provided no. cnly or the ant.'iorage o

al ctrical equipment but also the entire support that provides a load oath (such as bracing znd frames),

as xelI zs support for'nternally attached ccmconen s,

The lat.er is ospecially important for cabinet or panel type electrical equipment (such as control panels, instrument panels, etc.) which hzs internally supper.tad components.

An example of a potential mproperly supper ed interral ccr;.-cnent would be a, heavy ccmpcnent cantilevered off a front shee metal panel wi:hout additional suppcrt to a stronger and sti f er location.

mesa iradaqua e supports for internal ccmpcnents also should be iden ified and corrected be,ore Oecember 31, 1980.

2.

In order toe rify that zn anchoraca or a support of safety related electrical equipment has aC uata capacity, provide justification by

test, or analytical means.

I expansion anchor bolts axis

, justifi-cation provided previously for I= Bulletin 79-02 can be utilized if applicable.

The acceptance criteria for substantiating these judgem nts should be provided, this may involve speci ying the

>actor of sa aty arid allcwable stress limits used for d sign and justifying the overturning moment arid shear force used.

3.

Provide a table listing all (to include both floor and wall mounted) safety reIzted electrical equip;.en in tha plant.

For e-ch piece of equipment,

'rovide the infor.ation described in the attached table (attachment 2).

These inves ig tions o

each pieco of equiprent shou'ld determine:

a.

Mhether posi ive znchorace or su-port exists b.

The type o= anchorage c.

1lhether internally a tach d componen.s are properly supported d.

Identify non-seismic Czt.gory I equipment, the dislodaament of which during an earthquake may be detrimental to safety relz ed equipment and render them inoperable.

Inspection of the anchorages of such non-seismic Category I equipment should be conducted.

I positive anchorages do no exist, they should be identified and modifi'ed be ore Gecerrber 31, 1980.

4.

Mherever modif'.czticns oF znchorzges or supports are required, these mcdificaticns shculd be implemented znd thoroughly documented, 5.

The seismi c design o, cable trays

-.zy be treated as a separate prcbIo~,

because of its complexity.

Each licensee or Da SEP Owner's Group. should provide a separate action plan for the resolution of this issue within 30 days o

receipt of this latter.

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AftACiiNNT 2 Sttiyi<iilf OF INTCStlf+TIOH DF NIOiOBhljf AND SUfI'OilT OF SAfCT'f fiCUgCP CLCCIIIICQI. CilUII'Ntllt Aiib iiON-SCI5illC CAICGOIIT l ttENs TitAt IRT D<umu rois CqulpNENr Equip.

Nano E<tuip.

IP 'ys tc<a In iihlch Ins la I Iud location Typo of bldg.

A Clov.

Ancl<orage'as Anchorage Wdlftcd Since Jan. I, lg00 Internally T<IUII<. Nst<e I IP Ml~llppuri Cvaluated type-of'upport ttachad Co<<<ponents

.Ii!Iccsc Naacp l ID NIL~i type of 5.C<lg lpa lias 5upporl Elr~ luatc<i Non-Se I s Nlc Ca t I I tcps that could potcnllal ly I.O. of PucL<<<ent Suppurtlng Conclusion

'Cxaap les of Typo of Anchorage I I.

bolted to EqulIN<cut bolted to Concrete Mal) 5.. bolted to Concrete Slab bolted to bloci Matt 5.

Q idcd io tat<added Channc)

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SEP SAFETY TOPIC EVALUATION GINNA NUCLEAR POWER PLANT Contral@az.oz.c yz>z+-

TOPICS:

III-6,. SEISMIC DESIGN CONSIDERATION III-ll,COMPONENT INTEGRITY INTRODUCTION The nuclear power plant facilities under review in the SEP received construc-tion permits between 1956 and 1967.

Seismic design procedures evolved ; "

significantly during and after this period.

The Standard Review Plan (SRP) first issued in 1975, along with the Regulations 10 CFR Part 50, Appendix A and 10 CFR Part 100, Appendix A constitute-current'licensing criteria for seismic design reviews.

As a result, the original seismic design of the SEP, facilities vary in degree from the Uniform Building Code up through and approaching current standards.

Recognizing this evolution, the staff found that it is necessary to make a reassessment of the seismic safety of these plants.

Under SEP seismic reevaluation, these eleven plants were categorized into two groups based upon the original seismic design and the availability of seismic design documentation.

Different approaches were used to review the plant facilities in each group.

The approaches were:

Group I:

Detailed NRC review of existing seismic design documents with limited reevaluation of the existing facility to confirm judg-ments on the adequacy of original design with respect to cur-rent requirrements';-

REULATMYBUCKET FILE NPK I

Group II:

Licensees were required to reanalyze their facilities and to upgrade, if necessary, the seismic capacity of their facili-ty.

The staff will review the licensee's reanalysis

methods, scope, and results.

Limited independent NRC analysis will be performed to confirm the adequacy of the licensee's method and results.

Based upon the staff's assessment of the original seismic design; the Ginna plant was placed in Group I for review.

The Ginna plant, a

pr essurized light-water moderated and cooled nuclear re-actor, is located on the south shore of Lake Ontario, about 16 miles east of Rochester, New York.

Westinghouse Electric Corporation was the prime con-tractor for the plant.

The Westinghouse engaged Gilbert Associates,. Inc. as the architect-engineer responsible for the plant design and construction specifications.

Bechtel Power Corporation was the general contractor for con-struction.

The plant received its Construction Permit on April 25, 1966 and Provisional Operating License on September 19, 1969.

Rochester Gas and Elec-tric Corporation (RGSE), the owner, filed its application for a Full-term Operating License on August 9, 1972.

The Ginna plant was originally designed for an operating basis earthquake (OBE) with a peak ground acceleration (PGA) of 0.08g and reviewed for a safe shutdown earthquake (SSE) with a PGA of 0.2g.

Housner ground response spectra scaled to the specified PGA's were used as seismic input for the analyses and design.

The vertical component of ground motion was assumed to be the same as the hori-zontal components.

For the anlayses of most seismic Class I structures (con-

tainment shell, containment internal structures, generator building), the buildings weremodelled with fixed bases for calculating the fundamental auxiliary building;. and 'diesel as lumped-mass-spring systems frequency of each building; then, the corresponding spectral accelerations were used for performing the equivalent static analysis and seismic design.

For the control building and intermediate building, only the seismic resisting mechanisms (shear walls and steel bracings) were checked to determine if they were capable of resist-ing the equivalent seismic loads.

The same approach used for Class I structures,was applied for the analysis and design of the seismic Class I piping systems and equipment with the Housner ground response spectra used as input.

The damping ratios recommended by Housner were used for struc-tural and system analyses.

Chapter 3 of NRC NUREG/CR-1821 report, "Seismic Review of the Robert E. Ginna Nuclear Power Plant as Part of the Systematic Evaluation Program" (ref. 1) summarizes the details of the original analysis and design.

The SEP seismic review of Ginna facilities addressed only the Safe Shutdown Earthquake, since it represents the most severe event that must be consider-ed in the plant design.

The scope of the review included three major areas:

the integrity of the reactor coolant pressure boundary; the integrity of fluid and electrical distribution systems related to safe shutdown; and the integrity and functionability of mechanical and electrical equipment and engineered safety features systems (including containment).

A detailed re-view of the facilities was not conducted by the staff; rather our evaluations relied upon sampling representative structures,

systems, and components.

S 8

Confirmatory analyses using a conservative seismic input were performed for the sampled structures,

systems,

~ and components.

The results of these anal-yses served as the principal input for our evaluation of the seismic capacity of the facility.

REVIEW CRITERIA Since the SEP plants were not designed to current codes, standards, and NRC requirements, it was necessary to'erform "more, realistic" or "best estimate" assessments of the seismic capacity of the facility and to consider the con-::.'-

servatisms associated with original analysis methods and design criteria.

A set of review criteria and guidelines was developed for the SEP plants.

These review criteria and guidelines are described in the following documents:

1.

NUREG/CR-0098, "Development of Criteria for Seismic Review of Selected Nuclear Power Plants",

by N. H.

Newmark and W. J. Hall, May 1978.

2.

"SEP Guidelines for Soil-Structure Inter action Review",

by SEP Senior Seismic Review Team, December 8, 1980.

for the cases that are not cover ed by the criteria stated

above, the following SRPs.'and Regulatory Guides were used for the review:

1.

Standard Review Plan, Sections 2.5, 3.7, 3.8, 3.9 and 3.10 2.

Regulatory Guides 1.26, 1.29, 1.60, 1.61, 1.92, 1.100 and 1.T22.

RELATED TOPICS AND INTERfACES The related SEP topics to the review of Seismic design considerations and component integrity are II-4, II-4.A, II-4.B, and II-4.C These topics relate to specification of seismic hazard at the site, i.e. site specific ground response spectrum for the. Ginna site.

The seismic input selected for the confirmatory analysis of Ginna facility, namely the Regulatory Guide 1.60 spectrum scaled to 0.2g peak ground acceleration, envelopes the Ginna site specific gound response

as shown in Fig. 1, therefore the results for these four safety topic eval-uation will not affect the review of seismic design considerations and component integrity.

EVALUATION A.

GENERAL APPROACH The seismic reevaluation of Ginna Nuclear Power Plant was initiated by conducting a detailed review of the plant seismic documentation.

The results of this review are summarized in the draft report, "Seismic Review of Ginna Nuclear Power Plant - Phase I Report".

Then, the staff and our consultants conducted a site-visit.

The purposes of this site visit were:

(1) to observe the as-built plant specific) features rela-tive to the seismic design of the faciltiy, (2) to obtain seismic design information which was not availabl,e to the staff in the docket, (3) to

discuss, wi'th the licensee, seismic design information that the staff and our consultants had reviewed, and (4) based on the results of this field inspection, experience and judgement, to identify sample struc-

tures, systems, and components for 'which the confirmatory analyses (or audit analyses) would be performed.

The results of these

analyses, then, served as the basis for safety assessment of the plant facility.

When a structure was evaluated, it was judged adequately designed if the results from the structural analysis

'met one of the following three cri-teria.

1.

The loads generated from confirmatory analysis were less than orig-inal loads; 2.

The seismic stresses from confirmatory analysis were low compared to the yield stress of steel or the compressive strength of concrete; on<'.

3.

The seismic stresses from confirmatory analysis exceeded the steel yield stress or the concrete compressive

strength, but estimated re-served capacity (dr ductility) of the structure was such that in-.".

elastic deformation without failure would be expected.

If one of-the above criteria were not. satisfied, a more comprehensive reanalysis was required to demonstrate its design adequacy.

For piping reevaluation, the results from the audit analysis of each of the sampled piping systems were compared with ASIDE Code requirements for Class 2 piping systems at appropriate service conditions.

This compar-ison provided the basis for reevaluating the structural adequacy of piping systems.

Because limited documentation exists regarding the original specifica-tions applicable to procurement of equipment, as we]l as for the qualification of the equipment, the seismic review of equipment was based on expert experience and judgement.

Two levels of qualification were performed, structural integrity and functionability.

The results of this reevaluation of equipment served as the basis for modifications or reanalysis to be undertaken by the licensee.

B.

CONFI RMATORY ANALYSIS In order to provide independent analytical results for the reevaluation, a relatively complete seismic confirmatory analysis, which started with a definition of seismic input ground motion and ended with responses of the safety related structures and selected systems and components, during the postulated earthquake

event, was performed.

The analysis procedures and results are briefly discussed on the following sections.

1.

SEISMIC INPUT When seismic review of'Ginna plant started in mid 1979, the site specific gound response spectra were not avilable.

In order to per-form the review on a sampling basis that could be applied with confidence, a more conservative ground motion, namely Regulatory Guide 1.60 horizontal ground response spectrum (R.

G. 1.60 spectra) scaled to 0.2g, the original design peak ground acceleration (PGA),

was used as the horizontal component o'f postulated ground motion for analysis.

The input motion in the vertical direction was taken as 2/3 of the value in horizontal direction across the entire frequency range.

Recently, the site specific spectra development program was completed, arid the spectrum generated for'he Ginna site was issued to the licensee on June 17, 1981 (ref. 2) for any future work that may be required.

The basis for the development of site specific spectra was documented in NRC NUREG/CR-1582 report, "Seismic Hazard Aiialysis" (ref. 3).

This site specific spectrum is appropriate for assessing the actual safety margizs present for any structures,

systems, and components that have been identified as open items.

In Figure 1, a

comparison is made for the ground response spectra that were used for the original plant design and for SEP seismic reevaluation (Reg.

Guide 1.60 spectrum and the site specific spectra).

2.

ACCEPTANCE CRITERIA AND SCOPE The specific SEP reevaluation criteria are documented in NUREG/CR-0098 and SEP Guidelines for Soil-Structures Interaction Review.

These documents provide guidance for:

a) selection of the ear thquake hazard; b) design seismic loadings; c) soil-structure interaction; d) damping and energy absorption; e) methods of dynamic analysis; f) review analysis and design procedures; and g) special topics such as under ground piping, tanks and vaÃts, equipment qualification, etc.

These criteria are felt to more accurately represent the actual stress level in structures, systems and components during a postulated earth-quake event and consider, to certain extent, nonlinear behavior of the systems.

'he SEP shismic reevalu'ation of Ginna facility'as a limited review centering on:

o Assessment of the general integrity of the reactor coolant pressure boundary.

o Evaluation of the capability of essential structures,

systems, and corn'ponents required to shutdown the reactor safely and to maintain it in a safe shutdown condition (including the capability for removal of r esidual heat) during and after a postulated seismic event.

A total of two (2) structures, two (2) piping systems, seventeen (17) equipment components (mechanical and electrical) were fully evlauated.

They were:

o Structures

- Containment building (containment shell and internal structures) and the interconnected auxiliary, turbine, intermediate, control, service, and diesel generator building complex.

o Piping Systems - Portions of residual heat removal line and safety

'njection line.

o Equipment - 12 mechanical items and 5 electrical items.

Addi'tjonal samples will be selected if any open items cannot be resolved by analysis.

3.

ANALYSIS OF STRUCTURES Analytical procedures and methods conforming with the current state of the art were used.

These procedures and methods considered the three-dimension dynamic response effects of buildings, interaction between buildings, equipment masses

, structural damping in accordance with calculated stress

levels, and so forth.

A.

ANALYSIS OF CONTAINMENT BUILDING The containment building is a Oertical, cylindrical concrete struc-ture with a flat base mat and a hemispherical dome.

The building is 99 ft. high (from base mat to spring line) and has a 105 ft.

inside diameter.

The concrete wall, which is pr estressed vertically and reinforced horizontally, is 3.5 ft. thick.

The thickness of the reinforced concrete dome and base mat are 2.5 ft. and 2 ft. respec-tivelg.

Housed by containment shell, the internal reinforced con-crete structur es. are supported by the same base mat which is founded on bedrock. by means of post-tensioned rock anchors.

A hybrid computer-model* was used for the containment building (containment shell, internal structures, and base Oat).

The contain-ment shell was modelled as a fixed-base lumped mass-spring system and the internal structures were modelled as a fixed-base three-dimension-al finite element model.

These two models are coupled through the crane structure and the NSSS.

Because the building is founded on rock, soil-structure interaction effect corrections are not required.

The detailed discussion of modelling techniques and the final dynamic model used for the confirmatory analysis are found in NRC NUREG/CR-1821 report.

• The mode was original y developed by the'icensee and their consult-anC (Gilbert Assoc., Inc.) and reviewed by the staff.

I

10

In order to generate the building responses (dynamic moments,

shears, and axial forces) for the structural evaluation, the model was anal-yzed thorough the response spectrum analysis method with R.

G. 1.60 spectrum scaled to 0.2g as seismic input.

The time-history analysis approach together with an artificial time history record (acceler-ation)."scaled to the same PGA, namely 0.2g, was used for generating in-structure (or floor) response spectra.

After the peaks were broadened

+155 of corresponding frequency in accordance with R.

G.

1.122, the smoothed response spectra were used as input motions for the evaluation of piping systems and equipment.

All in-structure response spectra were summarized in Chapter 4 of NUREG/CR-1821 report.

The results of structural evaluation showed that containment building is capable of withstanding the postulated SSE event.

B.

ANALYSIS OF INTERCONNECTED AUXILIARY, INTERMEDIATE, TURBINE, CONTROL,

SERVICE, AND DIESEL GENERATOR BUILDING COMPLEX As shown in the plot plan (Fig.

2 of NUREG/CR-1821 report), the auxiliary, intermediate, control, and diesel generator buildings were classified as Class I structures.and the turbine servide buildings Class III'tructures.

Most of these buildings are steel frame struc-tur es with reinforced concrete basements that are structurally con-nected together.

Since the staff and its consultants believed that the coupling between all these buildings would effect the dynamic response of structur es, systems and components, the buildings were modeled as a U-shape-'hree dimensional space frame model with a fixed base to simulate the rock foundation.

The same approaches, ap-plied for the containment building analysis, were used here for

generating 'the building responses (dynamic moments, shears, member forces, etc.)

and in-structure response spectra. that were used as input for the evaluation of"the piping systems and equipment.

The details of modelling techniques, analysis procedures and an-alysis results are found in Chapter 4 of Ginna NUREG report.

The results of evaluation showed that the buildings have sufficient capacity to withstand -the postulated.SSE event.

However, four 1

sets of steel bracing (bracing at northeast corner of auxiliary building and bracings in the south, north, and west walls, of tur-bine building) were found to exceed the allowable stress level for the postulated SSE.

The licensee provided additional infor-mation for review on October 28, 1981 and November 13, 1981 (Ref.

4 8 5).

This open item is expected to be resolved by January 31, 1982 and will be addressed in a supplement to this Safety Evalua-tion Report.

4.

ANALYSIS OF PIPING SYSTEMS As a result of SEP preliminary seismic review of Ginna plant, NRC IE Bulletin 79-14, and other NRC Seismic requirements, the licensee initiated a seismic upgrade program after the completion of piping support modifications requir ed by IE Bulletin 79-14.

In order to conservatively respond to the SEP seismic review and possible future NRC seismic requirements, a set.of analysis procedures and criteria that conform with current NRC review criteria (namely, R.G. 1.60 Spectrum, R.G.

1.61

damping, SRP criteria, etc.)

were used for the piping anal-ysis.

To 'date; 'the..anhlysis of all safety related piping systems inside containment has been completed.

The overall upgrade program is scheduled for the completion by 1984 refueling outage.

As discussed in the section B.2 of this report, two pipe lines from those piping systems completed to date were selected and analyzed independently to verify the adequacy of the as-built design and con-firm the upgrade analysis results.

The pipe lines selected were portions of residual heat removal (RHR) and safety injection (SI),

system piping.

Audit analyses which incorporated current ASME Code and Regulatory Guide Criteria and used the floor response spectra as input motion were performed for each portion of piping system se-lected.

The results from these an'alyses were compared to ASME Code requirments for Class 2 piping systems at the appropriate service cond~tions.

This comparison provided the bases for assessing the structural adequacy of the piping under the postulated seismic load-ing condition.

Assumptions made for the an'alysis, methodology em-ployed and analysis results are found in the INEL'eport (Ref. 9).

The results from the confirmatory analysis showed that the sampled piping systems are capable.of withstanding the postulated SSE seismic input.

5.

ANALYSES OF SELECTED MECHANICAL AND ELECTRICAL E UIPMENT.'he evaluation of equipment was done on a sampling basis.

Safety re-

-e lated components required for safe shutdown, the primary pressure

boundary, and engineering safeguard features were categor ized as ac-tive or passive and as rigid or flexible according. to the criteria in R.

G. 1.45 and SRP 3. 9.3.

A representative sample (or samples)

from each group was selected and evaluated to determine the seismic design margin or adequacy of each group.

In this way, groups of similar components were evaluated without the need for detailed re-evaluations of all individual components.

The licensee was asked to provide seismic qualifications data for each sampled component including design drawings, specific'ations, and design calculations.

After a detailed evaluation of each com-ponent was completed, conclusions were drawn as to the overall seis-mic capacity of the safety related equipment at the Ginna facility.

-The description of selected components; analytical procedures and evaluations are found in Chapter 5 of the Ginna NUREG report.

As discussed in the NUREG report, a total of 13 open items (stnuc-tur al and/or functional integrity) out of 18 sampled equipment were I

addressed as a result ofj the evaluation; Some of. these 13 items remain 4

open due to lack of design information.

After the review and incorp-oration of additional information submitted by the licensee (Ref.

1'0-15), the results are summar ized below:

o 3 Mechanical equipment items and one electrical item wer e found to be adequately designed.

o The component cooling surge tank support system was found to re-quire upgrading.

The staff accepted licensee's design criteria and analysis results.

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Refueling Water Storage Tank (RRST)'as found to require upgrading.

This item will be resolved as part of the integrated assessment.

o Reactor Coolant Pumps were left open (structural integrity) due to lack of design information.

The licensee agreed to provide addi-tional information by January 29, 1982.

0 o

The licensee's structural 'integrity evaluation of motor operated valves (both valves and piping) larger than 2" under their seismic upgrade program was considered to be adequate.

The licen-see included the reanalyses of small pipe line (2" in diameter and smaller), to which motor operated valves are attached,;in the on-going seismic upgrade program.

A separate evaluation will be per-formed to determine the effect of valve eccentricity on the pipe stresses when the analysis results become available.

The licensee has demonstrated that the functional integrity of motor operated valves will be maintained under the postulated SSE.

o The existing essential service water pumps were determined to be not qualified (structural, and functional integrity) due to the lack of support near the suction of the pumps, resulting in over stress in the pump casing support.

These pumps are unique to the service water system.

o The modified:anchorage and support systems for safety related electrical equipment as well as the evaluhtions and modifications of internally mounted elements of safety related electrical equip-ment are found to be adequate.

o Rotor Control Centers and Switchgears - The structural design ade-quacy of the load path between an internally mounted component or device through the panel frame and.-bracing to the anchorage system was not evaluated due to lack of design information.

The licensee agreed to provide this information by January 27, 1 982.

This item is expected to be closed out by January 31, 1982.

o Control Room Panels - In order to demonstrate the structural inte-giity (load path from a internally mounted element to anchorage and support system) of panels, the licensee agreed to conduct a

low impedance test for a sample panel to determine the dynamic char-acteristics of the panels and to perform seismic anaXysis to demon-strate the design adequacy in the near future.

o The functionality of all safety related electrical equipment as well as the structura'i integrity of internal components of all safety related electrical equipemtn is being evaluated through SEP Owner Group program.

This program is scheduled for the completion by the end of 1982.

o gualification of electrical cable trays is being evaluated by test-ing through SEP Owners Group program.

This program is scheduled for completion by June of 1982.

CONCLUSI'ON Based on the review of the original design analyses, the results of con-firmatory analyses performed by the staff and its consultants, and the licensee's responses to the SEP seismic related safety issues, the follow-ing conclusions can be drawn:

Structure - All safety related structures and structural elements of the Ginna facility are adequately designed to resist the postulated seismic event.

However, four (4) sets of steel bracing system were found to exceed the a'llowable stress level for the postulated SSE.

The licensee provided additional analysis information for review on October 28, 1981 and November 13, 1981.

This open item is expected to be resolved by January 31, 1982.

Piping Systems - According to the results of SEP piping audit analysis performed for the sampled piping systems (Ref. 9), the piping systems have been found to be capable of withstanding the postulated SSE.

Mechanical Equipment - A total of 12 mechanical equipment items were sampled.

From the 12 items, 7 have been determined to be adequate and two were determined to be inadequate.

Generally, the'remaining open items are due to lack of design information.

This does not necessarily imply that safety deficiencies exist.

Rather, it is the staff's judge-ment that documentation of the adequacy of these open items can be ac-complished by February 28, 1982 and will be addressed in a supplement to this evaluation (Attachment 1).

However, our evaluation on three (3) sampled safety related tanks (namely, component cooling surge tank, boric acid storage

tank, and refuel,ing water storage tank} zhowed that the sup-prot of component cooling surge tank needs to be upgraded and the refuel-ing water storage tank requires both with regard to support and structural

-integrity.

Since two of the sampled tanks were found to require upgrading, the seismic review of safety related tanks should be performed by the licensee to demonstrate the design adequacy of the remaining safety re-lated tanks (volume control tank and NaOH spray additive tank).

h Electrical Equipment - As a result of SEP seismic review, three (3) ac.-'"

tivities have been or-are being completed by the licensee:

a) upgrading of anchorage and support of all safety related electrical equipment re-quired by NRC letters dated January 1,

and July 28 of 1980 (Refs.

16 8 17) has been completed, and found to be adequately designed (Attachment 1),

(b) a program has been initiated for the documentation of seismic quali-fication (functionality of the equipment and structural integrity of internal components) of all safety related electrical equipment, namely the SEP Owners Group program, and (c) a program for seismic qualification of electrical cable trays based upon testing by the SEP Owners has been implemented.

These latter two programs are intended to confirm the ade-quacy of existing designs and equipment.

Recently, NRC has initiated a generic program to develop criteria for the seismic qualifications of equipment in operating plant; Unresolved Safety Issue (USI) A-46.

This program is scheduled for the completion in'arch 1983.

Under this program, an explicit set of guidelines (or criteria) that

could be used to judge the adequacy of the seismic qualifications (both functional capability and structural integrity) of safety related mechan-ical and electr ical equipment at all operating plants will be developed.

Considering that:

(1)

All safety related electrical equipment has been properly anchored; (2)

Past experience and testing results (from both nuclear and nonnuclear facilities) indicate in general that electrical equipment will con-

- ti'nue to operate under dynamic loading conditions with only limited transient behavior, if the equipment is adequately anchored; and (3) the SEP Owners Group programs from which a set of general analytical methodologies is being developed for the seismic qualifications of cable trays and for documentation of other safety related electrical equipment (functionability);

it is our judgement that for the interim period until a technical resolu-tion of USI A-46 is reached regarding methods for assessing seismic qualification of equipment in operating plants, the safety r elated elec-.

trical equipment at Ginna plant will function during and after an earth-quake up to and including the postulated SSE.

If additional requirements are imposed, as a result of USI A-46, regarding functional capability of safety related electrical equipment, the Ginna facility will be required to address these new requirements along with other operating reactors.

Furthermore, since the ground response spectrum (0.2g R.

G. 1.60 spectrum) used for GihAa seismic reevaluation envelopes the Ginna site specific ground response spectrum, additional safety margins'in the structures, sys-

tems, and components do exist for resisting seismic loadings.

Thus, the staff concludes that Ginna plant has an adequate seismic capacity to resist a postulated

SSE, and therefore, there is reasonabl.e assurance that the operation of the facility will not be inimical to health and safety of the public.

REFERENCES 1.

NRC NUREG/CR-1821

Report, "Seismic Review of the Robert E. Ginna Nuclear Power Plant a Patt of 'the Systematic Evaluation Program",

December 1980.

2.

Letter,".from NRC 'to RG&E "dated Jiine:17',,1981.

3.

NRC NUREG/CR-1582 Report, "Seismic Hazard Analysis", Vol. 4, October 1981.

4.

Letter from RG&E to NRC dated October 28, 1981.

5.

Letter from RG&E to NRC dated November 13, 1981.

6.

Letter from RG&E to NRC dated February 27, 1981.

7.

Letter from NRC to RG&E dated February 20, 1981.

8.

Letter from RG&E to NRC dated April 1, 1981.

9, EGG-EA-5513 Report, "Summary of the R.

E. Ginna Piping Calculations Per-formed for the Systematic Evaluation Program", July 1981.

10.

Letter from NRC to RG&E dated January 7, 1981.

ll.

Letter from RG&E to NRC dated February 6, 1981.

12.

Letter from RG&E to NRC dated May 26, 1981.

13.

Letter from RG&E to NRC dated September 24, 1981.

14.

Summary of September 9,

1981 meeting held at Rochester,.-":New Yor k dated December 14, 1981.

15.

Summary of Integrated Assessment held on December 1,

1981 at Bethesda, Maryland (to be issued in the near future).

16.

Letter from NRC to RG&E dated January 1, 1980.

17.

Letter from NRC to RG&E dated July. 28, 1980.

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UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 January 29, 1982 Docket No. 50-244 LS05-82-01-070 Mr. John E. Maier Vice President Electric and 'Steam Production Rochester Gas 8 Electric Corp.

89 East Avenue Rochester, New York 14649 io

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SUBJECT:

SEP SAFETY TOPICS III-6, SEISMIC DESIGN CONSIDERATION AND III-ll,COMPONENT'NTEGRITY - GINNA NUCLEAR POWER PLANT We have completed our seismic review of Ginna Nuclear Power Plant.

Enclosed is a copy of our draft combined evaluation report of the two subject topics.

As discussed in this draft report, four items are required to be upgraded to meet SEP requirements for the postulated SSE:

(1) steel bracing at north-east corner of auxiliary building, (2) the support system of component cool-ing surge tank, (3) refueling water storage tank-; and (4) essential ser vice water pumps.

Six items still r emain open due to lack of design information.

According to mutual agreement between the staff and your representative, the responses to these items are scheduled by January 31, 1982.

A supplement to this report will be issued after the review of your responses for'he six open items are completed.

This evaluation will be a basic input to the integrated safety assessment for your facility unless you identify changes needed to reflect the as-built 5~i conditions at your facility.

With respect to the potential modifications

/j outlined in the conclusion of this report, a determination of the need to actually implement these changes milli be made during the same integrated as-tss>> ~s sessment.,

This topic assessment may be revised in the future if. your facil-"

ity design is changed or if NRC criteria relating to this topic are modified 400'efore the integrated assessment is completed.

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Sincerely, Encl osur es As stated Dennis tI. Crutchffeld, Chief Operating Reactors Branch No.

6 Divfs)on of L)censing cc w/enclosure:

See next page OFFICEI SURNAME/

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Docket No. 50-244 LS05-Mr. John E. Maier Vice President Electricaand Steam Production Rochester Gas II Electric Corp.

89 East Avenue'ochester, New York 14649

SUBJECT:

SEP SAFETY TOPICS III-6, SEISMIC DESIGN CONSIDERATION, AND III-ll, COMPONENT INTEGRITY - GINNA NUCLEAR POHER PLANT o

lie have comp/@ed our seismic review of Ginna Nuclear Power Plant.

Enclosed is a copy of our draft combined evaluation report of the two subject, topics.

As discussed in this draft report, some equipment items still remain open due to lack of design information.

According to mutual agreement between the staff and your representative, the responses to these open items are scheduled by January 31, 1982.

A supplement'o this evaluation report will be issued after the review of your responses is completed.

You are requested to examine the factgsupon which the staff has based its evaluation and respond either by confirming that the facts are correct, or by identifying errors and supplying the corrected information.

Ite encourage you to supply any. other material that might affect the staff's evaluatiAn of these topics or be significahh in the integrated assessment of your facility.

Your response is requested within 30 days of receipt of this letter. If no response is received within that time, we will assume that you have no com-ments or corrections.

Sincer ely,

Enclosure:

As stated Dennis M. Crutchfield, Chief Oped@ting Reactors Branch No.

5 Division of Licensing OFFICE/

SURNAME/

See next DATE)

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Mr. John E. Maier CC Harry H. Voigt, Esquire LeBoeuf Lamb, Leiby and MacRae 1333 New Hampshire Avenue N. W.

Suite 1100 Washington D. C.

20036 Mr. Michael Slade 12 Trailwood Circle Rochester, New York 14618 Ezra Bialik Assistant Attorney General Environmental Protection Bureau Mew York State Department of Law 2 World Trade Center New York New York 10047 Resident Inspector R. E. Ginna Plant c/o U. S.

NRC 1503 Lake Road

Ontario, New York 14519 Director, Bureau of Nuclear Operations State of New York Energy Office Agency Building 2 Empire State Plaza Albany, New York 12223 Rochester Public Library 115 South Avenue Rochester, New York 14604 Supervisor of the Town of Ontario 107 Ridge Road West

Ontario, New York 14519 U. S. Environmental Protection Agency Region II Office ATTN:

Regional Radiation Representative 26 Federal Plaza New York, New York 10007 Herbert Grossman, Esq.,

Chairman Atomic Safety and Licensing Board U. S. Nuclear Regulatory Comnission Washington, D. C.

20555 Or.

Emmeth A. Luebke Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.

20555 Dr. Richard F. Cole Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.

20555

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