ML17261A173
| ML17261A173 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 02/11/1980 |
| From: | White L ROCHESTER GAS & ELECTRIC CORP. |
| To: | Ziemann D Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8002140565 | |
| Download: ML17261A173 (34) | |
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February 11, 1980
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Division of Reactor Regulation Attention: l
. Oe.<<nis I,. Ziemann, Chic Cperating Reactors Branch No.
2 U.S. Nuclear Regulatory Commission washington, DC 20555 r
Subject:
Archo age and Support of Safety-Related Electrical
=cp." pment, R. E.~7'~>> Nuclear Paver Plant Docket H~~O-244 Dear Fz. Ziem~
In accordance vith Nr. =isenhut's lette on the above subject dated January 1,
1980 which was received January 11,
- 1980, RGH has developed the enclosed Electrical Equipment Anciior Se'smiq eriiication Plm.
Th's p3.an addresses the concews outlined in bo~h the letter and the attached dra t I.E. In ormation Notice.
Item (1).
Does positive anchorage exis- (load carryirig mechanism othe" than friction)?
A field su zey or all Class LE'lectrical ecpiipment at the Ginna Plant h~e~
made.
The findings of this survey are
- l979, on. th subjec of S"-~ Seismic Seviev.
This survey shoved'a" posi "ive anchoraue of all class l=- euu'pment eris-s.
The "efo e no temoorary anchors or supper s vill be added at this time.
Item (2). If positive anchorage
- exists, has the anchorage system been engineered vith adequate capac'ty?
Item (3).
Was the anchorage fabricated to auality standards?
Items (2) and (3) are addressed in the program described helo~.
~OCHESTER S wHO E'IC CORP February 7, 1980 Mr. Dennis L. Ziema~, Chief SHEET HO.'0 The R.E. Ginna Nuclear Power Plant was designed and con-structed prior to the first
~ ssuance of I:-=E Std.
344-}.971, "Guide for Seismic Qualificat'on of Class I.E. Equipment".
Therefore to pzovide assurances that minimum design retirements have been met, an analysis and verification program has been developed.
The design verification will be accompl'shed in three phases, inspection,
- analysis, and test and modification phases.
The output of t~'s plan will be a compaz'son of the Reouired Anchor Load Capacity (RALC) as determined by the analysis
- phase, with the Ve ified Anchor K.oad Capacity (VAT.C) for the anchor bolts associated with &at component or assembly as determined by the test and modification phase.
If the VM.C is found to be ecual to or greater than the RALC, then no modification is recuized.
however, if the VAE,C is found to be less than the R'ZC foz an electr-'cal
- assembly, addi-tional anchors will be added.
~
It is 2e intent of'h's program to resolve the overall
'ssue of elec"rical ec "pnent anchorage se'smic capability by Sep~eWe" 1,
1980.
The sched"le for any re~ized modifications will be depencent on their extent and ecuipment delivery schedules.
Enclosuzes
ZL=CTRICAL:-QUIP~< S:"ISMIC ACTION PLAN
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Description A thzee phase program vill be initiated to provide assurances that the anchorage system bolts vill pe form their design function during the SSE.
Phase I vill consist of inspecting, and preparing "as-built" sketches for all sa ety related electrical equipment as listed belov.
Anchor bolts used on th's equipment vill be field inspected.
"As buil"" sketches vill be pre!a ed showing all necessary info~ation to perrorm Phase II.
Phase II vill consist of an analysis of each elect=ca3.
equipm nt, anchoring
- system, the results of vhich will be compazed to the test infor-any result'ng modizicat'ons required,to upgrade the existing anchoring sys" m to the cr'teria described in Anal.ysis Phase I,I section.
Ecuinment Addressed The action plan vill include all Class 1E electrical systems and components.
Certain Class 1E.equipment installed during recent modifications in accordance vith IEEE 344-1975 requirements is known to be seismically anchored and vill not be considered in this s tudv. 'hese components aze listed belov; Elec rical Assemblies P"evouslv Oual'ied Foxboro Instnunent Racks 7.5 KVA Constant Voltage Transformers 7.5 iCVA Inveztezs Tp~e W 480 MCC (MM )
The folloving e'ectrical assemblies and/or component vil3. be evaluated by the Seism'c Act'on Plan:
Electrical 'Assemblies Covered b
this action plan Re'y Rack Assemblies 480 volt IE buses 480 volt (ac)
I=-
MCC 125 Volt (dc) IE Stazte=s Pave>> Panels I=. Battery Racks I=- Batte~,Chargers Instalment Racks Control Panels DG Panels Non.IE Items (Ancilla g It. s)
L
~ f Inspection/Phase I:
C.
An "as bui t" sketch vill be developed for each piece of electrical e@ '@ment a fected by this program.
The diagrams vill detail the size and shape of the component bases, the type, size and spacing of the anchor bolts and the physical dimensions and mass of the equipment.
Analysis Phase II This phase vill consist of'n analysis of each anchoring system to determine the minimum anchoring requirement to safely secur'e the equipm.t during a seismic even" using the folloving crite ia and assumptions.
The static anal sis descr'bed in Section Se3 of I:-:-E 344-L975
~ vill be the basis for establishing shear az.d tensile stresses expected in elect ical equipment anchors be'ng evaluated.
Specific-ally, the seismic response of all floor-mounted equipment is tion; using damping values in accordance vith R.Q. L.6L, multi-plied by a static coefficient of L.5 to account for multifrequency and multimode responses.
The inertial forces acting on the equipment center'f mass are then evaluated.
A multianchor computer model vilL then be used to determine the shear and tensile stresses for all floor.mounted em'.ipment using data from Phase 1.
he st esses thus determined vill establish the Required
~oa Caoacity (RALC) vhich will be
- compa, ed to the Verzzxed Anchor Load Capacx~y (VALC) determined. in Phase III, to establish anchor adequacy.
ri id and the zero period acceleration (ZPA) values vill e used o
ermine the seismic forces.
The tensile,and shear stresses vi13.. be calculat d using the multi anchor model.
Test and Mod'cation Phase III:
~
~ ~
i, p JJ< e) e After the "as built" ske<chos are completed, a test of
~le.
determine bolt LehgM, erwecment, and the bolt verified anchor load capability (VAK.C).
i Vhere the Required Anchor Load Capacity of an anchoring system is found to be g eater than the Verified Anchor E.oad.
f'apacities, a recommendation for additional anchor bolts or suppor-s vilL be made.
Equipment modifications vill be made using exis ing anchor bolt installation procedures and the "as bu't" sketches vill ref'ect all modificat'ons.
Results:
The results of the analyses and tests v'll show h~e~actors of ne ~~v fo the exieting ocher.b.olt-hy coepertng the
~RAL to the VAe.C values.
I C
I 3
Cable Tra and Conduit Support Anchors The cable tray and conduit support anchors vere installed using the manufacture s 'recommended procedures.
To verify the adequacy of Mesc anchors, a testing and verification.program vill be conducted to, (1) determine the nominal span betveen anchor centers.
This vill be documented by sketches of representative cable tray and conduit runs vith dimension and anchor locations.
(2) verify.the installed capacity of anchors.
(3) compare the verified t"ay and conduit support anchor configu ation with the configurations tested by Bechtel Power 'Corporation, described in ANCO Report 51053-21.1<<4, "Cable.'.Tray and Conduit Raceway Seismic Test Program".
It should be'oted that the existing cable <<"ay system at the Ginna Plant is a braced, strut supported system similar to those described in the test report.
These tests, vhich vere funded in part by RG&"=,
show d that in such tray systems, no major smctural
.fa'ures occurred during tests more severe than those required for Me Ginna site.
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~ Director of Nuclear Reactor Rem~lation Attention:
Mr. Dennis L. Ziemann, Ch'f Operating Reactors Branch No.
2 U.S. Nuclear Regulatory Commission Washington, DC 20555 t'ubject:
Systematic Evaluation Program - Seismic Review R. E. Ginna Nuclear Power Plant Docket, No. 50-244
Dear Mr. Ziemann:
During the April 10-3.3.,
1979 site visit by the NRC Seismic Review Team, men3ezs of the team requested that ve supply addi-tional information relat'ng to the seismic qualification of mechanical and electrical equipment; and fluid and electric distri-bution systems.
A subsequent meeting was held in Pittsburgh, Pennsylvania on June 12, 1979 between the
- NRC, RG&:- and their respective con-sultants.
It ~as agreed at the. Ju..e 12 neeting that RG&= would submit additional information and expected subm'ttal dates for mechanical and electrical equipment and systems by about June 29, 1979.
Accordingly, Enclosure I lists attached material and submittal lists attached material and st:mittal dates for electrical equip-ment. and systems.
As requested by your Staff, eight copies"of this letter and enclosures are being supplied for your use.
, If there are any questions regarding.this
- material, please contact us.
v Very truly yours, Enclosures
ROCHESTER GAS p
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July 3, 1979 D rec "r oz Nuclea= Reacto Regula 'n A-- n-ion:
Pw. De""'s L. Zie"a"m, Ch'ef Operating R ac"ars 3 anch No. 2 U.S. Nuclea=
Re~ "a"ory Ca~ssion Has~~
- gton, DC 20555 Sub~ ect:
Systematic =valuat on Proc=ra - Se sm c Review R. E.
G ~".a N cle= Pove Plant Docket No.
SC-24-".
Dear LL~. Ziem~"~:
During
~He April 10 llew 1979 sit v'sit by De HRC Seismic Rev='ev ~e~, m~~'rs o~ De team rerzes ed ~l-at ve apply addi-tioral in=o=aticr rel "
g to -'"e seismic ~ali:i ation o~
mechanical and elec-"ical er"=.en". a"d Quid and electric dis"-i-bution systems.
A SL' ecue t meet~ ~g >as held in PitAbl'h, Pe sy vania on June 12, 1979 between
~~e NRC, RCH and t".eir r spec-~ve can-sultar=s.
It ~as a.
d at t".e Jme 12 meeting that RG"= would subm't add'"'oral. in o~ation a"d expec
=" submittal dates
~or mechanical a"d electrical eq'ipment and system by 'about June 29, 1979.
Ac ord' y, ="closl -e
~ Lists a tac" t rat r al ar." semi ta'ates
<ar -~e mecha>>ca e~'->>me - and sys ems and acloswe IX me - a".d syst="s.
As r ~'es d by you S~~
eigh~ cori s o
t~'s let er an<
enlosur s
a=e be'rg st~plied ccr you-use.
LS. Dere are any questions rega "'ng th's mat ia'l, please contact us.
Ve y ~~~v yours,
+~4Q-,
L. D.
Enclosure I Mechanical Ecuipment
& Systems The folloving is a list of mechanical equipment and systems; and the tentative dates vhen RG&E plans to submit additional information on geese items.
Essential service vater pumps:
The pump specification, outline dzavings and foundation dravings are enclosed.
Verification of h
the installation and seismic integrity vill be submitted about October 1.
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Component cooling vater surge tank:
The tarmac and foundation drawings are enclosed.
Verification of the installation and seismic integrity v'll be provided about September 1.
Component cooling vater heat exchanger:
The heat exchanger and support dravings are enclosed.
Verification of the installation and seism'c integrity vill be provided about September 1.
Diesel generator air tanks:
C The tank and found'ation dravings vill be provided~
about July 15.
Verification of the installation and seismic integrity vill be submitted about Sept reer 1.
Boric acid storage tank:
The tank and foundation dravings vill be provided'bout July 15.
Verification of the installation and seismic integ=ity villbe. submitted about
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6.
Refueling eater storage tank:
The tank and foundation drawings vill be submitted about Julv 15.
Ver'fication of the installation and se'smic integrity vill be provided about September 1
7.
Motor op rator valves (electric/air) on lines
< 4" diameter:
8.
Det'ails of a typical installation and verification of<.seismic integrity vi13. be provided about September 1.
Prima~ equipment insid containment (zeactor coolant pump, pressurizer, steam generator,and control rod I
drive mechanism):
Equipment drawings, suvziary of equipment seismic analysis and ecuipment support dza ngs vill be subm'tted about July 15.
Verification of support insta'aticn and seismic integ ity'illbe provided about September 1.
9.
Interaction of seismic and non-seismic equipment (VAC ab'ove panel in d'esel generato" room and steel platform over oil line to feed pump in diesel generator room):
Verification of the insta, lation and seismic 10.
integrity of these 2 items vill be provided about Septenker l.
r V'-8 system dzwazic analysis (inside containment)
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C aY Three cop'es of Gilbert Associates draving C-381-354 J
Sheet 1, Rev'sion A vere forwarded to Nr. K. Jabbour JN on June 29, 1979 by express mail.
That drawing sho~s the basic as-built geometry of tho "A" R~R
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~ r system inside containment.
Ad"'tional cop'es of the piping draving, pipe s
"p-ert drawngs, and pip'ng system design.data necessary for analysis of the as-built conditions' vill be submitted about July 15.
I Naia Steam sysiem dynamic analys's (ins'de conta'nment):
- h as-built piping isomeirics of the B Na.'n Steam 1'r,e inside containment, suoport dravings, and p'oing system design daia necessary for analysis of 'Se as-huil condition vill be suhm'"ied about August. 1.
~ 12.
Ch mical analvses The and Volume Conirol System - equivalent siatic (ouiside conia'ament) as-built. piping isomeirics, supoort dravings,'iping sys"em d sign data for a portion of the CVCS system vill be suhmiited about Aug'st L.
Nemhers of the hRC Staff indicated in our June 12 13.
meeting that ihis information should be suhmitied insiead of the daia specified in the meeting minuies of the site visit of April 10-11, 1979
~
Sample field run of 2" piping:
Cooi s of an as-built piping isometric, support.
- dravings, and system design data vill be submitted I
about Augus 1.
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Lawrence Livermore National Laboratory September 11, 1981 SH 81-247 Docket t50-244 FIN A0415 g
TAN/mg 0131m Enclosure Thomas A. Nelson Structural Mechanics Group Nuclear Test Engineering Division Hr. William T. Russell, Branch Chief Systematic Evaluation Program Branch Division of Licensing Office of Nuclear Reactor Reg.
Washington, D.C.
20555
Dear Bill:
I have enclosed information regarding the seismic capacity of a reactor coolant pump at the Ginna plant.
While this open item has not been completely
- resolved, a rather simple action is proposed which should close this item.
Sincerely,
~ WesitycfCalfonia ~ P Q Bar 808 L'am Caflonia 94550 ~ Telephone(4f5) 422-f 100 ~ Twx910486-8339 BALLLLVMR
STRUCTURAI.
mECHAlllCS ASSOCIATES
~KKZ5~&3k A c alit. corp.
5160 Birch Street, Newport Beach, Cafif. 92660 (714) 833-7552 August 27, 1981 SMA 12205.20 Mr. Thomas A. Nelson L-9Q}
Lawrence Livermore Laboratory Nuclear Test Engineering Division P. 0. Box%08 Livermore, Califo'rnia-94550
Subject:
Resoluti.on of Open I'tems on Ginna Equipment I
Dear Tom:
In my June 22 letter, there were still two open items for Ginna equipment which were:
1.
Control Rod Drives and.Supports 2.
Primary Coolant Pump This letter is to advise you of the current status of these items.
George Wrobel of Rochester Gas E Electric QGE} Company arranged for me to talk directly to Robert Kelly of Westinghouse, who was responsible for providing much of the Ginna equipment seismic documentation.
Regarding loading documentation on the CRD assemblies, R. Kelly will send his analysis of CRD housing and seismic support loading for 'a 0.8g static coefficient loading.
A quick review of the analysis should resolve all outstanding issues on the CRD system.
A legible copy of the primary coolant pump report was transmitted to me from RGE.
Conclusions from the review of that report are contained in the attachment, Tom Cheng suggested that I be present at an upcoming open items meeting for Ginna in September.
In view of the current status of open items for which I have been responsible, I don't feel it necessary for me to attend;the meeting.
After my conversation with R. Kelly of Westinghouse, I feeI that.
the CRD and primary coolant pump issues will be resolved very quickly and can be handled without a meeting.
It is therefore recommended that these
'tems not be a topic of the meeting.
f Very truly yours, STRUCTURAL MECHANICS ASSOCIATES, INC.
iZQ Robert D. Campbell Project Manager RDC:lca cc:
T. Cheng (NRC)
ATTACHMENT'EACTOR COOLANT PUMP SEISMIC DESIGN REYIEM The original stress report submitted for review, was illegible.
A more legible copy was submitted to SMA by RGE on August 17, 1981, and the following conclusions can be reached from review of that submittal.
'l.
1.
The pump was analyzed for a 0.8g horizontal static coef-ficient and a 0.54g.vertical static coefficient.
As reported in NUREG/CR-3.821, Ae 7" damped peak spectral acceleration for both horizontal directions is 0.55 g's resulting in a vector sum of 0.78 g's.
- Thus, the equiva-lent static coefficient used in the original analysis is conservative by a small margin.
2.
All stresses calculated fdr the 0.8g H and 0.54g V static coefficients are within allowables designated for the original design basis.
3.
Analytical methods used in the design analysis are reason-able except in the case of pump nozzles.
4.
The pump nozzles analysis is unrealistic and inadequate for the following reasons:
'I Z
a.
Only the straight pipe portion of the nozzles were evaluated.
Local membrane stresses in the pump casing were not computed.
ATTACHHEHT Continued 0) b.
The derived pump nozzle loads have no resemblance to actual achievable loads.
The pump was assumed to be supported by the piping for purposes of deriving nozzle loads.
This is probably highly conservative but not necessarily so.
Actual piping reactions are available and should be used in an evaluation of the pump case.
A conversation between R. Kelly of Mestinghouse and R. Campbell of SMA revealed that:.
I 1.
The San Onofre pumps are very similar to the Ginna pumps and that a detailed finite element analysis was conducted for the San Onofre Units for a specified set of nozzle loads.
2.
A loop analysis of Ginna was conducted by Mestinghouse for seismic loading.
Actual pump nozzle loads'are obtainable
~ from the analysis.
Recommended Actions RGE should have Westinghouse make a comparison of Ginna vs San Onofre pump casing geometry and pump nozzle loads and scale resulting stresses from the San Onofre pump analysis for Ginna nozzle loa'ding conditions.
The load/stress comparison and a comparison of nozzle and
'asing geometry should be sufficient to demonstrate seismic capability of the Ginna primary coolant pumps.
~ Q STRUCTURAL mRCHA{llCS ASS OCl ATES 5t 60 Erch Street, Newport Beach, CaN. 92660 (714) 833-7552 SMA 12205.20 June 22, 1981 Mr. Thomas A. Nelson (L-90)
Lawrence Livermore Laboratory Nuclear Test Engineering Division P. 0. Box 808 Livermore, California 94550
Dear Tom:
In the Reference 1'ubmittal regarding review of open items on Robert E.
Ginna Nuclear Power Plant, there were three (3) open items still remaining.
1.
Control Rod Drive Mechanism 2.
Reactor Coolant Pump 3.
Steam Generator Tube Supports t
Tom Cheng of the USNRC contacted me on June 16, to inquire ff use of a.
site specific spectra for Ginna, in lieu of the regulatory guide spectrum anchored to 0.2g, would eliminate the oustanding items. 'I.indicated that for two of the three outstanding items, lack of information, not marginal stress conditions, was the principal concern.'s a result of the conversa-tion, I have reexamined the thr ee open items and can eliminate the steam generator tube support overstress problem by using Level 0 Service (faulted condition) allowable stresses and scaling up the original analysis results to applicable floor spectral accglerations.
The control rod drive and primary coolant pump items. still remain open but should be easily resolved with submittal of the necessary information.
The remaining action items and resolution of the steam generato'r tube support stress problem are sum-marized in the attachment.
Very truly yours, STRUCTURAL MECHANICS ASSOCIATES, INC-Robert D. Campbell Project Manager RDC:lca Attachment cc:
J. Stevenson
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lg ATTACHMENT
SUMMARY
OF OPEN ITEMS FOR ROBERT E.
GINNA NUCLEAR POMR PLANT AS PART OF THE SEP PROGRAM CONTROL ROO DRIYE MECHANISM The submittals provided by the licensee, References 2 and 3, do not contain a correlation betwee~
loading used in the analyses and accel-erations at the RPY support.
A conclusion regarding seismic resistance of the CRO system cannot be reached without such a correlation.
Mesting-house should be able to supply the necessary information to Rochester Gas.
and Electric.
REACTOR COOLANT PUMPS The Reactor Coolant Pump Stress Report submitted for review, Reference 4, is illegible due to poor reproduction quality.
A legible report needs to be submitted for review.
STEAM GENERATOR TUBE SUPPORTS Section 16 of Reference 5 documents the stress analysis of Series 44 steam generator internals for seismic loading.
A conservative analysis of the tube supports resulted in a primary membrane stress of.
,7900 psi for an equivalent static coefficient of 0;19g horizontal.
The stress was calculated in the ligaments between the tube holes and circu-lating holes in a local area adjacent to a wrapper channel.
The analysis.
'conservatively ignored a redundant load path from the. edge of the tube support to the tube holes.
Reference 6, submitted for review, is an update analysis of the Series 44 steam generator but does not address the
tube support for horizontal loading.
Neither Reference 5 or 6 describe the dynamic characteristics of the steam generator and its internals nor identify the material of construction for the tube support plates.
)
Reference 7, obtained during the SSMRP program, documents a
generic dynamic analysis of the Series 51 steam generator for varying support stiffnesses and locations.
The Series 44 steam generator is similar, but smaller.
The fundamental frequency of the Series 51 steam generator ranges from 4.8 to 9.6 Hz depending upon the support stiffness and location.
The fundamental mode is predominantly translation and rocking of the steam generator shell.
A generic response spectrum that is flat through most of this frequency range was used to compute response accelerations and loadings in the steam generator.
A review was conducted of Reference 7 to determine the degree of coupling between the steam generator shell and the internal structures and to establish the validity of considering the steam generator as a
SDOF system for estimatinq an appropriate equivalent, static coefficient for evaluation of the tube support plates.
It was concluded from review of Reference 7 that at the most critical support location, as determined in Reference 5, that the tubes and tube supports would accelerate as rigid bodies with the shell;
- thus, using the spectral acceleration from the Ginna spectrum for a funda-mental frequency of about 5 Hz is a reasonable approximation of an equiva-lent static coefficient to use for evaluation of the tube supports.
From the response spectra in Reference 8, for 7X darrping, the maximum spectral accelerations at 5 Hz are 0.58 g's in each of two ortho-gonal directions.
Combining the two directional accelerations, the resulting maximum vector is 0.82g.
Using this value and scaling the ligament stress computed in Reference 5, the resulting ligament stress is 34,095 psi.
Reference 7 indicates that Series 51 steam generator internals are constructed of SA 285-Grade C carbon steel.
A coayarison of the allowable stress for this material at the design temperature of 556 F
to the allowable stress quoted for the Series 44 steam generator tube
supports in Reference 5, indicates that the Series 44 tube supports are also constructed of SA 285-Grade C or equivalent.
This material and a
design teayerature of 556 F are used as a basis for establishing, allowable stresses for the Safe Shutdown Earthquake.
The tube supports are considered to be Class 1 plate and shell type cooyonent supports and the allowable primary men@rane stress is computed for Level D Service from Appendix F of the ASME Code, Reference 9.
The allowable stress fs the greater of 1.5 Sm and 1 2 Sy Fol SA 285-Grade C material, 1.2 Sy governs and the allowable stress is 27840 psi.
Note that the original design criteria limited the tube support stress to S.
Comparison of the calculated and allowable str ess for Level D
Service-results in a 22% overstress condition. lf site specific spectra anchored to 0.172g are considered in lieu of regulatory guide spectra anchored to 0.2g, the calculated stress deere'ases.
Decreasing the calcu-lated stress by the ratios of the site specific peak ground acceleration divided by the 0.2g peak ground acceleration used to generate floor
. spectra, the resulting ligament stress is 29,320 psi.
This is still about 5.3% over the Level D Service allowable of 27,840 psi.
In consideration of the conservatism inherent in obtaining the calculated
- stress, the computed 5.3% overstress condition is considered acceptable for several reasons.
1.
The site specific spectrum envelope has lower spectral accelerations in the frequency range of the containment structure than the regulatory guide spectrum if both are anchored to the same peak ground acceleration.
Conse-quent'Iy, the in-structure response spectra will be lower than those in Reference 8.
2.
The static analysis from Reference 5 did not account for'he r'edundant load path between the outside diameter of the tube support plate and the outer row of tube holes. 'he degree. of conservatism could not be evaluated since perti-nent dimensions are not provided in Reference 5.
The degree of conservatism is certainly greater than 5X though.
3.
The evaluation considered the applicable acceleration to be the vector. sum of the two orthogonal directional acceler-ations.
This assumes that both directional responses are in phase and that the resulting vector is aligned in the worst direction.
4.
The in-structure response spectra were peak broadened
+155 and smoothed so that the resulting spectra are essentially flat from 2-1/2 to 9 Hz, which covers the range of funda-mental frequency for the steam generators.
Items 3 and 4 are consistent with current regulatory criteria and are prudent conservatisms to cover many of the uncertainties in the simplified treatment of the tube support.
Recommended acceptance is, therefore, based on the conservatism of Items 1 and 2 being sufficient to overcome the estimated 5.3C overstress condition.
Further analyses or submittals from the licensee are not considered necessary.
1
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REFERENCES SMA letter, R. 0.
Campbell to T. A. Nelson, Review of Cpen Items Resulting fram Seismic Review of the Robert E. Gfnna Huclear Power Plant as Part of the SEP Program, 4 May 1981.
2.
High Speed Control Rod Drive Stress Analysis June 26, 1968.
3.
Control Rod Drive Mechanism Seismic Frame Calculations, August 13, 1968.
4.
5.
Static Seismic Load Stress Analysis, Model RGE Pump (V-11001-81),
July 30, 1968.
Vertical Steam Generator Stress
- Report, Mestfnghouse Electric Corporation, Tanqa Division, April, 1969.
~(
6.
MTO-SM-75-028, 44 Series Steam Cenerator Stress Report, External Load Analysis Update, May, 1975.
7.
Stress
- Report, 51 Series Steam Generator, Generic Seismic Analysis, Mestfnghouse Electric Corporation, Tanya Division, December, 1974.
8.
NUREG/CR-182l, Seismic Review of the Robert E. Ginna Nuclear Power Plant as Part of the Systematic Evaluation Program, 15 November 1980.
9.
ASME Boiler and Pressure Vessel Code,Section III, Nuclear Power Plant Components, Appendices, 1980.
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STRUCTURAL mECHAnICS ASSOCIATES
~ EKK~M A c e I If. c os o.
5160 Birch Street, Newport Beach, Calif: 92660 (714) 833-7552 SMA 12205.20 May 4, 1981 Mr. Thomas A. Nelson (L-90)
Lawrence Livermore Laboratory Nuclear Test Engineering Oivi'sion P.O.
Sox 808 Livermore,California 94550 Oear Tom:
SMA has reviewed the package of documents transmitted with your April 15 letter addressing op'n items on Ginna.
Our comments and recommended action are contained in the enclosure.
Very truly yours, STRUCTURAL MECHANICS ASSOCIATES, IHC.
p~
Robert 0. Campbell Project Manager ROC:mw Enclosure cc:
J. Stevenson w/encl.
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REVIEW OF OPEN ITEMS RESULTING FROM SEISMIC REYIEM OF THE ROSERT E.
GINNA NUCLEAR POMER PLANT AS PART OF THE SEP PROGRAM Reference 1 documents a review conducted by the Lawrence Livermore Laboratory and their consultants on the seismic adequacy of the Robert E. Ginna Huclear Power Plant.
Conclusions of the adequacy of the several items in the HSSS system were based upon summary information provided; however, the sources of the summary informa-tion were not available for independent review'.
Those items listed in Section 5.4 of the report as components for which seismic design analyses.
have not been independently verified are:
Reactor Control Rod Orive Reactor Vessel Supports Steam Generator Reactor Coolant Pumps Pressurizer and its Supports References 2 through 7 were provided to SMA in response to the above identified open items.- The following summa'ry and conclusions resulted from SMA's review'of the submittals.
1)
Control Rod Drive Mechanism In reference 7, the allowable bending moment in the CRDM due to seismic loads is developed.
This report does not,
- however, provide 'a correlation between bending moments and acceleration levels.
- Thus, no conclusion can be reached on the basis of the submittal.
r Reference 2 contains a stress analysis of the control rod drive support structure.
The analysis provides s.resses as '
function of a static load "P" in pounds.
There is no corre-lation between this static load and acceleration level. Therefore, a conclusion on the seismic capability of the support structur'e cannot be reached based upon the submittal,
SMA's experience with the SSMRP reference plant, which uses 106A full length control rod drive mechanisms, indicated that a large margin of safety exists for a 1.15g spectral accel-eration loading condition and we would not anticipate a
seismic problem with the Ginna CRDM.
Recommended action -
CROM loads documentation applicable to Ginna need to be supplied to SMA for final resolution.
2J Reactor'Vessel Su orts Documentation verifying the seismic adequacy of the reactor vessel.'supports was not submitted.
Based upon SMA's SSMRP experience for nozzle supported RPV's the seismic induced stresses in the nozzles and adjacent shell are very small and the governing element for RPV support is the concrete
- shield wall.
The shield wall was considered in Ref.
1 to be adequate to withstand the 0.2g SSE.
3 j.
Steam Generator Reference 6 contains a 1969 static analysis of the Series 44 steam geneator.
The most critical. area due to seismic load-ing identified in this report is the tube support baffle ligaments which are stressed past yield for a 0.38g horizon-tal static load.
The SEP revised spectra result in an SRSS response of 0.85g horizontal which will increase the stresses by a factor of 2.24.
- Thus, based on the submittal, the tube support baffles are overstressed for the 0.2g SSE.
.The static analysis at these tube support baffles was.done quite conser-
'atively,however, 'and a more rigorous 'analysis will most likely result in lower state of stress.
In 1975, an update on the series 44 steam generator was con-ducted (Reference 4)'ut the tube support area was only evaluated fot vertical seismic loading and not for the
C horizontal seismic loading.
Thus, the results of Reference 4 cannot be used to update the results of Reference 6 in the area of concern.
Recommended Action -
Documentation evaluating the tube support baffle ligaments for horizontal seismic loading should be submitted for review, 4)
Reactor Coolant Pum s
Reference 3 summarizes the stresses induced in the reactor
.coolant pump by a 0.8g horizontal and a 0.54g vertical load-
,ing.
The reported stresses are below'the ASME Code allowables, but. SMA is unable to 'evaluate the model or the analysis due to poor reproduction quality of Reference 3, lf the static
..analysis of the pump can be shown to be valid, then the stresses due to the revised Ginna spectra loading will be
'ess than those contained within Reference 3,'nd thus accept-able.
Recommended Action - A legible copy of Reference.3 should be provided to SMA for review.
5k Pr essurizer Reference 5
contains a 1969 stress report of an 1800 cubic
. foot pressurizer.
All pressurizers from 800 to 1800 cubic feet with cast and'fabricated heads utilize the same support skirts, thus conservative generic analysis was conducted for the heavier 1800 cubic foot models.
Based on this report, the loads resulting from the new Ginna spectra will. cause an over-stressed condition within the support flange.
This is a very conser vative conclusion, however, since the Ginna pr essurizer
,,...$ s a smaller 800 cubic foot model plus, the flange analysis
.. )tself was.very conservatively conducted using a
beam bending model in lieu of a more rigorous finite element model.
The 1973 pressurizer report refer red to in Reference 1
and the pressurizer summary stress report (reference 45 of Refer-ence 1) were not supplied to SHA for review; thus, the ade-quate capacity for a 6.7g equivalent static load portrayed in reference 45 of Reference 1 cannot be verified.
Later analysis (References 8 and 9.) obtained in the SSHRP program. for the Series 51 1800 cubic foot cast and fabricated head pressurizers showed the supports to be acceptable for a 0.96g. horizontal and 0.64g vertical loading condition.
This later analysis utilized a finite element ~odel of the skirt and flange as opposed to the conservative beam theory-used in reference 5.
Recommended Action - Ho action required;
-The pr essurizer stress analysis within references 8 and 9 and the statements in References 5,
8 and 9 that the support skirts are identical for the 1800 cubic foot and the 800 cubic foot designs sub-
- stantiate the adequacy of the Ginna pressurizer for a 0.2g.SSE.
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.REFERENCES b
1)
NUREG/CR -,1821 (UCRL.- 53014} Seismic Review of the Robert E
Ginna Nuclear Power, Plant as Part of the Systematic Evaluation
- Program, Novemb'er 1980.
2)
Control Rod Drive Mechanism Seismic Frame Calculations, 8/13/68.
'3)
Static Seismic,.Load Stress Analysis,Model RGE Pump (V-11001-81},
7/30/68.
4}
44 Series Steam -Generator External Load Analysis Update,5/75.
5)
RGE -Pressurizer Stress Report,9/69.
6)
Vertical Steam Generator Stress Report, 4/20/69.
7)
High Speed. Control Rod Drive Hechanism Stress Analysis, 6/26/68.
8) 51 Series Pressurizer Generic Seismic Analysis, Mestinghouse Electric Corporation, August 1974.
9) 51 Series'ressurizer Support Skirt and'lange Analysis, Mestinghouse'lectric Corporation, May 1974.
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