ML19242C952
| ML19242C952 | |
| Person / Time | |
|---|---|
| Site: | Vallecitos File:GEH Hitachi icon.png |
| Issue date: | 07/09/1979 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | Darmitzel R GENERAL ELECTRIC CO. |
| References | |
| NUDOCS 7908140242 | |
| Download: ML19242C952 (6) | |
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f* MCue yk UNITED STATES
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y ^ c(3),i NUCLEAR REGULATORY COMMISSION
'.; E WASHINGTON. D. C, 20555 y
k8 July 9, 1979 Docket No.:
50-70 Mr. R. W. Darmitrel, Manager Irradiation Processing Product Section General Electric Ccmpany Vallecitos Nuclear Center P. O. Box 460 Pleasanton, California 94566
Dear Mr. Darmitzel:
Based on our review of the seismic design inforaation submitted in res;:ense to the Order to Show Cause dated October 24, 1977, we have determir,ed that the additional infor ation identified in the enclosure is necessary to ccm-plete our review. Please prcvide your response by July 31, 1979.
Sincerely, j<
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Robert W. Reid, Chief Operating Reactors Branch 14 Division of Cperating Reactors
Enclosure:
Request for Additional Information cc w/ enclosure:
See next pege
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f General Electric Company cc w/ enclosure (s):
California Department of Health ATTN: Chief, Environmental Radiation Dr. Harry Foreman, Member Control Unit Atomic Safety and Licensing Board I
Radiologic Health Section Box 395, Mayo l
714 P Street, Room 498 University of Minnesota
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Sacramento, California 95184 Minneapolis, Minnesota 55455 Honcrable Ronald V. Dellums Ms. Barbara Shockley ATTN: Ms. Nancy Snow 1890 Bockman Road General Delivery, Civic Center San Loren:c, California 94580 Station Oakland, California 94604 Advisory Committee on Reactor i
Safeguards Friends of the Earth U. S. Nuclear Regulatory Ccmmission ATTN:
W. Andrew Baldwin, Esquire Washington, D. C.
20555 Leg:'l Director 124 Spear Street San Francisco, Cali ornia 94105 Jed Somit, Esquire 100 Bush Street Suite 304 San Francisco, California 94iO4 Edward Luton, Esquire, Chairman Atcmic Safety and Licensing Board U. S. Nuclear Rerulatory Ccemission Wasningtcn, D. C.
2C555 Mr. Gustave A. Lin snberger, Member Atcaic Safety and Licensing Scard U. S. Nuclear Regulatory Ccemission Washing:cn, D. C.
2L555 George Edgar, Esquire Morgan, Lewis & Sockius 1800 M Street, VJ Washing cn, D. C.
20036 S
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ENCLOSURE 2 GENERAL' ELECTRIC TEST REACTOR (GETR)
COCKET N0 50-70 REVIEW CF REPCRTS SUSMITTED IN RESPONSE TO NRC CRCER TO SHOW CAUSE DATED 10/2a/77 REQUEST FOR ADDITIONAL INFORMATION ENGINEERING BRANCH DIVISICN OF OPERATING REACTORS 1.
Discuss hcw extensive the cracking of the floor slabs is due to the overturning moments. Verify that there is no impact on :afety related components due to saalling or cracking.
2.
Verify, with a cetailed discussion addre; sing generated missiles, pipe and support deformation capabilities, structural stiffness and strength degradation, and reactor building leaktightness integrity; that the extensive cracking and failure resulting from surface rupture will not impact safety related components or systems.
3.
Verify that the extensive failure resulting from surface rupture will not ccmpromise the integrity of the interior radia~. wall, the circumferential wall connection or the ability of the containment to suoport the required loadings witnout impacting the integrity of any safety related components, systems or equipment.
Discuss the extent of the predicted containment damage in detail to substantiate your statement that its defcrmations are acceptable.
Specifically, adpress the possibility of a punch-ing mode of failure.
4 Justify the material croperties used for the soil spring model, including damping values and poisson's ra-io. To what level in the actual subgrade do these values correscend? Discuss the impact on the factors of safety provided fcr the forces and ficer accelerations if tne "most realistic case" of subgrace parameters (as describec in your Phase 2 Seismic Analysis Rescrt) is not present.
That is, what safety censiderations are provitec in tne event nat the actual response is greater nan precicted by Case i parameters?
5.
Verify that the calculations of tne sliding and everturning resistance have acccunted fcr tne recuction of the weignt of tne building due to vertical uplif t.
If uplift due :: ver tical excitation is not consicered, justify the accropriateness of tne unccnservative analysis.
5.
Verify that the maximum sliding disclacement of 1.3 inches results in no failure of safety related pioing, ccmponents, or equipment.
7.
In your Seismic Analysis of Reacter Euilding - phase 2 report, you state that "tnere is no structural contiruity between the foundaticn mat and the rest of the reactor building." Cescribe hcw this is represented in the mathematical mcdel.
Provice tne prcperties of the -' ember between the fcundation mat and the basement slab, and c' ascribe hcw they were determined.
(Provice the tems of the local stif# ness matrix).
Alsc, verify that th2 results in Tabic 2-9 for slicing at tne interior concrete -
fcundation slab interface reflect these preserties and the bounding case ccnsidering res;cnse variations due tc all potential variations ccnsidered in your analyses.
If relative acticn is predicted, discuss the imcact en the results of your analyses.
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_2 3.
Describe the procedures utilized in the determination of the soil spring boundary conditions in your Model B nonlinear analysis.
Also demonstrate that this type
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of representation of the subgrade is appropriate considering soil depth, layering, etc.
Discuss the acceptability of your modeling as opposed to using the current finite element techniques.
9.
Provide a d'escription of the core structure displacements associated with the yielding and settlement of the foundation mat.
Verify that these displacements were considered in the design of the core structure and safety related comconents, systems and equipment, and that the integrity of these safety related items is not compromised.
10.
In the post-offsat analyses, provide the acceptance critaria for the seismic dis-placements and forces. Also, provide the factors of safety against sliding and overtur.ing for this condition and sumarize how they were determined.
11.
Describe in detail the methods by which the allowable shear and tensile stresses were deternined frca the referenced test data.
Justify the correspondence between the GETR walls and these test samples since the PCA tests were flarge reinforced s;.:cimens. Verify that the stresses calculated via your finite element representa-t" ;n of the GETR are directly ccmparable to your stated allowables.
Provide the ba ses for your statements.
Include a discussion of hcw construction joints were crasidered in your evaluatien and the possibility of degradation of these joints dt.
to water seepage weakening the shear transfer pcross the joint.
- 12. k ify tna.t the effects of the primary piping which is ancnored to the concrete structJre,ncve been considered in the seismic analysis ans design of the concrete structure.
13.
Discuss the pr:cedt.ces used to detsrmine the iccaticn of imcact of the cask drcp cn tne canal sia:: wnicn cr: duces the maximum ctent en tne slab. Provide this ccrent, and verify that tne slab is cacable Cf witnstanding tnis load. Als0, ver'fy : Pat spalling cf ccncrete due c the cask crc: :n tne canal slab dces
.ct im::10: any safety related items.
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!f a cask crc; r::sults in damage to ne liner and cracking of the concrete, verity rat ace:aate canal sa:er is maintainec.
15.
'ecvide justificaticn that a ncn-mechanistic icwer head nozzle ru;:ture cccurs with ufficiantly Icw ;;rccaMiity to assure the ac:e::tability of the consequences of this event. Provide :,icilar justificatica fcr rupture of the reactor pcol.
Include a iscussicn, in tems Of radiaticn levels and stress levels, verifying that no em:rittlement cccurs, sucr as :: preclude costulating the above f ailures.
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I:acify.he maxiTum inner ar.d outer fuel storage tank displacements, and verify that
-hese maximum ciscie:eman:3 are cctained using the Tcre realistic configuration wnere sli-ing is ;;enitted.
!srify tnat nese tisclacements do not adversely imcact tne safe y related f.rc;ic.s of thc tanks. Discuss the : nsecuences of a 1.1 incn ircer tank maxim.m c cki ;
fic ti:n. AISO, verify that slicing Of the tanks CCes ct resul-in i cac :: tne :cral liner, w. x? d '.
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- 17.
In the inner fuel storage tank rocking analyses, provide justification for the use of a factor of 57*.' to reflect the energy losses and fluid inertia effects.
18.
In the inner fuel storage tank rocking analysis, describe hcw the 123.25 lb/in.
live load on the cuter tank was resolved into the concentrated loads applied at nodes 22, 23, 24, and 25.
- 19. Explain how the response spectra for three percent damping used in the se'smic analysis of'the primary cooling system and PR'l envelcp the respense spectra obtained for one percent damping, by a factor of 1.2.
(See ECAC-ll7-217.05, page 2-2) 20.
Cescribe the picing disclacements resulting frca the analysis of Run 1 & Run 2.
Provide the design and acctatance criteria for pipe displacements, and verify that the maximum displacements are within design alicwables. Also veri that seismic excitation dces not result in impact between piping systems anu any safety related equipment or compenents.
21.
Discuss how the effects of a surface rupture offset have been censidered, and verify that they will not ccapremise the integrity, of the primary c: cling system and reactor pressure vessel.
- 22. List the types of restraint antnorages used for the GETR piping and equipment, and describe the procedures used in the design of these anchorages.
'/erify that cyclic loads have been Ocnsidered, and describe and justify the anchor bolt and rock bolt cyclic load design requirements. Cescribe any inservice inspections which are planned for the bolts and justify the extent of the program.
- 23. '/erify that the piping restraints and anchors are in the :Orrect locations, as Jesigned.
21
'lerify that thermal loads and fluio transients were ::nsidered in the analysis arc testing of the valves.
25.
In ycur 3 ruc. ural Analysis of Third Ficar Missile :::act System, as ciscussed
- n ; age 17, ciscuss how and wny tne ncr-ai i::a t 1:20s are a:;iied :: One
- en-in the Z-directi:n, insteac cf the y-cirecti:n anica at; ears to be ::n-sistent with a resultant fcree ?x.
Also, discuss wny the lateral Icacing was ac;;iiec in the x-directicn. '/erify that a lateral 1:10 a:Cliac ;er;endicular
- the x-cirecti:n is not a Tcra critical case. Provide :ent ail:wable stresses, including buckling stresses, if a;;;r:criate, 00 verify that esign stresses are small. Explain tne inc nsistencies of Figures 5 and 5 notations.
25.
'lerify that maximum tensile f:rce in the :sse plate tolts due to lateral tent lcading with ; ward seismic mction (and nc ncr-al i :act Ita:ing) have been censide-ed and are within allowabies.
2).
Discuss the in-service surveillance programs wnich will :e concucted on all safety related c cacnents.
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- 23. Justify the acceptability of bolted base plates where the jam nut is placed inside of the main nut. Specifically, verify that the system will not fail at the jam nut when loaded due to vibratory motion, 'hus unlocking the main nut and allowing it to back off, i
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