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Sandia National LaboratoriesAlbuquerque. New Mexico 87185*

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May 2, 1983

Dr. James F. CostelloMechanical / Structural Engineering BranchMail Stop NL 5650U.S. Nuclear Regulatory CommissionWashington, DC 20555

Dear Jim:

Enclosed is a quick-look report for the test of model SC-1.Some data reduction was conducted during the test. Additionaldata reduction is underway. As soon as possible, acomprehensive data set will be distributed to EPRI and theindustrial organizations that have expressed a willingness tomake analytical predictions on models SC-2 and SC-3

Sincerely,

[d edx uT -

T. E. BlejwasSystems Safety TechnologyDivision 9442

Copy to:9440 D. A. Dahlgren ,

9442 W. A. von Riesemann9442 D. S. Horschel9442 J. Jung9442 R. T. Reese9442 R. L. Woodfin9442 File 1047.013

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''* ~ Quick-Look Report

Test of Model SC-1

Containment Integrity Program '

!

Model Description

Model SC-1 was a baseline (clean shell) configuration for the'

small steel containment tests. It consisted of a dome andcylinder attached to a " rigid" base. The cylindrical portionhad a height nad diameter of about 43 inches. The thickness

,

!

.

was about 45 mils. The spun dome was hemispherical with athickness that varied from about 40 to 70 mils.

Model Pressurization,

The model was internally pressurized in a step-wise fashion;

; using nitrogen gas. Pressure steps to 20, 40, 50, 60, 70 and80 psig were specified to the pressure control system. Model4

behavior was interpreted during the test to be primarilyelastic. Because significant plastic behavior was noted at 90;

psig, smaller increments were requested. Steps to 93, 96, 99,4

102 and 105 psig were taken with several minutes required ateach step to allow the godel and pressure controller tostabilize. A step to 150 psig was requested. After about 3

f minutes the pressure in the model was 110 psig but the modelcontinued to grow (a sequence of events seen in previoussteps). The pressure dropped about 3 psig in slightly morethan two minutes despite the apparently correct functioning ofthe pressure controller. At this point, the pressure began todrop rapidly (about 10 ps4 per minute). When the pressurereached about 90 psig, a hold of the pressure controller was

,

requested (preventing further nitrogen input to the model).i

The r.odel depressurized from 90 to 20 psig in about 10 minutesand continued to zero pressure.

Structural Behavior;

During pressure eps up to 60 psig the model behaved in anessentially lin elastic manner. The circumferential strainin the cylinder as measured as 0.08, 0.07, 0.08 and 0.08% at'

vertical positions of 6.8, 10.8, 14.8 and 18.8 inches above thebottom of the base ring. The analytically predicted elasticstrain at 60 psig is 0.09% at all locations. At 80 psig thestrains at the same locations were 0.18, 0.26, 0.23 and 0.17%respectively. Thus, because the strains more than doubled,i

some plastic flow occurred during the step from 60 to 80 psig.Unfortunately the displacements were too small at both pressurelevels to obtain reliable results.

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Nearly five minutes was required for stabilization of pressure-

and strain during the pressure step to 90 psig. Afterstabilization, a mid-height cylinder strain of 1.6% wasrecorded. The maximum radial displacement was measured as 0.34inches. Clearly, general membrane yielding occurred at orbelow 90 psig. The analytically predicted membrane yieldpressure is about 95 psig.

After membrane yielding, significant time (as much as 20minutes) was required for stabili=ation at each pressure. Themid-height circumferential strain was 1.9, 2.4, 3.0, 3.8 and4.9% at pressures of 93, 96, 99, 102 and 105 psig,respectively. The maximum radial displacements were 0.35,0.50, 0.63, 0.78, and 1.04 inches, respectively.

Because stabilization at 110 psig did not occur prior todepressurization, a full set of data was not recorded. Howeverafter full depressurization to zero, a mid-height residualstrain of 6.0% was measured. The maximum residual radialdisplacement was 2.36 inches.

Post-Test Examinatiof.

An examination of the model after testing revealed a tear nearthe mid-height of the cylinder. The tear is about 1-1/4 incheslong and 1/16 inches wide. The tear is adjacent to themeridional seam weld. In particular, the tear appears to be inthe heat affected zone next to a portion of the weld that wasrepaired several times. An examination of the interior of themodel revealed that excess repair weld material has been groundaway. Some cylinder parent material was also ground; the areaadjacent to the tear is thinner due to the grinding.

When the bottom head was r'emoved from the base fixture, theLVDT bracket was noted to be slightly loose. The interior ofthe model has marks from the ends of the LVDT's (apparently)moving back and forth during movement of the model to thesite. One LVTD was pressing hard enough against the cylinderto cause some gouging.

Continuing Efforts

Data reduction for the test is continuing. Displacementmeasurements will be calculated from readings taken withtheodolites during the test. Expected data includes strainsand displacements (both LVDT and theodolite) as functions ofpressure and location on the model.

A more complete examination of the tear in the model will bemade. However, repair of the model does not appear to bereasonable at this time. Consideration will be given tofabricating one or more additional baseline configurationmodels.

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Once all displacement measurements are available, an attempt-

will be made to determine the effect of the loose LVDTbracket. Because the measured movement is relative to the zeropressure readings, movement before testing would not haveaffected the readings during the test.

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Sandia National Lab 0ratoriesAlbuquerque, New Mexice 87185

.

May 2, 1983

Dr. James F. CostelloMechanical / Structural Engineering Branch-Mail Stop NL 5650U.S. Nuclear Regulatory CommissionWashington, DC 20555

Dear Jim:

Enclosed is a quick-look report for the test of model SC-1.Some data reduction was conducted during the test. Additionaldata reduction is underway. As soon as possible, acomprehensive data set will be distributed to EPRI and theindustrial organizations that have expressed a willingness tomake analytical predictions on models SC-2 and SC-3

Sincerely,

M ub- Tf

W' T . E. BlejwasS Systems Safety Technology

Division 9442

Copy to:9440 D.A. Dahlgren9442 W. A. von Riesemann

*

9442 D. S. Horschel94'' J. Jung9442 R. T. Reese9442 R. L. Woodfin9442 File 1047.013 -

.

.

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.

_ _ _ _ _ _ _ . _ _ _ _ - - - _ _ - -

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. _ . . - - . . . . _ - - - _ - . - _ . . -

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:

Quick-Look Report.

Test of Model SC-1

Containment Integrity Program|

Model Description

Model SC-1 was a bas'eline (clean shell) configuration for thesmall steel containment tests. It consisted of a dome'andcylinder attached to a " rigid" base. The cylindrical p'ortionhad a height and diameter of about 43 inches. The thicknesswas about 45 mils. The spun dome was hemispherical with athickness that varied from abcut 40 to 70 mils.

| Model Pressurization

The model was interna 12y pressurized in a step-wise fashion! using nitrogen gas. Pressure steps to 20, 40, 50, 60, 70 and

80 psig were specified to the pressure control system. Modelbehavior was interpreted during the test to be primarily.

elastic. Because significant plastic behavior was noted at 90psig, smaller increments were requested. Steps to 93, 96, 99,,

102 and 105 psig were taken with several minutes required ateach step to allow the model and pressure controller to

i stabilize. A step to 110 psig was requested. After about 3^

minutes the pressure in the model was 110 psig but the modelcontinued to grow (a sequence of events seen in previoussteps). The pressure dropped about 3 psig in slightly more

, than two minutes despite the apparently correct functioning ofj. the pressure controller. At this point, the pressure began to

drop rapidly (about 10 psi per minute) . When the pressurereached about 90 psig, a hold of the pressure controller wasrequested (preventing fur'ther nitrogen input to the model).The model depressurized from 90 to 20 psig in about 10 minutes:

! and continued to zero pressure.

Structural Behavior

During pressure jteps up to 60 psig the model behaved in anessentially ling-elastic manner. The circumferential strainin the cylinder was measured as 0.08, 0.07, 0.08 and 0.08% atvertical positions of 6.8, 10.8, 14.8 and 18.8 inches above the -.

bottom of the base ring. The analytically predicted elasticstrain at 60 psig is 0.09% at all locations. At 80 psig thestrains at the same locations were 0.18, 0.26, 0.23 and 0.17%respectively. Thus, because the strains more than doubled,some plastic flow occurred during the step from 60 to 80 psig.Unfortunately the displacements were too small at both pressurelevels to obtain reliable results. '

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Nearly five minutes was required for stabilization of pressure.

and strain during the pressure step to 90 psig. Afterstabilization, a mid-height cylinder strain of 1.6% wasrecorded. The maximum radial displacement was measured as 0.34inches. Clearly, general membrane yielding occurred at orbelow 90 psig. The analytically predicted membrane yieldpressure is about 95 psig.

After membrane yielding, significant time (as much as 20minutes) was required for stabilization at each pressure. Themid-height circumferential strain was 1.9, 2.4, 3.0, 3.8 and4.9% at pressures of 93, 96, 99, 102 and 105 psig,

; respectively. The maximum radial displacements were 0.35,0.50, 0.63, 0.78, and 1.04 inches, respectively.

Because stabilization at 110 psig did not occur prior todepressurization, a full set of data was not recorded. Howeverafter full depressurization to zero, a mid-height residualstrain of 6.0% was measured. The maximum residual radial'

displacement was 1.36 inches.

Post-Test Examination

An examination of the model after testing revealed a tear nearthe mid-height of the cylinder. The tear is about 1-1/4 incheslong and 1/16 inches wide. The tear is adjacent to themeridional seam weld. In particular, the tear appears to be inthe heat affected zone next to a portion of the weld that wasrepaired several times. An examination of the interior of themodel revealed that excess repair weld material has been groundaway. Some cylinder parent material was also ground; the areaadjacent to the tear is thinner due to the grinding. -

When the bottom head was removed from the base fixture, theLVDT bracket was noted to*be slightly loose. The interior ofthe model has marks from the ends of the LVDT's (apparently)moving back and forth during movement of the model to thesite. One LVTD was pressing hard enough against the cylinderto cause some gouging.

.

Continuing Efforts -

Data reduction for the test is continuing. Displacementmeasuremenu. will be calculated from readings taken withtheodolites during the test. Expected data includ2s strains

,

and displacements (both LVDT and theodolite) as functions ofpressure and location on the model.

A more complete examination of the tear in the model will bemade. However, repair of the model does not appear to bereasonable at this time. Consideration will be given tofabricating one or more additional baseline configurationmodels.

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Once all displacement measurements are available, an attempt*

will be made to determine the effect of the loose LVDTbracket. Because the measured movement is relative to the zeropressure readings, movement before testing would not haveaffected the readings during the test.

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Quick-Look Report

Test of Model SC-1

Containment Integrity Program

Model Description

Model SC-1 was a baseline (clean shell) configuration for thesmall steel containment tests. It consisted of a dome andcylinder attached to a " rigid" base. The cylindrical portionhad a height and diameter of about 43 inches. The thicknesswas about 45 mils. The spun dome was hemispherical with athickness that varied from about 40 to 70 mils.

Model Pressurization

The model was internally pressurized in a step-wise fashionusing nitrogen gas on April 21, 1983. Pressure steps to 20,40, 50, 60, 70 and 80 psig were specified to the pressurecontrol system. Model behavior was interpreted during the testto be primarily elastic. Because significant plastic behaviorwas noted at 90 psig, smaller increments were requested. Stepsto 93, 96, 99, 102 and 105 psig were taken with several minutesrequired at each step to allow the model and pressurecontroller to stabilize. A step to 110 psig was requested.After about 3 minutes the pressure in the model was 110 psigbut the model continued to grow (a sequence of events seen inprevious steps). The pressure dropped about 3 psig in slightlymore than two minutes despite the apparently correctfunctioning of the pressure controller. At this point, thepressure began to drop rapidly (about 10 psi per minute). Whenthe pressure reached about 90 psig, a hold of the pressurecontroller was requested (preventing further nitrogen input totne model). The model depressurized from 90 to 20 psig inabout 10 minutes and continued to zero pressure.

Structural Behavior

During pressure steps up to 60 psig the model behaved in anessentially liner-elastic manner. The circumferential strainin the cylinder was measured as 0.08, 0.07, 0.08 and 0.08% atvertical positions of 6.8, 10.8, 14.8 and 18.8 inches above thebottom of the base ring. The analytically predicted elasticstrain at 60 psig is 0.09% at all locations. At 80 psig thestrains at the same locations were 0 .18 , 0. 26, 0 . 2 3 a nd 0.17 %

respectively. Thus, because the strains more than doubled,some plastic flow occurred during the step from 60 to 80 psig.Unfortunately the displacements were too small at both pressurelevels to obtain reliable results.

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_ _. . _ . _ - _ _ . - .

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Nearly five minutes was required for stabilization of pressureand strain during the pressure step to 90 psig. Afterstabilization, a mid-height cylinder strain of 1.6% wasrecorded. The maximum radial displacement was measured as 0.34inches. Clearly, general membrane yielding occurred at orbelow 90 psig. The analytically predicted membrane yieldpressure is about 95 psig.

After membrane yielding, significant time (as much as 20minutes) was required for stabilization at each pressure. Themid-height circumferential strain was 1.9, 2.4, 3.0, 3.8 and4.9% at pressures of 93, 96, 99, 102 and 105 psig,respectively. The maximum radial displacements were 0.35,0.50, 0.63, 0.78, and 1.04 inches, respectively.

Because stabilization at 110 psig did not occur prior todepressurization, a full set of data was not recorded. Howeverafter full depressurization to zero, a mid-height residualstrain of 6.0% was measured. The maximum residual radialdisplacement was 1.36 inches.

Post-Test Examination

An examination of the model after testing revealed a tear nearthe mid-height of the cylinder. The tear is about 1-1/4 incheslong and 1/16 inches wide. The tear is adjacent to themeridional seam weld. In particular, the tear appears to be inthe heat affected zone next to a portion of the weld that wasrepaired several times. An examination of the interior of themodel revealed that excess repair weld material had been groundaway. Some cylinder parent material was also ground; the areaadjacent to the tear is thinner due to the grinding.When the bottom head was removed from the base fixture, theLVDT bracket was noted to"be slightly loose. The interi.or ofthe model has marks from the ends of the LVDT's (apparently)moving back and forth during movement of the model to thesite. One LVTD was pressing hard enough against the cylinde;to cause some gouging.

Continuing Efforts

Data reduction for the test is continuing. Displacementmeasurements will be calculated from readings taken withtheodolites during the test. Expected data includes strainsand displacements (both LVDT and theodolite) ds functions ofpressure and location on the model.

A more complete examination of the tear in the model will be'

made. However, repair of the model does not appear to bereasonable at this time. Consideration will be given tofabricating one or more additional baseline configurationmodels.

I!

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.

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,

i

once all displacement measurements are available, an attempt,

- will be made to determine the effect of the loose LVDTbracket. Because the measured movement is relative to the zeropressure readings, movement before testing would not haveaffected the readings during the test.

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ACT IV nY NTb YTb ci'e n y ro

sp a Lc(etc) spes, t,e c pys) po/pg 7,gg

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S 'ite & Fac. Pre p. R ~1 li . 23 I o 2. 7 6 (f.Q o 1oa

Inste - en% % n 2o li (2.t iB4 76 (i .h 3 is7

3 Exp. 27 2o(2.-D 57 y o (,73 (2 lo9Conducb

3+ eel models . -' l 7 4 (. 53 2.23 2 2 (.43 (M 34o

Conecch Models _.3 (.'D lo lo 62) Ito i204

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Stee\ - ab 15 tz (I.@ G4 49 (n.o) o 64

Stee) -3D 5 0 33 3/.l) 9 42.

Concreke 9 9(I.0 2'i 24 6'd O 24

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Managemen4.fepods, 2o I6(l.83 14o /07(I.8) 76 2 I S'Peer Revice, etc. -

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TOTAL 92 8 5-(11.9 )863 *is (s.2) 375 izze,

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Sandia National Laboratories !

date: June 17, 1983 Altiuquerque, New Mexico 87185

.

to: Containment Integrity Peer Review Group

from: R. T. Reese, 9442~

subjem: Peer Review Group Meeting, Sandia National Laboratories,Albuquerque, New Mexico, February 24-25, 1983.

Enclosed is a summary of this meeting including the discussionsand agreements / action items. Also enclosed are copies of theagenda and the transparencies used in the presentations. Someof the activities described on the schedule have changed sincethe peer review meeting. New dates will be announced later.

Enclosures1. Meeting Agenda2. Copies of Transparencies Used in the Meeting3. Y. R. Rashid, Review of the WASH-1400 Surry Containment

Assessment, Draft for Comments

Copy to:Peer Review Group; enc 1,2,3J. Costello, NRC; enc 1,2,3J. Burns, NRC; w/o encW. Anderson, NRC; w/o enc1523 D. B. Clauss; enc 1,2

E. H. Conley; enc 1,29440 D. A. Dahlgren;'w/o enc9442 W. A. von Riesemann; enc 1,2,39442 T. E. Blejwas; enc 1,2,39442 D. S. Horschel; enc 1,2,39442 J. Jung; enc 1,2,39442 R. T. Reese; enc 1,2,39442 W. A. Sebrell; w/o enc9442 File 1047.011

.

.

...

Enclosure 1,

AGENDA

Peer Review GroupContainment Integrity Program

'

February 24-25, 1983Sandia National Laboratories

Albuquerque, New Mexico

Thursday, February 24 Bldg 822, Room 8

9:00 a.m. Introduction Jim CostelloWalt von Riesemann (2a)

9:30 a.m. Program Overview Tom Blejwas (26.1)

10:00 a.m. Break

10:15 a.m. Small Steel Experiments

Models Tom Blejwas

Preliminary Test Ron Woodfin

Analysis Dan Horschel (2b.2)

11:15 a.m. Large Steel Experiment Tom Blejwas (2c)

11:30 a.m. Lunch - Coronado Club

12:45 p.m. Badge Office i

1:00 p.m. Concrete Experiments

Analysis Joe Jung (2d)

Model Schedules Tom Blejwas (2e)

Summary of Peer ReviewWritten Comments Tom Blejwas

Questions for OvernightConsideration Jim Costello

2:15 p.m. Tour of Model Prep. Lab,Bldg 867 Ron Woodfin

3:00 p.m. Tour of Containment TestSite Ron Woodfin

4:45 p.m. End of Day's Activities

.

'.,

Friday, February 25 Bldg 822, Room A,

8:30 a.m. Related Programs

BCL Studies Richard Denning (2f)

EPRI Concrete Tests Ian Wall (2g & 3)

Electric PenetrationAssemblies Wayne Sebrell (2h)

Penetration Program andIsolation Valves Jim Costello

9:30 a.m. Far-Term Plans

10:00 a.m. Break

10:15 a.m. Discussions of Concrete Models

11:15 a.m. Lunch - Coronado Club

12:45 p.m. Discussion of Concrete Models.

2:15 p.m. Break

2:30 p.m. Consensus of Concrete Models

3:00 p.m. End of Meeting

.

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Minutes of the Peer Review Group MeetingSandia National Laboratories.

Albuquerque, New MexicoFebruary 24-25, 1983

Attendees

Peer Review Group

Richard S. Denning, Battelle Columbus Labs

John D. Stevenson, Stevenson & Associates

Joseph J. Ucciferro, United Engineers & Constructors, Inc.

Ian Wall, Electric Power Research Institute

Richard N. White, Cornell University

James S. Wilbeck for Wilfred Baker,Southwest Research Institute

Nuclear Regulatory Commission

John Burns

James Costello

Sandia National Laboratories

Walter A. von Riesemann

Thomas E. Blejwas -

Daniel S. Horschel

Joseph Jung

Robert T. Reese, Secretary

Wayne A. Sebrell

Ronald L. Woodfin

David B. Clauss

E. H. Conley

.

. . . _ _ __ . _. . . _ . __

* -

: .,

|This summary is organized inte the following sections:

3' a. Program Direction,

b. Steel ModelsL c. Concrete Models

Each section contains discussions, agreements, and/oraction items. Materials used in presentations arecontained in the appendixes in the order given. Otherinformation is also contained in the appendixes.

t

j Program Direction

The Containment Integrity Program has two major objectives: to'

qualify analytical methods for reliably predicting thestructural capability of LWR containments subjected to extremeloading conditions caused by severe accidents and extreme4

environments and, in conjunction with other NRC sponsoredprograms, to provide a basis for prediction of containmentbehavior including leakage. The initial loading conditionconsidered is static internal pressure which will be applied tothree small-scale (1/32) steel model containments, a largesteel model (1/8 scale) and to model containments constructed .Ii

of reinforced concrete.

!A schedule of activities is given below (as of February 1983):

!

Steel ModelsPressure Test on Steel Model SC-1 Mar 1983Pressure Test on Steel Model SC-2 Jun 1983

: Pressure Test on Steel Model SC-3 Aug 1983Site Preparation for Large Steel Model Tests Apr-Jul 1983Fabrication of Large Steel Models Mar-Jul 1983

| Testing of Large Steel Model1 a. Erection / Leak Check Jul 1983: b. Instrumentation Jul-Oct 1983

c. Pressure Testing Nov 1983;

Concrete ModelsModeling Studies of Reinforced Containment Mar-May 1983Presentation of Modeling Studies to NRC Jun 1983Recommendations for Designs Jul 1983Design of Concrete Model Containments Aug-Oct 1983

Nov-Dec 1983Design ReviewSite Preparation, Construction of Model Dec 1983-May 1984Final Concrete Curing Jun 1984

.

Testing of Concrete Model Containmenta. Instrumentation Mar-Jul 1984

: b. Pressure Testing Aug-Sept 1984,

i

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_ . _ . . _ - . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _.___, . , _ _ _ _ , . _ _

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

-,

i

-Agreement / Action Items / Major Discussion Items

There was considerable discussion regarding the possiblei failure modes of the steel containments subjected to internal; pressure. Testing of the large steel model should provide; information to address the basic questions regarding failure -

either: catastrophic or sustained leakage.

The other discussion centered on the concrete models. Thesediscussions are summarized in the concrete model section.

The following agreements were made:

1. There should be interaction with EPRI and their sponsoredj work and testing program at the Construction Technology1 Laboratories (a division of Portland Cement Association)..

2. Conceptual features for concrete models were discussed andare outlined in that section.;

i

3. Leakage rates are to be measured.;

4. Tests on separate features of concrete containments arei necessary.

Jim Costello raised the following question which was discussed,

but no firm answers were formed. The questions was: Supposethat the British do not test a SNUPPS-type containment(prestressed concrete). What can be done to bridge the gap ofinformation between the behavior of reinforced concretecontainments and prestressed containments? This question wasdiscussed but no consensus was reached.'

(Steel Models

,

The tests on the amall steel models (SC-0 through SC-3) were

{ discussed. Tom Blejwas summarized changes on the large steelmodels including gussets or ring stiffeners, a constrainedpiping penetration, painting, and an internal support structure.j.

A suggestion was made regarding the constrained pipingpenetration in which an internal tie rod was proposed insteadof an external restraint. This suggestion will be incorporated

| into the model.I

Concrete Modelst

As indicated previously, one of the purposes of the peer reviewmeeting was to discuss the concrete models. Jim Costello posedthe following questions.

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.,

1. What must be included in the concrete models?.

2. What can or should be done in separate tests?

3. What would be the design and construction costs for 1and 2 above?

4. What additional tests would be required to " cover"prestressed containments?

With respect to 1, there was a lengthy discussion on the needand requirement for a liner in the concrete model. JoeUcciferro suggested that the liners and liner-concreteinteraction could be tested with separate component tests. If

a liner was not used then (1) the difficulties involved infabricating the liner and preserving it through theconstruction of the containment would be eliminated; (2) thelack of understanding on the concrete-liner interaction and theapplicability of scale model tests to verify this interactionwould be eliminated; (3) and the types of liner anchoragesystem that should be used to represent the existingcontainments would not require investigation. The otherapproach is to have a liner because the presence of a linerwill be a better simulation of a real containment and the linerwill, in all likelihood, affect leakage through the containmentboundary as continued applications of internal pressure forceyielding and failure of the containment. It was agreed that aliner is needed and that the modeling studies need to include aliner in their work.

There was also considerable discussion regarding thepenetrations in the concrete model. There was no disagreementthat penetrations were needed - only that the number and typeneed to be resolved in the modeling study.

,

There was also discussion in which agreement was made thatseparate tests are likely to be required on the following: (1)penetrations, (2) cylinder basemat interaction, and (3)confirmation of scaling laws in terms of replication of theflaws occurring in the construction processes.

The discussion on design and construction costs for concretemodels resulted in various estimates. The liner alone isexpected to cost between $125K and $200K. The cost for themodel (without liner) depends on the scale and ranges from

6 for 1/2 scale.$250K to as large as $1.3 x 10

The similarities and differences of the prestressed andreinforced concrete containments subjected to internal

.

J

pressures was discussed in some detail with emphasis on theultimate capacity.

-3-

.. - . . - . . __ . -. . _ . - _ _- - .- . _ . - - . .-

#s

The major question involving the additional tests required forprestressed concrete resulted in the following concerns andregions of the containment that are in,

(1)likely requirements: (2) discontinuities,transition from prestressed to reinforced,(4) evaluationcrack formation and propagation, andReference was made to work by Adolf(3)

inconcrete differences. ItWalsar '(Sargent & Lundy) on cracks in prestressed concrete.,

was suggested that a meeting be held in June in Chicago toreview the EPRI sponsored tests and the relationships involvingcracking prestressed and reinforced concrete section of modeledcontainments.

Separate Presentation

R. S. Denning - BCL

This presentation focused on the fission product release to theatmosphere and is based on a study funded by NRC to be reportedThe status of the study was reported along with

in June.possible implications for the Containment Integrity Program.Included were detailed discussions on the overestimates ofIn this reactor safety study,releases predicted in WASH 1400. (1)there were three sources of release of fission products:transport in the primary system, and (3)(2)from the fuel, The WASH 1400release through the containment boundary.methodology and the study presented agreed on the source termsWASH 1400 did not account for retention in theprimary system and likely overestimated the release from theThese two factors will lead to a likely reductionfrom the fuel.containment.in the overall source term.One factor governing the potential release of materials is the(particularly steamrelationships among pressure increases and the times atfailure modes of the containment,which the failures could occur in terms of the introduction ofspikes),

,

There arepossible source materials into the containment.in the amounts of potential sourcesignificant differencesmaterials retained in the primary system and in the containmentas a function of the timing of a failure mode.

The vugraphs used are found in Appendix 2F and describe therelationship for release fractions for cesium and other radio-nuclides in the primary system and in the development andcomparisons with the results from WASH 1400.

The implications for the containment integrity program are thatthe failure modes of containments and their timing areimportant in evaluating the release fractions which are thefactors in the assessment of reactor safety subjected toextreme loading conditions.

-4-

_ - . _ - _ -_ _ . . _ _ _ . - .

*,.

Ian Wall - EPRI

This presentation described the EPRI sponsored work onevaluating retention in the primary coolant sytem; measuringsteam spikes; hydrogen deflagration and tests at the PortlandCement Association's Construction Technology Laboratories(CTL) . " The tests at CTL are to substantiate the source termswith a target of reducing them by a factor of 10 by 1984 fromthe 0.6 range outlined by R. S. Denning. Besides the expectedretention of source materials in the primary cooling system andcontainments, Wall stated that the likely failure mode of acontainment will be to leak rather than fail catastrophically.

Tests were described in which a simulated wall of a containment(a rectangular reinforced concrete bloch with a steel liner)will be placed over a pressurized chamber which includesaerosol size particles. The block will be loaded in biaxialtension through the reinforcing bars or prestressing tendons.Measurements of leakage and aerosols will be made on theopposite side of the block to substantiate the source term.

Comments were also requested on the paper by Joe Rashid whichis also enclosed with the following request from Ian Wall:

" Enclosed please find a copy of a draf t EPRI report entitled' Review of the WASH-1400 Surry Containment CapabilityAssessment' by Dr. Y. R. (Joe) Rashid, President AnatechCorporation. EPRI requested Dr. Rashid to critically reviewthe structural analysis for the internal pressurizationcapacity of the Surry containment as presented in appendix VIIIof the WASH-1400. They encouraged him to evaluate inparticular the realism of the failure mode assumed in theWASH-1400 study. In this brief report, Joe argues that theSurry containment would most likely leak before it can rupturecatastrophically. EPRI is' sponsoring an experimental projectwith Portland Cement Association in order to substantiate manykey points in this argument. If Joe's argument can besustained, it would reduce the source term for LWRs withconcrete containments by orders of magnitude.

An issue of this significance warrants careful review, and Dr.Ian Wall solicits your comments which should be sent to him at

I the Electric Power Research Institute, 3412 Hillview, P.O. Box

! 10412, Palo Alto, CA 94303." (Please include a copy of these| comments to W. A. von Riesemann, Division 9442, Sandia National

i Laboratories, P.O. Box 5800, Albuquerque, NM 87185.) Specific"

questions that he would like to address include:'

!

! 1. Is there merit to Dr. Rashid's argument?

!

! 2. What are the critical assumptions?

I

!

!

-5-!

,

=,m-- . , . . . . y ,~~-,,-.m - - - , .,c. ,.-,-,.,--,,,-,-,,e,-.--,--~~__-,,-w- - - . _ ..y,.-..- - --,-e,w...,--ne.-_ _=--=-v~w-=-----

- . ... .n.. . . .. ... . .

,

-.

,

3. What experiments would your recommend to validate the,

calculations?

4. What additional calculations should be done?

5. How generic is the conclusion to other reactor containmenttypes?"

spectfully submitted,

$ s sL,

Robert T. Reese, Secretary

.

-6-

. .- - . - - . - . - _ , _____ __.- __, .___ __ - -- _ _ - . _ . .-

3A'

Sandh Nbdonal Labomtodes--

^'b" "''4"*' "' * " **'' 87'85date: June 1, 1983,

to: Steve Hatch, 9411

from: Joseph Jung, 9442

l. sy!

subject: Analyses of containments

At your request we have performed quick reviews or analysesto estimate the ultimate internal static pressurecapabilities of several nuclear power plant containments. Asummary of the results follows:

plant Unit Desion Estimated Staticpressure pressure Capability

psia psia

Grand Gulf 1 15 EbSurry 1 45 85Sequoyah 1 10.8 60Calvert Cliffs 1 50 124Oconee 1 59 151peach Bottom 2 56 123

The estimated failure pressures for the first three plantswere obtained from a literature search of reports ( see theattached memo from C. Conley,1523). The Grand Gulf andSequoyah failure pressure estimates were derived from finiteelement analyses while the Surry estimates were based onhand calculations. The Grand Gulf analysis is probably asgood an estimate of the failure pressure of the containmentshell as can be obtained with the current state of the artmethods. None of these analyses addressed possible majorpenetration failures.

For the Calvert Cliffs, Oconee and peach Bottom

analyses, containment information was obtained from therespective FSAR's. Since the average and minimum propertiesof the materials used in the containments were notavailable, the minimum allowable yield stresses of thematerials (as specified by ASTM) were used in handcalculations to estimate the internal pressures which wouldcause general yielding of the containments.

Although it can be argued that some containments can havesignificant pressure capacity beyond that which causesgeneral yielding, estimates beyond general yielding require

_ . - _ , _ _ _ _ - _ . ._ . . _- __ ___ .__ _ - _ _ . _ _ _ ___- _ _ _ _ . . . _

__ _ . _ _ _ _ _ _ _ _ _ _ _ . _ _

.

.

( -,

advanced analysis techniques. In addition, as the pressureincreases the probability of penetration failures increases*

so that analyses of penetrations would also be prudent toinsure that a major penetration does not fail before thecontainment shell fails. These considerations lead to fairlysophisticated analyses which are not appropriate at thistime due to time constraints.

The Calvert Cliffs and Oconee containments have many'

similarities such as both being prestressed concretecontainments with the same number of butresses and tendoni

types. In addition, both containments have 1/4" liners madeof A36 steel.The design pressure for Calvert Cliffs is 50'

psig while that for Oconee is 59 psig. The numbers givenin the first table of this memo refer to the internalpressure which would cause the liner, hoop reinforcing bars,and hoop tendons to yield in the cylindrical shell portion'

| of the containment. My calculations are in attachment 2.

The analysis of the peach Bottom containment (62 psig designpressure) consisted of estimating the the yield pressuresfor various parts of the containment. The ellipsoidal head," neck", sy 'rical section, and suppression chamber wereconsidered. The " neck" and suppression chamber have thelowest yield pressures of 123 psig. possible dynamic fluid-structure interaction in the suppression chamber was notconsidered. The calculations for peach Bottom are also in . ,

,

'

attachment 2.

To perform analysis which are of higher reliability thanthose performed here finite element analyses, as-builtmaterial data and detailed drawings of the structure and

;

major penetrations are needed.'

Rules-of-thumb using dehign pressures and containment typesj to estimate containment ultimate capacity can be misleading.

( A good example of this is the difference between theSequoyah and Hatts Bar ultimate pressure capabilities.Although both of these containments are steel and have 15 '

psig design pressures, the Watts Bar containment has apressure capacity between 120 and 140 psig, roughly twicethat of Sequoyah. This implies that rule-of-thumb estimatesof containment capacities should be applied with great care.

;

| copy to:.

9442 H. A. von Riesemann! 9442 T. E. Blejwas

9442 D. S. Horschel! 1523 C. H. Conley

t

;

||4

'

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

.

..

.

Attachment 1

Ultimate Pressure Capability Estimates forGrand Gulf, Surry, and Sequoyah

.

5

.

Sandia National Laboratories''

,

dateJ u ne 1, 1983 Albuquerque. Ne w Me xico 8 718 5

toJ . J ung , 9442

',

fromC. H. Conl ey ,1523

v

subjet: Current 'Best Estimates' for Containment Pressure Capacity

Listed below are the best estimates that are currentlyavailable for the internal pressure capacity of the GrandGulf, Surry, and Sequoyah containments. These estimateswere taken directly from the quoted references. All of the

analyses treat the increased internal pressure as a staticor quasi-static phenomenon. Following the estimates arecomments that may be helpful in judging their quality.

CONTAINMENT ESTIMATED MAXIMUM INTERNAL PRESSURE

1) Grand Gulf 54.5 psig85 psig2) Surry

3) Sequoyah 60 psig

COMMENTS

1) Grand Gulf: This an'alysis [1] was performed byBrookhaven National Laboratory in 1982. It is quite

rigorous in comparison to other comparable analysesof reinforced concrete containments for overpressure.They predict a liner failure, due to excessivestraining, in the region above the vessel midheightand below the springline.

Copy to:1500 W. Herrmann 1523 Route and File1510 D. B. Hayes 1524 W. N. Sullivan1520 T. B. Lane 1530 L. W. Davison1521 R. D. Krieg 1540 W. C. Luth1522 T. G. Priddy 9442 W. A. Von Riesemann1523 R. C. Reuter, J r. 9411 S. W. Hatch

1523 C. H. Conley

w- . . _

- . - - - _ _ _ . - , . _ __

-

-.

-2- June 1, 1983,jyng, 9442,

.

The finite element program developed by BNL uses one ofthe concrete material models currently in vogue, andalso appears to model the reinforcing and the reinforc-ing steel-concrete combination quite accurately. Theonly real criticism is that they model the reducedshear stiffness of the cracked reinforced concrete byapplying a constant reduction factor to the elasticshear modulus. Recent research has shown this to beunconservative. As with all of the computer codesthat incorporate sophisticated concrete models, onlylimited verification has been performed. Al so , the

stochastic nature of the problem was not addressed.

It is doubtful that a significantly better estimatecould be obtained given the current state of the artin reinforced concrete analysis.

2) Surry: The analyses for this containment were performedin 1973 by Batte11e's Columbus Laboratories and Paul Mast,Consulting Engineer. They are reported in W ASH 1400 [2].Both analyses consist of simple hand calculationsbased upon the authors' opinion of how the containmentwould behave when subjected to internal overpressure.The reported most probable f ailure in both studies istearing or ' blow out' of the liner due to excessivelywide cracks in the reinforced concrete. The failurepressures predicted by the two sources differed byonly 5 psi. Bounds for the failure pressure werecalculated by Battelle as the pressure at which theliner and the inner layer of reinforcing first yield,and the pressure at which they reach their ultimatetensile strength.

,

Later research on bond behavior and tension stiffeningeffects in reinforced concrete indicate that the authors'i

|predictions on cracking in the vessel are unrealistic.They predict that the cracks observed in the structural!

acceptance pressure test will widen excessively withoutsignificant new cracks developing. More likely, theextent of cracking will increase far beyond that observedin the pressure test. In that case, the predictedreinforcing and liner strains will be distributed overa larger number of cracks with subsequently smallercrack widths. If the strain in the reinforcing (andliner) between cracks were accounted for, the resultingpredicted crack widths would be even smaller.

m -T_ ' s a 7 .. m m h ,

:i

-. ,

I

J. Jung, 9442 -3- June 1, 1983,

The bounds developed for the f ailure load discount anystrength provided by the outer reinforcing due tospalling of the concrete on the outside face. Thereis currently no way to determine how severe the spallingwould be but assuming complete loss of bond seemsextreme. Also, unless the bars actually rupture, theymust provide some restraint to the expanding cylinderthey enclose.

Considering the above, the reported failure load isprobably conservative, but considering the time framefor this study, a better estimate is not obtainable.The inherent uncertainty in the material properties,geometry, etc., were not addressed.

3) Sequoyah: The report reviewed was prepared by AmesLaboratory in 1979-1980. It contains their estimateof the failure pressure along with the criticisms andestimates of five others. The reported failurepressures, when normalized to account for actual materialproperties, range from 40 to 60 psig. The failure ineach case was the result of excessive straining in thesteel, which is to be expected for a steel containment.The 60 psig value was calculated at Ames and is basedon the assumption that the steel of the containmentcan withstand strains 2 times larger than the elasticstrain without fracturing.

The Ames analysis addresses the stochastic nature ofthe problem and its accuracy might better be judged bysomeone with more expertise in that area. State ofthe art analysis techniques for steel are betterestablished than those for reinforced concrete and itis doubtful that this analysis could be greatly improvedupon. A preliminary study of the dynamic effects ofthe loading was included in the report.

Most of the reported failure pressures were within 15percent of each other. Also, the uncertainty in theAmes assumption on ductility is reportedly small . Itis therefore felt that 60 psig is a current best estimate.

4) General: As noted above, none of the reported failurepressures are based upon analyses addressing the possibledynamic nature of the loading. Further, a fractureanalysis of the steel liner in the reinforced concretecontainments was not performed, and only limited attentionhas been given to the penetrations. In all of the analyses,only a limited number of possible failure modes wereconsidered.

- .-.

,

.

-.

,

J. Jung, 9442 -4- June 1, 1983.

REFERENCES

[1] Sharma, S., Reich, M., Chang, T. Y., and Shteyngart, S.,

" Failure Evaluation of a Reinforced Concrete Mark IIIContainment Under Uniform Pressure," Department ofNuclear Energy, Brookhaven National Laboratory,BNL-NUREG-51543 and NUREG/CR-1967, Upton, NY, September1982.

[2] WASH 1400

[3] Greimann, L. , Fanou s , F. , Sabri , A. , Ketal aar, D.,

Wolde-Tinsae, A., and Bluhm, D., " Reliability Analysisof Containment Strength; Sequoyah and McGuire IceCondenser Containments," Ames Laboratory, FIN No.A4131, Ames, 10, November 1980.

.

6*

.

-.

Attachment 2*

Calculations for the Ultimate Strengths ofCalvert Cliffs, Oconee, and Peach Bottom

.

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