AECM-83-0669, Forwards Responses to Informal NRC Concerns on Containment Structural Integrity Test Per 830706 Conference W/Bechtel & Nrc.Test Performed in Compliance W/Provisions of Reg Guide 1.18,Rev 1

Forwards Responses to Informal NRC Concerns on Containment Structural Integrity Test Per 830706 Conference W/Bechtel & Nrc.Test Performed in Compliance W/Provisions of Reg Guide 1.18,Rev 1

ML20080S220
Person / Time
Site:	Grand Gulf
Issue date:	10/14/1983
From:	Dale L MISSISSIPPI POWER & LIGHT CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
AECM-83-0669, AECM-83-669, TAC-51866, NUDOCS 8310180306
Download: ML20080S220 (30)

Text

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MISSISSIPPI POWER & LIGHT COMPANY Helping Build Mississippi P. O. B O X 16 4 0, J A C K S O N, MIS SIS SIP PI 3 9 2 0 5 October 14, 1983 NUCLEAR PRODUCTION DEPARTMENT U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, D. C. 20555 Attention: Mr. Harold R. Denton, Director

SUBJECT:

Grand Gulf Nuclear Station Units 1 and 2 Docket Nos. 50-416 and -5C ':17-License No. NPF-13 File: 0260/C-197.0/L-334.0 Transmittal of Responses to Infortral NRC-SEB Concerns on the Containment SIT AECM-83/0669 On July 6, 1983, a conference call was held between representatives of MP&L, Bechtel, and the NRC Structural Engineering Branch to discuss NRC concerns pertaining to the Grand Gulf Nuclear Station - Unit 1 (GGNS)

Containment Structural Integrity Test (SIT). The NRC expressed concerns about (1) the basis for the GGNS containment being tested as a nonprototype containment; and 2) the SIT descriptions in FSAR subsection 3.8.1.7 and the "GGNS Unit 1 Primary Containment Structural Integrity Test Report," dated January 1982. MP&L responses to these concerns are provided as attachments.

Additionally, MP&L maintains that the GGNS-1 SIT was performed in cotsnliance with the provisions of Regulatory Guide 1.18, Revision 1, and Article CC-6000 of the ASME Boiler and Pressure Vessel Code, Section Ill, Division 2, 1980 Edition, except as noted in FSAR Appendix 3A.

Should you require additional information, please contact us.

Yours truly,

. Dale Manager of Nuclear Services MLC/JGC:rg Attachments A00\

cc: See next page [V (E'G10180306 031014 PDR ADOCK 05000416 P PDR Member Middle South Utilities System

AECM-83/0669 MISSISSIPPI POWER Q LIGHT COMPANY cc: Mr. J. B. Richard (w/a) ",

Mr. R. B. McGehee (w/o)

Mr. T. B. Conner (w/o)

Mr. G. B. Taylor (w/o)

L Mr. Richard C. DeYoung, Director (w/a)

Office of Inspection & Enforcement U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Mr. J. P. O'Reilly, Regional Administrator (w/a)

U.S. Nuclear Regulatory Commission Region II 101 Marietta St., N.W., Suite 2900 Atlanta, Georgia 30303

, Attachment 1 to AECM-83/0669

. Page 1 of 5 Containment Design Comparison NRC Concern:

The FSAR states in subsection 3.8.1.7.1 that the containment SIT test procedure will be based on the containment being a nonprototype structure.

Provide a comparison of the Grand Gulf containment design to similar containments.

Response

The Grand Gulf Nuclear Station (GGNS) Structural Integrity Test (SIT) is based on the containment being a nonprototype structure. The SIT procedure complies with the requirements of NRC Regulatory Guide 1.18, except as noted in FSAR Appendix 3A.

The Grand Gulf containment is a conventionally reinforced concrete structure.

It does not incorporate any new or unusual design features as defined in Regulatory Guide 1.18, Appendix A, which would require it to be considered a prototype containment. A summary of Appendix A to Regulatory Guide 1.18 is given below.

"A concrete primary reactor containment is considered a prototype if it is the first design to incorporate any of the following features:

a. A dome with a shape other than hemispherical;

b. An opening larger than 0.2D3;

c. Two openings with a diameter greater than 0.15D that are separated by a distance less than 0.2D;

d. A connection of the cylindrical wall to the bottom slab or to the dome by a sliding joint, a hinge, or a combination of hinge and sliding joint;

e. A pattern of main reinforcing other than vertical straight bars and horizontal hoops;

f. An intermediate interior floor connected to the wall; or 9 Any other structural design feature that may decrease the safety margins from those of a containment confirmed by an acceptance test.

3D = internal diameter of the cylindrical part of the containment."

The GGNS containment consists of a flat circular foundation mat, right circular cylinder, and hemispherical dome with major dimensions as shown in GGNS FSAR Figure 3.8-1 (attached). The inside diameter (D) of the circular cylinder is 124'-0". The largest opening, and the only opening greater than 0.150, is the 19 foot (0.153D) diameter equipment hatch. Sliding or hinged joints are not used to connect the cylinder wall to the bottom slab or the dome. No intermediate interior floors are fixed to the wall. The main reinforcing in the wall consists of inside and outside layers of hoop reinforcement, inside and outside vertical reinforcement, and diagonal reinforcement placed in two directions to form a helix with an angle of approximately 45 degrees from the vertical axis of the shell.

Attechment I to AECM-83/0669 Page 2 of 5 This design is similar to many other conventionally reinforced concrete containments. Table 1 (attached) lists several examples of plants using this type of containment.

The variations noted in containment dimensions are due to differences in design loads and/or design preferences.  ;

l The Salem Nuclear Generating Station, Unit 1, is a prototype conventionally reinforced concrete containment similar to GGNS. Its containment consists of I a flat concrete base mat at the wall to mat interface, right circular cylinder, and hemispherical dome with major dimensions as shown in Salem FSAR Figure 140 feet.

3.8-1 (attached). The inside diameter (D) of the circular cylinder is hatch.

The largest opening is the 19 foot diameter (0.1360) equipment No openings exceed 0.15D. Sliding or hinged joints are not used to connect the cylinder wall to the botton slab or the dome. No intermediate interior floors are fixed to the wall. The main reinfcrcing in the wall consists of inside and outside layers of hoop reinforcement, inside and outside vertical reinforcement, and diagonal reinforcement placed in two i l

directions to form a helix with at angle of approximately 45 degrees from the vertical axis of the shell. The Salem SIT is in accordance with the prototype test requirements of Regulatory Guide 1.18, as detailed in the Salem FSAR, subsection 6.2.1.2. Strain data was obtained based on a SIT air pressure of '

54 psig (115 percent of design pressure) as compared to the GGNS SIT air pressure of 17.25 psig.

As described above, the Salem design is similar to Grand Gulf. Based on this similarity in design, the successful prototype testing of Salem and the i

successful nonprototype testing of Grand Gulf, it can be concluded that the GGNS Unit I containment meets the requirements of Regulatory Guide 1.18, Revision 1, for SIT testing.

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Attachment 1 to AECM-83/0669 Page 3 of 5

TABLE 1 REINFORCED CONCRETE CONTAINMENTS Design Parameters s Internal Inside Design Diameter Pressure Name of Nuclear Unit (ft) (psi)

Grand Gulf 1&2 124 15 Connecticut Yankee 135 40 Indian Point 2&3 135 47 Salem 1&2 -

140 47 Diablo Canyon 1&2 140 47 Surry 1&2 126 45 Maine Yankee 135 55 Donald C. Cook 182 115 12 Beaver Valley 1&2 126 45 North Anna 1&2 126 45 Harris 1&2 130 45 Millstone 3 140 45 Comanche Peak 1&2 135 50 Seabrook 182 140 51 WPPSS 1 150 46.4

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Attachm:nt 2 to AECM-83/0669 Page 1 of 6 Extensometer Location Clarification NRC Concern:

Table 4-1, "Extensometer Locations," of the Grand Gulf Nuclear Station (6GNS)

Unit 1 Primary Containment Structural Integrity Test (SIT) Report, issued in January 1982, shows extensometers H-13, H-14, and H-15 at azimuths 210 , 270 ,

and 150 , respectively, at El. 230'-0". The table also shows that six extensometers were located at both El.103'-0" and El.167'-0". However, Section C.2 of Regulatory Guide 1.18, Revision 1, " Structural Acceptance Test for Concrete Primary Reactor Containments," requires that radial deflections be measured at three points along six meridians. The containment SIT performed does not meet this criteria since El. 230'-0" only measured radial deflections at three meridians.

Response

The test method at El. 230'-0" used to measure radial deflections consisted of placing extensometers at azimuths 210 , 270 , and 150 and connecting them to their " fixed" reference points at azimuths 30 , 90', and 330 , respectively.

The azimuth diametrically opposed to the extensometer location was used as the reference point because there are no internal structures suitable for use as reference points at this elevation. The radial displacement at each azimuth was assumed to be equal to one-half of the measured change in diameter.

Therefore, the displacements at azimuths 30 and 210 , 90 and 270 , and 150 and 330 are one-half of the measured differential displacements between each respective pair of azimuths. Tables 4-1, 5-13, 5-14, and 5-15 and Figure 5-1 of the GGNS Unit I containment SIT report have been revised to clarify the measurement locations and test results. The revised pages are attached.

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Attachm:nt 2 to AECM-83/0669 Page 2 of 5 MISSISSIPPI POWER AND LIGHT COMPANY GRAND GULF NUCLEAR STATION UNIT 1 PRIMARY CDNTAINNENT STRUCTURAL INTEGRITY TEST REPORT i

Be htel Power Corporation San Francisco, California July 1983 (Rev. 1) i

Attachment 2 to AECM-83/0669 Page 3 of 6 TABLE 4-1 EXTENSOMETER LOCATIOM Elevation Radials (H) and Hatch (E)

Table or Radius Extensometer Verticals (V) and Dome (D) Azimuth A-2 H-1 A-3 EL. 103'-0" 30' H-2 EL. 103'-0" A-4 H-3 90*

A-5 H-4 EL. 103'-0" 150' A-6 EL. 103'-0" 210' H-5 2L. 103'-0" A-7 H-6 270*

A-8 EL. 103'-0" 330" H-7 EL. 167'-0" A-9 H-8 30*

A-10 H-9 EL. 167'-0" 90*

A-ll EL. 167'-0" 150' H-10 EL. 167'-0" A-12 H-11 211*-08' A-13 EL. 167'-0" 270*

H-12 EL. 167'-0" A-14 H-13 330*

A-15 EL. 230'-0" 210* (30')*

H-14 EL. 230'-0" A-16 H-15 270* (90*)*

A-17 EL. 230'-0" 150' (330*)*

H-16 EL. 131'-0" A-18 H-17 30' A-19 EL. 131'-0" 150' H-18 EL. 131'-0" A-20 H-19 270*

A-21 EL. 202'-0" 30*

H-20 EL. 202'-0" A-22 H-21 150*

A-23 EL. 202'-0" 270*

D-1 R 3'-1" A-24 D-2 165'-30' A-25 R 31'-0" 75*

D-3 R 45'-0" A-26 V-1 75*

A-27 R 61'-8" 46' V-3 R 61'-8" A-28 V-4 225*

A-29 R 61 '- 8" 315*

E-1 EL. 198'-6" A-30 E-2 240' A-31 E-3 EL. 191'-2-1/2" 240' A-32 EL. 182'-9" 240' E-4 EL. 161'-9" A-33 E-5 240' A-34 E-6 EL. 153'-10-1/2" 240" A-35 EL. 146'-6" 240' E-7 EL. 172'-6" A-36 E-8 217*-47' A-37 EL. 172'-3" 223*-14' E-9 EL. 172'-6" l

A-38 E-10 230*-14' A-39 EL. 168'-2" 247*-07' E-ll A-40 E-12 EL. 172'-7-1/4" 259' EL. 172'-0" 264*

• Extensometers H-13, H-14, and H-15 measure dif ferential displacements between six attachment points on the containnent wall (i.e. , between Azimuth 210* and 30*, 270* and 90*, and 150* and 330' respectively).

DH-105 4-8 (Rev 1)

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.' Page 4 of 6 DEFLECTIONS (INCHES)

WALL EL MIN MAX AVG PRED1 PRED 2

"'"* 103'-6" .024 .044 .034 .08 01

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DISPLACEMENT

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!* e MEASURED VALUE i I l I i l NOTES:

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EL 230' ARE EQUAL TO HALF OF MEASURED l

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, FIGURE 5-1 CONTAINMENT STRUCTURE AVERAGE DEFLECTIONS AT 17.25 PSIG - WALL AND DOME 54 (REV.1)

Attachmznt 2 to AECM-83/0669 Page 5 of 6 TABLE 5-13

SUMMARY

OF DATA FOR TRANSDUCER H-13 (Psig) Displacement (diametric)

Date Time Pressure 1 (inches) 1/1 0000 0 0.000

' 1/1 0205 2.5 0.000 1/1 0343 5.0 0.000 1/1 0622 7.5 0.011 1/1 0930 10.0 0.026 1/1 1041 12.6 0.043 1/1 1246 15.0 0.125 1/1 1349 17.3 0.185 1/1 1630 17.3 0.210 1/1 2000 15.1 0.210 1/1 2249 12.5 0.198 1/2 0153 9.8 0.173 1/2 1458 5.1 0.130 1/2 1802 2.5 0.110 1/2 2114 0 0.088 TABLE 5-14

SUMMARY

OF DATA FOR TRANSDUCER H-14 (Psig) Displacement (diametric)

Date Time Pressure 1

(inches) 1/1 0000 0 0.000 1/1 0205 2.5 0.006 1/1 0343 5.0 0.013 1/1 0622 7.5 0.023 1/1 0930 10.0 0.039 1/1 1041 12.6 0.057 1/1 1246 15.0 0.154 1/1 1349 17.3

• 1/1 1630 17.3 0.249 1/1 2000 15.1 0.244 1/1 2249 12.5

• 1/2 0153 9.8 0.197 1/2 1458 5.1 0.147 1/2 1802 2.5 0.124 1/2 2114 0 0.102

• Invalid Reading - Out of Scale DR-105 5-22 (Rev 1)

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

Attachment 2 to AECM-83/0669 Page 6'of 6 j

TABLE 5-15

SUMMARY

OF DATA FOR TRANSDUCER H-15 (Psig) Displacement (diametric)

Date Time Pressure 1 (inches) 1/1 0000 0 1/1 0.000

, 0205 2.5 0.005 1/1 0343 5.0 0.009 1/1 0622 7.5 0.017 1/1 0930 10.0 0.029 1/1 1041 12.6 0.040 1/1 1246 15.0 0.108 1/1 1349 17.3 0.165 1/1 1630 17.3 0.184 1/1 2000 15.1 0.184 1/1 2249 12.5 1/2 0.171 0153 9.8 0.150 1/2 1458 5.1 0.112 1/2 1802 2.5 0.097 1/2 2114 0 0.080 IABLE 5-16

SUMMARY

OF DATA FOR TRANSDUCER H-16 (Psig) Displacement d

Date , Time Pressure (inches) 1/1 0000 0 0.000 1/1 0205 2.5 0.004 1/1 0343 5.0 0.009 1/1 0622 7.5 0.014 1/1 0930 10.0 0.019 1/1 1041 12.6 0.024 1/1 1246 15.0 0.044 1/1 1349 17.3 0.051 1/1 1630 17.3 0.056 1/1 2000 15.1 0.055 1/1 2249 12.5 0.051 1/2 0153 9.8 0.046 1/2 1458 5.1 0.037 1/2 1802 2.5 0.032 1/2 2114 0 0.026 t

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. Attachment 3 to AECM-83/0669

1. Page'1 of-17

,t Missing FSAR Figure ,

FSAR Figure 3.18-121 " Containment Structural Integrity Test Instrumentation

.and Crr.ck Mapping Locations," is referenced'in subsection 3.8.1.7. However, the figure wa's-not provided in Amendment 51 to the FSAR.

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Response

Figure 3.~8-121 was inadvertently omitted from Amendment 51 due to an administrative oversight. However, Figure 3.8-121 is being deleted from the FSAR. New-Figures 3.8-123, 3.8-124, 3.8-125, and 3.8-126 will be added to 3 - : replace Figure 3.8-121. ' Also, new Figure 3.8-122 was added to present the time versus pressure plot during the SlT. Additional changes to subsection

,, 3.8.1.7 to reflect the final results of the' containment' SIT are also planned.

' A draft of these changes is attached. These changes are currently scheduled

[for incorporation in the.FSAR in Amendment 57.

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Attachment 3 to AECM-83/0669 Page 2 of 17 FSAR

5. Compressive Strength (ASTM C 31-69, tested in accordance with ASTM C 39-72): Standard 6- by 12-inch concrete test cylinders will be molded and tested at 7, and 28 or 91 days from each 25

' 100 cubic yards, or a minimum of one set per

, day for each class of concrete.

Correlation tests for air content, slump, and temperature of concrete shall be made at the end 25 of the last piece of conveying equipment for the first batch produced each day and for each 200 u

c'bic yards or fraction over 100 cubic yards placed of each class of concrete.

b. Concrete Cylinders A strength test is the average of the strengths of the two specinens from each sample tested at 28 or 91 days. Only one cylinder shall be made and tested at 7 days. The provisions of ACI 31 -71, Paragraph 4.3.3, shall apply in determining the satisfactory strength le e se at 28 or 91 days. 9ng g ,

. j,j 3.8.1.6.6 Splices of Reinforcement 7

a. Provisions of ACI 318-71, Sections 7.5, 7.6, and 7.7, will be met for all reinforcing bar splices.

Splices are designed to develop the specified minimum yield strength.

b. All mechanical splices for reinforcing bars are made by.Cadweld process, using clamping devices, sleeves, charges, etc. , as specified by the Cadweld Splice Instruction Sheets for B and T series connections. C series materials are not permitted.

Qualification of operators, visual inspection, tensile testing, tensile test frequency, and procedure for substandard tensile test results are in accordance with the Cadweld instruction sheets, and will comply with NRC Regulatory Guide 1.10, Revision 1, dated 1/2/73.

3.8.1.7 Testing and Inservice Surveillance Requirements 3.8.1.7.1 Structural Integrity Pressure Test (Unit 1)

Following construction, the containment was proor-tested at 57 115 percent of the design pressure. During this test, deflec-tion and concrete crack measurements were made to determine Amend. 57 3.8-34

Attachment 3 to AECM-83/0669

' 8' 'U GG FSAR that the actual structural response was within the limits predicted by the design analyses.

A test procedure was issued prior to the structural Integrity Test and was based on the containment being a non-prototype structure. This procedure was in compliance with the require-ments of NRC Regulatory Guide 1.18, to the extent noted in Appendix 3A.

The containment was pressurized pneumatically in four pres-sure increments from atmospheric pressure to the maximum test pressure of 17.25 psig, as shown in Figure 3.8-122.

Normal operational water volumes were present, and the drywell

__ structure was vented 4 Pressurization and depressurization 6 were halted at 5.0, 10.0, 15.0, 17.25, 15.0, 10.0, and 5.0 psig to allow acquisition of the required structural integrity test data, i.e. , deflection and crack patterns. At these plateaus, the pressure was maintained for a minimum of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to determine if indicated displacements lagged the actual pressurization /depressurization cycle. In addition, deflec-tion data were recorded at each 2.5 psig change in pressure.

Gross structural deformations were measured using taut invar 57 wire extensometers that spanned between points on the contain-ment wall, dome, and springline and fixed points within the structure. The extensometers were located to measure radial displacements along typical wall sections and around the lower equipment hatch, vertical displacement of the dome relative to the operating floor, and vertical displacement of the spring-line relative to the foundation slab. The layout of the extensometer system is shown in Figures 3.8-123 through 3.8-125.

Concrete crack patterns were mapped in the areas shown in Figure 3.8-126.

The lengths and widths (measured by optical comparator) of all vis2ble cracks within these areas were recorded at speci-fied pressure levels.

A detailed description of the SIT test is given in the " Final Report On Primary Reactor Containment Structural Integrity Test Performed At The Grand Gulf Nuclear Station Unit 1 for Mississippi Power and Light Company," dated January 1982.

A summary-discussion of the results of the SIT test is given in the following paragraphs.

hroughthedrywellpersonnellock]

PRELIMINARY Amend. 57  !

3.8-34a l 57

Attachment 3 to 2cM-83/0669 GG Page 4 of 17 FSAR A. Radial and Vertical Deflections Tables 3.8-41 through 3.8-43 provide the radial and vertical deflections observed during the test at each displacement pred2cted transducer location, and the corresponding values.

tions were below theThe actual predicted vertical and radial fordeflec-locations. maximum values all The response of the containment to the maximum test pressure of 17.25 psig is illustrated in Figure 3.8-127, which shows the measured radial and vertical deflections of the of thedome.

cylinder wall and the measured vertical deflection Also shown are the predicted deflections based on both elastic and cracked section analyses. The measured less than thevertical deflections elastic analysisofprediction.

the dome and wall are This demon-strates that the membrane stresses in the dome and the vertical membrane stress in the cylinder wall were not sufficient to cause significant tensile cracking.

the modulus of elasticity used in elastic deflectionAlso, prediction calculations was based on the specified concrete compressive strength rather than the actual strength, which is higher, as demonstrated by concrete cylinder break tests.

Higher concrete strength, and, 57 consequently, higher modulus of elasticity, will result in deflections lower than the predicted as observed from the actual vertical deflection measureme,nts.

The radial deflection and cracked of the wall section casewall is between the elastic predictions. This demon-strates that hoop stress not develop the complete caused tensile cracking but did in the cracked section analysis.cracked section used as a model The wall radial deflections listed and plotted in Figure 3.8-127 show averages of the measurements made on several azimuths at each elevation. The individual measurements made at each elevation show a variation from azimuth to azimuth, which can be attributed to the containment shell not being constructed perfectly axisymmetric. On the dome, the measured deflections show a smooth trend. The measured wall vertical deflection listed in Figure

! 3.8-127 is the average of the measurements at extenso-meters V 2 , Vs, and V 4 (V2 ralfunctioned). The individual measurements are tightly grouped, as indicated in Table 3.8-41, demonstrating a uniform vertical elongation of j the cylindrical portion of the containment.

3.8-34b PRRMNARY Amend. 57 l 57

Attachment 3 to AECM-83/0669 gg rage 5 of 17 FSAR Figure 3.8-128 also shows the measured radial growth of the wall in the vicinity of the equipment hatch. The deflections measured along the horizontal center plane of the hatch are approximately symmetrical about the vertical center line.

of the hatch show that the outward movement in theMeasurements al vicinity of the hatch increased with elevation. This behavior is typical for a large opening located close to the containment base mat.

Measurements of 24-hour residual deflections showed an 85 percent recovery of the maximum recorded vertical deflection at the dome apex extensometer D ), an 88 per-cent recovery of the average (maximum vertical deflections 3

recorded at the springline (extensometers V , V3 , V 4 ),

and a 72 percent recovery of the average maximum radial 3

deflections recorded at the elevation that exhibited the maximum average peak radial deflection (extensometers Hg,i B20, H22)-

Recovery values are based on an average of the 24-hour residual deflections recorded at the appropriate extenso-meters, as tabulated in the " Primary Containment Structural Integrity Test Report" dated January 1982. In all cases, 57 the deflection recovery within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after complete depressurization, at the points of maximum recorded deflection for the dome vertica] direction and for the cylinder wall radial and vertical directions, was greater than 70 percent.

i B. Containment Concrete Cracking The surface concrete crack mapping areas on the contain-ment wall are shown in Figure 3.8-126. A variety of surface cracks observed with widths immediately priorless than 0.010 inch were to pressurization. I i

Concrete crack widths in the various surveillance areas did not change in measured width by more than 0.010 inch. This is consistent with the low deflections recorded and does not adversely affect the integrity of the containment structure.

i During the sWbsequent ILRT pressurization cycle to 12.5 psig, cracks that appeared in Crack Mapping Area 2 were l monitored. Those cracks enlarged slightly from the zero pressure recovery value and returned to hairline width not accurately measurable with the optical comparators.

Growth during the ILRT pressurization cycle was estimated at 0.002 inch at the locations observed.

3.8-34c PRELERRV (57 1

Attachment 3 to AECM-83/0669 Page 6 of 17 gg FSAR C. Summary and Conclusions The deflections met the acceptance criteria set in the Grand Gulf test procedure.

The containment structural integrity test provided proof of the structure's ability to contain the internal design pressure and provided measurement of structural response to changes in internal pressure. Test measurements for the containment included gross structural deformations and concrete crack growth.

The structure withstood the internal pressures with no observable indications of structural distress. All measured structural deformations were less than the maximum predicted values, and all recovery deflections 57 at the points of maximum average deflection recovered more than 70 percent of their maximum deflection. No cracks with widths equal to or exceeding 0.010 inch were observed in the crack mapping areas prior to pressuriza-tion. Changes in concrete cracks did not change in measured width by more than 0.010 inch during pressuri-zation, and those cracks that developed during pressuriza-tion closed to below measurable values at zero pressure.

The results of the structural integrity test provide direct experimental evidence that the containment struc-ture can contain the internal design pressure with sufficient margin of safety and that the gross response to pressure is predictable.

PREllMINARY Amend. 57 3.8-34d l57

GG l

FSAR 1 -

TABLE 3.8-41 i

VERTICAL DEFLECTIONS OF CONTAINMENT 2

CONTAINMENT SIT EXTENSO- AZIMUTH METER RADIAL LOCATION FROM (DEGREES) CONTAINMENT CENTERLINE DEFLECTION (INCHES)

Predictedi Predicted 2 Actu_a_l D: 165"-30' 3'-1" 0.1482 0.541 0.114 D2 75 31'-0" 0.145 0.507 0.112 '

l D3 75 45'-0" 0.139

~i 0.463 0.081 57

V (6 61'-8" 0.079

! 0.197 0.027 Va 225 61'-8" 0.079 0.197 0.031 V4 315 61'-8" 0.079 0.197 0.034

?%

i %C j 1 2

Based on uncracked analysis. "

Based on cracked analysis. gg i

MW N

9 PREllMINARY t.

i Amend. 57

CG

. . . FSAR -

TABLE 3.8-42 RADIAL DEFLECTIONS OF CONTAINMENT CONTAINMENT SIT EXTENSO- AZIMUTH METER ELEVATION DEFLECTION (INCHES)

(DEGREES) Predicted 1 Predicted 2 Actual H 103'0" 30 0.0104 H2 103'0" 0.080 0.029 90 0.0104 0.080 0.031 H3 103'0" 150 0.0104 H4 0.080 0.024 103'0" 210 0.0104 0.080 0.039 Hs 103'0" 270 0.0104 0.080 0.036 Hs 103'0" 330 0.0104 0.080 0.044 Hs i 131'0" 30 0.0502 0.380 0.056 H7 131'0" 150 0.0502 Hs 131'0" 270 0.380 0.082 1 0.0502 0.380 0.074 H7 167'0" 30 0.0477 57 1 0.420 0.101 Hs 167'0" 90 0.0477 0.420 Hg 167'0" 0.283 150 0.0477 0.420 0.263 Hao 167'0" 211*-08' O.0477 0.420 H 167'0" 0.166 270 0.0477 0.420 0.168 H 12 167'0" 330 s' R 0.0477 0.420 0.084 %E H9 202'0" 30 0.0484 0.420 0.201 H 2O 202'0" 150 0.0484 me H 21 202'0" 270 0.420 0.271 oR 0.0484 0.420 0.206 *E m H3 230'0" 210 (30) 0.0402 0.390 0.105 0m H4 230'0" 270 (90) 0.0402 0.390 0.1245 Hs*

i 230'0" 150 (330) 0.0402 0.390 0.092 n

o N M 1

Bas 6d on uncracked analysis 9 3.v 2

Based on cracked analysis S

• Extensometers H-13, H-14, and H-15 measure differential displacements between six points on

. . R g

the containment wall (i.e., between azimuths 210 and 30 , 270 and 90 , and 150 and 330,, g respectively). The deflections indicated for these extensometers are one-half of the measured diametric displacements.

Amend. 57

GG FSAR ,

TABLE 3.8-43 RADIAL DEFLECTIONS OF CONTAINMENT AT EQUIPMENT HATCH

, CONTAINMENT SIT EXTENSO- AZIMUTH METER DEFLECTION (INCHES)

ELEVATION (DEGREES) Predicted! Predicted 2 Actual Eg 198'-6" 240 0.0477 0.420 0.233 1

E2 191'-24" 240 0.0477 0.420 0.258 E3 182'-9" 240 0.0477 0.420 0.232 E4 161'-9" 240 0.0477 0.420 0.213 Es 153'-10 " 240 0.0477 0.420 0.187 57 WE% E Es 146'-6" 240 0.0477 0.420 0.167 E7 172'-6" 217*-47' O.0477 0.420 0.200 172'-3" m>

Es 223'-14' O.0477 0.420 0.210 $"

m=

lfEEEEEE E, 172'-6" 230*-14' m Ego O.0477 0.420 0.235

{ma 168'-2" 247'-07' O.0477 0.420 0.205 "

i Et 172'-7 " 259 0.0477 0.420 0.173 E E 12 172'0" 264 0.0477 0.420 f mems 0.170 y e

y 1 Based on uncracked analysis.

2 Based on cracked analysis. ,

Amend. 57

, g Attachment 3 to AECM-83/0669 pgg Page 10 of 17 l

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Figure 3.8-121 is deleted.

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l PRELIMINARY Amend. 57

Attachment 3 to AECM-83/0669

'. Page 11 of 17

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UNIT 1 TAUT WIRE EXTENSOMETER y LOCATIONS - RADIALS AT ELEVATIONS g 103*-6",131*, AND 230' FIGURE 3.8-123 i

Amend. 57 i

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167* AND 202', DOMES, AND VERTICALS FIGURE 3.8124 Amend. 57

Attachment 3 to AECM-83/0669 Page 14 of 17 DRYWELL WALL ~

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Amend. 57 l

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CRACK MAPPING AREA 6 ,

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--aC Amend. 57

___ _ __ - ______m _ - - _ _ - _

' Attachment 3 to AECM-83/0569 Pabe16of17 EFLECTIONS IINCHES)

WALL EL MIN MAX AVG PRED1 PRED 2

, ** 103'-6" .024 .044 .034 .08 .01

\ 131*-0" .056 .082 .071 .38 .05 167*-0" .084 .283 .178 .42 .05

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DISPLACEMENT

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8 DISPLACEMENT e MEASURED VALUE i I i I

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- ANALYSIS.

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! 2. PREDICTED DISPLACEMENT FROM ELASTIC ANALYSIS.

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j ! 3. MIN. MAX AND AVG DEFLECTIONS AT EL 230' ARE EQUAL TO HALF OF MEASURED j/ DI AMETRIC DISPLACEMENTS.

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L AVERAGE DEFLECTIONS AT 17.25 PSIG-WALL AND DOME FIGURE 3.8127 t

Amend. 57

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FIGURE 3.8-128 Amend. 57 i