ML17309A175

Agrees W/Sep Topic III-7.C Info Re Delamination of Prestressed Concrete Containment Structures,Per NRC 810624 Ltr

ML17309A175
Person / Time
Site:	Ginna
Issue date:	07/07/1981
From:	Maier J ROCHESTER GAS & ELECTRIC CORP.
To:	Crutchfield D Office of Nuclear Reactor Regulation
References
TASK-03-07.C, TASK-3-7.C, TASK-RR NUDOCS 8107150058
Download: ML17309A175 (11)

Text

ROCi".ESTER GAS ANO ELECTR!C CORPORATION ~ 89 EAST Av".'IUE. ROChESTEPc, N.Y 1-'6s9 C C~ C" C J>Cl sC ~ ~ 0 c July 7, 19 81 Director of Nuclear Reactor Regulation Attention: Mr. Dennis M. Crutchfield, Chief Operating Projects Branch 45 U. S.'uclear Regulatory Commission Washington, D.C. 20555

Subject:

SEP Topic ZZZ-7.C, "Delamination of Prestressed Concrete Containment Structures"

Dear Mr. Crutchfield:

Rochester Gas and Electric has reviewed the NRC's draft evaluation of this SEP topic, trans-mitted by letter dated June 24, 1981.

We concur in the factual information presented, and agree with the NRC's conclusion that the containment at Ginna would not experience delamina-tion.

Very truly yours, 8107150058 810707 PDR ADOCK 05000244 P PDR

RECT (4~gR

~ y 0 UNITED STATES NUCLEAR REGULATORY COMMISSION M~Grd WASHINGTON, D. C. 20555 June 24, 1981 "C

+w*w+

Docket No. 50-244 LS05 06-093 Mr. John E. Maier Vice President Electric and Steam Production Rochester Gas 5 Electric Corporation 89 East Avenue Rochester, New York 14649

SUBJECT:

SYSTEMATIC EVALUATION PROGRAM TOPIC III-7.C, DELAMINATION OF PRESTRESSED CONCRETE CONTAINMENT STRUCTURES - GINHA Enclosed is a copy of our draft evaluation of'ystematic Evaluation Program Topic III-7.C.

You are requested to examine the facts upon which the staff has based its evaluation and respond either by confirming that the facts are correct, or by identifying errors and supplying the corrected information. We encourage you to supply any other material that might affect the staff's evaluation of this topic or be significant in the integrated assessment of your, facility.

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

In futur'e correspondence regarding Systematic Evaluation Program topics, please refer to the topic numbers in your cover letter.

Sincerely, M.

37.'nnis CrutchT>el , ief Operating Reactors Bran'ch No. 5 Division of Licensing

Enclosure:

As stated cc w/enclosure:

See next page

Mr. John E. Ma i er CC Harry H. Voigt, Esquire Director, Criteria and Standards LeBoeuf, Lamb, Leiby and MacRae Division 1333 New Hampshire Avenue, N. W. Office of Radiation Programs Suite 1100 0 (ANR-460)

Washington, D. C. 20036 U. S. Environmental Protection Agency Mr. Michael Slade Washington, D. C. 20460 12 Trailwood Circle Rochester, New York 14618 U. S. Environmental Protection Agency Ezra.Bialik Region II Office Assistant Attorney General ATTN: E I S COORDINATOR Environmental Protection Bureau 26 Federal Plaza New York State Department of Law New York, New York 10007 2 World Trade Center New York, New York 10047 Herbert Grossman, Esq., Chairman Atomic Safety and Licensing Board Jeffrey Cohen U. S. Nuclear Regu1atory Cornnission New York State Energy Office Washington, D. C. 20555 Swan Street Building Core 1, Second Floor Dr. Richard F. Cole Empire State Plaza Atomic Safety and Licensing Board Albany, New York 12223 U. S. Nuclear Regulatory Comnission Washington, D. C. 20555 Director, Technical Development Programs Dr. Emmeth A. Luebke State of New York Energy Office Atomic Safety and Licensing Board Agency Building 2 U. S. Nuclear Regulatory Comnission Empire State Plaza Washington, D. C. 20555 Albany, New York 12223 Mr- Thomas B. Cochran Rochester Public Library Natural Resources Defense Council, Inc.

115 South Avenue 1725 I Street, N. W.

Rochester, New York 14604 Suite 600 Washington, D. C. 20006 Supervisor of the Tovin of Ontario 107 Ridge Road West Ontario, Nevi York 14519.

Resident Inspector R. E. Ginna Plant c/o U. S. NRC 1503 Lake Road Ontar i o, New Y or k 14519

ENCLOSURE GINNA PLANT DOCKET HO. 50-244 SEP TOPIC III-7.C DELAMINATION OF PRESTRESSED CONCRETE CONTAIHMEHT STRUCTURE I. INTRODUCTION Delaminations of concrete have occurred in the domes of two prestressed concrete containments, Crystal River and Turkey Point. The safety objec-tive of this 'review is to assure that the containment will maintain its structural integrity in order that function.

it may perform its intended safety I I. REVIEW CRITERIA REFERENCES

a. SER Turkey Point No. 3, Docket No. 50-250

b. Containment Dome Report, Turkey Point No. 3, dated February, 1972.

c. SER Crystal River Ho. 3, Docket No. 50-302

. d. Reactor Building Dome Delamination Final Report, Crystal River No. 3 dated December 10, 1976.

Problem history, analyses and repair procedures are described in the above references for the plants where dome delaminations occurred. The containment at Ginna, as described in the FSAR was compared with the containments referenced above in order to determine if such a failure could occur at Ginna.

III. RELATED SAFETY TOPICS AND INTERFACES

1. Containment structural integrity tests are reviewed under SEP Topic III-7.d.

2. Containment tendon inservice inspection program is reviewed under SEP Topic III-7.A.

IV. REVIEW GUIDELINES The containment design and configuration are reviewed in order to assess the possibility that delamination might occur. Recommendations, based on that assessment are noted below.

V. EVALUATION Delamination (cracks in planes parallel to inner and outer concrete surfaces) is caused by radial tension developed in the concrete by the forces from curved prestressing tendons. The curved prestressing tendons attempt to relieve the stresses in them and as a result may cause the concrete delaminate.

It appears that the two most significant factors which led to the

. delamination of the Turkey Point -,3 and Crystal River 3 domes were

~

radial tension in the concrete above the prestressing tendons and the use of a marginal strength coarse aggregate in the concrete.

The containment at Ginna is substantially different from Turkey Point

=3 and Crystal River =.".3 in that the Ginna containment only contains straight, vertical prestressing tendons in the containment wall. There is no prestressi ng in the hoop direc ion of the containment wall or in the dome. Since there is no curvature in the prestressi ng tendons at Ginna, there would be no mechanism to cause radial tension in the concrete due to orestressing forces.

VI. CONCLUSION The containment, at Ginna would not, experience delamination because the containmen. has no curved prestressina tendons to cause radial tension and delamination in the concrete due .o prestressing forces.

~P,5 REMI

+ 0 UNITED STATES NUCLEAR REGULATORY COMMISSION 7 +~1 WASHINGTON, O. C. 20555 0 + May 8, 1980 Docket No. 50-244 Mr. Leon D- White, Jr.

Yice President Electric and Steam Production Rochester Gas 8 Electric Corporation 89 East Avenue Rochester, New York 14649

Dear Mr. White:

RE: SEP TOPIC III-7.D CONTAINMENT STRUCTURAL INTEGRITY TEST Enclosed is a copy of our evaluation of Systematic Evaluation Program Topic III-7.D Containment Structural Integrity Test. This assessment CQHpares your facility, as described in Docket No. 50-244 with the criteria currently used by the regulatory staff for licensing new if facilities. Please inform us your as-built facility differs from .

the licensing basis assumed in our assessment.

, We have discussed this assessment with your staff and believe the facts concerning your plant are correct. Therefore, our review of this topic is complete and this evaluation will be a basic input to the integrated safety assessment for your facility unless you identify changes needed to reflect the as-built conditions at your facility. This topic assess-ment may be revised in the future if your facility design is changed or o I RC criteria relating to this topic are modified before the integrated assessment is completed.

S i nerely, n is M- Crutchfield, C ef Operating Reactors Bran g5 Division of Licensing

Enclosure:

COIKIleted SEP Topic III-7.0

"" w/enclosure:

See next page

Mr.'eon D. White, Jr. May 8, 1980 CC Harry H. Voigt, Esquire Director, Techni ca 1 Assessment LeBoeuf, Lamb, Leiby 8 MacRae Division 1757 H Street, N. W- Office of Radiation Programs Washington, D. C. 20036 (AW-459)

U. S. Environmental Protection Mr. Michael Slade Agency 12 Trailwood Circle Crystal MalI 82 Rochester, New York 14618 Arlington, Virginia 20460 Rochester Comnittee for U. S. Environmental Protection Scientific Information Agency Robert E. Lee, Ph.D.

P. 0. Box 5236 River Campus Region II Office ATTN: EIS COORDINATOR Station 26 Feder al Plaza Rochester, New York 14627 New York, Hew York 10007 f

J ef rey Cohen Herbert: Grossman, Esq., Chairman New York State Energy Office Atomic Safety and Licensing Board Swan Street Bui I di ng U. S. Nuclear Regulatory Coamission Core I, Second Floor Washington, D. C. 20555 Empire State Plaza Albany, New York 12223 Dr. Richard F. Cole Atomic Safety and Licensing Board Director, Technical Development Programs U. S. Nuclear Regulatory Coamission State of New York Energy Office Washington, D. C. 20555 Agency Building 2 Empire State Plaza Dr. Eameth A. Luebke Albany, New York 12223 Atomic Safety and L i cens ing Board U. S. Nuclear Regulatory Comnission Rochester Public Library Mashington, D. C. 20555 115 South Avenue Rochester, New York 14604 Mr. Thomas B. Cochran Natural Resources Defense Council, Inc.

Supervisor of the Town 1725 I Street, N. W.

of Ontario ~ Suite 600 107 Ridge Road West Washington, D. C. 20006 Ontario, New York 14519

S:-P SAFETY TOPIC EVALUAT 0:i R. E. "=I!'."A NUCLEAR POMER STATIQ!'. RGKE Topic III-7.0 Con.ainme.".t S ructural Integrity Tests Introduction In order to. assure tl at a -oncrete containment structure will respond satis-factorily to the postulate'esign pressure loads, a program of measurements, namely the Containment S ruc ural Integrity Test Program, is required to

.demonstrate the correlation with theoretically predicted responses and to

= prove the adequacy of the structure with respect to he quality of construction and material. The scope o this safety topic evaluation is to review the adequacy of the structur 1 integrity testing'procedure used by the licensee and, using current review criteria as a basis, to evaluate the measurements taken during the testing.

Current Review Criteria The current review criteria -.or this soecific safe y opic are:

1. Standard Peview Plan, Section 3.8.1; Z. Regulatory Guide l.lB;

3. ACI 359 (ASHE BPV-III-2) Code Art. 6000.

Rel'ated Sa et To ics ard In erfaces The con.ainment structure in..egrity test, of Ginna nuclear station was oerformed based on the original calculated design pressure o 60 psig. llithin the scope of he SEP sa ety Topic VI-3, "Containment Pressure and Heat Removal Capability",

this original design pressure will be reviewed to assure it's adequacy. Thus, the validity of this saf ty evaluation is continaent upon whether or not a positive conclusion can -'

drawn in the review of Topic VI-3. A reevaluation of this topical review will be necessary if the original calculated design pressure is increased.

Evalua.ion Oescriotion of Structure The con.ainment structure is a ver.ical prestressed concrete cylinder with a reinforced concrete flat '=ase and a hemispherical dome. A welded s.eel liner (3/8" in thickress for he dome and cylinder and 1/4" for the base) is at. ached to the inside face of tre concrete containment str.c ure. The principle dimens-'.ons include an ins;='e diameter of 105'-0" and a height (from too base to spring line, "-. 99'-0". The nominal thickness dimensions of

the reinforced concrete are 3'-6" =or the wall and 2'-6" for the dome. The concrete base slab is 2 ft. thick, with an additional 2 ft. lean concrete fill over the bot-om 1-i;.er plate. A 'etai led description of the struc.ure can be found in the "Final Facility Description and Safety Analysis Report'Ref. 2).

Test Procedure ard Assess.-.ent oi Test Results A de:ailed description of the structural inteority .est for the Ginna.contain-ment is contained in CAI ;".eport 81720, dated October 3, 1969 (Ref.,l). A number of different types of instruments (jio transit, i nvar tapes, LYDT strain gages, photoelastic discs, load cells, etc.) were utilized and are described in the .est report.'he containment vess 1 was pressurized to 69 psig (116 percent of'he design pressure of 60 psig) in five pressure steps (increments) and then depressurized in threo steps. At the maximum test pressure level (69 psig), the pressure was maintained for approximately four hours before the readings, measurements and observations were taken.

measurements and observations were also made at the other pressure step increments. .At these steps, the vessel pressure was slightly'increased above the level at which the measurements were taken and then the pressure was reduced to the specified level and observations made after at least ten minutes to permit an adjustment of strains within the structure. The detailed procedures can be found in the test report.

Based on our review of this report, no unusual response of the containment structure showed up.during the process of pressurization and depressurization.

The displacements (vertical and radial displacements) and the rebar and liner stresses calculated fro-... measured strains were always within allowable limits, except for one displacement which was sliohtly higher than predicted. The observed concrete crack widths and the recovery abater depressurization were also below the acceptabl limits.

Sicnificance of Oeviations from Current Review Criteria The test procedure and tne assessments of measurements described in the report were compared with the requirements stated in the curren review criteria.

The following deviations have been identified:

1. Curent criteria requires more measuring locations for global dis-placement and less =or local displacement.

2. A larger surface area is required by current criteria for observing

.he concrete crack patterns.

3. Current criteria requires the meas r ments of strain near the base of the cylinder and under the prestressed tenden anchor point and vertical displace-.ants on the do-e.. No such measurements were "escribed in the report.

rrert crit=-ria requires that ~he measurements to confirm the covel " of the str cture should be take.. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> arter depres-s". 1zai.io"le ns si.a-;ed in .he repor-:, these ;,.easurements were takeo ~

to 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> a-.ter depress.riization with a slightly lower recovery raie than .-"a requll ed by current cl iteria.

It is .he s.aff's <.-'"e...ent that the deviations identified above are not sicni=icant and wi~ll ."," affect the assessmients made in the section o>

"'"e ..est 're."or. eni.itle='Test Procedur arid Assess ient c Test Results",

since no un Jsual respo .se oi the structure was i ound dur ing the test, Ccnclusion Based on the infor;..a.icn provided in the test report and the FSAR and the evaluation stated abov , we conclude that the test procedure used is.

adeouate and the test results provide a 'basis to assure that the containment structure will safely erform. its intended functions and wi 11 withstand the design pressure load of 60 psig.

'eferences

1. "Structural In.egrity Test of Reactor Containment Structure - R. E,.

. Ginna Nuclear Po~er Station", GAI Report -=1720, October 3, 1969.

2. "Final Facility Description and Safety Analysis.'Report", 'R. E. Ginna Nuclear Power Station Unit No. l.