ML20080M694

Affidavit of Wf Gundaker Supporting Util Opposition to Joint Intervenors Motion to Reopen Record on safety-related Concrete.Corrosion of Reinforcing Steel in Concrete Mat Will Not Occur to Significant Degree.Certificate of Svc Encl

ML20080M694
Person / Time
Site:	Waterford
Issue date:	09/27/1983
From:	Gundaker W EBASCO SERVICES, INC., LOUISIANA POWER & LIGHT CO.
To:
Shared Package
ML20080M667	List: ML20080M663 ML20080M676 ML20080M694
References
NUDOCS 8310040245
Download: ML20080M694 (150)

Text

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. develop a film on the surface of the steel called "passivating film." This passivating film, consisting of gamma ferric oxide, .:

is formed by cement hydration and is maintained in the alkaline environment of the concrete. This film acts like a protective coating applied to the rebar which precludes corrosion from taking l I

place on the steel rebars.

4. For corrosion of any significance to take place on rebars embedded in concrete, the protective passivation filim must first be destroyed. The primary causa for a breakdown of the passivation film is the presence of a high concentration of chlorides, as would be the case in marine environments or in roadways where salts are .

used for snow melting purposes. Laboratory studies conducted by D. A. Hausmann of the American Pipe and Construction Company in South Gate, California, show that the corrosion threshold for rebars is reached when the passivating film starts breaking down.

This threshold is observed when the chloride concentration reaches

-710 ppm in the presence of free oxygen.

5. The possible sources of chlorides contained in the water seeping through the cracks are the concrete mix water, the concrete I itself and the groundwater. Analyses of the concrete mix water taken l

l on 2/3/76 and 4/15/77 showed chloride contents of 28 ppm and 24 ppm respectively. The main sources of chlorides in the concrete are the l

l mix water and the addition of calcium chlorides. The concrete

! specifications for the Waterford 3 Project states that the total l soluble chloride ion content in the water extracted from the concrete l'

mix should not exceed 250 ppm and that calcium chloride shall not be 0310040245 830930 PDR ADOCK 05000382 G PDR o . - - - -- -- - -- - - _ -

added unless specifically authorized. According to information obtained from ESI personnel, no calcium chlorides were added to the concrete in the Waterford 3 Project. An analysis of a groundwater sample recently taken at the Waterford 3 site showed that the chloride content was 34.9 ppm. Thus, it is not possible in my opinion for the chloride level to even approximate the 710 ppm corrosion threshold level necessary to initiate corrosion.

6. Attempts were also recently made by ESI personnel to remove water samples from the concrete mat cracks on the floor of the auxiliary building for analysis. In the first attempt only about three drops of water could be extracted with a hypodermic needle.

No analysis was possible with this amount of water. During a second attempt which lasted several hours, approximately 10 ml of water was gathered. This amount of water allowed the U.S. Testing Laboratory to make only a partial analysis of the sample. A comparison of some

of the parameters measured in the two water samples is as follows

Ground Water Crack Water l pH 6.75 7.95 Iron. 0.07 ppm 0.08 ppm Calcium Hardness (CACO3 ) 169 ppm 212 ppm Chlorides 34.9 ppm Calcium 67.5 ppm 85 ppm

7. The 710 ppm chloride corrosion threshold requires the presence of free oxygen. Hausmann's studies showed that no cor-

• Sample too small for measurement.

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rosion took place in concentrations of up to 3550 ppm of chloride when free oxygen was not present. While water contains dissolved oxygen, the minute amount of groundwater penetrating the cracks would in my opinion be basically classified as stagnant water, and oxygen replenishment would be a very slow process. Thus, it is likely that the chloride corrosion threshold for the rebars in the foundation mat would be significantly higher than the 710 ppm value reported by Hausmann for corrosion in a free oxygen environ-ment.

8. Another possible cause of the breakdown of the passivating film is direct chemical attack on the rebars. Dissolved hydrogen sulfide and dissolved carbon dioxide have been known to attack i

rebars in concrete. Analysis of a groundwater sample recently taken at the Waterford' 3 site indicated a content of less than 0.01 ppm of hydrogen sulfide and a content of 1.32 ppm of carbon dioxide.

Both these amounts are negligible and would have no deleterious effect on the concrete mat rebars. Since the concrete is not a source for the H 2S r the CO2, n direct chemical attack of the rebars is to be expected.

9. A common and direct indication of rust developing in con-crete rebars is the presence of a brownish stain on the surface of the concrete. With the water penetrating through the concrete from the bottom of the mat, such a rust colored stain would tend to deposit on the top surface of the concrete mat if any measurable corrosion is taking place on the rebars. ESI personnel who have

inspected the cracks have reported that no such stains are evident at the Waterford 3 concrete mat surface in the areas of the cracks.

In addition, the minimal-difference between the ferrous and ferric oxide contents on the two water samples analyzed (0.07 ppm for the groundwater and 0.08 ppm for the crack water) indicates that if any corrosion of the rebar is taking place, it is negligible.

10. In my opinion, the amount and nature of the water in the cracks precludes the possibility of eny significant corrosion of the concrete mat rebars.

11. The steel liner of the containment, which is 2-1/2 feet above the surface of the mat at its lowest point, is installed over fill concrete. The same protective type of passivating film would develop on the surface of the liner in contact with the con-crete, and the same statements outlined above for the rebars can be made for the liner.

12. When the cracking was discovered in 1977, my predecessor, A. W. Peabody (now retired) , and M. D. Oliveira (who has since left ESI to return to his native land) were asked to analyze the poten-tial for corrosion of the rebars and the outside bottom plates of the containment vessel. Their opinions were stated in memorandum COR-LW3-77-55M dated August 5,1977, a copy of which is attached as Exhibit B. On the basis of my own evaluation, I agree with the conclusions stated in their memorandum.

13. Based on my own knowledge and experience of corrosion matters, on my review of the literature on the subject, and on the

6-data that I have obtained related to the Waterford 3 project, I can state that there is no reason for me to believe that corrosion of the reinforcing steel in the concrete mat at the Waterford 3 Nuclear Plant would occur to a degree that would have any sig-nificance. Nor do I have any reason to believe that the integrity of the containment liner in those areas attached to the concrete would be affected.

- [L, a, _. _ r-v-u WILLIAM F. GUNDAKER Subscribed and sworn to before me this 7 day of A, , 1983.

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NOTARY PUBLIC KATHERIME if. HAARKE New York I My Commission Expires: NOTARY P SL Sg.

Quaufled in Nassau County p i

Commission Expires March 30,19 e i

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, EXHIBIT A WILLIAM F. GUCAKER . .

3/2 Director of Corrosion Engineering .

EXPERIDCE SLM4RY -

Registered Professional Engineer with over thirty years experience with the g Ebasco Organization of whicn over twenty have been in corrosion desicn -

engineering of fossil and nuclear field electric generating stations, of all kinds of pipeline systems (oil, gas, water, etc.), of marine structures (sheet piling, piers, docks, offshore platforms, etc.), and of miscellaneous industries such as refineries, pulp and paper, chemical plants, tank farms, etc.. Experience also ircludes design of gromd electrodes for HVOC power transmission lines.

Responsible for all designs on cathodic protection, grounding, coatings and lightning protection systems including preparation of design criterias, guides, stancards, coating specifications and plans, specifications and bill j

of materials for all desi G red systems. tctual experience includes the above ,

plus conducting all kinds of field surveys such as soil resistivity, cathodic  !

protection, stray currents, interference testing, construction surveillance, '

system energization and conducting coating evaluation tests.

Administrative responsibilities include preparation of budget, salary administration, staff requirements, etc., for the New York Corrosion Department, and technical responsibilities for the Houston ' Corrosion '

Department. Additionally, performs as Project Manager for all New York office corrosion stand alone projects. .

REPRESENTATIVE EXPERIDCE .

Client Project Florida Power & Light Co. -

St Lucie thit Nos.

. 'l a 2 Carolina Power & Light to. Shearon Harris thit Pbs.1 & 2 Louisiana Power & Light Co. Waterford thit Nos.

1,2&3 Minnesota Power & Light Co. Square Ebtte HVDC ,

Texas Eastern Transm. Corp. Miscellaneous Ibckeye Pipeline Company Miscellaneous Texaco Petroleum Domestic & Fore 15 p Miscellaneous Exxon Company USA & Foreign Miscellaneous Marine Contractors Coating Evaluation DOE Richland Facilities Cathodic Protection e

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. 3/2 l WILLIAM F. CUNDAKER

, REPRESENTATIVE EXPERIEPCE (Cont'd) .

, client . . Project Gulf 011 -.

Miscellaneous Consolidated Edison Grounding Study

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! Jeddah 011 Refining Company Cathodic Protection f

Monrovia Port Authority Chthodic Protection of

, Marine Structures  !

Centromin, Peru Soil Resistivity Survey Long Island LiWiting Co gany Stray Current Study

. DPLOYW.NT HIS10RY Ebasco Services Incorporated, %w York,* N.Y.; 1960-Present' o Director of Corrosion Engineering,1980-Present o Supervising Corrosion Engineer, 1971-1980 o Principal Corrosion Engineer, 1967-1971 o Senior Corrosion Engineer, 1964-1967 o Corzosion Engineer, 1962-1964 o Assistant Corrosion Engineer, 1960-1962 .

Chban Electric Company, Havana, Cuba; 1950-1960 o SsstationEngiber, 1955-1960 -

, o Electrical Technician,

• 1950-1955 l .

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EDLCATION Lhiversity of Havana, Cuba - B!EE - 1955 Adelphi thiversity - NBA - 1979 EGISTRATIONS Ptoressional Engineer - New York, New Jersey, California PROFESSIONAL AFFILIATIONS

, IflCE Accredited Corrosion Specialisi (Present Trustee and Past Chairman of Greater Metropolitan New York  ;

Section; .Present and Past 01 airman, Vice-Chairman and Wrrber of several Technical Committees)

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3h WILLIAM F. GUPOAKER -

Mt0FESSIONAL AFFILIATIONS (Cont'd) -

IEEE - Senior N er -

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(Past Chairman of work Grow on Generatirg Station Grounding Practices) . _

TEDNICAL PAPERS ms written and presented many papers and articles for NACE's National Conference, for the Apalachian Lhderground Corrosion Course, for the Cklahoma, Purdue and Liberty Bell Corrosion Courses, for IGT and for several NACE local sections.

i LAN2JAGES Totally fluent in Spanish .

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. e .as EXHI3IT B August 5, 1977 COR-LW3-77-SSM. .

'* i P Grossman h**%

To: /)

Y/ 'J From: A W Peabody /M D Oliveica

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Subject:

LOUISIANA POWER 6 LIGHT COMPANY WATERFORD SES UNIT 3 -

. CORROSION OF REINFORCING STEEL AND STEEL CONZA7NMENE VESSEL PIATES IN CONTACT WITH WATER In accordance with your telephone request, we have analysed a possible ,

situation in the consnon nat where supposedly ground water weeping from concrete cracks found on the surface of the mat could corrode the  %

reinforcing steel and the outside bottom plates of the Steel Contain-i 1 ment Vessel.

It is a proven fact that concrete by its alkaline nature passivates

. carbon steel embedded in it.

It is also known that water in contact with concrets becomes alkaline and consequently its corrosivity to s' teel decreases considerably.

• In addition to these factors, assuming that ground water i= lafe inside the crack network to a certain extent, this water will be near stagnant and without replenishment.of oxygen. Consequently, the rate of corrosion under the above circumstances, if any, will be negl41ble This applias to the reinforcing rebars as well as* to the outside of the vessel bottom plates, in case the repairs presently being conducted do not fully prevent the water from reaching the vessel.

I MDO/hn cc: R K Stampley .

J O Beoth/B D Fowler .,

D N Galligan *

• L Skoblar W F Gundaker O

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ATTACnMENT 3 H

f HARSTEAD ENGINEERING ASSOCIATES o INC. _

O, 169 KINDERKAMACK ROAD, PARK RIDGE, N.J. 07656

• Phone:(201)391-2115 fgffED N OCT ~3 g,j .

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Il p{, i?tf WATERFORD III SES ANALYSIS OF CRACKS AND WATER SEEPAGE IN FOUNDATICN MAT LOUISIANA POWER & LIGHT COMPANY' REPORT NO. 8304-1 ,

SEPTEMBER 19, 1983 l

Prepared by:

A. V. du Bouchet i3 9 Reviewed by: Wn VM( V A. I. Unsal.

Approved by:

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2 TABLE OF CONTENTS

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1.0 Introduction 1 2.0 Site Inspections and Interviews 2 3.0 Foundation Mat Design Concept 3 i

3.1 Site 3

' 3. 2 Design 3 3.3 Construction -

4 4.0 Significant Events During Construction 6

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4 '.1 Stop Work Order No. 1 _

.6 4.2 Concrete Placement' of the Mat . 6 .

4.3 Change in Allowable Soil Bearing 7 Pressure During Construction ~.

4.4 Site Settlements 7 e

4.5 Cracks Observed in ihe Top of Mat in r

the Containment Area ,

8 I 4.6 Cracks in Mat Outside of Conta nment f:

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5.0 Analysis of Waterford III Structural --

Foundations ,.

11 5 .1. Structural Concept _

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6.0 Review ~ of Engineering Design and '

Construction '

. 13 6.1 Geologic Studies 13 7%

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A TABLE OF CONTENTS (Cont'd.)

l l 6.2 Development of Engineering Properties 14 6.3 Foundation Design Concept 15 6.4 Design of Combined Mat 17 6.5 Earth Pressure Considerations 18 6.6 Groundwater Environment 18 .

6.7 Excavation Sequence 19 6.8 Dewatering Systems 20 6.9 Subsurface Instrumentation Program 21 6.10 Construction of Mat 21 6.11 Summary of Movements Recorded During .

Construction 21

'7 . 0 Evaluation of Cracking 24 8.0 Corrosion Potential 29 8.1 Passivation Mechanism in Reinforced

{ Concrete 29 8.2 Job Specifications 30 8.3 Laboratory Testing 30 8.4 Steel Containment Corrosion 34 9.0 Steel Containment Stability 35 9.1 Ebasco Calculation 1352.063 35 10.0 Conclusions and Recommendations 38

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10.1 Containment Vessel 38 10.2 Foundation Mat 38

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0, TABLE OF CONTENTS (Cont'd.)

REFERENCES 40 APPENDICES A Basemat Crack Maps B. Properties of Subsurface Materials Design Values C Synposis/ Introduction of State-Of-The-Art of Floating Foundations by H. Q. Golder -

D Generalized Site Cross Section E Effects of Soil Modulus on Shear and Moment F Design Envelope's of Mat Shear and Moment G Effects of Foundation Stiffness on Dynamic Shears and Moments H Abasco Services Letter F-16919, W3-NY-1 dated June 29, 1977 I Composite Foundation Mat Differential Settlement Contours J . Composite Foundation Mat Settlement K Crack Width Calculation .

L Steel Containment Stability Calculation M . Laboratory Report 9

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1.0 Introduction This report summarizes a study undertaken by Harstead Engineering Associates on behalf of Louisiana Power and Light Company.

The following major evaluation items are addressed in this report:

a) The engineering criteria employed in the prepara-tion of the site and in the design and construction ,

of the Waterford III Nuclear Power Island Structure (NPIS) basemat.

b) Cracking and leakage in the basemat.

c) The laboratory tests performed on water and leachate samples extracted from the surface of the basemat.

d) The stability calculations performed for the Steel .

Containment Vessel.

As required, relevant source material is either refer-enced or contained as an appendix to this report.

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2.0 Site Inspections and Interviews HEA personnel have visited both the New York office of Ebasco, Inc. and the Waterford III site.

These visits are summarized in HEA Trip Reports Nos.

1-6 (References 1-5), and were conducted in order to meet with key personnel familiar with the design bases of the Waterford III NPIS basemat, to document first-hand the extent of cracking and leaking at the surface of the base-mat, to gather pertinent reports and drawings, and to confirm a scope of work and corresponding schedule.

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1 3.0 Foundation Mat Design Concept 3.1 Site 1 The site of the Waterford 3 plant is locnted next to the Mississippi River. Natural grade is at about Elevation

+ 15.0 feet. To a depth of about 55.0 feet from grade, the soil consists of alluvial deposits which are relatively soft. At greater depth are the Pleistocene Age soils. The upper parts of these soils are stiff.

3.2 Design The Safety class structures are supported on a con-tinuous mat 270 feet wide, 380 feet long and 12 feet thick.

The' bottom of the mat is at a depth of about 60 feet below natural grade. Support for the mat is provided on the

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stiff Pleistocene clays, where the natural soil pressures were about 3300 psf. After the completion of construction, the soil pressure under the foundation mat is about 3100 psf. These pre'ssures consist of the weight of soil and construction above the mat less the buoyant pressure due to ground water. The water table is generally at an ele-vation of +-8.0 feet; therefore, the buoyant pressure is about 3400 psf. The weight of the soil which was excavat-I ed was about 5700 psf, while the weight of the construc-tion now'in place is about 5500 psf. The interesting feature of this is that the soil below the plant is exper-iencing almost the same pressures that it has in recent history. Therefore, increased consolidation of soil and the accompanying. settlement that often occurs when new construction weight is added to soil does not occur in this case.

Inasmuch as the water table is at about Elevation i + 8.0 feet or almost at natural grade, walls were erected

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iQ around the perimeter of the mat. These walls must resist the lateral pressure of the surrounding backfill soil and the-hydrostatic pressure of the ground water. These walls extend up and provide flood protection up to Elevation

+ 30.0 feet, which is 22.0 feet above the normal water table.

The mat and the walls form a reinforced concrete box structure, with interior walls and concrete placements referred to as counterforts providing additional stiffening.

The mat and the exterior walls are monolithic and there-fore prevent water flow through joints and in the sense that ground water is prevented from collecting inside the structure, the structure has been called a floating struc-ture.

3.3 Construction The steps-involved were:

a) Dewatering'the site b) Excavation down to Elev. - 47.0 feet c) Construction of the mat d) Construction of superstructure e) Gradual release of Dewatering f) Backfill of excavation surrounding the construction The soil pressures existing at Elevation - 47.0 feet vary considerably during construction. After dewatering the pressure increases to the weight of the soil above due l to' loss'of buoyancy and then.after excavation the pressure, of course, reduces to zero. When the pressure is reduced, ,

the soil heaves or rises due to the removal of the weight of the overlying soil.

As construction proceeds, the structural weight causes the soil pressure to increase and the soil begins to re-1 i settle. In order to provide additional compaction, the g

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soil bearing pressure is allowed to increase. The pressure was allowed to increase to 4500 psf. As construction con-tinued, the gradual release of dewatering offset the in-creasing structural weight.

The construction was planned to maintain a maximum differential soil bearing pressure of 2000 psf. In the final condition, a maximum differential soil pressure of 1000 psf was established by the designers (Reference.6).

During construction, settlement and water pressure readings were taken in order to encure that control was being maintained over differential mat settlements.

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4.0 Significant Events During Construction 4.1 Stop Work Order No. 1 LP&L issued SWO No. 1 on December 16, 1975 in order

-to correct deficiencies and nonconforming work in the in-spection and control of concrete mixing, transporting and placing of concrete, and curing and finishing. This

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resulted from observation of Placement No. 6.

4.2 Concrete Placement of the Mat During placement of Section 10B, a rain storm broke

,out. The placement was completed; however, because the o concrete quality was unknown due to dilution with rain water, NCR No. W3-39 was filed.

A repair program was established which included cor-ing, strength testing, pressure grouting of drilled holes, repair of surfaces, and waterproofing of the west face.

Discrepancy Notice C-13 dated 12-16-75 noted cracks in the west fac'e of Placement No. 2. Cracks were chipped out and surface roughened prior to making adjacent Place-i ment No. 4.. .

During placement of concrete in Placement No. 19, concrete was placed over a previous layer while it was no longer plastic. This surface was' raked and fresh concrete was placed. Concrete was later shipped out in certain areas to a depth of 6 inches to 12' inches below the mat top rebar and replaced with fresh concrete. Still later, 11 cores were taken to a minimum depth of 5 feet. The cores were tested and the core holes grouted.

The " cold joints" and dilution of concrete are undesir-able because of voids and weaknesses. The extensive and methodical repair program that took place as indicated in the documentation and subsequent observations of the 'foun-dation mat indicate that the repair was effective and that L

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there is no concern about the strength or corrosion pro-tection of the concrete.

4.3 Change in Allowable Soil Bearing Pressure During Construction On March 15, 1977, Ebasco requested that the maximum temporary bearing pressure during construction be increased from 4000 psf to 4500 psf.

This recommendation was based on the fact that the maximum allowable bearing pressure of the soil is 15,000 psf, the desire to accelerate recom-pression of the soil that heaved after dewatering and excavation, and the need to permit backfilling under the Turbine Building.

Due to scheduling difficulties, the dewatering system was not in place during the initial removal of 20 feet of soil.

The remaining soil heaved between 1.5 and 3.5 in- ,

ches.

After the dewatering was under way and about 1.0 inch of the heave was recovered, the job was shut down.

The dewatering was not operating long enough to balance the total heave.

In November 1974, the dewatering was reinstated. In January 1975, the remainder of ' excavation was restarted and the heave increased to between 4.0 and 9.0 inches. '~

When concrete construction proceeded, the heave reduced -

to between 1.0 and 6.0 inches.

The above compares t'o a heave projected as 2.0 inches. "

. Rebound is a function of both load removed and time of *l load removal.

The differences in schedule and loading -

were cited as reasons for the difference. .

In order to ensure full recompression of the rebound, a greater soil pressure was recommended. .

4.4 Site ~ Settlements In September 1978, a report " Review of Site Settle- .

ments" by M.

Pavone and J. L. Ehasz, was issued (Reference 9

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7) . It was noticed that there was a total settlement of about 11.0 inches from the maximum heave position.

Since the maximum heave was previously noted as being between 4.0 and 9.0 inches, the overall settlements were therefore beyond the original zero position of the soil.

The' settlements have remained constant since early 1979.

A curvature in the North-South direction was noted, the center of the mat being 2.5 and 1.5 inches higher than the south and north edges respectively.

4.5 Cracks Observed in the Top of Mat in the containment Area Nonconformance Report NCR W3-535, dated August 3, 1977 reported that cracks were discovered in the' top of mat, which were weeping water. The rate of weeping water was enough to show the crack and to moisten the surrounding

• concrete. A crack map was prepared (Reference 8) and the crack widths were noted as being between 2 to 5 mils. The piezometer levei was kept below Elevation - 50.0 feet since the start of 1975, until September 1977. The concrete in this region'was placed in December of 1975 and January of 1976. The lower concrete ring of the shield building was also in place. The cracks were chipped out to a 1.0 inch depth and to 12 inches on either side and repaired with epoxy grout. It was hypothesized in Ebasco Letter of July 27, 1977 (Reference 9) that the general curvature of the mat caused tension in the upper portion of the mat.

Locally the lower ring wall of the shield building would

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'was termed as a stress reversal and the possibility of an intertie between cracks could exist providing a direct leak path. It was also stated that leakage of water through the mat was undesirable because:

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a) a' film of water could be presumed to develop be-i tween the mat and the fill concrete beneath the containment vessel. This could require a reanalysis of containment stability (see Sec-tion 9.0).

b) if the leakage increased and water found its way out of the fill concrete it would be collected in the mat drainage system and run through the waste treatment system (see -Section 5.1) .

The r.epair apparently stopped further leakage.

4.6 Cracks in Mat Outside of Containment Area Ebasco Nonconformance Report W3NCR-16143 (Reference

,10) noted that:

I "there are concrete cracks in the base mat of the Reactor Auxilliary Building. This is evidenced -

by the percolation of water in small amounts, up through.

these cracks. These cracks are located in the Gas Surge Tank Room, Waste Gas Tank Room, and Waste Gas Compressor "B" Room, all at elevation - 35.0 feet.

These are examples of where cracks were found:

G. Harstead of HEA observed the above-mentioned loca-tions where cracks had been observed, as well as other areas, during the period of July 11-14, 1983 (Reference 1) .

All accessible areas of the basemat were subsequently inspected and any cracks found were mapped during the period of August 30 - September 2, 1983 by A. du Bouchet of HEA with the assistance of LP&L and Ebasco personnel (Reference 5).

The crack maps generated during this inspection are contained in Appendix A. The reference points employed to locate these cracks accord with the geometry detailed in Ebasco Drawings LOU-1564 G-499 S01, -02 and -03.

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,s In addition to mapping the orientation and extent of each crack, notations were made concerning any prior re-pairs to the crack, floor finish or lack thereof, evidence of dampness or seepage, and crack width.

As noted in Reference 5, " Crack width dimensions could not be quantified, but.are designated throughout as

' hairline'. In several instances, the existence of a crack could only be inferred by the darker coloration caused by the presence of moisture. No actual gaps were noted."

.The amount of moisture noted during this inspection period was minimal. In some instances dampness / moisture were present. There was, however, no evidence of seepage .

or migration that might have been deduced by the presence of standing water or draining along.the local slope of the basemat.

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5.0 Analysis of Waterford III Structural Foundations i 5.1 Structural Concept The foundation concept is an ingenious solution of the site problems in meeting the safety criteria estab-lished for the nuclear safety related structures. The most significant factor in assessing the adequacy of the design is that the final soil pressure after construction is actually less than the soil pressure which existed prior to the start of construction. The stability and safety that this' implies has been demonstrated, in that, the settle-ment has not changed for the past several years except for changes that would be expected by changes ih the water table.

As part of this concept, the mat and exterior walls were to remain watertight. If water could readily flow in and no provision was made to pump water out, conceivably the water inside would increase the effective soil pressure and result in further soil settlements, although the effec-tive soil pressure would still be less than one half of the maximum allowable. For many reasons, flooding would be in-tolerable; however, there would probably be little detri-mental effect upon the structure. The differential hydro-static pressure on the exterior walls would be eliminated, thereby reducing the lateral load on the building. The loading on the mat would remain approximately the same because the increased effective soil pressure would be equal to the weight of the water which leaked into the structure. There would be long term settlements in this case and perhaps some differential settlements, which is not an unusual situation in many structures.

Section 11. 2 of the Waterford III FSAR (Reference 11) details the capacity of the Waste Management System (WMS).

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Table 11.2-4 specifies a total expected waste flow of 1425 gpd based on the following flow sources:

Containment Building sump (40 gpd), Auxiliary Build-ing floor drains (200 gpd), laboratory drains and waste water (400 gpd), sampling drains (35 gpd), miscellaneous (700 gpd) and blowdown (50 gpd).

Table 11. 2-2 specifies a total useful internal vol-ume for the two WMS waste removal tanks of 7200 gals.

HEA therefore concludes that the capacity of the Waste Management System as detailed above effictively eliminates the possibility of ground water accumulation within the NPIS. .

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O-6.0 Review of Engineering Design and Construction To determine if implementation of - the unique floating foundation concept resulted in excessive differential move-ments during construction, documents pertaining to the design, engineering, and construction were reviewed. Data included related sections of the FSAR,' instrumentation re-l ports, calculations related to the design, formulation and .

application of the foundation design principle, and rele-vant correspondence. The following specific areas were addressed:

a) Geologic studies-b) Development of engineering properties for founda-tion soils ,

c) Foundation design concept d) Design of combined mat e) Earth pressure considerations-Groundwater environment

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f) g) Excavation sequence h) Dewatering system i) Construction of mat j) Summary of movements recorded during construction.

Pertinent data related to the above will be analyzed to establish that the design concept was developed and imple-mented successfully.

6.1 Geologic Studies The Waterford Nuclear Power plant is located on the west bank of the Mississippi River about 20 miles from New Orleans. The sito consists of over 3,000 acres with sur-face elevations ranging from approximately sea level in the

(v) 4

~. , _ _ . _ . _ _

southwest to about elevation plus 14 feet MSL at the base of the flood control levee.

The crest of the levee is the highest point of the site and is about elevation plus 30 feet MSL. .The

=

Mississippi River is 110 feet deep and about 2,200 feet in width adjacent to the plant.

Geologic studies conducted at the site included review and interpretation of geologic literature, subsurface bor-ings,-geophysical logs, cross-hole data and laboratory tests. The stratigraphic sequence is described as follows (FSAR Section 2.5. 4.1) ;

Sequence

• Depth (Feet)

Recent alluvium deposits 0 - 50 ,

Pleistocene sands and clays 50 - 1,100 Plio-Pleistocene interbedded sands and clays 1,100 - 1,900 Pliocene alternating sands and clays 1,900 - 4,900 Massive sandstone interbedded with shale 4,900 - 7,500 Shale alternating with thin sandstone layers 7,500'-10,500 Marine shales 10,500 -40,000 This review will be confined to the upper 500 feet of soil strata.

6.2 Development of Engineering Properties A' total of 74 soil borings were drilled at the site to determine the detailed stratigraphy applicable to the upper 500 feet of subsurface materials. Static and dynamic engineering properties were established based on laboratory tests on selected samples and in situ geophysical measure-ments. A brief visual description of the principal' soil strata is provid'ed below:

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. ... ~ . ..  :--.-. .-. . . . . . - - - .

==

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Sequence Depth (MSL) 1 Soft clay and silty clay with silt and sand GS to -40 2 Stiff tan and grey fissured clay -40 to -77 3 very dense tan silty sand -77 to -92 4 Medium stiff grey clay with silt lenses -92 to -108 5 Stiff. dark grey clay - organic -108 to -116 6 Soft to medium stiff tan and grey clay -116 to -127 7 Very stiff clays with silts and sands -127 to -317 8 Very dense sands and silty sands -317 to -500 Design values applicable to each stratum are defined in Appendix B (FSAR Table 2.5-14).

Foundation Design Concept

~

6.3 -

In reviewing the consolidation characteristics of the

, potential foundation bearing strata it is apparent that excessive settlement could be anticipated if the Nuclear Plant Island structures were founded directly in the stiff Pleistocene

• deposits (layer 2) unless the total bearing pressure from the structures were reduced during construc-tion by buoyancy effects. Of particular concern are the clays with relatively Low Overconsolidation Ratios (OCR),

such as layer 4, which has an OCR of 1.4.

By excavating a depth of soil approximately equal to ,

the weight of the structure, the effective pressure at the base remains unchanged, thereby reducing. the potential for underlying clay layers to settle. The floating foundation principle has not been used previously in nuclear power plant design; however, historically, many large structures have been constructed using this concept. See, for example, an extract from a pape entitled " State of The Art of Floating Foundations" by H. Golder, which is contained in j.q Appendix C.

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. _ . - _ . - - - _ _ - - _ - - _ . . - - - - - - - - . - - - - - ~ - - - - - - - - --

C)

The following is a brief summary of the significant control parameters developed for the Waterford Plant incor-porating'the floating foundation principle.

a) All Category I structures were combined in a nuclear plant island on a common mat.

b) Base of mat foundation is in the stiff Pleistocene clays at Elevation - 47 MSL.

c) Effective bearing pressure of the Nuclear Plant struc-ture is 3,100 psf. compared to the existing overburden pressure of 3,300 psf.

d) Dewatering systems were required to minimize potential for heave at base of excavation, control pore pressure in layer-3 and stabilize excavation slopes.

e) During construction the total pressure at base of mat

• may have increased to 4,000 psf resulting in an additional pressure of 700 psf. It was estimated that the h, eave after excavation would be in the order of 2 inches and recompression from this additional

. pressure would be complete by the end of construction.

f) A filter' layer of compacted sand and shell (18 inches) was placed at base of excavation underlying mat to permit distribution of pore pressure in the underlying clay on application of load.

g) As soon as the total load from the Nuclear Plant Island and surrounding granular backfill reached 4,000 psf, segments of the dewatering system were released in stages to achieve buoyancy of the struc-tures and backfill and permit construction to continue.

Details of the proposed Recharge Program were devel-oped during construction related to well efficiency and piezametric response.

h) A maximum differential loading of 1 ksf was applied

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to base of structures to minimize tilting, heaving, and settlement.

- 1) A detailed instrumentation program was required to monitor movement of structures and groundwater levels.  !

j) Long term settlement was anticipated to be less than 1 inch due primarily to local pore pressure adjust-ments in the clays.

A generalized site cross-section showing the Nuclear Island structures and adjacent non-Category I Turbine Building is outlined in Appendix D (FSAR Fig. 2.5-80).

6.4 Design of Combined Mat Details of the parametric and sensitivity studies con-

, ducted to establish the appropriate mat thickness and re-quired reinforcement for static and dynamic loading condi- .

tions are outlined in Reference 12. The selection of the subgrade modulus applicable to the foundation soils and mat geometry ir judgemental. The values used to estimate impact on mat thicknesses of 10, 12, and 15 feet (k,=150 pci and 125'pci) are considered reasonable. The twelve foot thickness finally selected was based on an economic compromise between the cost of' additional concrete to elim-inate shear reinforcement and providing some shear rein-forcement in local areas.

The influence of a constant or variable modulus on the shear and moment diagrams is shown in Appendices E and F (Figures 7 and 8 - Reference 12) . The design en-

~

velope selected covers all possible support conditions.

An inherently conservative approach was also adopted in analyzing the mat for seismic loadings resulting from the SSE (0.lg) and OBE (0.05g) . As shown in Appendix G (Figure 9, Reference 12) the total shear and moment in-creases rapidly with increasing foundation stiffness to t

approximately G = 3,000 ksf where G = the Dynamic Shear Modulus.

Although the indications of soil stiffness based upon geophysical site measurements indicated that the value of G should be 1000 ksf, the seismic responses in -

the plant structures would be greater for increased soil stiffness. In order to be very conservative, the seismic analysis was based upon a dynamic model using a G = 3000 ksf which resulted in peak seismic response and therefore peak moments'in the mat. 1 t

The seismic analysis mathematical model contains elastic springs representing the stiffness'of the soil.

The results of the analysis include soil spring deforma- ,

tions which represent soil movement with respect to some j origin point of earthquake. The peak horizontal deforma-tions were used to calculate the passive earth pressure on the perimeter walls.

6.'5 Earth Pressure Considerations i

The procedures outlined in section 2.5 4.10.3 of the l

f FSAR related to determination of static and dynamic earth l

i pressures on the structural walls were reviewed. An "At Rest" earth. pressure coefficient of K = 0.5 was selected for the compacted granular fill.

, This highly conservative approach adopted in determin-ing the earth pressure for dynamic loading conditions (cor-relating movement of structure from dynamic analysis with strain obtained from a typical earth pressure diagram)

! combined with static loads results in a heavily reinforced perimeter wall.

6.6 Groundwater Environment Evaluation of piezometric response in the Recent allu-vium to fluctuations in the river level indicated that the i.

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= .= m=- . - - - . . . .

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clays, silts, and sands were discontinuous and unresponsive.

Average permeability of these deposits is estimated to be in the order of 1.5 x 10-6 cms /sec (FSAR Fig. 2. 5-12) .

Similar conclusions were reached regarding the trans-missibility of potential sand layers in the Pleistocene clays. Below the stiff clays at Elev. -77 MSL it has been stated that all strata are responsive to river level fluc-tuations. The major source of recharge for the granular backfill surrounding the Power Plant is expected to be from rain and run-off, possible interconnections with dis-continuous sand layers extending to the river and the tan silty sand layer at Elev. -77 MSL.

• Water quality has been analyzed and no corrosive ele-ments were detected which could impact the reinforcing steel embedded in the concrete mat (see Section 8.0). The pos-sibility of the water becoming saline at some future date was considered; however, lack of oxygen would prevent cor-rosion from this water source. '

, The greatest potential for corrosive elements to be present in the groundwater immediately adjacent to the concrete mat would be from the Mississippi River; however, the seepage path required in the assumed continuous silty sand stratum at Elev. -77 MSL, relatively low permeability, estimated gradient of 0.008 and probable filter medium f would~ result in a groundwater environment surrounding the l

l Plant with all corrosive elements removed or highly diluted.

f 6.7 Excavation Sequence Unfortunately due to schedule and legal problems it was not possible to complete the required excavation for the Plant Island structures until October, 1975. The fol-lowing is a brief -summary of the major excavation phases i commencing with the initial excavation in April, 1972 l ., s (FSAR Section 2.5.4.5.1).

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Excavated R Stage From To Started Finished Phase I Grade -5 April 72 July 72 II -5 -22 January 75 June 75 III -22 -40* April 75 August 75 ,

i IV -40 -48 October 75 March 76 Turbine Bldg. Grade -40 January 77 March 77 Concurrent with the excavation phases, extensive dewatering systems were installed and operated.

6.8 Dewatering Systems The dewatering systems were- installed by Moortrench-American Corporation based on performance specifications prepared by Ebasco. A total of 251 dewatering wells were .

located around the perimeter of the excavation with 217 pumping from the Recent alluvium and the remaining 34 from the Elev. -77 Pleistocene sands. A second series of 12 deep pump relief wells were located around the combined structure mat and pumped from the Elev. -77 sands. No de-tails were provided on the design of the well systems. It is. assumed these studies were performed by Mooretrench. On evaluating the Instrumentation Reports covering the monitor-ing of the dewatering operation it appears that the sys-tems generally performed as intended. It was noted in a letter from Ebasco to Boh Bros. dated June 29, 1977 (Ap-pendix H) that significant operational and maintenance problems had developed. Corrective action was taken by the contractor and it is understood that the wells were stabilized and instrumentation readings were obtained in conformance with specifications. An extensive Recharge Program (required to achieve buoyancy of the structures) was implemented successfully in October, 1977 and completed by July, 1979 when normal groundwater pressure levels were

(]) achieved.

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

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..j-6.9 ' Subsurface Instrumentation Program The scope of the Instrumentation Program consisted of monitoring. piezometric levels, foundation soil heave, settlement, excavation slope movements and potential site subsidence due to dewatering. A total of 24 piezameters were installed to measure groundwater response in select-ed soil strata. Five additional piezometers were located in the filter layer underlying the combined mat. A total of nine (9) heave points, two (2) extensometers, 'six (6) inclinometers, and twenty-eight (28) settlement monuments were installed to measure movement of structures and ex-cavation slopes. -

6.10 Constructio'n of Mat The excavation for the combined mat was performed with

• backhoes by making an eight (8) foot vertical cut in the stiff Pleistocene clay from Elev. -40 to Elev. -48. The excavation was performed in strips. The initial strip was located under the Reactor Building area approximately 120 feet wide running to the full width of the mat. Subse-quent. strips '(Nos. 2 and 6) were cut north and south of strip No. 1 as shown in Appendix I (Fig. 2.5-118). Con-crete placements were made simultaneously in alternate strips. All exposed vertical cut faces were gunited within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of exposure to prevent dessication prior to concrete mat placement.

6.11 Sunmary of Movements Recorded During Construction On completion.of excavation in October, 1975 for the common mat it was noted that the clays had heaved a total of 5 to 10 inches with the maximum amount occurring at the north end closest to the river. This magnitude of heave was considerably greater than anticipated in orig-inal design (approximately 2 inches) and was due primarily

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to general relaxation of the clays due to the number of

' excavation phases and the stop/ start operation of the.de-

) watering systems.

To compensate for the additional heave, the permissible

. ' overload of 700 psf was increased to 1,200 psf in order to accomplish most of the recompression during construc-tion. This increase' permitted the load from the structures and backfill to be increased to 4,500 psf. prior to com-mencing the Recharge Program to achieve buoyancy. By July, 1977 recompression of the heave had occurred due to loading from structures and backfill, assisted by larger and more efficient dewatering pumps. The dewatering system was con-tinued until October, 1977 when the average net settlement ^

was approximately 2 inches. During the period October, 1977 to July, 1979 the rate of movement was controlled by re-leasing the dewatering system in stages, permitting con-struction of th'e plant to continue. The average net settle-ment increased to approximately 5 inches during this period.

Readings have stabilized ~at that level for the past.four years with onl'y minor fluctuations noted due to change in river level. The' composite foundation mat settlement is L shown in Appendix J (FSAR Fig. 2.5-117).

Detailed review of the instrumentation records cover-

" ing construction of the Plant indicated that the applied structural load was sufficiently controlled so that the permissible maximum differential loading of 1 ksf across the base of mat was not exceeded. Adherence to this cri-terion resulted in minimum deflections and minimum cur-vature (for the mat geometry) at the surface and base of mat. Maximum-differential movement was recorded at 2.5 inches with the maximum settlement occurring at the north and south ends.

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Although the maximum recorded heave and subsequent set-tiement was considerably in excess of original design '

estimates, careful control had been exercised in applying load from structures and equipment in a uniform sequence.

By conforming to the maximum differential loading criter-ion of 1 ksf recompression of the heave and consequently rate of strain was controlled. This procedure minimized unusual and severe distortions of the mat.

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7.0 Evaluation of Cracking While it is not possible to precisely predict ~ stresses in reinforcing bars, an upper bound estimate is possible by estimating the strain as the crack width divided by crack spacing. Assuming crack widths of 5 mils spaced 10 ft.

on center, it may be shown (Appendix K) that the approxi-mate stress in the top rebar is 1200 psi. The actual crack width and spacing would indicate a much lower stress.

Nevertheless, if the conservative value of 1200 psi ten-sion in the top reinforcing bars is conservatively assumed to be constant for the entire 330 feet of length of the mat, the indicated differential settlement 'would be some-what less than 1.0 inch. This provides added assurance that differential soil pressures were very well controlled .

during construction. This also indicates that the mat is quite tolerant of such differential settlements.

Furthermore, settlement stresses are considered secondary stresses in that they do not impair the struc-j tural capacity to carry other imposed loads such as dead load and seismic loads. This is possible provided that

~

there is no failure of the supporting soil. In the case of Waterford III the soil is loaded at about one fourth of the design load and in fact, less than the previous in-situ condition. When this is compared to the reinforcing bar yield stress of 60,000 psi, it is clear that these cracks did not give any evidence at all of any structural distress.

Cracks are expected in reinforced' concrete structures, and are caused by many factors, such as:

application of tensile forces, drying shrinkage of concrete thermal gradients,

. , .s and differential settlements.

!d .

~24-

The last three effects are the result of geometric constraints, which do not limit- the ability of a properly

. designed reinforced concrete structure supported on com-petent soil to carry imposed loads. By " properly designed" ,

it is meant that sufficient reinforcing steel is placed )

in the concrete to prevent.large tensile cracking of the  !

concrete, crushing of concrete,.or diagonal tension shear l failure. s The cracks that were reported 'are :of little concern with- respect to. the structural adequacy of the mat; there-fore, the precise cause of the cracks is not important.

The cracks could be the result of:

shrinkage

~

temperature gradient, -

settlement, or a combination of the above.

However, it is'concludbd that the origin of the cracks detected during construction was not due to severe' differ-ential movements occurring during or immediately_.a'fter application of loads from struNtures and equipment.

The water reported to have / surfaced through the - cracks.

is probably ground water under a. pressure' head. Based on ,

records'of dewatering, there does appear to have been suf-ficient hydrostatic head available to forca water through the cracks observed in the mat. Regardless of the hydraulic '

process, very.little water was' observed. It was described

./

i as "not resulting in generally enough water-to ' form a sheen but enough to definitely show the cracks and to$ moisten surrounding concrete". With the low rat'e .of water we$p-ing and the rather limited cracking, there is no reason for concern.

~ ' In 1983, additional cracks in the mat were reported in O .

2 25- _

-* w pr w + + v *- ---me- -www - g.g-e-- -*ge-r --e-W, m--=mgg - 4e e ' mm ,  ?*e~4 -*s--e+-- 7 -m-m aeP=n --t *re--J

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t Oj areas outside of the Reactor Building (see Section 4.6 and Appendix A). These cracks probably; developed several years ago. During an interview.with Mr. J. Sleger, he stated that one crack was obseived with evidence of seep-age during the late summer of 1979. It is very probable that all of the cracks discovered in 1983 were present for some years. Indeed, several .of these cracks gave in-dications of epoxy repairs.

All of these cracks appear to be the same; namely,' a crack which is either a hairline crack 'or which is idvis-

'ible to the naked eye. Many of the cracks are associated with "leachate", moisture and/or evidence of an epoxy re-pair at the top surface of the mat. Both "leachate" and moisture are observed in very small quantities. These "

cracks are not indicative of any high stress in the rein-forging bars. In fact, based upon the observed cracking, one could conc 1,ude that the foundation mat is virtually unloaded.. If the foundation mat was actually loaded as

. assumed in the design calculation, one would expect con-

. siderably more cracking. This tends to' confirm the statement that the calculations for the mat are indeed con-servative. ,

. Crack widths of anywhere from 1C to 80 mils, depend-ing upon crack spacing, which would not be beyond expecta-tion, are not cause for concern of the structural integrity of the mat.

While cracking of concrete is expected, it is, of course, important to evaluate the cracking for several j reasons: ,

I a) If.the crack width becomes very large and there are corrosive' chemicals and oxygen present the rein-forcing steel may be subject to rusting.

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)

b) Large and extensive cracking may be indicative of forces acting on the structure which can cause damage such that the ability of the structure to resist loads, due to service, is compromised.

c) For the case of the Waterford III mat in particular, seepage of water from cracks may invalidate the

" floating mat" concept and affect the containment vessel stability.

The cracks in the mat have widths that are so small that there is no chance of intrusion of corrosive materials and that corrosive materials are not in the environment within the plant or outside. In the Commentary to ACI 318-71 Section 10.6 it is stated that "To assure protec-tion of reinforcement against corrosion and for aesthetic

• reasons, many fine hair cracks are preferable to a few wide cracks." From the observation of the Waterford III mat, one would 'have to describe the situation as one of a few hair cracks much less than the many fine hair cracks y envisaged as a preferable condition.

The observations of the cracks indicate the seepage of water up through cracks carries with it "leachate" which contains primarily calcium carbonate and magnetic iron. The leachate apparently seals the cracks because many of the cracks show leachate deposits which are now dry. This self sealing process may eventually eliminate leakage; however, seepage is still in evidence even though the process has probably been underway for several years.

Nevertheless, the present seepage is minor and poses no dif ficulties. ,

Since the advent of Portland cement in construction, it has been known that steel reinforcing bars embedded in I

Portland cement concrete are protected from corrosion.

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O Quoting from the Commentary to ACI-318-71 Section 10.6 "Recent extensive laboratory work involving modern de-formed bars has confirmed that crack width at service loads is proportional to steel stress." As noted above, the observed cracks indicated a very low stress in the re-inforcing steel.-

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G 8.0 Corrosion Potential 8.1 Passivation Mechanism in Reinforced Concrets In order to assess the potential for corrosion in the reinforcing steel of the NPIS basemat, several references concerning corrosion of steel in concrete were reviewed (References 14-18) .

As noted in Reference 14, "the corrosion resistance of steel in Portland cement concrete has been recognized for more than a century. The protective mechanism, not des-cribed until recent years, is due to a passivating film of gamma ferric oxide which is formed and maintained in the alkaline environment produced by cement hydration".

As noted in Reference 15, " Iron and steel are not thermodynamica11y stable in water. Either acid or neutral .

water corrodes iron and forms a ferrous solution. This solution, in contact with oxygen, oxidizes to form hydrated ferric oxide -- a major constituent of rust. If the water is sufficiently alkaline, at pH 8 to 14 for example, the Fe23 O and Fe 340 which form are relatively insoluble and

. deposit a protective film on the metal surface. The metal is then said to be passivated".

The passivating mechanism, therefore, requires an alkaline environment (pH of about 12. 5) and an absence of oxygen in order to form a protective film on the surface of the reinforcing steel.

The alkalinity of the water derives from the hydra-l tion of the cement, which generates calcium hydroxide.

[

A relatively oxygen-free environment is generally insured by careful control of the concrete mix and its subse'quent placement. Depth of concrete cover is also a i

factor.

As noted in Reference 16, "In addition, concrete of 5 -

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. _ _ . _ . _ _ _ _ ,n . ._ .

[h low water-cement ratio and well cured has a low perme-ability which minimizes penetration of corrosion inducing factors -- oxygen, chloride ion, carbon dioxide, and water."

8.2 Job SpecificationsSection I, Paragraph 7.3 of the Ebasco Concrete Masonry specification (Reference 19) stipulates that: "The aggre-gate, sand and water combined in the same amounts as in the concrete mix shall not contain a total soluble chlor-ide ion content of more than 250 ppm water when water is extracted from the combination after being thoroughly mixed, unless the Engineer allows a deviation in writ-ing...".

• Section I, Paragraph 9.7 of that specification further requires that: "No hdmixture containing chlorides to an extent that the requirements of Paragraph 7.3, with the admixture mixed with the water, are exceeded shall be ac-ceptable unless the Engineer allows a deviation in writ-ing...".

Section II, Paragraph 8.4 of that specification also stipulates that: " Calcium Chloride shall not be used for accelerating the set of the cement.in any concreta con-l taining reinforcement or embedded metal parts".

The. limitation on the maximum allowable soluble chloride contained-in the' concrete mix defined in the Reference 19 specification is subsequently verified by the sampling and testing procedures mandated by that i

specification.

8.3 Laboratory Testing l In order to deduce any evidence of corrosion in the basemat reinforcing steel, several water samples and a solid (leachate) sample were subjected to laboratory I

1 analysis.

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-.-.j O 9 The three water samples subjected to lchoratory analysis were obtained at the following locations:

a) Water rising in Conduit No. 33074, which rises near the West Temporary Electrical Pit, runs to the southeast for approximately 90 feet, and again rises above the basemat. At the south end, no water was rising, indicating a blockage to'the flow of water. The conduit is located approxi-mately 3 feet below the top of the basemat.

b) Ground water flowing through conduits which extend from the side of the mat to the East Temporary Electrical Pit. '

c) Water collecting at a crack in the Waste Gas Tank ,

Compressor B room.

The solid sample was collected along the top surface of a crack located along an east-west axis between column lines R and Q 1 , and straddling column line I '

M The laboratory report summarizing the results of the analyses performed on these samples is contained as Ap-pendix M.

As noted under ' Testing Methods and Results' each of the three liquid samples were subjected to analysis for .

pH, chloride, alkalinity, iron, calcium and sodium. The results of these analyses are subsequently tabulated on page 2 (note that samples designated 'l', '2' and '3' accord with the order in which the' sample locations are defined herein).

The value of the pH obtained for sample 1, 12.5, ac-l cords with the pH of concrete, as previously noted. The pH of 7.5 obtained for samples 2 and 3 is due to tre car-bonation process which normally occurs at the surface of concrete exposed to open air.

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

O-l As noted in Reference 14, " Free carbon dioxide re-duces pH by carbonation, but only to a depth of a few millimeters in sound concrete". .

The report results indicate the virtual absence of iron in the three liquid samples, a clear indicator that the chemical constituents of rust are not present. The ppm of chloride are also well,within the maximum allow-able 250 ppm mandated in the Ebasco Concrete Masonry

]

4 specification (Reference 19), as previously noted.

The solid (leachate) sample was subjected to spec-trographic and X-ray diffraction techniques. Iron and .

Calcium'are identified as the two major chemical consti-tuents contained in the solid sample.

I As noted in the appended laboratory report under

' Remarks', the' calcium hf droxide liberated during the hydration of Portland cement will form calcium carbonate l'n the presence of carbon dioxide; the iron content con-tained in the solid sample is identified as magnetite.

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The I esults of the testing of the water samples and leachate are consistent with the process of corrosion pro-tection of the steel reinforcing bars embedded in the concrete. As a matter of interest, it should be noted that the reinforcing bars are large. In general, the top

~ .

reinforcing bar diameter is 1-3/8 inches while the bottom reinforcing bar diameter is 2-1/4 inches.

These properties accord with the properties of the iron compound which (under properly controlled conditions) forms a passivating film on the surface of the reinforc-ing steel (see the initial extract from Reference 15) .

It is interesting to note that this deposition mechan-ism also occurs in boilers, and is succinctly stated in Section 6, page 129 of Mark's Standard Handbook for Mech- -

anical Engineers (Seventh Edition) :

"At saturation temperatures above moder-ately low' pressures, a second mechanism

, predominates, in which iron removes oxygen I

from water or steam, forming iron oxide and releasing hydrogen:

3 Fe + 4H 2O --- Fe 03 4 + 8H It is noteworthy that this mechanism does not require the intervention of dissolved gaseous oxygen in the water, which is often the rate-limiting factor in the electrochemical i

i corrosion discussed earlier in this sub-section.

The stable oxide at boiler temperatures in a non-oxidizing environment is magnetite, Fe 34 0 (ferrous ferrite). A normal protective skin of magnetite is formed from the underly- ,

ing steel".

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l L _- , _ _ _ _ . - _ _ _ ._- _ - _ _ _ _ .

- O, On the basis of the foregoing evaluation, it is therefore concluded that there is no evidence to infer the existence of basemat rebar corrosion in the vicinity of a crack.

8.4 Steel Containment Corrosion As noted in HEA Trip Report No. 6 (Reference 5), an inspection of the annular area between the Containment vessel and the Shield Building revealed some surface cor-rosion at the base of the Containment Vessel, which might be due to'the presence of water generated by construction activity.

As soon as this area can be adequatel? controlled with respect to the presence of such construction-related water, it is the recommendation of HEA that a program be implemented to clean and field paint the base of the Con-tainment Vessel to insure that'the corrosion process has been eliminated in this area.

9 t . ._ _ , ._ -

. iO) 9.0 Steel Containment Stability -

9.1 Ebasco Calculation 1352.063 Ebasco Calculation 1352.063 (Reference 13) was ex-ecuted as a consequence of Ebasco Nonconformance Report W3NCR-16143, dated May 11, 1983 (Reference 10).

Attachment III to that NCR notes that "The effect of i postulated widespread hairline cracking of the basemat has been investigated by Civil Engineering for stability of the Containment vessel against flotation and overturning under buo'yant conditions caused by postulated groundwater intrusion...".

An attached memorandum from P.-C. Liu*to B. Grant dated May 24, 1983 specifically indicates that the stabil-ity of the containment vessel has been reviewed for a

• postulated hydrostatic infiltration to Elev. -1.50 feet.

An examination of Ebasco Drawing No. 1564 G-817, Rev. 13,

dated 02/03/83' designates El. -1.50 ft. as Top of Pier,

-4.5 feet below the tangent line of the cylindrical shell and the ellipsoidal base of the Containment vessel.

Ebasco Calculation 1352.063 (Reference 13) assumes that the base of the Containment Vessel is flat, and com-putes the safety factors against uplift, sliding and over-turning due to the effects of E-W DBE, vertical DBE and j buoyancy.

l The factors of safety against uplift, sliding and overturning initially computed are 2_.44, 2.51 and 6.77.  :

At a meeting held at Ebasco's New York office (see Section 2.0) it was agreed that Ebasco would revise the stability calculation to reflect'the SRSS of the E-W and l

N-S DBE's, and to reduce the dead load of the Containment Vessel by the magnitude of the buoyant force. ,

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

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O-The Revision 1 calculation, dated 07/28/83, computes revised factors of safety against sliding and overturning l of 1.17 and 3.16.

In order to confirm the stability of the containment Vessel, a simple stability model was formulated by HEA (Appendix L) which takes the curvature of the base of the Containment Vessel into account.

This stability model is formulated on the basis of two intrinsic properties of the ellipsoidal base of the Containment Vessel: that sliding and translation of the base of the Containment Vessel with respect to the mass concrete support cannot be uncoupled, and that any dis-placed configuration of the base of the Containment Vessel will result in "two-point" contact (points designated 'j' and 'k' on page 3 of the Appendix L calculation) . The lat-ter assumption' derivrs a from the fact that the radius of curvature of the ellipsoidal base of the containment Vessel is not a constant.

4 As shown in the computation, the critical stability mode for the Containment Vessel is overturning and not sliding. The factor of safety computed against overturn-ing is 2.34.

HEA'therefore confirms the stability of the Contain-ment Vessel under the action of the postulated earthquake and buoyancy forces.

The HEA computation also confirms the structural adequacy of the underlying mass concrete supporting the Containment Vessel as shown in Detail "B" of Ebasco Draw-ing LOU-1564-G-502, Rev. No. 6, dated 12/17/78.

L Factors of safety against uplif t, sliding and over-

[ turning were also computed for the Shield Building with l respect to the top of the Mat. The respective factors of l

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safety calediated were.3.23, 1.35 and 1.32, which do hiot take into account the additional shear and axial restraint that would be generated by the reinforcing steel tieing the Shield Building and the Mat together.

HEA therefore additionally confirms the stability of the Shield Building with respect to the top of the mat.

1 ,

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10.0 Conclusions and Recommendations 10.1 Containment Vessel The steel Containment Vessel is seated on a concrete dish. If it is assumed that hydrostatic pressure develops on' the interface between the bottom head of the Contain-ment Vessel and the supporting fill concrete, .there would be a reduction in stability. Calculations were performed which indicate a more than adequate margin of safety.

Therefore, it can be concluded that the cracking and seep-age in the foundation mat could extend into the' supporting fill concrete without causing any concern about the Con-tainment vessel stability. *

~

Quite independent of the cracking in the foundation mat, some surface corrosion was noted on the lower cylin-

'drical portion of the containment vessel. This surface corrosion has not affected the strength of the Containment vessel. However, this surface corrosion should be cleaned and the steel protected to prevent future corrosion.

10.2 Foundation Mat While certain difficulties were encountered during construction and procedural changes were made, they were resolved in a controlled manner so that there were no ad-verse effects upon the structural integrity of the founda-tion mat.

Cracks in the mat were reported in 1977 and again in 1983. However, it is likely that the cracks reported in 1983 were in existence for some time but were only noticed in'1983. In fact, if it weren't for the moisture associ-ated with the cracks, the cracks might not have been noticed at all. The extent of cracking is minor and is certainly within expectations for a structure of this type. The specific causes of the cracks are probably-a

combination of temperature effects, drying shrinkage and differential soil settlements under imposed loads.

While the cracking can be considered minor, the seep-age of water through the foundation mat contrasted with statements that the foundation mat was a " watertight barrier". However, the limited amount of water seepage does not invalidate the fundamental assumption that the foundation mat can support and maintain the imposed hydro-static pressure of the groundwater.

It w'as also determined that there is a self sealing of the cracks by the leachate. The leachate has two major components; calcium carbonate and magnetic

• iron. This magnetic iron is probably magnetite, Fe 0 34 which is the passivating oxide which forms on and protects the steel embedded in the concrete from rusting. The water taken from a crack is not very dissimilar to water taken from the ground surrounding the foundation mat. In neither case is the water considered aggressive.

Furthermore, visual inspections of cracks reveal no evidence of rusting. If corrosion of reinforcing bars in

-tie concrete were a problem it would be expected that the cracking would be extensive. This is because corrosion products of. iron occupy a much larger volume than that of the iron. The resulting expansive forces would cause addi-tional cracking and open up existing cracks and a rust discoloration would appear. The inspection and testing revealed no indications of such a corrosion pror.tess.

As a matter of fact, the cracking in the foundation is minor and there are no corrosive agents within the NPIS nor are any expected in the future. Therefore, there is no need to perform a program of crack repair or peri-odic inspection. Indeed, the leachate appears to provide O. .

j

", -. . . . . . - . .. . .: .-. . . . . ~ .- - . - ~ . - - . -

(3' for a self sealing process.

While the laboratory test results indicated that there was iron in the leachate, the sample of pit water in-dicates virtually no iron. This strongly suggests that the iron is not currently waterborn and therefore is not now coming.from the reinforcing bars. While the source of the iron is not known, it probably occurred over the past seven years of construction. Possible sources include pipe threading and sweeping of the floor with steel bristled brooms.

In conclusion, there is no evidence of any process which has been or could be. detrimental to the structural integrity of the foundation mat. .

e 6

. _. -~ _ 1 _. __ .. . ~ . _ . . . _ . . . . ..

F D.

• 1 REFERENCES

1. HEA Trip Report No. 1, W3-HE-LP-001, July 15, 1983.

2. HEA Trip Report No. 2, W3-HE-LP-002, August 1, 1983.

l

3. HEA Trip Report No. 3, W3-HE-LP-003, August 22, 1983. .

S

4. HEA Trip Reports Nos. 4& 5, W3-HE-LP-004, August 24, 1983.

5. HEA Trip Report No. 6, W3-HE-LP-006, September 6, 1983.

6. Foundation Design of the Waterford Nuclear Plant, by l J. L. Ehasz and E. Radin, December, 1973.

7. Review of Site Settlements, by M. Pavone and J. L. Ehasz, September, 1978.

8. RCB Foundation Crack Map, Ebasco Drawing No. SK 1564-4.1-

G-28, August 17, 1977.

9. Ebasco Letter D'oc: CH-039-77, File: 60-R-4, July 27, 1977.

10. Ebasco Nonconformance Report W3NCR-16143, May 27, 1983.

11. WSES-FSAR-UNIT-3, Section 11.2, Liquid Waste Management System.

l 12. Compatibility of Large Mat Design to Foundation Conditions, by J. L. Ehasz and P.-C. Liu

13. Ebasto Calculation OFS No. 1352.063, Steel Containment Stability, Rev. 1, July 28, 1983.

i

14. Steel Corrosion in Concrete, by D. A. Hausmann, Materials i

Protection, November, 1967, pp. 19-23.

15. The Mechanism of Steel Corrosion in Concrete Structures, -

.by C. T. Ishikawa and B. Bresler, Materials Protection,

S..

March, 1968, pp. 45-47.

ty i

-e _ _

'.)4

16. Mechanisms of Corrosion of Steel in Co.ncrete, by G. J. Verbeck, ACI Publication SP-49, June, 1975, pp. 21-38.

17. Criteria for Cathodic Protection of Steel in Concrete l

Structures, by D. A. Hansmann, Materials Protection, October, 1969, pp. 23-25.

18. Cathodic Protection of Steel in Concrete, by R. C. Robinson, ACI Publication SP-49, June, 1975, pp. 83-93. ,

19. Ebasco Specification Concrete Masonry, Project Identifica-tion No. LOU-1564.472, Issue Date: December 31, 1971.

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APPENDIX B Properties of Subsurface Materials Design Values

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6fSEC- FSA R-!MIT. 3 TABLE 2.5-14 .

PROPERTIES OF SURSURFACE MATF.RTALS _

DESIGN VALUES Visual Unified Coefficient Unconfined Overcon- Average Stratus . Soil Specific Natural of Permes- Compresolve Undrained Drained solidation Shear Young's rois-Descrip- Descrip- Gravity Denalty bility Strength Shear Shear Ratio Modulus Modulus son's {

tion Elev.(MSL) tion (Cs) (PCF) (k) cm/sec qu(kaf) Strength Strenath (OCR) Coas(ksf) E (kaf) Racio - M* l

-6 Clay and sitty Grade,to e ' .Cu . . 2.70 li t. 1.5 10 , 1.0 e = 0.5 KSF c' = 0 KSF 1.5 1200 3'600 0.48 clay with allt -40 e0

{} and sand (Recent f = 0' p'=25' 2.0 material **) .

Stiff tan and -40 to Cal 2.72  !!9 1:10 ~8 3.0 c - 1.5 KSF c' = 0.8 KSr 3.4 3900  !!,600 0.49 gray fissured -77 a p = 0* p=12.5*

2) cl*F

3) Very dense can -77 to --

2.70 125 3 10 -5 _ _,

,, = 0 KSF -

3900 11,500 0.48 eitty sand -92 y = 41*

i Hedium stiff

! -92 to CH 2.74 119 --

2.4 e = 1.2 KSF c' = 0.8 KSF 1.4 3900  !!,600 0.49 gray clay -108 f = 0' p'=12.5*

l 4) with salt l lenses l u l

i. 5) Stif f dark -10s to un & Cu 2.6s 104 --

3.6 e = 1.s KSr c' = 0.s KSF 1.7 3900  !!,600 0.49 l

/, gray clay - -!!6 / = 0* A= 12.5* l

$ organic l Soft to me- -!!6 to HL & Ct. 2.69 119 -

1.4 c = 0.7 KSF c' = 0.8 KSF 2.0 3900 11,600 0.49

6) dium stiff can -127 g = 0* J' = 12.5" ,

and gray clay '

with sand

• lenses l Very stiff -127 to CH & CL 2.71 119 --

4.0 c = 2.0 KSF c' = 0.8 KSF 1.5 3900 11,500 0.48

-317 y'

7) clays

,4ic, with to soa p.o o p. 12.5 2.4 eande ,

Very dense -317 to --

2.70 119 -- - -- -- -- 10,000 29,000 0.45 sande and -500 to

0) sitty sande 125 *

• Computed from field Vp and Vs measurements

• • Eucavated and replaced with compacted backfill Note: The average sheer moduli values are averaged from maximum eheer moduli obtained from field geophysical test results. They are repre-eentative only for low shear straine of approaimately 10~6 in./in.

O Y

I APPENDIX C Synopsis / Introduction of .

State-Of-The-Art of Floating Foundations

, by H. Q. Golder l

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544 SETTLEMENT CONTHOL

.lOtti'llill O[ llW dation is supported partly by the shear strength ami partly lay "hunyancy," .

1. e., the loumlation is partly floating. The ihnet to a lloatmg itanulation is a Soll. MITilANICS AND l'Ol'NDATIONS DIVISION ship in waier.

. This analogy of a ship andteates certain uselut tarts, namely:

1 % c c cli.mgs eil lhe .\merican h.oricly of (..wi.l I..ngmeers

1. The ship displares a volume of water equal to its weight;

2. after the initial settlement there as no further settlement;

3. the pressure on the horieuntal base is uniform;

4. if the water level ccasca to be uniforan, li. c., wave artson) the pres-STATI:-OF-Tilli-AltT OF FLOATING FOUNDATIONSil s& M W.Mtb MN dg M *nImag

5. slwre are lateral pressures on the side of 6he hull; ily Hugh Q. Gohler I kl. ASCL. *

6. ilm water has no shear strengthe i. e., shear strength is not necessary for suppsers; ;uul

7. the hull as complete wha:nlaunched intothe water ami launching stresses {

, may lw high.

SYNOl' SIS Tiw preceding facts imlicate that, for .: lloating hmmlatinus, the sont must For a floating foundathwithe soilmust have weight but ilneednot have shear have weight last it need not have shear strength. The soun.lation must he able

trength. TVw foundation must tw ableto resistpressure on its base andsides to resist pressure on ats base and sides and :( tie wesght or level on the soil l and.if the weaght or levelo! the soll varies.the pressures willnot he uniform. varies, the pressures will not tw uniform, and alwar ami hendmg torres will and shear and lending forett will art on the foundation. In practice, most act on the foundation. Once in position, the foundation will not settle further foundations are partly I'oating. ami almmd all so-called floating foundations (il fully floatmg) unless further load is athled. In soil of low shear strength ase only partly floating because a small residual pressure is usually left on constitaction oi lloating foundations mayle thiticult laat thedepth is not linnted the soll. After a history of their development amt the reuons for their use by shear strength if suitable construction procedures are used; however, are t iven, the problems to be considered in using floating boundations are " launching" stresses in the founuation may be high. In practice, most foun-exantined. Anmng the most important problems are excavation.hottom heave, dations are partly floating, and almost all of the so-ralled floating foundattuns settlewent and tilling, and structural problems, are only partly floating twcause a sm.ill ressdual pressure is usually left on the soil.

A floater.g toundation will he considered herein as one in uharh the greater part of the butidmg load is balanced against the weight of excavated maternal

"" " "N"" '

" ""E' "' #

INTL 10Dt'CTION In conshlering an engmecring problem it is often helpful to begin with,the ll!STOltY i limits between which the problem lies, in a physical sense, although these limits are not necessarily practicut in an enghicering sense. The conce!x of using !!oating foundations as not new. Tlwre is some evi-When a foundation rests on the ground surlare it is supported by the shear dencethat they were used in the tilthcentury,andit is probable that they were strength of the soil or rock of which the ground is composed. When the foun- used intuitively before that date.

dation is placed below the ground surface, for frost or drying protection, and In the discussion2of a paper by Casagrande and Fadum K.'Terzaght refers the weight of the overburden is deducted from the applied pressure, the foun- to a German work by G. Itagenddated 1t170, he which there is reterence to the use of floating foundations by Jolmitennie in Losuhui at the Albion klills. There Noto. .-lhscunsieme pen em it Aus;n>l I, I!wai To ustusul tisee rinseng elates amo tn.. nth, s no ou 1 gen reMy knN Mnat a htang fammWWn was. He says a written ruquest musthe filuel with the Owrutivu Scerutary, ASCM. Tlus tsapur is part ". . .a heavy building can still Iw safelybuilt by shiking it partly autothegroumi of thocopyinhted.f ern:ilot the S.il h h.mics and Finnalati ma lhvision, procuualmgs so that at actually 11 oats. The complete weight ut the lasitiling must not he .

..f the American S.ricar of civil 1:ns:nrer*.. Vol. 91. No. Skl2, AI.irrh, Itms. greater than that cf the excavated 'naterial."

d This is osio est the "sLite...f-the%ert"papeers 1.retschb=.1 at Out ASCH Sell klcohannew The referenice to Ite mic is taken from " A Treatwe on the Steam Engine" an.1 r..un.tatinna laivise m conf. .m *:usun es F.anutili.ni* Inr Contrat os Statiumunts.

• hy J. Farey (1827). Accordmg to llagen. Farcy says hulst at Northwesturn tiniv., Evanstem. Ill., .haeur. I! nit, time comsulmt popurn wnres pret-suitud ist the Septemhur.1961, ativixion .lournal.

8 Cons. Civ. Engr.. II. 4 Coldor anil Anacs.1.til.. Toremto. Cana<ta. ., Tarzaghi, K.srt, discummimi uf"Applicatime os Suel kluch.unca. us inumigning IMhng Finnulations,* 12y A. C.amagramlo aimi it. E. F.a. lum Transacisuais AsCE. Vut. Iou, auI 8,

p. -ta7.

3 Blagun. G., "llandlascialur W.saimurtunukunst." Krust U. Korn, Inurhas, luiu.

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O APPENDIX D Generalized Site Cross Section 6

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o APPENDIX G Effects of Foundation Stiffness on Dynamic Shears and Moments e

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O .d APPENDIX H Ebasco Services Letter F-16919, W3-NY-1

• dated June 29, 1977 e

e G

4 e

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• 4 "N EBASCO SERVICES l INC0RPO8ATED

• ~

UTIL!rY C O N sU LTANTs -

r N GINEE R s -

CONSTRUCTORS

, P. O. Box 70 K111ona, Louisiana 70066 June 29, 1977 F-16919 W3-NY-1 Boh Brcthers Construction Company .

P. O. Drawer 53266 New Orleans, Louisiana 70153 Attention: Mr. R. J. Drueding .

LOUISIANA POWER AND LIGHT COMPAhT WATERFORD STEAM ELECTRIC STATION 1980 - 1165 TfW INSTALLATION - UNIT NO. 3 CONTRACT W3-NY-1, EXCAVATION AND DEWATERING Raf: (1) Ebasco letter F-7419, dated May 18, 1977' '

(2) Bob Brothers letter (H. G. Chapman to J. O. Booth), dated May 27, 1977 Gentlemen:

As you ara aware, tha afficiant operation ot' the antire dawatering system (ejector wells and pumped relief wells) is critical to continued unhampered construction of the Waterford Nuclear Project. During the last few' months, it appears that the maintenance of the dewatering system and monitoring of the site instrumentation has deteriorated to a point that the dewatering system cannot support the recharging effort. This situation is totally unacceptable i

and corrective action must be taken immediately.

! The implementation of the Recharge Program is dependent on two primary factors.

t

, First, the effective foundation loading on the underlying Pleistocene soils must

( be controlled within certain limits as established by the PSAR (Preliminary Safety Analysis Report) and the job specifications. This will be accomplished by controlling the hydrostatic uplift or buoyant weight of the nuclear plant island structure. Second, the foundation soil response of the clay strata will be monitored with the site instrumentation and adjustments made to the Recharge Program as necessary to meet the design intent with respect to controlling heave and settlement of the combined structure. The Recharge Program will be im-plemented at the direction of and will be coordinated by the Ebasco Site Soils Engineer.

Due to the uncontrolled rise in piezametric levels oeneath the nuclear plant island during the last two months, initizi implementation of the Recharge Program must now be delayed. The rise in piezometric levels has been actributed by Moretrench-American to be entirely the result of water used for compaction of the fd,_ , backfill. Although evidence has shown that the compaction water from the

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. . l EDA 3CU SERVICES l

sseereauss -

I p Boh Brothers Construction Company June 29, 1977 J .

backfill has infiltrated the slotted casing of the p. umped relief wells which extend through the backfill, we do not believe that this is the total cause for the rise in piezametric levels.

We note that the flow rate of the primary dewatering systen (ejector wells),

as report,ed to Ebasco by Boh Brothers, remained essentially unchanged at 180 gym from January 9, 1977 (when the new south leg of the system went into operation) until June 9, 1977. On June 9, 1977 Boh Brothers reported the flow rate of the primary system was 82 gym, a drop of more than one-half from the readings of the previous five months. Subsequent checks of the system by Boh Brothers and by Ebasco indicate that the primary system is presently operating at an average rate of 100 to 110 gym, or a little less than two-thirds of the flow race that has been reported for the last five months. On the basis of the above information, it is obvious and was admitted by Moretrench-American in a meeting on June 16, 1977 that maintenance during that time period has been minimal. We find it difficult to believe that there has been no reduction in the efficiency of the system as contended by Moretrench. The apathy indicated by such performance is intolerable.

Implementation of the Recharge Program is scheduled to begin in early August.

The entire program will take from 6 to 12, months, depending on the criteria stated above. It is imperative chat the entire dewatering system (both the .

ejector system and pumped relief wells) remain in top working order until the Racharge Prog:.am is complet'ed and you are . directed to remove the system.

In addition, as noted above, monitoring of site instrumentation and reporting of results has been very poor throughout the proj ect. The following table indicates the frequency of readings reported by Boh Brothers from March 1977 to ifay 1977: '

r Instrument -

Date Total March April May Required Actual

1. Observation Wells 15 16 16 65 47

2. Piezonettra 1-10 0 3 3 '13 6

3. Piezameters :11-c. 4 4 5 '13 13

4. - Extensometers 3 1 13 8 5.. Heave Points 4 2 13 9

6. Dewatering Flow Race 15 11 9 65 35

7. River Elevation 14 12 12 65 38

, 3 8. Inclinometers (Compressions) 1 0 1 13 2 v

9. Inclinometers (Deflections) 5 . 2 4 13 11 Total 62 54 53 273 169

EBA5r; SERVICES 838GAPORATE3 Boh Brothers Construction Company June 29, 1977 A quick review of the table indicates only 62% of the required readings were taken during the three month period. This frequency of monitoring is un-acceptable. This problem has been addressed before (Ebasco letter F-7419, dated May 18, 1976) and acknowledged by Boh Brothers letter dated May 27, 1976. The effectiveness of the Recharge Program depends on timely collection and com-pilation of the data. It is expected that all instrum'e ntation will be read on time and the results reported within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as required by Contract and as committed to by Boh Brothers Construction Company.

In conclusion, it is apparent that. the caliber of service has diminished with respect to the dewatering ' system and the site instrumentation. Implementation of ene Recharge Program cannot start until control of the piezometric pressures has been re-established. A successful recharge program is essential to this project and all parties must adhere rigidly to the Contract requirements and responsibilities.

Yours very truly, a V.

J. O. Booth Project Superintendent GFG/)ah l cc: J. M. Brooks I

L. Elliott -

E. Henderson -

e R. Dawes

• E. Boyd S. Sha11 cross, R. Watt D. Mallette /

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APPENDIX I Composite Foundation Mat Differential Settlement Contours "

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l MAT DIFFERENTIAL SETTLEMENT CONTOURS IN INCHES VERSUS BLOCK 6 MOVEMENT

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4-1-76 8

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KEY:

BLOCK NUMBER f BLOCK PLACEMENT

/ DATE 3-30-76 3

AMENOMENT NO. 19.(6/811 LOUISIANA COMPOSITE FOUNDATION MAT -

Figure POWER & LIGHT CO. OlFFERENTI AL SETTLEMENT CONTOURS l Waterford Steam 2.5 118

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I APPENDIX J Composite Foundation Mat Settlement

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!.9 APPENDIX K Crack Width Calculation ep l

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'HARSTEAD ENGINEERING ASSOCIATES o INC.

PR OJ. NO. 83O&

A 169 KINDERKAMACK ROAD. PARK RIDGE. N. J. 07656 C- \ -

Z- l WATEMFoR.O R seq SUBJ. SUBDIV. SHEET

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APPENDIX L Steel Containment Stability Calculation J

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PROJ.No. 2 ? O "

'HARSTEAD ENGINEERING ASSOCIATES o INC.

A 169 KINDERKAMACK ROAD. PARK RIDGE. N. J. 07656 C-I -1 -1 W3 - M AT SUBJ. SU BDIV. SHEET

,] PROJECT

'- 1 '1 : * ; ' 'l ^ M Y/ I 1- *;*I PR EP. BY Ad0

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'HARSTEAD ENGINEERING ASSOCIATES o INC.

PR OJ. NO. Do4

• A 169 KINDERKAMACK ROADL, PARK RIDGE. N. J 07656 C- l -

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W 3 - ed A T SUBJ. SUBDIV. SHEET g PROJECT

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HARSTEAD ENGINEERING ASSOCIATES

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PR OJ. NO. h)0 &

p 169 KINDERKAMACK ROAD, PARK RIDGE. N. J. 07656 0- \

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( H PROJ.No. 8 L 0 "

'HARSTEAD ENGINEERING ASSOCIATES o INC.

169 KINDERKAMACK ROAh PARK RIDGE N. J. 07656 C-t -l ~

5 '

W 5. - H A T SUBJ. SU BDIV. SHEET PROJECT CLIENT ' 1'1:~: s in A F r,i t r ; ' ; G x -

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O APPENDIX M Laboratory Report . G l O .

g twin citv testanca no - --,nc. 1

                                                                              ;n =.u.,,=-
                                                                              .o.
      .ih,.h                  neronT or:             IDENTIFICATION OF LEACHATE W

PreascT LOUISIANA POWER & LIGHT PROJECT NUMBER 8304 DATE: September 9, 1983 REPORTED TO: Harstead Engineering Assoc Inc runnisseo aY: Attn: Gunner Harstead copics To: 169 Kinderkamack Rd

     .                      Park Ridge, NJ       07656 LABORATORY No.                1-34799 MTRODUCTION This report presents the results of our recent testing of samples you submitted for analy-sis. We received four samples; three liquid and one solid, for testing. The samples
 ,      were identified as follows:

1. Liquid Conduit

2. Liquid, Pit

3. Liquid, Crack -

4. Leachate We understand the samples -were taken from a reinforced concrete mat foundation. The foundation is under hydrostatic pressure from an elevated water table. The purpose of our testing is to evaluate the likelihood of corrosion in the reinforced concrete.

CONCLUSIONS Based on the results of our testing, it is our opinion the following conclusions are appropriate:

1. The leachate consists primarily of calcium ., carbonate and iron. Much of the iron is magnetic, suggesting a form such as , magnetite. The iron appears as fine wire 41ke pieces under magnification.

2. The water removed from the conduit is substantially different than the water

[ obtained from the crack and the pit. The high pH and alkalinity of the conduit l sample suggests the water has been in ' contact with the concrete for an extended period of time.

3. The chloride level in the water is sufficiently low to classify the fluid as not being aggressive.

TESTING METHODS AND RESULTS [ On August 18, 1983, we received four samples for analysis. The samples consisted of

, thr

a plastic containers of liquid and one solid leachate sample. Each of the fluid samples was tested for pH using colorphast indicator sticks. Also, each fluid sample was analyzed for chloride using the Standard Methods for Water Analysis, 407A. In addi-tion, alkalinity, iron, calcium and sodium was determined for each of the fluid samples l

using EPA Method 600/4-79-020. The following results were obtained: l

e *

                                               %               cmin ciev testincs O           .no .no.n   ,no    or.coeu. nc.
                                                        '*                         .Au     N $ 14
                  ., -                                                         ,~ o m   .2m-h h$hjf                  neront or:          IDENTIFICATION OF LEACHATE Nar                                                                                                    o4Tr. September 9, 1983 PAGE:         2 LABORATORY No.             1-34799 TESTING METHODS AND RESULTS (cont.)                                  ,

Sample Sample Sample Constituent 1 2 3 pH 12.5 7.5 7.5 Iron (ppm) ND* ND 1.7 Calcium (ppm) 375 71 31 Sodium (ppm) 2400 1400 5100 Chloride (ppm) 78 20 22 Alkalinity (CACO3) (ppm) 1300 - -

                                                              *ND = Not Detected The leachate sample was analyzed using a Jarrell Ash Emission Spectrograph. The sample was placed in carbon electrodes, and a filin of the spectra was obtained with a D.C. arc.

The following constituents were identified: Concentration Constituent (s) Major Constituent Iron, Calcium (10% or greater)

                                                                                         ~

Minor Constituent Sodium, Aluminum (10% to 1%) Trace Constituent Aluminum, Magnesium, (1% or less) Manganese, Titanium, Barium, Copper l Th2 leachate sample was also analyzed using X-ray diffraction techniques. The diffraction analysis identifies crystalline material which is present in the sample. The sample contains a majo'r amount of calcium carbonate. 1 l REMARKS i l Scale found on the surface of Portland cement concrete is typically comprised of calcium carbonate. During the hydration of Portland cement, calcium hydroxide is liberated.

   -     In the presence of carbon dioxide, the calcium hydroxide will form calcium carbonate.

The carbonation layer is generally limited to the top 1/8" of a quality concrete. l l

   .       -                                                                                                                                    l

b cwin ciev carscinca l O .mo .o u. .

                                              ..                            W50,lC?<

g.a, .o. . . , mm. h, hh j neront or: IDENTIFICATION OF LEACHATE

     ' Q s"                                                                                            D AT E: September 9, 1983 PAGE:            3 1.ABORATORY No. 1-34799 REMARKS (cont.)                             -

The corrosion of reinforcing steel may form a magnetic residue such as magnetite. This formation requires an aqueous environment where oxygen levels are low. The very low c iron content of the water samples suggests the water was not in contact with steel active-ly corroding. The formation of magnetite is observed frequently when steel corrodes in a chloride contaminated cementitious material and is then exposed to air. The low chloride levels found in the water suggest the presence of the iron in the leachate is not from such a condition. The test results are consistent with the iron originating from the surface of the slab. TWIN CITY TESTING AND ENGINEERING LABORATORY INC

s. $6 L '

I Richard D Stehly, P.E. Chief Engineer RDS/st O

                                                             . . , , . .   .       . . . _ .    . . ~          . _ _ .

89 o h UNITED STATES

                                    ' NUCLEAR REGULATORY COMMISSION
                                                                                                      . ATTACHMENT 4
    $                a                                REGION IV I*                    611 RYAN PLAZA DRIVE, SUITE 1000 ARLINGTON, TEXAS 76011

[ ED 000** ~

                      ,                              JUN 3 0lss3 5 001-3 Ni :22 In Reply Refer To:                 -

Docket: 50-382/83-18 [oss;.,%,,4 OmcE or sEcatif,p - p ..., ggg3 O MfyjpcyEmi Louis.iana Power & Light Company L. V. Maurin,.Vice President

                                                                     \8a%# [

aA ATTN: m .3p' Nuclear Operations 142 Delaronde Street New Orleans, LA 70174 4 Gentlemen: This refers to the inspection conducted under the Resident Inspector Program by Messrs. G. L. Constable and T. A. Flippo of this office during the period May 1-31, 1983, of activities authorized by NRC Construction Permit CPPR-103 for the Waterford Steam Electric Station, Unit 3, and to the discussion of our findings with members - of your staff at the conclusion of the inspection. Areas examined during the inspection included preoperational test procedure review, preoperational test witnessing, emergency drill postponement, emergency operating procedures review, Hayward Tyler pump question, training, concrete comon foundation mat, and human

engineering walkdown. Within these areas, the inspection consisted of selective examination of procedures and representative records, interviews with personnel, and observations by the inspectors. These findings are documented in the enclosed inspection report.

During this inspection, it was found that certain of your activities were in violation of NRC requirements. Consequently, you are required to respond to this violation, in writing, in accordance with the provisions of Section 2.20,1 of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regula'tions. Your response should be based on the specifics contained in the Notice of Violation enclosed with this letter.' , One open item is identified in paragraph 7 of the enclosed inspection report. In accordance with 10 CFR 2.790(a), a copy of this letter and the enclosures will be placed in the NRC Public Document Room unless you notify this office, by telephone, within 10 days of the date of this Ig.g% letter, and submit written application to withhold information contained therein within 30 days of the date of this letter. Such

                                                                                                              .f
                                                                                                                       $,g   h$

application must be consistent with the requirements of 2.790(b)(1). $yty[T@ s

                                                                                                              $ YdW           E CUM

                     =                                                                                                                                 '

Y

            .Loeisiana power & '                                                                                                                                                                                                                                                             JUN 3 01983 Light Company                                                                                                                  .

The response directed by this letter and the accompanying Notice is not subject to the clearance ' procedures of the Office of Management and Budget as required by the Paperwork Reduction Action of 1980, PL 96-511. Should you have any questions concerning this inspection, we will be pleased to discuss them with you. Sincerely, G. L. Madsen, Chief CthW Reactor Project Branch 1

Enclosures:

1. Appendix A - Notice of Violation

2. Appendix B - NRC Inspection Report 50-382/83-18 .

cc w/ enclosures: Louisiana Power & Light Company ATTN: R. P. Barkhurst, Plant Manager P. O. Box B Killona, LA 70066 Louisiana Power & Light Company ATTN: F. J. Drummond,' Project Manager, Nuclear 142 Delaronde Street . New Orleans, LA 70174 Louisiana Power &'. Light Company ATTN: Tom Gerrets, QA Manger 142 Delaronde Street New Orleans, LA 70174 - o O e 5 4

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o a APPENDIX A NOTICE OF VIOLATION Louisiana Power & Light Company Docket: 50-382/83-18 Waterford Steam Electric Station, Unit 3 Permit: CPPR-103 Based'on the results of an NRC inspection conducted during the period of May 1 through May 31, 1983, and in accordance with the NRC Enforcement Policy (10 CFR Part 2, Appendix C), 47 FR 9987, dated March 9,1982, the following violation was identified: Failure to Follow Procedures Durino Startuo Testing Criterion V of 10 CFR 50, Appendix B, requires that activities affecting quality be accomplished in accordance with prescribed instructions or procedures. Startup Administrative Procedure (SAP-20), " Orientation, Training, Qualification, and Checkout of Startup Personnel," - Section 4.2.2, requires that training shall be required for personnel who perfonn inspection, examination, and testing activities that assure the quality of components in the nuclear power plant during the startup phase. Also, Section 4.3.2.1 requires that technicians be familiar with the tools to be used and demonstrate proficiency in their use. Contrary to the above,. on May 17, 1983, the NRC resident inspector observed that two Louisiana Power & Light Company mechanical maintenance personnel were attempting to obtain vibration readings on safety-related CCW Makeup Pump A. Neither of the individuals was familiar with the task assigned, as they did not know how to read the instrument properly. m This is a Severity Level V Violation. (SupplementII.E) , (382/8318-01) l Pursuant.to the provisions of 10 CFR 2.201, Louisiana Power & Light . Company is hereby required to submit to this office, within 30 days of the date of this Notice, a written statement or explanation in reply, including: (1) the corrective steps which have been taken and the results achieved; (2) corrective steps which will be taken to avoid further violations; and (3) the date when full compliance will be achieved. Consideration may be given to extending your response time . l for good cause shown. l Dated: JUid 9 n mg,

                                                                                                                                      .                     O 9
                         ,    , . , , - _ . . -            .-n-,-     , , _ . , , - , . . , - - - . . _ ,          .,

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                                                                                                                                          \

i APPENDIX B U. S. NUCLEAR REGULATORY COMMISSION REGION IV NRC Inspection Report: 50-382/83-18 Docke,t: 50-332 Construction Permit: CPPR-103 . Licensee: Louisiana Power & Light Company (LP&L) 142 Delaronde Street New Orleans, Louisiana 70174 F5cility: Waterford Steam Electric Station, Unit 3 Inspection At: Taft, Louisiana Inspection Conducted: May 1-31, 1983 _ [A : J, m3 Inspectors: m / G n . TonstabT W Resident Inspector Date

b. L h$3 N k W 1972 '

T. A. Flippo,'IResident Inspector 'Date Approved: W u=w W. A. Crossman, Chief - 6/L9/83 Date Reactor Project Section B Inspection Sumarh Inspection Conducted May 1-31, 1983 (Report 50-382/83.18) - I'- l Areas Inspected: Routine, announced inspection of: (1)Preoperational Test Procedure Review; (2) Preoperational Test Witnessing; (3) Emergency Drill Postponement; (4) Emergency Operating Procedures Review; - (5) Hayward Tyler Pump Question; (6) Training; (7) Concrete Common Foundation Mat; and (8) Human Engineering Walkdown. This inspection involved 193 inspector-hours onsite by two NRC inspectors. ! Results: Within the eight areas inspected, one violation was identified , l (Failure to Follow Procedures During Startup Testing, paragraph 4). -

                                                                                                                           ~

e e

                   -___- __ --                                      .                _                      m.___                                   _....m
                                                                                                                                                                        .._                  w               _.
  .                          .                                                                                                                             y                                                    -
                                                                                                      -     2-
                               .                                                              - DETAILS

1. . Persons Contacted Principal Licensee Personnel

                          *R. Barkhurst, Plant Manager B. Wier, Preoperational Test Director T. K. Armington, Lead Startup Engineer T. Gerrets, QA Manager

0. D. Hayes, Opera.tions Superintendent, Nuclear C. J. Toth, Manager, Nuclear Training D. Packer, Acting Training Center Manager Z. Sabri, Ph.D., Director, Nuclear Training

                            *Present at exit interview.

In addition to the above personnel, the NRC inspectors held discussions with various operations, construction, engineering, technical support, and administrative members of the licensee's staff. -

2. Plant Status as of May 31, 1983 The Waterford 3 site is presently in the preoperational testing phase.

3. Preoperational Test Procedure Review The NRC inspectors reviewed the preoperational test procedure for performing response time testing of the plant protection system (SPO-66-004,."PPS Response Time Test"). The procedure was reviewed for technical content, compliance with the Final Safety Analysis Report (FSAR), and compliance with licensee's administrative procedures.

  ..                       No violations or deviations were identified.

4. Preooerational Test Witnessing The NRC inspectors witnessed the performance of portions of the following preoperational test procedures:

SPO-36-002 . Component Cooling Water (CCW) Flow Balance and Pump Performance Test SPO-58-001 Refueling Water System SPO-64-001 Control Element Drive Reed Switch Actuation 5

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s g The NRC inspectors verifi.ed that observed portions of the testing were conducted in accordance with approved procedures and evaluated the perfomance of licensee personnel conducting the tests. While witnessing portions of SFO-36-002, " Component Cooling Water (CCW) Flow Balance and Pump Performance Test," the NRC inspector observed two LP&L mechanical maintenance personnel attempting to obtain vibration readings on CCW Makeup Pump A, a safety-related pump. It became apparent to the NRC inspector that neither of the individuals was familiar with the task assigned, as they did not know how to read the instrument properly. Startup Administrative Procedure SAP-20, " Orientation, Training, Qualification, and Checkout of Startup Personnel," Section 4.3.2.1, states, in part:

                  " Familiarity with the tools to be employed 'and demonstrated proficiency in their use."

This is a violation. (382/8318-01) In addition, SAP 20, Attachment 5.9, " Mechanical Maintenance Qualification Certificate" for the two LP&L mechanical maintenance personnel was not properly filled out. - The NRC inspectors had no further questions in this area.

5. Emergency Drill Postponement LP&L's Waterford 3 exercise-for-score, scheduled for May 25, 1983, was postponed by the request of the state of Louisiana because of severe flooding conditions in the state.

This announcement was made on May 25, 1983, jointly by LP&L, the Nuclear Regulatory Commission, the Federal Emergency Management Agency (FEMA), and the Louisiana Departments of Public Safety and Natural Resources. The reason for the postponement was the existence of actual emergencies in the state due to heavy rains and flooding. It was stressed by all agencies that if a real emergency had existed at Waterford 3, the state would have responded as required. Had the exercise been held as scheduled, the state emergency forces would have been required to participate in a mock drill at the expense of diverting their efforts from the real emergencies then occurring in Louisiana. At this time, a new date has not been established for the . emergency drill. No violations or deviations were identified. l . .

             ..               ..                              .-----                          p .: -                   :     : = = .=. ; -- ,            = =        .- a            =--

4_

6. Emergency Operating. Procedures Review

                                         'The NRC conducted a 2-day meeting with senior members of LP&L's operations department to discuss comments on the emergency operating procedures (EOP),which were submitted recently. It was the observation of the NRC that, while the revised procedures

(.May 1983) were a great improvement over the earlier versions, there remained a number of technical and human factors problems with the procedures. Each of the nine E0P's were reviewed with the licensee, step-by-step, to try to help identify the problem areas. It was concluded at the end of the meeting that a revised set of procedures would have to be submitted. The NRC was concerned that the Waterford 3 personnel did not appear to be aware of the importance of having good E0P's. Their principal concerns appeared to be getting the procedures approved so they could get their operators licensed, and to minimize the impact on their administrative staff. The importance of allowing their operators to participate in the development of the E0P's was . also pointed out to LP&L personnel. While they agreed that this was important, they indicated that their operators were required to work full time in ' preparing for their exams and could not assist in preparing or reviewing procedures. The NRC representatives stated that using the operators in the development and validation process for the E0P's is' vital to ensuring that the procedures will be acceptable to the. operators. This item will remain open as identified in opert item (8315-01). . No violations or deviations were* identified.

7. Hayward Tyler Pump Question

! In response to Region IV management request and IE Bulletin 83-05, i

                                            "ASME Nuclear Code Pumps and Spare Parts Manufactured by The Hayward Tyler Pump Company," the NRC inspectors found that there are no Hayward Tyler pumps on the Waterford 3 site, but LP&L has used, and is planning to continue using, spare parts from the Hayward Tyler Pump Company (HTPC).                                                 IE Bulletin 83-05 requests that each '

licensee or construction pemit holder planning to use spare parts for ASME Code pumps address the HTPC recommendations on installation of replacement parts and to conduct a pump performance / endurance test, if necessary, to ensure reliability of the pump. LP&L representatives are now addressing the questions raised by

• l the IE. Bulletin.

This is an open item (8318-02). No violations or deviations were identified. r--_, , _ . . _ , - _ . , . . . . _ _ , , _ . . , , , , , . _ _ . . . , . . . . .,,.,,._,,..,,m.y.,.w. ..y_ ..____,,,__7 ,,_,.,m, _,...,_,.,o.,,,,m-m._, ,,,,. . ,m..,m__

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8. Training During this inspection period, prospective reactor operator (RO) and senior reactor operator (SRO) candidates were given NRC-type l written and walk-through exams. These exams were given by QES l Corporation (QES), Gaithesburg, Maryland. 1 Written Exam SRO R0 Candidates 29 15 Average Grade 71.8% 72%

Passed (> 80%) 2 1 Failed with all categories > 70% 6 3 Failed with one or more categories <70% 21 11 The low grades were explained as having been the result of the operators working on shift for several months and not having had their intensive review training. In add 1 tion, it was felt that they had trouble answering comprehensive essay questions of the type on the . exam. Based on the results of the oral walk-tnrough exams, QES felt that 27 individuals (20 SRO, 7 RO) would have passed an NRC exam even without the planned intensive review. Oral Walk-Through Exam Results SRO R0, Candidates 29 15 High (Probability of Success) 2 - Good (Probability of Success) 13 3 Fair (Probability of Success) 9 5 i Low (Probability of Success) 4 7 No (Probability of Success) 1 - No violations or deviations were noted.

9. Concrete Comon Foundation Mat On May 17, 1983, the NRC senior resident inspector received a call from a reporter for Gambit newspaper. The reporter stated that he had received a phone call from an individual who said that LP&L had found cracks in its 12-foot thick comon foundation mat on May 11, 1983.

The NRC inspector learned that Nonconformance Report (NCR) 6212 had been written as stated describing cracks through which water was percolating into the Reactor Auxiliary Building. The NRC inspector examined

• the foundation mat and found damp spots in the concrete that formed an irregular line. When viewed from a distance of 6-10' feet, one might speculate from the wet spots that a crack existed; however, on close examination, no crack is visible. LP&L chipped into the concrete surface at one location to expose an area for better visual inspection. Although moisture seepage was. present, no crack was '

visible. The amount of seepage is very small, usually evaporating

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s i immediately, resulting in a damp spot on the concrete. LP&L representatives evaluated the indications and determined that they were not significant. The NRC inspector reviewed records of previous, similar problems and discussed the matter with civil engineers experienced in concrete work. The NRC inspector had no further questions at this time.

10. Human Engineering Walkdown I On May 3,1983, the NRC inspector reviewed the completed Human Engineering Deficiencies (HED) listed below and verified that each deficiency was completed.

B-1-24 B-4-14 B-6-24 B-8-2 B-2-2 B-4-15 B-6-25 B-8-4 - B-2-3 B-4-16 B-6-26 B-8-5 B-3-1 B-5-9 B-6-32 B-8-7 B-3-3 B-5-13 B-6-33 B-8-13 B-3-4 B-5-15 B-6-35 B-8-15

                                                                                                                                                                                                                                                        ~

B-3-7 B-5-26 B-6-36 B-8-16 B-3-10 B-6-8 B-6-37 B-8-18 B-3-12 B-6-9 B-6-38 B-8-19 B-3-13 B-6-12 B-6-41 B-8-20 B-3-35 B-6-15 B-6-44 B-8-31 B-3-17 B-6-17 B-6-45 B-8-34 B-4-3 B-6-19 B-6-46 B-8-35 B-4-4 B-6-20 B-6-48 B-9-3 B-4-8 B-6-23 . No violations or deviations were identified.

11. Site Tour 1

   -.                   At various times during the course of this inspection period, the NRC inspectors conducted general tours of the Fuel Handling Building, Reactor Auxiliary Building, Turbine Building, and Reactor Building to observe ongoing construction and testing.                                                                                                                                         .

No violations or deviations were identified.

12. Exit Interviews

 ~

The NRC inspectors met with the licensee representatives (denoted in - paragraph 1) at various times during the ccurse of the inspection. The scope and findings of the inspection were discussed. G

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                                              -..       , . . - - , . . , - - - - , - - , , - ~ - . , - . , . - - , . . - ,
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1 s .. - . . . ~ __ .__.m - mw . . _ _ ._ , l ATTACHMENT 5 d

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          "   -m   Lzace l OU1SI APJ A                34,ocLAnOnce srnser P O W E F1 & L i G H T! R Q. BOX GOC 8 + NEW ORLEMS. *LOutSIANA          (50413GS 234570174 5' Pun!!?s2                                                                                  13 00T -3 All :22 August 30, 1977 0FFl!,E        Stcet!Ag 00CMETtNG & SERvirI LPL 73989 RANCH h     CQ-3-A35. 02.3 3-A1.04.01 Mr. E. Morris Howard, Director, Region IV U. S. Nuclear Regulatory Comission Office of Inspection & Enforcement 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas            76012        ,

SUBJECT:

Waterford SES Unit No. 3 Docket No. 50-382 Incident No. 8 Potenti21 Significant Construct. ion Deficiency Radial Cracks - Common Founda:: ion Mac -

REFERENCE:

Telephone comunication dated August 1,1977, A. E. Henderson (LP&L) to W. G. Hubacek (NRC).

Dear Mr. Howard:

Thio will confirm the telephone conversation this date between A. E. Henderson (LP&L) and your Bob Taylor i~n which LP&L reported that Incident No. 8, " Radial Cracks - Common Foundation Mac" has been evaluated as unreportable as a signi-fictne construction deficiency under 10CFR50.55e. Documentation of the evalua-tion is available at the Waterford 3 construction site. Yturs very truly, l D. L. Aswell l Vico President-Power Production DLA:AEH:jh1 bas: Ebasco (2), J. M. ' Brooks, J. O. Booth (2), D. L. Aswell, L. V. Maurin, A. E. Henderson, D. B. Lester, P. V. Prasankumar , H. W. Otillio, J. A. Reine, L. Biondolillo, C. G. Chezem, T. F. Correts, D. N. Calligan, C. J. Decareaux, W. N. Fadden, S. A. Alleman i l

(," ' ,'. / .. - ~ - ^ ~ - ~ ' - - ~ ' '~ '

                                                              ,_ _. _ _ - 7 ,                              ' ' ' ~ ~ '     ~

UNITED STATES

  '       0 '-      .
                           ,,                   NUCLEAN REGULATORY CoMMISSICN _              ATTACHMENT 6
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REGlcN IV -

                           $                     411 RYAN PLAZA DHIVE.SulTE 1000                    '.',

ARLINGTON, TEXAS 7G011

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September 21, 1977 UDRC Reply Refer To: ' 83 00T -3 AU :22 Docket !!o. 50-332/Rpt. 77-08 rerice F SEcrty

                                                                                    ": cum ;y      A o : "E Louisiana Power and Light Company ATTH: Mr. D. L. Aswell Vice President Power Production 142 Delaronde Street New Orleans, Louisiana           70174 Gentlemen:

This refers to the inspection conducted by Mr. W. G. Hubacek and other-members of our staff during the period August 30 through September 2, 1977, of activities authorized by NRC Construction Permit ilo. CPPR-103 for Waterford Steam Electric Station, Unit flo. 3 and to the discussion of our findings with Mr. A. E. Henderson, Jr., and other members of your staff at the conclusion of the inspection.

• Areas examined during the inspection and our findings are discussed in the enclosed inspection report. Within t' ese areas, the inspection consisted of selective examination of procedures and representative records, interviews with personnel, and observations by the inspectors.

During the inspection, it was found that certain activities under your license appear to be in noncompliance with Appendix B to 10 CFR 50 of the !!RC Regulations, " Quality Assurance Criteria for fluclear Power Plants." The items of noncompliance and references to pertinent requirements are-identified in the enclosed Notice of Violation. We have also examined actions you have taken with regard to previously

               ' identified inspection findings. The status of these items is identified in paragraph 2 of the enclosed report.

One new unresolved item is identified in paragraph 13 of the enclosed report. This notice is sent to you pursuant to the provisions of Section 2.201 < of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations. Section 2.201 requires you to submit to this office, with-in 30 days of your receipt of this notice, a written statement or expla- ' nation in reply including: (1) corrective steps which have been taken by '

                                                                                                                                 \

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_...,.___ . . . . _ _ _ _ . - . . _ _ _ . _ _ _ _ = ._ _ . LouisianaPower&O.ghtCompany Q Scr. ember 21, 1977

                                                                                                                                                                   ^

you, and the results achieved; (2) corrective steps which will be taken I to avoid further noncompliance; and (3) the date when full co'mpliance I will be achieved. You are not required to submit a written statement or explanation for the item of noncompliance for which corrective actions to prevent recurrence were completed prior to the conclusion of this inspection. In accordance with Section 2.790 of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations, a copy of this letter and the enclosed inspection report will be placed in the NRC's Public Document Room. If the report contains any information that you believe to be proprietary, it is necessary that you submit a written application to this office, within 20 days of the date of this letter,' requesting that such information be withheld from public disclosure. The application must include a full statement of the reasons why it is claimed that the information is proprietary. The application should be prepared so that any proprietary information identified is contained in an enclosure to the application, since the application without the enclosure will also be placed in the Public Document Room. If we do not hear from you in this regard within the specified period, the report will be placed in the Public Document Room. .,

Should you have any questions concerning this inspection, we will be pleased to discuss them with you.

                                                                                                            ' Sincerely,         ,
                                                                                                                      . s
                                                                                              -           W.C.SeidN, Chief Reactor Construction and                     *
                                                 ,                                                                Engineering Support Section

Enclosure:

IE Inspection Report No. 50-382/77-08 i t G e ens -* ' I i _ _ , - . ., - _ . . _ . . - . _ . . . . _ . . .. . . - - -_ _. _--- --_ 3. _ .

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50-382/77-08 O  :> . Appendix A NOTICE OF VIOLATIO?! Based on the results of the NRC inspection conducted on August 30 through September 2,1977, it appears that certain of your activities were not conducted in full compliance with conditions of your NRC Construction Permit No. CPPR-103 as indicated below:

1. Failure to Prov'ide Adecuate Procedures for the Control of Soecial Processes

            ~

10 CFR Part 50, Appendix B, Criterion IX requires that measures shall be established to assure special processes including welding are controlled and accomplished in accordance with applicable codes. The LP&L QA manual, requirement QR 9.0, paragraphs 9.1 and 9.2, requires that suppliers perform special processes in accordance with applicable codes and use written process control procedures specifying conditions to be maintained during the steps of the process. . Contrary to the above: The welding procedure specifications contained in Construction - Fracedure CP-203, " Welding Carbon Stael Structures," as Exhibits - F&M AWS-1, 2 and 3 and in use by the Fischbach and Moore welders 7 on August 31,1977, to control welding. conditions for seismic l support structures did not prescribe preheat requirements ~as ! established by the AWS Structural Welding Code, Dl.1,1975 edition. This is an infraction. - l The WPSs in use by the welders were corrected to include preheat requirements. A review of weld. records by F&M confirmed that no welds had been r:ade requiring preheat. Additionally, to avoid recurrence, Construction Procedure C.P-203, " Welding Carbon Steel

 ~-                                   Structures," including. all WPSs for use at the site were reviewed and changed to incorporate appropriate preheat requirements in conformance with AWS Dl.1 prior to completion of the inspection.

I

2. Failure to Follow Procedures for the Control of Soecial Processes 10 CFR Part 50, Appendix B, Criterion IX requires that measures shall be established to assure special processes including welding.are controlled.

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     , ,                   Th2 LP&L QA manual, requirem:nt QR 9.0, paragraphs 9.1 and 9.2, requires that suppliers perform special processes in accordance with written process control procedures specifying processing details and inspection and test requirements.                                       ,          ,

i Contrary to the above: . l On September 1,1977, the three following examples of failure to  ! control special processes were identified by the inspector:

a. The welds of C-245 seismic support were observed by the inspector to be welded on three sides rather than all around as prescribed by drawing "RAB Seismic Supports," G-695 S08, Rev. 4. The welds discussed had been inspected and accepted by F&M in accordance with QCI-10lW3, paragraph 5.2.3, " Quality Control Instruction for Weld Inspection," which requires that welds be placed as specified by the drawings.

b. The base metal adjacent to the accepted fillet welds in C-245 seismic support was undercut by excessive grinding greater than the allowable limit of 1/32" as prescribed by the AWS Structural Welding Code, Dl.1,1975 edition, paragraph 3.6.4, in accordance with the requirements of QCI-10lW3, paragraph 5.2.3.

c. Weld electrode control was not as prescribed by CP-205, " Construction '

Procedure for Control of Welding Filler Material," as demonstrated by u'ncontrolled welding electrodes observed in the work area of a tack welded support structure and neld electrode distribution records for September 1 not reflecting actual number of rods distributed and collected. This is an infraction with multiple examples'. s

3. Failure to Provide Adequate Tagging of fMnconforming Item ~

10 CFR Part 50, .t.ppendix B, Criterion V requires that " activities affecting quality shall be prescribed by documented instructions, procedures, or drawings of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, s, procedures, or drawings." Ebasco procedure ASP 111-14, paragraph 6.2.5.6.2, requires that status identifying tags shall be securely affixed to items. Contrary to the above: On August 31, 1977, an Ebasco QC hold tag was observed to be detached from.a section of nonconforming containment spray piping to which it previously had been affixed. - This is an infraction.

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  - -                                                         U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT REGION IV Report No.. 50-382/77-08 Docket No. 50-382-                                                               Category A2 Licensee:                  Louisiana Power & Light Company 142 Delarande Street New Orleans, Louisiana 70174 Facility Name:                     Waterford Steam Electric Station, Unit No. 3 Inspection at: Waterford Site, Taft, Louisiana                                 .

Inspection conducted: August 30 - September 2,1977 Inspectors: j' . /  ? /.[17 W. G. Hubacex Reactor Inspector, Projects Section Date

                                             -Paragraphs         1,)2, 3, 4, 5, 13 & 14                                                                     -

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                                              .C'A. Hermann, Reactor Inspector, Engineering Support 64, Date

• Section, Paragraphs 6, 7, 8 & 9

                                                                   /f l          'b W L. D. GTlbert, Reactor Inspector, Engineering Support 9/.;2/ /77 Date    /

Section, Paragraphs 6, 7, 8 & 9 i

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1 i '/ O h . A .A.- 'l- 2. I n/I J. I. Tapia, Reactor Inspectordntern, Engineering Date ' Support Section',far'agraphs 10, 11 & 12 Approved: ' - 40!A. - o!I/ 7'7 Date' W. A. Crossmyn, Chief, Projects Section 9 / 77 f' R. Hal.l E. . Chief, Engineering Support Section Date l ']"v

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Inspection Summary: Inspection on Aucust 30 - Seotember 2,1977 (Report No. 50-382/77-08) Areas Inspected: Routine, unannounced inspection of construction activities including the QA program of the piping contractor; containment erection and welding; welding of safety related structures and supports; backfill activ-ities; review of 10 CFR 50.55(e) items; and review of previously identified enforcement items and unresolved items. The inspection involved one hundred and eight inspector-hours by four NRC inspectors. Results: Of the six areas inspected, no items of noncompliance or deviations were found in four areas; three apparent items of noncompliance were found in two areas (infraction - failure to provide adequate procedures for the " control of special processes - paragraph 9; infraction - failure to follow procedures for the control of special processes - paragraph 9; infraction - failure to provide adequate tagging of nonconforming item - paragraph 5). e l e 1 e e

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Q Q . DETAILS

1. Persons Contacted -

Principal Licensee Ecoloyees

                      *A. E. Henderson, Jr., QA Manager
                      *T. F. Gerrets, Project QA Engineer
                      *P. V. Prasankumar, Site Construction Coordinating Engineer
                      *B. P. Brcwn, QA Engineer
                      *J. Woods, QA Engineer
                      *C. J. Chatelain, QA Engineer                     .
                      *B. M. Toups, QA Technician                                      -

Other Personnel

                      *J. O. Booth, Project Superintendent, Ebasco
                      *B. D. Fowler, Senior Resident Engineer, Ebasco                     -
                      *R. J. Milhiser, Assistant to Project Superintendent, Ebasco
                      *R. A. Hartnett, QA Site Supervisor
                      *F. D. Rose, Senior QC Supervisor, Ebasco
                      *T. Getzlaff, Lead QC Engineer, Material Control, Ebasco
                      *E. Breedlove, Project Site Manager, Tompkins-Beckwith (T-B)
                                                                  ~
                      *R. Hollenbeck, QA Supervisor, T-B                                            .
                      *G. Baker, QC Supervisor, Waldinger
                      *C. Boswell, QA Manager, Waldinger
                      *P. Smith, QA Officer, Ar.erican Bridge
                      *T. Kendrick, Site Welding & QA Supervisor, CB&I
                      *R. M. Ronquillo, QA Manager, Gulf Engineering
                      *L. Meerman, Project Manager, Fischbach & Moore (F&M)       -

G. L. Roshy, QC Project Manager, F&M

                      *E. Taylor, Chief Engineer, F&M E. White, Welding QC Engineer, F&M
                      *E. J. Wermann, Sr., Electrical QC Engineer, F&M                    -
                      *V. Farnsworth, QC Inspector, QA Corporation
                      *L. R. Wilson, QA Manager, J. A. Jones
                      *T. N. McAllister, QA Supervisor, J. A. Jones
'...                  The inspectors also talked with and interviewed other licensee and contractor personnel including members of the engineering and QA/QC staffs.

• denotes those attending exit interview.

s l l y L _ , _

2. Licensee Action on Previous Inspection Findinas (Closed) Unresolved Item (50-382/77-05-1) QA Program Inadequacies.

The inspector observed that the revised American Bridge Construction Department QA manual addresses previous inspector identified concerns regarding qualification of auditors, implementation of document control and control of welding. (Closed)UnresolvedItem(50-382/77-06-2) Personnel Access Control. Ebasco has issued Issue M of procedure ASP-lV-10, " Material Receiving, Warehousing and Control," which describes measures for controlling personnel access to storage areas. The inspector observed that measures described in ASP-lV-10 have been implemented.

3. Site Tour The inspectors walked through various are,as of the site to observe construction activities in progress and to inspect housekeeping and equipment storage.

No items of noncompliance or deviations were identified.

4. QA Manual Review .

With the issuance of the revised Tompkins-Beckwith (T-B) QA manual,

  ,             the inspector resumed review of the manual to ascertain whether activities asscciated with safety related piping are contrailed and performed in accordance with NRC requirements and PSAR commitments.

Documents which were reviewed included: the T-B QA manual, ' thirteen implementing procedures, and records of a site facility audit of T-B performed by Ebasco QA on August 12, 15 & 16, 1977. The inspector observed that some implementing procedures for audits and special processes were.still under development at the time of the inspection. Unresolved item 77-06-1, "QA Program Inadequacies," will remain open pending issuance of the implementing procedures and subsequent review by IE.

5. Inspection of Storace Areas During inspection of the Ebasco controlled outdoor storage area, the inspector observed that a QC hold tag had become detached from a section of nonconforming ten inch stainless steel containment spray piping and was lying on dunnage near the pipe. The tag, dated 10/6/76, MRIR No. 76-8970, apparently had been attached to the piping with tape that had become so severely weathered that its adhesive properties were 'los t. Another section of containment spray pipe in the same area x

l

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Q Q was found to'have a tag affixed to it with badly wrathered tape. The inspector informed the licensee that. failure to securely apply the tags was contrary to the requirements of Critorion V of Appendix B to 10 CFR 50 and Ebasco procedure ASP-III-14, paragraph 6.2.5.6.2, which states that status identifying tags shall be securely affixed to items.

6. HVAC Seismic Supcort Structures

a. Review of Procedures ,

                                                                            ~

The inspector reviewed the work and QA/QC procedures related to the on-site fabrication and welding of seismic support structures for safety related heating, air conditioning and ventilation systems. In particular, the following procedures relating to the Waldinger welding program were reviewed for conformance with the AWS Structural Code, Dl.l. ASME and appropriate QA require-ments: QA/QC Activities The Walding Corporation Quality Assurance Manual, Rev. O, draft 3, controlled copy 1, dated January 3,1977 Paragraph 8.0 - Identification and Control of Materials . Paragraph 9.0 - Special Processes Paragraph 10.0 - Inspection Paragraph 10.1 - Welding Inspection -

                       " Structural Support Hanger Fabrication and Installation Inspection,"

FQCP 10.2-3, Rev. 3 ,

                       " Fabrication - Weld Repair Field Quality Control," FQCP 2.2-11
                             \.

Work Activities ,.., " Installation - Structural Support Han'gers," Field Work Procedure l - Manual, P.4P 7.2-1 ( " Fabrication - Structural Support Hangers," Field Work Procedure Manual, FWP 2.2-12 AWS Prequalified Weld Procedure Specifications PQW-1 .SMAW 6011 electrode PQW-2 SMAW 7018 electrode i l e

        .-                                                                             r}

b Q , PQW-3 GMAW CO2 gas PQW-4 GMAW 75 Ar/25 CO2 , i

                 ~

b. Observation of Work Seismic support hangers, F271 and F161, were inspected for con-formance with requirements of drawing LOU-5817 G-922 501, Rev. 2, and the AWS 01.1 Structural Welding Ccde. In addition, the shop work activities and QA/QC activities for special processes were inspected for implementation of the Waldinger program.

No items of noncompliance or deviations were identified.

7. Imolementing Procedures - Safety Related Structures (Welding)

During this inspection, the inspector reviewed the "USS - Am. Bridge - Constnuction Department QA manual for Nuclear Power Station Structural Erection," QAM-III, controlled copy 14, Rev. 3, which had been approved for use. The revised QA manual and procedures contained provisions to

address the questions previously raised concerning qualifications of auditors, implementation of document control, and control of welding.

This item is considered resolved. ,

8. Fabrication and Erection of the Containment Vessel and Suocorts The inspector reviewed the work activities for the installation, erection and welding of selected samples of the top head first course and the crane girder. The inspector examined selected components, weldments, NDE reports, drawings and dimensional records as follows:

(1) Weldrent - Top Head Seam TH~1-1-H welded per procedure E7018, Rev. 2 Record Drawing - R82 - Rev. 1 Welder Qualifications - 2 welders per ASME B&PV Code, Section IX

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l Weld Filler Material - 7018 electrode l r Radiography per RTP-71-2426 98, Rev. 717 films - increments 0-18

                   -(2) Cocoonents i
                         . Crane Girder      Assy 67-A Stiffener - Assy 67-B Installation - temporary records for location, nonconformance reports regarding elevation (NCR-75) and location of center l                                             of rail (NCR-69)
                                                                 -6                         .

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                                                                      <3 Receiving Report - GC681 and material heat sheet
                                                    \

Testing and NDE - MT-Repdt 41 CB&I Detail Check List DCL-16 Inspection - per drawing 71-2426-65 No items of noncompliance or deviations were identified.

9. Fabrication, Installation and Weldina of Electrical Seismic Supports

a. Review of Procedures The inspector reviewed the construction and QA/QC procedures controlling the on-site fabrication and welding of electrical seismic support structures for safety related applications.

The Fischbach and Moore welding program was. selectively re-viewed for conformance with the Structural Welding Code AWS Dl.1 and applicable QA requirements as follows: (1) Construction Procedures CP-203, " Welding Carbon Steel Structures" CP-206, " Control of Welding Filler Miterial"

  ,                      (2) QA/QC Procedures Quality Assurance Manual, Section 9, " Control of Special Processes" Quality Control Instruction, QCI-10lW3, " Weld Inspection" During the review of the construction procedure CP-203, the in-spector noted that the procedure did not specify the welding preheat requirerrents as p'rescribed by the AWS Structural Welding Code Dl.1,1975 edition. The construction procedure CP-203, in paragraph 7 " Exhibits," referenced a nonexisting procedure for the preheat temperature. The welding procedure specification (WPS), listed as " Exhibits" in paragraph 7 of CP-203, was in use
 ..,'                    by the welders. The above stated procedure is inadequate to provide control of special processes as required by 10 CFR 50, Appendix B, Criterion IX and is considered an item of noncompli-ance.

The inspector interviewed the Fischbach and Moore QA manager con-cerning the noncompliance. The QA manager identified the seismic supports that had been welded and attested to the fact that no welding had be'en accomplished on material thicker than 3/4 inch. It was noted that preheating would not have been required prior to welding to meet code requirements for material 3/4 inch and less for the ambient temperature at the site._ e N R

                                                          -=               2.          .     -   _.

Prior to completing the inspection, the licensee provided the

                         ,     inspector with a revision to CP-203 which included preheat con-ditions in conformance with AWS Dl.l. Furthermore, the inspector reviewed the changes to the WPSs which included appropriate pre-heat rcquirements. These activities were considered suitable to achieve compliance and prevent recurrence.

This item is closed.

b. Observation of Work The inspector observed the on-site shop and field welding and inspection being performed by Fischbach and Moore for the fabri-cation and installation of electrical seismic supports for safety related applications. A list of completed seismic supports was obtained from the F&M QA Manager for selection of a random in-spection sample. The inspector selected support C 245 identified on Sheet 8 of Drawing G-605 FG8, Rev. 4, "RAB Seismic Supports."

The inspector verified that the welder and . inspector had marked the .:upport as being complete and acceptable. The fillet size, workmanship and location of weld were checked for conformance to the drawing. Two deficiencies were found that degraded the strength of 'the ' , support. They were: . (1) A fillet weld which was reouired to be all the way around by the drawing listed above was omitted on one side. (2) A steel plate of the support structure was undercut approxi-mately 3/15 inch adjacent to the toe of a fillet weld whereas the Structural Welding Code AWS Dl.1-75 permits a maximum of 1/32 inch for an undercut condition. While inspecting a tack welded support structure, the inspector found. E7018 welding rods (one new and one half used) laying next to the tack welded structure after the welders had left the con-struction site. During an interview with the weld rod room atten-t dant, the inspector learned that the electrode accountability ( system specified in Fa!! procedure CP-206 was not being followed since'the attendant was not counting the weld rods and stubs being i turned in by the welders to determine if all used and new weld l electrodes were accounted for. The inspector observed that the control of welding electrodes was not being performed by F&M personnel as required by F&M procedure CP-206.

      -                           The licensee was informed that they . were in noncompliance with Criterion IX of Appendix B to 10 CFR Part 50 in that the activities described above were not performed as stipulated by the procedures for controlling welding.

i l \ W

m A - ' Q & During the inspeqtion of the above fillet welds, it was noted tnat the zinc-rich paint applied to the welds as a protective coating in accordance with Fischbach and Moore procedure CP-203, Rev.'2,. contained cracks. The inspector reviewed construction procedure CP-203 and quality control instruction QCI-10lW3 to determine the requirements defining an acceptable painted surface and could not ascertain well defined acceptance criteria. The inspector discussed the matter with the licensee's Quality Assurance Technician and the contractor's Project QC Manager and was informed that painting was inspected during the final inspection of the installed supports. The inspector expressed concern to the licensee regarding the definition of quality requirements for the zinc coating. The licensee ccamitted to redefine the quality requirements for the coating and review the components already painted to insure the coatings were not cracked. _ This item is considered unresolved and will be reviewed during subsequent inspections.

10. Significant Construction Deficiencies Reported by the Licensee The inspector reviewed licensee action " elated to the following items which were previously reported as signi'icant or potentially significant construction deficiencies in accordance with the requirements of 10 CFR 50.S5(e). -

a. Corron Foundation Mat Cracks After an unsuccessful attempt at pressure injection of Concressive 1380 epoxy into hairline cracks caused by mat flexure, a more effective procedure was initiated to control the leakage of water through the cracks. This procedure consisted of chipping a one l inch deep trench along the length of the crack, roughening and cleaning of the surface as well as a one foot strip on either side of the crack, and filling of the trench with SIKADlJR Hi-Mod-LV epoxy. All repairs beneath placecent 502-6 were monitored for one day and no indication of water leakage was observed. The in-spector viewed the results of further sealing operations performed in anticipation of future concrete fill placements which should, when placed, reverse the flexure and minimize the cracks. This matter is considered closed.

b. Excessive Air Entrainment Additional borings in wall placements G-571-501-5B and -8B have I'dentified the area where concrete compressive strength is less than the design strength of 4000 pounds per square inch as' an area from one to four feet below the top of wall SB and up to and includ-ing thirteen feet frem the extreme east end of this placement. The total area involved is therefore approximately fifty-two square feet out of a total wall area of eight hundred and eighty-two square feet.

1 The wall is three feet thick. l -

y - . - - - . a.= : :=_ Ov 9_i Subsequent inspection and testing of the admixture dispensing system was performed by batch plant personnel and it was con- ' cluded that a three-way dispensing valve had malftn'ctioned and

'                                           allowed an excess of Air Entraining Agent to be batched. In-spection of the dispensing system is now b~eing conducted on a daily basis and calibration' will be performed at ninety day intervals. This mat.ter remains open pending suasequent review of licensee's evaluation ay.d disposition of the' deficiency.

c. Engineered Backfill Damages \ .

N s . r s A review was ' conducted of'the damage and repairs to backfill in the Northeast quadrant of the excavation caused by 'the rupture of an eight inch construction water main. The, rupture caused i sand boils to occur which resulted in areas oft less than seventy-five percent relative density. A series' of twenty-two' soil borings' gave standard penetration < rates' which verified the degree l and extent of the damage. Redsnsification of the damaged area was obtained through the use of a vibrating proberwhich consisted of a vibratory pile hamer mounted on . top of a thirty feet long-

                                           , section of thirty inch diatteter' pipe. Probing was
                                           ' five feet centers with.a minimum vibrating time                                                        fiveof

minutesconducted on. t at each location. Confirmatory  : oil borings indicated an ic: grove--

                                          'mentiof the density to at least sey'enty-fi' e percent relative                                                                            .

density over the entire area. The inspector reviewed the Eustis' Engineeripig' Ccepany; Soil Borings Report. and' compared the standard penetration test results- at three/ ocationf.

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l

• This matter remains cpen/ phnding a review of formal documentatio'n. '

                                                      ;<     jj                             -

Crane Damaged Wall .- '-

d. .,

t , l

                                                                                                                                )           ,

l l The repairs 'to the wall which was damage.f by 'tha collapse of a crane boom have oeen inadverter.tly affected ty, a potential . deficiency in the cement chen:ical analysis. (See paragraph 11 2 below) The average twenty-eightjday ccecrassive< strength was ' threesthousand five h'undred and sixty. pounds per. square inch.

                               '               A spare cylirider is being held for a ninety .cty- break. Thisi matter rema, ins open pending a verification of the actual i                                                                                 ,

strength and its .elatioriship to the cement corposition. i

                                                           <                                                      .                                          ~

11. Cement Cherfcal Analysis - ( , c/ ( ' A continuir:Mdegradation of concrete ccitpressive strength was. identified by the on-site testing laboratory. Sr.atisticeFanalysis indicated .a , coefficient of variation of nine and ninety-seven hundredths percent as of' July 29, 1977: The maximum allowa m isiten percent. A, s top-work order affecting the use cf ^three concrete mix' designs in Cless Estructures.. was issued *af ter tRe labonitory identified (a discrepancy with' the isuppliers chemical analysis. Samplas of. the cement were sent to the Portlarid Cement Association in order,to vdrify trie quality of the cement beingr med in all i ,j design mixes. - 1 1 l Jl

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                                         .. . - _ ,- .l , a     L,._,,                                                 ,3                     _

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_ _ . - . _ _ _ _ _. m _._. _ . . .__ O 3;-

 .     .        k2. Containment Nill Concrete Grouting The inspector reviewed Nonconformance Report W3-553 which concerned
                      'itself with approximately seventy-five feet of nylon rope and approxi-mately one hundred and thirty feet of PVC grout tubes that were left embedded in placement number 502-6A.             A mock up test block was con-structed to illustrate the epoxy grouting operation with the PVC U -

tubes and rope in place. Results of the test indicated that the epoxy would encase the rope and PVC and also gain access to any separation between the concrete and the containment vessel when the grout was pumped at five pounds per square inch pressure. The in-spector witnessed the chipping out of five, three feet deep test probes in the actual placement. No gaps between the concrete and vessel were observed. Two hundred and thirty-eight gallons of epoxy grout were used in the thirteen feet radius placement. No items of noncompliance or deviations were identified.

13. Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, items of noncompliance, or deviations. The following unresolved item was dis-closed during this inspection regarding the painting of electrical seismic supports: -

Identifier Title Reference 77-08-1 Painting of Seismic Supports Paragraph 9

14. Exit Interview i The inspectors met with licensee representatives (denoted'in paragraph l 1) at the conclusion of the inspection on September 2,1977. The j inspectors summarized the purpose and the scope of the inspection find-ings. A licensee representative acknowledged the statements of the inspectors 'concerning th.e items of noncompliance (paragraphs 5 and 9) and the unresolved item (paragraph 9).

l . 1 - 1 l l t' ( l

                                                                                                                                                                                                   -- ~
                   'MWW                                             &                                EATELINE                                                                                           L Not To Worry                                                                ===>-
        ~

Nuclear Regulatory . Commission will - Water scepage through the founda- Waterford's design calls for the plant to By RON RIDEN!IOUR _ agree, tion was not anticipated by Waterford's be a sort of steel and concrete boat floating im Fort, Louisiana Power nat remains to be seen. designers, raising questions about the ade- on the water-impregnated sands beneath nereis,in fact,a swa dealdintonna- quacy M

• wiginal
• sign and * .and surmunding it. In order to get their and Light Company's ilon available, including comments made engineering assumptions k is basedon. boat in the water, however, Waterford's L public spokesman, told by LP&L officials on the day before the - New cracks in the foundation, engineers had to lnstall a complicated water

  ..      GWHhit last'wcek that; LP&L                      newest cracks were discovered, which sug-           discovered on May 11 suggest that the            control and pumping system designed to
  '                                                        gests that the cracks in Waterford's foun-          design and engineering problems responbl-        keep the excavation dog for the common oTicials ' met on Monday, May                                                                        ble for the first cracks are yet to be resolv-   foundation mat dry while the huge slab was dation raise fundamental questions about j    . 23, and agreed to take a clear-                  % i,,,,,i,y g % pi,,,,s design and the              ed, raising fu ther questions about LPAL's       put together and the three structures k sup j    eyed hard hne about the weep-                  7ffect it may have on Waterford's promise             assumptions about their abillfy to               ports went up.

Ing, hairline cracks that have beek appear- to operate safely. guarantee Waterford's safe operation.

• According to Waterford's design ilEory l .

1' ing in the foundation of the Waterford til - Although the water seeping through Waterford's " floating" design is unique the ground water at the plant site would be I ^ l; nuclear power plant since 1977i Don't the cracks in Waterford's foundation is, among U.S. nuclear reactors. Emuislana's allowed to flow back into the area beneath

; worry about the cracks Fort said. They

currently a minimal amount, the plant's peculiar geology, particularly the mushy 'and around the nuclear Island on a con-

don't mean anything. design requires the foundation to be water- " jelly ground" nature of the soils nearest trolled basis as the plant went up, a process "nese hairline cracks... are absolutely tight throughout its expected 3(M0 year the river, induced Waterford's designers to caBed re-charging. Under the rechairsing insignificant to the structural integrity of plan a foundation for the plant that includ- theory, the flow of groundwater would be

l. ! operating life.

                                                                - LP&L internal dm .., . 5 written in .         ed no pilings. Hree of Waterford's four         controlled in such a way that the water sur-i thIt plant," Fort said. "We met Monday.

several of us did, on this particular thing. 1977 aner. cracks in the foundation directly ma}or buildings, all of those dealing with roun Eng the nuclear Island would be in a

      '    Of course, as everybody who's knowledge-          below Waterford's nuclear reactor were             the handling of nuclear materials, were          state of equilibrium vis a vis the weight of

! roble about it said, it's just absolutely first discovered called them a "significant bulk on what is called a " common founda- the nuclear Island, creating a buoyant ef-nothing to worry about and we're not 30- deviation from performance specifica- tion mat", an enormous steel-reinforced feet that would actually " float" the island i i ing to worry about it." Fort also said that g tions" which could affect the " safe opera- concrete slab measuring 270 feet in width, on the hydrostatic sand around k. 380 feet in length and 12 feet in Ihickness. In the spring of 1977, however, not long {' LP&L officials are sure that the U.S. l' tion of the plant." l r i', i I . 89 cc cx3 ^ w i wgo y 8 eg c3 !, ~ p I M*f x;y b c y \ \ IC a N ii D > \ ; N

                                                                                                                                                                      ~
                                                                                                                                                                                  ~~                       O M                        $

E

l >J k . . . - - ., ,.

s s R. - t Despite this. Wuctford's builders never partM given amount' because h comes i addressed the underlying cause of the , , before the first cracks beneath the plant's ' in several varieties... At this time m the cracks. Instead they patched the cracks !! reactor were discovered at Waterford there was a problem with the water re from the top of the foundation mat to halt plant life you'd expect k to be pretty stable and I understand that's exacdy the case." .  ! the immediate water seepage through them j j charging system. Waterford's engineers On the day after this conversation occur-and proceeded with their construction t f were not able to control the rate of re- schedule, which required pouring huge red, May 11, 1983, EBASCO engineers wrote a new non-conformance report an-charging in the way their design and con-

quantities of concrete over the places where struction plan called for. According to a nouncing the discovery of new cracks in the l]

the cracks first appeared. Brian Grant, the  ;

civil engineer for Waterford's prime con- EBASCO civil engineer, and Tom Gerrets, , foundation of Waterford's nuclear island. , I tractor, the architect engineering firm of LP&L's Quality Assurance Manager for According to nonconformance report

EBASCO Services

• Inc., there was a las in Watafad,& fen &dthissolutiminaMay El2,uponedinlanguage miming that I

l' the called for state of equilibrium between used back in 1977 to report the first cracks ' 10,1983, interview with Gambit, insisting j the weight of the plant and that of the water discovered, water was discovered to be per-4 ' that the cracks were indies.tive of no struc- , i in the soils surroundmg it. 'that lag, accor- tural problems with Waterford's founda. colatins up through new eracka discovered I ding to Brian Grant

• the EBASCO civil in &e Su of the reactw auxihary '

i tim. W mly rease hre was any pro-l enginsier, caused the common foundation blem at all, they said, was the possibility buildings, which is in turn the top of the mat to flex in an unanticipated way sat se ng wata might affect h N f 6 m et.

f. creating a condition EBASCO's enginee s structuralintegrityof theconcreteschedul. While LP&L officials and spokesmen i: ' have since analyzed as "nnss revasalr,

• ed m be pound om the cracks.Since pat- connue2minimizehimponanceofme Stress reversal, a reflection of motion in th cracks and the water seeping through them, ,

ching the crack surfaces halted the water  : foundation, created cracia in the founda- flow through them, at least long enough they also say they are considering seeking tion that ran all the way through k, which fw me new cacrem 2 be poured and sete NRC approval to change their Final Safety

l' in turn allowed water to come through.

Waterford's boat, in other words, was then the problem was adequately resolved. i Analysis Repon (FSAR) in a way which

                                                                                                                     . Aslongas therewas no more movement                      ' doesn't require Waterford to be watertight.

I springing leaks. m Watafwd s fundadon, Grant and - Ihe FSAR is LP&L's basic blueprint for l 4 Despite the sanguine attitude towards Gerrets said, then there was nothms to the construction of Waterford, a sort of i i the problem currently being displayed by wary abat. The cracks discovered in 1977 contract with the NRC which outlines how i LP&L officials, the discovery of those were indicat,ons i of a kind of movement Waterford is Imilt and guarantees its l 3 cracks in 1977 caused a great deal of con- &at was not expected, both men agreed, agegy, , cern then. In a report written on July 29 but hy &nied that they wem an indication i 1977, an unidentified EBASCO official of any swims stmetural problems wkh wrote that: "The 12 foot thick common stafwd's fundade. W stras ww-c foundation mat is considered to be thick mlproblems satcaused emecracks,by

enough to not require waterproofing to indicated, had been arrested.

I prevent leakage, therefore, this defect is 'l think that at this stage in the plant , I considered to have possibly adversely af- life," Grant said, "further settlement is not j fected the safe' operation of the plant and is [eally anticipated. I don t beh, eve it's mov-ii considered a significant deviation from ms. My understandirgt of the records that j' performance specifications which will re- were, that are being kept is that it (the J . quire extensive repairs to establish the ade- ] 'quacy of the structure." A few days later f"8*p island) has essentially stopped setti- ! EBASCO/LP&L reported the problem to wn a to my, W, sat any i the Nuclear Regulatory Commission as a desoHunhasetensinthemelear ! significant construction deficiency one of n ma tu. "I think , 1 the agency's most serious categories of con- the fact of significant settlement at this ! struction failures in a nuclear power plant- point would perhaps be the kind of thing ! .i that would be a problem, rather than any - . 1

                                            --          ~ - -               ~,..                                     . . ,

H*' W  % .m . 4

o - , 00CKETED USH9C

                                         ~ UNITED STATES OF AMERICA                                        -

NUCLEAR REGULATORY COMMISSION OFFlcE OF SECRETAiW Before the Atomic Safety and Licensing Appea109fddgyEgytr.l. In the Matter of )

                                                              )

LOUISIANA POWER & LIGHT COMPANY ) Docket No. 50-382

                                                              )

(Waterford Steam Electric Station, ) Unit 3) ) CERTIFICATE OF SERVICE This is to certify that copies of the foregoing

               " Applicant's Answer to Joint Intervenors' Motion to Reopen Contention" were served by deposit in the U.S. Mail, first class, postage prepaid, to all those on the attached Service List this 30th day of September, 1983.

s AW l

                                                       ' W W. Churchill,7.C .

Dated: September 30, 1983 i l i

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION , Before the Atomic Safety and Licensing Appeal Board In the Matter of. )

                                                                                                                 )

LOUISIANA POWER & LIGHT COMPANY ) Docket No. 50-382

                                                                                                                 )

(Waterford Steam Electric ) i Station, Unit 3) ) l SERVICE LIST i Christine N. Kohl Sheldon J. Wolfe Administrative Judge Administrative Judge Chairman, Atomic Safety and Chairman, Atomic Safety-and f Licensing Appeal Board Licensing Board + U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Washington, D.C. 20555 W. Reed Johnson Harry Foreman Administrative Judge Administrative Judge Atomic Safety and Licensing Atomic Safety and Licensing Appeal Board Board U.S. Nuclear Regulatory Commission Director, Center for Population Washington, D.C. 20555 Studies Box 395, Mayo Howard A. Wilber University of Minnesota r Administrative Judge Minneapolis, MN 55455 Atomic Safety and Licensing , , Appeal Board Walter H. Jordan U.S.-Nuclear Regulatory Commission Administrative Judge Washington, D.C. 20555 Atomic Safety and Licensing Board Sherwin E. Turk, Esquire 881 West Outer Drive Office of the Executive Oak Ridge, TN 37830 Legal Director U.S. Nuclear Regulatory Commission Docketing & Service Section (3) Washington, D.C. 20555 Office of the Secretary i

                                                                     - _                                         U.S. Nuclear Regulatory Commission Atomic Safety and Licensing                                                                          Washington, D.C. 20555 Appeal Board Panel U.S. Nuclear' Regulatory Commission                                                                  Atomic Safety and Licensing Washington, D.C. .20555                                                                                  Board Panel U.S. Nuclear Regulatory Commission Washington, D.C. 20555

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LP&L Service List-ASLAB Page Two

                                                                                         ~

Mr. Gary Groesch Luke B. Fontana, Esquire 2257 Bayou Road 824 Esplanade Avenue New Orleans, LA 70119 New Orleans, LA 70116 Brian Cassidy, Esquire Spence W. Perry, Esquire Federal Emergency Management Federal Emergency Management Agency Agency Region I Office of the General Counsel 422 J. W. McCormack 500 C Street, S.W., Room 840 Boston, MA 03109 Washington, D.C. 20472 Carole H. Burstein, Esquire 445 Walnut Street New Orleans,-LA 70118 i 4

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