ML20214Q273

Provides Results of Author 860903-05 & 09-12 Investigation of Whether Cracks in Base Mat Pose Safety or Operational Hazards.Base Mat Will Safely Carry Design Static & Seismic Loads & Exfiltration of Water to Ground Is Impossible

ML20214Q273
Person / Time
Site:	Fermi
Issue date:	09/11/1986
From:	Philleo R PHILLEO, R.E.
To:	Harrison J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
Shared Package
ML20214Q257	List: ML20214Q254 ML20214Q273
References
NUDOCS 8609240244
Download: ML20214Q273 (13)

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Ten.mmons poos ese.asse ptoBERT E. PHIL. LEO. P.E.

oowouiswee me einssa 74M3A.C^2M M OOURT ANNANDALa. WNDSDNA $$ooS September 11, 1986 l

Mr. J. J. Barrison, Chief Engineering Branch, DRS Begico III Buclear Regulatory Commission 799 moosevelt moed Glen Ellyn, IL 60137 l

Dear W . Barrison:

9-12, 1986, at I spent the periods September 3-5 ard septa =h=e the Fermi Power Plant lavestigating whether the Ono.acks in the3.Unit September I act2 base sat pose safety or operational hasards.

with representatives of Detroit Edisce, Sargent and Lundy, Hopper and Associates, and the NHC. Thereafter, I made two inspections of the

,4 base amt, such of the top surfleoe of which is accessible, and reviewed l the extensive documentation which has accumulated relevant to situation. Net of the pertinent documents were available in the However, some were not. I requested and Detroit Edison files. I reviewed the received additional dooumonts firam Sargent and Lundy.

methodology for all the design and analysis calculations and checked many of the calculations in detail.

In order to evaluate the significance of the cracks, the following spaastions must be answered:

1. Do the cracks impair the structural performance of the base mats

a. under static loads?

b. under the original design seismic loads? ,

o. uname subsequently revised seismic loads?

d. amaar the Perry earthquake?

2. Does infiltration of ground mater:

a. pose safety or operational problems?

b. permit corroeico of the reinforcing steel?

8609240244 860919 r

PDR ADOCK 05000341 PDR l* P

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l W. J. J. Barrison, Chief l

September 11, 1986 page a i

3 Is enfiltration of contaminated unter from the plant into the ground unter possible? l It is evident that the cracks were not caused by structural '

loads. The static loads for which the mat is designed include the

, deed seight of the sat, the uplift force of the ground unter (tetich is 4

four times as great as the dead weight), The and sat the was weight of the placed during 'salls l

l and Juneacaponents and July,1971.of the reactor building. Reference to the cracks first appeared

' record on April 12, 1972. At that time, only about one-third of l either the uplift load or the building load had been applied to the structure. The cracks had to be the result of restrained volume changes in the concrete.

Cracks in the reinforced concrete any occurCracks in compression in compression annan, tension sones, or zones of high shear stress. Concrete must I

scoes are of no concern because they close under load.

crack in tensile sones in order for the steel toMoraally, achieve the anycracks significant portion of its tensile strength. But if form as a result of load and are numerous and very narrow.Thus, pre-existing cracks exist, they serve the same purpose.

pre-existing cracks in the tensile Their widthsone may do raise nota hinder question the about functioning steel of reinforced concrete. Cracks in zones of high shear stress are a potential corrosion. If such cracks, or the reinforcing steel passing source of concern.

through them, are unable to transmit the sheer stress, there will be a relative dispianmaavit of the concrete on the two sides of the crack. ~:

I could detect no much dispimaamant in checking the cracks with a  !

=traightedge. Thus, without making any calculations, it any be determined that the base mat is adequate for carrying its dealgn static loads.

Calculations support the visual observations. Since the cracks I

are randon and unrelated to critical stress locations, it would be a f esincidence if any occurred in critical flexure or shear sones. Calculations ar Calculations demonstrate that recew did. A finite element  ;

g feasible means for evaluating seismic response.

L analysis has demanatrated that the mat will successfully withstand  ;

j toth the original design earthquake and the subsequently developed 1

site-specific earthquake under the assumption that the oracks extend completely through the mat and that there is no shear transfer othe than by the shear friction phenomenon.

sich the Ferry earthquake uns applied to the Fermi structure, but a esaparison of the response spectra indicates that for the frequencies er interest, the displacements, velocities, and accelecations of the '

' purry earthquake are between 0 3 and 0.6 of those for the Fermi site-specific earthquake. ,

.' SEP 181986

4 Ik. J. J. Barrison, Oilef september 11, 1986 Page 3 t=*== in The oracks noted in 1972 were of the order of 0.03 width. This is about twice the width recommended by the The oracks American more grouted in Concrete 1972 with aInstitute coment-fly forash corrosion grout. control.Such oracks as have been d more recently are too narrow to be ==w===fu11y There treatedwere with46ooment such grout. They have been sealed with epoxy resin.The operation has been s orneks seated in 1984 and 1995 Almost all the unter present is The top of the mat is generally dry.from leaking mechanical syste I observed seven floor cracks which are seeping some been sealed. Thus, inflow of ground unter unter, but there is no measurable flow.

is an almost nonwixistent problem and one that is well within the capability of the project unter-handling facilities, The width i There is virtually no potential for steel corrosion.

of the oracks in the repaired condition is mell below AmericanAnd sinc nttions.

Concrete Institute as to manied or filled with unter, there is no opportunity for oxygen Without oxygen there can be no corrosion.

reach the steel.

Rufiltration of unter from the reactor building into the ground coald occur only if a tremendous quantity of unter poured into the building. Since the base mat is 35 feet below the unter table, over But the most severe 35 feet of unter would have to secumulate.

l pipe-break accident postulated in the F3AR puts only 8.6

in the torus room. The reactor building is

( beftre the maximum unter height is reached.

flood-proofed to a level more than a foot above the floodAlllevel doors and predicted for the probable maximum meteorlogical event. In the penetrations below this level are of untertight design. ld unitkely event that a door mere left open during a f the outside unter.

In order to evaluate the safety of the structure, it is not it is Bowever, necessary to pinpoint the exact cause of the cracks.

almost certainly the result of thermal cracking associated with the The annneste uns higi temperature rise of the very strong concrete.Although the use of fly ash much stronger than it needed to be.

ra h a rate of strength gain soseuhat, mimaat all the concrete reached its required 28 day strength in less than seven days, and the one 90-day test that uns made resulted in a strength over twice that repired.

IRifle a maximim temperature limit uns imposed ubich re p ired the use of ice, there una probably a temperature rise in f

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W . J. J. Barrison, Oilef

- September 11, 1986 Page 4 this anasive structure of over 100*F. In cooling 100' to ambient In carrying out such l temperature, concrete must inevitably crack.

i placements, ersaking can be minimized by following the advice in ACI 207.2R, ' Effects of Restraint, Volume Change, The search forand an Reinforcement on improved concrete, Cracking of Massive Concrete."

suggested in the correspondence, has not materialised because Detroit Edison has not undertaken any nuclear reactor construction since l Forni 2.

In summary, the base sat will safely carry the design static and seismio loads, infiltration of ground unter is very small and easily manageable, corrosion of reinforcing steel is essentially impossible, and enfiltration of unter to the ground is impossible either in a pipe-break accident or a flood.

Sincere 1?; yours, Robert E. Philleo, P.E.-

REP /jj

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,e W. S. NUCLEAR REGULATORY C99(IS$10N OFFICE OF INSPECTION AND EhTORCDEENT

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IE Inspection Report No. 050-341/75-01 Licensce: Detroit Edison Company 2000 Second Avenue Detroit, Michigan 48226 License No. CPPR-87 Enrico Fermi Unit 2 Category: A Monroe, Michigan

' Type of Licensee: BWR (GE) - 1150 MWe Type of Inspection: Routine, Unannounced F

Dates of Inspection: March 5-6, 1975 Dates of Previous Inspection: November 25-26, 1974 (Construction)

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Principal Inspector: T. E. Vandel ~

(Dat6)

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Accompanying Inspectors: C. E. Jones (Date) t 1

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E. W. K. Lee (Date)

Other Accompanying Personnel: None l A Reviewed By:

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ayes Ty A Se61or Reactor Inspector gh43 ,

Construction Projects 7 '(Date) 4 l

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2. Storsae and Identifiestion of Sprina Rangers and Supports (R0 Inspection Report No. 050-341/74-08)

• Aspreviously$ reported,pipebangersandsupportswereapparently not being properly stored and protected.. e .J ..,

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During this inspection, the inspector noted that several spring bangers and supports stored outdoors were badly rusted, and the identifying paint was fading. This matter is now considered (Report to be in noncompliance with NRC rules and regulations.

Details Section 1. Paragraph 2)

3. Reactor Building Wall and Floor Slab Crackins (R0 Inspection Report No. 050-341/74-08, and Inquiry Report No. 050-341/72-02)_

The inspectors observed the area of the floor and walls where cracking had been identified. In' addition, they reviewed the licensee's final report. EF-2-29-537, submitted pursuant to

' 10 CFR Part 50.55(e), and informed the licensee they had no I further questions in regard to this matter. (Report Details, Section 11. Paragraph 5)

Management Interview A. The following persons attended the management interview at the I close of the inspection.

1 Detroit Edison Company (Edison) l T. A. Alessi. Quality Assurance Division Director

!i A. Alexiou, Project Quality Assurance Director

~ T. C. Byrd, Field Quality Assurance Engineer -

W. E. Everett, Assistant Construction Supervisor J. Ferguson, Supervisor - Major Projects A. B. Barris, Project Manager L. F. A. Reed Field' Quality Assurance Engineer E. C. Sliper, Project Engineer

[ Matters ' discussed and comments, on the part of management personnel.

3 B.

• were as follows: ,

1. The inspector stated that it appears that the following

[ activities were not conducted in full compliance with RRC rules and regulations and Edison procedures, The equipment, maintenance, and record control (EM&RC) file j a.

was not maintained properly. Consequently, the RCIC system steam turbine periodic inspection was overdue. (Recort l

Deta,ils,Section I, Paragraph 1)

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-- Spray Pumps, identification No. E-21-01-C-0015 and No. E-23-01-C-001A.

The inspector verified that procedure WP-M-019 was being followed l to provide storage protection and to perform surveillance inspections.

4. Esposed Sutidina Structures he sospector stated his observations of the well, where construction terminated in the reactor building, (about 660' elevation) that ice had l gathered in the construction joint and appeared to have damaged the )

concrete. The licensee stated that the area would be cleaned of loose concrete and the reinforcing bars cleaned, prior to resumption of construction.

The licensee was also asked about imbedoents lef t in the equipment mounting pedestals. Ice was observed in the recess and in the sleever around the anchor bolts. The licensee stated these would be cleaned and inspected prior to resumption of construction.

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5. Beactor Building Walls and Floor Slab Cracking The Edison final report nucher ET2-29,537 issued Novecher 8,1974 was discussed with Licensee personnel. In response to questioning the licensee indicated that the grouting repairs are complete and that although some slight leakage (0.13 gal per minute maximum) may still y

i occur it is fully expected that any leakage will seal itself off in time.

• In addition, the inspectors were assured that any minimal leakage would be handled through the radweste system, and that, the containment

- building negative pressure inside and the ground water hydostatic pressert on the outside, would prevent any outlenkage.

B The inspector stated he had no further questions in regard to this ratter.

ri Buildings and Installed Equipment As-builts l i. 6. -

! Information was requested relative to as built plant conditions records and plans to develop such documents. The inspectors were i informed that a computer oriented Nuclear Power System (NPS) has been developed as a status system and that the accuracy of the '

information is being checked by actual plant survey and documentation.

The installation status of other mechanical equipnent, piping etc.,

is also being manually recorded on drawings or by lists.

The licensee added that a system type of bar charts are being developed and expanded that will provide a method for commencing construction again once the shutdown period is over.

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5 U.S. NUCIE.AR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT REGION III Report No. 50-341/79-04 Docket No. 50-341 License No. CPPR-87 Licensee: Detroit Edison Company 2000 Second Avenue Detroit, MI 48226 Facility Name: Enrico Fermi 2 Investigation At: Enrico Fermi 2 Investigation Conducted: February 15-March 2, 1979

> Investigators:

(SC.5 F. C. Hawkin f en tor Inspector

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[.d. N 1/17 l~71 f.J. Marsh, Investigator (date) j f.C. W'#

H.S.Phillips,RbactorInspector W

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R. M. Wescott, Reactor Inspector -/ /4 7 v

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4. f.7)C 7/Z7h7 Reviewed By: C. E. Morelius Assistant to the Director (date)

(f{ ( ** t R. C. Enop, Chief 7 / f i

j Projects Section 1 (date)

Investinative Susanary i

Investination on February 15 - March 2, 1979 (Report No. 50-341/79-04)

Areas Inspected: heenty allegations were made relative to management and construction practice. In many instances these allegations pertained to non-safety related equipment or work. This inspection involved 146

+ faspector-hours which includes 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> on site and 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> of investigation time meeting with several allegers at appointed meeting places.

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SilMMARY OF FACT On February 8 and 9, 1979, Mr. Frank Kuron was interviewed by Messrs.

Robert Marsh (MRC Investigator, Region III), and Shannon Phillips (NRC Reactor Inspector, Region III), regarding his earlier, statements before the Fermi 2 Prehearing' Conference.

Mr. Euron provided the NRC representatives with information on twelve areas which be considered as potential health and safety concerns regarding the construction and future operational capabilities of the Fermi 2 site.

Through both question and answer and in narrative statements it was disclosed that Mr. Euron's concerns covered broad areas and in several cases were dated or were of a nonspecific nature. Mr. Kuron indicated much of his information was second or third hand and/or founded solely on heresay. Mr. Euron agreed to review his own records and contact his

" sources" in an attempt to provide the NRC more definitive information.

At the close of the interview, twelve (12) potential areas of investigation were identified. Following a review of Mr. Euron's allegations, an investigation was initiated on February 15, 1979.

On February 20, 1979, the investigation was continued at the Fermi 2 site. On February 21, 1979, Mr. Kuron was brought on site and in a walking tour of the facility, further defined his allegations.

l The orginal list of twelve allegations / areas of concern resulting from the February 8 and 9,1979, interview of Mr. Euron was expanded to twenty (20) items and additional detail was aquired from Mr. Euron and other sources. The areas investigated and the conclusions reached are suanarized as follows:

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1. Lack of Quality Control - No evidence to support or substantiate this allegation was identified.

2. Destruction of two trailer loads of quality control records - In 1974, documents from two trailers that contained personal records and copies of working drawings, specifications, and milestone charts were burned. No permanent QA records were destroyed. The investigation was unable to substantiate this allegation.

3. A recent fire in Building 45A was more extensive than reported to the NRC - No evidence to support or substantiate this allegation was identified. Continued as open ites (341/79-04-02) pending licensee identification of records burned.

4. Interference fit of a 24" Globe Valve - No evidence to substantiate this allegation or concern regarding improper construction practice was identified.

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5. Poor housekeeping in the drywell area - The investigation was anable to substantiate that overall housekeeping (drywell area included) was unacceptable at the time of this investigation.

6. Improper installation of reflective shielding - The investigation substantiated that this shielding is nonsafety related and therefore it is not a safety concern.

7. Pipe hangers improperly installed - Allegation determined to be valid and previously identified by the NRC. Corrective action is continuing.

8. Reactor Feed Pump Turbine damaged in early fire not properly repaired -

Investigation disclosed concerned equipment to be QA Level II, nonsafety related. Allegation considered a nonsafety construction issue.

9. Nozzles in main condenser improperly welded - Insufficient detail availsble to identify specific piping involved. Investigation disclosed no safety related piping in area designated by alleger.

10. Improper storage of turbine parts - Determined to be nonsafety construction issue.

11. Inadequate posting of work areas as required by 10 CFR 21 (Paragraph 21.6) - Allegation not substantiated.

12. Improper welding of Main Steam Line spool piece - No evidence to support or substantiate this allegation was identified.

13. Use of improper weld rod - Allegation not substantiated. System

t involved determined to be nonsafety related.

14. Improper pipe whip restraint weld - Allegation was determined to be unsubstantiated.

15. Improper installation of concrete anchors (Red Neads) - The investigation was unable to substantiate this allegation. Continued as open ites (341/79-04-03) pending additional testing by licensee.

16. Voids in grout of sacrificial shield wall - Allegation substantiated.

Two void areas identified by investigation and licensee DDR 1187 found to have been inadequately completed (incomplete repair).

These items cited as items of noncompliance (341/79-04-04).

17. Improper cadweld sleeves in Reactor Building - No evidence found to support or substantiate this allegation.

18. Emirline cracks in Reactor Building structural steel - No evidence found to support or substantiate this allegation.

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19. Surplus structural steel from RER Building considered by alleger to represent construction "short cuts" - No evidence to support or substantiate this allegation was identified.

- 20. Cracks in the concrete of the base' slab of the Reactor Building - The investigation revealed that an early history of cracking had existed with the base slab but that this previously addressed matter ind been satisfactorily resolved by licensee action. l CONCIESIONS  !

One ites of noncompliance (341/79-04-04) was identified as a result of this investigation of Mr. Euron's allegations. In the other nineteen (19) instances, the allegations / areas of concern were found to be either unfounded, pretriously identified, or addressing nonsafety related areas.

In the latter case, the available details of the allegation and findings of the investigative team were provided to the licensee- for their information and corrective action as deemed appropriate.

in the identified ites of noncompliance (allegation No. 16) the identified voids in the grouting of the sacrificial shield and incomplete corrective action previously initiated by the licensee under DDR 1186 were cited as examples of noncompliance with 10 CFR 50, Appendix B, Criterion IVI.

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a The investigation associated with this allegation did not reveal any evidence that would lead to the conclusion that reinforcing steel was omitted from the RER Building. No items of noncompliance or deviations were identified. -

A11eastion No. 20: Cracks in the concrete base met of the Reactor Building.

Mr. Euron expressed his concern of the concrete cracks which developed in the reactor building base met at elevation 540'. He felt that the cracking might " allow radiation to leak out of the reactor building" and that the structural integrity of the base mat may have been impaired.

Findina: Deco had previously identified the cracks in the reactor building base slab in accordance with 10 CFR 50, Paragraph 50.55(e)(3).

The final technical report from DECO was dated November 8,1974, No.

EFZ-29,537.

Deco sumnarized the reactor building base mat cracking problem as being one of ground water, which was seeping through the radial and circunferential cracks present in the base slab. Evaluation and disposition of the cracking problem by the licensee included the following actions:

- Building Outleakage - In the case of a pipe rupture in the Reactor Building, there would be no outward leakage of radioactive water through the cracks in the floor of the building unless the basement areas became flooded to such a depth that the head of water inside was equal to or higher than that of the ground water outside. Under normal plant operation conditions, this would require flooding in the basement to a depth of approximately 30 feet before reaching the same head at the normal external l ground water. If this flooding began to occur, the reactor 3

would be brought to a safe shutdown and the water contained within the building would be processed through the radwaste system. It should be noted that this case is only valid if the cracks were not repaired.

- Sargent and Lundy, the structural designers for the reactor building, performed a thorough anlysis and concluded that the observed cracks did not impair the structural strength of the base slab.

- A program was initiated to monitor the width and length of e-

' selected cracks for an increase in length or width and to identify any new cracks which might develop.

- Crack width and the penetration into the base slab was determined by taking random concrete cores at various specified locations.

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- Developed, approved and execute procedures for the drilling, pressure testing and grouting of all cracks present in the base mat. ,

- As of the date of this investigation, MCo personnel indicated that they felt the grouting program had effectively sealed the cracks in the base slab due to the lack of infiltrating ground water. The NRC team toured the elevation 540 base slab on February 22, 1979, and found no evidence of continued water seepage. No items of noncompliance or deviations were noted.

6. 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. Unresolved items disclosed during the investigation are discussed on pages 12 and 13 (Allegations 2 and 3),

and on page 22 (Allegation No. 15) of this report.

7. Exit Interview The investigators met with site staff representatives (denoted in the Persons Contacted paragraph) at the conclusion of the investigation on March 2, 1979. The investigators summarized the scope and findings of the investigation, including the apparent items of noncompliance identified in the Results section of this report. The licensee acknowledged the findings.

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