05000316/LER-1978-101-01, /01T-0 on 781223:both Control Room Pressurization Fans Were Inoperable.Caused by Water Seepage Into a Fire Protection Sys Switch,Resulting in Actuation of Fire Spray Sys & Lockout of the Fans

/01T-0 on 781223:both Control Room Pressurization Fans Were Inoperable.Caused by Water Seepage Into a Fire Protection Sys Switch,Resulting in Actuation of Fire Spray Sys & Lockout of the Fans

ML17317A864
Person / Time
Site:	Cook
Issue date:	01/05/1979
From:	Lease R INDIANA MICHIGAN POWER CO.
To:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
Shared Package
ML17317A865	List: 05000316/LER-1978-101, Forwards LER 78-101/01T-0 05000316/LER-1978-101-01, /01T-0 on 781223:both Control Room Pressurization Fans Were Inoperable.Caused by Water Seepage Into a Fire Protection Sys Switch,Resulting in Actuation of Fire Spray Sys & Lockout of the Fans
References
LER-78-101-01T, LER-78-101-1T, NUDOCS 7901150092
Download: ML17317A864 (116)

LER-1978-101, /01T-0 on 781223:both Control Room Pressurization Fans Were Inoperable.Caused by Water Seepage Into a Fire Protection Sys Switch,Resulting in Actuation of Fire Spray Sys & Lockout of the Fans

Event date:
Report date:
3161978101R01 - NRC Website

text

NRC FORMi366 i t7 %7)

CONTROL BLOCK:

LlCENSEE'VENT REPORT U. S. NUCLEAR REGULATORYCOMMISSION

,Q

{PLEASE PRINT OR TYPE ALLREQUIRED INFORMATION) 7 8

M'.-I n

2 Q2 9

LICENSEE CODE 14 15 LICENSE NUMBER 25 26 LICENSE TYPE 30 57 CAT 58 Q

3 0~Q CON'T

~O 7

8 QB 0

I 0

5 7

9 QB EVENT DATE 74 75 REPORT DATE 80 SO~U~RCE MLQ 60 61 DOCKET NUMBER 68 69 EVENT DESCRIPTION AND PROBABLE CONSEQUENCES QIO WITH THE UNIT IN MODE 1

OPERATION BOTH CONTROL ROOM PRESSURIZATION FANS WERE

~O3

~O4

~OS DETERMINED TO BE INOPERABLE CONTRARY TO THE REQUIREMENTS OF TECH.

SPEC. 3.7.5.1.

ONE FAN WAS RETURNED TO OPERABLE STATUS IN 56 MINUTES AND THE OTHER FAN IN 62 MINUTES.

NO PROBABLE CONSE(UENCES TO THE HEALTH AND SAFETY OF THE PUBLIC.

~OB

~07

~OB 7

8 80

~OB 7

8 SYSTEM

CAUSE

CAUSE COMP.

VALVE CODE CODE SUBCODE COMPONENT CODE SUBCODE SUBCODE A

Q>t A

QTs

~E.

Q1s 0

K T

8 R

K Q14

~Ebs LZJ Qs 9

10 11 12 13 18 19 20 SEQUENTIAL OCCURRENCE REPORT REVISION LER/RO EVENT YEAR REPORT NO.

CODE TYPE NO.

Rssos~

~78

~

~10 1

~w: ~01

~T 0

21 22 23 24 26 27 28 29 30 31 32 ACTION FUTURE EFFECT SHUTDOWN ATTACHMENT NPRDC PRIME COMP.

COMPONENT TAKEN ACTION ON PLANT METHOD HOURS Q22 SUBMITTED FORM SUB.

SUPPLIER MANUFACTURER

~Q>8 LJQ LJQss MQ 0

0 0

0 T

Qss

~NQ24

~ZQss A

1 9

8 Qs 33 34 35 36 37 40 41 42 43 44 47 CAUSE DESCRIPTION AND CORRECTIVE ACTIONS Q27 o

WATER SEEPED INTO THE LOCAL ACTUATION SWITCH OF FIRE PROTECTION W

P THE CHARCOAL FILTER ASSOCIATED WITH THE CONTROL ROOM PRESSURIZATI CAUSED ACTUATION OF THE FIRE SPRAY SYSTEM AND LOCKOUT OF THE FANS.

THE SPRAY 3

ACTUATION VALVED OFF BUT THE FANS WOULD NOT RESET UNTIL THE LOCAL CONTROL BOX WAS 4

DRAINED AND DRYED OUT.

THE WATER SEEPAGE INTO THE CONTROL BOX CONTINUED NEXT PG.

7 8

9 FACILITY STATUS

% POWER OTHER STATUS QBO E

2B 1

0 0

29 NA'ETHOD OF DISCOVERY LAJQ>>

44 45 46 Qss EK LJQ ~Q ACTIVITY CONTENT RELEASED OF RELEASE AMOUNTOF ACTIVITY 6

Z Q33

~ZQ34 NA 7

8 9

10 PERSONNEL EXPOSURES NUMBER TYPE DESCRIPTION Q39

~l7 ~00 0

Qs7 ~ZQss NA 7

8 9

11 12 13 PERSONNEL INJURIES 45 44 NA

. DISCOVERY DESCRIPTION Q32 ALARM OF FIRE SPRAY OPERATIONS LOCATIONOF RELEASE Qs 80 80 80 80 7

8 9

11 12 LOSS OF OR DAMAGETO FACILITYQ43 TYPE

DESCRIPTION

s 2

Qss NA 7

8 9

10 P U8 LIC ITY 7

8 9

10 NAME OF PREPARER 71O(/5 17171Z R.

S.

Lease 68 69 PHONE 616 80 NRC USE ONLY 80 80 ss 465-5901 X-313

gC I'

~ e 1

ATTACHMENT TO LER A'8-101/01T-0 JANUARY 5, 1979 PAGE 2

"CAUSE DESCRIPTION.:AND. CORRECTIVE ACTIONS 'ONTINUED):

WAS DIRECTLY TRACED TO A CONCRETE CORE DRILLING OPERATION BY CONSTRUCTION PERSONNEL.

THE CONSTRUCTION MANAGER HAS ISSUED A LETTER TO ALL CONSTRUCTION CONTRACTORS AND CONSTRUCTION ENGINEERING SECTION HEADS DIRECTING THEM THAT WATER RUN OFF FROM ALL CONSTRUCTION JOBS MUST BE DIRECTED AND CONTROLLED.

I t

REVISION NO.

REVISION DATE DATE ISSUED ISSUED BY RECORD OF REVISIONS

/ Qp~.~@4

/o pc// gp November 4, 1986 March 30, 1987 July 24, 1987 Se tember 25, 1987 Octob 20, 1987 November 4, 1986 March 30, 1987 July 24, 1987 September 25, 1987 October 20, 1987 T.

G. Harshbarger T.

G. Harshbarger T.

G. Harshbarger T.

G. Harshbarger T.

G. Harshbarger

1 1

4

LIST OF SUPPLEMENTS SUPPLEMENT TITLE Single-Failure-Proof Crane Information NRC Request for Additional Information (August 19, 1987) via Revision 3

I

.J

FIGURE 1.3"2 DONALD C. COOK UNIT 2 STEAM GENERATOR REPAIR PROJECT ORGANIZATIONCHART MANAGER OU ALITY ASSURANCE PROJECT MANAGER O.C. COOK ASS'T. PLANT MANAGER SGR ENGINEERING MANAGER PROJECT CONTROLS CONSTRUCTION CONTRACTS RAOIATION SUPPORT CIVIL ENGINEERING MECHANICAL ENGINEERING ELECTRICAL ENGINEERING PROJECT MANAGEMENT STAFF MATERIALS HANDLING OESIGN NUCLEAR OPERATIONS SITE MANAGER OUALII' ASSURANCE SUPERVISOR HEALTH PHYSICIST CONSTRUCTION ENGINEERING MANAGER AOMINISTRATIVE MANAGER PLANT SGR PROJECT ENGINEER ENGINEERINO SUPERVISOR CONSTRUCTION SUPERVISOR

C ~

TA

- 1 INDUSTRY CODES AND ST ARDS APPLICABLE TO TllE STEAH GFNERATOR RFPAIR PROJECT COD n

A O'AR A C

T 0 Aril tt)B ~ 8 l, "lr<<'<<min<>><l<~ <1 I'r'n<r't. l c>> for Carr frig Corr-cret<."

Amer i cn>>

M< 1<if rrp Soc 1 < t y D. 1. 1-1986, "Structural Mel ling Cod<

St.ee1."

1 I

ACI 301-8I<,

",Sl)ecf ffcntions for Structural Concrete

(

Bufldfngs, Clrnpters 2 and 3."

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I A<:1 3r)4

~ B'>,

"lr<<<<<mn< ~ rr<l<<l pr'n<;t lcc a for'<<nsurfrrp, I

Hlxlrrp. Trnr>sspnr't fng, nrrd Placing, Conct'ete."

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I ACI '3l'i-BO, "0< t ni! 'rr<l Detnf ling of Concrete I

R< lrrfnr'e< m< ~ r>r."

I

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I ACI 318-83, "Buildinp Code Requireraents for Rein-forced Concrete, Clrnpters 3, 4, and 5."

~

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I Amer 1 cnn Me 1 <1 l np Snc f e ty D. 1. 3. -1981, "Structural I

Mi'<lit'lp ( <1<l<<,

Sl'l<'et 8'l >><<1.

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ASHE Bnl1< r nrr<l Prosr<rrr<<Vessel

Code,Section II, "Hnterfnl Specifications," edition and addenda in use nt tfme of mntertnl procureraent.

.I Q~j~o:

Hix propot tions shall be, selected (1) utilizing laboratory or field trial batches, (2) previous satisfactory perforraance on sfrtrflar work using the same or sfmflar materials, or (3) prior

.I eRperience Mitlr these or sfmilar materfals to provide concrete of the required strength,. durability, vork-abf.lity, econoray, etc.

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C<rr frrg <<lrnl 1 h>> for a p< r in<i nf s< v< rr (7)

I days or untf.l standnrd cured cylinders reach a comp-rehensive strenptlr of 3500 PSI, Mhfchever fs first.

Adherence tn thfs criterfn shall be arrfffcfnnt to I

preclude testing for "Evaluation of Procedures,"

I "Curing Crfterfn Fffectiveness" or "Haturfty Factor Basis."

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g~~o

.Hfx proportions shall be selected (1) utilizing laboratory or field trial batches, (2) previous satisfactory performance on similar cwork using the same or sirailar raaterials, or (3) prior experfence Mfth these or similar materials to provide concrete of the re<jufred stranpth, durability, work-ability, econoray, etc.

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TABLE 3.2 ntinued)

.0 ADDITIONAL C.

ASHF. Bol lrr nnd Pressure Vessel

Code, Secti.on III-,

"Rules for Construction of Nuclear Vessels/Rules for Constr<rctl~n of Nuclear Power Plant Components,"

rdl tiorr'n<l n<ldrndn ns discussed below.

The orip,innl Construction code for D.

C.

Cook Urrl t 2

r<<r< lrnr vr rrsnln isSection III, 1968 Edl'on plus Addenda through Minter 196&, and for pipinp componer>ts ls ANSI B31.1-1967 and ANSI B31.7-1969.

As allowed by ASHF.Section XI, Subart.icle IMA-7210, select< d portions of the original Constrtrctlon Codes dealing vith installation and tt stlnp wl11 br updated to applicable portions of Srct.rotr ill, 1983 Edit.ion plus Addenda through Summer 1986.

ASMF. Boiler arid Pressure Vessel

Code,Section IX, "Mrldin~; nnd Brazing Qualifications," edition and n<l<lrndn ln <> at t lmr of procedure qual 1 f lent ion.

ASME Boiler nnd Pressure Vessel

Code,Section XI, "Rirlr f<rr Insrrvlrn lnsprctlon of Nuclear Powrrr Plant Compon>>nts,"

1983 Fdl.tion plus Addenda through Summer.1983.

ANSI B31.1, "Power Plpinp", edition and addenda in use nt. time of. contract award for field piping, services.

T ANSI Nr<5.2 1977 Q<<alit.y Assurance Program Req<<irement.s for Nuclear Facilities USAS (ANSI) B31.l-1967, "Power Piping".

USAS (ANSI) 8'11.7-1969, "Nuclear Power Piping".

fracture toughness requirements will not apply.

N-stamping of fabricated piping components vill not. be req<<lred.

f~>~to:

- Consistent with the plant design basis, frnct<<rr to<<phnens requi rnrarnts vill not npply.

I Q~e'LLoO1:

- This code applies only to power piping not classi.fied under ASME Section III, Division l.

Q~gr~t:

- As noted under ASME Boiler and Pressure Vessel

Code,Section III, "Rules for'onstruction of Nuclear Vessels/Rules for.Construction of Nuclear Power Plant Cormponents"

above, these codes represent the original Construction Code for

CO>. nit ASTM C31 "Stnndnrd Method of Making and Curfng Concret.e Spec fmens fn the Field".

ASTM C33 "Stnncfnrtl Specffication for Coarse Apprepnt.es".

ASTH C39 "Teit Method for Compressive Strength of'ylfndrfcnl Specimens".

ASTM CliO "Test: Method for Organic Impurities f n Vine App regn t es f'r Conc re te".

ADDITIONAL I

'0 C

I I

nuclear pfpfnp components.

Portions dealing with materials nnd fabrication for new nuclear pressure retainfno; components, and installation and testing, of all nuclenr pressure retaining components, will be I

updated to ASME Section III, with'he exception that fracture toughness i tf.rements vill not apply.

I The piping desig oasis and any additional design

- I nctfvftfes refntfng to nuclear.piping systei s will be in nccordnncc wfth USAS.(ANSI) 831.1-1961.

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f. fine agpregate mny be bet.ween 2.5 and 3.0, however fndividunl snmples sltnll not vnry more thnn t).20 from the nvernpe.

I Compliance with grndntfon and ffnoness modulus requirements for fine apgrepate shal'1 consist of 4 out of 5 sitccessive test results meeting the specf.ffcatfons.

I Coarse aggregate gradation shall be.Number 57, I inch x ¹4.

I Coarse agprepnte sodium.sulfate soundness loss shall be a

10 percent maximum at 5 cycles.

I

Coarse, aggregnte Los Angeles Abrasion loss shall I

be a maximum of 40 percent at 500 revolutions.

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0 F.O ASTM C3] "Stnndnrd Hethod of Making and Curing Concrete specimens f.n the Field".

ASTH C33 "Standard Specification for Coarse Aggrc gnti s".

ASTM C39 "Ti st Hethod for Compressive Strength of Cylindrical Specimens".

ASTM CliO "Test Method for Orf;anic Impurities in Fine Aggregates for Concrete".

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I ADDITIONAL O

C for nuclear piping components.

Portions dealing with materi.als and fabrication for new nuclear pressure retaining components, and installation and testing of all nuclear pressure retaining components, will be updated to ASMF.Section III, with the exception that fracture toughness requirements wt.ll not apply.

The piping design basis and any additional design activities relating to nuclear piping systems will be in accordance with USAS (ANSI) B31.1-1967.

fine aggregate may be between 2.5 and 3.0, however lndlvi dual ¹iiliiiilii¹Shall no( v¹ry mor¹ t h¹n 0. i 0 from the average.

- Compliance with gradation and fineness modulus requirements for fine aggregate shall consist of 4 out of 5 successive test results meeting the specifications.

- Coarse aggregate gradation shall be Number 57; 1 inch x ¹4.

- Coarse aggregate sodium sulfate soundness loss shall be a

10 percent maximum at 5 cycles.

Coarse aggregate Los'Angeles Abrasion loss shall be a maximiim of 40 percent at 500 revolutions.

TABLE 3.

ntinued)

ADDITIONAL ASTM C88 "Test Method for Soundness of Aggregates by Use of Sodium Sulfate or Magnesium Sulfate".

ASTM C96 "Scanclard Specification for Ready Mix Concrete".

ASTM C117 "Tc st Mc thod for Materials Finer Than No.

?00 Sieve in Mineral Aggregates by Mashing".

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ASTM C123 "Tc st Hc tl>ocl for Lightwei.ght Pieces I n Agp rc pa t.e".

ASTM C127 "Test Method for Specific Gravity and Adsorption of Coarse Aggregate".

ASTM Cl PII "Tc st M@I heel for Speci.f ic Cravl ty and Adsorption for Fine Aggregate".

ASTH C131 "Test Mc thod of Resistance to Dccgradation of Small-Size Coarse Aggiegate by Abrnslon and Impact. in the Los Angeles Machine" ASTH C136 "Hc thocl for Sic.ve Analysis of Fine and Coarse Appregates".

ASTM C138 "Test Hethod for Unit Meight, Yield, and Air Content (Gravimetric) of Concrete".

ASTM Clli2 "Test Method fox Cl'ay Lumps and Friable Partlcl<<s ln Apprcpatc".

ASTM Clci3 "Test Method for Slump of Portland Cement Concrete".

ASTM C150 "Specification for Portland Cement".

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I lieu of glass plate for unit weight determination.

Except "Yield" and "Ai.r Content (Gravimetric)"

portions will not be utilized.

set when tested in accordance with ASTH Cli51.

TABLE 3.

ontinued)

COD 0

I ASTH C231 "Test Method for Air Content of Freshly Mixed Concrete by the Pressure Method".

ASTH (:260 "Spc c f flent f ons for Air-Entrained Admfxtures for (.oncrrt.e".

Compounds for Curinp Concrete".

in Portland Cemctit Concrete".

ASTM C494 "Specification for Chemical Admixtures in Concrc tc '.

ASTM C566 "Test Method for Total Moisture Content of Apprc pnt r by Drying".

ASTM C6'.7 "Practice for Capping Cylindrical Concrete Sprcimc ns".

ASTM C618 "Sprcf ffcntfon for Fly Ash and Rain or Cnlclnecl Nnturnl'ozzolan for Use as a Hineral Admfxturr in Portlnncl Cement Concrete".

ASTM (:7t)2 "Het.hods for Reducing Field Samples of Apprepnte to Testing Size".

ASTH C289 "Trst Mc thod for Potential Reactivity of'ppr<<gnt.c s ((:hrmfcnl Method)".

AS'I'H t-'ftp) "Sficcl f lent,lotts f'r Liquid Membrane-Forming ASTM C311 "Metliods of Sampling and Testing Fly Ash or Natural Pozzolnns

. or.Use as a Mineral Admixture I

ASTH C112 "Hethod of Sampling Freshly Mixed Concrete".

I I

Except total alkalies shall not exceed 0.60 percet>t by vofpht when calculnted as the percentage of Na20 plus 0.658 times the percentage of K20.

Q~e

~o

- Only the Type B Apparatus shall be utilized.

TABLE 3.2-1 (Continued)

I

(0

)) '()

~

~

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I SSPC-SPl throuyh SP10

- 1982 Steel Structures Painting Council Specifications for Surface Preparat ion of Steel Surfaces I

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Note:

1)

All ASTMs are latest edition.

ll~'Salt~~~~&~~~~~%4llavar~a~i4'A4 li~

TABLE 2

USNRC REGULATORY CUIDPS APPLICABLE TO THE STEAM GFNERATOR REPAIR PROJECT FIE1.D MARK REGUIATORY CUIDF:

U

))FR R EGU LATORY GUIDE REGULATORY GUIDE ADDITIONAL 0

C 1.8 Personnel Selection and Training I-R (9/7S)

Committed to in UFSAR, Section 1.7, "QAPD", Appendi.x A.

).26 Quality Croup Classification and Standards for Mater,

Steam, and Rnd-wnstrt Containing Components of Nuclear Power Plants 3 (2/76)

Classification of Class 2 and 3.

components for the purpose of fmplamenting ASME Section XI requireraents was made in accordance wl C)1 t)l)n Ruflie.

Snfety Gu 1 r)r 30 Q<<nifty Assurance Requirements for Installation, Inspection

=

and Testing of Instrumentation nrrd E)ectrfcnl Equipment (B/72)

Committed to in UFSAR, Section 1.7, "QA)'D", Appndfx. A, 1.31 Safety Cuidr 33 Cont.rol of. Ferrite Content fn Strrfnless Steel Meld Metal Quality Assurance Program Requirements (Operatf.onal) 3 (4/78)

(11/72)

The requirements of this guide are now covered by ASME Section III.

Field work relating to the steam generator rrtpafr pro]ect wf,ll he -fn compliance wfth this regulatory guide.

Committed to in UFSAR; Section 1.7, "QAPD", Appendix A.

1.31 Quality Assurance Requirements for Cleaning of Fluid Systems nnd Associated Components of Mater-Cooled Nuclear Power Plants 0 (3/73) 'ommitted to irr VFSAR, Section 1.7, "QAPD", Appendix A.

TABLE 3.2-inued)

R FGU]ATORY GUI DF.

U B.R REGULATORY GUIDE REGU]ATORY GUIDE 0

ADDITIONAL iV C.

0 1.38 Quality Assurance Requirements for. Packing,

Shipping, Receiving

Storage, and ))andling of Items for Water-Cooled Nuclear Power

)'lants 1 (10/76)

C~mmitted to in UFSAR, Section 1.7, "QAPD", Appendix A.

1.39

)!ousekeeping Requirements for Mater-Cooled Nuclear Power Plants 1 (10/76)

Committed to in UFSAR, Section 1.7, "QAPD", Appendix AD Control of Sensitized Stainless Stee]

0 (5/73)

If applicable to this repair pro)ect, the field vork vill comply to this pu) de.

].48 Design Limits and Loading Corn)> f natl ono for Sa iam] c Category I Fluid System Components This regulatory, guide was vithdravn

]/4/85 (onn 50FR9732).

l. 50 Control of Preheat Temperature for Melding of Lov-Alloy St<<el 0 (5/73)

Prnj<<ct repair woik wf)) be perform<<d in compliance with this regulatory guide.

1. 54 Qual]ty Assurance Requirements for Protective Coatings Applied 0 (6/73)

Committed to in UFSAR, Section 1.7, "QAPD", Appendix A.

Exceptfon:

Committed only to ANSI N101.4-1972.

].58 Qua]ificati.on of Nuclear Pover Plant Inspection Examination and Testing Personnel 1 (9/80) 'ommitted to in UFSAR, Section 1.7, "QAPD", Appendix A.

0 TABLE 3

~ 2-inued)

REGUIATORY GUI DF.

O'E REGUIATORY GUIDE REGULATORY GUI DF.

0 ADDITIONAL N

C 1.64 Quality Assurance Requirements for the Design of Nuclear.

I'ower Plants 0 (10/73)

Committed to in UFSAR, Section 1.7, "QAPD", Appendix A.

1.68 Initial Test Program for Water-Cooled Nuclear Power Plants 2 (8/78)

This-regulatory guide will be used only for guidance in developing a test program for those components and systems affected by the Steam Generator Repair Pro]ect.

1.71 Welder Qualifications for Areas of'.fmfte~I Accessibility 0 (12/73)

Welders making velds in areas of re..tricted accessibility will be required to practice and qualify on a similar configuration to the weld being made.

1. 14 1.88 Qual It y Aasuranre Terms and Defi>>iti.ons Collection, Storage, and Mnfntenance of Nuclear Power Plants Quali.ty Assurance Records 0 (?/74) 2 (10/76)'ommitted to in VFSAR, Section 1.7, "QAPD", Appendix A.

Committed to in UFSAR, Section 1.7,

",QAPD", Appendix A.

1.89 I:.nvironmental Qualification of "c rtain Electrical Equipment Important to Safety for Nuclear

.Power Plants

'1 (7/84)

Pro)ect repair work vill be performed in accordance vith this regulatory guide.

TABLE 3.2 inued)

RFGUI~T<NY GU I D V.

NU BFR R FGU IATORY GUIDE RFGUIATORY GUIDE

'V 0

ADDITIONAI.

0 ON F C.

ONS I

A ~

~ ~

Quality Assurance Requirements for Installation, Inspection, and Testing of Structural Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants 1 (4/76)

C109) included in Table B of ANSI N45.2.5-1974 is inappropriate for field testing as it is a sophisticated 1abnratory test utilized for cement evaluation.

In lieu of daily tests, pre-packaged non-shrink grouts shall be accepted for use on the basis of manufacturer's certification or compressive strength tests made in the field.

Confirmation compressive strength tests shall be made during the first day's production and Ihttrssaftef on a basis of ssilhssr once per day of every one-hundred (100) bags

used, whichever is least.

Mater an<! ice sha11 be sampled and

'ested to ensure either potability or certified to contain not more than. 2,000 parts per million of chlorides a'l, nor more than 1,500 parts per million of sulfates as SO4.

Acceptability of this water or ice shall be per this certification and preclude the ASTM's referenced in Table B of ANSI N45.2.5-1974.

The ref.erence, in Table B of ANSI N45.2.5-1974, to soft fragment testing per ASTM changed designations to ASTM C851 which was deleted in 1985.'o testing for soft fragments is intended.

~ J

~Pxcs t e: Sister splices vill be substituted for production splice requi.red for tensile testing under Section 4. 9 of ANSI N45. 2. 5-1974.

TABLE 3.2-inued)

R EGU1 ATOPY GUIDE

'tUHBF.

RFGULATORY GUIDE REGUIATORY GUIDE

'V S 0 ADDITIONAL 0

C.

S 1.100 Seismic Qualification of Flectric Fqulpment Important to Safety f or Nuclear Power Plant; I (8/77)

Pro]ect repair work will be performed in accordance with this reftrlatory guide.

1.116 Qirnlit.y Assurance Requirements for Installation, Inspection, nnd Testing of Mechanical Equipment and Systems.

0-R (5/77)

Exception:

Commi tted to ANSI N45. 2. 8 (1975),

"Supplementary Quality Assurance Reqirirements for Installation, Inspection nnd Testing of Mechanical Fquipment and Systems for the Cnnntruction Pharrn of Nuclonr Pownr Plants" per UFSAR, Section 1.7, "QAPD",

Appendix A.

Not committed to this rr pulnt.nry guide.

1.173 Qirnl1ty Assurance Requirements for Control of Procurement.

of Items and Services for Nuclear P1nrr t' I (7/77)

Committed to in UFSAR, Section 1.7, "QAPD", Appendix A.

1.131 Qualification Tests of Electric

Cables, Field Splices, and Connections for Light-Mater-Cooled Nuclr nr Power Plants 0 (8/77)

Pro]ect field work will be performed in accordance with this regulatory guide.

1.144 Auditing of Quality Assurance Programs for Nuclear Power Plants 0 (1/79)

Committed to in UFSAR, Section 1.7, "QAPD", Appendix A.

1. 146 Qunl ification of Quality Assurance Program Audit Personnel for Nirclear Power Plants 0 (8/8ri)

Committed to in UFSAR, Section 1.7, "QAPD", Appendix A.

TARLE 3.2.3 D. C. UXX UNIT 2 TECHNICAL SPECTFlCATlONS NOT APPLLCABLE DLAIINQ TK STEAN GENERATOR RKPAlR PR(LIKCT TKCHNlCAL SPECtflCAllON R

T lTLE ADOlT lONAL Nf HATION 3.1.1.3 Reactivtty Control System.

Sorm D~tutton This TechnicaL Speci',ication ensures adequate

~txtng of coolant u'I so thc tou boron caatntratim streaa being introduced. into t' systea.

'This aixing prcvcnts a targe ccrecntratim gradient in the core Mich uoutd caae Locat tacd poucr excursions.

tftth no fuel.

in the reactor vessel, there is no corx:em about decay heat reeovat or boron mixing.

Rcactivsty Contat Systees-Soratior. Systcos

- floe Paths

~

Shutdown".

This Technical Specification requires that me borm injection fto~ path reroins operable.

This ensures that negative reactivity control ls avai table.

Mith no fuel in the reactor vessel there is no need for negatiw reactivity control.

3.1.2.5 Reactivity Control Systcm-Soric Acid Transfer Pepa-Shutdcan E

This Technical Specification ~iree that at Least one boric acid transfer ~ nmin opcrabt e, Th 'is cnswcs that negat ive reactivity control is avaitable.

Mith no fuel in the reactor vessel there is no need for negatiw reactivity cmtrot.

3.3.3.9 l ns 'trURtltat iEFt Radl oac t 'Ive Liquid KffLuen: tnstnacntat ion, speci ficatly t:~ fot toeing survei t tame reaircmnts:

C.3.3.92, lb Steea Generator dtovdovn Line (2-R-19)

C.3.3.9.2, Tc Stem Cenerator Stouoovn 1rea~t Efftrent (2

R 2C)

%accuse there Nitt be no steea or stem generators these GJO MAltors N: Ll not be

~'Ihtalned cperabte, eVLSXO:t 2

TARLE 3.2i3 (Contin <)

TED I CAL SPECI F ICATIIN ACOITI ORAL.

N I

3.3.3.10 Iratruscntatie

- Radioactive gaseous Process and Effluent Non'or ing Ins:nrentat ion, Specifically ~ follcaing arvti LLaree reyairmnts:

48.3.10.2, 2a Condenser E~t ion Svstoa acbl e Cas Activity Noni to (SRA.2905) 4.3.3.10.2, 2b Condemcr Kv~ticn Systei Effl~t Flov Rate (SFR.40'i, 2%-054, SRA.2g10) 4.3.3.10.2, 6a gland Seal Exhaust Noble gas Ac:Ivity (SRA 2%5) 4.3.3.10.2, bb System Effluent Flov Rate (SFR 2CI, 2.IO 054, SRA 2510) gecause there viLL be no stem or stem generators these eight eanitors uILL not be mintained operable.

3.4.7 Reactor Coolanc Systm - Choaistry This 'technical speci'f ication provides odphte corrosicn protection to ensure the structural integrity of the Reactor Coolmt Systea over the life of the plant.

Ouring the atcae generator repair there ~ILL be a period of asproximtely six aeths shen the Reactor CooLant Systca uiLL be drained to half-loop, the reactor vessel head vill be in place and the Residual Reat Rceoval PLaps uILL be shutckan.

Our ing this portion of the outage it vlI l ret be possible to obtain a chcaistry saaple free the Reactor Coolant Systea.

Therefore the Reactor Coolant Systm ski!l be placed althin speci fIce;<m liaits prior to this sh tdoatl and lsolatlcA pef iod.

saapLIng cm be reestablishad foLLowing the stem ycr>>rator repair it uiLL be verified that the Reactor Coolant Systee is still within the awaistry Liaits.

If the Reactor Coolant Systee is not aithin the cheaistry Liaits, the systea uiLL be cleaner' prior to reloading "fmL into the reactor.

Our engineering

j TABLE 3.2-3 (Cont inued)

TECIIRLCAL SPEC l F lCAT LOW T lTLE ADOITLQQL N

TCW 1

C

-.t evaluation has detof%ined that the stfl&tural integrity of the Reactor Coolmt Syst<< lliLL not be diainishad by an csllikely increase in chlorides or flmrides above the technicaL

~pecificatim Liaits of 0.16 Fpa.

This is based on the Reactor Coolnt Syst<<being at

~sbient t<<peratuce during this period and that stress corrosion crackLIS) (SCC) does not occur belcal 8FF and I arely at Less than 145 F ~

Also, SCC does not occur mtiL the concentraticxl of chloride and fluoride reaches several ordcl s of %agni tude above 'the technicaL specification L lait of 0.15 ppa; the Level belch erich the Reactor Coolant Syst<<vill be Left at during the period of shutdan and iSOLatiCSI.

3.9.1 Rtfue Ling Operat icols Ioron Concent rat ical Since there IllLL be no fuel in the reactor vessel Liaitations on reactivity condltiore in the reactor vesseL are no Longer ~ ccxlcern 3.9.2 Refueling Operaticre-Tnstrl<<ntatiel Since there lliLL be no fueL in the reactor vessel there illLl be no charge in the reactivity condition of the core, therefore, the aarce range ceutron flux anitors are not needed 3.9.8.1 Refmlirg Operatiels. - Residull Neat Remval and Coolnt Circulation Vith cxl fmL in the reactor vessel there sill be no residual heat to~.

Therefore, there is no need to mintain an operatimaL residual heat r~L Loop.

3.9.5.2 Refueling Operaticxls. Los Mater Level Vith re fmL in the reactor vessel there ill'L be no residual heat to reeeve.

Therefore, there is no need to saintain n operational residuIL heat removal loop.

Fov s'on 2

TASLE 3.2-3 (Cont inued)

TECNICAL SPECIFICATI&

TITLE A&I7 IOKAL T

C NT ~

d.S.T.Qa)

Adslnistratiw Controls - Plant Nuclear Safety Reviecc Caasitte>>-

Responslbil ities The P>>SRC viLL revieu the foLLoccirg item generator repair project doccsctents:

1.

The Stem Generator Repair Report 2.

The Stem Generator, Repair Ouality Assurmae Progrm 3.

I ceeedures covering rat~ to service testtng.

6.8.2 Adccinistrat ive Controls.

Procechres The PaSRC viLL reviecc the procccisres w'itten covering return to service testing.

d. II.3 Adsinistratiw Con'.role.

P rocechres Tmporary changes aced>> to proctdw%$ covering retcsn to ~ice testing provided it~ ~, b,

~nd'c.of technical spec ification d.8.3 are satisf ied.

d.12.2 Adainistratiw'Controls. Nigh Radiatin Area The keys to those h'igh radiation~ turned over to the stem generator project tern shall be mintainad seder the adainistra:ive control of the Project Neelth Physicist.

- 49c-Revis.on 2

Figure 8.8.1 D.C. COOK NUCLEAR POmR STATION SmAM GENERATOR REPAIR PROJECT RADIATION PRO'mCTION ORGAMZATION VICE PRESIDENT HUGEAR OPERAllOHS ANT OIVISI ANACER NUCLEAR OPERAllONS I

I A.E.PWC.

COLULQUS O.C. COOK NUCEAR STATION PLANT NANADER (COOK PLANT)

PLANT RADIATION PROTECllOH SUPERVISOR SECDON LlAHACER RADIOLOCCAL SUPPORT IT PROJECT HEALTH pHYslasT S/C REPAIR PROJECT HAHACER SITE HAHACER ROBOT RADIATI PROTECTION COORDINATOR XI Pl V)

O RADIATION PROlECllOH SUPERVISOR RADIAllON PROTECllOH TEcHMaANs RADIATIOH PROlECllOH SUPPORT SER'ACES COORDINATOR PROJECT ALARA COORDDIATOR EHCHEERS (AS NEEDED)

RADIATIOH PROTECTION SUPERVISOR RADIATION PROTECllOH TECHNIaANS DOSIMETRY AND RECORDS SUPERVISOR OOSDJETRY AHD RECORDS CLERKS l

C SUPERVISOR I

I ALARA J

PERSONNEL

Section 2.

.2 j 7/8 Qp/

+ir@T

~Pa e

T Zr Parametric Comparison ~

28 2.2.3 2.3 Materials Comparison Component Design Improvements 28 32 2.3.1 Design Improvements to Minimize otential for Tube Degradation 32 2.3.2 2.3.3 Desi n Improvements to Increase.

Performance Design mprovements to Enhance Maintainability and Reli bility 37 38 2.4 40 2'.1 2.4.2 2.5 Industry Codes nd Standards USNRC Regulatory qides Shop Tests am% Inspe ons 40 40b 42 SECTIOR 3 -

hIR PROJECT 3.1 47 3.2 3.3 Guidelines and Criteria

=

Preshutdown hctivities 47 50 3.3.1 3.3.2 3.4 3.4.1 Site Preparation Shipment and Storage of Replacement Com nents Post Shutdown hetivities Containment Preparations 50 54 55 3.4.2 Removal of Concrete, Structural and Equipment Interferences 57 3.5 3.5.1 3.5.2 Steaa Generator Removal hetivities Steam Generator Cutting Methods and Locations Removal and Handling of the Steam Generator Upper Assemblies 62 62 64 3.5.3 Removal and Handling of the Steam Generator Lower Assemblies 66 Revision 1

~Seceio 6.2 '

6.2.2 6.3 Handling of Heavy Loads Shared System Analysis Analysis of Significant I nzards

~Pa e

.155 162 163 6.3.1 6.3.2 6.3.3 Criterion 1 Criterion 2 Criterion 3 164 164 165 SECTXOZ 7 - EHVI3KHHHGM REPORT 7.1 7.2 7.2.1 Purpose of the EmrLroxxaental Report The Plant and 2hmiroxuacntal Xnterfaccs Geography and Demography 166 166 166 7.2.2 7.2.3 7.2.4 7.2.5 7.2.6 7.3 7.3.1 7.3.2 Regional Historic, Archaeological, Architectural,

Scenic, Cultural, and Natural Features Hydrology Geology Ecology Noise Eon-RacHological Enviroxmental Effects Geography and Demography Regional Historic, Archaeological, Architectural, Scenic, Cultural, and Natural Features 167 167 168 168 169 169 169 170 7.3.3 7.3.4 Hydrology Geology 170 171 7.3.5 7.3.6 7.4 7.4.1 Ecology Noise Radiological Emrironmcntal Effects Occupational Exposure 171 172 172 172 Revision 0

~

~

IZST OP TABLES Tab e

Title

~Pa e

1.1-1 D.

C.

Cook Nuclear Plant Secondary Side Water Chemistry Specification History-Steam Generator 2.2-1 Comparison Between the Original and Repaired Steam Generators 27 2.2-2 Comparison of Design Data Between the Original and Repaired Steam Generators 29 2.2-3 Comparison of Materials of Construction Between the Original and Repaired Steam Generators 31 3.2-1 Industry Codes and Standards Applicable to the Steam Generator Repair Project Field Work 49a 3.2-2 USNRC Regulatory Guides Applicable to the Steam Generator Repair Project Field Work 49f 3.6-1 3.8-1 3'-2 Steam Generator Repair Welds Repair Project Manrem Estimates Pro)ected Project Totals by Phase for Man-hours and Man-rem 75 98 103 7.4-1 7.4-2 7.4-3 Donald C.

Cook Annual Man-rem Expenditures Steam Generator Man-rem Expenditure Comparison Gross Contamination Levels by Location in Piping and Steam Generator 173 174 180 7.4-4 Donald C.

Cook Nuclear Plant Unit 2 Estimated Steam Generator Curie Content 181 7.4-5 Effluent Release Isotopic Distributors, Steam Generator Replacement

Project, Surry Power Station

- Unit No.

2 182 7.4-6 Comparison of Gaseous Effluent Releases from Donald C.

Cook Nuclear Plant 183 7.4-7 7.4-8 7.4-9 Radionuclide Concentrations in Reactor Coolant 184 Estimated Radionuclide Releases Due to Discharge of Reactor Coolant Water 186 Estimated Specific Activities of Laundry Waste Water 185 vii Revision 1

~

~

8

XZST OF XhSIXS cont'd.

Table 7.4-10 Title Estimated Radioactive Liquid Effluent Releases During the Donald C.

Cook Unit 2 Steam Generator Repair Proj ect

~Pa e

188 7.4-11 Comparison of Radioactive Liquid Effluent Releases 189

7. 8.-1 Summary Cost-Benefit Analysis for the Unit 2 Steam Generator Repair Project 200 viia Revision 1

FIGURE 2.2-2

.MODIFICATIONS TO UPPER ASSEMBLY INTERNALS UPPER ASSEMBLY NEW STEAM WATER DEFLECTORS (3 TOTAL)

EXISTING SWIRL VANE ASSEMBLY

???@Spy

+<X???. i 0??."~i'W'v'?.~

")i,.Ccg'4>??:?.v%,AV..

4 0 0 0 0 0 0 00 0 0 00 0

04

• 40 4

0 4 4 4 0

4 0 4 4 O4O4 44 4 I

EXISTING SECONDARY SEPARATOR SECONDARY MANWAY (2-180'PART)

,NEW DRYER DRAINS (8 TOTAL NEW STEAM CHIMNEYS

'(2 TOTAL SHELL

,NEW FEEDWATER RING AND INCONEL J-NOZZLES I~I~~> IK4MEQNIM:N'>".;,:.l NEW LOWER ASSEMBLY

'1',h1,; 'lb?49/ "'

"', "',,Oi ??0%'l',4? i'5, '?

XEC4+4~eNCiw'~4Jw'l.'Cv s;

..~mk~iiwiiV~wgi4A 44 REVISION 0

1 I

'I

TIFF.

2-1 XNIXJBXRY OODES AND PQQKlARDS APPIZCMKF~ TO 'LHE STEAN GEN1RAXOR REPAIR PRMZCZ I0 ACI 301-84, >>Specifications for Structural Oancxete Buildincpa, QMLptexs 2 and 3.>>

ACZ 304-85, "Reccaznended Practices for Measuring, Mixing, Transportirg, and Placing Oancxete.>>

ACZ 315-80, "Details and Detailing of Oancrete Reinforaernent."

@~Lion Mix proportions shall be selected (1) utilizing laboratory or field trial batches, (2) previous satisfactory performance on similar work using the same or similar materials, or (3) prior mqoerience with these or similar materials to pxovide concrete of the xequixed strength, durability, work-ability, economy, etc...

ACZ 308-81, cv~.>>

Practice for Curing Can-

~Mon:

Curing shall be for a period of seven (7) days or until standard curd cylinders xeach a camp-xehensive strength of 3500 PSI, whichever is first.

Adherence to this criteria shall be sufficient to preclude testing for "Evaluation of Pxoceduxes,>>

"Curing Criteria Effectiveness" or 'Maturity Factor Basis."

"MldfW foxced Oancxete, Chapters 3, 4, and 5..>>

American Welding Society D.l.1-1986, >>Stxuctural Welding Oode Steel.>>

American Welding Society D.1.3.-1981, >>Stxuctural Welding Oode, Sheet Steel."

ASHE Boiler and Pressure Vesse1 Oode,Section II,

'~terial Specifications," edition ancl addenda in use

)

at time of material pxocurenent; E>my<ion:

Mix proportions shall be selected (1) utilizing laboratory or field trial batches, (2) pxeviaus satisfactory performance an similar work using the same or similar materials, or (3) prior experience with these or similar materials to pxovide

)

concrete of the required strength, durability, work-ability, econamy, etc.

t4

TABLE 3.2-ntinued)

CODE OR PVQlQARD ASME Boiler and Pressure Vessel Code,Section III, "Rules for Oonstruction of Nuclear Vessels/Rules for Construction of Nuclear Power Plant Ocmponents,"

edition and addenda as discussed below.

'Ihe original Oonstruction code for D. C. Gook Unit 2 nuclear vessels isSection III, 1968 Edition plus Addenda throb Winter 1968, and for piping cxzqxments is ANSI B31.1-1967 and ANSI B31. 7-1969.

~Exes iona: Consistent with the plant design basis, 1m~ wy.

N~mping of fabricated piping camponents willnot be ra~

As allowed by ASME Section XI, Subarticle IWA.-7210, portions of the original Construction Godes dealing with installation and testing will be updated to applicable portions of Section III, 1983 Edition plus Addenda through Saner 1984.

ASME Boiler and Pressure Vessel Code,Section IX, "Welding and Brazing Qualifications," edition and addenda in use at time of procedure qualification.

ASME Boiler and Pressure Vessel Oode,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components " 1983 Mition plus Addenda through Sunmer 1983.

~Exes ion: Consistent with the plant design basis, fracture toughness requirements willnot apply.

ANSI B31 1 g "E~ Pipirg", edition and addenda in use at time of contract aware for field piping services.

Exce+ion: - 'this code applies only to pcarer piping not classified under ASME Section III, Division l.

USAS (ANSI) B31. 1-1967 Il~~ Pipingn USAS (ANSI) B31.7-1969'Nuclear Power Piping

~Exes ion: As noted under ASME Boiler ani pressure Vessel Code,Section III, "Rules for Construction of Nuclear Vessels/Rules for Co~ction of Nuclear~ Plant Gamponents"

abave, these codes represent the original Oonstruction Gode for for nuclear piping components.

Portions dealing with materials and fabrication for new nuclear piping components, and installation and testing of all nuclear piping acaqmnents, willbe updated to ASME Section III, with the exception that fracture toughness requirenents willnot apply.

TAIKZ 3.2-ntinued)

APIN C31 "Standard Method of Malcing and Curing Goncrete Specimens in the Field".

APIN C33 "Standard Specification for Coarse Aggregates".

%he piping design basis and any additional design activities relating to nuclear piping systems will be in accordame with USAS (ANSI) B31.1-1967.-

~Zxoe

'ons:

'lhe average fineness mcdulns of the fine aggregate may be between 2.5 and 3.0, however individual samples shall not vary more than 0.20 fram the average.

- Gampliance with gradation and fineness mxhQ.us gg lit ~1 of 4 out of 5 successive test results meeting the specifications.

- Ooarse aggregate gradation shall be Nuaher 57, 1 inch x N4.

ASIN C39 "Test Method for Oampressive Strergth of Cylindrical Specimens".

AKH C40 "Test Method for Organic Impurities in Fine Aggregates for Ooncrete".

- Ooarse aggrecpte sodium sulfate soundness loss shall be a 10 percent maxim at 5 cycles.

- Ooarse aggregate los Angeles Abrasion loss shall be a maxhmzn of 40 percent at 500 revolutions.

APIM C88 "Test Method for Soundness of Aggregates by Use of Sodium Sulfate or Magnesium Sulfate".

ASIM C94 "Stardard Specification for Ready Mix Concrete".

ASXN C117 "Test Method for Materials Finer

'Khan No. 200 Sieve in Mineral Aggregates by Mashing".

~ ~

TABLE 3.2 tinued)

CODE OR SZANIRR3 ASXM Cl23 "Test Hethod for Lightweight Pieces in Aggregate".

ASXM'C127 "Test Method for Specific Gravity and Adsorption of Ooarse Aggregate".

ASXM C128 "Test Method for Specific Gravity and Adsorption for Fine Acgv~te".

ASXN C131 nTest Hethod of ResistaIKe to Degradation of Small-Size Ooarse Aggregate by Abrasion and Impact in the tus Angeles Hachine".

ASXM C136 'Method for Sieve Analysis of Fine arxl rse 2~;egates>

t ~

ASXM C138 "Test Method for Unit Weight, Yield, and Air Content (Gravimetric) of Concrete".

ASXN C142 "Test Method for Clay Imops and Friable Particles in Aggregate" ASXM C143 "Test Hethod for Slump of Portland Oement Ooncrete".

ASXM C150 "Specification for Portland Oement".

ASXN C172 'Method of Sampling Fre@Q.y Mixed Concreten.

Exee~Mons: - Except strike off bar utilized in lieu of glass pl.ate for unit weight deterzaination.

E~t 'tYieldn and nAir Content (Gravimetric) portions willnot be utilized.

~Esne ons:

Recept cement shall te free of false set when tested in accordance with ASXM C451.

E~k total alkalies shall not e>mee5 0.60 percent by weight when calculated as the percentage of Na0 plus 0.658 times the percentage of K 0.

~ ~

TABZZ 3.2 Continued)

CODE OR FRNIRRD ASIN C231 "Test Method for Air Content of Freshly Mixed Concrete hy the Pressure NethocV'.

ASIN C260 "Specifications for Air Ehtzained ikhbd f

APIN C289 "Test Method for Potential Reactivity of Aggregates (Chemical Method) ".

APIN C309 "Specifications for Liquid Membrane-Forming Campounds for Curing Concrete".

ASIN C311 'methods of Sampling and Testing Fly Ash or Natural Pozzolans for Use as a Mineral Admixture in Portland Cement Oonco~".

ASIA G494 "Specification for Chemical in Conc:Eaten.

ASIN C566 "Test Method for Total Moisture Conte'f Aggregate by DryixxP.

APIN C617 "Practice for Capping Cylindrical Concrete Specimens".

ASIN C618 "Specification for Fly Ash and Rain or Calcined Natural Pozzolan for Use as a Mineral Admixture in Portland anent Concrete" ASIN C702 'Methods for Reducing Field Samples of Aggregate to Testing Size".

SSPC-SPl thxough SP10 - 1982 Steel Structures Pahxtiag Council Specificatians for Surface Preparatian of Steel Surfaces Exec~on: - Only the Type B ApImratus shall be utilized.

Notes:

1) All AS%Ms are latest edition.

TABIF

- 2 USNRC REGUIATORY GUIDES ~CABLE KO 'LHE STEAM 61XERAER REPAIR HKOECT HZID WORK 1.8 REGUZAXGRY GUIDE TITIZ HEGUIAIQKf GUIDE REVISION 1-R (9/75)

OQK6tted to in UFKR, Section 1.7, "QAPD", Appendix A.

1.26 Safety Guide 30 1.31 Quality Gzuup Classification and Standards for Water, Steam, and Rad-waste Oantaining Ccaaponents of Nuclear Paver Plants

~ttf for Installation, Inspection I

ttft f~tf and Electrical Equipment Oontrol of Ferrite Oord~

in Stainless Steel Weld Metal 3 (2/76}

(8/72) 3 (4/78)

Classification of Class 2 and 3 campceents for the purpose of implementing ASHE Section XZ I I wi.th this guide.

Oammitted to in UFSAR, Section 1.7, "QAPD", hppr~ A.

f ttt tft new covered by ASME Section III.

Field work relating to the stean generator repair project willbe in ccepliance with this regulatory guide.

Safety Guide 33 Quality Assurance Program Itt tf ~I

(~72)

~tted to in UFSAR, Section 1.7, "QAPD", Apped A.

1.37 Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Oamponents of WateL~ooled Nuclear Power Plants 0 (3/73)

Oammitted to in UFKB, Section 1.7, "969", Appendix A.

TABIZ 3.2-ntinued)

RHGUIAIORY GUIDE NUMEN 1.38 MGOEAIQRY GUIDE TZKZ Quality Assurance Requirenents for Pacify, Ships@, Receiving

Storage, ancl Handling of Items for WaM~ooled Nuclear Pc@mr Plants 1 (10/76)

Oammitted to in UFSAR, Section 1.7, nQAPDn, Appendix A 1.39 1.44 1.48 1.50 1.54 1.58 d~jw~ f Water-Oooled Nuclear Poem'lants Oantml of Sensitized. Stainless Design Limits and Ioading Oambinatians for Seismic Category I Fluid System Oamponents Oontrol of Preheat Teatperature for Weldizg of law"Allay Stee1 for Pmtective Coatings Applied Qualification of Nuclear Pamr Plant Inspectian Bamination and Testing Personnel 1 (10/76) 0 (5/73) 0 (5/73) 0 (6/73) 1 (9/80)

Oammitted to in UFBAR, Section 1.7, nQAPDn, AppeDdiX A If applicable to this reImir project, the field work will camply to this guide.

'Ihis regulatory guide was withdrawn 3/4/85 (see 50FR9732).

Project repair work willbe performed in campliance with this regulatory guide.

Oammitted to in UFSAR, ~on 1.7, nQAPDn ApI~ndix A.

Exceptio Committed only to ANSI N101.4-1972.

Oammitted to in UFBAR, Section 1.7, nQAPDn Apped A.

TMKH 3.2-2

)

RHGUM1QEK GUIDE REVISION tWltb for the Design of Nuclear Power Plants 0 (10/73)

Committed to in UFBAR~ Section 1.7g nQAPD'r Appendix A Initial Test Pxogram for Watex Cooled Nuclear P0wer Plants Welder Qualifications for Areas of Limited Accessibility Quality Assurance Texan and Definitions Collection, Storage, and IRDk f

CI Plants Quality Assurance Records 2 (8/78) 0 (12/73) 0 (2/74) 2 (10/76) this regulatory guide willbe used only for guidance in developing a test program for those components and systems affected by the Steam Generator ReLair Pmject.

Welders ma)de welds in areas of restricted accessibility willbe required to practice and qualify on a similar configuration to the weld being made.

Cam6tted to in UFKK, Sectian 1.7, nQAPDn Appendix A Oammitted to in UFKR, Section 1.7, nQAPDn-Appendix A Envtmranental Qualification of CBl.tain Electrical EquipIIPAIt Important to Safety for Nuclear Power Plants 1 (7/84)

Pnrject repair work willbe performed in accordance with this regulatory guide o

~ ~

TABIZ 3. 2-2

'd)

RIGUIATORY GUIDE NUMBER 1.94 R1MUIAIORY GUIDE TENEZ twjl~

for Installation, Inspection, and Testing of Structural Qm~e and Structural Steel During the Const~ction Ehase of Nuclear Paver Plants RHGUIATORY GUIDE REVISION 1 (4/76)

Exceptions:

<< "Grout testing" (AFK C109) included in Table B of ANSI N45.2.5-1974 is inappropriate for field testing as it is a sophisticated laboratory test utilized for cement evaluation.

In lieu of daily tests, pre-packaged non-shrink grouts ~ be accepted for use on the basis of manufacturer's ~fication or campressive strength tests made in the field.

Confirmation compressive strength tests shall be made during the first day's production and thereafter on a basis of either once per day of every ane-hundred (100) bags used, whichever is least.

- Mater and ice shall be sampled and tested to ensure either potability or certified to contain not more than 2,000 parts per million of chlorides as Cl, nor more than 1,500 p ~ p mls of ~fat as S04 A~kahilityof this water or ice shall be per this certMication and preclude the AKH's referenced in Table B of ANSI N45.2.5-1974.

- %he reference, in Table B of ANSI N45.2.5-1974, to soft fragment testing per APIH changed designations to ASIA C851 which was deleted in 1985.

No testing for soft fragments is intended.

TABLE 3.2-2 tinued)

RHGUZAXORY GUIDE NUMBER 1.100

1. 116

1. 123

1. 131 E EGUIAIORY GUIDE TZIIZ Seismic Qualification of Electric Ecpupnent Important to Safety for Nuclear Power Plants t&1!11 for Installation, Inspection, and Testing of Meclanical Eguignent and Systems.

~ftf 1~

Items and SerVices for Nuclear Plants

- Qualification Tests of Electric Cables, Field Splices, and Connections for LightWater~oled Nuclear Palmr Plants EKGUIATORY GUIDE REVISION 1 (8/77)

Pmject repair work willbe performed in accordance with this regulatory guide.

0-R (5/77)

Emption:

Oammitted to ANSI N45.2.8 (1975)

"Supplementary Quality Assurance f

11'nspection and Testing of Mechanical Equipment and Systems for the Oanstruction Ruse of Nuclear Palm Plants" per ASAR, Section 1.7, "QAPD" Apgerx6x A.

Not committed to this regulatoxy guide.

1 (7/77)

Oammitted to in UFSAR, Section 1.7, nQAPDn, 3+)endix A 0 (8/77)

Pnrject field work willbe performed in accordance with this regulatory guide.

1. 144

1. 146 Auditing of Quality Assurance Pxogralns for Nuclear Power Plants Qualification of Quality Assurance Program Audit. Pmmonnel for Nuclear Pawer Plants 0 (+79) 0 (8/80)

Oammitted to in UFBAR, Section 1.7, QAPDn Appendix A.

Oamitted to in UFBAR, Section 1.7, nQAPDrt AppelxUx A

The replacement lower assemblies will be transported to the Donald C.

Cook Plant by barge/railroad combination.

They will be barged to Mt. Vernon,

Indiana, where they will be transferred to railroad cars for transportation by rail to the plant.

The lower assemblies will be drained, dried and sealed prior to shipment.

A nitrogen blanket will be maintained on the primary and secondary side during shipment and storage.

During transportation the assemblies will be supported on the barge/car deck on specially fabricated

saddles, tied down by cables and restrained by end braces secured to the deck.

Post Shutdown ActLvS.ti.es 3.4.1 Containment Pre arations 3.4.1.1 Reactor Vessel Prior to the start of repair project the reactor will be defueled.

The upper internals will be returned to the reactor vessel and the reactor vessel head reinstalled.

The missile shields will be reinstalled and a heavy steel work platform will be assembled over the refueling cavity.

Lay-up procedures to insure reactor vessel cleanliness, prevent foreign objects from entering the reactor vessel, and minimize corrosion of the reactor coolant system will be developed.

3.4.1

~ 2 Polar Crane The polar crane is equipped with a 250-ton capacity main hoist and 35-ton auxiliary hoist mounted on a single trolley.

The polar crane possesses sufficient capacity to handle all major lifting requirements for the steam generator project inside containment and can be rerated to a higher capacity as required; however, rerating of the hoists is not anticipated. Revision 0

Some circuits of the following systems will be temporarily disconnected and/or removed:

o Fire Detection Communication Steam Generator Process Instrumentation Containment Ventilation Fuel Handling Hydrogen Recombiner 600 V Non-Ess Dist. 6 120/208 V Lighting Seismic Instrumentation Equipment determined to be essential during the Steam Generator Repair Project will be relocated, and/or its cable,

conduit, and cable trays will be re-routed as required to maiptain the equipment in proper operating condition.

3.4.2.7 Heating, Ventilation and Air Conditioning Ductwork Ductwork in the removal pathway will be removed or temporary relocat'ed.

Duct pieces removed will be cleaned, marked and placed in temporary storage outside containment until needed for reinstallation.

3.4.2.8 Steam Generator Insulation The existing steam generator metallic insulation will be reused.

The outer dimensions of the replacement steam generators duplicates the original steam generators, although some insulation sections will require modifications to accommodate the additional hand holes and inspection ports.

Sections of insulation shall be removed,

cleaned, wrapped in plastic bags and stored in wooden crates.

Storage crates will be stored outside containment off the Revision 1

~ 4

ground and protected from the weather.

Sequence of removal and storage location will be documented to facilitate installation.

Those sections requiring modifications will be stored separately to allow rework prior to installation.

The original equipment supplier, Diamond Power Speciality Corp., will provide procedures and technical supervision for insulation

removal, storage, modifications and installation.

3.4.2.9 Seismic Restraints Removal The steam generator snubbers will be removed to provide access for handling and movement of the steam generators.

In addition, the pipe whip restraint at the main steam pipe will also be removed.

Removal and storage of the snubbers and restraints will be in accordance with approved procedures and/or specifications.

Snubbers are periodically removed for ISI testing and off-site disassembly and inspection by an independent laboratory.

Removal and reinstallation procedures will be similar to those established for the periodic inspections.

3.4.2.10 Fire Sensors Thermistor cable tray fire sensors will be pulled back where they extend beyond removed cable tray sections.

These sensor circuits will remain in service during the steam generator project and will be reinstalled in accordance with approved procedures.

3.5 Stcam Generator Removal ActM.ti.es 3.5.1 Steam Generator Cuttin Methods and Locations t

3.5.1.1 Feedwater and Main Steam Line Piping Cuts The feedwater and main steam lines will be mechanically cut in two places.

The location of the cuts, the equipment to be used, and the method of cutting Revision 0

After the lifting assembly is installed, the crane shall take the weight of the lower assembly while the lower assembly is still supported by the temporary lateral support and the steam generator support columns.

The temporary lateral support will be removed and the lower assembly then lifted slightly off its support columns.

The lower assembly shall be raised until the lifting assembly is approximately 2'-0" below the underside of the steam generator doghouse enclosure roof and then moved horizontally until it is within approximately 6 inches of the opening in the steam generator doghouse enclosure wall. It will be lifted again until the bottom of the lower assembly clears the horizontal wall cut.

It will then be moved horizontally out of the steam generator enclosure.

After clearing the steam generator doghouse enclosure a downending fixture will be attached to the steam generator lower assembly audit will be lowered onto a set of low profile saddles.

After the lower assembly has bden secured to t'e saddles and the saddLes have been placed on rollers, the upper assembly will be winched through the equipment hatch.

Once the lower assembly is through the Unit 2 equipment hatch and resting on the transport deck in the auxiliary building between the Unit 1 and Unit 2 equipment hatches, it will be attached to the tandem auxiliary building bridge cranes.

The lower assembly will then be lifted, rotated and moved in a southeast direction until it has passed the southwest corner of the spent fuel pool.

After the upper assembly has passed by the southwest corner of the spent fuel pool it will be oriented in an east-west direction and moved to the eastern edge of the elevation 650'loor.

At the eastern edge of the elevation 650'loor, the lower assembly will be moved out into the railroad t

bay and oriented in a north-south direction, lowered to the 609'levation and secured to a wheeled transporter.

The lower assembly will then be transported Revision 1

It

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to Pip SQ06,~

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to Pipe Bdn Sbean Pipe to Pipe Sh-355,CL-1 Gr. IC70 (SMBl,CL-32 Ge CKR}

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ma to p1abe CL-1

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Intenal Rxa Prize Cnpamts 175 3/4n 3 62 r ~ U gag hx9zprp cx cx SN Sinijle U withhx3ug Rmxe hx9drg RR9gxxp 124.ZP 3~~

ange V em flatroot EKHS B3 Z~lrr's Qxdce 5x'N F00-3200 Gard fee 2hr30ndn UZemn Raa 800 hat & cooL AD/he HP,UZ,PZ cx MP 50 Rt ag.

Cd', Blair MP 1

2 v

3 I@id ~i! ~~~~1l&h Mthth~ f Outside diameter mm~ as noted.

Meld fillermetals and electrodes to be ordered in accordance with ASME Code Section II, Fart C.

Austenitic stainless steel to meet delta ferrite rec{uirements in ASME Code Sectian III,NB-2433.

Oovered electrodes to meet analysis tests of ASME Code Section III, NB-2420.

NDE to be in accordance with ASME Sectian V with acceptance staixlards in accordance with ASME Code Section III.

~ ~

In addition, a Plant/Project interface document shall be implemented to define areas of responsibility, communications,

control, and interface between teh Project Radiation Protection/AIBA Group and the Plant Radiation Protection Section.

Regular meetings between members of these two groups will be held to insure adequate communications and dissemination of information.

- 82a-Revision 1

o No changes are expected due-to differences in initial conditions (zero load steam temperature and pressure are identical for the unit with repaired steam generators).

The no load steam generator mass decreases insignificantly (-2.0 percent)

Therefore the conclusions of the existing steam line break analyses remain valid for the repaired steam generators.

6.1.2.5 Steam Syst: em Piping Failures

~ r Refer to Section 6.1.2.4 for discussion that applies to this accident as well.

6.1.2.6 Loss of External Load Donald C.

Cook Unit 2 is designed to have full load rejection capability, and a reactor trip may not occur following a loss of external load. It is expected that steam dump valves would open in such a load rejection, dumping steam directly to the condenser.

Reactor coolant temperature and pressure do not significantly increase if the turbine bypass system and pressurizer pressure control system are functioning properly. If the steam dump valves do not operate, the reactor will trip due to high pressurizer signal, high pressurizer level signal, or overtemperature 6 T signal.

Primarily to show the adequacy of the pressure-relieving devices and to demonstrate core protection margins, the Donald C.

Cook FSAR and analysis of record analyze cases where the steam dump valves do not operate, and there, is no direct reactor trip due to a turbine trip. It 'is shown in the FSAR and the analysis of record that the accident criteria on syst: em pressure and DNB are not violated in any of the loss-of-load cases.

- 144-Revision 0

An accident involving the dropping.or tipping of the steam generators during the removal process is considered highly unlikely because of the strict I

controls which will be placed on the movement process.

In the unlikely event that an accident involving the steam generators does occur, our reviews have determined that the only potential interactions with shared systems of significant concern involve the spent fuel pool cooling equipment located in the vicinity of the load path.

However, the slight potential for damaging spent fuel pool cooling equipment is not considered to represent an unreviewed safety question as defined in 10 CFR 50.59.

This conclusion is based on the various malfunction analyses presented in Chapter 9.4 of the FSAR.

These analyses conclude that-it is not possible for a piping failure to cause drainage of the pool below the top of the stored fuel elements.

In the event all cooling for the pool is lost, it would take a minimum of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for the temperature in the pool to each 180 F (which still allows 32 F margin to 0

0 Thus, sufficient time exists to either restore cooling capability or replace water which could be lost through boiloff to prevent damage to the stored fuel elements.

6.3 Analysis of Significant Hazards Consideration This section presents, pursuant to 10 CFR 50.91, the analysis which sets forth the determination that the Steam Generator Repair Project does not involve any Significant Hazard Consideration as defined by 10 CFR 50.92.

In addition to the appraisal on the significant hazards issue using the 0

standards in 10 CFR 50.92, which are presented below, it is important to note that the Steam Generator Repair Project proposed by I&MECo involves practices 1

that have been successfully implemented at two other commercial nuclear power

plants, namely, the steam generator repairs completed by the Virginia Electric

- 163-Revision 0

and Power Company for the Surry Power Station and by the Wisconsin Electric Power Company for the Point Beach Nuclear Plant, Unit 1.

The repair project is also similar to the repair projects conducted by the Carolina Power and Light Company for the H. B. gobinson Steam Electric Plant, Unit No.

2 and by the Florida Power and Light Company for the Turkey Point Plant Units 3 and 4.

Involve a significant increase in the probability or consequences of an accident.

The Steam Generator Repair Project does not affect the probability or consequence of an accident.

The probability or consequence of an accident is determined by the design and operation of plant systems.

The repair project involves the replacement of the Donald C.

Cook Unit 2 Steam Generator Lower Assemblies.

Due to the almost identical design of the replacement lower assemblies the repair of the Donald C.

Cook Unit 2 steam generators is a replacement in kind and will not change the design or operation of plant systems.

Thus, this repair does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Create the possibility of a new or different kind of accident from any accident previously evaluated.

The possibility of a new or different kind of accident is not created by the repair to the Donald C.

Cook Unit 2 steam generators.

All components and l

piping will be reinstalled to meet the original design and configurations and installation requirements.

Therefore, because there will be no changes to the plant and plant systems design no new or different accidents are created.

- 164-Revision 0

6.3.3 Criterion 3

Involve a significant reduction in a margin of safety.

Section 2.2 of this report illustrates that, although certain design enhancements have been

made, the steam generator repair will result in very little change to the original operating parameters.

Therefore, the impact on the accident analysis, as shown in Section 6.1 will be insignificant and there will be no significant resolution in the margin of safety.

- 165-Revision 0

7.3.2 e ional Historic Archeolo ical Architectural Scenic Cultural and Natural Features No known historic, archeological, architectural or natural resources exist on the portion of the plant site affected by the Steam Generator Repair Proj ect.

The access road used during plant construction parallels the beach and will be used for light construction traffic during the repair project.

This traffic may pose an aesthetic impact to individuals using the beach for recreation, however, this is a temporary impact that will end with the completion of the repair project.

7.3.3 gydro~lo y 7.3.3.1 Ground Water No impact to the site ground water is expected to occur as a result of the Steam Generator Repair Project.

7.3.3.2 Surface Water No impact to the surface water associated with the plant site is expected to occur as a result of the construction phase of the Steam Generator Repair Project.

In addition, the repaired steam generators will have essentially the same amount of blowdown discharged during operation as do the original steam generators and it anticipates that there will be no changes to the plant NPDES permit.

- 170-Revision 0

7.3.4

~Geolo There will be no geological impacts as the result of the Steam Generator Repair Project.

Excavation,

grading, and compaction will occur in limited amounts and these actions will occur in areas previously disturbed (i.e. parking lots,

roadways, and laydown areas).

7.3.5 Ecol~oy 7.3.5.1 Terrestrial Ecology There will be no impacts to the terrestrial ecology surrounding the plant site for the following reasons:

o No habitat will be removed as a result of the Steam Generator Repair Project since all activities related to the repair project will occur on previously disturbed area (i.e. existing access

roads, parking lots, laydown area.

o Since the area affected is already subjected to the intrusion of man and machinery (i.e. security patrols, existing security lights, and normal plant operations),

animals residing in the areas adjacent to the construction related activities should not be disturbed by the increased activity.

7 '.5.2 Aquatic Ecology As discussed in Section 7.3.3.2 neither the construction phase of the Steam Generator Repair Program and the operation of the repaired steam generators will not impact the aquatic ecology associated with the plant site.

Revision 0

TABLE 7.4-1 DONALD C.

COOK PER UNIT AVERAGE ANNUAL MAN-REM EXPENDITURES YEAR Exposure Man-rem 1980 246 1981 327 1982 321 1983 283 1984 1985 344 448

- 173-Revision 0

TABLE 7.4-6 COMPARISON OF GASEOUS EFFLUENT RELEASES FROM DONALD C.

COOK NUCLEAR PLANT Radioactive

~Secre Noble gases Iodines Particulates Tritium Average 1985 Release/Unit C

2.47 x 10 6'6 x 10 3.72 x 10 Estimated Release During the SG Repair Effort Ci Negligible 9 x 10-6 (1) 2.92 x 10 Negligible Notes (1) Estimated from Surry Unit 2 Data.

- 183-Revision 0

7.9 Emrkxoxaaental. Control The following environmental controls shall be utilized.to minimize the environmental impacts associated with the steam generator repair program.

These environmental controls shall be reviewed by the contractor prior to the start of work.

In addition, it, is recommended that these environmental controls be included as part of the contractor work specifications.

7.9.1 Noise To reduce the impact of.noise on the surrounding community, the majority of the construction activities involving the use of heavy machinery will take place only during the day shift. If second shift construction activity involving heavy machinery must occur, it will end by 9:00 p.m.

Noise from internal combustion engines will be controlled by the use of exhaust mufflers.

7.9.2 Limitations of Machiner Movement No machinery will be allowed to operate in areas not previously distributed by construction activities. If areas not previously disturbed are inadvertently impacted by machinery, it will be the responsibility of the contractor operating the machinery to restore the disturbed area to its original state.

7.9.3 Handlin and Stora e of Oil and Pollutin Materials The handling and storage and oil and polluting materials will be conducted in accordance with the D.

C.

Cook, "Oil Spill Prevention Control and Countermeasure Plan,"

and the D.

C.

Cook, "Pollution Incident Prevention Plan."

- 202-Revision 0

7.9.4 nvironmental Monitorin Periodic inspections of the construction activities will be conducted.

If any of the construction activities appear to be causing significant environmental

impacts, appropriate actions will be taken.

7.9.5 Permits A list of State and local permits needed to begin construction activities at D.

C.

gook will be developed by the D.

C.

Cook Environmental Section and the AEPSC Radiological Support Section.

However, it will be the responsibility of the contractor to obtain the required permits.

7.10 Conclusion It is concluded that with the proper mitigation practices as outlined in the Environmental Controls Section of this report, no significant adverse environmental impact will result from the proposed activity, that there are no preferable alternatives to the proposed action and that the impacts associated with the repair program are outweighed by its benefits.

It is further concluded that the site preparation

work, as described in Section 3,

does not involve an unreviewed environmental question pursuant to Part II, Section 3.1 of the Donald C.

Cook Plant Environmental Technical Specifications.

- 203-Revision 0

I r

r~,'I Y p I

~-"~5~~'

D.

C.

COOK PLANT UNIT NO ~

2 STEAM GENERATOR REPAIR REPORT SUPPLEMENT 1 TABLE OF CONTENTS

~SECT ON 1.2 GENERAL EVALUATIONS

~fLFc

~PGE 1-3 1-3 1.F 1 1.2.2 1.2.3 1.2.4 1.2.5 Crane Manufacturer and Design-Rated Load Comparison to NUREG-0554 and NUREG-0612 Seismic Analysis Lifting Beams Interfacing Lift Points 1-3 1-3 1-14 1-17 1-18 1.3 CONCLUSION LIST OF TABLES

I~ABL 2.2-1 2.2-2

~ITLE 150-Ton Capacity Single-Failure-Proof Crane Design Factors Steam Generator Repair Project Auxiliary Building Crane Lifts Over 60 Tons

~PGE 1-5 1-7 g'TGGg+E 1.2-1 LIST OF FIGURES

~TTTL Mathematical Model of Crane Trolley at Mid Span gAGE 1-2 Rewision 1

~ ~

design, fabrication, inspection, testing and operation as delineated in NUREG-0554 and supplemented by NUREG-0612.

This evaluation is presented in the form of a point-by-point comparison to NUREG-0554 which was developed by AEPSC and Whiting Corporation.

The new crane will meet all applicable sections of CMAA Specification ¹70, Revision 75 and ANSI B30.2.0

- 1967.

For ease in making a point-by-point comparison the following section numbers correspond to the section numbers in NUREG-0554:

2.

SPECIFICA 0

AND ESIGN CRITER 2.1 Const uct n and 0 erat n Periods Since the Donald C.

Cook Nuclear Plant is an operating

plant, the construction portion of this section is not applicable.

For the repair project and subsequent operating period the new crane will be designed pei CMAA

¹70, Revision 75.

Dynamic loads are considered due to load accelerations associated with a 150-ton load but not seismic loadings.

Simultaneous static and dynamic loading will not stress the equipment beyond the material yield.

2.2 a

imum C

tica Load Since the new crane will be operating indoors, degradation due to exposure will not be considered a factor in the crane design.

However, items subject to wear will have an additional design factor applied to them (see Table 2.2-1 of this supplement).

1-4 Revision 1

2.2 aximum Cr t ca pads (cont'd.)

The crane is being designed per CMAA ¹70, Revision 75 for dynamic loads due to the load accelerations associated with 150 ton load.

Considering dynamic loads due only to load accelerations, the maximum critical load is 150 tons.

However, as presented in the preliminary seismic analysis discussion, Section 1.2.3, when dynamic loads due to a seismic event (safe shutdown earthquake) are applied to the crane the maximum critical load is 60 tons.

A maximum critical load of 60 tons is sufficient for all but 24 lifts associated with the repair project.

Because these 24 lifts are one time only special lifts the provisions of NUREG-0612 Section 5.1.1(4) will apply.

This section states that for special lifts, loads imposed by the safe shutdown earthquake need not be included in the dynamic loads imposed on the lifting device.

Therefore, for these 24 special lifts the maximum critical load will be the same as the design rated load of 150 tons.

The design rated load and the maximum critical load will be marked on the crane.

1-6 Revision 1

~ ~

S

~ ~

TABLE 2.2-2 STEAN GENERATOR REPAIR PROJECT AUXILIARYBUILDING CRANE LIFTS OVER 60 TONS tern Est.

Wt.

~owns Number

~ifts Steam Generator Concrete Doghouse Front Roof Section 70 Steam Generator Concrete Doghouse Back Roof Section 60 Old Steam Generator Upper Assembly 112 Old Steam Generator

• Lower Assembly 247 New Steam Generator

• Lower Assembly 240 Refurbished Steam Generator Upper Assembly 112 24 Total

• These lifts will be made using the upgraded existing crane and the new crane tandem.

1-7 Revision 1

TABLE 2.2-2 STEAN GENERATOR REPAIR PROJECT AUXILIARYBUILD %G CRANE LIFTS OVER 6D TONS Est.

Wt.

LXmal Number Lif~'team Generator Concrete Doghouse Front Roof Section 70 Steam Generator Concrete Doghouse Back Roof Section 60 Old Steam Generator Upper Assembly 112 Old Steam Generator

• Lover Assembly 240 Nev Steam Generator

• Lover Assembly 247 Refurbished Steam Generator Upper Assembly

= 112 24 Total

• These lifts vill be made using the upgraded existing, crane aad the nev crane in a tandem configuration.

Re sion 2

4 0

2.3 0 eratin E vi nment Since the crane will be operated in the auxiliary building the crane will not be subjected to design basis accident type changes in pressure, temperature, humidity or exposed to corrosive or hazardous conditions.

Therefore, such considerations have not been included in the design of the crane.

As discussed in the following section, a minimum operating temperature will be determined.

2.4 Material Pro erties In addition to impact testing requirements on the main hook, structural members essential to structural integrity and greater in thickness than 5/8 inches are fabricated of impact tested material in accordance with the Section III of the ASME code.

The minimum operating temperature of the crane will be established by the crane manufacturer.

Any necessary steps to prevent operation of the crane below the minimum operating temperature will be taken.

In

addition, low alloy steels are not used in the fabrication of the crane, and cast iron is restricted to non-load bearing components.

2 '

Se ic Desi See Section 1.2.3.

2.6 The main bridge girders and structural load support members of the trolley, specifically those members supporting the critical load, are fabricated from structural plate.

Welded, rolled structural shapes are not used for these members'oreover, weld joints associated with the structural members within the main hoist load path are typically oriented such that the induced stresses will not be manifested in lamellar tearing at the weld zone.

All weld joints whose failure could result in the drop of a critical load will be nondestructively examined.

If any of these weld joint geometries would be susceptible to lamellar tearing, the base metal at the joints will be nondestructively

examined, 2.7 St ctural at e

As stated in Section 2.1, the crane will not be used for plant construction lifts.

A fatigue analysis will not be performed on the structures of this crane nor does it seem reasonable that the results of such an investigation would prove meaningful.

Designing for endurance in consideration of cyclic loading and material fatigue 1-8 I

Revision 1

limits has generally not proven to be governing in overhead crane design.

Moreover, the fatigue stress level of materials is typically beyond normal design stress allowables.

2.8 Weldin Procedures Welding, welding procedures (pre heat, post weld heat treatments),

and welder qualifications are in accordance with AWS Dl.l "Structural Welding Code."

SAFETY FEATURES 3.2 Auxiliary Systems The auxiliary hoist is of single-failure-proof design.

Where dual components are not provided within either hoist mechanical load path, redundancy is provided through an increased design factor on such components as required per NUREG-0612.

3.3 Electric Control S stems Limit controls are incorporated to minimize the likelihood of inflicting damage to the hoisting drive machinery and structure that otherwise might occur through inattentive and/or unskilled operator action.

An emergency stop button will be added to the control pendant that will interrupt the power supply to the crane and stop all crane motion.

3.4 Emer enc Re airs This crane is designed so that, should a malfunction or failure of controls or components occur, it will be able to hold the load while repairs and adjustments are made.

4.]

HOISTING MACHINERY Reevin S stem The static-inertia design factor of the wire rope, with all parts in the dual system supporting the DRL is 11 to 1.

Such conservative design more than surpasses requirements to sustain the dynamic effects of load transfer due to the loss of one of the two independent rope systems with an ample design margin remaining in the six parts supporting the load.

Compliance to this 1-9 Revision 1

recommendation requires high alloy rope.

By definition, reverse bends do not exist in the reeving system of the main hoist.

Studies have been conducted to establish the effects of reverse bend on fatigue life.

In consideration for the geometry of wire rope (helix) construction, unless the distance between the sheaves in the load block and head block are under one lead of the wire rope, a reverse bend cycle is not incurred.

Moreover, the ratio of rope to sheave diameter in the only qualifying area of the hoist mechanism is related to the drum, which is 30 to 1; 125$ of minimum requirement per CMAA Spec.

870, Rev.

75.

The pitch diameter of running sheaves and drums shall be in accordance with CMAA Spec.

470, Rev.

75.

All fleet angles within the main hoist reeving are within the recommended 3 1/2 degrees.

The crane is equipped with an equalizer beam/fixed sheave arrangement that provides two separate and complete reeving systems.

4.2 m

u o t The indicated drum support provisions are included in the design which, as required, would insure against disengagement of the drum from its braking control system.

4.3 ead a d ad Blocks Both reeving systems associated with this crane are designed with dual reeving.

This design will ensure the vertical load balance is maintained.

Each load-attaching point (sister hook and eye bolt) is amply designed to sustain 2008 of the 150-ton DRL.

The overhead crane shall be load tested at 1258 of the 150-ton DRL.

Nondestructive examination of the sister hook and eye bolt will be performed.

After successful completion of the

load test, a complete inspection of the crane, including a nondestructive examination of the sister hook and eye bolt, will be performed.

4.4 The main hoist full rated load speed at 4.5 FPM is considered to be "slow" for this rated load.

Further, the rope line speed at the drum at approximately 27 FPM is considered to be conservative.

4.5 esi A ainst Two-Block n The main hoist is equipped with two independent travel limit control devices in addition to a load sensing

system, as suggested, to insure against two-blocking.

Actuation of hoist travel limit switches or load sensing devices will deenergize the hoist drive.

In addition, Revision 1

the mechanical holding brake will have the capability to withstand the maximum torque of the driving motor.

4.6 Li tin Device Lifting devices for attachment to the main hook will meet or exceed these specified requirements.

4.7 Mire Ro e Protection Operation of the hoist is only to be attempted with the trolley and block aligned over the center of the load for a vertical lift.

4.8 Machiner Ali nment The provisions of this paragraph are incorporated in the design of the overhead crane.

4.9 Hoist Brakin S stem The provisions of this paragraph are incorporated in the design of the overhead crane.

BRIDGE AND TROLLEY 5.1 Brakin Ca acit The bridge and trolley drives will each be provided with an appropriately sized electric holding brake which, upon interruption of power, is applied whether through operator action or violation of travel limit provisions on the trolley and restrict area limit controls for the bridge.

Further, these brakes are capable of being operated manually.

The AC induction-motors and magnetic controls utilized for these drives are not prone to an overspeed condition, which is attributed.to inherent operating characteristics.

Therefore, overspeed limit controls for the bridge and trolley motion equipped with this type of drive would represent a needless feature.

Moreover,'the motor controls are provided with adequate overload protection.

The mechanical drive components are designed to sustain maximum peak loadings capable of being transmitted by either the motor or brake under all attitudes of normal crane operation.

All other recommendations of this section are compatible with the design of the crane.

1-11 Revision 1

5.2 Safe Sto s

As stated in Section 5.1, an overspeed condition considering the type of drive used for the bridge and trolley is not a concern with this equipment.

Appropriately designed and sized bumpers and stops are provided in accordance with CMAA Spec.

870 Rev.

75 and are adequate to absorb the energy of the trolley and bridge in the event of limit switch malfunction.

6.

RIVERS AND CO OLS 6.1 ive Se ectio The main hoist motor was selected on the basis of hoisting the design-rated load (150 tons) at the design hoisting speed.

Further, all proper and due consideration was given to the design of related mechanical and structural components to adequately resist peak torques transmitted by this motor within normal design limits.

Hoist overspeed and overload sensing-limit control provisions have been incorporated to guard against such occurrences.

Additionally, the hoist holding brakes are capable of controlling the design rated load within the 3

inches (8 cm) specified stopping distance.

In addition, an emergency stop button will be located at ground level to interrupt power to the crane independent of the crane controls.

Since the MCL is less than the DRL, administrative controls will be established to reset the overloading sensing device.

6.2 rive Cont ol S stems The design considerations discussed in this section have been addressed and incorporated as appropriate except for the restriction of simultaneous operation of motions.

The crane is not used to handle spent fuel assemblies.

6.3 funct o P otectio Features to sense, respond to, and secure the load in the event of hoist overspeed, overcurrent,

overload, over

travel, and loss of one rope of the dual reeving system have been incorporated.

6.4 Slow S eed D ives Features recommended in this paragraph will be incorporated as part of the motion control circuitry.

6.5 Safet evices Each hoist is equipped with two independent hoist overtravel limit controls.

1-12 Revision 1

~>

6.6 Cont o

Sta ons Since this crane is not equipped with a cab, the complete operating control system and emergency controls for the crane will be located on a pendant control.

In addition, as stated earlier an emergency stop button will be located at ground level to interrupt power to the crane independent of the pendant control.

Since the design rated load is greater than the maximum critical load, administrative controls will be established to ensure that the resetting of the overload sensing device is properly conducted.

7.

S IO NS UCTIO S 7.1 Gene a

Complete operation, maintenance, installation and testing instructions will be provided for the overhead crane by the crane manufacturer.

7.2 Co st etio and 0 eratin Pe ods As discussed in Section 2.1 this crane will not be used for plant construction.

The crane will be designed for Class A-1 service as defined in CMAA Specification 870, Revision 75.

The allowable design stress limits will not be exceeded during the repair project.

During and after installation of the crane, the proper assembly of electrical and structural components should be verified.

8.

ST NG AND P EVE VE I

EN CE 8.1

~Gne~a'3, A complete check will be made of all the crane's mechanical and electrical systems to verify the proper installation and to prepare the crane for testing.

The only components that will have been proof-tested at the time of installation are the main hook, eye bolt and wire rope.

8.2 tatic a d amic Load Tests The crane will be load tested at 125% of the design rated load.

The design rated load of this crane is 150 tons.

During the 1258 load test, the crane motions shall be limited due to the physical restrictions of the auxiliary building.

During the no-load test,

however, each crane motion shall be operated to its full travel limit.

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8.3 Two-Block est Although the hoist is equipped with an overload sensing

device, under no circumstances should such a test be conducted for the mere purpose of demonstrating design adequacy.

The purpose in providing numerous limit control devices is to ensure against such an occurrence.

The two load travel limit control switches will be checked prior to lifting a load.

The overload sensing device can be operationally checked within the design rated load of the crane without the need to secure the hoist to a fixed anchor for the purpose of generating an excessive load.

8.4 0 e at on ests Whiting's standard procedures require a no-load running test before shipment.

Calibration and adjustments for hoist overload and overspeed will be done after installation.

8.5 a

tena e

A maintenance program including periodic inspections of the crane will be developed.

This maintenance program will ensure that the crane is maintained at the design rated load.

Both the maximum critical load and the design rated load will be plainly marked on each side of the crane.

9.

OP RATING MANUA Whiting's standard Operations and Maintenance

Manual, which is to be provided for the overhead crane, will provide sufficient information in the proper operation of the overhead crane, lubrication instructions, parts ordering information, and periodic inspection points.

10.

UALITY ASSURANCE The Whiting Corporation is on the Donald C.

Cook Nuclear Plant Qualified Suppliers List for spare and replacement crane parts.

Whiting has a QA program that complies with ANSI N.45.2-1971/NRC Regulatory Guide 1.28.

This program applies also to the fabrication of new cranes for nuclear power plants.

Whiting will be audited for QSL recertification in April 1987.

Donald C.

Cook Nuclear Procedure MHI 2071, "Qualification and Training of Crane Operators,"

covers qualification requirements of crane operators and will be revised as necessary to reflect the single-failure-proof features of the new crane.

1.2.3 Seisaic Analysis This section presents the preliminary seismic analysis conducted to demonstrate the largest load the new crane 1-14 Revision 1

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can stop and hold during a safe shutdown earthquake.

The following information provides a description of the method of analysis, the assumptions

used, and the mathematical model evaluated in the analysis.

hna1ysia Description The crane was analyzed to determine the effect of seismic excitations.

For this analysis, the matrix displacement method was used based upon finite element techniques.

The crane was mathematically modeled as a system of node points interconnected by various finite elements representing straight beams.

All masses and inertias were distributed among the nodes whose degrees of freedom characterize the response of the structure.

The interconnecting finite elements were assigned stiffness equivalent to that of the actual structure.

The mathematical model represents as accurately as possible the flexibilityof the bridge girders, hoist

rope, and girder end connection.

The trolley, the drive units and the bridge trucks were represented as rigid bodies.

The crane was analyzed with the trolley positioned at mid-span.

This was done with loads of 50 and 60 tons in the down position.

Preliminary calculations showed that this condition would produce the maximum girder stress for a given load.

The dynamic analysis was of the mode frequency (MODAL) type, solving for the resonant frequencies and the mode shapes that characterize the crane.

The modes with meaningful participation in a given direction are directly expanded by the computer program to yield the expanded mode shapes, the element stresses azid the reaction values.

This type of analysis is linear and plastic deformation, sliding, friction, and slack rope are not taken into account.

The normal mode approach was employed for the analysis of the components.

All significant eigen-values and eigen-vectors were extracted, and these modes were combined by the method specified by the U.

S. Nuclear Regulatory Commission, Regulatory Guide 1.29, Rev.

1, Section 1.2.2 (Combination of Modal Responses with Closely Spaced Modes by the 108 Method).

Those modes with mode coefficient ratios less than 1$ in the x direction or 0.5%

in the y and z directions were dropped because their contribution is proportionally small when compared to the largest mode coefficient of the related directional excitation.

The results of the three orthogonal dynamic excitations were combined by the square root of the sum of the squares method (SRSS) and then absolutely added to the results of the static condition.

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Because the y reaction exceeds the frictional resistance of those bridge wheels that are braked, slip will occur.

The maximum acceleration in the y direction will be reduced from that predicted by the modal analysis.

The primary y mode was therefore reduced by a scale factor such that the resulting y reaction approaches the maximum that could be sustained before slip.

The results were then resummed as previously described.

In order to assure structural integrity, the job specification requires that the maximum stresses not exceed the minimum yield strength of the material divided by 1.5 for the OBE and 1.1 for the SSE.

The crane is constructed of ASTM A36 structural steel except for components which are specifically noted in the report.

A36 material has a specified minimum yield strength of 36 ksi.

The combined bending and axial stresses are limited to 24 ksi for the OBE and 32.7 ksi for the SSE.

The actual properties of the specified materials show a great deal of variation and are generally considerably higher than the minimum required by the material specification.

Also the maximum stresses occur only at a point on a section and cannot be themselves be indicative of the tendency of the section to permanently

deform, especially when the nominal stresses on the extreme fibers of the adjoining faces are significantly lower. It is therefore conservative to compare the combined bending and axial stresses at the corners with the specified allowables to assure structural integrity.

Impact factors for wheel flange to rail contact, etc.,

have been consider negligible.

The state of the art is such that these impacts cannot rigorously be studied;

however, independent time history analyses have been run in many cases, all indicating slow relative motion between the rail and the wheel.

This is because of the time dependency of the forcing function coming from the building into the crane.

Note that the only coupling through which these forces can be. transmitted is dynamic friction.

Upon reaching the rail the wheel will first rise through the corner radius and then contact the rail.

During this period, the structure is starting to deflect as the end of the crane in this direction is flexible.

The computer analysis was performed using ANSYS, a large scale finite element program.

S~zy of Remalts The crane was mathematically modeled using finite elements.

On the basis of preliminary runs, the number of degrees of freedom and the significance criteria for modal expansion were adjusted.

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c, reduced modal runs were made and the results summed.

Because slip occurs, the y excitation was proportioned and these results resummed.

The crane was,analyzed with the main trolley at mid-span (see Figure 1.2-1).

For this position the analysis was done with 50 and 60 ton loads on the main hook in the low position.

From preliminary studies, the load case considered should yield the maximum stresses in the girders.

Because of the seismic acceleration a slack rope condition was found to exist under certain conditions.

This cannot be truly simulated with a linear modal analysis.

However our experience with time history analyses shows that a modal analysis tends to produce conservative results.

The rope load predicated by the modal analysis is well below the allowable rope load.

When the excess dynamic rope load (that which produces a

slack rope) is deducted, a small upkick is produced by the loading conditions examined.

When the wheel loads

. parallel to the runway are compared with the vertical wheel load times the coefficient of friction, it is found that the crane bridge will tend to slide under certain loading conditions examined.

This sliding is oscillatory in nature and the loadings predicted by a modal analysis are conservative.

The ~heel loads have been adjusted to account for frictional effects.

Although some non-linearities are produced by the specified excitations the specified linear analysis will conservatively predict the behavior of the crane during a seismic excitation.

The crane was found to meet the requirements for a seismic excitation with a 60 ton load on the main hook.

LLX~Seams Stress levels of all load-bearing members of the lifting beam will not exceed 6,000 psi under rated load.

This low stress level meets requirements of NUREG-0612 and ANSI N14.6 specifications for increased design factors for single-load-path components.

Further, this design stress level qualifies for material test exemptions per Paragraph AH 218 of the ASNE Boiler and Pressure Uessel

Code,Section III, Division 2, as referenced in Paragraph 3.3.6 of ANSI N14.6-1978.

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Proposed lifting beam will not be subject to high amounts of radiation, 200 mili-rem/hour maximum, nor will it be submerged at any time.

Based on this criteria the proposed lifting beam design will not be subject to any sections of ANSI N14.6-1978 which refers to submerged duty, decontamination or radiation degradation.

Application of any coating system onto the lifting beam must not violate ED P.A. codes.

Under Section 6 of ANSI N14.6-1978 the main beam section and the hooks swivel are single path designed with stress levels below 6,000 psi.

Since the materials for these items will have mill certification and that 100$ of critical welds will undergo nondestructive examination to ensure structural integrity, these two items will not be subject to load test of three times their rated capacity.

These two items will however be subjected to a 150% load test.

Interfacing Life Points Interfacing liftpoints will be dual-load-path and will be designed to.shear stress levels not to exceed 4,500 psi under rated load.

This design stress levels qualifies for material test exemptions per Paragraph AM 218 of the ASME Boiler and Pressure Vessel

Code,Section III, Division 2 as referenced in Paragraph 3.2.6 of ANSI N14.6-1978.

COHCMSIOH The new crane being purchased by the Indiana

& Michigan Electric Company for use during the Steam Generator Repair Project has been evaluated against the criteria of NUREG-0554 and NUREG-0612.

Results of this evaluation have shown that the crane being purchased meets the guidelines and criteria of NUREG-0554 and NUREG-0612 and therefore will be classified and used as a

single-failure-proof crane.

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o No changes are expected due to differences in initial conditions (zero load steam temperature and pressure are identical for the unit with repaired steam generators).

The no load steam generator mass decreases insignificantly (-2.0 percent).

Therefore the conclusions of the existing steara line break analyses remain

'alid for the repaired steam generators.

6.1.2.5 Steam System Piping. Failures Refer to Section 6.1.2.4 for discussion that applies to this accident as veil.

6.1.2.6 Loss of External Load k

Donald C.

Cook Unit 2 is designed to have full load rejection capability, and a reactor trip may not occur follo~ing a loss of external load. It is.

expected that steam dump valves auld open in such a load rejection, dumping steam directly to the condenser.

Reactor coolant temperature and pressure do no significantly increase if the turbine bypass system and pressurizer pressure control system are functioning properly.. If the steam dump valves do not operate, the reactor vill trip due to high pressurizer pressure

signal, high pressurizer level signal, or overtemperature T signal.

Primarily to shov the adequacy of the pressure-relieving devices and to demonstrate core protection margins, the Donald C.

Cook FSAR and analysis of record analyze cases vhere the steam dump valves do not operate, and there is no direc reactor trip due to a turbine trip. It is shown in the FSAR and the analysis of record that the accident criteria on system pressure and D?iB are not

ivlated in any of the loss-of-load casos.

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