ML17317A864

LER 78-101/01T-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

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: ML17317A864 ML18095A006
References
LER-78-101-01T, LER-78-101-1T, NUDOCS 7901150092
Download: ML17317A864 (116)

Text

NRC FORMi366 i U. S. NUCLEAR REGULATORY COMMISSION t7 %7)

LlCENSEE'VENT REPORT

{PLEASE PRINT OR TYPE ALL REQUIRED INFORMATION)

CONTROL BLOCK:

,Q 7 8 9 M'.-I n LICENSEE CODE 2

14 Q2 15 LICENSE NUMBER 25 Q 3 26 LICENSE TYPE 30 0~Q 57 CAT 58 CON'T

~O SO~U~RCE MLQ QB 0 I 0 5 7 9 QB 68 69 EVENT DATE 74 75 REPORT DATE 80 7 8 60 61 DOCKET NUMBER EVENT DESCRIPTION AND PROBABLE CONSEQUENCES QIO WITH THE UNIT IN MODE 1 OPERATION BOTH CONTROL ROOM PRESSURIZATION FANS WERE

~O3 DETERMINED TO BE INOPERABLE CONTRARY TO THE REQUIREMENTS OF TECH. SPEC. 3.7.5.1.

~O4 ONE FAN WAS RETURNED TO OPERABLE STATUS IN 56 MINUTES AND THE OTHER FAN IN 62

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

~OB

~07

~OB 80 7 8 SYSTEM CAUSE CAUSE COMP. VALVE CODE CODE SUBCODE COMPONENT CODE SUBCODE SUBCODE

~OB 9

A 10 Q>t A QTs ~E.

12 Q1s 13 0 K T 8 R K 18 Q14

~Ebs 19 LZJ 20 Qs 7 8 11 SEQUENTIAL OCCURRENCE REPORT REVISION LER/RO Rssos~

ACTION FUTURE EVENT YEAR

~78 21 22 EFFECT

~23 SHUTDOWN

~10 24 REPORT NO.

1 26

~w:

27

~01 28 ATTACHMENT CODE 29 NPRDC TYPE

~T 30 PRIME COMP.

31 NO.

0 32 COMPONENT TAKEN ACTION ON PLANT METHOD HOURS Q22 SUBMITTED FORM SUB. SUPPLIER MANUFACTURER

~Q>8 33 LJQ 34 LJQss 35 MQ 36 37 0 0 0 0 40 41 T Qss ~NQ24 42

~ZQss 43 44 A 1 9 8 Qs 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.

80

~Q 7 8 9 FACILITY STATUS  % POWER NA'ETHOD OTHER STATUS QBO DISCOVERY OF

. DISCOVERY DESCRIPTION Q32 EK LJQ E 2B 1 0 0 29 44 LAJQ>>

45 46 ALARM OF FIRE SPRAY OPERATIONS 80 ACTIVITY CONTENT RELEASED OF RELEASE AMOUNT OF ACTIVITYQss LOCATION OF RELEASE Qs 6 Z Q33 ~ZQ34 NA NA 8 9 10 44 45 80 7

PERSONNEL EXPOSURES NUMBER TYPE DESCRIPTION Q39

~l7 ~00 0 Qs7 ~ZQss 13 NA 80 7 8 9 11 12 PERSONNEL INJURIES 8 9 11 12 80 7

LOSS OF OR DAMAGE TO FACILITY DESCRIPTION Q43 TYPE s 2 Qss NA 8 9 10 80 7

P U 8 L I C I TY NRC USE ONLY 71O(/5 17171Z 68 69 80 ss 7 8 9 10 R. S. Lease PHONE 616 465-5901 X-313 NAME OF PREPARER

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.

It RECORD OF REVISIONS / Qp~.~@4

/o pc// gp REVISION REVISION DATE ISSUED NO. DATE ISSUED BY November 4, 1986 November 4, 1986 T. G. Harshbarger March 30, 1987 March 30, 1987 T. G. Harshbarger July 24, 1987 July 24, 1987 T. G. Harshbarger Se tember 25, 1987 September 25, 1987 T. G. Harshbarger Octob 20, 1987 October 20, 1987 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 ORGANIZATION CHART MANAGER O.C. COOK OU ALIT Y PROJECT MANAGER ASS'T. PLANT ASSURANCE MANAGER SGR ENGINEERING MANAGER PROJECT CONSTRUCTION RAOIATION CIVIL MECHANICAL ELECTRICAL CONTROLS CONTRACTS SUPPORT ENGINEERING ENGINEERING ENGINEERING MATERIALS NUCLEAR OESIGN HANDLING OPERATIONS PROJECT MANAGEMENT STAFF SITE MANAGER OUALII' CONSTRUCTION HEALTH AOMINISTRATIVE PLANT SGR ASSURANCE ENGINEERING PHYSICIST MANAGER PROJECT SUPERVISOR MANAGER ENGINEER ENGINEERINO CONSTRUCTION SUPERVISOR 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 I

Q~j~o:

1 ACI 301-8I<, ",Sl)ecf f fcntions for Structural Concrete Hix propot tions shall be, selected

( Bufldfngs, Clrnpters 2 and 3." I (1) utilizing laboratory or field trial batches, 1 I (2) previous satisfactory perforraance on sfrtrflar work 1 I using the same or sfmflar materials, or (3) prior .I

'I I eRperience Mitlr these or sfmilar materfals to provide I I concrete of the required strength,. durability, vork-I I abf.lity, econoray, etc. I I I A<:1 3r)4 B'>,

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

~

I I Hlxlrrp. Trnr>sspnr't fng, nrrd Placing, Conct'ete." I I I I I ACI '3l'i-BO, "0< t ni! 'rr<l Detnf ling of Concrete I I R< lrrfnr'e< m< ~ r>r ." I I

,I

.I I

Aril tt)B cret< ."

~ 8 l, "lr<<'<<min<>><l< <1 I'r'n<r't. l c>>

~

for Carr frig Corr-I I

I

~i:untf.l standnrd days or C<rr frrg <<lrnl 1 h>> for a p< r in<i nf cured cylinders reach s< v< rr a comp-(7)

I I

I I rehensive strenptlr of 3500 PSI, Mhfchever fs first.

1 I Adherence tn thfs criterfn shall be arrfffcfnnt to I I I preclude testing for "Evaluation of Procedures," I I "Curing Crfterfn Fffectiveness" or "Haturfty Factor I Basis." I I I ACI 318-83, "Buildinp Code Requireraents for Rein- g~~o  : .Hfx proportions shall be selected trial batches, forced Concrete, Clrnpters 3, 4, and 5." ~

I (1) utilizing laboratory or field I I (2) previous satisfactory performance on similar cwork I I using the same or sirailar raaterials, or (3) prior I I experfence Mfth these or similar materials to provide I I concrete of the re<jufred stranpth, durability, work-I I ability, econoray, etc. I I I I I Amer i cn>> M< 1<if rrp Soc 1 < t y D. 1. 1-1986, "Structural I I I Mel ling Cod< St.ee1." I I 1 I I Amer cnn Me 1 1 <1 l np Snc f e ty D. 1. 3. -1981, "Structural I I Mi'<lit'lp ( <1<l<<, Sl'l<'et 8'l >><<1 . I I I I I I ASHE Bnl1< r nrr<l Prosr<rrr<<Vessel Code,Section II, I "Hnterfnl Specifications," edition and addenda in use I nt tfme of mntertnl procureraent. .I I

TABLE 3.2 ntinued)

ADDITIONAL

.0 C.

ASHF. Bol lrr nnd Pressure Vessel Code, Secti.on III-,

"Rules for Construction of Nuclear Vessels/Rules for fracture toughness requirements will not apply.

Constr<rctl~n of Nuclear Power Plant Components,"

rdl tiorr'n<l n<ldrndn ns discussed below. N-stamping of fabricated piping components vill not. be req<<lred.

The orip,innl Construction code for D. C. Cook Urrl t 2 r<<r< lrnr vr rrsnln is Section 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, f~>~to: - Consistent with the plant design basis, "Rirlr f<rr Insrrvlrn lnsprctlon of Nuclear Powrrr f rnct<<rr to<<phnens requi rnrarnts vill not npply.

Plant Compon>>nts," 1983 Fdl.tion plus Addenda through Summer.1983.

I ANSI B31.1, "Power Plpinp", edition and addenda in Q~e'LLoO1: - This code applies only to power use nt. time of. contract award for field piping, piping not classi.fied under ASME Section III, services. Division l.

T ANSI Nr<5.2 - 1977 Q<<alit.y Assurance Program Req<<irement.s for Nuclear Facilities USAS (ANSI) B31.l-1967, "Power Piping". Q~gr~t: - As noted under ASME Boiler and Pressure USAS (ANSI) 8'11.7-1969, "Nuclear Power Piping". 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

ADDITIONAL CO>. nit ", 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.

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

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

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

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

I I

ADDITIONAL 0 F.O O C I

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

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

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

ASTH C33 "Standard Specification for Coarse I Aggrc gnti s". I fine aggregate may be between 2.5 and 3.0, however I lndlv i dual ¹iiliiiilii¹Shall no( v¹ry mor¹ t h¹n 0. i 0 I from the average.

I I

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

I I

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

I I

- Coarse aggregate sodium sulfate soundness loss I shall be a 10 percent maximum at 5 cycles.

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

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

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

TABLE 3. ntinued)

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

I ASTM C96 "Scanclard Specification for Ready Mix I Concrete". I I

ASTM C117 "Tc st Mc thod for Materials Finer I Than No. ?00 Sieve in Mineral Aggregates by Mashing". I I

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

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

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

ASTH C131 "Test Mc thod of Resistance to I Dccgradation of Small-Size Coarse Aggiegate by ,I Abrnslon and Impact. in the Los Angeles Machine" I I

ASTH C136 "Hc thocl for Sic.ve Analysis of Fine and I Coarse Appregates". -

I I

ASTM C138 "Test Hethod for Unit Meight, Yield, I and Air Content (Gravimetric) of Concrete". I lieu of glass plate for unit weight determination.

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

I portions will not be utilized.

ASTM Clli2 "Test Method fox Cl'ay Lumps and Friable I Partlcl<t by vofpht when calculnted as the percentage of Na20 plus 0.658 times the percentage of K20.

I ASTH C112 "Hethod of Sampling Freshly Mixed Concrete". I I

ASTH C231 "Test Method for Air Content of Freshly Q~e ~o  : - Only the Type B Apparatus shall be Mixed Concrete by the Pressure Method". utilized.

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

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 Compounds for Curinp Concrete".

ASTM C311 "Metliods of Sampling and Testing Fly Ash or Natural Pozzolnns . or .Use as a Mineral Admixture 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".

TABLE 3.2-1 (Continued)

I

(0 I

)) '() ~ ~ j I

I I SSPC-SPl throuyh SP10 - 1982 Steel Structures I Painting Council Specifications for Surface I Preparat ion of Steel Surfaces I 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 R EGU LATORY REGULATORY CUIDF: GUIDE GUIDE ADDITIONAL U ))FR 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 3 (2/76) Classification of Class 2 and 3.

Standards for Mater, Steam, and components for the purpose of Rnd-wnstrt Containing Components fmplamenting ASME Section XI of Nuclear Power Plants requireraents was made in accordance wl C)1 t)l)n Ruflie.

Snfety Q<<nifty Assurance Requirements (B/72) Committed to in UFSAR, Section 1.7, Gu 1 r)r 30 for Installation, Inspection =

"QA)'D", Appndfx. A, and Testing of Instrumentation nrrd E)ectrfcnl Equipment 1.31 Cont.rol of. Ferrite Content 3 (4/78) The requirements of this guide are fn Strrfnless Steel Meld Metal 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.

Safety Quality Assurance Program (11/72) Committed to in UFSAR; Section 1.7, Cuidr 33 Requirements (Operatf.onal) "QAPD", Appendix A.

1.31 Quality Assurance Requirements 0 (3/73) 'ommitted to irr VFSAR, Section 1.7, for Cleaning of Fluid Systems "QAPD", Appendix A.

nnd Associated Components of Mater-Cooled Nuclear Power Plants

TABLE 3.2- inued)

R FGU]ATORY REGULATORY REGU]ATORY GUI DF. GUIDE GUIDE ADDITIONAL U B.R 0 iV C. 0 1.38 Quality Assurance Requirements 1 (10/76) C~mmitted to in UFSAR, Section 1.7, for. Packing, Shipping, Receiving "QAPD", Appendix A.

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

)'l ants 1.39 )!ousekeeping Requirements for 1 (10/76) Committed to in UFSAR, Section 1.7, Mater-Cooled Nuclear Power Plants "QAPD", Appendix AD Control of Sensitized Stainless 0 (5/73) If applicable to this repair pro)ect, Stee] the field vork vill comply to this pu) de.

].48 Design Limits and Loading This regulatory, guide was vithdravn Corn)> f natl ono for Sa i am] c ]/4/85 (onn 50FR9732).

Category I Fluid System Components l . 50 Control of Preheat Temperature 0 (5/73) Prnj<<ct repair woik wf)) be perform<<d for Melding of Lov-Alloy in compliance with this regulatory St<<el guide.

1. 54 Qual]ty Assurance Requirements 0 (6/73) Committed to in UFSAR, Section 1.7, for Protective Coatings Applied "QAPD", Appendix A. Exceptfon:

Committed only to ANSI N101.4-1972.

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

and Testing Personnel

0 TABLE 3 2-

~ inued)

REGUIATORY REGUIATORY REGULATORY GUI DF. GUIDE GUI DF. ADDITIONAL O'E 0 N C 1.64 Quality Assurance Requirements 0 (10/73) Committed to in UFSAR, Section 1.7, for the Design of Nuclear. "QAPD", Appendix A.

I'ower Plants 1.68 Initial Test Program for Water- 2 (8/78) This-regulatory guide will be used Cooled Nuclear Power Plants 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 0 (12/73) Welders making velds in areas of of'.fmfte~I Accessibility re..tricted accessibility will be required to practice and qualify on a similar configuration to the weld being made.

1. 14 Qual It y Aasuranre Terms and 0 (?/74) to in VFSAR, Section 1.7, Defi>>iti.ons "QAPD", Appendix A.

(10/76)'ommitted 1.88 Collection, Storage, and 2 Committed to in UFSAR, Section 1.7, Mnfntenance of Nuclear Power ",QAPD", Appendix A.

Plants Quali.ty Assurance Records 1.89 I:.nvironmental Qualification of '1 (7/84) Pro)ect repair work vill be performed "c rtain Electrical Equipment in accordance vith this regulatory Important to Safety for Nuclear guide.

.Power Plants

TABLE 3.2 inued)

RFGUI~T<NY R FGU IATORY RFGUIATORY GU I D V. GUIDE GUIDE ADD I T IONAI.

NU BFR

'V 0 0 ON F C. ONS Quality Assurance Requirements 1 (4/76) for Installation, Inspection, and C109) included in Table B of ANSI Testing of Structural Concrete N45.2.5-1974 is inappropriate for and Structural Steel During the field testing as it is a sophisticated Construction Phase of Nuclear 1abnratory test utilized for cement Power Plants 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)

I bags used, whichever is least.

A~

~ ~

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 ATOP Y RFGULATORY REGUIATORY GUIDE GUIDE GUIDE ADDITIONAL

'tUHBF.

'V S 0 0 C. S 1.100 Seismic Qualification of Flectric I (8/77) Pro]ect repair work will be performed Fqulpment Important to Safety in accordance with this reftrlatory f or Nuclear Power Plant; guide.

1.116 Qirnlit.y Assurance Requirements 0-R (5/77) Exception: Commi t ted to ANSI N45. 2. 8 for Installation, Inspection, (1975), "Supplementary Quality Assurance nnd Testing of Mechanical Reqirirements for Installation, Equipment and Systems. 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 I (7/77) Committed to in UFSAR, Section 1.7, for Control of Procurement. of "QAPD", Appendix A.

Items and Services for Nuclear P1nrr t' 1.131 Qualification Tests of Electric 0 (8/77) Pro]ect field work will be performed Cables, Field Splices, and in accordance with this regulatory Connections for Light-Mater-Cooled guide.

Nuclr nr Power Plants 1.144 Auditing of Quality Assurance 0 (1/79) Committed to in UFSAR, Section 1.7, Programs for Nuclear Power Plants "QAPD", Appendix A.

1. 146 Qunl ification of Quality Assurance 0 (8/8ri) Committed to in UFSAR, Section 1.7, Program Audit Personnel for "QAPD", Appendix A.

Nirclear Power Plants

TARLE 3.2.3 D. C. UXX UNIT 2 TECHNICAL SPECTFlCATlONS NOT APPLLCABLE DLAIINQ TK STEAN GENERATOR RKPAlR PR(LIKCT TKCHNlCAL SPECtflCAllON T lTLE l ADO T lONAL R Nf HAT ION 3.1.1.3 Reactivtty Control System . This TechnicaL Speci',ication ensures adequate Sorm D~tutton ~ 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. tft th no fuel.

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

Shutdown".

Rcactivsty Contat Systees- This Technical Specification requires that me Soratior. Systcos - floe Paths . ~ borm injection f to~ 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.

E

~iree 3.1.2.5 Reactivity Control Systcm-Soric Acid Transfer Pepa-Shutdcan This Technical Specification Least one opcrabt e, boric acid transfer ~

Th 'i s cnswcs that negat i ve that at nmin 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 't rURtlta t i EFt Radl oac t 'I ve %accuse there Nit t be no steea or stem Liquid Kff Luen: tnstnacntat i on, generators these GJO MAltors N: Ll not be speci ficatly t:~ fot toeing ~'Ihtalned cperabte, 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 Ef f trent (2 R 2C) eVLSXO:t 2

TARLE 3.2i3 (Contin <)

TED I CAL SPECI F I CAT IIN ACO IT I ORAL.

N I 3.3.3.10 Iratruscntatie - Radioactive gecause there viLL be no stem or stem gaseous Process and Effluent generators these eight eanitors uILL not be Non'or ing Ins:nrentat ion, mintained operable.

Specifically ~ follcaing arvti LLaree reyairmnts:

48.3.10.2, 2a Condenser E~t ion Svstoa acbl e Cas Act ivity 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) 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 vl I 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 f Ice;<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

TABLE 3.2-3 (Cont inued) j TECIIRLCAL SPEC l F l CAT LOW T lTLE ADO I TLQQL N TCW 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 1 based on the Reactor Coolnt Syst<<being at C

.t

- ~ 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 L ing Operat i cols Ioron Since there Ill LL be no fuel in the reactor Concent rat vessel Liaitations on reactivity condltiore in ical the reactor vesseL are no Longer ~ ccxlcern 3.9.2 Refueling Operaticre- Since there lliLL be no fueL in the reactor Tnstrl<<ntatiel vessel there ill Ll 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 be no cxl residual heat to ~.

fmL in the reactor vessel there sill Therefore, there is no need to mintain an operatimaL residual heat r~L Loop.

3.9.5.2 Refueling Operaticxls . Los Mater Vith re fmL in the reactor vessel there ill'L Level 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 I OKAL T C NT ~

d.S.T.Qa) Adslnistratiw Controls - Plant The P>>SRC viLL revieu the foLLoccirg item Nuclear Safety Reviecc Caasitte>>- generator repair project doccsctents:

Responslbil ities

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

d. II.3 Adsinistratiw Con'.role . Tmporary changes aced>> to proctdw%$ covering P rocechres retcsn to ~ice testing provided it~ ~, b,

~ nd'c.of technical spec if ication d.8.3 are satisf i ed.

d.12.2 Adainistratiw'Controls . Nigh Radiatin Area The keys to those h'igh radiation ~

over to the stem generator project tern shall turned 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 VICE PRESIDENT SmAM GENERATOR REPAIR PROJECT HUGEAR RADIATION PRO'mCTION ORGAMZATION OPERAllOHS ANT OIVISI ANACER NUCLEAR OPERAllONS SECDON LlAHACER S/C REPAIR RADIOLOCCAL PROJECT SUPPORT HAHACER A.E.PWC.

COLULQUS I

O.C. COOK NUCEAR STATION PLANT T

NANADER SITE (COOK PLANT) HAHACER I

PLANT RADIATION PROJECT PROTECllOH HEALTH I SUPERVISOR pHYslasT ROBOT RADIATI PROTECTION COORDINATOR SUPERVISOR l

RADIATION RADIATIOH PROJECT RADIATIOH DOSIMETRY C PROlECllOH PROlECllOH ALARA PROTECTION AND RECORDS SUPERVISOR SUPPORT SER'ACES COORDDIATOR SUPERVISOR SUPERVISOR COORDINATOR I

I XI RADIAllON RADIATION OOSDJETRY Pl PROTECllOH EH CHEERS PROTECllOH AHD RECORDS ALARA J TEcHMaANs (AS NEEDED) TECHNIaANS CLERKS PERSONNEL V)

O

j 7/8 Qp/

Section +ir@T ~Pa e

2. .2 Parametric Comparison ~T Zr 28 2.2.3 Materials Comparison 28 2.3 Component Design Improvements 32 2.3.1 Design Improvements to Minimize otential for Tube Degradation 32 2.3.2 Desi n Improvements to Increase. Performance 37 2.3.3 Design mprovements to Enhance Maintainability and Reli bility 38 2.4 40 2 '.1 Industry Codes nd Standards 40 2.4.2 USNRC Regulatory qides 40b 2.5 Shop Tests am% Inspe ons 42 SECTIOR 3 - hIR PROJECT 3.1 47 3.2 Guidelines and Criteria =

47 3.3 Preshutdown hctivities 50 3.3.1 Site Preparation 50 3.3.2 Shipment and Storage of Replacement Com nents 54 3.4 Post Shutdown hetivities 55 3.4.1 Containment Preparations 3.4.2 Removal of Concrete, Structural and Equipment Interferences 57 3.5 Steaa Generator Removal hetivities 62 3.5.1 Steam Generator Cutting Methods and Locations 62 3.5.2 Removal and Handling of the Steam Generator Upper Assemblies 64 3.5.3 Removal and Handling of the Steam Generator Lower Assemblies 66 Revision 1

~Seceio ~Pa e 6.2 ' Handling of Heavy Loads .155 6.2.2 Shared System Analysis 162 6.3 Analysis of Significant I nzards 163 6.3.1 Criterion 1 164 6.3.2 Criterion 2 164 6.3.3 Criterion 3 165 SECTXOZ 7 - EHVI3KHHHGM REPORT 7.1 Purpose of the EmrLroxxaental Report 166 7.2 The Plant and 2hmiroxuacntal Xnterfaccs 166 7.2.1 Geography and Demography 166 7.2.2 Regional Historic, Archaeological, Architectural, Scenic, Cultural, and Natural Features 167 7.2.3 Hydrology 167 7.2.4 Geology 168 7.2.5 Ecology 168 7.2.6 Noise 169 7.3 Eon-RacHological Enviroxmental Effects 169 7.3.1 Geography and Demography 169 7.3.2 Regional Historic, Archaeological, Architectural, Scenic, Cultural, and Natural Features 170 7.3.3 Hydrology 170 7.3.4 Geology 171 7.3.5 Ecology 171 7.3.6 Noise 172 7.4 Radiological Emrironmcntal Effects 172 7.4.1 Occupational Exposure 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 27 Repaired Steam Generators 2.2-2 Comparison of Design Data Between the 29 Original and Repaired Steam Generators 2.2-3 Comparison of Materials of Construction 31 Between the Original and Repaired Steam Generators 3.2-1 Industry Codes and Standards Applicable to 49a the Steam Generator Repair Project Field Work 3.2-2 USNRC Regulatory Guides Applicable to the 49f Steam Generator Repair Project Field Work 3.6-1 Steam Generator Repair Welds 75 3.8-1 Repair Project Manrem Estimates 98 3 '-2 Pro)ected Project Totals by Phase for 103 Man-hours and Man-rem 7.4-1 Donald C. Cook Annual Man-rem Expenditures 173 7.4-2 Steam Generator Man-rem Expenditure Comparison 174 7.4-3 Gross Contamination Levels by Location 180 in Piping and Steam Generator 7.4-4 Donald C. Cook Nuclear Plant Unit 2 181 Estimated Steam Generator Curie Content 7.4-5 Effluent Release Isotopic Distributors, 182 Steam Generator Replacement Project, Surry Power Station - Unit No. 2 7.4-6 Comparison of Gaseous Effluent Releases 183 from Donald C. Cook Nuclear Plant 7.4-7 Radionuclide Concentrations in Reactor Coolant 184 7.4-8 Estimated Specific Activities of Laundry Waste Water 185 7.4-9 Estimated Radionuclide Releases Due to 186 Discharge of Reactor Coolant Water vii Revision 1

~ ~

8

XZST OF XhSIXS cont'd.

Table Title ~Pa e 7.4-10 Estimated Radioactive Liquid Effluent Releases 188 During the Donald C. Cook Unit 2 Steam Generator Repair Proj ect 7.4-11 Comparison of Radioactive Liquid Effluent Releases 189

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

FIGURE 2.2-2

.MODIFICATIONS TO UPPER ASSEMBLY INTERNALS UPPER ASSEMBLY 4 0 0 0 0 0 0 00 EXISTING SECONDARY 0 0 SEPARATOR 00 0

04

• 40 4 SECONDARY MANWAY 0 4 4 4 (2-180'PART) 0 4 0 4 4 ,NEW DRYER DRAINS 44 O4O4 4 (8 TOTAL NEW STEAM WATER DEFLECTORS (3 TOTAL)

???@Spy +<X??? . i NEW STEAM CHIMNEYS 0??."~i'W'v'?.~ '(2 TOTAL

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

SHELL EXISTING SWIRL I

VANE ASSEMBLY

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

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

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

"',,Oi  ??0%'l',4?

..~mk~iiwiiV~wgi4A i'5, '?

44 REVISION 0

I 'I 1

I TIFF. 2-1 XNIXJBXRY OODES AND PQQKlARDS APPIZCMKF~ TO 'LHE 0

STEAN GEN1RAXOR REPAIR PRMZCZ ACI 301-84, >>Specifications for Structural Oancxete @~Lion Mix proportions shall be selected Buildincpa, QMLptexs 2 and 3.>> (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 304-85, "Reccaznended Practices for Measuring, Mixing, Transportirg, and Placing Oancxete.>>

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

ACZ cv~.>>

308-81, 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 E>my<ion: Mix proportions shall be selected foxced Oancxete, Chapters 3, 4, and 5..>> (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.

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;

t4 TABLE 3.2- ntinued)

CODE OR PVQlQARD ASME Boiler and Pressure Vessel Code,Section III, ~Exes iona: Consistent with the plant design basis, "Rules for Oonstruction of Nuclear Vessels/Rules for 1m~ wy.

Construction of Nuclear Power Plant Ocmponents,"

edition and addenda as discussed below. N~mping of fabricated piping camponents

'Ihe original Oonstruction code for D. C. Gook will not be ra~

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.

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, ~Exes ion: Consistent with the plant design basis, "Rules for Inservice Inspection of Nuclear Power fracture toughness requirements will not apply.

Plant Components " 1983 Mition plus Addenda through Sunmer 1983.

ANSI B3 1 1 g "E~ Pipirg", edition and addenda in Exce+ion: - 'this code applies only to pcarer use at time of contract aware for field piping piping not classified under ASME Section III, services. Division l.

USAS (ANSI) B31. 1-1967 Il~~ Pipingn ~Exes ion: As noted under ASME Boiler ani pressure USAS (ANSI) B31.7-1969'Nuclear Power III, Piping

~

Vessel Code, Section "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, will be updated to ASME Section III, with the exception that fracture toughness requirenents will not apply.

TAIKZ 3.2- ntinued)

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

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

APIN C33 "Standard Specification for Coarse ~Zxoe 'ons: 'lhe average fineness mcdulns of the Aggregates" . 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 of 4 gg lit ~1 out of 5 successive test results meeting the specifications.

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

- 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.

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

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

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, Exee~Mons: - Except strike off bar utilized in and Air Content (Gravimetric) of Concrete". lieu of glass pl.ate for unit weight deterzaination.

E~t 'tYieldn and nAir Content portions will not be utilized.

(Gravimetric)

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". ~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.

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

~ ~

TABZZ 3.2 Continued)

CODE OR FRNIRRD ASIN C231 "Test Method for Air Content of Freshly Exec~on: - Only the Type B ApImratus shall be Mixed Concrete hy the Pressure NethocV'. utilized.

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 Notes: 1) All AS%Ms are latest edition.

TABIF -2 USNRC REGUIATORY GUIDES ~CABLE KO 'LHE STEAM 61XERAER REPAIR HKOECT HZID WORK REGUZAXGRY HEGUIAIQKf GUIDE GUIDE TITIZ REVISION 1.8 1-R (9/75) OQK6tted to in UFKR, Section 1.7, "QAPD", Appendix A.

1.26 Quality Gzuup Classification and 3 (2/76} Classification of Class 2 and 3 Standards for Water, Steam, and campceents for the purpose of Rad-waste Oantaining Ccaaponents implementing ASHE Section XZ of Nuclear Paver Plants I I wi.th this guide.

Safety Guide 30

~ttf and I ttft f Electrical

~tf for Installation, Inspection Equipment (8/72) Oammitted to in UFSAR, Section 1.7, "QAPD", hppr~ A.

1.31 Oontrol of Ferrite Oord~ 3 (4/78) f ttt tft in Stainless Steel Weld Metal new covered by ASME Section III.

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

Safety Quality Assurance Program (~72) ~tted to in UFSAR, Section 1.7, Guide 33 Itt tf ~I "QAPD", Apped A.

1.37 Quality Assurance Requirements 0 (3/73) Oammitted to in UFKB, Section 1.7, for Cleaning of Fluid Systems "969", Appendix A.

and Associated Oamponents of WateL~ooled Nuclear Power Plants

TABIZ 3.2- ntinued)

RHGUIAIORY MGOEAIQRY GUIDE GUIDE NUMEN TZKZ

~

1.38 Quality Assurance Requirenents . 1 (10/76) Oammitted to in UFSAR, Section 1.7, for Pacify, Ships@, Receiving nQAPDn, Appendix A Storage, ancl Handling of Items for WaM~ooled Nuclear Pc@mr Plants 1.39 d~jw Water-Oooled Nuclear f

Poem'lants 1 (10/76) Oammitted to in UFBAR, Section 1.7, nQAPDn, AppeDdiX A 1.44 Oantml of Sensitized. Stainless 0 (5/73) If applicable to this reImir project, the field work will camply to this guide.

1.48 Design Limits and Ioading 'Ihis regulatory guide was withdrawn Oambinatians for Seismic 3/4/85 (see 50FR9732).

Category I Fluid System Oamponents 1.50 Oontrol of Preheat Teatperature 0 (5/73) Project repair work will be performed for Weldizg of law"Allay in campliance with this regulatory Stee1 guide.

1.54 for Pmtective Coatings Applied 0 (6/73) Oammitted nQAPDn to in UFSAR, ApI~ndix A.

~on Exceptio 1.7, Committed only to ANSI N101.4-1972.

1.58 Qualification of Nuclear Pamr 1 (9/80) Oammitted to in UFBAR, Section 1.7, Plant Inspectian Bamination nQAPDn Apped A.

and Testing Personnel

TMKH 3.2-2 )

RHGUM1QEK GUIDE REVISION tWltb 0 (10/73) Committed to in UFBAR~ Section 1.7g for the Design of Nuclear nQAPD'r Appendix A Power Plants Initial Test Pxogram for Watex 2 (8/78) this regulatory guide will be used Cooled Nuclear P0wer Plants only for guidance in developing a test program for those components and systems affected by the Steam Generator ReLair Pmject.

Welder Qualifications for Areas 0 (12/73) Welders ma)de welds in areas of of Limited Accessibility restricted accessibility will be required to practice and qualify on a similar configuration to the weld being made.

Quality Assurance Texan and 0 (2/74) Cam6tted to in UFKK, Sectian 1.7, Definitions nQAPDn Appendix A Collection, Storage, and 2 (10/76) Oammitted to in UFKR, Section 1.7, IRDk f CI nQAPDn- Appendix A Plants Quality Assurance Records Envtmranental Qualification of 1 (7/84) Pnrject repair work will be performed CBl. tain Electrical EquipIIPAIt in accordance with this regulatory Important to Safety for Nuclear guide o Power Plants

~ ~

TABIZ 3. 2-2 'd)

RIGUIATORY R1MUIAIORY RHGUIATORY GUIDE GUIDE GUIDE NUMBER TENEZ REVISION 1.94 twjl~ 1 (4/76) Exceptions: << "Grout testing" (AFK for Installation, Inspection, and C109) included in Table B of ANSI Testing of Structural Qm~e N45.2.5-1974 is inappropriate for and Structural Steel During the field testing as it is a sophisticated laboratory test utilized for cement Const~ction Ehase of Nuclear Paver Plants pre-packaged non-shrink grouts accepted for use on the basis of manufacturer's ~fication or

~

evaluation. In lieu of daily tests, be 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 A~kahility of this water or ice as S04 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 E EGUIAIORY EKGUIATORY GUIDE GUIDE GUIDE NUMBER TZIIZ REVISION 1.100 Seismic Qualification of Electric 1 (8/77) Pmject repair work will be performed Ecpupnent Important to Safety in accordance with this regulatory for Nuclear Power Plants guide.

1. 116 t&1!11 0-R (5/77) Emption: Oammitted to ANSI N45.2.8 for Installation, Inspection, (1975) "Supplementary Quality Assurance and Testing of Meclanical f Eguignent and Systems. and Testing of Mechanical 11'nspection 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. 123 ~ftf 1 1 (7/77) Oammitted nQAPDn, to in UFSAR, Section 1.7, 3+)endix A Items and SerVices for Nuclear Plants

1. 131 -

Qualification Tests of Electric 0 (8/77) Pnrject field work will be performed Cables, Field Splices, and in accordance with this regulatory Connections for LightWater~oled guide.

Nuclear Palmr Plants

1. 144 Auditing of Quality Assurance 0 (+79) Oammitted to in UFBAR, Section 1.7, Pxogralns for Nuclear Power Plants QAPDn Appendix A.

1. 146 Qualification of Quality Assurance 0 (8/80) Oamitted to in UFBAR, Section 1.7, Program Audit. Pmmonnel for nQAPDrt AppelxUx A Nuclear Pawer Plants

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 The The Feedwater and Main Steam Line Piping Cuts feedwater and main steam lines will be location of the cuts, the equipment to mechanically cut be used, in two places.

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 t elevation 650'loor, bay and oriented in a the lower assembly will be moved north-south direction, lowered to the secured to a wheeled transporter. The lower assembly out into the railroad 609'levation will then be transported and Revision 1

It 0

QME 3.6-1 KDN CNEBHKR HEEMR NZEB CU RKL Kba in. XHNP Eb3clBhee Ehirer to Pip ce Sh-1C6 Sh-306,~

to II

~

Sjrg1g V with hx9mg rhea 84%

with GIM E7C8-2 1100-3200 11xue Raa 600 Pa wel&6 Qp hat & axiL 400/hr Pipe to Pipe Sh-396,~ 14" .709I Sa abaca M ckaa E7QIS

~

Gird to X53di'rGhK'A50B,CL-2 .843" Girxgs V GIM lKO-3200 HP root N:ezle to to 35-40 without root 1 her mnxa AM HT firal ptipe SQ06,~ lxdcing zbg PW Rne 600 HP (Qat root) 631 hat & cmL GIM 400jhr.

CEP 86,CL-3EI. 2" ce wziEr SCCht to laa SQ06,~

Ehamm Rat. 8KB,CL-3a 2 CL ME&33 BX994 PTR E7t3IS 2g to Iiat. to 3EEa C3., CK Pip SQ06,~ GIM E7CB-2 2" cr mrna Rdxt PR/ E701S Nxz3e to to lEEa Pipe SQ06,~ GIM E7C8-2 Hlaiibm Pip Sh-306,~

to Pipe Bdn Sbean Sh-355,CL-1 32 I 1 i/8" Shale V 84% E703S 1100-1200 hs weld' Pipe to Pipe Gr. IC70 35-40" with Mth 1xurs (SMBl,CL-32 hx9drg rhea GIM Raa 600 Ge CKR} CEP hat & axiL 350~.

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%HE 3.6-1 (cent.)

0 SERM CKNERMICR EKBQR %KB SHH,C1-2 32'i 1 3/Bri Shrug V BSf EN1B 3300-1200 As aahRQ to 35-40 & with 2 hazs SKL6,CL-1 30-lP GIM E7CB-2 Raa 600 GL IC70 CXP hat & axiL (SA691,CL-32 350jhr.

GL" CKH)

RBc~ ~51,CARI 31 3D 2.88 Shg1e U GIM 6016 50 Not ~. Gdn2giLLh HP,UP,PZ Qxi1ak Pipe (3l6) ta XS flat root root & with 360 grit to Stem weld max lay 8%7 %16 CX'bi" gzmcabcx nmQe

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I SR31 SKN,CL-3 175 3/4n 3 62 r U gag EKHS B3 F00-3200 Gard fee HP,UZ,PZ Ttamitial ta 8633,~ hx9zprp cx cx SN Z~lrr's 2hr30ndn UZemn cx MP ma to p1abe CL-1 Sinijle U Qxdce Raa 800 with hx3ug 5x'N hat & cooL Rmxe hx9drg AD/he RR9gxxp

'akB Sh-285,GPC 124.ZP 3~~ ange V em 50 Rt ag. Cd', Blair MP

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arQ HLe. flat root Intenal Rxa Prize Cnpamts 1

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

Mthth f v 3 Meld filler metals 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.

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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.

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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 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 6T 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.

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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 0 F (which still allows 32 0 F margin to 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 0

hazards issue using the 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 that have been successfully implemented at two other commercial nuclear power 1 plants, namely, the steam generator repairs completed by the Virginia Electric

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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 l repair to the Donald piping will be reinstalled installation requirements.

plant and C. Cook plant systems design Unit 2 steam generators.

Therefore, because there no new All components to meet the original design and configurations will be no changes or different accidents are created.

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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.

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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.

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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.

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TABLE 7.4-1 DONALD C. COOK PER UNIT AVERAGE ANNUAL MAN-REM EXPENDITURES Exposure YEAR Man-rem 1980 246 1981 327 1982 321 1983 283 1984 344 1985 448

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TABLE 7.4-6 COMPARISON OF GASEOUS EFFLUENT RELEASES FROM DONALD C. COOK NUCLEAR PLANT Estimated Release Average 1985 During the SG Radioactive Release/Unit Repair Effort

~Secre C Ci Noble gases 2.47 x 10 Negligible Iodines 6 '6 x 10 9 x 10-6 (1)

Particulates 3.72 x 10 2.92 x 10 Tritium Negligible Notes (1) Estimated from Surry Unit 2 Data.

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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."

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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.

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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 ~f LFc ~PGE GENERAL 1-3 1.2 EVALUATIONS 1-3 1.F 1 Crane Manufacturer and Design-Rated Load 1-3 1.2.2 Comparison to NUREG-0554 and NUREG-0612 1-3 1.2.3 Seismic Analysis 1-14 1.2.4 Lifting Beams 1-17 1.2.5 Interfacing Lift Points 1-18

1.3 CONCLUSION

LIST OF TABLES

I~ABL ~ITLE ~PGE 2.2-1 150-Ton Capacity Single-Failure-Proof 1-5 Crane Design Factors 2.2-2 Steam Generator Repair Project 1-7 Auxiliary Building Crane Lifts Over 60 Tons LIST OF FIGURES g'TGGg+E ~TTTL gAGE 1.2-1 Mathematical Model of Crane Trolley at Mid Span 1-2 Rewision 1

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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.

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S

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TABLE 2.2-2 STEAN GENERATOR REPAIR PROJECT AUXILIARYBUILDING CRANE LIFTS OVER 60 TONS Est. Wt. Number tern ~owns ~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 AUXILIARY BUILD %G CRANE LIFTS OVER 6D TONS Est. Wt. Number LXmal 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.

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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 I

1-8 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.

HOISTING MACHINERY 4.] 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.

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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.

Revision 1

Cb 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

,~

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 flexibility of 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.

Revision 1

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. Static and three load step 1-16 Revision 1

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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 lift points will be dual-load-path and designed to.shear stress levels not to exceed 4,500 psi will be 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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jpi' 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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