ML19256B690

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Forwards Summary Chronology of TMI-2 in-plant Activities, TMI-2 Test Program Chronology Bar Chart & List of Delays to TMI-2 Test Program,In Response to Request for Detailed Analysis of TMI-2 Startup Test Program
ML19256B690
Person / Time
Site: Crane 
Issue date: 01/23/1979
From: Cutler R
GENERAL PUBLIC UTILITIES CORP.
To: Arnold R
GENERAL PUBLIC UTILITIES CORP.
Shared Package
ML19256B685 List:
References
TASK-TF, TASK-TMR G-712-8-4, NUDOCS 7908230266
Download: ML19256B690 (47)
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Text

{{#Wiki_filter:i mi d .r. M /,. M 7[N e' 'e '~c J7 l' . mer=0771C3 i,l3Gmorandum 3 q{ U" 4/yrp-b $sg5c3 22e January 23, 1979 ,,j S ::e

  • Three Mile Island : uclear Station Unit 2-4 lf)-

Startup Test Program Hi-tory and Delay Analysta aim,l / -'w#~ c 2 Mr. R. C. Arnold Q d-< Lo mo-Parsippany n!I-2/4022 g

  • f' nclosures:

e (1) Summary Chronology of D1!-2 In Plant Activities; 2/8/78 through 11/16/78 (2) DII-2 Test Program Chronology 3ar Charr (3) Power /:!cde Histogram of n!I-2, 2/1/78 through 11/16/78 (4) List of Delays to TMI-2 Test Program Due to Problems Encountered (2/1/78 - 11/16/78) (5) Test Program Critical Path Assumin; luin Steam Safety Valves Function Properly - Worst Case (6) Test Program Critical Path Assuming Main Steam Safety Valves Function Properly - Most Optimistic This is in response to your request for a detailed analysis of the T'!I-2 Startup Test Program and delays thereto which have caused interruptions to the schedule. The GPUSC Startup and Test Scheduling Engineer, Tom Faulkner, and I reviewed the Met-Ed n!I-2 Shift Fore =an's Log and the GPUSC Startup and Test Shif Test Engineer's Log for the period 2/S/78 through 11/16/73 and developed a su=ary chronology, Enclosure (1), which lists the major activities as they occurred. From this chronology, a D11-2 Test Program Chronology 3ar Chart, Enclocure (2), was developed. This chart graphically displays the Test Pro-gram as it actually occurred along wi:h problems uhich were encountered. (!:ote tha: the original Test Program Schedule envisioned a 120 day program. This was based on the n!I-l Test Program experience and did provide so=e minimal amount of time for delays. Various conditions can be expec:ed to cause delays during any startup program. I have attempted to select only those equipment problems which caused, or could have caused, delays in the ~:!I-2 Test Program.) Enclosure (3) is a Power / Mode Histogram for your information. On Enclosure (2) I have indicated in bold lines what I believe to be the a::ual cri:ical path between 2/1/73 and 11/16/78. For clarity, I have chosen not to indicate on this critical path those problems which occurred during any test phase if they, by themselves, only amounted to a few days delay at a time. In these cases, the cri:1 cal path is shown to con:inue alen; the test phase path. t9[6210 g950

r e Mr. R. C. Arnold January 22, 1979 Enclosure (4) is a listing of those problems shewn on 2nclosure (2) along with my evaluation of the period of time delay actually associated with each problem and the potential delay it, by itself, would have caused had it not been entirel,r or partially included under an umbrella of other problems at the same tima. Enclosures (5) and (O project :ritical path scenarios predicting what I believe would have been the course of events had we not eng. 1-aced the Main Steam Safety Valve problem and subsequent retrofit modification. In these cases, the inadaquacy of the steam line steam ha==e restraints, discovered in nid-1978. and the notification by 3&W of potential loose parts (orifice rod 1.semblies and burnable poison rod assemblies) in the reactor interr.als and subsequent rectification, would have had a much more serious irpact on our Test Program schedule. Enclosure (5) shows a 'NJ ;st Case" sr.enario and Enclosure (6) shows a "Most Optimistic" scenario. The resulta of these studies show that, because of problems encountered in the Test Progran other than the safety valve failure, the safety valve problem itself was solely responsible for a period of delay on the order of 20 to 39 days. An explanacion of these hypothetical critical paths and the actual critical path is attached to the respective enclosures. The folicwing persons have been consulted in performing this review: J. J. Barton - G?USC TMI-2 Project Manager T. Faulkner - G?USC Startup and Test Scheduling Engineer A. S. Dan - BLR TMI-2 Proj ect Manager W. R. Cobean - B&R, Vice President R. W. Hewird, Jr. - G?USC Manager of Nuclear Projects ,A / sf f.,y )cG c tet. t-RCC/brh R. C. Cutler CC:J. J. Barton (w/encls.) W. R. Cobean A. S. Da= R. W. IIcitar d ~;. H. iilrst R. J. Toola T. Faulkner I916 271

Enclosure (1) SCO!ARY CHRO:' OLOGY OF n!I-2 III PLMIT ACTIVITIES ~ 2/3/78 THROUGH 11/16/78 9 4 9 l'916 2T2' > s,. 1

ACT!VITY ~CA"' : FE3 lit.~ t ':ormal preparations for fueling - Reactor Vessel filled t 8 (Recetved NRC Operating License) 8 1 1/3 t .q... _.. -_ 9 ,:1EC working on fuel transfer carriages to support fuel load 1 1 f l I 10 1 i ,') r.- 6 1st Fuel asse:2. in core i f i .D -....... _.._, _ g 12 13 i i i Last Fuel.tssem. in core Co cence preparations for R:t head installation 14 15 I Core barrel installation complete 16 Rx head installation begins ', 'l ~. 1 1916 273

. -. ~ ACTI.ITY FE3 T' 2x head installed, nor torqued 4 17 i Rx head bolt torquing begins L 18 5 Rx head installation complete Ccemenced RCS fill & vent & CRDM coupling ? Started installing incore closures 1 19 ~ l Cocpleted coupling APSR's Complated torquing all incere closures i 20

  • e ergi:ed pressuri:ers for loop fill I

l RCS filled - to be vented . i Cc=cenced venting CRD's & RCS instrumentation t i 21 l Completed venting CRD's at low RCS temp./ pressure i Ee31a raising RCS temp./ pressure 22 ; m@N \\b 'l 9 9 23 D s\\ ,i ? .* = ~ .il. 2A l' j 3egan running RCF's. No ted noise on LP'.?!. R.J advised secure pumps 'j until analyzed. Pumps out of servica caused slower heatup rate 1/3 7 I _/s-25 Re startec RC?'s and heated up to <200*? [ C ::enced CRD Functional testing

  • I k

ACTUJITY A,4 a,, FE3 T Completed venting CRD's at RCS temp. <200'F, >400 psig Corner. cad filling OTSC f or secondary side hydro te-test 26 2/3 Pressuri:cd A&B OTSG's to 900 psig - >!anways/Handholes satisfactory, 1/3 other packing & seat leaks noted 27 Pressuri:ed A&B OTSG's to 950 psig, noted leaks, vented OTSC to fix leaks I Continued CRD Functional Testing 2S t 5 TAR 1 RCP operational tests completed CRD Functional Testing Conpleted, Start CRD Trip Tests l t I i l I i 3

I i

i Completed CRD Trip Tests, Paising RCS pressure for hydro 1/3

l. Commenced RCS hydro, completed inspection, noted leaks, co=cenced depras-j surization (Cooldown & drain to fix RCP Seal injection Grayloc flange leaks);

4 I l n m \\ yF 9 0. 1 5 Grayloc flanges fi:<ed, ccc=ence RCS fill g hk \\\\ " Fill & vent in pro;ress i 6 1 7. l'916 275

ACT!'.* ITY M.\\ R. - 5 i!l 7l l, I

I t

.i Cc:pleted CPO:: Ventin:; ..... i.a l Discovered gland steam leaks. Secured F.!l. Heating, gland stea: ,;j: RC3 =aintained <200". to fix (can' t heat up) 3' t- .i 1 3 I ! l l Commenced glsnd steas leak repair i f l Found caustic leaking from N, header in Aux. Bldg. 5 into bleed tanks Commenced cleanup mode. Completed gland steam leak repair 9, j 1 8 t l \\ l '/3 10 l Ccapleted cleanup ot Caustic (NaC'I) in RC Bleed Tanks 1 Co cence heat-up to Mode 4 a. i 4i Reach l-!cde 4, first ti=e l, I r i i I 11 ' Rx 31dg. Personnel Access Hatch door seal failed. (Can' t change node ' 2/3 I 'till fixed) - Tested Sat. at 2230, recommenced heat up I i ! ? l Chemistry problems with RCS for high socius - cleaning up two make-up dcmineralizers l, I *; h.9 1 i q. i l m N \\\\\\y(@ QJ)u " f Commanced heat up 9 i 3 Reached :! ode 3, first time 13 ' t 7,CP 2A Clutch failure, all pumps secured, plant coastin; down 1 f I e. 5 l.'l Installed anti-rotation 'I' beam on RCP 2A 1 14. 1 lI 1.o s t } gg aux. boiler, causir.; F.". out of spec. Cor.=ence ocerat h on 3 RCP's i-1 n1 t' '1 1 / I . ining O!3G due to high phosphates I/IU LIU 1/3 i Catmenced heat up i 15i 4 1/3 fi 3 ';oise noted in CTSG. Ccoled down, depressurized, for instru entatior 1/3 i r o.u.t..: recat:

ACT U.' ITY . ~, ".A R. 4 Plant cooldown continues 5 16 l: 1 . ii - i. 8 _....-....i_'. (Also looked at RCP-2A during this period) t i a 17 ' 1 1, ! !I e i-CC:!PLETED OTSC INSTR'O!ENTATIO:I REPAIRS i I.' Commenced fill & vent l !! 18 3 l' Completed venting CRD's I 35' doing RCP-2A pump shaft runouts & :otor checks & seal replacement ,Ij i l i i 19.i i 1 i l Co:pleted work on 2A & 1 locked it. Commenced heating & pressuri::ing RCS

  • /

g Cor.tinued heat-up i 4 !~ 20 ' !i i 1 i i i 6 i i l 3l I 21 j l CRD Rod tests performed 1 l, lj I i i. I j ;' Cooled slightl;< to fix RCP-!A seismic restraint which bottomed out i , 2/3 i Co:cenced Pressuri:ce Spray test ] 532'F, 1400 psi; b, ( C. ant.inuina CCD t.e.s tina) :!ad.e o..t.he r r. epai r s Mk\\ N f \\\\\\\\j 9a-i 9 @.' e 23 l b 1 j Completed group rod drop testing i llt

RCP,flov meas'urc=ents with 3 RCP's conducted, completed.

1/3 24 : I e i ?' Depressuricia; 5 cooling down due to leaking Conax connectors ... j 1916 277

.f.".'..,._., ..t .. - - - - - - - - -. -.. - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -- - - - - - - - - - - - - ~ -.2 ,, \\,.,. I 25 5 1 Comoleted Canax connector replacement, tightening C0:=enced heat up 26 1 i i 4 3. ~ 1/3

e. l i j i Started pulling reds to critica111ty 2

Initial Criticallity Achieved, first time j Con =enced prepara: ions for :ero power physics testing 28 i Star: reactimeter checkout i Compluce reacti=eter checkout a

I; Ran all rods out critical boron concentration test 29 ' 3 ~

MEC ES testing-fuse blev on 2-IV inverter, Ax trip, ES actuation, (Pressur-1:ce Electrocatic Relief Valve Lif ted) NaCH contaminated RCS Ccamenced cleanup of RCS 1/3 I 30 1 1 p Disecvered Cl" in RCS due to NaOH contamination, cc=mence plant coolde.n i 31 ':9'. Cocidown, cleanup continues 9, QM.NS# I 3 _ Gs \\ 64 AP R ' ) \\\\"$ 3rcakdow. vacuus to repair sua. stea leak in Uni: 9 y) 1 Satisfactorily tested repaired inverter 1 Torqued OTSG conax connectora RCS cleanup continues -egun heating up aux. stea: line frc Uni: 1 l V. Oc n tinu flushin: '[ f ren 't'; & CH s y s r.s i a, ,s r

ACT "l !!'.'... .m: AP R ~' ~ 5 6 3 Discoverec DR-V-LS6B/RC-V-149 problems preventing proper pressuri:er Spray o eration, thus slowing down RC3 cleanup progress 1 i Star:2d venting,CRD:t's; ?C-V-II.9 problems resolved i CRD:! ven:ing to clean CR!::1's of Ma continues I. i 4 1 1 l j i l 1 i Comolete CROM venting Condensate & Feedwater cleanup in process } f 1 5 1 1 t L j I.ost one Aux. Boiler causing F.W. to cool down, delaying F.U. cleanup i

  • progress j

I i 1 6' i 'l l S t a r t e d F.*.J. heating again after Aux. Boiler was fixed j Continuing F.U. cleanup i 7 (' ; i RCS heat up in progress 1 ( 4 3 i 1/3 ! i Re-established prerequisites for zero power physics testing ,8l 2 Zero Po-er Physics Testing (ZPPT) in progress < t I g$$b 3 Croup 8 Rod ': orth M.casucc=ents completed N 2 Preparing for Stock Rod Worth !easurements 9 3 Qiustionable ?.eactimeter results discovered gY Reciaccc card in c.1actimeter - neccare f or_ chec'cout o 4

1. J s

y 9 j Reae:iceter still uunsat. 1 10 i

1-3 5 a swapped on reactinc:er prepare for checkout l

,, i (Spare reactimeter rec'd on site) t' 3, ' New reac:iceter de~. castrates same unsat. characteristics 1/3 i ,r

CGntinued CFPT waile at:capting to resolve reactineter problems i

11 3 ( D f916 279

I ACT!'.'!!Y ..:.:... ~ lJi... APR 3 - :.? 2 PPT continues 12 3 1. 3 3 2 PPT continues 9-s 13 3 d :> !2,- b 2 l ZEPT continues 9 D' 3 s\\

  • t, Q

s y i l l i i j 3l Verified & rechecked reactimeter results from control room 15 : 2 i ! f' i

3 2 PPT continues

,c 3 i 2, i N1 cable unction box discovered to have bad connector causing 16, 3 some unsat. reacti=eter readings. Problem was corrected. 3 2 PPT continues T i 3l 17 - 3 r 2 PPT continues t ,I -i 3I Completed 2 PPT 8 I r 2 i 3 segin 0-15% power escalation / testing - Rx Trip 9 2.50 power on 1/3 2 13 ' 1 7 Power / Imbalance / Flow 9-15 Power escalation testin; in progress .. J, ...l ? Rx trip when blowing down suction strainer on Condensate Pump i 3 (tripped on high pressure) 1/3 19 ; 2 Improper sampling procedure caused delay in recovery

1 10 Power Escalation Testing in progress

! l3 15' Completed turbine data for 3 I ._........t.._ i RPS Channel C =anually tripped due to bad NI l !' 7 Rn tripped on lii flu:c, Channel D (NI-3) Delay in recovery due to ';I-8 again 1/3 3 1 15 ? wer Iscalatic: 1: pr:: ess e \\ t

.Ci:*:T TT

...., : n L.~ :! '.. - - ---. ~.d.' u.. JJ R 1 15 Turbine Generator synchroni:cd to ; rid for first ti.ne ( 10 5 '".le ) 21 i 157. Power Testing continues l e-I Leakage noted at B OTSG Conax connectors y i i Perforned Loss of Offsite Power Test 22 3 I Delav in recoverv due to NI nain g 2/3 2 1 15 T.G. back on line ( 100 MWe), Coepleted 0-15'; Testing (W/3 RCP's) 23 20 3egin 15-407. escalation / testing 1 30 Rx trip on NI power spike, rapid cooldown, 51.S. relief salve excessive 3 blowdown, ES actuation on low pressure, NaOH injection, bellows liners ! blown from St.S. relief valve dischar3e_ stacks. ' 1/3 4 Cooldown 5 investigation in progress ,f Inspection fou::d Conax connectors leaking - will depressuri:e to 1 24 I repair. Also found tube leaks in 3A FW heater 5 ..i. RCS cleanup in progress e-25 ' I t qN>O I s b C' s v. a ?- v)c W i. ( i sec 9 26 Drained 3A FU heater for repair 1 9 ~ y y RCS cleanup continuing j. 27 1 i - me CS depressurized f or Cona:: repair, discovered all leaks an RC?-23 "otor egan working Conax connectors 1 f Coatinuing to investigate bellows liner problem j i -..n 7 RCS cleanup in progress 29 1 ,p., t I e --+ - --- ---*=***~ - 1916 281

ACTU/!!Y ~ ~ ~. APRI ~i Preparing for RCP-2A Clutch replacement i. I' 30 ; I Bellows liners evaluated-all hori:ontals & verticals to be replaced 1 -l 1

i

.1-.. nY !' 1l' Replaced Conax connectors seals where needed 1 Moved RCF-2A clutch into R.B. I i e i: e ..-s I! '. Torqued all Conax connectors 2 Cont ~ nued to wo rk M.S. relief slve discharge lines 1 i ii t . i' t i i I 3i Continued to work M.S. r lief valve discharge lines 1 ~ 3 t .' l Cc=nenced F.W., heating i. 4l l Unit I Lost Aux. Soiler - couldn' t =sintain teep. on 1 boiler, delayed 1 ! i F.W. cleanup i ! L~ nit I now at 757. - extraction steam available for F.W. heat-up 5! 1 i l Co=pleted RCF-2A moto r checka, ran wa ter ey\\x.(sks Complated RCP-2A noto r/ pump coupling c4e^ 'g' g'pj RCS fill / vent vnive lineup in progress <c 6 (N@Q 9 N'; -.4..GL { O' Started filling a venting RCS 7 i Filled pressuriser, started drwing bubble 1 i t i Started venting C't0's ._L.....--.-......-.. - ~ - - - ~

I l

Completed initial CRD venting l Func tionally tested RCP-2A sat. 1916 2821 8 l ? l 7 Found C.,nax to nectors leaking. tor:ued to stop 12aks 3

ACT!'7:TY . a. ... a >!AY ' ~5 Started second vent & flush of CRD's 9. Found Conax connectors leaking, cc=cenced cooldown/depressuri:stion to 1 repair Removed nuts feca Conax connectors, apolied Molvkote & retorcue'd completeu venting CR6s, commenced raising RCS pressure ~- 1 10 i. ' ! i, I li Continued to flush CRD's at pressure i ; Completed venting CRD's 1 11 1 Discovered Purge Valve AH-V-4A could not close f .7.C working on repair to AH-V-4A Continuing F.W. cleanup, checking out RCP operations, etc. 12 il 1 I 9 b ~ I m AH-V-4A repaired & tested gkY t / 13 P. ode 4 prerequisites completed 1 4 3 RCP-2A upper oil reservoir leaked empty. Splash shield drain lines clogged, i 4 cil spill to 230' el. in D-ring. Started cooldown, depressurication to fix. 1 14 5 i. Etxed RCP-2A backstop filter gasket, added new oil Oil cleanuo in progress t 15 Oil cleanup in progress. Also fixing oil leaks on RCP-23 motor i l l l flushing oil drain lines l Oil leaks f t::ed, drain lines cleaned 16 Found leaks reversed on four RCP oil pumps caused' ea. verse rotation 1 Fixed leada, commenced plant heat up 4* !3 aegan r'ost E.L., pre-critical testing w/4 RCP's not previausly perfor ed 1 IT O i i t 1916 283

ACTIVITY MAY 3 l Congle:cd I. pump testing (pre-requisite to return to criticality) 13 i Cocmenced M.S. relief valve :estina 2/3 .I i l I i i I i 19 1 1 I i . -. _ ~ ....e.. I s I 1 j i. I 20 1 .I I I i I 21 1 22 !! K 1 % c\\ g %.e... -.. ..-...,I. r s S'N i 9 NS u 99 3 i l 23 t 8 5 ** t 25 1 1.916 284 i ._n.._.._._.__..... , i l Ca.:inuec Su7000: Of 0 2 '. 2 f T i 'l e OeStin4 26 ?. 2 : '. - 2. R::c r :ncientin; J?R.; ;;roole: pcten:ial i: plan: is v 6 l run en 'ess : nan 3 ?.C?'.s.

ACT '/~~Y STAY 3 Cantinued :es:in:; St.S. relici valves 3 Discontinued valve testing and cooleo down to work :1.S. valves 1 27 4 5 l 1 28 !i l t i 1 -- 29 i i i . i I. a ._u._.-.-.- !l 30 ! -1 I y, Ccemenced heat (>!. S.-R-4 A is installed la >!.S.-R-1 A location) i up of RCS 4 l 3 l I 31 l Sect ed St.S. valve testing 1 .IU::E : Continued St.S. valte testing .I 4 1 II i 1 l A Y t i* i3 j T-g m 9 1 3

i. i I'

l '. 1 !? l i i i d 1 (i f I

Discentinued
!.S. valve testing, cc renced cooldown 1

14 '5 1916 285

.iCT ,' ITY i.g.: .w., 7 JL::E' 5 Plant cooled down and holding pending resolution to M..S. relief valve 1 evalutation i 5 t I i i 6 1 4 7 i 1 i i 3 Co cenced cooldown for Rx Head removal in anticipation of 1 removing ORA's & installing 3PRA Holddown devices r - r 9 I 1 i j 10 1 M# gy o k 11 ? .ess.11 head 'wlts detensioned i ,6 9 q I 12 l 1 i .._.......I.. I h head recoved 13 I 1 i l'916 286 i .. ~........ ,F e

ACT !'.* !TY ru.. 4 .!L':;? 6 Care plenua removed preparar. ions for incore work 14 1 i .411 40 OR.\\'s recoved...... -....... - - 3 l 15 1 i 16 All EP?l holddown devices installed, verification in progress 1 t .I Con =ercial plenum installation 17 l 1 I f .........r------------------------------- i 18 1 t 4 4 I (Firming up plans to codify M.S. lines for new valves) h 19 q$ $<h j O e .n: 12.;;:11., :::. rq cd D 20 1 9 GM .ill M.S. valves removed from steaalines 'l ~ s i 22 l Rx vessel was overflowed without head torqued, spilled boric acid over 1 I -vessel flange & C-rings preparing to lif t head & clean up i e f-s. 1916 287

.\\ C u... I.,., - eu.. - JL... .. :. o i 3 3i i t 24 1 23 1 i ! i i Crouse commenced to =ake =odifications to M.S. lines. I 26 1 l i l S' l-g %}$)h. __ 1 27 l' 93 l I . g. i i i O 23 9 i I t 29 l 1 i t ! i i I 30 l t 1916 288 ww-. L 4 a. ._e ..._.._.-=w. ...ma

ACTUlITY i Tu.... JtLY E Con:inue codifications to it.S. lines 2 l i t 3 1 I i. I i 1 4 t i t t l t l 5 1 i f I 4 I i 6 1 6 .. m i A 1 7 v 9 / 3 1 6 i i 1 l 9 i 1 t. ~ i l l-4 4 4 I9l6'289

ACTI'J ITY Trca:. - JUL.. o' Cantinue !!.S. Line codifications i 11 i 1 l l I 't ^ 'i i 12 1 .I t i i l l i 13 } i i' l t t j ! i

  • -i 1

y l' I 1 l ~ l t I d $YhY 15 i 9 l t-- ,k*

5 i Ps 11ead Tensioned 16 t

I 17 (Func:ianally tested cable reca halon sys:cm sa:.) 1 i i 13 i .s 1906'290 l, i. l a. 8 i ? '1 1

1 ACT!VITY ,N JULY b. Continue 't.5. line modift:ations i g 20 i. i l L. .c. I c f 1 21 i } t, ese l. !i g $ _9 l i i k 3 I i 1 1 23 l 1 f,! l t 'l l t

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j { hI 24 i i l! i t l ~ -4 9 t 9g i l Y i 9

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6@Mu a @*ee Weemi eyaw. g.g g g g p a,ggg g ,, g g gg, I .i. [ I 0 t b 4 l e I i i t 1916 29)

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ACTu'ITY na.., s JCI Y 5 Continue ".S. line modifications t 1 29 l I i _. t _ - ..L'-- t I 1 3 1 i. I 1 31 i a i i i ACG. [ 1 1 i i I. I i. 2 t 1 T# 6 9 gg% ~ S I l l 2 i 9 e 5 I I f916 292 t t _.4.

===---*=e--.- = .g

ACT lITY .,..ir,- ~ AUG. i Continue M.S. line codifications 1 7 i - - - - - - - - - - - ~ ~ - l 8 (Conducted retest of RR-P-13/C/D-Sat.(License condition)) l i i f I 9 I i 1 1 i t t 8 I 1 10 I i 1 11 n l N'c 1 l 14 y e e g nR C j S A hM g 13 l u... _..s.---..-....-----..-------- 4 6 i 1. Cr:use begins work on ne; '!.S. line hant;ers 1 I I l I .... {. -= = 1916 293

ACTIVIT'i ...~...- - n,.. - AUG. 5 Con:inue it.S. line nadifications 1 16 i -- -.. i -..... t 17 1 1 i. 1 l. i i 1 13 i ? i i I 19 i i i t 20 1 I l I 1 I n i Begin preparations for M.S. hydro -.t 4, i h. b #, { %e w ' ne, % o\\% ~ \\ s-t i I s 9 1 1 j 1916 294 t 2 e v i t _.i I COnciated '!.S. 3ifa:7 talve modifictti:ns y de;an .cra af ".5. .ines. "alve ;a.;s aid nc: '. .d. 1 2a I

ACTIVITY AUG. 5 Continued working on hydro gag pro'clems 25 1 26 M.S. relief valve gags re-installed for hydro. Completed Sat. I concenced draining OTSG's & Steam lines. s Crouse continues M.S. line insulation & new hangers. 27 Cc plated restora: ion of secondary af:er hydro 1 Cc=cence heat up for lift testing. t 4 3 {I f1 23

  • i^

I Star:ed lifting M.S. safety with h'ydro set l Started lifting valves with steam pressure to check blowdown ~ 29 I 1 t l i Con:inued safety valve testing 30 1 i l Comoleted s:tiety valve testing & removed all gags i Performed full flow & coast dont tests of RCS 31 i Co:cenced cooldown for M.S. hanger installation (continuation) h (F> y yy SE?. 'danpr installa ica continues 9 I; 9 o ..s Q ! I lla n 3'e t installa:Lon continues ? h .) 1916:295 ~ j 4, .-._i _ - _-- ... - ~. o

A CT I'.* I!*f -- _;. s m, - SE?. 5 Hanger ins talla: ion continues 3 1 i

i l

{ I 1 4 I e I I 5 1 f f l s l l i. I 6 i

  • 1 I \\

i t t ... -- 2.0. _ 4 2 I l Cleaning up feedwater to spec. x3 3 i j' Comeenced plant heat up i j 10 f* C'ntinue plan heat up 1.'91 296

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ACT I*.* I T ... ~ - -.. - - -.. - -........ -..... -... - - - - - -.. - - - - -. ~-

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~~ S E P.' 3 Cantinuin.; cooldoun k i 12 5, 1 1 i l liegin repairing RC-V-3 i 13 'l 1 i .Coc=enced heat up, RCS filled & vented >{, ', CRC:1 venting co=pleted t l 14 4 1 3,,* i.RC-V-3 tested OK Found steam laak on ' B' E.F. recirc line - cracked fitting not isola, ale ( 4l Coe ence ecoldown to fix. 5-15 l j 1 1 e 8 t i Repair completed, commence heat up 4-3 G fContinueheat 17 up to power p 1 2. Pefotted reactimeter testing Sat. O. Y-l.10. 13 i 15 T.C. on line (103 lMle) g'd 2.' 3 i ; ' 3eg in l'7. with 4 RCP's Q / i .... - ~.-- -~ l ! l l I'e r f a r: ed !3% power testing 19' ,t I._ L 'Perf orced shutdown outside control ecom tes t i 3 e : ICS being tuned 15' data reviewed prior to power escalation i f1 13 ba ck on 'iino ( 183 '.*.?E) 20. I 20 Rn trip en lii Press due to lost cond. 5 F.'.,*. pumps hen blowing dcwn cond. 2/3 t booster pump. J 2 j?ac;*zer f-m tria '.-. Tr :ine va line-1916 297

ACT!'.* I M*

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  • :... -

SEP. 1 15 Rx trip due to F..i. instability in startup range (Trippec on Hi pressure) 3 21 2_ Recover from trip 1/3 1 LO Turbine on line

:30 Progressed thru startup range, settled out oscillations

! 34 ReducinoDpo..we..r. f or 30*. Turbine trip test ! 30 30*: Turbine trip test I.15. 22, Manually tripped reactor to com=ence cooldown for work outage 2 i 3 ', Exceriencing eroblems backflushing co-cumo strainers 1/3 4: Also fixing MSR crossover piping steas leaks (fer=enite) and 5' leaking conax connectors and bad RTD in RCS hot leg. 23 1 = f I I o a 24 I 1 i 4 ,Cc= ence heat up 3 t i g. 2 25 170' I j a Discovered steam leaks in M.S. cross under piping due to vibration 3 Rx trip on Hi pressure due to C0 pump trip when reducing load to fix leak. St._am leak re gired e 2 Recover from trip 1.'O T.G. back on line 26 F'.1-P-13 has broken 1CS controller, FU-P-1 A, Hi vibration 1' Can' t escalate in power until resolved. p . U FW-?-IA cot. ed, tested Sat. N D Ccemence power escalation to 40*.' ( i N 1 y ).'S o u Keeping F.W. in spec. as power increases i ...... b 5.

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!! I l 40 ' 40.~. powe r te sting commences f' I !: I _ Repaired FW-?-13 controller P ; '., testing in progress 1916 298 -v. I i I 6

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! l I Bo th F.'I. strings in service 4

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l I i1 5 ' 407. Rx trip test performed 3 Commenced cooldown for work outage s 4.a r,u a.; ud pump,sucmion vatve a stra ner prooien _ b. 1, 3 Inspecting main concenser (- l (E Inspecting main condenser I t i i n. ~ i s Inspecting cain condenser s i t 3 1916~299

aC....., s ...r.t ~c u. - .u.. .~ 0CT. 5 ~F ~~ 'Co cee nce =.'.4. clean up 5 l e. 9 l 'F.CS heat up 1 4 ................ ~ - - 3 i 10 i . ioted leak in Conax connector 1 4 Com=enced cooldewn to fix. 5 RC-V-1 would not operate 11 Completed conax repairs 1 l. t y. Co:pleted CRC:l venting, begin heat-up + = - -. - - - ~ - ~ - ~ ~ - - - - ~ ~ - 4 Co=pleted electrical repair of RC-V-1 3 i 12 ' 1 I. , ~.__ 1.10 Turbine on line .g i 21.?tC-V-1 inoperable AW- .' j, ~ 15 i A @N Oh 13 f 2 Found 5 fi:<cd loose connection on RC-V-1 h 1 1 20 Turbine on line A- ---- ..-.. - ~ __ . ~. - -_ ~

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! :s ' 3 V 14 ~:t< trip on Icw pressure following turbine trip aecover from trip 1 i 2 1 20 Turbine on line s (.40~aperiencingF.W. p oscillations & Fe plugging Ca pump strainers i 15 i l ?.ese tting M.S. sa fe ty valve s .i Heater Drain systca being fine tuned i 1 3 tuntng continues !lI 16 4 Pe rf a rs 4 th PCP t rip f rom 40". i !i 1916..,00 3 0* 30 All 40~. power testing ccepleted. 3 15 Repea: 30." turbine trip to collect tissing data due to inst. pr:blen 1,,. 1.s .t a,.ast test v

15 Cc.rr.ence power escalatian ta ~ i*; pcwer

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.o ACT!'.'!!Y T -. O r ~ OCT. 1 Power escalation in pro.gress 50 (Fe in system limiting escalation rate.) 18 52 discovered leak in FW-V-17A 3ody r/ HD drain tank level controller problem has 1 HD pump out of service, further 3 53 delayin; escalation rate 57 t/ 19 58 H? level control valve fixed, both HD pumps back in service C 31o'n packing on MU-V-17 (Switched to bypass valve) 63 ' 67 20 ij ' 75 Turbine trip due to phase comparison relay on generator output <3 15 , g -- ~35 . 55 21 1 63 Turbir,e trip-repeat of above j 15 20 33 70 o i ' 75 Holding Stable for equlibrium zenon 22 ~ l 23 Comnenced 7 5,'.' power tes ting <'8 ^1 t t i -. - ~. i. n 1 i

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ACT I'7 IT'* c.. OCT. 1 75 Comple ted 7 5; poue: t.9 sting 27 Received T'..'C approval to escalate to 1007. plateau with intermediate testing I at 90 90 ::oted vibrations on turbine generator exciter bearing :o. 9, took turbine off line f 28 Ccamenced work on exciter bearing i e i I ( i-I' 8 ,/ 29 3 9 Experienced ratchet trip of CRD rods of group 5 when reducing power i i _........C A tinuing to repai.r exc.i.t.e..r bea.. ring-. - - 5 1 i (Iapair of RC-V-1 in progress - replaced operator gear box) 30 [ t .. - - -. Replaced faulty relay on CRD s which had caused ratchet trip -.-----.---.-.A l, ', Continue work on exciter bearing repair l s 8 v SCV. 2 ' Completed repsir af exciter bearing. s 1 15 Replaced gaskets on erciter cooling lines. g{ 7 I i l* i l i 70 l 90 (Zxperienced condensate booster pu=p trip on low pressure, Rx run back to f 2 53% and recovered.) Coceenced 90?. power testin.;. I i i l i 3 1 !I Rx trip on high pressure alter operator inadvertantly turned of f power sup-l4' 3 0 ply to condensate po11shing system tripping condensate booster and feedwater ptmps !y Repaired RC-V-1 11.it switch I r e / i / 's [7 Rerairs completed, retur.ed critical )h b .. ' I

ACT!'l TY s _,c -..

0V. ! 13. Returned to power operations, began escalation to 90%

5 . '90 Waiting for xenon equilibriu:2 1 .-.?..'....... !l / 6 i l e s.- Commenced 90% data collection. I l Rx trip on variable low press. af ter heater drain pu=p tripped causing I. 7 3 0 feedwater pump trip, low press. caused NaOH injection. 3 RC-V-1 again needs repair - operator to be changed out. i 11/7 trip evaluation in progress i t 8 l Decision made to begin turbine screen outage on 11/11 in parallel with [ plant cleanup. 1 - l l i 9 "S-V-26A yoke found broken (probably caused excessively low press. and / resultant :TaOH injection on 11/7) RCS cleanup in progress a ... ~.. ~ ~ -... -. - - - - Concenced XCS cooldown 4' Replaced gaskets on exciter cooling lines. / 10 5 l Turbine screen oui. age in progress l l \\ V i y 9 12 l i 13. i l i ( EF's bell reducer cut: frc~ line for replacement) I I Ii 1 1916 303

I 1 ACTIVITY 7._. ,.e.

iOV. 5 0 Turstne screen removal outage continues l!+

t - ~ ~ ~ ~ ~ ' - ~ ~ - I e h 15 16 9 T I e 1 i t I e h e e h e s 1 8 h @g) g hN e 9 e e 1 i i i e 1-916 304

. Attachment to Encl. (2) Analesis of.ictual Critical Path Shown on Enclosure (2) 1. Cc=pleting URC Operating License pre-requisites delayed issuance of the license by 3 days. The Operating License is required before fuel load can begin. 2. Fuel transfer mechanism problems experienced at the onset of fuel load- 'ing is not shown on the critical path. Although it was a factor causing the fuel loading activity to be extended longer than it might otherwise need to have been, fuel loading was, in fact, completed in the number of days allotted. 3. The Reactor Coolant Pu=p 2A clutch failure which occurred on.3/13/78, caused a delay of 6 days of pre-critical testing. After 3/19/78, the test program was resumed with only 3 of 4 Reactor Coolant Pu=ps in operation. This resulted in additional pre-critical testing on 5/15-16 to complete test require =ents with four pu=ps in operation. Installa-tion of the repaired clutch in early May was accomplished in parallel with the repair of the bellow liners and therefore did not affect the critical path. 4 Steam Generator Instrumentation "Conax" fitting leaks have been a continuous probles throughout the test program. Generally, the leaks occurred due to pressure / temperature cycles of the system caused either purposely or inadvertently. Therefore, in many cases, this proble arose in parallel with another problem and I am considering, for the purpose o: :his review, only 9 out of a possible 23 delay days attribut-able to these fittings. 5. We have experienced three inadvertent sodium hydroxide safety injec-tions into the Reactor Core. One on 3/29/78 caused a delay of S days to the zero power physics testing. This first injection recovery delay was further co= pounded by chloride contamination of.the Reactor Coolant Systen due to the use of Lapure Sodius Hydroxide chemicals. The second injection on 4/23/78 is not considered critical path because of the over-riding bellow liners recovery program which also resulted from the 4/23/73 transient. The third injection on 11/7/78 caused a direct critical path delay of 4 days to the power escalation to 100% activity before it was decided to connence turbine suc ten removal. On 11/11/73, therefore, the screen outage became controlling. 6. Nuclear Instrumentation /Reactimeter problems caused an approximate 8 day celay to scro power physics tasting and 3 mora days during 153 power testing. 7. Various Condensate System Strainer / Pump Suction Valve problems caused further delays to 15% power testing (2 days) and' power escalation testing (9 days) and was the prime reason for an eight day work outage perf arned in October. Problems in this area caused several plant trip.. 1916 305

t 3. Although failure of the Main Steam Safety Valves to function properly on 4/23/73 was the root cause of safety injection, I am considering only that period of time from 5/13/73 through 9/1/73 (105 days) as being critical path due to safety valves. The reason for this is be-cause, even if the valves had functioned properly on 4/23, the valve di. charge line bellow liners would have failed and the time it took to fix them, from 4/23 to 5/10 (17 days) was controlling at that point. A subsequent problem with oil leaking from two Reactor Coolant Pump motors caused an additional 3 day delay before plant operations could resume to the point of discovering that the safety valves could not be adjusted or modified to function properly. 9. Removal of the Reactor Internals Orifice Rod Assemblies and Burnable Poison Red Assemblies was not critical path since it was performed completely in parallel with the steam line modifications. 10. In September 1977, it was discovered that Main Steamline snubbers were not provided to accommodate steam ha==er vibrations. Apparently when Gilbert Assodiates (GAI) performed a. steam hammer analysis ~ for TMI-1, ir was desided that GAI should also do a similar analysis for TMI-2 because, at that time, B&R did not have the in house capa-bility to do it themselves. In that same year, Section 9.3 of the TMI-2 FSAR was written indicating that: "In the Main Steam System, special attention was given to the dynamic effects of the fast closure of the turbine stop valves on the piping between the steam generators and the turbine steam chest. Hydraulic snubbers are provided to minimi:e stea= ha==er while allowing normal system thermal move =ent." (Subsection 3.9.1.1, 2nd paragraph.) These words were written prior to the analysis based on the assumption that due to the similarity of piping arrangements between Units 1 and 2, snubbers would be required on Unit 2, as on Unit 1, for the suppression of steam hammer effects. The actual steam hammer analysis was to be performed at so=e later date after 3SR finalized the Main Steam piping arrangement and the location and sizing of thermal and seismic pipe supports and transmitted that information to GAI. For reasons unknown, there are no records of either the agreement to have GAI do the steam hammer analysis or of B&R ever having transmitted the required design infornation to GAI. As a result of this apparent oversight, no 3ccan hammer anubbers were designed or provided for in the early days of the projcct. The lack of such snubbers was first noted by inspection during Hot 7unctional Testing in September 1977. Since 1972, 3&R has developed the in house capability of performing steam hammer analysis, so upon discovery of this omission, BLR was diracted to proceed with the analysis and design of additional snubbers as required. ~ Because fuel loading and startup testing was scheduled for late 1977 and early 1973, BSR was also requested to calculate the maximum power level the Unit could safely be operated at without having the new snubbers installed in the event lead time for material procurement became critical. B&R subsequently estimated that power level to be about 307.. (3SR letter 3934-G?, 9/13/77 and Project change Notice 2457 refer.) i f916 306

4 B&R perf ormed the steam hae=er analysis using as the basis for the analy-sis a 130 milliseconds turbine stop valve closure time. That infor:2-tion was received verbally from Westinghouse on 9/6/77, with confirma-tion trom Westinghouse by =emorandum of the sa e date. The acto also included a flow vs. closure time curve purported to be typical of these valves which tended to support the use 'of 150 milliseconds in that the closure was fairly linear with respect to flow decrease. The results of this analysis showed the need for additional snubbers on steam line piping in the Turbine Building area. (Existing restraints provided fcr dead wei;ht, ther=al and seismic loads were found to be capable of accc==odating the added steam hanner loads in the steanlines routed through the Control 3uilding Area, except for one seismic restraint with a PSA-35/35 kip snubber which was upgraded with a PSA-35/50 kip snubber.) Material for these new snubbers was received on site and inst'allation. co==enced around 3/1/78. Installation was ce=pleted by 4/15/78, prior to reaching that point in the Test Program requiring power levels in excess of 30%. Therefore, these snubbers never became controlling on the critical path. In January 1973, the General Office Review Board (GORB) reviewed the steam hammer analysis snubber problem described above and concluded that since the analysis, design and installation was done on an expedited basis at the end of construe _on, an independent review of 3&R's efforts was warranted in view of the importance of the snubbers. Met-Ed was requested to perform this review. Meanwhile, 3&R had been attempting to justify the use cf an increased turbine stop valve closure ti=e of about 200-250 milliseconds (which B&R had seen for some other plants) in order to de=enstrate additional conservatism in the snubber Jesign because one of the snubbers in the Control Building Area was marginally capable of sustaining the calculated loads and increased pewer levels without snubbers installed were being sought. To do so, 3&R requested Westinghouse to check the information previously provided (150 milliseconds) to see if actual test data was available. Westinghouse prepared a flow vs. closing time curve and tele-copied it to S&R on 4/25/78. This curve, based on test data, differed from the previously received " typical" curve in that it no longer showed a linear relationship between time and flow. Various interpretations of this curve demonstrated closure times anywhere between 50 to 150 milliseconds. On 5/17/73, when Met-Ed first met with 3&R to review the steam ha==er work as requested by the G023, this new curve was discussed. No agree-ment was reached as to the cost appropriate closure tiac to use in the steam han=er analysis at that time. During the next several weeks it was concluded that, based on the new curve, an effective valve closure time to use for an appropriately conservative steam hammer analysis of the main steamlines should be 30 milliseconds vs. 150 milliseconds actual closure time used in the previous analysis. Met-Ed made this reco==enda-tion to CPU on 6/12/73 (GEM 2544) and G?U forwarded the reco==endations to SLR (TMI-II/7025, 6/23/73) requesting an evaluation of the ef fcets of W' ~ 1916'307

9 the new closure time. On 7/13/78 B&R discussed with CPU the results of seme limited ec-analysis using the 50 milliseconds closure time. This re-analysis showed that the loads on the main steam seismic snubbers in the Control Buildin; Arca increase by about 507., which was in excess of their rated espacity in most cases. B&R recommended that additional snubbers be purchased chile they completed the analysis and design of new or upgraded restraints. (The steam hammer snubbers in the Turbine 3uilding previously installed were found to be acceptable without modi-fication.) A;ain, 3&R recommended limiting plant operations to no more than 302 power until the rastreints were modified. (At this point in time, however, the plant was shut down for modifications to the steamlines for new safety valves.) These recommendations were forwarded to GPU by latter on 7/26/73 (4249-GP). The new seismic snubbers were subsequently designed, fabricated and dot tvered to the site by 8/14/78. Installatien was not completed by - the time the Main Steamline/ Safety Valve modifications uere completed on 9/1, so the balance of the installation became critical path and caused an additional 7 day delay to the Test Program. (It should be noted that the snubber installation required the steamlines and surround-in; work areas to be cool, so no plant power operations could have taken place in parallel.) The following Field Change Requests and Engineering Change memos des-c:ibe the modified or added steam hammer snubbers and seismic snubbers: Turbine Building - FCR 2457.1 - ECM 5899, 5948 Control Building Area - FCR 2457.2 ECM 9047, 9052, 9053, 9062, 9071 11. After completion of the Main Steam Line restraint installat;on, the plant was returning to power operations for resumption of the Test Program when the plant experienced problems with pressurizer spray valves and also dis covered an unisolable crack in a fitting in an Emergency Feedwater lir.e that serves to route water to a steam generator recirculation line. These two problems caused an additional critical prth delay of 6 days. 12. The f ailure of a Turbine Generator Enciter Bearing during power escala-tica testing on 10/27 caused another 4 day delay after completion of the 75': power platlau testing. \\ Conr.iderin; the aforementicned proble=s to be equipment probices causing, or having the potantial to cause, delays to a normal Test Program, it can be seen that these problems, taken in series, could have caused.a delay of about 365 days. In the sequence of occurrence, however, the net d411" is about 201 days. il9l6 308 4 4 m

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Enclosura (*) Delays to T:!I-2 Test, Program Due to Problems Encountered (2/1/73 thru 11/16/78) Delav Davs Actual Potential 1. Operating License Pre-requisites 8 8 2. Lack of >hin Steam Line Restraims for Steam Hanmer Forces (affected S-II portion of lines in Turbine Bldg.) 0 12(1) 3. Inadequate Main Steam Line restraints for Steam Ha==er Forces based on "new" criteria (affecced S-I portion of lines in Control Bldg. Area) 7 71(2) 4. Fuel Transfer Mechanism Probless 0 3 5. Reactor Coolant Pu=p (2A) Clutch Failure 3 16( ) 6. Stess Generator Instrumentation Fitting Leaks 9 23 7. Sodium Hydroxide Injection Transients 12 31 ~ S.

uclear Instru=entation Proble=s 11 11 9.

Condensate System Strainer Blowdown problems 11 13 10. Main Steam Safety Valve Problems 105 120 11. Safety Valve Discharge Bellow Liner Failures 17 17 12. Reactor Ccolant Pump (2A/B) Motor 011 Leaks 3 3 13. Reactor Internala potential loose parts 0 26 14. Pressuri er Spray Valve problems 3 4 15. EEU Fitting failure in Steam Generator recirculation line 3 3 16. Turbine Cencrator "nciter bearing failure 4 i Totals 201 365 Footnotes: 1. Assu=es 30% power icvel exceeded by 4/3/73 based on original schedule. AssumespoUerlevelwasgreaterthan30%on6/28/73whenconcernwasdis-2. covered and plant operations were innediately terminated. 3. Assumes delay tire for :lutch removal and reinstallation plus 2 extra test days af:ar reinstallation. ~ 2 ->i! 3

r . ~. Attachment to Encl. (5) Analvsis of Test Procram Critical Path Assumin; ?!ain Steam Safetv !alves Functicn Proneriv '. ors t Case Referring to the attached Test Program Chronology 3ar Chart, it can be seen that, making the above assunptiens, events up to 5/13/73 would have re-mained the same except that the Sodium Hydroxide Safety injection of 4/23/78 would not have cccurred. Cleanup from that injection, however, was not critical path. After 5/13/73, plant testing wculd have resumed with all feur Reactor Coolant Tunps back in operation. (Reinstallation of the clutch was done in pcrallel with bellow liner repairs.) on 5/26/78 B5U notified G7U of potential loose parts in the core based on recently discovered failures at another plant. Based on the severity of the consequences of such an event, we would undoubtedly have decided not to continue plant operations until the problem was resolved. The Crifice Rods were subse'- quently removed and hold down devices installed on Burnable Poison Rod Assemblies in an expeditious manner, so it can be assumed that the least delay possible to the Test Program would have been 26 days as shown. About a week later, on 6/23 as events actually occurred, GPU concluded that the steamlines were in jeopardy and again would have ter=inated operations (since icwer than 30% power level testing had already been completed) until ade-quate supports were installed. This would have caused another direct delay to the program of 71 days. All other problems actually encountered are considered to have happened at one point or another during plant testing, possibly in the sequence shown on the critical path. The end result shews that, had events occurred as described abeve, we would have been 5 days into the Turbine Screen Removal Outage on 10/26/73 rather than 11/16/73. In other words, under this scenario, the Test Program would have been 20 days shorter. 1916 313

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s, Attachment to Encl. (6) Analvsis of Test Procran Critical Path Assumine Stain Steam S1fer 'Jalves Function Proneriv '!a s t Ootimistic Referring to the attached Test Program Chronology Bar Chart, it can be seen that, nakin'; the above assumptions, events up to 5/13/73 would have remained the same except that the Sodium Hydroxide Safety injecti:n of 4/23/78 would not have occurred. Cleanup from that injection, however, was not critical path. After 3/13/73, plant testing would have resumed with all four Reactor Coolant Punes back in operation. (Reinstallati:n of the clutch was done in paralici with bellow liner repairs.) On 3/26/78 3&W notified CPU of potential loose parts in the core based on recently discovered failures at another plant. Based on the severity of the consequences of such an event, we would undoubtedly have decided not to continue plant operations until the problem was resolved. The Orifice Rods were subsequently removed and hold down devices installed on 3urnable Poison Rod Assemblies in an expeditious ennner, so it can be assumed that the least delay possible to the Test Program would have been 26 days as shown. Since the assumption is made here that the Main Steam Safety Valve prob-les did not exist, it is reasonable to also assume that greater emphasis would have been placed on the ste--line restraint probles at an earlier point in time. Recognizing the fact c..e t, after 5/26/78, the plant was into an extended outage, we would have taken extraordinary steps to resolve the restraint problems at the same time. Therefore, assuming this problem was identified as controlling on about 6/2/78, the critical path would have been shortened by about 19 days frca the " worst case" scenario described on Enclosure (5) due to the parallel activities. The restraint installation would then have been completed on about 8/12/73. All other proble=s actually encountered are considered to have happened at one point or another during plant testing, possibly in the sequence shown on the critical path. The end result shows that, had events occurred as described above, we would have been 5 days into the Turbine Screen Removal Cutage on 10/3/73 rather than 11/16/73. In other words, under this scenario, the Test Program would have been 39 days shorter. e m 19~16 315 '~

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