ML20099C953

Forwards Proposed Alternative Pipe Break Criteria for Immediate Review.Criteria Would Obviate Need to Postulate Arbitrary Intermediate Pipe Breaks.Meeting Can Be Arranged to Discuss Submittal

ML20099C953
Person / Time
Site:	Byron, Braidwood, 05000000
Issue date:	11/15/1984
From:	Swartz E COMMONWEALTH EDISON CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
9439N, NUDOCS 8411200128
Download: ML20099C953 (27)

Text

F Commonwealth Edison

. One First Natrnal Plaz', Chic'go, ilknots

/

Address Reply to: Post Office Box 767

'[

Chicago. Illinois 60690 November 15, 1984 l

l L

Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

Byron Station Units 1 and 2 Braidwood Station Units 1 and 2 Elimination of Arbitrary Intermediate Pipe Breaks NRC Docket Nos. 50-454/455 and 50-456/457'

Dear Mr. Denton:

Representatives of the Commonwealth Edison Company met with members of the NRC Staff on. September 17, 1984 to discuss the possibility of applying alternative pipe break criteria in the design of our Braidwood Station.

At that meeting, we discussed our approach toward elimination of arbitrary intermediate pipe breaks for various piping systems which was consistent with the approach accepted by the NRC for the Catawba and Vogtle Stations.

The NRC Staff was receptive to our approach and encouraged our formal submittal for NRC approval.

Enclosed for NRC Staff's immediate review are the alternative pipe break criteria which we propose to apply to our Braidwood Station, and now also to our Byron Station which would obviate the need to postulate arbitrary intermediate pipe breaks.

Arbitrary intermediate pipe breaks are those break locations, which based on piping stress analysis results are below the stress and fatigue limits specified in Branch Technical Position (BTP) MEB 3-1, but which are arbitrarily selected as the two highest stress locations between the terminal ends of a piping system as required by the BTP.

It has become apparent to both the NRC Staff and the nuclear industry that this particular criterion requiring the postulation of arbitrary intermediate pipe breaks can be overly restrictive and result in an excessive number of pipe rupture l

l protection devices which do not provide a compensating level of safety.

It is for this reason as further explained and justified in l

detail in the Enclosure to this letter that the Commonwealth Edison l

Company is pursuing the application of alternative pipe break l

criteria in the design of our Byron and Braidwood Stations.

As discussed with members of the NRC Staf f during a meeting on November 14, 1984, recent developments concerning the crush strength of the energy absorbing material (EAM) utilized in certain pipe whip restraints at our Byron Unit 1 may financially impact the Commonwealth Edison Company beyond that presented in the Enclosure.

O!00d4!4 Doc PDR

\\\\

up

' d[ >, 1 ~

^

P I

7~

7

,y 8

c

=

W L

5

^

~

+#

,T ~

-3 A

v U M, ' p[

,i6 n

j?

r

.g.

s s

]

4'V

~

l

~

y s

, W ' y. -,

n w.-

y,.

s y,,

y

H[R. $Deriton t-E2; I

• =

Novembe 15,519.84;

~

y

,Q+

p

}.

,m.

o x

(

- 3 x

.c 1, '....... _.

.n s

s

i...

.z-

We-Jhavoidecided

to; replace the potentiallyidefective:' E M ?conta'ined,N

^

yin 1213 pipe; whip 3 restraints at' Byron Unit,1 which!areLtechnically; ' '

~

frequired and whichica'n;be'replacedfwith'relativeieaseh However,(

there are"42) additional pipetwhip-~ restraints with potentially; idefective;EAM;at?arbitraryjbreak-locations.

Expeditious:NRC E f.

impprova1Yof thisyrequest to ? eliminate arbitrary -intermediate :breaksi '

2

' would-resolverthe?EAM concern for?these:42, whip;; restraints withinoJ u-JfinancialJimpaction Commonwealth Edis'on Company 1However,(if: timely; m

y

.NRC' approval'offthis1requesttis-not granted, alternateLplans/toi j

2

'resolveEthe<EAM;issueJinclu' ding'further analysis.and; probable...

V

-replacementDofLadditional EAMLwould be required.; : Implementing these.

~

~3 9'

alternate' plans' prior-to completion-of;the Byron Uniteitstartup.

1, program, if;requiredrwodld delayithefcompletion of.the~'startup; m

g

-programtand;cause a financialsimpact.on Commonwealth Edison'

.JCompanyt Therefore['it;is imperative.ithat-the NRCJexpeditiouslyq y

- review - the' Enclosure and. provide us with' approvalcand.or; comments _as; soon,astpossible.-

r w

-Attachments A and B? provide a: list-by piping system.:of th~h

~

O ASME Class 1,: 2 and l3' piping : intermediate: break locations which are-

~

candidates to'be eliminated. -Attachment C-provides;the1 technical jusi.ification for the employment of. the' alternative pipe break,

~;

criteria.- ' Attachments D 'E F 'and H provide detailed descriptions' of-our provisions'forzmin,im}.zin,g' stress corrosion cracking in high"'

energyL11nes, minimizing the effects of thermal and:v'ibration J

induced pipin'g fatigue, minimizing steam and water hammer-effects, and minimizing local stresses from welded attachments.-

Attachment I' provides a summary of the benefits derived from elimination of-the arbitrary' breaks.

Attachment G is provided to supplement Attachment F and provides.a detailed-discussion of the water hammer. prevention features of our main ' feedwater system.

Although we-recognize that the NRC has not approved the elimination of.the intermediate breaks in the'feedwater system at other plants due to concerns with water-hammer, we believe that our Byron and Braidwood plants have adequate provisions for minimizing such affects to justify their elimination.

l

.. Itais'important to note that the Enclosure is based on j

Byron' Unit 1.as-built pipe whip restraint locations and final' pipe' s

-stress analyses.

While the actual number of' pipe. whip restraints and specific break locations are finalized on Byron Unit 1, they are-not yet finalized on Byron Unit 2 and Braidwood. Units.1 and 2.

We therefore request that:the NRC review and approval.of.the application of the alternative pipe-break criteria on Byron. Unit 2 andLBraidwood. Units'1 and 2 be in terms of piping systems and methodology, an_d not in terms of the actual numbers and locations of pipe; whip. restraints.

~

_.--..m.....,.~~,...,____.2-A

rp. -

.7-

,, m li..

.'M'

.r

-r

7

+

s

...-t 3

M'.

}

~

W*

H. R.LDenton

-l3'-u

-November 15, 1984<

1

Although we'discussedLths enclosed submittal-with members ~

U of,the1NRC Starfiat.the. November 14,:1984 me'eting, we,are.available-

'to, disc'us's-this submittal.in further^ detail.'as necessary..-Please-advise ~-thisl-office as'soon as possibleias;to your intentionstand further" requirements =in:this matt'er.

'-Very trulyLyours-g

's.(

n

.y.

E. Douglas Swartz Nuclear; Licensing Administrator

-EDS/1m-cc:

J.'G.

Keppler - RIII J. F. Streeter - RIII J.~A.

Stevens - LBI R. J.-Bosnak - MEB 9439N

.y

.,y

.t

q. 7, '

1 g..

V; 4,

ag 4

n

! BYRON /BRAIDWOOD> STATIONS I'

~ ARBITRARY INT ERM EDI ATE PIPE BREAKS.-

v i

Commonwealth - Edison Company -f( Ceco) l has' f_ollowed. closelyf the 1

= recent1 activities :of ' the1 Nuclear. Regulatory : Commission ;(NRC):

F

."staf f and the nuclearnindustry. related to the $ treatment 7 of. design basis pipe : breaks,in high; energy pipingisystems.. ' Incparticular, ittis'notodithatithe.NRCfstaff:has expressed an: interest?in the:

industry's3proposalitojmo' ify the current: pipe : break ; criteria 1to d

eliminate from design' consideration those intermediate breaks'

generally. referred to as1 arbitrary ; intermediate breaks, ii.e.,-

thoseibreak locations which,. based on stress analysis,iare;below

the stress ' limits and/or tthe ' cumulative usage : f actors specified - '

in the current NRC;criterialibut,are; selected _toLprovide;a7 -

minimum of two breaksibetween terminal' ends.'

NRCistaff and i

industrys discussions 1with.the Advisory CommitteeLon Reactor

~

Safeguards 1( ACRS) con March 129 and June 2, ' 1983.have findicated ?

general agreament withothese objective's 'and recognition' that ?

F elimination of;theJarbitrary intermediate breaks offers-considerable-benefits due to the' deletion of the associatedipipe r:

whip; restraints and otherfprovisions currently incorporatedain plant designs to' mitigate the ef fects of such breaks.

~

}

The break selection cr'iteria currently employed by CBCo for-the-Byron /Braidwood; Stations is.taken-from-NRC Branch ~ Technical I

Positions ASB 3-1 and ~MEB 3-1.

'These documents requiresthat pipe

~

breaks be considered at tensinal ends and'at intermediate f-locations where' stresses or cumulative usage factors exceed j

specified limits.. If.two; intermediate locations cannot be determined based on the above, i.e. stresses and cumulativeLusage 4

factors are below specified limits, then the two highest ' stress--

locations are selected.'.

i CECO concurs with -the nuclear industry in the belief - that current -

knowledge and experience supports the conclusion that designing for the arbitrary intermediate breaks is not justified and.that i-this requirement should be deleted.

This conclusion is supported by extensive operating. experience in over 80 operating U.S.

11 plants and a number of similar plants overseas -in which no piping f ailures have been known1to occur that would suggest 'the need to design protective features to mitigate the dynamic effects of.

i arbitrary intermediate breaks.

Arbitrary intermediate breaks are often postulated at locations where~ stresses are well below the ASME Code allowables and within a few percent of the stress levels at other points in the same system.

This results in l

complicated protective features being provided for specific break 9

locations in the piping system that provide little to enhance j

overall plant safety.

i In practice, consideration of these two arbitrary intermediate j-breaks. is particularly' dif ficult because the location of the high.

[

stress points may move several times as the seismic design and i~

1 i

--.--_----.-_-=..-a..--,

.~

s y

H N..

4. ymMWR Ms

~

a m

ay d' -J?

G n

q T.

O analysis of ' structures and. piping develops. 1The 'industryJ..

recognizes. that :the : revised ' M EB 3-1., ;which ? was included in the.

July :1981, revisions' to the Standard; Review Planc (NUREG-0800),

provides -criteria (for. : noti having L toirelocate intermediate > break

_ points when highest' stress : locations;shif t as airesult of, piping reanalysis..'As a practi' cal matter,/however, tliese criteri'a *'

provide little: relief,;since_the burdeh.is on' the desfgherTt"di

prove that, not ' postulating breaks at s.relocat.ed highest stress 1 s,

points -does not degradEsafety. -. ThisLa'apcrequite extensive w.;

j

- additional. analysis;ofibreak/targetLinteractions?forithe-

/" _

g relocated-break points and could resultiin' design,1 fabrication

' and installation of: additional _pipeLwhipirestraintsjat the:.

relocated break points ~and elimination of.previously installed restraints 'at abandian'ed break point's.' ~ Early determination exact break locations is'quite'important because of~al,1 the(of-4 y

. secondary effects of'the pipe breakito be considered'. f c:

> ~

' The benefits to be realized from othe delimination of1ths arbitrary -

intermediate break : locations center primarily 'around _ the a

elimination of theJassociated pipe whip.restraintsLand other- '

.f7 structural provisions to _ mitigate the consequences of -_ the'se 4

i breaks.

While a substantial' reduction in capital-costs for these' -"

L

-restraints and. structures can be : realized immediately, there are,~ '

also significant operational benefits tx) be realized over,the-40-year; life;of each plant.. As identified in NUREG CR-2136,

.);.

these effects are particularly in the areas-of overall plant i

reliability and -exposure of plant personnel ~ tx) radiation when.

~

excessive pipe whip ' restraints are installed.

s,e s

)

' Access during plant operation for au'ch" activities ~ as maintenance Y

and inservice inspection is improved due to the elimination of.

congestion created by these restraints and the supporting structural steel,' and in _ some cases due tof the need to remove some restraints to gain acce'ss to welds.

In addition to the decro in maintenance effort, a signiilcant reduction in nuOO-

}

rem exps ire can be realized through. fewer manhours spent in.,

~

radiation areas.

Also, the need to verify appropriate cold and -

hot clearances between pipes and restraints during ' initial.

l heatup, which. requires additional hold points during the startup phase, can be, dispensed.

I

~

. Recovery from unusual plant conditions would al-o be improved by elimination of this congestion.

In the. event c a radioactive

- release or spill inside the plant, decontamination operations would be much more.ef fective if the complex shapes, represented i

by the structuralaframeworks supporting the restraints,.were i

t eliminated.

This results in decreasing man-remlexposures associated with decontamination and restoration activities.

~.

similarly, access for control of' fires within these areas of the 1

J

\\

u

~

s c 4

y-yr-e

---

w

-v*--*-e----*Fa-e g

we w ay--

-ew e

c**e e

e-iagi-"a eMw-e M T ie me e wea b e-

-e ermwwg*

e--ee er-e e e ih

'me-yP=w-*

• we Wm w-eelme-gur-pie se e ac--mMe-m---hm-rw w eepit er'

~

+

r:

s plant - would. be improved,Jespecially. under ilow visibil'ity.

1 conditions. -Substantial overall benefits in these. areas would.be realized by Lreducingf the. number c(f[ whip restraints required..

By : design, whip ' restraints-fit ~ closely around ithe high energy

piping with gaps typically;being on the ' order of ' half-an" inch.

These. restraints and _ theirf supporting ~ steel increase 1 the -heat :

loss to ;the surrounding ' environmentj significantly., L Also, because

' thermal movementnof the piping, system during. start-up t and :

shutdown-~could deform the piping--insulation against: the fixed 1

-whip restraint, the insulation must,be. cut back in these areas',-

creating; convection gaps adjacent to,the restraint,. which ' also Lincreases. heat lossL to the environment.. This is a_ major:

contributor to the tendency 'of many ; containments to operate 7at -

temperatures near_ technical specification limits.- -The elimination of whip : restraints. associated _ withf arbitraryJ vintermediate breaks.would assist in controlling the normal-

~

environmental -temperatures and improving. system operational, efficiency.

For the above re'asons,.CBCo. requests NRCL approval fof the...

following -fo'r the application of alternative. pipe break criteria which twould ' eliminate the need to -postulate arbitrary intermediate pipe ' breaks, ci.e., --those _ break' locations which, based on stress' analysis, are below the stress limits and the cumulative usage _ factors specified in the current NRC criteria, but are selected to provide a' minimum of two breaks between terminal ends:

ASME Section III Piping 'Inside Containment o

Piping systems shall. be designed to accommodate pipe

~

breaks at terminal ends and locations where the stress or 4

usage f actor' criteria of MEB 3-1 are exceeded.

No arbitrary. intermediate breaks will be postulated when the.

~ stress and/or usage factor criterion. are not exceeded.

L o

For breaks-that must be taken, the design will accommodate pipe whip, jet impi ngement, and compartment pressurization i

resulting from mechanistic treatment of the break.

i Current acceptable methods for limiting break opening, moderate and low energy exclusions, limited duration operation, etc..may still be applied.

o For flooding evaluation, environmental qualification of.

i equipment and structural design of areas traversed by-high.

q energy piping systems, breaks will continue to be postulated in accordance with the present project

criteria, i.e., -in each area traversed by. the high energy piping system, non-mechanistic breaks are postulated at-the location that results in the most severe environmental conseque nces.

Therefore, elimination of the arbitrary l:

-3 i

2. ---_, _ _._ ~...-_ -.__ __ _.-.__ ~

-m-

, I

._'(>

b g

e "

'- [

T z N- ? <

' % ~

~

e 3

a x

. f*

~

g 3

.e c..

i

,s 6

Kintermediate breaks will? notJimpact.the 9 flooding:

evaluation,J environmental qualification program -or plant"

! structural ^ desig n.4 J ASME'Section III and' Seismicallyi Designed-- Nor,- SME Section

?

.III PipingtOutside Containment

- y 1

(o' LPipingi systems ~.'shallLbe T designed : to' accommodate " pipe.

' breaks 1 at J terminal fends'and -locations: where f then stress criteria of c MEB.-3-1 areL exceeded. -Noiarbitraryi_

~

! intermediate; breaks willibe. postulated; whens the' stress:

criterionDare not? exceeded ?

o : l For breaks.thatt must be ' taken,-' the Ldesign will accommodate ;

pipe whip cand. jet ; impingement fef fects L resulting J from'

~

~

mechanistic: treatment;of the.. break... Compartmenth

- pressurization.c andj flooding > ef fects/ from : breaks Lpostulated:

+

' int accordance. with M EB 3-1 will? bef accommodated: inithe -

design.: Current-acceptable methods /forJ11miting breaku opening', moderate andilow energylexclusions,111mited;

~

durationioperation, etc._may.stilltbeJapplied._

o. For environmentalL qualification of. equipmentgand structural l design of. areas traversed.' by high; energyc piping?

systems,_ breaks-will continueE to;be7 postulated in

_accordance(with'the_present? project' criteria,Ji.e.,,in-3

~

F 1each ; area traversed by the :high.- energy piping T system,J non-mechanistic breaks are postulated at the location that.

~

results'in'the most-severe:environmentalEconsequenc_es.'

Therefore, Lelimination of1the arbitrary ' intermediate 1

breaks will not: impact _the environmental ~ qualification s

program or plant" structural ~ design.

Applicaton of the alternative pipe break criteria described above4 i

will'not alter'the'commitmant to quality in the design of '

structures, systems, and ?? +nnents ' important to' safety.

The quality assurance progr<a ill continue-to ensure that 1ponents important to' safety are structures, systems F

i designed, fabricatfj.

M-

. ad, and tested to the quality u

- standards commensura e wa;e. the' safety functionEto1be performed.

Attachment A lists byTsubsystem the Class 1 arbitrary.

intermediate breaks and pipe' whip restraints which'can be eliminated 'from the design (since the stress and' usage f actor limits are not exceeded).

The~FSAR'will be revised after NRC approval of this' submittal to show the physical' location of'the restraints: within a given system.

A total of approximately 154

' breaks per unit are to be eliminated.

Attachment B lists the ASME Class 2 and 3. piping intermediate b'reak locations that are to be eliminated.

A-total of

- approximately 81 break's per unit are to be eliminated.

L 4

l-

~

---w+m*

mw

• .,arr-o.

wd e s 4

+--w-+,

,y,.we-,

-wr.,.g,,%---,,p,...p.

,_.,p,m-,eg.p,

,,w, as hr at

-y F t Mwrww--v d >

• e w w+-tyf t-

c r

hY

^

" ~

1

~

s y

.,s

+

Jn th'is Llsubmiitt al? wei are ( prov idi ng Eaddi tional i tiecluiicall tinformationitoMjustify further,that' request.;? Specific NRCl iconcerns ;areladdressed in1 Attachments. C through ;H 'as follows::

z t 1. ' : Tech nicalk j usti f i catilonif or - elimi nat ion

" Attachment lCL ofsarbitraryl intermediate: breaks' W:

2'. " - Provisionsifor minimizing fatress cor s

T ~tachmentsD At

. rosionieracking11n-high.. energy -'linesi 3..

. ProvisiOnsifor lniinimizing.the: ef f acts _ -

' Attachment.E:

~

(off thermal and vibration -induced ' piping a

. fatigue J-c 4. -

Provisions _' for: minimizing swater/ steam?

1 Attachment,F 2 _

'_ hammer '-e f fects 5.

Provisions foraminimizingflocal' stresses

" Attachment H-

=from welded attachments' The application of.the1 proposed' criteria Echangesf williresul't ~ in.

the deletion of approximately 235 break locations and 167 pipe

~

-whip restraints in Class 1,-2 &: 3 -piping. - The' breaks and-restraints currently targeted-for elimination'?areolisted in l-

~ Attachments -A & L B.

-However, it should be f noted1that. piping and '

U system design is an iterative. process and that ' postulated break e

locations could potentiallyLmove as the system' design, and analysis of structures and piping develops over the course of the design process.

Owing tof the : iterative nature of the - design process and its potential for af fecting postulated break locations, changes affecting high energy systems are continuously-monitore'd and evaluated to determine 1theJimpact'on break 1

. location.."We-propose to apply these alternative criteria to any.

potential break locations in the systems identified herein,'

provided the stresses at those-locations are below the' break selection threshold, and the operational concerns in attachments-i

( E) through (H) _are adequately adressed.

This flexibility is j

nececsary to minimize ~ future requests for break. elimination as the ' location of intermediate break points change during ' the -

evolution of the plant' design.

o Also, for those piping systems, or portions thereof,. which are not ' included in this submittal, -the existirq guidelines in MEB 3-1 of-the SRP (NUREG-0800) Revision 1 'will-be met.

If other~

piping. subsystems 1 included in the systems ' identified in Table l

.D-1, butznot specifically identified in this submittal, L

-subsequently qualify for the conditions described herein, the L

' implementation ~of the proposed elimination of'the arbitrary-

~

intermediate break! criteria'may be used.

If.this criteria-is to be applied to additional ~ systems not-included.in Table-D-l, those

~

3

systems will be appropriately identified to the staf f.

5 l

v v

L l'

.u.

' Ceco has evaluated the potential cost savings ;and operational benefits that result!from the elimination of arbitrary intermediate. breaks.

These benefits include!$11.5 million Jsav'ings L in analysis, - design ~- fabrication, and installation of t

associated pipe whip restraints and jet' impingement-barriers. and -

1000 man-rem in~ dose reductions for both Byron and Braidwood Stations over their?40-year' plant lives.

A detailed breakdown of-the benefits realized bysthe-elimination of the arbitrary.

intermediate breaks'is;provided in-AttachmentLI. 'The actual benefits that. CECO -will; realize are~ expected to be higher than these. due - to the hidden f actors and intangibles-that. are.

difficult to. identify at this time.

It is clear, however, that elimination of the arbitrary-intermediate breaks is both safety-

~

effective and cost effective.

The percentage of the total potential benefits that can be realized by CECO. for 'the Byron and Braidwood Stations becomes-a matter of timing due: to the advanced stage of design and construction at - Byron-2._ and. Braidwood, and the pending completion of the Byron-1 startup - program which' may be~af fected ' by, the : pipe

' whip restraint energy absorbing material ( EAM) issue.

To make it possible for CECO to realize -the maximum benefitsi af forded by-this. proposed change in the pipe break criteria, immediate attention by the NRC is requested with a favorable response to the proposed change in the pipe break criteria by December 31 1984.

l l

\\

I 6

..=

ATTACHMENT A-Summaary. of Class l'. Piping -

Intermediate Break Reductions-No.-of' Pipe Whip:

Intermediate No. of Breaks

-Restraints System Subsystem Break Locations Elimi nated

. Elimi nated Seal Water ICV 34, 36 Socket' Welds at Valves 8

0 Injection 40, 41 1CV8772A-D Loop Fill ICV 12, 13

.SocketLWelds'between 20' O.

-14, 24 Crossover Leg Nozzle and 1st Valve-Safety ISIl0, 11 Elbow Butt Welds.between 4 '-

0 Injection Hot Leg Nozzle and 1st l

to Hot Leg Valve l

Normal ICV 06,.23 Butt Welds at Class 1

'6' l'-

Letdown

-Valves Residual' 1RH02 2nd Elbow -from Hot Leg '

[2 2-Heat Removal 12" X 12" X 6"' Tee'

_2;

'0"

16 f4 _

RC Bypass.

IRC01-04,

.lst and 3rd : 8"-

Elbow 16-19 Welds from Cold Leg Loop Stop Valve 1-1/2" Valve-Welds 16:

O.

1-1/2" Flange Welds 8'

0 i

4 A-1.

ATTACHMENT A Summaary of Class 1 Piping Intermediate Break Reductions No.'of~

Pipe Whip.

. I nte rmed iate -

No. of Breaks Restraints:

System Subsystem Break' Locations

- Elimi nated

' Elimi na ted Excess ICV 09, 11, Socket Welds between 12 0:

i Letdown 15, 16, 25 Hot-Leg Nozzles and 1st valve-Socket.. Welds between.

40 0

Crossover Leg Nozzle and 2nd Valve Surge Line 1RYO5 Elbow / Welds 2

0:

Accumulator ISIO1, 03, 04-10" X'10".X 6"' Tee 24 4.

and Cold Leg 09, 16, 17 Socket Welds between'

' 14.

'O' Injection 6 X 2 Branch and 1st Valve.on~2" S1 TOTAL 154 11:

4 4

e A '

.z.

s

- ATTACHMENT B '-

8 Summary of Class 2 and 3 Piping Intermediate Break Reductions No. of' Pipe No of Breaks-JWhip Restraints:-

System Building Subsystems' Eliminated Elim i nated '-

CVCS C

.1CV04,05, 07, 22, 23 5

0 CVCS A

1CV01, 18, 38, 39, 44 15 0

53, 71, 72 Main Steam C

IMS05, 06,-07, 08

.'8 12 Main Steam

-A IMS01

20-23-Main C

1FWO2, 03, 04, 05

~8 117 Feedwater Bypass

-C IFWO6, 07,.08, 09

-8 0

Feedwater SG Blowdown C

ISD01-06, 11,,12

'15 43 SG Blowdown A

ISD67-10,.25

'2 0'

i TOTAL 81 56' 4

t-4

+

I B-1 J

_ _ _ _ _ _ _.. _ _ _ _ _t. j

ATTACBMBfT C TIK'HNICAL JUSTIFICATION FOR ELIMINATION OF ARBITRARY INTERMEDI' ATE BREAKS The following items provide generic technical justification for ' the

~ elimination.of-arbitrary intermediate' pipe breaks and-the associated pipe. whip restraints.

1.

The operating ' procedures and piping and system designs minimize the possibility of stress corrosion cracking, thermal and vibration induced fatigue, and water / steam hammer in~these lines in which arbitrary pipe breaks are currently postulated.- Detailed descriptions of the design provisions for these phenomena;are provided in Attachments D, E '&

F, respectively.

2.

Welded attachments are not located-in close proximity to the breaks to be eliminated.

Consequently, local' bending stresses resulting from these attachments will not.significantly af f ect the stress levels at the break locations (refer to Attachment H)..

3.

The remaining postulated pipe; breaks and whip restraints provide an adequate level of protection in areas containing high energy lines.

Potential environmental effects are still considered in the design.

4.

Pipe breaks are postulated to occur at locations where stresses are only 80% of Code allowables (Class 2 and 3) or where the cumulative usage factor is only 10% of the allowable 1.0.

The arbitrary breaks to be eliminated all exhibit stresses and usage factors below these conservative thresholds.

5.

Pipe rupture is recognized in Branch Technical Position MEB 3-1 as being a " rare event which may only occur under unanticipated conditions".

6.

Arbitrary intermediate breaks are only postulated to provide additional conservatism in the design.

There is no technical justification for postulating these breaks.

7.

Elimination of pipe whip restraints associated with the arbitrary breaks will facilitate in-service inspection, reduce. heat losses from the restrained piping, and reduce the potential for restraining pipe due to unanticipated thermal growth and seismic motion.

8.

Pipe break related equipment qualification (BQ) requirements will not be affected by the elimination of the arbitrary breaks.

Breaks are postulated non-mechanistically for DQ purposes.

It is concluded that the elimination of arbitrary intermediate breaks is technically justified, based on the reasons stated above.

C-1

,s c

ATTACHMENT D PROVISIONS FOR MINIMIZING STRESS CORROSION CRACKING IN HIGH BEERGY LINES

~ Industry experience has _ shown- (NUREG-0691) thatLstress corrosion

- cracking (SCC) will not' occur unless the -following conditions exist.

simultaneously; _high tensile stresses', susceptible piping material, and-a corrosive environment. 'Although any stainless or_ carbon steel piping will exhibit some degree of residual stress and material susceptibility, Commonwealth Edison Company minimizes the. potential for SCC by. choosing - piping material with low susceptibility to stress corrosion. and by preventing the - existence of _ a corrosive -

environment.. The material specifications consider compatibility with

' the system's operating environment (both internal and external), as well as other materials in the system, applicable ASME code requirements, fracture toughness characteristics,. and welding,.

processing, and f abrication techniques.

The likelihood of' stress corrosion cracking in stainless steel increases with carbon content.

Consequently,'only the lower ~ carbon

~

content stainless steels-(304,-304L,.316, 316L) have been used for the primary systems

• at the Byron /Braidwood, Stations.

The existence of-a corrosive environment is prevented by strict criteria for internal and external pipe cleaning, and water chemistry control during start-up and normal operation.

For the secondary systems **, ferritic type carbon steel has been the choice for the piping, fitting s, and valve bodies forming the pressure boundaries.

This ferritic material has been found satisfactory from the standpoint of non-susceptibility to stress corrosion cracking for the service conditions encountered.

Since in the case of PWR's the secondary systems are not made of stainless steels, the question of stress corrosion cracking as reported in stainless steels 'does not arise.

All piping involved in the elimination of arbitrary intermediate breaks will be cleaned and flushed as part of the stait-up test program.

The. piping will be flushed with demineralized water subject to written criteria for limits on total dissolved solids, conductivity, chlorides, fluorides, and pH.

Plush water quality is monitored periodically.

The flushing is controlled by detailed procedures written for each system.

Water chemistry for pre-operational ~ testing is controlled by written specifications.

• Primary Systems: Reactor Coolant (RCS), Chemical and Voltrne Control (CVCS), Safety l

Injection (SI), Residual Heat Removal System (RHRS).

f Secondary Systems: Main Steam (MS), Main Feedwater (MEW), Auxiliary Feedwater ( AEW),

l Steam Gemrator Blowdown (SGBDS) l D-1

4

'.ATTACHNBIT D=

During l plant operation,; primary and secondary s'ide. water chemistry.

' will be monitored in the stainless -steel and carbon steel piping.

Contaminant concentrations will-be kept :below.the thresholds known to -

be conducive ~ to' stress corrosion cracking.

The major water chemistry contro11 standards will-be included in the plant operating procedures for-the lines-in which arbitrary breaks were previouslympostulated.

Oxygen content is expected to.be 'less than 0.005 ppm.during ; normal.

power-operation, thus further minimizing the' likelihood of stress corrosion ' cracki ng.

Table D-1 summarizes the4 systems in whichEcurrently postulated arbitrary intermediate breaks are to be eliminated.

Note that a.

number of these systems operate at temperatures below 200 F.

-Industry wide experience shows that= stress corrosion isynot a problem at temperatures this low..The recommended water chemistry requirements for primary systems are provided in Table 5.2-3 of the.FSAR.

Operating water chemistry guidelines for secondary side piping are-given in Table 10.3-1 of the FSAR.

Commonwealth Edison - has developed :

and implemented a secondary water : chemistry program based upon.the Steam Generator. Owners. Group Secondary Water Chemistry Guidelines.

l

\\

D-2

_.7 1

ATTACHMENT D Table.D-1 l

Elimination of Arbitrary Break Systems Summaary c

Piping Operati ng No. of. Breaks:

Piping System Material

. Temp. (OF).

Deleted (Per Unit)

Safety Injection SS' 120 22

~

CVCS - Charg ing SS

'S16/130

-8 CVCS.- Letdown SS 556 68:

CVCS - RCP. Seal I nj.

SS 130

.10-

^

RCS - Surge Line SS 653 2-

. RCS.- Bypass SS 619 40 CVCS - Loop Fill SS 120 20.

~ ~ ~

Bypass Feedwater CS 445 8:

Steam Generator Blowdown CS' 454 17 Main Steam CS 545-28' Main Feedwater CS 445

.8 Residual Heat Removal SS 619/350

'4 235 SS - Stainless Steel CS-- Carbon Steel D-3 u.

r-.

m L..

J

i

'ATTACHN ENT E

. PROVISIONS FOR MINIMIZING THE EFFECTS OF '

THERNAL AND VIBRATION INDUCID PIPING FATIGUE

. I.'

G EN ERAL FATIGU E DESIGN' CONSIDERATIONS For Class 1 lines, fa'tigue1 considerations are-addressed ~by the cumulative usage factor (CUF).- To ensure thatipiping -will 'notl f ail due to fatigue, the ASME Code has set the CUF limit at 1.0.

By definition,'all2 arbitrary intermediate break: locations; have CUFs below 0.1.

For Class 2 and 3' lines, fatigue is considered 'in the allowable

~

stress range checkL for thermal; expansion - stresses.

This stress is included in the total stress value used to. determine postulated break locations.

All' arbitrary break locations exhibit stresses less than 80% of the' code allowables. -If - the number of thermal cycles is expected to be greater than 7,000, 4

then the allowable stresses are'further reduced by an amount dependent on the number of cycles..

II.

TH ERMAL DESIGN CONSIDERATIONS By -limiting the mixing of low ' velocity, low temperature.

auxiliary feedwater with high temperature water in the steam.

generator inlet nozzle, cyclic thermal' stresses'in the auxiliary feedwater piping are minimized.

Mixing is prevented in the auxiliary feedwater supply to the 6-inch auxiliary feedwater steam generator inlet nozzle with a vertical piping - arrangemant -followed by a 90-degree elbow welded to the 6-inch inlet nozzle.

Feedwater temperature instrumentation is provided in the vertical run of the inlet elbow to the 6-inch steam generator inlet nozzle to monitor and alarm the backflow of high temperature water.

Mixing of the low velocity, low temperature main feedwater with high temperature water in the steam generator is prevented in the main 16-inch feedwater nozzle by isolating flow to the main nozzle and introducing feedwater to the 6-inch auxiliary feedwater steam generator inlet nozzle for power levels below 20 percent.

The physical layout of the auxiliary feedwater piping, temperature monitoring / alarm instrumentation, and minimum feedwater flow rates are in compliance with the Westinghouse l

design criteria for the main / auxiliary feedwater supply piping to the steam generators.

Cyclic thermal stress is prevented in the other lines containing l

arbitrary intermediate breaks by maintaining uniform

-temperatures with no mixing.

l E-1 i

r m

f, 7 M EENBrf E'~

ui s,

.III.: : VIBRATION DESIGN CONSID ERATIONS~'.

...y

r

Pipingiin' the Byron /Braidwood [ Stations (is"' esignedia'nd I supported.

~

d to-- minimize transie.nt 'and ' steady" state vibration.. -Testi.ng : will.

tbe0 performed"as1 described:in: Section:3.9.2 of the.FSAR to ensure-ithat' vibration of. the _ piping systems is:.within allowables

~

levels. : f Plant personnel,willLbe trained :to recognizelexcessive-

. piping -vibrat'icn :so :that potential. problems 1 ca.n. be -resolved. ;In

-addition, a, formal - testi program, - a's toutlinedL in : the-FSAR,- will

, be ~ completed to. verify the acceptability of - thefpiping'. steady.

n sta'te vibration ~.

i-I a

8 E-2

r J

ATTACBMENT F PROVISIONS FOR MINIMIZING ' STEAM / WATER HAMMER EFFIK'rS

Systems within Westinghouse' scope of Isupply are not in general, susceptible to water' hammer.- The reactor' coolant, chemical 1and. volume-

-control and_ residual heat removal systems have been specifically designed to preclude water hammer.- Preoperational testing.and, operating experience-have verifie'd the Westinghouse design approach and furthermore,.have indicated that significant water hammer events have usually been-initiated _in: secondary systems within.the Balance of--

Plant (BOP)EscopeJof! supply.-

In these-systems, anticipated hydraulic transients have been included in the design loads and design features have been incorporated to prevent water hammer.

Westinghouse' has conducted a. number of. investigations into the 'causes and consequences of water hammer-events.

The results of these investigations have been reported to Westinghouse operatirg plant customers and have been reflected in design interf ace requirements to the BOP designer for plantu under construction, to assure that water hammer events initiated in the secondary systems do not compromise the performance of the Westinghouse-supplied safety-related systems and components.

Some of the lines in which arbitrary intermediate breaks are to be eliminated have the potential for water / steam hammer effects.

These lines have been designed to minimize or preclude such effects.

Water hammer in each of the systems involved in the elimination of arbitrary breaks is described below:

1.

Safety Injection System The safety injection lines are all water solid and at ambient temperature, thus no water hammer is expected.

2.

Chemical and Volume Control System (CVCS)

Normally, the CVCS is water. solid.

In the low temperature lines (less than 125 F) water hammer would not be expected because of the small probability of steam void formation.

In the high temperature lines, the piping has been designed to maintain water solid conditions during normal operation, thus minimizing the possibility of water hammer effects.

3.

Reactor Coolant System There is a low potential for water hammer in the reactor coolant-system, because it is designed to preclude steam void formation.

However, excessive cooling of the reactor coolant system, which initiates safety injection, could potentially result in water hammer.

If any problems are experienced during preoperational testing, they will be eliminated by modifying operating procedures.

F-1 q

~..

. ~

o 4

~,

+

+

s.

JurTACIMErrjFf s_<

u 14 '. '

'Ma[ncSteams The main; steamj piping - f romi th'e f 5-way - restraints Ijust7outdide

~

1

-Ccontainment~.to the main.JturbineLis. sloped;aty1/16~of'anLi'nch perc M

foot to -assurei properJ drainag~e 'during ithe various phases ? of:.

$ operation.

18-inch ediameter. drip -legs : approximately; five ~ feet longiare installed - upstream 'of.the main iturbine ' inlet on the c 736-inch mand 38-inch main : steam elines-to collectlandidispense drainagecto the condenser. lThe branchilines?that.teet.off'thei r

(main:' steam:1in' s Lare properly sloped with. drain provisionsitoi e

-eliminate;the-possibility'of. water hammer to occur:due to-icondensate-drain waternpockets collecting in low points or pipe R

loops.

5..

Steam' Generator ~ Blowdown (SGBS)

~

i

~

. Blowdown-flow ~from-thefsteamigenerator'is)normallyLtwo-phase and' of,0-10; percent Equality.

The s piping : 1~ayout is generally routed o

downward starting. from the ' steam generator blowdown e nozzle connection. and1 continuing ;to~ the containment penetrationithus

. minimizing 1 the. formation of waterL pockets.

Therefore, thei potential for water hammerTis minimized for-the. blowdown-lines-

~

within containment. --Water hammer may occur downstream of the.

isolation valves. upon reinitiation of blowdown flow following isolation.

Operating procedures will p' ovide Lforigradual r

_ repressurization of the downstream piping before establishing full flow, 'thereby ' minimizing any potential water hammer.

3 problems.

1 6.

Auxiliary Feedwater i

.The Auxiliary Feedwater - ( AP) system provides;feedwater to the.

steam generator auxiliary: nozzle via aLeonnection to the F

feedwater bypass piping.

Each -steam generator: auxiliary nozzle

}.

utilizes a 900 elbow connected immediately to a vertical-run of.

j pipe to minimize steam voids. 'Under normal. operating conditions,.

the main feedwater split flow arrangement (described in Section-

/) ensures that the bypass line is kept filled with water,'and

}

steam is thereby ' prevented from leaking back._intoL the ' Auxiliary Feedwatera piping.

)

i The following design features are included to avoid a bubble i,

collapse water hammer event n

1.

Temperature sensors are; installed on the bypass piping.close

.to the auxiliary nozzle to~ detect backleakage of hot water or steam.

f 2.

'If backleakage is detected, the piping will be-slowly

. refilled or the plant brought to a cold shutdown condition, i

depending on the circumstances.

An analytical study' l -

F-2

-e-..,,,w,.,-

• --,,,,,,-,,,,e,,,.,-.

,r.,-,

~.,,,+.,c

,e s.,-.

-.---e,,-,.,y

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

..wc

,,,y-%

wyw

-v,--e,...,,y

1 ATTACHNBFf F performed' by Westinghouse.shows;that.the' bypass piping can

.be-slowly ~ refilled-safely.- The recommended flowrate--is on the' order of 15 gpm..

3.

The steam generator water level _ should = tme maintained above the; auxiliary _ nozzle discharge pipe so that-if backleakage does occur, water-instead of steam 'will L leak back'. into.the

. pipe.

4.

'The Auxiliary lFeedwater System' check valves willibe-

'maintainedLto minimize backleakage.

5.

Consistent ' with Westinghouse - recommendations, there Lare 'at~-

~1 east.two check valves in each flow path by which backleakage could" occur into the Auxiliary Feedwater or Main Feedwater System.

-7.

Main Feedwater.

The routing, of the main feedwater -giping, which Lvaries in -

temperature from approximately 300 F it low load to 445 P at full load into the steam generators, which operate between 545 F to 557 F during normal operation,, is in compliance with the-Westinghouse criteria for layout, temperature monitoring / alarm, and_ operational procedures to minimize or eliminate water.

hammer.-

Water hammer prevention features of the mainJfeedwater.

system are described in detail in Section 10.4.7.3Dof the Byron /Braidwood FSAR (Attachment G).

The Byron /Braidwood Stations have Westinghouse Model D preheater type steam generators.

The main supply of feedwater enters the preheater through the main 16-inch nozzle in the lower shell.

The other-supply of feedwater is through the 6-inch diameter auxiliary nozzle located in the upper shell.

The Feedwater Bypass System is designed to prevent the introduction of cold water into the preheater section.

In those circumstances where it is necessary to introduce cold water into the steam generator, the Feedwater Bypass System operates to direct the cold water to the upper, auxiliary nozzle.

This bypass consists of a 6-inch diameter line which connects the main feedwater line to the auxiliary nozzle. ' The Auxiliary Feedwater System also provides feedwater to the steam generator through the bypass piping and the auxiliary nozzle in the event of a loss of main feedwater.

Steam backleakage into the bypass piping is very unlikely.

During power operation, the Byron /Braidwood Stations utilize a split flow scheme which pro. ides a continuous flow through the F-3

r}

' i 1

4 1

+

q_

t

bypass piping. to -thesaudiliary nozzle of f about :10% of 'the main j

~feedwater-flow.

This1 continuous flow effectively prevents (the

backflow of steam from thef steam generator.:

i During 7 the ' normal _ operations 'of. heatup,icooldown L and - hot standby :

)

0 (rated flow less than 1154 "and temperatures cless'.than 250 F),_

feedwater is ' supplied ~ only through : the 6" auxiliary fnozzle.-

'However, only-relatively small, amounts of feedwatercare' required',

not' enough : to-always permit a' continuous' flow so J that the:

opportunity.for steam backleakage does exist.if'the7checkJvalves

, c fail _and thel steam generator. water level falls below the Eauxiliary_ nozzle internal extension..Possible steam backleakage,

i is detected:by. surface mounted resistance. temperature-detectors

- which are provided on each feedwater-pipe.

These Jare monitored by the~ plant process computer and are. alarmed in.the main control room so that. actions : can -be taken to ; initiate feedwater ' flow to-the upper nozzle ~before'potentialtfeedwater hammer conditions'may

~

-develop.: Also, the plant operator is' instructed to feed continuously-rather than intermittently 4as muchias possible...

This practice' reduces /the. likelihood of steam backleakage-and,_

therefore, water. hammer.

'In the eventuality'that the presence of. steam is suspected in the bypass-line of one or more loop, based on temperature data and water level status and history,.the recommended. course.of action-is to slowly refill one loop at.a time with the Auxiliary Feedwater System.' An L analytical study by _ Westinghouse shows.that the-safe refilling flow rate is in the-rangelof 15 to - 123 'gpm 'pers steam generator.

To be conservative, Westinghouse has

~

recommended the value of 15 gpm or as close to this as can be

~

provided.

Based on another - analysis performed - by. Westinghouse which considered the-classical water hammer case of feedwater line break-followed by check valve closure,. Westinghouse recommended that the valve close to the auxiliary nozzle should be removed:

and the'other check valve:in the bypass line should be replaced with a slow closing valve.

Commonwealth Edison has implemented -

this recommendation.

The design features and operating procedures described-above lwill preclude or minimize the effects of water' hammer.

t' F-4

ATTACEMBfT G'

)

MATER HAMMER PREVBETION F ERTUR ES (B/B-FSAR'AM.44)

~

10.4.7.3 Water Hammer Prevention' Features-Several water hammer preventionL features have been designed intolthe feedwater system.- These features'are provided to minimize the possibility.of:various water hammer phenomena,in the steam generator preheater, steam generator main.feedwater inlet piping and the steam generator upper nozzle feedwater piping.. The - following discussion is

-typical for each of"the--four-steam generators and-their associated feedwater piping.

10.4.7.3.1. Start-Up, Low Load Conditions a.

Under start-up and low load conditions when NSSS rated.

flow is less than 15% and temperatures'are less than 250 F, feedwater will only be admitted to.the upper nozzle of.the' steam generator by the'use of flow through the feedwater bypass tempering line and/or flow-through the feedwater preheater bypass line via

the feedwater bypass control valve and feedwater preheater bypass valve.

The 6-inch diameter' upper nozzle is located on the upper shell of the steam generator, below the normal, full power water level.

Level control in the steam generator is provided by the feedwater bypass control valve at these conditions.

b.

Surf ace mounted resistance temperature detectors (RTD) are provided on each of the feedwater pipes, leading to and very near the steam generator's upper nozzle to detect during start-up and low load conditions as well as other operating conditions, possible back leakage of' steam from the steam generator into the feedwater piping.

These RTD's are monitored by the plant process computer and alarmed in the main control rocu so that actions can be taken to initiate'feedwater flow to the upper nozzle before potential feedwater i

hammer conditions may develop.

10.4.7.3.2 I ncreasing Load

(

a.

As load increases about 15% of NSSS rated flow and 0

l feedwater temperatures rise above 250 F, forward feedwater flushing of the main feedwater piping may be initiated by opening the feedwater isolation bypass

}

valve.

A small controlled flow through the 3-inch feedwater isolation bypass line is provided to flush l

G-1 L

,ATTACHNBrf C the main feedwater. piping between.the iholation(valve and the steam generator.

b.

. Three sets.of three RTD's-are provided.on the main

-feedwater piping upstream and downstream of the feedwater isolation valve and near the' steam-generator-feedwater nozzle - to detect when the feedwater flushing-

~

in temperature 1 rises ~above 255 F.

Two out of 'three logic is provided ' for each set,of three RTD's and all three

.must be. satisfied -to meet the forward flushing temperature requirements..

c.

If flow in the'3-inch feedwater. isolation. valve ~ bypass

'line = ( forward flushing flow)3 remains above a preset

. minimum and below a preset maximum and the f flushing temperatures remain satisfied, a timed; period. occurs af ter which a permissive signal. is provided to automatically open.the feedwater isolation valves.

Automatic opening of a feedwater isolation valve can be blocked by placing its control switch in the main control room in-the closed: position. 'This automatic =

permissive to.open occurs after a timed period to' allow.approximately two volumes of water to'be purged from the ~ piping between the ' feedwater isolation valve and the steam generator main feedwater nozzle.

Feedwater. flow at the main.feedwater flow-element must-also be above a preset minimum in order for the feedwater isolation valve.to open.

d.

After the feedwater isolation valve has opened, the-feedwater isolation bypass valve will be manually closed.

e.

Prior to openir.g of the feedwater isolation valve, transfer from the feedwater bypass control valve to the feedwater control valve will occur in order to provide steam generator level control at the higher feedwater flow conditions.

f.

If flow to the steam generators remains continuous during a load transient and above a minimum flow rate, feedwater will not be terminated to the main'feedwater nozzle even if temperature of the feedwater has dropped below 250 F.

Interruption or a reduction in flow below the minimum rate however, will cause the feedwater preheater section of the steam generator to be bypassed.

g.

Steam generator low level trips are provided to close all of the feedwater isolation valves, feedwater isolation bypass valves and feedwater preheater bypass valves.

Steam generator low pressure trips'are provided to.close all of the feedwater isolation valves, feedwater isolation bypass valves, feedwater G-2

ATTACHNBrf O preheater' bypass. valves and the,feedwater bypass tempering valves. -

- 10.4.7.3.3 Split Feedwater Flow Prihr ito opening of the feedwaterf isolation' valve, the a.-

majority of feedwater: flow at_the lower power; level is-

~

introduced to'the upper nozzle of-the. steam generator.

by:.the preheater bypass pipe.

b.-

At higher powervlevels after the feedwater isolation valve has opened, only'a small' portion of the

-feedwater flow bypasses the preheater, with the bypass' portion contributing to approximately 10% of tull 1

feedwater flow at 100%-power.

.This' split;feedwater flow arrangement provides. an approximate 90% of full flow limit to the main feedwater nozzle at higher power levels in order to minimize the potential for tubing vibration in the steam; generator.

The feedwater1 flow rate to the' steam-generator nozzle'.is monitorad and alarmed, if flow. rises above approximately 90%, in order for' actions to be taken to reduce flow.

c.

The preheater bypass valve remains.open throughout _ the

. start-up and low load conditions, as well as up to-and including full power operation.

10.4.7.3.4 Other Upper Nozzle Feedwater-Line Uses Inasmuch as there is water flowing to the upper nozzle of the steam generator during normal' operation, and it is the required location for introducing cold fluid into the steam generator, auxiliary feedwater and chemical feed are connected to the upper nozzle feedwater lines rather than to the main feedwater lines.

The chemical feed. lines are used to add chemicals directly to the steam generators under low load conditions prior to wet layup.

The chemical feed and auxiliary feedwater' lines are safety Category I, Quality Group B out to, and including their isolation valves.

t i

l l~

G-3 I

ATTACHMBfT H PROVISION FOR MININIIING LOCAL STRESES FROM NELDE ATTACHNBfTS CECO-has reviewed all arbitrary intermediate break locations, to be eliminated and.has determined'that in no case are welded attachments placed'in close proximity to postulated break locations.-;As a: result,

. local' bending stresses induced by the. attachment. will not af fect the stressos at the postulated' break point.

To ensure that this is the case, the local stresses have been determined and added to the primary stress report.

f l

l H-1 L

$1 WTACWIWF I SUGGIARY OF SWWITS FOR TEB E.!NIEMION OF ARBITRARY IWT W WIM E PIPE BR MES -

BT W /3RAIDuCOD ST MI WS Changes Resulting from Cost Savings Break Elimination (1983 Rates)

Operational Benefits Elimination' of 67 Pipe o Design, Fabrication and o Potential improvement in Whip Restraints per Unit Installation Costs

• quality of inservice inspection (ISI)-

o Dose Reduction Costs o Dose reduction from improved personnel access during maintenance, ISI and recovery from unusual plant conditions, e.g.,

radioactive spills, fires, etc.

o Improved capability to recover from unusual plant conditions, e.g.,

decontamination following radioactive spills, access for fire lighting, etc.

o Reduced system heat loss resulting from improved insulation design.

o Dose reduction and improved construction schedule by eliminating the need to set and maintain restraint clearance gaps, Elimination of Jet o Barrier Design, o Dose reductica from improved Barriers and/or Fabrication, and personnel access during Equipment Relocation Installation maintenance and recovery from unusual plant conditions, e.g.,

radioactive spills, ftres, etc.

o Dose Reduction Costs o Relocation Costs o Improved capability to recover f rom unusual plant -

conditions, e.g.,

decontamination following radioactive spills, access for fire fighting, etc.

Elimination of Analyses o Jet Impingement Load o Improved system layout and Associated With the and Pipe Whip Analyses design for future plant Dynamic Ef f ect s a nd Costs

• modifications Loading Conditions TOTAL SAVINGS 811.5 Million 1000 man-rom in dose (UNITS 1 AND 21 reduction over the 40-year plant life.

"one cost Appitcnole to both Units.

1-1