ML20202B311

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Confirmation of Design Adequacy for Jet Impingement Effects,Braidwood 1
ML20202B311
Person / Time
Site: Byron, Braidwood, 05000000
Issue date: 07/31/1986
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COMMONWEALTH EDISON CO.
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NUDOCS 8607100272
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BRAIDWOOD 1 CONFIRMATION OF DESIGN ADEOUACY FOR JET IMPINGEMENT EFFECTS 4

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JULY 1986 i

8607100272 860702 PDR ADOCK 05000455 A PDR ,

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EXECUTIVE

SUMMARY

The potential effects of High Energy Line Breaks (HELB's)' lave l been reviewed for Braidwood Unit 1 with the same level of detail as was done for Byron Unit 1. This report has been prepared as

. an overview and summary t'o document the completion of the review and to point out the significant differences between the Byron-Unit 1 and Braidwood Unit 1 analyses. Rotential intera.ctions between the various effects of HELB's, such as pressurization, flooding, and pipe whip, have been considered with jet impingement effects. .However ,. only the jet impingement ef fects are covered in this report. y There are no significant differences between the equipment

. required for safe , shutdown for Byron Unit 1 and Braidwood Unit 1.

However, because of the reduction in the number of breaks and-some differences in location of breaks and equipment or r6uting of cables or piping, there may.be differences in the components which could be affected by a break. The majority of.these differences are jet effects $hich were evaluated only on Byron Unit 1 because the break was eliminated on Braidwood Unit 1.,

These jet effects are not specifically identified in this report ,

because of the large number of differences. The few instances of a unique jet effect on Brai.dwood Unit 1 equipment which did not occur on Byron Unit 1 are covqred in the report.

The Braidwcod Unit 1 desig'n includes an inherent protection against the effects of jet impingement. However, a detailed , ~

review of the design is required. The procedure used to perform this review included a re. view of the potential effects of c individual jets as well as the resultant effect of the jet on components used to safely shutdown the plant. Break locations were defined using the current methodology. Safe Shutdown components were identifies and locations were compared with break locations.  ; -

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  • The determination of po'tential.da* mage was made by comparing,the component locations and the pr6ak locations and defining.the: '

potential interactions.. .'These were examined in detai.l'as '

described in the text of the report. ,

For the potential jet impingement damage, the effect on safe shutdown was examined and, if problems existed, more detailed calculations of the jet influence and loading were completed.

The evaluations summarized here are documented fully in calculations and reports which are referenced in the report.

These documents directly correspond to calculations and reports which were completed for Byron Unit 1.

Differences between e"quipment ' location ahd cable and pipe .

routings for Byron Unit 1 and Braidwood Unit 1 are relatively

  • minor. In the auxiliary building, a design change made after submittal of the Byron Unit' I report added temperature sensors to prevent, environmental qualification probl' ems. The additional equipment was evaluated for jht impingement at the time of the redesign and is discussed in this report. Also, breaks in the AS

. and CV systems caused impingement effects for Braidwood Unit 1 .

which are different from those identified for Byron Unit 1 -. In Containment, a pressutizer pressure sensor was af fected on Bra,idwood Unit 1 but,not on Byron Unit 1. ' Evaluations ~showed .-,

that these differenceh will nog result in safe shutdown concerns.

e As a result of this evaluation it has been demonstrated that Braidwood ' Unit 1,can be safely shutdown,afier a HELB considering the combined effects of jet impingement and other effects of the break and a limiting single . failure. Because of the reduction in break postulation requirements, the number.of potential jet ef fects on components has peen significantly reduced. The common d'esign basis utilized for-Byron Unit 1 and Braidwood Unit.l.has resulted in relativ,ely few dif,ferences between the HELB effects on the units.

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BRAIDNOOD 1 o Confirmation of Design Ad~equacy 5 --

~For. Jet Impingement Effedts i

. 1.0 Intr'o duction

-' s 2.0 Definitions 3.0 ,. Braidwood Design Approach ,

. '4 . 0 Confirmatory Study.

4.1 Scope

. *4. 2 Safe Shutdown Success Criteria Safe Shutdown 4.2.1 . -

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4.2.2 Cold Shutdown . ,,.

4 2.3 Reactivity Control

. 4.2.4 Decay Heat. Removal

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  • 4.2.5 Offsite Release .- .

4.3 Single Failure Criteria

. 4.4 Confirmation Procedure -

4,5 '

Safe. Shutdown Systems and Components ,

4.5.1 .. Identifi-cat, ion of , Safe Shutdown Systems ,

s. . 4.5.2- ' Safe Shutdown' System Design Features
  • 4.6 High Energy Lines 4.7 High Energy Line Breaks.

4".7.1 Jet Impingement Load Definition-5.0 Results of Confirmatory Study .

5.1 Auxiliary Building High Energy'Line Breaks 5.1.1 Auxiliary Steam Line Breaks 5 .-l . 2 Steam Generator Blowdown System Breaks 5.1.3 Chemical and Volume Control System Brea.ks" 5.2 Containment Building High Energy Line Breaks _-

. 5.2.1- Safe' Shutdown Systems 5.2.2* Summary of Jet Impingement Effects ~

4 6.0 Concitisions ,

7.0 References -

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1.0 INTRODUCTION

The design of the Braidwood station includes extensive separation of redundant _ mechanical and electrical systems to insure that plant safety will not be compromised by damage resulting from design basis events including High Energy Line Breaks (HELB's),

Moderate Ener.gy Line Breaks (MELB's), external flood-ing, fire, tornados, and turbine missiles. This confirmatory report specifically addresses the subject of potential jet impingement effects-which could.

result from high energy line breaks. However,'the ,

approach used to incorporate separation, redundancy and diversity i.nto the design of the safety systems provides a high-degree of protection against -' <

postulated events which could damage safe shutdown

, equipment. .

This report describes the approach taken in the design

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process and major design features which were incorpo-rated as a result., A review of potential jet effects . -

on safe shutdown components has been completed to confirm that the. design approach was, indeed, effective in protecting the plant from potential jet impingement effects.

This study addresses specifically Braidwood Uni'h 1.

The Byron Unit 1 Confirmatory Report submitted to the NRC in August 1984 is generally applicable to. Byron Unit 2 as well as Braidwood Units 1 & 2. This document does not unnecessarily repeat the generic information provided in the Byron 1 report. .Instead, sufficient background has been included to make clear the steps taken and the evaluations completed for Braidwood 1, the differences between the procedures used and the results of the Confirmatory Studies for. -

Byron Unit 1 and Braidwood Unit 1 are reported in

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detail. .

2.0 DEFINITIONS Diversity - A plant feature whereby an independent, non-identical system or component is available in the event of a failure of a system or component.

Emergency Core Cooling System (ECCS) - Those systems -

which function, in the event.cf a L.OCA, to prevent -

- co're damage. This includes the Safety Injection System and portions of.the Chemical and Volume Control System and the Residual Heat Removal System.

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Hazard Zone - A defined bounded area of the plant to be used to investigate.the potential extent of damage and system failure following an event which has a physical.eff.ect which may be spatially limited (e.g.,

fire, HELB, missile generation). The initiating event

.may or may'not be limited to one zone depending upon the nature of the event and the nature of the zone -

boundaries.

High Energy Line - A pipe line which operates during normal plant operations at temperatures in excess of -

2000 F and/or pressures in excess of 275 psia. Lines-which operate at high. energy conditions less than 2%

of the system operating time are not. considered high energy (S.tandard Review Plan Section 3.6.2). .

High Energy Line Break (HELB) - A locati'on wit $in a +

piping system where, per the guidelines of Standard Review Plan (SRP) Section 3.6.2, a break is to'be postulated.

HELB Zone - A hazard zone which contains a postulated HELB. ,

Loss of Coolant Accident (LOCA) - A HELB in the piping .

which forms the boundary of the reactor coolant system. For the purpose of this study large LOCA's are dgfined as those with a break area of greater than 1.0ft andsmalgLOCA'sarethosewithabreakarea less than 1.0ft Redundancy - A' plant design" feature whereby an -

independent, functionally identical system or component is availabl'e'in the event of a failure of a ,,

system or component.

Safe' Shutdown - A plant condition such that:

1) The reactor can be maintained subcritical,
2) Decay heat can be removed. -
3) Offsite release in excess.of allowable limits -

is prevented.

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item of structure, .

' Safe Shutdown equipment, cable, Component or piping- Any'equired r to maintain integrity or functionality to achieve safe shutdown following at least one postulated event. scenario within the plant design basis.

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Safe Shutdown Equipment - Mechanical and electrical.

. . equipment (e.g.., pumps,'. valves, switches, instruments) required to function to achieve . safe shutdown follow-ing at least one postulated" event scenario within the

  • plant design basis.

Safety Evaluatiom Report (SER) - The Braidwood Safety Evaluation Report (NUREG-1002). .

Separation - Physical isolation by distance or barrier of a safe shutdown system'or component from a redun- -

dant component or hazards'such as high energy' lines.

' Single Failure - Arbitrary failure of'a single -

component to perform its safety function following a postulated initiating event (See Section 4.3) -

Standard Review Plan (SRP) -

NUREG-75/087. The 1981 ~

revision of the SRP (NUREG-0800) is utilized where it

'provides clarificationsof the intent of NUREG-75/087.

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3.0 BRAIDWOOD DESIGN APPROACH The Braidwood design includes many features which

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eliminate or mitigate damag.ing effects of postulated High Energy Line Breaks (HELB's). This is a result of

  • ' a design. approach which addressed the requirements of General Qesign Criteria (GDC) 4 of 10EFR50. This -

design approach followed the guidelines of Branch .

Technical Posi. tion APCSB 3-1 and.Section 3.6.1 of the Standard Review Plan '(SRP) (Reference 1). These guideline ~s state that plant designs should protect 4 essential,. systems and. components from the effects of high energy line failure. The preferred methods of

' protection are separation of the essential systems ,

f rom high . energy line brea.ks by an adequate distance -

or by structures. In the' event.these methods cannot . .

.be used, redundant design' features which are protected should be provided. If these methods are not used, restraints or barriers should be provided.

The' safe shutdown systems.and components in the Braidwood design have been separated from high-energy.

lines and a1so separated from redundant systems to the extent practicable. As a result,- relatively few protective restraints and barriers have been required.

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4.0 -CONFIRMATORY STUDY ..

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In 1984 the Byron 1 Jet Impingement Confirmatory Study was completed to resol've questions raised by the NRC Integrated ^ Design Inspection Team. This study extehds,.'

the Byron 1 work to Braidwo'od 1. Although the desig'n' of the two units is almost identical, porti~ns o of the Confirmatory Study utilized "As-Built" information

.which can be unique to one unit. Also, certain changes in NRC requirements in.the area of break definition resulted in a change of scope of the study.

This section furnishes an' overview of the approach ,

taken in the Braidwood 1 Jet Impingement evaluation .

and-describes the differences with the Byron 1 effort. Sect

  • ion 5 summarizes the results and provides ~

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an assessment of the differences between the two '

. units.

4.1 SCOPE .

This Confirmatory Study considers potential jets from postulated high energy line breaks (HELB's) in the Braidwood l' Containment and in the Auxiliary Building. ,

HELB's'are assumdA to occur in piping following the .'

guidelines in SRP Section 3.6.2 with the following~two" - .-

exceptions: ,,

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- 1) Breaks.are not postulated in the large piping of the main coolant loops in the Reactor

- Coolant System. These breaks were eliminated-- .

from the evaluation of dynamic effects'

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because of the.results of stddies employing the " Leak-Before Break" concept. Use of this .

approach was approved for use on Braidwood by the NRC in Reference 5. .. ,,

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Arbitrary Intermediate Breaks at low stress 'N

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  • level focations, as provided for in the SRP Section 3.6.2, are not postulated.'This

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deviation from the SRP approach was approved- ,

by the NRC in Reference 6.

j The scope of the j'et impingement evaluation on. -

l Braidwood 1 was reduced considerably by these changes. Approximately 544 breaks were evaluated in the Byron 1 study. After" elimination of the Primary l

n Loop breaks and the Arbitrary Intermediate Breaks, 290 HELB's remained to be evaluated on Braidwood 1.

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s Com$o.nents dhich might.be used to safely shutdown the; plant following a. postulated HELB (as described above)

  • are included as potential jet targets. ,

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'4 '.- 2 .. Safe Shutdown Success Criteria s -

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In accordance with the requireme,nts of.GDC 4 to pro- ,

tect against the dynamic effects of line break, this

' dstudy

will show that the HELB's i.n ques. tion can be ..

mitigated and the unit brought to a safe. shutdown' .

condition. The criteria for achieving, safe shutdown

- are as follows: ,

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, 1. Reactivity is controlled such that the ~

reactor is subcrit.ical- . .

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2. Mechanisms are provide'd to remove decay heat. .

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Offsite releases of radioactivity.are - ..

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restricted to the limits of 10CFR100. , , ,

Safe. Shutdown 4.2.1 '

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._ Safe shutdown following'a'LOCA is def'ined as attaining

. cold leg recirculation. using only qu'alified (Safety.

s-Related) equipment and_ instrumentation, an6 maintain-

  • ing offsite' releases within the.gegulatory limits. -

Limiting offsite radioactive'fe.le'ases within the

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regulatory limits is accomplished by maintaining.at least one barrier between the radioacti,v.ity and the .

environment (.i.e., rqactor coolant pressure boundary .

or reactor containment). -

For non-LOCA. breaks, -safe shutdown is defined as hot , ^

standby .(T a

. zero percenE9. greater rated thermal than or equal

~ power and to . k 350. degrees F,eff of less than- '

0.99). The reactor coolant pressure boundary must be.

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maintained intact using only qualifidd (Safety. .

- Related) equipment. , .

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  • 4.2.2 Cold Shutdown ,

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Braidwood's licensing basis is hot shutdown, ~

therefore, it is not necessary to demonstrate

,, capabilitytoreachcoldshutdowncohdition'sgreactor F', 0%

coolant temperature less than or equal to 200 -

-- *Aav rated. thermal power, and k of.less than or equal to 0.99) using only safety re$a ed equipment'. Howeve r', , *

.. bhe existence'of a method for reaching cold shutdown ~

without repair or replacement of equipment has been

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revigyed arul is described in this stedy. Non-safety re' lated. equipment may be used t6' attain cold shutdown. , , , ,

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4.2.3 - Reactivity Control

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" - Sufficient negative reactivity can be provided for hot ~

shutdown by rod insertion with or without h= single active failure ofJA worst cas.e' stuck control rod. .The- -

Braidwood Ref'ueling Water Storage Tank (RWST) has sufficient boron concentration to assure that' ~

r'activit'y e can be conetol. led in a cold shutdown ..

. condition without use of the botic acid . transfer

' system except in'a case which combines an unfavorable

-. ' core. history with a. single active. failure of a stuck control rod. The-additional boration can be achieved through operation of. boric aci.d trans'fer pumps 0AB03P

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.. and 1AB03P to utilize the boric acid tank 1AB03T as a- .-

- s.ource o'f,boratlon. J 4.2.4 Decay Heat Remov.al y _

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, . Decay heat can b'e remdved from .the reactor in several , .

ways.- T h e p r i m a r y m o d e o f h e a t .,r e m o v a l i s t h r o u g h t h e --

. steam generators. The Reactor Coolant (RC) system is

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_ designed to transfer heat to the steam generators by -

- natural circulation (i-f forced flow 6 sing RC pumps is 3

not'available) in all events.except large break LOCA's. Following a large break LOCA event, the core

- is cooled by the Emergency Core Cooling System (ECCS).

No active components inside containment are required to function to remove heat when using either steam generator cooldown'or ECCS. Ins'trumentation inside 3A# '*"

containment is used to monitor the' conditions and system functi'ons, but all pumps and valves ~(other than

.. - check valves) which'must function for heat removal are located in the Auxiliary Building or Main Steam Tunnel.a - ,s.

- + Normal cooldown with the primary system in the natural "

circulation mode removes heat by supplying cool

,, operated relief valves to reject heat to the ,

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atmosphere. One operable stgam generator is adequate .

  • - , . t o r'e m o v e d e c a y h e a t (Reference 8).

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  • The ECCS function is to provide cooling water to the' core after a LOCA4 The sources of water are the ,

ac,cumulator tanks -in containment,.the Refueling Water - .

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Storage Tank (RWST) which is located external to the Auxiliary Building, and the containment recirculation sump which collects leakage from the. break.

To' bring the ' plant t'o a cold shutdown condition, the . .

RHR system is nor^mally ,used.- After a non-LOCA HELBi .

the RHR system will take suction from the Loop 1 or 3 hot. legi cool the fluid in the RHR heat exchangers , -

y (' transferring heat _to the component cooling system) and reinject the fluid into the reahtor' coolant-system cold legs. Followin'g a LOCX,'the RWST is used.as a -

suction source followed by the use of th'e

  • recirculgt%on sump. The only active me'chanical '

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components'inside containdent used for cold shutdown. .

decgy' heat removal are the RHR hot leg suction valves., -

These valves are used only in non-LOCA evbnts.

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Other options.e'xist fog remov51.of decay h at. Cool s.

down to cold shutdown conditions.can be accomplished

by increasing the febdwater level in the steam gene- .

. rators with cooler water. This method eliminates the .

need for any, active equipment inside containment'to

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remove. decay heat. This method, although available after a HELB, was not found to be required by the .

. postulaEed events in the scope,of this study. ,

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It is also possi,ble td reach cold shutdown conditions u, by adding cool water to the reactor vessel via the charging system and removing. heat via the letdown

  • system,,the excess letdow'n system, or, if these paths are, unavailable, the pressurizer power operated relief -

This cooldown method-(primary system feed and valves.

bleed) is included in the'Braidwood Emergency .

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Operating Procedures'bu't is not necessary for any '

event within"the.. scope of.this study. .

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4.2.5 Offsite Release- *

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w To' prev,e'nt offsite radioactive" release, a barrier must -

be maintained between radioactive material such as reactor coolant and the atmosphere. For ndn-LOCA

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HELB's the reactor coolant system beundary forms this 1 WP barrier. No additional barriers are required. Aft.er l a LOCA, the containment integrity must be prese'rved.

Systems which penetrate the containment must be iso- .

o lated if they are open to both the primary system (or the containment atmosphere)*and the atmosphere.outside.

~ - containment. The Cont,ainmen,t Spray System is used to .,

- remove' radionuclides from the containment atmpspher,e *

  • after a LOCA and:to control the' sump pH. The a o

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. Containment Spray, as well as the Reactor Containment ~ * #',

Fan Coolers and passive heat sinks, remove heat from  ;

the containment atmosphere to maintain containment l

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4.3 ~

Single Failure Criteria Thg Standard Review'Pl'an (Reference 1.) is explicit in

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e its defi.nition of the Single Failure Crit 6ria_.for high , ,

and moderate energy'line. break.. Section 3.6.1 refers

< in ssveral places -tx) the assumption of a " Single active

  • component failure". This. clearly refers to fa,ilure of a component which must perform ~an active (as opposed to passive) function to support operation of a safe shutdown systems Active. components are ~

those which- must mechanically move or electrically change state to perform the required function.

Examples of active components would be pumps.which  ;

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must run or valves which must open'or_close."~ Examples l

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of passive components are pipes, valves which are not. * ,

required to function, cables, breakers, and sw-itches ,.,

which do not change electrical state or mechanical position. ,

'The definition'of single failure in 10CFR50 Appendix A 4 is slightly different from that in Reference l'. A 2

footnote to "the Appendix A definition indicates that . ,

passive failures of electrical equ,ipment should be assumed and that.the requirements for single passive failures of fluid systems are under-review. Section' s,

  • 3.6.7 of Reference ~l clarifies the fluid systems .

single failure requirements. Under loss of offsite power conditions 'the uncertainty "about considerati*on ' "

of passive electrical failures is ~ of no significance because a single active mechanical failure (die'sel *

" generator failure) causes loss of one electrical

. division and bounds all potential '

active and passive -

- elec'trical failures. .

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Events wh.ich do not resul,t "in loss of offsite power are'less well defined with respect to single failures.

. . Loss of an-entire electrical division would require a .,

, passive failure when offsite power is not lost.' .

j Although it is bel'ieved that the intent of the SRP is .

to consider failure of a single active component, for t'he purpose of this confirmatory study, loss of an electrical division as a single failure has been

  • - considered.

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-4 . 4 'onfirmatiof C Procedure .

. The procedure used to cdhfirm safe shutdown capability ' -

- varies depending.upon the nature of the component and u- .

the area of the plant undpr investigation. Some components, by their nature, may be assessed independikk

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- i dently ofiother. components. However, ,t.he operation of .

redundant system components *must be evaluated fh '

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-4 relation to other~ system functions in the event of component failure. These potential interactions have been considered as. required. This procedure assures that'a review of potential jet effects on safe

. shutdown components is pdrformed. . .

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The fgetors considered in the evaluation can be

, ademonstrated by a brief listing of the major steps ih:

the process: ,

,1. Electrical and Mechanical equipment, and power and control cables in a defined HELB

.. pone are assumed to be unavailable du'e to the specifi.c break in the arga. A matrix'of s -

. da' mage vs. break is maintained. -

'2. Instruments, instrument lines, and instrument ,

cables are located.with respect-to breaks and .

potential damage for individual breaks is .

, , determined.,

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, ,3. Safe shutdown piping and supports in proxi * ,

mitylto.pELB's'is evaluated for possible < -

loading and for verification that Westinghouse' System Standard Criteria

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. (Refer.ence 4) is not violated and;that redundant safe shutdown piping is available. . .

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4. Struct'uraI components subject'to jet' loading' '

- las well-as pressurization) are determined-i and checked for adequacy. Components such-as '

block walls which may fail are evalu.tted fo'r, ef fects on other safe shutdown comp'onents~ **:-

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  • s,uch as those listed above.

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L 5. For each defined break, all potential *

failures determined 'in this procedure are

  • considered simultaneously along with the limiting Single Failure. Safe ~ shutdown

.m capability is then evaluat'd. e e

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, 6. In the event safe shutdown capability cannot be shown, a more detailed r'eview of the geo.-

metric relationship of/the components and the -

, breaks is performed to show safe shutdown capability. _

If this procedure was unsuccessful a design change may have been required to meet the, design basis.. .

4.5 - Safe Shutdown' Components Components. required to withstandfor be protected from the effect of jet impi'ngement have been determined by iden.tifying equipment potentially used-to reach safe shutdown, as defined in Section 4.2. It should.be

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noted that, because of the redundancy and diversity of

. the Braidwood safety systems design, no single component or system is. required for safe shutdown' unless failures

  • occur in one or more independent 1 systems. As a result,.a unique safe shutdown ,

component list can be established for each postulated combination'of initiating event and single failure. -

To facilitate this confirmatory study,,a' single list

~ has been established which encompasses the events. If necessary the list can be modified and -edited for ..

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specific events.to establish safe shutdown _ capability. ,

4 . 5 . l** ~ . Identification of' Safe Shutdown Systems Safe shutdown systems can be' categorized in s.everal ways. .A group of fluid safety systems assure the cappbility to remove decay heat. These systems are:

Chemical and. Volume Control .(CV)

-- Safety Injection (SI) -

. , Residual Heat Removal (RH)

Auxiliary Feedwater ,(AF)

These systems are supported by two fluid support -

systems: ,,

Essential Service Water (SX) '

Compo,nent Cooling. (CC) r To rem'ove heat from the. core in non-LOCA events, the ..

Main Steam (MS) and Reac' tor Coolant (RC, RY) systems must retain the integrity of pressore boundar,ies and power operated relief valve operability to the extent that decay heat is removed. .

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For~certain severe HELB events, portions of the Reactor Protection System must be operable t'o initiate mitigation.

Electrical and HVAC support systems are required to

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, assure operability of fluid systems." The Containment Spray (CS) and HVAC systems may be required to control environmental conditions.

The systems listed here have been designed to assure i that safe shutdown canete achieved following initiat-( . ing events which may disable certain portions of safe shutdown systems because of'the physical location or system configuration.

4 '. 6 High Energy Lines High Energy-Lines are defined in Section 3.6.2 of the SRP (Reference 1) as those lines which, in normal plant operations, operate at conditions above 200 0F

! ^

and/or 275 psia for.more than 2% of the system operat-ing time. The Braidwood design purposely limited the number of HELB's in the Auxiliary Building to reduce the hazards associated with these lines. Startup feedwater pumps were installed.to assure that Auxiliary Feedwater lines are not required during normal plant operations. Tunnels were designed to contain Main Steam, Feedwater, and Auxiliary Steam' lines and to isolate them from safety related equipment, io

! As a result, in the Braidwood design, only 6 systems

contain piping which qualified as high energy. These .

systems are: ..

Reactor Coolant (RC, RY, SI Accumulators)

Feedwater (FW)

Main Steam (MS) l ~

Chemical and Volume Control (CV)

Auxiliary Steam (AS)

Steam Generator Blowdown (SD) l These 6 systems are designed to minimize the number of areas where safe shutdown systems and equipment could -

be affected by the results of a'high energy line break. This is accomplished by utilizing physical separation (distance and barriers) to isolate safe shutdown systems from high energy lines, and by protective features such as pipe whip restraints and

% ^ e 9

,.y, p. .--.m,.. - - . -.--_w,_r..,_-_____y. ..-..w-,,,,,-__-..,_-,w 7,,.-.w___.,m__-,m_.,,,-_.,,_-,e%.y , , , _ , . -, -- . . _ , . .

\ -.

1 Jet impingement. shields to restrict or eliminate ,

effects of high energy line breaks.

1 Only the last 3 of these system-(CV, AS, SD) are. . .

. located in the Auxiliary Building and the AS and SD ,

routing in safety related areas is very limited.

1

  • 4.7 High Energy Lins Breaks .

I

~

In the early pha.se of design, b'reaks were postulated in high energy systems following Reg. Guide 1.46.

i

' This resulted in breaks postulated at 1.ocations judged 4 to potentially threaten safe shutdown components. For _ , ,

this confirmatory study, breaks have been postulated

j. in accordance with the guidelines of Section 3.6.2 of F - the SRP (Reference 1) with the exceptions noted in

~

Section 4.1 of this study.

. 4.7.1 Jet Impingement Load Defini~ tion . ,

! The potential loads and. region of influence of high i' . energy line break jet impingement can be defined using -

the information available in'ANS 58.2 (Reference 2),

I and NUREG-CR/2913 (Reference3). Jets can be classi-fibd as either-subcooled, non-flashing liquid ' jets, o.r two-phase and steam jets.

4

  • ANS 58.2 is used to predict liquid jet loads. These jets are predicted from the charging portion of the CV i , system and the SI system accumulator piping. The CV -

system lines are pump discharge' lines which are limit-ed in discharge. flow by the pump runout and the piping configuration. Calculations (Reference 14) demon-strated that the loads from breaks in these lines are

! relatively low (less than 500 lbf total). The SI accumulator line breaks could potentially result in i,

higher loads because they are fed from a pressure ,

l vesse.l. However, these are' located inside containment i such that they*do not pose a safe shutdown hazard.

i NUREG-CR/2913 provides a simplified method for determining loads due to two phase and steam jets.

. The range.of conditions applicable to Braidwo~od is covered. Two general conclusions can be reached from

the report
'

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h 9

13 -

5- g wew6 n., -y 7-7-&9 -- pg --g-e- -19%---. -q-,gy9p m -e.g.m ww-g---y g-w>-eg, gen,eygi9.y.e-9mgr.-.,em97g- w g y a, wwge,y.-p.i .esep, _ mywgg, g pgav e m&m-Mse->a

-. . - . - = . .. .- . . ..

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1) Loads decrease rapidly as.the break to target distance-increases with the jet pressure ,

becomingl insignificant ~at some distance '

- - between 5 and 10-pipe' break diameters from the break.

4

2) Loads ar'e lower than predicted by previously used methodologies at distances greater than

. . 1 to.3 pipe break diameters (depending cn1

break conditions) .- -

1 ~

l References 2 and 3 were used to confirm that the ..

Braidwood design approach has resulted in acceptable protection against the effects of high energy line breaks. When the design was reviewed it was found in .

many cases that the required components would not be

- affectWd by postulated jets. In these cases, a .

l further review of the separation'of redundant 5

components was not performsd since adequacy was already demonstrated. Separation of components

' 'provides additional protection against HELB and other-hazards.

i l 5.0 Results of Conf.irmatory Study

1. The differences between the evaluation results pre-

~

viously reported for, Byron 1 in the 1984 confirmatory

~

report and the corresponding results for Braidwood 1 are summarized in this section. This is done in.a l manner which parallels the Byron 1 work. The components in the plant were divided into categories

~

of related. components. These categories are equipment and cables, instrument lines and cables, piping and ,

4 supports, and structure. Each-group was reviewed to

- .. determine the extent to which the components were .

~~

vulnerable to jet impingement and the potential interactions between breaks and components were

- identified. Then the individual breaks were reviewed to evaluate the total effect.of each break on the

.y, types of component 5 and, in turn, on the capability of the plant safeby 4ystems.

. This'was a# %f(icient approach to the confirmatory

!.

  • effort because most' equipment is not affected by HELB effects.- The original layout of the plant separates

- physically most safe shutdown components from the HELB locations. To fully determine the effects of a break on safe shutdown, it is necessary to consider the sum l

effects on the types of equipment and the resultant effe' cts on the functioh.of safety systems. With the ,

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~

8 1

(*$ .

individual components already -review 5d'for all HELB effects, the results.are easily found for the breaks. ,

.The process was considerably smaller in scope for ... ~

Braidwood-l because a number of: breaks'were eliminated

  • as a result of the Arbitrary Intermediate Break and

-Leak Before Break programs.

Section 5.1 describes'the effects of the postulated HELB's in the Auxiliary Building and Section-5 12, describes the effects of the postulated HELB's in the Containment. Appendices were included with-the Byron

'l repdrt. These were extensive calculations or '

summaries of calculations which contained the review of each safe shutdown ~ component. These calculations (References 9, 10, 11, 12, 137 and 15) have also been.

completed for Braidwood 1 but are not included with

  • this. report. Results of the calculations form the basis of'this section since the methodologies.and.

approaches are the'same as.those.used on Byron Unit 1. .

5.1 Auxiliary Building High-Energy Line Breaks ,

?.

-Relatively few areasiin the Auxiliary Building ~are potentially exposed to.HELB's and jet impingement.

The main steam, feedwater and portions of the auxiliary steam and steam generator blowdown systems are located in piping tunnels which contain no safe

^ shutdown components. >~

In the Auxiliary Building, high 3,nergy portions of the Auxiliary Steam, Steam Generator Blowdown, and--

~

Chemical and Volume Control Systems are located'in 17 HELB Zones. This section will summarize the ef.fects of .

HELB's in the Auxiliary Building.

-5.1.1 Auxiliary Steam Line Breaks ,

The auxiliary steam (AS) system provides low pressure (50 psig) steam for various plant process uses. The -

AS system is not a safe shutdown system. It is located in areas near the turbine building and in the radwaste areas. To allow routing'of some large .

diameter AS system piping through the auxiliary building without creating a HELB hazard,~a pipe tunnel is used.

. b

+

  • l N F'$

~. ._ _ _ __

e 5.1.1.1 Additional Braidwood 1 Analysis .

A design m6dification has been installed which ir.terlocks temperature switches, located near postu-lated break locations in auxiliary steam lines-in the auxiliary building with the steam supply valves to limit the en,vironmental temperature and provide automatic AS isolation. ,

The following safe shutdown equipment and , components were-identified as differences between postulated-jet

- impingement damages for Braidwood Unit I when compared to Byron Unit 1: _ ,,

o Pipelines ICC32A2.and 1CC34AB3/4 o Cables which serve - the auxiliary building HVAC exhaust and supply f an motors and the charcoal booster f ans.

o Power cable to Motor Control Center (MCC) 2AP23E -

o AS system temperature switches OTS-AS031C/D/E and 0TS-AS032C/D/E.

Safe Shutdown Evaluation Component Cooling system piping ICC32A2 and ICC34AB3/4 supply cooling water to the boric acid system - vent condenser which is not required for safe shutdown. In addition, damage to the CC Lines will not degrade the performance of the component cooling system.

Cables which serve either the auxiliary building HVAC exhaust and supply f an motors or charcoal booster fan motors may be damaged. However, the auxiliary building HVAC exhaust and supply fans ~can be started manually and the charcoal booster f ans are only required subsequent to a LOCA. The AS system break will not result in.a LOCA.

The loss of power to MCC 2AP'23E will affect the operation of Motor Operate'd Valve (MOV) OSX147. This. .

is ths only safe shutdown. equipment supported by MCC 2AP.2 3 E . The. MOV. is one of two normally closed valves

. J ',

which a're l'ocated i'n th'e parallel discharge paths f r'om .

~

  • the common component cooling heat exchangers leading

'to the cooling lake. Following the non-LOCA HELB, the ,

? - postulated single failure will result in one CC heat exchanger being available to achieve' safe shutdown.

The use of one CC He.at exchanger will only extend the time it takes to achieve safe shutdown.

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

. I

.. . <?.. s Following postulated damage to the AS system , ,

temperature switches, the control logic circuitry for l the AS isolation valves is designed to fail safe if

-the signal from the temperature switches is inter-rupted. Therefore, if the switches or cables are rendered inoperable, safe shutdown is not adversely [,

affected. Redundant isolation valves are also include'd in the design to accommodate single failures.

The safe. shutdown systems damaged by the postulated AS system breaks will not preclude safe shutdown subsequent to postulated AS system breaks.

5.1.2 Steam Generator Blowdown System Breaks The steam generator blowdown-(SD) system consists of lines from each steam generator which are routed from the Containment through the main steam tunnel.and from the Auxiliary Building to the blowdown condenser. The SD system is not required for safe shutdown.

A postulated HELB in the SD system may affect safeJ'*-

shutdown. capability if.the steam source (Steam Generator Blowdown) is not isolated to prevent exposure of safe shutdown equipment in the Auxiliary Building to temperatures in excess of their qualification.

5.1.2.1 Additional Braidwood 1 Analysis A design modification has been installed with interlock temperature switches which are located near postulated breaks on SD lines routed in the auxiliary building with a series arrangement for SD automatic isolation.

The safe shutdown equipment located in this zone a're the temperature switches which are used for SD system break isolation. These switches and their associated cables have been located such that they are not affected by jet impingement. There is no safe shut-down piping in th,is zone. Also, there*are no .

postulated damages to be added to those identified for '

Byron Unit 1. .

5. l'. 3 ' Chemical and Volum'e Control System Breaks ,

The chemical and vol'ume control (CV) system is a large and complex system with many functions. However, only a limited portion of the system is considered high

  • 4 -

energy and only a lim-ited portion of the system is

, required to safely shutdown the plant.

The high energy portions of the CV system are from the .

charging pump discharge nozzle to the reactor coolant system and to the RC pump seals, and the letdown flow path.

Fifty two HELBs were evaluated for the CV system for Byron Unit 1.* However, due to the elimination of Arbitrary Intermediate Breaks, thirty four HELB's are evaluated for Braidwood Unit 1. In addition, the postulated jet impingement damages due to CV system breaks are caused primarily by t.erminal end breaks, which resulted in a major reduction in equipment and components which could poten.tially be damaged by CV s,y, stem, breaks.

5.1.3.1 Additonal Braidwood 1 Analysis The following safe shutdown equipment and components were identified as differences between postulated jet impingement damages for Braidwood Unit 1 when compared to Byron Unit 1:

o 'Miniflow MOV's ICV 8114 and 1CV8116 -

o Pipelines 00G39A/B3 and 00G42A/B3 o AS system temperature switches OTS-AS031'F and 0TS-AS032F

.. . Safe Shutdown Evaluation MOV's ICV 8114 and ICV 8116 are two additional CV pump-miniflow valves which may be damaged along with valves-1CV8110 and ICV 8111 which were identified as being damaged for Byron Unit 1.. Jet impingement damages to all four miniflow valves will not cause a safe shutdown concern, as was stated in the Byron Unit 1-confirmatory report, since the valves will fail in the open position and thereby maintain the miniflow -

., ' circuit. In addition, the valves are required to

- close on an ESF signal subsequent .to.a LOCA.

.Therefore, since'the breaks causing the initiating event will not. result in a LOCA the. valves are not required to operate.

The Of f gas -(OG) system components support the operation of the hydrogren recombiners which are used during post-LOCA operations to control hydrogen concentration inside containment. H.ence, since the e

Off gas system piping and components are required for initiating events inside containment only, jet impingement damages due to outside containment non-LOCA HELB's will not affect safe shutdown.

Cables for temperature switches OTS-AS031F and OTS-AS032F which are required to isolate the AS system subsequent to a AS system break may sustain jet impingement' damages. However, since the AS system is not~ required for safe shutdown, inadvertent isolation of the system due to a CV system b,reak will not affect i safe shutdown.

As stated in the Byron Unit 1 confirmatory report safe shutdown can be achieved subsequent to a postulated CV system break in the auxiliary building.

5.2 Containment Building High Energy Line Breaks In the Containment, HELB's are postulated in the Reactor Coolant System (RC, RY), the Chemical and Volume Control System (CV),.the Main Steam System (MS), the Feedwater System (FW), the Steam Generator Blowdown System (SD) and the high pressure portion of the SI (Accumulator) System. Breaks in these systems will be categorized according to the effects of the initiating failure and the functions required to  %

mitigate the break and safely shut down the plant.

Breaks which cause a LOCA are classified as Reactor Coolant breaks regardless of the specific system identification of the failed piping.

5.2.1 Safe Shutdown Systems Systems used for shutdown following a HELB inside Containment may be required for all, part, or none of the postulated. events. The need for some of the systems is based on availability of other systems.

Some of the more important safe shutdown systems can be shown to be unaffected by any postulated HELB's o inside containment as a result of the design of the systems. In this section',.uses and design features of safe shutdown systems are supmarised. Those systems or system functions which are.shown to be.available '

after all HELB',s will not'thdn be repetitiously

' discussed for each type of break.

C t

4

3. . .

1 - ~ ,,

.s 5.2.1.1 Main Steam (MS) System Following a HELB, the MS System is used in conjunction with the AF System to remove decay heat. The steam generator power operated relief valves and/or safety valves are used to release steam to'the atmosphere.

The valves are located in the valve rooms of the Main Steam' Tunnel. Equipment, instruments, and cables required for the MS system function are not located inside the containment. The MS system will be available for the applicable break cases examined in Section 5.2.2.

5.2.1.2 -

Feed' water (FW) System ,

The FW System has no active components inside containment. The only required function of the FW System following a HELB in containment is to provide a secondary steam system pressure boundary. The FW System will fulfill its safety function for the applicable break cases examined in Section 5.2.2.

5.2.1.3 Essential Service Water (SX) System The SX System has only one safety function which includes components inside the containment. This is the cooling water supply to the Reactor Containment Fan Coolers (RCFC's). There are no active components inside Containment. The SX System will fulfull its safety function for the applicable break cases examined in Section 5.2.2.

5.2.1.4 Containment Spray (CS) System The CS System is used following a LOCA. The CS System will remove heat from the Containment atmosphere and control the concentration of radiation in the Containment atmosphere both by washing the atmosphere and by controlling the containment sump pH. There are no active components inside containment. The CS system will accomplish its safety function for the applicable break cases examined in Section 5.2.2.

5.2.l'.5 Residual Heat Remt /al (RH) System- .

The RH System functions in two distinct modes. follow-

'ing a HELB Inside. Containment. F611owi'nq a LOCA,.the RH pumps serve as low head ECCS pumps, . initially 4 taking suction from the Refueling Water Storage Tank (RWST) and subsequently from the Containment Sump .

< s,

". e2 0 -

6 e gf

  • 4 8 1 j

(recirculation mode). Following a LOCA, RH System equipment, instrumentation, and cables inside .

containment are not required for safe shutdown. The RH System will fulfill its safety function for all break cases examined in Section 5.2.2.

~

The RH System is not required to operate to achieve hot shutdown following a non-LOCA HELB event.

However, following the non-LOCA HELB, the RH system may be utilized to achieve cold shutdown. In addition, the RHR loop suction valves and associated cables located inside Containment are used for cold shutdown after these events.

5.2.1.6 Reactor Coolant (RC/RY) System The RC System is considered to include the primary system portion of the RY System and portions of other systems which are connected to the primary coolant system. The RC system can perform its safety functions of heat removal and prevention of radioactive releases since it has no active components which are used during safe shutdown. For each break case in Section 5.2.2, the potential effects on integrity of the RC System have been reviewed and

~ resolved.

5.2.1.7 Safety Injection (SI) System The SI System includes injection paths to supply water to the RC System from the centrifugal charging pumps, safety injection pumps, and residual heat removal pumps. The SI System is used following LOCA's. The SI system will fulfill its safety function for the applicable break cases examined in Section 5.2.2.

5.2.1.8 Chemical and Volume Control (CV) System The CV System inside Containment consists of the normal charging, seal injection and letdown paths.

Jet impingement effects on the CV System are addressed for.the applicable break cases examined in Section

- 5.2.2. -

502.1.9 .. Component Cooling (CC) System .

- The CC System has o'nly one. function inslde Containment

. which may be required for safe shutdown. This is supply of cooling water to the Reactor Coolant Pumps

. ,(RCP's), thermal barriers. If seal injection (CV b

't g

~

System) flow is interrupted in a non-LOCA event,.the CC flow to the thermal barrier insures seal integ'rity and prevents leakage of primary coolant. Jet impingement effects on the CC system are addressed in the applicable non-LOCA break cases examined in Section 5.2.2.

5.2.1.10 ESF/ Reactor Trip Following a HELB, automatic reactor trip and safety system initiation will occur as required based on -

signals from qualified instrumentation. After the automatic functions are initiated, manual actions are '"

taken by the plant operators based on qualified instrument readings and the Braidwood Emergency Operating Procedures. Each type of accident will cause a unique response of the reactor and steam supply system, and therefore, requires a different set of functional instruments for automatic actions and monitored output for manual actions. For the breaks postulated in containment, ESF/ Reactor Trip instrumentation will be available as required. This is summarized for the applicable breaks in Section 5.2.2.

5.2.1.11 Containment Isolation .

Fluid Systems which penetrate Containment but do not have a safety function following a LOCA are automati-cally isolated foll.owing the break if high containment pressure or radiation signals are generated.

Containment isolation will be achieved following postulated LOCA's.

5.2.1.12 Off Gas (OG) System ,

The OG System is designed to maintain the free hydrogen concentration in the containment atmosphere below the flammability limit of 4.0 volume percent following a LOCA. The OG System is not adversely affected by postulated jet impingement affects.

5.2'l.13

. HVAC Inside Containment .

TheHVACdfsteminsideContainmentconsists,ofthe Reactor Containment Fan Coolers (RCFC's). The RCFC's are supplied with cooling water by the Essential Service. Water (SX) and Chilled Wat.er (WO) Systems.

Only the SX is required after a HELB. The Containment e

a g

Spray system provides~a backup means of heat removal

. from the Containment. The availability of SX water has been addressed in Section 5.2.1.3.

5.2.1.14 Auxiliary Feedwater (AF) System The Auxiliary Feedwater System is'used to supply water to the steam generators to remove decay heat either to maintain the reactor in a dot standby condition or to proceed toward cold shutdown. The AF System contains no active components inside containment.

5.2.2 Summary of Jet Impingement Effects In this section, the postulated HELB's inside Containment are classified according to the break effects and the systems and components required for subsequent safe shutdown. For each type of break the systems required and the potential effects of jet impingement are reviewed. Single failure is considered and the resulting safe shutdown capability is reviewed to assure that jet impingement from HELB's inside Containment does not adversely affect safe shutdown.

5.2.2.1 Types of HELB's Inside Containment The postulated HELB's inside containment have been classified into LOCA and non-LOCA events. LOCA's have been divided into three types: Large Liquid LOCA's, Small Liquid LOCA's, and Steam Space (Pressurizer)

LOCA's. The non-LOCA HELB's have been divided into .

six types: Main Feedwater,-Main Steam, Bypass Feedwater, Charging, Steam Generator Blowdown, and Safety Injection (Accumulator). .

5.2.2.2 LOCA LOCA's are those HELB events which result in a loss of primary coolant to the Cont.ainment.. LOCA's which occur in liquid lines may result in a two phase blowdown while those occuring in steam lines result in steam release. LOCA's may or,may not be isol'able -

depending upon break location.

  • 5.2.2.2.1 Large Liquid LOCA'~s Large'liquidLOCA'saredefinegasthosebreakswith' These breaks occur an area of grea'ter than 1.0 ft .

in the pressurizer surge line only. All breaks in the

- 23 -

., , wm-- - ,c- e- ---~e.- -,,on

main loop of the Reactor Coolant system have been deleted based on the Leak-Before-Break program. In addition, th,e number of breaks occuring in the pressurizer surge line have been reduced to two terminal end breaks due to the elimination of Arbitrary Intermediate Break's (AIB's). As a result, the consequences due to damage from jet impingement 6 for Braidwood Unit 1 are enveloped by those for Byron

~

Unit 1.

5.2.2.2.1.1 Safe Shutdown Requirements To bring the plant to a safe shutdown condition following a large liquid LOCA, the reactor must be ,

, tripped and necessary plant parameters monitored.

Containment isolation as required to prevent offsite release must be accomplished. Heat must be removed from the containment atmosphere and decay heat must be removed from the reactor vessel. To assure that the event stays within the analyzed designed basis, break propagation must be controlled as described in Westinghouse Design Criteria SS 1.19 (Reference 4).

Pressurizer pressure and containment pressure signals

  • will trip the reactor and initiate containment isolation and emergency core cooling (ECCS).. In addition, the wide range reactor coolant system (RCS) pressure, the Containment pressure, the Main Steam pressure, the refueling water storage tank (RWST) level, and Containment Radiation level are used to monitor the plant conditions.

Following this event, the CS system is used to cool the containment and clean the Containment atmosphere.

The RCFC's are also used to cool the Containment. The

~

OG system (Hydrogen Recombiners) may be used during the long term containment atmosphere cl.eanup.

Initial and long term decay heat removal is provided by the ECCS System operating initially in an injection mode (RWST) and ultimately in a recirculation mode (containment sump). For this event, the SI accumula -

tors are required (three injecting and one spilling' -

through break) to reflood the core as well as one of a the following three systems or combinations of systems to replace core coolant boil-off: ,

a. one train of the residual heat removal system, or S

k

+

4 I b. one train of the high head. safety injection

system in conjunction with the use of one

]

residual heat removal pump and one residual heat exchanger (of-the same train as the high head safety injection system) to provide .

suction from,the sump, or

c. one train of the charging / safety injection l'

system-in conjunction with the use of.one

- residual heat removal pump and one residual

    • heat exchanger (of the same. train as the-

! charging / safety injection system) to provide

! suction from the sump.

I 5.2.2.2.2 Small Liquid LOCA's a o

l SmallliquidLOgA's.arethosewithabreakareaof These breaks are similar in effects less than 1.0ft .

to the_large breaks except the rate of break flow,.RC t

system depressurization, and Containment pressuriza-1 tion are all slower. The wide range of break sizes add to the total list of equipment and components

- which may be used because of the variety of options ,

available to achieve safe shutdown. These breaks are located in the lines connected to the reactor coolant i

loops. ,Most"are lo~cated in short sections of piping ,.

f- between the loop and an isolation valve. The RC loop t

bypass, piping and the RTD manifold piping is located' 1 between the hot and cold legs of the loop which restricts the breaks to an area near the faulted i l'oop. The small liquid LOCA break outside the secondary shield is in the letdown line. The effects .

i i , of this break are minimi~ zed due to the flow sestricting orifices in the line. .

I- Breaks postulated to cause small liquid LOCA's are reduced by over forty percent for Braidwood Unit 1

~

when compired to Byron Unit 1 due to the elimination j of Arbitrary In.termediate Breaks. This reduction in l postulated break , locations resulted in less safe ,*.

l shutdown piping, equipment and components being ,

-affected by jet impingement for Braidwood. Unit 1 when j * '

c compared'to Byron Unit 1. In addition',.the safe -

. e .

shutdown targets. identified and evaluated.for .

}, ., ',Braidwood Uni.t 1 were also evaluated for Byron' Unit '

' - 1. Therefore, as. determined for Byron Unit 1 the safe

]

shutdown requirements for a small liquid LOCA will not -

i I be' violated.

J =

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_ _ _ _ - _ , . . , - _ _ - - _ _ _ . _ - _ _ _ - _ . _ _ _ -_ _1_. _ .

. I e l 5.2.2.2.2.1 Safe Shutdown Requirements To bring the plant to safe shutdown condition following a small liquid LOCA, the reactor must-be.

tripped and necessary plant parameters monitored.

Containment isolation must be accomplishedas required to prevent offsite releases. Heat must be removed from the containment atmosphere and decay heat mus,t be removed from the reactor vessel. To limit the severity of the event, break propagation must be restricted..

Instrumentation required for ESF initiation and for monitoring after the event are listed in Reference 7.

Pressurizer pressure and containment pressure signals will trip the reactor and initiate Containment -

isolation and emergency core cooling (ECCS). In

~

addition, the wide range RCS pressure, Containment pressure, main steam pressure, RWST level, pr,essurizer*

level, narrow range steam generator level, core exit

  • temperature, and containment radiation level'are used to monitor the plant conditions. .

- Following this postulated event, the CS system may be used to cool the Containment and clean the Containment atmosphere. The RCFC's are also used to cool the Containment. The OG (Hydrogen Recombiners) system may be used during the long term containment atmosphere _

cleanup.

Initial and long term decay heat removal is provided by the ECCS system operating initially in an injection mode (RWST) and ultimately in a recirculation mode (containment sump). For most of these postulated events, the secondary system (steam generators) will -

remove decay heat also. For these events, the required flow to the reactor vessel is dependent upon break size. For the smallest breaks, the centrifugal charging pumps op'erating in the safety injection mode . . .

can maintain the RC system inventory. For larger breaks,.the accumulators (three injecting and one

. . spilling through the fault.ed. ling) may be required. .

Therefore, availability of the accumulators'and one l train of charging / safety injection, high he'ad sa$bty,. -

injection, and we'sidual ,hea,t removal was evaluated.

  • a 9

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= .

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5.2.2.2.3 Steam Space LOCA's These LOCA's are postulated to occur when a pipe attached to the. upper portion of the pressurizer is ruptured. This type of break can occur in the pressurizer spray line,.the pressurizer Power Operated Relief Valve (PORV) lines, and the pressurizer safety valve lines. The mass flow rate'is less from these

~

breaks than an equivalent liquid break because.of the .

reduced density of the steam. The targets affected due to steam space LOCA's for Braidwood Unit 1 are the same as those affected for Byron Unit 1. This is because there are no breaks deleted by the Arbitrary Intermediate Break or Leak Before Break criteria which caused steam space LOCA's. Therefore, as proven in the Byron Unit 1 Confirmatory Report safe shutdown capability will not be adversely affected by jet impingement since all the required safe shutdown systems will remain operable subsequent to the HELB.

5.2.2.2.3.1 Safe Shutdown Requirements ,

wr To bring the plant to a safe shutdown c'ondition . .

following a steam space LOCA, the reactor must be tripped and necessary plant parameters monitored.

Containment isolation as required to prevent off-site release must be accomplished. Heat must be removed from the containment atmosphere and decay heat must be 4 removed from the reactor vessel. As. discussed in .

Westinghouse Design Criteria SSI.19, these breaks are allowed to cause additional primary system' steam space breaks but should not cause a liquid LOCA or secondary, system breaks. ,,

Instrumentation required for ESF initiation and.for monitoring after the event are listed in Reference 7.

Pressurizer pressure-and containment pressure signals will trip the reactor and initiate containm'ent isolation and emergency core cooling (ECCS). In

  • addition, the wide range RCS press'ure, the Containment -

,. pressure, the main steam pressure, the RWST, level, the narrow range steam generator level, the cor,e exit ,.

. . temperature, and-containm,ent radiation are,used to-monitor the plang conditions. - *

  • Foll'owing this event.,' the CS system is used to cool the Containment and clean the containment. atmosphere.

The RCFC's are also used to cool the containment. The c -

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27 -

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OG system (Hydrogen Recombiners) may be used during long term containment atmosphere cleanup.

Initial and long term decay' heat removal is provid'ed

  • by the ECCS operating initially in an injection mode (RWST) and ultimately in a recirculation mode (Containment sump). Also, the secondary system (steam .

generators) is available to remove decay heat. As was noted for the small liquid breaks, the SI components used are, to some extent, dependent on'the break

  • size and the rate and extent of primary system depressurization. The accumulators and one of the pumps (Charging, Safety Injection or RHR) are adequate -

to maintain RCS Inventory. The SI system, as noted in Section 5.2.1, is designed such that required equipment or instrumentation is not located inside Containment. ,

5.2.2.3 Non-LOCA HELB's . .

HELB's which do not result in a loss of primary '

coolant occur in the secondary coolant system (Main

. Steam, Feedwater, Steam Generator Blowdown) and the-systems which serve the' primary system (charging, Safety Injection). For these events, decay heat is removed via the Auxiliary Feedwater and Main Steam Systems (see Section 5.2.1.1 and 5.2.1.2). Because the primary coolant boundary is intact, the containment isolation function is not required.*

5.'2.2.3.1 ' Main Feedwater Line Break The Main Feedwater lines are four 16-igch lines which supply the four steam generators. Based on the deletion of Arbitrary Intermediate Breaks only two breaks remain per loop. These are located at steam generator nozzles and at containment penetrations.

The postulated breaks will cause a reduction in water level and pressure in one steam generator, an'd subsequently an increase in containment pressure. Due to the reduction in postulated break locations very few safe. shutdown targets *are impi,nged by HELB jets

' and.those which-are' determined to incur, impingement "

were al'so identified by.the Byron Unit 1 Confirmatory'.

study. Therefore,, as determinbd for Byron Uni.t 1 the .~.-

. . . safe shutdown requirements for Braidwood Unit 1 following a Main Feedwater line break will not be-violated. ~

- 28'-

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5 5.2.2.3.1.1 Safe Shutdown Requirements To reach a safe shutdown condi'tionIf ollowing' the

,' event, the reactor must be tripped and plant

- conditions monitored., Heat must be removed from the

~

. e- Containment atmosphere and decay heat must be removed. -

from the reactor coolant system. The break must be confin'ed to the secondary system and not cause a -

release of. primary coolant. ,

Instru' mentation required for ESF initiation and for '

monitoring af ter the event are listed in Reference 7.

.Mai.n steam pressure and' narrow range steam. generator level provide the signals' which trip the reactor and ~

initiate TSP function's. Although the containment ~is' isolated on high containment pressure, this is not necessary following~a non-LOCA' event. Con'tainment pressure is..used to monitor the plan,t,conditi6ns, as-well as wide range RCS pressure, pressurizer level,-

and' core exit temperature. Containment radiation is

  • mon (tored to verify the HELB is not ,a LOCA.

~

- The RCFC'.'s rsmove containment atmosphere heat. ;The ,

Containment Spray System, although it is available for heat removal, is not requiret following a main

  1. - feedwater line break. One functio,nal Auxiliary Feedwater train and one functional steam generator remove decay h' eat to maintain the re. actor at hot -

standby conditions. '

~ . . ., ~ , -

5.2.2.3.2 ' Main Steam Break .

2 The four Main Steam lines transport steam from each

, steam' generator to the various system 'compon'ents ,

located in the, turbine building. A total _of twenty breaks were postulated in the main steam lines for the By ron ,U n'i t 1 evaluations; however, b'ased'on the deletion of ' Arbitrary. Intermediate Breaks o'nly eight

~

tbrminal. end breaks (two per loo evaluated for Braidwood Unit D. p) 'These remained breakand were locations '

are postulated t'o occur at the steam generator hozzles and at containment penetrations. The. jet impingement

. analyses for these breaks determined thad no safe shutdown equipment and components required subsequent .

to a Main Steam Line Break will be damaged by the -

requirements"as discussed below will not. be violated *

'and safe shutdown.can be achiev,e'd. '

. 29 '

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5.2~.2.3.2.1 Safe' Shutdown Requirements  !

1 3

To reach a safe shutdown condition following'this~ -

l event, the reactor must be tripped and' plant condi2*~

~

,^ tions must be monitored. Heat must be removed from the containment atmosphere and decay heat must be removed from the reactor coolant system. The break '

must be confined to the secondary system and not cause, *'-

a release of primary coolant.

> tv Instrumentation required for ESF initiation and for monitoring after the event are listed in Reference 7.

Main steam and pressurizer pressure reductions and containment pressure increase will cause reactor trip. The containment will also be isolated but this -

,is not,necessary following this non-LOCA event.

Additional parameters which are monitored are wide range RCS pressure, pressurizer level, narrow range steam generator level, core exit temperature, and containmen,t radiation. ,

The RCFC's remove containment atmosphere heat. The Containment Spray System, although it is available for -

heat removal, is not required following a main steam -

break.

One functional Auxiliary Feedwater system train and one functional steam generator removes decay heat' after a main steam line break. The charging and safety injection systems, which can be used to maintain RC system volume and boration level,during shutdown, contain only piping components inside containment. -

The other systems used for safe-shutdown ar.e not located in the Containment. -

Bypas's Feed' water Line Break

~

5.2.2.3.3 ,

~

Based on the elimination of Arbitrary Intermediate Breaks, postulated breaks in the bypass feedwater ,

lines are reduced to only two per loop which are

. located at steam generator nozzles and at containment penetrations. These breaks are in 6-inch lines and would ihitially release two phase fluid, but, as the '

steam gene,rator. level drops this would change to -

steam. Therefore, the. jet impingement zone of influence would be limited to 10 pipe diameters. Due .

-to the rednction in postulated breaks and' the limit'ed jet impingement zone of influence no safe shutdown e

4

.e equipment and components which are required subsequent to a bypass feedwater line break will be damaged.

Therefore, as determined for Byron Unit 1 the safe shutdown requirements will not be violated.

5.2.2.3.3.1 Safe Shutdown Requirements To reach a safe shutdown condition follo, wing this ,

event, the rea.ctor must be tripped and plant condi '

tions must be monitored. Heat must be removed from the containment atmosphere and decay heat must be removed from the reactor coolant system. The break- .,

must be confined to the secondary system and not cause

- - a release o.f primary coolant. ,

Instrumentation required for ESP initiation and for

. monitoring after the event are listed in Reference ,7.

Containment pressure, main steam pressure, and the narrow range RCS temperature RTD's will provide input. ' i to trip the reactor. The containment pressure, main -

steam pressure, wide range RCS' pressure, pressurizer .

- level, narrow range' steam generator level, core exit temperature, and containment radiation will be used to ....

- monitor the plant condition.

The RCFC's remove containment atmosphere heat. The Containment Spray system, although it is available for.

heat removal, is not required following a feedwater bypass line break. .

One functional Auxiliary Feedwater system train and -

one functional steam generat~or will remove 'ecay d heat ,

after a feedwater bypass line break.

The charging and safety injection systems, which can be used to maintain RC system volume and boration ,

during shutdown, conta'in only piping components inside containment.

  • The other qystems used for safe shutdown are not located in the containment.

~

5.2.2.3.4 Charging Line Break Charging line breaks are postulated on the normal charging and' seal injection lines upsfeam of*the -

isolation valves at the RC system and RC pump connections. Other postulated Chemical and Volume .

Control (CV) System piping breaks will result in a loss of reactor coolant and were addressed in Section

e O

w T {

e 5.2.2.2.2, (Small- Liquid'LOCA's). Based on the

- deletion of AIB's, Non .LOCA'CV system breaks inside#

containment were reduced to 33 breaks from a total of -

53 for Byron Unit 1. Due to this reduction in postu-4 lated break l'ocations fewer safe shutdown equipment'

  • and conponents are identified as being impinged for Braidwood Unit 1 when compared to Byron Unit 1. In addition, the safe shutdown equipment and components ~ r identified for Braidwood Unit 1 were also evaluated in the Byron'Un'it 1 Confirmatory , report. However,

~, pressurizer pressure transmitter 1PT-458 may be affected by jet impingement from a charging line break for Braidwood Unit 1. This transmitter is not

~

required following this type of break. Therefore, the safe shutdown requirements as presented below can be achieved" subsequent to a CV system break. ,,

5.2.2.3.4.1 Safe Shutdown Requirements f*'

~

Following a. charging line break', t e reactor will not ~

be automatically tripped because no ESF signal will be

~

generated. To bring the plant to a safe shutdown condition normal plant procedures can be used. . .

Charging is still availa,ble because two of three paths (Normal, Charging /SI, Seal Injection) will remain funct_lonal. .

Th'e RCFC's remove contai,nment atmosphere heat. The

- normal Feedwater system or one auxiliary feedwater train in conjunction with at least one functional

- steam generator remove decay heat. If the break is in the seal ~ injection system, component cooling supply to the RC Pum'p thermal barriers must be provided to .

prevent RCP seal damage.

~

. Instrumentation requi. red for monitoring after the break ar'e listed in Reference 7. The containment pressure, main steam pressure, and containment radiation instrumentation are outside of the' ~

containment. Equipment, cables, and/or sensing lines fo'r the wide range RCS prsssure, pressurizer level, ,

narrow range steam generator. level, and core exit temperature are located inside cdHfhinment.

S' team Generator Blowdown (SD) Line Break

~

5.2.2.3'.5 Steam Generator Blowdown line' breaks are 1 1/2 inch or ~ .

2 inch breaks in the liquid Steam Generator boundary;

'There were eight breaks per loop previously identifie'd and considered for the Byron Unit 1 analyses.

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However, these have been reduced to four terminal end "

breaks per loop which are located at the steam generator nozzles and at containment penetrations.

~

Based on their locations'these breaks will not cause -

the, impingement of safe shutdown equipment which are ,

required to function subsequent to SD system breaks.

Therefore, the safe shutdown requirements subsequent to SD system breaks will not be violated.

5.2.2.3.5.1 Safe Shutdown Requirements Following a SD line break, the reactor will be. tripped on low level in the affected' Steam Generator. A normal shutdown procedure is then used because of the small size of'this break.

The Main Steam pressure instrumentation is located .

outside the Containment. Equipment, cables, and/or~ '~

sensing linbs for the wide range RCS pressure, -

Pressurizer level, narrow range Steam Generator level; -

and Core Exit temperature are located inside the Containment. ,

The RCFC's remove Containment atmosphere heat. One

- Auxiliary Feedwater Train in conjunction with one functional Steam Generator will remove decay heat. -

5.2.2.3.6 Safety Injection Line Break Safety Injection line breaks are postulated.to occur in the portion of piping normally pressurized by the .

accumulators. The pipes contain ambient temperature liquid at.700 psi. A postulated HELB occuring in SI.

piping does not cause reactor trip or affect equipment -

which are required following Safety Injection.line breaks. A total of sixty four breaks were analyzed .

for Byron Unit 1, however, due to the elimination of Arbitrary Intermediate Breaks only eight breaks remain'de and were evaluated for Braidwood Unit 1.

This resulted in very little safe shutdown equipment being impacted and of those which are affected none -

are required to oper' ate subsequent to a safety

- injection line break. Therefore, the safe shutdown requirements subsequent to a safety injection line' break will not be violated.

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5.2.2.3.T.1 Safe Shutdown Rsquirements ',

Following a SI 1ine break, the reactor will not,.be' ' :-

automatically.. tripped because no ESF signal will ,,

~

result. To bring the. plant to a safe shutdown ,

can'dition,g normal plant procedures gan be used. -

n 4 .

~'

The RCFC's will remove the normal containment heat- .

load. . The normal feedwater system or one auxiliary feed' water ~ train'in conjunction with one functjonal., *

-> The main steam pressure instrumentation is located ' '

. outside containment...Eguipment,, cables, and/or sens-ing lines for the wide range RCS pressure, pressurizer .

- level, narrow range steam generator level, and core exit temperature.are located inside containment. , ,

6.0 Conclusion . . .

A detailed,evaiuation of potential jbt impingement effects utilizing the current requirements for break postulation and the location and design of. safe shutdown components and structures has demonstrated the adequacy of the Braidygod Unit 1 design. .

Postulated jet impingement effects will not result.in -

an inability'to safely shutdown the plant.

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7.0 REPERENCES* -

._- . . - \

...gNUREG-75/087,,

~ n..-

1. " NRC Standard Revie:w Plan for the Review of

+ Safe'ty Angl*ysis Reports for Nuclea'r Power Plantsf September 1985. -

. ANS 58.2, American National Standard , Design Basis for,,

2. .

Protection of' Nuclear Power Plants ' Against Ef fects. of Postulate'd Pipe Rupture. '

o,

~

J. 3. NUREG/,CR-2913, "Two Phase* 3 Jet foads ," "Jhnuary 1983.

4.
  • Westinghouse Design Criteria SS 1.19' - CTiteria Tor; brotection

.*., A* gainst Dynamic Effects Resulting from Pipe Rupture, Revi"sion -

.No. 1, August ~1980. .

, _ x.~'_

5. NRC-Letter, " Issuance of Amendment No. I to constructi'on -

pe rmits CPPI't-131, CPPR-1,32, andCPPR-133 fiir thb By'ron .

~

. Station, Unit 2 and Braidwood Station, Units 1 and 1, ce April.29,. 1986. -

3 .

6. NRC Letter, " Byron /Brdidwood - Elimination of Arbitrary

. Intermediate Breaks", January -7, 1985. _

7. -

Bhron 1 Confirdion of. Design Adequacy for Jet Impingem'en t.

    • :Ef fec ts , Augus t 198 4.. .,,

-+

8. 'l;)raf t Technical Specification fos' Byron Station ~ Unit l' - " ~ '

(December, 1983)'.. -

ape;

9. Sargent & Lundy Calculation No. 3C8-0785-003, Revision 0,

~

" Verification of High Energy. Line Break Design Approach for.

Jet Impingement Effects on Safe Shutdown Equipment, ..

Instrdmentation, and Cables (nutside' containme,.nt) - Braidwo'od~ - -

Unit 1", J6ne 12, 1986

  • Revfsion'0, "Th'e

~

10.

Sa rgen t & Lundy Calculation No. 3C8-0 Influence.of Partition Wall Integritp,98,5-001," on Plant, Safe Shutdown'..

Br.aidwood Unit 1", June-12, 1986 _

11. Sargent & Lundy Calculation No. 3C8-0186-002, Revision 0, .

" Verification of HELB De' sign ' Approach' for Jet Impirigement Ef fects on Safe Shutdow'n Piping", Braidwood Unit 1",

1986.

June , -

7 ,. . .. ,. . .w

12. Sargent.& Lundy Calcblation No. 3C8-0486-002, Revision 0,'- .

" Verification of HELB Design Approach- for Jet Impingement *-

Effects on Safe Shutdown Pipind' Supports - Braidwood Unit 1, July,1, 1986. - --

^

~ ". '

35 ,. , , . .

9

  • ,A y a f n

.~

. u .

~

. , , t. '#_ '

, p. .

. e .o .

e.1 13.. Sargent'& Lundy Calculation No. HELB-24, Revision 0, Verificdt' ion of Design Adequacy for-Jet Impingement Effec'ts on, Safe Shutdown Sensing Lines and Cables located Inside ' '

Containment'and. Sensing Lines located Outside Containment - ,

Braisiwood Unit,1", June.9, 1986.

  • e.. *
14. SargeriP& Lundy Calculation No. EMD-052558, Revision 0, " Loads -

< on Structures Due to High Energy Line %reaks in the Auxiliary Building", July 1, 1985. .

15. Sargent - & ,Lun$y Calculati*off"No. HELB-22, Revision 0, -

" Determination of Differences Betkeen Braidwood 1 and Byron'1 -

StructQral Steel Loadings Inside Containment", April 14, 1986.

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