ML17194B267

Forwards Safety Evaluation Addressing Open Items in 820117 Draft Safety Evaluation of SEP Topic III-5.B, Pipe Break Outside Containment. Facility Adequately Protected from Pipe Break Effects,Subj to Listed Resolution

ML17194B267
Person / Time
Site:	Dresden
Issue date:	08/20/1982
From:	Oconnor P Office of Nuclear Reactor Regulation
To:	Delgeorge L COMMONWEALTH EDISON CO.
References
TASK-03-05.B, TASK-3-5.B, TASK-RR LSO5-82-02-045, LSO5-82-2-45, NUDOCS 8208250147
Download: ML17194B267 (17)

Text

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Docket No. 50-237 LS05-82-08-045 Mr. L. DelGeorge Director of Nuclear Licensing Commonwealth Edison Company Post Office Box 767 Chicago, Illinois 60690

Dear Mr. DelGeorge:

.~

August 20, 1982

SUBJECT:

SEP TOPIC 111-5.B, PIPE BREAK OUTSIDE CONTAINMENT DRESDEN NUCLEAR POWER STATION UNIT 2 By letter dated January 17, 1980, the staff issued a draft safety evaluation on this topic. Your letters of March<?), 1980 and July 16, 1982, responded to open items from the draft evaluation.

The enclosed safety evaluation address~s the open items from our draft evalua-tion. The staff concludes that Dresden Unit 2 is adequately protected from the dynamic effects of pipe breaks outside containment, subject to the resolution of the following fn the Integrated Assessment; For main steam, isolation condenser and cleanup piping, provide resolution for staff concern that postulated pipe breaks in the containment penetration-area with an assumed single failure of the inboard isolation valve would result in an unisolable p~ipe, break outside of containment (i.e., a loss of coolant acci~ent).

The need to actually implement chang~s as a result of this_ evaluation w111 be detennined during the Integrated §lafety Assessment. This topic.assessment

• may be revi~d in the future if your facility design is changed or if NRC criteria relating to this topic are modified before the Integrated Assessment is completed.

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Sincerely, Paul W. O'Connor, Project Manager-Operating Reactors Branch No. 5 Division of Licensing cc_ wlenclosur..e.:

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• See pr~v1ou~ ye11ow~r ao~t1~al concurrences.

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SEPB:DL NRC FORM 316 (10-60) NACM*0240 USGPO: 1981-335-960

Docket No. 50-237 LSOS Mr. L. DelGeorge

~

.Director of Nuclear Licensing Commonwealth Edison Company Post Office Box 767 Chicago, Illinois 60690

SUBJECT:

SEP TOPIC III-5.B, PIPE BREAK OUTSIDE CONTAINMENT DRESDEN NUCLEAR POWER STATION UNIT 2 By letter dated January 17, 1980~ the staff issued a draft safety evaluation on this topic. Your letters of March 21, 1980 and July 16, 1982, responded to open items from the draft evaluation.

The ~nclosed safety ev~luation addresses the open items from our draft eval-

-uation. The staff concludes that Dresden Unit 2 is adequately protected from the dynamic effects of pipe breaks outside cont<i"inment, subject to the resolution of the following in the Integrated Assessment.

The various options (as listed in the enclosure) to resolve the issue.

of the effects* of pipe br~i;lkS outside containment- *in the containment penetration area for thenmaijnssteam, isolation condenser and cleanup pfpingg The need to actually implement changes as a res~~t of this evaluation wfll be detennined during the Integrated Safety Assessment. This topic assessment may be revised f n the future if your facility des.ign f s changed or if Nile::

criteria relating to this topic are modified before the Integrated Assessment is completed.

Enclosure:

As stated SincereJy, Paul W'~ O'Connor, Project Manager Operating Reactors Branch No. 5 Division of Licensing m

• See previous concurrence NRC FORM 318 (10-80) NRCM 0240 OFFICIAL RECORD COPY AD:SA:DL Tippolito 8/ /82 USGPO: 1981~960

Docket No. 50-237 LSOS Mr. L. DelGeorge Director of Nuclear Licensing Commonwealth Edison Company Post Office Box 767 Chicago, Illinoi~ 60690

SUBJECT:

SEP TOPIC III-5.B, ~IP£ BREAK OUTSIDE CONTAINMENT DRESDEN NUCLEAR POWER STATION UNIT 2 By letter dated January 17, 1980, the* staff issued a draft safety evaluation on this topic. Your letters of March 21, 1980 and July 16, 1982, responded to open items from the draft evaluation.

Enclosed is the final ~afety evalu~tfon for Topic III-5.B. ihe ~taff con-eludes that Dresden Unf t 2 is adequately protected from the dynamic effects of pipe breaks outside contairnnent, subject to the resolution of the follow-ing in the Integrated Assessment.

The various,options (as listed in the enclosure) to resolve the *issue of the effects of pipe breaks outside containment in the containment penetration area for the main steam, isolatfpn condenser and cleanup piping.

The need to actua11Y implement chang~s as a result of this evaluation will be determi.ned dur~ri9 the Integrated pafety Assessment. This topic assessment majlbe revised in the future if your 'facility design is changed or ff NRC criteria relating to this topic are modified before the Integrated Assessment f s completed.

Enclosure:

as-stated Sincerely, Paul W. O'Connor' Project Manager Operating Reactors Branch No. 5 Division of Licensing AD:SA:DL Tippol ito 8/

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NRCFORM318(10-80)NRCM0240 OFFICIAL RECORD COPY USGP0:1981-335-960

'*1 Docket No. 50-237

• LSOS -..<..... -*--*--

Mr. L. OelGeorge Director of Nuclear Licensing Commonwealth Edison Company Post ~ff1ce Box 76*7 Chicago, Illinois* 60690

SUBJECT:

SEP TOPIC III-5.B, PIPE BREAK OUTSIDE CONTAINMENT DRESDEN NUCEEAR POWER STATION UNIT 2 By letter dated January 17, 1980, the staff issued a draft safety evaluation on this. topic. Your letters of March 21, 1980 and July J6, 1982, responMed-to open items from the draft evaluation.

EnClosed fs the fina.1 safety evalu~tion for Topic III-5.B. The staff con-cludes that Dresden Unit 2 fs adequately protected from the ~(ynam1c effects of pipe breaks outside containment, subject to th~ resolutfo*n of the follow-ing in the Integrated Assessment.

The various options (as. listed in the enclosure) to resolve the issue of the effects of *pipe breaks outside eontainment between the contain-ment and the first containment isolation valves-fao the main steam, isolat'ion condenser and cleanup ~fping.

The need to actually fmpl~ent*changes as a result of this evaluation will be detennf ned during the Integrated Safety Assessment. This topic assessment may be revised fn the future if your fac11ity design is changed or ff NRC criteria relating to this topfc* are modified before the Integrated Assessment 1s completed.

Enclosuret As stated cc w/enclosure:

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. Sincerely, Paul W. O'Connor, Project Manager Operating Reactors.Branch No. 5 Division of Licensing AD:SA: DL Tlppol ito 8/ /82 v

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NAC FOAM 318 (10-80) NACM 0240 OFFICIAL RECORD COPY USGPO: 1981-335-960

f Mr. L. De1George cc Robert G. Fitzgibbons Jr.

Isham, Lincoln & Beale Counselors at Law Three First National Plaza Suite 5200 Chicago, Illinois 60602 Mr. B. B. Stephenson Plant Superintendent Dresden Nuclear Power Station Rural Route #1 Morri~, Illinois 60450

• The Honorable Tom Corcoran United States House of Representatives Washington, D. c. 20515

• u *. s. Nuclear Regulatory Commission Resident Inspectors Office Dresden Station RR 11 Morris, Illinois. 60450 Mary Jo Murray Assistant Attorney General Environmental Control Division 188 W. Randolph Street Suite 2315 Chicago, Illinois 60601 Chainnan Board of Supervisors of

~Grundy County Grunay-tounty Courthouse Morris, Illinois 60450 John F. Wolf, Esquire 3409 Shepherd.Street Chevy Chas.e, Maryl and 20015 Dr. Linda W. Little 500 Hennitage Drive Raleigh, North Carolina 27612.

Judge Forrest J. Remick

.The Carriage House - Apartment 205 2201 L Street, N. W.

Washington, D. c. 20037

.Y

• Dresden 2 Docket No. 50-237 Revised 5/19/82 Illinois.Department of Nuclear Safety 1035 Outer Park Drive, 5th Floor Springfield, Illinois 62704.

U. S. Environmental Protection Agency Federal Activities Branch.

Region V Office ATTN:

Regional Radiation Representative 230 South Dearborn Street Chicago, Illinois 60604 James G. Keppler,* Regional Administrator Nuclear Regulatory Commission, Region III 799 Roosevelt Street Glen Ellyn, Illinois 60137

SYSTEMATIC EVALUATION OF PIPE BREAK OUTSIDE CONTAINMENT TOPIC III-5.B FOR DRESDEN NUCLEAR GENERATING STATION UNIT 2

TABLE OF CONTENTS I.

INTRODUCTION I I.

REVIEW CRITERIA I I I.

RELATED SAFETY TOPICS 1v~

REVIEW GUIDELINES

• • v..

EVALUATION VI.

CONCLUSION VII.

REFERENCES

SYSTEMATIC EVALUATION PROGRAM TOPIC IIl-5.B DRESDEN.NUCLEAR POWER STATION UNIT 2 TOPIC:.111-5.B, Pipe Break Outsid~ Containment I.

INTRODUCTION The safety objective of Systematic Evaluation Program (SEP) Topic 111-5.B, "Pipe Break Outside Containment, 11 is to assure that pipe breaks would not cause the loss of required function of "safety-related" structures, systems and components and to assure that the plant can be safely shutdown in the event of such breaks.

The required function of safety-related systems are those functions required to mitigate the ef-fects of the pipe break and safely shutdown the reactor plant.

II.

REVIEW CRITERIA General Design Criteria 4 (Appendix A to 10 CFR Part 50}, requires in part that structures, systems and components important to safety be ap-propriately protected against dynamic effects, such as pipe whip and discharging fluids, that may result from equipment failures.

III.

RELATED SAFETY TOPICS A.

This review complements that of SEP Topic VII-3, "Systems Required for Safe Shutdown."-

B.

The environmental effects of pressure, temperature, humidity and flooding due to post~lated pipe breaks are evaluated under Unresolved*

Safety Issue A-24, 11Qual ification of Class 1 E Safety-Related Equipment.

11 C.

The effects of potential missiles generated by fluid system ruptures and rotating machinery were also considered and are evaluated under SEP Topic III-4.C, 11 Internal~y Generated Missiles.

11 D.

The original plant design criteria in the areas of seismi~ input analysis design criteria are evaluated under SEP Topic III-6, "Seismic Design Considerations.

11 IV.

REVIEW GUIDELINES The current criteria for review of pipe breaks outside containment are continued fo Standard Review ~lan 3.6.1, 11Postulated Piping Failures in Fluid Systems Outside of Containment, 11 including its attached Branch Technical P.osition, Auxiliary System Branch 3-1 (BTP ASB 3-1) and Standard Review Plan 3.6.2, 11Determination of Break Locations and Dynamic Effects Ass6ciated with the Postulated Rupture of Piping, 11 in-cluding its attached Branch Technical Position; Mechanical Engineering Branch 3-1 (BTP MEB 3-1).

The staff used an 11effects ori.ented 11 approach to determine the accept~

ability of plant response to pipe breaks, i.e., each structure, system, component, and power supply which must function to mitigate the effects of the pipe break and to safely shutdown the plant was examined to de-tennine its sµscept.ibility to the effects of the postulated break.

Break effects considered were:

compartment pressur.ization, pipe whip, jet impingement, spray, and flooding.

V.

EVALUATION The staff issued a draft safety evaluation for this topic on January 17, 1980 (Reference 1). A request for additional information was issued on the same day in a separate letter (Reference 2).

The licensee responded to these letters in March 19PO (References 3, 4 and 5).

ln response to staff questions, the licensee provided additional informati'"on i'n Reference

6.

Each open item from the draft evaluation and staff requests along wi*th the licensee response and the'staff's resolution are presented below.

Staff Q~estion 1

1.

Provide a comparison of the design of the containment penetration piping outside containment between the containment and the first containme~t isolation valves for the main steam lines, tsolation con-denser steam and condensate lines, and reactor water cleanup inlet line with the provisions of Section B.l~b of Branch Technical Posi~ton MEB 3-1 (appended to Standard Review Plan 3.6."2} in sufficient detail to identify the degree of conformance with and deviations from these provisions.

Licensee Response The piping outside containment for the main steam lines, isolation condenser steam and condensate lines, and the reactor water cleanup inl~t line meet the provisions ASMt.Code Section III, Sub-article NE-1120.

All lines are designated as Quality Group A (Class 1) up to the outboard containment isolation.

The stress analysis performed on these piping lines was in accordance with the r~gulations in effect at the time the plant was constructed.

The stress limits, denoted for Class 1 piping in Section B.l.b(l).of Branch Technical Position MEB 3-1,.could not be compared to the original stress analysis since there was no ASME Se~tion III Class 1 Analysis performed on these lines.

The stress limits which most closely* represent what has been done on these lines are the limits described in Section B.l.b(l)(e). For the stress limits in this section, the lines are in conformance.

All Dther sections of B.l.b(l) could not be compared as indicated below:

A.

Section B.l:b(l)(~) required a maximum to the stress range based upon maximum allowable stress.

The maximum allowable stress limit as re-quired in the original analysis is different than Sm and no. comparison can be made.

The piping systems did meet the maximum allowable stress

. limit requirement and stress range of the original analysis.

B.

Section B. l ;b(l ).(b) coulc! not be compared for the same reason as in Item A.

C.

Section B. l.b(l)(c) discusses cumulative usage factors.

No cumulative usage factors were calculated in the original analysis.

D.

Section B. l.b(l){d) discusses allowable stress limits outside the containment.

These stresses could not be* compared for the same* reason as in Item A.

Staff Resolution As noted above, no stress data are available to demonstrate that these piping systems between the containment penetration arid the isolation valve

.outside containment meet the stress limit requirements of BTP MEB 3-1.

The following options are recommended for consid.eration in the Integrated Assessment:

A.

The licensee should provide information to demonstrate that the maximum stress in the portion of the pipe between the flued head and the out-board isolation valve, as calculated by Eg.(9) in paragraph NC-3652 of ASME Settion III code under th~ loadings resulting from a postulated piping failure of fluid system piping beyond the isolation valve does not exceed 1.8 Sh.

If the maximum stress exceeds the 1.8 Sh value, the licensee can either add supports to reduce the maximum stress or proceed with the option (B) below.

B.

The licensee should demonstrate by fracture mechanics analysis that a double ended pipe break could not occur beyond the flued head, or proceed with Option G:

C.

The 1 icensee should demoflstrate that the p.robabil ity of pipe breaks for these four lines beyond the containment penetration with a single active failure of the inside containment isolation valve and considering the radiological consequences of this event would not result in an unacceptably high risk.

Staff Question 2 Provide a comparison of the design of the containment penetration piping outside containment for the isolation condenser steam line and reactor water cleanup inlet li~e with the provi~ions of Section B.2.C of Branch Technical Position ASB 3-1 (appended to Standard Review Plan 3.6.1) in sufficient detail to identify the degree of confonnance with or the deviations from these provisions.

Licensee Response Section B.2.C of Branch Technical Position ASB 3-1 contains four subsections

~hich addres~ the requirement of the fluid system piping in containment

• penetration areas.

Each subsystems requirements are compared below...

1.

Subsection B.2.C (1) discusses the stres~ limits in the piping near the containment penetrations and the need for pipe whip restraints in the area of the containment isolation valves.

The lines were analyzed in accordance with codes in effect at the time of construction.

Hence, the stress*limits could not be compared (See Response 1).

The restraints in the area near the isolation valves were not designed with respect to pipe whip and operability of the isolation valves.

However, the physical routing of the lines and the routing of the power and control cabling for the isolation of the respective line is not ~ridangered by a break of that line.

2.

Subsection B.2.C (2) discusses piping that. penetrates both sides of a dual barrier containment structure. Because of the destgn of the primary and secondary containment of Dresden~_2, this subsection is not applicable to*these lines.

3.

Subsection B.2.C (3) discusses the tenninal ends considered for the extended piping runs.

Because of the design of the pipe lines this sub-section is not applicable to the Dresden-2 destgn.

The lines in question are high energy lines along their entire run. and hence, this question is not applicable.

4.

Subsection B.2.C (4}* discusses the piping classification of the lines...

• The isolation condenser steam line and the reactor water cleanup line*

are both considered to be Quality Group A lines up to and including the outboard containment isolation va~ve as required by Section 50.55a of*

10 CFR 50.

This classification was made without regard to pipe whip restraints. Because of the absence of the pipe whip restraints the break in classification of the piping was made after the outboard.con-tainment isolation valve.

Staff Resolution The draft evaluation (Reference 1) indicated that, although a detailed review had not been performed, a break in the cleanup line may be able to damage the cable tray with control and power for the outboard tsola-ti_on valve, or the valve itself. Similarly an isolation condenser steam llne or condensate line break may damage its respe$=tive outside containment isolation valve.

The licensee has stated that they have surveyed the targets in th.e break vicinity and their review shows thatthe power and control cabling would not be affected by the postulated breaks.

However, since pioe whip restraints are not provided,* the stress limits of ASB 3-l cannot be met so operability of the outboard valve following a break downstream of the valve cannot be assured.

With a~ assumed single failure of the ~nside coritainment isolation valve this situ~tion is then comparable to that'*

discussed above in Item 1, soi similar resolution applies.

Staff Question 3 Provide an evaluation of the consequences of a postulated main feed system high energy line break on the mezzanine floor of the turbine building.

Consider the postulated loss of offsite power and a concurrent single active failure-in accordance with the pro~isions of Section B.3 of Branch Technical Position ASB 3-1.

Licensee Respons~

In References 5 and 5, tne licensee discussed.the capability for r.eaching safe shutdown following a feedwater line break at the Feedwater Regulating Station (FRS).

The licensee e~aluated the effects of the feedwater line break on the CRD pumps, isolation cond~nser and ADS valves.

No CRD pump.cables are routed near* the Feedwater Regulatin_g Station.

However, a postulated break could aff~ct the 4KV ti~ between switchgear 23-1 and 23.

Therefore, CRD Pump 2A wouJd be unable to receive diesel power.

I It is possible to power CRD pump 20 for safe shutdown in the event of a

• feedwater line break in the feedwater regulation station (FRS) area.

However, the following operator actions would be required due to the loss of control cables in the FRS area:

f

1.

2.

3.

Manual start of the Unit 2 diesel, since autostart could be *affected by the postulated break.

Manual control of the 4KV feed breakers.

Manual transfer of Division II 125 VDC control power from the main to ~

reserve supply.

• Despite the above actions, a single failure of CRD Pump 2B would preclude the use of the CRD Pumps without a crosstie between the CRD pumps of Unit 2 and Unit 3.

This cr.osstie is proposed in the Fire Protection study on associated circuits.

Control cables to the Isolation Condenser isolation valves do pass through the FRS area.

A feedwater line break damaging all cables in the FRS area is considered to re~ult in a spunious signal closing the in-containment Isolatton Cdndenser valves and loss of control of those valves. Since the feedwater break affects only control cables to these valves, the operator could restore the valves to the open position by jumpi'ng contacts at the motor control center. The valves outside containment can be manually operated.

ADS cables also pass through the FRS area, but ADS ts not needed to bring the reactor to a safe shutdown condition. if the i:solation condenser 'is avai'lable.

The licensee further stated that with respect to the flooding consequences, the only safety-related equipment affected would be the Dresden-2 Diesel Generator located below the regulating stations. Measures have been taken

. to seal all possible drainage paths to the Diesel Generator Room.

Thus shorting of electrical equipment is not possible. Also, the water would not.

affect the diesel generator and the auxiliaries once it reach.ed *the 517' ele-vation afte~ draining through the stairs.

The other consideration is control room habitability.

No path extsts wh.ere the water could drain to the control room level of 534 1

• from tfte 538 1 ele-vation.

The stairs connecting the floors would allow th~ water to dratn to the 517 1 elevation. This i's because the stai'rs are made of grating material.

The positiv.e'pressure in the control room would prevent the noncondensable radioactive gases from entering the control room.

Also, the activity present i'n the feedwater line would be low thus k.eeping control room doses down.

Staff Resolution Th~*staff concern ~as that a postulated break at ihts locati~n could damage redundant Division I and Di'visi'on Il cable trays and thus dtsrupt power supplies to both trains.of emergency core cool tng.

lf tile i'solation con.:..

denser iS available, the ECCS would not be needed for ini'ttgation of the feedwater break or for reaching safe shutdown.

However, ~ith an assumed single failure of an isolation condenser valve, some ECCS would be requi-red.

For isolation condenser operation, only one isolation valve (outside containment) must be opened since the other valves are normally open.

As discussed above, operator action to control these valves is pos-sible. Therefore, the isolation condenser would probably be available.

Another alternative wo~ld be the turbine driven high pressure coolant injection system (HPCI) which would use reactor steam as motive power and provide mail<eup water. All necessary valve manipulations can be*

performed manually.

Therefore, the staff concludes that the effects of a feedwater pipe break at the PRS would not prevent safe shutdown.

Staff Question 4 Provide additional details of the corrective actions implemented to prevent flooding of the Unit 2 diesel generator control cabinet as.descri'bed tn Reportable Occurrence Report 79-52 of October 17; 1979.

Dtscuss the*

capabrt l ity of the corrective measures (RTV seal ant}_ to prevent disab 1 fog the diesel generator fol lowing a postulated mafo feed system pi'pe brea.k. on the mezzanine floor.

Licensee Response Reportable Occurrence Report 79-52 for Dresden-2 was submitted on October 17, 1979.

The D/G was declared inoperable due to a*short in the control cabinet caused by water which had permeated through. an HVAC duct due to a

. spill on the mezzanine floor.

In detail this ts the scenarto of the accident:

1.

While installing the sprinkler system for the cable. pans i'n the mezzanine area*; water leaked out of the end of the spri"nkler li'ne onto the floor.*

2.

The drains in the area were sealed due to a modi'ficati'on of changfog the discharge of the drains from the river to radwaste.

Consequently, the water collected on the floor.

3.

The water drained through the HVAC duct into the diesel generato~ room where some of the water seeped into the diesel generator control cabinet.

This caused the short in the circuitry and an* auto start block alarm.

4.

The operating engineer declared the diesel inoperattv~ and the requited surveillances were performed.

5.

The control cabinet was dried and the diesel was demonstrated to be*

operable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

\\

0

• In order to prevent this occurrence again, the area around the HVAC duct was sealed with GE-1200 silicone rubber sealant. The sealant was also.

used on the exhaust and air intake piping.

This was to prevent any leakage on the floor from draining into the diesel generator room.

~n addHion, the drain pipe rerouting to radwaste has been completed.

Also, as part of the fire protection modifications the HVAC duct has received a fire damper and an 18 11 curb around the damper.

The damper is provided with spray protection to keep the water from entering the curb area.

Therefore, neither spraying nor flooding will po'se a problem.to the diesel generator control cabinet.

Staff Resolution This question is related to the previous one wtth respect to prevention

• of water damage of the Unit 2 di~sel generator.

The modtfi~attons described above have been implemented by the Hcensee and provide reasonable assurance that spraying and flooding from a matn feed ltne break will not prevent operation of the Unit 2 diesel generator.

Staff Question 5 Provide an evaluatton of the consequences of flooding both 4 KV swttchgear 23-1 and 24-1 to tlte depth of the curb surrounding the switchgear panels from a postulated reactor buildi'ng closed cooling water system leak (spray) on the 545~ elevation of the reactor building.

Licensee Response A postulated leakage crack in the RBCCW piping on elevation 545 1-6 11 would result in a leakage rate of approximately 623 gpm.

Both 4 KV switchgears could be ~isabled if flood~d to the curb surrounding them.

Some of the breaker auxiliary components would be submerged and subject to failure..

Floor drains at this elevation can collect at least 300 gpm.

The rell!ainder would flow through two pipe penetrati'on sleeves to the 517' eleva.tion below.

The first is an 8 11 sleeve with a 6 11 (wi*th insulation) heating steam pipe in the penetration, and the other is a 12 11 pipe in'a 20 11 sleeve.

The 8 11 penetra-tion has no lip on the sleeve so drainage can begin as soon as water flows to the penetration.

The 20 11 penetration has al 1/2 11 lip on a sleeve,.-*but thts will be notched to penlitt drainage. Therefore, any, level of water in the curb area would quickly drai"n and prevent floodi'ng.

Nearly all the water reaching elevation 517 1-6 11 will flow into floor drains on that elevation. A negligible amount of water will flow through floor openings on elevation 517 1-6 11 to elevation 476 1-6 11 No accumulation of water is expected on any elevation.

MCC's 28-1, 28-7, 29-1, 29-4, and 29-7 are located on elevation 517'-6".

These MCC's are mounted on low pads but are not surrounded by curbing.

However, since no water will accumulate on elevation 517'-6", the only possible hazard to an MCC ts water falltng directly onto the MCC from a floor opening above.

Only MCC 28-1 is under floor openings that could al)ow water to fall on it. Loss of MCC 28-1 would disable operation of the Isolation Condenser valves and disable safety systems powered by Division I. However, Division II would be unaffected and safe shutdown could be reached wtth the Division II HPCI and LPCI Systems and the ADS.

Staff Resolution Based on the above discussion, the staff concludes that the effects of a 1 eak in the RBCCW would not prevent reachi-ng a safe shutdown conditi*on-.

VI.

CONCLUSION Based on the pre~ious staff reviews and the above discussion~ we conclude that the Dresden Unit 2 facility is adequately protected from the effects of pipe break outside containment subject to resolution of the following:

For main steam, isolation condenser and cleanup pi~ing, provide resolution for staff concern that postulated pipe breaks in the contafnment penetration area with an assumed single failure of the inboard isolation valve would result in an untsolable pipe break outside of containment (Le., a loss of coolant accident} *

.. ) '

• VII.

REFERENCES

1.

Letter, D. Ziemann (NRC) to D.L. Peoples (CECo), transmitting Draft Evaluation of Topic 111-5. B, dated January* 17, _1980.

2.

Letter, D. Zi"eniann (NRC) to D.L. Peoples* (CECo), transmitting' Request for Additional Information, dated January 17, 1980.

3. Letter, R. Janecek (CECo) to D. Ziemann (NRC),

dated March 10, 1980.

4.

Letter, R. Janecek (CECo} to P.W. O'Connor (NRC)*,

dated March 10, 1980.

5.

Letter, R. Janecek (CECo) to D. Ziemann (NRC),

dated March 21, 1980.

6.

Lette*r, T.J. Rausch (CECo) to P.W. O'Connor. (NRC),

dated July 16, 1982.

... '. :~

• .. ***