ML20113H712

Discusses Potential Condition Adverse to Quality Rept Initiated by Util to Document Potential Environ Qualification Problems Resulting from Errors Identified in Calculations of Temp Profiles Following Postulated HELB

ML20113H712
Person / Time
Site:	Davis Besse
Issue date:	07/28/1992
From:	Shelton D CENTERIOR ENERGY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
2056, TAC-M84152, NUDOCS 9208050199
Download: ML20113H712 (19)

Text

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Donekt C. SheMon 3CC Madison Avenue Vre PrcWent Nuclear Toledo OH 43652A101 Dads-Besse (419)249 2300 Docket Number 50-346

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License Number NPF-3 Serial Number 2056 July 28,.1992 i

United States Nuclear' Regulatory Commission Document Control Desk-j Vashington, DC 20555

Subject:

High Energy Line Break (HELB) Analyses Modeling Errors Cen tlemen t -

On April 27,,1992,' a Potential Condition, Mvcrse to Quality Report (PCA0R) was_ initiated by Tcledo Edison ("r.) to document potential environmental qualification problems resulting from errors identified in calculations of temperature profiles following postulated HELDs at the-Davis-Besse Nuclear. Power Station (DBNPS).

Specifically, the analyses' performed to predict environmentel conditions following c 11ELB outside containment associated with lines containing cuperheated steam, used non-conservative. techniques regarding heat transfer coefficients.

Further detail on the apparent cause of the errors-is'provided in Licensee Event Report-(LER)92-004..which uns initially submitted on May 8, 1992, and-revised _on June 18. __1992 - by_ TE.

.The HELB analyses affected by the_modeling. error. include calculations l

for breaks in the following high energy lines located in the Auxiliary Building (1) Main-Steam to Auxiliary Feedvkter Pump Turbines (AFPTs);

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-(2) Hain Feedvater; (3) Steam Generator Blovdown; and-(4) Auxiliary.

Steam Supply. LA_ Justification for Continued Operation (JCO) was completed by TE on-May 1, 1992. The JC0_ concluded that, following postulated HELBs, it is reasonable:to sssume that-equipment required to

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mitigate the consequences of.~the break vill perform the.ir intended-safety. functions _and continued operation of the plant-is. justified.

The-JC0 was revised on-July 17s 1992, to incorporate the results of an independent review and.Other:ongoint_ activities. 'A copy of-the revised

-JC0.is' attached.

As noted in the JCO, as an interim measure, credit'is taken for fire protect',on sprinklers to mitigate the_ effects of a_ Main Steam to AFPT Idesk in Room 124. Additional administrative controls were established

-v f a a Standing Order to ensure the operaoility/: functionality.of the fire protection system. The Standing Order,-issued on May 4, 1992, requires-the plant to be placed in! Mode 4 (Hot Shutdown) if the sprinklers in Room 124 are not functional for greator than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

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'Dockot Number 50-346 License Number NPF-3 7~

Serial Number 2056 Page 2 The Standing Order also requires operators, during rounds, to inspect suoms containing portions of the AFP Turbine Steam Supply Lines (Rooms 124, 235, 236, 237, 238, 304, 314, 404, 427, and 501) for indications of a leak-before-break of main steam piping to the AFPTs, and to also verify that Door 212 (at the entrance to Room 235) is closed.

Following identification of the calculational error, a multi-disciplined task force was appointed to evaluate long term permanent solutions to this problem.

The recommendations from this task force. required,a significant effort to evaluate the leasibility and acceptability of the long term. solution. These efforts, which aimed at restoring environmental qualification of certain equipment and at elimination of postulated breaks, are nov complete.- In order to resolve identified concerns, TE is planning to perform the following j

modifications to the plant r

1..

Replace Safety' Features Actuation System containmens.ressure transmitters PT2001 and PT2002 vith transmitters qualified to the L

newly developed temperature profile.

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-2.- Remove anchor A001 on the Hasn Steam piping in Room 124 to l

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eliminate a postulated high energy line break in Room 124, and mraffy pipe supports os necessary to climinate postulated critical et 'ka in selected areas.

This vill support the environmental t

x quanfication.of epulpment required to mitigate the effects of a

.Hsin_ Steam to APPT break.

These modifications vill be completed by the end of the eighth refueling outage. The eighth refueling outage is tentatively scheduled to commence March 1, 1993.

Following completion of these activities, the~ compensatory measures identified in the-JC0 vill no longer be necessary for power-operation.--

Toledo' Edison vill enntinue to keep the NRC' Senior Resident Inspector Informed of progress on this issue.

In addit' ion, Toledo Edison would l

like to meet with the NRC Staff to present the resolution of this issue, and requests that the NRC Staff advise Toledo Edison when such a meeting could be held.

If'you have any questions, please contact Mr. R. V. Schraudr-Hanager - Nuclear Licensing at (419) 249-2366.

Very truly.yours,-

-HKL/dic cci LA. B. Davis, Regional Administrator, NRC Region ITI J. B. Hophins, NRC/NRR DB-1 Senior Project Manager V. Levis, NRC Region III, DB-1. Senior Resident Inspector Utility Radiological Safety Board

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, Docket Number 50-346 Lic'.ase Number NPT-3 Serial Number 2056 Attachment page 1 7/17/92 Rev. 1 JUSTIFICATION FOR CONTINUE 0 OPERATION PCAOR 92-0195

1. Introdottion 10CFR50.49 tequires that time dependent compartment temperatures and pressures be calculated for equipment which la required to remain functional following postulated high energy line breaks (llELBs).

Environmental qualification of equipment for HELBs is based on these compartment temperatures.

It has been learned that the analysis performed to predict environmental conditions (i.e., pressure and tempetature) following IIELB outside containment associated with lines containing superheated steam, used non-conservative techniques vith regard to heat transfer coefficients. Recent calculations, with a different computer program and revised heat transfer coefficients, ptedict significantly different room temperatures.

PCA0R 92-0195 vas initiated to document this discrepancy.

The HELB analyses affected by the modeling error include analyses for breaks in the following high energy lines located in the Auxiliary Building:

1.

Main Steam to Auxiliary Feedvater Pump Turbines (AFPTs) 2.

Main Feedvater (MFV) 3.

Steam Generator Blovdown 4.

Auxilicr/ Steam Supply Duc to the coinpartmentalization of the Auxiliary Building, a postulated hicak of these lines could affect the environment of various rooms in the Auxiliary Building.

The ditfetences in predicted environmental conditions are significant enough to question the original qualification of affected equipment located in rooms affected by the postulated breaks.

This justification for continued operation has been prepared to demonstrate reasonable assurance that the equipment required to mitigate the consequences of a postulated high energy line break vill perform its intended safety Iunetion when called upon.

II. Systems Affected Equipment located in various rooms of the Auxiliary Building affected by postulated breaks of the above mentioned high energy lines are affected.

Specific equipment is discussed in the effects or safety section below.

III. Safety Function of Affected Systems The cystems and equipment located in the Auxiliary Building provides a variety of functicns including support of t'ormal power operation and support of safe plant shutdova during design basis events.

Affected rooms and equipment required to raitigate the consequences of the break are discussed in the effects on safety sectien below.

,' Docket Number 50-346 Licente Number !JPF-3

' Serial Number 2056 Attachment Page 2 IV. Effects on Safety The llELB criteria given in USAR Section 3.6.2.1 requires that a double-ended rupture be postulated.n piping operating above 275 psig and 200'F.

The pipe bre-ks curtently analyzed in the USAR are postulated in l

accordance with NRC Branch Technical position MEB-3-1 at the terminal ends and at the intermediate locations where the pipe stre s exceeds 80%

of the ASME code allovables.

If the pipe stress did not exceed 80% of the ASME Code allovables, breaks are postulated at two arbitrary intermediate locations where the stresses are high.

In Generic Letter 87-11, the NRC provided a relaxation for arbitrary intermediate pipe rupture requirements. The Generic Letter specifically eliminates the requirement for postulation of arbitrary intermediate pipe breaks for pressure, temperature, bumidity and flooding considerations.

The impact on safety due to the modeling error is described below.

See Appendix 1 for additional information regarding the Fire Sprinkler System.

See Appendix 2 for additional information regarding PRA.

1.

Main Steam to AFPTs The two 6" steam supply lines to the AFPTs originate in the main steam penetration rooms on elevation 643' and terminate at the AFPT in the AFP rooms on elevation 56b' (See.igure 1).

These lines traverse a total of 15 rooms in the Auxiliary Building.

The af fected roo'ns are listed in Table 1.

Revised postulated l

break locations are also summarized in Table 1, taking credit for the relaxation for arbitrary intermediate pipe rupture requirements allowed by Generic Letter 87-11.

The Main Steam piping to the AFPTs is provided with the

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following instrumentation to detect breaks and mitigate the effects (Refer to the simplified diagram of the Auxiliary Feedvater System in Figure 2):

PSL-106A thru D und PSL-107A thru D These pressure switches, located in rooms 237 and 238 on the AFPT steam inlet piping just upstream of the steam admission valves (MS5889A and MS58P9B), are designed to detect breaks in the portion of pre urized 6" AFPT main steam piping in the Auxiliary Building that is downstream of the chek valves MS734 and MS735 in the crossover piping and check valves MS726 and MS727 in the AFPT Main Steam piping.

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Docket Number 50-346 License Number NPF-3 Serial Humber 2056 Attachment Page 3 A double-ended rupture in this portion of the 6" main steam piping would depressurire the piping up to the AFPT steam admission valves.

Upon detection of a lov pressure condit.on the pressure switch interlock vill activate an i

alarm in the control room and isolate the appropriate steam line valves MS106 and MS106A or MS107 and MS107A. The pressure svitch interlock is designed such that two parallel trip schemes exist for each AFV train.

During normal operation only MS-106A and MS-107A are open.

The closute of the isolation valves via this pressure svitch interlock terminates the steam blovdoun in the Auxiliary Building and mitigates the consequences of the HELB.

pSLS894A and B and PSLS895A and B

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These pressure switches installed on the main steam piping to the AFPTs in rooms 500 and 501, upstream of MS106A and HS107A respectively. These pressure switchen are designed to detect breaks upstteam (f MS734 and MS735.

Upon detection of a lov pressure condition, they annunciate an alatm in the control room to alert the operator to take appropriate actions.

The analyses assume that the pressure switches vill detect the breaks and actuate an alarm in the control room to alert the operator.

The analyses assume a limiting single failure of the open main steam isolation valve (MS106A or HS107A) to close.

The Auxiliary Building environmental conditions vere d(termined using the assumption that within ten minutes of detection of the htcak the control room operator takes necessary aanual actions to terminate the steam blovdovn to the Auxiliary Building.

Impact of postulated breaks in various rooms is as follows:

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Rooms 237 and "T8_

These rooms are separated such that a break in either room does not affect the other AFV train or any other equipment required for safe shutdown. The rooms are ditectly vented to the turbine building.

Since a break occurring in either of these rooms does not co:amunicate with the other (adjacent) room, the equipn.ent qualification in the adjacent AFp room is not affected.

A break in one of th?se rooms affects the same train of AFV, but since l

that train of AFV is already disabled due to the break, there is no additional impact.

A few cables needed to mitigate the break are located in each room.

However, these cables are capable of vithstanding projected revised temperatures for these rooms.

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Docket IJumber 50-346 License 14 umber 11PP-3 Serial Number 2056 Attachment Page 4 Room 124 Room 124 does not contain any safety-related equipment.

The qualification information with the safety-related cables contained in this room indicates that these cables are capable of performing their safety function.

Room 124 communicates with room 235.

A break in either toom vill be detected by the above mentioned pressure switches located in toom 237 and vill close MS106A, te -eby terminating the blevdown.

Ilovever, in the event of a single failure of HS106A to close, the steam blovdawn vill continue. Room 235 also does not contain any safety-related equipment but communicates to room 236 through a blowout panel.

-Room 236 does contain safety-related equipment necessary to mitigate the !! ELD and ensure the safe shutdown of the plant.

Room 236 also communicates with the rest of the building which contains safety-related equipment required to mitigate this break.

The blovout panel between rooms 235 and 236 is provided i

vith a vatet curtain. The water curtain is part of the fire suppression system and vill be actuated (10 minute time delay) i by the steam environment.

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Room 124 contains fire protection sprinklers and a water curtain between room 124 and 235 which actuate at 212'F.

The actuation of ' sprinklers (360 gpm) nd water curtain provide $25 gpm of total fire suppr'ssion discharge into room 124. The 360 gpm I

flow rate is ex,;ccted to provide adequate cooling to the steam leaving the room, such that the temperature of room 236 vill be bounded by the existing analyses.

l Room 227 contains equipment required to mitigate the consequences of this break, however, room 227 is isolated from the break by a closed door. The differential pressure required to allov significant steam.through the door opening exceeds the differential pressure required for the blovout panel to room 236.

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Room 401 Room 401 communicates with rooms 300, 400, 404, and 405.

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300 communicates to the atmosphere. There is no safety-related

. equipment required to mitigate the consequences of this break located in rooms 300, 400, 401, or 404. Room 405 is connected by small ventilation openings, but is a " dead end" compartment and vill undergo a negligible temperature excursion, j

Room 500, 501-Root 315 communicates directly vith rooms 500 tnd 501.

Breaks in rooms 560 and 301 vere originally modeled to directly affeet

- equipment rooms 500, 501, and 515, and to affect other areas in the at lary building through a ventilation duct flow path in-room 31-.. There is no opening in this duct vork in the affected rooms.

Ilovever, further investigation has revealed ll that a 3 x 10 foot door (#518), which opr.ns from room 515 into the turbine building, vill fail at approximately 1 psid thereby l'

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Docket Number 50-346 License Number NpF-3

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I providing a communication path to the turbine building.

Failure of this door should prevent full collapse of the ducting leading to room 314 and vill eliminate or reduce the quantity of steam which would enter in other areas of the auxiliary building.

Any residual steam which enters room 314 through potentially distorted ducting vould be easily accommodated by condensation, allowing conditions to remain within existing environmental qualifications for the room. Vhile steam venting into the turbine building through door #518 vill prevent impacting other auxiliary building compartment temperatures, temperatures within rooms 500, 501, and $15 vill not be reduced.

Breaks must be considered in room 500 upstream and dovnstream of the normally open cross connect isolation valve MS107A.

Bteaks must also be considered in room 501 upstream of HS106A.

For a break downstream of HS107-, MS107A vould normally be isolated either automatically by the above mentioned redundant pressure switches, or by procedural operator action based on teceipt of a lov steam line pressure annunciator in the control room. These actions rhould occur prior to the time when the projected elevation in compartment temperature could cause loss of the ability of the valves to respond. However, assuming a failure of MS107A to close, (as the assumed single failure), the l

operator vould continue executing the steps prescribed in DB-0P-02525.

These steps involve tripping the reactor, initiating auxiliary feedvater flov, isolation of both steam generators by pressing the SFRCS manual actuation switches, closing AF608 to isolate auxiliary feedvater to steam generator

1. 1, and then blowing down steam generator #1 using an atmospheric vcat valve to terminate the leak.

Following the manual SFPCS trip, auxiliary feedvater pump 2 vill be aligned to steam generator #2 (steam and feed sides). Auxiliary feedvater

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pump 1 vill receive steam from steam generator #2, but vill not necessarily feed either steam generator.

Breaks upstream of the motor operated isolation valves are l

considered highly unlikely based on analyzed piping stress levels.

F,ven if this occurs since the analyses assumed single failure of MS107A to close, the analysed are not affected.

Additionally the motor driven feedvater pump would remain capable of supplying either steam generator vi h water, for single failures postulated in the auxiliary feedvater system.

While steam generator #1 is bloving down into rooms 500, 501, and 515, the compartment temperatures vill exceed the original equipment qualification temperatures for some of the equipment.

The calculations for this break with continued blovdown reveal l

that the temperature ir, the room rises to 350'F in approximately 100 seconds and increases to a maximum 370'F over the next 500 seconds. Temperature decreases after the reactor is tripped and decreases further after 1200 seconds.

i Docket Number 50-346 i

• , License ilumber NPF-3 i

' Serial Number 2056 Attachment Page 6 In the analysis it is conservatively assumed that-operator action vill commence 10 minutes after the break.

Since the operators are trained on (his scenario and receive several indications of the break occurrence (annunciator alarms from PSL5895A,B and alarms from PSL107-D, fire detection system alarms, etc.), 10 minute operator action is extremely conservative.

Additionally these analyses did not take credit for any fire protection sprinklers in the room 501.

Although, for the analysis purposes this area is divided into three rooms,

- in reality it is one long volume.

The fire protection sprinklers in room 501 voeld reduce the average temperature in l

these rooms'to much lover than 370*F stated above.

The following equipment in these rooms is required to function to mitigate the consequences of this break, assuming single failure of the MDFP and non-isolable break upstream of HS106, MS106A, MS107 and MS107A.

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l Steam _ inlet valves HS106, HS106A, MS107, and HS107A including, limitorque operator, pover and control cable, and terminal blocks.

Hain steam line pressure ivitches that provide input to SFRCS including terminal blocks and associated cables.

Pressure switches on the main steam to auxiliary feedvater pump turbines which provide control room annunciation.

Local control stationa for the steam inlet valves.

The available test data-for the Environmental Qualifications of l.

this equipment is as follovst-Limitorque test report B-0027 "Limitorque Valve Actuator Temperature Related to liigh Superheat Ambient Temperature" d

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concludes that the valve actuator could he subjected to i :

ambient temt.aratures of up to 492*F in excess of 17 minutes vithout exceeding 315'F for electrical components and motor vindings. Applying this methodology to the limitorque l

operators associated with the steam inlet valves, it is concluded that i f the valve operator is subjected to 370'F the electrical component and motor vinding temperature vill i

not' exceed the qualification temperature of 250'F in 25 ll minutes.

Valve control circuit pressure switch terminal i

blocks-tested to 350*F for one-half-hour after an initial spike to 380'F.

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Docket !Jumber 50-346 License 11 umber NPF-3 Serial thimber 2056 Attachment Page 7 Pressure svitches for SFRCS inputs and control room alarm-tested to 430'F for 4 minutes followed by 370*F for 10 minutes followed by 350'F for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> followed by increase to 430*F for a short period.

Valve power terminal block for H5106 (DC valve)-tested to 345'F for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Valve local controls stations-ramped to 318'F in 66 minutes fol?oved by a decrease to 175'F over 3.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 2ollowed by a hold at 175'F for 4.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Based on evaluation of this data it is technically judged that all of the above equipment and associated cables (pover and control), with the possible exception of valve local control stations, vill function with the temperature profile predicted by the pteliminary calculations. The valve local control stations vere electrically l

disconnected from the control circuit by temporary modifications to the facility. The automatic operation of these valves from the control room vill not be affected by this change.

The only other equipment of imru tance for this transient involves the SFAS containment pressure transmitters.

While these transmitters ate not required to mitigate this transient, spurious actuation could complicate the event by initiating SFAS Incident Levels 1 through 4 which vould result in containment spray actuation and containment isolation.

The SFAS containment pressute transmitters vere insulated as a precautionary measure.

Additional testing performed on local control switches and indicating light assemblies aboved that these components are functional when subjected to 375F to 400'F for approximately one half hour.

Information provided by the vendor of the SFAS pressure transmitters indicates that the transmitters vill not cause an SFAS Level 4 initiation due to zero and span shifts associated with high temperatures.

Rooms 600 and 601 These rooms are mainsteam line areas and are vented directly to atmosphere via blovout panels in the roof.

In addition pressure retaining doors are provided between each steam line room and test of the Auxiliary Building.

Breaks in these rooms are bounded by a 36 inch main steam line break.

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' Docket. Number 50-346 License Number NPF-3 Serial Number 2056 Attachment i

Page 8 2.

Hall _ Feedvater Line Breaks The two 18" main feedvater lines in the Auxiliary Building enter from the Turbine Building on elevation 585' and end at the containment penetrations on that elevation.

The line to SG 1 enters the Auxiliary Building in Room 313, turns upward to elevation 603' into room 404, traverses-Room 404, then turns downward into room 303 on elevation 585', where the containment penetration is located.

The line to SG 2 is located wholly in Room 314.

The affected rooms are listed in Table 2.

Postulated break locations are also summarized in Table 2.

The impact of breaks in these rooms is as follovst

- Rooms 303/314 Rooms'303/314 are affected by breaks of the HFV line.

Since the i

HFV lines.contain liquid at approximately 455'F, the break does not evolve significant superheated steam.

Because of the very large mass flov rate of saturated mixture, the effect of overestimating condensing heat transfer is not as significant for.these liner.

However, the eventual blovdown of the steam space in the steam generators does introduce some superheated steam into the rooms.

In the previous analyses the superheated blevdown only occurs for approximately 75 seconds-and raises the temperature from 215'F to 270*F.

Additional analyses using the corrected model has indicated that the peak should be approximately 330'F.

The maximum environmental qualification l

teiperature for the room is 300'F for greater than 10 minutes.

i llovever, because the room temperature returns to approx!mately i

212"F vithin two minutes, it is reasonable to conclude that the equipment remains qualified.

In addition these rooms contain fire protection sprinklers which actuate at 212'F.

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other' Rooms i

Ne significant temperature increase vas noted in other. rooms l

affected by steam flooding.

3.

steam _Cc3 n t_or Blovdown Line Breaks

. The two 4" steam gen 9 tac Movdown lines in the Auxiliary Building' originate at tbh.c ntainment penetrations in Room 236 L

on elevation 5658 for both' lines. The lines then turn up-into l

Room ?l4_on elevation 585' then penetrate the east vall of that room into the Turbine Building. The affected rooms are listed in Table 3.--

Fostulated break locations are also summarized in Table 3.

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Docket Number 50-346 License 11 umber 11PP-3 Serial Number 20$6 j

Attachment Page 9 The Steam Generator Blovdown Line rupture analysis was run for 30 minutes at a flow rate of 200 lbm/sec vith an enthalpy of $60 BTU /lbm.

Since the enthalpy can not produce superheated steam, the results of the room temperature analysis vill not be affected by the error in the original analysis.

During the timt period following manual action to terminate the line break, enthalpy m.sy exceed 1200 BTU /lbm, but only when the flow is less l

than 5 lbm/sec.

Kealistic heat transfer rates are sufficient to prevent significant increase in room temperature results.

l 4.

Auxiliary Steam Supply These are low pressure, high temperature moderate energy lines.

Based on criteria described in USAR Section 3.6, critical cracks must be postulated for these lines.

These lines contain supetheated steam, however, previously analysed non seismic

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portions of these lines have been isolated. Therefore no area of the auxiliary Building vill be affected by breaks in this line.

Further, the room 124 where non-seismic auxiliary steam line is located contains fire protection sprinklers.

V. Summary g

As an interim measure, credit is taken for fire protection sprinklers to mitigate the effects of a Main Steam to AFPT break in room 124.

Additional administrative controls vill be promptly established to ensure the operability / functionality of the fire protection system.

The plant vill be placed in Mode 4 (llot Shutdown) if the sprinklers are inoperable for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

No credit is taken for fire sprinklers to mitigate the effects of a Main Steam to AFPT break in rooms 500 or 501. Vith the exception of steam inlet valve local control stations and SFAS containment pressure transmitters, equipment qualification analyses demonstrates that the affected equipment remains capable of performing its function. The steam inlet valve local control stations were electrica)1y disconnected from the control circuit. The SFAS containment pressure transmitters were insulated as a precautionary measure to assure that the environmental qualification is maintain '

The impact of Main Steam to AFPT breaks in other r,.sms has been evaluated and the consequences remain acceptable.

The impact of a Main Feedvater Line b'cak or a Steam Generator Blowdovn Line break in the Auxiliary milding has been evaluated and the consequences remain acceptable.

Significant non-seismic

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Class I Auxiliary Steam Supply lines to the Auxiliary Building have been isolated previously, therefore no area of the Auxiliary Building vill be affect?d by breaks in these lines.

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l Docket 11 umber 50-346 License 11 umber NPF-3

' Serdal 11 umber 2056 Attachment Page 10 VI. Conclusion Based on the above, it is concluded that following postulated breaks, it is reasonable to assume that the equipment required to mitigate the consequences of the break vill perform its intended safety function and continued operation of the plant is justified.

Toledo Edison is continuing to evaluate long term resolution of this issue and vill provide the NRC vith interim reports of progress.

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Docket Number 50-346 License Number NPF-3 Serial Number 2056 Attachment Page 11 TABLE 1 MAIN STEAM TO AFPT LINES -- AFFECTED ROOMS ROOH NUMBER DESCRIPTIOJJ NOTES (1) 124 Clean Vaste Receiver Tank Room 1_1 B

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227 Passage S

F 2.35 Boric Acid Evap Room 1-1 (2) 236 No. 2 Hechanical Penetration Room S

F 237 Aux Feed Pump Unit Room 1 1 B

238 Aux Feed Pump Unit Room 1-2 B

300 Fuel Handling Area 304 Corridor to Hech. Pent Room 364 S

F 314 No. 14 Hechanical Penetration Room (2) S F

400 Equipment Hatch Aree Passage 401 Fuel Handling Exhaust Unit Room B

404 Spent Fuel Pool Corridor (2) 427 No. 2 Electrical Penetration Room S

F 500 Radvaste & Fuel Handling & Air Supply B

S Equipment 501 Radvaste Exhaust Equip & Hain Station B

S F

Exhaust Fan Room 515 Purge Exhaust Equipment Room 601 No. 1 Main Steam Line Area B

S 602 No. 2 Main Steam Line Area B

S (1) B. Break Postulated S. Contains safety-Related Equipmer Required to Mitigate Break F. Contains Fire Sprinklers (2) Dreak not postulated due to relaxation allowed by Generic Letter 87-11

• Docket Number ~a0-346 License Nurnber NPF-3

' recial Number 2056 Attachment Page 12 TABLE 2

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MAIN FEF.C.ATER LINES -- AFFECTED ROOMS 6

Roua Number Description Notes.(l' 303 No. 3 Hechanical Penetration Room J

F 313 Mix Tanks & Hatch Area No. 4 Hechanical Penetration Room B

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-(1) B. B eak Postulated F-Contains Fira Sprinklers A

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'.Page 13-TA3LB 3 STEAM GENERATOR BLOVDOVN LINES -_ AFFECTED ROOMS r

Room Number.

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,'_ Docket Number 50-346

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Serial Number 2056-Attachment Page 16 APTENDIX 1 Fire Sprinkler System --Supplementary Information For systems located in rooms exposed to elevated temperatures caused by a HELB.vhere the maximum expected room temperature exceeds th:: temperature rating of the sprink1ces, all of the sprinklers-in the room vill be assumed to actuate. The rooms exposed to temperatures sufficient to actuate all sprinklers and included an the analysis of breaks for environmental qualification include: 124, 236, 314, and 501.

The sprinkler system design flow rate for room 124 vater discharge with all sprinklers actuating due to elevated temperatures caused by a HELB is approximately 525 gpm. There are no safety-related equipment in this room and therefore safe shutdovn is not affected.

The sprinkler system design flovrate for room 236 vater discharge with all sprinklers and the vater curtain actuating due to the elevated temperatures caused by a HELB_is approximately 535 GPM.

Drainage vould be through the floor drainage system te the miscellaneous vaste drain

-tank and through a pipe chase into room 115. The additional vater discharge due to-sprinkler system actuation vill not adversely effect any safety related equipment required to mitig:ite the llELB.

Therefore, actuation of the sprinHer system due to the HELB vill not affect safe shutdown.

The sprinkler system design flovrate for-room 314 vater diccharge with all sprinklers and the vater-curtain actuating due to the elevated temperatures cau' sed by a HELB-is approximately 2200 GPH.

Drainage paths vould include auxiliary building rooms 115, 236, 300, 303, 304, 3D, and 313 and through blowout' panels into room 236. The additional vater discharge due to sprinkler system actuation vill not adversely-effect any

-additional safety related equipment required.to' mitigate the HELB.

Therefore, actuation of the sprinkler system.due to the HELB vill not affeet' safe shutdown.

-The sprinkler tystem design flovrate for room 501 vater discharge with

'all sprinklers actuating due'to the elevated temperatures caused by a HELB isfapproximately 1650 GPM.-

Drainage vould be out'_two doors leading i.

to the heater bay area and-through 13 floor drains located throughout the 623' elevation to the miscellaneous vaste drain tank and to a sump in L

room 105.

Localized vater accumulation may occur as the water spreads to E

,the drains:in room-500 and 515, howevet, due to the openings leading.to the' heater bay area, resultant flooding would be negligible.

Safety related equipment in room 105 may be adversely a:fected by water flood ig, however, the required sate shutdown eqt Ipment vould not be l;

affected.

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, Docket _ Number-50-346-7..

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,JLicense Number NPF-3' f

Seria! ' lumber 2056

't A t t au. men t Page 17 APPENDIX 2 Probabilistic Risk Assessment (PRA) -- Supplementary Information A risk-evaluation utilizing the pr?liminary Davis-Besse PRA has been

-performed and-is summarized as follows:

First, an evaluation of the stsam supply piping located in Ro m 500, 501, and 515 vas performed to estimate a pipe rupture probability.-

Utilizing generic pipe rupture probability data from the Davis-Besse IPE 4

generic database, the-probability of a rupture in the piping was calculated to be 4.lE-5/yr (4.67E-9/ segment-hr *'24 hr/ day

• 365 hr/ day).

[ Note that the gener2, pipe rupture probability data does not take credit r

for the seismic class I qualification, design stress margins, etc. cf the piping.]

Por; comparison purposes, this value can be compared against the i

appropriate' initiating event frequency from the overall PRA.

Of the initiating events, the feedvater/steamline break initiating event applies, with a frequency of 3.62E-3/yr.

As can be seen, the failure prooability of the piping under consideration is more than an order of l-

- magnitude (less than 10%) below that for the feedvater/steamline break L

initiating event frequency.

[It should be noted that while the above numbers are based on somewhat_different data sources and assumptions, the relative.nagnitudes Lare judged to be representative.]

Next, the impact of feedvater/steamline breaks on the overall core damage frequency (CDP) was' estimated to ascertain the effect on plant safety.

While all-the PRA sequences have not been finalized, the transient sequences:vhich include feedvater/steamline breaks are dominated by the "TBU" sequence _(Transient / Loss:of All Feedvater/ Failure of Feed & Bleed).

-This-sequence is also a dominant contributor to the overall CDF. Of those'init1ating. events-included in the TBU sequence, feedvater/steamline breaks make up about:l% of the total. Therefore,-the contribution of L

feedvater/steamline-breaks to the overall CDF is on the order of 1%.

l.

_Given that ithe piping under consideration contributes less than 10%'of feedvater/steamline breaks, and the contribution of feedvater/steamline

, breaks to CDF is on the order of 1%, the impact on CDF from a rupture of this piping is less-than 0.1%.

-In conclusion, a high energy line break of this piping has a negligible impact.-_(i.e.=less-than 0.1%) on the overall plant core damage frequency, h

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