ML20100F041

Proposed Tech Specs Allowing Station to Begin Reactor Coolant Heatup & Conduct Zero Power Physics Testing Prior to Entering Mode 1 of New Fuel Cycle (Cycle 5) Utilizing Present Configuration of Startup Feedwater Pump

ML20100F041
Person / Time
Site:	Davis Besse
Issue date:	12/03/1984
From:	TOLEDO EDISON CO.
To:
Shared Package
ML20100E978	List: ML20100E976 ML20100E993 ML20100F041
References
TAC-56390, NUDOCS 8412060438
Download: ML20100F041 (18)

Text

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l p 1.0' DEFINITIONS DEFINED TERMS l.1 The DEFINED TERMS of this section appear in capitalized type and are

[ applicable throughout these Technical Specifications.

THERMAL POWER 1.2 THERMAL POWER shall be the total reactor core heat transfer rate to I:

the reactor coolant.

E RATED THERMAL POWER 1.3 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2772 MWt.

OPERATIONAL MODE 1.4 An OPERATIONAL- MODE shall correspond to any one inclusive combina-

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\ tion of core-reactivity condition, power level and average reactor coolant temperature specified in Table 1.1.

ACTION 1.5 ACTION shall be those additional requirements specified as corollary i statements to each principle specification and shall be part of the j; specifications.

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OPERABLE - OPERABILITY 1.6 A system, subsystem, train, component or; device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s).

Implicit in this definition shall be the assumption that all necessary '

l attendant instrumentation, controis, normal and emergency electrical power-sources, cooling or seal water, lubrication or ether auxiliary equipment,

' that are required for the system, subsystem, trai,n, component or device l

to perform its function (s), are also capable of performing their related l support function (s).-Pejor to ede;,w g e s / ./;,. (7 /e y d.u.aGl5sr [eedusader sys/ art OPERf81L/7Y p),a)) ge, c{e  ;,,g w;+/r& censiclerabion o f f h e. .5Ya $ s of & y/ac hp

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DAVIS-BESSE. UNIT 1 -

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I 8412060438 841203 PDR ADOCK 05000346 PDR

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, : Docket N).. 50-346L Licano N3. NPF-3:

' Serial'No. 1105 Attachment"B SN ' Emergency Re' quest Circumstances and Justification-The following explains the circ'umstances leading to this emergency request and justification of the need for prompt NRC action.

Potential Consequences of Auxiliary Boiler Use During Heatup and Zero

. Power Testing

'The1 Davis-Besse plant has been shutdown for 95 days during which time approximately one-third of the reactor core has'been replaced with new fuel. After'this extended shutdown period and corresponding redo ' ion in decay heat.: insufficient thermal. energy remains in the RCS to maintain

reliablefoperation of the~ Main.Feedwater Pumps for plant startup. Under these JLme decay heat conditions (coupled with the unavailability of the SUFP),~it,is necessary to use the Auxiliary Boiler to provide steam to the Main Feedwater Pumps.

.When the Auxiliary Boiler.is-supplying steam to the Main Feedwater Pumps,.

a boiler trip.will result in loss of feedwater which actuates the safety

. grade' Steam and Feedwater Rupture Control System (SFRCS). The SFRCS will

.then cause a reactor trip (reportable to the NRC) and actuation of the AFW system.- This will change the steam generator level which.will affect the RCS pressure.and temperature. Such a change in RCS conditions will have a major impact on the physics testing program because the time required to re-established test conditions could be between eight and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for "

-each such trip.'

The Auxiliary' Boiler is normally used-to supply other auxiliary system

-~ steam-loads when main steam is not available Past experience has shown that use of the boiler when supplying steam to the Main Feedwater Pumps cin. addition to other auxiliary steam requirements can result in sufficiently

, high steam demand to affect the performance of the boiler.

3 Thus, it can be determined that use of the Auxiliary Boiler can result

'in:

1. Increased challenges to safety systems (SFRCS, AFW).

2. Increased reported reactor trips.

-3. Significant lengthening of the zero power physics testing program, with a corresponding increase in need for replacement power.

Based upon the above, it is clearly undesirable to use the Main Feedwater Pumps ~and Auxiliary Boiler to supply feedwater in lieu of the Startup

'Feedwater Pump.

Chronology of-Events Necessitating An Emergency Request In May'1984, while performing a review related to the SUFP system it was discovered that use of the SUFP. presented a high/ moderate energy line

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DocketlNo. 50-346

, Y TLic*ina Ns. NPF-3

' Serial'No; 1105- l

Attachment-B.

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p break hazard to-the Auxiliary Feedwater.'(AFW)l system. TEDS internal

. mechanism"(SurveillanceLReport) for capturing and tracking such. potential

> pr'obles items .through resolution was implemented. TED committed to the.

it - LNKC-.to11mplement'certain operational restrictions and compensatory i: faeasures concerning operationfo the SUFP.

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~ Efforts were=then initist'ed to resolve the prchlem from a technical

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. standpoint. JA;1ead engineer was assigned to; follow this problem.and Bechtelf(TEDS l architect / engineer) was requested to develop potential

! . options for early resolution of<this problem. This engineering review

. effort resulted in a determination that no viable solution could be.

developed and' implemented in the time remaining before completion of the LFall. refueling outage.

L . Concurrent with the above engineering effort, TED personnel performed t

further safety reviews to quantity the.. significance of this problem to the Davis-Besseldesign basis. A second engineer was' assigned to perform an analysis to quantify the. significance of the risk to the AFW system due to the'use of the SUFP..

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- This -review of risk resulted-in a determination that the real risk to the

?AFWS was not significant and that continued use-of the SUFP for normal

'startups and shutdowns was technically justified. - This determination was

~ documented-in a safety evaluation and was reviewed and approved by TEDS

onsite.and corporate safety review boards.

n g Work then commenced on a request for permission from the NRC to use the '.

i SUFP for startup and shutdowns. .TlH) recognized the operational importance

j. offthe SUFP to Davis-Besse and took great effort to assure, through a'n ,

extensive technical review and comment period..that the technical review Lwas complete and' accurate. The request for a Safety Evaluation Report

- was' submitted on October- 18, 1984. 'At the time ~of submittal, TED considered l itsLtiming to be adequate to allow appropriate NRC review and subsequent approval prior-to the restart of Davis-Besse.

It was subsequently determined, however, that a license amendment was the appropriate licensing vehicle to resolve this problem. Accordingly, TED n  : expedited'its internal review and submittal of a license amendment request.

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'In' addition to the specific activities' mentioned above, discussions were ,

P,  : held with both NRR and OI&E personnel to keep them appraised of the i

situation. This chronology demonstrates that Toledo Ediron (TED) has shown prudent and proper management prioritization with regards to, efforts

to achieve a timely resolution of'the SUFP issue. Since sufficient time

-'does not remain prior to the present scheduled need for the SUFP to meet the noticing procedure of 10 CFR Part 50, Section 50.91 cannot be met and b..

gg an' emergency circumstance exists. Toledo Edison firmly believes that an '

< expedited review and approval of this license amendment is warranted.

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a . Dockat N).:50-346.

M iLic5nts'No. NPF-3

Serial No.'.1105- '

' Att'achment : B '.

A 4-

,' Justification for an Emergency Request' Tole'do: Edison's share of the Davis-Besse station isl428 HWe which is approximately 32 percent-of the~ projected winter. season demand of 1350.

MWe. lut extended unavailability of.the Davis-Besse. station will create-

.the-need for~TED to. purchase'approximately 150 MWe until the Davis-Besse:

, station is' synchronized.to-the grid. In addition,_ Cleveland Electric and illiuminatingicompany will be required to purchase 170:MWe. replacement i  : power. These~ figures are contingent upon winter weather conditions within the combined service areas"and the continued operation of other electric generating units. The cost of. replacement power will most likely be high

'if.other' electric utilities are experiencing heavy load demand, and.the

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-possibility exists that power might not be=available should other utilities

experience generation problems or should weather conditions cause'an increase in electrical' demand.

.At this time there are no major restraints'to the station's scheduled' December 284 1984, synchronization to the grid.other than the standard NRC restart approval and NRC approval of this Technical ~ Specification change.

Emergency action on the attached amendment request is required to remove

/t he latter. restraint.

k Based upon the above informations we believe Toledo Edison has acted prudently and expeditiously in its attempts to make a timely application

.for this amendment and resolve the-SUFP issue.. Furthermore, in recognition lof the power demand situation in our service area and the need_-for power from the Davis-Besse station, we firrly believe an emergency' request and <

NRC approval'is justified.

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_' /,f Lic:n:1 N;. NPF-3 gQ Serial No.111051 * ' -

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if.l ' Attachmeiit' C  ? ,

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p#$si 2 ' I '_Startup Feedwater Pump Issue Chronology e

, 1984  ;

May- ' June July- ' August September October ' November 1

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ProblemLStudied. = SRB /CNRB Meeting l Submittal NRC Review, ler.i JLER Generated.- Review. with NRC. :to NRC.- License Idepti- > Solutions Identi- ~__..

NRC Review. -Amendment ffled.. -fied.. kequest.

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.Date' ACTIVITY DESCRIPTION

,g8 NY. 5 i May 15.L 1984' -

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~$oledoEdisonpreparesSurveillanceReportonStartup Feedwater Pump (SUFP) operating contrary to USAR x description.- _

1May June,~1984 Toledo Edison reYiew of SUFP sittiation and prepar-ation of a. safety evaluation.

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..T V Juce 18, 1984 Toledo Edison:: generates Deviation Report on SUFP

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situation. _ .

July-11,- 1984 Toledo' Edison initiates implementation of Facility Change Request (FCR) 84--119 to perform a 10 CFR 50.59 review and obtain-NRC concurrence for interim ,

use of the.SUFP with administrative procedtires.

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July l'8, 1984 .

Toledo Edison Licensee Event Report (LER) No',84-009 submitted reporting the discovery of'the SUFP concern.

Probabilistic Risk AssessmetSt (PRA) - based justific-Augist 3, 1984

'ation for SUFP operation codpleted.

August 8, 1984 Station Review Board (SRB) meeting and approval of i

e FCR 84-119.

E, ,- _ , & ..

• - Au' gust 10, 1984 Company. Nuclear Review Board (CNRB) meeting and s- s approval of FCR 84-119. .

. Sept. 17, 1984 Toledo. Edison internal meeting on SUFP modification.

,' ' Sept'. 19, 1984- Toledo, Edison meeting.with NRC Staff regarding SUFP problem.

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Sept. 26, 1984 Toledo Edison internal Design Review meeting with Operations Department.

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> >.*; .,-. Docket Ns. 50-346-:

-Lican n No. NPF-3

-Serial No. :1105:

' Attachment C-Date ACTIVITY DESCRIPTION 0ctobei[1'8, '1984 ' Tol'edo Edison letter (Serial 1070) submitted to the i' NRC, with the PRA and Safety Evaluation, requesting expedited review'of the interim use of the SUFP.

November 5,:1984'  : NRC Staff notifies Toledo Edison that a License Condition on the SUFP is necessary.

November 7, 1984 Toledo' Edison' initiates implementation of FCR 84-192, on the proposed SUFP License Condition' (2.C. (3)(t)).

. November:7, 1984 SRB meeting and approval of FCR 84-192.

n November 9. 1984- - CNRB meeting and approval of FCR 84-192.-

November 12, 1984. Toledo Edison letter (Serial 1100) submitted to the NRC requesting a License Condition for the SUFP.

< November 20, 1984 NRC Staff notifies Toledo Edison that a 30-day public

. notice period will be necessary for the proposed License Condition.

- Nov.' 20 -29', 1984 Toledo Edison /NRC discussions concerning a proposed license amendment.

-November 21, 1984 Toledo Edison letter (Serial No. 1093)' submitted general system description and flow diagram for the proposed new SUFP.

November 28, 1984 Toledo Edison initiates implementation of FCR 84-208, on an. emergency request for a license amendment modifying Technical Specification Section 1.6.

. November 28, 1984 SRB meeting and approval-of FCR 84-208.

' November' 29,-1984' CNRB meeting and approval of FCR 84-208.

~ December: 3, 1984 Toledo Edison requests NRC approval of a change to Technical Specification Section 1.6 on an emergency

+ basis.

December 19, 1984: Scheduled-Davis-Besse need for SUFP for zero power physics testing.-

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- .  ; -Docket:N2.:$0-346 ,,

iL ic;nca:Na, NPF-3: ,

Serial No. 1105-

" Attachment D M c  ;

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,;, Event-Evaluation 7 A description of the postulated event use'd to evaluate-the; interim i

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(December-19, 1984 until entering-Mode.1) Startup Feedwater Pump (SUFP) hae is as follows:

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OIAI Initial Plant Conditions:

l 1.= . Reactor Coolant System (RCS) Temperature - The startup.feedwater

~ pump'will be used to support the RCS ' neat-up from a RCS temper-f _ ature of 280*F,(minimum temperature for entry into-Mode 3) to

-530*F.

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2.. LRCS Pressure - App'roximately 2155 psi.

3.- - Relevent Equipment Status - Steam and Feedwater' htupture Control System (SFRCS) and Auxiliary Feedwater (AFW) system lined up and i

-.available on demand; Emergency Core-Cooling System (ECCS)

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E subsystems'available on demand; SUFP. lined up and id operation; Reactor Coolant. Pumps _(RCP) operating.

14. Calculated decay heat load in the core.- 2.3 MWt-(Equivalent to effective heat available from the 113 pre-cycle 5 fuel assemblies F'

.after-95Ldays shutdown.)'

o b B. ' Evaluation of Postulated Failure Events::

1. 'The~ assumed initiating event being valuated is a loss of all

. feedwater (main,. auxiliary and startup) with ne, loss of offsite

< power. For this case, offsite power is assumed c.o remain S 'e

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'available for maximizing the aaount of energy contained in the RCS. ' The non-seismic! suction and discharge piping.of the SUFP > ,

!- ~is assumed to fail in such a manner as to render'both trains of-L Auxiliary Feedwater (AFW). inoperable." This is postulated to r result in a ' total . loss of secoridary ' side cooling. The; inter-

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p ruption of water to the Steam Generators will result in a-

[ reactor trip.from the Anticipatory Reactor Trip System (ARiS)

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I dueto a Steam and Feedwster Ruptu're Control System (SFRCS)

L Actuation' signal. The SFRCS trip will be" generated as a result

._of either Low-Steam Generator' Level or Steam to Feedwater Differential' Pressure. Due to the reactor trip, the Control

.' Roon operators will i.nplement, the: appropriate steps of Station

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~ Emergency Procedure EP 1202.01, "RPS, SFAS, SFRCS Trip or SG Tube-Rupture".

Sinceit(isassumedthatAFWisunavailableduetotheSUFP

,s+,' piping-ftilure, RCS pressure will begin to rise due to the lack 4 '

,of heat removal. As RCS' pressure approaches the sub-cooling d!Q r. _ limits the operator will trip the Reactor Coolant Pumps as per W ' EP 1202.01. This will limit the energy input to the RCS to

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Licinse No.-NPF ,

- Serial _No. 1105 L ' Attachment D ,

k IEP 1202.'01fdirects the~ operator to respond to a loss of main and auxiliary'

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, feedwater .by taking the following actions:

1.1 .Open the, Pilot ~ Operated Relief Valve.

, 2 .~ Open the RCS Hot: Leg High-Point Ve'nts

13. -Open_the' Pressurizer Vent

- 4.1 Start the. Makeup Pumps.-

. 5. . Ensure .that the High ' Pressure : Injection Pumps start when~

the'RCS pressure drops-to 1650 psig.

LFor the purpose ~of; conservative evaluation, acceptability.of handling this

, ' postulated event is ensured assuming only safety-grade equipment with the Disposition'of-a'limitingLaingle failure. . For this postulated event the No.12 Diesel Generator is assumed'to' fail and, therefore, removes one High Pressure Injection (HPI)T Pump and the pressurizer vent path. This vent

path could remove-more energy..than the Hot Leg High Point Vents due to its steam. space location..

LVenting the RCS'to depressurize, and injecting coolant into the RCS would

. " continue within-appropriate pressure temperature limits until the Decay _

Heat, Removal System can be put. in service. EP 1202.01 identifies several

.  : combinations to: accomplish this process. For this evaluation the minimum safety' grade'eq'uipment2(Table 2) capability available with the single

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._ failure of No. 2' Diesel Generator would be'one HPI train and.two Hot Leg

.'High' Point Vents-(Table ~1, Option'1).- This combination of one:HPI pump

.ani two Hot Leg High PointL Vents results in a 4.8 MWt removal, more than

-tw' ice the required heat removal capability.; No excess inventory -loss '

would result;'therefore, no^ core uncovery would take place. Therefore,

'even'in the unlikely event of a loss of startup feedwater concurrent with

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c La loss of both trains of Auxiliary Feedwater, the concern can safely be mitigated.

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. Table 1: ' Comparison of Selective' Options'Available* '

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' Excess Heat' Estimated . Initial Rate. Required:. . Removal.

Flow Through , of Heat . Heat . :yAvailable 2 '

Option- Description RCS-(gpm)- Removal (MWt)' Removal (MWt)' (MWt) 4 4

9

{ 1 1:HPI Train with'2 RCS' Hot 78 .4.8 2.3 2.5.

Leg High Point. Vents f 2 1 Makeup Train with 2 RCS Hot. .90 5.5 2.3 3.2' l

Leg.High Point Vents -

3 1 HPI train with 1 RCS Hot 39 2. 4 . 2.3 0.1 .,

j Leg High. Point . Vent s

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• Options utilizing two pumps will result in more conservative flows and heat removal rates. -

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-Table 2: Potential Cooling Components-3 Onsite 1E Non-1E Component Seismic Powered- Powered Power Supply Description ' Comments Yes . No. 1, Train Non-Essential

-PORV No -

r 120 VAC PORV Block No Yes -

No. 1 Train Essential Valva 480 VAC Prescurizer Yes Yes -

No. 2 Train Essential Vant RC 200 480 VAC Prazzurizer Yes Yes -

No. 2. Train Essential Vant RC 230A 480 VAC Hot Leg No. 1 Yes Yes -

No. 1 Train Essential Each High Point Vent has an AC powered valve 3 High Point-Vent 120 VAC, 120 VDC- and a DC powered valve in series.

Hot Leg No. 2 Yes Yes -

No. 2 Train Essential Each High Point 1 Vent has an.AC powered valve High Point Vent 120 VAC, 120 VDC and a DC powered valve in series.

Makrup Pump No. 1 No Yes -

No. l' Train Essential 4160.VAC Makeup Pump No. 2 No Yes -

No. 2 Train Essential-4160 AC HPI Pump No. 1 Yes Yes -

No. 1 Train Essential 4160 VAC HPI Pump No. 2 Yes Yes -

No. 2 Train Essential 4160 VAC Letdown Coolers No NA NA . Cooled by fully safety grade, seismic, com-ponent cooling water. Letdown is isolated-by the Safety Features Actuation System on low RCS. Pressure (1650 psig)..

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Docket N3.~50-346-lLicin W N3. NPF-3

A Serial-No. 1105:

iAttachment E Safety Evaluation iThe requested change to Technical-Specification Section 1.6 would allow the Davis-Besse Nuclear Power Station Unit No. 1 to complete reactor coolant heatup~and conduct zero power physics testing prior to entering Mode 1 of the new;fuelicycle (Cycle 5) utilizing the present configuration

of ~ the :SUFP and associated pipelines and valves.

The safety function of the-AFWS is-to provide an alternate source of water to the steam generators for heat removal. The SUFP system itself performs

.no; safety function.> It is, however, used as a backup to the main and auxiliary feedwater systems for supplying water to the steam' generators in case of the total loss of these two' systems.

The S. U FP is' located in Room 238.which is the same room as one of the auxiliary feedwater pumps (AFP), pump 1_-2. During'the revier of the

- high/ moderate ' energy line break criteria as it relates to the AFW rooms,

it was determined that the SUFP system can jeopardize AFP 1-2 in the event

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.of a pipe leak or rupture.

'The SUFP system is non-seismic downstream of valve FW91. The suction of the SUFP system utilizes the Deaerator or the Condensate Storage Tank (CST)'as'its water' source. The normal source is the Deaerator. It also uses the non-seismic Turbine Plant Cooling Water (TPCW) system for pump cooling.~ The'line from the Deaerator to the SUFP also runs through Room 237. Room 237 contains AFP 1-1.. The discharge line from the SUFP is aligned to the inlet of the high pressure feedwater heaters.. In the past, these systems-vere not valved off so as to provide immediate backup for the Main Feedwater System (MFWS) if needed.

,Since this problem was identified, the SUFP system and the TPCW system in Rooms 237 & 238 have been isol'ted.

a This isolation has been accomplished by closing valve FW91 from the CST, valve FW32 from the Deaerator, valves

! CW196 and CW197 used for pump cooling, and valve FW106 in the discharge

l. 'line.

L i- The concern with the location of the SUFP is the potential for pipe whip

[ and jet impingement in Room 238 and flooding and high temperature in

[: Rooms 237 and 238. The~ concern with the~TPCW system is the potantial I' for flooding.these rooms. These concerns will only be realized when the SUFP system is available for operation.

! During the period that the startup pump is operating, the suction piping L to the.SUFP is a moderate energy line based on the criteria in USAR H Section 3.6 which states that a line outside containment operating above l 275-psig or 200'F is a moderate energy line. The discharge piping from

'the SUFP is a high energy line. based on the USAR Section 3.6 which states

that a line outside containment operating both above 275 psig and 200*F j; is a high energy _line. If a high energy line is in service more than six

l. hours, Section 3.6 requires that it must be analyzed for pipe rupture.

l The SUFP system could be.in operation for as long as 336 hours0.00389 days <br />0.0933 hours <br />5.555556e-4 weeks <br />1.27848e-4 months <br /> (14 days) l.

during reactor coolant heatup and zero power physics testing. The TPCW system supply and return piping are neither moderate nor high energy lines bu t', since the lines are non-seismic, a flooding concern remains.

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.a / Docket Na. 50-346 l

_ ILican=3 Ns.-NPF-3

-Serial No. 1105 Attachment E:

cit has~been-postulated.that during a' seismic event the pipes would rupture

in such a manner as to damage AFP 1-2 or possibly AFP 1-1 due to flooding.

lit -has also been postulated that with a.high energy pipe break the sheared

- pipe could damage AFP 1-2 due to jet impingement.or pipe whip and would

-cause high temperatures and pressure in Room 238. A moderate energy pipe break could-damage either AFP 1-1 or-1-2 due to flooding and high temper-1sture in eithe'r Room'237 or 238. 'A-rupture / break in the TPCW piping could damage either AFP 1-1 or 1-2:due to flooding in either Room 237 or 238.

.A Probability Risk Assessment'(PRA) study (Attachment G)'has been' performed-

~

!since this situation was discovered. The PRA documents that the worst esse probability for a rupture / break in the-SUFP and the TPCW piping

= causing the failure of AFP 1-2 is of the order of 1.45E-6 per the 14-day

. periodf (conservatively the: period of time under this amen dment during Ewhich the~SUFP would be operating). The probability for failure'of

-AFP 1-1 in Room 237 is smaller due to less SUFP piping in the room. The probability for failure of 'the AFWS due to pipe . rupture / break in the SUFP s

and the TPCW system is documented.in Calculation Number C-NSA-45.02-03.

This probability is insignificant in light of thetAFWS unavailability on the order of IE-2/yr. for each train which was submitted to the NRC in December, 1981.

Although these risks to the AFWS from the SUFP systems are considered insignificant, the SUFP suction and discharge p'iping were hydrotested in the Fall,'1984, to'the original acceptance criteria (ANSI B31.1) to ensure-the integrity of the SUFP suction and discharge piping. In-addition, certain precautionary measures will be observed during SUFP operation, when the SUFP is not- being used as a ' source of auxiliary feedwater. An-operator shall be positioned at the AFW room area when the SUFP is operating in Modes 2~and 3. Upon indication of'a pipe leak the operator will either trip the SUFP locally or contact the control room to trip the SUFP.

This may .not reduce the probability for a pipe break in the SUFP system, however, it will reduce significantly the impact of a SUFP system failure

-resulting in a AFWS failure. He would then close'all SUFP isolation valves which are external to the AFW rooms. This operator action is being

taken since piping leaks are expected to occur prior to,any complete piping rupture. If the'SUFP is being used as a source of auxiliary feedwater, specific direction appropriate to the situation will be provided to the operator by the shif t supervisor.

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In the unlikely event that the SUFP system should fail and render the AFWS

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inoperable, Calculation Number C-NSA-45.02-005' shows that the reactor decay

~ heat"(2.3 MWe) after a 95 day shutdown could be adequately removed (see Attachment D). A heat load of 2.3 MWt requires a flow of approximately 40 gpm for decay heat removal. Three alternatives for the required decay

_ heat removal have been investigated and are presented in Table 1 (see

Attachment D). The first, a combination of one HPI and two hot leg high point vents (this combination is both safety grade and meets the single

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failure: criteria) is capable of a flow of 78 gpm and a heat removal of 4.8 MWt.- This'is a factor of 1.1 times more than is necessary. The second, a combination of one makeup pump and two hot leg high point vents is capable g

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.- .. ' Docket Nos 50-346 g

'Lic~nta Ns. NPF-3 Serial-No.' 1105 rAttachment E=

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_- Jof.a flow of 90 spa and a heat removal ~ rate of 5.5 MWt. 'This is a factor s

ofl1~.4 times more than'is necessary for decay heat-removal. The-third, a

p' ~ combination of one HPI with one hot leg high point. vent, is capable of a flow of 39 gpm and a heat removal rete of 2.4 MWt.' This is also more than

is necessary for decay heat removal.

lThe events postulated'above result'in-a plant. response.thati assures-the

ability of. residual heat removal utilizing safety grade equipment under single failure conditions. This is done without violating full design slimits on-RCS pressure boundary design conditions and-therefore meets the-

' General Design Criteria.34. Additionally, in 1979, an analysis was t performed by Babcock & Wilcox relating to loss of all'fecdwater events.-

An NRC evaluation of the.B&W analysis was provided showing adequate residual-heat removal capability with the injection of coolant into the RCS to assure no core uncovery. This evaluation recognized-that the

-initiating' event w'as highly unlikely but that never-the-less the results were acceptable. Based-on the above, it is concluded that this Technical

_ Specification change does not involve an unreviewed safety question.

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= Significant Hazards Consideration-

~n y The; proposed license amendment would allow the. Davis-Besse Nuclear Power-9 Station. Unit No.1:to conduct reactor. coolant heatup' and cero power jphysics testing prior to entering Mode 1 of the new fuel cycle-(Cycle 5)

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. utilizing the present. configuration of the.Startup'Feedwater Pump-(SUFP) 7 .'and its associated. pipelines and valves.. As' explained below,--this proposed-amendment >does not involve ~a significant hazard.

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iThe safety: function ofLehe_AFWS.is-to provide an alternate source of water to-the steam generators _for heat removal. The'SUFP system itself

_n  : performs no . safety function.- It is, however, used as a backup to the main i'D; n. and auxiliary feedwater systems for supplying water to the steam generators Jin' case of'the total loss of these.two systems.- ,

~ The.SUFP is' located in Room 238 which is the same room as one of.the T auxiliary! fee' dwater pumps (AFP).-pump'l-2. During the TED review of'the high/ moderate energy line break criteria as 'it relates to the AFW rooms, i .it was determined that the SUFP system can jeopardize AFP pump 1-2 in the event of.a pipe leak or rupture.

TheiSUFP system is non-seismic downstream of valve FW91. The' suction of

- the SUFP system utilizes the Deserator or the Condensate Storage Tank N :(CST) as'its water source. The normal. source is the Deaerator. It also

'uses the non-seismic Turbine Plant Cooling Water (TPCW) system for pump

, cooling.::The line from the'Deaerator to the SUFP also runs through

' Room 237. -Room 237 contains AFP:1-1. The discharge line from the SUFP is

. Laligned to the inlet of the high pressure feedwater heaters. In the past, 4 these systems were not valved off so as to provide.immediate backup for' t-the Main Feedwater System (MFWS) if needed.

g Since this problem was identifie'd, the SUFP system and the TPCW system in Rooms 237'& 238 have been isolated by TED. This isolat!.on has been accomplished by closing valve FW91 from the CST, valve FW32 from the

'Deserator, valves CW196 and CW197 used for pump cooling, and valve FW106

- in the discharge line.

4 The concern with the location of the SUFP is the potential for pipe whip 1and jet impingement 'in- Room 238 and flooding and high temperature in Room 237 or 238. The' concern with the TPCW system is the potential for m . . flooding these rooms. These concerns will only be realized when the

SUFP system is available.-

, ;During the period that.the startup pump is operating, the suction piping to the SUFP is a moderate energy,line based'on the criteria in USAR Section 3.6 which states that a line outside containment operating above 275 psig or.200*F is a moderate energy-line. The discharge piping from the SUFP is a high energy-line based ~cn the USAR Section 3.6 which' states that a line outside-containment operating both above 275 psig'and 200*F is a high 5 energy line. If"a high~ energy line is'in service more than six hours,

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.. ' Docket N). 50-346 EC Liern;n No.~NPF-3

'Seriel No. 1105 Attachment F

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Section 3.6 requires that it must be analyzed for pipe rupture. .The'SUFP

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system could be in operation for as long as 336 hours0.00389 days <br />0.0933 hours <br />5.555556e-4 weeks <br />1.27848e-4 months <br /> (14 days) during reactor coolant heatup and zero. power testing. -The TPCW system supply and return piping are neither moderate nor high energy lines but, since the lines are non-seismic, a flooding concern remains.

.It has been postulated that during a~ seismic event the pipes would rupture in such a manner'as to damage AFP 1-2.or possibly AFP 1-1 due te flooding.

lIt has also=been postulated that with a high' energy pipe break the sheared pipe could damage AFP 1-2 due to jet impingement or pipe whip and would

.cause high temperatures and' pressure in Room.238. A moderate energy pipe

' break could damage either AFP 1-1 or 1-2 due to flooding and high tempera-ture in either Room 237 or 238. 'A rupture / break in the TPCW piping could Ldamage either AFP 1-1 or 1-2 due to flooding in either Room 237 or 238.

No' Significant' Increase in Probability or Consequences A Probability Risk Assessment (PRA) study has been performed since this situation was discovered. The PRA documents that the worst case probability for a rupture / break in the SUFP and the TPCW piping causing the failure of AFP 1-2 is of the order of'1.45E-6 per the 14-day period (conservatively the period of time under this amendment during which the SUFP would be operating). The probability for failure of AFP 1-1 in Room 237 is smaller due to less SUFP piping in the room. The probability for failure of the AFWS due to pipe rupture / break in the SUFP and the TPCW system is documented in Calculation Number C-NSA-45.02-03. This probability is insignificant

.in' light of the AFWS unavailability on the order of IE-2/yr. for each train which was submitted to the NRC in December, 1981. Accordingly, this amendment request does not involve a significant increase in the probability or consequences of an accident previously evaluated (loss of feedwater).

Although risks to the AFWS from the SUFP system are considered insignificant,

he SUFP suction and discharge piping were hydrotested in the Fall, 1984, to the original acceptance criteria (ANSI B31.1) to ensure the integrity of the SUFP suction and discharge piping. In addition, certain precautionary

- measures will be observed during SUFP operation, when the SUFP is not being used as a source of auxiliary feedwater. An operator shall be positioned in the AFP room area when the SUFP is operating in Modes 2 and 3.

Upon indication of a pipe leak the operator will either trip the SUFP locally or contact the control room to trip the SUFP. This may not reduce the probability for a pipe break in the SUFP system, however, it will reduce significantly the impact of a SUFP system failure causing a AFWS failure. He would then close all SUFP isolation valves which are external to the AFP rooms. This operator action is being taken since piping leaks are expected to occur prior to any complete piping rupture. If the SUFP is being used as a source of auxiliary feedwater, apecific direction appropriate to the situation will be provided to the operator by the shif t supervisor.

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'Lic7ise Ms. NPF-3:

d# 1 Serial No. 1105'

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'No"Significant deduction in Marain of~ Safety iInithe'unlikelj'eventithat the!SUFP system'should fail and render the AFWS l

_ inoperable, Calculation' Number C-NSA-45.02-005- shows-that'the reactors .  :

decay. heat,(2.3 MWt) after.a 95 day; shutdown could be adequately removed l (see Attachment. D). A heat load.of 2.3 MWt requires -a flow of approximately -

140 sys for decay heat l removal. Three -alternatives: for the required. decay . i (heat. removal *hav'e been investigated.- The firs _t, a combination-of one HPI pump andftwo hot leg high pointhents-(this combination is safety-grade and, meets the.~ single failure criteria).is' capable of a flow'of 78 gym and-a heat-removal rate of 4.8 MWt.1 This is a factor of 1.1 times more than

is necessary." The second, a combination of.one makeup and two hot leg-

- thigh: point vents is capable of a flow of 90.gpm and a heat removal rate of

.5.5 MWt. iThis is a factor of 1.4 times more than is necessary for' decay

.  : heat removal..'The third, a combination of one HPI pump with one hot. leg-thigh' point: vent, is capable of~a flow of-39 spm and a heat removal rate of

2.4 MWt. This.is also more than is necessary for decay heat removal.

-Accordingly .there is not a significant reduction in the margin of safety.

~

No Creation of' the Possibility of a New or Different Accident

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, iAn analysis of a' total loss of Main and Auxiliary Feedwater was performed at.the NRC's request, by B&W;in 1979. This analysis, which showed that a loss:offall feedwater did.not result in core uncovery and resulted.in a

! - calculated peak fuel cladding temperature of less than 800*F, was evaluated -

! by the,NRC and the'conseauences deemed acceptable. This evaluation is ,

i~ ' documented'in Harold.Denton's letter to John'J. Mattimoe (Sacramento i Municipal. Utility District) dated June 27, 1979.'- This previous.NRC

-evaluation of the B&W analysis is applicable to the current' Davis-Besse

situa, tion.. .. Therefore, the license amendment being requested does not

_ create .the' possibility of new or- different accident from any. previously l' evaluated.o ' . ,

! -Based'on'the above information and the attached Safety Evaluation this ,

L Amendment Request would not: (1) involve a:significant increase in the probability.'or cons'equences of an accident previously. evaluated; or (2)

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create the possibilityTof a new or different kind of, accident from any

? accident previously evaluated;- or 3) involve. a .significant reduction in a

- margin of safety. Therefore, the requested license amendment does not- .

, lpresent a significant hazard.

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'I W Docket N>. 50-346:

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- Lic m 3 N3.1 NPF-3 l Serial (No. 1105 Attachment G-PRA' Based Justification for Operation of the SUFP System During SUFP system operation, a portion of the SUFP discharge line and the minimum recirculation line renders itself.to consideration as a high

energy line. .Similarly..a portion,of the suction piping because of use

-with the Deserstor Storage Tank water requires consideration as a moderate energy-line.' The immediate safety impact of l'ack of such high/ moderate

-energy:line considerations is on the Auxiliary Feedwater System (AFWS) since the SUFP is housed in the same room as Auxiliary Feedwater Pump t(AFP) 1-2.; A portion of the discharge.and minimum recirculation lines is routed in this same. room.. In addition, the suction path from the Deaerator

-runs.through both AFP rooms.- With the SUFP in operation while taking suction from the Deserstor the'following lengths of pipes may, pose a Lchallenge 'to the AFWS in view of the high/ moderate energy line breaks.

Furthermore, the'non-seismic TPCW piping, in the AFP rooms (listed below).

poses a potential' flooding concern.

-17 feet of 4" discharge line in AFP Room 1-2

18. feet of 1 " minimum recirculation line in AFP Room 1-2 27 feet.of 6"~ suction line in AFW Room 1-2

.27 feet of 6" suction line in AFP Room-1-1~

62 feet of 4" TPCW line in AFP Room 1-1 40 feet of 4" in AFP Room 1-2 24 feet'of less thanfor equal to 2" in AFP Room 1-2

.The overall figure of merit for any one train of the Davis-Besse AFWS is of the order.of IE-2 per year as deduced from the AFWS PRA study (EDS p Report No. 02-1040-0195, Revision 1) submitted to the NRC in December 1981. This implies that one train of the Davis-Besse AFWS will be unavail-

' 31 able with a frequency of IE-2 per year for all initiating events which may

- require availability of AFWS.

i l- The'SUFP could be in operation for as long as 336-hours (14 days) during l- heatup and zero power physics testing. Assuming the duration of SUFP operation to be 336 hours0.00389 days <br />0.0933 hours <br />5.555556e-4 weeks <br />1.27848e-4 months <br />,.the total worst case probability of any break L (whether high energy, moderate energy or seismic) in the unanalyzed piping which may' challenge the availability of an AFW train is of the order of

. 1.45 E-6 per the 14-day period. The worst case probability is for AFW train 1-2 because of significantly larger overall length of piping in this room. This evaluation conservatively assumes that any rupture of this non-seismic piping will flood the room to the extent of ccusing train inoperability with a probability of unity. For further conservatism this assumes that both AFW trains are rendered inoperable.

' Since the SUFP system failure as postulated above poses a challenge to the AFW. train at a frequency of l'.45 E-6 per the 14-day period, the probability of such SUFP system ruptures / breaks leading to inoperability of an AFW train is insignificant as compared to other failures that may render the AFW system inoperable. It is, therefore, concluded that the above issue L . poses a very minimal risk to. the accomplishment of the safety function of

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Lir nca N3.:NPF-3

-Serial:No.-1105. .

Attachment G.

the'AFWS and an extremely negligible risk to public health and safety.

. Operation ~of the SUFP system during'the time period evaluated above is, therefore, adequately justified.

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