NRC-89-0259, LER 89-017-01:on 890810,new Analysis of Feedwater Line Break Scenario Assumes Failure of Feedwater Startup Control Valve in Open Position.Review of Pipe Break Documentation Conducted.Updated FSAR Change to Be submitted.W/891215 Ltr

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LER 89-017-01:on 890810,new Analysis of Feedwater Line Break Scenario Assumes Failure of Feedwater Startup Control Valve in Open Position.Review of Pipe Break Documentation Conducted.Updated FSAR Change to Be submitted.W/891215 Ltr
ML19332F798
Person / Time
Site: Fermi DTE Energy icon.png
Issue date: 12/15/1989
From: Orser W, Pendergast J
DETROIT EDISON CO.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
CON-NRC-89-0259, CON-NRC-89-259 LER-89-017, LER-89-17, NUDOCS 8912190057
Download: ML19332F798 (12)
v • d • e


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Wilham S. orser E!!.'i's."kons d 10CFRS0.73 _ 4

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oece.ber 15. 1989 NRC-89-0259 l

1 U. S. Nuclear Regulatory Commission Attention: Document Control' Desk  !

Washington, D.C. 20555 l

Reference:

(1) Fermi 2 NRC Docket No. 50-341 Facility Operating License No. NPF-43 (2) Transmittal of Licensee Event Report 89-017-00 dated September 11 1989.

NRC-89-0178

Subject:

Licensee Event Report _ _(LER) No. 89-017-01 Please-find enclosed LER No. 89-017-01, dated December ,

15, 1989 for a reportable event that occurred on August  !

8 1989. This LER has been revised to reflect the finalized Feedwater and Main Steam Line Break analyses '

that-have been. approved by Detroit Edison. A copy of i

-this'LER is also beire sent to the Regional  ;

-Administrator. USNRC wogion III.

If you have any questions, please contact Joseph Pendergast at (313) 586-1682.

Sincerely j N  :

Enclosure:

NRC Forms- 366. 366A cc: A. B. Davis J. R. Eckert R. W. Defayette/W. L. Axelson W. G. Rogers J. F. Stang Wayne County EmerBency Management Division hDR912190057 891213 ;2Tgf;Q;g

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U.S. NUCLEMd E.EgULATOaY COMM10SION s.PM4VED OMS NO. 31504104 5x*'a58 8'*v" LICENSEE EVENT REPORT (LER) ,

F ACILITY NAME m DOCKET NUMSER (21 FAGE (3i Fermi.2' o i s' l o l o I o l3 l 4 l1 i lor l1 l 1 TITLEtes, llevised Feedwater Line Break Analysis ivlNT DATE 166 LER NUMSE R IG) Pr. PORT DATE 17) OTHER F ACILITIES INVOLVED (St YEAR 5 O i' MONTH DAY YEAR F ACILITV NAMES DOLKET NUMBERtSt f

MONT H DAY YEAR n N/A o1sl01010 1 l l

.0l8 1l0 8' 9 8l9 0 l ' 1l 7 0l 1 1l2 1l5 8] 9 N/A oisto r otoi i i TH18 REPORT IS SUSMITTED FURSUANT TO THE REQUIREMENTS OF 10 CFR {: ICheck one or more of the fonouring) 011 OPE R ATING "oot t 1 20 402m) 20 m., so.7:an2Hi.i n.7i mi PO ER 20 406(sH1H4) 60.341sH1) 50.73teH2Hvl 13.711e1 0 06 09i9 t 20 40.a>H Ha

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- LICENSEE CONT ACT FOR THIS LER Uti NAME TELEPHONE NUM6ER Joseph Pendergast, Licensing Engineer ^ " ' ' ' '

3 ! 11 3 Si8 161-11 1618i2

- COMPLETE ONE LINE FOR E ACH COMPONENT F AILURE DESCRISED IN THIS REPORT (13)

[ CAUSE SYSTEM COMPONENT h U~

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l G vES ur e . more.o rnectrO svowssion Oa re) v ~ NO l l l A88TKACT ILuwt so 1400 spaces. le., soorommatory Muen sine'e-soece troewsreen f>nes! nel l r 1 1 The analysis of the postulated f eedwater line break in the steam tunnel is described in Updated Final Saf e ty Analysis Report (U FS AR)- s e c tion 3. 6. 2. 2. 2. The UFSAR evaluation was completed in the early 1970's. Since documentation for the UFSAR conclusions l could not be retrieved, a new analysis was conducted. The I difference between the new analysis and the UFSAR is that the single failure of the fast closing feedwater pump discharge valves as described in the UFSAR could not be confirmed as the most conservative flood scenario. A failure in the open position of the feedwater start-up control valve was found to be the most conservative flood scenario. With this failure of the feedwater start-up control valve, the break will not be icolated until the tripping of the condensate and hcater feed pumps on low hotwell level. Therefore the consequences of the revised analysis are  ! additional flooding beyond that previously analyzed for this event. The review of the new feedwater and main steam line break analyses has been completed by Detroit Edison. A UFSAR change will be submitted with the next annual UFSAR update in March 1990. 

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U GIIBC Pekm 8BBA U.S. NUCLEA3 EEIULATORY COMM80860N  ; j **" ai._c,._.UCENSEE EVENT-REPORT (LER) TEXT CONTINUATION " 

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Ce EXPIRES; B!3UN j l FACILtTV feAmet (H . DOCKER NtA4th (2) LER NUMSER 14) th0E (3) ~ 

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o f510 lo lo 131411 81 9 0l1 l 7 - 0 11 0l2 OF 1 l.1 1xtin - a, <. .acr asuunn ' I q s .- y i Initial' Plant' Conditions: Operational Condition: 1- (Power Operation) Reactor Power 99.5 Percent Reactor Temperature: 535 degrees Fahrenheit Reactor Pressure: 1010 psig { Description of' Event: l 

                          - The analysis of'the postulated feedwater (SJ) line break in the steam 1 tunnel is described in Updated Final Safety Analysis Report                                                                                   .

(U F S AR) section 3.'6.2.2.2. The UFSAR analysis was-completed-in the early 1970's. Documentatic, supporting the conclusions reached,hasEbeen searched for # could not be retrieved. Therefore a new analysis w a s d w. to. support the UFSAR conclusions;--A difference was found between the new analysis : and- 

                            ' the'UFSAR. conclusions.                                      For the feedwater line-break scenario..the single. failure assumed in the UFSAR.was failure of a-fast closure
                             . (8~ seconds) inboard feedwater pump discharge valve. The break scenario ends when the. slow (88 second) outboard feedwater' pump discharge valves close (ISV) and High Pressure Coolant: Injection
                '                                                                 ~
                             = (HPCI)l (BJ) , restores Reactor Pressure Vessel (RPV) level.
                           - The new. analysis of the feedwater line break se'enario assumes failure of the feedwater startup control valve (TV) in the o' pen                                                                                   L,
                           - positiou.                         This single failure-was relected to maximize flooding' in affected areas of the plant. . The f eedwa ter lin e break scenario with the single failure of the startup control valve is-therefore p                             more conservative. The new feedwater line break would not
 ;                            terminate until'the tripping of the condensate (SG) and heater feedipumps (P) on low hotwell level (SD). No operator action was assumed for the 8 1/2 minutes during which water flows through the E                             postulated pipe break.                                         Supporting documentation is provided in Tables-I. II..and III. and the figures attached to this Licensee

[; Event Report (LER). > l-Cause o f _Ey.en t : A review of pipe break documentation supporting the original FSAR and' existing UFSAR analyses was conducted. Documentation for all pipe break scenarios exists with the exception of the main steam (SB) and feedwater line breaks in the steam tunnel. The UFSAR 1 [ NRC FORM 360A , *U.S. CPos 1968 5J0489 000 70 , 

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LICENSEE EVENT REPORT (LER) TEXT CONTINUATION AreaOvEo ous NO. siso.oio. EXPIRES: 8?31/88 

    #ACILITV NAME tu                                      DOCKEY NUMBER (2)               LER NUMBER (S)                     PAGE (3)
                                                                                   ,E .

w =.v Fermi 2 I o my w ., . , 4 o l5 Io Io l0131411 3l9 - 0l1l 7 - 0l1 0l3 OF 1l1 mac i-- =w nn conclusions resulted from an analysis completed in the early ' 1970's. Documentation supporting these existing UFSAR. conclusions could not be retrieved, therefore a new UFSAR analysis was developed for the main steam and f e edwa t er line br eaks. The UFSAR analysis for feedwater line break was found.to be , non-conservative. Analysis of-Eventi , The new analysis for a feedwater line break in the Steam Tunnel has different. consequences. The Steam Tunnel (NM) and Auxiliary Building (NF) first floor, where the Reactor Building Closed Cooling Water (RBCCW) (CC) heat exchangers are located, reach peak flood depths of 53.7" and'45.8". respectively. The steam Tunnel and RBCCW hec: exchanger room flood elevations then begin to decrease due to flow through the Steam Tunnel floor and equipment drains (DRN). The water in the RBCCW heat exchanger room drains to the southeast corner room of the Auxiliary Building. The northeast corner room peak flood depth of 80.9" is reached due to.. floor drain-flow into the sump. Equipment drain flow causes flooding:in.the southeast corner room to a depth of 169". Approximately 10 hours after the lin e break s, water would begin to spill into the torus and HPC1 rooms. The Torus room peak flood depth reaches 14.8". The HPCI room would reach a peak flood depth of 78.8". With the RBCCW heat exchanger floor drains open, the evaluation concludes that the' control air compressor room could flood. However, when preliminary results of the analysis were received in 1987, to mitigate possible flooding in this room plugs were promptly installed in the RBCCW room floor drains. Plugging of the floor drains eliminates flooding in the control air compressor room and reduces northeast corner room flooding. However, it results in substantially more flooding in the sou*heast corner room. More flooding of the southeast corner room will have no impact on safe shutdown of the plant since Division 2 of Core Spray is assumed to be lost even without plugging the floor drains. The original UFSAR analysis assumed that the drainage system is sufficient to remove the flood water from the Reactor / Auxiliary Buildings. As a result, the sub-basement corner rooms. air compressor room, and Torus room would not have flooded. Reanalysis of the main steam line break in the Steam Tunnel shows that the break' flow and flow path are similar to those described in UFSAR 3.6.2.2.1. Flooding levels due to the main steam line break are bounded by the feedwater line break. The resultant peak. temperature, pressure and humidity are shown in Table III. N%C FORM 34ea *U.s. cro,1960 *>20 $80 000 :0 4343).

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. Texta - . ~ - ,,c,                  .i. u m, The safe shutdown path considers systems necessary to scram the reactor, depressurize the reactor, and to establish and maintain the core cooling utilizing the Residual Heat Removal (RHR) (BO) system.         For feedwater line breaks. the availability of offsite power maximizes the consequential effects of flooding.                                     Therefore.

offsite power was assumed to be available. If offsite power is not available, the Condensate and Heater Feed pumps will trip, thus ending the break scenario sooner. No water will be lost from the reactor since the feedwater check valves are designed to close immediately. 1An evaluation of the effects of a feedwater and main steam line break has been completed, with the exception of Equipment Qualification Evaluation and Hydrostatic Loads on the plant structures. These evaluations will be completed and reflected in the March 1990 UFSAR Update. The feedwater evaluation shows the first floor Auxiliary Building containing the RBCCW. the northeast corner room containing Division 1 Core Spray (BM) and Reactor Core Isolation Cooling (RCIC) (BN). the southeast corner room containing Division 2 Core Spray. HPCI and the Torus room would flood to the elevations shown in Table II. Under these conditions the plant can achieve safe shutdown by using the Automatic Depressurization System (ADS) (BF) and both divisions of RHR. Also, since the HPCI room is not predicted to flood until ten hours after the line breaks. HPCI is available for the first ten hours and ADS and RHR are available throughout the event. HPCI' will restore water level to compensate the loss of feedwater flow. The vessel can be manually depressurized by using the Main Steam Safety Relief Valves. After pressure reduction, the operator places the RHR System (Division 1 or Division 2) in the Low Pressure Coolant Injection (BO) mode. The Residual Heat Removal Service Water (RHRSW) (BI) System is used as the heat sink in the RHR cooling mode. The effects of a main steam line break are confined to the Steam Tunnel, first floor Auxiliary Building and the Turbine Building. The Steam Tunnel temperature is boun41ed by the existing temperature in the UFSAR. The highe r Steam Tunnel pressure has been evaluated and found to be acceptable. The higher temperature and pressure in the first floor Auxiliary Building containing RBCCW has no impact to safe shutdown-since no safety related equipment is located in this area. 4 

  . NRC FORU 366A                                                                                                  *v.s. CFO 1988 SN-W M0 70 69 832
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                                                                                                               - EXPIRES. 8'31/08 FACleliv NAmt til                                            DOCKET NUMSE R (2)              LER NUMBER lel                        PAGE131 naa      $40 .W.        $rJ3:

Fermi 2 0 l5 l0 l0 l013 l 4l1 8l 9 - 011l7 - 011 01 5 0F 111 TEKt I:more spen a requeod, un ennonel NRC form JesA M (1n . i I In UFSAR Section 3.4.4.4 Internal Flood Protection, under site l flooding conditions both divisions of Core Spray, and HPCI, RCIC,  ; the Torus room and control air rooms are flooded. With the loss i of Divisions 1 and 2 of Core Spray. RCIC and HPCI, safe shutdown capability of the reactor is not affected. ADS and both divisions of RHR are still available for safe shutdown of the plant. I Therefore the consequences of the feedwater line break are bounded l by this analysis.  ! 

                    , Corrective Actions:

When the UFSAR feedwater/ main steam line break analyses I documentation was determined to be missing, a further review of other pipe break analysis documentation was conducted for other systems. This review concluded that appropriate documentation is available for all other postulated breaks. i i The feedwater and main steam line break analyses have been I reviewed and approved by Detroit Edison. A UFSAR change will be submitted with the next annual UFSAR update in March 1990. l Previous Similar Events: This is the first LER describing an inaccu' rate UFSAR analysis.  ; 1 1 

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LICENSEE EVENT REPORT (LER) TEXT C3NTINUAT10N . unovto ous no sico-oioo EXPlRES 4?31/N f ACILITV feAsef of DOCKET NunsetR us LER NUMeER (6) PA06 (31 vi.a a m ,rt' -tre.y: Fermi 2: 0 5 0 0 01314 l 1 819 - 0l117 - 0l1 0l8 O F. 1 l1 l rext er ., ,w. -unc F mswnn TABLE II s , FEEDWATER LINE BREAK IN STEAM TUNNEL  ; 

                                                         . FLOOD HEIGHT WITH RBCCW-                                                                                    '

HEAT: EXCHANGER ROOM FLOOR l -DRAINS PLUGGED - \ 

                                                                                                                                                                        \
     '                        Affacted                                           Flood Area                                                                                      Flood                                          4 Elevation
  • Depth
  • Steam Tunnel 587.988 u -- 'RBCCW Room 53.7" 507.318 45.8" I

Control Air Compressor Room 551.00'** NE Corner Room 0.0"** 546.74' 80.9" l 

                            'SE Corner Room                                      554.088                                                                              1 Torus Room                                                                                169"                    l HPCI' Room.

541.238 14.8" l 546.57' 78.8" l l j;

  • This table-lists maximum flood heights for each area, maximum l

height do not occur simultaneously for all rooms. l This room is unaffected by flooding'but is listed because of l- its significance to the evaluations. - L i l-k f l NAC FOMM 366A *d.5. CFOs L 968 4 2 0 - 3M 9, W010 [ [ g 63)

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a l TABLE III MAIN STEAM LINE BREAK IN STEAM TUNNEL J i 

                -l                                         l                                     I                                         l      i TEMPERATURE   _

l PRESSURE j,___ HUMIDITY _ l REVISED'J UFSAR 1 REVISED l__UFSAR____j_ R_EVISED 1_UFSAR__l~ J, Steam _ Tunnel 1_227, F 1_26_6_ F l'4.9 PSIG__1 3.4_PSIG l_ 100% J,_100% l l1st Aux. Bldgl l l 1 l l l l l(RBCCW Roomlgj,_215__F__J_175 F l 0.8 PSIG l 2 25 PSIG J_ 100% J,100% l ~ ,!'l 1 I i ( I ( '. lf 

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