ML20235G662

Package of Documents Re 870101 Plant Trip,Including Region III Daily Rept for 870102,plant Daily Status Rept for 870102 & Main Feedwater Pump Turbine Rept

ML20235G662
Person / Time
Site:	Davis Besse
Issue date:	09/25/1987
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20235G648	List: IA-87-589, Package of Documents Re 870101 Plant Trip,Including Region III Daily Rept for 870102,plant Daily Status Rept for 870102 & Main Feedwater Pump Turbine Rept ML20235G645 ML20235G662
References
FOIA-87-589 NUDOCS 8709300143
Download: ML20235G662 (8)

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. W . . ._ . _ CtNfRATOR 7H 8

_ . _. _ _ _ Output ma e.ti _ move iseees) 6ee 6 pet old 6es .

1. he teoctor hos been erttice! g *ders and the generator has been se line O dets sinto the test ehvideva. *

. I haya etnce lost 1.lcense Seest Report (LES) wee generated _ days.* NumbeI of Ltse* fper to Date *

7. PLANT CtIDittif f F# ttiAR T s In Spee. O.l Out of $ pee. l 18FeDe leproefne a f Stt211eDiBY s le Spec. dj that of Spee. l Stable 1.30 Th&NDe taproving $ table l,U peretterattae Deterferettas ${Ni~l Consentet

t. Curlet.rtr.D tmJUp 19fMt3 FAff 24_tatues The turbine /ge,nerator une eychronssed to the grid et 0234 seues en 1/1/87. Apprestostely 402 Rs power mee schelved prior to the trip of the unit et 1039 hours0.012 days <br />0.289 hours <br />0.00172 weeks <br />3.953395e-4 months <br /> (See description under of f.oorest evented Ceepleted poet trip recovery operettee. recuum wee restored, the meter driven f eed pump supplying f eedwater to the Steen Cenerstere.

Coeristed a solesse f ree the elesetteneses Weste etenitor tank e

T 4 ftANNtp putton tyters atzt to wUng

. Troubleeheet the enuse of the finen Feed pomp /forbine trip.

. perait the it hain feed peep turbine e

tesnote11 the sir esetet eloos of the Mate Steen testatise telees.

t err WetlAt f VENTS PAST 14 ftM8 ,, ,

a The unit tripped f res 601 power et 8039 bevre en 1/1/87 due to e spurieue trip of the fl nele reed Pump seich caused a teatter Trip free the Anticipatory Beacter Trip 3retes (ARTS). As a result of the lose of feedwater, e stese and f eedwater Rupture Centrol System (STtCS) trip occurred first en Steen Generator low level and then on Stems to feed dif f erentiel pressure. It is ovepected that the il MrP forbine has securred bearing dessee. PCAQ $7 0008 ese subeltted to doevoetit this ocewrence.

• CRD Breaker 8 will set stay t!ssed. Meletenance le currently working this problem.

. PCAQ 87-0002 wee submitted to document the pub!!shteg and diettsbuttee of en Lacesplete teeteten to $p lle6.28.

6. ftt.seNICAL SPECIFICAflUp AL"fluu ffAf tMErr9 aprtRED InitCW HAT Dt5ULf IN POWt3 DEf telt'YIup Ut gulf SNUtutnal S tat es ,

tech. spee. hate tech. Date/Tlen Present prAq - Ceres *G sow haber se c. totered Artion to teostred testeett e:eewe st ed t Ac t les I. Fire terriere 3.7.10 pusersue Setes p/A he fee Seguires Appendis R. Modifice.

tlos.

2. Coseous Effluent lastrumentation 3.3.3.10 10/30/85 N/A No me seguires noteamence of FT.5090 and FT 50906

3. Ligute Ef fluent instrumentation 3.3.3.9 12/12/86 N/A We No Seguires heintemence en fiel?OO & 17013

4. Fire Dectacties tese 3.3.3.4 uveereue betes s/A pe Yes Tech spes Choose 8eguired S. Ctet toeletten selves 3.8.3.1 12/*/?t N/A no ne seguired notatemence on su 1357.

7. $r*1005 03 Cort 1NUIpc EQUIPMLNT F308LDi5

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_ , M (' i':s'3 DAVIS-BESSE MAINTENANCE MANAGEMENT SYSTEM 12/31/8 DM 20 PUP 4' .l OPERATIONS' NOTES UPDATE i 0F 10

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D/'E ENTERED #12/31/06 BY- MCLYNO ERASE: N l

1. PLANSfARE TO ESLiiLATE PCWER TO APPROXIMATELY 40% AND HOLD FOR SRB t,PPPOVAL 10 INCREASE POWER.

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2. CONTII;UE MAXIMUh" DILUTION FLOW FOR THE ENVIRONMENTAL GROUP UNTIL APPROVAL BY EPA TO LESSEN LIMITS.

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3. JIM ALBERT HAS FURNISHED THE SHIFT SUPERVISOR WITH A GROUP OF C0hPUTER., POINTS THAT MUST BE PRINTED OUT ONCE PER DAY FOR THE NEXT TWO WEEKS. THIS IS TO DETERMINE IF WE HAVE AN AMBIENT

.Tf;MPERATURE SHIF7 ON THE RCS WIDE RANGE PRESSURE TRANSMITTERS.

y < c 4., ATTENTION FIRST AID TEAM - EMERGENCY MEDICAL RESPONSE CABINET Ci

[ HAS BEEN RELOCATE l) FROM THE 623' LUNCHROOM TO THE 3RD FLOOR PSF (JUST OFF THE TURBINE DECK). ALSO, DUE TO THE TEMPERATURE INCREASE THE 02 IN CABINET ?3 ON 603' HAS BEEN REMOVED. 02 IS

? AVAILABLE IN CADINET ?i PSF, ?2 STATION LOBBY AND ?4 PPF OUT-BOUND.- .

DAVIS-BESSE MAINTENANCE MANAGEMENT SYSTEM 12/31/86 DM020 PUP OPERATIONS NOTES UPDATE 2 0F 10 ERASE: N r

. MAINTENANCE WILL BE WORKING ON THE CIRC PUMP COOLING LINES TO THE THRUST BEARING., ilWO ?'S 1-86-4625-00 AND i-86-4627-00.

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6. SYSTEMS ENGINEEh1NG HAS INVESTIGATED THE CAUSE OF THE. MAIN GENERATOR SEAL DIL PRESSURE SWINGS AND FOUND IT TO BE IN THE MAIN REGULnTOR PCW 2470. IT IS BEING REPAIRED UNDER MWO i-86-4631-00.

y .

T.- STEAM SEAL REGULATOR VALVE GS 2338 IS BEING WORKEp,UNDER MWO j 1-86-4624-00.

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8. TURBIME BYPASS VALVE SP 13B3 IS BEING WORKED UNDER NWO 1-86-4596-00. IF COMPLETED, THIS SHOULD CORRECT THE PROBLEM FOR WORK REQUEST #86-4425 ON CD 430.

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REACTOR TRIPS DEFINITION: Unplanned automatic scram while the reactor is critical.

l (

Reference:

SECY-86-317)

'This Week Since 12/22/86 11/01/86-12/21/86 During SALP Period 0 3 0 3 3- On Mtrch 13, 1987, at 0822 hours0.00951 days <br />0.228 hours <br />0.00136 weeks <br />3.12771e-4 months <br />, the reactor tripped from approx-imately 92% power on high Reactor Coolant System (RCS) pressure.

Cause: The accidental closure of Steam Generator 1-2 Main Feedwater Stop Valve (FW-601), due to a painter inadvertently bumping the valve's local control switch, caused a partial loss of RCS heat sink (the SC) and resulted in increasing RCS pressure and subsequent Reactor Protection System 2300 psig High Pressure Rrzetor Trip.

Other Actuations: A Steam Feedwater Rupture Control System (SFRCS) actuation took place on low SG 1-2 water level.

Plant Response: Plant response was normal except as follows:

Subsequent to the reactor trip, one of the secondary side main steam safety valves (SP17B2) stuck partially open, resulting in increased RCS cooldown. The plant stabilized at 500*F, which is within the RCS pressure-temperature curves.

OEET E 6 at 1039 hours0.012 days <br />0.289 hours <br />0.00172 weeks <br />3.953395e-4 months <br />, the reactor tripped from 40%

power on an Anticipatory Reactor Trip System (ARTS) initiated actua-tion of the Reactor Protection System trip circuitry.

Cause: The il Main Feedwater Pump experienced a loss of lube oil (due to a plug in the lube oil line) and subse-quent MFP bearing damage and vibration, thereby i tripping the MFP. Upon loss of the MFP, ARTS  !

actuated to cause a reactor trip.

Other Actuations: Steam Feedwater Rupture Control System (SFRCS) actuated subsequent to the MFP trip due to 1) Steam Generator low water level due to feedvater loss, and

2) steam to feodwater differential pressure trip ,

197.6 paid also due to loss of feedwater. {

l Plant Response: All systems actuated and performed per design.

1 Davis-Besse l Report No. 18 l

_ _ _ _ . _ _ _ _ _ 4

ENGINEERED SAFETY FEATURES ACTUATIONS DEFINITION: Any event or condition that resulted or should have resulted in manual or automatic unplanned actuation of any Engineered Safety Feature (ESF), including the Reactor Protection System.

(

Reference:

10CFR50.72(b)(1)(iv), 10CFR50.72(b)(2)(ii))

Davis-Besse includes actuation of the AFW System (Steam and Feedwater Rupture Control System (SFRCS)) in this category.

l This Week Since 12/22/86 11/01/86-12/21/86 During SALP Period l

0 6 0 6 l 3- Reactor trip of March 13, 1987

a. Reactor Protection System (RPS) high Reactor Coolant System (RCS) pressure - response normal.

b. Steam Feedwater Rupture Control System (SFRCS) actuation on SG 1-2 low level - response normal.

M7Ta"$$Itibi!ofrJanuarygl@S7/

a. Reactor Protection System (RPS) initiation from the Anticipatory Reactor Trip System (ARTS) - response normal.

b. Steam Feedwater Rupture Control System (SFRCS) actuation on

1) SGs low level and 2) steam and feedwater differential pressure - response normal.

1- Reactor trip of December 22, 1986

a. Steam Feedwater Rupture Control System (SFRCS) actuation on SG 1-2 low level - response normal.

b. Reactor Protection System (RPS) initiation from the Anticipatory Reactor Trip System (ARTS) - response normal.

Engineered Safety Features actuations which occurred during 11/01/86 - 12/21/86

- None 3 Davis-Lesse Report No, IB

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MFPT #1 i

MFPT #1 tripped at 10:39 on 1-1-87 which resulted in an ARTS trip / Reactor trip.

Indication of MFPT #1 turbine bearing vibration increase prior to the trip was indicated by TS1 point recorder.

Investigation into the event showed that the MFPT mechanically tripped due to high vibrations experienced. MFPT did not reach overspeed conditions nor did it experience an electrical trip.

High vibration was due to failure of stub shaft radial / thrust bearing. Blockage of the oil supply by a pipe plug vac the root cause. Situation may have existed since 1984 refueling outage.

Disassembly has been completed. Repairs needed to components in front standard.

Stub shaft, bearing pillow block, and jack shaft bracket are most critical items.

Delivery of stub shaft is currently 1/12/87. Reassembly estimated at approximately 5-9 days with realignment checks. .

L Designing pillow block thermocouple installation to improve monitoring capabilities.

May need to install during refueling outage along with other enhancements of MFPT monitoring capabilities.

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%se EDISON so,so~ ,uu 300 SAAotSoN AVENUE tautoo. oso esse February 13, 1987 KB87-00019 File: RR 2 (P-6-87-01)

Docket No. 50-346 License No. NPF-3 paloR111ROUTIF ggu scond b mi K- --

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Mr. Harold Denton, Director I Office of Reactor Regulation skal u U. S. !?uclear Regulatory Commission E ,lL Washington, D.C. 20555

Dear Mr. Haller:

Nonthly Operating Report, jJahuary[1'987-Davis-Besse Nuclear Power Station Unit 1 Enclosed are ten copies of the Monthly Operating Report for Davis-Besse Nuclear Power Station Unit 1 for the month of January 1987. I If you have any questions, please feel free to contact Morteza Khazrai at (419) 249-5000 Extension 7290.

Yours truly, oY h  ?

P Manager Davis-Besse Nuclear Power Station LFS/MK/1jk Enclosures cc: Mr. A. Bert Davis, v/l Regional Administrator, Region III l Mr. James H. Taylor, Director, v/2 Office of Inspection and Enforcement Mr. Paul Byron, v/l l NRC Resident inspector f0I /)- P 7- $T'/

l Nuclear Records Management, Stop 3220 D/4 LTK/002 ,

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FvrrD 9 97 Gi?

AVERAGE DAILY UNIT POWER LEVEL WKET NO. 50-346 WT Davis-Besse Unit 1 DATE February 13. 1987 COMPLETED BY _prteza Khazrai TELEPilONE . _(419) 249-5000 Ext. 7290 Mom January 1987 DAY AVERAGE DAILY FOWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe Net) (MWe-Neti 50 -

411

_ 37 2 0 18 413 186 g, 410 3

397 20 411 4 .

404 21 410 5

406 22 412 6

409 23 412 7

8 410 24 410 398 25 418 9

to 386 26 438 33 412 27 469 411 492 12 28 l 410 29 466 33 412 3o 461 l 34 15 410 33 471 16 414 INSTRUCTIONS On this format.hst the average d Dy unit power level in MWe-Net for each day in the reporting mimth. Gimpute o the marest whole messwatt.

1 (9/77) l l

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OPERATNG DATA REPORT l

. EDOCKETNO. 50-346 DATE February 13, 1987 I COMPLETED BY Mortera Khazrai TELEPHONE 419 ,249-5000. Ext.

7290 OPERATING STATUS Notes

1. Unh Name. Davis-Besse Unit 1

2. Reporting PW Januarv 1987 2772

3. Lhemsed ThennalPower (MW ): *

4. Nameplate Rating (Gross MWe):

925 906

5. Design Elecencal Raths (Net MWe1: -

904

- 6. Mashnum Dependable Capscity (Gross MWe):

860

7. m=Imum Dependable Capacity (Net MWe):

. S. If Changes Occur in Capacity Ratings (items Number 3 Through 7)Since Last Repon.Give Reasons:

9. Power 1svel To Which Restricted. If Amy (Net MWe):

30. Reasons For Restnctapas. If Any:

This Month YrAo Date Camelative

31. Bours in Reporting Period 744 744 74.568

, 708.5 708.5 36,763.6

12. Number Of Houn Reactor Was Critica!

13. Rosetor Reserve Shutdown Houn 35.5 35.5 4.660.3 )

697.6 697.6 35,186.2

14. Hours Generator On-Lune

15. Unit Reserve Shutdown Hours 0.0 0.0 1.732.5 36, Gross Thermal Energy Genersted (MWH) 1.009.895 ... 1.009.895 82.436.559 .

17. Gross Electrica1 Energy Generated (MWHI 319.592,_. 319.592 27.281.979

18. Net Electrical Energy Genersted (MWH) 287,599 287.599 25,524,262 93.8 93.8 47.2

19. Unit Service Factor 93.8 93.8 49.5 l

20. Unit Availability Factor 44.9 44.9 39.8

21. Unit Capacity Factor (Using MDC Net) - _

42.7 42.7 37.8

22. Unit Capacity Factor (Using DER Net)

23. Unit Forced Outage Rate 6.2 6.2 36.7 l

24. Sheldomus Scheduled Over Next 6 Monttu (Type. Date.and Duration of Each 3:

25. If Shut Dens At End Of Report Period. Estimated Date of Startup:

26. Units in Test States (Prior to Comuneren! Operation): Forecast Acharted MrTRAL CRITICALITY '

INITIAL ELECTRICITY COMMERCI AL OPERATION .

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OPERATIONAL

SUMMARY

JANUARY 1987 The turbine generator was synchronized on line at 0235 hours0.00272 days <br />0.0653 hours <br />3.885582e-4 weeks <br />8.94175e-5 months <br /> on January 1 ,

1987. l The reactor power was maintained at approximately 401 power until 1039 I hours on January 1,1987, when a reactor trip occurred. The trip was the result of the loss of Main Feedwater Pump Turbine (MFFT) il. The MFFT il tripped as a result of high vibration due to bearing failure. MFPT (2 was out of service at the time. The loss of feedwater initiated an Anticipatory Reactor Trip System (ARTS)/ Reactor Protection System (RPS) trip which shutdown the unit.

The reactor criticality was established at 2212 hours0.0256 days <br />0.614 hours <br />0.00366 weeks <br />8.41666e-4 months <br /> on January 2, 1987.

The turbine generator was synchronized on line at 0630 hours0.00729 days <br />0.175 hours <br />0.00104 weeks <br />2.39715e-4 months <br /> on January 3, 1987.

Reactor power was slowly increased to 50% which was attained at approximately 1900 hours0.022 days <br />0.528 hours <br />0.00314 weeks <br />7.2295e-4 months <br /> on February 3, 1987. The power increase was limited between 50% and 701 power the rest of the month due to the testing and troubleshooting of MFPT #1.

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REFUELING INFORMATION DATE: January 1987

1. Name of facility: Davis-Besse Unit 1

2. Scheduled date for next refueling shutdown: February 1988

3. Scheduled date for restart following refueling: April 1988

4. Will refueling or resumption of operation thereafter require a technical specification change or other license amendment? If answer is yes, what in general will these be? If answer is no, has the reload fuel design and core configuration been reviewed by your Plant Safety Review Committee to determine whether any unreviewed safety questions are associated with the core reload (Ref. 10 CFR Section 50.59)?

Ans: Expect the Reload Report to require standard reload fuel design Technical Specification changes (3/4.1 Reactivity Control Systera and 3/4.2 Power Distribution Limits).

5. Scheduled date(s) for submitting proposed licensing action and supporting information: Summer, 1987

6. Important licensing considerations associated with refueling, e.g.,

new or different fuel design or supplier, unreviewed design or performance analysis methods, significant changes in fuel design, new operating procedures.

Ans: None identified to date.

7. The number of fuel assemblies (a) in the core and (b) in the spent fuel storage pool.

(a) 177 (b) 204 - Spent Fuel Assemblies B. The present licensed spent fuel pool storage capacity and the size of ,

any increase in licensed storage capacity that has been requested or  !

is planned, in number of fuel assemblies. l I

Present: 735 Increase size by: 0 (zero)

9. The projected date of the last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity.

Date: 1996 - assuming ability to unload the entire core into the i

spent fuel pool is maintained, l

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., COMPIETED FACILITY CHANGE REQUEST y

{

FCR 50.85-027 SYSTEM ' Auxiliary Steam

, COMPONENT Supports No. HBD 87-H9 and BBD 87-H10 l .j CHANGE, TEST OR EXPIREMENT j i

FCR 84-027 performed the following changes and modifications; l i

1. Replaced the bottom channel on support BBD 87-H9

2. Repisced the top and bottom channels on support BBD 87-H10 l

3. Modified the weld design wh:lch had welds around the entire perimeter I of the channels to welding only three (3) sides of the channels.  !

This FCR 84-027 was closed November 4, 1986.  ;

REASON FOR CHANGE FCR 84-027 was initiated due to NCR 83-123 and 83-0068 which identified nonconforming welds on structural members of subject supports.

SAFETY EVALUATION

SUMMARY

This FCR 84-027 replaced horizontal structural members on supports HBD 87-H9 and HBD 87-H10 and modified the weld design from all around perimeter to three sides only.

The safety functions of the supports is to restrain the piping under a postulated seismic loading. The supports have been analyzed and found to be acceptable for short term operation (BT-14402). Modification performed by FCR 84-027 will allow long text operstions. Based on the above issue an unresolved safety question does not exist.

TEMP 1 E/236/1

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i COMPIETED FACILITY CHANGE REQUEST l FCR No.45-131 SYSTEM Pr. actor Coolant System

\

COMPONENT PSV-RC02 l

CHANGE. TEST OR EXPERIMENT FCR 85-131 modified the PORV discharge line by adding a pair of flanges in j the PORV discharge line. 1 l

This FCR 85-131 was closed October 23, 1986. j 4

REASON FOR CHANGE $

j This modification was incorporated to provide easier access to the PORV for removal and reinstallation.

SAFETY EVALUATION

SUMMARY

The purpose of this modification improved conditions related to PORV maintenance. The installation of these flanges permitted the removal of ,

the valve and spool piece as a unit, eliminating lengthy exposure and {

maneuvering of the valve around existing piping and equipment. The j installation of these flanges does not alter the intended design function j of the piping system.

Therefore, the occurrence, consequences, malfunction, probability and possibility of an accident previously evaluated in the Updated Safety Analysis Report (USAR) is not increased as a result of this modification. q The margin of safety as defined in Tech. Spec. has not been reduced. This J change does not involve an unreviewed safety question.  ;

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TEMP 1 E/236/2

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[ COMPLETED FACILITY GANGE REQUEST l

l b FCR NO.85-071

-i SYSTEM Main Steam System {

COMPONENT Various .

CHANGE, TEST OR EXPERIMENT FOR 85-071 modified the various pipe supports on the main steam inlet I lines to the Auxiliary Feedwater Steam Turbines 1-1 and 1-2. .)

This FCR 85-071 was closed on November 11, 1986.

1 REASON FOR CHANGE i This modification was implemented due to Nonconformance Report 85-0019, which identified several supports needing repairs, ranging from replace-ment of expansion anchors to replacement of pipe saddles.

SAFETY EVAL 11ATION

SUMMARY

The function of a pipe support is to provide support for all design loads.

The pipe support from the main steam inlet lines to the auxiliary feedwater  ;

pumps was analyzed by Bechtel with one bolt inactive. This was found to a meet all short and long term conditions. However, the base plate for this support was found to be damaged and required rework per NCR 85-0071. The damaged anchor bolt was replaced and this restored the support to the original design.

This modification does not create a reduction in the margin of safety as defined in the basis for any Technical Specification. Therefore, an unreviewed safety question does not exist.

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TEMP 1 E/236/4

l r-H CONPI2TED FACILITT CRANGE REQUEST y FCR NO.84-193 }

SYSTEM Moisture Separator and Reheater System COMPONENT HSR Safety Valve Vents '

f 0

CHANGE, TEST OR EXPERIMENT FCR 84-0193 allowed the installation of a threaded reheat safety valve l vent line to the condenser.. i i

This FCR 84-193 was closed February 21, 1986.

REASON FOR CHANGE The safety valve vent lines were hard piped to the condenser during valve maintenance these lines needed to be cut. Installing unions facilitates maintenance work on the safety valves.

SAFETY EVALUATION

SUMMARY

The subject vent lines are used to vent the area above the bellows in the case of a bellows leak in these valves. These lines serve no nuclear safety related function. Addition of unions does not effect any nuclear safety related items. This change does not constitute an unreviewed safety question.

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TEMP 1 R/236/5  !

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