ML19253A960

Monthly Operating Rept for Aug 1979

ML19253A960
Person / Time
Site:	Davis Besse
Issue date:	09/07/1979
From:	Caba E TOLEDO EDISON CO.
To:
Shared Package
ML19253A958	List: ML19253A956 ML19253A960
References
NUDOCS 7909130438
Download: ML19253A960 (23)

Text

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OPERATING DATA REPORT DOCKET NO.

50-346 DATE Septercer 7, 1979 C05tPLETED BY Erda1 Caoa TELEPHONE 419-250-5000. Ext.

236 OPERATING STATUS Notes

1. Unit Name:

Davis-Besse Unit 1 D D

2. Reporting Period: AuRust. 107o

3. Licensed Thermal Power t.\ lwr):

2772 gg 925 g - _ -

4. Nameplate Rating (Gross 51%e): 0 906

5. Design Electrical Rating (Net 51We):

6. Staximum Dependable Capacity (Gross S!We):

to be deteminec a , _[ g t be deteminec

7. 5!aximum Dependable Capacity (Net SlWe):

8. If Changes Occur in Capacity Ratings (!tems Number 3 Through 7) Since Last Report.Give Reasons:

9. Power Level To which Restricted. If Any (Net Ntwe):

N ne

10. Reasons For Restrictions. If Any:

This 5fonth Yr..to.Date Cumulative 744 . 5.331 17,596

11. Hours In Reporting Period 9,621.2 744 2,989.4

12. Number Of Hours Reactor Was Critical 1,655.2 2, e +6. 3

13. Reactor Reserve Shutdown Hours 744 2,898.9 8,632.1

14. Hours Generator On.Line 0 1,728.2 1.728.2

15. Unit Resene Shutdown Hours 2,022.114 7,131,662 17,319.232

16. Gross Thermal Energy Generated 15tWH) 674,179 2,377,397 3,761,152

17. Gross Electrical Ener;y Generated (31%H) ,_,

641,855 2,236,227 5,277,687

18. Net Electrical Ener;y Generated (N!% H) 100 49.7 50.8

19. Unit Senice Factor 100 79.4 61,9

20. Unit Asailabdity Factor

21. Unit Capacity Factor iUsing ilDC Net) to be detemined

22. Unit Capacity Factor (Using DER Net) 95.2 42.3 36.7 0 - 2.8 19.6

23. Unit Forced Outage Rate

24. Shutdowns Scheduled Oser Next 6 Stonths (Type.Date,and Duration of Eachi:

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

26. Units In Test Status iPrior to Commercial Operation): Forecast Achiesed INITIAL CR!T!CALITY INITIAL ELECTRICITY .

CON 15tERCIAL OPER ATION __.

7 909130 V30

AVERAGE DAILY UNIT POWER LEVEL DOCKET NO. 50-346 gg.37 Davis-Besse Unit 1 DATE Septexiber 7, 1979 COMPLETED BY Erdal Caba 419-259-5000, Ext.

TELEPilONE 236 MONT11 August, 1979 DAY AVERAGE DAILY POWER LEVEL DAY AVER AGE DAILY POWER LEVEL (MWe-Net) (MWe-Net) g 874 g7 883 877 gg 877 2

879 39 381 3

878 878 4 20 876 874 5 21 .

863 22 857 6

7 671 23 870 677 873 8 .24 875 , 876 9 3 872 26 878 10 11 881 27 877 12 882 .3 868 880 29 80 13 61 14 30 884 877 IS 31 16 881 INSTRUCTIONS On this format,!ist the average daily unit power !evelin MWe Net for each day in the reporting month. Compute to the nearest whose megawatt.

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OPERATIO:!AL SDCtARY FOR AUGUST, 1979 8/1/79 - 8/6/79 The reactor power level was =aintained between 99 and 100 percent with the generator gross load at 915 i 10 sie.

8/7/79 - 8/8/79 The 90-40-90 percent reactor power transient testing was performed on August 7, 1979 as part of the Unit Load Transient Test, TP 800.23. The 60-30-60 transient with three reactor coolant pumps running was perfor=ed on August 8,1979 to complete the Unit Load Transient Test. This marked the completion of the power escalation test program.

The unit experienced feedwater swings periodically f rom August 8, 1979 to August 13, 1979. The problem was found to be caused when the speed control for the main feed pump turbine would go high causing the Integrated Control System (ICS) to recover and bring the speed back down. The speed control problea was attributed to high temperature in the speed control cabinet for the main feed pu=p turbine 1-1. A fan was tenporarily installed to cool the speed control cabinets.

8/9/79 - 8/21/79 The reactor power level was maintained between 99 and 100 percer.-

with the generator gross load at 915 t 10 Ge. At 1113 hours0.0129 days <br />0.309 hours <br />0.00184 weeks <br />4.234965e-4 months <br /> on August 22, 1979, the ICS went into the tracking mode and followed a reduction in power to approximately 82%. 'The runback was caused when an erroneous low pressure signal from the turbine header pres-sure limiter transducer overrode the ICS signal and began to close the turbine control valves. The turbine load limit was reduced until it controlled header pressure and the throttle pressure itsiter setpoint was increased to the maxi =um. The turbine header pres-sure limiter circuit output was grounded to assure the circuit would not affect the unit if the signal totally failed.

The unit was returned to 100% full power operation by 1515 hours0.0175 days <br />0.421 hours <br />0.0025 weeks <br />5.764575e-4 months <br /> on August 22, 1979.

8/23/79 - 8/31/79 The reactor power level was maintained at 100% with the generator gross load at 915 i 10 INe.

On August 28, 1979, high airborne activity was noticed in all negative pressure areas. The problem was traced to the makeup pump seal drain and a temporary fix was initiated on September 1, 1979, until a more per=anent solution could be found.

%G152

DATE: Aueust. 1979 RERJELING INFORMATI0'i

1. Nane of f acility: Davis-Besse Nuclear Power Station Unit 1 ,

March, 1980

2. Scheduled date for next refueling shutdown:

May, 1980

3. Scheduled date for restart following refueling:

4. Will refueling or resunption of operation thereafter require a technical If answer is yes, what, specificatica change or other license a=endment?

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)?

Yes, see attached

5. Scheduled date(s) for submitting proposed licensing action and supporting information. Decenber, 1979

6. Importcnt licensing considerations associated with refueling, e.g., new or different fuel design or supplier, unreviewed design or perf ormance analysis methods, significant changes in fuel design,.new operating procedures.

The scent fuel pool capacity expansion program was approved by the NRC in Anendnent 19 to the operating license received August 1,19 i9.

7. The nu=ber of fuel asse=blies (a) in the core and (b) in the spent fuel storage pool.

0 (zero)

(a) 177 (b)

8. 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 nu=ber of fuel assemblies.

Increase size by 475 (735 total)

Present 260

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

1989 (assuming ability to unload the entire core into the spent Date fuel pool is =aintained and the unit goes to an 18 month refueling cycle) 33G153 1

REFUELDiG DiFORR\ TION (Continued)

August, 1979 , ,

Page 2 of 2

4. The following Technical Specifications (Part A) will require revision:

2.1.1 & 2.1.2 - Reactor Core Safety Limits (and Bases) 2.2.1 - Reactor Protection System Instrunentation Setpoints (and Bases) 3.1.3.6 - Regulating Rod Insertion Limits 3.1.3.7 - Rod Program' 3.2.1 - Axial Power Imbalance (and Bases)

The f ollowing Technical Specifications (Part A) may also require revision:

3.1.2.8 & 3.1.2.9 - Borated Water Sources (and Bases) 3.2.4 - Quadrant Power Tilt (and Bases) 3.2.5 - DNB Paraneters (and Bases) 0 e.

0001G4

FACILITY CHANGE REQUESTS COMPLETED DURING AUGUST, 1979 FCR NO: 79-247 SYSTEM: Containment Air Coolers COMPONENT: Motor power lead conduits CHANGE, TEST, OR EXPERIMENT: This FCR was written to approve cutting the conduit between the f an casing and the motor on all three containment air coolers. Upon reinstalling the fan motors, the subject conduits were reconnected with threaded couplings. This change was done with the prior approval of the fan vendor, Joy Manufacturing Company, as well as the unit architect-engineer, Bechtel Company.

REASON FOR THE FCR: Due to the conduit configuration outside of the fan housing, the conduit could not be removed in one piece to allow the disconnection of power leads to the fan motor. The motor leads must be removed to allow the fan motors to be removed in order to repack the bearing grease.

SAFETY EVALUATION: The addition of these couplings will not affect the function of the containnent air coolers, nor their seismic qualification. This is,not an unreviewed safety question.

336165

FACILITY CHANCE REQUESTS COMPLETED DURING AUGUST, 1979 FCR NO: 79-265 SYSTEM: Main Feedwater and Main Steam COMPONENT: Seismic Supports CHANGE, TEST, OR EXPERIMENT: On June 29, 1979, under Maintenance Work Order 79-265 the physical work fot (CR 79-265 was completed. This FCR added shims to the hydraulic snubber mounting brackets of seismic snubbers SR-9A and SR-32. These changes have been documented by the unit architect-engineer, Bechtel Company by revisions to Bechtel drawings C-611, C-617 and vendor (lTT Grinnell) drswings CC-12-146-4 and CC-12-128-6.

REASON FOR THE FCR: The shims were added to extend the hydraulic snubber mounting brackets in order to bring the piston position during normal operation closer to the center of its travel.

SAFETY EVALUATION: This change adds shims to seism'ic supports SR-32 and SR-9A. The added shims will assure that the cold and hot piston settings fo,r these hydraulic snubbers meet Toledo Edison acceptance criteria for piston location. Snubber reliability will be enhanced by this change. There will be no adverse effect on the safety function' of the main feedwater and main steam systems. This is not an unreviewed safety question.

33515G

FACILITY CHANGE REQUESTS COMPLETED DURINC AUGUST, 1979 FCR NO: 78-264 SYSTEM: Neutron Flux Monitors COMPONENT: Amphenol connectors in penetrations associated with neutron flux monitors CHANGE, TEST, OR EXPERDIENT: On May 16, 1979, work was completed on Maintenance Work Order 78-1346 which ccmpleted the work called for in FCR 78-264. This change installed Ray Chem type WCSF-N heat shrink tubing over all Amphenol connectors located in pene-trations PlL1L, PZL4G, P3L4S, and P4LlG which are associated with neutron flux monitor-ing. Ray Chem type WCSF-N heat shrink tubing has been qualified for nuclear service and has been successfully tested by Ray Chem for LOCA conditions.

REASON FOR THE FCR: The connectors were covered with heat shrink tubing to ensure their hermiticity during small stecn line breaks and rod ejection accidents within contain-ment during which the reactor neutron flux monitors may be required.

SAFETY EVALUATION: The subject Amphenol connectors are used on the neutron flux moni-tors, which are not needed during or after a LOCA. The connectors are qualified for the design temperatures which can be expected during small stemn' line breaks and rod ejection accidents. The addition of the Ray Chem heat shrink tubing to these connectors will no t af f ec t their function.

83G15'?

FACILITY CRCCE RECUESTS COMPLETED DURING AUGUST, 1979 FCR NO: 79-259 SYSTEM: Containment Air Coolers C0KPON ENT: Temperature Indicating Controllers (TICS) 1356,1357 and 1358 CHANGE, TEST, OR EXPERIMENT: On June 25, 1979, work was completed which changed the 0

setpoints of TIC-1356, TIC-1357, and TIC-1358 frc= 120 F to 500F. Prior to this change, the controllers regulated the Service Water outlet valves of their associated contain-ment air cooler to maintain a proper contaimment air temperature. This change forces the valves to remain wide open during normal operation thereby eliminating the res-ponse tbse requirement. The Davis-Besse setpoint index was revised to reflect this

,aange.

This change is being =ade on a temporary basis until FCR 79-280, which will modify the pneumatic actuators on these valves to bring their response ti=e within the required limits, is implemented. FCR 79-2S0 is scheduled to be i=ple=ented during the 1980 refueling outage.

REASON FOR THE FCR: The stroke time of the service water outlet valves of the coolers was measured to be approxt=ately 75 seconds fro = the fully closed position to the fully open position. For=erly, when the valves were being regulated by their associated temperature controllers to maintain proper containment air te=perature, the valves =ay have been at times throttled to the extent that upcn receipt of a Scf ety Features Actua-tion Systes signal, they =ay not have fully opened within the allowable time (see Licensee Event Report NP-32-79-10).

SAFETY EVALUATION: This FCR provides for the revision to the setpoints on TIC-1356, 1357 and 1358 from 1200F to 500 F. This change will f orce the service water flow control valves in the outlet of containment air coolers to a wide open position during normal operation. With the valves wide open initially, the full service water flow would occur when the service water pump is started. This'FCR will not degrade the cooling function of the contain=ent air coolers,and it is not an unreviewed safety issue.

D36158

FACILITY CHANGE REQUESTS COMPLETED DURING AUGUST, 19H FCR NO: 77-309 SYSTEM: Emergency Diesel Generators COMPONENT: Fuel Oil Storage Tanks CHANCE, TEST, OR EXPERIMENT: FCR 77-509 was written to review work done on Neve=ber 23, 1977, under Maintenance Work Order 77-1709. Under this Maintenance Work Order, the three inch diameter, four bolt flanges on the sounding connections of both fuel oil storage tanks were modified to provide for a 1/2" pipe plug for level measuring purposes. This change has been documented in the appropriate drawings by the unit-architect engineer, Bechtel Company.

REASON FOR TIIE FCR: Removing the three inch diameter, fcur bolt, flange to measure the tank level as required to fulfill surveillance requirements was found to be extremely inconvenient.

SAFETY EVALUATION: This change, which installs a plug for measuring tank level, will have no adverse ef fect on the safety function of the emergency diesel generator system.

6g

FACILITY CHANGE REGUESIS COMPLETED DURING AUGUST, 1979 FCR NO: 77-358 s

SYSTEM: Service Water COMPONENT: Service Uater Pump Motors 1-1, 1-2 and 1-3 CHANGE, TEST, OR EXPERIMENT: This FCR was written to provide drain valves on the upper bearing oil reservoirs of the three service water pumps. On June 23, 1979, Maintenance Work Order 78-2045, which added the drain valves to the already existing oil drain piping, was completed. The affected drawings and piping class sheets were revised by the unit architect-engineer, Bechtel Company, to document the addition of these drain valves.

The service water system procedure and surveillance test, as well as the capped valve procedure, have been modified to reflect the addition of these valves.

REASON FOR THE FCR: The valves were added to facilitate the changing of the motor lubri-cating oil. Formerly, the oil had to be drained by removing a pipe plug in the oil drain line, which lead to oil being spilled on personnel or equipment. The addition of these drain valves eliminates this oil spillage.

SAFETY EVALUATION: This change adds a drain valve in the drain line to the upper bearing oil reservo.r on service water pump motors 1-1,1-2 and 1-3. The valve will make it easier to drain the reservoir. A pipe cap is installed downstream of the valve, and the valve is held in the closed position by a lock wire. This change will not affect the safety function of the service water pumps, but will reduce oil spills when draining the reservoir for maintenance.

336160

FACILITY CHANGE REOUESTS COMPLETED DURING AUGUST, 1979 FCR NO: 79-172 SY STEM: Various Systems COMPONENT: Motor starter overload heaters CHANGE, TEST, OR EXPERIMENT: Maintenance Work Order 79-172, completed on May 15, 1979, completed the physical work associated with FCR 79-172. This FCR replaced overload heaters on Q-listed motor starters, which were found to be not in accordance with the overload heaters size specified on the relay setting data sheet.

REASON FOR THE FCR: This FCR is part of the corrective action fcr Licensee Event Report NP-32-79-04 (see also FCR 79-163). This change replaces the overload heaters which were not in accordance with the relay setting data sheets.

SAFETY EVALUATION: The function of the overload heaters which are being changed under this FCR bear no action on the control and indication circuits. .The controls of the nucicar safety related motors are not affected by the function of these heaters. How-ever, these heaters are designed properly for the given application. No overload relay is used for tripping the essential motor operated valves and dampers. The fault p ro te c tion is provided by the molded case breaker. The duty cycle of the essential motor operated valve decide the-selection of the overload heaters.

It is thus concluded that the safety function of the systems is not affected by changing to the required thermal overload heater size. The changes do not involve an unresolved safety question.

3361G1

FACILITY CHANGE REQUESTS COMPLETED DURING AUGUST, 1979 FCR NO: 79-163 SYSTEM: Various Systems COMPONENT: 'fotor starter overload heaters CHANGE, TEST, OR EXPERDENT: Maintenance b*ork Order 79-163 completed on April 24, 1979, completed the physical work associated with FCR 79-163. This FCR replaced overload heaters on Q-listed motor starters, which were found to be not in accordance with the size overload heaters size specified on the relay setting data sheet.

REASON FOR THE FCR: This FCR is part of the corrective action for Licensee Event Report NP-32-79-04 (see also FCR 79-172) . This change replaces the overload heaters which were not in accordance with the relay setting data sheets.

SAFETY EVALUATION: The overload heaters being changed perform no action on the control or indication circuits. Functionally the overload heaters have no ef fect on the nuclear safety related motors. For essential =otor operated valves and dampers, overload relays are not used for tripping. Molded case breakers provide fault pro _tection. The operation time of the valves is considered in the selection of essential MOV overload heaters.

Therefore, the sa'fety function of the systems is not affected by changing to the proper thermal overload heater size per notes on relay setting sheet 54A. The changes do not involve an unreviewed safety question.

335162

FACILITY CHANGE REQUESTS COMPLETED DURING AUGUST, 1979 FCR NO: 78-039 SYSTEM: Service Water (SW)

COMPO!iENT: Pressure switches PSH 2917, PSH 2917A, PSH 2918, PSH 2918A, PSH 2919, PSH 2919A CHANGE, TEST, OR EXPERIMINT: The setpoint index, Bechtel drawing 7749-M-620S, was revised to document the changing of the setpoints of the above mentioned pressure switches as follows:

Pressure Switch From ))t PSH 2917 115 psig 95 psig PSH 2917A 125 psig 110 psig PSH 2918 115 psig 95 psig PSH 2918A 125 psig 110 psig PSH 2919 115 psig 95 psig PSH 2919A 125 psig 110 psig These changes were made with the approval of the unit architect-engineer, Bechtel Company, and are documented in a revision to the Davis-Besse Uni ~t 1 Setpoint Index.

REASON FOR THE FCR: The settings of these pressure switches were changed in order to allow the service water strainer blowdown valves to operate as designed. PSH 2917A, 2918A and 2919A operate to open their associated strainer blowdown valve when the service water pump discharge pressure valve increases to the switch setpoint. Since the setpoint of the relief valves are 120 psig, the former pressure switch setpoint of 125 psig caused the relief valve to actuate prior to the strainer blowdown valve. This change corrects this off design condition.

SAFETY EVALUATION: The changed setpoints have been tested and resulted in the blowdewn valves opening before the relief valves lif t. This is in accordance with system design.

The setpoint changes will not adversely affect the safety function of the system. This is not an unreviewed safety question.

33G1G3

FACILITY CHANGE REQUESTS COMPLETED DURING AUGUST, 1979 FCR NO: 79-224 SYSTEM: Pressurizer Relief System COMPONENT: Hydraulic snubbers on hanger 30-CCA-8-H12 CHANGE, TEST, OR EXPERIMENT: L' hen the two hydraulic snubbers located on pipe hanger 30-CCA-8-H12 were replaced af ter undergoing functional testing on May 27, 1979, it was found that it was nearly Onpossible to replace and to purge air from the equaliz-ing tubing between the two snubbers. Consulation with the snubber vendor, ITT Grinnell, revealed that replacement of the equalizing tubing was not necessary for the proper operation of the snubbers. As the tubing was not necessary, and since f ailure to purge the air from the tubing would not allow the snubbers to operate properly, the tubing was not replaced. This change is documented on the applicable pipe hanger drawing.

REASON FOR THE FCR: It was nearly physically impossible to replace the equalizing tubing in its prior configuration and no means for bleeding of the air from the tubing was provided. Since the tubing is not necessary for proper snubber operation, and air not purged from the tubing may impair the operation of the snubbers, the tubing was not replaced.

SAFETY EVALUATION: FCR 79-224 involves removal of a pressure equalization line on two snubbers located at hanger CCA-8-H12. Per ITT Grinnell, Providence, Rhode Island, the equalization line is not required; hence, its re= oval will not prevent the snubbers from performing their intended safety function. Since the line is not required for proper snubber performance, no unreviewed safety q:-estion is created.

OONS O

FACILITY CFM GE REOUESTS COMPLETED DURING AUGUST, 1979 FCR No: 79-258 SYSTEM: Startup Test Program COMPONE'iT: 100% Turbine Trip Test CHANGE, TEST, OR EXPERIMENT: Facility Change Request 79-258 was written to initiate a saf ety evluation to verif y that the expected results of a turbine trip from 100%

full power (af ter the addition of the Anticipatory Reactor Trip System (ARTS) do not present any unreviewed safety questions. The basis for this evaluation was reactor trip from 100% of full power which occurred on January 17, 1979. This evaluation also shows that the re.quirements of a 100% turbine trip in the startup test program can be satisfied by using the data from the aforementioned reactor trip f rom 100%

of full power.

REASON FOR THE FCR: FCR 78-496 performed a safety evaluation to use the test results of the 100% of full power load rejection test to satisfy the requirement of a turbine trip test from 100% of full power. Since the addition of ARTS, which initiates a reactor trip on any turbine trip which occurs from greater than or equal to 15% of full reactor power, FCR 73-496 no longer satisfies the test program requirement for a turbine trip test f rom 100% of full power.

SAFETY EVALUATION: Toledo Edison Power Engineering had performed a safety evaluation (FCR 78-496) on performing the 100% load rejection test (TP 800.13) in placa of the 100% turbine trip test (TP 800.14). In light of changes made to Davis-Besse Unit 1 and installation of the Anticipatory Reactor Trip System (ARTS), 100% turbine trip cannot be replaced by the 100% load rejection test. Uith the installation of ARTS, the reactor is tripped almost instantaneously when the turbine is tripped and the transient for all practical purposes is identical if either the reactor or the turbine is tripped first. A new safety evaluation is performed below, utilizing data for a reactor and turbine trip from 100% thermal power to substitute for the 100% turbine tri.p test. Specifically, the acceptance criteria for Reactor Trip Test as outlined in TP 800.14 are shown to be successfully met. The acceptance criteria for reactor trip (instead of turbine trip) were used since with ARTS, a turbine trip causes a simultaneous reactor trip. The reactor and turbine trip transient used is the January 12, 1979 event.

Brief Descriotion of the January 12, 1979 Event: The unit was in Mode 1 with reactor power 2772 MWT and load 900 MWE (gross). The event was initiated by an accidental grounding of containment hydrogen analyzer.

The ground resulted in blowing of the 200 amp internal fuse on the inverter feeding Y2 essential instrument bus. The resultant loss of Y2 ,

bus caused a loss of power to Reactor Protection System (RPS) Channel 2, Safety Features Actuation Systes (SFAS) Channel 2 and Steam and Feedwater Rupture Control System (SFRCS) Channel 2. Prior to this, RPS Channel 3 #

was bypassed for surveillance testing resulting in average nuclear instrumentation (NI) power frem Channel 1 and Channel 3 to be locked out of high auctioneer power circuit such that the NI power signal to the ICS was only the average of Channel 2 and Channel 4. With the loss of MN EM

FACILITY CHA:;GE REQUESTS CCMPLETED DURI"G AUGUST,1979 PAGE 2 FCR 79-258

. power to RPS Channel 2 the average NI power signal to the ICS dropped to approximately 50%. The ICS, attempting to increase the power to the desired 100% called for a rod withdrawal.

Wich the loss of power to RTS Channel 2 the Reacter Coolant System (RCS) flow indicator to the ICS fed through RPS Channel 2 also went to zero.

This caused the BTU limits of the ICS to i= mediately reduce =ain feed-vater flow.

The result of the above sequence of events was a reactor trip (on high flux / delta flux / flow) and an SFRCS trip (on steam ganerator-2 lov level).

The SFRCS trip was caused by the loss of Y2 bus, resulting in ICS cutting back main feedwater to steam generators an BTU limits. The turbine tripped almost at the same ti=e (within 210 milliseconds) the reactor tripped. Fer further details, see Licensee Event Report :!P-33-79-13.

Acccotance Criteria for Reactor Trio Test as Outlined in TP 800.14 8.1.1 The turbine tripped when the reactor was tripped.

8.1.2 The reactor coolant system pressure re=ains above~ the high pressurc injection setpoint of 1600 psig after a reactor trip.

8.1.3 None of the reactor coolant pe=ps trip as a result of the reactor tr.4p.

8.1.4 The turbine bypass system setpoint is transferred to maintain header pressure at 995 + 25 psig.

The header pressure is controlled at the steam generator outlet folloving a reactor trip.

8.1.5 Steam generator outlet steam pressure re=ains below 1155 psig

~

after the reactor trip.

8.1.6 During the reactor trip transient, the pressuri:er level =ust re=ain between ten (10) and 320 inches and be returned to above 40 irches when the plant has stabilized. (Reactor Trip only)

J 8.1.7 Steam Generator Level remains above the initiation point of the SFRCS and beJew 375 inches indication and controls at the low level limit when the plant has stabilized.

~

8.1.8 The steam generator feedwater te=perature remains above 110 F after the reactor trip.

3M [5b I

FACILI?f CHANGE REQUESTS COMPLETED DURING AUGUST,1979 PAGE 3 FCR 79-258 The following discussion demonstrates that all of the above acceptance criteria were satisfactorily met.

1. The turbine tripped 0.21 sec. after the reactor trip.

2. High pressure injection setpoint of 1600 psig was never reached during the transient. The minimum RCS pressure available from the reacti=cter data is 1867 psig.

3. All four reactor coolant pumps were running through the transient and none was tripped as a result of the reactor trip.

4. The ste.:m generator outlet pressure was maintained at 995 + 25 psig following the reactor trip excepting the initial pressure rise caused by ICS runback of main feedwater flow. This acceptance criteria is not required to be met af ter the SFRCS trip, since actuation of SFRCS disables ICS control of the atmospheric vent valves and the Main Steam Isolation Valves are closed.

5. The steam generator outlet steam pressure did not exceed 1155 psig after the reactor trip. The maximum steam generator pressure available on the reacti=eter data is 1079 psig observed on steam generator 2.

6. During the reactor trip transient the pressurizer level reusined within the range of 10-320 inches and was controlled between 88-128 inches (from control room strip chart) after the plant returned to stable conditions.

7. During the applicable period, cain feedwater temperature was of the order of 450-460 F satisfactorily =eeting the 110 F limit.

8. SFRCS tripp.3 on low steam generator level. The trip was not a resuit of the reactor trip but was caused by the loss of Y2 bus which resulted in ICS cutting back main feedwater to the steam generators to balance the mis =atch between reacter power signal to ICS and main feedvater flow. At no ti=e did the level exceed 375" on the startup range. Also, pressure on the secondary side of either steam generator was never lost and sufficient heat sink was maintained. Since the low steam generator level trip of SFRCS was not a result of the reactor trip this acceptance criterion is considered to be satisfactorily met. In addition, Davis-Besse Unit I has successfully demonstrated this capability (steam generator levels above SFRCS trip set point and controlled at low level limit after plant stabilization) on another reactor trip from 99% full power on Nove=ber 13, 1978.

OM3ICa' O

FACILITY CHANGE REQUESTS COMPLETED DURING AUGUST,1979 PAGE 4 FCR 79-258 Pursuant to the above, it is concluded that with the change to the test program proposed in this FCR (79-258): 1 i) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report is not created.

11) A possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report is not created.

iii) The margin of safety as defined in the bases for any Technical Specificaticn is not reduced.

This is not an unreviewed safety question.

6

FACILITY CHANGE REQUESTS COMPLETED DURING AUGUST, 1979 FCR No: 79-216 SYSTEM: Decay Heat Femoval and Low Pressure Injection COMPONENT: Hydraulic snubber on pipe hanger CCB-6-H8 CHANGE, TEST, OR EXPERIMENT: On June 29, 1979, work was completed which added an ex-tension to the mounting of the hydraulic snubber CCB-6-H8 which is on the decay heat removal / low pressure injection systes piping. The affected pipe hanger drawings have been revised to document this change.

REASON FOR THE FCR: The extension to the snubber mounting plate was added to bring the piston pos!" ion during normal operation closer to the center of its travel.

SAFETY EVALUATION: This FCR involves the repositioning of a Q-listed snubber to its correct position. This change enhances the ability of the snubber to perform its safety related function. No unreviewed safety question is involved.

6 936169

}ACILITY CHANGE REOUESTS COMPLETED DURING AUGUST. 1979 FCR NO: 79-185 SYSTEM: Reactor Coolant 3ystem (RCS)

COMPONENT: Pressurizer CHANGE, TEST, OR EXPERIMENT: This FCR was written to authorize the te=porary installa-tion of special test apparatus and to authorize the conduct of an experiment using that apparatus to determine the ef f ectiveness of using the pressurizer Resistance Te=pera-ture Detector (RTD) and heater bridge to measure approximate pressurizer level. The experiment was conducted under Test Procedure TP 550.03, " Pressurizer Low Level Deter-mination Using Pressurizer RTD and Heater Bridge", which was written expressly for this purpose. The experiment was conducted on May 2 and May 3, 1979, under the dir-ection of personnel frem the Oak Ridge National Laboratory, as well as the Reactor Coolant System vendor, Sabcock and Wilcox.

REASON FOR THE FCR: The purpose of the aforementioned experiment was to collect data to assist the personnel at Three Mile Island Unit 2 in measurement of pressurizer level by alternate means.

SAFETY EVALUATION: The pressurizer low level determination using pressurizer RTD and heater bridge will he accomplished in operational mode 5 and will include the following:

Phase 1 - Insrallation of pressurizer RTD circuitry Phase 2 - Instellation of pressurizer heater circuitry Phase 3 - Varying pressurizer level Phase 4 - System restoration This experiment will be conducted within the Ibnits of Tec5nical Specification 3.4.9.1 and 3.4.9.2. This experiment is not described in the Davis-Besse Unit 1 FSAR, however, as the procedure limits and precautions state, the pressurizer heater bank control switch will be in the "0N" position and RCS pressure will be controlled manually with heater banks 23, 4. The pressurizer is rcquired by Technical Specification 3.4.4 to be operable only in Modes 1 and 2. Level will be manually maintained in the pressurizer by monitoring digital voltmeter (DVM) readings, and/or converting uncompensated pres-surizer level to actual level. During the experiment, the level in the pressurizer will be malatained above the pressurizer lower delta P nozzle, to avoid having to re-vent the RCS and vent the level transmitter lower taps. The experiment will be terminated if the level goes below the low level tap. Tests by the RTD vendor and others have indicated that the RTDs can withstand the applied current without adverse effects.

The instrumentation involved in this experiment is Q-listed from pressure boundary of RCS standpoint, but otherwise it is not nuclear safety related.

Based upon a detailed review of tnis experiment procedure, it is concluded that tais experimen t is not an unreviewed safety question.

336170

Rev. 1 - 9/7/79 D OM WW OPERATING DATA REPORT

p. _ _

50-346 DOCKET NO.

f DATE ^uzust i, 1979 b b .,

_k g' C051PLETED BY

t. caca 259-5000 Ext. 7.36 TELEPHONE OPERA TING STATUS Notes

1. Unit Name:

Davis-Besse t' nit 1

2. Reporting Period: Julv. 1979 2772

3. Licen>ed Thermal Power (Mht):

4. Nameplate Rating iGross MWei: 025 906

5. Design Electrical Rating (Net 51We):

6. 512ximum Dependable Capacity (Gross 51We): To be det.

7. Maximum Dependable Capacity (Net 5the): To be det.

8. If Changes Occur in Capacity Ratin;s (Items Number 3 Through 7) Since Last Report.Gise Reasons:

9. Power Level To hhich Restricted.If Any (Net MWe): Zero (uneil Julv 6. 1979)

10. Reasons For Restrictions. !f Any:

NRC OIE Bulletins and Shutdown Orders This 51onth Yr.-to.Date Cumulative 744 5,087 16,852

11. Hours In Reporting Period 8,877.2 49S 2,245.4

12. Number Of Hours Reactor Was Critical 2,648.5 1 246.0 1,858.2

13. Reactor Reserve Shutdown Hours 7,838.1 479.8 2,154.9

14. Hours Generator on.Line 1,723.2 264.2 1,728.2

15. Unit Rerene Shutdown Hours 15.297,118 1,230,451 5,109,548

16. Gross Thermal Energy Generated 151WH)

17. Gross Electrical Energy Generated (MhH) 409,950 _,

1.703.218 5.096.473 381,814 1,594,372 4,635.832

18. Net Electrical Ener;y Generated (5tWH) 64.5 42.4 43.3

19. Unit Senice Factor 100 76.3 59.9

20. Unit Availability Factor To be det.

21. Unit Capacity Factor (Usin; MDC Net)

22. Unit Capacity Factor (Usin; DER Net) 56.6 34.6 33,7 0 3.8 21.3

23. Unit Forced Outage Rate

24. Shutdowns Scheduled Oser Nest 6 51onths iType. Date.and Duration of Each):

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

26. Umt> In Test Status iPnor to Commercial Operation): Forecast Achiesed INITIA L CRIT!CA LITY INITIAL ELECTRICITY COMMERCIAL OPER A TION 336171 (9/77 )

.