ML19305D309

Monthly Operating Rept for Mar 1980

ML19305D309
Person / Time
Site:	Davis Besse
Issue date:	04/07/1980
From:	Sarsour B TOLEDO EDISON CO.
To:
Shared Package
ML19305D308	List: ML19305D307 ML19305D309
References
NUDOCS 8004140353
Download: ML19305D309 (12)

Text

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AVERAGE DAILY UNIT POWER LEVEL 50-346 DOCKET NO.

Davis-Besse Unit 1 WIT April 7, 1980 DATE Bilal Sarsour COMPLETED BY

, Ext.

TELEPHONE 251 March, 1980 MONTH c DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)

(Mwe. Net) 628-632 g

37 635 633 18 2

652 626 39 3

663 632 3

4 664 631 21 5

664 630 22 6

610 634 23 7

632 629 24 3

640 631 25 9

635 632 26 10 256 630 27 11 0

634 28 12 0

629 29 13 54 635 30 14 562 632 33 15 638 16 i

INSTRUCTIONS On this format, list the average daily unit power levelin MWe. Net for each day in the reporting month. Compute to the nearest whole megawatt.

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OPERATI G DATA REPORT 50-346 DOCKET NO.

DATE April /, 1980 COMPLETED BY Bilal Sarsour TELEPHONE 419-259-5000, Ext.'

251 '

OPERATING STATUS Notes Davis-Besse Unit 1

1. Unit Name:

March, 1980

2. Reporting Period:

2772

3. Licensed Thermal Power (MWt):

925

4. Nameplate Rating (Gross MWe):

906

5. Design Electrical Rating (Net MWe):

934

6. Maximum Dependable Capacity (Gross MWe):

890

7. Maximum Dependable Capacity (Net MWe):

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

None

9. Power Level To which Restricted,if Any (Net MWe):

10. Reasons For Restrictions. If Any:

This Month Yr.-to-Date Cumulative 744 2,184 22,709 i1. Hours In Reporting Period 680.25 1,918.95 12,883.15

12. Number Of Hours Reactor Was Critical 0

0 2.875.8

13. Reactor Reserve Shutdown Hours 668.88 1,349.38 11,724.18

14. Hours Generator On-Line

15. Unit Reserve Shutdown Hours 0

0 1.728.2

16. Gross Thermal Energy Generated (MWH) 1.305.851.5 4,371,629.5 24,571,136.5

17. Gross Electrical Energy Generated (MWH) 446,623_

1,479,072 8,202,583 416,323 1,389,440 7,560,018

18. Net Electrical Energy Generated (MWH) 89.9 84.7 53.2

19. Unit Senice Factor

20. Unit Availability Factor 89.9 84.7 61.5 62.9 71.5 37.5

21. Unit Capacity Factor (Using MDC Net) 61.8 70.2 36.9

22. Unit Capacity Factor (Using DER Net) 10.1 15.3 25.8

23. Unit Forced Outage Rate

24. Shutdowns Scheduled Over Next 6 Months (Type.Date.and Duration of Each):

April 10, 1980 12 Week Refueling Outage

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

26. Units In Test Status (Prior to Commercial Operation):

Forecast Achiesed INITIAL CRITICALITY INITIA L ELECTRICITY COMMERCIAL OPERATION (uf77)

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50-34 DOCKET NO.

Davis-Besse Unit 1

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UNIT SHUTDOWNS AND IOWER REDUCT:ONS UNITNAME DATE April 7, 1980 g

1 Bilal Sarsour COMPLETED by REPORT MON 111 March. 1980 TELEPilONE 419-259-5000. Ext. 251 t

E C-Cause & Corrective

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j.y 5 Licensee h

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Action to No.

Date g

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.E s s Event g

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Prevent Recurrence mu H

y 3 g=,3 5 Report a o

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2 80 03 27 F

75.12 A

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!!P-33-80-32 RB INSTRU Manual reactor trip after control rod safety g,roup 3 drop.

See Operational Summary for further details.

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Method:

Exhibit G Instructions I

2 F: Forced Reason:

1-Manual for Prepara'..on of Data S: Schedu!cd A Equipment Failure (Explain) 2 Manual Scram.

Entry Sheets for Licensee B Maintenance of Test 3 Automatic Scram.

Event Iteport (LER) File (NUREG.

C.Refuelir g 4 Other (Explain) 0161)

D Itefulatory Restriction 12.-Operator Training & License Examination 5

F.Administratise Exhibit 1 Same Source G operational Enor (Explain)

(9/77) ll-Othes (Esplaud

OPERATIONAL

SUMMARY

MARCH, 1980 3/1/80 - 3/18/80 Reactor power was maintained at 71% with the turbine generator gross load at 675 10 MWe.

Power was reduced to 71% to extend the core to the desired outage date and to enable a shutdown of Reactor Coolant Pump (RCP) 1-1 without a transient if seal problems developed.

3/19/80 The reactor power level was maintained at 71% with the generator gross load at 675 i 10 MWe until 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> on March 19, 1980 when reactor power was increased to 74% with the generator gross load at 700 i 10 MWe to more closely match the desired core burnup before the refueling outage.

3/20/80 - 3/23/80 The reactor power was maintained at 74% until the morning of March 23, 1980, when additional seal problems indicated the need to remove the RCP l-1 from service. The reactor power was then reduced to approximately 68%.

RCP 1-1 was shutdown Sunday, March 23, 1980 at 1320 hours0.0153 days <br />0.367 hours <br />0.00218 weeks <br />5.0226e-4 months <br /> due to further degradation of the second stage seal.

Steam Generator 1-1 level reduced to the low level limit setpoint and the resultant feedwater swings placed the integrated control system in track.

The unit stabilized out at 55% approximately six minutes after the pump was shutdown. Power was returned to approxi-mately 71% at 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> on March 24, 1980.

3/24/80 - 3/29/80 The reactor power was maintained at approximately 72% with the generator gross load at 670 1 10 MWe until 1007 hours0.0117 days <br />0.28 hours <br />0.00167 weeks <br />3.831635e-4 months <br /> on March 27, 1980 when control rod safety group 3 dropped causing reactor power to drop to a low of 43% of full power. The turbine control valves began closing to compensate for the reduction of steam generator outlet pressure, causing the generator output to fall. At approxi-mately 1009 hours0.0117 days <br />0.28 hours <br />0.00167 weeks <br />3.839245e-4 months <br />, control rod safety group 4 also began to insert in jog speed (3 in/ min). The safety control rods could not be with-drawn by the control room operators, therefore, at 1014 hours0.0117 days <br />0.282 hours <br />0.00168 weeks <br />3.85827e-4 months <br />, the reactor was manually tripped.

3/30/80 - 3/31/80 The reactor was critical at 0207 hours0.0024 days <br />0.0575 hours <br />3.422619e-4 weeks <br />7.87635e-5 months <br /> on March 30, 1980. The tur-Reactor bine generator was synchronized on line at 1322 hours0.0153 days <br />0.367 hours <br />0.00219 weeks <br />5.03021e-4 months <br />.

power was slowly increased and attained approximately 72% full power on March 31, 1980 with the generator gross load at approxi-mately. 650 i 10 MWe, I

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I March, 1980 DATE:

REFUELING INFOP 1ATIOR i

Davis-Besse Nuclear Power Station Unit 1

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1.

Name of facility:

April, 1980 Scheduled date for next refueling shutdown:

2.

July, 1980 Scheduled date for restart following refueling:

l 3.

Will refueling or resumption of operation thereaf ter require a technical If answer is yes, what, 4.

specification change or other license amendment?If answer is no, has the reload !

in general, will these be?

and core configuration been reviewed by your Plant Safety Review Committee i h to determine whether any unreviewed safety questions are associated w t l

the core reload (Ref.10 CFR Section 50.59)?

i Yes, see attached i

I l

Scheduled date(s) for submitting proposed licensing action and supporting February, 1980 (revision submittal expected) 5.

l information.

Important licensing considerations associated with refueling, e.g., new or

'different fuel design or supplier, unreviewed design or performance analysis 6.

methods, significant changes in fuel design, new operating procedures.

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

storage pool.

0 (zero) 177 (b).

(a) l The present licensed spent fuel pool storage capacity an 8.

in number of fuel assemblies.

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0 (zero) 1 Increase size by Present 735 The projected date of the last refueling that can be discharged to the spent 9.

fuel pool assuming the present licensed capacity.

1989 (assuming ability to unload the entire core into the spent fuel pool is maintained and the unit goes to an lo monta retueling cycle >

Date

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r REFUELING INFORMATION Continued Page 2 of 2 The following Technical Specifications (Part A) will require revision:

4.

Reactor Core Safety Limits (and Bases) 2.1.1 & 2.1.2 -

2.2.1 - Reactor Protection System Instrumentation 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 following 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 Parameters (and Bases) g p

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COMPLETED FACILITY CHANGE REQUESTS FCR NO:

77-070 SYSTEM: Nuclear Sampling System COMPONENT: Grab sample source valves CHANGE, TEST, OR EXPERIMENT:

On August 7, 1979, drawing change notices were issued to change the indicated position of 104 grab sample source valves on the piping and instrument diagrams (P& ids) from nomally closed to normally open.

REASON FOR THE FCR: The '

changes were requested via FCR 77-070 to minimize in-plant radiation exposures in compliance with the NRC Regulatory Guide 8.8 "as low as reasonably achievable" guidelines.

Having the sample source valves normally open eliminates the need to enter radiation areas every time a sample is to be taken in order to open these valves.

i SAFETY EVALUATION: A case by case analysis of the safety significance of leaving these source valves open was performed. The analysis shows that safety is not com-promised by this change and it is not an unreviewed safety question.

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COMPLETED FACILITY CHANGE REQUESTS FCR NO:

79-178 SYSTEM: Reactor Coolant System (RCS)

COMPONENT: Reactor Coolant Pumps (RCP) starting interlock -

CHANGE, TEST, OR EXPERIMENT: On August 3, 1979, the setting of the RCP starting inter-This lock was raised from a power level of 22% of full power to 60% of full power.

allows an RCP to be started at a power level up to 60% of full power. This change was made under the guidance of the nuclear steam supply system (NSSS) vendor, Babcock cnd Wilcox. The following procedures have been revised to reflect this change:

SP 1103.06, "RCP Operation"; PP 1102.04, " Power Operation"; PP 1101.01, "NSSS Limits cnd Precautions".

REASON FOR THE FCR: A review by Babcock and Wilcox of the requirements for starting a fourth RCP while at power, revealed that for starting the RCP when at a higher power level than the'22% limit, a smaller pressure transient will result.

The set-point was then raised in order to eliminate the requirement of lowering reactor power l

below 22% before starting a fourth RCP.

SAFETY EVALUATION: A 10CFR50.59 review for changing the interlock setting so that a fourth reactor coolant pump can be started for power levels up to 60% of full power has been conducted by the NSSS vendor, Babcock and Wilcox. The scope of the review included simulation runs as well as a study of data from another plant to determine the magnitude of the reactor coolant pressure peaks for fourth RCP restarts at various 1 power levels for both the beginning and end of core life. The consequences of tripping the reactor during the fourth RCP restart at any power level as well as other potential cafety concerns were reviewed as well From the results of the review it was concluded that the change does not constitute an unreviewed safety question because:

i 1.

The probability of occurrence or the consequences of an accident or mal-function of equipment important to safety, previously evaluated in FSAR, has not been increased.

2.

The possibility of an accident or malfunction of a different type other than any evaluated previously in the FSAR has not been created.

3.

The margin of safety as defined in the basis for any technical specification has not been reduced.

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COMPLETED FACILITY CIIANGE REQUESTS FCR NO:

79-407 SYSTEM: Integrated Control System (ICS)

COMPONENT:

Setpoints CHANGE, TEST, OR EXPERIMENT: On November 17, 1979, the ICS low level setpoint for cteam generator (SG) level control was raised from 28 inches to 35 inches. This was performed in order to provide adequate margin between this setpoint and the Steam and Feedwater Rupture Control System (SFRCS) low SG level trip setpoint.

In addition, the ICS setpoints for the reactor low limit for the reactor /SG demand signal and for the reactor power low level limit for Tave control were both raised 10% (from 15 to 25% and from 10 to 20%, respectively).

REASON FOR THE FCR: The SFRCS low level trip setpoints were adjusted to comply with the limits of Technical Specification 3.3.2.2 (see Licensee Event Report NP-32-79-12).

To provide a margin between the SFRCS low SG level trip setpoint and the ICS low SG level setpoint to an adequate amount, the ICS low SG 1evel setpoint was-raised from 28" to 35".

This in turn raised the reactor power level at which Tave reaches 5820F to approximately 25% thus necessitating the above changes to the ICS.

SAFETY EVALUATION:

See attached f

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Safety Evaluation FCR 79-407 l

l The SERCS Steam Generator low level' trip set point index value is 17"+2" above the Startup range low level tap.

For reactor power < 15%, the actual Steam. Generator level was 19.17" (IGS I

sees a corresponding value of 28" indicated which is the Steam Generator LLL) l prior to this change. This value is very close to the setpoint index upper limit (19"-).

This could mean unnecessary SFRCS actuations on Steam Generator low levels.

To avoid this situation and at the same time adhering to the technical specification limits, the low level setpoint for the ICS Steam Generator low level limit should be increased (to 35").

This corresponds to 31.5" of actual 5500F water in the Steam Generator.

l The above change needs to change: 1) the integral master control low limit (IG 8.11) from 15% to 25% which will now cause the " Limited Steam Generator l

Reactor Demand to Reactor Control" signal to always be above 25%., 2) the low limit in the reactor control analog logic (RC 11.9) from 10% to 20%

which will now prevent Tavg. control to derive the Reactor demand signal below 20%.

i The following cases are analyzed due to the proposed changes:

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1.

Reactor Power run back to 25%.- Previously the power run back was l

to 15% in case of load rejections With the proposed change,

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this run back will be to 25%.

The' temperature of the primary coolant would still be 5820F and thus there won't be any unnecessary l

skrinkage in the primary coolant.

i 2.

Reactor Tripping 0 - 25% Power - Previously if the reactor tripped between 0 - 15%, the most41T caused in the primary reactor coolant l

was approximately 5820F - 540 F y 42 F.

In th.e new case, if the l

reactor trips between 0 - 25%, this max.11T (the reactor, trips at 25%), j

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vould be less than 420F.

Hence, the reactor ' coolant system is not i

subjected to any excessive shrinkage and pressurizer level changes i

should be reduced.

l 3.

Reactor Trip at Power >25% - The4T for the primary coolant for l

power differential between 25% and the power at which the reactor i

tripped would be the same as before due to the proposed changes,

11he following changes are proposed in the FSAR.

These changes are analyzed above and no unreviewed safety question is involved.

i 1)

Pg. 7 - 33e, Item 5 Steam Generator level - The minimum operational level in each j

steam generator is 35-inches as indicated on S/U range.

I 2)

Pg, 7 - 60. Para. 3 l

6 The reactor control is designed for automatic or manual operation above 25% output and for manual operation below 25%.

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3)

Pg. 7 - 62, Para. 7.7.1.2.4 Wherever a value of 15% appears, it should be changed to 25%. -

4)

Pg. 7 - 63 Item 3 A minimum water limit is provided for 25% low load control in the downcomer section.

I 5)

Pg. 7 - 59, Section 7.7.1.2 (ICS), 2nd para.

The ICS maintains constant average RC temperature between 25 and 100% rated power and constant steam pressure at all loads.

6) Table 7 - 10, Pg. 7 - 97. Section C (5b)

Steam Generator Level Control (SG Subsystem):

The ICS Controls Steam Generator Water Level to prevent overpumping (high level). This limit ensures Superheated Steam under all operating conditions between 25% and 100%.

The above changes will result in procedural control changes and the operators will be retrained for these.

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By doing the above, the margin of safety as defined in the bases for any technical specification has not been reduced.

Pursuant to the above, this should be considered as a 10CFR50.59 review of the i

proposed changes.

i l

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COMPLETED FACILITY CHANGE REQUESTS FCR NO: 79-441 (including Supplement 1)

SYSTEM: Reactor Coolant System (RCS)

COMPONENT: Hot leg whip restraints R-1, R-3 and R-4 on both hot legs L

CHANGE, TEST, OR EXPERIM'ENT: On December 29, 1979, all modifications to RCS hot leg whip restraints requested under FCR 79-441, including Supplement 1, were completed.

Specifically, the modifications involved the complete removal of the shims on the R-3 whip restraints on both hot legs and the removal of the adjustable portions of the north and south shims on the R-1 and R-4 whip restraints installed on both hot legs.

These modifications make the gap between the shims and the hot leg, or between the restraint ring and the hot leg in the case of the R-3 restraints, large enough that unrestrained thermal and seismic movement may take place consistent with the assump-tions used in the RCS seismic analysis.

The restraint modifications were made under the guidance of the unit architect /

engineer, Bechtel Company.

REASON FOR THE FCR: Originally FCR 79-441 involved only a modification to the north-south shims of the R-4 restraints to prevent them from making contact with the boss of the hot leg resistance temperature detector (RTD).

In the process of implementing the modification, it was found that the shims could not be moved without contacting the A subsequent review found that, due to a design error, the spaces between hot leg.

the shims of the R-1, R-3 and R-4 restraint and the hot legs were not adequate to per-nit unrestrained thermal and seismic movement of the hot legs as was assumed by the RCS vendor's seismic analysis, to exist at all times including a design basis s,eismic (The whip restraints serve only to control a pipe whip during a LOCA).

This event.

The modifications deficiency was reported in Licensee Event Report NP-32-79-14.

these deficiencies.

implemented under FCR 79-441 Supplement 1 correct SAFETY EVALUATION: Modifying the north and south shim bars on RCS pipe whip restraints R-1, R-3 and R-3 so that during the design basis earthquake the RCS pipe will not con-tact the whip restraint does not affect the ability of the RCS pipe to perform its The RCS pipe is designed to withstand seismic loads without lateral-design function.

l support.

These modifications will not result in an unreviewed safety question.

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