ML20237J861

Monthly Operating Rept for Jul 1987

ML20237J861
Person / Time
Site:	Davis Besse
Issue date:	07/31/1987
From:	Khazrai M TOLEDO EDISON CO.
To:
Shared Package
ML20237J535	List: ML20237J534 ML20237J861 ML20237J887
References
NUDOCS 8708180379
Download: ML20237J861 (26)

Text

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AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.

50-346 UNIT D-B Unit 1 DATE August 14, 1987 COMPLETED BY TELEPHONE (419) 249-5000 Ext. 7290 MONT11 July 1987 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)

(MWe-Net) 607 37 864 1

2 604 834 gg 3

104 806 g9 748 4

20 817 804 5

21 l

i 840 798 6

22 824 790 7

804 769 8

24 797 780 9

25 l

10 26 814 l

800 836 11 27 1

795 856 l

12 28 13 713 841 29 863

,4 30 15 871 807 3

l 16 869 I

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

(9/77 )

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OPERATING DATA REPORT.

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DOCKET NO. 50 346 DATE. August 14' 1987 COMPLETED BY Mnrt Khazrai TELEPHONE 619-249-5000 Ext'. 7290 OPERATING STATL -

1. Unit Name:

Davis-Besse Unit 1 Notes

, 2. Reporting Period: July 1987 l 3. Licensed Thermal Power (MWt):

2772

4. Nameplate Rating (Gross MWe):

925

5. Design Electrical Rating (Net MWe):

906 904

6. Maximum Dependable Capacity (Gross MWe):

860

7. Maximum Dependable Capacity (Net MWe):

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

9. Power Level To Which Restricted. If Any (Net MWe):

' 10. Reasons For Restrictions.lf Any:

This Month Yr.-to-Date Cumulative

!!. Hours in Reporting Period 744 5,087 78,983-

'.12. Number Of Hours Reactor Was Critical 744

.4,061.2 40,116.3

13. Reactor Reserve Shutdown Hours 0.0

'143.9 4,768.7

14. Hours Generator On-Line 731.3 3,980.3 38,468.9

15. Unit Reserve Shutdown Hours 0.0-0.0 1, m.5

16. Gross Thermal Energy Generated (MWH) 1,876,090 8,586,104 90,012,768

17. Gross Electrical Energy Generated (MWH) 6.QE d,42_,

2,796,478 29,758,865

18. Net Electrical Energy Generated (MWH) 575,951 2,603,006 27.839,669

19. Unit Service Factor 98.3 78.2 48.7

20. Unit Availability Factor 98.3 78.2 50.9

21. Unit Capacity Factor (Using MDC Net) 90.0 59.5 41.0

22. Unit Capacity Factor (Using DER Nrt)

MA 56.5 38.9

,23. Unit Forced Outage Rate 0.0 4.6 35.0

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

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

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

Forecast Achiesed 1

INITIAL CRITICALITY I

INITIAL ELECTRICITY COMMERCIA L OPER ATION (4/77 )

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C' REFUELING INFORMATION DATE: July 1987 1.

Name of facility: Davis-Besse Unit 1 2.

Scheduled date for next refueling shutdown:

February 1988 3.

Scheduled date for rastart 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 vill 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 l

50.59)?

Ans: Expect the Reload Report to. require standard reload fuel design Technical Specifications changes (2.' Safety Limits and Limiting Safety System Settings, 3/4.1 Reactivity Control Systems, 3/4.2 Power Distribution Limits and 324.4 Reactor Cool &nt System.)

5.

Scheduled date(s) for submitting proposed licensing action and supporting inforr.ation: December, 1987 6.

luportant 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 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 number of fuel assemblies.

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:

1995 - assuming ability to unload the entire core into the spent fuel pool is maintained.

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j OPERATIONAL

SUMMARY

l July 1987 The reactor power was maintained at'approximately 73% power until 2200 l

hours on July 2,'1987 when initiated a manual power reduction to approximately 6% was initiated to investigate suspected Reactor Coolant

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Pump (RCP) motor 1-2-high upper thrust bearing. temperature.

The turbine generator-was.taken off line at 0600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> ~on July 3, 1987.

The Reactor Coolant Pump,(RCP) 1-2 motor bearing. oil systems, bearing housing and bearing temperature. sensor vere inspected and no root cause was evident. RCP 1-2 was restarted and upper bearing temperatures stabilized at normal values. The reactor power was slowly increased to approximately.99% power at 1740 hours0.0201 days <br />0.483 hours <br />0.00288 weeks <br />6.6207e-4 months <br /> on July 4, 1987.

The reactor power was maintained at approximately 99% until 0820 hours0.00949 days <br />0.228 hours <br />0.00136 weeks <br />3.1201e-4 months <br /> on July 5, 1987 when reactor power was reduced to 93% due to the low load requirement.

The reactor. power was slowly increased at 0300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br /> on July 6,-1987 and reached 99% power at 0700 hours0.0081 days <br />0.194 hours <br />0.00116 weeks <br />2.6635e-4 months <br /> on July 6, 1987. At 2108 hours0.0244 days <br />0.586 hours <br />0.00349 weeks <br />8.02094e-4 months <br /> on July 12, 1987 initiated. manual power reduction to-approximately 71% was initiated due to the suspected RCP motor 1-2 high upper thrust bearing temperature. A containment entry was made to investigate the problem and it was-determined that an instrument problem rather than a pump problem-existed.

The reactor power'vas slowly increased at 1012 hours0.0117 days <br />0.281 hours <br />0.00167 weeks <br />3.85066e-4 months <br /> on. July 13, 1987 and reached 100% power at 2100 hours0.0243 days <br />0.583 hours <br />0.00347 weeks <br />7.9905e-4 months <br /> on July 14, 1987. The reactor power was fluctuating between 100% and 84% due to the high condenser pressure caused by high' circulating inlet water temperature resulting from the summer atmospheric conditions.

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COMPLETED FACILITY CHANGE REQUEST FCR NO 81-067 3

i SYSTEM.

Emergency Diesel Generator l

COMPONENT Diesel Generators 1-1 and 1-2 CHANGE, TEST OR EXPERIMENT FCR 81-067 installed an idle start /stop feature on the Emergency Diesel

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j' Generators (EDG).

FCR 81-067 was closed March 2, 1987.

REASON FOR CHANGE The idle start /stop feature allows the EDG to be started to an idle speed for testing purposes. The modification also places the EDG in an idle stop condition after a normal stop signal is received which allows for a gradual cooldown of the EDG. This will prolong the life of the Emergency Diesel Generators.

SAFETY EVALUATION

SUMMARY

The safety function of the EDG is to provide on-site standby power for 4

essential loads required for safe plant shutdown.

The idle start /stop feature installed by FCR 81-067 will be overridden when a normal or emergency start signal occurs and the EDG will start to full speed.

The modification also changed the field flashing circuit so the field receives a flash at 450 rpm vice 400 rpm.

This change was made so the field would not be flashed on an idle start. The safety functions of the field flash circuit is to apply an external DC voltage to the generator field. The field flash voltage is applied when the generator reaches rated speed and is maintained.until generator build-up occurs.

All changes to the control circuitry of the EDG have been reviewed and do not adversely affect or change the safety function of the individual circuits or the entire EDG. Therefore, an unreviewed safety question u

does not exist.

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h COMPIITED FACILITY CHANGE REQUEST FCR'No 83-135 SYSTEM-Station Air COMPONENT SA 501, SA 2010 CHANGE, TEST OR EXPERIMENT FCR 83-135 changed the normal position of valves SA 501 and SA 2010 from

- open to closed.

FCR 83-135 was closed' June 10, 1987.

REAS'ON FOR CHANGE Rust and dirt buildup in the station air line to containment was causing valve SA 2010 to become fouled and not seat properly.

(Ref. LER 83-044)

Closing these two valves will prevent this buildup on the internals of valve SA 2010.

SAFETY EVALUATION SUMMAltY The safety function of SA 2010 is to provide containment isolation on a safety actuation'aignal.

By closing this valve during normal operation its safety function is not violated. Valve SA 501 is a gate valve immediately upstream of SA 2010 and has no safety function.

Closing valve SA 501 eliminates the problems encountered with valve SA 2010 by ensuring.that the moist station air and dirt buildup will not accumulate on the internals of SA 2010.

Closing these two valves does.not increase the probability of occurrence or consequences or the possibility of an accident, or the occurrence of a malfunction of any safety related equipment as previously analyzed or evaluated in the Safety Analysis Report. Therefore, an unreviewed safety question does not exist.

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9 COMPLETED FACILITY CHANGE REQUEST FCR NO 85-0317 SYSTEM High Pressure Injection System COMPONENT Spring Hanger FSK-M-CCA-18-1-H CHANGE, TEST OR EXPERIMENT FCR 85-0317 removed spring hanger FSK-M-CCA-18-1-H from the High Pressure Injection System.

FCR 85-0317 was closed June 10, 1987.

REASON FOR CHANGE NCR 85-1516 detailed deficiencies in spring hanger FSK-M-CCA-18-1-H.

The system will perform its design function without this support, therefore the support was removed.

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SAFETY EVALUATION

SUMMARY

I It was determined that pipe support FSK-M-CCA-18-1-H is not required for the system to perform its design function and it therefore was removed.

The removal of this support does not increase the probability of occurrence or the consequences of an accident or malfunction of safety related equipment as evaluated in the Safety Analysis Report.

Nor does the removal create a possibility for an accident different than any evaluated previously in the SAR, or reduce 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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COMPLETED FACILITY CHANGE REQUEST l

.FCR NO 86-0366 i

SYSTEM.

Hydrogen Control l

COMP 0NENT' M-1020/HI l

CHANGE, TEST OR EXPERIMENT.

FCR 86-0366 removed: weld interference with an attachment bolt on a I

support for the Hydrogen Control System locate'd inside containment.

.i FCR 86-0366 was closed June 26, 1987.

j REASON FOR CHANGE A small length of weld on a support for the Hydrogen Control System had to be removed to facilitate installation of valve CV 5011B seismic support.

i SAFETY EVALUATION'

SUMMARY

The. safety function of this support is to restrain / support valve CV 5011B and motor under normal and seismic loading conditions. The engineering l

justification for this modification was that the small amount of weld removed does not reduce the capacity of the connection since the connection is still stronger than the actual member.

Since the repair returned the support to its design function intent, there are no adverse affects on

.the valve.or motor.

This modification does not increase the probability of occurrence or consequences or the possibility of an accident, or the occurrence of a malfunction of any safety related equipment as analyzed or-evaluated in the Safety Analysis Report. Therefore, an unreviewed safety question does not exist.

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COMPLETED FACILITY CHANGE REQUEST FCR NO 81-323 SYSTEM Control Room Chlorine Detectors COMPONENT AE-5358A, AE-5358B CHANGE, TEST OR EXPERIhENT Delete chlorine detectors AE-5358A and AE-5358B at the cont-i. room ventilation intake from the 'Q'-List.

FCR 81-323 was closed June 29, 1987.

REASON FOR CHANGE I

Control room intake chlorine detectors (AE-5358A and AE-5358B) are not required to be on the 'Q' List.

SAFETY EVALUATION

SUMMARY

The function of the 'Q' List is to identify nuclear safety related equipment that requires Quality Assurance in accordance with the requirements of 10CFR50 Appendix B.

The function of the chlorine detectors at the control room HVAC intake is to initiate automatic control room isolation upon a chlorine gas release, however control room HVAC intake chlorine detectors are not required for the l

accidents which have been analyzed in the FSAR. The detectors will not i

actuate for the chlorine line break accident because the detector setpoint will not be reached. For the tan!. car rupture accident the chlorine l

detectors located near the tank car will initiate control room isolation significantly before the control room intake chlorine detectors. Therefore, the control room HVAC intake chlorine detectors are not nuclear safety related.

The criteria for the chlorine detectors as specified in the FSAR is that, "the means used to initiate automatic isolation should meet single active failure and seismic criteria". The detectors were origina~11y included in the 'Q' List because of the requirement that they be redun-dant, seismic class I and have independent essential power supplies.

However, the existing control room intake chlorine detectors while they are still required to be seismic class I, do not possess the current requirements for a Q-component.

Based'on the above no unreviewed safety question exists.

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COMPLETED FACILITY CHANGE REQUEST FCR NO 85-0060 SYSTEM Process.and Area Radiation Monitoring COMPONENT

-Various CHANGE, TEST-OR EXPERIMENT FCR 85-0060 modified the electrical conduit supports for the Process and Area Radiation Monitoring system.

FCR 85-0060 was closed June 29, 1987.

REASON FOR CHANGE The modifications were required in order to compensate the rigid conduit which crosses building seismic joints and to upgrade the existing supports.and conduit system to meet the long term operability requirements.

SAFETY EVALUATION

SUMMARY

The conduit supports, as originally designed, assumed no load contribution from the adjacent seismic zones due to the relative displacements of the buildings and differences in seismic zones.

In order to upgrade the supports to meet the long term operability requirements, FCR 85-0060 modified conduit supports by replacing existing 1/2 inch diameter HDI's with 3/4' inch diameter HILTI kwik bolts, hex nuts and flat washers.

The described modification does not increase the probability of occurrence or the consequences of an accident or malfunction of any safety related equipment as evaluated in the Safety Analysis Report. Therefore, an unreviewed safety question does not exist.

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i COMPLETED FACILITY CHANGE REQUEST FCR N0 81-013 SYSTEM Condensate and Demineralized Water Transfer and Storage COMPONENT SV 6831B CHANGE, TEST OR EXPERIMENT FCR 81-013 replaced solenoid valve SV 6831B with a solenoid valve which has a larger orifice.

FCR 81-013 was closed July 1, 1987.

REASON FOR CHANGE DW 6831B has periodically had trouble meeting the required stroke time of 10 seconds to isolate demineralized water to containment. Replacement of SV 6831B with a solenoid valve with a larter orifice reduces the stroke time of DW 6831B.

SAFETY EVALUATION

SUMMARY

Solenoid valve SV 6831B admits instrument air to actuate DW 6831B. The safety function of the solenoid valve is to open upon a Safety Features Actuation signal, venting the air to close DW 6831B.

DW 6831B is one of two containment isolation valves for the line supplying demineralized water to containment. The solenoid valve was replaced with a similar valve containing a larger orifice to reduce the stroke time of DW 6831B.

Replacement of solenoid valve SV 6831B does not increase the probability of occurrence or the consequences of an accident or malfunction of any safety related equipment as evaluated in the SAR.

Nor does it create the possibility for an accident different than any evaluated previously in the SAR, or reduce the margin of safety as defined in the basis for any l

technical specification. Therefore, an unreviewed safety question does

)

not exist.

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3 C0!iPLETED FACILITY CHANGE REQUEST FCR NO 86-0070 SYSTEM Various COMPONENT Various Solenoid Valves CHANGE, TEST OR EXPERIMENT FCR 86-0070 modified raceways to various solenoid valves by adding a.

' seal and drainage hole in the conduit.

FCR 86-0070 was closed July 1, 1987.

REASON FOR CHANGE 10 CFR 50.49 requires equipment be qualified for the environment in which it is installed. The modifications were required for the solenoid valves

.to be environmentally qualified.

SAFETY EVALUATION

SUMMARY

In order for the solenoid. valves to be considered qualified for their environment, qualified seals (Bisco LOCASEAL) were placed at the conduit i

entrance to the solenoids to prevent moisture intrusion into the solenoid L

coil.

In addition, weep holes were installed at a point in the connecting flexible conduit or conduit fitting to prevent any moisture accumulation

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at the seals. The table below lists the solenoid valves modified and

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their respective safety function:

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~I VALVE SYSTEM SAFETY FUNCTION l

I. SVICS11A1 Main Steam Failure of the solenoid valves SVICS11A2 Main Steam would prevent the steam generators

.SVICS11B1 Main Steam from being able to vent to atmosphere SVICS11B2 Main Steam when plant conditions warrant.

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II. SV 4633 Reactor Sample Provide isolation boundary between j

SV 4636 Reactor Sample Decay Heat System and Sampling System.

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Also provides for post LOCA sampling.

III. SV 4642 Reactor Sample None.

IV. SV 1467 Component Cooling Allow CCW flow through DHR coolers l

SV 1469 Water (CCW) on a Safety Features Actuation System (SFAS) signal.,

--L-L_------_.-_.-..

VALVE SYSTEM SAFETY FUNCTION V. SVDH 13A Decay Heat Removal (DHR) Allow full DHR flow through DHR SVDH 13B DHR coolers on a SFAS signal by SVDH 14A DHR opening cooler outlets and SVDH 14B DHR shutting cooler bypasses.

The described modificat. ion does not increase the probability of occurrence or the consequences of an accident or malfunction of any safety related equipment as evaluated in the Safety Analysis Report. Nor does it create a possibility for an accident different than any previously evaluated in the SAR, or reduce 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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COMPLETED FACILITY CHANGE REQUEST l

FCR NO 86-0319 l'

SYSTEM; Makeup and Purification l

COMPONENT:

MU-1A and MU-1B Ls

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' CHANGE, TEST OR EXPERIMENT 1

L FCR 86-0319 changed the time delay setting for the Letdown Cooler inlet valves, MU-1A and'MU-1B.

FCR.86-0319 was closed July 1, 1987.

REASON FOR CHANGE f

Due to interlock. circuitry design and. valve stroke time, the' ten second

' time delay did not maintain MU-1A or MU-1B in the closed position, nor close CC 1409 and CC 1410 (Letdown Cooler Component Cooling Water (CCW) inlet' valves). The purpose of the ' time delay is to prevent valve cycling.

SAFETY EVALUATION

SUMMARY

The safety function of valves MU-1A and MU-1B is to isolate their respective letdown cooler on high letdown fluid temperature. The coolers are isolated by stopping.the RC letdown flow followed by interrupting the cooler CCW flow. The closing control circuitry for MU-1A and MU-1B is designed such that if MD-1A or MU-1B is stroked fully closed, with their respective CCW inlet valve fully open, they will' reopen.

Consequently the setting of the time relay must be limited to a value which clears the full open indication'of the CCW inlet valve prior to full close indication of the letdown inlet valve. The time relay does not affect the function of.any other safety or non-safety related systems.

Surveillance procedures _(ST 5011.06, ST 5099.07) have demonstrated proper valve operation in the past. However, the tire relays were replaced and set to the" design setting, the as-found setting is not available. FCR 86-0319 changed the design setting to ensure proper valve operation.

Based on the above, an unreviewed safety question does not exist.

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COMPLETED FACILITY CHANGE REQUEST I

FCR No 85-0075 l

. SYSTEM Boric Acid Addition COMPONENT MU-346, MU-361 l

CHANGE, TEST OR EXPERIMENT FCR 85-0075 replaced check valves MU-346 and MU-361.

FCR 85-0075 was closed July 10, 1987.

FEASON FOR CHANGE The two check valves were allowing backleakage from the Makeup System into the Boric Acid Addition Tank (BAAT),

This leakage could cause the Boron concentration in the BAAT-to drop below the technical specification limit.

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SAFETY EVALUATION

SUMMARY

The Boric Acid Addition System is non-nuclear safety related and I

therefore has no safety function. The check valves were replaced with similar check valves with the exception of the piston movement. The new valves do not inhibit the flow requirements of the Boric Acid Addition Syrtem.

Replacement of the valves does not increase the probability of occurrence or the consequences of an accident or malfunction of any safety related equipment as evaluated in the Safety Analysis Report. Nor does this change create a possibility for an accident different than any evaluated previously in the SAR, or reduce the margin'of safety as defined in the basis for any technical specification. Therefore,.n unreviewed safety question does not exist.

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4 COMPLETED FACILITY CHANGE REQUEST FCR NO 86-0325

~ SYSTEM Startup Feed Pump and Auxiliaries-

' COMPONENT FW 106

. CHANGE, TEST OR EXPERIMENT FCR 86-0325 locked closed FW 106'which is a normally closed valve.

FCR 86-0325 was closed July 10, 1987.

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REASON'FOR CHANGE Locking closed FW-106 will ensure approximately 15 feet of non-seismic Startup_ Feed Pump.(SUFP) discharge piping from the SUFP discharge check valve'to the main feedwater header will not be inadvertently pressurized during modes 1, 2 or 3, to-a pressure greater than 275 psig.

SAFETY EVALUATION

SUMMARY

- 1 FW 106'will protect nuclear safety related Auxiliary Feedwater (AFW) equipment from damage by preventing the creation of an unar.alyzed moderate energy line within the AFW Room. The SUFP is electrically disabled and isolated and a new motor driven feed pump, located outside of the AFW room, has replaced the SUFP.

FCR 86-0325 does-not increase the probability of occurrence or the con-sequences of an accident or malfunction of safety related equipment as evaluated in the Safety Analysis Report. Nor does it create a possibility for an accident different than any previously evaluated in'the Sla, or f

reduce the margin of safety as defined in the basis for any technical l

specification. Therefore, an unreviewed safety question does not exist.

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e COMPLETED FACILITY CHANGE REQUEST-FCR NO-85-0292 I'

SYSTEM-Borated Water Storage Tank-C,0MPONENT BW 1525B, D/ Level Instrument Shelter t.

l CHANGE, TEST OR EXPERIMENT FCR 85-0292 improved insulation on BWST level transmitter source valves BW 1525B and BW 1525D. The FCR also insulated the enclosure around BWST level transe aters LT 1525A and LT 15250.

l-FCR 85-0292 was closed July 15, 1987.

REASON FOR CHANGE-Modifications were.made to provide greater assurance against freezing the BWST level instruments and' sensing lines.

l SAFETY EVALUATION SUMMAJR l

.The. Borated Water Storage Tank (BWST). stores borated water for emergency core cooling and safe shutdown of the plant. The Safety Features

Actuation System monitors tank level by level transmitters LT 1525A, B, C

'i and D. ~ Valves BW 1525B and BW 1525D are. isolation valves for level

. transmitters LT 1525B and LT 1525D. This.FCR upgraded the insulation on these valves to include the upper bonnet assemblies and handwheels.

.The insulation can be removed from LT 1525B and D, should'it.be necessary to operate these valves. This FCR also insulated a seismic enclosure which protects level transmitters LT 1525A and LT 1525C.

The addition of insulation on the level transmitter' source valves and the enclosure do not affect the Seismic I capability or safety grade integrity of the components.

Based on the above, an unreviewed safety question does not exist. _ _ _ _ _ _ _

9 COMPLETED FACILITY CHANGE REQUEST FCR NO 86-0062 SYSTEM Various COMPONENT Various Terminal Block Boxes CHANGE, TEST OR EXPERIMENT FCR 86-0062 drilled 1/4 inch diameter drain holes in the bottom of 125 terminal block boxes located in the auxiliary building.

FrR 86-0062 was closed July 15, 1987.

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REASON FOR CHANGE Drain holes were installed to prevent the possibility of condeni, ate or water from collecting in the terminal boxes. This makes the installed j

configuration the same as what was used in the Environmental Qualification tests.

SAFETY EVALUATION

SUMMARY

The Environmental Qualification analysis of Buchanan NQ0211, NQ0511 and Stanwick Type G terminal blocks were. performed exposing these blocks to a steam environment. The Buchanan blocks had a 1/4 inch weep hole and the Stanwick blocks were exposed to a steam spray.

In either case, water was not allowed to accumulate around the terminal blocks. The weep holes prevent moisture accumulation and comply with 10CRF50.49 which requires equipment be qualified for its environment.

Tlie safety function of the terminal block boxes is to provide physical protection to the blocks and cables.

The 1/4 inch diameter weep holes installed do not adversely affect the operation of the terminal blocks or the plant and satisfy 10CRF50.49 requirements.

I Based on the above, an unreviewed safety question does not exist.

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l COMPLETED FACILITY CHANGE REQUEST FCR NO 86-0124 I

l SYSTEM 125 VDC COMPONENT Jumpers 1, 2 and cable ICY 110A l

CllANGE, TEST OR EXPERIMENT FCR 86-0124 swapped the black and white wires of Jumper #1 and relabeled Jumper #1 and #2 in the field and on the drawing.

Cable ICY 110A was also relabeled on the drawing and in the field.

FCR 86-0124 was closed July 15, 1987.

REASON FOR CHANGE FCR 86-0124 corrected wiring discrepancies documented in NCR 86-0979.

SAFETY EVALUATION

SUMMARY

Safety related cable ICY 110A was terminated correctly but was labeled incorrectly both on the drawing and in the field. FCR 86-0124 changed the drawing and made the necessary field corrections.

This FCR also swapped the white and black wires of non-safety related Jumper #1 on the drawing and corrected the identification of non-safety related Jumpers #1 and #2 both on the drawing and in the field.

This FCR does not increase the probability of occurrence or the consequences of an accioent or malfunction of any safety related equipment as evaluated in the Safety Analysis Report.

The ciargin of safety as defined in any technical specification is not reduced.

Therefore, an unreviewed safety question does not exist. L__ _ _ _ _

COMPLETED FACILITY CHANGE REQUEST FCR NO 85-0184 SYSTEM Containment Vessel and Penetrations COMPONENT Modules D, G of P2L4G and F, G of PIL1L CHANGE, TEST OR EXPERIMENT YCR 85-0184 replaced Amphenol triaxial modules D, G of P2L4G and F, G of P1L1L with Conax triaxial modules P/N 7198-21011-01.

FCR 85-0184 was closed July 18, 1987.

REASON FOR CHANGE The modules vere replaced as part of a corrective measure for Rescart Action Plan No. 15 which was to correct the repeated inoperability of the Source Range Nuclear Instrument Channels.

SAFETY EVALUATION

SUMMARY

The safety function of a containment electrical penetration assembly is to carry several safety related circuits through the containment vessel while maintaining containment integrity under operating or accident conditions. FCR 79-401 purchased several modules from Conax Corporation for installation into electrical penetration assemblies manufactured by Amphenol when required for replacement due to maintenance or modifications. The Conax modules used to replace the Amphenol modules are functionally equivalent relative to electrical characteristics and sealing capability. This change will not affect the safety function of the electrical penetration assemblies.

1 Based on the above, replacement of the modules will not increase the probability of occurrence or consequences of an accident or the I

occurrence of a malfunction of any safety related equipment as analyzed or evaluated in the SAR. Therefore, an unreviewed safety question does not exist.

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COMPLETED FACILITY CHANGE REQUEST FCR NO 85-0344 SYSTEM Various COMPONENT Various Limitorque Actuators

. CHANGE, TEST OR EXPERIMENT FCR 85-0344 inspected and replaced suspect wires of Limitorque actuators not proven environmentally qualified.

FCR 85-0344 was closed July 18, 1987.

REASON FOR CHANGE Inspection of some NSR Limitorque actuators sbawed that in some cases wire qualification could not be proven.

SAFETY EVALUATION

SUMMARY

The safety function of the replacement wire is to ensure that the Limitorque operators will operate electrically, if required, in an environment equivalent to which they were environmentally qualified. FCR 85-0344 assured any unqualified wire was replaced with the proper qualified wire and provided a method of documenting and controlling environmental qualification of the wiring on Plant Q Limitorque operators.

The replacement of suspect wires with those that are qualified does not increase the probability of occurrence or consequences or the possibility for an accident or malfunction of any safety related equipment as previously evaluated in the Safety Analysis Report. Nor does it reduce 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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I COMPLETED FACILITY CHANGE REQUEST FCR NO 86-0089 l

SYSTEM Water Treatment, Degasifier COMPONENT PSH 6777 CHANGE, TEST OR EXPERIMENT FCR 86-0089 reconnected pressure switch PSH-6777 to the suction of the l

Degasifier Vacuum Pumps with stainless steel tubing.

l FCR 86-0089 was closed July 20, 1987.

l REASON FOR CHANGE The source tap for PSH-6777 was not added when the vacuum pumps were l

changed from Kinney pumps to Nash.

The installation allows remote indication of vacuum pressure.

SAFETY EVALUATION

SUMMARY

Pressure switch PSH-6777 provides remote annunciation when suction pressure for the degasifier vacuum pumps falls below the setpoint.

PSH-6777 does not perform a safety function.

FCR 83-028 replaced the vacuum pumps and modified the suction header. At that time PSH-6777 was inadvertently left disconnected. This FCR drilled a pressure tap on the suction header and installed stainless steel tubing from the tap to the pressure switch. This modification restored the system to its original design.

Based on the above, an unreviewed safety question does not exist. ;

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a COMPLETED FACILITY CHANGE REQUEST FCR NO 84-011 l

SYSTEM Containment Recirculation COMPONENT Technical Specification 3.6.4.2 l

CHANGE, TEST OR EXPERIMENT FCR 84-011 deleted references to the Containment Recirculation System (for hydrogen mixing) from the Technical Specifications and its basis.

It also removed reference to equipment and components associated with the Containment Recirculation System from the 'Q' List.

FCR 84-011 was closed July 21, 1987.

REASON FOR CHANGE The Recirculation System 4.s not required for hydrogen mixing since natural recirculation of containment atmosphere is sufficient and always available.

i SAFETY EVALUATION

SUMMARY

The original safety function of the Containment Recirculation System was

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to assure proper mixing of hydrogen in containment so localized hydrogen 1

burns would be prevented folloying a LOCA.

Since the design of j

Davis-Besse, additional research has shown that containment dome j

recirculation systems are no longer required and provide no safety j

function. A report (NUREG/CR-0304) from a U.S. DOE sponsored study provided detailed mathematical and experimental proof that natural recirculation of containment will be sufficient to ensure proper mixing of hydrogen.

The containment recirculating fans are used during normal operation to prevent temperature stratification of the air. This temperature stratification has caused grease to melt off from the polar crane.

This is an operational problem with.no safety significance. No other equipment requiring low temperatures is located in the top of containment.

l Based on the above, deletion of the Containment Recirculation System from the Technical Specifications and Plant "Qi List does not constitute an unreviewed safety question.

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9 COMPLETED FACILITY CHANGE REQUEST FCR-NO.

85-0106 SYSTEM Quench Tank Recirculation COMPONENT RC229A CHANGE, TEST OR EXPERIMENT FCR 85-0106 replaced SV 229A, solenoid valve for control air to RC229A actuator,.and the solenoid valve-to-actuator control air line.

FCR 85-0106 was closed July 21, 1987.

REASON FOR CHANGE RC 229A is required to close in ten seconds per Technical Specifications. This. valve had a history of stroking in nine seconds and

. was even declared inoperable on February 20, 1985 when it could not close in the required period of time.

FCR 79-311 replaced the solenoid valve for RC 229B and the valve has consistently achieved closure times less than the required ten seconds.

SAFETY EVALUATION

SUMMARY

The safety function of RC 229A, quench tank isolation valve, is to provide containment isolation. The safety function of SV 229A is to

- allow air bleed off from the actuator allowing RC 229A to close within the specified time period. This FCR replaced SV 229A with the same type valve except with.a larger diameter discharge. The modifications made are similar to that which was performed for RC 229B, the first containment isolation valve for the quench tank outlet. Response time testing has demonstrated RC 229B consistently meets the ten second f

closure requirement with the installation of a similar type solenoid valve. Post modification response time testing of RC 229A has confirmed the valve's ability to meet the specified closure time requirement.

Based on the above, it is concluded that modifications made by this FCR do not adversely affect the safety function of RC 229A. Therefore, an l

-unreviewed safety question does not exist.

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COMPLETED FACILITY CHANGE REQUEST FCR NO 86-0249 SYSTEM Intake Structure / Travelling Screen COMPONENT CDF12C CHANGE, TEST OR EXPERIMENT FCR 86-0249 lowered the curbing directly in front of cabinet CDF12C located in the Intake Structure.

FCR 86-0249 was closed July 21, 1987.

-REASON FOR CHANGE Due to curbing, the door on cabinet CDF12C was >nable to be opened fully. This presented a safety hazard when perxorming work inside the cabinet.

SAFETY EVALUATION

SUMMARY

The Intake Structure is a safety related structure, however, the concrete curb has no safety function.

The curb's function is to contain the oil from the Service Water pumps should there be a major leak.

FCR 86-0249 chipped and lowered the existing curb. The structural integrity of the Intake Structure was not affected by this change. The fire hazard analysis was not affected as each Service Water pump has its own curb around it.

The modifications made do not increase the probability of occurrence or consequences of an accident or malfunction of any safety related equipment as evaluated in the Safety Analysis Report. Nor do the modifications described create a possibility for an accident different than any evaluated previously in the SAR, or reduce 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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