ML20235N430
| ML20235N430 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 06/30/1987 |
| From: | Khazrai M, Storz L TOLEDO EDISON CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| KB87-00490, KB87-490, NUDOCS 8707200067 | |
| Download: ML20235N430 (29) | |
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e OPERATING DATA REPORT DOCKET NO. 3 0-346 DATE 7-14-87 COMPLETED BY Mort Khazrai TELEPlf 0NE 410-?44-5000 Ex't. 7290 OPERATING STATUS
- 1. Unit Name:
Dav f c:-Besse Urif t 1 Notes
- 2. Reporting Period:
. lune 1987
- 3. Licensed Thermal Power (MWt):
2772
- 4. Nameplate Rating (Gross MWe):
925 906
- 5. Design Electrical Rating (Net MWe):
- 6. Maximum Dependable Capacity (Gross MWe): 904 860
- 7. Maximum Dependable Capacity (Net MWe):
- 8. If Changes Occur in Capacity Ratings (Items Neunber 3 Through 7) Since Last Report, Give Reasons:
- 9. Power Level To Which Restricted,If Any (Net MWe):
- 10. Reasons For Restrictions.If Any:
This Month Yr.-to.Date Cumulative
- 11. Ilours in Reporting Period 7?O 4343 78239
- 12. Number Of Hours Reactor Was Critical 405.6 3317.2 39372.3
- 13. Reactor Reserve Shutdown Hours 0.0 143.9 4768.7
- 14. Hours Generator On-Line 386.6 3249 37737.6
- 15. Unit Reserve Shutdown liours 0.0 0.0 1732.5"~
- 16. Gross Thermal Energy Generated (MWH) 850,228 6710014 88136678-17, Gross Electrical Energy Generated (MWH) 271,644 2188336 29150/z.s
- 18. Net Electrical Energy Generated (MWH) 250,301 2027055 272fi3718
- 19. Unit Senice Factor s1 7 74.8 48.2
- 20. Unit Availability Factor 53.7 74.8 50.4
- 21. Unit Capacity factor (Using MDC Net) 40.4 54.3 40.5
- 22. Unit Capacity Factor (Using DER Net) 38.4 51.5 38.5
- 23. Unit Forced Outage Rate 33.9
.10.7 35.6
- 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 Startup:
- 26. Units In Test Status (Prior o Commercial Operation 1:
Forecast Achiesed INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCI A L OPER ATION 4
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. AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.
50-346 UNIT Davis-Besse 7~14-87' DATE COMPLETED BY Mort Khazrai TELEPHONE 419-249-5000 Ext. 7290-June 1987 MONTH-DAY AVERAGE DAILY POWER LEVEL DAY AVER AGE DAILY POWER LEVEL (MWe Net)
IMWe-Net) -
1 0
17 538
.2 0
'740 18 0
845 3
19 0
858 4
20 0
5 21 854 0
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'7 0
0 614 8
24 0
616 9
25 10 0
26 609 O
II 27 611 0
614 12 28 0
608 13 39
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13 I4 30 606 15-352 31 16 756
- INSTRUCTIONS On this format, list the average daily unit power level in MWe Net for each day in the reporting month. Compute to i the nearest whole negawatt.
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OPERATIONAL REPORT June 1987 s
The unit remained shutdown until 0223 hours0.00258 days <br />0.0619 hours <br />3.687169e-4 weeks <br />8.48515e-5 months <br /> on June 14, 1967, when the reactor was deborated to criticality.- The turbine generator was synchronized on line j
at approximately 1815 hours0.021 days <br />0.504 hours <br />0.003 weeks <br />6.906075e-4 months <br /> on June 14, 1987, marking the completion of the unit scheduled outage which began on May 8, 1987.
At 2316 hours0.0268 days <br />0.643 hours <br />0.00383 weeks <br />8.81238e-4 months <br /> cni June 14, 1987, while the unit was at approximately 24%, the turbine generator was manually tripped to perform turbine generator overspeed trip test.
The turbine generator was re-synchronized on line at 0223 hours0.00258 days <br />0.0619 hours <br />3.687169e-4 weeks <br />8.48515e-5 months <br /> on June 15, 1987.
The reactor power was increased to approximately 100% at 1300 hours0.015 days <br />0.361 hours <br />0.00215 weeks <br />4.9465e-4 months <br /> on June 16, 1987.
The reactor power was maintained at approximately 100% until 2050 hours0.0237 days <br />0.569 hours <br />0.00339 weeks <br />7.80025e-4 months <br /> on June 16, 1987, when Rod Group 7 Rod'10 (7-10) dropped into the core. A plant runback was initiated to approximately 47%. On June 17, 1987 at 0208 hours0.00241 days <br />0.0578 hours <br />3.439153e-4 weeks <br />7.9144e-5 months <br /> Rod 7-10 was repaired and-repulled, the reactor power was slowly increased l
to approximately 99%. Reactor power was maintained at approximately 99% until 2303 hours0.0267 days <br />0.64 hours <br />0.00381 weeks <br />8.762915e-4 months <br /> on June 22, 1987, when reactor power was reduced to approximately 73%. The. power reduction came as a result of an indication of a high temperature on an upper thrust bearing of reactor coolant pump 1-2.
The reactor power was maintained at approximately 73% power for the rest of the. month.
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REFUELING INFORMATION DATE: June 1987 1.
Name of facility: Davis-Besse Unit 1 2.
Scheduled date for next refueling shutdown: February 1988 3.
Scheduled date for restart following refueling: April 1988 4.
Will refueling or resumption of operation thereafter require a l
technical specification change or other license amendment? If answer is yes, what in general will these be? If answer is no, has the reload fuel design and core configuration been reviewed by your Plant Safety Review Committee to determine whether any unreviewed safety questions are associated with the core reload (Ref. 10 CFR Section 50.59)?
Ans: Expect the Reload Report to require standard reload fuel design Technical Specifications changes (2. Safety Limits and Limiting Safety System Settings, 3/4.1 Reactivity Control Systems, 3/4.2 Power Dis-tribution Limits and 3/4.4 Reactor Coolant System.)
5.
Scheduled date(s) for submitting proposed licensing action and supporting information: December, 1987 6.
Important licensing considerations associated with refueling, e.g.,
new or different fuel design or supplier, unreviewed design or performance analysis methods, significant changes in fuel design, new operating procedures.
Ans: None identified to date.
7.
The number of fuel assemblies (a) in the core and (b) in the spent fuel storage pool.
(a) 177 (b) 204 - Spent Fuel Assemblies 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.
BMS/005
l f-COMPLETED FACILITY CHANGE REQUEST FCR N0'
'85-0045 SYSTEM Diesel Generator and Auxiliaries COMPONENT PCV 2987, PCV 2988, PCV 2989 and PCV 2994 CHANGE, TEST OR EXPERIMENT FCR 85-0045 replaced PCV 2987 and PCV 2988 on Emergency Diesel Generator (EDG) #1, and PCV~2989 and PCV 2994 on EDG #2.
FCR 85-0045 was closed March 17, 1987.
REASON FOR CHANGE Numerous manhours had been expended rebuilding and repairing the regulators
'and their associated pilots. The regulators were obsolete and could not be l
replaced with a like model. No parts were available other than rebuild kits.
SAFETY EVALUATION
SUMMARY
This change replaced the existing air pressure regulators in the Diesel Generator Air Start System with new pressure regulators capable of maintaining a pressure of approximately 180 psig to the diesel generator air start motors. The new regulators equal or exceed the old regulators both in design and functional capability.
The design and safety functions of the Diesel Generator Air Start System remain unchanged.
This change does not create the probability for an accident of a different-type than any evaluated previously in the Safety Analysis Report (SAR),
This change does not involve an unreviewed safety question.
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COMPLETED FACILITY CHANGE REQUEST FCR NO 81-0054 SYSTEM Service Water System COMPONENT Backup Service Water Pump CHANGE, TEST OR EXPERIMENT FCR 81-0054 added a new service water pump, which does not require offsite power, and cross-connection piping, with manual valving to the existing l
Service Water System.
FCR 81-0054 was closed February 2, 1987..
REASON FOR CHANGE.
This modification complied with the alternate shutdown capability requirements defined in'10CFR50 Appendix R, Fire Protection.
SAFETY EVALUATION
SUMMARY
FCR 81-0054 provided a backup service water pump which does not require offsite power. The purpose of this modification is to satisfy the. separation
-with fire barrier criteria relating to fire protection for safe shutdown capability.
l The wor'r authorized by FCR 81-0054 did not degrade the Service Water System. The addition of the service water backup pump, piping, and valves enhanced the capability of the Service Water S'jstem. This change provided the necessary separation' criteria to comply with the requirements of 10CFR50 Appendix R.
l Therefore, based on the above, an unreviewed safety question does not exist.
COMPLETED FACILITY CHANGE REQUEST FCR NO 82-0011
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SYSTEM Core Flood Tanks COMPONENT N/A CHANGE, TEST OR EXPERIHENT FCR 82-0011 changed the Core Flood Tank (CFT) temperature minimum limit from 70*F to 50*F.
FCR 82-0011 was closed May 7, 1987.
REASON FOR CHANGE This change was made due to the difficulty in maintaining minimum core flood tank temperature during the winter months while the reactor is shutdown.
SAFETY EVALUATION
SUMMARY
Analysis by B&W allowed a reduction in the minimmn pressurization tempera-ture limit for core flood tanks from 70*F to 50 F..
By demonstrating that no cracks are initiated at the new minimum temperature, the B&W analysis also concluded that the minimum NDTT is +10*F.
Therefore, lovering the minimum pressurization temperature for the CFT's from 70 F to 50*F and subsequent revision to the Technical Specifications does not create an unreviewed safety question.
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COMPLETED FACILITY CHANGE REQUEST FCR NO 79-357.
SYSTEM Anticipatory Reactor Trip System (ARTS)
COMPONENT Turbine Trip Input Signal CHANGE, TEST OR EXPERIMENT FCR 79-357 modified the main turbine trip by providing a parallel contact to XKT-1182.and' associated wiring to sense a turbine' trip from a loss of Electro-Hydraulic Control (ERC) oil pressure.
FCR 79-357 was closed February 23, 1987.
REASON FOR CHANGE This modification will temporarily satisfy NRC Commitment #540 dated October'3, 1979 until a safety grade ARTS can be procured and installed.
SAFETY EVALUATIO4
SUMMARY
FCR 79-357 upgr4ded the ARTS by providing a sensing signal from loss of EHC oil pressure..This change will provide redundancy to the ARTS system without degrad ng the existing safety systems. FCR 79-357 will not i
increase the probability of occurrence or the consequences of an accident or malfunction of safety related equipment as e<eluated in the Updated-
. Safety. Analysis Report. An unreviewed safety question does not exist.
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. COMPLETED FACILITY CHANGE REQUEST FCR NO 85-0302 1
SYSTEM Reactor Coolant Letdown COMPONENT HOV MU-11 CHANGE, Tf.ST OR EXPERIMENT FCR 85-0302 changed MOV MU-11 wiring to allow the valve to torque out in the open direction rather than shutting off on' limit switch signal.
This FCR 85-0302 was. closed May 6, 1987.
REASON FOR CHANGE MOV MU-11 did not seat tightly when utilizing the limit switch in the open direction.
SAFETY EVALUATION
SUMMARY
Valve-MU-11 is a double poppet globe valve designed to close tightly in both the open and closed directions. Ensuring MU-11 will close tightly in the open direction is the correct method of operation, and will add reliability and safety to the system. The margin of safety as defined in the bases for any Technical Specification is not reduced.
Therefore an unreviewed safety questien is not involved.
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COMPLETED FACILITY CHANGE REQUEST FCR NO 84-0178 SYSTEM Reactor. Coolant Sys.
COMPONENT N/A CHANGE, TEST OR EXPERIMENT FCR 84-0178 rerouted cable ICBE1602F for the PORV Block Valve (RC 11),
to separate it from the PORV (RCZA) circuit.
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FCR 84-0178'was closed April 2, 1987.
REASON FOR CHANGE This modification was implemented in order to comply with electrical separation requirements of 10CFR50 Appendix R.
(Ref. Appendix R Compliance Assessment Report No. 02-1040-1153 Rev. O Page 4-97.)
SAFETY EVALUATION
SUMMARY
The safety function of Cable ICBE1602F provides circuitry for PORV Block Valve RC 11.
The. change rerouted cable ICBE1602F through fire area V (Room 401) where it will not impact safe shutdown.
There will be no change in the safety function or the cable. This change will not create an adverse environment. The operability of the PROV Block Valve RC 11 will not be limited or effected.
An unreviewed safety question does not exist.
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7 COMPLETED FACILITY CHANGE REQUEST FCR NO 80-0030 SYSTEM RPS and NI's C0?fPONENT.
NIM BIN CHANCE,. TEST OR EXPERIMENT FCR 80-0030, installed NIM BIN on NI-1/or NI-2 per instruction from B&W in order.to provide audible indication of neutron count rate from the control room to containment.
FCR 80-0030 was closed April 7,1987.
REASON FOR CHANGE FCR 80-0030 was implemented in order to comply with Refueling Operations Technical Specification 3-9.2.
SAFETY EVALUATION
SUMMARY
FCR 80-0030 provided a temporary connection from a RPS cabinet to a-counter scaler for' the purpose of providing an audible indication of increasing neutron level in containment during refueling. The installation only exists at a time when the reactor is tripped and will not effect the operation of the RPS which is not required in Modes 5 or 6.
Based on the above.no unreviewed safety question exists.
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COMPLETED FACILITY CHANGE REQUEST i
L FCR NO 85-0057 SYSTEM Emergency Ventilation System COMPONENT
.Various CHANGE, TEST OR EXPERIMENT FCR 85-0057 modified the electrical conduit supports for the emergency l
ventilation system.
This FCR 85-0057 was closed March 16, 1987.
REASON FOR CHANGE j
The modifications were required in order to compensate the rigid conduit
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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 85-0057 modified conduit supports by replacing existing 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 Updated Safety Analysis Report.
Therefore, an unreviewed safety question does not exist.
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i COMPLETED FACILITY CHANGE REQUEST L-1 FCR NO 85-0058 SYSTEM Auxiliary Feedwater System 1
COMPONENT Various 1
CHANGE, TEST OR EXPERIMENT
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FCR 85-0058 modified the electrical conduit supports for the auxiliary feedwater system.
This FCR 85-0058 was closed March 16, 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 i
from the adjacent seismic zones due to the relative displacements of the l
buildings and differences in' seismic zones.
In order to upgrade the l-supports to meet the long term operability requirements, FCR 85-0058 added a vertical bracing to the conduit supports. 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 Updated Safety Analysis Report. Therefore, an unreviewed safety question does not exist.
30 COMPLETED FACILITY CHANGE REQUEST FCR NO 84-0189 SYSTEM Emergency Diesel Generators COMPONENT K-005-01 and K-005-02 i
CHANGE, TEST OR EXPERIMENT FCR 84-0189 provided the following changes to the Emergency Diesel Generators'(EDG).
1.
Permits manual control of the hy.fraulic governor upon loss of electronic governor.
2.
Corrects the actuation of the low fuel oil pressure alarm during idle speed operation.
3.
Corrects the actuation of the over frequency alarm caused by engine acceleration during emergency starts.
FCR 84-0189 was closed February 25, 1987.
REASON FOR CHANGE To correct Emergency Diesel Generator operational problems.
SAFETY EVALUATION
SUMMARY
The EDG cont.rol circuitry did not al3ow the Operator to lower the speed of the diesel via the governor control switch unless wires were lifted. FCR 84-189 installed a governor mode selector switch which permits manual control of the hydraulic governor when the electronic governor is inoperable.
FCR 84-0189 does not increase the probability.of occurrence or the conse-quences of an accident or malfunction of safety related equipment as evaluated in the Updated Safety Analysis Report. Therefore, an unreviewed safety question does not exist.
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COMPLETED FACILITY CHANGE REQUEST
.FCR NO 83-0095 SYSTEM Auxiliary Feedwater COMPONENT Woodward Governors CHANGE, TEST OR EXPERIMENT FCR 83-0095 modified the Woodward PG-PL auxiliary feedwater pump governors i
as follows:
2 1.
Added independently adjustable high and low mechanical stops.
l 2.
Installed new design slip clutch.
3.
Added dowel pins to eliminate run out between the motor and clutch.
FCR 83-0095 was closed March 30, 1987.
REASON FOR CHANGE This change was necessary due to problems with the Woodward governors jamming on the high and low stops and failure to hold torque settings during operations. Modifications to the governors improve reliability of this safety grade system.
SiFETY EVALUATION
SUMMARY
The modification accomplished the following changer:
i 1.
Eliminated the internal governor linkage binding by installing independently adjustable high and low speed mechanical stops. This enables each stop to be adjusted for maximum contact with the respective roll pins.
2.
The speed clutch reliability was improved by installing a redesigned slip clutch.
The above' listed modifications increase the reliability of the AFW pump governors. -This improves the overall reliability of the AFW system, thereby better ensuring it fulfills its safety function.
Based on the above it is concluded that the change proposed by FCR 83-0095 does not constitute an unreviewed safety question.
' COMPLETED FACILITY CHANGE REQUEST FCR NO 79-0293
. SYSTEM.
Auxiliary Feedwater System COMPONENT Pressure Switches CHANGE, TEST OR EXPERIMENT FCR 79-0293 replaced the " Static 0 Ring" pressure switches on the Auxilia-ry Feedwater. System with new " Static 0 Ring" pressure switches which have l
stainless steel diaphragms and stainless steel pressure ports.
j FCR 79-0293 was closed March 20, 1987.
REASON FOR CHANGE Investigation of the drift and failure problems with the " Static 0 Ring" pressure switches revealed the cause to be corrosion of the piston due to permeation of condensate (with ammonia treatment) through the "Buna-N" diaphragm onto.the piston.. Replacement of the " Static 0 Ring" pressure i
switches with stainless steel " Static 0 Ring" pressure switches corrected the drift and failure. problems.
SAFETY EVALUATION
SUMMARY
The new " Static 0 Ring" pressure switches, with stainless steel diaphragms and ports, have the same pressure ratings and setpoints as the origina!'
" Static 0 Ring" pressure switches. The new switches reset at a higher pressure due to their higher dead band. This higher dead band does not degrade the safety function of the Auxiliary Feedwater System, as the normal operating pressure on the switches is much higher then their trip setpoint.
This modification does not increase the probability of occurrence or the
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consequences of an accident or malfunction of any safety related equipment as evaluated in the Updated Safety Analysis Report. Based on the above, no unreviewed safety question exists.
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COMPLETED FACILITY CHANGE REQUEST FCR NO 86-0053 SYSTEM Auxiliary Bldg. CTRM Air COMPONENT SV 5301 & SV 5311 CHANGE, TEST OR EXPEP.IMENT FCR 86-0053 replaced solenoid velves SV 5301 and SV 5311 with nuclear qualified ASCO valves.
FCR 86-0053 was closed May 14, 1987.
REASON FOR CHANGE Replacement of solenoid valves JV 5301 and SV 5311 was necessary due to failure and unavailability of materials.
ASCO valves #NP8316A75E are
- acceptable replacements for the existing valves.
' SAFETY EVALUATION
SUMMARY
The safety function of the Emergency Ventilation System is to ensure a suitable environment for plant equipment and to ensure station operator comfort and safety.
Isolation of the normal ventilation system precludes the admission of airborne contaminants into the control room and is accomplished by deenergizing solenoids which actuate isolation dampers.
The bleed-off function for the two redundant trains is performed by solenoid valves SV 5301, SV 5301A, SV 5311 and SV 5311A. A pair of solenoid valves is used'in parallel for each train to meet the response time requirements for isolation of each train of control room air.
The installation of ASCO model NP8315A75E exceeds the system requirements for IEEE Class IE solenoid valves, seismic and environmental qualification
. and overall system response time.
This 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 Updated Safety Analysis Report. Therefore, an unreviewed safety question does not exist.
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COMPLETED FACILITY CHANGE REQUEST FCR NO 86-0065 SYSTEM' CTMT Purge COHPONENT SV 5004 CHANGE, TEST OR EXPERIMENT FCR 86-0065 installed a union between solenoid valve SV 5004 and the actuator of CV.5004.
FCR 86-0065_ was closed May 18, 1987.
REASON FOR CHANGE This modification was necessary as solenoid valve SV 5004 could not be removed for repairs due to interference between the valve body and a shield protecting the valve.
SAFETY EVALUATION
SUMMARY
FCR 86-0065 added a union between solenoid valve SV 5004 ' nd CV 5004.
a This permits removal of the solenoid for maintenance. The consequences of an accident previously evaluated in the SAR are not increased as a result of-this change. The probability of a malfunction of equipment important to safety previously evaluated in the SAR is not. increased as a result of this modification. The installation of a union in the air supply line between SV 5004 and CV 5004 does not involve an unreviewed safety question.
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p COMPLETED FACILITY CHANGE REQUEST.
FCR NO-86-0064
. SYSTEM 13'.8 KV System COMPONENT 81-1 81-2/HA05 (HB05)
A(B) A(B) 1 CHANGE, TEST OR EXPERIMENT
[
FCR.86-0064-changed the setpoint of the underfrequency relays from 2 cycles per second to 20 cycles per second.
j This FCR 86-0064 was closed June 6, 1987.
REASON FOR CHANGE The underfrequency relays are causing failure of fast transfer of 13.8 KV bus from one startup transformer to another because of a time delay of 2 cycles. This cycle setpoint is been changed to 20 cycles which will alleviate the problem.
SAFETY EVALUATION
SUMMARY
FCR 86-006'rallows fast transfer from one startup transformer to another for all modes of plant' operations, and allows testing of the transfer scheme under plant shutdown conditions. The modification of the 13.8 KV bus underfrequency relays from'2 cycles to 20 cycles does not increase the probability of occurrence of an accident and does not increase the conse-quences of an accident.
It does not increase the probability of occur-rence of a malfunction of equipment important to safety as defined in the bases for any Technical Specification.
Based on the above an unreviewed safety question does not exist.
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4 COMPLETED FACILITY CHANGE REQUEST FCR NO 82-063 SYSTEM Waste Gas System COMPONENT-GE D3060K13-509 SH 4 of 18 i
CHANGE, TEST OR EXPERIMFNT l
- FCR 82-063 incorporated.the "as-built" configuration for radwaste gas flow to the station vent.
' FCR 82-063 was closed June 1, 1987.
REASON FOR CHANGE FCR 82-063 updated GE print D3060V13-509 SH 4 of 18 to reflect actual configuration.
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SAFETY EVALUATION
SUMMARY
i FCR 82-063 was issued to incorporate the "as-built" configuration for radwaste gas flow to the station vent. Voltage dividers that were re-
~
quired te properly match components in the flow measurement / control loops were not' shown.on early vendor prints.
Updating _the appropriate drawings will not increase the probability of l
occurrence or the consequences of an accident or malfunction of any safety related equipment as evaluated in the Updated Safety Analysis Report. An unreviewed safety question does not exist.
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1 COMPLETED FACILITY CHANGE REQUEST l
I FCR NO 85-0242 Rev. A l
SYSTEM Decay Heat and Low Pressure Injection l
COMPONENT' FTDH2A and FTDH2B l
CHANGE, TEST OR EXPERIMENT
. FCR 85-0242 replaced the amplifier circuit board on flow transmitters FTDH2A and FTDH2B.
I FCR 85-0242 was closed June 1, 1987.
REASON FOR CHANGE The relocation of the steam admission valves necessitated the change out of the transmitter circuit boards, as a result of new high energy line break analysis.
SAFETY EVALUATION
SUMMARY
. FCR 85-0242 replaced the amplifier circuit _ boards and calibration circuit boards in low pressure injection flow transmitters FTDH2A and FIDH2B.
The amplifier circuit boards and calibration circuit boards were replaced in the existing flow transmitters due to thermal aging conr.iderations addressed in the environmental qualification program. The flow transmit-ters with the new boards are equivalent to the previous transmitters.
This change does not affect the safety function of the flow transmitters.
This modification does not increase the probability of occurrence or consequences or the possibility of an accident, or the occurrence of a malfunction of equipment to safety as previously analyzed or evaluated in the SAR.
Based ou the foregoing consideration, it is concluded that this modifica-tion does not involve an unreviewed safety question.
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. COMPLETED FACILITY CHANGE REQUEST.
FCR K0 86-0057 SYSTEM Low Pressure l Injection System COMPONENT Anchor Bolts, Shims and Grout
' CHANGE, TEST OR EXPERIMENT FCR 86-0057 removed the 1/2" shell anchor bolts and replaced them with 3/4" shell anchor. bolts and 3/4" - 1 1/2" tap bolts.
It also provides for.
shims and grout behind the base plate.
FCR-0057 was closed June' 11, 1987.
REASON FOR CHANGE Deficiencies with the 1/2" shell anchor bolts were identified while performing MWO 1-86-0096-00 in accordance with disposition of NCR 85-0660.
SAFETY-EVALUATION
SUMMARY
The. repair to this pipe support consisted of replacing 1/2" diameter shell anchors with 3/4" diameter shell anchors and grouting or installing between the base. plate.and concrete..Since the repair is intended to assure that the supports meet the Safety Analysis Report allowables upon completion of repair, there will be no adverse affects on the LPI system piping. Therefore an unreviewed safety question does not exist.
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C_0MPLETED FACILITY CHANGE REQUEST FCR NO 83-0141 SYSTEM-Service Water System COMPONENT LSH 2805, LSH 2806 and LSE.2807 CHANGE, TEST OR EXPERIMENT FCR B3-0141 disconnected the containannt air cooler leak detectors and alarms for LSH 2805, LSH 2806 and LSH 2807. FCR 83-0141 also removed and plugged the 1/8" normal condensate drain line.
FCR 83-0141 was closed June 15, 1987.
REASON FOR CHANGE This system has been a nuisance alarm and in continuously alarming since the condensate drain line was reduced to 1/8".
This method for determining service water leakage through the CTMT air coolers has not been practical.
SAFETY EVALUATION
SUMMARY
l The containment air cooler leak detection system is not a required system and due to plugged tubes has served no function since startup. The i
monitoring requirements per criterion 30 of the general design criteria l
have been satisfied. Based on this, the removal of the containment air cooler leak detection system will not increase the probability of occur-rence or the consequences of an accident or malfunction of any safety related_ equipment'as evaluated in the Updated Safety Analysis Report.
This modification will not create a possibility for an accident different than any evaluated previously in the Updated Safety Analysis Report, or l
reduce the margin of safety as defined in the basis for any Technical Specification. 'An unreviewed safety qucstion does not exist.
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COMPLETED FACILITY CHANGE REQUEST FCR NO 85-0319 SYSTEM High Pressure Injection
. COMPONENT Support Anchor A-190 CHANGE, TEST OR EXPERIMENT FCR_85-0319 repaired High Pressure Injection support anchor A-190 by adding shim plates to restore support functions.
FCR 85-0319 was closed June 10, 1987.
REASON FOR CHANGE FCR 85-0319 was incorporated to correct the deficiencies identified by NCR 85-0257.
SAFETY EVALUATION
SUMMARY
Support A-190 is. located on piping in the High Pressure Injection (HPI) system. The items identified in this NCR are being repaired to meet the original design requirement for the support. The repair of this support will'not increase the probability of occurrence, consequences of an accident or malfunction of safety related equipment as evaluated in the Updated Safety Analysis Report.
This repair does not reduce the margin of' safety as defined in the basis for any Technical Specification. An unreviewed safety question does not exist.
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COMPLETED FACILITY CHANGE REQUEST FCR NO 85-0010 SYSTEM Auxiliary Feedwater System i
COMPONENT Support 6C-EBB-4-H11
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1 CHANGE, TEST OR' EXPERIMENT FCR 85-0010 modified support 6C-EBB-4-H11 located on AFW piping to SG 1-2 by:
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Rotating the two (2) snubber end brackets 90.
2.
Reworking support items 2, 7x8 to permit snubber position in the hot condition to be horizontal.
l FCR 85-0010 was closed June 11, 1987.
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REASON FOR' CHANGE The D1334 outage inspection revealed a non-conforming condition with resped'4 to the design drawings.
SAFETY EVALUATION
SUMMARY
Support 6C-EBB-4-H11 was inspected by engineering personnel on 1/11/85, "i
and the actual discnsions recorded. From this information engineering 4 3 reviewed system / plant operation'with the support"lu.the ab-found condition
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and found it to be acceptatle for interim operation (BT-15140). For long term operation minor modifications were required. These modifications complied with the original support design criteria and allows long term plant operation.
This modification does not increase the probability of occurrence, the consequences of an accident or malfunction of safety related equipment as evaluated in the Updated Safety Analysis Report.
This modification does not reduce the margin of safety as defined in the basis for any Technical Specification. Therefoie, an unreviewed safety question does not exist.
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COMPLETED FACILITY CHANGE REQUEST FCR'N0' 85-0150 SYSTEM Diesel Generator and Auxiliaries COMPONENT Diesel Generators 1-1, 1-2 CHANGE, TEST OR EXPERIMENT FCR 85-0150 reviewed and updated Opdated Safety Analysis Report (USAR)
Table 8.3-1 (Loading of Emergency Diesel Generator during LOCA coincident with loss of offsite power.)
FCR 85-0150 was closed February 23, 1987.
REASON FgR CHANGE This change to USAR Table 8.3-1 was completed to accurately reflect plant as-built conditions pertaining to the loading of Emergency Diesel Generators (EDG) 1-1 and 1-2.
SAFETY EVALUATION
SUMMARY
FCR 85-0150 reviewed the (USAR) Table 8.3-1.
This change updated the USAR to accurately reflect the plant as-built condition. The safety function of the affected section is to ensure that the loads required for ope. ration of each safety train during the unlikely event of LOCA condi-tions are less than the continuous rating of the EDG for each train. The above changes do not create any new adverse environment and do not cons-titute an unreviewed safety question.
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,*s COMPLETED FACILITY CHANGE REQUEST FCR NO 84-0220 SYSTEM-High Pressure Injection System COMPONENT H.P.I. Check Valves'HP-48, HP-49, HP-50, HP-51, HP-56, HP-57, HP-58 and HP-59 CHANGE, TEST OR EXPERIMENT FCR 84-0220 performed the HPI Check Valves Leak Test using leak measuring devices (LMD) and acoustic emission devices. This testing is in addition to the requirements of the Inservice Inspection (ISI) Valve Program.
FCR 84-0220 was closed February 4,1986.
REASON FOR CHANGE In response to the ISI Safety Evaluation Report received from the NRC, on i
May 1984 (Log 1521) Toledo Edison was to determine the condition of each check valve as part of the additional testing. Since individual testing is not possible, the valves will be tested in pairs.
SAFETY EVALUATION
SUMMARY
The test to the HPI System is an Inservice Inspection of the Class 1 piping and valves in Containment as committed to the NRC.
It was con-ducted by providing a measurable leakage path through drain valves to floor drains. The test involved adding an LMD with a pressure regulator and a gauge to hold a 10 to 20 psig pressure on the drain. The piping for the LMD has a restriction orifice capable of limiting flow to 5 gpm at full Reactor Coolant System (RCS) pressure. Also during the time of the test, an operator was on standby to close the drain valves in case of HPI actuation. There was no' increase in the probability of an accident previously evaluated in the Safety Analysis Report. There was no decrease in the margin of safety as defined in the basis for any Technical Specifi-cation. Therefore, the testing performed.above did not constitute an unreviewed safety question.
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TOLEDO EDISON EDISON PLAZA 300 MADISON AVENUE TOLEDO. OHIO 43652-0001 July 14, 1987 l
Docket No. 50-346 KB87-00490 License No. NPF-3 File: RR 2 (P-6-87-06)
Document Control Desk U. S. Nuclear Regulatory Commission Washington, D.C.
20555 Gentlemen:
Monthly Operating Report, June 1987 Davis-Besse Nuclear Power Station Unit 1 i
Enclosed are ten copies of the Monthly Operating, Report for Davis-Besse Nuclear Power Station Unit 1 for the month of June 1987.
If you have any questions, please feel free to contact Morteza Khazrai at (419) 249-5000, Extension 7290.
Very truly yours,
/
Louis F. Storz Plant Manager Davis-Besse Nuclear Power Station LFS/MK/jmh Enclosures cc:
Mr. A. Bert Davis, w/l Regional Administrator, Region III Mr. Paul Byron, w/l NRC Resident Inspector Nuclear Records Management, Stop 3220 9(
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