ML20215C188
| ML20215C188 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 05/31/1987 |
| From: | Khazrai M, Storz L TOLEDO EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| KB87-00187, KB87-187, NUDOCS 8706180018 | |
| Download: ML20215C188 (12) | |
Text
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AVERAGE Dally UNIT POWER LEVEL 50-346 DOCKET NO.
v337. Davis-Besse Unit'1:
DATE June 12. 1987 Morteza Khazrai COMPLETED 8Y TELEPHONE d19-249-5000. Fxt.
ido ay, N
. MONTH i
DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe Net) -
IMWe Net).
875 0
t
,7 2
875 0
18 879
'O 3
19 877 20' 0
q 4
L 878 0
s 21 6
873 0
22 7
860 0
.]
23 74 0
8 24 0
0 9
'S t
~
Y 0
0 10 26 i
0 0
ll 27 i
0 0
12 28 0
0 13
.29 0
0 14 30 0
0 15 31 0
'l 16 4
INSTRUCTIONS On this format, list the average daily unit power level in MWe Net for each day in the reporting month. Compure t the nearest whole megawatt.
i 19177) 8706180018 B70531 PDR ADOCK 05000346 R
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l OPERATING DATA REPORT -
50-M6.
1 DOCKET NO.
DATE dune u,1987 COMPLETED BY norteza Khazrai TELEPHONE 419-249 4000,-
'j OPERATING STATUS Ext. 7290'-
1 4
s Davis-Besse linit 1 Notes
- 1. Unir Name:
- 2. Reporting Period:
MaV,.1987
- 3. Licensed Thermal Power (MWt):
2772
- 4. Nameplate Rating (Gross MWe):
925
- 5. Design Electrical Rating (Net MWe):
906-
- 6. Maximum Dependable Capacity (Gross MWe):
904-
- 7. Maximum Dependable Capacity (Net MWe):
860
- 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,if Any:
This Month Yr.-to-Date Cumulative I1. Hours in Reporting Period 744 3,623.'
77,519
- 12. Number Of Hours Reactor Was Critical 176.5 2,911.6 38,966.7
- 13. Reactor Reserve Shutdown Hours 0.0-143.9 4,768.7
- 14. Hours Generator On Line 174.2 2.862.4 37,351-
- 15. Unit Reserve Shutdown Hours 0.0 0.0 1,732.5
- 16. Gross Thermal Energy Generated (MWH) 471.018-5.859.786 87.286,450
- 17. Gross Electrical Energy Generated (MWH) 156,382-1,916,692 28,879,079
- 18. Net Electrical Energy Generated (MWH) 140,766'_,
1,776,754 27,013.417 i
- 19. Unit Service Factor 23=4 8.U 48.2
- 20. Unit Availability Factor 23.4 79.0 50.4
- 21. Unit Capacity Factor (Using MDC Net) 22.0 57.0 40.b
- 22. Unit Capacity Factor (Using DER Net)
'20.9 54.1 38.5
- 23. Unit Forced Outage Rate 0
6.2 35.6
- 24. Shutdowns Scheduled Over Next 6 Months (Type. Date, and Duration of Each t:
- 25. If Shut Down At End Of Report Period. Estimated Date of Startup:
June 13. 1987
- 26. Units in Test Status (Prior to Commercial Operation):_
Forecast :
. Achieved; INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERATION (Q/77)
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REFUELING INFORMATION DATE:.May 1987 l
i 1.
Name of facility: Davis-Besse Unit 1 f
2.
Scheduled date for next refueling shutdown:
February 1988 s
3.
Scheduled date for restart following refueling: April'1988 4.
Will refueling or resumption of operation thereafter require a
]
technical specification change or other license amendment?
If' answer j
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.,
j new or different fuel design or supplier, unreviewed design or 1
performance analysis methods, significant changes in fuel design, new operating procedures.
Ans: None identified to date.
.j 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
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9 OPERATIONAL'
SUMMARY
May 1987~
.]
Reactor power was maintained at approximately 100% power'until 2155 hours0.0249 days <br />0.599 hours <br />0.00356 weeks <br />8.199775e-4 months <br /> on May.7, 1987, when power reduction resumed for a planned maintenance outage.
The turbine-generator:was taken off line at 0613 hours0.00709 days <br />0.17 hours <br />0.00101 weeks <br />2.332465e-4 months <br /> on May 8, 1987.
i
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Reactor power was reduced to 0% power _ at 0830 hours0.00961 days <br />0.231 hours <br />0.00137 weeks <br />3.15815e-4 months <br /> on May 8. -1987, and -
)
the unit remained shutdown the rest of the month.'
The following are the more significant. outage activities performed during-
?
the outage:
.i 1)
Replaced Pressurizer Code Safety Valves 2)
Replaced valve RC-11' 3)
Repaired the Motor Driven Feedwater Pump
,)
4)
Replaced the. Power Operated Relief Valve (PORV)?
5)
. Repaired Core Flood Check Valve CF-30 L
a6)
Replaced all-four Reactor Coolant Pump. seals j
7)
Repaired all six Turbine Bypass Valves-(TBVs)'
8).
Adjusted packing or repaired packing on 35 manual ReactorL Coolant System valves 9)
Replaced Auxiliary Feed Pump #2 bearing.
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COMPLETED FACILITY' CHANGE REQUEST-FCR No 85-0174~
SYSTEM-Emergency Diesel Generator (EDG) Fuel Oil-Storage Tank High Level Alarm-
' COMPONENT-LSH 4891 and LSH 4892
-CHANGE,-TEST OR EXPERIMENT.
This FCR 85-0174 deleted the high level' switches LSH 4891 and LSH_4892.
,j from the circuit of the Control Room annunciator alarm and computer alarm-q points L389.and L394. '(EMERG DG FOST.1-1 HI/LO-and EMERG DG FOST LVL~
l HI/LO.)
I This FCR 85-0174 was closed March 2, 1987.
REASON FOR CHANGE The annunciator alarms L389 and L394'were often in alarm condition due to full storage tanks. Modifying the alarm circuit by.. removing the high level switch input resulted in the alarm actuating only on low. level which is the true safety-concern.
SAFETY EVALUATION
SUMMARY
The deletion of the high level input from th'e circuit' removed the distrac-tion created by the annunciator being in the alarmed state. This will'-
improve the operators' ability to recognize the valid alarm concerning low J
Fuel Oil Storage Tank' level.
1 This change does not increase the probability of occurrence or consequenc -
es of an accident or malfunction of any safety related' equipment as-evaluated in the USAR. Therefore, an unreviewed safety question does not exist.
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a COMPLETED FACILITY CHANGE REQUEST l
l FCR NO 34-132 T
SYSTEM
' Miscellaneous Control Switches COMPONENT Listed Below j
CHANGE, TEST OR EXPERIMENT FCR 84-132 performed.the following modifications:
1.
Added flip guards on reactor trip pushbuttons HSN1-45 and-HSN1-46.
2.
Added flip guards on main feed pump trip pushbuttons HS-797 and HS-798.
.i 3.
Added guard on turbine trip pushbutton on EHC console.
4.
Added slide guards on SFAS pushbuttons HIS-2020A, HIS-2022A, HIS-2021A and HIS-2023A.
.I i
5.
Added flip guards on SFAS pushbuttons HIS-2020B, HIS-2022B, HIS-2021B and HIS-2023B.
1 6.
Added plexiglass cover over CRD breaker trip switches for CRD breakers A, B, C and D.
7.
Added plexiglass guard to side of Electrical Distribution Panel C5715.
This FCR 84-132 was closed March 4, 1987.
REASON FOR CHANGE i
As a result of the Detailed Control Room Design Review Project, the above modifications were instituted to prevent accidental actuation of switches and are required to be completed by NRC commitment HED 4.1-1.
SAFETY EVALUATION
SUMMARY
The Detailed Control Room Design Review Project determined what control switches should be provided with guards to prevent accidental' operation.
The controls identified were reactor trip, main feed pump trip, turbine trip, SFAS actuation, control rod drive breaker trip and electrical distribution panel. The covers and guards do not change the safety 4
related functions of these systems.
Based on the above discussion, FCR 84-132 will not increase.the probabil-ity of occurrence or the consequences of an. accident or malfunction of safety.related equipment as evaluated in the USAR. Therefore, an un-reviewed safety question does not exist.
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' COMPLETED FACILITY CHANGE REQUEST
-l l
4 FCR NO 84-026 Rev. A SYSTEM Control Rod' Drive System j
COMPONENT Reactor Trip. Breakers CRANGE, TEST OR EXPERIMENT
'I FCR 84-026 modified the following:
a 1.
Automatic actuation of the shunt trip in reactor trip _ breakers A, B, 1
C and D.
1 1
2.
Independent testability of the undervoltage and shunt trip circuit.
a 3.
Class IE 125VDC control power supplies for.the shunt trip operations..
4.
Remote alarm capability for loss of control power.:
LI J
5.
Trip sequence of breakers A(B) and C(D).-
-This FCR 84-026 was closed March 2, 1987.
l REASON FOR CHANGE I
These modifications were implemented in response to NRC generic i
letter 83-28 (Appendix A NUREG-1000, Vol. 2) dated July 8,:1983, regarding actions based on generic implementations of Salem ATWS events.:
FCR 84-026 included' revision of drawings 26370-2, 26371-2, 26372-1 and other. changes.
ll SAFETY EVALUATION
SUMMARY
The function of the shunt trip device is to provide :dditional assurance that the reactor trip breakers will open when required. There is no safety function associated with the shunt trip devices.
As concluded from review of the design shown on drawings 26372-1, 26370-2 and 26371-2, the addition of a safety related shunt trip enhanced the ability of the breaker to trip.
Isolation is provided in breakers A and B between the Class 1E and non-1E portion of the breaker control circuit. The trip signal to the shunt trip coil was changed to Class 1E from non-1E. A source incerruption device (SID) was provided. An are suppression diode was connected across the coil to avoid arcing of the contacts of the relays within the SID..
It is' concluded that FCR 84-026 Rev. A enhanced _the reliability.of the reactor trip breaker trip function rather than adversely affecting it.
Based on the above, an unreviewed safety question-does not exist.
'I
a
- 9 COMPLETED FACILITY CHANGE REQUEST FCR NO
.85-0095 SYSTEM Main Steam System COMPONENT N/A CHANGE, TEST OR EXPERIMENT FCR 85-0095 modified the main steam line anchors outside of Containment for both main steam lines.
This FCR 85-0095 was closed March 5, 1987.
REASON FOR CHANGE Non-Conformance Report (NCR) 85-0031 identified the need for beveled shims to be placed beneath the washers for the bolts placed at an angle on these anchors.
SAFETY EVALUATION
SUMMARY
The modification of the anchors, by placing tapered shims benetth the nuts and washers for bolts where gaps exist, will provide full load bearing.
It will allow the anchors to continue meeting longterm conditions by preventing the possibility of any nuts loosening due to thermal expansion and contraction of the anchor's structural. members.
Based on the above discussion, there will be no" increase in_the probabil-ity of occurrence, or consequences of an accident or malfunction of safety related equipment. Therefore, an unreviewed safety question does not exist.
a 4
COMPLETED FACILITY CHANGE REQUEST j
- y i
FCR NO 85-0331
]
SYSTEM Low Pressure Injection COMPONENT Anchor A-41' CHANGE, TEST OR EXPERIMENT FCR 85-0331 repaired low pressure injection anchor A-41 by the addition of shims which' restored function to design condition.
This FCR 85-0331 was closed February, 26 1987.
REASON FOR CHANGE FCR 85-0331 was implemented to correct a deficiency identified in Non-Conformance Report (NCR) 85-0791.
1 SAFETY EVALUTION
SUMMARY
Anchor A-41 is located on piping in the Low Pressure Injection / Decay Heat (LPI/DH) System.
The LPI/DH System is'part of the emergency core cooling system which functions to provide cooling water to the reactor at low RCS pressure.
FCR 85-0331 repaired anchor A-41 to meet the original design condition for the support.
The support is now able to perform its design function.
The repair of this support will not increase the probability of occurrence or the consequence of an accident or malfunction of safety-related equip-ment. Therefore, an unreviewed safety question does not exist.
8 TOLEDO EDISON EDISON PLAZA O ED, H 43652 June 12, 1987 KB87-00187 File:
RR 2 (P-6-87-05)
Docket No. 50-346 License No. NPF-3' Mr. Harold Denton, Director Office of Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Mr. Denton:
Monthly Operating Report, May 1987 Davis-Besso Nuclear Power Station Unit-1 Enclosed are ten copies of the Monthly Operating Report'for Davis-Besse Nuclear Power Station Unit 1 for the month of May 1987.
If you have any questions, please feel free to contact Morteza Khazrai at.
(419) 249-5000, Extension 7290.
Yours truly, 44GPo
(
sr Louis F. Storz s
Plant Manager Davis-Besse Nuclear Power Station LFS/MK/ljk Enclosures cc:
Mr. A. Bert Davis, w/1 Regional Administrator, Region III Mr. James M. Taylor, Director, w/2 Office of Inspection and Enforcement Mr. Paul Byron, w/1 NRC Resident Inspector Nuclear Records Management, Stop 3220 h(
LJK/002 s\\\\
L i
f TOLEDO IEE)lEIC)rd EDISON PLAZA 300 MADISON AVENUE TOLEDO. oHlo 43652 June 12, 1987 KB87-00187 i
File:
RR 2 (P-6-87-05)
Docket No. 50-346 License No. NPF-3 Mr. Harold Denton, Director Office of Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Mr. Denton:
Monthly Operating Report, May 1987 Davis-Besse Nuclear Power Station Unit 1 Enclosed are ten copies of the Monthly Operating Report for Davis-Besse Nuclear Power Station Unit 1 for the month of May 1987.
If you have any questions, please feel free to contact Morteza Khazrai at (419) 249-5000 Extension 7290.
Yours truly, f hy/!:.,
- .1 tj)
Louis F. Storz Plant Manager Davis-Besse Nuclear Power Station LFS/MK/ljk Enclosures cc:
Mr. A. Bert Davis, w/1 Regional Administrator, Region III Mr. James M. Taylor, Director, w/2 Office of Inspection and Enforcement Mr. Paul Byron, w/1 NRC Resider.t Inspector Nuclear Records Management, Stop 3220 LJK/002
}[b '
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