ML19211C913
| ML19211C913 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 01/08/1980 |
| From: | Murray T TOLEDO EDISON CO. |
| To: | Stello V NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| Shared Package | |
| ML19211C914 | List: |
| References | |
| L80-018, L80-18, NUDOCS 8001150467 | |
| Download: ML19211C913 (1) | |
Text
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_.;] h*.' b i C N U January 8, 1980 L80-018 FILE:
RR 2 (P-6-79-12)
Docket No. 50-346 License No. NP F-3 Mr. Victor Stello, Jr., Director Office of Inspection and Enforcement U. S. Nuclear Regulatory Commission Washington, D. C.
20555
Dear Mr. Stello:
Monthly Operating Report, December,1979 Davis-Besse Nuclear Power Station Unit 1 Enclosed find ten (10) copies of the Monthly Operating Report for Davis-Besse N0 clear Power Station Unit 1, for the month of December,1979.
Also enclosed is a revised copy of the Operational Summary from the Monthly Operating Report for the month of November, 1979. The revision to this summary is indicated by a "1" in the lef t margin.
Yours truly, o"
-(-m Terry D. Murray Station Superintendent Davis-Besse Nuclear Power Station TDM/lj k Enclosures cc:
Mr. James G. Keppler Regional Director, Region III Encl:
1 Mr. Norman Haller, Director Office of Management Program Analysis 1765 161 Encl:
3 THE TOLEDO ED! SON COMPANY EDISON PLAZA 300 MADISON AVENUE TOLEDO, OHIO 43652 soo,no #6 9
AVERAGE DAILY UNIT POWER LEVEL
~
DOCKET NO.
Davis-Besse Unit 1
.IT January 8, 1980 DATE Erdal Caba CO5fPLETED BY 419-259-5000, Ext.
TELEPilONE 236 December, 1979
~
MONTil DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL O!We-Net)
(stwe. Net!
0 1
0 37 0
2 0
gg 0
0 39 3
0 0
4 20 0
0 23 5
0 0
22 6
0 0
7 23 0
0 24 8
0 0
9 25 0
10 0
26 0
0 11 27 0
0 12 28 0
0 29 13 0
0 30 14 0
0 15 3g 0
16 1765 162 INSTRUCTIONS On this forma:, list the aserage duly unit power leselin 51We-Net for each day in the reporting month. Compute to the nearest whole megawatt.
(9/77) 6 3
46.0 g d 1765 163 J
OPERATING DAT, A REPORT DOCKET NO. 50-346 -
DATE January 6, 1980 COMPLETED 13Y Erdal Caba TELEPilONE 419-259-5000. Ext.
236 OPERATING STATUS Davis-Besse Unit 1 Notes e
- 1. Unit Name:
December, 1979 f
- 2. Reporting Pericd:
I 2772 f
- 3. Licensed Thermal Power t. twt):
\\
925
. 4. Nameplate Rating (Gross. lwe):
\\
3
,5. Design Electrical Rating (Net StWe):
E,' 6. Stasimum Dependable Capacity (Gross StWe): to be determined to be determined
- 7. Stasimum Dependable Capacity (Net MWe);
- 8. If Changes Occur in Capacity Ratings (Items Number 3 Through 7) Since Last Report.Gise Reasons:
c None
- 9. Power LeselTo Which Restricted,if Any (Net $1We):
- 10. Reasons For Restrictions.If Any:
- i-This 51onth Yr. to.Date Cumulatise 744 8,760 20,525
- 11. Ilours in Reporting Period 29.1 4,332.4
,10,964.2
- 12. Number Of flours Reactor Was Critical 0
2,085.5 1875.8
- 13. Reactor Resene Shutdown flours 0
4,141.6 9,674.8
- 14. Hours Generator On Line 0
1.728.2 1.7.8.2
- 15. Unit Reserve Shutdown flours 806 10,011,937 20.199,507
- 16. Gross Thermal Energy Generated (5thil) 0 3,339,756 6,723,511
- 17. Gross Electrical Energy Generated 13thil) 0 3,129,118 6,170,578
- 18. Net Electrical Energy Generated (. tWH)
\\
0 47.3 49.5
- 19. Unit Senice Factor 0
67.0 58.8
- 20. Unit Asailability Fa: tor to be determined
- 21. Unit Capacity Factor (Using SIDC Net) 0 39.4 36.2
- 22. Unit Capacity Factor (Using DER Net)
,_28.5 27.6 100
- 23. Unit Forced Outage Rate
- 24. Shutdowns Scheduled Oser Next 6 Stonths (Type. Date.and Duration of Each):
Refueling Outage, March 1980 12 weeks January 5, 1980
- 25. If Shut Down At End Of Report Period. Estimated Date of Startup:
- 26. Units in Test Status (Prior to Commercial Operation):
Forecast Achieved INITIAL CRITICALITY INITIAL ELECTRICITY CO. t3tERCl \\ L OPER ATION '
\\
1765 164 m 2)
OPERATIONAL SIDD!ARY DECEMBER, 1979 The unit shutdown which was initiated on November 30, 1979 to investigate the motor lower bearing oil level alarm on Reactor Coolant Pump (RCP) 1-2; to fix Group 7 Rod 5 and Group 5 Rod 11 APIS was still in progress throughout the month of December. The following is a list of the major work items performed during the outage:
1.
The RCP l-2 motor lower bearings oil leak was thoroughly investigated.
No concrete reason for the low IcVel was found.
It was possibly due to filling the motor when it.was running which caused a syphon effect.
2.
Group 7 Rod 5 and Group 5 Rod 11 APIS were fixed.
3.
Shortly after the shutdown, RTD RC3A3 failed, and it was thought that there was a primary leak through the RTD.
It was later determined that the Icakage was through the furmanited gasket and into the thermocouple extension piece. The furmanite box had come into contact with a Reactor Coolant System (RCS) piping whip restraint bending the RTD boss and RTD, cracking the extension piece. This in turn allowed water to leak into the RTD head and short out the signal.
Subsequently, it was found that the two furmanited RTDs on Loop 2 were both damaged by the furmanite boxes contacting whip restraints.
On Loop 1 there was a slight dent in the bosses due to the whip restraints contacting the pipe. Three of the bosses were replaced and new well type RTDs were also installed.
The repairs to the bosses were completed on December 26, 1979.
4.
Due to the extended outage several eighteen month surveillance tests were completed.
\\
UNIT S!!UTDOWNS AND FOW L!t REDUCTIONS
- Dog,
{0'
' a a linit 1 E
Januarv H.
1(18 0 DATE f5,1 raba d
COMPLETED BY REPORT MONT!! December, 1979 WPitONE 410-M0 5000- Ext. 236 c
.] g 3
=
4Yl Licensec Et y%
Cause & Corrective o
No.
Date o-E2
_E 5 5 Event s?
93 Action to j@
3 g =g Report a di0 EU Prevent Recurrence u
20 79 11.30 S
11 4 B
1 NA NA NA Maintenance outage due to a low (Continued) bearing oil level alarm on Reactor F
732.6 Coolant Pump 1-2.
See Operational Summary for further details.
N Ch
( Tl 1
CP-I Ch 3
4 l
2 Method:
Exhibit G - Instructions F: Forced Reason:
S: Schedu!ed A Equipment Failure (Explain) 1 -Manual for Preparation of Data B. Maintenance of Test 2-Maaua! Scram.
En:ry Sheets for Licentee 3-Automatie Scram.
Event Report (LERilYe(NUREG-C Refueling D Regulatory Restriction 4-Ot her (Explain) 0161) 1:-Operator Training & License E.samination 5
F Adnunistrative Exhibit 1 Same Source G Operational 1.iror (Explain)
(')/77) lI Other (laplain)
DATE:
Decenber. 1979 REFUELI';G INFOR"ATICN_
1 Davis-Besse Nuclear Power Station Unit 1.
Name of facility:
March, 1980 Scheduled date for next refueling shutdown:
2.
3.
Scheduled date for restart following refueling:
June. 198f_ _ _
Will refueling or resumption of operation thereaf ter require a technical If ancwcr is yes, what, 4.
specificatica change or other license amendment?If ancwer is no, has the reload f in general, will these be?
Safety Review Committee and core configuration been reviewed by your Plant to determine whether any unreviewcd safety questions are associated with the core reload (Ref.10 CFR Section 50.59)?
Yes, see attached Scheduled date(s) for sub=itting proposed licensing action and supporting 5.
information.
Dec ember, 197 9 Important licensing considerations associated with refueling, e.g., new or different fuel design or suppider, unreviewed design or perf ormance analysis 6.
changes in fuci design, new operating procedures.
methods, significant fuel pool capacity expansion program was approved by the NRC The spent 1, 1979.
in Amendment 19 to the operating license received August (a) in the core and (b) in the spent fuel 7.
The number of fuel assc=blics storage pool.
0 (zero) 177
_ (b).
(a)
The present licensed spent fuel pool storage capacity and the size of any has been requested or is planned, 8.
increase in licensed storage capacity that in number of fuel assemblies.
0 (zero)
Present 735 Increase size by refueling that can be discharged to.the spent The projected date of the last 9.
fuci pool assuming the present licensed capacity.
fuci 1989 (assuming ability to unload the entire core into the spent retuellug cycle)
Date the untt goes to an 16 month pool is maintained and 1765 167
i REFUELING INFOPJL\\ TION Continued Page 2 of 2
, December, 1979 4.
The following Technical Specifications (Part A) will require revision:
2.1.1 & 2.1.2 - Reactor Core Safety Limits (and Bases) 2.2.1 - Reactor Protection System 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) i7165f68
COMPLETED FACILITY CHANGE REOUESTS FCR NO:
79-378 (including Supplements 1 through 4)
SYSTEM: Various Class IE Systems COMPONENT:
125 VDC Control Relays CHANGE, TEST, OR EXPERIMENT: On November 19, 1979, the installation and testing associated with the implementation of FCR 79-378 and its four supplements was com-pleted. As requested by the FCR, arc suppression diodes were installed in various safety related and non-safety related circuits where the contacts of Couch relays are controlling highly inductive loads (generally the coils of other relays). A total of 26 safety related circuits was involved including 4 control relays in the auxiliary f eedwater pump controls,18 in switchgear controls, and 4 in emergency diesel generator controls.
REASON FOR THE FCR:
It was found that failures of Couch relays which occurred in the reactor coolant pump interlock circuitry (see Licensee Event Report NP-33-7 9-12 6) had been caused by the high transient voltage which occurs when the Couch relay contacts interrupt the current flow through a highly inductive load.
The high voltage arcs across the contacts of the Couch relays resulting in a short circuit. The diodes added under this FCR preclude this from occurring by shunting out the transient voltages through a path separate of the relay contacts.
SAFETY EVALUATION: This change will not adversely affect the function of the relays.
It will improve the reliability of the circuits by providing a path for the discharge of the energy in the coil through the diode, rather than through the contacts of the interrupting relay. This is not an unreviewed saf ety question.
1765 169
COMPLETED FACILITY CHANGE REQUESTS FCR NO: 79-405 SYSTEM: Reactor Protection System 'RPS)
COMPONENT: Power Range Nuclear Instrument (NI) 5 CHANGE, TEST, OR EXPERIMENT: On November 15, 1979, the physical work and testing associated with the implementation of FCR 79-405 was completed. This FCR moved cables 2LRPSA01C and 2LRPSA0lX, which are the leads for the bottom ion chamber signals of NI-5, from penetration cable CW3 to penetration cable DW3 both located in penetration P2L4EX/P2L4GI.
REASON FOR THE FCR: During the performance of IC 2002.03, NI Detector Post Installa-tion Test, it was found that the inner and outer shicids on penetration cable GW3 were shorted together.
SAFETY EXPLANATION: The change outlined was reviewed on Bechtel Drawing E-530 and determined to have no adverse impact on safety. The change of modules within This penetration P2L4G will utilize a spare which was provided for this purpose.
is not an unreviewed safety question.
1765 170