NUREG-0366, Response to 790102 Request Providing Tables & Individual Plant Summaries from Draft NUREG-0366 Nuclear Power Plant Operating Experience for 1977
| ML19312A137 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 01/04/1979 |
| From: | Barth C NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| To: | Smith I Atomic Safety and Licensing Board Panel |
| References | |
| NUDOCS 7902050240 | |
| Download: ML19312A137 (21) | |
Text
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o NRC PUBLIC DOCUMENT ROOM UNITE 0 STATES
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/4 NUCLEAR REGULATORY COMMisslON y
3 WASHING TON, D. C. 20555
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January 4,1979 c-g T 4 d, p y
a Ivan W. Smith, Esq., Chairman
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d gg ef Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission 9
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Washington, D.C.
20555
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In the Matter of Carolina Power and Light Company (Shearon Harris Nuclear Power Plant, Units 1, 2, 3 and 4)
Docket Nos. 50-400, 50-401, 50-402 and 50-403
Dear Chairman Smith:
In response to your request of January 2,1979 we are providing you, and all parties, with Tables 4.1, 4.2 and the individual plant summaries for Brunswick and Robinson from the draft NUREG-0366 Nuclear Power Plant Operating Experience for 1977.
If the final version is issued prior to the hearing and the figures enclosed herewith change, we will so inform the Board and all parties.
Sincerely, W$
Charles A. Barth Counsel for NRC Staff
Enclosure:
As Stated cc:
(w/ enclosure)
Atomic Safety and Licensing Mr. Glenn 0. Bright Board Panel Dr. J. V. Leeds, Jr.
Docketing and Service Section Richard E. Jones, Esq.
Wake County Public Library Thomas Erwin, Esq.
George F. Trowbridge, Esq.
Atomic Safety and Licensing Dennis P. Myers Appeal Board 7 9 0 2 0 5 0N/d'
l i
DRAFT NUCLEAR PO',1ER PLANT OPERATING EXPERIENCE i
1977 i
e t
i i
. Manuscript Completed:
Date Published:
Office of Management and Program Analysis ll.S. Nuclear Regulatory Commission
'>ashington, D.C.
20555 se
.l_.__..
4.2.3 Operational LERs - 1977 Introduction The data for the LER file was tabulated by reactor type, i.e., BWRs and PWRs, and categorized by plant systems in Tables 4.1 and 4.2.
In Appendix B-1, these systems are presented in a list with the subsystems included.
In addition, components identified in the licensee report are categorized in accordance with the components list presented in Appendix B-2.
In general, these system and component categories correspond to thnw developed by subco=1ittee il 18-20 of the American National Standar-Institute, fluclear Plant Reliability Data.
The 61 commercially operating plants covered in this report submitted 2918 LERs during 1977, an increase of 906 over 1976.
This can be partially attributed to two new BWRs (Browns Ferry 3 and Brunswick 1) and three new PWRs (Crystal River 3, Calvert Cliffs 2, and Salem 1), that started commercial operation in 1977.
In addition, St. Lucie which started operation in December,1976, is included in this report.
The new PWR plants submitted 397 LERs, while the BWRs submitted 139.
The already existing BWRs increased their submittals by 87 LERs, and 'he PWRs t
increased theirs by 283.
There are various reasons for this increase, including more stringent requirements in some areas.
Systems and Components (See Appendix B-1 and B-2 for list of System and Components)
LERs are classified into 14 major system headings including 72 subsystems.
As in previous years, the major systems most frequently reported were the Engineered Safety Features, Reactor Coolant and Connected Systems and Instrumentation and Control.
~
% of Reports
% of Reports Engineered Safety Features 28.1 18.00 Reactor Coolant a Connected Systems 19.1 12.38 Instrumentation & Control 14.4
, 15.97 The five new plants reported most frequently the same systems reported by previously operating plants.
The components most frequently reported were:
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0 2
18 0
0 4
0 0
C 16
,0 BROWNS FERRY 1 5
0 7
1 1
3 2
0 0
3 3
3 1
2 31 2.23 ;
BROWNS FERRY 2 2
0 1
0 0
0 3
1 0
4 4
0 0
1 16 1.23 '
BROWNS FERRY 3 4
0 3
1 0
0 3
2 0
1 7
,0 0
1 f 22 1.09 i l4 117 l 8.99 0
6 32 2
0 BRUNSWICK 1 2
0 56 5
0 1
6 3
i l0 l 70 l 5.23 BRUNSWICK 2 0
0 30 1
0 0
3 1
1 2
30 i 2 0
COOPER 1 0
0 11 3
0 0
13 6
0 1
12 6
2 0
54 4.15 DRESDEN 1 2
0 5
1 1
1 8
10 0
3 5
0 0 l0 36 l 2.75 76l5.04 DRESDEN 2 0
0 15 11 0
1 12 2
1 9
23 1
0 1
54l4.15 DRESDEN 3 0
0 14 10 0
2 17 1
3 1
8 0
0 1
03 l 714 DUANE ARNOLD 0
1 22 10 0
1 l 10 4
0 5
30 0 l 1 9
l71l5.45 l2 3
11 4
3 14 FIT 2 PATRICK 1 0
1 1G 3
1 3
6 4
10 8
88 l 6.70 l1 HATCH 1 5
0 16 11 1
4 13 5
1 1
12 l 0 31 HUMBOLOT BAY 0
0 0
0 0
0 0
2 0
0 2
0 0
0 l 4
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0 2
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1 1
0 0
0 6
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1 3 l8 l 2 5
1 12 0
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46 l 3 52 l0 l 11 i
4 MILLSTONE 1 0
0 42l3.23 4
0 7
21 0
0 0
MONTICELLO O
O 6
1 0
1 2
l 4.C2 NINE MILE POINT 1 0
1 16 0
0 1
20 2
,3 3
13 1
0 4
61 29 l 2.22 0
O 3
4 0
2 14 I O
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l3 OYSTER CREEK 0
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l93lI.1-PEACH BOTTOM 2 1
0 12 6
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1 23 18 0 l 0
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1 15 4
1 3
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0 3
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67 PILCRtM 0
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1 5
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0 9
5 0
4 2
2 2
3 21 2
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OU AD CITIES 2 0
1 9
3 0
1 5
0 0
1 15 0
1 0
33 2.7 '
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VERMONT Y ANKEE O
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0 6
5 1
0 15 0
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'3 71 366 42 20 52 f 33 TOTAL 24 5
300 109 5
4.00 f100.0d 1 30 5.45 28.11 3.23 1.54 8.37 l 0.38 l 2.5414.36 f 5.76
% Or 1302 1.84 0.33 23.04 gDIL{
TABLE 4.1 BWR PLANT VS. SYSTEM
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gl 9
8 UEAVER VALLEY 1 0
1 19 CALVE RT CLIF FS 1 5 l 2 6 l 10 2
13 28 3
4 19 H
10 0
13 123 l 7.61 CALVERT CLIF FS 2 0
2 17 9
0 9l 21 1
6
/
15 2
1 3
93 l 5.76 l
l CRYSTAL RIVER 3
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4 21 13 0
30l 16 9
5 13 23 l 5 2'
11 153 9 84 J.C. COO K 1
1 4
4 0
2 13 0
4 3
4 1
0 5 )
42 2 60 i
FORT CALHOUN 1 1 l 0
3l 9l 0 l 1 l 8 2
1l 2
12 1l 3
4 l 47 l 2.91 H. B. ROulNSON 1 Ol 0l l 2 1
5l 4
5l 2 l 0
2 l 30 l 126 4
1 l 0l 4 HADOAM NECK 0l 0
8 1
1 l 6 l 6 3
5 1 l 3
0l 0
1 l 35 l 2.17 INDIAN POINT 2 1
0
/ l 5l 0l 2l 3 1 1 10 2
4 2l 01 2l 33 l 4.41 INDIAN POINT 3 0
0l 2 l 4l 0l 1 l 3l 1
6 3
2 0l 0l 1 l 23 l 1.42 KEWAUNEE 2l 0 3 l 6l 0
0l 7 0
2l 0l 14 l 6 0l 1 l 41 l 2.54 MAINE YANKEE Ol 0
1 l 0
0 0l 3 1l 1
1 l 3l 0
1 l 2 l 13 1080 MILLSTONE 2 0l 0l 2 l 9 l 0l 9l 11 l 1
2 0l 19 l 4 l 1
4 G2 l 3.84 OCONEE1 2l 1 l 5 3l 0l 5l 5l 0l 5 1 l 6l 1 l 0
6 4's I 2.48 OCONFE 2 0l 0l 1 0l 0l 3 3l 0l 1l 5 4 l 0l 0l 0 l 17 1 1.05 OCONEE 3 0l 0l 2
0 1 l 1 l 5 0l 2l 2
Sl 0
0l 0 l 18 l 1.11 PALISADES 0l 0
1 8l 0l
- 2 l 7l 2l 2l 3
15 2
0l 1 l 53 l 3.28 POINT BE ACH 1 2
0l 2 2l 0l 2 2l 2 0l 0l 0l 3l 0l 0 l 15 l 0 93 POINT 8E ACH 2 0l 0l 1
0l 0l 2 1
l 0l 4l 0l 4l 0
0l 0 l 12 l 0.74 PRAIRIE 6SLAND 1 0l 0 l 2l 3l 0l14l 3
Ol 2l 0l 15l 2l 0l 3 l 44 l 2.72 I
PR AIRIE ISLANO 2 0l 0l 4 l 2l 0
0l 6 0l 0
0l 9l 0 0l 0l 21 1.30 i
l J
1l R ANCHO SECO 1
0l 3 3
0 l 1 l
2l 2l 0l 2 l 1
20 l 1.24 4
R. E. CINN A 0
0 7
2 0
0 l 0 0
5 4
6 1
0 2
27 l 1.67 SAINT LUCIE I O
O 3
5 0
7 10 1
4 8
5 2
3 5
53 3.28 SALEMI 2
3 19 12 0,
3 19 2
1 1
17
'2 3
8 32 5.69 l 0l 0l 0 0
2 0
3 0
1 1
13 } 0.30 SAN ONOFRE 1 0
0 5[ 1 SURRY1 0
0 4
0 0
1 2
0 4l 1
4 i
2 l 1 2
21 l 1.30 SURRY 2 0l 0
2 0
0 1
2 0l 6l 1 l 5
2l-0 l 20 l 1.24 THREE MILE ISLAND 1 0
1 2
2 0
2 1
1 1
0 5
0 2
6 23 1.42 TROJAN O
Ol 13 5l 1
1 4 l 2 2
3l 10l 1 l
2 0l 44 l 2.72 TURKEY POINT 3 0l 0l 1
3 0l 0l 2l 0l 0 1
1 l 0l 0l 0l 810.50 TURKEY POINT 4 0l 0l 3 0
0l 0l 3l 1 l 3
1l 2l 1 l 0l 0l 14 ! 0.87 Y AN K E E.ROVr E O ! 15 l l 7 0
4 l 5 l 5 l 8I 2l 10 l 0l 0l 1 l 61 13.77 4
ZION 1 0l 8 l 8
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9l 18 l 5!
3l 7e 19 i 2 l 0l 21 76 l 4.70 w
ZION 2 1 l 6 l 4l 5l 0l 3l 15 l 2l 9I 3l 14 f 0f 01 0! C2 I 3 34 TOTAL 27 l 44 l 200 l 149 l 8 l 184 l 258l 48 l 117l 106l 291I 61 l 23 l 100 1 1616i100.00 167 l 2.70 { 12.3Sl 9.22 l 0.50 l 11.39115.97 l 2.97 l 7 25 l 6 56 l 18 00 i 3.77 l 1.42 l 6.22 ?00.001
% OF 1616
l BURS PWRs Component
% of Reports
% of Reoorts Instrumentation & Controls 34.6 19.8
~
Valves and Valve Operators 17.7 15.6 Pumps 2.1 6.2 Pipes / Fittings 4.1 4.1 Circuit Closers / Interrupters 3.8 3.6 Cause, Discovery, and Reactor Status Table ?.3 presents a breakdown of LERs by proximate cause, method of discovery, and reactor status at the time of the event.
The table incicate: the following:
(1) Component failure was the cause of more than half (53%) of the reports, down slightly from the 58% of 1976.
(2) Those causes not fitting into specific categories and termed "Other" accounted for 17%.
This represents an increase from the 9% of 1976.
(3)
Personnel error was the cause of approximately 16% of the events, about the same as for 1976.
(4)
Slightly more than half of the events (51",) were discovered during routine and special tests or inspections.
Approximately 44% were operational events.
As in 1976, BWRs had significantly fewer operational events than did PWRs.
(5) Approximately 25% of the events occurred while the plant was shutdown.
Most of the events (61%) occurred during steady state operation.
4-7
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BRUNSWICK 1 I
f 1
I.
Summary I
Description Performance
- Outages Location: Southport, North Carolina Net Electrical Energy Total No.
30 Docket No:
50-325 Generated (tiV10 :
2,515,789 Forced 25 Reactor Type: BUR Unit Availability Scheduled 5
Capacity (MWe-Net):
821 Factor (%):
56.7 Total:
3399 Hours, 38.8%
Commercial Operation: 3/18/77 Unit Capacity Factor (%)
Forced 3056 Ilours, 34.9%
Plant Age:
1.1 Years (Using }DC):
46.0 Scheduled 343 Hours, 3.9%
~
Unit Capacity Factor (%)
(Using Design IME):
44.3
- Data is for period from date of commercial operation to end of year.
II.
Highlights At the beginning of the year, the unit was in the preoperational testing phase which lasted until March 18, when the unit was declared commercial. On April 27, a massive generator failure occurrod which required complete replacement of the generator stator.
In November, another long outage was necessary l
to replace reactor recirculation pump seals on lA and IB pumps.
O W
11 c1.
I BRUNSWICK 1 I
DETAILS OF PLANT OUTAGES No.
Type Description Cause u
vn ystem Component (1977)
(firs)
Hethod involved Involved i
1 1/1 22 F
During startup testing, the A
3 Reactor' coolant Valves bypass valves went full open (CC) causing the trip.
2
.1/4 58 S
During startup turbine trip B
3 Steam & Power Turbines testing, the turbine was (IIA) tripped. Maintenance was then performed.
3 1/7 67 S
The reactor was manually shut B
1 (Not given)
(Not given) down to continue reactor start-up testing and mainten'nce.
a 4
1/14 52 F
EHC oil leak.
A 2
Steam & Power Turbines (llA) 5 1/17 9
F The station batteries were placed A
3 Electric Power Generators on'the equalizer to raise the (EC) specific gravity.
Thio raised D.C.
Bus Voltage and caused several inverters to trip, causing the reactor to scram.
Inverters were recalibrated.
6 2/3 19 F
Reactor scram - I & C was checking A 2
Electric'Powe'r Circuit closers computer points from E4-DG watt-(EB) interrupters meter.
In connecting the test equipment, Phase B fuse blew, causing all E4 loads to trip.
Lube water to the circulating water pu=ps was lost causing circulating water pumps to trip and the reactor scrammed on low conden-
~~",,,.
ser vacuum.
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BRUNSWICK 1 DETAILS OF PLANT OUTAGES (continued)
No.
Type Description Cause u
wn ystem Component (1977)
(Hrs)
Method Involved Involved 7
2/21 160 F
Reactor Scram - I & C technician A
3 (Not giver.)
(Not given) failed to valve out a Icvel transmitter while performing PT 3.15, causing a high reactor water Icvel signal and a scram.
Outage was continued for main-tenance.
8 3/18 20 F
During startup testing of reactor A
3 Instrumentation Instrumentati pressure Regulator, APRM hi-flux
& controls (IA) 6 controls Icvel caused a scran.
9 4/1 89 S
Reactor scrammed during startup B
3 (Hot given)
(Not given) test #27/8.2-load reject test at full power. Maintenance was performed.
10 4/6 39 F
Control operator reset recire G
3 Reactor coolant Circuit closer:
pump run-back too quickly causing (CB) interrupter FW flow and flux spikes. Scram on high flux 1cvel resulted.
11 4/27 1698 F
Scram caused by turbine trip due A
3 Steam & Power Generators to ground in main generator. The
- * (HA) generator was repaired.
12 7/7 0.4 F
Generator lock out relays tripped. A 4
Steam & Power (llA)
Relays 13 7/8 0.3 F
Turbine overspeed trip test A
4
, Steam & Power (RA) Turbines 14 7/8 22 F
Generator removed from grid to A
4 Steam & Power (HA) Generatore degas and rebalance rotor.
6 Reactor remained critical.
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k BRUNSWICK 1 DETAILS OF PLANT OUTAGES (continued)
"## I" No.
- Type Description Cause 78
- P """E (1977)
(Hrs)
Method Involved Involved 15 7/17 30 F
High leakage in drywell.
A 1
Reactor coolant Pipes, fitti (CB) 16 7/22 55 F
Investigation of drywell A
1 Engineered Safety
- Pipes, floor drain Icakage.
Features (SA) fittings 17 7/28 313 F
Shorted windings in Generator A
3 Steam & Power Generatora stator. A new stator was in-(IIA) stalled.
18 8/12 19 F.
Generator was separated from grid A
1 Steam & Power Electrical to clear a ground in the No. 6 (RA) conductors lift pump.
Reactor remained in hot standby.
19 8/14 11 F
Ground on No. 6 lift pump.
A 4
Steam & Power Electrical (llA)
Conductors 20 8/28 14 F
Steam leak in turbine bldg.
A 4
Reactor coolant Pipes, fittit (CC) 21 9/16 18 F
Operator error caused condensate G
3 Steam & Power NA booster pump low suction and scram (HH) on reactor low water level.
22 9/30 128 S
Startup test #25 (lSIV closure at B
3 (Not given)
(Not given) full power).
23
~10/8 14 F
Clean A-S and A-N water boxes.
A 1
Steam & Power Heat Exchange (HC) 24 30/14 53 F
High drywell leakage through floor A 1
Engineered Safety Pipes, fitti drains.
Features (SA) 25 10/29 37 F
Steam leaks on valve stems and A
1 Reactor coolant Valves ik, repairs to automatic voltage (CC) sc5[
regulator on main encra to r.
t
~
BRUNSWICK 1 DETAILS OF PLANT OUTAGES (continued) 6 No.
Type Description Cause si m ystem Component (1977)
(Hrs)
Method Involved Involved t
26 11/13 333 F
During a test, the seals A
3 Reactor coolant Pumps were drnaged on both recire.
(CB) pumps, requiring replacement.
27 12/2 45 F
Repairs to IA Reactor A
4 Reactor coolant Pumps Feedwater Pump.
(CH) 28 12/16 29 F
Power lost to Unit No. 1 A
3 Elcetric Power Circuit closers E!!C during ECCS Test. Blown (ED) interrupter fuse in E2 bus.
29' 12/21 44 F
Operator reducing' power with H
3 (Not given)
(Not given) recire. pumps. Rx low water icvel and Group 1 isolation caused scram.
30 12/31 1
S Maintenance on condenser B
1 Steam & Power Heat Exchange waterboxes.
(HC)
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BRUNSWICK 2 I
b I.
Summary l
i Description Performance Outages i
i Location: Southport, Notth Carolina Net. Electrical Energy Total No.
23 Docket No:
50-324 Cencrated (tMI):
2,436,597 Forced 19 l
Reactor Type:
BUR Unit Availability Scheduled 4 Capacity (fee-Net):
821 Factor (%):
55.7 Total:
3884 llours, 44.3%
Commercial Operation:
11/3/75 Unit Capacity Factor (%)
Fcrced 787 }{ours, 9.0%
Plant Age: 2.7 Years (Using MDC):
35.2 Scheduled 3097 Ilours, 35.3%
Unit Capacity Factor (%)
(Using Design FME):
33,.9 II.
liighlights Operations for the year ~were routine except for an outage in April for maintenance on a recirculation pump and a 107 day refueling outage from September to December.
i e
6 5
a
r Bl.3NSWICK 2
~,
DETAILS OF PLANT OUTAGES
^*
"##*I "
u un ystem Component No.
Type Description Cause (1977)
(Hrs)
Method Involved Involved 1
1/2 19 F
Reactor feedwater pump 2A A
3 Reactor coolant Pumps tripped during test P.T. 40.2.5 (CH)
(Turbine Control Valve Test),
causing a low water level.
I 2
1/7 23 F
Technician error during test G
3 Reactor (RB)
NA P.T. 1.1.4 caused a Group 1 Isolation and scram.
Engineered 3
2/2 74 F
Eurnt drywell drain pump A
1 Motors motor safety Features (SA) 4 2/14 83 F
Turbine trip caused by a H
3 Steam & Power Turbines spurious signal (high exhaust (HA) hood temp.)
5 2/23 17 F
Reactor vessel low level caused G
3 Reactor coolant Pumps when B feedwater pump tripped (CH) while the other pump was oper-ating at minimum speed.
6 4/5 27 F
Loss of instrument air caused A
3 Auxiliary Process Blowers i
loss of condensate booster pump (PA) supply pressure and loss of feed-water causing scram'on reactor low water level.
7 4/15 503 S
Routine maintenance including B
2 Reactor coolant Pumps
~
~ ~ ~ ~ ' ' '
modification of 2A reactor (CB) recire pump.
8_
5/7 26 F
Reactor feedwater pump and A
3 Reactor coolant Pumps N:5 38 booster pump tripped.
WE**
L (CH)
-9 5/21 19
'F Operator error during test G
3 Reactor C3{
NA a-=
P.T. 40.2.5.
(RB)
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BRUNSWICK 2 9
b
~
DETAILS OF PLANT OUTAGES (continued)
Date h ation Shutdown System Component No.
Type Description Cause (1977)
(Hrs)
Method
.nvolved Involved l
10 5/31 41 F
Mechanic error while cleaning-G 3
Steam 6 Power NA EllC oil strainers caused low (llA)
EllC oil pressure and scram.
11 6/2 8
F Spurious upscale on 2C Radi-11 3
Instrumentation Electrical ation Monitor caused by a
& controls (IA)
Conductors man climbing on cable tray in the cabic spreading room near affected channel cables. New cables installed.
12 6/14 24 F
Group 1 isolation due to reactor A
3 Reactor coolant Instrumentatio-low water level caused by heater (Cil)
& controls drain deacrator level oscillations.
13 7/11 26 F
Lou condenser vacuum due to in-A 3
Steam & Power (Unknown) sufficient circulation water flow.
(IIF) 14 7/15 76 F
liigh turbine exhaust temperature A
3 Steam & Power Turbines caused Group 1 isolation and scram.
(llA) 15 7/19 9
S Investigation of 7-15 scram.
B 4
Steam & Power Turbines Repaired EllC system.
(llA) 16 7/31 19 F
Technician error while per-G 3
(Unknown)
(Unknown) forming test caused scram.
17 ~,8/14
~
39
'~ F -
I6C error while performing test G
3 Instrumentation Instrumentation P.T. 3.1.5 caused Group I Isola-
& controls (IA)
& control tion and scram.
18 8/17 1,07 F
Con' denser leaks and high chlorides A 1
Steam 6' Power licat Excha gerJ in main condenser.
(HC) y 4
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Descri : ion Perfog.ance Outages Location: Hartsville, S.C.
Net Electrical Energy To :al No.
15 i
Docket No:
50-261 Gcnerated (M.H): 4.230,398 Porced 12 Reactor Type: PWR Unit Availability Scheduled' 3
Capa city (1."..~e-::e t) : 712 Factor (%):
85.2 Total:
1293 llours,14.8%
Co:=ercial C;eration:
3/7/71 Unit Capacity Factor (%)
' " ' Fo'rce'd 8 1236 hours0.0143 days <br />0.343 hours <br />0.00204 weeks <br />4.70298e-4 months <br />,14.1%
Plant Age:
7.3 Years (Using MDC): 72.6 Scheduled 57 Hours, 0.7%
Unit Caracity Factor (%)
(Using Design >!WE): 69.0 l
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Highlich:s The unit was operated routinely at 100% of rated power for the first half of the year and then at reduced power from July through November to extend the fuel cycle. There were two months in which operation was uninterrupted. During December, the unit. operated at 100% of rated power and the year ended with the unit at full power.
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-' ROBINSON 2 DETAILS OF PLANT OUTAGES Date Duration No.
Shutdown System Component (1977)
(Hrs)
Type Description Cause Method Involved Involved 1
01/11 3
F Loss of motor control A
2 Electric Electrical
- center, Power (EB) conductors 2
01/24 4
F liigh pressurizer pressure caused A 3
Steam & Power Valves by turbine valves closing.
(11B) 3 01/25 3
F Loss of "B" inverter.
A 3
Electric Power Generators (ED) 4 02/05 278 F
"C" RCP high seal leakage.
A 1
Reactor Coolant Pumps (CB) 5 02/18 34 F
Inspect turbine generator for A
1 Steam & Power Generators insulator failure.
(IIA) 6 03/23 30 F
E II. governor valves failed.to A
3 Steam & Power Valves close causing reactor trip.
(IIA) 7 03/25 26 S
Repair turbine trip block.-
B 1
Steam & Power Turbines (ilA) 8 04/24 148 F
Loss of leak-off on "C" Reactor A
3 Reactor Pumps Coolant Pump.
Coolant (CB) 9 06/17 21 S
Operator braining.
E 1
Reactor (RB)
Instrumen-tation &
Controla-sC" 10 08/17 11 F
Safeguard trip during test
,A 3
C Steam &' Power Valves C
cycling of steam driven aux.
(Illi)
A feedwater pump discharge valve.
11 08/23 7
'F Repair feedwater heater tube A
1 Steam & Power Heat Exchangers leaks (101)
f 1 ROBINSON 2 DETAILS OF PLANT OUTAGES (continued)
Date h ration No.
Type Description Caus e Shutdown System Component (1977)
(IIrs)
Method Involved Involved 12 09/30 10 S
Repair Feedwater IIcater B
1 Steam & Fower IIcat Exchangers tube leaks.-
(!!!!)
13 10/05 4
F Turbine trip due to high level A
3 Steam & Power llea t Exchangers in B Steam Generator.
(IIB) 14 10/25 712 F
RIIR Valve No. 750 experienced a A
Engineered Accumulators blown packing. Outage was ex-Safety Features tended to replace a boron in-(SF) jection tank.
15 12/19 2
F Overtemperature delta-T while A
3 Instrumentation Instrumentation performing periodic testing.
& Controls (IA) & Controls O
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BRUNSWICK 2 DETAILS OF PLANT OUTAGES (continued)
No.
Type Description Cause Shutdown System Component (1977)
(Hrs)
Method Involved Involved 19 9 /4 25 F
Generator automatic voltage A
3 Steam & Power Generators regulator was out of service.
(HA)
While attempting to adjust manual rheostat, generator
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tripped on undervoltage followed by a load reject and scram.
20 9/8 14 F
Feedwater pump control oscil-A 3
Reactor coolant Instrumentation lation caused scram.
(CH)
& controls 20a 9/10 1817 S
Reft < ling and maintenance.
C 2
Reactor (RC)
Fuel Elements 20b 9/10 contd. 768 S
Repaired cracked core spray B
4
,"E " h PI es, fittings P
res piping.
37) 21 12/26 120 F
Steam leaks on main steam by-A 4
Reactor coolant Valves pass valves and problems with (CC)
HPCI system.
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