ML19309B022
| ML19309B022 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 05/08/1975 |
| From: | SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| Shared Package | |
| ML19309B021 | List: |
| References | |
| NUDOCS 8004020499 | |
| Download: ML19309B022 (8) | |
Text
.O O
(
.o
_~
.c 4
./
' NL,i -
1.1 REPORT FOR JANUARY 8004020 lff 1-2 6
C u
(
V, l.1.1 OPERATIONS
SUMMARY
1.1.1.1 There were no significant changes in design of' Rancho Seco Unit 1 made during January, 1975 l.1.1.2 Performance Characteristics 1.
Primary effort this month was the escalation of power from 40% plateau through the 75% on up to the 92.5% plateau. Operator Training and Licensing was also carried out this month.
2.
During flow coastdown tests the data revealed that for the four pump coastdown the flow decrease transient was below the acceptance criteria.
The criterion for the loss of coolant flow accident is that the minimum DNBR experienced by the core shall not be less than 1.30.
The analysis used for the acceptance criteria in the test and reported in the FSAR used the expected design flow coastdown characteristics. This criteria did not represent the minimum flow required to maintain a constant DNBR of 1.30 during the transient. The flow to meet this criteria which was used for the safety analysis accident was received from B&W and revealed that the actual flow was well above the requirements of a minimum DNBR of 1.30 or greater during the transient. Therefore, it was concluded that the test results were acceptable an'd that the revised acceptance criteria would not increase the probability of occurrence or the con-sequences of an accident or malfunction of the equipment important to safety previously evaluated in the FSAR.
3 During flow coastdown tests the data revealed that'for the three pump coastdown the flow decrease transient was below the acceptance criteria.
The criterion for the loss
,f coolant flow accident is that the minimum DNBR experienced by the -sro shall not be less than 1.30.
The analysis used for the acceptance criteria in the test and reported in the FSAR used the expected design flow coastdown characteristics.
This criteria did not represent the minumum flow required to maintain a constant DNBR of 1.30 during the transient. The flow to meet this criteria which was used for the safety analysis accident was received from B&W and revealed that j
the actual flow was well above the requirements of a minimum DNBR of 1.30 or greater during the transient. Therefore, it was concluded that the test results were acceptable and that the revised acceptance criteria would not increase the probability of occurrence or the consequences of an accident or malfunction of the equipment important to safety previously j
evaluated in the FSAR.
1.1.1.3 Changes in Operating Methods The following significant safety related changes were made to Rancho Seco Unit 1 Operating Procedures in January, 1975:
1.
Procedure A.8, " Decay Heat Removal System" Revision 2 - Updated valve lineup to show latest operating configuration.
j 1
L
.n 1
r.
- u...,
C s
(
(. )
.,. 2 Procedgf3 A.33, " Fire Water Protection System" 9hNision 2 - Revised 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />' fill load fuel requirement from 200 to 250 gallons minimum in order to comply with NEPlA requirements.
1.4.1.4 Sueveillance Tests
~
1.
wenactor iullding isolation valve SFV-53610 failed to open on January 19, 9575 dutitig performance of Surveillance Procedure SP 205.07 An Abnormal Occurrentu Report No. A0-75-1 was flied with NRC.
1.1.1.5 Periodic Containment Leak Rate Tests The following local leak rate testing was performed during January,1975:
1.
The Personnel Hatch cuter seal was tested on January 10, 1975 No leakage I
was found.
1.1.1.6 Changes, Tests, and Experiments Authorized by NRC No changes, tests, or experiments requiring authorization from Nuclear Regulatory Commission, pursuant to 10 CFR 50.59a were performed during January, 1975.
l l
b e
f
>v i
gs
(_
([)
1.1.2 POWER GENERATION Nuclear Ja,nuary 1975 cumulative 1.
Number of hours reactor was critical 737 2,022 2.
Number of times reactor was made critical 1
13 3
Gross thermal power generated (MWH) 1,314,568 2,260,830 4.
Effective full power days 19 76 34.11 Elp:trical 1.
Gross electrical power generated (MWH) 437,639 709,199 2.
Net electrical power generated (MWH) 404,709 601,841
.3 Number of hours the generator was on line 723 1,740 1
o e
4 4
=
- e.
v
... ~I e
\\
1801 2 3 4 5 6 7 8 9 '10 11 12 13 14 15 15 17 18 19 20 21 22 23 24 25 28 27 28 29 30 31110 105 105 j
I
~
.1
_l
_.7,
-y
_y 3 70
._ [
i l
73 m..
_.f f
r y
65 65 r
~
i, 1
s
. j.
d v
__. F w
o O
Od 45 45
._t_
O.
Ne Ng q!
40 7
'T""
40 Ta.
g n p b
35 35
. p, dg
,. o,
m4..>J 30 30
- 2..>. l r e os.
a.
25 23 32
- g3 v
a
20 20 S
5 t 2 3 4 S 6 7 8 9 10 Il 12 13 14 lb 16 17 15 19 20 21 22 23 24 25 26 27 28 29 30 31 Month _.bMY 19 5-5
1.1.3 UNIT 1 SHUTDOWNS - January, 1975 Status During Corrective Action Duration Reason Date Type Outage Hoggers and air ejec-8 hours 1-11 Forced Reduced power Lew Vacuum tors placed in service Reduc-from 75% FP until vacuum was re-tion to 55% FP stored.
b) 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> 1-15 Scheduled Manual Shut-
' Conduct Licensing Startup Exam-down inations for Operators.
1-20 Genera-Reactor ran Generator tripped on over ex-Electrical Maintenance 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> tor trip back from 40%
citation and loss of field.
Dept. Investigated problem but could not FP to 25% FP locate any failed components.
1-25 Forced Reactor ran Steam leakage in High Pressure Furmanite compound 51 hours5.902778e-4 days <br />0.0142 hours <br />8.43254e-5 weeks <br />1.94055e-5 months <br /> Reduc-back from 75%
Turbine casing flange.
Injected in flanged asca to eliminate steai i tion FP to 45% FP leaks.
-s,
o tW
..+.
nU.
M,
, o. -
4 c
I,}.~ N M ' A i.. " ' e.'ip January,-1975 l.1.4' MAINHNANCE'0F SAFETY RELATED EQUIPMENT..
Description h
J.1 A. I
. Failure of Reactor Building isolation Valve
. Amer 4agiperformance of surveillance test SP 205 07, Isolation valve Survelliance 4
a Test, the Reactor Building equalization valve SFV-53610 failed to close' from j
lts normal open position. Upon investigation, it_was found that the air line to the 4alve operator had filled with water. The line was drained.and blown dry.
Me weive was stroked six times af ter the repair and operated each time. When I~
the problem was discovered, the redundant isolation valve SFV-53609 was tested successfully. An examination was conducted to increase the performance of the instrument air dryer system. As a result of this examination, the instru-1 ment air dryer silica bed was replaced with 125 lbs of desiccant, cycle time l
was changed from 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the desiccant drying temperature was in-creased to 155*F to optimize the purge flow drying cycle, and the purge flow rate was increased from 11 scfm to 14 scfm. The Reactor was operating at 75%
.Jull Power and no transient was associated with this occurrence.
f 1.1.4. 2.
Turbine Bypass Valve Malfunction As the. result of a plant transient caused by a reactor trip, Turbine Bypass valve PV-20564 stuck open and had to be isolated manually.
Removal of the valve internals revealed damage to the valve cage and plug. Minute amounts of metal were machined off the plug and cage. The valve was reassembled and l
tested satisfactority pri,or to returning to service.
No plant transients
{
occurred as a result of this problem.
1.1.4.3 Failure of Turbine / Motor Driven Auxiliary Feed Pump to Start The Turbine / Motor Driven Auxiliary Feed pump P-318 failed to start when the i
discharge pressure of both main feed pumps dropped below 850 psig during a j'
4 transient caused by a Turbine-Generator / Reactor Trip.
Motor driven auxiliary feed pump P-319 did start as the pressure dropped below 850 psig.
Investigation revealed the actual set point for the pressure switches that start P-318 was found to be lower than the switches that start P-319 The switches were re-calibrated and the automatic start circuit for P-318 was tested and found to operate correctly. Since Auxiliary Feed Pump P-319 did -start on the low pressure signal, no Technical Specification requirements were violated and no further plant transient occurred as a result of this problem.
i.
1.1.4.4 Main Generator Protection System Breaker Failure Main Generator Protection tripping unit differential and -turbine breaker 2C144. internally shorted 480 volts to bell alarm switch annunciator system Edestroying two flasher cards on H2ES system. The damaged breaker was removed and replaced with a new hreaker. New flasher cards were installed and the-4 system was tested successfully prior to returning to service. No plant tran-sient occurred as a result of this problem.
'h m
,~.
A
. L)
O January, 1975 1.1.4 MAINTENANCE OF SAFETY RELATED EQUIPMENT Description 1.1.4.5 SFAS Channel "C" Power Supply Failure SFAS Channel "C" power supply failed during a trip of Vital AC Power Bus inverter "C".
Power supply voltage intermittently dropped to 13 VDC. The failed power
. supply was replaced with a spare from stock. Troubleshooting the failed unit revealed a failed capacitor in the power supply. The failed capacitor was re-placed and the power supply retested successfully. The repair unit was re-Installed and the spiere returned to stock. This component failure occurred simultaneous with t,he "C" inverter trip which resulted in a reactor trip on December 28, 1974.
8 D
e 6
e l
l
l'1
(_'
/d,.i e
I.1.5 CHANGES IN PLANT OPERATlNG OP.GAhlZATION - JANUARY, 1975 There were no changes in the Plant Operating Staff for those positions which are designated as key supervisory personnel on Fig. 6.2-1 of the Technical Specification.
1 4 -
e l
l 4
l 1
,