ML20117P823
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UNITED STATES 8
NUCLEAR REGULATORY COMMISSION o
y,
,E WASHINGTON, D. C. 20655 DEC 7 III l
l MEMORANDUM FOR:
Williams F. Kane Director l
Division of Reactor Projects, RI i
FROM:
Thomas M. Novak, Director Division of Safety Programs Office for Analysis and Evaluation of Operational Data
SUBJECT:
AE0D INPUT FOR THE VISIT OF COM4ISSIONER CARR TO l
THE CALVERT CLIFFS SITE ON DECEMBER 21, 1987 l
Per your staff's (L. Tripp) request on November 30, 1987, enclosed is a brief sumary analysis of the recent operational experience of Calvert Cliffs Units 1 and 2 from January 1,1987 to date.
l If you need an additional infonnation, please contact Marc Haprer (x24497) of my staff.
l Thomas M. Novak, Director Division of Safety Programs
)
Office for Analysis and Evaluation of Operational Data
Enclosure:
As stated l
cc w/ enclosure:
L. Tripp, RI 4
l i
l 6 2090cf 0 p' *hh' 6//c 1
4 Enclosure AEOD Analysis of Operational Experience a~t Calvert Cliffs Units 1 and 2 Since January 1, 1987 I.
Ovarvian Since January 1,1987, Unit 1 has experienced six reactor scrams, and Unit 2 has experienced five.
Unit i has experienced only three ESF actuations, and Unit 2 only four.
All seven of these ESF actuations were associated with reactor scrams. In addition, two events were judged to be significant by the AEOD LER screening process: a Pressuriser Quench Tank Rupture Disk rupture at Unit 1 on March 10 and a loss of all offsite non-emergency AC power on July 23.
II.
Ahnnraal Oneurrannan III. There have been no events classified, or considered, as Abnormal Occurrences (AOs) at Calvert Cliffs Units 1 and 2 to date in 1987.
Other Onarational Data LER Rev.igga To date for 1987, a total of 20 LERs have been submitted by the two Calvert Cliffs units.
Several of the LERs have been updated at each unit.
Calvert Cliffs Unit i has submitted 14 LERs.
Of these, two, both involving equipment problems, were classified as significant by the AEOD LER screening process and are discussed below.
A broad six of causes accounted for the remaining 12 LERs:
six personnel errors; two inadequate procedures; two equipment problems; one design deficiency; and one QA personnel problem.
For Unit 2, six LERs have been submitted.
Of these, none were I
considered significant.
Of the six LERs submitted, four were caused by equipment problems and the causes for the other two have not yet been reported.
The LERs judged to be significant by the ABOD LER screening process
,i were:
Unit 1 LER 87-006, event date March 10, 1987; with the unit at 100 percent power, the Pressuriser Quench Tank Rupture Disk ruptured.
The unit was shut down and cooled down.
The cause of the leakage to the quench tank which had caused the disk to rupture was determined to be from one or both Pressuriser Safety Valves.
Surveillance tests found that both safety valve setpoints were out of specification.
The installed safety valves were replaced with spare valves that had been rebuilt by the manufacturer and subsequently tested.
The cause of the drifting safety valve setpoints is being pursued with the manufacturer.
Beyond review of the LER, discussions with the resident inspector indicated that the
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leakage from one safety valve is believed to have existed for six to eight weeks prior to the rupture of the Quench Tank Rupture 2
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i Unit 1 LER 87-012 (Unit 2 also affected), event date July 23, 1987; j
a fault on a transmission line (line 5052) caused the circuit i
breakers to open to isolate the fault.
In addition, circuit j
breakers for the other transmission line (line 5051) incorrectly tripped open.
The result was a loss of all offsite non-emergency j
AC power.
This resulted in both Unit 1 and Unit 2 reactors i
tripping from 100 percent power on loss of load.
All three emergency diesel generators started and loaded automatically.
Natural circulation was observed on both reactors.
An alert was 4
declared at 3:30 p.m.
i l
1 The initiating fault was caused by a tree coming in contact with the transmission line.
The cause of the breakers tripping at the i
l site on transmission line 5051 was a defective logic circuit card
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in the primary static protective relay circuit.
The defective card i
allowed the primary relays to trip despite the absence of a permissive signal from the other end of the transmission line.
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At 5:23 p.m. alternate offsite power was established to engineered 4
safety bus # 21.
Establishment of electrical power was delayed due I
to a circuit breaker trip for unknown reasons ( operator error may have contributed).
At 7:10 p.m. normal offsite electrical power was restored.
Forced circulation was restored to both units at j
approximately 8:45 p.m.
1 Rmantor Enrama 1
l From January 1 through November 22, 1987, Calvert Cliffs Unit i has j
experienced six reactor scrans from power.
Four of the six scrans were automatic, and two were manual.
Three of the six scrans were
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the result of hardware failures related to three different systems:
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the reactor coolant, chemical and volume control, and main
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generator systems.
Of the remaining scrams, one was caused by a personnel error ( by an unlicensed operator), one by a deficient procedure, and one by a loss of offsite power described above.
Unit 2 experienced five reactor scrans from power so far in 1987.
Three of the scrans were automatic and two were manual.
Two of the five scrans were caused by equipment problems (one in the turbine electro-hydraulic control system and the other in the feedwater system), one by an unknown cause, one by an unknown cause, and one by the loss of offsite power described above.
ESF Actuations Calvert Cliffs Unit 1 experienced only three engineered safety feature (ESF) actuations in 1987, a rate considerably lower than the industry average.
One of the ESF actuations was a loss of offsite power resulting in an automatic emergency power (diesel generator) actuation on July 23, 1987, described above.
The other
ESF actuations were an isolation and an inadvertent boration associated with reactor scrama described above.
Unit 2 experienced only four ESF actuations so far in 1987, a rate well below the industry average.
One of the events was the loss of offsite power resulting in an emergency power actuation described above.
The three other events involved actuation of auxiliary feedwater after reactor scrams.
IV.
Performance Indicator Data Performance Indicator (PI) data extending through September 1987 are attached.
NOTE: The data presented in the PI curves cover only the first three quarters, whereas the data for reactor scrans and ESF actuations above were to the present.
Also note that PI reactor scram data does not count manual scrams.
Finally, note that the PI safety system actuations are only a specific subset of all ESF actuations, namely emergency core cooling system (ECCS) actuations and emergency power (diesel generator) actuations in response to a dead bus.
PREDECISIONAL
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mm 4.5 55533"buscator CALVERT CLIFFS 1
Legend:
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