ML18150A049
| ML18150A049 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 04/17/1987 |
| From: | Cantrell F, Holland W, Larry Nicholson NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18150A047 | List: |
| References | |
| 50-280-87-05, 50-280-87-5, 50-281-87-05, 50-281-87-5, NUDOCS 8704280246 | |
| Download: ML18150A049 (15) | |
See also: IR 05000280/1987005
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTA STREET, N.W.
ATLANTA, GEORGIA 30323
Report Nos.:
50-280/87-05 and 50-281/87-05
Licensee:
Virginia Electric and Power Company
Richmond, Virginia 23261
Docket Nos.:
50-280 and 50-281
License Nos.:
Facility Name:
Surry 1 and 2
Inspection Conducted: March
through April 4, 1987
Inspectors:
.L//; }/}:, )
W.
Date S%gned
L;/; / 1~-}
Date Sighed
Accompanying Inspector:
S. G. Tingen; F. S. Cantrell, Section Chief on 3/13/87
. ,
~ /-//'
7
Approved by: F. S. d~<<,1i\\~filj?'.£hief
Division of Reactor Projects
L//r)j{(J
Date Signed
SUMMARY
Scope:
This routine inspection was conducted in the areas of licensee action
on previous enforcement matters, plant operations, plant maintenance, plant *
surveillance, followup on inspector identified items, and licensee event report
review.
Results: One violation was identified in this inspection report (see
paragraphs 3 and 5) .
8704280246 870420
ADOCK 05000280
G
1.
Persons Contacted
Licensee Employees
REPORT DETAILS
- R. F. Saunders, Station Manager
D. L. Benson, Assistant Station Manager
- H. L. Miller, Assistant Station Manager
D. A. Christian, Acting Assistant Station Manager
J. A. Bailey, Superintendent of Operations
- E. S. Grecheck, Superintendent of Technical Services/
Acting Assistant Station Manager
D. J. Burke, Superintendent of Maintenance
S. P. Sarver, Superintendent of Health Physics
R. L. Johnson, Operations Supervisor
J. A. Price, Site Quality Assurance Manager
- W. D. Craft, Licensing Coordinator
J. B. Logan, Supervisor, Safety and Licensing
- R.H. B1runt, Acting Superintendent of Technical Services
- Attended exit meeting.
Other licensee employees contacted included control room operators, shift
technical advisors, shift supervisors and other plant personnel.
2.
Exit Interview
The inspection scope and findings were summarized on April 6, 1987, with
those individuals identified by an asterisk in paragraph 1.
The following
new items were identified by the inspectors during this exit.
One violation (paragraphs 3 and 5) was identified for failure to provide
adequate procedure and/or failure to foll ow procedure for corrective
maintenance, surveillance testing, and operational evolutions (280;
281/87-05-01).
The Region II Section Chief met with licensee management
at the site on April 8, 1987, to discuss this violation and emphasize
that inattention to detail is a major concern.
One inspector fo11owup item (paragraph 6) was identified for fol1owup on
corrective action for Unit 2 RHR pumps during the next appropriate outage
for Unit 2 (281/87-05-02).
The 1 i censee acknowledged the inspection findings with no dissenting
comments.
The 1 i cense did not identify as proprietary any of the
materials provided to or reviewed by the inspectors during this
inspection.
\\
2
3;
Licensee Action on Previous Enforcement Matters
(92702)
(Closed) Violation 280, 281/ 86-02-02, Failure to establish measures and
procedures for review of 10 CFR Part 21 reports.
The issue involved the.
licensees failure to provide documents which addressed corrective action
taken for a Part 21 report sent to Vepco on October 11, 1982.
Corrective
action taken by the licensee included revision of administrative
procedures to ensure that review of vendor recommendations is addressed
and documented during the review of vendor information. Also, additional
instructions were provided to affected employees on processing of
10 CFR Part 21 reports.
Also, 10 CFR 21 report status is- tracked in
the licensees commitment tracking system. - The inspector verified that
corrective actions were implemented. This item,is closed.
( Closed) Violation 280/86-11-01, Failure to establish measures and
procedures for maintenance operations for vital battery upgrade.
The
issue involved the inc.1dequacy of a design change procedure (DC-85-32-1).
This procedure was the controlling procedure for replacement of the
station 18 battery.
Step 4.3 of the procedure incorrectly loaded the
battery with vital loads prior to lifting the battery leads.
When the
leads were lifted, a v~ltage transient occurred resulting in a train B
safety injection, letd"own isolation, containment phase I and phase II
isolation and loss of annunciator panels F through J.
Corrective action
taken by the 1 icensee included revision of the procedure to prevent
reoccurrence.
Also, the importance of appropriate preparation and review
of design change procedures was reemphasized with management and
individuals responsible for the design change process.
During the Unit 2
refueling outage, the 28 battery was replaced with the revised procedure.
No transients were encountered. This item is closed.
(Closed) Violation 281/86-20-01, Failure to provide adequate procedure
which documents corrective maintenance on safety-related equipment.
The
issue involved inadequate documentation of corrective maintenance which
was accompli.shed on Unit 2 recirculation spray heat exchanger B during the
period of July 23-27, 1986.
Licensee corrective action included complete
revision of appropriate maintenance procedures to include additional
controls on component disassembly and reassembly, additional controls on
mechanical joint reassembly which cannot be operationally checked after
completion of work, and . specific retesting requirements to verify
component operability.
The inspector reviewed the revised procedures and
considers that appropriate corrective action has been taken. This item is
closed.
(Closed) Violation 281/86-42-02, Inadequate Procedure for the Maintenance
of the Main Steam Trip Valve. Inspection reports 280;281/86-41 & 42
address in detail the improper assembly and testing of the main steam trip
valves (MSTV) that resulted in a reactor trip on December 9, 1986. The
licensee response to this violation was documented in a letter, dated
3
March 11, 1987, acknowledging the violation as written and committing to
both a short term corrective action of revising the subject procedure and
a long term project involving updating all *safety related procedures. The
inspectors reviewed the revised * MSTV procedures as documented in
inspection report 280;281/87-04 with no discrepancies noted. The resident
inspectors wil 1 continue to monitor progress in the overa 11 procedure
improvement program during regular inspections. This item is closed.
(Closed) Unresolved Item 281/86-41-01, Justification for deletion of step
5.4.4.3 of corrective maintenance procedure MMP-C-RH-015 during repair of
RHR pump 2-RH-P-18 in October, 1986_.
The issue involved a procedure
deviation (change) to the corrective maintenance procedure which
eliminated a step from the procedure which would check the motor mounting
fit for centralization.
The acceptable tolerance for this step was 0.003
- inch.
However, the reason listed for deleting this step from the
procedure was that the runout was from 0.020 to 0.025 - inch due to rust
and pitting.
The pump was reassemi:> led with out correcting the above
condition.
The inspector's review of the pump technical manual determined
that the step deleted was one of the steps listed in th~ manual to assure
proper pump alignment.
Discussions with plant management resulted in a
conclusion that management was aware cf the condition of the pump when the
decision was made to reassemble the pump; however, appropriate engineering
documentation was not included in the work package to allow the inspector
to complete his review of the activity. This issue was unresolved pending
licensee action to provide engineering justification of the procedure
deviation at the end of the inspection period in January 1987.
Since the issue was identified, several meetings have been held between
station management and the inspector. The licensee provided the inspector
with an "Analysis of 2-RH-P-18 Procedure Deviation" on March 10, 1987.
In
that analysis, the licensee stated that after the repairs conducted 1n
October 1986, full operability of the pump was verified in accordance with
ASME Code,Section XI requirements by performance of 2-PT-30 .1.
In
addition, the licensee stated that although rust and pitting were evident
on the measured surface, constant errors were observed, indicating that
one side was not significantly high or low. _ They also stated that
increasing vi.bra ti on i ndi cations on the RHR motor 1 ed to the staff's
decision to remove the pump from service and send the motor, with shaft,
to Westinghouse for analysis and examination.
During the forced outage to
repair feed and condensate piping, the licensee has overhauled the
residual heat pump 2-RH-P-18 and corrected the surface condition of the
pump stand.
After repairs were completed, the pump and motor post
maintenance testing verified that vibration levels were well within the
acceptable range.
The inspector agreed with the licensee that the pump post maintenance
testing did verify operability and acceptable vibration levels after the
last corrective maintenance activity.
However, the inspector also
reviewed the vibration data recorded for the pump motor bearings taken
-- ---
-
4
before the corrective maintenance in October 1986, and after the
corrective maintenance in November 1986; and concluded that the data
indicated that an abnormal condition existed in the pump-motor assembly
after the maintenance was completed.
Discussions with the licensee and
vendcir representative on March 11, 1987, provided the following.
conclusion.
After disassembly of the pump during the forced outage, the impelJer
wear ring was noted to be deformed in a manner which would indicate
that the motor mounting surface was not mating with the pump stand to
assure proper alignment of the pump/motor assembly.
This condition
helped to explain the higher vibration indications after pump
reassembly in October 1986.
Based on the preceding conclusion, the inspector considers that corrective
maintenance procedure MMP-C-RH-015 did not provide adequate instructions
to assure that proper pump/motor alignment was achieved after deletion of
step 5.4.4.3 from the procedure on October 26, 1986.
Technical
specification 6.4.A.7 requires that detailed written. procedures with
appropriate check-off lists and i nstructi ans sha 11 be provided for
preventati on of corrective maintenance opera ti ans w~i ch would have an
effect on the safety of the reactor.
Failure to provide an adequate
procedure for corrective maintenance on RHR pump 2-RH-P~lB is a violation
,(280; 281/87-05-01).
(Closed)
Unresolved
Item
280/86-41-01;
281/86-41-02,
Licensee
determination of appropriate curves for determining operability pressure
differential for the RHR pump(s).
The issue involved the inspectors
review of a work order (Job Number 3800042352) during a previous
maintenance inspection.
The work had been identified as necessary during
performance of periodic testing (PT) of RHR pump 2-RH-P-lB.
The test,
which was conducted on October 6, 1986, declared the pump inoperable due
to a high differential pressure as required in the acceptance criteria of
the PT.
However the work order was voided prior to* any work being
performed due to an engineering evaluation of the test results.
This
evaluation was documented in engineering work request (EWR)86-414.
The
inspector reviewed the EWR and the PT acceptance criteria and concluded
that different pump curve information was used for the establishment of
the acceptance criterion in the PT as compared to the criterion used to
evaluate the EWR.
This issue was identified to the licensee and a meeting
was held between licensee supervisory engineering personnel and the
inspector on February 24, 1987.
In that meeting the licensee stated that
the information documented in the EWR was correct information based on
engineering review of the specific pump curves and that the PT had been
recently revised to incorporate testing of the RHR pumps at a flow rate of
3000 GPM in lieu of 4000 GPM per the manufacturer general reference data.
The EWR concluded that based on the pump curve at a flow rate of 3000 GPM,
the pump can develop a total head of 280 feet or 121.2 psid instead of the
manufacturer general reference value of 117 psi d.
The acceptable
1
~
. 5
range of operation, as specified in ASME Section XI, Table IWP-3100-2,
should be 112.7 to 123.6 psid.
Since RHR pump 2-RH-P-lB delta pressure
was 122.7 psid when tested on October 6, 1986, the pump was declared fully
Following the results of the EWR on October 7, 1986, the Unit 2 RHR pumps
.were again tested in accordance with 2-PT-30.1 on October 11, 1986.
During that test the delta pressure recorded for RHR pumps 2-RH-P-lA and
2-RH-P-lB were 115.6 psid and 115.6 psid respectively.
This data .fell
within the acceptance criterion of 2-PT-30.1 for full operability.
The
acceptance criterion had not been changed from the manufacturers general
reference value* of 117 psid used when the PT was performed on October 6,
1986.
The inspector reviewed the acceptance criterion of 2-PT-30.1 and
determined the following:
The reference delta pressure used to evaluate pump acceptability was
117 psid.
This reference pressure did not agree with the '.Jalue
determined in EWR 86-414.
Acceptance criteria listed in steps 6.1.1.1 and 6.2.1.1 of 2-PT-30.1
listed a range of 108 to 118 psid for declaring the RHR pumps fully
The inspector calculated this range based on a reference
delta pressure of 117 psid in accordance with ASME Section XI, Table
IWP-3100-2 and determined that the range for- declaring th!? pumps
fully operable should be* 108.8 to 119.3 psid.
Acceptance criteria listed in steps 6.1.2.1 and 6.2.2.1 of 2-PT-30.1
listed a range of 104 psid to less than 108 psid OR greater than 118
psid to 119.5 psid as the ALERT range for the RHR pumps.
The
inspectors calculations based on the ASME code determined that the
ALERT range should be 105.3 psid to less than 108.8 psid OR greater
that 119.3 psid to 120.5 psid.
Acceptance criteria listed in steps 6.1.4.1 and 6.2.4.1 of 2-PT-30.1
listed the.INOPERABLE range for the RHR pumps as less than 104 psid
OR greater than 119.5 psid.
The inspectors calculations based on
ASME code determined that the INOPERABLE range should be less than
105.3 psid to greater that 120.5 psid.
Technital Specification 6.4.A.2 requires that detailed written proced~res
with appropriate check-off 1 i sts and i nstructi ans shal 1 be provided for
testing of components and systems involving nuclear safety of the station.
The inspector concluded that 2-PT-30.1 did not provide adequate
instructions to insure that the RHR pump operabiltty was correctly tested
and verified.
This item is identified as a further example of violation
280; 281/87-05-01.
It should be noted that both pumps were actually
operable at all times based on the final acceptance criteria.
In a related matter, the inspector reviewed two station deviations which
were written on December 12, 1986.
The deviations (Sl-86-843 and
Sl-86-844) indicated that during performance of 1-PT-30.1 for Unit 1 RHR
pumps, the acceptance criterion of the PT declared 1-RH-P-lA *inoperable
6
and 1-RH-P-18 in the ALERT range.
Work Orders (Job Numbers 3800046940 and
3800046941) were issued for engineering evaluation of these conditions.
The work orders were subsequently voided after engineering determined that
the incorrect flow rate was being used for the test. Additional review of
this condition and discussion with licensee management provided
information that the PT tested the pumps at 4000 GPM instead of the
recommended fl ow rate of 3000 GPM.
Si nee 4000 GPM was above the shutoff
head of the pumps, the PT was deviated (changed) to adjust the flow rate
to 3500 GPM and the test was reperformed on December 13, 1986.
Test
results of this PT appeared to be within the acceptance criteria of the
test.
However, during review of the completed test by the inspector, it
was determined that the procedure deviation only corrected the test for
- flow rate and did not correct the acceptance criteria for the new test
- flow rate.
The PT reference delta pressure of 97.4 psid appeared to be
the correct pressure for 4000 GPM based on pump curves.
However, the
inspector determined that the reference delta pressure using a flow rate
of 3500 GPM should be 108.2 psid.
The delta pressure determined by
1-PT-30.1 on December 13, 1986, was 96.1 psid for 1-RH-P-lA and 92.6 psid
for 1-RH-P-18.
The PT dec)ared both pumps operable.
The inspector, based
on a reference delta pressure of 108.2 psid, calculated a value of 97.4
psid as the lowest pressure which would allow the pumps to be declared
operable based on the ASME Section XI, Table IWP-3100-2.
The inspector
also noted that the completed PT was reviewed by the Surveillance and Test
Engineering Group in December 1986/January 1987. This review is conducted
to verify that the test results are satisfactory; however,
no
discrepancies were noted by the engineering review.
The inspector
considers that deviated 1-PT-30.1 which was performed on December 13,
1986, is a further example of violation 280; 281/86-05-01.
This violation was brought to the attention of the licensee and immediate
corrective action was taken.
Based on the final acceptance criteria both
pumps were technically inoperable; however, the inspector verified that
the next PT performed to verify operability on January 14, 1987~ was
properly revised and declared the pumps fully operable.
One violatio~ was identified during these inspections.
4.
Unresolved Items
Unresolved items were not identified during this inspection.
5.
Plant Operations
Operational Safety Verification (71707)
The inspector conducted daily inspections in the following areas:
control
room staffing, access, and operator behavior; operator adherence to
approved procedures, technical specifications, and limiting conditions for
operations; examination of panels containing instrumentation and other
reactor protection system elements to determine that required channels are
7 .
operable; review of control room operator logs, operating orders, plant
deviation reports, tagout logs, jumper logs, and tags on components to
verify compliance with approved procedures.
The inspector conducted weekly inspections in the following areas:
verification of operability of selected ESF systems by valve alignment,
breaker positions, condition of equ*i pment or component( s), and operability
of instrumentation and support stems essential to system actuation or
performance.
Plant tours which included observation of general _plant/
equipment conditions, fire protection and preventative measures, control
of activities in progress, radiation protection controls, physical
security controls, plant housekeeping conditions/cleanliness, and missile
hazards.
The inspector conducted biweekly inspections in the following areas:
verification review and walkdown of safety-related tagout(s) in effect;
review of sampling program (e.g.~ primary and secondary coolant samples,
boric acid tank samples, plant liquid and gaseous samples); observation of
control room shift turnover; review of implementation of the plant problem
identification system; verification of selected portions of containment
isolation lineu~(s); and verification that notices to workers are posted
as required by 10 CFR 19.
Certain tours were conducted on backshifts.
Inspections included areas in
the Units 1 and 2 cable vaults, Vital battery rooms, Steam Safeguards
areas, emergency switchgear rorims, diesel generator rooms, control room,
auxiliary building, cable penetration areas, independent spent fuel
storage facility, low level intake structure, and Safeguards Valve Pit
areas. Reactor coolant system leak rates were reviewed to ensure that
detected or suspected leakage from the system was recorded, investigated,
and evaluated and that appropriate actions were taken, if required.
On a
regular basis, radiation work permits (RWPs) were reviewed and specific
work activities were monitored to assure they were being conducted per the
RWPs.
Selected radiation protection instruments were periodically
checked, and equipment operability and calibration frequency were
verified.
In the course of monthly activities, the inspectors included a review
of the licensee's physical security program.
The performance of various
shifts of the security force was observed in the conduct of daily
activities to include: protected and vital areas access controls;
searching of personnel, packages and vehicles; badge issuance and
retrieval; escorting of visitors; and patrols and compensatory posts.
8
Unit 1 began the reporting period at power.
The unit remained at power
throughout the reporting period; however, the following deficiency was *
identified by the inspectors during this period:
On March 22, 1987, an *operator noticed that the main steam tr.ap root*
valves l-MS-74, 106, and 143 were open in lieu of the required closed
position.
These valves are part of the steam trap system off the
Unit 1 main steam trip valves. They were verified closed on
February 16 with independent verification on February 17.
The
mispositioning of the above valves created an approximately one and a
half inch bypass flow path around each of the three main steam trip
valves. The licensee is currently evaluating the safety significance
of this system configuration. Station administrative procedure
SUADM-0-10 "Operations Department Procedures, specifically requires
that plant equipment shall be operated in accordance with *written
procedures. The licensee can find no documentation subsequent to the
independent verification regarding operation of the above valves.
This failure to follow procedure is a further example of violation
280; 281/87-05-01.
Unit 2 began the reporting period in cold shutdown.
Heatup of the unit
above 200 degrees F. began on March 16 with the reactor reaching
criticality on March 19. The inspector reviewed the licensee's
calculations for estimated critical rod position per .operating procedure
1-0P-lC, "Estimated Rod Bank Position", and witnessed the startup to
verify actual core performance.
The fo 11 owing events occurred during
preparation for power ascension:
On March 12, 1987, the unit was in cold shutdown (approximately 195
degrees F.) with decay heat being removed by steaming to the
condenser.
Condenser vacuum was being maintained by the condenser
vacuum pumps.
The C reactor coolant pump (RCP) was running and the C
charging pump was running to provide seal injection flow to .the
operating RCP.
At approximately 1700 hours0.0197 days <br />0.472 hours <br />0.00281 weeks <br />6.4685e-4 months <br />, the C charging pump
tripped due to operations personnel racking out the B charging pump
breaker for maintenance.
This evolution resulted in loss of seal
injection to the operating RCP.
The unit operator tripped C RCP and
started A residual heat removal (RHR) pump to provide for decay heat
removal.
The control room operator directed that the B charging pump
breaker be racked in, and when the breaker was racked back in, C
charging pump restarted .. The control room operator restarted C RCP
and then tripped A RHR pump.
During the event plant parameters
remained constant. However, a similar event occurred on unit 1 while
at full power in June 1985.
Based on that event, the licensee took
extensive corrective action to prevent recurrence.
That corrective
action included signs being installed on the B charging pump breaker
cabinets warning of the condition and precautions added to the
procedure (Maintenance Operating Procedure, MOP-8.3, "Removal of
9
2-CH-P-B Charging Pump for Maintenance") which is used to remove th*e
pump from service for maintenance. Licensee fo 11 owup determined that
miscommunication between the control room operators and the operator
racking out the breaker, and operations personnel not using procedure
(MOP-8.3) to establish the maintenance condition for B charging pump,
resulted in recurrence of the event.
Failure to follow procedure
when removing from service the B charging pump for maintenance is a
further example of violation 280; 281/87-05-01.
On March 16, 1987, Unit 2 tripped while making preparations for
reactor startup. The unit was at intermediate shutdown, zero percent
power, with shutdown banks A and B withdrawn. An instrument
technician was troubleshooting a faulty steam flow indication and
unblocked the
11at power" P-7 permissive by simulating a reactor power
greater than ten percent. The turbine was unlatched at the time,
resulting iri a reactor trip on a turbine trip. The technician
performing this work per the guidance of Periodic Test procedure 2.9A
"Steam Flow (F-2-485)
11 was not aware that the reactor trip breakers
had been closed and that the shutdown banks were withdrawn. The
procedure recognizes that a reactor trip wi 11 occur and contains
instructions to either latch the turbine or insure operations are
aware that a trip will occut. The technician marked both his and the
shift supervisor 1s signoff
11 N/A
11 for this step without confirming
plant status (i.e. reactor trip breakers closed) with the O?erators.
This failure to follow procedure is a further example of vi6lation
280; 281/87-05-01.
During the startup, water was noted in the bearing lubricating oil
system for auxi 1 i ary feedwater pump P-3A.
The 1 i censee postulated,
subsequent to troubleshooting the 1 ube oil cool er and finding no
leakage path, that the water collected in a clogged pump seal leakoff
reservoir and entered the oil system through the pump bearing. The
inspector witnessed the retest of the pump to declare it operable for
unit startup. No deficiencies were noted in the pump performance.
The inspectors noted that the 1 i censee took prompt and aggressive
corrective action when each issue was identified.
Unit 2 reached full power on March 24, and remained at power until
Apri 1 3, when the unit commenced a rampdown to conduct ba 1 anci ng
evolutions on the main turbine.
During rampdown, the unit tripped at
approximately 2:20 A.M. on April 4, due to a turbine trip on low
differential pressure across the high pressure turbine.
The unit was
restarted and after turbine balancing the operators were preparing to
latch the turbine when the inspection period ended.
Engineered Safety Feature System Walkdown (71710)
The inspector performed a walkdown of the accessible areas of the
emergency diesel generator and fuel oil system for both units to verify
its operability.
This verification included the following: confirmation
J
10
that the licensee*~ system lineup procedure matches plant drawings and
actual plant configuration; hangers and supports ar~ operable;
housekeeping is adequate; valves and/or breakers in the system a.re
installed correctly and appear to be operable; fire protection/prevention
is adequate; major system components are properly labeled and appear to be
instrumentation is properly installed, calibrated and
functioning; and valves and/or breakers are in correct position as
required by plant procedure and unit status. *
Within the areas inspected, additional examples of the violation noted in
paragraph 3 were identified.
6.
Maintenance Inspections (62703)
During the reporting period, the inspectors reviewed maintenance
activities to assure compliance with the appropriate procedures.
Inspections areas included the following:
Inspection of Unit 2 Steam Generator Main Steam Trip Valves (MSTV)'
The inspectoi'S witnessed portions of the maintenance and testing of MSTV
201A, B, and C .. Inspection reports 280; 281/86-41, 86-42, and 87-04
address in detail the improper assembly and testing of the MSTV that
resulted in a reactor trip on December 9, 1986.
On March 3, 1987, the
inspectors witnessed post maintenance testing of the above Unit 2 valves
that assured full opening of the MSTVs.
No discrepancies in testing were
noted.
Review of Unit 2 Residual Heat Removal (RHR) Pump Repairs
During the past forced outage for Unit 2, the licensee decided to overhaul
RHR pump 2-RH-P-18 due to the pump exhibiting high vibrations when
predictive analysis testing was conducted.
The overhaul was completed on
February 20, 1987, and the pump was returned to service on on the same
day.
However, after ten days of operation; the lower radial bearing of*
the pump failed resulting in pump failure.
The inspector reviewed the
completed procedure used to overhaul and rep~ir RHR pump 2-RH-P-lB during
the period of January/February 1987.
All test data indicated that the
pump had been. properly overhauled.
The pump was again tagged out for
repair on March 2, 1987.
The pump motor was removed and shipped to a
repair facility for corrective maintenance.
During the timeframe that the B pump motor was undergoing repair, the
licensee decided to allow the unit to heat up to approximately 190 degrees
F and remove residual heat by steaming to the condenser.
This decision
was made, in part, because vibration analysis of the A RHR pump indicated
that bearing degradation was occurring.
The inspector questioned the
licensee as to whether the pump was considered operable based on bearing
11
wear.
A meeting was held between the NRC and licensee management on
March 12, 1987.
At that meeting, the licensee stated A RHR pump was
experiencing bearing wear: however, the pump was considered operable based
on surveillance testing.
The licensee also stated that they were in the*
process of procuring new parts to properly repair the A pump; however,
these parts would not be available prior to unit restart.
The licensee
also committed tomake appropriate repairs to the A pump during the first
outage of appropriate duration after receiving replacement parts.
The NRC
agreed at the meeting that the A pump was operable and the licensee was
taking a proper approach to correcting any condition which may exist.
After repair the B pump motor was returned to the station and reinstalled.
The pump was tested and declared operable on March 15, 1987.
However,
during plant heatup the upper motor-bearing temperature continued to rise
to greater that 190 degrees F prior to the pump being secured in order to
continue the unit heatup.
NRC concern with this high temperature resulted
in a phone conversation between the licensee and NRC management in Atlanta
on March 17, 1987.
The results of that discussion were that the pump had
been demonstrated to be operable on March 15, 1987, and no degradation of
the bearing was indicated based on vibration analysi_s.
Therefore, NRC
concurred with the. licensee position that the pump was operable and that
the startup could continue.
Foll owup on corrective action for Unit 2 RHR pumps during the next
appropriate outage is identified as an inspector followup item
(281/87-05-02) for Unit 2 only.
Within the areas inspected, no violations or deviations were identified.
7.
Surveillance Inspections (61726)
During the reporting period, the inspectors reviewed various surveillance
activities to assure compliance with the appropriate *proced*ures as
follows:
Test prerequisites were met.
Tests were performed in accordance with approved procedures.
Adequate coordination existed among personnel involved in the test~
Test data was properly collected and recorded.
Inspection areas included the following:
On March 2, 1987, the inspector witnessed portions of the performance of
periodic test 1-PT-28.2, "Reactor Core Flux Map". This test performed flux
map #22 for Unit 1. No discrepancies were identified.
12
On March 19,1987, the inspector witnessed surveillance testing of the
turbine-driven auxiliary feedwater pump 2-FW-P-2 per periodic test
2-PT-15.lC.
This test demonstrates the operability of the subject AFW
pump with the unit stable at greater than 2% power. The inspector noted
that condensate from the turbine exhaust ran down the inside of the
safeguards room wall onto the room emergency lights.
The licensee is
currently evaluating this condition.
Within the areas inspected, no violations or deviations were identified.
8.
Followup on Inspector Identified Items
(92701)
(Closed) Inspector Followup Item (IF!) 280/85-20-02, Followup on action
taken to correct problems encountered while performing containment leak
rate collection and calculations. The issue involved numerous steps which
could allow for potential errors and delays in handling and computing
leakage rate values.
The licensee has upgraded the intergrated leak rate.
test (ILRT) program and installed the upgr~ded program on Unit 1 and 2
computers.
This upgraded program was satisfactorily used during the ILRT
performed on Unit 2 in 1986.
This item is closed.
(Closed) !FI 280; 281/86~02-0l, *Followup on review of procedures and
training for the installation of EQ transmitters and assemblies,
The
issue involved improper installation of a Conax electrical seal assembly.
Corrective action by the licensee included revision of applicable
procedure and additional training for personnel performing this type of
work.
The inspector reviewed the revised.procedures. This item is closed.
(Closed) !FI 280/86-20-01, Followup on review of turbine driven auxiliary
feedwater (AFW) pump maintenance procedure for adjusting governor linkage
and periodic testing {PT) procedure review.
The issue involved several
governor linkage problems experienced by the licensee during return to
service of the Unit 1 turbine driven AFW pump during the latter part of
August 1986.
The inspector reviewed the revised maintenance procedure and
the revised PT (1-PT-15.lC) and determined that appropriate corrective
actions were included in the procedures. This item is closed.
9.
Licensee Event Report (LER) Review. (92700)
The inspector reviewed the LERs listed below to ascertain whether NRC
reporting requirements were being met and to determine appropriateness of
the corrective action(s).
The inspector's review also included followup
on implementation of corrective action and review of licensee
documentation that all required corrective action(s) were complete.
(Closed) LER 280/86-12, Loss of Boric Acid Flow Path. The issue involved
isolation of a required flow path for maintenance of the boric acid
filter.
The cause of the event was personnel error due to an unusual
lineup of the boric acid system.
Licensee corrective action included an
1--
13
addendum to shift orders to increase operator awareness to operational.
detail.
Also, improved labeling was implemented to minimize recurrence.
The inspector reviewed the addendum.
This item is closed.
(Closed) LER 280/86-14, Inadvertent ESF Actuation.
The cause of the event
and corrective actions are addressed in paragraph 3 of this report under
closeout of violation 280/86-11-01.
This item is closed.
(Closed) LER 280/86-17, Loss of RHR and Actuation of ESF.
This issue
involved loss of power to an emergency bus due to personnel error in
breaker testing.
Corrective action included verification of automatic
actuation of required safety systems, and restoration of normal reserve
station servi'ce power.
Also, a human factors analysis was performed and
the conclusion of the report was that the operator, after being
interrupted from the evolution in progress, became complacent and actuated
the incorrect relay causing the event.
The report recommended that each
control operations technician be reinstructed in the correct methods of
work continuation after interruptions and breaks.
The inspector verified
that corrective actions ~ere accomplished.
This item is closed.
(Closed) LER 280/86-18, ESF Actuation - #3 EDG Auto St"'.rt.
The issue
involved an automatic start of the #3 emergency diesel generator due to
inadequate procedure, which was being used to install additional fuses in
the 3 charging pump control circuits.
Corrective action by the licensee
was to correct the procedure and properly complete the modification. This
item is closed.
(Closed) LER 280/86-19, Spurious Operation of Reactor Trip Breakers .. The
issue involved an automatic opening of the reactor trip breakers during
preparation to withdraw shutdown bank rods in preparation for restart.
Alarms and annunciators indicated that the trip had been generated by
signals which were blocked by permissive (P-7) interlocks.
Further
investigation indicated that the trip signal was generated by a turbine
first stage pressure spike which would unblock permissive P-7. * The
licensee concluded that the sensing line for one of the first stage
turbine pressure transmitters received a sharp blow causing the pressure
spike.
Additional corrective action included testing of permissive P-7
logic relays.
No discrepancies were noted. This item is closed.
(Closed) LER 281/86-07, Manual Reactor Trip Due to High Steam Generator
Level.
The issue involved failure of feed reg valve (FCV-2478) to close
during shutdown.
After shutdown, the valve was disassembled and metal
debris was found which prevented full valve closure.
The valve was
reassembled, tested satisfactorily, and returned to operable status.
This item is closed.
(Closed) LER 281/86-09, Spurious Operation of a Reactor Trip Breaker.
The
issue involved automatic opening of the B reactor trip breaker which was
caused by an electrication improperly jumpering a relay during replacement
of the relay.
Corrective action included proper reinstallation of the
i
..1
14
jumper, and subsequent replacement of the failed relay. The procedure was
reviewed with electrician who improperly placed the relay, and he was *
cautioned to be more attentive to terminal identification.
This item is
closed.
(Closed) LER 281/86-10, Inoperable Charging Pump Component Cooling Water
Pumps.
The issue involved a loss of cooling water supply from both the A
and B charging pump component cooling water pumps due to air being
introduced into the system during maintenance on the A pump.
Corrective
action included venting of the system and return of the pumps to service.
This item is closed.
(Closed)
LER 281/86-11, Service Water Leak in Unit 2 Containment.
The
issue involved service water leaking into the recirculation spray side of
the recirculation spray heat exchanger (RSHX) B due to tube degradation.
The unit was shut down for repair of the heat exchanger.
Eight tubes were
plugged in the RSHX and appropriate testing was conducted to verify tube
integrity. The inspector reviewed the completed maintenance procedure and
also verified that the licensee completed testing on other RSHX to verify
tube integrity. This item is closed.
(Closed) LER 281/86-12, Consequence Limiting Safeguards Relay Failures.
The issue involved failure of the A and B train relays which provide one
of the four actuation signals for containment High High pressure.
The
licensee immediately jumpered the relays to provide for the Safeguards
signal and subsequently replaced both relays and the power supply for* the
relays.
After replacement, the relays were tested and declared operable.
This item is closed.
(Closed) LER 281/86-20, Reactor Trip and Main Feedwater Pipe Failure. The
issue involved the Surry Unit 2 feedwater pipe rupture event of *
December 9, 1986.
The 1 icensee report dated January 14, 1987, entitled
11Surry Unit 2 Reactor Trip and Feedwater Pipe Failure,
11 provided detailed
information on the December 9, 1986 event with a recovery plan and
corrective actions for NRC review and concurrence prior to unit startup.
The results of the NRC Augmented Inspection Team inspection was documented
in report 280;281/86-42 dated February 10, 1987.
The resident inspectors
have continually monitored the recovery process to ensure compliance with
the above reports.
The violation associated with the subject event has
been reviewed and closed per paragraph 3 of this report.
This item is
closed.