ML18142A176
| ML18142A176 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 10/04/1984 |
| From: | Burke D, Marlone Davis, Elrod S, William Orders NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18142A177 | List: |
| References | |
| 50-280-84-24, 50-281-84-24, 50-281-84-25, NUDOCS 8502060332 | |
| Download: ML18142A176 (6) | |
See also: IR 05000280/1984024
Text
Report Nos. :
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTA STREET, N.W.
ATLANTA, GEORGIA 30303
50-280/84-24 and 50-281/84-24
Licensee:
Virginia Electric and Power Company
Richmond, VA
23261
Docket Nos.:
50-280 and 50-281
License Nos.:
Facility Name:
Surry 1 and 2
Inspection Conducted:
August 1-31, 1984
Inspectors:
/) *. /
~A.Je_
D/0. ~6rkefenior Resident Inspector
dJ-.,L. /'1dle
W. Orqyrs/;Senior Resident Inspector
M. J. ~4, Resident Inspector
Approved by: &/ 6~L.IZ--
- r S. 1:lrgµ:f, s'ection Chief
Division of Reactor Projects
/f./&uAe--
SUMMARY
Date Si ned
- /ILi,Utl
bite Sighed
Scope:
This inspection involved 150 inspector-hours on site in the areas of
plant operations and operating records, plant maintenance and surveillance, plant
security, and followup of events.
Results:
In the areas inspected, two violations were identified; failure to
follow procedures during
replacement of an
relay -
paragraph 6.d;
10 CFR 50.59 safety evaluation not performed/documented for change to facility as
described in FSAR - paragraph 5.e .
- aso2060332 asorf6
PDR ADOCK 05000280
G
REPORT DETAILS
1.
Licensee Employees Contacted
R. F. Saunders, Station Manager
D. L. Benson, Assistant Station Manager
H. L. Miller, Assistant Station Manager
D. A. Christian, Superintendent of Operations
M. R. Kansler, Superintendent of Technical Services
H. W. Kibler, Superintendent of Maintenance
D. Rickeard, Supervisor, Safety Engineering Staff
S. Sarver, Health Physics Supervisor
R. Johnson, Operations Supervisor
R. Driscoll, Director, QA, Nuclear Operations
Other 1 i censee emp 1 oyees contacted included contra 1 room operators, shift
technical advisors (STAs), shift supervisors, chemistry, health physics,
plant maintenance, security, engineering, administrative, records, and
contractor personnel and supervisors.
2.
Exit Interview
The inspection scope and findings were summarized on a biweekly basis with
certain individuals in paragraph 1 above.
3.
Licensee Action on Previous Enforcement Matters
This subject was not addressed in the inspection.
4.
Unresolved Items*
Unresolved items were not identified during this inspection.
5.
Operations
Unit 1 and 2 operations were inspected and reviewed duri~g the inspection
period.
The inspectors routinely toured the control room and other plant
areas to verify that plant operations, testing and maintenance were being
conducted in accordance with the facility Technical Specifications (TS) and
procedures.
Within the areas inspected, one violation was identi,fied
(paragraph 5.e).
Specific areas of inspection and review included *the
following:
a.
Review was made of annunciated alarms in the control room and
inspection of safety-related valve, pump, and equipment alignments on
the consoles and in the plant.
- An Unreso 1 ved Item is a matter about which more information is re qui red to
determine whether it is acceptable or may involve a violation or deviation.
2
b.
Unit 1 began the reporting period at 80 percent power.
Maximum power
is limited to 80 percent due to immovable control rod B-6 (see previous
report, 50-280/84-20 and 50-280/84-21).
Unit 1 operated at power for
the duration of the reporting period; no trips or shutdowns occurred.
c.
Unit 2 began the reporting period operating at full power.
On
August 9, 1984, a reactor trip occurred from the comp 1 et ion of the 2
out of 3 logic matrix on overpower delta T protection.
Prior to the
event, the unit was at ful 1 power with rod contro 1 in manua 1, and
technicians were troubleshooting a failed delta flux indicator (NI-43).
Improper use of an ungrounded 120 VAC power cord with a digital volt-
meter to obtain detector current measurements caused one of the control
power fuses to fail on Nuclear Instrument NI-43.
The loss of power to
the drawer resulted in completion of the NIS dropped rod protection
circuitry, causing a turbine runback to approximately 70%.
Immediately
following the dropped rod runback, a series of overpower (OP) delta T
runbacks ramped the turbine down to 40% of full power.
The initiation
of the OP delta T trip function was caused by a decreasing OP delta T
setpoint due to negative delta flux (as sensed by nuclear instrumenta-
tion).
Approximately two minutes after the start of the runback, a
reactor trip occurred from overpower delta T protection.
Immediately
after the reactor trip, an overtemperature delta T trip was received.
Following the trip, all safety systems functioned normally except for
MOV-FW-251C (auxiliary feedwater pump discharge valve) which would not
remain closed after the operator manually closed it. The cause of the
valve malfunction was a timing control relay which was subsequently
replaced.
ALER will be submitted on the event and additional testing
will be conducted during an upcoming outage.
Instructions for proper
setup of test equipment are being revised.
The unit was subsequently
restarted and returned to power operations. The unit operated at power
for the remainder of the reporting period.
d.
At 10:50 a.m., on August 9, 1984, two men, an electrician employed by a
contractor and a mechanic employed by the licensee, were electrocuted
in a non-radiological accident in the turbine building.
Two contractor
employees were drilling anchor bolt holes into a turbine building wall
electrical duct bank near the emergency switchgear room to support Fire
Protection System electrical conduits.
The electrician was knocked
unconscious when the drill he was using struck one of the 4160 volt
reserve station service transformer power lines in the concrete duct
bank.
No reactor or electrical trips occurred.
The electrician was
taken by helicopter to Norfolk General Hospital and was pronounced dead
at 12:25 p.m.
The electrician
1s helper was taken to a Smithfield
physician for treatment of an ankle injury.
When the first aid team responded to the accident, one of the team
members involved in the rescue attempt contacted the drill which was
still embedded in the 4160 volt cable.
The mechanic was pronounced
dead at the scene by the Surry County medical examiner.
The licensee de-energized the reserve station service transformer and
powered affected loads with the emergency diesel generators while an
evaluation of the safety, damage,
and repairs were made.
An
3
investigation by the licensee and the Virginia Department of Labor and
Industry (OSHA) is in progress.
e.
While reviewing the Unit 1 and 2 component cooling water (CCW) system,
the inspectors noted that the CCW outlet trip valve from the reactor
coolant pump thermal barrier was not numbered on the CCW valve opera-
ting numbers print FM-72A.
Following investigations, the licensee
stated that the three air-operated trip valves (per-unit) on the RCP 1 s
were never installed in accordance with an October 11, 1972 proposal by
the AE and Design Change 73-106; however, the ori gi na 1 hi fl ow trip
signal to CCW common trip valve TV-CC-107 (and 207 on Unit 2) was
defeated.
Relief valves are installed on each RCP thermal barrier CCW
line and discharge inside containment, and annunicator alarm procedures
direct the reactor operators to close TV-CC-107 (and 207) on high flow
indication. The licensee also stated that the check valves on the CCW
inlet lines to the RCP thermal barriers, described on FM-72A, were
never installed.
In addition, the air-operated trip valve on the CCW
outlet from the primary drain coolers (HCV-CC-114) was blocked open
several years ago due to spurious but frequent trip valve closures.
The trip valves are described in Section 9.4 of the Surry updated FSAR
as follows:
11 In the event that a leak occurs in the RCP thermal
barrier cooling coil, an alarm annunciates in the control room and the
high pressure reactor coolant is safely contained by closing the
appropriate stop valve.
A high cooling water outlet flow signal from
either the thermal barrier cooling header, the excess letdown heat
exchanger, or the primary drain cooler automatically closes the isola-
tion [trip] valves.
11
The removal of the automatic trip valve isolation
function on high outlet flow from the RCP thermal barriers and primary
drain coolers constitute a change in the facility as described in the
FSAR and thus requires a written safety evaluation in accordance with
Contrary to these requirements, a written safety evalua-
tion to determine that the change did not involve an unreviewed safety
question was apparently not performed or documented, and is a violation
(280
and
281/84-24-01).
Subsequent
review determined that an
unreviewed safety question did not exist concerning this change, and
that NRC approval of the change was not required.
The reactor coolant
is safely contained with the existing components and procedures.
6.
Surveillance and Maintenance Activities
During the reporting period, the inspectors reviewed various surveillance
and maintenance activities to assure comp 1 i ance with the appropriate
procedures and TS, and verified the operability of major plant systems.
One
violation was identified in the electrical maintenance area (paragraph 6.d).
Inspection areas included the following:
a.
Inspections of the auxiliary building, subsurface drain systems, cable
penetration areas, switchgear and cab 1 e rooms, outside areas, steam
safeguards and the turbine building were conducted to verify equipment
operability and alignment.
No violations were identified in the areas
insp.ected.
4
b.
The inspectors reviewed the control room logs and operations daily and
reviewed the reactor coolant system leak rates on a daily bas-is.
Severa 1 LCOs in Section 3 of the TS were a 1 so verified on a periodic
basis to ensure compliance with the requirements.
The inspectors also
verified that at least two Senior Reactor Operators (SRO) were on duty
at all times during reactor operations, and at least one of the SR0 1 s
was in the reactor control room at all times.
c.
The inspector requested that the periodic Units 1 and 2 pressurizer
power operated relief valves (PORV 1s) stroke testing be discontinued
during power operations in accordance with NRC policy recommendations.
The licensee is revising Periodic Test Procedure PT 2.26 to limit PORV
stroke timing and testing to outages and operability verification
requirements.
The MOV b 1 ock va 1 ves wil 1 continue to be peri odi ca lly,
tested in accordance with TS requirements.
d.
On the evening of August 21, 1984, during replacement of a failed
Westinghouse BF relay in the Unit 2 RPS logic during power operation,
the following occurred when one of the relay leads was lifted:
(1)
Motor driven auxiliary feedwater pump 2~FW-P-3B started.
(2)
The three steam generator blowdown trip valves outside containment
went closed.
(3)
Both source range NI
1 s (N-31 and 32) were reenergized.
(4)
Five first-out reactor or turbine annunciator alarms were
activated on the annunciator panels.
The above conditions were reset and corrected within a few minutes.
The licensee did not promptly report the actuation of the Engineered
Safety Feature (AFW pump 38) as required by 10 CFR 50.74(b)(2)(ii), but
subsequently determined that a 1 though the maintenance and procedures
were preplanned, the pump start was not.
The event was reported to the
NRC using the ENS phone on the morning of August 22, 1984.
A written
LER wi 11 a 1 so be submitted.
Although procedures, prints, and jumper
logs were used to verify the electrical wire or lead removal, the
common lead lifting led to the loss of additional BF relays due to a
misunderstanding of the series
11da i sy-cha i n
11 wiring i nsta 11 at ion and
inadequate control wiring diagram electrical prints.
While reviewing
the event, the in~pector noted that no reactor trip breakers opened,
although the RPS 1 ogi c indicated that the 'B I reactor trip breaker
should have opened.
An approved electrical jumper had been installed
to bypass the
I B
I train reactor trip 1 ogi c; however, the procedures
used did not specify this jumper or bypass, which is a violation of
procedures ADM-29.5 and EMP-C-RT-24 (281/84-24-02).
The electricians
installed *the jumper to prevent the
18 1 reactor trip breaker from
inadvertently cycling during the maintenance.
The
1A1 train remained
operable, and the 18 1 reactor trip breaker was closed per EMP-C-RT-24.
The licensee
1 s Safety Engineering Staff is performing a failure
analysis on Westinghouse BF and BFD relays which have failed recently.
(Open Item ?80/84-24-03).