ML18139A195
| ML18139A195 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 03/06/1980 |
| From: | Burke D, Kellogg P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18139A192 | List: |
| References | |
| 50-280-80-01, 50-280-80-1, 50-281-80-03, 50-281-80-3, NUDOCS 8005070768 | |
| Download: ML18139A195 (10) | |
See also: IR 05000280/1980001
Text
- *
-8005 070 76 0
e
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTA ST., N.W., SUITE 3100
ATLANTA, GEORGIA 30303
Report Nos. 50-280/80-01 and 50-281/8003
Licensee:
Virginia Electric and Power Company
Richmond, Virginia
23261
Facility Name:
Surry Units 1 and 2
Docket Nos. 50-280 and 50-281
License Nos. DPR-32 and DPR-37
SUMMARY
Inspection on December 17, 1979 - February 4, 1980
Areas Inspected
This routine, announced inspection involved 180 inspector-hours on site in the
areas of plant operations and operating records, plant modification and main-
tenance, followup of LER's and plant security.
Results
Of the four areas inspected, no apparent items of noncompliance or deviations
were found in three areas; two apparent items of noncompliance were found in the
plant maintenance area (Infraction - inadequate maintenance procedures for valve
1-RC-107 control, and Infraction-inadequate review of substitute gaskets used
during maintenance of Unit 1 pressurizer power operated relief valves - para-
graph 6) .
- *
DETAILS
1.
Persons Contacted
Licensee Employees
Virginia Electric and Power Company (VEPCO)
- W. L. Stewart, Station Manager
- J. L. Wilson, Superintendent, Operations
- T. A. Peebles, Superintendent, Technical Services
- R. F. Saunders, Superintendent, Maintenance
R. M. Smith, Superivsor, Health Physics
R. L. Baldwin, Supervisor, Administrative Services
G. Kane, Operating Supervisor
- F. L. Rentz, ResidentQC Engineer
M. R. Kansler, Acting Engineering Supervisor
Other licensee employees contacted during this inspection included control
room operators, shift supervisors, QC, HP, plant maintenance, security,
engineering, chemistry, and administrative personnel.
- Attended exit interview.
2.
Management Interviews
The scope and findings were summarized on a weekly basis with those persons
indicated in paragraph 1 above.
The items of noncompliance were discussed
and prompt licensee action was taken to resolve the specific discrepancies.
3.
Licensee Action on Previous Findings
Not inspected.
4.
Unresolved Items
5.
Unresolved items were not identified during this inspection.
Unit 1 Operations
The inspector routinely toured the Unit 1 control room and other plant
areas to verify that the plant operations were in accordance with the
Technical Specifications (TS) and facility procedures.
Plant logs, main-
tenance records, equipment and instrumentation were also reviewed.
Within
the areas inspected no items of noncompliance were identified.
Specific
areas of inspection and results included the following:
a.
The inspector reviewed and discussed with licensee and NRC personnel,
the Unit 1 "A" reactor coolant pump motor failure and subsequent
- *
b.
c.
d.
e.
-2-
reactor trip from full power which occurred at 1:59 p.m. on December 19,
1979.
The inspector was in the control room when the differential
pressure (DP) between the "A" steam generator and the common steam
header reached 100 psi and automatically initiated Safety Injection
(SI), some 3.5 minutes after the reactor trip.
The inspector verified
that the Emergency Procedures (EP) were properly implemented and that
the Engineered Safeguards (ES) systems and components functioned as
required; the core was adequately subcooled at all times.
Since SI
was not required in this instance, the licensee has initiated an
Engineering Study to determine if the steam header to line DP setpoint
.should be increased to account for steam generator tube plugging.
The
TS 3.7 differential pressure setpoint limit is 150 psi (max.).
The
inspector verified complete NRC review and concurrence prior to Unit 1
restart on January 8, 1980.
Following the discovery of scoring or scratching on the shafts of
large bore (12") B-P hydraulic snubbers, the NRC requested that the
licensee functionally test two of the large bore snubbers which are
mounted on each reactor coolant pump.
Following review of the testing,
the NRC approved Unit 1 restart and interim operation on January 5,
1980.
The snubber reservoir levels are being checked and documented
weekly in accordance with PT 39.3; significant increases in leakage
have not been observed, although some fluid leakage is present.
The inspector observed Unit 1 restart on January 8, 1980;
the pre-
startup checklists and procedures were completed as required, and the
ECP was accurate.
The inspector verified that the steam generator
(SG) low-low level trip setpoints had been conservatively increased
from some 5% to 17% (PT 2.5) to compensate for the generic non-
conservatism identified in the accident analysis for a high energy
line break (see LER 280/79-22).
However, the reduction in the SG
level operating band or range, requires closer control of SG ievels to
prevent a reactor trip during startup. Initial criticality occurred
at 1:07 p.m.
The inspector followed up on the whole body counts (WBC) of certain
employees who worked in Unit 1 containment on December 19 and 20,
1979, following the shutdown: -The licensee identified a calculational
error in the percent maximum permissible body burden (MPBB) of iodine
131 for these employees, corrected the records, and notified the
emplpyees of the correct lower values.
The MPBB ranged from one to
several percent.
The inspector had no further questions.
The inspector noted that the Unit 1 containment particulate and gaseou~r'_..---*-----,*
.
radiation monitors (RM-159 and -160) were out of service for certain
short periods in December and January due to the condensation of
moisture in the sampled air.
The licensee sampled the containment air
...
and monitored the direct radiation monitors in the containment (RM on
manipulator crane, etc.) when RM-159 and -160 were not operable; no
significant increases in radiation levels were observed.
The licensee.
plans to insulate and/or heat trace the sampling system to reduce -
~-
condensation (Open item 280/80-01-03).
. ------ ---~--
- -*
e
e
-3-
6.
Unit 1 Maintenance
The inspector observed and reviewed certain maintenance activities to
assure that they were conducted in accordance with established procedures
and TS requirements.
Areas inspected included the following:
a.
On January 11, 1980, the Unit 1 reactor operator temporarily placed
the excess letdown in service and noted an immediate decrease in
pressurizer level and pressure.
The level decreased from 46% to 39%
and the pressure from 2230 to some 2100 psig.
The operator immediately
isolated the excess letdown and terminated the decrease; the pressure
and level returned to normal.
Subsequent investigation determined
that valve 1-RC-107 from the primary loop drain header (which connects.
to the excess letdown heat exchanger) to the primary drains transfer
tank (PDTT) was not closed.
The valve was previously opened to 'drain
the "A" reactor coolant loop during replacement of the RCP motor, a.nd
apparently not reclosed, although certain station personnel recalled
closing the valve prior to startup.
Since the maintenance operating
procedures (MOP 5.6) for refilling the RCS loops did not contain
appropriate specific steps or instructions to cl-0se 1-RC~107, the
procedures were determined to be inadequate and contrary to Technical
Specification 6.4.A.7, which requires detailed written procedures with
appropriate check-off lists and instructions for maintenance operations
of this type (Infraction - 280/80-01-01).
b.
The inspector reviewed maintenance work performed on the Copes-Vulcan
pressurizer power operated relief valves (PORV's), PCV-1455c and
PCV-1456, which were r~built on December 28, 1979, through January 4,
1980.
The valves had been leaking slightly during previous operations,
and had indication of weeping shortly after the Unit 1 restart on
January 8, 1980.
The PORV's are currently isolated.
While reviewing
Maintenance Reports (MR) S1912041010 and S1912041011, the inspector
noted that the replacement parts used were not listed on the MR.
Review of the storekeeping records indicated that the following
gaskets were used on MR S1912041010 and MR 61912041011:
The
were
Quantity
Stock No.
2
1740493G
1
1740547G
.1
1740514G
above Flexitallic gaskets were
used on PCV-1455C:
Quantity
1
2
Stock No.
1740547G
1740514G
used on
Style
R3-3F
CG-IF
R3-15H
PCV-1456,
Style
CG-lF
R3-15H
the following
__ ,. __ _
* --~*
c.
e
e
-4-
Each PORV required two gaskets, which are stock replacement gaskets
170514G and 1740604G; these gaskets are 1500 psi (min) rated gaskets.
The R3-3F and CG-IF gaskets are rated 300 psi and 150 psi, respectively.
In addition, plant maintenance personnel told the inspector that
additional gaskets were used and not properly logged out of the store-
room.
At least two 1740850, style CG-6F (600 psi) gaskets were used,
and may have been one of the final gaskets installed in each PORV.
The use of the lower rated gaskets in the PORV's once the stock of
specified replacement gaskets was depleted, is contrary to plant
procedures for design control (QA manual, Section 3) of safety-related
components and is an infraction (280/80-01-02).
After reviewing plant
records and documentation, the inspector could not determine which
four of the above gaskets are in the PORV's.
The inspector also noted
that the licensee considers the majority of the plant gaskets to be
"off-the-shelf" type items which are not specifically identified',
accepted and tagged by the QA staff'.
The inspector is reviewing this
matter with NRC personnel (open item 280/80-01-04).
The* inspector observed additional maintenance work and testing such as
the removal and cleaning of the condensate filter screens during
operation and periodic testing of the recirculation spray pumps.
Within the areas inspected, no further items of noncompliance were
identified.
7.
Followup on BFD Relay Testing
8.
Prior to Unit 1 restart on January 8, 1980, the inspector reviewed licensee
actions taken in regard to IE Bulletin 79-25, "Failures of Westinghouse BFD
Relays in Safety-Related Systems." The inspector reviewed BFD relaty testing
performed in March of 1976 under ST-41, which verified relay drop-out times
of 25 msec or less.
Of the 89 Units 1 relays tested, three relays failed
the 25 msec test time by a few milliseconds and were replaced.
No relay
sticking or degradation has been observed since the testing.
Three new
style 5072A49 relays have been installed, but have not been tested for
No failure of the three relays would interfere with or affect
the engineered safeguards systems operability. All BFD relays will be
tested and/or replaced in accordance with IEB 79-25, which remains open.
Implementation of TMI-2 Lessons Learned Short-Term Recommendations (NUREG -
0578)
The licensee completed initial installation of many of the NUREG-0578
short-term recommendations (section 2) on Unit 1 prior to the January 8,
1980, restart.
Certain items such as the high range radiateion monitors
were subsequently installed as the materials became available.
Specific
items inspected included the following:
a.
Modifications to the pressurizer heater power supplies were not neces-
sary since certain heater banks were on the emergency power system.
Direct position indication of pressurizer safety and power-operated.~:
* -* -
--* - ---
.
9.
e
e
-5-
relief valves (PORV's) was accomplished by installing an acoustic
monitoring system to monitor each valve (Design change 79-S54).
In
addition to the valve tailpipe temperature monitors, the PORV's have
limit switches to monitor valve (operator) position, however environ-
mental qualification data is currently not available for these limit
switches.
The NRC will review the seismic and environmental qualifi-
cation data on the acoustic monitoring systems when available.
An
additional source of safety-grade motive power (emergency air) has
been installed on the PORV's (DC-79-S53).
b.
The Unit 1 auxiliary feedwater flow indication instrumentation was
modified by DC 79-S55 to provide each channel with power supply diversity
and physical independence.
The electrical modifications are being
reviewed by the NRC. The inspector noted that*although certain nesting
was performed after the modifications, the flow instrumentation was
not recalibrated at this time.
Since certain flow channels such as
FT-lOOA had not been calibrated since 5/12/78, the licensee stated
that the instrumentation would be calibrated during the next outage.
c.
Two core cooling monitors (saturation meters) were installed in Unit 1
to monitor RCS pressures and temperatures.
The control room instru~
mentation monitors the auctioneered high in-core T/C or loop RTD
temperature and the wide range (PI-1-402) or protection channel
(PI-1-457) pressure instrumentation pressure for subcooling.
The NRC
is reviewing the instrumentation and inputs.*
d.
Piping and systems outside containment, such as the outside recirculation
spray and low head safety injection pumps and piping, are periodically
leak tested.
e.
High range radiation monitors have been installed on the process vent,
ventilation vent, and main steam relief lines (DC 79-59), and procedures
have been developed to improve post-accident sampling capabilities
(CP-66 and 66A).
NRC review of these items is in progress.
The inspector noted that updated procedures or instructions have been
issued to operate and utilize the above new equipment.
Detailed
procedure review will be conducted at a later date.
Unit 2 SGRP
The Unit 2 steam generator replacement project was completed on December 31,
1979; testing of various Unit 2 systems is in progress.
During routine
tours of Unit 2, the inspector observed certain maintenance work and testing;
no items of noncompliance were identified.
While reviewing ETA 50006-P-4-ULC,
"Primary System Hydro and Flush", the inspector noted that the flush was
through 2-DG-78, but could not locate this valve on the P&ID FM-83B. The
licensee determined that the valve tagged 2-DG-78 is.actually 2-DG-82, and
is re tagging the valve.
.* __
v
e
e
-6-
10.
Review of Reportable Occurrences
The inspector reviewed the Reportable Occurrence (RO) reports listed below
to ascertain that NRC *reporting requirements were being met and to determine
the appropriateneess of corrective action taken and planned.
Certain
Licensee Event Reports (LER) were reviewed in greater detail to verify
corrective action and determine compliance with the Technical Specifications
and other regutatory requirements.
The review included examination of log
books, internal correspondence and records, review of SNSOC meeting minutes,
and discussions with various staff members.
Within the areas inspected, no items of noncompliance or deviations were
identified.
a.
LER 28-/79-08, concerned an apparent under power condition on the 480
VAC emergency bus that could exist during accident conditions with or
without loss of off-site power.
The condition was identified during a
load study by the AE.
Additional transformers and components were
installed and are operable in Unit 1; the Utit 2 modifications and
testing are in progress and will be reviewed prior to startup.
This
LER is closed.
b.
LER 280/79-10 identified discrepancies in the location and configuration
of certain pipe supports discovered during an investigation of the
as-built piping conditions.
The discrepancies identified have been
corrected as required by the NRC Show Cause Order of March 13, 1979,
and IE Bulletin 79-14.
Reanalysis and corrective action (if required)
continues on piping systems outside containment (e.g:
component
cooling water).
This LER is closed.
c.
LER 280/79-11 has been supplemented and the correct codes applied.
This LER is closed.
d.
LER 280/79-20 concerned the abnormal degradation of battery charger
2B2 internal power cable insulation.
The dried and cracked insulation
was determined to be the result of environmental conditions.
Reduced
air flow to the battery chargers 2Bl and 2B2 may have lead to increased
internal temperatures and teh consequent deterioration of the cable
insulation.
No insulation deterioration was noted on the stations six
similar battery chargers.
The air flow to all the battery chargers
was increased, and the units are periodically inspected.
This LER is
closed.
e.
LER 280/79-21 addressed the failure of MOV-SW-104C to fully close
during exercise.
This valve is normally open and remains open during
accident conditions; the valve operated properly after manual stroking.
This LER is closed .
f.
LER 280/79-22 concerned a non-conservatism in accident analysis:
the
steam generator level instrumentation indications would increase if a
-~--. *-** - ..... - __ ::--__ .._ ... -.:.*--*-**---*---*---**-*-~*---~- .....
e
e
-7-
rise in containment temperature decreased the reference leg water
column density.
To compensate for the temperature effects on the SG
reference leg, the low-low level trip setpoint has been increased to
15% minimum (see ~ara. 5.c).
This LER is closed.
g.
LER 280/79-23 concerned the discovery of circumferential crack indications
in the steam generator feedwater piping nozzle to reducer areas.
All
indications were repaired and inspected in accordance with IE Bulletin
79-13.
This LER is closed.
h.
LER 280/79-24 addressed the identification of 64 of the approximately
1800 plant penetration fire barriers as potentially inoperable, during
initial conduct of Periodic Test 24.11 which inspects each fire barrier.
The fire barriers were repaired as required.
This item is clos~d.
i.
LER 280/79-25 concerned a late report of RCS activity, which rose
slightly above 1.0 uci/gm for about 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after a reactor trip
from 19% power.
The LER reference to TS 6.9.1 was incorrect.
This
LER is closed.
j.
LER 280/79-27 addressed overstress in the three RTD bypass lines and
in a safety injection line inside containment.
Reanalysis and modifi-
cations were made to the lines as required.
The LER was submitted one
day late.
This LER is closed.
k.
LER 280/79-28 concerned the failure of 10 of the 209 anchor bolts
tested in accordance with IE Bulletin 79-02.
The failures were reviewed
and corrected as neces&ary.
This LER is closed.
1.
LER 280/79-30 addressed the failure to properly reset the alarm setpoints
on radiation monitors RM-CC-105&106, which monitor the component
cooling water.
Radiation monitoring equipment is now checked daily.
This LER is closed.
m.
LER 280/79-31 concerned a liquid waste release whose isotope activity
summation exceeded the radiation monitor setpoint of 1.5E-3uci/cc
without tripping.the alarm (actual summation was 1.6E-3uci/cc).
The
acutal summation is calculated since the detector is calibrated for
one* isotope and administrative control is necessary to prevent isotope
releases which the radiation monitor may not see or detect.
The
release did not exceed TS 3.11.A or 10 CFR Part 20 limits.
This LER
is closed.
n.
LER 280/79-32 addressed a Westinghouse identified non-conservatism in
the analysis which could affect the calculated peak clad temperatures.
Reanalysis for Surry 1 determined that sufficient margin existed to*
assure that the PCT limit of 2200°F was not reached.
Unit 2 will be
evaluated prior to startup and FQ reduced if necessary.
This LER is
Closed.
.*
, .......
- , .**. 'IJ.
o.
p.
q.
r.
s.
t.
u.
v.
w.
e
e
-8-
LER 280/79-33 concerned the failure of service water valve 1-SW-107
which supplies cooling water to the charging pump seal and oil coolers;
the redundant train was operable, and the valve was immediately repaired.
Degradation of these 1.5 and 2 inch valves has been identified as a
generic problem at Surry, and all bronze ball valves are being installed;
the Unit 2 installation is complete and the Unit 1 changeout is in
progress.
This LER is closed.
LER 280/79-34 addressed a non-conservative design feature discovered
in trip valve TV-SV-102 logic which diverts SJAE discharge into contain-
ment when the SJAE radiation alarm is actuated.
A high containment
(CLS) pressure signal closes TV-SV-102 to isolate containment, however,
if the high CLS signal is reset and the radiation alarm is still
present, the valve will reopen.
The Emergency Procedures were revised
to assure TV-SV-102 is closed and will not reopen prior to resetting
CLS. This LER is closed.
LER 280/79-35 concerned a low current on heat tracing circuit 2B
(Panef 8).
The redundant circuit was operable.
A supplemental LER
has been submitted to address the generic implications of heat tracing
tape failure; corrective actions are in progress on the BAST pump
leaks and the heat tracting systems.
This LER is closed.
LER 280/79-36 addressed an error in the containment spray effectiveness
values used for containment depressurization values in accident analysis.
Until the analysis or modifications are complete, interim, more conser-
vative values for the maximum allowable service water temperatures
have been implemented.* Until 1 is operating within the revised limits,
aided by the seasonally cool weather.
This LER is closed.
LER 280/79-37 concerned a failure in the new type heat tracting tape
due to overheating.
The heat tracing.circuit was redesigned to imple-
ment more local control over each circuit.
Systematic evaluation
continues.
This LER is closed.
LER 280/79-38 addressed the inoperability of one train of the RWST
chemical addition tank discharge due to manual isolation of MOV-CS-102B;
valve MOV-CS-102A was verified operable.
This item was discussed in
IE Inspection Report 280/79-67 and is closed.
LER 280/79-39 concerned a heat tracing tape failure; the redundant
circuit was operable.
This LER is closed.
LER 280/79-40 addressed a component cooling water p1p1ng (support)
overstress condition.
Modifications were made to relieve the overstress
conditions.
This LER is closed.
LER 280/79-41 addressed the "A" RCP motor fault, the reactor trip, and
subsequent safety injectin which occurred on 12/19/79.
See para. *'S .--a_;.
of this report.
This LER is closed.
'
(
-9-
x.
LER 280/79-42 concerned an overstress condition on a component cooling
water pipe hanger; the condition was corrected.
This LER is closed.
y.
LER 280/79-43 addressed the temporary inoperability
recorder RR-175; the drive wheel string disengaged.
monitors and alarms were operable, and the recorder
LER is closed.
of radiation
The radiation
repaired.
This
In addition to the above, the inspector reviewed the licensee's October 18,
1979, response to the D. G. Eisenhut letter of 21, 1979, which requested
information on the potential for multiple equipment failures and challenges
to safety systems at PWR's.
Licensee review of trip reports and operating
reports indicated that no common mode failures occurred which prevented
operability of safety systems.
A review of periodic test (PT) procedures
determined that adequate provisions have been incorported into the PT's to
prevent challenges to.the safety-related systems; the licensee also revised
existing administrative procedures to re-emphasize the importance of avoiding
challenges to the protective features of the facility.
Some events which
might be considered multiple failures or challenges to safety systems
occurred some years ago due to inadequate procedures or personnel error;
however, the problems were revised at that time and corrective actions were
taken.
This item is closed.
6.
Plant Physical Protection
The inspector verified the following by observation:
a ..
Gates and doors in protected and vital area barriers were closed and
locked when not attended.
b.
Isolation zones described in the physical security plans were nqt
compromised or obstructed.
c.
Personnel were properly identified, searched, authorized, badged and
escorted as necessary for plant access control.
Within the areas inspected, no items of noncompliance were identified.
-
~ _;...