ML18153C562
| ML18153C562 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 02/20/1991 |
| From: | Holland W, Tingen S, York J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18153C560 | List: |
| References | |
| 50-280-90-41, 50-281-90-41, NUDOCS 9103120090 | |
| Download: ML18153C562 (12) | |
See also: IR 05000280/1990041
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTA STREET, N.W.
ATLANTA, GEORGIA 30323
Re~ort Nos.:
50-280/90-41 and 50-281/90-41
Licensee:
Virginia Electric and Power Company
5000 Dominion Boulevard
Glen Allen, VA
23060
Oocket Nos.~
50-280 and 50-281
License Nos.:
Facility Name:
Surry 1 and 2
Inspection Conducted:
December 30, 1990 through January 26, 1991
J. W. York, Resi
Inspector
S. ~in~, R~nff.ctor
Accompanying
Morris Branch
A. B. Ruff
Approved
?
SUMMARY
Scope:
..7-/,;. .oh/
Date ;Signed
~oh/*
ate 'Signed
This routine resident inspection was conducted on site in the areas 6f plant*
operations, plant maintenance, licensee event repoft closeout, and action on
- previous inspection findings.
During the performance of thi~ inspection, the
resident inspectors conducted review of the licensee's backshift or weekend
operations on January 15, 17, 20, and 24.
Results:
In the surveillance functional area, the failure to correctly classify service
water pumps 1-VS-P-lA, B, and C and chilled water pumps 1-VS-P-2A, B, and C in
accordance with Regulatory Guide L26 is identified as a violation (paragraph
3. d).
In the surveillance functional area, a non-cited violation was identified_ for
---*- -fai~lu-re- to-p*erform-cont-afnment-Type-:-B-and--G-leal<-rate.-testing in accordance
with the Appendix J specified test interval of two years (paragraph 5.a).
9103120090 910220
ADOQ'i. 05000280
G
I
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I
1.
Persons Contacted
Licensee Employees
REPORT DETAILS
- W~ Benthall,.Supervisor, Licensing
- R. Bi.lyeu, Licensing Engineer
R. Boles, System Engineer
. *S. Burgold, Supervisor, Instrument and Control
D. Chiistian, Assistant Station Manag~r
E. Cosby, System Engineer
- J. Downs, Superintendent of Outage and Planning
. D. Ericksori, S~perintendent of Health Physics
- D. Hart, Supervisor, Quality Assurance
- G. H~nnicutt, Assistant Shift Supervisor
- R. Gwaltney, Superintendent of Maintenance
- M. Kansler, Station Manager
T. Kendzia, Supervisor, Safety Engineering
- J. Kin, Corporate, Inservice Test Engineer
J. McCarthy, Superintendent of Operations
- A. Price, Assistant Station Manager
- S. Poage, Quality Assurance Performance Coordinator
E. Smith, Site Quality Assurance Manager
- T. Sowers, Superintendent of Engineering
S. Stanley, Supervisor, System Engineering
Other licensee employees contacted included control room operators, shift
technical advisors,_ shift supervisors and other plant personnel.
NRC Personnel
W. Holland, Senior R~sident Inspector
- S. Tingen, Resident Inspector
J. York, Resident Inspector
- A. Ruff, Project Engineer
M. Branch, Senior Resident Inspector - Watts Bar
- Attended exit interview.
Acronyms and initialisms used thro~ghout this report are listed in the
last paragraph.
2.
Plant Status
Unit 1 began the reporting period at power.
The unit operated at power
for the duration of the inspection period.
_: -------- - -Uni-t, 2 -began -t-he -repor-t-i-ng-:-per-i-od- -at --2%- power~-- -on- D*e*cember -29-; -the-crn;-t ____ - -- - --
was ramped up to 90% power and operated at this power for the duration of
2 .
the inspection period.
The unit was at 2% power in order to replace
- degraded jumper straps on the C phase isolation bus duct .which is
discussed in Inspection Report,50-280,281/90-39.
The tinit was limited to
90% because control fod MlZ was stuck.
This issue is also discussed in
Inspection R~port 50~280~281/90-39~
3.
Operational Safety Verification (71707 & 42700)
a.
Daily Inspections
The inspectors conducted daily inspections in the following areas:
control room staffing, access, and operator behavior; operator .
adherence :to approved procedures, TS, and LCOs; examination of panels
containing instrumentation and other re~ctor protection system
elements to determine that required channels are.operable; and review
of control room operator logs, operating orders, plant*~eviation
reports, tagout logs, temporary. modification *logs, and tags on
components to verify comp 1 i ance with approved procedures.
The
inspectors also routinely accompanied station management on plant
tours and bbierved the effectiveness of their influence on a~tivities
being performed by plant personnel.
No discrepancies were noted.
b. * ~eekly Inspections
c.
The inspeitors conducted weekly inspections in the following areas:
operability verification of selected ESF systems by valve alignment,
breaker positions, condition of equipment or component, and
operability of instrumentation and support items essential to system
actuation or performance.
Plant tours were conducted which included
observation of genera 1 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
inspectors routinely noted the temperature of the AFW pump ~ischarge
piping to ensure increases in temperature were being properly
monitored and evaluated. by the licensee.
No discrepancies were
noted.
Biweekly Inspections
T~e inspectors conducted biweekly inspections iri the following areas:
verification review and walkdown*of safety-related tagouts 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 lineups; and verification that
notices to workers are posted as required by 10 CFR 19.
- --* ._:_ -* -- -
In December-1-998 ,-the...:.*inspectors-noted*-that--durfog-routi11e-*testi n-g-*o-f --- - ---- -
the ED Gs, the fuel oil filters were fouling at a higher rate than
I
i I ____
_
'
.
.3
nonnal.
T_he fouling was detected by high fuel oil filter inlet
,
pressure during EOG operation.
In December and early January~. fuel *
- oil samples were obtained from the #1 EDG fuel oil filters, the three
EDG day tanks, and both. EOG underground storage tanks.
Laboratory
.analysis of these samples indicated that clear lacquer like flakes
were present in samples obta*i ned from the #l EDG fi 1 ters, day tanks,
-and in one of the underground storage ta.nks.
The 1 aboratory was
unab 1 e to detenni ne the composition of the 1 acquer 1 i ke flakes.
,On January 14, the licensee performed an SE {9i-004) which addressed
the effects of fouling fuel oil filters on operations of the EDGs.
The SE indicated that the fuel oil problem effected filter
performance and that with a duplex filter on the EDGs, a fouled
filter could be replaced on line, after first switching. to _the
standby filter.
The SE indicated. that this operation was in
a'ccordance with the manufacturers I
specification and that* the
s~itching and replacement evolution had been demonstrated on two of
the EDGs.
.
.
The SE di~ not address the posiibility that the fuel filters for the
diesels that drive the ESW pumps may be subject to this same problem.
The in~pectors inspected the ESW diesels and determined that the fuel.
oil filters were single filters and not duplex strainers.
The
inspectors were concerned that the fuel oil supply for the ESW
di~sels may contain similar lacquer flakes as found in the EDG fuel
oil supply.
The inspectors were informed that the fuel oil for the
ESW diesels was obtained from a different supplier than the EOG fuel
. oil and that ESW diesels had been operated monthly without any
apparent fuel filter fouling.
In order to resolve the issue of the
possibility of the ESW diesel fuel filters fouling, the licensee
removed and inspected the A ESW diesel fuel filter.
The visual
results bf this inspection indicated that the fuel filter was clean;
A sample from the fue 1 oil filter was sent to a 1 aboratory for
_analysis.
As a precautionary measure, the licensee staged extra ESW
diesel fuel filters in case fouling did become a problem.
In order to determine the cause of the EDG filter fouling, . the
licensee has sampled the EOG fuel oil base tanks; day tanks,
underground storage tanks and the aboveground storage tank.
The
licensee has also sampled the ESW diesels, security diesel, ISFSI *
diesel, and fire pump diesel fuel oil storage tanks.
All fuel oil
sa~ples have been sent to laboratories for analysis.
The licensee
has not received all the results of the fuel oil sample analysis,
however, preliminary results indicate that all sample are withi_n ASTM
specifications. Preliminary results indicate that the probable cause
rif the clear lacquer flakes is a mixture of LTSA and AFFF.
LTSA
contains an amihe, and AFFF contains a polymer.
Combining a.polymer
with an amine forms a lacquer like substance.
LTSA is routinely
added to the fuel oil by station personnel to control corrosion.
In
_____ . ___ - . Ju ne---1-9-90,- .Al:-f-1:- -(-1 e-s s----tha.n -30-- -ga 1-1 ons-) -wa-s-un-i nten-t-iona 1-ly--'inj ee-ted-~---- - --
into the EDGs'. aboveground fuel oil storage tank (210,000 gallon
.,~;
4
capacity tank).
The purpose of the AFFF system is to combat a fuel.
oil fiie in the aboveground storage tank; The aboveg~ound storage
tank is only a sciurce*of fuel oil fcir the EDGs.
LTSA and AFFF have
also been identified as surfactants.
Surfactants tend to coat
filters with a substance that increases the filter's fouling -rate.*
The li~ensee has discontinued the use of LTSA and is fnvestigating
methods for removing fuel oil contaminants from all EOG fuel storage
tanks.
This issue will be monitored during the upcoming inspection
period.
d.
Other Inspectfon Activities
Inspections included areas in the Units 1 and 2 cable vaults, vital
battery rooms, steam safeguards areas, emergency switchgear rooms,
diesel generator rooms, control room, auxiliary building, Unit 1
containment, cable penetration areas, independent spent fuel storage
facility, low level intake structure, and the safeguards valve pit
_ and pump pit areas. RCS leak rates were reviewed tci ensure that
detected or suspected leakage from the system was. recorded,
investigated, and evaluated; and that appropriate actions were taken,
if required.
The inspectors routinely independently calculated .RCS
leak rates using the NRC Independent Measurements Leak Rate Program
(RCSLK9).
On a regular basis~ 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 opera bi 1 i ty and cal i bra ti on frequency were
verified .
. On January 16, SW pump 1-VS-P-18 vibration levels significantly
increased.
The pump was secured within several hours after noting
the increased vibration, butwas not declared inoperable.
On January
19 the pump was placed in operation in order to investigate the
source of vibration.
The pump was then declared inoperable based on
excessive vibration.
The pump was replaced the following week.
Inspection Report 50-280,281/90-30 discussed an October 1990, similar
occurrence where the vibration of SW pump 1-VS-P-lC increased
significantly.
. This pump was also secured but not declared
The pump was eventually replaced.
Inspection Report
50-280,281/90-30 also stated that SW pumps 1~vs-P-lA, Band C were
not in the licensee's IST program because they were not installed in
an ASME Class 1, 2 or 3 system.
There are three SW pumps (1-VS-P-lA, B, and C) that supply cooling
water to three chiller units.
Each chiller also ~as a chilled water
pump (1-VS-P-2A, B, and C) that circulates chilled water to air
handling units.
The purpose of this system is to maintain
temperature in the contro 1 room * and Units 1 and 2 emergency
switchgear and relay rooms.
The system is designed to maintain
temperature in a specified band to ensure the proper operation of
___ i11st_rumentation._ -.,TS )_,_23_ r.equ.i.res._tha-t-aJJ_ -th~ee- ch-i-ller-un-it-s-be--- -- - - -
operable when the .Plant temperature is above cold shutdown
J' .
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e.
5
conditions.
The inspectors consider that this system is required to
mitigate the consequences of an atcident and to maintain ~ither unit
in hot shutdown conditions for an extended period of time.
Regulatory Guide 1.26, Quality Group Classifications and Standards
for Water, Steam, and Radi cacti ve-Waste-Conta i rii ng Components of
Nuclear Power Plants, states that cooling.w~ter systems important to
safety that are designed *for functioning of components and systems.
important to safety such as reactor cool ant pumps, diesels, and
control room be clas*sified as ASME Class 3 components.
The
licensee's Topical Report,* VEP l-5A, states that the licensee's
Quality Assurance Program meets or exceeds the applicable guides and
standards described in Regulatory Guide 1.26.
Based on the safety
importance of SW pumps 1-VS-P-lA, B, and C a~d chilled water pumps
1_..;.VS-P-2A, B, and C w.ith regards to maintaining contro_l room
instrumentation temperatures within a specified band and the
licensee'*s Topical Report that co1T111its to Regulatory Guide 1.26
guidelines, the inspectors consider the control room chiller* SW and
chilled water pumps are improperly classified as non Class 3 compo-
nents and therefore erroneously omitted from the 1ST program.
TS 4.0.3.a tequires inservice testing of ASME Code .Class 1, 2 and 3
pumps and valves be performed. in accordance with Section XI of the
ASME Boiler and .Pressure Vessel Code.
Failure to correctly classify
SW pumps 1~vs-P~lA~ B, and t and chilled water pu~ps 1-VS-P-2A, B,
and C in accordance with Regulatory Guide 1.26 is identified as
Violation 280,281/90-41-01. * This inadequate classification resulted
in the pumps not being tested in accordance with Section XI require-
ments.
Physical Security Program Inspections
In the course of monthly activities, the inspectors included a review
of the l 1 censee
I s phys i ca 1 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.
No discrepancies were noted.
Within the areas inspected, one violation was identified.
4.
Maintenance Inspections (62703 & 42700)
During_ the reporting period, the inspectors reviewed maintenance
activities to assure compliance with the appropriate proceduresw
Inspection areas included the following:
a.
B BAST Temperature Detector Replacement
On January 15, the inspectors witnessed the replacement of B BAST
. ___ ----- -- tempe~a-ture-detec-to~.----T.he pu-1".'pose-o-f--the---tempe-rature-dete<;-tol'- --i-s-t-o---- ---
provide local indication, and provide input to a controller that
regulates the B BAST heaters to maintain temperature of the tank
.. *
b.
- -------
6
within *the TSs allowable limits.
WO 104723 and procedure PT-2.23,
Boric Acid Storage Tank Temperature Calibration, dated August 31,
1990, was used to.accomplish this maintenance. * The inspectors
reviewed the WO and calibration procedure at the job site while the
maintenance was in progress. * As a result of this review, the
inspectors noted several examples where maintenance personnel were
not following* the requirements of station administration procedures.
Paragraph 3.3.9 of procedure SUADM-ADM-47,. Operation of the
Instrument Department, requires that permission to perform safety
related work without a procedure be granted by instrument department .
supervision and noted a~ such on the WO.
The technicians performing.*
this maintenance had a procedure to calibrate the temperature
detector but did not have a procedure to install the detector.
Distussion with the lead techhician fndicated that installation of
the detector was within the skill of the craft,.however, the WO was
not annotated by instrument department supervision that the detector
could be installed without a* procedure.
. Paragraph 6.7.2 of
VPAP-0501, Procedure Admi ni strati ve Control Program, requires that
the procedure signoffs be completed as the procedure step is
completed.
When the inspectors visited the job site, the technicians
had completed the installation of the temperature detector and were
in the process of calibrating the detector in accordance with
PT-2.23.
With the exception of the signoff being made for notifying
the Shift Supervisor, none of the remaining signoffs in the initial
conditions and precautions section of PT-22.3 were made.
- These
examples where personnel did not foll ow station administrative
procedures were identified as a weakness in the area of maintenance
and were discussed with the maintenance department superintendent
and instrument supervisor.
Inspection Reports 50-280,281/90-07 and
20 previousli identified examples where personnel were not following
station administrative procedures.
The inspectors are concerned that
personnel do not always follow station administrative procedures when
performing maintenance related activities and consider thit this area
warrants additional management attention.
ATWS Mitigation System (TI 280/2500-20)
The
NRC* ma-ndated the implementation *of common inode failure
protection for the reactor in order to reduce the* risk of an
ATWS event (10 CFR 50.62).
The ATWS Rule requires specific
improvements in the design and operation of nuclear plants to
reduce the probability of failure to shut .down the reactor
following anticipated transients* and to mitigate the* conse-
quences of ari ATWS event.
The licensee's-modifications to
incorporate these improvements were* inspected during.* the last
two inspection periods. *
At the end of the last inspection period, the AMSAC. system was
placed in the normal mode at 35% power.
Approximately two
-- -minute S-- -a f-te-r--plac i n g--th e -sys-tern- i-n - s e Y'-V i ee,--t he-AMS AC--a-rmed--.,.------ --- -- ---
annunciator was activated and the system was placed in bypass.
The inspectors monitored the l icensee
1 s corrective action in
I-
I
I **
7
response to the AMSAC alarm.
An analysis of the problem
revealed that several analog input cards had failed.
Further
i nvesti gati on showed that the cards* were rated for approximately
30 to 50 volts DC usage.
The Surry control room annunciator
system which . interface with these cards is a, 120 volts. DC
system.
This high voltage caused the cards to fail.
The.inspectors met with the licensee on January.24, 1990; to
discuss the .design process for this project.
The licensee had
an architect engineering firm perfprm the design for the ATWS
(AMSAC) system.
The project involved the design for this system
at both North Anna and Surry.
The licensee stated that from the
onset of the Surry project, North Anna system-specific items
kept appearing in * the Surry design.
While some of these i terns
were ide_ntified to the AE-, the 1 icensee' s design process did not
identify the fact that the North Anna annunciator system
utilizes 48 volts DC and that the Surry ATWS system was designed
for this va 1 ue.
The Surry annunciators' 120 vol ts DC caused the
failure of the analog input cards previously .mentioned.
This
lack of attention to det.ail in the licensee's design process is
identified as a weakness in the area of engineering support.
The cards were replaced, the AMSAC system was modified so the
cards would not receive the higher voltage and the system was
retested.
The inspectors observed part of the AMSAC retest.
In
addition, the inspectors observed the periodic test that was
used* to return* the system to service, 1-IPM~AMS-PNL-001, Rev .* 0,
AMSAC. Functional Test, dated December 2, 1990.
The system was
placed in service at 1229 hours0.0142 days <br />0.341 hours <br />0.00203 weeks <br />4.676345e-4 months <br /> on January 23 .. The next day at
0657 two annunciators,
AMSAC trouble and AMSAC armed, were
received and* the system was pl aced in bypass.
The 1 i censee has
initiated a field change to place monitoring equipment inside
the AMSAC panel . in order to determine the cause of the a 1 arms.
The inspectors will continue to monitor this problem area during
subsequent inspection periods~
Withi~ the areas inspected, no violations were identified.
5~
Licensee Event Report Review
(92700)
The inspectors reviewed the LER listed below to ascertain whether NRC
reporting requirements were being met and to ev*aluate initial adequacy of
the corrective actions.
The inspectors' review also included followup on
implementation of corrective action and review of licensee documentation
that all required corrective actions were complete.
(Closed) LER. 280/90-010, Late Type B and C Testing in Accordance with 10
CFR 50 Appendix J Requirements Due to Improper Interpretation of Criteria.
The licensee discovered, as a result of one of their recent QA audits,
that the period of time between some lOCFRSO Appendix J Type Band C tests
exceeded the 2 year maximum time limit. The audit showed that the Type C
,8
testing interval for Unit 1 was exceededfor approximately two months in
1990; and, prioi to 1986, both Units 1 and 2 had.experienced extended
intervals greater than that allowed by Appendix J for Type B testing.
.
-
.
.
The licensee indicated that the primary caus~ of these ~vents was the .
improper interpretation of 10 CFR 50 Appendix J, Section III. D. criteria
and the improper application of the+/-
25% grace period allowed by TS
4.0~2.
Previous to this event, the licensee's Appendix J testing was
based on an 18 month refueling cycle and was tracked on an aggregate
rather than a component basis.
Recording Type C testing dates on* an
aggregate basis. did not pose a problem si nee the time periods between the
completion of the first and last components were relatively small when
compared to the six month margin between the 18 month refueling outage and
the two year maximum test mandate of Appendix J. However, due to a recent.
extended dual unit outage a .significant amount of time passed between the.*
testing of the first componerits and the last c6mponents. This exceeded the
time that was previously considered to be conservative for these 10 CFR 50
Appendix J tests.
At .the time this issue was identified, the licensee verified thal the most
recent test results were acceptable.
A one time extension request,
received from the NRC on June 22, 1990, had been granted for Unit 1 based
on technical assessments of the issue.
A subsequent TS amendment, also
provided a one~time extension of the 10 CFR 50 Appendix J.Section III.D
tests for .Unit 2 in accordance with NRC exemption dated September 25,
1990.
,
The licensee has changed their Type C testing programs to track testing of
i ndivi dua 1 components and the survei 11 ance tracking program has been
updated to reflect that the TS 4.0.2 extension period of+/- 25% is not to
be applied to this testing.
TS 4.4~0 states that Type A, B, and C tests will be in accordance with
- 10 CFR 50 Appendix J Section III. D.
Contrary to this, the time periods
between tests for some of th~ Ap~endix J tests exceeded the maximum two
year intervals.
This was reported by the licensee and is identified as
Non-Cited Violation 280,281/91-04-02. This licensee identified violation
. is not being cited because the criteria specified in Section V.G.1. of the
NRC Enforcement Policy were satisfied.
6.
Action on Previous Inspection findings
(92701, 92702)
a.
(Closed) IFI 280,281/90-14-01, Followup on Licensee Leak Reduction
Program.
Technical Specification section 6.4.K.1 requires the
establishment of PM and inspection requirements as a part of a leak
reduction program for systems outside containment that would or coula
contain highly radioactive fluids during a serious transient. There
are numerous procedures and periodic tests . regarding this 1 eak
reduction program, but there was no single document that insured that
this program would be implemented~ i.e., a single document did not
tabulate or list the required procedures, PM's, and visual
b.
9
inspections to insure that the TS wa~ fully met and that tracking and
performance could be implemented effectively.
The licensee changed
. ENG-40, Quantification of External System Leakage, Rev. 1, dated
November 13, 1990, to include all of necessary information *to satisfy
this concernL * The inspectors*reviewed ENG-40 and.consider that this
issue h~s been satisfactorily resolved.
(Closed) IFI 280,281/90-14'."02, Followup on Lic_ensee Action for
Replacement of Westinghouse Type BFD Relays.
This IFI was opened
when three BFD relays failed in the spring of 1990.
These .relays are
used in the reactor protection system and engineered safeguards
system and because of these and previous relay failures, the licensee
was considering the rep 1 acement of a 11 these type of relays during
the riext outage.
Replacement would be made with a newer Westinghouse
type NBFD65NR
relay that is more. reliable.
In their investigation
of this problem, the licensee det.ermined that the degradation of
r~lay insulation and subseq~ent failure is related to heat and large
voltage spikes across the coil terminals that are induced when the
current is interrupted. The licensee stated that mass replacement of
the old BFD type relays with the newer type is no longer being
considered for the following reasons: 1) Failure rates are fairly low
(12 failures, 4 in Unit 1 and 8 in Unit 2, in 1990 out of
approximately 400 telays), and failures usually o~cur after an outage
or plant shutdown when relay -state (current is being interrupted) is
changed; 2) Failures do not normally occur during plant steady state
power operations when relays are in a non-changi~g st~te; 3) The
relays are used in redundant trains of protection and, when they
fail, they fail in a safe direction, i.e., that is in the direction
that would not keep the relay from perfonning its safety function; 4)
There is insufficient space in the present cabinets and area to
accommodate the newer type relays and extra relays that would have to
be added for circuit modifications and operability for a mass
replacement design change; 5) The ambient heat of the relay rooms and
the relay energization heat has been reduced.
Ventilation louvers
have been added to relay panels and the operating voltages of the
relays has been reduced to the lower portion of their operating band
to decrease the heat generated by the energized relays.
Air
conditioning, ventilation, and upgraded modifications in 1990 have
reduced ambient and hot-spot room temperatures from 16 to 20 degrees*
(room ambient temperature reductions from the low 90 degree range to
the mid 70 degree range and hot-spots temperature reductions from
approximately 120 degrees to 104 degrees). Additional modifications
were made to Unit 1 systems during the last refueling outage that
enhanced and stabilized temperature conditions and further
modifications are scheduled for Unit 2 during the upcoming .refueling
~utage in April 1991.
Since the heat. reduction efforts have been fairly effective and
further upgrades are planned, the voltage spikes are presently
consider~d to be the significant contributor for relay failures.
The
licensee is doing further investigations in this area. Site Nuclear
c.
10
Engineering Services has proposed to Corporate Engineering, via NES
- 2414, that .a volt~ge surge suppressor, such as a diode or varistor,
be placed in parallel across the coil leads.
This would safely
dissipat~ * the voltage spikes when the current is interrupted and
could minimize or prevent coil failure.
The. licensee is also
tracking the coil resistances of some of the relays to determine if
any trends are evident that could be used to predict coil failure.
(Closed) Violation 280,281/90-14-04, Failure to Follow Procedure for
Testing of Systems and Components as Required by TS 6.4.D.
This
violation was issued for a failure to follow procedural requirements.
A voltage check for periodic tests on one of the main station
batteries was not performed and ah electrical reading for an*
emergency service water pump was performed at a location other than
that specified in the procedure.
The licensee responded to this
violation in* a letter dated May 29, 1990.
In that letter, the
licensee stated that the following corrective actions would be.
implemented: personnel were counseled as to the importance :of
properly recording* and reporting surveillance data; a permanent
voltmeter was installed on all of the service water pumps; a
memorandum was circulated*that emphasized attention to detail, prompt ..
reporting of off-normal conditions, and careful documentation of test
results; and. required a thorough review of test data by
supervisory/reviewing personnel.
The inspectors rev*iewed the
corrective actions and consider that the licensee's actions were
satisfactorily implemented.
7.
Exit Interview
The inspection scope and results were summarized on January 30, 1991, with
those individuals identified by an asterisk .in paragraph 1.
The following
summary of inspection activity was discussed*by the inspectors during this
exit ..
Item Number
VIO 280,281/90-41-01
NCV 280,281/90-41-02
Description and Reference
Failure to correctly classify SW pumps
1-VS-P-lA,B,C and chilled water pumps
1-VS-P-2A,B,C
in
accordance with
Regulatory Guide 1.26 (paragraph 3.d).
Failure to perform containment Type B
and C leak rate testing within Appendix
- J two year test interval (paragraph
. 5.a)~
. * *
The licensee acknowledged the inspection conclusions with no dissenting
comments .. _The 1 i censee did not identify as proprietary any of the
materials provided to or reviewed by the inspectors during this
inspection.
'11 *
.* 8.
Index .of Acronyms amd Initial isms
AE
AFFF
. ATWS .
. BAST
CFR
EOG
EsF*
ESWP
!FI
. IST
LCD
LER
LTSA
NRC
TI
TS
- VIO
. VPAP
-*
ARCHITECT ENGINEER
AQUEOUS FIRE FIGHTING FOAM
AUXILIARY f~EDWATER
.
ANTICIPATED TRANSiENT WITHOUT SC~A~ *
ATWS MITIGATION SYSTEM ACTUATION CIRCUIT
AMERICAN SOCIETY FOR TESTING MATERIALS
BORIC ACID STORAGE TANK
CODE OF FEDERAL REGULATIONS
DIRECT CURRENT
~ . * EMERGENCY DIESEL GENERATOR
ENGINEERED SAFETY FEATURE
EMERGENCY SERVICE WATER PUMP
- -
-*
INDEPENDENT SPENT FUEL STORAGE INSTALLATION
INSPECTOR FOLLOWUP ITEM
. INSERVICE TESTING
LIMITING CONDITIONS OF OPERATION
LICENSEE EVENT REPORT
LONG TERM STABILITY ADDITIVES
NON-CITED VIOLATION.. .
.
NUCLEAR REGULATORY COMMISSION
PREVENTIVE MAINTENANCE
QUALITY ASSURANCE
RADIATION WORK PERMIT
SAFETY EVALUATION
.
TEMPORARY INSTRUCTION
TECHNICAL SPECIFICATIONS
VIOLATION
VIRGINIA POWER ADMINISTRATIVE PROCEDURES
WORK ORDER