ML18151A061
| ML18151A061 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 09/15/1987 |
| From: | Cantrell F, Holland W, Larry Nicholson NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18151A059 | List: |
| References | |
| 50-280-87-21, 50-281-87-21, GL-83-28, IEB-84-02, IEB-84-2, NUDOCS 8709240552 | |
| Download: ML18151A061 (19) | |
See also: IR 05000280/1987021
Text
Report Nos.:
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTA STREET, N.W.
ATLANTA, GEORGIA 30323
50-280/87-21 and 50-281/87-21
Licensee:
Virginia Electric and Power Company
Richmond, VA
23261
Docket Nos.:
50-280 and 50-281
Facility Name:
Surry 1 and 2
License Nos.: DPR-32 and DPR-37
Inspection Conducted:
Approved
SUMMARY
Date 'Signed
°//11~s/f 7
Date Signed
S7/; .5/~" 7
Date Signed
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, licensee event report
review, 10 CFR Part 21 review, and closeout of temporary instruction T2500/19.
Results: Two violations were identified in this inspection report which are
being considered for escalated enforcement action and will be forwarded under
separate cover.
In addition one violation is listed in this report from
findings identified in inspection report 280; 281/87-11 .
Or.::5r-, 970916
870924
~ ~ 05000280
ADOCK
'
Gl
REPORT DETAILS
1.
Persons Contacted
Licensee Employees
2.
- 0. L. Benson, Station Manager
H. L. Miller, Assistant Station Manager
- E. S. Grecheck, Assistant Station Manager
J. A. Bailey, Superintendent of Operations
D. J. Burke, Superintendent-of Maintenance
S. P. Sarver, Superintendent of Health Physics
- R. H. Blount, Superintendent of Technical Services
R. L. Johnson, Operations Supervisor
J. A. Price, Site Quality Assurance Manager
J. B. Logan, Supervisor, Safety and Licensing
G. 0. Miller, Licensing Coordinator
- F. P. Mone, Supervisor, Quality Assurance
- M. A. Griffin, Administrative Assistant
- Attended exit meeting.
Other licensee employees contacted included control room operators, shift
technical advisors, shift supervisors and other plant personnel.
Exit Interview
The inspection scope and findings were summarized on August 31, 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 (paragraph 4) was identified for failure to conduct
evaluations for unreviewed safety question determination as required by
technical specifications and 10 CFR 50.59. This item is being considered
for escalated enforcement.
One Violation (paragraph 4) was identified for inadequate procedures,
failure to follow procedures in testing the safety injection system, and
failure to perform a technical specification-required portion of the
diesel generator surviellance.
This item is being considered for
escalated enforcement.
One Unresolved Item (paragraph 9) was identified for review of the
licensee's revised response to IE Bulletin No. 84-02.
In addition, one violation (paragraph 13) was identified from findings of
an NRC headquarters vendor inspection as documented in inspection report
280; 281/87-11.
2
The licensee acknowledged the inspection findings with no desenting
comments in the findings relating to this report.
However, the licensee
did take exception to the findings identified in this report relating to
inspection report 280; 281/87-11.
The licensee did not identify as
proprietary any of the materials provided to or reviewed by the inspectors
during this inspection.
3.
Plant Status
Unit 1
4.
Unit 1 began the reporting period at power.
The unit operated at power
until August 7 when at 1:20 p.m. the unit was manually tripped from 100
percent power due to failure _of the cooling capability of the lB main
transformer.
Repairs were made to the transformer and the unit returned
to power operation the morning of August 8, 1987.
The unit operated at
power for the remainder of the inspection period.
Unit 2
Unit 2 began the reporting period at power.
The unit operated at power
for the duration of the inspection period .
Licensee Action on Previous Enforcement Matters
(92702)
(Closed) Unresolved Item (URI) 280; 281/87-17-01, Review of 10 CFR 50.59
safety
evaluation
for
emergency
busses
cross-connect
breaker
configuration.
The subject issue was discussed in inspection report 280; 281/87-17.
In
that* report, the inspector determined that the
emergency
busses
cross-connect breakers (5Hl) for both units were racked out; however, the
breakers remained in the cubicles. This condition was in conflict with
the condition described in the FSAR, paragraph 8.4.1.
The inspector
requested that the licensee provide the safety evaluation which was
required by 10 CFR 50.59 when the decision was made to leave the subject
breakers in their cubicles.
This request was being evaluated when the
last inspection period ended.
During this inspection period, several additional items relating to the
above concern have come to the attention of the inspectors.
One item
involved the licensee's decision to furmanite a leaking valve (2-WT-177)
which was a manual isolation valve for the chemical addition system to the
B steam generator main feed line in containment.
This repair, on July 4,
1987, resulted in the valve being left in the open position and
The
inspector requested that the licensee provide the
10 CFR 50.59 safety evaluation for this change in system configuration
and, at the time of the request, the evaluation had not been accomplished.
The inspector then reviewed the temporary modification (TM)
log and
determined that this condition had been. logged the evening of July 4,
1987.
However, additional review of the log entry determined that no
3
technical/safety .evaluation review was conducted for the modification.
Also, the inspectors were reviewing station deviation reports and noted
that report number Sl-87-512 dated June 19, 1987, identified a condition
which indicated that testing of the turbine inlet valves as identified in
the FSAR, Section 14.2.13 was not being accomplished.
The inspector also
requested the safety evaluation for this condition and was informed that
the deviation report was still under review.
These concerns were brought to the attention of station management on
July 8 and again on July 10, 1987.
The inspectors stressed that the main
issue was ev2.luation of plant configuration changes to assure that an
unreviewed safety question had not resulted when changes were made.
It
appears that deleting the turbine inlet valve test does involve an
unreviewed safety question and prior Commission approval should have been
sought prior to deleting the test.
After these meetings, another concern was i dent i fi ed to the 1 i cen see
regarding the use of the station fire protection system for purposes not
recognized in the FSAR.
A ring header is periodically installed on the
top of both containments during the hot summer months and fire water
pumped through the header and allowed to run down the outside to assist in
containment cooling. The inspectors questioned if an evaluation had been
performed to determine the impact this usage would have on the capability
to respond to a fire, as well as the consequence of the water running into
the p 1 ant.
A tour of the p 1 ant rev ea 1 ed water in safety-re 1 ated pump
rooms as well as water running down the wall beside the electrical
penetrations in the cable vault.
The inspector a 1 so conducted *a review of the admi n i strati ve procedure
which provides for requirements to conduct evaluations for unreviewed
safety questions.
Those procedures were:
SUADM-ENG-01
SUADM-ENG-03
SUADM-0-11
Engineering Work Request
Design Changes
Functional Bypass and Temporary Modification Control
The inspector concluded from the review that procedures do require
evaluation for unreviewed safety questions;
however,
the procedure
guidance did not always clarify those areas which could require reviews as
required by 10 CFR 50.59 or Technical Specifications.
Additional discussions with station management were held on August 21,
1987,
on
these issues.
Discussion of the issue involving station
deviation report Sl-87-512 resulted in the conclusion that an unreviewed
safety question determination was not conducted when the decision was made
4
to delete testing of the turbine inlet valves.
The licensee provided a
copy of a new station deviation to the inspector on Auqust 21, 1987, which
indicated that an adequate safety evaluation per 10 CFR 50.59 was not
performed when the periodic test for testing of the turbine inlet valves
was discontinued.
Based on the previous findings, the inspector determined that safety
committee (SNSOC) evaluations for unreviewed safety questions had not been
performed and documented in the above cases as required by Technical
Specification 6.1.7.f and g; nor .had the licensee followed the correct
process for conducting a safety evaluation in the above cases as required
by 10 CFR 50.59.
Deleting the turbine inlet valve test appears to involve
an unreviewed safety question.
This item is identified as a violation
(280; 281/87-21-0J) for both units.
(Closed) Unresolved Item 280; 281/87-17-02, Inadequate evaluation of
deficiencies noted during.surveillance testing.
Inspection report 280; 281/87-17 identified numerous concerns regarding
the documentation, evaluation, and corrective actions as a result of the
safety injection train undervoltage functional tests performed during the
1986 refueling outages for both units.
The specific test procedures
reviewed qy the inspector were PT-18.2 A & B for both units. Discussions
with the licensee following their search for additional documentation
revealed the following:
a.
Technical Specification 4.6.A.l.b requires testing demonstrate that
the loss of voltage and degraded voltage protection is defeated
whenever the emergency diesel is the sole source of power to an
emergency bus and that this protection is automatically reinstated
when the diesel output breaker is opened.
This requirement is not
included in the above procedures and consequently has not been
performed.
This is in violation of the Technical Specifications and
the licensee stated the test would be performed during the next
outages.
b.
1-PT-18.2A, "Safety Injection Train A - H Bus Undervoltage Functional
Test,
11 completed 7-7-86.
The completed test results were not reviewed by the surveillance
and test engineering group as required by paragraph 5 .1. 5 of
station administrative procedure SUADM-0-23.
Acceptance criteria was deleted with no reason for deviation
stated as required by paragraph 5.4.3 of station administrative
procedure SUADM-0-21 .
Verification that the emergency diesel generator was secured and
restored was not performed as required by step 5.24.4 of the
above test procedure.
5
The use of a special test to satisfy surveillance testing is
inadequate in that this procedure does not receive the review
and approval required by a normal periodic test.
Also, the
special test system is inadequate in that the performance of
these tests have routinely not been reported to the NRC in the
Monthly Operating Report as required by 10 CFR 50.59 and local
administrative procedure SUADM-0-18.
This is another example of
the violation of 10 CFR 50.59 cited above in this p-aragraph
(280; 281/87-21-01).
c.
1-PT-18 . .?B,
11Safety Injection Train B -
J Bus
Functional Test," completed 7-6-86.
Test results were
11 unsati sfactory
11 and no corrective action was
performed; however, the unsatisfactory results were determined
to be from a procedure problem.
The licensee could not locate
any procedure change request forms which are required by
admi ni str.at i ve procedure SUADM-0-21
and
speculated that a
procedure change was at one* time initiated and subsequently
lost.
The required post-test position of high-head safety injection
pump 1-CH-P-lA was changed in Attachment I of the above test
procedure without a
procedure deviation
as
required by
administrative procedure SUADM-0-21.
d.
2-PT-18.2A,
11Safety Injection Train A
-
H Bus
Functional Test
11 , completed 11-23-86.
Problems identified on the test critique sheet and changed on a
procedure deviation were evaluated as
11 procedure problems.
11
No
procedure change request form was initiated nor were the
problems corrected in the next revision as required by
administrative procedure SUADM-ADM-21.
The licensee stated that
the changes were submitted, but lost.
e.
2-PT-18.2B,
11Safety Injection Train B - J Bus Undervoltage Functional
Test,
11 completed 11-21-86.
The test critique sheet states that high-head safety injection
pump 2-CH-P-lA requires retesting.
No documentation can be
found that either retests this pump or evaluates the deficiency.
Technical Specification 6.4 requires that detailed written procedures with
appropriate check-off lists and instructions shall be provided and shall
be followed for the testing of components and systems involving nuclear
safety of the station.
The above findings represent both an inadequate
procedure and failure to follow procedures with regard to testing,
6
documenting,
and
evaluating
results
of
safety
injection
system
surveillance tests.
In add.ition, a portion of the emergency diesel
generator survei 11 ance test apparently has never been performed.
This
item is identified as a violation of Technical Specifications
(280; 281/87-21-02) for both units.
Within the areas inspected, two violations were identified.
5.
Unresolved Items
Wnresolved items are matters about which more information is required to
determine whether they are acceptable or may involve violations or
deviations.
One new unresolved item is identified in paragraph 9.
6.
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
operable; review of control room operator logs, operating orders, plant
deviation reports, tag out 1 ogs, jumi:ier 1 ogs, 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 equipment or component(s), and operability
of instrumentation and support items 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 lineup(s); and verification that notices to workers are posted
as required by 10 CFR 19.
7
Certain tours were conducted on backshifts or weekends.
Backshift or
weekend tours were conducted on July 9, 11, 29, August 5, 7, 8, and 25.
Inspections included areas in the Unit 1 and 2 cable vaults, vital battery
rooms,
steam safeguards areas, emergency switchgear rooms,
diesel
generator rooms, control room, auxiliary building, cable penetration
areas,
independent spent fuel
storage facility,
low level
intake
structure, and safeguards valve pit and pump pit areas. Reactor coolant
system 1 eak 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.
The inspectors routinely
independently. calculated RCS
leak rates using the NRC
Independent
Measurements Leak Rate Program (RCSLK9).
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.
Se 1 ected
radiation protection instruments were periodically checked, and equipment
operability and calibration frequency were verified.
The Plant Risk Status Information Management System (PRISIM) was installed
in the resident inspectors' office during this inspection period.
This
personal computer program provides PRA results and other risk-related
information to the inspector for use on deciding inspection priorities.
This version of PRISIM is based on the PRA of Surry 1 performed by EI
International as part of the Accident Sequence Evaluation Program.
National Engineering Laboratory modified the PRA in response to comments
made by the Virginia Electric and Power Company.
The PRA results obtained
from PRIS IM are based on core damage frequency.
PRIS IM does not
incorporate the results of assessments of plant damage, containment
responses, or public health consequences.
A meeting was. held on July 14,
1987, at Region II to introduce the Surry 1 PRISIM and to discuss the plan
for testing and evaluating the program.
On July 20, 1987, the inspector witnessed the 1 oadi ng of four spent fue 1
assemblies into a dry cask for storage at the Independent Spent Fuel
Storage Installation (ISFSI) located onsite at Surry. This fuel movement
from the spent fuel pool was being performed by licensed reactor operators
using a written procedure.
This was the fourth Castor V/21 Cask to be
loaded with twenty-one (21) spent fuel assemblies.
The three previously
loaded casks are presently in storage at the ISFSI storage pad.
On July 27, a malfunction of the Kaman radiation monitor RM-GW-103-1
caused the automatic isolation of the containment vacuum system to
atmosphere.
The subject monitor surveys the gaseous effluent release path
of the process vent stack in conjunction with the Victoreen monitors. A
radiation level above setpoint causes the containment vacuum
pump
discharge valves FCV-GW-160 & 260 to shut. The licensee notified the NRC
pursuant to 10 CFR 72 of an engineered safety system actuation, then
subsequently determined that this was not an ESF actuation since the
subject valves are not required for containment isolation.
The shutting
of these valves did, however, remove the one containment vacuum flow path
required by Technical Specification 3.15.8, and required the licensee to
reestablish a flow path or be in hot shutdown in at least six hours.
The
8
inspector expressed concern that during the troubleshooting of this
radiation monitor, a jumper was installed essentially around the monitor
to permit the reestablishment of the containment vaccum flow path.
This
jumper was not controlled by an entry to the station jumper log or
specified in an applicable approved procedure, as required by administra-
tive procedure SUADM-0-11.
The omission of the required administrative
controls prevented this temporary plant modification- from being properly
evaluated as required by 10 CFR 50.59.
This is another example of
violation (280, 281 87-27-01) identified in paragraph 4.
On several occasions during this inspection period, a high chlorine alarm
was received in the main control room from the
11A
11 train chlorine monitor.
This monitor is one of two installed in the main control room to sense
chlorine that could leak from the storage tanks at the site sewage
treatment plant.
Investigation revealed the alarm to be caused by an
erroneous spike on that detector.
This alarm did, however, actuate a
train to isolate the control room ventilation as required.
The licensee
notified the NRC of these ESF actuations as required by 10 CFR 50.72.
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.
Engineered Safety Feature System Walkdown
(71710)
The inspector performed a walkdown of the accessible areas of the safety-
related portions of the Emergency Diesel Generator system and the
Auxiliary Feedwater System for both units to verify their operability.
This verification
included the
following:
confirmation that the
licensee's system lineup procedure matches plant drawings and actual plant
configuration;
hangers
and
supports are operable;
housekeeping is
adequate; valves and/or breakers in the system are installed correctly and
appear to be operable; fire protection/prevention is adequate; major
system components are properly labeled and appear to be operable;
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, one additional example of a violation cited in
paragraph 4 was identified.
7.
Maintenance Inspections (62703)
During
the
reporting period,
the inspectors reviewed
activities to
assure compliance with
the appropriate
Inspection areas included the following:
maintenance
procedures.
9
The inspectors followed the repair of a defective weld in the Unit 2
letdown line inside containment.
On July 22, the unidentified leakage
from the reactor coolant system increased from approximately 0.2 gpm to
0.6 gpm.
The operators also noted a slight increase in containment
activity. Containment entries were m~de and the leakage was determined to
be primarily from a defective socket weld on a two-inch pipe tee just
downstream of the letdown isolation valve HCV-2200C.
This condition was
documented on station deviation report S2-87-342. * On July 28, the
licensee had determined that a permanent repair as required by the ASME
Code Section XI was not possible due to the inability to successfully
isolate the leak; therefore, an Engineering Work Request (EWR 87-288) was
issued approving a temporary repair until the 1988 refueling outage. This
temporary repair consisted of encasing the defective weld in a can and
injecting a Furmanite sealant. The EWR considered this a temporary repair
as a gasket joint seal and included a safety analysis.
The inspector
reviewed all the documentation for this effort and expressed the following
concerns:
a.
An attempt to permanently repair the defective weld should be made no
later than the upcoming 10 day snubber outage scheduled for late this
year in lieu of the 1988 refueling outage.
The licensee agreed and
committed to a repair no later than the snubber outage; and
b.
A formal evaluation control with regard to containment integrity was
not performed for the time between the discovery of the defect and
the accomplishment of the temporary corrective action.
The subject
defect is located inside the containment between the inside
containment isolation valves and the letdown line containment
penetration. This results in the outside containment isolation valve
being the only barrier for this penetration.
These concerns and the resulting enforcement action are addressed in a
special inspection report (280;* 281/87-26).
8.
Surveillance Inspections
(61726, 61700)
During the reporting period, the inspectors reviewed various surveillance
activities to assure compliance with the appropriate procedures as
fo 11 ows:
Test prerequisites were met.
Tests were performed in accordance with approved procedures.
Test procedures appeared to perform their intended function.
Adequate coordination existed among personnel involved in the test .
Test data was properly collected and recorded.
10
Inspection areas included the following:
On July 29, the inspector witnessed portions of periodic test 2-PT-18.7,
11 Charging Pump Operability and Performance Test.
11
This test was performed
following the replacement of oil on charging pump 2-CH-P-lC.
The licensee
noticed foaming of the oil in this pump and suspected it to be
contaminated with water.
Analysis of the oil removed, however, did not
indicate a presence of water.
The samples have been sent to an
independant lab for further evaluation.
No discrepancies were noted.
On August 7, 1987, the inspector witnessed portions of periodic test
1-PT-14.2,
11Main Steam Trip Valves and Main Steam.Non-Return Valves.
11
No
discrepancies were noted; however, the results observed did not meet the
requirements of the acceptance criteria. The test requires that the MSTVs
close from actuation of the switch to indication of closure in 5 seconds
or less.
Results being obtained were in the 6 to 7 second range.
The
inspector noted that this test is not a realistic test of closure time of
the valves in an accident.situation and also was informed that a Technical
Specification change had been submitted to resolve this condition.
A
deviated PT-14.2 was performed satisfactorily later that evening.
On August 18, 1987, the inspector witnessed portions of the periodic test
1-PT-15.lC,
11Steam Generator Auxiliary Feedwater Pumps.
11
This test
performed the monthly surveillance on the turbine driven auxiliary
feedwater pump l-FW-P-2.
The pump was started using the
11A
11 train steam
admission valve and appeared to start and perform within the acceptance
criteria specified for alert in the above procedure.
The recorded
vibration for the turbine to pump bearing was measured to be 0.19 in/sec,
thus placing the pump in the alert range. A vibration measurement greater
than 0.2 in/sec would constitute an inoperable pump.
The licensee is
continuing to evaluate this condition.
No discrepancies were noted.
Within the areas inspected, no violations or deviations were identified.
9.
Followup on Inspector Identified Items and IE Bulletins
(92701)
(Closed) Inspector Followup Item (IFI) 280, 281/84-31-01:
Implementation
of Licensed Operator Requalification Program (LORP) and its submittal to
the NRC.
The LORP required that persons failing to score satisfactorily
on requalification examinations be removed from shift duty, participate in
an
accelerated requalification program,
and pass a requalification
examination.
The accelerated training program required nothing other than
self-study.
The. inspector followup item involved upgrading accelerated
requalification to include more than self-study, and the submittal of a
revised requalification program for
NRC
review and approval.
The
inspectors
reviewed
the
present
status
of
1 i censed
operator
requalification and determined that in 1985 the licensee implemented
accelerated requalification training with individualized requirements
beyond self-study. The licensee stited that the LORP is being updated to
incorporate revised 10 CFR 55 requirements, and should be submitted to the
NRC within two months.
This revised LORP contains expanded requirements
11
for accelerated requalification.
Based on this information, the item is
closed.
(Closed) I FI 280, 281/85-31-01:
Admi n i strati ve Contro 1 s for Changes
Affecting the Work Planning and Tracking System (WPTS).
A previous NRC
inspection identified that the licensee was using a computerized Work
Planning
and
Tracking
System without a formal
method to ensure
incorporation of changes to applicable plant procedures, policies or
hardware.
To ensure work orders adhere to current requirements, the
1 i cen see returned to verifying information entered on the work orders
against the Technical* Specifications and hard copies of the procedures
addressing Environmental Qualification, Safety Related Equipment (Q-List),
and Inservice Inspection Requir~ments.
B~cause these inputs should have
adequate administrative controls, this item is closed.
The licensee is in
the process of updating the computerized WPTS to assure accuracy, improve
software contro 1 s, and enhance the tracking of work orders.
Ful 1
implementation of the computerized WPTS is expected to be completed within
two months.
(Closed) I FI 280, 281/85-31-04:
Preventive Maintenance Overdue Criterion
and Management Review of Overdue Items.
Licensee procedures had not
defined an overdue criterion for preventive maintenance or re qui red
management review of delinquent items.
The inspector verified that
Procedure SUADM-M-30,
Planned Maintenance System Manual,
which was
approved August 18, 1987, contains overdue criteria and management review
requirements. The item is closed.
(Closed) IFI 280, 281/85-31-05:
Review of Preventive Maintenance Program
for Adequate Procedures as Recommended by Vendor Manuals. The inspector
followup item involved assuring that preventive maintenance is required at
-the frequencies recommended by the vendors, and that the program provides
verification of the design basis dewpoint of the containment instrument
air system.
The licensee stated that all safety-related preventive
maintenance procedures will be updated and approved by September 1987 in
accordance with their commitment of October 1985 in response to Generic
Letter 83-28.
The inspectors reviewed a draft revision of procedure
IA-C-M/M entitled Nash Model 8045 Containment Instrument Air Compressor
Preventive Maintenance (Non-Safety Related).
The draft revision included
a post maintenance check that the air dryer is operating properly, and was
scheduled for final review on August 21, 1987.
Based on this information,
the item is closed.
(Open) IE Bulletin No. 84-02, Failure of General Electric Type HFA Relays
Used in Class IE Safety Systems.
During a review of the subject bulletin
for closeout, the inspector questioned the 1 icensee regarding current
status.
A review of the status was conducted by the licensee and they
determined that the original response to this bulletin, dated July 31,
1984, was incorrect in that eight relays per unit were not identified as
12
candidates for replacement.
This omission prevented these relays from
being inspected and subsequently replaced as required by the bulletin.
The licensee is in the process of preparing a revised response to the
bulletin.
This issue is identified as an
unresolved item (280;
281/87-21-03) pending review of the licensee's revised response.
Within the areas inspected, no violations or deviations were identified.
10.
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 review also included a followup on
implementation of corrective actions anp review of licensee documentation
that all required corrective action(s) were complete.
LERs that identify violation(s) of regulation(s) and that meet the
criteria of 10 CFR, Part 2, Appendix C,Section V are identified as
Licensee Identified Violations (LIV) in the following closeout paragraphs.
LIVs are considered first-time occurrence violations which meet the NRC
Enforcement Policy criteria for exemption from issuance of a Notice of
Violation. These items are identified to allow for proper evaluation of
corrective actions in the event that similar events occur in the future.
(Closed)
LER 280/84-10, Appendix R Review.
The
issue involved a
reanalysis of the station's compliance with 10 CFR 50, Appendix R design
requirements.
The reanalysis identified five areas which did not meet
Appendix R requirements.
The 1 i censee corrected the discrepancies.
Region II conducted an inspection in May 1987 of the Surry Power Station
which evaluated the licensee's actions regarding the implementation of the
requirement~ of Appendix R.
Based on that inspection, this LER is closed.
(Closed)
LER 280/87-01, Pressurizer PORVs Declared Inoperable Due to
Excessive Stroke Time.
The issue involved failure of the subject valves
to open within the required time frame when the emergency backup air
supply was used as the only pressure source.
The test deficiency was
identified by the licensee during review of IEN 86-50.
The licensee
determined that the air supply lines from the emergency air supplies were
undersized.
Corrective action included redesign and installation of
properly sized air lines and related components.
The corrective action
and retest was verified as complete by the inspector.
This item is
identified as LIV 280/87-21-04 for failure to provide adequate design and
testing of a safety-related system.
This LER is closed.
(Closed)
LER 280/87-03, Control/Relay Room Chiller Inoperable Due to
Inadequate Service Water Flow.
The issue involved loss of operation of
the subject chiller on three different occasions contrary to Technical
Specification 3.14.B.
The licensee has identified the root cause to be
inadequate flow of service water.
The inadequate* flow was due to pump
13
suction
strainer clogging,
pump
failure,
excessive
throttling of service water, respectively, for the occasions.
Licensee
corrective action is to replace the three control room chillers with new
units.
In addition, an engineering evaluation has been conducted of the
control/relay room ventilation system,
including the service water
subsystem.
The
inspector reviewed the engineering conclusions and
considers that the licensee is taking appropriate long-term actions
(service water system upgrade) to eliminate this condition. This item is
closed.
(Closed) LER 280/87-04, Potential for Bypass of Safety-Related Filters Due
to Inadequate
Fan
Shaft Seal.
The
issue involved the licensee 1 s
determination that the control room ventilation fans have the potential to
admit more than the Technical Specification allowed limit ~f unfiltered
air into the control room due to inadequate fan shaft seals.
Corrective
action included installation of temporary shaft seals until permanent
bellows seals were procurred.
The inspector verified installation of the
new seals~
This item is closed.
(Closed) LER 280/87-05, Control Room Chillers Tripped Due to Clogging of
the Service Water Y-Type Strainers.
The issue involved clogging of the
Y-type strainers for the Band C chillers when the rotating strainer
upstream of the Y strainers was returned to service following maintenance.
Marine growth inside the rotating strainer became dislodged when the
strainer was
returned to service clogging
the Y-type
strainers.
Corrective action included cleaning of the Y-type strainers and returning
the chillers to service. Also, the Y-type strainers are now cleaned on a
bi-weekly basis and the maintenance procedure for the rotating strainer
was changed to require cleaning of the Y-type strainers prior to returning
the rotating strainer baci to service.
This item is closed.
(Closed)
LER 280/87-06, Control/Relay Room Chiller Inoperable Due to
Chiller Service Water Pump Trip.
The issue involved operator observation
that the service water pump for B, the control/relay room chiller, had
tripped rendering the chiller inoperable. Corrective action included the
operator reseting the thermal overload on the pump motor and restarting
the pump.
The exact cause of the pump trip could not be determined;
however, it was suspected that the thermal overload device at the motor
control center activated, tripping the pump.
The licensee considers that
this event was a random event due to no abnormal condition being
discovered during checkout of the motor control center.
This item is
closed.
(Closed)
LER 280/87-07, Control/Relay Room Chiller Inoperable Due to
Insufficient Service Water Flow.
The issue involved a trip of two of the
control room chillers.
The trips were due to insufficient service water
flow.
Corrective action included manual adjustment of the service water
discharge flow in addition to repairing a seal leak on one of the service
water
pumps.
An
engineering
evaluation
of
the
service
water
14
system has been conducted.
The inspector reviewed the engineering
conclusions and considers that the licensee is taking appropriate
long-term actions (servic.e water system upgrade) to eliminate this
condition.
This item is closed.
(Closed) LER 280/87-08, Control/Relay Room Chiller Tripped Due to Valve
Positioning Error.
The issue involved improper operation (repositioning)
of the condenser service water discharge valve causing the chiller to trip
on high condenser discharge pressure. The cause of the valve positioning
error was using the same operator (valve handle) to position several
valves.
Corrective action included installing of permanent valve handles
to each valve requiring operation. The inspector verified that corrective
action was accomplished. This item is identified as LIV 280/87-21-05 for
failure to provide adequate control of a safety-related system.
This LER
is closed.
11.
10 CFR Part 21 Inspections (36100)
(Closed) 280/P2186-0l, Power Supply Failures of GE SLV Relays. The issue
involved failure of the subject power supply due to potential material
problem of the core in the power supply transformer.
Corrective action
taken by the licensee included replacement of the subject relays with
newly designed versions of the relay.
The new relays were installed by
design change packages 86-01 and 86-02.
This action was completed in
1986.
This .item is closed.
(Closed) 280/P2186-04, Contromatics Actuators on Dampers Furnished by
Pacific Air Products (PAPCO) may have Jackscrew/Handwheel Installations
which are Improperly used for Routine Cycling.
The issue involved regular
cycling of the dampers by some plants using the jackscrew/handwheel which
was designed only for emergency manual operation of the component.
The
issue was identified to various utilities by the vendor along with a
listing of replacement parts available to allow for continued manual
cycling of the dampers.
The licensee reviewed this condition and
determined that dampers purchased from PAPCO that are air operated do not
have handwheels for manual operation. The licensee is also in the process
of determining whether any .PAPCO air-operated dampers
have
been
subsequently modified to include handwheels, and if so., whether proper
components were used.
Based on the licensee's action in this area, the
inspector considers that this item is being properly resolved. This item
is closed.
12.
Followup on Temporary Instruction 2500/19 (25019)
(Closed) Temporary Instruction 280; 281/T2500/19, Inspection of Licensee's
Actions Taken to Implement Unresolved Issue A-26: Reactor Vessel Pressure
Transient Protection for Pressurized Water Reactors .
During this inspection period, the inspector reviewed the licensee's
installed systems
and
program
for mitigation
of
low-temperature
overpressure transients in accordance with commitments.
This review
15
included verification of the design,
administrative controls and
procedures, training, equipment modification, and surveillance programs
implemented due to the subject concern.
The licensee completed all
r~quired commitments in 1985 and considers that the isiue is closed.
The
inspector reviewed each area as described below and concurs with the
licensee's assessment.
Design
The
inspector
reviewed
documentation
including
the
Technical
Specifications which requires that the overpressure protection system be
operable and in operation prior to the reactor coolant system temperature
decreasing below 351 degrees F if a bubble does not exist in the
pressurizer or if the RCS is not vented through an open power operated
relief valve.
Also, all but one charging pump must be demonstrated
inoperable every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
A review was also conducted of applicable
design drawings to verify that the two trains of overpressure protection
are independent and a single failure in one train does not disable the
other train.
Finally, documentation was reviewed to verify that the
pressure chosen for systems actuation was conservative to ensure that
10 CFR 50, Appendix G limits were not exceeded. No discrepancies were
identified.
Administrative Controls and Procedures
The inspector reviewed. administrative and operational procedures and
verified that:*
Controls are in place to minimize time in a water solid condition.
Controls are in place to minimize temperature differentials between
the steam generators and reactor vessel while solid and prior to
unisolating a reactor coolant loop or starting of a reactor coolant
pump.
Controls are in place to limit the number of operable charging pumps
during the time that the reactor coolant system is less than 351
degrees F.
Also, the inspector verified that the installed system is in accordance
with license requirements and that procedures for manual alignment of the
system when required, and removal of the system when not required, are in
place and being used during each startup and shutdown.
No descrepancies
were identified.
Training and Equipment Modifications
The inspector held discussions with the operations superintendent and
verified that all operators have received training in RCS low-temperature
overpressure events and that procedures require proper alignment of the
16
mitigation system and removal from service of unneeded pressure sources.
Also, modifications that are made to the overpressure mitigation systems
receive proper design reviews and are tested after installation to ensure
operability in accordance with requirements.
The inspector also verified
that alarms are installed in the control room to warn the operators of
pressure transients which could challenge the overpressure mitigation
systems.
Also, backup air bottle supplies are required to be pressurized
to 1000 psig prior to the system being declared operable.
No
discrepancies were identified.
Surveillance
The inspector reviewed the appropriate surveillance instructions and
determined that system operability is verified prior to entering a
condition requiring the systems. After performing corrective maintenance
on the system, and stroke times on the PORVs are verified using the backup
air bottle supplies.
No discrepancies were identified.
Based on the preceding review, the inspector considers that all action
necessary to clos~ this item has been accomplished.
13.
Inspection Report 280; 281/87-11 Findings
Inspection Report 280; 281/87-11 documented three potential enforcement
findings.
These findings were discuised during the inspection and in a
conference call between Region II, Vendor Inspection Branch, NRR, and the
licensee on August 31, 1987.
The findings are discussed in detail in
IR 280; 281/87-11 and constitute a violation (280; 281/87-21-06) of NRC
requirements.
ENCLOSURE 3
PROPOSED MEETING AGENDA
Virginia Electric and Power Co~pany Meeting with NRC
September 24, 1987
I. Opening Remarks
II. Issues of Concern
A.
Failure to perform a 10 CFR 50.59 safety evaluation
prior to deleting the turbine valve freedom test
described in the Final Safety Analysis Report.
1.
Discuss the VEPCO review process prior to
deviating from commitments described in the
FSAR.
Discuss the review process for deleting the
turbine valve freedom test and reason for
deleting the test.
3.
Does VEPCO believe that deleting this test
increases the probability of occurrence of
4.
Was obtaining prior Commission approval for
deleting this test considered?
5.
What actions will be taken by VEPCO concerning
performance of this test and performing
10 CFR 50.59 evaluations in general?
B.
Failure to perform a surveillance test required by
Technical Specifications 4.6.A.l.b and continued
operation without performing the test.
1.
Discuss the circumstances resulting in the
failure to perform the surveillance procedures.
2.
Discuss actions taken once the NRC pointed
out, on June 29, 1987, that the surveillance
had not been performed including actions taken
to ensure proper diesel generator operation .
NRC
Enclosure 3
3.
4.
5.
6.
2
Discuss operability as defined in the Technical
Specifications and how it relates to operability
of the diesels considering this surveillance
had not been done.
Why were the diesels not declared
Does VEPCO consider Surry Power Station to be in
violation of the Technical Specifications with
regard to this missed surveillance.
Discuss VEPCOs justification for continuing to
operate with a Technical Specification required
surveillance not performed and without seeking
any temporary waiver of the requirement.
Discuss the effect of a failure of the components
required to be tested by this surveillance.
III. Closing Remarks
.
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I
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NRC