ML18152A372
| ML18152A372 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 02/25/1988 |
| From: | Cantrell F, Holland W, Larry Nicholson NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18152A373 | List: |
| References | |
| 50-280-88-01, 50-280-88-1, 50-281-88-01, 50-281-88-1, NUDOCS 8803160308 | |
| Download: ML18152A372 (13) | |
See also: IR 05000280/1988001
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTA STREET, N.W.
ATLANTA, GEORGIA 30323
Report Nos.:
50-280/88-01 and 50-281/88-01
Licensee:
Virginia Electric and Power Company
Richmond, Virginia 23261
Docket Nos.:
50-280 and 50-281
Facility Name:
Surry 1 and 2
License Nos.:
Inspection Conducted: January 3, through January 31, 1988.
'
.
Inspectors: *)i r (;.>--LHl
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-W~. -E-. -H~o_l_l_a_n~d-,-Se-n-,-. o_r_R_e_s_i_d_e_n_t_I_n_s_p_e_c_t_o_r _____ _
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L. E. Nicholson1, /Resident Inspector
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SUMMARY
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Date Silgned
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Date .Signed
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Date Signed
Scope:
This routine inspection was conducted in the areas of plant operations,
plant maintenance, plant surveillance, followup on inspector identified items,
licensee event report review, and quality assurance program review.
Results: Two violations were identified in this inspection report.
8803160308 880225
ADOCK 05000280
G
REPORT DETAILS
1.
PERSONS CONTACTED
Licensee Employees
- W, L. Stewart, Vice President, Nuclear Operations
- 0. L. Benson, Station Manager
H. L. Miller, Assistant Station Manager
- E. S. Grecheck, Assistant Station Manager
- J. A. Bailey, Superintendent of Operations*
- 0. 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 Engineering
- G. D. Miller, Licensing Coordinator, Surry
- G. G. Harkness, Licensing Coordinator, North Anna
- F. P. Mone, Supervisor, Quality
- Attended exit meeting.
Other licensee employees contacted included control room operators,
shift technical advisors, shift supervisors and other plant personnel.
The* NRC,
Region II Branch Chief, B. A. Wilson, and Section Chief,
F. S. Cantrell, attended the exit meeting on February 2, 1988, in
conjunction with a plant visit.
2.
Exit Interview
The inspection scope and findings were summarized on February 2, 1988,
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 provide adequate
procedure in the implementation of design control over substitution of
material for safety-related components (280; 281/88-01-01).
One violation (paragraph 6) was id~ntified for failure to follow procedure
during performance of surveillance testing by plant operators (280;
281/88-01-02).
The licensee acknowledged the inspection findings with no dissenting.
comments.
The license did not identify as proprietary any of the
materials provided to or reviewed by the inspectors during this
inspection.
' *
2
3.
Plant Status
Unit 1
4.
Unit 1 began the reporting period at power.
The unit operated at power
for the duration 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)
(Open)
Unresolved Item 280; 281/87-09-01, Clarification of Requirements
for Flushing Sensitized Stainless Steel Pipe.
The issue involved the
adequacy of flushes performed by Periodic* Test Procedure *1-PT-18.5, *
11 Flushing of Sensitized Stainless Steel Piping", as required by Technical
Specification 4.1.E.
The licensee performed a detailed engineering study
(Techni ca 1 Report ME-0009) which concluded t.hat a 1 though the current flush
procedures do not cover all sensitized piping, adequate flushing is being
performed if credit is taken for the normal testing of various pumps.
The
report suggested that a revision to the Technical Specification be
submitted to clarify the issue.
The inspector reviewed the engineering
study and discussed the results with the licensee.
This issue is made
more difficult due to the lack of specific,
detailed technical
specification requirements and the absence of formal documents detailing
the problem during plant construction.
The licensee is preparing a
revision to the Technical Specification that incorporates the engineering
conclusions; therefore, this item will remain open pending the submittal
and approval of this revision.
(Closed)
Violation
280; 281/87-26-01,
Failure to provide adequate
procedure and/or administrative controls to ensure containment integrity.
The subject violation was identified in inspection report 280; 281/87-05
and discussed with regi ona 1 management in an enforcement conference on
August 10, 1987.
The con cl usi on of the enforcement conference was that
the licensee failed to recognize, evaluate, and control conditions which
resulted in a degradation of containment integrity.
In the response to the violation dated October 29, 1987, the licensee
stated that a failure to document management evaluations and a weakness in
the operations department administrative controls were the general reasons
for the violation.
The inspectors reviewed the enhanced minutes of the
Station Nuclear Safety and Operating Committee (SNSOC) and attended
selected meetings to evaluate the management review of off-normal
conditions. The improvements made in this area are evident and effective.
The inspector also reviewed the overall administrative control upgrade
implemented by the licensee as a result of this violation and consider
them to be adequate.
This item is closed.
\\ *
3
(Closed)
Unresolved
Item
281/87-36-02,
Material
substitution
on
safety-related components.
This issue involves the discovery of hand-made
replacement parts found in a safety~related Limitorque valve actuator.
As
discussed in Inspection Report 280; 281/87-36, these parts were apparently
installed without any documented engineering evaluation,
review, or
approval; and ultimately contributed to the failure of a safety-related
component.
The inspectors discussed this situation with the licensee and
expressed concern with the system that allowed this to occur.
Although
Administrative Procedure SUADM-M-27,
"Requirements For A Repair Or
Replacement Follower
11 ,
requires that repairs which change original
specifications should be documented on an Engineering Work Request (EWR),
the review and evaluation performed in accordance with this procedure
occurs prior to starting the job, primarily for ASME Code Section XI
concerns.
No guidance is included in this administrative procedure
specifying the course of action that should be fo 11 owed if a materi a 1
substitution is required during the repair process.
The licensee Quality Assurance Program Operations Phase Topical Report VEP
1-5A, Amendment 5, states in paragraph 17.2.3 that the Nuclear Design
Control Interface Manual establishes the requirements for the review for
suitability of materials in safety-related components.
The inspectors
reviewed this corporate document and could not find guidance applicable to
the case discussed above .
A review of the information indicates that the substitution of the
hand-made tripper fingers discussed above appears to be an isolated
occurrence. Discussions with selected mechanics and foremans did reveal
that they were aware of the need to obtain approval prior to the
- substitution of parts, and this was supported by a large sample of
completed work orders reviewed by the in specters.
Although it can be
concluded that the failure to perform a review for suitability of
application for the hand-made tripper fingers as required by 10 CFR50,
Appendix B, Criterion III does not constitute the norm,
there currently
is no procedural requirement that specifies such review is necessary. The
requirements of 10 CFR 50, Appendix B, Criterion V, states in part that
activities affecting
quality
shall
be
prescribed
by
documented
instructions. The failure to provide and to implement adequate procedure
in the implementation of design control over substitution of material for
safety-related
components
is
identified
as
a
violation
(280;
281/88-01-01).
Within the areas inspected, one violation was identified.
5.
Unreso 1 ved Items
Unresolved items were not identified during this inspection period.
6.
4
Plant Operations
Operational Safety Verification (71707)
The inspectors 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; and review of control room operator logs, operating orders,
plant deviation reports, tagout logs, jumper logs, and tags on components
to verify compliance with approved procedures.
The inspectors conducted weekly inspections in the following areas:
verification of operability of selected Emergency Safety Features (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.
Daily 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
were performed by the inspectors.
The inspectors routinely monitor the
temperature of the auxi 1 i ary feedwater pump discharge piping to ensure
steam binding is prevented.
The inspectors conducted biweekly inspections in the following areas:
verification re~iew 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.
Certain tours were conducted on backshi fts or weekends.
Backshift or
weekend tours were conducted on January 6, 9, 11, 18, 19, 25, 28, and 30.
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, cable penetration
areas,
independent spent fuel
storage facility,
low level
intake
structure, and the safeguards valve pit and pump pit areas. Reactor
Coolant System (RCS) leak rates were reviewed to ensure that detected or
suspected leakage from the system was recorded, investigated, and
evaluated and that appropriate actions were taken, if required.
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.
Selected
' *
5
radiation protection instruments were periodically checked, and equipment
operability and calibration frequencies were verified.
In the course of monthly activities, the inspectors included a review of
the licensee's physical security program.
The performance of various
shifts of the security force was observed in the conduct of daily
activities to include: protected and vital areas access controls;
searching of personnel, packages and vehicles; badge issuance and
retrieval; escorting of visitors; and patrols and compensatory posts.
On January 5, 1988 the inspector witnessed the movement of a dry cask
loaded with spent fuel to the Independent Spent Fuel Storage Installation
(ISFSI). This is the fifth dry storage cask that is located at the ISFSI
site. The inspector reviewed the procedures being used at the jobsite and
monitored the coordination among the workers involved.
No discrepancies
were noted.
On January 9, the inspector conducted an inspection of the Unit 1 turbine
in 1 et va 1 ve freedom test in accordance with Periodic Test Procedure
l-PT-29.1, "Turbine Inlet Valve Test".
The inspector witnessed portions
of the unit rarnpdown to less than 70% power, the pretest briefing ana
actual valve testing. One deficiency, noted with the #2 left reheat and
intercept valves, was determined to be an instrumentation problem only .
The test was declared successful and the unit was returned to full power
operation.
The inspector noted during the test that the rod control system was in the
manual control mode.
Initial condition 3.5 of test procedure l-PT-29.1
requires that this test be conducted with the rod control system in the
automatic control mode.
The shift supervisor stated that he was aware of
the requirement and did not want to pl ace the rod contra 1 system in
automatic because of a relatively large mismatch between Tave and Tref.
The inspectors subsequently reviewed Operations Procedure l-OP-2 .1. 2,
11Decreasing Power From Existing Power Level To 2%11 , that was -used to
reduce power below the 75% level as required for the above test.
This
procedure does not differentiate between actions required to reduce power
to 70% as opposed to 2%, therefore, the operator must review this
procedure and submit a procedure deviation (temporary change) to remove
the steps not applicable each time this procedure is used to support
PT-29.1.
The inspector noted the following discrepancies in section 3.0,
initial conditions:
Initial condition 3.4 requires that the portable narrow range steam
generator level recorders be in place and operating. This step was
initialed by the control room operator indicating compliance with the
condition.
The inspector observed that no
1 eve 1 recorders, as
required by this step, were in use during the performance of the
procedure.
On January 26, the in specters reviewed the comp 1 eted
procedure 2-0P-2.1.2 that was used for pow~r reduction of Unit 2 on
January 23 in order to perform 2-PT-29.1, "Turbine Inlet Valve Test"
6
and found that the same initial condition 3.4 was initialed as being
performed by a different control room operator.
A review of the
strip charts from the portable level recorders that are used to
satisfy this requirement indicated that this equipment was last used
in December, 1987.
The licensee acknowledges that the recorders were
not used during the performance of.these procedures.
Initial condition 3.7.1 through 3 .. 7.4 requires that jumpers be
installed to remove the seal-in function from controllers of moisture
separator reheater steam supply isolation valves.
This initial
condition was not met or resolved prior to proceeding with Operations
Procedure 1-0P-2.1.2 on Janurary 9, 1988.
The shift supervisor
stated at the time that this requirement was only applicable for
power reductions.much greater than planned for this test.
Technical Specifications and paragraph 5.4.3.a of Surry Administrative
Procedure SUADM-ADM-21,
11Station Procedures
11 , states in part that *all
temporary changes or deviations to procedures shall be approved prior to
implementation.
Paragraph 3.1.2 of Administrative Procedure SUADM-0-10,
110perations Department Procedures
11 , further states that if a procedural
step cannot be fo 11 owed as written, then no procedura 1 steps sha 11 be
accomplished until a procedure change is approved.
In the instances
discussed above, no change or deviation to the appropriate procedures had
been implemented to delete these initial conditions prior to proceeding.
Procedure deviations were submitted for the rod control and jumper
requirements after the Unit 1 test was completed and the unit returned to
full power operation.
No reference was made in the procedure deviations
regarding the failure to use the portable steam generator level recorders.
Surry Technical Specification 6.4.A requires detailed written procedures
with appropriate check-off lists and instructions shall be provided for
testing of components and systems involving nuclear safety of the station.
Surry Technical Specification 6.4.D specifies that all
procedures
described in Technical Specification 6.4.A shall be followed.
Surry
Technical Specification 6.4.E and F specifies that temporary changes to
procedures described in Technical Specification 6.4.A may be made,
provided such changes are approved prior to implementation.
Therefore,
failure to properly establish or disposition initial condition 3.5 of
1-PT-29.1, initial conditions 3.4 and 3.7 of procedure 1-0P-2.1.2, and
initial condition 3.4 of procedure 2-0P-2.1.2 prior to procedure
performance is a violation of Technical Specification 6.4 (280; 281/
88-01-02).
- The inspectors were particularly concerned that after identification of
the initial condition disparity, and prior .to test performance, the
licensed control room operators did not take appropriate administrative
action in accordance with requirements to either satisfy the initi-al
conditions of the procedures or change the procedures prior to proceeding
with the Unit 1 test on January 9, 1988.
This issue was identified to
licensee management (operations superintendent) on January 11, 1988.
Licensee action at the time included re-instruction of the appropriate
7
shift personnel with regard to procedure adherence.
The verification of
initial conditions (events of the 11th and 26th) that had not been
performed on two separate occasions by two different licensed operators
was also of particular concern.
The inspectors concern was identified to
licensee management (station manager and both assistant station managers)
on January 26, 1988.
The inspectors voiced the findings identified above
and expressed particular concern for an apparent immediate problem of
licensed operators verifying action steps in procedure without actual
accomplishment of the step.
The
station manager acknowledged the
inspectors concern, stated that actions of this nature are not in
accordance with management requirements, and took immediate action to
personally discuss the issues with each licensed shift beginning with the
shift following the meeting. Additional followup action included issuance
of a memorandum to all station supervisors emphasizing that full
compliance with procedures is required. Additional followup meetings with
ope rat i ans shift personne 1 and station management were conducted after
interviews with operators indicated additional clarifications of procedure
requirements were warranted.
The inspectors consider that the immediate
actions taken by station management were adequate in providing an interim
solution to their concerns.
Engineered Safety Feature System Walkdown
(71710)
The inspector performed a walkdown of the accessible areas of the safety
related portions of the Component Cooling Water System.
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 violation was identified.
7.
Maintenance Inspections (62703)
During the reporting period, the inspectors reviewed maintenance
activities involving Motor Operated Valves (MOV) to assure compliance with
the appropriate procedures.
The inspector conducted a comprehensive review of the maintenance
procedures used to repair and adjust safety related MOVs.
This inspection
included work performed during both scheduled actuator overhauls and
emergency repairs required during -normal power operations.
In the cases
where the instructions would be the same regardless of the actuator model,
the inspector selected the L imi torque Mode 1 SMB-00 as a representative
actuator due to the large quantity of these actuators used in the plant.
The specific maintenance procedures reviewed during this effort were as
fo 11 ows:
8
MMP-C-MOV-178,
11 Removal And Overhaul Of Limitorque Model SMB-000
Through SMB-00 And SBOOO Through S800
11 , dated 12/11/86
MMP-C-MOV-178.1,
"Removal And Overhaul Of Limitorque Model SMB-0
Through SMB-4 And SB-0 Through SB-4
11 , dated 12/11/86
EMP-C-MOV-11,
11 Disconnect And 'Reconnect Safety Related MOV 1 s
11 , dated
10/15/86
EMP-C-MOV-18,
"Safety Related MOV's - Repair, Replacement, Checkout
and Adjustments", dated 9/18/86
EMP-C-MCC-152,
11 Rep 1 a cement Of TOL I s [Therma 1 Over Load] For MDV
Feeder Breakers In Safety Related Motor Control Centers 11, dated
8/18/87
The review of the above procedures included examining selected completed
procedures and their corresponding work order packages retrieved from
station records for work that had been performed during the 1 ast two
years.
The general areas examined during this review included the
following:
Conformance
of
the
procedure
to
the
appropriate
licensee
administrative requirements, format, and control.
Post-maintenance testing performed was adequate for the repairs made.
Guidance provided in the procedures were technically adequate,
including incorporating lessons learned that are not specifically
covered in the vendor manuals.
Compliance with the procedures and controls.
The inspector found the 1 i censee program for repair of motor operated
valves to be in compliance with accepted standards and techniques.
The
procedures include extensive detail and reflect a thorough knowledge of
the subject.
Completed work packages reviewed from station records were
found to be acceptable in accordance with administrative procedures.
No
discrepancies were noted.
Within the areas inspected, no violations or.deviations were identified.
8.
Surveillance Inspections (61726)
During the reporting period, the inspectors reviewed various
surveillance activities to assure compliance with the appropriate
procedures as follows:
Test prerequisites were met.
Tests were performed in accordance with approved procedures.
9.
9
Test procedures appeared to perform their intended function.
Adequate coordination existed among personnel involved in the test.
Test data was properly collected and recorded.
Inspection areas included the following:
On January 4,. the inspector reviewed the completed Unit 1 monthly periodic
test procedure (l-PT-18.8), "Charging Pump Component Cooling and Service
Water Performance".
No discrepancies were noted.
On January 5, the inspector witnessed portions of the monthly operability
test of the emergency diesel generator #3 in accordance with test
procedure 1-PT-22.3C, "Diesel Generator No. 3 Test".
No discrepancies
were noted.
On
January 21,
the inspectors witnessed porti-0ns of the monthly
surveillance test 2-PT-8.1, "Reactor Protection Logic".
This test is
performed to ensure the continued proper operability of the reactor trip
portion of the reactor protection system as required by Technical
Specification Table 4.1-1.
No discrepancies were noted .
On January 25, the inspectors witnessed the testing of the containment
outside recirculation spray pump 2-RS-P-2A in accordance with Periodic
Test Procedure 2-PT-17. 3.
This monthly operabi 1 ity test verifies the
shutoff head and vibration of the pumps as required by Technical
Specification 4.5.
The pump and* its associated components responded as
required.
No discrepancies were noted.
Within the areas inspected, no violations or deviations were identified.
Followup on Inspector Identified Items
C92701)
(Closed)
Inspector Followup
Item (IFI)
281/87-05-02,
Followup
on
corrective action for Unit 2 RHR pumps during the Unit 2 outage in
December, 1987.
This item was identified in Inspection Report 280;
281/87-05.
In that report the licensee committed to make repairs to the A
RHR pump during the subject outage. Also, the licensee confirmed, during
the shutdown of Unit 2 in December, that the B RHR pump motor upper
bearing was operating within limits prior to starting work on the A pump.
Corrective action for the A RHR pump included complete replacement of the
pump motor, impeller, and stand.
The inspector reviewed the completed
work packa~e which was used to replate the pump (MMP-C-RH-015), and also
reviewed the post maintenance test results which demonstrated operability.
No discrepancies were noted.
This item is closed.
(Closed) I FI 280; 281/87-09-02, Fo 11 owup on review of the corrective
action process.
This item was identified in Inspection Report 280;
281/87-09.
In that report the inspectors noted that the licensee was
conducting a review of the implementing procedure for identification of
- '
10.
10
deviations and the use of the deviation reports for tracking and trending
of corrective actions implemented.
This issue was also discussed in the
cover letter to Inspection Report 280; 281/87-26 dated September 29, 1987.
The letter requested that Virginia Power describe those actions taken or
planned to improve the effectiveness of the corrective action process for
identification and evaluation of nonconformance conditions.
Licensee
response to that report dated October 29, 1987, stated in part
11 increased
management attention has been devoted to the procedural requirements for,
and the functioning of, the corrective action process.
Enhancements to
the Nuclear Operations Department Standard (NODS-QA-01)
have
been
completed....
Corresponding
rev1s1ons
to
appropriate
station
administrative procedures will be made.
Similarly, the Engineering and
Construction Department has adopted a new procedure (NDCM Proc. 6.1) which
provides for a formal
process to identify and evaluate potential
discrepancies.
11
The inspector reviewed NODS-QA-01, Rev. 1, and the new NDCM Procedure 6.1.
I~ addition, the inspector reviewed Surry Administrative Procedure
SUADM-0-12,
revision dated December 14, 1987, which implements the
requirements of NODS-QA-01.
The
inspector considers that the two
corporate level procedures provide adequate guidance to the stations in
implementation of requirements for identification and correction of
conditions adverse to qua 1 i ty.
The inspector . discussed the current
revision of SUADM-0-12 with station management and considers that it also
provides for appropriate guidance to station personnel in implementation
of a program which will identify and correct conditions adverse to
quality.
This upgraded program has already been observed by the
inspectors as providing more uniform guidance in the identification of
station deviations.
This item is closed.
(Closed)
IF! 280; 281/87-31-01, Followup on emergency diesel generator
adjustments and testing.
This item involved the corrective actions for
the
11 start failure
11
alarm that was noted on the EOG #3 during the
performance of the October 1987 monthly operability surveil 1 ance test.
The inspectors witnessed troubleshooting and adjustments of the start
circuit speed and time relays in accordance with maintenance procedure
EM~-C-EE-215,
and reviewed the subsequent conclusions documented on
Engineering Work Request (EWR)87-387. The specific recommendations made
in this EWR were discussed by the licensee in the station safety committee
and an appropriate safety eva 1 uat ion was performed.
The inspectors
witnessed two starts of this EOG following the adjustments and concluded
that the circuitry is functioning as required.
This item is closed.
Licensee Event Report (LER) Review
(92700)
The inspector reviewed the LERs listed below to ascertain whether NRC
reporting requirements were being met and to determine appropriateness of
the corrective action(s).
The inspector's review also included followup
on
implementation
of corrective action and
review of licensee
documentation that all required corrective actions were complete.
11
LERs that identify violation(s) of regulation(s) and that meet the
criteria of 10 CFR, Part 2, Appendix C,Section V shall be 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 280/87-28, EOG Output Breaker Protection Circuit Design
Deficiency.
The issue involved identification of a condition in that the
emergency di ese 1 generator ( EOG) output breakers contained instantaneous
overcurrent tripping relays in their protection circuits.
The relay
- settings were such that a fault on an individual safety bus load could
cause the EOG output breaker to trip before the load breaker would open to
isolate the fault. The instantaneous overcurrent trip relay settings were
increased on a 11 four emergency supply breakers to a setting that would
prevent this from occurring.
The inspector discussed this situation and
verified the implementation of corrective action.
This LER is closed.
(Closed) LER 280/87-31, Containment Isolation Valve Inoperable Due to
Mechanical Binding of Valve Operator. The issue involved inoperability~of
~ontainment instrument air isolation trip valve (1-IA-TV-100) during
surveillance testing. The cause of the problem was blockage of movement
of the valve by its electrical connector.
The electrical connector was
properly relocated and the valve was retested satisfactorily. Additional
inspection by the licensee for both units did not identify any similar
problems.
The inspector reviewed the
LER
and* conducted a random
inspection of containment isolation valves to verify the licensee 1 s
conclusions.
This LER is closed.
(Closed) LER 280/87-32, Protection System Channel Inoperable Due to Failed
Summator in Signal Conditioning Circuit.* The* issue involved failure of
the reactor coolant system Bloop averag~ temp~rature (Tave) protection
channel to fail high.
The operators, in response to the failure, properly
diagnosed the condition and placed the affected protection channels in
trip until repairs were completed.
The inspector reviewed the LER and
technical specifications and considers that conservative actions were
taken.
This LER is closed.
(Closed) LER 280/87-34, Iodine Spike Due to Defective Fuel Element.
The
issue fnvolved an increase in specific activity of the reactor coolant to
greater than 1. 0 mi crocuri es/cc dose equivalent I-131.
This condition
exceeds Technical Specification 3.1.D.2 and requires a special report as
outlined in Technical Specification 3.1.D.4.
The inspector reviewed the
LER and the requirements of Technical Specification 3.1.D.
This LER is
closed.
(Closed) LER 280/87-35, Required Surveillance Not Performed Due To Human
Error. The issue involved the failure to perform testing of the emergency
di ese 1 generator degraded/undervo ltage protection circuit as re qui red by
Technical Specification 4.6.1.b. This item was identified to the licensee
'
12
by the resident inspectors in Inspection Report 280; 281/87-17, and a
violation was issued in Inspection Report 280; 281/87-21.
The inspectors
witnessed the performance of and reviewed the results 'of speci a 1 tests
1&2-ST-206 that functionally tested the circuit enable/disable interlocks.
This LER is closed.
(Closed) LER 280/87-37, Emergency Diesel Generator Auto-Start Due to
Failed Relay and Blown Fuse.
The issue involved an engineered safety
feature actuation that occurred due to a capacitor failing to ground in
the bus IJ overvoltage circuit and causing a positive DC fuse in the
degraded voltage circuit to blow.
The licensee manually loaded the #3
emergency diesel generator onto the IJ Bus while repairs to the
undervoltage circuit were being implemented.
The inspector witnessed the
troubleshooting and repair of the above circuits and reviewed the circuit
diagrams with the licensee.
This item is closed.
11.
Quality Assurance Program Review
(35701)
On January 14, 1988, the inspectors visited the Virginia Power corporate
Quality Assurance (QA) Office in Richmond, Virginia, to meet the current
QA management, and to tour the QA offices.
A short presentation of the
Quality Assurance organization was presented to the inspectors by
Mr. R. J. Hardwick, Corporate Manager, Quality Assurance.
The inspectors
then asked questions with regards to present and future quality assurance
efforts at the nuclear stations, and discussed the requirements of the
current Quality Assurance Topical Report
11Topical Report Quality Assurance
Program Operations Phase VEP 1-SA, Amendment Five dated June 1986 11 *
Presently, the QA organization management is reviewing the Quality
Assurance Department Standards to insure they are current and provide
appropriate direction.
The inspectors considered that the meeting was
mutually beneficial and will continue to monitor program enhancements.