ML20151T790
| ML20151T790 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 01/31/1986 |
| From: | Jenison K, Linda Watson, Weise S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20151T743 | List: |
| References | |
| 50-327-85-47, 50-328-85-47, NUDOCS 8602100363 | |
| Download: ML20151T790 (10) | |
See also: IR 05000327/1985047
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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101 MARIETTA STREET, N.W.
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ATLANTA, GEORGI A 30323
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Report Nos.:
50-327/85-47, 50-328/85-47
Licensee:
Tennessee Valley Authority
6N38 A Lookout Place
1101 Market Street
Chattanooga, TN 37402-2801
Docket Nos.:
50-327 and 50-328
License Nos.: DPR-77 and DPR-79
Facility Name:
Sequoyah Units I and 2
Inspection Conducted:
December 6 1985 thru January 5, 1985
Inspectors:
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K. M. Jenison, Seng Refident inspector
Date Signed
MA
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L. J. Watson, Res en
nspector
Dat Si ned
3/ N
' Approved by:
S. P. Weise, Section Chief
Date 71gned
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Division of Reactor Projects
SUMMARY
Scope: This routine, announced inspection involved 276 resident inspector-hours
onsite in the areas of operational safety verification including operations
performance, system lineups, radiation protection, security and housekeeping
inspections; surveillance and maintenance observations; review of previous
inspection findings; followup of events; review of licensee identified items; and
review of inspector followup items.
Results: One violation for failure to follow procedure was identified with three
examples.
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REPORT DETAILS
1.
Licensee Employees Contacted
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H. L. Abercrombie, Site Director
- P. R. Wallace, Plant Manager
- L. M. Nobles, Operations and Engineering Superintendent
- B. M. Patterson,- Maintenance Superintendent
- J. M. Anthony, Operations Group Supervisor
- R. W. Olson, Modifications-Branch Manager
- M. R. Sedlacik, Electrical Section Manager, Modifications Branch
- H. D. Elkins, Instrument Maintenance Grouo Manager
- M.~ A. Scarzinski, Electrical Maintenance Supervisor
- M. R. Harding, Engineering Group Manager.
D. C. Craven, Quality Assurance Staff Supervisor
- D. L. Cowart, Quality Surveillance Supervisor
- D. E. Crawley, Health Physics Supervisor
- G. B. Kirk, Compliance Supervisor
M. L. Frye, Compliance Engineer
H. R. Rogers, Compliance Engineer
- R. C. Birchell, Compliance Engineer
- D. H. Tullis, Mechanical Maintenance Group Supervisor
J. H. Sullivan, Regulatory Engineering Supervisor
- R. J. Griffin, NSRS Sequoyah Site Representative
- C. L. Wilson, Nuclear Engineer, NSS
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- E. W. Whitaker, Nuclear Licensing Engineer, TVA-NLB
- C. E. Chmielewski, Nuclear Engineer 2,
- L. D. Alexander, Mechanical Section Manager, Modifications Branch
- W. Murhead, QA Specialist, QAB
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- E. Bosley, QA Evaluator, QAB
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Other licensee employees contacted included technicians, operators, shift
engineers, security' force members, engineers and maintenance personnel.
- Attended' exit interview
2.
Exit Interview (30703)
The inspection scope and findings were summarized with,the Plant Manager and
members of his staff on January 8,1986. One violation hith three examples
described in paragraphs 5 and 7 were discussed. The licensee acknowledged
the inspection findings.
The inspectors acknowledged that . the safety
significance of the violation examples was low, but pointed out that they
were indicative of either lack of attention to detail or lack of an
( understanding of the administrative control . requiremants for procedure
changes versus adherence. The licensee did not identify as proprietary any
of the material reviewed by the inspectors during this inspection. During
the reporting period, frequent discussions were held with the Site Director,
Plant Manager and other managers concerning inspection findings. At no time
during the inspection was written material providad to the licensee by the
inspector.
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3.
Licensee Action on Previous Inspection Findings (92702)
(Closed) Violation 327,328/85-17-01.
Failure to Follow Procedure for
Installation.of Vital Battery V.
The inspector reviewed MR A-545415 and MR
A-545416 and verified.that intercell and end spacers were installed with no
more than one quarter inch clearance.
4.
Unresolved Items
Unresolved items are matters about which more information is required to
determine whether they are acceptable or may involve violations or
deviations.
One unresolved item identified during this inspection is
discussed in paragraph 9.
5.
Operational Safety Verification (71707)
a.
Plant Tours
The inspectors observed control room operations, reviewed applicable
logs, conducted discussions with control room operators, observed shift
turnovers,
and confirmed operability of instrumentation.
The
inspectors verified the operability of selected emergency systems,
reviewed tagout records, verified compliance with Technical Specifi-
cation (TS) Limiting Conditions for Operation (LCO) and verified return
to service of affected components.
The inspectors verified that
maintenance work orders had been submitted as required and that
followup activities and prioritization of work was accomplished by the
licensee.
Tours - of the diesel generator, auxiliary, control, and turbine
buildings and containment were conducted to observe plant equipment
conditions, including potential fire hazards, fluid leaks, and
excessive vibrations and plant housekeeping / cleanliness conditions.
During walkdowns of the Auxiliary Building and Turbine Buildings, two
cases were identified where pipe hangers were inadequate.
One case
involved a pipe hanger on Safety Injection System piping that had
apparently been removed for valve maintenance during the current outage
and had not been replaced. The second case involved two missing nuts
on the pipe supports for feedwater bypass line on the Main Feedwater
System.
Both of these items have been referred to NRC regional
specialist inspectors for followup.
The inspectors walked down accessible portions and/or reviewed control
room indications of the following safety-related systems on Unit I and
Unit 2 to verify operability and proper valve alignment:
Residual Heat Removal System (Units I and 2)
Charging Pump Flowpath (Units 1 and 2)
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Control Room Ventilation Chlorine Detection System (Common)
Auxiliary Building Gas Treatment System (Unit 1 and 2)
Auxiliary Control Air (Units 1 and 2)
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The inspectors conducted a walkdown to review Abnormal Operating
Instruction, AOI-27, Control Room Inaccessibility, Units 1 and 2.
The
inspectors noted the following items:
(1) The reactor operator. providing information on the procedure was
well informed.in the use of the procedure and had participated in
annual drills of the procedure.
(2) The procedure stated that certain material (controlled copies)
were to be maintained in the Auxiliary Control Room ( ACR). These
materials were maintained in a locked cabinet outside the ACR.
The licensee committed to correct this administrative error in the
procedure. The list of material included telephone directories.
The inspectors noted that these directories were not up-to-date.
In addition, an administrative error consisting of the entry of a
procedure revision into the wrong procedure was noted.
The
licensee stated that these items would be corrected.
Followup on
corrective action for these administrative errors is identified as
Inspector Followup-Item (327,328/85-47-01).
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(3) A caution in the procedure, which would require an operator to
swap suction for the turbine driven Auxiliary Feedwater pumps from
the Condensate Storage Tank to the Essential Raw Cooling Water
System at 11.5 psig, was incorrect.
The reactor operator stated
that the value had been changed from 11.5 to 13 psig. The inspec-
tors asked the licensee to provide additional information on
the methods used to update procedures when critical parameters
change. The inspector reviewed Administrative Instruction AI-19,
Plant Modifications, which requires the review of-plant procedures
when modifications .are made to the plant or critical parameters
are changed. For Operations procedures, the licensee stated ths.t
an SRO or RO reviewed procedures that he/she determined to be
affected.
This omission appears to be an isolated case,
therefore, the item will be identified as Inspector Followup Item
(327,328/85-47-02) for additional NRC review of control of pro-
cedure revisions due to modifications and update of appropriate
procedures to reflect the correct setpoint discussed above.
No' violations or deviations were identified.
b.
Security
During the course of the inspection, observations relative to protected
and vital area security were made, including access controls, boundary
integrity, search, escort, and badgin3
No violations or deviations
were identified.
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c.
Radiation Protection
The inspectors observed Health Physics (HP) practices and verified
implementation of radiation protection control. On a regular basis,
radiation work permits (RWPs) were reviewed and specific work
activities were monitored to assure the activities were being conducted
in accordance with applicable RWPs.
Selected radiation protection
instruments were verified operable and calibration frequencies were
reviewed.
On January 2, 1986, the inspectors observed three TVA employees clean-
ing borated water from the spill basin surrounding the boric acid
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transfer pumps. RWP 02-0-86668 required the workers to wear a top and
bottom plastic suit. Two of the three individuals failed to meet the
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' RWP requirement .in that they had unfastened the top front portion of
the plastic suit to increase air circulation to their upper bodies.
Because of the nature of the work that was involved (bailing contami-
nated water with buckets) the above actions increased the potential of
skin contamination.
In addition, the foreman who was in charge of the
cleanup activities, was standing just outside of the contaminated area
and was observing the workers.
When questioned as to what the
requirements of the RWP were, he did not appear to know.
Technical Specification 6.11 states that procedures for personnel radiation
protection shall be prepared consistent with the requirements of
10 CFR 20 and shall be approved, maintained and adhered to for all
operations involving personnel radiation exposure.
Sequoyah Nuclear
Plant Radiological Control Instruction RCI-14, Radiation Work Permit
(RWP) Program, requires'each individual to wear all required protective
clothing and equipment as required when entering an RWP area. Failure
to properly wear protective clothing required by an RWP is a Violation
(327/328-85-47-03).
On December 12, 1985, a sheetmetal worker was internally contaminated
while working on auxiliary building ventilation ductwork.
He was
decontaminated by the SNP Health Physics staff and his body burden is
being evaluated. This item will be reviewed by an NRC Region II Health
Physics specialist during the next scheduled inspection (Inspection
Report 50-327, 328/85-04) and will be tracked as Inspector Followup
Item (327, 328/85-47-04).
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6. i Monthly' Surveillance Observations (61726)
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The inspectors observed Technical Specification (TS) required surveillance
testing and verified that testing was performed in accordance with adequate
procedures; that test instrumentation was calibrated; that Limiting Condi-
tions for Operation were met; that test results met acceptance criteria
requirements and were reviewed by personnel other that the individual
directing the test; that deficiencies were identified, as appropriate, and
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that any deficiencies identified during the testing were properly reviewed
and resolved by management personnel; and that . system restoration was
adequate. - For . complete tests, the inspector verified that testing
frequencies were met and tests were performed by qualified individuals.
The inspector witnessed / reviewed portions of the following surveillance test
activities:
SI-2, Operations Shift Log
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Functional Testing of 1-FCV-63-22 in accordance with WP 11866
.SI-7, Electrical Power System: Diesel Generator
IMI-99, Reactor Protection System
No violations or deviations were identified.
7.
Monthly Maintenance Observations (62703)
Station maintenance- activities of safety-related systems and components
were observed / reviewed to ascertain that they were conducted in accordance
with approved procedures, regulatory guides, industry codes and standards,
and in conformance with TS.
.The following items were considered during this review:
LCOs met while
components or systems were removed from service, redundant components
operable, approvals obtained prior to initiating the work, activities
accomplished using approved procedures and inspected as applicable, pro-
cedures adequate to control the activity, troubleshooting activities
controlled and repair records accurately reflected what actually took place,
functional testing and/or calibrations performed prior to returning com-
ponents or systems to service, quality control records maintained, activ-
~1 ties accomplished by qualified personnel, parts and materials used properly
certified, radiological controls implemented, QC hold points established
where required and observed, fire prevention controls implemented, outside
contractor force activities controlled in accordance with the approved
Quality Assurance (QA) program, and housekeeping actively pursued.
The following maintenance activities were observed / reviewed by the
inspectors:
WP 11473
Installation of Fire Protection Piping to Correct Areas of
Adequate Spray
WP 11850
Torquing of ERCW Check Valves to Upper Containment Compart-
ment Coolers (2-VLV-67-786A, 786C)
.WP 11850 required that four check valves be installed in the ERCW lines
.to the upper compartment coolers. During the inspection of WP 11850, the
inspector noted that the' work plan required, "at completion of the check
valve installation, QC inspector shall verify that valve internals meet
Class D cleanliness requirements per TI-70."
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A QC inspector signature was found for valves 2-VLV-67-786B and -786D.
The blanks had not been signed by a QC inspector for valves 2-VLV-67-786A
and -786C, for which torquing had just been completed. The WP had not been
followed in that a QC inspection had not been performed, however, Mainte-
nance Instruction MI 11.4, Maintenance of CSSC Valves, had been used by the
pipefitters to verify cleanliness.
The inspector discussed the problem with the cognizant engineer.
The
cognizant engineer had been notified that the Work Plan had not been imple-
mented and determined that since the piping and valves were Class C that a
QC signoff was not required and therefore there was no reason to reverify
cleanliness. The engineer was processing an intent change to delete the QC
holdpoint.
The inspector reviewed Administrative Instructions AI-19, Plant Modifica-
tions, and AI-4, Plant Instructions - Document Control. The instructions
state that upon discovery of a discrepancy, the cognizant engineer will be
notified and will resolve the discrepancy. The engineer's actions in this
case appear to be appropriate. The inspector noted that the procedures do
not specifically address the corrective action to be taken or how it will
be documented. The inspector brought this to the attention of the licensee.
Failure to follow a procedure for modification to a safety-related system
-is a violation; however, since the failure to follow procedure was identi-
fied by the licensee and corrective action was being taken, this is con-
sidered licensee identified and will not be cited.
Also during the review of WP 11850, the inspector determined that data
sheets documenting the disassembly .of the valve had not been completed
as required by MI 11.4.
Failure to follow the procedure for installation
of the ERCW check valves is identified as an example of Violation (327,
328/85-47-03).
MR A 563333
RHR Pump 2A-A Minimum Flow Valve 1-VLV-74-12 MOVATS.
Qualification
Maintenance Data
System maintenance
on
Rosemount Pressure Transmitter 1-FT-3-147, Steam Generator
- 3 Auxiliary Feedwater Inlet Flow Transmitter
MR A 593499
Replacement of NAMCO limit switch on Steam Generator
Blowdown Isolation Valve
Maintenance Instruction MI 10.37, NAMCO Limit Switches, was used during the
performance of Maintenance Request A-593499. Measuring and Test Equipment
(M&TE) prescribed in step 5.2 of MI 10.37 included a 0-30 inch pound torque
screwdriver. The torque screwdriver used in step 6.1.1.2 of MI 10.37 to
replace the bottom cover gasket was not a 0-30 inch pound torque screw-
driver, and a "non-intent" change was not made per AI-4 to reflect as such.
However, the value to which the cover was torqued appeared to be within the
accuracy range of the tool actually used. This is identified as a failure
to follow procedure and is a further example of Violation (327, 328/85-
47-03).
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8.
Licensee Event-Report (LER) Followup (92700)
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The lfollowing LERs were reviewed and closed. The inspector verified that:
reporting requirements had been met; .causes had been identified; corrective
actions appeared appropriate; generic applicability had been considered; the
LER forms were complete; the licensee had reviewed the event; no unreviewed
safety questions were' involved; and violations of regulations or Technical
Specification conditions had been identified.
LERs Unit 1
327/83007
Auxiliary Feedwater Level Control Valve Inoperable
327/83009
. Control Rod Position ~ Indication Inoperable
327/83154
Waste Gas Decay Tanks With High Oxygen Concentration
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327/83167
Waste Gas Decay Tanks With High Oxygen Concentration
327/84019
Ice Condenser Ice Weights Below Design Requirements
327/84044
Diesel Generator Start During Surveillance Testing
327/84062
Control Room Isolation
327/84065
Auxiliary Building Isolation
LERs Unit 2
328/82115
Containment Personnel Air Lock Inoperable (Rev. 1)
328/83049
Inoperable Intermediate Ice Condenser Door (Rev. '1)
328/83093
Condenser Vacuum Flow Rate Monitor Inoperable (Rev. 1)
328/84017
Reactor Trip on Low-Low Steam Generator Level
328/84018
Three Reactor Trips on Low-Low Steam Generator Level
9.
Event Followup (93702, 62703, 61726)
a.
On December 9,
1985, with Unit 2 in Mode 5 at 127F and 150 psig,
instrument technicians were calibrating the Unit 2 reactor coolant loop
- 3 wide
range
pressure transmitter,
2-PT-68-66.
. Surveillance
Instruction, SI-484, Periodic Calibration of the Reactor Vessel Level
Instrumentation (RVLIS) and RCS Wide Range Pressure Channels (P-403,
P-406) (Refueling Outage), was being utilized.
Step 1.1 of Appendix A.1 of the procedure requires the removal of input
wires to PS-68-66B/E to~ prevent closing of valve 2-FCV-74-2, the
Residual Heat Removal (RHR) suction valve.
This interlock protects
against overpressurization of the RHR System.
During performance of
the calibration, -operators noted that the RHR suction valve,
2-FCV-74-2, started closing. The operator promptly removed the Train A
RHR pump from service and notified the instrument technicians of the
valve closure.
The technicians re-terminated the input wires and
stopped the work. The system was restored to operation in 8 minutes.
There was no increase in RCS temperature. Maintenance Request (MR)
AS46561 was written to determine the source of the problem.
The
reactor operators took RHR out of service for approximately 50 minutes
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during the troubleshooting work.
The licensee simulated the event by
relifting the input wires and determined that when the wires were
lifted together that a circuit was created that caused a valve ~ closure-
signal. .The licensee is revising the procedure to prevent a similar-
. occurrence in the future. Notification of all-Westinghouse plants was
Lperformed through the Westinghouse news letter.
No1 violations or deviations were identified.
b.
- 0n December 13,-1985, it was . discovered that vendor axial flux W(z)
curve data for Unit 2 incore calculations was entered incorrectly into
the prime' computer code. This W(z) curve data had-been entered for use
in Unit 2 - Cycle 3 axial flux and local power peaking factor
calculations.
After the error' was identified, preliminary data. was
analyzed and .it was determined by the licensee that no safety limits
were violated. The error was identified by the licensee during the
performance of Surveillance Instruction SI-126, Incore Calculation.
NRC review of whether or not a safety limit was violated is Unresolved'
Item.(327,328/85-47-05).
c.
On December 18, 1985, the non-safety-related station air supply to a
door gasket, located in the barrier door separating the spent fuel pool
and the~ transfer canal, was lost. The deflated door began to leak
spent fuel pool. water into the transfer canal resulting in a two foot
' drop in spent fuel pool level in approximately forty minutes. Thr. leak
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was essentially stopped by repressurizing the door seal with safety
related air.
Spent fuel pool level was subsequently returned to a
normal level of greater than twenty three feet above the irradiated
fuel.
The practice of inflating the door with non-safety-related air
and operating with the transfer canal dry will be reviewed as Inspector
Followup Item (327, 328/85-47-06).
No violations or deviations were identified.
10.
Inspector Followup Items (92701)
(Closed) IFI 327,328/85-16-05.
Surveillance Testing for Diesel Generator
Batteries.
The inspector reviewed the performance testing of the four
diesel generator batteries.
The performance tests were conducted using
Surveillance Instruction SI-238.2, Diesel Generator Battery Capacity Test,
Units 1 and 2,
for batteries IA-A, IB-B, 2A-A and 2B-B on October 18,
September 11, October 19, and September 13, 1985, respectively. The results
of these tests.were reviewed. One error was identified in the calculation
of the capacity of battery 1A-A. The capacity had been entered as 97% when
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the actual capacity indicated (with the chosen test parameters) was 91%.
This value would not have resulted in additional testing per IEEE 450-1980.
'It should be noted that the test was term 1nated at a conservative value of
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105 volts when IEEE 450-1980 would have allowed a discharge to 99 volts.
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In addition, the inspector reviewed SI 238.2 -against IEEE 450-1980. The
licensee has an internal commitment to conduct diesel generator _ battery
service and performance discharge testing in accordance with this standard.
.Two typographical errors involving a reference to an incorrect step number
and an incorrect title for a key' were noted which did not affect the results
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of the tests. Also, one reference to a table in IEEE 450-1975 in lieu of
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the 1975 table was utilized to eliminate requirements for calculations. of
temperature corrected test voltage levels prior to the test. Instead, the
test results are corrected for temperature after the test using the method
in IEEE 450-1975.
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