ML20138N206
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| ML20138N206 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 12/06/1985 |
| From: | Jenison K, Linda Watson, Weise S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20138N155 | List: |
| References | |
| 50-327-85-35, 50-328-85-35, NUDOCS 8512230406 | |
| Download: ML20138N206 (15) | |
See also: IR 05000327/1985035
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UNITgD STATES '
g .o NUCLEAR REGU'.ATORY COMMISSION
[[ - , REGION 11.
. g -- -j 101 MARIETTA STREET,N.W.
~*- 2 ATLANTA, GEORGIA 30323
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' Report Nos~.: 50.-327/85-35, 50-328/85-35
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Licensee: Tennessee Valley Authority
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6N11 B Vissionary Ridge Place-
"- --1101.Ma'rket Street
, Chattanooga, TN 37402-2801
- Doc ket < No s'. : 50-327'and 50-328 ' License Nos.: DPR-77 and DPR-79
' Facility Name: Sequoyah Units 1 and 2
. Inspection Conductea: Octeer 6 through November 5, 1985
Inspectors: 6Qd <W
K. M. Wnisof, Senior Resident Inspector
/A/05/B5
Dat'e Si'gned
. G 0. nd .Ww /G-loS/A5
L. J. W4tson,gResident Irspector Dat'e Signed
Accompanying. Personnel: G. . Pi
Approved by: 7/ ~
II
S. P. Weise,~ Section Chief
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DatE Signed
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Division of Reactor Projects
Summary
Scope: .This routine, announced inspection involv'ed-349 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;
walkdown-of Engineered Safety Features;; and review of inspector followup items.
:Results: One violation was identified - Failure to implement procedures'in the
areas of reactor trip response time testing (paragraph 7), installation, of, a
containment penetration'(paragraph 8), radiation monitor testing (paragraph'10); ;
and, configuration control of a radiation monitor power source (paragraph 10). ,
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B512230406 851210
PDR- ADOCK 05000327
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REPORT DETAILS
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- 1. Licensee Employees
Persons Contacted
H. L. A'oercrombie, 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
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M. R. Sedlacik, Electrical Section Manager, Modifications Branch
*H._D. Elkins, Instrument Maintenance Group Manager
G. B. Tiner,. Instrument Maintenance Engineer
*M. R. Harding, Engineering Group Manager
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*D. C. Craven, Quality Assurance Supervisor
*G. B. Kirk, Compliance Supervisor
, M. L. - Frye, Compliance Engineer
D. H.:Tullis, Mechanical-Maintenance Group Supervisor
J. H. -Sullivan, Regulatory Engineering Supervisor
*C, E. Bosley, Quality Assurance. Auditor
-Other licensee employees contacted included technicians, operators, shift
engineers, security force members, engineers and maintenance personnel.
* Attended exit interview
2. Exit Interview
The inspection scope and findings were summarized with the Plant Manager and
members of his staff on November 6, 1985. A violation with examples
described in paragraphs 7, 8 and 10 was discussed. The licensee
acknowledged the inspection findings and identified as proprietary a portion
of- the material reviewed by the inspectors- in regard to the negative rate
trip application as discussed in paragraph 10. The information in this
report 'does not include that proprietary information. During the reporting
-period, frequent discussions were held with the Site Director, Plant Manager
and his assistants concerning inspection findings. At no time during the
inspection was written material provided to the licensee by the inspector.
3. Licensee A>: tion on Previous Inspection Findings (92702)
.This subject was not addressed in this inspection.
4. Unresolved Items
No unresolved-items were' identified during this inspection.
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'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
-Specification (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.
The inspectors walked down accessible portions of the following
safety-related systems on Unit I and Unit 2 to verify operability and
proper valve alignment:
Residual Heat Removal System (Units 1 and 2)
Charging Pump Flowpath (Units 1_ and 2)
Control Room Ventilation Chlorine Detection System (Common)
Spent Fuel Pool Cooling System (Common)
. 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 badging.
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On November 1, 1985, the licensee declared a moderate security
. degradation as a result of the actions of a security officer posted at
-the entrance of -the Unit 2 containment hatch on the 690 level.
Appropriate ' compensatory actions were taken and the licensee's
personnel administrative process was implemented. The inspector
reviewed the above incident and had no further questions. .This item
will be reviewed by NRC specialist inspectors at a later date. No
violations or deviations were identified,
c. Radiation Protection
The inspectors observed Health Physics (HP) practices and verified
implementation of radiation protection control. On a regular basia,
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
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: instruments were verified operable and calibration frequencies were
. reviewed.
6. Engineered Safety Features Walkdown (71710)
EThe , inspector verified. operability of the Component Cooling Water system
(CCS) on' Units 1'and 2: by continuing a walkdown of the accessible portions
of a the systems.~ -Inspection Report 327,328/85-32 documents the previous
inspection.of this.' system. The following specifics were : reviewed and/or .
' observed'as_ appropriate:
a. _that the licensee's system lineup procedures. matched plant drawings and
the as-built configuration;
p b. :that equipment _ conditions were sati sfactory and items that might
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' degrade ' performance were identified and evaluated (e.g. hangers and
supports were operable, housekeeping etc, was adequate);
c. ~with assistance 'from licensee personnel, the interior of-the breakers
and electrical or instrumentation cabinets were inspected for debris,
loose material, jumpers, evidence of rodents, etc;
;d. .that instrumentation was properly valved in and functioning and
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calibration ~date's were appropriate;
e. that -valves were in proper position, breaker alignment was correct,
power was available, and valves were locked as required; and
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f. local and remote instrumentation was compared, and remote instru-
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mentation was functional.
- No violations or deviations were identified.
' 7' . -Monthly Surveillance Observations (61726)
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
Conditions 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
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 inspectors witnessed / reviewed portions of the following surveillance
test activities:
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SI-82.2 Functional Tests for the Radiation Monitoring System
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.SI-67 . Periodic Calibration of the RPI System
.The inspectors reviewed the results of reactor trip response time testing.
The following procedures were reviewed:
:IMI-99 ' Reactor . Protection System RT 11.6, Response Time Test of
dT/Tavg Channel II, Rack 6
IMI-99 Reactor ' Protection System RT 11.8, Response Time Test of
dT/Tavg Channel 4, Rack 13
IMI-99 Reactor Protection System RT 7.14 Response Time Test of Loop
1 Steam Generator Level Channel III (L-518) (L-3-39)
IMI-99 Reactor Protection System RT 7.17 Response Time Test of Loop
2 Steam Generator Level Channel III (L-528)
IMI-99 Reactor Protection System RT 7.20 Response Time Test of Loop
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3 Steam Generator Level Channel III (L-538)
IMI-99 Reactor Protection System RT 7.23, Response Time Test of Loop
4 Steam Generator Level Channel III (L-548)
IMI-99 Reactor Protection System RT 611A, Response Time Testing
Engineered Safety Feature Actuation Slave Relay K611
The inspector observed a. portion of the performance of the response time
testing for loop 1 steam generator level Channel III under procedure RT
7.14, The technician stopped the test when he could not complete step 4.4
which required that he insure that a test indicator light on the train he
was testing was -lit. The test indicator light was not lit. The technician
took the procedure-to his foreman for guidance. The foreman discussed the
step with an instrument maintenance engineer and determined that the light
would not illuminate because the reactor trip breakers were not closed. A
nonintent change was requested to revise the procedure.
During these discussions, the inspector observed that a piece of scratch
paper with a note written on it had been inserted into RT 7.14 indicating
that Step 55, which had not been performed at that point, could not be
. performed because of plant conditions. Step 55 requires verification of
certain block switches by confirmation that the block switch lights were
-lit. The technician stated that he had been instructed to place the remark
N/A (not applicable) adjacent to this step, and to continue with the test;
however, the technician stopped the procedure performance prior to reaching
this step.
The inspector reviewed additional procedures and determined that certain
steps had been marked N/A. In RT 611A, Step 5.5.6 requires that certain
equipment be returned to normal position. This is an independent verifi-
. cation signoff. Twenty-six of these steps were marked N/A with a note that
the components were tagged under various hold orders not specified in. the
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data-sheets. It should be noted that hold orders require independent veri-
fication of return to service. One additional non-safety-related component
was marked N/A with no reference to a hold order or other explanation. The
procedure requires that - if a device cannot be returned to normal, the
information should be entered as discrepancies on the data cover sheet.
This_information had not been entered in the data sheet as a discrepancy.
Additional . steps in procedures RT 7.17 and 7.23 require verification that
the status and alarm lights are not lit except as allowed by Step 2 of the
procedure which states that lights marked by an asterisk may be normally lit
if the unit is offline. Lights had been verified to be in a status not
allowed by the procedure and signed off as acceptable due to plant
conditions. The licensee stated that the status and alarm light
verification did not affect test performance in Modes 5 and 6 but was to
assure that if the test were performed in modes 1 through 4, the reactor
would not be tripped.
The failure to follow procedures RT 611A, RT 7.17, and RT 7.23 constitute a
violation 327,328/85-35-01.
In addition, the inspector noted that the following steps in procedure
RT 611A had been marked N/A in the data sheet when it appears that they
.should have not been marked that way.
Step 4.1.10 requires that an annunciator window for the low pressure
indication from the Condensate Storage Tank to the Auxiliary Feedwater
Pump (AFWP) be cleared. The pressure switch would automatically
initiate Essential Raw Cooling Water (ERCW) flow to the AFWP if the
pressure reached the low setpoint. The step was marked N/A with a note
that H0 1073 had power off of all ERCW valves. In this case, the reason
for the annuniciator window indication was clearly indicated in the
data sheet.
Step 5.2.1 was marked N/A. This step had a double entry for signing
off one handswitch position in the data sheet. This entry was clearly
a typographical error and should be corrected.
8. Monthly Maintenance Observations (62703)
a. 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 were met
while components or systems were removed from service; redundant
components were operable; approvals were obtained prior to initiating
the work; activities were accomplished using approved procedures and
were inspected as applicable; procedures used were adequate to control
the activity; troubleshooting activities were controlled and the repair
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record accurately reflected what actually took place; functional
testing and/or calibrations were performed prior to returning
components ~ or systems to service; quality control records were
maintained; activities were accomplished by qualified personnel; parts
and materials used were properly certified; radiological controls were
implemented; QC hold points were established .where required and were
observed;-fire prevention controls were implemented; outside contractor
' force activities were controlled ir accordance with the approved
Quality Assurance (QA) program; and housekeeping was actively pursued.
b. The inspectors reviewed the modification of feedring J-tubes in the
four steam generators. The licensee had planned a modification
involving replacement of the carbon steel J-tubes with Inconel J-tubes
due to wall thinning in the J-tubes. Upon examination of the tubes and
feedring after removal of the J-tubes, the licensee determined that the
carbon steel feedring had been eroded by high velocity flow at the base
of the J-tube.
The modification was revised to include oversized boring of the holes
in the feedring to eliminate the eroded areas and buildup of the J-tube
wall with Inconel in this area to fit the larger hole. The inside
diameter of the J-tube remained the same except that the entrance to
the tube from the feedring was machined to a smooth rounded edge to
prevent turbulance. The J-tube was welded to the feedring with Inconel
weld filler metal.
The inspector examined J-tubes removed from the steam generators and
examined a portion of one feedring upon removal of the J-tubes. The
inspector also observed an inspection of J-tube welds by the vendor's
QA representative. The inspector reviewed WP 11829 which covered all
the J-tube replacements in the four Unit 1 steam generators and
drawings D-246-941-1 and -2. The licensee will be presenting the
findings on- the feedring degradation to the Westinghouse Steam
Generator Owners Group during November,1985.
No violation or deviations were identified.
c. The inspector observed preparation for a leak test of a containment
penetration which had been replaced to meet environmental qualificaton
requirements. The inspector reviewed Work Plan 11802, dated
October 28, 1985, which required that the penetration be assembled and
tested in accordance with a validated vendor manual. The inspector
determined that the licensee had not received the updated vendor manual
describing the assembly of the feed thru tubes for the penetration. The
vendor manual at the work site did not describe the assembly of the
feedthru tubes, but did describe the leak test requirements. The
inspectors determined that the manual used had not been reviewed and
validated by PORC. The penetration was installed and assembled based
on verbal instructions received from the vendor at an earlier date.
The vendor, subsequent to this inspection providea written instructions
to the licensee which included requirements for QC hold points not
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performed during -installation. Based on ' this new information, the '
licensee reworked the penetration in. accordance with the vendor's
instructions and a validated vendor manual. Failure to' implement the
work 1 plan for. assembly of containment electrical penetrations is a
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further example of violation 327, 328/85-35-01.
d. The replacement of electrical relays in.the 6.9KV shutdown boards was
observed. The following documents were reviewed:
Special Maintenance Instruction SMI-0-202-1
Maintenance Instruction'6.20
- Maintenance Request A284454
Procurement Documents (575) 5886000390, 5886000771
The maintenance appeared to be adequate and no violations or deviations
were identified.
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;9. Licensee Event Report (LER) Followup (92700)
'The following LER's 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,
a. LER Unit 1
327/85021 Control Room Ventilation Isolation
327/85023 Auxiliary Building Isolation
327/85026 Failure to Obtain a. Noble Gas Sample
327/85027 Main Steam Line I:olation
327/85029 Reactor Trip on Loss of Power to Main Feedwater Pump
327/85030 Auxiliary Feedwater Initiation
327/85033 Main Control Room Isolation Due to Failure to Follow
Procedure
327/85034 Diesel Generator Operability
327/85035 Emergency Diesel Generator Start While Trouble Shooting
Control Power
327/85037 Main Control Room Isolation Due to Spike on Radiation
Monitor
327/85038 Auxiliary Building Isolation From SFP Rad Monitor During
Filter Changeout
b. LER Unit 2
328/85007 Inadvertent trip 2A-A Shutdown Board Feeder Breaker
328/85008 Failuresto Complete Hourly Fire Watch
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'10. Event Followup-(93702, 62703, 61726)
a. On October 1,1985,- the inspectors received a copy of a Westinghouse
Technical . Bulletin which dealt with- negative and positive flux rate
reactor trip setpoint calibration methodology. Discussions were held
with'several levels of plant management including cognizant engineers
and technicians. As a result of this process the following documents
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.were reviewed:
Westinghouse Technical Bulletin NSID-TB-85-13
Westinghouse Technical Manual N2M-2-1-X, Nuclear Instrument-System
Surveillance Instruction (SI) E0, Power Range Nuclear Flux Channel
Calibration and Functional Test
Standard Practice SQA26 Attachment 4, Operating Experience Review
Recommended Action Sheet
Instrument Maintenance Instruction (IMI) 92-PRM-CAL, NIS Power Range
Standard Practice SQA26 Attachment 3, Experience Review Evaluation
Form
Standard Practice SQA26 Attachment 2, from Supervisor, Regulatory
Engineering to Supervisor, Instrument Maintenance and Lead
Instrument Engineer (D. Elkins, R. Gladney)
Standard Practice SQA26 Attachment 1, Operating Experience Review
Screening Sheet
TVA memo McGriff to Brimer, Sullivan of September 3,1985
TVA memo Gibbs to Wilson copy to Sauer of July 17, 1985 (Note: This
is a Watts Bar site memo)
Technical Specification Change Request 85-122
Management A: tion Tracking System (MATS) Assignment Sheet dated June
26, 1985
TVA memo McCloud to McGriff dated June-25, 1985
Precautions, Limitations and Setpoints for Sequoyah Nuclear Plant
NRR memo, T. Dunning to Dunenfeld, Westinghouse Neutron Flux Rate
Setpoints
Sequoyah Nuclear Plant Startup Test 9.5 Evaluation Report
WCAP-10297-P-A Westinghouse Dropped Rod Methodology for Negative
Flux Rate Trip Plants
Technical Specification Table 2.2-1
The SNP management and staff were knowledgable of the WTB about the
second week in July. The WTB discussed the alignment procedure for the
Nuclear Instrumentation system power range positive and negative rate
trip bistables and explained that some plants had misinterpretated the
procedure as outlined in the Westinghouse Nuclear Instrumentation
System Technical manual. The electrical circuit addressed by both the
Nuclear Instrumentation Technical manual and the WTB consists of an
upper and lower detector whose signals are indicated on nuclear
instrument (NI) meters 301 and 302. The signals are added together
and averaged through a level and averaging circuit. A resulting
adjusted signal is then read on full percent power meter 303.
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The adjusted electrical signal passes through two subsections of the
power range rate and delay circuit (NM311), resulting in a potential
difference on a downstream operational amplifier. The output of the
: operational amplifier is fed to the input of a bistable which will trip
when a given input value-is reached, resulting in a reactor trip.
The- Westinghouse Nuclear Instrumentation (WNI) Technical Manual
described the process used to calibrate this power range rate and delay
' circuit to ensure that bistable NC301 has the proper TS reactor trip
setpoint. Surveillance Instruction SI 80 was reviewed and appeared to
conform with what was indicated in the WNI Technical Manual.
Performance of the steps described in the WNI Technical Manual and SI
80 resulted in a stepped potential difference of 3% (negative rate
trip) and 5*4 (positive rate trip) being applied to the operational
amplifier in the rate and delay circuit.
The Westinghouse Technical Bulletin stated that the power range
detector A test signal is used to create a step signal which is the
input to the power range rate and delay circuit (NM311) and that the
detector A test signal should be set numerically equivalent to the
value of percent full power change given~ in the plant Precautions,
Limitations and Setpoints(PLS) document. For Sequoyah, the PLS
document disagrees with the Technical Specifications, and the licensee
used the Technical Specification values. The WTB also stated that due
to possible misinterpretation of the Nuclear Instrumentation System
manual, plants may have doubled the Detector A test signal in order to
compensate for the summing the level amplifier.
Additionally, the WTB ~ requires maintenance personnel to set the
detector A test signal in power units or percent of full power detector
current, to the value given in the PLS document for tne percent full
power change for the rate trip. For example, if the PLS document
requires a rate trip on a 5% change of full power, then the detector A
test signal should be set to 5 power units or 5% of detector A full
power current.
The difference between the Westinghouse Technical Bulletin (WTB) and
the Westinghouse Nuclear Instrument Technical Manual is in the
amplitude of the potential applied. The WTB requires that the
amplitude be read on the meter after the leveling circuit.
The initial TVA management review determined that the bulletin could
not be complied with because the operational amplifier input would have
to be set to 1.5*. and that this value would not allow sufficient margin
from' normally present nuclear flux circuit noise (approximately l*s).
The licensee interpreted that the trip value of the TS should be equal
to the magnitude of the detector input since this is consistent with
standard TS trip setpoint methodology.
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A TS change request was processed through the Plant Operations Review
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Committee (PORC) on. September 11, 1985 to implement the standard TS
trip setpoint valves. It stated that implementing the calibration
method stated in the WTB would significantly increase the chances of
inadvertant trip actuations caused by nuclear noise using the current
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TS valves.
-The licensee requested Westinghouse to perform a study and determine
whether the WTB applied to Sequoyah. The TS change request was-
submitted to the NRC by letter dated October 22, 1985. Westinghouse's
response addressed the conservatism of the current Sequoyah TS compared
to the PLS values and performed some calculations on the power range
rate and- delay circuit. Although Westinghouse calculations were
provided for several cases, the results appeared to be only
conditionally acceptable. Conversations were held between NRC Region
II and the licensee and NRR personnel. The licensee's interpretation
on setpoint methodology for testing was consistent with TS intent. In
light of the WTB, the NRC determined that the TS were in error and that
this issue appeared to be generic. Resolution of this TS issue prior
to the startup is an Inspector Followup Item 327, 328/85-35-02.
On October 30, 1985, The inspector witnessed a surveillance test which
verified the Digital Rate Circuit time constant on Power Range Monitor
channels N-41 and N-43. The work was requested under MR A-539515 and
A-539516 for channels N-41 and N-43, respectively. The technicians
utilized Instrument Maintenance Instruction, IMI-92-PRM-CAL steps
5.2.9.12 through 5.2.9.14 to perform the test and IMI-134 to record the
data. The test was conducted by inputing a negative three percent
change in power level and upon reaching the desired level determining
the decay time to reach 37% of the initial value, f.e., one time
constant. The test determined that the time constant for N-41 was 1.31
seconds for for N-43 was 1.30 seconds. The time constant is required
per TS table 2.2-1 to be greater than 1 second.
No violations or deviations were identified.
b. On October 26, 1985, the licensee discovered a leak in the reactor
cavity liner. The cavity was drained and a nozzle cover was repaired
and reseated and the cavity was refilled. On October 28, 1985, the
licensee discovered that the cavity liner was again leaking. The
leakage was going to the keyway sump-under the vessel and through the
number 2 cold leg penetration to the containment sump. The licensee
evaluated the leakage, which remained steady at approximately I gpm
through the end of the report period, and determined that the liner
itself was probably the source of the leakage. The inspector reviewed
the procedure for failure of the reactor cavity seal, Abnormal
Operating Instruction, A0I-290 and discussed the leak rate with the
cognizant engineer. The licensee stated that based on the present
indications that refueling operations would proceed with close
monitoring of the leakage. At the end of refueling the licensee will
drain the reactor cavity and repair the leak.
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No violations or deviations were identified.
c. On October 10, 1985, the licensee conducted Surveillance Instruction,
SI-82.2 as part of a post modification test to restore radiation
monitors 2-RM-90-106B and -112B to service. Work plan 11793 had been
written to incorperate changes descriosd in the Engineering Change
Notice 5198 and Field Change Request 3785. The maintenance consisted
of a modification to an electrical ground point. The Instrument
Technician placed switch HS-90-136A in the block position on Unit 2 and
then inserted a test signal into the Unit I circuit in error. This
action resulted in a containment ventilation isolation. Failure to
adequately implement SI-82.2 is a further example of violation
327,328/85-35-01,
d. On October 31, 1985, while transferring start bus.1B from normal to
alternate supply, the alternate breaker faiied to latch. This resulted
in a loss of power to the 1A Shutdown Board and a start signal to the
diesel generators. Two of the diesel generators started; the other two
diesel generators were out of service for maintenance. The licensee
attributed the failure of the alternate breaker to mechanical binding
at the end of travel resulting in the failure to latch. The breaker
was subsequently relatched; however, the licensee stated that main-
tenance would be performed on the breaker to investigate the problem.
In conjunction with this failure, a "B" Train Auxiliary Building
Isolation occurred due to loss of power to spent fuel pool monitor
0-RM-90-103. This monitor is required to operate to prevent a release
of radioactive material from the Auxiliary Building in the event of a
fuel handling accident in the spent fuel pool. As identified on
drawing PL J281-53 the monitor should have been powered from a Train B
power source inside the radiation monitor cabinet. The licensee
determined that the monitor was plugged into a nonessential power
source.
The inspector reviewed MR A-530620, which the licensee identified as
the latest maintenance involving unplugging of the power source. The
MR required maintenance to be done in accordance with Instrument
Maintenance Instruction IMI-134, Configuration Centrol of Instrument
Maintenance Activities. This procedure required the use of a
configuration control sheet to assure that equipment was returned to
its proper orientation. The requirements for use of the configuration
control sheet were not properly followed in that the sheet did not
identify the specific plug mold from which the monitor was unplugged
and returned. Failure to implement configuration control procedures is
a.further example of violation 327, 328/85-35-01,
11. Inspector Followup Items (92701)
Based on inspection activities in the affected functional areas the following items were determined to require no additional specific followup
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and are closed. Discussions were held with the licensee with regard to the
tinieliness of corrective actions.
83-23-04 (units 1 and 2)
84-11-03 (unit 1)
12. Review of Part 21 Reports (36100)
a. The inspector reviewed a 10 CFR Part 21 report, provided to the NRC in
a letter dated March 13, 1984, on Brown Bovari Corporation Type ITE-27N
undervoltage sensing relays. Correction . of the design deficiency
required replacement of a 100 kilchm resistor with a 200 kilohm
resister on fourteen relays provided to Sequoyah. The inspector
reviewed MRs A-082428, A-082427, A-082426 and A-082424 which replaced
12 of the resistors on the subject relays which are utilized for
undervoltage protection on the 6.9 KV shutdown boards. The inspector
randomly selected six of the relays and verified replacement of the
resistors. Two additional relays maintained at replacement parts were
also verified to be modified. This item, identified as 327,
328/P21-85-03 is closed.
b. The inspector reviewed a 10 CFR Part 21 report, provided to the NRC on
June 15, 1984, on the use of Crawford Fitting Company Swagelock
fittings. Crawford Fitting Company determined that this issue was not
of safety concern as documented in their November 16, 1984 letter to
the NRC. This item, identified as 327,328/P21-85-02, is closed. Note
that vendor recommendations on the use of Swagelock fittings was
reviewed in Inspection Report 327/85-27, 328/85-28 and an Inspector
Followup Item was left open regarding the licensee's evaluation of high
pressure seal fitting adequacy.
13. Refueling Activities (60710)
Unit 1 began removing fuel from the reactor for the Cycle 4 fuel load on
October 23, 1985. Reload of the core was in progress at the end of this
inspection report period. The inspector observed preparations for
refueling, fuel handling operations in containment and in the spent fuel
pool, movement of thimble plugs and rod cluster control assemblies in the
spent fuel pit, and other ongoing activities associated with the rifueling.
The inspector verified that_ selected Technical Specification requirements
were met, that appropriate procedures were being utilized, that containment
integrity was being maintained, that housekeeping and control of materials
entering containment was adequate and that staffing was in accordance with
the Technical Specification requirements. The following documer,:s were
reviewed:
Fuel Handling Instruction FHI-5, RCC Change Fixture
Fuel Handling Instruction FH'.-6, Preparation for Refueling
Fuel Handling Instruction FHI-7, Refueling Operation
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Fuel Handling Instruction FHI-13, Burnable Poison Rod Assembly Handling
Tool
Fuel Handling Instruction FHI-14, Thimble Plug Handling Tool
Fuel. Handling Instruction FHI-17, Rod Cluster Control Change Tool
Administrative Instruction AI-26, Prevention of Foreign Material in the
Primary System -
Restart Test Instruction (RTI)-2, Core Loading
Technical Instruction (TI)-1, SNM Control and Accountability System
No violations or deviations were identified.
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14. Inspection Plan for Followup of Sequoyah Nonconformance Report
A staff review was conducted, by a team of NRR technical reviewers and
Region II personnel,.of the management processes involved in the resolution
of Nonconformance Report (NCR) SQNNEB 8501 and its associated Failure
Evaluation Engineering -Report (FEER). Attendant to this staff review,
selected NCRs and FEERs were collected for additional evaluation. As a
result of this additional review several cases were identified where
potential safety questions were raised. Safety Evaluations were made by the
staff for each safety question and required inspection effort was identified
in a staff memo (Verrelli et al to Denton) dated August 9, 1985.
An ' inspection plan for followup of the Sequoyah NCR open concerns was
established by Region II in a staff memo (Weise to Walker) dated
September 23, 1985, that identified several items which required resident
inspector followup. The status of those items which required resident
inspector followup is indicated below:
a. NCR SQN CEB 8406 involved two air clean up units that were not welded
to their steel supports in accordance with TVA. drawing 48N726. The
welds were later upgraded to the requirements of drawing 48N726 under
Maintenance Request A236959. The welding discrepancy was an undersized
weld which was later determined to have been a temporary fit-up weld
that should have been replaced with a permanent weld after
installation. The licensee inspected all applicable welds in the
mechanical equipment room and identified no other welds which were
undersized. These particular welds, because of their temporary nature,
did not have strike numbers or other means with which to identify the
crew that performed the welds. The licensee's corrective action
appeared to be adequate in this instance, and this item is closed,
b. NCR SQN EEB 8406 involved some Class 1E 480 volt switchgear breakers
and motor control center molded case circuit breakers which could be
subjected to fault currents beyond their design capability. A FEER was
issued by the licensee identifying this condition as a Category III. A
Category III indicates that a component is unable to perform its
required design function unless corrective modifications are made.
Subsequently a safety evaluation was performed and found that the
condition did not impact the safety of the plant and that no
operational limitations were required. As a result of staff review it
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was determined that certain aspects of the FEER were deficient and the
licensee committed to revise the NCR. The inspector obtained a copy of
-the revised NCR 'and transmitted it to the appropriate Region II
personnel. -In addition, it appeared that the original NCR was written
before a calculated load study was completed and there was. no
i statistical validity for the assumptions made in the FEER. As a result
of the~ revised NCR, this item was reduced. in condition to a Category I,
acceptable for all modes of operation and design conditions. For the
. purpose of this. inspection, this item is considered closed.
c. NCR SQN NEB 8407 involved eight Class IE radiation monitors which had
been miswired or had their identification tags interchanged. This item
was the subject of ' Region II enforcement action (327,327/84-38). The
. licensee's response to this enforcement action was reviewed by the
inspector. A team inspection is planned to address the NRC order EA
85-49 which will include a review of the licensee's NCR corrective
' actions. After the team inspection is complete the inspector. will
review the licensee's corrective actions for the previous violation.
For the purposes of this review plan, this item is closed.
d. NCR SQN NEB 8408 involved a relative humidity control component which
could fail as a result of high radiation during a reactor accident.
The licensee's resolution to this issue was to allow the relative
humidity heater to energize when the fan starts and reaches full speed.
The relative humidity control component would be used for alarm
purposes only. A ' review - of the adequacy of .TS surveillance was
conducted by reviewing Surveillance Instructions SI-141 and -142 and
. Technical Instruction TI-9. The surveillances conducted on the
Emergency Gas Treatment System appear to be adequate.. This issue is
closed.
e. NCR SQN EEB 8412 involved Bettis Actuators with potential deficiencies.
This issue was resolved in Inspection Report 327,3P8/85-26.
f. NCR SQN NEB 8413 involved a discrepancy between the as found spent fuel
pool alignment and that alignment described in the FSAR. A review of
the reportablity aspects of this issue was conducted, and the issue was
determined not to be reportable. An update was made on the most recent
FSAR amendment submittal by the licensee to reflect current spent fuel
. pool alignment. A review of the established makeup sources and
applicable procedures, System Operating Instructions 501-70.1 and -78.1
and Abnormal Operating Instruction A0I-15, was conducted. The
procedures and system alignments appear to be adequate and in
compliance with TS. This issue is closed.
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