ML20149E452
| ML20149E452 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 12/22/1987 |
| From: | Jenison K, Mccoy F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II), NRC OFFICE OF SPECIAL PROJECTS |
| To: | |
| Shared Package | |
| ML20149E379 | List: |
| References | |
| 50-327-87-65, 50-328-87-65, IEB-83-01, IEB-83-1, NUDOCS 8802110055 | |
| Download: ML20149E452 (34) | |
See also: IR 05000327/1987065
Text
UNITED STATES
[v.0tg% NUCLEAR REGULATORY COMMISSION
y" REGION 11
g j 101 MARIETTA STREET,N.W.
O r ATLANTA, GEORGt A 30323
'+9 . . . . . ,o
Report Nos.: 50-327/87-65, 50-328/87-65
Licensee: Tennessee Valley Authority
500A Chestnut Street
Chattanooga, TN 37401
Docket Nos.: 50-327 and 50-328 License hos.: DPR-77 and DPR-79
Facility Name: Sequoyah Units 1 and 2
Inspection Conducted: October 6, 1987 thru Novembir 5, 1987
Lead Inspector: N]/8. M /of/2//87
K.M.Jenison,SeniorReg#intyspect)r Date Si'gned
Accompanying Inspectors: P. E. Harmon, Resident Inspector
D. P. Loveless, Resident Inspector
W. K. Poertner, Resident Inspector
W. C. Bearden, Resident Inspector
M. W. Branc , Sequ h Restart Coordinator
Approved by:
-
M
F. R. McCoy, Chief, Projects Section 1
/ /M.:2[F 7
Date SVgned'
DivisionofTVAProjects
SUMMARY
Scope: This routine, announced inspection involved inspection onsite by the
Resident Inspectors in the areas of operational safety verification including
operations performance, system lineups, radiation protection, safeguards and
housekeeping inspections; maintenance observations; review of previous
inspection findings; followua of events; review of licensee identified items;
review of IE Information Notices; and review of inspector followup items.
Results: Three violations were identified.
(327,328/87-65-01), Inadequate Corrective Actions paragraph 12
(327,328/87-65-02), Inadequate Response Time Test - paragraph 6
(327,328/87-65-03), Failure to Adequately Control Changes to
Control Room Orawings paragraph 3
One unresolved item was identified.
(327,328/87-65-04), Surveillance Discrepancies - paragraph 13
$0 kOO!k O!O 7
J
_ __. - - __
, _ _ _ _ ___ _ _ _ _ _ - ____ _ _ _ _ _ - _ _ _ _ _ _ _ _ _
REPORT DETAILS
-
1. Licensee Employees Contacted
H. L. Abercrombie, Site Director
~
J. T. La Point, Deputy Site Director
- L. M. Nobles, Plant Manager
- B. M. Willis, Operations and Engineering Superintendent
- B. M. Patterson, Maintenance Superintendent
R. J. Prince, Radiological Control Superintendent
- M. R. Harding, Licensing Group Manager
L. E. Martin, Site Quality Manager
D. W. Wilson, Project Engineer
R. W. Olson, Modifications Branch Manager
J. M. Anthony, Operations Group Supervisor
- R. V. Pierce, Mechanical Maintenance Supervisor
M. A. Scarzinski, Electrical Maintenance Supervisor
- H. D. Elkins, Instrument Maintenance Group Manager
R. S. Kaplan, Site Security Manager
J. T. Crittenden, Public Safety Service Chief
- R. W. Fortenberry, Technical Support Supervisor
- G. B. Kirk, Compliance Supervisor
D. C. Craven, Quality Assurance Staff Supervisor
J. H. Sullivan, Regulatory Engineering Supervisor
J. L. Hamilton, Quality Engineering Manager
D. L. Cowart, Quality Engineering Supervisor
H. R. Rogers, Plant Operations Review Staff
R. H. Buchholz, Sequoyah Site Representative
M. A. Cooper, Compliance Licensing Engineer
Other licensee employees contacted included technicians, operators, shif t
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 staf f on November 5,1986. Three violations described
in this report's summary paragraph were discussed. No deviations were
discussed. The licensee acknowledged the inspection findings. The
licensee did not identify is 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.
During the exit interview plant management committed to revising A01-27,
- Control Room Inaccessibility, per paragraph 3 below, in order to meet the
!
guidance of regulatory guides 1.68.2 and 1.68.
. a
r -
'
2
3. Licensee Action on Previous Inspection Findings (92702)
(Closed) Unresolved Item (URI) 327, 328/87-24-02, Control of Temporary
Changes to Drawings. This URI has been determined to constitute a
violation. Drawings in the control room are marked by the modifications
engineers immediately following the completion of physical changes to the
~
plant's as-built condition. These temporary drawing changes ~ consist of
red marks for additions to the drawings, and green marks to designate
removals. During the inspection described in IR 327, 328/87-24, several
instances were identified where errors were introduced to the control room
drawings by the modifications engineers when they marked up the drawings.
10 CFR 50, Appendix B, Criterion VI requires that changes to documents
(including drawings) be reviewed for adequacy and approved for release by
authorized persons. This requirement was not met in that changes to
primary control room drawings were routinely made with no verification or
review by second parties. _This is a violation VIO 327,328/87-65-03.
(Closed) URI 327, 328/87-08-02, Abnormal Operating Instruction (A01)
Personnel Required for Remote Shutdown. The inspector reviewed A01-27,
Control Room Inaccessibility. The procedure describes actions to be taken
should the control room become uninhabitable. The procedure requires the
dispatching of more personnel than are required as a minimum on-shift in
TS 6.2.2.a. Appendix A of 10 CFR 50 requires in GDC 19 that equipment at
appropriate locations outside the control room be provided with a design
capability for prompt hot shutdown of the reactor, including necessary
instrumentation and controls to maintain the unit in a safe condition
during hot shutdown. This criteria is addressed in regulatory guide (RG)
1.68.2, Initial Startup, Nuclear Power Plants. This RG states that
startup testing should demonstrate that the number of personnel available
to conduct the shutdown operation is sufficient to perform the many
actions required by the procedure in a timely, coordinated manner.
Startup test documentation, as described in item SU-1.2A of Table 14.1-3
in the final safety analysis report (FSAR), was reviewed. In addition,
the following issues, identified in inspection report 327, 328/87-08, were
also reviewed:
-
Whether there is sufficient personnel and guidance to perform a safe
and orderly shutdown from outside the control room with a minimum
shift crew per the TS.
-
Whether procedures are adequate to address the limiting case of
minimum shift manning.
-
Whether the initial startup test was performed utilizing the
personnel indicated in the procedure or the TS minimum, and whether
the procedure was adequate.
The inspector walked down the procedure utilizing licensed personnel and
determined that the plant could be shutdown using minimum shift crew.
However, performance with a minimum crew would require some step sequence
-
3
modifications ; to A01-27. A draf t update to A01-27 was prepared and
reviewed making it possible to run with minimum shift crew. During the
exit interview, plant management committed to revising A01-27 in order to
meet the guidance of regulatory guides 1.68.1 and 1.68.
The inspector cid determine that the initial startup test utilized normal
shift manning as opposed to the minimum shift crew in TS. This action to
use the normal shif t manning was according to the TVA commitment set in
the FSAR. This item is closed.
(0 pen) Violation (VIO) 327, 328/86-37-06, Containment Sump Leve' Test
Deficiencies. The inspector reviewed this issue which involved a
prolonged history of test deficiencies with respect to the calibration of
the containment sump level transmitters. These particular transmitters
were identified as being out of TS tolerance six times during the
surveillances conducted between 1984 and 1986. The NRC identified that
the licensee did not initiate a quality assurance document identified as a
corrective action report (CAR). Failure to initiate a CAR was identified
as a violation and the cover letter for inspection report 327, 328/86-37-
stated that, "violation 327, 328/86-37-06 was also identified and is
described in paragraph 13 of the enclosed inspection report. This
additional violation is under consideration for escalated enforcement
action. Accordingly, a notice of violation addressing this particular
violation is not being issued at this time, and therefore no response to
this violation is required." The cover letter for the inspection report
further stated that "the number and characterization of violations
described in paragraph 13 of the enclosed inspection report may change as
a result of further NRC review".
The licensee's corrective actions with respect to the containment sump
level transmitters included issuance of a condition adverse to quality
report (CAQR) CAQR-SQP870043 and the establishment of a plant tracking and
trending program under standard practice SQM-58, maintenance history and
trending. The general adequacy of the QA .CAQR and the maintenance
tracking and trending processes are subjects of separate NRC inspections.
The specific issues dealing with the identification of the containment
sump level transmitters as a condition adverse to quality and tracking and
trending of those specific material deficiencies appear to have been
adequately addressed by the licensee. This item will remain open, pending
escalated enforcement action.
(Closed) VIO 327, 328/86-39-01, Failure to Report Computer Program Errors.
This violation involved the use of incorrect axial flux curve and rod bow
penalty data in monthly surveillances of incore reactor parameters. In
response to the violation, the licensee wro;e a licensee event report
(LER) 322 '86-004 which was closed in NRC inspection report 327, 328/87-08.
The LER appears to address adequate correctivt action to violation 327,
328/86-39-01. The inspector examined the root cause of this issue in
detail and evaluated the corrective action witt respect to the security
and accuracy of data being placed into saist3 related plant computer
systems in general and the INCORE program in particular. In LER 328/86-004 the licensee committed that before installing vendor-supplied
data into sof tware in the future, verification will be provided that the
_ _ _ _ _ _ _ _ - . _ _
_
4
- correct data is being used. This is accomplished through an independent-
verification process, This_ item'is closed.
(Closed)URI 327,.328/87-30-03, Reactor Coolant-System Sight Glass Design.
The design .of the sight _ glass .used to indicate . reactor ' coolant system
(RCS) inventory level during partial- loop drain down ' conditions was-
questioned by the NRC inspection staff as well as the licensee'.s nuclear
manager's review group (NMRG) following the RCS spill event of January 28,
1987. Several design deficiencies were . identified and later determined
by the NRC . not to be safety related. In particular, the monitoring
arrangement using a TV camera and monitor was not well designed to allow
the control room operator to readily determine the level in the sight
glass. Resolution of the above deficiency and .several other minor defi-
.ciencies is scheduled during the next refueling outage for each unit. The
corrective actions are presently being tracked under the licensee's
tracking and reporting of open items system '(TROI). The actual site
glass modifications will be reviewed when the appropriate reviews and
corrections have been accomplished during the unit 2 cycle 3 outage. This
item is closed.
(Closed) URI 327, 328/87-30-04, Change to TS Basis. Sequoyah issued
Technical Specification (TS) basis change 87-14 directly to 'the NRC.
without a nuclear safety review board (NSRB) review. The licensee's
position is that the TS basis is not an actual part of the TS as defined
by 10 CFR 50.36(a). Therefore, TVA's position is that changing the basis
does not require an NSRB review. NRC approval of the change to the TS
Basis was issued on August 18, 1987. .The inspector . discussed the
licensee's position with OSP Technical Programs management, and determined
that this generic issue is currently under NRC review. TVA indicated
that they would have the NSRB perform such reviews until such time this
generic issue is resolved. This item is closed.
-(0 pen) VIO 327, 328/86-73-04, Failure to Properly Evaluate the Generic
Applicability at Other Nuclear TVA Facilities of Conditions Adverse to
Quality (CAQ). The violation pertained to the adequacy of engineering
evaluations. Deficiencies identified included Bellefonte CAQR BLN4929,
dated June 30, 1986. Sequoyah did not properly analyze the cause of the
CAQR and did not have documentation for justification of the determination
that the CAQR did not apply to Sequoyah. TVA gave as reasons for the
violation "the failure on TVA's part to provide adequate and sufficient
problem descriptions and detailed information in the assessment of the
potential generic condition as recorded in the P&CE memorandum so that
other TVA facilities could properly assess the generic implications of the
cited conditions." Their corrective action includes a revision 3 to YVA's
- t nuclear quality assurance menual (NQAM) part I, Section 2.16, which
requires generic reviews to be completed af ter root caJse analysis and
'
recurrence control actions have been determined. This was implemented
i
July 1, 1987. TVA further revised procedures for more controlled and ;
centralized requirements for the conduct and timing of generic reviews.
The inspector reviewed CAQRs at Sequoyah that have been determined by TVA
-
5
to be generic. This was a random review of CAQRs initiated by Sequoyah
and by other plants. The foll_owing deficiencies were noted.
-
CAQR number WBP870420 was initiated on June 6, 1987, at Watts Bar and
concerned the storage of electrical cabies in the warehouse and craft
storage area at Watts Bar. The CAQR listed 8 items to address and
gave as the root cause of the CAQ, "inadequate procedures for cable
transactions and storage which lead to personnel being trained
incorrect." On August 1,1987, the generic review sheet was signed
with a determination that this CAQ did not exist at Sequoyah. The
justification was based on inspection of cable storage areas and
addressed the 8 items, but failed to address the root cause, i.e. ,
inadequate procedures and training.
-
CAQRs WBP870420 and BFQ870375 attachment 7 CAQ generic review sheets
were not properly reviewed. NQAM Part 1, Section 2.16, step 10.3
states," the justification for determining that a CAQ is not generic
shall be documented. The individual who prepares the justification
shall provide a dated signature, and the individual's supervisor
shall approve the justification by dated signature." On August 1,
1987, the same individual signed WBP 870420 as reviewer and as
supervisor of reviewer. On June 26, 1987, the same individual signed
BFQ 870396 as reviewer and as supervisor of reviewer. This is a
failure to follow procedures.
-
On July 9,1986, Bellefonte (BFN) identified a CAQ and initiated SCR
BLN4929. The root cause of this CAQ was an operator attempting to
close the diesel generator (D/G) output breaker out of phase.
Sequoyah misidentified the root cause. The recommendations made at
Bellefonte to prevent recurrence are: (1) additional simulator
training for operators, (2) review of the incident with operators,
and (3) addition of a permissive synchronize and voltage check scheme
on the diesel generator output breaker.
Sequoyah CAQR SQP870943, signed on July 2, 1987, states in the
description of proposed disposition, that "items 1 and 2-POTC
provides D/G simulator training during RO certification. Operators
are already aware that the D/G should be in phase before
synchronization. No additional training is proposed. Item 3, we do
not recommend adding an interlock to the D/G breaker. This accident
appears to have been due to human error and was an isolated
incident." The recommended actions at Sequoyah appear to be
inconsistent with those at Bellefonte. Further, NQAM Section 1,
Part 2.16 step 10.5.2 states "for CAQs determined to be generic,
affected organizations should communicate with each other in the
development of corrective action plans to ensure that, when
appropriate, such plans are consistent."
Further inspections of the generic review issue will be conducted prior to
the startup of either unit. This violation remains open.
6
(Closed) 327, 328/87-06 observation MEB-8, Inconsistent Equipment
Qualification Temperature. Calculation NEB 811007235 had been prepared to
analyze a deficiency in the pre-operational test results for the turbine
driven auxiliary feedwater pump (TDAFW) room ventilation system and to
provide suggestions to reduce the temperature rise in the room. The
calculation uses a temperature rise to 125 degrees F, which the
calculation states is the equipment qualification temperature. This is_
not consistent with plant data sheet, 47E235. TVA recalculated the room
heat loads in 844870609010, dated April 10, 1987, and the required air
flow rates in B44870716007, dated July 16, 1987. The conclusion is that
the installed exhaust' fans are adequate to meet the requirements of normal
operation and LOCA, with and without loss of offsite power, with a maximum
of 110 degrees F in the room. The corrective action in PIRSQNMEB8773 is
to change FSAR section 9.4.2.2.7 to delete the room temperature rise
criterion and to replace it with a maximum room temperature of 110 degrees
F. This corrective action appears to be an adequate response to NRC's
concern. This item is closed.
(Closed) 327, 328/87-06 observation GEN-1, Substantiated Condition for a -
CAQ. Nuciaar engineering procedure (NEP)-9.1, Corrective Actions, revised
July 1, 1986, defines the controlled system within ONE to document,
evaluate and resolve conditions adverse to quality (CAQs). NEP-9.1 was
being revised to agree with the corporate QA procedure, NQAM, Part I,
Section 2.6, which includes in the definition of CAQs the statement that
"unsubstantiated conditions are not defined as CAQs," The concern was
that this statement had the net effect of eliminating a set of CAQs that
had previously been identified and resolved. NEP-9.1, Revision 2,
Section 2.2, dated June 30, 1987, eliminates the questionable statement
from the definition. Section 1.lb of NEP-9.1, Revision 2, provides for a
problem identification report (PIR) system to document problems, and
potential problems that are not CAQs as defined. The licensee's action
resolves the concern. This item is closed.
(Closed) 327, 328/87-14 observation 6.19, 480 Volt Board Room Air Handling
Unit Control Logic. Cut-out switches had been installed in each of the
heating, ventilation and air conditioning (HVAC) systems for the 480V
board room and the 6.9KV shutdown board room to shut off the room's
cooling system on a high temperature signal in the room. The NRC concern
was that the close location of the temperature sensors was such that a
common mode failure could disable all cooling. CAQR871011 for the 480V
board room and CAQR871279 for the 6.9KV shutdown board room were issued to
disconnect the room high temperature cut-out switches since the switches
had been installed for economic reasons to protect equipment in the event
of refrigerant loss, and had no safety function. ECN7263 followed up on
this item and the switches have been disconnected. This item is closed.
(Closed) VIO 327, 328/87-02-02, Failure to Establish, Maintain and
Implement Safety-Related Procedures. This violation resulted from three
engineered safety feature (ESF) initiations reported in licensee event
reports (LERs) 328/86-08, 328/86-09 and 328/86-10, and from observations
of the performance of general operating instruction (G01)-6H, Freeze
7
Protection. The violation consisted of an inadequate work plan (WP),-
instrument maintenance instruction, G0I documents, and personnel errors
during procedure performance. The inspector reviewed TVA's corrective
actions detailed in the June 12,-1987 written response to this violation
and in the LERs. The specific procedural errors and weakt. esses have been
corrected. Site procedures such as surveillance instructions have been
through an extensive review process and are now developed and revised
using checklists and a writer's guide. Maintenance procedures are
currently undergoing a similar process. Personnel involved in the
specific incidents have been reinstructed and more extensive training has
been conducted regarding adherence to procedures. The licensee's
corrective actions appear to be adequate. This item is closed.
(Closed) VIO 327, 328/86-68-05, paragraph c, Section 2.4.5, deficiency
D-2,4-6, Loose and Broken Flexible Conduits. NRC inspectors reviewed the
July 16, 1987 TVA response to this portion of the violation and the
corrective action taken. All discrepancies except the loose conduit on
valve 2-FSV-30-14 were addressed in NRC report 87-57. Electrical
maintenance request 8234106 completed corrective action on valve
2-FSV-30-14. NRC inspectors performed field inspection of the valve and
found corrective action to be adequate. This violation is closed.
(0 pen) URI 327, 328/86-68, section 2.4.10 U-2,4-2, Flamastic Thickness on
Cable Trays. NRC report 327, 328/86-68 noted saveral areas in
safety-related cable trays that had flamastic thicknesses in excess of the
1/4-inch that was assumed for ampacity derating due to thermal loading. '
During discussions between NRC inspectors and site management. TVA
committed to an examination of the as-installed flamastic thicknesses at
Sequoyah. The TVA draft commitment was reviewed by the NRC and
discussions were held with TVA's department of nuclear engineering
personnel. TVA sampled flamastic thicknesses at 68 node locations of the
720 total node locations associated with class 1E control power and power
cable trays. The node locations were selected by random number
generation. Actual flamastic thickness examination identified 16 cables
that had thicknesses in exce;* of 1/4-inch, but that no thickness exceeded
1/2-inch. Calculation SQN-t2-025 was performed on all class 1E control
power and power cables based on a thickness of 1/2-inch. The results
indicated that only three cables (2PL3091A, 2PL40918, and 2PL4958A) would
be inadequately sized if the flamastic exceeded the 1/4-inch maximum
assumed thickness. A supplemental field walkdown measured flamastic ;
thickness at 22 locations on these three cables. The supplemental
walkdown identified that only one cable (2PL4901B) of the three has a
flamastic coating exceeding 1/4-inch. That cable was determined to be r
'
adequate because it was a normally deenergized alternate feeder; it was
routed in a mild environment; the flamastic coating exceeded 1/4-inch for
only part of the length of the c4ble; and the ampacity reduction was only
0.67% below full load current. NRC performed inspections on cable trays
that had been previously identified as having thickness greater than
1/4-inch and noted that the trays were control and signal cable trays. A
satisfactory re-inspection of class ?E control power and power cable trays
was conducteci by TVA. The licensea's corrective actions appear to be
-_.
--
_ ._.
.. , .-- ___ .. -
8
adequate. However, this item remains open pending review of the final TVA
resolution memo B25 87-1029-008 and tracking numbers NCO 87-0029-002 and
NCO 0484-005.
(Closed) URI 327, 328/85-45-18, Provide Complete Description of Corrective
Actions Taken To Preclude Circumstances Which Resulted In The Installation
of Upper Head Injection (UHI) Level Switches With Incorrect QA Level
Designations. The licensee identified in PRO-2-85-008 that 3 UHI level
switches did not have proper documentation of seismic qualification and QA
level designation. The subject switches were replaced and a review
commenced on both IE components that had been replaced and components in
stock in the warehouse. NRC report 327, 328/86-48 reported status of the
review and the ongoing resolution of 138 components that had inadequate
seismic qualification documentation. NRC report 327, 328/87-37 provided
further status and reported that not all devices that required replacement
had been replaced. TVA has completed the review and replacement of
required items that are documented in TVA memorandum, D. W. Wilson,
DNE/H. L. Abercrombie/0NP dated March 19, 1987. NRC inspectors reviewed
seismic and QA level designation documentation for the devices and found
them to be acceptable. This item was reviewed against the NRC enforcement
criteria, and was determined to be licensee identified. Therefore, this
item is closed.
(Closed) URI 327, 328/87-24-01, Drawing Control. TVA administrative
instruction (AI)-25, part I, Drawing Control After Unit Licensing,
revision 20, details the methods utilized by Sequoyah for control of
drawings. This procedure has been revised in response to NRC concerns
regarding drawing control. AI-25 provides definition; of the various
drawings; responsibilities for control of drawings; procedures for
receipt, inspection, filming, distribution, and filing; guidelines for
determination of primary drawings and utilization of drawing criteria.
AI-25 specifically addresses guidelines for selection of those drawings
considered primary drawings (ie. , required in the control room for safe
startup, operation and shutdown of the Unit. Appendix 1 to AI-25 lists
those drawings considered to be primary. Included in Appendix 1 are
drawings designated as critical drawings (required for technical support
center and Chattanooga emergency operations center). Primary drawing
changes will be controlled by the operations group manager and the plant
manager as provided in AI-25. This procedure was approved by the plant
operations review committee (PORC) on September 25, 1987. The primcry
drawings in the control room are required to be inventoried against
Appendix 1 of AI-25 annually. Currently, the site document control
section is ensuring prima./y drawings are maintained up to date. This URI
is closed.
(Closed) VIO 328/85-28-05, Accumulator Boron Concentration out of TS
Limit. The following actions were accomplished by the licensee: caution
statements were added to the "drain" and "fill" portions of 501-63.1,
concerning the potential for uncertainties of boron samples taken during
these processes; and a training letter was issued to licensed personnel,
discussing the occurrence and the 501 revision. The licensee's actions
appear to be adequate. This item is closed.
9
(0 pen) URI. 327, 328/87-18-01, Potential for Secondary Bridging, Potential
Use of Flammastic As An Alternative To Solid Barriers, and The Lack of
Criteria for Separation of Safety-Related conduits. This unresolved item
contains three ~ specific examples related to the three generic concerns
stated above. The examples include a divisional cable that was misrouted
in a non-divisional tray, two non-divisional cables that were misrouted in
a divisional tray, and a division A conduit that was routed nearly in
contact with an uncovered division B tray. The concern for potential
secondary bridging (a non-divisional cable routed with one division and
then routed with the opposite division elsewhere in the plant) results
from the indeterminate path of the misrouted cables. A second potential
for secondary bridging existed in free air routing which is not documented
in cable routing records. Long vertical runs of flammastic cable bundles
in the cable spreading room contain both divisional and non-divisional
cables. Because routing is not documented in free air, the potential for
secondary bridging exists. TVA has instituted a change to the cable
routing computer program which adds free air nodes to the cable tray
system. A change in the computer program will prohibit routing a
non-divisional cable which has been run with a divisional cable, from
being routed with the opposite division. This action is adequate to
satisfy the concern regarding secondary bridging in free air. The concern
for secondary bridging due to misrouting of cables in cable trays is not
resolved and requires further review. Documentation reviewed indicated
that the non-divisional cables running in the divisional tray had been
corrected by condition adverse to quality report (CAQR) SQP870702 which
changed the routing cards for the misrouted cables. As a result of the
CAQR, a category D FCR (5572) was written and completed to change the
routing cards. This example is closed. The potential use of flammastic as
a fire barrier which is prohibited by subparagraph 8.3.1.4.2, but was
allowed by section 4.2.9 of Sequoyah design criteria SQN-DC-V-12.2, has
been corrected. A design input memo (temporary change to the design
specification) entitled, Design Input Memo On Separation Of Electrical <
Equipment And Wiring Design Criteria, SQN-OL-V-12.2, dated June 25, 1987, '
added information to the design snecification that precludes use of
flamastic as a fire barrier. This generic issue is closed. The
conduit / cable tray separation criteria issue is not resolved and will be ;
the subject of further review. Items that remain open include the
'
divisional cable that is routed in a non-divisional tray, the potential
for secondary bridging due to misrouting of cables in cable trays, and
both the example and the generic concern regarding the lack of criteria
for conduit / cable tray separation. The correction of the divisional cable
that is in the non-divisional tray, and the potential for secondary
bridging due to misrouting of cables in cable trays will be followed up
under violations 87-52-01 and 02. The conduit / cable tray separation issue :
is a startup item and will be followed under URI 87-18-01. Accordingly,
this item remains open.
(Closed) Violation 327, 328/86-68-05, paragraph d, section 2.4.14,
deficiency 0-2.4-16, Pipe Support Discrepancies. TVA's July 16, 1987,
response to this violation and completed corrective actions were reviewed. .
Field change requests (FCRs) were issued to document the as-built clamp
10
gap and change weld sizes and configuration. Space plates were added to
correct the potential rotation problem with the one snubber and bracket to
. clamp assembly and weld data sheets documented that six undersized welds
were built up to drawing requirements. Programmatic changes related to
procedure adherence by-craft technicians and- quality control-(QC)
inspectors, detailed in the Sequoyah nuclear performance plan and
elsewhere, have addressed the root causes of this violation. TVA's
corrective actions are adequate and this portion of the violation is
closed.
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 was identified during this inspection,
and is identified in paragraph 13.
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 cor. firmed operability of instrumentation. The
inspectors verified the operability of selected emergency systems,
and verified compliance with Technical Specification (TS) Limiting
Conditions for Operation (LCO). 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 safety
injection system on Unit 2 to verify operability and proper valve
alignment.
No violations or deviations were identified,
b. Safeguards Inspection
In the course of the monthly activities, the inspectors included a
review of the licensee's physical security program. The performance
of various shif ts of the security 6rce was observed in the conduct
of daily activities including protected and vital area access
controls; searching of personnel and packages; badge issuance and
retrieval; patrols and compensatory posts; and escorting of visitors.
11
In addition, the inspectors observed protected area lighting,
protected and vital areas barrier integrity. The inspectors verified
an interface between the security organization and operations or
maintenance. Specifically, the resident inspectors: responded to
bomb W eats, fires, etc. ; interviewed individuals with . security
concerns; inspected security during outages; reviewed licensee
security event report; visited central or secondary alarm stations.
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 basis,
radiation work permits (RWPs) were reviewed and specific work
activities were monitored to ensure the activities were being
conducted in accordance with applicable RWPs. Selected radiation
protection instruments were verified operable and calibration
frequencies were reviewed.
During a tour of the auxiliary building the inspector identified two
separate deviations from the licensee's health physics procedures.
These involved dress out noncompliances which were minor in nature
and did not result in any personnel contamination or overexposure.
The licensee currently has a long term health physics corrective
action effort in progress which includes improvements in dress-out
and frisking issues. The two issues were discussed with operations
3
section management, NRC m,'nagemnt, and Sequoyah site management,
during the exit conducted for this inspection period. Sequoyah
health physics management initiated two radiological incident reports
(RIR) et the time the two noncompliances were identified by the
inspector. The inspector will review the resolution of RIR 87-24
dated October .6,1987, and RIR 87-23 dated October 16, 1987, after
completion of licensee corrective action. The inspector had no
further questions.
No violations or deviations were identified.
, 6. Monthly Surveillance Observations (61726)
The inspectors observed / reviewed the TS required surveillance testing
'
Tisted below and verified that testing was performed in accordance with
adequate procedures; that test instrumentation was calibrated; that L%
were met; that test results met acceptance criteria requirements and were
reviewed by personnel other than 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.
,- (
(;J '
l
12
a. SI-166.36, Diesel Starting Air Valve Test
SI-7, Electrical Power System: Diesel Generators
The above two surveillances were performed in tandem with a test
director in charge of SI-166.36 and an_ auxiliary unit operator (AV0)
in charge of SI-7. During the performance of the two surveillances
the test director of SI-166.36, assumed that the AVO performing SI-7
would record a piece of data that was necessary .for the performance
of SI-166.36 (the time for the diesel to reach and attain 900 rpm).
After the diesel generator had been started and was running at the
required rpm, the test director realized that the AVO was not
required by SI-7 to record the time it took the diesel to reach the
required speed. The test director retrieved the information from an
alternate data source (control room operator) without repeating the
diesel generator starting sequence. The inspector had no further
questions.
b. SI-94.5, Reactor Trip Instrumentation Refueling Outage Channel
Calibration. Portions of this SI were observed and the inspector had
no questions.
c. SI-98.1, Channel Calibration for Engineered Safety Feature
Instrumentation (Steam Flow & Pressure).
IMI-99 CC 9.108, Offline Channel Calibration of Loop 3 Steam
Generator Steam Pressure Channel II.
On October 7, 1987, the inspector observed testing in progress on CC
9.10B of IMI-99. The inspector noted that a step in the procedure
was signed off stating that the precautions and prerequisites of
FT/CC-9 were completed. FT/CC-9 was not at the work site and the
technician that had signed it off stated that she knew what the steps
were. The following were stated:
(1) Assure double person signoff for critical steps
(2) Notify operations prior to starting work
(3) Place orange stickers in the control room alarm windows
(4) Assure other loops are in service as necessary
'
At a later date the inspector determined that the precautions and
prerequisites of FT/CC-9 were somewhat different than what he had
been told. The actual procedure stated the following:
3. PREREQUISITES
3.1 Copy of functionti test or channel calibration procedure
and its associated data sheet (s).
3.2 Provide adeouate communication between instrument
locations.
-
13
3.3 Equipment required - see data sheet of functional test or.
'
channel calibration being performed.
3.4 For channel calibrations only, calibration data card for
each instrument listed in component list at the beginning of
each channel calibration procedure.
4. PRECAUTIONS
4.1 Calibration date on test instruments must be current.
4.2 Test may be performed on only one protection set at a time. '
When one protection set is being tested, the remaining
protectior, sets must be in nornal (untripped) mode.
4 . 'J Notify shift engineer of maintenance, calibration, ' or
functional test to be performed.
4.4 Instrumentation and test equipment must be energized at
least the minimum length of time to achieve stability in
accordance with applicable manufacturers instruction manual'.
4.5 If during any at power test, a related reactor protection ;
channel trip is actuated from another protection set, the test j
must be terminated and all channels returned to normal i
(untripped) condition.
!
4.6 Control systems are to be placed in the manual control mode i
before any change is made in a channel defeat and/or channel
transfer switch position. After the change is made, the control
system may be returned to automatic control.
4.7 Observe all posted health physics precautions obtaining a
special work permit when required or if work to be performed ,
could result in changes in radiation levels in work area. Tools ,
or equipment being moved from contaminated to regulated zones or
from regulated to clean zones must be surveyed by health physics
l unit prior to removal.
4.8 Use IMI-118 for backflushing and filling of possible
contaminated instrument lines.
l 4.9 Loops containing Barton transmitters, Models 763 and 764, ,
'
used in conjunction with the Foxboro 610A power supply shall be
l- de-energized prior to switching the analog channel test to
l normal, and then re-energized af ter placing the test switch in ,
'
the normal operating position.
Although no violation of the procedure was observed, there was an !
appearance that the procedure was being performed without appropriate '
controls. The inspector discussed with licensee management the
l
l
<
'
>
'
i
,f. ,
f a-
, ! l
e
.
14
I
importance of 'the technicians either knowing explicitly, through
'
p'. training, their responsibilities or having a copy of an approved,
J
/ procedure in-hand.
t d. SI 247.900,,E691neered Safety Features Response Time' Verification
7
1 ', , During review of SI-247.900,. Engineered Safety Features / Response Time
/
'
Verification, the inspector determined that the SI does not meet the
2 - requirements of TS 4.3.2.1.3. TS 4.3.2.1.3 requires that containment
spray responsa ' time be demonstrated to be within the limit at least
- once per 18 months. The inspector determined that the time period
for the containment spray pump start interlock . to close was not
included as part of the response time for the containment spray
isolation valve to open. The pump start interlock must be satisfied
before valve movement will begin. TVA procedures bypassed this
interlock during response time testing. The failure to have an
adequate SI to measure containment spray response time is identified
as violation 327,328/87-65-02.
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 I
with approved procedures, regulato y 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 record
c, 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; radiologicdl controls were implemented; QC hold points <
were established where required and were observed; fire prevention '
controls were implemented; outside contractor force activities were
controlled in accordance with the approved Quality Assurance (QA) program;
and housekeeping was actively pursued.
a. Work Request (WR) B298870; Maintenance on Temperature Monitor
TM-68-43K :
t
During the performance of the above corrective maintenance, the i
technicians implemented configuration control through the use of
instrument maintenance instruction IMI-134, Configuration Control of
Instrument Maintenance Activities. When this maintenance was [
' observed the TM-68-43K circuit card had been placed into its cabinet
and troubleshooting was in progress. The configuration indicated on
,
L-
_ _ ___-
15
i
the IMI-134 data sheet did not indicate the placement af the circuit
card into the cabinet.
IMI-134 section 3.2.7 states that the technician is to list on the
data sheet (work performance sheet) any configuration changes. "This
includes: jumpers, wire lifts, inhibits, temporary instrument
settings, unbolting flanges, disconnecting tubing and pipe fittings,
temporary connection, etc."
During this troubleshooting period, the configuration work
performance sheet did not accurately reflect the configuration of the
equipment covered by the work activity or the trouble shooting that
was conducted due to the fact that the card was in the rack when the
configuration log indicated it was out of the rack. This is due '.o
the licensee policy of logging the card one time for trouble shooting
and then removing and reinstalling the card at will to accomplish
that troubleshooting. The corporate maintenance u.anager had
identified a similar programmatic issue involvina adequate functional
testing of equi 7 ment after maintenance is performed. During a
meeting which was attended by the plant maintenance supe ri nter. den t ,
corporate maintenance manager, and NRC management, it was agreed thn.t
selection of adequate post maintenance testing (PMT) deperded en in
accurate understanding of the troubleshootino that was conoucted and
the cor. figuration changes that occurred in the equipment. This issue
will not result in a violation because the equipment was out of
service during this periud and the licensee had previousi/ identified
tha need for functional testing and was implementing corrective
action when this issue was identified. The inspector had no further
questions,
b. Preventive Maintenance (PM) 0961-068; Main Control Room Recorder
2-M-5
Portions of the 2-REC-068-VAR scaling were observed by the inspector.
The PM referenced the vendor's manual and indicated that the scaling
shculd be accomplish " within the limits set by the vendor. The
va' car set the scaling limits between zero and fif ty milli-amps on
tne wand type indicator pens. The inspector observed that the
technicians had apolied cixty milli-amps to one of the indicator
pens. The techniciins had applied the sixty milli-amps to the pen in
order to mo(6 it aside and had decided that the additional amperage
would not damage the recordar. This particular equipment is not a
critical safety system component (CSSC) and therefore the require-
ments of TS 6.S.1 do not apply and no violation will be issued. This
issue was discussed with the plant maintenance superintendent, the
corporate maintenence superintendent, NRC management and other plant
management. The inspector had no further questions.
J
a r ~.
-- - - - - - - - - - -
- - - - - - - - - - - - - - - - - - -
i;fl7
(?
l < . ,f I
g.m ,.
.
fs, -
n 5 (' .
- ,9 c ?? ,
'
.
'
16
o
,
c. ' Work r4quett B28468-d; Calibration of Indicating Instruments for
'
480 volt Shutdown Board (282-B)
This item was' performed to validate the voltage, amperage, and watt
~
meters for the 480 volt shutdown boards. - All three meters were found.
within ttlerar.ce. .No adjustments were necessary,
d. hork Plan-(VP) 12360; Fire Protection Connections to the Control Room
Computer.
On Septerrber 28, 1987, the inspector observed work in progress on WP
12360. This WP implements a portion of engineering change notice
(ECN) L5841 on connections between fire detection panel 0-L-633 and
the. plant computer. An engineer was in charge of testing and was
acting as'the test director. Procedures were out and being followed
, by the Individuals performing the WP. The engineer and testing
personnel appeared to be knowledgeable of the work and their
responsibilities. The inspector had no further questions. l,
Ho violations or deviations were identified.
3. Licenneo Event Report (LER) Followup (92700)
The fel;owing LERs were reviewed and closed. The inspector verified that:
reporting requiremente. 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 t
merit; no unreviewed safety questions were involved; and no violations of .
regulations or TS condition had been identified. [
a. Closed LERs
_
LER 327/86-044 Inadequate Verification of ECCS Flow. As previously
s'.ated i n report 87-36, the remaining required licensee action in
this matter way to revise procedure SI-137.3 to include reactor i
coolant pump (RCP) seal differential pressure requirements. Revision
4 to this procedure was approved on May 29, 1987. NRC review of the r
revised procedure reveals that the necessary changes have been
adequately incorporated. This item is closed. p
LER 327/87-001, Trip Setpoints for Air Circuit Breakers (ACBs)
Incorrect. The tri; setpoints for ACB's on shutdown boards that feed
control and auxiliary building vent boards were incorrect due to a
desig, error. ECN L6SS3 has been issued and the loads have been
! analysed to deterc.11ne proper trip setpoints. WP 12636 has been
issueo and is being worked. The work required to satisfy this LER *
has heq c v.pleted. Licensee's corrective actions appear to be ,
acce; trble. This. item is closed, i
1
LER ??.7/87-On, Fa dure to Cycle Test Six Fire Protection Valves per
TS Surveillance Requirement 4.7.11.1.d Due to an Inadequate
L _ , , . --
- - - - _ _ .-__. _ _
17
,
Procedure. The inspector reviewed revision 11 to surveillance
instruction SI-172, Fire System Testable Valve Cycling, and verified
that the six valves have been included. These valves are not
critical safety system components per SQA-134, Critical Structures,
Systems and Componeras (CSSC) list. The inspector also reviewed
completed data sheets indicating these valves had been satisfactorily
cycle tested per SI-172 on August 5, 1987, This item is closed.
LER 327/87-035; Diesel Generator 2B-B Start During Replacement of
Fuses. The licensee was unable to reproduct the event. Subsequent
testing verified that all diesel generators started as required when
a fuse was removed. Conclusion: event may have been generated by
momentary high impedance in relay (ES28Y) circuit causing only that
relay to pickup. Licensee's investigation and results appear to be
adequate. This issue is closed.
LER 327/87-038; 6.9.KV Circuit Breakers Not Tested. The 6.9 KV
circuit breakers for pressurizer control heaters that are used to
protect IE busses from faults on non-1E loads have not been tested
.. because of an engineering oversight. The licensee has generated
surveillance instruction (SI) 737-1A, 18, 2A, 2B to cover TS
surveillance requirement 4.8.3.3 for the 6.9 KV pressurizer control
~
~ -
heater circuit breakus. Thes~e' h n e been completed. Licensee's
corrective actions appear to be acceptable.
LER 327/87-042; Inadvertent Starting of The Fire Pumps During a loss
of Coolant Accident. The licensee submitted an information LER to
identify a potential problem with the shutdown power capabilities at
the Sequoyah plant. The licensee determined, as a result of an
electrical calculation program review, that an event could occur
which was not analyzed for in the final safety analysis report. This
event was a loss of coolant accident concurrent with the inadvertent
starting and running of the station fire pumps. The licensee
determined that starting the fire pumps concurrent with a LOCA could
potentially degrade the auxiliary electric power system voltage and
thereby prevent safety related equipment from performing its intended
function. The NRC is reviewing this and other issues in the
electrical calculations design review process to determine if the
potential problems identified by the licensee are adequately
resolved. The licensee determined that no immediate corrective
action was necessary because of the mode condition of each of the
units. Long term licensee corrective action will include
administrative compensatory measures on the control of the unit 2
fire pumps. During its operational readiness inspection, the NRC
will examine all administrative compensatory measures employed by the
licensee prior to the startup of either units. The generic issue of
compensating issues for degraded auxiliary electric power systems
will be addressed as a topic in the NRC operational readiness
inspection 327, 328/87-73. This LER is closed.
,
_
i.
18
LER 327/87-024;-2-FCV-74-3 Exceeded Allowable Stroke Time. .The. local
stroke time of a flow control valve 2-FCV-74-3 exceeded the maximum
allowable stroke time due to an inadequate procedure. The inspector
reviewed the stroke time test results for 2-FCV-74-3, which showed an
acceptable local and remote test after maintenance on June 12, 1987.
TVA's review data comparing the most recent local stroke times for
all power operated valves which did not identify any additional
discrepancies, was also reviewed. A review of the applicable
surveillance instructions indicated that they have been revised to
require comparison of local stroke times to the maximum allowable.
The licensee's corrective actions appear to be adequate. This item
is closed.
LER 327/87-023; Ir.spection of Ice Condenser Assemblies. Inspection
of the ice condenser vent assemblies has not been performed in
accordance with the TS due to lack of guidance in the surveillance
instruction. The licensee determined that the total requirements of
TS surveillance requirement 4.6.5.3.2.b was not being met due to ice
condenser vent assemblies being incorrectly interpreted as being the
intermediate deck doors. This interpretation did not support an
inspection of the vent curtains for free movement. The immediate
corrective action was to verify free movement of the vent curtains.
Long term corrective action revised surveillance instruction
(SI)-108 Ice Condenser Doors, to provide proper guidance for
ensuring free movement of the vent curtain during conduct of the TS
surveillance requirement. The inspector reviewed revision 9 to
SI-108 and found that it contained adequate guidance for ensuring
free movement. of the vent curtain in step 6.2.4. The licensee's
corrective action appears adequate. This item is closed.
LER 327/87-046; Possible Dilution of The Emergency Core Cooling
System During Large Break LOCA Events. This LER was issued by the
licensee for information only. The licensee is currently reviewing
the possible safety implications that a given plant configuration
would have on a post-LOCA long term cooling condition. The concern
involves the possibility tnat nonborated water from such sources as
the essential raw cooling water system, high pressure fire protection
system, primary water system and component cooling water system will
affect the long term shutdown capability of the unit. This issue is
being handled as a startup item by the licensee. The potential
generic issue has been transferred to the NRC, Office of Special
Projects for disposition. This LER is closed.
b. Reviewed LERs which Remain Open
LER 327/87-030; Blown Fuse in Emergency Start Circuits Result in
Spurious Emergency Diesel Generator Starts on Two Occasions. A
review of the fuses indicates that the f ailed fuses came from
(FLAS-5) lot no. 3. As of July 13, 1987, 69 FLAS-5 fuses have failed
with 67 confirmed from lots 2 and 3 and two indeterminate. A change
in manufacturing process was initiated (between lots 3 and 4) by the
l
l
_
19
. manufacturer as improvements in the production process. The licensee
is in the process of changing out all FLAS-5, lot 2 and 3 fuses. The
emergency diesel generator, emergency start circuit fuses have been
verified or replaced with fuses from lots manufactured after lot' 3.
The licensee's corrective actions appear to be acceptable. This
issue remains open pending completion of NRC review of the fuse-
replacement program.
(0 pen) LER 327/86-039 Surveillance Requirements Not Performed
Because of Inadequate Procedures. Licensee event report (LER) 327/96-39 reported that procedures were not adequate to test all
interlock functions of the reactor trip system interlocks. Reactor
trip, P-4 permissive (technical specification 4.3.1.1.2, Table
3.3-1.22G) or to test the ice condenser inlet door position at the
local panel during- the functional test (technical specification 4.6.5.3.1). NRC inspectors reviewed surveillance instruction SI-108,
Ice Condenser Doors. The procedure had been revised to include
verification of the inlet door position at the local panel as
required by the technical specification. This item is closed.
Surveillance instruction SI-268-3, periodic verification of the P-4
interlock, was written and provides required testing for the "turbine
trip on reactor trip" function of the reactor trip P-4 permissive.
Testing of the "main feedwater valve closure on low reactor coolant
system average temperature with reactor trip" function of the reactor
trip, P-.4 permissive interlock is in the process of being evaluated.
A draft technical specification interpretation entitled "technical
specification interpretation log No. 94 revision 1-definition of
total interlock function" dated October 7, 1987 was written and
discussed the total interlock function as including the process
input, solid state protection system (SSPS) logic (if any), and
output function. The output function would include verifying the
output of the logic circuitry including the energizing of the SSPS
master relays for ESF permissives and verifying the logic through-the
output of the SSPS undervoltage card for reactor trip permissives.
The requirement to test the slave relay output contacts as a
surveillance instruction (SI) requirement of the technical
specifications is the issue under TVA review. The P-4 permissive
slave relay contact for the main feedwater valve closure is presently
tested but not by a technical specification surveillance instruction.
This item will remain open sending review of TVA action concerning
the slave relays.
(0 pen) LER 327/86-048; Inadequate Verification of ECCS Flow Due to
Procedural Inadequacy. As previously stated in report 87-36,
procedure SI-260.2.1 is to be instituted and procedure SI-260.2 is to
be revised to ensure that Centrifigal Charging Pumps are tested under
conditions as specified in the TS. Completion of these actions is
not required for Unit 2 restart, but is required for restart of
Unit 1. Conversations with licensee personnel reveal that these
corrective actions are not yet complete. This item remains open.
__-. _ _
.
20
(0 pen) LER 328/87-041 revision 1, Loss of RHR Flow Resulting From
Mispositioning of a Breaker Due to Personnel Error. A student
assistant unit operator (AVO) opened a circuit breaker .while
performing a routine surveillance to verify alignment of breaker 2 on
the 120 VAC vital instrument power board 1-IT. Licenste's actions
prevented damage to RHR and flow was restored within 4 minutes. The
student AVO was counselled on the necessity for good communications,
of following procedures (attention to detail) and the consequences of
failure to do so. The licensee's specific corrective actions appear
to be acceptable. The generic issue of allowing student operators to
unilaterally operate plant equipment will be reviewed at a later'
date. This item remains open.
(0 pen) LER 327/86-042, Two Surveillance Requirements Not Performed
~
Because of Inadequate Procedures. As previously stated in report
87-36, the licensee has requested relief from American Society of
Mechanical Engineers (ASME), section XI, subsection IWP-3100, for
several safety related pumps because of possible damage to the pump
by throttling the pump miniflow recirculation valves during testing.
Current information -obtained from the licensee indicates that the
granting of such relief is in the final approval process at NRC, and
will be issued in the near future. This item remains open until the
relief request is granted.
(0 pen) LER 327/86-020, Failure to Perform a TS Required Quarterly
Functional Test. As previously stated in report 87-36, procedures
SI-244, Periodic Functional Tests of Radioactive Effluent Monitoring
Instruments and 51-244.2, Peridodic Functional Tests of Radioactive
Effluent Monitoring Instruments were to be revised to include a
functional test for channel F-15-43. These procedures have been
revised by the licensee. However, NRC review has identified several
questions concerning the adequacy of the tests. These questions are;
(1) the procedures provide no guidance as to what position flow
control salve FCV-15-43 is to be lef t upon completion of the test; ;
(2) is flow indicating controller FIC-15-43 required to function in
both "auto" and "manual" modes, or in the "manual" mode only. The
latter question is due to conflicting statements in licensee ,
generated reports. Attachment 1 to PRO 3-86-031, under "additional i
comments", stated "F-15-43 is used for auto isolation flow control", '
while the "analysis of event" section of the LER report states that
"this instrument provides only an a'.rm on the plant process
computer" and that "the flow channel does not provide an isolation
function"; and (3) presently,51-244 requires that this channel in
unit 1 be tested in the "manual" mode only, while ?I-244.2 requires
that both "manual" and "auto" modes be tested in Unit 2. As a result
of these questions, additional information and/or action is required
from the licensee before this item can be considered complete. This
item remains open.
(0 pen) LER 327/87-027; Surveillance Requirement Was Not Fulfilled
Because Four Essential Raw Cooling Water (ERCW) Valves Were Not
_,
21
Verified in the Correct Position. This event occurred because the
subject valves were only verified to be in the correct position
(throttled) every 90 days in accordance with surveillance instruction
(SI-682), ERCW Flow Balance Valve Position Verification. Technical
Specification surveillance requirement 4.7.4.a required them to be
verified in the correct position every 31 days. These valves had
been verified to be'.in the open position every 31 days in accordance
with SI-33, "ERCW Valves Servicing Safety-Related Equipment." TVA
concluded that because the valves were tagged as throttled valves and
the knowledge of plant operators, these valves had remained in the
throttled position. Corrective action was to delete the four subject
valves from SI-33 and revise SI-682 to ensure that these valves are
verified in the correct position every 31 days. SI-682 was also to
be rerun prior to restart.
The NRC inspector confirmed that SI-33 and SI-682 have been revised
as indicated. However, review of the pertinent documentation has
resulted in the following concerns related to the performance of
these sis and the evaluations of discrepancies identified.
The LER analysis does not address the question of how operators
could sign off 51-33, verifying valves in the opsn position, if
they were in fact in a throttled position.
- A review of SI-682 data packages performed in 1986 and 1987
indicates that a number of valves have been found improperly
positioned at each inspection. Hispositioning ranged from one
turn or less . to the completely opposite position from that
specified (for open or closed valves). Numerous potential
reportable occurrence (PRO) reports have been generated with no
apparent identification or correction of the root cause. PRO
investigations do not appear to be accurate or adequate.
Completed SI-682 data packages had not always listed improperly
positioned valves in a deficiency log and there were inadequate
dispositions of the discrepancies (June, September and December
SI-682 performances).
The plant operations review staff (PORS) is reviewing these concerns
and is performing an indepth root cause analysis. Pending NRC review
of this reassessment by TVA this item remains open.
9. Sequoyah Requalification Program Corrective Actions
Two license examiners of Region II's Operator Licensing Section conducted
an unannounced inspection of Sequoyah's corrective actions for
deficiencies noted during the December 1986 requalification examination.
Of particular interest were actions taken to upgrade training of Reactor
Operators on the bases for steps in the Emergency Operating Procedures
(EOPs).
- _ _ _ _
.
.
22
On October 8-9, 1987, the inspectors observed ongoing requalification
training for a group of operators, involved in the last week of the 1987
requalification cycle. This consisted of simulator and classroom
training, a ' final written examination and a graded simulator evaluation.
Lesson plans, examination questions and simulator scenarios utilized in
requalification training were also reviewed with the focus being on E0P
related areas.
The findings of this inspection are as follows:
a. The requalification lesson plans for E0P training contained learning
objectives which specifically addressed questions asked on previous
NRC examinations and did not incorporate any further objectives which
would be appropriate for this training. The objectives should be
i
more broadly based to cover the pertinent topics within the lesson
l plan, not just previously asked NRC examination questions,
b. Specific requalification lesson plans did not exist for coverage of
individual emergency procedures. It was nr.ted that some questions
related to procedural bases were developed, however, there were no
requalification learning objectives associated with these questions.
Lesson plans utilized in the hot license training program were
reviewed and could readily be adapted to the requalification program,
'
c. The written examination questions utilized for evaluating the
requalification training in E0P usage were phrased exactly the same
as the learning objectives. No other questions were developed to
examine this area, and further effort should be conducted to create a
broader sample of questions to ensure a valid examination can be
generated for each requalification group.
d. Simulator training covered the learning objectives, however, the
instructor-to-student ratio for training sessions was 1:5 and for
final evaluations was 2:4. Consideration should be given to
improving this ratio in order to provide more attention to
individuals within the crew and equalize the burden on the
instructors to control the simulator scenario as well as evaluate the
operators.
During this inspection, the corrective actions initiated by the
facility were identified and their implementation was found to be
adequate. As noted above, however, further effort is required to
improve requalification training on the E0Ps. These findings were
presented to facility staff members at an exit meeting on October 9,
1987.
10. Plant Operations Review Committee (PORC) (40700)
The inspector conducted a functional review of the PORC which performs as
the onsite safety review committee. This functional review was intended
to evaluate if the activities of PORC could support the heatup and
u-----______-._-.------- - - - - - - _ _ _ _ _ - - - - - - - - - - - - - - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
r
23
eventual startup of unit 2. The site is currently changing to a qualified
reviewer concept as a result of TS change 87-34. This review however, has
been conducted prior to the implementation of the TS change. The
inspectors observed the PORC review proposals which affected nuclear
safety including issuance of plant procedures and changes thereof,
modifications to systems and equipment, and unreviewed safety question
determinations (USQDs). The inspectors attended PORC meetings conducted
on October 7, 8, 13, 14, 26, and 28. The inspectors had the following
comments:
One plant modification reviewed by PORC on October 8, 1987,
addressing manholes which held safety related cables, was presented
to PORC without an approved USQD. PORC rejected the modification.
' One plant special maintenance instruction (SMI) addressing thermal
excursion of piping had been rejected by a previous PORC. The
observed PORC was not aware of why the SMI had been rejected and had
to request information from the individual that was presenting the
SMI to explain why the previous PORC meeting had rejected it. This
is an example of a loss of continuity with respect to PORC oversite
activities. In addition, the individual presenting the SMI stated
that the thermal property being tested for would result again because
of a lack of a program to control field routed lines and conduits
with respect to thermal interference. The SMI was rejected by PORC
a second time.
No violations or deviations were identified.
11. IE Bulletins (92701)
IE bulletins (IEB) are documents issued by the NRC which require certain
specific actions of the addressee. The inspector has reviewed the actions
taken by the licensee as a response to the below listed IE bulletins. The
inspector verified that: corrective actions appeared appropriate; generic
applicability had been considered; the licensee had reviewed the event and
that appropriate plant personnel were knowledgeable; no unreviewed safety
questions were involved; and that violations of regulations or TS
conditions did not appear to occur.
IEB 83-01, Salem Anticipated Transient Without Scram. The inspector
reviewed licensee response dated March 4, 1983, A27 830304 010, and
maintenance instruction MI-10.9, Removal, Inspection, Lubrication, and
Replacement of Control Rod Drive MG Set, Reactor Trip, and Bypass Circuit
Breakers. After a review of the supporting documentation and the
implementation of the actions required initially by this bulletin the
licensee's actions were determined to be adequate. This item is closed.
12. Inspector Followup Items
Inspector Followup Items (IFIs) are matters of concern to the inspector
which are documented and tracked in inspection reports to allow further
-
24
review and evaluation by the inspector. The following IFIs have been
reviewed and evalcited by the inspector, ~The inspector has either
resolved the concern identified, determined that the licensee has
performed adequately in the area, and/or determined that actions taken by -
the licensee have resolved the concern,
t
.(Closed) IFI-327/87-11-01, Employee Concern Program (ECP) Element 11301,
Design of Plates Rev. 6. The inspector determined that the specific
concerns addressed in this item were to be closed out by the safety
evaluation report (SER) to be completed by NRC. This item is redundant to
that closure process. Therefore, this item is closed and element report
11301 will remain open until the issuance of the SER.
(Closed) IFI 327/87-11-03, Pipe / Fittings As Related To Construction. This
i is part of the on going employee concern element report 17105 Revision 2.
Parts of the essential raw cooling water system were changed from carbon
steel to stainless steel without a quality seismic analysis being
performed. This has now been accomplished by the civil section of TVA's
division of nuclear engineering. IFI 327/87-11-03 is redundant to element
report 17105 Revision 2 which will be addressed by the NRC in an SER. The
following references were reviewed: (1) TVA's Engineering Change Notice .
L-5009; (2) Memorandum from H. L. Abercrombie to D. W. Wilson dated,
February 6, 1986, (RIMS B 25 870312040); (3) Memorandum from
J. A. Southers to Sequoyah Engineering Project Files dated March 25, 1987,
- - (RIMS B 25 870325048); and (4) Memorandum from D. C. Hatcher to Sequoyah
Engineering Project Files dated, April 8,1987, (RIMS B 25 870408078).
This IFI is closed. The element report 17105 will remain open pending
issuance of the SER.
(Closed) Observation 327,328/87-06-EEB-1, Battery and Charger Sizing.
This observation was related to errors in the calculation for sizing of
the class 1E batteries. The NRC design calculation review team _ reviewed
TVA's revised calculations and found them acceptable. This calculation '.
was performed using a maximum inverter load of 17.5 KVA instead of its
name plate rating of 20 KVA. TVA's electrical engineering branch informed
the team that they intend to establish design and administrative limits to
keep inverter loads less than 17.5 KVA. TVA also committed to reevaluate
sizing criteria for the battery and charger whenever the DC system load ,
changes. Design criteria procedure SQN-DC-V-11.6 has been changed to
include in step 1.2.1 system description. . ."The maximum allowable load to
each inverter is 15 KVA." Also, drawings 45N703-2, 45N703-3, 45N703-4,
and 45N706-1, 45N706-2 45N706-3, 45N706-4, and 45N703-1 have been revised
to include the note "although each vital inverter is rated to deliver 20
KVA, the maximum design load on each is 15 KVA." Sequoyah engineering
,
procedure SQEP-09 has been revised to adequately require a calculation for ,
the battery and charger whenever a change in loading of the DC system '
'
occurs. Also, assumption 2.7 in calculation procedure SQN-CPC-004
addresses calculating inverter lord at 17.5 KVA and a design limit of 15
KVA. Licensee action on this item is Adequate. This item is closed. ;
,
f
- - --
___ __
25
(Closed) IFI 327, 328/85-26-02, Ice Condenser Door Blocks. This inspector
followup item addressed the control of the devices used to block shut the
ice condenser intermediate doors when the units were in modes 4 and 5.
The devices are controlled through the use of work requests and verified
through routine auxiliary unit operator performance. This item is closed.
(Closed) IFI _ 327, 328/86-69-02, - Effluent Releases In Accordance With
Surveillance Instruction- SI-400.2. On November 19, 1986, a ten minute
transfer was conducted from the high crud tank B to the turbine building
sump without the appropriate setpoint change adjustment to the in-line
radiation monitor. 0-RM-90-225, being performed _as required by
surveillance instruction SI-400.2, Condensate-Demineralizer Waste Effluent-
To The Turbine Building Sump - Periodic Continuous Releases. The system,
by design, provides for a pathway from the tank to cooling tower blowdown
which goes immediately to the river. In this case, however, the transfer
was made to the turbine building sump which is only released itself-
through another monitored pathway. The interpretation given_ the
inspector from the OSP technical staff was that a release occurred only
if it went directly to the environment. Therefore, the transfer of
,
'
liquid from the high crud tank to the turbine building sump was not
considered to be a release even though it left the radiological
controlled area.
The licensee discovered this occurrence 10 minutes into the event and the
transfer was immediately terminated. Once this condition was. discovered
the licensee complied with the action statements of Technical
Specification 3.3.3.9 for this event. Additionally, licensee samples
reflected that the activity of the transferred water was within the
release limits of Technical Specification 3.11.1.
Personnel involved were reinstructed in the requirements of SI-400.2 for
setpoint changes on rad monitors at a January 7,1987 safety meeting.
In addition, the individual involved was privately counselled in
following procedures. Because this issue was identified by the licensee,
the licensee took immediate corrective actions, and no release to the
environment occurred, no violation will be issued and IFI 327,
328/86-69-02 is closed.
i (Closed) IFI 327, 328/86-28-19, Unanalyzed Installation of Piping
Insulation. This item involved the processing of nonconformance report
SQN-QAB-8105 and the necessity to determine if pipe insulation affected
the structural integrity of safety related piping. The issue was
addressed by the Watts Bar nuclear plant employee concerns task group
(ECTG) and in a letter (Brown /Abercrombie, T25 870106 855) dated
January 6, 1987. No structural issues as a result of insulation appear to
exist at Sequoyah. This item is closed.
(Closed) IFI 327, 328/86-71-08, Surry Plant Feedwater Line Rupture.
Following the Surry feedwater line rupture this IFI was opened to ensure
that the licensee initiated a program to evaluate the occurrence of wall
thinning in the feedwater system. The licensee initiated a voluntary
l
. _ - _ -
1
.
26
,
program employing various inspection and testing techniques. This
voluntary program was implemented prior to the issuance of IEB 87-01,
Thinning of Pipe Walls In Nuclear Power Plants, dated July 9,1987. The. ',
liconsee responded to the bulletin- in - a letter (Gridley/NRC,
L44 870918 806) dated September 18, 1987. All further NRC. questions
concerning the adequacy of the licensee's feedwater system pipe -thinning
will be addressed within the bounds of the IEB. This item is closed.
(Closed) IFI 327, 328/84-24-04, and 327, 328/86-60-07, Generic Fittings. '
This item was initially closed in inspection- 327,328/86-15- and
subsequently reopened in inspection 327, 328/87-43 to reevaluate all i
aspects of the seal table tube ejection event. The inspector reviewed the
licensee's evaluation of the generic (mixed) fittings as ~ contained in
problem identification report (PIR) SQNEEB87131. This PIR referenced a
report from Singleton Materials Engineering Laboratory (SME). This report
was conducted to determine the adequacy and safety of compression fitting
assemblies which were installed contrary to manufacturers' recommendations- ;
'
and/or using mixed manufacturer fitting components. The SME report
concluded that regardless of various improper assembly techniques, the ;
assembly is acceptable if the joint was assembled to leak tightness as >
displayed during hydrostatic testing or in-service leak checking. Various
low-amplitude (service) vibration and high amplitude (seismic) vibration
tests were performed, as well as axial tension (pullout) tests to support t
this conclusion. The licensee has performed walk-down inspections of !
various plant areas and found no leaking fittings on instruments or drain F
lines, A procedure (MMT-28) was written to implement formal fitting i
assembly requirements. Site personnel have been trained in the use of
this procedure and are scheduled for periodic retraining. This item is
closed.
(Closed) 327, 328/87-06 Observation EEB-3,120V AC and DC Solenoid Valve
Voltage. This observation noted that electrical calculations of cable ,
impedance do not consider cable slack, high ambient temperature, and drops !
, across connections. Additionally, assumptions are not listed in a !
dedicated section. TVA's technical justification is documented in !
"
B43870529908 and is included in TVA's response to the NRC dated July 2,
1987. Nuclear engineering procedure (NEP)-3.1, Calculations, requires t
assumptions to be listed and documented and includes a sample format. The ;
licensee's response, as described in the July 2, 1987 letter, is adequate.
This item is closed. .
!
(0 pen) 327, 328/86-27 0.3.3-1, Pipe Support Friction Design. This item ,
described TVA's failure to analyze pipe supports for friction forces due t
to thermal displacements. TVA has responded to the NRC in letters dated !
July 28, 1986, December 31, 1986 and August 21, 1987, with a program to (
evaluate 60 randomly selected supports on major systems. Design criteria a
SQN-DC-V-24.2 Supports For Rigorously Analyzed Category I Piping, defines .
the requirements used in the evaluation program. The general, restart I
criteria contained in volume 2 of TVA's nuclear performance plan (NPP) [
have been made specific to pipe supports in CEB-CI 21.89, Modification i
Priorities for Pipe Supports on Rigorously Analyzed Category I Piping,
r
!
I
i
---
, , , - , - - - .,, - - - - _ . . -
-__ _
27
These criteria have been transinitted to the NRC in a letter dated,
August 31,. 1987, and will be used to determine whether an action is
required pre-or/ post-startup. This item will remain open until the
evaluations are complete.
(Closed) IFI 327, 328/87-02-06, Review Supplemental Report on Glycol Valve
Stroke Failure. The inspector reviewed revision 1 to LER 327/084-070.
, TVA's further investigation into the valve failure concluded that failure
was due to a buildup of sediment on the stem, was an isolated incident and
that no further action was required. The maintenance history of all
valves of this type were reviewed by this inspector. Total failure to
operate has not been a problem with these valves. Maintenance attention
to these valves has been increased and improved since the incident. This
inspector also reviewed the complete history of stroke times for the two
valves that failed and the work requests that overhauled them after
failure. No degradation trends in stroke times were apparent, either
before or after the event. Although the inspector does not concur that
the f ailure to operate in this event can clearly be blamed on "sediment
buildup," the f ailure does appear to be an isolated case. This item is
closed.
(Closed) 327, 328/86-55 observation 6.15, Testing of 0.5 Second Time Delay
Relays. This item noted that the 0.5 second time delay relays on critical
safety systems component (CSSC) systems had not been calibrated. TVA had
transmitted to the NRC reportable occurrence report SQR0-50-327/871010 on
February 27, 1987, which stated that a QA audit had found that numerous
relays, switches and controllers had not been identified on any procedure
to require calibration. Revision 1 to Maintenance Instruction MI-13.1.3,
Setpoint Verification and Calibration for Time Delay Relays Associated
with Load Shedding logic, was issued on June 9,1987, and these relays
have now been calibrated. TVA has committed, in its response to NRC, to
incorporate the subject instrumentation into Sequoyah standard practice
SQE-8, Control of Installed permanent Process Instrumentation, by
December 31, 1987. The licensee's actions are satisfactory and this item
is closed.
(Closed) VIO 327, 328/86-68-05, paragraph c, section 2.4.4 deficiency
0-2.4-5, Loose Debris in Valves 2-FCV-1-17 and 2-FCV-1-18. NRC inspectors
reviewed the July 16, 1987 TVA response to this portion of the violation
and the corrective action taken. The loose debris in the limit switch
area of the valves was corrected under work plan 12305. Licensee action
on these items is adequate. This portion of the violation is closed.
(Closed) IFI 327, 328/86-48-05, Reportable Occurences. Ten potential
reportable occurrences (PR0s), identified by the licensee, with the status
of corrective actions were reviewed by the inspector. Five of the items
were reported as completed, the remaining five were reported as in
progress. The five items still in progress are:
Item 3 - MOV (CSSC) MOVATs testing
Item 7 - Pressurizer instrumentation
. .
.. __
_ _ _ _ _ _ _ .
.
28
Item 8 -1RHR switches
Item 9 - Backfill instrument sensing lines procedures t
'
Item 10 - Condenser vacuum exhaust flow monitor
The_ licensee assigned Sequoyah Activities List (SAL) tracking numbers to *
the above items as follows:
Item 3 - SAL-0317
Item 7 - SAL 0023' .
Item 8 - SAL 0699
Item 10 - SAL 0699
Item 9 was assigned a corporate commitment tracking system (CCTS) number
NCO 870114003 and NCO 870114005. All items are reported as complete for
Unit 2 except item 9 (NCO 870114003) which is being evaluated by TVA
licensing for satisfactory completion per the commitment. A third
commitment, evaluation of outgassing of sense lines for devices required
to operate during and after design basis accident (DBA) was not provided
for review. Discussions with TVA personnel indicated that the outgassing
~
evaluation was not part of this PRO concern. Based on the above, and
TVA's completion of the procedures required for backfill, this item is
closed.
(Closed) IFI 327,328/86-49-05, Apparent Deficiency of Adequate
Instrumentation to Monitor Cooling Water to the Emergency Diesel
Generators (EDGs). A valid emergency start, coincident with a blackout
could result in severe damage to the EDGs under certain operational ERCW
conditions (reliance on cross connect system). Sequoyah has taken steps
to reduce the risk by implementing attachment 1, page 3 of OSLA-30, dated
June 11, 1985. This attachment requires the EDG to be made inoperable if
normal emergency raw cooling water (ERCW) supply is removed from service.
It also requires the assigning of a dedicated operator for the EDG's, if
two (2) trains are made inoperable by the action. This corrective action
should eliminate damage to the EDGs, due to administrative action making
ERCW unavailable. A review of the FSAR did not locate a requirement for
flow measurement of ERCW to the EDGs. The generic issues associated with
this issue will be addressed under "compensatory measures for defeated
safety functions" during the NRC operational readiness review. This item
is closed.
(Closed) IFI 327, 328/84-43-06, Training For Reactor Vessel Head Vent
(RVHV) System. The training for the RVHV system was reviewed as part of a
, recent emergency operation procedures insepetion conducted November 2
through 5,1987. However, during a review of Sequoyah action list (SAL)
Ite'n 13, NRC inspectors noted that there were several discrepancies with
the operation of the RVHV system. The discrepancies were noted in the
performance of work plan (WP) 10597 which was written to perform post
modification testing (PMT) of the system. This PMT tested the operability
i of the RVHV system. During the performance of this PMI many exceptions
l and deficiencies were noted. The testing to prove the operability of the
system was conducted in October of 1983. All deficiencies except for one
l
- _ _ - _
!
,
29
was addressed by a department of nuclear engineering document entitled.
Interim Review and Approval of Post Modification Test Results, PMT-39,
dated June 6, 1986. The final discrepancy, a change to the controller to
provide more accurate valve position for-the head vent throttle valves, is
to be completed under ECN 5160 during - a future outage. The post
modification testing ECN 2777 along with workpackage 19597 was closed out
on a partial modification completion form which included an attached USQD
dated January 13, 1987. During the inspectors review of. the test
deficiencies and associated documentation, it was noted that during
testing when the block valve (2-FSV-68-395) was opened, both throttle
valves (2-FSV-68-396, 2-FSV-68-397) inadvertently opened to 6D% of full
~
open and then shut within about 5 seconds. This inadvertent opening was
noted during the resolution of deficiency DN-6 in work package 10597,
PMT-39, Appendix 0, Deficiencies and Exceptions. However, the inadvertent
opening did rot appear to have been made part of the deficiency reviewed
by DNE in either the PMT or USQD evaluation. A review of the functional
restoration guides (FR-I.3, revision 0, page 9 and 10, and FR-H.1, page
10) indicated that the procedures do not address the system- abnormality.
During discussions with the TVA staff, TVA indicated that the problem with -
this target rock valve was well known. However, when questioned by NRC
inspectors, some control room operators were not aware of the problem.
The NRC inspectors requested the licensee address the following concerns:
a. Why does the ACTION / EXPECTED RESPONSE column, step 19 of SQNP FR-I.3,
Unit 1 or 2, revision 0, page 9, and step 13 RESPONSE NOT OBTAINED,
SQNP FR-H.1, Unit 1 or 2, revision 1, page 10, not provide
precautions or describe the expected system response?
b. Provide documentation which indicates that the vendor has evaluated
the fact that the Target Rock throttle valves may "pop" open when
subjected to inlet pressure transients.
c. Verify that DNE agrees that the valve performance is acceptable and
documented by an unreviewed safety question determination (USQD).
d. Provide documentation that the problem has been addressed to the
operating crew.
Items (a) and (d) above, regarding the functional restoration guides
and training on valve response, was addressed by the recent NRC E0P
inspection. Item (b) above was addressed by the licensee by providing
the inspector with copies of Target Rock report #2866, Solenoid Valve
Response to Inlet Pressure Transient (December 17, 1980), along with
ASME publication 81-PVP-39 (April 1981). Item (c) above was addressed
by the licensee by providing the inspector a copy of the test deficiency
review and approval memorandum (D. W. Wilson to H. L. Abercrombie,
RIM B25 860606 012). .
The inspectors reviewed the above information and determined that items
(a), (b), and (d) were adequately addressed. However, item (c) above was
~
L
_ _ _ _ _ ___ _______
i
30
determined to be a violation of corrective action requirements, in that,
the department of nuclear engineering review of the test deficiencies
documented in deficiency report 2-PT-789 (dated April 12, 1984) did not
specify that emergency procedures should be changed and personnel trained
to cope with the described condition. Additionally, the evaluation of the
test deficiency did not reference the vendor's test report as
justification for accepting the test deficiency. This item, which involved
questionable performance of equipment is identified as a violation of the
10 CFR part 50, appendix B requirements for effective corrective action,
and is identified as violation 327, 328/87-65-01.
(Closed) IFI 328/86-62-04, Review of Final Revision of AI-19, part IV,
plant Modification After Licensing, To Ensure Time Commitments for ECN
Closecut and Drawing Update Is Established. During a review of the
transitional design change program conducted in November 1986, and
June 1987, (inspection reports 327, 328/86-62 and 327, 328/87-42) the
inspectors identified a weakness in the new program. Specifically, there
was no commitment on TVA's part to ensure timely closure of ECNs. This
issue was seen as a significant weakness, because the lack of timely
closure was determined by TVA and the NRC to be a contributor to past
design control problems. Additionally, the inspectors determined that
without timely closure of ECNs the requirements of 10 CFR 50.71, which
requires the annual update of the FSAR to be current within 6 months of
the modification, could not be assured. The licensee determined that it
would take approximately 6 months to complete the closure of the ECNs from
the point of field completion and operations acceptance of the
modification. This time commitment was established by revision 24 of
AI-19, part IV and appears to reflect the actual time needed by the
licensee to complete all reviews and closure effort. However, this 6 month
closure of the ECNs will not ensure that the requirements of 10 CFR 50.71
are satisfied. Specifically, TVA uses the closure of the ECN and not the
field completion of physical work, as the starting point of updating the
FSAR. This process could cause as much an 18 montF lag between the actual
plant configuration and the FSAR.
The above issue was discussed with the licensee and a commitment to start
the FSAR update process at field completion vs. ECN closeout was made by
the licensee. However, the licensee should have changed the FSAR update
process as part of the transitional design control program based on
corrective action report (CAR) 86-04-21 and the findings in inspection
report 327, 328/86-62. This lack of complete correct've action is
identified as a violation of 10 CFR part 50, appendix B, criterion XVI and
is a second example of violation 327, 328/87-65-01. IFI 328/86-62-04 is
closed.
13. Restart Test
a. SI-26.2B Loss of Offsite Power With Safety Injection D/G 2B-B
Containment Isolation Test, Revision 14, and STI-78, D/G 2B-B Load
Sequence Tes'., Revision 0. STI-78 was performed in conjunction with
51-26.28. N1C inspector > observed the performance of all of STi-78
_ _ _ _ _ _ _ _ _ _ _ _ - _
31
and test 1 and test 2 of SI-26.28. Test sequence 1 and test sequence
2 tested loading / shedding and safety injection / phase B actuation.
NRC intpectors attended the shift turnover briefing prior to the
testing. The of f going and on-coming shift supervisors discussed
plant rctivities for the previous shift, maintenance activities
perfo'.meri during the previous ibif t and the performance of the load
shnding, and the containment isolation test. NRC inspectors
6 tended a second briefinn for the on-coming operating crew for the
performance of the test. The briefing covered personnel assignments,
location of personnel, comrnur;ications, and the details of how the
timing would be conducted. The importance of good comsunication was
stressed by the shift supervisor. Each person responsible for
sections of the test was given highlighted copies of applicable
sections of the test. The following observations were noted during
the testing:
-
All pressurizer heaters were energized rather than just group B
and C heaters as specified by the test procedure. The assistant
test director told the unit operator to energize pressurizer
heaters. The unit operator energized all heaters and then
questioned the assistant test director as to whether he wantd
all heaters on. The lineup was promptly restored to the co' rect
lineup. The unit operator did not, like other test per:,onnel,
have a copy of the test procedure to check his pn cicular
actions. During subsequent steps it was noted that the unit
operator was careful in requesting information on what switches
and controls to operate by number. However, he was not given a
copy of the test. Having a copy of the procedure c>r repeating
back exact verbage of the step would have prevented the
occurrence.
-
During the initiation of the safety injection (SI) signal all
personnel were on nation to start recorders from a countdown
and "go" signal from the assistant test director. The assistant
test director did not check to see that the unit operator was
ready, consequently all recorders were started on the "go"
signal, but the safety injection signal was not activated. The
test was delayed to change paper in the recorders and rerun the
step. The test director was not directly overseeing this
portion of the test which may have contributed to the error.
-
During the tests, the unit operator secured an SI pump with the
SI signal still inserted. This resulted in the SI pump stopping
and immediately restarting. The assistant test director noted
the mistake and reminded the unit operator to place the pump in
the "lockout" position. The pump was secured due to isolation
l valve leakage that was increasing pressurizer level. Securing
the pump was not part of the test procedure, but was allowed by
precautions in the procedure to control plant parameters.
l
l
L
.. .
32
.
- Loads normally powered from vital inverter 1-4 was inadvertently
deenergized during the test when the 6.9KV shutd(wn board was
blacked out in step 6.2.8. During the shift turnover briefing,
it was discussed that maintenance was being performed on the 1-4
vital inverter. It was not brought out that the inverter was
powered from its maintenance supply (the 480V shutdown board)
5 nor that the maintenance supply would be deenergized during the
blackout of the 6.9 KV shutdown board. When power was lost, the
,
CO 2 fire doors for the 2B-B board room tripped. This caused
s
r considerable confusion in the control room as the cause was not
known. The test was stopped ano eventually it was determined
E that the vital inverter had been deenergized. The problen could
h have been prevented by a careful review of the effects of plant
maintenance on testing that was to be conducted.
i The following material deficiencies or potential material
jr deficiencies were noted during the testing:
a
f -
The 28-8 diesel exhaust fans did not restart when the diesel
-
started and returned powec to the 6.9KV shutdown boards. The
J probable cause was determined to be a f ailure of relay R1X2. A
work request was written to investigate and correct the cause of
the failure. The relay problem did not effect continuing the
test and was noted as a test deficiency.
-
During a review of the recorder data from the recorder connected
to the 6.9KV shutdown board, it was determined that the recorder
1 had stopped working during the SI activation test.
4 Investigation revealed that a lead had become loose causing the
malfunction. The malfunction resulted in having to rerun a
section of STI-78, Load Shedding. Steps 6.5.3 through step
r 6.5.6 had to be repeated. These steps pertained to the
simultaneous starting of a main fire pump and containment spray
_
pump on the 6.9KV shutdown board that was being supplied by the
,, 2B-B diesel generator. The inspector determined that the
- licensee's actions were in compliance with AI-47.
s
-
One procedure problem was noted with 51-26.28, step 6.2.2.2 in
f[w resetting the phase A isolation signal. The step read to reset
the phase A isolation signal using reset button HS-30-63D rather
than using HS-30-630 and HS-30-63E. The result was that some of
the relays in the following steps were not deenergized as
-
required by the procedure. The test was stopped while a
temporary change to the test procedure was obtained. The test
was recommenced after the temporary change was processed and
approved.
-
During the test the 2B-B diesel generator fuel oil transfer pump
did not restart as expected upon a low day tank level. The 2B-1
diesel generator fuel oil transfer pump was tagged out for
maintenance. Fuel oil to support the diesel during the test was
. . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
,
33
transferred manually. The licensee is investigating why the
fuel oil transfer operation did not function as expected. The
failure of the fuel oil transfer pump to start did not affect
the performance of the test.
The discrepancies identified during the performance of the above
. testing activities was discussed during the exit. meeting with
Sequoyah plant management. The NRC will further review these
discrepancies under unresolved item (URI) 327, 328/87-65-04.
b. Followup of open Sequoyah activity list (SAL) item 51 L-finger Gap
Adjustment (MI-11.2A). This SAL item addressed the resolution of
Limitorque valve maintenance regarding adjustment of limit switch
L-fingers. The inspector reviewed the limitorque vendor manuals
available in Sequoyah document control facility. No provisions or
reference to the process of measuring the gap or bending the
L-Fingers could be found in the manuals reviewed. The vendor manuals
do suggest that the geared limit switches should be replaced as an
assembly. The diagrams pictured show the L-fingers as part of the
assembly. Discussions with TVA personnel indicated that the vendor
is planning to include the measurement of the L-finger gap in future
maintenance instructions. TVA also presented a Commonwealth Edison
internal memo that, in part, states: the vendor. Limitorque, has
addressed the concern of insufficient contact pressure by inspection
of the L-finger gap and bending the L-finger to achieve a 1/32-inch
nominal gap. This issue is closed.
.
L.