ML20056A478
| ML20056A478 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 07/26/1990 |
| From: | Harmon P, Little W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20056A476 | List: |
| References | |
| 50-327-90-22, 50-328-90-22, NUDOCS 9008080003 | |
| Download: ML20056A478 (23) | |
See also: IR 05000327/1990022
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L UNITED STATES 1
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NUCLEAR REGULATORY COMMISSION
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101 MARIETTA STREET.N.W.
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x Report Nos.: _50-327/90-22 and 50-328/90-22-
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Licensee': Tennessee Valley Authority
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6N.38A Lookout Place-
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1101_ Market Street
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Chattanooga, TN. 37402-2801
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lDocketNos.:
50-327'and 50-3281
License Nos.: DPR-77 and DPRa79
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Facility Name:
Sequoyah Units 1 and 2
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Inspection Conducted: June-6, 1990 thru July-5,- 6 0
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' Lead Inspector: LAAMA @
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Pf/ Harmon~, pi@ Rbsident Inspector
Date Signed
. Inspectors:
D. Loveless, Resident Inspector.
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J. Brady, Project Engineer
. Approved by:
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W. 5. Little, Chief, Projec't Section_1 -
Date Signed
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TVA Projects
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. SUMMARY'
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1 Scope:-
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-This> announced inspection involved inspection' effort .by, the Resident Inspectors
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in 1 the E area; of . operational safety verification? including _ control room
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, observations.Loperations performance, system lineups, radiation protection,
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safeguards, and conditions. adverse.to quality. LO.ther areas inspected included
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surveillance testing observations maintenance observations,, review of previous
inspection findings, foilow-up of events, review of, licensee identified items,_
- and revi_ew .of inspector follow-up items.
.Also included in this report is
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- documentation of special inspection findingc in'the area.of cable testing.
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< : Results:
iWea'knesses were noted in the licensee's control of overtime for plant personnel
' as described in_ paragraph 7, 'and' use of ' administrative control programs in
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response to unanticipated changes in core differential temperature.'
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The areas-of Operations, Maintenance,.and Surveillance were a_dequate and fully
capable ~ to supporti current plant operations.
The observ_ed activities of the
control room operators were professional and well executed.
9008080003 900726
ADOCK 05000327
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'One violation was identified which involved a failure to control overtime for
plant' personnel.
-327,328/90-22-01, Exceeding of Overtime Limits Without Proper Approval
- paragraph 7.
Two unresolved items * were identified pertaining to cable testing issues.
327,328/90-22-04, Unissued Calculation for 1E Cable Testing - paragraph
10.
327,328/90-22-05, Assumption Not Identified in Calculation - paragraph 10.
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One non-cited violation was identified involving a loss of a shutdown board
during surveillance testing.
327,328/90-22-03, Partial Blackout Caused By Energizing Circuit With
Testing Rig Still Connected - paragraph 9.c.
One. licensee-identified violation was identified involving an inadvertent entry
- into TS 3.0.3' and placing the plant in an unanalyzed condition during
containment purge operations.
327,328/90-22-02, Unit 1 in Unanalyzed Condition During Containment Purge
Operations - paragraph 9.b.
No deviations or inspector follow-up items were identified.
Three events that occurred during the inspection ; period are described in
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paragraph 9.
The events were a partial blackout and emergency diesel start due
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to improper testing of a shutdown board; inadvertent entry into TS 3.0.3 by
placing Unit 1 in an unanalyzed condition during containment purge operations;.
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and. identification and resolution of anomalies in core delta T identified
during the Unit 1 startup from the refueling outage.
Unresolved items are matters for which more information is required to
determine whether they are acceptable or may involve violations or
deviations.
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REPORT DETAILS
1.
Persons Contacted
Licensee Employees
- J. Bynum Vice President, Nuclear Power Production
W. Byrd, Manager, Project Controls / Fir.ancial Officer
C. Vondra, Plant Manager
- R. Beecken, Maintenance Manager
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L. Bryant, Work-Control Superintendent
- M. Burzynski, Site Licensing Manager
- M. Cooper, Compliance Licensing Manager
J. Gates, Technical Support Manager
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- W. Lagergren,-Jr., Operations Manager
- M. Lorek, Operations Superintendent
R. Lumpkin.-Site Quality Manager
- T. Flippo,, Quality Assurance / Quality Engineering Manager
R. Proffitt, Licensing Engineer
- H. Rogers, Technical Support Program Manager
- M.'Sullivan, Radiological Control Manager
P. Trudel, Project Engineer
C. Whittemore, Licensing Engineer
- G. Hipp, Licensing Engineer
- R. Thompson, Licensing Engineer
NRC Employees
- L. J. Watson,.Chitf, Project Section 1
- Attended exit interview.
Acronyms and initialisms used in this report are listed. in the last
paragraph.
2. -
Operational Safety Verification (71707)
a.
Control Room Observations
The inspectors conducted discussions with control room operators,
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verified that proper control room staffing was maintained, verified
that access to the control room was properly controlled, and that
operator attentiveness was commensurate with the plant cor#iguration
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and plant activities in progress, and with on-going contro' room.
operations.
The operators were observed adhering to approp iate,.
approved. procedures, including Emergency Operating Procedure 3, for
the.on-going activities. The inspectors-observed upper management in
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the control room on a number of occasions.
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The inspector verified that the licensee was operating the plant in a
normal plant configuration as required by T3 end. when abnormal
conditions existed, that the operators were complying with the
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appropriate LC0 action statements.
The inspector verified that RCS
leak rate calculations were -performed and- that leakage rates were
within the TS limits.
The inspectors observed instrumentation and recorder traces for
abnormalities and verified the status of selected control room
annunciators to ensure that control room. operators understood the
-status of the plant. ' Panel indications were reviewed for the nuclear
instruments, the emergency power. sources, the safety parameter .
display system and the radiation monitors to ensure operability'and
operation within TS limits.
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No violations or deviations were observed,
b.
Control Room Logs
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The inspectors observed control room operations and reviewed
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applicable logs including the shift logs, operating orders, night
order book, ' clearance hold order- book, and , configuration log to
obtain information concerning operating trends and activities.
The
TACF -log was reviewed to verify.that the use of jumpers and lifted
leads causing ' equipment to be inoperable was clearly. noted and
understood. No issues were identified with these specific logs.
Plant secondary chemistry reports were reviewed.
The inspector
verified that primary plant chemistry was within TS limits.
The implementation of the licensee's sampling program was observed.
Plant specific monitoring systems including seismic, meteorological
and fire detection indications were reviewed for operability.
A
review of surveillance records and tagout logs was performed to
confirm the operability of the RPS.
'No violations or deviations were observed.
c.
ECCS System Alignment
The' inspectors walked down accessible portions of the Unit 1 charging
and letdown portions of the CVCS System, to verify operability, flow
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path, heat sink, water supply, power supply, and proper' valve and
breaker alignment.
The inspectors verified that a selected portion of the containment
isolation lineup was correct.
No devietions or violations were identified.
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d.
Plant Tours
Tours of the diesel generator,- auxiliary, control, and turbine
buildings, and exterior areas were conducted to observe plant
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equipment conditions, potential fire hazards, control of ignition
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sources, fluid leaks, excessive vibrations, missile hazards and plant
housekeeping 'and cleanliness conditions.
The plant was observed to
be clean and in adequate condition.
The inspectors verified thet
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maintenance work orders had been submitted as required and that
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followup activities and prioritization of work was accomplished ty
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The inspector visually inspected the major components for leakage,
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proper lubrication, cooling ' water supply, and any general condition
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that might prevent fulfilling their functional requirements.
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The inspector observed shift turnovers and determined that necessary.
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information concerning the plant systems status was adoressed.
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No violations or deviations.were observed.
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Radiation Protection
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The inspectors observed HP practices and verified the implementation
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of radiation protection controls.
On a regular basis, RWPs were
reviewed and specific work activities were monitored to ensure the
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. activities were being conducted in accordance with the applicable
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RWPs.
Workers were observed for proper frisking upon exiting
contaminated areas and the radiologically controlled area.
Selected
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radiation protection instruments _were verified operable and
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calibration frequencies were reviewed.
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-No violations or deviations were identified,
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f.
Safeguards Inspection
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In the course of the monthly activities,- the inspectors included a
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review of the licensee's physical security program. The performance
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of various shifts of the security force.was observed in the conduct
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of daily activities including: protected and vital ' area access
controls; searching of personnel and packages; escorting of visitors;
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badge issuance and retrieval; and patrols and compensatory posts,
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The . inspectors observed protected area lighting, and protected and
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vital areas barrier integrity.
The inspectors verified interfaces
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between the security organization and both operations and
maintenance. Specifically, the Resident Inspectors:
witnessed firearms training and qualification
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interviewed individuals with security concerns
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visited central and secondary alarm station
verified protection of Safeguards Information
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verified onsite/offsite communication capabilities
No violations or deviations were identified.
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Conditions Adverse to Quality-
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The inspectors reviewed selected items to determine that the
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licensee's problem identification system as defined in STD 3.1.1,
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Corrective Actions, which supersedes AI-12, Corrective Action, was
functioning.
CAQR's1 were routinely reviewed for adequacy in
addressing a problem or event.
A sample of the following documents
were reviewed for adequate handling:
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Work Requests
Potential Reportable Occurrences
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Radiological Incident Reports
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Problem Reporting Documents
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Correct-on-the-Spot Documents
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Licensee Event Reports
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Of the items reviewed, each was found to have been identified by the
licensee with immediate corrective action in place.
For those issues
that ~ required long term corrective action the licensee was making
adequate progress.
One exception to this was identified in .
paragraph 10 relative to.1E cables.
Corrective actions' teken in response to the unanticipated core
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differential temperature were effective.
Details-are described in
paragraph 9.a.
No violations or deviations were observed,
ho trends were identified in the operational safety verification area.
General con'ditions in the plant were adequate.
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Radiation protection and security are adequate to continue two unit
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3.
Surveillance Observations and Review (61726)
Licensee activities were directly. observed / reviewed to ascertain that
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surveillance of safety-related systems and components was being conducted
in accordance with TS requirements'.
The inspectors verified that: testing
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performed in accordance with
adequate procedures; test instrumentation was calibrated; LCOs were met;
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test results met acceptance criteria and were reviewed by personnel other
.than the: individual directing the test; deficiencies were identified, as
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appropriate, and any deficiencies identified- during the testing were
properly reviewed and resolved by management personnel; and system
- restoration was adequate.
For completed tests, the inspector verified
that testing frequencies were met and tests were performed by qualified
individuals,
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During surveillance testing on June 25, 1990 at 9:21 a.m., with both units
at 100 percent power, an improper test sequence resulted in loss of the
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6.9 kv shutdown board 1A-A and the start. of all emergency diesel
generators.
The IA-A shutdown board was reenergized by the 1A-A EDG and
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loads were sequenced back onto the board. . Details of this event are
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described in paragraph 9.c.
The following activity was observed / reviewed with no deficiencies
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identified:
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DPS0-SMI-4C, Indicating Voltmeter Calibrations for 6900 Volt Shutdown
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Boards.
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No trends were-identified in the area of surveillance performance during
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this inspection period.
The area of surveillance scheduling and
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management was observed to be adequate.
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One non-cited violation in the area of surveillance activities -is
described in paragraph 9.c.
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4.
Monthly Maintenance Observations and Review (62703)
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Station maintenance activities on safety-related systems and components
were observed / reviewed to ascertain that they were conducted in accordance
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with approved procedures, regulatory guides, industry codes and standards,
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and in conformance with T.S.
The following items were considered during this review:
LCOs were met
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while components or systems were removed from service; redundant
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components were operable; approvals were obtained prior to initiating the
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work;: activities were accomplished using approved procedures and were
inspected' as applicable; procedures used were adequate to control the
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activity, troubleshooting - activities were controlled and the repair-
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records accurately reflected the activities, functional testing and/or
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calibrations were performed prior to returning. components or systems- to
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service, QC records were maintained, activities were accomplished by
qualified personnel, parts and materials used were ' properly certified,
radiological controls were implemented, QC hold points were established
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where required and were observed, fire prevention controls were
implemented, outside contractor force activities were controlled .in
accordance with the approved QA program, and housekeeping was actively
pursued.
The following work request was reviewed:
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WR C001716 -
Following an ABI initiated by performance of an SI,
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the licensee proceeded to search for the leak that
had been causing auxiliary building airborne
radiation problems for several months.
A leak was
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found on 2-FT-62-137, emergency boration flow
transmitter, causing 3800 MPC of noble gas in the
local area.
This leak only existed during certain
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evolutions and the Auxiliary Building ventilation
system caused quick dispersion.
Rusted bolts were
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removed from the fitnge and new bolts torqued to
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normal specifications. This isolated the leak and the
airborne zone was cleared the next day.
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Management Activities in Support of Plant Operations
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TVA management activities were reviewed on a daily basis by the
inspectors.
Resident Inspectors observed that planning, scheduling, work;
control and other management meetings were effective in controlling plant
activities.
First. line supervisors appear to be knowledgeable and
involved in the day to day activities of the plant.
During the inspection
period, management involvement in resolving the problems associated with
the observed core Differential -Temperature was conservative, and
effective.
The event was well managed with the exception of the failure
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to promptly initiate the approved problem reporting procedure as described
in paragraph 9 a.
First line supervisor involvement. in the field has.
been observed and appeared adequate.
6.
Site Quality Assurance Activities in Support of Operations (71707)
The inspectors discussed the overtime problems addressed below with site
QA.
Following monitoring activities QA issued a CAQR (SQQ900287) on
overtime.
The ' CAQR indicated that Technical Support, Maintenance,
Radiation Control,' and Modifications were not meeting the guidelines of
Al-30 Section 23.0.
The inspector reviewed additional data and determined
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that-Operations and Chemistry were also not meeting these guidelines. QA
was questioned'as to why their monitoring did not pick up these critical
organizations. . They stated that . only organizations that were not
implementing the' criteria at all were documented in the CAQR.
In
this. specific instance, QA did not appear to be aggressive in pursuing the
issue.
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NRC Inspector Follow-up Items,-Unresolved Items, Violations (92701, 92702)
(Closei.) URI- 327,328/90-20-01, ' Overtime Rules Inadequate and not
Implemented.
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The inspector reviewed Al-30, Nuclear Plant Conduct of Operations,
Section 23.0, Plant. Staff Overtime Limits.
These requirements were
reviewed in three areas from Generic Letter 82-12 as follows:
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Enough plant operating personnel should be employed to maintain-
adequate shif t' coverage without heavy use of overtime. The objective
is to have operating personnel work a normal 8-hour day, 40-hour week
while the plant is operating.
AI-30 states that rormal shifts will be 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> in length.
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effort will be made to eliminate 16 hour1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> shifts.
No 16 hour1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> shifts
will be approved unless the circumstances are extreme.
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Nuclear Power' Standard STD-2,1.7, Administration of- Overtime,
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Revision 0, interim. change 2, restates the above NRC guideline in its~
entirety in Section 2.1.3.
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2)
In the event that unforeseen problems require substantial amounts of
. overtime to be used, or during extended periods of shutdown for
refueling, major maintenance l or major plant modifications. on a
temporary basis, the following guidelines shall be followed:
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An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />
straight (excluding shift turnover time).
b)
An individual-should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />
in any 24-hour period, nor more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48-hour
period, nor more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any seven day period (all
excluding shift turnover time),
c)
A break of at least eight hours should be allowed between work
periods (including shift' turnover time).
d)
Except during extended shutdown periods, the use of overtime
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should be considered on an individual basis and not for the
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-entire staff on shift,
Recognizing that very -unusual circumstances may arise requiring
deviation from the -above guidelines, such deviation shall be
authorized by the plant manager or his deputy, or higher levels of
management.
AI-30 embraces items a through c of the above guidelines, and among
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others states that the Maintenance, Operations, Technical Support,
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Division of Nuclear Engineering, Radiation Control and Chemistry
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organizations are required to follow these guidelines. Deviations to
~ these guidelines require prior plant manager (or designee) verbal
approval, with the exception that the SOS may approve shifts greater
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than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with written notification to the Operations
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STD-2.1.7 restates the above NRC- guidelines in their entirety in
Section 2.1.3.A for the same staff as listed in AI-30. Additionally,
for deviations' to these guidelines the standard requires that' the
following criteria be met:
Very unusual circumstances exist.
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Significant reduction in the effectiveness of operating
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personnel would be highly unlikely.
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Document the reason for authorizing the deviation.
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3)
Procedures-a're encouraged that would allow licensed operators at the
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controls to be periodically relieved and assigned to other duties
away from the control board during their tours of duty.
AI-30 states that if an operator is required to work in excess of 12
continuous hours, their duties should be carefully selected.
It is-
preferable that they not be assigned any task that affects core
reactivity or could possibly endanger the safe operation of the
plant.
The inspector reviewed the requirements under item 1) above and fotermined
that while the operations organization is staffed at levels conensurate
with two unit operation, often the licensed operators are not utilized to
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the extent necessary to reduce excessive levels of overtime.
During the
recent Unit 1 outage operations personnel were pooled for overtime usage.
Thus the usual high levels of overtime seen during an outage were also
seen for operators running the operating unit.
From March 14 through
March 27,1990, while Unit 2 was running at 100% power the ASOS's were
working six or seven 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shifts per week as a result of planned
overtime for the entire SR0 staff.
Additionally, the inspector noted that overtime was used regularly during
times of_ both units 'in operation.
In a period from January 1 through
March 4,1990 a random selection of 72 man-weeks were reviewed and found
that 21.-included overtime in excess of normal shif t turnover.
Staffing
levels and normal usage of overtime were not reviewed for other plant
sections.
The _ inspector reviewed plant usage of the guidelines listed under item 2)-
.above.- Following questions asked by the inspector, site QA issued CAQR
SQQ900287 for " numerous ' cases of personnel exceed lng the overtime
(requirements) of Al-30 with no, documented authorization" in Technical
Support, Maintenance, Radiation Control,- and Modification. Therefore, the
inspector did not review these sections further.
Numerous examples were found in the Operations section .where' overtime
requirements - were exceeded without documentation or meeting' other
guidelines as required.
From March 12 through April 29, 1990, the
inspector noted- 14 examples of individuals working more than 72. hours
in .a seven day period without an AI-30 Attachment E being prepared.
A review of AI-30 Attachment E's dating from November 13, 1989 through
March 23, 1990 showed fourteen examples of prior verbal approval not being
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obtained and no ex'planation as to why prior verbal approval was not
obtained; six examples of sixteen hour shifts being authorized for reasons
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that appeared' to' be routine when the procedure requires extreme cases
only; and four cases where overtime in excess of the guidelines was
authorized without a documented reason per procedure.
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Additionally, a review :of payroll records for March 1990 showed a
Chemistry first-line supervisor worked greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> _ in a seven
day period without proper documentation or approval.
During tnis inspection period the inspector discussed a six hour-
turn-around time scheduled for an AUO. The SOS stated that the individual
had a normal- 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> turn-around scheduled.
However, he had worked 4
overtime hours that morning.
The shift clerk _had tried to reach him to.
tell him not to return until he had been off six hours, but had failed.
The inspector asked the SOS the significance of six hours when AI-30'said
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eight._ The SOS replied that this was a Union requirement, that the Al-30
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requirements were broken all the time, and that six hour turn around times
were a normal occurrence. When asked if management needed to be informed,
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the SOS stated that it was not necessary.
After further discussion, the
SOS called the Plant Manager who denied the six hour turn ~ around and
informed the'S0S that these were considered unaccepteble.
The inspector . reviewed the implementation of the guideline stated in
item 3) above and did not find evidence of any formal implementation.
Discussions with operators, senior operators and duty plant managers
showed that consideration of overtime for job assignment was not being
performed.
Individuals interviewed stated that they would consider it in
the future but .no formal method or program existed other than the
suggestion in Al-30._ The inspector identified four cases on May 1, 14, 22
and 24 respectively where Unit Operators on the operating unit'were
required to work; greater than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> without being transferred from lead
operator to B0P or.to the non-operating unit.
The exceeding of the Al-30 overtime limits without proper approval is
considered a violation of T.S. 6.8.1 for failure to follow procedure
and is designated as VIO. 327, 328/90-22-01.
This violation'is a repeat
of Violation 327,328/87-78-01 issued March 14,.1988.
Generic Letter 82-16 requested that plants submit a TS requiring
compliance with the overtime guidelines as stated in GL 82-12.
TVA did
not submit a technical specification at that time.
This position was
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accepted by the NRC .in the SER documenting review of GL 82-16.
The SER
stated that Al-30 procedural -limits are considered sufficient for Sequoyah
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to meet the_ requirements of the GL.
However, the SER stated that if the
staff. should identify any significant problems in the implementction of
this procedure, the staff would request TVA to propose that the madel TSs
on overtime in the GL be - added to the Sequoyah TSs.
Because of the
broad ~ scope and apparent lack of management attention for the overtime -
problems, this item is being forwarded to NRR with a recommendation .that -
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they request that TVA submit the standard TS as written in GL 82-16 for
addition to the Sequoyah TSs.
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'The followup of plant overtime problems will be addressed in
VIO 327,328/90-22-01 above; therefore, URI 327,328/90-20-01 is closed.
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(0 pen)URI 327,328/90-06-05, Resolution of SSONI Issues,
a.
The following . observations were noted during Inspection Report.
327,328/86-68- as referenced in the stated report: sections.
These
items were addressed in Inspection Report 327,328/90-06 as needing
4
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further NRC' review:
Section
Observation
Description
2.1.2.5:
0-2.1-1
Lack of controls .to properly
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identify and. segregate material storage.
areas.
2.5.1
0-2.5-1
Root cause of valve damage not
identified.
2.5.2
0-2.5-2
No evaluation of maintenance
deficiency.
.
2.5.3
0-2.5-3
Vendor instructions not included
in site work procedures.
2.5.6
0-2.5-4
Root causes of hardware failures
not identified.
.;
2.6'
0-2.6-1
QA/QC coverage of modification
work,
2.6
0-2.6-2
Two-party verification of
i
maintenance activities.
The above listed. items represent inspector observations at the time
of the inspection and are not regulatory requirements.
These were
addressed in an internal NRC memorandum dated May 31,1990 from
Pierson to Watson.
Given the breadth and depth of TVA's Sequoyah-
Nuclear Performance Plan, the causative' factors precipitating these
observations were determined to have been corrected.
These items
are, therefore, administratively closed,
b.
Observation 0-2.1-1, Material Control-Issues, wcs resolved during the
procurement-inspection documented in IR 327,328/88-07 which provides
justification for. closure of this observation,
c.-
IR .327,328/86-68, Section 2.3.5,
addressed deficiency 0-2.3-5,
Installation Different from Flow Diagram. This item was addressed in
'IR'327,328/87-40 and left open.
The inspector stated that, while
this deficiency was not used to support a violation, it was
identified in Section 2.3.4 of the Report (IR 327,328/86-68) as a
deficiency, as well as in Table I of the report.
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On June 12, 1990 TVA provided the Resident Inspector with a copy of
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Drawing Deviation #87 DD 2219 and the new drawing to correct this
situation.
D-2.3-5 is closed.
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d.
The inspector reviewed licensee action. on VIO 327,328/86-68-05,
Safety Related Equipment Was Not Installed in Accordance with the
Instructions, Procedures or Drawings, associated- with Deficiency
D-2.4-7, Resistors Not Properly Mounted to Terminal Blocks.
The
+
violation indicated that field inspection found that resistors added
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to the control circuits in Auxiliary Control Panel 2L-11A on Terminal
Block 3-30 were not mounted to the panel but were held in place only
by their jumper wires.
TVA admitted the violation stating that
improperly mounted resistors discovered during the inspection were
the result of a lack of design-approved installation criteria for
terminal _ blocks, ' resistors, and diodes.
TVA stated that the
resistors were' remounted in a, ~ edance with a revision to WP 12188
and TVA drawing 47A348-284.
... IR 327,328/87-40 the inspectors
reviewed the corrective actions for this deiiciency and identified no
problems.
The item was left open for verification of corrective
action to be accomplished during a future inspection. The inspector-
observed the mounting of the resistors in question and determined
that they were now properly mounted to the panel. Deficiency D-2,4-7
of VIO 327,328/86-68-05 is closed.
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e.
The inspector reviewed Unresolved item 327,328/86-68-14, documented
and tracked in IR 327,328/86-68 as U-2.4-1, Unclear Status of Freshly
Painted Welds.
The inspectors observed a number of welds on reworked
pipe supports which were freshly painted.
Although this indicated'-
that the welds were new, there was no identification of these joints
on the WP weld maps, nor were there any-'new weld documentation
records.
The inspection team was unable to determine from the
e
'
records whether these were new, undocumented welds, or old welds that.
,
had been repainted by the craft.
'
The licensee has contacted personnel that were associated with the
SSOMI inspection.
TVA personnel contacted had no knowledge of what
welds were involved and many. people that may have had knowledge of
this item have left Sequoyah and could not be contacted.
The
licensee stated that they were not aware of any new undocumented
,
Following this 1986 inspection TVA committed in Section III.8
of the Sequoyah Nuclear Performance Plan to evaluate the adequacy of
the TVA welding program for all of the TVA plants and the suitability
of welded structures and systems for service,
This program was reviewed, inspected and approved by the NRC in
NUREG-1232, Volume 2, Safety Evaluation Report on Tennessee Valley
'
Authority: Sequoyah N iclear Performance Plan, Section 3.5 dated
May 18, 1988, and its supplement dated February 3, 1989. ^ herefore,
.
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U-2.4-1(327,328/86-68-14) is considered closed.
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f.
The inspector reviewed the. item identified in IR 327,328/90-06 as
327,328/86-68-17,- referred to -in IR 327,328/86-68 as U-2.4-4,
Question on Concrete Expansion Anchor-Sizes.
WP 11400 (ECN L6322)
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modified or added-16 supports to seismically qualify ERCW piping in
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the Diesel Generator Building.
Eight supports were inspected and no
hardware discrepancies .were observed.
The expension anchor test
report for support 17A586-2-15 shows 1/2 inch diameter anchors were
tested and specified the appropriate test load for 1/2 inch anchors
'
as required.
However, 3/4 inch diameter anchors were installed and
the indicated actual test load was for 3/4 inch anchors.
.,
The inspector reviewed procedure G-32. Bolt Anchors Set In Hardened
Concrete.
Section -3.8 states that, unless specifically prohibited,
anchor substitution may be made if the load capacity of the
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substitute anchor equals or exceeds the load capacity of the called
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for anchor in both tension loading alone and shear loading alone.
TVA determined that these criteria were met in the above referenced
substitution, and that the testing met the requirements of M&Al-10,
Installation, Testing and Documentation of Anchors Set in Hardened
Concrete.
The ' inspector had no further questions.
U-2.4-4
(327,328/86-68-17) is considered closed.
The items remaining open and requiring closure before URI 327,328/90-06-05-
can be closed are as follows:
Item
86-68 #
Description and Resolution Required
86-68-05
D-2.4-2
Improperly documented work. This
item was denied by TVA, and the NRC denied
1
this denial in.IR 327,328/87-40.
Resolution
is still pending.-
86-68-05
D-2.4-15
Missing anchor bolt pull test
records.
Item- 2 of this deficiency is
pending inspection of new installation
inside containment.
TVA denied item 1 of
this deficiency, and the NRC denied this
denial in IR 327,328/87-40.
Resolution is
E
still pending.
86-68-11
U-2.1-5
Missing snubber test and procurement
documentation.
indicated
that
corrective actions are comple te.
The
inspector requested additional information
for closure during this inspection period.
86-68-13
U-2.3-1
Failure to include vendor instructions in
work procedures.
This item remains open
pending further TVA corrective action'.
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86-68-18
U-2.5-1
Effect of loose controi room electrical
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panel. doors on seismic ' qualification.
,
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is currently performing a calculation
related to this condition.
This item-
-;
remains open 'pending NRC review of the-
calculation.-
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The additional items addressed above from item 327,328/90-06-05 are still
under NRC review. Therefore, this item remains open.
,
(Closed) NCV 327,328/89-18-05, Failure to Supply Licensed Operators With
Updated TS Within the Appropriate Time Period
The licensee revised Al-30, Nuclear Plant Conduct of Operation, to address
t
the implementation of TS changes including emergency and exigent changes..
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These corrective actions are adequate. This NOV is closed.
8.
Licensee Event Report Followup (92700)
i
The following LERs wereLreviewed and closed. The inspector verified that:
reporting requirements had been met, causes had been identified,
corrective actions appeared appropriate, generic applicability had been
.
considered, the LER forms were completed, no unreviewed safety questions
were involved, and violations of regulations or Technical Specification
'
,
conditions had been identified.
UNIT 1
327/89-006
Auxiliary Building Fire Door Was Breached Without
Appropriate Compensatory Measures in Place Due To
Inadequate Training
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327/89-020
Failure To Perform A Proper Monthly Check For the
Reactor Vessel Level Instrumentation ' System. Upper-range
Indicators
327/89-035
Sequoyah Unit 1 Reactor Trip Because of ~a High-high.
Steam Generator Level
.
327/90-002
Main Control Room (MCR) Isolation and LC0 3.0.3
Entry Because Both MCR Air Intake Radiation Monitors Were
Inoperable As a Result of Accidental Bumping of Circuit
Breaker.
327/90-003
Two Inadvertent Containment Ventilation Isolations
Caused From Power Supply Failures of the Upper Containment
and Containment Purge Radiation Monitors
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9.
- EventFollow-up-(93702)
a.
On June 12 during power escalation following the refueling outage on
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Sequoyah Unit 1 an anomaly was not:J in the data taken for the-
-
primary / secondary calorimetric.
The. observed- core differential
temperature, delta T, was highcr than expected. This increased delta
,
.T resulted in a calculated value of less than the TS allowable for
J
RCS flow when the delta T Ws used to calculate flow. The licensee
observed this anomalous high delta T at each increased power level..
.
At 90% - power (based on toe- Hghest indicating delta T) the power
ascension was stopped.
Calorimetric data indicated that true core
power was substantially less, approximately 84%.-~Further analysis of
the apparent' discrepancy resulted in a conclusion that the RCS flow
had not changed from previous cycles, was in fact well above the TS
value, and the errors introduced in the flow / power calculations were
caused by the hot leg RTDs reading higher than actual core outlet
,
temperatures. The hot leg RTDs were checked using bridge networks to
verify that the 'new signal processing system, Eagle 21, was not
causing - the error. .
The RTDs were determined to be reading
accurately, and the Eagle 21 system was translating that information
correctly.
The . licensee concluded that the RTDs were not sampling a truly
representative version of the RCS flow in the hot legs. Although the
RTDs are arranged to sample the RCS water temperature at three points
around the' circumference of the pipe, TVA has concluded that the flow
in: the hot legs at the RTD locations' is experiencing a streaming
phenomenon. . The three RTDs, located at the 0,120, and- 240 degree
-locations around.the hot leg pipe, are sampling the water in a manner
that results in higher than actual temperatures.
This is not
occurring at the cold leg RTDs because of the mixing effect of the
RCS pumps prior to the cold leg RTDs. _The licensee removed the RTD
'
bypass manifolds for Unit 1 during the refueling outage.
.This
replaced the flow nozzles in the - hot leg- pipe with RTDs in
thermowells.
The flow nozzles _had multiple inlet holes which
probably resulted in a better bulk temperature reading than the
single point RTDs.
After discussing _the issue with NRC on June 13 and June 14, power was-
escalated to 98% (U1118), with the highest delta T indication
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calculated to be 104%.
Further increases in power would challenge
the turbine runback circuit, which occurs at approximately 106%
,
delta T on two of four (2/4) delta T channels.
Correction of the
delta T channels was accomplished by rescaling the delta T channels
to reflect the 100% delta T value for the 63 degree delta T condition
versus the 60 degree condition.
This was accomplished using TI-2.1,
Calorimetric Calculation for Unit 1, on June 14, 1990.
The technical resolution of this issue was reviewed and followed
closely by the NRC and problems with that resolution were not
identified.
However, the administrative processes designed for such
issue resolution were not taken by TVA until NRC requested the formal
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documentation of the evaluations, corrective actions, and problem
identification forms supporting this issue's resolution. A CAQR was
written detailing'the anomalous delta T condition after the condition
had already been corrected by rescaling of the delta T circuit for-
the ' new delta T.
TVA managers stated that a CAQR does - not
>
necessarily have to be initiated prior to resolution of_ the problem,
but did concede that the issue should have been reported using .the
form for reporting problems, i.e., Problem Evaluation Report (PER),
9
as specified in Site Standard Practice SSP 3.2.
TVA standard
~
STD 3.1.1, Corrective Actions, stipulates that corrective actions are
to ~ be determined and documented in the CAQR within 30 days of
,
discovery.
Resolution of this issue is an example of the use of
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administrative programs merely to document the identification and
resolution of problems, and not as .the process used to initiate and
control corrective actions after discovery.
Although no problems
were-identified' concerning the technical decisions in this instance,
identification of the problem and its resolution were not
accomplished using the corrective action program.
Rather, the
CAQR/PRD process was employed as an afterthought, primarily to
document the corrective actions arrived at outside the corrective
action program.. Similar instances of a reluctance to initiate the
administrative programs have occurred previously and includc such
issues as the failure to document the RHR pump-to-pump interaction
problem on a CAQR, and repeated instances of failure to initiate
Safety Evaluations-for identified deficiencies. Although violations
of NRC requirements are not identified in this instance, lack of a
commitment to aggressively use the proper administrative programs is.
considered a weakness,
b.
On June 28, 1990, at 10:35 p.m., a Unit i upper containment purge was
initiated in accordance with S01-30.2, Containment Purge System
Operation.
At 7:20 on June 29, the oncoming operating shift crew
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discovered that the purge was being ' performed without the
Environmental- ' Allowance Monitor -(EAM) disabled. as required by
Instruction Change Form (ICF) 90-0326 to the procedure.
The ICF
requires that the EAM is to be disabled during. purge operations to
change the SG level trip setpoint from the normal 13 percent to a
more conservative'19 percent setpoint.
The EAM automatically resets
this setpoint during accident conditions due to harsh environment
conditions inside containment that could cause the . indicated SG
levels to read non-conservatively.
Analysis by Westinghouse
determined that the automatic reset function could be delayed for
certain accidents including a feedline break while the purge system
is operating. Therefore, the more conservative-setting of 19 percent
should-be in effect while the purge system is operating. The change
to the purge system procedure was in the form of a temporary ICF
which had not been interfiled in the procedure.
When the operators determined that the EAM had not been disabled, all
SG level trip channels were declared inoperable until the EAM was
disabled and TS 3.0.3 was entered. After disebling the EAM, TS 3.0.3
was exited at 8:16 a.m. on June 29.
A DNE evaluation was initiated
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and the assessment concluded that during the time the level trip
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channels .were at the non-conservative setpoint of 13 percent cith
purge in effect
Unit I was in an unanalyzed condition.
This
determination was made at 1:10 p.m. on June 29, 1990, and 1-hour
notification made to the NRC at 1:40 p.m.
The : licensee's investigation determined that the operator did not
check the front of the controlled copy of the procedure to ascertain
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whether a temporary ICF was attached.-
The normal method of -
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. conducting the purge requires the operators to obtain a controlled
.;
copy of the procedure from the Technical Information Center (TIC).
which has copies of all ICFs interfiled in the procedure.
In this
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instance, a controlled copy from the control room was used which does
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not routinely interfile temporary ICFs, but simply attaches the 'ICF
to the front of the procedure.
Corrective actions included making the ICF permanent and interfiling
the ICF in the procedure.
The licensee is investigating the
possibility of a design change to disable the EAM at. all power
conditions above the point at which 'the reset function is needed,.
that is,'during plant startup while the SG level _ control system is in
.
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manual' control and level transients routinely occur. Once the level
control system is in automatic, there is no real need to lower the
trip setpoints to-the 13 percent'value.
Response to this event and the corrective actions taken were timely
and appropriate. .- This licensee identified violation is not being
cited because criteria specified in Section V.G.1 of the NRC
Enforcement Policy were satisfied.
.This item is designated as NCV
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327',328/90-22-02.
c.
During surveillance testing of control board voltmeters on June 25,
1990, an improper test sequence resulted in blown fuses in the .
'
shutdown board logic relays which deenergized the IA-A 6.9 kv board
and startup of_the emergency diesel generators.
The event occurred during performance of procedure DPS0-SMI-4C,
Indicating Voltmeter Calibrations for 6900' Volt Shutdown Boards. As
required by the' procedure, 'the bus voltmeter. selector switch was-
placed in the OFF position to electrically isolate the voltmeter for
testing and calibration.
A voltage source with a test meter- was
applied to check the installed voltmeter.
- After completing the.
calibration, the test requires a check of the voltmeter for upscale
,
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readings: on all three phases by placing the installed voltmeter back
in service and selecting each of the three phases in sequence. The
p
test director did not remove the test equipment from the circuit, but
'
instead only turned off the test meter used with the test equipment.
The installed voltmeter was returned to service with the test
I
equipment still connected and applying a test signal. This tied the
test voltage to the installed voltmeter circuit and overloaded the
entire bus potential transformer circuit because a short circuit was
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induced around the meter.
This blew two fuses which protect the
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voltmeter and' the IA-A logic panel relays', 'deenergizing the relays.
- Deenergizing the -logic relays causes. the bus to- trip free from the
normal supply and actuate the blackout relays and start the F.DGs.
'
All required loads were then ' sequenced on after the 1A-A EDG
reenergized the board.
'
After finding the cause of the-event, the operators replaced the
blown fuses, paralleled the board with the normal supply, then
,
' stopped the EDGs'.
The event was reported and investigated as.
,
required.
The cause of the event was determined to be a deficient
procedure used in the surveillance, DPS0-SMI-4C.
The procedure did
not specify when the test equipment was to be removed, and lift the-
removal up to the performer.
There was also no precaution statement
to alert the user to the possibility of causing a shutdown board
blackout.
Corrective actions included revising the procedure, reviewing similar
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procedures for deficiencies of this type, briefing of test personnel
,
on.the event, and submitting a Design Change Request for considering
,
a .eparate fuse indication circuit.
This ' violation is not' being cited because criteria . specified in
Section V.A of the NRC Enforcement Policy were satisfied. This item
a
will be tracked as NCV 327,328/90-22-_03.
10.
Electrical Cable Issues (37700);
During- a review of the electrical calculations which support the
licensee's March 28, 1990 submittal pertaining to their cable test
-program, it was revealed by the licensee that calculation SQN-CSS-009 for
selecting the worst case 1E conduits relative to pullbys, jaming, and '
lack of support of vertical cable runs .was never issued by TVA.
The
_ inspector verified that'the review and approval-blocks on the calculation'
_
were not signed.
The licensee stated in Enclosure 2 of the July 31, 1987
submittal titled, " Revised Cable Test Program," that the implementation of
the cable selection criteria would be in a calculation.
The licensee
1
stated to the inspector that the unapproved and unissued calculation
SQN-CSS-009 was the document that implemented that statement.
The _ licensee told the inspector that they were aware that calculation
SQN-CSS-009 was not approved as early as March of 1988. This calculation
implemented the conduit selection criteria for 1E cables to select the
.
worst case conduits for high-potential testing. The test results were the
basis,for NRC and TVA concluding prior to restart in 1988 that 1E cables
were properly installed in the plant and that_ the associated
safety-related equipment would perform its intended function.
Nuclear
Power Standard 3.1.1, Appendix B, states that one of the criteria for
preparing a CAQR is failure to comply with licensing commitments to NRC
where it is apparent the discrepancy directly affects quality.
Since the
adequacy of this calcuiation was part of the basis for justifying
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equipment operability, and since tne calculation was not reviewed and
approved by the licensee, the basis for concluding that the licensee
. tested the '" worst case" conduits is in question. The inspector concluded
that this issue affected quality.
The licensee stated that they do not
know why the calculation was not approved-and. issued.
The inspector was
told by the licensee that a CAQR had not been issued ~ the failure to
approve and issue this calculation.
This issue ~ is designated as URI-
327,328/90-22-04, Unisssued Calculation for IE Cable Testing, pending
further review of why this calculation was not issued.
During the review of Calculation SQN-CSS-031,- Electrical Cable
Installation Pull-by Calculations Comparison, the inspector-noted that the
,
conclusion was depeMent on an assumption which was not stated in the
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Calculation.
The assumption was that the cable reviewed using the
criteria developed from Watts Bar plus the cables tested from the original
criteria were representative of the population of IE caWas from
SQN-CSS-005.
The inspector did not find this assumption 'tated in the
calculation as required by NEP 3.1, Calculations. Attachment 4 to_NEP.3.1
states that assumptions shall be adequately identified and documented.
This . item is designated as URI- 327,328/90-22-05, Assumption Not Identified
in Calculation, pending further review of why this assumption was not
stated.
11. - Exit Interview (30703)
-- The inspection scope and findings were summarized on July 3,1990, with
those persons indicated in paragraph 1.
The Senior Resident Inspector.
-described the areas inspected and discussed in detail the inspection
findings listed below.
A weakness was also identified in the use of
administrative-control programs to correct conditions adverse to quality.
The licensee acknowledged the inspection findings and did not-identify as
proprietary any of the material reviewed by the inspectors during the
inspection.
Inspection Findings:
One violation was identified which involved a failure to control overtime
for plant personnel.
327,328/90-22-01, Exceeding of Overtime Limits Without Proper
Approval - paragraph 7.
Two unresolved items were identified pertaining to cable testing issues.
327,328/90-22-04, Unissued Calculation for 1E Cable Testing -
paragraph 10.
327,328/90-22-05, Assumption Not Identified in Calculation
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paragraph 10.
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One non-cited violation was identified , involving a loss of a shutdown
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board'during surveillance testing.
327,328/90-22-03, Partial Blackout Caused By Energizing Circuit With
Testing Rig Still Connected - paragraph 9.c.
One licensee-identified violation was identified involving an inadvertent-
entry into TS 3.0.3 and placing the plant in an unanalyzed condition
)
.
'during containment purge operations.
327,328/90-22-02, Unit 1 'in Unanalyzed Condition During Containment
Purge Operations - paragraph 9.b.
No deviations or inspector follow-up items were identified.
During the reporting period, frequent discussions were held with the
Interim Site Director, Plant Manager and other managers concerning
inspection findings.
12. List of Acronyms and Initialisms
ABGTS-
Auxiliary Building Gas Treatment System
ABI' -
Auxiliary Building Isolation
ABSCE-
Auxiliary Building Secondary Containment Enclosure
-
Administrative Instruction
AI
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Abnormal Operating Instruction
A01
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.
Auxiliary Unit Operator
AVO
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'ASOS -
Assistant Shift Operating Supervisor
ASTM -
American Society of Testing and Materials
Boron Injection Tank
BIT
.
-
Browns Ferry Nuclear Plant
-
Control and Auxiliary Buildings
C&A
--
CAQR -
Conditions Adverse to Quality Report
Component Cooling Water System
-
Centrifugal Charging Pump
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CCTS -
Corporate Commitment Tracking System
Code of Federal Regulations-
CFR
-
COPS -
Cold Overpressure Protection System
4
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CSSC -
Critical Structures, Systems and Components
CVCS -
Chemical and Volume Control System
CVI- -
Containment Ventilation Isolation
Direct Current
DC-
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Design Change Notice
DCN
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Diesel Generator
-
Division of Nuclear Engineering
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Engineering Change Notice
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ECCS -
Emergency Core Ccoling System
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Emergency Instructions
El
-
Emergency Notification System
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Emergency Operating Procedure
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E0P'
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Emergency Operating Instruction
'
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ERCW -
Essential Raw Cooling Water
'
Engineered-Safety Feature
-
Flow Control Valve
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FSAR -
Final Safety Analysis Report
GDC .-
General Design Criteria
General Operating Instruction
G01
-
Generic Letter
GL
-
U
HVAC -
Heating Ventilation and Air Conditioning-
Hic -
Hand-operated Indicating Controller
'
H0
Hold Order
-
,
Health Physics
-
Instruction Change Form
ICF
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Independent Design Inspection
IDI
-
NRC Information Notic.e
'IN
-
Inspector Followup Item
j
IFI
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IM
Instrument Maintenance
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Instrument Maintenance Instruction
I
IMI
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Inspection Report
1R
-
Kilovolt-Amp
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KVA
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f;
KW
Kilowatt
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,
Kilovolt
i
KV
-
Licensee Event Report
i
LER
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LC0 -
Limiting Condition for Operation
j
Licensee Identified Violation-
=
LIV
-
LOCA -
Loss of Coolant Accident
MCR= -
Main Control Room
I
Maintenance Instruction
i
- MI
-
Maintenance Report
I
--
-MSIV -
Main. Steam Isolation Valve
NB-
NRC Bulletin
-
-
NQAM --
Nuclear Quality Assurance Manual
i
NRC
Nuclear Regulatory Commission
-
OSLA -
Operations-Section Letter - Administrative
'
OSLT -
Operations Section Letter - Training
4
Office of Special Projects
- 0SP
-
Precautions, Limitations, and Setpoints
Preventive Maintenance
l
-
PPM' -
Parts Per Million
l
PMT-
Post Modification Test
-
PORC -
Plant Operations Review Conmittee
4
P0RS -
Plant Operation Review Staff
Problem Reporting Document
-
Patentially Reportable Occurrence
PRO
-
41ality Assurance
-
Quality Control
-
I
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Radiation Control Area
-
RCDT -
Reactor Coolant Drain Tank
Reactor Coolant Pump
-
-
Regulatory Guide
-
-
Radiation Monitor
-
Reactor Operator
R0
-
.'
Rod Position Indication
-
Revolutions Per Minute
-
Resistivity Temperature Device Detector
-
Radiation Work Permit
-
RWST -
Refueling Water Storage Tank
Safety Evaluation Report
-
-
Surveillance Instruction
-
Special Maintenance Instruction
SMI
-
,
System Operating Instructions
501
-
Shift 0)erating Supervisor
SOS
-
Sequoyal Standard Practice Maintenance
SQM
-
SQRT -
Seismic Qualification Review Team
Nrvallance Requirements
SR
-
Senior Reactor Operator
590
-
S!0MI-
Safety Systems Outage Modification Inspection
SSIE -
Safety System Quality Evaluation
SSIS -
Solid State Protection System
-
ST1 -
Special Test Instruction
TACF -
Temporary Alteration Control Form
TAVE -
Average Reactor Coolant Temperature
TDAFW- --
Turbine Driven Auxiliary Feedwater
T1
Technical Instruction
-
TREF -
Reference Temperature
TROI -
Tracking Open items
Technical Specifications
TS
-
Tennessee Valley Authority
-
UHI
Upper Head Injection
-
Unit Operator
!
UO
-
Unresolved Item
-
,
USQD -
Unreviewed Safety Question Determination
i
Volts Direct Current
i
VDC
-
Volts Alternating Current
'
VAC
-
Work Control Group
.L
WCG
-
Work Plan
WP
-
.WR
Work Request
-
-
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