ML20148K746
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| ML20148K746 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 03/22/1988 |
| From: | Bearden W, Brooks C, Christnot E, Ignatonis A, Andrea Johnson, Paulk G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20148K733 | List: |
| References | |
| 50-259-88-04, 50-259-88-4, 50-260-88-04, 50-260-88-4, 50-296-88-04, 50-296-88-4, NUDOCS 8803310292 | |
| Download: ML20148K746 (23) | |
See also: IR 05000259/1988004
Text
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UNITED STATES
NUCLEAR REGULATORY COMMISSION
[[p Mcg]o /p REGION il
e . o 101 MARIET TA STRE ET, N.W.
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Report Nos. 50-259/88-04, 50-260/88-04, 50-296/88-04
Licensee: Tennessee Valley Authority
6N 38A Lookout Place
1101 Market Street
Chattanooga, TN 37402-2801
Docket Nos. 50-259, 50-260, and 50-296
License Nos. OPR-33, DPR-52, and DPR-68
Facility Name: Browns Ferry Nuclear Plent
Inspection Conducted: February 1-29, 1988
Inspectors: Of[dh
G. L. Pfulk, Senior R(s'iderip Inspector
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Date Signed
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Date /$igned
C.R. Bro 6k.s,ResidehtIngector
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Date Signed
E. F. Chrthtnot, Resipent [nspector
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W. C. Bepirden, Resid(nt In{pectbr
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Date Signed
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D' ate Signed
A. H. Johtison, Project Engipeer
Approved by: $ b . ,~)h,4hdm ~
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A. J. Itifatchis'f 5ection 7 Chief, Da e Si e
Inspection Programs
TVA Projects Division
l SUMMARY
Scope: This routine inspection was in the areas of operational safety,
maintenance observation, surveillance testing observation, reportable
,
occurrences, previous enforcement action, restart test program, employee
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concerns, design controls and quality surveillance report reviews.
Results: Two violations were identified: (1) two examples of inadequate
procedures or failure to follow procedures for QA records preparation and
workplan control, (2) failure to correctly translate the design basis into
specifications and drawings.
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REPORT DETAILS
1. Licensee Employees Contacted:
C. C. Mason, Senior Manager, Operations Center
H. G. Pomrehn, Site Director
*J. G. Walker, Plant Manager
P. J. Speidel, Project Engineer
*J. D. Martin, Assistant to the Plant Manager
*R. M. McKeon, Operations Superintendent
J. S. Olsen, Superintendent - Units 1 and 3
T. F. Ziegler, Superintendent - Maintenance
*0. C. Mims, Manager - Technical Services Supervisor
J. G. Turner, Manager - Site Quality Assurance
M. J. May, Manager - Site Licensing
"J. A. Savage, Compliance Supervisor
A. W. Sorrell, Site Radiological Control Superintendent
R. M. Tuttle, Site Security Manager
*L. E. Retzer, Fire Protection Supervisor
*H. J. Kuhnert, Of fice of Nuclear Power, Site Representative
Other licensee employees contacted included licensed reactor operators,
auxiliary operators, craftsmen, technicians, public safety officers,
quality assurance, design and engineering personnel.
* Attended exit interview
2. Exit Interview (30703)
The inspection scope and findings were summarized on February 26 and
March 4, 1988 with the Plant Manager and Superintendents, and other
members of his staff. New items identified:
a. Violation (259.260,296/88-04-02) Two examples for failure to follow
procedures for QA records preparation and work plan control,
paragraphs 5.a and 11.
b. Violation (259,260,296/88-04-03) Failure to correctly translate
design requirements into drawings, paragraph 11.
c. Unresolved Item *(259,260,296/88-04-01) A problem with security
administrative controls and documentation for out of service
equipment, paragraph 5.a.
d. Inspector Followup Item (259,260,296/88-04-04) Single failure
criteria involving emergency core cooling systems. Identified as
part of restart test program.
*An Unresolved Item is a matter about which more information is
required to determine whether it is acceptable or may involve a
violation or deviation.
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e. Inspector Followup Item (259,260,296/88-04-05) Diesel Generator
field breaker sizing and excessive heat in DG panel. Identified as
part of restart test program,
f. Inspector Followup Item (259,260,296/88-04-06) Filter capacitors
missing from battery charger B.
The itcensee acknowledged the findings and took no exceptions. The
licensee did not identify as proprietary any of the materials provided to
or reviewed by the inspectors during this inspection.
Commissioners Carr and Bernthal made visits to the site during this
l report period. General discussions were held with TVA management and NRC l staff.
3. Licensee Action on Previous Enforcement Matters (92702)
(CLOSED) Violation (259,260,296/84-15-08) This violatior resulted from
inspector review of plant drawings. The violation stated that Revision 2
to TVA drawing 47W847-10 was found in the Technical Support Center (TSC).
However, the current revision at that time should have been Revision 3.
The out-of-date drawing was removed and the current revision of drawing
r7W847-10 placed in the TSC. All Drawing Control Center issue clerks have
been cautioned concerning the need for absolute accuracy in the drawing
fi'ing process. Additionally, the licensee is involved in an ongoing
prote",s of review and improvement of plant drawings including replacement
of r.xisting drawings with Configuration Control Drawings (CCDs).
The licensee has addressed the inspectors concern as stated in the
original inspection report and corrective action should be adequate to
preclude recurrence. This violation is closed.
4. Followup of Open Inspection Items (92701)
(CLOSED) Inspector Follow-Up Item (259/87-14-01) This item resulted from
an inspector followup of an event which occurred during the performance of
the monthly surveillance test on the 3 ED Diesel Generaict (DG). Failure
of the C phase fuse conta:ts in the DG exciter potential transformer
circuitry resulted in initiation of Standby Gas Treatment, refutling zone
isolation, control room emergency ventilation, Unit 3 half-sc am, and
primary containment isolation. The cause of the failure was the line
side of the fuse stab for the C phase becoming worn making little or
rio contact. This condition resulted in arcing across the contacts,
subsequent cable damage and phase-to-phase fault. Numerous components,
wiring and cabinet structures were damaged and the DG was secured and
made inoperable.
The licensee subsequently issued LER 87-08 which identified this failure
and outlined the following corrective action:
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Inspection of Control Cabinets of remaining 7 DGs
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All 4160 volt potential transformer fuse contacts in plant to be
inspected
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Maintenance procedures to be revised to require inspection and
maintenance of all 4160 volt fuse contacts on regular basis
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Based on the results of an engineering evaluation which deter-
mined that these fuses were unnecessary, the fuse and spring
finger contacts will be bypassed on all 8 DGs
The licensee's failure evaluation and assessment of generic ramifications
associated with this event are adequate, This item is being closed and
the implementation of the above correc;1ve action will be tracked as part
of the followup on LER 87-03. This item is closed.
(CLOSED) Unresolved Item (259,260,296/87-30-02) CREV Inoperable Due To
Excessive Flow. This item was opened when the inspector noted that the
licensee's original LER 87-14 was not adequate i .e. , did not explain
the problem with Control Room Emergency Ventilation or discuss safety
consequences. Also missing were dates and times the equipment was
inoperable.
Subsequent to this finding LER 259/87-14 Rev. I was issued. The inspector
reviewed the newer revision to the LER and found it adequate, addressing
the inspector concerns. Additionally, the inspector reviewed the
licensee's evaluation of effects on 10 CFR 50 Appendix A, accident
exposure limits which would have resulted from reduced filter efficiency.
The LER will remain open pending further review by the staff; however, the
inspector determined that the licensees evaluation was adequate for the
purpose of closing this item. This item is closed.
5. Operational Safety (71707, 71710)
The inspectors were kept informed of the overall plant status and any
significant safety matters related to plant operations. Daily discussions
were held with plant management and various members of the plant operating
staff. ,
The inspectors made routine visits to the control rooms when an inspector
was on site. Observations included instrument readings, setpoints and
recordings; status of operating systems; status and alignments of
emergency standby systems; onsite and of f site emergency power sources
available for automatic operation; purpose of temporary tags on equipment
controls and switches; annunciator alarm status; adherence to procedures;
adherence to limiting conditions for operations; nuclear instruments
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operable; temporary alter &tions in effect; daily journals and logs; stack
monitor recorder traces; and control room manning. This inspection
activity also included numerous informal discussions with operators and
their supervisors.
Generai plant tours were conducted on at least a weekly basis. Portions
of the turbine building, each reactor building and outside areas were
visited. Observations included valve positions and system alignment;
snubber and hanger conditions; containment isolation alignments;
instrument readings; housekeeping; proper power supply and breaker;
alignments; radiation area controls; tag controls on equipment; work
activities in progress; and radiation protection controls. Informal
discussions were held with selected plant personnel in their functional
areas during these tours.
In the course of the monthly activities, the inspectors included a review
of the licensee's physical security program. The performance of various
shifts of the security force was observed in the conduct of daily activi-
ties to include; protected and vital areas access controls, searching of
personnel, packages and vehicles, badge issuance and ret rieval, escorting
of visitors, patrols and compensatory posts. In addition, the inspectors
observed protected area lighting, protected and vital areas barrier
integrity.
a. Security Concerns
During a routine tour on February 15, 1988, the inspector inter-
viewed the secondary alarm station (SAS) watchstander regarding the
operability of his closed-circuit television (CCTV) display screens.
The screen for a certain can.e ra was blank but no maintenance
request (MR) sticker or deficiency tag was hung. The watchstander
was unable to determine why the camera was out of service so he
contacted the Central Alarm System (CAS). CAS personnel knew that
the camera was out of service but they did not know which program
controlled the work activity or when the camera could be expected to
be available. The inspector was able to trace the work to Engineer-
ing Change Notice (ECN) 286 and Workplan 0017-86. Work started in
July 1987, and the camera had been out of service since December 1987.
The workplan had been bogged down with problems for some time and had
recently been put on complete hold until a coordinating activity was
completed. This was expected to keep the camera out of service until
about June 1988. The inspector learned from the modification
engineer that the camera could be restored to an operable status by
simple tasks requiring only a few man-hours of work. After being
asked for a re-evaluation of the situation, the licensee immediately
implemented the necessary work and returned the camera to service on
February 19, 1988. In following up on this activity the inspector
categorized and reported his concerns to plant management as follows:
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(1) There was a lack of aggressive action to restore security
equipment to service. Equipment out of service necessitates
some compensatory action, The number of compensatory measures
on-site are excessive and each additional one adds to the
overall vulnerability of the facility. This was a finding by
the Regulatory Effectiveness Review (RER).
(2) There was a lack of understanding by some on shif t security
personnel of the status of security equipment. Although
appropriate compensatory measures were in effect for the camera,
at least one individual thought the camera was in service, one
individual thought the camera could be turned on if needed and
others didn't know the status of the camera at ell.
(3) There was a lack of attention to detail in completing the
programmatic paperwork associatecl with removal of security
equipment from service. Individuals requesting removal of
security equipment from se'"ice or degradation of a security
barrier are required to 4:i out Form BF-117 as described in
Standard Practice 11.5, Removal of Security Equipment From
Service, A portion of this form is used to document any
applicable compensatory measures and an approval signature is
required from the security shif t supervisor along with the time
and date approval is given. The BF-117 form for removal of a
certain camera did not have the required compensatory measures
indicated nor was any signature obtained. This was also the
case for approximately half of the active BF-117 forms reviewed.
Another abuse of the form was the expected duration block. This
block was listed as indefinite for the camera re-wiring work and
several other jobs in progress as well.
(4) There was a lack of feedback to the security shift supervisor
when jobs were completed which required removal of 'a security '
device from service or partial degradation of a security
feature. This resulted in several 8F-117 forms being maintained
in an active status long after completion of the job. As an
example, the BF-117 form for the opening of a vital area was
initiated by the project engineer for the recirculation system
safe-end replacement job. Authorization was requested for the
duration of the safe-end replacement; however, this job had been
finished for about nine months.
(5) An attitude existed among security shif t personnel such that
they "lived with" a recurring equipment deficiency as opposed to
pursuing a permanent repair. There was an additional attitude
among the maintenance personnel responsible for security
equipment such that they f ailed to believe a problem existed.
The inspector witnessed such a problem on a camera on
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February 15, 1988. The picture gradually lost focus until the
picture was lost and replaced by a pattern of bars for a brief
period of time and then the picture returned. This cycle repeated
itself several times while the inspector was interviewing the
SAS watchstander. The SAS watch was completely unaware of why
the camera performed in that manner. Watchstanders in the CAS;
however, were very familiar with the problem and indicated
that it was a common problem among some specific cameras not
restricted to just a certain camera. It was termed "crow-
barring" and was apparently caused by a problem with the
automatic iris adjustment. Maintenance personnel were initially
unfamiliar with a description of the problem and af ter a short
period of evaluation concluded that the problem only exhibited
itself during certain times of the day. This explanation was
rejected by the inspector since the problem was witnessed on an
overcast day in the morning. The inspector requested a further
evaluation of the problem.
It should be noted that throughout this review, appropriate compensatory
measures were being maintained when required. The problem was with the
administrative controls and documentation which were established in order
to assure appropriate control was maintained. The number and character-
ization of findings in this area indicate a definite potential for an
actual breakdown in compensatory measures to occur. These findings will
be tracked as an Unresolved Item (259,260,296/88-04-01) for failure to
adhere to security procedures pending a followup inspection by Region II
Security Inspectors.
One additional problem outside the security organizations responsibility
was noted during this review. The workplan control form (Form BF-62) for
workplan WP0017-86, Cable Pull and Camera Support for Permanent Power
Installation, was not properly completed. Step IV.B required a check on
whether any plant equipment is to be removed from service by the workplan
and Step XI required the Shif t Engineer to give permission to take any
equipment out of service. The workplan did not specify that the security
CCTV's would be taken out of service nor was the shift engineer's
permission obtained. This deficiency was discussed with the plant
management and modifications representatives as an example cf a violation
for f ailure to adhere to procedures (259,260,296/88-04-02).
b. Operator Logs
The quality of Reactor Operator logs was dis:ussed with plant
nanagement personnel on February 19, 1988. The need for continuing
oversight by operations supervisor > was stressed. Some logs continue
to have legibility problems and even logs that are legible are
sometimes unintrepretable even by personnel with a good understanding
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of plant equipment, programs, and procedures. Entries are made which
identify a procedure in progress by number without stating the title
of the procedure so that proper documentation of the activity is
made. Many abbr 2viations, acror.yms and initialisms are used without
a list of such a;. proved shortcuts. Carbon copies are maintained in
the control room logbook while original sheets are routed for review.
All of these factors make it difficult to interpret activities
performed during the shift without asking for a line-by-line
interpretation by the operator on-shift. Another recurring
deficiency was the use of a temporary "scratch pads" by the unit
operators. Contemporaneous log entries are not made at the time of
an occurrence. Operators make entries on a temporary scratch pad
during the shift and then at some point prior to shif t turnover
transfer these entries into the official log. This process lends
itself to potential abuse in that a temptation exists that embarras-
sing or sensitive entries may not be transferred to the legal
logbook. Guidance as contained in Standard Practice 12.24, Conduct
of Operations, was found to be confusing. In one portion the
Standard Practice reads:
"Lags are legal records and shall be kept in a neat, legible
manner. All entries shall be made at the time indicated on the
log. If any log readings are missed, the reason shall be stated
on the log. Write-overs, white-out, or erasures shall not be
allowed on any station logs. Mistakes shall be crossed out with
a single line, initialed and dated by the person maintaining the
log. All log entries, as well as other documentation,
signatures and initials shall be made in black ink only. All
log entries (Shif t Engineer, Unit Operator or Radwaste Logs)
must be written clearly, precisely and completely. The log
books are an official record and as such all entries should be
thought out as to understanding of an event and future reada-
bility by others. The records of events must be documented to
the fullest extent".
Another section of the Standard Practice; however, reads:
"Log Books, Charts, Turnover Sheets, etc. are official records
and mirror the conduct of the operations at Browns Ferry. They
must be maintained in a professional manner. Continuity of log
entries, specifically systems logged out of service and entries
identifying problems, must be maintained. To this end,
operators are encouraged to maintain "scratch pads" to maintain
records of out-of-service equipment and eturn to service for
subsequent entry into the logs. The scratch pads may be
destroyed after use. Formal logs are to be maintained as up to
date as practical during the shift."
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Section 4.1 of the Nuclear Quality Assurance Manual, Part III
contains the licensee's requirements for permanent QA records. l
Paragraph 6.1 states that written instructions that cover QA records l
preparation shall include requirements to ensure that QA records are ;
complete, legible, and in black ink or other permanent medium. An
exception is allowed to the permanent medium requirement which allows
some documents to be prepared in a nonpermanent medium. The document
must be converted to a p?rmanent medium prior to final approval and
the nonpermanent document must remain under the control and
responsibility of the supervisor who gives final approval of the
document. Standard Practice 12.24 does not contain any control ,
measures over the nonpermanent scratch pad. This is a violation of
10 CFR 50 Appendix B, Criterion V for failure to have an adequate
procedure for preparation of operating logs (259,260,296/88-04-02).
c. Posting of Notices and Information to Workers
The inspector verified posting of information as required by the
Nuclear Regulatory Commission was met in accordance with 10 CFR 19,
10 CFR 20 and 10 CFR 21. The Browns Ferry Site Director Standard
Practice 2.3 delineates posting requirements as specified in the Code
of Federal Regulations.
The following deficiencies were noted:
(1) The Notice to Employees (Attachment 1) of SDSP 2.3
references employees to use of the Nuclear Safety Review
Staff (NSRS) for employee concerns not able to be handled
through line management. This attachment should be updated
since the NSRS no longer exists.
(2) The annual summary of work injuries and illnesses was not
posted by February 1 as required by SDSP 2.3. Additionally
errors exist on the location where posted column of SDSP
2.3 in that new areas for posting have been selected
without updating the procedure.
(3) Documents required for pnsting by the 10 CFR 19, 20, and 21
cannot be examined at the location given in SDSP 2.3.
Corrections to the procedure should be made to correct
these location errors.
Upon notification by the inspector the licensee took corrective
action to address these deficiencies. I
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d. The inspectors audited a training course conducted by Westinghouse
for Operations Department personnel entitled "Conduct of Operations".
The goal of the class was to increase awareness of the importance of
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conducting Browns Ferry Station operations in an attentive, diligent
and conscientious manner. The concept of professionalism was
discussed throughout the session. Events in other industries l
including airlines, railroads, shipping and chemical plants were 1
assessed and operator errors which contributed to those events were
discussed. The Chernobyl event was discussed in detail. Barrier
analysis was used to analyze these events with emphasis on operator
performance in event mitigation. The course was very well received
by the attendees and was considered to be outstanding.
6. Maintenance Observation (62703)
Plant maintenance activities of selected safety-related systems and
components were observed / reviewed to ascertain that they were conducted in
accordance with requirements. The following items were considered during
this review: the limiting conditions for operations were met; activities !
were accomplished using approved procedures; functional testing and/or
calibrations were performed prior to returning components or system to
service; quality control records were maintained; activities were
accomplished by qualified personnel; parts and materials used were
properly certified; proper tagout clearance procedures were adhered to;
Technical Specification adherence; and radiological controls were
implemented as required.
l Maintenance requests were reviewed to determine status of outstanding jobs ,
and to assure that priority was assigned to safety-relat1d equipment t
maintenance which might affect plant safety. The inspectors observed the l
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below listed maintenance activities during this report period:
a. RHRSW Coupling Replacements -
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b. 3A Diesel Generator Yearly Inspection
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j In response to an inspector concern raised at an NRC daily management
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meeting on February 7, 1986, the licensee has completed an inspection and
evaluation of a structural crack noted in the plant off gas stack, TVA
engineers conducted a physical / visual survey of the off-gas stack's
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exterior and all accessible interior surfaces. The survey of the exterior
i surf aces (f rom the base to the first platform, first platform to the
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second platform and se on), was conducted or December 11, 1987, and the
accessible interior surfaces on December 14, 1987. The results/ findings
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of this survey are documented in BFEPC1 Calculation No. CD-Q0066-871856, '
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showing all turface cracks visually identified and located in plans and
developed elevations of the stack. ;
TVA's conclusion was that the cracks are not evidence of a structural
defect. The cracks are minimum (mostly hairline) ir, size, no concrete
spalling ar.d no Stain found in cracks to signify rusting of rebars. It
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was TVA's judgement that the most probable cause of these surface cracks
would be the temperature variation that the stack has been exposed to
since construction. Browns Ferry Engineering Project considers this
effort complete and the subject closed /dispositioned.
No violations or deviations were observed in this area.
7. Surveillance Testing Observation (61726)
The inspectors observed and/or reviewed the below listed surveillance
procedures. The inspection consisted of a review of the procedures for
technical adequacy, conformance to technical specifications, verification
of test instrument calibration, observation on the conduct of the test,
removal from service and return to service of the system, a review of te>t
data, limiting condition for operation met, testing accomplished by
qualified personnel, and that the surveillance was completed at the
required frequency.
During a review of Surveillance Instruction 4.7 F.3 for the SI upgrade
program, the licensee noticed that the corresponding technical specifica-
tion (4.7.F.2.6) required cold DOP for in-place leak testing of HEPA
filters. It was thought that the test method in use since 1976 was a hot
DOP test. As a result the licensee declared the Standby Gas Treatment
System inoperable as well as the Primary Containment Purge and Control
Room Emergency Ventilation Systems which have similar testing require-
ments. After further review the licensee now considers its test method to
comply with the cold D0P test requirement but no explicit definition of a
hot versus cold DOP test can be found. System operability was reinstated.
To avoid a future misunderstanding, the licensee intends to submit a
technical specification change to clarify the required testing method.
8. Quality Surveillance Report Reviews (40704)
The licensee program of quality surveillance survey results was reviewed
by the inspector to assure proper review adequacy, quality assurance
program satisfactorily implemented, and quality control program surveys
conducted in accordance with procedures. The following quality surveil-
lance section results were reviewed:
a. Restart Test Program - QBF-S-SS-0074
b. Poner Stores Package Searches - QBF-5-SS-0057
c. Qaality Requirements for 1/4 inch and smaller OD tubing -
Q3F-5-SS-0032
d. Purchase Orders / Specifications-QBF-5-SS-0086
e. Grouting and Dry - Packing of Base Plate and Joints - QBF-S-
-99-0063
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f. Radiological Emergency Plan Training: Operations Support Center
Staff - QBF-S-88-0063
g. NRC Commitment Verification - QBF-S-88-0067
h. NRC Bulletin 79-14 Phase II Walkdown-QBF-S-88-0076
1. Condenser Tube Pullout - (BF-S-88-0058
The surveys reviewed were thorough and noted significant deficiencies that
; required correction. The surveys adequately addressed the QA and
programmatic requirements. No concerns were noted,
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9. Reportable Occurrences (90712,92700)
The below listed licensee events reports (LERs) were reviewed to determine
if the information provided met NRC requirements, The determination
included: adequacy of event description, verification of compliance with
technical specifications and regulatory requirements, corrective action
taken, existence of potential generic problems, reporting requirements
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satisfied, and the relative safety significance of each event. The
following licensee event reports are closed:
LER No. Date Event
259/84-23 Rev. 1 5/18/84 Primary Containment
Isolation System ;
Initiation ,
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259/85-49 and Rev. I 6/23/85 Inoperability of Diesel
Generators Because of i
Seismically Unqualified
Battery Racks
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259/87-01 and Rev. 1 1/28/87 Personnel Errors Cause
Fire Watch Violations
259/87-12 and Rev. 1 5/6/87 Inadequate Maintenance
Procedure Cause Breaker
Failure to Close and
Enginected Safety
Actuation
259/87-23 8/26/87 Personnel Error Results
in Unrepresentative
Radiological Release
Assessment Data
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LER No. Date Event
259/87-29 11/5/87 Personnel Error in
Writing Equipment Tag-Out
Clearance Results in
Actuations of Engineered
Safety Features
296/87-04 10/13/87 Diesel Generator 3EB
Start Due to Personnel
Error During Maintenance
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296/87-05 10/19/87 Unplanned Diesel
Generator Start Due to
Relay Failure
296/87-06 11/24/87 Unplanned Engineering
Safety Features
Actuations Due to Relay
Failure and Personnel
Error During Corrective
Maintenance
During replacement of a failed relay (LER 259/84-23) an adjacent wire came
loose and caused primary containment isolation system i solations on
Unit 2. The wire had been incorrectly terminated. The wiring on Units 2
and 3 were inspected and Unit 3 was found to have similar problems, The
, incorrect termination on Unit 3 was determir.ed to be the result of two
field changes. The wiring was corrected through engineering change notice
P5135 and workplan 3048-84.
'
Ouring a post maintenance review of a maintenance request to repair the
diesel generator battery rack studs (LER 259/85-49) it was determined the
wrong stud material was used during the 1976 and 1980 installation of
studs for the new battery racks. All the diesel generator battery rack
studs were replaced with certified studs and seismic qualification was
restored by workplan 1224-85 and 3054-85.
A fire watch was not posted (LER 259/87-01) in the area of the diesel
generator building when portions of a fire protection system was isolated
and an individual (fire watch) reported to the wrong cable spreading room.
The plant fire protection unit de/ eloped a listing of areas requiring fire
watches and the reasons for them. The listing will be routinely updated
and maintained in the shift engineer's office. A fire protection engineer
his been placed on the call out list.
Rring transfer of a shutdown bus to its alternate power supply, the
alternate power supply failed to close the breaker, causing the engineered
safety feature actuation (LER 259/87-12). An inspection of the breaker
.;
found grease and dirt buildup on the control cell lintage which prevented
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proper breaker operation. The control cell linkages on the four kilovolt
plant breakers were inspected and cleaned. Procedure BF EMI-7, Mainte-
nance of Medium and Low Voltage Switchgear was revised to include cleaning
of suspect parts.
The composite portion for the month of August 1987, was inadvertently
disposed of which resulted in an unrepresentative quarterly sample for
radiological release assessment data for the third quarter of 1987 (LER
259/87-23). The surveillance instruction was revised to emphasize
labelling and storage of samples.
,
A fuse was removed from a panel in order to deenergize a primary contain- '
ment isolation valve modifications activities which resulted in engineered
safety features actuations (LER 259/87-29). The switching necessary to
electrically isolate the valve was inadequate during preparation of the
clearance. The individuals involved were cautioned and a critique of the
incident was reviewed by operations personnel.
On November 24, 19o7, control room personnel observed the B train of the
CREV system to be running from an auto start signal (LER 259/87-30).
Investigation by electrical maintenance personnel revealed that a burned
electrical relay caused the auto start. The relay coil was assessed as
an end-of-life failure failure and was replaced.
During the performance of maintenance on the 3EB diesel generator an
electrician accidently shorted the auto start relay, while connecting an
oscilloscope, which caused an auto start (LER 296/87-04). The personnel
involved were critiqued on the event and all electrical technicians
reviewed and signed the critique on November 30, 1987.
After completing a degraded logic test on a 4160 volt shutdown board the
3D diesel generator received an auto start signal (LER 296/87-05) an
investigation revealed two failed voltage relays that caused the auto
start signal which were monitoring the shutdown board. The failed relays l
were replaced and the logic test was rerun. '
On November 22, 1987, the Unit 3 inboard containment isolation valves were
automatically closed due to a failed relay coil in the primary containment
isolation system. On November 24, 1987, during replacement of the fsiled
relay coil the electrical jumper used was inadvertently dislodged by
maintenance personnel (LER 296/87-06). Corrective maintenance was
completed and the systems were returned to normal.
The following licensee event reports were reviewed and remain open pending ;
further review: I
LER No. DATE EVENT
259/85-51 9/25/85 Deteriorated Cable
in Reactor 1
Protection System '
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.-__-_ _______ ___ ___ ______________________ ___ _________________ _ __-_ _ ___________-_______ ___________________ _ _________________
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LER No. DATE EVENT
296/83-47 and Rev. 1 7/30/83 Seal Flow Through RHR
Seal Cooler Less Than
Required Minimum Flow Rate
296/86-08 9/5/86 Shorted Generator Coil
Reduces RHR Capability
No violations or deviations were observed in this area.
10. Employee Concerns ( Allegation) '
The inspector reviewed an employee concern involving soldering identified
,
as employee concern ECP-87-BF-897-01. This concern was initially
identified by a TVA instrument technician assigned to the Browns Ferry
(BFN) Instrument Shop and a copy of the concern was sent to the Region II
office in June 1987. The specific concern involved soldering performed by
a vendor on solid state electronic cards used in the neutron monitoring
system which initiates scrams and rod blocks. The inspector was informed
by BFN personnel that the electronic cards were received from General
Electric (GE) on site for a modification (mod) and the mod was installed
in 1983. Under a seven (7) power microscope the inspector observed an
electronic card similar to the cards used for the mod. The soldering
contacts on the pins of chips and operation amplifiers (0p Amps) appeared
contaminated, i.e. poor pre-soldering clearing and/or post soldering
cleaning, and discontinuing, i.e. sometimes referred to as "cold solder
joints", where there are gaps between the pins and the solder. The
technician informed the inspector that these problems were identified on
nine of eleven Tnermo Trip Cards and on nine of nine Direct Currect (DC)
amplifier cards, all used in Unit 2. The technician also stated that the
- I&C shop personnel were making repairs to the Thermo Trip Cards and DC
amplifier cards in the I&C shop until they were instructed to return all
cards to power stores. The employee concerns representative indicated
that the soldering specification may or may not have been stipulated in
the procurement process used for the modification. The employee refer-
enced several military specifications as possible soldering standards,
j This item will remain open and tracked as an allegation.
P
11. Design Control (37702)
On January 27, 1988, the licensee made a non-emergency report per 10 CFR
50.72 for an unanalyzed condition outside the design basis of the plant.
The finding concerned the seismic qualification of Emergency Equipment
Cooling Water (EECW) and Residual Heat Removal Service Water (RHRSW)
buried piping where the piping exits and enters building structures.
Appendix C, Section C.2.1 of the FSAR describes this feature as follows.
Class I buried piping, at penetrations into secondary containment and at
entry points into the intake structure, is protected from dif ferential
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movement of the soil and structure by a guard box and flexible joints.
The guard box is supporteu by, and moves with, the soil. One open end
butts against, but is n7t connected to, the building. Large pipes which
may be overstressed by the differential movement of the structure and the
soil-bearing end of the guard box are provided with two flexible
couplings. One coupling is located near the structure and one near the
soil-bearing end of the guard box. The guard pipe provides adequate
clearance to permit one joint to move with the structure and one with the
,
soil, without contacting the pipe.
Analyses of seismically-induced soil motions on Class I buried piping were
made, and the seismic stresses were determined to be small. Therefore,
differential movement at support points, at containment penetrations, and
at entry points into other structures is the primary concern in designing
buried piping at the Browns Ferry Nuclear Plant.
In October 1986, the licensee discovered that the function of some of the
flexible (Dresser) couplings had been defeated at some time in the past.
Tie-rod harnesses were installed across the couplings on the B RHRSW and ,
FECW piping at the intake structure. These htenesses were apparently
j installed in order to provide additional axial support for the coupling.
9 Since no drawings could be found which showed the harnesses, +.he Division ;
of Nuclear Engineering (DNE) was asked to evaluate the as-found condition ,
and provide the necessary design information in order to repair some of
< the tie-rods which were found damaged. DNE responded that the harnesses
; were not required and could either be repaired or removed at the option of
the plant maintenance organi:ation.
About a year later, during preparation for a hydrostatic test in
June 1987, similar damage was found on tie-rods and lugs on harnesses on
the A RHRSW couplings. The tie-rods were removed f rom these couplings
'
prior to the hydrostatic test. Later, while the system was being brought
>
up to hydrostatic test pressure, the coupling failed due to the excessive
axial load at about normal system pressure. This failure was aided by the
fact that a hanger (M-30) near the coupling had been previously removed.
It was later determined that this M-30 hanger was underdesigned for the
expected axial loading during a seismic event.
In December 1937, a Cono. tion Adverse to Quality Report (CAQR-871126) was
written which concluded that the original design intent as stated in the
FSAR had been defeated by the use of harnesses which did not preserve the
required flexibility of the Dresser coupling. As a result of the rigid
connection, relative movement between the building and buried piping as
would occur during a seismic event, would produce loads far greater than
the components were designed to withstand. A loss of both the EECW and
4
RHRSW systems could occur during a seismic event. One problem with this
4
CAQR noted by the inspector was that the check on potential af fect on l
) operability of the nuclear plant was marked "no" Justification for this
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conclusion was contained on the CAQR continuation sheet which simply
stated that it hao been determined that the adverse condition would not
prevent EECW/RHRSW from performing as designed; therefore, operability was
not affected. This was contradictory with the CAQR paragraphs.
The subsequent red phone report was prompted by maintenance and technical
support personnel review of the preliminary design change to the Dresser
coupling. The design was so radically different from the previous design
that a question of operability of the original design and as-found condi_
tion was raised. A high level management review team was assembled to
evaluate the ramifications of the findings and immediate work was started
to cut out eight of the Dresser couplings and replace them with rigid
welds. Although an in-depth analysis is still underway by the licensee,
the following deficiencies are currently known:
(1) Hanger M-30 which provides the only axial restraint for the couplings
in question was significantly underdesigned and could not withstand
the thrust load. An analysis determined that the hanger would bend
and deflect out of position during a seismic event. This would allow
the coupling (without the tie-rod harness) to spring apart and fail
the pressure boundary of the coupling.
(2) The original design function of the Dresser coupling was defeated at
some point by the installation of tie-rod harnesses. No design
analysis was performed to justify this modification. The harnesses
were probably installed in 1973 time frame in order to maintain the
coupling integrity during water hammer events which were common at
that time. A recent analysis showed that the harnesses were
underdesigned for the required load.
(3) The design evaluation of the as-found condition of the couplings in
October 1986 was erroneous. This evaluation, contained in a
memorandum from S. R. Lawson to R. H. Wall dated 10/30/86, concluded
that the M-30 hanger provided the necessary axial restraint that the
Dresser coupling lacked and therefore the tie-rod harnesses could be
eliminated.
The safety significance of these findings are still under evaluation by
the licensee. At a minimum, a seismic event would have resulted in a
degradation of the RHRSW and EECW available due to a loss of the coupling
integrity. At the most, a complete loss of the Ultimate Heat Sink (VHS)
would have resulted. This is a violation of the Design Control require-
ments of 10 CFR 50, Apper.di x B (259,260,296/88-04-03), Although the
problem was identified by the licensee it is considered a self-disclosing
violation in that a failure of the coupling had to occur during a
hydrostatic test to force full reconciliation of the deficiency. The
opportunity existed in October of 1986 to fully evaluate the problem
but a perfunctory analysis failed to correct the design deficiency.
_ _ _ _
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During a review of the documentation associated with this problem the
inspector noted that delinquent reviews were performed on the hydrostatic
test data on the RHRSW system. SI 3.3,13.A.2, ASME Section XI Hydrostatic
Pressure Testing of the RHRSW System Buried Piping, which was performed on
July 30,1987, did not pass the acceptance criteria due to the failed
Dresser coupling. The Shift Engineer did not review the data to determine
if an LCO was violated until November 25, 1987. The results were not
reviewed by the Mechanical Engineer until November 30, 1987, and the
Cognizant Engineer did not review the data until February 18, 1988. For
the next trial of SI 3.3.13.A.2 performed on October 31, 1987, most of the
reviews were more timely; however, the data sheet did not specify that the
acceptance crii.eria was not satisfied nor was the Shitt Engineer's
signature obtained to document that a review for an LCO violation had
occurred. These deficiencies were discussed with plant management as
further problems with attention to cetail and meticulous compliance with
procedures. Similar examples of nine delinquent reviews had been identi-
fied by the licensee and documented in CAQR 880102.
,
12. Restart Test Pregram
a. Restart Testing Status
The inspector attended RTP status meetings, reviewed RTP test
procedures, observed RTP tests and associated tests performances, and
reviewed selected RTP test results. The following specific RTP
activities and associated activities were monitored during this
reporting period:
(1) RTP-002, Condensate, The system was released for testing by
the Joint Test Group on February 23. A special test (ST-99
Condensate Demineralizer) is in progress which involves the
condensate polishers clean and precoat sequence and is being
,
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conducted under the chemistry departments responsibility.
(2) RTP-023, Residual Heat Removal Service Water (RHRSW), The system
'
is being impacted by the header outages due to the Dresser
couplings. Several Maintenance Requests, Hold Orders and Design
Change Notices are outstanding as well as hydrostatic tests
completions.
I (3) RTP-024, Raw Cooling Water (RCW), The system is being restrained
somewhat by repairs required to 2A and 2B RCW pumps, time delay
relays and various system valves. The system is also impacted
by the Emergency Equipment Cooling Water (System 067) header
outages.
I (4) RTP-030, Diesel Generator and Reactor Building Ventilation (DG &
1 RX BLDG VENT.), Section 5.1, DG Building Ventilation Flow
l
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Verification Units 1,2, & 3 was performed during this reporting
period; however, initial review of data indicates a possible
retest of some of the twenty-six fans involved in this section
of the test may be required.
(5) RTP-031, Control Building Heating Ventilation and Air
Conditioning (Cont. Bldg HVAC), The actual restart test
procedure is still in the draft stage. However, primary
activities are in process which includes repairs to a tear in
ductwork and the installation of ductwork, conduit and cable
seals. Both of these are being worked partly under Engineering
Change Notices 0031A and P0647 respectively. Restart test
procedure performance is scheduled to start on March 29.
(6) RTP-057-4, 480 Volt Distribution System (480 V Dist.), The
system is closely related to system 82 standby Diesel Generators
(DG) in that load shedding verification is performed in
conjunction with OG load acceptance test. Logic function tests
are performed to plant procedures such as SMI-1-48SD.
(7) RTP-057-5 4.16 KV Distribution System (4 KV DIST.), The system
is also cicsely related to the DGs and has been identified as
system necessary to support Loss of Power / Loss of Coolant
Accident (LOP /LOCA). A schedule for the performance of
procedures (Logic functions, etc.) indicates th4t the 4.16 KV
shutdown boards till be completed prior to the associated DG Low
Acceptance test. Plans are in place to perform a battery
discharge (ampere hour) test when the 4.16 KV Shutdown Board 3EB
is out of service for functional testing anC neintenance.
(8) RTP-057-7, 250 VOLT OC Shutdown Board Battery Chargers (250 VOC
S/0 Batt.), The system has received upper management attention
through the Restart Operations Center (War Room) and every
effcet is being made to complete the RTP r.o later than March 6.
The major hold ups for the test has been a lack of material
(filter capacitors) and craft support.
(9) RTP-065, Standby Gas Treatment (SGTS), The system is under test
not only to support LOP /LOCA, but also to establish secondary
containment for the fuel reconstitution. Several dampers have
not satisfied the time to close test specification, but also
some dampers have been disabled by linkages and motors removed
and did not function when initiation signals were present.
(10) RTP-067, Emergency Equipment Cooling Water (EECW), The system is
affected by the Dresser coupling issue due to header outages to
remove selected couplings. Preparations were being made to
hydrostatic test the system once the couplings were removed and
replaced by welds.
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(11) RTP-070, Reactor Building Closed Cooling Water (RBCCW), The
system is in the pre-test stage with ASME Section 11,
Hydrostatic tests in progress and Mechanical Testing group is
performing 10 CFR 50, Appendix "J" tests, i.e, local leak rate '
tests. The system is scheduled for restart test release the
first week of March,1988.
(12) RTP-075, Core Spray (CS), The system has successfully tested
section 5.7, local Operation of 2A, 2B, 2C and 20 core spray
"
pumps. Several ECNs, MRs and Surveillance Instruction updates
( are being processed to support completion of the test. ,
(13) RTP-082, Standby Diesel Generators (STDBY DG), Several load
acceptance tests on the eight (8) DGs have been performed. Two
, of Unit 3 DGs require degraded voltage tests and are scheduled
to be performed during the monthly sis. Additional special
testing involving the DGs speed governors and voltage regulators
are scheduled for March, 1988. This special test will involve
direct vendor input with DNE supplying the test methodology.
The RTP will be revised to reflect this test.
b. Design Deficiencies Identified By RTP !
,
(1) Systems 57-3, 250 Volt DC Distribution and 57-4, 480 Volt AC
Distribution
During a review of subsystem 280, Battery Boards and subsystem
231, 480 Volt AC shutdown boards by a system engineer in January +
of 1987, in preparation for the restart test the following was
discovered:
- (a) The loss of 250 Volt DC Battery Board #1 would cause a loss '
of 4S0 V toad shed logic signals to 480 V shutdown boards
1A & 18.
(b) The loss of Battery Board #1 would cause a loss of Safety
i Divison II core spray logic.
The above condition was determined by the licensee as a breach
in single failure criterta and documented by CAQR's. This item
is identified as an Inspector Followup Item (259,250,296/88- '
04-04).
i (2) System 82 Diesel Generators (DGs)
l
The DGs are designec for 3050 KW for 1/2 hour, 2950 KW for seven
(7) days and 2850 KW indefinite. The RTP calls for a 24 hour
j fuel consumption run with the first 2 hours 0 2950 KW and the
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4 next 22 hours 0 2850 KW. During the fuel consumption run the
field breaker for the generator tripped shortly af ter the test
was started while the DG was 0 2950 KW. This has been
attributed to the following:
(a) The field an.perage is undersized 0100 amps and a DCN No. l
3532 has been issued to upgrade the capacity to 125 amps. '
(b) The control cabinets where the breakers are located have
too high of an internal temperature and a DCN No. 3531 has
been issued to add ventilation louvers to the cabinets.
<
This item is identified as an Inspector Followup Item
(259,260,296/88-04-05).
. (3) System 32, Control Air System
During the test of drywell control air suction valves FCV-32-62
& 63, the cylinder operated valves failed to close on loss of
power to the solenoid valve and upon loss of control air as
,
required by the Safety Design Basis contained in the F3AR. This
item was previously identified in NRC Report (259,260,296/87-
-33).
'
(4) System 57-7, 250 Volt DC Shutdown Batteries
The filter capacitors in the battery charges did not pass the
ripple voltage tests. All capacitors are being changed and
as of the end of this reporting period, Charger A has new
,
capacitors, successfully passed the ripple voltage test and
' "
is back in service. The restart test and 5ystem engineering
personnel shifted to Charger B for the next ripple test.
1 Hewever, when maintenance removed the cov6r of the B Charger
the filter capacitors wert issine. This item is identified
: as an Inspector Followup Item ( ..,26. . 296/88-04-06). ;
c. Deficiencies Identified By RTP That Are Under Review
,
(1) System 82, Diesel Generators (DCs) ,
t
The RTP requirement is that a seven (7) day supply of diesel
fuel oil be available for three (3) DGs. The fuel oil transfer
pump that may be called upon is supplied from a IE source; i
however, it must be primed from service air, which is not a
systera important to safety.
(2) System 82 Diesel Generators (DG) i
,
During the paralleling of the Unit 1/2 DGs with the Unit 3 DGs l
the KVAR snaring was not present. It was initially determined i
that this may be a wiring error; however, further evaluation is
ongoing,
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(3) System 65, Standby Gas Treatment (SBGT)
During the performance of an ANSI required test the phase to
phase current readings for the relative humidity heater were
greater than the 5 percent relationship required by the ANSI
Standard. N510-1975, Section 14.2,3.
(4) System 23, Sump Level Switches
The RHR service water building sump level switches cannot be
adjusted to meet the high level pump st .*t requirement and the
requirement has no band.
}}