ML20245J053
| ML20245J053 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 02/01/1989 |
| From: | Carpenter D, Little W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20245J040 | List: |
| References | |
| TASK-2.F.2, TASK-TM 50-259-88-32, 50-260-88-32, 50-296-88-32, GL-84-23, IEB-84-02, IEB-84-2, IEB-85-003, IEB-85-3, IEB-88-003, IEB-88-004, IEB-88-3, IEB-88-4, NUDOCS 8903060147 | |
| Download: ML20245J053 (36) | |
See also: IR 05000259/1988032
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UNITED STATES
. ',. F e**cy,'h
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, NUCLEAR REGULATORY COMMISSION
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101 MARIETTA STREET. N W
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Report Nos.: 50-259/88-32, 50-260/88-32, and 50-296/88-32
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.: DPR-33, OPR-57,
and DPR-68
Facility Name: Browns Ferry 1, 2, and 3
anspection at Browns Ferry Site near Decatur, Alabama
Inspection Conducted: October 1 - 31, 1988 .
Inspector-
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'U.'R. Carpeerter, NRC Site Manager _
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DstV51gned
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Accompanied by: C. Brooks, Resident Inspector
E. Christnot, Resident Inspector
W. Bearden, Resident Inspector
A. Johnson, Project Engineer
J. York, Senior Resident Inspector, Bellefonte
A. Ignatonis, Technical Assistant, Inspection Programs
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Approved by: [ j f
W. 5. Littl#, Section Chief, UK,Y1gned
Inspection Programs,
TVA Projects Division
SUMMARY
Scope: This routine resident inspection included the areas of operationa'
safety verification, surveillance observation, modifications, system
return to service, reoortable occurrences, restart test. program,
followup of NRC Bulletins, followuo of open inspection items, and
licensee action on previnus enforcement matters.
Resul ts : One violation with two examples, were identified:
259, 260/88-32-01, Failure to follow procedures while tagging out
components for maintenance - Example 1 (Paragraph 2.b.)
259, 260/88-32-01, Failure to follow procedures while performre i
surveillance test - Example 2 (Paragraph 3.b.)
8903060147 890217 1
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One inspector followup item (IFI) was identified:
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259,1260, 296/88-32-02; Review of system
82, diesel Generator,
RTP results. (Paragraph'7.b.)
The. violation, and the IFI described above ~are required :to 'be
resolved prior to the restart of Unit'2.
The- program for system return to service has not totally' met the
expectation of the NRC with regard to' meticulous attention to detail
and thoroughness of open item resolution. - Followup . inspection
activity will be performed during future NRC resident inspections.
(Refer to paragraph 5 for details.) .
In - paragraph 5.c. , a concern is documented regarding operator access
to local control panels. The NRC considers the: licensee's response
to this concern to be thorough, prompt, and we11' directed.
Paragraph 2.a. documents examples of minor administrative errors
discovered in the Temporary Alteration Control Program. These
omissions, although' not constituting actual uncontrolled temporary
alterations, are examples of the type of errors that could occur if a
large future. backlog were allowed to redevelop. The NRC' inspector
noted that the licensee had continued to_make progress in;the ongoing
program to continue to reduce the' current backlog.
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REPORT DETAILS
1. Licensee Employees Contacted:
S. White, Senior Vice President, Nuclear Power
C. Fox, Vice President and Nuclear Technical Director
- J. Bynum, Vice President, Nuclear Power Production
- C. Mason, Acting Site Director
- G. Campbell, Plant Manager.
H. Bounds, Project Engineer
s- *R. McKeon, Assistant to the Plant Manager
- J. Hutton, Operations Superintendent
- R. Laverne, Maintenance S0perintendent
"D. Mims, Technical Services Supervisor
G. Turner, Site Quality Assurance Manager
- P. Carier, Site Licensing Manager
- J. Savage, Compliance Supervisor
A. Sorrell, Site Radiological Control Superintendent
R. Tuttle, Site Security Manager
L. Rett.er, Fire Protection Supervisor
H. Kuhnert, Office of Nuclear Power, Site Representative
T. Valenzano, Restart Director
Other licensee employees or contractors contacted included licensed
reactor operators, auxiliary operators, craftsmen, technicians, and public
safety officers; and quality assurance, design, and engineering personnel.
NRC Exit Interview Attendees
- D. Carpenter
- E. Christnot
- C. Brooks
- W. Bearden
- A. Johnson
- Attended exit interview
Acronyms and initialisms used throughout this report are listed in the
last paragraph.
2. Operational Safety Verification (71707)
The NRC 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. Inspection
observations included instrument readings, setpoints and recordings;
status of operating systems; status and alignments of emergency standby
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+ systems; onsite and offsite emergency power sources available for auto-
- matic operation; purpose of temporary tags on equipment controlsf and
switches; annunciator alarm status; adherence to procedures; adherence to:
limiting conditions for operations; nuclear instrument operability;
temporary alterations in effect; daily journals and logs; stack monitor
recorder traces; and control room staffing. This inspection activity also
included numerous discussions with operators and supervisors.
Ongoing - general plant tours were conducted. Portions of the turbine
buildings, each reactor building, and general plant 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. Discussions were held with selected plant
personnel in their functional areas during these tours.
a. . Temporary Alteration Control
An NRC inspector reviewed the temporary alteration change form (TACF)
file' located in the main control room area, and noted that.the number
- of open and outstanding- Unit 2 and. common TACFs was continuing to
decrease in accordance with the schedule published by the licensee as
part of an ongoing management program. However, the NRC inspector
noted that the three TACFs listed below had an indeterminate status.
Although the TACFs were no longer present in the' Unit 2 TACF file and
the inspector believed them to be inactive, no closure dates were
entered in the closure column of the TACF index.
TACF Subject
2-83-029 Steam packing exhauster
2-83-030 HPCI steam packing exhauster bypass
2-83-031 Test gauge on FE-32-75 to monitor leak
in drywell control air system
The inspector brought this issue to the attention of licensee manage-
ment. In response, the licensee ';erified from other documentation
and by actual hardware walkdowns that the TACFs in question were no
longer installed in the plant. The Unit 2 TACF index was
subsequently updated to reflect that the TACFs were closed. After.
further evaluation, the licensee determined that the missing dates
were an oversight and that each of the TACFs had been closed out
during 1983. The omission had not been detected earlier because all
outstanding open TACFs (pre-1984) were closed in 1984 and reissued
under a new numbering system which included a new index. A large i
number of open TACFs existed at that time and the pre-1984 portion of
the index contained many entries identifying TACFs which had been
closed and reissued under new numbers. The NRC inspector considered
' that the omission. was an isolated occurrence with no safety
significance. However, it was recommended that the licensee review
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all TACF' files.to verify that no other similar omissions. existed that-
would constitute an: uncontrolled temporary Lalteration- to' the
. facility. The licensee agreed to perform an audit to verify that no
other problems existed. This issue will be followed.up during future
resident inspector coverage.
b. Inadvertent Removal From Service Of Wrong Component
On October 16, 1988, diesel generators "A", "B" and "D" for Units 1
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and 2 were out of service for maintenance. One'of the' air compres-
sors ("B" compressor) supplying starting air for the operable "C" DG
was out of service. Two EECW pumps were operable as required by
Technical Specifications (TS) supplying cooling water to the eight
Units 1, 2 and 3 DGs.'
On October 16, 1988, the reactor operators were directed to tag out
the "A" Diesel Air Start System air compressor for the Units 1 and 2
diesel generator "B". However, the "A" air compressor for the Unit 1
and 2 (DG) "C" was tagg?d out instead. Independent verification of
the tagging was not performed and the fact that the compressor had-
been tagged out on the wrong D3 was not detected until an alarm was
received which indicated that the Units 1 and 2 DG "C" had low
starting air pressure. This resulted in DG "C" . being . technically.
inoperable according to Technical Specifications. Loss of the four
DGs resulted in EECW pump B-3 being TS inoperable. With the B-3 pump
inoperable, only one EECW pump, D3, remained ' TS operable. In this
configuration, with only one EECW pump operable : rather than two
required by TS, all eight diesel generators became .TS -inoperable.
With no fuel in the vessel or fuel handling in progress, TSs did
not. require any operable DGs. SDSP 14.9, ' Equipment. Clearance
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Procedure, requires component positioning and tagging to be ' as
provided on the clearance sheet. SDSP 14.9 and SDSP 3.15, Indepen-
dent Verification, require an independent . verification of the
position of the component. The independent verification must be
completely separate and independent of the initial alignment,
installation, or verification. SDSP 3.15, Attachment A, Systems and
Components Requiring Independent Verification, lists System 86,
Diesel Air Start System, as requiring independent verification. The
failure to tag the correct component and perform an independent
verification when tagging out the Diesel Air Start System air com-
pressor was a violation of TS 6.8.1 for failure to follow SDSP 3.15
and SDSP 14.9, and was identified as the first example of Violation
(VIO) 259, 260/88-32-01.
One violation was identified in the Operational Safety Verification
program area.
3. Surveillance Observation (61726)
The NRC inspector observed and/or reviewed the surveillance instructions
(SI) discussed below. The inspection consisted of a review of the
procedures for technical adequacy and conformance to TS, verification of
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test instrument calibration, observation of the conduct - of the test,
confirmation.of proper removal from service and return to service _ of the
system, and a review of the test data. The inspector also verified that
limiting conditions for operation were met, testing was accomplished by
qualified personnel, and the surveillance were completed at the required
frequency,
a. Fire Protection Surveillance Discrepancies
An NRC inspector accompanied licensee fire protection personnel
during the performance of 0-SI-4.11.A.5, High. Pressure Fire
Protection Valve Position Verification. During the SI performance,
the inspector identified the following discrepancies:
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Battery . Room 2 sprinkler isolation valve 2-26-1358 was not
. included in the valve lineup, although the same valve for
battery Room I was included.
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Hose Station 2-26-807A had a connection wrench attached vith a
chain that was too short to allow use.
These items were discussed with licensee management and it was agreed
that the issues would be investigated and corrective action taken.
Resolution of this issue will be reviewed 'during future resident
inspection coverage,
b. Performance of Incorrect Step in Surveillance Instruction
On.0ctober 17, 1988, an unplanned Engineered Safety Features -(ESF)
actuation occurred while performing 2-SI-4.2. A-10, Reactor Building
and Refueling Floor Ventilation Radiation Monitor Calibration and
Functional Test. After performing step 7.6.110 ' of the SI, the
technician turned to page 38A instead of page 38 and performed step
7.6.111.6 instead of step 7.6.111.1. Step 7.6.111.6 was the step to
reset the radiation monitor for the reactor zone exhaust. When step
7.6.111.6 was performed without first performing steps 7.6.111.1 thru
7.6.111.5, an ESF actuation occurred. Failure to perform the SI in
the proper sequence was identified as a second example of VIO
259,260/88-32-01.
One violation was identified in the Surveillance Observation Area.
4. Modifications (37700)
An NRC inspector 'ollowed the licensee's ongoing work associated with
Engineering Change Notice (ECN) E-2-P7131, which was related to
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NUREG-0737, Item II.F.2. This modification was to reroute the Unit 2
reactor water level reference legs, in order to reduce the routing of the
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reference legs inside the drywell. This would minimize the potential of
l erroneous reactor water level indications in the event of post-accident
boiling in the reference legs. When completed, the modification will
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satisfy the actions identified in Generic Letter (GL) 84-23, as presented
in the TVA Nuclear Performance Plan (NPP).
Four reactor vessel water level lines were being rerouted outside of the
drywell. . Two were to be routed through existing penetration X-17, an j
abandoned Residual Heat Removal (RHR) system penetration containing capped q
piping. The other lines were to be rerouted inside .the drywell through '
the debris screen into existing 18 inch diameter containment atmospheric
dilution (CAD) ducting. The completed ducting with the two lines were to
exit the drywell through existing penetration X-26. The existing reactor
water level sensing line penetrations (X-28A, X-280, X-29A, and X-290)
were to be capped.
The inspector reviewed 'the documentation associated with the ECN,
including the licensee safety evaluation, and accompanied the system
engineer on a walkdown of the ongoing work. No problems were noted with
the ECN documentation or the observed work.
The activity inspected in this area appeared to be effective with respect
to meeting the objectives of the NUREG-0737 modification. However, at the
time of the' inspection, the work was not yet complete. Further review and
evaluation will be performed during future reporting periods as part of
the normal NRC resident inspector activities.
No violations or deviations were identified in the modification area.
5. System Return to Service (71711)
In preparation for fuel loading, the licensee was completing a systematic
evaluation of known restart issues and deficiencies, establishing pre-
requisites, and completing specific work required to ensure fuel loading
would be conducted in a safe and reliable manner. For each system
required by TS to support fuel loading, the licensee was to complete
modifications, correct known deficiencies, and complete work requests that
would impact the safety function or operability of the system. For those
NPP Volume III Special Programs where the discovery and corrective action
implementation were incomplete, the licensee was to prepare written
justification that system operability was not likely to be impaired by
undiscovered deficiencies or unfinished corrective actions.
The NRC review of a sample of the licensee's return to service activities
a- included the following aspects of the program:
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The licensee's position papers developed to justify the acceptability
of fuel load, given the status of the major NPP programs such as
Electrical Issues, Seismic Issues, Instrument Line Slope, and
Procedure Upgrades
The scope of systems required for Fuel Load and system boundaries
required to be reviewed under the system pre-operation checklist
(SPOC) process
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System design completion verification as controlled by the Department
of Nuclear Engineering (DNE) system acceptance evaluation process
System alignment, status assessment, end operability determination,
and
System configuration control and control over special operating.
conditions following declaration of system operability.
a. System Safety Evaluations
On October 4, .1988, the NRC inspector observed a meeting between a
plant system engineer and a DNE system engineer to review the
configuration of system 78, Fuel Pool Cooling System, as part of the
DNE system acceptance for fuel load per Browns Ferry Engineering
project (BFE?) PI 88-07,- Systems Plant Acceptance Evaluation. The
engineers reviewed the results of the safety evaluation and an
unreviewed safety question determination as , part of the return to
service process. The plant system engineer identified an apparent
contradiction between the system safety functions identified in the
DNE safety evaluation and the safety functions described in the Final
Safety Analysis ; Report (FSAR). The DNE safety evaluation identified
the. spent fuel pool heat removal function as non-safety related,
whereas the FSAR described this function as part of the safety design
basis. The DNE safety evaluation also described the spent fuel pool
water level monitoring, maintenance, and prevention of inadvertent
drainage function as safety related, whereas the FSAR described this
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function as a power generation design basis. The contradiction
resulted from the Design Baseline Verification Program (DBVP), which
included a reconstitution of design basis criteria documents. These
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documents were used as the basis of the DNE system safety evaluation.
The apparent contradiction led to revision 1 to the safety evalua-
tion, which was submitted by the DNE project engineer to the Piant
Manager for use in the SP0C for system 78 on October 18, 1988. This
revision stated that the fuel pool cooling function of the system was
a safety function but that this safety function was not required for
fuel loading. No further justification was included to document why
this function could be excluded for fuel load. Subsequently,
following completion of the SP0C and establishment of system status
and configuration control, the NRC inspector learned that the DNE
position on the safety function had not changed and that the restart
design criteria still listed the heat removal function as non-safety
related. During a meeting on October 21, 1988, the NRC inspector
' informed the Plant Manager of this problem and expressed concern that
something as fundamental as system safety function could be in
question at this point in the restart process. This specific issue
had not been resolved by the end of this inspection period.
The NRC inspector met with members of licensee management in order to
determine management contrels over similar contradictions. There was
apparently no planned transition for promulgating an effective date
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.for when the DBVP design criteria documents would supersede the FSAR
for conflicts such as this. This step was considered necessary by.
the NRC ' inspector, since the FSAR update is . one of the last
activities in the DBVP' and a fairly lengthy period may transpire
before the' FSAR is brought into conformance with the reconstituted
facility design basis. Subseouent to this meeting, the DNE safety
evaluation was revised to clarify the system safety functions. That
revision states that fuel pool heat retroval is the system's primary
function, but still- does not list the functica as a safety function
(refer to paragraph 10.q of this report for a discussion of the FSAR
long term update program,- Unresolved Item (URI) 88-02-03). The
licensee stated that contradictions of this type would be _ corrected
by conducting a review of-the DBVP punchlist to identify all the FSAR
changes currently. kn'own, and providing this list ~ to ' all personnel
qualified to perform safety evaluations per 10 CFR 50.59. Licensee
management further stated that the.se discrepancies' are punchlisted
for revision of _ the FSAR to bring the two documents into agreement.
This issue will be tracked along with URI 88-02-03 and must be :
resolved prior to restart.
- Resolution; of these issues will be reviewed in conjunction with
j future resident ~ inspector coverage of system return to service.
b. 10 CFR 21 Reports
The licensee's program' for addressing outstanding 10 CFR Part 21
reports for fuel load systems was reviewed. Specifically, activities
related to IE Bulletin 88-03, Inadequate Latch Engagement in HFA-Type
Latching Relays Manufactured by General Electric, were reviewed. The
. licensee indicated in their response to this bulletin, dated July 6,
1988,- that inspections and any necessary repair or replacement of~the
relays would be accomplished prior to restart. The NRC inspector
observed that this item was not on the licensee's Site Master Punch
Li st (SMPL) for tracking, and confirmed through discussions with
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licensee management that the activities were not considered to be
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required for the fuel load systems. The NRC inspector considered
this position to be unacceptable, given the age of this issue.
General Electric (GE) first made purchasers of the subject relays
aware of the potential deficiencies via a November 16, 1987, Service
Advice Letter (SAL). .The prompt notification requirements of
10 CFR 21 are meaningless if prompt action is not taken by the
! licensee. Licensee management representatives were informed of the
NRC inspector's position, and at the end of the inspection period had
not decided how the issue would be resolved. The licensee was
L expected to evaluate the results of the limited inspection activities
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that had been accomplished and perform an engineering evaluation of
the remaining inspection attributes in order to determine whether a
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failed latching relay could adversely impact a system required for
fuel load.
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Due to the apparently excessive time period that this Part 21 report-
had remained open at Browns Ferry, the inspector requested a listing
from the TVA licensing organization of any other 10 CFR Part 21
reports that remained open pending final corrective action. Thi s
listing was not available at the end of the inspection period. When
this listing becomes available, the inspector will review and assess
the effectiveness of the licensee's' program for Part 21 report
resolution.
Followup on these issues will be part of the continuing resident
inspector coverage of system return to service,
c. System Preoperability Checklist (SP0C)
The SP0C package for system 69, Reactor Water Cleanup (RWCU) System,
was reviewed. Operability Item Deferral Number 69-1 documented
deferral of the approval by the Joint Test Group (JTG) of the restart
test results until system operability declaration. The NRC. inspector
held discussions with the system engineer, the Restart Test Manager,
and the Return to Service Manager regarding this deferral and learned
.that not only had the JTG review of the results package been
deferred, but also the performance of the entire restart test for
system 69. The test was not just deferred until the declaration of
system operability (required before fuel load) but was in fact
deferred until af ter fuel load. The licensee position was supported
by an engineering justification attached to the deferral form which
concluded that the restart test.was not required for fuel load.
The NRC inspector held discussions with licensee management on this
issue, and stated that the Restart Test Program (RTP) tests were
considered by the NRC to be the foundation for system operability
declaration and system return to service. Furthermore, the licensee
had stated that all discovery programs of the NPP would be complete
at fuel load or a justification would be provided for considering
that possible unidentified discrepancies would not impact system
operability. The NRC considered the RTP to be an effective means for
identifying system operability and functional' concerns.
Further discussions were held with licensee management representa-
tives, who indicated that a more cohesive decision making process had
generally been used to justify exceptions to RTP testing. Basically,
all functions of the systems which must be operable for fuel loading,
as identified by the fuel load system boundaries, were to be verified
by the RTP. Other system functions which were not required to be
operable until just prior to restart might not be confirmed by
performance of the RTP. This logic was considered by the NRC
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inspector to be technically acceptable and applicable to all RTP
l deferrals except in the case of the RWCU system. The licensee
indicated that a review of the need for deferral of RWCU system RTP
testing would be accomplished and the RTP completed, if possible,
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prior to fuel load. Followup on - this specific item will be
accomplished during future resident inspection coverage.
The NRC inspector reviewed the SMPL entries associated with the
return to service of system 78, Fuel Pool Cooling (FPC) System. The
contractor recommendations report for system 78 stated that the local
control panel (panel 25-16) was inside a contaminated area. T11s
impeded operator access to local control of FPC system pumps and
valves. The NRC inspector determined that this condition had not
been corrected. When informed, the Radiological Controls Manager
reviewed the feasibility of decontaminating the immediate area around
the panel. Operations and radiological controls personnel performed
several joint plant walkdowns in order to identify other areas vhich
should be decontaminated to facilitate operator access to key plant
equipment. The NRC inspector observed examples where this had been
adequately accomplished. The inspector judged the licensee's '
response to this concern to be through and well directed.
In summary, the System Return to Service program had not totally met the
expectations of the NRC with regard to meticulous attention to detail and
thoroughness of open item resolution. The NRC inspectors concluded that
further review and evaluation were required, and that the following
weaknesses described above will be reviewed during upcoming inspections.
1) System safety function definition
2) Outstanding Part 21 reports
3) Completion of RTP testing for fuel load functions
4) Review of open contractor recommendations for fuel load systems
No violations or deviations were identified in the area of system return
to service.
6. Reportable Occurrences (90712, 92700)
The Licensee Event Reports (LERs) listed below were reviewed to determine
if the information provided met NRC requirements. The determination
included the verification of compliance with TS and regulatory require-
ments, and addressed the adequacy of the event description, the corrective
action taken, the existence of potential generic problems, compliance with
reporting requirements, and the relative safety significance of each
event. Additional in plant reviews and discussions with plant personnel,
as appropriate, were conducted,
a. (CLOSED) LER No. 296/82-35: Failure of Drywell Floor Drain Sump
Outboard Isolation Valve To Close.
While the licen~see was confirming the operability of the water flow
integrator from the drywell floor drain sump, the outboard isolation
valve failed to close because of a stuck piston in the three-way
solenoid valve operator. To correct the problem, the licensee
replaced the three way solenoid valves on all three units.
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The NRC inspector reviewed and evaluated the completed work plans and
concluded-that corrective actions were adequate. This'LER is closed.-
,7 b. (CLOSED) LER No. 260/85-15: ' Insufficient Voltage To High Pressure
Coolant Injection Controls.
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Injection (HPCI) control circuitry, the licensee determined that the
electric . governor motor (EGM) control box was . not . receiving the
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voltage required to. meet minimum voltage' specifications for'the HPCI
turbine controller when the input ' voltage was at design minimum. Tne
licensee replaced the voltage dropping network feeding the HPCI
governor.with a 48 volt DC power supply which would meet the voltage.
. requirements for all' analyzed conditions.
The NRC inspector reviewed and evaluated the ECN, engineering
analysis, and work plan completion and verification form, and
considered the corrective action adequate. This LER is closed.
c. (CLOSED) LER No. 259/85-17: Lack' of Environmental Qualification for
H202 Analyzer Valves.
A licensee design evaluation of the . teflon valve seats and valve
packing in the Hzoz analyzers had determined that' accident. radiation
levels would exceed the radiation failure threshold 'of teflon. The
licensee changed the valve stem packing and replaced the valves as
applicable.
The NRC inspector reviewed and evaluated the work plan specification
and the completed work plans for the valve and valve stem packing
replacements. Corrective actions .taken were considered adequate. ;
This LER is closed. !
d. (CLOSED) LER No. 259/85-50: Failure to Perform Surveillance
Instructions.
During an October 1985, licensee management review of surveillance ,
scheduling, the licensee identified eight sis that were not being
performed as required by TS for a unit in shutdown for refueling.
Violation 259, 260, 296/85-57-09 was issued on February 11, 1986, for
failure to perform sixteen required sis, including the eight
identified by the licensee (See paragraph 10.c). Closure of the 1
violation closes this LER.
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e. (CLOSED) LER No. 259/85-55 and Rev. 1: Open Fire Barrier
During licensee maintenance activities, a spare sleeve penetration in
a fire barrier was found to be unsealed.
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Violation 259,.260, 296/86-09-03 was issued on May 21, 1986, for
' failure to include piping fire barrier penetrations in a surveillance
- n program. The violation was closed in NRC Inspection Report 259, 260,
296/87-21, based on appropriate licensee corrective action. This LER
is therefore closed.
f. (CLOSED) LER No. 259/86-06, Rev. I and Rev. -2: Tornado Missile
Protection for Vent Towers.
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During a 1986 design evaluation _ of control bay ventilation modifica-
tions, design engineers identified an unanalyzed condition involving
tornado / missile protection for equipment located in the- control bay
vent towers. The design basis evaluation .for protecting existing
equipment had been previously established, and the results were used.
to perform a site specific risk assessment. The assessment results
indicated that the. risk to the equipment was extremely low and that
no modifications to the vent towers were required.
The NRC _ Materials Engineering Branch .of NRR reviewed the LER and'
" Calculations of Probability of Occurrence and Consequences of
Tornado-Generated Missile Strike of Safety-Related Equipment in -Vent-
Towers", and all NRC questions were resolved through a series of
discussions with the licensee. Corrective actions were considered
adequate and this LER is closed.
g. (CLOSED) LER No. 260/86-10, Rev.1 and Rev. 2: Recirculation Inlet
Nozzle Safe End Cracks.
In July 1986, the licensee determined by ultrasonic inspection that
all ten of the Unit 2 recirculation system reactor vessel inlet
nozzle safe. ends were cracked. The licensee replaced the Unit 2
inlet nozzle safe ends and a portion of the associated recirculation-
system piping.
The NRC inspector reviewed and evaluated the ECN, the specifications
for the replacement of the recirculation inlet nozzle safe ends, the
work plans, and the completion notifications for the work plans. The
corrective actions taken were considered adequate. This LER is
closed.
h. (CLOSED) LER No. 259/86-22: Nonsafety Grade Air Actuators on
Containr.ient Isolation Testable Check Valves.
During a design review, the licensee determined that nonsafety grade
air t.ctuators on containment isolation testable check valves could
fail, causing the check valve to open and relieve reactor coolant
into piping not ' designed for reactor system temperature and pressure.
The licensee removed the air supplies to the valves to prevent
inadvertent actuation during plant operations, and installed quick
disconnect couplings to allow easy reconnection for testing during
shutdown.
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The NRC inspector reviewed and evaluated the work plans and work _ plan
closures. The corrective actions taken were considered adequate.
.This LER is closed.
1. (CLOSED) LER No. 259/88-01: Unplanned Reactor Water Cleanup Isolation
Due to Loose Connection.
The licensee had determined during troubleshooting that the cause of
an unplanned RWCU isolation was a loose solder connection in the RWCU
temperature indication circuitry. The licensee repaired the solder
connection. and recalibrates the switch. A brush recorder was
temporarily connected for thirty hours to monitor switch behavior.
No abnormalities were observed.
The NRC inspector reviewed and evaluated the operator logs and the
completed maintenance requests. Corrective action taken by the
licensee was considered adequate. This LER is closed.
j. (CLOSED) LER No'. 260/88-02: Trip of Reactor Protection System Bus 2B
Feeder Breaker Initiates Engineered Safety Features Actuations.
On May 26, and May 27, 1988, the breaker (952) feeding reactor
protection system (RPS) bus 2B tripped and caused an ESF actuation.
.The licensee performed a failure investigation and no root cause
could be determined. On May 27, 1988, breaker 952 was replaced with
a molded case switch that was previously approved by an ECN and work
plan. No trip of RPS bus 2B feeder had occurred since breaker 952
was replaced.
!
The NRC inspector reviewed the ECN, failure investigation, and work '
plan. Actions taken were considered adequate. This LER is closed.
k. (CLOSED) LER No. 296/88-02: Unplanned Reactor Water Cleanup System
Isolation Due to Personnel Error. ;
1
An isolation of the RWCU system resulted from a personnel error, when
the temperature trip setpoint knob was accidentally turned during the
decontamination of instrument panels.
The NRC inspector reviewed and evaluated the maintenance request and l
the recalibration of the setpoint from the "as found" value of 52 i
degrees F to the correct RWCU trip setpoint of 140 degrees F. The i
NRC inspector also reviewed the training attendance records for the
training given to all decontamination crews reminding them to use )
caution when decontaminating any plant panels. This corrective !
action was considered appropriate. This LER is closed. !
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1. (CLOSED) LER No. 260/88-10: Inadequate p ocedures Cause Two Cases of l
Missed Samples That Were Required to ;ompensate for Inoperable
Radiation Monitors.
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In-' January 1988, two - similar events' occurred involving missed
compensatory - sampling ~ for inoperable effluent - radiation monitors.
.The licensee determined that inadequate' procedures were the 'cause
of both events. The licensee revised the applicable operating
instruction and SI to ensure that the chemistry lab would be notified
when any. required sampling should be initiated or stopped to meet TS
requirements.
The ' NRC inspector - reviewed the . revised procedures- .for' all three
units, and concluded that the corrective actions taken were adequate.,
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This LER is closed.
No violations or deviations were identified in the area' of. Reportable
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Occurrence s'.'
7. - Restart Test Program (99030B)
The. inspector attended-RTP status meetings, reviewed RTP test procedures,.
. observed RTP tests ari associated test. performances, . reviewed RTP test
results (including test exceptions), and attended selected Restart Opera-
tions Center (War Room) and Joint Test Group (JTG) meetings. The' RTP
activities and associated activities monitored, and status of testing,
during the period of the_ inspection, are discussed below.
- a. RTP Status =and Restart Test Performances
The . inspector maintained cognizance of ongoing restart . test
activities, and monitored particular activities in detail as appro-
priate. Specific inspection observations'are discussed in paragraphs
7.b'and 7.c below.
The following information summarizes the status of procedures, tests
performed, and the hardware related test exceptions identified by the
RTP group, at the time of the inspection:
Required for Required for
Fuel Load Criticality Total
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Procedures Issued and Approved 28 15 43
Tests completed as of 10/31/88 22 4 26
Completed test approved by the 20 1 21
Plant Manager i
Unresolved Hardware TEs 18 35 53
The total number of procedures required for fuel load was originally
identified as 26, but had been increased to 28. However, discussions
with licensee management indicated that this number might later be
reduced as a result of the return to service of systems for fuel i
load.
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The following restart tests were in progress during this reporting
period:
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o RTP-023, Residual Heat Removal Service Water
I o RTP-030, Diesel Generator and Reactor Building Ventilation
o RTP-031A, Control Building HVAC (Water Side)
o RTP-031B, Control Building HVAC (Air Side) ,
i
o RTP-047, Turbine Generator / Electro-Hydraulic Control
o RTP-57-3, 250 Volt DC Unit Battery
o RTP-06A, Primary Containment Isolation
o RTP-069, Reactor Water Cleanup
o RTP-070, Reactor Building Closed Cooling Water
o RTP-082, Diesel Generators i
o RTP-085, Control Rod Drive ,
o RTP-099, Reactor Protection System
o RTP-ICF, Integrated Cold Functional
The above tests were either in the prerequisite stages, system
performance stages, initial RTP group reviews, DNE reviews or final
JTG reviews.
b. Diesel Generator Testing
Although it was previously reported in NRC Inspection Report 259,
260, 296/88-28 that field activities involved in test 2-BFN-RTP-082,
Diesel Generators, had been completed, further review by the RTP
procedures review group indicated that the load reject test of DG 3A
had either not been adequately documented or had not been performed.
The test was subsequently reperformed. The failure of the DG to trip
from 2600 KW, when the output breaker was opened was identified as a
test exception. NRC review of the DG RTP results is identified as
IFI 259, 260, 296/88-32-02. This issue must be resolved prior to .
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Unit 2 restart.
c. Specific Test Wttnessing and Results Evaluation
The NRC inspectors observed and reviewed portions of the performance
of 2-BFN-RTP-099, Reactor Protection System. Section 5.6 involved
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testing the RPS high water level trips in the . scram discha'rge tanks, -
and Section 5.26 involved cable voltage drop testing to verify that
the voltage available at-the system components will be greater than.
-the component minimum operating voltage. No deficiencies were
identified.
No violattons or deviations were identified in the Restart Test . Program
area.
- 8. Followup.of NRC Bulletins (92703)
a. (CLOSED, Unit 2 only) IE Bulletin No. 84-02: Failure of General
Electric (GE) Type HFA Relays in Use in Class 1E Safety Systems.
'This bulletin addressed similar failures of GE HFA relays, which had
been reported in several GE service reports. The licensees were.
requested to inform the NRC of their plans, including schedules, for
implementing the manufacturer's recommendation in the~ ' subject? GE
reports. . The licensee had completed all but the following two items
documented in IE Report 88-28:
(1) Completion of the replacement of all normally-deengerized relays
,in Unit'2
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(2) Completion of the replacement of normally-energized relays in
Unit 1 Systems required for the startup of Unit 2
>
The NRC inspector reviewed the completed work requests and computer
printout documenting the completion of the relay replacement'for the
two-items. This action was considered adequate to support Unit 2
restart.
This bulletin is closed for Unit 2 only.
b. (OPEN) IE Bulletin No. 85-03 and Supplement I: Motor-Operated Valve
Common Mode Failures During Plant Transients Due to Improper Switch
Settings.
As requested by action item e. of Bulletin 85-03 and Supplement I,
the licensee identified the selected safety-related valves, the 1
maximum differential pressures of the valves, and the program to l
assure valve operability in letters dated May 13, 1988, September 30,
1986, and May 1, 1987. Review of these responses indicated the need
for additional information, which was reouested in an NRC Region II
letter dated April 1,1988.
Review of the licensee's response dated July 15, 1988, to the request
for additional information indicated that the licensee's selection of i
the applicable safety-related valves, and the maximum differential I
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pressures of the valves, met the requirements of the bulletin. The i
program to assure valve operability ver action item e. of the ,
bulletin was considered acceptable,
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The results of the inspections to verify proper implementation of 1
this program, and the review of the final response required by action
item f. of the bulletin, will be addressed in future inspection
reports. Resolution is required for restart.
c. (OPEN) IE Bulletin No. 88-03: Inadequate Latch Engagement in HFA-Type
Latching Relays manufactured by General Electric Company.
This bulletin was issued as a result of a report from GE which stated
that some HFA type ' latching relays were malfunctioning. The NRC
stated that operability of all HFA-151B, -1548, and -154E relays with
a manufacturing date code prior to November, 1987 should be
inspected. The licensee response to the bulletin committed to
inspect, repair, or replace. the relays failing the inspection
criteria before the restart of each unit. During the period of the
inspection, the licensee was completing portions of the inspections
on the systems required to support fuel load. An engineering evalua-
tion of the remaining inspection attributes will be accomplished
prior to fuel load. (See paragraph 5.b. of this report), therefore,
this item remains open.
d. (OPEN) IE Bulletin No. 88-04: Potentiel Safety-Related Pump Loss.
This bulletin addressed the issue that when two centrifugal pumps are
'
operated in parallel and one of the pumps is stronger than the other,
the weaker pump may be dead-headed when the pumps are operating in
the minimum. flow mode. This could cause excessive pump impeller
wear. The phenomenon is manifested at low flow rates because of the
flatness of the pump characteristic curve in this range.
The licensee's response stated that verification of the adequacy of
the miniflow line sizing for the Residual Heat Removal Service
Water / Emergency Equipment Cooling Water (RHRSW/EECW), RHR, and Core
Spray (CS) pumps is considered to be a portion of the essential ,
design calculations for Browns Ferry. These calculations are under !
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TVA's Design Calculation Review Program for essential calculations,
which are commitment items 78, 78A, and 78B of the Browns Ferry NPP,
Volume 3, revision 1. This program is required to be completed prior
to restart of Unit 2. The licensee has committed to a supplemental
response of 30 days upon review completion of items 78, 78A, and 78B
of the NPP program. Based on the above, the NRC inspector concluded
that this item is acceptable for fuel load but will remain open l
pending completion of the licensee commitments. This item must be j
completed prior to unit restart. i
No violations or deviations were identified in the area of NRC Bulletins.
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9. Followup of Open-Inspection Items (92701)
a. (CLOSED) Inspector Followup ~ Item (259, 260,- 296/86-05-08),.
Questionable Instrument Calibration Techniques for Radiation
Monitors.
The original issues associated with 'this item were inspected and
reported in NRC Inspection Report No. 259, 260, 296/86-32; however, a
new issue related to .the revised SI was detected which prevented
closure at that time. The new-issue related to the possibility that !
the SI could cause a reactor trip if performed during power opera-
tions. Such a trip could result from = a high main steam tunnel
temperature condition created because the SI called for isolating the
normal ventilation in the reactor zone under test. The SI had been
written in this manner in order to avoid inadvertent ESF actuations
during the SI by manually initiating the ESF as a prerequisite to the
test. The NRC inspector had asked the licensee in September 1986,-to=
reevaluate this approach. The licensee implemented Design Change-
Request '(DCR) D3311, which installed permanent test boxes to allow
testing of the radiation monitors with the ESF. actuation . relays
defeated.in order to prevent the spurious actions which had occurred
too frequently in the past.
The NRC inspector. reviewed documentation . associated with this.
modification and inspected the installation of the test boxes in the
field. The inspector confirmed that 2-SI-4.2.A-10, Reactor Building
and Refueling Floor Ventilation Radiation Monitor Calibration -and
Functional Test,-was revised on September 9, 1988, to incorporate.the
hardware changes and that the reactor zone normal' ventilation system
would remain in service throughout performance of the SI. These
changes should eliminate the potential for a reactor trip to occur
from high main steam tunnel temperatures. This item is therefore
closed.
b. (CLOSED) Inspector Follcwup Item (259, 260, 296/86-06-07), Design
Requirements for Instrument Sensing Line Slope.
This item was written to ensure that engineering requirements were
established for the slope of instrument sensing lines at the Browns
Ferry Nuclear Plant (BFNP). In February 1986, the NRC inspector
learned that the only source of requirements for instrument sensing
lines was TVA General Construction Specification G-60. G-60 required
a target slope of 1-inch per foot with a one-eighth inch per foot
absolute minimum. The preface to G-60 stated that the specification
was only applicable to Bellefonte Nuclear Plant, therefore leaving
BFNP with no requirements. The licensee was using G-60 for work at
BFNP since there was no other applicable document. The NRC inspector
reviewed Specification Revision Notice (SRN) G-60-1, dated May 22,
1986, which made G-60 applicable to future modifications at Browns
Ferry. The inspector concurred that the specifications in G-60 were
appropriate for Browns Ferry, and the item is closed.
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c. (CLOSED) Inspector Followup Item (259, 260, 296/86-32-06),
Deficiencies in Diesel Fire Pump Building
An NRC -inspector identified several material and housekeeping
deficiencies that existed in the Diesel Fire Pump Building.
The NRC inspector reviewed documentation provided by the licensee to l
support actions taken to correct the identified deficiencies.
Additionally, the NRC inspector conducted a tour of the Diesel Fire
Pump Building to observe actual conditions. During the tour, the
inspector observed that all previously identified housekeeping and
material deficiencies, with one exception, had been corrected. The
one exception was the battery mounting rack not being secured to the
building foundation.' This condition still existed and had been
evaluated by the licensee as acceptable. FSAR section 10.11.5.1
stated that the High Pressure Raw Water Fire Protection System is not
designed Class I seismic and does not necessarily remain functional
in an earthquake. However, a fire in any component of an essential
system will not prevent safe shutdown of the reactor because
essential components are redundant and meet separation criteria and
the plant construction does not easily propagate fires. Portable
fi re protection equipment is provided for use following an
The inspector noted a requirement in Plant Manager Instruction (PMI)
12.12, Conduct of Operations, for a daily tour of the building by an
operator during routine rounds. Additionally, several new minor
material deficiencies were noted by the inspector, which were pointed
out to licensee fire protection personnel accompanying the tour. The
deficiencies included damaged piping insulation, deteriorated rubber
boots, painted rubber expansion joints, and apparently damaged or
unused heat tracing. The deficiencies were documented by the
licensee on MRs 902120, 902121, 902122, and 902123. The overall
condition of the building was much improved and the NRC inspector
agreed with the licensee's evaluation that mounting was not required
(the construction of the battery racks was otherwise adequate).
Corrective actions taken by the licensee were considered to be
adequate. This item is closed.
d. (CLOSED) Inspector Followup Item (260/86-40-05), Material
Discrepancies and Housekeeping Problems in the Main Steam Valve
Vault.
An attempt was made by the NRC inspector to close this item in
April 1988, following notification by the licensee that a cleanup had
been performed. The inspector toured the area in April and found
conditions still unacceptable (refer to Inspection Report t
259,260,296/88-10). On October 18, 1988, the inspector made a
followup tour of the area and noted a significant improvement. All
previously identified items were corrected and no new concerns were
detected. This item is closed.
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e. (OPEN) Inspector Followup Item (259, 260, 296/86-40-12), Potential
For Overpressurization of Residual Heat Removal System Piping.
A modif.ication was installed in order to-reduce an excessive' pressure
drop across a throttling valve in the RHR' system. An . orifice plate
was installed in a section of piping rated at 150 psig and increased- i'
the pressure in this section of pipe :to an undetermined valve.
Although the section of. pipe in question (the test return line) was
not instrumented during the ' post modification test, the . nearest-
portion wit' p.ressure indication exceeded 300 psig.during the test.
~ The NRC?dentified that the potential .for exceeding the pipe design
pressure had' not been analyzed as part of the modification. This
7" finding was made as part of the Unit 3 modification package review.-
.
The Llicensee completed the. modification on Unit 2 and performed a
more extensive post modification test on October 4, .1988. This PMT
duplicated the worst condition of both RHR pumps operating with. full
flow through the' crifice, and measured the pressure in.the suspect
,
piping. .The test results indicated that the worst case. pressure L
increase with the drywell at atmospheric pressure was 137 psig, which
Additionally, a design calculation
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was within the 150 psig rating.
was performed by the licensee which resulted.in'an expected pressure
-
drop of 133 psid across the orifice under maximum flow conditions.
.The NRC inspector reviewed the test data and the design calculation
and confirmed that these values were . appropriately derived. The-
inspector determined tha.t under normal conditions the piping would
remain within its design rating; however, following the design basis
LOCA, FSAR Section 14.'6.3.3.2 indicates the. Torus pressure can be as
high as 271psig. Under accident conditions ; this ' pipi.ng ' section '
pressure could be as high as 164 psig exceeding the design pressure
of 150 psig. The licensee's review of the test data did not result
in their identifying this problem.
This IFI remains open pending TVAi s evaluation of the acceptability
of the piping design for the potential accident conditions,
f. (CLOSED) Inspector Followup Item (259, 260, 296/88-05-05), Close out
of Restart Test Program Maintenance Work Request.
This item documented a concern identified during the RTP testing of
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air dampers in the DG rooms. It was identified by the Site Quality
Assurance Monitoring Group that maintenance requests (MRs) not being
addressed was a site wide problem and not just an RTP problem. The
QA inspector' initiated Condition Adverse to Quality (CAQR) BFQ 88
0143 to document missing MR's. Although the initial CAQR was aimed
at MRs generated as a result of environmental qualification (EQ)
walkdowns, a further revision of the CAQR was aimed at the site in
a general. The NRC inspector will monitor the followup of the CAQR.
This item is closed.
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g. (OPEN)' Inspector Fo116wup Item (260/88-10-01)', . Main Steam Tun ~nel'
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Blowout Panel Function Possibly Defeated.
'The licensee identified,and documented on CAQR 880293-that the Unit 2
Main St_eam Tunnel . blowout panels were, not installed in accordance.
t: ;with'the. drawings. Specifically, RTV . sealant had been used to. fill
large gaps between the panels and framing. 'The RTV could act ~1ike an
adhesive .to prevent l blowout of the panels' at their design
differential pressure.
Corrective action was completed by the licensee with theiexception
that many of the explosive bolts . were inadequate and had been
replaced on a<short term basis with non-explosive bolts. Only the-
secondary containment integrity function of the panels is required
-.for fuel load. The -blowout function needed only to be ' operable to
mitigate a steam pipe' break, which would not be possible until'after.
restart. This . item was therefore acceptable for fuel load but'
r s remained open pending. completion of. corrective action by the.
licensee. Final corrective action is required prior to' restart..
h. (CLOSED) Inspector Followup Item (259, 260, 296/88-10-03), Lack:of
"' Understanding of the Restart . Test Program by On-Shift Senior
E Personnel '.
This item documented a concern that on-shift senior reactor operators
(SR0s) upon returning from extended periods of training were not
fully aware of the RTP. The NRC inspector reviewed a memo dated
July 13, 1988, from the Operations . Superintendent to all operation
personnel, in which the purpose- of the RTP was clarified. . The
inspector continued to ' observe ' operations personnel, especially
. senior or shift personnel, and the RTP Test Director's activities,
and determined that the RTP program was adequately understood by
operations personnel. This item is closed.
One IFI was upgraded to a violation in the area of Followup on Open
Inspection Items.
10. Licensee Action on Previous Enforcement Matters (92702)
a. (CLOSED) Violation (259/85-13-03), Failures to Follow Procedures and
an Inadequate Procedure During Retest of Control Rod 34-03.
This violation was identified in Inspection Report 259/85-13-03, but !
was not cited at that time. Subsequently, Violation 259, 260,
296/85-36-01 was issued to address this finding. The violation
comprised examples of failure to follow procedures and an inadequate
procedure regarding Unit I control rod 34-03 maintenance activities.
(' Resolution of this item is addressed through the followup on
'
Violation 85-36-01, which is discussed in paragreb 10 b of this
report. This violation is therefore closed.
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b. (OPEN) Violation (259, 260, 296/85-36-01), Failure to Follow
Procedures and an Inadequate MR for CRD 34-03 Post-Maintenance. H
Testing.
Violation 259, 260, 296/85-36-01 consisted of four examples in which
procedures were not adherea to or were inadequate. Three of the four
examples pertained to maintenance work done on Unit 1 CRD Module
34-03. The fourth exsmple involved the licensee's failure to perform
a safety evaluation in order to determine HPCI system operability ,
with failed-open resistors on HPCI steam line drain isolation valves.
The 1icensee's corrective actions for the fourth example of the
violation were being reviewed separately and will be addressed in a i
future inspection report.
The licensee's response to the violation was provided in a letter to
the NRC dated September 27, 1985. The NRC inspector reviewed the
licensee's reasons for the violation and their corrective actions.
1) Example 1: Inadequate CRDH Maintenance Request
.{
The first example of the violation dealt with an inadequate
maintenance request, MR A126652, which failed to contain
functional and post maintenance testing (PMT) requirements as
required by the Mechanical Maintenance Instruction (MMI) 28,
Control Rod Drive Hydraulic Unit (Repair, Removal, and Replace-
ment). The licensee's reason for the violation was that the ;
foreman failed to follow MMI-28, and that MMI-28 lacked clarity ;
and did not adequately cross-reference applicable sections l
within the procedure (e.g., testing). MMI-28 was revised j
accordingly.
The NRC inspector reviewed the most recent version of MMI-28, l
revision 6, dated August 23, 1986, and verified that appropriate
revisions had been made for clarification of PMT requirements.
Section 10.3 of MMI-28 provided a listing of PMTs for different '
areas of maintenance performed on HCU units and delineated the
individual responsible for performing the PMT. Also, the NRC
inspector reviewed the training attendance record dated
November 24, 1987, verifying receipt of training for draft
personnel on MMI-28 requirements. The NRC inspector concluded
that adequate corrective action had been taken for this example
of the violation.
2) Example 2: Failure to Exercise Control Rod within Time Limit
The second example of the violation involved the licensee's
failure to exercise control rod 34-03 within the required time
limits specified in MMI-28, Section 10.3, and Operating
Instruction (OI) 85, Control Rod Drive (CRD) System, Section
3.H.1.e.2.e. Reasons given for the violation in the licensee's
response were that MMI-28 and 01-85 required control rod
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insertion and withdrawal times (i.e. 48 plus or minus 3 seconds) {
which were too restrictive and were inconsistent with the RTI-5- 4
and vendor recommendation criterion of 40-60 seconds.
The inspector reviewed the revised procedures and verified that
they incorporated the acceptance criterion recommended by the j
vendor. Section 8.8 of. the upgraded Unit 2 0I-85, revision 3,
provided timing adjustment of contrcl rods within the tolerance. j
of 40-60 seconds. Technical Instruction (TI) 20, Control Rod j
Drive System Testing, Revision 0, provided the same requirement ]
reflected in Sections 7.3.7.4 and 7.3.8. Also, the NRC- )
inspector verified that the same criterion was provided by the
vendor (GE) in the GEK-9585/9586 document. The NRC inspector i
concluded that this example of the violation had been adequately .!
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resolved.
3) Example 3: Failure to Follow Procedure Limits on CRD Pressure -l
and Control Rod Position
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The third example of the violation involved the failure to ' f
follow procedure 01-85, Control Rod Drive System, Section 3.D.9,
in that during withdrawal of control rod 34-03 from the fully
inserted position (00 notch position) the drive water pressure
was not returned to normal limits before the rod passed the 02 i
notch position. The drive water pressure was determined'to be j
approximately 50 psi above the normal limits when the rod passed i
notch position 02. The licensee's reason for this viniation was i
that procedure 01-85 was too restrictive in its limitations of
CRD pressure and control rod position, which resulted in an i
inadvertent failure to-follow procedures. ]
For corrective action, 01-85 was revised to permit drive
pressure to remain above normal levels until the 06 notch
position is reached. Based on the CRD design, which is of a 1
finger / collet configuration with a traveling distance of 3 l
inches per step and a normal withdrawal / insertion speed of 3 1
inches per seconds + 20 percent, the inspector agreed with the i
licensee's position that 0I-85 had previously been too !
restrictive in the limitations on CRD pressure control with i
respect to control rod notch positions. Section 8.16 of the
upgraded Unit 2 procedure (0I-85, revision 3), provided clearer
and more detailed instructions on what to do when a control rod #
becomes difficult to withdraw. Caution statements in 01-85 were
changed to return the CRD drive water header differential ,
. pressure to between 250 and 270 psid (normal limits) as soon as I
l possible in order to prevent a drive from double notching in a !
high rod worth region, and to reduce exposure of drive seals and j
directional control valves to excessive pressures. Furthermore, i
the licensee incorporated GE's recommendation from the GE
contractor recommendations work to use the double clutch method
l of withdrawing a control rod from notch 00 to notch 02 just
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- applying elevated drive water pressure. ' 0I-85' for' Units
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.. :11and 3:had also been revised,' but not through the upgraded
procedure; process. ;The NRC inspector expected that: the-
procedure would be revised in the. future so'asto be consistent-
with the Unit 2 procedure. .This example of the violation was.
considered. resolved.
'
,,
The violation ; remains open pending inspection, of; corrective ~ actions
for the fourth example, which ' concerned HPCI operability.
. c' . (CLOSED) Violation (259, 260, .296/85-57-09), Failure to Conduct
Sixteen Surveillance Instructions During. Shutdown and Refueling.
~
The'11censee determined that' the root cause included the failure to.
fully implement TS requirements in plant procedures L and : personnel
error. in TS interpretation. The licensee subsequently performed all
-the SI!s not performed on Units 1 and 3, with the exception-of'SI ,.
Recirculation Pump Trip Reactor High . Pressure, on Unit 1. This SI-
was not performed because the recirculation. pumps were not operating
'and the reactor vessel head had.-been removed. Unit 2 SI's were not
~
"
performed because the fuel had been unloaded 'and! the ' applicable'.
systems were' no longer required by TS.
The licensee corrective action to avoid further violations was to
update SI-1, Surveillance Program, Appendix C to accurately ' reflect
the TS. requirements for performance of these SI's during shutdowns
and refueling. On August 23, 1988, the licensee also updated SI-1
.because of issuance of TS Amendments 136 through 144 for the restart
of Unit 2.
The 'NRC inspector ' reviewed and evaluated the corrective action
documented above, and considered it ~ appropriate to support Unit.2
restart. This item is closed.
d. (OPEN) Violation (259, 260, 296/86-25-01), Failure to Follow
Procedures (Three Examples).
1) Example A: Fire' Protection Sprinklers not Configured in
Accordance with Approved Plant Drawings
Example A of Violation 86-25-01 was attributed to inadequate ;
coordination in work plan preparation in December 1976, and. lack '
of a. post-modification test in 1977, which resulted in a failure l
l
to remove a welded blank in the fire protection line. The
condition went undetected until approximately July 14, 1986,
when the licensee was prompted to determine why a section of
piping in the Unit 3 reactor building could not be flushed
during a non-routine flush of the fire protection pre-action ;
sprinkler system for removal of any accumulation of mud or ;
clams. The affected section contained 19 sprinkler heads which 4
were rendered inoperable.
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For. corrective' action, - the : licensee impro_ved the control' and
- documentation . of work plans and control of . temporary altera-
tions, and implemented a ' program requiring. the . Fire Protection
Unit's overview of all modifications and . post-modification
testing pertaining to fire protection systems. Further, an
approximate 30 percent random sample of pre-action system branch
_
. lines were selected for a L special -air test. More' than 100
branch lines were tested and no blockages were. identified. The
testing was ; completed by November 30, -1986. The -licensee's ,
basis for selection of a 30 percent random testing scheme was
.provided in their supplemental response to the. violation, dated
March 2,1987,. - The supplemental response was reviewed by the
inspector.and found to be acceptable.
This item is also addressed in LER 296/86-06, which'was reviewed
and closed in NRC Inspection Report 259, 260, 296/87-20, 1
Subsequently, .another inspection was - performed in which .the :l
inspector reviewed all licensee -gene' rated Licensee Reportable
Event Determinations (LREDs), LERs and CAQRs/ CARS issued .after
December 1, 1986, in the area of fire protection. No recurrence
'e
of a similar event was identified.
"
The corrective actions taken by the licensee in response to q
Example A of the violation were considered. adequate.
2) Example B: Control Rod Drive Hydraulic Control Units not. -
Installed in the Design Support Mounting Configuration Required
by Design Drawings
This example of. the violation was the outcome of URI 259, 260,
296/85-25-01, which was closed and upg~raded to a violation for
failure to have hydraulic control' units (HCUs) mounted as
required by design drawing 919D615. The inspector found loose )
bolting, several free-standing HCUs, misa11gned channel nuts, j
j
and missing washers. )
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I
In the response to the violation, dated October 16, 1986, the l
licensee stated that the violation resulted from poor work I
practices and inadequate inspections conducted during the
plant's construction phase. For corrective action, the licensee
stated that they replaced all CRD HCU mounting bolts and that
the floor mounting hardware had been installed in accordance
} with design drawing 919D615.
The inspector followed up on the licensee's corrective actions )
by reviewing all MRs associated with the inspection, replace- '
ment, and torque work done on base mounting bolts. Also, the
inspector performed a walkdown of CRD HCUs, in particular those
for Unit 2, and verified that hardware installation was in
accordance with design drawing 919D615. Vertical back-to-back l
mounting and horizontal mounting of the HCUs were checked. A ;
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number of 3/8 inch and 1/2 inch base mounting - bolts, flat
washers, and lockwashers were found to be relatively new and
properly installed. Work had been completed in late 1985 and
1986. Per the review of MRs A-570904 and A-581910 (for Unit 2)
the 1/2 inch bolts were torqued _50 ft-lbs as recommended by
Unistrut, and the 3/8 inch HCV back mount bolts were torqued to
19 f t-l bs . For Units 1 and 3, the same work was performed per
'MRs A-570905, A-706829 and A-592255. The toroue work received
quality control verifications. No discrepancies were found and
the inspector considered the corrective actions to be adequate.
for Example B of the violation, and the issue is resolved.
This violation ' remains open pending NRC inspection of the third
example, which pertained to CREV mounting details. Resolution of the
final violation example is required prior to restart.
e. (OPEN) Violation (259, 260, 296/86-25-06), Failure to Maintain
Records of Facility Changes, Including the 10 CFR 50.59 Safety
Evaluation.
This violation resulted from a change to plant flood protection
features. Originally, flood doors to the Reactor Building and
Radwaste Building were normally maintained closed except for
personnel and equipment access as stated in the FSAR. In 1981, the
licensee changed the normal practice such that the doors were
maintained normally open. When questioned by the NRC inspector in
1986, no safety evaluation could be retrieved which would document
acceptability of the change per 10 CFR 50.59.
The licensee's corrective action consisted of reevaluating the
condition and performing a new safety evaluation. The NRC inspector
reviewed revision 2 of the safety evaluation, dated June 18, 1987.
The evaluation adequately justified changing the FSAR to reflect the
current practice of leaving the doors open. This change was made in i
Amendraent 5 to the FSAR in August 1987. The evaluation further l
recommended that the Bases for Section 3.2 of the TS be changed to !
delete the statement, " Plant flood protection is always in place and I
does not depend in any way on advanced warning." This statement was
not accurate under the current circumstances, which required operator
action to close the flood doors when required. As of October 18,
1988, this change had not been made. The evaluation additionally
recommended that an administrative instruction be developed to ensure
that operators close the flood doors whenever the Wheeler Reservoir
elevation reaches 558 feet. The plant responded by adding the
necessary operator action to Annunciator Response Procedure (ARP)
9-20. The NRC considered this to be inappropriate since the entry
condition into the procedure was the actuation of the " Lake Elevation
l High" alarm which occurs at 564 feet, 6 feet above that at which
l
operator action is required. This item will remain open pending
- resolution of tho above two outstanding deficiencies by the licensee.
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This item is acceptable for fuel load, because the plant will be in
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.the action statement (shutdown) required by TS upon high water'1evel
l- conditions. Final closure of the item is required prior to restart.
f. (CLOSED) Deviation (259, 260, 296/87-30-04), Failure to Maintain
. Written Justification for Changes to the FSAR.
In response to this Notice of Deviation, the licensee committed to
the following:
o Review the 1987 annual FSAR update to ensure proper justifica-
tion existed for each change and correct any changes which could
not be justified by the 1988 annual update.
o Revise the administrative procedure governing FSAR updates to
require formal justification for all changes. .
o Reinstate the commitment to perform a periodic examination of
the site surroundings to provide a reasonable representation of
area population and land use in the next FSAR.
o Submit a letter to the NRC describing the program for-
periodically updating the FSAR chapter which deals with area
population and land use.
The NRC inspector reviewed documentation associated with the
licensee's commitments and determined that they had been adequately
implemented. The licensee's review of the 1987 annual FSAR update
detected three changes which could not be justified. The NRC
inspector confirmed that these changes had been reinserted into the
FSAR in Amendment 6. The NRC inspector also confirmed that generic
implications for other TVA f acilities had been addressed through
issuance of a TVA corporate-wide Office of Nuclear Power Standard.
This standard (0NP-STD-6.1.6 Rev. O, Maintaining and Controlling
Safety Analysis Reports) contained specific guidance on periodically
evaluating and updating the FSAR chapter dealing with site
description, land use, and representation of area population. As a
final check on the adequacy of the administrative procedure Laverning
FSAR changes, the NRC inspector selected a sample of 16 changes made
to the FSAR in Amendment 6 issued in July,1988. The licensee was
able to provide adequate justification and safety evaluation for all
of these changes. This deviation is closed.
g. (CLOSED) Violation (260/87-37-04), Control of Measuring and Test
Equipment.
This violation identified the unauthorized adjustment of the zero
adjust screw on.TVA pressure gauge # E82214, which was being used in
'
a post-modification test performed on instrumentation associated with
safety-related systems. A licensee craf tsman performed the field
adjustment because the gauge was reading off-zero with no pressure
applied. The only authorized adjustment of this gauge was during a
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multi point calibration procedure traceable to the National Bureau of
Standards. i
When informed of the event, the Measuring and Test Equipment (M&TE)
_. Coordinator and the Unit 2 Instrumentation and Controls Section
Supervisor took immediate action to have the gauge retrieved and
tested. Additionally, a memorandum was initiated to inform all
instrument mechanics on proper use of gauges.
The NRC inspector reviewed the licensee's response to the violation,
dated January 15, 1988, and determined that the stated corrective
f actions should be adequate to prevent recurrence. The licensee
e
9
evaluated the circumstances associated with the violation and ceter-
mined that field adjustment of the gauge zero, although acceptable on
'
some types of M&TE pressure gauge, was not acceptable on those
manufactured by Wallace & Tiernan, Inc. The improper adjustment of
the gauge was attributed to inadequate training. Browns Ferry
instrument mechanics had since received additional training on proper
actions using Wallace & Tiernan gauges. Special caution tags have
been prepared for use with any M&TE pressure gauge that can- not be
zero adjusted in the field. Additionally, SDSP-29.1, Control of
Measuring and Test Equipment, was revised to include the requirement
to attach the special caution tags to all associated analog and
digital M&TE pressure gauges. This item is closed.
h. . (CLOSED) Violation (259, 260, 296/88-05-01), Failure to Control the
Issuance of Documents and Changes.
This violation identified the failure by the licensee to properly
control revisions to TACFs. Revision 1 to TACF number 3-88-001-111
was not properly reviewed for adequacy, approved for release, or
properly distributed. Similar problems were also noted on other
licensee TACFs.
The NRC inspector reviewed the licensee's response to the violation,
dated June 24, 1988, and determined that the stated corrective
actions should be adequate to prevent recurrence. The licensee has
corrected the deficiencies as noted in the original NRC inspection
report. Additionally, a licensee review of all open TACFs to verify
proper handling and correct documentation was conducted and all
identified deficiencies were corrected. There was also an orgoing
licensee program to reduce the number of outstanding TACFs, with a
goal of zero open for Unit 2 and common systems TACFs prior to
restart.
PMI - 8.1, Temporary Alterations, has been revised to clarify the
TACF revision process. Requirements have been included to distribute
copies of revised TACFs to appropriate organizations. This item is
closed.
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1. (CLOSED) Violation (260/88-05-02), EECW Check Valve Installed
Reversed.
This violation identified the licensee's failure to properly verify ,
'
correct valve operation by inspection following the maintenance '
activity performed under MR# 792717. EECW check valve 2-67-659 was
removed during MR 792717 and reinstalled backwards, resulting in the
inability of the north EECW header to supply . cooling water to
safety-related components. Existing licensee procedure, B F-3 . 2 ,
Quality Control Inspection Program, Section 5.2.1, contained examples
of activities that should be verified by: using QC hold points.
Satisfactory operation of a valve following maintenance was one of
the examples.
As corrective action, the licensee conducted additional training for-
maintenance personnel to review the event and emphasize the need to
understand and fully implement the appropriate procedures when
performing the work. HMI-51, Maintenance of CSSC/Non-CSSC Valves and
Flanges, was revised to include the requirement, as step 8.2.2.5, to
check for proper valve orientation whenever a valve is reinstalled in
a system.
The NRC inspector reviewed the licensee's response to the violation,
dated May 25, 1988, and additional supporting documentation which
verified the performance of the corrective actions. The licensee
evaluated the failure and attributed it to the following causes:
o Failure of maintenance personnel involved in reinstalling the
check valve to follow existing procedural guidelines to ensure
the valve was properly installed
o Failure of maintenance planners for the work activity to include
a verification step to check and document valve orientation in
the work instructions '
o Difficulty in determining actual direction of EECW flow in
piping adjacent to the valve location
The check valve was removed and reinstalled in the proper orientation
under MR 886541. Proper orientation was verified by the licensee as
part of the corrective action of CAQR BFP 880193. Additionally, an
NRC inspector observed the correct valve orientation. The licensee
has labeled the EECW piping adjacent to the valve to show actual flow
direction. The NRC inspector concluded that adequate corrective
action had been taken to prevent recurrence. This item is closed.
j. (CLOSED) Violation (259, 260, 296/88-05-04), Failure to Comply With
the PORC Composition Requirements of Technical Specifications and
Failure to Maintain PORC Meeting Minutes.
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The NRC inspector reviewed the licensee's response to this violation
and concurred with actions taken to prevent recurrence. The NRC
inspector attended a Plant Operations Review Committee (PORC) meeting
on October 18, 1988, and confirmed that the membership complied with
the TS quorum requirements. This violation is closed,
k. (CLOSED) Violation (259, 260, 296/88-05-08)., Failure to Provide
Adequate Training for Craft Personnel.
This violation resulted from an inspector followup of TVA's
implementation of the Browns Ferry Regulatory Performance Improvement
Program (RPIP) action items. The violation pertained to non-
compliance with procedure BF PMI-4.3, Specialized Training, in which
a certain number of 'craf t personnel were found to be delinquent in
receiving periodic general employee training (GET) retraining,
including regulatory compliance.
The inspector reviewed the licensee's response to the violation,
dated June 24, 1988, stating the reason for the violation and the
corrective actions taken to preclude further violations. The cause
of the violation was attributed to inadequate supervisory enforcement
of training attendance requirements. Corrective action included the
following: (1) Issuance of a memorandum by the Site Director
requiring action to correct delinquent GET training; (2) Development
by the Training Department of a training schedule to ensure personnel
attendance; and (3) Consolidation of retraining in regulatory
compliance (RPI 1.383), Introduction to QA/QC (GET 4), and Plant
Procedures and Instructions (GET 6) into one course.
The inspector followed up on the licensee's corrective actions by
reviewing their memoranda issued to correct GET absences and other
delinquencies in the training plan for the new consolidated course,
and the training schedule. The licensee implemented the revised
training schedule on July 3,1988, which required retraining on an
annual basis. Per procedure, delinquent individuals will be informed
in writing, and be required to attend rescheduled GET by the end of
the calendar quarter.
The corrective actions should be adequate to preclude recurrence of
the violation. This item is closed.
1. (OPEN) Violation (259, 260, 296/88-22-01), Inadequate Corrective
Action.
In January 1987, licensee management became aware that four
temporarily promoted shift engineers did not satisfy the minimum
qualifications delineated in Nuclear Plant Operator Training Program
procedure PMP 0202.05. Adequate corrective action was not taken to
disposition the issue. PMP 0202.05 required that the candidate for
the position of Shift Engineer must pass the shift engineer
j accrediting examination unless waived by the Chief, Operator Training
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, Branch, and the Plant Training Review Board or Accrediting
Subcommittee. In addition, there were four other permanently
assigned shift engineers for whom records could n'ot be found to show
thac their certification examinations were successfully passed.
In August 1988, the four temporarily promoted shift engineers
received and passed the accrediting examination for the Shift Opera-
tions Supervisor position and were interviewed by the Site Director.
The licensee subsequently searched for plant records to show that the
other four permanently assigned shift engineers had . passed their
accrediting examinations. The examination cover sheets were found on
microfiche, and this provided documentation to show that their
certification examinations had been successfully passed.
The NRC inspector reviewed the documented corrective action and
considered it acceptable to support Unit 2 restart. However, the
item remains open pending NRC acceptance and inspection, if
applicable, of the formal licensee response to the violation.
m. (CLOSED) Unresolved Item (259, 260, 296/86-06-08), Inadequate
Slope on Instrumen+. Sensing Lines.
The NRC inspector closed all aspects of this URI in July 1988 (refer
to Inspection Report 259,260,296/88-21), with the exception of one
instrument line which did not comply with the established minimum
slope requirements. The licensee performed an in-depth engineering
evaluation of the suspect line, and concluded that the as-built
configuration was acceptable. The evaluation assessed the entire
length of sensing line from the pressure transmitter to its
dead-ended termination in the drywell (the parameter being monitored
was drywell pressure). Although, a low point did exist in the line,
the geometric configuration would prevent buildup of condensate to
more than one-half of the sensing line's internal diameter.
Although, this could create an orifice effect in the sensing line,
there would be essentially no flow in this situation and therefore no
detrimental pressure drop. The inspector interviewed the engineers
responsible for the engineering evaluation to ascertain whether
corrosion from the~ standing water in the low point had been
considered and to discuss the evaluation in general. The inspector
j
concluded that the evaluation adequately assessed the as-built
i condition.
1
This concern was originally identified during the field work phase of
the modification, and the licensee demonstrated compliance with the
slope requirements or adequately evaluated any nonconformances prior
to completion of the modification. Therefore, no violations existed
l and this item is closed,
n. (OPEN) Unresolved Item (259, 260, 296,/86-28-02), Discrepant Scram
Valve Opening Times.
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The licensee discovered during the performance of Special Test 86-10
in July 1986, that several scram inlet and outlet valves delayed
opening 'for up ' to 20 seconds. This item was reviewed in NRC
Inspection Report 259, 260, 296/68-16, and the following issues
remained open at that time:
1) Acceptance criteria for scram pilot valve timing upon scrcm air
header blowdown should be established. The data already
accumulated should be shown to support compliance with this time
or followup tests should be performed to demonstrate compliance.
2) Either single rod scram testing prior to plant startup or scram
valve time tests prior to plant startup should be accomplished
for each scram solenoid pilot valve that has been refurbished in
accordance with the GE recommendations in Service Information
Letter (SIL) No. 441. This is to ensure HCU operability and to
detect any further anomalies.
3) The licensee should check the adjustment of all scram valve
opening air pressures which have indicated a potential for
noncompliance with the recommended spring tension settings in GE
The inspector determined that item 1) would be addressed prior to
restart by the performance of the post modification test for the
alternate rod injection (ARI) modification, which includes the
acceptance criterion of 15 seconds on scram outlet valve opening.
Item 2) would be satisfied by tne performance of single rod scram
testing during the Unit 2 power ascension program. Item 3) had been
completed and no values were known to be in noncompliance with the
recommended spring tension settings in GE SIL No. 373. This URI was
considered by the inspector to be adequately resolved for fuel load
but will require additional followup during the power ascension
phase.
o. (OPEN) Unresolved Item (259, 263, 296/87-26-03), RHR Pump Nozzle
Stress Exceeds Allowables
The licensee's Engineering Assurance (EA) organization identified a
deficiency concerning an assumption by DNE engineers that blanket
approval was authorized for nozzle load calculations to result in a
20 percent overstressed condition. The actual requirement was a
specific case-by-case justification, analysis, and approval of each
condition. The problem was identified during a review of the IE
Bulletin 79-14 calculations and was documented in EA Audit 87-13.
The NRC inspector reviewed the licensee's corrective action in
response to this finding, and confirmed that the generic implications
had been addressed (two additional examples were found) and
ccrrected. Long term corrective action had been addressed through
procedure changes and training. The licensee's EA orgtnizatior.
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verified acceptable implementation of the corrective action and
closed this item on August 26, 1988. However, no reanalysis had been
perf ormed on the specific calculation where the deficiency was
detected. The NRC will maintain this item open pending a reanalysis
and/or a case specific justification for the RHR pump nozzle in
question. The status of this issue is acceptable for fuel load but j
will require additional followup prior to restart. i
1
p. (CLOSED) Unresolved Item (259, 260, 296/87-26-04), SNM Control.
This item identified irregularities involving a shipment of Special
Nuclear Material (SNM) to another licensed facility. Violation 259,
260, 296/87-29-01, which identified the licensee's failure to perform )
an adequate physical inventory and follow TI - 14, was subsequently !
issued to address this issue. Resolution of this item will be l
tracted by the followup of the violation. This item is closed.
q. (OPEN) Unresolved Item (259, 260, 296/88-02-03), Long Term
Corrective Action and Interim Controls for FSAR Deficiencies.
The licensee's Nuclear Safety Review Board (NSRB) had identified that
the FSAR was so deficient that it could not be relied upon for the
purpose of making 10 CFR 50.59 safety evaluations and Unreviewed
Safety Question Determinations (USQD). The problem resulted from
inadequate controls over annual FSAR updates for many years. (Refer
to NRC Deviation 259, 260, 296/87-30-04 for details on this problem).
The licensee documented this condition on CAQR BFF 870088 and
prepared and approved an FSAR Update and Verification Plan (B22
87088827 007) as part of the corrective action for the CAQR. The
plan involved a review of many documents, including outputs from the
DBVP program, to identify required changes to the FSAR.
Additionally, the licensee planned to perform a review to verify the
accuracy of substantial statements in the FSAR. The target
completion date for this activity was July 22, 1990. During the
review period, deviations identified by the program will be
identified as CAQR's, if appropriate, and USQD's will be prepared and
approved by the PORC and Plant Manager. Issues which are identified
as being unreviewed safety questions will require approval by the
NRC.
The licensee recognized that in the interim period prior to
completion of the long term program, the FSAR could not be relied
upon for reviews of changes, tests, and experiments per 10 CFR 50.59.
On March 23, 1988, the licensee's Site Director issued a memorandum
which detailed the condition of the FSAR and required verification of
information by an independent source (such as DBVP design criteria
documents) when performing 10 CFR 50.59 screening reviews and safety
evaluations. SDSp 27.1 " Evaluation of Changes, Tests, and Experi-
ments" was revised to provide guidance on additional documents to be
used for conducting safety evaluations including TS Bases, NRC safety
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eval ~uations, ' licensing submittals to the NRC, Commitments and
Requirements Data Base, and NRC regulatory guides as committed to in.
the QA Manual.
The NRC inspector reviewed the licensee's training lesson plans . '"
associated with.10 CFR 50.59 evaluations and' found that the training'
- specifically highlighted these concepts. .The NRC inspector further *.
sampled : some of the more recent " Screening Review Forms for i -
Documenting. Applicability of a Safety Evaluation" per SDSP 27.1 to
confirm that other 1 sources were being appropriately referenced. Of
-the 25 screening review forms' sampled, none referenced sources other
". ."
than the FSAR and TS. During discussions with licensee engineers -and
management, the NRC inspector learned that many still relied solely-
upon the FSAR for ID CFR 50.59 information. The TVA licensing
organization ' acknowledged. .that ' additional corrective action was -
needed in this area; An impromptu training session was promptly held
for upper' site management. Changes were initiated to;SDSP'27.1 to
~
include,further_ explicit guidance and' controls. This URI has been
reviewed in detail by the NRC inspectors over. a several month time
frame and has been evaluated as being adequately addressed by the
licensee for fuel load operations but will remain open pending
additional corrective action in the interim controls area. This
issue along with the related issue discussed in paragraph Sa. will
require followup evaluation prior to restart.
No violations or deviations were ' identified in the area 'of Licensee
Actions on Previous Enforcement Matters.
11. Exit Interview (30703)
The inspection scope and findings were summarized on October 28,~1988,
with those persons indicated in paragraph 1. The inspectors described the
areas inspected and discussed in detail the inspection findings listed
below. The licensee did not identify as proprietary any of the material
provided to or reviewed by the inspectors during this inspection.
Dissenting comments were not received from the licensee.
Inspection Findings:
VIO 88-32-01: Failure to Follow Procedures For. Equipment
Tag-out and Independent Verification (paragraphs 2.b
and 3.b)
IFI 88-32-02: Review of system 82, Diesel Generator, RTP Results
(paragraph 7.b)
12. . Acronyms and Initialisms
ARI Alternate Rod Injection
ARP Annunciator Response procedure
BFEP Browns Ferry Engineering Project
BFNP Browns Ferry Nuclear Plant
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CAD ' Containment Atmospheric Dilution
CAQR Condition Adverse to Quality Report
CAR Corrective Action Report
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-CRD Control Rod Drive
CREV' Control Room Emergency Ventilation
.CS . Core Spray
CSSC Critical Structures, Systems, and Comporents
DCN Design Change Nctice '
DCR Design Change Request
DG Diesel Generator
- DNE , Department of Nuclear Engineering
DBVP Design Baseline and Verification Program
EA Engineering Assurance g
ECN Engineering Change Notice
EECW: Emergency Equipment Cooling Water
EGM Electric Governor Motor
EQ- Environmental Qualification
ESF Engineered Safety Feature
FCV Flow Control Valve
FPC Fuel Pool Cooling
FSAR Final Safety Analysis Report
GET General Employee Training
HCU Hydraulic Control Unit
HPCI High Pressure Coolant Inspection
HPFP High Pressure Fire' Protection
HVAC Heating, Ventilation, & Air Conditioning
IE. Inspection and Enforcement
I FI ' Inspector Followup Item
JTG -Joint Test Group-
KW Kilowatt.
LER .
Licensee Event Report
LRED
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Licensee Reportable Event Determination
LOCA Loss of Coolant Accident
MMI Mechanical Maintenance. Instruction
MR Maintenance Request
M&TE Measuring & Test Equipment
.NPP Nuclear Performance Plan
NQAM Nuclear Quality Assurance Manual
NRC Nuclear Regulatory Commission
l NRR- Nuclear Reactor Regulation.
NSRB Nuclear Safety Review Board
0I Operating Instructions
PMI Plant Manager Instruction
PMT Post Maintenance Test
PORC Plant Operations Review Committee
! QA Quality Assurance
l' QC Quality Control
l RPIP Regulatory Performance Improvement Program
RHRSW Residual Heat Removal Service Water
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