ML20195B237
| ML20195B237 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 10/21/1988 |
| From: | Carpenter D, Little W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II), NRC OFFICE OF SPECIAL PROJECTS |
| To: | |
| Shared Package | |
| ML20195B228 | List: |
| References | |
| 50-259-88-21, 50-260-88-21, 50-296-88-21, NUDOCS 8811010408 | |
| Download: ML20195B237 (25) | |
See also: IR 05000259/1988021
Text
l
- >
,
l
.
.y.m UNITED STATES 'l
f* '
1* E NUCLEAR REGULATORY COMMISSION
9' . iE '
REGION H
101 MARIETTA ST N.W. ;
\e'.... ATLAH7A. GEOROtA 30323
l
' deport Nos. : 50-259/88-21, 50-260/88-21, and 50-296/88-21
Licensee: Tennessee Valley Authority
6N 38A Lookout Place
1101 Market Street
Chattanooga, TN 37402-2SJ1
Docket Nos.: 50-259, 50-260 and 50-296 License Nos.: DPR-33, DPR-52, ,
and DPR-68 i
. Facility Name: Browns Ferry 1, 2, and 3
Inspection at Browns Ferry Site near Athens, Alabama
l
'
Inspection Conducted: July 1-31, 1988
Inspector':j 2 / /AP
Ri Carpfaterf Senior Resident Inspector Ofte Signed
Accompanying Personnel: C. Brooks, Resident Inspector
,
E. Christnot. Resident Inspector
'
W. Bearden, Resident Inspector
A. Johnson, Project Engineer
J. Yo k,.
/ 9nior Resident Inspector, Bellefolnto
Approved by. _. . 6h ^ MAY _, l
W. S. C tie ( Section Chief, ___ Date 5'igned i
Inspection Programs,
,
TVA Projects D.yision
t
~
SUMMARY
Scope: This rout.ine inspection was in the areas of operational safety,
maintenance abservations, surveillance testing observations, restart
test program, and licensee action on previous inspectior findings,
Plant Operations Review Committee, reportable occurrences, corrective
action program, condenser retubing, and General Electric con'.ractor
- recommendations
Results: One violation was identified involving failure to perf orm CAQR
generic reviews in a timely manner. One unresolved item was identi-
fied concerning the procedures controling keys for access to high
radiation (*1000 mrem /hr) areas. Four Inspection Followup Items
,
'
I (IFIs) were identified involving RHRSW corrosion, deficiencies
identified during the restart testing of lop /LOCA C, vaulting of
completed and approved test results, and adequacy of identifying at:d
closing out of significant hardware test exceptions.
,, All of these issues are to be resolved prior to restart,
i
! 8811010408 881025
,
'
G
ADOCM 05000259
PNU
-
\t
- _ _ - _ _ _ _ _ _ __
'
.
,
\ . .
.
-
t
l
!
l
REPORT DETAILS
1. Persons Contacted
fi Licensee Employees
- J. Walker, Plant Manager
P. Spiedel, Project Engineer
J. Martin, Assistant to the Plant Manager
- R. McKeon, Operations Superintendent
T. Ziegler, Superintendent - Maintenance
- D. Mims, Manager - Technical Services Supervisor
J. Turner, Manager - Site Quality Assurance
M. May, Manager - Site Licensing
- J. Savage, Compliance Supervisor
A. Sorrell, Site Radiological Control Superintendent
R. Tuttle, Site Security Manager
L. Retzer, Fi.e Protection Supervisor
H. Kuhnert, Office of Nuclear Power, Site Representative
T. Valenzano, Director - Restart Operations Center
- C, McFall, Compliance Engineer
Other licensee employees or contractors contacted included licensed
reactor operators, auxiliary operators, craftsmen, technicians, public
safety officers, quality assurance, design, and engineering personnel.
- NRC Attendees
- D. Carpenter
- E. Christnot
- C. Brooks
b
- W. Bearden
- Attended exit interview.
Acronyms and initialisms used throughout this report are listed in the
last paragraph.
2. Operational Safety (71707,71710)
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 NRC inspectors made routine visits to the control rooms. Observations
included instrument readings, setpoints and recordings; status of
operating systems; status and alignments of emergency standby systems;
onsite and offsite emergency power sources available for automatic
_.
_ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ -_
i .
,
. .
.
.
2
operation; purpose of temporary tags on equipment controls and switches;
annunciator alarm status; adherence to procedures; adherence to limiting
conditions for operations; nuclear instruments operability; temporary
alterations in effect; daily journals and logs; stack monitor recorder
traces; and control room manning. This inspection activity also included
numerous informal discussions with operators and supervisors.
Onge ng general plant tours were conducted. Portions of the turbine
bui' dings, each reactor building and general plant areas were visited.
Obse-vations included valve positions and system alignment; snubber and ;
hangea conditions; containment isolation alignments; instrument readings;
housek0eping; proper power supply and breaker alignments; radiation area '
controb; tag controls on equipment; work activities in progress; and
radiation protection controls. Informal discussions were held with
selected plant personnel in their functional areas during these tours.
!
The NRC inspector toured the residual heat removal service water (RHRSW)
pump building on July 12, 1938, and found the condition unacceptable. The
B3 pump which was in service at the time exhibited gross shaf t seal
leakage. This condition had been in existence since at least May 21,
1988, and was doc" ted on MR No. 869190. Another deficient condition
which was docume a on a tag hanging r,n the pump, MR No. 865395, since
May 23, 1988, wa. .essive vibration. The apparent reason that this pump
was in service was that of the 12 RHRSW pumps, seven pumps were
out-of-service for various reasons. General corrosion was evident on all l
piping and components with the most severe being the valve bonnets for the i
A1, 81, and B2 pump discharge valves. The licensee was asked to determine i
the minimum acceptable wall thickness for these valves and establish the
acceptability of both the current wall thickness and projected end of life '
wall thickness. This will be tracked as an Inspector Fnllowup Item (IFI)
(259,260,296/88-El-03) RHRSW Corrosion. Licensee representatives wers [
cautioned not to allow the flexibility and redundancy available with this l
system, particularly with only one unit to be placed in operation, to L
translate into a lack of aggressiveness in system maintenance. l
On July 15, 1988, the NRC inspector observed the controls established over h
high radiation areas which exceed 1,000 mrem /hr per BFN Technical !
Specification (TS) 6.8.3.2. Fuel reconstitutien activities released l
activated corrosion products which were deposited in fuel pool cooling [
(FPC) system components. As a result, a high radiation area greater than ,
1,000 mrem /hr was created around the Unit 1 FPC heat exchangers. Access I
to the area was not secured by locks with the keys under control of the
Shift Engineer. This condition is allowed by TS for a period of up to 30 !
days provided that the area is controlled by direct surveillance to (
prevent unauthorized entry. The inspector interviewed the high radiation ;
area watch and cofirmed that he was knowledgeable of his duties and [
responsibilities. The inspector observed that the area was properly [
posted and confirmed by review of the surveys and by independent measure- !
ment that boundaries were properly established, however, the condition I
was not properly annotated on the survey maps posted at the Radiation i
Worker Information Boards located at the entrance to the Radiologically
i
!
!
- _ _ - . _ -
_ _ _ _ _ _ . _ ____ _
-
.
. .
.
.
3
Controlled Area (RCA), nor was it listed as a "Miscellaneous Problem" or
"Unusual Condition" on the Shift Operating Supervisor's (SOS) Status
Board. The inspector questioned whether SOS permission was required for
entrance into the area since no key control was currently available.
Licensee representatives responded that 505 permission was not required
but would review the issue. The intent of '.he TS is that management, via
the Shift Engineer, be made aware of and exert positive control over each
individual entrance into a greater than 1,000 mrem /hr high radiation area.
The NRC inspector also assessed the licensee's routine program for control
over locked high radiation areas. The program is described in Browns
Ferry Standard Practice BF-19.26, Key Control and Accountability;
Radiological Controls Instruction RCI-17 High Radiation Area Door
Control; and Operations Section Instruction Letter OSIL-16 Keys. The
instructions were found to be contradictory and confusing. OSIL 16,
Section 3.1 requires high radiation area dour keys to be "under the strict
administrative control of the Shif t Operations Supervisor" (505) (new
designation for the Shif t Engineer); however, in Section 4.1 it also
allows high radiation area keys to be assigned to each control room,
assistant shift operations supervisor, RadCon, and Nuclear Security
Services with the SOS maintaining "accountability for them". Section 4.2
of this OSIL states that high radiation area keys may be used by
"Operations, Operations Training, and NRC Resident Inspection personnel
only". No allowance is made for RadCon or Nuclear Security use of their
assigned keys. RCI-17 requires in Step 6.3 that authorization to unlock
high radiation area doors will be obtained from the SOS, but in Step 6.4
it indicates that RadCon has pre-authorized use of their keys and need
only inform the Unit Reactor Operator and not the Shift Engineer.
The NRC inspector ascertained through interviews with RadCon personnel and
a SOS that high radiation area door keys were maintained by RadCon and
that the SOS permission was not sought or obtained for all entries into
the locked high radiation areas. The only control exercised by the 505
over the high radiation area door keys was a once per shif t acknowledge-
ment that the 505 clerk had performed a survey of all the keys and all
keys were accounted for. The inspector is concerned that the procedures
that control the keys controlling access to high radiation areas (11000
mrem /hr) are confusing and contradictory and this is identified as an
unresolved item pending clarification of this issue (259, 260, 296/
88-21-01).
3. Survaillance Observation (61726)
On July 12, 1988, during the performance of Surveillance Instruction (SI)
2-SI-4.9. A.2.a-2, Weekly Check for Shutdown Board C and D Batteries the
licensee declared the shutdown boards inoperable. The battery electrolyte
l temperature was found to be in excess of the 90 F acceptance criteria on
both the C and D shutdown board batteries. Later that same day, the Units
1 and 2 A, B, and C diesel generators were declared inoperable for the
same reason during the performance of 0-SI-4.9. A.2 a Weekly Check for
Diesel Generator Batteries. The licensee initiated CAQR No. 880470 in
order to document proper resol ation of the condition. The source of the
acceptance criteria on battery electrolyte temperature was the vendor
- - - - . - _ - - - - - - - - - - - - - - _
_ _ - ,
.
. .
,
.
4
manual which incivded a normal operating temperature range of 60-90 F.
The only adverse affect of higher temperatures is a 50 % reduction in
expected life for each 18 F above 77 F in the event of continued
exposure to the elevated temperature. The only limitation contained in
the vendor manual with regard to temperature is an absolute limit of
170 F during battery charging. The license initiated changes to all
appropriate sis to raise the upper temperature limit accordingly. The NRC
inspector followed the licensee's activities and confirmed the
appropriateness of the corrective action through a review of the vendor
documentation.
No deviations or violations were identified.
4. Plant Operations Review Committee (40700)
On July 14, 1988, the Plant Operations Review Committee (PORC) conducted a
meeting by telephone conference call at 8:00 pm. As requested, the NRC
inspector was notified and monitored the meeting by phone. TS 6.5.1.4
authorizes PORC business to be conducted by phone for expedited meetings
when it is not practical to convene as a group. Several PORC members
questioned the necessity of the telephone meeting given the issues on the
agenda. Two CAQR's requiring PORC approval in order to release
non-conforming material for testing and installation f rom the warehouse
were discussed (BFN 880476, BFN 880481). The telephone conference was
deemed necessary in order to meet schedules established for systems return
to service to support the targeted fuel load date. In both cases, the
deficiency which prevented release of the material was the lack of seismic
qualification documentation. The CAQR's clearly documented a prohibition
on considering the systems in which the components were to be installed as
operable unless and until the seismic documentation was obtained and
approved by design engineers. Part B of the CAQR's, however, took a
contradictory position to this. The "No" block was checked on both CAQR's
in answer to the question of whether the CAQR impacted unit operability.
This was a point of discussion among several PORC members, but all
subsequently approved the CAQR's as written. Several NRC observations as
a result of this meeting were discussed with the Plant Manager and members
of his staff during a routine weekly meeting. Among these concerns were,
1) Telephone conference PORC meetings should not be conducted as a matter
of convenience but should be reserved for events, in;idents or conditions
having true safety significance; and 2) The apparent inconsistency of not
checking the block marked "Yes" in response to the question of impact on
operability. These observations were considered by the inspector to be
isolated occurrences.
No deviations or violations were identified.
5. Maintenance Observation (62703)
During a control room observation on July 24, 1988, the inspector noted on
out-of-service amber indicating light for the 4KV shutdown bus no. 2
auto-transfer lockout relay 43-2. The inspector tracked the deficiency to
_ _ _ _ _ _ _ _ - .
,
~
'
.
'
. -
,
I
l
i i
MR No. 811066 dated December,1985. The current status was reported as !
being in the Electrical Technical organization for engineering evaluation f
- (since March 1988). Upon interviewing the responsible personnel, the i
- inspector learned that the MR was being turned over to operations for l
J post-maintenance testing and closeout following a visual inspection which (
i found no problem. The NRC inspector accompanied the cognizant engineer on ;
'
an inspection of the light and noted that the wrong light had been tagged ;
1 as being deficient. The engineer had been troubleshooting the light for !
] the 4KV shutdown board A transfer switch 435A. The orange MR sticker with !
the MR number was erroneously applied to an adjacent light. The deficient i
i condition went uncorrected for 2 1/2 years due to erroneous positioning of r
,
an MR sticker and poor trouble shooting of a properly functioning com-
'
ponent. This is considered to be an isolated breakdown of the MR process
j that occurred 21/2 years ago. Recent improvements in the licensee's i
- programs should preclude issues such as this in the future.
No deviations or violations were identified. l
,. :
6. Reportable Occurrences (90712, 92700) -
1
i The below listed licensee event reports (LERs) were reviewed to determine l
! if the information provided met NRC requirements. The determination t
i included: adequacy of event description, verification of compliance with (
) technical specifications and regulatory requirements, corrective action f
,
taken, existence of potential generic problems, reporting requirements I
1 satisfied, and the relative safety significance of each event. Additional
- in plant reviews and discussion with plant personnel, as appropriate, were i
conducted. l
] t
- (Closed) LER No. 296/83-04 Rev.1, Residual Heat Exchanger Tube Leak. A l
] leaking tube was found in the RHR 30 heat exchanger. Metallurgical i
,
examination revealed a circumferential1y criented crack in an area of the l
! tube where mechanical damage had occurred. Metallographic examinations !
did not reveal any evidence of corrosion assistance to the failure. Eddy f
- current testing was performed on 380 tubes of which 12 tubes were found to !
be mechanically damaged (dented). All 12 tubes including the leaking tube
l were plugged. The NRC inspector reviewed the completed work plans. [,
(Closed) LER No. 259/84-08 Rev. 1, Reported Failures of the Unit 1 High l
d
Main Steam Line Flow Differential Pressure Transmitters (DPT) 1-25 A l
l through D. The f ailures occurred during January and Februacy of 1984. l
l Following testing of the transmitter, the licensee concluded that the ;
j failure was due to the behavior of the pulse dampening devices (snubbers) {
- installed in the instrument sensing lines. The snubbers were removed and
Unit I was operated for approximately seven months with no further
problems noted. The snubber removal occurred under Temporary Alteration
] Change Form (TACF) 1-84-079-1. The temporary alteration was made
l permanent by ECN P0126 as implemented under WP 10370. The NRC inspector
i reviewed the completed TACF, ECN and work plan.
!
I
l
--
. _ _ _ . _ _ _ _ _ _ . ______- _ _______-_____________________ - _ _ _ -
l . .
.
.
6
(Closed) LER No. 259/84-15 Rev. 1, Removal of Under Designed Vacuum
Priming Valve. During a design review of the EECW system it was
discovered that the vacuum priming valve on the north header (none on the
south header) was under designed for the system pressure. This valve had
been isolated from the system for two and a half years and testing
determined that the valve's function was unnecessary. The NRC inspector >
reviewed Work Plan and Inspection Report No. 3026-87 and performed a field
inspection to ascertain that the valve had been removed and the piping
capped.
(0 pen) LER No. 296/85-17 Rev. 1, Failed Supports On the Residual Heat
Removai System. Three hangers located on loop 1 of the RHR system of Unit
3 failed due to high vibration caused by the throttling action (resulting
in cavitation) of an injection valve during shutdown cooling mode
operation. The licensee decided to reduce the vibration to an acceptable
level by replacing the original valve disc with a fluted disc. This has
been completed for Unit 2 and vibrational measurements have been made on
the Unit 2 torus to torus portion of the RHR system and found acceptable.
However, the vibrational measurements have not been made on a part of the
'
line that is used during the normal reactor cooling mode. The vendor does
not recommend running water into the reactor vessel when the fuel is not
present and the head is not on the vessel because of potential damage to
some of the reactor internals. The vibrational tests for the remaining
portion of the RHR piping will be made during post modification test (PMT)
No. 139 and will be required prior to Unit 2 criticality.
(Closed) LER No. 296/86-07, Loop I of RHR System Inoperable Af ter Two
Damaged Hangers Discovered. Hanger H3 had a 4-inch crack at a structural
tubing weld and hanger H8 had a support lug broken of f from the ceiling.
Both hangers are located on the 18-inch diameter RHR injection line above
the torus. The high vibration of this piping is caused by the throttling
action of loop I RHR injection valve 3-FCV-74-52 that occurs during the
shut down cooling mode. Metallurgical examination revealed that the parts
The NRC inspectors performed a visual
,
'
had failed due to fatigue.
inspection of the two repaired hangers. In order to reduce the magnitude
of the vibration, the injection valves were to be modified by replacing
the present valve discs with fluted discs. The modification of the valves
has been completed on Units 1 and 2, but not on Unit 3. The modification
of the valve on Unit 3 will be followed under LER No. 296/85-17.
(Closed) LER No. 296/86-09, Damaged RHR Hanger Causes Prohibited
Operability Configuration. Hanger H10 (loop I) located on the 24-inch RHR
shutdown cooling and low pressure coolant injection line above the torus,
had a crack in a load bearing tube steel member. Metallurgical analysis
of the crack showed a fatigue mode of failure. As noted in LER No.
296/86-07, this piping on Unit 3 (loop 1) has considerable vibration
caused by the throttling action of injection valve 3-FCV 74-52. The NRC
,
inspectors performed a visual inspection of the reinstalled hanger H10.
'
The modification of the valve on Unit 3 will be followed under LER No.
296/85-17.
I
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _-
-
,
.
- -
l
-
> ,
7
i (Closed) LER No.296/87-02, Unplanned Reactor Water Cleanup Isolation
During Testing Due to Fuse Failure. When chanael Al was de-energized
during a functional test, there was an unexpected isolation of the reactor ;
water cleanup system. This happened because channel B2 was de-energized I
due to a blown fuse. The functional test procedure was revised to add '
- steps which verify that the trip relays are energized at the start of the
'
i test. The NRC inspector reviewed the revised functional test procedure
and verified that it requires verification thot the trip relays are i
energized. I
(Closed) LER No. 259/87-08 and Rev.1, Failure of Potential Transformer
,
Fuse Contact Cause E1cetrical Fault and Engincering Safety Feature
Actuation. During the performance of a monthly surveillance test a phase
to phase short occurred between contacts in the diesel generator control l
cabinet for the 200 diesel generator (DG). The fault caused a refueling
! zone isolation, initiation of standby gas treatment and control room i
emergency ventilation, and on Unit 3 a half scram and primary containment l
1solations. The cause of the fault was a f silure of the potential 3
transformer fuse contacts. The fuse and spring finger contact were
bypassed on all eight DGs after an engineering ovaluation determined that
the fuse in the DG exciter potential transformer ci rcuitry was un-
necessary. The NRC inspector reviewed the engineering evaluation and the
completed work plans that bypassed fuse and spring finger contacts for all
eight DGs.
.
- (Closed) LER No. 259/87-22 and Rev. 1, Engineered Safety Feature Actuation l
'
Due to Personnel Error During Switch Calibration. During calibration of a '
raw cooling water pressure switch, two emergency equipment cooling water
pumps were inadvertently started due to a personnel error. The cali-
bration procedure was revised to provide an improved method of isolating
the switch during calibration. The instrument mechanics involved were
counseled on the need for increased caution when working with energized
j equipment. The NRC inspector reviewed the revised calibration procedure
1 and the documented counseling of the instrument mechanics.
, ,
1 (Closed) LER No.259/88-03, Inadequate Procedure Causes Inadvertent Start (
- of Emergency Equipment Cooling Water Pumps. During an attempt to put a i
'
raw cocling water (RCW) pump into service and the taking out of ser/ ice !
d
another RCW pump, the RCW header pressure dropped below the low pressure i
setpoint. The operating instructions for the RCW system were revised to !
provide instructions for alternating pumps in and out of service. A l
review of this event will be provided to current operations personnel. l'
! The NRC inspector reviewed the revised instruction and the event
description provided to the operations staff.
l
!
No deviations or violations were identified, j
,
,
,
7. Restart Test Program (RTP) i
- ;
! The inspector attended RTP status meetings, reviewed RTP test procedures,
observed RTP tests and associated test performances, reviewed RTP test !
,
i
I I
i
_ _ _ _ - _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _
. _ _ _ _ _ _ _ _ _ . - .- _ _ _ _ _ _ _ _ . _ _ _ _ _
-__ _ _ _ _ _ _ _ _ _ _ _ _ _ ____
.
.
. .
8
results (including test exceptions), and attended selected Restart
Operations Center (War Room) and Joint Test Group (JTG) meetings. The
following are the RTP activities and associated activities monitored, and
the status of testing during this reporting period:
a. Restart Tests Performances (RTP)
The following restart tests were in progress during this reporting
period:
RTP-03 A and B, Reactor Feedwater
'
RTP-023, Residual Heat Removal Service Water
RTP-030, Diesel Generator and Reactor Building Ventilation
RTP-031 A and B, Control Building Heating Ventilation and Air
Conditiening
RTP-077, Turbine-generator / Electro Hydraulic Control
RTP-067, Emergency Equipment Cooling Water
RTP-064, Primary Containment Isolation
RTP-070, Reactor Building Closed Cooling Water
l
RTP-073, High Pressure Core Injection
l RTP-074, Residual Heat Removal System
RTP-082, Diesel Generators
RTP-084, Containment Atmosphere Dilution
l
RTP-085, Control Rod Drive
I
RTP-092, Neutron Monitoring (SRM, IRM, LPRM, APRM)
RTP-099, Reac+or Protection System
,
RTP-357-1, 120 V DC Diesel Generator Batteries
i
! RTP-057-3, 250 V DC Unit Battery
RTP-057-5, 4 KV Distribution
RTP-BVC, Backup Control
RTP-L/L C, LOP /LOCA "C" Rev. 2
. - - _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _
. .
.
.
9
,
I
The above tests were either in the prerequest stages, system
performance stages, initial RTP Group reviews, DNE reviews, or final ;
'
JTG reviews.
b. Loss of Power / Loss of Coolant Accident (LOP /LOCA) Testing
During the LOP /LOCA series of tests a significant test exception was ;
identified involving the Unit 3 diesel g'enerators outpuc breakers 3A, .
3C and 3D. When performing LOP /LOCA C", these breakers, which j
supply standby power to shutdcwn boards 3EA, 3EC and 3ED !
respectively, locked out, i.e. would not close onto their shutdown .
boards, when requirbd by the LOCA signal. The licensee modified the !
breaker control logic circuits (See paragraph 8 of this report for
details). Also during the LOP /LOCA "C" test, a major electrical
switchboard (480V Reactor MOV Bd 2A) was not aligned as called for in !
the test pr1 requisites. The licensee initiated CAQR 88-0399 to
document this misalignment. These two items prompted the licensee to
perform a LOP /LOCA "C" re-test referred to as LOP /LOCA "C" Revision
2. This test called for turning off the incoming plant power using
the plant unit startup switchboard breakers, which are located in the
plant, instead of the three main power feeds coming from the
switchyards. NRC Inspectors monitored this retest from various plant
locations. To perform the procedure reviews, test witnessing, and
post test evaluations, the inspectors used the NRC 2513 program
inspection modules as guidance. No problems with the test procedures
were identified.
The retest was mainly for those items that did not test satisfac-
torily during the original LOP /LOCA "C" Test. However, a new
addition to the test was added and referred to as Section 5.5, Diesel
Generator / Paralleling System. This new test was to verify a design
i
feature that allowed for the paralleling of all eight (8) DGs with
of f site power while the LOCA signal was still present. The following
NRC inspector observations were made and discussed with the licensee:
(1) Two inspectors were stationed in the Unit 1 and 2 control rooms
and made the following observations:
1
- RHR Pump 2A breaker failed to close
- The following loads could not be verified due to their
breakers being racted out: 480 V Shutdown Boards IA and 1B;
250 V Battery Charger 1; Fuel Pool Cooling Pump 1A; Drywell
Blower 1A; LPCI MG Set IDN; and LPCI MG Set IEN.
I
- Low Pressure Coolant Injection (LPCI) MG Set 1EA should
have remained energized; it's breaker was found opened.
i
l
- Control bay water chiller "B" should have load shed; it's
breaker was found closed.
i
!
. -- . _ _.. _ _ _ _ _ . _ ,. _ _ .
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.
. .
.
10
-
Step 8.3.13 of 01-82 (restoring of f-site power to the 4KV
Shutdown Boards) could not be executed. The auto transfer !
lockout relay could not be reset (apparently due to the >
accident signal).
-
I & C Breaker 204 inadvertently tripped during the test. <
-
A blown fuse in the control power prevented operation of
the alternate feeder to 4KV Unit Board 2A.
-
A lack of communication and proper operational coordina-
tion was observed while conducting the diesel generator
paralleling operation. A major load was re-energized (480
Shutdown Board 2B) which produced a voltage transient while
attempting to establish proper synchronization. This came
as a complete surprise to the operator who was attempting I
to parallel the Unit 1 and 2 DGs.
'
-
Improvements were noted in control room ccmmunications and
control. The unit operator announced the receipt of
alarms, provided reports on completed actions, established
professional face-to-face communications through "repeat- ,
backs", and verified expected indications following switch l
manipulations. Additionally, a prompt review of loads lost
'
following r.n inadvertent trip of an I & C breaker was
conducted. Although not everyone practiced these e
'
techniques, the Unit 2 operators established and projected
a close team-work relationship using these methods. ;
-
The amber indicating light on panel 9-23-8 for the 4KV i
Shutdown Bus 2 auto transfar lockout relay 43-2 was out of
'
service with a December 1985 MR No. 811066 attached. This '
delay is considered to be unacceptable. Refer to paragraph
5 for additional information on this subject. i
(2) An NRC inspector was assigned to the Unit 3 control room and l
made the following observation: i
i
of
'
-
Operations personnel wer a knowledgeable testing
requirements and plant conditions, and were attentive to
their duties, The Unit 3 diesel generators were paralleled
to of fsite power in accordance with step 6.2 of BF 3-01-82,
Standby Diesel Generator System Operating Instructions; and
step 5.5 of 2-BFN-RTP-L/L-C, LOP /LOCA testing. No discre-
pancies were noted.
___ ____ ___ __ _ . _ _ _ _ _ _ _ . _ _ _ _ .
.
-
.
.
11
!
<
'
(3) An NRC inspector was initially assigned to the Unit 2 Auxiliary
instrument Room and made the following observations: ;
'
-
In order to indicate isolation signals had been transmitted
to Unit 2, the inspector expected to observe that 5 HFA ,
relays de-energized; however, only one HFA Relay (16A-K26) l
was de-energized. At the start of the test, the inspector ;
was told that a test exception would be written.
-
Equipment on the 4-KV Shutcown Board B performed as
required - No discrepancies were identified.
- Equipment on the 4-KV Shutdown Board D performed as !
required. 480 V Diesel /.uxiliary Board B transformer l
TDB ACB (compartment 13) was sucposed to remain closed; ;
however, only the yellow lir s* was "on" indicating an open
condition. ;
'
- The licensee performed these tests in a knowledgable and
professional manner QA inspectors were noted at each
location that the NRC inspector monitored.
(4) An NRC inspector was assigned to the 3EC shutdown board feeder
breaker from the 3C DG. This breaker had malfunctioned during !
the LOP /LOCA "A" and "C" tests. The inspector made the i
following observations: ;
- The normal feeder breaker opened upon the initiation of the
LOP signal; approxirately six seconds later the the standby
feeder breaker from 3C DG closed and then immediately
opened; the charging motor recharged the breaker springs; .
and, unlike what occurred during the original LOP /LOCA C"
test, the breaker closed and remained closed. The in-
l spector then observed the newly installed time delay relay
I
and it indicated that it had activated, timed out and ;
'
closed it's logic. !
,
- No deficiencies were ider tified at the 3E0 shutdown boards; i
diesel generator auxiliary boards 3A, 30. A Md B; and :
'
shutdown board B.
l
l
- At the Unit 3 control room, the paralleling of the Unit 3
! DGs with offsite power while the LOCA simulated signal was
l still present were observed. No deficiencies were
identified.
The failure of the RHR pump 2A breaker to close is identified as IFI
259,260,296/88-21-04, pending determination of why it ftiled and any
necesury corrective action.
At the beginning of the LOP /LOCA test series, several site
departments appeared to not give these tests the type of attention
,
- __ ___.__-. __ s ,. , _ _ _ _ _ _ _ - . _ , , _ _ , _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ____. _ ___ __ _ _ __
_ ___
i
. .
> ;
.
.
,
l
12
I
t
', needed; however, by the end of the testing series this attitude was
turned around and the final test was conducted with the type of pro- l
fessionalism expected at a nuclear power facility. .
t
c. Specific Test Witn'ssing - Reactor Protective Systen, !
,
The inspec+or reviewed 2-BFN-RTP-099, Reactor Protective System
(RPS), aw observed portions of the scram testing involving various >
scrams such as high reactor vessel pressure, high main steam line !
radiation, and low reactor vessel level. The Browns Ferry RPS
- consists of two trains, A and B, with each train sub-divided into two
channels A ' # and B , B . Each channel has the same number of
l 2 3 2 l
scram signals. This arrangement allows for a half scram logic if any .
one of the channels activate, and it takes a logic of one out of two '
trips occurring twice (one out of two, twice) to initiate a full l
Scram. The overall intent of the test is to calibrate each scram
parameter such as high reactor vessel pressure and low reactor level; *
check the functions of the reactor mode switch, such as refuel and l
startup; verify the scram logic from each scram signal; verify that
,
the reactor can be scramned remotely by turning of f the power supply
j electrical breaker to the RPS Motor generator sets (a total of two),
and co-duct time response testing for each channel. l
, ,
The NRC inspector observed the performance of Section 5.5 of the
test, Low Reactor Water Level. This section was verified using sis
2-SI-A,B,C and 0 (o* for each channel). The sis required that a ;
'
J calibration device be hooked up to the Rosemount level transmitters
j located in the reactor building and that a calibr.ition of the trans- >
l mitters be performed. The $! also required that the Rosemount Analog
- Trip Units (ATU) located in the Unit 2 auxiliary instrument room be ;
checked. The RPS testing will continue untti each scram signal has
been verified (a total of 39 signals). It should be noted that the l
- inspector Milized NRL 2513 inspection program modules as guidance in !
'
performin, thess inspection activities. No deficiencies were l
observed. L
[
d. Test Results Review ;
The NRC inspector reviewed the results of 2-SFN-A" ;~), Standay Gas f
Treatment System, which were reviewed by the .loi it Y ' t. Group and !
approved by the Plant Manager on July 6, 1938, r s tdst was ,
J reviewed by the inspector from the Baseline To .equirements i
Document (BTRO) to the final results. Act W field ti,ervations were f
conducted and were documented in previon residen+. inspection f
reports. Throughout the testing activities the inspector used NRC
,
i
,
'
2313 Inspection Prograr, Modules as ;uidance in performing the '
inspection activities.
It was noted by the inspector that a total of twenty-one (21) test
exceptions (TE) were de mentid by the Test Director, with five
i outstanding when the tr J. r$5ults were approved by the Plant Manager.
The follcwing outste-S v; 'ZS were reviened in depth:
4
- - - . - - - . - - - - _ . - . _ _ _ _ , . _ . _ _ _ _ _ _ _ _ _ -
--
.
. l
,
- . .
'
\
,
l
'
13
u
(1) TE-7 involved the zonal dampers located in stairwells and
elevator shafts separating the three unit reactor building zones >
.from the refuel zone. The test director initiated MR No. 779834
to repa,r these dampers. However, the FSAR indicates that
credit is not taken for these zona' dampers because credit is
taken for tk entire secondary containment, i.e. the three
reactor building zones plus the refueling zone. The BTRD was
revised to reflect this.
(2) TE-8 involved the four (two per train) dampers located in the
equipment access area located between Units 1 and 2. The test
requirement called for the dampers to close in ten (10) seconds,
however, all four dampers closed in approximately twenty-eight
(28) seconds. The test director initiated CAQR-880177 to
document this deficiency.
(3) TE-11 involved the attempt to document stack effect by blowing
smoke into the duct work located in the stack. This item was "
' discussed in a previous resident inspection .eport. The test
director initiated CA0R-880304 because of possible seismic
considerations with an unmonitored ground release.
(4) TE-20 involved a hold order on both A and B OG auxiliary elec-
trical boards, breakers numbered SC, which supply power to the
dilution fans.
(5) TE-21 involved an NRC Violation (259,260,296/88-05) which
documented the perforniance of iodine testing of the filters.
The inspector noted, on July 26, 1988 (20 days after final review and
approval), that the approved test results were not turned over to the <
i
QA vault in a timely manner. This item is identified as IFI
259,260,296/88-21-05, Vaulting of Completed and Approved Test
Results, pending the development of more timely administrative r
controls,
e. Test Exceptions ,
The inspector continued to followup the licensee's handling of the l
TEs identified by the RTP. As of mid July 1988, four hundred forty ;
four (444) TE's were identified, of which one hundred and one (101)
were still outstanding. The restart testing group has reviewed the !
TE! and have categorized them into six areas as follows.
(1) Equipment Deficiencies, which is subdh ided into equipment
malfunction (1.1) and equipment performance (1.2).
(2) Procedural 31fficulties, which is subdivided into procedure
errors /edi W rial (2.1); procedure method / performance (2.2); ,
plant condition / equip ~nt availability (2.3); and prerequisite /
initial conditions (2.4).
I
.
'
.
, . .
4
f
14
(3) Personnel Errors, which is subdivided into test director errors
(3.1) and support personnel errors (3.2).
(4) Partial Release (4.0), which is used when the JTG releases a
particular section of a tast for performance.
(5) Calibration Deficiencies, which is subdivided into measuring and
test equipment (5.1) and process instruments (5.2).
(6) Other
The inspector raviewed selected TEs from Test Procedures. 052-4,
057-5, 065, and 082 against this categorization. It was noted that
-
of the four hundred forty four TEs, ninety-two (92) involved
equipment malfunction (category 1.1) and forty nine (49) involved
equipment performance (category 1.2). The inspector observed that
maintenance requests (MRs), procedure changes (intent and non-intent),
and CAQRs were used to document and close out TEs. The inspector
expressed concern that there MRs, procedure changes and CAQRs be
given appropriate consideration in the final review and approval of
each test results package. The inspector will verify this in his
review of the approved test results packages. This is identified as
IFI 259,260,296/88-21-06. Adequacy of Closing Out of Significant
Hardware Test Exceptions (Categories 1.1 and 1.2.).
No deviations or violations were identified.
8. Modifications (37700)
The NRC inspector reviewed CAQR 880394 which documented the locking out of
3EA, 3EC and 3E0 DG breakers during the initial LOP /L"CA C test and
,
>
subsequently generated a design change request. This change involved the
installation of time delay relays manufactured by ASEA Brown Boveri in
i each of the eight (8) shutdown boards in order to allow the DG output .
> breakers to close on to their respective shutdown boards during a loss of I
powe- followed by a loss of coolant accident. The inspector also reytewed
the design change implementation which was accomplished by Work Package
No. 0095-88, Engineering Change Notice (ECN) E-0-P7150, Test Scoping
Document for test no. PMT-195, and Post Modification Test 195 (PMT-195).
The inspector observed the installation process involving shutdown boards
C, B, and 3EC The licensee did not install the time delay relay into
shutdown bos- D, because that board would be deliberately disabled for
spectors mon the installation of the time delay relays. The review
of the test coment and the PMT-195 indicated that the delay was
to be set for 2.o - or - 1.35 seconds, because the time to recharge the
closing springs of the breakers was to be two (2) seconds or less.
H)vever, the closirg spring recharging time on some of the breakers was
'
greater than two (2) seconds. The Post Modification Testing Group ini-
tiated a CAQR which resulted in changing the time delay to three (3)
seconds. During the LOP /LOCA C retest the inspector noted that the
'
modification worked as designed. The inspector will follow up on the
installation of this design on shutdown board D.
,
.. . . .
.
- .
...
.
15
No deviations or violations were identified.
9. Generic Applicability of Conditions Adverse to Quality
. The NRC inspector selected various CAQRs for review and identified four
CAQRs that had not received a Browns Ferry generic review within the time
frame required by the NQAM Part I Section 2.16, paragraph 10.5 which
requires that potentially affected organizations complete a generic review
within 70 calendar days from the origination of the CAQR. Specifically
the inspector noted the following generic reviews that were performed
late: .
l
CAQR Due Performed -
SQN 871347 11/1/87 2/9/88
SQP 871003 8/10/87 2/24/88 /
SQP 871066 7/15/87 4/5/88 ,
SQT 871347 10/15/87 2/23/88
The inspector determined that the problem with late generic reviews has
existed for some time. The inspector reviewed various Site Quality
Manager memos which identified many overdue generic review items. The
memos identifying overdue generic reviews have been issued on a routine
basis. The inspector was informed by the Site CAQR Coordinator that the
l number of outstanding late generic review items has decreased from 137
during January 1988, to 24 during July 1988, and that NQAM Part I, Cection
2.16 is in the process of being revised to re-structure the time limits to
l a more workable structure. The NQAM will also require escalation to
! higher management based on late generic review items. SDSP 3.7 will be
revised to reflect the new NQAM revision within 90 days of NQAM issue.
The NRC inspector believes the problem of late generic reviews has
received insufficient management attention. This has occurred inspite of
,
the fact that licensee QA personnel have documented the problem in 2
different CAQRs during 1987 (BFP 87 0830 and BFP 87 0326). Additionally,
separate violations in this drea are documented in NRC Inspection Reports
86-43 and 87-41. Although the late generic reviews were identified by the
licensee, timely and effective corrective action has not resulted. This
constitutes an apparent violation for failure to follow procedure,
Violation 259, 260, 296/88-21-02, Failure to Perform CAQR Generic Peviews.
10. Condenser Retabing
Ouring July 18-19, 1988, an NRC inspector observed ongoing work associated
with DCN M0075A, Unit 2 Main Condensor replacement of Admirality Brass
tubes with Alleghtry Ludlum AL-6XN stainless tubes. The observed work was
being accomplished in accordance with WP 2165-88. The inspector reviewed
WP 2165-88 and noted that it included a special installation instruction
which stated various requi-ements intended to control the tubing
installation work and prevent damage to the new tubes prior to instal-
lation. The inspeccor noted various poor work practices which were not in
- . - - _ .
P
., . ..
.
16
'
.
'
'.
accordance with the special instruction included in the work plan.
Spec.fically the inspector observed the following:
Unsupported sections of new tubing lengths of 25 to 30 feet Juring i
'
installation and storage at the work location. Special instruction
required no more than 10 feet of unsupported span.
Failure to clean the full length of-new tubes with freon during the
, actual installation process.
WorAers were observed walking on new tubes located in the tube
storage rack.
Sections of heavy wood lumber used as temporary walkways were drug ,
'
across new tubes located in the tube ttorage rack, i
The inspector discussed the observed poor work practices with licensee
management and the licensee agreed to investigate the event and take
'
corrective action.
The licensee responded by way of a Modifications Manager memo dated
July 26, 1988 (R06 88 0725 876). As corrective action the licensee agreed
to revise the workplan, counsel the craft involved with the job, and take
additional measures to protect the new tubes during the ongoing work. ,
t
Although the condenser retubing wor ( is not associated with' CSSC ;
components, there exists a significant concern that similar poor work
,
practices could exist on other jobs associated with safety-related systems ,
or components. [
No deviations or violations were iaentified.
,
11. Licensee Action on Previous Enforcement Matters (92702) ;
i
(OPEN) Violation 259,260,296/88-04-02, Procedures improperly encouraged *
use of temporary "scratchpads" before making formal operating log
entries. The licensee admitted that this practice did not comply with the i
Nuclear Quality Assurance Manual and recently re.: 3ed Standard Practice ;
12.24, Conduct of Operations, to eliminate this practice. Currently, the
procedure requires log entries to be made directly into the official
record at the time of the event. The inspector observed several operators '
over the cou-se of this inspection period and in all except two instances .
they were found to be in compliance with the new guidelines. In these l
two instances occurrences were not recorded in the log, even though
significant time had elapsed since they had occurred, and operators were
still using temporary "scratchpads." The operators did not appear to be ,
involved in activities that would hinder them from making prompt log *
.ntries. These cases were reported to the Operations Superintendent and
'
are considered to be example's of the continuing violation. This item l
will remain open pending furtner corrective action by the licensee and ;
additional :smpling for compliance by all operators. $
.
!
_ , __ ~ ~ _ _ _ _ _ _ _ _ _ . , _ _ ____
~
.
I
. .
.
17
(CLOSED) Violation 259,260,296/84-34-04, Failure to adhere to procedures
and inadequate procedures which contributed to an incident involving
overpressurization of the Core Spray Systam on August 14, 1984. One
example of this violation involved an SI step which erroneously designated
M0V Board 2A or 2B vice the correct MOV Board 1A or 18. This SI has been
corrected. The second example involved failure on the part of an operator
to properly open a circuit breaker as required by the SI. In response to
this violation, the licensee conducted operator training on proper breaker
manipulation, clarified the step in the procedure, and added second party
verification steps throughout the instruction in order to assure that
similar operator errors would be minimized. Further, the licensee ini-
tiated a change to the TS which would eliminate the need to perform these
types of tests at power and allow testing to be performed during outages.
Thus the potential for a recurrence of this type of event and pcssible
interfacing system loss of coolant accident is diminished. The NRC
. inspector reviewed the associated procedure changes and the TS change.
The TS change was approved by the NRC on February 12, 1988. This
violation is considered closed.
(Closed) Deviation 259,260,296/87-02-01, Curbs and floor drains in
battery rooms and battery board rooms. This item concerned Section 10.11
cf the FSAR requiring that each battery room and battery board room
contain drains and curbs. A tour by a NRC inspector was made of the
battery and battery board rooms for all three units to ascertain that the
curbs had been installed. It was also noted that (a) Unit 2 had floor
drains in both rooms, (b) Units 1 and 3 had floor drains in the battery
rooms, and (c) Units 1 and 3 have a 2 inch x 6 inch opening in the wall
between the battery room and the battery board room to allow the water to
flow from the battery board room to the battery room floor drain. This
deviation in considered closed.
12. Followup of Open Inspection Items (92701)
(Closed) Inspector Followup Item 296/83-19-03, Shutdown board room conduit
damage. During a tour of the Unit 3 shutdown board rooms, the inspector
noted that several electrical conduits extending through the 3E0 room
south wall to the outside area were displaced 2 inches lower from the
- ;osition as indicated on the applicable TVA drawing 45N888-12RA.
The inspector reviewed the TVA drawing discrepancy report associated with
this item. The licensee evaluated the condition and determined the
installation as existing as acceptable and that the physical location of
the conduit was incorrectly shown on the original drawing. The inspector
reviewed the revised drawing modified to comply with drawing discrepancy
package No. 3-86-0772 which shows the correct conduit locations. This
item is closed.
(Closed) Inspector Followup Item 260/84-41-04, Relocation of HPCI EGM
control boxes. The licensee had identified the need to relocate the HPCI
EGM control bnxes (Panel 25-49 Turbine Control Panel) due to the harsh
er.vironment of high temperature and high humidity in which the control box
1
.
-
.
18
was located. This item was opened to track corrective action until a
design change request (DCR) could be approved and the EGM control box
relocated.
The inspector reviewed ECN P3184, associated with DCR 2349, which called
for replacement and/or relocation of various components of the HPCI system
and determined that tne ECN was field complete and that the EGM control
box for the Unit 2 HPCI system had been relocated to a new location away
form the HPCI turbine skid. ECN P3184 had been written to upgrade the
environmental qualification of the HPCI System for Unit 2. The inspector
concludes that adequatc licensee corrective action has occurred to resolve
the original concerns as identified in the original inspection report for
Unit 2 cnly. This item is closed for Unit 2 but will remain open for
Units 1 and 3 pending future review of licensee actions.
(Closed) Unresolved Item 259,260,296/85-15-01, NDT curve out of date; no
surveillance required, and TS discrepancy between ur,its. The inspector
had identified three concerns during a review of Unit 1 TS. The concerns
identified were:
-
Out of date Figure 3.6-1 contained in TS
- No documented verification contained in G01-100-1, Integrated Plant
Operations, that reactor vessel shell temperatures were at or above
the temperature of curve no. 3 of Figure 3.6-1 as required by TS 3.6.A.2
- Inconsistency between Unit 1 and Unit 2 T: and Unit 3 TS for TS 3.6.A.1.
The 'qspector reviewed the documentation associated with the licensee's
reply to the identified concerns. The inspector noted that the following
corrective action had occurred:
-
TS for all units have been revised to remove the inconsis' /
identified above and to update Figure 3.6-1 for use until 12 LFPY
(effective full power years) of irradiation
- G01-100-1 has been revised to require that vessel shell and primary
wster temoeratures are greater than 180 F on all working indicators.
The inspector feels that the licensee's corrective actions are sufficient
to address the concerns as identified in the original inspection report.
This item is closed.
(OPEN) Unresolved Item 259,260,296/86-06-08, Inadequate slope on
instrument sensing lines. This item documented a failure to comply with
the targeted s' ope of 1 inch per foot during installation of Workplan No.
2040-35. Although 1 inch per foot downward slope is the goal, Construc-
tion $pecification G-60 allows as little as 1/8 inch per foot slope when
it is impractical to achieve the desired 1 inch per foot. The NRC
s
. . - . - . . . - . . -_ - .- .
.
,- . .
.
0
1
19
i
inspector accompanied the modification engineer on his final . walkdown
inspection and witnessed selected slope measurements in the field.
, Revision 14 to the Workplan added a QC holdpoint to verify slopes on all
'
sensing lines. Revisions 21, 23, and 25 to the Workplan documented rework l'
necessary to meet the sloping requirement. Only one line failed to meet
the minimum slope criteria of Specifications G-60. Field change reqLest .
(FCR)86-178 documented zero slope on line B-5 shown on drawing 47W600-58.
Approval of the FCR on April 16, 1986 accepted the zero slope.
The instrument line having zero slope serves three instruments; 9T-64-160A
High Range Drywell Pressure, PT-64-57B High Drywell Pressi:re, and
PT-64-5BB High Orywell Pressure. The NRC inspector located the instrument
line on July 20, 1988, and found that it was not impossiblo to re-route
the line and achieve some downward slope. No fur;her engineering analysis
or safety evaluation was performed by the licensee to justify acceptance
of this deficient condition. Tnis aspect of the unrosolved item will
remain open pending further justification to be provided by the licensee.
All other aspects of the unresolved ite.. are esosidered closed.
$ (Closed) Unresolved item 259/260/296/86-14-02, Tuinel inspections. During
,
an April 1986 tour of the CST tunnel, NRC inspectors noted corroding
,
support base plates, pipe clamps with missing or loose fasteners, and a
'
general deterioration of the tunnel. During an NRC tour of this tunnel in
June 1988, it was noted that the area had been cleaned, painted a9d some
repairs completed. However, in a discussion with the licensee and a
review of the area drawings it was determined that this tunnel and related
piping are not safety-related and chis unresolved issue is closed.
, (Closed) Inspector Followup Item 259,260,296/86-16-03, Usage of non-
licensed operators. The inspectors had identified a concern about the use
of unlicensed operators in the control room. Although unlicensed
operators had sometimes been used to maintain the unit operator log book,
BF-12.24, Conduct of Operations, specifies that the Shift Engineer is
,
'
responsible for insuring that proper records and logs are maintained and
that pericdic reviews are performed. These reviews shall be at least once
per shift, and documented by initials or signatures on each logsheet. In
the present plant condition with all units defueled, a licensed unit
operator is not required in the control room but only to be on site.
This area was reinspected in NRC Inspection Report 259,260,296/86-28 where
- the inspector noted that implementation of the shif t engineer reviews was
only being done 50% of the time. This item was left open in 86-28 pending
further review.
During recent routine tours of the control room, the inspector noten chat
the unit operator log indicated proper reviews in all cases. Additionally
the inspector was informed by the licensee that the practice of using
unlicensed operators fcr this purpose has been stopped and will no longer
be necessary. This item is closed.
,
,
- w
-
-r-%..%,-.--.#--- , -.,wr, --m%..~,m.,- , , , , . . , , , _
,, . . _. . - . . _ . --
. +.
l
. .
..
-
.,1
! 20
o .
y
I
(CLOSED) Unresolved Item 259,260,296/86-40-08, Failure to verify auto-
I matic initiation of primary containment isolation valves during
surveillance testing. This deficiency was identified by.the licensee's SI -
review and. upgrade program concurrently with the NRC's identification of
'
this finding. Valves 74-57, -58, -60, -61, -71, -74 and -75 were not in TS :
Table 3.7.A which TS 3.7.D.1 references as listing the isolation valves '
required t, be operable. These valves were listed in TS Table 3.7.F.
! Primary Cw.cainment Isolation Valves located in Water Sealed Seismic Class
i 1 Lines. Table 3.7.F is not referenced in the text of tFe TS which is
,
silent on the operability and survsillance requirements for these valves.
'
The licensee has revised their sis to asure that these talves will be
tested for automatic initiation and closure. The licensee has submit ed a
! revised TS table 3.7.A incl;/ing all primary containme'it isolation valves
required to be operable.
No violation will be issued for this item since the TS d.J not clearly
l identify the operability requirements for these valves, and the licensee
l has taken action to revise the TS and include the valves in the SI test
program.
l
,
This item is closed.
(Closed) Unresolved Item 259,260,296/86-43-02, Adequacy of CAQs Reviewed
i This item was associated with the adequacy of generic review of CAQs to
other nuclear facilities which TVA has in operation or under construction.
'
The concern was associated with 4 CAQs docur.. anted at Bellefonte and Browns
l Ferry Nuclear Plants and possibly applicable to other TVA sites. At the
- time of the inspection there did not appear to be evidence of an adequate
l generic review for the identified CAQs due to a lack of documentation to
l verify performance.
l
l The inspector has reviewed the licensee's response to the concerns
l
identified in the original inspection report and determined that a
violation did not exist. Subsequent to the December 1986 inspection
licensee ONE personnel have provided documented evidence that potential
generic reviews had been performed or that the conditions were site
specific and did not atoly to other sites. This item is closec .
(Closed) Inspector Followup Item 259,260,296/87-09-07, Security lighting
DG building walkdown. This item had been openec' to identify <arious
t
concerns observed by the inspector during a walkdown performed in the
security lighting diesel generator building. The inspector reviewed a
licensee's Fire Protection Engineer memoranduto dated August 24,1987 (R43
87-324 882), with the attached supporting documentation. The NRC
inspector conducted a tour of the building and noted the following:
-
There was no noticeable diesel fuel odor.
-
DG mounting frame bolts and battery mounting bolts were secure.
-
Firo extinguishers indicated current inspection.
-
There has been a general improvement in housekeeping in the building.
r-
i. : r
.
.
. .
,
. 1
&
21
.
A
The inspector noted that the applicable fire extinguishers and fire
. protection valves have been added to the respective fire protection
instructions to insure routine verification. The NRC inspector concluded
that sufficient actions have occurred to add:ess the concerns as identi-
fied in the original inspection reports. This item is closed.
p (Closed) Inspector Followup Item 260/87-09-08, Deficiencies identified
during a review of the operating instruction (OI) upgrade program. A
rather extensive list of procedure deficiencies and enhancements were
noted during the inspection and tracked as a single IFI. The licensee
addressed each item and initiated procedure changes as necessary. The
inspector reviewed the changes and considered this item closed.
(Closed) Inspector Followup Item 259,260,296/87-30-01, Missed surveil-
lance due to discarded chemistry composite samples. On August 26, 1987
after completion of the monthly surveillance on radiation monitor filter
activity, the sample was inadvertently discarded. The sample would
normally have been retained for a quarterly strontium composite as
required by TS 4.8.8.3. This resulted in an unrepresentative quarterly
sample to characterize the third quarter of 1987. The licensee determined
that the error was due to personnel error i.e., failure by a chemist to
properly table the samples resulting in another chemist mistaking it for
waste. This resulted in toe issuance of LER 87-023 to report the missed
surveillance.
As corrective action to prevent reoccurrence of this event, the licensee
has revised SI 4.3.B.2-2, Airborne Effluent - Particulate Filter Analysis
(Monthly Gross Alpha), to include the requirement to properly table the
sample and store in a proper storage location. The inspector was informed
by licensee management personnel that there have been no similar
recurrences of inadvertently discarding semples since SI-4.8.B.2-2 was
revised. The inspector concluded that the licensee's actions were
adequate to cddress the concern as identified in the original inspection
report. This item is closed.
(0 pen) Inspector Followup Item 260/87-42-05, Insulation cut away around
tailpiece valves75-646 and 647. A testing manifold installed on contain-
ment Spray Loop II is equipped with a three-/luarter inch test / drain
tailpiece containing the above two valves and comes so close to a longer
insulated pipe that part of the insulation had to be removed. The
clearance between the two pipes had been questioned. The inspector
performed a visual inspection of the two pipes and noted that the larger
pipe had only dead weight supports and could be easily moved horizontally
N by hand. The inspector questioned wh6ther in a seismic event the larger
non-safety-related pipe could move he. izontally and shear of f the smaller
safety-related pipe. A CAQR was written to have design evaluate this
condition. This condition will remain open.
- -
. ~
, ,
F # ,
. .
.
22
.
13. General Electric Contractor Recommendations
The NRC inspector continued to review the status of the licensee's
resolution / implementation of recommendations made by General Electric as
part of the system review program as discussed in NRC Inspection Report
88-16. In particular the inspectors reviewed the adequacy of the
licensee's classification of closed and outstanding open items for proper
determination as to whether the item was required to be resolved prior to
Unit 2 restart. Items classified as Category A are required to be
resolved prior to unit startup while items classified as category B
through E are not required to be completed prior to restart or'may not be
completed at all.
The inspector selected system 63, standby liquid control system, for
review. The GE systems review punchlist includes a total of 36 items
classified as category B, C, D or E. No items were classified as Cate-
gory A. The ' classification of each item was compared to the restart
criteria as defined in TVA system engineering memorandums dated August 10,
1987, and November 12, 1987 (R40 870810 976, R40 87110 997). The inspec-
tor did not identify any punchlist items which did not appear to be
properly classified. However two concerns were identified and discussed
with the licensee.
-
Several items associated with revisions to system operating
instructions, system drawings, and surveillance instructions were
classified as category B and showed no completion date under status
column indicating that the items were not complete. The licensee
stated that the significance of the revisions had not rated a
classification of category A fcr the items; however, the corrective
action in each case was complete. The open status was to reflect
penoing management review of each item.
-
Items 63-30, 63-33 and 63-35 were associated with control of GE
Design Specifications. Design Specifications are vendor supplied
design information provided for each GE system during plant
construction. TVA has not updated or controlled the GE Design
Specifications for Browns Ferry since initial construction. This
concern exists for other GE systems also and is documented as
outstanding items on other GE systems. The items are closed on the
system review punchlist and status is stated as "TVA cloes not Control
GE Design Specifications and will not be changed." The licensee has
no plans to revise the Design Specifications and licensee management
stated that 10 CFR 50, Appendix B, Criteria III, Design Control,
requirements are met by separate design basis information resulting
from the Browns Ferry Design Baseline and Verification Program
(DBVP). The licensee further stated that the GE Desi' *pecifica-
tions were used only as information and not as basis foi Jesign work.
The inspector will review this area during the upcoming inspection
periods.
_ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ . _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _
. . .
,
. .
.
.
.23
.
6
14. Exit Interview
-
The. inspection scope and findings were summarized on August 1, 1988, with
those persons indicated in paragraph 1 above. 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.
, Item Number Description and Reference ,
URI 88-21-01 Administrative control of high radiation area door
keys as required by TS 6.8.3.2
VIO 88-21-02 Failure to perform CAQR Generic Reviews in
timely manner as required by NQAM Part~1, Section
2.16.
IFI 88-21-03 RHRSV Corrosion
IFI 88-21-04 Deficiency identified during the retest of
LOP /LOCA C
IFI 88-21-05 Vaulting of Completed and Approved Test Results
IFI 88-21-06 Adequacy of identifying and closing out of
significant hardware test exceptions; Licensee's
TE Categories 1.1 and 1.2.
l
15. Acronyms and Abbreviations
- ATU - Analog Trip Unit
l
BFN - Browns Ferry Nuclear
BTRO - Baseline Test Requirement Document
CAQR - Condition Adverse to Quality Report
CSSC - Critical Structures, Systems, and Components
DCR - Design Change Request
DG - Diesel Generator
DNE - Department of Nuclear Engineering
DPT - Differential Pressure Transmitter
ECN - Engineering Change Notice
EECW - Emergency Equipment Cooling Water
EFPY - Effective Full Power Years
FCR - Field Change Request
FPC - Fuel Pool Cooling
IFI - Inspector Followup Item
JTG - Joint Test Group
LER - Licensee Event Report
LOP /LOCA - Loss of Power / Loss of Coolant Accident
MG - Motor Generator
MR - Maintenance Request
- - _ _ - - - - - .
-
g.---_.--_-_.-.-___----- , ,
-.
_, .< l
,. . .
. a i
L
24 -
..
$'
NQAM - Nuclear Quality Assurance Manual
OI - Operating Instructions
OSIL - Operations Section Instruction Letter *
' Post Modification Test -
PMT -
PORC - Plant Operations Review Committee
QA - Quality Assurance . '
. m'
RCA - Radiologically Controlled Area
Radiologically Control Instruction
'
RCI -
RCW - Raw Cooling Water
RHRSW - Residual Heat Removal Service Water
RPS - Reactor Protection System-
RTP - Restart Test Program
SDSP - Site Director Standard Practice
SI - Surveillance Instruction
SOS - Shift Operations Supervisor
TACF - Temporary Alteration Change Form
TE - Test Expectations
TS - Technical Specifications
'
.
e
4
e
f
\