ML20215K342
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| ML20215K342 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 06/09/1987 |
| From: | Bearden W, Brooks C, Ignatonis A, Andrea Johnson, Patterson C, Paulk G, York J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20215K262 | List: |
| References | |
| TASK-1.B.1.2, TASK-TM 50-259-87-14, 50-260-87-14, NUDOCS 8706250296 | |
| Download: ML20215K342 (21) | |
See also: IR 05000259/1987014
Text
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, **E' , . t NUCLEAR REGULATORY COMMISSION
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', iReportNSs.(50-259/87-14,.50-260/87-14,;and50-296/87-14
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TLicensee: Tennessee Valley. Authority '
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6N 38A: Lookout Place
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Doc'ket'Nos 50-259,:50-260,"and 50-296 ,
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' A' uLicense,Nos._.0PR-33,-OPR-52,'and DPR-68 ^
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Facil_ity Name: Browns-Ferry lNuclearPlant(
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In'spectior) at Browns Ferry Site' ne'ar Athens, ' Alabama -
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; Inspect 1on Coriducted: March 1 -' April 30,11987' s
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Da'te S'igned
T .G.L.Pau$,SeniorRegdettInspector .
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C.A.PattFrson,.Residebt'Ingector ~ Da'te Signed
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.C. R. Broo6t, Resident (,InspQtoh j
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, JJ. W : York [/ Resident Ir(ppepor '
.Bellefonte Nuclear PlantL
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04kkA Bea$en, Resident Itispgetor
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Bellefonte Nuclear' Plant.
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Date Signed
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. A,' H. Johr{fon, Project Erg 1tneer _
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. Approved.by: 'd;n e eh [[dJ!gf
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. A.- J. :Ignhtonis, Srection Chief, Inspection '
- ' Program's, TVA 'rojects
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8706250296 870615
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SUMMARY
Scope: This routine inspection was performed in the areas of operational
safety, maintenance observation, surveillance testing observation, reportable
occurrences, configuration management, Restart Review Board and Independent
Safety Engineering Group (ISEG) activity, welding modifications, the layup
. program and low pressure turbine disc cracking.
Results: One . violation was identified for failure to comply with the
operability requirements of Technical Specification 3.7.E, Control Room
Emergency Ventilation System.
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REPORT DETAILS I
1. . Licensee Employees Contacted:
H G. Pomrehn, Site Director
J~. G. Walker, Deputy Site Director
P. J. Spiedel, Project Engineer .;
*R. L. Lewis, Plant Manager !
J. D. Martin, Assistant to the Plant Manager i
*R. M. McKeon, Superintendent - Unit Two
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J. S. Olsen, Superintendent - Units One and Three !
T. F. Ziegler, Superintendent - Maintenance
D. C. Mims, Technical Services Supervisor l
J. G. Turner, Manager - Site Quality' Assurance '
M. J. May, Manager - Site Licensing
*P. P. Carier, Compliance Supervisor
A. W. Sorrell, Health Physics Supervisor
R. M. Tuttle, Site Security Manager ;
*D. Short, Project Management Configuration .
*B. R. McPherson, Technical Support Services .j
*A. J. Everitt, Mods Supervisor
*J. W. Shaver, Technical Support l
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*D. R. Gallien, Chemical Technical Support
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Other licensee employees contacted included licensed reactor operators,
auxiliary operators, craftsmen, technicians, public safety officers, s
quality assurance, design and engineering personnel. !
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2. Exit Interview (30703) l
The inspection scope and findings were summarized on May 1,1987,
with the Plant Manager and other members of his staff as indicated by an
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asteeisk.
The licensee acknowledged the findings and took no exceptions.
The licensee did not identify as proprietary any of the materials provided
to or reviewed by the inspectors during this inspection.
* Attended exit interview i
3. Licensee Action on Previous Enforcement Matters (92702)
(0 pen) Unresolved Item (259, 260, 296/86-25-11) and Violation (259, 260,
296/86-32-01)
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Both of. these : items- concern Technical Specification requir.ed flow rate l
testing .of . thel Control Room Emergency Ventilation System (CREVs). '
- Paragraph 5 of. this report, Operational Safety, contains an update on
these . items. ,They. remain open pending resolution of additional items i
raised during this inspection. l
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. (Closed) Violation (259/84-34-01) This violation was for inadequate
quali ty . control of rebui.it parts in a solenoid valve. During the.
followup 'after the . Unit 1 core spray system over pressurization . of
August.14, 1984, it was found that the air-operated testable check
valve (75-26) was being partially held open by its air actuator,
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The solenoid' actuator (8344 ASCO Series) had been improperly reassemtsled
during maintenance activities sometime in the past'and an incorrect pilot
valve insert had been used. The incorrect insert caused misoperation of
the; valve such that it operated the reverse direction from that expected.
No approved maintenance procedures were available for.the mechanic'to use
in ' reassembling-the valve. . The licensee had been using rebuild kits.from
power stores: to ' rebuild the 8344 ASCO series valves. Records indicated
that the - rebuild' kits .were ordered and stocked under inadequate QA
requirements although the original valves were designated QA . level II by
the; licensee. Copies of three procedures which were revised to resolve-
this problem were provided to the. inspector. Mechanical' Maintenance
Instruction (MMI) 51 was revised to state that it was a policy to replace
solenoid valves rather than-rebuild them. If . rebuilding was required, a i
detailed instruction must be written and approved to perform the work. 'A
- post-maintenance test is also-required. Standard Practice 16.4; Material,
Components, and Spare Parts Receipt, Handling, Storaye, Issuing, Return to a
Storeroom, and -Transfer, . was revised to detail receipt inspection i
responsibilities and certification of personnel. Standard Practice 16.2,
Procurement, war revised to detail procurement document preparation,
review, and changes. This item is closed.
(Closed) Violation (259/84-34-05) This violatior was against 10 CFR 50,
Appendix B, Criteria VII, for inadequate receipt inspection of solenoid
valve parts; .The power stores procurement information for these solenoid 1
valves was revised to reflect the latest vendor.part identification number !
on_4/12/85. The inspector reviewed a copy of this information provided by
the. licensee. Training for power stores personnel for receipt inspection
was conducted and completed on 3/2/85. Browns Ferry's Standard
' Practice 16.4; Material, Components, and Spare parts Receipt, Handling
Storage Issuing, Return to Storeroom, and Transfer; has been revised to
require receipt inspection of site and engineering change notice
procurements for QA levels I, II, and QA level III items by QC inspectors '
who report to the Site Quality Manager. This item is closed. ;
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(Closed) Violation (259/84-34-06) This violation was for three examples
of activities- affecting quality, which were not in accordance with plant i
drawings or procedures. These items related to the over pressurization of
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the Unit I core spray system on August 14, 1984. First, the Mechanical
Maintenance Instruction (MMI) 51; Maintenance of Critical System
Structures Components (CSSC)/Non-CSSC Valves and Flanges, did not contain l
adequate post maintenance instructions to ensure proper valve operation; !
mechanically or electrically. The solenoid valve on the testable check i
valve of the core spray system was replaced and correct operation !
demonstrated. The licensee stated that all remaining core spray, residual
heat removal, high pressure coolant injection, and reactor core isolation
cooling . testable check valves were inspected and no similar problems l
noted. MMI-51 was revised to include detailed instructions describing
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post maintenance test requirements. The inspector reviewed these changes
which apply to the systems noted above.
After the solenoid valve was replaced the wiring to the position
indicating lights was corrected to be as shown in the plant drawing. 1
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MMI-51 was revised to include verification of proper assembly and
operation. ;
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Finally the solenoid valve part number used in the testable check valve ;
was not as required by plant drawings. The new model was substituted by )
the vendor as a replacement in 1978. The licensee performed an evaluation i
and determined the new model to be an acceptable substitute. This was !
clarified by additional correspondence from the vendor in April 1985. The ]
inspector reviewed this correspondence. Procurement procedures were i
strengthened in 1982 in an unrelated matter and now require an evaluation ,
if deviations are noted from the procurement documents. This should
minimize the likelihood of recurrence. This item is closed.
(Closed) Followup Item (259, 260, 296/84-41-03) This item was to review I
the revision to plant electrical maintenance instructions (EMI) as
recommended by a task force reviewing problems with the high pressure ,
coolant injection (HPCI) system. The HPCI task force recommended !
including the Terry Turbine governor calibration procedure in EMI-36. !
EMI-36 was revised and upgraded into three procedures. These incorporated
the Terry Turbine governor calibration procedure using the vendor manual. l
The inspector reviewed the three procedures: Calibration of an Installed,
operating HPCI Turbine Governor Control System, ECI-0-073-G0V003;
Calibration of the Ramp Generator Signal Converter (RGSC) of the HPCI ,
Governor Control System, EIC-0-073-GOV 002; Calibration of EGM Control Box l
of the HPCI Governor Control System, ECI-0-073-GOV 001. These procedures l
were performed on a HPCI governor simulator in the Nuclear Diagnostic l
Section (NDS) laboratory. The procedures were found technically correct
and .resulted in a satisfactory governor calibration. Some minor i
procedural revisions were recommended by the NDS and incorporated into the
procedures. This item is closed.
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-(Closed). Open: Item -(259/84-44-01) .This : item . addressed several
deficiencies.noted'during the review of. methods used by plant personnel to
take reactor - . water conductivity measurements. The deficiencies : 'and
corrective. actions are listed below:
a. Technical.. Instruction (TI) 38, page 702b incorrectlyL uses' a-
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. . conductivity baseline reference temperature as- 35 degrees. C-
vice 25 degrees C as; required. -
- .TI 38 was deleted and new instructions were written . to. use
25 degrees C.
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b. Plant' operators using the demineralizer operating log,- (page- 702B,.
TI 38)' were not familiar with the . requirement for temperature.
compensation'in obtaining correct conductivity readings'.
TI.38 was delethd and the operators now only log readings from inline.
' instruments or have the chemical section take. measurements.
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c. _ An unapproved computer program was apparently being..'used by lab . ;
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-analysts- . to- cetermine temperature corrected conductivity
measurements.
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The procedure was numed '" Temp Comp" and. is now~ designated in; a PORC
approved procedure.
d. Lab analysts were .not familiar with Section 1100, TI 38, procedure.
for taking conductivity measurements on the Leeds and Northrup
conductivity instrument although the measurements are taken daily.
Specifically, they were not familiar with the requirements to use a
flow cell constant,-step 5.
' TI 38 was' deleted. Chemistry. Instruction 617 provides measurement
steps using' a dip cell (page 4, XII A5) and a flow cell (page 5,
XII B4).
e. Plant o'perators and lab analysts were not familiar with the flush
requirements for the reactor water cleanup system conductivity
measurements although these measurements are taken numerous times
daily.
Flush requirements were added to the procedures, CI 469, 469.1,
469.2, 469.4A, 469.5, 496.6.
f. In TI 38, page 107, the conductivity temperature corrective chart is
inadequate in that th.e y-axis data was apparently omitted during
copy. No official procedure observed had a correct chart.
TI 38 was deleted and replaced by CI 617, Figure 1. This is-the
correct chart and is adequate.
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g. -An unapproved data recording sheet was being used to record
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conductivity data 'The sheet was not part of any procedure.
Data sheets are now in CI 500, CI 617-1 and 617-2. ;
The above items- were not listed as a violation since the licensee was
. currently taking action in this area in accordance with the Regulatory !
Improvement Plant for the plant. The inspector reviewed -the applicable
procedures revisions. . This item is closed.
(Closed) Violation (259, 260, 296/84-44-03) This violation was for
failure to adhere to Radiological Control Instruction (RCI)4 (Periodic
Inspection and Maintenance of Radiological Emergency. Plan Equipment and
Supplies). The administrative controis for ensuring the current . and-
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correct contact of RCI-4 was inadequate. Control of emergency eoulpment
was divided between the health physics section and the radiological-
emergency l plan (REP) section. Cognizance over . the equipment .was
transferred and the revisions to implamenting procedure IP-17,' Emergency
Equipment and Supplies, were made to combine all emergency requirements
into one procedure. .The inspector. reviewed the procedure changes'and the
latest inventory sheets of emergency equipment. No problems were noted.
This item _is closed.
(Closed) Violation (296/84-45-03) This was a technical specification
3.6.E.1 violation for failure to have all jet pumps operable when in the
startup mode. The jet pumps were not operable due to two flow.
transmitters not valved in properly. The instrument index was revised to
identi fy requirements for alignment and operability checks of jet pump
differential pressure instruments following maintenance. Also,
Surveillance Instruction SI 3.6.E.1 was upgraded regarding requirements
for instrument alignment. This SI covers calibration and return to
. service of instruments including first and second party verificatior, on
all valves. Copies of the procedure revisions were provided to the
i n specto r.. This item is closed.
L (Closed) Violation (259, 260, 296/86-05-02) This violation'was against '
10 CFR 50, Appendix B, Criteria III and related to design control. Four
plant drawings were found to reference the incorrect design specification
for the separation, isolation, and identification of engineered
safeguards. Design specification 22A1421 was referenced instead of the
current specification 22A2809. The four drawings were revised to
reference the current specification. The inspector reviewed the current
h_ drawings. The licensee conducted an evaluation of the differences between [:
the design specification and determined that the integrity of separation
E requirements as applied to the plant design had not been compromised. The I Nuclear Performance Plan Vol III provides actions related to the
improvement in the design control process. This item is closed.
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(Closed) Open Item (259/84-48-03) This item was a concern that there was
-no: control to prevent mixing potentially incompatible greases when using~ l
Mechanical Maintenance Instruction (MMI) 17, Preventive and Corrective
Maintenance of 'Limitorque Operations, and Electrical Maintenance j
Instrumentation (EMI) 18, Limit Switch and Torque Switch Setting Procedure '
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for' Motor Operators. EMI-18 was revised to reference MMI-17 for gear
operator lubricant. MMI-17, Table A provides a table of valve number and
type of lubricant. Attachment A provides a data sheet for recording any
grease change out. The inspector reviewed EMI-18 and _MMI-17 for the
procedure revisions which direct all lubricant control to the single i
This item is closed.
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procedure MMI-17.
(Closed). Followup Item (259, 260, 296/85-57-08) This item was to review
the failure evaluation for the "B" diesel generator failure to start on
December 16, 1985. During troubleshooting, the licensee found a switch
stuck in a position that would not allow a fast start signal initiated
from the control' room to reach the auto start relays. The switch failure i
was attributed to normal component failure. After the switch was replaced
the surveillance instruction for operability was performed with no
problems noted. This item is closed.
(Closed) Open Item (259/86-40-14) Tracking and Closecut of NRC-forwarded
Allegations. The inspector reviewed the tracking system used by the
licensee to followup on employee' allegations. Periodic reports are
forwarded by the licensee to give a status of all TVA allegations. The
tracking g rid close-out system methodology seemed to be adequate to
provide for documentation review. This item fs closed.
4. Unresolved Items (92701) ,
No unresolved items were identified during this inspection period. l
S. Operational Safety (71707,7171.0)
The inspectors were kept informed of the overall plant status and any
significant safety matters related to plant operations. Daily discussions
were held with plant management and various members of the plant operating
staff.
The inspectors made routine visits to the control rooms. Observations
included instrument readings, setpoints and recordings; status of opera-
ting systems; status and alignments of emergency standby systems; onsite
and offsite emergency power sources available for automatic operation; ,
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purpose of temporary tags on equipment controls and switches; annunciator
alarm status; adherence to procedures; adherence to limiting conditions '
for operations; nuclear instruments operable; 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 the operators and their supervisors.
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General plant tours were conducted on at least a weekly basis. Portions of )
the turbine building, . each reactor building and outside areas were ;
visited. Observations included valve positions and system alignment;
snubber and hanger conditions; containment isolation alignments; i
: instrument ' readings; housekeeping; proper power supply and breaker;
alignments; radiation area controls; tag controls on equipment; work
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activities in progress; and radiation protection controls. Informal )
discussions were held ~with selected plant personnel in their functional i
areas during these tours. i
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Biweekly verifications of system status which included major flow path
valve alignment, instrument alignment, and switch position alignments were
performed on the RHRSW/EECW systems. On March 8, 1987, the inspector i
observed that all of the baseplate bolts on C1 RHRSW pump were loose. l
The system engineer took immediate action. to correct the situation in l
accordance with MMI-29, RHRSW Pump Inspection and Maintenance. The pump l
was declared inoperable until the bolts were properly torqued. These i
bolts have 'apparently been loose since the last overhaul completed on j
December 2, 1985. A similar concern on RHRSW pump baseplate bolts was
identified as a violation last month (259, 260, 296/87-09-01). Therefore,
no violation will be issued for these additional examples. The licensee
is responding to the violation which includes this additional concern. ;
For corrective action, the licensee has checked the torque on all remain- l
ing RHRSW/EECW baseplate bolts. )
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In the course of the monthly activities, the inspectors included a review
of the licensee's physical security program. The performance of various
shifts of the security force was observed in the conduct of daily
activities to include: protected and vital areas access controls;
searching of personnel, packages and vehicles; badge issuance and
retrieval; escorting of visitors; patrols and compensatory posts. In
addition, the inspectors observed protected area lighting, protected and
vital areas barrier integrity. Visual vehicle search of the NRC NDE Van {
was observed entering and leaving the protected plant area and found to be
adequate,
a. Diesel Generator Electrical Malfunction l
On April 20, 1987 during a routine diesel generator surveillance '
test, the 3D diesel generator relay protection circuit alarmed and
tripped the output breaker. The local operator notified the control
room of an explosion and smoke from the generator control cabinet. ,
The diesel was manually tripped and all electric circuits tagged and !
de-energized. The diesel had been running loaded at 2500 KW, j
1875 KVA for approximately 10 minutes prior to the trip.
The licensee's investigation of the event revealed that the original
fault occurred in the upper compartment containing the fuse drawer
for the exciter and fuel transfer pump potential transformers. The
line side of fuse stab for C phase was apparently worn out and making
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'littleornotcontactwhspthetest. began. An arc wJs drawn across- l
this ' contact',- which _eventudily' caused heat' damage to 'the cohnected ' j
cable's insulation. Thej sat and indulation_ damage a'pparently caused: 'l
- similar damage to the twv remainihg. phases resulting in 'a phase.to
phase f fault that damaged numerous coniponents,, wiring, and the cabinet ,
-structure.
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The moving portion of this'C 'phse contact kas found to be bW adf t
: down~ to where a'. gap approximattily 1/2 inch wide exists betweer * rh;l
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stationtry co'ntact. The station &M contact shows sign 6cf burni,.; F
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.and wear. The A phase and B phasq contacts'are very worn,Lwith the B ;
phase making ' very minimal. contadt 6n dhe p@e ,of the moving . contact. 1
Because.of the poor condition of these contacts, maintenanc_e requests
were written for an sins similar contacts on the.
remaining seven diesels,ncs $picHpn ' of all were found on some
similar conditions
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of the the other dieset .cortrol cabinets, .the licensee called in a j
4-hour non-emergency report / per.10-CFR-50.72 (b)(2)(1) on April 28, !
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1987. An inspector folloeup i item will be assigned to track i
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D completion of the -. failure evaluation and assessment . of generic
. ramification of tivis_ problem (259,260,296/87-14-01).
b. Unusual-' Event. - Suspected Fire Urf t Two Drywell
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On. April. 23,'1987, atT3:41 p.m. an Unusual Event was declared due to
. smoke being ; found in theMnit 2 drywell. .. The " phnt fire brigade
responded. Welding had been. in progress 'but 'no _ source of the' smoke
was.found. The source..of the smoke Wuld not be determined ' initially
and a news release Awa,s, made by. the licensee.i The: smoke ;was removed- 1
from the drywell threugn the drywell purge system. The source of the 1
smoke was identified ys a< portable _ power supply (powtr/nack) located i
outside the1drywellinext to the equipment accew piug. The> i
ventilation' flow 3cacNed the' smoke into _ the drywell. iThe Unusual
- Event was cancelled 'ht' SiO5 p.nt. of the' same .date.' No personnel:
injuries or damage tb safety eo'.sipment was noteO The rejetor is
defueled and the dryn11 equipmant access plugs are removW
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c. "antrol Room Emergency Nnt' nation ,
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In' July 1986, the inspector became concerned that the licensee's, I
method of testing thL Control Room Emergency Ventilation system I
(CREV) flow rate was not in acdordance with Technical Specifications l
(TS). TS 3.7,E.2.C requires that system flow rate shall be within {
plus or minus -(+/-) 19% design. flow rate when tested in accordance j
with ANSI N510-1975, Tht FSAR describes the design flow requirement l
to be 500 SCFM 'com;o' sed of.4135 SCFM for door ard damper leakage ' and * i
the remaining 365 SCfK for piping an'd elect ical penetrations ~ pius a ;
margin band. Unresolved item 259,??60, 296/86-25-11 described this j
issue which was iater upgraded to a ' violation (259, 260, i
> 296/86-32-01). la response to the violation, the license changed the j
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@q 1 . testing 1 method : contained in Surveillance Instruction (SI)' 4.7.E.5,
cjb; Control Room Emergency Ventilation System Flow Rate Test. During'the
first performance of this SI on March 1 and 2,1987, problems were
@K 6 encountered which, when analyzed, showed that as a. result of the
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# < deficient flow test method, indicated CREV flow rate was higher than
f % the actual flow rate. Although' the test method indicated about
oj . 530 SCFM, the actual flow was determined to be about 400 SCFM based
on the aew accurate test method. The CREV system flow rate .is
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MU, ' controlled by ' positioning a flow control. damper at the discharge of
n 'c theCREV blower. This damper had been adjusted. in the past by.
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Mechanical Test Section personnel as. needed to obtain 500+10% SCFM.
4 flow as, measured by SI 4.7.E.5. Thus, actual CREV flow has been below
'the TS requirements for as long as this method for flow testing has -
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been usea. The above discussion refers only to CREV train B since
the flow test method used on train A was different and provided'more )
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accurate results. Except for periods when the A train was inoper-
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~a ble, the CREV system could have fulfilled its intended function of 1
maintaining a positive pr :,sure in the control room. Timeliness of
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the licensee's- actions in response to this issue is also a concern.
t. Procedural deficiencies were raised in. July 1986, .a violation was
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issued in September 1986, but yet the actual low flow condition on
the B train of' CREV was not identified. until March 1987, nearly eight
s ' months after.the original concern. A violation for failure to comply '
with'TS 3.7.E.2.C is issued in this report (259, 260, 296/87-14-02),
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a Faulty , implementation of- the new ' testing methodology . on March 1,
( 1987,' created an additional situation of TS noncompliance. 'The
updated version of SI 4.7,E.5 was used as post maintenan'ce testing on
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CREV' train B in. order to declare the train operable. ,0nce train B
'T was operable, train A was taken out of service. for maintenance at :
6:20 a.m. on March 2,1987. .Later that same day, Mechanical Test l
personnel relized that the new test instrumentation. (a micromano- ,
meter)- n s not properly zeroed during the ' testing on train B l
conduKed at. 6:30 p.m. .. on the previous evening and as a result of !
this the flow control. damper had been improperly adjusted to provide
only about .400 SCFM. The improperly zeroed instrument had
,' errone.ously indicated an acceptable value of 474 SCFM. Thus, while m the A train was inoperable for maintenance, the B train was
simultaneously inoperable for an unknown low flow condition.
TS 3.7.E.4 prohibits reactor or- refueling operations with no CREV
units operable. This requirement was violated .since fuel movement !
was being conducted in the unit 1 spent fuel pool during this period. i
An - LRED (Licensee Reportable Event Determination) was issued on
March 2,1987, regarding inoperability of the B train due to low :
flow condition; however, due to a lingering evaluation of the LRED, i
the above event was not identified until much later. This discrep- )
ancy was identified by the licensee greater than a month after the 4
event occurred described in this paragraph. The inoperability of ;
both CREV trains is treated as another example of the previously <
described violation.
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5 6. , Maintenance--Ob'servation ' (62703)
Pla'nt ' maintenance' activities of selected safety-related systems and-
'm components were~. observed / reviewed to ascertain that they were, conducted in
accordance with requirements. The following items' were considered during
thi s L review: .the: limiting . conditions. for plant operations were -met;
activities were' accomplished using approved _precedures; functional' testing
'and/or calibrations were performed prior to returning components.or system
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to service; quality control records ' were maintained; l activities were
accomplished by qualified' personnel; parts and materials used were
properly. certified; proper clearance procedures were adhered to; Technical
LSpecification adherence; and radiological controls were implemented 'as
required.- ;
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Maintenance requests were reviewed to determine status of outstanding
jobs and to assure that' priority was assigned to safety-related equipment
maintenance which might affect plant safety. The inspectors observed the ,
below listed maintenance activities during this report period:
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- a '. Reactor. Building Roof Repair
..b. Unit'2 Safe-end Replacement
.c. Units 1 and'3 Layup Operations.
d. Unit 1 Turbine Rotor Rebuild Work at TVA Service Shop, Muscle Shoals.
.No violations or deviations were observed in this area.
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7. Surveillance; Testing' Observation-(61726)
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. Surveillance Instruction ~ Review and Upgrade
In the area of- Surveillance . Instructions- (sis) revision the. inspector
reviewed:the licensee's quality surveillance report performed by the plant
quality assurance. staff. The Browns Ferry (BFN) Quality Assurance staff
recently performed a technical review of BFN Surveillance Instructions.
The. subject report- entitled " Surveillance Instruction Review,
QBF-S-87-0066.was issued to update plant management of findings associated
with the review of sis'by Quality Assurance.
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The sis were in instrument
maintenance, chemical engineering, electrical maintenance, fire protec-
tion. The intent of the review was to select about 10 percent of the
430 sis that BFN had committed to review for technical adequacy prior to
restart. The .QA review was temporarily halted after about 5 percent of I
the~ sis had been evaluated. A Condition Adverse to Quality Report (QR) I
was written which ' concluded that the existing technical adequacy review
program was' inadequate. Also, for five out of six of the sis reviewed by 4
QA, a condition adverse to quality report was written. The program has I
been halted as corrective action is being developed. Initial plans are to i
review all sis again no matter what stage of review and/or upgrade has l
been' previously completed. The new program will be completed in phases. I
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The first phase involves reviewing the SI against,a verification checklist
which will guide- the' revision through the steps . necessary to assure
technical. adequacy. This checklist ' was committed to in the Nuclear
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' Performance Plan, Section'2.4, Procedure. Upgrades in. order to assure
consistency and documentation of results. The QA Survey found.that the
. checklist was not consistently used by all personnel involved in SI
' reviews and upgrade. The 'second phase is an independent review of the
-procedure by.a qual.ified reviewer. The third phase involves a walkdown of .
the. procedure in. the field with a' qualified performer, the cognizant
engineer and the procedure ~ upgrade writer. This step assures workability
of the procedure and ensures consistency between plant nomenclature, the ,
, . facility layout, and the work environment. The fourth phase is a.valida- {
tion of the procedure by actual performance.
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The cognizant engineer
must monitor .the validation performance. The SI review program has been
ongoing'since January 1986 and has been inspected-and addressed in inspec-
tion report numbers 259, 260, 296/86-05; 86-25; and 86-36. This program,
was originally conceived and implemented in early- 1986 with a July 1986
. target completion.date. The SI review and revision program implementa-
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tion' schedule continues to slip and based on'the above described findings
it appears that much of the past work will have to be redone. More -
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management involvement. in the implementation of. this program will be
required to assure satisfactory completion of this program.
8. Reportable Occurrences (90712,92700)~
The. below listed licensee events reports (LERs) were reviewed to
determine if the - information' provided met NRC requirements. The
determination ' included: adequacy of event description, verification
of compliance with Technical Specifications and regulatory requirements
corrective action taken, existence 'of potential generic. problems,
reporting requirements satisfied, and the relative ' safety significance of
each event. Additional in plant- reviews and discussion with plant
personnel, as-appropriate,.were conducted.
The following licensee event reports are closed:
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LER NO. Date Event
260/85-07 7-26-85 Containment Isolation
Due to Breaker Failure
260/85-16 12-6-85 Excessive brift of ,
Pressure Switches
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(ASCO)
260/86-01 1/31/86 Inadequate Procedures
Leads to Lapses in t
Special Requirements
for Use of Temporary 4
Lead Shielding 1
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259/86-03 8/29/86 Cable Fault Results in
Shutdown Bus Transfer
296/86-03 2/21/86 Personnel Error in
Voltage Measurement
Results in Inadvertent
Containment Isolation
259/86-09 3/17/86 Failures Experienced
with Reactor Building f
Ventilation Radiation
Monitor Circuits
260/86-09 6/27/86 Incorrect Performance '
of Local Leak Rate Test
296/86-10 10/17/86 Inadvertent Secondary I
Contair nent Isolation . 1
from Numerous Monitor ;
Spikes
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260/86-11 11/6/86 Reactor Protection
System Trips Due to
MG Set problems
260/86-13 10/24/86 Engineered Safety [
Features Actuation
260/86-15 1/2/87 Inadvertent Secondary
Containment Isolation
from a Failed Relay
Coil
The cause of the breaker failure (LER 260/85-07) was loosened bolts
holding the gear box assembly. The bolts were tightened and a complete
maintenance service was performed.
It was determined that all (unit 1, 2, and 3) ASCO (12) switches
(LER 260/85-16) for HPCI turbine exhaust had their internal
cylinder / piston supports inadvertently omitted during the original
manufacturing process. The manufacturer repaired the switches and they !
were bench tested prior to installation. !
Inadvertent procedure controls (LER 260/86-01) to ensure maintenance of ,
the safety evaluation for special requirements of keeping spent fuel pool I
gates closed while lead blankets were in use were used. The spent fuel
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gates were closed with a hold order. All outstanding temporary alternatives
were reviewed and the administrative procedure controlling temporary
alterations was revised to assure special requirement area implemented and
maintained.
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.The cause of the. bus transfer-(LER 259/86-03) was a cable being shorted to
ground. J An independent investigation by:Wyle Laborat'ories classified the
. cable. fault'as a random failure.
The inadvertent containment isolation was' (LER 296/86-03) caused by the
voltmeter ~ lead L slipping and momentarily shorting the power supply ~ to
. ground. Corrective action. consisted of a rewriteL of the . Surveillance
Instruction, critique of the' event' by the instrument maintenance person-
nel, and verifying corrective. functioning of the logic.
The . failure of 'the radiation monitor circuit (LER 259/86-09) included a
wiring ' discrepancy, a relay failure and an improper jumper installation.
The drawing was revised, the K2 relay was replace, and the personnel
involved.were cautioned.
The incorrect performance of the local leak rate test (LER 260/86-09)was
caused by an; error in the' surveillance test procedure. The procedure was
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technically reviewed and revised. ,
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The inadvertent secondary containment isolations (LER 296/86-10) were
caused by radiation monitor- spikes. The radiation monitor detector and-
convertor unit were replaced and functionally tested.
The motor generator (MG) set problems (LER 260/86-11) were caused by dust
build-up on the motor intake and exhaust vents. The MG set's intake'and
exhaust vents were , cleaned and placed on a preventative maintenance
schedule.
The engineered safety features actuation (LER 260/86-13) was caused by a
burned coil in the relay. The. failed relay:was replaced. i
Tho inadvertent secondary containment isolation (LER 260/86-15) was caused
by a failed relay coil. ~The failed relay was replaced.
The following licensee event reports were reviewed and remain open pending !
further review: '
LER No. DATE EVENT
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259/85-36 7/23/85 Ongoing 10 CFR Appendix "J" !
Reviews
259/85-53 12/17/85 Failure to Meet 10 CFR
Appendix "J" Criteria
9. Restart Review Board
The inspector reviewed the function of the Restart Review Board and
applicable procedures. The board was created as a subcommittee of the ;
Restart Task Force. The Task Force was established by the Manager of !
Nuclear Power on March 19, 1986, to verify the identification of problems
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'and initiate actions for resolution where necessary prior to restart. The <
Restart Review' Board assists .the Task Force in determining the '
required-for-restart status of'line items from the major tracking system
lists. :The. lists include the.following:
(1) Significant1 Condition Reports; (2) Corrective Action Reports;
4 (3) Discrepancy Reports; -(4) NRC Commitments; (5) Site Licensing
Tracking of :NRC Inspection Items; (6) Browns Ferry Commitments
Relative to Division of Nuclear Quality Assurance Audits and Other-
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TVA-sponsered Non-regulatory Reviews- by Offsite Agencies;
(7) Engineering Change Notices; and (8) Conditions Adverse to
Quality Reports'
The Review' Board Jis composed of five members appointed by the Site
Director. These members come from the plant staff, site engineering, site
licensing, and the Task Force, The board normally reviews each list every f
month for new items. The required-for-restart status is based on a . list i
of restart review criteria. The Restart Review Board procedure, Site )
Director's Standard Practice 7.2, was cubmitted on December 16, 1986 and
approved on March 25,.1987. ,
10. Configuration Management Program :
An inspector continued to review the licensee's ongoing Design Baseline
o Prog ram.' This program is designed to improve the configuration management
system at_ Browns Ferry by ensuring that the actual plant configuration is.
reflected on plant documents and conforms to the design requirements. In
particular, the inspector: reviewed the status of the issuance of the new
Configuration Control Drawings -(CCDs). ;
For the 47 systems involved to support Unit 2 restart a total of. 550 new
' drawings will result. ALL CCDs were originally scheduled to be issued by i
April 1,1987, however, delays due to manpower and hardware restraints
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have prevented completion of that effort. As of April 24, 1987, new
. drawings for 14 systems for a total of 107 new drawings have been issued. .j
This . includes .5 systems that were not considered necessary for Unit 2
restart. The licensee indicated that very little work was left and that i
the remaining CCDs were scheduled to be issued by April 30, 1987. When
issued each new CCD immediately supercedes that old "as constructed"
drawing. The CCDs will replace the "as designed" drawings only after
validation when each system evaluation is complete. No drawings will be i
considered validated in accordance with SDSP 9.2 until the system is
completely field verified and design evaluated. The evaluation process is
scheduled to start in April and be complete by June 30, 1987, and any ,
identified plant modification work is to be completed during the second !
half of 1987.
The inspector reviewed various recently issued CCDs for the Reactor Core i
Isolation Cooling (RCIC) and Residual heat Removal (RHR) Service Water l
Systems and performed walkdowns of selected portions of the Diesel Gen-
erator and Standby Liquid Control Systems. No significant discrepancies ,
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were note'd and the quality and functional utility of the. drawings appeared
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to be much better~than the drawings that were being replaced.
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11. Independent Safety Engineering Group
TVA committed to implement an Independent Safety Engineering. Group (ISEG):
in the Nuclear Performance Plan', Volume 3, Section II 1.2.7.1. The ISEG / '
per NUREG-0737, Item 1.B.1.2 is 'an additional . independent. group of a !
minimum of L five dedicated,- full-time engineers, located onsite, but
reporting offsite to a. corporate official who is not in the power s
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production management ' chain. The ISEG is to maintain surveillance of
plant-operation and maintenance activities to provide independent verifi-
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cation that these activities are correctly performed. The ISEG shall then
. be in a position to' advise utility management on the overall quali.ty and. i
safety of operation. i
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Although the ISEG is not yet staffed at Browns Ferry, procedures have been !
developed to implement the ISEG' process. These procedures consisted .of
' Nuclear Power Requirements . Procedure No. 0604.05 k ependent Safety
Engineering Group Evaluations, Rev. O. in whic$ che corporate policy,
responsibilities and requirements are outlined and a series of implement-
ing procedures. within .the Division of Nuclear Safety and Licensing
. (DNSL-ISEGI-6.1 Series . of Procedures). The implement.ing procedures were
found to be. very prescriptive and even provided boilerplate report
~ formats. The .ISEG process will consist of two categories; ISEG Reviews
and.'Surveillances. .ISEG Reviews are in-depth evaluations of a specific '
scope iperformed' by a team of reviewers whereas Surve111ances are daily
routine . reviews .off activities such as operations, maintenance, testing,
and11og reviews; The ISEG review subjects are selected by the Corporate
Manager for ISEG from a list of candidate reviews which come from various
sources. Surveillance activities are directed by the Site Lead Reviewer.
The .ISEG reviewsDappear to be structured very similar to inspections
conducted by - the NRC, INPO, and .TVA QA with entrance meetings, exit
meetings, and reports. The line organizations . are responsible for
implementing corrective action and officially respond to reports. :A
quick turnaround is assured by requiring ISEG reports to be issued within
30 days and ~ responses are required also within 30 days. The daily
surveillance activities conducted by the ISEG are to be documented in 9
monthly summary reports to the Corporate Manager of Safety. ;
One concern with the licensee's program has to do with staffing. The BFNP-
ISEG is to consist of three engineers onsite (a lead reviewer and two l
staff reviewers) and two engineers in the corporate office. NUREG-0737 q
clearly specifies five dedicated, full-time engineers, located onsite. l
The BFNP approach is questionable especially when considering th.e j
qualifications and positions of the BFNP reviewers. They are to be some- ;
what specialists (typically an engineer in a specific discipline with a l
minimum of three years of nuclear plant experience) as opposed to gener-
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alists with many years experience in various db 'olines. The licensee
polled six utilities with established ISEG groups and found that five of
the six had five or more engineers onsite with one utility even having
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eight onsite engineers. Given BFNP's past performance and 'considering
that this group'is part of an'overall. performance improvement plan, it is
Edifficult,to justify acceptance of less than the industry consensus' for
onsite engineers. This concern was addressed in a request for additional
information to the Nuclear Performance : Plan,. Volume 3 Browns Ferry-
Nuclear Plant, from D. R. Muller, NRR to S. A. White TVA's manager of
.. Nuclea'r Power dated 0ctober 21, 1986. ;
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Another concern stems from the prescriptiveness of.the implementing proce-
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dures. ' Valuable improvement in safety can- be achieved by allowing .an ,
experienced engineer to " follow his nose." With formal reviews and '
surveillance assignm'ents being made by ISEG supervisory personnel, it is
unclear how much time will be available to the ISEG engineer for such
independent inspection activities.
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Overall, the ISEG activities should be geared to maintaining awareness of
plant status and current activities. and performing reviews in order ^ to i
improve plant. safety. These. activities also allow the ISEG to' be in a
position to advise' utility management on the overall quality, safety, and
trends in plant operation. It is this advisory function (as detailed. in
NUREG-0737) which is not evident .from the TVA' program and 'ir.iplementing l
documents. It is .possible that the ISEG process could become merely !
another inspection group similar to Quality Assurance, Quality Surveil-
. lance, Nuclear Managers Review Group, Nuclear Safety. Review Board, the
Office of Nuclear Power. Site' Representatives, INPO, and' the. NRC. In the
absence ~of periodic. programmatic reviews of ISEG activities. and findin'gs
to assess the implication on safety of plant operation's, the ISEG could
become another routine program with routine findings and reporting. The a
above concerns were brought to the attention of the Manager,,ISEG, during
a meeting on April 2, 1987.
Although the ISEG reports through the licensing chain to upper management,
a strong communication link with the Nuclear Safety review Board-(NSRB) is
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-anticipated. The Manager, ISEG, will routinely report findings to the
NSRB and the NRSB intends to submit candidate topics for ISEG reviews to
the' Manager, ISEG. In addition, the ISEG ' lead site reviewer will be a
member of the NSRB Unreviewed Safety Question (10 CFR 50.59) Subcommittee.
This ISEG-NSRB coordination should strengthen both organization's ability
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.to keep top management informed of overall plant safety.
12. Welding Modifications (37700)
The welding program at the site was examined by two different NRC groups.
A brief summary of the areas follow:
a. Examination of Welds by Nondestructive Examination (NDE) Van
NDEs were performed April 13 - 24, 1987. The examinations were :
concentrated on the recirculation piping and safe-end replacement for i
Unit 2, but NDE was also performed on other samples. The following l
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tests were performed:
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Piping
Reradiographed '12 welds; magnetic particle inspected 4 welds;' ~ dye
penetrant inspected 18 welds; ultrasonically inspected 2 welds;-
.. visually inspected 27 welds; hardness' tested 18 welds; ferrite' tested
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L18 welds; a'nd reviewed 24 licensee radiographs of welds.
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Supports (Visual Inspection Only)
EECW System.- 41 welds; RHR System'- 53 welds; cable-tray supports -
25 welds.
Material Certification Test Reports for the weld filler material, .
safe ends, and piping used for the. Recirculation Piping project were' l
reviewed- for conformance to the applicable ASME Code and TVA..pur- -l'
chasing specifications.
No violations, deviations, unresolved items, or inspector . followup
items were. opened during this inspection. Details will be documented
.in Inspection Report Nos. 259, 260, 296/87-16. q
b. Review of Welding Program, Phase I
A. team of'eight people from the Office of Special Projects, consul-
.tants, and RegionLII specialists reviewed Phase I of the welding-
program ' at= Browns Ferry during April 20 - 24, 1987. phase I is a
programmatic evaluation of the welding performed at the . 'si te .
Several potential shortcomings were discussed with the licensee. At
the exit, it &;as stated that the group would further evaluate the ;
information submitted and the details of the evaluation would be'in
Inspection Report Nos.- 259, 260, 296/87-19.
Within the areas inspected, no apparent violations or deviations were~
identified.
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13'. Layup Program Status
The inspectors discussed the status of the layup program with the i
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licensee, The licensee is placing emphasis on the layup of the oldest
unit, Unit 1. The efforts on Unit 1 are being reviewed for applicability
to Unit 3 layup. Unit 2 is in the long cycle or "as is" condition and is
not being considered for layup, since it will be the first unit back
online. Following is the layup status for the various Unit 1 systems:
Layup Status - Unit 1 (dated 4/28/87)
Condensate /Feedwater Currently being vented and
drained. Should be under
dry air purge this week.
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Feedwater Heater Shellside/ Drained, valve problem. !
Extraction Steam Should be under dry air l
purge today. !
Condenser Tube Cleaning To be worked with hotwell l
layup.
Main Turbine Generator _ Being worked, majority l
complete, j.
Offgas Charcoal Bed Purge Procedure being prepared.
Plant Equipment Inspection In place and being worked for:
and Rotation Program -Core spray pump motor
-RHR pump motor -CR0 pump
motor -EHC pump motor -Cond. ,
pump motor -Cond. booster !
pump motor -Cond. Vacuum
pump motor -Raw cooling
water pump -Raw service 1
water pump -Raw cooling !
water booster pump -Reactor
recirc pump motor -Reactor recirc
M/G motor and generator -Recirc
MG set oil pump motor
Heat Exchanger Cleaning Program in place. 1
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HPCI Turbine Layup In place.
RCIC Turbine Layup In place.
HPCI Waterside Piping Being vented and drained,
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RCIC Waterside Piping Being vented and drained. !
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Motor Heater Operability In place on
-Core spray pump motor -RHR
pump motor -Reactor recirc ;
pump motor -Aux. raw cooling '
pump motor
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A tour was made to observe the setup for dry air purge on the unit 1 and 2
main turbine generator.
Since microbiologically induced corrosion (MIC) is of concern, the f
licensee briefed the inspectors on the control of activities in this area. !
Ninety-five (95) stainless steel welas in the Emergency Equipment Cooling
Water System (EECW) which is common to all three units have been radio-
graphed. Eight of the welds had MIC indications. Three of the eight
radiographs had positive indications and five radiographs showed suspect
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indications. . These . eight '~are. being ' further evaluated using UT and ,l
metallography. sThe residents will followup on this evaluation.
The: inspectors ' reviewed the radiograph for weld no. = T-EECW-2-BD-378... The
radiograph showed pitted areas plus a characteristic marking produced by
the- corrosion nodule. The following are the results of the radiography
. for.the detection of MIC:in the sample of welds for the EECW system:
-6 welds found with MIC indications ,
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-25 welds found with other 25 welding indications, lack of fusion, sugar-
ing, slag nonmetallic inclusions, incomplete excessive reinforcement- q
-2 welds found with both MIC and other welding indications l
-1 socket weld was found with no end gap between pipe and fitting- l
-60 welds were found to be acceptable
14. Low Pressure. Turbine Disc Cracking
In a meeting at NRC Headquarters in Bethesda, Maryland on January 6,'.1982,.
General Electric (GE) stated that some stress corrosion cracking emanating
- from keyways on Lcw. Pressure Turbine Wheels had been detected. As:a
result of the detection-of this condition, the Region II office Materials
and Processes Section followed the' inspection and repair of GE Low Pres-
- sure Turbine Wheels (see Inspection Report Nos. 259,260,296/83-22).
On April 29, 1987 a followup on this project was conducted by.the resident
inspectors when they toured. the TVA equipment ' repairs facility at Muscle
Shoals, AL. .A combination of TVA and GE personnel were repairing the. low-
pressure turbines for Units 1, 2 and 3. The inspectors saw the machining, .
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handling, and equipment necessary for removing.and replacing the wheels on i
the turbine shaft. j
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To prevent the stress corrosion cracks from recurring, the key ways are J
being machined. out, of the bore area on each wheel . After machining the !
keys out,. the bore is dye penetrant inspected and if any . cracks are .
. detected,' machining continues in increments until the cracks disappear.
'The acceptance criteria for the amount of material that can _ be removed
from the bore are that sixty thousandths of metal must remain as a buffer 1
zone between the maximum design diameter allowed. Six out of 148 did not l
have this required thickness of metal and had to be scrapped. These
refurbished wheels are then shrink-fitted onto new larger diameter shafts.
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