ML20150E530
| ML20150E530 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 03/24/1988 |
| From: | Ignatonis A, Paulk G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20150E524 | List: |
| References | |
| 50-259-88-02, 50-259-88-2, 50-260-88-02, 50-260-88-2, 50-296-88-02, 50-296-88-2, NUDOCS 8804010033 | |
| Download: ML20150E530 (40) | |
See also: IR 05000259/1988002
Text
50 Clog UNITED STATES
Do
/ NUCLEAR REGULATORY COMMISSION
'f fe
/ j
REGloN ll '
101 MARIETTA STREET,N.W.
f
'* ATLANT A, GEORGI A 30323
e
k ....,/
Report Nos. 50-259/88-02, 50-260/88-02, and 50-296/88-02
Licensee: Tennessee Valley Authority
6N 38A Lookout Place
1101 Market Street
Chattanooga, TN 37402-2801
Docket Nos. 50-269, 50-260, and 50-296
License Nos. OPR-33, DPR-52, and DPR-68
Facility Name: Browns Ferry Nuclear Plant
Inspection at Browns Ferry Site near Decatur, Alabama
Inspection Conducted: January 1-31, 1988
Inspectors: 6 O. Mad 4 3!d4/17
G.L.Pau%,SehprResidentInspector Date Sitned
Accompanied by:
E. F. Christnot, Resident Inspector-
K. D. Ivey, Project Engineer
C. A. Patterson, Resident Inspector
Surveillance Instruction Review Team Members:
W. C. Bearden, Resident Inspector
C. R. Brooks, Resident Inspector
- R. C. Butcher, Senior Resident Inspector, Grand Gulf
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A. H. Johnson, Project Engineer
l E. Lea, Jr. , Reactor Inspector
R. D. Starkey, Reactor Inspector
i NRC Contractor Assistance:
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Donald A. Beckman, Engineering Change Notices / Modifications
David H. Schultz, Engineering Change Notices / Modifications
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Approved by: 6 d -/3 - u a 3/ot 4 /XT
! A. J. Ignatofyi s,'Tec ion ChWtf Ode Sisned
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Inspection Programs,
TVA Projects Division
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8804010033 880324
PDR ADOCK 05000259
@ TQCTQ
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SUMMARY
Scope: This routine inspection was in the areas of: open inspection item
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followup; operational safety; maintenance observation, restart test program;
maintenance improvement; management meetings, covering Nuclear Safety Review
Board, piece parts, and Joint Test Group activities; Surveillance Instruction
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(SI) upgrade program review (team inspection); and engineering changes and
modifications.
Results: One violation of 10 CFR 50, Appendix B, Criterion V was identified
in the engineering changes and modifications area.
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REPORT DETAILS
1. Licensee Employees Contacted:
C. C. Mason, Senior Manager, Operations Center
H. G. Pomrehn, Site Director
- J. G. Walker, Plant Manager
P. J. Speidel, Project Engineer
J. D. Martin, Assistant to the Plant Manager
R. M. McKeon, Operations Superintendent
J. S. Olsen, Superintendent - Units 1 and 3
T. F. Ziegler, Superintendent - Maintenance
- D. C. Mims, Technical Services Supervisor
J. G. Turner, Manager - Site Quality Assurance
M. J. May, Manager - Site Licensing
- J. A. Savage, Compliance Supervisor
A. W. Sorrell, Health Physics Supervisor
R. M. Tuttle, Site Security Manager
J. R. Kern, Fire Protection Supervisor
H. J. Kuhnert, Office of Nuclear Power, Site Representative
Other licensee employees contacted included licensed reactor operators,
auxiliary operators, craftsmen, technicians, public safety officers,
quality assurance, design and engineering personnel.
- Attended exit interview
2. Exit Interview (30703)
The inspection scope and findings were summarized on January 29 and
February 5,1988 with the Plant Manager and Superintendents, and other
members of his staff. New items identified:
a. Violatica (260/88-02-04) ra;iors to follow Procedures, paragraphs
11.a.5 and 11.a.7.
b. Unresolved Item (259,260,296/88-02-03) Control of FSAR Updates,
paraaraph 9.a.5.
c. Inspector Followup Item (259,260,296/88-02-02) Temporary Alterations
Change Forms, paragraph 9.a.4.
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d. Inspector Followup Item (260/88-02-01) HPCI Valve Failure, paragraph
6.b.
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.
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3. Followup of Open Inspection Items (92701)
(Closed) Open Item, 259/84-07-07, Rosemount Failures. This item was to
followup on the probable cause of the Unit 1 main steam line high flow
differential pressure detector failures. This item was reported in
licensee event-report (LER) 259/84-08. Following extensive testing of the
transmitters, it was concluded that the problems were attributed 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 seven months with no further problems noted. Further details
are provided in revision one of LER 259/84-01 dated July 13, 1987. This
item is closed.
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(Closed) Inspector Followup Item, 259,260,296/86-25-09, Failure of High
Radiation Door Control. A violation was subsequently issued for failure
to control access to high radiation areas in inspection report 86-26. The
violation item number 259,260,296/86-26-09, was closed in report 87-28
after the corrective action was found adequate. This item is closed.
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(Closed) Inspector Followup Item, 259,260,296/83-56-03, Snubber Failures.
This item resulted from a review of the snubber inspection and testing
program for Unit 1, in particular the failure analysis that is performed
on failed snubbers. The inspector found that no specific guidance was
provided for site personnel to indicate all areas to be investigated when
a failed snubber was encountered. The licensee presented information in a
Commitment Closure Summary consisting of modified procedures SI-4.6.H-1
and SI-4.6.H-2, Functional Test of Hydraulic and Mechanical Snubbers.
The modification to procedure SI-4.6.H-1 is not pertinent to this IFI.
However, the modifications to procedure SI-4.6.H-2 improve the engineering
evaluation of each inoperable snubber by requiring that the Division of
Nuclear Engineering be notified of the failure (in case a reanalysis of
piping or restraint requirements is required); asking if the vendor needs
- to be contacted in case the failure is not obvious; and, requiring the
l initiation of a Condition Adverse to Quality Report (CAQR). There is not,
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however, a requirement to review previous test data to determine if
! unusual behavior or marginal but acceptable performace was exhibited.
- Additionally, no specific guidance has been incorporated to indicate areas
to be investigated when snubber failures are encountered. The licensee
committed to modify procedure SI-4.6.H-2 and Data Sheet 4.6 H-2-3 under
procedure change request SI-4.6.H-2-06 to add these requirements. The
inspector reviewed the procedure change dated January 1, 1988 and found it
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to be satisfactory. This item is closed.
l (Closed) Inspector Followup Item, 259,260, 296/86-25-12, Inspect SI Review
l Programs. This was a tracking item to evaluate the implications of an
oversight by the licensee during their upgrade of SI 4.7.E.5 where the
flow test method did not comply with ANSI N510-1975. This was an NRC ,
finding after the licensee had completed their upgrade and review of this
SI. No further examples of this oversight were identified and therefore
this item is closed and is considered an isolated case.
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(0 pen) Inspector Followup Item, 259,260,296/86-05-05 Inspect SI Upgrades.
This was one of the original concerns that were precursors to the SI
review and upgrade program. The concern was that some procedures were not
sufficiently detailed given the level of training by the technicians to
allow them to be followed as written on a consistent basis. The specific
procedure cited as an example of this has been corrected; however, the
validation step in the upgrade program is the key to resolving the issue
on a generic basis. Since weaknesses were found in the validation
process, this item will remain open pending further observation of sis
during the plant start-up phase.
(0 pen) Inspector Followup Item, 259,260,296/86-05-08. This concern was
updated and modified somewhat in Inspection Report 86-32. The issue
related to the approach taken by the licensee in correcting procedure (s)
which have a recurring problem with inadvertent ESF actuations. The
technique used in SI 4.2.A.10 and other cases is to avoid an inadvertent
actuation by manually initiating the ESF as a prerequisite to performing
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the procedure. The method masks personnel errors such as unintentional
electrical shorting or bumping of sensitive instrumentation. Since their
approach does not correct root causes it is not considered an appropriate
response to the concern. A permanent fix is being installed for SI
4.2. A.10 by adding banana plug test jacks; however, the policy for this
type of response in other areas is unclear. The plant management was
asked for a policy decision on this issue. It will remain open pending
, policy development.
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4. Unresolved Items *
One unresolved item was identi*ied in paragraph 9, regarding long term
FSAR update program.
5. Operational Safety (71707, 71710)
The inspectors were kept informed of the overall plant stat.us and any
significant safety matters related to plant operations. Daily discussions
l were held with plant management and various members of the plant operating
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The inspectors made routine visits to the control rooms when an inspector
was on site. Observations included instrument readings, setpoints and
recordings; status of operating systems; status and alignmants of
emergency standby systems; onsite and offsite emergency power sources
6<ailable for automatic operation; purpose of temporary tags on equipment
controls and switches; annunciator alarm status; adherence to procedures;
- An Unresolved Item is a matter about which more information is
required to determine whether it is acceptable or may involve a
violation or d.viation.
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adherence to limiting conditions for operations; nuclear instruments ;
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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 operators and
their supervisors.
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;
instrument readings; housekeeping; proper power supply and breaker;
alignments; radiation area controls; tag controls on equipment; work j
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activities in progress; and radiation protection controls.
discussions were held with selected plant personnel in their functional ;
areas during these tours.
In the course of the monthly activities, the inspectors included a review
of the licensee's physical security program. The performance of var ious
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shifts of the security force was observed in the conduct of daily ,
activities to include; protected and vital areas access controls,
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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. ;
During a monthly review of quality surveillance section survey results
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(QBF-S-88-0040, as an example) the inspector noted that plant security r
plan requirements are discussed in detail in the summary section of the
non-safeguard report. The inspector indicated his concern to the plant
, manager that safeguard information may be disseminated without proper
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review. !
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- On January 21, 1988, at approximately 3
- 50 a.m. the sump in the i
i Standby Gas Treatment (SBGT) Building was discovered overflowing with
the water coming out of the sump, onto the floor and ultimately out :
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j of the building. The SBGT building is a Radiological Controlled
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Area. Approximately 200 gallons flowed outside the building into the i
- yard drainage system (which discharges to the Tennessee River) and i
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onto a roadway. This was initially classified as an uncontrolled
, release of potentially contaminated water to the environment.
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The event was caused by a stuck float switch which controls the sump pump !
i and demineralized water flow to the sump. Demineralized water is added to
the sump as necessary to maintain a loop seal in the floor drain system.
1 The switch was jammed by pieces of insulation from a nearby work activity !
in a position that locked-out the pump and added demineralized water.
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Health Physics Technicians conducted radiological surveys of the
roadway, the entrance to the yard drain, the door to the Standby Gas
Treatment Building, and the floor of the building. The surveys
indicated no contamination above minimum detectable activity.
6. Maintenance Observation (62703)
Plant maintenance activities of selected safety-related systems and
components were observed / reviewed to ascertain that they were conducted in
accordance with requirements. The following items were considered during
this review: the limiting conditions for operations were met; activities '
were accomplished using approved procedures; functional testing and/or
calibrations were performed prior to returning components or system to
service; quality control records were maintained; activities were
accomplished by qualified personnel; parts and materials used were
properly certified; proper tagout clearance procedures were adhered to;
Technical Specification adherence; and radiological controls were
implemented as required.
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:
a. Discharge tunnels inspection
On January 5,1988, the inspector conducted a walkdown with licensee
personnel of regrouting activities inside of the Unit 2 circulating
water system discharge tunnel. Regrouting of the tunnel connections
was being performed to repair leaks between all three BFN discharge
tunnels and leaks to the environment. The work was conducted in
accordance with ECN P0985 and workplan no. 2240-87.
From the walkdown, review of documentation, and discussions with the
licensee's rept esentative, the inspector concluded that the work was
performed in accordance with approved work instrucions.
b. HPCI Valve Failure
Divers performed further inspection on the failed manual isolation
valve 1-HCV-73-25. This normally locked open valve provides
isolation between the suppression pool ring header and the High
Pressure Coolant Injection (HPCI) pump suction for maintenance
purposes. The valve disc was found separated from its stem during a
boroscopic examination in September 1987 as reported in LER 87-27.
The failure mode discovered by the divers was apparently corrosion
induced f ailure of the bolt heads on the key-way cover plate which
maintains the stem-to-di sc key in place. Since all of the core
standby coolant systems (CSCS) have a similar maintenance valve,
inspections were also perfomed on the Unit 1 Residual Heat Removal
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(RHR) and Core Spray (CS) systems. Although some corrosion was
evident on these bolts, all were in place and functioning. Work is
still continuing on inspection of the 1-HCV-73-25 shaf t and keyway
in order to define the full extent of the corrective maintenance
necessary. These manual isolation valves are not part of the Nuclear
Steam Supply System (NSSS) design and reportedly are not found at
other BWR-4 facilities. BFNP was not required nor had they ever
performed a flow path verification through this valve since the
normal alignment and test alignment for HPCI suction was from the
condensate storage tank (CST). The Unit 2 valves will be inspected
prior to startup and this item will be acked as an Inspector
Followup Item (260/88-02-01). This finding raised a concern that
the manual isolation valve position may not be reliably determined
when based on stem position only. The applicability of this type of
valve position indication needs to be reassessed by the licensee. ;
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No violations or deviations were observed in this area.
7. Restart Test Program (RTP)
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The inspector attended the RTP status meeting, reviewed RTP test
procedures, observed RTP tests and associated tests performances, and
reviewed selected RTP tests results. The following specific RTP L
activities were monitored during this reporting period:
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s a. RTP-023 Residual Heat Removal Service Water System (RHRSW).
Section 5.6 of the test stipulated that SI-4.2.H.1 Reservoir Level
- Monitoring Functional Test and Calibration be performed and be
included as Appendix P of the RTP. The SI was performed on
November 27, 1987, and signed of f on December 12, 1987. The RTP
Engineer obtained a copy of the completed SI and inserted the item .
, into the RTP. Section 5.9 of the RTP requiras a sump pump and level ;
i switch operability check; however, several of the level switches were i
- rusted out and require replacement. -
b. RTP-024-Raw Cooling Water (RCW). Section 5.6 of the test requires 7
4 that the RCW booster pumps demonstrate operability and Section 5.5
requires that the RCW pumps demonstrate operability. Section 5.6 was
successfully completed; however, Section 5.5 is awaiting repair and ,
calibration of time delay relays,
c. RTP-30, Diesel Generator Building and Reactor Building Ventilation ;
System. Section 2.0 Prerequisites were currently being verified.
Section 5.1 ventilation for the diesel generator buildings for all
three units is scheduled to start on February 1,1988.
d. RTP-032, Control Air /Orywell Control Air. The equipment access inner ;
doors, numbers 3 and 4, successfully completed the seven day
air-inflated test on January 22, 1988. One CAQR Number 87-0711 was
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written to document a test deficiency.
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e. RTP-57-4, 480 Volt Distribution System. Section 5.3 requires that
the 480 volt distribution system load shed and sequence when an
accident signal and DG voltage are present. Preparations for this
section was in progress and the critical item involved an air
handling motor and fan for shutdown board "B". The actual test for
the "B" DG is scheduled for the week of February 1-5, 1988.
f. RTP-57-7, 250 Volt DC Shutdown Batteries. The test was held up due
to excessive ripple voltage from the battery chargers. New filter
capacitors were ordered and installed. The test will resume in
February 1988.
g. RTP-67, Emergency Equipment Cooling Water. Section 5.8 requires that
the various temperature pressure control valves function to control
cooling water flow through various chillers. However, reveral valves
did not function, consequently maintenance work requests were written
to document the repairs needed.
h. Special Test 87-34, Control Rod Drive. This special test involved
diagnostic testing of each control rod. The test required that each
control rod be withdrawn; timed with adjustments made if necessary;
the hydrolic water fluid flow and pressure be recorded; and any
deficiencies identified. As of January 26, 1988, one control rod
mechanism would not function as required. This test was performed to
status the rod drive mechanisms prior to the performance of RTP-085,
Control Red Drive.
The RTP continues to utilize specific plant SI and/or specifir sections
thereof. A separate section of this report documents which sis or por-
tions of sis were performed.
As part of the Restart Test Program the RTP Test Engineers are using Plant
sis to meet some of the test requirements, test prerequisites and test
data. During this reporting period the following sis were utilized:
(1) SI-4.2.B-39A, Core Spray System Logic, Loop 1. This SI was not
performed in its entirety. Gnly those sections needed to support
2-BFN-RTP-82, Standby Diesel Generators were performed as requested
by MR 859763.
(2) SI-4.2.B-45A, LPCI - System Logic, Loop 1. This SI was not performed
in its entirety. Only those sections needed to support
2-BFN-RTP-082, Standby Diesel Generators were performed as requested
by MR 859763.
The above sis were also performed under plant conditions as they are. In
step 4.1.24 of SI 4.2.B-45A the SI stipulates checking interlocks on
valves FCV-74-52 and 53, at present both valves are out of service.
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8. Maintenance Improvement Program (62700)
The Maintenance Improvement Program (MIP) is discussed in Nuclear
Performance Plan Volume III, Part II, Section 4.0. The MIP was
developed to correct past programmatic problems in maintenance. Ten
points are covered by the MIP. A previous inspection was conducted of the
MIP during July 1987, documented in inspection report 87-27. This
inspection focussed on the overall program status and emphasis on
maintenance procedures.
Part of the items in the MIP are designated to be completed prior to-
startup of Unit two. The percentage completion as of 12/28/87 of each of
the ten points for the startup items is given below:
Area % Complete
(a) Organir-+1on and administration. 80
(b) Training. 58
(c) Facilities and tools. 55
(d) Procedures / programs. 72
(e) Materials. 32
(f) Work control. 59
(g) Maintenance information. 40
(h) Maintenance problem analysis. 63
(i) Radiological control. 95
(j) Monitoring and evaluation of maintenance. 46
Average Percent Complete 60
Also, the complete MIP consisting of startup items and items to be
completed after startup has an average percent completion of 52%.
The procedures upgrade for the MIP is divided into mechanical, electrical,
and I & C disciplines. The progress of the procedure upgrade was reviewed
including incorporation of vendor manual information into the procedures.
The progress was found as follows:
- Complete /# Required Vendor Manual Review
Mechanical 113/122 11
Electrical 69/160 4
Instrument & Control 28/45 0
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The inspector reviewed the lists of the electrical and the Instrument and
Control (I&C) procedures indicated as part of the improvement program i
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i phase 2. Of the 69 electrical procedures listed only 4 were Vendor Manual
(VM) certified and of the 28 I&C procedures listed none were VM certified.
The four electrical procedures reviewed were Electrical Preventive
Instructions (EPI) and Electrical Corrective Instructions (ECI) involving
the condensate head tank level switches and the fire protection electrical ,
storage batteries. These procedures referenced three vendor manuals and a
review of the manuals indicated that they were the correct manuals and
each manual was available for check out from the VM control section. '
Various typographical errors were noted in some of the sections of the
procedures and these were pointed out to the licensee's representatives.
> As additional electrical and I&C maintenance procedures are VM certified a
more detail inspection wili be performed.
The inspector reviewed the status of the program to upgrade the mechanical 7
corrective maintenance instructions (MCI). The inspector discussed the
methodology used with the Mechanical Procedures Section Supervisor and
noted that the process included: procedure technical ar.d style reviews;
mechanical maintenance technical and craft reviews; vendor manual
validation and reviews; and procedure validation during first time use.
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i The inspector noted that while most of the existing instructions had
been upgraded and many new instructions had been written, only 11 had been
reviewed for conformance to the vendor technical manuals at the time of ,
the inspection. The inspector reviewed five of the completed MCIs against '
the applicable ver. dor nianuals for incorporation of vendor information.
The MCIs reviewed were:
! * MCI-0-032-VLV002 "Drywell Air Operated Suction Flow Control
Valve (FCV) 32-62, 32-63, Disassembly, inspection, Rework and
i Reassembly" l
- MCI-0-032-VLV003 "Control Air Supply To Drywell Valve 32-332, -
Disassembly, Inspection, Rework and Reassembly"
- MCI-0-032-VLV004 "Control Air Supply To Drywell Check Valvo :
32-333, Disassembly, Inspection, Rework and Reassembly" ;
- MCI-0-070-VLV001 "Reactor Building Closed Cooling water,
Pacific 10" Gate Valve, FCV-70-48, Disassembly, Inspection,
Rework and Reassembly"
- MCI-0-023-PMP001 "Residual Heat Removal Service Water Pump
Room Unwatering Pump Disassembly, Inspection, Rework and
Reassembly"
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From review of the MCIs, the inspector noted that most of the work steps
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in the vendor manuals as well as appropriate caution and warning
- statements were included in the instructions. However, there were some
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apparent discrepancies: (1) MCI-0-070-VLVC01 did not contain detailed
valve lapping instructions as outlined in the vendor manual;
(2) MCI-0-070-VLV001 did not contain a step requiring personnel to ensure
that no residual pressure remained in the line prior to working on the
valve even though the vendor manual included this as a caution; and
(3) MCI-0-023-PMP001 did not contain steps requiring lubrication of seals
and 0-rings prior to installation in the pump as specified in the vendor
manual. These items were discussed with the licensee for follow-up
action.
The inspector concluded that the methodology and reviews utilized in the
procedure upgrade process should ensure that acceptable instructions will
be available to perform corrective mechanical maintenance activities.
However, the apparent discrepancies noted by the inspector raised concern i
. about reviews of the instructions against the vendo, manuals. This
concern will be reviewed during further routine inspections of the
licensee's Maintenance Improvement Program (MIP).
? The inspector discussed the status of the vendor manual program with the
cognizant supervisor. For phase II of the program 200 of 320 manuals have
undergone a technical review. Once the technical review is complete the
manual is compared to the applicable plant procedures. The licensee
plans to provide a letter to the NRC updating their schedule for
completion of the program. !
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9. Management Meetings (40701)
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a. Nuclear Safety Review Board (NSRB) !
The inspector attended the Nuclear Safety Review Board (NSFds)
meetings held on January 14 and 15, 1988, at the BFN site. This
, NSRB audit was conducted per the required Technical Specifications
l Section 6, Administrative Controls, Subsection 6.5.2, and commitments
made in the Nuclear Performance Plan, Volume 3,Section V. !
Operational Readiness, paragraph 8.0. The topics discussed during
the January 15 meeting included the following:
- (1) Radwaste - The NSRB discussion of the licensee'- plan to reduce
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the amount of discharge especially liquid wasto indicated that
the program should be placed under operations &cause of the
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1 (2) Unit 2 condenser; the retubing schedule for Unit 2 condenser was
discussed and that an aggressive program to reduce air
in-leakage will also be started at the same time,
i (3) Control Room - The annunciators do not have reflash. Many
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alarms with multiple inputs need to be reassessed for
I appropriate operator response. For example, on a specific
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occasion a fuel pool level came in as Hi; however, it was later
thought to be low because both have the same control room alarm.
Alarms in noncritical areas need to sound for only 20 seconds
with critical alarms sounding until acknowledged. A study is to
be performed prior to startup of Unit 2. The Safety Parameter
Display System (SPDS) was discussed and it was disclosed that
the commitment was to complete the installttion during the next
cycle; however, the study for the SPDS is not as far along as it
should be.
(4) Temporary Alterations Change Forms (TACF). There is a large ,
backlog of TACFs and Hold Orders. For Unit 2, there are
approximately 200 TACFs, with approximately 90 being safety-
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related and between 700 and 800 for all 3 units. There are
upwards of 600 hold orders for all three units. In the past tne
TACFs system was abused; however, new controls are in place and
only eight were approved last year. .This large number of TACFs
calls into question the effectiveness or status of the configu-
ration control program. The large number of Hold Orders makes
it very difficult for the operators to control system lineups
and system rearrangements. The Change Control Board and Manage-
ment must work together to get the TACFs situation resolved;
however, at the present there is no identified schedule to
resolve TACFs. This item will be tracked as Inspector Followup
Item (IFI 259,260,296/88-02-02). '
(5) Final Safety Analysis Report (FSAR) - Controls over the annual
FSAR update have been deficient in the pest. This has resulted
in an FSAR which cannot be relied upon for 10 CFR 50.59
purposes. The NSRB concluded that safety evaluction required by
10 CFR 50.59 must be only partially based upon the FSAR with
supplemental validation required by the use of other licensing
documents. The USQD subcommittee was not aware of additional
. training required of the USQD preparers or reviewers to ensure
i that this dilemma is fully understcod on a site-wide basis. A
pieliminary schedule of three years has been established for
updating the FSAR. An Unresolved Item will be initiated to
track this program and ensure ARC concurrence is obtained on
the long term corrective action and interim controls (UNR 259,-
260,296/88-02-03).
(6) Implementation of new PORC procedures - The NSRB reviewed PORC
activities in light of the recent technical specification
changes. The change eliminated one of the majur burdens on ,
PORC. The "qualified reviewer" concept was expected to allow
P0nC to focus on events and issues with true oparational
consequences. Concerns were expressed regarding ti.e excessive
number of PORC alternate designees and the lack of consistency
i of PORC reviews that may result from this. ,
!
--r-- - -
. _. . . _ . . , . _ , . . , . . , , . _ _ . , . . _ . _ . . . . - . _ , - - - , - , . _ . _ , . . . _ , . _ _ , _ - - _ - _ , - . - _ . , _ _ . ~ . , - , - . , _ _ -
_ _ _ - _ _ _ - - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ . - _ _ _ _ _ _ _ - . .-. _
.
.
12
(7) Labeling Program -
Completed approximately 15,000 to 16,000
mostly in the mechanical area, electrical area continues to be
a problem.
(8) Maintenance - Maintenance Work Requests (MWR) are at about 5,000
for all units, about 2,000 of these are Unit 2 startup items.
The total number appears to be holding steady (as old ones are
worked, new ones are written). Maintenance appears to be
fragmer.tary. ,The licensee is reorganizing the maintenance area
currently which should be a benefit.
(9) Engineering Support -
Support from Department of Nuclear
Engineering (DNE) is still slow for design concerning altera-
tions and testing activities. System Engineers should help;
however, due to their recent addition they are not fully
effective yet.
b. Piece Parts
The inspector attended a piece parts briefing at the Trinity School
in Athens, Alabama. TVA has leased the schoolhouse and is currently
using it to house a portion of the Piece Parts program. The group
consisted mainly of Wyle Laboratories (WL) and Bechtel North America
(BNA) personnel under TVA supervision. The group was working to
procedure SDSP 16.1 "Dedication Program For Commercial Grave
Material, Parts or Components for EQ Related Applications" and were
tasked with the following:
(1) Review past, prier to January 14, 1988, maintenance work
requests (MRs) for material used in performing maintenance on
selected equipments.
(2) Review past stores issue forms for materials issued for use in
selected equipments.
(3) Review replacement / repair parts in the warehouses ordered for
selected equipments.
The above reviews were made in crder to determine the status of
material installed in the plant and material in the various warehouse
locations The inspector was informed that approximately 260 items
of material were identified as indeterminate to date. Where
discrepancies are found, CAQRs or other traceable documents are
generated in order that technical resolutions are completed. The
combined personnel from both WL and BNA, known as the Item Evaluation
Group (IEG) numbered approximately 65, with specific tasks assigned
to each subgroup. During the briefing the inspector was informed
that all future procurements will be add essed by the Contract
. ._ _ _ _ . _ . - _ - _ _ . . ._
.
.
13
Engineering Group (CEG). Currently as material arrives and if any
discrepancies are discovered through the receipt inspection program
the discrepancies will be forwarded to the CEG for resolution. The
interface between the IEG and the CEG is one of consultation. The
IEG is generating packages with each package tied to a specific item
of equipment. It is estimated that approximately 5,000 packages will
be generated by the IEG and 500 have been completed to date. A
10 CFR 50.49 inspection will be performed by the NRC during the a
later visit. The inspector asked about Quality Assurance (QA)
coverage of the program and was informed that QA was being provided
by BNA.
c. Jo!;; Test Group (JTG) Meeting ,
On January 19, 1988, the inspector attended a JTG meeting which
involved change notices to three Restart Test Program procedures,
two system test specification and test procedure. All ite.ns were
recommended to the Plant Manager for approval. Ten open/ closed
items were discussed involving various tests, department interfaces
and research of past maintenance activities to determine recurring
problems. The inspector had no concerns on the conduct of the
meeting or techr.ical issues addressed.
! 10. Surveillance Instruction Review and Upgrade Program
i
During the week of January 25, 198R, a six man team inspection was
'
perfermed on the licensee's SI review and upgrade program. The program is
defined in Section 2.4, 5.1, and 5.2 of the Browns Ferry Nuclear Perfor-
'
mance Plann (NPP). The two major objectives of the inspection were: 1) to
. verify that uegraded sis accurately reflect and accomplish the require-
ments established by Technical Specifications and 2) to verify workability
of the sis by direct observation of SI performance by licensee personnel. ;
j Additional aspects to be covered during the inspection included a sampling i
of the documentation packages assembled by the licensee in order to verify
'
programmatic compliance with the NPP commitments, a review of completed SI
data for completeness and adequacy of the licensee's technical review and
closecut of previous open items in this area. Previous inspection
findings in this area are documented in Inspection Reports 259/260/296,
86-25, 86-36, 87-14, and 87-30.
The results of the inspection indicate that in all cases reviewed, the sis
4
, adequately implement the technical specification requirements. Results on
the second objective (workability) indicate that additional attention is
needed in order to force a timely completion of this aspect. In one SI
observed (2-SI-4.2.J-28, Biaxial Seismic Switch Calibration and Function ,
Test) the test had to be aborted since it could not be performed as
This procedure had not been through the final step of the
I written.
upgrade process as outlined in Site Director Standard Practice 2.14, [
,
I
i
- . - - - - , - ..---__..- ,- _., - ._. - - ., .- - -- _ , _ - .-
__ _ _. _ _ . - ._
i
14
i
SI Evaluation Program. This final step is a validation performance where
a qualified procedure performer, the author, and the cognizant engineer
actually perform and witness the procedure. The program as outlined in
SDSP 2.14 appears to be breaking down at this point. Although the
cognizant engineer is responsible for scheduling the validation perfor-
mance as soon as the procedure has been upgraded and approved for use, ,
only a small percentage of the approved sis have been validated. SDSP ,
2.14 further states that the intention is to complete the validation ,
performance on the first time the procedure is used after approval. Many
of- the sis reviewed had not been validated even though they had been ;
performed on at least one occasion since approval. Further, at least one !
cognizant engineer was not totally familiar with his role in scheduling '
the validation run. The l'censee stated that of a total of 197 approved
upgraded sis, only 40 have been validated and all of those 40 were hydro-
static test procedures. Twenty-five sis are in some form of partial
,
validation pending procedure changes and re-witnessing of certain
- portions. That leaves 132 procedures not validated with no tracking or
scheduling mechanism to attempt to comply with the SDSP 2.14 provision ,
of performing a validation run on the first performance. This area of t
'
concern was stressed as the major finding of the inspection during the
<
exit with the Plant Manager on January 29, 1988. f
The remaining findings or concerns are grouped into four categories.
'
First are concerns of a technical nature then programmatic or policy
issues followed by procedure enhancement and then several items detected
during the inspection that are not SI related. .
a. Technical Concerns ,
(1) A discrepancy exists between SI 4.9.A.2.C, Shutdown Board )
'
Battery Discharge Test, and IEEE 450-1980, Recommended Practice
ter Large Lead Storage Batteries for Generating Stations. The I
- SI requires a cell to be jumpered out to prevent cell reversal
prior to it's voltage reaching 0.5 volts whereas the IEEE
Standard requires this action at 1.0 volt.
"
, (2) The acceptance criteria for maximum battery voltage contained in
SI 4.2.J-28, Biaxial Seismic Switch Calibration and Functional
Test, was 0.5 volts higher than that listed in the vendor '
,! manual. This comment also applies to SI 4.2.J-18, Triaxial Time t
j History Accelerographs.
(3) SI 4. 5. A.1.B(I), Core Spray Pumps Monthly Operability states
that this surveillance must be performed immediately in the j
event that both loops of Low Pressure Coolant Injection are
found to be inoperable. This action is not required by -
TS and may be competing for the operators attention. The TS
required action statement is an immediate plant shutdown in this
condition. Licensee representatives agreed to review the
.
advisability and basis for this statement.
l
!
i
i
- ._ _ _ -, _ _ , _ _ , _ - , _ _ _ _ _ _ . _ , _ . ~ . _ . . _ ___
15
(4) The calibration of biaxial seismic switches performed through
SI 4.2.J-2A is an electronic calibration which uses a field
calibrator unit to impress a signal substitution on the
circuitry. The TS definition of a calibration requires that the
device output must correspond within an acceptable accuracy to a
known value of input. In the case of a seismic instrument, the
input is acceleration (or displacement and velocity which may be
translated to acceleration). The licensee depends upon the
instrument manufacturer to provide the calculated conversion of
input acceleration to the test signal. The manufacturer is not
on the QA approval suppliers list nor is there any QA coverage
on the calculation. Since this calculation is critical to the
calibration of the instrument, the licensee was asked to review
this concern for compliance with the QA manual requirement on
coverage and traceability of calibration of TS equipment.
(5) The inspector noted that the SI 4.2.C.4-1, SRM functional test
did not require that each circuit card to be pulled to func-
tionally test the inoperative trip channel of each instrument.
The inspector noted that the condition also applies to the IRM
and APRM channels and had already been identified by the
licensee and CAQR BFQ870978 was written for dispostion of the
condition. A TS Interpretation has also been developed to allow
this particular function to be tested by the injection of a test
signal simulating a pulled card. The inspector intends to
further review the dispor.ition of this CAQR.
b. Programmatic or Policy Issues
(1) It was noted during the review of SI 4.2.B-21 A, Core Spray Pump
Discharge Pressure Switch Functional Test that there were no
informational statement which would make the technician
performing the test aware of the trip setpoint. Although the
functional test is only intended to verify that the trip
actuates, it would be unacceptable to consider a switch func-
tional if its setpoint had drif ted in a gross manner. The
technician should be made aware of the setpoint so that
corrective action could be initiated if, for instance, a switch
actuated at 50 psig when its setpoint is 200 psig. This concern
was found to be generic to all functional testing of instruments
with trips.
(2) Steps 7.6.29 - 7.6.33 of SI 4.2.C-4.1, SRM Functional Test
provide for functionally testing the detector retract permit
interlock. The interlock is demonstrated to function between
50 cps and 300 cps as read on the SRM drawer LCR meter.
Although note 3 to table 3.2.C of the TSs specifies that the
-- -_ - - ..
.
!
t
16
permissive interlock is bypassed when count rate is greater than
or equal to 100 cps, no actual setpoint is specified in this SI. .
The licensee stated to the inspector that there was no intent to
verify the sepoint which was checked under another calibration
instruction only to functionally . test the equipment. The .
inspector feels that the functional test instruction could be
improved by including the actual setpoint.
(3) SI 4.2.J-2A contained a step to "perDrm the _necessary steps of
SI 4.2.J-2B" . This general statement has the potential for
circumventing the requirements of having pre-approved step-by-
step procedures. It is left to the judgment of the performer as
to what steps are necessary and may not be consistently applied
by different technicians. A licensee representative stated that '
l
the . management policy is to not allow such statements in
,
surveillance instructions,
i
(4) Some inconsistencies were noted in the manner that initial
as-found data outside the acceptance criteria is documented on
the SI cover sheet. Only one check mark is provided in the !
blank for recording acceptance criteria satisfied (i.e., yes '
'
no _ _) it is unclear whether this was intended to be completed '
for the as-found data or the as-left data following any
necessary adjustment. On at least one occasion when the ,
as-found data was out-of-tolerance but the as-left data was '
acceptable, the mark was made that the as-left was satisfied.
This led to the SRO not being informed of & deficient condition
, and he therefore failed to perform a review to identify any ,
relevant limiting conditions of operation (LCOs). A clarifica-
[
tion of the cover sheet or addition of a check for the as-found ,
as well as the as-left data may be in order. l
(5) The plant management initiated a policy on how operators were to
l handle the interface between the plant labeling program and '
- procedure upgrade. A short time period is allowed to exist with
'
discrepancies between component and annunciator labeling and
'
procedures. The interval was to be minimized by weekly
revisions to the Annunciator Response procedures (ARPs). Many
j annunciator labels were found to disagree with the sis reviewed.
l c. Enhancements
l (1) Several unit prefix identifiers were missing on some components f
l identified in the sis. '
! ;
l (2) Some illustratien improvements were recommended (add a North- l
,
South orientation on the seismic switch drawings and denote L
i specific fasteners on the control rod drive housing support
drawings where tolerances are taken).
1
1
F
!
. - - - ,-- . - - - - ,_-
_ .. . - - - . -
.
. 17
_
"
(3) A note was recommended to be added to takeup the cassette tape
leader on the triaxial time history surveillance since some
-
initial seismic data could be lost on the non-magnetic leader.
The instrument mechanic performing this surveillance was msde
aware of this during a recent training class. A note to repiace
the dessicant would also be in order on this procedure. The
inspector witnessing this procedure noted that although the
mechanics were prepared to do this, it was not done until
prompted due to the lack of a specific procedural step.
~
(4) On some instrument surveillances, (i.e. , SI 4.2.B-21A) instru-
ment isolation and drain valves are two party verified during
the procedure and then a final independent verification of valve
position is made at the conclusion of the procedure. In the
absence of a valve or instrument list, it is unclear how the
third, independent party knows which valves were operated during
! the procedure. Only one signoff step is provided to document
'
that all the valves manipulated during the procedure were
returned to normal, instead of a two party verified signoff.
! (5) Several steps were suggested to be verified by a second party
on various procedures.
j
d. Miscellaneous Issues
.
(1) A licensee reportable event determination (LRED) performed
'
following failure of the biaxial seismic switches on January 5,
1988, contained an erroneous statement that the instruments were
totally passive with no inputs or trip outputs or annunciation.
Lack of knowledge on this instrumentation may be due to the lack
of documentation in the FSAR. There is surprisingly little
information available on these instruments although they are
required by TS and by 10 CFR 100.
(2) It was noted that although the Radiological Emergency F'.an was
revised in the fall of 1987 to delete Implementing Procedures
(IPs) in favor of Emergency Plan Implementing Procedures
(EPIPs), the Annunciator Response Procedures for seismic
instruments still referenced IP-24. This IP no longer exists.
The correct reference should be Emergency Plans Manual (EPM-9).
The following surveillance procedures were reviewed during the course
of the inspection:
4.1. A-1 RPS Mode Switch in Shutdown Functional Test
4.1.A-2 RPS Manual Scram Functional Test
4.2.B-21 Inst. Controlling CSCS, CS Pump discharge pressure
4.2.B-21A Inst. Controlling CSCS, CS Pump discharge pressure
Functional Test
4.2.8-31 Inst. Controlling CSCS, RCIC Steam line high flow
18
4.2.C-4.1 SRM Functional
4.2.C-6 SRM/IRM Detector Calibration
4.2.0-3B RCW Effluent Radiation Monitor Inoperable
4.2.J-1A Triax time history accelerographs. Channel check
4.2.J-1B Triax time history accelerographs Functional Test
4.2.J-2A Biax seismic switches channel check and Functional Test
4.2.J-28 Biax seismic switches Calibration and Functional Test
4.2.J-3 Triax Peak Acceleographs Functional Test
4.2.K-1 Stack Rod Monitor Calibration
4.3.A-2 Control Rod exercise
4.3.V.1A Control Rod Coupling Integrity
4.3.B.1B Control Rod Coupling Integrity After Refuel or Maintenance
4.3.B.2 CRD Housing Supports
4.3.B.3.A RWM/RSCS Functional Test for Startup
4.3.B.3.B RWM/RSCS for Shutdown
4.3.B.3.B.3 RhH Program Verification
4.7.A.5.C Calibration of Gas Chromatograph used for 02 Measurement
4.7.B-6 SBGTS Iodine removal efficiency.
4.8.B.1. A.1 Airborne release rate by continuous air monitor
4.8.B.1. A.2 Airborne release rate by Manual sampling
4.9. A.1.B.4 D/G emergency load acceptance test
4.9.A.2.A-1 Weekly check of 250V Main Batteries
4.9.A.2.B-1 Quarterly check of 250V Main Batteries
4.9. A.2.C (II) Shutdown Board Battery Discharge Test
4.5. A.I.B (I) CS Pump Operability
4.5.A.1.C (II) CS Mov Operability
11. Inspection of Engineering Changes and Modifications (37700)
The inspection was conducted to ensure that Engineering Change Notices
(ECNs) and resulting plant modifications required to support Browns Ferry
Unit 2 restart are being adequately accomplished in accordance with the
licensee's quality and engineering assurance p rgrams. This assessment is
based upon inspection of ECN and plant modification packages that have
progressed completely through the design process and are being (or have
been) installed.
The inspection activities reported herein involved represent the second
phase of a two part inspection effort begun in October 1987, and
reported in NRC Inspection Report Nos. 50-259/87-42, 50-260/87-42, and
50-296/87-42.
The l '. c e n s e e has previously experienced breakdowns in the facility's
configuration management as documented by prior NRC inspections,
Systematic Assessments of Licensee Performance, and the licensee's Nuclear
Performance Plan. The NRC Technical Liaison personnel provided the
inspectors with a general inspection plan addressing the following major
areas to be assessed. The inspection plan was applied to a sample of ECNs
and modification packages for the High Pressure Coolant Injection (HPCI)
System, Reactor Core Isolation Cooling (RCIC) System, the Core Spray (CS)
System, and Reactor Suilding Closed Cooling Water (RBCCW) System.
_ - .. -
. . _ _ . . . _ _
19
.
.
'
This inspection continued with field inspections of ECNs and the asso-
ciated work plans for the systems listed above. Findings and concerns
resulting from this inspection arc provided in herein, and summarized
in subparagraph e. Further NRC followup inspections are addressed in
subparagraph f. Specific reference material reviewed during the inspec-
tion is listed in subparagraph g. Persons contacted by the inspectors and
who provided material input to the inspection are listed in subparagraph
,
h.
The following ECN attributes were specifically reviewed and inspected '
during this two phase inspection. Additional background information and
findings were provided in the Phase 1 report.
(1) ECN/ Modification Processes
(2) Review of ECN/ Modification Packages
(3) Inspection of Field Activities
I !
'
(4) Quality Organization Involvement
(5) Organizational Interfaces
(6) Progrkm Status and Overview
'
4 Engineering Change Notices are the vehicle established to control the :
development, approval and transmittal of design ir f orn,ation for plant ,
modifications. Work Plans and Inspection Records (WP& irs) are the work
instructions for installation of the modifications. For the purposes of i
'
this report, the ECNs and WP& irs are collectively referred to as
"modification packages" or "work plans" respectively. ,
- The findings presented below reeresent either new findings developed during the ,
! second inspection phase or the results of additional inspector and licensee
i followup on initial phase findings (as indicated). Previous inspection !
findings in this area are provided in Inspection Report Number 07-42. The
'
! following modifications were initially inspected during the prior site visits;
l inspection was completed during this inspection period. ;
l [ NOTE: Unit 2 only; all valve & equipment numbers prefixed by 2-] ;
i
'
! ECN Subject
P0157 Removal of Core Spray (CS) Pump Motor Oil Coolers .
--
WP 6619 & 2018-84 - Remove Coolers !
!
! P0795 Installation of Valves to Permit 10 CFR 50, App. J Testing i
! of CS Valves75-606, -607, -609, and -610 ,
j --
WP 2117-86 - Piping & Valves -
i
--
WP 2118-86 - Piping & Valves
l
--
WP 2119-86 - Piping Supports !
I --
WP 2120-86 - Hydrostatic Test ;
1
1
_ _ _ _ _ _ - _ _ - _ _
20
P02039 Replacement of CS Solenoid Valves FSV 75-57 and -58 with
environmentally qualified units
--
WP 2174-84 - Electrical
P0959 Installation of Valves to Permit 10 CFR 50, App. J.
Testing of RBCCW Valves
--
WP 2121-87 - Piping & Valves
--
WP 2122-87 - Drywell Area Supports
--
WP 2123-87 - Torus Area Supports
--
WP 2124-87 - Reactor Building Area Supports
--
WP 2165-87 - Hydrostatic Test
L2003 Replacement of Stainless Steel Core Spray Piping and Safe
Ends with Carbon Steel
--
WP 9244 - Same as Above
P0162 Remove Auto Initiation Opening logic from Normally Open
HPCI Steam Line Isolation Valves
--
WP 2005-86 - Test FCV 73-3
--
WP 2100-85 - Test FCV 73-2
--
WP 2144-85 - Electrical Work
P0652 Replace RCIC Valve FCV 71-40 with Pneumatic Operated Soft
Seat Check Valve
- -- WP 2054-84 - Replace Valve
l -- WP 2148-85 - Electrical Work
P06S1 Replace HPCI Valve FCV 73-45 with Pneumatic Operated Soft
.
Seat Check Valve
l - 2151-84 - Replace Valve
i -- 214. 35 - Electrical Work
P0153 Reroute Cabling to Separate HPCI and Automatic
Depressurization (ADS) Components (App. R)
j -- WP 2166-85 - Cable Relabeling
-- WP 2005-85 - Install New Conduit
-- WP 2084-85 - Pull / Terminate New Cables
l -- WP 2085-85 - Rework Cable / Internal Wiring
-- WP 2100-85 - Perform Modification Testing
-- WP 2162-85 - Install Cable Tags
P0965 Installation of Valves to Permit 10 CFR 50, App. J.
Testing of HPCI Valves
-- WP 2126-87 - Install Valves & Piping
-- WP 2127-87 - Install Supports
P03061 Replace Level Switches (IEB 79-018)
-- WP 2193-84 - Replace Level Switches
-- WP 2034-86 - Construct Access Platform
-- WP 2066-86 - Perform Functional Test
!
_ _ _ _ _ _ _ _ - _ _ _ _ _ _
..
21
P03116 Replace Limitorque Operator Components (IEB 79-018)
-- WP 2226-84 - Replace Components FCV 73-2
-- WP 2152-87 - Modify Pipe Support
- -- WP 2166-87 - Rework Hangers
-- WP 2224-87 - Replace Gears (FCR 73-34, -44) ;
a. ECN/ Work Plan Inspection Findings
] (1) During review of ECN P0162, the inspector noted that internal
plant correspondence, dated July 29, 1986 (R41 860725 898) l
requested design authorization not to install keylock type !
switches for valves 2FCV 73-2 and -3, but to use ". . . existing
Class 1E SBM switches ... still available for ready *
installation." The inspector noted that associated ECN P0955,
Install HPCI Steam Line Isolation Valve Switches" was written as '
a complement to ECN P0162 to delete the keylock switch
requirement of ECN P0162. *
,
L
On reviewing the WP&IR 2171-87 associated with ECN P0955, the '
inspector noted that the work plan was written to install new
switches, not yet received, beccuse material traceability on the
"existing" switches (i.e. , Class 1E SBM switches) had been lost.
The existing switches had apparently been improperly stored and
tracked after removal for modification work. Replacement
switches had been ordered but were on a very long lead time back
.
order. Additional licensee action appears r.ecessary to ensure
I that material traceability is not lost on talvageable material
and to improve material procurement for replaten,ent switches.
'
- (2) Inspector review of WP 2054-84 (ECN P0652) determined that a
weld defect requiring repair was detected by nondestructive
examination (NDE) via NED Procecure Report No. R0878 on
April 22, 1986, at 8:03 a.m., for weld TRCIC-2-7X1 at valve FCV
71-40. Notwithstanding the defect, a hydrostatic test was
conducted at about 11:00 a.m., hours on the same day because the
cognizant engineer conducting the test was not aware of the
defect. During the test, with the system at test pressure, the
cognizant engineer was advised of the defect and immediately
stopped the test. Subsequantly, the defect was repaired and
another hydrostatic test successfully conducted on May 14, 1986.
Additional licensee action appears necessary to assure that
outstanding inspection and work items are properly dispositioned
prior to pressure testing of systems.
(3) While performing field inspection of installations per WPs
2054-84 and 2153-84, the inspector noted that the two test con-
nection lines connected to valves FCV 71-40 and FCV 73-45 were
equipped with open ended tailpieces, i.e. , without cleanliness
_ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _
n
'
22
,
caps of any kind. A check of plant mechanical drawings for the
subject installation showed no requireme':ts for installation of
such caps on these and most similar installations. Open ended ;
tailpieces provide a pathway for introduction of foreign <
material to systems subject to cleanliness controls and, in some
circumstances, can result in the induction of stagnant water
from trenches, sumps, etc. if system pressure drops below. ;
atmospheric. Additional licensee evaluation appears warranted .
to determine the acceptability of such installations.
(4) Valve FCV 71-40 (Atwood and Morrill Co., Inc., Orawing 15184-02)
has a cap (Piece #5) with a large notch in its edge which is not [
pictured on the vendor drawing. The carbon steel interior of -
the valve top cavity (under Piece #5) was found flooded with
i water and significantly corroded. The notch is large enough to !
, permit the entry of water and other foreign material. .
! Engineering evaluation of the conditions appears warranted to :
determine the acceptability of the apparent deviation from the
vendor drawing and the as found conditions of the valve.
- (5) ECN P0651, WP 2153-84, replaced HPCI check valve FCV 73-45 with
j a pneumatic operated, soft seat check valve and reinstalled two .!
J test connection lines with two 3/4 inch isolation valves in each
line. The test lines exit the valve's body via sockolet weld !
joints ens then run horizontally without support for either j
line. This configuration results in about 20 inches of l
i unsupported moment arm (piping and 3/4 inch valves) cantilevered !
,
f rom the sockolet welds. The design documents for this instal- :
lation did not provide for supporting these lines and the :
inadequate installation was not recognized by design and ;
'
) installation personnel.
i
The inspector observed that, a) piping support is provided for l
a similar test line configuration of valve FCV 71-40, and b) a
~
3
general small bore pipe study for adequacy of hangers (ECN i
- p0625, Branch Line Vibration, 1984) identified that FCV 73-45 !
- ' . configuration as inadequate. Additional licensee evaluation ;
appears warranted to determine the reasons for the omission of !
l piping supports for ECN P0651 and for the failure of the design
installation personnel to recognize the omission. ;
-
!
l (6) The conduit and cable installation of ECN P0753 was inspected in !
. the Auxiliary Instrument Room #2, 593' elevation; the Cable !
Spreading Room, 606' elevation; and the control room, Panel 9-3, l
1 617' elevation. Numerous deficiencies were noted as follcws: [
i
- (a) Conduit #2ES200-IS2 was found loose (penetration at 593' !
. level in concrete pad, drawing 45B2895-165). [
- !
i
! [
L i
, i
' '
<
i
-. . -
i
23
.
(b) Junction of conduits 2ES200-IS2, 2ES-204-IS2, and
2ES-211-IS2 was found to have all fittings loose.
. (c) One foot west of the junction noted in (b) above, conduit
2ES-204-IS2 was hard installed against the support
structure of the adjacent overhead light. The inspector
noted that the inspection criteria (BF MAI-27, Attachment
-
1, Conduit and Junction Box Inspection Requirements)
concerning proximity of conduit to other plant structures l
states, "No minimt.:m physical separation is required between
a conduit and any other conduit, tray, or device unless r
otherwise specified on design documents." This criteria
'
appears tv be inappropriate for prevention of macceptable .
'
stresses and chafing.
(d) Flexible conduit 2ES-211-IS2 (at overhead) was missing the
1 cover on a condulet (LB) fitting and the fitting was loose.
j (e) Numerous other conduit fittings in the area (not worked !
under ECN PU753) were loose, e.g. 2GS2059 at the East t
Control Rod Drive Scram Discharge Volume Tank Not Drained- l
Junction box. ;
4 ;
l (f) ine condulet body on conduit 2ES-211-IS2 at the 606' l
elevation is missing the cover, and is loose at the conduit :'
reducer and coupling (drawing 4582895-166).
1
.
(g) The LB fitting at the junction of conduits 2ES-201-IS2, !
'
2ES-211-IS2, and 2ES-1436-IS2 was mit, sing its cover.
t
'
(h) The condulet at the top of flexible conduit 2ES-201-IS2 f
,
entering Panel 9-3 was missing its cover, f
(1) Conduit fittings in Panel 9-3 were found loose; debris from
the installation was noted in the bottom of the panel. ;
The inspector noted that the Conduit Installation Data Sheets of BF I
MAI-27, Attachment 2, were signed off complete for the above listed
conduits by the craft and QC except for Data Sheet Item 17, Cleanup, !
j for conduit 2ES-200-IS2. The inspector was advised by the licensee ;
i that this conduit had additional work in progress and therefore ;
j Iten 17 had not been signed off. j
!
3
The inspector also noted that the Cable Pulling Data Sheets, BF [
4 MAI-44, Attachment 2, were signed of f complete by craft and QC for 3
the cables in at least one of the conduits (P:.nel 9-3, conduit l
It was also noted that subsequent, but not yet i
2ES-201-IS2). i
completed work plan steps (WP 2084-85, ;teps 17 & 18) referring to
I
'
t
) I
! t
t
i
1 k
24
close cuts and walkdowns may have identified or corrected the above
incomplete work items. The deficiencies notes in this paragraph are
lists J as a violation, example B of the 10 CFR 50, Appendix B,
Criterion V violation (260/88-02-04).
(7) Work Plan 2152-87, issued to accomplish part of ECN P3116,
performed a verification and correction of fielo conditions for
various hangers and pipe supports for the HPCI System. At the
time of inspection, WP 2152-87 was field complete and reviewed
by the cognizant engineer. The following discrepancies were
noted by inspector review of the work plan:
(a) Five work plan steps stated, "Verification of support
configuration will be done by ONE". These steps had not
been completed at the time the work plan was signed off
complete by the modifications engineer. The engineer
advised that the steps were intended as information only
notes and were not intended to control the DNE activities.
As they appeared, the steps were part of the procedure
action steps and sign off of the work plan without their
completion appears inappropriate without revision of the
work plan to accommodate it.
(b) Step II.E.1 did not list drawing 467B455-125, Revision 2 as
the reference drawing for accomplishing verification of
support identification and stencilling. This appears to be
contrary to SF 8.3, Appendix E, Item 3, Plant
Modifications, which requires that the work plan steps
include references needed to accomplish the work.
(c) Support H-94 required adjustment of the spring can load to
bring the load in accordance with plan re qui rer.e n t s .
However, the Hanger and Restraint Inspection Data Sheet
(MAI-23, Attachment A), Step 5.1.3 requires "Verification
that all threaded connections are installed snug" was >
checked "No" in the "Inspection Required Check list",
indicating that no final inspection of threaded connection
integrity was required.
Since the threaded connections on the hanger were broken to
make the adjustments per the work plan, the inspection 3
requirements of MAI-23, Attachment A, Section 5.1 should
apply. The inspector further noted that step 5.1.4 for
verification that all structural shapes are the correct
size and installed correctly was also checked "No" even
though the installation shape was modified (see additional
field inspection finding below).
4
-- -.
__ _ _ _ _ _ - - _ _ _ - _ _ - _ - _ _ _ _ - _ __ _. _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ __ _______ _ _ _ . _ _ _
r
25
During discussions with the cognizant engineer concerning
k? 2152-57, the inspector was advised that the final hanger
inspection of MAI-23, Section 5.1 was not applicable at the
time of work plan closure for isolated hangers such as
those under this WP and the-"some other" mechanism would
accomplish the'"final inspection". The engineer was unable
to identify what mechanism would apply. The inspector
concluded that MAI-23 was the applicable criteria af ter.
further: discussion with licensee management. Failure to
specify and implement the proper inspections appears to be
contrary to the requirements of MAI-23.
(d) Hanger H-94 was laterally repositioned on the supported
.HPCI pipe and was adjusted to design values for spring can
load. There was no supporting documentation to document
this . hangar movement. The inspector noted that the
repositioned and adjusted installation resulted in a
misalignment of about 10 degrees of the strut, turnbuckle
and pipe clamp assembly with respect to the spring can
assembly
The work plan referenced MAI-23 for the installation.
Section 2.6 of MAI-23 requires that "Support points may be
relocated laterally to line up with the pipe if the support
can be fabricated within the allowable support fabrication
tolerances of Section 2.7". Section 2.7.3 requires that a
'
2 degree tolerance be applied to specified angles unless
tolerances provided by current applicable codes are less
'
restrictive. The inspector concluded that the modified
installation was not made in accordance with a work plan or
referenced requirements. Items 7.c and 7.d are included
in the Notice of Violation of 10 CFR 50, Appendix B,
Criterion V as example A (260/85-02-04).
(8) WP 2166-87 (ECN P3116) performed installation verifications and
corrected installations for two hangers in the Residual Heat
'
Removal (RHR) System. The work plan was field complete and
reviewed by the cognizant engineer at the time of the
'
inspection.
(a) Step II.A stated, "DNE to 'as construct' portion of
47B2452-118, Revision 0, that does not require physical
work per PI 87-49." This step had not been completed at
the time that the work plan was signed off. This appears
to be an additional example of improper work plan closure
as discussed for WP 2152-87 above.
(b) Hanger H-93 in the RHR System (drawing 4782452-119) was
removed by WP 2166-81. Although the work was field
complete, the area of the large bore (about 24") pipe under
the removed pipe clamp was unpreserved (unpainted), carbon
steel. Adjacent pipe surfaces were painted. The inspector
- _ - - - - - - - - - -
_ ,
26 !
<
P
noted that the work plan did not include a work step to
-preserve the exposed piping and the cognizant modifications :
personnel had apparently overlooked the need to evaluate >
and address the condition. l
<
.
(c) The hanger H-93 material and fittings removed, including
pipe clamps, strut, etc. was found lying on the grating >
under the point of original installation notwithstanding
.
work plan steps requiring maintenance or proper house-
'
keeping. This further represents inadequate control of
'
abandoned material which has the potential for its
unauthorized reuse.
(9) ECN L2003 involved replacement of stainless steel Core Spray s
piping and reactor vessel penetration safe ends with carbon
4
steel to reduce the potential for intergranular stress corrosion
cracking (IGSCC) in accordance with WP 9244. WP 9244 was
released for work on March 31, 1978 and installation appears to
have been substantially completed in June,1978. However, the
4 work package was not properly closed. ,
1
The package had been assigned to the Special Projects Group
'
i (Backlog Group) for review and closure. The inspector conducted ,
a preliminary review of the current status of WP 9244 on l
November 4,1987 finding that the work plan had not yet been
reviewed nor processed for disposition, that the work plan
records were not sufficiently organized to permit review, but
4
that portions of the workplan had not been signed off as
complete. The partially executed work plan was divided into .
'
Parts I and II, addressing installation of piping from FCV 75-26 '
to -54 and and from FCV 26 and -54 to FCV 75-23 & -51
'
i
- respectively. The work plan had not been signed off for key
installation and testing steps. The hydrostatic test for Part I
'
had not been signed off by QA. Dye Penetrant Testing for Part I
! welds for FCVs 75-23, -25, -26, -51, -53, -54 had not been
signed off by QC. The hydrostatic test for Part II had not been
l signed off by either the cognizant engineer nor QA. The data
l package for Part II, including major installation, NDE and ,
4
testing steps had not been signed off by the cognizant engineer l
nor QA. l
i The work plan as found status did not appear adequate to have
! supported a licensee determination of Core Spray System
TS operability for periods of operations since mid-1978.
- Although the work plan file may contain sufficient information
- to substantiate completion of the items above, the condition of .
<
the records did not permit a ready determination. The licensee ,
,
was unable to provide any information which would ' Jicate that
-
,
4
1
'
.
I
_ _ _ _ _ _ _ _ _ _ _ .
27
a system operability review had been conducted for the work plan
and its status at any time since 1978 except for an incomplete
attempt during 1984 to identify missing closecut elements.
During this insrection, the licensee had completed a review of
the large bore piping installation records finding minor dis-
crepancies in the weld data sheets and NDE records. The
inspector reviewed the informal evaluation prepared by the
Modifications Welding Engineering Supervisor finding that the
licensee's evaluation and proposed discrepancy dispositions
appeared reasonable. The discrepancies involved missing or
incomplete records for code required NDE; in each case the
licensee had either found records of equivalent examinations or
planned to formally disposition the discrepancies per the piping
design requirements. The licensee was in the process of formal
disposition at the close of this inspection.
During this review, the inspector noted that the portions of the
work plan with the incomplete verification signatures (above)
were missing from the original package. Upon identification by
the inspector, the licensee attempted to locate the missing
pages but was unsuccessful. The missing pages appear to be the
last two pages of the Part I and Part II work plan travellers.
The licensee believes that equivalent information available
elsewhere in the records will permit equivalent disposition of
the work plan.
During December 1987, the licensee also performed as-built
walkdown verifications of the small bore piping installed by the
work plan. The field sketches resulting from the initial
walkdowns were compared to the work plan and welding records
with significant disagreement identified. Additional walkdowns
were performed during January 5-8, 1988. The documented results
of the walkdown to work package comparison were not available to
the inspector at the end of the inspection but the licensee
orally advised that the discrepancies above were confirmed
including the apparent use of improper material, disagreement
between weld /NDE data sheets and installed piping joints, etc.
The licensee was planning their course of action at the close of
this inspection.
(10) The Nuclear Performance Plan (NPP), Volume III,Section II and
Appendix A, include licensee commitments to evaluate, disposi-
tion and close partially implemented ECNs and work plans such
as ECN L2003 eboe. NPP, Volume III, Appendix 0, Revision 1,
includes a listing of specific ECNs which will be completed
(closed) prior to restart from the Unit 2 Cycle 5 outage.
However, ECN L2003 and other "backlog" ECNs and workplans are
. _ . . _ . . -
_ _ _ _ _ _ _ - _ - _ . - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
28
r,t listed for closecut in Appendix 0. Discussions with the
licensee indicate that prior practices for maintaining the
outage schedule data base did not include these activities as
"active" ECNs. Therefore they were not previously included. As
discussed further below, the licensee has recently performed a
major revision of the management information system data base
and these, and other newly issued or planned ECNs are now
included in the list from which Appendix D is derived.
In order to accurately reflect those ECNs which must be closed
prior to restart, the licensee should consider revising NPP,
Volume III, Appendix 0.
(11) The prior inspections of WP 2117-86 and 2118-86 (Installation of
CS Test Valves) found an interference between the new piping and
insulation on a plant heating system pipe, overspray of fire
retardant material on CS valves, and unpainted piping weld
joints. The two former discrepancies were properly corrected by
the licensee.
The latter item involving the unpainted piping resulted from the
craft painting foreman signing off the work plan repainting
steps although all piping joints had not been repainted. The
licensee had then made a nonintent change to the subject work
plan authorizing the repainting. During this inspection, the
Quality Engineering (QE) Manager advised that previous problems
had been experienced with nonintent changes being used for
rework / system reentry and that a Condition Adverse to Quality
Report (CAQR) should have been issued instead. The inspector
advised the QE Manager that the matter should be reviewed by the
TVA staff and appropriate action taken in accordance with their
procedures.
(12) ECH P3069, Replacement of CS Solenoid Valves PSV-75-57 and -58
in per IE Bulletin 79-01B (Environmental Qualifications),
involved replacement of solenoid valves, wiring splices, and
miscellaneous conduit to upgrade the environmental
qualifications of tiie installation.
During review of WP 2124-84, the inspector found that the
Nuclear Power Storeroom Requisitions No. 1385DC00623 (solenoid
valves), 5688-16363 and 5688-16349 (splice material) did not
have the appropriate blocks checked designating the environ-
mental qualification requirements nor application of IEEE Class
1E requirements. The cognizant engineer and storeroom personnel
were able to demonstrate that properly certified material was
actually issued. However, this example is similar to that
addressed in Inspection Report 87-42, involving designation of
__ _ _ _ _ _ _ _ _ _ _ _ _
29
quality level for self drilling anchor bolts. Further, the
cognizant engineer was unable to identify the instructions
applicable to processing the forms and was apparently unaware
of BF 16.4 (below).
The inspector reviewed BF 16.4, Material, Components, and Spare
Parts Receipt, Handling, Storage, Issuing, Return to Storeroom
and Transfer, Revision 7, finding that only cursory instruc-
tions exist for completion of the forms. BF 16.4, Section 6.2,
provides for the Item Evaluation Group to review each CSSC
requisition to ensure that all issued items are properly
evaluated for qualification requirements providing some
additional assurance that an improperly completed requisition
would not resuit in improper material issuance and installation.
Based on snis and the previously identified example, additional
licensee action appears warranted to ensure that personnel
are properly trained and supervised in the completion of
requisitions for safety grade material.
b. Procedure and FSAR Updates
(1) The NRC requires periodic updating of Final Safety Analysis
Reports to reflect changes permitted by 10 CFR 50.59. The
licensee's program was reviewed for incorporation of the changes
made by the ECNs discussed herein. SDSP 15.7, Periodic FSAR
Updating, Revision 1, provides for semiannual updating and
identifies departmental responsibilities for the various FSAR
sections. It requires Licensing to issue listings of ECNs and
Drawing Discrepancies (DDs) to the responsible departments. The
latter departments then evaluate the changes for FSAR impact,
prepare proposed FSAR changes, etc.
None of the ECNs inspected are currently in this process. SDSP
15.7 applies only to s.losed ECNs; all of the subject ECNs are
open. The current FSAR update cycle ends on January 22, 1988.
Licensing plans to issue the above listings for cognizant
department evaluation on or about February 1, 1988.
(2) The large numbers of work plans being processed for Unit 2
restart reference associated system instructions and procedures
that require revision upon completion of the work. Discussions
with cognizant operations personnel, as an example, indicated
that a large number of operating and surveillance instructions
required revision after closure of work plans currently in
progress. Operations maintains a cross reference file of work
plan impact versus system for work plans started af ter about
mid-1985. For work plans predating 1985, no cross reference
exists. Further, the current cross reference is only at
the system level. The specific procedures impacted by the
- - - _ - _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
____. _ j
___ - _ _ _ _ _ _ _ _ _ _ _ _ _
i
i
1
2 30
i
}
'
!
modifications have not yet been fully identified. At the time ;
of this inspection, the overall level of effort necessary to ;
update plant procedures had not been quantified or scheduled l
l although the plant manager stated that a detailed scheduling .
4 effort was underway. No plans apparently exist to prepare draft !
procedure updates while work on its respective modifications are !
still in progress, leading to a potentially severe backlog of
,
!
i
! unprepared revisions immediately prior to fuel reinstallation
1
and restart. ;
,
'
c. ECN and Work Plan Status {
.
The licensee employs a computer data base management information
system (MIS) which develops activity lists and status, schedules,
- budget, and performance information. During the initial 1987 phase
1 of this inspection, DNE and the plant /DNC maintained separate data i
,
bases that were not reconciled. As a result, while DNE information
i indicated that about 600 ECNs applied to the U2C5 restart, the '
- Modifications Group information showed only about 500 ECNs.
4
Additionally, about 50-60 tasks and commitments to NRC had been
- identified but not fully scoped and were not consiste9tly input to {
,
either MIS; each of these had the potential for creating one or more ;
ECNs and numerous work plans,
i
l In the interim, the line departments, working with the Long Range
Planning Section, have reconciled the data bases for content and work ,
1 scope to establish common data for DNE, Modifications, and Plant t
l Maintenance. The data bases now also reflect the "backlog" listings l
l of incomplete ECNs predating the 1985 plant shutdowns. Although +
! separate data bases are still maintained (due to computer hardware
i and software considerations) they m reconciled weekly, The
j inspector reviewed curre.t En and work plan listings and status
- printouts confirming the above. ;
i 4
! The data, current as of about December 30, 1987, indicated the !
I
following status: .
t
Total ECNs Currently Identified - U2C5 = 758
t
- Design Not Yet Started 134 l
- Design in Progress 145
i
Design Complete 440 t
! Design Status Not Available 39 ;
Implementation Not Started 327 [
'
Implementation in Progress 336 i
) Implementation Complete 95 ;
4
l Testing Not Started 151 ,
,
Testing in Progress 8 !
l !
,
i
i
!
- , . - - . - - - :
_ _ - . __ _________ ________________ _
31
Testing Complete 6
Test Status Not Available 593
ECN Closeout Not Started 296
ECN Closecut in Progress 99
ECN Closecut Complete 94
ECN Closecut Status Not Available 269
Work Plan Status
Total Open Work Plans 994
Total Open V2C5 Work Plans 808
Work Plans to be Writte, 391
Work Plans in Writing 38
Work Plans in Approval 50
Work Plans in Work 455
Work Plans in Closeout 303
Work Plans Closed 253
d. QA/QC Overview and Involvement
The general scope and levels of effort applied to ECN and
modification processes by QA, QC, and EA were reviewed. The purpose
of this review was to sub.iectively assess the involvement and impact
of the quality organization on the processes.
(1) The QA audit program was reviewed through discussions with the
site audit supervisor, review of the FY 1988 TVA Internal Audit
Plan and Schedule, review of recent audits of plant
modifications and design control, and review of program scoping
procedures (listed in Attachment A). One audit (BF-A87-0013)
had been performed for BFNP modification and design control
activities during June - July, 1987. Another annual audit was
scheduled for August, 1988 and was to be a joint audit performed
The 1987 audit appeared to be sufficiently comprehensive and had
resulted in one CAQR involving repair / replacements made by the
plant mechanical modification section without formal transmittal
of weld maps.
(2) QA Surveillances are conducted by an onsite group using
corporate Management Review Guidelines (MRGs). The inspector
reviewed the 1987-88 schedules and plans and twenty surveillance
reports from late 1987 involving ECN and modification
activities, finding them to address a combination of
documentation reviews (audits) and field observations of in
process activities (surveillance).
_ _ _ _ _ .
32
The inspector noted that the current surveillance activities are
more like audits than observations of in process activities in
that they concentrate on review of documentation rather than
live activities. The group supervisor acknowledged the above
and provided the inspector plans to shift this emphasis during ,
1988. The QA Surveillance section is further in the process of
computerizing their scheduling, tracking and program management
activities.
(3) The EA Program was reviewed through discussions with site EA
personnel, review of the audit schedule, review of recent
audits, and review of EA procedures. Several audits of BFNP
design change and modification activities during the past two
years have been conducted. EA audits are scheduled in a manner
which results in each DNE organizational element to be audited
annually against each applicable 10 CFR 50, Appendix B,
criteria. Two audits were reviewed by the inspector: No.
87-01, Modification Control and Corrective Action and No. 87-22,
Undervoltage and Analog Trip Modifications.
The audits were evaluated as substantive with numerous
deficiencies identified. Areas audited included calculations,
design input, interf ace control, ECNs and FCRs, drawings,
requisitions, post modification reviews, and bills of material.
Formal responses to the identified deficiencies (treated as
CAQRs) were required, including EA approval of the planned
corrective action by the Project Engineer.
t
In addition to audit activities, surveillances are performed in
accordance with EA-I 65.02, Performing and Documenting Surveil-
lance Activities. Only one surveillance has been performed by
the BFNP Site EA Staff and did not deal with modifications. The
inspector noted that no topical suggestions are provided by the
procedure; rather the Manager of EA identifies activities
requiring surveillance.
Yet another activity, Reviews, are performed by EA. The intent
of reviews is a continuing status evaluation of NPP commitments,
such as the HPCI system reliability improvements. The inspec'.or
reviewed a recently accomplished report on the HPCI system
improvements and noted that the contents adequately provided
status of the subject matter to plant management.
Most of the EA Technical Audit Staff performs audit functions of
the Design Basis Verification Program (DBVP). A relatively small
EA Oversight Group functions in accordance with BFN8102-00. EA L
Oversight Review Plan. The Oversight Group has addressed itself
to the adequacy of plant modification implementation during two i
.. .- ._ _ _- -. .-- _.
_ _ _ _ _ - _ _ _ _ _ - . .
33
periods, January - June, 1937 and post June 1987. The first
period was selected because January 1,1987, established the
DBVP functional configuration by walkdown program; approximately
June 1,1987, initiated the Transitional Change Control (TCC)
program, i.e., PI 86-03 implemented for ECN preparation.
The Oversight Group evaluated a sample of four work plans (and
related ECNs) of 55 work plans completed during the first
period; none have completed evaluation for the latter period.
The evaluation included configuration checks on work plan
drawings and referenced drawings, and evaluation of procedures
and compliance with procedures. The inspector reviewed the
draft preliminary report of Activity 1 - Control of Change
Configuration (undated), and noted several substantive, negative
findings. Action items were noted to have been prepared from
each deficient condition and forwarded to DNE.
The inspector concluded that, based on the results of the EA
Oversight review team findings, the function was adequately
fulfilled. However, not withstanding the adverse nature and
number of findings, the inspector noted that the satrple size of
surveyed work plans was not increased. BFNP should consider
increasing the sample size of evaluated work plans for the
transitional period of January - June 1987, to assure that
configuration control is adequate.
e. Summary
Items 1- 12 below represent the important findings and concerns
resulting from this and the prior inspection intervals and are
recommended for additional licensee action or response.
(1) Multiple examples of craft and/or QC verifications of completed
work and final inspections for modification activities subse-
quently found incomplete. Examples include loose conduit
fittings and condulet +.ight covers, electrical panel construc-
tien debris, and incomplete painting of weld joints. [Section
C.6, this report]
(2) Significant pipe support inspection rec,uirements we-e omitted
from work plans. Pipe support rework materially affected
the physical integrity without provision for reinspection.
Inadequate training and experience appeared to contribute to
responsible personnel not re:ognizing the applicability of the
inspection requirements. [Section C,7.c, this report]
(3) An RHR pipe support was installt .vith misalignment exceeding
procedural limits. The condition was not recognized by field
installation personnel and was not subject to final inspection
(see item 3 above). [Section C.7.d, this report]
4
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
34
(4) Design output documents failed to include necessary piping
supports for HPCI valve test valves and tail pieces. Design and
"
field personnel failed to recognize the omission and its
potential for fatigue or seismic failure. [Section C.5, this
report)
(5) The number and significance of EA findings to date (from
relatively limited site EA surveillances and oversight
activities) warrants an increase in the scrutiny available
through the EA programs directed at the ECN and modification
activities. [Section F.3, this report)
(6) Post modification test control was found weak in several areas.
Post modification testing administered by Work Plan did not
provide for sufficient notification of the Shift Engineer,
including one example wherein a test was delayed for 3 weeks
without subsequent pretest notification. In a second example, a
hydrostatic test was conducted prior to disposition and repair
of weld (NDE) defects. [ Inspection Report 87-42 and Section
C.2, this report)
(7) An Unreviewed Safety Question Determination (USQD) found that
the design 1: sued by ECN could result in system misoperation
outside the bases of the safety analysis (spurious valve
operation). The ECNs P0651 and P0652 have not been revised and
the installation is mechanical field complete with no further
action to date. [ Inspection Report 87-42]
(8) The large number of modifications will necessitate a
proportionately large number of procedure revisions. Specific
procedure impacts are largely unidentified and detailed planning '
for specific procedure changes is not yet in place. [Section
O.2, this report]
(9) Piping and cable tray support design required numerous Field
Change Requests indicating that more prodesign and preconstruc-
tion attention to existing conditions should be exercised.
,
This trend appeared to be improving during this inspection.
[ Inspection Report 87-42)
(10) Weaknesses in administration of work packages including improper
cross referencing (DCR and ECN to work plan, work plan to work
plan, work plan to test, etc.) leading to potential misappli-
cation of references; inappropriate work plan steps; failure to
provide reference drawings for work steps; failure to include
inspection, painting, etc. steps in work plans. [ Inspection
Report 87-42]
35
(11) Weakness in the completion of Nuclear Storeroom Requisitions for
the proper entry of quality requirements with the potential for
issuance of unqualified material. [11/6/87 Report and Section
C.12,thisreport)
(12) Nuclear Performance Plan, Volume III, Appendix D, lists ECNs
required to be completed prior to Unit 2, Cycle 5 restart but
does not currently list backlog ECNs required by to be closed by
Appendix A and Section II.2. [SectionC.10,thisreport)
l
f. In addition to those items listed above, additional NRC followup
'
inspections will also be performed for the items listed below. A
summary of these latter items were provided to the TVA Compliance
Group via telephone subsequent to the inspection.
(1) Completion of licensee action to correct installations and
completion of documentation for "backlog" Core Spray ECN L2003.
[SectionC.7,thisreport]
(2) Licensee evaluation of open ended vent and drain tail pieces not
equipped with piping (cleanliness) caps. [S4ction C.3, this
report)
(3) License evaluation and disposition of valve cap configuration
and foreign material intrusion for HPCI Valve FCV 71-40.
[Section C.6.c, this report]
(4) FSAR and procedure updates resulting from ECNs. [Section D.1,
this report)
(5) Acceptability of MAI-27 provisions for installation of safety '
grade conduit with :ero clearance from other structures and
'
components. [Section C.6.c, this report]
g. References Used During ECN/ Mod Inspection
--
Browns Ferry Unit 2 Technical Specifications
BF3.2 QC Inspection Program, Revision 1
BF8.3 Plant Modifications, Revision 10
SDSP8.1 Plant Modification / Design Change Approval Revision 5
SDSP8.4 Preparation and Processing of Work Plan and Inspection
Records Forms, Revision 6
SDSP8.8 Conversion of Temporary Alteration to Permanent Plant
Mcdification, Revision 0
, SDSP8.9 Field Change Requests, Revision 0
SDSP9.8 Site Walkdown Program, Revision 2
SDSP13.3 Implementation of ASME Section XI, Revision 4
SDSP17.2 post Modification Programs Test, Revision 2
i
l
l
..______ _________ ________ ______ __ __
36
SDSP15.7 Periodic FSAR Updating, Revision 1
MAI-54 Pressure Testing of Piping Systems, Revision 0
BFEP PI 86-03 Preparation and Control of Engineering Change Notice
Modification Package, Revision 2
BFEP PI 87-27 Procedure for Origination of Configuration Control
Drawings, Revision 0, with Supplements
NEP 6.1 Change Control, Revision 0, with Supplements
NEP 3.2 Design Input, Revision 0, with Supplements
NEP 5.1 Design Output, Revision 0, with Supplements
CAQR 870727 Condition Adverse to Quality Report, Additional
We'iding on RBCCW Pipe Support af ter final QC
Inspection
LP 4N 45A-L TVA Internal Audit Plan for FY88 and First Quarter
Audit Schedule
BF-A-87-0013 Plant Modifications and Design Control QA Audit Report
SDSP 12.4 Returi or Systems to Operable Status for Restart
Follswing Modifications, Revision 1
PAI-23 Support ind Installation of Piping Systems in
Categoi I Structures, Revision 5
MAI-27 Insta11ath i of Electrical Conduit Systems and i
Junction b.xes, Revision 4
PAI-44 Cable Pulling for Insulated Cables up to 15KV,
Revision 0 '
--
BFNP QA Staff Trend Report Data 12/86 - 11/87
--
BFNP Operational Readiness Issues - DNQA/EA Integrated *
"
Verification Plan
QMI-602.6 Surveillance, Revision 1
QMP-118.1 TVA Internal Audit System Plan and Audit Scheduling,
Revision 0
QMI-312 Quality Audit Program - NQA&EB
QMI-102.1 Quality Surveillance (Monitoring) Program - Site,
Revision 1
QMI-311 Standard Audit Module Scoping Document - Preparation
and Control, Revision 2
-- QA Surveillance Log, 1987
-- QA Surveillance Schedule 1987-88
-- QA Surveillance Report and Status, 1987
-- Modification Status Report, U2C5, November 2, 1987
-- BFNP Work Plan Status Charts, 10/15/87, 1/4/88
-- DNE Modification Engineering Status Charts, various dates 10 -
12/87
-- Post Mod Test Assignments Listing, 10/15/87
-- Work Plan Status List by Milestone, 12/29/87
-- U2C5 Modifications List, SAS Data Base, 1/4/88, 12/1/87, 11/3/87
QA Surveillance Reports:
QBF-587-0454 QA Surveillance Report - Operational Readiness
Restart Issues
QBF-S-87-0427 QA Surveillance Report - Electrical Maintenance
_ _ _ _
37
QBF-S-87-0244 WP&IR Preparation, Review and Approval
QBF-S-87-0054 WP&lR Work Performance
QBF-S-87-0320 WP&IR Work Performance
QBF-S-87-0321 WP&IR Work Performance
QBF-S-87-0331 WP&IR Work Performance
QBF-S-87-0342 WP&IR Work Performance
QBF-S-87-0369 WP&IR Work Performance
QBF-S-87-0385 WP&lR Work Performance
QBF-S-87-0429 WP&lR Work Performance
QBF-S-87-0441 WP&lR Work Performance
QBF-S-87-0442 WP&IR Work Performance
QBF-E-87-0444 WP&IR Work Performance
QBF-S-87-0444 Prefabrication Workplans
QBF-S-87-0279 WP&IR Work Performance
QBF-S-87-0082 ECNs - Corrective Actions
QBF-S-87-0023 ECNs - Followup on Corrective Actions
for Cancelling ECNS and Drawing Discrepancies
QBF-S-87-0013 ECNs for Drawing Discrepancies
QBF-S-87-0081 U2 HP Raw Water Fire Protection System. Walkdown
Drawings:
Flow Diagrams Mechanical Diagrams
Mechanical Control Diagrams
Isometric Analysis Diagrams
Wiring Diagrams
Conduit & Cable Schedules
Conduit Routing Diagrams
Pipe Support Details
h. Persons Contacted During ECN/ Mod Inspection
A. Ballard Acting Principal Engineer, DNE
R. Bice Modifications Field Engineer
,
R. Burt Mechanical Modification Section Supervisor
A. Chapman Asst. Modification Manager
L. Clardy QA Surveillance Supervisor
P. Crabb Work Plan Coordinator
D. Deyer Electrical Engineer
B. Garner Electrical Modification Field Engineer
L. Hargett Task Engineer, DNE
H. Hodges Mechanical Engineer (Mech. Test)
C. Hsieh Compliance Engineer
D. Langley diectrical Design Engineer
E. Lorg QC Supervisor, Mechanical
R. Martin Asst. Modification Manager
J. McCaleb Mechanical Modification Field Engineer
M. McCord System Engineer
S. McRight PMT Supervisor
.
38
J. Nelson QA Specialist
D. Nilius System Engineer
M. Oliver Mechanical Engineer (Lead Technical Supvr.)
J. Pettit Mechanical Modification Field Engineer
C. Rickard Mechanical Engineer, PMT
G. Robert Mechanical Modification Field Engineer
S. Rowe System Engineer
J. Savage Compliance Supervisor
K. Sheppley Mechanical Modification Field Engineer
D. Skridulis Compliance Engineer
0. Simmons Mechanical Engineer, PMT
C. Simms Orincipal Civi' Engineer
W. Spader Principal Mechanical Engineer
S. Thomas Electrical Engineer, PMT
E. Winters Task Engineer, DNE
J. Wright DNE Project Services
R. Young Modification Manager