ML20236W107
| ML20236W107 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 11/20/1987 |
| From: | Brooks C, Christnot E, Ignatonis A, Patterson C, Paulk G, Vias S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II), NRC OFFICE OF SPECIAL PROJECTS |
| To: | |
| Shared Package | |
| ML20236W086 | List: |
| References | |
| 50-259-87-37, 50-260-87-37, 50-296-87-37, GL-84-23, IEB-79-02, IEB-79-04, IEB-79-2, IEB-79-4, NUDOCS 8712070281 | |
| Download: ML20236W107 (25) | |
See also: IR 05000259/1987037
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION .
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101 MARIETTA STRE ET, N.W.
ATLANTA. GEORGI A 30323
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Report Nos.: 50-259/87-37, 50-260/87-37, and 50-296/87-37
Licensee: Tennessee Valley Authority
6N 38A Lookout Place
1101 Market Street
Chattanooga, TN 37402-2801
Docket Nos. 50-259, 50-260, and 50-296
License Nos. . DPR-33, DPR-52, and DPR-68
Facility Name: Browns Ferry Nuclear Plant
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Inspection Conducted: October 1-30, 1987
Inspectors: [<kM+ b mA R ..-, / OD
G. L. PatDK, Senior R$sidefit Inspector Date Signed
G4 Lod A s
C. A. Patt'erson, Res'iddht inspector
H/soh 7
Date Signed
G< & (L _ g e,
C. R. Brc'6ks, Resident 0 In(pector
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Dat'e Signed
D& ab &
E. F. Ch91stnot, ResidentVInspector
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Date Signed
0'3 wr.Q h ///20/E ?
S. J. Viki, Project inspe"ctor Date Signed
Approved by: 8. d r ~m' ., //b o/M
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A.J.Ig@tonW,(JSectionChief Dat'e Si'gned
Inspection Programs,
TVA Projects Division
SUMMARY
Scope: This routine inspection was in the areas of operational safety,
maintenance observation, surveillance testing observation, reportable
occurrences, Engineering Assurance (EA) organization involvement on Browns
Ferry matters, inadvertent initiation of fire protection system, restart review
subcommittee, diesel generator overload evaluation, tests and experiments
program, design baseline verification program, inadequate core cooling instru-
mentation, containment coatings, environmental qualification testing, facility
modifications, document control program, Q-list, and companion drawing
discrepancy /CAQR review.
B712070281 871203 9
PDR ADOCK 0500
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Results: Three violations were identified: (1) failure to properly control
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measuring and test equipment; (2) failure to properly indicate the operating .
statu's of a component important to safety such that inadvertent operation is l
L prevented; and (3) failure to have an adequate procedure for controlling core
l drilling operations through secondary containment boundaries.
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REPORT DETAILS
1. Licensee Employees Contacted:
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, Superintendent - Unit 2 -
- 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
- D. A. Pullen, 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 .
2. Exit Interview (30703) ,
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The inspection scope and findings were summarized on October 30, and l
November 3,1987, with the Plant Manager and/or Superintendents and other
members of this staff.
The licensee acknowledged the findings and took no exceptions. The ;
licensee did not identify as proprietary any of the materials provided to
or reviewed by the inspectors during this inspection.
- Attended exit interview
3. Licensee Action on Previous Enforcement Matters (92702)
(CLOSED) Followup Item (259,260,296/86-40-02), This item was to review an
operations critique of the fire protection system initiation and spray
down of the Unit 2 reactor building that occurred on December 23, 1986.
The inspector reviewed the critique and supplemental response to the
critique which clarified the action taken under various maintenance
requests. The inadvertent initiation is a continuing problem as discussed
in paragraph 5 of this report. The review of the critique closes this
item.
(CLOSED) Violation (259,260,296/84-15-04), This violation was against 10
CFR 50, Appendix B, Criterion V because the control air system drawings
did not reflect the system in the plant and discrepancies were found in ;
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fire protection drawings, pressure switch setpoints and annunciators.
First, for the fire protection concerns, drawings 67M4-7-47B601-026 and
45N644-1 were revised to show a correct setpoint of 100 psig instead of
120 psig. The inspector reviewed five other drawings which corrected the
annunicator title and supply. Second, for the control air drawings the
entire control air system has been walked down and the drawings revised to
reflect the control air piping and valves. Five drawings were provided to
the inspector for review. The control air system is one of the systems
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within the scope of the design baseline verification program being
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evaluated to determine whether the original design or subsequent
modifications have affected the capability of the system to fulfill its
safe shutdown function. Correction of the drawing deficiencies closes
this item.
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(CLOSED) Violation (259,260,296/85-45-08), This violation was for failure
to take corrective action to preclude repetition of a significant
condition adverse to quality for a diesel generator failure to start.
During performance of monthly surveillance testing a diesel generator
failed to start with the cause not being determined, documented, or
reported to management. The licensee stated no definitive reason for the
failure could be found but was thought to be corrected during maintenance
I activities. The diesel was removed from service to complete scheduled
vendor recommended maintenance. TVA issued a report describing the
maintenance and any problems. No problems were found that could have
caused the failure to start. This event was put in the compliance
bulletin to insure those responsible for failure investigations were aware
of this problem. The inspector reviewed the TVA maintenance report for
the diesel and the vendor inspection and evaluation reports. The diesels
were generally found to be in good condition. This item is closed.
(CLOSED) Followup Item (259,260,296/86-36-02), This item identified the
reactor protection system panels being anchored to the floor differently
on each unit. This item is similar to unresolved item
259,260,296/85-57-03 closed in this report. Licensee event report (LER) 260/85-20 addresses the panels for each unit and will be tracked until
closure of the LER. This item is closed.
(CLOSED) Unresolved Item (259,260,296/85-57-03), This item was that the
reactor protection system instrument panels were not seismically
qualified. The licensee reported this under LER 260/85-20. This
condition was the result of the actual constructed configuration never
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being properly documented during construction. An analysis showed that
the panel anchors would see high loads during a seismic events and because
the anchor bolt material could not be identified, a f ailure was assumed.
The configuration control program currently in place identified this
problem. The panels will be corrected prior to restart and will be
tracked with closure of the LER. This item is closed.
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(CLOSED) Unresolved Item (260/83-36-02), This item was to evaluate
additional modifications and retests concerning the main steam relief
valve tailpipe 10 inch vacuum relief valves. The inspector observed
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testing of these valves at Wyle Laboratories on August 30-31, 1983. The
operational cycling test indicated the valve seat would not reseat after
operation due to hinge mechanism binding and loss of spring preload to the
movement of the knurled hinge pin within the arm assembly. After
modifications to the hinge shaft the valves were returned to Wyle
Laboratories for a second series of tests. During this test it was
observed that the bumper spring was permanently deformed. This was
believed due to improper setup of the test equipment and contact between
the bumper and load cell bolt head. The test was stopped and the valve
returned to TVA for repair on September 16, 1983. A third series of test
was performed on September 23, 1983, with acceptable results. The
inspector reviewed the test results and unreviewed safety question
determination (USQD) associated with engineering change . notice 653.
Revision five of the USQD accepted the modified vacuum breaker valve as
adequate. This item is closed.
(CLOSED) Unresolved Item (50-259,260,296/87-07-02), Pipe Support
Discrepancies. This item involved pipe supports that were inspected
against their detailed drawings for configuration, identification,
location, fastener installation, welds and damage / protection. These
supports were associated with the torus attached piping. The findings
were discussed with the QA/QC inspectors and engineers during a review to
determine the effectiveness of the Bulletin 79-02 and 79-4 programs. The
two supports that were found to have discrepancies, were reviewed during
this' inspection. For support No. R-12, Drawing No. 47B458-404 (Rev. 13)
was revised to show the proper weld configuration as installed in the
field. For support Nos. R-12A & R-12B, Drawing No. 478920-91 was revised
.to show the correct as-built condition of the base plate and a new
calculation was performed. The licensee is using calculation No.
BWPC20879, Rev. 4, to design and document anchor bolt repairs and
replacement details for existing substandard baseplates and/or bolt
anchors. This calculation verifies use of general construction
specification No. G-32 for " Bolt Anchors Set in Hardened Concrete",
Section 3.6.3.5, which gives requirements for movement of new anchor bolts
adjacent to existing abandoned anchors or pulled anchor bolt holes. This
item is considered closed.
(CLOSED) Unresolved Item (50-259,260,296/87-07-01), Discrepancy in
Insta11aton of Designed Conduit Supports. This item involved a conduit
support, 48B810-14, Rev. 1, that was inspected against its detailed
drawings for configuration, identification, location, fastener
installation, dimension, clearance, member size, welding, clamp, clamp
bolt edge distance in unistrut, and damage protection. The results were
discussed with the QA/QC inspectors and engineers, to determine the i
effectiveness of the conduit and support program. The inspector reviewed l
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the licensee's evaluation and corrective actions for the discrepancies
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noted in the inspection report for this item. The licensee issued a field
change' request on March 19, 1987, to address Corrective Action Report
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l (CAR) 87-0037. CAR 87-0037 was issued to evaluate the installed I
condition. In Attachment "A",
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in the " corrective action" it stated, "The
layer "as-is" is acceptable and no further physical work is required".
Under the " Actions to Prevent Recurrence" section, the licensee is .to
revise all procedures that require QC verification on torquing to require
" torque seal" to be applied after torquing. The inspector reviewed the
calculi ions for support 48B810-14, Revision Log BF EPC8-0013 dated
March .6, 1987, found them to be acceptable. This item is considered
closeJ.
(CLOSED) Inspector Followup Item (50-259,26,0,296/85-30-01), Corrective
Action on Inspected Pipe Supports. This item involved pipe supports that
were inspected against their detailed drawings for configuration., >
identification, fastener installation, and damage /protecticn. The
inspection report stated that numerous supports had some discrepancies
with respect to the installed condition and hanger sketches. The licensee
issued Maintenance Request Form A-748620 dated 11/4/86, to correct the
discrepancies noted in IFI. This item is considered closed.
4. Unresolved Items (92701)
There are no unresolved items identified in this inspection report.
5. Operational Safety (71707, 7.1710)
Daily discussions were held with plant management and various members of
the plant operating staff. The inspectors were kept informed of the
overall plant status and any significant safety matters related to plant
operations.
The inspectors made routine visits to the control rooms when an inspector
was on site. Observations included instrument readings, setpoints and
recordings; status of operating systems; status and alignments of
emergency standby systems; onsite and off site emergency power sources
available for automatic operation; purpose of temporary tags on equipment
controls and switches; annunciator alarm status; adherence to procedures;
adherence to limiting conditions for operations; nuclear instruments
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 cor.ditions; containment isolation
alignments; instrument readings; housekeeping; proper power supply and
breaker; alignments; radiation area controls; tag controls on equipment;
work activities in progress; and radiation protection controls. Informal
discussions were held with selected plant personnel in their functional j
areas during these tours.
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In the course of thd monthly activities, the inspecto'rs included a review ;
of the licensee's physical sea.urity program.' Thre performance of various
shifts of the security force was observed in the cmduct ' of daily
activities' to include; protected and vital areas? access -controls,
searching of personns1, packages and -vehicles, badge N ssuance and
retrieval, escorting of' visitors, j patrols and comnensat#7.' posts. In:
addition, .the inspectors observed ptrtected area lighting, pentected sW .[
vital areas barrier u tegrity. ,
a. Inadvertent Fire. Protection Initiation
On Septem er 29, 1987, an it:a vertent initiation of the fhed spray
fire protection system occurred in the Unit 3 reactor bunding~ At
3:40 p ~m. , a flow alarm was received in the control room indicating
flow from zone 3L system deluge valve 3-26-78L. An operator found j
water spraying from the nozzles and isolated the system. Inspection
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found that. tbe _ valve had not tripped from the detection circWtry and
was not manuaVy initiated. The valve cover plate was removed and
revealed that 'the. rubber seat disc had been forced passed the valva
lapper- by system pressure. This allowed water to enter the
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downstream piping and dischaNe through the open nozzles. The valvo *
is manufactur'ed by Star Sprint er Corporation. -
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As a result of the continuing problem of inadvertent initiations the
Plant Manager if ormed the inspector on October 2, 1987, that all-
Star valves were being taken out of service and a fire watch posted.
Past problems with theie va1ves are discussed in Inspection Report
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b. Improper'ControldfComponentOperability
On October 13,.1987.,,the reactor operator in the Unit 2 Control Room
was aligning componei.ts of the Residual Heat Removal Syste'n (RHR) in'
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preparatiba for a restart test. RHR pump suction. valve 2-74-24 was'
noted to be in an intermediate position. When the handswitch for
this valve was placed in the closed position it completed'its stroke;
however, because of a misadjustment of the limit switches, the ,
control circuitry did not de-energize the valve and as it continued
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to- be drhen into its stops the breaker tripped on ove-load. This f
'H valve had been under a maintenance hold order for about 'a year prior
to the. event. In September 1986, the limit switch gear box grease
, was changed out. Since this operation displaces the limit switches,
Electrical Maintenance Instruction (EMI).18, Limit Switch Adiminent
~ for. Limitorque Valve Actuator, must be performed prior to re' turning
' the valve to service. This EMI had not been performed and un
September 14, 1987, when plant operators requested that the valve be
turned over for operation, maintenance personnel communicar.ed this
problem to them. It was agreed that the valve would be reletased from
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the hold order for manual op4aration under the condition that it would ,
not be suitable for elegt ricel operation until EMI-18 could be I
performed. This restriction was subsequently lost in the many I
turnovers that occur over the course of a month and electrical power
was eventually restored to the valve. Since the limit switches are
used in the control circuitry as well as for position indicators in
the control room, the closing cyc e of the valve remained energized
., ne mattert What the actual valve rosition was. Failure to maintain
the operatroaal restriction on the electrical operation of the valve
is a violaHon of 10 CFR E0. Appendix B, Criterion XIV
(260/87-37-D'l).
3. Maintenance Observation (62703)
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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 redew: :ne 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.
iiaintenance 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
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below list d maintenance activities during this report period:
a. Plant Non preferred Auto-throwover Switch, MR No. 811099
b. Limitorque Valve Operator Maintenance on Valves 2-74-12 and 2-74-24
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Troubleshooting Diesel Generator 3A and 3B Overspeed Problems
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No violations or deviations were observed in this area.
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T. Surveillance Testing Observation (61726)
The inspectors observed and/or rev ewed the below listed surveillance
procedures. The inspection consisteJ of a review of the procedures for
technical adeque.cy, conformance::o technical specifications, verification
of test instrument calibration, observation on the conduct of the test,
removal from service and return to service of the system, a review of test
data, limiting condition for, operation met, testing accomplished by
qualified personnel, and tnat the surveillance was complete at the
required frequency.
Restart Test Procedure-032, Control Air System discovered a deficient
condition associated Nith the Reactor Building Equipment Access Door Seals
on October 9, 1987. ihese secondary containment airlock double doors are
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maintained airtight by inflatable seals. Rec'..nda n t air receivers are
supplied in order to provide about' one weeks worth of air should the
normal air supply to- the seals fail . During the test, 'the seals were
found to deflate in less than a day following removal of the normal air
supply. This test deficiency will require resolution prior to restart.
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s- The~ Test Engineer hts recommended that the leaking seals be replaced and a
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Design Change Notice (DCN) will be prepared by DNE to affect this
, recommendation. This item is being tracked by a condition adverse to
quality report (CAQR).
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8. Reportable Occurrences (90712,92700)
The below listed licensee event report (LER) yas ; reviewed to determine if
the information provided met NRC requirements. The determination
included; adeauacy of event description, verification of compliance with
technf ral specifications and regulatory requirements, corrective action
taken, existence of potential generic problems, reporting requirements
satisfied,'an'd the relative safety significance of each event.
LER No. Date Event
1260N7-07 10-2-87 Drywell Lortrol Air Isolation
Valves Outside Design Basis
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This LER described a finding that the two. primary containment isolation
valves on .the drywell control air system suction piping would not go
e closed upon a loss of control air pressure. The licensee reported this !
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finding under 10 CFR 50.73(a)(2)(ii) as a condition outside the design
basis of the plant. Since
independent trains and wad. this deficient
caused.hy a singlecondition existedanon
problem; namely both of the
erroneous
modification, the inspecto;e questioned why the event was not also reported
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under the criteria of 10 CFR 50.73(a)(2)(viW A licensee representative
itated that an oversight had been made and that a revised LER would be
i submitted including both reporting criteria. This LER will remain open
pending revision by the licensee.
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9. Restart.' Review Subcommittee j
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On October 1, US7, the inspector attended a meeting of the restart review
subcommittee. The purpor,e of this meeting was to determine if certain NRC
open inspection Ttecs we e restart items. The subcommittee is one of the
e) subcommittees of the charge control board as designated in Site Directors
j Standard Practice (SDSP) tW.1,' Plant Modifications / Design Change Approval.
.The chairman of the subcommittee at tne beginning of the meeting checked
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the attendance for a quorum of raembers. The meeting was conducted in
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'accordance with SDSP 8.1 using a checklist of the restart criteria for
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each item discussed. In general the decisions made were conservative.
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The inspector also attended a Restart Review Subcommittee (RRCS) meeting
conducted on October 8,1987. The items reviewed were from the employee
concerns program. The inspector observed the RRCS using both the TVA and
NRC draft proposed restart criteria during the reviews of the employee
concern items. The inspector observed several examples of difficulty in
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analyzing an item using the restart criteria prior to and during the
voting process of the board. For instance employee concerns, which are
uniquely identified would have condition adverse to quality reports
(CAQRs) attached and the board members questioned if the concerns were
being voted on or the CAQRs.
The inspector observed a total of fourteen employee concerns being
reviewed and voted on. The items were presented to the board by various
site organizations such as Division of Nuclear Engineering (DNE), Quality
Assurance (QA), Employee Safety and System Engineering. Each item
presented by the organizations were either recommended as "yes" (a restart
item) or "no" (not a restart item). The inspector noted that several of
the employee concern items were from other TVA facilities, such as Watts
Bar and Sequoyah.
10. Diesel Generator Overload Evaluation
TVA contracted with General Electric to perform a dynamic analysis of the
diesel generator system. This study was initiated after a static load
study by Bechtel indicated an overload problem wit.h the diesel generators
due to addition of system loads without adequate design control. Special
test 87-23 was conducted in June, 1987 to obtain data on the diesel
generator excitation system to be used in the model of the dynamic
loading. General Electric issued a preliminary report dated August 12,
1987, which concluded the "D" diesel generator was overloaded (Reference
IE Report 87-33). Also, the time required to start a residual heat
removal pump (RHR) during test 87-23 was greater than 4 seconds
conflicting with 3 to 3.5 seconds found during past surveillance testing.
To resolve this difference special test 87-30 was conducted to obtain
additional excitation system data.
Special test 87-30 consisted of testing diesels 3C, 18, and 10. The
inspector observed the testing of IB on October 4, 1987. The test
procedure was approved on September 29, 1987, followed by an immediate
temporary change (ITC) which was approved on October 1,1987. The ITC
changed the position of the diesel generator operational mode selector
switch and uncoupled the RHR pump to be powered from the 3C diesel. The
operational mode selector switch has the following positions:
Single Unit
Units in Parallel
Paralleled with the System
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The position of the switch modifies the function of the diesel generator
governor and voltage regulator to suit the operational condition required.
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For the RHR puro start during test 87-23 the switch was left in the
paralleled with the system position, but for test 87-30 the position was
changed to single unit. After a review of the test data for 87-30 the
time for the RHR pump start was between 3 and 3.5 seconds agreeing with
past data. The recorder traces of field voltage and generator terminal
voltage likewise indicated the diesel generator was more responsive when
in single unit mode. While in paralleled with the system mode the
governor and voltage regulator is less responsive as more load is expected
to be absorbed by the system or grid.
The inspector found that the procedure change was not properly made using
an ITC. A procedure change can be processed as a formal change requiring
the Plant Operations Committee (PORC) approval or an ITC. The ITC
guidelines required completion of a checklist to evaluate if the change
can be made as an ITC. One question is whether the change is a change in
scope, technique, or sequential order of instruction steps that would
affect the result or nuclear safety. This question was answered no.
However, the switch position was changed for the purpose of affecting the
results and the uncoupling of the pump changed the technique.
A meeting was held with the applicable system engineer and technical
services manager to discuss the testing and ITC on October 5,1987. At
this time, it was identified by the licensee that an ITC should not have
been used for the change. Plant Managers Instruction (PMI) 17.1, Conduct
of Testing, requires that non intent changes not requiring PORC approval
be, approved by a Senior Reactor Operator by drawing a single line through j
the portions to be changed, writing the change above the line, and dating '
plus initialing the entry. An ITC form is not to be used. PMI 17.1 was
not followed. The licensee indicated that PMI 17.1 would be changed to
allow usage of an ITC. Also, the inspector questioned one of the approval
signatures on the ITC. Approval is required by the section supervisor
responsible for the procedure. An operations supervisor had signed the
ITC but this should have been someone from technical services.
Collectively, these items indicate a general non familiarity during this
test with the special test procedure PMI 17.1. Also, the inspector was
particularly concerned that even though special test 87-30 was conducted
to resolve questionable test data, PORC approved 87-30 but did not approve
the ITC which changed the operational mode switch position for the diesel
generator and changed the test results. In Inspection Report 87-33 a
violation was issued, for losing the PORC approved copy of special test
87-23 which is a quality assurance record. Collectively, the lost record,
following PMI 17.1, and PORC approval of procedure changes indicate a lack
of management control and involvement in the area of special tests. (See
paragraph 12 of this report).
11. Engineering Assurance (EA) Organization
The EA organization is responsible for administration and management of
the QA program as applied to TVA nuclear engineering and design
activities. The resident staff met with the Manager of EA and his staff
on October 6, 1987, to discuss overall program goals, objectives, and
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project implementation at Browns Ferry. The:following program areas were
discussed and will be reviewed during future inspections.
a. Program Areas Addressed by EA include:
Design and Procurement Control
Training for Design Engineers
Program Audits
Technical Audits
Procured Services
Trending
Problem Reporting
Project Assignments
Design Baseline and Verification Programs
b. Goals and Objectives of EA
Develop, issue, and maintain quality-related procedures for control
of DNE activities.
Provide technical training and training in the use of quality-related
Nuclear Engineering Procedures and selected project procedures.
Conduct program audits to assess compliance to DNE procedures and
engineering / design aspects of the TVA quality assurance program.
Conduct in-depth technical audits to assess the technical adequacy of
engineering work.
Review procurement documentation and perform surveys and audits of
suppliers of engineering services.
Develop, implement, and maintain the DNE trend analysis progran.
Develop, administer, and maintain the DNE corrective action program
including reviews of DNE-related problems for generic applicability.
c. Implementation Progress of EA Program Areas
DNE Nuclear Engineering Procedures (NEP) Manual issued July 1,1986.
Continuous upgrading is in process.
EA training activities commenced in June 1986 and are continuing on a
scheduled basis.
EA program audits were restructured in the third quarter of 1986 to a
modular approach using standardized checklists. Audits of project
activities and subject areas have been routinely conducted.
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In-depth technical audits by EA have been primarily directed to the
independent review of the design baseline and verification program
although technical audits have been. performed in other selected neas
such as calculations and environmental qualification.
Review of procurement documentation for engineering services has been
instituted and approximately 47 audits / surveys of engineering service
suppliers have been conducted to date.
An NEP (Nuclear Engineering Procedure) for performance of trending of
CAQs has been developed and issued. As an interim measure, the DNE
trend data base is being maintained. Trend data reports were issued
in August 1987, to initiate the third semiannual cycle of the DNE
trend analysis program. A survey of TROI is currently underway to
determine whether the TROI data base is sufficiently developed to
facilitate a transition from the DNE data base to TROI. EA is
utilizing data from both the DNE and TROI data bases in trending
activities.
EA corrective action activities have been restructured to be
compatible with the next CAQ system established by the NQAM. NEP-9.1
has been issued to reflect the new system. NQAM requirements that
CAQRs be evaluated for potential generic applicability have been
implemented by EA.
d. Design Baseline Verification Program (DBVP)
The majority of EA support for Browns Ferry has been in the DBVP
program area. Eighteen contractors have been assigned the DBVP
program review responsibility.
(1) EA Program Objectives for DBVP
(a) Engineering Assurance will perform an independent review of
the DBVP.
(b) An EA oversight review team of experienced technical
personnel will independently review, on a sample basis, the
DBVP as the project completes its document preparation,
document revisions, and/or review. The review objectives
are:
Confirm and validate that engineering activities are being
conducted in accordance with the approved program plan and
procedures established for the DBVP.
Confirm functional and technical adequacy of system
alterations and completeness / correctness of supporting
i documentation.
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Verify that corrective actions resulting from these
evaluations have been documented and properly implemented
or scheduled for postrestart.
(c) This review will provide added assurance that the
engineering activities associated with the program are
conducted in a technically adequate manner and in
accordance with the written. procedures prepared
specifically for this effort. ;
(d) The results and conclusions of this review will be
documented in a final EA report to the Director, DNE.
(2) Engineering Assurance Independent Oversight Review for the
Browns Ferry Plant Design Baseline and Verification Program
(DB&VP).
(a) Goals and Objectives j
(
The Design Baseline and Verification Program for the Browns
Ferry Plant, which has been issued and docketed, will
receive two levels of oversight by Engineering Assurance.
During the initial stages of the program, a full-time
surveillance team monitored procedure development and
application to ensure that systems were in place to achieve' !
the objectives of.the DB&VP.
An EA Independent Oversight Review Team (ORT) was
established March 30, 1987, to review the technical
adequacy of the output of the DB&VP.
(b) Implementation Progress
A surveillance team leader from EA has been assigned, along
with five experienced engineers from DNE branches and
contractors. Surveillance have be.en in progress for about
three months. Evaluations have been performed on portions
of_ the system walkdown process, identification and
capturing of commitments and requirements, test scoping
documents and design criteria.
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The EA surveillance discontinued its review of baseline
activities on April 10, transferring the responsibility for
followup and closure of open items to the ORT.
The ORT has developed and issued an oversight review ]
program plan, discipline review plans, and attribute lists 1
for each activity being reviewed. Interfaces have been !
established with baseline, project and site management with l
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periodic meetings being conducted to keep management
informed on the status of the ORT review.
Reviews have been completed or are underway. in areas of:
commitment / requirement data base, system walkdowns, design
criteria documents, shutdown analysis, and system
requirement calculations.
The baseline work which is tasked out to task performance
contractors will be performed at multiple locations and may
have an impact on the ORT.
12. Tests and Experiments Program (37703)
The inspector reviewed the licensee's activities related to lessons
learned from the 1986 accident at the Chernobyl Nuclear Power Plant.
Various activities have been suggested by the NRC as outlined in
NUREG/BR-0032, Vol. 7, No. 36, U.S. NRC News Releases for the week ending
September 15, 1987. Of these activities, only one item has been addressed
by the' licensee as of. yet. . The suggestion to review administrative
controls in order to strengthen technical reviews and approvals of tests
and experiments was completed on April 23, 1987, by the experience review
program. The licensee's activities were prompted by the Institute of
Nuclear Power Operations (INPO) Significant Operating Experience Report
(SOER) 87-1 entitled Core Damaging Accident Following an Improperly
Conducted Test. Areas recommended for evaluation were as follows:
a. Positive management control is exercised over the approval and
conduct of special tests,
b. Test procedures should include sufficient detail for conduct of the
planned configuration as well as unexpected plant responses,
c. Operators are properly prepared and briefed prior to conducting the
test.
d. Across the board training is in place for managers, operators,
engineers and supervisors. This area should receive special
attention in the initial as well as continuing training program.
The licensee's review concluded that under recommendation number 1,
management control and technical support was already adequate and no
activities were initiated. Under recommendation number 2, test procedures
were judged adequate with the exception that guidance for aborting a
special test should unexpected plant responses occur was not explicitly
addressed. PMI 17.1, Conduct of Testing was revised to include a
statemen that "when necessary" guidelines for aborting the test should be
included. The evaluation of recommendation number 3 concluded that
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special tests are discussed during the routine shift briefing and no
further actions were warranted. In response to the training
recommendation, various changes were made to the operator training and
shift technical advisor training program.
The inspector's evaluation of the licensee's activities in this area
concluded that the level of management involvement in this area was
insufficient. Outside of the training area, only one activity was
initiated in response to this 50ER; that being a minor procedure revision.
No input was sought or provided by the Plant Operations Review Committee
(PORC), Nuclear Safety Review Board (NSRB), Plant Manager or Site
Director. Training upgrades were made only to the operators and STA
programs. Training for managers, supervisors and engineers was not
implemented. No review was initiated to evaluate past performances of
special test. This type of review would either tend to confirm the
licensee's conclusion that control is already adequate or it would
highlight problems with interfaces and coordination that may not be
evident by a simple review of program documents. Contrary to the
licensee's conclusion that PMI 17.1, Conduct of Testing procedure was
adequate, the inspector found that the procedure did not focus adequate
attention on special tests with true safety significance. This was due to
the same instruction attempting to cover both safety-related special tests
and non safety-related tests. Since the majority of special tests are 4
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either non safety-related or of very little safety significance, a
complacency may tend to develop such that insufficient attention is
focussed on the truly significant test. Some managers- have pointed out i
that most special tests are written simply because no procedure previously
existed. This can be exemplified by the recent special test written to
allow dumping non-contaminated chemicals from a tank car to the river.
This was a non-routine evolution requiring special controls but no test or
experiment was associated with this activity. Licensee representatives
agree that these types of procedures should not be controlled as a special !
test. The licensee's activities in resoonse to the Chernobyl accident
will continue to be tracked under an Inspector Followup Item
(259,260,296/87-37-02).
13. Design Baseline Verification Program (DBVP)
During the month the status of the DBVP was reviewed. Major elements of
the program that are complete include the safe shutdown analysis, system
requirements calculations, design criteria, and system test requirements.
The remaining elements are design calculation review and system evaluation
reports. The evaluation reports will consist of comparing the plant i
configuration to the design basis to determine if the plant configuration
is as desired or acceptable.
The work on the DBVP is shifting to a managed task assignment using a
contractor. The present DBVP manager relocated to the corporate office
and is no longer on site. Also, 14 out of 18 contractor personnel l
assigned to the Engineering Assurance organization for Browns Ferry DBVP l
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oversight are located at the corporate office. This approach appears to
be . a shif t away from the Project Engineer team approach located at the
plant site. This is discussed in Volume I of the Nuclear Performance
Plan,Section IV.E.2. Stated in this section is that the project team is
. principally located at the plant site. Work and resources are being
shifted from the central staff to the . project teams as necessary to
implement the project engineering concept. These statements are in
conflict with the present arrangement.
14. Inadequate Core Cooling Instrumentation (Generic Letter 84-23)
The inspector reviewed a letter dated September 25, 1987, from R, Gridley
to the NRC Document Control Desk providing a re-evaluation of the comple-
tion dates for the reactor water level instrument reference leg modifi-
cation required by Generic Letter 84-23., The NRC requested the re-
evaluation because of a concern that a relationship may exist between
Generic Letter 84-23 and the water level mismatch events which occurred on
Unit 3 in 1985. TVA concluded that the two were not related and provided
justification for operation of Unit 2 for one cycle before performing the
modification.
Missing from the re-evaluation was a discussion of a General Electric
report contracted by TVA concerning the water level mismatch events. The
inspector felt this information was needed for a NRC reviewer to make an
adequate evaluation of the situation. General Electric agreed with the
plant staff conclusion of the possible cause of the November 5, 1985 water
level event. But, the probable cause of the February 13, 1985 event was
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determined by General Electric to be the rigid instrument piping system.
A' site inspection revealed the existence of a piping configuration which
does not permit movement with the reactor pressure vessel thermal growth.
It was expected that the rigid system can perform its intended function
for one additional fuel cycle. This was based on the acceptable results
of a non-destructive examination performed on the instrument piping.
General Electric design documents require the loading to be " substantially
none" upon the reactor pressure vessel two inch instrument nozzles. The
conclusions, causes, adjustments and recommendations were presented to the
Plant Operations Review Committee on July 18, 1986. However, the discus-
sion of the February 1985 event in the letter makes no mention of these.
A meeting was held with members of the plant licensing and technical
services staff to discuss this omission in the letter. From this meeting
it was not clear whether the licensee agrees with the General Electric
report or how the findings will be resolved. Disposition of these items
will be tracked under an Inspector Followup Item (259,260,296/87-37-03).
1
15. Containment Coatings )
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On October 9,1987, a meeting to discuss containment coatings was held l
with applicable licensee representatives. This issue is identified in
section 14.3 of the Browns Ferry Nuclear Performance Plan. An FSAR
commitment to use coatings qualified to the requirements of the American
National Standards Institute has been followed for the inside containment
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surface and the biological shield, Section 5.2.3.2 of the FSAR states
that the exposed portions of the interior of the drywell have been
sandblasted and provided with a protective coating consisting of an
inorganic zinc primer (Amercoat Dimetcoat 6) with an epoxy topcoat
(Amercoat 66). Section 5.2.3.2 states that as part of the torus
modification program, the interior of the suppression chamber has been
recoated for corrosion protection with Valspar Hi-Baild Epoxy 78.00. This
coating has passed test criteria for a design basis accident as outlined
in ANSI N101.2-1972. This coating system is expected to withstand
temperatures and pressures of the steam environment during a design basis
loss-of-coolant accident.
However, some purchased' components have been installed inside primary
containment with unqualified coatings. A walkdown of the drywell and
torus for Unit 2 is being conducted to determine the areas and amount of
unqualified coatings. An estimated eight square feet of unqualified
coatings was identified on the torus vacuum breaker actuators. Examples
in the drywell are snubbers and electrical connection boxes. A
calculation of the allowable fraction of unqualified coatings is planned.
This will be based on plugging the emergency core cooling system suction
strainers. Any required corrective action has been identified as a
restart item.
16. Facility Modifications (37701)
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a. Analog Trip Unit l
On October 6, 1987, the inspector witnessed portions of Post
Modification Test Instruction PMT-116, Rosemount Trip Calibration
System. The test was intended to verify the operability of the
recently installed Analog Transmitter and Trip Units (ATTU). The
ATTU replaced the troublesome electro-mechanical switches in the
Reactor Protection System (RPS) and various Emergency Core Cooling
Systems (ECCS) control functions. During the performance of step
5.4.12.18 of the PMT, the inspector questioned the adequacy of a
piece of measuring and test equipment (M&TE). A calibration check of
drywell pressure transmitter 2-PT-64-56c was being performed by
isolating the transmitter and providing a pressure supply locally at
the transmitter. The calibration pressure was being recorded and
controlled by a 0-4.5 psig range pressure gauge, TVA serial number
E82214. With no pressure supplied to the gauge, it indicated about
two divisions upscale. When the instrument mechanic was quizzed as
to why the gauge didn't indicate zero pressure, he first ensured that
the gauge was vented and then turned the zero adjust screw unt4' the
meter read zero. This was confirmed by the inspector on the
following day to be an improper adjustment of M&TE. The M&TE
coordinator and the Unit 2 I&C Section Supervisor took immediate
action to have the gauge retrieved and tested. A memorandum was
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initiated to inform all instrument mechanics on the proper use of
this gauge. The gauge should have first been exercised by
pressurizing it up to it's full scale, held for several minutes and
then vented. If the zero had been different than that specified on
it's latest calibration data card, it should have been returned to
the calibration lab for testing. The only authorized adjustment of
the zero adjust screw is made during a multi point calibration
procedure traceable to the National Bureau of Standards. The
erroneous adjustment of the pressure gauge was a violation of 10 CFR
50, Appendix B, Criterion XII, Control of Measuring and Test
Equipment (260/87-37-04).
b. Appendix R Modifications
The inspector reviewed the administrative controls applicable to core
drilling through secondary containment walls. Engineering Change
Notice (ECN) P0953 authorized the drilling of a 6-inch diameter hele
through the Unit I reactor building west wall for the purpose of
pulling cable for fire protection (Appendix R) modifications. The
work was being performed under Work Plan and Inspection Record (WP &
IR) number 0001-87. Standard Practice 14.4, Drilling, Chipping, or
Altering Concrete or Masonry and Excavation, contains the
prerequisites for the core drilling operation. The procedure
requires that the Engineering Supervisor review the proposed drilling
permit and ensure that secondary containment is maintained.
Attachment I to the standard practice contains the anaiysis
methodology by which this is done. The method involves
mathematically modeling the hole through secondary containment as a
square-edged orifice in a pipe of infinite diameter and calculating
the air flow past the orifice with a 0.25-inch W.C. pressure drop.
The additional in-leakage calculated is then compared to the most
recent secondary containment performance surveillance to determine if
the available margin to the technical specification limit would be
exceeded. The sample calculations contained in Standard Practice
14.4 were deficient. Variables used in the orifice equation were not
defined; some variables (those with greek characters not normally
found on typewriters) were not shown in the equation; a cursory
dimensional analysis of the flow rate equation yielded a unitiess
number as opposed to the ft/ min value supposedly obtained; and no
basis was given for the value chosen for the flow coefficient.
Although the flow coefficient may range from 0.5 to 0.9 and the flow
rate is directly proportional to the flow coefficient, a value of 0.5
was chosen and stated to be conservative. Since this would yield
lower flow rates and allow larger hole sizes as compared to a value
of 0.9, this is a non-conservative assumption. Additionally, the
references listed for the analysis methodology were not auditable.
One of the references was a " Crane Manual". Neither the correct
title nor the appropriate edition was cited. Another reference was
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made to previous calculations by licensee personnel. This refer,ence
could not be traced to a date, document number or any other method of I
identification. Portions of this document were reviewed and found to l
contain faulty assumptions for the Reynolds Number which would have
been appropriate only for laminar flow regions. Actual calculations
of the Reynolds Number results in f. low rates well into the. turbulent
regions.
In addition to these problems with the sample calculations, the
inspector questioned the applicability of modeling the hole as an
orifice in an infinite diameter pipe. The inspector performed an
independent calculation by modeling the hole as a 4.5 ft. length of
6-inch diameter pipe with a 0.25-inch W.C. pressure drop This
calculation resulted in in-leakage values consider. ably higher than
that obtained by the orifice method. The " Crane Manual" referenced
in Standard Practice 14.4 provides guidance on when an orifice model
can be used to approximate flow through short length tubes. . The
criteria that should be satisfied is L/D (length over diameter)
should be less than 2.5. This was not the case for the core drilling
operation currently in progress since L/D was 9. (Reference Flow of
Fluids Through Valves, Fittings, and Pipe, Crane Technical Paper
No. 410, 1980)
Several discussions were held with licensee representatives who
conceded that the standard practice was deficient and sould be
revised. At the close of this reporting period, no consensus had
been reached on what value of in-leakage should be assigned for the
core drilling in progress. The various deficiencies associated with
Standard Practice 14.4 are a violation of 10 CFR 50, Appendix B,
Criterion V, which requires adequate written procedures for
activities affecting quality (259,260,296/87-37-05).
17. Environmer.tal Qualification Testing
The inspectors accompanied two licensee representatives during their
observation of environmental qualification testing performed at a local
test facility on October 1, 1987. Qualification Plan No. 17460-54,
Revision A, Interim Revision 3, Step 3.6.1, Accident Simulation was
performed. This procedure was a high energy line break (HELB) simulation
of three molded case circuit breakers used in various Motor Control
Centers (MCC) at Browns Ferry. The breakers were installed in the
accident simulation chamber in a manner similar to the plant configuration
and subjected to a temperature, humidity and pressure environment which
peaked at 227 degrees F, 0.97 psig and 100 percent relative humidity.
During the transient the breakers were operated under a resistive loading
network intended to simulate a motor load. Load resistors were switched
such that an inrush current of about 150 percent was present for one
second followed by a steady state operating current of 30 percent breaker
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ratinge The inspectors noteJ that adequate margin. had been applied to the
expected service environment to obtain the test profiles and that adequate
. acceptance criteria was1specified prior to the test.' Measuring and. Test
Equipment was properly controlled and maintained calibrated within the
required frequency. The-frequency.of licensee surveillance and audit of
the' test-facility was adequate and effective. The qualification test ~ plan
generally conformed to the requirement of .10 CFR 50.49.- .With the
exception of minor accident chamber temperature control problems c:used by
at loose electrical connector, the test equipment and qualification test
specimens performed flawlessly. No violations or. deviations were
identified.
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18. Document Contro'1 Program (39702)
An inspection was conducted to verify the accuracy, legibility, control, ;
and distribution of plant drawings. . Illegible copies of recently issued - !
Configuration Control Drawings (CCD) were found in the control room which. l
prompted this special inspection.
The licensee has threa basic classifications of drawings as defined in
Part III,!Section 1.1 of the Nuclear Quality Assurance Manual (NQAM).
Critical Drawings--Drawings depicting system features which are- used by
the plant Technical Support Center (TSC) and the Chattanooga Emergency
Control Cent'er (CECC) to determine system operation and function. The
drawings are _ for- use in a radiological emergency to analyze problems and
make recommendations for the mitigation of the consequences of an
accident. The following types of drawings may be included: schematics,
electrical single lines, control and logic, flow diagrams, structural, and
equipment arrangement drawings.
Primary Drawings--All drawings which are necessary to startup, operate, ,
and. shutdown the plant. Drawings for both emergency and normal shutdown
are to be included. Primary drawings are located in the unit control room
and are referred to during daily operation of the plant.
Secondary Drawings--All other drawings that are not primary drawings.
The list of critical drawings is included as Attachment A to Standard
Practice 2.5, Drawing Control. These drawings consist of the following:
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47W800 Series - Flow Diagrams for Nuclear Steam Supply Systems (NSSS)
and Balance of Plant Systems (BOP)
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47W610 Series - Mechanical Control Diagrams for NSSS and B0P Systems
(note that although TVA does not issue Piping and Instrumentation
Diagrams, (P&ID) a composite of the 47W800 series and its
corresponding 47W610 series would essentially be identical to a
P&ID).
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47W611 Series - Mechanical Logic Diagrams (these are logic trees for
control functions such as valve controls which depict in a ' tree
format the nece:sary and sufficient condition for an action or output
to occur).
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15N500 Series - Normal and Standby Auxiliary Power. One-line
electrical drawings.
Two problems were identified with the critical drawings. First, no
structural or equipment arrangement drawings are on the critical drawing
list. Floor plans and equipment layout drawings would be useful to
individuals in Chattanooga directing a response to a radiological
emergency at the site. The 47W200. Series contains the equipment
plan / layout diagrams. Secondly, the 47W611 Series of logic diagrams are
stamped "For Information Only" and "Not Controlled Copy". This is
reportedly due to the fact that these drawings have not been changed and
updated.with plant changes and modifications. It is inconsistent to have
a drawing important enough to be on the critical drawing list but yet be
uncontrolled. The inspector was aware of a similar condition that existed
at the Sequoyah Nuclear Plant (SQN) earlier this year. SQN l
representatives decided to remove the drawings from the control room and
the critical drawing list until they were properly evaluated and updated.
Browns Ferry personnel were informed of this and were requested to respond
with their plans.
The primary drawing list is maintained by the Plant Manager. As discussed
in previous inspection reports, plant drawings are being converted to
Configuration' Control Drawings (CCD). These CCDs will be the single
drawing of record and replace the old "as-designed" and "as-constructed"
versions of the drawing. Most CCDs are being issued as drawings from the
licensee's Computer Aided Drafting (CAD) System. This has virtually
eliminated the legibility problems which were associated with the old
manual mylar originals. The CCD commitment for restart however, is
extremely small, covering only those flow diagrams, mechanical control
diagrams and certain electrical drawings associated with the Safe Shutdown
Systems as defined in the Design Baseline and Verification Program (DBVP).
The licensee undertook a major CAD restoration activity in the last year
and has all primary drawings now CAD restored; however, issuance of these
drawings is not expected in the near term due to competing priorities for
restart manpower. The remaining activities involve review, approval and
issuance of these drawings. Manpower will be mainly dedicated to drawing
changes made as a result of plant modifications. Work plans which have
been designated as restart prerequisites currently number 899. These
workplans are estimated to contain at least 8376 drawings for update. 235
workplans were completed in the last 18 months. Miscellaneous
deficiencies detected during the drawing inspection are listed below:
a. Control of field drawings has recently changed; however, no explicit
direction exists for field drawings issued under previous controls.
Part III, Section 1.1 of the NQAM now requires that field drawings be
approved for use by DNE. Previously these drawings were approved by
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the plant manager. It was also noted that the lower tier
implementing documents have yet to be changed to require DNE approval
of field drawings. ' A field drawing is a sketch or drawing made 'to
clarify maintenance, modification testing or operation where no
existing drawing previously existed. Field Drawing No. BF-85-P-133
of the Diesel Generator Starting Air System was reviewed and found to
be ~ redundant to the recently issued CCD for this system. No
cross-reference or other mechanism was found to be in existence which
would have cancelled this field drawing.
b. Some legibility problems have been traced to problems with
reproduction equipment. A recurring problem with fuser control has
yet to be fully corrected. Since the ink is not fully fused to the
paper, a good copy may be generated but portions are easily rubbed
off with use. The inspector noted that portions of 3-47W1610-85-5.
CRD Hydraulic System in the control room had illegible portions and
that ink could be rubbed of f by hand contact.
c. The inspector witnessed changes being made to 2-47E610-75-1, Core
Spray System Mechanical Control drawing in response to Drawing
Discrepancy Package -075-034. Changes were being improperly depicted
for valve position indicators and position switches. Licensee
representatives concurred with this finding and instituted immediate
corrective action.
d. Drawing 15W500-2,- Key Diagram of Normal Auxiliary Power is a-
safeguards drawing and is not readily available to operators in the
control room. Only one component on the drawing relates to
safeguards information. Operators have expressed a desire to have
this drawing in the control rooms since it provides an excellent
source of normal and alternate power sources for numerous components
throughout the plant. The inspector learned that future
implementation of design revision 20 will declassify this drawing;
however, no work plan or engineering change notice could be located
that would implement this change. The inspector discussed this
matter with licensee representatives in an effort to facilitate this
change.
e. Drawing changes forced by modification workplans are required to be
formally issued within 15 days of receipt by draf ting services for
primary and critical drawings and within 90 days for secondary
drawings. It is unclear when this requirement (contained in Project
Instruction PI-87-48) will be routinely satisfied. The current
backlog makes this virtually impossible. The goal is to have primary
drawings issued by restart with the host of secondary drawings to
' follow within 90 days.
f. Although CCDs are to be the single drawing of record, several
versions of CCDs are allowed which may confuse drawing users. The
first version of an issued CCD simply replaces the previous
"as-constructed" drawing. The "as-designed" drawing remains i
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controlled and is used in certain design and modification related
acti vi ti e s.. Following comparison of the as-designed drawing with the
CCD and various licensing documects, the CCD becomes " Validated" and
replaces the as-designed drawing. An interim stage is also
authorized in .which only portions of a drawing have been field
verified to reflect the as-built configuration. In this case the
field verification boundaries will be clearly annotated. Following i
design evaluation of the partially field verified drawing the CCD
becomes " Design Verified". Prior to this design verification
activity, the as-designed drawing remains valid for partially field
verified CCDs. The architects of this phraseology intended for
people who use drawings to be able to ascertain the proper use of the
drawing by noting whether a sticker entitled " Design Verified" or
" Validated" appears on the drawing. Informal discussions with plant
personnel did not satisfy the inspector that these fine points are
clearly understood.
19. Q-List
The inspector reviewed the Q-List status and noted that Nuclear Quality
Assurance Manual procedure ID-QAP-2.7 describes the development, control
and application of a "Q List" for documenting and classifying structures,
systems, and components that fall within the scope of TVA's design,
construction, and operation quality assurance programs for nuclear plants.
Section 4 of the procedure assigns the responsibility of developing,
issuing and maintaining the Q-Lists to the Division of Nuclear Engineering
(DNE). Section 5.5 " IMPLEMENTATION" and subsection 5.5.3 of the procedure
states: "The BFN (Browns Ferry Nuclear) and SQN (Sequoyah Nuclear) Q
Lists will be put into effect.by the NSD (Nuclear Site Director) at a date
mutually agreed to by DNE and NSD.
On October 9, 1987, a status briefing was presented by licensee
representatives to the inspector. This briefing indicated that a DNE
procedure, Browns Ferry Nuclear Plant BFNP-PI-87-52, Development and
Control on Browns Ferry Nuclear Plant, was in draft form and would be out
shortly. The briefing also indicated that this procedure would be
addressing Unit 2 specifically and that it would provide the procedures
and requirements for development of the official BFN Unit 2 Q-List. As of
yet a date mutually agreeable between DNE and NSD for placing the Unit 2
Q-List in effect has not been determined.
20. Companion Drawing Discrepancy /CAQR Review
The following area that deals with concerns related to the structural
adequacy of various safety-related systems was partially reviewed. The
issues addressed encompass responses to CAQRs in the area of pipe
supports, conduit supports, and QA Audit of CAQRs.
! ,
CAQR No. BFP 870207 R/0 dated May 4, 1987, involved adequacy of closed l
CAQRs. CAQR No. BFP 870209 R/0 dated May 4, 1987, involved various
details shown on Drawing 48W1241-1 R/9. Documents reviewed in conjunction
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with the above CAQRs were: Problem identification report'(PIR) No. BFN
CEB 8628 dated July 28, 1986; General Construction Spec. No. G-32; ECN
P-0268; Dwgs 47B920-501 R/0, 47B458-404, R/4, 478920-91 R/6; and one
calculation for ECN P-0361 Rev. 5, dated 7/30/87 for the torus attached
piping supports. No discrepancies were noted in the above two CAQRs.
CAQR No. BFP 870205 was reviewed for adequacy of corrective actions. This l
, CAQR dealt with conflicting requirements between companion drawings and i
referenced general notes. The drawings involved in this matter ' .re
48B800-1, general notes, and 48B810-2 series which are companion drawings
for installation of conduit supports. In review of the CAQR and drawings,
it was noted that the licensee's response to the CAQR stated the CAQR was
not significant. Review by the inspector identified one area of concern;
there were conflicting tolerances on the two drawings where Engineering
Design might not be aware of the craft installing conduit supports to two
tolerances.
During the review of the CAQR, the inspector reviewed the following
documents:
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OES 7.01, Drafting Standards, Sections 5.5.1 & 5.52,- dated April 8,
1986
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General Construction Spec., G-40, Installing Electrical Conduit
Systems and Conduit Boxes. Rev/9 dated 1/15/86, sections 1.1,
3.2.2.2 and 4.0
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Design Criteria No. BFN-50-C-7104, Design of Supports R/0 dated
July 1,1987, attachmend D sections 2.5, 6.1, Appendix A
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Construction Spec No. G-3, Section E.2 (this has been superceded by
General Construction Spec. G-40).
In reviewing the above documents it was clear that there is a conflict
between the drawings and the various documents, to such a degree it could
be confusing as to which document takes precedent and could affect the
craft in the proper installation of conduit supports. The inspector held
discussions with appropriate licensee personnel and they agreed the CAQR
was improperly dispositioned and that inadequate corrective action was
taken. During the discussion the licensee committed to initiating a new
CAQR to make clear the hierachy of drawings so that it would be clear to
all concerned which takes a higher priority during construction and to
ensure design drawings and companion drawings do not conflict. The
licensee initiated CAQR BFN 870451 dated 10/29/87. This item is being
identified as Inspector Followup Item (50-259,260,296/87-37-06).
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