ML20244C780
| ML20244C780 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 04/14/1989 |
| From: | Livermore H, Runyan M Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20244C772 | List: |
| References | |
| 50-445-89-19, 50-446-89-19, NUDOCS 8904200412 | |
| Download: ML20244C780 (17) | |
See also: IR 05000445/1989019
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APPENDIX B
U.
S. NUCLEAR REGULATORY COMMISSION
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OFFICE OF NUCLEAR REACTOR REGULATION
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NRC' Inspection Report:
50-445/89-19
Permits: CPPR-126
50-446/89-19
CPPR-127
Dockets: 50-445
Category: A2
50-446
Construction Termit
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Expiration Dates:
Unit 1: August 1, 1991
Unit 2: August 1, 1992
Applicant: .TU Electric
Skyway Tower
400 North Olive Street-
Lock Box 81
Dallas, Texas
75201
Facility Name:
Comanche Peak Steam Electric Station (CPSES),
Units 1 & 2
Inspection At:
Comanche Peak Site, Glen Rose, Texas
Inspection Conducted: March 8 through April 4, 1989
Inspector:
NVctf-E -
I-/ /- 83
g M. F. Runyan, Resident Inspector,
Date
Civil Structural
(paragraphs 2, 4, 5, 6, and 7)
Consultants: W. Richins, Parameter (paragraphs 2, 3, 5, and 6)
J. Dale, EG&G (paragraphs 2, 5, and 6)
J. Birmingham, RTS (paragraph 3)
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Reviewed by:
Auf.A.udtr2P ,
V- /Y'- 8S
H. H. Livermore, Lead Senior Inspector
Date
8904200412 890414
ADOCK 05000445
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Inspection Summary:
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Inspection Conducted:
March 8 through-April 4,
1989 (Report
50-445/89-19; 50-446/89-19)
Areas' Inspected:: Unannounced, resident safety inspection of the
applicant's actionsfon previous. inspection findings, follow-up on
. violations / deviations, action on 10 CFR 50.55 (e) deficiencies
identified by.the applicant, Post-Construction Hardware Validation
Program (PCHVP),-and. general plant areas (tours).
Results: . Within the' areas inspected,.a weakness was identified
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regarding QC measurement' techniques (paragraph ~2.g).
One violation.
Je.
.was identified regarding methods used by QC to measure valve stem;
orientation andLincore instrument tubing dimensions within the
PCHVP (paragraph.2.g)..
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DETAILS
1.'
Persons Contacted
- R. W. Ackley, Jr., Director, CECO
- J. L. Barker, Manager, Engineering Assurance, TU Electric
- D.
P. Barry, Senior Manager, Engineering, Stone and. Webster-
Engineering Corporation (SWEC)
- J. W. Beck, Vice President, Nuclear Engineering, TU Electric
- M. R. Blevins, Manager, Technical Support, TU Electric
- H. D. Bruner, Senior Vice President, TU Electric
- W.
J.
Cahill, Executive Vice President, Nuclear, TU Electric
M. R. Clem, CAP, Structural, SWEC
- J. T. Conly, APE-Licensing, SWEC
- C.'G.
Creamer, Instrumentation & Control (I&C) Engineering
Manager, TU Electric
H. M. Crockett, Engineering, TU Electric
- G.
G. Davis, Nuclear Operations Inspection Report Item
Coordinator, TU Electric
- J.
C.
Finneran, Jr., Manager, Civil Engineering,
TU Electric
- C.
A. Fonseca, Deputy Director, CECO
- W.
G. Guldemond, Manager of Site Licensing, TU Electric
N. D. Hammett, Engineering Assurance, Brown and Root (B&R)
- T.
L. Heatherly, Licensing Compliance Engineer,
TU Electric
- J.
C. Hicks, Licensing Compliance Manager, TU Electric
- C.
B. Hogg, Engineering Manager, TU Electric
- A.
Husain, Director, Reactor Engineering, TU Electric
- S.
D. Karpyak, Nuclear Engineering, TU Electric
- J.
J. Kelley, Manager, Plant Operations, TU Electric
- O. W. Lowe, Director of Engineering, TU Electric
- D.
M. McAfee, Manager, Quality Assurance (QA), TU Electric
- S.
G. McBee, NRC Interface, TU Electric
- J. W. Muffett, Manager of Engineering, TU Electric
D. Noss, Licensing, Daniel
- E.
F. Ottney, Program Manager, CASE
- J.
D. Redding, Executive Assistant, TU Electric
- D.
M. Reynerson, Director of Construction, TU Electric
- M.
J. Riggs, Plant Evaluation Manager, Operations, TU Electric
- A.
H. Saunders, EA Evaluations Manager, TU Electric
- A.
B. Scott, Vice President, Nuclear Operations, TU Electric
- C.
E. Scott, Manager, Startup, TU Electric
- J.
C.
Smith, Plant Operations Staff, TU Electric
- M.
R. Steelman, Licensing, TU Electric
- P.
B.
Stevens, Manager, Electrical Engineering, TU Electric
- J.
F. Streeter, Director, QA, TU Electric
- C.
L. Terry, Unit 1 Project Manager, TU Electric
- T.
G.
Tyler, Director of Projects, TU Electric
- R.
D. Walker, Manager of Nuclear Licensing, TU Electric
- R.
G. Withrow, EA Systems Manager, TU Electric
- D.
R. Woodlan, Docket Licensing Manager, TU Electric
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- J.
E. Wren, Assistant Director QA for Administration,
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TU Electric
The NRC inspectors also interviewed other applicant employees
~during this inspection period.
- Denotes personnel present at the April 4, 1989, exit
meeting.
2.
Applicant Action on Previous Inspection Findings (92701)~
a.
(Closed) Open Item (445/8514-O-18):
'During'an NRC
inspection, the CPRT inspector identified bent flange
bolts, oversized flange bolt holes, and missing corner
These items were reported on
out-of-scope No. 199 and Deficiency Report (DR)
I-M-DUPL-021-DR-1 and subsequently on Nonconformance
Reports (NCRs) CM-87-5820-X and M-85-101718-SX.
Due to
the implementation of TU Electric's Hardware Validation
Program and Corrective Action Request (CAR)-111, this
duct piece was scrapped and a new duct piece installed.
The new duct piece was as-built, detailed, and inspected
(See Inspection Report B-1-756-196-01) by EBASCO and
found acceptable.
The NRC inspector reviewed the above listed out-of-scope,
DR, NCR, and inspection reports.
The NRC inspector also
reinspected the duct piece and found it to be acceptable.
No further violations or deviations were found.
This
open item is closed.
b.
(closed) Open Item (445/8514-O-28):
During an NRC
inspection of duct support package I-S-HVDS-023, the CPRT
inspector identified possible deviations dealing with
incorrect member size and undersized welds.
These items
were addressed on DRs I-S-HVDS-023-DR-1 thru DR-11 and
subsequently on NCR M86-100321SX.
The items were then
corrected per CAR-111 and verified through the
implementation of TU Electric Procedure CPE-FVM-CS-029.
The NRC inspector reviewed NCR M-86-100321SX and
Construction Operation Traveler DH-1-1701-830-1N-1R and
verified by field inspection that all rework had been
correctly implemented and the open item resolved.
No
further deviations or violations were identified.
This
open item is closed.
c.
(Closed) Open Item (445/8514-O-31):
This item addressed
potential deviations regarding incorrect member size and
undersized welds for structural steel member
AFCO-MK-C182-7-RB.
This member is part of a platform
structure located in the Unit 1 Reactor
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building.
These conditions were identified by CPRT
duringian inspection of Verification Package I-S-STEL-88
witnessed by the NRC inspector, documented on'CPRT DRs
I-S-STEL-88-DR1 and I-S-STEL-88-DR2, and incorporated
into NCRs M86-100295SX, Revision 3, and M89-01030,
Revision 0.
These NCRs were subsequently dispositioned
"use-as-is" based on review.of calculation
16345-CS(B)-215, Revision 0, and comparison of member and
weld stresses.
The NRC inspector reviewed the
calculation and the above NCRs and concurs with the
use-as-is dispositions.
This item is closed.
d.
(Closed) Open Item (445/8615-0-11):
Tais item addressed'
an heating, ventilation, .and air-conditioning (HVAC)
chiller foundation anchor bolt,which was not. bearing a
load due to a 1/4-inch gap between the nut and the mating
surface.
A CPRT inspector identified the nonconforming
condition during the walkdown of Inspection Package
1-M-MEIN-035.
This deficiency was transferred to NCR
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PM87-00667, Revision 1, and was dispositioned " rework"
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per Specification 2323-MS-101.
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The NRC inspector examined the subject bolt in the field
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and concurred that the nonconforming condition had been
corrected.
This open item is closed.
e.
(closed) Open Item (445/8828-0-02):
This item addressed
the apparent discrepancy between the Technical
Specification (TS) limit of 150 degrees F at the reactor
cavity exhaust and the Final Safety Analysis Report
(FSAR) (9.4A.1.3) limit of 135 degrees F in the neutron
detector well, which is located in the reactor cavity.
The applicant had revised the TS to the higher limit
after demonstrating that by limiting the exhaust
temperature to 150 degrees F, the neutron detector
temperature would not exceed 135 degrees F.
The open
item was based on an understanding that an explanation of
the above would be included in future revisions of both
the FSAR and the TS.
The applicant has since decided
that the revisions are unnecessary.
The NRC inspector
concurs with this position, inasmuch as no real conflict
exists between tho TS and the FSAR.
This open item is
closed.
f.
(Closed) Open Item (445/8905-0-01):
This item addressed
methods used to evaluate the locations of Richmond
inserts located behind baseplates or which were otherwise
obscured.. The concern pertains to the overall effort to
evaluate the r.dequacy of concrete attachments per Field
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Verification Method (FVM)-075.
The item was opened when
the NRC questioned the degree of conservatism employed in
locating the probable location of inaccessible Richmond
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inserts.
-As a result of further review of FVM-075
calculations, the NRC inspector has concluded that the-
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method used to estimate Richmond insert locations was
sufficiently accurate to provide confidence in the
resulting calculated bolt capacities.
Therefore, the
original issue and cause for concern no longer exists.
This open item is. closed.
g.
(Open) Unresolved Item (445/8911-U-01):
This unresolved
item addressed the concern that the method by which QC
inspectors had measured valve stem angles as depicted on-
isometric drawings did not result in the accuracy
required by Procedure AQP-11.5.
The as-built valve stem
orientation is required to be within i 5 degrees of the
angle shown on the isometric; and, thus, the measurement
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technique must attain at least this accuracy.
Valve stem
orientation was checked pursuant to the PCHVP inspection
of piping and in-line components within the Mechanical
Project Status Report.
The example specific to the ~
unresolved item was Valve ICC-692 shown' rotated
13 degrees from vertical on Isometric Drawing
BRP-CC-1-RB-003.
The as-built orientation of this valve
at'the time of the NRC inspection was about 22 degrees
from vertical.
Further information revealed that the
stem orientation of Valve ICC-692 had changed between the
time of the PCHVP inspection and the NRC inspector's
review due to construction activity.
A seal weld was
applied by Traveler CC-1-RB-003. . The traveler did not
specify a verification of the valve stem angle after the
work was accomplished; therefore, the nonconforming
condition escaped QC notice.
This discrepancy was
documented on NCR 89-2564.
The above information
suggests that the subject valve stem orientation
discrepancy was not the result of an inaccurate QC
measurement.
The above scenario suggests a deficiency in
the area of post-modification inspection and testing, and
the unresolved item will remain open pending further
review of this matter.
As part of the investigation of the original issue, a
lead QC inspector was selected to demonstrate his, and
presumably the standard, method of measuring valve stem
angles during the PCHVP walkdowns.. Afterwards, an
engineering survey crew measured two of the same angles.
The table below shows the results along with the angle as
depicted on the isometric.
Angles given are corrected to
the same orientation as measured by the QC inspector.
Valve
OC Inspector
Survey Crew
Isometric
XSF-131
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51.3
51
SFX-085
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N/A
14.3
ICS8382B
30
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The results of this exercise clearly demonstrated that
the method used by the'QC inspector was not accurate to
within the required tolerance of i 5 degrees.
The QC
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inspector used a protractor with a bubble, centered the
bubble, and then tilted the protractor over to better
align the leading arm to the centerline of the valve
stem. HBy tilting the protractor, he was measuring an
angle in a skewed plane and not the angle' depicted on the
isometric which is in a vertical plane.
Also, the angle
was measured from the top rather than the centerline of
the pipe.
The method was, therefore, not only
inaccurate, but also invalid.
The NRC inspector observed
two additional QC inspectors demonstrate valve stem angle
measurements, but their techniques also appeared to be
far less accurate than i 5 degrees.
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A major hardware issue probably does not exist because
all safety-related valves are being' measured by the more
accurate survey method in response to NRC Bulletin 79-14,
which is. essentially a verification of the as-built
condition of the plant.
However, any subsequent
modification or maintenance which changes valve stem
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orientation could be affected by the invalid measurement
techniques.
The failure to train QC inspectors to
properly measure. valve' stem angles is considered a
violation (445/8919-V-01).
A second example to this
violation involving the measurement of incore instrument
tubing is discussed in paragraph'5.c.
These examples are
considered a weakness in QC measurement methods involving
such programmatic issues as techniques, equipment, and
training.
h.
(Closed) Unresolved Item (445/8842-U-01):
This item
addressed concerns regarding the effectiveness of FVM
CPE-SWEC-FVM-CS-075 (FVM-075), part of the PCHVP.
FVM-075 was developed to identify concrete attachment
spacing violations based on requirements presented in
Specification 2323-SS-30, " Structural Embedments."
The
NRC inspector questioned the adequacy of FVM-075.
The applicant issued Project Technical Report (PTR)-03,
Revision 0, " Evaluation of Seismic Category I and II
Concrete Embedments" dated February 10, 1989.
PTR-03
summarizes the historical background of deficiencies
identified in Specification 2323-SS-30, and the changes
made to Specification 2323-SS-30 and Design Base Document
(DBD)-CS-015 to provide controls for in-procsss and
future concrete embedment installation activities.
In
addition, PTR-03 describes the implementation and results
of FVM-075 which addressed concrete embedments installed
prior to the above changes to Specification 2323-SS-30
and DBD-CS-015.
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FVM-075 evaluated Hilti bolts, Richmond inserts (used and
unused), Nelson studs for embedded plates and shapes,
grouted-in bolts, embedded bolts, and through-bolts, all
referred to collectively as concrete embedments.
The
- applicant segregated the concrete surfaces population
into two categories:
(1) walls and slabs consisting of
5265 surfaces, and (2) beams and columns consisting of
1659 surfaces.
A total of 53 surfaces consisting of
22 walls / slabs and.31 beam / columns were evaluated using
423 field as-built sketches prepared by field engineering
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personnel.
The sketches identified'9017 concrete
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embedments used for 3088 supports-(Category I,
II and
nonsafety'related).
All spacing' violations were
identified and evaluated in 53 calculations using current
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requirements andL49 NCRs were generated.
Stone and
Webster Engineering Corporation (SWEC).obtained actual
loads for each concrete embedment deviating from
specification requirements.- The 53 calculations were
prepared to document that the factor of safety for each
embedment meets or exceeds the requirements of
DBD-CS-015.
PTR-03 concluded that:
"The completed evaluations provide sufficient data to
assess the condition of the existing CPSES Seismic
Category I and II concrete embedments.
The evaluations
performed to date provide a 99% confidence
level .
. that the current design loads for at least
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99.8% of the seismic Category I and'II concrete
embedmonts in the unexamined balance of the plant will
provide the required factor of safety."
Based on the results of FVM-075 from evaluating the
53 surfaces sampled, FVM-075 activities will be closed by
the applicant.
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The NRC inspector previously performed field inspections
of completed as-built drawings for one beam, two columns,
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and two wall surface packages.
The NRC inspector also
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reviewed the calculations for each package which
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identified and evaluated concrete attachment spacing
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violations.
These inspections are documented in the
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following NRC inspection reports:
50-445/88-82, 50-446/88-78
50-445/89-05, 50-446/89-05
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50-445/89-11, 50-446/89-11
During this report period, the NRC inspector reviewed
Calculation 16345/6-CS(S)-462, Revision 0, " Qualification
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of IBM Personal Computer Program CES."
The computer
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program CES computes the ' pacing between concrete
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embedments and compares
.le.results with the minimum
spacing requirements of Specification 2323-SS-30 and
DBD-CS-015 (Design Basis Document) to determine if
spacing violations exist.. The calculation package
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contains a description of the' program,-user instructions,
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limitations, explanation of the program output, and.
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qualification of the program.
The.NRC inspector reviewed
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the calculation package and performed supplemental
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calculations to verify the results obtained by the
program CES. 'The NRC inspector concluded that the
progrta CES correctly identifies potential concrete
attachment spacing violations.
Based on the NRC inspection activity summarized above and
review of pTR-03, the NRC inspector concluded that the
FVM-75 program was developed and implemented in an
effective manner and adequately identifies concrete
attachment spacing violations.
This item is closed.
3.
Follow-up on Violations / Deviations (92702)
a.
(Closed) Violation (445/8603-V-04):
The disposition of
NCR M-80-00161 failed to address all aspects of the
reported nonconformance.
Specifically, the disposition
of NCR M-80-00161 addressed authorization to cut a single
layer of concrete reinforcing steel at a 2 1/2-inch
depth, but failed to address the cutting of a template
bar (i.e., a construction aid) and the notching of
another bar of reinforcing steel at a 7-inch depth.
The applicant admitted to the violation and performed the
following corrective actions.
NCR M-80-00161 was
transferred to NCR CM-87-4182 and additional
justification for the "use-as-is" disposition was
provided; however, NCR CM-87-4182 still did not address
all aspects of the nonconformance.
Subsequently, the
applicant issued NCR 89-02715 to clarify the disposition
of NCR CM-87-4182.
The NRC inspector has reviewed
NCR 89-02715 and applicable documentation.
The NRC
inspector concurs that the disposition of NCR 89-02715
addresses all aspects or the identified nonconformance.
Effective December 22, 1986, the applicant provided a
revised program for the reporting and control of
nonconformances and deficiencies.
The revised program
has been reviewed by the NRC and determined to require
proper justification of "use-as-is" dispositions.
Further, the applicant has instituted a program to review
NCRs dispositioned prior to the effective date of the
revised program.
The program has been reported to the
NRC as part of SDAR CP-86-48.
NRC inspection of that
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.SDAR CP-86-48.
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Since (1).the applicant;has taken' appropriate corrective-
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action:for'the identified violation, (2) the program for
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-nonconformance control has.been revised to preclude-
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recurrence, and.(3) a: program for thefreview'of previous-
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NCR~ dispositions has been established, this. item is
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closed.
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b.-
-(Closed) Violation (445/8882-V-01):
This violation
addressed a weld. size :for a connection on a monorail
structural framework. identified by PCHVP
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Package CS-090-SG1-834-94-S2/I3.
The weld size was
increased from.3/16-inchoto 1/4-inch by Design. Change
Authorization'(DCA)S48, Revision 5,
(issued September
31988).
The. installed weld is a 3/16-inch fillet'.
The
weld was determined by.the NRC inspector to.be undersized
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per DCA'548, Revision.5, and the criteria in
Specification 2323-SS-16B, Revision 2, " Structural
Steel / Miscellaneous' Steel."-
'The applicant stated in response to the violation
(letter 1TXX-89082, dated Februaryf24,11989) that'the
violation occurred due to personnel error.
The engineer
who prepared DCA 548, Revision 5, incorrectly used the
weld size shown on an as-built drawing for.the weld size-
on a typical section used in Revision 5 to DCA 548.
Both
the engineer and the designichange. reviewer made the
assumption that the as-built weld size applied to
numerous.similar connections shown in the DCA.
As
' corrective action, the applicant issued DCAL548,
Revision 6, to correct the weld sizes shown.
In
addition, . t.ne applicant reviewed all other DCAs
originated by the engineer who prepared DCA 548,
Revision 5 and a sample of 15 civil / structural DCAs.
No
other similar instances.were identified.
DCA trend
reports issued during the fourth quarter of 1988 did not
identify an: adverse trend for the civil / structural
discipline.'
The NRC inspector. reviewed the applicant's response,
DCA-548, Revision 6, and the above trend reports.
The
NRC inspector concluded that the applicant had taken
satisfactory measures to correct the original problem and
-provide assurance that a programmatic weakness does not
exist.
This violation is closed.
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4.
Action on 10 CFR.Part 50.55(e) Deficien6ies Identified by the
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Applicant (92700)
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a.
(Closed) Construction Deficiency (SDAR-CP-86-54):
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" Original Design of Control Room Ceiling.". By letter
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TKX-6007 dated October 3, 1986, the applicant informed
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the NRC that a deficiency involving the design and
installation of the control room ceiling was.a reportable
item.
The control room ceiling as originally designed
and installed did not comply with the provisions.of
FSAR 3.7B.2.8, in that evaluations had not been performed.
to assure that the failure of the ceiling would not
adversely affect safety-related components or operator
actions.in the control room.
As a result, a.new ceiling
was designed and installed in accordance with'the FSAR-
requirements.
Previous NRC inspection of this issue, which was
encompassed in the review of ISAP II.d, is documented in
NRC Inspection Reports 50-445/85-13, 50-446/85-09;
50-445/86-01, 50-446/86-01; 50-445/87-11, 50-446/87-09;
50-445/87-13, 50-446/87-10; 50-445/88-64, 50-446/88-60;
and 50-445/88-70, 50-446/88-66.
The NRC staff review is
summarized in NUREG-0797, Supplement No. 20 (SSER 20)
pp. 33-34.
The NRC staff concluded that this issue had
been adequately resolved; consequently, this construction
deficiency is closed.
b.
(Closed) Construction Deficiency (SDAR-CP-87-106):
" Component Design Specification Discrepancy."
By
letter TXX-7127 dated December 29, 1987, the applicant
informed the NRC that a deficiency involving the
configuration of foundation' bolting for safety-related
mechanical equipment was not a reportable item.
The
applicant determined that Specification 2323-MS-101,
Component ~ Design Specification, did not adequately define
equipment foundation bolting configurations which
subsequently resulted in inadequate bolting
installations.
In their investigation performed under
ISAP VII.c, Appendix 17, Mechanical Equipment
Installation, CPRT identified a large quantity of bolting
configuration discrepancies, but determined by
engineering calculations that none of them affected the
safety function of the equipment.
Corrective action for
this issue was performed under CAR S-53.
Specification 2323-MS-101 and associated site procedures
were upgraded and training was conducted for all
responsible QA and QC personnel.
All permanently
installed equipment was walked down by the applicant to
verify foundation bolting configurations and all
deficiencies were identified on NCRs.
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The NRC inspector reviewed file correspondence,' CAR S-53,
and ISAP VII.c, Appendix 17, Results Report, andi
concurred with the applicant's categorization and
disposition of this issue.
This construction deficiency
is closed.
5.
Post-Construction-Hardware Validation Program (PCEVP) (46053,
46055, 46071, 48051, 48053, 48055, 50073)
a.
Concrete Attachments (CPE-SWEC-FVM-CS-075)
The NRC inspector has completed inspection activities
regarding.the implementation and conclusions of FVM-075
(see the closure of unresolved item 445/8842-U-01 in this
y
inspection report).
The NRC inspector concurs with the
applicant's conclusion that FVM-075 demonstrates the
adequacy of concrete attachments used for Category I
and II supports.
b.
Piping and In-line Components (CPE-SWEC-FVM-PS-081 and
CPE-SWEC-FVM-EE/ME/IC/CS-089 and -090)
(1)
BRP-CH-X-FB-010, Fuel building, Chilled Water.
This
PCEVP inspection entailed approximately 20 feet of
predominantly 1 1/2-inch piping depicted on
isometric drawing No. BRP-CH-X-FB-010,
Revision CP-1.
The only problem noted by the QC
inspector was a missing handwheel on valve XCH-310.
This discrepancy was documented on NCR 88-07883 and
~
the handwheel was replaced.
The NRC inspector walked down all piping relevant to
the isometric drawing and verified that applicable
attributes were satisfactory.
The only variance
with the isometric identified by the NRC inspector
was some temporary flexible hosing installed per
DCA 75933, Revision 0.
This DCA was issued after
the date of the PCHVP inspection.
The NRC inspector
concurred with the results of this PCHVP package.
(2)
BRP-AF-1-SB-009, Safeguards building, Auxiliary
Feeddater.
This PCHVP inspection entailed
approximately 40 feet of 6-inch piping depicted on
I
isometric drawing BRP-AF-1-SB-009.
The QC inspector
documented that all applicable attributes were
satisfactory.
The NRC inspector walked dcwn all piping relevant to
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the isometric drawing and verified that all
j
applicable attributes were satisfactory.
The NRC
inspector concurred with the results of this PCHVP
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inspection.
However, DCA 71641 issued after the
PCHVP inspection, appeared to erroneously change a
dimension shown on isometric drawing
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BRP-AF-1-SB-009.
As depicted on p. 12 of 14 of
DCA 71641, the dimensional length of piping
segment 10 at elevation 792'
3" was changed from
o
2 - 0 1/8' to l'
- 10 1/8" in response'to .
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DR-P87-00355.
The NRC inspector verified the length
as 2'
0 1/8" and questioned the applicant to' explain
the reason for-the change.
The applicant stated
that the dimensional change was needed to provide
closure for a three-dimension computer aided drawing
(CAD) of the auxiliary feedwater system.
The
specified tolerance of 2 inches between the
isometric drawing and the as-built condition was
maintained and will in this and other similar cases
be verified by QC inspection.
The NRC inspector
determined that this explanation was satisfactory.
(3)
BRP-CS-1-AB-005, Auxiliary building,. Chemical and
Volume Control System'(CVCS) positive displacement
pump discharge.
This PCHVP inspection entailed
approximately 45 feet of predominantly 3-inch piping-
depicted on isometric drawing BRP-CS-1-AB-005,
Revision CP-1.r The only problem noted by the QC
inspector was a single dimensional discrepancy which
exceeded the allowed 2-inch tolerance by 3/4 inches.
This condition was. documented on NCR PM-87-00790 and
!
- dispositioned "use-as-is."
The NRC inspector walked down all piping relevant to
the isometric drawing and verified that the
dimensional discrepancy described above was the only
unsatisfactory attribute pursuant to this
inspection.
The NRC inspector reviewed
NCR PM-87-00790, agreed with the "use-as-is"
disposition, and concurred with the results of this
PCHVP package.
c.
Incore Instrument Tubing (CPE-SWEC-FVM-EE/ME/IC/ CS-090)
The scope of this PCHVP activity included dimensions and
installation routing of incore instrument tubing as
depicted on drawings FSI-0043 and 1207E02 (sheets 1, 2,
and 3).
The QC inspector documented numerous
unsatisfactory dimensions pertaining to the 58 incore
instrument tubes on NCR 89-1095, which to date has not
been dispositioned.
The specified tolerance for this
inspection is i 2 inches, as required by Procedure
AQP-11.5.
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The NRC inspector attempted to duplicate 10 examples of
the QC inspector's measurements pertaining to dimensions
"C" and "D"'as shown on drawing;1207E02, sheet 2 of 3.
These dimensions' define the linear distance either side
of weld union' coupling to the point where the. tube begins
to form a 12-foot arc.
The measurements.taken by the NRC
inspector differed from those 'ecorded by the QC
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inspector by more than two inches for the seven examples
listed below.
Tube No.
Dimension Checked
JZ2
.NRC
34
C
54 3/4"
4' 11 3/4"
24
C
5'
3"
5'
0"
32
C
7'
5 1/16' 7'
2"
38.
C
6' 8 1/2"
'6' 6 1/4"
40
D
l' 3 1/2"
0'
10"
49
D
O' 9 3/8"
0' 5 3/4"
51
C
3' 10 1/4" 3'
6 1/4"
Inasmuch as the NRC inspector's measurements were
consistently shorter than those recorded by the QC
inspector, clarification was requested regarding the'QC
measurement' technique.
The NRC inspector witnessed a demonstration of the method-
used to measure dimensions
"C" and "D" by the same QC
inspector who performed the PCHVP inspection of record.
The QC inspector applied a portion of a straightedge
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flush to the linear portion of the tube and then wedged a
0.01-inch feeler gauge to the point where it stopped.
This point was then marked and used as a measurement
point.
Mathematically, a 12-foot radius bend elevates
0.01 inches in 1.7 inches of horizontal distance.
Thus,
the point marked by the QC inspector was at least
1.7 inches from the actual tangent point.
Since the arc
appears to flatten near the point of tangent, the
horizontal error most probably exceeded 2.0 inches.
Additionally, other sources of error exist, such as not
placing the straightedge directly under the tube.
The
NRC inspector measured the tangent point by observing
where light first passed between the pipe and
straightedge (i.e., no feeler gauge was used).
This
explains why the NRC measurements were consistently
shorter than the QC measurements.
Based on the above
discussion and the NRC measurements, the NRC inspector
determined that the above-mentioned QC measurements
failed to achieve the required tolerance.
This issue
constitutes the second example of Violation 445/8919-V-01
discussed in paragraph 2.g.
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d.
Heating, Ventilation, and Air-Conditioning (HVAC)
(CPE-EB-FVM-CS-029)
The NRC inspector performed field and document
inspections of the following documentation packages
generated as a result of construction, engineering, and
inspection activities related to'pCHVP:
Seismic Duct Hangers
Unit
Room
DH-1-792-1N-1A
1
127
DH-1-792-1N-1B
1
127
DH-1-792-1N-1C
1
127
DH-1-792-1N-1D
1
127
DH-1-792-1N-1E
1
127
,
DH-1-792-1N-4A
1
127.
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Duct Segments.
Unit
Room
B-1-751-141
1
129
B-1-751-142
1
129
B-1-751-143
1
129
'B-1-751-144
1
129
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Duct segments B-1-751-141, 142, 143, and 144 were
initially identified as Seismic Category II.
Concerning
these duct segments, the NRC inspector identified the
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following discrepancies:
unpainted surfaces, bent access
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door handles, oversized washers on bolts, distorted hem
flanges due to overtorque, loose register bolts, lack of
fusion on welding, and welding over galvanizing.
Subsequent to this inspection, the NRC inspector
determined that the duct segments in question are
actually non-nuclear safety.
The previous classification
was due to a typographical error.
The NRC inspector
questioned this safety classification in that Technical
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specification components (battery chargers) are located
in the same room.
The applicant provided the explanation
that the battery chargers are assumed to fail in the
accident analysis and, therefore, the ventilation of
this space is not essential for the safe shutdown of the
unit.
The NRC inspector concurred with this explanation.
In that the duct segments are not safety-related and the
identified discrepancies are not likely to affect the
!
ventilation of this space, correction of the identified
'
discrepancies is left to the discretion of the applicant.
No other deficiencies were identified among the items
referenced above.
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e.
Structural and Miscellaneous Steel (CPE-SWEC-FVM-
EE/ME/IC/CS-086 and CPE-SWEC-FVM-ME/EE/IC/CS-090)
FB-271B-01, platform, Auxiliary building, Unit 1:
The
NRC inspector reviewed the PCHVP package'and inspected
the platform.
The QC inspector had previously identified
. unsatisfactory conditions related to:
(1) item size and
length, (2) item location and orientation,
(3) alterations, (4) items not installed, and (5) weld
,
location.
These conditions were generally identified on
'
NCR 88-11162, Revision 0, which has not been
dispositioned.
With the exception of weld location, the
specific details of the above unsatisfactory conditions
are not provided on NCR 88-11162 or the corresponding
inspection reports.
The NRC inspector, therefore, could
not verify the location of'the unsatisfactory conditions.
r
The NRC inspector. identified two additional concerns:
(1) The spacing between two bolts on the baseplate shown
in Section A-A on drawing MFB-0803-DCA, Sheet 2,
i
1
Revision 1, is specified to be 14" (1 2").
The actual
spacing is 11.3/8".
(2) The weld type and size for the
installed toe boards are not specified.
.
l
The above. concerns will be addressed as part of.the continuing
review of unresolved item 445/8905-U-02, which identified
similar problems.
6.
Plant Tours (46053, 48053, 42051C, 50100, 55150, 45051, 45053,
46061)
The NRC inspectors made frequent tours of Unit 1, Unit 2, and
common areas of the facility to observe items such as
in-process work activities, equipment protection, and
housekeeping activities.
The NRC inspector' observed work in
progress related to the closure of a temporary access opening
used for cleaning and removal of debris from the seismic air
gap.
The access opening was approximately 8 feet x 8 feet in
the Auxiliary building outer wall.
The closure work was being
performed under Traveler CCE-88-1239-8404A, package
02C-8904A-1855-3 and 02C-8904A-1855-4.
The work included
'
concrete chipping, cleaning, grouting, installation of
concrete forms, cad-welding, and concrete pouring.
The NRC
l
inspector will continue observation of related work during
subsequent inspection periods.
No violations or deviations
were identified and no items of significance were observed.
7.
Exit Meeting (30703)
An exit meeting was conducted April 4, 1989, with the
applicant's representatives identified in paragraph 1 of this
report.
No written material was provided to the applicant by
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did.not identify as proprietary any.of the materials provided.
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.to'or reviewed by:the' inspectors.'during this inspection.
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During this meeting,ithe NRC inspectors summarized the scope
Land findings'of the inspection.'
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