IR 05000445/1989088
| ML20005G117 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 01/10/1990 |
| From: | Latta R, Livermore H Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20005G114 | List: |
| References | |
| 50-445-89-88, 50-446-89-88, NUDOCS 9001180148 | |
| Download: ML20005G117 (19) | |
Text
.
,
'
o
.
,
,
,
'
L
..
.
'
'
.
i U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION NRC Inspection Report:
50-445/89-88 Permits: CPPR-126 50-446/89-88 CPPR-127
Dockets: 50-445 Construction. Permit 50-446 Expiration Dates:
Unit 1: August 1, 1991 Unit 2: August 1, 1992 Applicant:
TU Electric Skyway Tower 400 North Olive Street
,
Lock Box 81 r
Dallas, Texas 75201 Facility Names Comanche Peak Steam Electric Station (CPSES),
.
Units 1 & 2
.*
Inspection At:
Comanche Peak Site, Glen Rose, Texas Inspection Conducted:
December 6, 1969, through January 2, 1990 AJ/
Inspector:
N 46O
/-/o '/o l
R. M. Latta, Resident Inspector Date l
(Electrical) (paragraphs 4, 5 and 6)
Consultant:
J. L. Birmingham - RTS (paragraphs 3 and 4).
W. D. Richins, Parameter (paragraph 2)
l P. Stanish -' Parameter (paragraphs 3 and 4)
-
!
!
Reviewed by:
fdw #2 V
/-/O-74 H. H. Livermore, Lead Senior Inspector Date l
-
,
i l
l 9001180148 900110 PDR ADOCK 05000445 PDC g
.
-
.
.
-
L
-
'
l
+
.
.
'
.,
.'
-2-
.
Inspection Summary:
Inspection Conducted:
December 6, 1989, through January 2, 1990 (Report 50-445/89-88: 50-446/89-88)
Areas Inspected: Unannounced, resident safety inspection of the applicant's actions on previous inspection findings; follow-up on
,
violations / deviations; 10 CFR 50.55(e) deficiencies identified by the applicant; and plant tours.
Results:
Within the areas inspected no significant strengths or weaknesses were identified.
During the inspection period, no significant safety matter, violation, deviation, or unresolved item was identified.
<
h i
_
s
'
s
.
,
'
.,
.'
-3-
,
DETAILS 1.
Persons Contacted
- J. W. Beck, Vice President, Nuclear Engineering, TU Electric
!
- 0.
Bhatty, Issue Interface Coordinator, TU Electric
,
- M. R. Blevins, Manager of Nuclear Operations Support,
{
TU Electric
- H. D. Bruner, Senior Vice President, TU Electric
- A.
R. Buhl, IAG~
- W.
J. Cahill, Executive Vice President, Nuclear, TU Electric
- H. M. Carmichael, Senior Quality Assurance (QA) Program Manager, CECO
- J.
T.
Conly, APE-Licensing, SWEC
- D.
E. Deviney, Deputy Director, QA, TU Electric
- F.
Dunham, QA Issue Interface, TU Electric
,
- J.
C.
Finneran, Jr., Manager, Civil Engineering,
!
TU Electric
,
- J.
L. French, IndepOndent Advisory Group l
- B.
P. Garde, Attorney, CASE i
- W.
G. Guldemond, Manager of Site Licensing, TU Electric
- T.
L. Heatherly, Licensing Compliance Engineer, TU Electric
- J.
C. Hicks, Licensing Compliance Manager, TU Electric
- C, B. Hogg, Chief Engineer, TU Electric
- J.
L. LaMarca, Manager of Electrical and I&C Engineering, TU Electric
- F. W. Madden, Mechanical Engineering Manager, TU-Electric
- J. W. Muffett, Manager of Project Engineering, TU Electric
- S.
S. Palmer, Project Manager, TU Electric
_
,
- P. Raysircar, Deputy Director / Senior Engineer Manager, CECO
)
- J. D. Redding, Executive Assistant, TU Electric
- M. J. Riggs, Plant Evaluation Manager, Operations, TU Electric
'
- A.
B. Scott, Vice President, Nuclear Operations,-TU Electric
- J.
C.
Smith, Plant Operations-Staff, TU Electric
- R.
L. Spence, TU/QA Senior Advisor, TU Electric
'
- P.
B. Stevens, Manager of Operations Support Engineering, TU Electric
- C.
L. Terry, Manager of Projects, TU Electric
- R.
D. Walker, Manager of Nuclear Licensing, TU Electric
- R.
G. Withrow, EA Manager, TU Electric
- D.
R. Woodlan, Docket Licensing Manager, TU Electric
<
The NRC inspectors also interviewed other applicant employees during this inspection period.
- Denotes personnel present at the January 2, 1990, exit meeting.
.
- ~
.
.
-
-
.
--.
.
'%
s.
,
.
..
.
-4~
-
,
2.
Applicant's Action on Previous Inspection Findings (92701B)
l (Closed) Open Item (445/8973-O-11):
This item addressed the apparent lack of understanding of control room personnel
l regarding task manpower requirements.
Immediately prior to the April 23, 1989, auxiliary feedwater.(AFW) backflow event (see
'
NRC Inspection Report 50-445/89-30; 50-446/89-30),. the control room operators sent only one auxiliary operator, near shift change, to operate valves 1AF-041 and 1AF-042.
As documented in the above referenced inspection report, additional time and/or manpower was required and the control room operators should have been aware of the time required for one individual to operate these valves.
Subsequent to this event, the applicant has prepared three documents to address this open item.
Procedure ODA-302, Revision 8, dated October 20, 1989, discusses the relief of personnel at shift changes.
Specifically, ODA-302_ states that:
(1) the shift supervisor may delay shift turnover and (2) the personnel involved in an activity requiring shift turnover on station should be aware that they will be relieved on location.
Procedure OWI-206, Revision 1, dated December 1, 1989, provides guidelines for the operation of manual and power operated valves.
This procedure includes a list (Attachment 6) of
.
valves with remote operators which are difficult to operate and i
provides estimates of the time and number of turns required to operate each valve.
The applicant has also developed an informational operator aid which contains data sheets for each safety-related valve.
Included on each data sheet are the valve identification and location, the direction to close, an estimate.of the difficulty in operating the valve and the approximate time required to operate the valve, as well as a statement concerning valve accessibility.
As stated by the applicant, these-data sheets will be available in the control room for reference.
The inspector has reviewed the above documents and' concluded that they collectively address the identified concern.
Therefore, this item is closed.
3.
Follow-up on Violations / Deviations (92702)
a.
(closed) Violation (445/8513-V-01):
This violation involved multiple examples of the applicant's failure to assure the proper identification of embedded conduit at
"through-wall sleeves" (TWS) and "through-floor. sleeves" l
(TFS).
l
'
The applicant acknowledged the violation and has implemented corrective and preventive actions.
These
.
l
-
.
.
.-
'
i
.
.
,
,
.,
.
.
-5
'
.
,
i actions included the revision and clarification of the I
applicabic construction and-construction inspection procedures, as well as the retraining of the affected
personnel.
Additionally, in order to identify all of the
,
'
discrepant TWS and TFS installations, an attribute for the verification of the proper identification of all Unit 1 and common TWS and TFS was included in Field' Verification Method (FVM)-CPE-FVM-EE-023, " Acquire Data for Cable
Percent-Fill Calculations and Identification of Thru-Floor and Thru-Wall Embedded Conduit Sleeves."
In accordance
"
with this FVM, incomplete or discrepant TFS and TWS
'
conduit identifications were scheduled for rework by the issuance of a design change authorization (DCA).
As determined by the NRC inspector, the actual field rework
'
was accomplished by the implementation of a construction traveler with Quality control (QC) verification required as a step on the traveler for all safety-related conduits.
The QC verification was documented on the appropriate QC inspection reports and the required verification step was signed as complete on the traveler.
The implementation of CPE-FVM-EE-023 was evaluated by the NRC as part of the verification of the applicant's Corrective Action Program (CAP).
The NRC inspection ~of the CAP is: documented:in various NRC reports and summarized in NRC Inspection Reports 50-445/89-14, 50-446/89-14; 50-445/89-28, 50-446/89-28;-and 50-445/89-61, 50-446/89-61..
Based on these inspection results, FVM-023 was determined to have been properly
'
implemented.
The NRC inspector reviewed 15 DCAs, as well as the accompanying travelers which documented the
,
implementation of corrective actions.
These travelers were also found-to have been satisfactorily implemented.
Based on the above reviews'and documented inspection activities, the NRC inspector determined that the applicant's corrective actions regarding this issue were i
acceptable.
Therefore, this violation is closed.
.b.
(open) Violation (446/8604-V-03):
This violation concerned the failure to install. cable support grips as required by Specification 2323-ES-100.
In particular, this discrepancy was identified for cables installed in Cable Tray Sections T23GECX91 and T24GEDG98.
The applicant has evaluated this violation and has issued Deficiency Report C-87-2743 to document.the-procedural violation.
Furthermore, Specification 2323-ES-100 has been revised to state that engineering is responsible for the identification of required Class lE cable support grip-installations.
.
-
- -
- -
.
.
. - _ -
.
._
.-.
.
.
,
.
-6-
'
.
.
Although the violation concerned deficiencies identified in Unit 2, the generic implications for Unit 1 and common areas were addressed in the applicant's response contained in TU Electric letter TXX-6481 dated. June 30, 1989.
In particular, FVM-089 was developed to require engineering walkdowns of the installed cable to determine if engineering requirements for cable support grips had been satisfied.
The NRC inspector determined that FVM-089 had been completed for Unit 1 and common areas and that as a result of the associated engineering walkdowns, 15 design changes requiring the installation of cable suppo; grips had been initiated.
,
The NRC inspector reviewed the'above1information it.,luding l
the 15 issued design changes and determined that the subject FVM and the issued DCA appeared adequate'to
'
>
address the concern.: Furthermore,-it is noted that FVM-089 was-implemented as part of the applicant's Post-Construction Hardware Validation Program (PCHVP).
NRC inspection of the PCHVP determined it to have been implemented satisfactorily and although NRC inspection documentation of the PCHVP appeared in various reports, the inspection results were summarized in 50-445/89-14, L
50-446/89-14; 50-445/89-28, 50-446/89-28; and a
50-445/89-61, 50-446/89-61.
Based on the review of the above information, the NRC inspector determined that the applicant's corrective and preventive actions for Unit 1 and common areas appeared adequate to prevent reoccurrence.
However, the violation-will remain open for Unit 2 pending the completion of corrective actions and a satisfactory review of those actions by the NRC.
Additionally, in that a protracted
,
l length of time will have elapsed prior to completion of j
corrective actions for Unit 2, it.is recommended that an evaluation of the potential degradation of any deficient cable installations be included'as part of the Unit 2 corrective actions.
'
c.
(Closed) Violation (445/8966-V-01; 446/8966-V-01):
This-violation involved manufacturing defects in structural
!.
tubing installed in the service water tunnel.
These defects which required more than 150 feet of weld repair d
had been incorrectly identified by'the applicant as nonextensive installation deficiencies..This
'
characterization severely limited.the review of this issue for generic implications.
Furthermore, for structural tubing found to violate the minimum requirements of ASTM A 500, the applicant's review was limited to specific sections of identified tubing with no review-performed for potential generic implications.
s
,
--
- - -,
.
c-
.
.,
.
-7-
.
t As stated in the applicant's response to this violation contained in TU Electric's letter TXX-89812, the basis for the engineering conclusions had not been adequately described.
Specifically, the determination that indications identified by ultrasonic testing in the tubing weld seams and thin wall tubing which did not represent reportable deficiencies had not been sufficiently-documented.
The NRC inspector reviewed the applicant's corrective actions including the technical evaluations contained in the Consolidated Engineering Construction Organization (CECO) report No. 15454-N(C)-021, Revisions l=and 2.
Additionally, the NRC inspector determined that structural'
tubing repairs have been effectively implemented on the affected supports and that the tubing manufacturer has been notified of the subject defect as documented in the applicant's correspondence CPSES-8901968.
Based on the above reviews which included structural tubing considerations for both Units 1 and 2, the NRC i
inspector determined that the applicant's corrective i
measures appeared adequate.
Therefore, this violation is closed for both Units 1 and 2.
For other actions taken in response to this violation and the associated construction deficiency (see paragraph 4.c).
>
4.
Action on 10 CFR Part 50.55(e) Deficiencies Identified by the
Applicant (92700)
'
a.
(Closed - Unit 1 only) Construction Deficiency'
(SDAR CP-86-19): ~This construction deficiency was issued to address the incorrect installation.of steam service
'
pressure transmitters.
However, the scope of this item
- '
was expanded to include corrective and preventive actions for five related construction deficiencies.
For the-sake of clarity, each issue will be addressed separately referencing the initial SDAR number and title.
(1)
SDAR CP-86-16, " Fire Effects on Instrumentation Tubing."
This issue concerned the potential for zinc embrittlement of stainless steel instrument tubing during a fire.
In their approach to this issue, TU Electric identified those plant fire zones in which combustible loading was sufficient to result in room temperatures, in excess of the melting point of i
zinc, during a fire.
Those areas identified as having a high temperature potential were reevaluated
.
taking into consideration the effect of the fire
protection / suppression systems available in those
,
.
.
-_. -
.
_.
.
..
-
l
.
!
.
o.
w
-
.
'
l
.
i-8-
'
.
l areas.
The applicant's evaluation determined that in I
all but two of those areas the fire alarm and fire suppression systems were sufficient to ensure timely
..t I
control of fires and prevention of elevated temperatures.
The two areas excepted were.the
,
l battery rooms and the uninterruptible power supply
,
rooms.
The evaluation of these rooms determined that
'
they did not contain instrument tubing of concern.
l
.
Accordingly, TU Electric determined the issue-not to I
be reportable.
'
,
The NRC inspector assessed the applicant's approach to this issue, the methodology used to evaluate the i
'
areas, as well as the applicant's final determination as to the affected areas.
The NRC inspectc1 i
determined that the applicant's assessment was adequate and that the nonreportability determination was acceptable.
Accordingly, this item-is closed for Unit 1.
(2)
SDAR CP-86-19, " Improper Installation of Steam
,
Service Pressure Transmitters."
The applicant determined that certain steam service pressure transmitters mounted above:their associated root valves without steam traps or wet legs could be subject to condensation flashing resulting in erratic signal outputs.
To address this issue the cpplicant developed design modifications to relocate or-to provide wetlegs for these instruments.
Design Modification-Request Construction (DMRC) 87-1-010 was issued to address changesifor'the Unit l'
safety-related steam systems and DMRC-87-1-158 was issued to address changes for the Unit 1 nonsafety-related' steam-systems.
These DMRCs have been completed.
Configuration of instrument
,
installations was validated.during the PCHVP.
The NRC inspector reviewed the above documentation which verified the' completion of DMRC 87-1-010 and 87-1-158 and determined that the applicant had satisfactorily addressed this issue for Unit 1.
However, this issue-romains open for Unit 2 pending the satisfactory completion of DMRC 87-2-009 for Unit 2-systems.
(3)
SDAR CP-86-50, "Unistrut Spring Nuts on Instrument Supports."
This SDAR addressed a deficiency concerning nut alignment and torque requirements for Unistrut spring nuts on instrument mounts.
As a l
result of this issue, the applicant revised the applicable drawings and specifications to incorporate
,-
l the specific torque values and spring nut alignment
.
-. - -
-
.
. - -
-
-
.
-
-.
-.
. - _ _. - - _,
.
o
,
'
.
-9
-
criteria.
Additionally, FVM-086 was impletaented to perform a reinspection of bolts, torquing and spring nut alignment for instrument tubing supports.
NRC inspection of the implementation of FVM-086 determined that it had been performed satisfactorily
for Unit 1.
Accordingly, this item is closed for Unit 1.
(4)
SDAR CP-86-70, " Elevated Temperature Effects on Tubing."
This deficiency specifically addressed the potential that elevated temperatures produced during a high energy line break-(HELB) or a loss of coolant
.
accident (LOCA) would be detrimental to instrument l
supports and tubing rendering them incapable of
,
l performing their safe shutdown function.
.
The applicant has completed the engineering evaluation of this issue and has determined that the elevated temperature effects that occur during a HELB or LOCA are adequately accounted for in the methodology and conservatism of existing design j
criteria.' This determination was based, in part, on calculations documented in Engineering Evaluation
'
Report 01-0210-1065, Revision 4, dated March 1987.
.
!
That evaluation performed by Impell determined the
!
i proper configuration =of tubing, clamps, and'
l instruments both at normal 1 operating conditions or at
i conditions expected to occur during a high energy l
line break.
,
The NRC inspector reviewed the above documentation and determined that the applicant had adequately demonstrated that elevated temperatures would not prevent the instrument. tubing and supports from
.
performing their safe shutdown function.
{
Accordingly,.the inspector agrees with the
applicant's determination that this issue was not
)
reportable under the provisions of 10 CFR,
!
i
'
Part 50.55(e).
This item is closed for Unit.1.
(5)
SDAR CP-86-77, " Instrument Tubing Minimum Wall i
l Thickness."
This deficiency concerned.the potential i
l that for certain size tubing in.high temperature and l
pressure applications, the wall thickness of.the tubing may be insufficient.
The applicant addressed this issue by performing an
!
analysis of the required minimum wall thickness versus the design temperature and pressure.: The results of the evaluation were issued as calculation
'
No. 16345/6-1C-(B)-001, Revision 1.-
The NRC
!
l
!
l
L l
E
,
.-
C
-
.
'
.,
'
.
,
inspector determined that the applicable information from this calculation was incorporated into Instrumentation Installation Specification CPES1-1018, which was used, in part, as i
the basis for QC reinspections performed in accordance with FVM-086, "PCHVP Construction / Quality Control Reverifications."
The NRC inspector reviewed calculation No. 16345/6-1C-(B)-001 and determined that it properly evaluated the minimum wall thickness
requirements as prescribed by ASME Section III,
Article NC-3676 and NC-3641.
This information.was incorporated into Appendix C of Specification CPEG-1-1018 and into the list of attributes verified i
!
by the implementation of FVM-086.
Based on the above-reviews and the satisfactory completion of FVM-086, the NRC inspector determined that the applicant's
!
actions were acceptable for~ Unit 1.
However, the
!
closure of this issue for Unit 2 has not yet been addressed by the applicant.
l In that each of the cumulative instrumentation related issues included in the expanded SDAR CP-86-19 were i
satisfactorily addressed-for Unit 1, construction deficiency SDAR CP-86-19 is considered closed for Unit 1.
i b.
(Closed - Unit 1 only) Construction Deficiencies (SDAR CP-86-52 and SDAR CP-87-76):
" Cable Tray Splices / Connections" and " Field. Drilled Cable Tray Holes."
By Letters TXX-6415 and-TXX-6763, the applicant notified I
the NRC of a reportable deficiency involving
-
splice / connections used for cable trays in Units 1 and 2.
Specifically, splices / connections used for some' tray ~ types and sizes were neither installed in accordance with the
manufacturers requirements nor the approved CPSES design.-
'
Further, certain field drilled cable tray holes.were not I
in accordance with the specified design.
.
Based on the similarity of these issues, the applicant provided a combined response which attributed the root cause to inadequate construction ~and construction inspection procedures.' Accordingly, the applicant revised the controlling Electrical Construction Procedures ECP-10-for Unit 1 and ECP-10A for Unit 2.
Relative to'the
misdrilled/ unused holes identified in SDAR CP-87-76, the applicant has incorporated the criteria'for the holes'into Electrical Specification ES-100 and has revised the applicable construction and inspection procedures.
Additionally, the NRC inspector ascertained that the-applicant has performed both testing and analytical
1
..
-
-
-
.
.-.
.
.
,
.
.
- '
.,
?
-11-
.
evaluations to determine the acceptability of alternate configurations for cable tray splice / connections.
Regarding the misdrilled holes,- the. applicant based 11ts.
l evaluation on a worst-case configuration and assimilated the as-built data for the cable tray configurations under
'
'
the auspices of FVM CPE-EB-FVM-048.
As a result of this action, the unacceptable splice or unused hole configurations were modified'to meet the approved design criteria.. Also, Electrical Specification 2323-ES-100 was updated to provide the drawing references for the location i
of nonstandard but acceptable splice / connections.
It is noted that when this issue was identified, the applicant assumed that virtually all of the Unit 1 cable trays were supplied by one manufacturer, T. J. Cape, i
However, during the evaluation of Unit 1 and common areas,
,
instances of cable trays supplied by Burndy-Husky were l
identified.
For those instances where the cable trays had been supplied by Burndy-Husky, the applicant performed i
individual evaluations and the necessary modifications i
were implemented.
j i
l The NRC has inspected the implementation of the PCHVP i
including FVM-EB-048.
The NRC inspection results were in j
general satisfactory as reported-in NRC Inspection Reports 50-445/89-14, 50-446/89-14; 50-445/89-28, i
50-446/89-20; and 50-445/89-61,.50-446/89-61.
Based on a review of the revised construction and
inspection procedures, the applicable sections of FVM-EB-048, and the conclusions stated in Impell l
Report 11-0210-0026, the NRC inspector determined that the'
applicant's corrective and preventive actions for these-
,
'
two deficiencies were satisfactory.
Therefore, construction deficiencies SDAR CP-86-52 and SDAR CP-87-76 are closed for Unit 1.
c.
(Closed) Construction Deficiency (SDAR CP-88-28):
During
'
the assembly of a Unit'2 pipe support,-a crack on the internal radius of a piece of tube steel (structural
'
tubing)'was identified.
Further investigation identified-similar defects on tube steel supplied by the same manufacturer which had:been documented on the same receiving inspection-report (RIR).
In addition, the wall thickness of some tube steel appears undersize.
In the applicant's letter TXX-88562 covering the topic of tube'
steel manufacturing defects, it was reported that
.
split-seam welds on tube steel in the service water pipe tunnel had been identified on nonconformance reports
(NCRs); and the scope of this construction deficiency evaluation was increased to address this issue.
- =
_..
_
.
_
.
.
_.
.. _.
,
,
.
A
.
' '
.,
>
-12-
,-
As previously documented in NRC Inspection-Report 50-445/89-66; 50-446/89-66, the NRC inspector determined that the applicant's efforts and evaluation of the corner cracking problem identified in structural tubing supplied l
by UNR/Leavitt appeared adequate..However, on the two additional issues addressed in this construction deficiency, the NRC inspector determined that the applicant's approach to the' resolution of these issues.was incomplete.
!
Specifically, for the tubing found to have a wall thicknesses of less than the specified requirements, the NRC inspector's concern was that no extended review had been performed to-quantify the issue.
However, during-this inspection period, the applicant provided additional information concerning this issue as delineated in Revision 1 to Report 15454-N(C)-020 (see paragraph 3.c).
The original review of this-issue by the applicant was limited to the two heats of-material-initially identified-with no discussion of the rationale for not extending this review to other nonconforming structural tubing provided by the vendor or other vendors.. The current revision of the above referenced report details several reasons why.
further evaluation of this nonconforming condition was not performed.
In particular, the, applicant's rationale included the assertion that the original population-of supports evaluated were regarded as representative of the entire population and that the analysis made use of
,
conservative design parameters.
'
'
Relative to the NRC concern as to why this condition'had i
not been identified by receiving inspection, the applicant's current report stated that; during the period of 1980 through January 1988, receiving. inspection only.
verified the thickness of structural tubing at the ends.of i
the sections.
However,' based ~on the findings identified on the two original heats of material, ultrasonic (UT)
examination was instituted as a part of receiving inspection in January 1988.
The NRC inspector's review of
the receiving inspection reports (from January'1988 through September 1989) for. wall-thickness violations indicated that four additional NCRs had been initiated.-
These NCRs were related to the heats of material supplied
,
by UNR/Leavitt (the same vendor that had produced tubing
'
with cracks in the corners discussed above).
The NRC i
inspector determined that the concern relative to-minimum.
wall-thickness violations in relation to UNR/Leavitt had
'
been identified by quality control who had initiated an evaluation of UNR/Leavitt supply independent of the engineering evaluation.
The four subject.NCRs had documented minimum wall-thickness violations ranging from 2.5 to 3.6 percent below specification minimums which l
,
_ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _. - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _. _,
_
.
___ _
.;
~'.-
.
.
j
.
,,
.
'
-13-
-
-.
appeared to be consistent with the previous findings. 'As a result of the extended QC evaluation of UNR/Leavitt i
supplied tubing, which included-the reinspection of tubing installed in safety-related applications, four additional NCRs affecting three heats of material _were issued.
The i
largest deviation identified during the applicant's reevaluation was 4.4 percent below specification minimum
'
which exceeded the minimums previously_ evaluated in Revision 0 of the subject report.
Consequently, the four
.
supports which utilized the nonconforming material were reevaluated and determined to meet code allowables.
The NRC inspector reviewed the data presented in Report 15454-N(C)-020, Revision 1, relative to this issue; and based on this expanded review has determined that the applicant's conclusion that this issue was not reportable is acceptable.
Regarding the issue of the split seams on the service water tunnel pipe support frames, the NRC inspector questioned the assumptions and conclusions of the
'
applicant's review of this issue as previously documented in NRC Inspection Report 50-445/89-66; 50-446/89-66.
In response to this concern, the applicant provided the NRC inspector with amplifying information contained in TU Electric letter CPSES-8901968 to Welded Tube of America (WTA) notifying them of the manufacturing defect identified in the tubing used in the subject frames.
,
l Also, the: applicant's ReportL15454-N(C)-021, Revision 2, j
documented the testing of 56 heats of WTA supplied tubing
'
with no additional failures of the seam weld reported.
,
Based on the above reviews and inspection related activities,'the NRC inspector determined that_the applicant's actions which. included: system' repairs and the notification to the vendor were acceptable.
Therefore, this construction deficiency is closed.
d.
(Closed - Unit 1 only) Construction Deficiency
,-
l (SDAR CP-88-042):- During the conduct of maintenance i
activities on valve lHV-4514 in the component cooling l.
water (CCW) system, a flange misalignment was discovered.
l Subsequently, it was determined that excessive forces had
!
been utilized in the realignment of the piping by
!
employing a hydraulic jack to reposition the piping which i
was contrary to site procedures.
The initial NRC review of this issue was documented in NRC l
Inspection Report 50-445/89-66; 50-446/89-66.
As identified by the applicant, the root'cause of this issue
was attributed to the welding process, and in light: of the high stresses induced by this abnormal event, the
'
l
'
^
,. __
!
'
..
q o
.
,
' '
s,
.
-14-
-
applicant-committed to review the attendant circumstances and_ determine if additional controls on the welding process were necessary.
As a result of the applicant's continuing actions in this mattor, the inspector reviewed DCA 87041, Revision 6,
,
which revised Specification 2323-MS-100, Revision 10.
This DCA addressed the subject issue by adding controls to-
the welding of support attachments to nuclear safety-related piping.
Specifically, this specification now requires that these attachments be reviewed by welding engineering prior to release to construction, and defines the type of review required.
It also stipulates.that the welding engineering group is responsible 'for specifying the proper weld procedure to be utilized to minimize _the i
potential for misalignment due to weld. shrinkage, as well
as allowing welding engineering to specify the
'l construction sequence, if necessary.
Based on the review of this additional information, the l
NRC inspector determined that adequate controls are now in-
';
place so that the potential for piping misalignment due to-weld shrinkage has'been minimized.
Therefore, this construction deficiency is closed for Unit 1 only pending the implementation of similar corrective actions-for
,
Unit 2.
j i
c.
(closed - Unit 1 only) Construction Deficiency'
.i (SDAR CP-89-01):
By letter'TCO-89001 dated January 16,
!
(
1989, TU Electric notified the NRC of a deficiency found
'
in certain Unit 1 limit switches.
Specifically, TU Electric had identified cracks in the contact strips of approximately nine NAMCO limit switches._ TU Electric has performed an in-depth evaluation of this deficiency:and determined the item to beLreportable in accordance with
10 CFR, part 50.55(e).
Investigatory actions taken by-TU Electric included
,
enlisting the assistance of the manufacturer to determine i
the extent and most probable cause of the deficiency as:
well as submitting six contact' strip assemblies to an
independent laboratory for testing.
Tests conducted by
.i the independent laboratory consisted of:
!
Visual examination and photography.
.
Scanning electron microscopy of contact strip
.
fracture surfaces.
Energy Dispersive X-ray (EDX) and Auger elemental
.
analysis of deposits on the contact strip fracture
,
!
surface.
_ -
_.. _ _ _
_
l
..
.
.,
.
'
-15-
,
.
Cutting, mounting, polishing, and metallographic
,
.
examination of the contact strips.
,
l Chemical analysis of the material from a cracked
'
.
'
contact strip using NBS standards and comparison to specifications.
Hardness testing of the contact strips.
.
Over-torque tests of screws into the contact strip
.
inserts.
No definitive cause was determined from=the above tests; however, the laboratory concluded that the probable cause was residual stress in the contact strip and the presence
,
of preplating contaminants.
The evaluation performed by TU Electric determined that-the deficient contact strips were found-in only NAMCO Model EA-170 and EA-180 limit switches.
Furthermore, it was determined that all of the deficient limit switches were manufactured with a-date code of.1882, indicating'the-18th week of 1982.- TU Electric requestedLinformation from the manufacturer as to whether there was anything unique
'
about the manufacturing process for limit switches EA-170 and EA-180~for that time frame. 'NAMCO responded that there was no record of any variation in the manufacturing process for that week.and that similar contact strip.
assemblies were used in switches manufactured between January 1982 and February =1984.
Accordingly, TU Electric-focused its corrective and preventive actions on limit switches manufactured'during the above time frame.
TU Electric inspected all of the switches'with date codes t
between January 1982.and February.1984 installed in Unit l'
I or still in site stores.
That. inspection identified an additional ten cracked' contact' strips each'with a date code of'1882.
Correspondingly,JTU Electric replaced all of the limit switch NAMCO Models EA-170 and EA-180 with
~
L date codes of 1882, and the TU Electric Nuclear Operations-
,
Defective Item List (NODIC) was revised to restrict the procurement of NAMCO limit switch Models EA-170 and EA-180 with'a date code of 1882.
l The NRC inspector reviewed the above-information as well
!
as additional closeout documentation in the' subject SDAR
!
file and determined that the applicant's corrective
!
actions were acceptable.
Therefore, this construction deficiency is closed for Unit 1 only pending the implementation of similar. corrective' actions for Unit 2.
f.
(Closed) Construction Deficiency (SDAR CP-89-20):
This issue involved Limitorque-actuator mounting bolts which
.
.
-
-
.
.-
-
. - - - _
-.. _ _. _ -_.- -_
_ _ _ - _ _ _ _ _ _
i
.
o
-#
,
'*
.o,
,
.'
-16-
.
,
were required to meet the material specification of ASTM A193 Grade B8M, Contrary to this requirement, the e
'
bolts installed in the field were not marked "B8M" on the head-but were found to be inscribed with the Number 316 which is not a recognized symbol for any type or grade of standard bolting.
In response to this issue, six. bolts were removed from
!
Unit 2 Limitorque actuators and were destructively tested.
The test results indicate that the bolts meet the chemical composition.and mechanical properties specified per ASTM A193/A193M-87 Grade B8M.
Based on these test
.
results, the applicant concluded that in the event that i
this deficiency had remained uncorrected, it would not
.
I have adversely affected the safe operation of-the plant;
'
and, therefore, was not reportable.
The NRC inspector reviewed NCR PM-87-00357 which documented this condition and specified the replacement of the subject fasteners with the properly identified material.
Additionally, the NRC-inspector reviewed the metallurgical evaluation of-the bolts and determined that this issue had been appropriately characterized as nonreportable.
Therefore, this construction deficiency is-closed.
,
g.
(Closed) Construction Deficiency (SDAR CP-89-27):
This deficiency involved a " crack like" condition which was-l identified on an ASTM A 500 tube steel instrument air support.
In particular, this' deficiency was observed on the inside wall surface-parallel to and outside the heat affected zone of the weld.
Subsequently, Deficiency' Report (DR) C-88-00558 was written-to document this condition and initiate an evaluation.
As established 1by the applicant's supporting engineering calculations contained in Report PTR-017, this specific deficiency would not have.resulted'in a failure under the design loads.
Additionally, the investigations
-
performed.by the applicant indicated that the condition found was apparently not generic and that there was reasonable assurance that the' condition was limited to the 100 feet of the subject material'(heat No.. 57423) received at CPSES.
As determined by the NRC inspector, all
-
identified supports containing the suspect material for Unit 1, common and Unit 2 have been reworked ~to remove the H
tube steel fabricated from heat No. 57423.
,
Based on the review of the above referenced report and the'
'
evaluations performed, the NRC inspector determined that the applicant's actions were acceptable.
Therefore, this construction deficiency is closed for Units 1 and 2.
,
4,__m__
---
.
.
-
.
.
-
-
e i
.-
.
- r,
,.
'
-17-
.
'h.
(closed - Unit 1 only) Construction Deficiency (SDAR CP-89-30):
"AFW Pump Turbine Low Lube Oil Pressure Switch."
By letter TCO-89113, TU Electric notified the NRC of a deficiency concerning pressure switch 1-PS-2452-4 installed on the auxiliary feedwater pump turbine.
Specifically, documentation to certify the pressure switch as a qualified component for Class lE application could not be substantiated.
,
The applicant evaluated the pressure switch and its I
application-and determined that an acceptable electrical
'
separation between'the Class 1E power source and the unqualified pressure switch did not exist.
It is noted
that the pressure switch itself does not provide a i
safety-related function; however, in that the switch is, L
powered from a lE bus, a double isolation device between the lE bus and the pressure switch is required.
At the time the deficiency was identified, the pressure switch was isolated by a single fuse.
The NRC inspector determined that this deficiency was documented on NCR 89-10360 and that a disposition had been issued to-reconfigure the isolation circuit to provide double fuse isolation.
The double fuse isolation option.is an acceptable means of isolation as defined by the CPSES Final Safety Analysis Report (FSAR) and Design Basis Document (DBD)-EE-057.
l
.
I The NRC inspector reviewed CPSES FSAR, Section 8.3, and DBD-EE-057 and determined that the double fuse configuration would allow the pressure switch circuit to meet the requirements of IEEE Standard 384 as committed to by CPSES.
Furthermore, the NRC inspector' reviewed Work'
f Order C-89-14638 which documented the installation in
'
auxiliary relay rack No. 4 of the Unit 1 auxiliary feedwater pump turbine of spare fuse No. 4 in series with fuse No. 5 per the disposition of NCR 89-10360, l
Revision 0.
Based on the review of the above documents, the NRC inspector determined that the applicant had satisfactorily addressed the issue for Unit 1.
Accordingly, this item is l
considered closed for Unit 1 only pending the
!
implementation of similar corrective actions for Unit 2.
,
5.
Plant Tours (42051C, 51053, 52053)
l
.
'
The inspectors conducted routine plant tours during this inspection period which included evaluation of work in progress as well as completed work to determine if. activities involving l
safety-related electrical systems and components including L
electrical cable were being controlled and accomplished in i
l l
.
-
- - -
-
...
o
.
.
'
o,
,,
'
-18-
.
l accordance with regulatory requirements, industry standards, and the applicant's procedures.
No violations or deviations were identified.
6.
Exit Meeting (30703B)
An exit meeting was conducted January 2, 1989, with the i
applicant's representatives identified in, paragraph 1 of this report.
No written material was provided to the applicant by the inspectors during this reporting period.
The applicant did not identify as proprietary any of the materials provided to or reviewed by the inspectors during this inspection.
During this meeting, the NRC inspectors summarized the scope and findings of the inspection.
!
i
<
i i
{
l
!
i I
_
_ _ _. -. -
.
.
.
'
C -
-
'
50-445/79-Pf;50-446/f7-ff DISTRIBUTION:
Docket File'(50-445/446)-.
'
NRC PDR
'
LPDR
CPPD-LA
$
CPPD Reading (HQ)
ADSP Reading
'
- Site Reading File
- R. Warnick
- J. Wiebe
- H. Livermore
- MIS System, RIV
- RSTS Operator, RIV RPB, RIV
'
RIV Docket File J. Gilliland, RIV
,
- L.
Shea, ARM /LFMB C.
I. Grimes J. H. Wilson J.,Lyons M. Malloy J. Moore, OGC
,
'
M. Fields D. Crutchfield T. Quay E. Jordan B. Grimes B. Hayes C. Vandenburgh J.
Partlow
- w/766 L: