IR 05000445/1990003
| ML20006F380 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 02/16/1990 |
| From: | Latta R, Livermore H, Runyan M Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20006F378 | List: |
| References | |
| 50-445-90-03, 50-445-90-3, 50-446-90-03, 50-446-90-3, NUDOCS 9002270451 | |
| Download: ML20006F380 (96) | |
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l APPENDIX B
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NUCLEAR REGULATORY COMMISSION
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OFFICE OF' NUCLEAR REACTOR REGULATION v
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t NRC Inspection Report:
50-445/90-03 Permits: CPPR-126.
50-446/90
CPPR-127.
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Dockets: 50-445 Construction ~ Permit 50-446 Expiration Dates:
Unit 1: August 1, 1991 Unit 2: August.1,.1992 s
Applicant:
TU. Electric.
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' Skyway Tower.
- l-400 North Olive Street Lock Box 81 i
Dallas,- Texas 75201 l
l Facility Name:
Comanche Peak Steam Electric Station (CPSES),.
' Units 1 & 2
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l Inspection At:
Comanche Peak Site, Glen Rose, Texas-
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Inspection. Conducted:. January 3 through February 6, 1990
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Inspectors:
[W/-
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R. M. Latt's, Resident Inspector Date l.
Electrical and Mechanical (paragraphs 2, 3, 4,.8,.
l and 9)
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\\L c1w 2-W M. F. Run}yan,'tural and Mechanical (paragraphs 2
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Resident Inspector,.
Date
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Civil Struc 3, 4,
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and 11)
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fother NRC Inspector:-
D. P._Norkin, NRR.(paragraphs 5 and 6)
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Consultants:
J. Birmingha:n, RTS (paragraphs 3, 4 and'10)-
W. P. Chen, Energy Technology Engineering Center (paragraphs 2 and 4).
W. Richins, Parameter (paragraph'7)
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Reviewed by:
, b r-S&kb H. Livermore, Lead Shnior Inspector Bath
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Inspection S ary:--
Inspection' Conducted: January 3 through February 6, 1990 (Report
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50-445/90-03; 50-446/90-03)
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Areas Inspected: Unannounced, resident safety inspection ~of a!
= applicant's actions-on' previous inspection findings; follow-up on'
. violations / deviations;Eaction on 10 CFR 50.55(e) deficiencies identified by the applicant; Commodity Clearance Program; follow-up of. Corrective Action-Program commitments; testing of piping supports
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"duringLhot functional testing; mechanical-components and equipment;
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allegation follow-up; applicant meetings;.and plant tours.
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. weaknesses were observed.
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i Results:-'Within the areas inspected, no significant strengths or-l
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'l Two violations were identified.
One violation involved inadequate
corrective actionLfor'an NRC-identified error in a structural l
calculation for the service water intake structureE(paragraph.2.d).
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The other' violation involved the failure to report a deficiency in.
q accordance with 10-CFR=50.55(e)-(paragraph 8.b)~
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An unresolved item was. identified questioning whether' nonconforming
i items documented'on an "unsat-IR" would be satisfactorily evaluated i
for 10 CFR Part 21 reportability (paragraph 10.b).
q An open~ item was identified to examine the generic. implications of an
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L inadequately patched drilled hole-in a concrete wall (paragraphril).
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.An additional open> item was identified to investigate the potential
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use of erroneous uncalibratable plant parameter gauges in safety
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applications (paragraph 4.r).
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DETAILS I
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Persons Contacted
- M.
Axelrad, Newman and Holtzinger
- J.
L.' Barker, Manager, Independent. Safety' Evaluation Group (ISEG), TU Electric
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- J._W.
Beck,LVice President,. Nuclear Engineering, TU Electric
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Bhatty,_ Issue-Interface Coordinator, TU Electric
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- M.
R. Blevins, Manager of Nuclear Operations Support,
'TU Electric.
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D. Bruner, Senior Vice President,LTU Electric
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- J.
H. Buck, IAG o
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C..Byrd, Manager, Quality Control-(QC), TU Electric:
- W.
J..Cahill, Executive Vice President,-Nuclear, TU Electric
- H. M. Carmichael, DBO EA Manager
- J. T. Conly, APE-Licensing, SWEC
- B.
S. Dacko, Licensing Engineer, TU Electric
- D. M. Ehat, Consultant, Bentham Group H
- D.
R. Ferguson, NESG,-TU Electric j
- S.
P._Frantz, Newman and Holtzinger
- B.
P. Garde, Attorney, CASE =
- W. G. Guldemond, Manager of Site Licensing, TU Electric
- T.-L.
Heatherly,_' Licensing Compliance Engineer, TU Electric'
- C. ~ 1B. Hogg, Chief Engineer, TU Electric
- J. J. Kelley, Plant Manager, TU Electric
- J.
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LaMarca, Manager of; Electrical and I&C Engineering, TU Electric j
L*H. Lawroski, Consultant, TU Electric j
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- J.:F.iMcMahon, Manager Nuclear Training, TULElectric
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- J. W. Muffett, Manager.of Project Engineering, TU Electric
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Ottney,. Project Manager, CASE
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- P.1Raysircar, Deputy Manager, Project Engineering,-TU Electric j
- M.
J. Riggs, Plant Evaluation Manager, Operations, TU Electric j
. A. B. Scott, Vice President, Nucicar Operations, TU Electric
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- J..C.' Smith,. Plant' Operations _ Staff, TU' Electric
- P.LB'.
Stevens, Manager of Operations Support Engineering,
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TU Electric d
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F._Streeter,-Director, QA, TU Electric
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Terry,-Manager.of Projects, TU Electric
- T.~G.
Tyler, Director, Management Services, TU Electric j
- J.
R. Walker, Operations / Engineering Training Manager, i
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TU Electric
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The NRC inspectors also interviewed other applicant employees
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during-this inspection period.
- Denotes personnel present at the February 6, 1990, exit meeting.
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NRC. personnel present at February 6,,1990, exit meeting are as j
follows:~~
S.
D. Bitter, Resident Inspector
W.
D. Johnson, Senior Resident Inspector
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R. M..Latta,. Resident Inspector H. H. Livermore, Lead Senior Inspector
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M. F. Runyan, Resident Inspector R.. F. Warnick, Assistant-Director forLInspection Programs.
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J..S.
Wiebe,. Senior' Project Inspector 2. -
Applicant's Action on Previous Inspection Findings (92701)
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(Closed) Open Item (445/8601-0-11):
This open item concerned the applicant's inspection program for certain'
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aspects of electrical separation.
Specifically,'the open
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item addressed:
(1) the installation of' separation barrier
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material (.SBM), (2) the inspection or acceptability.of
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abandoned holes in cable tray side rails, (3) the issuance
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of an engineering testing and evaluation;roport~regarding minimum acceptable' separation / criteria,'and'(4) a proposed-room / area electrical separation inspection'prcgram.-
Since~the electrical separation criteria was to be derived
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from the engineering report, the NRC inspector reviewed the engineering. report and other. data pertinent,to the report
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and-its issuance.
The evaluation and testing of: separation criteria was conducted by Wyle Laboratories;and was performed'specifically for.CPSES.
The testing used
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material and configurations. typical of CPSES-installations.
oThe criteria resulting from the Wyle Laboratories testing was subsequently incorporated into.CPSES Design-Basis
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Document =(DBD) EE-057, " Separation' Criteria."- As reported
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in SSER-17, the NRC staff reviewed DBD EE-057 and the L
attendant testing and analysis documentation and determined
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i-that this' testing and analysis documentation was applicable to~the CPSES electrical separation design' criteria and that L
the criteria are in accordance with Revision 1 of
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Regulatory Guide:1.75.
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As: stated above, the separation criteria resulting from the test report was incorporated into DBD EE-057.
Subsequently-the applicable construction and inspection procedures were a
revised accordingly.
The NRC inspector reviewed a sample
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of the-revised procedures; e.g.,.NQA 3.09-3.02,
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" Electrical / Raceway - Cable Tray," and determined that the criteria were properly incorporated or referenced in DBD EE-057.
The revised inspection procedures also addressed the installation of separation barrier material (SBM),'and the inspection of abandoned holes in cable tray side rai's, (Note: SBM was removed from most electrical an adverse-effect on ampacity ratings.)
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The inspection'of electrical separation was; accomplish'ed by
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Field Verification Method-(FVM)-088, " Electrical
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Separations,"
which was performed as part of the-
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Post-Construction Hardware Validation Program (PCHVP)'and-
'is'completeifor Unit 14 and Common.. The:NRC inspector deems
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the above actions to satisfactorily address the. concerns of the original open: item.
Since the various aspects of.the.
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open item have been adequately addressed, this.itemlis-
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(Closed) Open Item i445/8852-0-08):
This item was
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identified during the closeout' inspection efforts
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' associated with IEB 77-07..In particular,Lthis open item t
concerned the use of Kapton insulated wiring which is used
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in various' conductor sizes for;the: electrical pigtails on both safety-related and nonsafety-related Conax electrical
. penetration assemblies ~.
As previously documented in NRC-Inspection Reports 50-445/89-04, 50-446/89-04;
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50-445/89-73, 50-446/89-73; and 50-445/89-84, 50-446/89-84;
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- the technical issues regarding_the use of Kapton insulatedL
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conductors at CPSES as well as specific installation practices have been addressed and were determined to be generally _ acceptable.
Therefore,.this open item is-closed.
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(Closed) Open Item (445/8920-0-02):
' Loose shims on-q pedestal baseplate on the fuel transfer-mechanism.
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Specifically, the NRC inspector observed loose shims under L
the pedestal baseplate which supports the fuel transfer lis mechanism.
This condition was identified-to the applicant
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and1was documented in applicant ~ Deviation Report Ul-01388..
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Subsequent-to the identification of this item, the
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C89-0007671, Revision 0.
The NRC inspector reviewed the completed = work order, including the associated weld data record, and determined-that the applicant's actions-
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(Closed) Open Item (445/8948-0-01; 446/8948-0-01): 'This item involved an NRC identified ~ error'in the design structural calculation of the Service Water Intake
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E Structure (SWIS).
In calculation 16345/6-CS(B)-058, a groundwater level of 780 feet had been assumed, whereas
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piezometer readings adjacent to the north wall indicated a-
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J1 groundwater level of approximately 783 feet.
In-response
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a theJbase calculation which reanalyzed the SWIS north wall -
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for a groundwater level of 783 feet.
This open item, which was closed;in NRC Inspection Report 50-445/89-89;
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50-446/89-89,'was reopened during this inspection period as
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the result'of new information which revealed that Change
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Notice 2'was inadequate.
The NRC inspector reviewed a i
. requested copy of actualipiezometer readings taken'during e
1988 and noted that.both piezometers-13 and 24 (located-
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just off.the north wall'of the SWIS) indicated groundwater levels of greater than 783 feet.
Piezometer 24,. located'
closest to_the wall, had a peak reading of 783.'2 feet.
-piezometer 13, located a few feet.from piezometer.24 and o-farther from the wall, had a peak reading of 783.1: feet.
The_ assumed groundwater level;for the SWIS structural-calculation should have r arted with the highest reading of
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-these two piezometers (763.2 feet) and added additional
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margins to account for instrument error and'the-expected-
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changes in-groundwater levels over a 40-year service life.
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Clearly, the assumed groundwater level of 783 feet did not meet' this' objective.
Adding to the. concern is that 1988,
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the year in which.the measurements were_taken, was relatively dry with approximately 20 inches of precipitation.- During 1989, when approximately 40 inches of precipitation fell, no groundwater measurements were
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taken.
The average precipitation at the site _is around f'
,30 inches per year.
The failure to take~ adequate corrective action on an NRC-identified.deficiencyjis a-violation-(445/9003-V-01; 446/9003-V-01).
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~ (Closed) Open Item (455/8948-0-03):
This' item involved'the e..
apparent lack of a weld fit-up inspection corresponding to six fillet welds observed by the NRC inspector on
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- structural steel platform AB-206-01.
The six-joints in question.were specified-on design drawings'as requiring single bevel groove welds in lieu of the-fillet welds which were actually installed.
The' applicant concurredtwith this finding and-performed. ultrasonic tests <(UT) on'all
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accessible-full and partial penetration welds.' located on
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L the six Seismic Category I-platforms in. Unit-1.
'A? total'of
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L 46 welds were inspected'and, in each case, the required
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L penetration was achieved.
This' contrasted with-the'six.
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subject welds where the required penetration was not
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achieved.
On this basis, the applicant concluded that the
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observed discrepancies were isolated.-
The applicant supplied documentation showing that a fit-up inspection for these welds ~had taken place on October 1, 1980, and i'
admitted that this inspection was inadequate.
The applicant's position was that the inadequate fit-up L
inspection was isolated as evidenced by the results of the f
weld reinspections.
Platform AB-206-01 was evaluated for structural adequacy in light of the discrepant welds
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Therefore, the
applicant determined that no further corrective actions
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were necessary.
The NRC inspector reviewed documentation
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supporting the ultrasonic tests referenced above and i
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Calculation' Change Notice 002 to' Calculation l
16345-CS(B)-178, Revision 3.
The NRC inspector also
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I witnessed a demonstration of the ultrasonic' test technique used-to determine. weld penetrations.
Based on the fact that the discrepant welds'and poor fit-up inspections were L
'shown to be isolated:and that the affected platform was i
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structurally adequate in spite of the discrepant welds, the-NRC inspector concluded that the applicant hadltaken adequate action ~to resolve this item. ~This open item is f'
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(Closed) Unresolved Item (445/8965-U-04):
During-the NRC review of the applicant's room,' area, and system turnover.
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programs, several questions were. raised concerning the overall adequacy of these programs to: identify and correct hardware dicerepancies which remained after the completion of construction.
This unresolved item. tracked the NRC's
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continuing assessment of these programs.
Previous NRC
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inspection of this issue is documented in NRC Inspection Reports 50-445/89-65, 50-446/89-65;?50-445/89-76, 50-446/89-76;.and 50-445/89-89,iS0-446/89-89. ;NRC l
Inspection Report-50-445/89-89, 50-446/89-89. documents the NRC's final acceptance of the applicant's turnover
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programs.
All issues associated with this unresolved item were resolved in this previous NRC review.
Consequently,
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this unresolved item is closed, g.
(Open) open Item (445/8973-o-04):1 Following the AFW check'
valve failures (NRC Inspection Report 50-445/89-30; 50-446/89-30), the applicant developed an inspection and reassembly. procedure and post-installation test procedures l'
to demonstrate the operability of Borg-Warner check valves.
In several instances,-the post-installation backflow tests failed to meet the acceptance criteria, revealing l areas that had not been fully corrected by.the original:
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procedures.
This open item addressed the root cause analyses and generic implications of these l:
second-generation check valve failures.
A summary of the suspected root.cause and the corrective I
J action taken for each check valve failure is provided below:
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Valve 1AF-0083 (valve body / bonnet)-was rotatively.
misaligned and the disc-stud was1 bent.
.A new disc-stud
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L assembly was installed, the valve internals were
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. reinstalled, and the reverse flow-leak testing was i
satisfactory.
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Valve ICA-0016 exhibited excessive seat leakage.
The swing L
arm.and bushing were replaced and the valve was blue
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-6 checked. -The valve internals-were reinstalled and-the
- subsequent' reverse flow: leak testing was satisfactory.
Valve 1AF-0057 exhibited' unacceptable valve body / bonnet 1 rotational misalignment and. incorrect bonnet elevation.
'The valve was disassembled and supplemental measurements.
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were taken, the valve internals were reinstalled using the-
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new height specification, and the valve was successfully
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tested in-the reverse-flow direction.
Valve'1SW-0048 was determined-to have an-excessively long swing arm-bushing.
The bushing lengthLwas reduced by.0.08"
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and replaced'in the disc-stud assembly.-.Theivalve'
internals.were. reinstalled and the. valve'was successfullys tested in-the reverse flow direction.-
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valve IMS-143-was determined by radiography to have-the disk lodged underlthe seat, ring.
.The disk had apparently become' lodged under the seat during the reassembly process.
The valve didinot' experience forward flew after the reassembly process _ The valve was disassembled and-then
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reassembled taking care.tol ensure that the. disk did not.
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' lodge;under the seat.
Theoreverse flow (air). test was then successful.- - A reverse flow steam' test will be. conducte.d in-Mode'3.
In conjunction with the above documented activities, the
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applicant has revised the Borg-Warner check valve'
reassembly procedure and designed a specialized set of tools 1to allow-for the establishment of more precise rotational alignment of the bonnet.to the valve body.
The NRC' inspector witnessed a demonstration.of the new tools-and technique in the mechanical maintenance shop and the reassembly;of valve 1AF-045 in the plant.
The'NRC inspector concluded that the new procedure will enhance the rotational alignment between the valve bonnetJand body.
Approximately 13 Borg-Warner check valves in the auxiliary-feedwater and feedwater systems were' identified <by the applicant as having excessive body to bonnet external-leakage. 'These valves were disassembled, honed to-remove
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scratches in the valve body throat and provide better sealing surfaces, and reassembled..
In most. cases, this corrective' action' essentially stopped the leakage.
Several check' valves, including 1AF-038, which continue to' leak, are scheduled to be " hot torqued" in Mode 3.
The applicant anticipates that the extra pressure will seal the valve.
Each valve that war disassembled was retested for backleakage upon ;ntssembly with satisfactory results.
This open item will be left open pending successful Mode 3 testing of valve 1MS-143 and demonstration that the hot
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torquing referenced.above corrects the remaining. body to bonnet leakage problems'.
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(closed) open Item (445/8973-0-06):o This item addressed
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the apparent lack of adequate flushing capability in the auxiliary feedwaterL(AFW) system-using existing ~ drains.
This concern resulted from NRC interviews, conducted within the. Augmented Inspection Team (AIT)2 inspection documented in NRC Inspection Report:50-445/89-30; 50-446/89-30,- during.
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'which plantcpersonnel stated"that' check: valve internals
O were.routinelyLremoved to provide the appropriate drain-i-paths.
At the. time,.both NRCiand the applicant speculated
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that-the frequent. disassembly and: reassembly of check'
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valves-mayfhave contributed-to their' eventual failure.
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TheLapplicant's response to this issue is documented in
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TU Electric. memorandum CpSES-9001379, Davisito Guldemond.
This document presents the following points:
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The startup practice of using check valves for flush exit / entrance points is an industry accepted evolution.
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Check valve f ailure was: due to inadequate installation-procedures.in the Borg-Warner instruction manual and.
i was'not related to the frequency with which-these
procedures were used.
(3)
Additional drains and vents will be installed during-
.the Unit 1 first. refueling outage to facilitate-the
.i planned periodic inspections of Borg-Warner check valves, t
The NRC inspector agreed that the frequency with which check valves were used as drain and vent points was not a contributing cause of the AFW backflow eventn.
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- applicant's intent to install-new drains ad mnts and the fact that the-plant.is moving into the or
- tone phase should greatly lessen the need in the future to utilize check valves in this manner.
This open item is closed.
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(closed) open Item (445/8973-0-07):
This item identified the NRC's concern that no apparent provisions were made for continued maintenance and system preservation for the.AFW-system during the period from completion of.preoperational.
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testing in'1984-until completion of hot functional testing in 1989.
This perception was based on NRC reviews of maintenance histories and discussions with personnel during the AIT inspection documented in NRC Inspection Report 50-445/89-30; 50-446/89-30.
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at f" 2 The-applicant stated that maintenance and preservation;of
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.the.AFW systemiduring?this. time period was controlled _by.
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Procedure MDA-301, " Protective Maintenance Program," and
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' Procedure-MEI-043, " Performance of Activities 1 Required by
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H ANSI N45.2.2."
Procedure.MEI-043 applies to equipment-
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The applicant
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_provided a' list of work orders _on the'AFW system covering
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late 1985 to-late 1989 which included;some: preservation
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activities such as oil changes, filter examinations, inspection of bearings, " major" inspections, and " teardown"
inepections.
The applicant stated'that:the AFW system was-
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'in wetLlay-up with adequate concentration of hydrazine:to
' prevent corrosion.until December 1986 when the system was placed in-' dry lay-up...Hydrazine was also used in' dry.
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lay-up: for those ~ areas which could noi-' be ' dr".in'ed.
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The NRC inspector _ reviewed Procedures MDA-301'and MEI-043 and information regarding lay-up conditions of the AFW.
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It appears that maintenance and_ preservation of
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the AFW system, though not extensive, was adequate to-
. item is closed.-
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This open-J ensure the continued operability ~of the' system.
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.(Closed) Open Item (445/8973'-O-08):
During;the.auxiliar'i
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-feedwater (AFW) backflow' events'(see NRC Inspection Report:
50-445/89-30) 50-446/89-30), steam generator water. flowed Jin the reverse-direction through the feedwater isolation bypass = valves and in the forward directi'n through the o
preheater bypass valvesitotthe AFW piping.
The applicant informed the NRC of.their-intent to-administratively isolate the feedsater isolation bypass valves during startup and shutdown-conditions to preclude.the possibility-t for similar backflow events in the future.
The. applicant-
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has' revised Procedure IPO-004A (Revision-3), " Plant L
Shutdown from Minimum; Load to Hot Standby," and Procedure L
IPO-002A (Revision 4),'" Plant Startup from Hot Standby to Minimum Load," to require the feedwater isolation bypass valve downstream' manual isolation valves-to remain closed H
whenever.the AFW system is being used to' feed the steam n
generators.
On startup these manual valves are opened upon.
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transfer from the AFW system to the-main feedwater system, and.on shutdown the valves are closed on transfer back to
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the AFW system.
.If operators adhere to-these L.
administrative controls, backflow events similar to those experienced on April 23 and May 5, 1989, should not recur.
As a backup, the applicant has also revised Proceduras IPO-004A and IPO-002A to require closure of the preheater
' bypass valves whenever the AFW system is providing feedwater to the steam generators.
In order to effect this change-, the applicant.had to modify the interlock between l
the: preheater_ bypass valves and the feedwater isolation
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Design.Chanhe" Authorization (DCA)-92571.wasissued
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to reconfigure contacts to; permit.the preheater bypass
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valves:to: remain closed when its control. switch is in the
' closed' position regardless of the position of the feedwater q
isolation valves.' The inter 1cck'between these two valves s'
is restored when the:preheater hypass valve control is
returned to " AUTO."
The1preheater bypass valves.will provide a redundant pressure: boundary to prevent backflow
"
from the steam: generators to1the AFW system.
q
,,
~
L The NRC; inspector reviewed-the revisions to Procedures
,
q IPO-004A and IPO-002A, DCA 92571', and relevant changes made.
H y
to DBD-ME-203, "Feedwater System," andiconcluded that the
'!
.y
. applicant. has taken sufficient action on.this iitem.
This-p open-item isiclosed.
k.
(Closed),Open Item (445/8973-0-10):
This item addressed-
,
p the applicant's evaluationaof the human factors associated'
,
L with remote valve operators.. Valves.1AF-041 and 1AF-054 j
b (AFW pumptdischarge~ isolation valves), due to the-difficulty of operating their reach-rod valve-operators,.
q indirectly contributed to the AFW backflow events reported
'
in.NRC1 Inspection Report :50-445/89-30; 50-446/89-30.
These-valves required approximately 30 minutes to close from full?
L.
open or to: open from full closed..The applicant. conducted a plant walkdown.to locate-and evaluate all valves operatedi
'!
. ith reach-rod operators.
In Unit 1 and common, 398 valves L
w-were checked, of which 190 were safety related.
Each valve was checked'forLlabeling, stroke time, ease of, operation,
-
number of~ turns per stroke, accessibility, and direction of-operation. 'Each valve checked was determined to be operable and-the eight safety-related valves which could not be operated.(due to-plant conditions) were judged to be=
operable based'on comparison with similar valves.
Howeve.r,-
40' valves were classified as " difficult to operate"1due j.
-
mainly to long stroke times or difficulty in turning the valve operator.
To date, the-applicant has modified.only l
one valve, 1AF-041 (see-paragraph'l), reducing the gear ratio and the time to operate from 30 minutes to 2 minutes.
The applicant intends to modify'the other two ArW pump discharge isolation valves (1AF-054 and 1AF-066) during the
'
first refueling outage and will schedule other valve modifications on a case-by-case basis.
A list of
,
difficult-to-operate valves har, been included in ProcedureL OWI-206, " Guidelines for Operation of Manual and Power
Operated Valves," to alert operators and control room-personnel to the schedule and manpower requirements
'
associated with these valves.
The NRC' inspector reviewed data sheets from the plant walkdown, the revisions made to Procedure OWI-206, and a summary of the applicant's actions on this issue documented
'
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in memorandum CPSES-90001405.
The NRC' inspector concluded that the applicant has-taken adequate action to address-this issue.
This open item'is closed.
1.
(closed)'Open Item (445/897?-O-12):
This item addressed the applicant's ections to make valves 1AF-041 and 1AF-054-easier to operate.
During NRC investigation of the April 23 and May 5, 1989, AFW events involvina,the failure.
of several Borg-Warner check valves, the NRC determined that the difficulty of operationLof these two valves was a_
contributing cause.
For. valve 1AF-041,~the applicant issued DCA 91717,
.
. Revision 1, to modify the existing 24:11 ratio manual gear-
-
operator to a 6:1 ratio operator.
This reduces the-number of turns required to open the valve from approximately 404 turns to 89l turns. 'The valve rim-pull 11s still1within
-
a-the specification limit of ed ft/lbs.
This work was completed via Work: Order C890015384.
Design Modification (DM)89-403 requires-the, reduction of the operator gear.
ratio for valve 1AF-054 (as'well as valve'1AF-066).
-Valve 1AF-054 currently has'a gear ratio of 18:1 and the
' difficulty of operation is not as great'as that for valve:1AF-041.
In addition, the applicant has developed an operator aid which contains information for. operations personnel on the difficulty.and length ofl time required-to operate each valve (see the closure of 445/8973-0-11, NRC Inspection--Report 50-445/89-88; 50-445/89-88).
Based on'
the above applicant actions, this item is closed.
m.
(Closed) Open Item (445/8973-C-13):
This item addressed theLapplicant's review of check. valve min / max axial gap (play) criteria developed by Borg-Warner in response to check. valve failures associated with the AFW-backflow events discussed in NRC (AIT) Inspection Report 50-445/89-30; 50-446/89-30.
Early in the investigation of'
the check valve' failures, axial gap was thought to have been a-significant contributor to the failure mechanism.
Later'research established valve bonnet height;as the primary cause'with axial gap as a less important, secondary factor.
The applicant has completed review of Borg-Warner's axial gap criteria and has incorporated these values (with some conservative changes) into Procedure MSM-CO-8801,
"Borg-Warner Check Valve Maintenance," Revision 2.
Some of the Borg-Warner check valves currently installed have axial gaps outside the envelopes specified in Procedure MSM-CO-8801.
Each of these valves have individual calculations verifying that the axial gap will not affect operability of the valve.
Nonconformance Report i
(NCR) 89-7476~ documents the axial gap range of the s
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Lcurrently installed valves-and functions (along with the:.
calculations) as a use-as-is disposition where gap length
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does not conform to Procedure MSM-CO-8801.' The. applicant stated that any future modifications'tocthe check valves ~
would likely; involve' complete: replacement ~of the
-
.
bonnet-swing arm assembly at which time-the axial gap criteria of Procedure MSM-CO-8801 would be fully
'
incorporated.
The'NRC. inspector determined that the applicant has
established adequate control of the axial gap-dimension and that the operability-of check valves with axial gaps i
outside the procedural envelope is adequately assured by.
both calculation and functional, backflow tests.. This open
' item is.closca.
,
n.
(Closed)'open Item (445/8973-0-14):
Training to' increase operator awareness.
As previously documented in NRC
~
Inspection -Report -50-445/89-30 r 50-446/89-30, this item was I
identified during the NRC'AIT evaluation of multiple check valve failurestin'the AFW system experienced during hot functional testing.
In particular,.the AFW'backleakage events reflected negatively on the quality of training.
received-by the plant operations staff..The necessity of-
,
performing in-sequence valve operation was apparently not
,
adequately emphasized.
A second. training-related concern d
wasnidentified in that the failure of operations personnel.
J to document the discovery of three failed AFW check valves 1on a Plant Incident Report (PIR) or on an NCR.'
In response to these issues the applicant committed to enhancing the awareness of plant operations personnel to
,
operability issues by conducting additional training in this area.
This additional training encompassed the'
a following: elements:
(1) an operations management and
,
senior reactor operator workshop,.(2)fauxiliary operator requalifying course (" Plant Incident Reports"), and
(3) auxiliary operator requalifying course ("Recent Plant Incidents").
+
The NRC inspector reviewed course outlines, lesson plans, andEattendance verification records for the three training sessions referenced above and concluded that the applicant's retraining.cffort has fully addressed the
personnel issues associated with the AFW backflow events.-
This open item is closed.
'
o.
(Closed)'open Item (445/8973-0-15):
This item addressed service life degradation of the AFW minimum flow
- .
recirculation check ralves (1AF-045, -057, and -069).due to turbulent flow conditions resulting from proximity to
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breakdown'fiow orifices.
This issue was raised during the
,
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AIT' inspection (NRC Inspection Report 50-445/89-30; 50-446/89-30) in' association with NRC review of the-
..
applicant's' action to-address the failure of valve 1AF-069 which occurred on April 5, 1989.
The failure of this valve l
'
was probably the result of bonnet height elevation
!
discrepancies through flow turbulence downstream =of the
>
orifice causing the valve disk to slam repeatedly against s
,
the stop may have been a contributing cause.
s
.. p
'
At the time of the AFW backflow-events, the applicant's.
p
. consultant, Kalsi Engineering, Inc., was' performing a comprehensive review of safety-related check' valves in
-
,
response to Significant Operating: Event Report:.
,
(SOER)~86-03, " Check Valve Failure:or Degradation."
.
Kalsi's final report, "SOER 86-03: Check l Valve. App 31 cation E'
Review," dated November 30,.1989,- recommended (fc; valves 1AF-045, -057, and -069) the' replacement of the l
existing 3/8" x-5/8" (step) disk studs with 5/8" (straight)
.!
disk ~ studs to reduce the probability of disk stud fatigue
failure.
The applicant adopted this' recommendation in
,
design modification (DM)-89-?16 and Design Change Notico
. !
(DCN)-000103... The! disk stud % were modified under work orders:C890014336, C890014469, and C890014470 for
.
,
't
' valves lAF-045, -057, and -069 respectively.
All three
'
valves: subsequently passed backflow-tests' conducted in
- accordance with Procedure EGT-328A, " Reverse Flow
..
~
!
!
operability Testing for Auxiliary Feedwater Check Valves."
p
,
The NRC inspector reviewediall of the documentation L
referenced above and concluded that, for at least the'short l'
term,,the disk stud modification was a viableLalternative-to' increasing the distance between the orifice and the check valve.
The applicant 1 plans to inspect the condition
l ofithe AFW minimum flowLrecirculation check valves during.
L the first refueling outage and. plans at some future date to'
b relocate the check valves.. This open item is closed.-
,
p.
(Closed) Open Item (445/8973-0-09):
During a previous NRC-inspection-of theLbackflow events in the AFW system piping,
.;
the NRC had concern 3 relating to high stresses in an cibow
West of support No. AF-1-096-023-S33R and in two instrumentation connections.
These items of concern were
<
in-the pipe evaluated in Stress Problem 1-100 and L.
determined to have been highly' stressed during the events.
.;
fl During this-' inspection period, the inspector reviewed the-
analyses forLStress Problem 1-10C documented in C
Attachment 9 to Calculation 15454-NP(S)-GENX 343.
This attachment documents the results of thermal expansion stress evaluations in accordance with ASME Code
- q Section III, Class 2 and 3 criteria (except that ASME Code stress allowables were not used).
The evaluations showed
'
that:
(1) the highest stresses due to thermal expansion
.
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effects-(97.41 ksi),z and to the-combined effects of:
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- pressure, weight,'and thermal expansion ~(104.0 ksi) during J
the events were: attained in the subject elbow; and (2)'the W
highest stress in the piping in the vicinity of..the subject two instrumentation connections due to thermal expansion
'
.only was 47.0 ksi, and to'the combination of-sustained.
loads'and thermal expansion was-52.9~ksi which exceeded the
,
ASME Code allowable stresses.
In-addition,1 stresses in
several other. locations in the piping due'to thermal o
.
expansion onlyJand the combination of sustained loads.and i
f thermal expansion exceededtthe' ASME Code allowable l
stresses.
Subsequently, the second event was reevaluated-
'
' -
to account for as-built gaps in four supports'in the l-vicinity'of-the piping. adjacent to the subject-
.i t
instrumentation connections.
The reevaluation demonstrated that the highest stress in this piping due to thermal.
!
expansion only was reduced to 8.0 ksi and to the
' combination of sustained loads and thermal expansion to j
13.8 ksi which were less-than the-ASME Code stress
'
allowables.
Stresses in.the piping, including the highly-stressed subject elbow, remote from the supports where n
>
p as-built gaps were included in the analysis,'were.
?
unaffected in this reevaluation.
.
Subsequently, TU Electric performed radiographic ar.d n
ultrasonic inspections of areas in the piping, incladh,
.,
the pipjng in the; vicinity of the' subject instrumentation-connections and elbow, and verified _that no damage had been L
incurred during the events and the ASME < Code: minittum
.
'
L wall-thickness: requirement'not violated during the events.
. Based on the preceding inspectioncresults, the-' inspector
-
found-that the TU Electric evaluations and'inspectionsL L
described'in'the preceding.were sufficient to resolve the
)
~
previous NRC concerns.-
Although the'ASME Code allowable:
'
L stresses were exceeded during the events, measures are-L
'being instituted.by TU Electric to prevent reoccurrence of b
backflow in the AFW piping system thereby limiting future'
L stresses in the piping system-to no more than in their L
design.
Consequently, given that the number of' load. cycles during which-some areas of.the AFW piping systems have been
exposed to the high stresses experienced during the events'
are few-(no more than two) and no damage was found in t.hese,
i areas, the NRC' inspector determined that the APW ' piping.
f
. system is adequate to serve its-intended function durjng
plant life.
This open item is closed.
<
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(Closed) Open Item (445/8975-0-01):
As part of the evaluation of the impact on the integrity of the affected-piping system, pipe supports, containment penetrations, and instrumentation due to the events of April 23 and May 5,
'
F 1989, events involving backflow through the AFW system, I
,
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-.
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.
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.
fU Electric performed analyses ~to identify areas ^in the l
piping system and pipe'supportLwhere ASME Code stress
'
allowables had been exceeded.
These areas were subsequently ~ inspected for damage.and~ components found:to
.
Ebe damaged were replaced.
The NRC had determined.that-the
!
'
TU Electric evaluation was deficient since:
(1)' areas of'
the AFW piping system which could have been damaged dus~toi
!
thermal' transient effects had not'been. satisfactorily
' identified and insp'ected - the TU Electric stress: analysis s
was. based on-steady state temperature distributions only;
. and-(2) the effects of greater than designLthermal
,
'
movements on branch piping and1 instrument-loops had'not, g
been evaluated satisfactorily.
Subsequently, TU Electric-i performed additional evaluations and inspections in i
response ~to these NRC identified deficiencies.-
l With respect to thermal transient effects, TU Electric-conducted additional evaluations and inspections and
,
.
verified that no damage had been incurred due to-these
"
.
effects during the backflow events.
Results of these evaluations ~were documented in Attachment 38, " Thermal a
Shock Evaluation," to Calculation 23454-NP(S)-GENX-343,-
~.
g i
"Evaluat' ion of Piping and' Pipe Supports Due~to'Reverso Flow-
Incidents During HFT," Revision 0, dated September'5, 1989.
.
This attachment indicates that thermal transient-effects
.
were evaluated only at' structural discontinuities..
-
+
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'.
. Material discontinuities were not considered since all
' piping,affected was' constructed-from carbon steel material
only.
In addition:
(1) pipigg which experienced
.
.
J temperatures of'less than 250 F, and (2) dead-ended piping
,
- with discontinuities at distances at least (dt)1/2 (where=
d =spipe diameter,_t = pipe wall thickness) from high-temperature fluid' flows'were excluded from the evaluation.
E E
p~
'The NRC inspector concurred with these exclusions-since:-
'
- (1)evaluationsofdiscontinugtiesinpipingexperiencing-
,
. temperatures greater than 250 F will bogndi hose in piping t
a L
'
n experiencing temperatures less than 250 F, andD(2)'the.ASME l
Code requires that effeccs due to nearby discontinuitieu be L
consideredJonly if they are at distances less than-2.5.(rt)1/2 (where r = pipe mean radius),
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>
b The remaining structural discontinuities were identified.
u and evaluated-in. Attachment 38.1 to Calculation-GENXL343.
p?
'This attachment contains tabulations of the discontinuities
~ defined in the.13 stress problems for the AFW piping
"
-
.;
involved in=the' backflow events.
The BRP-series piping j
drawings.for the stress problems were also identified in
!
the attachment.
TheLinspector reviewed the tabulations for
'
""
l 1-008, 1-152, and 1-169;and found them to Stress' Problems".All'the discontinuities were properly
l
be acceptable.
-
identified.and: included.
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All 'of the structural discontinuities identified were characterized by_providing the. appropriate." main run" and
" branch / valve"-data.
These data included pipe ~ diameters,_
pipe' wall: thicknesses?and conservative average temperatures
i achieved during the events.
In addition,cthe wall'
i or.t /t, (where t, and ty are the thickness ratios t,/tb b
,
main run and branch / valve wall' thickness,1respectively)'
"
were provided (the ratios.were selected such that the
- '
'
tabulated values were always less than or equal to l.0)fand
'
the disposition to inspect or not inspect the discontinuity.
' documented.
of the 119 structural discontinuities
evaluated, 12 were selected for NDE inspections.
The NRC inspector found that the TU Electriccevaluations were acceptable. -Although stresses in the structural discontinuities were not explicitly calculated, the
,
evaluations were sufficient to identify and provide a
'
relative ranking of the susceptibility of the discontinuities-to be damaged by thermal transients.
In
the absence of material discontinuities, the ASME Code criteria for therm &l transient stresses in structural discontinuities are based on LlT and Abs (T -T ) effects
3 b
(where flT := through-wall. temperature gradient and y
Abs (T,-T ) = difference in average temperatures in adjacent b
areas of the discontinuity).
In_ general, for a given transient 41T increases with increasing wall thickness and
-
y p
l Abs (T -T ) increases with increasing differences 11n wall a
b l.
thickness of adjacent areas.
Accordingly, in the
TU Electric evaluation,-the greater the wall thickness in
the' discontinuities evaluated and the smaller the t,/tb
'
t /t,; ratios, the higher will be the thernal transient j
b
'
-stresses.
s L
The-12 discontinuities selected by TU Electric for Ls inspection were characterized as having been exposed to the-Inaximum temperature difference' during the transient (from ambient to 557F) and ranged in diameter size from 3 inches
.i p<
to 18l inches and adjacent wall-thickness ratios of
betweer. 0.1 and 0.7.
In addition, configurations selected L
included weldolets, sockolets, valves, flanges, and.
,
(; '
. penetrations.
The inspector found that the discontinuities
-
/J selected for inspection were sufficient in number and type and appropriate to determine whether or not the affected-
"
piping had been damaged due to the transient effects.
The inspector also reviewed NDE inspection records for all but one of the 12 discontinuities inspected and verified that the results were satisfactory.
No unacceptable indications
.
p.
were found.
.
Based on the preceding results of inspection, the inspector O
[i found that TU Electric has satisfactorily identified and inspected areas.in-the AFW piping system which could have
'
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Lbeen' damaged by thermal transient effects during the AFW, backflow 'eventsland verif 4.ed that no damage. was incurred.:
.
lWith respect to greater than design-thermal movements'for'
-
branch piping and instrument loops, thermal movements
'
c1 during.the events were evaluated by TU Electric and g
determined to be' acceptable.
Movements during.the events
,
'wer'e" compared with previously calculated displacements used1 to qualify the branch piping and instrument loops.
,.
Movements equal to or less than previously calculated
"
displacements were. considered acceptable.
Movements-
,
y greater than previously calculated displacement werez
.
evaluated to assess-their acceptability.
The evaluation verified that.the increased stresses were less than. piping
and loop.ASME Code,Section III, Class 2 and 3 allowable stresses.
These TU Electric evaluations were documented-in
'
Attachment-37 to' Calculation 15454-NP(S)-GENX 343.
The inspector found that these TU Electric evaluations were
.
acceptable.. Evaluations for 7 branch 1 piping and
,
30 instrument loops were documented. 'Thormal movementsf during the event for six of the seven branch piping were
.'
-less than the displacements used for their qualification.
For the seventh, the total thermal movement: (0.058' inches)
'
exceeded the qualification movement (0.022 inches) by
"
0.036-inches.
The TU Electric evaluation'noted that since-
the stresses in the branch piping qualification analysis
<
were low, the movements were acceptable.
Based on the small magnitude of the movements involved 1and the ample
-
stress margins.(0.27 ksi'versus 17.50 ksi and 5.22 ksi versus-37.50 ksi) available in the. qualification' analysis, the inspector found that the results of the evaluation'were acceptable.
For the~30. instrument l loops evaluated, the
movements /during the events for:23 were.less than the
.
"
displacements used for their qualification and 1 contained a flexible hose.
Movements greater than the qualification
'
displacements in the remaining.six loops were evaluated on the basis,of applicable flexibility criteria or stress margin considerations.
The inspector found the methods and'
,
.results of these evaluations to be acceptable.
g lL'1 Based on the preceding, the inspector found that.
,
p'
TU. Electric-has satisfactorily evaluated the effects'of R
greater than design thermal movements experienced during
,
L
- the AFW system: backflow event-by-branch: piping and
.
instrument loops and determined that these components were E
not adversely affected during the events.
This open item i
G-is closed.
~
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[
r.
(closed) open Item (445/8978-0-04):
The results of a previous NRC review of the TU Electric Train C Two Inch Diameter-and Smaller' Conduit System Maintenance Program s
,
L which was to be' performed as part of their Corrective
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Action Program (CAPk were reported iniNRC Inspection Report
-
50-445/89-78; 50-446/89-78. -This NRC review'had found-3-
tthat:
(1) the: program had been terminated after only
'
approximately'50, percent,of all applicable supports 11ocated:
in approximatelyi60 percent of all-the applicable rooms-in
-
the plant had.been evaluated;iand (2) subsequent
'e responsibility.for the maintenanceiprogram was to be:
,
transferred to the Systems Interaction-Group.
TU Electric
.[
'-
1etter NE-27883 (previously, referenced as DE.27883)
e -
identified the~ rooms'that had:not been evaluated under the-
~
program and advised that if; safety-related components were L!
added in the future to-these rooms, evaluations of all L
Train C less than or equal to 2-inch. diameter that were within the zone of influence:of the;new safety-related
.
'
components would have to be evaluated for unacceptable interactions.
This letter had-defined the applicable procedures:as CPE-EAP-CS-018, which was the procedure
"
applicable to the maintenance program up;to the time of its termi' nation.-
- ,
-During the review, the inspector'had requested TU Electric
!
o to provide the procedures which would control the future-
!:
installation.of safety-related equipment'to ensure that
,
L provisions for performing evaluations.in accordance with-the~ requirements of CPE-EAP-CS-018 (if the equipment was to
'
o t
'be installed in a-room not previously included in the.
truncated maintenance program) had-been adequately
"
addressed.
TU Electric had advised that the-requirements
to such future evaluations were in the process of being-added to the next revision of~ Procedure ECE-2.24.
-
During.this. review period, TU Electric provided the
.
inspector with three procedures which had been revtsed and.
issued to assure:that the maintenancefprogram evaluations-l will beLperformed'for future equipment installation.- The
'
procedures'were EME 2.24-01,.EME 2.24-13, and DEO-DEO-EAP-CS-041 which were more appropriate for-future
'
maintenance program evaluations than Procedure ECE 2.24 previously identified in NRC Inspection Report 50-445/89-78; 50-446/89-78.
The inspector reviewed these
procedures and found the following:
'
(1)
EME.2.24-01, " Evaluation of Seismic /Non-Seismic Interaction," Revision 2, dated April 4, 1988:
This procedure establishes the methodology for evaluating and. resolving " seismic /non-seismic" interactions.
,
Engineering Document Change Notice (EDCN)-04 dated-January 5, 1990, revised this procedure to require that: -(1) field walkdowns be performed in accordance with Procedure CPE-EB-FVM-SI-40 or EME-2.24-13.to identify. potential non-nuclear safety /nonseismic sources.and. seismic nonseismic interactions
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l (Section"6.1.1); and-(2) unacceptable' interactions due
!
to 2-inch'and smaller diameter Train'C conduits and
' '
their supports be resolved'by performing: analyses in i
accordance with the appropriate Design Procedure'
'
ECS-511 (Section 6.2.5).,
.(2) -EME 2.24-13, " Systems Interaction Program
,
Maintenance," Revision 0, dated June 8, 1989:
This.
+
v procedure establishes-the methodology for the performance and update of Systems Interaction Program.
,'
.walkdowns' required to reflect the as-built configuration.
EDCN-01 dated January-5, 1990, revised
"
.
Lthis procedure tx) require-that:
(1)'2-inch and-
!
smaller. diameter Train C conduit systems be considered
.
at sources (Section 4.6.3); and (2) maintenance
[
'
' activities (including source identification, walkdown,
interaction evaluation,'and documentation)-for these
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conduit systems were to be in'accordance with
. Procedure DEO-DEO-EAP-CS-041 (Section 6.7.1.b).
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DEO-DEO-EAP-CS-041, " Train C Two Inch Diameter and
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Smaller Conduit System Maintenance," Revision 0, dated-
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January 5, 1990:
This procedure 11s based ~on the requirements of Procedure CPE-CPE-EAP-CS-018 used
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previously.in the truncated maintenance: program.
This procedure requires, in part, that:
(1) if.new
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safety-related components are added in the future in L
any of the 164 rooms previously identified'" unique" L
supports in the conduit: system need to'be_ evaluated; (2) if new' safety-related components are added in.any
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ofsthe 118 rooms not_previously evaluated, all:
conduits.(and their supports).within the influence
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zone of the~ components are to be evaluated for'
L seismic /nonseismic: interactions.
Rooms previously L
evaluated and~not evaluated were identified in
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Tables 6 and~7, respectively, of the. procedure.-
In addition, evaluations were to be in accordance=with
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the requirements specifded in the procedure..
L, Based on the preceding results of this review, the NRC inspector found that measures taken by TU Electric to assure that the maintenance program,for the two inch and ci smaller diameter Train C' conduit systems will be
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satisfactorily-implemented when new safety-related components are added to the plant.
This open item is closed.
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(Closed) Open Item (445/8978-0-05):
This open item was r
issued pending NRC review of the applicant's actions regarding certain defective tube steel.
Specifically the applicant identified a crack-like indication in a piece of
4 x-4 x 3/8-inch tube steel.
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The. applicant's. actions.regardingathe tube steel were.
-performed.under Significant Deficiency Analysis Report
.(SDAR) CP-89-27.
The NRC review of.those actions has.been
.performediand'was reported in-NRC Inspection' Report-
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50-445/89-89; 50-446/89-89.' ~Inithat report, the NRC
inspector determined that the applicant's actions regarding
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Accordingly, this open item is closed.
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3.
Follow-up on Violations / Deviations (92702)
a.
(Closed) Deviation (445/8516-D-30):
Thin deviation-concerned four instances where field reinspections by CPRT personnel were not in accordance with approved instructions.
Specifically, CPRT reinspections either did not inspect or did-not identify:
(1) an incorrectly marked
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identification tag on valve No. 1-RC-8061B, (2) a less than required air. gap for one location each in Verification Packages 1-E-ININ-049 and -059, (3) the adequacy of bend
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radii for~ instrument tubing vent or drain-lines in.
Verification Packages 1-E-ININ-041 and -051,,and (4) an-improper hookup between the low pressure sensing line attachment.and-a differential pressure type instrument in Verification Package 1-E-ININ-049.
_g The applicant evaluated this deviation and determined.that.
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certain corrective and preventive. actions were appropriate.
For items 2, 3, and 4, the corrective actions concisted of amending the verification packages as applicable or issuing an out-of-scope observation-to: document the deficiency.
'For item-1, the applicant identified and corrected the
't improperly marked tag due to a different program,
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accordingly no further corrective action for item 1 was deemed necessary.
Due to a generic reanalysis of the adequacy of inspection instructions,-each'ERC. Quality
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Instruction was reviewed for clarity in describing both what was to be inspected and what was-the accept / reject criteria.
Additionally, an overview Inspection Program was i
implemented by CPRT to monitor the adequacy of
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reinspections.
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The NRC inspector reviewed both the corrective and preventive actions and determined the actions were acceptable..Accordingly, this deviation is closed.
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b.
(Closed) Violation (445/8518-V-ll):
This violation concerned the failure to maintain conduit and cable tray
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identification for thermolagged cable tray and conduit raceways.
The applicant evaluated this violation and determined the root cause to be a failure of building management y
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.(construction)'to understand that maintenance of cable tray and conduit identification was required after the
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installation of the thermolag.
The misunderstanding appears to have been caused by a-_ failure to' properly-incorporate the requirement into procedures.
The applicant _ documented the deficient-cable tray'and conduits on nonconformance reports (NCRs) with the-intention of performing walkdowns-and reinspections to q,
ensure proper identification oflthe.thermolagged raceways'.
j However, due to1 subsequent engineering evaluations which-
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indicated that the thermolag could adversely impact the
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design allowables for ampacity:and loading, the majority of-installed thermolag was removed.
Additionally, the y
remaining installed thermolag was later removed and
replaced due to a determination that the installed
J's thermolag had variations in thickness-and, therefore, it's-
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adequacy was_ indeterminate.
The reinstallation of l'
thermolag.was performed to revised = construction and'
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-inspection' procedures which clearly-required that cable tray and conduit identification be maintained by appropriate marking on the outside of the thermolag.
The NRC inspector reviewed the. revised construction ~and
inspection procedures and determined that they properly incorporated the design requirement to identify..
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thermolagged raceways.'
Further,-NRC tours of those areas
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- where:the.thermolag had been removed-and replaced, found the raceways to be properly identified.-
l Since the'thermolagged raceways are now properly identified j
and revision of the: applicable construction and inspection j
is deemed adequate'to preclude recurrence, this violation
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is closed..
c.
'(Closed)' Violation (445/8847-V-01, Items a, b, and d):
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This violation concerned the failure of TU Electric to
adequately implement the requirements of Appendix.BLto 10 CFR Part 50.
The failure occurred during the.
_procuremont, implementation, and follow-up of vendor
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.i" services to remove plasite coating from the station service water system (SSWS) piping.
Item'a, " Failure to Establish QA/QC and Technical
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Requirements in Procurement Documents."
The applicant.
l cvaluated the implementation of the SSWS coating L
removal and,-as a result, identified the need'for
enhancements to the site procurement process for l-Code
"V" service procurements.
Those enhancements are identified in Engineering Report ER-ME-19 and have been incorporated into site procurement procedures.
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The applicantiperformed a review of all' previous
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TU: Electric Code
"V" service procurements and!
-determined-that the detai1Jof_TU Electric QAL involvement'specified=in_the procurement documents varied. =However,.all=of the_procurements. required
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GN Electric QA to monitor the vendor's activities.
TU Electric's: review of the associated inspection' and
survaillance reports determined that the requisitioned i
services were successfully completed and documented.
j Although the procurement for the SSWS coating removal originally failed to recognize the necessity for
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compliance with the controls imposed by.CPSES ASME'
Section XI procedures, those controls were imposed during subsequent SSWS corrective actions, such as the
replacement of damaged sections of piping..
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Additionally, the applicant determined that the
failure to reference applicable source documents-(NCRs, and DCAs) did not contribute to the other problems that occurred with the SSWS coating removal process.
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The NRC inspector has reviewed the program
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enhancements referenced in~ER-ME-19 andithose
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incorporated into current procurement procedures.
l These enhancements appear to provide the necessary
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L guidance-to ensure the inclusion of appropriate technical and QA requirements into future Code "V"
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service procurements.
Further, NRC review of the.
.other Code
"V" service procurements verified that
.TU Electric QA-was' required to monitor the contractor's activities and;that the processes-involved,-such as chemical cleaning,.did+not require prequalification testing.
The NRC-inspector concurred
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that-the other Code
"V" service procurements did not exhibit the same deficiencies that were. identified in L
the procurement ~for the SSWS coating-removal.
However, since the other service requisitions. lacked a detailed specification of QA involvement,;the.NRC inspector requested additional information-regarding-l the adequacy.of QA overview for those services.
The'
L applicant provided a document, "QA Department Review L
of Six Code V Services Procurements.'" -That document
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l provided an in-depth review of.the adequacy of QA involvement during the procurement, implementation, and follow-up of the six Code V services.
The NRC inspector reviewed this document and determined that it accurately identified certain weaknesses-in the specif;. cation of the details for QA involvement and vendor interaction with site QA.
However, y
documentation of work completion and QA overview appeared acceptable for the services performed.
The l
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. document concurred'with ER-ME-19 that program U
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enhancements.wereinecessary.
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The NRC inspector,has. reviewed the applicant's actions
to' enhance the program and determined that, since the applicant has revised-the procurement program to
require Code
"V" service procurements to provide a
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more detailed delineation of technical and QA-
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requirements in procurement documents'and since.
corrective actions taken for~ rework of the damaged-y portionslof the SSWS are complete with proper technical and QA requirements implemented, item a of-this violation is closed.
Item b, " Failure to Control Special' Processes."
This
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item identified the applicant's failure to control the SSWS ciating removal as a special process.
The applicant evaluated this item and determined that, as
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noted for Item a, the failure to clearly define the relationship between the contractor's activities, the
'i TU Electric QA program, and other affected TU~ Electric organizations was mainly responsible-for this failure.
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The contractor's procedure for the removal of the-SSWS-
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coating was reviewed and approved by TU Electric
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engineering.
However, contrary to site requirements, TU Electric QA was not involved in the approval
- process.
q As stated in Item a, the NRC inspector has reviewed the applicant's procurement program enhancements and concurs that they now require a clear definition.of
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the-relationship'between the contractor's= activities and the TU Electric QA Program.
Further, the NRC
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L inspector reviewed the revisions =to1the applicable
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h construction' department procedure whichJrequire that F
construction-contractor work procedures he' approved as B
construction department procedures prior.to use.
The NRC inspector concurs that those procedure revisions
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and program enhancements incorporated to ensure =the
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involvement of both TU Electric QA and other affected r
TU Electric organizations in the review and. approval L
of_the contractor's procedures for unusual or special-
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processes are adequate.
The NRC inspector notesLthat
. an underlying cause of the failure to properly control n
the spinblasting process was a slowness on L
- TU Electric's part to recognize the spinblasting as a Ly special process.
Therefore, the NRC inspector discussed with applicant personnel what actions had
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been taken to increase the sensitivity and awareness of procurement personnel for this type of issue.
The applicant's personnel explained that the issue is
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included in-the required; training for procurement
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engineers and that the detail of figure 7.4 to.
Procedure MMO 6.02
.02,-" Procurement Engineering
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Review:and Processing of Procurement Documents",
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entitled " Services Review Summary," had-been increased-l
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tx) ensure that such' issues would be considered.- The
.NRC_ inspector reviewed figure.7.~4 and considered it to l
befadequate. 'Accordingly, item b of this violation is closed.
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Item d, " Failure to Identify and Correct Conditions
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Adverse to Quality."
The applicant failed to identify.
.l and correct certain conditions adverse to quality that i
t occurred during the coating removal on the SSWS.
This i
failure occurred even though the NRC inspector had
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identified concerns for the coating removal process.
Subsequently, it was determined that the coating
removal-process had resulted in several areas of wall thinning and one hole, approximately 1/2-inch in
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- diameter, being blasted in the service water piping..
Corrective actions for this violation included M
suspension of the remote sandblasting process until-
modifications ta) the process were developed to resolve the problems.- Further programmatic changes to R
procurement procedures were added to better define the
responsibility of the contractor and TU Electric organizations to document and resolve nonconforming-o conditions.- The corrective' actions for the remote sandblasting-process were implemented during the coating removal process for the Unit 2.SSWS'which
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. resulted in no significant damage to the piping.other than one case due to equipment failure.
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As stated above, in. items a and b, the NRC inspector reviewed'the revisions to the procurement procedures
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and deemed them to be appropriate.
Further, q
corrective: actions for the damaged portions of piping-have been implemented.
The applicant disagrees with a R
part of this violation; i.e.,-failure to respond to-l NRC concerns.
However, the applicant has directed by-l memorandum NP-13211 that a project or task manager be appointed, as appropriate, for specified types of
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activities.
The task or project manager is
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specifically responsible for serving as a focal point
for:NRC concerns and for ensuring effective
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communication.
Additionally, a site licensing manager
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P has-been appointed to ensure that responses to NRC concerns are effectively and promptly provided.
The NRC inspector deemed these and the other corrective actions to adequately address this violation.
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- Accordingly, item d of this violation is closed.
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-(Closed) Violation (445/8938-V-04):
This violation e
identified <that-one of-the_ applicant's nondestructive examination procedures was less conservative-than ASME Code Section XI-regarding the recording-of ultrasonic testing (UT) reflectors.
The Code requires all reflectors exceeding 50 percent of the reference level to be recorded.
The applicant's procedure, ISI-206, Revision 0, " Ultrasonic Testing," stated that reflectors exceeding the reference level need be recorded only if they are determined to represent valid flaws.- Reflectors, due to the geometry of the material were, therefore, excluded from the reporting requirements.
The violation stated that due~to this; procedural inadequacy, geometric reflectors.were-not-recorded during the Unit 1 Preservice Inspection (PSI)
Program.
The= recording of geometric reflectors is intended-to create baseline information for comparison with later examinations.
The applicant's supplemental respons'e to this violation (TXX-89859) identified-that Relief Request ~B-13, documented j
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in CPSES SSER 12, Appendix 5, paragraph 3'.3, provided.
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relief from recording geometric reflectors in-the Unit 1 PSI.
As a' result, the applicant withdrew a commitment made
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in their initial response (TXX-89730);to revise.the Unit 1 l
Final Safety Analysis Report (FSAR) to indicate that geometric reflectors would be recorded during Unit 1
inservice inspections (ISI).
Despite their intent to leave
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the FSAR intact, the applicant has' revised Procedure ISI-206 (as Revision 1) to require ~the' recording _of geometric. reflectors.within the Unit 1 ISI program.
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Therefore, the desired baseline information will be constructed during the first series of Unit 1 ISI-examinations.
The NRC inspector reviewed all correspondence related to this violation, participated in a conference call with NRC and applicant representatives, and reviewed Procedure ISI-206, Revision 1.
Based on this information, the applicant's response to this issue appears to be adequate i
and, consequently, this violation is closed.
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e.
(Closed) Violation (445/8960-V-02):
This violation concerned certain electrical tape splices which were not in accordance with requirements.
Specifically V-configuration
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- tape splices in Limitorque motor valve operator 1-8808C were not installed in accordance with the requirements of f)
Electrical Specification 2323-ES-100, Revision 6.
The l.
deficient tape splices could have resulted in a reduction L
of environmental protection for the spliced conductors.
L The applicant has reviewed the violation and taken the following corrective and preventive actions.
A review of y
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installation _ records >and design documents was performed to-identify plant equipment containing this' type of taped j
splice.. Subsequently, a field examination of that.
-equipment identified 20 locations in which the tape splices
were improper.
These-improper spli'ces were-documented on
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NCRs and dispositioned according to. environmental requirements and--reworked as necessary.
u As action to preclude recurrence, the applicant issued a revision-to.the electrical specification to clarify the installation of the tape splices.
Applicable construction
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and QC personnel were retrained in order to be more aware
- of-the tape splice requirements.
- The NRC inspector reviewed DCA 78713, Revision 12,-which
. revised the section of the' Electrical Specification ES-100
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that addresses V-configuration tape splice / installation.
The NRC inspector also reviewed sections of lesson plan EM 31.D89. X A1.LP, the lesson plan used for retraining:
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personnel to be more aware of tape splice requirements..
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These documents were determined to be appropriate for clarifying and emphasizing the requirements for V-configuration tape splices.
Since the applicant has implemented acceptable corrective and preventive actions, this violation is closed.
f.
(closed) Violation (445/8976-V-01):
This violation identified two snubbers with end cap / dust cover assemblies D
that could rotate about their travel indicators.
This.
condition was apparently the result of failing to follow the installation Procedure, ACP-ll.5, " Component Support
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LFabrication and= Installation."
After further review, the
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applicant determined that the procedure or adherence to the
procedure was not the major cause for this condition. LThe
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snubbers have a locking device which functions to prohibit i
rotation of_the dust cover relative to the snubber body.
If sufficient torque is applied to the dust cover, the
E locking device can-deform such that the dust cover can rotate freely.
The applicant stated that the torque
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required to deform the locking device is not excessive and
'4 would not damage other snubber components.
The applicant J
believes the deformation of the locking devices occurred
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when mechanical assistance was used to reconnect the i
)-l snubbers to the end connections.
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i In consultation with the snubber manufacturer, Pacific-Scientific, the applicant determined that the l4 unlocked dust covers did not affect the operability of the snubbers.
This was based on verification that there still existed adequate thread engagement on the load column.
Due to the limited rotation permitted by the snubber end-m
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-28-connections, the thread engagement existing at the time of-installation will remain essentially unchanged regardless of movement of the dust cover.. The applicant inspected
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60_' snubbers selected, randomly and discovered that 5 snubbers in this' sample exhibited simil~arldust cover-rotation. ~All five snubbers were verified to have-adequately threaded load columns and'thus were considered acceptable without rework.
The applicant-has revised construction drawings pertaining-to snubber installation to' require QC. verification of
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-proper load column' thread engagement during snubber installation. -Additionally, Procedure MSM-60-0215,
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" Mechanical Snubber Maintenance," Revision 0, has been
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revised to require notification of engineering whenever a snubber duct cap is-found to rotate relative to-the snubber
body,
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The NRC inspector reviewed correspondence from the applicant and Pacific-Scientific, NCRs 89-9642 and 89-9643 (issued for the original two-snubbers), Project Technical l
Report (PTR)-15, " Engineering Evaluation of Snubber Rotation," a-sample of the revised construction drawings, and the referenced revision to Procedure MSM-GO-0215.
Based on this information, the NRC inspector determined-that the applicant has taken adequate corrective action for this issue.
This violation is closed.
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(closed) Vl.olation (445/8976-V-02):
This violation
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identified a-stud which was underengaged on its associated nut'on flange No. 1, valve 1-8010A, isometric drawing BRP-RC-1-RB-028B.
This condition was in violation of-Specification 2323-MS-100 which requires studs to be at a minimum flush with the top of the nut. face..The applicant interviewed the individuals who last installed this-flange and discovered that they were aware of the subject
'3 requirement.
They had not noticed the discrepant condition partly due.to accessibility limitations resulting from temporary staging that had been erected around the valve.
The amount of underengagement of the stud was approximately 1/16' inch.
Considering that there are 12 studs on this flange, the as-found condition would not have affected the safety-related function of the flange.
The subject stud was repositioned and the nut retorqued under Work Order C890015226.
The applicant reinspected flanges on 17 valves similar to the affected valve and 55 additional valves selected at random.
Inasmuch as no additional thread engagement
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conditions were identified, the applicant concluded that the-identified case was isolated.
The mechanics involved were counseled on the importance of procedural compliance
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=andLattention to detail..The Manager of operations,-OC, issued:a-memorandum to Operations QC personnel making them
. aware-of the violation.
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'The NRC inspector' reviewed Work Order C890015226.and
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verified'in the field that the stud had been properly
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repositioned.
Additionally, the NRC inspector reviewed
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documentation supporting the applicant's flange reinspection effort and the memorandum generated-by the.
f Manager of Operations, QC.
Based on.this information, the_
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f, NRC inspector concluded that the applicant had taken adequate corrective and preventive measures for the i
identified discrepancy.
This violation in closed.
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h.
(Closed)._. Violation (445/8978-V-02):
This violation identified'the. improper' removal of lugs from a pipe spool.z The lugs were an integral part of the design of a pipe: whip restraint.
Additionally, five DCAs_were'necessary to restore a proper configuration.
The applicant replaced the-lugs-and_ determined that this was'the only. application Ti where lugs were used as part of a pipe whip restraint.
The-applicant. stated that due to'this unique configuration, no other preventive actions were necessary.
The NRC inspector requested additional'information to support this position
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and to ensure that this event was not an indication;of.a-larger configuration control problem.
The applicant-
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supplied the original plant drawing (BRP-CS-1-AB-005, CP-1)
which.had been modified to remove.the lugs.
Note 12 on-this-drawing clearly states that the lugs are part of the pipe whip restraint.
The engineer failed to acknowledge this_ note while completing the DCA (71228).
The removal of
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z the lugs, therefore, appeared to be the result of a personnel error rather than a programmatic weakness in the design. control. process.
The applicant agreed that the-first two revisions to the DCA did not adequately resolve the pipe support interference problem which originally prompted the removal of the lugs.
Later revisions were appropriate and adequate according'to the applicant.
The NRC inspector reviewed drawing BRP-CS-1-AB-005, CP-1, and all revisions to DCA 71228 and verified that personnel-5; still on site who had responsibility for this issue had
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been made aware of the violation.
The NRC inspector
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item were adequate.
This violation is closed.
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(closed) Violation (445/8978-V-03):
This violation concerned numerous errors which were found in Appendix E of Specification 2323-SS-16B and DCA 74249, Revision 7.
(DCA 74249 was an engineering change to Appendix E).
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The app 1'icant'has evaluated this deficiency and determined
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- that the' errors were~ attributable to oversight or lack of:
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U attention to detail on the part'of the engineering.
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personnel ~ preparing and reviewing Specification u-L2323-SS-153, Revision 2, and,DCA 74249, Revision 7.
H As corrective _ action, the applicant reviewed Appendix E to identify other errors.
The applicant determined that
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approximately 85 errors were identified out of
-approximately 2300 attributes.
DCA 74249, Revision-19, was
issued to correct those errors.
The errors-generally-occurred in three categories of attributes:
(1) restraint
number, (2) restraint type, and'(3) bounding problem'
number.
The applicant stated that these categories of
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attributes were provided in the specification as. general
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information.and'were not used in the performance of field work.
The applicant concluded that existing programs.and procedures were adequate to preclude such documentation e1 discrepancies from causing inadequate installations..
As action toLpreclude recurrence, the chiefLengineer issued a memorandum to engineering-personnel emphasizing the responsibilities and requirements-regarding DCA preparation
and review.
The memorandum was distributed to both'
TU. Electric and contract engineering personnel.
The NRC inspector reviewed the above actions and information and, in'this case, concurs that-the type of
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errors' identified are unlikely to result in-inadequate'
restraint installations.
The NRC inspector recognizes
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that, in general, construction performs work to controlled
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drawings distributed by the site document control group.
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'Recent NRC~ inspections have= determined that the current document control program appears to be effective in providingl controlled drawings.
However, the NRC inspector questioned the applicant'regarding the generic implications of the numerous errors identified in this DCA'.
1The applicant provided the following additional-information.
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Site programs.such as the Specification, Procedure and Drawing Update (SPADU), Design Validation Program (DVP),
Technical Audit Program (TAP), and-the Engineering
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Functional Evaluation (EFE) have provided reviews and evaluations of existing specifications.
Those programs.
h have effectively established a baseline of technically.
adequate specifications.
Other programs, such as surveillances of DCA adequacy and the trending of DCA related deficiencies have identified few problems due to DCA errors and indicate that the content of DCAs is generally satisfactory.
The applicant concluded that the errors in DCA 74249 appear to be an isolated occurrenco and the program for design control is adequate.
The NRC
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inspector concurs'with the. applicant's' conclusion.-
Accordingly, this violation is' closed.-
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-(closed) Violation (445/8984-V-01):. This violation concerned a cable grip which was, improperly suspended from
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the3end of electrical ~ penetration assembly E-76.-
The-improperly suspended cable grip bail was located at the module ~ polymer seal which could have damaged the
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penetration. seal and/or the Kapton insulated wires
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. associated with the penetration.
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The app'licant's response to this violation which-was provided in TU Electric's letter TXX-90024, dated.
January 19, 1990, stated that the subject cable grip'was initially installed in December 1987,. removed in January 1988, and reinstalled in March 1988.
The applicant's-
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review also indicated that the cable grip bail was properly
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located at=the completion of both the initial installation and the subsequent reinstallation.
Based-on these efforts',
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the applicant was unable to conclusively _ determine how thez
cable grip became mispositioned.
However, as1 stated in the
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above referenced response letter, the configuration'of-this-cable grip assembly is such that the bail could be
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relocated without disassembly.
In the absence of a
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definitive root'cause, the applicant hasLpostulated that the cable grip bail may have been relocated during other-
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work activities in the area.
Given that.the cause of the mispositioned cable-grip was indeterminate, the applicant's corrective actions were-focused cn1 the effects of this condition.
Specifically, the Kapton insulation on the associated conductors was-inspected by QC personnel in accordance with Electrical.
,
-Specification ES-100'and was also evaluated'by engineering for damage prior to the installation of WT-65 Siltemp wrapping.- In addition to' repositioning,.the cable grip bail for the subject penetration in accordance with Work order C890015639, the applicant conducted an evaluation of similar cable grip installations to determine if any additional configuration discrepancies existed.
This evaluation' indicated that all similar cable grip installations were properly configured..
Based on a review of the applicant's actions in response to i'
this-violation, the NRC inspector determined that
,
TU Electric's corrective measure appeared adequate to
_ prevent future reoccurrence.
Accordingly, this violation
L is closed.
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Action'on 10 CFR Part 50.55(e)' Deficiencies Identified by the
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Applicant (92700)-
n a.-
(Closed - Unit'l only) Construction Deficiency (SDAR CP-83-15):'
" Bolting Material-Used;for Cable Tray
'
Clamps."
This deficiency involved the improper use of mild
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steel bolting (ASTM A-307) in lieu of high strength bolting to attach cable tray clamps to cable-tray supports..
Additionally, due to a^ procedural discrepancy, the bolts l:
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L were only wrench tightened instead of being. tensioned by the turn-of-the-nut method as required by Specification 2323-SS-16B.
Previous NRC review of this issue is documented in NRC. Inspection Report' 50-445/89-75;-
50-445/89-75.
This construction deficiency was left-open n
in the referenced report pending completion of all.
modifications resulting.from the applicant's review of this
o E
issue.
During'this inspection period, the NRC inspector.
t reviewed documentation provided by the applicant showing that all work'related to this item is'now complete.
p Consequently, this construction deficiency is closed'for t
o L
Unit 1.
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b.
(Closed - Unit 1 only) Construction Leficiency (SDAR CP-85-35):
" Cable Tray Hanger Design." 'This issue
' involved the resolution of concerns. raised by: the' NRC and
'
CYGNA regarding-the design of cable tray supports.
Additionally, discrepancies existed between as-built and
,L as-designed configurations.
Previous NRC review of this-
. issue is documented in NRC Inspection Report 50-445/89-75; 50-445/89-75.
This item was left open in the referenced
.
report.pending completion of all modifications'resulting from the applicant's review of this issue.
During this inspection period,1the;NRC inspector reviewed documentation provided by the applicant showing that all. work related to this item is.now complete.- Accordingly, this construction deficiency is closed for Unit 1.
l c.
(closed - Unit 1 only) Construction Deficiency (SDAR CP-85-50):
" Cable' Tray Tee Fittings."
The applicant's. actions to resolve this construction deficiency were evaluated in NRC Inspection Report 50-445/89-66; 50-446/89-66.
That evaluation found the applicant's
-
corrective and: preventive actions generally satisfactory.
However, to complete the review additional information was
i.
required.. Specifically, the inspector required:
-(1) an explanation of why apparent nonconforming conditions were documented on design change authorizations rather than on standard discrepancy documents such as NCRs; (2) documenta-tion indicating how the vendor test data (data resulting from testing performed by Corporate Consulting and Develop-i ment Company (CCL]) and engineering calculations were adopted into the design criteria contained in ECS-5101; and
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(3) confirmation that all necessary rework was complete'and
.QC inspected as required..
The' applicant provided the above requested information and L
the NRC inspector has completed review of the information.
H For item (1), the applicant' explained that the scope of L
Field Validation Method (FVM)-CS-050 was to obtain and-record "as-built" data regarding field dimensions and weld p
information.
The information gathered by the ETH was subsequently compared to the minimum weld requirements
specified on the vendor. drawing and DCA 41204.
However, i
L acceptability of the cable tray tee fittings was evaluated.
l by comparison to the applicant's cable tray tee fitting L
test'results.
Those cable tray tee fittings identified to L
require modification or replacement after comparison to the
['
test results were documented on NCRs..
The NRC inspector determined this explanation to be acceptable.
Regarding the incorporation of vendor data, test data, and I.
engineering calculations into Procedure ECS-5101, " Cable
Tray and-Cable Tray Hanger Design," the applicant explained
)
h that Impell Calculations M-36 and M-86 and Ebasco Volume'1,
,
Book 1, part 5, item 3, generated allowables based-on' test data derived from CCL Test Report CCL-A742-87.
These developed allowables were then incorporated into the design criteria contained in ECS-5101.
Although vendor data was considered, the calculations relied more on the-CCL test i
data as'it was determined to be more applicable to the
.
. plant installed condition.. This approach was determined to
!
be acceptable by the NRC. inspector.
.
!
As' evidence that all' required rework of the cable tray tee fittings had-been accomplished andlthat applicable QC inspections performed, the applicant provided a printout of
the package. status for each affected cable tray tee fitting and an example of a typical package with QC inspection s
requirements.
Each of the work packages for Unit 1 were identified as complete.
The sample package provided l
evidence that typical QC involvement was appropriate.
'
i Based on the initial NRC inspectors. satisfactory review and
a review of the additional information provided by the applicant, this item is closed for Unit 1 and common.
j
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Closure for Unit:2 will be dependent upon a satisfactory
4 completion of required work packages for Unit 2.
j
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d.
(Closed - Unit'l Only) Construction Deficiency (SDAR'CP-86-45):
" Seismic Category II systems and Components."
This construction deficiency incorporated the resolution of deficiencies involving the potential unfavorable interaction of Seismic Category II hardware with Seismic Category I hardware.
Previous NRC review of
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.this issue is documented in'NRC Inspection, Report
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50-445/89-76; 50-446/89-76. LThis item was-left open inLthe
.>
referenced report pending-completion of DCAs resulting from the applicant's review of this issue.. During this inspection period, the'NRC inspector. reviewed documentation
'
l.-
provided by the applicant showing ~that all DCAs related'to-t this issue are now complete.
Accordingly, this
-
construction deficiency is closed for Unit 1.
e.
(Closed:
Unit 1 only) Construction Deficiency (SDAR CP-86-75):. "ASCO Solenoid Valves-in Piston Air Actuators."
By letter TXX-6097 dated November 20, 1986, the applicant notified the NRC of a potentially reportable-
!
item involving ASCO solenoid valve internals being exposed
to petroleum based lubricants.. The deficiency was-
,
identified during the applicant's actions in response to Generic NRC Inspection and Enforcement Notice (IEN) 80-11,
"
Problems with ASCO Valves in Nuclear Applications Including Fire Protection-Systems.".Of specific concern were ASCO solenoid valves equipped with ethylene propylene elastomers potentially exposed to petroleum lubricants-used in piston air-actuators.of Pacific Air Products Company (PAPCO)
'
Ethylene Propylene polymers are resistant to degradation due to radiation exposure but experience accelerated degradation when exposed to petroleum based lubricants.
!
The. applicant presented information detailing actions taken to: resolve the concern.
Actions taken included review of safety-related systems to determine the applicability of t
'the concern.to other piston operated dampers, issuance of NCRs and DCAs to document deficiencies and'to.effect changes in hardware and design documents, contacting the manufacturers of purchased piston air actuators to determine the type of lubricant used (silicon based lubricants do not have the same degrading effect),
replacement lof certain ASCO seals with Viton material (Viton is resistant to petroleum degradation), updating the
' Equipment Qualification Manual. List (EQML) to reflect.the design changes, assuring that maintenance procedures for piston actuators specify the use of silicon based lubricants, and revision'of'the lubricant specification to ensure that departures.from the use of silicon based lubricants are subject to review for material compatibility.
.
In-evaluating this response, the NRC inspector reviewed the NCRs and1the DCAs issued.for the affected solenoid valves, a sample of the travelers and QC inspection reports used to document the changes, correspondence between TU Electric and PAPCO, maintenance procedures for site actuator maintenance requiring the use of non-petroleum based
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lubricants, and the site EQML-to verify that all ASCO solenoids converted to Viton seals were properly-identified.
Based upon this review the NRC inspector
.'
determined thatEthe app 1dcant's approach was_ sound and 1 appeared thorough.
However, the NEC inspector requested-
,
additional information regarding the following:~
(1) were
-
the Viton replacement seals procured as safety-related i
material, (2) have all safety-related ASCO solenoid valves'
requiring Viton substitution been changed, and
-.
(3) resolution of.certain' minor discrepancies-noted during the NRC inspector's. review.
In response to the above request, the applicant provided the following:
-(l). copies.of the supplements to the ASCO purchase orders referencing the applicable NCRs and requiring that the material be procured safety-related,-
(2) the review of HVAC damper. data sheets used to determine
,
the-solenoid valves requiring Viton. substitution, and (3) satisfactory answers.regarding each of-the minor
[
discrepancies..
Based on the satisfactory review of the applicant's actions.
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taken to address this issue and-to preclude recurrence, this item is closed for Unit 1 and common.-
Closure of-SDAR CP-86-75 for Unit 2 is dependent'primarily on satisfactory completion of hardware changes for Unit 2 since the programmatic issues are satisfactorily addressed here.
,
f.
- (Closed - Unit 1 only) Construction Deficiency-(SDAR CP-87-13):
" Class lE Separation Violations."
By letter TXX-6519 dated June _15, 1987,- the applicant notified the NRC1 ofLa deficiency involving Class lE separation violations which resulted from the removal or modification-of nonsafety class cable tray covers after acceptance by; n
QC.
The deficiency was documented by NCR-E-86-101601 and corrective action report (CAR)-064.
The applicant has-completed an evaluation of this deficiency and-determined that the deficiency was reportable in accordance with the requirements of 10 CFR 50.55(e)..
Corrective actions for this deficiency included the J
replacement of required cable tray covers.
The replacement m
was performed as part of the Corrective Action Program and was conducted _using FVM CPE-SWEC-FVM-EE/ME/IC/CS-088,
?
" Electrical Separation."
Additionally, the applicant documented resolution of the deficiency in the closure of CAR-064.
,
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Preventive actions for this deficiency included revision of Procedure CP-CPM-6.10, " Inspected Item Removal Notice (IIRN)," to require that an IIRN be issued for the removal o
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.of;any cable tray l covers.
The.IIRN would result in.
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inspection of the cable tray. covers for conformance,to-
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separation criteria.
Applicable personnel-were trained-in accordance with the procedure revisions.
The Procedure
,
CP-CPM-6.10 was; subsequently superseded by' construction procedures requiring the issuance of construction travelers or other controlled documents-to'effect the removal:of-
'
cable tray covers and to_ assure'their replacement or~that an evaluation of their removal would be performed.
.The change from-Procedure CP-CPM-6.10 to the other construction
Jprocedures was accomplished by. processing Commitment
'
. Variance Request CVR-89-0-22 to assure:that the intent of
,
the original commitment was maintained.
CAR-064 also l
verified:
(1) training of personnel, (2)' adequate
'dispositioning of NCR E-86-101601, and (3) reinstallation
,
,
of cable tray. covers (as verified by implementation of FVM-088).
'
'The NRC inspector reviewed the documentation of the above actions and determined that the actions and-the supporting-documentation were satisfactory.
Additionally, NRC
,
inspection of the implementation of FVMs under the PCHVP determined that they were satisfactorily performed (see 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.-
,
Note:
FVM-088 addressed Unit 1 and common only.
Based'on the satisfactory implementation of corrective and preventive actions, this construction deficiency is considered closed for Unit 1 only.
Closure for Unit 2 will be held'open pending the satisfactory-implementation of
,
J corrective actions for' Unit 2.
g.
(Closed - Unit 1 only) Construction Deficiency P
(SDAR CP-87-61):
" Snug Tight Torquing of Structural
. Bolts."
This issue involved the resolution of deficiencies i
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involving high strength bolts that were not torqued to AISC-standards.
Previous NRC review of this-issue is documented E'
in NRC Inspection Report 50-445/89-74; 50-446/89-74.
This
,
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item was left open in the referenced report pending
'
completion of NCRs resulting from the applicant's. review of-this issue.
During this inspection period, the NRC
.
inspector reviewed documentation provided by the applicant showing that all NCRs related to this issue are now j
complete.
Accordingly, this construction deficiency is
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L closed for Unit 1.
.
L h.
(Closed) Construction Deficiency (SDAR CP-87-62):
By l
letter TXX-6731 dated September 30, 1987, TU Electric h
reported an item conservatively deemed to be reportable
!
under the provisions of 10 CFR 50.55(e).
This item related J
to design deficiencies in the use of nomographs to design l
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seismically nonsafety-related supports; i.e.,Ifor non-ASME
,
. Code, Class 5,.large bore'(greater than 2-inches in diameter) piping systems.
! !
As stated in TXX-6731, thisfitem was identified-during
third-party inspections.
Some of the-subject Class 5-
~
' piping had been seismically qualified using a nomograph method.: InLsome cases, this approach resulted in inadequate analysis. documentation to support the design-basis for seismic category II pipe supports and the
.
generation of~ loads on these supports and insufficient
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methods for isolating the seismically supported piping from
_
.the effects of interconnected nonseismic piping.
The root cause df the. condition was determined to be
'
inadequate engineering requirements in Specification
.
. 2323-MS-46B, " Piping Hangers (Non-Nuclear)."
This-t specification-did not provide adequate guidelines on proper-e use of the nomograph for Seismic Category II large bore pipe supports.
The deficiency was generic to all Seismic
~
Category II large bore pipe supports and associated Class 5 piping systems that were not rigorously analyzed as part of p
the SWEC requalification process (e.g., high energy piping L
and extensions to ASME Code stress problems).
L L
The method employed in the design of the seismic Category II large bore pipe supports for Class 5 piping systems could have resulted in failures in these systems during a seismic event..This condition, if. uncorrected, could have resulted in. adverse interactions affecting the operation or functioning of safety-related equipment or-
,
systems,
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As corrective action, TU Electric had proposed to:
(1) revise-Specification 2323-MS-46B to include appropriate
- criteria for the use'of nomographs; and (2) implement a j
hardware' validation program via FVM CPE-SWEC-FVM-PS-082,
" Validation of Seismic Category II Large Bore Pipe Supports," for the systems involved.
The -IR44 program was to include engineering walkdowns, engineering evaluations and documentation of the completed evaluations.
>
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As preventive action, TU Electric had proposed to establish
'
the-following measures:
(1)
Products of the hardware validation program were to be documented and controlled by Procedure ECE 2.13,
" Retention and Control of Engineering Documents," to provide traceability of the design basis.
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(2): Procedure,ECE 5.05, " Drawing Control," was to be
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issued to assure that configuration (consistent with
the design documentation) is maintained.
(3)
Design Basis-Documont-(DBD)-CS-068, "Non-ASME Piping
.
and Support Design," and Procedures ECS 5.03-05, i
"Non-ASME Pipe Stress Analysis," and ECS 5.03-06,
"Non-ASME Pipe Support Design," were to be issued to c
assure. Seismic Category IIflarge bore pipe supports and the associated Class 5 piping.are designed to
)
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. acceptable and approved engineering. criteria.
Subsequently, TU Electric letter TKX-88471 dated June 16, 1988, provided the status and updated the previously proposed. corrective.and preventive actions.
Changes to j
these actions were primarily procedural in nature.
Finally,_by letter TKX-89651' dated. September 7, 1989, TU Electric revisedithe corrective action as proposed in TKX-88471.-lTU Electric elected to implement.a new program
,
to demonstrate the adequacy of the total population'of.
'
-large bore Class 5 pipe supports by selecting a bounding sample consisting of'the least conservative seismic design configurations, and proving the acceptability of those configurations by analysis.
Demonstration of the acceptability of the bounding sample was to provide H
adequate assurance that the remaining pipe support systems wereLalso acceptable and would not catastrophically fail during a seismic. event.
'
To implement this new. program, a walkdown was to be conducted for Unit 1 and common areas within Seismic
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Category.I structures to identify-design configuration
'
anomalies in large bore Class 5 piping systems which, based on earthquake engineering database experience, have less conservative seismic designs.
A bounding sample was-then to be selected from the observations recorded during the walkdown.
A detailed j
analysis was to be performed on each anomaly in the
bounding sample.to demonstrate its structural design
adequacy during a seismic event.
If any anomaly in'the
bounding sample was determined by analysis to be unacceptable, the affected piping / supports were to be modified as required.
In that event, the anomaly with the l-next least conservative design would then be selected for
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analysis from the walkdown observations to replace the unacceptable anomaly in the bounding sample.
The above program superseded the corrective action specified in TKX-88471.
However, as indicated in
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TKX-88471, recurrence of this deficiency was to be j
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E precluded by changing the_ applicable design basis document'
Eto require-that nomographs be utilized only in~accordance
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with approved engineering procedures.
During this inspection period,' reviews of the new TU Electric program were continued and assessments of the:
i corrective-and preventive actions performed. fFurther,-
.i during a previous-inspection, theLNRC inspector had:found l
that although no1 concerns relating to class 5 small bore
- piping systems had been: identified.by external sources, TU Electric had prudently-elected to also extend their new program to include small bore (i.e., 2-inch and smaller-
diameter) piping systems.
Accordingly, the NRC reviews _and-
,
assessments were extended to include both.small bore and
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large. bore Class 5 piping systems.
Results of reviews and assessments-for these two sizes of systems are provided separately in the following.
l Small Bore Piping Systems
'
Documents reviewed and results of these reviews are as-follows:
i (1)
EQE Procedure 52006.01-P-002, " Field Walkdown
-
Procedures for Selection of Two Inch and Under
.Non-Nuclear Safety Non-Seismic Piping for Analytical Evaluation," Revision 1, dated April 26, 1988:
l This procedure was to be used in walkdowns to:
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(1) assess, on the basis of seismic experience
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database and the in-structure seismic response spectra for CPSES, the potential for failures in the subject piping system due to seismic-loadings; and.(2)? select
,
examples of these piping systems.which have the
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L highest-vulnerability for failure.
The example I
systems were to be analyzed by TU Electrici.In. fact, not all examples identified were analyzed.
Instead,_
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i the examples were assessed and the most vulnerable of
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representative types selected to constitute a bounding
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L sample.
Only systems in-the bounding sample were p
analyzed by TU Electric.
~
i-Procedure EQE 52006.01-P-002. included requirements for-(
walkdown training, field walkdown methodology,Jand the
/
selection of piping systems for analytical evaluation.
Selection criteria were based on strength of piping, i
fittings, and supports as demonstrated in past
!
This EQE procedure also provided specific piping span spacing guidelines (Appendix B)
L and general criteria for identifying seismic L
valnerabilities in piping systems (Appendix C).
To l'
assess the adequacy of.the EQE walkdowns, the NRC l'
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inspector 1 conducted independent walkdowns.and reviewed-with TU Electricfand EQE personnel piping systems
,
selected by EQE. LThe NRC-inspector found that the~
example systems identified were among the worst.from a seismic vulnerability viewpoint.
Based on the
. guidelines and criteria provided in the procedure.and.
'
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the acceptability of the piping systems' selected i
during implementation of the procedure, the NRC.
inspector found that the procedure and.results-
~
achieved were acceptable.
(2)
TU Electric Procedure ECS 5.01-29, " Design. Criteria for Small Bore Non-Nuclear Safety-Related Piping and Supports Assessment in Support-of EQE Systems.
ti Interaction. Program," Revision 0, dated March 30, 1988:
.
.This procedure established the requirements for stress
analysis and support design for the subject piping and supports.
The procedure specified thatn (1) stresses
.
in the piping. system were to be limited to the lesser'
a of.3.6 S,or 1.05 S I" *#* Sm and s are the primary u
u membrane' stress limit auf the ultimate strength,.
'
respectively, of the material under consideration) as prescribed in the ASME Code, 1986 Edition, Appendix.F, Section (actually subsubparagraph) F1331.l(c)(1); and.
.(2) pipe support structural members were to remain within alastic limits.
Deviations.to the specified limits were'to be. permitted.on a case-by-case basis provided.they were evaluated using appropriate seismic margin assessment methods developed by EQE.
,
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The NRC inspector found that the stress 111mit
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specified for piping'by this procedure was'the faulted condition primary membrane plus primary: bending stress intensity (Pg+P) limit criterion specified for b
components ~by Appendix F of the 1986 Code.
As an alternative, Subsubarticle F-1430 specified that for piping, when applying thetprocedures of'NB-3652, criterion)
Equation (9) of NB-3652 (the piping Pg + LP3 Es shall be limited to the more conservative lesser-of
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is the' yield strength of the-3.0 S,and 2 Sy (where Sy material) criterion.
However, the NRC inspector found.
E that in practice the more conservative lesser of 3.0 S,and 2 S piping stress limit was applied to a y
conservative modification of the ASME Code
'
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Section III, Class 2 and 3 load combination in the TU Electric piping system analyses, but that-deviations were allowed for threaded and socket welded joints based on EQE margin assessments.
As documented below, these TU Electric piping analyses were acceptable to the NRC inspector.
Consequently, L
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although the piping stress' limits-specified in this procedure may be acceptable, the NRC inspector would.
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recommend that, for~the sake ~of consistency,-future-
>
analyses performed under this procedure be based on
the more conservative lesser of 3.0 S,and 2 S limit
y and the EQE derived limits for, threaded and' socket
,
welded joints utilized to date.
Moreover,'the NRC'
inspector.would recommend that deviations from the
_
actual criteria utilizedLin-the analyses be limited on-
.
a case-by-case basis to the lesser of 3.6 S*
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and 1.05 S specified in this procedure.
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Further,.the NRC inspector found that in contrast to-I
'theLvery specific criterion provided-for piping,-the criterion provided by'this procedure for supports was limited to structural members in supports only and was
. a generalization of the actual criteria utilized in
~
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Lthe analyses.
Criteria for other elements of the supports-including welds, standard components,"and anchorages.were notJspecified.
However, as documented below-and similar to the case for piping, the NRC inspector found.that the. actual TU Electric support analyses,were acceptable.
Moreover, the NRC. inspector found-that the. limits specified for deviations from-the acceptance criteria in the actual support i
calculations were also acceptable,
.
Subject to modifications to the acceptance-criteria for piping and-supports and the specification-of-
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limits for deviations from these criteria described in
the preceding, the NRC inspector found that this procedure was adequate to~ assure that the subject piping systems and supports:will not fail catastrophically during seismic events.
(3)
EQE Report 52006.01-R-001, " Seismic Evaluation of Non-NuclearESafety Non-Seismic Small Bore Piping at Comanche Peak Steam Electric Station," Revision 1, dated December 1989:
."".
.This report:
(1) described the walkdowns conducted to-identify.the sample piping systems; (2) identified the 20 bounding systems (selected'from the sample systems
'
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identified'during the walkdowns) to be analyzed by j
TU Electric; (3) described the.results of a seismic
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margin. assessment of the selected piping systems which was performed to demonstrate the' reserve strength of the small bore piping systems; and (4) described the results of a comparison between CPSES piping and piping in industrial plants that had experienced earthquakes up to four times the CPSES Safe Shutdown Earthquake (SSE) without evidence of failure which was
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performed to further demonstrate the' ruggedness,of the-
- CPSES small bore _ piping systems.
"
Relative to all'except the first of the preceding *
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(a)
The 20 bounding piping systems were. identified.in
Table 2-2 of this report.
The table also o
identified-the stress packages (C402 through j
C421) in which the bounding systems were analyzed-
and the characteristics.of and items,of concerns relating to each of the bounding systems..
. Types of piping systems in the bounding. sample included fire protection, demineralized, chilled,
.
and turbine water systems; and waste processing, chemical and volume control, service air and instrument air systems. -These bounding sample
"
systems were distributed throughout-Unit 1 and common areas' Seismic Category I buildings.
Characteristics tabulated included piping
,
. materials,-predominant and secondary support j
"
types, and predominant and secondary types of pipe fittings.-
Concerns tabulated included stiff branch with flexible header configurations, inertial loads, Jcorrosion,Thanger failure, low capacity joints,-
'
in-line mass, differential building motion and equipment anchorage..
Based on~the preceding, the.NRC inspector found.
that the piping systems in the bounding sample were acceptable.
The number and types of piping j
systems included their distribution through the plant; the range of the types of' supports, pipe fittings, and concerns represented; and the peak
.]
seismic' accelerations.(ranging between 2.13g
,
and 3.37g except for one case of 1.37 ) to be
sustained were adequate to' assure that the-l 20 systems constituted a bounding sample of the subject small bore piping systems.
'
k (b)
The seismic margin assessment performed was intended to demonstrate that the TU Electric.
analytical evaluations for the 20 bounding piping systems were more conservative than required by
margin assessment methods developed'by both the NRC and the Electric Power Research Institute (EPRI) for the evaluation of nuclear power plants for earthquakes greater than their design basis.
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These; methods'were.to demonstrate a
"High-Confidence-of-Low-Probability-of-Failure i
-
(HCLPF)" corresponding approximately to a 95%
'
confidence of less'than 5% probability of failure should the defined earthquake. occur.-
j This assessment purported to:
(1) identify'_
,
compounding conservations inherent'in the piping
' design analysis chain utilized by TU Electric in
,
their analysis.of the 20 bounding--systems; (2) provide approximate quantifications ofsthese
'
conservatisms as they-pertain to the TU Electric analysis method;-and (3) demonstrate that the criteria utilized in the TU Electric analyses
'
,
were conservative.
-.
The NRC inspector'found that the' basic stress:
acceptance' criterion proposed in the EQE-report for piping systems included the-following-conservatively modified occasional load.
'
combination of the ASME Code for~ Class-2 and 3 or the ASME/ ANSI:B31.1 Code-piping systems:
= Lesser (3'S ' 28 PD
+ 0.751 (Mg+MB m
y
[
gg 2t Z
'
where all terms are as defined in the referenced codes and M is the M resultant moment term-e B
B defined in referenced codes modified:to include
'
moments due to seismic anchor movement 1 (SAM).
The SAM > induced moments'and the code earthquake.
t criteria induced M m ment weresto be combined by
B the square root of*the' sum of the_ squares (SRSS)'
method 7on the basis of-probabilistic arguments.
- However,.the NRC inspector found in practice the
. moments were combined such that the 0.751' stress
-
intensification factorn(SIF) term did'not' apply
~
to-the SAM induced' moment. -However, the NRC
'
inspectorLfound.that this was acceptable for the analyses performed.
The NRCLinspector recommends that: -(l) for. future TU_ Electric ^ analyses, the 0.751 factor be also.applicableLto SAM. induced
f moments; (2) the results of the: existing analyses
"
be not used elsewhere (e.g., for comparison-purposes) unless modified in accordance with recommendation (1).
Based'on the arguments presented in Chapter 3 of
'
the EQE report in support of the basic piping acceptance criterion, but subject to the restrictions described in the preceding paragraph
'
in the application to this criterion, the NRC
'
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, inspector found that'the proposed criterion was
,
acceptable.
Details of:
(1) the-relationchtp:
.;
'
-between the TU Electric, analysis method.and the
,
,
,
~
NRC and EPRI margin l assessment programst and
(2) the resultstof analyses of piping: systems :in i
the EQE database. analyzed,in'accordance with the (
'
TU Electric methodology provided in Chapter 3 of
the EQE report were-adequate'to assure that the
"
,4 overall TU Electric /EQE program will provide j
acceptable reserve margins in the subject piping g
systems.
't
,
In addition, as observed in the essessment of-y piping strength margins, arguments for the use of increased stress'allowabics in margin studies-
,
have all been based on large-bore welded piping-
-l systems.
Consequently, since some of the-
concerns for the CPSES small bore. piping systems were' associated with threaded and socket-wolded
-
joints primarily in fire protection piping systems, supplemental criteria were developed for
these joints..The approach adopted in
!
establishing.the supplemental criteria for these
!
joints was to use a combination of piping experience data, fatigue test data and fatigue; j
analysis.
Based on this approach, the allowable-i stresses-(based'on the gross pipe section)
developed for' threaded and socket welded joints-in carbon'and stainless steel piping systems with moderate operating temperatures were 44.7 ksi and 54.2.ksi, respectively-(as compared ~to the otherwise proposed basic allowable stress of 60.0 ksi). ' Alternately, when multiplied by the 0.751 SIF factor as required by the TU Electric analysis methodology, the allowable stresses were 77.1 ksi and 85.4 ksi, respectively (1 = 2.3 and 2.1 for threaded and socket-welded joints, e
respectively).
The NRC inspector found that the allowable stresses developed by EQE for threaded and-i socket-welded joints were acceptable.
The fatigue data and the results of fatigue analyses presented justify the proposed allowables.for f'y their intended application, viz.:
to assure
-
,
survivability of the joints when subjected.to 10 full cycles of the allowable stress during an
'
SSE event with a factor of safety of 20 on the number of cycles.
On an equivalent basis the more conservative load controlled allowable alternating stress would be 150 ksi.
,
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The comparison between CPSES piping and_the EQE database-piping systems which had' experienced
,y'
earthquakes up to 4' timer the CPSES SSE without.
evidence of failures was adequate to' demonstrate-
the ruggedness.of the<CPSES' piping systems.- The basis for the comparison included support o
details,~ type of' construction, system energy type and support-spans.
The NRC inspector found that, in contrast to.the; case of piping, the EQE report did not provide acceptance criteria to be utilized in the analyses of.the pipe supports in tne bounding sample piping systems.
~Further, discussions of supports were' limited'
primarily to rod type hangers.
s
')
'
(4)- TU Electric Report ER-DBD-CS-02, " Assessment of Small Bore-Non-Nuclesr Safety Non-Seismic Piping,"
Revision 1, dated December 1989:
This report documented'the details of walkdowns,'
surveying, and other efforts performed by TU Electric
.
- to obtain isometrics of the 20 bounding piping. systems
and other data in preparaticn for the piping system stress analyses.
The report also-described the methods of analysis and documented the:results of the_
,
piping and support stress analyses performed.on the
.l
'-
20-bounding-systems.
.
The piping system analyses were performed using the a
requirements of Procedure CPPP-7 as guidelines.
The i
NRC inspector found that the use of Procedure CPPP-7
.]
which was applicable to CPSES safety-related piping
~
systems was appropriate. -Loads' considered in the analyses included pressure, dead weight,. thermal expansion, SSE inertial loads and SSE SAMs..The SSE
-;
inertial loads were based on applicable CPSES envelope-
!
response spectra and ASME Code Case (CC) N-411 damping
values, and SSE SAM displacements on.the results of-i]
appropriate SWEC performed building seismic analyses,
!
During a previous NRC inspection relating to this
_
l construction deficiency; the.NRC inspector found that TU Electric was performing analyses of some piping
g~
- systems utilizing time. history inputs and CC N-411-f damping values.- The NRC inspector had found that
,
although the use of CC-N411 damping values in time
history analyses may be acceptable, such use~was not
!
in accordance with current NRC requirements.
Subsequently, TU Electric modified their method of
analysis to be as described above.
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In addition, as described in Section 4.3 of the s
report, an iterative analysis' approach was utilized int the analysis.for several of the bounding piping
-
,
systems..
Initial. analyses for these systems resulted in loadings.on some-supports in these systems.in excess ofitheir capacities-calculated in'accordance-with the allowable-stresses-specified.for supports.
!
. Subsequently,: analyses of these systems in'which the
!
'
restraint provided,by theseL" overloaded" supports were
,
ignored (i.e., the analysis conservativelyzasaumed1
,
'
=that these supports had failed) were. performed.
Ir1 general,Hthe subsequentcanalyses resulted in acceptable: piping stress: levels and' acceptable loads.
on theLremaining supports.
Stresstlevels and loadings
- ' -
were reduced at critical locations due possibly to a
,
o
- combination of'the removal of constraints.and a
L reduction in seismic loadings as a result of reduced.
.
natural frequencies.
The NRC inspector found thats in
<,
.
-
view of the purpose of the TU Electric /EQE program and-
'
seismic margin considerations,-this iterative method.
of analyses;was acceptable.' Although neglected in the:
analysis,-the-masses added to the piping system as a result of failure of the " overloaded" supports'were J
judged to be not significant1and hence not in need of
consideration.
p F
The results-of the piping analyses were-tabulated in
'
Table 3-1 of the report.
This table indicated that L
iterative analyses were performed for six of the systems and hand calculations were' performed for one.
The results indicated that subsequent to iterative
g'
type _ analyses where necessary,1the basic piping.
L allowable stress was: exceeded in only.three of the
!
bounding systems.
These piping systems were analyzed
.t
,
L in Stress Calculation ECE-WP-1-C406, ECE-CA-1-C410, L
and ECE-CA-X-C414 and the maximum stresses were L
" > 60,000 (psi)," 80,753 psi and 74,723 psi,
'
.
respectively.
During a previous inspection, the NRC inspector reviewed the hand calculations in calculation ECE-Wp-1-C406 and found that the maximum stress-j
..
calculated was 193,398 psi which was unacceptable.
'
\\P The calculation was for a short run of 3/4-inch _ branch 6>
piping between two larger sized (2-inch and 4-inch)
l headers.
The calculation conservatively evaluated the
!
stresses induced in the branch piping due to the-l differential seismic motion between the flexibly l-supported headers.
The branch piping was accepted on the bcsis of the formation of plastic hinges in the piping which was contrary to the applicable acceptance criteria.
Subsequently, TU Electric reanalyzed the L
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subject bounding piping system.. This analysis
?
' included. portions of the headers and gaps in the supports in'_the; piping analyzed.- The maximum calculated 4 stress. exceeded the basic EQE> proposed
,
~ llowable stress. criteria but satisfied the previously
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,
Yr discussed-less conservative,-alternate criteria permitted by Appendix F of the 1936 ASME. Code..The
.
acceptability of this analysis is discussed below.
The pipe support its.the small bore non-nuclear safetyL
!
'
nonseismic piping systems,-except the fire. protection system, at CPSES were typical type designs originally
-
developed by Gibbs and Hill.
These supports were
. designed to carry dead weight only and were-predominately simple rod hanger supports or U-bolts on j
simple steel angle frames. -The fire protection,
l
'
supports were designed by Grinnell Fire. Protection Systems-(GFPS) to carry-piping dead weight in accordance with the criteria " National Fire Protection Association Standard ~for the Installation of Sprinkler'
Systems," NFPA-13,1973, 1978, and 1985. editions. :
!
!
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The analysis of all the supports in.the bounding sample analyses was performed in three phases:-
stiffness assessment, capacity calculations, and'
comparison calculations.
Stiffness assessments were-
!
g performed on a stress package basis. 'After completion of the detailed walkdowns,. pipe support sketches were
. supplied to the support engineers.
The engineers-j assessed the supports to assure that they-met minimum i
stiffness criteria and, if not,: to calculate Jactual
'I stiffness values.
Appropriate support stiffness-
"
values were supplied to the stress analysts for use in.
their analysis.
Stiffness assessments were documented
'
'
in Calculation ~ECE-PSE-009.
-The second phase of support analysis was the. capacity i
calculations.
To perform these calculations, typical-L designs were reviewed and sorted into.like categories.
l For example, simple cantilevers were placed in-one
{
category, double cantilevers-in another, etc.--For-J each category, calculations were performed to..
determine the load capacity for that particular group
'
i of like supports.
Capacity assessment calculations; L L were documented in Calculation ECE-PSE-020 n
through -022, -026, -029, -030, -034, -045 through -047, -077, and -078.
The last two of these
.
calculations assessed the capacity of Grinnell pipe
{
ring / rod coupling type supports and the ultimate capacity of U-bolts, respectively.
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-The? third and final phase of the-support analysis was-
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.the generation of comparison 1 calculations..During this phase, final support loads generated'during.the j
stress analysis of the bounding sample piping systems j
.
-
weretcompared to the previously calculated support capacity ~1oads to assure that maximum allowable. loads were not exceeded.
These calculations were performed
- on a.stressipackage basis.. Individual-qualifying j
'
calculations: for supports were performed where
_*
required.
Comparison' calculations were documented-in Calculations ECE-PSE-049 through--052, and -054
through -068.
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The; supports-were qualified'on the basis of.two load
_ combinations.. The first, which was optional, was the normal-conditionL(i.e., dead ~ weight plus thermal
.
.
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Lexpansion) loading,-and the.second was the SSE (i.e.,
!
dead weight plus thermal' expansion'plus SSE inertial
,
plus-SSE SAM) loading condition.
Loads were conservatively combined on an absolute' sum basis.
.
Qualification criteria for the normal condition loading were the normal AISC code-allowable' stresses.
.
Those for the SSE loading condition were generally-
[
held to less than 150% of the normal AISC code
- allowable-but otherwise broadly based on Subsection NF-and Appendix F to the'1986 ASME' Code. :The factor of safety for Hilti bolts was generally maintained >at 3, as required by DBD-CS-068, but was-permitted to be as low:as 2 in some cases.
- The NRC. inspector'found that in view'of the purpose of
,
the TU Electric /EQE. program, the acceptance' criteria-for pipe: supports and the criteria for their
'
exceedance on alcase-by-case basis and their.
anchorages.were adequate..The criteria for the-supports will assure that except.for limited localized
yielding due to bending, the; supports.will remain-essentially elastic.
Further, the criteria'for the anchorage might result in localized.ylelding or slippage in the anchorages'only but not catastrophic-
'
failure,
,
The pipe support calculations generated during the course of the support qualification process were identified in Table 4-1 of this TU Electric report.
The type of the calculations and the stress problem or support type group number associated with the calculations were also provided in Table 4-1.
The TU Electric reported that approximately 375 typical designs were assessed as part of this effort.
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Subsequently, to-assess the adequacy of the
'
implomontation of the TU Electric analyses, the NRC inspector-reviewed several of the piping and support-analyses performed by TU' Electric as part of this.
program..Recults for only'some of the reviews
- 4
'
performed are provided in the following to indicate g
their extent.
<
?
(a). Pipe. Stress Calculation ECE-CA-C410, Revision 1,.
dated December 13, 1989, and Pipe Support
..
Calculation ECE-PSE-057,. Revision 0, dated
.
,
k May.27,- 1988.
'
' i These analyses'were for a 100-foot vertical run of 2-inch socket welded carbon steel, service air
,
i.
pipe which was not supported axially except for
1-inch; branch lines approximately 30-feet apart.
The vertical-pipe run.was U-bolted to the
.,
..
containment shell and the branch lines U-bolted ~
'
"
to-internal structure.
The primary' concern was'
with possible (later determined to be 7/8-inch)
', '
differential SSE building motions.
Based on the results of initial analyses, a dead weight support was requested to bo installedLnear the low end of.the vertical run of1the 2-inch
,
pipe.
This support was added to. reduce the
stresses in the piping underLnormal (sustained)
loads.
The NRC, inspector was informed that'this
bounding piping system was unique with respect to
- '
being poorly supported for dead weight loads;
'
the three other systems of'similar configurations.
in Unit 1 were adequately' supported for. dead.
.'
weight.
Based on this explanation, the NRC inspector found thatthe inclusion of;this
modified piping system infthe bounding sample was acceptable, n
In addition, iterative seismic analyses were L-performed for this piping system.
In the initial
"-
analysis the maximum stress (80.8 ksi) in the socket welded joints was less than the 85.4' ksi allowable stress developed by EQE, but the loads e
Np >
on three of the supports exceeded their
/. W '
capacitics.
Subsequently, the piping system was l
reanalyzed with the most " overloaded" support
'
eliminated.
The results of this subsequent
analysis were acceptable.
The maximum socket weld joint stress (50.0 ksi) was less than the
.
l
'
initial analysis, and the loads on the remaining
,
L supports in the analysis less than their capacities.
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Based on'the preceding results of analyses and.
the satisfactoryLresults of review of the hand U
calculations for the SRSS combination-of the SSE.
inertia-and SSE' SAM induced moments,_the NRC
~
inspector found that'the piping analyses-performed for this bounding sample piping system'
had been implemented satisfactorily.
'
,
The NRC inspector also found that the supports I
for this bounding sample problem were qualified, i
on the basis of the worst-case loads obtained
during the iterative analysis.
In addition:.
P
. (1) for rod hangers, downward SSE inertial loads-
were increased by a factor of l'.3 to account for-possible impact due.to lift-off-due to upward SSE:
loads; (2)1where applicable, loads due to self-
. weight excitations factored to include multi-modal effects andiloads due to other piping attached to the support but not part of_the i
bounding sample (" gang hanger") were;also
'
included; and (3) methods for calculating _.
,
stresses-in the supports including base plates-and loads on'the anchorages utilized computer
-
programs approved for use for CPSES safety-related systems.
The support analysis concluded that-all the supports-in this bounding sample system were acceptable except the one "over -
-
loaded" support identified in the iterative.
!
analysis.
{
"
J Lased on the preceding results of review,.the NRC inspector found that the method of analyses:and
.the conclusion of the analysis as stated were
't
-
acceptable.
(b)
Pipe Stress Calculation ECE-Wp-1-C406, Revision 1, dated January 9, 1990:
'
,
,
As discussed in the preceding, the maximum stress in the piping in the. reanalysis of this bounding
- sample piping system exceeded the Lower but.was
.;
less~than the higher allowable stresses pre-H
. scribed by the 1986 ASME Code.
However, y
TU Electric personnel informed the NRC inspector a
hc that this. bounding sample system was a unique-example of rigidly supported branch lines attached to flexibly supported headers.
Further-more, several examples of this type of configura-
tion were included among the 20 bounding samples
.
and shown to be acceptable.
I i
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-
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.
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j On the basis of the preceding, the NRC inspector
'found that this bounding sample was unique and J
,
k utilization of the higher allowable stress
!
prescribed by the 1986 ASME Code was acceptable.
,
(c)- pipe Stress Calculation ECE-CI-X-C419, Revision 0, dated May 18, 1988.
[
This analysis was.for a long apparently
' unsupported length.of instrument air 2-inch
-
copper tubing.
The concern was related to the i
as-built restraint spacing resulting from
'
'
possible missing hangers.
s-However, the analysis, which did not include the effects of supports with excessive gaps, demonstrated that the maximum stress in the tubing was less than the allowable.
Design loads.
on supports not included in the analysis were conservatively based on loads on adjacent supports included in the analysis.
-
,
The NRC inspector found that the piping analysis for this bounding sample system was acceptable.
(d)
Support Stiffness Calculation ECE-pSE-009,
Revision 0, dated March 1988:
,
The NRC inspector found that the methods of
analysis used to calculate support stiffness for Stress Problem C408-and the support stiffness
!
transmitted for use in the pipe stress analysis
were acceptable to the extent reviewad.
Support
';
stiffness calculations were based on methods used i
for safety-related systems and stiffnesses
'
transmitted for use in the piping analysis considered generic support stiffness requirements
'
for safety-related systems.
The use of methods and criteria for safety-related systems was acceptable.
.
Based on the results of the preceding reviews and similar reviews for other bounding sample systems, the NRC
.
inspector found th_t the TU Electric analyses were being implemented in an acceptable manner.
Large Bore piping
The documents reviewed and the results of reviews were similar to those for small bore piping.
The documents received are identified in the following, but only large bore specific review results are provided.
I l
-
_. )
>
-52-
,
(1)_ EQE Procedure 52019.01-P-002, " Field Walkdown x
Procedures, Seismic Adequacy Evaluation of Non-Nuclear Safety Non-ASME Large Bore Piping at Comanche Peak Steam Electric Station," Revision 1, dated March 10, 1989:
This EQE large bore piping procedure was similar to the small bore piping procedure and hence acceptable.
!
'
(2)
TU Electric Procedure ECS-5117, " Assessment of seismic o
Category II and Non-Nuclear Safety Related Piping and i
Supports," Revision 0, dated May 4, 1989.
This TU Electric large bore piping procedure was i
similar to the small bore piping procedure.
,
Previously identified NRC inspector recommendations
'
for the small bore piping procedure relating to i
acceptance criteria also apply to this large bore piping procedure, j
(3)
EQE Report 52019.01-R-001, " Seismic Evaluation of i
Non-Nuclear Safety Non-ASME Large Bore Piping at Comanche Peak Steam Electric Station," Revision 1, dated December 1989:
l This EQE large bore piping report was similar to the small bore piping report except for the following:
(a)
The 25 bounding sample systems were identified in
!
Table 2-2.
Similar to the small bore piping systems, a large percentage of the bounding systems were fire protection system piping.
Pipe l
sizes varied between 2 1/2 inch and 10 inch in i
diameter.
Similar to the sc.all bore-piping
'
systems, the predominant primary supports were
,
,
rod hangers.
Unique supports identified included
tank nozzle, stanchion, saddle and " baling wire" types.
In addition, the presence of aircraft cable type i
restraints was also documented in Table 2-2 of this report.
Their presence was noted in most fire protection system piping.
These restraints were not designed to be active during normal loading conditions but were intended to offer
.
.
vertical restraint in the event of pipe-support failure.
These restraints were considered in the original Seismic Category II over I Systems
,
Interaction program and in the piping system qualification analyses to be performed as part of the initial Corrective Action Program to resolve the subject construction deficiency.
These
.
.F
i
!
!
i-53-
i restraints were not considered in this
'
TU Electric /EQE program; consequently, Class 5
)
piping systems, in which these restraints are i
present and analyzed as part of the TU Electric
,
EQE program, will have added assurance that
'
failures in such systems will not result in
unacceptable interactions in the safety-related
!
equipment or systems.
In addition to the types of joints identified for small. bore piping, cast iron bell and spigot type i
joints in 4-inch and 6-inch diameter drain system-
,
piping were identified in four of the bounding i
sample systems.
Metal band type connections were also identified in a 3-inch diameter bounding system.
Items of concern for the large bore '
'
piping were similar to those for small bore piping.
The NRC: inspector found that the bounding sample systems for the subject large bore piping systems were acceptable on.a basis similar to that for t
the small bore piping systems.
j Of the 25 bounding sample piping systems identified, 13 were analyzed-in 10 piping system
analyses:
3 pairs of the 13 systems identified
were parts of 3 extended runs of piping which were evaluated in 3 of the 10 analyses.
Reasons for not analyzing the remaining 12 systems were i
documented in this EQE report..Thus, for example:
(1) three systems with bell.and spigot
joints and metal type connections were of unique l
l configurations and were rodified to resolve the
l items of concern; (2) two systems attached to vessels were not analyzed on the basis of requiring the Systems Interaction Program assure
,
that the vessels not be dislodged during the SSE; (3) one system was a seismic Category I system in t
l which installation of required support was
,
ongoing; and (4) the " baling wire" support identified in one of the systems was determined to be a temporary construction aid.
The NRC inspector found that the rationale provided for
,
'
not analyzing these systems was acceptable.
However, the NRC inspector found that three systems with aircraft cable restraints were not analyzed on the basis that they had been j
evaluated and found acceptable in the previous
-
qualification program.
The NRC inspector found that although this explanation is acceptable, the
l explanation was not in accordance with the l
l L
'
.~.- -
.
. - -
- - ---- -
__
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!
.
~
-54-i
,
e program guidelines of TU Electric Procedure ECS-S117.
Accordingly, the NRC inspector
recommends that this TU Electric procedure be
modified to provide guidance for not requiring
!
analyses for systems identified as having
concerns.
Similarly, the NRC inspector also
'
,
found that two-other systems containing copper
!
tubing material were not analyzed and no
explanations provided in the report.
_
!
Subsequently, the NRC inspector was informed by i
.
TU Electric that these systems had been qualified
under a separate Instrumentation and Control (I&C) program.
This explanation was also acceptable to the NRC inspector but also not permitted by this TU Electric procedure as
currently written.
'
(b)
The seismic margins assessments included the development of allowable stresses for threaded-joints in stainless and carbon steel piping systems identical to those. developed for small
,
bore piping systems.
These allowable stresses were acceptable to the NRC inspector.
The assessments also included results of EQE analyses
-
'
of database piping systems similar to those provided for small bore piping systems.
Systems analyzed in this report included piping of up to 18 inches in diameter.
Accordingly, the NRC
'
inspector found that the systems analyzed were
,
appropriate for this_large bore piping system program.
(c)
The comparison between CPSES large bore and large
-
bore database piping which had survived earthquakes up to_four times the CPSES SSE was
,
similar to that provided for small bore piping systems and was acceptable to the NRC inspector.
The comparison included bell and spigot type piping.
(4)
TU Electric Report ER-DBE-CS-009, " Assessment of Large Bore Non-Nuclear Safety Non-ASME Piping," Revision 0, i
dated December 1989.
P This report was similar to the small bore piping report.
The report included:
(1) descriptions of the piping and support analyses; (2) analyses acceptance criteria; and (3) results of analyses for the 10 large
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bore bounding sample systems analyzed.
In contrast to the case of small bore piping systems, the analysos for the large bore systems did not
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require the use of iterative analyses for qualification of the systems.
Furthermore, generic support stiffness and capacity calculations were not performed.
Instead, support calculations were performed on a system basis.
The report identified the stress problem and the support calculations for and the maximum stresses in the 10 bounding systems analyzed.
The maximum stress reported for all the systems was 54,591 psi and occurred in Stress Problem Calculation ECE-FP-1-C648.
The NRC inspector reviewed three stress problem calculations and their associated support calculations.
The results of the reviews were similar to those obtained for the small bore piping system analyses (stresses were less than the allowable stresses) and thus were acceptable.
During the inspection, the NRC. inspector also found that the DBD-CS-008 governing this TU Electric /EQE program wab outdated.
The revision current during the inspection, Revision 0, dated May 6, 1988, was not applicable to the new TU Electric /EQE program but did relate to the previous corrective action program for resolution of this construction deficiency.
Subsequently, the NRC inspector was informed that the DBD was in the process of being updated to reflect the new program.
The update was to be accomplished by DCA 93403, Revision 0, which was being processed for release.
In summary, the NRC inspector found that the TU Electric EQE program for this construction deficiency was technically adequate to correct the deficiency but that the two TU Electric procedures for the program will require modifications to be consistent with the program as implemented.
In addition, the NRC inspector found that DBD-CS-008,.when modified by DCA 93403 to reflect the program as implemented, will be acceptable to prevent further occurrences of this deficiency, i
1.
(Closed - Unit 1 only) Construction Deficiency (SDAR CP-87-67):
" Undersized Bolts and Missing Jam Nuts."
This item incorporated the resolution of deficiencies involving undersized bolts and missing jam nuts in I~
structural steel supports.
Previous NRC review of this issue is documented in NRC Inspection Report 50-445/89-74; 50-446/89-74.
This item was left open in the referenced
.
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report pending completion of NCRs resulting from the applicant's review of this issue.
During this inspection period, the NRC inspector reviewed documentation provided by the applicant showing that all NCRs related to this
P l
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I-56-l issue'are.now complete.
Accordingly, this construction
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r deficiency is closed for Unit 1.
j.
(Closed - Unit 1 only)
Construction. Deficiency-(SDAR CP-87-68):
"Hilti Bolt Inadequacies."
This e
construction deficiency involved the incorrect use of Hilti
bolts to anchor rotating equipment and improper torquing of
Hilti bolts.
This issue was previously reviewed by the NRC
as documented in NRC Inspection Report 50-445/89-74; 50-446/89-74, and was left open in that report pending completion of NCRs associated with this item.. During this
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inspectica period, the applicant supplied documentation
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showing that all NCRs associated with SDAR CP-87-68 are now complete.
This construction deficiency is closed for
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Unit 1.
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(closed - Unit 1 only) Construction Deficiency.
(SDAR CP-87-71):
" Missing Welds and Undersized Members."
This issue involved the resolution of' deficiencies
'
regarding undersized welds, missing welds, and undersized structural steel members.
Previous NRC review of this
issue is documented in NRC Inspection Report 50-445/89-743 l
50-446/89-74.
This item was left open in the referenced report pending completion of NCRs resulting from the
applicant's review of this issue.
During this inspection period, the NRC inspector reviewed documentation provided
'
by the applicant showing that all NCRs related to this issue are now complete.
Accordingly, this construction deficiency is closed for Unit 1.
1.
(Closed - Unit 1 only) Construction Deficiency (SDAR CP-87-122):
" Nozzle Load Interfaces."
This deficiency involved nozzle load stresses (for various L
nozzles) which may have failed to meet FSAR limits.
The issue was identified during seismic equipment qualification reviews performed by the applicant.
Specific conditions identified for Unit 1 were all safety-related HVAC fan coil unit nozzles, L
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L l
the hydrogen purge exhaust filtration unit
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nozzles, f
the containment spray heat exchanger shell and
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support structure, and the emergency diesel generator inlet lube oil
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pressure strainer clip angles.
The applicant's evaluation of Unit 2 equipment is not yet complete.
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The applicant has completed its evaluation of the above potentially deficient conditions for Unit I with the i
following results.
Regarding the HVAC fan coil units, the applicant determined that it was necessary to add flexible connections to the fan coil units.
]
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The hydrogen purge exhaust filtration unit was determined to meet the requirements of USNRC Regulatory Guide 1.7, a
" Control of Combustible Gas Concentrations in Containment R
Following a, Loss-of-Coolant Accident."
This allowed the applicant to declassify the hydrogen purge system from an engineered safety feature (ESP) safety-related seismic Category I system to a non-ESF nonsafety-related seismic.
Category II system.
This declassification was reviewed and determined to be acceptable by the NRC.
'
The containment spray heat exchanger shell and support structure were determined to require modifications to eliminate the overstress conditions.
Finally, the emergency diesel generator inlet lube oil
'
pressure strainer clip angles were determined to be adequate to meet "as-built" nozzle loads.
Therefore, the applicant determined the mounting to be adequi2tely
'
qualified and to~ require no further corrective action.
The NRC inspector reviewed the documentation associated with the resolution of each of the above items.
Documents reviewed included (1) calculations performed to determine
'
the acceptability of norzle load interfaces, (2) nonconformance reports issued to document the
,
unsatisfactory conditions, (3) travelers and inspection
~
reports documenting the completion of required actions, and (4) updates to design documents to incorporate the hardware modifications and the declaasification of the hydrogen
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purge system.
From this review, the NRC inspector i
'
determined that the applicant had satisfactorily resolved the deficiencies.for Unit 1 and common areas.
Corrective actions for Unit 2 are as yet incomplete.
Accordingly, the construction deficiency is closed for Unit l'and common only.
m.
(closed - Unit 1 only) Construction Deficiency
-(SDAR CP-87-131):
TU Electric notified the NRC by letter TXX-88177 dated February 5, 1989, of an item relating to
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unacceptable subcompartment pressurization effects resulting from a postulated pressurizer surge line break.
This item was determined to be reportable under the
provisions of 10 CFR 50.55(e).
During a design validation of the steam generator structural steel platforms and the
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reinforced concrete pressurizer support slab, it was determined that the calculated stresses resulting from the pressurizer surge line break would exceed the applicable allowable stresses.
.This exceedance was a deficiency in
,
fin:1 design which was approved and released for construction and also not.in accordance with the design bases and criteria commitments in the CPSES FSAR.
Had the
'
deficiency. remained-uncorrected, structural elements overstressed as a result of a postulated pressurizer surge
'
line. break could subsequently have adversely affected the safety of plant operations.
Subsequently, analyses of the pressurizer surgu lines were performed by TU Electric to eliminate the need to consider the dynamic effects associated with the postulated breaks.
.
These analyses were performed in accordance with the
recently modified Ceneral' Design criterion (GDC) 4 o.
Appendix A to 10 CFR 50 using leak-before-break (LBB)
i methods.
Additional analyses.of the pressuriger surge
.
lines to evaluate the effects of the thermal stratification t
and LBB analyses of the pressurizer surge lines were
,
conducted as part of the TU Electric program for the j
elimination of the evaluation of the dynamic effects for
'
certain main reactor coolant loop branch line breaks using i
advanced fracture mechanics techniques (LBB analyses)
allowed by the broad scope GDC 4 rule change published on
,
October 27, 1987, (52 FR 41288).
Lines included in this i
program were the pressurizer surge, the residual heat
removal suction hot leg, and the accumulator injection
lines.
This program has'been under NRR review and is
scheduled to be approved in SSER No. 23 to the CPSES SER.
contingent on inclusion of this approval in SSER No. 23,
,
the inspector determined that the deficiency has been
resolved satisfactorily.
The recent changes in GDC-4 provide an acceptable analytical alternative to the postulation of pipe breaks..
The cause of the deficiency was determined to be tr.e
,
failure of the original design calculations to include the
pressurizer surge line break loads.
TU Electric determined that this deficiency was isolated since all postulated high energy line break loads were reviewed as part of their
,
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Corrective Action Program and no other omissions of these
loads in plant design identified.
[
As part of the corrective action for this deficiency, TU Electric has noted that design criteria for the consideration of loadings resulting from high energy line breaks were included in the DBDs which were used as the basis for design validation.
These DBDs include DBD-CS-073, -074, -081, -083, -084, and -085.
Future plant modifications will be controlled by these design documents.
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Bared en the preceding, the inspector concurred that the deficiency was isolated and the corrective action
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acceptable to prevent future occurrences of the deficiency.
n.
(Closed ~ Unit 1 on3y) Construction Deficiency i
(SDAR CP-87-132):
" Unqualified Limitorque Actuators."
By
,
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letter TXX-88-027-dated January 5, 1988, the applicant notified the NRC of a deficiency involving unqualified Limitorque actuators.
Specifically, the applicant identified deficiencies such ast (1) unqualified terminal blocks installed in power and control applications, (2) unqualified nylon insulated wire connectors, (3) unidentified.or damsged wire inside the limit switch
,
compartment, (4) cracked limit switch rotors; and (5) moters inside containment which are inadequately
'
drained.
These deficiencies were identified during a review of the equipment qualification for Limitorque actuators and-the performance of FVM TE-EQ-047, "Limitorque Actuator Walkdowns."
The applicant's approach and~ methodology to recolve this issue were evaluated and found satisfactory during an NRC team inspection of equipment qualification.
That
'
ssressment is documented in NRC Inspection Report 50-445/89-60; 50-446/89-60.
'However, not all required work
-
had been implemented and, therefore, the SDAR remained
'
open.
During this inspection period, the NRC inspector reviewed the closure status of the NCRs documenting the deficiencies-for the Limitorque actuators.
These were determined to be
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acceptable for Unit 1 and common although further work is
>
required for Unit 2.
Accordingly, the SDAR is considered closed for Unit 1 with Unit 2 closure dependent on the satisfactory completion of work for Unit 2.
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o.
(Closed - Unit 1 only) Construction Deficiency (SDAR CP-87-133):
"High Energy Line Break (HELB)
Analysis."
This construction deficiency was issued to address various potentially deficient aspects found in the original HELB analysis.
Originally, the-deficiencios were identified and reported to the NRC separately by the following SDARs.
SDAR CP-85-20 Containment Spray Headers - Feedwater
,
Line Break.
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SDAR Cp-85-46 Damage Study Evaluation of Westinghouse
Analyzed Piping.
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SDAR CP-86-13 Jet Impingement Load Review.
)
i 4e SDAR CP-87-53 Pipe Whip Rostraint Design Methodology.
y
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SDAR'CP-87-57 Computer Modeling Error - Compare i
Mod 1A.
- SDAR CP-87-58 Computer Modeling Error - Relap - 3.
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During the performance of an NRC team inspection on
<
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equipment qualification (NRC Inspection Report i
i 50-445/89-60; 50-446/89-60) the applicant's methodology and i
approach to resolve these deficiencies were reviewed and determined to be acceptablo; however, pending-the i
completion of corrective actions, the SDAR was to remain
'
open.
At the time of the equipment qualification
inspection, the SDAR was estimated to be 100% complete for work insido Unit 1 containment and 80% complete for work outside containment.
During this inspection period, the NRC inspector reviewed the updated status of work for the 3DAR.
By letter CECO-3855 dated December 5, 1989, the site Consolidated Engineering Contractor Organization (CECO) stated that all
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interection ovaluations had been completed with the
'
exception of minor restraint adjustments.
Further, the letter stated that.all HELB restraints had been evaluated and determined to be acceptable and, therefore, recommended
,
the SDAR be presented to the NRC for closure.
The NRC
,
inspector asked the applicant for the status of the minor restraint adjustments.
The applicant indicated that three modifications had been mado primarily to provide clearance for thermal and seismic displacements or to remove pipe
whip restraints determined to be not required.
The NRC inspector reviewed the applicable DCAs, construction travelers, and QC inspection reports for each of the modifications.- These documents were determined to satisfactorily verify that the required work had been completed and QC inspected.
Therefore, based on the aatisfactory review of the applicant's approach and methodology performcd during the EQ inspection, CECO letter 3855 stating that all evaluations were complete,
=
u view of the documentation for the completed
)'
modifications, the NRC inspector considers the SDAR closed for Unit 1.
Closure of SDAR CP-87-133 for Unit 2 is pending upon satisfactory completion of similar actions for Unit 2.
p..
(Closed - Unit 1 only) Construction Deficiency
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(SDAR CP-88-36):
" Charging Pump Miniflow Linos."
By letter TXX-89663 dated September 5, 1989, the applicant
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informed the NRC that a deficiency involving the Chemical Volume and Control System (CVCS) charging pumps' alternate
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minimum flow recirculation piping was not a reportable item.
This piping is required to remain intact during a
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Safe Shutdown Earthquake (SSE) and the concern was identified that compliance with this requirement could not
be demonstrated.
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The applicant performed engineering walkdowns of the
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affected lines to verify dimensions, support function,_and
support locations.
A pipe stress analysis-and pipe support
assessment was then performed to determine functional
.
capability.
All commodities met the applicable stress
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l limits with the exception of three pipe supports.which were calculated to yield but not break in an SSE.
Additional
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f calculations redistributing loads from the above supports
(which were assumed ~ removed)'to adjacent supports showed that the adjacent supports would still meet stress limits and that the overall operability of the line would be maintained during an SSE.
The applicant summarized these
'
findings in Engineering Report (ER)-CS-008.
The NRC' inspector reviewed documentation contained in the.
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SDAR' file and ER-CS-008 and concluded that the applicant
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had taken adequate corrective action for the identified problem.
This construction deficiency is closed for
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Unit 1.
q.
(Closed - Unit 1 Only) Construction Deficiency (SDAR CP-88-40):
"NPSI Sway Struts."
This issue involved
,
the resolution of deficiencies involving restriction of sway strut movement to less than the specified degree of freedom.
Previous NRC review of this issue is documented
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in NRC Inspection Report 50-445/89-75; 50-446/89-75.
This
,
item was left open in the referenced report pending completion of the inspection and rework of sway struts
resulting from the applicant's disposition of DCA 87041.
During this inspection period, the NRC inspector reviewed documentation provided by the applicant showing that all inspections and rework related to this issue are now complete.
Accordingly, this construction deficiency is closed for Unit 1.
r.
(closed - Unit 1 only) Construction Deficiency
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(SDAR CP-89-08):
" Fisher Control Valve Actuators."
This
deficiency involved Fisher valve discrepancies.
Specifically, three component cooling water valves failed a
,
I stroke test and two main steam valves were determined to have inadequately sized valve stems.
Corrective action involved the replacement of motors and limiter plates, valve stem replacement, and/or an increase in air supply j
pressure in several of the affected valves.
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Previous NRC review of this issue is documented in NRC Inspection Report 50-445/89-78; 50-446/89-78.
In that j
report, the NRC inspector concluded that the applicant's J
technical disposition regarding this issue was adequate, I
though the item was left open pending completion of the l
Work described above.
During this inspection period, the
"
applicant supplied documentation showing that all work
,
associated with this SDAR was complete.
]
The NRC inspector reviewed this information and checked some of the valves in the field.
One apparent discrepancy
was identified concerning valve 1-HV-2452-1, main steam supply valve to the turbine driven AFW pump.
'n analysis performed by Fisher showed that the air supply for valves 1-HV-2452-1 and 1-HV-2452-2.needed to be increased
-
from 65 psig to 70 psig.
Completed work documents showed
"
that this change had been implemented.
The NRC inspector noted, however, that the local gage indicating the air
,
supply for valve 1-HV-2452-1 read 66 psig.
The gage for
valve 1-HV-2452-2 read 70 psig.
The applicant explained
,
that the local gage for valve 1-HV-2452-1 was known to be
erratic and inaccurate, but that the air supply pressure
!
!
for this valve had been set to 70 psig by calibrated instruments.
The local gage was not used for this i
calibration and is not used for surveillance testing.
The
^
!
applicant stated-that there is no intent to repair this faulty local gage.
The NRC inspector questioned the propriety of not. repairing a local gage which is known to i
be inaccurate, especially one that indicates a
,
safety-related parameter.
The applicant stated that these gages are used plant-wide and are uncalibratable
,
" throw-away" devices installed not to accurately measure a parameter, but rather;to give "go-no go" information to individuals (i.e., system engineers) walking through the
.
plant.
Further, the valves in question are full stroke tested quarterly or prior to entry to Mode 3 and are
,
calibrated whenever rework is performed.
However, the NRC
'
inspector questions whether gauges such as these which are known to be inaccurate should be discarded or tagged to preclude their potential use in a safety application.
This
'
is an open item (445/9003-0-02).
Since all other issues are resolved, this construction deficiency is closed for Unit 1.
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s.
(Closed) Construction Deficiency (SDAR CP-89-11):
" Pipe Support Discrepancies."
This construction deficiency
,
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addressed the high number (1500) of discreparacies found in i
!
safety-related pipe supports by the applicant in response to an NRC violation (445/8912-V-03).
Previous NRC inspection of SDAR CP-89-11 is documented in NRC Inspection i
t l
Report 50-445/89-75; 50-446/89-75.
In this report, this construction deficiency was left open pending NRC
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L-63-j
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inspection of the applicant's room / area and system turnover programs, the commodity clearance program, and the
"
. completion of hardware rework associated with this issue.
The NRC completed inspection of the applicant's room / area j
and system turnover programs and accepted these programs as t'
l-documented in NRC Inspection Report 50-445/89-89; 50-446/89-89.
Also, as documented in this inspection
.
report (paragraph 5), the NRC inspected the applicant's i
Commodity Clearance Program and concluded that this program
,
was effectively implemented.
On January 8, 1990, the
.
applicant informed the NRC that all construction activities
' associated with SDAR CP-89-11 were complete.
On the basis
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of these developments, this construction deficiency is
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closed.
.
t.
(Closed) Construction Deficiency (SDAR CP-89-26):
"IEN 89-56, Questionable certification of Material."
By letter TXX-89775 dated October 23, 1989, the applicant informed the.NRC of a potentially reportable deficiency
,
'
involving questionable certification of procured material.
This' issue was originally identified in NRC Information Notice 89-56.
In-that notice the NRC advised all holders i
of operating licenses or construction permits for nuclear power reactors that corporate officers of two material i
suppliers had been indicted'for their alleged roles in
'
selling commercial-grade steel as being military-grade to
'
the U.S. Defense Department.
The NRC advised that
,
applicants and licensees review any nuclear.procurements from these vendors to ensure that appropriate bases exist for the use of the supplied material.
TU Electric has completed an investigation of the subject
,
!
l of the IE Notice and.has determined that the condition is L
not reportable for CPSES.
This determination was based'on L
the followings (1) TU Electric had issued six procurement
+
orders (POs) to one of the suppliers, (2) of the six POs, one was cancelled and one was used for a nonsafety-related
.
'
application, (3) the certified material test reports (CMTRs) for two of the remaining Pos were validated by the i
l original manufacturers, and (4) samples of the material supplied by the last two POs were tested by an independent laboratory and the results indicated that the steel met the chemical and physical requirements of the applicable
)
material specifications.
l l
The NRC inspector reviewed the contents of the six purchase orders, applicable material request forms, the associated
)
,
E CMTRs, correspondence between TU Electric and six of the
'
material manufacturers, and the results of the tests from e
L the independent laboratory.
Based on a review of these l
documents, the NRC inspector concurs with the applicant's
l
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H
-64-c determination that this issue is nonreportable.
'
Accordingly this item is closed.
l u.
(Closed) Construction Deficiency (SDAR CP-89-32):
,
" Pressurizer Pressure Transmitter."
By letter TXX-89834 dated December 4, 1989, the applicant notified the NRC of a
'
deficiency involving a single pressurizer pressure transmitter installed-in Unit 1.
Specifically, an i
ITT Barton pressure transmitter 1-PT-457 was identified as l
not having been modified to prevent thermal nonrepeatability.
j
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i The applicant has completed an evaluation of this deficiency and determined that the condition did not meet the criteria for reportability pursuant to 10 CFR 50.55(e)
'
or 10 CFR Part 21.
Further, the applicant has completed
corrective actions for the identified deficient transmitter and a review was performed to identify if other transmitters were similarly deficient.
The review determined that pressure transmitter 1-PT-457 was the only
,
affected transmitter.
The NRC inspector reviewed the completed work order which documented the replacement of the unmodified transmitter with a new qualified transmitter from stores.
Additionally, the NRC inspector reviewed office memorandum i
CPSES 8900756 dated December 15, 1989, which documented that a review of other applicable transmitters had been performed.
The review identified no other unmodified
-
transmitters.and therefore concluded that the incident was isolated.
The NRC inspector considered these actions to have been appropriate and concurs with the applicant's conclusions.
Accordingly, this item is closed.
L 5.
Commodity Clearance Program (49063, 49065, 50073, 50075)
Introduction The purpose of the Commodity Clearance Program is to assure that adequate seismic and thermal clearance exists between (1) safety-related commodities, (2) safety-related and seismically supported nonsafety commodities, and (3) certain
.
nonseismic commodities in order to preclude interactions that g
could cause damage severe enough to reduce safety function to an
/
unacceptable level.
CPE-SWEC-FVM-CS-068 (FVM-068), Revision 1, L
dated May 17, 1989, establishes the methods and sequence by which the field verification of adequate clearance is performed
,
and documented.
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I-65-l Scope
)
FVM-068 applies to Seismic Category I structures, and defines commodity as any component, equipment, structure, pipe, HVAC
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duct, conduit, cable tray, support, etc., located in such structures.
The FVM is applicable to safety-related commodities and seismically supported nonsafety commodities to account for potential interactions due to seismic and thermal induced displacements as well as the effects of hot pipe on Class 1E
,
cables.
Nonseismic commoditics not addressed in the Systems
'
Interaction Program are addressed in the Commodity Clearance Program.
r Program Implementation
,
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A walkdown package is developed for each Seismic Category I room / area.
The package includes a Clearance Evaluation Form, which is used to record and disposition deviations from the l
l Clearance Matrix.
The Clearance Matrix identifies the
'
l displacements of commodities due to seismic and thermal motion l
and the required clearance between commodities to preclude an i
l interaction.
The Clearance Matrix is used as a screen for identifying potential interactions.
Further engineering analysis is required to make the determination as to whether the interaction will.actually occur and whether the interaction can reduce safety function to an unacceptable level.
The following are
'
some of the factors considered by the walkdown engineers in these analyses:
,
a.
Stress' calculations are used to calculate pipe
,
displacements.
b.
For Clearance Matrix deviations concerning hot pipe / Class lE cable, attachment 1 to FVM-068 references calculations (e.g., 16345-ME(B)-390) which include
,
algorithms for use in developing the actual pipe surface
,
temperature and/or the target surface temperature.
c.
Commodities are evaluated for damage vulnerability; e.g.,
high damage vulnerability commodities, such as instrumentation and valve operators, are most credibly damaged by an interaction.
Such interactions can be acceptable if it can be demonstrated that the commodity remains functional or an unsafe condition is not created by an assumed loss of the component function.
d.
FVM-068 lists acceptance criteria for situations such as seismically induced impact between commodities of equal size / type, intervening barrier between commodities, source n
has less mass and stiffness (e.g., flexible conduits) than E
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F e-66-E target, source impacts target with glancing blow, restraint to commodity movement, etc.
A walkdown for a room / area is considered complete when the identified clearance deviations are documented and resolved either as acceptable in accordance with the acceptance criteria or by issuance of an NCR and/or DCA to make a hardware modification.
Subsequent to room / area completion, a DCA to Specification CPES-S-1021 is issued to notify engineering, construction, and QC that the room / area has been approved by engineering, and, subsequent to the DCA date, any new work shall meet the requirements of CPES-S-1021.
Since CPES-S-1021 includes the same clearance criteria which is included in FVM-068, its implementation by DCA assures that design work subsequent to the commodity clearance walkdown complies with the same criteria upon which the walkdown was based.
Technical Audit Program (TAP) Engineering Surveillance The inspector reviewed TAP Engineering Surveillance Reports ES-89-09 and ES-89-22.
These reports. documented surveillance by the TAP Engineering Surveillance Group (ESG) to verify the adequacy of implementation of FVM-068 in the identification and resolution of potential interactions involving safety-related commodities.
ESG reviewed Commodity Clearance packages for 17 of the 285 areas included in the program scope.
Portions of six Hot Functional Testing (HFT) stress problem walkdown packages, which documented piping-specific commodity clearance walkdowns performed to support HFT, were included within these 17 area walkdown packages.
ESG evaluated the walkdown packages by performing technical reviews and/or independent walkdowns and by reviewing associated DCAs and NCRs.
Deficiencies were identified in the following categories:
a.
Some deviations from the clearance Matrix and unusual spatial relationships were not being evaluated to demonstrate that functionality is maintained or that an unsafe condition is not created.
ESG walkdowns identified cases where available clearances were less than those required by the Clearance Matrix and unusual spatial relationships were not being evaluated to demonstrate that functionality is maintained or that an unsafe condition is not created.
ESG walkdowns identified cases where available clearances were less than those required by the Clearance Matrix.
b.
Inappropriate or questionable acceptance criteria were used
'
in evaluating deviations from the Clearance Matrix.
c.
A " draft" Revision 1 to FVM-068 was being implemented.
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DCAs were not being issued against Specification
CPES-S-1021 upon room / area completion.
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e.
Clearance Matrix values were inconsistent with supporting
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calculations.
For example, the displacement and clearance requirements for instrument tubing and supports contained
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in the matrix were inconsistent with calculation 16345-EM(B)-023, Revision 1, " Clearance Q
,
Requirements for Tubing and Tubing Supports."
f.
DCAs were being used to authorize some hardware modifications necessitated by clearance evaluation, although FVM-068 stipulated use of NCRs (rather than DCAs)
for this purpose.
g.
FVM-068 did not consider relative building motions.
!
'
'
The NRC inspector reviewed the Consolidated Engineering Contractor Organization (CECO) responses to the deficiencies identified by ESG.
The responses considered the cause, preventive action, corrective action, and generic implications.
Some of the deficiencies were formally accepted as resolved by ESG, while others.were still open at the time of the NRC inspection.
For resolved items, the inspector agreed that CECO
,
actions justified closing the items.
Walkdown Packages Reviewed by NRC The inspector reviewed walkdown packages for room / area 54, 77S I
'
and 88 in the safeguards building.
The inspector also reviewed stress packages 1-069 and 1-052V because the walkdown packages
.
for room / area 77S and 88 stated that these stress packages
!
(required for HFT and located in these rooms / areas) had been walked and evaluated for commodity clearances.
These walkdown
and stress packages were not reviewed by ESG.
The inspector reviewed Clearance Evaluation Forms for both the
[
room / area walkdown packages and the stress packages, and determined that deviations were properly identified relative to the clearance Matrix.
Evaluations were documented for cases where actual clearances were less than those indicated in-the Clearance Matrix.
Such evaluations sometimes demonstrated the
.
acceptability of actual clearances by using pipe stress calculations to provide displacements explicit to the actual commodities, or determined that the interaction would not have an adverse impact on safety.
For cases where interactions were deemed to have an adverse impact on safety, NCRs and DCAs were
-
originated to make appropriate hardware modifications.
The inspector's review of Clearance Evaluation Forms, NCRs, and
,
DCAs for the above walkdown and stress packages determined that
.
IL l
i
in-68-L
.
they complied with the requirements of CPE-SWEC-FVM-CS-068, Revision 1.
Summary l
The inspector reviewed CPE-SWEC-FVM-CS-068, Revision 1, and CPES-S-1021, nevision 0, and concluded that the commodity clearance Program was well conceived and had the necessary i
elements to identify unacceptable interactions within the
program scope.
The inspector concluded that this program was effectively implemented based on his review of the two ESG j
o reports and his independent review of walkdown packages and
-
stress packages.
6.
Follow-up on TU Electric Commitments Resulting from NRC i
Inspections of Corrective Action Procram (46053, 48053, 50073, 50075)
.
Inspection Reports 50-445/87-19, 50-446/87-15; 50-445/87-37, 50-446/87-28; and 50-445/88-29, 50-446/88-25 covered
implementation of the design portion of the corrective Action Program by TU Electric's contractors.
Supplement 17 to the Safety Evaluation Report (NUREG-0797) dated November 1988,
>
Appendix A, details open items resulting from these inspections, responses by TU Electric aimed at resolving these open items,
.
'
and NRC evaluations of the TU Electric responses.
The NRC evaluations in SSER 17 were based upon follow-up
"
inspections performed by the original inspectors.
During the follow-up inspections, the inspectors. reviewed new and revised
calculations intended to resolve the open items, as well as commitments for calculations for the same purpose.
The inspectors closed out some open items if they agreed with the methodology for resolving the open item, even though the
.
methodology had not been fully implemented.
All open items covered in SSER 17 were closed out in that document.
'
' -
TU-Electric letter TXX-89780 dated November 8, 1989 stated that approximately 200 commitments were associated with the above r
inspection reports;'all but five had been implemented.
The remaining five were scheduled for completion prior to Unit 1
-
fuel load.
During the week of December 4, 1989, the team leader for the above NRC inspections followed up on implementation of
{
TU Electric commitments pertaining to 24 open items from the above inspection reports.
These open items covered the following-disciplines civil / structural, electrical, instrumentation / controls, mechanical systems, and systems interaction.
As described below, the inspector determined in
'
each case that the commitment had been adequately implemented.
Since this sampling process addressed approximately 12% of the 200 commitments, the inspector concluded that there is e
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E
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L
-
-69-t
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reasonable assurance that all such commitments would be i
I l
adequately implemented.
!
CS-08
,
,
a.
Documents
.
(1)
Calculation 16345-CS(B)-172, Revision 0, Reactor Makeup Storage Tank
h (2)
Calculation 16345-CS(B)-171, Revision 0, Concrete Design - Condensate Storage and Refueling Water
,'
Storage Tanks.
b.
Commitment i
Revise the above calculations to confirm the adequacy of the method used to approximate the dynamic fluid loads.
l c.
Resolution of Commitment
,
,
The hydrodynamic (seismic) forces for the fluid in the'
tanks were developed in Revision 3 of the above calculations on the basic of TID-7024, " Nuclear Reactors and Earthquakes", August 1963, U.
S. Atomic Energy Commission, Division of Reactor Containment, Chapter G and Appendix F.-
These forces were distributed to the cylindrical tank wall as a horizontal cosine distribution which varies with height in accordance with Appendix F of the TID.
This revision also accounts for the effect of
'
vertical seismic forces on the lateral fluid pressure. The resulting hydrostatic and hydrodynamic forces were applied to a thin shell mathematical model using the SHELL-1 computer code to determine element forces in the walls of the tank.
A comparison of the resulting forces and moments i
in Revision 0 and-Revision 3 showed that the original analysis was conservative. These actions confirmed the adequacy of the method used to approximate the dynamic a
-
fluid loads.
CS-11 a.
Document
-
DBD-CS-081, Revision 0, General Structural Design Criteria
,
b.
Commitment Revise the DBD to state that probable maximum precipitation shall not be combined with seismic loads.
-
_
_
_ _,
.
F~
!
!
!
l-70-t b
!
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c.
Resolution of Commitment i
'
Section 4.3.2 of DBD CS-081, currently Revision 4, has been i
'
revised as shown below:
" Flat roofs with continuous parapets shall be designed for the hydrostatic load resulting.from eight inches of
standing water caused by the probable maximum precipitation (PMP).
This load is considered to be an abnormal extreme environmental load and should not be combined with seismic
,
loading."
l CS-16
!
a.
. Document DBD-CS-081, Revision 0, General Structural Design Criteria
.
b.
Commitment
,
Revise the DBD to clarify the definitions of Ro and Ra to
,
be consistent with the FSAR which includes only pipe reactions.
'
c.
Resolution of Commitment
!
The descriptions of Ro and Ra have been revised to be consistent with the FSAR.
The current wording in Revision 4 is shown below:
"Ro Pipe reactions during normal operating or shutdown conditions, based on the most critical transient or steady state condition.
Ra Pipe reactions under thermal conditions generated by
.
the postulated break and including Ro."
i CS-30 and CS-54 a.
Documents (1)
Calculation 16345-CS(C)-083, Revision 0, Safeguards Building, Unit 1, Wall Design, East - West (2)
Calculation 16345-CS(C)-084, Revision 0, Safeguards Building, Foundation Mat Analysis b.
Commitment Revise slab calculations 16345-CS(C)-070 through 076 to discuss the shear transfer capability of the slabs.
-
s'
..
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_
-
c:
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-71-
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'
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!
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c c.
Resolution of Commitment
'
An alternate calculation, 16345-CS(C)-143, was prepared to L
supplement the above calculations and address the shear transfer capability of the slabs.
This supplemental g
calculation determines the area of steel. reinforcement i
required for in-plane shear loads, combines it with the.
'
area of steel reinforcement required for out-of-plane loads
',
from the original calculations, and compares it to the areas of steel provided for the critical wall / slab interfaces.
This comparison verified'that the slab horizontal reinforcement is adequate to carry the combined.
.
effect of the out-of-plane and in-plane seismic loads.
,
'
CS-40
,
a.
Document Calculation 16345-CS(B)-025, Revision 0, Penetration Anchorage Analysis b.
Commitments
.
(1)
Revise DBD-CS-073, " Concrete Containment Structure",
to limit,the allowable punching shear stress under biaxial membrane tension without additional reinforcement to twice the square root of the l
compressive 3 strength of concrete.
(2)
Revise Calculation 16345-CS(B)-025 to incorporate the above criterion.
'
c.
Resolution of Commitments (1)
DBD-CS-073 has been revised.
The current wording in Revision 2 is "The containment structure concrete and reinforcing steel shall be in accordance with ACI-318, Section 11.10.3 for punching shear, except in areas of biaxial membrane tension where the allowable punching shear stress is further limited to twice the square root of the compressive strength of concrete without
'
the inclusion of additional reinforcement."
,
(2)
Revision 3 of the calculation has been issued and the shear has been checked for an allowable of twice the square root of 4000 psi, or 126 psi.
t
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- _.
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Im
Y i
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--
r-72-
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T CS-42 i
a.
Document Calculation 16345-CS(B)-067, Revision 0,^ Moment
,
' Distribution for Beams and Columns in Floors at Elevation
,
832'-6", 806'-0", and 905'-9" (Containment concrete i
Internal Structures).
,
b.
Commitment
,
,
The above calculation will be revised to provide an adequate description of the limitation and use of the results given on page 29 of the calculation.
c.
Resolution of Commitment The summary of the fixed end moments (FEM) on page 29 of calculation 16345-CS(B)-067, Revision 0, indicates that the FEM coefficients at the beam to wall interface are for information only.
A description of the limitations on their use was added in Revision 1 and is summarized below.
l The vertical load distribution analysis performed in Calculations 16345-CS(B)-088 to 091 and 096 was based on
the assumption that radial beams were fully fixed at the steam generator compartment walls and pinned at the perimeter columns.
The same approach was used in Calculations 16345-CS(B)-029; -075; -072; -076 - analysis and capacity evaluation of beams (and slabs) for floor elevations 832'-6"; 860'-0"; 885'-6" and 905'-9",
respectively. 'This assumption resulted in a conservative
!
estimate of forces and moments for structural adequacy review of the= beams and compartment walls.
However, this approach is not conservative for the structural adequacy review of the perimeter columns, since it~results in decreased axial forces and eliminates bending moments to the columns. Therefore,-it is necessary to provide a calculation to address the effect of the' beam-column moment connections.
Calculation 16345-CS(B)-067 provides moment coefficients to be used in adjusting the previous vertical load distribution analysis results for evaluation of the
'
perimeter columns. Results of the analysis are used in Calculation 16345-CS(B)-020 for the structural adequacy review of the internal structure column II
'
-73-i
>
CS-46.
,
E a.
Document h
Calculation 16345-CS(C)-130, Revision 0, Reactor Building -
Containment Shell Design b.
Commitment Revise the above calculation to show that the additional l.
stress in the vertical reinforcing bars resulting from the.
radial shear is within the allowable rebar stress.
L c.
Resolution of Commitment Revision 2 of the above document calculates the additional rebar stress in accordance with ACI 318-71 Section 11.15,
" Shear Friction", and shows that the rebar stresses are within allowable limits.
CS-48 a.
Document Calculation 16345-CS(C)-130, Revision 0, Reactor Building -
Containment Shell Design b.
Commitment Revise the above calculation to demonstrate that no additional reinforcing is required to maintain the hoop bars in: place and that adequate bond exists between the concrete and the rebar to-resist the load which would tend to spall the concrete.
'
c.
Resolution of Commitment Revision 3 of the calculation shows that the dome can resist the pressures exerted by the hoop bars, the concrete will not split due to these pressures and the meridional-bars can transfer their force adequately to the concrete through bond stress without failure.
This calculation is based on equations from a technical paper attached to the calculation entitled "Rebar to Concrete Forces in Membrane
.
- Behavior Regions of R/C Domes".
CS-49
a.
Document L
Calculation 16345-CS(C)-130, Revision 0, Reactor Building, y
Containment Shell Design c
e
'!
L
!
f-74-i i
b.
Commitment
!
'The above calculation will be revised to demonstrate the I
adequacy'uf the rebar development lengths provided using a quantitative justification based on actual bar stresses.
c.
Resolution of Commitment Revision 2 of the calculation determined the maximum stress for the vertical #18 bars in the base mat using the NEWSECT
"
computer program.
Once the stress was determined, the required development length was calculated and compared with the provided development lengths.
This comparison demonstrated that the vertical 618 bars have adequate anchorage.
!
CS-51 a.
Document l
Calculation 16345-CS(C)-086, Revision 0, Column Design,
,
Safeguards building
b.
Commitments (1)
DBD-CS-081, " General Structural Design Criteria", is to be revised to include final design jet Impingement
.
'
loads and design inputs for evaluating structures for impacts'due to pipe breaks.
(2)
The CPSES structural elements are to be evaluated for the above loads.
(3)
Each column is.to be evaluated for the pipe whip restraint loads given in DBD-CS-081, attachment 4.
!
(4)
The' load verification program will have random samples of column elements with more than five significant
'
,
attachment loads (exclusive of pipe whip restraint loads) per Project Procedure PP-210.
c.
Resolution of Commitments (1)
The design inputs for jet impingement and pipe impact
>
loads have been developed and were added to DBD-CS-081, attachments 5 and 6 respectively by DCA 84569, Revision 1.
(2)
The CPSES structural elements are evaluated for the above loads in Calculations 16345-CS(B)-113, -145,
,
and -146.
,
-
_, _ _ =-
-
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-* -
y
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"
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,
>
!
-75-t (3)
The CPSES columns are evaluated for the effects of
pipe rupture restraints in the calculations referenced
'
!
in 2. above.
(4)
The original response to this item indicated that the load verification program would have a random sample
.
of column elements with more than five significant j
,
attachment loads.
The load verification program for
columns was revised to perform a walkdown of all
'
L colunm elements and select the two worst cases,,1.e.,
the most heavily loaded with respect to attachments.
p This program is described in PP-210.
The selection of
'
F the two worst case columns is documented in calculation 16345-CS(C)-638.
The analysis of the two
.
,
L" columns'(documented in calculation 16345-CS(S)-640)
confirms that the column designs, which were validated
based upon DBD design criteria, remain valid when
.
,
actual attachment loads are.used. Since these are the most heavily loaded columns, this conclusion can be
!
'
extrapolated to other columns which were validated
-
based upon the DBD criteria.
l
!
E-1 l
a.
Document Calculation 16345-EE(B)-031, Revision 0, Protective Relay Settings for 6.9 kv Safeguard Buses
'
b.
Commitment For the 1,000 hp component cooling water pump, an overcurrent relay tap setting of 120 percent was selected.
For the auxiliary feedwater pumps.and the station service i
water pumps, 124 percent was selected.
For these pumps, relay settings will be increased to a minimum of 125 percent as part of the final relay setting program.
c.
Resolution of Commitment DBD-EE-051 has been revised to specifically require that the overload protection for 6.9 kV Class lE motors be 125%
of full load amperes.
Calculation TNE-EE-CA-0008-265 has
\\
been revised to ensure that all 6.9 kV Class lE motor
/
overload protection devices are set at 125% (minimum) of full load amperes.
.
>
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..
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_
_
_
.., _
_
,
.
-76-i t
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E-13 a.
Document.
!
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Calculation 16345-EE(B)-069, Revision 0, Voltage Drop Verification - Miscellaneous DC Control b.
Commitment
- -
.
All vendors have been requested to provide device
!
I information for the validation of components.
In-rush currents, if different from steady state current, will be.
,
included in this information.
'
c.
Resolution of Commitment Device data from vendors was reviewed to establish in-rush
-
and steady state values.
Calculation 16345-EE(B)-69 has been revised to update vendor data for devices to reflect
'
in-rush / steady state current values and to account for
,
,
in-rush current when it differed from steady state current.
l E-16 a.
Document Calculation 16345-EE(S)-147, Revision 0, Cable Sizing l
Calculation - DC System r
b.
Commitment Calculation EE(S)-147 will be modified to the extent required by the determination of in-rush currents as described in the response to open item E-13.
,
c.
Resolution of Commitment Calculation 16345-EE(S)-147 was issued for a design change involving solenoid operated valves 1HV-2397, 2398, 2399-and 2400.
Vendor data received did not distinguish between
'
in-rush and steady state current and therefore 16345-EE(S)-147 was not revised.
However, this calculation has been voided and its contents are incorporated into
. Calculation 16345-EE(S)-69.
I-10 l
r.
a.
Document l:
DBD-ME-229, Revision 0, Component Cooling Water System
\\
Ilm
.
_~.
r
'
i L
'
-77-
b.
Commitment The DBD will be revised to include a scenario for isolating the nonsafeguards. loop should one safeguards loop fail.
i c.
Resolution of Commitment r
I The DBD revision indicates that single failure of one CCW pump, while. operating in a cooldown mode or upon an
"S" signal, may cause the remaining operating pump to operate
'
in the runout condition and may result in flow deficiencies
.
to safety-related equipment.
This condition can be alleviated by hydraulic load trimming to reduce the pump
>
flow.
Trimming may be accomplished by alignment of
,
components to the other operating unit, by flow termination
or by throttling.
L I-22 a.
Documents
,
' Calculations 16345-IC(B)-029 and 030, Revision 0, Station
L Service Water Supply Header Pressure Lo Channels 1-PIS-4250 i
and 1-PIS-4251 b.
Commitment l
The setpoint calculations for the remaining 24 ITT Barton
-
switches will be revised to-establish maximum reset points.
c.
Resolution of Commitment l
I DBD-EE-037, Revision 3, " BOP Safety Related Setpoints" has
been revised to include the following statement l
l
"For all instruments having fixed or non-adjustable
deadband, the maximum reset point (based on the maximum possible actuation point) shall be determined and checked to assure that it does not adversely affect system operation."
L All setpoint calculations for ITT Barton switches have been L
reviewed and/or revised.
The Instrument Setpoint t
\\:
Calculation Acceptability Checklist specifies minimum and
maximum reset values.
The maximum reset values for all'ITT Barton setpoint calculations have been checked relative to the normal process operating ranges to assure that system operation is not adversely affected.
,
k l
- -
.
-
,
't s
.
-78-i
.
F-33
,
a.
Document
,
Calculation 16345/6-NU(B)-023, Ultimate Heat Sink and Maximum Sump Temperature
,
b.
Commitment
,
The component cooling water heat exchanger performance will be monitored using a testing program which is being l
developed.
The instrumentation accuracies will be evaluated by standard error analysis to assure that the monitoring program accurately predicts heat exchanger performance.
The test data obtained will be analyzed to correlate to a fouling factor for the-heat exchanger.
This
fouling factor will be compared to a set of curves which provide acceptable fouling factors corresponding to SSW
.'
inlet temperatures. If the fouling factor is determined to i
be unacceptable, then the affected component cooling' water-
.
^
system train will be declared INOPERABLE and the appropriate Technical Specification requirements satisfied.
t The CCW heat exchanger tubes will be cleaned.
A retest of
,'
the performance will be made to verify an acceptable fouling factor.
c.
Resolution of Commitment Calculation 16345-ME(B)-609, Revision 2, " Performance
' Prediction and Fouling Factor Determination of-CCW Cooler",
predicts during normal' operation when the CCW cooler requires cleaning in order to be capable of meeting t
accident performance requirements.
Test Procedure Number
'
..
l EGT-339A,. Revision 0, "CCW Heat Exchanger Fouling Monitoring",. addresses the test for obtaining temperature.
t i
l and flow data from the CCW heat exchangers for evaluation to determine when cleaning-is necessary.
All test instrumentation is in place, and-the actual tests will be implemented at a frequency (monthly, weekly, or daily).
dependent upon proximity of the test data to the curve of unacceptable fouling factors.
F-44 and F-46
'
a.
Document Calculation 16345-ME(B)-196, Revision 0, December 31, 1987, CCW Worst Case Non-Seismic Pipe Break i
-,
-
- -.
.
~
~
~
l N,,jf, j
1
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e-
-
9T(
"
-
'
-.
,
o
.
,
>
v i
L_ l b.:
" Commitments i
'(1). The posision for. valve XCC-080 will not be set-at-
.
32 degrees.open which-cannot be set accurately.in the.
l field.
Instead, the valve-position will be adjusted based'on.flowLrequirements such-that the flow required
'
'for chiller operation is obtained during system. flow balancing.
Administrative controls, as described in'
i s
the response to Item F-49, will-be_ implemented by TU
!
Electric to' restrict' valve operation.
-
i (2)
.CalculationLNo. 16345-ME(B)-196 was performed to-
!
< maximize the flow out of the nonseismic pipe break..
r
"
-
" Loss of NPSH during this event was not a concern for;
~
this: operating condition.
The reduction of suction 1:
. pressure for the-length of time postulated-(30 i
seconds) is not a severe. transient'and is expected to
.
be acceptable.. The pump vendor will be contacted'to-
,~
verify that this transient is acceptable.
i c.
-Resolution of Commitments
The above two commitments were based upon the assumption, in the above calculation, of two guillotine-breaks in 10-inch nonseismic piping to and from ventilation chillers.-
Further. review. concluded that'a full guillotine break was not required to be postulated and a. reanalysis, based upon the worst case: crack, determined'that this failure'would.be enveloped by other postulated breaks in the system.'The new analysis (Calculation 16345-ME(B)-73 and memo.SW-2503) was provided~to support this conclusion.
'
-F-49 y
a.
Jocument-l Calculation.16345-ME(B)-267, Revision 0,. December 10, 1987, Component Cooling Water System Flow Distributions-i b.
Commitment
,
DBD-ME-0229 will be revised to include throttle valve requirements and the CCW flow diagram will be revised to-reflect.the; locking requirements.
c.
Resolution of Commitment DBD-ME-229 and the CCW flow diagrams (M1-0229-CP-9-AD-4, i
M1-0229-CP-5-AD-4, and M1-0229-ACP-5-AD-5) were revised to-implement the requirements for locked-in-place valves.
,
..
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I'6[
.
'
'
[~'
'
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,
'
a
,
pf f
.-80-
';
e
,
.
y
- F-50
m
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a.-
Document:
.
,
Calculation 16345-ME(B)-166, Revision 2,-February! 25, 1988,
.
p'
Effect on CCW System of a Thermal Barrier Tube Rupture
- "
b.
Commitment-
In lieu of further transient analysis, water hammer
,r+
analysis, and changes to FSAR radiological results, a L
design change is.being implemented to automatically isolate i
the event rather than relying on later manual action.
t L
Redundant, automatic isolation valves'and detection-instrumentation will provide timely isolation to limit the amount of steam entering CCW and eliminate any significant affects on the safeguards portion of the system.
.
c.
Resolution of Commitment DBD-ME-229 and component cooling water system flow diagrams M1-0231 and M1-0231A were revised to implement automatic j
redundant isolation capability.
Installation of valves and instrumentation was accomplished under Traveller
CP1-ECPROR-20 and DCAs 65520 and 73773.
!
S-16
.
a.-
Document Calculation CPE-SI-CA-0000-666, Revision 0, December 2, 1987, HELB' System Analysis - Room 113.
'
b.
Commitment
,
This calculation is being revised to clarify the steam generator blowdown (SGBD) break mitigation scheme.
i
'
c.'
Resolution of Commitment Calculation CPE-SI-CA-0000-666-has been revised to clarify
.the.SGBD mitigation scheme.
Additional explanation in this
.
regard is provided on P.
30 of the shutdown analysis calculation CPE-SI-CA-0000-665, Revision 3.
i S-18 a.
. Document Calculation CPE-SI-CA-0000-714, Revision 0, October 26, 1987, Pipe Rupture Analysis - Auxiliary Feedwater System Outside Containment, Problem 10 B and C Unrestrained.
,j >'
mL b
.
if-81-b.
Commitment
,,
'
Procedure CPE-EB-FVM-SI-34, Revision 1 did not explicitly
.
' require pipe whip restraints to be-listed as HELB targets.
-
In-order to eliminate any possible confusion, Procedure
-
FVM-SI-34 i'4 currently being revised to explicitly require pipe whip restraints associated with the source line to be m
identified as targets.
In addition,'all pipe whip
' restraints within the zone of influence of a HELB shall be reviewed to ensure that jet loads are considered, if i
h appropriate.
!
l To evaluate the s.
ificance of failure to account for the
-.
actual load on a-pipe whip restraint, Ebasco is currently
'
identifying all design load combinations for the pipe whip restraints.
These loads will be compared to the maximum
design loads of the restraints.
Any-corrective action q
required as a result of this evaluation will be documented in the final closure report of SDAR CP-87-133.
'
c.-
Resolution of' Commitment Proc'edure CPE-EB-FVM-SI-34 was, revised to require the identification of pipe-whip restraints associated with the source'line as. targets.- 'All'HELB restraints have been-
,
evaluated as acceptable.
These evaluations considered all design loads including jet impingement loads.
]
S-22 a
a.
-Document
.
!
!
Calculation CPE-SI-CA-0000-779
L I
)
b.
Commitment This calculation will be revised as follows:
-
,
,
!
(1)
The objective will include addressing of mass point y
spacing criteria.
(:2 )
Page 33, Section 7, will indicate that there should be
!
lumped mass in the model to represent concentrated
\\.
mass for valves and valve operators and other piping i
r
. system in-line components; e.g.,
strainers.
I i
'
c.
Resolution of Commitment L
Calculation CPE-SI-CA-0000-779, Revision 3, addresses mass point spacing criteria as the objective of its analysis.
l t
L- -
{EN
'
.
k j
'
-82-V:
PageL33 of the calculation indicates that there will-be
-)
lumped mass in the'model to represent concentrated mass for piping ~ system in-line components.
7.
Testing of piping,' pipe supports, and restraint systems durino j
"
'
hot functional testing (70370)-
During hot functional testing 7 n 1989, the applicant performed
extensive testing and monitoring of:-
(1) pipe vibrations resulting from various operating conditions, and (2) the displacements of piping, pipe supports, and pipe restraint systems due to thermal expansion at various reactor coolant system (RCS) temperature plateaus.
'The NRC inspector discussed-these activities with the applicant's test engineers and reviewed the inspection requirements and records.
,
Pipe Vibrations
'
Pipe vibration monitoring was controlled by Test Procedure 1CP-PT-90-01, Revision 0, " Piping Vibration Monitoring," and Project' Procedure CPPP-25, Revision 2, " Piping-t Vibration Test Procedure."
The objective of the piping vibration monitoring was to verify that the vibration level of selected Class 1, 2,'3, and other high energy piping systems was acceptable.
Both. steady-state and operational transient conditions were monitored.
Procedure 1CP-PT-90-01-requires a displacement criteria evaluation if-the steady-state vibration velocities exceed'
O.5 inches per second_(ips) peak to peak.
A total of 67 velocity field measurements were made using Procedure
,
1CP-PT-90-01 oftwhich 26 exceeded the 0.5 ips limit.
of the-L 26 measurements requiring evaluation of.the displacement l
criteria, only one' failed to pass both'the velocity and
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displacement criteria.
Test Deficiency. Report (TDR) 8220 was written to address steady-state vibrations for the motor driven-auxiliary feedwater Train B test line flow mode.
TDR 8220 was closed by NCR 89-9188 and dispositioned "use as is" by
- engineering.
Transient brations were monitored both by visual observation and by com;tring measured loads at selected locations with the j
calculated loads.
TDRs 7417 and 7419 address an
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(:
out-of-tolerance condition for the pressurizer relief valve
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_ transient testing where measured loads exceeded the acceptance criteria at three load-pin locations.
These conditions were-addressed by NCR 89-6405 and dispositioned "use-as-is" by engineering and Westinghouse after subsequent retesting.
TDR 7374 documents the failure of two loading measurements to meet acceptance criteria during the transient testing of.the turbine driven auxiliary feedwater pump piping.
TDR 7374 was L
dispositioned "use-as-is" after extensive retesting and l-l
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4-83-engineering analysis of the results.
The deficiency identified in TDR 7374 was due to the multiple failure of Borg-Warner check valves.
TDR 8677 addresses the use of-visual monitoring on the
- safety injection system in lieu-of instrumented measurements due 1to a data acquisition problem at one specific location.- No other significant problems were identified during transient vibration testing.
The NRC inspector discussed the piping vibration tests with the
' applicant's test engineers and reviewed the above procedures and TDRs.
In addition, the NRC inspector reviewed portions of-the-test logs, data sheets, and associated _ calculations comparing these with the requirements stated in the Comanche Peak FSAR (Section 3.9B.2 and Table 14.2-2).
The NRC inspector concluded that-the tests were performed according to the FSAR with adequateLprocedures and records.
No concerns were identified.
Thermal Expansion Monitoring of the displacements of piping, pipe supports, and
pipe restraint systems due to' thermal expansion at specific RCS temperature plateaus was controlled by Test Procedure-lCP-PT-55-ll-SFT, Revision 0, " Hot Functional Thermal Expansion."
The objectives included:
to_ verify that selected systems (see FSAR,
.
Section 3.9B.2.1.1) can respond thermally without obstruction or interference.during system heatup from ambient conditions to operating conditions and return to ambient conditions.
to verify that snubbers and spring cans remain within
.
their working range.
'
j to verify that pipe whip restraints are properly set.
.
. Pipe displacements were primarily monitored using an automated
!
data acquisition system programmed to measure displacements at 226 locations.
Measurements of the positions of snubbers, springs, and whip restraints were done by inspection.
Following completion of the-prerequisite portion of tge test procedure,
-
the plant was heated up at approximatelg 50 F ger hou5 and held at tempgrature plateaus of ambient, 250 F, 350 F, 450 F,
.and 557 F.
Thermal expansion measurements, visual walkdowns, and data collection were conducted at these temperature plateaus.
During the test, 300 problems were identified on Problem Reports resulting in 137 test deficiency reports (TDRs).
In addition, 33 pipe whip restraints required gap adjustment.
All TDR's have now been completed.
The problems identified included insulation
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tand scaffolding interference, support binding, inadequate pipe l
whip restraint gaps, and pipe monitor points out-of-tolerance.
q The NRC inspector discussed the thermal expansion. monitoring'
with the applicant's test engineer and reviewed Procedure 1CP-PT-55-ll-SFT and the associated test logs. 'In addition, the
. NRC inspector-reviewed the followings.
'
.
Problem' Sheets P-001 to P-047 and associated TDRs.
.
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Problem Sheets T-001 to T-030 and associated TDRs.
.
A selected sample of Predicted Data Sheets for pipe
[
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node points, snubbers, and spring cans.
,
A selected sample of Data Sheets which record'the
.
,
positions of snubbers and spring cans (primarily for the main steam system).
Piping displacement data'recogded by the data
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acquisition system at the 250 plateau.-
Datasheetsforsnugberandspringcanpositions
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recorded'at the 250 plateau.
Two concerns were identified.
Problem Sheet T-30 recorded that several (12) monitors for pipe displacements showed
,
out-of-tolerance conditions. -The resolution for monitor
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MS-1-002-04X was' missing the calculated. displacement value where
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the value was being compared to the measured value (-0.55 inches)..The calculated value-(-0.52 inches) was
subsequently supplied to the NRC inspector'and entered into.
L Problem'SheetLT-30 by the, test _ engineer.
This concern was
determined by the NRC. inspector to.be an isolated oversight.
'
The second concern-involved'the small amount gf snubber.and L
spring can position data collected at the 250.F plateau.
Only data from snubbers and spring can positions for the steam generator supports and the RCS-supports was recorded as directed by Westinghouse.,Apparently, no pipe support positionLdata was recorded for other systems. 'FSAR Table 14.2-2, Sheet 52a, requires the applicant to " Inspect sgubbers and spring cans at l
s g
g specified temperature intervals (250 F, 350 F, and 450 F) to ensure their expected thermal movements and swing clearances are
such that from cold to hot condition and return to ambient
<'
L condition are within the criteria per applicable design
'
drawings."' The applicant stated that the extensive inspectiong g
350 F,gand 557 F of snubbers and springs at ambient temperatu5,and 450 F.
bounded the conditions which occurred at 250 F Also, according'to the applicant, theFSARdoesnotspgcificafly require that all inspections be performed at 250 F, 350 F, and 450 F.
These temperatures are given inside a parenthesis i
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ffz t
t w-85-N,
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- and represent suggested values.- The NRC inspector concluded Si
'thatfthe applicant met the intent of the FSAR-requirement.
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Except.for the above concerns, the NRC inspector concluded that the' monitoring was performed according to the FSAR'with adequate procedures.and records.
2
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....
(50071, 50073, 50075)
8.
Mechanical Components and Equipment a.
On January 5, 1990, subsequent to the completion of repair activities conducted in June 1989 on valves 1MS-142 and
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1MS-143, the-applicant performed radiographic examinations to confirm the orientation of the disc to the seat.
It is noted that similar Borg-Warner swing check valve failures
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in the auxiliary feedwater system, discovered during hot functional testing, were the subject of an Augmented Inspection Team inspection, as documented in NRC Inspection
'
Report 50-445/89-30; 50-446/89-30 which resulted in
,
escalated enforcement action.
The radiographs indicated that the valve disc for valve 1MS-142 was laying off the
- i seat and that the disc for valva 1MS-143 was lodged under.
the seat ring.
Both of these valves-are Borg-Warner, t
4-inch pressure seal swing check valves which are located-on the main steam supply lines from steam generators Nos. 1 and 4 to the turbine driven auxiliary feed water pump.
The significance of this occurrence was that valve 1MS-143 was stuck open without ever having experienced forward flow
conditions. 'Thus, the disc had-apparently become lodged under the seat. ring during a reassembly process.
On-January 11, 1990, the NRC inspector witnessed the
,
disassembly of valve 1MS-143.
This activity was conducted-in accordance-with Maintenance Procedure MSM-CO-8801,
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Revision-2"and Work Order C900000202.
Subsequent to the L
disassembly of valve 1MS-143, which did-not reveal any significant deficiencies, the-applicant reassembled the-valve in accordance with the above reference procedure.
In-addition to evaluating the applicant's trouble-shooting activities relative to valve 1MS-143, the NRC inspector reviewed' Technical Evaluation 1WC-90-79 and witnessed major portions of the reverse flow air testing for valves:1MS-142 and 1MS-143 which was conducted in accordance with Test-
. Procedure EGT-740A, Revision 0.
As stated by the applicant in the above referenced
'
L technical evaluation, valve 1MS-143 was reassembled taking L
care to insure that the disc did not lodge under the seat.
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L Furthermore, it was stated that "All Borg-Warner pressure L
seal check valves may be susceptible to the same condition.
L However, the remaining valves have been shown by forward l
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-86-andireverse flow testing and informational RT's not.t'o exhibit this problem generally."
In the-as-left-condition both valves 1MS-142 and 1MS-143
'
had been successfully air tested in the reverse flow
direction; however, it is noted that both of.these valves i
are scheduled'to be operationally: tested during startup.-
b.
The'NRC' inspector also evaluated the applicant's response to PIR 89-243.
In particular, this PIR identified several pieces of foreign material which were discovered in the-piping immediately down stream of check valve 1CT-145.
This check valve, which is a Borg-Warner, 16-inch, bolted-bonnet, swing. check valve is located downstream of the Containment Spray Heat Exchanger CP1-CTAHCS-02 and;is-on the. supply line to the Train B containment spray
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D headers.
The foreign material which was identified
consisted of a grinding wheel approximately 3 3/4 inches in diameter, a dressing stone approximately i
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1 1/2 inches X 2 inches, and grinding grit.
This' issue was addressed by the applicant on 50.55(e)'
Reportability Evaluation Form SN-463 which concluded that this deficiency was not reportable.
The basis for this
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determination was that this condition did not represent ~a
significant breakdown in programmatic requirements.
The
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justification-for nonreportability also stated that the
,
most probable activity that could-have introduced the-
' debris was a 1983 rework of the up-stream heat exchanger.
This justification also indicated that adequate cleanliness controls were specified on.the traveler utilized to perform the 1983 work,and that this was an inadvertent incident not
'
attributable-to a programmatic breakdown.
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In that the applicant's conclusion that the system cleanliness controls were adequate did not correspond to the reported condition; the NRC-. inspector requested the
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engineering' evaluation that was used to arrive at the determination of.nonreportability.
The applicant responded
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that.although they had performed a preliminary engineering i
assessment to address this deficiency, the formal r
documentation was not retained because the issue was determined nonreportable.-
Subsequently, the applicant
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provided an amendment to SN-463'which provided additional
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information pertaining to the nonreportability determination.
Additionally, the applicant maintained that as a result of the implementation of corrective actions
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associated with Borg-Warner check valve failures (see NRC Inspection Report 50-445/89-30: 50-446/89-30) that numerous check. valves had been recently disassembled and that no
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other examples of foreign material had been identified.
However, it was established by the NRC inspector that there
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were'no programmatic controls specified in'the Borg-Warner
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Check Valve Maintenance Procedure MSM-CO-8801,. Revision 2,-
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which provided verification of the system cleanliness other
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than in the valve body.
The NRC inspector reviewed the amendment to SN-463 which stated, in part, that the polishing wheel could have blecked the flow to one of the horizontal spray nozzle
,
headers which was outside the bounds of the containment
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pressure / temperature analysis.- In that this event
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represented an unanalyzed condition, the applicant
' i concluded that the safety of operation of the plant could
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have been adversely affected.
With the initial criteria of
10 CFR 50.55(e) satisfied-(i.e., adversely affecting the safety of operation) the applicant then proceeded to evaluate and eliminate the remaining criteria which need'to be satisfied in order to establish the technical requirements for reportability.
The applicant concluded
that although this deficiency met the first criterion of
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adversely affecting the safety of operations, it did not-l met any one of the criteria in paragraphs -(i), (ii), (iii),
i or (iv); therefore, did not meet the criteria set forth in 10 CFR 50.55(e)'for notification to the Commission.
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The NRC inspector reviewed the applicant's amended response to SN-463.and determined that it was unacceptable.
This-
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determination was based on the apparent failure of the
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applicant to adequately consider the implications of a
potentially degraded safety-related system with regard to a
Specifically, this paragraph
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provides.for notification of a significant deviation from
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performance specifications which will require extensive l
evaluation, extensive-redesign, or extensive repair to l
establish the adequacy of.a structure, cystem or component
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to meet the criteria and basis stated in the safety l
analysis report or construction permit or to otherwise
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establish'the adequacy of the structure, system, or component to perform its intended function (emphasis added).
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The applicant's justification that the retrieval of the
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debris removed the unanalyzed condition thus allowing the
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- containment. spray system to perform its specified safety l
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function, overlooked the generic implications of what
allowed'this loss of' system cleanliness to occur, where else it occurred,'and what was done to verify that
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programmatic controls were adequate.
The applicant's failure.to report a deficiency in accordance with 10 CFR 50.55(e)', and to perform adequate corrective-measures involving a significant condition adverse to quality is-identified as a violation (445/9003-V-03).
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9.
Allegation Follow up (99014)
(Closed)! Allegation (OSP-89-A-061):
As previously documented in NRC Inspection Report 50-445/89-64; 50-446/89-64, this
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allegation involved a former worker's' concerns relative to-safety issues which were identified to members of'the NRC resident staff at Comanche Peak on July 13, 1989.
Specifically, the two items which remained open at the conclusion of the above referenced report involved plant electrical components and
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systems.
The following dispositions are;provided for these L
issues:
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(a)
The alleger identified a concern regarding the. utilities
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apparently erratic preventative maintenance practice.
!
4 relative to the cleaning and preservation of. lighting and switch gear panels.
The alleger stated that his experiences were that some panels were cleaned / dusted on a regular basis while others were not maintained.
This concern was general iri nature in that the alleger could not specify any specific electrical panels in the plant that were affected by this practice.
,
The NRC inspector reviewed the applicant's. preventive
,I maintenance program and preservation schedule for lighting and switchgear panels including electrical breakers.
This
program defines the periodic surveillance testing and
'
maintenance to be performed both for safety-related and nonsafety-related equipment and components.
The NRC.
inspector examined a sample of the controlling procedures and verified the completeness of the. corresponding records which established the frequency 1of these activities.
b1 Based on these reviews, the.NRC inspector concluded that i
- i the applicant's program for-preventive maintenance of-lighting and switchgear panels appeared.to be adequate.
Therefore, this portion of the allegation could not be substantiated.
No further action is recommended regarding this issue.
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I (b')
The alleger also stated that because of his personal i
knowledge of a specific utility employee involved with l,
station battery maintenance, he was concerned with that
'
l individuals qualifications and ability to perform his job (
on safety-related equipment.
An associated issue involved O
concerns: over the applicant's procedural controls and usage (depletion) of' required battery cells.
Specifically, the
alleger felt that the existing procedures did not provide
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notification of the potentially degraded batteries subsequent to their use on the weekend.
The alleger was O
also concerned with the applicant's battery maintenance program and' questioned the adequacy of all cell a---
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In response to-this issue, the NRC inspector reviewed the
'
procedures-governing the surveillance test program for both-
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1E and non~-1ELstation batteries as defined in Procedure Revision Title
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'MSE-SO-5702
Battery Surveillance. Test
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MSE-SO-5706
Battery-Inter-Cell Ccnnector
'4
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Resistance Test
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MSE-SO-5710
Battery Performance Discharge Test MSE-SO-5701
Surveillance of Safety Related Station Batteries Additionally, the NRC inspector reviewed a representative-sample of the-test results documented on the sign-off
sheets for the above listed procedures for the last two years.
Based on these procedural reviews and documentation evaluations, the NRC inspector determined that the
'
applicant's program for the control of station batteries appears-to be adequate in that all required surveillance'
testing has been satisfactorily" performed within the prescribed' periodicity.. Therefore, this portion of-the allegation could not_be substantiated.
No further action is, recommended regarding this issue.
10.
Applicant' Meetings '(92700)
'
On January 5,'
1990, the NRC inspector' met with applicant l
a.
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personnel to discussithe effect of a water leak'that
.i occurred December 23, 1989., The leak was apparently due to colder than normal' temperatures causing a fire sprinkler system to freeze and subsequently break.
The break
.i n
occurred in an annex area of PSE building QR-123 and L
resulted in the-wetting of certain stored documents.
p.
The applicant presented information detailing actions taken
i following a similar incident that occurred in February-
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.1989.
Those actions included:
(1)
The addition offinsulation to the side walls in'the
{
area of the sprinkler system.
,
(2)
Redirecting ventilation ducting to provide air circulation to the sprinkler area.
(3)
Elevation of file cabinets off of the annex floor by installation of 2 x 4s.
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(4); Repair Af.the annex roof.
(5), The placement of protective plastic over: the file
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cabinets.
. These actions were deeraed to be~ adequate to prevent 'a recurrence of the problem and-to better provide.for the-protection of the~ stored documents.
However, as stated
"
above,-the unanticipated cold temperatures experienced on December 23, 1989, resulted in a water leak-in the.same
'
area.
.The applicant then presented examples of the typical water damage sustained by some of the documents as well as a worst-case example of water damage.- The typical 1 examples-
indicated that a slight wetting of the documents, generally in the corner areas, had occurred..The NRC inspector noted
!
.that the worst-case example was.still legible and had apparently absorbed some tinting:from a file cover-or
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folder.
The applicant explained that-the records had been-
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removed from the storage aitnex to f acilitate their
.;
s examination and to promote the drying of those documents
.which had sustained wetting.
The applicant further stated t
that. currently the documents appear to be dry.and to have sustained little or no permanent damage.
The. applicant
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intends to closely examine those documents that sustained i
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the worst wetting for any unacceptable damage and.if
. j necessary'to replace damaged documents from duplicates
maintained at other locations.
The applicant's prompt attention to'this occurrence and the involvement of appropriate-personnel. appears to have-resulted in the salvaging of the records and the
elimination of'the'need to restore the-records.from.
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' duplicate. storage areas.: Actions proposed to prevent
~
L recurrence include insulation of.the roof area over-the annex, closure-of some openings found in the eaves of the-annex-roof, installation of ceiling grills to allow room air into the: sprinkler piping area,.and the potential installation of: thermostat 1cally. controlled roof vents in
,,
lieu of the passive roof vents" currently installed.
Theue.
actions should be effective in preventing a recurrence of t
the condition..
-'
The.NRC. inspector notes that the applicant's actions and
!
L quick response to this issue indicate a strength in the
applicant's program.
'
b.
On February 5, 1990, the NRC inspector met with applicant ji personnel regarding the adequacy of receipt inspection procedures with respect to the requirements of 10 CFR'
Part 21.
Specifically discussed were recent changes to
i
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nm.
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-91-Procedure:NQA 3.09-11.03, " Receiving-Inspection." ' The procedural changes were accomplished by Document ^ Change'
' Notices (DCN) 1 andL2 to Revision-1 of the procedure and directed that_" Items which do not conform to the specified'
requirements shall be documented;in accordance with-References 3.13 and/or 3.2.as applicable." _ References 3.13'
and 3.2 are CPSES Procedure NQA 3.09 - 9.02, " Inspection-Reports / Inspection Plans," and NQA 3.05 ~
'! Reporting and
,
Control of_Nonconformances," respectively.
NQA'3.09 --9.02-provides for the documentation of certain nonconforming conditions-by marking the inspection ~ report attribute
~"unsat."
_ Control of the nonconformance is.then accomplished by the "Unsat-IR."
'
The NRC inspector's concern was that nonconforming items documented on an "Unsat IR" would fail to receive an-l evaluation for reportability as required by-10 CFR=Part-21.
10 CFR Part 21 requires that for-safety-related items, an-
!
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evaluation be performed of deviations or departures from-the technical requirements included in the procurement.
documents to determine if the deviation was reportable to:
,
the NRC.
"Unsat irs" do.not receive such an evaluation'.-
!
.The. applicant discussed the issue with the NRC inspector
,
and agreed to review the receipt inspection procedure for
,
adequacy.
However, pending further NRC evaluation t'o
!
determine if the changes resulted in failure to comply with l
the= requirements'of 10 CFR Part 21, this issue is considered unresolved (445/9003-U-04; 446/9003-U-02).
a Plant Tours (71707)
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During a. plant tour, the NRC inspector was informed-by a security guard that:while.walkinga on a stairway.in.the
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. Safeguards building, he felt /a' strong flow of air against the-l
'
back=of his hand.
- After locating the air flow, the_NRC i
inspector _ requested the applicant'to chip the concrete to
~
determine'the source of the air leakage.
The chipping removed a.
l loose surface patch leading to a 1-inch diameter through-wall
(2'6 thick) snap tie hole, apparently related to the form work used for the wall construction.
The hole, which opened into the i
seismic gap between the Auxiliary and Safeguards building, is
!
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located'at elevation-883'4" and 2'7" cast of ES.
The applicant documented the discrepancy on ONE form FX90-640 and
,
.dispositioned the hole to be reworked as a nonstructural defect i
with'no additional corrective action.- The NRC inspector b
questioned whether other corrective actions are needed to L
determine (1) whether a multitude of surface-patched snap tie holes could exist,-(2) whether this condition would have any L
effect on the structure, (3) whether any Hilti bolts installed L
close to surface-patched snap tie holes could be affected, and H
(4) whether surface patching of holes could have occurred in E
other applications.
Until these questions are resolved, this l:
issued will be tracked as open item (445/9003-0-05).
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12.
Unresolved Items i
Unresolved. items-are matters about which more information is required in order to ascertain whether they are acceptable items, violations, or deviations.
One-unresolved item was disclosed during the inspection and.is discussed in paragraph 10 b.
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13.-
open Items
'
open items are matters which have been discussed with the-applicant, which will be reviewed further by the inspector,.and
. which involve some action on the part of the NRC or applicant or both.
Two open items disclosed during the inspection are discussed in paragraphs 4.r and 11.
14.
Exit Meetino (30703)
An exit meeting was conducted February 6, 1990', with the-
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applicant's representatives identified in paragraph 1 of this t
report.
No written material was provided to the applicant by
'the inspectors during this reporting period.
The applicant did
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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.
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