IR 05000313/1989200
| ML19325C501 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 09/22/1989 |
| From: | Imbro E, Lanning W, Parkhill R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19325C494 | List: |
| References | |
| 50-313-89-200, IEB-79-02, IEB-79-14, IEB-79-2, NUDOCS 8910160318 | |
| Download: ML19325C501 (28) | |
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t U.S.-NUCLEAR REGULATORY COMMISSION-
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0FFICE OF NUCLEAR REACTOR REGULATION l
Division of Reactor Inspection and Safeguards
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Report ho.: 50-313/89-200 l
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. Docket No.:- 50-313.
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~ Licensee:.
Arkansas oowerandLight(AP&L) Company
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'P.O. Box 551
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Little Rock, Arkansas. 72203 l
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Facility Name: Arkansas Nuclear One (ANO)
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Inspection at: AP&L Corporate Offices in Little Rock Arkansas and AND Site.
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Inspection Conducted: July 10, 1989 through July 21, 1989 Inspection Team Members:
. R. W. Parkhill, R$1B, NRR - Team Leader i
A. H. Lee, EHEB, NRR
J. S. Ma, ESGB, NRR
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L. E. E11ershaw, RIV
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L. D. Gilbert, Region IV S. W. Korde, Consultant
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R. E. Serb, Censultent
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Attende' Exit Meeting:
E. V. Imbro, RSIB, NRR-J d
I. Barnes, RIV
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-o D. D. Chamberlain, RIV
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M. WfhY f M /S Ronald W. Parkhill, Team Leader Date Sig
4 Reviewed By:
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f Lugene V. Imbro, Chief, Dde 5jgned Team Inspection Development Section B I
Approved By:
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Wyny' al Inspection Bran h, Da[e Sygnet!
D',' Larining, Chief Spec RR
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8910160318 894006 I
PDR ADOCK 05000313
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TABLE OF CONTENTS l
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I 1.
BACKGROUND INFORMATION.....................................
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2.
PURPOSE OF INSPECTION......................................
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INSPECTION EFFORT AND REPORT ORGANIZATION..................
3.1 Inspection Effort.....................................
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3.2 Report Organization...................................
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LICENSEE WEAKNESSES........................................
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4.1 56niple Size Associated with Bulletin 79-02............
I 4.2 Consistency Between the As-Built and As-Designed Piping..............................................
4.3 Hissing Pipe Support Calculations.....................
'4.4 Design input and Modeling Practices.........'.........
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4.5 Code Compliance......................................
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LICENSEE STRENGTHS.........................................
5.1 Implementation of Bulletin 79-02 - Excluding Sample Size................................................
5.2 Staffing and Inspection Support.......................
5.3 Current Analytical Techn' ques and Design Criteria.....
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SAFETY ANALYSIS REPORT (SAR) COMPLIANCE.....................
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CONCLUSION.................................................
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i APPENDIX A' SIGNIFICANT DISCREPANCIES...........................
A-1 APPENDIX B MINOR DISCREPANCIES.................................
B-1 APPENDIX C PERSONNEL CONTACTED DURING INSPECTION...............
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f Arkansas Nuclear One (ANO) - Unit 1
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Bulletin 79-02 and 79-14 Follow-up Inspection July 10 through 21, 1989
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BxCKGROUND INFORMATION I
i InspectionandEnforcement(IE)Culletin79-02, Revision 0,"PipeSupport l
Base Plate Designs Using Concrete Expansion Anchor Bolts," was issued on
March 8,1979 to ensure the adequacy of the design and installation of
pipe support base plates using concrete expansion anchor boltn.
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Bulletin 79-14 Revision 1, " Seismic Aaalysis for As-Built Safety-Related Piping Systems," was issued on July 2, 1979 (16 days after the initial
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issuance) to ensure cor.formance between the as-built safety-related piping systems and the associated seismic analyses (i.e., the as-design piping.
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configuration).
As a result of inspections.recently conducted at varinus facilities, the i
NRC has been getting indications that the actions requested by Bulletins 79-02 and 79-14 may not have been properly completed by all licensees.
J Consequently, the NRC staff has decided to review implementation of these bulletins at a few representative plants. The first plant selected for this review was ANO, Unit 1.
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PURPOSE OF INSPECTION The purpose of this inspection tvas to assess Arkansas Power and Light (AP&L) Company's commitment to implement IE Bulletin 79-02 and
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3.
INSPECHONEFFORTANDREPORTORGANIZATION 3.1 Inspection Effort
for Bulletin 79-02 the inspection team reviewed the licensee's program as i
well as iraplementation of that program to verify conformance with the l
subject bulletin. A site walkdown of various pipe supports was performed
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to evaluate the adequacy of randomly selected installed baseplate and anchor bolts, as well as a review of pipe teprt designs selected from the piping analyses reviewed by the team.
For Bulletin 79-14 six piping analyses were identified by the intpection team as potential inspection subjects during a pre-inspection visit co the i
AP&L offices on June 28 and 29, 1989. These analyses were identified to permit AP&L to assemble copies of the associated piping calculations and i
i pipe support drcwings such that the inspection team could begin work immediately upon arrival. The six chosen analyses included the following
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portions of systems-
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l (a) Emergencyfeedwater(EFW)frompumpdischargenozzlestocontainment penetrations.
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(b) EFW turbine steara supply.
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i l(c) Decayheatremoval(DHR)suctionpipingfromthecontainment penetruions to the DHR pump suction nozzles.
d Main feedwater safety-related piping located outside of containment, i
e DHR discharge piping from the DHR pumps to the DHR heat exchanger.
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f Service water piping located in the ir.take structure building.
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The inspection methodology was composed of a comprehensive review of ccsign criteria documents. design inputs, drawings, computer models,
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walkdowns to verify consistency between the field installations and the
analyzed configuration and review of other design basis documents, as
~i applicable, to ensure compliance. The piping runs modeled in the six analyses were' walked down on a sampling basis to verify agreement between the installed configuration and the analytical model.
Due to time con-i strainte, only the first four analyses [1.e., (a) through (d)) had a
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detailed design review performed by the inspection team.
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3.2 Report Organization
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Section 4 lists those practices which were viewed by the inspection team to be indicative of licensee weaknesses. Section 5 provides those
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licensee good piactices that were favorably viewed by the inspection team to be indicative of licensee strengths. Observations cp SAR compliance are
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5ection 7 provides the overall conclusions reacied by the inspection team, t
Three appendices are attached to the report. Appendix A lists the most significt.nt discrepancies identified by the inspection team. Appendix B lists the discrepancies primarily identified as a result of the inspection
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team's walkdown that'are not included in Appendix A.
Appendix C lists the
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personnel contacted during the inspection.
4.
LICENSEE WEAKNESSES
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4.1 Sample Size Associated With Bulletin 79-02
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The number of bolt sampics selected and tested was approximately one tenth
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that required by IE blietin 79-02.
IE Bulletin 79-02 required that bolt samples would be selected raiadomly and tested to achieve a Si percent confidence level that less than 5 percent defective bolts n.e installed
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in any one of the safety-rSlated piping systems, and that this level of confidence be achieved for each system. The contractor who performed this
task for the licensee had randomly se % cted and tested bolts to achieve the 95 percent confidence leeel with less than a 5 percent defective bolt criterion from the whole plat,t, not system by system. Since there were II. safety-r(?ated piping systems at ANO-1 Unit 1, the total nunber of bolt
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samples required by IE Bulletin 79-02 would have been about 10 times more than that selected and tested. The licensee has acknowledged the sampling deficiency and conunitted to perform additional sample testing to correct the deficiency.
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4.'2 Consistency Between the As-Built and As-Designed Piping AP&L acknowledged they did not possess nor could the architect-engineer
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readily retrieve the original 79-1* walkdown documentation. Based on the
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lack of original 79-14 documentation and the continuing identification of field discrepancies, AP&L initiated the Iso-Update Program in 1986 to improve the quality of design drawings and later in 1987 expanded the
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scope of the program to be a reconciliation between the as-built seismic
Class i piping / supports and the governing design.
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At the time of the inspection AP&L had completed approximately 10 percent i
of the Iso-Update Program for ANO Unit 1.
AP&L indicateo that in its
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srevious walkdown program approximately 50 percent of ihe plant systems
1ad been completed, but due to a change in walkdown criteria the program
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was started over.
Based on the discrepancies that continued to be identi-
fied by this program, as well as the discrepancies identified by the intpection team (see Appendices A and B), it is apparent that the action
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in res)onse to Bulletin 79-14 is incomplete and AP&L wiil not have assur-ance t ut the piping design matches its field inst 611ation until the
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Iso-Update Program is completed. Current planning by AP&L was to complete
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the program between five and eight years from the date of this inspection.
The team concludes that a higher priority on the completion of this
programiswarranted. AP&L has not completed the actions requested by NRC dulletin 79-14.
4.3 Missiw Pipe Support Calculations The original pipe cupports were designed by Grinnell through a centract with the architect-engineer. The design loads for pipe supports were
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shown on the pipe support sketches, but the design calculations to sub-
.stantiate the adequacy of these supports were not available. AP&L stated that efforts had been made to recover the pipe support design calculations L
from Grinnell but were not successful.
Since ANO Unit I was one of the early-vintage nuclear pawer plants tne i
inspectionteambelievedthatthesituationofunavailableormissIng design records for pipe supports was not uncommon. The design practice for pipe supports at that time wcs mainly a catalog type of selection and the need of_ record keeping was not universally followed among pipe support designers.
AP&L and its contractors had analyzed some of the as-built piping systems and the analysis results had yielded new loads for pipe-supports. These new loads were then compared to the design loads shown on the Grinnell pipe support sketches.
If the new load was 5 percent greater than the original design load, new design calculations would be required and l
generated for the pipe support. Otherwise, the pipe support was consid-
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ered to be adequate and was not reanalyzed.
If the new des,ign calcula-
tions resulted in the need for physical modifications to the pipe supports, modifications would be performed. The inspection team believed th'.t the licensee's logic for assessing the adequacy of pipe supports was
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rea.onable because the 5 percent tolerance limit was small enough to avoid overloading the pipe support.
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'4. 4 Design input and Modeling Practices
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In reviewing the piping analyses the inspection team checked the coding of
computer input against design information, and identified some modeling
inconsistencies which amounted to incorrect and/or ender prediction of
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design loads and stresses..The representative examples included inaccu-
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rately modeling thermal modes, failure to consider the effects of mass
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eccentricity, zero period acceleration, seismic anchor movements, and using various truncated models to amend and reduce model size. AP&L-attempted to defend some of the items by demonstrating consistency with-
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original modeling methods, engineering judgments, or documented internal memorandums. The inspection team took exception to these arguments and
concluded that modeling techniques based on nonconservative engineering
judgments were not acceptable.
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4.5 Code Compliance
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The original code of record for ANO Unit I nuclear piping was ANSI B31.7
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draft Edition of February 1968, with Errata of June 1968. The current
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analysis code utilized by AP&L was the ASME-III 1977, including Winter
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1979 Addenda. However, AP&L indicated that they still had the caveat to
use the original code despite their current analysis intentions. The inspection team took excestion to this approach. AP&L needs to estcblish the code of record, and tien to consistently apply it. AP&L had hired a consultant to do the code reconciliation and the associated report was
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scheduled to be issued in January 1990.
5.
LICENSEE STRENGTHS t
5.1 Implementation of Bulletin 79-02 - Excluding Sample Size
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A description of the work performed for IE Bulletin 79-02 at ANO Unit I t
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was contained in a report entitled " Arkansas Nuc? ear One NRC IE Bulletins 79-02 and 79-14 Summary Report" pre)ared by Bechtel Power Corporation, February 1985 and revised Septem)er 1985.
The scope of the
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work for IE Bu14 tin 79-02 was handled by two Bechtel subgroups; enaineer-ing and ennstruction. The engineering group identified all the safety-related pipe supports with ficxible baseplates, evaluated them for adequacy, and redesigned the inadequate pipe supports. The construction group performed plant walkdowns, and tested expansion anchor bolts using torque verificatior methods, verified the as-built pipe support configu-rations against tLe design, and physically modified pipe supports in accordance with the redesign by the engineering group.
All the safety-related pipe supports with flex 41e baseplates were reana-1 ped for the adequacy of expansion anchor bolu.
The pipe support loads were taken from the Grinnell pipe support sketches and factors of safety for expansion anchor bolts were then calculated by a computer program which had accounted for baseplate Tlexibility. Expansion anchor bolts having a factor of safety less thert two required irinediate or short-term pipe support modification. Pipe supports with expansion anchor bolts having a factor of safety between two ar.d four (nonshe'l bolts) and five (shell bolts) were given a longer time for the completh u of modification.
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- All the associated pipe support modifications had been completed before i
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December 12, 1979, and had been determined to meet IE Bulletin 79-02
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requirements.
Except for the deficiency in the number of bolts sampled, the licensee j
responded rapidly and performed well with respect to IE Bulletin 79-02.
All the pipe supports anchored by expansion bolts had been upgraded with a
factor of safety equal to, or greater than four for nonshell bolts and i
five for shell bolts. Such a high factor of safety for expansion anchor bolts had been consnonly recona. ended by bolt manufacturers for design, and normally followed by designers. The factor of safety makes allowances for I
any differences between actual field conditions and laboratory test
conditions. Bolt behavior studies have shown that significant differences could occur due to the baseplate flexibility effect, concrete cracking j
near bolts. and improper installation of bolts. The baseplate flexibility j
effect was generally.not considered by Bechtal during original design due
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to its complexity, but was considered by Bechtel during its re-analysis for the adequacy of expansion anchor bolts at ANO-1 Unit 1.
No concrete cracking near bolts was observed by the inspection team during walkdowns, j
The aforementioned evidence indicated that sufficient safety margins
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existed in the field with bolts still having a minimum factor of safety of
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four. The proper installation of bolts was verified at ANO-1, although the sample size was not up to the stendard specified in the IE
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Bulletin 79-02. The effect of the small sample size in determining the i
safety margin of expansion anchor bolts was secondary compared to the i
primary effects of baseplate flexibility, cor.ccete cracking near bolts,
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and improper installation of bolts. Therefore, with all things consid-ered, the licensee's activity with respect to IE Bulletin 79-02 had reasonably assured the safety of pipe supports with expansion anchor bolts
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in spite of relatively low sample size.
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5.2 Staffing and Inspection Support AP&L was directly responsible for the performance, supervision, control, and management of nuclear safety-related piping analysis and support j
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. design work; having taken over the scope of this work from the original
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. architect-engineer firm, Bechtel, in October 1986.
The inspection team was generally satisfied that AP&L had established a competent design and management team to conduct nuclear safety-related piping design work. The design organization appeared to be very respon-
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sive to operational issues and where broed problems were identified, a o
L program was already in place which would eventually correct the situation if fully and properly implemented (i.e., Iso-Update, progrr a configuration management, snub >er reconciliation and specific piping technical
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guidance). Without exception, all AP&L personnel interfaced during the inspection showed a high degree of professionalism, competence in j
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understanding of technical issues, willingness to provide information, and contributed to discussions in an objective manner to resolve comments.
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5.3 Current Analytical Techniques and Design Criteria
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The inspection team found the current piping analysis methods, such as modeling techniques and loads specification, recently employed by AP&L to
be appropriate and comparable to methods used elsewhere in the industry.
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Contributing to the adequacy of analytical techniques was the apparent competence and concern with ANO piping design quality exhibited by the
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project engineering staff. Also, AP&'. had taken steps to improve the piping design process during the past two yeara. The standardization and l
guidance provided by Specification APL-M 2514, entitled, " Technical
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Specification for the Design of Piping," Revision 0, dated June 10, 1987
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and the yet to be issued document 88-E-0125-01, " Structural Design Guide,"
i were examples in this regard. The specification included topics such as closely-spaced modes, 3-directional earthquake, use of zero period accel-
eration (2PA) in assuring minimum pipe motion, seismic anchor moveuent.
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use of actual support stiffness, deflection check at restraints, backside
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anchor load calculation, adequacy)of the multi
>11 cation factor between
operational basis earthquake (OBE anddesignl>asisearthquake(DBE)
responses, friction force in support design, and load combinations.
Although the specification did include a caveat which sercitted exceptions to the criteria AP&L statements and inspection team o>tervations indicatedthatItwasseldominvoked.
However, the inspection team did
conclude the specification should be revised to eliminate exception and thereby ensure consistent application of project requirenents. The
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planned release of the design guidelines as controlled documents would
also promote piping design quality.
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SAFETYANALYSISREPORT(SARLCOMPLIANCE The inspection team identified no areas of noncompliance with the SAR.
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This was mainly due to the vintage of the plant and that ANO Unit I was not connitted to the standard review plan format. Thus, the AND Unit 1
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SAR was rather terse and did not address design specifics, such as load l'
combinations, which were included in lower level design input documents referenced in Section 5.3.
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The ins)ection team reviewed the smoothed response spectrum curves and l
found tiem acceptable on the basis of the vintagr of the plar.t and the SAR l
cohnitments.
Unlike the current position in Regulatory Guid:: 1.122, there was no specific guidance on cpectrum peak broadening when the plant was constructed. The spectrum peaks were therefore smoothed, but not
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broadened.
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The ground response spectra used for ANO-1 design was less conservative than the Regulatory Suide 1.60 spectra, but typical for plants of that I
vintage. However, the damping values used for piping analyses were very conservative, which is 0.5 percent for both DBE and OBE, for pipes of
various sizes.
Another exception regarding current licensing requirenents concerned the cut-off frequency in the dynamic analysis, which was specified at 30 Hertz in lieu of 33 Hertz. Consideringthatzeroperiodacceleration(ZPA)was planned to be included in the piping analysis for assuring adequate representation of support motion, the staff feels that the absence of
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higher modes would not have significant effect on the piping response,
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and, therefore, was acceptable.
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CONCLUSIONS
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For. Bulleth 79 02, the inspection team concluded that AP&L had not
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adequately completed the actions specified in the bulletin because of
having tested too few bolts. All other aspects of the program were l
satisdactory and the inspection team concluded that adequate anchorage for
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baseplates had been provided, even though some additional testing is l
appropriate.
'I For Bulletin 79-14, AP&L had not adequately completed the actions speci-fied in the bulletin. The inspectfon ter.m identified numerous inconsir,-
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tencies between the constructed pip lng r,ystem and the piping analyses, the walkdown program was only 10 percent c'.,mplete, good engineering practices
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were inconsistently considered (e.g., ZPA, eccentric rasses, SAM, etc.),
r and additional analyses that were needed (i.e., local stresses for lugs,
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valve seismic qualification, water hanmer, etc.).
In addition, AP&L had
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indepe iently identified that the wrong rerponse spectra had been used in
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the serv 1ce water piping analysis for the intake structure. Similar l
problems may exist in the piping analyses of other systems. Therefore, the Iso-Update program needs to be completed in a timely and thorough t
manner to bring ANO Unit 1 into conformance with Bulletin 79-14.
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APPENDIX'A - Significant Discrepancies
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Bulletin 79-02 Discrepancies
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Sanple Size of Bolts'to be Tested A-1
Bulletin 79-14 Discrepancies
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Piping Analysis Nonconservatisms l
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Pump Nozzle Loadings A-1 l
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SeismicAnchorMotion(SAM)
A-2 Thermal Expansion for Vary (ZPA) perating Modes 3.
ing O A-3
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Zero Period Acceleration A-4 4.
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Eccantric Mass of Valve Actuators A-4
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Containment Pe,1etration Di: placement A-5
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B.
As-Designed and As-Built Piping Differences
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Nonfunctional Pipe Supporte in Service Water System A-5 2.
Recently Reworked DHR Pipe Support Not In Agreement with Design A-6 3.
Spring Hangers A-6 4.
Snubber Settings A-6 5.
Main Feedwater Containment Isolation Valve Interaction
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with Structural Platform A-7
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C.
Additional Analyses To Justify Current Plant Configuration 1.
Main Feedwater Water Hawner Analysis A-7
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Seismic Qualification of Main Feedwater Isolation Valve A-8
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Damaged DHR Piping A-9 4.
Code Reconciliation A-10
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5.
Updating Stress Analysis Calculations A-10
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APPENDIX A
SIGNIFICANT DISCREPANCIES
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The discrepancies identified in this appendix are considered to be the most significant issues resulting from this inspection and are numbered separately.
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They are categorized by their applicability to the individual bulletins.
For Bulletin 79-14 the discrepancies are further broken down into headings of
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piping analysis nonconservatisms, as-designed and as-built piping differences,
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and additional analyses needed to justify curren' plant configuration.
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AP&L action item number that was assigned to the individual issues during the l
inspection are also provided as an aid in providing continuity between the l
inspection and the report. The inspection team acknowledges AP&L's commitment to individually resolve and track all issues identified during this inspection such that their resolution could be verified at a later date.
Bulletin 79-02 Discrepancies 1.
Sample Size of Bolts to be Tested 50-313/89-200-01 (OPEN)
Bulletin 79-02' required that bolt samples should have been selected i
randomly and tested to achieve a 95 percent confidence level that less
than 5 percent defective bolts were installed in any one of the safety-related piaing systems, and the sampling program conducted on a system-by-syatem 2 asis. However, AP&L's architect-engineer had randomly selected and tested bolts to achieve the aforementioned confidence level from the whole plant, not on a system basis, without increasing the size of the sample.
Since there were 11 safety-related piping systems at
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ANO-1 l' nit 1, the total number of bolts sampled should have been increased by a factor of 10 to comply with the original scope of this bulletin, The
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work done for Bulletin 79-02 at ANO Unit I was contained in a report entitled " Arkansas Nuclear One NRC IE Bulletins 79-02 and 79-14 Summary Report," prepared by Bechtel Power Corporation, dated February 1985 and revised September 1985.
(Refer to AP&L Action Item No. 54.)
Bulletin 79-14 Discrepancies
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A.
Piping Analysis Nonconservatis.as (1) Pump Nozzle Loadings 50-313/89-200-02(OPEN)
In reviewing the Decay Heat Removal (DHR) suction piping analyses, the inspection team identified three unjustified modeling discrepan-cies which if corrected would have increased the suction nozzle loadings on the DHR pumps and the reactor building spray pumps.
First, in modeling the normal shutdown cooling mode, a temperature of 300'F was assumed for almost 15 feet beyond closed valves DH-8A and DH-8B without any justification for this assumption. A realistic temperature gradient in these lines would result in a temperature distribution of less than 300'F and would increase the loading on the DHRsuctionnozzles(RefertoAP&LActionItem32).
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Second, the effect of eccentricity of aess for the actuator of valve BW-8A~which was oriented in the horizontal position was not consid-ered in the enalysis model. AP&L indicated that the mass of manually operated valves could be ignored as indicated in the Pistar modeling guidelines manual. The inspection team took exception to these i
guicelines and advised AP&L that no guidelines should be followed
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blindly.
In this case, the subject valve although manually operated,
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had a significant amount of eccentric mass due'to the handwheel gear
box, unlike most other manually operated valves and needed to be
evaluated.
In eddition, the actuator was oriented in the horizontal
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position and was located in close proximity to DHR Pump 34A suction
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nozzle (Refer to AP&L Action Item No. 41).
j Third, according to AP&L guidelines, valves were modeled as equive-
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lent pipes with twice the nominal wall thickness and flanger were modeled as pipes with additional mass.
Both of these practices make
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the piping system more flexible than it would have been with realis-tically modeled flanges and valves. While this practice may be conservative from a seismic standpoint, it underpredicts the loads
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and stresses in the thermal expansion analysis. Specifically, the
inspection team was concerned that the thermal loads on the nozzles
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of reactor building spray (RBS) Pump 35A had been similarly underes-timated.
(RefertoAP&LActionItemNo.66.)
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During the inspection, AP&L did not provide information to the team that demonstrated that operability of the DHR and RBS pumps had not been adversely affected or provide assurance that similar modeling nonconservatism was adequately considered for other analyzed safety-
,
related systems. Also, conservative guidance with regard to the consideration of eccentric masses of valve actuators and thermal
,
analysis modeling was not yet established.
,
References:
a.
DHR Original Stress Analysis Calculation No. 669 by Bechtel; Math Model Revision 6. September 22, 1979; Computer Input, j
May 16, 1980; Check Sheets, June 16, 1980, i
b.
Stress Analysis Calculation Package 870-1098-02, File Number ARK 21.0300, Revision 4, pages 1 through 114 and cover sheet.
c.
Stress Analysis Calculation Package 870-1098-20, File Number ARK 21.0307, Revision 1, pages 1 through 64 and covar sheet.
.(2) SeismicAnchorMotion(SAM) 50-313/89-200-03(OPEN)
The piping analysis for the emergency feedwater (EFW) turbine steam supply line (AP&L calculation 87D-1099-02, Qualification of EFW Turbine Steam Supply for M0 VATS Changes, Revision 0, dated June 21,1989) assumed that the subject lines were anchored at the
connection to the main steam piping. However, no SAM displacements were modeled because the previous 1980 analysis by the architect-engineer had incorrectly omitted them, and the newer analysis made A-2
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I the same assumptions. This is viewed by the inspection team to be an l
example of improper verification of inp(Refer to AP&L Action Item No.
ut data, as well as a
no3 conservative modeling assumption.
53.)
j L
'
Another example of not considering SAM was identified by the inspec-tion team for piping analysis associated with the safety-related main
feedwater piping.from the containment penetration. including the containwent isolation valve (Reference Computer Problem No. 441 dated November 30,1972 and No. 439 dated March 27,1973). As a result of i
this and other discrepancies identified by the inspection team and i
AP&L project personnel, AP&L reran the associated computer analysis.
Preliminary, results from AP&L indicate that associated feedwater piping was qualified even with the failure of one existing snubber.
As part of their general response to having considered SAMs. AP&L could only. verify that the calculation associated with.the high pressure core spray had included SAM.
(RefertoAP&LActionItem t
No.6.)
For the EFW turbine steam piping, AP&L did not evaluate the effect of
!
the seismic displacement effects associated with main steam line i
piping on the qualification of this piping.
In addition AP&L has
.
not evaluated other safety-related piping analyses to ensure that piping models terminated at nonrigid piping and anchors have been correctly modeled to include the associated seismic anchor movements.-
(3) Thermal Expansion for Varying Operating Modes _50-313/89-200-04 (OPEN)
The inspection team identified that a single thermal expansion load case was evaluated for the piping analysis ossociated with the EFW
'
turbine steam supply line. The case evaluated was associated with i
the subject piping heated uniformly to the same maximum operating temperature.
However, the inspection team identified one example of a thermal mcde which was not analyzed where the piping downstream of thesteamadmissionvalves(CV/SV2613and.CV/SV2663)wasrelatively
cold compared to the upstream piping. An AP&L evaluation of the system operating modes for this piping was not available at the time of the inspection. Although approved on June 21, 1989, the subject piping analysis was actually performed in early 1988. AP&L noted
that current analysis practice would require evaluation of thermal expansion load cases. A draft project guideline for the review of operating modes was awaiting approval and available for information.
,
,
AP&L had not reviewed and identified all associeted EFW turbine steam supply operating modes which may require thermal ex>ansion analysis.
Piping stress reanalysis may be required based on tae results of this review. Additionally, AP&L had not provided assurance that other safety-related piping analyses included all thermal expansion load cases.
(Refer to AP&L Action Item No. 64.)
.
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(4) ZeroPeriodAcceleration(ZPA) 50-333/89-?00-05 (OPEN)
In tt
)HR suction piping analysis for the seismic loading condition, f
'
the inspection team noticed that AP&L.used a model extcaction cut-off l
of 30 hertz or 100 modes, whichever c; curred first. Generally, in
!
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p most analyses the frequency limit would be reached before the modal i
f limit. However, if the analysis was v6ry large, like the DHR suction i
piping or involved small bore piping, the 100 mode limit could i
l n
possibly govern.
If the modal limit governed, the dynsmic analysis i
would underpredict support loads and. pipe stresses. This practice
was further aggravated since AP&L was apparently only considering the
!
larger of the dynamic solution or the ZPA, as' suggested in Section
6.3.1 of document APL-M-2514, " Technical Specification for the Design
of Piping," Revision 0, dated June 10, 1987.
For the four analyses i
reviewed by the inspection team only two had properly considered ZPA:
DHR suction piping and EFW pump discharge piping. Neither the main
feedwater piping analysis dated March 27, 1973 nor the EFW turbine i
steam supply piping analysis dated June 21, 1989 included ZPA. AP&L did not have any justification that other dynamic analyses had been reviewed to ensure that their frequency cut-off comitment of
30 hertz.had been met and that it had been combined with ZPA.
(Refer
!
to AP&L Action Item 50 which discussed nodal cut-off. ZPA was i
discussed but no action item number was assigned.)
j i
(5) Eccentric Mass of Valve Actustors 50-313/89-200-0,6 (OPEN)
'
The inspection team identified two examples where AP&L piping analy-I ses han failed to consider the eccentric mass of valve actuators.
The first was previously discussed in discrepancy number 2 of this i
a ppendix, in w11ch the eccentric mass of the valve BW-8A, located in tie decay heat removal suction piping, was inappropriately omitted in
!
calculating the DHR pump suction nozzle loadings. The second example was associated with the main feedwater containnent isolation valve.
Preliminary reanalysis of the main feedwater piping by AP&L indicated
.
that the effects to eccentric mass in this application were minimal, however, this was not apparent until the analysis was performed.
AP&L's generic response to this issue, included a Bechtel letter numbered MCO-00850 and undated which stated that valve eccentricity was not addressed as > art of Bulletin 79-14 and that the NRC had
.
accepted this approac1. The inspection team strongly disagreed with L
exclusion of valvo eccentricity from the scope of Bulletin 79-14.
Page A-2 of the subject bulletin dated July 18, 1979 clearly stated
'
that valve and valve operator locations and weights were to be l
included.
Based on Bechtel's lack of consideration of the effects of
'
eccentric valve masses AP&L failed to consider these eccentric mass
'
effects on the operability of other safety-related systems.
(Refer to AP&L Action Item No. 5.) Also AP&L did not have conservative g
procedural guidance with regard to the consideration of eccentric masses in piping analyses.
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(6) Containment Penetration Displacements 50-313/89-200-07 (OPEN)
-
In reviewing the DHR sucuon piping analysis, th inspection team
,
noted that containment penetration displacements were not considered
'
s
,
associated with post-LOCA temperature rise and pressurization of the reactor containment building. Sir.llarly, pressurization effect associated with integrated leak rate testing was not addressed as well as, pressure / temperature increases associated with a main streamline leak. AP&L had not analyzed the DHR :,ystem and other safety-related systems that penetrate the containment for these effects.
(Refer to AP&L Action Item No. 39.)
B.
As-Designed and As 83uilt Piping Differences (1) Nonfunctional Pipe Sepports in the Service Water System 50-313/
89-200-08 (OPEN)
The inspection team performed a walkdown of the service water system piping in the intake structure building as shown on piping isometric drawings 14-SW-132, Revision 3, and 13-SW-133, Revision 4.
Two separate pipe supports, each designated as HBD-2-H1 were identified by the inspection team as not being load bearing. Both pipe supports i
were designed as vertical supports configured as pipe stanchions l
'
wd ded to the subject piping and in bearing with the floor via l i r*,
basepictes. The inspection team identified that a b inch gap e4r.ed between the stanchion and the floor. Therefore the as-built Nnfiguration did not agree with the existing analysis. AP&L indi-cated that these two supports were immediately grouted once the issue was identified. Additionally, in performing a system operability evaluation to support plar.t operations with one service water pump removed and in preparation for 'his inspection, AP&L indicated that i
the wrong response spectra curves had been utilized by their architect-engineer in designing the service water system piping (0.8g used instead of 2.1g). AP&L indicated that the reanalysis of the service water piping using the correct response spectra ard with one service water pump removed resulted in an operable system.
However, j
the team concluded that the following would be needed to clearly l
establish that the operability of the system is unaffecteo.
j a.
Verify system operability with the two supperts not included in the design analyses.
b.
Assurance that the supports when grouted became load bearir.y and that they were not just cosmetically changed to remove +.he apparent gap.
I c.
Include in the reanalysis consideration of all sliding supports that do not use Teflon, Lubrite or other friction reducing material. These frictional considerations apply to the pipe support design as well as the stress analysis.
(Refer AP&L Act h Item %?. 68)
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(2)T Recently Re-Worked VHR Pipe Support Not In Agreement With Design.-
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T-50-313/89-200-09 (OPEN)
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As;part of the corrective actions ' associated with a water haanner E
' event, pipe support _DH-122 was to be re-worked to its original l
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L configuration.=.The water hanner event and corrective actions were m
.
_ described in an undated and unsigned compilation report entitled
'
'
" Probable ~ Cause Evaluation Arkansas ~ Nuclear One-Unit 1, Decay Heat Piping and Support Deficiencies." A walkdown by the inspection team identified that this wide flange box restraint (i.e., DH-122) had a 4-inch' gap at'the top of'the pipe in lieu of the design'value of-
,
-1/16-inch. Therefore, from a design perspective the_ pipe support
--wouldinot be assumed to restrain the piping in the vertical upward
,
direction. AP&L did-not have information to'desonstrate the
'
acceptability of the excessive gap for hanger DH-122 including an explanation from a programmatic perspective why QA/QC did not
,
' identify and correct this discrepancy.
(RefertoAP&LActionItem l
No. 5.)s i
(3)'SpringHangers 50-313/89-2.3-10 (OPEN)
. As a result of the inspection team's field walkdown numerous discrep-j ancies were identified with regard-to spring hanger design. The issues identified ranged from s range by as much as 8 percent (pring settings outside of the design a
see AP&L Action Item No's.8,-9, 20, 27.29,56and58),maintenancetagsinstalledforaslongas6 months on potentially: inoperable springs (see AP&L Action Item No.
-20), extra unanalyzed mass attached to s-trical conduit (see AP&L Action Item 20)pring hanger, viz. an elec-
. no design setting provided
)
-on hanger: drawing (see AP&L Action Item No. 37), interference with
.
adjacent.pipinginsulation(seeAP&LActionItemNo.18),anda spring setting could not be verified due to missing or illegible scales (seeAP&LActicnItemNo's-23and?.9). AP&L did not have an evaluation of-all of the aforementioned. discrepancies including l
timeliness of maintenance activities, nor did AP&L have a program to
-)
ensure that spring settings were verified and mai:itained within an q
acceptable tolerance.
(SeeAP&LActionItemNo.17.)
,
- (4) Snubber Settings 50-313/89-200-11 (OPEN)
During'the walkdown of the main feedwater system, the inspection team W
identified one pipe support, EBD-10/HS-39, Revision 1, whose snubber hot load pin-to-pin dimension exceeded the design value by an amount l
. equivalent to the predicted design movement. AP&L subsequently n
~
'
reviewed the snubber configuration to ensure that the snubber was not bottom-out in either its hot load or cold Mad position and reviewed
"
_"
the last outage inservice inspection (ISI) data which had indepen-dently evaluated the subject snubber's travel and verified that this f'
snubber had an acceptable travel range. Also, AP&L indicated that a
snubber reconciliation program had currently reviewed 39 Unit 1
]
snubbers. The purpose of the snubber reconciliation )rogram was to update the pipe support drawings in accordance with tie existing stress analyses and create a data base to augment the ISI evaluation
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F of snubbers. The-inspection team concurred with the implementation l
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w manner. The: inspection team also noted that based on the-issues I
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T identified-in this inspection report _and the non-rigorous status of
.W-the AP&L piping analyses,-the calculated snubber movement may be
>
underestimated.,(Referto'AP&LActionItemNo.24.)
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'(5)j Main Feedwater Containment I wlation' Valve Interaction With
'
,
Structural-Platform 50-313/89-200-12 (OPEN)
.
The' inspection' team,;during the system walkdown, identified that the l
actuator of the main-feedwater iso'ation valve, CV 2680, was actually in contact with the handrail of the ladder of the structural plat-i N
form.- This intarference.needed to be eliminated to. ensure that seismic. interaction with the platform would not have any deleterious effect on the. valves-operability. Additionally, AP&L did not-describe what programs had been implemented for ANO Unit I to review
,
similar unacceptable seismic interactions and why this example was not previously identified.
,
C.
. Additional Analyses To Justify Current Plant Configuration (1) Main Feedwater Water Hammer Analysis 50-313/89-200-13 (OPEN)
j In reviewing the main feedwater stress analysis, th: inspection team questioned whether any water hammer evaluations had been performed.
Specifically, the inspection team was concerned that AP&L had no
. justification to ensure that the pressure _ integrity of the safety-related )ortion of the main feedwater system would not'be breached
,
due to tie severe loadings resulting from a water hammer event.
Piping integrity:is required to-prmit the emergency feedwater system
,
t to fulfill its intended safety function'.
'
AP&L's response from their architect-engineer indicated that the main
"
,
,
feedwater water hammer analysis piping-design did not consider water hammer as a: design transient. 'Also, the response further stated that
,
the loss-of nonseismic piping as a result of a design basis earth-t quake:was not required to be postulated as justified in NRC Generic
'
Letter-GL 87-02,'" Verification of Seismic Adequacy of Mechanical and
!
Electrical Equipment In Operating Reactors."
This response was not acceptable since it inappropriately references GL 87-02-as the justification for assuming that piping designated as
-
nonseismic Category I and located in a nonseismic Category I building would maintain its pressure integrity following a design basis 4 s seismic event. The classification of the main feedwater piping located in the turbine building has consistently, since the construc-tion permit was issued, been classified as nonseismic Category I and,
>'
as such, was not designed for a seismic design basis event. There-
. fore, if a design basis seismic event occurred, the nonseismic ;aain feedwater piping located in the tarbine building would be postulated to fail instantaneously and the safety-related portion of the main feedwater piping should have been analyzed for the associated water
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hassner.c Therefore, AP&L:is requested to provide assurance that.the-
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J safety-related portion of the main feedwater piping would maintain
its: pressure. integrity subsequent to a water hammer resulting from a
design basis seismic event. Without such an analysis or suitable
'
,
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justification,. AP&L would not have assurance that the EFW ' system could'meetlits licensing design requirements. Also,:the team was-_
'
,
,
' concerned that other systems may have similarly)not considered water
hanner: loadings..(Refer t6 AP&L Action Item 3.
y (2).reismicQualificationofMainFeedwaterIso1LtionValve.
..
50-313/89-200-14-(OPEN)
,
.
The inspection team reviewed the seismic qualification of the main
feedwater isolation. valves. The qualification analysis was performed -
for AP&L by Anchor / Darling Valve Company and entitled " Seismic Design-
>
Calculations, Bechtel Order 6600-M-258-BC Rev. 8, Valves CV-2630'and-
- CV-2680,"' September 10, 1974. Bechtel Specification No. 6600-M-123,
'
" Nuclear Class Gate, Globe'and Check Valves-For Arkansas Nuclear One
-
For The Arkansas Power and Light Company," Revision 2, February 17,
";
1972, in Section 2.5 required that safety-related valves with opera-tors would be able to withstand en inertial load of 3g in any direc-
-
tion in addition to normal operating-loads.
In addition, the specification required;the extended parts of'the valves have a~
E frequency of vibration greater _than 20 hertz.
In reviewing these L
_ requirements, the inspection team noted that the subject valves were not qualified'_for the orientation in which the valves were. installed J
(i.e...withtheactuatorsinthehorizontalposition).
AP&L was able to show that the actuator bolt stresses were below the
'
' maximum allowable stresses. However, for the yoke, AP&L was unable
E to demonstrate that the actuator was qualified for the specification requirement of seismic and thrust forces acting concurrently. AP&L o
initiated a Condition Report 1-69-0415 to verify that the seismic
~
,
accelerators et the valve were less than 39's and to have
"
y Anchor /Darl.ing produce a new seismic qualification calculation.
In addition:to these actions, AP&L is requested to review the seismic qualification of other similar valves including containment isolation
- w valves, to determine whether or not this issue is pervasive.
(Refer
~
.to AP&L Action Item 4B.);
A~ review of.the vendor Technical Manual TM A 391.0010 Section 2.3.1.g
,
,
I was made by AP&L at the request of the insrection team, which veri-L fied:that these valves may be installed in positions othar than with A
the stem vertical. A review of Anchor / Darling document entitled
'
" Description of Modifications Needed to Convert 18 inch S150DD Valves For Use in Horizontal Lines and Horizontal Steam Positions" dated R
,
October 1, 1973 and Anchor / Darling drawirg 93-13095, "18 No. 5150WDD F
'
R Series 600...With Limitorque Operator" Revision G, revealed that the
,
b additional parts necessary for horizontal steam installation had been
.added to the subject valves..Therefore, the valves have been v
installed 'n an orientation acceptable to the valve manufacturer but needed an u, dated seismic qualification analyses to demonstrate operability under the geforning design conditions.
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- 4 (3)-DamagedDHRPiping 50-313/89-200-15(OPEN)
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'In response to questions'regarding damaged DHR piping'and pipe
.supportsidue:to postulated water namer related events,- AP&L provided the. inspection' team a copy of an undated report entitled " Probable ~
~I Cause Evaluation ANO-Unit 1, Decay Heat Piping and Support Deficien-cies." The report contained a sumary of_ each support deficiency, I
'
review of operational history, root cause determination, corrective j
-
actions, and conclusions / recommendations.
Based on the evaluation by
.
'
.their operations and engineering staff, AP&L concluded that
Category I deficiencies were caused by independent water hammer i
related events. occurring at different times, and under different i
-
-operational scenarios. The inspection team only considered certain s
aspects of this report as part of the-inspection effort. Specifi-l
-
cally, AP&L's evaluation and refurbishment of piping and pipc sup-ports.inc'uding their adequacy for continued service was reviewed.
J However, certain other aspects of the events such as root cause
~ determination, operational corrective actions, and long-term recom-mendations to eliminate such events remained outside of the inspec-
,
tion team's scope-of review.
The report classified the observed DHR piping and support deficien-J cies identified by AP&L engineering and QC personnel in three'
l categories:
- a.
Deficiencies'resulting from inadvertent pipe movement due to an unanalyzed mode of operation of the DHR piping system.
Included
,
in this ~ category were observed failures such as pipe dents,
- failed'1ugs, bent shoes, pulled anchor bolts, and damage to L,
support steel._
n
b.'
Deficiencies due to configuration reviews. This category mainly
,
l
- incimied grouted penetrations, which should not have been-l h
. grouted per the. original design.
cs. Deficiencies due to construction and maintenance related itew, g
such as, arc strikes, missing washers, poor welds, and bent rods not related to thermal and dynamic conditions.
,
f The report specified corrective actions for all support deficiencies,
. in-the form of physical repair / rework or technical evaluation. Also j
- included were corrective actions from an operational standpoint to
,
,
avoid future recurrence of these events. AP&L provided a copy of
'
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Memorandum HCS-89-0187 from W. Eaton to R. Lane dated March 18, 1989, l
'
that included an assessment of the pipe dents identified in the DHR piping in the vicinity of certain welded attachments. This evalua-
tion. considered existing pipe : tresses at dented areas, adequacy of
'
wall thickness.under pressure, reduced section modulus, and ovalling l
effect. AP&L concluded that the dents were acceptable.
'The inspection team was generally satisfied with the report consider-ing material evidence and technical reasoning. Honver, the report lacked details regarding reportability requirements, operability
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_ analysis review and results, and' integrated systems review to. elimi-
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nate water hammer.. The inspection team focused its attention mainly-
-tu the. suitability of.the damaged piping and ',opports.
In general, y-the. supports were reworked to their originai design configuration and-d
'
.the pipe denting (up to 3/8" density over an approximate'16: square
- i
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incharea)wasaccepted-asis.fWithregard-tothisreview,AP&L
,
comitted to perform the following two. actions.
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First, AP&L vill review'the local stresses on all lugs attached to
)
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the decay heat removal-piping. This action was predicated on lug i
failures as a result of the-postulated water hammer and the u" of-J relatively thin piping (i.e. - schedule 105):in the DHR. system. - This generic action was further justified since AP&L failed to recognize a
-
+
-potential; trend associated with'a specific lug design.
In evaluating
.
a pipeLsupport, DH-113, at a result of a moderate load increase of
,
11 percent :AP&L determined that the associated. lugs had to be-reconfigured to satisfy local stress considerations. - However, AP&L
,
.
~did not review any other-lug pipe seaports to verify their design
'
L basis ~'was adequate. -This is accepta)1e since it is the inspection
team's view that load. ircreases of only 11 percent should be within E
the design' margin for properly designed lugs. Also, only supports that have had their load increased by 5 percent would have calculations substantiating.their configuration since AP&L does not have the original' pipe support designs performed by ITT Grinnel.
g Second, to confirm that. local stain hardening did not occur in the i
area of the dented DHR pi ing, AP&L committed to perform hardness i
L
- testing.using a hand held instrument such as an Equitip.
(Referto
,
L-
- AP&L Action Item Ko. 26.)
D
. (4)-Code' Reconciliation 50-313/89-200-16 (0 pen)
.
?
lip &L was' committed to design nuclear piping in accordance with the
'
K
- ANSIL B31.T Nuclear Piping Code,1968 Edition.
However, most reanalyses ~of piping: systems had been performed in accordance with a later Edition and Addenda of the ASME Code Section III, without performing any reconciliation between.the two Codes. AP&L had recognized the need to-perform the required reconciliation-and issued a request for proposal in late 1988. At the beginning of the'inspec-
~
tion, AP&L had selected a contractor to perform the required task, but had not authorized work to begin. However, AP&L has indicated to the inspection team that a code reconciliation document would be
'
completed in early~1990. Also AP&L needs to update the FSAR to indicate the governing Code (s),it intends to use for the design of
-
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piping systems.
(RefertoAP&LActionItemNo.48.)
(5) Updating Stress Analysis Calculations-50-313/89-200-17 (OPEN)
The' inspection team's' review of the stress analysec for the EFW turbine steam supply was hindered by the fragmented status of the calculation. ~Three previous calt.ulations, References b through d and
- the results of those four calculations [i.e., References a, b, c, and d') needed to be considered jointly with one other overlapping calculation, Reference a, to assess the qualification of the subject piping..The relationship of these calculations to each other was
.
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r difficult to determine. The inspection team found that it was not
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k possible to determine the' relationship of the qualification documen-
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.
itation compiled from'these calculations to che sunnary piping quali-fication as documented in-the calculation file without guidance of.
,
AP&L personnel familiar withithe specific calculation. The inspec--
t
'
. tion team believed that this situation was the result of incomplete e "@
. documents, incorrect references, and incomplete explanation.of file
- contents. The following. paragraphs further illustrate this' concern.
During the. inspection, neither the inspection team nor project personne1'could determine by reference to the calculation file the.
y basis for the pipe support design loads. -The pipe stress and support
. load sumr.; ries included in the dile for Reference a refer to only one-
.of the-included calculations, Reference c.
References to the other.
- ..
',
applicable calculations, References b and d, were not included.
,
The allowable' loads.and qualification'for the EFW turbine pump nozzle
' K3 were not included in Reference a.
The r.ozzle load summary sheet ye
_ included in Reference a refers to Reference b for this information.
E
'
However, Reference b included neither the allowable nor qualification l
for the nozzle. The nozzle qualification,_ including specified p
allowables, was found to be maintained separately in Reference f.
g b
The inspection. team strongly felt that the-status of documentation L
for the piping' system was inadequate and that inadequacies would be
.
unlikely to be corrected without guidelines defining when to-totally revise-pi (Refer to AP&L Action item
_No. 49.) ping stress calculations.
o l
\\
References:-
a.-
' AP&L' Calculation Nurnber 87D-1099-02, Qualification of EFW l-w
<
/
Turbine Steam Supply for M0 VATS Changes, Revision 0, dated l
E June 21', 1989.
!-
L b.-
AP&L~ Calculation Number 800-1083-01, Revision 3, dated November 14,31984.
c..
AP&L Calculation Number C6D-ir,05-01, Revision 3, date unknown.
L-d.-
AP&L Calculation _ Number 86D-1005-07, Revision 0, dated l
L January 27 1986.
I e.
Bechtel Calculation Number 636, revisian unknown, dated
.
May 30,1980.
L f.
AP&L Calculation Number 88E-0104-26, Nozzle Load Review for the Inlet and Outlet Nozzles on the EFW Pump Turbine Driver K-3, s
,
Revision 0, dated November 14, 1988.
,
A-11 u
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_
_
.
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i AK CNDIX B
<
t MINOR DISCREPANCIES 50-313/89-200-18 (OPEN)
>
' This appendix lists the' discrepancies identified by the inspection team by
'
system that are not included in Appendix A.
The Arkansas Power and Light Company (AP&L) action'itemnumberisprovidedaswellasthestatusofthe specific discrepancy. Eventhough some of these discrepancies are closed, they e
are-still examples of differences between the as-built and as-designed piping
.
that had not been reconciled no'r documented prior to the inspection. Also, the inspection: team acknowledges AP&L's commitment to individually resolve and-track all issues identified during this inspection such that their resolution
.could be. verified at'a later date.
'
t
,
.
AP&L Action Drawing / Hanger No..
DiscrepancE Status Ittem No.
.
' Main' Feedwater System -
' Hanger MFW-1, Spring hanger #2Lscale read 4860f Open 29-
- Revision l'
'
.when drawing specified a hot load H
.of 4323f and a cold load C
of #593f.
Hanger MFW-35, Spring' hanger load scale was not Open
~ Revision 0 visible, i.e. : either painted over ur does not exist.
. Drawing 3-EFW-108, Drawing showed two conflicting dimen-Open 13-Revisiun 0 sions for location of Hanger H3.
Hanger 3-EFW-108-H8, Drawing showed dimension from center Open None Revision 0 line'of pipe to wall as 3'8" when-it
.
was ectually 2'11".-
,:
Hanger 3-EFW-109-H8, S) ring hanger il scale read 395#
Closed
Revision 0 w1en drawing specified the' loading
.
as HL 409f and CL 415f.
L Hang'er 3-EFW-109-H5, Spring hanger scale read 1170f Closed
l Revision 0 when drawing specified the
'
loading as HL 1129f and CL 1132f.
'
h Hanger 3-EFW-111-H3, Spring hanger scale read 675# when Closed
l:
Revision 0 drawing specified-the loading as l
HL 654f and CL 670f.
~
Hanger.3-EFW-111-H7,.
Spring hanger scale read 900f when Closed
Revision-drawing specified the loading as l:
HL-868f and CL 873i.
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Actioni l
^
Drawing / Hanger No~.
Discrepancy /
Status Item No._
l 0>
liAnger3-EFW-115-H1,;
Spring ~ hanger scale. read 252# when-Closed
Revision 0-drawing specified the loading as q
fS HLi240f and CL 243f.
J
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,
.
,, ~ :Hani,er 3-EFW-115-Htv Nut loose on one of four anchor Closed 15:
~
-
.
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- Revision'0 bolts for wallisupport plate.,
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.
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_
, ; Hanger 3-EFW-117-H4 y No-lateral clearance was' observed on Open 14-
'y
'
' Revision 0 either side of-pipe when the drawing
>
'
specified'1/16".
-
- t, Emergency Feedwater Turbine Steam Supply System
!
. '
-e Hanger 1-MS-5-H3, l' 'ersirad welds (1/8" vs.1/4")
0 pen l'
u
. Revision 0"
'
' Hanger 1-HS-5 HS, Undersized welds (1/8"'vs. 1/4") and Open
j
-
Revision-0-
-weld' size could not be verified as
.
'
'
.all:around on baseplate.
.
q<
irle: ger!1'-MS-118-H6,>
Undersized welds on bracket'(3/16" Open 55-
- Revision 0 vs1/4").
'
%
- . Hanger 1-HS-118-H3,'
Isometric drawing shows hanger Open
e i Revision 0 Llocation'from vertical pipe as.
"
^4-1/8" when it was actually 10".
E*
Hanger 1-MS-118-H16'
Drawing specifies all.around weld at Opt.i
'
channel attachment.to' embedded steel,.
.
but it was only. welded on'outside of
O
' channel not tne.intide. Drawing-
Especified:all around fillet weld on l'.
brace,_ but backside of channel could
.
' '
>
'1 not accomodate a fillet weld.
s l
Decay Heat Removal System-Hanger DH-14,
Drawing showed dimension from Open
p
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' Revision 2'
centerline of pipe to the symetrical K
support rods as l' 1" when they were
actually l' 10".
Hanger was attached li to baseplates bolted to the concrete
%
ceiling, not to structural steel as
=shown. Also, the nut on top of the n
p spring can was locse.
O L.
'Drawiiig 7-DH-6, Drawing showed dimension from elbow Open None
' Revision 16 to tee on line GCB-1-6 to be 3' 8-1/2"
"
p when it was actually 2' 2-1/2".
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- Discrepancy Status:
Item No.
w l Drawing ~ 7-DH-6 :
. Drawing showed dimension from
- Closed
~
' Revision 16; Hanger,DH-307:to centerline of tee as 4' 6" when it was cctually 5' 4-1/2".
'
,
,
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[* Note _that this discrepa.wy was 1-included in'the stress analysis but not'
'
_
documentedonthefabricationisometric(fabiso).)
'
- Drawing 7-DH-6"
-Drawingishowed dimension from
- Closed
'
lRevisiun 161 Hanger DH 16 to centerline of albow
-
as 3' 7" when'it was actually 3' 2"..
.
-
,
(* Note that. this. discrepancy:was
,
included in'the stress analysis but not
'
documented ~on the fab iso).
.l
R
- " Hanger DH-?05 Spring hanger scale read 640f when Open
- Revision 0:.
drawing specified no HL or CL
settings;.. centerline of' pipe elevation o
specified as 325' 11:on hanger
-
' drawing did not agree with 1sometric drawing;' weld missing between pipe and support plate; weld symbolism problems,
' spring support drawing was 9-3/4",
.
verses actual measurement of 14".
'
Location dimension of 8' 7" shown r *
)
L plan was actually 7' 0.
'
Hanger DH-306 Drawing dimension between upper Open
i Revision'2-support rod and lower support rod l
indicated as 8"'when it was actually.
2-1/2"; ~ Drawing-dimension between l
,
L centerline.of pipe and su) port angle
.
E iron indicated as 3' 0" w1en it was L
actually 3' 4".
Clarify note on l-drawing which. stated that' hanger i
,
L should"not be loaded.
J
$
lDrawingGCB-4, Sheet 1 of the drawing showed a
- Closed
l'
- Sheet 1, Revision 6 dimension between the tee (ltem 156)
and the elbow'(Item 170) as 3' 4" l
J when it was actually 2' 4".
(* Note
>
that this= discrepancy was included
' '
'
,
-in the stress analysis but not docu-L, rented on the' fab iso.)
y
..
.
Drawing showed dimension from Hanger Open
.
Drawing HCB-2,
,
,
L-Sheet 2 Revision 6 DH-106 to centerline of elbow as
'
L',
2' 0" when it was actually 2'
4".
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. Drawing / Hanger No.-
Discrepancy Status Item No.
ya
.
.
.
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'
Hanger DH-103, Drawing did not -show weld symbols Open 42'
l
- -Revision 2 or weld-sizes.- Top wall baseplate I<
t,ad 1/4_" gap at lowe.r 'ryht anchor.
!
'
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Bottom wall baseplate had 3/8" gap _
?
at upper right anchor.- Floor base-l...
. plate did not have anchor. bolts as
- specified for; items 9 and 10 but
had installed stods with nuts.
'
A nger DH-106, Upper left anchor bolt on lower wall.
Open
L ::t Revision 4 baseplate was installed at an P
8 degree angle with no beveled washer
- as required by the installation
.
. specification,
'
o L
' Hanger DH-106 Drawing showed.2'
l'" span for Open
'
Revision 1 location of support on ceiling I
!
beam when it was actually 3' 9"; End
'
.of-I beam (pc.-8) ends at mid-point.
s
,
,'
l of 6"' channel 1.nbed, therefore, weld-
'
length is approximately 3"11nstead of 6" as indicated on drawing.-
Hanger DH-108, Drawing showed^all around-fillet weld.Open 45-
- Revision 6:
for six locations and only three sides, were welded; 1/4" shim plate used but not-specif_ied on. drawing.
Hanger DH-116,-
Drawing showed hanger assembly flange -Open
- l R.tvision 1 welded to 6" embedded inserts. The
-
i flange was actually welded to a 4" by 6 plate which.was welded:to the
.
,
insert, with the p1 tie oriented in
,
sucn a way that the lanie to plate
,
weld was just 4" on two sides rather than the 6" required on the drawing.
,
",
- Hanger DH-117 -
- Spring load was 3600# versus Open
'
des gn va ues of CL = 3860f and i
l Revision 1
,
HL = 3814#
'
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LHanger'DH-124, Spring hanger scale read 3050f when Open
,
'
Revision None drawing specified the loading as'
-
Go H'. P.7466 and CL 2896f y
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Discrepancy
~ Status Item No.
Action
"
. ? Drawing / Hanger No.-
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Han Revgert'i-125,
= Drawing.showed a 1/4" fillet weld, Open
,
sico l'
all around, at-each connection
~*
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between Items 1 and: Item 4.(top and'
?
- bottom),andbetweenItem4(topand
' "
bottom) and column when the actual
'
s
.'
Hanger HCB-2-H1 Item f8 on bill of material-was a Open 60-j
'
,
- -
Revision O.
smaller plate than that installed,
'
'
.also it appeared to be stainless:
'
,
'
on the saddle were_not specified
'on the'drawinge
- Hanger HCB-2aH2,_
Lug welds were unders::ed; 1/8" Open 62-
! Revision O'
, to 3/16" verses 1/4 specified. -
c
- Service Water SystemL
'
- Drawing'13-SW-133, The location fcr Valve SW-1B was Open
Revision 4 shown as 2' 5"~from Valve SW-2B when it was actually 2' 0"; two locations were shown-for Support HBD-2-H1, one
,
Iocation-nei r Pump A and the other
near Pump C.
"
Hanger HBD-2-H1, Baseplate anchors were not in Onen
' Revision-unknown-
. agreement with hanger drawing for
.
,
support located near Pump.C; Gaps up
.to 1/4"-existed between entire surface of " free,to slide" plate.and base-
,
plate; No drawing for support located-
,
'
near Pump A which was also designated as HBD-2-Hl.
(Note that 2 suoports
,,
were-notloadbearing.)
I E
. Hanger MBD-2-H2, Baseplate anchors on slide support Open 68&69
. Revision 2 not as shown;on drawing, bolts were s
s apparently welded and ground flush
-with baseplate.
- Hanger HBD-2-H3, Two hanger sketches were provided for Open
e
- Revision 1F2 Support HBD-2-H3 and both designated
'
~
as Revision IF2, each indicated different locations for the support.
k l
.
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Action
'
Drawing / Hanger No.
Discrepancy.
Status Item No.
-
,
,
,
'
1 Hanger HSD-2-H4,c Pipe restraint'shown on Revision 1F
.0 pen-71
'
-
Revision 1F2 of. drawing:did not exist.=
,
Henger'HBD-20-H51,.
'
Revision 2
'
Baseplate anchors on slide' support-Open
'
not as shown on drawing, bolts were apparently welded and ground flush
.;
,q with baseplate.-
'
Various In'the package provided to the NRC Open 73.
-inspection team-there were'several
'
,
,
discrepancies regarding hanger
..
-locaticns and.some outdated. hanger
-1-
'
' sketches were provided. hat were-not
,
. stamped obsolete or-superseded.
,
,
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- -
APPENDIX C-PERSONNEL CONTACTED DURING INSPECTION 7,
- Bame-Title-Affiliation
- ,
- D.-Bauman Design Configuration Document Pioject Lead-AP&L MECrowning
.Mechar.1 cal ~ Engineer AP&L y.
'
- C. Chadhourn Mechanical Engineer AP&L
- M. Cimock Mechanical Engineer
..
'
- W.?Eaton.
Manager of Civil & Structural t:ngineering AP&L R. Edmunds Pipe Stress Engineer AP&L
- H.iGreene Quality Assurance Superintendent
-
- B. Greeson Superintendent of Structural D6 sign AP&L A. Halbert Supervisor Engineering - Site-AP&L-
'i L. Howard:
Plant Engineer - Site-APal
'
"
.M. Huff.;
Supervisor Mechanical Systems AP&L
'D.-James Licensing Engineer AP&L
- G. Jones General Manager of Engia" ring.
AP&L R. Lane Manager of ANO Enginee t"; - Site AP&L A. McGregor Superintenden.t of Engi..F.ng Services AP&L
- P. Hovero-Senior Enpineer - Pipe Supports AP&L-D.
Peschong
- Supervisor Structural Engineering AP&L
.
..B. Redgers-Supervisor'Aechanical Engineering AP&L
' Pipe Stress Engineer-AP&L M. Tull Licensing AP&L
'
.. _
~*W. Turk Manager of Licensing AP&L W. Watson.
Project. Engineer Bechtel
'G. Wiedstein Pipe Stress Engineer AP&L M. Wood Iso Update - Engineering Technician AP&L
- A..Wrope:
Manager of Electrical, Instrumentation &
AP&L Control Engineering
- Attended. Exit Meeting s
i-
)
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b
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