ML20099F373
| ML20099F373 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 11/21/1984 |
| From: | Cain J LOUISIANA POWER & LIGHT CO. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| W3B84-0817, W3B84-817, NUDOCS 8411260388 | |
| Download: ML20099F373 (183) | |
Text
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P O W c R & L I G H T / f4vOALEANS LCXBANA 70100 e p l5952204 INE0.IS SONE November 21, 1984 J.M. CAlt!
President W3B84-0817 A4.05 Director of Nuclear Reactor Regulation ATTN: Mr. Darrell G. Eisenhut, Director Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555
SUBJECT:
Waterford 3 3ES Partial Response to Items from Waterford Review Team
REFERENCES:
- 1) Letter W3B84-0807, J.M. Cain to D.G. Eisenhut, dated October 31, 1984
- 2) Letter, D.G. Eisenhut to J.H. Cain, dated June 13, 1984
Dear Mr. Eisenhut:
The purpose of this letter is to submit revised responses supplementing Issues 6, 7, 19, 20 and the assessment of Collective Significance. The revision to Issuc 6 is provided in accordance with reference 1. The remaining limited revisions reflect information developed since the original submittals and limited technical corrections. The logic and the approaches to resolution of the issues remain unchanged. These revisions have been discussed with your staff.
To facilitate your review, change bars have been provided in the right hand margins of the revised responses to indicate the portions which have been revised.
The revisions to the responses have been reviewed and verified by LP&L QA in accordance with Procedure QASP 19-13. The designated subcommittee of the Waterford Safety Review Committee also has reviewed the adequa;y of the revised responses for resolving the issues raised. The subcommittee scope af responsibility does not include independent validation of the facts.
The complete responses to Issues 1 and 10 with respect to QC inspectors will be submitted shortly.
8411260388 841121 <3 00!
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Mr. - Darrell-G; Eisenhut '
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Page 2 l W3B84-0817 7 November 21. 1984 , , . . , . ,,
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The Task Force has not yet-dompleted its independent. validation of the
, facts. The Task Force h.ts'committTed to' notifying me and the NRC
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immediately should it find cignificant-deviations in the course of its
. validation. Inzthe ever[c $f such notification, LP&L will amend individual responses as may be neces'aary.
Sincerely. .
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, , . s M. Cain JMC:DED:pc1 ,
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Attachments
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Mr.~Darrell G. Eisenhut, Director. Page 3 l W3B84-0817' November- 21, 1984 cc: 'Mr. R.S. Leddick- Mr. J. Harrison Waterford 3 QA Team Leader Mr. D.E. Dobson Region,III.
700 Roosevelt Rd.
Mr. R F. Burski' Glen Ellyn, IL 60137-Mr. K.W. Cook Mr. J.E. Gagliardo Director of Waterford 3 Task i
Mr. T.F. Gerrata Force Region IV Mr. A.S. Lockhart 611 Ryan Plaza Suite 1000 t
Arlington, TX 76011 Mr. R.P. Barkhurst Mr.l?. Couchman Mr. L. Constable NUS Corporation 4- USNRC - Waterford 3 910_Clopper Road-GaitherrLurg, MD 20878 Mr. R.D. Martin 1 U.S. Nuclear Regulatory Commission Mr. R.L. Ferguson Region IV JNC Nuc. lear Industries 611 Ryan Plaza Suite 1000 1200 Jadwin,' Suite 425 i Arlington, TX 76011 Richland, WA 99352
) Mr. D. Crutchfield Mr. L.L. Humphreys l JU.S. Nuclear Rer21 story Commission UNC Nuclear Industries
- Washington, D.C. 20555 1200 Jaawin, Suite 425 Richland, WA 99352 i Mr. G. Knighton, Chief Licensing Branch No. 3 Mr. G. Charnoff Division of Licensing Shaw, Pittman, Potts &
. Washington, D.C. 20555 Trowbridge 1800 M. St. N.W.
Mr. M. Peranich Washington, D.C. 20555 Waterford 3 Investigation and i Evaluation Inquiry Report Team Dr. J. Hendrie
, Leader 50 Bellport Lane 4340 E.W. Hwy. MS-EWS-358 Bellport, NY' 11713 Bethesda, MD 20114 Mr. R. Douglass Mr. D. Thatcher Baltimore Gas & Electric Waterford 3 Instrumentation & Control 8013 Ft. Smallwood Road Leader Baltimore, MD 21226
- 7920 Norfolk Ave. MS-216 Pethesda, MD 20114 Mr. M.K. Yates, Project Manager Ebasco Services, Inc.
- l. Mr. L. Shao Two World Trade Center, 80th Waterford 3 Civil / Structure Team New York, NY 10048 Leader.-
5650 Nicholson Ln. Mr. R. Christesen, President Rockville,'HD Ebacco Services, Inc.
Two World Trade Center
, New York, NY 10048
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PROGRAM. PLAN ISSUE: 20 DATE: 11/21/84 TITLE:
Construction Materials Testing (CMT) Personnel Qualification. Records.
DESCRIPTION'0F ISSUE:
Verify the proper certification of construction materials testing personnel.
LP&L APPROACll TO RESOLUTION:
CEO has been contacted to assist in providing additional background information or justification for qualification of QA/QC personnel identified as part of.NCR W3-F7-Il6.
A verification program has been established to review the professional credentials of 100% of the CEO CMT site. QA/QC personnel, including supervisors and managers who performed safety related functions at Waterford III during its.
construction. Criteria for certifications or qualification:of QA/QC personnel will be based on ANSI.N45.2.6-1973 and ,
SNT-TC-1A for QC inspection personnel and construction QA program requirements for QA personnel.
In addition background inscstigations will be performed for personnel in all groups.- If qualification of an individualDo can not be verified, appropriate site nonconformance documentation will be initiated to document evaluation of safety a.
eignificance and corrective actions, including reinspection of work performed'as necessary. , -
x For GEO QC Inspectors remaining on site, a reverification is being completed of proper qualification in accordance with ANSI-N45.2.6-1973. -
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WORK INSTRUCTIONS AND PROCEDURES EMPLOYED:
COMPANY PROCEDURE NUMBER TITLE Ebasco QAI No. 32 .
Instructions for Verifications of QA/QC PersonncI Qualifications.
LP&L QASP 19.12 Review of Contractor.yA/QC Personnel Qualification Verification.
QASP 19.13 Response Validation
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ORGANIZATIONS INVOLVED:
ORGANIZATION FUNCTIONS PERFORMED PERSONNEL QUALIFICATION / TRAINING REQUIREMENTS Ebasco 1) Verify Education / Experience of QA/ 1). Training requirements to QAI-32.
QC personnel.
2a) Review program requirements of CEO, 2) Ebasco's Quality Resources Trainic3 Manual-1 (QRTM-1) review and collect data and delineates the requirements for qualifying records identify inspectors whose reviewer. QAI-14, " Training and Qualification qualifications are not verifiable Requirements for Quality Assurance Records Personnel"'
against ANSI N45.2.6-1973. endorses QRTM-1 and requires all reviewers have training SNT-TC-1A and QA program on procedures they are reviewing to. For qualification /
1 requirements for QA personnel. certification ' filed training requirements are QAI-32 and ANSI N45.2.6.
b) Determine, to the extent feasible, inspections performed by personnel whose qualifications are not verifiable.
c) Disposition quality documentation generated by LP&L in item (5) below. ,
LP&L 1) Audit Ebasco's implementation of 1) (a) Indoctrination / training to LP&L & Ebasco QAI-32. procedures, ANSI N45.2.6-1973 & 1978, ANSI N45.2.23-78 SNT-TC-1A-75 & interpretations.
(b) rJentation as to task objectives, organizations,-
and associated responsibilities and dut ies.
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ORGANIZATION 3 INVOLVED: (CONT'D)
ORGANIZATION FUNCTIJNS PERFORMED PERSONNEL QUALIFICATION / TRAINING REQUIREMENTS LP&L Cont'd .
(c) OJT for three days to assure knowledge, understanding, and proficiency demonstration.
(d) Individuals selected have inspection related and/or.
were involved in the training / certification or review of inspection personnel types.
(e) Personnel involved in this ptacess have noc worked.
tor Ebasco or GEO.
- 2) Review all those verified by 2) See item 1 above.
Ebasco.
- 3) Sample Education / Experience 3) See item 1 above.
verification of GEO performed by Ebasco.
- 4) Perform final management 4) Review Board - Three Senior LP&L QA personnel qualified .
determination of the qualifications to ANSI N45.2.23 (1978).
of individuals who are potentially unqualified.
- 5) Initiate suitable quality 5) LP&L Lead Auditor who is qualified .tc ANSI N45.2.23 documentation in cases where (1978).
inspections were performed by personnel where qualifications could not be verified.
- 6) Make final determination on 6) LP&L QA and Project Management, dispositioning of quality documentation mentioned in 4) above by Ebasco.
- 7) Validate response per QASP 19.13 to 7) Validation will be performed under the dirset assure positive statements of fact supervision of the LP&L Lead auditor who is qualified to are substantiated. ANSI N45.2.23 (1978).
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ATTACHMENT 1 FLOW CHART-INSPECTOR QUALIFICATION REVIEW-1 Initial LP&L Reviev /dditional Background LP&L Review Board Group Leterminations Investigations Determinations Final Results.
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' Qualified _ Qualified j A f Adrinistrative d Qual;fied Background Deficiency Only 3 '
Verificatior (Qualified) File Merger I
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Not Inspector Files Reviewed From Qualified' Qualified Ebasco & LP&L -
V Poter,tially Qualified hot Not Qualified D I Qualified ,
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' V Indeterminate -> Not CAR P j
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Qualified) y Dispositioned
- Not 9 Qualified 4
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RESPON3E ITEM N0;: 6 (Revision 1)
TITLE: Dispositioning of Nonconformance and Dicercpancy Reports NRC' DESCRIPTION OF CCNCERN:
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The staff conducted a review of Ebasco nonconformance reports (NCRs) randomly selected from the Ebasco QA vault and the NCR tracking system. The selected NCRs _ were reviewed for content, compliance with procedures, accuracy, completeness of the dispos-ition and final closure. Of the NCRs reviewed it is the staff's judgement. that approximately one third contained questioncble disposicions. Ocher NCRs were found still open.
The implied safety significance is that improperly dispositioned NCRs or lack of NCR closure could place the quality of installation in question.
For example, Ebasco NCR-W3-5364 identifies that welds were painted before the final weld inspection was performed. The NCR was closed out with a letter stating that the final inspection will be performed to inspect only for
< undersizing and lack af weld material where installation drawing calls for weld material. No paint was to be removed therefore the inspector could not inspect for welding defects. .
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The NCRs reviewed by the staff dealt with a wide variety of issues. The following is a list of example Ebasco NCRs that th2 staff feels contain
. questionable dispositions or exceeded closure tire requirements.
Ebasco W3 NCRs
, NCR-7139 NCR-7177 FCR-3912 .NCR-7182 NCR-5563 NCR-7181 NCR-7184 NCR-6159 NCR-6723 NCR-3919 NCR-7547 NCR-6221 NCR-1650 NCR-6511 NCR-6623 NCR-4219 NCR-5586 NCR-7432 NCR-7180 NCR-4137 NCR-6165 NCR-4088 NCR-7099 NCR 0786 NCR-6597 NCR-7533 NCR-7179 NCR-7140 NCR-5565 The staff also found similar type problems related to Mercury NCRs in that"the dispositions were questionable; supporting documentation could not be located; rework appears to have not been accomplished; NCRs were not processed; a sufficient basis was not provided; and closure basis was inadequate.
i The following NCRs fall into these categories:
Mercury NCRs 180 420 528 568 625 255 429 540 591 656 268 438 554 594 6 5?,
363 487 560 595 380 491 565 614 6-1
Addicirnnily duting thin ravi;w thm ctaff found probl ms with Eba::co diecrc;p'.ncy reports (DRs) in that . it appears some DRs should have been elevated to NCRs:
closure references were incorrect or inappropriate; closure action was improper; documentation was inaccurate; closure was via a DR, should have been an NCR; disposition failed to address the discrepancy; and the disposition of "use-as-is" had insu.'ficient basis.
The following DRs fall into these categories:
Ebasco DRs Related to Turnover Packages Q2-CS-1C-27 ED-1C-1143 Q2/3-FW/1C-851 Q1-RC-LWS-RC-2 Q2-SI-1C-89 LW3-RC-29 QMC-APO-P47E Q2-LW3-SI-10F/E C(W)-lC-342 CC-lC-6 The staff concludes that some Ebasco and Mercury NCRs and Ebasco DRs vere questionably dispositioned and that LP&L shall (1) Propose a program that assures that all NCRs and DRs are appropriately upgraded and adequately dispoeitioned and corrective action cornole ted , and (2) correct any problem detected.
DISCUSSION:
LP&L initiated a program, beginning in February 1984, to review Ebasco site Nonconformance Reports (NCRs) to verify the effectiveness of the Waterford 3
" deficiency reporting /disposid on programs during construction. Inat program consisted of a review of Ebasco site NCRs clor,ed prior to initiation of the ,
program (approximately 7100). Each Eba.ico site NCR was reviewed and independently assessed by LP&L to determine if:
o The disposition addressed the described discrepancy; o The NCR was reviewed for reportability 10CFR50.55(e) and 10CFR21; and c The NCR had received the appropriate signatures.
This response discusses and presents summary results of the original review and a significantly expanded program addressing dispositioned NCRs/DRr (voided and administrative 1y closed NCRs are addressed in the respo.1se to Issue 13). This program provides adequate confidence that the overall construction deficiency reporting / disposition system was effectively implemented. Corrective action as a result of the expanded review is also discussed. Discussion of the issue is structured along :he lines of the major elements of the expanded program as follows:
I. Review of the specific nonconformance reports and deficiency reports identified by the NRC.
II. Review of Ebasco Nonconfornance Reports III. Review of Mercury Nonconformance Reports IV. Review of Ebasco Deficiency Repor s.
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Thrc'J ~ gtn3ral .conclu2 ions htvo' r2sultad ' to ~dato1 from the original- and axpanded ~
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Jreviews, asfollows:
11: No a'dditienal condition was identified in these' reviews which, were 'it
- to have'-remained ~ uncorrected, would have affected adversely the safety ~
"of operations of Waterford'3.
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- 2. Corrective action ~' required as La result .of- the reviews involved correction of documentation deficiencies, reinspection or enginee*ing
-evaluation and only limited hardware rework.
i' 3.- Due to tihe structure of ~ the1 filing system, eysteaatic review of - the' Waterford L construction - deficiesy.' documentation: .is dif fie r.lt , but
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is achievable.-
I.- Review of the Specific NCRs and DRs identified by the NRC
- The Ebasco and breury NCRs and :the Ebasco DRs identified by the NRC were first reviewed by Ebasco tQuality . Assurance Engineers. The NCRs 'and - DRs .
were . reviewed for ' proper disposition, corrective action completion, L appropriate documentation, and proper closure.- Upon completion of Ebasco's review and ' required corrective actions, LP&L QA reviewed the NCRs . and l corrective actions taken by Ebasco, and sampled the~ Ebasco review of DRs.
LP&L Project Engineering reviewed the NCR's for technical content.- The-I, review of NRC identified Ebasco and Mercury NCRs . and Ebascc DRs was scoped
- as follows
t A. Ebagco Nonconformance Reports
[ Thirty Ebasco NCRs are identified by the NRC in this issue. In .
l addition, seven Ebasco NChs related to this issu'e are specifically
- identified in Fupplement 7 - to the Safety Evaluation Report (SSER)*-
l- which was issued on October 1, 1984. Attachment 1 summarizes the i
4 results of' the review of NRC identified Ebasco NCRs.
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- B. Mercury Nonconformance Reports 2
L Twenty-three Mercury NCRs are identified by the NRC in this issue. An additional fifteen Mercury NCRs related to this issue are specifically identified in the SSER. Attachment 2 summarizes the . results of the
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review of NRC identified Mercury NCRs.
!- C. Ebasco Deficiency Reports 1-h Ten Ibasco DRs are identified by the NRC in this issue. An additional j three Ebasco DRs related to .this issue are specifically identified in :
) the SSER. Limited documentation deficiencies were identified and j corrected, none of which were safety significant.
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- NUREG 0787 (SER Supplement 7 - September 1984) b j
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--The review of tha..NRC ideritified documsnts has besn -completed. While QA program
- procedurali deficiencies existed, no safety significant deficiencies have been identified..
II. Review of Ebasco Nonconformarce Reports The review of Ebasco site'Nonconformance Reports encompassed approximately
'98'i of the site - NCR numbers issued by Ebasco during the construction of
' Waterf ord . 3. The review consisted of several elements, each with it:5 own particular level of review. Figure 6-1 depicts the. elements-of Ebasco NCR
. review - process ' in the form of a flow diagram, in order to facilitate understanding of the process.
FIGURE 6-1 REVIEW OF EBASCO FCRs 3
NCR'S ISSUED BY E6ASCO 7,800 NCR'S INITI AL ' LPSL REVIEW DCJWLED LPSL REVIEW NCR'S CLCSED PRICR ' NCR'S CLOSED TO 2/84 AFTER 2/84
.% 7,10 0 NCR'S . 532 hC R' S g
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P C*EN TI A LLY POTENYt A LLY
, CEFICIENT QCFICIENT S A' 'i FACTC R Y ESASCO
- REVIEW S ATISFAC TOR Y
& RESPONSES geg S ,v 461 NCR'S
.s6663 NCR'S .s 437 NC R' S v 7, NCR'S Y
._ ~ $08 l
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t LP&L R EV IE W a 08 5 NCR'S DETAILED SAMPLE REWW S ATIS FACTO R Y EFICIENT . PROGR AM DEFIC'#NCIES t46 NCR*$
S AFETY StGNIFic TNT DEFICIENCES.0 NCR !
124 N C R'S y N362 NCR'S S AT6SFACT04 v DEFICIENT. PROGR AM OEFICIENCIES - 33 N Ci? ' S SAFETY S I G N IFIC AN T DEFICIENCIES - 0 NCR'S
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Tha following parrgr:phs discu;s ths individual clen;nto of th9 rcvi1w of Ebasco NCRs:
A. LP&L QA Review of Ebasco NCRs closed prior to February 1984
- 1. Initial Revie In February 1984, LP&L QA initiated a review of Ebasco NCRs. This review was undertaken to verify, by way of a Work Instruction, that:
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- a. The disposition addressed the described discrepancy;
- b. The NCR was reviewed for reportability under 10CFR50.55(e) and 10CFR21; and
- c. The NCR had received the approptiate signatures.
Approximately 7100 Ebasco NCRs were reviewed and 437 potentially deficient NChs were identified. Upon completion of the evaluation, it was n termined that 122 NCRs were deficient in disposition, corrective action, software or closure, or combinations thereof. Corrective action required as a result of tnis review involved only limited hardware rework and correction of documc..tation deficiencies.
Seventy-two of the NCRs were considered potentially deficient for lack of documented evidence that they had been reviewed for reportability per 10CFR50.55(e) or 10CFR21. Subsequent documented reviews of these NCRs determined that none were reportable.
- 2. Det* ailed Review LP&L selected 124 (a'pproximccely 28%) of the potentially deficient NCRs identified in the initial review for an in-depth review. This review included hardware verification for rework / repair, software verification for updating as-built drawings and specifications and evaluation of documentation for the required corrective actions and retrievability of documentation.
As a result of this detailed review, 33 NCRs were found to be deficient, and seven CIWAs were initiated to address the deficiencies. None of these deficiencies met the criterion for safety signific, nce. Corrective action for 30 of the deficient NCRs involved correction of documentation deficiencies, reinspection or engineering evaluation. For the remaining three, limited discretionary rework is being performed.
B. Detailed LP&L QA Review of Ebasco NCRs closed after February 1984 Ebacco NCRs closed after February 1984 were reviewed as a separate group by LP&L QA. Review of thesc NCRs was in-depth and was for the purpose of verifyiag proper disposition, ariequate documentation to support the required corrective action, required software changes completed and proper closure. Five hundred thirty two (532) NCRs were reviewed with 71 NCRs requiring resolution of comments, Of those 71 NCRs, 24 vere determined to have valid deficiencies. ' Corrective action for 22 of the deficient NCRs involved correction of documentation deficiencies, reinspection or engineering evalua tion. For the remaining two, limited discretionary rework is being parformed.
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C. Ebasc' NCR Clo'ur, Timelin m
'Jith respect to the NRC concern regarding timeliness of Ebasco NCR closure, Ebasco procedure ASP-III-7, " Processing of Nonconformance", required completion of corrective action within twenty (20) days of receipt of the dig ositioned hCR. If the verification of corrective action was not completed within the allotted twenty days, a written regaest for extension was to be filed with the Ebasco Quality Assutance Department for approval.
The twenty dc) time period did not begin until the nonconformance report had been dispositioned and evaluated by the appropriate departments. The twenty day requirement was for administrative control only and did not adversely af fect the quality of Waterford 3. In December, 1983, Ebasco procedure ASP-IIl-7 was revised to delete this requirement.
All Ebasco SCRs closed as of approximately the end of September, 1984 (Approximately 98% of the Ebasco NCRs issued) were subjected to an LP&L review as described above. While program de2f ciencies existed, and minor rework was required, no safety significant deficiet.cies have been identified.
III. Mercury Nonconformance Reports Mercury dispositioned approximately 3700 Mercury NCRs. Of these, approximately 1700 'iere upgraded to Ebasco NCRs and, as such, were reviewed as Ebasco NCRs (See Section II of t his response). The remaining Mercury NCRs were reviewed as follows:
A. Mercury NCRs dispositioned "Use-As-Is" were reviewed to assure that they were upgraded to Ebasco NCRs, as required. As a result of this review, eleven NCRs were deemed to require upgrading to Ebasco#NCRs.
These eleven NCRs are' now identified on Ebasco NCRs, and were
, processed under the Ebasco NCR proggam.
B. Approximately 1850 Mercury NCRs were dispositioned " rework / repair" or
" rej e c t . " In most cases, when Mercury designated a deficiency to be corrected by " repair", it was, in fact, a " rework." For example, in dispositioning rejected welds, Mercury would specify the weld be
" repaired" in accordance with procedures to meet the design requirements. This is actually a " rework" disposition. Mercury procedures did state that deviatione from original design or technical specAfication outside the tolerances allowed was a " repair". Mercury procedures required nonconformances meeting this criteria to be upgraded to Ebasco NCRs so that these Javiations would be reviewed and approved by E5asco.
A random sample of 66 Mercury NCRs from those dispositioned " rework /
repair" was selected for review. These NCRs were reviewed for proper disposition, adequate documentation of corrective actions required and proper closure. LP&L QA reviewed each sampled Mercury NCR in accordance with QASP 19.17. Deficiencies were corrected and docum2need. None were found to be of safety significance.
C. Seven hundred twenty f!ve (725) o' the 1850 Mercury NCRs dispositioned
" rework / repair" and " reject" were reviewed by Ebasco for reportability per 10CIM150.55(c). None of the NCRs were' determined to be reporta51e.
LP&L QA selected a random sample of 64 of thes* NCRs for a reportability review and the Ebasco conclusions were confirmed.
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D. Mercury documented material conditionally released from Ebasco cn Material Receiving Reports (MRR) and assigned Mercury NCR numbers to each such MRR in accordance with Mercury Procedure SP664.
Approximately 120 Mercury NCRs of this type were identified by Ebasco.
LP&L reviewed the Mercury files and, although the conditional releases appeared to have been properly handled, there were instances where supporting information (Ebasco NCRs, DNs) was neither referenced nor included in the documentation package. Tbc supporting information is available and will be either included er referenced, in the NCR packages, as appropriate.
This review of dispositioned Mercury NCrs is complete. While program deficiencies existed, no safety significant deficiencies have been identified.
The results of these sample reviews establish a 95% confidence level that at least 95% of the total population of Mercury NCRs do not contain unreported conditions reportable under 10CFR50.55(e) or 10CFR21.
IV. Review of Ebasco Deficiency Reports The Ebasco QAIRG review of contractors records required tnat deficiencies be documented on Deficiency Reports in accordance with QAI-9, " Review and Handling of Construction Installation (DRs) Records". A random sample of DRs generated as result of the review of Mercury and Tompkins-Beckwith records was reviewed for proper closure. For each contractor, 230 QAl 9.2 Deficiency Repert Sheets were selected and reviewed as follows:
A. The revicw of Deficiency Reports on Tompkins-Beckwith included 115 Deficiency Report Sheets on piping and one hundred fifteca QAI 9.2 Deficiency, Report Sheets on seismic hangers and supportg. These QAI 9.2 Deficiency Report Sheets included approximately 856 DRs. This review identified 12 DRs which required engineering evaluation and concurrence. Although minor deficiencies, such as missing references, signatures or dates were identified, the JR closures were satisfactory.
B. The review of the 230 Mercury QAI 9.2 Deficiency Report Sheets was divided equally among P-2 and P-3 tubing, and tube track supports.
These QAI 9.2 Deficiency Report Sheets included approximately 1173 The review identified 31 DRs which required engineering
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DRs.
evaluation. The engineering evaluations are in progress. Although minor deficiencies, such as missing references, signatures or dates were identified, the DR closures were satisfactory.
LP&L QA performed audits of the Ebasco review. These audits included random samples of the Mercury and Tompkins-Beckwith DRs reviewed by Ebe.co.
While documentation deficiencies existed, no safety significant deficiencies, or deficiencies requiring ra ork, have been identified.
CAUSE The review program verified that deficiencies were generally processed in accordance with the site procedures. However, those procedures did not' provide adequately rpecific guidelines for the implementation of procedural requirements which led to excessive need for judgements and interpretations.
This program weakness led to the inconsistenciec in handling deficiencies at Waterford 3 which have been identified by 1.P6L and the NRC.
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I GENERIC IMPLICATIONS The_ review program encompassed apprerimately 98% of the Ebasco NCRs and statistically justified samples of Mercury NCRs and Ebasco DRs. The results of an in-depth review and verification of a conservative sample of NC"a and DRs has provided adequate confidence that the deficiency system did not allow conditions in dispositioned NCRs/DRs to remain unreported per 10CFR50.55(c) and 10CFR21.
SAFETY SIGNIFICANCE .
LP&L has performed a review of major elements of the construction deficiency reporting / disposition system. The results of this review indicate th:t, in general, _the system was ef fectively implemented. The procedures contained the basic requirements for documenting and controlling deficient conditions. The deficiencies identified during the review of nonconformances are considered minor in nature and were generally resolved with the addition of documentation or further evaluation. The items dispositioned as rework were based on good engineering practice or management conservatism rather than on safety significance. There is no recognized reason that this issue should constrain fuel load or power operation.
CORRECTIVE ACTICN PLAN / SCHEDULE All reviews and required corrective actions are completed.
ATTACHMENTS
- 1. Ebasco Nonconformance Reports Identified by the NRC.
- 2. Mercury Nonconformance Reports Identified by the NRC. '
PEFERENCES None.
l 6-8
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l ATTACHMENT I EBASCO NONCONFORMANCE REPORTS IDENTIFIED BY THE NRC The alawing is a list of EBASCO Nonconformance Reports (NCRs) identified by the NRC in Issue No. 6 and in Supplement 7 to the Safety Evaluation Report (SSER). The list identifies the NRC Concerns with each NCR and the Resolution or Corrective Action. The list also summarizes additional concerns identified as a result of the LP&L Review and the Resolution or Corrective Actica. It should be noted that dispositioned NCR's were reviewed for reportability under iJCFR50.55(e) and 10CFR21 and none were found to be re; ortable.
A. Ebasco NCR's Identified in issue No. 6
- 1. NCR U3-1650 (a) NRC CONCERN How was it determined which bolts to retest when QCP 309 did not require the recording of tester serial number on previous tests?
RESOLUTION LA CORRECTIVE ACT1oN All uses of gauge QC 4.2.2 by F&M (QCP-309) were accepted-as-is by ESSE with no further action required.
Tension tester gauge QC 4.2.2 was issued and tracked on Ebasco's M&TE Master Log. Review of this log indicated each contractor that was issued QC 4.2.2 during the timp it was out' of '
calibration. Each contraccor reviewed his installation records to see if tension testing was done during this time. If so, a description of the work wa . given to ESSE for evaluation. Each use was accepted as-is by ESSE based upon the small Agree of error found during recalibration.
(b) LP&L IDENTIFIED CONCERNS
- 1. All issuances of subject pressure gauge not properly cddressed in NCR.
- 2. Statement by user of pressure gauge is no* acceptabic for dispositioning of NCR.
RESOLUTION OR CORRECTIVE ACTION
- 1. Records of tension tests were esaluated by ESSE of those users (contractors) of gauge not previously addressed.
- 2. Review conducted of contractor tonsion test records did not reveal any use of pressure gauge by this individual.
Documentation of this review attached to NCR.
6-9
ATTACHMENT 1
- 2. NCR W3-3912 (a) NRC CONCERNS
- 1. Involved nine 23J-2 type supports discovered during walkdown for which the fit-up inspection was by-passed. The original NuR disposition failed to address the actions required to prevent the reuse of the items. Attachment No. 14 of this NCR identified this issue which was resolved by stating "it was not required for .the disposition of this NCR..." No other NCR was recpened or referenced to resolve the issue.
RESOLUTION OR CORRECTIVE ACTION
- 1. Support #8 was not removed because of HVAC interferences.
This support will be properly tagged as "not to be utilized-nonconforming".
- 2. Sup;> ort #13 (angle to plate) would be acceptable - for' reuse in its intended design application since it would not be possible to cluster enough tubing attachments to reach the yield point of the stru:ture.
< 3. The remainder of the supports (angle M existing steel) were removed. Since the material is traceable by heat number, it j is approved for safety-related applicatione.
- 3. ::CR W3-3919 (a) NRC CONCERNS 1, 530' more tubing installed than received.
- 2. Requisition on warehouse (ROW) changed using Liquid Paper.
- 3. 10% of OCR Packages selected to verify heat number of j installed tubing. Only one (1) OCR Package actually reflected heat in question.
t RESOLUTION OR CORRECTIVE ACTION i
NCR re-opened and re-evaluated by QA Engineering and ESSE. Final i
evaluation was to accept-as-is based upon the contractor's
, Material Control Program.
- 4. NCR W3-4088 (Mercury 491)
L '
(a) :lRC CONCERNS There was no description attached to the NCR to verify that corrective action was accomplished or completec.
6-10
ATTACHMEFT~l
- 4. NCR' W3-4088 (Continued;
> s-RESOLUTION OR CORRECTIVE ACTION
'1. _ Found . and attached' a ' copy of LP&L CIWA - 828372, which was issued to perform the corrective action for NCR-W3-4088,
- 2. Found and attached a Mercury Q.C. . report which verifies-adequate completion of corrective action.
- 3. Found ' and attached a Mercury weld ~ data report for the replacement welds.
- 4. - Found and -attached a copy of - drawing 100-T-035-A, - which reflects the replacement welds described in #3 above.
(b) LP&I IDENTIFIED CONCERNS
- 1. Inadequate ~ "use-as-is" .just111 cation provided by engineering, .for discrepant items B, C, & G on NCR
-attachment #1.
- 2. Drawing 100-T-035-A showing the affected' instrument line was not attached to the NCR.
! 3. Supporting weld data documentation was not attached to the NCR.
l RESOLUTION OR CORRECTIVE ACTION
- 1. C%cained and attached additional ESSE evaluations to the
! NCR.
- 2. Obtained and attached copy of drawing 100-T-035A to the NCR.
- 3. Obtained 1nd attached a copy of Mercury's weld data report for the replacement welds.
I
- 5. NCR W3-4137 (Mercury #420)
I (a) NRC CONCERNS I
- 1. Improper NCR closure and reopening.
- 2. Incorrect reporting system (DN in lieu of NCR).
l RESOLUTION OR CORRECTIVE ACTION
- 1. NCR-W3-4137 was reopened and processed in accordance with npplicable procedures.
I (b) LP&L IDENTIFIED CONCERNS
- 1. NCR Cor ective Action did not adequately correct thi, discrepancies.
- 2. DN-SQ-1991 was not properly processed in accordance with the i applicable procedures.
i 6-11
b ATTACHMENT 1
- 5. . NCR W3-4137:(liercury #420)
' RESOLUTION OR CORRECTIVE ACTION
- 1. Eupport was reinspected te provide "as-built" and submitted to engineering for design evaluation.. ESSE evaluated the condition to be acceptable and drawing was revised to i reflect existing field condition.
- 2. . Corrective action for violation of Procedure WQC-150 (DN in lieu of NCR) cannot be accomplished since subject procedure has been retired.
- 6. NCR W3-4219 (a) NRC CONCERNS i There are no records for rework or . reinspection to indicate satisfactory reinstallation of supports and sample lines.
RESOLUTION OR CORRECTIVE ACTION Sample line was reworked to original design and tracked on Mercury NCR 684. Reference Attachment #3 of NCR W3-4219 for an
. acceptable evaluation by Construction Engineering.
- 7. NCR W3-5563 -
(a) NRC CONCERNE i 1. Inspections signed off by an unqualified inspector.
! 2. Inspection reports co-signed by Level II inspector 3 years and 5 months later.
) RESOLUTION OR CORRECTIVE ACTIONS
. NCR reopened and CIWA #011340 written to re-inspect Fuel Handling '
Building (FHB) Crane. This work was completed and CIWA closed on l 11/15/84 The installation was found to be acceptable.
8.
NCR W3-5564 (a) NRC CONCERNS Disposition of NCR for inspection through paint is unacceptable, due to paint precludes adequate visual inspection of the welds.
RESOLU p ;N OR CORRECTIVE ACTION Downgrading of FHB stairways from seismic class 1 to seismic class II eliminaten the requirements for visual inspection.
6-12
Y 3 ATTACHMENT 1 -
, 8. NCR W3-5564'(Continued)
'(b) LP&L IDENTIFIED CONCERNS,
- 1. No QC verific Ition signature on the sketches provided in
' attachment #23 of the NCR.
- 2. Insufficient ESSE avaluation for downgrading . seismic Class I stairs in the FHB, to seismic class II.
RESOLUTION OR CORRECTIVE ACTION
- 1. Ebasco Q.C. performed and documented a verification of the items' identified in the stairwell on NCR attachment #23, and attached the results to the NCR as attachment #24.
- 2. ESSE Electrical and HVAC revieved the information in NCR attachments #23 and #24, and determined them to be non-safety.
- 9. NCR W3-5565 -
l (a) NRC CONCERNS j 1. 'Ihe Qualification of the Q.C. inspector who performed . rhe l inspection o( reeving of the F.H.B. Crane.
- 2. the documentation of the reinspection was not attached to the NCR as directed by the NCR.
RESOLUTION OR CORRECTIVE ACTION i 1. The Fuel Handling Puilding Crane was turned over to the j LP&L with subsequent testing and reinspection performed by I
the LP&L on 1/29/83 per proceduro SPO-40-002,
- 2. The testing and inspection data performed by LP&L has been attached to the NCR.
(b) LP&L IDENTIFIED CONCERI'S, !
Nonconformance was reopened on April 26, 1984 to add attachment l
IA and cicoed the same day without documented evidence that the investigation as required in the attachment was actually performed.
RESOLUTION OR CORRECTIVE ACTION l
i
- Attachment 5 has been added to the NCR to reference LP&L test procedure SP0-40-002 which documentel the final functional testing of the subject crane.
6-13
ATTACHMENT 1
- 10. NCR W3-5586 (a) NRC CONC 53NS
- 1. kelders Test Lab was not on Mercury's gen 11 fled suppliers list, and this item was not addressed in the NCR disposition.
- 2. Statement provided by Welders Test Lab, that "a Mercury Inspector reviewed all tests", is not adequate.
RESOLUTION OR CORRECTIVE ACTION
~ 1. Mercury audits of Welders Test La' for years 1979, 1980, 1981 & 1982 added as information to verify Mercury surveillance of supplier't. activities.
- 2. Statements from present ard former contractor employees and corporate officials added to support the fact that qualified contractor personnel reviewed all tests.
- 11. NCR W3-6159 (a) 3 C CONCERNS:
- 1. Traceabil,ity problems were not identified and addressed by
~
the NCR.
- 2. The sample used for tensile testing the welds was questionable in that the worst case example should have been used for the test.
RESOLUTION OR CORRECTIVE ACTION <
- l. All tubetrack materials are purchased, received and maintained by Ebasco's QA Program. Material is requisitioned by subcontractors from the Ebasco warehouse.
- 2. Calculated st ress levels imposed on the weld were conservatively established, taking credit for only 50% of the specified weld length and assuming dortgn basin earthquake.
(b) LP&L IDENTIFIED CCNCERNS
- 1. Six (6) out of twenty-two (22) welds were found to contain weld defects. What was done to increase the sample si:e7 ,
- 2. No evidence to indicate the test sampics were selected from '
" Worst Case" installations. I 6-14
1 l
l ATTACHMENT I l
)
- 11. NCR W3-6159 (Continued)
RESOLUTION OR CORRECTIVE ACTION
- 1. QAIRG records review required reinspection of 67". of tube track. No other rejectable conditions were found.
- 2. Calculated stress levels imposed on the weld were conservatively established, taking credit for only 50% of the specified weld length and assuming design basis earthquake.
- 12. NC2 W3-6165 (a) NRC CONCERN
- 1. There is no indication of measures taken to preclude recurrence.
RESOLUTION OR CORRECTIVE ACTION
- 1. A review of filler metal requisitions and T&B time sheets indicates that welder R-7 not R-1 made the weld concerned -
.nd R-1 was not employed during the time the weld wts.made, therefore, measures taken t o. preclude recurrence was not necessary.
(b) LP&L IDENTIFIED CONCERNS
- 1. Documented verification that welder R-1 was not on site should be included.
RESOLUTION OR CORRECTIVE ACTION
- 1. Review attached to NCR indicating welder R-1 not on site during the time period wold was made.
- 13. NCR W3-6221 (a) NRC CONCERN
- 1. Wold control records signed off by Level I Inspector.
- 2. Letter of designation based on revision of QA Manual not if effect at the time of letter issuance.
RES0t.UTION OR CORRECTIVE ACTION
- 1. l.P& L QA ovaluated inspectora experience, education. nd training and determined the inspec to r was qualified to .
perform the designated activities.
6-15
ATTACHMENT 1
- 14. NCR W3-6511 (a) NRC CONCERNS
- 1. The NCR only addressed the fact that the maximum gap was violated, should have included undersiz e weld; lack of fusion; are strikes and undercut.
- 2. There are no records of rework or reinspection.
RESOLUTION OR CORRECTIVE ACTION
- 1. Support wac reinspected by Ebasco. QC and 1s 1.i.ilt . data supplied to ESSE. ESSE accepted support "as-is".
- 2. Documentation posted to Mercury installation package to assure update to as-built installation documentation.
- 15. NCR W3-6597 (Mercury #2870)
(a) NRC CONCERNS
- 1. NCR exceeded the closure time requirements of ASP-III-7, section 6.1.3.a.
RESOLUTION GR CORRECTIVE ACTION
- 1. The closure timo requirement is generically addressed in Issue '#6 report.
(b) LP&L IDENTIFIED CONCERNS
- 1. No tra:eability for installed bolt, nut and lockwasher.
- 2. No torquing for the bolting above.
- 3. DCN not refirenced on drawint.
- 4. Were new Hilti's installed?
If this was a ro-verification of torque, where is original torque documentation?
RESOLUTION OR CORRECTIVE ACTION
- 1. None required -
purchased commercial grade with C of C provided by supplier.
- 2. No torque value required.
- 3. DCN was incorporated on drawing.
- 4. New Hilti's were not installed. This was the original torque iuspection.
t 6-16
v.
ATTACHMENT 1
- 16. NCR W3-6613 (a) NRC CONCERNS
- 1. What actions were done to assure that no additional heat numbers were falsified?
- 2. Identity of the persvn who forged the signature and entered l the incorrect heat numbers on tha Oun!.ity Records.
RESOLUTION OR CORRECTIVE ACTION
- 1. A review of all installed process tubing records back. to their applicabic CMR was performed by Ebasco QAIRG.
- 2. Identity of person is unknown and cannot be ascertained since contractor is no longer on site.
(b) LP&L IDENTIFIED CONCEFMS
! 1. Ev.'dence that the " untraceable" material was returned to the warehouse at scrapped.
- 2. Evidence that a search for adaitional falsified records war performed with regard to ti.a Mercury program.
RESOLUTION OR CORRECTIVE ACTION
- 1. Warehouse records were researched no evidence of return was a
found. . .
- 2. QAIRC and LP&L turnover review found no other cases of
- falsification.
l
- 17. NCR W3-6723 l (a) NRC CONCERNS l
i F6M procedure QC-309 violated ANSI N45.2 Section 13, because t' did not require the tension tester serial number, pressure gage numbe r or calibration date to be recorded.
RESOLUTION OR CORRECTIVE ACTION During the time frame involved there were only two (2) pressure gauges / tension teaters that were utilized sitewide, QC 4.2.1 & QC 4.2.2. These gauges were maintal' icd under Ebasco's M&TE procedure WQC-4. Copies of the calibration records are attached to NCR-W3-7184.
e 6-17
~~-- --
T-ATTACHMENT 1,
- 18. NCR W3-6786 i (a) NRC C0h LRN
- 1. Possible heat numbers not recorded on the as-built drawings.
- 2. NCR did not address where the required heat numbers were
.rocorded.
- 3. NCR did not address how traceability was maintained.
RESOLUTION OR CORRECTIVE ACTION
- 1. NCR-W3-4593 was reopened and addressed the following:
- a. Verified that any tubi g purchased non-safety was not used it a safety application or was repinced.
- b. Site procedure required material purchased non-safety to be identified (i.e. painting. marking, etc.)
- c. NCR-W3-4593 S/1 was referenced in all Mercury P2 and P3 OCR packages where direct traceability is not documented.
- d. A itat of manufacturers and heat numbers for tubing is attached to NCR-W3-6786 and 4593.
(b) LP&L IDENTIFIED CONCERNS
- 1. Heat numbers not posted to "As-Built" drawings.
- 2. NCR did not adeque ely address if the "PAB" (Preliminary As Cailt) Program. ,
- 3. The NCR did not determine if all possible heat numbers were traceable to the safety /non-safety installati ons and/or to the applicable P.O.
RECOLUTION OR CORRECTIVE ACTION 1,2&3 NCR-W3-4593 was re-opened, tedispositioned and addressed the concerns as stated above for NCR-W3-6786. NCR-W3-4593 S/1 nith attachments addressing heat numbers added to NCR-W3-6786.
- 19. NCR W3-70,99 (a) NRC CONCERNS
- 1. No documentation to adequately support the NCR Disposition.
RES01.l' TION OR CORRECTIVE ACTION
- 1. Stress eniculations utilized as a basis for disposition have been attached to thu NCR.
6-18
ATTACHMENT'l l-
- 19. NCR W3-7099 (Contirued)
(b) LP&L IDENTIFIED CONCERNS
- 1. Cracks in heat af fected zone of :abinecs 48A & B. '
- 2. Smallur than design embed plates.
- 3. Flare bevel in lieu of fillet welds.
i.
RESOLUTION OR CORRECTIVE ACTION f 1. Cracks 2 valuated and accepted by ESSE based on low stress.
- 2. Embed plates are the correct size; cabinet 48A requires a split 4"x4"x3/8 tube steel (which leaves 3" wide exposure) and ' calunet 488 required a 4" vide plate.
l
- 3. Flare bevels, fillets and lengths accepted by ESSE based on design calculations indicating lott stresses in weld.
- 20. NCR W3-7139 (a) NRC CONCERNS
- QC data in NCR was incorrect for 2 of 3 radiation monitors.
RESOLUTION OR CORRECTIVE ACTION l
l NCR re-opened and letter of clarification and inspection report
- l added to NCR. ,
(b) LP&L IDENTIFIED CONCERNS F&M Inspection Report #303-71-u24 contains only sheet i of 3 and does not include a list of the discrepant supports.
RES0!UTION OR CORRECTIVE ACTION '
Shoots 2 and 3 of Inspection Report added. ;
l !
l 21. NCR W3-7140 (a) NRC CONCERNS l None were identified in the allegations associated with this i issue in Supplement 7 to the Safety Evaluation Report (SSER).
l
- f
. 1 6-19
ATTACHMENT 1
- 21. NCR W3-7140 (Continued)
(b) y;L IDENTTFIED CONCERUS
- 1. Traceability of rework materials not recorded.
RESOLUTION OR CORRECTIVE ACTION
- 1. Rework consisted of additional welding only, filler netal requisition form enclosed in docur.intation cf NCR.
- 22. E R W3-7177 (a) NRC COMERNS
- 1. No calibration of pressure gauge used on expansion anchor tension rester.
- 2. Requirement that three additional anchors be tested af ter failure of one not adhered to.
RESOLUTION OR CORRECTIVE ACTION ,
- 1. Inspectors si. nature attests that tension testing was performed per governing specification.
- 2. Subsequent retests were performed eith acceptable results.
l
- 23. NCR W3-7179 ,
(a) NRC CONCERN None were identified in the allegations associated with this ,
issue in Supplem nt 7 to the Safety Evaluation Report (SSER).
(b) LP&L IDENTIFIED CONCERNS l NCR is acceptable RESOLUTION OR CORRECTIVE ACTION None required.
6-20
t ATTACHMENT 1
- 24. NCR W3-7180 (a) NRC CONCEdiS F&M procedure QC-309 violated ANSI N45.2 Section 13, . because it did not requira the tension tester serial number, pressure gage l number or calibration date to be recorded.
RESOLUTION OR CORRECTIVE ACTION During the time frame involved there were only two (2) pressure gauges / tension testers that were utilized sitewide, QC 4.2.1 & QC 4.2.2. These gauges were maintaincd under .Ebasco's M&TE procedure WQC-4. Copies of the calibraticn records are attached to NCR-W3-7184.
- 25. NCR W3-7181 l
(a) NRC CONCERNS F&M procedure QC-300 violated ANSI N45.2 Section 13, because it did not require the tension tester serial number, pressure gage number or calibration date to be recorded.
RESOLUTION OR CORRECTIVF ACTION
, During the time frame involved there were only two (2) pressure gauges / tension testers that were utilized sitewide, QC 4.2.1 & QC-4.2.2. These gauges were naintained under Ebasco's M&TE precedure WQC-4. Copies of the calibration records are attached to NCR-W -7184.
- 26. NCR W3-7182 (a) NRC CONCERNS F&M procedure QC-309 violated ANSI N45.2 Section 13, because it did not require the tension tester serial number, pressure gage number or calibration date to be recorded.
RESOLifTION OR CORRECTIVE ACTION During the tino f rame involved there were only two (2) pressure gauges / tension testers that were utilized sitewide. QC 4.2.1 & QC l 4.2.2. These gaugas were maintained under Ebasco's M&TE i
procedure WQC-4. Copies of the calibration records are attached i to NCR-W3-7184.
6-21
p ,
ATTACEMENT 1
?l 7, ~
.~1 a
'(+ *~ !a f,.
- 27. NCR W3-7194
/l (a) NRECONCERNS y
F&M procedure QC-309, violated ANSI N45.2 Section 13, because it did not. require thi? tensien tenter serial , number, pressure gage A rumber vr,calibra, tion date to be recorded. '
s '/ fr s
RE60LUTION OR CORBtCTIVE ACTION ,
'I During the time frame involved there were only two (2', pressure
', 'i diuir,es/ tension testers that were 'utilis:edf sitewida . QC 4.2.1 & QC
- 4.2.2. These gauges wers maintaine.d under Ebasco's M&TE procedure WQt-4. Copies of the calibrstion records are attached ~
to NCR-W3-7184..' 4 n .. o~
- 28. NCR W3-7132 .
~ .s , .
(a) NRC CONCERNS , ,
- 1. Concrete preplEicement & post-placement documentation could
~ '
not be matched. -
< 2. No spac{fic references were used tur voiding the NCR.
'..3. QA Engineer approved the recommended disposition and then voided the NCR.
' RESOLUTION OR CORRECTIVE ACTION
- 1. NCR-W3-7431 R1 addressed curing violations. NCR-U?-7435 addressed the placement docimentation.
- 2. Late entry added to NCR-W3-7432 referencing NCR's W3 7431 R1
& W3-7435.
- 3. Not a procedural violation per ASP-llI-7 Rev. 5. The t'ecommended disposition was approved 11/23/83; NCR was voided 1/16/84. .
- 29. NCR W3-7533 (a) NRC C0iCERNS ._
Noac f<<dre iden tif i.'d in the allegations associated with this tanue in supplement 7 to the Safety Evaluation Report (SSER).
(b) LP4L TDENTIFIED. CONCERNS -
NCR is acceptable. , ,
/ /
RESOLUTION OR CORRECTIVE ACTION None required. ~
9 6-22
l l
. ATTACHMENT 1
- 30. NCR W3-7547
-(a) N,RC CONCERNS'
- l. Inproper engineering evaluation is demonstrated with an accept-as-is disposition based on an acceptable hydrostatic test.
. RESOLUTION OR CORRECTIVE ACTION The disposition was based on prior acceptance of fit-up and final weld inspection and that the pressure boundary had not been violated therefore no hydrostatic test is required.
(b) LP&L IDENTIFIED CONCERNS:
- 1. Is the fit-up of FW-5 acceptable?
RESOLUTION OR CORRECTIVE ACTION
- 1. Radiographic examination of FW-5 was performed and ' fit-tip gap engagement reqairements were met.
B. Ebasco NCR'S Identified in Supplement 7 to the SSER The following Ebasco NCR's were identified by the NRC in Supplement .7 to the Safety Evaluation Report published,0ctober 1, 1984. ~
- 1. NCR W3-3947
, a) NRC CONCERN l
l Fit-up inspection was by-passed and the support had been completely welded out with only the welder's identification number.
RESOLUTION OR CORRECTIVE ACTION Inspection revealed an acceptable heat number (15537) of k" angle and filler metal withdrawal authorization slip furnished for hanger. An additional visual inspection revealed an acceptable final weld.
6-23 ,
'q ATTACHMENT 1
.s - .v
[ _f s
[ '2 . . NCR W3-4593 _ l a,p7 a) NRC CONCERN ,
Disposition inadequate.
p RESOLUTION OR CORRECTIVE ACTION NCR was ~ re-opened as Supplement 1 (S/1) since original disposition of NCR-W3-4593 .had r.o t been correctly implemented.
.. Mercury's material control program was analyzed based on purchase of mt.terials , material identification and dimensional verification.
In. April, 1984, NCR-W3-4593. S/l was closed. Based on this analysis, .it can be shown. that. safety-related tubing of correct '
sf 2. and wall ' thickness was installed by Mercury. Therefore, he ing addressed. the requirements of a- material control program and identified and corrected deficiencies - noted, direct heat traceability is not required for Mercury tubing installation.
In addition,- NCR-U3-4593 S/l was referenced in all of Mercury's P2 and P3 OCR packages 'where direct traceability was not documented, and a; document was attac.hed, which provided a list of manufacturers of tubing, and heat numbers furnished.
- 3. NCR W3-5819 .
a) NRC CONCERN Identified the problem of instrumentation supports being painted l
prior to final weld visual inspection. Disposition had been to inspect the welds through paint which was unacceptable. l.
RESOLUTION AND CORRECTIVE ACTION l'
NCR supplemented with ESSE evaluation " Reinspection of Welds.
through Paint for Sine and Profile" for additional justification, j
( 4. NCR W3-5973 , h I
a) NRC CONCERN None were identified in the Allegations associated with this issue in Supplement 7 to the Safety Evalurtion Report (SSER).
l -
e
(
6-24
~
L ATTACHMENT 1
- 4. NCR W3-5973 (Continued) b)' LP&L CONCERN NCR is acceptable.
RESOLUTION OR CORRECTIVE ACTION None required.
5 .' . NCR W3-5974 a) .NRC CONCERNS
-The NCR's ' disposition is questionable as the problem still existed in that safety and non-safety grade material could have been mixed.
RESOLUTION OR CORRECTIVE ACTION The attachments added to NCR as a result of corrective action were the back-up data used in verifying whether or not the material was safety related. Each Seismic I hanger / piping system component was verified by the QAIRG group as being safety related. Those items which were found to be non-safety were removed and safety material installed.
l
- 6. NCR W3-6514 a) NRC CONCERN Mercury installed supports without material traceability.
RESOLUTION OR CORRECTIVE ACTION Bergen Patterson designed supports, other than ASME NF supports, ,
do not require traceability. The structural members were ;
supplied by Bergen Patterson and were received with a certificate !
of compliance. '
b) LP&L CONCERN Attachment No. 6, Item 1 is not justification for closure of NCR.
RESOLUTION OR Og{f5CTIVE ACTION A late entry note added to Attachment No. 6 provided an expanded discussion on the use and acceptance of letter F-61147E. The i
. statement (Item 1) of Attachment No. 6, in conjunction with Items 2 and 3 of the Attachment, were the basis for closing this NCR.
6-25
4 ATTACHMENT 1 l
i
- 7. NCR W3-6719 a). NRC CONCERNS f
: The hydrostatic test conditions were assumed by Ebasco to be the
" worst case" and therefore that "all" other hydrostatic tests pa.rformed by Mercury were deemed satisfactory. This was not the case, since only'one test was reviewed by Ebasco.
RESOLUTION OR CORRECTIVE ACTION Attachment No. 17 written by ESSE: clarifying -justification of selection of worst case ~ condition and ' . providing - support calculations.
6 8 9 i
s j' (
6-26
ATTACHMENT 2 l MERCURY NONCONFORMANCE REPORTS IDENTIFIED BY THE NRC The following is a list of Mercury Nonconfcrmar.::e Reports (NCR's) identified by the NES in Issue No. 6 and in Supplement 7 to the Safety Evaluation Report (SSER). The list identifies the NRC concerns with each NCR and 'the Resolution or Corrective- Action. The list also summarizes any additional concerns identified as a result of the LP&L Review and the Resolution or Corrective Actioa. It should be noted that -dispositioned NCR's were reviewed for
- reportability under 10CFR50.55(e) and 10CFR21 and -none were found to be reportable. l A. Mercury NCR's Identified in Issue No. 6
- 1. NCR-180 (Ebasco NCR W3-6839)
(a) NRC CONCERNS J
None were identified in the allegations associated with - this -
issue in Supplement 7 to the Safety Evaluation Report (SSER).
(b) LP&L IDENTIFIED CONCERNS
. '1. No objective evidence provided for "as-built" condition of t the discrepant Hilti's for the Engineering Evaluation.
RESOLUTION OR CORRECTIVE ACTION
- 1. Testing was performed on bolts with an embedment of 3" where field installation procedures required 3 ". Results of re-inspection of Hilti bolts under records review and N1 instrument walkdowns have found the as-built conditions to be generally acceptable. Any Hilti bolts without letter designation were ultrasonicly tested for length to determine proper embedment.
- 2. NCR-255 (a) NRC CONCERN None were identified in the allegations associated with this issue in Supplement 7 to the Safety Evaluation Report (SSER).
(b) LP&L IDENTIFIED CONCERNS The documentation of the corrective action was not availabl.e for eight of the fourteen supports requiring retorque.
RESOLUTION OR CORRECTIVE ACTION The supports identified as having misplaced documentation wer; reinspected. This action has been completed with acceptable results and attached within the N.C.R. package.
6-27
e ATTACHMENT 2-
- 3. Mercury NCR-268_
(a)' NRC CONCERN None were identified. in the allegations = associated with this issue in Supplement 7 to the Safety Evaluation Report (SSER).
.(b)- LP&L IDENTIFIED CONCERNS
- 1. This NCR is not a rework as stated, it is a "use-as-is" since as-built information is to be ridlined.
- 2. Should have been up-graded to an Ebasco NCR.
- 3. No objective evidence Ebasco Engineering has . approved the as-built conditions.
- 4. All deficiencies ~ identified in the description are not
. addressed in the disposition completed section of the NCR.
- 5. - There is not obj ective evidence to indicate that all existing field conditions have been incorporated into the redline drawfrg.
, 6. NCR - was - written - 1/26/82 and closed 12/22/82. Training records . supplied for ' corrective action' are dated 'l/29/82 (due to updated revision of five procedures released this date) and 6/17/84 (due to Ebasco audit) there is to evidence of timely retraining of personnel per disposition of NCR.
RESOLUTION OR CORRECTIVE ACTION
- 1. Tite NCR represents a procedural violation for failure to redline the drawing prior to the installation of the supports. -There was no physical rework due to the actual installation being acceptable. This NCR was written as an in-process deficiency due to the inspector's findings during walkdown inspection.
- 2. The NCR was not used to accept a deviation from design requirements, thus, did not require upgrading to an Ebasco NCR.
- 3. As-built conditions were in accordance with Ebasco guidelines provided to Mercury in the specifications and drawings.
- 4. The deficiencies identified were addressed by redlining the drawing and requiring the training to address the procedural violation.
- 5. Copy of the drawing is attached.
- 6. No specific training records could be located for this NCR.
However, as a result of SCD #57, all Mercury personnel were retrained. This training addressed redlining.
l l
a 6-28
ATTACHMENT 2
- 4. NCR-363 (a) NRC COSCERN An Authorized Nuclear Inspector (MiI) -review was not perfo rmed for installation of strongback support-lugs to ASME process pipe.
RESOLUTION OR CORRECTIVE ACTION ASME process pipe is class .3 and doec not require ANI review.
(b) LP&L IDENTIFIED CONCERNS
- 1. Mercury NCR should have been upgraded to an Ebasco NCR.
- 2. Mercury Proj ect . Engineer did not verify similar
. installations for like condition.
RESOLUTION OR CORRECTIVE ACTION
- 1. ESSE approved the existing condition by issuance of an DCN.
- 2. Ebasco QA reviewed similar installations and the review results were placed with the Mercury NCR File.
- 5. NCR-380 (Ebasco NCR-W3-4015)
(a) NRC CONCERNS None were' identified in the allegations associated with this issue in Supplement 7 to the Safety Evaluation Report (SSER).
(b) LP&L IDENTIFIED CONCERNS
- 1. Three sets of weld data records for support 604-70 are attached to the NCR. Unable to determine which record is being used as a basis for acceptability.
- 2. Mercury documentation cannot be found for welding performed by welder M-229.
RESOLUTION OR CORRECTIVE ACTION
- 1. NCR-W3-4015 was revised to NCR-W3-4015 R-1 for clarification of this discrepancy.
- 2. Research by Ebasco revealed that welder M-229 was qualified to perform the welding on the anchor plates.
- 6. NCR-420 (Ebasco NCR W3-4137)
See Ebasco NCR W3-4137 - (Attachment 1, Item A.5) 6-29
E l
l I
ATTACHMENT 2
- 7. ~NCR-429 (Ebasco NCR W3-3965)
(a) NCR CONCERNS Nont were identified in the allegations associated with this issue-in Supplement 7 to the Safety Evaluation Report (SSER),
(b) LP&L IDENTIFIED CONCERNS NCR is acceptable.
RESOLUTION OR CORRECTIVE ACTION None required.
- 8. NCR-438 (Ebasco NCR W3-4013)
(a) NRC CONCERN l None were identified in tue allegations associated with this issue in Supplement 7 to the Safety Evaluation Report (SSER).
(b) LP&L IDENTIFIED CONCERNS
- 1. The disposition did not address the action taken to preclude the,use of the~angl'e iron that was removed from the Mercury support. ,
RESOLUTION OR CORRECTIVE ACTION The piece of angle was removed from the Mercury support, thereby resolving the nonconforming condition. Maintaining traceability of non-safety material (angle) is not required.
- 9. NCR-487 (Ebasco NCR W3-4044)
(a) NRC CONCERNS None were identified in the allegations associated with this issue in Supplement 7 to the Safety Evaluation Report (SSER).
(b) LP&L IDENTIFIED CONCERNS
- 1. Item No. 15 - Attachment #3 - Evaluation does not provide ,
evidence that drawing has been redlined to reflect field I conditions. Calculations should also be attached to verify additional loads for the attachment steel.
- 2. Per field verification, tubing for pressure indicator PI-SI-7140 has reverse slope and loose clamp.
+
6-30
ATTACHMENT 2
- 9. NCR-487 (Ebasco NCR W3-4044) (Continued)
RESOLUTION OR CORRECTIVE ACTICN
- 1. The referenced item conforms to the hanger detail, therefore, Mercury drawing 160-T-033A does not require redlinin2 Calculations for the attachment steel have been attached co the NCR.
. 2. Additional engineering evaluation has been added to address the reverse slope and the loose clamp has been corrected.
- 10. NCR-491 (Ebasco NCR W3-4088)
See Ebasco'NCR W3-4088 - (Attachment 1, Item A.4)
- 11. MERCURY NCR-528 -(Ebasco NCR W3-4824)
(a)' NRC CONCERNS None ' vere identified in the allegations associnted with this issue in' Supplement 7 to the Safety Evaluation Report (SSER).- -
(b) LP&L IDENTIFIED CONCERNS
- 1. No statement or documentation was attached to the NCR to '
resolve traceability of heat #M2245. ,
- 2. Disposition of NCR fails to state whether the correct ID#
was etched on the plate.
- 3. No documentation was attached to the NCR to verify corrective action taken.
RESOLUTION OR CORRECTIVE ACTION 1 & 3. Attached a copy of MRR-77-11206 to NCR, indicating heat code MZ-245 (M2245), and associated supplier C of C.
- 2. Field verified heat number 7428779 on anchor plate.
- 12. NCR-540 (a) NRC CONCERNS None were identified in the allegations associated with this issue in Supplement 7 to the Safety Evaluatien Report (SSER).
6-31
~
E ATTACHMENT'2
- 12. NCR-540 (Continued)
(b) LP&L IDENTIFIED CONCERNS -
- 1. ' Documentation not attached to.NCR for replacement of support locator #31.
.2. Documentation not attached to NCR for replacement of tubing that had cold spring.
RESOLUTION OR CORRECTIVE ACTION
- 1. Mercury documentation was attached to NCR for replacement of support locator 631 with an~ acceptable support locator.#33.
- 2. Mercury documentation was attached to NCR for replacement of tubing with cold spring.
- 13. NCR-554 (a) NRC CONCERNS No documented evidence of corrective action for hanger deficiencies identified during walkdown.
_ RESOLUTION OR CORRECTIVE ACTION Documentation search and re-inspection established rework was accomplished. ,
(b) LP&L IDENTIFIED CONCERNS
- 1. No welding documentation for repair of supports.
- 2. No inspection documentation for repair of supports.
- 3. Inadequate documentation of corrective action to correct elongated holes in tube track.
RESOLUTION OR CORRECTIVE ACTION 1 & 2. Documentation search and reinspection established rework was a:complished.
1 Reinspection established rework was accomplished.
- 14. NCR-560 (Ebasco NCR W3-5428)
(a) NRC CONCERNS None identified in the allegation associated with this issue in Supplcnent 7 to the Safety Evaluation Report (SSER).
I.
6-32
ATTACHMENT 2
- 14. NCR-360 (Ebasco NCR W3-5428) (Continued)
(b) LP&L IDENTIFIED CONCERNS
'1. The NCR was = closed without the ' ~ appropriate documentation being attached to verify revision of drawing #163-L-003A and E Support Inspection Reports.
RESOLUTION OR CORRECTIVE ACTI0h;
- 1. A review of drawing #163-L-003A' revealed the required' revision to reflect locators 3, 4, and 5 to be 000-H-150-N.
A copy of the drawing has been attached.
- 2. Copies of the. Siipport . Inspectirn Reports :'for each support .
locator 3, 4, anu 5 have been cttached.
- 3. CIWA 011645 was issn *.d for reverification of the torque on Hilti bolts for supports 3 and 4.
- 15. NCR-565 (Ebasco NCR W3-4730)
(a) NRL CONCERNS None identified in the allegation associated with this issue in Supplement 7 to the Safety Evaluation Report (SSER).
(b) LP&L IDENTIFIED CONCERNS ,
, The review of Mercury NCR-3243 which was issued to resolve items
- 1 and 2 of NCR-565 fails to provide adequate documentation to determine resolution.
RESOLUTION OR CORRECTIVE ACTION The required documentation has been obtained irom Mercury files and added to the NCR to resolve comments.
-16. NCR-568 (Ebasco NCR-W3-4730)
(a) NRC CONCERNS No documentation was attached to the NCR as objective evidence for corrective action taken.
(b) LP&L IDENTIFIED CONCERNS t
The disposition of items #2, 3, 4, and 5 fail to provide adequate
[ . engineering basis for a~ccept-as-is.
ll ~
6-33
-AT*ACHMENT'2~
- 16. -NCR-568 (Ebasco NCR-W3-4730) (Continued)
- RESOLUTION OR CORRECTIVE ACTION ,
l
-Items #2", 3, 4, and 5 were inspected for compliance to FCR-IC-579 (basis ' for accept-as-is of elongated holes) . Items 3, 4, and 5 -
were acceptable. Item 2 was acceptab_e . af ter evaluation by Design Engineering.
- 17. NCR-591 (Ebasco NCR-W3-4206)
(a) ,N2C CONCERNS None were identified in the allegations associated, with this issue in Sunclement 7 to the Safety Eva3uation Report (SSER).
(b) LP&L IDENTIFIED CONCERNS
- 1. The analysis conducted for' this NCR was not attached, including ESSE concurrence.
RESOLUTIONS OR CORRECTIVE ACTION
- 1. Calculations were performed by ESSE to substantiate analysis described in NCR. Analysis was attached to the NCR.
- 18. NCR-594 (Ebasco NCR-W3-5557)
(a) NFC CONCERNS t
None were identified in the allegations associated with this issue in Supplement 7 to the Safety Evaluation Report (SSER).
(b) LP&L IDENTIFIED CONCERNS Ne documentation that drawing has been redlined.
RESOLUTION OR CORRECTIVE ACTION Support in question is a typical detail and therefore not red-lined. Deviation is referenced appropriately in OCR package.
- 19. NCR-595 (Ebasco NCR-W3-4197)
(a) NRC CONCERNS i
None were identified in the allegations associated with this issue in Supplement 7 to the Safety Evaluation Report (SSER).
6-34
ATTACHMENT 2
- 19. NCR-595 (Ebasco NCR-W3-4197) _ (Continued)
(b) LP&L IDENTIFIED 'CCECERNS
- 1. Several supports installed -which are not per an approved installation detail.
RESOLUTION OR CORRECTIVE ACTION
- 1. Description of NCR incorrectly written as Locator . "5" was -
actually installed as Locator "23".
- 2. The anchor plate installation for Locator "23" is acceptable per the general ' notes section of the B-430 series detail drawings.
- 3. Attachments 'to 'NCR were - mcde to clarify installation details.
- 20. NCR-614 (Ebasco NCR W3-4219)
See Ebasco NCR W3-4219 - (Attachment 1, Item A.6)
- 21. NCR-625 (Ebasco NCR-W3-5282)
(a) NRC CONCERNS None were identified in the allegations associated with - this .
issue in Supplement 7 to the Safety Evaluation Report (SSER).
(b) LP&L IDENTIFIED CONCERNS
- 1. One weld sign-off for two welds.
- 2. Reason for voiding installation and location information.
RESOLUTION OR CORRECTIVE ACTION
- 1. Inspection reports identify welder of both joints.
- 2. Information voided due to redline #6.
- 22. NCR-656 (Ebasco NCR-W3-4303)
(a) NRC CONCERNS None were identified in the allegations associated with this
- issue in Supplement 7 to the Satety Fraluation Report (SSER).
1 i
l 6-35
l l
ATTACHMENT 2 l
- 22. NCR-656 (Ebasco NCR-W3-4303) (Continued)
, (b) .LP&L IDENTIFIED CONCERNS
- 1. Process tubing. supports installed without- approved installation details.
RESOLUTION OR CORRECTIVE ACTION
- 1. Design Er31 neering . reevaluate.d to accept-as-is per tatation on installation detail of supports.
- 2. 'The current _as-built condition was reverified by _ Ebasco QA Surveillance Engineering.
- 23. MERCURY NCR-658 (a) NRC CONCERNS No documentation was attached to the NCR as obj ective evidence for corrective action taken.
RESOLUTION OR CORRECTIVE ACTION
- 1. A field verification by EBASCO revealed that corrective action ner the NCR's disposition had been correctly performed.
- 2. Found and attached to the NCR, a Mercury anchor inspection report for retorquing of Hilti bolts.
(b) LP&L IDENTIFIED CONCERNS No documentation was attached to the NCR as obj ective evidence for corrective action taken.
p3LUTIO OR CORRECTIVE ACTION
- 1. Ebasco ficid verification revealed that corrective action per the NCR's recommended disposition had been properly
. performed (see Ebasco General Inspection report SW-913).
- 2. Found and attached to the NCR, a Mercury anchor inspection report for retorquing of Hilti bolts.
? -
I' l
i
! 6-36
3 ATTACHMENT 2
+ ,
i B. MERCURY NCR's IN SUPPLEMENT 7 TO THE SSER !
I Th'e following Mercury NCR' 3. were identified by the NRC in Supplement 7 to the Safety Evaluation . Report (SSER) published October - 1, 1984. Mercury ,
NCRs 888 and 889 were deterrined to have been administratively closed and accordingly are addressed in the response to Issue 13. ;
i.
- 1. NCR-313 !
i
.(a) NRC CONCERNS Identified seven h inch stainless steel lines for P2 instruments that were damaged by weld spatter. The-NCR stated that the lines wets replaced. and documented as such in operational control record (OCR) 945 and OCR 1020, but it could not be ascertained from these rework packages that the repair and reirspection was either started or completed. There was no documentation with these NCR's to prove that corrective action was completed.
(b) LP&L IDENTIFIED CONCEPES The documentation of the corrective action was not included in the Mercury NCR package.
RESOLUTION'OR CORRECTIVE ACTION
- 1. Documentation was copied from the referenced OCR packages, reviewed and added to,the NCR package.
- 2. A reinspection was performed by Ebasco QC Inspector and the satisfactory QC Ir.spection Report was added to the NCR ;
package. !
- 2. NCR-322 (a) NRC CONCERNS j Identified seven h inch stainless steel lines for P2 instruments i that were damaged by weld spatter. The NCR stated that the lines were replaced and documented as such in operational control record (OCR) 995 and OCR 1020, but it could not be ascertained from these rework packages that the repair se.d reinspection was either started or completed. There was no documentation with -
these NCR's te proved that corrective action was completed, f I
(b) LP&L IDENTIFIED CONCERNS g
. The NCR package was lacking documentation to support closure of '
the NCR. k 6-37
ATTACHMENT 2 l
- 2. NCR-322 (Continued) }
RESOLUTION OR CCRRECTIVE ACTION
- 1. Documentation was retrieved from the referenced OCR package <
and added to the NCR package. ,
- 2. A reinspection was performed by Ebasco QC Inspector and the ,
satisfactory QC Ir.spection . Report was added to the ' NCR package.
- 3. NCR-337 (a) NRC CONCERNS Identified 'seven is inch stainless steel lines for P2 instruments that were damaged by weld spatter. The NCR stated that the lines were replaced and documented as such in operational control record (0CR) 995 and OCR 10.!0, ' but it could not be ascertained from these rework packages that the repair and reinspection _was either started or completed. There was no documentation with these NCR's to proved that corrective c.ction was completed.
(b) LP&L IDENTIFIED CONCERNS The NCR package was lacking documentation to support - closure of the NCR.
RESOLUTION OR CORRECTIVE ACTION
- 1. The referenced 0C3 package was researched and records needed to support closure of the NCR were reviewed and found to be acceptable.
- 2. An inspection was performed by Ebasco QC Inspector with satisfactory results. QC Inspe tion Report was .tdded to the NCR package.
- 4. _NCR-572 (a) NRC CONCERNS No.ed that the weld on support lccator #26 was undersized. The NCR stated that the weld was reworked and weld metal added to !
bring weld to sufficient size. There was no reference as to what OCR was issued to perform this rework or traceability of weld metal used in the performance of this job. Also, there were no inspection reports identified o.r contained in the package.
6-38
ATTACHMENT 2 1
- 4. NCR-572 (Continued) ,
'I
! l RESOLUTION OR CORRECTIVE .* TION ~j i
- 1. Support No. 26 was.' redesignated as support No. 1714-33_by Redline No. 6 of Drawing No. 163-T-013-A.
- 2. A copy of documentation for weld bufld up us located and ,
placed in file.
'5. NCR-673 (a) NRC CONCERNS Identified probler.s - with instrument tubing installed by OCR
- 723. ,
(b) ~ LP&L IDENTIFIED CCNCERNS The lines identified by Mercury NCR-673 were identified as P7N3 class lines and are covered by the requirements of ANSI B31.1.
The corrective action was to be tracked and resolved by Mercury Co. Engineering Department. Documentation was not in NCR folder !
j to show that the problem was tracked and resolved. j RESOLUTION CD. CORRECTIVE ACTION .
- 1. Ebasco re-inspected these lines on 8/2/84 and found that the discrepancies noted in this NCR had been corrected, and the condition no longer existed. ,
i
- 2. Copies of documentation to verify the re-inspection . were placed in the NCR folder.
j i
- 6. NCR-674 l (a) NRC CONCERNS Identified problems with the electromagnetic control panel worked by OCR #1246.
(b) LP&L IDENTIFIED CONCERNS Documentation was missing from NCR folder to support disposition and closure of NCR.
6-39
A,TTACHMENT 2
,>u.
- 6. NCR-674 (Continued)
RESOLUTION OR CORRECTIVE ACTION .
- i. Ebasco rainspected the supports and tubing addressed on this ;
NCR, and ESSE accepted the installation as-is. i 1
- 2. . Copies of _ the inspection and evaluation were placed - in the !
NCR foldar for. support documentation to justify disposition !,
and closure of this NCR.
- 7. NCR-675 (a) NRC CONCERNS Identified problems with instrument tubing installed by ' OCR
- 720.
(b) LP&L IDENTIFIED CONCERNS Documentation was not in NCR folder to capport disposition and ,
closure of the NCR.
RESOLUTION OR CORRECTIVE ACTION v
- 1. Documentation was located to show that Ebasco performed an '
inspection and copy of the inspection report was placed in' the NCR folder.
- 8. NCR-676 (a) NRC CONCERNS Identified problems with instrument tubing inscalled by OCR
- 720.
1 (b) LP&L IDENTIFIED CONCERNS Documentation was not in the NCR folder to justify closure of 4
this NCR.
RESOLUTION OR CORRECTIVE ACTION
- 1. Ebasco mspected the tubing and found that the minor 1 ow 4 woulo not affect the applicable pressure switch. ESSE [
concurred and accepted the installation as-is.
- 2. Copies of the evaluation have been placed in the NCR folder to support closure of the NCR. g 6-40
ATTACHMENT 2
- 9. 'NCRi 677 (a) NCR CONCERNS ,
i
' Identified problems .with instrument tubing installed by OCR {
- 1332.
(b) LP&L' IDENTIFIED CONCERNS Documentation not' available in NCR folder to support disposition and closure of this NCR.
RESOLUTION OR CORRECTIVE ACTION ;
- 1. Ebasco re-inspected the tubing addressed by this NCR and ESSE accepted the installation as-is.
- 2. Copies of'the inspection and evaluation have been placed in the NCR folder to support disposition and closure of ' this NCR.
- 10. NCR-678 (a) NRC CONCERNS
>=
Identified problems with instrument tubing installed' by OCR
, #723. !
(b) LP&L IDENTIFIED CONCERNS l Sufficient documentation not in NCR folder to support disposition l and closure of NCR. !
RESOLUTION OR CORRECTIVE ACTION
- 1. Ebasco re-inspected the tubing addressed in this NCR, and j the results were evaluated by ESSE to use-as-is.
- 2. Copics of the inspection and evaluation have been placed in the NCR folder to support disposition and closure of this NCR.
'll. NCR-806 (Ebasco NCR W3-7547)
(a) NRC CONCERNS Ebasco NCR W3-7547 noted discrepancies against OCR#1830 and Mercury NCR-806. The disposition of this NCR is unsatisfactory due to the system passing a hydrostatic test is used as the basio !
for accountability of fit-up discrepancy. l 6-41
4 ATTACHMENT 2
- 11. NCR-806 (Ebasco.NCR W3-7547) (Continued)~
. RESOLUTION OR CORRECTIVE ACTION i l
See Attachment 1, Item A.30 (Ebasco NCR W3-7547). ,
.- 12. 'NCR-2234 (Ebasco NCR W3-4593) ,
3-
.(a)' NRC CONCERNS' Stated that no heat numbers could be verified between FW13. and I FW13R. .This is for :0CR#666, System 52B.. The recommended disposition was per Attachment #4 of NCR W3-4593.
(b) LP&L IDENTIFIED CONCERNS Documentation not available is NCR folder to support disposition and closure of NCR.
RESOLUTION OR CORRLCTIVE ACTION
- 1. Cepies of the referenced attachment of Ebasco NCR W3-4593 were placed in this NCR package.
2.
Documentation necessary to support ef'osure of this NCR was added to the package as supplemental information.
- 13. NCR-3149 (a) NRC CONCERNS r Indicated that there was no documented indications that welder M-343 was qualified to welding procedure specification D (WPS-D).
Disposition of this problem was by use of ' a weld test coupon subsequently found on April 27, 1983, but no longer available.
No documentation existed on the qualification of this welder or on his retest. Thus, all velds made by this welder were suspect.
(b) LP&L IDENTIFIED CONCERNS Documentation was not available in NCR folder to. rapport justification and closure of this NCR.
RESOLUTION OR CORRECTIVE ACTION The welder's (M-343) certification records were located and placed in the NCR folder.
i 4
6-42
RESP 0t'1E ITEM.NO:' '7-(Revision 1)- 'i TITLE: BACKFILL SOIL DENSITIES NRC DESCRIPTION OF CONCERN:
'The staff found that records are missing for the in-place density test of
. backfill in Area 5 (first- 5' starting at Elevation -41.25'). These. documents I- are important because the seismic response of the plant is a function of the soil-densities.
LP&L shall'(1) Conduct a review of all soil packages for completeness and
- , technical adequacy and locate all records and provide closure _on technica]
4 questions, or (2) conduct a review of all soil packages for completeness and technical adequacy and where-soil volumes cannot be verified by records as meeting criteria, perform and document actual soil conditions by utilizing.
penetration tests or other methods, or (3) Justify by analysis that the soil j volumes with missing records, or technical problems as defined af ter the 4
records review, are not critical in the structural capability of~the plant under seismic loads.
DISCUSSION: ,
- LP&L has reviewed all soils packages for completeness and technical adequacy, has located the items found missing by the staff, has identified those soil volumes for which complate records were not found, and has justified by analysis i that the structural capability of the plant under seismic loads is assured. . A
}' detailed engineering report has b6an prepared and attached to this response i describing the review and analysis of the soil backfill densities, which j' reconfirms the adaquacy of the backfill. This was also repescedly demonstrated j in the seven (7) statistical studies of backfill densities performed during the ,
j construction period, which showed good control of the work was achieved and
- specification requirements generally exceeded.
The following discussion is a summary of the findings of the attached report.
- The design criterion for the backfill was to obtain a liquefaction free material-
- at 75% relative density. To confirm compliance with this design criterion, a
! detailed three stage program was implemented to perform a review for ;
l completeness and analysis of backfill soil density and inspect 10n reports for j technical adequacy which verifies the structural capability of the plant under
- seismic loading conditions.
i h The program effort was conducted under the direction of the Ebasco Site Soils l Engineer who was present during t'ae performance of the majority of the actual l backfilling operations. Two basic sets of evaluations were performed, the i firsc on soil backfill test records, and the second on the corresponding l inspection Reports.
l During the Stage I effort, a detailed search was made of all locations l containing soil backfill data. Additionr.1 test records and inspection reports
' were obtained from contractor and laboratory files and also Engineering, Laboratory and Quality control indices and tabulations were retrieved.
i 7-1
i Once the packages of soil data were located and collected, Stage II activities concentrated on a review of the documents for completeness and a compilation of the data into a format amenable to . review of the NRC concerns.
Included in the review were each type of Inspection Report and each type of te6t record in the soil packages. It_was determined that the complete set of test records and a nearly complete set of inspection reports had been located.
In direct response to the first paragraph of the Description of the NRC Concern, the data for the 34 in-place density tests performed in the first 5.5' of Class A fill placed in Fill Area #5 from Elevation -41.75 to EL -36.25, has been located.
Stage III activities consisted of engineering evaluation of the data gathered and organized in Stages I and II. The results of the Stage II and III evaluations for completeness and technical adequacy for both the test records and inspection reports are summarized at follows:
(A) ffALUATION OF TEST RECORDS Test records deal with quantitative attributes of the fill such as density, moisture content and gradation. The test most indicative of quality is density, since it. relates directly to liquefaction entential, hesever, the other attributes were also revfcwed for acceptability.
Utilizing the complete package of final backfill test records, totalling approximately 3100 tests, overlay plots of relative density were constructed at each one foot interval of elevation and laboratory test data were tabulated during the Stage II effort. These documents represent a ;
graphical plot of density test frequency and distribution, and tabulate and display the final insitu relative densities.
The Stage III review and evaluation of the technical adequacy of the Class A backfill to provide structural stability of the plant under seismic loadings was based upon a comparison of the design requirements as stated
- in the Ebasco Specification LOU-1564.482 with existing documcatation and with the relative density plots prepared in this review. These plots are available in the Site Quality Assurance Records Vault. These plots demonstrate satisfaction of requirements for test frequency and distribution throughout the fill volume.
The evaluation included each type of test record required by the governing specifications and procedures and analyzed:
The completeness of all test records i
The testing frequency and distribution of in place density tests The frequency of laboratory control tests The performance of statistical studies The Class A Backfill relative density l
l l
i 7-2
- - - __. . . _- -. _ _ _ _ __ . ~ _ _
. l
--l l
4 The results of these analyses are as follows:-
(1) The Class A br.ekfill soil testing records are complete.
1 i (2) Field density and laboratory density and gradation tests wre l
{.
generally performed in accordance with the specified frequancies. l In lees than 8% of the cass:. reviewed, the laboratory control '
F tests were run-at intervals slightly larger than the specified (one control set per ten field density tests) criteria. The ,
backfill placed during these periods was randomly located 'l throughout.the fills and the relative densities obtained during these intervals were found to be in compliance with the specification requirements. This variance was therefore evaluati.d to be acceptable.
(3) Field tests were located in accordance with the specified random j distribution. In less than'5% of the tests reviewed, the location coordinates of the inplace deasity tests were found to-be in error. These tests were still a valid indicator of the relative density of the backfill at a random spot at a known
- elevation in a kn un fill area and were therefore deemed to be j acceptable tests.
(4) Statistical studies of relative density were performed in accordance with the specification requirements.
(5) The Class A backfill soil densities are in accordance with the specification requirements and will provide the required design structural capability to the plant under seisnic loads.
]
l (B) EVALUATION OF INSPECTION REPORTS I
j Inspection records generally deal with qualitative attributes of the fill a such as proper preparation of the fill surface and cleanliness of fill
) received. PreJaction-related quantitative information such as fill 4
location, elevation and area are also p:ovided.
} During the Stage II review activity, the total file of inspection reports for Class A backfill was inventoried and combined into conpatible soil pacxages. Included in the inventory were approximately 12,000 inspection
- reports ranging from EL -44 to EL+20 throughout all seven fill areas. The
{ reports were grouped and compiled by fill location, elevation and placement i date for each of the five. types of inspection forms and susmarized in j several tabulations.
The evaluation of these inspection reports was divided into two phases:
, the evaluation of the inspection reports to determine their overall l completeness, and the evaluation of the frequency and distribution of
- inspection reports to determine their content.
l Two comparative analyses were performed to determine the relative
, completeress of the inspection documentation. The first analysis performed
{ was a comparison of the qu.tntity of inspection packages to testing packages throughout the fills, while the second compared the documented surface area of inspection to the total surface areas of the fill placement.
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Once completeness of inspections was established, an additional analysis was performed to define the magnitude, the d*stribution and significance of l
.the documentation.found to be missing. -This analysis evaluated the I distributten of each type of: inspection report by fill location and elevation, and determined types of missirg documentation and the amounts.of backfill by volume affected. The results of this analyals-are as follows:
(1) The distribution of the existing. inspection documentation throughout the backfill is essentially identical to the distribution of the field L testing effort in that where inspection reports are found for s 'given i- fill area and elevatiouc a density test report is also found, thus
, indicatira a one to one relationship between inspection and testing l activitias._ This is an expected trend since the inspection activity included ordering tests performed. It is therefore concluded that the. -
- j. inspection activity took place whenever tests are found and that . ,
- missing inspection reports are not indicative of lack of inspection activity, e
1 (2) Eighty percent of the volume of the backfill has a sufficient quantity of each type of inspection report to fulfill the requirements of the 3
. specification and inspection procedures.
(3) For the 20% of the volume of the backfill which was missing some of j the required inspection reports, 16% hat an average of.81% of the i
reports required. 3.8% has one or more typa of inspection missing, and 0.2% :onsisting of six one foot lifts in four fills have no inspection. reports at all.
l For details, see the Report Section 4.B.' and Table No. 2.
i
! The effect on each of these types of deficiencies was evaluated based upon the j quantity and type of inspection documentation existing above, below and around
- the affected fill areas, the relative density results in the affected areas and i the relatively reall volume of fill affected. It was concluded that the
} deficiencies found in the inspection documentation are most probably due to lost folders, are not indicative of a lack of inspection effort, and will have no effect on the structural capability of the plant under seismic loeds.
l CAUSE:
The cause of this concern was the fact that some of the field inspection and laboratory test records for the Class A backfill were still in the contractor's j QA records vaults. This contractor is still active on site and had not initiated the transf er of documentation to the LP&L-Ebasco Quality Assurance Vault. All available soil records are now permeaently stored in this vault.
! GENERIC IMPLICATIONS:
f' Based upon the results of the detailed review and analysis of backfill soil-i densities and corresponding inspection reports described in the discussion above, the Class A backfill was found to be sufficiently in compliance with the
! specification requirements.
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The large effort required to establish the completecons of the records is due to the intrinsic difficulty of scoping a bulk process such as backfill in the absence or an administrative control tool, such as a logbook of inspections, which was not required by the implementing procedures. This scoping problem is believed to be unique to the soils / backfill effort.
Difficulty in establishing records completeness also was due to incomplete records turnover from the onsite coteractor involved. Therefore, a generic concern exists as to the extent to which there has been incomplete records turnover on the part of remaining site contractors. This is addressed in the CORRECTIVE ACTION PI41 below.
SAFETY SIGNIFIEWCE:
Test records and inspection reports were located and analyzed demonstrating compliance with the specification. Therefore, the Class A backfill will perform its function with respect to structural design capability under seismic losds.
. LP&L therefore believes that this issue is of no safety significance with j respect to fuel load, power ascension or operation.
t
~ CORRECTIVE ACTION PLAN / SCHEDULE:
1 The complete set of laboratory test records, along with the attached report and 1
corresponding documents, has been transmitted to the LP&L-Ebasco Quality
{ Assurance Records Vault.
1 -
The retaining site subcontractor records for com)1sted work have also tcen
, transferred to Ebasco. Records for the minimal construction and testing l activities are being turned over as work is completed. This will assure 1 accessibility and retrievability of subcontractor records and ultimate turnover to LP&L in accordance with the established records turnover program.
ATTACHMENTS:
1
! " Report on the Review and Analysis of Soil Backfill Densities" - NRC Concern "
l No. 7.
i l
i l
i l
7-5 f
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1 REPORT ON THE REVIEW AND ANALYSIS OF SOIL BACKFILL DENSITIES.
IN RESPONSE TO NRC CONCERN NO. 7 FOR LOUISIANA POWER & LIGHT COMPANY WATERFORD STEAM ELECTRIC STATION UNIT #3 i
EBASCO SERVICES INCORPORATED i AUGUST, 1984 4
REVISION 1 NOVEMBER, 1984 i
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, REVIEW AND ANALYSIS OF SOIL I BACKFILL DENSITIES i
, NRC CONCERN NO. 7 TABLE OF CONTENTS
, PAGE
- 1. INTRODUCTION ,' 1 Table No. 1 - Study Plan Flow Chart 1
- 2.
SUMMARY
AND CONCLUSION. 1 i
A. Test Records 2
- B. Inspection Reports 2
- 3. STAGE I - LOCATION OF EXISTING DATA 3-5 Document 1 - Ebasco Specification, LOU-1564.482 3 ,
Document 2 - Ebasco Quality Control Inspection 3 Procedures, QCIP-2/WQC-1 Document 3 - J.A. Jones Site inspection & 4 Testina. Procedure for Backfill &
Testing, W-SITP-12 Document 4 - Soils Laboratory - Class A Backfill 4 '
, Test Index Document 5 - Soils Laboratory - Class A Backfill 4 Field and Laboratory Test Summary 1
Document 6 - Ebasco Statistical St dies of Class 4 A Backfill Relative Dermities
\
l Document 7 - Class A Backfill Samples Forms . 5
- 4. STAGE II - REVIEW OF SOIL PACKAGES FOR COMPLETENESS 5-12 A - Test Records 5 Document 8 - Class A Backfill Test Index By Fill 6 Number'In Ascending Elevation Document S - Class A Backfill Relative Density 6 Overlay Plots 4
B - Inspection Reports 7
- 1. Description of Inspection Forms 8 l
- 1. Completeness and Distribution of Inspectiers 9 Table No. 2 - Inventory of Soil 9 Inspection Reports f *
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, REVIEW A13 ANALYSIS OF' SOIL BACKFILL DENSITIES x
- .. NRC CONCERN NO. 7 TABLE OF CONTENTS
./**
, PAGE Table No. 3 - Evaluation of Soil 10 InspectionReportsby},
SurfaceAreaCoverEhel L . ../ ~
Table No. 4h - Relative Distribution of 11 Inspection-Reports to Density
, Tests
- 5. STAGE III - REVIEW AND EVALUATION OF SOIL PACKAGES FOR 17-21' TECHNICAL ADEQUACY AND-SPECIFICATION COMPLIANCE A. Test Records .. 12
- 1. Testing Frequency and Distribution of 13 In-Place Densities Tests ,
~
Table No. 5 - Comparison of In-Place ~
14 Density Test Frequency and Distribution
- 2. Frequency of Laboratory Control Te'ats 15 Table No. 6 - Frequency Che Proctors / 16 Sieves to Densities Table No. 7 - Nonconforming Intervals - 16 Proctors / Sieves to Densities .
Table No. 8 - Analysis of Nohconfo ming 16 Control Test Frequencies
- 3. Performance of Statistical Studies 16 Table No. 9 - Schedule of Relative Density 16 Correlation Testing
- 4. Claes A Backfill Relative Density 17 j B. Inspection Reports 19 Completeness 19 Distribution 19 TABLES - 1 THROUGH 9 APPENDIX A - In-Place Density Tests - Fill $ EL -41.75 to EL -36.25 DOCUMENTS _ - 1 THROUGH 9 (Available at Waterford-3 for Review) c-
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REVIEW AND ANALYSIS OF SOIL BACKFILL DENSITIES-NRC CONCERN NO. 7 i
- 1. _ INTRODUCTION l
t In the NRC letter of June 13, 1984, the following Concern Ko. 7 was expressed re3ative to the Soil Backfill Densities:
ITEM NO: 7 i
TITLE: BACKFILL SOIL DENSITIES l l
NRC DESCRIPTION OF CONCERN: '
The staff found that records are missing for the in-place density test of backfill in Area 5 (first 5' starting at Elevation -41.25'). These documents are important because the seismic response of the plant is a function of the soil densities.
LP&L shall (1) conduct a review of all soil packages for comSlateness i
at.d technical adequacy and locate all records and provide closure on technical questions, or (2) conduct a review of all soil packages for completeness and technical adequacy and where soil volume cannot be verified by records as meeting criteria, perform and document actual soil conditions by utilizing penetration tests or other motifbds, or (3) justify by analysis that the soil volumes with sh sing records, or technical problems as defined after the records review, are not critical in the structural capability c2 the plant under seismic loads.
4
- ' In response to the above stated concern, the Ebasco Civil ESSE Department implemented a three stage program to resolve this concern. The review and evaluation of soil test records was conducted in accordance with approach (1) of the concern while the review and evaluation of inspection reports was conductad in accordance with appre.ach (3) of the concern.
The study plan depicted in Table.1 and described herein, was implemen:ed to determine if the deficiencies that do exist in the soil packages will critically effect the structural - - '
capacity of the plant under seismic loadings.
Stage I of the program consisted of a data acquisition effort. After the data was located and collected, the Stage II effort consisted of a review for completeness and data compilation. Finally, the Stage III activity consisted of an overall review and evaluation of the soil packages for technical adequacy and spacification compliance.
The program effo::t was conducted under the direction of M. Teachin. the Resident Sr. Site Soils Engineer, who was present during the performance of the majority of the actual backfilling operations.
t
- 2.
SUMMARY
AND CONCLUSIONS i
As a result of the study program described herein, it has been concluded that:
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REVIEW AND: ANALYSIS OF SOIL
< BACKFILL DENSITIES j
' NRC CONCERN NO.- 7 A. Test Records j
.(1) The Class A Backfill soil test records are complete.
(2) Field and laboratory tests were performed in accordance with the spe.ified frequencies. In less than 8% of the cases reviewed, the laboratory control tests were run at intervals slightly ',
larger than the specified, one set per ten inplace density test l criteria. The backfin placed during these periods was randomly located throughout the fills and the relative densities obtained during these intervair -:ere found to be acceptable when compared to the specification iequirements.
(3) Field tests were located in accordance with the specified random distribution. In less than 5% of the tests reviewed, the location coordinates of the inplace density tests were found to
. be in error. These tests were still a valid indicator of the relative density of the backfill at a random spot at a known elevation in a known fill area and were therefore found to be acceptable tests.
(4) Statistical studies of relative density were performed in accordance with the specification requirements.
(5) The Class A backfill soil densities are in accordance with the specification requirements and win provide the design structural capability to the plant under seismic loads.
B. Inspection Reports (1) The distribution of the existing documentation throughout the backfin is essentially identical to the distribution of the field testing effort, thus indicating a one to one relationship between inspection and testing activities. Since the field testing activity is known to be complete, the inspection activity is also believed to be complete.
The majority of the missier 't.spction reports are enerefore believed to be misplaced.
- i. Spection trends based upon evaluation of inspecmt. 9e; ency and distribution indicate that the majority of the n; a.es :spections were performed.
(2) 80% of the volume of ene backfin has a sufficient quantity of each type of inspection report to fulfill the requirements of the l specification and inspection procedures.
(3) For the remainder of the volume of the backfin which has missing inspection reports:
l
. -(a) 16.0% of the volume of the backfill'has an average of.81% of ;
the quantity of inspection reports required with at least '
one of each type of inspection report on each fill at each elevation in its volume.
.- 2 --
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REVIEW AND ANALYSIS OF SOIL BACKFILL DENSITIES ,
NRC CONCERN NO. 7
'(b) 3.8% of the volume of the backfin has a partiany complete L
representation of inspection reports with one or more type
, of inspection missing on each fill at each elevation in its
, volume.
l (c) 0.2% of the volume of the backfill has no inspection reports at the flu locations and elevations included in this volume.
The effect on each of these types of deficiencies has been l
evaluated and found to have no effect on the structural capability of the plant under seismic loads.
- 3. STAGE I - LOCATION OF EXISTING DATA The primary emphasis of.the Stage I activity was the collection of soils.
data which in addition to specifications and procedures, includes test records and inspection reports. To accomplish this task, a detailed review was performed of the following data locations:
Ebasco Quality Assurance Records Vault Ebasco Engineering Files Ebasco Warehouse On-Site Laboratory Files (G.E.O.)
Contractor Quality Assurance Records Vault (J. A. Jones)
As a result of this effort, several key document packages were located end are attached to this report for permanent storage. A brief description of each of these document packages is presented below. Ihe hierachy of the documents is depicted in the Study Plan Flow Chart, Table N,o.1 attached.
DOCUMENT 1 -
Ebasco Specification LOU-1564.482, R7 Filter and Backfill.
-TKis is the latest revision of the specification under which an soil backfin was selected, placed, compacted and tested. The document presents the design requirements of the backfill activity and served as the basis for the development of the two Quality Inspection Procedures summarized below.
I DOCUMENT 2 - Ebasco Quality Control Inspection Procedures, QCIP-2, RH and WQC-1, RA l These are the Ebasco Quality Control Inspection Procedures under which the soil backfin material was selected, placed, compacted, tes'ted, documented and approved.
I
. _ . _ i REVIEW AND ANALYSIS OF SOIL
. BACKFILL DEN 21 TIES NRC CONCERN NO. 7
, DOCUMENT 3 - J. A. Jones Site Inspection and Test Procedure for Backfill and Compaction, W-SITP-12. R8 l
This is the latest-revision of the Contractor's Quality Verification procedure under which all soil backfill material was selected, placed, ,
compacted, tested-and documented. '
Each of these documents required the performance of routine field and laboratory testing of the backfill material. The actual soil testing was
. performed by an onsite laboratory in accordance with these requirements.
The following control documents were generated by the soils laboratory in addition to the standard set of test reports.
DOCUMENT 4 - Soils Laboratcry - Class A Bar.kfill Test Index This index was developed by the test laboratory as a working record of each Class A test performed. This hardcover, bound notebook lists the test number, location coordinate, elevation date and type of test performed. It was developed as a system of assigning numbers to and documenting the counletion of all Class A tests.
DOCUMENT 5 - Soils Laboratory - Class A Backfill Field and Laboratory Test Summary This summary was developed by the soil testing laboratory as a daily tabulation of the results of soil testing performed. Contained in this document are the lab test number, fill number, test location, field density, Ian density, grain size and relative density test results for each day of work, recorded on a single page for supervisory review and study.
Utilizing these records, Ebasco performed the required periodic statistical studies of insitu relative density of the backfill as described in brief-in Document 6 below.
DOCUMENT 6 - Ebasco Statistical Studies of Class A Backfill Relative Densities This document contains all of the seven statistical studies performed on the Class A backfill relativc densities which document the backfills overall acceptability. It also contains letters to the earthwork contractors regulating the percent compaction criteria based upon the results of these studies.
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REVIEW AND ANALYSIS OF SOIL BACKFILL DENSITIES NRC CONCERN NO. 7 DOCUMENT 7 - Class A Backfill Inspection Reports In order to review the large quantity of inspection reports which aske up La soil packages in the files, nine basic types of fores' were identified. Document 7 contains samples of the typical forms found in each cf the soil packages in the vault. These forms are discussed in detail in Stage.II of the report.
After locating and collecting the data, Stage II activities concentrated on a review of the documents for completeness and on compiling the data into a format compatible for review of NRC Concerns.
In order te perform this task, the 17,000 existing soil documents were divided into the following two types:
(1) Soil Inspection Reports (Forms 1-5)
(2) Soil Test Records (Fcras 6-9)
Since-the test records provide a direct measure of the capability of the backfill to provide the required structural support to the plant island under seismic loadings, they were the first records to be reviewed. The remaining inspection reports were reviewed after the completion of the test record stady. The details of these activities are presented below.
- 4. STAGE II - REVIEW OF SOIL PACKAGES FOR COMPLETENESS 1
A. Test Records The first step in the review of the documentation was a detailed review of all soils laboratory documentation on site for completeness. Included in i the review were:
In-Place Density Tests - ASTM 2167 Forn 6 Proctor Tests - AS'Di 1557 Forn 7 i
Moisture Content Tests - ASTM D2216 Forn 8 Sieve Tests - ASIM D422 Form 9 Relative Density Tests - ASTM D2049 (Off Site Lab)
By comparing the Class A Backfill Test Index (Document 4) and the Field and i
Laboratory Soil Test Summary (Document 5) ta the actual files of soil test data at the onsite laboratory, a complete set of field and laboratory test records was found to exist.
In direct response to the first paragraph of the NRC Concern No. 7, attached in Appendix "A" are copies of the 34 in-place density tests j performed in the first 5.5' of fill placed in Fill Area f5 from Elevation t
-41.75 to EL -36.25. In addition to the density tests records Table A-1 summarizes the elevation of the test, the test coordinate, the test number, the date the test was performed and, documents the number of the reference proctor and grain size lab tests used to determine specification compliance. Each test location and relative density are plotted on the corresponding overlay plots in Document 9 of this report.
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C REVIEW APQ ANALYSIS OF SOIL RAcrFILL DENSITIES NRC CONCERN NO. 7 Utilizing the complete set of backfin density test records and the Class A Backfin Field and Laioratory Test Summary (Documecst No. 5), and keeping in mind the goals of completeness and technical adequacy, two new documents were developed for subsequent evaluation. 'A brief description of sach of these documents and methodology used to prepare the documents is presented below.
DOCUMENT 8 - Class A Backfill Test Inder by Fill Number in Ascending Elevation This document is a complete listing of all Class A density tests categorized by fill area in order of ascending elevation. It lists for each fill area, the field density test location, number and date of performance in order of ascending elevation.
This tabulation served as the basis for the preparation of the overlays of relative density by elevation, Document 9 discussed below.-
DOCUMENT 9 - Class A Backfill Relative Density Overlay Plots By Elevation In order to evaluate the frequency and distribution of field test and relative density, the following procedure was used to construct the overlay plots:
(1) All Class A density tests were regrouped by fill numbe'r in order of ascending elevation (Document No. 8).
(2) A key plan drawing of the plant island excavation was constructed containing the soil backfin grid system. One original sheet was used for each one foot interval of backfill. Relative dunsity overlay plots were then constructed from EL -44 to Elevation +20 to encompass all Class A backfill density tests.
(3) Using Document 8, each det.aity test was plotted on the form using
' the test coordinates and elevation. A different symbol was used for each respective fill number. The test number was recorded adjacent to each data plot. It should be noted that the boundaries of each fill area are not represented. This is 1
because the boundaries were somewhat arbitrary and changed in exact location at different elevations in the fill. In addition, t
backfill activities typically involved areas smaller than the numbered fill area, and in some cases, was carried across fill boundaries.
(4) The test number was then recorded in the test schedule on the side of the overlay along with the relative density value for each test found from the Class A backfill Test Suasary (Document 5).
(5) For Class A backfill placed above Elevation +13 (See Statistical i Study No. 7 Document 6), the percent compaction value for each i
' field test was found in the Class A Backfill Field and Laboratory Test Summary-(Document 5) and recorded in the test schedule with as asterisk.
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-REVIEW AND ANALYSIS OF SOIL BAJKFILL DENSITIES NRC CONCERN NO. 7 (6) Once the data was plotted and tabulated, ~ the theoretical surface boundaries of the backfill were approximated utilizing the fill boundaries and -the Nuclear Plant Island exterior walls. The surface area of the backfill at each elevation was then calculated with a planimater and recorded on the overlay.
~(7) In cases'where the actual distribution of the pic?ted density tests indicated backfill placement outside of the theoretical-houndaries, the fill boundary was extended to include that material.
(8) By dividing the surface area by 20,000 ft , the minimum number of density tests required by the Specification LOU-1564.482 was calculated and recorded on the overlay.
(9) Finally, the actual number of density tests performed at each elevation was recorded, completing the overlay.
The completed overlay plots are a graphical presentation of the density I ter:: frequency and distribution, and most importantly, they tabulate and
. display the final insitu relative densities and/or percent compaction of the backfill.
These plots were utilized in the review and evaluation of Test Records for technical adequacy and specification compliance in the Stage III-A of the 4 Study Program.
- B. Inspection Reports i
I In the review and evaluation of the completeness of the inspection i documentation, the following factors were considered:
j i
The requirements of the Quality Control Inspection Procedure in force at the time the work was done. Three different Ebasco procedures and '
one Contractor procedure existed during the eight years of placement.
Each procedure was ravised numerous times. Therefore, different inspection report forms were in use at different times during backfilling operations.
4 The location and elevation of the fill. Some forms were used to
, documene inspections of activities which were not cosumon to all fill j placements. Therefore all forms were not required in all packages.
The frequency of inspection. Some backfilling activities required
- 100% Ebasco inspection and others not. Since the work was done by a i contractor that had an acceptabla quality assurance program, Ebasco l inspection was designated as "once per day, by Checklist, when work is i in progress." (QCIP-2, Section 8.4.2 - Document 2).
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- REVIEW AND ANALYSIS GF SOIL BACKFILL DENSITIES l NRC CONCERN NO.. 7 ,
(1) Description of Inspection Forms Considering these variations in procedures, fill locations and inspection frequencies, the following basic inspection report forms were found to exist, samples of which are found in Document 7:
Form #1 - J. A. Jones Daily Backfill Inspection Reports W-SITP-12 (R1-R8)
These forms suaserized the overall acceptability of the daily backfill operation including material acceptability, excavation, backfill placement and compaction, and field testing. They were completed by the contractor on a daily basis for each backfill area of major earthwork.
Form #2 - Eoasec, Borrow Material Inspection Reports QCIP-2-1$QC-1-9
- These forms summarized the acceptability of the borrow material used for Class A backfill including the material source, moisture content and gradat. ion check test results. This inspection was performed by Ebasco daily.
Form #3 - Ebasco Excavation and Stripping Inspection Reports
~
QCIP-2-2NQC-1-17 These forms summarized the acceptability of the activities performed in ptoparing the fill area for the new backfill placemact. Included on this form are drainage conditions, stripping, excavation, cleanup i
and moisture and density testing of exposed materials. The form was primarily utilized for excavation stripping and grubbing when the Class A backfill abutted and joined the natural clay slopes (below EL
-5). Above this elevation, the use of this form was up to the discretion of the Ebasco Inspector.
Form #4 - Ebasco Daily Backfill Inspection Reports QCIP-2-3NQC-18 These forms summarized the acceptability of the daily backfill operation emphasizing the backfill placement, compaction and field testing. It.is very similar to the Form #1 completed daily by the .1.
I A. Jones, quality verification inspection force and was utilized daily by Ebasco for all major Class A backfills.
4 Form #5 - Ebasco Backfill Acceptance Report l
- QCIP-2-4 l l
This form summarized the findings of the Ebasco inspection repcet forms #2, 3 & 4 and the soil laboratory test results resulting La the overall acceptance of a particular fill. The form was discontinued in revision H of QCIP-2 (12/6/77).
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l REVIEW AND ANALYSIS OF SOIL BACKFILL DENSITIES l NRC CONCERN NO. 7 (2) Completeness and Distribution of Inspections During the Stage II review activity, the total file of inspection reports for Class A backfill was inventoried and combined into compatiMa soil packages as exemplified in Document 7. Included in the inventory were approximately 12,000 inspection. reports ranging from EL -44 to EL +20 throughout all seven fill areas. The reporte were grouped and compiled by fill location, elevation aR placement date for each of the five types of ,
inspection forms summarized above.- The resulting inventory of inspection reports-is presented in Table No. 2 and discussed below. l The evaluation of these inspection reports was further divided into two phases; the evaluation of the inspection reports to dete sina.their overall completeness and the evaluation of. the frequency and distribution of inspection reports to determine their content. The following discussions summarize the results of these evaluations: l
- a. Completeness cf Inspections
, In the evaluation of- the com;.leteness cf the inspection documentation, it misst be noted that the exact aumbers of I- . inspection documentatien required by the governing procedures cannot be reconstructed. Certain of the five types of inspections were required on a daily basis (100% coverage - Forms 1, 2 & 4) shile others were required on a partial coverage basis i (Form 3 & 5). For this reason several comparitive analyses were performed to evaluate relative completeness of the documentation.
When evaluating the total number of . forms existing for each type-
- of inspection (Table 2), it is found that Forms 2 and 4, which are representative of the required 100% inspection, number an average of 2900 each, and that Forms 3 and 5, which are '
' representative of a partial inspection, number as average of 2000 each inspections. Tne Form 1 inspection (J. A. Jones Daily Inspection Report)-which was performed at a 100% coverage and
{ :.tus should have resulted in approximately 2900 forms, appears to j be incomplete. It must be noted, however, that the Form 1 daily inspections by J. A.' Jones and the Form 4, Daily Inspections by Ebasco, were duplicate inspections of the same placement and compaction activities. Since the missing Form 1 data is found on the duplicate Form 4 Inspection Reports, which eppear to be complete, the missing Form 1 Reports constitute no loss of quality documentation and have no further significance to the
! inspection report evaluation unless the corresponding Form 4 is j- missing. Thus'the existing inspection documentation woulf.
indicate that 100% inspection coverage consists of 2900 l inspections. -
4 In order to evaluate the validity of this number, consideration was given to the complete set of field density test records presented in Table No. 5 (which will be discussed in more detail in the evaluation discussions of density testing). This table indicates that 3076 Class A density tests were performed when i_ _, . ~~. .___.__.i__. ____1..___.___.__._..__ _ _ _ _ _ _ _ . _ _ _ _
- - ~ - - - - . _ . .
-REVIEW AND ANALYSIS OF SOIL BACKFILT DENSITIES NRC C0JCERN NC. 7 on{y858testswererequiredbasedupontheonetestper20,000 ft specif16.0 frequency. Thus approximately three times as many tests were psrformed as the fin surface area would require.
i Sincethespecgficationalsorequiresonetestforeacharealess than 20,000 ft placed in any one day, the existence of so reuy extra tests would indicate that2 the large majority of fills placed were less than 20,000 ft and thgt the testing frequency was governed by the less than 20,000 ft placed in any one day
. criterion. This is further subetantiated by a review of the density overlay plots (Document 9) which clearly indicate small fin placements at the upper elevations and around specific construction items. This bging the case, since each seau fin area of less than 20.000 ft worked required a test, it would also rnquire a set of inspections for the same fin area. Noting that the 3076 field density tests constitute a complete set of test records and considering the correlation developed above it is reasonable to conclude that the total number of inspection report packages for 100% coverage should also number around 3076.
Taking into account that a small percentage of fills had more thanonedensitytgetperfill,becausetheirsurfaceareas exceed (d 20,000 ft., tha number of required inspection packages should be slightly less. By comparing the 2900 existing inspections that represent the 100% inspection frequency to the 3076(-) packages which should have existed. It is concluded that ,
1 based on this comparison, the inspection documentation files are substantially complete.
To further evaluate and better define the completeness of the inspection reports, e comparative analysis ws? performed of the i
surface area indicated on the Inspection Reports to the total surface area of~the fill areas.
! In this analysis, the surface area recorded in each of the daily inspection report packages (Form 1 or 4) was totalled and compared to the total surface area of the backfill at each j elevation as calculated on the overlay plots (Document 9). By comparing the actual surface area of backfin inspected to the
{ total surface area of backfin placed, the percentage of inspection coverage was calculated. The results of this analysis are summarized in Table No. 3 and discussed below:
(1) The actual inspected surface area in some cases was larger 3 i
than the theoretical surface area (overlay plots). This is because many fill areas were constructed .on more than one day, thus generating two reports for the same area.
~
- (2) Evaluation of the percent of inspection coverage colu an of -
{ Table 3 indicates that for PO% of the volume of the j backfill, there exists a sufficient quantity of each type of inspection to document the acceptability of the backfin represented by the inspected surface area.
- i 6
l
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W, REVIEW AND ANALYSIS OF SOIL-'
. BACKFILL DENSITIES
'NRC CONCERN NO. 7
- -(3) For the remaining.20% of the volume of.the backfill which was found to have missing inspection reports, the average percent of inspection coverage was found to be'81%.
As a reault of these ana3yses of:the completeness of the inspectica documentation, it is concluded that the documentation j is basically complete with 80% of the volume of the backfill 1 documented with complete soil packages and the remaining 20% of I
.the backfill containing partial deficiencies in the inspection
- reports..
- b. Distribution of Inspections As part of the evaluation of the significance of the missing inspection' reports, the distribution of the existing inspection.
documentation was evaluated.
. To consider the distribution of the existing inspection reports throughout the fill area, Table No. 4 was developed. It compares i the distribution'of the inspection effort to the distribution of
{ the field testing effort which is known to be complete. By
- . comparing the percent of inspections on each fill area to the i percent of field density tasting on each fill area, it is found that both the inspection and testing activities 1
- ave essentially identical distributions of effort. This observation further supports the correlation that approximately one inspection report should exist for each-density test and strengthens the conclusions that the inspection report documentation is basically
- complete.
In the further evaluation and definition of the distribution of the types of inspection reports shown in Table No. 2, two 1 distinct trends are immediately apparant, with the divir. ion in j trend at elevation -25.00.
4 (a) Between elevation -25 and the bottom of the excavation, there exist 52 fills with partial distribution of inspection i
- report documentation, or none at all. Of these 52 fills
25 fill areas have some types of inspections-by both
- the Contcactor and Ebasco. These fills constitute 6.3%
- of the total number of fills constructed and account for 1.8% of the total volume of Class A backfill constructed.
21 fill areas havs inspection documentation only by the l Contractor. These fills constitute 5.3% of the total number of fills constructed and account for 2.0% of the j total volume of Class A ba?kfill constructed.
i 4
-,---w- - -
-, w-----., , - -,---.,,,,-,-..-n.-rr--.y ,..--...,,,m._ ,----.,-ver--m**=- =- "ve------"=+- +-w - - *F m- e * '-
REVIEW AND ANALYSIS OF SOIL BACKFILL DENSITIES NRC CONCERN NO.'7 6 fin areas have no inspection documentation. - These fills constitute 1.5% of the total number of fills constructed and account for only 0.2% of the total volume of backfill constructed.
(b) -For the ramminder of the fi n placements between elevation
-25 and plant grade with minor exception, the data in Table 2 indicates that each type of inspection was performed at least once on each fin area at each elevation. In some cases, as many as 60 inspections of a particular' type were i
performed or. one fill at one elevation (Fin #6, EL 13.00 -
13.99).
Thus, a review of the distribution of the types of inspection >
repcres th.t are missing indicates that the 52 fill areas with an incomplete distribution of inspection documentation are concentrated in 13.1% of the total number of fin areas constructed and account for only 4% of the total volume of-backfin pieced.
The impact of these findings on the evaluation of the techaical adequacy of the inspect $on reports is discussed in Stage III-B of
( ,
this report.
- 4. STAGE III - REVIEW AND EVALUATION OF SOIL PACKAGES FOR TECHNICAL ADEQUACY AND SPECIFICATION COMPLIANCE ,
A. Test Rscords The review and evaluation of the technical adequacy of the Class A l backfin to provide structural capability of the plant under seismic
- loadings was based upon the design requirements as stated in the Ebasco Specification LOU-1564.482. Those sections pertinent to the 1 Class A backfin soil density are as fo nows
i I
In-Place Density and Testina i
Sand materials and clan shell to be used as Class A backfin shall j have an in-place relative density of 75 percent. The variation for Class A fill from the above specified degrees of compaction shall be a maxim a of one standard deviation less than the specified relative d a.c.sity . Thn nuaarical value of the standard deviation from Class A '
I 1111 win be established by a series of field tests to be conducted during the initial compaction operations and will be reported in terms of minism allowable density required.
' The minimum allowable density for the basis of field control at the start of work and until establishment of the standard deviation for i
Class A fin shall be 95 percent of Modified Proctor. The required percent compaction will be adjusted either up or down, depending upon j the results of statistical studies which win be made during the
- backfilling operations in order to maintain the 75 percent relative density requirement.
Me
. _ , , - - - - . . _ . - . _ . . . , . . , - . .,_.- - ~.-. .,. - , _ _ - .___-,.____-- . - - - - . - . , - - - - _ . _
REVIEW AND ANALYSIS OF SOIL BACKFILL DENSITIES E2C CONCEkN NO. 7
" Clay materials to be used for' Class A backfilj shall have in-place density of 90 percent of the maximum density obtained in the Modified Proctor Compaction Test. All materials to be used for Class B backfill shall have an in-place density of 90 percent of the maxima density obtained in the Modified Proctor Compaction T6st. The variation from the'above specified degrees of compaction shall be a
- maximum of 10 percent of the density test results falling a maximum of 5 percent less than the specified density in a random distribution as determined by rhe Engineer.
d-
.1 Control tests of desities and moisture contents shall be made by the Engineer as the work progresses, to assure trat required '
densities and moisture contents are being achieved.
.2 The in-place' density stall be tested in accordance with AS1M-D1556 ASTM-D2167, ASTM-D2922 and any.other method suitable in the judgment of the Engineer to inst.re that the backfill has j-been properly compacted. One test shall be made in each layer for every 20,000 sq.ft. of compacted Class A fill area and one test for every area of less than 20,000 sq. ft. pisted in one day.
.3 The optimum conditionc for both moist.ure and density will be determined by the Engineer for the fill materials. One.
i laboratory density test (ASTH-D1557) and one. mechanical gradation test (ASTM-D422) shall be. performed ou samples taken fron
' in-place density test holes for each ten in-place density tests performed. The results of these tests made during the backfilling operation shall be made available to the Contzactor."
d In summary, the basic criterion of the specification were to:
Obtain 75% relative density in the Class A fill.
To check the compaction of the fill with field in-place
+
' density and moisture tests and laboratory density and i
gradation tests at specified frequencies.
To perform periodic statistical studies of the Class A backfill relative density in order to evaluate the results.
Compliance with these requirements is discussed in the following sections.
(1) Test Frequency and Distribution of In-Place Densities By using the completed density overlay plots (Document 9), the frequency of Class A in-place density tests (ASTM D-2167) performed for each one foot elevation of backfill was compared to j
the backfill specification criteria stated above. Since each in-i place density test includes a moisture tent, verification of moisture testo sas simultaneously deve loped with the density j revice.
- e e>6ee e or -- .w-* wgr----*,v w--ew-gr*-y-ey,vr yrw hg-arevy-myw-
- . -~ .- - .. . . . _. -
REVIEW AND ANALYSIS OF SOILL ,
BACKFILL TNSITIES NRC CONChatN NO. 7 l.
In addition to this criteria, inherent in the' requirement for the '{
performance.of statistical studies'is the need to demonstrate a' '
random distribution of test data.- By studying the location of i
tests on each fill, an evaluation of the random distribution of the test pattern was also performed.
g Table No. 5 and Document 9, the overlay plots present a summary -'
of the results of these reviews. The minimum number of field density tests required for each fill was tabulated along with the actual number of, tests performed and the distribution of those-tests.by fill number. @
i SincetherelativedensityoverlayplotswerecoII:structedateven one foot inteu als.and the backfill was placed in 15" lifts, density tests at an elevation one foot above and be'.et each plot i
were reviewed to detera!.de specification compliance. In addition, backfill placed in adjaceng fills was also evaluated since each test represents 20,000 ft of backfill. Thus, by superimposing three overlay sheets (36" of compacted fill), a i
-three dimension test distribution was reviewed for each lift of backfill.
The results of a simultaneous review of Table No. 5 and the overlay plots indicates the following:
(a). A comparison of the total volume of the Class A backfill shown on the overlays to the neatlina quantity shown on the j
design drawing (LOU-1564-G-497S01, Ro) indicates that the overlay Class A soil volume is 33% larger than the design ~
quantity. This is duc to the actual expansion of the Class A fill '.oundaries into Class B fill areas at the higher
+
- elevations during construction (as shown c7 the overlays as indicated by actual test locations). Taking the expanded
' backfill boundaries into account, the following evaluations were made (b) Based on ghe testing frequency of me field density test per
', 20,000 ft of fill, 2794 in-place density tests were performed in fill areas requiring 858 tests. Thus, approximataly three times as many density tests were run as the surface area of the fills required. This was due to the placement of numerous smaller fills each day at the higher elevations, as described in Section 4.B.2.a above.
, l (c) On only one fill of the 385 fills studied, was there an inadequate number of density tests performed in the 3 l
' foot wedge of ba'ckfill reviewed (Fill #2, EL -19). In this case, the size of the fill was small and the relative densities of the fills on both sides and above and below this fill all met the specification requirements.
Therefore, it is concluded that this' deficiency will have no i significance on the stability of the Plant Island under the j event of seismic loadings-i
(
- l t
REVIEW /ED ANALYSIS OF SOIL BACKFILL DENSITIES NRC CONCERN NO. 7 *
(d) Visual analysis of the location of the density tests shows them to be completely random and distributed without pattern throughout the bac1.C111. It should be noted that some test locations on the lab forms were found to be in error (approximately 5%) when plotted on the ovetlays. This is certainly due to the inaccuracies of visually locating ones position in the field off of sign posts hundreds of feet away and tens of feet above the actual test elevation.
Since these test locations were still indicative of the relative density at a random spot on the fill, the density values were accepted' as valid and included in the density analyses.
Taking these factors into consideration, it has been (stermined that the spenfication requirements n>r in-place test frequency and distribution have been complied with.
(2) Frequency of Laboratory Control Tests By using the Class A Backfill Test Index (Document 4) and the
! Field and Laboratory Soil Test Summary (Document 5), the frequency of the laboratory density control tests performed (ASTM D1557) and the mechanical gradation control tests performed (ASTM i D-422) was compared to the specification requirements.
4 Table No. 6 presents the results of a detailed review of the laboratory testing frequency compared to the nusber of in-place density tests performed between laboratory check tests. Using the specification requirement of one set of control tests per tan in-place density tests, all nonconforming test intervals were tabulated in Table No. 7.
An evaluation of the data presented in these tables indicates the i following:
(a) From the start of Class A backfilling operation in January, 3 1976 to the present date, a total of 3137 Class A in-place dansity tests have osen performed. Jf these 2794 tests are in backfill ~ subject to potential liquefaction while the i
remaining 282 test are above this sone. During the same period of time, 361 sets of control tests (Proctor, Sieve and Moisture Tests) have been performed, thus averaging one.
set of tests per 8.6 in-place density tests compared to one set per 10 in-place density tests as required in the specification.
I (b) During the performance of tha 361 sets of ccetrol tests, in l only 27 instances were the tests performed at intervals '
larger than the specification requirements. Thus, the j control test frequency was adhered to 92.5% of :ne time in '
the last eight and one half years of backfilling activity.
1 1 I
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REVIEW AND ANALYSIS Of SOIL BACKFILL DENSITIES NRC CONCERN NO. 7 *
\
(c) Analysis of the nonconforming intervals indicates that in 20 ;
of the 27 cases, the test interval was extended from 10 to a !
maximum of 13 field tests per set of control tests. Since in each of these cases, the extra in-place density tests included in the extended intervel were in material on the sans fills, already tested in the allowable 10 density I'
tests, the intent of the specification was complied with in these cases.- By accepting these intervals, die intent of the specification req.airement on control test frequency was 4
adhered to 99.8% of the time.
(d) -In the remaining seven cases, where the control test interval was extended from 15 to a maximum of 29, a review of the test locations and relative density test results presented in Table No. 8 1ndicates that the test intervals are completely random through the fill as a whole and that the relative densities obtained during these intervals are -
all acceptable within the statistical tolerance of the specificaticc.
1
^
Taking these factort, into consideration, it h.s been determined that the specification requirements for the performance of laboratory control tests relative to Class A backfill in-place
, density testing, has been c.aplied with.
- (3) Performance of Statistical Studies 4
Document 6 presents copies of all seven statistical studies performed during the actual backfilling operation, in addition to
, letters to the backfilling contractors informing them of the results. In addition. Table No. 9 presents the schedule of relative density correlation testing showing the periodic updating of these correlation curves during the major period of backfilling operations.
j From these documents it has been concluded that
' (a) The specification requirements for the periodic performance of statistical studies during the backfilling operations has j been complied with and that; i
I (b) The value of the field control (percent compaction) was adjusted either up or down, depending on the results of the statistical studies. ,
- Taking these factors into consideration, it has been concluded
{ that the statistical review of the relative densities of the
' Class A backfill was performed during the backfilling operations i
in accordance with the specification requirements.
4 T
g
-4 . . _ . , - _ . . _r_, - . , . . . . _~._m x . .._. _.-_.,...-._,- - ,--.,-_-.___. --.....__ .. __.,_ -.. _.. __., -
-; ,- . . .~ - - - -- . .
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- '. . l
-REVIEW AND ANALYSIS 0F SOIL. i TBACKFILL' DENSITIES ~ '
NRC CONCERN NO. 7 .
. -(4) ' Class'A Backfill Relative Dansity- l j.
-1
, In analyzing the relative density of,the compacted Class'A 3 backfill as a whole, the following statistical approach was .j adopted to. comply with the specification requirements. 1 f '
The specification required the in-place compacted Class A
. backfill to have' a relative density.of 75 percent. The - all naable .j variation for the Class A fill less'than the specified density t
^
g
~was a maximum cf one. standard deviation. The numerical value of f; the standard deviation for this material was periodically
^ '
established by, conducting a series of studies on field tests and was reported in terms of minimum allowable proctor density-- !
requittd to yield'the required relative density.'. l During the performance'of these statistica1 studies, the field densities were converted to rciative densities by the use of the correlation curves. The correlation curves were constructed '!
using cut.ulative test. data from random samples .taken. from tne !
fill. The following procedure was used to develop these curves. 1
- For each f amily of materials:
[ (a) A representative 300 lb. sample was obtained from the fill for every 200 to 250 in-place density tests petformed. ,
(b) A'100 lb sample was,sent to the field l'ab and a 200 lb saeple was sent to the home of fice lab (Peabody Testing) for i parallel testing to determine a modified proctor compaction [
curve and percet finer than a #200 sieve.
(c) The parallel results were compared. The Proctor densities were found to agree within 12.pcf and the percents finer.
' " than the #200 sieve within 13 percent.= Thyrefore, the home 1 oGice lab proceeded to perform maximum (O max) and minimum (0nin) density determinations on the material.
(d) The following equation was used to plot the correlation curves.
Dry Density = ( max.) x ( 5 min)
{ max.- Dr ( fmax.- g min).
Where:
Dry' Density = field dry density Dr = relative density i Imax,dmin.=measuredinthehomelabforthismaterial type.
{ i.
!~ .
. . , , . _ . , _ -.~ ,~ ,.- ---.-.- _ _ ,..~ - . _ - . - . ~ . . _ _ . . - - , _ - - - _ _ - . .-
I REVIEW AND ANALYSIS OF SOIL BACKFILL DEEJITIES
'NRC CONCERN NO. 7 -
Each curve was established by assuming various D values and calculating Dry Densities.
- Cumulative S;atistical Study No. 6 (Document No. 6) was performed in August of 1978, and represented all class A backfill placed to i that date. Statistical Study No. 7 was performed in July,1984 i
'and includes the remainder of Class A tests in the backfill subject to potential liquefaction. For both studies, correlation curves of field density to proctor density were developed for three family of materials. The results of these studies are .
summariaed as follows: * '
.itudy No. 6
- Based upon the standard properties of the normal bell curve, the j cumulative Study No. 6 was performed on 2499 Class A backfill tests. The density values of the original failing Class A density tests (that were retested) were not included in this study since those tests did not represent the final densit/ o.!
the backfill which formed the seismic support of the Flaat
, Island.
1 i The study determined that the standard deviation for all Class A i backfill was 12.4%. The specification tolerances were then defined'by this standard deviction (in a thr e standard ' deviation 4
universe) ast (a) 13% of the Class A backfill tests could have relative densities ranging from 62.6% to 75.0% and (b) 3% of the Class A backfill tests could have relative densities ranging from 50.2% to 62.6%.
Using these definitions, cumulative Study No. 6 concluded that i
the Class A backfill was constructed in accordance with the 75%
- relative density requirement. In addition, those tests which
) fell below 75%, were found to be within the specification j
tolerances when compared to an allowable tolerances of 16%.
{ Therefore, the backfill was found :o be 1e .:ompliance with the specification requirements.
- Study No. 7 i
Study No. 7 consisted of 251 in-place density tests taken in backfill placed since August 1978 up to elsvation +13.00 (the upper boundary above nich liquefaction will not occur, see Study l No. 7. Document 6). The results of this study indicate a mean relative density of 91.7% with a standard deviation of 18.6%.
4 4 e -heDee
-,.v.-w---.-- ,,_ --.,,.- - ,,-..
,,,n,.y, -mm em.-.... w.-w.,--..-w%.w w-, w %. w,,,,,,.,
REVIEW AND ANALYSIS OF SOIL BACKFILL DENSITIES
- NRC CONCERN NO. 7 The mean relative density is well above the .pecification requirements and is somewhat higher than the mean relative density from study No. 6 (83.8%). The standard deviation for the I
current work is larger tha for previous studies. This is certainly not surprising considering the large variation in c.,apaction techniques utilised to construct backfill in the six years of operations included in this study.
The actual number (12.4%) and values of in-place density tests in !
Study No. 7 which fell below the =4=i=== density of 75% was fetad to be within the 16% allowable tolerance.
- In i.ummary, the backfill included in Study No. 7 was found to be in conformance with the specification requirements. Taking this into account und considering thats (1) All the backfilled placed prior to this study also was in compliance with the specification requirements; and (2) Study No. 7 completes the series of studies on backfill subject to potential liquefaction; it is concluded that all backfill was placed in compliance with j the specification requirements and that the final insitu soil 1
densities will provide the required design structural capacity to i
the plant under seismic loadings.
B. Inspection Reports The results of the Stage II evaluations on completeness and distribution of the existing inspection documentation, determined the followingt
, (1) Completeness of Inspections ,
j Although no exact method exists for determining the quantity of l inspections that were required during the backfill operations,
- two comparative analyses were performed to evaluate the relative
} completeness of the inspection documentation. These analyses i
concluded that the existing documentation is basically complete and that 80% of the volume of the backfill is documented with complete inspection packages while the remaining 20% of the backfill has some deficiency in the inspection packages.
(2) Distribution of Inspections The dist.ibution of the existing inspection documentation throughout the backfill is essentially identical to the distribution of the field testing effort by fill location, thus confirming a one to one relationship between inspection and testing activities.
i f
i ..m., - -
4.m.-.-.
l REVIEW AND ANALYSIS OF SOIL BACKFILL DENSITIES NRC CONCERN NO. 7
\
For the 20% of the inspection packages found to be incomplete, ;
three distinct types of discrepancies were found to exist. The j following discussions and conclusions are presented relative to the effect of these discrepancies on the technical adequacy of the inspections.
(a) 16.0% of the volume of the backfill has an average of 82% of the quantity of inspection reports required with at least one of each type of inspection report on each fill at each ,
elevation in this volume.
For utsmple, although there are 28 existing Form 2 Inspection Reports, in the vault for Fill No. 3 at elevation
'+12 (Table No. 3), 6 Form 2 inspection reports are believed i to be missing. In all these cases however, the 81 of existing documentation of each type of inspection clearly establishes that the Quality Control and Quality Verification processes were implemented during the construction process. In addition, the backfill relative density study documents that the required density tests were performed and resulting relative density for the fills included in this 16% volume were found to be within specification requirements. Thus the existing inspection repor.c coupled with the satisfactory density records indic. ate that this deficiency will have no signifzcance on 4
the stability of the Plant Island under seismic loadings.
(b) 3.8% of the volume of the backfill has a partially complete representation of inspection reports with one or more type of inspection missing on each fill at each elevation in this volume. Included in this volume of backfill aret
! 25 fi31a which have inspection records from both the j Contractor and Ebasco. Although some of the five
, required inspection reports are missing, there exists a sufficient quantity of data on the existing reports to determine that the Quality Control and Quality i
Verification processes were implemented during the construe. tion of each of these fill areas. In addition, the design specified relative densities were schieved i
within the specified tolerances (Section IIIA) for all
- the fills affected. Therefore, it has been concluded that this deficiency, which effects 1.8% of the j
backfill, will have no significance on the stability of the Plant Island under the event of seismic loading.
Also, included in these fill areas are 21 fills which
' have documentation of inspections by either Ebasco or the Contractor. Since Ebasce did a 100% duplicate l inspection of the contractors inspection, the fact that i contractor inspection reports are missing does not l
l l
1
, 1
--__.-.____..___,____,_..._____,__.__.-_.._______.___,_____,___.__--.__m--
,.______,m___m_.
- .-_ _ _ ~ - - . - . -
l REVIEW AND ANALYSIS OF S0IL l TACKFILL DENSIliES i NRC CONCERN NO. 7 necessarily lead to a loss in the documentation of quality. As stated before, the existing inspections on i
these fills clearly establish that the quality control process was implemented during the construction
, process. In addition, it should be noted that in accordance with the Quality Control procedures (Document 2 & 3), the in-place density tests performed on each c,f these fills were ordered by and directed by the Ebasco Q.C. Inspector. He witnesced and evaluated each field test for specification compliance while the test was ceing performed in the field. If the percent compaction was not in compliance with the ryecified
=inimag , the Ebasco QC Inspector directed the Contractor's QC Inspector to irplement rework (recompaction).- The rework was witnessed by the Ebasco
- Inspector and at its completion, ratests were taken at his direction. Thus, the existing inspection documentation, coupled with the complete file of test records for each fill involved (indicating acceptable i
relative density and quality control involvement) i indicate that this deficiency, which effects 2.0% of
- the backfill, will have no significance on the *
] stability of the Plant Island under the event of seismic loadings.
( !
1 (c) 0.2% of the volume of the backfill has no f aspection
- reports at the fill locations and elevations included
! in this volume.
Fer these 6 fill areas, there was no inspection
! documentation found onsite. The material in these fills is found to be concentsated below elevation -37 in ==a11 drainage ditches and trenches which have very
- little volume or in fills. As stated above, the l complete record of density testing testifies to the i
total involvement of the quality control inspectors and to the achievement of~th,e relative density. The fact that the majority of the nissing reports are clustered i together in groups on three fills indicates a high
, probability of lost foleers of soil packages. Thus,
) even if the records are lost, the acceptability of the relative density, the indication of Q.C. involvement, j and the fact that the affected fills account for for j
only 0.2% of the backfill placed provides sufficient evidence to conclude that this deficiency will have no
! significance on the stability of the Plant Island under the event of seismic loadings.
i Considering the discussions above, it has been concluded that the 1
deticiencies found to exist in the inspection documentation are of minor significancs and will have no effect on the structural 3
capability of the plant under seismic loads.
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REV/.*eD TABLE NO.3 PAGE I 0F Z
~
NRC CONCERN NO.7 . . :_ .
l ANALYSIS OF SOIL-INSPECTION REPORTS l __ SY FILL SURFACE AREA (FTa) ,
ELEVATION 1 2 5 FlLL NO.
4- 5 6 7 TOTAL SURFACE AREA (FT9 7. COVERAGE fW C00.^ DENTS
? - INSPFCTION imm DENSITYOWERAM GICTION1U0m5
-44o0 ~ -43.0I er e m a := A # '
N/A j j'I -4340 ~-42.01 m 4
- A W 4700 800 4800 -
N/V -
4 80 T ittous w o
}
-42.00 ~-4 8.01 +
- A 200 h 3500 100 3s00 --'_N/A -
a APPROPRIATE
, -4l.00 - -40.01 a R 25800 __.200 _ E 2450 100 28S50 WA_ - -
INSPEGTION msmw i -4040 ~ - 39.01 6800 A P5800 10800 a 5900 14 Goo 63900 55000 II& MIS $1HG.
, - 39.00 - -38.03 G800 - A 26IOC 300 - A 10050 48100 91950 G5000~ 143
-38.00~ - 57.01 300 A 30500 10300 500 2200 45600 89400 18600 I;.5
-37.00 ~-56.01 1550 300 27700 IJG06 - 82 5400 G0900 -
- -36.00 ~-3',.01 107300- 80000 - 134
' 16350 1700 84600 1000 1500 19t00 61500 185650 94000 i 19s
-3 5.00 ~ -3441 33000 1700 48000 400 500 8000 42300 158800 106000 It9
-34 00~-33.0I IG000 1700 29500 5000 1850 33700 61500 160550 99000 I6'Z.
l -3340 ~-52.0 I 2000 2300 t9000 5000 18000 53000 10000 859300 -
108000 14 8
-3240 ~ -31.0 8 16000 1700 50500 6450 17500 t1000 50500 151650 114000 133 1 -31.00 ~ -30 01 15000 IG100 .50000 .5000 .11500 4000 60500 166700' _._.331800 I 6500u les
-30.00~ -29.01 IG100 41000 2500 17500 68000 60500 '215200 14G000 141
) -2 9.00 ~-28.01 M000 25700 43000 5000 _35000 35000 51500 tll200 133000 ~ ~ - 15 0
-28.00 - -27.01 IG000 9000 77200 9700 35000 2l000 14500 16t400 -
158000 ll5 ' '
-2.7.00~ -2G.01 IG000 9C00 254000 2 5000 35000 5575 o 73500 smoteo 163000
~~- 84 0 ~
- 2 GD0 ~ -Z5.01 16000 9000 41500 9850 -l8000 11128 68500 186 578 -'-' 168000 ~ - ~ ~ liI
, ! -25.00 ~ -24.01 2l00 9900 5t000 5000 _69500 399t8 73500 ._ 250928
.3310007 - --- 159- -
i - 2AA0_~_ - 23.01 3000 2150 --95100 _70600 _ss250 70000 51000 ._ sess00 _ _ 1e5000- ~ -
_ 200-
! -2 3.00 ~ -22.01 4100 5600 .54000 . 41800 _33750 58000 57000 2590c 197300. 13 2 -- --
! ;2.2.00 ~-21.01 5000 5600 52500 41000 : 67500 34200 57004
'36ts00 t19800-J ' ~ 12 0 __ __ _ .
! 8
-2. l AO ~.-20.01 5000 .As00 68000 37500 101SS1 if300 57000 270100 . 258500 ~- -- l 15 "--
- -2 0.00 ~ -19.01 4W .3100 2S2500 T36300 JIbeJ 3000 57000 tem ._ _t41900--- ---
3700 525C3 _438GI _35150 92
- 19.00 ~ -16.01 2G00 39700 40000 -
tsSc50 165700 T-- _ _ . _ st . - -
i
- 1840 ~ - t 1.01 3700 3700 52500 36500 .37700 14900 58200 207?00 - 161500 -
79-
-I F.00 ~-lG.OI 7G00 3700 112300 41000 35100 11600 15000 SFy600 - - 175400 ' ~ ~~ M '
- Ire.00 ~ - 15.0 I eG00 3700 1It000 31000 :35700 39100 47000 271500
'! __.304100 - 98.
- 15.00 ~ -I4.01 2000 2600 96950 38100 _57000 12800 I4300 .323950 193500 I
76'-
- I4.00 ~ - 13.O I 4G500 10000 G9500 58000 ~50500 40000 47800 ~~ ' 385300 t98000 I s0G
, - 13.00 ~ - 12.0 I t1300 16500 G9500 47000 10000 60000 15700 250000 386500 I -~-~79 :
4 I
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- m eev 4 TAP >LE NO.5 NE2W NRC CONCERN NO. 7 j . _
ANALYSIS OF SOfL-INSPECTION REPORTS 1
i
,}
BY. FILL SURFACE AREA (FT?) .
4 '
ELEVATION FILL NO. loTAL SUN.ME #KA GTM 7,coem DC r"."wr5 _
i '
I 2 :"> 4- S 6 'l = C710D0 f8535 OEN5mroWERm5 A-mt Mh
- 12.00 ~ -f f .0 0 4500 Y -T-G 49500 56000 5^ -:--- Maan 55100 268700 I
.-l 1.00 ~ -10.01 3600 80006 77500 r.9 400 3s000 .64000 66000 465000 58 336700 369000 ei
. r 10.00 ~ -q.01 51000 1e000 aa m ige 00 37 = 104500 9te" _-._: 471300. _
- 9.00 ~ -8.01 32GGG __. 3t6500__. _ ___14 6 .
16000 118S00 19000 37500 94000 103aaa -- ~ ~ 492000 - --~~~ 3t7 000 ;
= 8.00 ~ ._ _. 150 l.01 19C+J 27000 809-- - 13s000 89600 . 63-:-:-J .62000 - 437500 . _ __< _ 3 5500.___
.554 _
2 7.00 ~ -G.01 35 - i 41C+- 86*aa 183100 14100 .18000 96 2 j lI 464800... .___ .. 332000 __ _ 14 0_.
f 5 6.00.^* -5.01 34000 ; -:-:-3 114- - 95000 ' ^ - -3 13 6 2 40500 - ' 505300 ~
!: _ 5.00 ~ -4.01 _7(+:-i 88500 slot 00 8s300 . e0e-- ess -:- 106600 .__ 2 539450___. 465500 : ett
.:._ _ _42 8 500.__ _ :~ 1 : Its
'! .4.00 ~ -3.01 61500 ~8500 73950 e05 :00 '805GG I40850 air.00 Zr emaaa. 4t7800; '
- 3.00 ~ - 2.01 72 4 .21500 89 -@ .94900 ilataa 163030 18100 . :_ 5ta sso _ .
-- .. ll9 ,
.; 2.00 ~ -f.Of 439500 130
?*-:-:-3 9500 94450 180500 IE-:- 98350 81000 1 .- 484900
} - f . 00 ~ -0.01 86393 80500 96000 6 9300 de:E - 389250 41000 444000~ 109 i 484360 dege00__ . 1_ __ 505 -
! _ 0.0 0 ~ 0.99 Iwm 44800 158500 808800 41i @ 825300 15900 - 673000 - 484G00 l
_ I .OO' ar 1.99 1318D0 ' Mena 138050 646000 ~47maa :17400 '45300 - -- 749250-~ 139 -
_ t.00 ~ 2.99 197100 Sf600 948650 148 100 4Gi., 12^'aa 7?*aa ~~ - " 0" ^ '*a - ~ --- ' a - -M - --- 155
. 3.00_~7_3.99 3 ga- -3 "84600 868150 858000 48000 881100 33 ~ _m 17t850 - ~~ 45G800 -
'lSt j _._d.0c_~_. 4.99 35000 10400 367800 e30800 56900 .80150 .46300 _."_St9050. 499800_: _ : : . lso
! > ~ 5.00 4 5 99 48000 ~24400 226900 ses00 69500 9S ISO 90100 1-~1 636250 ~ ~ 2 464500Z- 458000 tis l - 6.00 W 6.99 32900 i'a 289 -- 183600 88500 103*aa : ^-3 _ _ . .702300 ~ "-~ 451100. _ - 13 1-~-
{ , - - -
' --- ~ ' ISS " - -
} t _ ?.00.~ 1.94 36:93 S L-_- - 144 -- -- l i' a"' . 4 *: - 852550 148*^ ._ _ :::^^ 445100. _ _ISS.
_~ 8/10 ~ 4.99 4G000 9E5GG 842300 8045GG $6150 IL , r fl970J -- Z7adaaa' l
g l ~ an.00. ~ 9.99 806200 86000 841G00 143fauw :tt800 158350 76300 ~ 135650J 197t00._. Z: _111._
i ?
-- 8 0LOOW 40.99 et6800 .114 000 et200 It'aaa '16600 150"",0 104peo -- 368100~ ~ ~203---
i ? .
1 1.00_~ _ 81.19 3335GG 134 2-3 _96600 826'aa 29800 .191850 _85800 . ._ _791050..
" -- 783650 - - - 34t700:-- :Z--". 220 --~
__ 391100__.
~
,j _ _ s t00v't.99 isle 3 14t500 278900 6 ^-- - 1tt000 199900 193+^: ~ too f ; 665900--- 3 89 i e --- _208
- .13.00 +_13.99 219800 .346 250 261300 :~ R '.**"a -- - - - 95 ISO Z 71083650-~
j
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i 8 4.00 3C.14.99 .75000 74400 266800 J28400 r 4---- at450 .7c Z Z S" "30_. . . . _ 181 1:
404450 _ ____.303000~-- .__ _ _ _ 3 34 - _ .
l ZI$.00 Y15.99 e4000 -"1050 "3690D SC-~ 4 ^ S "^; 101450 ilt_ _rr.-.
l i 16.00 ~ 36.99 .S6150 .41500 1 a a'"a - 0' "*8 F -- : $3. IGG :? M- ---- tl 5MI--
3eawMD:- 2 27*{+ _
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PR5GUENOY MO D/3TRIBUTCM l
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-38.on~ - s7.01 7I, Boo 4 73 ' 7 G 7 4 D /G Si
-w ~ AG.or 80,000 4 AD' G $ G G D /2 7
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- -Moo ~ Mol loa. coo G /8- I / 5 / 4 3 3 ;
\ -M m-- Moi 86,000 E th O / Z / 2 b 4 1 l
-1300~ 1?os 10A0ao G /7 /. / 8 / % 4- S \
-h2Da~ bl.0/ //4,000 G /8 / 2 4- / Z Z G -
l -3 /.c o ~ A o o l /S/.800 7 2/ / O 4- / 5 6 4 l -10.oo~ -29.0/ 14G,066 8 Z/ ' O I D /
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[9. m ~ 2 7 0 / 158 000 8- /4 / / Z. / 9 4 %
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D /7 / / 3 / 2 4 5 l -Moo ~ 25,ot 168,060 8 /G / Z Z O O 7 4-1
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FR5iSUE/YCY NYC O/fRIOJ791Y FREQUENCY Ol&TRIBUT/Ot1 "
ELEVAT/OtY SUN %c8 tYO.Of i=STS F/M /YO /YO M O
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+/,00 ~ +/. 99 73 8 5 6 $ 10 // It i 484.000 25 72 8 + 20 D G // /4 .
"2.co~ +2. 99 4E6,.800 2S SS O b 26 0 8 20 la "2.m~ *b.DD d28.800 '2 2 80 7 5 14 5- //- 21 0 Mon ~+4.0D 458D00 ZD 82 -
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+10. DD~ +/0.DB b42,700 /B Ib5
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- 78/DZt D ll 72 ID // 71 27
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7 I5 959 6 4(, I8Almo) 0 82-465 9 *'
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26 15 9 FG 272 c> 98 SG6 7 .
27 22 C'* 59 22l 8 . loo 579 LI Inov 23 o^ FB 2 91 9 IC2 505 5 26 31 6~ 40 302 10 535 10 5 9 2g 42 lo - 61 312 8 106 6 05 6 l 30 47 Y G3 526 16 loS Sts &
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17 4 ll73 10. 191 15l2. 10 i 134 7 93 lo- I- 175 lifG 6 Eli 1524 lo ,_.'
11 5 R04 10 17 7 lis7 49 217 13 7 216 10 1539 to l
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9 1639 9 148 F91 Igg 11 o 5 -10 213 16 51 lo 149 904 10 -- Iso isti S 150 915 9- 2:4 I&fc8 to i
' 1 91 I ?> l2. O i 22G 1677 10 151 917 lo - i les laiS G
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153 953 1 10 - _ 19 3 1322. O T w o n.r 2.29 154 981 7'- 171z lo (9(,; in2. 9 ' 2 31 1724 10 15 5 ci73 9 1 197 1344 10 232 156 9 1735 9 :
CR3 19 8 1357 10 233 1747 to 157 #19 6 S i 199 1370 lo 2as 175g 9 l
152 i I007 10 - Sol la82 10 159 1o16 8 23/; 1724 C(ceerov ;,
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24l 1619 10 277 2194- 10 ;
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Sal 2586 7 246 Il865 10 225 2245 10 SSE 2526 10 243 1677 l0 287 2:56 9 814 2&lo l0 '
245 1869 16 288 2265 3 BEG 2Gza 10 -
249 l.90I \0 289 226 7 3 BS7 2G3s la
- ,, y 250 l912 \0 290
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251 922 S :
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257 l99l .0 l 289 233G 10 35\ 2750 10 i
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259 2Ol5 l0(cwnW4 30\ 2.564 9 354 1754 10 260 2026 l0CLNs4) Ba5 2.579 8 356 3767 IO .
2Gi 2027 0 ao7 2892 lo ..,_ _ 157 277S Ie -
- . 262. 2c38 10 5 09 FAOS IO 3 Ss. 2.79I lo />auunJ, 26E Soso 1 l0 ~ _ Sl A. 1418 l0 SSA L,92. O twov);
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82.3 295s 10 45z R%9 10 -
agg 2M7 11 4 53 3374 10 39I 2978 9 454 52S lO ,
292 2990 10 l 455 1994 IO 35>3 Roo2. 10 l 456 .N07 io 39G 30l+ 10 451 34tP, Ib 397 3oz7 10 45S 3436 11 4 00 3053 74(*!W's.d, _,
460 8443 l2 '
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sos 352.9 4 2275 4 32. 10 50 6 359Z l2-4s4- 31 % 10 so'7 K6o 10 439 3197 l0(>mtD sop, 3Gol IO
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TABLE NO. 7 MRO CMCEMY NO,7 -
~
l ACo'CONFOMPbht3 7D F/ELO MNSITY //fMRVALe PROCWl S t
t 1
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CURVE 7537 ho WNN /YO ELEV twxstyy 1 tgWE 699 F4 45 N ord 5B -416 l-21-7 6 78 70 0 F4 80Vl 285 58 -ss.G \ 9C 72 70i F4 44W 215 SB -408) 1 21 76 7/+
102. T4 A0W IGA 58 -39 0 l 2l 7G 84-703 F4 ~/0W 17 4 5B -35.3 l zi- 7 6 79 704 F4 2-3W los 58 - 37.~7 l-2176 ~l8
'705 F4 42W IoAl- 5B -36 4 121-7G 78 l 7C6 f4 'EW 55 SB -38 4 1-22-76 66 :
. 10 7 E2 6E I45 4 -36 4 122-76 6(,
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- 41.26 l 23 74 12 7l3 E2~ GW ES 4 -
41.45 l 13 7G 75 714 F4 45N e4W SB -4+.75 1-za .74 7z 715 ' F4 49W 2bl 56 -34 75 l#217C 82 716 F4 Glw 12 s 58 -34.75 17.3 7G Gg -_
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- 721 F4 45N .fgW SA - 41 75 1-26-% 8I 723 ,D?. ISW 13 5 4 - 30 50 l 2 6-74 73 l 72 4 F4 Gr.N 43W FA - 4050 12G14 79 '
725 ' F3 7N 45W ,
5A - 3 5 2.0 l 2.6 74 79 72G F4 40W 53N 5A-s - 3920 1-27 ~)G 77 727 EG 355 66E 38 - 42.50 ~ 1-E7 76 81 731 EG 78N 6gG 3B - 42.50 1-27-74 97 733 EG 225 30N 33 - 41 50 1 2K-76 7+
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76 0 34 2tS 33E 68 - 36,75 2-l2.-761 73 7K Bi- 7N 52E 2 ~8dNo 2-IE. 76 6s 76 2 B4 Ig5 56E GB -36.75 2.l2 76 68 763 B4 ICS 37E GB -3 5,50 2-12r76 64 ;
766 B4 295 55E 6B -35 50 2 -15 7 6 88 747 84 oki 50E 2. -85.50 2-JS -)G 84 sur
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W CURVE TEST ho WT70N MO "' "V mystyy 5 LRWE BIS F4 gow 8ss SA .-37.25 3 25-74 a) c+
9 19 55 2DN DE 88 -38.25 3 1-76 6+
Bro CG iss loN GA ~40.2.5 3-l 76 79 Bal E.5 2SN 325 1 38 -37Z5 3I76 57
_ B 2.E E5 304 305 38 -36 25 3-l74 G9 846 CG 40S 25'E 7 -38 25 E-17 96 71 966 C2 37S ISN I -39 7S 4 Z3-74 65 f 865 61 G85 42s t- -39 75 4. z2c76 5s 1 863 RET *EST 73 ~
BG7 C 2. (6S ISE \ - %.~15 423 76 da
.. 6'69 C2 IOS SE l - 38,75 42G c76 80 873 B2. SSS 40E I - 3g.75 4 -25,76 se
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-37.50 4 2776 5s "
876 32 40E 80S I - 37.50 42776 -
911 E3 son sN 4 - 4027 5 ijf-76 G9 i
912 EB 54N 20W 4 - 4025 5 IB 76 627
_ BIS E3 80N 30W 4 - 39 25 5 I&76 69 915 E3 SIN 80W 4 - 3700 5'-ls-7C 6G 917 E3 4cM glW 4 ,3700 519-1G 91 1
918 E3 255 roc 4 - 37.25 52074 30 9 19 E3 2.75 48W 4 ~ S7.25 52076 70 920 2 DB, 899 30E 4 - 35.25 5 20-76 GI
- 92I Es ins Eda Sh - 3G E 5 20 76 sq 922 E3 25E 2aN ' Sh - -36,25 520% Go il l _
92? ES 605 lsN Sk -35.25 s ii 76 Se l 924 E3 2SE 15N SA -3f.25 52176 94 l 1
%52. CG 350 30W GA -3S.25 G E< 76 555 931 CG 47W 30N 6A -39 25 6276 74
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6274 Go
- 934 C6 OM 45W Gk 17 2S 6376 61
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.- 2B8 h 86 90M GSG le -39 50 lFG.16 Iil,I.5 l 2-39 A BG 93N .gsE G -4l.25 11G.76 se.5 24DA BG 95N G45 G - 8900 \\ -G.74 (51.5 1 24lARR M 55 5 85 G - 4025 II -67& 140.5 242 A BG 98N MG 6 - 3775 ll-G 7G Izq..o 243 A BG 914 44E 6 - 37. E5 ll- 7 7G 112.5
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TA 8! P N O &. w w7 c N m co/ycr_7 w Nd 7 hMALYSIS OFh0MCOMMINQ 'cch&&L 72?CT'Mulia%sirncm> ;
P%iA LAO F/LL TEST
%W MELA7N CURVE TEST no MTION MO ELEV Metyy 96 SO477 A E7 BON 966 S - 27.30 3 l6 77 88.5 478A E7 bon ?SE 7 _- 30.20 31677 6g,5 4 79A F7 48N 0E 3 - 2G.30 31777 65 0 480A E6 20M 555 3 -2620 31777 79.o ,
48lA E7 l5S ISE 7 - 29:30 2 I7 17 77 0 \
482.A ER Sln 225 7 - ?g.30 31777 72.0 l 483A D3 70N Can ? - 32 20 3-17 77 67 5 484h D7 305 12 5 7C
- 32 80 2-18 72 S T.O .
4gsA m ON 55 7A - 4.30 3 1577 925 42Gb E6 800 SEE 3S - 25 20 8 l$'7] 67.0 487A E7 SSN 435 7A - 25.20 3 -lE-71 79 0 419 A D7 455 24E >c - 2l fo 219,77- 66 5 ,
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TA3LE NO. 9-NRC CCNCIES NC. 7 SCHD U7.I 07 RELA!.77I DINSIri CO?J.ILATiCN IISTING TIST S*UM3IR
- TIST DATE L2%2 815 2/25/76 1040 3 0023A 8/12/76 50A 9/9/76 78A 9/22/76 256A 10/8/76 271A 11/9/76 377A L2/15/76 444A 2/2/77 532A 2/23/77 621A 4/1/7" 4 / 22 /','7 255A ICS7A ~5/31/77 7/7/77 g#g 1321A '
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BACKFILL DENSITIES
~NRC CONCERN NO. 7 APPENDIX A IN-PLACE DENSITY TESTS FILL 5 Ei.,-41.75 to EL -36.25 e'
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- IN-PLACE DENSITY TESTS - FILL d5 EL -41.75 TO EL -36.25 TEST TEST TEST - TEST PROCTOR TEST l ' EVALUATION LOCATION- NUMBER DATE ' CURVE No.
-41.75 F4'45N 38W 'LRWE721 ;1/26/76 'l
-41.60 F4.ON 45W LRWE699 1/21/76 1
-40.50 F4 62N 43W LRWE724 1/26/76 1
-40.30 F4 215 44W , LRWE701- 1/21/76 1-
-39.60 F4 285 80W LRWE700 1/21/76 1
-39.25 F4 20N 80W LRWE808 2/24/76 6
-39.25 ' F4 18N 20W LRWE807 -2/24/76 6
-39.20 ~ F4 53N 40W LRWE726 1/27/76 1 4 -39.20 F3 7N 43W LRWE725 1/26/76 1
-39.00 E3 30N 33E LRW1031 8/12/76 15/18
-39.00 F4 16N 40W LRWE702 1/21/76 1
-38.75 E4 10N 33E LRW1036 8/12/76 15/18
-38.30 F4 17N 70W LRWE703 1/21/76 1
-38.25
- F4 30N 50W LRWE811 .2/25/76 3 l
-38.25 FA 35N 43W LRWE812- 2/25/76 3
. -38.25 E4 10N 31E LRW1037 8/12/76 15/18
-38.00- E3 31N 32E LRW1033 8/12/76 15/18
-37.75 E3 31N 34E LRW1035- 8/12/76 15/18 h -37.70
-37.50 F4 105 43W E4 11N 32E LRWE704 LRW1038 1/21/76 8/12/76 1
15/18 4
-37.50 E4 69N 27E B0102A 10/13/76 34
-37.25 F3 80S 70W LRWE 813 2/26/76 $
, ., -37.25 F4 805 84W LRWE 816 2/26/76 6
. -37.25 F4 85S 80W LRWE 818 2/26/76 6
-37.25 ES 40N 27E B0089A 10/11/76 34
-37.00 E4 60N 27E B0101AR9 10/14/76 34/36
-36.76 E4 6CN 28E' B0110AR4 10/14/76 36
-36.75 ES _42N.1IE 30090AR2 10/12/76 34
- -36.40 F4 ISS 78W LAWE706 1/22/76 .1
-36.40 F4 10N 42W LRWE705 1/22/76 1
-36.25 E4 45N 27E B0116AR 10/15/76 36
-36.25 E3 24N ISE LRWE922 5/20/76 7
-36.25 E3 24r 37E LRW4921 5/20/76 2 g
-36.25 ES 38N 27E B0097AR 10/12/76 34 NOTE. Actual In-Place Density Test sheets are available at the Waterford 3 Site 4
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RESPONSE
ITEM NO. : 19 (Revision 1)'
Water in Basemat Instrumentation Conduit
~
TITLE:
'NRC DESCRIPTION OF.CONCEPN:
la examining the safety. significance of the allegations, the NRC staff performed
~
system walkdowns asja;means of verifying the as-built conditions. During one of those walkdowns, the staff.noted that'tbere was water-infan electrical conduit that penetrated:the basemat.- If the seals in that conduit should fail there'is a potential' direct. path for ground water.to flood the auxiliary building basement. - LP&L should review all conduit that penetrates the basemat and
. terminates above:the top of the basemat to assure that these potential direct
! access paths cf.watar are -properly sealed.
DISCUSSION:
During the construction period, several permanent conduits embedded in the basemat were observed to seep water at the stub ~-up couplings. None of them leaked in a quantity sufficient to cause flooding concerns during construction.
Silicone foam seals were placed in these coaduits beginning in late 1983.
, In May, 1984, a walkdown, as' described in Attachment 1, was performed by Ebasco j which identified 29 places where wetness due'to seepage from conduits or 1 conduits within 9 boxes plus one piezometer riser were found and 12 places where evidence of past leaking from conduits and piezometer risers were found. These l cases will be addressed by LP&L by removing the existing seals and replacing them with a light density silicone elastomer which hcs the capability to stop the seepage as required. This work will be performed as a routine maintenance item ar directed by the Plant Operations Staff, since the slow seepage through
- the seals is a maintenance inconvenience and not a flooding hazard. This is l reflected in Attachment 1.
i The 12 sheet table that is part of Attachment 1 is in fact 2 related listings.
- The first 2 sheets list 36 iters (27 conduite including one piezometer riser and 9 pull boxes). These items were checked off in the listing as either having a
~
i leak or giving evidence of once having a leak. The remaining 10 sheets detail I what conduits come into each of the 9 pull boxes listed on the first 2 sheets (Items 4.5,7,9,10.23,27,28 and 32). These 10 sheets have listed on them 56 i conduits (within pull boxes) which when combined with the 27 conduits (not in
! pull-boxes) on-the first 2 sheets makes a total of 83 identified conduits.
I
_(Note: Attachment 1/Paragrcph I indicates that 8 pull boxes were identified.
Subsequent to issuance of Attachment 1, additional condaits and one pull box
[
were added to the table. The first sentence of Attachment 1/ Paragraph 1 requires correction. The first walkdown resulting in the memo consisted of an inventory of-individual conduits which had seepage or evidence of past seepage and pull boxes containing numerous conduits which had a potential for seepage or evidence of-past seepage. Subsequent to the first walkdown, the covers were removed from the pull boxes to identify individual conduits within the pull boxes with seepage or evidence of past seepage. This reduced the totals
. reflected.in the sentence and provides the actual numbers of conduits with evidence of current or past seepage as shown in the tables.)
Ic+1
~
. Temporary conduits whichLonter th'e basemat' from outside,:and'which once allowed passage of ground water in quantities that required periodic. pumping, have now.
all been pressura grouted as part of the normal design requirement and their temporary blockout pits filled with concrete as shown on Drawing LOU-1564-G-499 S09. Therefore, they no longer serve .as ' leak paths ' for ground water.
Attachment 2 discusses the sealing of a piezometer. riser and a piezometer standpipe.. The piezometer riser (Item 8 of Attachment 1) consists of pietcmeters in a conduit down in.the aquifer (surrounded by a well pipe). ~The conduit was internally sealed behind the piezometers and was sealed again.in the portion of conduit that transverses-the basemat.- As recommended in Attachment 2, this conduit will be sealed with a light density silicone elastomer since two L of-the piezometers are still operable. The piezometer. standpipe is basically a-
~
, well pipe filled with water under pressure from the aquifer with piezometers
! attached at the -35 level. .This standpipe has been pressure grouted. The location of the riser is just south of ' the J wall, between SA and 6A (i.e. , in
) corridor south of EFW pump A - see FSAR Figure 1.2-11). The location of the standpipe is north of the L wall .between 6A and 7A (i.e., in the radioactive _-
pipe chase - see FSAR Figure 1.2-19).
1 CAUSE:
Except in the case of the piezometer riser, the seal material in place does not provide total waterstop characteristics.
GENERIC IMPLICATIONS: ,
There are no generic implications since the potential paths for ground water to flow in appreciable quantities had already been addressed.
SAFETY SIGNIFICANCE:
J There was never a path for ground water to flow in sufficient quantity to flood the auxiliary building basement, even before the seals were installed and before
- i. the temporary conduits were grouted. The floor drain and sump pump system vas j more than adequate to handle the quantJty of water which entered the building i during construction, and is adequate to handle the much reduced quantity presently observed, most of which evaporates before ever reaching a floor drain.
On this basis, there is no recognized reason that this issue should constrain j fuel load or pouer operation.
L CORRECTIVE ACTION PLAN / SCHEDULE:
As stated above, there is no safety significance associated with this issue.
- Corrective action will be taken as part of good construction practice. The j decision to replace the seals on the conduits will be based strictly on operating and maintenance considerations. Any replacement seals will consist of a light density silicone elastomer which has the capability to stop the seepage.
- l.TTACHMENTb :
(1) Memorandum ES-9160-84 of May 18,1984 (2) Memorandum ES-9409-84 of June 1, 1984 I 19-2 l
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REFERENCES:
(1) Drawing LOU-1564-G-499 so9 (2) FSAR Figure 1.2-11 (3) FSAR Figure 1.2-19 l
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May 18, legt ,
. IS-9160-84 . .
Io: J.' Ieugh:aling
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1 Subjee:: LOUISIANA ?CWIR & LIGII CJ2 ANT ,
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- f-WA4:.x.SII?AGI TROM CONDUITS,- j.
ELIVATION -35 l
I acec:da=:e vi:h your reques:, Civil and Ifee: ical ISSI cenducted a valkdo a c of :he :::dui:s v'..ich pene.:: ate :he =a: a: Elevatio -35 of the I.A.3, . .s a:4 Cooli=g : vers to da:er=ine which cc:dui:s are leaking varer. A: he sa=e time NI!I was 'recuested :o reviev the type of =aterial : hat teuld be e= ployed no sea.1 .
. :he :: dui:s'and elirica:e seepage of water on,:e the ficer. -
r Er* resui:s of :h:.s s:udy are as fc11 vs:
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/>Esc.h Iesults of Walkdev: .
A :::prehensive valkdev. cf t cenduits which penetra:e the "a at -[
Ilevaric: -23 revealed 41ther seepage af va:e: c: eviden:a :har va:er has '
leaked i::= 76 of :hese cenduits. The at: ached table prevides a ===ple:e [
11s:1 g of .he affe :ed :::duits including their Ice::1c a=d cables '
. ec::ained. A lar;;e nu=ber of :bese c nduits (53) penetrate the Man and ;
e:: fice =eunted pull b:xes. There are eight such pull bezes tha: hai e :
b e er. id en:1.fied . {,
II.' Iesults of NIII 5::dy WII was requestad to reviev this p :ble= and identify :he :.pe of fix :ha -
veuld preven: vs.:e: f::= pene::t-ing these cc dui:s. 1: vas deter =i=ed i
-ha: seali=g :he :: duits vi:h Ligh: De:si y Silicene Ilas::=e: (LDSI)
- 9.ich has been provided by 3&3 is ace : dance -1:h existing specifica:1c:
ICU 1364.249W vill prevent the senpage of va e th cugh the conduit's. '
Assuming :he va:e table :c be equal to grade eleva:ics of +17.5 fee: and i
- he affected ccedui:s and flush vi:h slab elevc: ion -35 feet (verse case), ~ -" !
- ba prassure ::,:ep .
cf the cenduit c;ening ean be ccicula:ed as fc11cvs: .
essure .(?S~) - Iead (f:.)/2.31(!:/ psi) where Isad (f .) = d{ -d y, :herefere j
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(.5.)e =: 1,e.5-( .5)/2. A. c: 22..epsi
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A f :: (4) inch .:hickness of LDSI has bee: :es:ed by 3&3 :o be a fire ra:ed seaJ.
20 psi
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Is-9160-84
~( Jince the pressure on the conduit is 22.7 psi, it is recemmended' t' hat a six
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inch thickness of DSI in each conduit and vill provide a margin for flood 1=g at g: de; elevation. . The existing Silicone Team fire barrier material must be y ec=pletal :te=cved prior to pouring the LDSI. Also, upon .ur1=g the LDSI boe==es hard and can culy be're=cved by using a chisel. -'
It should be :oted that. the seepage ci vater onto the floor of Elevation -35 through these conduit .is not an i= mediate hazard- to the safety of the plant or '
. its person =el, but rather a nuisance to maintenance. On this basis, it is ;
racc== ended that replace =ent of the Silicone foam fire barrier =aterM vi-h the .
LDSI be scheduled as a pos fuel load task at a time conva"4="t to LP C b e
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SC20ZOT: LOU'"5.".'.A'a PC*nn & LIGET COMPANY . *
*n A'"I?JCRD SIS - CN;T No. 3 l'
WA"I?. SEI?AGI ??.CM ?II::OMETI?.S IN 3ASI MAT .., f
.s P.efore .::: .ta.=o--T5-9160-34 cf 5-13-04, Grille to Eoeghtalbg, *WATIR SIIP. AGE
.vv. w.. w...~..5, e.r.-..c.! va . . 3.c.
- a= 3 cf the attachment to referenced ===o ide . ifies a leaki=g 4"4 :* see f== l
;in::=eters, e.s she. . ca iravi .7 *.0U-1564-C-4995':9, Detail "X".
j J.c cf de pie ===e:ars in his riser are still cperating and they should ===ti.ue is servi =e se long.es $ay give-reading =. A;plicarice,cf Lasz sealis; foam as
- j' Tc === ended in de =e= vill not interfere vid c:sti=ued sa vica. '
h4:o is also a pie:c=etar standpipe', No'. 7-23, which 'is .shown es the sa=e desving, kJ.j2vhichwa.sn::
. d:ss he: leak. 6:1uded in the referenced =e=o because it is set a. c==d..is and
'- i s Ohip s.a .dpipe a.lse' rereiras sealhg even thcugh it' does not laak at ,- resent, since ility is ever (it =enit=:s a deep arcifer cf ao present i=:arast) a:d in its
;:ese.n: ::nfigura.:10: (full ef water under .:essere) represents a ;ctantial leak pa2. if it were :: .he b:cken by accident c ::::osien. .
' .* : should he sealed by hjectie: ef pressure 7:=ut h suffic:.e== ccantity to fe t t
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- is suggested
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RESPONSE
ITEM NO: 20 (Revision 1)
. TITLE: Construction Materials Testing (CMT) Personnel Qualification Records NRC DESCRIPTION OF CONCERN:
The Inquiry Team ' ef fort included a review of the disposition of the generic
-problem identified . . during the LPLL Task Force verification relative to GEO Construction Testing (GEO) documentation for personnel qualifications in the area of CMT.
The utility should conduct a review of supporting documentation ' for. GEO - corrective action sts:ed in Attachment 6 of NCR W3-F7-116 - (Ebasco W3-6487). This review should focus on the identification of CMT personnel placed in GEO Categories 1, 2, or 3_who were apparently qualified solely on written statements by other individuals attesting to the individuals training and qualificutions. For such individuals, the applicant should pursue- any new -information or evaluations which could provide further assurance in support of the actual past work experience and training referenced by the written statements. DISCUSSION: As requested by the staff, LP&L has pursued and obtained additional information on the GEO individuals performing inspections and tests as will be explained in the sections of this response entitled " Collection and Verification of Personnel Data" and " Disposition of Deficiencies". Also, evaluations have been made of work performed by GEO personnel a' briefly outlined herein. A verification program was implemented to review the professional credentials of e 100% of the site QA/QC perscnnel who may have performed safety-related functions at Waterford 3, including supervisors, managers and remaining QA/QC personnel. Assessment of the qualifications of all GEO Construction Material Testing (CMT) personnel, including those identified in Attachment 6 of Ebasco NCR W3-6497 (the NRC reference to Ebasco NCR W3-6487 is apparently a typographical error), was a part of that verification program. The responses to Issues No. I and 10 discuss inspector qualifications for other Waterford 3 contractor personnel. The program, which is beinh Performed under the overall direction of LP&L, consists of three major elements: o Collection and verification of personnel data. . o Evaluation of qualifications against specif ed standcrt's. I o Dispositioning of deficiencies resulting from cases where inspections, tests or data collection were conducted by personnel whose qualifications against the appropriate standards 'could not be confirmed. 1 20-1
"h,.,
r Q'
-Collection and Verification of Personnel Data
- Personne1~ data were ' collected from various ; sources, including site files,-
; contracter Lhome office files, personal 1 contact:nith individuais or supervisors
-and a thorough background verification program.
~
Ef forts 1 were made to verify ~ ? the education: and work experience < of 100% of the GEO-CMT- QA/QCf personnel by researching ' Wa'terford. 3 j GEO-CMT records - L and H by contacting schools, former employers and others' -While the success rate of the1 background verification? effort for . GEO-CMT was. hood. there were cases where - confirmatory-information was not obtainable. 'In such cases, the judgement of
- the LP&L Review Board, as described . below, wasL used to rule on the reliability of:the.available information.-
Evaluation of'Qualif'ications to Specified Standards
' QA/QC ~ personnel -data were evaluated in order to classify individuals as either.
t having verified qualifications or not. -Training, education and work experience were the qualifications-of primary concern. These qualifications were verified
- against.the~following criteria
i (1) Inspectors - ANSI N45.2.6-1973 l I- (2) ;0ther QA/QC Personnel - QA Program requirements i Initial qualification determinations for GEO-CMT personnel were. performed first
- by.Ebasco and then separately by an LP&L review group. In order to control the consistency of. these determinations, approved procedures were utilized. -
Determinations related primarily to balancing education, experien*ce and j . training factors. The LP&L reviev ' group qualification determinations were rendered in two j categories: " qualified" and "potentially not- qualified".- "Potentially not ). qualified" determinations were referred to an LP&L Review Board comprised of
- j. senior LP&L QA personnel. The Review Board determinations were further reviewed i by a consultant very familiar with inspector qualification ~ and related l standards. This process resulted in a final determination for all QA/QC i . personnel as either " qualified", or " unqualified".
i.
- The qualification review process is described in QASP 19.12 and QAI-32. The following points further clarify the process
- 1. The meaning of the term " unqualified" must be amplified. In some cases determinations were made that, based- on verified data, individuals' backgrounds did not warrant qualification to ANSI N45.2.6-1973. In other cases, however, individuals were considered
" unqualified" as an expedient in reaching resolution to tLe concern.
l- This occurred in~ cases in which: ? ! a. Research of rc:ords, inquiries to past employers and employees l c utact with schools and verification of training received was j either not possible or co'ild not be concluded in a reasonable period of time. 20-2
-- ~ _ _ _ . _ . _ _
b, ? Apparent discrepancies' existed between background information br ~ 'provided byfsome: individuals.and that obtained in the 6 verification process', an'd resolution could not' be ' achieved on a
- timely / basis.- Minor .- discrepancies L were excused;- however, significant; discrepancies _ generally rendered any. ~ other significant but unverified data'as suspect.
2 .- In the process _used,1being judged as " unqualified" to ANSI: , N45.2.6-19737 did not Lautomatically render the individual's - work as - invalid.- For_ ' example, E an individual . may not have the education and I ' experience qualifications _for-.all inspection work, yet--be -fully competent through specific training to perform the particular : tasks
! assigned to - him. : which might -have been very simple .and; repetitive in .
nature.' Such an individual potentially _ satisfies ANSI requirements - which. ultimatelyl require l that: an individual's ' qualifications. be sufficient to provide reasonable' assurance that the-individual can.- competently-perform.a particular task. . Whether~or not the' individual. is technically qualified,1the individual's work can-be deemed valid.-
' 3.-
During the construction period, ' GEO made- undocumented judgements with < respectf to the ,. need. for eye examinations for- inspection personnel. Such ' judgerants were based o'n the '. level of visual acuity or color
~
.-perception required to achieve competent inspections. Such _udgements
~
vere also made ' as part of . the verification program and uisposition
- process and will be. documented. 'It is noted that such judgements. are specifically suggested 'in ANSI N45.2.6-1978. This iactor was not deemed disqualifying.
^
- 4. Some individuals were classified as inspectors but performed no safety-related inspections' and were otherwise not involved in quality related work. To the extent such individuals were identified, they. were I
excluded from the overall inspector population. Disposition of Deficiencies For - those - individuals - found " unqualified" .the LP&L review board initiated i Corrective Action Request (CAR) EQA84-21S1 to formally disposition the identified- deficiencies. Ebasco NCR-W3-6497 has been reopened and when 4 reclosed, will reflect the disposition of that CAR . i Disposition of CAR EQA84-21S1 was accomplished by 3 methods as follows:
- 1) Assessment of Key CMT tests and of skills required to perform these tests.
The key tests were as follows: ', a) Concrete - The most important t;st is the final cylinder break test as this test serves to confirm the strength of the concrete actually , placed in the' structure. Other tents on concrete are generally either performed as measures to avoid subsequent replacement of sub-l specification concrete or were performed in collecting the concrete
'for and preparing of the test cylinders. The break test requires ll minimal skill in setting up and starting a compression device which
! compresses a' pre-molded cylinder to failure. A large gauge records
- .the force required which is easily translated into the data recuired.
i. i 20-3 s
, --- .,.-a,m,e r e< ., , - -er--,m-+.,-y--,+. -s- - - - - +------~,-,w, ~r- - , . ,-r-e-3 w. ---w.- ~ww. 3-,-- ---+- -
s Further confidence in the quality of the as-built material is provided by the fact that improper operator action would tend-to degrade test results, i.e. , improper testing would cause the concrete to appear less strong than it actually is.
~
b) Soils - The most important test is the field-density test as it measures whether the backff11 material has been compacted to specific requirements. The field portion of the work, which was performed by the technician, consisted of digging a small hole and placing the removed.' soil in an airtight container, positioning a rubber balloon apparatus over the hole, inflating the balloon to a pre' te rmined .- pressure and reading a volume indicator scale. Further, confidence in the quc,lity. of the as-built material' is provided by the quantity "o f tests. conducted. As .. stated in the engineering report supporting the response to issue 7, to insure control of backfill placement approximately three times as many field density tests were conducted as required by the technical specifications. c) Cadwelds - There was only one type of test on cadwelds conducted by l GEO-CMT and that was the break test. This test is as simple as the concrete break test. The test specimens are secured in a tension device, tension is applied and the failure strength is read from a gauge and recorded. I It has been determined that only minimal training would be required i for an unskilled individual to become profic'ient in performing the above cests. A single demonstrition coupled with minimal practice under proper-supervision is sufficient. GEO has formally confirmed that " Prior to being assigned to production work, all personnel were trained to perform the work ( required." On the basis of the above, though not strictly qualified to ANSI N45.2.6-1973, individuals could be considered competent to perform the technician or data collection type functions described.
- 2) Quality of Testins, Performed by Personnel in Question A detailed analysis was conducted of inspection / testing performed by a large sample of Level I personnel in question. This sample is felt to l include the most significant exposure in terms of potential for inferior l inspection / testing. Level II and III personnel either performing or directly supervising the performance of the tests dtscribed above should be competent to perform such functions.
1 20-4
- 3) Engineering Evaluation A statistical analysis was conducted, using industry standard techniques, to evaluate test results for concrete and the class A backfill (Reference 3). In the case of concrete both the overall and within-test coefficients of variation demonstrated excellent control of the product which would not be the case had the tests not been well conducted. Backfill test results also demonstrate good consistency. A review of cadweld data and test results described in Issue 11 indicates reliability of the test data and confirms the adequacy of the cadweld testing. This evaluation verifies the overall adequacy of the work of all levels, Levels (I, II and III) of GE0-CMT QC personnel.
CAUSE: Implementation of ANSI N45.2.6-1973 allows substitution for education and experience levels by noting that "... education and experience requirements specified for the various levels should not be treated as absolute when other factors previde reasonable aasurance that a person can competently perform a particular task." GEO and its predecessor organizations issued certifications of qualifications for testing personnel under successive programs which employed such substitutions and which became more detailed and better documented with time. The program in place since 1978 generally parallels the ANSI Standard for inspector certification. However, the verification program revealed that verification of background data was not adequate or documented, documentation of the justification for substitution of other factors for the requisite degree of training, education or experience was sometimes not provided, lacked depth, was not totally in accord with contractor procedures or the ANSI standard, as currently interpreted.. GENERIC 1MPLICATIONS: This issue has been treated generically. The scope of the verification program included 100% of the QA/QC personnel of all site contractors who may have performed safety-related work, including GEO CMT personnel. With regard to future work, qualification and certification of inspectors (including NDE personnel) will be administered through strict compliance with LP&L Nuclear Operations Procedures which meet the requirements of Regulatory Guide 1.58 Rev. 1 (ANSI N45.2.6-1978) and SNT-TC-1A-1975, as applicable. SAFETY SIGNIFICANCE: The results of the verification program and evaluation of the work performed by
" unqualified" GEO CMT personnel provides reasonable assurance that the related installations will perform satisf actorily in service. There is no recognized reason that this issue should constrain fuel load or power operation.
20-5
CORRECTIVE ACTION PLAN / SCHEDULE: On the basis of-Reference 3, CAR EQAS4-21S1 has been dispositioned.
REFERENCES:
1.; QASP 19.12 Review of Contractor QA/QC Personnel Qualification Verification
- 2. QAI-32. Instructions for Verification of QA/QC Personnel Qualifications
- 3. Engineering Evaluation of Report on the Review and Analysis of the work of GEO - Construction Material Testing.
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' - i S'L s ITEM: : COLLECTIVE SIGNIFICANCE ~ (REVISION -1)
PURPOSE:_'
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In response to the twenty-three issues identified in the NRC letter of -June._l'3,-_ i 1984, ,LP&L : has provided the ' NRC 1with a program - plan describing the L ongoing - '
~ activities to resolve the NRC's concerns. The twenty-three' responses 1 developed
. in accordance with -that program plan have addre ssed the specific. NRC concerns.
- As'part of tnat effort, the findings,of each' issue were evaluated to determine
.the "cause" and " generic implications". That evaluation process was conducted
.in- a' manner that allowed commonalities between ' the = various issues' to be- -
considered and factored into the generic . implications'of f one or more issues,:
. where appropriate.
1 The. purpose of this assessment of collective significance . is to evaluate the -
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overall ; significance of ;the findings from the . twenty-three evaluations to-
. achieve.the following objectives:
Identify and. assess the significance to safety and to the construction _ program - of the . findings f rom the evaluations of .the twenty-three issucc. , t Identify actions that could have prevented occurrence of~ the
- twenty-three issues and thereby identify the lessons learned which, if-implemented, would provide reasonable assurance that such deficiencies ,
;_ would be precluded from occurrinE in the future.
1-phase Quality Assurance Program to.
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3 Review the' LP&L operational l ' determine whrther the lessons learned are reflected in the Program or whether additional modifications to the Program are warranted. . t
. The conclusions that have been reached in this assessment of collective l significance are discussed in the following sections. The principal conclusions are as follows:
4 ! In response to Issue 23 "QA Program Breakdown Between Ebasco and j Mercury". LP&L committed to further address areas needing improvement t -in the QA program in this assessment of the collective significance of , the 23 issues. Having completed the assessment, and in consideratien ! of ' problems . related to Mercury in 'many of the other issues, it is j apparent that programmatically the corrective action was not i sufficiently thorough. Thus the partial breakdown acknowledged in i 1982 with respect to Mercury was not totally corrected. However,
-overall site performat v.s improved, particularly with respect to the
} quality of installed hardware, and there was no escalation into ar. 1- overall breakdown of the QA program. i 1 i , l 1 j-I' [ i i
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-The 23 issues have'been thoroughly analyzed.- The process las involved more 6an 1000 man-months of effort, exclusive of over 100 man-months _.
expenced by the NUS Task Force Support Group. The results, reflecting
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the general quality of the ~ QA = program and of the construction work itself, provide a high degree of confidence that _ the structures, systems . ani components as constructed are adequate to protect - the public heal;.n and safety during operation.~ .Only very limited hardware
. . rework has been undertaken as a result of the twenty-three concerns, and in several cases this revork has been discretionary.
L The lessons learned from .the twenty-three : concerns provide 'a reasonable basis. to determine whether the operational' phase . of th e -
. Quality Assurance Program adequately addresses the problems which occurred during construction.
The assessment of the operational phase Quality Assurance Program has. provided reasonable assurance that the program is adequate to preclude similar problems. This process, though extensive, clearly has been valuable to LP&L. The process has identified areas for improvement in the LP&L QA program and has reconfirmed the safety of the as-built plant. This discussion of collective significance is divided into the following three parts:
- 1. Assessment of Construction Program and Safety Significance
- 2. Identification of Lessons Learned
- 3. Operational Phase QA Program Assessment ASSESSMENT OF CONSTRUCTION PROGRAM AND SAFETY SIGNIFICANCE To assess the safety significance of th6 23 issues to the as-built plant, the issues have been categorized according to the effort needed to resolve the concern (See Table 1). Four categories have been created as follows:
Mercury: Those issues involving resolution of work within the scope of Mercury's effort. With the exception of Issue 23, all are also discussed in the following three categories. Software: Those issues involving records reviews or limited action i such as clarification / correlation of records, engineering evaluation, record analysis, or procedural changes. Inspection / Evaluation: Those issues involving reinspections and engineering evaluations for resolution. Hardware: Those issues involving physical rework to address the findings. The significance to the construction program in term of whether u.knesses have . been corrected and the nature of the weakness is treated on' a case by case basis.
L ,
- 1. Mercury Work:
Ten of the 23 iseues dealt in varying degrees of specificity with the Mercury program. Issue 23 "QA Program Breakdown between Ebasco and Mercury" dealt expressly with the o ffectiveness of the corrective action program undertaken by LP&L as a result of the problems identified in the Mercury program in 1982. Additional questions as to the effectiveness of the QA review of Mercury work are included in the following NRC concerns: Issue Title 1 Inspection Personnel Issues 2 Missing N1 Instrument Line Documentation 3 :nstrumentation Expansion Loop Separation 4 Lower Tier Correctise Actions 6 Dispositioning of Nonconformance & Discrepancy Reports 13 Missing NCRs 14 J.A. Jones Speed Letters and EIRs 17 QC Verification of Expansior Anchor Characteristics
2 Welder Qualifications (Mercury) & Filler Material Control (Site Wide)
Analysis of these concerns shows (a) improvement in, but continuing problems with, the control of Mercury efforts during construction, and (b) ultimate success in assuring the adequacy of the work within the Mercury scope. Improvements in the control of Mercury work are detailed in response to Issue 23. These include a June 1982 LP&L order for Mercury to cease safety related installations until there had been extensive Mercury organizational changes, additional staffing to addresa quality inspections / reviews, training to provide the guidance / direction needed for quality results, and the establishment of an Ebasco Management team to provide support and management oversight of the Mercury program. Subsequent improvements in control over Mercury included both ongoing administrative and qulity program changes, and gradual reductions in the Mercury scope until a full demobilization by November 1983. A review of the post June 1982 work demonstrated a significa.nt improvement in both the quality of installations and the quality of documentation. Notwithstanding improvements in the Mercury program, problems continued. Most importantly, generic implications of identified problems were not sufficiently addressed. Had they been, many of the problems identified by the NRC would have been identified by LP&L. For example, a significant number of QC inspectors hired by Mercury as part of the 1982 corrective action were apparently not sufficiently qualified to ANSI N45.2.6-1973, and this was not discovered in the QA process. As an indication of the ongoing problem, Mercury did not process NCR-888 to address concerns that QC personnel were not properly qualified. This action could have then resulted in a more effective corrective action to address the Mercury concerns as well as early identification of the issues frand in Issues 1,
- l 10 and 20. -
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7 While there were~ continuing problems ~with' control of Mercury, the'as-built-
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' condition; of Mercury work,. as determined by LP&L,;is adequate to assure the
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public hea* th ' and1 satsty. :This _. is demonstrated by. reverification . and; s, , testing activities both'as;a part'of the Mercury corrective action program established in ' 1982 L and as' a 'part - of - the' responses l to the twenty three Eissues. -The reverification activities-' encompass _all types e of Mercury safety-related work. '(See .Resoonses to Issue and Issue 23)- As shown in
~
- :he response to Issue' 1, an extensival reinspection of all M1. instrument.
lines ,resulted in'~ a .small amount of rework, 6 ast of .which was elective and none of which was significact to safety. , 2.- Software:
"The resolution of six ot' the twenty-three identified l issues ' was achieved
~
through actions limited to such tasks as. reconciliation / - correlation of
-- records, records' analysis, . records - reviews, . statistical : analysis, engineering analyses,.etc. ' Collectively, the evaluations of these concerns indicate that the past actions to address weaknesses in . plant records had shortcomings but that these did not- . result. in . - problems implying -
inadequacies in picnt hardware.
. In responding to Issue 5 " Vendor Documentation - Conditional Releases", a review was performed of the material receiving and control. systems as well as other areas .with = a poc.ential u fnr a- similar situation (i.e. concerns noted on Release for Shipment Forms, Ebasco Home Office controlled NCR's, and material received under manufacture, deliver and e:ect type contracts).
It was determined that the Trslams were limited to the absence of the formal tracking required by . existing procedures for conditional certifications in Combustion Engineering documentation packages. There was an undetected violation of procedures but based on a review of CE purchase orders, it was concluded that there would have been no safety consequences if the deficiency had-remained uncorrected. Issues 7 " Backfill Soil Densities" and 11 "Cadwelding" involved analyses of records. For Issue 7, records correlation had not been completed because some were in the Ebasco vaults and some had not yet been obtained from the contractor who, it should be noted, was still onsite and active. The correlation, review and analysis demonstrcted that there was Food work control, that specification requirements were generally exceeded, and that the backfill was adequate to perform its design fuaction. In Issue 11, the [ quantity of' data did not' allow ready analysis to demonstrate the attributes desired. Therefore,12&L transcribed cadweld data onto computer storage to demonstrate compliance with Regulatory Guide 1.10 and specification sampling frequencies. De review identified three minor discrepancies not 1 identified in the prior NCR and these were evaluated and found to be l acceptable. i , Issue 8 " Visual Examination of Shop Welds Luring Hydrostatic Testing", was l' the result of a checklist that only identified field welds. This concern , had been previously identified in June 1983 and dispositioned to , demonstrate the adequacy of the visual examination of a. hop welds and the
- j. lack of any safety impact. The review gives no indication of deficiencias.
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. The! ' records. reviews 1 for : Issue i13 ~"Missingl NCR's"' included site f NCR's, -
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Ebasco Home 0ffice:NCR's, and Mercury NCR's andfdemonstrates'that,s although. 1 documentation' was not readily available :to.. answer some Lof the concerns, l there was no : loss of tcontroliover NCRrs that L would. currently l imply-- open . ' fquestions abo'uti the ? ac'ceptability L of ' installed safety systems. i The : cause Lof most of the " concerns 'related to Ebasco !NCR's ?was identified as 'a chage ' - nin record' keeping ~ in 1979, a' temporary practice that :: allowed NCR. numbers.L to , Jbe issued priori to ; the' NCR being : written... and the use Lof ia preassigned - , . iblock of-NCRlt.habersb The review of. Mercury NCR's' concluded that there was- .l lone ' missing ? NCR1 which idid -- not / represent an iunresolved condition, fone - j superceeded > NCR,', and ;three NCR's Lwhich :had not been processed by Mercury. ;
~
These L three f NCR's, one fof ' which - is - covered by 11ssue E 1,i have ~ now been
.r' esolved. <The 'cause 1was1 Mercury's fimpropet .jpplication . of s their own procedures.
1 Issue j l6'. " Surveys and Exit ; interviews ofI Qt Personnel" involved .an LP&Li
- initiative for obtaining' employeel feedback ;on potentia 1 ' safety : concerns.
The shortcomings, of the initial program haveJ been; adJeessed.:' The exit
- interview program.has been completely restructured and is providing.a very-
{- useful service; in-~ obtaining.. feedback ' on individual's concerns. Feedback j retaived prior to the restructuring is. being ireanalyzed and concerns ~ are being closed through an orderly closure process.
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- 3. Inspection / Evaluation: . ;
{ Nine of the twenty three issues were resolved by reinspections, engineering evaluation,- statistica1L sampling, or s.inilar efforts -but required -no
- changes to the plant hardware.. An evaluation of these concerns' leads to a h conqueion there were weaknesses in plant records but
- these weaknesses-have
+ now isen addressed and do not represent a potential hardware deficiency. i-Three 'of the Issues, 1 " Inspection Personnel Issuaa", 10 " Inspector t i Qualification - J.A. Jones & Fegles", and 20 " Construction Material-Testing ^ (CMT) Personnel Qualification Records" invol/c4 a review of professional j credential and education / employment checks on 100% of the site QA/QC ' ;' personnel involved in safety related activities. In this review, QA/QC r personnel have been classified using conservative and standardized , i acceptance criteria. as " qualified" and " unqualified". -These + 3 classifications were revicwed and finalized by an LP&L Review Boar? - of . I senior QA personnel with the assistance of contractor and consultant {. support. For " unqualified inspector personnel Corrective Action Requests , i were written to formally track and disposition potential deficiencies. For , j Mercury, substantial reinspection was initiated, particularly for the N1 j tubing installation, and rework is covered in the next section. For most 1 contractors reviewed under Issues 1 and 10 the disposition of deficiencies , j has not required reinspection. In the case of Issue 20, ar. engineering
- . evaluation of the work of CMT personnel has established that questions l t
l about personnel qualifications have not rendered the work indeterminate.
- There have been many other methods (e.g. ANI, NDE, prerequisite l
- - preoperations/ integrated testing, overinspections, etc.) which provide
) assurance that quality has been built into the plant. There have been no ) , safetf significant hardware changes found and this provides positive !. evidence as to the adequacy of the overall construction program. . 1 L -s-e e}
.- - ~ . - _ _ _ - - _ . . _ . , _ , _ , - . , _ _ _ . . - _ , . , _ , . , , . - . , - - , - , _ , - . , ~ . -
M _. Issue 4, " Lower Tier Corrective Actiona Are Not Being Upgraded to NCR's" required an extensive effort to review document packages, based on a - statistical sample, to ascertain whether they had been properly upgraded to . - NCRs, whether the disposition was adequate, and whether proper reporting per 10CFR50.55(e) and 10CFR21 had occurred. The review identified minor weaknesses in the construction program in following procedural criteria for lower tier documents with regard to voiding and upgrading to NCR's. While it does indicate a deficiency in the construction program, it does not indhate that there was a loss of control over non-conforming materials, parte, or components. This conclusion is supported by the results of a statistically justified sampling program. The resolution of Issue 9 " Welder Certification" identified adequate welder certification but found that the records for seven instrument cabinets were incomplete or missing. The adequacy of the welding performed by J.A. Jones has been reviewed. In cases where welding deficiencies were iuentified, the welds were dispositioned to be acceptable as is. The missing or incomplete documentation identifies a loss of control in records management but the acceptable dispositioning of the welds and the results of the complete review of the J. A. Jones welding scope demonstrates the overall adequacy of the J.A. Jones welding. A sampling program of the information request documentation used by contractors was undertaken in order to resolve issue 14 "J.A. Jones Speed Letters and EIRs". In the case of approximately one third of the contra.cors, instances were identified where design changes were made by Information requ.sts without appropriate documentation. This was determined by taking a minimum 10% r mdom sample of each contractors information requests (for fifty or less such documents, there was a total review) and expanding that sample by 10% increments wherever there was a violation of design control. Approximately 5% of the total 1R's evaluated (approximately 6000) involved design control but no rework was required except for that being conducted within the scope of SCD-78 (lmerican Bridge Welding Deficiencies). It was concluded that the lack of control exercised over these contractors was a deficiency in controlling records in accordance with the construction program procecures. There are no remaining open issues. The response to Issue 17 "QC Verification of Expansion Anchor Characteristics" recognizes a shortcoming in not specifically delineating all characteristics on an inspection checklist although the necessary characteristics were listed elsewhere. The expansion anchors were the
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subject of several different corrective action programs as part of th e overall effort to verify the adequacy of Mercury 's work. These corrective actions previously addressed the NRC concern except for several technical questions which have been resolved. A 100% reinspection of Mercury N1 instrument installations has been completed and provides further evidence of expansion anchor adequacy. The shortcomings in the original inspection checklist are conside red a procedural deficiency in the construction program, but a current lack of safety significance was demonstrated.
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' Issue 118 ~~ " Documentation' of . Walkdowns of Non-Safety Related Equipment" resulted from.the documentation by, exception practices. used during previous : _
. plant' "two -Jover [one" ivalkdowns.. To resolve it his concern, a- detailed
. reinspection under1 a' formal engineering - procedure : was ~ performed L of ' the
-instrueent' air' system and two; p1 ant areas to provide additional confidence
-in : .the original : design' and--walkdowns.- This-. reinspection found-.no deficiencies and, supported a ' conclusion that the construction program was
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' adequate and there are no unresolved safety deficiencies.
.The resolution Lof Issue 21. "LP&L QA Construction System. Status -and Transfer Reviews" involved' demonstrating adequate control of comments and open-items
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#- riv 'the system tranafer and' testing process. 'As a result of extensive
' efforts on; this matter, including' confirmatory field verification of three items. it. was determined that no significant_ comments or open items -were runtracked andcchat there was no impact =on testing or system operation.
There were two separate issues in Issue 22 " Welder Qualification (Mercury)
-and Filler Material . Control -(Site _ wid e) '.' . ;The 'first, welder qualifications', was resolved by a thorough review of welder documentation and welder ' qualification. -No significant deficiencies .were identified and -
those minar deficiencies identified were properly dispositioned. . Concerns over weld filler : metal controls were addressed by a review which showed site : practices to be unclear with regard to ambiguities between .various code requirements. . Further, ~ justification of several past corrective actions was provided where there had -been deviations 'from the site procedure. In both cases, the evaluation demonstrated that, although there were deficiencics in procedural clarity and the control of site practices, no unresolved safety issues exist. 4 '. Hardware: Seven of - the twenty-three issues involved hardware changes in addition to - inspections, evaluations or other software . activities to resolve the concerns. A review'of these concerns has shown that, if lef t uncorrected, two of the reworked items presented a potential safety concern. Of these twa, one was related to rework on a three foot section of tubing and the second represented a case where the safety significance was not determined. It has been concluded that while construction program deficiencies existed these did not warrant an Unplication that the corrective action system as currently implemented was inadequate to provide assurance that the plant is safely constructed.- The N1 instrumentation walkdown initiatad in response to Issue 1,
" Inspection Personnel It. sues" ' has identified deficiencies that, if left uncorrected, would not have ef fected the safety 'of plant operations. The conclusions on Mercury correction actions were discussed earlier.
6 7 4
I A lack of. documentation consistent with 10CFR50 Appendix B requirements for ' local mounted instruments installed to ANSI B31.1 was evaluated in Issue 2 -
'" Missing N1 Instrument Line Documentation". In responding to the concern,
- 18 installations -were identified as having documentation insufficient to meet the objective ~ requirements of Appendix B. Based - on documentation reviewed, the as-built- installations were considered capable .of performing their intended functions. - Nevertheless. a decision was made to rework the installations to standardize compliance with ASME code requirements. This.
records deficiency in. the construction program was found to have resulted in no safety significant deficiencies. The rework was performed as part of a conservative corrective action. Issue 3 " Instrumentation Expansion Loop Separation" identified a procedural implementation deficiency in the construction program occurring when ' insufficient attention was .given by Mercury personnel to specified' installation separation criteria. Reinspections of those installations identified by the NRC as well as installations where tubing lines were run in proximity to each - other resulted in the identification of additional deviations to the separation criteria. With the exception of one-three foot section of ~ tube track all were found acceptable "as-is". The necessary rework has been completed. It was concluded that this was a deficiency in the Mercury corrective action but was of limited safety significance because of the isolated nature of the rework. Issue 6 "Dispositioning of Nonconformance and Discrepancy Reports" identified specific Ebasco and Mercury NCRs and Ebasco DRs in which the NRC had concerns relative to dispositioning, lack of supporting documentation, accomplishment of related rework and sufficiency of engineering justification of dispositions. A review of these Waterford 3 records was conducted and no condition was found which, were it to have remained uncorrected would have adversely affected the safety of operations of Waterford 3. LP&L had previously initiated a program in February 1984 to address Ebasco NCRs. This program was expanded to encomprass the NRC request and is nearly complete. While some discrepancies were noted and several reinspections performed, rework was performed in only a f ew cases. The most significant amount of rework occurred as a result of the findings in Issue 12 " Main Steamline Framing Restraints". In this case it was found that additional rework was identified from the review of American Bridge information requests and the incomplete scoping for cpen Significant Construction Deficiency 78. Rework was required to replace the framing bolts where documentation was not available and bolt identification could not be readily verified. Upon identification of the concern a conservative management decision was made to replace the bolts in lieu of attempting to test or sample test the bolts in question to determine their usability. Thus no determination was made regarding the safety significance of the existing condition. A rescoping of other significant open SCD's has been conducted to address potential concerns related to scoping practices. Deficiencies were corrected and no further safety concerns remain in this area.
s
.Is .e 15 " Welding of "D" Level Material Inside Containment" resulted in a r aspection of the most significant "D" level welds. The finc'ings _
- , intify a deficiency in the construction program because no record keeping t quirements were specified in the CB&I QA program for these type welds.
The reinspecticn of welds identified weld deficiencies that were evaluated to be acceptable "as, is" and a number at arc strikes that required rework (grinding) to demonstrate that no damage to base metal. had occurred. It was concluded that the construction program weakness created no significant safety concerns and raised no unr2 solved implications with regard to- the adequacy of the "as-built" plant. 7 ssue 19 _ " Water In Basemat Instrumentation conduit" was evaluated by a valkdown to identify .ireas of seepage and potential direct paths for ground wai.:r As a result of this walkdown a piezometer standpipe will be pressure grouted prior to fue] load to limit further seepage. This. rework was identified even thcugh the evaluation showed that there was no potential for flooding the auxiliary base sat. It was concluded that no construction program deficiencies or safety concerns exist.
- 4.
Conclusions:
The twenty three issues have been assessed and corrective actions have been or are being taken to correct deficiencies found. The safety significance of ongoing activities and completed activities is being assessed for each of the plant systems required by technical specifications to be operable during the various operational modes. Those safety evaluations needed to support any phase of operation will be a prerequisite to LP&L requests for a license to operate in that phase. The responses to the 23 issues, when assessed together, lead to two generic conclusions: (a) The QA program during the construction phase continued to have shortcomings, but with current corrective action the objectives and criteria of the construction program have now been met. The deficiencies fell primarily into the categories of records management and control of corrective actions. (b) The overall adequacy of the plant in the areas of the 23 issues is confirmed by the extensive re-evaluations and reinspections conducted in response to the 23 issues and by the minimal rework required as s' result of the concerns. The plant as-built can be operated without undue risk to public hetlch and safety. S
.9
s IDENTIFICATION OF LESSONS LEARNED Lessons learned were developed from the twenty-three issues for. the purpose of evaluating the ability of the operational phase Quality Assurance Program to preclude the mistakes made during construction. These lessons learned are
-intended to define the types'of actions which could have been taken to avoid the safety impacts that were identified. Table 2 preset.t3 the lessons learned as well as a brief description of the manner in which the operational phase Quality
' Assurance Program addresses the lessons learned. This approach allows u definition of the actions needed to anticipate problems. The need to identify emerging QC problems in a timely manner and to take effective and timely corrective actions is also recognized. The next section provides a more complete description of the operational phase QA program to supplement the lessons learned table and to describe the management oversight, trending and corrective action programs that allow for prompt identification and action on problems.
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' TABLE 1
. ACTIVITIES REQUIRED TO RESOLVE THE TWENTY THREE ISSUES 1
. Inspection /- .
(1) > Concern. Software' Evaluation Hardware 1 D-
.2 D 3 L 4 X 5 -X' 6 D '
7 X-8' X
.9 X
10 X 11 X , 12 . PS 13 Y 14 X 15 D 16 X 17 X 18 X 19 D I 20 X 21 X , 22 X NOTES: (1) The safety significance of the hardware impacts has been indicated by a "D" where hardware changes were discretionary or in accordance with gcod ptactices, a "1S" where the safety significance was not fully , evaluated, and an "L" where there was safety significance if lef t uncorrected but the significance was Itmited because of the isolated nature or limited extent of the deficiency.
+
i l? y N] . s ,- , , z TABLE 2 . OPEkATIONAL READINESS ASSESSMENT- +
, PAST FUTURE 4 .
- Actions Which Could Have Prevented Occurrence ;
, Issue (Lessons Learned) Reflection in Operational QA Program
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j 1 This concern could have been avoided if a .During the operations phase..LP&L and contractor inspection'- !' uniform and conservative standard had been personnel will be certified to' ANSI M45.2.6-19/8 and j imposed for judging QA/QC personnel . Regulatory Guide l.58 Rev. 1.- Prior to certification a-
- qualifications and for documentation of those background investigation must be satisfactorily completed 4
qualifications. documenting a candidate's education'and employment experience l as described in Section II.D. l 2 Recognize that quality records required by Documentation L (objective evidence ofi accepesace)' requir ements
- 10CFR50 Appendix B sometimes exceed the record during normal operations are defined in drawings,.
l keeping requirements of industry codes. The specifications, and procedures. Review of specified' ] concern could have been avoided if the documentation requirements' associated with station i contractors had been required to supply the mcdifications is an integral part of the operations phase i proper documentation. dasign process. This review assures the appropriateness andf j completeness of required documentation. The Station 4 Modification process is described in Section II.H. I j 3 This concern, which dealt with ff eld run Under the operations phase .QA Program field run items will be i installations, could have been avoided by . minimized and controlled by procedure. The Station j increased training of design / installation / Modification Package .JSMP) . process includes a checklist of 4 inspection personnel.in order to increase generic criteria to be addressed. _ Additionally, L the Detailed . l their understanding of generic criteria and Construction Package will contain necessary acceptance } their ability to recognize deficiencies. l criteria to direct the installer and inspector (see'Section
- II.H). ,
!' 4 The basic causes of this concern (which are .During the operations phase a uniform" program'for quality 1 not felt to be unique to Waterford 3) relate deficiency identification and resolution will be employed.- to the large number of specialty type quality The~ Condition Identification ~and Work Authorization (CIWA)' i contractors employed during the cocetruction .will- be the primary means of identification 'and L l phase, coupled with inherent desiga/ Laplementation offco:rective action at Waterford'3. . The. { construction interface problems ar scciated -quality deficiency mechanisms utilized by LP&L are described
- with parallel design and construction. The in' detail in Sections'II.B.1.a-c.
l problems in this issue accruing from the above
- situation could have been avoided nad a more-definitive and standardized quality deficiency program been developed and implemented.
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JF' u J. s TABLE 2 - OPERATIONAL READINESS ASSESSMENT'
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PAST' EUTURE < r .. a. Actions Which Could Have Prevented Occurrence - <
- Issue (Lessons Learnedl Reflection in Operational QA Program 7 -5 The concern cot 1d have been avoided if it had Any quality related . material received on site with . . ,
s been recognized that while CE handled . conditional certification .is tracked ~ in' accordance with the : certifications differently than other vendors . procedures for Discrepancy. Notices 7as described in Section that did not eliminate the requirement to II.B.I.b. track conditional certifications in order to ensure closure. 6 a. Some of the concerns could have been a. Under the operations phase- QA Program. in order .to provide avoided by recognizing the need to have'a stanJardization. hardware deficiencies will be identified more uniform process (LPSL, Ebasco, and through use of the LP&L CIWA (plant identified) or. DN contractors) for the disposition and (receipt -inspection identified) as -noted in Section , resolution of deficiencies. II.G.3.
- b. Some of the concerns could have been .b. All quality related deficiencies identified during the-avoided by establishment of a routine operations phase undergo verification review of1the - _
process for additional verification corrective action and disposition prior to closing out the (including field verification) of the -deficiency. The deficiency. identification ~and resolution resolution to assess the adequacy of mechanisms are described in detail in Sections II.3.1.a-f. : dispositions and corrective actions. More As part .of the semi-annual audit of ths ' corrective action emphasis should have been placed on a QA. . process. the"QA Program will include's field verification" management overview designed to distinguish audit of-the.CIWA closure process. In' addition, Operations generic trends and root caises of QA utilizes a QA Trending Programs to> identify' adverse 1 , k deficiencies from. isolated significant quality. trends and generic quality problems as described i occurrences ot repetitious occurrences of in Section II.B.l... l less significance. l c. Given the need for more consistent c. During.the'opecations phase,.the Quality I.ssurance Section'
. engineering judgement, some concerns could -holds monthly-training sessions.- Lessons learned'or
! have been avoided by_the use in training of corrective actions as a result of quality deficiencies or j specific disposition of.past problems. undesirable programmatic trends identified at Waterford 3: ' l will be _ reviewed during these sessions. as described. in , Section 11.E.2..-Additionally..the QA Section will prepare,' for distribution to plant . staff performing. j quality related'vark. .similar briefing' materias as a l . feedback mechanism for current quality concerns. l . ( - r w
- - , _ .2
_ ._.- _ _ .. _ _ _ _ .~ _ _ . _ _ _ _ _ . . _-_ _ . . . _ _ . _ _ .~-. . y.- _ ,, a: .m
.c y
- - ~
TABLE 2 -
,3
- ~ ' OPERATIONAL READINESS ASSESSMENT. ~
. y . . .,
~
} PAST FUTURE- - O
- g
! Actions Which could Eave s f } Prevented Occurrecce ' Issue- (Lessons Learned) Reflection in Operational QA Program' *
- . ~
- d. Records are processed:upon completion of the activity!and
~
- d. Recognize th( need for ready retrieval /
3 control of records. This would be assisted verified complete byl cognizant-supervisory personnel',All( . ! by processing records as-the work is Quality records'during the operations phase are maintained: i completed'through all required reviews, by LP&L's Project Files. Documents Lare stored and cross-L t l resalutions of comments, and necessary indeaed to facilitate timely. retrieval. Records- . verification and then vaulting-the' records, management;is further described in Section 11.1. iThe.
~
- This approach would have avoided some of current programs of record management at Waterfordl3 aren- _
i the concerns that arose because of records under: review by LP&L' management to ensure proper -
~
i retrievabiliity. discipline and optimum: utility. exists.- This review isi j expected to be complete and.any necassary' programmatic: ] changes. will. be initiated by ~ Noves.ber:30,-1984. f [ 7 This concern could have been avoided if :as . Records are processed upon~ completion.of'the activityja'nd work was completed, records were retrieved - verified complete by cognizant supervisory l personnel..Qualityl' '
- from the contractor, processed through the records.during the operations' phase are maintainea'by LP&L's i required reviews, any necessary verification . Project Files.. Records managementLis further described
- in l completed and then vaulted. Section 11.1. , .
1 4 I 8 Shop. welds, the subject of this. concern..were : N/A. J I hydrostatically tested and inspected and, ! therefore, no deficiency exists. , i 1 9 This concern could have been avoided if, as During the operations phese, any; change in scope of the' l- work was completed, records were verified as contractor's responsibilities.would initiate an LP&L review- ! complete against the scope of wonk. of>the applicable portions of the: contractor's QA' program- Y
- j. ' similarly.to.what:is required for a new' contract.L. Such~
! review would includefdocument generation requirements. > l Section II.G further. discusses the review of< contractor QA
- programs. ~-
1 { 10 This concern could have been avoided.if a During the operations phase, LP&L and contracto'r inspectionL
- uniform and conservative standard had been- personnel will be! certified to ANSI N45.L 6-1978 and
i- imposed for judging QA/QC personnel. Regulatory; Guide.1.58 Rev.l1. . Prior /t o certification a j ^ qualifications and for documentation of-those - background investigation must be satisfactorily completed qualifications. - documenting a candidate's education and_ employment experiencei as' described in-Section 11.D., - , O _. --e .m.i. -- - e- w * =--. , e > , +=- v- -o- w *
- s
. _ --. ~ _ - . _ __ _ _ . . _ _ . ._ _ _ . _ _ _ ._ . - -_ .
7 y __
'"ABLE 2 < -
OPERATIONAL READINESS ASSESSMENT c'
. L
!- PAST x , , FUTURE i -
, i - i -
1 a i N t n sg. [ ;
. , Actions Which Could Have 3i I
4
?. #V Issce .
Prevented Occurrence (Lessons Learned) S. )I 4 Reflection in Operaticual(OA Program [p [ l[ j; MU co., T
.-1 u+ 'eni f v ;(;- V ,( s (g
, p g g Y', This concern could have been avoided if, in This concern relates to bulk doasductionxand' is not- Y Rf 11,[, -f addition tp in-process analysis conducted, a applicable to the operations pha's'e![ # pf <
. ' mean( ta[ track the completion and correlation - e y ,
< ~,
j.
~' '
j of datr/ records needed to verify compli nce ' ' ' i 9~ p h ith sFecificationk had been implecented.
~
\ ,
1 , i
+
s1 _ y .,1 . .e s .-n . u I 12' This ccacarn could have been avoided if it had Multihle levels of pre .ndl post 4 implementatica review lof been recognized that scoping of complex x . ccrrective act ions occur during the operations phase. - ~ > corrective actions (e.g. mulriple contractors, Corrective action must<be implemented 'and-tracked through one'. y , complex drawings, and,consertetion ' of the' deficiency, identification mechanisms described in- y ~ ',
. Sections _II.B.1.a-e.- Brodd scope and complex'correctiva -
, interferences) required commetsurate care in
^
f : >- _uctions will;be cause forcdevelopment of a Special-Procedure >hj.
~,
assuring that the scoping of the corrective s @p action is accurate cnd tracked to assure as described in'OP-005-0015 "I6strhetions, Procedures and # " ' 2 corpletion. ' y Drawings", in criter ~ to control'sedping and interfaces, and to
. establist. a stracking mechanism to aneure completion .and f - '
(j. ' 7 .1 < '. 1 , closure'. 4 ,
'~
?, f
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'13 Some concerns could have been avoided through The operations phaec QA Progen provides for different means
{ the use of a rigidly controlled tracking from the construction pha'se ^to identify, trackV and resolve - i j system to control special purposa hardware quality problems. The quality deficiency iden'tification,- j deficiency documents that have characteristics mechanisms, all of.which provide for's"controllebtrackibj;. such as: multiple interfaces; requite system, are discussed in Sections II28.1.a-ec /
. 'Of [ yf j
tracking during processing; and/or are needed yf X [
; to control quality related questions in.a timely manner.
../ ' r/ }~ ~' ,
14 This concern could have been avoided if Plant modifications during the' operations' phase are ..
.i procedures regarding information requests had . accomplished-through the Station Modification Prog <an (SMP),
been standardized and controlled. The described in Section II.H. Work is directed by the Detailed; s
! procedures should have been the subject-of Construction Package,(DCP) assembled under the Program. .For ~ '
1 training to ensure a proper understanding and cases where work cannot be done in accordance with.the DCP,' awareness of the procedure and limitations of changes may-be allowed only upon approval _of a change-to the i the IR instrument. Audits could have been Station Modification -Package or, for minor changes, through '
, more comprehensive to assure that the program approval of a Detailed Construction Package Change (DCPC).
All work documentation, including DCPCs,:isl included in.the
~
and procedures were being properly followed. CIWA post implementation. review' described in~Section. j
, II.B.I.a,'~as well;as the.SMP closure review described ini Section'II.H. '
m TABLE 2 OPERATIONAL READINESS ASSESSMENT. PAST FUTURE-Actions Which could Have Prevented Occurrence _ .
- b; Issue (Lessons Learned) Reflection in Operational QA Program ,
15 The concern could have been avoided if Documentation (objective evidence of acceptanc'e) requirements contractors had been required to ensure during normal operations are well defined in. drawings. adequate inspection documentation for . specifications and' procedures. Review of specified Seismic Category I work outside the ASME Code documentation requirements; associated withLstation jurisdictional boundaries. modifications.is an integral part of the operations phase design process. This review assures the ' appropriateness and
, completeness of required documentation. The Station-Modification process is described in Section II.H.-
16 This concern could have been. avoided if the The LP&L Quality Team has.been constituted-to' allow'any program had been auditable, if more formal individual to express quality concerns on a' confidential training had been provided to the basis, and be assured of: ,(1) investigation of_the concern, interviewers, and if more detailed followup (2) substantiation of the concerns and . (3)" correction ofithe' had occurred.
~
. concern. The . Quality Team program-is described in detail in Section II.A.ll.
17 The concern might have been avoided if, durine .The FSAR and the'LP&L QA Manual require that inspection' the preparation of construction / inspection procedures, instructions and' checklists contain acceptance procedures more care was taken to explicitly' and rejection criteria. Prior to' implementation..there is an list the characteristics necessary to ensure appropriate review to assure that necessary acceptance proper verification of installation in the ' criteria are' adequately-transposed from the design disclosure. inspection sections and checklists, documents to the' inspection procedures,sinstructions;and checklists. 18 The two-over-one problems uncovered in the .Under~the operations phase'QA Program the Station previous inspections were documented on an Modification Package process includes a checklist _of all exception basis. The concern over the generic.criteric to be addressed during the design and-adequacy of those inspections could have been verification stege. This procese is described in Section avoided by a requirement to ensure adequate II.H. and more auditable documentation of the inspections. ; 19 There is no path for groundwater.to flow in N/A sufficient quantity to flood'the auxiliary building basement and, therefore, no dpficiency exists. i.
- m. , . ,
TABLE E OPERATIONAL REALINESS ASSESSMEh1 PAST FUTURE Actions Which Could Have Prevented Occurrence Issue (Lessons Learned) Reflection in Operational QA Program 20 This concern could have been avoided if a During the operations phase, LP&L and contractor inspection uniform and conservative standard had been personnel will be certified to ANSI-Md5.2.6-1978 and imposed for judging QA/QC personnel Regulatory Guide 1.58 Rev. 1. Prior to certification a qualifications and for documentation of those background investigation must be satisfactorily completed qualifications. documenting a candidate's education and employment experience as described in Section 11.D. ' 21 During the system transfer and testing During the operations phase LP&L will retain control and process, Waterford 3 had several groups with responsibility for new and existing systems. No system generally discrete responsibilities for transfer.outside of LP&L will occur. identifying and resolving quality related issues. This resulted in the achievement of ;0 optimum hardware quality however full .- understanding of the day-to-day coordination between those groups of the open items and their status could have been enhanced by , better documentation and training on that process. 22 a. Concerns could have been avoided if records a. As a result of this issue, LP&L is evaluating the Waterford had readily allowed the hierarchy of welder 3 welding program to identify areas of potential position and process qualifications to be improvement. As part of this evaluation, welder records demonstrated for audits and verification will be configured to readily allow the hierarchy of of compliance with requirements. welder position and process qualifications to be demonstrated. _
- b. Recognizing the need to provide clear b. Deviations from applicable codes and standards may not be_
~~
justification when there are apparent taken under the operations phase QA Program unless conflicts with cede requirements could have evaluated in accordance with 10CFR50.59. avoided this concern.
TABLE 2 OPERATIONAL READINESS ASSESSMENT f PAST FUTURE Actions Which Could Have Prevented Occurrence Icxue (Lessons Learned) . Reflection in Operational QA Program .; L 23 a. This concern could have been avoided by . a. LP&L retains and exercises responsibility for the recognizing that' delegation to Ebasco of operational phase QA Program. The QA Program of the routine QA auditing overview of Mercury. contractors / vendors performing work for Waterford 3 during without adequate LP&L involvement inhibited the operations phase must meet all applicable requirements the timely recognition by LP&L of quality of: the LP&L QA Program (see Section II.G). The , problems. Engineering and Systems Development QA Group conducts-4 audits and surveys of off-site contractors, vendors, and . quality related suppliers. The Operations QA and Plant
. Quality Groups conduct on-site. audits and surveillances of quality related: activities as described in Sections II.F.1 and II.F.2.
) b. More emphasis should have been placed on a b. Operations QA utilizes a QA Trending Program to identify-QA management overview designed to adverse quality trends and generic quality problems. This distinguish generic problem trends and root is discussed in detail in Section II.B.2.a. The yearly causes of audit findings from isolated . audits schedule is approved by the full Safety Review occurrences. Committee _( SRC). Operations QA audits are reviewed by-an SRC Suncommittee and results reported.tc the full SRC as described in Section II.A.I.
- c. Staffing levels should have been higher. c. During the operations phase LP&L retains direct control of
, its QA Program. This resulted in a significant increase I in staffing over that employed by LP&L Construction QA.
.The current staffing levels.of selected Waterford 3 groups--
including the operations phase QA organization is
~
described.in Section II.C.- i 1
; i .
r -
)
}. 1 1 ? i-3 i. OPERATIONAL PHASE QA PROGRAM ASSESSMENT
s TABLE OF CONTENTS
~
SECTION PAGE : LI. ' QA Program Overview- 1 A. . Organization .1 B. QA Program Scope 2 C. Quality Training 2 D. Inspection / Audits- _
- 3' E. Corrective Action Implementation and Verification = 3 II. Selected Aspects of the Operations QA Program 3 A. Management Oversight _ 4
.1. Safety Review Committee 4
- 2. . Yearly Management Audits of the'QA Program 5
- 3. QA Trending. Program Quarterly Rep. orts 5
- 4. Quality Assurance Probram Status Summaries- 6
- 5. Plant Operations Review Committee d 6.- Quality Inspection Activities Status Reports '6
- 7. Licensee Event Reports _
7
- 8. Availability Improvement Program Reports 7 4
- 9. Independent Safety Engineering Group _ 7
- 10. Operations Assessment and Information Dissemination Group 8
- 11. Quality Team 3 B. . Quality Deficiency Identification and Resolution- 10
- 1. Isolated Quality Deficiencies 10 a.
^
C1WAs 10
- o. DNs 12
- c. QNs 13-
-d. CARS 13
- e. AFRs 14 i
- f. NRC Inspection. Reports- 15
- 2. Generic Quality Deficiencies 16
- a. QA Trending Program 16
- b. Availability Improvement Program 18
- c. Hardware Trending 19
.C.- Staffing. 20
. D. Certification of Inspection Personnel 21 I
i __ 1 s- .- e- -,--en , -- - - - , 4 - - r- - - - - e.w ~ w ,-, y ->w -
P 3 TABLE OF CONTENTS r SECTION PAGE - E E. Quality Assurance Indoctrination and Training 21 = 1. Plant Staff Quality Related Training 21 .
- 2. Quality Assurance Section Training 22 23 L 3. Contractor Training F. Audit / Review Programs 24
{ [ 1. Nuclear Oper.ations QA Audit / Monitoring Programs 24
- a. Audit Program 24 4
- b. Monitoring Program 25 g.
- 2. Plant Quality Group Review and Verification c Process 25
- a. Plant Quality Inspection Reports 25
? 26 i- b. Hold Points
- c. Quality Instructions 27
- d. Plant Quality Surveillance 29
- e. Stop Work 29 E
E G. Control of Contractor Quality Related Activities 29 P
- l. Evaluation of Supplier's Quality Assurance T Program 29
- 2. Conduct of Contractor Quality Assurance Audits 30 Deficiency Reporting by Contractors
- 3. 31 E
H. Station Modification Program 31 - =. g
- 1. Records 33 i'
~ m i-- K-k. e E
'5 g r-ir- .,
da
's A
2
t
' OPERATIONAL PHASE'QA PROGRAM ASSESSMENT _
.The individual responses and the prior discussions in this analysis of
" collective significance" establish ~that, with respect to ths'23 issues, the plantias-built:is adequate to assura public health and safety during operation.
At the same time, the review identified various areee :Ln which the construction phase QA Program could have been improved. , While the construction phase is.
. essentially . complete, the operations phase will shortly commence. .Ia thi's light, it is appropriate to review the Waterford 3 operations phase-QA Program with a focus on the lassons learned from the 23 issues.
LP&L has establiehed a comprehensive program for quality assurance during the operating phase of Waterford 3. The Nuclear Operations Quality Assurance
- Program is applied to activities affecting the quality of.those items which-
. prevent or-mitigate the consequences of postulated accidents which could cause undue risk to public health and safety. Those activities include plant.
operation, maintenance, repair, modification and refueling. The QA Program is described in Chapter 17.2 of_the Waterford FSAR and in the . _QualityLAssurance Manual. Section I of this assessment provides an overview of the QA Program, not_a detailed discussion. In Section II selected aspects of the QA Program will be covered in detail in counterpoint to .the issues raised in the 23 NRC concerns. I, QA Program Overview A. Organization LPhL retains and exercises responsibility for the OA Program at Waterford 3. The Senior Vice President Nuclear Operations, who reports to the President of LP&L, is responsible for defining quality assurance
- policy. Reporting to him are the Plant Manager-Nuclear, Nuclear Services Manager, Project Manager-Nuclear, Corporate Quality Assurance Manager, and the Safety Review Committee (the members of which are appointed by the Senior Vice President Nuclear Operations). The corporcte organization for implementation of the QA Program is shown in Figure 17.2-1 of the FSAR.
While quality is a concern of all Nuclear Opera tions personnel, the Quality Assurance and Plant Quality Groups within Nuclear Operations deserve special mention. The Quality Assurance (QA) organization is responsible for developing, coordinating, and assaring implementation of the LP&L QA Program. Although most quality related activities are performed by personnel cutside the QA organization, an overview of the performance of these activities relative to QA-Program compliance is accomplished by QA personnel through reviews and audits. e e l
, e ._ , _ _ - . _
.QA 'is' divided into two groups.- The Engineering and Systems. _
Development QA Group conducts ' surveys and audits of contractors and vendors,-maintains the Qualified Suppliers List, reviews procurement packages..and conducts surveillance of quality related suppliers. The
-Nuclear Operations'QA Group assures that the QA~ Program at.the site is
;be{ngeffectivelyimplemented.
Operations QA is a relatively new organization. 'It became a functional
~
,uality q management tool with its first audit in January,-1982 of the system turnover process. In fact, it snus as a direct result of this-audit that the problem with Mercury (Issue #23) was first identified and reported:to the NRC. Its responsibilities include the audit',
monitoring, review and' quality trending programs.for Waterford 3. The Plant Quality Department reports to the Plant Manager-Nuclear.
\ This Department has direct responsibility to implement the requirements of the QA Program related to onsite-initiated activities including review, inspection, verification and surveillanci requirements.
B. QA Program Scope As described in the LP&L QA Manual, the QA Program is applied to all quality related areas of plant operation. For safety-related items, all applicable portions of the QA Program (i.e. Appendix B) criteria are applied. The QA Manual also provides a separate section of Special-Scope QA Policies,. defining application of selected-10CFR50 Appendix B criteria as necessary. Currently, such areas as fire protection, radiological environmental monitoring, the Availability Improvement Program, computer sof tware, radiation protection and emergency preparedness are covered as special scope policies. Special scope. policies will be issued to cover additional areas such as security and radioactive waste management. C. Quality Training Training is fundamental to quality. As a result, indoctrination and training programs are established for Nuclear Operations personnel performing quality related . activities. The programs are designed to ensure that personnel are knowledgeable in quality. assurance procedures / requirements and have the necessary proficiency to implement the requirements. The Quality Assurance Section assists l with the development and conduct of quality assurance indoctrination and training with the Corporate Quality Assurance Manager reviewing I and concurring with the program content.
+e
_ __.~ __. _ . _ , __.
qg. .. - - - . l l 1 D. b.apection/ Audits
~
Monitoring'of quality program implementation is performed through
~
inspection and'surveillances^during operation, maintenance,. modification, repair, material receiving, and storage activities.
-Maintenance and modification instruction, and work plans are reviewed
-by. Plant Quality personnel to assure the inclusion of inspection requirements and to verifygthat methods and . acceptance criteria ere defined. Inspections-are performed by qualified Plant Quality E personnel.- For quality 1related' activities (e.g. surveillance' testing) where direct inspection is not utilized, the Plant Quality Group.
surveil the activities in accordance with established procedures. Audits are conducted by the Quality Assurance Section to provide a . , comprehensive independent verification and evaluation of quality related procedures and activities. Additional audits are performed as required to-verify and evaluate _ supplier and contractor Qcality Assurance 1 Programs, procedures, activities, L and interf ace controls. E. Corrective Action Inplementation and Verification
. lor deficiencies identified by plant ~ staff or identified during the inspection / audit process, multiple means exist to implement corre;cive action. For each means of deficiency identification there exists a. process to implement, track, and verify as complete the' appropriate corrective
- action. Furthermore, through various crending programs the generic significance of individual. deficiencies _taken as a whole is identified, assessed and corrective action implemented. Such trending programs exist for the areas of programmatic, systematic and hardware deficiencies.
II. Selected Aspects of the Operations QA Program The 23 NRC issues have dealt with possible quality problems during the construction phase of Waterford 3. During the review of these issues LFLL ; has identified various lessons learned that, in retrospect, would have led-to changes in the construction QA Program. It is natural, therefore, to examine the operational phase QA Program for Waterford ) in light of the construction phase lessons learned. The discussions which follow are intended to amplify on selected aspects of the operational phase QA Program which reflect incorporation of the major lessons learned from the construction phase. It should be noted that the Operations QA Program was developed independently of the construction QA Program in order to meet the needs of an operating plant. With minor exceptions, the Operations QA Program was not changed as a result of the lessons learned from the 23 NRC < concerns, but rather anticipated and already encompassed those areas of concern.
=*
l
)
f
' The following discussions are ' divided -into' nine major areas:
A . -- Manageaient' Over sight B. -. Quality Deficiency Iddntification :ad Resolution-
'Cl . Staffing D. Certification of Inspection; Personnel.
- E. -Quality Assurance Indoctrination and Training F. --Audit / Review Programs . f .
G. . Control of' Contractor Quality-Related Activities
'H. Station Modification ~ Program
^~..
-I. . Records:
~
c A. ' Management Oversight = Maintaining a high level of quality 1 at an operating. nuclear power. plant requires continuous management involvement in the QA Program.
.LP&L management has' structured the opetetional QA' Program to ensure-management oversight sad control of all aspects'of quality at Waterford 3.
The Plant Manager, reporting directly to the Senior Vice President.
- Nuclear. Operations, is responsible for the primary implementation of .
- . quality-related 4easures'during the operation activities-at Waterford
- 3. .The Senior Vice President Nuclear Operations, the Plant Manager, and other utility executives esp.loy ~a number cf management tools to implement and validate ebe operational QA Program.
d
- 1. Safaty Review Committee-g The Waterford 3 Safety Review Committee (SRC), of chich the Flant j Manager is a member, reports directly to the Senior Vice
! President Nuclear Operations through monthly reports of SRC activities. It is primarily responsible for the management level , overview of the operation of the Waterford 3 plant to assure that the plant is operated in accordance with the Technical Specifications and to review significant safety issues. One of the key functions of the SRC is to review the audit program as defined by the plant Technical Specifications. At Waterford 3 the SRC has established a subcommittee responsible for reviewing all QA audits specified by the Technical l- Specifications as well as reviewing any special audit or additf onal-audits performed by the QA organization. The SRC Charter requires a minimum of quarterly reviews of the resolts l of the audits performed. As a matter of practice, the audit l- subcommittee generally has review meetf ugs scheduled concurrent l with the monthly meetings of the full SRC. These subcommittee , meetings include a review of the results of all audits performed since the last subcommittee meeting. Significant issues raised in these audits are brought to the attention of the full SRC. In addition to reviewing the individual audits and their - rindings, the subcommittee reviews the schedule of audits as __ prepared by-the Operations QA Group to assure that it is'in conformance with the requirements of the Technical Specifications and to casure that audits are being conducted on a timely basis in accordance with that schedule. )
, l
-- ~, .- , - . . - . . . . , . - - . . . - . - . . - . , . . - , - - - . - _ - - . . . . . - . . . . , , . . - . . - - . - - - , - - - -
Because the SRC is. concerned with an overview of plant-operation, and' identification and review of significact safety a issues, the SRC review of the operational QA audits serves. to provide an additional-review of root cause, generic implications, and safety-significance of'the findings in those audits. In addition, the SRC receives regular reports by the Corporate Quality Assurance' Manager of significant issues and occurrences in the QA area. The combination of an overview ofl the QA program and the QA-audit findings provides an-opportunity to assess the quality of the audits in determining.and evaluating QA issues at a management level.
- 2. Yearly Managecent Audits of the QA Program Audits of the Quality Assurance Program are ccnducted as 4
specified in the QA Manual, Chapter 18.7, and in the FSAR, Section 17.2. These' audits are currently scheduled in accordance with QA procedure QASP 18.12. Management audits are conducted by an independent audit team from the Middle South Services Quality Assurance group. Members of the audit team are qualified to appropriate standards. The review topics cover all activities associated with the a,' ministration and execution'of LP&L's QA Program. Findings are reported to the Senior Vice-President level and assigned to the appropriate LP&L QA managers for corrective action. Findings'
, -are tracked using approved procedures and forms. Audit findings are reviewed for underlying causes to determine corrective action to prevent recurrence. Those deficiencies requiring long i term action to correct, or which have the potential for recurrence, are reinspected in follow cn management audits to determine the effectiveness in addressing identified problems.
It is anticipated thet the yearly management audit of the QA Program will be an effective management tool in assessing and maintair.iug the adequacy and effectiveness of the operations phase QA Program.
- 3. QA 'frending Program Quarterly Reports The Operations QA Group administers a QA Trending Program intended to identify adverse programmatic quality trends and initiate corrective action. While other mechanisms exist to identify and correct individual quality concerns, the'QA Trending Program will-allow management a tool to identify l- underlying " common mode" sources of quality deficiencies. The QA j Trending Program is described in detail in Section II.B.2.a.
i i
l h: = A Trend analysis reports will be. issued quarterly by the Corporate
-Q4. Manager to the Safety Review; Committee and the. Senior Vice ~
President Nuclear Operations. ItLis expected that the QA. Trending Program will prove a valuable senior management tool
- for assessing _and controlling the level of quality-at Waterford-3.
.4. Quality Assurance Program Status Sununaries Summaries of QA Program activities at Waterford-3 are provided
^ to the Senior Vice President' Nuclear Operations on a weekly and monthly basis. a) Weekly Report ' provides a status _as~of the last: day of the week reviewed for various QA Program subjects of interest . which include Audits & Reviews, NRC Site Activities, and QA Training. These reports are posted in all QA office locations. b) Monthly Report - presented .to ' the Chief Executive Officer I. and Senior Vice President Nuclear Operations during the monthly Program Review meeting. It provides a summary of site-related QA. activities similar to the weekly report and includes statistical studies where applicable.
- 5. Plant Operations Review Ccamittee The function of the Plant Operations Review Committee (PORC) is to advise the Plant Manager on all matters related to nuclear safety. In fulfilling this function the PORC reviews, among others, plant procedures that affect the public health and safety, propased hardware m)difications that affect nuclear a safety and all reportable e"ents. The PORC provides the Plant Manager, prior to implementation, with written recommendations and 10CFR50.59 safety evaluations with respect to the acceptability of procedural and hardware changes. The minutes of each PORC meeting, documenting the results of all PORC activities performed under the provisions of the Technical Specifications, are provided to the Plant Manager, Senior Vice President Nuclear
- Operations, and the Safety Review Committee.
- 6. Quality Inspection Activities Status Reports I The Plant Quality Department will provide quarterly reports to
- ne Plant Manager-Nuclear. Included in the reporting is an analysis of quality trends with respect to deficiencies
- identified during processing of Discrepancy Notices, Quality Notices, and Plant Quality Department reviews / inspections of CIWAs, procedures and procurement documents. Reporting in this area-has recently commenced. The frequency, format, and i- categories reported in the Quality Inspection Activities Status ,
! Reports are expected to change to fulfill the needs of the Plant , Manager in detecting adverse trends in quality related activities on site. i i
~
i-W
- 7. 1 Licensee Event Reports-LP&L has established.a permanent'onsite: Event Evaluation JCommittee (EEC) for the purpo'se of coordinating the evaluation, reporting.and, closure of correctiveLactions: associated with 5 reportable events described in 10CFR50 73.- The EEC is s ';
- responsible,tofthe Plant Operations Review Committee (PORC) and
. the . Plant Manager.
3 Any inidividual identifying a reactor. trip, transient, safety-
.related equipment failure;or malfunction, radiological-event, security; event,Jviolation of a technical specification, or'other events deemed to be potentially reportable,.are responsible for
' initiating _a. potential reportable event:(PRE). report. .Following.
any necessary -immediate corrective actions and/or modifications, the EEC ensures that a prompt, thorough PRE investigation.is conducted.~. During the investigation, the=cause.of the event is
-identified and corrective action' initiated to prevent recurrence.
Generally, corrective action is. documenteu and tracked .via one of
~
the deficiency identification mechanists discussed in Section1 TI.B.I.a-e.. In addition to the standard closure verification processes,1the EEC independently tracks and confirms adequacy of corrective action. The EEC provides the PORC with a report of the completel investigation and recommendations. --Following PORC review the l- Plant Manager is responsibli for approving disposition of PRES as Licensee Event Reports /for transmittal'to the NRC. 4 8. Availability Improvement Program Reports The ' Availability Improvement Program (A~.P) is currently under -
. development by LP&L for implementation during the operations phase at Waterford 3. Quality related problems, as described later in this submittal, will be periodically reported to senior management.- Whereas.the QA Trending Program will-provide i
management input as to advarse programmatic trends, the AIP will i provide adverse trend information on the system / hardware level.
- 9. Independent Safety Engineering Group-One of the functions of the Independent Safety Engineering Group
~
(ISEG) is to prepara and conduct independent reviews of plant activir.ies which may result in recommendations to plant staff and corporate management. These recommendations include corrective actions such as procedure revisions,' equipment modifications and
~
1 additional training necessary for improving overall quality nasurance and plant safety. Evaluations of plant operations, maintenance and modification are documented through ISEG reports. i: These reports, as well as any action item resulting from them are ! logged by the ISEG group for purposes of tracking and resolution. - f To keep management appraised of ISEG activities, an ISEG Monthly _ p Summary is provided to the Senior Vice President Nuclear " L Operations and the Engineering and Nuclear Safety Manager listing u -evaluations performed that month and areas of ongoing review. l. L _ _ . - . _.-_ 2 .. . __ , _ _ . _ _ _ - - _
- a'wa
+ N .I
- 10. - Operations Asse:,sment and Information Dissemination Group
~~
The Operations Assessment and Information Dissemination' Group (OA&ID) is responsible to the Nuclear. Safety Supervisor for-screening,. evaluating, and disseminating operational. experience information. A significant management overview function that thn'- OA&ID group.will provide'is the detailed evaluation of selected LP&L' Licensee Event Reports (LERs).- This evaluation.will explore. generic implications or special aspects' of the~ event which are outside.the' scope of normal LER evaluation and review. Periodic status reports will be provided to management.
~ 1[. Auality Team The LP&L-Quality Team offera concerned individuals the-opportunity:to voice quality concerns on a. confidential basis.
Reporting directly to the Senior Vice President Nuclear _ Operations,-the Quality Team has been empowered with the authority to conduct investigations of any quality concerns , brought _to their attention; investigate instances of. -
- intimidation and harassment of individuals providing informati n -
to the. Quality Team; and maintain strict independence and corfidentiality.. Following preparatory work the Quality Team was staffed and began full operation at the beginning of August, 1984.- The Team acquires quality concern information through the following methods: 7
^
- a. Local and toll free hotline telephones are established to receive quality concern calls. The numbers are published
, widely to project personnel. Quality Team personnel man the
; phones during working hours, while calls are recorded at
! other times.
- b. All personnel terminating employment from Waterford 3 exit i through Quality Taam headquarters. Personnel are afforded g~
the opportunity to express quality concerns on a confidential basir,. Any individuals who t.arminate employment off site' or during other than wors;ing haars are sent a letter requesting any quality concerns they may have.
- c. All Waterford 3 personnel can " walk in" the Quality Team j headquarters at any time to discuss quality concerns,
- d. Concerns received by the Quality Team from sources external to Waterford 3 are documented and processed in the same 4
manner as internal concerns.
- e. Tha Quality Team is re-evaluating all interviews conducted prior to the presenc Team-configuration (see NRC Concern ,
#16). ..
4
, . . -_ . . . ~ - - - - ._ _ . . . , _ _ _
- ??
Regardless of how the quality concern was identified, each is _ addressed in the same manner. An initial revies is conducted for reportability and safety significance requiring immediate corrective action. An Investigative Plan, intendei to tasolve
- each concern identified, is-then developed and t Quality Team investigator assigned for completion. Once the investigative actions are completed and the concern is resolved all ~ documentation is retained s an auditable filu. The specific procedural steps are contained in QASP 19.ll.." Quality Team Operating Procedure".
Substantiated quality concerns are documented for corrective action and verification on a Quality Team Deficiency Report (QTD2). The QTDR is~very similar in form and handling to the Corrective Action Report (CAR) discussed in cection ll.B. I .d. The Quality Team reviews the results of implementing the QTDR findings and, where the corrective action is unsatisfactory and/or attempts at resolution have been unacceptable, the Quality Team notifies the Senior Vice President Nuclear Operations by letter requesting resolution and action (s) to prevent recurrence. Final reports for all concerns are directed
~
to the Senior Vice President Nuclear Operations with copies to appropriate senior managers. The Quality Team is committed to investigate concerns in a manner that focuses on determining root cause and complete implementation of corrective action. To support root cause determination the Quality Team maintains a trending program categorized by type of quality concern (e.g. unqualified personnel, inadequate training) and means of identification (e.g. hotline, " walk-in"). The basic elements of the trending 1 progres center around data retrievability and sorting to suit
. management needs. The key attributes are:
- a. Concern categorization a9d coding
- b. Statistical data gathering
- c. Evaluation and analysis.
The Senior Vice President Nuclear Operations, and other appropriate senior management, are provided with timely Quality Team information to assist in their assessment of the status of the QA Program. The Quality Team transmits, among others, the f ollowing reports:
- a. Weekly Status Report of the Quality Team Program .
Activities
- b. Quality Team Monthly Status Report
.c. Quality Team Deficiency Trends Status Report (weekly)
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3; B.- Quality Deficiency Identification and Resolution , In maintaining and improving quality a comprehensive program must exist to identify cnd correct quality deficiencies. Two components are important for. successful . implementation' of - such a program. 1First, sufficient means and opportunity should be available to identify and correct individaal quality. concerns as they occur. Secondly, a capability should exist to as ess the identified . deficiencies as a whole to determine whether .they are isolated occurrences or.due to underlying common causes. The LP&L QA Program incorporates provisions for both components of quality deficiency identification.
- 1. Isolated Quality Deficiencies LP&L employs a hierarchical' system for identification of individual quality deficiencies. 'At the first level of the d
hierarchy it.is intended that adverre quality con'itions will be ! identified by plant staff using CIWAs (Condition Identification and Work Authorization), DNs (Discrepancy Notices) and QNs (Quality Notices). The second level of detection includes CARS (Carrective Action Request) and AFRs (Audit Finding Reports) issued by the Operations QA Group during monitoring and audits. Finally, at the third' level are NRC Inspection Reports. Upon identification af the quality problem, sps ific action is necessary for effective resolution: 1) cause is identified either explicitly or as part of the trending program, 2) appropriate corrective action is implemented, 3) a means of , tracking the deficiency and corrective action (s)- to completice is available, and 4) verification of completion and l 4 effectiveness of corrective action is documented. These steps are included for the deficiency identification mechanisms at Waterford 3 and are described in the discussions which follow. i
- a. CIWAs PURPOSE: The Condition Identification and Work Authorization (CIWA) is the primary vehicle through which abnormal plant conditions are identified, evaluated and corrected, as well e.s the means for implementiag routine maintenance.
ORIGINATION: If, during the course of inspection, testing or operation, a condition adverse to quality is identified by any Waterford 3 personnel, it is required that a CIWA be generated. Routine maintenance must also be performed via a CIWA.
-e
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4 # CORRECTIVE ACTION' IMPLEMENTATION: ; Except in cases requiring-immediate attention',Hcorrective' maintenance nay not commence _ without a processed CIWA in accordaner with UNT-5-002. . Any maintenance or' adverse quality.cor.dition favolving the basic
. power'piant.is forwarded to the Control Room Supervisor (CRS)/Shif t Supervisor (SS) .for ' review. sThe- CIWA is then;
. forwarded to Planning and Scheduling Department . (P&S) for evaluation,' dispositioning and work plcnning. CIWAs are evaluate.1 as nonconformances when ;the adverse quality -
condition is. determined to be a. departure from specified requirements and,S1) is not the result of normal wear or,
'(2) is not a secondary effect due to failurefof another component, or (3) is not identified as a routine part of the work process and will be corrected as a continuing.part of the work process, or (4) 'is dispositioned as " repair" or, "use-as-1s", or. (5) is a suspected generic problem. If . the
~
CIWA is dispositioned as " repair" or ;".use-as-is", it . must obtain concurrence from Plant Engineering. Plant Engineering performs a technical evaluation in such cases (including a Safety. Evaluation, if necessary) to determine . cause and corrective action'and documents the results on the CIWA. If a design: change is necessary, a Station Modification Request number is entered on the CIWA. When-the CIWA has been dispositioned, a copy is forwarded to On-Site Licensing for.a 10CFR21 evaluation. i lhe CIWA is then processed as a work package by the appropriate discipline. The CIWA work package is reviewed and approved prior.co commencemeat of work by the responsible Maintenance Supervisor and Plant Quality Group
. (for quality related work packages) to ensure inclusion of accurate and complete work instructions and/or inspection
- l Hold Points. Subsequent changes which change the scope of work or acceptance criteria are reviewed by the same review organizations.
.Upon completion of work, the responsible department Supervisor reviews the work package for completeness and i
forwards the CIWA work package to P&S for closure on the MTS (Master Tracking System). The MTS identifies all archived and active CIWAs at the plant site. Tight administratPre controls are instituted to assure prcper input and extraction of data to/from the MTS. CORRECTIVE ACTION VERIFICATI0h: Post closure review by the Plant Quality Group and Plant Engineering consists of an overall review of the adequacy of the CIWA and corrective action. All CIWAs identified as Non-Conformance are periodically analyzed by Operations QA for adverse quality
- trends. The Nuclear Safety Section of the Project I:
Management Department also provides an independent review of _ , non-conformances, dispositions, and close-outs. 11- _...---g-_-.., , , 7 , , , .-,y. - -
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'PNRPOSE: , The Discrepancy. Notice (DN)[is the mech'anism'
.through which discrepancies are identified:during receipt.
inspectionsfof, quality.related parts,. material, and components by LP&L Plant Quality personnel at Waterford_3. ORIGINATION: .-Upon receipt of quality related items, Stores. personnel notify _the-Plant; Quality Group.and initiate a Material Receipt Inspection Repor t. :For,those. items
, .specified'in,the procurement package as requiring tailored' i Lor Speciall Receipt Instructions, a "Special' Receipt Instruction Sheet" will be initiated'by Plant Quality _
. personnel'., The inspector examines incoming materials in-accordance.with approved inspection instructions. In the event a-discrepancy-is identified during the inspection,' a DN is issued by Plant Quality which maintains a log sad
~
status of al1~DNs. The DN is also forwarded .co Licensing for 10CFR21' evaluation. CORRECTIVE ACTION IMPLEMENTATION: A " hold tag" is attached. to:the' discrepant item (s) inspected which is then.placad in-a segregated area. A Material' Review Board (MRB) exists to - ensure proper disposition of discrepant material. . . Representatives ~to the MRB, which is chaired by.the Plant-Quality. Manager, include personnel from Maintenance,~ Plant. Engineering and Purchaaing. .Upon completion of. review and concurrence with the final dispotition, members of'the MRB sign and date the DN. If the discrepancy can be corrected after installation,-the' item eey be. released for-installation on a ." Conditional Release" (CR) basis subsequeat to approval of the " Request for Conditional Release" (RCR). Once the RCR is-approved and granted, the. - CR- is : sequentially numbered and logged in the CR Log and stated as such on the CR tag and the RCR. The." hold tag"' will be removed from the item in exchange for 6 "CR tag". j- The original RCR stays with the DN and a copy is attached to the~ CIWA with special instructions (limitations) for installation. Conditionally released items may not be , placed in-service until the DN is satisfactorily closed. l Closure of the CR is a pre-condition for cloaure of the DN. In those cases where a design change was necessary'to close the CR, a Plaat Engineering representative has joint approval responsibility, i L. CORRECTIVE ACTION VERIFICATION: The Plant Quality Manager. is ultimately responsible for approval of DNs through inspection / reinspection, as applicable. DNs are i periodically analyzed by the Operations QA Group for quality ! trends. The Nuclear Safety Section of the Project Management Department will also. provide an independent " review.of non-conformances (DNs), dispositions, and - close-outs.
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' c. QNs PURPOSE: Conditions adverse to quality which are due.to a lackfaf, or a breakdown.in, administrative 1 controls 1 are
' documented with a-Quality Notice (QN). This document identifies non-conformances indicating a breakdown or -
substantial departure from required procedures or - instructions to the extent that a loss of control ~is evident. ORIGINATION: :Any Waterford 3 employee may initiate a QN and request a sequential number from Plant Quality who maintains the log and status of each QN. Within'30'dayslof the-identification of a.QN, the responsible department'is required to report the actions taken or proposed to cover the following: a) the.cauee of the condition, b) correction of the conditions identified, c) action to prevent recurrence, and d) schedule of implementation. CORRECTIVE ACTION VERIFICATION: The Plant Quality. Group is responsible-for verification of corrective actions committed to in the 30-day response-supplied by the affected discipline (s). The Licensing Group reviews QNs for reportability under 10CFR21. QNs are periodically analyzed by the Operations QA Group for quality trends. The Onsite Safety Review Subgroup' of the Project Management Department provides an independent review of non-conformances, dispositions and close-oute.
- d. CARS PURPOSE: The purpose of a Corrective Action Request (CAR) is to provide a mechanism through which the Operations QA Group can document deficiencies based on monitoring of plant activities or conditions, and present such findings to the affected Manager for a timely and effective resolution of the concern.
ORIGINATION: A CAR originates as the result of monitoring , or. observation of a quality affecting activity or condition
, which could be detrimental to the safe operation of the plant and/or safety of personnel. QA personnel assess the cause and significance of the deficiency to determine if an ; immediate corrective action is required. Where such a
, determination is made, a "Stop Work Order" may be initiated, or other steps taken for immediate implementation. The CAR includes a description of the identified deficiency, and a l' requirement that corrective action, underlying cause and action to preclude recurrence be documented by the " responding organization. - l 13-
1 A 1 1
- l CORRECTIVE ACTION-IMPLEMENTATION: The delivery date of the l
. CAR to the affected organization is the start of the 30-day j period during which _ the cognizant' group must resolve the i
deficiency, or define steps to be taken to effect-resolution and provide a schedule for completion. CORRECTIVE ACTION VERIFICATION: If the resolution and corrective action'are considered acceptable, the QA
. Representative indicates so on the. CAR'and recommends approval and closeout of the CAR. The ori& inal CAR is given -
to the applicable QA Supervisor for final approval and filing. -If the resolution and corrective action are_not considered applicable, the cognizant Group Head will be so informed and a schedule' arranged for satisfactory disposition. The action taken will be filed in the Open CAR File. If corrective action and the schedale for resolution are acceptable, but such action has-not yet been taken,_the QA Representative may accept the proposed resolution on the original CAR and maintain it in the Open CAR File. .After satisfactory resolution and closeout as attested to by;the applicable QA Supervisor's signature, the original' CAR will be maintained.
- e. AFRs PURPOSE: The Audit Finding Report (AFR) is the Operations
, QA mechanism for documenting deficiencies identified during audits of organizations performing quality related activities at Waterford 3. These AFRs-are then forwarded to appropriate levels of management.
4 ORIGINATION: An audit is structured around a checklist prepared by the auditor and concurred with by the supervisor. The checklist is used during the audit to compare the audited organization's mode of operation against procedures, standards and other documents which govern its domain of operation. CORRECTIVE ACTION IMPLEMENTATION: The audited organization is required to complete the following actions upon receipt of the audit report: a) Review and investigate the condition described in each audit finding, b) Schedule appropriate immediate corrective action to correct the deficiency and to prevent recurrence, and c) Respond to all findings within (30) days af ter acknowledging the audit finding. The response must clearly state the corrective action inplemented and/or the scheduled date targeted for the completion.
^t x
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CORRECTIVE ACTION VERIFICATION: . The QA Audit Sup'ervisor assures that corrective action is_being accomplished in a _
. timely manner by maintaining a tracking system of all unresolved items. The Lead Auditor confirms through personal observation or verification, that corrective action is' accomplished as scheduled. The verification review also assures that the corrective action is adequately identifiedL and implemented for each finding, including considerations for:
'a) Similar conditions b)~ Corrections as to cause c) Software aspects d) ' Hardware aspects e) Schedule f) Completeness f .- NRC Inspection Reports i
ORIGINATION: These. reports are transmitted to LP&L by'the NRC Region IV office. A summary of NRC inspected _ areas of operations, maintenance, administrative controls, and license activities are contained therein and may identify open items, unresolved items, and/or Violations / Deviations. S CORRECTIVE ACTION IMPLEMENTATION: The Nuclear Services Manager and the Nuclear Support and Licensing Manager are responsible for the coordination of reviews and preparation
! of responses to NRC Inspection Reports. This task is performed by the Onsite Licensing Unit of the Licensing Section.
The specific task is performed by the Licensing Engineer (LE) through the development of a Licensing Action Plan (LAP). This plan may necessitate input from other departments and is transmitted to them through the use of a Licensing Information Request (LIR) form. The LIR .s responded to and certified by the respective departments via the Task Review And Certification (TRAC) form. The response is reviewed by the LE for consistency with the LAP, LP&L commitments, completeness and the FSAR. Inspection
, Report responses are reviewed by the Plant Manager, l
Licensing Manager, and the Nuclear Support and Licensing l llanager prior to transmittal to the NRC. CORRECTIVE ACTION VERIFICATION: This is accomplished through receipt of signed off TRAC forms from responsible departments as well as a confirmatory review by the LE. LIRs are tracked from inception through completion by the LE via the computerized Licensing Commitment Tracking System. Responses to the NRC pertaining to Inspection Reports and . 10CFR21 are further validated by the Operations QA group via _ QASP 19.13 prior to transmittal to the NRC. - l _ .__- 1.__. _ - _ - _ _ . . _._ _ _ _ - - . - ___ _ , _ . ~ _ . _ _ . . _ . _
.]
-l 2'. Generic Quality Deficiencies There may be cases where correcting individual' quality -
deficiencies is. insufficient to assure overall quality. Such cases occur where there are underlying causes common to more than one deficiency..-Therefore, LP&L-has established programs to provide timely, identification and correction for such generic. deficiencies. The'following.three sections will discuss the QA Trending Program, the Availability Improvement Program, and Hardware Trending.
- a. QA' Trending Program-
-Recognizing the need for early identification and correction of generic quality problems the Operations QA-Group initiated a Quality' Trending Program in May, 1984 with the publication of procedure QASP 16.1.
Data Reduction The Operacions QA' Group collects and analyzes quality data for the purpose of identifying adverse trends. Responsible organizations initiate corrective action for Waterford 3. programmatic deficiencies. Documents to be incorporated into the trend analysis-include, but are not limited.co:
~
CIWAs (Condition Identification and Work Authorizations) QNs (Quality Notices) DNs (Discrepancy Notices) AFRs (Audit Finding Reports) CARS (Corrective Action Reports) NRC Inspection Reports For each document the assigned QA representative will review and identify any deficiency in the effectiveness of the QA Program. The identified deficiency will then be categorized according to the following scheme: Equipment Control Training and Qualification Design Control Maintenance and Modification Control Procedure Adherence Plant Records Management Control of Purchased Materials and Services Identification and Control of Materials, Parts and Components
. . ~. . - , . . - - . . _. - - - . - . - . -. - - .. .-
. - - .- .. - -. . . ~ .- . _ .
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- Control of Special Processes - '
Inspection Test Control . Control of Measurement-and Test Equipment- . Surveillance Testing and Inspection-Schedule Plant Security' '. Corrective Action
- As' experien'ce is gained in the trending program, categories '
will be addad and deleted as necessary. Trend Analysis The Operations QA representative will evaluate the trend reports.to determine-if a possible adverse. trend exists based on the following:
- a. 'A significant-increase in'the number of occurrences of a specific adverse . condition category is noted as
, compared to the previous reporting period.
- b. 'A continuing and significant rise in the overall trend
- of adverse conditions forfa responsible organization over the-last three months is noted.-
- Further investigation to confirm possible adverse trends
- may be indicated and accomplished by monitoring the specific activity or program in question.
4 Correct'ive' Action i' . Corrective action will generally be -in the form of issuance .
- of a Corrective Action Request (CAR) to the Manager of the responsible organization. Future trending reports will.be used (in addition to standard QA confirmatory actions) to verify the adequacy of the corrective actions.
Reporting The trend analysis report will be issued on a quarterly basis in the form of graphs and summary reports (including i summaries of CARS and corrective nctions) toL the Safety Review Committee and to the Senior Vice President Nuclear Operations through the Corporate QA Manager. The reports ; I will be formatted in a manner to facilitate the identification of trends in programmatic deficiencies. T f-
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g . Management: Overview. . The, trending program provides a_ valuable senior management tool for. assessing;the effectiveness of.the qualityJprogram
-at'Waterford'3. Trends whose root cause'may-lie in.the
' areas lof staffing,' corporate: philosophy,: management
- deficiencies, and the
- like, 'can-most approprintely be
. resolved through;the Senior Vice President Nuclear- ._
f Operations'following his quarterly. review of the trending reports. CurrentLStatus The trending progras has'been recentlyl initiated'at. .
- ~
Waterford 3 with the first_ quarterly-report to the Senior-
- j. Vice' President: issued in October,~ 1984.
L b.' -Availability Improvement Program 7 [ The Availability Improvement Program (AIP)'for'Waterford 3-ll will be_ implemented-to: improve overa11' plant reliability. l In so doing, quality related problems will be' identified to-
- management and corrective action implemented on a system / component level. While the QA Trending Program'wilin identify generic programmatic deficiencies,-it is expected 4
that problems identified by the AIP will be predominately in - j- the hardware area. l The AIP centers around a computerized model of.the Waterford 3 plant. The plant will l>s divided-into generic t functions, which will be further subdivided into [ subfun'tions, c equipment systems, and, finally,' equipment
- items. The model database will be regularly updated to i reflect actual plant performance data, enabling the a calculation of reliability / availability for any' l_ hierarchical level of the computer model. Availability-
- goals will be set initially based upon industry performance of similar plants. As the AIP proceeds, and the database is extended, plant-specific availability goals will be j' utilized.
L l When an unusual characteristse affecting some met.surement. ! -- of availability is identified, or a problem is recommended for investigation, a Unit Availability Investigation (UAI) will be undertaken. The UAI will focus on a group, or , individual piece, of hardware as appropriate. A root cause l' analysis will be performed to determine the reasons for [ abnormal performance. The analysis may make use of plant [- personnel-interviews, vendor interviews, consultant interviews, investigation of environmental conditions, special testing, etc. - L
~ ' - ---
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34 s1+ t + Upon determination of the root cause of the problem, corrective action will' be ' implemented as necessary and _ tracked to completion. Verification of ef fectiveness of the carrective action will be evidenced through . improved availability. performance.under the AIP. Periodic reports of the resalts of the.AIP will be provided to Nuclear Operations ~ management, including the Senior Vice-President Nuclear Operations.. Such_ reports will identify advr.rse availability trends, the root cause of such trends, corrective ~ action taken, and confirmation of effectiveness of the corrective action. As with any trending program,;an operational database is required prior to effective implementation of the AIP. LP&L expects thel AIP to be fully implemented within two
. years.
- c. Hardware Trending The purpose of the Maintenance History System .(hHS) _is to identify potential improvements in the preventive
, maintenance program, to suggest Jmprovements to corrective maintenance procedures, to. identify. equipment requiring-upgrade, and to provide a tool for assessing adequacy of spare part inventory levels. Af ter completion of a plant modification, repair or maintenance, . a MHS form is filled out on the affected component describing the nature of the work perfcrmed. The MHS form is attached to the CIWA before routing for closure review. These forms are used for data entry into the NHS computer system. The MHS data base is currently under extensive review to update and verify accuracy and adequacy of input data. This data base will i provide a complete preventive and corrective maintenance history of all plant system components. This will enable LP&L managers to detect equipment trends 1.1. systems under
; their control. Once operating time is accumulated on plant systems the Plant Maintenance Superintendent will select key systems to review the frequency and scope of preventive maintenance for changes as necessary to improve systew operability.
Pump and valve testing performed under the requirements of the ASME Boiler and Pressure Vessel Code is another source of trending information. A list of Section XI tests performed on safety related equipment under'this Code for which data nust be recorded to identify failure trends has [ been established at Waterford 3. This list includes such equipment as the Emergency Diesel Generator, Charging Pump, Containment Spray Pump, Reactor Coolant System (RCS) Pumps, RCS Instrumentation, MSIVs and containment isolation . boundary valves. This trend information will provide plant _ management with advance notice sufficient to take the ~ necessary corrective actions te prevent failure of such l equipment vital to nuclear safety. I _t9-
-- e In programs of this magnitude it is inevitable that changes _
will be necessary. As LP&L gains mete experience in quality trending, program refinements will be made to support the program purpose of identifying adverse quality trends. It is also-important to note that the effectiveness of any trending program is a direct function of.its database. The identification of trends requires a detailed previous i history. By initiating _the trending program at this time LP&L expects it to become a useful management tool going into commercial operation. ,
-I C. Statfing j l
The organization, staffing levels and personnel qualifications for ) Waterford 3 are described in Chapter 13.1 of the FSAR. Staffing of i key areas of plant operations and quality include: Authorized Actual Level Staff Staffing Level- as of 9/84 Plant Operations and Maintenance 211 191 Plant Technical Services 96 92 Plant Training 31 28 Plant Quality 13 13 Quality Assurance 46 42 The operations phase QA organization is divided into two main groups - Nuclear Operations QA and Engineering / System Development QA each of which is further subdivided into 3 sections. QA staffing for the operations phase is detailed below: Autborized Staff Staffing Level Nuclear Operations QA Manager 1
- QA Audits 9 - QA Support 6 - QA Analysis 9 - Total 25 Engineering / System Development QA Manager 1 - Audit / Surveillance 5 - System Development 7 - Engineering / Procurement _ ,4_ - Total 17 QA Management 4 l
D. Certification of Inspection Eersonnel _ Inspection I.ersonnel during the operations phase of Waterford 3 including *. hose provided by contractors are certified in accordance with QI-10-001, " Qualifications of Inspection Personnel". Certification for Level I, II and III qualifications is done in accordance with ANSI N45.2.6-1978, and Regulatory Guide 1.58 Rev. 1. Prior to certification a background investigation must be satisfactorily completed verifying a candidate's education and employment expetience. Recertification is performed every two years. E. Quality Assurance Indoctrination and Training
- 1. Plant Staff Quality Related Training An indoctrination' and training program has been established for the Nuclear Operations Department personnel performing quality related activities. It is designed to ensure that personnel involved are knowledgeable in quality assurance procedures / requirements as well as the overall functional responsibilities in the plant, and have the necessary proficiency to implement the requircaents. The scope, objective, and method of implementing the indoctrination and training program are documented in procedures developed by the Training Department. The Quality Assurance Training and Indoctrination Program requires that:
a) Personnel responsible for performing activities that affect quality are instructed on the purpose, scope, and implementation of quality related manuals, instructions, and procedures; b) Personnel performing activities that affect quality are trained and qualified in the principles, techniques, and requirements of the activity being performed; c) Proficiency and requalification of personnel performing activities requiring certification are maintained by retraining, re-examination, and/or recertification on a periodic basis; d) Proficiency tests be givea to those personnel performing i and verifying activities affecting quality, and acceptance criteria developed to determine if individuals are properly trained and qualified; e) Certificates of qualification clearly delineate (1) the specific functions personnel are qualified to perform and (2) the criteria used to qualify personnel in each tunction; and e
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-2. fQuailty Issut$nce Section.Tr$1ning
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I'>, -. j "..+: QA Procedure QASP;,;.l!1, directs'the development, implementation and documentation of;the QA Section< training program to
? reasonably assure that'LP&L-QA personnel have sufficient ~
knowledge and experiience to lerform assigned tasks at Waterford
. (6 3. Training is implurented through: ~ ,
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-e, Completion 'of 'a QA required reading list;
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Formal classroom training (onsite and offsite) in specific c topical and procedurab areas to enable and enhance
. ;.f performance and effectiveiness;.
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Performance oL cy-the-job training assignments by
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individuals at thefr supervisor's discretion where formal ~ courses cannot* provide the level of training necessary for
, a particular quality-related task;
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m Special training where ' unique skills are needed for
' performance of' specific finctions such as monitoring of 4NDE,weldingandffyeprotection;
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Periodic. training such as th( monthly QA Section training sessions or group cessions ca an as,-needed basis where e' ' ~
' change's, revisions or new-requirements from LP&L QA Program
' ;, " documents, re'gulatory> codes and standards are brought to
- . the attention of QA pEsonuel.il.esdu'ns learned or brrective actigus as aNesu'lt of quality deficiencies or s undesirable programmat.ic trends identified at Waterford 3 and other nuclear generating facilities will be reviewed j-during these sessions.; ;
) 3 l - Th3 Quality Assurance Section, Training Committee was formed on 12/(6/83sto review the goals, objectives, effectiveness, and ..
, implementati9n of the traininlg program for the Quality Assurance Sectione,ilt ,is composed of supervisory members from
*, . Enginek q y/ Systems Development, Nuclear Operat. ions, and Nuclear e Construction QA Groups to act as a steering comraittee to provide management vich, an overview for. evaluating the ef fectiveness and i future ' direction of ths QA Training Program.
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i An evaluatica of the 1983 QA Training Program by this "ad hoc" group stressed three areas of concern for additional improvement: presentation and preparation of training lessons, attendance, and attitude and participation during training. As part of an effort to remain innovative and improve the skills ef QA personnel two new training formats emphasizing professional development and corporate awareness were introduced. Under professional development, college professors and outside consultants provide instruction in stress management, leadership, oral communication, technical writing, time management, problem solving and negotiating skills. To enhance corporate awareness, representatives from various organizations within LP&L and the } Middle South System will occasionally present their group's workscope to provide better understanding among QA personnel of company operations. The success achieved by the Quality Assurance Section in meeting their trafning goals is evidenced in a Good Practice noted by INP0 during a recent corporate acsistance visit (December 1983). While evaluating senior corporate management attention and support of programs for developing experienced, trained, and qualified personnel required for the operation and support of Waterford 3, INPO stated in Good Practice 2.5A-1:
"An excellent continuing professional training program has been developed for the Nuclear Operations Quality Assurance Group. This program is intended to enhance the inspecting, interviewing, and general management skills of QA personnel and has been well received by QA personnel."
- 3. Contractor Training Contractors supplying quality related se: vices to LP&L for which they conduct their own quality inspection and surveillance functions, are responsible for training their inspection personnel and documenting their qualifications under their own QA programs. These programs must meet or exceed the requirements of LP&L's QA Program, including training, before such vendors can be placed on the Qualified Suppliers List and enter into contract agreements with LPaL. QA program assessments of QSL vendors are made through Annual Evaluations and Triennial Audits (refer to Section II.G.1). Additionally, whenever contract personnel are performing quality related work onsite, i?plementation audits of vendor activities are conducted by Operations QA personnel (refer to Section II.G.3).
Contract personnel who perform quality related werk under LP&L's QA Program must be trained in accordance with LP&L Procedures. LP&L managers directly supervising these personael are responsible for ensuring they receive the proper QA training. Cont act personnel performing inspection and monitoring functions are periodically evaluated by LP&L. Evaluation documentation is - retained in individual training files in LP&L Project Files.
7 9 F. Audit / Review Programs
- 1. Nuclear Operations QA Audit / Monitoring Programs
- a. Audit Program As part of its charter to assure th.t the QA Program at Waterford 3 is adequate and being effectively implemented, the Operations QA Group administers an audit program of on-site quality related activities.
The QA Audit Supervisor, within the Operations QA Group, maintains a yearly audit schedule. Audit subject and frequency are based upon 10CFR50 Appendix B, the L?&L QA Manual, Technical Specifiestion 6.5.2.8, Regulatory Guide 1.33, Rev. 2-1978, paragraph C.4, and Regulatory Guide 1.144, Rev.-1980, paragraph C.3. These documents establish minimum requirements which are generally exceeded. For instance, whereas the Technical Scacifications require a,udits of Appendix B criteria to ue conductev at least once s per 24 months, such audits are presently scheduled on a yearly basis. The annual audit schedule is updated every six months to incorporate any changes since the previously issued schedule. For example, when an unscheduled audit is performed it is added to the schedule as a record of the
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audit having been performed. In revising the schedule, the QA Audit Sopervisor considers the need for redirection of auditing efforts in response to problems identified as a result of the audit program, regulatory inspection findings, Site QA Reviews, Safety Review Committee direction, etc. Regularly scheduled audits are supplemented by scheduling additional audits for reasons such as:
- a. Significant changes are maae in functional areas of 2
the QA Program such as significant reorganization or procedure revisions;
- b. A systematic, independent assessment of program effectiveness is considered necessary; or
- c. Verification of implementation of required corrective action is necessary.
The Corrective Action Audit, which is performed twice annually, includes items of noncompliance previously identified to the NRC between the two preceding Corrective Action Audits. Those items are also included within the audit checklist of the Corrective Action Audit conducted one year later to ensure that the corrective action for those items remains in compliance with commitments made to the NRC. i
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The overall . scheduling and audit of unit activities is - performed under the management cognizance of the Safety Review Committeel(SRC) as previously described in Section II.A.l. In addition to periodic' reports of audit activities from the SRC, the Senior Vice President Nuclear Operations-receives the audit reports within 30 days of completion of the audit by Operations QA.. o The audit process is described in detail in QA Procedure. QASP 18.10 " Conduct of On-Site Internal and External Nuclear-Operations Quality Assurance Audits".
.b.- Monitoring Program Monitoring of plant activities is carried out by the Operations QA Group in order to provide additional observation of various aspects of. plant quality related activities.
Monitoring may be initiated for e variety.of reasons. For example, the QA Trending Program may identify an adverse quality trend; audit personnel may note a potential quality problem area outside the scope of their audit; or, during the course'of review of CIWAs or procurement documents, QA personnel may identify areas of questionable quality. Deficiencies identified during monitoring activities are documented throuCh the use of a Corrective Action Report
-(CAR). The origination, tracking and verification of corrective actions for CARS has been previously described in Section } ..B.l.d. The overall monitoring process is covered in QA Procedure QASP 18.9 "Corduct of Nuclear Operations Quality Assurance Monitoring of Quality Activities".
- 2. Plant Quality Group Review and Verification Frocess The Plant Quality Group has responsibility to review and verify implementation of the quality requirements related to Waterford 3 on-site activities.
l ! Plant Quality Inspection a. Quality inspections are performed at designated inspection Hold Points. Quality and Technical Reviews are performed by the responsible department head and Plant Quality Group cm l all quality related maintenance, modification and testing procedures and work packages. This review ensures that the procedure or work package addresses applicable NRC requirements Technical Specifications, applicable quality requirements and commitments made to the NRC. As a result of these reviews, Hold Points are designated in the . procedure / work package, during which a Plant Quality Inspector:
IO l j y 1) ~ Ensures necessary test and inspection equipment.is ~ properly'calibrited before use,
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2). . Checks that the procedure is applicable to.:he work being performed,
- 3) Performs ~ inspection in accordance with the work procedure, ,
- 4) Reinspects items found unacceptable _during previous inspection,
'5) Documents the results on the work instructions, attached data sheets or. Quality-inspection Report, and
- 6) Writes or ~ directs a CIWA be written to correct an
-unacceptable condition unless the iten can be reworked.
Completed work packages /CIWAs are reviewed by the Plant Quality Group to ensure thatLinspections/ verifications were~
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properly performed and documented. 'In the unlikely case that an inspection required by an established Hold Point is missed or not documented, then a Quality Notice (QN) is initiated. The work package will remain incomplete until the QN is verified' as closed by rescheduling and completing the inspection', or producing valid documentation-of the inspection, or obtaining approval to delete the Hold Point.
- b. Hold Points Inspection Hold Points are required whenever there is a reasonable possibility that an undetected deviation could occur that affects plant safety. In determining.
probability for an undetected deviation, post-maintenance testibility, complexity, criticality, and uniqueness of the work being performed are considered. Information concerning Inspection Hold Points is obtained from related design drawings, specifications, codes, standards and
, controlled documents.
The fellowing are examples of activities which would normally require Inspection Hold Points:
- 1) Activitias which could affect the integrity of the reactor coolant pressure boundary of safety / quality related components (e.g., installation and/or setting of pipe or component hangers; bolt-up and torquing of closure studs; installation of locking devices; welding, including fit-up and welding / welder qualifications; heat treatment; and hydrostatic testing.)
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- _ [2) Nondestructive examination.
- 3) 'Clea'liness.and' n foreign material.~ exclusion.. including cleanliness of components with' tight clearance..such as control rod drive mechanism ~ internals and_ major. pump seals,1and system-or component closure following-
. main cenance . --
4)- Characteristics'ofielectrical components or circuits such as cable / routing, splicing, lugging and potting,
' tightness of._ connections, and penetrations and fire ,
- stop installation which'cannot be verified _by -
, post-maintenance and/orimodification' testing.
5)- Characteristics of materials or components,-such'as surface finish, _ hardness, dimensions, leveling . alignment, torque, and clearance'when such. . characteristics are critical to safety ana vien they
-will not be verified in subsequent tests or inspections.
-c. -Quality Instructions Quality Instructions (QIs) are provided -for those cuality
,!- related activities'of-the Plant Quality organization _outside of maintenance, modification and testing procedures / work packagc that' require quality inspection / review. Scae of the key' instructions are: 1). Quality Review of Procurement Documents - The Quality Reviewer (QR), as designated by the Plant Quality. . Manager, conducts a quality review of purchase:and
- s. contract requisitions which include: Local Emergency Orders, Spare Parts Equivalency Reports, Major i Changes, Major Exceptions and Transfer Requests. The t
QR verifies during his review that the procurement document: - a) Meets the guidelines of the Purchsse Requisition-Quality Review Guide, l b) Has a review by the Requirements Engineer to- , ensure the' technical requirements are included and meet or exceed pre'iously imposed specifications, c) Contains applicable references, F d) Contains a statement concerning vendor requirements, 10CFR50 Appendix B requirements, QA. Program requirements, 10CFR21 Reporting, Right of Access and Nonconformance Reporting, and ',, i
e e) C nfirms that the recommended vendor is on the - Qualified Suppliers List. Reviews which result in comments are documented on a Purchase Requisition Review Comments sheet and tracked on the Outstanding Plant Quality Review Comments Sheet until resolved.
- 2) Materials Receipt Inspection - Quality related materials received on site are controlled through ths use of a Materials Receipt Inspection Report (MRIR) initiated by Plant Stores personnel. A plant Quality Inspector will verify on the MRIR that.
a) Identification and markings are in accordance with codes, specifications, purchase orders and drawings, b) The manufacturer documented fabrication and
, testing requirements, c) Protective covers and seals are in place, d) Coatings rad preservatives meet specifications, e) Dessicants are in place and unsaturated, f) No physical damage exists, g) Cleanliness has been maintained, and h) Other checks including weld preparations, workmanship, insulation resistance checks and dimensional checks have been conducted as appropriate.
Items passing review are affixed with a RELEASE tag. Discrepant items are identified with h0LD tags. Discrepancies are documented by Discrepancy Notices which are logged and tracked by 'the Plant Quality Group until resolved or dispositioned by the Material Review Board (MRB) as described in Section II.B.L.b.
- 3) Material Storage Inspection - This instruction provides Quality Inspectors with detailed procedures for verifying proper classification, packing, storage, cleanliness and seFregation of mater!als received.
- 4) Cleanliness Inspections - This instruction provides for cleanliness verification of materials, equipment and components as required by work package instructions. -
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, 5)- HousskeepingTInspections - This--instruction provides . _ .
+ ' < for the use_of: Quality: Inspectionfchecklists to verify.
-prescribed standards _of cleanliness in various plant ~
artas for the purposes of personnel safety, morale,
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contamination-'ation control,_ fire prevention and 16egradation of; plant operability. Discrepancies are t
. note'd.on the Quality Inspection Checklists an? tracked
.and resolved .through the Inspection .
Comments / Resolution; Sheet. -
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d.- ' Plant Quality:Surveillances
. In additionito: Quality Inspections, Quality. Surveillances =
provide for: observations of quality related activities. These' surveys are documented on Quality-Surveillance Report' I' - (QSR) forms. When deficiencies are noted during the Surveillance, a QN shall be written' requiring corrective
. action.- Plant Quality Surveillances provide sampling of:a' t
portion of station activities, whereas Quality Inspections. provida_ for- checks iof specific quality ,aff ecting activities.- 4
. e. _Stop Work The Plant Manager or. Plant Quality Manager may issue verbal
-stop work orders-(SW0s)fto. halt unsatisfactory work and to--
control _the processing, delivery, Lor installation of nonconforming material at Waterford 3. A' verbal SWO is , followed up_with a written SWO which is documented on at
- . SWO form, and logged-for tracking. Notification of the SWO is made to the Senior Vice President Nuclear iperations,
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Corporate QA Manager, Safety Review Committee, Control Room Supervisor, individual- company involved, Plant Manager, applicable departmentisupervisor,-and the Plant Operations I' Review Committee. When the deficiency'is corrected, or. sufficient steps have been taken to ensure'that further noncompliance will not occur,~a Stop Work Order Release (SWOR) ' form is. issued by the Plant Quality Manager to allow i-i work to resume. A SWOR form notes the corrective action taken and the reason for release. i l - G. Control of Contractor Quality Related Activities.
- 1. Evaluation of Supplier's Quality Assurance Program Suppliers providing safety related material or services must be on the LP&L Qualified Suppliers List (QSL). Before a vendor car.
, be placed-on the QSL, that vendor must be evaluated for acceptability by the LP&L Engineering / Systems Development QA [_ Group.- . j
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.l An initial evaluation of a prospective contractor is performed j
.by reviewing the contractor's:
- a. Current quality assurance program manual, procedures and records;
- b. Capaoility to conduct quality activities as revealed through examination of the facilities for performing-such work and ability of- the supplier's personnel;
- c. Past performance _'oased on experience that LP&L and other users have gained using identical or similar products and services.
Based on results of the above evaluatian process, a supplier is classified:
- a. Acceptable - no questions / concerns were raised during evaluation, or questions / concerns have either been resolved or have an insignificant impact on the item /scrvice to be provided,
- b. Unacceptable - the supplier's program doesn't meet procurement document requirements, or is not adequately o implemented and review questions rat satisfactorily addressel/ resolved.
- c. Conditionally Acceptable - only certain portions of a supplier's program are acceptable and purchase activities are limited to restrictions as imposed by_the Engineering / System Development QA Group and noted on the QSL and are to be reflected in procurement. documents. Full acceptability will be based on satisfactory supplier resolution of questions / concerns.
Once a contractor is on the QSL, a documented evaluation of the supplier will be performed annually and kept in that vendor's file. While an audit is not necessary for a satisfactory annual evaluation, an audit must be performed evsry three years for a vendor to remain on the QSL.
- 2. Conduct of Contractor Quality Assurance Audits
- a. Off-Site QA Audits The Engineering / Systems Development group is responsible .
for ensuring all QSL iisted contractors' offsite activities are audited to requirements of 10CFR50 Appendix B and LP&L's QA Program. Either they themselves will audit these contractors, or a vendor audit group will be contracted __ which has been qualified to LP&L's QA Program to conduct
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these audits. Andits will be conducted triennially per NRC Regulatory Guide 1.44.
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- b. On Site Auditing and Monitoring of Contractors-The Nuclear Operations Quality Assurance Manager directs audits of those organizations not within LP&L that are
. performing quality-related services at Waterford 3. These-type of contractor audits are designated as "On-Site-External' Audits" and are conducted as previously described in Section II.F.1.a.
Periodic monitoring of on-site contract:r activities is done through the use of Monitoring Reports as assigned by: the QA Analysis Supervisor under the Operations QA program previously described in Section II.F.1.b.
- 3. Deficiency Reporting by Contractors All vendor personnel performing on-site quality inspections of their company's work under LP&L's QA Program are required to report deficiencies identified for inclusion on a CIWA. This includes deficiencies discovered outside the scope of work-being performed. A CIWA, which documents a deficiency and its corrective action / rework, is approved and tracked by LP&L management as described in Section II.B.l.a. Corrective action verification.is provided by post closure review of the CIWA by the Plant Quality Group.
H. Station Modification Program The purpose of the Station Medificatica program is to provide a mechanism through which design modifications to Waterford 3 are controlled and tracked. The Station Modification Package serves as a comprehensive, stand alone decign change document which has undergone the appropriate interdisciplinary reviews. The process assures that no changes are made to the plant structures, systems and components which may introduce an unreviewed cafety question per the criteria delineated in 10CFR50.59. Any individual with the concurrence of che department head may request a design modification. Reasons for the change could include enhancement of the plant structures, systems, or components as a result of engineering preference, regulatory requirements, licensing commitments, ALARA, Human Engineering Design considerations, etc. { Upon management approval of the request, a Sta: ion Modification Package (SMP) is assembled and receives appropriate interdisciplinary review. During the course of the design and review process checklists are used to ensure that, among other things, generic criteria such as separation, failure effects, fire protection, etc., are taken into account. The LP&L Quality Assurance Program requires that documentation appropriate to satisfy 10CFR50 Appendix B will be generated and retained. l I
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Typical SMP Contents include: . -
- 1. : Sununary. Functional Descriptionj 2.' List'of-Attachments .
. La) Purchase' Orders / Requisitions i-b) . Recommended Spare Parts c) New or. Revised Drawings / Description Documents / Tech Manuals / Equipment Specification / System
Description:
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. d)' Vendor Information-
- . e). : Design Calculations / Analyses-l, ~f). Work Procedures
- 3. . List of[
References:
4 '. Bill of: Material
- 5. ' Installation Instructions l . 6.- Examinations (e.g. NDE requirements, PSI /ISI surveillance requirements)
- 7. Testing'(including acceptance criteria)
- 8. Nuclear Safety Evaluation checklist (10CFR50.59 review) 4 .
Modification is performed via the Condition Identification and Work Authorization (CIWA) process described in Section.II.B.1.a. Detailed. Construction . Packages (DCPs) are prepared for work activities. Pertinent design and ' reference information (e.g.' isometric ' drawings, ;
; engineering. instructions, code type testing requirements, l installation procedures) is included in the DCP as well as
( instruc'.Lans for implementation documentation. Acceptance j criteria / tests / checks are developed and included as part of the DCP prior -o implementation. With the exception of minor changes, alterations (or field changes) i to the DCP. may not be made without approval of a revision to the SMP. L For minor changer, the Action Engineer may authorize a Detailed Construction Package Change (DCPC) in which case a detailed description of the change is documented prior to Laplementation of the change. All DCPC documentation is retained as part of the work package and subject to post-implementation review. Verification of implementation is first performed by the Station i Coordinator and the Action Engineer who had the responsibility for , developing the package. The Action Engineer assures that all work
- j. was accomplished according to the 1DH' and that acceptance criteria 4
are met. Control Room controlled drawings are redlined to reflect the change. The Action Engineer then initiates a Modification Project Closeout Revie .orm, and forwards it to the SM Coordinator 4 r + , 4----,,-- --,,r,m. _ ,-,,dm., , + , - -. , , e ,. ---+-,-g., - + ,..-,---,w-c-wy-,,w,--- -,w.--w---., - . - , . -,re,,e4 w
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i (SMC). The SMC forwards a Work Completion Notice to all af fected _
~ disciplines so that appropriate documents are revised. Completed Document' Update Forma are returned to the SMC to certify that all. -
4 affected drawings, procedures, programs,'and/or training plans have been revised and approved. At this time the CIWA-is closed and the SM. Closeout Review form initiated and sent to the Systems Engineering Department-Head for review and approval of. the Modification Project Closure Review form. See Sectica II,I.3 for quality review and storage of.5MPs. I. Records
- 1. Project Files Project Files is the focal point foristorage and maintenance of.
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uncontrolled records and documents. The filing system used is a computerized document retrieval system. Completed records forwarded to Project Files -are indexed enf the computer, then microfilmed and stored by Film Access Number. This number indicates the roll and frame' number of a particular' document or-
.its hard copy location.~ . Records are thus effectively filed under document number, record type, date, title, vendor, subj ect , equipment number,-etc., allowing a user to retrieve ;
documents in a timely manner. _ _ . Records processed by Project Files are received under a standard transmittal form which lists the contents forwarded. The records transmitted are inspected to ensure that all of the records on the transmittal form are'present, complete, and validated. If the records _are complete and' agree with the transmittal form, then the form is signed by the package reviewer, filed, and a copy sent to the originator. Unlimited access to Project Files is granted only to personnel assigned to the Project Files Group. This minimizes the possibility of lost / misplaced records by personnel who have not
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been indoctrinated in the proper procedures for control.of ducuments. The Project. Files Supervisor may authorize temporary access when individual requirements cannot be handled by the Project Files personcel. QA records may be accessed by request for work / review, but may only be reviewed in designated controlled areas.
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. 2.: Document. Control"
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Document Control is the organization responsible for processing . l controlled documents such as' approved-drawings, specifications,-
-technical manuals,LFSARs,-SMPs and some procedures. . This-process includes receiving, recording,-distributing, updating-and retrieval ~of those~ documents affecting quality to ensure only the ' latest applicable 1 revision is used for operation an'd-
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maintenance at Waterford 3. Controlled issue is maintained by. the use of transmittal forms which mu'st be signed and returned l 1 by assigned copy holders'on established distribution lists. l -Direct access to files maintained by the Document Control'is l limited to group personnel and their supervisors. '~- 3.- Records Quality Review Quality-related Station Modification Packages (SMPs) are ! reviewed by the Operations QA group before final closure and trcusaittal to Project Filea. . A Quality Reviewer-(QR) completes
- a QA Review Checklist on the SMP to ensure that records establishing proper review and other necessary records are
- retained. :The QR review scope ensures that documents required by the_ SMP index and controlling procedures are included, proper review and approval is indicated on the records, applicable codes and-quality standards are identified, test and inspection requirements are documented, and aafety evaluation and design verification is performed.
4 Comments from this review are tracked and closed out on a 4 standard Procedure Review Comments sheet, ensuring completeness of the SMP. The Checklist, comments sheet and any additional records generated by the QR's review are filed for ctorage. Similarly, qcality related documents generated by the Plant Quality and Quality Assurance groups in the performance of their i duties are reviewed and retained in Project Files. -These records include audit reports, nonconformance reports, receipt
- ' inspection reports, CIWAs, QNs, DNs, Stop Work Orders, QC
; surveillances, QC Inspector certification, hold tags, i conditional release tags, various NDE documents, calibration
- records, and NDE personnel qualification and training records.
- (NOTE
- Some aspects of Records Quality Review, particularly l records storage, are not yet fully implemented due to their
- j. recent adoption by Wateriard 3.)
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- 4. Status During the construction phase, records management was primarily handled by the architect / engineer. As a result, although current records are' handled and-processed as described above, there remains a backlog of construction phase records to process through the LP&L Lacords System. Additionally, to assure centinued high quality in records storage and retrieval, LP&L management is : evaluating the current recorde management process for Waterford 3 to identify any areas needing improvement. _It is expected that appropriate racommendations of this evaluation will be initiated by November 30, '1984.
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