ML20237L110
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'/ 'e UNIT ED STATES
.!". e NUCLEAR REGULATORY COMMISSION
'* $y ,. $ WASHINGTON, D. C. 20555
%, , JAN 131987
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'PEPODANDUM FOR: John G. Davis, Chairman i Comanche Peak Report Review Group i
FP0M: Vincent S. Noonan, Director Project Directorate No. 5 Division of PWR I.icensino-A, NRR
SUBJECT:
INFORMATION FOR COMANCHE PEAK OTA REPORT 86-10 (REV.11 This is a revision to my memorandum to vou of January 6,1987. This revision is beino issued to include information concernino two special panels that were established for Comanche Peak (Contention 5 Panel; Intimidation Paneli. This information has been added under the response to reauest No. 2. Since this memorandum contains all the information previously transmitted, this supersedes my memorandum of Januarv 6, 1987 The attached information and included documents is hereby transmitted !
to you as reouested in your memorandum of December 22, 1986. In addition, enclosed for vour information is a summary table of the issues raised in the OIA report that was prepared by CPPO and IE prior to the formation of the Comancha Peak Report Review Group.
Much of the work associated with the Comanche Peak proiect is unique and complex. For this reason, we suagest that we meet with you at some point to discuss these activities and this report, e
/;"? / .. /
- /
t S. onan,' or
,8 Project Directorate No. 5 Division of PWR 1.icensino-A, NRR 1
Attachments: .
l
- 1. Responses to Questions )
.2. Summary Table cc w/ attachments: )
C. Heltemes, Jr. l G. Arlotto ;
R. Erickson J. 1.ieherman C. Paceri911n, Recion Til l J. Tavlor R. Part.in, Reaion IV P. Denton R. Vollmer T. Novak 8708200084 870312 PDR ADOCK 05000445 G PDR U
____.___-_-_.._-_.----_-_-_-_a-
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At+achment 1 "
PESPONSE T0 OVESTIONS i
Recuest No. 1:
A report on the extent to which the activities of the Comanche. Peak Pro.iect Office (CPP01 covered the scope and requi ements of the current TE quality assurance construction inspection prograg.. This report should identify that
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which the CPP0 has inspected and specify whether that which-CPPO has insnectad is recuired to be inspected as'part of the IE ouality assurance construction inspection program.
Response
It is important to.urderstand from the outset that the CPP0 was formed to serve a purpose that differed from completion of all or parts of the IE con-struction inspection program (MC 2512). However, as the CPP0 efforts evolved and as the licensee,' Texas Utility Generating Company (TUGCO), responded to CPP0 reports, findings, and reauests for information, a comprehensive evaluation of the cuality of construction and desien of Comanche Peak has been undertaken. Not all the.results of this evaluation has been completed, but the scope of this effort has been so extensive that it should provide a complete and accurate picture"of Comanche Peak desion and construction adecuacy. l To help understand how the CPP0 efforts and TUGC0 corrective actions have evolved and how these efforts could be related to the standard IE 2512 proaram, the following inforretion is presented.
On March 1?, 1984, the NRC Executive Director for Orerations directed that the L~ NRR Comanche Peak Pro.iect Director be responsible "to assure the overall )
coordination /4rtegration of licension issues and to assure that the issues are resolved on a schedule to satisfy hearino and licensina decision needs." As a result of these directives a series of inspectinrs were conducted at the )
, ?
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direction of the CPPO, The first of these inspections was conducted by a aroup _
! a called the Special Review Team (SRT), whose charter was to determine whether
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1 site construction activities could continue while the CPPO resolved allegations )
i of improper construction and ouality assurance (0A). The team concluded that l i
the licensee's procrams were being sufficiently controlled to allow continued I olant construction while the iRC completed its review and inspection n' the !
\
fecility. The team evaluated the implementation of 1.icensee's manaaerent con-l trol of the construction, inspection ard test programs and fcund it generallv !
l effective and receivino proper manaoement attention, The team also probed the axtent of problems with design documentation and construction deficiencies.
Roth strengths and weaknesses were found, but no weaknesses of such severity as to warrant a halt in construction were identified. The second of these irspec-tions was conducted by a nroup called the Technical Review Team (TRT) whcse charter was to resolve specific issues such as allegations and concerns similarly dealine with (principally) the adecuacy of construction and the TUGC0 OA Prcaram, As mentioned above, these inspections were not performed to satisfy the TE ouality assurance construction inspection program but were per-formed to assure the CPP0 that TUGC0 had met its FSAR commitments, The CPP0's ma.4cr inspection efforts to resolve the approximately 1000 allegations were conducted at the Comanche Peek site from July to October 1084, The CPP0's evaluations of the accuracy and safety significance of these allegations were documented in five Supplemental Safety Evaluation Reports (SSER'sl in the araas of electricel/ instrumentation and test proarams (SSER d7',
civil / structural and miscellaneous (SSER API, protective coatings (SSER #9),
mechanical /pipino ISSER *101, and 0A/0C (SSEP #111,
3 TI'GC0 responded to staff ccncerns resulting from these allegations by forming --
the Comanche Peak Response Team (CPRT) in October 1984 The CPRT scope and effort expanded greatly over the ensuing months as the SSERs were published and as concerne about the facility from all external sources (such as the NPC staff; the ASI.R; and TUCCO's Independent Assessment Proaram contractor, CYGNA Enerqy Services} became better understood. The CPRT issued a' program plan to describe its efforts, and this plan has undergone three ma.ior revisions to j date. -Revisior 3 was issued by TUGC0 in January 1986, and was reviewed and approved by the CPP0, subject tn some items for continuino sta#f review as the CPRT implemented its program. 'The CDP 0 documented the review and approval o# I the CPRT program plan in SSER 13, published in May 1986. The CPRT program plar contains subtiered action plans to resolve issues from soecific greuos.of l
allegations as well as a major re-inspection of construction and documentation that was developed and implemented to provide assuranca that the. plant was constructed safely and that there was' reasonable assurance that any defects were discovered and corrected. The CPRT effort has been implemented by an independent third-party consultant and the effort is beino closelv monitored hv RIV and CPPO engineers and consultants.
Table 1 summerizes those inspections conducted or directad by the CPP0 with the explanatory information.
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R i ni em pa G E sr O I ng R i o P nf r oo f p E i o I t n n co eo Y ei pi B pt s yt sce t c D nel u E i su t r R - d nt I l so es U asm rn
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n t r oi oi C f e a c d u d d f n u s f u as en en f aol n cn sor h s co co o ti i i r e uys t A nc nc st t f e 0 a a y t n ic scn C y mt mt c ee ao cn sao Gt rl rl a h pe h c eo i uc Ui oi oi u t s t pc q Tl t f u f u q y s T ed ac nb nb e fi 0 a d Rd n f ue o- o- d o P m en T aa iQf cs cs a P t a n f a a l a C n a t eno eye e e e a o gs unoi nr no no n vgs yi in oiit i on it it im on ft sc t o si erct mt a mti rc m
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- - 0 - r 0 1 - 2 euo E 4 71 6P 1 1 1 1 esf T s n A no D e ec s s c m n t i dR a o r l n m e pn o a T s a m p e e el a p h s T w RP r u t r e g S s w i k m o t w yt e v aa r & r e b c i e er P o i v R P g s p v d p N e o A y p e esw O R l er Q a u AR t n I a hP r S Q c i T l c c - T m u C a i n- w & - a dVr E i n a 0e e r nI P c h mm Ci l C T g o-S e c oa G v b A R o CR N p e C e U e a V P r I S T
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-4 Reouest No. 2: _
As a separate section of the report, an identification of items and activities of the program implemented by CPP0 that are in addition to or different frem
' those in IE- program.
Response
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'1 The; inspections conducted end directed by CPP0 were different from the IE inspection procram in that the IE program is implemented over the entire >
construction period o# the plant encompassing many in-process construction activities. The CDP 0 inspections did not start until the plant was essentially completed. This was the ma.ior difference between the twc inspection programs. >
in spite of the fact that the purpose of the CPP0 efforts was different from l
the TE construction inspection procram, in our view the activities of the
- CPPn program are in essential agreement with the TE program except that the ma.ior CPP0 program (the license's third-party CPRT effort) included a fer larger sample si7e (in some instances 100%) than is implemented in the IE proaram. While the CPRT effort was being performed, Recion IV inspectors overviewed its activities and in addition performed independent inspections to confirm CPRT results. To put the IE and CPDT efforts in perspective, a review of the respective man hours required is appropriate. The IE inspection effort reouires approximately 10 - 15,000 man hours. The licensee CPRT quality of construction effort reouired approximately 200,000 man hours involv-ino some 650,000 observations relative to construction hardware and the review of 9650 hardware-related packages of documents and records.
. 6 fr addition to the above, the Executive. Director for Operations established _ .
two special panels hacause of the complexity and special problems at Comanche Peak: the Cortention 5 Panel and the intimidation Panel. Each is described below.
Contention 5 Panel The sinole remainino contention for the OI. hearing at Comanche Peak was, and is, Contention No. 5, which states that applicants' failure to adhere to QA/0C provisions of the construction permit and Appendix R to 10 CFR 50 has raised substantial cuestions as to the adecuacy of construction. As a result, the Commission cannot make the findinos reovired in order to issue an 01..
To oroperly evaluate this contertion and to prepare a comprehensive staff position, the EDO established the Contention 5 Panel on December 24, 1984 The panel was composed of senior NRC maragers together with experienced advisors. The Panel provided valuable cuidance to the CPP0 after its establishment. With the formation of applicants' Comanche Peak Pesponse Team Procram Plan and its approval by CPP0 in SSER 13, the activities of this panel has correspondingly diminished, and the CPPO is_ recommending that it be disbanded at this time. '(Its charter is listed in response to Request No. 5.)
i intimidation Panel The Intimidation Panel was also established by EDO on December 24, 198d. This I panel was established to evaluate whether intimidation or harassment of
- 7 employees or contractors occurred and make a determination concernino the - -
effects on the overall OA/0C procran at Comanche Peak. This panel issued its report to CPP0 on October 18, 1965, and was sent to the applicants on November 4, 1985. (The report is listed in response to Recuest No. 5.)
Pecause the panel has finished its work, CDP 0 is currently recommending that it he disbanded also.
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. 8 Deauest No. 3:
The CPP0 inspection activities described aaainst the requirements o' the TE program, i.e., the specific item inspected, the skills applied, the schedule of the CPPO activities, the site activities, the " depth" of CPPO activities, the size of." samples," etc. In sum, the CPRRG desires an expression bv CPPO of what inspections CPPO actually performed in all its aspects, and a comparison with what the IF construction oualitv assurance program reauires or describes.
Response
i i
The ma.ior inspection - orts by the CPPO were the inspections conducted by the Technical Review Team (TPT) and RIV inspections o' the Comanche Peak Resocose Team Program Plan * (CPRT PPl. The TRT inspection was principally concerned with identified issues and concerns and specific allocations. Over 1000 allegations were reviewed by the TRT. This resulted in a series of findinas l and concerns that led to the CPRT-Program Plan. The CPRT PP was the licensee's !
independent third-party effort, monitored bv RTV and CPPO, to provide assurance that'the Comanche Peak Steam Electric Station (CPSES) was constructed in such a manner that reaulatory requirements and licensee commitments were met.
4 The IE construction quality assurance inspection proaram is desianed to be implemented over the life of the construction praiect and therefore most of its procedures require in-process inspections whereas the CPRT-PP effort was implemented at the conclusion of construction. The orocrams should best be considered independently and the satisfactory completion of either procram should provide confidence that the plant was safely constructed and that ,
concerns related to the cuality of construction of the hardware at CPSES would be identified, evaluated, and resolved.
- The CPRT PP includes both a desian and construction reinspection program. '
This report and documents address only the construction portion.
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'. In addition to the two ma.ine inspection efforts there also were a series of U ___ _
inspections. One review involved a comparison of the TUGC0 OA Program aaree- )
)
. ment with oricinal PSAR commitments as well as current NRC OA requirements and ]
resulted in the conclusion that the TUGC0 OA Program is satisfactory.
The specifics of each inspection for the TPT and CPRT PP effort are available (
in the reports cf the TRT (SSERs 7-11) and the inspection procedures and results repcrts of the CPRT PP effort.
The followirg is a list of the CPPO-related inspections and a brief synopsis of the purpose of each inspection: l Special Review Team - The purpose of this inspection was to evaluate the implementation of the liennsee's manacement control of the construction. {
inspection and test orocrams and also the extent of problems with desian
' documentation and construction deficiencies. The Special Review Team was composed of nine cualified NRC inspectors from Reof on II, lead bv CPP0, which conducted the field inspection in April 1984 The inspection effort entailed about 640 staff-hours.
Technical Review Team Reports - A series of five SSERs dealipp with various technical concerns and allegations about construction practices at Comanche Peak.. These inspections covered the technical disciplines of electrical and test procrams, civil / structural, coatings, mechanical /pipino, and quality assurance. A sixth report (SSEP 13) evaluated the licersee's CPRT Progran l I
Plan. The TPT was composed of over 50 technical experts drawn from NRR and i i
l the Reaions, rational lat' oratories, and consultino organizations. TFe TRT
10 ccoducted most of its onsite inspections during five, two-week sessions from _-
July throuch October 1984. We estimate the on-site inspection effort involved about P0,000 staff hours.
l
- Comanche Peak Response Team Procram Plan - A licensee-developed corrective action proaram covering both desion and construction, includina specific concerns and issues, that was aporoved bv the CPPO. A review and evaluation of this plan is contained in SSER 13. In addition to the CPRT efforts that are resolvina specific, identified issues, a separate effort consists of a comprehensive re-inspection of comoleted hardware and a doc;m9ntation review of those hardware attributes that are either inaccessible (like reinforcing steel) or non-recreatable (like PT of a weld root pass). This iicensee re-inspection effort has been accorolished from August 1985 throuch December 19P6, Sv about 150 ANSI N45.2.6 oualified third-party OC inspectors. The effort consisted of samples from about thirty populations listed below
El.ECTRI CAI.
l Conduit Cable Cable Trav Electrical Equipment Instrumentation Equipment I.iahting NIS Cable Terminations MECPANICAI.
PVAC Ducts and Plenums PVAC Equipment Installation Field Fabricated Tanks Mechanical Ecuipment Installation I.arge Bore Pipina Configuration Small Pore Piping Configuration
' Pipe - Welds and Paterial Piping System Polted Joints / Material Tubino Welds /Paterial
r
- 11 STRUCTV,R3 Concrete Piacement Structural Steel I.iners
' Fuel Pool I.iner Fill and Backfill Placement Cement Grout Epoxv Grout 1.arae Rore Pipe Supports - Picid f.arce Pore Pipe Stonorts - Non Rigid Small Fore Pipe Supports I.erge Bore Pipe Whip Restreints Instrument Pipe / Tube Supports CAT 1 Conduit Supports HVAC Ouct Supports Equipment Supports Each of these populations consists of many items recuiring re-inspection or document review (attributesi. The CPPT selected a random statistical sample from each popalation (at least 60) and an additional sample of items from safe shutdown systems. A description of the samplino plan is described in SSEP 13, Section 2. The third-party field inspections involved about 200,000 s ta f f-hours and the inspection of abcut 650,000 items on inspection checklists. >
During the course of these third-partv re-inspections, R-IV developed a special inspection progran whereby their inspectors and consultants were to witness about 10% of the re-inspections and were to separately inspect about 5%.
TUGC0 Guality Assurance Proaram Review - An inspection of the Texas Utilities Generatino Company (TUGCO) Quality Assurance Program to determine if the program meets NRC requirements. This effort consisted o# an item-by-item comparison of TUGCO's 0A Procram with cast and present Standard Review Plan requirements, and was conducted by three cualified NPC OA inspectors from IE expending about two and one-half staff months in late 1985.
. 12
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Cable Trav and Hanaers Report,- A special inspection conducted to evaluate tha conformance of CPSES, l' nit l'as-built cable trays and supports with drawings.
TI'is inspection encompassed about 220 staff-hours in November 1985 and was j conducted by one NRR engineer and three consultants experienced _in structural engineering. The team inspected 32 cable tray supports.
HVAC and Supports - Report - A.special inspection conducted to evaluate the conformance o' CPSES, as-built HVAC support system with drawinos. This inspection was conducted by four NPC consultants in February 1986. It encompassed 32 HVAC supports and 250 staff-hours.
CPRT PA Procram Review - A review of the CPRT Procram Plan cuality assurance program to assure that all elements of the CPRT PP are conducted in accordance with applicable NRC requirements and licensee commitments. The CPRT PP is an extensive corrective action procram developed by the licensee with review and concurrence by the NRC-CPPO. A review and evaluation of this OA proaram is contaired in section 4.0 of SSEP 13. This review was conducted hv 3 NRC OA inspectors and reviawers and took 2 staff months, mostly in December 1985.
Pecuest No, d:
For each CPP0 inspection requirement, provide a reference to a CDPD document conveyino the requirement includino the relevant pace number (s).
Response
Each inspection conducted by the CPPO was #or a specific purpose as described and was not in response to a specific irspection requirement such as exists in the IE Marual.
l 13 1
Recuest No. 5: _. . <
A copy of appropriate reports or relevant pages #cr each of the CPP0 inspections.
Response: .
The reports submitted in response to this recuest are the followino:
Atta c hmen_ t,* Document 1- frecial Review Team Report I
?- Technical Review Team Reports - SSERs, 7, 8, 0, 10,11, and 13 3- Comanche Peak Response Team Proaram Plan and completed results reports. (Some of the results reports are not yet completed.)
4- Tl!GC0 Ouality Assurance Proaram Review 5- Cable trav and banger inspection report 6- PVAC and support inspection report 7- CPRT-0A Program Peview Report 8- Contention 5 Panel (memo from E00 dated February 28, 1985, superseding earlier memo dated December 74, 1984) .
i 9- Intimidation Panel (memn from EDO dated December 24, 1984)
)
10- Report of Intimidation Panel (letter, NRC to TUGC0 dated November 4, 1985) !
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- These attachments are ornvided to the Chairman, CPRRG only becauSP of their volume.
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, Attachment 2
SUMMARY
TABLE OF ISSUES RAISED IN OIA REPORT 86-10 Introduction Prior to formation of the Comanche Peak Report Review Group, both NRR and IE reviewed select portions of OIA Report 86-10 and prepared sumary tables.
NRR's effort focused on listing all of the issues raised in the OIA report, and beingdetermining)whether addressed which may similar issues envelope thehave been raised particular issue, ifand notaddressed specificO1y, (or are generically. This could be done through work performed by the TRT, CPRT '
-(licensees response program) or other licensing efforts, etc. IE's effort focused on a select nuraber of issues raised in the OIA report (primarily the 16 identified in the summary) and the determination of whether final resolution of the issue-(i.e. violation, unresolved item, open item, etc.) was indeed proper and safety significance. Attached is a table which sumarizes the initial work ~of the two offices.
Notes i Issues underlined are the 16 issues referred to in the OIA sumary report.
Determination of whether resolution was proper and safety significarice was made by IE for a select number of issues.
Current status was provided by Region IV. )
~ 1 Goldberg (G) and Scarbourg (S) determinations are provided to show the ;
varying opinions of the staff when making determinations of whether an j issue should be a violation, unresolved item, etc.
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Comments section have been raised andisaddressed provided or by(the CPP0 are being to identifywhich addressed) that may similar issues
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envelope the particular issue; if not specifically, generically.
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. . E87 , g k NUCLEAR REGULATORY COMMISSION ,
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MAR 121981 pya .,
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. _ - MEMORANDUM FOR: John T. cot 11'n's,
-RWpientf Administrator
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. - - . Region IV' ; .
~~
. Harold R. Denton, lirector .-i. OS Office of Nuclear ieactor Regulation Richard C. DeYoung, Director
- Office of Inspection & Enforcement FROM: William J. Dircks Executive Director for Ope ations
SUBJECT:
COMPLETION OF OUTSTANDING REGULATORY ACTIONS ON COMANCHE PEAK AND WATEP.FOR'S Construction of the Comanche Peak and Waterford facilities-is nearing completion. There remain a number of issucs that need to be resolved before the staff can maks its licensing decisicas. Tht, issues remaining for these plants are quite complex and span more the ,ne Office. In order to assure.
the overall coordination / integration of : a issues and to.cnure issues are resolved on a schedule to satisfy he ,,
a d licensing decision needs, I am directing NRR to manage all necessary riRC actions leading to prcmpt licensing decisions. Darrell Eisenhut, 'Jirector, Divirion of Licensing, NRR j is being assigned the lead responsibility for this activity. He will '
coordinate the efforts of NRR, IE, and Region IV, and will coordinate this activity with 01 and OELD. Prior to any of the affected Offices undertaking r.najor activities (e.g., inspections) or raking decisions on these plants, that activity should be concurred in by MR.
We are presently in the process of assigring a dedicated senior manager to
- assist Mr. Eisenhut in the management of these activities.
1 The first phase of this program will be the identification of issues needed l to be resolved for each plant prior to hearing and licensing decisions.
Once the issues have been identified a Program Plan for resolution of each item should be developed and implemented. The Program Plan should address the scope of the work needed, the identification of the responsible line organization, and the schedule for cogletion. In principle, this effort -
will therefore be similar to the effort undertaken regarding the allegation ~
review on Diablo Canyon except that this effort should encompass all licensing, inspection, hearing, and allegation issues.
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, 9-Each affected Office will assign a full time senior manager to work with NRR to define, schedule and complete the issues. I expect these managers to be - 1 identified by each of you within a few days. All affected offices should
. _- provide dedicated resources and..qive' their full support to this effod, to ass fre that al.1 ex.isting issue's are exped.iticusly handled and all new issues are pr:nr.ptiy provided .to NRR, so'~Estnotjo delay the ' licensing decisions. In
..e__. .
'- addition, copies' of all infonnation, doctinents, depositions, etc. should be promptly prerided to HRR to ensure a coordinated approach.
I. anticipate that the approach utilized here will be necessary for a number !
. of' >pcoming OL prcjects, and am directir.g NRR to take the lead for carrying cet this activity.
e s Willi- J. Dircks Executive Director for Operations cc: G. Cunningham, ELD
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C____________.__._____________.___ _
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. /'. . .. '4,3 UNITED STATES c/h 5- / /
8 -o NUCLEAR REGULATORY COMMGSION a ,I - wasmucTow. o. c.2osss Eg'&
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- October 17, 1984
[
. . .. MEMORANDUM FOR:' Office Directo'rs' ". ,'
" Regional Administrators FROM: William J.'Dircks Executive Director for Operations
- 6
SUBJECT:
COMANCHE PEAK PROJECT DIRECTOR 4 I have appointed Vincent Noonan as the Comanche Peak Project Director effective innediately. He replaces Thomas Ippolito, who resigned on October 4, 1984. Mr. Ippolito's resignation was for personal reasons and not beca u e of any concerns regarding the Comanche Peak project..
Mr. Noonan will continue to coordinate and direct the overall licensing review effort that had been begun following the same organization as
- before. (See my memorandum dated March 12, 1984, copy attached).
The technical review team will remain intact. Mr. Noonan will report to Darrell Eisenhut with regard to the overall adequacy of the.
Comanche Peak project.
Mr. Robert Martin, who has assumed the position of Regional Administrator for Region IV, will assure the continued review and coordination of construction and operation issues for the Comanche Peak project.
/
bO William J. Dircks t .
Executive Director for Operations
Attachment:
Memo, Dircks for Collins, Denton,
& DeYoung, dated 3/12/84 i
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[ -:z7 [ i NUCLEAR REGULATORY COMMISSION 3 WASHINGTON, D C. 20555
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% . , , . . -} -t gC GE? 15 ma Dockets: 50-445 50-446 l
Texas Utilities Electric Company 1 Attn: M. D. Spence, President, TUGC0 )
Skyway Tower * )
)
400 North Olive Street i Lock Box 81 .
Dallas, Texas 75201 <
Dear Mr. Spence:
SUBJECT:
COMANCHE PEAK REVIEW On July 9, 1984, the staff began an intensive onsite effort designed '
to complete a portion of the reviews necessary for the staff to reach its i Ldecision regarding the licensing of Comanche Peak Unit 1. The onsite
' effort covered a number of areas, including allegations of improper construction practices at the facility. l 1
The NRC assembled a Technical Review Team (TRT) responsible for evaluating
-most of the technical issues at Comanche Peak, including allegations. The TRT has recently identified a number.of items that have potential safety .
implications for which we require additional information. These items are listed in the enclosure to this letter. Further background information i
l regarding these' issues will be published in a Supplement to a Safety i
Evaluation Report (SSER), which will document the overall TRT's assessment of the significance of the issues examined. -
The items in the enclosure to this letter, which are in the general areas of electrical / instrumentation, civil / structural'and test programs, cover only a portion of the TRT's effort. The TRT evaluation of items in the areas of mechanical. 0A/QC, and coatings, and its consideration of the programmatic implications of these findings, are still is progress. A summary of these -
issues will be provided to you at a later date. I You are requested to submit additional information to the NRC, in writing, including a program and schedule for completing a detailed and thorough assessment of the issues identified. This program plan and its implemen-tation will be evaluated by the staff before NRC considers the issuance of an operating license for Comanche Peak, Unit 1. The program plan should address the root cause of each problem identified and its generic implic-ations on safety-related systems, programs, or areas. The collective significance of these deficiencies should a'so be addressed. Your program plan should also include the proposeo TUGC0 action to assure that such problems will be precluded from occurring in the future,
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SE? 18 n34 Mr. M. D. Spence _ . . .
This reouest is' submitted to you in keeping with the NRC practice of promptly notifying applicants of outstanding information/evaluatien needs that could potentially affect the safe operation of their plant. Fu r'ther requests for additional information of this nature will be made, if necessary, as the activities of the TRT progress.
. Sincerely, -
t , ,
a 'eli G. Eisenhut,' DIN [ tor Division of Licensing, NRR
Enclosure:
As stated -
cc w/ enclosure See.next page l
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COMANCHE PEAK l Mr. M. D. Spence President Texas Utilities Generating Company 400 N. Olive St., L.B. 81
~
Dallas, Texas 75201 cc: Nicholas S. Reynolds, Esq. Mr. James E. Cummins
- Bishop, LiberTnan, Cook, Resident Inspector / Comanche Peak
- i Purcell & Reynolds Nuclear Power Station l 1200 Seventeenth Street, N. W. c/o U. S. Nuclear Regulatory l Washington, D. C. 20036 Comission P. O. Box 38 Robert A. Wooldridge, Esq. Glen Rose, Texas 76043 Worsham, Forsythe, Sampels &
Wooldridge Mr. John.T. Collins 2001 Bryan Tower, Suite 2500 U. S. NRC, Region IV ~
Dallas, Texas 75201 611 Ryan Plaza Drive '
Suite 1000 Mr.. Homer.C. Schmidt Arlington, Texas 76011 Manager - Nuclear Services '
Texas Utilities Generating Company Mr. Lanny Alan Sinkin Skyway Tower . 114 W. 7th, Suite 220 ;
400 North Olive Street Austin, Texas 78701 L . B . 81 -
Dallas, Texas 75201 . B. R. Clements <
Vice President Wuclear Mr. H. R. Rock Texas Utilities Generating Cc.d any Gibbs and Hill, Inc. Skyway Tower 393 Seventh Avenue 400 North Olive Street New York, New York 10001 L. B.'81 Dallas, Texas 75201 Mr. A. T. Parker Westinghouse Electric Corporation William A. Burchette, Esq.
P. O. Box 355 1200 New Hampshire Avenue, N. W.
Pittsburgh, Pennsylvania 15230 Suite 420 Washington, D. C, 20036 Renea Hicks, Esq.
Assistant Attorney General Ms. Billie Pirner Garde Environmental Protection Dnision Citizens Clinic Director P. O. Box 12548, Capitol Station Government Accountabi.lity Project Austin, Texas 78711 1901 Que Street, N. W.
Washington, D. C. 20009 Mrs. Juanita Ellis, President Citizens Association for Sound David R. Pigott, Esq.
Energy Orrick, Herrington & Sutcliffe 1426 South Polk 600 Montgomery Street Dallas, Texas 75224 San Francisco, California 94111 Ms. Nancy H. Williams AnthonyZ.RoismaN,Esq.
CYGNA Trial Lawyers for Public Justice 101 California Street 2000 P. Street, N. W.
San Francisco, California 94111 Suite 611 Washington, D. C. 20036 4
ENCLOSURE 1 RE0 VEST FOR ADDITIONAL INFORMATION I. Electrical / Instrumentation Area
- a. Electrical Cable Terminations t The Technical Review Team (TRT) inspected r6ndam samples of safety-related terminations, butt splices inside panels, and l vendor-installed terminal lugs in General Electric (GE) motor (
control centers, and reviewed documentation relative to the installations.
- 1. The TRT found a lack of awareness on the part of quality control ~,
(QC) electrical inspectors to document in the inspection reports when the installation of the " nuclear heat-shrinkable cable insulation sleeves" was required to be witnessed.
Accordingly, TUEC shall clarify procedural requirements and provide additional inspector training with respect to the areas in which nuclear heat-shrinkable sleeves are required on splices and assure that such sleeves are installed where required.
- 2. The TRT found inspection reports that did not indicate that the required witnessing of splice installation was done. Examples are as follows:
IR ET-1-0005393 IR ET-1-0005396 IR ET-1-0005394 IR ET-1-0006776
, IR ET-1-0C05395 IR ET-1-0014790 Accordingly, TUEC will assure that all QC inspections recuiring witnessing for butt splices have been performed and properly l documented; and verify that all butt splices are properly I identified on the appropriate drawings and are physically i identified within the, appropriate panels.
- 3. The TRT foJnd a lack of splice qualification requirements ano ;
provisions in the installation procedures to verify the . I operability of those circuits for which splices were being used.
l Accordingly. TUEC shall develop adequite installation / inspection procedures to assure that the wiring splicing materials are qualified for the appropriate service conditions, and that splices are not located adjacent to each other,
- 4. Selected cable terminations were found that did not agree with their locations on drawings. Examples are as follows:
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__ 1 Panel'CP1-ECPRCB-14, Cable E0139880 Panel CPI-ECPRTC-16, Cable E0110040 Panel CP1-ECPRTC-16, Cable E0118262 Panel CP1-ECPRTC-27, Cable EG104796 Panel CPX-ECPRCV-01, Cable EG021856 '
Panel CPI-ECPRCB-02, Cable NK139853 (nonsafety)
Accord-ingly, TUEC shall reinspect all safety-related and associated terminations in the control room panels and in the termination cabinets in the cable spreading room to verify that their locations are accurately depicted on drawings. Should the results of this reinspection reveal an unacceptable level of nonconformance to drawings, the scope of this reinspection effort shall be expanded to include all safety-related and asseciated terminations at CPSES. '
- 5. The TRT found cases where nonconformance reports (NCRs) concerning vender-installed terminal lugs in GE motor control centers had been improperly closed. Examples are NCR Nos.
E-84-01066 through NCR E-84-01076, inclusive.
Accordingly TUEC shall reevaluate and predisposition all NCRs related to vendor-installed terminal lugs in GE motor control centers.
- b. Electrical Eouiement Seoaration The.TRT reviewed the separation criteria between separate cables, trays and conduits in the main control room and cable spreading room in Unit l', and the compatibility of the f.lectrical erection specifications with regulatory requirements. The TRT reviewed documentation and inspected random samples of separation between I safety-related cables, trays and conduits and between them and
( nonsafety-related cables, trays and conduits.
- 1. In numerous cases, safety-related cables within flexible conduits inside main control room panels did not meet minimum separation requirements. Examples are as follows:
Panel CPI-EC-PRCB-02 Panel CP1-EC-PRCB-07 Panel CPI-EC-PRCP-06 Panel CP1-EC-PRCB-08 Panel CP1-EC-PRCB-09 .
Accordingly, TUEC shall reinspect all panels at CPSES, in addition to those in the main control room for Unit 1, that l contain redundant safety-related cables within conduits, or ;
safety and non-safety related cables within conduits, and either !
correct each violation of the separation criteria, or
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demonstrate by analysis the acceptability of_the conduit as.a j barrier for each case where the minimum separation is not met.- j
- 2. In several' cases, separate safety and nonsafety-related cables and. safety and nonsafety-related cables within flexible conduits inside main control room panels did not meet minimum )(
-separation requirements (Table 1 identifies examples of these i cases). No evidence ~ was found that justified the lack of separation'.
Accordingly, TUEC shall reinspect all panels at'CPSES, in.
addition to those in the main control room of Unit 1, and either ,
correct each violation of the separation criteria concerning separate cables and cables within flexible conduits, or -
demonstrate by analysis the adequacy of the flexible conduit as a barrier.
- 3. The TRT found that the existing TUEC analysis substantiating the adequacy of the criteria for separation between conduits and cable trays had not been reviewed by the NRC staff.
Accordingly, TUEC shall submit the analysis that substantiates the acceptability of the' criteria stated in.the electrical erection. specifications governing the separation between -
independent conduits and cable trays.
4 The TRT found two minor violations of the separation criteria s
inside panels CP1-EC-PRCB-09 and CPI-EC-PRCB-03 concerning a-barrier that had been removed and redundant field wiring not meeting minimum seperatien. The devices involved with the barrier were FI-2456A, PI-2453A, PI-2475A, and 172450, associated with Train A; and FI-2457A, PI-2454A, PI-2476A, and IT-2451, !
associated with Train B. The field wiring was associated with devices HS-5423 of Train B and HS-5574, nonsafety-related.
Accordingly, TUEC shall correct two minor violations .of the separation criteria inside panels CP1-EC-PRCB-09 and CP1-EC-PRCP-03 concerning a barrier that had been removed and-redundant field wiring not meeting minimum separation.
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Table 1 Examples of Cases of Safety or Nonsafety-Related Cables l 1
In Contact With Other Safety-Related Cables Within Conduits in Control Room i
~
Panels
- 1. ' Control Panel CP1-EC-PRCB Containment Spray System Cable No. Train Related Instrument EG139373 8(green) Undetermined .!
E0139010 A (orange) Undetermined .
{
- 2. Control Panel CPI-EC-PRCB Reactor Control System l Cable No. Train Related Instrument EG139383 B (green) Reactor manual trip switch E0139311 A(orange) Undetermined
- 3. Control Panel CP1-EC-PRCP Chemical & Volume Control System !
Cable No. Train Related Instrument EG139335 B (green) LCV-112C E0139301 A (orange) Undetermined 4 Control Panel CP1-EC-PRCB Auxiliary Feedwater Control System Cable No. Train Related Instrument E0139753 A (orange) FK-2453A E0139754 A (orange) FK-2453B ;
E0130756 B (green) FK-2454A !
EG139288 B (green) FK-2454B i i
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- c. Electrical Conduit Sucoorts l
-The TRT examined.the nonsafety-related conduit support installation in selected seismic Category I areas of the plant. The support installation for non-safety related conduits less than or. equal to 2 inches was inconsistent with seismic requirements and no evidence could be found that substantiated the adequacy of the installation for nonsafety-related conduit of any size. According to-Regulatory Guide 1.29 and FSAR Section 3.78.2.8, the seismic Category II and nonseismic items should be designed in such a way that their failure would not adversely affect the function of safety-related components or cause injury to plant personnel.
Accordingly, TUEC shall propose a program that assures the adequacy ,
of the seismic support system installation for nonsafety-related conduit in all seismic Category I areas of the plant as follows:
i
- 1. Provide the resul.ts of seismic analysis which demonstrate that all nonsafety-related conduits and their support systems, satisfy the provisions of Regulatory Guide 1.29 and FSAR Section 3.78.2.8.
- 2. Verify that nonsafety-related conduits less than or equal to 2 inches in diameter, not installed in accordance with the ,
requirements of Regulatory Guide 1.29, satisfy applicable design {
requirements. '
- d. Electrical OC Insoector Trainino/0 qualifications, The TRT examined electrical QC inspector training and certification l files, and requirements for personnel testing, on-the-job training, and decertification. The TRT also interviewed selected electrical QA/QC personnel.
)
- 1. The TRT found a lack of supportive documentation regarding personnel qualifications in the training and certification files, as required by procedures and regulatory requirements.
Also, the TRT found a lack of documentation for assuring that i the requirements for electrical QC inspector decertification l were being met. Specific examples are: -
I One case of no documentation of a high school diploma or General Equivalency Diploma.
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One case of no documentation to waive the remaining 2 l months of the requirsd 1 year experience.
One case where a QC technician.had not passed the required color vision examination administered by a professional eye specialist. A makeup test using colored pencils was administered by a QC supervisor, was passed,
. and then a waiver was given.
L
- TJo cases where the experience requirements to become a Level 1 technician were only marginally met.
One case of no documentation in the training and certification files substantiating that the person met the experience requirements.
s Accordingly, TUEC shall review all the electrical QC inspector training, qualification, certification and decertification files against the project requirements and provide the information in such a form that each requirement is clearly shown to have been
/'
met by each inspector. If an inspector is found to not meet the training, Qualification, certification, or decertification requirements, TUEC shall then review the records to determine the-adequacy of inspections made by the unqualified individuals .
and provide a statem'ent on the impact of the deficiencies noted on the safety of the project.
- 2. The TRT found a lack of guidelines and procedural requirements for the testing and certifying of electrical QC inspectors. Specifically, it was found that:
No time limit or additional training requirements existed between a failed test and retest.
No controls existed to assure that the same test would not be given if an individual previously failed that test. !
l No consistency existed in test scoring.
No guidelines or procedures were available to control the i disqualification of questions from the test. j j
No program was available for estotli:hing new tests (except when procedures changed). The same tests h:d been utilized for the last 2 years.
Accordingly, TUEC shall develop a testing program for electrical QC inspectors which provides adequate administrative guidelines, procedural requirements and test flexibility to assure that suitable proficiency is achieved and maintained.
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- i The deficiencies identified with the electrical QC inspections have !
generic implications to other construction disciplines. The implications of these findings will be further assessed as part of
' the overall progrartcatic review oo QC inspector training and qualification and the results of ':his review will be reported under the QA/QC category on " Training a'id Qualification."
,I I . Civil / Structural Area
- a. Unable to Justify Reinforcing Qui Omitted in the Reactor Cavity The TRT investigated a documente! occurrence in which reinforcing steel was omitted from a Unit I retctor cavity concrete placement l between the 812-foot and 819-foot t-inch elevations. This J reinforcement was installed and inspected according to drawing 2323-51-0572, Revision 2. However, after the concrete was placed, Revision 3 to the drawing was issued showing a substantial increase 1
in reinforcing steel over that which was installed. Gibbs &' Hill Engineering was informed of the omission by Brown & Root i Nonconformance Report CP-77-6.. Gibbs & Hill Engineering replied that the omission in no way impaired the structural integrity of the structure. Nevertheless, the additional reinforcing steel was added as a precaution against cracking which might occur in the
' vicinity of the neutron detector slots should a loss of coolant )
accident (LOCA) occur. A portion of the emitted reinforcing steel !
was also placed in the next concrete lift above the 819-foot i-inch 1 level. This was done to partially compensate for the reinforcing steel omitted in the previous concrete lift and to minimize the overall area potentially subject to cracking.
The TRT requested documentation indicating that an analysis was performed supporting the Gibbs & Hill conclusion. The TRT was subsequently informed that an analysis had not been performed.
Therefore, the TRT cannot determine the safety significance of this issue until an analysis is performed verifying the adequacy of the reinforcing steel as installed.
Accordingly, TUEC shall provide an analysis of the as-built condition of the Unit I reactor cavity that verifies the adequacy of the reinforcing steel between the 812-foot and 819-foot i-inch -
elevations. The analysis shall consider all required load combinations.
b ., Fa,11 1fication of Concrete Compression Strenoth Test Results The TRT investigated allegations that concrete strength tests were falsified. The TRT reviewed an NRC Re Report No. 50-445/79-09; 50-446/79-09)gion cf thisIV investigation matter that included (IE i
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t interviews with fifteen individuals. Of these, only the alleger and one other individual stated they thought that falsification occurred, but they did not know when or by whom. The ,
TRT also reviewed slump and air entrainment test results of concrete placed Februaryduring)the 1977 andperiod did notthe alleger find was employed any apparent variation(J'nuary in the 1976 to uniformity of the parameters for concrete placed during this period.
Although the uniformity of the concrete placed appears to minimize the likelihood that low concrete strengths were obtained, other allegations were raised concerning the falsification of records associated with slump and air content tests. The Region IV staff addressed these allegations by assuming that cencrete strength test results were adequate. Furthermore, a number of other allegations dealing with concrete olacement problems (such as deficient aggregate .
grading and concrete in the mixer too long) were also resolved by s assuming that concrete strength test results were adequate. The TRT agrees with Region IV that, while the preponderance of evidence suggests that falsification of results did not take place, the matter cannot be resolved completely on the basis of concrete strength test results, especially if there is any doubt about whether they may have been falsified. Due to the importance of the concrete strength test results, the TRT believes that additional action by TUEC is necessary to provide confirmatory evidence that the reported concrete strength test results are indeed representative of the strength of the concrete installed in the Category I concrete structures.
Accordingly, TUEC shall determine areas where safety-related concrete was placed between January 1976 and February 1977, and provide a program to aesure acceptable concrete strength. The program shall include tests such as the use of random Schmidt hamer tests on the ,
concrete in areas where safety is critical. The program shall include a comparison of the results with the results of tests per-formed on concreto of the same design strength in areas where the strength of the concrete is not questioned, to deterinine if any significant variance in strength occurs. TUEC shall submit the program for performing these tests to the NRC for review and approval prior to performing the tests.
- c. Maintenance of Air Gao Between Concrete Structures
,, The TRT investigated the requirements to mainthin an air gap between concrete structures. Based on the review of available inspection reports and related documents, on field observations, and on discussions with TUEC engineers, the TRT cannot determine whether an adequate air gap has been provided between concrete structures. Field investigations by B&R QC inspectors indicated unsatisfactory conditions due to the presence of debri,s in the air 1
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gap, such as wood wedges, rocks, clumps of concrete and rotofoam.
The disposition of.the NCR relating to this matter states that the
" field investigation reveals that most of the material has'been removed." However, the TRT cannot determine from this report (NCR.
C-83-01067) the extent and location of the debris remaining between the structures.
Based on discussions with TUEC engineers, it is the TRT's understanding that field investigations.were made but that no permanent records were maintained. In addition, it is not apparent that the permanent installation of elastic joint filler material
("rotofoam") between the Safeguards Building and the Reactor Building, and below grade for the other concrete structures, is consistent with the seismic analysis assumptions and dynamic models _
used to analyze the buildings, as these analyses are delineated in the Final Safety Analysis Report (FSAR). The TRT, therefore, concludes !
that TUEC has not adequately demonstrated compliance with FSAR Sections 3.4.1.1.1, 3.8.4.5.1, and 3.7.B.2.8, which require separation of Seismic Category I buildings to prevent seismic interaction during an earthquake.
Accordingly, TUEC shall: -
- 1. Perform an inspection of the as-built condition to confirm that -
adequate separation for all seismic category I structures has been provided.
- 2. Provide the results of analyses which demonstrate that the 'l
, presence of rotofoam and other debris between all concrete structures (as determined by inspections of the as-built
.. conditions) does not result in any significant increase in 1 seismic response or alter the dynamic response characteristics (
of the Category I structures. components and piping when compared with the results of the original analyses.
- d. Seismic Desion of Control Room Ceiling Elements The TRT investigated the seismic design of the ceiling elements installed in the control room. The following matrix designates those ceiling elements present in the control room and their seismic category designation:
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- 1. Heating, Ventilating and Air Conditioning - Seismic Category I
- 2. Safety-Related Conduits - Seismic Category I
- 3. Nonsafety-Related Conduits - Seismic Category II
- 4. Lighting Fixtures - Seismic Category II
- 5. Sloping Suspended Drywall Ceiling - Non-Seismic i
- 6. Acoustical Suspended Ceiling - Non-Seismic )
- 7. Lowered Suspended Ceiling - Non-Seismic j l
According to Regulatory Guide 1.29 and FSAR Section 3.78.2.8, the l seismic Category II and nonseismic items should be designed in such a way that their failure would not adversely affect the functions of i safety-related components or cause injury to operators. 1 For the nonseismic items (other than the sloping suspended drywall '
ceiling), and for nonsafety-related conduits whose j diameter is 2 inches or less, the TRT could find no evidence that the possible effects of a failure of these items had been considered. In addition, the TRT determined that calculations for seismic Ca.tegory II components (e.g., lighting fixtures) and the calculations fnr the sloping suspended drywall ceiling did not adequately reflect the rotational interaction with the nonseismic l items, nor were the fundamental frequencies of the supported j masses determined to assbss the influence of the seismic 1 response spectrum at the control room ceiling elevation would have on I the seismic response of the ceiling elements.
Accordingly, TUEC shall provide:
- 1. The results of seismic antlysis which demonstrate that the nonseismic items in .the control room (other than the sloping suspended drywall ceiling) satisfy the provisions of Regulatory Guide 1.29 and FSAR Section 3.73.2.8.
- 2. An evaluation of seismic design adequacy of support systems for the lighting fixtures (seismic Category II) and the suspended drywall ceiling (nonseismic item with modification) which accounts for pertinent floor response characteristics of the systems. {
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- 3. Verification that those items in the control room ceiling l not installed in accordance with the requirements of Regulatory Guide 1.29 satisfy applicable design requirements.
4 The results of an analysis that justify the adequacy of the nonsafety-related conduit support system in the control room for conduit whose diameter is 2 inches or less.
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- 5. The results of an analysis which demonstrate that the foregoing problems are not applicable to other Category II and nonseismic structures, systems and components elsewhere in the plant.
- e. Unauthorized Cutting of Rebar in the Fuel Handling Building The TRT investigated an alleged instance of unauthorized cutting of rebar associated with the installation of the trolley process aisle rails in the Fuel Handling Building. The claim is that during installation of 22 metal plates in January 1983, a core drill was used to drill about 10 holes approximately 9 inches deep. The TRT reviewed the reinforcement drawings for the Fuel Handling Building and determined that there were tnree layers of reinforcing steel in ~
the top reinforcement layer of the slab. This reinforcement layer consisted of a No.18 bar running in the east-west direction in the first and third layers, and a No. 11 bar, running in the north-south .
direction on the second layer. The review also revealed that the layout of the reinforcement and the trolley rails was such that the east-west reinforcement would' interfere with the drilling of holes-l along only one rail location. However, if 9-inch holes were drilled, both the first and third layers of No.18 reinforcement would be cut.
Design Change Authorization No. 7041 was written for authorization to cut the uppermest No. 18 bar at only one rail location, but did not l reference authorization to cut the lower No.18 bar. DCA-7041 also stated that the expansion bolts and base plates may be moved in the east-west direction to avoid interference with reinforcement running in the north-south direction. The infomation, described in DCA-7041, was substantiated by Gibbs & Hill calculations. If the ten
, holes were a::tually drilled 5 inches deep, then the allegation that the reinforcement was cut without proper authorization would be valid.
Accordingly, TUEC shall provide:
- 1. Information to demonstrate that only the No. 18 reinforcing steel in the first layer was cut, or
- 2. Design calculations to demonstrate that structural integrity is maintained if the No.18 reinforcing steel on both the first and third layers was cut.
III. Test Programs Area
- a. Hot Functional Testing (HFT)
The TRT reviewed a sample of the completed data packages for HFT preoperational test procedures, pertinent startup administrative procedures, NRC inspection reports, and the preoperational test index and its schedule. The TRT also inspected test deficiency reports
f 12 (TDRs) that were generated as a result of test deficiencies found prior to and during HFT. .
- 1. Chapter 14 of the FSAR and Regulatory Guide 1.68 provide requirements for the conduct of preoperational testing.
In reviewing test data packages, the TRT found that certain test objectives were'not met. It appears that the Joint Test Group approved incomplete data packages for at least three preoperational hot functinal tests. These were:
Test Procedure Deficiency ICP-PT-02-12, " Bus Because acceptable voltages Voltage and Load Survey" could not be achieved with the -
specified transformer taps, they were s changed. A subsequent engineering evaluation required returning to the original taps, but no retest was performed.
ICP-PT-34-05, " Steam Level detectors 1-LT-517, 518 Generator Narrow Range and 529 were rep, laced with Level Verification" temporary equipment of a design that was different from that j which was to be eventually installed l
)
ICP-PT-55-05 Level detector 1-LT-461 apaeared
" Pressurizer Level to be out of calibration during the Control" test and was replaced after the test. 4 The retest approved by the JTG was a {
cold calibration rather than a test )
consistent with the original test )
objective, which was to obtain satisfactory data under hot conditions.
Accordingly, TUEC shall review all complete preoperational test data packages to ensure there are no other instances where test objectives were not met, or prerequisite conditions were not satisfied. The three items identified by the TRT shall be included, along with appropriate justification, in the test i deferral packages presented to the NRC. !
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- 2. The TRT noted during a review of HFT completed test data that the JTG did not approve the data until after cooldown from the i test. The tests are not considered complete until this approval is obtained. In order to complete tne proposed post-fueling, c'eferred preoperational HFT, the JTG, or a similarly qualified group, must approve the data prior to proceeding to initial criticality. The TRT did not find any document providing assurance that TUEC is comitted to do this.
Accordingly, TUEC shall comit to having a JTG, c ifmilarly qualified group, review and approve all post-fueling preoperational test results prior to declaring the system operable in accordance with the technical specifications.
- 3. The TRT pointed out that in order to conduct preoperational tests at the necessary temperatures and pressures after fuel load, certain limiting conditions of the proposed technical specifications cannot be met, e.g., all snubbers will not be operable since some will not have been tested.
Accordingly, TUEC shall evaluate the required plant conditions for the deferred preoperational tests against limiting conditions in the proposed technical specifications and obtain NRC approval where deviations from the technical specifications "
are necessary.
4 Data for the thermal expansion tests (which have not yet been approved by the JTG) did not provide for traceability between the calibration of the measuring instruments and the monitored locetions, es recuired by Startup Adminis*.rstive Prcetdure-7.
The information was separately available in a personal log held by Engineering.
Accordingly, TUEC shall incorporate the infomation necessary to provide traceability between themal exp-ansion test monitoring locations and measuring instruments. TUEC shall also establish administrative controls to assure appropriate test and measuring equipment traceability during future testing,
- b. Containment Intercrated Leak Rate Testino (CILRT) -
The TRT reviewed the data package for the CILRT performed on Unit 1, and discussed the conduct of the test with TUEC and NRC {
personnel who participated in or witnessed it.
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Apparently after repairing leaks found during the first two attempts, the third attempt at a CILRT was successful. It was successfully completed after three electrical penetrations were isolated because the leakage through them could not be stopped.
Though the leaks were subsequently repaired and individually tested.with satisfactory results, NRC approval.was not obtained to perform the CILRT with these penetrations isolated. In addition, leak rate calculations were performed using ANSI /ANS 56.8, which-is neither endorsed by the NRC nor in accordance with FSAR comitments. ,
Accordingly, TUEC shall identify to NRC any other differences in the conduct of the CILRT as a result of using ANSI /ANS 56.8 rather than ANSI N45.4-1972. Additionally, TUEC shall identify to NRC all other deviations from FSAR commitments. .s
- c. prerequisite Testino The TRT reviewed FSAR commitments, startup administrative procedures, prerequisite' test records, craft personnel qualification records, and discussed them with startup'and craft management personnel. The TRT also observed test support craft personnel at work and interviewed some of them to gain familiarity with their attitudes and capabilities. .
The review of test records revealed that craft personnel were signing to verify initial conditions for tests in violation of startup Administrative Procedure-21, entitled: " Conduct of Testing" (CP-SAP-21). This procedure requires this function to be performed by System Test Engineers (STE). Startup management had issued a memorandum improperly authorizing craft personnel to perform these verifications on selected tests.
Accordingly, TUEC shall rescind the startup memorandum (STM-83084),
which was issued in conflict with CP-SAP-21, and ensure that no other I memoranda were issued which are in conflict with approved procedures.
- d. preoperational Testino The TRT assessed the preoperational test program by reviewing administrative procedures, interviewing startup personnel, and examining test records, schedules, system assignments subsystem
- definition packages, and the master data base.
problems found with test data are addressed in section III.a of this enclosure. The TRT also found that STEs were not being provided with current design ir. formation on a routine, centrolled basis, and had to update their own material when they considered it appropriate.
Accordingly, TUEC shail establi:h measures to provide greater assurance that STE's and other responsible personnel are provided with current controlled design documents and change notices.
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- 44 0 hff ' " fh -
UNITED STATES
[
ll; p kE NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 2o555 V*M:e ****
NOV 2
- 1984 Docket Nos.:
and 50-446 50-445 g3)pLCL 7 1/
Mr. M. D. Spence J
President Texas Utilities Generating Company )
i 400 North Olive Street Lock Box 81 ,
Dallas, Texas 75201 l
Dear Mr Spence:
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Subject:
Comanche Peak Review
)
- 1 On July 9,1984, the staff began an intensive onsite effort to complete a per- 1 tion of the reviews necessary for the staff to reach its decision regarding the l licensing of Comanche Peak, Unit 1. The onsite effort covered a number of areas, d including allegations of improper construction practices at the facility.
On September 18,1984,' the NRC met with you and other Texas Utilities Electric' i Company representatives to provide you with a number of technical issues in the q electrical / instrumentation, civil / structural, and test program areas having j potential safety implications. The issues discussed constitute a portion of j the technical issues and allegations.being evaluated by the Technical Review e Team (TRT).
The activities of the TRT have progrested to the point where it is appropriate !
to provide you with a status of additional items under review and to request additional information. These items, in the coatings, mechanical, and miscel-lanecus areac, are l'sted ir the enclosure to this letter. Further backg ound infor'mation regarding these issues will be published in a Supplement to a Safety j Evaluation Report (SSER), which will document the TRT's overall assessment of i the significance of the issues examined. j l
The items in the enclosure t'o this letter cover only a portion of the TRT's j effort. The TRT's ongoing evaluation, QA/QC review and conversations-with {
allegers may reveal additional items in the coatings, mechanical, and mis- '
cellaneous areas for which additional requests for information may be appro-priate. Also, the TRT evaluation of QA/QC issues, and its consideration of ,
the programmatic implications of these findings, are still in progress. A )
summary of these issues will be provided to you at a later date.
You are requested to submit additional information to the NRC, in writing, in- l cluding a program and schedule for completing a detailed and thorough assess-ment of the issues identified in the enclosure to this letter. This program plan and its implementation will be evaluated by the staff before NRC considers the issuance of an operating license for Comanche Peak, Unit 1. The program plan should address the root cause of each problem identified and its generic kcNN ,,
- _ - _ _ -___ - _ A
2-implications on safety-related systems, programs, or areas. You should also address the collective significance of these deficiencies. Your program plan should also include the proposed TUEC action to assure that such problems will not occur in the future.
This request is submitted to you in keeping with the NRC practice of promptly notifying applicants of outstanding information needs that could potentially affect the safe operation of their plant. Future requests for additional information of this nature will be made, if necessary, as the activities of the TRT progress.
Sincerely,
$1 .
Darreli G CEise hu* trector
/ Division of Licensfn<g Office of Nuclear Reactor Regulation
Enclosure:
As stated cc w/ enclosure:
See next page I
I
, COMANCHE PEAK-Mr..M. D. Spence President Texas Utilities Generating Company 400 N. Olive St., L.B. 81 Dallas, Texas 75201 )
)
cc: Nicholas S. Reynolds, Esq. Mr. James E. Cummins Bishop, Liberman, Cook, Resident Inspector / Comanche Peak Purcell & Reynolds Nuclear Power Station 1200 Seventeenth Street, N. W. c/o U. S. Nuclear Regulatnry Washington, D. C. 20036 Comission P. O. Box 38 Robert A. Wooldridge, Esq. Glen Rose, Texas 76043 Worsham, Forsythe, Sampels &
Wooldridge i Mr. Robert D. Martin 2001 Bryan Tower, Suite 2500 U. S. NRC, Region IV -
Dallas, Texas 75201 611 Ryan Plaza Drive Suite 1000 Mr. Homer C. Schmidt Arlington, Texas ' 76011 Manager - Nuclear Services Texas Utilities Generating Company Mr. Lanny Alan Sinkin Skyway Tower 114 W. 7th, Suite 220 400 North Olive Street Austin, Texas 78701 l L. B. 81 Dallas, Texas 75201 B. R. Clements '
Vice President Nuclear Mr. H. R. Rock Texas Utilities Generating Company Gibbs and Hill, Inc. Skyway Tower 393 Seventh Avenue 400 North Olive Street New York, New York 10001 L. B. 81 Dallas, Texas 75201 iMr. A. T. Parker Westinghouse Electric Corporation William A. Burchette, Esq.
P. O. Box 355 1200 New Hampshire Avenue, N. W.
Pittsburgh, Pennsylvania 15230 Suite 420 Washington, D. C. 20036 Renea Hicks. Esq.
Assistant Attorney General Ms. Billie Pirner Garde Environraental Protection Division Citizens Clinic Director P. O. Box 12548, Capitol Station Government Accountability Project Austin, Texas 78711 1901 Que Street, N. W. -
Washington, D. C. 20009 Mrs. Juanita Ellis, President Citizens Association for Sound David R. Pigott, Esq.
Energy Orrick, Herrington & Sutcliffe 1426 South Polk 600 Montgomery Street Dallas, Texas 75224 San Francisco, California 94111 Ms. Nancy H. Williams Anthony Z. Roisman, Esq.
I CYGNA Trial Lawyers for Public Justice 101 California Street 2000 P. Street, N. W.
San Francisco, California 94111 Suite 611 Washington, D. C. 20036 i
+: .
COMANCHE PEAK cc: Mr. Dennis Kelley Resident ~ Inspector'- Comanche Peak clo U. S. NRC l P. O. Box'1029 I Granbury, Texas 76048
.Mr. John W. Beck Manager - Licensing Texas Utilities Electric Company Skyway Tower
-l 400 N. Olive Street i L. B. 81 Dallas, Texas 75201' Mr. Jack Redding -
Licensing s Texas Utilities Generating Company 4901 Fairmont Avenue Bethesda, Maryland 20014
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REQUEST FOR ADDITIONAL INFORMATION IV. protective _Coatinos Area
- a. Surveillance and Test procram for Coatinos The protective coatings Technical Review Team (TRT) reviewed the backfit program, design basis accident qualifications, traceability, application and repair procedures, training, coating exempt log and dispositioning of non-conformance reports. Concurrently, the staff is evaluating the effects I
on containment emergency sump performance of paint and insulation debris.
The results of the two concurrent reviews will be combined in one supple-mental safety evaluation which is scheduled to be issued by January 1985.
Actions required for resolution of protective coatings issues will be delineated in the supplement. ,
V. Mechanical Area
- a. Inspection for Certain Types of Skewed Welds in NF Scoperts The TRT investigated inspection procedures of Brown & Root (B&R) for welds in pipe supports designed to ASME III Code, Subsection NF. The TRT found that no fillet weld inspection criteria existed for certain types of skewed welds. By definition, skewed welds are thote welds joining (1) two non-perpendicular or non-colinear structural members, or (2) two members with curved surfaces or curved cross sections, such as a pipe stanchion (a sec-tion of pipe used as a structural member) welded to another pipe stanchion or to a curved pipe pad. Notice that for type (2), the effect of curva-ture at the weld connection induces skewed considerations, even though the two joining members are physically perpendicular. The B&R weld inspection procedares CP-QA?-12.1 and QI-QA?-11.1-22 for NF supports have addrassed
' type (1) skewed welds; however, the TRT found that QI-QAp-11.1-28 did not include weld inspection criteria for type (2) skewed welds. Although the TRT was told by B&R personnel that orocedure QI-QAP-11.1-26 for piping weld inspection was ~used, since such weld connections were similar in con-figuration to a pressure boundary stanchien attachment weld, no evidence documenting the use of this inspection procedure was provided to.the TRT.
According to records reviewed by the TRT, these welds were actually cate-gorized as "all other welds" rather than " skewed welds" on the required QC chec kli st. Instead of using fillet weld gauges for measuring the size of nonskewed welds, welders were supposed to use a straight edge and a steel',
scale for measurement of a type (2) skewed weld, as described in QI-QAP-11.1-28. In addition, due to the variable profile along its curved weld connection, the weld size should have been measured at several dif-ferent locations. The lack of inspection criteria and lack of verification of proper inspection procedures being conducted for type (2) skewed welds are a violation of ASME Code for NF supports committed to by TUEC in FSAR Section 5.2.1 and a violation of Criterion XVII in Appendix B of 10 CFR 50.
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The TRT reviewed weld inspection procedures, weld data cards, and visually inspected several type (2) skewed welds in randomly sampled NF supports where pipe stanchions were used. Although the small sample of welds l l
inspected by the TRT are acceptable, due to deficiencies in inspection i records and the apparent lack of inspection criteria, the TRT is not cer- I tain whether other type (2) skewed welds were inspected properly. This is a generic issue involving many NF supports in various safety-related sys- t tems. The lack of documented inspections and criteria for type (2) skewed j welds in NF supports represents a safety concern regarding the possible {
existence of under-sized welds in supports which are required to resist {
various design loads. '
Accordingly, TUEC shall (1) Revise B&R weld inspection procedures CP-QAP-21.1 and QI-QAP-11.1-28 to properly address type (2) skewed welds of stanchion to stanchien N and stanchion to pipe pad; and, (2) provide evidence to verify that previous inspections of these types of skewed welds were performed to the appropriate procedures.
- b. Improper Shortenine of Anchor Bolts in Steam Generator Upoer Lateral Supports The TRT was informed that some a'nchor bolts in the steam generator upper support beams were shortened during installation to less than the length shown on the design drawing without proper authorization. The TRT was told that the bolt cutting incident occurred either because the hole of ;
the, anchor device was filled with debris, or the threaded portion of the ,j bolt had concrete mix stuck to it. There are 18 bolts at each end of each i of 4 beams, totalling 144 bolts. There is one oeam for each steam geriera-tor. The bolt threads into an anchor device embedded in the concrete wall. 4 The acceptable bolt length or the length of bolt available for threading into the anchor device is vital to ensure structural capability of the j support beams.
i The TRT attempted to review TUEC records for ultrasonic (UT) measurement results and general installation practices. The TRT was told that ultra- i sonic testing of these types of bolts was not a procedural requirement; l however, TUEC was unable to provide any other installation records for TRT l
review. The TRT concludes that such unauthorized bolt cutting and lack of '
installation inspection records is a violation of QA procedures and Cri-terion XVII in Appendix B of 10 CFR 50. Since the support beams are essen-tial to provide lateral restraint for the steam generator during a LOCA or i seismic event, adequate anchoring capability of the bolts has safety sig- l nificance and, as a result, appropriate measures are needed to ensure conformance with General Design Criterion 1 of 10 CFR 50.
Accordingly, TUEC shall provide evidence, such as ultrasonic measurement results, to verify acceptable bolt length. Should unauthorized bolt l
cutting be verified, TUEC shall:
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(1) replace shortened bolts with bolts of proper length, or provide analysis to justify the adequacy of shortened bolts as installed; and, 4 (2) provide justification or propose measures to ensure that no similar concern exists for bolting,
- c. Design Consideration for Picing Systems Between Seismic Category I and Non-Seismic Category I Buildings In April 1984 the Comanche Peak Special Review Team (SRT), formed and coor-dinated between NRR, IE and Region II and IV, performed a limited review of Comanche Peak. The TRT, in reviewing the SRT findings in the area of piping design considerations, has discovered that piping systems, such as -
Main Steam, Auxiliary Steam and Feedwater, are routed from the Electrical Control Building (seismic category I) to the Turbine Building (non-seismic category I) without any isolation. To be acceptable, each seismic cate-gory I piping system should be isolated from any non-seismic category I piping system by separation, barrier or constraint.
If isolation is not feasible, then the effect on the seismic category I piping of the failure in the non-seismic category I piping must be considered (CPSES FSAR 3.78.3-13.1).
For CPSES, FSAR section 3.78.2.8 establishes that the Turbine Building is a non seismic category I structure and failure is postulated during the seismic (SSE) event. The effect of Turbine Building failure on any non-isolated piping routed through the Turoine Building from any seismic category I building must be considered.
iIn addition, for non-seismic category I piping connected to Seismic Category I piping, the dynamic effects of the non-seismic category I piping must be considered in the seismic design of the seismic category I piping and supports, unless TUEC can shoiv that the dynamic e.ffects of the non-seismic category I piping are isolated by anchors or restraints. The anchors or restraints used for isolation purposes must be designed to withstand the combined loading imposed by both the seismic category I and
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non-seismic category I piping.
Accordingly, TUEC shall ~ ovide analysis and documentation that the piping systems routed from seism 1c category I to non-seismic category I buildings meet the stated FSAR criteria. 4 i
- d. Plug Welds 1
The TRT investigated allegec generic problems regarding uncontrolled !
repairs to holes existing in pipe supports, cable tray supports and base )
plates in Units 1 and 2. These holes, which had been misdrilled during j fabrication, were repaired by plug welds. Since these supports are Seismic l l
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Category I supports and th: effects of the welds have not been evaluated, this constitutes a violation of Criteria IX and XVI of Appendix B to 10 CFR 50. Region IV inspections have confirmed the existence of such welds in cable tray supports located in the Unit 2 Cable Spreading Room.
Although the effects of unauthorized, undocumented and uninspected plug welds in some locations (e.g., the webs of I-beams or in structural members in compression) will be inconsequential, their effects in critical loca-tions (e.g., flanges of I-beams in flexure or in structural members in tension) in critically loaded supports or base plates could affect their structural integrity and intended function. ,
Accordingly, TUEC shall perform one of the following:
(1) Modify its proposed plan to Region IV (TXX-4183 and TXX-4259) to include a sampling inspection of all areas of the plant having plug '
welds, to include cable tray supports, pipe supports and base plates. Propose alternate methods of inspection.where the oblique lighting method is not viable (e.g., locations covered by heavy coats of paint). Perform an assessment of the effects on quality due to uncontrolled plug welds found during the proposed inspection, as modified above. Submit a report documenting the results of the in-spection and assessment to the NRC for review. .
(2) perform bounding analyses to~ assess the generic effects of uncon-trolled plug welds on the ability of pipe supports, cable tray sup-ports and base plates to serve their intended function. Submit a report documenting the esults of the assessment to the NRC for review.
- e. Installation of Main Steam pioes The TRT investigated an allegation that a Unit 1 main steam line had been installed incorrectly and had been forced into proper alignment after flush-ing operations by use of the main polar crane and come-alongs. It was also clairned that pipe supports had been modified to maintain the line in its forced position and vibrations following detachment of the flushing line could have damaged the main stea'n line. Based on its investigation, the TRT determined that the alleged incident pertained to restoration of the Unit 1, loop 1 main steam line to its initial, correct installation posi-tion. (The line had shifted during flushing operations due to the weight of the added water and because the temporary supports sagged.) The TRT j also determined that the modifications to permanent pipe supports were ;
necessary to provide proper support to the main steam line in its restored i position (initial designs for and construction of the supports had been {
l based on the shifted position of the line) and, although the alleged vi-
) brations could not be confirmed, their associated stresses might not have !
damaged the main steam line. (The highest stresses would have occurred in <
the weaker, temporary flushing line.) The TRT review of a TUEC analysis, performed 1 year after the incident, concluded that the analysis was incom-plete. An evaluation for the full sequence of events leading up to the l
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incident had not been performed. The TRT review of Gibbs & Hill Specifica-tion No. 2323-MS-100 indicated that there were. inadequate requirements and construction practices for the support of the main steam line during
{
flushing, and for temporary supports for piping and equipment in. general. j In particular, evaluations to assure the ade'quacy of temporary supports during flushing and installation were not required. The deficiencies'in ,
the analyses, specifications and construction practice identified above. (
- constitute a violation of Criterion V of Appendix B to 10 CFR 50. !
Accordingly,.TUEC shall:
(1) Modify Gibbs & Hill Specification No. 2323-MS-100, and institute pro-cedures for. support of the main. steam line during flushing and for temporary supports for piping and equipment in general to assure that the quality of piping and equipment are not affected. ~
(2) Perform an assessment of stresses in the portions of the Unit 1, loop 1, main steam and feedwater lines that were affected in the sequence of events involved during their initial installation, flushing and final. installation. Conditions requiring stress analysis are:
(a). Flushing condition when the lines were full of water and temporary supports had sagged or settled. .
(b) Disconnecting condition when vibrations of the temporary line i could have occurred.
(c) Lifting condition when forces were applied by the polar crane and come-alongs.
These assessments shall be based on appropriate piping configurations involved.
(3) Perform a non-destructive examination of locations in the Unit 1, loop 1, main steam and feedwater piping where stresses were exceeded {4 during the conditions of concern in a. through c. above. l (4) Review the existing baseline UT examinations for those portions of the Unit 1, loop 1, main steam and feedwater involved in all the conditions of concern in a. through c., above, for unacceptable '
indications.
(5) Review records of hydrostatic testing of the main steam and feedwater line to verify the quality of piping involved in the incident.
(6) Provide similar assessments for circumstances involved in a lifting incident identified during the TRT inspection for the Unit 1, loop 4, main steam line.
1
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(7) Provide assessments of effects on quality of safety-related piping and equipment whic! were involved in similar incidents of sagging, g settlements and failures, if any, of temporary supports.
(8) Submit the results of analyses, examinations and reviews in a docu-mented report for NRC ruiew.
VI. Miscellaneous Area
- a. Gap Between Reactor pressure Vessel Reflective Insulation (RpVRI) and the Biological Shield Wall '
The TRT investigated an allegation that the Unit I reactor pressure vessel outer wall was touching the concrete biological shield wall. .
A TRT review of existing documentation and discussions with TUEC personnel indicated that this allegation was not factual. However, N a significant-construction deficiency report, submitted pursuant to 10'CFR Part 50.55(e), on August 25, 1983, documented that unacceptable cooling occurred in the annuius between the RPVRI and the shield wall ;
during hot functional testing, apparently because of the existence of an inadequately sized annulus gap and possibly because the presence of construction debris in the annulus. TUEC corrected the situation by modifications to allow-increased air flow for proper heat dissi-pation and by removal of the construction debris. TUEC representa- 4 tives indicated that testin'g to verify the adequacy of the cooling flow will take place when additional. hot functional testing is con-ducted. Information gathered by the TRT during the investigation indicated that a design change in-the RPVRI support ring (i.e., loca-
. ting the ring outside rather than inside the insulation) resulted in a limited clearance between the RPVRI and the shield wall. The TRT review of tne 50.55(e) repurt rev641ed that. TdEC failed tc: (1) ad-dress the fundamental issue of the design change impact on annulus cooling flow, and (2) determine whether Unit 2 was similarly affected.
Accordingly, TUEC shall: ;
}
(1) Review their procedures for approval of design changes to non-nuclear safety related equipment, such as the RPVRI, and make revisions as necessary to assure that such design changes do not adversely affect safety-related systems.
J (2) Review procedures for reporting significant design and construc-tion deficiencies, pursuant to 10 CFR Part 50.55(e), and make changes as necessary to assure that complete evaluations are conducted.
(3) Provide an analysis which verifies that the cooling flow in the annulus between the RPVRI and the shield wall of Unit 2 is adequate for the as-built condition.
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t (4) Finally, verify during future Unit I het functional testing that completed modifications to the RPVRI support ring now allow adequate cooling air flow.
The TRT noted that control of debris in critical spaces between components and/or structures was identified as an issue, both in the investigation of this allegation and the civil /structaral area item II.c (Maintenance of Air Gap Between Concrete Structures), contained in Darrell G. Eisenhut's September 18, 1984, letter to TUEC. Accord-ingly, TUEC shall also:
(1) Identify areas in the plant having critical spacing between components and/or structures that are necessary for proper func-tioning of safety-related components, systems or structures in .
which unwanted debris may collect and be undetected or be dif-ficult to remove; (2) Prior to fuel load, inspect the areas and spaces identified and remove debris; and, (3) Subsequent to fuei load, institute a program to minimize the collection of debris in critical spaces and periodically inspect these spaces and remove any debris which may be present.
- b. Polar Crane Shimmino The TRT investigated the installation of the polar crane rail support system by visual inspection, review of associated documentation, and discussions with TUEC representatives and their contractors. Region IV Inspection Report 50-445/84-08; 50-446/84-04 and Notice of Violation,
, dated July 26, 1984, documented that gaps on the Unit 1 polar crine bracket and seismic connections exceeded deLign requirements. In Texas Utilities Generating Company responses of August 23, 1984, and September 7, 1984, the gaps were attributed to crane and bolting self-adjustment resulting frea crane operation. A site design change (OCA-9872, Revision 4, dated August 24,1984) was issued to document the acceptability of the gaps in excess of 1/16 irich which were identified in the above NRC inspection report.
During further investigation of the allegation that shims for the ,
rail support system of the polar crane had been altered during installation, the TRT observed gaps which may have been excessive between the crane girder and the girder support bracket. Detailed ;
specifications addressing the gap tolerance in the girder seat con-nections did not exist; however, Gibbs & Hill letter GHF-2207, dated November 28, 1977, stated that the " seated connections will not require shimming since the area in bearing is at least the width of the bottom flange of the crane girder." Contrary to this Gibbs &
Hill assumption, the TRT observed nine girders with gaps which
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extended under the bottom flange that reduced the bearing surface to-less than the 20-inch flange width stated in the letter. The TRT i also observed conditions which indicated that the crane rail may i still be moving in a circumferential direction, that three rail-to- !
- rail ground Wires were broken, that two shims have partially worked )
out from under the rail, and that two Cadwelds were broken.
Accordingly, TUEC shall j
\
- 1. Inspect the polar crane rail girder seat connections for the '
presence of gaps which reduce the bearing surface to less than the width of the bottom flange and perform an analysis which will determine whether existing gaps are acceptable or require corrective action, i -
i
- 2. Determine if additional rail movement is occurring and, if so, 'N provide an evaluation of safety significance and the need for corrective action.
- 3. Perform a general inspection of the polar crane rail and rail support ; system, correct identified deficiencies of safety sig-nificance, and provide"an assessment of the adequacy of existing maintenance and surveillance programs.
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'J-8 g+,3s gc e UNITED STATES NUCLEAR REGULATORY COMMISSION b/' fG ~ [4 '
- 4 W ASHINGTON. D. C. 20555 5
oL% % y/
{* C,T****
January 8,1985 Docket Nos. 50-445/446 Mr. H. D. Spence, President Texas Utilities Generating Company 400 North Olive Street Lock Box 81 Dallas, Texas 75201
Dear Mr. Spence:
Subject:
Comanche Peak Review On July 9, 1984, the domanche Peak Technical Review Team (TRT) began an intensive onsite effort to complete a portion of the reviews necessary for the ~
NRC staff to reach its decision regarding the licensing of Comanche Peak Unit
- 1. The onsite effort covered a number of areas, including the review of allegations of improper construction practices at'the facility.
On September 18, 1984, the NRC met with you and other Texas Utilities Electric Company representatives to provide you with a request for additional infor-mation in the electrical and instrumentation, civil and structural, and test program areas having potential safety implications. On November 29, 1984, we reported to you on the status of our technical review in the protective. .
coatings area and requested additional infomation in the mechanical, and miscellaneous areas. TRT reviews of construction QA/QC allegations and ,
technical issues have progressed to the point where we can now provide you '
with the status of our efforts in the construction-QA/QC area and a request for a program plan specifically addressing our concerns. Further background informcon regarding these allegations and technical issues will te published in Supplements to the Comanche Peak Safety Evaluation Report (SSER),
which will document the TRT's detailed assessment of the significance of all issues exanined.
The TRT effort constitutes one element in the process of the agency's review of the Comanche Peak license application. The OA review group on the TRT was comprised of about 20 individuals having a total of over 300 years experience in nuclear engineering, OA, and related fields. This group spent several months at the Comanche Peak site examining the construction QA program in -
depth.
The TRT findings are provided in the enclosure to this letter. We have not proposed specjfic TUEC corrective actions as we have in previous reports from the TRT. We request that you evaluate the TRT findings and consider the implications of these findings on construction quality at Comanche Peak. We request that y6u submit to the NRC, in writing, a program and schedule for completing a.4etailed and thorough assessment of the QA issues presented in the enclosure to this letter.
c M c=mrh l ) L13 pg.
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l Your programmatic plan and the plans for its implementation will be reviewed .
and evaluated by the staff before NRC considers the issuance of an operating license.for Comanche Peak Unit 1. The TRT considers the construction QA/QC j findings to tre generic to both Units 1 and 2 and your program plan and schedule i should address both units. This program plan shall: (1) address the root cause {
of each finding and its generic implications on safety-related systems, j programs, or areas, (2) address the collective significance of these ,j deficiencies, and (3) propose an action plan from TUEC that will ensure that j such problems do not occur in the future. Your actions should consider the use of management personnel with a fresh perspective to evaluate the TRT's findings and implement your corrective actions. Finally, you should consider the use of an independent censultant to provide oversight to your program.
The findings of TRT with respect to QA/QC allegations, along with the TRT's assessments of your response to this letter, will be provided to the Senior Management Panel on Contention 5 established by the Executive Director on '
December 24, 1984. The Senior Management Panel will detemine an overall NRC staff position on Centention 5 based on an integrated review of a number of sources of infomation concerning QA/QC at Comanche Peak in addition to the TRT findings, including information from the CAT team, the SRT team, 01, Region IV and the Hearing Board.
The TRT's overall evaluation of the technical issues and allegations is nearing ccepletion. As we finalize infomation received in conversations with allegers, and further assess the implications of our findings we will infom '
you of additional concerns, as they arise. In the mean time, your examination of the potential safety implications of the TRT findings should include, but nct be limited to the creas or activities selected by the TRT.
In order to fully discuss these concerns with you we are scheduling a meeting for Jenuary 17, 1985 which will be held in our office in Bethesda, Maryland.
This meeting will provide an opportunity to ask questions regarding these concerns prior to formulating your program plan. Additienal meetings will be held at NRC request as your program plan is fomulated.
This request is suhitted to you in keeping with the NRC practice of promptly notifying applicants of outstanding infomation needs that could potenti. ally affect the safe operation of their plant. Future requests for infomation of this nature will be made, if necessary, as TRT technical reviews continue.
Sincerely,
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or . Os or DivisionohLicensing Office of Nuclear Reactor Regulation
Enclosure:
As stated cc w/ enclosure:
See next page s i
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' COMANCHE PEAK Mr.;M. D. Spence
- 1
-President
- Texas' Utilities Generating Company 400 N.. Olive St., L.B. 81 Dalles. Texas 75201 cc: Nicholas S. Reynolds, Esq.~ Mr. James E. Cumins Bishop. Libeman, Cook - Resident Inspector / Comanche Peak
.Purcell & Reynolds Nuclear Power Station
'_1200 Seventeenth Street, N. W. c/o U. S. Nuclear Regulatory
-Washington,.D. C. 20036 .
Commission P. O. Box.38 Rabert A. Wooldridge, 'Esq. Glen Rose, Texas ,76043 Worsham Fntsythe, Sampels &
Wooldridge .
Mr. Robert D. Martin 2001 Bryan Tow,er, Suite ~2500 U. S. NRC, Region IV -
' Dallas, Texas 75201 611 Ryan Plaza Drive Suite 1000 l a Mr. Homer C. Schmidt- Arlington, Texas 76011 Manager - Nuclear Services
-Texas Utilities. Generating Company . Mr. Lanny Alan Sinkin -
Skyway Tower 114 W. 7th, Suite 220 400 North Olive Street Austin, Texas
- 78701 N" L. B. 81 Dallas, Texas 75201 B. R. Clements . .
Vice President Nuclear Mr. H. R. Rock : Texas Utilities Generating Company
- i. Gibbs and Hill, Inc. Skyway Tower i 393 Seventh Avenue 400 Ncrth Olive Street New York, New York IUD 01 L. B. 81 .
. Dallas, Texas 75201 Mr. A. T. Perker Westinghouse Electric Corporation William A. Burchette, Esq.
P. D. Box 355 1200 New Ha=pshire Avenue, N. W.
-Pittsburgh, Pennsylvania 15230 Suite 420 Washington, D. C. 20036 Renea Hicks, Esq.
Assistant Attorney General Ms. Billie Pirner Garde Environmental Protection Division Citizens Clinic Director !
P. D. Box 12548, Capitol Station Government Accountability Project Austin, Texas 75711 1901 Que Street, N. W.
Washington, O. C. 20009 Mrs. Juanita Ellis, President 4 Citizens- Association for Sound David R. Pigott, Esq.
Energy Orrick, Herrington & Sutcliffe 1426 South Folk 600 Montgomery Street Dallas,_ Texas 75224 San Francisco, Califormia 94111 Ms. Nancy H. Williams . Anthony Z. Roissan, Esq.
CYENA Trial Lawyers for Public Justice 101 California Stn et 2000 P. Street, N. W.
San Francisco, California 94111 Suite 611 '
Washington, D. C. 20036 L
u
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l -- ' COMANCHE PEAK cc: Mr. Dennis Kelley .
,, Resident Inspector - Comanche Peak
- c/o U. S. NRC P. O. Box'1029 Granbury. Texas 76048
'Mr. John W, Beck Manager - Licensing Texas Utilities Electric.fm=pany Skyway Tower 400 N. Olive Street ~
L. B. 81 Dallas, Texas 75201 Mr. Jack Redding -
N Licensing Texas Utilities Generating Company 4901 Fairmont Avenue Bethesda, Maryland 20014 d
0 O
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. i Enclosure l
Technical Review Team Findings Resulting From .
Quality Assurance / Quality Control Allegations In evcluating the QA/QC program at CPSES, the Technical Review Team (TRT) com-pleted the following: (1) interviewed Texas Utilities Electric Company (TUE')
l- and Brown & Root (B&R) personnel and allegers, (2) reviewed quality assurance records, selected affidavits, transcripts and depositions,~ and NRC Regional:and i Office of Investigations reports, and (3) physically insoected hardware to evaluate the safety significance of quality assurance / quality control (QA/QC) allegations at Comanche Peak Steam Electric Station (CPSES). i 1 QUALITY ASSURANCE PROGRAM The TRT found that although the TUEC QA program documentation met NRC require- '
ments, the weaknesses of its implementation in several areas demonstrate that TUEC lacked'the commitment to aggressively implement an effective QA/QC program in several areas:
A. TUEC failed to periodically assess the overall effectiveness of the site QA program in that there have been no regular reviews of program adequacy by senior management. Further, TUEC did not assess the effectiveness of its QC inspection program.
B. During the peak site construction period of 1981-2, TUEC employed '
only four auditors, all of whom had questionable qualifications in technical disciplines. Although charged with overview of all site construction and associated vendors, these Dallas based auditors provided only limited QA surveillance of construction activities.
l 'C. Repetitive NCRs were issued that identified the need to retrain con-struction personnel in the requirements and contents of QA procedures.
One corrective action request (CAR) dealing with inadequate construc-tion training and records remained open for one year. The ioentical problem was identified in a subsequent CAR, which still had not been closed at the time of the TRT's onsite review.
D. The'TRT found inany exacples of incomplete and inadequate workmanship and ineffective QC inspection in TUEC's evaluation of the as-built i program. (See Section 4 for a detailed discussion.) '
E. Some craft workers newly assigned as QC inspectors were in a position to inspect their own work and records. Site management did not view
- this lack of separation between producticn and inspection roles as a l
potential conflict-of-interest.
F. There were potential weaknesses in the TUEC 10 CFR 50.55(e) deficiency-reporting system. Applicable procedures did not identify what types 1
L_ _ - _ _ _ _ _ _ _ _ _
i of deficiencies constituted significant breakdowns in the QA program, t nor how they should be evaluated for deportability to the NRC. Evalu-ation guidelines for reporting hardware deficiencies lacked clarity f and definitive instructions and the threshold for reporting deficien-cies was too high. Specific past and present construction deficien-
. j cies that were not reported by TUEC are listed in Sections 4, 5 and !
11 of this enclosure. l G. The TUEC exit intervSw system for departing employees appeared to be j neither.well structured nor effective, as evidenced by.the lack of l empicyee confidence, limited implementation, failure.to document J explanations and rationale, and failure to complete corrective actions and to rietermine root causes.
H. The B&R corrective action system was generally ineffective and was ,
bypassed by the B&R QA Manager, as exemplified in the following i instances:
i l
- 1. There were no definitive instructions to describe the types of .
problems that required corrective action. Minimal procedural s instructions resulted in corrective action decisions frequently being left to the judgement of the QA Manager.
- 2. Since June 1983, B&R had issued no Corrective Action Requests (CARS), and was substituti.ng memos and letters of concern for this functio'n. This shortcut had become a regular method of operation and appeared to bypass the CAR system.
I. The TUEC corrective action" system was poorly structured and ineffec- -
tive in that:
- 1. Controlling procedures were brief and general.
- 2. There was no translation of FSAR requirements on trending and no details on how trend analyses were to be accomplished.
- 3. Quarterly reports were not issued in a timely manner.
- 4. The method of categorizing problems by building did not assure meaningful trend analysis.
- 5. A 1984 CAR report identified three items requiring action; how- 3 ever, none had been taken. i i
- 6. CAR 029 was used as a vehicle for a specific disposition rather than for generic action, as intended by the CAR system.
2 QUALITY CONTROL INSPECTION The TRT evaluated the CPSES QC program to determine if it was functionally l effective a'nd if the QC system and organization effectively ensured consistent j quality of-design, procedures, processes and product at the plant. The results !
of this review showed the following problems. '
s
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l A. ' Based on the TRT review of about 200 fuel pool travelers, TUEC was unable to maintain an effective and controlled QC program for fuel pool liner fabrication, installation, and inspection. Typical fuel pool traveler irregularities were:
j' 1. There was apparently a routine practice during construction of the fuel pool that allowed craft personnel to. complete a portion of the inspection report forms prior to'the actual inspection.
Craft personnel entered the word " SAT," dated the entry, and left blank only the space for the QC inspector'r, signature. It appeared that the craft personnel were judging the inspection results prior to inspections.
- 2. The'date accompanying the signature for visual examination of an inside weld was changed to a date that appeared to precede the examination.
- 3. Entri23 by the same inspector for two different inspections did not appear to match in that one entry appeared to be written by another person.
- 4. The procedure number for a dye penetrant inspection was changed -
.by an inspector different from the one who conducted the
' inspection. .
- 5. The date for a dye penetrant inspection was changed by'an ;
inspector other than the one who performed the inspcetion.
- 6. Fuel pool travelers were found with missing QC signoffs for fitup and cleaniness. No proof could be found that some of the required weld fitup and cleanliness inspections were ever i performed. i
. 7. The TRT review disclosed the following irregularities with j traveler entries in addition to those listed above:
1 (a) Date changes after the fact '
(b) Signoffs for functions out of sequence (c) Corrections after the fact (d) Changes to first party . inspector date signoffs (e) Missing signatur6s B. There were examples of limited corrective action, including vendere supplied pipe whip restraints that had received inadeqeate source inspections.- Tw1ve NCRs were issued involving weld defects on these restraints. TUEC corrective action included paint removal from only
- a. sample of the welds and 21 restraints were selected for reanalysis; however, the TRT found no basis or criteria for paint removal or how the worst case restraints were identified.
The reviews of allegations in the Civil and Structural, Coatings, Electrical, Test Programs, and Piping and Mechanical areas also incicate DC inspection deficiencies, as provided in our letters of September 18, and November 29, 1984.
3
3 T-SHIRT INCIDENT 1
The T-shirt incident has previously been explored in many f orums, including hearings before the Atomic Safety and Licensing Board. The TRT has examined this matter, but will not now describe all of the associated issues. Impor-
- tantly, however, the TRT believes that TUEC management failed to adequately investigate the incident to determine its root cause, but reacted as though the QC inspectors involved were guilty of disruptive behavior. Of particular concern to the TRT is the strong perception that TUEC QA management may have acquiesced to pressures and complaints from construction personnel and may have failed to adequately support their QC workforce.
4 INSPECTIONS OF AS-SUILT PIPE AND ELECTRICAL RACEWAY SUPPORTS The TRT conducted a series of inspections encompassing as-bdilt safety-related I pipe support and electrical raceway support installations. These inspections were of co'npleted systems or components that had been previously inspected and accepted by TUEC QCi as meeting the respective construction and installation requirements. j s i A. Pipe Succort Inspections I Tables 1 and 2 are indicative of the scope of the TRT pipe support as-built inspection effort. Of the 42 pipe supports inspected, 37 were randomly selected, wnile 5 originated from an alleger's list. Forty-six ceficien-cies were identified in the supports inspected. Following are examples of the deficiencies identified and the applicable criteria. TUEC's final QC inspections of tnis sample ranged fro:n Dece=ber 1982 to Octcber 1984.
- 1. Component Secoort Welds:
(a) Applicable c-iteria ASME Section TII, NF Subsection and subarticles NF-4424 and NF-5360 set forth rules for examining welds.
B&R 01-0AP-11.1-28 Revision 25. Paracraoh 3.5._5.1 delineates criteria for the examination of welcs,. including inspection parameters for acceptable weld sizes.
The TRT found supports exhibiting welds that did not appear to be in accordance with the above-refenneed codes and procedures.
(b) Ex.a=ples of deficient welds (1)' Suppert No. AF-1-001-001-533R. Discrepancies included porosity; insufficient weldleg; incumplete welds and
. insufficient fill. This support was removed, scrapped, and coepletely rebuilt subsequent to the TRT inspection.
mas 4
-___ _ _ - - - _ _ - - - - - . _ - - - - - - ------------------_-------_----------------------------------------J
c-.- --.- - - - - , - . , - - . - - - - - . , . - , - - - - , - . , . - - . - - - . , , - - - . . - - , . - - - - , , - - - - - -
Table 1' Pipe supports in unit 1 Supports: Inspected by TRT As-Built group -
842 Class'1 supports inspected 4 Class 2 supports inspected 14
~
Class 3 supports inspected 24 .
Hangers wit.h problems 26
. Total problems identific.i 46 Procedure adequacy problems 5 Hardware-related problems 16 As-built drawing related problems 8 Component identification problems 2 Weld-related problems 10 QC record problems 1 Material identification problems 4 Welds inspected without paint by TRT 305 Welds inspected with paint by TRT 89 Total welds inspected by TRT 394 Welds needing weld repair 10
% of welds inspected 2.5%
' Supports needing welding repair 6
% of supports inspected 14%
No. of Supports Bldo System Inspected Containment Safety Injection (SI) 1 Containment Reactor Coolant (RC) 6 Containment Residual Heat Removal (RHR) 2 Fuel Handling Component Cooling (CC) 11 j Safeguards Residual Heat Removal (RHR) 1 1 Safeguards Containment Spray (CT) 8 Safeguards Demineralized Water (DD) 1 Safeguards Auxiliary feedwater (AF) 8 Auxiliary Chemical Volume & Centrol (CS) 1 Safeguards Main Steam (MS) 2 Safeguards Chilled Water (CH) 1 "All 42 pipe supports inspected by the TRT had been previously accepted by site QC. .
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1 (2) Support No. AF-1-001-702-533R. Exhibited extraneous welding that was not documented on the as-built drawing. One of the required welds was' undercut. beyond the limits of acceptance (this weld was subsequently repaired).
(3) Support No. CC-1-126-013-F33R. Support drawing required a 1/4" fillet weld to connect item 5 to item 6. This weld was omitted in the field.
(4) Succort No. CC-X-039-007-F43R. A required 5/16" all-around fillet weld had an approximately 1/16" undersize weld leg for the length across the top flat of the tube steel.
(5) Support No. RH-1-006-012-C42R. An all-around 1/4" fillet.
weld connecting item.5 to item 7 was undersized by 1/32" to 1/16" across the top.
(6) . Support No. AF-1-037-002-533R. This support exhibited a 1/16" to 3/32" reduction in plate thickness and weld size .
due to excessive grinding of the weld at the base plate. s Base material thickness of the support plate was reduced beyond the limits of acceptance in three locations.
(7) Su cort No. CT-1-013-014-532R. ExcessiveoYergrindingof welas resulted in notching of the sway strut rear brackets.
This condition was repaired subsequent to the TRT inspection.
- 2. -Lockinc Device for Threno'ed Fasteners: .
(a) Applicable criteria Subarticle NF-4725 states in part that all threaded fasteners,.
except hign-strength bolts, shall be provided with locking devices to prevent loosening during service.
ASME Sect. III, Div.1. Interpretation No. III-1-83-49R provides that the user snculd satisfy himself tnat any other cevice than those described in NF-4725 is capable of acting as a locking device under all service conditions.
Brown & Root procedure 01-0AP-11.1-28, Attachment 2, Doeration 7, Insoection Attribute h., requires that all exposed tnreads be free of extrnneous cateriai.
CPSES/FSAR, Paracraoh 17.1.2 states that the design verification procedure assure that crawings, specifications, procedures, and
. instructions meet stipulations of related codes and standards.
10 CFR 50.55(e)(1) directs that the holder of the construction e pemit shall notify the NRC regarding each deficiency found in design and construction which, if not corretted, could adversely affect the safety of operations at any time throughout the expected lifetime of the plant.
s 8
1 I
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l l There appeared to be a difference in locking devices on threaded fasteners for similar pipe support hardware made by two separate vendors. Whereas in some cases Nuclear Power Service Incorporated (NPSI) specified only one l nut and no locking device, ITT-Grinnell required two nuts in those same I applications. If the design of NPSI models indeed should be found to need -
1 the locknuts or their equivalent, there could be hundreds of pipe supports i instal, led without adequate locking devices. ,
(
i The TRT found examples in Unit I where deficiencies existed so that i
' TUEC was in potential violation of the codes, proceduNs, guidelines, and commitments concerning locking devices for threaded fasteners.
1 In spite of the requirements pursuant to 10 CFR 50.55(e)(1), TUEC did not report to the NRC the omission of thread-locking devices in the Unit I nuclear safety systems and did not attempt correctiva action until May 1984, when TUEC tested previously applied paint for thread-lock capability. That test was inconclusive, since it did not estab-lish that the paint, an epoxy process., would reliably perfom as an effective locking device under all service conditions and throughout the expected lifetime of the plant. Further, TUEC could not identify ,
to the TRT which paint was the subject of testing.
TUEC had a potentially inadequate quality assurance specification No. 2323-AS-31, which did not cover inspection of painted threaded fas-teners. The paint was applied to ASME code-controlled, NF hardware per specification 2323-AS-30 (non-Q) which required no inspection. This issue appears to be generic for Unit 1.
The TRT notes that TUEC did not initiate an NCR identifying the widespread problem of missing locknuts; only a Request for Information was generated, which TUEC could not locate for the TRT. An NCR, required by procedure, would have brought the problem and its ramifications to management atten-tion and would have provided a vehicle for controlled, organized, and approved engineering disposition.
. (b) Examples of deficient locking devices.
Pipe support RC-1-901-702-CE25 had a load bolt at a beam attach- {
cient which did not exhibit an approved locking device. (The bolt j material type was SA-307 grade A.) Additionally, pipe support l CS-1-085-003-A42K had no approved locking device on the "special clamp" bolts, even though the design drawing for this clamp showed each bolt with a nut and'a locknut.
- 3. Minimum Edge Distance for Bolts: ' I (a) Applicable criteria OI-0AP 11.1-28 Revision 19, Paracraph 6.1 required that bolt j holes in structural members shall not be closer than 1-1/2 times l
. the bolt diumeter from the edge of the member to the center of I the bolt hole.
9
ASME Sect. III Div. 1, Subsection NA, Appendix XVII, Table XVII-2462-1(b)-1, gives specifically allowed minimum edge dis-tanc.as for bolt holes (reamed, p.unched or drilled) at sheared or rolled edges of plates, shapes, or bars.
(b) Example of minimum edge distance violation The baseplate for pipe support CC-X-039-006-F43R, located in the component cooling system, Room 249A, Fuel Handling Building, violated minimum edge distance criteria for bolt holes,
- 4. Base Plate Hole-Location Oimensions:
(a) Applicable criterion -
i QI-0AP-11.1-28, Revision 19, Attachment 4, Paracraph 2, under I fabrication tolerances, limits a " hole centerline location to 11/4" or as shown on the design drawing."
(b) Examples of hole-location dimension problems s
~
The TRT found the horizontal member of Support CC-1-126-010-F33R was 3 inches lower at its centerline relative to the upper bolt-hole centerline than shown on the vendor-certified drawing. The as-built drawing had not been revised to reflect the actual-instal 1ed condition in the plant. This support was located in the component cooling system, Room 247A, in the fuel Handling Building. Other supports with similar hole-locati'an violations found in the inspecti'ons.were: CC-X-039-007-F43R , .
CC-1-126-011-F33R, and CC-1-126-012-F33R.
- 5. Spherical Bearino Gap:
(a) Applicable criterion Brown & Root Procedure, OI-0Ap 11.1-28, Revision 25 ,
paracraoh 5.7.3.1 states that "a sufficient numoer of spacers shall be used to prevent the spherical bearings from becoming dislodged," and "in no case shall the resulting gap be more than ,
the thickness of one vendor-supplied spacer."
(b) Examples of spherical bearing gap deficiencies
]
An excessive free gap existed between spherical bearing and washers on the sway strut assembly of support CC-1-126-015-F43R.
Other supports with similar bearing gap anomalies found in TRT's inspections wert: RC-1-052-016-C41K, RC-1-052-020-C41X, and
~
MS-1-416-001-S33R. The frequency of this type of procedure vio-1ation in the TRT's limited inspection suggests that this problem is generic for Unit 1.
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- 6. Spherical:Bearino Contamination:
(a) Applicable criterion .
'0I-0AP-11.1-28 Revision 22, Paracraph 6.3.1 Note 2 states in part -'
that " bearing internal and external surfaces shall-be free of j rust and foreign material, and bearing shall move freely within the housing."
(b) . Examples of spherical bearing contamination
, , The TRT found paint contamination in the bearings of both snubber assemblies on component support SI-1-090-006-C41K that severely-
[ ' obstructed the bearing cavities and limited their movement. This Class I component support. is located in the Containment Building' )
of the Unit 1 safety injection system. A similar condition
- exists.on support MS-1-416-002-533R. q I
- 7. Snubber Adapter Plate Boltino - Lack of Full Thread Encacement: '
(a) Applicable criteria QI-0AP-11.1-28, Revision 22. Paragraoh 6.1, states that "all ~
bolts, stucs, or threaced rocs shall have full' thread engagement' in the' nut." '
ASME Sect. III, Div. 1. Subsection NF, Subarticle NF 4711 states that "the threacs of all bolts or studs shall be engaged for the
' full length of thread in the nut."
QI-0AP-11.1-28, Revision 25, Attachment 29 pemits less.than full thread engagement in tnreaced plates. This allowance for..less than full thread engagement is a potential violation of the ASME Code Sect. III, NF-4711; no code case was invoked to set i aside tnis procecure. The requirement of NF-4711 that "the j threads of all bolts or studs shall be engaged for the full l 1engtht of thread in the nut" also' implies that there be a full '
length of a threaded hole in plates, shapes, or bars where the required threaded hole length is the same as the bolt diameter.
Further, there is no evidence that partial thread engagement at the snubber adapter plate connection has been given consideration l in the design procedures for linear-type supports, nor does it .
appear that sufficient design margins have been introduced to allow for less than full-threaded connection. The TRT did not-check "as-built" analyses to detemine whether any such varia-tions from the design nors had been considered in the "as-built" i stress calculations.
What is in question is whether any calculations had been made to address this particular thread engagement condition for each size snubber being used in the plant.
l
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l (b) Examples of lack of full thread engagement Snubber (' shock arrester) adapter plate bolt threads were insuffi-ciently engaged in all four threaded holes of component support MS-1-416-002-533R. The worst condition was 0.095" short, or more - ,
i than 25% less than full thread engagement. Similar lack of full {
l thread engagement deficiencies was found on NF supports SI-1-090-006-C41K and CT-1-013-012-532K.'
- 8. Threaded Rod Thread Encanement:
(a) Applicable criterion 01-0AP-11.1-28, Revision 21, Paracrach 6.3. 2. a. directs that "QC shall verify thread engagement if site [ sight] holes are present !
in the strut body." '
(b) . Example of rod thread engagement deficiency Sight holes were present in the strut body to verify threaded ' I rod engagement. The rod was not visible through the sight hole for support RC-1-901-702-C825. -i
(a) Applicable criterion 01-0AP-11.1-28. Revision 22, Paraorroh 6.3.1.1.b states that "the size of tne cotter pins, wnen used, should be the maximum size i the hole will acccmodate and shall be fully opened."
(b) Exar:ple of locking device deficiency l
.5way strut No. AF-1-001-014-533R had a broken cotter pin.
- 10. Lead Side of Pioe Clamo Halves Not Parallel:
(a) Applicable criterion i
'01-0AP-11.1-28, Rev. 25, Sec. 3.7.3.1 states that " pipe clamp halves, in relation to attaching eyerod end, shall be parallel."
(b) Examples of halves not parallel Clamp halves for pipe supports AF-1-001-001-533R and AF-1-001-014-533R were not parallel.
- 11. P.ipe Clearances Outside of Allowable Tolerance:
(.a) Applicable criterion OI-0AP-11.1-28, Revision 19, Attachment 4, item 3.b states'"where l the design shows 0" on one side and 1/16" on the other, 0" must I be maintained while 1/16" i 1/32" is required on the other side." )
12
)
1 (b) Ext.7.ples of pipe clearance violations '
I
, Pipe support CC-1-125-013-F33R exhibited no clearance on top or bottom, while the hanger drawing' called out 0" on the bottom and 1/16" on top. A similar problem existed for pipe support .
AF-1-001-702-533R.
- 12. Pipe Clamp Locknut Loose: . .
(a) Applicable criterion OI-0AP-11.1-28 Revision 21, Sect. 6.1 states that "unless other-wise shown on the drawing, fasteners will be tightened securely."
(b) Example of loose locknut A pipe clamp locknut for pipe support AF-1-035-011-533R was found loose (less than finger-tight).
13.- Snubber / Sway Strut Misalignment: -
(a) Applicable criterion $,
01-0AP-11.1-28. Revision 18, Sect. 6.3.1.d states that " maximum sway strut misalignment shall not exceed 5* for ITT-Grine11 and .i NPSI from the centerline of the sway strut."
(b) Examples of misalignment Pipe support CC-1-126-014-F43R exhibited angularity that exceeded this requirement. A similar problem existed with pipe support 3 RC-1-052-020-C41R. l
(a) Applicable criterion l
01-0AP-11.1-28, Revision 24. Sec. 3.8.3.5.6 states that "devia-tion of more than 2 1/8" from the specified cold setting (AC dimension shown on the design drawing) is not permitted, unless autborir.ed by a design change."
(b) Example of incorrect AC dimension !
Pipe support CS-1-085-003-A42K deviated by approximately 1" f[un the cold set dimension shown on the design drawing.
- 15. Support Configuration Did Not Match Drawing:
(a) Applicable criterion
,_ QI-0AP-11.1-28, Revision 24, Attachment 2, Operation 3 lists the {
following inspection attribute: " support configuration complies 1 with the design drawing."
i i
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(b) Examples of configuration problems Pipe support snubber CT-1-005-004-522K was installed end-to-end opposite from the orientation shown on the drawing. A similar problem existed with pipe support CT-1-013-010-522X, where dimen -
sional discrepancies. existed on the support drawing that detailed the orientation of the snubber.
- 16. - Component Type /Model No. Installed Did Not Match Drawina:
(a) Applicable criterion 01-0AP-11.1-28, Revision 24, Sect. 3.2.1.1 states that " vendor-supplied NPT stamped component supports shall bear marking (i.e.,
name plate) traceable to the design drawing."
(b) Examples of component identification problems.
Model numbers of installed snubbers for pipe support .
SI-1-090-006-C41K did not match the model number on the design N drawing. A similar problem existed with pipe support RC-1-052-020-C41R. ',.
s
- 17. Weld Data Card Missino OC Initials For Welds: .
(a) Applicable criterion 01-0AP-11.1-28, Rev. 25, Paracraoh 3.5.3 Welder and Weldino Material verification states that "The QCI shall verify that the s welder is qualified to make the weld utilizing the welder quali-fication matrix (attachment 16, typical), that the use of the WPS (Attachment 17, typical), and the type of filler pMterial listed on the WFML [ weld filler material log] are the same as those listed on the weld data card (WDC), and the welder's symbol has been recorded on the WFML."
, (b) Example of deficient weld data card ,
Support number CC-1-126-013-F33R had some welds performed with no QC inspector initials or signature on the corresponding blocks of the weld data card for that support inspection package.
- 18. Identification of Materials and Parts:
(a) Applicable criteria 10 CFR 50 Accendix B, Criterion VIII states that " measures shall
. assure that identification of the item is maintained by heat
- number, part number, serial number or other appropriate means l' ,
either on item or on records traceable to the item, as required throughout fabrication, erection, installation and use of the item."
14
OI-0AP-11.1-28, Revision 19, Sect. 3.1.2 states that "at installation inspection, the QC inspector shall verify the hanger number, the material type, grade and heat number ... using the information provided on the Material Identification Log." 1
~
1 (b) Examples of material identification deficiencies l
~ l A replacement part (sway strut everod) for pipe support CT-1-013-014-532R had no apparent material identification either on the hardware or in the documentation package for the support, i The Material Identification Log (MIL) did not list any identi- !
fication traceable to the origin of the replacement part. A i similar problem existed with pipe supports CC-1-126-012-F33R, CC-X-039-005-F43R, and AF-1-035-011-533R.
1 B. Deficiencies with Hioh Rate of Occurrence j l
The following pipe support inspections by the TRT were in addition to those ;
already listed in the previous examples. Results of these ancillary -)
inspections are summarized in Table 3. .
j l
The TRT identified six specific deficient items which need further evalua- j tion to assess their generic implications. The TRT concern is that these ;
items may have a high rate of occurrence throughout plant safety-relat;ed )
systems. The specific " frequently occurring" items and relevant inspec-tion criteria were as follows: l J
(1) Stritt and snubber load pin spherical bearing clearance with washers was excessive (Ref. QI-QAP-11.1-28, Sec. 3. 7. 3.1 Rev. 25). ~
(2) Strut and snubber load pin locking devices (cotter pins or snap lock rings) were damaged or missing (Re'. QI-QAP-11.1-28 Rev. 25, which did not specifically address load pin locking devices).
.(3) Pipe clamp halves on load side were not parallel (Ref. QI-QAP-11.1-28, !
Sec. 3.7.3.1 Rev. 25). !
(4) Bolts threaded into tapped holes of snubber acapter plates had less than full thread engagement (a " frequently occurring" deficiency; see related discussions on pipe supports, example 7 " Snubber Adapter Plate Bolting - Lack of Full Thread Engagement" within Part A of this {
section on as-built inspection).
l (5) "Hilti Kwik" bolts (concrete expansion anchors) as installed did not meet minimum effective ed edment criteria (Ref QI-QP-11.2-1, Sec. 3.5.1 Rev. 16).
(6) Locking devices for threaded fasteners were missing or of a non-approved type (see item 2 " Locking devices for threaded fasteners" on ,
p.ipe support deficiencies within Part A of this section on as-built 1 inspection). J l
a
i Table 3 Summary of additional TRT inspections )
I Area: Room 77N, El 8'10'-6" Unit 1, Safeguards Bldg .
No. of Supports No. of. Supports Deficiency Inspected Deficient- % Deficient.
Item 1. Excessive 92 5 5.4%
Spherical ' Bearing Clearance' Item 2. _ Load Pin Locking 92 14 15.2%
Device Missing Item 3.- Pipe Clamp. Halves 40 9 22.5%
Not Paral,lel Item 4. Snubber Adapter 19 *13 to be s ^l Plate Bolts With determined Lass Than Full Thread Engagement.
Area: Cable Spread Room 133, El 807'-0" Unit 1, Auxiliary Bldg Deficiency Bolts Insoicted Nur.ber Deficient % Deficient c Item 5. Hilti Kwik Bolt 24 3 12.5%
Does Not Meet Minimum Embedment**
- Bolts had less than full thread engagement.
- Taking into account the " allowed" slippage of the bolt for a distance of cne nut thickness due to torquing (Ref. " Installation of 'Hilti' Drilled-In Bolts" 35-11c5-CEI-20, Rev. 3, Para. 3.1.4.1) and the minimum specified embedment, the above Hilti bolts violated the " effective" embedment requirements.
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e
/g The TRT undertook additional hardware inspections to ascertain the regu-larity with which these specific items may exist. All accessible pipe i supports in Room 77N, at the 810-foot, 6-inch elevation of the Unit 1 Safeguards Building, were inspected for " frequently occurring" defi- l ciencies 1, 2, 3 and 4 listed above. To assess the level of occurrence of -
I
" frequently occurring" deficiency 5, electrical support 'Hilti' baseplates i located in the Cable Spread Room 133, at.the 807-foot elevation of the Unit _ r Auxiliary Building, were inspected: For details on " frequently occurring" deficiency 6, see item A.2, " Locking Device for Threaded Fas-teners," of the pipe support deficiencies, described above.
C. Electrical Raceway Supoort Inspections The TRT inspected electrical conduit supports and cable tray hangers l to the requirements of QI-QP-2.1.10-1, Inspection of Seismic Electrical -
Suoport and Restraint Systems; QI-QP-11.21-1, Requirements of Visual j Weld Inspection; and other applicable instructions for conduit support and cable tray hanger inspections. .All electrical raceway supports included in TRT inspections had been previously QC accepted. Table 4 -
summarizes the results of the TRT inspections not previously provided as part of our letter of September 18, 1984. 1 1
The TRT found the following discrepancies during its inspection of selected ' electrical conduit supports and cable tray hangers in Unit 1:
- 1. Undersize Welds:
i (a) Applicable criterion DCA 3464, Rev. 23, oaoe 3 of 32, note 3 states in part that "welcing requirements as shown on various details should be read as the minir.um requirement."
(b) Era:=ples of undersize welds Three of four welds on conduit support C120-21-194-3 (cable spread room) were undersized. The required weld size was 1/4" at all weld joints, while the measured weld size was 7/32" to 5/32" for the full lengths of three out of the i four welds.
Similarly, cable tray hanger CTH 5824 (Containment Building) had 12 undersize welds. The all-around welds on the six horizontal beams should be 1/4" in size, according to '
details L and 2 L2 on Drawing FSE-00159, sheet 5824,1 of 2.
The measured size'of these welds was 3/15" to 5/32" at each connection. Also, support IN-SP-7b exhibited undersize
. welds measuring 7/22" to 5/32" instead of the required 1/4".
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1 Table 4 Summary of electrical raceway support inspection by the TRT - unit 1 I Support welds inspected -
59 I Supports inspected 5*
Supports with problems * !
3 (60%) i Types of problems l l
Hardware-related, other than welding 6 {
Unauthorized configuration change 1 Weld-related types of problems (categories) l 2
Welds requiring rework 41 ;
Welds made in field but not recorded on drawing 80** j Beam stiffeners added but not recorded on drawing 40 Building / Area Sucoorts Cable Spread Room CTH 12646 -
s C 130-21-250-3 .
C 120-21-194-3 Auxiliary Building CTH 6742 Containment CTH 5824 i
- All electrical supports inspected by the TRT had been previously inspected and accepted by QC. <
- Full visual inspection was not performed by the TRT on these extra welds. I i
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] , 4
- 2. Misplaced Welds: 1 (a) Applicable criterion .
01-0P-11.10-1, Revision 29, Paragraph 3.5.2, Assembly .
Inspection,-includes the requirement to inspect.a support for. configuration. ~ Paragraph 3.6.2 of the same. procedure
- requires that support welds' receive visual inspection and
-that nonconforming welds be reported.1 1
h (b) Examples of misplaced welds d During inspection' of Hanger CTH-6742, the TRT found that two -
structural welds were.made in the wrong. direction. The .
3/16" shop welds which join MK-10 and MK-11 were made hori- -
zonta11y instead of vertically, as shown on drawing- .
FSE-00159, sheet 6742. QC Inspection. Report ME-I-0024909, j
, dated February 16, 1984, accepted all inspectable attributes ;
as' satisfactory prior to 'the TRT . inspection. L
- 3. Unauthorized Configuration Changes-(a). Applicable criterion QI-QP-11.10-1.- Inspection of Seismic Electrical Support' and Restraint Systems, paragraon 3.5.2 includes the. requirement-for inspection of a support for configuration compliance. ,
(b)' Examples of configuration change -
The TRT found that cable. tray hanger CTH 5824 (Containment
. Building) 'had been fabricated to include 40 more stiffeners and 80 more welds than required or shown on drawing FSE-00159, . sheet 5824, 2 of 2, Detail L2 . -Inspection Report i HE-1-0006155 verified. final QC inspection and acceptance on l January 3, 1984.
Further, cable tray hanger CTH-6742 (Auxiliary Building),
Clip, MK-12, should be 6" x 6" x 3/4" angle stock in accord-ance with FSE-00159, sheet 6742. The actual flange thick-ness of MK-12 was 3/8".
- 4. Hilti Anchor Bolt Installation Deficiencies:
(a) Applicable criterion 01-0P-11.2-1, Concrete Anchor Bolt Installation, provided
, . requirements for proper installation and inspection of Hilti anchor bolts.
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L (b) Examples of Hilti bolt deficiencies i CTH-6742 (Auxiliary Building) anchor bolt torque was not verified (paragraoh 3.5 of the procedure). Hilti bolts were i
not marked in accordance with attachment 1 of the procedure, * '
nor was the length of these bolts verifiable (paragraph 3.2).
~
l CTH-5824 (Containment Building) base plate bolt holes had ,
violated minimum edge distance--edge distance cannot be~less '
than 1 7/8" (Attachment 2 of the procedure). Actual dis-tance was 15/8" to 13/8" from the nearest plate edge.
This condition affected five of the eight Hilti anchor bolt holes in the base plates for this hanger.
One Hilti bolt was skewed to more than 15 degrees. Maximum allowable skew was 6 degrees without corrective bevel washers (paragraoh 3.1.2).
'The Hilti bolt torque on this hanger CTH 6741 (Auxiliary .
Building) was not documented as being verified by QC N (paragraph 3.5).
- 5. Undersize Nuts:
There was inconsistency in the application of nuts for SA-325 bolts in that both standard and heavy hex nuts were used. No stipulation was found which would permit the use of standard (non-beavy) hex nuts., This condition is a potential violation of the Haterial Specification ASTM A325 (ASTM, Part 4-1974) -
paraoraon 1.5, whicn provides that " heavy nex structural bolts and heavy hex nuts shall be furnished unless other dimensional requirements are stipulated. .. ." B&R Drawing No. FSE-000159, sheet 5824, 2 of 2, required the use of ASTM A325 bolts for cable tray hanger number CTR-5824.
D. Sen?,ary of Pioe Supoort and Electrical Racewav Suceert Inspections The as-built verification effort conducted by the TRT provides evi-dence of faulty construction by craft personnel, installed hardware !
that does not match as-built drawings, and ineffective QA and QC inspections. Despite the small size of the TRT's sample, there appears to be a large number of deficiencies. The potential also exists that
. these deficiencies are not represented correctly in the final stress analysis.
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1 5 DOCUMENT CONTROL The TRT evaluaug Qe CPSES document control system to determine if it was -)
effective and if it ensured consistent quality 'of documents for construction practices and records. The results'of this review showed the following .
p robl e.ns.
A. The TRT found that there was a potential for document control center (DCC) )
field distribution centers (satellites) to issue deficient document packages i to craft personnel. Typical problems identified were: packages were not I thoroughly examined; procedures and guidelines were not specific or were I not followed; and documents controlling operation of the centers existed in the form of guidelines and charts rather than as controlled procedures.
B. The TRT found that many problems indicative of inadequate drawing control ;
existed at CPSES from September 1981 to April 1984. These problems had '
been identified prior to the TRT's evaluation by both TUEC and NRC Region IV audits and reviews.
Prior to placing the satellites in operation (a phased effort between ~
February and August 1983), DCC distributed drawings, component modifica-tion cards (CMCs), and design change authorizations (DCAs) to file custo-dians, welding engineering, .the pipe fabrication shop, QC, and the hanger task force. Document control through this system proved to be ineffective.
In an attempt to correct identified " problems, OCC satellites were created to distribute drawings to field personnel, rather than use the file custo-diens. However, between August 1983 and April 1984, recurring problems l
with document control were identified. Exar::ples of the types of document /
control problems that existed between August 1983 and April 1984 were as '
follows:
- 1. Drawings released to the field were not current.
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- 2. Drrwing and rpeci*ication changes were not cu rer.t.
- 3. Design documentation packaoes were inec=plete.
- 4. DCC did not provide the satellites with up-to-date drawings, CMCs, DCAs and document revisions.
- 5. Drawings hanging from an open rack, which had no checkout control, were available to craf t and QC personnel.
- 6. Design change logs were inaccurata.
- 7. Design documents were not always properly accounted for in DCC.
, 8. C'u rrent and superseded copies of design documents were filed l together.
- 9. 5.atellite distribution lists were inaccurate.
- 1D. There were discrepancies between drawings contained in the satellites and those in DCC. I 21
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- 11. Some drawings were missing from the catellite files. l l
- 12. Telephone requests for design documents resulted in the issuance of i documents that bypassed the controlled distribution system. j
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In April 1984, top management took a direct interest in recurring document control problems. Their efforts appear to have been successful.
For in~ stance, in April 1984 satellites 306 and 307 had error rates of 30% ;
and 10%, respectively; but by July 1984, these error rates had fallen to i l 1ess than 1% for both satellites. The TRT has found that TUEC document I
! control after July 1984 was adequate; however, the effects of document I control inadequacies prior to July 1984 have yet to be fully analyzed by j TUEC.
C. Deficiency repor'.ing procedure'CP-EP-16.3 appeared to relate only to craft and engineering personriel and was not directed to noncraft and nonengineer-ing personnel who may have had knowledge of reportable items. Procedure CP-EP-16.3 indicated that the applicable manager was responsible for docu-menting and reporting Deficiency and Disposition Reports (DDRs);' but there -
were no checks or balances to ensure that a annager or a designated substi'-
tute would process a DDR.
D. TUEC did not consider the CYGNA audit findings regarding the DCC as appropriate for formal reporting to the NRC pursuant to 10 CFR 50.55(e),
as required by procedure CP-EP-16.3, " Control of Reportable Deficiencies."
E. The TRT found that the DCC issued a controlled copy stamp to the QC depart-ment to expedite the flow of hanger packages to the Authorized Nuclear Inspector. Methods for this kind of issuance and control of such stamps
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were not described in TUEC's procedures.
6 TRAINING / QUALIFICATION The TET identified numercur weaknesres dur%g its revie+of the ASME and non-ASME training, certification, and qualification of QC and DCC personnel. TUEC's training and certification program lacked the progra:rnatic controls to ensure ;
that the requirements in 10 CFR 50, Appendix B were achieved and maintained.
The items identified by the TRT include those listed below, in addition to tne items previously provided in our letter of Septe=ber 18, 1984.
A. Twenty percent of the training records reviewed contained no verifica- l tion of education or work experience.
B. The results of Level I certification tests were used for some 4 Level II certifications rather than the re.sults of a Level II !
test. {
C. A.fter failing a certification test, a candidate could take the identical test again.
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,-. o a D. Certifications were not always signed or dated.
E. White-out was used on certification' tests. :
F. Seven inspectors had questionable qualifications. ~
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G. There was no limit or control on the number of times an examina-
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tion could be retaken.
H. No guidelines were provided for the use of waivers for on-the-job training.
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1 I. In some cases decertification was accomplished by a simple "yes" from a supervisor.
J. There was no formal orientation training for DCC personnel prior to August 1983.
K. The responsibility for administration of the non-ASME training -l l program was not clearly assigned to a single individual or group.
L. Non-ASME personne1~ capabilities were loosely defined by levels (I, II, III). \
M. There were numerous additional problems in non-ASME certification' testing, such as: no requirement for additional training between ;
a failed test'and the retest; no time limitation'between a failed o test and a retest; two different scoring methods to grade a test and a retest; no guidelines on how a. tert question should be- 1 disqualified; no program for. periodically establishing new tests except when procedures changed; and no details on how the administration of tests should be monitored. -
i N. The exemption provision in ANSI 445.2.6, which allowed substitution of previous experience or demonstrated capability, was the normal method for qualifying inspection personnel rather than the exceptional method. q 7 VALVE INSTALLATION The TRT found that installation of certain butt-welded valves in three systems
. required removal of the valve bonnets.and internals prior to welding to protect temperature-sensitive parts. The three systems involved were the spent fuel cooling and cleaning system, the boron recycle system, and the chemical and -
volume control system. This installation process was poorly controlled in that disassembled parts were piled in uncontrolled areas, resulting in lost, ,
damaged, or interchanged parts. This practice created the potential for inter- !
changing valve bonnets and internal parts having different pressure and temper- !
ature ratings. t ase
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l 8 'ONSITE FABRICATION The TRT findings regarding onsite fabrication shop activities indicated that:
A. The scrap and salvage pile in the fabrication (fab) shop laydown yard was not identifiec: and did not,have restricted access.
B.' daterial requisitions prepared in the fab shop did not comply with the applicable procedure.
C. The fab shop foremen were not familiar with procedures that controlled the work under their responsibility.
D. Fabrication and installation procedures-did not include information to ensure that B&R-fabricated threads conformed to design specifications or to'an applicable standard.
- E.
Indeterminate bulk materials that accumulated as a result of-site cleanup operations were mingled with controlled safety and nonsafety ' 4 material in the fab shop laydown yard.
- F. Site surveillance of material storage was not documented. <
G. Work in the fab shop was performed in response to memos and sketches instead of hanger packages, travelers, and controlled drawings.
9 HOUSEKEEPING AND SYSTEM CLEANLINESS TRT inspections at CPSES indicated that the facility was well maintained.
However, two issues were identified that indicate housekeeping and system cleanliness deficiencies.
A. The TRT reviewed the August 6, 1984, draft of flush procedure FP-55-08.
The purpose of this procedure was to verify the cleanliness of Unit 1 ;
reactor coolant loops, including the reactor vessel, by means of hand- '
wiping, visual inspection, and swipe testing. Tert: t: datermine surface chloride and fluoride contamination were perfomed by TUEC systems test engineers and Westinghouse representatives. The TRT notes, however, that FP-55-08 required only two swipe tests of the reactor vessel-one on the side and one on the bottom. This limited number of swipe tests may not provide adequ' ate assurance that the vessel had been properly cleaned.
B. In rooms 67, 72, and 74 of the Unit 2 Safeguards Building, the TRT observed that not all snubbers were wrapped with protective covering when welding was being done in close proximity to them. This practice 1
was a violation of B&R procedure CP-CPM-14.1, which required protec- t tion of installed equipment during welding. This condition was immediately corrected when the TRT reported it to TUEC QA management, and an inspection was performed by TUEC to correct similar conditions '
in other areas as well.
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10 - NONCONFORMANCE REPORTS (NCRs)
There were several weaknesses in the NCR and deficiency identification reporting systems. The TRT found that:
A. The TUEC procedure for preparation and processing of NCRs did not contain explicit instructions for handling voided NCRs.
B. NCRs were used as a tracking document to record removal of a part from equipment on a permanent equipment transfer rather than for reporting a nonconforming condition; such usage of the NCR was not defined in procedures. ii C. There was an inconsistency between paragraphs 2.1 and 3.2.1 in pro-cedure CP-QP-16.0. Paragraph 2.1 required all site employees to ,
report nonconformances to their. supervisor or to the site QA super-visor, while paragraph 3.2.1 required persons other than QA or QC personnel to submit a draft NCR to the Paper Flow Group.
D. The NCR form had no form number or revision date to indicate that the form was being adequately controlled.
E. There were two versions of the TUEC NCR form, one with and one with-out a space for the Authorized Nuclear Inspection (ANI) review.
F. The NCR form had no space to identify the cause of the nonconformance and the steps taken to prevent its recurrence.
G. The NCR form had no provision for quality assurance review. -
H. The TRT found approximately 40 different forms (other than NCRs) for recording deficiencies. Many of these forms and reports were not considered in trending nonconforming conditions.
11 MTERIALS The as-built review effort by the TRT included a material traceability check on 33 of the same pipe supports that the TRT had field inspected. The material traceability was adequate for those 33 pipe supports, with the exception of four material identification discrepancies, as noted in section 4 on as-built
. inspections.
In another case, TUEC failed to maintain material traceability for safety-related material and numerous hardware components. This QA breakdown was -
identified in an ASME Code survey in Octcber 1981 yet was not reported to the NRC in accordance with the requirements of 10 CFR 50.55(e).
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