ML20238A667
| ML20238A667 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 02/18/1987 |
| From: | Cheng C, Durr J, Wermiel J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I), Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20237K807 | List:
|
| References | |
| NUDOCS 8708210046 | |
| Download: ML20238A667 (68) | |
Text
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DRAFT REV. O CPRRG TASK GROUP 3 REPORT OF SAFETY SIGNIFICANCE i
1 February 1987 l
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8708210046 870312 PDR ADDCK 05000445 0 PDR
UNITED STATES NUCLEAR REGULATORY COMMISSION COMANCHE PEAK REPORT-REVIEW GROUP -
TASK GROUP-3 REPORT DRAFT REY. O FEBRUARY 18, 1987 Task Group Members:
b: n Jacque4P. Durr, Citief
- w n
- , . ,*
~k/N1
Date engineering Branch Division of Reactor; Safety Region I C "f ~ P y ' '
C. Y. Cheng, Section [ Eider" d.-/ (~) / b- ' 7 Date Engineering Branch Division of PWR Licensing-B Jaf)c S. Wem1el, Section Leader-Yllk7 PIEnt, Electrical, Instrumentation and [/yate
-Control Systems Branch Division of PWR Licensing-B L , . . . ., . c ;- - d O g /!~7/N M nnette L. Viett1-Cook, Project Manager Date FWR Project Directorate No. 5 Division of PWR Licensing-A i
i 3l. , Y .
Martin J. Vtrgilio, Chief Cate BWR Assessment Branch Operating Reactors Assessment Staff Task Grouo Leacer:
)
J r W, 7!V7 l 7ennis M. Cru ::nfleia, AssMtent 01 rector Oate Division of PWR LicensingtB
]
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CONTENTS TITLE PAGE-
- 3. EXECUTIVE
SUMMARY
. . . . . . . . . . . . . . . . . . . . . . 1-1 1.1 Background . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.2 Task Group . . . . . . . . . . . . . ... . . . . . . . ..
1-1 1.3 Safety Concerns from_ Other Task Grou 1-1 1.4 Methodology .. . . . . . . . . . . .ps .........
........-. 1-1 1.5 Conclusions and Recommendations . . . . . . . . . . . . . 1-1
- 2. INTRODUCTION ........................ 2-1
- 3. ISSUES ........................... 3-1 3.1 !ssue 1 fror. Inspection Report 85-05 . . . . . . . . . . 3-1 3.2 Issue E from Inspection Report 85-05 . . . . . . . . . . 3-5 3.3 Issue 3 from Irspection Report 85-05 . . . . . . . . . . 3-9 3.4- Issue 4 from Inspection Reports 85-07/05 . . . . . . . . 3-11 3.5 !ssue 5 from Irspection Reports 85-07/05 . . . . . . . . 3-la 3.6 Issue 6 from Inspection Reports 85-07/05 . . . . . . . . 3-16 3.7 Issues 1-9 and 11-1E from Inspection Reoorts 85-14/11 ................... 3-19 3.8 Issue 10 from Inspection Reports 85-14/11 ....... 3-24 3 . E- Issue.16 fror Inspection Reports 85-14/11 ....... 3-26 3.10 Issues 1-5 from Inspection Reports 85-16/13 ...... 3-28 3.11 Issue 6 from Insoection Reports 85-16/13 . . . . . . . . 3-31 3.11 !ssues 7, 9, and 10 from Inspection Reports 85-16/13 . . 3-35 3.13 Issue 11 from Inspection P.eports 85-16/13 ....... 3-38 3.14 Issue 8 from Irspection Reports 85-16/13 . . . . . . . . .2-39 3.15 Issue 12 from Inspec* tion Report 85-16/13 . . . . . . . . 3-45 ATTACHMENTS
SUMMARY
OF *SSLES RAISED IM O!A REPORT 86-10 APPENDIX MM
~6sk Group 3 Charter -
t e
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1.0 CXECUTIVE
SUMMARY
1.1 Background
By memorandum dated January 21, 1987, the NRC's Executive Director for Operations (EDO) charged the Comanche Peak Report Review Group (CPRRG) with the responsibility for (1) determining wilether the currer.t augmented review and inspection effort at Comanche Peak is sufficient to com for any identified weaknesses in Region IV's inspection programs, pensate (2) examining issues relating to the processing and disposition of inspection findings of OIA Report 86-10, and (3) examining the safety significance of issues identified in OIA Report 86-10. In addition, the CPRRG was requested to review NRC Form 766, make recommendations regarding its use, and offer the EDO an opinion on the possibility of broader implications involving Region IV. Task Groups were formed to evaluate these issues. This report documents the results of Task Group 3's examination of the safety signif-icance of the 34 issues identifiec ir OIA Report 86-10. -
1.2 Task Group Task Group 3 was formed on February 2,1987, and ccesists of a senior level NRR manager, a project manager, and senior technical experts in quality assurance, mechanical / structural engineering, auxiliary systems engineering and instrumentation and control systems. Individuals were selected for this effort based on demonstrated skill and ability in either a technical discipline or technical project manecement. The scope and depth in technical expertise of Task Group 3 providec the capability to address, in detail, each of the 34 issues.
1.3 Safety Concerns fren nther Task Groups Task Group 3's charter includeo provisions for incorporating new concerns icertified by either Task Group 1 or Task Group 2. Neitner Task Group 1 nor Task Group 2 identified additional safety concerns for Task Grcup 3 to accress. Accordingly, the scope of the Task Grouc 3 effort did r.ot expanc ceyonc the 34 issues identified in the CIA Report.
2.4 Methodology The Task Group organized the 34 issues by technical discipline and evaluated eacn to determine its safety significance based on tne assumo-tion that the condition was as stated in CIA Report 86-10. A worst-case consequence evaluation was performed assuming that the concern existed without recognition or correction. When possible, related individual issues were grouped together and evaluated as a single larger concern.
Hardware issues that developed from quality assurance concerns were ,
i evaluated to determine the worst-case consequence from both a nareware and a cuality assurance perspective.
1.5 Conclusions and Recommendations I The majori y of tne issues identifiec in :ne 0*A report revolve around cues
- dons regarc;ng crocecures , corrective action systems, accits. trace-abiitty, and documentation. In evaluating the worst case for safety significance, Tast Group 3 postulatec hareware ce#iciencies resul:1rg # rem l 1-1 i
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T se ,
these acmiristrative problems, although actual hardware deficiencies were '
rever identified by the inspectors. In developing recommendations for followup, Task Group 3 lookea at corrective actions broad enough to identify and accress both documentation deficiencies as well as any hardware ' deficiencies that may exist.
Table 1 provides a summary of the conclusions reached by Task Group 3 on the 34 issues evaluated by addressing the safety significance for the worst-case, the adequacy for actions planned by the applicants or taken to date, and recommencatiens for followup. The Table also cross references each of the 34 issues to the section where it is discussed in this report.
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2.0 TNTR000CTf0N By memorandum dated January 21, 1987 to J. G. Davis, V. Stello, NRC's Executive Director for Operations, established the Coesnche P6 n Report Review Group (CPRRG). The purpose of this Group is to review the issues raised by the Office of Inspectors and Auditor (OIA) as a result of an ,
investigation of wrong doing concerning the handling of safety issues by Region IV personnel at Comanche Peak. The CPRRG was assigned the task of reviewing the technical issues identified in OIA Report 86-10 and of determining the following:
Whether the current augmented review and inspection effort at Comanche Peak is sufficient to* compensate for any identified weakness in Region IV's QA inspection programs.
Whether the issues when identified were appropriately handled as to l process and disposition.
The safety significance of the 34 issues identified in OIA Report 86-10.
The purpose and significance of NRC Form 766 and to make appropriate recommendations concerning its use.
Finally, without expanding the specific tasks above, the CPRR was to offer the Executive Director for Operations (EDO) any judgerent on whether it is likely that there are broacer implications in Region IV.
This report assesses the safety significance and adeouacy of corrective ,
actions for the 34 issues icentifiec in Attachment MM to OIA Repert 85-10. I Section 3 of this report provides Task Group 3's evaluations of the issues:
Issues are identifiec by Inspection Report and numbered and restated as they appear in Attachment MM to OIA Report 86-10. Sectier ? incluces 15 separate evaluations covering the 3a issues. When possible, relatec incividual issues were grouped together and evaluated as a single larger concern. The format of the individual evaluations included in Section 3 makes those cases wnere issues were grouped obvious to the reacer.
Issues are explained in the sub-sections of Section 3 entitled " Description."
These " Description" sections elacerate en the concerns raised by the inspectors, reflecting additional information found througn a review cf the inspection reports, testimony and other documentation provided to the Task Group. Background information germane to the concerns is provided in the subsections entitled " Discussion."
For each issue the Task Group performed a worst-safety case assessment assuming that the issue exists as stated without recognition or correction.
The results of these assessments are provided in the subsections entitled
" Safety Significance." When sufficient information was available, the Task l
Group also performed a more realistic assessment of safety-significance.
For eacn issue, tse ~ask Group evaluated the adequacy of the actions taken or plannec by One acclicant to rescive tne concern. Where tnose 2-1 1 1
actions did net, in the Task Group opinion, adequately resolve the concerns, additional actions were recommenced. Recommendations include additional actions for both the applicant and NRC staff. The subsections entitled " Followups Actions and Recommendations" summarize this portien of the Task Group's effort. References for each evalua-tion are provided in the last sub-sections of eacn evaluation.
ACKNOWLEDGEMENT The principal members of the Task Group were supported by several staff members whose contributions permitted the Group to meet its established deadlines. We wish to specifically acknowledge the efforts of Walt Oliu, whc provided philosophical and editorial guidance; Pam Shea, who provided typing and overall administrative support; and Ron Lipinski and Martin Hum, who provided technical support in the areas of structural and mechanical engineering.
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. 3.0 ISSUES 3.1 Issue i from Insoection Recort 85-05:
Failure to translate design criteria frem NSSS vender into installation specifications, procedures and drawings; anc failure to control deviations from the requirements contained in these documents with regard to Unit 2 RPV installation.
Description:
During a routine inspection conducted from April 1,1985 through June 21, 1985 to verify final placement of the Unit 2 reactor pressure vessel (RPV) and internals, NRC inspectors found that neither site-prepared installation drawings nor specifications wnich implemented the Westinghouse ServicesDivision(WNSD)-recommendedprocedureswereavailable.glear The concerns raisec by the inspectors on this issue include lack of. specific installation criteria on centering tolerances, levelness tuierances, tnc clearances between support brackets and support shoes.
The inspectors believed that the traveler used by the constructor, Brown &
Root, was not an adecuate specification on which to base installation of the RcV. They believed the original WNSD installation procedure shuula have been transcribed into a site specification or procedure.
Discussion:
During the construction of a nuclear power plant, the manu*acturer's engineering design cepartment produces cesign output documents that translate the design into specifications, procecures, and drawings that are used to actually construct the plant. These design docur.ents must be controlled and any changes to them must be reviewed ano approved by the criginal manufacturer.
The Unit 2 RPV at Comanche Peak was set in place using a WNSD procecure!N as a guide to develop a cocument referred to as a traveler. The traveler
- s used at Comancne Peak to provide detailec instructions to personnel performing the work. It is intenced to fulfill the NRC requirement to provideproceduresandinstructionsapppriatetothecircumstances,as spec fied in a Browr & Root procecure 5500jO Brown & Root Traveler ME-79-246-contained the essential information to install the RPV and showeo concurrence signatures from the Westinghouse site representative and frert {
the applicant's quality control staff. '
The NRC inspectors reviewed Brown & Root Construction and Operatier Traveler ME-79-248-5500.
- Requirements in WNSD procedures were specified in the traveler. The inspectors found tnat worksheets attacned to the traveler snowed the pressure vessel to be centered and leveled within the established tolerances. However, Traveler Operation 19 recuired verifdca-tion of a 0.002" to 0"-0.005" clearance between the supoort cracket anc l succort snoe. c#ter tne snm plates were applied. Change 5 in ?M tr$veler 1 3-1 l
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- subsequently aproved a change in the allcwec clearance by pennitting i clearances from 0.015 to 0.025 inch. The installation data reflected in j attachment 3B of the traveler incicated an as-built clearance of 0.012 to !
0.026 inch, a space wnich exceeds both the original and reviseo tolerances.
The inspectors were concerned about the acceptability of these revised tolerances. .
The inspectors also believed that the traveler was neither an adequate means to control the installation of the RPV, as required by 10 CFR 50, Appendix B, nor that the system provided for adequate control of changes to that traveler. The inspectors believed that the Construction ' Operation Traveler, which was based on the WNSD-recomended installation criterie, was not equivalent to first translating WNSD criteria into a CPSES speci-fication, drawing, procedure, or instruction, and then issuing a traveler-based on these documents. The inspectors' concern was that without follow-ing this secuence of document development, changes oy site construction persennel to the traveler could be made without a formal desian change being processed. Basically, the inspectors believec that a traveler was not equivalent to a detailed procedure or instruction.
Safety Sientficance:
The safety significance of this issue cepends on whether the (; nit 2 RFV was procerly installed in cccordance with the WNSD-recomenced instalia-tion criteria.
There are two elements to this issue. One involves the Unit 2 RPV install-ation and the other relates to the adequacy of travelers as an effective docurrent to control the quality of installed or erected components. ,
The Unit 2 RPV installation hash;een t
determined to be acceptable by WNSO.
- n a letter to the TU Electric, WNSD states that variations sligntly in excess of the 0.020-inch (cold) clearance requirement have no effect en the design analysis of the RpV support or on the reactor coolant 1000 system. It is not clear wnether kNSD.has based their conclusions or ast exoerience or en a more detailed plant-specifip~ engineering evaluation.
- n response :: an NRC recuest for information, ' the applicants stated that WNSO has recently indicated an evaluation exists for accepting the tolerances; however, the evaluation has not been received at the joosite.
In addition, WNSD has indicated that the final as-built stress rescrts (which will incorporate deviations encountered during the construction .
phase) will be issued after completion cf construction.
In the worst case, a larger-than-specification clearance could : emit excessive movement of reactor coolant loop components during plant operations or during a seismic event. Since the RPV is supportec alter-natively by four hozzles, the unever, clearances between supoort bracxets anc support snoes could create binding in the supports and thus incuce local stresses larger than the design stress in the reactor no::les, supports, or portions of the reactor ecolant system. The worst-case scenario is that this excessive clearance cculo affect the design aralysis (e.g., fatigue lifel o# :ne reacter vessel no::les, supports, and reector coolant loop system. Consequently, the worst-case scenario woulo nave l
3-2 l _ _ _ -- _
safety significance because the design life of the affected components could be reduced. However, since the detailed engineering analysis of the as-treilt condition is not available, the actual safety significance of this change in tolerances cannot be assessed at this time.
With regard to the adequacy of travelers as measures to control activities effecting quality,.the operation traveler is designed to meet quality assurance criteria set forth in licensing documents and applicable codes and standards during the construction of CPSES. The fact that the WMSD vessel installation procedure was incorporated directly into the traveler has no safety significance. The major concern should be that the instruc-tions, no matter what fom they took, were accurate, clear, and correctly transmitted the design output criteria to the field where the work was performed. The controlling procedure states that when implemented as called for, the operation traveler system fully complies with the QA criteria and that no additional controls are required.
In the case of Traveler ME-79-2A8-5500, the tolerances and dimensions were taken directly from the WNSD generic document which recommended that the installation instructions be prepared by the applicants ano reviewed by WNSD site representatives. The concurrence by the WNSD field engineer represents their acceptance of the traveler as an appropriate instruction and/or drawing for installation of the reactor vessel. In addition, the controlling procedure for travelers recuired that all changes, regardless of their significance, must be documented on the traveler, denute the operations affected, describe and identify the authority for making the rectired enange, and include QA representative approval.
Followup Actions and Recommendations:
Without a review of WNSD's engineering evaluation, the adecuacy of the apolicants' corrective action cannot be assesseo. Therefore, the Task Group recommends that a detailed review of WNSD's engineering acceptance of the as-built tolerances be conducted when the report becomes available
' to ensure that the excessive clearance noted dces not have any affect on the design analysis of the reactor vessel supports or reactor coolant loep systen. In addition, a visual inspection should be trecc cf the accessible
~
reactor vessel sur*oundings, especially the reactor vessel noz:les and supports areas, for any sign of distress, construction debrds, or concrete cracking, either during or immediately after hot functional testing.
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References:
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- 1. Westinghouse Electric Corporation Westinghouse Nuclear Service )
Division, " Procedure for Setting of Major N555 Components," dated February 13, 1979. ,
- 2. Brown & Root, Inc. Proceoure MCP-1, Installation of Mechanical Ecuipment and Mo. CP-CPM-6.3, Revisien 6, Preparation. Approval erd i Control of Operation 'ravelers. -
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- 3. Brown & Reet Construction and Operation Traveler, ME-79-2a?-55CO, '
" Reactor Vessel Installation," dated April 10, 1979.
4 Letter from R. S. Howare (Westinghouse) to J. T. Merritt, Jr. (Tl!
Electric), dated January 10, 1986. .
- 5. Letter from W. G. Council (TV Electric) to U. S. Nuclear Regulatory Commission, Attn: Occument Control Desk, dated February 10, 1987.
o 4
4 3.2 Issue 2 from Inscection Pecort 85-05:
Failure to maintain tolerance reovired and failure to report tolerance ceviations en a nonconformance report with regard to Unit 2 RV support brackets and shoes.
Description:
The NRC ip{yectors believed that the control of changes in traveler ME 248-5500 Revision 5,(2jnwas accordance with and not adeouate the traveler control procedure that a nonconformance CP-CPM-6.3, report should have been prepared on the issue (See Section 3.1) of tolerances for Unit 2 reactor pressure vessel (RPV) claarances between support brackets ard support shoes.
Discussion:
The RP'! for Unit 2 was installed using the constructor's (Brown & Root) traveler ME-79-2aE-5500, the pertient pages of which hre appendet te the end of this section. (Additional details can be found in Section 3.1.)
Step 19 of the traveler states, " Verify cold clearance of .020" to 0" .005" for each side shim. See attachment 38 also." Attacteent 3B is a table of the measurements taken during installation of the RDY. Several cf the measurements exceed the allowable tolerance.
The original tolerances specifiec on Operation Traveler ME-79-248-5500, Revision 0, were extracted from Westinghouse Nuclear Services Division's (WNSD) generic procedures. The operation traveler's instal'ation tolerances were subsequently revisec to rc# lect site-specific field conditions.
WNSD's concurrence with the revised tolerarces was documented by signature en the operation traveler. In additiun, Westingbcuse Water Feactors Division proviced a letter, WPT-8148 dated January 10, 1956,'-' acceptine the Unit 2 RPV shim installation tolerances.
With regarc to the cuality cssurance aspects of the issue, the applicar*
recuested and received aporoval for a change to the tolerances from the WNSD. This approval was recordec on the traveler as Revisicn 5, which states, in part, "During the veri #ication of clearances it was revealed that the clearances at 64 (degrees} left and right shims, 153 (degrees) right shim only, 247 (cegrees) rignt shim only, and 338 (degrees) left shim orly dic not meet the tolerance specified in op. *19."
The column on the traveler aojacent to the operation numcer is titled
" Dept." and for operation 19 has the abbreviation "QCV." This means that the quality control staff verifiec the activity. Uncer those circumstances, the QC staff should have acknowledged that the operation deviated from the requirement. This deviation should have resulted in a nonconformance reoort being written against the activity. By way of acknowledging this fact, the aoplicants recogri:ed that en engineering disposition was necerrary and contacteo WNSD to obtain 't (in TU Electric's letter CPDA-4E'.13, as inddcated in the'r rescense o :he Task Grcut 3's recuest for infer atdor.
3-5
Safety Significance:
The safety significance of this issue depends on wnetner the FPV was croperly installed in accordance with (Brown & Root Construction and Operation Traveler ME-79-248-5500," Reactor vessel Installation."!" As ciscussed in .Section 3.1, the worst-case condition may have some safety sigr.i ficance As noted in Section 3.1, the Brown & Root Construction anc Operation Traveler No. ME-79-248-5500 was issued, controlled, and changed in accord-ance with Brown & Root Procedure CP-CPM-6.3. In addition, this traveler was reviewed and accepted by the WNSD site representatives, as were changes to it. Revision 5 to the traveler, steps 19 and 19A, clearly states that if specified clearances cannot be met, then the WNSD site represe'ntatise will determine specified the acceptability of the as-built clearance. In attachment 3B to the traveler, the WNSD representative reviewed and dock-rerted their acceptance of the clearance changes, thus satisfying all steps of the traveler. However, the practice of processing c'eviations outside of.
the nonconformance program subverts the intent of the program. The concor-fomance program performs several functions, including systematic c ccessing of deviaticns, controlled reviews anc disposition by the engineering staf#,
and program trending for mar.agement review.
Followuo Actions and Recommendations:
The Task Group concludes that the followup actions and recommerdstions discussed in Section 3.1 adecuately address the hardware ascects of this issue.
i j' The Task Group re:cmmends that the traveler procedure anc nonconfermance review and approval precedure be examined as to whether acecuate controls are soecified to assu e that noncon#nemances are recorted and procerly pr: cessed. The Task Group further recommends that changes, such as :ncse affectine the RpV, be documented in the plant-specific cerign base: uno maintained'by the applicant or an applicant designee.
References:
- 1. Brown & coot Construction and Operation Traveler, ME-79-Z'8-5500.
" Reactor Vessel Installation," dated April 10, 1979.
- 2. Brown & Root, Inc., Procedure CP-CPM-6,3, Revision 5, " Preparation, Accroval, and Control of Operation Travelers," dated December 13, 1978.
l
- 3. Letter from R. 5. Howard I, Westinghouse) to J. T. Merritt, Jr. UU Electric),
cated January 10, 1986.
I 4
Letter frem 7 Electric to NFC, February 13, 1987.
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CONSTRUCTION OPERATION TRAVELER CONTINUATION TRAVELER NO -
ACTIVITY DESCRIPTION ME-79-24B-5500 9 Reactor Vessel Installation PAG Q g g PREPARED BY-
_ D ATE U D*
REVIEWED BY b '
OATE ~#*W ANI REVIEW *-
DATE O# NO DEPT OPERATION QA/M
, ENO ant 14 M/W Take gauge readings of the spaces on either gcw side of the vessel support pads and record on 7{
Attachment i 3 .
Also recorded cavity, 730*O, temperature.
g C j: ,
15 M/W QCg Machine the side shims to the thickness recorded in Operation 0 14 D,
minus cold clearance. Tolerance is + 0, .020" .001" for 7, 39 temperature + S CT.
gg og 16 M/W QCV Coat be the surfaces in contact with the of vessel the side shimspads support which wilf?)
with "molykote type Z" d.mf film lubricant. yg Buff to improve) finish.
17 M/W Install T.he side shim in the support shoes.
Raise the vessel to a height necessary for top installation of the shims. Bolt the shim keepers onto the support shoes.
la M/W E08E0iWAil03 0s' Lower the vessel and check levelness and vessel Record axis. axis forandorientation with containment obtain approvals on Attachment i 4 .
19 M/W verify cold clearance of .020to 0*
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QA review of operation traveler.T[ TMcV U =
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3.3 Issue 3 from inspection Report 85-05:
Failure to perform ^ audits. or surveillance of reactor pressure vessel specifications, procedures, and installation.
Description:
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Based on their inspection, NRC' inspectors concluded that the failure to perform audits or surveillance of the Unit 2 reactor pressure vessel (RPV)
! installation, specifications, hardware placement, or as-built records, was a violation of. Appendix B, Criterion XVIII, " Audits."
Discussion:
The applicant is required to establish a quality assurance program that will function'during the construction and operation phases of the plant that complies with 10 CFR 50, Appendix B criteria. One requirement of the quality assurance program is that it audit the quality assurance function itself. The cuality assurance program, although designed to assure product Quality, is also a management mechanism for checking the proouction process and for verifying that the process is functioning properly. One methed of gathering the information to make this determination is an audit.
The audit program should be planned to cover those aspects of the quality and production process that are key indicators of the overall process.
This function is accomplished througn the development of an audit pier.
Such audits do not (because they cannot) examine every characteristic of all elements of the process. The selective audit information gathered should then be arewarranted.gviewedbymanagementtodetermineifcorrectiveactions I During a routine inspection, the NRC inspector reouested the applicert's reccrds of CA audits or surveillance for the Unit 2 RPV installation.
The applicant aid net provide any documentation of an aucit or surveillance which evaluatec specifieo placement criteria, placement precedures, harc-ware placement, or as-built rec 0rds.
Safety Significance:
The failure to per#orm an audit of the Unit 2 RFV installation may nct De significant by itself. The issue must be viewed in the context of tne overall audit plan to determine if the plan is comprehensive. The audit is intenceo te sample the cuality assurance program and verify that it is working, whereas a quality control activity is directly related to de+er-r'ining product acceptability.
There is no direct eouioment safety significance resulting from failure to audit the installation of the RPV. The cuality of the installation was monitored by cuality control personnel anc occumented en the traveler.
However, there may be a broacer ccncern i# the audit plan was ceficient or
.anagement was ne: reviewirg the aud4t results arc taking accrocriate corrective action.
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l If the audit plan was deficient, there would be inadequate monitcring cf ?
the overall CA progran, and deficient areas would remain undetected.
Because of this, management wculd not be informed abcut and thus urable to implement needed corrective actions.
This problem has been identified in NUREG-0797 Supplement Ho. 11, Aopendix 0, Section 3.2.11, anc Appendix P, Section 4.7.
The Comanche Peak Technical Reviews Team (TRT) found that the TU Electric audit group consisted of only four auditors during the 1981-1982 period and that TV Electric management was not sufficiently committed to cuality assurance. The TRT cited failure to perform management assessments anc overview of the effectiveness of the quality assurance program. Further, they determined that all aspects of safety-related activities were not audited, and that there was procedural implementation inadequacies, questionable auditor qualifications, incomplete assessments of the CA pregram on 3n annual basis, and inadequate corrective actions to prevert a recurrence of deficiencies.
The applicant: have responded to the issues raised by the [RT anc ethers by form issues.g the Comanche Peak Response Team which will address these Recommendations anc Followue:
The followup cf this issue will be adcressed by the Comanche Peak Response Team (CPRT) as stated in NUREG-0797, Supplement 13, Abstract, "The NRC staff concludes that the CPRT Program Plan provides an overall structure for addressing all existing and any future issues which may be identified from further eva'.uations..."
The CPCT has issued the Comanche Peak Response Team Program Plan ano Issue - Specific Action Plans (ISAP) which contain a section which e addresses cuality of construction and OA/CC adequacy program plansI #!.
The baris of this plan is reinspection of hardware tc estaolish that there are no undetected or uncorrected deficiencies. Further. a section exists vnich spe qualifications.\qfically acdresses applicant audi+ programs ar.c auditor
References:
- 1. Cuality Control Handbeck, J. M. Juran, Third Edition, PP 21-10 to 21-13.
- 2. ' etter from E. H. Johnsen (NRC) to Texas Utilities Electric Company, Attn: W. G. Counsil, catec Feoruary 3,1986.
- 3. Cemanche Peak Response Team Action Plan, ISAP VII.c, Revision 1, "Ccnstructicn Reinspection /0ccurrentation Review Dlan," catec 3 4
January 24, 1986.
Ocmanene Peak Fesconse ~eam Action Plan ISAP V::.a.4, Revis'er 1.
"Acit Program and Auciter Ct.alificatien," dated January N,1986.
3-10
. 3.4 Issue 4 from Inspection pecorts 85-07/05:
For the CVCS spool piece, failure to maintain traceability of item by applicable specification and grade of material'and heat number or heat Cooe.
Description:
Based on his inspection, the NRC inspector cencluded that the failure to mark the chemical volume control system (CVCS) spool piece 3Q1 (DWG No.
BRP-CS-2-RB-76) with material specification and grade, heat number, or heat code, was a violation of.10 CFR 50, Appendix B, Criterion VIII,
" Identification and Control of Materials, Parts, and Components," and American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section III, 1974 Article NA 3766.6.
Discussion:
Bulk piping material is manufacture with identifying markings located intermittently along its length. The markings, specified by the American Standare for Testing of Materials (ASTM), incluoe information about the type of material and the heat number. These markings permit the user to trace the product back to the certified material test report, which at ests to the chemical _ and physical characteristics of the piping. The piping material is manufactured in lengths ranging up to 30 feet. These lengths are cut and assembled into piping systems at the piping fabricators or at the construction site. Relatively short pieces of piping are referreo to as "spcol cieces."
When spool pieces are cut frcm bulk piping in a way that produces a sec. tion without the manufacturer's marking, then the user must transfer the marking er a unicue coce that permits it to be traced bacl' to the certified material test report. These cutting operations are witnessac by cuality control inspectors who can attest to the proper trarsfer of the marking.
Article NA 3766.6 of ASME Ccde Sectier !!!,1974 recuires that the piping tc be identified be marked with the acclicable specification and grade of material, heat number, or heat cede of the material, and any acditional rarking requireo by Section III, to facilitate traceability of the reports of the results of all tests and examinations perforred cn the material. Alternatively, the Code permits a marking symbol or cooe to be used which identifies the piping with the material's certification and requires that such a symbol or code shall be explaired in the certificate.
Se NRC inspector claimed that the spool piece ir cuestion had been marked with the spool piece number (301) anc the Brown & Root (B&P) crawing number but that he could not find the material specification number and type, heat cede, or other means of traceability. If it can be verified that the socci piece numcer previces s unicue icertificatier marking anc that BSD 3 *.1 1
drawing number number (as-built sketch) provides traceability to a tabulat'er of materials which contains all information required.by the Coce, then the speci piece markings have satisfied all the traceability requirements of the Ccce.
Sa'ety Significance:
This issue has no safety significance if the material identification can be positively established and the integrity of the spool piece field welcs can be ascertained; otherwise, this spool piece should be replaced.
With regard to the safety significance of the worst-case scenario, it depends on whether this spool piece is made of austenitic Stainless steel or a carbon steel. If it is carbon steel and the welder dic not identify the mistake during welding, then the issue is safety significant. For this case, the spcol piece will develop a leak or break rather cuickly, either during reoperation testing or early in the plant life, because tne bcrated reactogoolant water is known for aggressively cerroding carbon steel rapidly.
There are three pctential consequences from failure cf the spool piece.
!f it is located in the CVCS safety-related piping system (i.e., the safety injection system) and inside the reactor coclant system (RCS) pressure beundry, its failure would have safety significance. If it is located in the safety-related pertion of the CVCS, but outside the RCS pressure boundry, its failure could result in a total loss cf RCS makeup and safety injection.
If the spool piece is located in ncrsafety-related piping cf the CVCS. the consequences of a break would be minimal.
i Usually, the CVCS is isolated during a LOCA, except for the centrifugal charging cumps and the piping in the safety injection path. Hcwever, this problem can easily be detected by a simple check using a magnet to make sure that the spool piece is not made of carbon steel (austenetic-s ain-less steel is renmagnetic).
l If instead of carbon- steel, the wrong type of austenitic stainless steel p1cirg was usec, these degradations wil' not happen because stainless steel is Pighly resistant to corrosion in the primary ccclant environment.
!n summary, although the worst-case scenario would have safety signif'-
cance, the problem can easily be alieviateo by verifying that a carben steel spool piece was not used in the system. The FSAR, Table 9.3 J indicates that the piece in question is made of austenitic stainless steel.
, r ollowur Actions and Recommendations: ~
There are no corrective actions planned by the applicant with regard to this issue. The Task Group recommends that the traceability of this scoci piece be reverified, if not already core, so as to ensure that the Mgn:
type of steel nas been fitted inte Pe CVCS picing system.
l 3 *2
- Furtherr.cre, it shoulo be verified that an adequate quality control procedure exists for quality control inspectors to witness the transfer of markings on piping that would otherwise lose its trace-ability when cut into smaller sections.
References:
- 1. Letter from E. H. Johnson (NRC) to Texas Utilities Electric Conpany, l Attn: W. G. Coer.511,' dated February 3,1986.
- 2. Comanche Peak Steam Electric Station, Final Safety Analysis Report, Section 9.3.4.1.1.7, page 9.3-33.
- 3. Czajkowski, C. J. NUREG/CR-2827, " Boric Acid Corrosion of Ferritic Reactor Components," July 1982.
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3 '.3 i
3.5 Issue 5 frem Insoection Recorts 85-07/05:
Deferral of hydrostatic test on cold leg test subassembly.
Description:
Basedentheirinspection,NP.Cinspectorsconcludedgataccordingto their interpretation of:the ASME Code,Section III, the cold leg
~of the reactor coolant system should have been hydrostatically testec in the vendor shop prior to shipment to the site. They also maintained that the NPP-1 Form should not have been signed and that the flPT stamo should not have been applied.
Discussion:
Title 10 of the Code of Feoeral Regulations, Section 50.55a, recuires that reactor ecolant pressure boundary systems be designed, fabricated and. installed in accordance with.the American Society of Mechanical Engineers -(ASMEj Eoiler and Pressure Vessel Code,Section III.
Paragraph NB-6221 of ASME Code,Section III, requires that completed components ano appurtenances, with certain exceptions, be subjected to a hydrostatic test at a pressure not less than 1.25 times the. system design pressure prior to installation in the systen. This paragraoh also permits substitution of a system hydrostatic test for a component hydrostatic test provided, if required, that: (1) the component can be repaired by weloing as a result of the system hydrostatic test, (2) the component repair can be postweld' heat treated and nondestructively exarnined subsequent to the system hydrostatic test, and (3) the compon-ent is subjected to a minimum required system hydrostatic test follcwing the comoletiun of repair and examination. In addition, paragraphs NA-1210 anc NA-1232 of-the Code clearly specify that piping subasseeclies are sections of a pipirig system and, therefore, are not cc:rponents.
Since piping subassemblies are net defined as ccmponents, the recuire-ments of NS-6221 and NA-8220, which governs the ap:lication of tne appropriate code symbol only after the hydrostatic test, do not apply to the ccic leg oiping subassembly.
Safety 5ienincance:
This issue was raised because of an incorrect interpretation of the hydrostatic test requirements of tne ASME Code,Section I!I; it there-fcre has no safety significance. The worst case scenario would occur I
if this ciping subassembly was not hydrostatically tested, was subsequ-ently installeo in the piping system, anc curing the system hydrostatic test, ruotured due to overstress. Therefore, the worst case has ec safety significance.
Teilewuo Actions and Recen endations:
The WC inspection progran includes a pr:cecure. IE !rsoection :"cce-dure 7Ca62, " Reactor Coolant Sys em Hydrostatic Test Witnessing," whicn 3-14
(
' recuires the NRC inspector to witness this test. Therefore, it is extremely unlikely the piping would not be tested. In view'of this requirement, recommendations are~ unnecessary.
Reference:
- 1. American Society of Mechanical Engineers, Boiler and Pressure Vessel Code,Section III, Subsections NA and N8,1974 Edition, July 1, 1974.
G 2-15
1 3.6 Issue 6 fecm :nscection Reccets 85-07/05: ? l l
No objective evidence (records) that mixer blades hao been inspected !
quarterly since they were placed in service in 1077.
Description:
During inspections of Comanche Peak Steam Electric Station (CPSES) i conducted from April 1, 1985 through June 21, 1985, NRC inspectors identifiec a lack of records indicating mixershadbeenperiedicallyinspected.gatmixingbladesinconcrete Discussion:
In the FSAR the applicant (ACI) Standard ACI-304-73.ganitted
> to American ACI-304, Section Concrete Institute 4.2, states that
" Mixers shoulc be properly maintained to prevent mortar or dry material leakage and inner mixer surfaces should be kept clean ard worr. blades replaced." It does not define the frequency or the extent of the inspections in order to achieve ' proper maintenance."
The same document, in Section 4.4, further states that "The perfor-mance of mixers is usually determined by a series of uniformity tests made en samples taken from two to three locations within the concrete batch ceing mixed 'or a given time period." It also states that various tests, such as air content, slump, etc., are g d to check mixer performance. B&R Procedure No. 35-1195-CCP-10, Section 3.9.4, states that " Mixer blades shall be replacec when they have lost 10 percent of their original height," and that the " concrete superin-tendent or his representative shall perform a quarterly check to ensure performance." This requirement pertains to both, Section 3.9,
" Central Nixing", and to Section 3.11, " Delivery of Concrete in Truck Mixer tc Point of Placement," of Procedure No. 35-1195-CCP-10 by reference in Section 3.11.6. When the NRC inspector asked for evicence that the blades hao been checked for wear on a quarterly basis as requireo by the procedure, he was told that such records could not be precuced.
Safet'. Significance:
Neither S&R Procedure 35-1195-CCP-1C(3) nor ACI 304-73 reouire documentation for the inspection of mixer blaces. The B&R precedure is more specific cn this sub,4ect and stipulates that checks of the condition of the mixer, including the blades, should be made on a l quarterly basis. It would, therefore, be prudent on the part of the contractor to keep scme kind of records to verify that such maintenance inspections took place.
Insoection of mixer blades is one of the many precautionary measures taken during construction in order to ascertain that the concrete cuality will meet job specifications. The ultimate proof of cencrete adecuacy results ' rem the certressicn tests per#cr ed on concrete cylincer samcles taken from each cour. *
.n the worst case, if tne mixer 3-15
e blades were worn out to the extent that the cuality of the contrete could be affected, it would be evidenced by failure in compression tests and lack of uniformity of the concrete compressive strength.
Failure to meet the job specifications would constitute the " worst-case" scenario. This condition would have some safety significance if concrete below specified strength were used in the containment structure or in concrete Jupports for safety-related components.
Although ACI-304 recommt.nds that mixer blades be maintained, it emphasizes the fact th.t mixer performance is evident in its final product, i.e., the osality and consistency of concrete. Tests for air content, slump, unit weight of air-free mortar, compressive strength, water r.ontent, etc., are the most common ways to check mixer performance. Following the philosophy of the ACI, the Task Group requested that the applicant submit the following additional information:
- 1. Three sample cylinder test records spaced so that one is from the beginning, one in the middle, anc one at the end of the period censidered.
- 2. The retores of the minimum compressive concrete strength for the , period in question.
- 3. A reccrd of statistical distribution cf the concrete test results as far as available for the perioc considered.
By letter dated February 10, 1987, the applicant providet the information to the Task Group consisting of the records of compres-sion tests on concrete cylinders on July 29, 1977, August 19, 1977, February 5, 1950, and May 18, 1984 In all cases the 28-day compressive strength of the concrete exceeded the specifiec cesign strength of 400C :si.
In the same letter the applicant provided urifomity test resuits cated July 9, 1976. The results for weight per cubic foot, air content ',velume cercent of concrete), coarse aggregate certent, arc the average compressive strength at seven days, are all within the maximum cemissible differences, thus complying with project reovire-ments. During a telephone conference with the applicant's represen-tatives on February 11, 1987, che T M C oup was informed that the preparation of information related to tne c> efficient of variation of concrete mix requires more time anc will be provided in the future.
Followuc Actions and Recommendations:
It is not entirely possible to assess the cuality of concrete used at the plant with ecmplete confidence on the basis of the few test sar;'les reviewec by tne Task Group. This matter shculd be investi-gated in more cetail by reviewing test results available at tne site.
3-17 I l
The statistical cata regarding uniformity of concrete, when submitted by the applicant, should be reviewed by the Task Group in order to gain additicral confidence with regard to the quality of concrete used at the plant. Examination of this information, together with the evider.ce that periodic compression tests of concrete samples met the job specifications, will constitute sufficient evidence that the mixer blades did or did net affect quality of cencrete, and, therefore, safety of the plant.
References:
- 1. Letter from E. H. Johnson (NRC) to W. G. Counsil (TUEC), dateo February 3,'1986, Appendix B, NRC Inspection Report 50 445/85-07 and 50-446/85-05.
- 2. ACI Standard " Recommended Practice for Measuring, Mixing, Transporting, and Placing Concrete," ACI-304-73.
- 3. Brown & Root Procedure No. 35-1195-CCP-10, "Corcrete Batch Plant Operaticns," Rev. 5, dateo December 4, 1978. -
4 Letter from W. G. Counsi' (TUEC) to U.S. Nuclear Regulatory Commission, dated February 10, 1987.
- bII _ _ - _ - - _ _ _ _ _ _ _ _ _ . - - - _ - - - - - - - - - - - - - - _ - - - - -
3.7 Issues 1-9 and 11-15 of Inspection Report 85-14/11: i Issues 1 through 9 and 11 through 15 all relate to storage, trans-eission and handling of quality records. These issues were dealt {
l-i with collectively because the consequences of worst-case events have a common result.
- 1. FSAR 17.1.17 does not describe TV Electric records system.
- 2. 0A manual does not address ANSI- N 45.2.9 requirements / i commitments.
- 3. TV Electric faileo to have/use procedures to control shipment of original design recoros for piping to Stone 8 Webster, NY.
4 Original design records shipped in cardboard boxes to Stone &
l Webster.
'i. No backup copy of records made for records shipped to Ste'ne &
Webster.
I
- 6. Failure to contro' cnd account for cesign records transferred from 5 4:e to Stone & Webster, NY. TU Electric stated design record shipped without making backup copy because cost too much. Also stated it was comoany policy to proceed at own risk.
- 7. Site records of Chicago Bridge & Iron shipped to Houston, Texas in cardboard boxes.
- 8. No backup copy'of records made for records shipped t Chicago Bridge & Iron.
- 9. TU Electric failed to inventory records sert to Chicago Bridge &
Iron.
- 11. Failure to preclude rain from entering OA interim recoros vault ever several years time.
- 12. Failure to preclude food and coffee pot from OA interim recora vaults (fire hazard).
- 13. Fr.ilure to install fire suppression systems, drains, and a slopea floor at permanent vault.
14 Plant recercs stand in ' ciders or bincers in oper face cabinets at records center.
- 15. Failure to provide temporary or permanent storage facility for '
records entered into the permanent records center then co-mingled with in-process dccuments in paper flow grcup.
3-19
_ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ - - _ .------.---E
Cescriotion:
- 1. The Final Safety Analysis Report (FSAR), Chapter 17.1.17, does '
not adecuately describe the quality records system for Comanche Peak construction activities. The FSAR does not reflect the level of detail prescribec by the American National Standards Institute (ANSI) N 45.2.9. (1,2)
- 2. The Texas Utilities Electric Company (TU Electric) Corporate Quality Assurance,(QA) Program Manual and QA Plan de not address ANSI N 45.2.9 in all aspects. (1,2)
- 2. TU Electric shipped original quality records to Stone & Webster, Inc. without an implementing a procedure'to control the process.
(1,2) 4 Original design records were shipped to $ tone & 'r.'ebster, Inc., in carcboard boxes. These boxes did not afford the proper protection specified fer records cf this type. (1)
- 5. The records shipped to Stone & Webster did not have duplicate copies retained onsite. (1)
- 6. The records sent to Stone & Webster were not inventoried nor dccumented such that accountability could be maintained.
- 7. Original quality records were shipped to Chicago Bridge & Iron (CB&I) in Houston, Texas in cardcoard boxes. This is sitrilar to item 4 abcve. (1)
- 8. The recoros shipped to C3&I die not have duolicate cecies retained cnsite. This is similar to item 5 above. (1)
- 9. The records shipped to C5&I were net inventoried nor documerted sucn that accountability coule ce maintained. This is sdm4 ar
- item 6 above. (1)
- 11. There was e.iderce of rain leaking through a ventilation cuct in the interim record vault creating a hazard for stored records. (1,2)
- 12. Food stuffs and a cof#ee pot were found in the interim reccrc vault creating a potential rodent and fire hazard. (1,2)
- 13. There is no installed fire suppression system in the permanent records storage vault. In adoition, there is a 2-inch fire hose available to fight fires but there are no floor drains and the floor is not sloped to provide runoff. This cendition creatas a potentici for addeo water camage to the recorcs sbculd the fire hose ce used. (1,2) 2-20 l
t .___ _ - -__ _ _ _ _ _ . .
4 e
14 The TV Electric records center, one of two pemanent records storage areas, has records stored in folders and binders in openfaced cabinets. The records center is protected by a deluge water sprinkler ~ system for fire. In the event the 4
sprinkler is actuated, the records will be damaged or destroyed.
(1,2)
- 15. Pennanent plant records were withdrawn from record vaults to facilitate work flow. These records were stored temporarily in the Paper Flow Group trailers in fire-rated and nonfire-rated '
cabinets. (1,2)
Discussion:
Title 10 CFR 50.71 and Appendix B require that records be generated and maintained to confirm that certain activities have been satisfact-crily accomplished. These records pertain to plant activities, such as quality control and quality assurance functions, to plant design processes,'such as design repor.ts or design verifications, and to the results of ecuipment performance tests, such as preoperational and plant startup testing. These records provide proof that the facility was designed and constructed.as described in the FSAR and document the basis for. licensing the plant. Some records are designated es " life-time or pemanent" records while others will be retired after construc-tion is completed and plant operation begins. The permanent records are_ maintained to support safe plant operation, plant trodificaticr.s, accident analysis, and in-service examination.
The foregoing issues all deal with the proper storage and retentior, of thase kinds of records. The applicant has committed to store records in accordance with ANSI N 45.2.9, " Requirements for Cullection, Storage ano Maintenance of Quality Assurance Records for Nuclear Pcwer Plants." The jurisdiction of this standarc begins when a file is completed and is cesignated a cuality record.
Safety Significance:
The. " worst case" result of any of the above issues is that important design, construction and operations records would be lost, destroyed or missing. Most recorcs can be recreated by reinspection of acces-sible ecuipment, reconstruction of design documents, or replacement of the record or ecuipment. The condition of inaccessible equipment can also be determined by engineering analysis or more rigorous testing. .
The near-tem impact of missing records is a delay in completing occument packages for system turnover to establish operational readiness for plant preoperational and startup testing. The long- i tem effects coulc be that operating plant activities such as piant i modifications, in-service examinations and testing, event analysis. l and systers analysis would be hampered by incomplete irfermation. -
{
3-21 <
1
l Followuo Actions and Recommendation: ?
1
!ssues 4 through 6 and 12 were not transmitted as violations to the applicant, anc do not require a fermal response for corrective actions. Item 2, transmission of records without a procedure, was "esponded to by the applicant.
The response indicated that a procedure was developed, records were retrieved and copied, and personnel were trained to the procedure.
On February 13, 1987, the applicants, in responding to issues 1, 2, 7, 8, 9, 13, 14, and 15, provided the following additional information:
Issue 1 - The FSAR does not describe the TV Electric records system.
The applicants have concluded that._their procedures oo not adequately describe the final TU Electric review and the conversion of documents to records. TU Electric is drafting new procedures to correct this {
deficiency.
Issue 2 - CA manual does not address ANSI N d5.2.9 requirements. The applicant believes the QA manuals and procedures are adeouate.
Issues 7, 8 and 9 - Demonstrate that the CB&I records controls were implemented. The applicant essentially believes the CESI proceoure
- was acequate and correctly implemented. No further action is warranted.
Issue 11 - Failure to preclude rain frem entering QA interim records vault. The water leakage was corrected.
Issue 13 - Failure to install a fire suppression system, drains, and a slopec floor in the permanent records vault. TV Electric's positior. 's that a fire suppression system is neither needed nor reouired. Their survey of the floor has demonstrated that it is sloped.
- ssue 14 - Plant records stand in folders or binders in open faced cabinets at records cen er. TU Electric is installing an ala m to cetect operation cf tre sprinkler system arc alert personrel to the I
cotertial for ficccing. They consioer records storace in open faced cabinets is accettable.
Issue 15 - Permanent plant records were withdrawn from record vaults j
te fccilitate work flow. These recoros were stored temporarily in tne Paper Flow Group trailers in fire-rateo and nonfire-rated cabinets.
l The applicants maintain that documents in the Paper Flow Grcup are in-process er,c not sucject to ANSI N 45.2.9. Mcwever, TV Electric has placed the documents in fire-rated cabinets.
The Task Group recommends that the NRC staff review the revised Il; Electric NEO procedures upon their ectrpieticn anc verify the implemer-1 tation. Pending Task Group 2 reccemendations relating tc apolicacie
! an#cccement, t'.e NRC staf# snould veri #y correct i ve actiers ;rc;csec I by the acoldcants.
! 3-22 l
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l
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\
1
References:
l.
DIA Report of Investigation 86-10, Allegation of Misconduct by Region IV Management with Respect to the Comanche Peak Steart Electric Station. I
- 2. NRC Inspection Report 50 445/85-14 and 50-446/85-11.
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1 3-23 l
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3.8 Issue 10 of :nsoection P.eoorts 85-14/11:
TU Electric audited CBI Houston and in the scope of the report stated it included Criterion XVI, QA records but did not document the audit of records Violation Criterion XVI!!. * '
Oescriotion: .
Texas Utilities Electric Company (TUEC) documented an audit of the quality assurance (QA) program of the Chicago Bridge & Iron, Inc.
(CB&I), the contractor responsible for constructing the Unit No. 2 containment building liner, and failed to document the details of the record's portion of scope of the audit (ge audit, although thic pertion was listeo in the Discussion:
TUGC0 is required by 10 CFR 50, Appendix A, Criterion I, and Appendix E, te establish a quality assurance. program. One requirement of tne cuali+r assurance program is that TUEC.be responsible for auditing the work of its contractors, including the contractor's QA recoros. The major purposes of such an audit are to provide management with information regarding the effectiveness of the production process, to identify non-c:cfoming er deficient items, and to monitor the CA program itself.
Based on the audit finoings, management evaluates such conditions and initiates corrective actions where warranted.
Safety Significance:
There is no direct ecuipment safety significance for net per#oming an aucit. The worst case outccme is a programmatic breakdcwn of be audit crocess.
The issue is that the applicant failed to docurent one element of an audit, quality assurance recorcs. The central question then beceres, was tnis audit perfomeo?
The listecaudit in the record establishes scope that(poits were perfomed of other arut cf the audit. If the audit of 0A recoros was not performed, then at a minimum, the result c:uld be an inadecuate recorc's pregram. For a detailec discussion of the safety significance relating to audits, see Section 3.3 of this report, are for the record's program aspects, see Section 3.7.
!f the audit of records was cerfomed but not documented, then the effect on construction of the liner is of no conse-quence.
I 3-24
.-__-7_
Recommendations and Followup Actions:
See the Recommendations anc Followup Actions in Section 3.3 of this report.
References:
- 1. Texas utilities Generating Company letter to Chicago Bridge & s Iron, Inc., dated May 7, 1985, 0XX-2381; subject: TUGC0 QA.
Audit Report, QA Audit File: TCB-6
- 2. Draft Report 50-445/85-14 and 50 446/58-11, undated, CPRRG-17.
O b
3-25 1
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- - - - - - - J
i 0.9 Issue 16 from Inscect'cr Pecorts 85-14/11:
l Weld red not identified at main distribution station (i.e., labels taker cff or lost on E-309 electrodes at the main distribution station).
l Descr10 tion:
Based on inspections conducted, NRC inspectors concluded that there was l
I a failure to maintain the material classification, size, and heat lot number markings on several containers of Sandvick welding rods located in the main storage areas and that this failure was contrary to par I graph 3.2.1ofBrown& Root (B&R)ProcedureCP-CPM-6.9,AppendixB,g l and was also a violation of 10 CFR 50, Appendix B, Criterion V.
1 Discussion:
Weld rods used in safety-relatec applications are nonnally stored in the main distribution station anc then transferred in marked shipping cartons to three rod issue stations: red houses 2, 3, and 4. Leose labels 1
cbserved in the main distribution station were also observed in rod house 4 curing a routine NRC inspection. Even so, the inspector noted that :he material was identifiable because of marking on the storage bin anc shipping cartons. However, loss of identify was possiole wrer tne rods were removec from storage. Rods are issued for use cre lot at a time.
They are taken frcm their cartons only when needed to replenish the rtock in the drying oven or when issued for use.
The remaincer of a lot ret issued is put in the oven. (A heated statien-ary or energized portable oven was used in the fiel'd to;4eep coatea electrodes dry.) 887 Procedure C?-CPM-6.0, Appendix B' ' has requirements
'cr control'ir.c the identification cf weld filler material that is removed frem its trig 1nal container.
Saity Signi'icarce:
The ccatec welc rces in cuestien (E309 ccated electredes) are usually used f:r wela1ng dissimilar metals (e.g., carcer steel to stainless steel). E3C8 coa ec 'e ectrodes, en the other hand, are emoioyeo in i
welcing stainless steel to stainless steel.
in color, thickness of coating, and length.(2'othThey electroces areother differ frer' similcr ccmmon coatec electroces en site, such as E7018 or E8018 (which are usec to join car:cn steel to carbon steel) in colce, thickness of coating. ~
anc length. Weiders wouic recognize the difference immediately upcn striking an arc if E7018 or ESC 18 was substituted for E309 because the arc characteristics are so dif#erent. However, weiders may nave difficulty in cistinguisning tne unmarxeo E3C8 electreces from E C9 electrodes.
In the worst case scenario, a welcer coulo use an E7018 or E801E carbon steel electrode in a stainless steel system. Assuming that the weic ceccsit was adecuate anc :nat the weld cassec :ne required visual, nondestructive, anc hydrostatic testing, tne weld cev'c be cut inte service in the reactor coolant system. The careen steel wouic c:rroce 3-25 I
I I
- , veryqujykiybecauseof.theaggressivecorrosiveeffect!cfborated waters k . The corrosion would ultimately lead to a weld failure and tc a loss of coolant accident if the weld was in the reactor coolant l- pressure bouncary. If an E308 electrode is substituted for an E309
) electrode, the finished weld practically has the same quality as that made from_an E309 electrode; therefore, such a substitution wculd have no safety significance, i
Fellowup Actions and Recommendations: . .
The Task Group recommends that the NRC staff. determine if individual electrodes are identified individually according to AWS or MIL-E-22200 recommendations with a " Type Mark," such as "309". If each electrode is uniquely identified, the probability for misuse becomes very low, especially when coupled with the arc characteristics anc physical - - -
) appearance of these electrodes. If the electrodes are not uniquely identified, then the applicant should undertake a statistically valid samoling of stainless steel welds to determine whether and to what degree misidentification and misapplication occurrea.
References:
- 1. Brown & Roet, Inc., Procedure CP-CPM-6.98, Revision 2, " Weld Filler Material Control," dated September 21, 1984 '
- 2. NUREG-0797, Supplement 10, Safety Evaluation Report, pg. N-79. April 1985.
- 3. C:ajkowski, C. J., NUREG/CR-2827, " Boric Acid Corrosion of Ferritic Reactor Compenents," July 1982.
3-27
3.'.0 !ssues 1 throuch 5 ' rom Insoection Recorts 85-16/13:
. 1. Failure to develop / implement procedure to demonstrate 50.55(e) deficiencies corrected.
- 2. Failure to revise implementing procedures containing 50.55(e) reporting. TU Electric failed to revise implementing procedures before corporate NEO Precedure CS-1 was implemented resulting in conflict with five other procedures.
~~3. Failure to maintain retrievable 50.55(e) files (i.e., could not produce record in almost a month).
4 Failure to report to NRC actual corrective action taken on 50.55(e)s.
- 5. IV Electric's SC.55(e) files not auditable.
Cescriotion: -
1
- 1. TV Electric's procedure to process Construction Deficiency Reports (CDR) failed to recuire file information which would give evidence of issue Closure.
- 2. TV Electric failed to revise subtier implementing procecures before l corporate NEO Procedure CS-1 was issued, resulting in conflict with five other procedures.
- 2. TU Electric failed to maintain COR files that were retrievable.
'. TU Electric failed to report to NRC the corrective actions actusl'y tak.er. anc cnanges to commitments.
The reporting requirements under 10 CFR 50.55(e), Construction Ceficiency Reports (CCR), were instituted to previde NRC with promot notification cf significant construct 1on ceficiercies. They are to give NPC timely ir.tcr-mation on which to base an evaluation of the potential safety consequences of the ceficiency and determine if further regulatory action is recuirec(*}.
CORs are normally identi#ied by the applicerts' cuality assurance prograr, tnrough nonconformance recorts, design deficiency reports, venoor 10 CF?
I' reports. or other similar systems.
3-25 j l
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i Any breakdown in the CDR reporting and tracking system would affect the !
notification, evaluation and final clesure of construction deficiencies !
l as it relates to NRC. NRC requires t g selected construction de#iciency 1 l reports be closed through inspections . If detailed tracking files are not maintainec, closure becomes more difficult; however, the primary '
corrective action tracking document for the identifieo deficiency wculo be the original quality assurance report.
l The procedure identified in item 1 does not require certain information t to be retained in the applicant's tracking file which would permit the !
inspector to readily detemine if the item had been properly closed. l This makes the file unauditable for the inspector unless there are cross references to the corrective actions programs.
The failure to revise subtier procedures, item 2, results in nor. uni-fomity in the processing of CDRs, but does not necessarily affect reporting tc the NRC. This also affects the relationship between the l NRC and the applicant, and internal processing within the applicant's organization.
The failure to maintain CDR files that were retrievable, item 3, stems j
- rom the inspector's inability to cross reference between the CDR files and corrective action program files. This is similar to items 1 and 5 in that files that are not retrievable are also not auditable.
The failure to report corrective actions actually taken and any changes to coranitments, item 4, directly affects NRC/ applicant communications.
Not receiving this information affects NRC's ability to perform a mean-
, ingful evaluation and reach any decisiens to take further regulatory action.
Safety Significance:
The issues at TU Electric that were identified by the inspector all relate to reporting procecures between the NRC and the applicant.
Based on the fact that CPRs were written and filed, there is re indica-tion in the inspector's report that the identification mechanism for CDRs was deficiert. Therefore, the Task Groue assumes that the sour:es of input to the process were functioning satisfactorily. Under these circumstances deficient eouipment or controlling systems were apparently being corrected through other established mechanisms, such as the non-conformance corrective action process prescribed by 10 CFR 50, Appenoix S, Criterion XVI, Corrective Actions. Thus, if these reporting mechanisms are functioning, there is no safety significance relative to the plant equipment. ,
Followuo Actions and Recommendations:
The apolicant's response cf February 13, 1987, states that CCR activi-ie:
are currently controlled by TU Electric Procedure NEO CS-1, " Evaluation of and Reporting :# Items / Events unoer 10 CP 21 and 10 CFR 50.55:e,' "
TU Electric has established e licensing ccanitmert resolutier, precess to 3-29
track the timely ccmpletion of ccrnitments made to the NPC. Also, there is a task force chartered to identify, validate anc assure pcsitive c'iosure of CCPs.
It was notec by the Task Group that TV Electric Precedure NEO CS-1,
" Evaluation of ano Reporting of Items / Events Under 10 CFR 21 arc 10 CFR 50.55(e)," does not specify that all items reported under the proceaure should be first recorded in the established cerrective action systems.
The procedure states that inputs can be received from any source. Where the source is other than an established ecality tracking system, it is possible that a reported deficiency would not be properly processed under a formal co rective action system.
The Task Group recentnends that an audit be performed by the NRC staff of the current CDR program to verify its acequacy and implementation.
References:
- 1. 10 CFR 50:55(e) Statement of Considerations, 37 FR 5459.
- 2. :nspectior and Enforcement Manual, Inspection Procedure 927CO, S/13/Sa
- 2. Inspection anc' Enforcement Manual, Interpretations 10 CFR 50.55(e),
4/1/80 i
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3-30
, 3.11 Issue 6 from Inspection Report 85-16/13:
TV Electric never responded to cli aspects of IEB 79-14 Description During several NRC inspections of Comanche Peak Steam Electric Station (CPSES) from November 1 through 30, 1985, inspectors focused on tne applicant's response to IE Bulletin (IEB) 79-14. " Seismic Analysis for As-Built Safety-Related Piping System," was reviewed. Specifically, the issues in IEB 79-14 was evaivated and closed for nonconformances because paragraph 4 was not satis- t by TV Electric in 1983. The NRC inspectors indicated that the c1;.,ure was premature since Stone &
Webster was then analyzing Unit 1 seismic analysis against as-built drawings, an issue cirectly'related to this Bulletin.. In addition, the inspectors found that the same analysis had been completed for Unit 2.
TV Electric stated in response that the IEB 79-14 file would be reopened and a supplemental report would be submitted upon completion rf the ongoing erginee-ing werk relevant to this Eclietin. The status of IEE 79-14 statu 146/8513-0-03).gsstillconsideredopenbyNRC(M5/8516-0-03, Dis cus s _icn:
- This issue cannot(be, resolved by comparing !EB 79-14 with tne applicart's formal responses *
Engineering Company. Q and with Actions taken recent activities by the applicaritbyand Stone & Webster the staff te acdress this issue must be considered in the context of Commission practice with regard to evaluation of all responses by utilities to the Bulletin. The discussions in 01A Report 66-10 the applicant's "AgBuilt Verification Program"gnot and Comanche indicate Peakwhether Prd ect Procedures / change the status of this issue.
NRC issued IES 79-14, " Seismic Anclyses For As-Built Safety-Related Piping System:" on July 2, 1979 and Revision 1 on July 18, 1979. Supole-ments 1 and 2 :o the Eulletir were issuec or August 15 and September 7, 1979, respectively. Since TV Electric responses were nct based on the original Bulletin, the pertinent decument is Revision 1.
IEB 79-14 reouested that all power reactor facility licensees and con-struction permit holders verify, unless previously verified to an ecuivalent degree within the last 12 months, that their seismic analysis applies to the actual configuration of safety-related piping systems. The Bulletin was issued because inspections of safety-related piping systems adoressed in IEE 79-02, 79-04, and 79-07, and snow-cause orders for four nuclear power facilities, revealed some as-built
. deviations from design documents used for input to seismic design analyses. These deviations were significant enouoh to have an adverse effect on :he validity cf the seismic analyses. The Bulletin recuirec licensees of cperau ng facilities te perform walk-down inspections of safety-related pipirg systems, make compaciens to seismic analys t 1
3-31 \
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4 I input, identify nonconformances and evaluate their effect on system ,
operability, and either make hardware modifications er reanaly:e the '
as-built configuration to validate their seismic analyses. Polders o' construction permits were required to inspect and report on safety-relatec piping systens for compatibility cf seismic analyses with as-built configurations.
The _ Bulletin specified that all inspections were to be completed and the results reported to NRC within 120 cays. IE provided specific written cuidance to the Regions for immediately implementing IE inspections crd for reviewg applicant followup actions and written responses to IEB 79-14 Although these instructions focused on operating j facilities, the Bulletin was equally applicable to plants under construc- ;
tion. As the implementation of this bulletin progressed, a substantial i numberofquestipsysyereraisedbyIEinspectors, representatives.
g licensees,andindustry In response, NRC issued two supplements to the Bulletin.
Licensec reports were initially evaluated using a task groue ap roach involvinf6pgJsonnel from IE headquarters, NRR, and the Regional offices A contractor was subsequently retained to assist the 1 staH in reviewing responses to the Bulletin. As of June 1955, the !
bulletingasclosedoutforonly48cfthe124facilitiesunder i review.' >
The applicant's initial response (October 25,1979) stated that -hey were finalizing a precedure to adoress the Bulletin and recuested a waiver from the 120 day reportieg requirement because construction was )
not yet sufficiently complete to support a system inspection program.'2 -!
The apolicant subsequently provided (December 3, 1982) a formal response to IES 79 i program."g that Thedefined the scope safety class of the si:e, and "As-Built type of pipingVerification systems included in the program were defined. The applicant specifically stated that dccumentat1on in the form of picing and support construction orawings and support locatien iscmet cs would be field verified bv site CA personnel (emchasis acceo).p"' The apolicant implemented inis progrsm-tnrcugnout.
Bulletin."}he construction of Units 1.and 2 in response to the The acplicant also ccmissioned Stcne & Webster Engineering Corporation to perform a stress recualificatien of cooe piping and
! piping supports.
Safety Significance: '
The purpose of IEB 79-14 was to ensure that the seismic analysis input i
l information anrees with actual construction details at this fccility.
The Sulletin reouires that specific nonconformances be resolved by either making changes to the system, such that it conforms to the j design, or by correcting the seismic analysis to demonstrate the acceotatility of the as-built ecnfiguration criteria.
The issue frem OIA ?ecer 85-10 is that the acclicant never respended to all aspects of IES 79-14.
( 3-32 l
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)
' In the worst-case scenario CPSES receives its cperating license with I as-built pipe support systems that are not consistent with the plant's j seismic analysis. Under the.e circumstances, the safety significance of the worst-case scenario could be enormcus in that the plant might not be able to achieve a safe shutdown condition following a seismic event.
Based on the limited infomation available, the Task Group reached the following conclusions.
(1) When IEB 79-14 was issued on July 2, 1979, the design and construc- )
tion of CPSES was not sufficiently advanced to make a detailed response to the Bulletin meaningful.
(2) The program outlined by letter TXX-3597I3) may not have addressed all aspects of the bulletin. However, it dio explain the applicant's intentions. The applicant implemented and f C owed the "As-Built Piping Verification Program" (Instrue with the expressed purpose of addressing IEB 79-14 gn CP-EI-4.5-1)
'3' Etcpening the issue of IEB 79-14 in Inspection Peport 50 42 UPE-16; 50-446/85-13 was appropriate considering the majcr analyticel activities and plant modifications at CPSES, (4) The applicant has on-goi
.foentifiedinIEB79-14pg,)programstoaddresstheconcerns (E) The safety significance of the worst case could be enomous.
(6) At this time, there is no safety-significant issue since IEB 79-14 will be closed before licensing. l l
ollowur Actions anc Recommendations:
Eased on -he historical record of IE9 79-14, the actuab regulatory requirements and acceptance criteria are not obvicus.' The Task Group ciscussec this issue with a member o' the Comanche Peak Technical i Review Team, who concluded that TU Electric has essentially fulfilled l their commitment te IEB 79-14 anc considers the issue closed. However, based on a Task Group review of the records, the Task Group finds that l this issue has yet to be fully resolvec anc thus that this issue is stC'. '
open. The Task Group recommends that the NRC staff provice the applicant with a clear and concise written evaluation of the applicants' actions taken to date and specify additional actions required to close this issue. This written position should be consistent with the Commission policy that was apolied to other licensees and applicants that success-fully comp'eted the requirements of the Bulle-in.
References:
.. Letter to Texas Uti'ities Generating Comcany fron E. H. Johnsen WN.
cateo April 4, 1986.
- 2. Letter to NRC Rep en IV from R. w'. Gary JUEC), cated October 25, 19 9 JTXX-3062),
3-33
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4
. Letter to NRC-Region IV from W. G. Counsil (TVEC), dateo April 3, 1986(TXX-4729).
- 5. Letter to NRR frcm W. G. Counsil (by J. W. Beck, TUEC), dated
' September 19,1986(TXX-5034).
- 6. Memorandum to Regional offices from S. E. Bryan (NRC), dated August 6, 1979,
Subject:
TI 2515/29 - Inspection Requirements For IE Bulletin 79-14.
- 7. Memorandum to Regional ' office from N. C. Moseley (NRC), dated September 6.1979.
Subject:
Supplement 2 to IE Bulletin 79-14
- 8. Mcmcrandum to N. C. Moseley (NRC) to D. G. Eisenhut (NRC), dated August 28, 1979,
Subject:
Recommendation Concerning Inspection -
anc Implementation Requirements of IE Bulletin 79-14
- 9. Memorandum to D. G. Eisenhut .(NRC) to E. L. Jordan (NRC), dated September 11, 1979,
Subject:
Evaluation of Respnnses to IE Belletin 79-14
- 10. Letter to R. L. Baer (NRC) from R. A. Lofy (Parameter, Inc.),' dated June 28, 1985, related te HRC Contract 05-82-249.
- 11. Memorandum to Regional offices f rom E. L. Jordan, dateo rebruary 7 1986,
Subject:
IE Sulletin 79-02 and 79-14 Status Reports.
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l 3-34 l
3.12 Issues 7, 9 and 10 of Inspection Reports 85-16/13:
I
- 7. TV Electric's IEB record files were incomplete (1982 and 1985).
. 9. Deficiency in TU Electric's procedures to handle IEBs. They do not describe how construction management personnel handle IES requiring action especially hardware repair, replacement and modification.
l
- 10. ho focal point at TU Electric to track IEB actions. I
Description:
- 7. TV Electric Inspection and Enforcement Bulletin (IEB) record files ~~
were incomplete. The bulletin files were decentralized rather than 1ccated in the CA records center. Further, the engineering evaluations were retained by the incividual engineers rather i
than being centrally filed.
- 9. There were deficiencies in the proct: dure to process IEBs. (Note:
The record is not clear as to what the inspector meant by deficient.
The Task Group assumed that the deficiencies resulted from tne lack !
of a central coordination position for :EBs ano the perceived file deficiencies.)
- 10. No central coordination function at TU Electric to track IES actions.
Discussion: .
The NRC issues Inspection and Enforcement Bulletins {IEB) to cperating reactor f acilities and those under construction to transmit information or to recuest action er information regarding matters of sefety, safeguards or environmental significance.(1) The safety information transmitted may identify generic equipment or design deficiencies.
Licensees and applicants are expected to oetermine the applicability of the Bulletin to their plant and initiate appropriate corrective actions. IE5s usually recuire licensees and applicants to respono with such information as applicability to the plant, equipment affected, operability status of systems, corrective actions initiated, and schedules for completion. i l
The NRC inspection program requires that licensee and applicant implementing programs for IEB-requested actions be inspected to assure that appropriate actions have been taken.(2) The program reouires that all cocuments in the response to NRC be reviewed anc a determination made that the response was proper. It also requires l that onsite sampling inspections be made to verify that equipment changes were mace as re;crted.
3-35
)
Basec on the range of issues they encompass IE3s can be a mechanism for identifying nonconforming conditions. Consequently, licensees and applicants must make provisions for receiving, evaluating, initiating corrective actions, verifying corrective actions, arc reperting resuits to their c.anagement anc to NRC.
The Companche Peak applicant has a procedure for processing correspen-dence. inclucing IEB's, from NRC; Nuclear Operations Engineering Manual, Licensing, Procedure No. N0E-205. This procedure assigns responsibilities and discusses the processing and the retention of records.
Safety Significance:
Assuming the worst case to be the breakdown in the processing of IEBs. the impact on safety would be significPet. If the NRC issuee a Bulletin that affected the facility eouipment or operating procedures and it was not incorporated into the plant, the facility -
could cperate with an inherent undetectec/ uncorrected defect.
A Sulletin, by definition, is only issued when safety concerns nave been identi'fied and when NRC believes there is a threat to the public safety. Eulletins can affect eperating ano construction functions and a thorough and comprehensive controi program must be established to evaluate, track anc resolve Bulletin issues.
Followu: Actions and Recommendations:
Apolicant Procedure NOE-2C5, dated October 7, 1985, generally describes
-he process for controlling Bulletins through their receipt,1cgging in, "eview, plan development, response, and closecut. Paragraph 4.2.15 cf the procedure specifies that documents which provide source informatier
'or the Bulletin resocese 3e included in the decument package. If the
- rocedure is 1mplemented, it sr.culd provide un adeouate system for
- rocessing Bulletirs. However, the procedure does not clearly discuss the relationship between the operating organi:at.icn and the correctivt actiert systems. For TU Electric internal auditors, TU Electric management, anc NRC to be able to determine if appropedate actions have been taken, TU Electric must clarify the relationship between the ccrrective action system and the cperating organization.
The applicant suomitted a responso to unresolved Item 445/85-16-U-02 and 446/85-13-U-02 in the correspondence to NRC dated February 13, 1987. In this letter, the applicant statec that Bulletins recuiring licensee responses are processed by the Operations Support Section under crocedure ECE-AD-18. Al' other correspondence is routed to the Industrial Operating Experience Recore j 4
Coordinator fer evaluatien oer procedures N05-;C3.
~he acclicant has committed to perferm a croceoure and record review te ascer adn tre adecut:y :f ne 'EP eregran. ~be eview sheule as:ure {
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__ _ _ _ _ _ _ _ _ _ _ _ - - . _ _ . I
that all Bulletins were received and processed and that corrective actions wert initiated through the above established programs. It is the. Task Grcup's reconnendation that the NRC staff evaluate the effect-iveness of this program through an audit.
References:
. 1. NRC Inspection and Enforcement Manual, Chapter 0702, NRC Office'of Inspection and Enforcement Bulletins and Infonnation Notices, 2/21/86.
- 2. NRC' Inspection and Enforcement Manual, Chapter 92703, IE Bull'etin, Confirmatory Action Letter and Generic Letter Followup, 2/14/86 6
8 3-27
3.13 Issue 11 from Insoection P.ecert 85-16/13:
Description:
All reporting requirements of Inspection and Enforcement Bulletin 79-14 were not met.
Discussion:
Inspection and Enforcement Bulletin (IEB) 79-14 was issued because f:RC had identifed several operating facilities in which the as-built con-figuration of the piping systems did not agree with the seismic analysis design' inputs. (See Sections 3.11 and 3.12 for more details.) The Bulletin requested several actions of licensees and applicants, including that they identify and report to NRC on the status of nonconforming conditions noted during inspections that would cause safety systems to be inoperable under certain seismic events.
Sa'ety Significance:
This issue has ne safety significance baseo on the fact that facilities under construction were not intended to report nonconformances to the NRC in compliante with IEB 79-14. The Task Group discussed the Bulletin anc its scope with a cognizant inspectier and Enforcement staff member who confirmed triat the intent of reporting nonconformances was to assure that operating plants took appropriate corrective actions.
Followuo Actions and Reccmmendatinns:
This item was initially discussed as an unresolveo item in'the draft inspection report and subsequently dropped frcm the final inspection eport. The issue was never formally transmitted to the applicant; ! Pus, r.c corrective actions'wculd have been initiated. No further actions arc war anted cencerning tnis matter.
- -38 3
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I 3.14 Issue 8 from Inspection Reports P5-16/13:
~
NAMCO switches IEB 79-28 were not properly identified on installation travelers. {
1
Description:
Ouring)
(CPSES routine NRC inspections in November of Comanche 1985, inspectors Peak Steam Electric identified inconsistencies Station between certain tions NAMCO switch model numbers identified on the installation instruc-plant.gavelers) and the model numbers on the installed hardware in the Inspectors noted deficiencies in documentation and a delay associated with the filing of documents in the master data base and QA vault. In addition, DIA raised a concern related to the adequacy o the hardware when documenting the results of its review of this issue. [2 /
Discussion:
On December 7, 1979, NRC issued IES 79-28, "Possible Malfuretion of NAMC0 Mocei EA 180 Limit Switches at Elevated Temperatures." The purpose of the Bulietin was to alert the industry to a deficiency in certain manufactured lots of NAMCO EA 180 limit switches. NANCO CONTROLS, the switch manufacturer, determined that-the switch top cover gasket eritted a resin vapor at temperatures above 175'F. This vapor could condense into oeposits on the normally open contacts, possibly causing a switch malfunction.
By letter dated March 24, 1980, from R. J. Gary (TUGCO) to Xarl V. Seyfrit (NRC), the applicant for CPSES (hereafter the applicant) responded to !EB 79-28. In itr. response, the applicant stated that 14 EA-180 NAMCO switches required repli. cement of the top cover gasket, that none of the switches had been put in service or exposed to ambient temperatures of more than 175'F, and that the replacement gaskets were being ordered from NAMCO and would be insta11ec by June 30, 1980. Subsequently, by letter datec Ju h 30, 1981, from R. J. Gary to Karl V. Seyfrit, the applicant revised its earlier response to IEB 79-28, stating that due to difficulty resolving environmental qualification concerns, all NAMCO switches within the scope of IES 79-28 woule be replaced prior to plent operation.
During routine NRC inspectiers of CPSES in November 1985, the inspectors focused on the applicant's actions in response to IES 79-28, which apparently evolved into a broacer review of other NAMCO switches not addressed by the Bulletin. During the November 1985 inspection, inspectors identified inconsistencies between certain NAMCC switch model numbers identifiec cn the installation instructions (travelers) and the model numbers on tne installed hardware. Specifically, two NAMCO switches on residual heat removal (RHR) system valves 1-HCV-606 ano 1-FCV-618 were identified on travelers EE 82-1415-5801 ano EE 83-0373-5801 as EA 180-32302 anc EA 170-31302, respectively. The switches actually installed were stamped EA 180-31302 and EA 180-31302. It should be noted'that the Inspection 2-39
i Reoorts(85p'2/15)identindtheseswitchesasbe4rgwithinthescopeof e -
IES 79-25, newever, this is inconsist with the applicant's statement that these particular switches were replaced as part of the environmental cualif1 cation upgrade effort initiated and not in response to athe part of the fogeen switch replacement Bulletin Safety Significance:
In assessing the safety significance of the November 1985 findings, Task Group 3 assumec that the wrong switches were installed and that this situation existed without correction or recognition. Even based on this assumption, the Task Group determined that this issue is not safety significant. In performing its worst-case assessment, the Task Group reviewed the CPSES Final Safety Analysis Report (FSAR), information provided by the switch manufacturer (NANCO CONTROLS), anc TU Electric, and discussed the issue with their representatives.
As shown on Figure 5.4-6 of the CPSES FSAR, ( ) both valves 1-HCY-606 anc 1-FCV-618 are lccated in the A train of the Unit 1 RHR system outside the containment builcing. Valve 1-FCV-618 is an air diaphragm-operated butterfly valve utilized to control bypass flow arounc the A RHR heat exchanger. Valve 1-HVC-606 is an' air diaphragm-operated butterfly valve utilizec to control discharge flow fror. the A RHR heat exchanger. The W C0 switenes identified by the inspectors provide valve position indication reacouts in the emergency response facility anc control room.
!n assessing the safety significance of an error involving the installation cf these switches, Task Group 3 evaluated the cor. sequences of installation of switch model EA 170-31302 on RHR valve 1-FCV-618 in lieu of switch model numoer EA 180-31302. The Task Group learned that the rwitches differ in the environment. g construction Althougn bothandEAcapability 130 and EA to170 functien series in a hostile switches are cetignec to function in a high radiation field, switches in the EA 180 series have been desioned to function in a containment building follrwing a des';n bases accident under such harsh environmental conditicrs as elevatec temperatures and pressures. In :ne worst case scenario, a swi*cn l desigred to function in a mild environment cculd have been instal!6: :n i
ChR valve 1-FCV-613. Since Onis valve is located cutsloe ne certainmen; building in a mild environment. the Task Group concludes that such an error would not be safety significant.
In assessing the safety significance of an error involving the installation of switch EA 130-32302 en valve 1-HVC-606, rather than switch EA 180-31302, the Task Group determinec that the worst case woulc result in a malfunctioning valve position indicator. The Task Group learned that the switches ciffer in their internal return spring configuration and direction of rotation.
The EA 150-31302 switches are set up at the factory to operate in the clockwise fac: cry to direction opercte inand tnethe EA 180-32202 ccunter clockwise switchesareseyupatthe cirection. ' !nstalli ; a l
1 l 3 40
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l switch with the wrong direction of rotation woulo result in a false indication of valve position in the control room. Althcuch a false i inoication eculd be misleading to the plant operators, other information i is available to the control room operators about systet status, including RHR discharge to the reactor coolant system cold leg temperature recorder (TR-612), and RHR discharge .to the reactor coolant system cold leg flow indicator (F1-618), both of which can be read at the control board. T l theTaskGroupconcludesthatsuchanerrorisnotsafetysignificant.g, Moreover, an installation error that results in a malfunctioning valve position indicator should be detected and corrected as part of the pre-operational and startup test program. As a minimum, prior to declaring the RHR system valves operable, testing prescribed by Section XI of the ASME Code must be completed. One such test involves stroking each i valve and verifying the operability of the valve position indicator.
Followuo Actions and Recommendations:
As discussed in the OIA Report, according to the inspectors, the applicant provided NPC with two new travelers to demonstrate the acceptability of the as-built configuration, in addition, Region IV management provideo 01A copies of an applicant nonconformance report and four travehys that demonstrated the acceptability of the as-built configurations In response to an NRC request for information, I6 b D the applicart clarified the evolutionary nature of the NAFC0 switch replacement program.(5)
During the replacement precram for environmentally unqualified switches, the applicant replaced the existing type and mocel with eovivalent type and model limit switches. The last change to the switch on valve 1-FCV-618 was controlled by traveler EE 83-1851-5801 and completed in
, August, 1983. A review of this traveler by the applicant indicates that l switch EA 180-32302 was replaced with EA 180-31302. The last chance to the switch,on valve 1-HCV-606 was controlled by traveler EE B3-0459-5801 and completed in June, 1983. A review of this traveler by -he applicant indicates that switch EA 180-32302 was replaced with EA 180-31302. Based en a review of the appropriate recorcs and testinc performed, the applicent has confirmeo that the as-installed configuration satisfies the design requirements.
With regard to the oiscreparcy betweep,the travelers and the installed hardware identified by the inspector , the Task Group has concluded that it results from a delay in updating the official file. Changes made in mid-1983 had not yet been entered into the pemanent plant record in November 1985. The inspector initially obtained t NAMCO switch traveler information from the permanent plant record vault.g' None of the information available to the Task Group clarifies why outdated recoros j were being stored nor what assurances exist that the records would ever j have been updated. If the switch records were in the permanent plant record vault, this indicates that installation hac been completed and )
j possibly tnat a ;crtion of the system had been turnec cver to the plant. ]
i 3 -C. i 1
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Once piant records are completed, any charges to the equipment must be icentified in the records. !t is the Task Group's recommendation that the t closecut of the unresolved item should assure that the applicant understand: I the rcet cause of the traveler mismatch with the equipment. They should also assure that the records system is structured such that equipment records affected by in-plant eouipment chances are identified and tracked
~
to completion. The records system should assure that the changes are captured, that duplicate packages do not result, and that permar.ent records, such as ecuipment qualification files, are updated.
In reviewing the travelers associated with the NAMCO switch replacement for valves 1-FCV-618 and 1-HCV-6C6, the Task Group found at least three travelers that directed the installation of one or more switches designed to cperate in a particular direction of rotation in locatioris where the opposite direction of rotation was required. Travelers EE 82-1415-5801, ,
EE 83-0447-7802, and EE 83-0373-5201 directed the installation of switches with one direction of rotation ;nat were later replaced by a switch with the opposite direction of rotatien via travelers EE 83-0459-5801 anc EE 83-1851-5801. Because traveler EE 82-1415-5801 directed the removal of switches with the wrong direction of rotation, there is evidence that other travelers that proceeded EE E2-1415-5801 were in error. Althcugn the differences were identified and corrected, the circumstances raised a
- concern relative to the development cf travelers for the entire NAMCO switen replacement effort. It is the Task Group's recommendation that -he closecut of this unresolved item attempt to determine the cause of this ac;arent breakdown and its implications. The applicant has confirmed that all. switches affected by IEB 79-28 were identified and replaced ba:ed on
' programs in place for procurement and documentation review, on install-ation instructions utili:ed, on GA inspection recoros, and en walkdowns.
As a part of the current Corrective Action Program (CAP) in the equipment cualificatien area, a cceplete field verification of Class IE eouiprent anc documentation verification will be performed by the applicant. The purpose of this program is to icertify and resolve all as-ouflt environmental cutlification discrepancies. It is the Task Group's recommendation thr.:
the closecut of this unresolved item include monitoring of tne applicant's ongoing or: grams (including the CAP) to confirm that the correct switches Pave been in:talled.
References:
- 1. Letter to Texas Utilities Generating Company from E. H. Johnsen (NRC' dated April 4,1986 transmitting Inspection Reports 85-13 ana 85-16.
- 2. CIA Report 86-10, Attachment MM, Technical Review of Issues Centained in Cemanche Peak :nspection Reports.
- 3. Comanche Peak Stean Electric Station Final Safety Analysis Report, ;
Units 1 and 2. Docket Numeers 50-445 and 50 446. !
3 42
4
.. NAMC0 Limit Switches and Quick Connectors for Nuclear Environment, Series EA 180-302/602-Rev. and EA-170-302/602 Rev., 3M/5-85.
- 5. Letter from W. G. Counsil (TV Electric) to the NRC dated February 9, 1987.
- 6. Letter from V. S. Noonan (NRC) to W. G. Counsil (TV Electric) dated February 1, 1987
- 7. Letter from V. S. Noonan (NRC) to W. G. Ccunsil (TV Electric) dated February 6, 1987. ,
- 8. Memorandum from R. D. Martin (NRC) to J. G. Davis (NRC) dated January 20, 1987.
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3.15 Issue 12 from Inscec:sion Fecort 85-16/13:
Insufficient evidence of successful testing of BISCO fire seals'- filing of false report by BISCO - Validity of BISCO seal questioned.
Description:
In Comanche Peak Inspection Report 85-16/13(1) , NRC inspectors identified an unresolved item with regard to the qualification of eight Brand Industrie.1 Services, Inc. (BISCO) fire-rated electrical penetration seals (Nos. AB-790-174-1022A, EC-854-150A-1018A and B, EC-854-151A-2003A and -2004A, EC-854-151B-2025A and -2026A and TB-803-010A-1008A). Specifically, the inspectors could not verify fror available documentation that the actual seal installation met thedesignreg'rementsspecifiedintheComanchePeakFinalSafetyAnalysis Recor: (FSAR) .
~he drspectors found that documentation pertainir; to tne testing of the BISCO seals in accordance with the tes-ing stancards of ASTM E-119 anc IEEE 634, as specifiec in the FSAR, coulc not support the 3-nour rating certification statement provided by the applicant.
Discussicn:
FSAR Section 9.5 states that a minimum 3-hour fire-resistant barrier shall be orovideo that separates each cable spreading room from other plant areas anc that seperates recuncant safety divisions. This FSAP criteria
's in accercance with the requirements of Appendix R to 10 CFD Part 50 and the guiaelines cf Standaro Review Plan, Section 9.5.1. The incpection report inM cates tha the electrical penetration seals ir cuestion were intendec to provide e full 3-hour fire rated barrier between redundan- '
trains of safe shutcown system cabling in the icentified plen; areas. M so ratec, tre seals wouic crevent a fire in one room from screeding throu;b
- ne nail penetra*icn to Ine acjacent room ccntaining recundant safe snut-cown cabling for a mimimum of inree hours anc thereby creclude a loss of safe snutcown furctiuns. A typical electrical concuit fire resdstan senstration seal configurcticn is shewn in tne accompanying figure.
Safety Significance:
In the worst case, an unqualified penetration seal could rot adequately prevent the spread of fire between adjacent areas containing redundant safe shutdown caciing. Such a fire could. damage redundant cabling cesignec to acujeve cost-fire safe shutdown. Mcwever, b," 'e::er dated February 9, 1987, the aoolicant indicatec Inat of the eight penetration seals of concern, six (EC-554-150A-101SA anc B, EC-854-151A-2002A anc -200aA, and EC-854-151-2025A and 2026A) are installed with 1-hcur fire-ratec barriers, anc ne remaining two NB-CD3-CICA-1008A ard AE 790-174-1022AI are 'nstallec I a5
u ,
with 3-hour fire-rated boundaries. For the six 1-hour seals, no further concern exists since available documentation confirmed seal quelificaticn in excess of a 1-hour rating. Sheuld a fire breach the penetrations. of {
the remaining two boundaries, redundant trains of safe shutdown eouipment/ l caeling would not be affected since in one case (TB-803-010A-1008A), no safe shutdown ecuipment is in the fire area itself, and in the other- l
( AB-790-174-1027.A), the adjoining area contains only Unit 2 safe shutdown equipment. The impact on Unit 2 has not yet been assessed by the licensee, but the Unit 2 safe shutdown capability would be ensured by proper protection.
Thus, even if the seals in cuestion were to fail, the post-fire safe shut-down capability of the plant would not be lost. It should be further noted that installed automatic fire suppression systems in the areas of concern would provide additional protection against the spread of fire beyond that provided by the seals, anc that fire detectors would alert-olant operators to take any necessary manual actions to fight the-fire er initiate plant shutdown. Applicant calculations for the areas of concern aise indicate ccebustible leadings below the rating of the barriers.
Followuo Actions anc Recommendations:
In the February 9, 1987 letter, the applicant indicated that ccmpliance with licensing requirements for the seals will be achieved by reinstalling penetration seal AB-790-17a-1022A in accordance with the appropriate BISCO procedure prior to fuel load. No further action will be taken on penetra-tion seal T3-803-010A-1008A because no safe shutdown equipment would be affected by a fire in the area. For the six remaining seals in the 1-hour fire-rated barriers, no further action is planned since the documentation concern extended only to the 3-ncur fire rated seals. Finally, the applic-ant indicated that a verification walkdcwn of all as-built penetration seals will be cer#crmed. The Task Group concurs with the applicant's corrective actions anc concludes that ccmpliance with fire protection requirements will be achieved.
Sirce SISCC seals are used in a number of nuclear gewer plants, the above inspection #inding was treatec as a generic concern by Regicn P! and was referrec to tne Office of Inspection ano Enforcement, Vencor Branch, fer assdstance in resolution. The Vendor Branch reviewed the test cccumenta-tien concern and insoected at other plants and the BISCO facility. The results of their review will be issued shortly in the form of an inspection report. The report will provice general clarifying information en fire barrier poetration seal testing and will provide the necessary backup information tc assist inspectors when they review seal certification documentation in the future. Region P/ shoulc utilize this report wher reinspecting the SISCO seals and certification documentation at Comanche Peak to close out the unresolved inspection item and to confirm final compliance with fire protection requirements.
3-M ,
l
References:
- 1. Letter to Texas Ltilities Generating Ccmpany fror: E.H. Johnscr., Director, Division of Reactor Safety and Projects, NRC Regien IV, datec April 4, 1986.
- 2. Comanche Peak Steam Electric Station, Units 1 anc E Final Safety Analysis Report Docket Nes. 50-445 and 50-446.
- 3. Diagram of Typical Electrical Penetratier Fire Resistant Seal. '
a.. Letter to the NRC frem W.G. Counsil, Texas Utilities Electric Company, dated February 9, 1987.
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3-1 (1/20/86)
TASK GROUP 3 TASK FOR EXAMINATION RELATIVE TO SAFETY SIGNIFICANCE OF ISSUES OF OIA REPORT 86-10 This task relates to examination for safety significance of the 34 issues of Attachment 1 to Attachment MM of OIA Report 86-10 and the safety significance of any additional issues revealed by the activities of Task Group 1 and Task Group 2.
The efforts relative to Task Group 1 would ce to describe the safety significance of failure by the agency to perform inspection requirements of the IE construction inspection program, particularly these requirements which cannot now be fully performed cue to the extent of the plant completion. . These items would be the result of Task Group 1 activities under Items 4 and 5 (and described in Item 6d) of the task description TASK GROUP 1-TASK FOR DETERMINING
'dHETHER THE CURRENT AUGMENTED REVIEW AND INSPECTION EFFORT AT COMANCHE PEAK IS SUF::C!ENT TO COMPENSATE FOP ANY :DENTIFIED WEAKNESSES IN REGION IV's
!NSPECTION PROGRAMS. In addition, the efforts relative to Task Group 2 would be the safety significance of process and disposition issues identified by Task Grcup 2. These items would be the result of Task Group 2 activities under Itees 1 and 2 (and describoc in Items de and f) of task performance in tne task description TASK GPOUP 2-TASK FOR EXAMINATION OF ISSUES RELATING TO PROCESS AND DISPOSITION OF INSPECTION FINDINGS OF OIA REPORT 86-10.
The Task Group will.have available to it:
'.. The information of issues from CI A Report 66-10 (essentially the 24 issues in at achment ! to At acnment MM ef CIA Report 36-10).
1
_s . .
3-2 *
- 2. Any other ~ issues needing safety significance review frcm tne work of Task Group ~1.
- 3. ~Any other issues needing safety significance review from the work of Task Group 2.
4 Basic inspection and licensir.g information and individuals' knowledgeable of issues.
4 Insofar as possible, the performance of this task should rely on the written record where that record provides sufficient information to support '
conclusions. When it is necessary to extend this record, this should be done.
in writing or by interview. Transcripts are to be made of interviews.
In addition to skills available within the agency, Task Group 3, whenever nechssary or desirable, should use outside exoerts to provide needed skills anc objectivity.
This Task Group shall:
- 1. Analyze each issue and describe its safety significance to the Comanche Peak reactor assuming that the condition of the issue is as stated in OIA Report 86-10. The purpose of this analysis is to specify the " worst safety case" that r,ay result assuming that the issue exists as stated without correction or recognition that the adverse situation exists, that is, as if it existed ano was undetected. In this analysis, trovice complete and scecific references that serve as tne basis for determination of safety sign 1ficance.
4 i
3-3
- 2. Analyze the actions which were ta. ken or which are planned to be taken to alleviate any safety significance associated with each of the 34 issues at Comanche Peak identified in OIA Report 86-10. In this analysis, give references to reports and other dncuments that serve as the basis for determining the actions taken or planned and their effect in alleviating concerns. Evaluate the actions to determine the extent to which they alleviate or resolve the safety concern. If the actions taken or planned do not sufficiently alleviate the safety concerns, describe actions which, if taken, would alleyiate the concerns. Give references in support of opinions or conclusions reecrding corrective actions.
- 3. Evaluate the safety significance of the #ailure to completely perform the inspection requirement in accordance with the IE construction inspection program. (Note: The input for this item will come from Task Force 1. ) The analysis should take into account all inspection activities associated with an incomplete or non-performed item.
Safety concerns should be described. If agency actions alleviate concerns, these should be cescribec. If the actions taken or planned do not alleviate concerns, cescribe actions which, if taken, wouic alleviate concerns.
4 Evaluate the safety significance of items identified by Task Force 3 that were net orocessed accoroing to guidance or which are in dispute as to cor ectress of processing anc disposition. All safety concerns shculd De identifieo. :f actions have alleviatec these conceros, these actiens snoulc be cescribec. If no action has been taken or if actions taken or ;!annec cc not alleviate -he concern, cescribe actions wnien, if taken, wou': al'eviate concerns.
~
3-t i
- 5. Report on the task efforts: '
- a. Describe how the task was performed. '
- b. Identify the information base on which the task group's are based. This should describe any auditing to assure the accuracy ,
of the information base. .
L
- c. Provide complete and specific references to the'information that serves as the basis for conclusions and findings.
- d. Highlight any safety significance issues where agency actions or e s
plans may not alleviate safety concerns.
Schecule 1/ 0/57 Task Leacer accointed 1/I2/87 Membershio, Organi:ation 1. Approach established Start work on example (pilot effort)
- /22/E7 1/P/97 Feview oilet effert sitn CPRRG 2/5/87 Analyze reports of Tasks 1 and 2. Factor into Task 3
!?
2/7/37 Submit final report to CPRRG l
1
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- Y ' #" I ' t*i UNITED STATES
~[ c n NUCLEAR REGULATORY COMMISSION
- WASHINGTON, O. C. 20555 g g
\;. - f MEMORANDUM FOR: Guy A. Arlotto. Chairman Comanche Peak Report Review Group FROM:
SUBJECT:
William G. Mcdonald, Acting Director, IRM h
COMANCHE PEAK REPORT REVIEW TASK RG GROUP IRM-01 (CPRRG Attached is the draft report on the 766 System that you requested. I feel that the report answers the questions'the CPRRG raised about the 766 System, but not necessarily in the same order included in your letter of February 12. To expedite the draft report issuance, I have included the specific questions and our answers in this letter of transmittal. It could be incorporated into the executive summary of the final report.
The draft report also covers areas not specifically requested, but which I feel are germane to an overall assessment of the 766 System and its implementation.
{
Ouestion 1.
The intended purpose and use of the 766 System.
Answer The original purpose of the 766 System as stated in a 1976 issue of IE Manual Chapter 0535 was to be "a management tool used to capture, maintain and report statistical and plannino data concerning inspection, investigation, inquiry activities and associated enforcement actions conducted by IE".
L_______-____. _.
t .. ;
G. A. Arlotto 2 j
l Interviews with cognizant NRC personnel familiar with the 766 System development indicated the principal uses originally envisioned were for budgeting and resource allocation; to monitor Regional performance and to provide field data k the time required to perform inspection procedures in )
the then new module system. These uses were prima'rily for Headquarters rather than the Regions.
Detailed information in this area is included in Section IV of the report.
Question 2 Determine if the current use of the 766 System by both Headquarters and Regional Office management is consistent with the intended purpose. Does the existing 766 System meet the stated purpose? What reasonable use can be made of the 766 System?
Answer IE Headquarters is still using the 766 System for some cf the original uses discussed in the response to Question 1. Regional Office management has attempted to utilize the 766 data for its cwn purposes, but with mixed success, which appears to vary from Region to Region. Some possible Regional uses do not appear to have been considered in the original 1 purpose. The 766 System meets the somewhat general stated purposes, but !
{
does not meet current Headquarters and Regional needs. The existing 760
'l 1
_____ _-__- _ _ - ___ A
't b G. A. Arlotto 3
)
System can be used for overall assessment of Regional performance in such areas as inspection time versus SALP ratings, and average time to issue I
inspection reports. It can also be used for a history of most violations l to add perspective to enforcement actions. i Detailed information can be found in Section VI of the report. i l.
Ouestion 3 If the 766 System does not fulfill its intended ' purpose and if this purpose l 1s necessary for an effective Inspection and Enforcement program, how are these objectives currently being accomplished?
Answer It is believed that the 766 System fulfilled its original purposes, but the original purposes and uses are not adequate to meet the current needs of the Inspection and Enforcement program. A great many changes have taken place over the years and the 766 System has simply not kept up with the changes. This forced the Regions into developing their own systems and methods, which contributed to 766 problems.
4 Details can be found in Sections VI and VII.
,s a G. A. Arlotto 4
Question 4 Alternatives and recommendations for the continuation of the NRC 766' System in order to satisfy the needs of NRC Headquarters and Regicnal Office management.
Answer _.
Section VIII of the report includes alternatives and recommendations on management information needs.
4
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