ML20207K550

Rev 1 to TVA Employee Concerns Special Program,Sequoyah Nuclear Plant Element:Review of Nuclear Safety Review Staff Open Items Requiring Corrective Action at Sequoyah

ML20207K550
Person / Time
Site:	Sequoyah
Issue date:	12/16/1986
From:	Debbage A, Gass K, Knightly J TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML20207K525	List: ML20207K522 ML20207K540 ML20207K550
References
ECTG-NSRS-01, ECTG-NSRS-1, NUDOCS 8701090447
Download: ML20207K550 (41)

Text

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TVA EMPLOYEE CONCERNS REPORT NUMBER: ECTG-NSRS-01 SPECIAL PROGRAM REPORT TYPE: Sequoyah Nuclear Plant Element REVISION NUMBER: 1 TITLE: Review of Nuclear Safety Review Staff Open Items Requirir.g Corrective Action at Sequoyah F

REASON FOR REVISION: To incorporate TAS, SRP, and internal review comments.

i j SWEC

SUMMARY

STATEMENT: The 8 items in this report were identified by the l

Nuclear Safety Review Staff (NSRS) and were assigned to the Employee Concerns -

Task Group (ECTG) for verification and closure. The verification activities are in progress, with present status described in this report.

PREPARATION PREPARED BY:

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N/A i ECSP MANAGER DATE MANAGER OF NUCLEAR POWER DATE CONCURRENCE (FINAL REPORT ONLY) i

• SRP Secretary's signature denotes SRP concurrences are in files.

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TENNESSEE VALLEY AUTHORITY l

SEQUOYAH NUCLEAR PLANT EMPLOYEE CONCERNS TASK GROUP

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t Subcategory: Sequoyah-NSRS Classical Items i

r Element: Review of Nuclear Safety Review Staff Open Items Requiring Corrective Action At Sequoyah

! Report Number: ECTG-NSRS-01 k

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Evaluator: 11hdf6 7

A.G.Debbag Ddtel Reviewed by: Y/ ~ ~

//-/4- BL K. R.' Gass Date Approved by: N W rh/6[b J. J. Knightly Date f-f.

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Rsview cf Nuclear Safety Raview Staff Opsn Items Raquiring Corrective Action At Sequoyah i

I. Introduction i The Nuclear Safety Review Staff (NSRS) conducted reviews of activities at Sequoyah Nuclear Plant (SQN) during its span of overview responsibility from 1979 to 1986, and at the commencement of 1986 there were 42 open items requiring follow-up review for closure. In early i February 1986, NSRS conducted its final review of NSRS open items at SQN j and closed 38 in report R-86-01-SQN (reference 7). At.this same time, a i Design Quality Assurance Branch surveillance was conducted within the Division of Nuclear Engineering (DNE) to review all NSRS open items applicable to SQN, and those items which had generic applicability to SQN where OE had responsibility for corrective action or response, to determine which items must be resolved before Sequoyah restart. '

Surveillance report S86-10 (reference 2) identified the five items which j were required to be resolved before startup and these were transmitted to the Project Manager, Sequoyah Engineering Project on February 14 -

1986.

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A subsequent memorandum dated May 1, 1986, (reference 1), from the Director of the Nuclear Manager's Review Group (NMRG) to the Manager of Licensing identified eight open items affecting SQN restart. The eight included the five items identified by DNE and additional items r identified in NSRS report R-86-01-SQN. -

The eight open items are:

R-80-05-SQN-4B Configuration Control l

R-84-19-WBN-01 Clear Identification of Purposes and Uses of All l l Controlled Documents at All Plants >

R-84-19-WBN-07 Development of Criteria for Cable Tray Fill Level and QC  !

Inspection .

R-85-02-SQN/WBN-2 Maintenance, Operating, and Test Instruction  ;

I-85-06-WBN-01 The Adequacy of the Dispositions for Identified Cable f

, Bend Radius Problems [

[

l I-85-06-WBN-02 The Adequacy of the Program for Cable Pulling Activities

• I-85-06-WBN-05 Underdevelopment of Established and Documented Limited  !

l QA Program and Failure to Incorporate the Established ,

! Limited QA Program Into Nuclear Quality Assurance Manual  !

(NQAM) l R-86-01-SQN-01 Improvements in Overall-As Low as Reasonably Achievable f l

(ALARA) Program.  !

! i The review detailed in this report examined all eight items. {

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1 In Juns 1986, ths Wetts Bar Nuclear Plent's (WBN) spscial Employee ,

Concerns Task Group (ECTG) assumed responsibility for tracking and verification of corrective action on all NSRS classical report open l items (reference 96). This was in addition to the open items resulting j from the NSRS/ Quality Technology Company Employee Concern Program. No Corrective Action Tracking Documents (CATD's) will be issued for the I classical open items. Several ECTG reports include classical report l N

details. CATD's have been written, or are in the process of being g written for many areas similar to those identified in the classical reports. Also, SQN has plans in place for resolution of most of these I items. ECTG plans to work with SQN to resolve these items. Since the g classical open items are restart issues, this report will be revised periodically before restart to show the status of each item. l The following is the present ECTG status of these eight items as a result of the verification activities conducted by the ECTG.

Closed Remain Open .

R-85-02-SQN/WBN-02 R-80-05-SQN-04B R-86-01-SQN-01 R-84-19-WBN-01 Q R-84-19-WBN-07 I-85-06-WBN-01 I-85-06-WBN-02 I-85-06-WBN-05 Following preparation of this report, a review of all NSRS classical report open items was undertaken by SQN and ECTG. Many han been assigned to organizations other than SQN, but failure to resolve them would impact SQN operations. In addition to the 8 items in this report, 17 items were identified as requiring closure verification by ECTG before SON startup. An additional 69 items were considered to be applicable to SQN but do not appear to require closure before startup.

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These 17 items are reviewed in report ECTG-NSRS-02.

II. Verification of NSRS Open Items A. (0 pen) R-80-05-SQN-04B, Configuration Control

• l. Ba.ckground In the original review dated June 1980, NSRS concluded from electrical deficiencies it found that configuration control was inadequate and made the following recommendations:

Page 2 of 39

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Rsvisa the existing Configuration Control Pecgram including appropriate instructions to require more frequent and indepth inspections such that the entire plant has been inspected each refueling cycle.

During the February 1986 follow-up (reference 7), NSRS examined the Configuration Control Program in place at the plant and reviewed the status of the TVA changeover to the configuration control drawing system at SQN. A number of actions had been taken which improved the confidence that the "as-constructed" - .

drawings and the actual plant configuration of'CSSC systems agreed. Some of these were:

a. Control of temporary alterations had been tightened over the years, and a periodic PORC review of outstanding temporary alterations was required (AI-9, " Control of Temporary Alterations and Use of the Temporary Alterations Order").

b. Control of plant modifications (AI-19. Part IV, " Plant -

Modifications: After Licensing") required marking the required control drawings as soon as the modification was field complete and provided for marking the drawings for a partially completed modification if the system was to be operated.

c. All personnel were charged with immediately reporting discrepancies they found between the plant configuration and "as-constructed" drawings (AI-25, Part I, " Drawing Control After Unit Licensing").

d. Maintenance Instruction (MI) 6.20 " Configuration Control During Maintenanca Activities," provided a simple method of controlling temporary conditions during maintenance activities as an alternative to the Temporary Alteration Control Forms (TACFs). This reduced the number of TACFs which made that program easier to track, and made control of temporary conditions during maintenance easier, and therefore less likely to be subverted.

e. System operating instructions now require double verification of the operational alignment configuration of critical systems.

f. Unit 1 and unit 2 control drawings that shared common equipment were compared and discrepancies documented through AI-25.

g. Drawings with areas marked " incomplete" were updated with information obtained by researching construction work packages.

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In establishing tna ecnfiguration basslins, NSRS considsrad it important to verify that the plant configuration and the "as-constructed" drawings agree by performing walkdown inspections. In consideration of the considerable improvement in configuration control since 1980, NSRS no longer considered it necessary to verify plant configuration every refueling outage. This item remained open pending verification of the configuration baseline. To achieve this, NSRS recommended the following actions to be taken before restart.

a. Completion of phase I of the plan for conversion to configuration control drawings for those drawings ,

previously identified by the' plant as necessary for CSSC configuration control.

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b. Selection by the plant of a number of CSSC systems for complete walkdown to verify that the actual physical

configuration agrees with the CSSC configuration control. -

drawings verified as part of the phase I effort.  ;

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c. Walkdown of the selected systems and correction of any discrepancies found. If significant problems are found.

additional systems should be selected for walkdown.

i 1 2. Corrective Action Taken TVA has had two surveys of the design control program performed by Gilbert-Commonwealth. The first survey covered the program .

activities after June 1985 (reference 14). The second survey covered the program activities from operating license to June 1985. This second survey had extensive NRC involvement and j resulted in NRC inspection reports 50-327/86-27 and 50-328/86-27.

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To confirm the adequacy of past modification work, a SQN Design Baseline and Verification Program has been established. This program has been designed to address the design control issues j by:

! Verifying and establishing the plant configuration.

Reconstructing the design basis.

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Reviewing and' evaluating all modification changes since

,' Operating License (OL) issuance against the design basis.

j Performing the required tests or modifications developed from this review and evaluation.

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l All ECN's not cosplete b2 fore issuance of the OL which wers initially determined not to be within the scope of the Design Baseline and Verification Program will receive reviews commensurate with the importance of the system function. A more

-- indepth review will then follow to ensure that no interrelations exist that may need to be considered by the Design Baseline and Verification Program. An additional review will be performed on those Engineering Change Notices (ECNs) that are written against systems that support normal plant operations to ensure that the system's " function ability" is not compromised (reference 63).

By memorandum dated April 18, 1986 (reference 3), SQN stated that additional walkdowns would be performed as part of the pre-startup program. A program was also underway to develop and issue procedures for walkdowns, assign staffing, and provide training for walkdown personnel.

The implementation of this program has been structured into two .

distinct phases. The pre-restart phase is to carry out the program fer the systems required to mitigate Final Safety Analysis Report (FSAR) Chapter 15 accidents and provide for safe shutdown.

The post-restart phase will continue the engineering activities to a point that complete engineering documentation and evaluations will be developed describing the final as-constructed configurations with supporting engineering technical justifications.

3. Specific Verification Methodology The corrective action undertaken by the plant was examined, which included:

Procedures for Drawing Control Procedures for Walkdown Walkdowns Review of audits and support The procedures reviewed included the following Sequoyah I g/ '

Engineering Project (SQEP) procedures: I SQEP-AI-8 Drawing and Reproduction SQEP-AI-8A Equipment Qualification Program Documentation SQEP-8 Packaging and Controlling of Walkdown/ Test Documentation SQEP-2 Procedure For Evaluating Engineering Change Notice and Field Change Notice Documents SQEP-6 Procedures For System Evaluation and Development of System Evaluation Report Page 5 of 39

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Attechment 4 to SQEP-8 dstalls the System Walkdcwn List for unit 2 as follows:

Unit 2 l Walkdown Package No. System Name I

WDP- 1 Main Steam l

-3 Main and Auxiliary Feedwater System l

-7 Upper Head Injection  !

-15 Steam Generator Blowdown

-18 Fuel Oil  !

Fire Protection (Containment Isolation Only)  !

-26

-30 Ventilation -

1

-31 Air-Conditioning 1

-32 Control Air l

-43 Post Accident Sampling l WDP-46 Feedwater Control (AFWTDP Control) l gf

-59 Domineralized Water l

-61 Ice Condenser (Containment Isolation Only) 4

-62 Chemical and Volume Control

-63 Safety Injection System

-65 Emergency Gas Treatment System l

-67 Essential Raw Cooling system l

-68 Reactor Coolant System l

-70 Component Cooling Water System  !

-72 Containment Spray l

-74 Residual Heat Removal (RHR) System  !

-77 Waste Disposal System  !

-78 Spent Fuel Pit Cooling (Containment Isolation  !

Only) g Page 6 of 39 i

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-81 Primary Water (Containment Isolation Only) l

-82 Diesel Starting Air (Including DG) l

-83 Hydrogen Mitigation l

-88 Containment Isolation l

-90 Radiation Monitoring l 1

-92 Neutron Monitoring System I

-99 Reactor Protection g WDP-AP Auxiliary Power l

WDP-VP Vital Power .

4. Verification Analysis The staffing for the walkdown program was reviewed. It appeared that adequate numbers of personnel had been made available for this activity. The Electrical Section had 61 technical and 5 support personnel. Discussions with plant personnel involved in walkdowns indicated that progress was being made and was not limited by manpower shortages.

The SQN Nuclear Performance Plan was reviewed. The SQN activities list contains restart items with action list unique identification numbers:

0101 The pre-restart portion of the design baseline and verification program will consist of the following four elements for the systems or portions of systems listed in Table 6: ( 1.) Design criteria / design basis; (2.) System walkdown/ test; (3.) Evaluation of ECNs and other changes; and (4.) System evaluations and corrective actions.

0102 The Engineering Assurance Staff will perform an independent oversight review of the design baseline and verification program.

0096 A transitional design control system will be implemented prior to restart.

0085 Resolve Significant Condition Report (SCR) EQP8505 by establishing a program to maintain drawings in an as-configured status for the Environmental Qualification (EQ) program once the baseline is developed. -

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Tha status of the progres ieplamsntation w:s reviewsd. SQEP-4 program status for SQN-2 . dated August 19, 1986, showed for 37 systems - 15 had 100-percent completion of ECNs with 6 at 0-percent completion. Electrical Discipline Design Baseline Verification Status SQEP-2, dated August 18, 1986, showed for 35 systems - 20 had 100-percent completion of ECNs.with 4 greater than 90-percent. Discussions with design personnel indicated that there was still much work to be performed before the Configuration Control Program would be satisfactorily completed. It is evident in this review by ECTG that both commitment and resources are dedicated to SQN, which on resolution will adequately resolve the NSRS recommendation.

5. Completion Status Further progress is needed at SQN to complete the configuration i control program. The commitment and resources will ensure its completion. This item will remain open until program completion.

i B. (0 pen) R-84-19-WBN-01, Clear Identification of Purpose and Use of l

All Controlled Documents at All Plants

1. Background

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{ In the original review NSRS (reference 20) made the following

{ recommendations: "All controlled documents should be clearly i

identified for all plants. The purposes and uses for each of the documents should be delineated. Information contained in I documents designated to be controlled should be assessed for "

contribution to the intended purpose and use. Superfluous information should be deleted and discrepancies in documents 6 with overlapping information should be corrected. Establishing a verified as-built drawing control system should be assigned a I very high priority.

The NSRS review of the Black and Vestch (B&V) report noted that

task force category 3 contained 25 B&V findings where 4

logic / control drawings did not agree with the electrical i drawings. The identified cause for the category was failure to implement design review procedures as required by engineering procedure EP 4.25. The task force concluded the problems were generic to logic, control, schematic, and connection diagrams throughout Watts Bar Nuclear Plant (WBN) units 1 and 2. The review was extended to three additional systems where similar problems were found. It was determined that corrective action was required for both past and future work."

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The report stated that WBN issusd ECNs and Field Change Raquests

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(FCRs) to correct identified errors in hardware wiring, and training was conducted in the Instrumentation and Control section of Sequoyah/ Watts Bar Project (SWP) for Engineering a Procedure (EP) 4.25. The WBN drawings were to be stamped to restrict the use to the intended function. No further reviews of other systems were planned to determine if other systems had the same problems.

i By memorandum dated July 31, 1984 (reference 20), WBN agreed

} that the condition existed; however, they did not. agree that the condition had a detrimental effect on quality. They had determined that the users of the design informati~o'n were in fact utilizing the drawings with the correct information. The only exception to this conclusion was the use of logic drawings.

Division of Engineering Design (EN DES) had completed corrective 4

action by reviewing all WBN logic drawings.to remove discrepant i information; this was documented on ECN 4666, which was closed. .

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In follow-up review dated September 5, 1984 (reference 21), NSRS stated that the recommendation had not been fully complied with j since it referred to all TVA plants and the response only I addressed WBN. As part of the follow-up review, NSRS examined a

! number of drawings which were changed under ECNs 4666 and 4667.

4 The changes appeared to clarify and correct the drawings listed. A large number of logic and control drawings were j changed. Therefore, NSRS considers this issue satisfied for WBN i because of the corrective actions taken and verified. It was 4 left open until EN DES completed a similar review and made' 4

l corrections as needed for SQN, Bellefonte (BLN) and Browns Ferry Nuclear Plants (BFN) logic and control drawings versus electrical drawings and termination lists.

2. Corrective Action Taken i

f WBN corrective action was completed with closure of ECM 4666. J For SQN, OE placed a request on June 27, 1984, for funding by }

ONP to address applicability of B&Y findings (reference 68). [

! I l The SQN Site Director stated in a memorandum dated r l

September 12, 1984 (reference 22), that no action was necessary  !

by the plant. "0NP recognized that the logic prints were not i properly controlled. Electrical prints were used for actual j equipment operation. This negated the safety impact of the ,

discrepancies between the logic prints and electrical prints. t i The SQN drawings hau been subjected to more review than those of WBN because of the preoperation start-up and surveillance l testing programs. Operations personnel routinely reviewed logic l prints before using electrical drawings to operate equipment."

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3. Spscific Varification Methodology Discussions were held with former NSRS reviewers to clarify the nature of the discrepancies between logic / control diagram and detailed electrical drawings, and also, to determine the s corrective action needed at SQN to close out this item. The 8 Nuclear Performance Plan was examined to determino reference and I actions planned. Also, action item 0140 was reviewed. The status of this open item was discussed with SQN design supervisors.

} 4. Verification Analysis 1

The B&V task force noted that logic / control drawings did not always agree with the electrical drawings. In the Nuclear Performance Plan, section 3.2.2(b) the system walkdown/ test is  !

detailed as part of the design baseline and verification program. Drawings to be reviewed are flow diagrams, single-line drawings, and control and logic diagrams located in the control -

room depicting the systems listed in this review j (section II.A.3).

Nuclear Performance Plan, part IV, (revision June 17, 1986),

i gave a summary of SQN restart items. Restart item 551 stated:

u Address NSRS open item'R-84-191WBN-01.by resolving discrepancies between electrical logic / control diagrams and i detailed electrical drawings.

Review of correspondence showed that restart item 551 was reassigned to DNE-SQN Nay 14, 1986, for resolution  !

(reference 23).

DNE's Operational Readiness Plan was examined (reference 69).

Action item 221 stated:

B&V task force category 3 - Logic / control drawings do not agree with electrical drawings.

Action item 299 noted: l NSRS open item R-84-19-WBN-01, clear identification of purpose tnd uses of all controlled documents at all plants.

The SQN restart unimplemented design item evaluation for action item 299 was examined. The evaluation erroneously stated that j

it was not a safety problem, and any work performed on this item l would have no effect on quality. SQEP electrical design apparently did not regard this as a restart item at that time. .

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The SQN Nuclear Parformance Plan (July 17, 1986) was rsviewed.

The SQN activities list contains a restart item with the action list unique identification number 0140:

Address NSRS open R-84-19-WBN-01 involving discrepancies between electrical logic / control diagrams and detailed electrical drawings (B&V, Chapter 3)

5. Completion Status The status of R-84-19-WBN-01 needs to be examined at SQN. It should be verified that planned actions will ensure that the information errors on control and logic diagrams are corrected.

This will satisfy the NSRS recommendations. Until this verification is made the item will remain open.

C. (0 pen) R-84-19-WBN-07, Development of Criteria For Cable Tray Fill Level and OC Inspections

1. Background In the original review at WBN, NSRS concluded that criteria should be developed for field use to control actual cable tray fill levels and to provide a basis for Quality Control (QC) inspection. A feedback system should be included from the construction forces pulling cable to the designers routing cable to avoid the overfill problems to date. Although the cable tray levels at WBN unit 1 may already be established in many instances, expeditious action should be taken to upgrade the system for WBN unit 2 and BLN.

BSV review in task force category 36 found that cable tray fill criteria were not ensured of being met because of the less-than-conservative nominal values used for cable cross-sectional areas in the cable routing program. After evaluation by designers Division of Nuclear Engineering (DNE) concluded that the licensing requirements had been met and no corrective actions are required for either past or future work.

N3RS reported that it was not clear that the licensing basis had been met. The WBN FSAR stated that ". . . low-voltage power cable tray fill shall be limited to a maximum of 30 percent of the cross-sectional area of the tray, except when a single layer of cable is used. Cable tray fill for control and instrumentation cables shall be limited to a maximum fill of 60-percent of the cross-sectional area of the tray." The supporting EN DES documentation for the conclusion that the licensing requirements had been met was based upon considerations of dead weight, capacity,'and heating value of O

Page 11 of 39

ccLbustible in insulation and jacket materials. While NSRS agreed these were important considerations, there were others such as mechanical protection of the cables from missiles or casual hazards.

The FSAR described a fully automated computerized system to I route cables and to control cable fill using the criteria stated above. Contrary to this, there was no variable for control for cables of the same gauge but different diameter; there was no formal feedback procedure to alert the designer when for vagaries of construction, the tray was physically full before

! all the cables were installed as computer routed. Further, no l acceptance criteria were provided for either the installer or

the QC inspector to use in order to consistently determine that a tray was physically full. NSRS believed safety evaluations should be made of the conditions described before substantial plant operation.

By memorandum dated July 31, 1984 (reference 20), the Nanager of

• Power stated that Power did not agree with the concern on cable tray fill documentation. "The documentation of cable tray fill existed in the fully automated computerized system to route the cables." The computer system was controlled by WBP-EP-3.13. A change to the FSAR (emendment 52) was submitted to explain the use of nominal cable diameters in cable tray fill calculations.

NSRS recognized (reference 26) that TVA used a computerized system to route cables and to limit the fill in the cable trays. Although this system was used to assist and to document what was actually accomplished in the field, the computer system could not be used as a final acceptance vehicle without some verification of what existed in the field. The concern regarding the cable routing system was raised when NSRS observed that cabling in many areas had exceeded the height of the side rails of the cable trays, even though the tray seemed (in most cases) to have sufficient area to lay cable below the side rails. This physical condition at the plant also negated the natural protection the cable receives from the side rails, thereby unnecessarily exposing them to damage. NSRS recognized-that the National Electrical Code (NEC) did not specify tray fill criteria until 1975.

The reply by DNE dated October 3, 1984 (reference 27), stated:

NSRS Position a - Develop criteria for field use to control actual tray fill levels and to provide a basis for QC inspection.

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OE Response "This NSRS position is based on the idea that since cables have been found above side rails, the accuracy of the computerized cable routing program is questionable and should be verified. TVA's commitment relative to cable tray fill at WBN is contained in Section 8.3.1.4.1 of the FSAR. This commitment does not prevent TVA from filling cable trays above the side rails. Since its inception in the mid-1970's, the computerized cable routing program had successfully routed more that 56,000 cables, attesting to the accuracy of the program.

Lack of the natural protection afforded by the tray side rails was of no major significance when it is recognized that layout of the tray system had been predicted on or analyzed for missile  ;

protection, a seismic event, pipe whip / jet impingement, and fire. BLN was TVA's first nuclear plant committed to meeting l NRC Regulatory Guide 1.75 which required that cable trays not be filled above the side rails. For these reasons, OE still contended that there was no need for documentation of the actual cable tray fill." .  ;

i NSRS Position b - Either QC or the appropriate Quality Assurance I (QA) organization should, through an inspection and/or audit l process, determine if the existing installation meets the established criteria.

OE Response - A surveillance report on the computerized cable routing program had been conducted by the Quality Management Staff (Reference 70).

This report contained a recommendation to delete the Watts Bar Engineering-Construction Monitoring and Documentation (ECN&D)

Program User's Guide as a reference in WBP-EP 43.13 " Cable Schedule Handling Procedure - Watts Bar Nuclear Plant." This recommendation was made on the basis of the ECN&D program not being a controlled document.

NSRS Position c - Where deviations from the FSAR cammitment are-made, TVA should perform a safety analysia to justify the  ;

deviations. Such deviations should be examined for i reportability to NRC.

4 OE Response - There were no deviations from the FSAR commitment.

i A report by H. C. Rutherford dated May 14, 1985 (reference 28),

stated in part that "The ECM&D program was an integral part of the overall computerized cable routing and installation documentation program, and simply deleting reference to it in WBPEP 43.13 wculd also not resolve the concern."

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NSRS Stctus Rsvisw datsd August 21, 1985 (esfersnce 29), statsd that OE issued nonconformance reports (NCRs) concerning the lack of documentation to show verification and validation of the computerized cable routing programs and accompanying data tables.

NCR WBNECB8501 was issued on Apell 8, 1985. Plant specific NCRs were also issued on BFN, SQN, and BLN. According to H. C. Rutherford's memorandum (reference 28), the degree of noncompliance to the OE procedures would be different for each plant. Corrective actions had not teen established for the NCRs. However, Rutherford's memorandum stated, "If~ appropriate changes were made to provide independent verification of the overall process and to provide control over changes, then use of the computerized system in conjunction with established QC inspections of cable routing should provide a satisfactory process to assure that cable fill requirements were not exceeded." -

NSRS status review transmitted to ONP September 13, 1985 (reference 30), was as follows:

1. A satisfactory verification process must be established and maintained for the computer cable rcuting programs and data tables.

2. Appropriate change controls must be developed and implemented in order to control changes to the computer programs.

2. Corrective Action Taken The DNE on May 20, 1986 (reference 31) stated that further action or classification by DNE was required to close out the concerns. If appropriate changes were made to provide independent verification of the overall process and to provide control over changes, then use of computerized system would provide a satisfactory process to assure cable fill requirements were not exceeded.

H. C. Rutherford's memorandum (reference 28), outlined the steps necessary to establish verification and control of the computerized cable system. Resolution of these steps were being implemented as the corrective action under NCR WBNECB8501.

3. Specific Verification Methodology The original report and correspondence relating to this concern was reviewed. Principals involved in this activity were contacted to determine the status for SQN. .The NCR for WBN was WBNECB 8501.

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How2 War, the ccncarn applicable to SQN was mainly directed to ensuring that cable tray fill levels were satisfactory. The SQN Nuclear Performance Plan was reviewed to determine actions planned or taken to resolve this item. NCR's for SQN were examined and the status determined.

4. Verification Analysis Review of the SQN Nuclear Performance Plan (revision June 17, 1986) listed under " Cable Tray Support Analysis,"

j showed the following restart items.

! Restart Item 308 Resolve SCREEB8529 by demonstration that exceeding TVA conduit fill standards are acceptable or take corrective action.

Restart Item 194 Resolve SCREEB8620 by reanalysis of 1 overloaded cable trays. Current plan is to .

1 l reanalyze trays that were loaded beyond TVA l standard by a computer model which did not consider abandoned cable.

Significant Condition Report (SCR) EEB 8529 (reference 71) was i originated December 6, 1985, and closed; January 10, 1986. It

stated that WBN PIRNEEB 8546 had identified examples of WBN conduits containing 400 MCM cables that exceeded conduit fill as I specified in DS-E13.1.2 and DS-13.1.4. -A computer run was made of all 400 MCM 600-Y cables at SQN. Fifty-five conduits j exceeded allowable fill standards. The engineering report J (reference 72) stated that exceeding the percentage conduit fill

! did not in itself constitute a failure. Tests and inspections

! of the cables did not revsal any damage after installation. It i was also found that the percent fill was only violated for part 4

of the distance in some cable installation because of a change In conduit size. The evaluation concluded that no modifications were necessary and SCR completion verification sheet issued (reference 73).

l SCR SQNEEB 8620 R0 (reference 74) was originated 4 February 2, 1986. It stated that there had been no method of tracking or identifying abandoned cables in the cable trays when the original cable number was utilized to identify the revised

! cable. This could result in an unidentified violation of j procedure SQN-DC-Y-11.3, " Power Control and Signal Cables for Use in Category 1 Structure", Section 6.2.2, " Cable Tray l Loadins", which limits the cable tray loading to 30-percent fill i

for power and 60 percent fill for control and signal cables.

This could result in insufficiwnt dorating of power cable because of increased thermal consideration and exceeding of i

cable tray hanger limits because of increased cable weight. SCR SQNEEB 8620 R1 (reference 75) addressed the root cause and actions to prevent recurrence.

• Page 15 of 39

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Tha SQN Nuclear Parformance Plan (July 17, 1986) was reviewid.

I The SQN activities list contains a restart item with the action list unique identification number 0172:

i Resolve NSRS open item R-84-19-WBN-07 by. developing cable tray fill level criteria and performing a QC inspection.

I 5. Completion Status i Satisfactory completion of restart item 0172 should resolve the i j NSRS recommendations. Until corrective action to restart item 0172 is completed this NSRS item will remain open.  ;

D. (Closed) R-85-02-SQN/WBN-02, " Maintenance. Operatina, and Test Instructions" f 1. Background l .

} In the original review, NSRS concluded that SQN instructions

were not sufficiently clear and did not include sufficient 4

precautions and other measures to preclude degradation of the high-pressure seals. NSRS recommended changes to several

, instructions to fix those problems and also recommended that the primary system pressure not be increased while the thimble tubes l are disconnected from the overl.ead path transfer system. The j latter recommendation was intended to preclude the ejection of a thimble tube in the event of failure of a high pressure seal.

3 During the February 1986 folinw-up, NSRS in report R-86-01-SQM

determined that several recommendations had been incorporated

and others were being addressed in proposed procedure 1

revisions. The item had been left open pending completion of  ;

the following actions.  ;

i j a. Issuance of the proposed MI-1.11. " Thimble Tube j Installation" (reference 36), which will replace Special

!. - Maintenance Instruction (SMI)-1-94-5 and addresses several of the original recommendations.

b. Issuance of the proposed revision to SMI-0-94-3 l,

(reference 33) that requires the use of an appropriate thread lubricant and cautions against allowing fitting bodies to turn.

c. Further revision of SMI-0-94-3 to include a precaution against working on the high-pressure seals when the primary l system is pressurized above atmospheric, i

! d. Revision of appropriate instructions to preclude pressurizing the primary system with the thimble tubes i

disconnected from the overhead path transfer system or at least preclude an1 work on the seals with the primary system -

pressurized above atmospheric and the thimble tubes disconnected from the overhead path transfer system.

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! Page 16 of 39

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2. Ccerective Action Taken By memorandum dated April 18, 1986 (reference 3), SQN stated i that MI-1.11 had been issued and that SMI-0-94-3 would be I

revised by May 1, 1986. MI-1.9 (reference 34) and MI-1.10 (reference 35), along with the issuance of MI-1.11 and revision to SMI-0-94-3, would provide assurance that any work is precluded on the seals with the primary system pressurized above '

atmospheric and the thimble tubes disconnected from the overhead path transfer system.

3. Specific Verification Methodology The original report details were reviewed to determine all of i

the specific recommendations made following the thimble tube event. Instructions SMI-0-94-3, MI-1.9, -1.10, and 1.11 were reviewed and compared with the original review recommendations.  ;

Discussions were also held with cognizant personnel to determine. .

the level of understanding and caution gained as a result of the

accident. ,

! 4. Verification Analysis i SMI-0-94-3, Revision 2, " Seal Table High Pressure Seal Repair,"

was issued June 27, 1986. The work instructions include i prerequisites such as reactor shutdown, plant cooled down and ,

j depressurized, and tne reactor coolant level drained down below  !

! the level of the seal table. The work instructions are clear

! and the use of thread lubricant is specified. TVA QC hold i points are specified at critical parts of the repair procedure. i

MI-1.9, "Botton Mounted Instrument Thimble Tube Retraction and l l Reinsertion," revisica 7, was issued September 9, 1985. It had ,

j been revised to incorporate changes in response to NSRS Report 4

R-85-02-SQN/WBN in accordance with Corrective Action Tracking System (CATS) No. 85219. This instruction provides for QC hold points; inspection of the threads of high pressure fittings for signs of galling, wearing, or cross threading, and application ',

of specified lubricant to the threads.

i

MI-1.10. "Incore Flux Thimble Cleaning and Lubrication," ,

revision 3, was issued September 9, 1985. It had been revised i to incorporate changes in response to NSRS report l R-85-02-SQN/WBN in accordance with CATS No. 85219. The

, instruction warns that no maintenance is to be conducted on the 1 high-pressure fittings while the primary system is pressurized  !

above atmospheric or head pressure from the guide tube.  ;

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MI-1.11. " Thimble Tubs Inst 331stion," ravision 0, wIs issurd July 10, 1986. This new instruction replaced SMI-1-94-5. The prerequisites include that the reactor water level is below the vessel flange. The work instructions are clear, and provisions for verification by the cognizant engineer and QC inspector are included. Discussions with cognizant personnel confirm that they are well aware of the former hazards associated with the thimble tube event. It is concluded that from this review by ECTG that the NSRS recommendation has been adequately resolved.

5. Completion Status No further action is needed by SQN. This item is closed.

E. (0 pen) I-85-06-WBN.01, The Adequacy of the Dispositions For Identified Cable Bend Radius Problems

1. Background In the original review at Watts Bar, NSRS investigators determined that there was not sufficient manufacturers' documentation / justification / test data nor OE engineering basis to substantiate the final dispositions documented for NCRs or establishment of RT ain values for multiconductor cables. The available information, as well as Office of Engineering Design and Construction (OEDC)-developed acceptance criteria for the sampling program used to justify the as-installed conditions,

~

1acked engineering support. The report stated that OC was permitted by OE to bend the cables permanently to a radius of one-half of values which in 1981 were recognized to be an industry standard for minimum values to which a cable can be I

permanently bent.

In 1982. NRC questioned the acceptability of those cables that had been permanently bent to one-half of the standard values at WBN. OE and OC developed a plan of action which for all practical purposes accepted the installed condition of the cables although a number of design standard drawings and a construction specification were revised to delete the allowable excessive bend radius and include industry standards criteria.

These documents were released in late 1983 and early 1984. Many appendix R cables, as well as other cables for WBN unit 1, i pulled and accepted in accordance with ONP procedures may have l

been installed with excessive bends. The ONP procedure was not corrected to delete the one-half radius criterion until March 1985. Although OE and OC's final disposition accepted the i installed condition of these cables at WBN based on "0E/0C l developed" values, NSRS questioned the validity of these l

values.

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Page 18 of 39 i

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Ths NSRS report stated that banding cables to less than the industry standard values had the potential of irreversible I adverse affects on the cable spiral wound shleiding and cable-insulation resulting in reduction of the quellfled cable life f under normal conditions with probable accelerated affects because of harsh environment. Violation of industry standaros in installation of these cables could potentially invalidate the l environmental qualification certification of the cable. I NSRS recommended that OE should establish values for RTain, as f well as RPain, that are fully supported by manufacturers' test  ;

data. A sample size should be selected at WBN unit 1 for il class 1E cables, then a controlled inspection be conducted based on fully supported RPain and RTmin values, and the result used to support the adequacy of as-installed cables. If evaluation of the results of the sampling program failed to adequately justify the as-installed conditions, then the development of a formal in-service / inspection program should be -

considered.

OE's position (references 43 and 45) was as follows:

RPain standards and values are discussed in the following documents: Electrical Design Standard DS-E12.1.5, "Ninimum Radii for Field Installed Insulated Cables Rated 15,000 Volts 1 and Less," was issued on September 20, 1983, and was based on l sound engineering data, both obtained in a scientific manner.

from cable vendors and formulated by TVA engineering. The installation factors for cable bending radii shown in Table 1 of the standard are based on ICEA standards 8-19-51, 8-66-524 and 8-68-516. The factors are in strict compliance with ICEA with no relaxation for the minimum pulling radius. The purpose of establishing an RT ain is to avoid subjecting the cable to a i condition which could negate the vendor's qualification.

Bending to a radius less than R Tain could cause premature failure of the insulation because of overstressing in the area of the bend.

Documentation for the reduction in the minimum cable bend radius factor for shleided power cables is contained in a DNE calculation (reference 47). For multiconductor cables, the individual conductor outside diameter may be used to determine the minimum training radius values for 600-volt or less cable at any point along their route. This is acceptable because the outer jacket of the multiconductor cable is sacrificial and is not required for the cable to function electrically or for environmental qualifications.

Page 19 of 39

OE stcted that a stcaring sample was taksn during 1983 to identify the cable installations most likely to have bend radius problems. Condulets were identified as the worst cases. A total inspection was conducted of the VS-level cable trays at Watts Bar. Approximately 80 to 90 possible cable bend radius violations on class IE cables were found in this inspection i (reference 43). The OEDC task team's computer program based on

! condulet vendor's minimum condulet bend radius, the outer diameter of cable, and the established minimum bend radius of the cable was run on all cables in conduit at WBN. The run identified 175 " conditional" condulet which posed possible bend radius violations. A visual inspection walkdown of all .

" conditional" condulets revealed only nine violations which were nonconformed and dispositioned based on a letter from Okonite Company which allowed a bend radius of 3 x OD. This factor allowed a minimum cable bend radius of 3.153 inches in condulets 1 which only had possible minimum radii of 4 inches. Therefore, the cables could not possibly have been bent beyond their vendor-established minimum bend radii. -

2. Corrective Action Taken An OE calculation (reference 95) was issued August 6, 1985.

4 This calculation provided traceability of manufacturer's i information and further justification for the use of that information to relax the minimum V5 level cable bend radius factor from 12 times overall diameter to the factors presently-shown in Table 2 and 3 of DS-E12.1.5 (reference 43).

i OE further stated that Wyle Laboratories was currently

conducting environmental qualification testing of TVA cable samples. These samples were being tested on mandrels which l duplicated minimum bend radii based on the factors set forth in

Table 1 of DS-E12.1.5. Bend radius documentation based on these i results was to be issued April 1986 (reference 43).

f A request for funding was made by the Director of Engineering ,

j and Technical Services on January 15, 1986 (reference 46). ,

The scope of work was as follows:

I

1. Review and analyze the rationale developed by EEB for

, evaluating operation and maintenance data and surveillance l test data on existing cable installations to establish the i adequacy of cable subjected to some installation abuse with

{' respect to less-than-specified minimum cable bend radii, excessive sidewall pressures, and excessive pulling tension. l

2. Assess and evaluate the surveillance testing program and f consider potential improvements in the particular tests conducted and data collected.

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Page 20 of 39

_ . . _ . - _ _ _ _ _ - - ~ _ . . _ .

J Provida cceporato endorstment of ths TVA practicas and 3.

processes and implement this short-term program to verify installed cable adequacy.

4. Provide any recommendations for a longer term program to participate with other industry segments to collect, share, i and evaluate data relevant to establishment of adequacy of cable _ installed with some deviation of specified minimum bending radii, maximum allowable sidewall pressure, and maximum allowable pulling force.

3. Specific Verification Nethodology Report I-85-06-WB (reference 6), SQN Generic Concern Task Force 4

Report (reference 66), and all subsequent correspondence related to minimum bend radius were reviewed. Discussions were held 4 -

with the NSRS investigator and design supervisors and staff in

the Electrical Engineering Branch (EEB) and the SQN Design i -

Project.

4. Verification Analysis A memorandum dated June 11, 1986 (reference 47), reviewed the i class IE cable bend radius issues and stated in part:

i Based on recent studies and by utilizing recognized cable

! properties, we believe that the cable bend radii issue

< should not affect restart. The basis for our approach stems I from the fact that the property " retention of elongation" of cable insulation is the accepted method of measuring (and

establishing) the end of life in cables. Typical cable

insulation materials have unaged elongation properties which vary from 250 to 530 percent. It is accepted that after

! thermal aging, mechanical stress, radiation, and postulated

accident environment, elongation properties degrade to a

] degree depending on the specific insulation material.

j A cable bent to a given radius produces an elongation stress on the outer surface of its insulation. This can ba calculated. A cable having a bending radius of one time its

, diameter hcs an elongation stress of 33 percent. This stress has been shown to have a minor effect on the qualified life of the cable.

i

! Our analysis will show that for the worst-case (minimum

allowable) cable bending radius (one time its outer diameter), the resulting stress from the bend radius will not affect the qualified life of the cable. We strongly believe that this analysis will conclusively show that cables have a qualified life of at least 30 years with the

vast majority qualified for 40 years. )

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A mmrorandum dated Ssptsmbsr 2,1986 (refercnce 85) stated in part:

The EEB has evaluated the adequacy of the bend radius to which class 1E cables were installed. EEB's final report documenting the evaluation and providing conclusions and recommendations will be issued in September (1986).

i During this review it was determined that the report had not yet been completed.

The NRC in a memorandum dated August.4, 1986 (references 93 1 andM94) requested information concerning cable pulling and cable

[ bending radii at SQN. An NRC team attended a meeting at SQN on i September 24, 1986, to discuss cable issues. No report has been issued at this time. One attendee stated that all cable

- problems appeared to be resolved except the medium voltage shielded cable. .

l S. Completion Status While actions appear to have been taken to justify the reduced i bend radius factor, the engineering tests / reviews / analysis j appear to be fragmented. In addition, the original report j discusses the potential for irreversible adverse effects on j cable spiral wound shielding and cable insulation which could i result in reduction of qualified cable life. This did not i appear to be addressed in the correspondence reviewed.

Pending review of the NRC meeting details and resolution of the 7

above, this item remains open.

i F. (0 pen) I-85-06-WBN-02, The Adequacy of the Program For Cable Pulling

Activities -

1

1. Background

• f

, In the July 1985 report, NSRS concluded that OE and OC's established and documented program was inadequate to accomplish

, the cable pull activities. The past and present programs were 1 claimed to be inadequate, inconsistent, and in violation of the j accepted industry standards and practices. The report stated j that Construction specification G-38 had not defined the method i

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for calculating maximum allowable tension for multicable pulls

. nor had it finally established the method of pulling i multicables utilizing a break link. Also, the industry l considered the side wall pressure as a limiting factor for cable j pulling activities. Failure to ensure that maximum allowable tension, as well as sidewall pressure, was not exceeded in any cable pulling operation increased the potential for cable damage

^

(insulation degradation) which is not necessarily detected i .

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Page 22 of 39

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visibly ce through testing. Irpropar installaticn of cables could potentially invalidate the environmental qualification certification of the cable.

NSRS recommended that OE management should revise G-38 and G-40 to incorporate resolutions to the identified problems and

> subsequently WBN OE, OC, and ONP management should revise the relevant OE documents as well as OC and ONP procedures. The final adequacy of the_present cable' installation should then be evaluated according to revised acceptance criteria. A sampling -

approach should be considered. If evaluation of the result of the sampling program failed to adequately justify the as-installed conditions, then development of an in-service inspection program should be considered.

i On November 21, 1985, several WBN employee concerns related to cable pull noncompliances were transmitted by NSRS report I-85-466-WBN (reference 52). The report concluded that corrective action measures to correct report I-85-06-WB-02 would -

also resolve the referenced employee concerns.

OE's position was that any inadequacies, inconsistencies, or violations of accepted industry standards and practices in its past and present cable pulling program could, at most, result in 1 random cable failures. These random failures could not prevent j safe shutdown of the plant.

In memorandums dated July 8, 1985 (reference 42, 43), OE stated that'an upcoming revision of G-38 would form the basis for a complete response to NSRS item I-85-06-WBN-02. G-40 i (ref,erence 44) did not require revision. OE had contacted various utilities and architect-engineers in the spring of 1985

on cable pulling practices. The maximum allowable pulling force j for group cable pulls varied from 60 to 100 percent of the sum

! of the individual maximum allowable pulling forces. The survey l indicated that all of the utilities and A/E's used one break l link for cable pulling, instead of individual b eak links for l each cable.

G-38, revision 6 was issued September 15, 1985; it required cable sidewall pressure calculations to be performed before cable installation. OE stated that the adequacy of past installations (reference 44) was confirmed by Construction experience (reference 92). The justification for not using individual break links and the function of the outer jacket of multiconductor cables was discussed in a November 5, 1985 memorandum (reference 45).

. 2. Corrective Action Taken Page 23 of 39

l

. l A probles idsntification esport PIR WBNEEB8534 was issusd by OE, August 13, 1985, " Cable sidewall pressure calculations were not considered in the design process. .This condition was identified in the NSRS Report I-85-06-WBN. Construction specification G-38R5 did not address cable sidewall pressure."

An inquiry by OE of cable manufacturces (reference 48) was

-initiated in order to support OE's response to NSRS audit finding I-85-06-WBN-02. Letters dated July 26, 1985, were sent  !

to 30 cable manufacturers requesting information related to cable furnished to TVA on each contract by each manufacturer.

NCR OC NCR 6270 R0 was issued September 6, 1985. The NCR stated that cable insulation could fall or have shortened life if cable sidewall pressures had been exceeded during installation.

A request for funding (reference 46) was made by the Director of l Engineering and' Technical Services on January 15, 1986. .

3. Specific Verification Methodology For this follow-up action, report I-85-06-WBN was reviewed and discussions held with the NSRS investigator and others knowledgeable of cable handling activities. All subsequent correspondence, test reports, and the SQN Generic Concern Task Force reports, were examined. The status of the activity at SQN was determined from reviewing recent correspondence and discussions with design supervisors.

4. Verification Analysis Calculations were made by DNE on March 10, 1986 (reference 53),

to determine if cable sidewall pressures (SWP) were exceeded in conduits containing class 1E cables. SWP limits were exceeded in 12 of 82 conduits when compared with vendor data. A cable SWP testing program was to be initiated. '

Cable sidewall bearing pressure tests were made and reported June 11, 1986 (reference 47). The report in part states:

To evaluate the cables already installed in conduits and determine the impact of including SWP in the cable pull j tension limitation, an evaluation program was initiated in '

August 1985. This program surveyed WBN to determine the worst sections of conduit, with respect to cable pulling.

Using analytical methods and field data, SWP, which the class 1E cable was subjected to, was determined. Concurrent with this evaluation, a survey of the cable manufacturers on new SWP limits was made. Based on SWP limits obtained from cable manufacturers, 21 different cables in 12 conduits exceeded manufacturer's published SWF allowable values. The need for additional investigation was established. .

Page 24 of 39

TVA Central Laboratories perfor. sed extensive tests on 21 samples, in addition to representative s'amples selected from SQN, BFN, anu BLN nuclear power plants to include different cable types (power, control, signal and instrumentation, and coaxial), types of insulation, jacket materials and manufacturers. These tests established allowable SWP limits in excess of that required within margin. TVA test results are also consistent with EPRI Report No. EL-3333 where allowables were determined to be 4-5 times higher than previous manufacturers' limits.

The test performed in a fixture containing four 90*

horizontal bends was set up and cables were pulled through the conduit with tension forces from 2 to 12 times more than the recommended values in G-38. The cables were subjected to pulling tension values very near the ultimate breaking strength of cable. Each cable, after being pulled, was .

inspected, dimensioned, carefully stripped to examine individual conductors of multiconductor cable, and subjected to dielectric breakdown test. The dielectric breakdown values of the tensioned cable was compared with dielectric breakdown value of the virgir cable of same sample. The average dielectric breakdown Jalue of all the 32 cables ,

tested was within 20 percent of the average dielectric I breakdown value of the respective virgin cable sample, thus meeting the acceptance criteria set for the test per ASTM D 149. Further, none of the cables revealed any significant degradation of insulation. The results of the tests are presented in the ' Cable Sidewall Bearings Pressure Tests' report prepared by the Central Laboratories of TVA, dated May 30, 1986.

In a memorandum dated June 16, 1986, SQN Site Licensing stated that they transmitted information requested by NRC which included TVA Specification GCS-G 38, " Installing Insulated Cables Rated Up To 15,000 Volts (Revision 2 through 7),"

WBN-QCI-3.05, " Cable Installation (all revisions)," and WBN-QCP-3.05, "EQC Procedure For Cable Installation."

The NRC requested information regarding cable pulling (and cable bend radii) at SQN August 4 and 29, 1986 (references 93 and 94). NRC was at SQN on September 24, 1986, to discuss the cable question. Preliminary indications were that NRC was reasonably satisfied with the SQN cable pulling activities. No report on the official acceptance has been received.

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l S. Comp 1stica Status l It appears that adequate testing has been performed to verify the cable quality. Memorandums do not address whether or not an in-service inspection program has been considered or whether the tests performed adequately justify the as-installed conditions. l Pending statements to this effect this item remains open.

G. (0 pen) I-85-06-WBN-05, Undevelopment of Established and Documented Limited QA Program and Failure To Incorporate The Established Limited QA Program Into NOAN

1. Background i

In the original review at WBN, WSRS concluded that implementation of the limited QA program as established in Office of Engineering Design and Construction (OEDC) instruction OEDC-PRN-QAI-2 was inadequate. Also, the NQAM was determined to be inadequate in addressing the limited QA program as .

established by superseded OEDC-PRK-QAI-2 document. NSRS recommended that the Division of Quality Assurance (DQA) should identify the required "special controls" applicable to systems previously identified as " limited QA" in the NQAN.

Appropriate line organizations should incorporate those requirements into the DNE, DNC, and ONP procedures for implementation by the appropriate offices. In view of the fact that WBN had recognized security and fire protection as the only two systems for limited QA consideration, DNE should perform.an evaluation to ascertain the impact of not implementing the i

established " limited QA" program on the other systems identified in OEDC-PEN-QAI-W. The purpose of this evaluation should be to determine the extent and depth of the impact on the quality and '

reliability of " limited" QA features' contributions to the safety of the plant.

j

2. Corrective Action Taken i

The problem identified was for WBN, and responses to WBN are given in memorandum dated July 8, 1965. The preliminary

. response given April 28,_1985 (reference 60), was as follows: ,

The OEDC " Limited QA" program was established basically to

' define a management control scheme to control the design and a

construction of structures, systems, and components 4 important to the operation of TVA's nuclear plants. These structures, systems, and components--while important to the operation of the nuclear plant--are not classified as

" safety-related" and, therefore, are not subject to the

' management controls (i.e., QA program) mandated by 10 CFR 50, Appendix B.

I Page 26 of 39

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c.-r ~w y-ww -&-- , . - - + - -t c-e---yr-2+-vw-= &Or-

Bafora issuance of the NQAM, a careful chsek was mids to ensure that requirements previously established by the manuals to be

-canceled by the NQAM (i.e., OEDC-PRM, OP-QAM, and IPM) were carried forward in the NQAM. During this check, it was revealed that the NQAM did not incorporate the " limited QA" program requirements of OEDC-PRM-QAI-2. In addition, it was determined that the manual did not address the former requirements of the OP-QAM pertaining to management controls for certain important operations phase activities not subject to 10 CFR 50, appendix B. Because of time constraints associated with issuance of the NQAM, a decision was made to reestablish and delineate the management controls associated with the operations phase activities mentioned above. These controls are currently specified in NQAM, Part I, Section 1.3. "QA Program Requirements

- NRC - Regulated Programs Which are Conditions of the Operation License." With respect to the requirements previously established by 2QAI-2, the decision was made to not incorporate these existing requirements into the NQAM before its issuance on December 31, 1984. This decision was, for the most part, based -

• on the time constraints mentioned above and the fact that 2QAI-2 did not clearly delineate the management central schemes (i.e.,

QA program controls) that should be applied to the various structures, systems, and components within its scope.

In light of the division pertaining to 2QAI-2, DQA scheduled, tracked, and proceeded to develop an NQAM procedure that would clearly establish the management controls applicable to the design, construction, and operation of special features such as those formerly covered by 2QAI-2. Furthermore, this document was developed in concert with the Q-list specification to ensure that the identified management controls were clearly linked to the structures, systems, and components to which it applied. In addition. DNE investigation has determined that the NSRS staff finding at WBN was based upon the Limited QA Program for radwaste systems. The Limited QA Program for fire protection and security was reviewed by NSRS and found to be acceptable.

Several DNE documents and TVA documents define limited QA and' identify the components covered by the Limited QA Program.

OEDC-PRM-QAI-2 established the basic requirements of the Limited QA Program (QA(L)). This QAI, now titled, "QA Requirements -

NRC - Regulated Programs which are Conditions of the Operating License," was written to implement the requirements of NRC Regulatory Guide 1.143. The QAI required DNE (then OEDC) to:

A. Issue drawings, specifications, and purchase documents to define the requirements of QA(L).

I B. Maintain a QA(L) program.

C. Develop and maintain an auditable QA record system for structures, systems, and components subject to QA(L).

i Page 27 of 39

Since 1980, 75 ECNS have besn issutd for rzdwaste systers on WBN, 44 of which identify QA on the ECN cover sheet. The 31 remaining ECNs did not identify QA and should not have because of the scope of the ECN. For example, some of these 31 covered documentation corrections and others involved changing set points but no physical changes requiring QA.

A review of all ECNs related to the radwaste system for SQN is not being performed. This decision is based on no problems being found during the WBN ECN review. SQN utilizes similar ECN procedures as WBN.

TVA is committed to partially conform to the regulatory guide on SQN as follows:

Original equipment - TVA was not committed to comply with RG 1.143. However, major modifications to existing radwaste systems and new radwaste systems are committed to comply.

To meet requirements, DNE has issued general construction -

specifications (G-specs) for the various systems listed in the QAI. The systems and ths associated G-specs are listed on the attachment. These specs identify to ONP the QA(L) requirements for item control, fabrication and installation inspections, special construction processes control, testing control, nonconformance control and reporting, document control, and records control.

DNE further defines the limits of the QA(L) program by marking the boundaries on the design drawings or by adding notes to the drawings. Also, the G-spec is usually referenced on the drawing, and it is the responsibility of the system engineer to l

ensure that the ECN and design criteria accurately depict the QA requirements so the designer, constructor, and procurer are left l with little doubt as to what is required.

The procedures that control the issue of design, drawings meet i the requirements of 10 CFR 50 appendix B. DNE makes no

distinction in its handling of QA, QA(L), or non-QA drawings.

Therefore, QA(L) drawings and designs exceed the minimum QA

, requirements for QA(L). The proposed NQAM document has

, presently been submitted for review by responsible line organizations and will be issued in the near future.

III. Specific Verification Methcdology i

The NQAM procedure was reviewed. Discussions were held with Quality

! Assurance, Construction - Modifications, Design supervisors and staff l regarding the status of SQN limited QA programs, i

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4. Varification Analysis NQAM Part I, Section 1.3, " Limited Quality Assurance Program Requirements" was issued June 27, 1986. The procedure requirements were to be fully implemented lat SQN within 90 days, i.e., September 25, 1986. Programs to which the procedure applies are identified and audits are required for assessment of program compliance and effectiveness. A revision to this procedure with more specific details had been drafted and had been expected to be issued mid-September 1986. QA stated that they had had meetings with the Plant Manager, and it was agreed that a new plant procedure describing the limited QA program and l which would consolidate the existing fragmented program was to

be issued after the NQAM was revised. Design supervision stated i that a limited QA program summary referred to in the memorandum l dated April 28, 1986, was in the completion stage. No problems requiring resolution were identified la these discussions.

During this last review it was determined by ECTG that the plant procedure had not been issued. -

5. Completion Status Completion of this item is needed by SQN. Pending review of the plant procedure and revised NQAM when issued, this item remains open.

H. (Closed) R-86-01-SQN-01 Improvements in Overall As Low As Reasonabir Achievable (ALARA) Program

1. Background In the original review reported in I-84-12-SQN-13 NSRS recommended that it be emphasized to plant employees that preplanning as specified in plant instruction RCI-10 must be accomplished. During the February 1986 follow-up, NSRS 4 determined that sufficient corrective action had been taken to

satisfy the NSRS open item. However, discussions with the i Hecith Physics section staff, ALARA engineer, and review of

documentation led the NSRS reviewer to conclude that additional

specific actions should be taken to strengthen the ALARA program at SQN. The following suggestions were made

a. ALARA Engineer Staffing Support - Determine the appropriate Health Physics (HP) technician staffing level required to effectively perform ALARA duties during normal and off-normal working hours. This determination needs to consider all plant functions which require ALARA considerations; such as maintenance, operations, test, 4

modifications, outage planning, design, and site services.

i A job-task analysis could be used to determine an effective staffing level.

i i

Page 29 of 39

l b. ALARA Rwview Committse - Establish an ALARA review committee composed of members from the major functional areas with the l responsibility for overall coordination of the ALARA program. Specific functions would include:

(1) Review exposure reduction plans for specific jobs with

.4posure estimates greater than 25 man-rem.

(2) Direct the implementation of approved ALARA suggestions.

(3) Review planning schedules.

9 (4) Review specific and timely ALARA problems; such as, '

f reports of unnecessary loitering in dose areas.

(5) Review personnel contamination reports.

(6) Review corrective action on delinquent post job ALARA reports.

(7) Review status of ALARA projects.

1

! The ALARA Committee composition and responsibilities should i be incorporated into a plant instruction, e.g., a SQN Standard Practice or Radiological Control Instruction.

c. ALARA Employee Suggestion Program - Increase employee i participation in the ALARA employee suggestion program.

Adoption of an awards program could be a way to increase participation.

d. Department ALARA Coordinators - ALARA coordinators should be i assigned to all the functional organizations, e.g.,

modifications, operations, maintenance, test, design, and site services, to provide these groups with the expertise i necessary to support all aspects of the ALARA program. This would be an expansion of the current plans of the HP Section

to assign an N-3 HP to assist planners with ALARA.

e. Training - An ALARA training program should be prepared and given to those individuals directly responsible for the

. ALARA plant efforts, e.g. , ALARA Committee members, i department coordinators, plus those individuals responsible for preparation of ALARA preplans and post plans. The training program should be extensive and incorporate as basic elements: the physics of radiation; fundamentals of radiation attenuation; types of radiation sources; review of industry experience; methods to reduce exposure, e.g.,

changing test frequency or time of test, changing preventative maintenance frequency or time of maintenance, relocate components with high failure rates to lower radiation fields, and/or permanent shielding, flushing systems, etc.

l Page 30 of 39 l

2. Corrective Action Toksn By memorandum dated April 18, 1986 (reference 3) -SQN made the following response:

"a. The site HP section will be reorganized in the near future.

Staffing levels and job functions are a major part of this reorganization. This item will be completed and the new organization in place June 1, 1986.

b. SQN site management does not endorse the need for a formal ALARA committee at this time. Each manager, supervisor, and site worker is charged with the responsibility of implementing ALARA practices and philosophy in his daily activities. The site ALARA goals are achieved through the efforts of all site personnel.

c. The site HP section will do more to promote the ALARA .

suggestion program. Increased communication and feedback on man-rem savings should be made available to all site employees. A form of reward is under consideration by the HP supervisor at this time.

d. The HP section has made a shift supervisor available to the-maintenance planners; however, refer to the response t'o item b for management direction on separate committees and coordinators.

e. An ALARA training program that was previously developed is ,

to be revised and presented in a manner that will complement the responses to b, e, and d.

3. Specific Verification Methodology Since the corrective action to the original recommendation had )

been accomplished, this item was essentially c16 sed. However, -

the additional recommendations and responses were reviewed to see whether the recommendations were reasonable and whether the i plant had taken reasonable steps to improve the program l

further. Although five suggestions had been made, it was not considered essential that the plant had fulfilled all five. The intention was to determine that improvements had been made.- The

, suggestions had been arrived at by discussions with SQN 3

personnel combined with information contained in the following documents:

l l Institute of Nuclear Power Operations (INPO)-Operational l Ruperience Note REN/0E -8A, "A Good Practica For the ALARA Program" i

Quality Audit Branch Report number OSS-A-85-0016 RC-14, " Radiation Work Permit Program"

• l RCI-10, "Minisizing Occupational Radiation Exposure" l Page 31 of 39 l

l l

. _ - - . - , _ _ _ _ . _ , , , _ , _ . - . . . . _ _ _ _ . . . _ _ _~ . . . _ _ . . . _ . - , _ . . _

(

1 l -

Furthsr d2 tails supporting the raccamendaticns ers givsn in report R-86-01-SQN, pages 78-82.

4. -Verification Analysis

[ The plant was responsive to an HP reorganization, training need.

and HP shift supervisor availability to maintenance planners, but did not endorse a formal ALARA committee. However, l corporate changes identified in the Nuclear Performance Plan, 3 Volume II, have given improvements to the program and increased corporate involvement with the objectives of minimizing employee exposure and improve regulatory performance which include the i following:

J O

• A Radiological Assessor position has been established on the 1 Site Director's Staff. This function provides progra.amatic overview of the SQN Radiological Controls Program. The 4

Radiological Assessor interfaces with the Manager of Radiological Controls within the Division of Nuclear -

1 Services and the Manager of the Radiological Controls i Section at SQN. ,

4

~

f

• The Manager of the Radiological Control Section (formerly the Health Physics Section) now reports directly to the SQN j Plant Manager. Moving the reporting relationship higher in h the organization will provide plant management more -

L immediate access to radiological control performance data h and result in earlier identification and correction of problems.

SQN has implemented a contamination area control program to minimize contamination areas in nonoutage periods. This program is directed through biweekly meetings held by the j l Plant Manager with Operations, Maintenance, and Radiological i Control representatives. l 4

• i In March 1986, an HP shift supervisor was selected and assigned to work with the Maintenance Department to help coordinate radiological area work with Radiological Control 1

Section and crafts. This individual will provide special

, assistance and training to the craftsmen in areas such as frisking, radiation work permits, radiological survey maps, ALARA, and good HP work practices. Participation in maintenance planning / scheduling meetings provides Radiological Control Section input into the work pla.ns and schedules.

The SQN Radiological Control Section has implemented a management position of HP Training Officer. This person has the responsibility to ensure HP technicians, both American National Standards Institute qualified and new trainees. l l meet and complete their training and retraining requirements. '

5. Completion Status p .

1 No further action is needed by SQN. This item is closed.

i Page 32 of 39

f References

1. Memorandum R. K. Seiberling to R. L. Gridley, dated May 1, 1986, l

" Schedule of TVA/NRC Activities Prior to Sequoyah (SQN) Restart" (Q01 860501 223)

2. Memorandum E. Gray Beasley to J. P. Vineyard, dated February 14, 1986, "Sequoyah Nuclear Plant - QA Review of Operational Readiness - NSRS Open Items - Surveillance Report Number S86-10" (B05 86 0214 001)

3. Memorandum H. L. Abercrombie to R. K. Seiberling, dated April 18, 1986, "SQN-NSRS Follow-up Review of Open Items From Previous NSRS Reviews and Investigations" (SS3 860416 982)

4. Memorandum K. W. Whitt to H. L. Abercrombie dated February 14, 1986, "Open NSRS Items From Reviews And Investigations Requiring Resolution Prior To SQN Startup" (S06 860127 800)

5. NSRS report number R-84-19-WBN,' "NSRS Assessment of Results of the Black and Vestch Independent Design Review of the Watts Bar Nuclear Plant 2 Auxiliary Feedwater System," dated July 5, 1984 (GNS 840705 054)

6. NSRS report number I-85-06-WBN, "WBN - Investigation of an Employee Concern Regarding Cable Routing, Installation, and Inspection," dated July 9, 1985 (QO1 850709 050)

7. NSRS report number R-86-01-SQN, "SQN-NSRS Follow-up Review of Open Itenc From Previous NSRS Reviews and Investigations," dated March 21, 1986 (Q01 860321 051)

8. Memorandum R. L. Gridley to B. Y. Youngblood dated April 4, 1986 -

third report on resolution of safety-related NSRS reports (L44 860404 810)

9. Sequoyah Engineering Project (SQEP) Administrative Instructions (AI)

SQEP-AI-8, Revision 0, dated September 25, 1985, " Drawing And Reproduction"

10. SQEP-AI-8A, Revision 0, dated November 15, 1985, " Program Documentation"

11. SQ$P-8 Revision 3, dated June 28, 1986, " Packaging And Controlling of Walkdown/ Test Documentation"

12. SQEP-12 Revision 2, dated July 23, 1986, " Procedure For Evaluating Change Notice and Field Change Notice Documents"

~~~~

13. SQEP-16. Revision 0, dated June 21, 1986 " Procedure For System Evaluation and Development of System Evaluation Report"

14. Gilbert / Commonwealth. Incorporation report number 2600 dated October 1985. " Assessment of the Design Control Program For the Sequoyah Nuclear Plant" *

15. SQN-DEM (EA 85-49), " Gilbert - Commonwealth Design Control Survey,"

dated November 7, 1985 (L44 851107 804)

Page 33 of 39 e --

16. Memorandum J. P. Vineyard to H. B. Rankin, dated March 4, 1986, "SQN-Technical Review of Unimplemented and Partially Implemented Engineering Change Notices For Selected Systems" (B25 860304 010)

17. Memorandum M. T. Torney to Those listed, dated August 7, 1986, "SQN-Design Control Programmatic Requirements" (B25 860807 010)

18. Station Auxiliary Power System Drawing number 15E500-1, Revision 5

19. 12 V AC and 125 DC Vital Plant Control Power System Drawing Number 45N700-1, Revision 4

20. Memorandum H. G. Parris to H. N. Culver dated July 31, 1984, "NSRS Assessment of the Results of the Black and Veatch Independent Design Review of WBN Auxiliary Feedwater System - NSRS Report R-84-19-WBN" (EDC 84081 601)

21. Memorandum H. N. Culver to R. M. Pierce dated September 5, 1984, "NSRS -

Assessment of the Results of the Black and Veatch Independent Design Review of WBN Auxiliary Feedwater System - NSRS Report R-84-19-WBN" (GNS 840906 101)

22. Memorandum T. G. Campbell to J. P. Vineyard dated September 12, 1984, "SQN - Funding to Address Applicability of Black and Veatch Findings" (SQP-840913 002)

23. Memorandum J. E. Huston to W. C. Drotleff dated May 14, 1986, "SQN -

Operational Readiness Review" (LO3 960513 839)

24. Sequoyah Restart Unimplemented Design Item Evaluation. Action Item 299 dated March 10, 1986, "NSRS Open Items R-84-19-WBN-01" (B25 860311 002)

25. Memorandum R. W. Cantrell to H. N. Culver dated October 1, 1984, "BFN, SQN, and BLN - NSRS Report R-84-16-WBN-Item R-84-19-WBN-01" (QMS 840928 201) ,

26. NSRS Report R-84-26-WBN, "NSRS Routine Review of the Response to NSRS Recommendations Identified in NSRS Report R-84-19-WBN," dated September 5, 1984 (GNS 840906 101)

27. Memorandum R. M. Pierce to H. N. Culver dated October 3, 1984, "WBN -

NSRS Routine Review of the Respcase to NSRS Report R-84-18-WBN" (WBP 841010 021)

28. Memorandum H. C. Rutherford to Electrical Branch Files, dated May 14, 1985 "WBN - Assessment of the Computerized System for Routing Electrical Cables" (B43 850514 926)

29. Memorandum G. R. Owens to NSRS Files dated August 21, 1985, "WBN -

Status and Disposition of NSRS Concerns R-84-19-WBN-06 and -07 Resulting From Independent Assessment of the B&V Review" (QO1 850913 053)'

Page 34 of 39

. l

30. Memorandum K. W. Whitt to H. G. Parris dated September 13, 1985, "WBN -

Disposition of NSRS Concerns R-84-19-WBN-6 and -7 Resulting from Independent Assessment of the Black and Veatch Review" (Q01 850913 052)

31. Memorandum W. C. Drotleff to R. K. Seiberling dated May 70, 1986, "WBN -

Disposition of NSRS Concerns R-84-19-WBN-6 and -7 Resulting from Independent Assessment of the Black and Veatch " view" (B26 860502 010)

32. ;3RS Report R-85-02-SQN/WBN, "Sequoyah and Watts Bar Nuclear Plants Spet.'al Review of Manufacturer / identified potential misapplication of Swageiock tube fittings at Westinghouse Reactor Seal Tables, " dated March 15, 1985 (QOI 850325 051)

33. Special Maintenance Instruction SMI-0-94-3, " Seal Table High Pressure Seal Repair," Revision 2, dated June 27, 1986

34. Maintenance Instruction MI-1.9, Revision 7. " Bottom Mounted Instrument Thimble Tube Retraction and Reinsertion," dated September 9, 1985 -

35. Maintenance Instruction MI-1.10, Revision 3, "Incore Flux Thimble Cleaning and Lubrication," dated September 9, 1985

36. Maintenance Instruction MI-1.11 Revision 0, " Thimble Tube Installation," dated July 10, 1986

37. Mechanical Maintenance Sectico Instruction Letter HMSIL-A36 " Common Mode Failure - Maintenance Initiated," Revision 4, dated March 19, 1986

38. Nuclear Quality Assurance Manual Part III, Section 7.3, Revision 0, dated June 20, 1986, " Common Mode Failures, Maintenance-Initiated"

39. SQN Survey Sheet 4a-86-A, "QA Maintenance Surveillance"

40. SQN Surveillance Brief Report number 13 , " Post Maintenance / Modification Test Surveillance" ,

41. Memorandum titla " Watts Bar Nuclear Plant - NSRS Investigation of an Employee Concern Regarding Cable Routing Installation and Inspection Practices - NSRS Report I-85-06-WBN"

42. Memorandum J. C. Standifer to R. M. Pierce dated July 8, 1985, (B26 850708 023)

43. Memorandum R. M. Pierce to K. W. Whitt dated July 8, 1985 (QO1 850708 604)

44. Memorandum J. W. Coan to R. M. Pierce dated August 22, 1985 (B26 850822 003)

45. Memorandum R. W. Cantrell to C. C. Mason dated November 5, 1985 (B43 851105 915) .

46. Memorandum G. F. Dilworth to John Hutton dated January 15, 1986, t " Request For Funding of Short-Term Program to Evaluate Adequacy of l Installed Class IE-Cables" (B49 860115 002) l Page 35 of 39

n i

47. Mem:rendum W. S. R:ughlsy to J. A. Kirksbo datsd Juns 11, 1986, " Class i IE Cable Bend Radius"  ;

48. Memorandum J. W. Coan to Guenter Wadewitz dated August 29, 1985, "WBN -

Insulated Conductor Vendor Data Related To Sidewall Pressure Issue - i Class IE - NSRS Audit Finding I-85-06-WBN-02" (B43 85 0829 912) l

49. '

Memorandum F. W. Chandler to Those listed dated September 6, 1985,

" Potential Generic Condition Evaluation" (B43 850909 934) l

50. Memorandum F. W. Chandler to J. A. Raulston dated September 23, 1985, "WBN - Cable Sidewall Pressure Calculations - 50.55(e) Item" l

51. Memorandum J. W. Coan to Guenter Wadewitz dated October 7, 1985, "WBN -

Insulated Conductoe Vendor Data Related to Side wall Pressure Issue -  !

Class 1E - NSRS Audit Finding (-85-06-WBN-02)" (B43 851011 915)  !

52. Employee Concern Investigation Report Transmittal dated . l November 21, 1985, " Cable Pull Noncompliances" (I-85-466-WBN)

{

53. OE Calculations " Sidewall Pressures of Class 1E Cables in Conduits" (B43 860310 936)

54. Memorandum N. A. Taff to W. S. Raughley dated June 4, 1986, " Cable Sidewall Bearing Pressure Tests" (B12 860604 001) e

55. W. C. Drotleff to R. K. Seiberling dated June 18, 1986, " Watts Bar Nuclear Plant - NSRS Investigation of an Employee Concern Regarding Cable Routing, Installation, and Inspection Practices - NSRS Report I-85-06-WBN" (843 860609 927)

56. R. K. Seiberling to Richard P. Denise dated June 25, 1986 (QO1 860623 262) >

57. Memorandum R. M. Pierce to K. W. Whitt dated July 8, 1985. "WBN - NSRS Investigation of An Employee Concern Regarding Cable Routing. -

Installation, and Inspection Practices - NSRS Report I-85-06-WBN" (QO1 850708 604)

58. Memorandum J. C. Standifer to W. R. Brown dated December 20, 1985, "Wetts Bar Nuclear Plants Units 1 and 2 - Employee Concern - Limited QA Program" (B45 851220 257)

59. Memorandum, W. R. Brown to K. W. Whitt dated February 10, 1986, "NSRS Investigation of An Employee Concern Regarding Cable Routing.

Installation, and Inspection Practices - NSRS Report I-85-06-WBN" (F01 860210 601)

60. Memorandum D. W. Wilson to Those listed dated April 28, 1986, "SQN -

NSRS Open Item I-85-06-WBN Limited QA Program - Restart Item" (B25 860428 023)

61. Nuclear Quality Assurance Manual Part I, Section 1.3, Revision 0, dated June 27, 1986, " Limited Quality Assurance Program Requirements" (L16 860611 001)

Page 36 of 39

~

62. NSRS Report I-84-12-SQN, "SQN-NSRS Investigation of Incident of Incore Instrument Thimble Tube Ejection Accident on April 19, 1984," dated August 1, 1984 (GFN 84080)(050)

63. Nuclear Performance Plan, Volume 2, Revision dated July 22, 1986 Section 1.2.3, " Radiological Control"

64. NSRS Review of Generic Concern Issue Report I-86-251-SQN, " Electrical Cables" dated February 18, 1986

65. Sequoyah Nuclear Plant Generic Concern Task Force Concern WI-85-100-011

" Cable Tray Fill Criteria," dated June 6, 1986

66. Sequoyah Nuclear Plant Generic Concern Task Force EX-85-076-003 through IN-86-266-109 (26 concerns), "Overtensioning and Minimum Bend Radius Violations of Cables Due to Improper Cable Installation Methods," dated June 27, 1986

67. Watts Bar Nuclear Plant Employee Concerns Task Group Construction C010900 " Cable," dated August 7, 1986 - supplement for Sequoyah Nuclear Plant

68. Memorandum J. P. Vineyard to C. C. Mason dated June 27, 1984, "SQN -

Funding to Address Applicability of Black and Vestch Findings,"

(PWP 840627 004)

69. Memorandum J. P. Vineyard to Those listed, dated March 4, 1986, "SQN-Office of Engineering (OE) Operational Readiness Plan,"

(825 860304 001)

70. Memorandum E. Gray Beasley to J. C. Standifer, dated September 17, 1984 l (QMS 840917 205) l l 71. Significant Condition Report SQN EEB 8529 RO, dated December 6, 1985 l (843 851220 904) ,

72. Engineering Report SCR SQNEEBN 8529 RO, dated December 27, 1985 (301 860114 908)

73. Significant Condition Report Completion Verification Report SQN EEB8529 .

R0, dated, January 28, 1986 (843 860203 9002)

74. Significant Condition Report SCR SQNEEB 8620 R0, dated February 18, 1986 (B43 860226 921) l 75. Significant Condition Report SCR SQNEEB 8620R1, dated March 21, 1986 (843 860410 910)

76. Nemorandus D.W. Wilson to R. P. Denise, dated September 17, 1986, "SQN -

NSRS Open Item Number I-85-06-WBN Limited Quality Assurance Program (B25 860917 032).

l 77. Sequoyah Nucloar Plant Generic Concern Task Force (26 concerns)

" Overfill of Cable Trays and Conduits," dated June 3, 1986 Page 37 of 39

(

78. Sequoyah Nuclear Plant Generic Concern Task Force, Dated April 29,1986,

" Plant Procedures on Overfill of Cable Trays"

79. NSRS Report I-86-251-SQN, Review of Generic Concern Issue, " Electrical Cables," dated February 14, 1986

80. Memorandum W. S. Raughley to J. A. Kirkebo, dated May 21, 1986, " Cable Sidewall Bearing Pressure," (B43 860522 912)

81. Nemorandum W. S. Raughley to Those Listed, dated June 23, 1986

" Electrical Issues - Cable Sidewall Pressure" (B43 860626 931)

82. Memorandum R. L. Gridley to H. R. Denton, dated August 26, 1986, "NMRG Report Number R-86-02-NPS, Review of Maintenance at Browns Ferry, Sequoyah, and Watts Bar Nuclear Plants"

83. Memorandum J. A. Raulston to Those Listed, dated June 5, 1986, " Letter From Quality Technology Company to Chairman Dean dated May 30, 1986 -

84. Nemorandum J. A. 01shinski to S.A. White, dated March 5, 1986, " Report Numbers 50-390/86-03 and 50-391/86-03" (A02 860306 006)

85. Nemorandum W. S. Raughley to Those Listed, dated September 2, 1986, "All Nuclear Plants - Electrical Issues - Class 1E Cable Bend Radius" (B43 860903 904)

86. Nemorandum H. A. Taff to W. S. Raughley, dated June 4, 1986, " Cable Sidewall Bearing Pressure Tests" (E13 860604 001)

87. Memorandum C. H. Sudduth to Electrical Engineering Files, dated February 28, 1986, " Evaluation of the Adequacy of Installed Class 1E Cables " (B43 860331 916)

88. Memorandum WYLE Laboratories to E. Chitwood, dated March 19, 1986,

" Performance of a Bend Test of Electrical Cables For Watts Bar Nuclear Plant (B43 860324 003)

89. Report NUREG/CR - 4548, Sandia National Laboratories for US NRC,

" Correlation of Electrical Reactor Cable Failure With Materials De6radation"

90. Nemorandum NRC dated July 10, 1986, " Notice of Meeting With TVA Concerning Cable Pulling At Watts Bar"

91. TVA Electrical Engineering Branch, dated February 18, 1986, " Notes On Sidewall Pressure / Pulling Tension / Bend Radius"

92. Memorandum Guenter Wadewitz to J. W. Coan, dated August 83, 1985, (C24 850808 007).

Page 38 of 39

s

93. Memorandum B. J. Youngblood to S. A. White, dated August 29, 1986,

" Request For Information Concerning Cable Pulling and Cable Bending Radii at Sequoyah"

94. Memorandum N. R. Harding to D. W. Wilson, dated September 4, 1986, "SQN-Cable Pulling Questions" (S10 860904 859)

95. OE Calculation dated August 6, 1985 " Justification of TVA DS-E12.1.5 Table 2 and 3. Cable Bend Radius Factors" (B45 850806 927)

96. Memorandum R. K. Seibenling to R. P. Denise date June 2, 1986,

" Assumptions of Responsibility for Nuclear Safety Review Staff Open Items and Reports" (QO1 860602 252) 4 t

b f

h Page 39 of 39 l