ML20214A110
| ML20214A110 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 10/27/1986 |
| From: | Gardner G, Lagergren W TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20214A106 | List: |
| References | |
| 2195T, 301.15-SQN, 301.15-SQN-R, 301.15-SQN-R00, XX-85-102-005, XX-85-102-5, NUDOCS 8611190226 | |
| Download: ML20214A110 (17) | |
Text
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1 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT EMPLOYEE CONCERNS TASK GROUP OPERATIONS CEG Subcategory: Mechanical Equipment Reliability and Design Element: Hardware Not Properly Identified Report Number: 301.15-SQN Revision 0 XX-85-102-OO5 i
Evaluator:
2[betLE OrM IO - D
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G.
D.' Gar er Dato Revlewed by:
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4^-87"8[
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'mber Date Approved by:
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W. R. Lageh3 en Date k
i 8611190226 861112 PDR ADOCK 05000327 P
PDR 2195T
I.
HARDWARE NOT PROPERLY IDENTIFIED This ' report evaluates the issue that hardware is not properly identified in the field. The issue was identified for Browns Ferry Nuclear Plant and was evaluated generically for Sequoyah Nuclear Plant (SQN).
II.
SPECIFIC EVALUATION METHODOLOGY The following issue was identified by Quality Technology Company (QTC) for BFN and was determined generic to SQN:
XX-85-102-OO5 Brown's Ferry:
Hardware is not properly identified in the field.
A person needs a drawing to identify it.
Nuclear Power Dept concern. CI has additional information.
The scope of this issue for SQN is perceived as relating to incorrect or missing equipment identification tags. The NRC expurgated file for this issue was reviewed and no additional information was obtained.
The SQN Special Maintenance Instruction (SMI) for system walkdowns was reviewed to determine actions underway at SQN to identify problems with equipment tags. An interim report prepared by the Office of Nuclear Power Configuration Management Survey Team was also reviewed to provide supplemental information related to component identification.
Results of the evaluations at Watts Bar Nuclear Plant (WBN) and BFN were also reviewed.
Interviews with the walkdown manager at SQN and with the Corporate Configuration Manager were conducted for supplemental information.
III.
FINDINGS SQN SMI O-317-30 (Reference 1) requires equipment tags to be checked against the flow and control drawings during system walkdowns.
The walkdown manager at SQN confirmed that this was being done.
According to the manager, several hundred missing tags have been identified. Not many cases have been identified where tags were incorrect. These cases are being documented and corrective actions initiated in accordance with the SMI.
The manager stated that not all systems are being reviewed; only those systems or portions of systems identified by the Division of Nuclear Engineering (DNE) in walkdown packages are being verified against the design drawings.
Page 1 of 3
I The Configuration Management Survey and Analysis Report (Reference 2),
Section II.C states that proper identification of items and documents will enable TVA management to ensure that:
" Items are designed, manufactured, installed / erected, tested, operated, and maintained in accordance with the design criteria and licensing commitments".
The report notes in section III.C.1 that procedures exist for all sites, except BFN, which provide general instruction for assigning identification numbers to components, however, there is no consistent review process to ensure that unique ids are assigned in accordance with prescribed standards.
As a result, multiple data bases (with different owners) exist (instruments tabs, EQIS, Q-lists drawings, etc.) which contain data discrepancies.
According to the Corporate Configuration Manager, Division of Nuclear Services, and a consultant in that office, the interim report has been included as an appendix to the new Corporate Configuration Plan which is in review at this time. This new plan provides various task activities for plant sites and other central organizations to provide for new numbering schemes and new rules for maintaining data bases.
Conclusions The issue that hardware is not properly identified was validated for SQN. The walkdown effort at SQN has identified numerous missing equipment tags and some cases of incorrect tags have been identified and are being corrected.
The scopa of tha walkdown effort does not include verifying that data bases and drawings reflect the correct component identifiers.
Although not specifically evaluated, traceability for testing and maintenance activities and proper operation of plant equipment may be compromised and are not precluded by the current situation.
For these reasons, this issue is considered safety-related and may represent a significant condition adverse to quality. The new corporate configuration plan will address the issues related to drawing / data base discrepancies.
IV.
ROOT CAUSE No root cause could be determined for missing equipment tags at SQN.
The fact that multiple or incorrect ids exist is perceived to result from the lack of a central point for control and review of ID assignments.
Page 2 of 3
F
.g V.
GENERIC APPLICABILITY Evaluations have been conducted at SQN, WBN, and BFN with similar find ing s. The issue is considered generic to Bellefonte based on these findings, however, since BLN is presently in a limited construction phase, the new corporate configuration plan should be adequate to address and resolve these issues.
For this reason, an evaluation at BLN will not be conducted by ECTG.
VI.
REFERENCES
- 1. SQN SMI O-317-30, Revision 2, " System Walkdown," August 11, 1986
- 2. TVA Office of Nuclear Power Interim Report, " Configuration Management Survey and Analysis Report," March 1986 VII.
IMMEDIATE OR LONG-TERM CORRECTIVE ACTIONS Corrective Action Tracking Document (CATD) OP 30115-SQN-01 has been written to SQN to address the issues of missing and incorrect tags discussed in this report.
CATD OP 30115-TVA-01 was written to the corporate configuration manager's office to address the resolution of data base and drawing discrepancies.
Because the later issue is generic, the corrective action response from the corporate configuration manager will be included in subcategory report 301 rather than in this report.
Page 3 of 3
~
c' TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT EMPLOYEE CONCERNS TASK GROUP OPERATIONS CEG Subcategory: Maintenance Element: Corrective Maintenance Report Number: 308.03-SQN Revision 1 XX-85-096-004 XX-85-096-005 XX-85-096-N07 GSB-85-001 DHT-85-003 2850162005 SQP-6-014-002 XX-85-071-003 Evaluator:
C. H. Gilmor 10-20-86 C. H. Gi or Date
///[
h/-2#-If, Reviewed by:
%1-
/B-2 C-fS Approved by:
W. R. Lagerkden Date 1686T i
I.
Corrective Nainten nco This report evaluates eight employee issues. Three of the issues are lR1 related to the incore instrumentation thimble tube ejection accident in April 1984 at Sequoyah Nuclear Plant (SQN). One issue is related to corrective maintenance of Auxiliary Building Secondary Containment Enclosure (ABSCE) Fire and Security doors at SQNi one issue is related to post-use calibration checking of torque wrenches at SQN; one issue I
involves the performance of temporary leak repairs and compliance with l
ASNE code requirements; one issue involves the performance of prework lR1 review of maintenance work packages, and the last issue questions the I
hardware repair process. The overall element evaluation is related to i
the adequacy of the corrective maintenance program and activities at SQN.I II.
Specific Evaluation Nethodology concern XX-85-096-004 states:
The Radiation Nonitor Tube problem at Sequoyah could happen again.
The way it was designed, it cannot be properly fixed during operation.
Concern XX-85-096-005 states:
The Radiation Monitor Tube problem (Thimble Guide Tube Incident) in Unit 1 in Apell 1985 could occur again, because the equipment is not properly designed to be fixed during plant operation.
According to a note written on the concern, XX-85-096-004 was transferred to XX-85-096-005 once additional information was obtained by Quality Technology Company (QTC) from the concerned individual.
(R1 To evaluate this issue. Nuclear Safety Review Staff (NSRS) reports lR1 I-84-012-SQN dated August 1, 1984 (reference 1) and I-85-614-SQN dated 1
l November 25, 1985 (reference 2) were reviewed to determine the depth l
of the investigation, the status of any recommendations made, and Licensee Event Reports SQRO-50-327/84030 (reference 3) and i
I SQRO-50-327/84030 Revision 1 (reference 4) were reviewed.
The SQN procedures (references 5 through 7 and 9 through 11) governing incore instrumentation thimble guide assembly maintenance were reviewed j
to determine if they had been revised as recommended in the NSRS reports.
The TVA procedure on evaluation of special tools for reactor coolant l
system work (reference 8) was reviewed to verify that it had been written as committed in the responses to the NSRS report from line management.
Additionally, SQN Plant Compliance was contacted to determine the status of commitments made in the two Licensee Event Reports (LERs).
The third concern XX-85-096-N07 states:
l Page 1 of 13
-q
n Rsvisica 1 A review was also conducted of fire door operability related LERs to determine the effect of implementation of the commitments.
The sixth concern, SQN Generic Open Concern 2850162005 states:
TVA makes repairs to their nuclear plants which are not in l
accordance with ASNE codes, such as overlays, patches, and lR1 even furmanite (sophisticated glue).
l To evaluate this issue, cognizant maintenance engineering supervision l were interviewed and referenced maintenance procedures covering I
temporary leak repairs were reviewed.
l This issue (2850162005) is a shared concern. This report covers only l the use of temporary leak repairs using viscous fluids (e.g.,
l furmanite). The remaining portion of this issue is being addressed by l the Welding Concerns Evaluation Group l
The seventh concern, SQP-6-014-002 states:
lR1 Although the foremen are required to sign a document stating they l have reviewed the work package with the craft assigned to the l
Job, they in fact often do not review the work package at all I
with the applicable craft. Nuclear Power concern.
l To evaluate this issue, cognizant maintenance supervision was I
contacted and referenced maintenance procedures were reviewed.
l The eight concern, X1-85-071-003 states:
Sequoyah: CI has general questions about hardware repair process l
and requested that QTC investigate. Details known to QTC; I
withheld to maintain confidentiality.
l lR1 The NRC expurgated file for this issue was reviewed and I
contained no additional information. Because of the lack of l
information, an evaluation of this issue was not possible. The SQN l
Site Director was notified by memorandum (reference 33).
l III.
Findings The baseline requirement for all the evaluated concerns is established by Technical Specification 6.8.1. (reference 22) which states:
Page 3 of 13
m R; vision 1 c'
6.8 Procedures and Programs 6.8.1 Weltten procedures shall be established, implemented and maintained covering the activities referenced below:
The applicable procedures recommended in Appendix "A" a.
of Regulatory Guide 1.33. Revision 2. February 1978.
NRC Regulatory Guide 1.33 Revision 2, (reference 23) February 1978 Appendir A states:
Maintenance that can affect the performance of safety-related a.
equipment should be properly preplanned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances.
These requirements are implemented through the Nuclear Quality Assurance Manual (NQAM), (reference 24), part II, section 2.1, which states in part:
This procedure provides control over maintenance activities which could affect the capability of critical structures, systems, and components (CSSC) to perform their intended safety function...
The requirementF of this procedure apply to activities which are performed on CSSC items for the purpose of repairing, reworking, replacing or readjusting...
Maintenance which can affect the performance of CSSC items shall be preplanned and performed in accordance with written procedures or documented instruction appropriate to the circumstances and which conform to applicable codes, standards, specifications, and criteria.
A.
Concerns 11-85-096-004 and IX-85-096-005 The concerns as documented by NSRS reports I-84-012-SQN and I-84-614-SQN (references 1 and 2) were substantiated in that the thimble guide tubes are not designed to be maintained during power operation. However, review of the NSRS reports I-84-012-SQN and I-85-614-SQN; the recommendations included in the reports; Sequoyah Maintenance Instructions SMI-0-94-1 SMI-0-94-3, SMI-1-94-5, MI 1.9, MI 1.10 and 1.11; and procedure SQM-63 (references 7, 9. 5, 10, 11, 6, and 8, respectively) demonstrates that the employee concern has been adequately addressed and appropriate corrective actions taken, i.e., the design of the thimble guide tube has not been changed, but the incident has been prevented from recurring by the establishment of controls which prohibit maintenance from belns performed at any time when the reactor coolant system is above atmospheric pressure and 150*F.
Page 4 of 13
F-Revisien 1 Based upon the recommendation in NSRS report I-84-012-SQN (reference 1), the site provided specific commitments to the NRC by LER SQRO-50-327/84030 Revision 1, (reference 4).
These commitments included review and update of the applicable l
procedures, implementation of a special tool control program, and a lR1 prohibition on cleaning thimble guide tubes with the reactor coolant I
system above atmospheric pressure. These commitments were entered I
into the Commitment Action Tracking System (CATS).
Review of the CATS status on August 13, 1986, verified that each of these commitments has been closed.
B.
Concern XI-85-096-N07 This concern is substantiated as documented in NSRS Report I-84-012-SQN (reference 1) III.L.1 and IV.P.
The concern was subsequently closed by completion of the recommendation contained in I-84-012-SQN (reference 1), specifically the issuance of LER SQRO-50-327/84030 Revision 1 (reference 4) to SQRO-50-327/84030 on October 11, 1985.
lR1 It should be noted that Employee Concern Task Group (ECTG) Operations Category Report 307.08-SQN dated August 4, 1986 (reference 18),
documents a general concern regarding improper reporting to the NRC and determined that the concern could not be substantiated on a lR1 general basis.
l C.
Concern GSB-85-001 The requirements for using calibrated measuring and test equipment (M&TE) are included in Regulatory Guide 1.33 February 1978 (reference 23) and implemented in TVA through NQAM, Revision 0 (reference 24), part III, section 3.1 which states:
This proceduto establishes a calibration program to control and verify the accuracy of Measuring and Test Equipment (M&TE) which is used to ensure that critical systems, structures, and components (CSSC) are in conformance with prescribed technical requirements and that data provided by testing, inspection, calibration, or maintenance is valid.
There is no written specific requirement to perform a recalibration or calibration check immediately following the use of M&TE.
However, an NRC audit (referred to in reference 12) during the week of January 21, 1985, resulted in the following findings:
Page 5 of 13
R vision 1 1.
It is difficult to ascertain if any particular group is responsible for the handling of AI-31, attachment 5.
2.
The present tracking program is cumbersome because it requires the Materials Unit personnel to "run around" to get these cleared.
3.
Evaluations are sometimes difficult to make, and corrective action is hard to implement because a long period of time elapses before we know the tool is out of calibration.
4.
Some people apparently do not respond quickly enough to make the 60-day limit on evaluations.
The findings were documented in a memorandum from J. B. Krell to P. R. Wallace dated February 6, 1985 (SS3850201910) (reference 12) and document that the typical method of N&TE control, i.e.,
maintenance of use records and periodic calibration with follow-up on instruments found out of calibration, was not working at SQN.
The concern was thus substantiated. To resolve the post-use calibration issue (item 3 above), recommendation 2 of the J. B. Krell memorandum stated:
Problem 3 could be reduced if the toolroom would check tools such as torque wrenches, torque multipliers, or other
" problem" instruments each time they are returned. If that check shows that the tool is out of tolerance, the engineer could take immediate corrective action but would not have to re-evaluate any job except the one it was just used on.
The present system finds a problem (but it) may be months after a job is done and could conceivably require an outage to take corrective action. The Materials Unit would probably have to write an instruction and train people to do this checking.
This recommendation was implemented through a revision to Administrative Instruction AI-31 (Reference 13). Revision 5 of AI-31 (Reference 13), dated June 5, 1986, was reviewed and contains a requirement that upon return of torque wrenches to the toolroom a reference check shall be performed in accordance with Section Instructional Letter (SIL) SS/NU-3 to ensure the tool was in calibration at the required torque value and I
during the time it was used. This action effectively lR1 implements reconunendation 2 of J. B. Krell's memorandum.
l Page 6 of 13
R3risien 1 s'
D.
Concern DNT-85-003 The concern regarding maintenance of fire and security doors is IR1 substantiated as documented by LER SQRO-50-327/84073, which reported a total of 74 fire doors as being either nonfunctional or as not meeting Underwriters Laboratories' standards as the result ofmaintenance and modification activities performed...".
Additionally, 12 reportable occurrences documented on 5 LERs were issued because firewatches could not be performed because of inoperable fire doors preventing access to the areas to be checked.
The actions initiated and completed by SQN to resolve this concern (reference 15) only address the mechanical maintenance activities.
No documented action has been taken to resolve the concern regarding mechanical work being classified CSSC while electrical and instrument work on the same door is classified as non-CSSC.
According to the cognizant supervisor, this portion of the concern was resolved during training. Specifically, it was explained that the mechanical work could affect the ability of the door to perform its intended function, i.e., protect an area or room from spread of fire from another room or isolate it from the spread of contamination.
The electrical and instrumentation work, on the other hand, affect door operators and operator indicators which, although important, are not CSSC functions assigned to fire or ABSCE doors.
Documentation of this explanation, however, is not included in the training records, nor could any other documentation be found for the evaluation of this portion of the employee's concern.
A review of LERs entered into the records management system for 1984 through July of 1986 showed a total of nineteen instances where door failure or inoperability directly or indirectly resulted in a reportable occurrence. However, the last incident reported occurred on March 18, 1986, which is about the time the dedicated door crew was established and began to function.
E.
Concern 2850162005 The base line requirements delineated above, as well as the ASME Boiler l
and Pressure Vessel Code (reference 29) apply to work on CSSC components.
l However, there is no portion of the code which addresses temporary l
packing or gasket leak repair.
Interviews held with cognizant lR1 maintenance engineers and supervision and review of Special Maintenance l
Instructions (reference 30) documented that until mid-1986, furmanite i
repales had been conducted on CSSC in accordance with Plant Operations l
Review Committee (PORC) and Plant Manager Approved Special Maintenance l
Instructions (SNIs).
l Page 7 of 13
R3risien 1 In each case reviewed, the leak repair affected involved temporary l
repair of a leaking bonnet-to-body or cover gasketed mechanical l
Joint or the temporary repair of a leaking packing gland. In no l
case was a repair attempted on a hard metal pressure retaining part.
l 1.e., the leak temporarily repaired resulted from partial failure of I
a gasket or packing and not the ASME code covered valve body or l
bonnet. The temporary leak repair method involves the pressure l
injection of a small amount of viscous fluid into the leak area to l
re-establish the seal between the metal parts, and in the cases l
reviewed the maintenance work order covering the repair was held IR1 open until plant conditions were such that a permanent repair, e.g.,
I repacking or gasket replacement, could be effected.
l Effective in mid-1986, these repairs are to be handled in accordance l
with Administrative Instruction AI-19, Part IV (reference 31) as l
plant modifications and all previously issued SMI's have been lR1 cancelled.
l F.
There is no baseline requirement for concern SQP-6-014-002.
l However, there is a requirement, SQN-1 Appendix B paragraph 4.3.8 lR1 (reference 31) that maintenance instructions include prework l
briefing instructions. Additionally SQM-2, (reference 32) l paragraph I.2.F states:
i It shall be the responsibility of the foreman to review all l
WR packages for proper planning (See planning and review I
checklist Appendix D).
For CSSC WR packages it is his lR1 responsibility to ensure that an instruction review sheet I
(attachment E) is attached, and that he and the craftsman I
have reviewed the WR package and all instructions attached.
l The review sheet shall be signed by the foreman and the l
craftsman before work begins.
l Interviews with cognizant maintenance supervision confirm that each I
foreman is required to ensure completion of attachment E to SQN 2 by l
the craftsmen (reference 32) and that the completed form is returned l
to the cognizant maintenance supervisor. There is no requirement (R1 for the supervisor to sit down with each craftsman to review the WR l
package. When suggestions are made for, improvement of the l
procedure on the instruction review sheet, it is retained until the l
procedure is revised to incorporate the suggestions, otherwise l
the cognizant maintenance supervisor discards the instruction l
l review sheet when the job is completed.
l l
i i
l Page 8 of 13
R3rision 1
==
Conclusions:==
Concerns E-85-096-004 and U-85-096-005 are substantiated. The actions committed to by SQN to the NRC have been closed out and although the letter of the concern, i.e. design of the thimble guide tube, has not been changed, actions have been taken to prevent l
maintenance from being performed when the reactor coolant system is lR1 above atmospheric pressure and 1500F, thereby effectively l
preventing recurrence of the event.
l Concern H 096-N07 is substantiated. As shown in NSRS report I-84-012-SQN, LER SQRO-50-327/84030 (reference 1) was, by being incomplete, misleading. However, as documented in ECIG-Operations report 307.08 SQN (reference 18) this LER is the exception rather than the rule and was resolved by issuing a revision to LER SQRO-50-327/84030 (reference 4).
Concern GSB-85-001 is substantiated. The program in place at SQN met all regulatory requirements but was ineffective and cumbersome thereby inhibiting timely feedback to users when, during the periodic recalibration of a torque wrench, the wrench "as-found" data was determined to be out of tolerance. The actions taken and procedure changes implemented have effectively resolved both the concern and the user feed back problems of the earlier program by providing immediate post-use verification of torque wrench calibration.
Concern DHT-85-003 is substantiated by the number and frequency of reportable occurrences resulting from inoperable fire doors. The 4
actions taken and in progress and the procedure changes implemented, have effectively resolved the concern with the exception of the j
following:
(1) The question pertaining to the CSSC designation of work activities l l
needs to be addressed including the impact upon the access to lR1 potentially vital areas of the plant because of the non-CSSC work. l (2) Open Maintenance Action Tracking System (NATS) items 1294, 1295, l
l and 1298 (reference 26) should be completed and closed out.
lR1 (3) The training activity conducted for the dedicated door crew l
appears to be a one time only class, without provision for i
periodic retraining or training of new personnel assigned to the lR1 dedicated door crew. The need for recurring training should be l
evaluated.
l Note:
Element report 306.01 also addresses problems related to l
fire doors.
lR1 l
Page 9 of 13
R visicn 1 C:ncern 2850162005 was not validated with regard to leak repairs in l
that there is no ASME code requirement which addresses temporary l
packing or gasket leak repair using furmanite or any other sealing l
compound. The use of furmanite at SQN was for mechanical joint l
1eakage and in no cases was it used for repairing the ASME valve l
body or bonnet. The actions taken previously by SQN maintenance l
personnel to perform this work in accordance with specific approved lR1 procedures meets the requirements of Technical Specifications and l
the NQAM. The revised process implemented in mid-1986 to handle this l
work as a modification is sn improvement in the process and meets I
current NRC and INPO recommendations.
l Concern SQP-6-014-002 was not validated in that there is no l
requirement for craft supervision to review work packages with l
craft personnel individually. The supervisor documents, by lR1 his signature on attachment E of SQM-2, that the craftsman have l
Individually reviewed the work package and that any problems l
have been resolved.
l Of the concerns evaluated, XX-85-096-004, XX-85-096-005, l
XX-85-096-N07, and GSB-85-001 were validated and are considered l
to be safety-related based upon the fact that they either IR1 involved safety-related equipment (i.e., thimble guide tube) or they l
involved safety-related activities (i.e., calibration of M&TE).
l Corrective actions for these deficiencies were already in place l
prior to this evaluation and no further corrective action is l
required. Concern DHI-85-003 regarding fire door maintenance does l
not meet the current TVA definition of safety-related, however.
l some deficiencies were noted which will require evaluation by SQN.
l Concerno 2850162005 and SQP-6-014-002 are not considered safety-l related since they could not be validated and no deficiencies l
were identitled.
l IV.
Root Cause The common theme to the five concerns substantiated was a lack of adequate procedural control over the specific activity being performed resulting from unclear, inadequate, or incomplete procedures.
The root cause is supported by the findings of the Nuclear Management l
Review Group review of maintenance activities as reported in the exit lR1 meeting report dated July 29, 1986 (SS3 860805 256), (reference 27, l
(
finding F-2) which states:
l Some instructions were not clear, were not concise, and did not containl the information necessary for users to understand and perform work lR1 activities effectively."
l i
A root cause for concerns 2850162005 and SQP-6-014-002 could not be l
established since the concerns could not be validated.
lR1 Page 10 of 13
R visicn 1 V.
Generic Applicability The issue of performing maintenance on the reactor coolant system l
(XX-85-096-005) should be evaluated at WBN to ensure procedures do not lR1 0
allow this practice above atmospheric pressure and 150 F.
l The issue of torque wrench calibrations (GSB-85-001) is considered lR1 generic and will be evaluated at all sites.
l The issue of door maintenance (DHT-85-003) will be evaluated at l
WBN and BFN. Since BLN is in a construction phase, the issue is IR1 not considered generic at BLN.
l The improper use of furmanite for temporary packing and gasket leak l
repair was not validated and is not considered generic. The remaining l
1ssues of 2850162005 will be addressed by the Welding Category Evaluation lR1 Group.
l The issue that craft foreman do not review work packages with I
craftsman (SQP-6-015-002) was not validated and is not considered lR1 generic. No procedure violations were noted by this evaluator.
l VI.
References 1.
NSRS Report I-84-012-SQN dated August 1, 1984 2.
NSRS Report I-85-614-SQN dated November 25, 1985 3.
Licensee Event Report SQRO-50-327/84030 dated May 18, 1984 4.
Licensee Event Report SQRO-50-327/84030, Revision 1, dated October 11, 1984 5.
Sequoyah Maintenance Instruction SMI-1-94-5, Revision 1. " Thimble Tube Installation," dated July 10, 1986 (Cancelled) 6.
Sequoyah Maintenance Instruction MI-1.11, Revision 0, " Thimble Tube Installation," dated July 10, 1986 7.
Sequoyah Maintenance Instruction SMI-0-94-1, Revision 1 "RPV Bottom Mounted Instrument Thimble Tubes Cleaning and Flushing - Units 1 and 2 " dated October 9, 1984 (Cancelled) 8.
Sequoyah procedure SQM-63. Revision 0, "Special or Modified Tooling-Primary Systems " dated May 9, 1985 I
l 9.
Sequoyah Maintenance Instruction SMI 0-94-3, Revision 2. " Seal Table High Pressure Seal Repair," dated June 27, 1986 Page 11 of 13
R3risicn 1 10.
Sequoyah Maintenance Instruction MI-1.9 Revision 7 " Bottom Mounted Instrument Thimble Tube Retraction and Reservation," dated June 27, 1986 11.
Sequoyah Maintenance Instruction MI-1.10. Revision 3. "Incore Flur Thimble Cleaning and Lubrication" 12.
Memorandum J. B. Krell to P. R. Wallace dated 6, 1985 lR1 (SS3 850201 910) 13.
Sequoyah Administrative Instruction AI-31 Revision 5 " Control of Measuring and Test Equipment " dated June 5, 1986 14.
Section Instructional Letter SS/MU-3, Revision 0, " Measuring and Test Equipment," dated March 5, 1986 15.
Memorandum D. H. Tullis to CI dated January 30, 1986 (No RIMS Number) 16.
Sequoyah Surveillance Instruction SI-261 Revision 9 " Visual Inspection of Technical Specification Fir; Doors on a Periodic Basics" dated July 16, 1986 17.
Sequoyah Maintenance Instruction MI-16.1, Revision 0, " Repair and Maintenance of Fire Doors and Various Fire Door Hardware Units 1 and 2," dated May 29, 1986 18.
Employee Concerns Task Group-Operation Report 307.08 SQN dated August 4, 1986 19.
NQAM Part II Section 2.1, Revision 0 20.
Phy. SI-13. Revision 48, " Fire," May 5, 1986 21.
NMRG Maintenance Review Exit Meeting at SQN July 29,1986 S53 860805 256 22.
Sequoyah Technical Specifications Unit 1, Revision 48 1
23.
NRC Regulatory Guide 1.33 Revision 2. February 1978 24.
Nuclear Quality Assurance Manual Revision 1 25.
Licensee Event Report SQRO-50-327/84073 Dated 12/26/84 26.
Maintenance Action Tracking System Item 1294: Develop detailed drawing of each CSSC door Item 1295: Set up doors in Power Stores on initial stock Item 1298: Revise PHYSI-3 to distinguish ABSCE/ fire door branch Page 12 of 13
R:visica 1
\\
l 27.
NMRG Review of Maintenance Exit Neeting Notes dated July 25, 1986 (S53860805256) 28.
NUREG 1369-CR, Revision 1. September 1982 4
29.
American Society of Nechanical Engineers Boiler and Pressure Vessel l
l Code (1981) l 30.
Special Maintenance Instruction (SNI) l
- SMI-2-62-4, Revision 1, dated May 4, 1983 lR1
- SMI-1-1-4, Revision 5 dated November 19, 1982 l
- SMI-1-3-5, Revision 3, dated 5/24/82 l
- SMI-2-1-2, Revision 0, dated February 10, 1984 l
1 A
l 31.
Sequoyah Standard Practice SQM-1, Revision 6 l
32.
Sequoyah Standard Practice SQM-2, Revision 19 33.
Memorandum to H. L. Abercrombie from W. R. Brown, " Element Reports 1
301.06-SQN and 308.03-SQN," dated August 22, 1986 (T25 860822 900) l 1
j VII. Immediate and Long-Term Corrective Actions Concern DHT-85-003 corrective actions.
The remaining maintenance action tracking items 1294, 1295, and 1298 (reference 26) must be completed as scheduled. Additionally, the issue 1
of CSSC designation of door maintenance activities should be resolved and documented and the door maintenance training program should be more l
formalized with a defined lesson plan and requalification requirement.
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